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MCNPX User`s Manual

1. MCNPX User s Manual 99 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 47 Source Probability Card Variable Description Sets how the P s are interpreted Allowed values are blank same as D for an H or L on the SI card probabil ity density for an A distribution on the SI card option D bin probabilities for an H or L distribution C cumulative bin probabilities for an H or L distribution V for cell distributions probability is proportional to cell volume x P if P s are present P P source variable probabilities must be zero for 1st histo jee Lk gram bin designator negative number for a built in function a b parameters for the built in function Table 5 48 Default SPn D P4 Pk 5 6 1 3 SBn Source Bias Form SBn option B By or SBn f ab n option f a and b are the same as for the SPn card except that the only values allowed for fare 21 and 31 Bi By source variable biased probabilities Default SBn D B Bk Table 5 48 Special Source Probability Functions Source Variable Function No and Input Parameters Description ERG 2 a Maxwell fission spectrum ERG 3 ab Watt fission spectrum ERG 4 ab Gaussian fusion spectrum ERG 5 a Evaporation spectrum 100 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 48 Special Sou
2. surface number 1 lt j lt 99999 j If surface defines a cell that is transformed with TRCL 1 lt j lt 999 See Section j reflecting surface j white boundary surface absent or 0 for no coordinate transformation n gt 0 specifies number of a TRn card lt 0 specifies surface j is periodic with surface n a equation mnemonic from Table Table 5 6 list one to ten entries as required 60 MCNPX User s Manual Table 5 6 MCNPX Surface Cards MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Mnemonic Type Description Equation Card Entries IP Plane General Ax By Cz D 0 ABCD PX Normal to X axis x D 0 D PY Normal to Y axis jy D 0 D PZ Normal to Z axis iIz D 0 D SO Sphere Centered at Origin 12 ytz R 0 R S General j a xyZR SX Centered on X axis x x y y z z R 0 X R SY Centered on Y axis xx 4y24227 R 0 SZ Centered on Z axis 7 yn x y 9 z R 0 Z R y y 2 2 R 0 C X Cylinder Parallel to X axis ae a ae yzR C Y Parallel to Y axis 72 R 0 ZZR C Z Parallel to Z axis x 3 2 2 R 0 Da x y R CX On X axis Be A_R 9 CY On Y axis Ce rsa R CZ On Z axis y z R 0 R xX z R 0 R x y R 0 K X Cone Parallel to X axis 5 5 e lyy 5 x 1 K Y Parallel to Y axis YW E a 0 ee K Z Parallel to Z axis Ena 2 E xy zt hee 2
3. er Mean Lifetime Low Kinetic seconds IPT Name of Particle Symbol Mass MeV Energy Cutoff decayed on MeV production Original MCNP Particles 1 neutron n n 939 56563 0 0 887 0 1 anti neutron n n 939 56563 0 0 887 0 2 photon y p 0 0 0 001 huge 3 electron e e 0 51 1008 0 001 huge 3 positron et e 0 51 1008 0 001 huge Leptons 4 muon u l 105 658389 0 11261 2 19703 x 10 6 pipe sym bol 4 anti muon u 105 658389 0 11261 2 19703 x 108 5 tau T i 1777 1 1 894 2 92x10 38 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 Table 4 1 MCNPX Particles LA CP 02 408 are Mean Lifetime Low Kinetic seconds IPT Name of Particle Symbol Mass MeV Energy Cutoff d MeV ecaye on production 6 electron neutrino u 0 0 0 0 huge Ve 6 anti electron neu u 0 0 0 0 huge trino 7 muon neutrino Vm V 0 0 0 0 huge 8 tau neutrino v w 0 0 0 0 huge Baryons 9 proton p h 938 27231 1 0 huge 9 anti proton p h 938 27231 1 0 huge 10 lambda A9 1115 684 1 0 2 632 x 10 lower case L 11 sigma 1189 37 1 2676 7 99 x 103 12 sigma 1197 436 1 2676 1 479 x 102 13 cascade x 1314 9 1 0 2 9x 107 14 cascade y 1321 32 1 4082 1 639 x 102 15 omega Q o 1672 45 1 7825 8 22 x 108 16 lambda A c 2285 0 2
4. if omitted the default behav ior is system dependent the detected hardware soft ware platform and compilers determine what the default FFLAGS should be MCNPX User s Manual MCNPxX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 3 1 Configure Script Parameters Option Syntax Effect on the generated Makefile if requested Effect on the generated makefile if NOT requested with CFLAGS value There is a separate variable that is used for optimization switches See with COPT in this table If in doubt run the con figure script and examine the system default or system computed values that appear in the gener ated Makefile h You may want to include the defaults in the string you specify for CFLAGS with this mechanism when configure is run again substitute a quoted or double quoted string for value that represents allowable com piler switch settings these settings will override the system default or system computed values if omitted the default behav ior is system dependent the detected hardware soft ware platform and compilers determine what the default CFLAGS should be MCNPX User s Manual 23 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 3 1 Configure Script Parameters Option Syntax Effect on the generated Effect on the generated p y Makefile if requested makefile if NOT requested wi
5. 4 1 5 Level Densities As the excitation energy of a nucleus increases excited level states get closer together in energy Methods of statistical mechanics and thermodynamics have long been used to describe the structure of a highly excited nucleus At large excitation energy E the density of excited levels 1 D where D is the average distance between levels is of the form 1 4 2 ae d Ce where C and a are parameters which are functions of the mass number and must be empirically adjusted Generally C is evaluated from the observed level density at low exci tation E 1 MeV and a is adjusted to represent the spacing of levels found from the resonance capture of slow neutrons E 6 to 8 MeV Users of intermediate energy simu lations codes have long known that results are highly sensitive to how the a parameter is set Three options for level density parameters are offered by the Bertini and ISABEL codes Ignatyuk model The default evaluation of the level density parameter a uses the energy dependent formulation of Ignatyuk as implemented in GNASH ART88 with the provision that E 0 Where E is the excitation energy and ag is the Gilbert Cameron Cook level density parameter 1 Excellent discussions of level density physics can be found in many standard nuclear physics textbooks such as Chapter 11 of EVA55 44 MCNPX User s Manual MCNPX User s Manual Ap Version 2 3 0 April 2002 LA UR 02 2607 Accelerato
6. En Upper energy bin limit The lower bin limit is considered to be zero Sn Use Splitting if Sn gt 1 Splitting Use Roulette if 0 lt Sn lt 1 5 8 17 ESPLT Energy Splitting and Roulette Form ESPLT n N4 N5E5 Table 5 103 ESPLT Card Descriptor Description n any particle symbol or IPT number from Table 4 1 Ni number of tracks into which a particle will be split E TS MeV at which particles are to undergo split Default Omission of this card means that energy splitting will not take place for those particles for which the card is omitted Use Optional use energy dependent weight windows instead Example ESPLT N2 1 2 01 25 001 This example specifies a 2 for 1 split when the neutron energy falls below 0 1 MeV another 2 for 1 split when the energy falls below 0 01 MeV and Russian roulette when the energy falls below 0 001 MeV with a 25 chance of surviving MCNPX User s Manual 165 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 5 8 18 PWT Photon Weight Form PWT Wi Wo Wie Wi Table 5 104 PWT Card Variable Description W relative threshold weight of photons produced at neutron collisions in cell number of cells in the problem Use Recommended for MODE N P and MODE NPE problems without weight windows NOTE The PWT card is ignored if a WWP P photon weight window exists 5 9 OUTPUT CONTROL PRDMP LOST DBCN FILES
7. The MCNPX code and data effort represents the efforts of many people much of whose work is represented in this manual The primary team members are listed below Code Development Team H Grady Hughes team leader Harry W Egdorf Franz C Gallmeier John S Hendricks Robert C Little Gregg W McKinney Richard E Prael Teresa L Roberts Edward Snow Laurie S Waters Morgan C White Library Development Team Mark B Chadwick Stephanie C Frankle Gerald M Hale Robert C Little Robert MacFarlane Morgan C White Phillip G Young Physics Development Team David G Madland Stepan G Mashnik Richard E Prael Arnold J Sierk APT AAA Target Blanket Design and ED amp D Team LANSCE Team Michael W Cappiello Rhonda K Corzine Phillip D Ferguson Michael M Fikani Frank D Gac Michael R James Russell Kidman Stuart A Maloy Michael A Paciotti Eric J Pitcher Lawrence G Quintana Gary J Russell Beta Test Team 900 users from 200 institutions worldwide MCNPxX was originally conceived as an upgrade to the existing Los Alamos LAHET Code System LCS and our deepest thanks is extended to Dr Richard E Prael for his support and guidance Without his longtime vision of providing the highest quality simulation tools to the accelerator community the MCNPX project could not have happened MCNPX 2 3 0 is based on MCNP4B and we gratefully acknowledge the importance of that seminal code in our work The MCNP code series
8. Variable Description n tally number the alphabetic keyword identifier for a special treat ID ment keyword Description FRV fixed arbitrary reference direction for tally 1 cosine binning TMC time convolution INC identify the number of collisions ICD identify the cell from which each detector score is made GEB Gaussian energy broadening identify the sampled index of a specified source distri SCX bution SCD identify which of the specified source distributions was used PTT put different multigroup particle types in different user bins ELC electron current tally 132 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 75 FTn Card Special Treatment for Tallies Variable Description P parameters for that special treatment either a num J ber a parenthesis or a colon require FUn card Default If the FT card is absent there is no special treatment for tally n Use Optional as needed A description of the special treatments available follows with an explanation of the allowed parameters for each FRV V V V3 The V are the xyz components of vector V not necessarily normalized If the FRV special treatment is in effect for a type 1 tally the direction V is used in place of the vector normal to the surface as the reference direction for getting the cosine for binning GEB abc The parameters specify the full width at h
9. yes then we set a local script variable ac_oldxs to yes For completeness we define that local variable with a default value of no inthe AC_CLL_DEFAULTS macro This gives the variable a value even if the option was not used Later in a more strategic place in the code we will test ac_oldxs and do something appropriate In our case we put the code that acts to define the symbol into the AC_EXTRA_DEFINES macro which is called last during execution of the AC_ENV_FLAGS_VARS macro The code associated with our ac_oldxs defines an extra symbol OLDM that will appear on the compile line as DOLDM MCNPX User s Manual 31 MCNPX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium After saving all of the changed files you must regenerate the configure scripts by execut ing the following command from within the menpx_2 3 0 directory force regeneration of configure scripts at all levels autoreconf localdir config f Once the configure scripts at the various levels have been generated you can execute configure with the desired feature that were added For our example we would execute the following to use our new with OLDXS option in order to get old cross sections acti vated when the Fortran code is compiled from the top level of your working directory configure and request that the new option be used configure with OLDXS The configure will recursively descend the nece
10. 1 xtie ulnu yu A en o d oA 5 ae u where x is a positive real number specifying the line of integration Blunck and Leisegang BLU50 have extended Landau s result to include the second moment of the expansion of the cross section Their result can be expressed as a convo lution of Landau s distribution with a Gaussian distribution o0 n2 f s A a f Hs exp C5 an o0 o Blunck and Westphal BLU51 provided a simple form for the variance of the Gaussian 4 Ogwy 10eVZ A MCNPX User s Manual 59 MCNPX User s Manual Ap Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium Subsequently Chechin and Ermilova CHE76 investigated the Landau Blunck Leisegang theory and derived an estimate for the relative error ees 105 1 a i caused by the neglect of higher order moments Based on this work Seltzer SEL91 describes and recommends a correction to the Blunck Wesitphal variance Bees OBw 1 3 cR This is the value for the variance of the Gaussian which is used in MCNP MCNPX Electron Multiple Scattering ETRAN and MCNP MCNPxX rely on the Goudsmit Saunderson theory GOU40 for the probability distribution of angular deflections The angular deflection of the electron is sam pled once per subset according to the distribution w F s u 5 1 3 exp sG P u l 0 where s is the length of the substep u cos is the angular deflection from the directi
11. 35 ANSI ANS 6 1 1 1991 ISO isotropic a en Particle energy Interpolation method 1 logarithmic interpolation in energy linear in function 2 linear interpolation in energy and function 3 recommended analytic parameterization not available for ic 10 units of the result 1 rem hr particles em sec 2 sieverts hr particles cm2 sec 114 MCNPX User s Manual MCNPxX User s Manual Version 2 3 0 April 2002 E LA UR 02 2607 Accelerator Production of Tritium Table 8 9 DFACT Argument Descriptions Continued ARGUMENT DESCRIPTION acr Normalization factor for dose DFACT result will be multiplied by any factor greater or equal to 0 0 for exam ple acr 1 0 means no change The value must be a real number Certain special options are also available 1 0 normalize DFACT results to Q 20 by dividing out the parametric form of Q which equals 5 0 17 0 exp In 2E 2 6 from ICRP60 1990 paragraph A12 2 0 Apply LANSCE albatross response function MCNPX User s Manual 115 MCNPX User s Manual Ap Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium 8 5 HISTP and HTAPE3X In order to produce the LAHET compatible HISTP files the following card must be added to the inp deck HISTP no arguments HTAPESxX is a code for processing medium and high energy collision data written to the HISTP history file by MCNPX In add
12. 62 5 3 2 4 Surfaces Defined by Macrobodies 0 0000 63 5 3 2 4 1 BOX Arbitrarily oriented orthogonal box 63 5 3 2 4 2 RPP Rectangular Parallelepiped 63 5 3 2 4 3 SPH Sphere ideii 0c ee 64 5 3 2 4 4 RCC Right Circular Cylinder Can 64 5 3 2 4 5 RHP or HEX Right Hexagonal Prism 65 5 3 2 4 6 REC Right Elliptical Cylinder 65 5 3 2 4 7 TRC Truncated Right Angle Cone 66 5 3 2 4 8 ELL Ellipsoid 2 0 0 eee 66 5 3 2 4 9 WED Wedge 0 00 eee 67 5 3 2 4 10 ARB Arbitrary Polyhedron 00 000 00 67 5 3 3 Geometry Data 00 2 cee 68 5 3 3 1 VOL Cell Volume 0 eee 68 5 3 3 2 AREA Surface Area 000 ccs 69 SISS U UNiverse eaaa aa eres ea adie daa Hes 69 5 3 3 4 FIER Fillvas idea aaa el ani Sete tack dee Rife ee 70 5 3 3 5 TRCL Cell Transformation 0 000 c eee eee 71 DFSG LAT Lattice iiss deren hl eer eet nad 72 5 3 3 7 TRn Coordinate Transformation 0 0008 73 5 4 MatetiialS 23 3 ood Ca ee eee oe ee aan cere iia 74 MCNPX User s Manual vii viii MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 54 1 Mm Materi l s c 0 203 0 o 060 sion 2 2o aaa tol alae See ees 74 5 4 2 MTm 0 B Material 0 cece cece 76 5 4 3 MPNm_ Photonuclear Material 0200
13. MCNPX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium here given by source probability distributions SP1 and SP2 The z coordinate is left unchanged Z 0 There is no PAR option in this example therefore the particle generated by this source will be the one with the lowest IPT number in Table 5 1 neutron The SP cards have three entries The first entry is 41 which indicates sampling from a built in gaussian distribution note the function 41 is a gaussian in time in MCNP It has been modified for the purpose of MCNPX It has the following density function 2 2 poy exr 4 Z anan 1 T The parameters a and b are the standard deviations of the Gaussian in x and y The second entry fx or fy on the SP cards is the full width half maximum FWHM of the Gaussian in either the x or y direction and must be computed from a and b by the user as follows f 8In2 a 2 35482a hpi fy 8In2 b 2 35482b The third entry represents the centroid of the Gaussian in either the x or y direction We recommend that the user input 0 here and handle any transformations of the source with a TR card as described below Using a non zero value will interfere with the rejection func tion as specified by the cookie cutter option Note that in Table 10 in the MCNPX output file the definitions of a b and c are different from those discussed above however fwh
14. Must be followed by a single reference to a TR card that can be used to trans late and or rotate the entire mesh Only one TR card is permitted with a mesh card 96 MCNPX User s Manual MCNPX User s Manual Version 2 3 0 April 2002 F LA UR 02 2607 Accelerator Production of Tritium Source Mesh Tally type 2 The second type of Mesh Tally scores source point data in which the weight of the source particles P 1 P 2 P 3 are scored in mesh arrays 1 2 3 therefore a separate mesh tally grid will be produced for each particle chosen Currently it is not possible to chose more than one particle type in a type 2 Mesh Tally1 However some graphics pro grams will enable the user to add separate histograms together offline The usefulness of this method involves locating the source of particles entering a certain volume or crossing a certain surface The user asks the question If particles of a certain type are present where did they originally come from In shielding problems the user can then try to shield the particles at their source Refinements in this tally will be forthcoming in further versions of MCNPX as user feedback is received This mesh tally is normalized as number per SDEF source particle R C S MESHn P 1 P 2 P 3 P A trans n 2 12 22 32 note number must not duplicate one used for an F2 tally Table 8 2 Source Mesh Tally type 2 Keyword Descriptions Key
15. QO QO QO QO 0 QO QO QO QO 0 QO QO QO D Dy 0 0 9 8003E 00 1 3626E 02 5 7541E 02 4 9705E 01 6 8449E 00 QO QO 9 8600E 01 1 0482E 02 5 5000E 04 1 0972E 02 QO 0 1 9848E 01 3 4100E 02 cutoffs tco 1 0000E 34 eco 0 0000E 00 wcl 5 0000E 01 wc2 2 5000E 01 Net neutron production in this case is 18 364 n p or 0 5 above the base case value The difference is primarily due to the neutron multiplicity between 20 and 150 MeV in the new 150 MeV evaluations as compared to the multiplicity given by the LAHET physics models 196 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 in this energy range Since the data evaluations are considered more accurate than the LAHET physics models the base case value of 18 263 should be considered the better estimate Note the difference in net production by nuclear interactions 15 617 n p for the base case versus 17 897 n p for case 1 and by n xn reactions 3 785 n p for the base case versus 0 516 n p for case 1 for the two cases The difference of 2 280 n p between the two cases for net production by nuclear interactions is the value calculated by the LAHET modules within mcnpx for net neutron production by neutrons in the energy range 20 to 150 MeV Similarly the difference of 3 269 n p in the values for net n xn production is the value predicted by the new 150 MeV Pb data libraries for net neutron production by neutrons with energ
16. contains the regression test files for the various known platforms in use src contains the source code files for mcnpx and several related utilities miscellany contains things that don t fit into any other category of interest to developers config contains autoconf related macros scripts initialization files Second Level bertin builds and executes a program hcnv to translate LAHET text input to binary input phtlib builds and executes a program trx to translate LAHET text input to binary input gridconv converts output files generated by mesh tally and mctal files into a variety of different graphics formats htape3x reads the history tapes optionally generated by mcnpx and performs post processing on them makexs a cross section library management tool that converts type 1 cross sections to type 2 cross sections and vice versa xsex3 a utility associated with the new cross section generation mode for mcnpx which allows tabulation of cross section sets based on physics models include contains include files shared across directories and include files localized in subdirectories mcnpx the organizing root directory for the mcnpx program Third Level cem dedx etc directories that organize the Fortran90 and C source code files that are related to different aspects of the MCNPX program Fourth Level individual Fortran90 and C source code files for a particular aspect of MCNPX 18 MCNPX User s Manual MCNPX User s M
17. e energy angle correlated spectra for secondary light particles energy spectra for gammas and heavy recoil nuclei The lower energy neutron libraries do not always contain complete secondary charged particle emission data since they are based on earlier evaluations In these cases the library processing routines ignores the incomplete information Therefore the secondary charged particles be produced and tracked below 20 MeV only for certain isotopes Thresholds for particle emission are given in Table 4 4 Table 4 4 Charged Particle Production Thresholds for Low Energy Neutron Libraries MeV Isotope ZAID Proton Deuteron Triton Alpha H 1 1001 24c 1 0E 11 H 2 1002 24c 3 339 1 0E 11 Be 9 4009 24c 14 266 16 301 11 709 0 667 C 6000 24c 20 0 20 0 20 0 N 14 7014 24c 20 0 20 0 20 0 O 16 8016 24c 20 0 20 0 20 0 Al 27 13027 24c 1 897 6 274 11 29 3 25 Si 28 14028 24c 4 0 20 0 20 0 2 746 Si 29 14029 24c 3 0 20 0 20 0 1 3 Si 30 14030 24c 8 012 20 0 20 0 4 345 P 31 15031 24c 20 0 20 0 20 0 Ca 20000 24c 20 0 20 0 20 0 20 0 Cr 50 24050 24c 1 0 20 0 20 0 2 25 Cr 52 24052 24c 3 256 20 0 20 0 1 233 Cr 53 24053 24c 2 69 20 0 20 0 1 0 Cr 54 24054 24c 6 33 20 0 20 0 1 581 Fe 54 26054 24c 0 7 20 0 20 0 3 0 Fe 56 26056 24c 2 966 20 0 20 0 0 862 MCNPX User s Manual 49 MCNPX User s Manual E Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium Table 4 4 Charge
18. files at all levels After successful configuration you can now make mcnpx using your new compiler with the following command from the top level of your build directory make mcnpx 3 1 7 3 How to add a new feature via with Example 2 Add a new option to the configure script that will activate the use of the old cross section capability during the compilation of mcnpx by defining the symbol OLDM for the compiler to recognize yes it s already there but we will step through it This one is requires the use of mcnpx_2 3 0 config aclocal m4 and all of the configure in files at the various levels 30 MCNPX User s Manual MCNPX User s Manual Ap Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium configure in in the menpx_2 3 0 directory configure in in the mcnpx_2 3 0 sre directory configure in in the mcnpx_2 3 0 src mcnpx directory configure generic in in the mcnpx_2 3 0 config directory Examine one of the configure in files There are several examples of checking for options such as compiler link method and debug via the AC_ARG_WITH macro Decide where the new call to with OLDXS should be placed Since it is only going to define one extra symbol for the compile step it could probably be placed anywhere after the initial default environment settings have been done AC_CLL_DEFAULTS and before the environment variable adjustments will be made AC_ENV_FLAGS_VARS for the detected and reques
19. t y y 0 KX On X axis Se Ry z tet K ou yer NGaRy y 9 t z 2 0 Bes n Z axis Z Jy 27 t x x 0 7 2 1 ix 2 t y 0 re Ax y t z 2 0 1 used only for 1 sheet cone SQ Ellipsoid Axis not parallel 52 2 232 ABCDE Hyperboloid to X Y or Z axis nee TEYA TMERR FGx yz Paraboloid Pee Keely y 2F z z G 0 GQ Cylinder Axes not parallel eE O ebay EN ABCDE Cone to X Y or Z axis FGHJK Ellipsoid Fzx Gz Hy Jz K 0 Hyperboloid Paraboloid MCNPX User s Manual 61 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 TX Elliptical or 2 Xv z sirqular X X B lly 2 2 A C 1 C papas TY torus a es Pe ae 2 n2 zy ZABC a 1 y z Axis is y y BY N x x z 2 A C 1 C TZ Parallel to PTEN NAE E r E E SA z z B x X A Cf 1 LY orz 4 y Y xy zABC axis IXYZP Surfaces defined by points See sections 5 3 2 2 Example 1 j PY3 This describes a plane normal to the y axis at y 3 with positive sense for all points with y gt 3 Example 2 j KYO O 2 25 1 This specifies a cone whose vertex is at x y z 0 0 2 and whose axis is parallel to the y axis The tangent of the opening angle of the cone is 0 5 note that z is entered and only the positive right hand sheet of the cone is used Points outside the cone have a positive sense 5 3 2 2 Axisymmetric Surfaces Defined by Points Form j n a
20. the compile step for the gen erated Makefiles this option can be used in combination with other options such as with FC and with CC configure will search for a Fortran77 compiler and use the first one it finds this option can be used in combi nation with other options such as with DEBUG and with CC with CC value sub stitute the desired C compiler name for the value placeholder e g with CC gcc to use the gcc compiler value will be used to compile C source code location of binary directory containing value must be in your PATH environment variable configure will search for a C compiler and use the first one it finds this option can be used in combination with other options such as with DEBUG and with FC with LD value Sub stitute the desired link editor for the value placeholder e g with LD usr ccs bin ld to use the Standard Sun linker MCNPX User s Manual value will be used to link object code Unlike the with FC and with CC options whose names are used for more than just find ing the executable The value can be a full path to the loca tion of the desired Id program as well as being a single name like Id configure will search for a linker and use the first one it finds This is typically needed on systems with both a ven dor supplied compiler set and the GNU tool set In such cases there may be two ver sions of Id that mus
21. 100 0 0 0 83 optional PHYS E 10000001 1 1 84 optional PHYS H 100 00JOJO 85 optional PHYS lt pl gt 100 3J 0 other particles 86 e TMP 2 53 x 108 86 e THTME 0 87 optional COINC none 87 e neutron problems only optional CUT lt pl gt huge 0 0 0 5 0 25 min sre wt 88 optional ELPT cut card energy cutoff 88 optional NPS none 89 optional CTME none 90 optional LCA 211002311010 91 optional LCB 2500 2500 800 800 1 0 1 0 93 optional LEA 14101001 95 optional LEB 1 5 8 0 1 5 10 0 96 Source specification cards section 5 6 on page 97 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 113 Summary of MCNPX Input Cards optional SDEF ERG 14 TME 0 POS 0 0 0 WGT 1 PAR N 97 optional Sin H Ij Ik 99 optional SPn D P Px 99 optional SBn D Bi Bk 100 optional DSn HJ Jg 101 optional SCn none 102 b KCODE 1000 1 30 130 MAX 4500 2 NSRCK 0 102 6500 1 none c KSRC none 102 optional SSW SYM 0 103 optional SSR OLD NEW COL m 0 104 optional SOURCE amp SRCDX 107 b neutron criticality problems only c KCODE only Tally specification cards section 5 7 on page 111 optional Fna Ro O0forn 5 112 optional FCn none 121 optional En very large 122 optional Tn very large 122 optional Cn 1 122 optional FQn FDUSMCET 123 optional FMn 1 124 optional DEn DFn none 126 optional EMn 1 128 optional
22. 1996 MCNPX User s Manual 189 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 190 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 7 Appendix A Examples Example 1 Neutron production from a spallation target One of the fundamental quantities of interest in most spallation target applications is the number of neutrons produced per beam particle incident on target For targets fed by proton accelerators this quantity is typically denoted as n p Here we demonstrate how one goes about calculating this quantity for a simple target geometry using MCNPX The geometry consists of a simple right circular cylinder of lead 10 cm in diameter by 30 cm long A beam of 1 GeV protons is launched onto the target The beam has a spot size of 7 cm diameter with a parabolic spatial profile see Fig A 1 Figure A 1 Neutron production from a spallation target In MCNPX net neutron production is tallied implicitly and is provided by default in the problem summary for neutrons The problem summary shows net neutron production resulting from nuclear interactions this is the component that accounts for neutron production by all particles transported using INC Preequilibrium Evaporation physics and net production by n xn reactions these are neutrons created in inelastic nuclear interactions by neutrons below the transition energy using evaluated nuclear data Net producti
23. 3 ay 0 9 ap 0 2 a3 0 01 p r po fexp r c a 1 c 1 07A18 fm a 0 545 fm p r ajp 0 i 1 16 p r po exp r c a 1 c 1 07A1 fm a 0 545 fm Pn r pp r N Z p r aip 0 i 1 7 a4 0 95 a2 0 8 ag 0 5 a4 0 2 as 0 1 ag 0 05 a7 0 01 Nucleon Potential Vy Tf By Nucleon kinetic energy Vu Tf By Ty dependent potential Vu Vi 1 Th Tmax Pion Potential Vx Vy Vr 0 Vx 25 MeV Mean Nucleon Binding By 7 MeV initial By from mass table By 7 MeV Energy the same value is used throughout the calculation Elementary Cross Sections standard BERTINI INC old standard ISABEL old new CEM97 last update March 1999 A A interactions not considered allowed not considered yA interactions not considered not considered may be considered Condition for passing from the INC stage cutoff energy 7 MeV different cutoff energies for p and n as in VEGAS code P Wmod Wexp Wexp P 0 3 Nuclear density depletion not considered considered not considered Pre equilibrium stage MPM LAHET model MPM LAHET model Improved MEM CEM97 Equilibrium stage Dresner model for n p d t 3He He emission fission y Dresner model for n p d t 3He He emission fission y CEM97 model for n p d t He fHe emission fission y Level density 3 LAHET models for a a Z N E 3 LAHET models
24. 609 623 2000 COU97 J D Court Combining the Results of Multiple LCS Runs memo LANSCE 12 97 43 Los Alamos National Laboratory May 8 1997 COU97a J D Court More Derivations Combining Multiple Bins in a MCNP or LAHET Tally memo LANSCE 12 97 66 Los Alamos National Laboratory July 16 1997 CMU94 Carnegie Mellon University Software Engineering Institute The Capability Maturity Model Guidelines for Improving the Software Process Addison Wesley 1994 DRE81 L Dresner EVAP A Fortran Program for Calculating the Evaporation of Various Particles from Excited Compound Nuclei Oak Ridge National Laboratory Report ORNL TM 7882 July 1981 EVA55 R D Evans The Atomic Nucleus Robert E Krieger Publishing Co 1955 EVA98 T M Evans J S Hendricks An Enhanced Geometry Independent Mesh Weight Window Generator for MCNP 1998 Radiation Protection and Shielding Division Topical Conference on Technologies for the New Century Sheraton Music City Nashville TN vol l p 165 April 19 23 1998 FAS94a A Fasso A Ferrari J Ranft P R Sala G R Stevenson and J M Zazula FLUKA92 Proceedings of the Workshop on Simulating Accelerator Radiation Environments SARE1 Santa Fe New Mexico January 11 15 1993 A Palounek ed Los Alamos LA 12835 C p 134 144 1994 FAS94b A Fasso A Ferrari J Ranft and P R Sala FLUKA Present Status and Future Developments Proceedings of the IV Internatio
25. C Frankle G M Hale H G Hughes A J Koning R C Little R E MacFarlane R E Prael and L S Waters Cross Section Evaluations to 150 MeV for Accelerator Driven Systems and Implementation in MCNPX Nuclear Science and Engineering 131 Number 3 March 1999 293 M B Chadwick P G Young R E MacFarlane P Moller G M Hale R C Little A J Koning and S Chiba LA150 Documentation of Cross Sections Heating and Damage Part A Incident Neutrons and Part B Incident Protons LA UR 99 1222 1999 H G Hughes et al MCNPX for Neutron Proton Transport International Conference on Mathematics amp Computation Reactor Physics amp Environmental Analysis in Nuclear Applications American Nuclear Society Madrid Spain September 27 30 1999 S G Mashnik A J Sierk O Bersillon and T A Gabriel CCascade Exciton Model Detailed Analysis of Proton Spallation at Energies from 10 MeV to 5 GeV Nucl Instr Meth A414 1998 68 Los Alamos National Laboratory Report LA UR 97 2905 R E Prael and H Lichtenstein User Guide to LCS The LAHET Code System LA UR 89 3014 Revised September 15 1989 11 CONTENTS OF CODE PACKAGE Included are the referenced documents in 10 a and one distribution CD which contains a GNU compressed Unix tar file with the full source code for the MCNPX system executable files installation scripts and test sets for each of the supported architectures WinZIP 8
26. C Little R E MacFarlane R E Prael and L S Waters Cross Section Evaluations to 150 MeV for Accelerator Driven Systems and Implementation in MCNPX Nuclear Science and Engineering 131 Number 3 March 1999 293 CHA99b M B Chadwick P G Young R E MacFarlane P Moller G M Hale R C Little A J Koning and S Chiba LA150 Documentation of Cross Sections Heating and Damage Part A Incident Neutrons and Part Incident Protons Los Alamos National Laboratory Report LA UR 99 1222 1999 http t2 lanl gov data he html CHA81 A Chatterjee K H N Murphy and S K Gupta Pramana 16 1981 p 391 182 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 CHE76 V A Chechin and V C Ermilova The lonization Loss Distribution at Very Small Absorber Thickness Nucl Instr Meth 136 1976 551 CHE68 K Chen et al Phys Rev 166 1968 p 949 CLO83 P Cloth et al The KFA Version of the High Energy Transport Code HETC and the Generalized Evaluation Code SIMPEL Jul Spez 196 Kernforschungsanlage Julich GmbH MARCH 1983 CLO88 P Cloth et al HERMES A Monte Carlo Program System for Beam Materials Interaction Studies Kernforschungsanlage Julich GmbH Jul 2203 May 1988 COLOO G Collazuol A Ferrari A Guglielmi and P R Sala Hadronic Models and Experimental Data for the Neutrino Beam Production Nuclear Instruments amp Methods A449
27. Example PIKMT26000 55 1 102001 1 7014 0 29000 2 3001 2 3002 1 8016 1 This example results in normal sampling of all photon production reactions for 14N All photons from neutron collisions with Fe are from the reaction with MT identifier 102001 Two photon production reactions with Cu are allowed Because of the PMT parameters the reaction with MT identifier 3001 is sampled twice as frequently relative to the reaction with MT identifier 3002 than otherwise would be the case No photons are produced from 160 or from any other isotopes in the problem that are not listed on the PIKMT card MCNPX User s Manual 79 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 5 4 10 MGOPT Multigroup Adjoint Transport Option Form MGOPT MCAL IGM IPLT ISB ICW FNW RIM Table 5 30 Multigroup Adjoint Transport Option Keyword Description F for forward problem MCAL A for adjoint problem the total number of energy groups for all kinds of particles IGM in the problem A negative total indicates a special elec tron photon problem indicator of how weight windows are to be used 0 means that IMP values set cell importances Weight windows if any are ignored for cell importance splitting and Russian roulette default IPLT 1 means that weight windows must be provided and are transformed into energy dependent cell importances A zero weight window lower bound produces an impor tance equal to the
28. K Gudima S G Mashnik and V D Toneev Cascade Exciton Model of Nuclear Reactions Nucl Phys A 401 1983 329 HAL88 J Halbleib Structure and Operation of the ITS Code System in Monte Carlo Transport of Electrons and Photons edited by Theodore M Jenkins Walter R Nelson and Alessandro Rindi Plenum Press New York 1988 153 HENO00a J S Hendricks Advances in MCNP4C Radiation Protection for Our National Priorities Spokane Washington September 17 21 2000 LA UR 00 2643 HENOOb J S Hendricks Point and Click Plotting with MCNP Radiation Protection for Our National Priorities Sookane Washington September 17 21 2000 LA UR 00 2642 HENO1 J S Hendricks Superimposed Mesh Plotting in MCNP International Meeting on Mathematical Methods for Nuclear Applications M amp C 2001 American Nuclear Society Salt Lake City Utah September 9 13 2001 HENO2a J S Hendricks G W McKinney L S Waters H G Hughes E C Snow New MCNPX Developments LA UR 02 2181 12th Biennial Radiation Protection and Shielding Division Topical Meeting Santa Fe NM American Nuclear Society ISBN 8 89448 667 5 ANS Order No 700293 April 14 18 2002 184 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 HOW81 R J Howerton ENSL and CDRL Evaluated nuclear Structure Libraries UCRL 50400 Vol 23 Lawrence Livermore National Laboratory February 1981
29. MG w Pe am aw or ur ce nv IPSP ur 1 o Jo rR JR fo nN JN JO Jo fo fo 11 lo Jo rR rR Jo Jn n fo Jo Jo o 2 o Jo jr rm Nn n n Jo o Jo Jo 102 3 lo lo Jn Jo n Jo n n Jo n IN 103 o lo n R n fo In n Jo N JN 5 In n In Jo n lo n N n N JN 15 N In n R n fo n n JN N JN s n n In o n lo Jo In n N JN 108 N In n R n fo Jo n n N JN lo lo IR r lo In In Jo Jo fo lo 109 206 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table B 1 Applicability of Input Control Parameters Continued IOPT G wn Pe Rm Rm or ur ce nv RS ur 10 O O R R N N N O O O O 110 11 O N R R O N N O N N N 111 12 O N R R O N N O N N N 112 13 O O R O O N N O O O O 14 N N N O N N N N N N O 114 N N N R N N N N N N O 15 N N N O N O O N N N N 115 N N N R N O O N N N N 16 O N N 0 N O N N N N N 116 O N N R N O N N N N N R required O optional N not used IRS is optional with any value of IOPT IOPT defines the editing option to be applied as defined below For all but IOPT 13 100 may added to the basic option type to indicate that the tally over a list of cell surface or material numbers will be combined in a single tally Prefixing IOPT by a minus sign when allowed indicates an option dependent modifi
30. Table 5 1 Particles in MCNPX Low Kinetic greeks IPT Name of Particle Symbol Mass MeV Energy Cutoff x decayed on MeV production 21 neutral pion 0 z 134 9764 0 0 8 4 x 1017 22 kaon K k 493 677 0 52614 1 2386 x 108 23 Ko short 497 672 0 000001 0 8927 x 10 0 24 Ko long A 497 672 0 000001 5 17 x 108 25 pt g 1869 3 1 9923 1 05 x 104 26 p d 1864 5 1 0 415x108 28 g j 5278 7 5 626 1 54 x104 29 po b 5279 0 1 0 15x104 0 1 4 30 B q 5375 0 1 34 x 10 Light lons 31 deuteron d 1875 627 2 0 huge 32 triton t 2808 951 3 0 12 3 years 33 Helium 3 s 2808 421 3 0 huge 34 Helium 4 a a 3727 418 4 0 huge Particle tracking between interactions involves several physics considerations which are described below Atomic electron interactions will cause a charged particle to lose energy along its track length ionization Certain modifications to this energy loss are determined by energy straggling theory Multiple scattering of charged particles is also implemented Note that there is currently no delta ray production of knock on electrons for charged heavy particles now in MCNPX version 2 3 0 although it is present for electrons No option for electromagnetic field tracking is currently implemented in MCNPX Attempts are currently underway to develop this capability which will be fully implemented in a future version of the code FAV99 MCNPX User s Manual 67 MC
31. The MCNPX classes are a vital part of our code quality assurance program and we very much appreciate their help and support We would also like to thank members of the Los Alamos Export Controls Office particu larly Sarah Jane W Maynard Crystal Johnson and Steve H Remade for their outstanding help in dealing with the export issues for our foreign beta test team members Publishing Team Finally we wish to thank Berylene Rogers for copyediting and preparing the final docu ment and Patty Montoya Barbara Olguin Arlene Lopez and Jean Harlow for their help in reproducing and assembling the manual iv MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Dedication We dedicate this code to the memory of our respected colleague Dr Russell B Kidman Russ was an invaluable member of the APT Target Blanket design team and a computer simulations expert for many projects at Los Alamos His tragic and premature death has left us all with a deep sense of loss MCNPX User s Manual v LA CP 02 408 TABLE OF CONTENTS Acknowledgments cai rii rere be wd ka eaa a cheba eked iii DEdiCaliONns iini iarom oani nhn a a aa naa oa ee o Ea ia v PIGlaCe imoa oan aa a E E a eet esa A N xiii LAINTOdCUCHON cess Seen e a a a a a a aa 1 2 Warnings and Limitations 00 eeeeeeeeeeee 5 3 Installation ieie era wap ele lela le na wih Sw ee Soe a a ea 9 3 1 UNIX Build Systemi siec 2205 eee see ee ee ee eee
32. The code will use usr mcnpx lib as its default location for finding the data files When the user has an existing directory layout that does not follow the mcnpx default then the data path itself can be customized like this usr local src mcnpx_2 4 0 configure libdir usr mcnpx which will leave the default executable location as usr local bin and set the location for the data files to usr mcnpx Finally both the prefix and the libdir options can be used together with the libdir options taking precedence over the library directory implied by the prefix These options should remove the need to edit paths in the source code In fact with support for these options there are no longer any paths in the code to edit 3 1 3 3 Individual Private Installation For the purpose of the second illustration we will look at a single non privileged user Me on a computer loading and building a private copy of the code The local user building the private copy is username me whose home directory is the directory home me The user has fetched the distribution from CDROM or from the net and has it in the file home me mcnpx_2 4 0 tar gz The user will unload the distribution package into home me mcnpx_2 4 0 The user will build the system in the same directory as the source install the binary executable in home me bin and install the binary data files and eventually the mcnp cross sections in home me lib This method makes it hard
33. VOL x X Xi or VOLNO x X Xi Table 5 19 Cell Volume Card Argument Description volume of cell i where i 1 2 number of cells in the problem NO no volumes or areas are calculated Default MCNPxX attempts to calculate the volume of all cells unless NO appears on the VOL card If no value is entered for a cell on the VOL card the calculated volume is used Use Use only if required cell volumes are not properly calculated NOTE Ifthe number of entries does not equal the number of cells in the problem itis a fatal error Use the jump nJ feature to skip over cells for which you do not want to enter values 68 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 5 3 3 2 AREA Surface Area Form AREAX ie X j X n Table 5 20 Surface Area Card Argument Description area of surface i where i 1 2 number of surfaces in A the problem Default MCNP attempts to calculate the area of all surfaces If no value is entered for a surface on the AREA card the calculated area if any is used Use Use only if required surface areas for F2 tallies are not properly calculated Repeated Structures Cards 5 3 3 3 U Universe Form U n cell card entry or U n4 n2 nz n data card Table 5 21 Universe Card Argument Description arbitrary universe number Integer to which cell is assigne
34. and amp created but not transported Both can make TTB approxi mation photons Mode P no TTB energy of e e pair deposited locally e annihilated replaced by two photons Incoherent Compton Scattering Klein Nishina formula Regarded to be on free elec trons Uses form factors to account for electron binding effects MCNPX User s Manual 55 MCNPX User s Manual Version 2 3 0 April 2002 x LA UR 02 2607 Accelerator Production of Tritium Table 4 5 Summary of Photon Physics Options Continued Simple Detailed Process used above energy EMCPF used below energy EMCPF on the PHYS P card on the PHYS P card default 100 MeV default 100 MeV Coherent Thompson Scattering Not included Scattering angle of photon is Involves no energy loss therefore no computed and transport con electrons are produced for further tinues transport a In analog capture a particle is killed with probability equal to the ratio of the absorption oa to the total cross section oT Killed particles deposit their entire energy and weight in the collision cell b In implicit capture the particle weight Wh is reduced to W as follows W a 1 0 07 x Wp If the new weight W n is below the problem weight cutoff as specified on the CUT card the particle is rouletted resulting in fewer particles with larger weights A fraction oa oT will be deposited in the collision cell corresponding to t
35. and A S Tishin Sov J Nucl Phys 21 1975 p 256 JAN82 J F Janni Proton Range Energy Tables 1ke V 10GevV Atomic Data and Nuclear Data Tables 27 2 3 1982 KAL85 PRECO D2 Program for Calculating Preequilibrium and Direct Reaction Double Differential Cross Sections LA 10248 MS Los Alamos National Laboratory 1985 KOC59 H W Koch and J W Motz Bremsstrahlung Cross Section Formulas and Related Data Rev Mod Phys 31 1959 920 LAN44 L Landau On the Energy Loss of Fast Particles by lonization J Phys USSR 8 1944 201 MCNPX User s Manual 185 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 LITOO R C Little J M Campbell M B Chadwick S C Frankle J S Hendricks H G Hughes R E MacFarlane C J Werner M C White P G Young Modern Nuclear Data for Monte Carlo Codes LA UR 00 4979 MC2000 International Conference on Advanced Monte Carlo for Radiation Physics Particle Transport Simulation and applications Lisbon Portugal October 23 26 2000 MAC94 R E MacFarlane and D W Muir The NJOY Nuclear Data Processing System version 91 Los Alamos National Laboratory Report LA 12740 M October 1994 MAD88 amp D G Madland Recent Results in the Development of a Global Medium Energy Nucleon Nucleus Optical Model Potential in Proceedings of a Specialist s Meeting on Preequilibrium Reactions Semmering Austria February 10 12 1988 Edited
36. by B Strohmaier OECD lt Paris 1988 p 103 116 MAS74 S G Mashnik and V D Toneev MODEX the Program for Calculation of the Energy Spectra of Particles Emitted in the Reactions of Pre Equilibrium and Equilibrium Statistical Decays in Russian Communication JINR P4 8417 Dubna 1974 25 pp MAS98 S G Mashnik A J Sierk O Bersillon and T A Gabriel Cascade Exciton Model Detailed Analysis of Proton Spallation at Energies from 10 MeV to 5 GeV Nucl Instr Meth A414 1998 68 Los Alamos National Laboratory Report LA UR 97 2905 http t2 lanl gov publications publications html MCKOO G W McKinney T E Booth J F Briesmeister L J Cox R A Forster J S Hendricks R D Mosteller R E Prael A Sood MCNP Applications for the 21st Century Proceedings of the 4th International Conference on Supercomputing in Nuclear Applications September 4 7 Tokyo Japan 2000 MCK02 G W McKinney J S Hendricks L S Waters T H Prettyman Using MCNPX for Space Applications LA UR 02 2179 12th Biennial Radiation Protection and Shielding Division Topical Meeting Santa Fe NM American Nuclear Society ISBN 8 89448 667 5 ANS Order No 700293 April 14 18 2002 MOH83 H J Moehring Hadron nucleus Inelastic Cross sections for Use in Hadron cascade Calculations at high Energies CERN report TIS RP 116 October 1983 MOL48 G Moliere Theorie der Streuung schneller geladener Teilchen II Mehrfach
37. details BNUM Obremsstrahlung photons will not be produced gt 0 produce BNUM times the analog number of bremsstrahlung photons Radiative energy loss uses the bremsstrahlung energy of the first sampled photon gt 0 produce XNUM times the analog number of electron induced x rays ANUM 0 x ray photons will not be produced by electrons gt 0 produce RNOK times the analog number of knock on RNOK electrons 0 knock on electrons will not be produced 84 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 33 Electron Physics Options Keyword Description gt 0 produce ENUM times the analog number of photon induced secondary electrons ENUM 0 photon induced secondary electrons will not be produced NUMB gt 0 produce bremsstrahlung on each substep 0 nominal bremsstrahlung production Default PHYS E 100000011110 Use Optional 5 5 2 4 Protons Form PHYS h EMAX EAN TABL J ISTRG J RECL Table 5 34 Proton Physics Options Keyword Description EMAX Upper limit for proton energy MeV Analog energy limit MeV Implicit capture for E gt Ean EAN inetd implicit capture for E lt Ean Table based physics cutoff For E gt Tabl use model physics TABL E lt Tabl use physics from data tables WARNING If Tabl gt emax of a data table the cross section values at E emax will be used in the energy r
38. following form is required IFSEG NSEG FSEG 1 FSEG NSEG For IFSEG 1 the segmenting planes are perpendicular to the x axis for IFSEG 2 the y axis and for IFSEG 3 the z axis The FSEG I are the coordinates of the NSEG planes in increasing order Segmenting may also be accomplished by using segmenting cylinders The input has the same format as segmenting by planes however IFSEG negative designates cylindrical segmenting IFSEG 1 indicates that the segmenting cylinders are concentric with the x axis IFSEG 2 indicates that the segmenting cylinders are concentric with the y axis IFSEG 3 indicates that the segmenting cylinders are concentric with the z axis The values of the FSEG array are the radii of nested concentric cylinders and must be in increasing order Segmenting cylinders are concentric with an axis not just parallel For KOPT 4 or 9 an additional record must be supplied with the direction cosines of the arbitrary vector with which cosine binning is to be made The form of this record is CN 1 CN 2 CN 3 where the parameters input are the direction cosines of the arbitrary vector with respect to the x y and z axes The vector need not be normalized The surface current tally represents the time integrated current integrated over a surface area and an element of solid angle Unless otherwise normalized it is the weight of particles crossing a surface within a given bin per source particle As such i
39. gt sc 3 5 E Any particle generated within this cell is accepted any outside of the cell is rejected Any well defined surface may be selected and it is common to use a simple cylinder to represent the extent of a beampipe In this example a source is generated in an x y coordinate system with the distribution centered at the origin and the particles travelling in the z direction The particle coordinates can be modified to an x y coordinate system by translation and rotation according to the following equations where 0 lt 9 lt x x X sing y cosd Xo MCNPX User s Manual 109 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 y X coso y sing Yo Thus the angle is the angle of rotation of the major axis of the source distribution from the positive y direction in the laboratory coordinate system If cos 0 0 the angle is 90 and the major axis lies along the x axis The TRn card in the above example implements this rotation matrix however the user is warned that o in the TRn card is equal to I L 5 Defining Multiple Beams The opportunity to specify a probability distribution of transformations on the SDEF card is a new feature that goes beyond enabling the representation of LAHET beam sources It allows the formation of multiple beams which differ only in orientation and intensity a feature that may have applications in radiography or in the distribution of p
40. lowest IPT number or symbol repre sented on MODE card Example 1 SDEF no entries This card specifies a 14 MeV isotropic point source at position 0 0 0 at time 0 with weight 1 all defaults Example 2 SDEF par SF Cel d1 Pos d2 Rad Fpos d3 98 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Spontaneous fission source Source points will be selected in spheres Pos Rad and limited to fission cells by Cel Each sampled source point will be a spontaneous fission site par SF producing the appropriate number of spontaneous fission neutrons per fission at the appropriate energy with isotropic direction 5 6 1 1 Sin Source Information Form SIn option I Table 5 46 Source Information Card Variable Description distribution number n 1 999 from corresponding dis tribution number on SDEF card Sets how the l s are interpreted Allowed values are blank or H histogram bin upper boundaries scalar only L discrete source variable values opion A points where a probability density distribution is defined S distribution numbers I Ir source variables or distribution numbers Default SInH Ik 5 6 1 2 SPn Source Probability Form SPn option P Pk or SPn fa b Table 5 47 Source Probability Card Variable Description distribution number 1 999 from corresponding distribu tion number on SDEF and SI cards
41. options A summary of the cards follows The options controlling the Bertini and ISABEL physics modules are taken from the User Guide to LCS PRA89 The user is referred to that document for further information CEM allows neutrons and protons up to 5 GeV and pions up to 2 5 Gev to initiate nuclear reactions Valid targets are nuclei with a charge number greater than 5 and a mass num ber greater than 11 The light nuclei are passed to the Bertini ISABEL models that use the Fermi Breakup model in this regime CEM consists of an intranuclear cascade model fol lowed by a pre equilibrium model and an evaporation model Possible fission events are initiated in the equilibrium stage for compound nuclei with a charge number greater than 70 The fragmentation of the fission event is handled by modules from the RAL fission model Fission fragments undergo an evaporation stage that depends on their excitation energy After evaporation a de excitation of the residual nuclei follows generating gam mas using the PHT data 76 MCNPX User s Manual MCNPX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium Future developments of MCNPX will allow greater freedom in the selection of physics options INC pre equilibrium evaporation fission etc so the user may compare the effect of varying one parameter at a time In version 2 3 0 CEM is still relatively self contained All of the input values on the four
42. s Manual 87 Accelerator Production of Tritium 88 MCNPX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 MCNPX User s Manual MCNPxX User s Manual z Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium 7 New Variance Reduction Techniques The high energy cascade process generates numerous particles over a very broad range of energies at Super Collider energies 20 TeV 20 TeV proton collisions the average number of particles generated for a central collision is 19 000 This is a far different sit uation from what the typical low energy MCNP user is accustomed and standard methods such as fixed cell importance biasing applied equally to all particles is not always the best solution At a minimum one should consider biasing in both spatial cells and energy groups and the complexity of the problem leads one to consider semi automatic schemes such as the weight window generator DSA etc Special variance reduction techniques have also been developed in the industry to enhance the production of particles of interest One example is Leading Particle biasing where production of only the highest energy most promptly produced particles is enhanced In addition one cannot assume isotropy of particle emission at high energies and the actual emission pattern varies over a wide range This anisotropy causes problems in using detector techniques for neutral particles above library energies Closel
43. these settings will override the system default or system computed values if omitted the default behav ior is system dependent the detected hardware software platform and compilers deter mine what the default FOPT should be MCNPX User s Manual 25 MCNPxX User s Manual E Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium Table 3 1 Configure Script Parameters Option Syntax Effect on the generated Effect on the generated p y Makefile if requested makefile if NOT requested with COPT value substitute a quoted or double if omitted the default behav quoted string for value that ior is system dependent the There is a separate variable that represents allowable compiler detected hardware software is used for non optimization switches See with CFLAGSin Optimization switch settings platform and compilers deter this table If in doubt run the con these settings will override mine what the default COPT figure script and examine the houl system default or system com the system default or system should be puted values that appear in the computed values generated Makefile h You may want to include the defaults in the string you specify for COPT with this mechanism COPT settings are always appended to CFLAGS settings when configure is run again 3 1 6 Multiprocessing Many users have requested full multiprocessing including the b
44. ticle In the F6 and F6 tallies material density is available for the chosen cells and normalization is MeV gm source particle 112 MCNPX User s Manual MCNPxX User s Manual i Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium 8 4 Dose Conversion Coefficients The health physics industry and regulatory authorities have published a variety of fluence to dose conversion coefficients and it can be difficult for the unexperienced user to keep track of the latest versions In addition much new work is in progress for providing coeffi cients for particles other neutrons and photons as well as extending the limits of their upper ranges to the high energies needed in many accelerator applications A new function has been added to MCNPX which contains a number of standard dose conversion coefficients and efforts are being made to include the option to call this func tion in various tallies In MCNPX version 2 3 0 this function is directly used through the dose keyword of the Type 1 Mesh Tally section 8 1 1 If access to the MCNPX source code is available the user can add additional factors although this can also be done by individually inputting values into the de dfcards Func tion DFACT is an effort to hardwire in standard values since user input can be notoriously subject to error The MCNPX code developers will add more options as they become avail able The acr option can also be modified to add u
45. which attempts to score energy deposition by following individual particles 15 Continue Runs that include Mesh Tallies must use the last available complete restart dump The output file for mesh tallies is not integrated into the restart dump file Runtpe However they are written at each dump cycle Since the mesh tally file is overwritten at each dump care must be taken to ensure that the files used to continue a run were generated at the same dump cycle and that the last complete dump on the Runtpe file is used 16 An old version of FLUKA is implemented in MCNPX version 2 3 0 The version of FLUKA now in MCNPX is taken directly from the LAHET version 2 8 code and is known as FLUKA87 Only the high energy portion of FLUKA is present to handle interactions above the INC region This is not the latest version of FLUKA and does not contain any of the FLUKA code improvements added since that time See Section MCNPX User s Manual 7 MCNPxX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium 4 2 for further information The FLUKA code module will be upgraded in a future ver sion of MCNPX The contents of the HISTP file arising from interactions processed by the CEM module do not distinguish among evaporation particles emitted before or after fission All are labeled as pre fission Therefore the HTAPE edits that depend on this distinction will not produce the intended output e pre
46. with CC cc with LD usr ccs bin Id with DEBUG prefix home me libdir usr mcnpx data now make the executable mcnpx program We will omit the regression tests this time although it would be a good idea to run them again if different compiler optimization values are used make install That s all there is to it There are many other options available with this new version of mcnpx Please read the User s Notes or the Programmer s Notes for more details 3 1 4 Directory Reorganization In order to accommodate the use of the autoconf utility to generate the Makefiles it became necessary to arrange the source code and regression test directories a bit We also added a config directory to hold autoconf related code The new directory structure is depicted in Figure 3 1 Each of the levels contains a collection of autoconf files and links Removal of any of these files will break the automated configure and make capabilities First Level Data contains data used with the bertin phtlib makexs targets Docs con tains files describing this mcnpx distribution Test contains the regression test files for the various known platforms in use src contains the source code files for mcnpx and several related utilities miscellany contains things that don t fit into any other category of interest to developers config contains autoconf related macros scripts initialization files Second Level bertin builds and executes a progr
47. 0 September 2002 LA CP 02 408 Table 5 113 Summary of MCNPX Input Cards required Data cards plus blank terminator 31 optional C Comment card aA Problem type card Geometry cards section 5 3 on page 58 required Cell cards plus blank terminator 34 58 required Surface cards plus blank terminator 31 60 optional VOL 0 68 optional AREA 0 69 optional U 0 69 optional TRCL 0 71 optional LAT 0 72 optional FILL 0 70 optional TRn none 73 Material specification cards section 5 4 on page 74 optional Mm no ZAID default 0 set internally first match 7A in XSDIR 01p 01e e MTm none 76 MPNm 77 d DRXS fully continuous 81 d TOTNU prompt v for non KCODE total v for 77 KCODE d NONU fission treated as real fission 77 optional AWTAB _ atomic weights from cross section tables 78 optional XSn none 78 optional VOID none 78 optional PIKMT no photon production biasing 79 MCNPX User s Manual 175 176 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 113 Summary of MCNPX Input Cards optional MGOPT fully continuous 80 optional DRXS 81 d neutron problems only Physics Cards section 5 5 on page 82 a MODE lt pl gt 4 82 a Required for all but MODE N optional PHYS N huge 0 O 20 0 O 82 optional PHYS P
48. 0 QO QO gamma xn 0 0 0 particle decay 0 0 0 adjoint splitting 0 0 0 total 395962 1 9794E 01 3 4090E 02 total 395962 1 9794E 01 3 4090E 02 number of neutrons banked 370423 average time of shakes cutoffs neutron tracks per source particle 1 9798E 01 escape 5 7616E 00 tco 1 0000E 34 neutron collisions per source particle 2 7981E 01 capture 4 8708E 01 eco 0 0000E 00 total neutron collisions 559626 capture or escape 5 7574E 00 wcl 5 0000E 01 net multiplication 0 0000E 00 0000 any termination 5 3337E 00 we2 2 5000E 01 Calculated net neutron production for this case is 18 335 and examination of the net nuclear interactions and net n xn figures show very similar results to the base case The implication of this result is that we need not concern ourselves with light ion transport if the quantity with which we concerned is related solely to neutrons as neutron production by light ions is small when we start with a proton beam Case 3 In this variation we replace the Bertini INC model used in the base case for the simulation of nucleon and pion interactions with nuclei by the ISABEL INC model in this example both INC models utilize the same GCCI level density model We invoke the ISABEL INC model by including in the input deck the following card Base Case lca Case 3 lea j j 2 This changes the value of the variable IEXISA third value on the Ica card from its default value of 1 to 2 The neutron problem summary fo
49. 0 is required to expand this file under Windows 12 DATE OF ABSTRACT September 2002 KEYWORDS CHARGED PARTICLES COMPLEX GEOMETRY ELECTRON GAMMA RAY HIGH ENERGY KAON MONTE CARLO NEUTRON PION PROTON RADIOGRAPHY SPALLATION WORKSTATION Owor o2z b MCNPxX User s Manual Version 2 4 0 September 2002 LA CP 02 408 MCNPX USER S MANUAL Version 2 4 0 September 2002 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Disclaimer This report was prepared as an account of work sponsored by an agency of the United States Government Neither the United States Government nor any agency thereof nor any of their employees makes any warranty express or implied or assumes any legal liability or responsibility for the accuracy completeness or usefulness of any information apparatus product or process disclosed or represents that its use would not infringe privately owned rights Reference herein to any specific commercial product process or service by trade name trademark manufacturer or otherwise does not necessarily constitute or imply its endorsement recommendation or favoring by the United States Government or any agency thereof The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States government or any agency thereof ii MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Acknowledgments
50. 117 0 13000 1 22400D 04 2 50000D 01 0 4331 48 119 0 14640 1 62000D 02 6 00000D 01 0 2329 MCNPX User s Manual 145 MCNPX User s Manual Ap Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium 11 Edit Option IOPT 9 or 109 Surface Current with Collimating Window Option 9 is identical to option 1 except that a rectangular or circular window is imposed on each surface and the tally made within and without the window The window is defined by the intersection of a rectangular or circular tube parallel to the x y or z axis with the tally surface A window definition record appears in place of the segmenting record of option 1 For KOPT 0 1 2 3 or 4 the window is formed by the rectangular tube the window record has the following allowed forms parallel to x axis 1 y min y max z min z max parallel to y axis 2 z min z max x min x max parallel to z axis 3 x min x max y min y max For KOPT 5 6 7 8 or 9 the window is formed by a circular tube cylinder the window record has the following allowed forms parallel to x axis 1 y center z center radius parallel to y axis 2 z center x center radius parallel to z axis 3 x center y center radius 12 Edit Option IOPT 10 or 110 Surface Flux with Collimating Window Option 10 is identical to option 2 except that the edit is performed inside and outside a win dow defined as in option 9 Instead
51. 2 3 0 Particles will be transferred to the transportable category in future versions as appropriate models of interaction physics are obtained Obviously MCNPX has only one character to designate particle symbols therefore we had to resort to symbols after the regular alphabet ran out Output tables in the MCNPX OUTP file have been extended to support the additional tracked particles in a straightfor ward manner The list of particle properties as well as decay branching ratios for non tracked particles is derived from the 1998 Review of Particle Physics PDG98 The publication of the Review of Particle Physics is supported by the US Department of Energy the US National Science Foundation the European Laboratory for Particle Physics CERN by implement ing arrangement between the government of Japan and the United States on cooperative research and development and by the Italian National Institute of Nuclear Physics INFN It represents the current standard of international agreement on particle physics properties Table 5 1 Particles in MCNPX ae Mean Lifetime Low Kinetic seconds IPT Name of Particle Symbol Mass MeV Energy Cutoff decayed on MeV 5 production Original MCNP Particles 1 neutron n n 939 56563 0 0 887 0 1 anti neutron n n 939 56563 0 0 887 0 3 electron e e 0 511008 0 001 huge 3 positron e e 0 511008 0 001 huge Leptons 4 muon w l 105 65
52. 2 4 0 September 2002 LA CP 02 408 Table 5 36 Free Gas Thermal Temp Keyword Description n index of time on the THTME card Tin temperature of i cell at time n in MeV I number of cells in the problem Default 2 53 x 10 8 MeV room temperature Use Optional Required when THTME card is used Needed for low energy neutron transport at other than room temperature A fatal error occurs if a zero temperature is specified for a non void cell 5 5 4 THIME Thermal Times Form THTME ty to th ty Table 5 37 Thermal Times Keyword Description time in shakes 108 sec at which thermal temperatures tn are specified on the TMP card N total number of thermal times specified Default Zero temperature is not time dependent Use Optional Use with TMP card 5 5 5 COINC He Detector Coincidence Form COINC n h4 Ig lg Cell number for 3He coincidence detectors Cells listed on the COINC card neutrons only must contain 3He and the problem must be run in analog mode Print Table 118 will tabulate the weight and number of 3He captures per history along with the factorial moments for each listed cell This feature is proprietary to the sponsor and is available only in executable code versions until 4 1 03 MCNPX User s Manual 87 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Default 3He detector coincidence moments not tabulated Use Use whe
53. ANSI ANS 6 1 1 1997 31 ANSI ANS 6 1 1 1991 AP anterior posterior 32 ANSI ANS 6 1 1 1991 PA posterior anterior 33 ANSI ANS 6 1 1 1991 LAT side exposure 34 ANSI ANS 6 1 1 1991 ROT normal to length amp rotationally symmetric 35 ANSI ANS 6 1 1 1991 ISO isotropic aa oN AP en Particle energy Interpolation method 1 logarithmic interpolation in energy linear in function 2 linear interpolation in energy and function 3 recommended analytic parameterization not available for ic 10 units of the result 1 rem hr particles cm sec 2 sieverts hr particles cm2 sec MCNPX User s Manual 151 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 85 DFACT Argument Descriptions Continued ARGUMENT DESCRIPTION Normalization factor for dose DFACT result will be multiplied by any factor greater or equal to 0 0 for exam ple acr 1 0 means no change The value must be a real number acr Certain special options are also available 1 0 normalize DFACT results to Q 20 by dividing out the parametric form of Q which equals 5 0 17 0 exp In 2E 2 6 from ICRP60 1990 para graph A12 2 0 Apply LANSCE albatross response function 5 7 22 7 Processing the Mesh Tally Results The values of the coordinates the tally quantity within each mesh bin and the relative errors are all written by MCNPX to an unformatted binary fi
54. Accelerator Production of Tritium for most LA150 materials contains an extensive set of global tabular and graphical repre sentations of the new data tables 4 3 1 2 Photonuclear Production Data Recently work has begun on a program to evaluate photonuclear cross sections to 150 MeV for a range of materials important in accelerator components bremsstrahlung and spallation targets and shielding applications Considerable interest has also been shown by researchers involved in lower energy applications particularly the medical industry Until now such data have not been available in the ENDF B VI data library nor have radi ation transport codes such as MCNPX been able to use photonuclear data in a fully coupled manner The GNASH code has been extended to include photonuclear processes using a giant dipole resonance mechanism below 20 30 MeV and a quasideuteron mechanism at higher energies YOU98 Table 4 3 summarizes the currently available evaluations which are released with MCNPX version 2 3 0 Data for all secondary particle channels are included but particular emphasis has been placed on high accuracy neutron production cross sections We also note that data on photonculear interactions above 150 MeV will eventually be included in the CEM physics modules Work is now complete on the implementation of the new photonuclear data and physics into MCNPX version 2 3 0 WHI99 If photonuclear physics is enabled in a simulation see
55. F6 tally to be in error 2 Note that the Pi if included on the MODE card will be transported before it decays even though its life time is 8 4 x 10 seconds This allows the user to use MCNPX tallies for that particle MCNPX User s Manual 111 MCNPX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium b modehnp dtsau file runtpf tally 36 20 9 tally mev particle 10 20 10 11 sal i 100 200 300 aao energy nev In MCNPX version 2 3 0 the two forms of the F6 tally are F6 P C1 C2 Cm F6 C1 C2 Cm Table 8 8 Energy Deposition Card Argument Descriptions Argument Description C1 C2 Cell numbers in which to score energy deposition MCNPX has the standard F6n P tally where P can now be any particle In addition MCNPX has a new F6n tally which contains energy deposition from all particles in the problem It is not currently possible to have an F6 tally which will do energy deposition for more than one but less than all particles We will consider adding this capability in the future Note that the pedep keyword in a Type 1 Mesh Tally is analogous to the F6n P tally and the Type 3 Mesh Tally is analogous to the F6n tally although the normalizations will be different Since the mesh tallies score energy deposition within a mesh cell which may contain more than one material normalization is The units of this tally are MeV source par
56. ForNTYPE gt 0 a record containing NTYPE particle types in any order defined as the array ITIP I lI I NTYPE In the present MCNPX version 2 3 0 the contents of a sur face source file WSSA are insufficient to distinguish between a particle and its anti particle it is to be expected that this condition will be remedied in future releases of MCNPx The allowed particle types are listed in Table D3 which also indicates the overlapping particle antiparticle tally definition which follows the column MCNPX Usage For NPARM gt 0 a record containing NPARM user defined cell material or surface numbers integers in any order for which one wishes a tally to be made these are defined as the array LPARM I l 1 NPARM If a null record is supplied with NPARM gt 0 it is treated as 1 2 3 NPARM Note a different meaning for NPARM is used for lOPT 13 e ForNFPRM gt 0 a record containing NFPRM upper cosine bin boundaries defined as the array FPARM I l 1 NFPRM The first lower cosine boundary is always 1 0 If a null record is supplied equal cosine bin boundaries from 1 0 to 1 0 will be defined by default MCNPX User s Manual 141 MCNPX User s Manual Ap Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium e If NPARM is preceded by a minus sign a record containing NPARM or NPARM 1 nor malization divisors these are defined in HTAPE3X as the DNPARM array The NPARM values are in a one
57. GeV or 1 GeV per nucleon for composite particles although it may execute at higher energies MCNPX User s Manual 91 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 41 LCA Keyword Descriptions Continued Keyword Description ICHOIC 4 integers ijkl which control ISABEL INC Model default 0023 i 0 Use partial Pauli blocking i 1 Use total Pauli blocking i 2 No Pauli blocking not recommended j 0 No interaction between particles already excited above the Fermi Sea j gt 0 Number of time steps to elapse between such CAS CAS interactions k 0 Meyer s density prescription with 8 steps k 1 Original isobar density prescription with 8 steps k 2 Krappe s folded Yukawa prescription for radial density in 16 steps with a local density approximation to the Thomas Fermi distribution for the sharp cutoff momentum distribution k 3 The same as k 0 but using the larger nuclear radius of the Bertini model k 4 The same as k 1 but using the larger nuclear radius of the Bertini model k 5 The same as k 2 but using the larger nuclear radius of the Bertini model 1 Reflection and refraction at the nuclear surface but no escape cutoff for isobars 2 Reflection and refraction at the nuclear surface with escape cutoff for iso bars 3 No reflection or refraction with escape cutoff for isobars 4 The same as l 1 but using a 25 MeV potential well for pions 5 The
58. HUG95 H G Hughes and L S Waters Energy Straggling Module Prototype Los Alamos National Laboratory Memorandum XTM 95 305 U November 29 1995 HUG97 H G Hughes R E Prael R C Little MCNPX The LAHET MCNP Code Merger XTM RN U 97 012 April 22 1997 HUG98a H G Hughes K J Adams M B Chadwick J C Comly L J Cox H W Egdorf S C Frankle F X Gallmeier J S Hendricks R C Little R E MacFarlane R E Prael E C Snow L S Waters P G Young Jr Status of the MCNP M LCS Merger Project American Nuclear Society Radiation Protection and Shielding Topical Conference April 19 23 1998 Nashville TN HUG98b H G Hughes et al Recent Developments in MCNPX American Nuclear Society Topical Meeting on Nuclear Applications of Accelerator Technology Gatlinburg TN Sept 20 23 1998 HUG98c H G Hughes et al MCNPX Code Development 4th Workshop on Simulating Accelerator Radiation Environments Knoxville TN September 14 1998 HUG99 H G Hughes et al MCNPX for Neutron Proton Transport International Conference on Mathematics amp Computation Reactor Physics amp Environmental Analysis in Nuclear Applications American Nuclear Society Madrid Spain September 27 30 1999 ICR84 International Commission on Radiation Units and Measurements ICRU Report 37 Stopping Powers for Electrons and Positrons October 1984 IGN75 A V Ignatyuk G N Smirenkin
59. MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 nonorm tally 101 free e loglog xlims 0 1 1000 ytitle protons MeV file free c linlog xlims 1 0 1 0 ytitle protons steradian file tally 102 free e loglog xlims 0 1 1000 ytitle neutrons MeV file free c linlog xlims 1 0 1 0 ytitle neutrons steradian file tally 103 free e loglog xlims 0 1 1000 ytitle pi MeV file free c linlog xlims 1 0 1 0 ytitle pit steradian file tally 104 free e loglog xlims 0 1 1000 ytitle pi0 MeV file free c linlog xlims 1 0 1 0 ytitle pi0 steradian file tally 105 free e loglog xlims 0 1 1000 ytitle pi MeV file free c linlog xlims 1 0 1 0 ytitle pi steradian file tally 106 free e loglog xlims 0 1 1000 ytitle deuterons MeV file free c linlog xlims 1 0 1 0 ytitle deuterons steradian file tally 107 free e loglog xlims 0 1 1000 ytitle tritons MeV file free c linlog xlims 1 0 1 0 ytitle tritons steradian file tally 108 free e loglog xlims 0 1 1000 ytitle He 3 MeV file free c linlog xlims 1 0 1 0 ytitle He 3 steradian file tally 109 free e loglog xlims 0 1 1000 ytitle alphas MeV file free c linlog xlims 1 0 1 0 ytitle alphas steradian file tally 110 free e loglog xlims 0 1 100 ytitle photons MeV file free c linlog xlims 1 0 1 0 ytitle photons steradian file end 232 MCNPX User s Manual Owor o2z N Zz MCNPX User s Manual Accelerator Version 2 3 0 April 2002
60. MCTAL format plot file with default name XSTAL These file names may be changed by file replacement The most general execute line has the format XSEX3 INXS OUTXS HISTP XSTAL 4 Plotting Output from XSEX3 The source code for XSEX3 contains a plotting package using the LANL Common Graph ics System the latter is not generally available outside of Los Alamos National Laboratory A new feature has been added for this release whereby a nonzero value for the input quan tity KPLOT will cause the writing of a file XSTAL in the format of an MCNPX MCTAL file Plotting of XSTAL is performed by MCNPxX using the execution option menpx z followed by the required instructions rmctal xstal nonorm The latter is essential since the data are normalized in XSEX3 Each case in XSEX3 is expanded in the XSTAL file for each particle type produced The tallies are identified by the numbering scheme 100 case number particle type the latter defined in the table below The last in the sequence corresponds to the elastic scattering distribution of the incident particle When plotting XSEX3 output the appropriate y axis labels are barns MeV steradian parns MeV or barns steradian If the yield multiplicity option was used in XSEX3 the appropriate y axis labels are particles MeV steradian etc The energy axis may be either energy MeV or momentum MeV c according to the XSEX3 option emp
61. Merger XTM RN U 97 012 April 22 1997 HUG98a H G Hughes et al Status of the MCNP LCS Merger Project American Nuclear Society Radiation Protection and Shielding Topical Conference April 19 23 1998 Nashville TN HUG98b H G Hughes et al Recent Developments in MCNPX American Nuclear Society Topical Meeting on Nuclear Applications of Accelerator Technology Gatlinburg TN Sept 20 23 1998 HUG98c H G Hughes et al MCNPX Code Development 4th Workshop on Simulat ing Accelerator Radiation Environments Knoxville TN September 14 1998 HUG99 H G Hughes et al MCNPX for Neutron Proton Transport International Conference on Mathematics amp Computation Reactor Physics amp Environmental Analysis in Nuclear Applications American Nuclear Society Madrid Spain September 27 30 1999 120 MCNPX User s Manual MCNPX User s Manual Ap Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium ICR84 International Commission on Radiation Units and Measurements ICRU Report 37 Stopping Powers for Electrons and Positrons October 1984 IGN75 A V Ignatyuk G N Smirenkin and A S Tishin Sov J Nucl Phys 21 1975 p 256 JAN82 J F Janni Proton Range Energy Tables 1keV 10GeV Atomic Data and Nuclear Data Tables 27 2 3 1982 KAL85 PRECO D2 Program for Calculating Preequilibrium and Direct Reaction Dou ble Differential Cross Sec
62. Neutron Proton gae inensis a a a E MeV MeV base Bertini nh 150 0 1 Bertini nh 20 0 2 Bertini nh dtsa 150 0 3 ISABEL nh 150 0 4 Bertini nh 150 150 5 CEM nh 150 0 For the sake of brevity we reproduce here just the neutron problem summaries from the MCNPX output decks Base Case sample problem spallation target neutron production with 20 MeV neutron transition energy EJ Pitcher 1 Nov 99 c c c c cell cards c c Pb target 11 11 4 1 2 3 bounding sphere 20 1 2 3 4 c outside universe 30 4 i surface cards oO 1 pz 0 0 2 pz 30 0 3 cz 5 0 126 MCNPX User s Manual Accelerator Production of Tritium 4 so 90 0 c material cards c c Material 1 Pb without Pb 204 m1 82206 24c 0 255 82207 24c 0 221 82208 24c 0 524 c c data cards mode nh imp n h 1 1r 0 phys n 1000 j 150 phys h 1000 j 0 Ica jjj nps 20000 prdmp j 30j 1 c c c source definition 1 GeV proton beam 7 cm diam parabolic spatial profile sdef sur 1 erg 1000 dir 1 vec 0 0 1 rad d1 pos 0 0 0 par 9 sil a 0 00 1 0 2 0 3 0 4 0 5 0 6 0 7 0 8 0 9 1 0 1 1 1 2 1 3 sp1 1 4 1 5 1 6 1 7 1 8 1 9 2 0 2 1 2 2 2 3 2 4 2 5 2 6 2 7 2 8 2 9 3 0 3 1 3 2 3 3 3 4 3 5 0 00000 0 09992 0 19935 0 29780 0 39478 0 48980 0 58237 0 67200 0 75820 0 84049 0 91837 0 99135 1 05894 1 12065 1 17600 1 22449 1 26563 1 29894 1 32392 1 34008 1 34694 1 34400 1 3
63. P PI MCNPX User s Manual 163 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 100 DXTRAN Contribution Card Variable Description which DXTRAN sphere the DXC card applies to If 0 or m absent the DXC card applies to all the DXTRAN spheres in the problem Default m 0 n particle designator probability of contribution to DXTRAN spheres from ry cell i Default P 1 I number of cells in the problem Use Optional Consider also using the DD card Section 5 8 11 5 8 15 BBREM Bremsstrahlung Biasing Form BBREMb b2b3 b49m m32 Mm Table 5 101 Bremsstrahlung Biasing Card Variable Description by any positive value currently unused ep bias factors for the bremsstrahlung energy spec eee trum My Mp list of materials for which the biasing is invoked Default None Use Optional 5 8 16 SPABI Secondary Particle Biasing FORM SPABI p xxx E1 S1 E2 S2 164 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 102 Secondary Particle Biasing Argument Descriptions Argument Description p Secondary Particle Type see Table 4 1 XXX List of primary particles to be considered For example nphe represents reactions of neutrons pho tons protons and electrons No spaces are allowed e Ifall particles are to be considered the entry should be all
64. Pi if included on the MODE card will be transported before it decays even though its life time is 8 4 x 10 seconds This allows the user to use MCNPX tallies for that particle MCNPX User s Manual 115 MCNPxX User s Manual Version 2 4 0 September 2002 LA CP 02 408 especially important for the user to include all possible secondary particles on the MODE card especially photons and neutrinos in order to get the most accurate energy deposition tally MCNPxX has the track length heating F6n p tally where p can now be any particle In addition MCNPX also has a collision heating F6n tally which contains energy deposition from all particles in the problem It is not currently possible to have an F6 tally which will do energy deposition for more than one but less than all particles We will consider adding this capability in the future Note that the pedep keyword in a Type 1 Mesh Tally is analogous to the F6n p tally and the Type 3 Mesh Tally is analogous to the F6n tally although the normalizations will be different Since the mesh tallies score energy deposition within a mesh cell which may contain more than one material normalization is per unit volume The units of this tally are MeV source particle In the F6 and F6 tallies material density is available for the chosen cells and normalization is MeV gm source particle Example 1 F2 N 136T This card specifies four neutron flux tallies one across each of th
65. Prael J F Dicello and M Zaider Improved Calculations of Energy Deposition from Fast Neutrons in Proceedings Fourth Symposium on Neutron Dosimetry EUR 7448 Munich Neuherberg 1981 BRE89 D J Brenner and R E Prael Calculated Differential Secondary particle Pro duction Cross Sections after Nonelastic Neutron Interactions with Carbon and Oxygen between 10 and 60 MeV Atomic and Nuclear Data Tables 41 71 130 1989 BRI97 J F Briesmeister ed MCNP A General Monte Carlo N Particle Transport Code Los Alamos National Laboratory Report LA 12625 M Version 4B March 1997 http www xdiv lanl gov XCI PROJECTS MCNP manual html CHA98 M B Chadwick et al Reference Input Parameter Library handbook for Cal culations of Nuclear reaction Data IAEA TECDOC Draft IAEA Vienna March 1998 CHA99a_ M B Chadwick P G Young S Chiba S C Frankle Hale H G Hughes A J Koning R C Little R E MacFarlane R E Prael and L S Waters Cross Section Evaluations to 150 MeV for Accelerator Driven Systems and Implementation in MCNPX Nuclear Science and Engineering 131 Number 3 March 1999 293 CHA99b_ M B Chadwick P G Young R E MacFarlane P Moller G M Hale R C Little A J Koning and S Chiba LA 150 Documentation of Cross Sections Heating and Damage Part A Incident Neutrons and Part Incident Protons Los Alamos National Laboratory Report LA UR 99 1222 1999 http t2
66. Production of Tritium LA UR 02 2607 MCNPX USER S MANUAL Version 2 3 0 Laurie S Waters Editor Accelerator Production of Tritium MCNPX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 MCNPX User s Manual Zz Apr MCNPX User s Manual Accelerator Version 2 3 0 April 2002 Production LA UR 02 2607 of Tritium Acknowledgments The MCNPX code and data effort represents the efforts of many people much of whose work is represented in this manual The primary team members are listed below Code Development Team H Grady Hughes team leader Harry W Egdorf Franz C Gallmeier John S Hendricks Robert C Little Gregg W McKinney Richard E Prael Teresa L Roberts Edward Snow Laurie S Waters Morgan C White Library Development Team Mark B Chadwick Stephanie C Frankle Gerald M Hale Robert C Little Robert MacFarlane Morgan C White Phillip G Young Physics Development Team David G Madland Stepan G Mashnik Richard E Prael Arnold J Sierk APT AAA Target Blanket Design and ED amp D Team LANSCE Team Michael W Cappiello Rhonda K Corzine Phillip D Ferguson Michael M Fikani Frank D Gac Michael R James Russell Kidman Stuart A Maloy Michael A Paciotti Eric J Pitcher Lawrence G Quintana Gary J Russell Beta Test Team 800 users from 175 institutions worldwide MCNPX was originally conceived as an upgrade to the existing Los Alamos LAHET Code System LCS and
67. SYSTEM 3 1 1 In the Beginning Remember that your PATH environment variable governs the search order for finding utilities You should be aware of the value of your PATH environment variable by issuing the following command echo PATH You may find it useful to set your PATH environment variable to a strategic search order so that the utilities that are found first are the ones you intend to use Setting of environment variables is done differently depending upon what shell you use Please consult the appropriate manuals for your shell Most systems have more than one shell Any system can have more than one version of any utility You must know your utilities If you work on a UNIX or Linux operating system you can use the following inquiry commands to learn if you have more than one make utility which make MCNPX User s Manual 9 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 which gmake Many systems come with a make utility that is provided by the vendor On UNIX and Linux you must use the GNU make utility and it must be version 3 76 or later Sometimes the GNU make utility is installed in an executable file called gmake Sometimes system administrators make symbolic links called make that when resolved invoke the gmake utility You can make your own symbolic links in directories that you own and control so that when you execute the make command you will be executing the make you intend to use You
68. Since it must be read and stored by the MCNP subroutines it must not appear within the mesh data block between the tmesh and endmd cards The structure of the mesh as well as what quantities that are to be written to it are defined on two control cards in the MCNPX INP file The general forms of the two mesh cards are RMESHn P keyword i i 1 10 CMESHn P keyword i i 1 10 SMESHn P keyword i i 1 10 RMESH is a rectangular mesh CMESH is a cylindrical mesh and SMESH is a spherical mesh The n is a user defined mesh number The last digit of n defines the type of infor mation to be stored in the mesh P is the particle type being tallied which be absent depending on the type of mesh tally Up to 10 keywords are permitted depending on mesh type In MCNPX version 2 3 0 there are four general types of mesh tally cards each with a different set of keywords 1 The user should be warned that the mesh tally number must be different from any other tally in the prob lem For example an fl n tally will conflict with a RMESH1 n tally 94 MCNPX User s Manual MCNPX User s Manual i Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium Track Averaged Mesh Tally type 1 The first mesh type scores track averaged data flux fluence or current The values can be weighted by an MSHMF card through the DFACT dose conversion coefficient function or for energy deposition R C S MESHn P traks flux dose popul pede
69. TMn 1 128 optional CMn 1 128 MCNPX User s Manual 177 178 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 113 Summary of MCNPX Input Cards optional CFn none 129 optional SFn none 129 optional FSn none 130 optional SDn 0 131 optional FUn Requires SUBROUTINE TALLYX 131 optional TFn 1 1 last last 1 last last last 135 optional TIRn 136 optional PERT none 140 optional TMESH 143 optional FTn none 139 Variance reduction cards section 5 8 on page 153 required IMP required unless weight windows used 153 optional WWG none 154 optional WWGE single energy or time interval 155 optional WWP 535000 155 required WWN required unless importances used 156 optional WWE none 157 optional MESH none 158 optional EXT 0 159 optional VECT none 160 optional FCL 0 160 optional DDn 0 1 1000 161 optional PDn 1 162 optional DXT 000 163 optional DXC 1 163 optional BBREM none electron photon transport only 164 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 113 Summary of MCNPX Input Cards optional SPABI 164 optional ESPLT no energy splitting or roulette 165 optional PWT 1 MODE NPorNPE only 166 Output Control Cards section 5 9 on page 166 optional PRDMP end 15 0 all 10 rendezvou
70. The content of the SP and SB cards then follows the general MCNP rules 86 MCNPX User s Manual MCNPX User s Manual Ap Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium The following example shows a case of three intersection Gaussian parallel beams each defined with the parameters a 0 2cm b 0 1cm and c 2 in the notation previously dis cussed For each the beam is normal to the plane of definition e Beam 1 is centered at 0 0 2 with the major axis of the beam distribution along the x axis emitted in the z direction with relative intensity 1 e Beam 2 is centered at 2 0 0 with the major axis of the beam distribution along the y axis emitted in the x direction with relative intensity 2 e Beam 3 is centered at 0 2 0 with the major axis of the beam distribution along the line x z emitted in the y direction with relative intensity 3 The card SBn is used to provide equal sampling from the three beams which is indepen dent of the relative intensities This example demonstrates most of the new features The input cards are as follows Title c Cell cards 999 0 999 cookie cutter cell c Surface Cards 999 SQ 25 100 0000 4 000 cookie cutter surface c Control Cards SDEF DIR 1 VEC 001 X D1 Y D2 2Z 0 CCC 999 TR D3 SP1 41 470964 0 SP2 41 2358482 0 SI3 L123 SP3 123 SB3 111 TR1 00 2 100 010 001 TR2 200 010 001 100 TR3 0 20 707 0 707 707 0 707 010 MCNPX User
71. User s Manual Version 2 4 0 September 2002 LA CP 02 408 19 The Resource Option The RESOURCE option allows the user to edit the data available on a history file while altering the assumed spatial distribution of the source from that used in the original calculation For its application see reference 1 20 The Merge Option Not used in HTAPESX For any tally either the HISTP file or the HISTX file is edited but not both 21 The Time Convolution Option Assume that an initial calculation has been made with the default source time distribution i e all histories start at t O A time dependent tally for any of the allowed LAHET source time distributions may then be made with HTAPE3X without rerunning the transport calculation For details see reference 1 22 The Response Function Option Any non zero value of the IRSP parameter allows the user to apply an energy dependent response function F E where E is the particle energy to the current and flux tallies given by edit option types 1 2 4 9 10 and 13 The user supplies a tabulation of the function F E by the pairs of values FRESP I ERESP I which are input as the arrays ERESP I l 1 NRESP and FRESP I l 1 NRESP described in Section 2 above The element IRESP I of the third input array then specifies an interpolation scheme for computing the response function value within the interval ERESP l lt E lt ERESP I 1 For IRSP gt 0 the interpolated response
72. User s Manual Version 2 4 0 September 2002 LA CP 02 408 Title c Cell cards ccc 0 nnn cookie cutter cell c Surface Cards nnn SQa b0 0 0 0 c 0 0 0 cookie cutter surface c Control Cards SDEF DIR 1VEC 0 0 1X D1Y D2Z 0CCC cccTR n SP1 41 f0 SP2 41f 0 TRn Xo Yo Zo COS Sing Osing cos 0 0 0 1 The SDEF card sets up an initial beam of particles travelling along the Z axis DIR 1 VEC 0 0 1 Information on the x and y coordinates of particle position is detailed in the two SP cards X D1 Y D2 indicating that the code must look for distributions 1 and 2 here given by source probability distributions SP1 and SP2 The z coordinate is left unchanged Z 0 There is no PAR option in this example therefore the particle generated by this source will be the one with the lowest IPT number in Table 4 1 neutron The SP cards have three entries The first entry is 41 which indicates sampling from a built in gaussian distribution note the function 41 is a gaussian in time in MCNP It has been modified for the purpose of MCNPX It has the following density function 1 x 2 yn Ce pox y exp 3 8 E 2xab t exe2 The parameters a and b are the standard deviations of the Gaussian in x and y The second entry f or fy on the SP cards is the full width half maximum FWHM of the Gaussian in either the x or y direction and must be computed from a and b by the user as follows 108 MCNPX User s Manual MCNPX
73. User s Manual Version 2 4 0 September 2002 LA CP 02 408 f 8In2 a 2 35482a 1 fy 8In2 b 2 35482b The third entry represents the centroid of the Gaussian in either the x or y direction We recommend that the user input 0 here and handle any transformations of the source with a TR card as described below Using a non zero value will interfere with the rejection function as specified by the cookie cutter option Note that in Print Table 10 in the MCNPX output file the definitions of a b and c are different from those discussed above however fwhm will be the same as the 3rd entry on the SP cards The parameter a in Table 10 differs from the parameter a above by a factor of the square root of 2 This is a legacy item from the conversion of the 41 function from time to space and will be corrected in a future version The user generally does not want the beam Gaussian to extend infinitely in x and y therefore a cookie cutter option has been included to keep the distribution to a reasonable size CCC ccc tells MCNPX to look at the card labeled ccc ccc is a user specified cell number to define the cutoff volume The first entry on the ccc card is 0 which indicates a void cell The second number nnn nnn again is a user specified number indicates a surface card within which to accept particles In the example this is a SQ surface a 2 sheet hyperboloid is defined as follows 1 2 2 xX y 2
74. VVV WWW SUR Surface Zero means cell source ERG Energy MeV 14 MeV TME Time shakes 0 MCNPX User s Manual 97 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 45 General Source Variables Variable Description Default u the cosine of the angle between VEC Volume case u is sampled uniformly in 1 DIR and UUU VVV WWW Azimuthal angle to 1 isotropic Surface case p w 2u is always sampled uniformly in 0 to in 0 to 1 cosine distribution 360 Reference vector for DIR vector nota Volume case required unless isotropic VEC tion Surface case vector normal to the surface with sign determined by NRM NRM Sign of the surface normal 1 Reference point for position sampling 0 0 0 POS vector notation Radial distance of the position from POS 0 RAD or AXS EXT Cell case distance from POS along AXS 0 Surface case Cosine of angle from AXS Reference vector for EXT and RAD vec No direction AXS tor notation x coordinate of position 0 y coordinate of position 0 z coordinate of position 0 Area of surface required only for direct None ARA contributions to point detectors from plane surface source WGT Particle weight explicit value only 1 EFF Rejection efficiency criterion for position 01 sampling explicit value only Source particle type i e h or 9 neutron if no MODE card PAR SF invokes spontaneous fission
75. Z1 RO X2 Y2 Z2 F1 F2 F3 In MCNPX version 2 3 0 the form of the card has been changed old input files are back ward compatible if one replaces the control card symbol Pin P X1 Y1 Z1 RO X2 Y2 Z2 F1 F2 F3 n is the tally number and must be a multiple of 5 since this is a detector type tally P is the particle type for the tally Only neutrons or photons are allowed since detector techniques do not currently work for charged particles 102 MCNPX User s Manual MCNPX User s Manual i Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium Table 8 5 Pinhole Radiography Argument Descriptions Argument Description X1 Y1 Z1 The coordinates of the pinhole RO Always 0 zero for this application Note neither the pinhole nor the grid should be located within a highly scatter ing media X2 Y2 Z2 The reference coordinates that establish the reference direction cosines for the normal to the detector grid This direction is defined as being from X2 Y2 Z2 to the pinhole at X1 Y1 Z1 F1 If F1 gt 0 the radius of a cylindrical collimator centered on and parallel to the reference direction which establishes a radial field of view through the object F2 The radius of the pinhole perpendicular to the reference direction F2 0 represents a perfect pinhole F250 the point through which the particle contribution will pass is picked randomly This simulates a less than perfect pi
76. a build directory Call it mcnpx mkdir mcnpx go into that new empty working space cd mcnpx execute the configure script the prefix tells where to put the executables and libraries mcnpx_2 4 0 configure prefix home me now make the executable mcnpx program and the bertin and pht libraries run the tests and install in home me bin and home mel lib make all tests install 3 1 3 5 Individual Private Installation special compilers and debugging As a final example suppose you want basically the same thing as the previous example but you would like to have the debug option turned on during compilation The compiled 16 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 code will go into a private local library nome me bin but you wish to use the cross section files and LCS data files already on your system We will assume that these data files already exist in the directory usr mcnpx data We will assume that the source distribution has already been unpacked by a system administrator into usr local src mcnpx_2 4 0 If your system has only f90 it will be found and used We decide to specify the Sun f90 and cc compilers for this build go to your user home directory cd set an environment variable that identifies where the distribution lives This isn t really necessary but cuts down on typing later MCNPX_DIST usr local src mcnpx_2 4 0 export MCNPX_DIS
77. a coupled photon electron mode to get better results In fact in working with these type of coupled problems it was found that the most consis tent results as compared to a F8 p e tally could be achieved if the energy deposited by the electrons only was scored This seems to work very well since in photon energy dep osition most if not all of the energy lost by the photon goes into creating secondary electrons that then account for the energy deposited in the cell Electrons The electron energy deposition is evaluated as the de dx ionization uniformly distributed along track length dx Then several adjustments are made the first of which is for x ray production if photons are to be produced by including a p on the mode card The de dx term is decreased by the amount of energy that goes into the secondary x rays produced if they are being transported otherwise this adjustment is not made An adjustment is MCNPX User s Manual 109 MCNPX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium always made for the knock on electrons or delta rays produced since these will be banked and subsequently transported and their energy deposited during that transport process There are also adjustments made for any auger electrons produced In addition if the bremsstrahlung photons are not to be transported the electron energy that would be lost in their production is also distributed uniformly along the
78. added Section 3 1 6 Chapter 4 Physics and Data e Proton and photonuclear capability is added in the tabular region Photonuclear capability in the physics region will be included in an upcoming version See sections 4 3 1 1 and 4 3 1 2 e 150 MeV Neutron data libraries have been updated to include Mercury and Bismuth A 100 MeV library on Be has also been added e Charged Particle Production Threshold table added Table 4 4 e Nontracking change Higher Energy Table discussion has been updated to include barpol dat and OLDXS information Section 4 3 1 3 Use of the new cross sections is now the default This will result in a higher neutron production rate on some targets e Section 4 3 1 4 on Atomics Mass Tables added e Section 4 3 1 5 on Nuclear Structure Data Library PHTLIB added including discus sion of alternative SPEC1 file e Section 4 3 2 2 revised to correct mistypes Chapter 5 Multiparticle Extensions and General Tracking e Non tracked particles information has been included in Table 5 1 and Appendix B has been deleted e Mass of the neutron corrected in Table 5 1 e Corrected the symbol for charged pions in Table 5 1 from to Section 5 3 on Energy Straggling for Heavy Charged Particles has been revised to include discussion of Vavilov tracking improvements Chapter 6 MCNPX User s Manual 9 MCNPX User s Manual Ap Version 2 3 0 April 2002 LA UR 02 2607 Accelerator
79. added to the inp deck HISTP no arguments 5 9 6 DBCN Debug Information Form DBCN X Xo X3 X MCNPX User s Manual 171 172 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 110 Debug Information Card Variable Description yx the starting pseudorandom number l Default 519 152917 use Xg instead Xo debug print interval X3 and X4 history number limits for event log printing X maximum number of events in the event log to print per gt history Default 600 Xe unused x 1 produces a detailed print from the volume and surface 4 area calculations number of the history whose starting pseudoran Xg dom number is to be used to start the first history of this problem X closeness of coincident repeated structures surfaces 3 Default 1 E 4 X seconds between time interrupts Default 100 sec 10 onds 1 causes collision lines to print in lost particle event X11 log X42 expected number of random numbers X73 random number stride Default 152917 X14 random number multiplier Default 519 x 1 prints the shifted confidence interval and the variance of 13 the variance for all tally bins x scale the score grid for the accumulation of the empirical 18 f in print tables 161 and 162 0 default angular treatment for partial substeps to genera tion sites of secondary particles X47 gt 0 alternate angular treatm
80. also funded through this pro gram and preliminary capabilities were first included in MCNPX version 2 1 5 The MCNPX program began in 1994 when several groups in the Los Alamos X T and LANSCE divisions proposed a program of simulation and data tool development in support of the Accelerator Production of Tritium Project The work involved a formal extension of MCNFP to all particles and all energies improvement of physics simulation models exten sion of neutron proton and photonuclear libraries to 150 MeV and the formulation of new variance reduction and data analysis techniques The proposal also included a program of cross section measurements benchmark experiments deterministic code development and improvements in transmutation code and library tools through the CINDER 90 project Since the closure of the APT project work on the code has continued under the sponsor ship of the AAA and other programs Since the initial release of MCNPX version 2 1 on October 23 1997 an extensive beta test team has been formed to test the code versions prior to official release The initial release of MCNPX version 2 1 5 to the beta test team occurred on May 21 1999 Final corrections and supplements to the code were released to RSICC in November 1999 along with the current revision 1 of the User s Manual Approximately 800 users in 175 institutions worldwide have had an opportunity to test the improvements in the code lead ing to version 2 3
81. amp Log Short cuts MCNP xX allows five shortcuts to facilitate data input in some instances 1 nR means repeat the immediately preceding entry on the card n times For example 2 4R is the same as 2 2 2 2 2 nI means insert n linear interpolates between the entries immediately preceding and following this feature For example 1 5 2i 3 0 on a card is the same as 1 5 2 0 2 5 3 In the construct X ni Y if X and Y are integers and if Y X is an exact multiple of n 1 correct integer interpolates will be created Otherwise only real interpolates will be created but Y will be stored directly in all cases In the above example the 2 0 may not be exact but in the example 1 4i 6 1 2 3 4 5 6 all interpolates are exact xM means multiply the previous entry on the card M by the value x For example 112M 2M 2M 2M 4M 2M 2M is equivalent to 112 48 16 64 128 256 nJ means jump over the entry where used and take the default value As an example the following two cards are identical in their effect DD 1 1000 DD J 1000 J J J is also equivalent to 3J You can jump to a particular entry on a card without having to explicitly specify prior items on the card This feature is convenient if you know you want to use a default value but can t remember it DBCN 7J 5082 is another example nLOG means insert n logarithmic interpolates between the entries immediately preceding and following this feature For example 001 4Log 100 is equiva
82. and Match problem involves reworking of various data structures in the code This will not be completely implemented until the end of year 2002 4 3 1 1 The LA150 Proton and Neutron Libraries Table 4 3 summarizes the 150 MeV neutron proton and photonuclear libraries availble to date Table 4 3 Summary of LA150 Libraries Element Neutrons Protons Photonuclear Hydrogen 1H 2H 1H 2H Beryllium Be 100 MeV Carbon nato 12C 12C Nitrogen 14N 14N Oxygen 160 160 160 MCNPX User s Manual 47 MCNPX User s Manual E Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium Table 4 3 Summary of LA150 Libraries Continued Element Neutrons Protons Photonuclear Aluminum 27 Al 27 Al 27A Silicon 28g 29g 30g 28gj 29g 30g 28g Phosphorous 31p 31p Calcium natog 40ca 40caq Chromium 50 y 52cr 53 y 54Gr 50 y 52cr 53 y 54Gr lron 54Ee 56Fe 57Fe 54Ee 56Fe 57Fe 56Fe Nickel 58y 60N 61N 62Ni 64Ni 58y 604 61N 62N 64N Copper 63cu amp 5Cu 63cu amp 5Cu 63cu Niobium 93Nbp 23Nb Tantalum 18175 Tungsten 182 183 yy 184 yy 186 yy 182yy 183 yy 184 yy 186 yy 184W Mercury 196g 19819 199g 200g 196g 19819 199g 200g 201 Hg 202g 204Hg 201 Hg 202g 204Hg Lead 206 pp 207 Pp 208pp 206 pp 207 pb 208ppH 206Pp 207 Pp 208Pp Bismuth 2095 2095 a A much larger set of photonuclear dat
83. and changes A computer test farm of 20 different software hardware configurations is maintained to ensure that code development does not adversely affect any previously tested system We are also constantly moving toward a modular system whereby the user may easily implement alternative physics packages EGD01 Some restructuring of the code has already been done toward that goal including the development of an autconfiguration system In addition to describing the new interaction physics this manual contains a summary of information from recent MCNPX release notes memos publications and presentations It represents the work of the code development team the nuclear data team the physics development team and several outside collaborators The manual is updated and extended with each new code release 2 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 The reader must be aware of certain limitations in code usage These items are listed in Chapter 2 Chapter 3 covers code installation and general notes on software management Chapter 4 covers MCNPX Input cards Information supplemental to the text is included in the Appendices This manual is not intended to replace the existing user guides to MCNP4C BRIOO the LAHET Code System PRA89Q nor any other manual covering incorporated physics modules The user should become familiar with these works which are extensively referenced Workshops
84. are not properly specified In Example 1 above if all tallies that are positive with respect to surface 3 are also all positive with respect to surface 4 the third segment bin will have no scores Example 2 F2 N 1 FS2 3 4 The order and sense of the surfaces on the FS2 card are important This example produces the same numbers as does Example 1 but changes the order of the printed flux Bins two and three are interchanged Example 3 F1 N 12T FS1 3 T This example produces three current tallies 1 across surface 1 2 across surface 2 and 3 the sum across surfaces 1 and 2 Each tally will be subdivided into three parts 1 that with a negative sense with respect to surface 3 2 that with a positive sense with respect to surface 3 and 3 a total independent of surface 3 130 MCNPX User s Manual 5 7 15 SDn Segment Divisor tally types 1 2 4 6 7 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Form SDn D11 85 Dim D21 Doo ae Dom Dki Dko Dkm Table 5 73 Segment Divisor Card Variable Description n tally number n cannot be zero k number of cells or surfaces on Fn card including T if present number of segmenting bins on the FSn card including m the remainder segment and the total segment if FSn has aT D area volume or mass of j segment of the i surface or 1 cell bin for tally n The parentheses are optional Use Not with detectors May be required
85. before making their contribution to the cell 6 10 or 13 tally 5 7 13 SFn Surface Flagging tally types 1 2 4 6 7 Form SFn S4 Sk Table 5 71 Surface Flagging Card Variable Description n tally number S problem surface numbers whose tally contributions are to be flagged Default None Use Not with detectors Consider FQn card MCNPX User s Manual 129 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 5 7 14 FSn Tally Segment tally types 1 2 4 6 7 Form FSn Sj Sk Table 5 72 Tally Segment Card Variable Description tally number Sj signed problem number of a segmenting surface Default No segmenting Use Not with detectors May require SDn card Consider FQn card Example 1 F2 N 1 FS2 3 4 This example subdivides surface 1 into three sections and calculates the neutron flux across each of them There are three prints for the F2 tally 1 the flux across that part of surface 1 that has negative sense with respect to surface 3 2 the flux across that part of surface 1 that has negative sense with respect to surface 4 but that has not already been scored and so must have positive sense with respect to surface 3 3 everything else that is the flux across surface 1 with positive sense with respect to both surfaces 3 and 4 It is possible to get a zero score in some tally segments if the segmenting surfaces and their senses
86. both installed In the previous example the GNU g77 compiler would have been used because if it exists g77 will be found first when searching for Fortran com pilers on your system If your system has only f77 it will be found and used We decide to specify the Sun f77 and cc compilers over the GNU g77 and gcc compilers for this build The with LD flag may be needed in such a case because a full installation of the GNU compiler tools can also include a GNU version of the Id link editor Unfortunately the dif ferent Id commands take command line arguments whose syntax differs between the two systems As far as is known this ONLY affects certain experimental uses of MCNPX and should not be needed by normal users It is shown in this example as a sample of how it is used in the few cases where it is needed go to your user home directory cd set an environment variable that identifies where the distribution lives This isn t really necessary but cuts down on typing later MCNPX_DIST usr local src mcnpx_2 3 0 20 MCNPX User s Manual MCNPX User s Manual i Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium export MCNPX_DIST make a working space that reminds you it s a debug version mkdir mcnpx debug cd mcnpx debug execute the configure script request debug for the Makefiles also specify where to put the installed code and which compilers to use MCNPX_DIST configure with FC f77
87. can also establish an alias in the shell runtime control file whereby any make command you issue actually executes gmake You can also substitute the gmake command everywhere you see the make command in the examples that follow The important point of this discussion is to know your make and use the right one otherwise this automated build system can fail If no make or gmake is found you either have a PATH value problem or you need some help from your system administrator to install GNU make If both make and gmake exist query each of them to see what version you have make v gmake v Some vendor supplied make utilities do not understand the v option that requests that the version number be printed If you see an error or usage message then your make is one of the vendor supplied variety Make sure you have GNU make version 3 76 or later installed and that it is found in your search path first If you work on a Windows platform this distribution is not the correct one for your needs Please request a separate Windows distribution Until an automated build system for Windows is created binary images will be distributed 3 1 2 Automated Building The process used when building mcnpx varies greatly depending upon the following e hardware platform e g SPARC ALPHA 1386 operating system e g Solaris Linux HP UX e available compilers e g f90 cc g90 gcc pgf90 gcc mcnpx program options e g the default
88. capability of MCNP involving the i option is retained allowing a large number of regularly spaced mesh points to be defined with a minimum of entries on the coordinate lines All of the coordinate entries must be monotonically increasing for the tally mesh fea tures to work properly but do not need to be equally spaced It should be noted that the size of these meshes scales with the product of the number of entries for the three coor dinates Machine memory could become a problem for very large meshes with fine spacing 92 MCNPX User s Manual MCNPX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium 10 0 100 200 300 400 500 distance from center of first ladder c Figure 8 1 Mesh Tally depiction of a sample spallation target neutron fluence Additional cards which can be used with Mesh Tallies are ERGSHn E1 E2 MCNPX User s Manual 93 MCNPxX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium MSHMFn E1 F1 E2 F2 FMn Where E1 is the lower energy limit for information to be stored to the mesh n and E2 is the upper energy limit as they appear on the ERGSH card The default is to consider all energies The entries on the MSHMF card are pairs of energies and the corresponding response functions as many pairs can be designated as needed The FM card is the same as described in the MCNP4B users manual
89. card All cards before the blank line delimiter are continuation cards The syntax and components of the message are the same as for the regular execution line message Any filename substitution program module execution option or keyword entry on the execution line takes precedence over conflicting information in the message block INP filename is not a legitimate entry in the message block The name INP can be changed on the execution line only 4 1 4 Problem Title Card The first card in the file after the optional message block is the required problem title card It is limited to one 80 column line and is used as a title in various places in the MCNPX output It can contain any information the user desires or can even be blank and often contains information describing the particular problem Note that a blank card elsewhere is used as a delimiter or as a terminator 4 1 5 Card Format All input lines are limited to 80 columns Alphabetic characters can be upper lower or mixed case Most input is entered in horizontal form however a vertical input format is allowed for data cards A comment can be added to any input card A dollar sign terminates data entry and anything that follows the is interpreted as a comment Blank lines are used as delimiters and terminators Data entries are separated by one or more blanks 4 1 6 Comment Cards Comment cards can be used anywhere in the INP file after the problem title card and before the
90. cell labels colors will be temperature 1 wwn cell labels colors will be weight windows 1 by particle type axi cell labels colors will be exponential transform by particle type pd cell labels colors will be dxc cell labels colors will be dxtran contributions u cell labels colors will be universe numbers lat cell labels colors will be latices fill cell labels colors will be filling universes nonu cell labels colors will be fission turnoffs pac cell labels colors will be particle activity column PAR controls particle type displayed controls number on the cell quantity N Example wwn3 p would provide photon weight windows in the 3rd energy group and be clicked using wwn P amp N BOTTOM MARGIN COMMANDS MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 1 Interactive Geometry Plotter Commands Command Result Toggled on by Click here click Enter Data Allows entry of parameters per prior plotting methods e g Origin 0 0 0 will locate plot origin at x y z 0 0 0 Redraw redraws the picture when it needs refreshing returns control to the command window enabling traditional Plot gt plot commands to be entered End terminates the plot session Plotting Superimposed Weight Window Mesh MESH off can be toggled to MESH on position by clicking when a mesh has been generated by WWINP card entry wwn par N yields weight window par
91. correction in energy deposition which is not a strict linear function In MCNPX the procedure is to search through all cells and find the first one with the material in question and use that density for the correction factor for all cells using that material The effect is small so this is an adequate procedure however MCNPX does give a warning message when you encounter such situations In MCNPxX with more charged particles and greatly expanded energy range this for merly small correction now becomes increasingly important and the usual way of handling it is not sufficient MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 MCNPX User s Manual 8 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 3 Installation This chapter describes how to build MCNPX on a system The system will need a FORTRAN 90 compiler a C compiler and GNU Make 3 76 or higher MCNPX installs and runs on Windows amp Linux PC s and a variety of common Unix workstations Some of our supported systems include e IBM RS 6000 AIX e DEC Alpha Digital Unix e SGI IRIX 32 and 64 bit e HP HP UX version 10 e Sun Solaris e Intel 1386 Linux e Microsoft Windows PC The code distribution contains full source code for the MCNPX 2 4 0 system and test sets for each of the supported architectures The CDROM also contains a recent source distribution of the GNU make utility needed to properly build the system 3 1 UNIX BUILD
92. do not create delayed photons Photonuclear interactions from 1 0 to 150 0 MeV in tabular range for 12 isotopes No physics models outside the tabular range are available in MCNPX 2 4 0 For any incident particle where libraries exist neutrons protons and photonuclear MCNPX 2 4 0 users should not specify isotopes with different transition energies between tabular data and physics models The transition energies should be the same for each incident particle and should not exceed the maximum energy of the selected data library 7 TYPICAL RUNNING TIME Runtime for the test cases was 17 minutes for the test cases on a Dell PowerEdge6400 running Linux 37 minutes on an IBM RS 6000 Model 270 and 43 minutes on a HP B1000 PA 8500 8 COMPUTER HARDWARE REQUIREMENTS MCNPX runs under Unix Linux and Windows operating systems and has been implemented on IBM RS 6000 AIX DEC Alpha Digital Unix SGI IRIX 32 and 64 bit HP HP UX version 10 Sun Solaris Intel Linux and Windows based PC s The compiled version of the code tends to run 8 Mbytes Dynamic allocation makes memory demands variable on all platforms 9 COMPUTER SOFTWARE REQUIREMENTS C and Fortran 90 compilers are required to compile The GNU make utility is required to build the system on Unix and Linux platforms The GNU make exe utility is included for Windows users The only graphics support for this release is X11 http www x org Downloads_terms htm This is a Fortran 90 version of M
93. electron particle track Of course if these photons are to be transported no corrections to the electron energy deposition is made Heavy Neutral and Charged Particles In the energy range where tables are available the neutron and proton energy deposition is determined using the neutron heating numbers in the same manner as F6 tallies are done in MCNP4B These heating numbers are estimates of the energy deposited per unit track length In addition the de dx ionization contribution for the proton is added in similar to the electron treatment Above that tabular energy limit or when no tabular data is available energy deposition is determined by summing several factors For charged particles ionization de dx energy is deposited uniformly along the track length which is important to keep in mind when doing a mesh tally All other energy deposition is calculated at the time of a nuclear inter action The energies of secondary particles if they are not to be tracked i e not included on the MODE card will be deposited at the point of the interaction Nuclear recoil energy will always be deposited at the point of interaction In order to obtain the most accurate energy deposition tallies possible the user must include all potential secondary particles on the MODE card Electrons can be omitted provided the user fully understands how energy deposition for photons is done The han dling of energy deposition for non tracked secondar
94. electrons they are therefore elemental in nature Additionally the evaluators who work on photonuclear data are generally separate from those who work on photoatomic data For these and other rea sons it was decided to store photonuclear data for MCNPX on tables separate and distinct MCNPX User s Manual 51 MCNPxX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium from the tables providing photoatomic data A new u type table has been constructed for MCNPxX to hold the photnuclear data This table follows the logic that was established for handling multiple particle emission in neutron and proton data and codifies it in a manner consistent for any incoming particle with multiple ejectiles It uses established standard conventions for laying out the data blocks such that existing sampling algorithms can be applied If photonuclear physics has been enabled in either biased or analog modes the user must supply material descriptions that include phtonuclear tables The standard materials M card has been extended to allow specification of photonuclear library IDs in the expected manner see section 6 1 6 Photonuclear transport using physics modules above the tabular range is available and is being tested for a near term MCNPxX release 4 3 1 3 Higher Energy Tables MCNPX Version 2 1 5 MCNPxX includes an elastic scattering model for neutrons above 15 MeV and protons above 50 MeV sepa
95. equal sampling from the three beams which is independent of the relative intensities This example demonstrates most of the new features The input cards are as follows Title c Cell cards 999 0 999 cookie cutter cell c Surface Cards 999 Q251000 0 0 0 40 0 0 cookie cutter surface c Control Cards SDEF DIR 1VEC 0 0 1X D1Y D2Z 0CCC 999T R D3 SP1 41 4709640 SP2 41 23584820 SI3 L123 SP3 123 SB3 111 TR1 00 2100010001 TR2 200010001100 TR3 0 2 0 707 0 707 707 0 707010 5 7 TALLY SPECIFICATION Fna FCn En Tn Cn FQn FMn DEn DFn EMn TMn CMn CFn SFn FSn SDn FUn FTn TALLYX TFn TIRn PERTn TMESH MCNPX User s Manual 111 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 The tally cards are used to specify what type of information the user wants to gain from the Monte Carlo calculation that is current across a surface flux at a point heating ina region etc This information is requested by the user by using a combination of the following cards To obtain tally results only the Fn card is required the other tally cards provide various optional features The n is a user chosen tally number lt 999 choices of n are given in the following section When a choice of n is made for a particular tally type any other input card used with that tally such as En for energy bins is given the same value of n by the user Much of the information on these cards is used to describe
96. every source or scatter event a ray trace contribution is made to every bin in the detector grid This eliminates statistical fluctuations across the grid that would occur if the grid location of the contribution from each event were to be picked randomly as would be the case if one used a DXTRAN sphere and a segmented surface tally For each event source or scatter the direction to each of the grid points is determined and an attenuated ray trace contribution is made As in pinhole image projection there are no restrictions as to location or type of source used These tallies automatically bin in a source only and a total contribution but could be further binned as described for the pinhole tally The transmitted image projection is set up as follows TI R C n P X1 Y1 Z1 RO X2 Y2 Z2 F1 F2 F3 TIR is used to establish a grid on a plane surface TIC is used to establish a grid on a cylindrical surface n is the tally number and must be a multiple of 5 since this is a detector type tally P is the particle type for the tally Only neutrons or photons are allowed In MCNPX 2 x this card was called Fln P old input files are backward compatible Table 5 78 Transmitted Image Projection Argument Description Argument Description The coordinates used with the entries on the FSn and Cn cards to define the detector grid In the plane grid case this defines the center of the grid In the mel cylindrical grid case this defines
97. for a a Z N E CEM97 models for a a Z N E Multifragmentation of light nuclei Fermi breakup as in LAHET Fermi breakup as in LAHET Fermi breakup as in LAHET Fission models ORNL or RAL models ORNL or RAL models CEM model for of RAL fission fragmentation 40 MCNPX User s Manual Accelerator Production of Tritium MCNPX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 Table 4 2 Intermediate Energy Model Recommended Ranges Variable Bertini Isabel CEM Lower energy 20 150 MeV 20 1 50 MeV 100 MeV Upper Energy 3 5 GeV nucleon 1 GeV 5GeV nucleon 2 5 GeV pion nucleon Nuclei all all carbon and heavier Incident particles p n pions A lt 4 and antiprotons p n pions a All models will run outside their recommended energy limits however no detailed nuclear struc ture is contained at lower energies At higher energies the Bertini and CEM models will start to underpredict certain quantities although 10 GeV is a reasonable upper limit 4 1 1 Intranuclear Cascade Models The concept of an Intranuclear Cascade INC model is quite old and intuitively simple A particle incident on a nucleus will interact with individual nucleons with final states defined by a set of fundamental particle particle cross sections The nucleons are considered to be acold free gas confined within a potential that describes the nuclear den
98. forms a cursor to zoom into a part of the picture SCALES adds scales showing the dimensions of the plot ROTATE rotates the picture 90 PostScript creates a PostScript publication quality picture in the file plotm ps toggles colors on and off producing a line only drawing COLOR var var will ether register off with COLOR toggle or cel default or can be changed using any parameters in the right margin control string as appropriate to problem XY YZ ZX alter plot perspective to corresponding planar combinations LABEL controls surface and cell labels LEVEL Toggles through universe levels in repeated structures geometry Cell line Toggles through no lines cell lines ww mesh lines ww cell MCNPX User s Manual 45 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 1 Interactive Geometry Plotter Commands Command Result RIGHT MARGIN COMMANDS Used in Edit COLOR and LABEL cel cell labels colors will be cell numbers imp cell labels colors will be importances rho cell labels colors will be atom densities den cell labels colors will be mass densities l cell labels colors will be volumes calculated or user sup plied fcl cell labels colors will be forced collisions by particle type mas cell labels colors will be masses pwt cell labels colors will be photon production weights mat cell labels colors will be material numbers default tmp
99. function value multiplies the tally increment for IRSP lt 07 it divides the tally increment There are five interpolation schemes that may be specified individually for each energy interval in the response function tabulation using the following values for IRESP I 1 Constant the response function value is the value at the lower energy of the interval 2 Linear linear the response function is interpolated linearly in energy 3 Linear log the response function is interpolated linearly in the logarithm of the energy MCNPX User s Manual 221 MCNPxX User s Manual Version 2 4 0 September 2002 LA CP 02 408 4 Log linear the logarithm of the response function is interpolated linearly in energy 5 Log log the logarithm of the response function is interpolated linearly in the logarithm of the energy Any value of IRESP I outside the range 1 5 is treated as 1 i e constant over the interval The energy range for the specified response function need not span all possible particle energies in the problem If a particle energy falls below ERESP 1 then FRESP 1 is used as the value of the response function Similarly if a particle energy exceeds ERESP NRESP then FRESP NRESP is used as the value of the response function 23 Executing HTAPE3X The default file name for the input is INT the default file name for the output is OUTT the default file name for the history file is HISTP and the default file name for
100. gamma xn 0 0 0 particle decay 0 0 0 adjoint splitting 0 0 0 total 394337 1 9713E 01 3 4016E 02 total 394337 1 9713E 01 3 4016E 02 number of neutrons banked 368985 average time of shakes cutoffs neutron tracks per source particle 1 971 7E 01 escape 5 7458E 00 tco 1 0000E 34 neutron collisions per source particle 2 7874E 01 capture 4 6648E 01 eco 0 0000E 00 total neutron collisions 557485 capture or escape 5 7417E 00 wcl 5 0000E 01 net multiplication 0 0000E 00 0000 any termination 5 3201E 00 wc2 2 5000E 01 The two methods for calculating total neutron production give the following results net nuclear interactions net n xn 15 801 0 1834 3 9123 1 2660 18 263 n p escapes captures 18 249 0 014226 18 263 n p Both methods give the same answer Since escapes captures is easier to calculate this is the method typically used A reasonable upper limit on the relative uncertainty of n p is 20 000 0 7 Case 1 The first variation considered is the impact of the extension of the evaluated neutron cross sections to 150 MeV on total neutron production To evaluate this impact we set the transition energy between LAHET physics and neutron transport using evaluated nuclear data given by the third value on the phys n card to 20 MeV 1000 Base Case phys n j 150 Case 1 phys n1000 j 20 In this case neutron transport is done in the same manner as was done traditionally with LAHET and HMCNP The neutron pr
101. in MCNPX are also held on a regular basis http mcnpxworkshops com MCNPX User s Manual 3 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 MCNPX User s Manual 4 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 2 Warnings and Limitations All computer simulation codes must be validated for specific uses and the needs of one project may not overlap completely with the needs of other projects It is the responsibility of the user to ensure that his or her needs are adequately identified and that benchmarking activities are performed to ascertain how accurately the code will perform The benchmarking process for the Accelerator Production of Tritium project is extensive yet does not cover the entire range of possible output of MCNPX The results of these activities will be published separately and the code development team will strive to make available results from other projects We also solicit your input for potential code features MCNPX is a superset of MCNP4C3 and can generally be expected to track MCNP4C3 MCNP xX is guaranteed to do everything MCNP4C3 does as well or better The following warnings and known bugs apply to the energies and particles beyond MCNP 1 Pertubation methods used in MCNP have not yet been extended to the non tabular models present in MCNPX MCNPX crashes if run for problems that invoke the pertu bation capabilities above the MCNP energy range or beyond the MCNP
102. is recorded on the history tape The default is 0 however some options require that a value be supplied KOPT defines a sub option for tally option IOPT The default is 0 NPARM usually defines the number of cells materials or surfaces over which the tally is to be performed when applicable the maximum is 400 If NPARM is preceded by a minus sign NPARM I normalization divisors will be read in as described below The default is 0 however some options require that a value be supplied NFPRM at present is used only to define the number of cosine bin boundaries to read in for particle current tallies the maximum is 400 If NFPRM is preceded by a minus sign cosine bin tallies will be normalized per steradian the total over cosine bins will remain unnormalized i e angle integrated The default is 0 Table B 3 Particle Type Identification in HTAPE3X Type LAHET Usage MCNPX Usage 0 proton proton p 1 neutron neutron n 2 qt T T 3 9 n 4 7 5 ut 138 MCNPX User s Manual Accelerator Production of Tritium Table B 3 Particle Type Identification in HTAPE3X Continued MCNPX User s Manual Version 2 3 0 April 2002 Type LAHET Usage MCNPX Usage u Ww 7 deuteron deuteron 8 triton triton 9 3He 3He 10 alpha alpha 11 photon photon 12 Kt K KT 13 Klong Klong 14 K short K short 15 K 16 p 17 n 18 electron electron p
103. it needs to known on where to find individual data tables MCNPX uses the same procedure as MCNP to find the nuclear data libraries as described in Appendix F of the MCNP manual If XSDIR is not in your current directory MCNPX will search the following places for both the libraries and XSDIR file in order starting from 1 We repeat that portion of the MCNP manual here with annotations 1 xsdir datapath on the MCNPX execution line note datapath is truncated to 8 characters which means that it is really the name of a file not a path It is easiest to assign a name via a symbolic link e g In s home me lib data xsdir xsdir1 Then you can say mcnpx xsdir xsdir1 MCNPX User s Manual 27 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 2 DATAPATH datapath in the INP file message block Gt ON Ee this version of datapath can be a full description the current directory the DATAPATH entry on the first line of the XSDIR file the UNIX environmental variable setenv DATAPATH datapath the individual data table line in the XSDIR file the directory specified at MCNPX compile time in the blkdat f BLOCK DATA subrou tine This can be edited to change the directory but the code must be recompiled MCNPX has come up with the following slightly modified set of directions In the following cases if the desired file is found exit the list with the success 1 Look in the current working dire
104. lan gov data he html CHA81 A Chatterjee K H N Murphy and S K Gupta Pramana 16 1981 p 391 CHE76 V A Chechin and V C Ermilova The lonization Loss Distribution at Very Small Absorber Thickness Nucl Instr Meth 136 1976 551 CHE68 K Chen et al Phys Rev 166 1968 p 949 118 MCNPX User s Manual MCNPX User s Manual Ap Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium CLO83 P Cloth et al The KFA Version of the High Energy Transport Code HETC and the Generalized Evaluation Code SIMPEL Jul Spez 196 Kernforschungsanlage Julich GmbH MARCH 1983 CLO88 P Cloth et al HERMES A Monte Carlo Program System for Beam Materials Interaction Studies Kernforschungsanlage Julich GmbH Jul 2203 May 1988 COLOO G Collazuol A Ferrari A Guglielmi and P R Sala Hadronic Models and Experimental Data for the Neutrino Beam Production Nuclear Instruments amp Methods A449 609 623 2000 COU97 J D Court Combining the Results of Multiple LCS Runs memo LANSCE 12 97 43 Los Alamos National Laboratory May 8 1997 COU97a_ J D Court More Derivations Combining Multiple Bins in a MCNP or LAHET Tally memo LANSCE 12 97 66 Los Alamos National Laboratory July 16 1997 CMU94 Carnegie Mellon University Software Engineering Institute The Capability Maturity Model Guidelines for Improving the Software Process Addison Wesley 1994 DRE81 L Dresn
105. last blank terminator card These cards must have a C anywhere in columns 1 5 followed by at least one blank Comment cards are printed only with the input file listing and not anywhere else in the MCNPX output file The FCn input card is available for user comments and is printed as a heading for tally n as a tally title for example The SCn card is available for user comments and is printed as a heading for source probability distribution n 4 1 7 Horizontal Inout Format Cell surface and data cards all must begin within the first five columns The card name or number and particle designator is followed by data entries separated by one or more blanks Blanks in the first five columns indicate a continuation of the data from the last named card An amp preceded by at least one blank ending a line indicates data will continue on the following card Data on the continuation card can be in columns 1 80 Completely 34 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 blank cards are reserved as delimiters between major sections of the input file An individual entry cannot be split between two cards There can be only one card of any given type for a given particle designation see section 4 1 10 on page 38 Integers must be entered where integer input is required Other numerical data can be entered in any form acceptable to a FORTRAN E edit descriptor 4 1 8 Repeat Interpolate Multiply and Jump
106. le 9 3 1 1 In the Beginning 0 00sec eee eee eee eee eee eee 9 3 1 2 Automated Building 200 e eee eee 10 3 1 3 Build Examples 200 cece eee eee eee eee eee 12 3 1 3 1 System Wide Installation 0002 eee 12 3 1 3 2 System Wide Installation With Existing Directories 13 3 1 3 3 Individual Private Installation 0000 eee eae 14 3 1 3 4 Individual Private Installation Done Better 15 3 1 3 5 Individual Private Installation special compilers and debugging 16 3 1 4 Directory Reorganization 00c eee eee eee eee 18 3 125 User s Notes 302 see Se Sel amie opr neat ENEE ded ee ete ee 19 3 1 6 Multiprocessing 0 c cece eee eee 25 3 1 7 Programmer s NoteS 0 0 20 cece eee eee eee 25 3 2 Windows Build System 00 c eee 26 3 3 Libraries and Where to Find Them 000 cece ee eee eee eens 27 4INDUt Files iirid patented cet Oe he an eee heater hee eee 31 AN INP RILES ce a aes tere Sia ote ee etree ed ee cee wae eee eee a A 31 4 11 Initlate RUN ei anos oo ore Se A ee ae ee ee ee eta eerie 31 4 1 2 Continue Run oe bese ee ee eee ee eae 32 4 1 3 Message Block 200 c cece eee eee eee 34 4 1 4 Problem Title Card 000 c cee eee 34 4 1 5 Card Format 03604800 ees eee ae ee tee a dea ee 34 4 1 6 Comment Cards 0 0c cee 34 4 1 7 Horizontal Input Format 0 0 0 c eee eee 34 4 1 8 Repeat
107. level density parameters and fission models All of these are external to the particular intranuclear cascade pre equilibrium model chosen Bertini ISABEL or CEM and may be used with any of these choices Table 5 43 LEA Keyword Descriptions Keyword Description IPHT 0 Do not generate photons in the evaporation stage 1 Generate de excitation photons default ICC Defines the level of physics to be applied for the PHT physics 0 The continuum model 1 Troubetzkoy E1 model 2 Intermediate model hybrid between 1 and 2 3 The spin dependent model 4 The full model with experimental branching ratios default NOBALC 0 Use mass energy balancing in the cascade phase 1 Turn off mass energy balancing in the cascade phase default Note A forced energy balance may distort the intent of any intranuclear cas cade model Energy balancing for the INC is controlled by the input parame ter FLIMO NOBALE 0 Use mass energy balancing in the evaporation stage default 1 Turn off mass energy balancing in the evaporation stage IFBRK 1 Fermi breakup model for A lt 13 and for 14 lt A lt 20 with excitation below 44 MeV default 0 Use Fermi breakup model only for A lt 5 MCNPX User s Manual 95 MCNPxX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 43 LEA Keyword Descriptions Continued Keyword Description ILVDEN 1 Use original HETC level d
108. located without any transport the only possible outcomes are a nuclear interaction or no interaction The procedure may be used to calculate double dif ferential particle production cross sections from any of the interaction models in the code Bertini ISABEL CEM etc the procedure has no meaning if such a model is not allowed for the specified particle type at the specified energy 2 Input for MCNPX Since there is no way to avoid the MCNPX geometry input the user should define a region containing the material for which the cross sections are desired and locate the source in that region To avoid possible error only one material should be defined Note with N1COL 1 MCNPX will override the source specification and construct the source as a pencil beam in the z direction as required by XSEX3 Other MCNPX options may be used to suppress either nuclear elastic or nonelastic reactions 1 To create a HISTP file to be edited by XSEX3 include a HISTP card in the INP file 2 Define a volume parallel beam source in the z direction vec 0 0 1 which is com pletely contained inside a cell with the material for which the cross sections are to be calculated 3 Specify the incident particle type and kinetic energy on the SDEF card 4 Use NOACT 1 the 8th parameter on the LCA card The user may wish to suppress nuclear elastic scattering in the calculation by using IELAS 0 on the LCA card An AWTAB card may need to be supplied if t
109. lowest nonzero importance for that energy group 2 means that weight windows do what they normally do Controls adjoint biasing for adjoint problems only MCAL A 0 means collisions are biased by infinite medium fluxes ISB default 1 means collisions are biased by functions derived from weight windows which must be supplied 2 means collisions are not biased name of the reference cell for generated weight windows 0 means weight windows are not generated default ICW c 0 requires volumes be supplied or calculated for all cells of nonzero importance normalization value for generated weight windows The ENW value of the weight window lower bound in the most important energy group in cell ICW is set to FNW default 1 80 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 30 Multigroup Adjoint Transport Option Keyword Description compression limit for generated weight windows Before generated weight windows are printed out the weight RIM windows in each group separately are checked to see that the ratio of the highest to the lowest is less than RIM If not they are compressed default 1000 NOTE MCAL and IGM must be specified J is not an acceptable value for any of the parameters Use Required for multigroup calculation Presently the standard MCNPX multigroup neutron cross
110. lt 70 The default is 8 0 zero or negative is an error condition see YZERE above Note Applies only for ILVDEN 1 YZERO The YO parameter in the level density formula for Z 71 and all fission frag ments The default is 1 5 Zero and negative values are an error condition see YZERE above Note Applies only for ILVDEN 1 BZERO The BO parameter in the level density formula for Z 71 and all fission frag ments The default is 10 0 for IEVAP 0 and is also 10 0 for IEVAP 1 Zero and negative values are an error condition see YZERE above Note Applies only for ILVDEN 1 MCNPX User s Manual 83 MCNPX User s Manual E Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium 6 3 Extended Source Options The extended source option has been adapted from similar capabilities in the LAHET Code System as described in PRA99 Two features have been added in MCNPX to the MCNP general source routines The first is a simple modification that permits the use of an f 41 Gaussian probability distribution for the X Y or Z positional parameters on the SDEF card In MCNP the 41 option has been used for a time gaussian distribution in MCNPxX the fatal error for specifying the spa tial option has been removed This allows creation of a Gaussian beam profile however the user should keep in mind that many realistic accelerator beams are only approximately Gaussian and normally have enhanced tails d
111. mcnpx source code in which to do the build This can be done several times in different build directories with different options such as debugging non debugging versions or different compiler types The local user building the private copy is again username me whose home directory is the directory home me The user has fetched the distribution from CDROM or from the net and has it in the file nhome me mcnpx_2 4 0 tar gz The user will unload the distribution package into home me mcnpx_2 4 0 With this method the source can be anywhere as long as the user has the pathname to it The user will build the system in the local directory home me mcnpx install the binary executable in home me bin and install the binary data files and eventually the mcnp cross sections in nome me lib MCNPX User s Manual 15 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 The following example uses bourne shell commands to accomplish this task If you are more familiar with csh you will need to adjust things appropriately NOTE Comments about the shell commands start with the character Also don t be alarmed by the generous amount of output from the configure and make scripts They work hard so you don t have to go to your user home directory cd home me unpack the distribution that was copied from the net or a CDROM This creates home me mcnpx_2 4 0 gzip dc mcnpx_2 4 0 tar gz tar xf make a local directory for
112. more important limitations that have to be considered when setting up a problem It may be necessary to modify MCNPX to change one or more of these restrictions for a particular problem Table 4 2 Storage Limitations Entries in the description of a cell 1000 after processing Total number of tallies NTALMX 100 Detectors MXDT 20 Neutron DXTRAN spheres MXDX 5 Photon DXTRAN spheres MXDX 5 NSPLT or PSPLT card entries 10 Entries on IDUM card 50 Entries on RDUM card 50 Set as a dimension in an array MCNPX User s Manual 43 44 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 MCNPX User s Manual 5 Plotting 5 1 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 THE INTERACTIVE GEOMETRY PLOTTER Table 5 1 Interactive Geometry Plotter Commands Command Result TOP MARGIN COMMANDS UP RT DN LF When clicked moves the plot frame up right down left respectively Origin When clicked followed by click on some point in the plot moves the origin to that point 1 2 5 Zo0om125 When double clicked at any point on the continuum results in a zoom of the corresponding fraction multiple When clicked followed by clicking on a point in the pic ture will zoom to that point LEFT MARGIN COMMANDS provides information for the plot cell number and coordi Eolit nates at the most recent cursor click point CURSOR
113. n xn 25539 1 2753E 00 4 9548E 01 fission 0 0 0 loss to fission 0 0 0 photonuclear 0 0 0 nucl interaction 3667 1 8335E 01 6 2061E 01 tabular boundary 0 QO QO tabular boundary 0 QO 0 gamma xn 0 0 0 particle decay 0 0 0 adjoint splitting 0 0 0 total 395962 1 9794E 01 3 4090E 02 total 395962 1 9794E 01 3 4090E 02 number of neutrons banked 370423 average time of shakes cutoffs neutron tracks per source particle 1 9798E 01 escape 5 7616E 00 tco 1 0000E 34 neutron collisions per source particle 2 7981E 01 capture 4 8708E 01 eco 0 0000E 00 total neutron collisions 559626 capture or escape 5 7574E 00 wcl 5 0000E 01 net multiplication 0 0000E 00 0000 any termination 5 3337E 00 wc2 2 5000E 01 Calculated net neutron production for this case is 18 335 and examination of the net nuclear interactions and net n xn figures show very similar results to the base case The implication of this result is that we need not concern ourselves with light ion transport if the quantity with which we concerned is related solely to neutrons as neutron production by light ions is small when we start with a proton beam Case 3 In this variation we replace the Bertini INC model used in the base case for the simulation of nucleon and pion interactions with nuclei by the ISABEL INC model in this example both INC models utilize the same GCCI level density model We invoke the ISABEL INC model by including in the input deck the following card Base C
114. neu tron component on the earth s surface e Detection technology using charged particles i e abandoned landmines In addition to the activities of the beta test team the development of MCNPX is governed by several documents including MCNPX Software Management Plan e MCNPX Requirements MCNPX Design e MCNPX Functional Specifications Configuration management of the code is done through CVS which allows us to conve niently track issues and changes A computer test farm of 20 different software hardware configurations is maintained to ensure that code development does not adversely any pre viously tested system We are also constantly moving toward a modular system whereby the user may easily implement alternative physics packages EGD01 Some restructuring of the code has already been done toward that goal including the development of an autconfiguration system In addition to describing the new interaction physics this manual contains a summary of information from recent MCNPX release notes memos publications and presentations It represents the work of the code development team the nuclear data team the physics development team and several outside collaborators The manual is updated and extended with each new code release Not all of the capabilities of MCNP4B are fully present in MCNPX version 2 3 0 and in addition the reader must be aware of certain limitations in code usage These items are listed in Chapter
115. of tallies is a sum but for normalized tallies types 2 4 6 and 7 the union results in an average See Section 5 7 1 2 for an explanation of the repeated structure and lattice tally format The symbol T entered on surface or cell Fn cards is shorthand for a region that is the union of all of the other entries on the card A tally is made for the individual entries on the Fn card plus the union of all the entries If a tally label of the surfaces or cells in the output requires more than eleven characters including spaces MCNP defines an alphabetical or numerical designator for printing purposes The designator for example G is 1 2 3 4 5 6 is printed with the tally output This labeling scheme is usually required for tallies over the union of a long list of surfaces or cells Energy Deposition Tally F6 Note In the energy range where tables are available the neutron and proton energy deposition is determined using the neutron heating numbers in the same manner as F6 tallies are 114 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 done in MCNP These heating numbers are estimates of the energy deposited per unit track length In addition the de dx ionization contribution for the proton is added in similar to the electron treatment Above that tabular energy limit or when no tabular data is available energy deposition is determined by summing several factors For charged particles i
116. of the segmenting record of option 1 a window definition record appears whose form is described in option 9 For KOPT 0 the rectan gular form is used and for KOPT 1 the circular form is used Parameter NFPRM is unused 13 Edit Option lIOPT 11 or 111 Pulse Shape of Surface Current For each defined bin option 11 provides an edit of the current crossing a surface in an energy and angle bin the mean time t of crossing in the bin the standard deviation o of t given by t the figure of merit FOM1 given by current o and the figure of merit FOM2 given by current o Unless otherwise modified the current tally is dimensionless The units of t and o are nanoseconds while FOM1 is in ns and FOM2 is in ns The parameter FNORM is used to adjust the units of the time variable which are nanoseconds in LAHET3 and does not modify the surface current edit Thus to convert from nanoseconds to microseconds use FNORM 0 001 The bin definition is identical to option 1 including surface segmenting except that NTIM is unused 146 MCNPX User s Manual MCNPX User s Manual Ap Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium 14 Edit Option IOPT 12 or 112 Pulse Shape of Surface Current with Window Option 12 provides the same edits as option 11 with the same bin definition as option 9 using a collimating window The input is identical to option 9 with the exception that NTIM is
117. or 1 GeV per nucleon for composite particles although it may execute at higher energies MCNPX User s Manual 77 MCNPX User s Manual E Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium Table 6 3 LCA Keyword Descriptions Continued Keyword Description ICHOIC 4 integers ijkl which control ISABEL INC Model default 0023 i 0 Use partial Pauli blocking i 1 Use total Pauli blocking i 2 No Pauli blocking not recommended j 0 No interaction between particles already excited above the Fermi Sea j gt 0 Number of time steps to elapse between such CAS CAS interactions k 0 Meyer s density prescription with 8 steps k 1 Original isobar density prescription with 8 steps k 2 Krappe s folded Yukawa prescription for radial density in 16 steps with a local density approximation to the Thomas Fermi distribution for the sharp cutoff momen tum distribution k 3 The same as k 0 but using the larger nuclear radius of the Bertini model k 4 The same as k 1 but using the larger nuclear radius of the Bertini model k 5 The same as k 2 but using the larger nuclear radius of the Bertini model 1 Reflection and refraction at the nuclear surface but no escape cutoff for isobars 2 Reflection and refraction at the nuclear surface with escape cutoff for isobars 3 No reflection or refraction with escape cutoff for isobars 4 The same as l 1 but using a 2
118. our deepest thanks is extended to Dr Richard E Prael for his support and guidance Without his longtime vision of providing the highest quality simulation tools to the accelerator community the MCNPX project could not have happened MCNPX 2 3 0 is based on MCNP4B and we gratefully acknowledge the importance of that seminal code in our work The MCNP code series represents many thousand person years of effort over the past 30 years and we hope our efforts will add new vistas to this core capability Our special thanks goes to Dr John Hendricks and Dr Gregg McKinney MCNPX User s Manual iii MCNPX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium as well as the numerous contributors who over the years have made MCNP a world class code We also wish to express our appreciation to Dr Alfredo Ferrari currently with CERN for allowing the use of an early version of the FLUKA code in MCNPX permitting a significant expansion of our upper energy limits We will endeavor in future versions of the code to upgrade this capability In addition we wish to express our fond appreciation for the efforts of Dr Stepan Mashnik who has improved the CEM code for inclusion in MCNPX Dr Nikolai Mokhov of Fermi National Laboratory has provided improved high energy photonuclear physics routines that will be implemented in future versions of the code We also wish to thank him for his part in the formal review
119. particles is also implemented There is currently no delta ray production of knock on electrons for charged heavy particles in MCNPX although it is present for electrons 40 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 No option for electromagnetic field tracking is currently implemented in MCNPX 4 1 11 Default Values Many MCNPX input parameters have default values that are summarized in Section 5 10 Therefore you do not always have to specify explicitly every input parameter every time if the defaults match your needs If an input card is left out the default values for all parameters on the card are used However if you want to change a particular default parameter on a card where that parameter is preceded by others you have to specify the others or use the nJ jump feature to jump over the parameters for which you still want the defaults CUT P 3J 10 is a convenient way to use the defaults for the first three parameters on the photon cutoff card but change the fourth 4 2 INPUT ERROR MESSAGES MCNPX makes over 400 checks of the input file for user errors A fatal error message is printed both at the terminal and in the OUTP file if the user violates a basic constraint of the input specification and MCNPX will terminate before running any particles The first fatal error is real subsequent error messages may or may not be real because of the nature of the first fatal messag
120. path of cross sections and other data files A special autoconf generated configure script distributed with MCNPX will examine your computing environment adjust the necessary parameters then generate all Makefiles in your chosen build directory so that they all match your particular computing environment 10 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 The full structure is now in place to allow a graceful migration to individual feature tests during the autoconfiguration process in the future The autoconf generated configure script will search for GNU compilers first before attempting to locate any other compiler present on your computing environment Please be aware of exactly how many Fortran and C compilers exist in your computing environment It may be necessary to specify which Fortran and C compiler should be used You have that power via options given to the configure script See the with FC and with CC options later in this document Rather than having the one Build directory of past distributions one is now free to create as many build directories as desired anywhere one wants named anything one wants Through the use of options supplied to the configure script one can vary the resulting generated Makefiles to match a desired configuration Most software packages that use autoconf have a basic build procedure that looks like gzip dc PACKAGE tar gz tar xf cd PACKAGE
121. pre equilibrium models are used to describe this phase in which high energy particles and light ions are emitted able to interact with other nuclei In Sections 4 1 1 through 4 1 6 we give more detail on the various physics models used to simulate these processes Table 4 1 compares the three MCNPX options in terms of the differences in these components Table 4 2 gives the working range of validity for each 38 MCNPX User s Manual MCNPxX User s Manual Version 2 3 0 April 2002 ry LA UR 02 2607 Accelerator Production of Tritium first stage intranuclear cascade a T high energy proton intermediate stage preequilibrium Ss be second stage evaporation and or fission i s N I 4A eo a s a Se 7 i SAN e e y final stage residual deexictation Vane Figure 4 1 Interaction processes MCNPX User s Manual 39 Accelerator Production of Tritium MCNPxX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 Table 4 1 Summary of Physics in Intermediate Energy Models Physics Process Bertini ISABEL CEM Method INC EQ or INC EQ or INC PE EQ INC PE EQ INC PE EQ Intranuclear Cascade Model Bertini INC ISABEL INC improved Dubna INC Monte Carlo Technique spacelike timelike spacelike Nuclear Density Distribu tion P r polexp r c a 1 c 1 07A1 fm a 0 545 fm p r a ip 0 i 1
122. protons is also randomized The randomized cutoff energy is the default CTOFE 1 0 For the ISABEL INC the randomized cutoff energy is always used FLIMO The maximum correction allowed for mass energy balancing in the cascade stage used with NOBAL 1 and NOBAL 3 FLIMO gt 0 Kinetic energies of secondary particles will be reduced by no more than a fraction of FLIMO in attempting to obtain a non negative excitation of the residual nucleus and a consistent mass energy balance A cascade will be re sampled if the correction exceeds FLIMO FLIMO 0 No correction will be attempted and a cascade will be re sampled if a negative excitation is produced FLIMO lt 0 default 1 0 The maximum correction is 0 02 for incident energy above 250 MeV 0 05 for incident energy below 100 MeV and is set equal to 5 incident energy between those limits As an example consider LCB 3000 3000 2000 2000 1000 1000 For IEXISAQ 1 the default nucleons will switch to the BERTINI model from the FLUKA model below 3 GeV and Pions would switch below 2 GeV Kaons and anti nucleons would switch to the ISABEL model from the FLUKA model below 1 GeV lons use only the ISA BEL model and muons have no nuclear interactions For lIEXISA 2 nucleons and pions would also switch to the ISABEL model below 1 GeV Note that the nominal upper energy limit for the ISABEL model is about 1 GeV nucleon it may actually execute at higher energies without crash
123. represents many thousand person years of effort over the past 30 years and we hope our efforts will add new vistas to this core capability Our special thanks goes to Dr John Hendricks and Dr Gregg McKinney as well as the numerous contributors who over the years have made MCNP a world class code We also wish to express our appreciation to Dr Alfredo Ferrari currently with CERN for allowing the use of an early version of the FLUKA code in MCNPX permitting a significant expansion of our upper energy limits We will endeavor in future versions of the code to MCNPX User s Manual iii MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 upgrade this capability In addition we wish to express our fond appreciation for the efforts of Dr Stepan Mashnik who has improved the CEM code for inclusion in MCNPX Dr Nikolai Mokhov of Fermi National Laboratory has provided improved high energy photonuclear physics routines that will be implemented in future versions of the code We also wish to thank him for his part in the formal reviews of our work Several visitors have provided invaluable help to the nuclear data team with evaluations notably Dr Satoshi Chiba JAERI and Dr Arjan Koning ECN Petten Of special note is the valuable help given us by those sponsoring MCNPX classes includ ing William Hamilton of HQC Professional Services Inc Enrico Sartori of NEA Tadakazu Suzuki of JAERI and Pedro Vaz of ITN Portugal
124. return not immediately preceded by an amp or by a COPLOT command Commands consist of keywords usually followed by some parameters entered space or comma delimited MCNPX User s Manual 48 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Defaults are available for nearly everything If MCNP is run with Z as the execute line message and if file RUNTPE is present with more than one energy bin in the first tally and if a carriage return is entered in response to the MCPLOT prompt a lin log histogram plot of tally MeV vs energy with error bars and suitable labels will appear on the screen 5 2 2 Plot Conventions and Command Syntax 5 2 2 1 2D plot The origin of coordinates is at the lower left corner of the picture The horizontal axis is called the x axis It is the axis of the independent variable such as user bin or cell number or energy The vertical axis is called the y axis It is the axis of the dependent variable such as flux or current or dose Each axis can be either linear or logarithmic 5 2 2 2 Contour plot The origin of coordinates is at the lower left corner of the picture The horizontal axis is called the x axis It is the axis of the first of the two independent variables The vertical axis is called the y axis It is the axis of the second independent variable The contours represent the values of the dependent variable Only linear axes are available 5 2 2 3 Command syntax Each command con
125. run The tally numbers are entered on the TALNP card as negative numbers 8 2 4 Reading the Radiography Tally Output The output of the two radiography tally options is contained in the mctal file It can be for matted for use with external graphics programs with the gridconv routine The user is referred to Section 8 1 2 for information on how to use gridconv MCNPX User s Manual 107 MCNPX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium 8 3 Energy Deposition With the high energy extensions in MCNPX considerable thought has gone into the design and adoption of energy deposition tallies In particular we must address such issues as 108 Local energy deposition of non tracked particles is not valid as particle energy increases Heating numbers and Kerma factors do not exist in the physics modules Energy dep osition processes must be modeled online as interactions occur and the individual contributions summed This process is termed collision based estimate Track ionization for charged particles is not linearly distributed over a step but can increase or decrease as the particle slows down depending on initial energy MCNPX 2 3 0 always scores the energy of a particle at the beginning of a step In most cases step sizes for charged particles are small therefore little error is introduced in this pro cess However occasionally particles may lose so much energy in one t
126. sections are given in 30 groups and photons are given in 12 groups Thus an existing continuous energy input file can be converted to a multigroup input file simply by adding one of the following cards MGOPT F 30 MODEN MGOPT F 42 MODENP MGOPT F 12 MODE P 5 4 11 DRXS Discrete Reaction Cross Section Form DRXS ZAID ZAID ZAID or blank ZAID Identifying number of the form ZZAAA nn where ZZ is the atomic number AAA the mass number and nn the neutron library identifier Use Discouraged Default Continuous energy cross section treatment if DRXS is absent Example DRXS A blank DRXS card will use discrete reaction neutron data wherever possible MCNPX User s Manual 81 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 5 5 PHYSICS MODE PHYS TMP THTME COINC CUT ELPT NPS CTME LCA LCB LEA LEB 5 5 1 MODE Problem Type Form MODE x4 Xi xi particle designator The MODE card can take any argument listed in the Symbol column of Table 4 1 in any order It must list all particles that will be transported If a particle is designated the anti particle will also be transported For example MODE n e will transport neutrons and anti neutrons protons and anti protons u and uw electrons and positrons Default If the MODE card is omitted MODE N is assumed 5 5 2 PHYS Energy Physics Cutoff 5 5 2 1 Neutrons Form PHYS n EMAX EAN IUNR DNB TABL FISM RECL Table 5
127. signs are optional Table 5 52 Surface Source Write Card MCNPX User s Manual Variable Description problem surface number with the appropriate sense of s inward or outward particle direction for which parti cle crossing information is to be written to the surface source file WSSA Macrobody facets are allowed Ci problem cell number keyword Values m symmetry option flag m 0 no symmetry assumed m 1 spherical symmetry assumed The list of problem SYM surface numbers must contain only one surface and it must be a sphere m 2 write particles to a surface bidirectionally Otherwise only particles going out of a positive surface and into a negative surface are recorded nNyNp tracks to record absent record all tracks This is the default PTY nj N record neutron tracks n P record photon tracks n E record electron tracks 103 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 52 Surface Source Write Card Variable Description CCo gt gt Cy list of names of all the cells from which CEL KCODE fission source neutrons are to be written active cycles only Default SYM 0 PTY absent record all particle types Use Optional as needed 5 6 5 SSR Surface Source Read Form SSR keyword value keyword value The signs are optional Table 5 53 Surface Source Read Card Keyword Description S1 So Sp lis
128. source only and a total contribution but could be further binned as described for the pinhole tally The transmitted image projection is set up as follows in version 2 1 5 Fin P X1 Y1 Z1 RO X2 Y2 Z2 F1 F2 F3 Note that this form is the same as the pinhole image The transmitted image capability is turned on by setting F2 less than zero as described below Version 2 3 0 changes the form of the card old input files are backward compatible if one replaces the control card symbol 104 MCNPX User s Manual MCNPxX User s Manual i Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium TI R C n P X1 Y1 Z1 RO X2 Y2 Z2 F1 F2 F3 TIR is used to establish a grid on a plane surface TIC is used to establish a grid on a cylindrical surface n is the tally number and must be a multiple of 5 since this is a detector type tally P is the particle type for the tally Only neutrons or photons are allowed since detector techniques do not currently work for charged particles Table 8 6 Transmitted Image Projection Argument Description Argument Description X1 Y1 Z1 The coordinates used with the entries on the FSn and Cn cards to define the detector grid In the plane grid case this defines the center of the grid In the cylindrical grid case this defines the center of the cylinder on which the grid is established RO Always 0 zero in this application as in the pinhole case X2 Y2 Z2 The refer
129. stage HISTP may be edited as noted in comment 3 above 116 MCNPX User s Manual MCNPX User s Manual Ap Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium 9 References ARM73 T W Armstrong and K C Chandler SPAR A FORTRAN Program for Com puting Stopping Powers and Ranges for Muons Charged Pions and Heavy lons ORNL 4869 Oak Ridge National Laboratory May 1973 AAR86 P A Aarnio A Fasso H J Moehring J Ranft and G R Stevenson CERN TIS RP 186 1986 FLUKA 86 users guide AAR87 P A Aarnio J Lindgren A Fasso J Ranft and G R Stevenson CERN TIS RP 190 1987 FLUKA 87 AAR90 P A Aarnio e al FLUKA89 Consiel Europeene Organisation pour La Recherche Nucleaire informal report January 2 1990 ART88 E D Arthur The GNASH Preequilibrium Statistical Model Code LA UR 88 382 Los Alamos National Laboratory February 1988 ATC80 F Atchison Spallation and Fission in Heavy Metal Nuclei under Medium Energy Proton Bombardment in Targets for Neutron Beam Spallation Sources Jul Conf 34 Kernforschungsanlage Julich GmbH January 1980 BAR73 V S Barashenkov A S Iljinov N M Sobolevskii and V D Toneev Interac tion of Particles and Nuclei of High and Ultrahighy Energy with Nuclei Usp Fiz Nauk 109 1973 91 Sov Phys Usp 16 1973 31 BAR81 J Barish T A Gabriel F S Alsmiller and R G Alsmiller Jr HETFIS High Energy Nucleon
130. standard Lane model assumption and by accounting approximately for the Coulomb correction Final comparisons of predicted and measured elastic scattering observables for both protons and neutrons were made for 7AI Fe and 2 8Pb The results were generally good 52 MCNPX User s Manual MCNPX User s Manual i Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium The neutron and proton elastic cross sections so generated are tabulated for 9 mass val ues and 20 energies between 50 and 400 MeV Above 400 MeV the tabulations from HERMES are used and the HERMES neutron elastic cross section tabulation below 50 MeV has been extended to lower and higher masses to minimize mass extrapolation error Proton elastic scattering vanishes below 50 MeV in this implementation Examples of the use of the elastic scattering data can be found in PRA95 MCNPX 2 3 0 MCNPX 2 3 0 users will notice a new file called BARPOL dat This contains improved data on reaction and elastic cross sections which is used in the physics as opposed to the library regions of the code The old method described above has been retained although the new one is the default in compiling the code To access the old method com pile the code with the with OLDXS option from table 3 1 Previously the concept of a reaction cross section for use with the intranuclear cascade model has been implicit in the model and not explicitly defined fo
131. tally bins subdivisions of the tally space into discrete and contiguous increments such as cosine energy or time Usually when the user subdivides a tally into bins MCNP can also provide the total tally summed over appropriate bins Such as over energy bins Absence of any bin specification card results in one unbounded bin rather than one bin with a default bound No information is printed about the limits on the unbounded bin If there are reflecting surfaces or periodic boundaries in the problem the user may have to normalize the tallies in some special way this can be done by setting the weight of the source particles or by using the FMn card Printed with each tally bin is the relative error of the tally corresponding to one standard deviation These errors cannot be believed reliable hence neither can the tally itself unless the error is fairly low Results with errors greater than 50 are useless results between 20 and 50 can be believed to within a factor of a few results between 10 and 20 are questionable results less than 10 are generally but not always reliable except for detectors and detector results are generally reliable below 5 One bin of every tally is designated for the tally fluctuation charts at the end of the output file This bin is also used for the weight window generator It also is subject to ten statistical checks for tally convergence including calculation of the variance of the variance VOV Th
132. the evaporation phase production is edited For KOPT 6 or 7 only the total particle production is edited For KOPT 8 or 9 only the pre fission evaporation production is edited For KOPT 10 or 11 only the post fission evaporation production is edited If KOPT is even the edit is over cell numbers if KOPT is odd the edit is over material numbers If NPARM is zero the edit is over the entire system The parameters NTYPE and NFPRM are not used If KPLOT 1 a plot is made of each edit table With KOPT 0 or 1 the cascade production for neutrons and protons is simultaneously plotted as a dotted line with the total production Unless otherwise modified tally option 3 or 103 represents the weight of particles emitted in a given bin per source particle As such it is a dimensionless quantity 6 Edit Option IOPT 4 or 104 Track Length Estimate for Neutron Flux Option 4 is not available in this version use a standard F4 flux tally 7 Edit Option IOPT 5 or 105 Residual Masses and Average Excitation Option 5 provides an edit by mass number A of the calculated residual masses and the average excitation energy for each mass Only nonelastic interactions are included The option accesses the records on HISTP for all interacting particle types The edit is performed for both the final residual masses and the residuals after the cascade phase If IOPT is preceded by a minus sign the edit is performed for events initiated by primar
133. the secondaries are high or when the user is simulating thin volumes When secondary particles are indicated on the MODE card MCNPX will subtract their energies from the heating values and energy deposition will be handled in the regular process of tracking those particles Where there are no libraries available de dx nuclear recoil and the energies of some non tracked secondary particles are added to the F6 collision estimator A secondary particle can be produced either by collision or by particle decay In MCNPX the energies of neutral particles will never be added to the collision estimator this includes neutrons photons neutrinos pi0 and neutral Kaons This is not consistent with the library heating factor treatment and will be reconsidered in future versions of the code Therefore it is 1 In MCNPX residual nuclei cannot be tracked This is usually not a problem for heavy residuals however for light residuals such as a scattered hydrogen nucleus errors in energy deposition in small volumes can occur This has caused some users problems when tracking in small volumes where it is unlikely that the recoil hydrogen nucleus will not stop We will modify this practice in an upcoming release 2 Energies of particles which fall below minimum energy cutoffs will also be deposited locally The user must be certain that the value of these cutoff energies will not cause the results of the F6 tally to be in error 3 Note that the
134. the surface crossing file is HISTX for input into HTAPE3X The latter is written by MCNPX with the default file name WSSA If option 8 is requested the data file PHTLIB must be in the user s file space if option 16 is requested the data file BERTIN must be in the user s file space All these file names may be defined by file replacement on the execute line HTAPE3X INT my_input OUTT my_output HISTP file1 HISTX file2 References 1 R E Prael and H Lichtenstein User Guide to LCS The LAHET Code System LA UR 89 3014 Los Alamos National Laboratory September 1989 http www xdiv lanl gov XCI PROJECTS LCS lahet doc html 2 H G Hughes R E Prael and R C Little MCNPX The LAHET MCNP Code Merger X Division Research Note XTM RN U 97 012 LA UR 97 4891 Los Alamos National Laboratory April 1997 http www xdiv lanl gov XTM hughes LA UR 97 4891 cover html 3 J F Briesmeister editor MCNP A General Monte Carlo N Particle Transport Code Los Alamos National Laboratory report LA 12625 M March 1997 http Awww xdiv lanl gov XCI PROJECTS MCNP manual html 4 J Linhard V Nielsen and M Scharff Kg Dan Vidensk Selsk Mat Fys Medd 36 No 10 1968 222 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 5 M Robinson The Dependence of Radiation Effects on Primary Recoil Energy Radiation Induced Voids in Metals AEC Symp Ser 26 p 397 US Atomic Energy Comm
135. to 20 MeV Although the 150 MeV evaluations do include the detailed secondary infor mation in the 20 150 MeV range the lt 20 MeV data typically do not Therefore sec ondary production is ignored in processing that energy range Table 4 4 lists the actual secondary particle production thresholds in LA150N MCNPX User s Manual 5 10 11 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Fixing this situation is non trivial and involves a re evaluation of the low energy data Improved libraries will be issued but on an isotope by isotope basis No explicit generation of delta ray knockon electrons as trackable particles is done for heavy charged particles Delta rays will be produced for electrons Positrons may not be used as source particles Correcting this involves a change in the way the particle identification numbering system is handled for electrons and positrons Historically this has not been treated in the same way as the method used for neutrons in MCNP which forms the basis for the multiparticle extension of MCNPX Beware of the results of an F6 p tally in small cells when running a photon or photon electron problem Photon heating numbers include the energy deposited by electrons generated during photon collisions but assume that the electron energy is deposited locally In a cell where the majority of the electrons lose all of their energy before exit ing that cell this is a good approximatio
136. to make multiple versions with different options A better example will follow this one The following example uses bourne shell commands to accomplish this task If you are more familiar with csh you will need to adjust things appropriately NOTE Comments about the shell commands start with the character Also don t be alarmed by the generous amount of output from the configure and make scripts They work hard so you don t have to go to your user home directory cd home me 14 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 unpack the distribution that was copied from the net ora CDROM This creates home me mcnpx_2 4 0 gzip dc mcnpx_2 4 0 tar gz tar xf go into the unpacked distribution cd mcnpx_2 4 0 execute the configure script the prefix tells where to put the executables and libraries configure prefix home me Make the executable mcnpx program the bertin and pht libraries and run the regression tests make all make tests now install the executable mcnpx program and the bertin and pht libraries in home me bin and home me lib mcnpx make install 3 1 3 4 Individual Private Installation Done Better For a more flexible version of our second example we will look at the same single non privileged user Me on a computer loading and building a private copy of the code This time however the user will use a second directory away from the
137. to one correspondence with the LPARM array The last NPARM 1 entry applies to a total over the NPARM entities where applicable if omit ted it defaults to 1 0 Through this feature it is possible to input a list of volumes areas or masses as appropriate obtained from a MCNP calculation When IOPT gt 100 the NPARM cell surface or material identifiers are treated as a single entity in constructing a tally edit In this case the NPARM normalization divisors are summed to a single divisor Consequently one may supply the full list of divisors if appropriate or just supply one value for the common tally e For IRS gt 0 the original source definition record in LAHET format as described in Section 2 4 of reference 1 followed by the new source definition record also in LAHET format e ForlITCONV 0 a LAHET source time distribution record as described in Section 2 4 of reference 1 e For IRSP 0 three records defining the user supplied response function ERESP I l 1 NRESP a monotonically increasing energy grid on which the value of the response function is tabulated FRESP l l 1 NRESP the values of the response function at the above energies IRESP l l 1 NRESP 1 interpolation scheme indicators where IRESP I indi cates the interpolation scheme to be used for the response function in the I th energy interval The length NRESP lt 200 is obtained from the array ERESP input terminated by a The
138. translation and rotation according to the following equations where 0 lt 9 lt x x X sing y cosd Xo y X coso y sing Yo Thus the angle is the angle of rotation of the major axis of the source distribution from the positive y direction in the laboratory coordinate system If cos 0 0 the angle is 90 and the major axis lies along the x axis The TRn card in the above example imple ments this rotation matrix however the user is warned that in the TRn card is equal to IU OL 5 Defining Multiple Beams The opportunity to specify a probability distribution of transformations on the SDEF card is a new feature that goes beyond enabling the representation of LAHET beam sources It allows the formation of multiple beams which differ only in orientation and intensity a fea ture that may have applications in radiography or in the distribution of point sources of arbitrary intensity The use of a distribution of transformations is invoked by specifying TR Dn on the SDEF card The cards SI SP and optionally SB are used as specified for the SSR card on page 3 57 of the MCNP4B User s Guide Sin L l4 lk SPn option P4 Pk SBn option By The L option on the SI card is required new input checking has been implemented to ensure this usage for both the SDEF and SSR applications The option on the SP and SB cards may be blank D or C The values l4 l identify k transformations which must be supplied
139. typically do not Therefore secondary production is ignored in processing that energy range Table 4 4 lists the actual secondary particle production thresholds in LA150N Fixing this situation is non trivial and involves a re evaluation of the low energy data Improved libraries will be issued but on an isotope by isotope basis 6 MCNPX User s Manual MCNPX User s Manual i Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium 10 Light particle residual nuclei are not transported When a light charged particle is the residual nucleus in a nuclear reaction in the tabular range those charged particles are not produced banked and transported in MCNPX 2 3 0 Instead their energy is assumed to be deposited locally For example the residual proton from neutron elas tic scattering on H 1 is not produced or transported This will be resolved in a subse quent version of MCNPX 11 No explicit generation of delta ray knockon electrons as trackable particles is done for heavy charged particles in 2 3 0 Delta rays will be produced for electrons 12 The upper energy limit for photon transport is 100 GeV and for electron trans port is 1 GeV This is a standard feature of MCNP4B and has been inherited by MCNPX 2 3 0 Although adequate for most uses of MCNP4B higher energy problems often need increased upper energy ranges particularly at electron accelerators Future versions of MCNPX will remove these limitat
140. unused 15 Edit Option IOPT 13 Global Emission Spectrum The original definition I of option 13 was given by Option 13 tallies the number of particles per unit solid angle entering the external void region with direction cosine falling within a segment of solid angle as such it represents the angular distribution of the emitted particles at a very large distance from the interaction region The option uses any NCOL 4 leakage records on HISTP and all records on HISTX indiscriminately Surface crossing records appearing on a SSW written file are not distinguished as to whether they correspond to an internal surface crossing or to escape into the external void Therefore for use with MCNPX the original intent of this option may most easily be achieved by defining the external importance 0 leakage region as the exterior of a sphere containing the complete geometry then only specifying the defining spherical surface on the SSW card that controls the contents of the surface crossing file Energy binning is specified by the usual methods The number of energy bins is given by NERG The number of particle types for which surface crossing data are to be tallied is given by NTYPE and must be gt 0 The polar angle bins representing lines of latitude are defined by entering the NFPRM cosine values in the FPARM array Binning in the azi muthal angle corresponding to lines of longitude is determined by the value of NPARM which defines N
141. use extreme caution when doing this 72 MCNPX User s Manual MCNPX User s Manual Ap Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium Additional cards are needed when specifying photonuclear libraries Since there are a lim ited set of libraries available there may not be a photonuclear table corresponding to the neutron proton electron or photon ZAID s on the Mm card A new card has been devel oped to instruct the code what to do in such cases WHI00 MPNn PNZA PNZA PNZAgnn pairs For material n one can enter the isotope description from which to get photonuclear data for all elements listed on the Mn card A Zero entry for PNZA will turn off photonuclear interactions for that particular element 6 1 7 Energy and Thermal Treatment Cards PHYS TMP THTME MTm A PHYS card may be specified for any particle type and we recommend that they be included for all particles on the MODE card Charged and Neutral Particles except Photons The first entry on the PHYS card is the maximum energy for the specified particle Note that the default EMAX can be quite low and failing to reset this for high energy problems will result in code termination because particle energies exceed EMAX The code will note the largest EMAX from all the specified PHYS cards in the problem If a tracked particle does not have a PHYS card its EMAX will be set to this largest value If no PHYS cards are included in
142. user must maintain the proper correspondence among the arrays see Section 22 below e Any additional input required for the particular option For basic option types 1 2 or 11 this may be the specification of surface segmenting For basic option types 9 10 or 12 it is the collimating window definition Also for basic option types 1 9 11 or 12 an arbitrary vector for angular binning may be input The order of the input records as they appear in the INT file is illustrated in Table D4 3 Edit Option IOPT 1 or 101 Surface Current Option 1 tallies the particle current across the NPARM designated surfaces it is analo gous to the MCNP F1 tally If IOPT is preceded by a minus sign the weight binned is multiplied by the particle energy The number of energy bins is given by NERG The num ber of particle types for which surface crossing data is to be tallied is given by NTYPE and 142 MCNPX User s Manual MCNPX User s Manual i Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium must be gt 0 Current will be tallied on NPARM surfaces a total over surfaces is not per formed Any of the above particle types may be specified Binning into NFPRM cosine bins is defined by the value of KOPT For KOPT 0 or 5 the cosine is taken with respect to the normal to the surface at the crossing point For KOPT 1 or 6 the cosine is taken with respect to the x axis For KOPT 2 or 7 the cosine is taken with r
143. x y coordinates of the points in the current plot PRINTPTS PRINTPTS is not available for co plots or contour or 3D plots File Manipulation Commands Read dump n from RUNTPE file aa If the parameter n RUNTPE aan is omitted the last dump in the file is read DUMP n Read dump n of the current RUNTPE file MCNPX User s Manual 51 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 3 MPLOT amp MCPLOT Commands Command Description Write the tally data in the current RUNTPE dump to MCTAL WMCTAL aa F file aa Read MCTAL file aa Parameter setting Commands Parameters entered for one curve or plot remain in effect for subsequent curves and plots until they are either reset to their default values with the RESET com mand or are overridden either by the same command with new values by a con flicting command or by the FREE command that resets many parameters There are two exceptions FACTOR and LABEL are effective for the current curve only An example of a conflicting command is BAR which turns off HIST PLINEAR and SPLINE Define tally n as the current tally n is the n on the Fn card in the INP file of the problem rep resented by the current RUNTPE or MCTAL file TALLY n The default is the first tally in the problem which is the low est numbered neutron tally or if none then the lowest numbered photon tally or if none then the lowest nu
144. 0 0000 any termination 5 4273E 00 wc2 2 5000E 01 128 MCNPX User s Manual MCNPX User s Manual i Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium Net neutron production in this case is 18 364 n p or 0 5 above the base case value The difference is primarily due to the neutron multiplicity between 20 and 150 MeV in the new 150 MeV evaluations as compared to the multiplicity given by the LAHET physics models in this energy range Since the data evaluations are considered more accurate than the LAHET physics models the base case value of 18 263 should be considered the better estimate Note the difference in net production by nuclear interactions 15 617 n p for the base case versus 17 897 n p for case 1 and by n xn reactions 3 785 n p for the base case versus 0 516 n p for case 1 for the two cases The difference of 2 280 n p between the two cases for net production by nuclear interactions is the value calculated by the LAHET modules within mcnpx for net neutron production by neutrons in the energy range 20 to 150 MeV Similarly the difference of 3 269 n p in the values for net n xn production is the value pre dicted by the new 150 MeV Pb data libraries for net neutron production by neutrons with energies between 20 and 150 MeV Case 2 In the second variation we transport not only nucleons denoted by the symbols n and h on the mode card and charged pions but also light ions deuterons tr
145. 0 1969 BLU50 O Blunck and S Leisegang Zum Energieverlust schneller Elektronen in dunnen Schichten Z Physik 128 1950 500 BLU51 O Blunck and R Westphal Zum Energieverlust energiereicher Elektronen in dunnen Schichten Z Physik 130 1951 641 BRE81 D J Brenner R E Prael J F Dicello and M Zaider Improved Calculations of Energy Deposition from Fast Neutrons in Proceedings Fourth Symposium on Neutron Dosimetry EUR 7448 Munich Neuherberg 1981 BRE89 D J Brenner and R E Prael Calculated Differential Secondary particle Production Cross Sections after Nonelastic Neutron Interactions with Carbon and Oxygen between 10 and 60 MeV Atomic and Nuclear Data Tables 41 71 130 1989 BRIOOx J F Briesmeister ed MCNP A General Monte Carlo N Particle Transport Code Los Alamos National Laboratory Report LA 13709 M Version 4C March 2000 CHA98 M B Chadwick et al Reference Input Parameter Library handbook for Calculations of Nuclear reaction Data IAEA TECDOC Draft IAEA Vienna March 1998 CAR98 L L Carter R C Little J S Hendricks New Probability Table Treatment in MCNP for Unresolved Resonances 1998 Radiation Protection and Shielding Division Topical Conference on Technologies for the New Century Sheraton Music City Nashville TN vol Il p 341 April 19 23 1998 CHA99a M B Chadwick P G Young S Chiba S C Frankle Hale H G Hughes A J Koning R
146. 0 and to provide feedback to the developers This process is invaluable and we express our deepest appreciation to the participants in the beta test program Applications for the code among the beta test team are quite broad and constantly devel oping Examples include e Design of accelerator spallation targets particularly for neutron scattering facilities e Investigations for accelerator isotope production and destruction programs including the transmutation of nuclear waste e Research into accelerator driven energy sources e Medical physics especially proton and neutron therapy e Investigations of cosmic ray radiation backgrounds and shielding for high altitude air craft and spacecraft MCNPX User s Manual 1 MCNPX User s Manual E Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium e Accelerator based imaging technology such as neutron and proton radiography e Design of shielding in accelerator facilities e Activation of accelerator components and surrounding groundwater and air e Investigation of fully coupled neutron charged particle transport for lower energy applications e High energy dosimetry and neutron detection e Design of neutrino experiments e Comparison of physics based and table based data e Charged particle tracking in plasmas e Charged particle propulsion concepts for spaceflight Single event upset in semiconductors from cosmic rays in spacecraft or from the
147. 0 MeV libraries have been produced for use with MCNPxX The neutron library is called LA150n The proton and photonuclear libraries are called la150h and la150u respectively The LA150N library is the same as DLC200 with the addition of 150 MeV evaluations above the DLC200 energy limits Once the proton and photonu clear components are added the entire library will be reissued under the name DLC200X 4 Anumber of users are requesting secondary particle and recoil nuclei information for the lower energy portions of the libraries typically below 20 MeV Note that some information is available in the lower energy tables per table 4 4 in this manual but it is far from complete A proper fix to the problem will involve full re evaluations of the lower energy libraries which is a time consuming and often difficult task Nonethe less progress is being made and the user should look for improved library releases in the future The LANL group that formats libraries for MCNP MCNP xX is currently providing 64 bit type 2 binary files and MCNPX 2 4 0 will only accept these Therefore the user will find that older versions of 32 bit binary libraries won t work with the 2 4 0 The program MAKXS is provided with the MCNPxX distribution to do the reformatting and details can be found in Appendix C of the MCNP manual An alternative is to use type 1 formatted sequential access libraries The XSDIR file tells the code all the information
148. 0 ee eee eee 77 5 4 4 TOTNU Total Fission 0 00 ce eee 77 5 4 5 NONU Fission Turnoff 000 c eee ees 77 5 4 66 AWTAB Atomic Weight 002 cece eee eee 78 5 4 7 XSn Cross Section File nannan nannan 78 5 4 8 VOID Material Void 0 00 cece ees 78 5 4 9 PIKMT Photon Production Bias 2 0000e eee eeee 79 5 4 10 MGOPT Multigroup Adjoint Transport Option 80 5 4 11 DRXS Discrete Reaction Cross Section 81 5 5 PHYSICS 226 054 0205 e elected eskee ek ANE cie gaia eee oe ee 82 5 5 1 MODE Problem Type 00c cece eee e eee eee 82 5 5 2 PHYS Energy Physics Cutoff 000 e cece e eee eee 82 5 5 201 NEUIONS 402 eee else aa tet hed eee ee dae eee oe es 82 55 22 PNOLONSEs hs act Minh E A 2 tact trees Std Oeste hl bt toes 83 5 5 2 S Electron 2 2025 a Sd hese ee ene devel Ba ee eee oE 84 5 5 2 4Proltons e sxx0 vos beens Pete earde eed eae t eho sews Bete 85 5 5 2 5 Other Particles 234 6 eh ele ajo bbc Sa haere Viel eed Se 85 5 5 3 TMP Free Gas Thermal Temperature 00000ee eens 86 5 5 4 THTME Thermal Times 0000c cece eee ees 86 5 5 5 COINC 3He Detector Coincidence 0 00 eens 87 5 5 6 Problem Cutoff Cards 0 00 c eee eee eee eee 87 55 6 1 CUT Gutier renega te ined Cae EE uea ete oa RENA 87 5 5 6 2 ELPT Cell by cell Energy Cutoff 0 00 88 5 5 6 3 NPS Hist
149. 000eeeeeeee 89 7 1 Secondary Particle biasing 0 00 cece eens 89 8 New and Improved Tallies and Data Analysis 91 8 1 The Mesh Tally 2 0 0 0 cee eee eee 91 8 1 1 Setting up the Mesh in the INP File 0 0 0 e eae eae 92 8 1 2 Processing the Mesh Tally Results 0 000 e eee 100 8 2 The Radiography Tally 0 eects 102 8 2 1 Pinhole Image Projection 0 0 cee ee eee 102 8 2 2 Transmitted Image Projection 0 ee eee 104 8 2 3 Additional Radiography Input Cards 2 0 000 ee 107 8 2 4 Reading the Radiography Tally Output 00005 107 8 3 Energy Deposition 0 0 0c eee 108 viii MCNPX User s Manual MCNPX User s Manual x ae 8 4 Dose Conversion Coefficients 0 00 cece 113 8 5 ISTP and FVAPESXe 0shaeosackead f xen Mee bd a felt oe Ped Rene oe 116 9 RGIGFENCES 6 ie ean eke cdinae Stee hie eee COS ae Se hee 117 Appendix A Examples jc 0 sci ceased Saw eee ge Kew eae tw d 125 Appendix B HTAPESX for Use with MCNPX 00000es 135 AOSTA tee ia te Renee gare Savi Rie Soteen ke Pee war eee LY 135 4 The HTAPESX Cod yc it et he ete eh Behe tt oe aes 135 2 Input forHTAPE3X o n2hate soa hs Sareterde dada ts Jed ie A a Toke ee 135 3 Edit Option IOPT 1 or 101 Surface Current 0 0 0 0 eee eee 142 4 Edit Option IOPT 2 or 102 Surface Flux 0 00 c eee ee 14
150. 002 LA UR 02 2607 Accelerator Production of Tritium RAN85 J Ranft and S Ritter Z Phys C27 1985 412 569 RIL75 M E Riley C J MacCallum and F Biggs Theoretical Electron Atom Elastic Scattering Cross Sections Selected Elements 1 keV to 256 KeV Atom Data and Nucl Data Tables 15 1975 443 RUT11 E Rutherford The Scattering of a and b Particles by Matter and the Structure of the Atom Philos Mag 21 1911 669 SCH82 P Schwandt et al Phys Rev C 26 55 1982 SEL88 S M Seltzer An Overview of ETRAN Monte Carlo Methods in Monte Carlo Transport of Electrons and Photons edited by T M Jenkins W R Nelson and A Rindi Plenum Press New York 1988 p 153 SEL91 S M Seltzer Electron Photon Monte Carlo Calculations The ETRAN Code Appl Radiat Isot Vol 42 No 10 1991 pp 917 941 SNO96 E C Snow Radiography Image Detector Patch for MCNP private communication SNO98 E C Snow Mesh Tallies and Radiography Images for MCNPX Proceedings of the Fourth Workshop on Simulating Accelerator Radiation Environments SARE4 Tony A Gabriel ed 1998 113 STE71 R M Sternheimer and R F Peierls Phys Rev B3 no 11 June 1 1971 3681 TRI97a R K Tripathi F A Cucinotta J W Wilson Universal Parameterization of Absorption Cross Sections NASA Technical Paper 3621 January 1997 TRI97b R K Tripathi J W Wilson and f A Cucinotta New Para
151. 02 LA CP 02 408 or Cn 4 er k Table 5 62 Cosine Card Variable Description n tally number upper cosine limit of the ith angular bin for surface current C tally n Cy gt 1 Ck 1 bi upper angular limit expressed in degrees 1 lt 180 0 Default If the Cn card is absent there will be one bin over all angles unless this default has been changed by a CO card Use Tally type 1 and 2 Required if CMn card is used Consider FQn card Example C1i 866 5 0 5 866 1 or C1 150 120 90 60 30 0 This will tally currents within the angular limits 1 180 to 150 2 150 to 120 3 120 to 90 4 90 to 60 5 60 to 30 and 6 30 to 0 with respect to the positive normal No total will be provided 5 7 6 FQn Print Hierarchy Form FQn a dp ag Table 5 63 Variable Description n tally number F cell surface or detector D direct or flagged U user S segment M multiplier C cosine E energy T time aj Default Order as given above right to left MCNPX User s Manual 123 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Use Highly recommended Prints tallies in more easily readable blocks in the output file without affecting answers Example FQ4 ESM The output file printout will be tables with multiplier bins across the top segments listed vertically and these segment multipli
152. 1 non blank 7 The S must not appear anywhere else in the input file MCNPX User s Manual 37 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 8 The K are optional integers If any are non blank all must be non blank If the S are cell parameter card names the K if present must be valid cell names The same is true with surface parameters 10 If the K are present the D must not be multiple special syntax items such as OR 4 1 10 Particle Designators Several of the input cards require a particle designator to distinguish between input data for tracked particles Refer to the pertinent card information for instructions The particle designator consists of a colon followed by the particle symbol or IPT number s immediately after the name of the card At least one blank must follow the particle designator For example imp n signifies neutron importances follow enter photon importances on an IMP P card To specify the same value for more than one kind of particle a single card can be used instead of several Example IMP E P N 11 0 With a tally card the particle designator follows the card name including tally number For example F5 N indicates a neutron point detector energy tally In the heating tally case both particle designators may appear The syntax F6 N P indicates the combined heating tally for both neutrons and photons Table 4 1 MCNPX Particles
153. 15E 01 6 4394E 01 0000E 05 7 4505E 03 0 9011E 01 3 4571E 02 cutoffs tco 1 0000E 34 eco 0 0000E 00 wcl 5 0000E 01 wc2 2 5000E 01 130 MCNPX User s Manual MCNPxX User s Manual i Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium Note the net neutron production calculated with the ISABEL INC model is 17 569 which is 3 8 below the value predicted by the Bertini INC model This is consistent with other studies that reveal slightly lower neutron production resulting from ISABEL as compared to Bertini Case 4 In the next variation from the base case we use the new evaluated proton libraries for transporting protons below 150 MeV replacing the Bertini model used at all proton ener gies in the base case We invoke transport of protons with energies less than 150 MeV by including a phys h card to specify the transition energy between LAHET physics and data evaluations for proton transport Base Case phys h 1000 j 0 Case 4 phys h 1000 j 150 The neutron summary table for this case is shown below sample problem spallation target Case 4 neutron creation tracks weight energy neutron loss tracks weight energy per source particle per source particle source 0 0 QO escape 365199 1 8244E 01 2 1884E 02 nucl interaction 308299 1 5415E 01 3 2024E 02 energy cutoff 0 0 0 particle decay 0 0 0 time cutoff 0 0 0 weight window 0 0 0 weight window 0 0 0 cell importance 0 0 0 cell imp
154. 2 Chapter 3 covers code installation and general notes on software management 2 MCNPX User s Manual MCNPxX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium Chapter 4 gives an overview of the new high energy physics options in addition to discuss ing the extended 150 MeV nuclear data libraries and other tabular data available for MCNPX Chapter 5 describes the extended particle set with specific notes on particle tracking multiple scattering and energy straggling routines Chapter 6 contains information on modifications and enhancements to existing MCNP4B INP input cards while Chapter 7 covers new variance reduction techniques Chapter 8 describes new tallying capabilities Information supplemental to the text is included in the Appendices This manual is not intended to replace the existing user guides to MCNP4B BRI97 the LAHET Code System PRA89Q nor any other manual covering incorporated physics mod ules The user should become familiar with these works which are extensively referenced Work is now underway to fully upgrade MCNPX to MCNP4C and to explore the possibil ities inherent in conversion to Fortran 90 Classes in MCNPX are also held on a regular basis http mcnpxworkshops com MCNPX User s Manual 3 MCNPX User s Manual Version 2 3 0 April 2002 F LA UR 02 2607 Accelerator Production of Tritium 4 MCNPX User s Manual MCNPX User s Manual Ver
155. 2 pe Dy away from point 1 1 1 6 6V9 6 V9 p 6 away from origin 8 5V9 5 v9 p 5 toward origin 9 SZ S Z p L d along Z axis 10 4X 4 X p 4 along X axis 5 8 9 VECT Vector Input Form VECT Vm XmYmZm VN Xn Yn Zn Table 5 95 Vector Input Card Variable Description m n any numbers to uniquely identify vectors Vm Vn XmYmZm coordinate triplets to define vector Vm Default None Use Optional The entries on the VECT card are quadruplets which define any number of vectors for either the exponential transform or user patches See the EXT card Section 5 8 8 for a usage example 5 8 10 FCL Forced Collision Form FCL n x4 X2 Xj X 160 MCNPX User s Manual MCNPxX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 96 Forced Collision CArd Variable Description n particle designator forced collision control for cell i 1 lt x lt 1 gt 0 applies to particles surviving weight cutoff weight win dow games in the cell lt 0 applies only to particles entering the cell 0 no forced collision in cell Xi I number of cells in the problem Default x 0 no forced collisions Use Optional Exercise caution 5 8 11 DDn Detector Diagnostics Form DDn k m kp mp Table 5 97 Detector Diagnostics Card Variable Description tally number for specific detector tally n 1 for neutro
156. 3 5 Edit Option IOPT 3 or 103 Particle Production Spectra 144 6 Edit Option IOPT 4 or 104 Track Length Estimate for Neutron Flux 144 7 Edit Option IOPT 5 or 105 Residual Masses and Average Excitation 144 8 Edit Option IOPT 6 or 106 Energy Deposition 20000 0 eee 145 9 Edit Option IOPT 7 Mass and Energy Balance 00200000 eu 145 10 Edit Option IOPT 8 or 108 Detailed Residual Mass Edit 145 11 Edit Option IOPT 9 or 109 Surface Current with Collimating Window 146 12 Edit Option IOPT 10 or 110 Surface Flux with Collimating Window 146 13 Edit Option IOPT 11 or 111 Pulse Shape of Surface Current 146 14 Edit Option IOPT 12 or 112 Pulse Shape of Surface Current with Window 147 15 Edit Option IOPT 13 Global Emission Spectrum 20 000000 147 16 Edit Option IOPT 14 or 114 Gas Production 0 00 e eee 148 17 Edit Option IOPT 15 or 115 Isotopic Collision Rate 2 148 18 Edit Option IOPT 16 or 116 Recoil Energy and Damage Energy Spectra 149 19 The Resource Option 0 2 0 ccc tee 150 20 The Merge Optom weres iena ee ead atte ee Lew bs eee ee 150 21 The Time Convolution Option 0000 ce eee 150 22 The Response Function Option 000 ccc eee ee 151 23 Executing HTAPE3X 2 0 tees 151 References sc
157. 3078 1 30678 1 27151 1 22449 1 16522 1 09322 1 00800 0 90906 0 79592 0 66808 0 52506 0 36637 0 19151 0 00000 For the base case the neutron problem summary follows MCNPX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 sample problem spallation target base case neutron creation tracks weight energy neutron loss tracks per source particle source 0 0 QO escape 365317 nucl interaction 316017 1 5801E 01 3 2136E 02 energy cutoff particle decay 0 Os OF time cutoff 0 weight window 0 0 0 weight window 0 cell importance 0 0 0 cell importance 0 weight cutoff 0 Oe 0 weight cutoff 0 energy importance 0 QO 0 energy importance 0 dxtran 0 0 0 dxtran 0 forced collisions 0 0 0 forced collisions 0 exp transform 0 0 QO exp transform 0 upscattering 0 0 0 downscattering 0 tabular sampling 0 QO QO capture 0 n xn 78320 3 9123E 00 1 8804E 01 loss to n xn 25352 fission 0 0 0 loss to fission 0 photonuclear 0 0 0 nucl interaction 3668 tabular boundary 0 QO 0 tabular boundary 0 gamma xn 0 0 0 particle decay 0 adjoint splitting 0 0 s 394337 1 9713E 01 3 4016E 02 total 394337 number of neutrons banked 368985 average time of shakes MCNPX User s Manual weight energy per source particle 8249E 01 4266E 02 2660E 00 1 0 0 0 0 0 0 0 0 0 0 1 A 0 1 8340E 01 0 0 1 9713E 01 cutoffs 1995E 02 2 0 0 0 0 0 0 0 0 0 9 8498E 00 7 6455E 02 4 8878E 01 0 6 140
158. 31 Neutron Physics Options Keyword Description n particle designator EMAX Upper limit for neutron or proton energy MeV Analog energy limit MeV Implicit capture for E gt Ean EAN implicit capture for E lt Ean Unresolved resonance range probability table treatment when IUNR data tables exist 0 on 1 off Delayed neutron production when data tables exist 1 analog DNB 0 off gt 0 produce up to n delayed neutrons per fission n gt 0 Note in KCODE n lt 0 biasing disallowed 82 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 31 Neutron Physics Options Keyword Description Table based physics cutoff For E gt Tabl use model physics TABL E lt Tabl use physics from data tables WARNING If Tabl gt emax of a data table the cross section values at E emax will be used in the energy range emax Tabl Fission multiplicity control Sample the number of fission neu trons lt nu gt from a gaussian of width Fism FISM 0 sample lt nu gt as the integer value either above or below lt nubar gt 1 sample lt nu gt with the width appropriate for each nuclide Light ion recoil control Number of light ions h d t s a to be RECL created at each neutron elastic scatter off H D T 3He He CUT n 2J 0 is usually needed for n h d t s a 0 lt RECL lt 1 Use Encouraged D
159. 4 September 14 16 1998 Knoxville Tn ed by Tony A Gabriel ORNL pp 171 181 PRA98b R E Prael Upgrading Physics Packages for LAHET MCNPX Proceedings of the American Nuclear Society Topical Meeting on Nuclear Applications of Accelerator Technology Gatlinburg TN Sept 20 23 1998 PRA98c R E Prael and W B Wilson Nuclear Structure Libraries for LAHET and MCNPX Proceedings of the Fourth workshop on simulating Accelerator Radiation Environments SARE4 September 14 16 1998 Knoxville Tn ed by tony A Gabriel ORNL pp 183 PRA99 R E Prael Primary Beam Transport methods in LAHET Transactions of the June ANS Meeting Boston June 6 10 1999 PRAO00a R E Prael Proposed Modification to the Charged Hadron Tracking Algorithm in MCNPX Los Alamos Research Note X 5 RN U August 23 2000 LA UR 00 4027 PRAOOb R E Prael A New Nuclear Structure Library for MCNPX and LAHET3 Proceedings of the Fourth International topical Meeting on Nuclear Applications of Accelerator Technology Nov 12 15 2000 Washington DC pp 350 352 RAD77 Radiation Shielding Information Center HETC Monte Carlo High Energy Nucleon Meson Transport Code Report CCC 178 Oak Ridge National Laboratory August 1977 RAN85 J Ranft and S Ritter Z Phys C27 1985 412 569 MCNPX User s Manual 187 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 RIL75 M E Riley C J MacCallum and F Biggs
160. 4353 2 06 x10 18 cascade Z 2470 3 1 0 9 8 x 106 19 lambda Ap r 5641 1 0 1 07 x 104 Mesons MCNPX User s Manual 39 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 4 1 MCNPX Particles Low Kinetic Mean Lifetime IPT Name of Particle Symbol Mass MeV Energy Cutoff aaa Mev production 20 pion z 139 56995 0 14875 2 6033 x 10 8 20 pion 7 139 56995 0 14875 2 6033 x 10 8 21 neutral pion x Z 134 9764 0 0 8 4 x 10717 22 kaon K k 493 677 0 52614 1 2386 x 108 22 kaon K k 493 677 0 52614 1 2386 x 10 8 23 Kg short 497 672 0 000001 0 8927 x 19 19 24 Kglong 497 672 0 000001 517x108 25 pt g 1869 3 1 9923 1 05 x 104 26 p 1864 5 1 0 415x100 27 D f 1968 5 2 098 4 67x10 28 Bt j 5278 7 5 626 1 54x 107 29 Bo b 5279 0 1 0 1 5x104 30 B q 5375 1 0 1 34 x 104 Light lons 31 deuteron d 1875 627 2 0 huge 32 triton t 2808 951 3 0 12 3 years 33 Helium 3 s 2808 421 3 0 huge 34 Helium 4 a a 3727 418 4 0 huge Particle tracking between interactions involves several physics considerations which are described below Atomic electron interactions will cause a charged particle to lose energy along its track length ionization Certain modifications to this energy loss are determined by energy straggling theory Multiple scattering of charged
161. 5 7 17 FTn Special Treatments for Tallies 0055 131 5 7 18 Subroutine TALLYX User supplied Subroutine 135 5 7 19 TFn Tally Fluctuation 00 cece 135 5 7 20 TIRn The Radiography Tally 200 e ee eee eee eee 136 5 7 20 1 Pinhole Image Projection 0 c eee eee 136 5 7 20 2 Transmitted Image Projection 00 eee ee eae 138 5 7 20 3 Additional Radiography Input Cards 139 5 7 20 4 Reading the Radiography Tally Output 140 5 7 21 PERTn Perturbation 000 c eee eee 140 MCNPX User s Manual ix MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 5 7 22 TMESH The Mesh Tally 0 00e cece eee eee eee 143 5 7 22 1 Setting up the Mesh in the INP File 144 5 7 22 2 Track Averaged Mesh Tally Type 1 2 000 145 5 7 22 3 Source Mesh Tally Type 2 00 000 eee eee eee 147 5 7 22 4 Energy Deposition Mesh Tally Type 3 0 148 5 7 22 5 DXTRAN Mesh Tally Type 4 00 0c e eee eee 149 5 7 22 6 Dose Conversion Coefficients 0200 eee eee 150 5 7 22 7 Processing the Mesh Tally Results 004 152 5 8 Variance Reduction 00 ccc eee eee eee eee eee eens 153 5 8 1IMP Cell Importance 00 0c eee eee 153 5 8 2 WWG Weight Window Generator 00000eee eens 154 5 8 3 WWGE Weight Window Generation En
162. 5 I Range of one or more lattice ele ments Use the same format as on the FILL card I Ip I3 l4 Is Ig Indicating lattice element Jj Zo I3 I4 Is Ig etc See LAT and FILL cards for indices explanation Example F4 N 5 lt 4 lt 2 100 This example could specify an F4 tally in cell 5 when it is in cell 4 when cell 4 is in cell 2 which is a lattice and only in lattice element 1 0 0 While any cell lattice filled or simple can be entered as a tally cell e g S4 through S5 only cells filled with a universe can be used in higher levels e g C through Cs Important the arrows separate different universe levels Cell 5 in U 2 is inside cell 4 in U 1 For C4 lt C3 C4 must NOT be in the same universe as Co 5 7 1 2 1 Multiple bin format In addition to multiple levels multiple entries can be used in each level of the tally chain resulting in multiple output bins Within the parentheses required around the tally bin chain other sets of parentheses can be used to indicate the union of cells as in a simple tally description resulting in fewer output tally bins S4 S5 lt Cy C2 U Lal lt C3 C4 C5 This example results in one output tally bin and will be the union of the tally in Sz plus S5 that fill C or Cp elements J Z3 and when C or Co fills cells C3 Cz or C5 Removing the first and third inner parentheses S4 S5 lt C1 C 2 U4 I lt C3 C4 Co results in the crea
163. 5 MeV potential well for pions 5 The same as l 2 but using a 25 MeV potential well for pions 6 The same as l 2 but using a 25 MeV potential well for pions Note Not all the options for the ISABEL INC model have been thoroughly debugged JCOUL 1 Use Coulomb barrier on incident charged particle interactions default 0 No Coulomb barrier for incident charged particles NEXITE 1 Subtract nuclear recoil energy to obtain nuclear excitation energy default 2 Do not subtract nuclear recoil energy NPIDK 1 Force 1 to terminate by decay at the pion cutoff energy 0 Force av to interact by nuclear capture INC when cutoff is reached default Note The capture probability for any isotope in a material is proportional to the product of the number fraction and the charge of the isotope However capture on 1H leads to decay rather than interaction 78 MCNPX User s Manual Accelerator Production of Tritium MCNPxX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 Table 6 3 LCA Keyword Descriptions Continued Keyword Description NOACT Note The use of the NOACT option other than the default is intended as a diagnostic tool allowing other processes to be more easily observed PRA99 2 Attenuation mode transport primary source particles without nonelastic reactions 1 Do not turn off nonelastic reactions default 0 Turn off all nonelastic reactions 1 Compute nuc
164. 6 lt 78 V V5 V Vi Vo v Vy vi vy v 123 lt 456 lt 78 V V Vz 1 2 3 4 5 6 123 lt 456 lt 78 Viog Ving Viog V123 V123 V123 MCNPX User s Manual 119 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 V volume of cell i for bin j and Viss volume of the union of cells 1 2 and 3 for bin j If cell i is repeated the same number of times in all six bins generated by the first line above then all six SD values for this input bin will be the same V4 V V However if cell 1 is repeated a different number of times in each bin then different SD values should be entered The volume is multiplied by the number of times it is repeated In these cases the total cell 1 volume for each generated bin will not be calculated The bin generation order is explained previously in the Fn card section For the first line above the bin order is 1 lt 4 lt 7 1 lt 5 lt 7 1 lt 6 lt 7 1 lt 4 lt 8 1 lt 5 lt 8 and 1 lt 6 lt 8 The second line above generated 18 tally bins and 18 SD values are required in the proper sequence This option requires the knowledge of both the number and sequence of bins generated by each input tally bin 5 7 1 3 Detector Tallies tally type 5 Form for point detectors Fnip XYZ R Table 5 56 Point Detector Variable Description n tally number pl N for neutrons or P for photons XYZ location of the detector point radius of the sphere of exclusi
165. 8 18 Edit Option IOPT 16 or 116 Recoil Energy and Damage Energy Spectra Option 16 provides an edit of the spectra of total recoil energy elastic recoil energy total damage energy and elastic damage energy Also estimated are the mean weight of recoiling fragments per history mean weight of recoil or damage energy per history and the mean energy per fragment the ratio of the previous two estimates NERG specifies the number of energy bins for the spectra a minus sign on NERG will have the tabulation normed per MeV recommended to produce a true spectrum Input variables NTIM NTYP NFPRM IXOUT IRS IMERGE ITCONV and IRSP are unused KOPT 0 indicates tally by cell KOPT 1 indicates tally by material NPARM is the number of cells or materials to be read in for the tally If a minus sign flag is used with IOPT IOPT 16 the weights tallied for the spectra will be multiplied by corresponding recoil or damage energy At any collision the damage energy Egis obtained from the recoil energy E of nucleus A Z by the relation of Linhard 4 Ej E L E using the formulation of Robinson 5 Table 8 2 p 01337452202 A Ay 2 i i 3 3 4 9 g2l8 4 72 3 js fin su NSD OR AT A A Z 2 22 22 glei ei 0 40244e7 3 4008 n fi UES es ee re zi 2 1 kig i 1 1 where the summation is over the components of the material with atom fractions f 220 MCNPX User s Manual MCNPX
166. 8 2 9 3 0 3 1 3 2 3 3 3 4 3 5 sp1 0 00000 0 09992 0 19935 0 29780 0 39478 0 48980 0 58237 0 67200 0 75820 0 84049 0 91837 0 99135 1 05894 1 12065 1 17600 1 22449 1 26563 1 29894 1 32392 1 34008 1 34694 1 34400 1 33078 1 30678 1 27151 1 22449 1 16522 1 09322 1 00800 0 90906 0 79592 0 66808 0 52506 0 36637 0 19151 0 00000 For the base case the neutron problem summary follows sample problem spallation target base case neutron creation tracks weight energy neutron loss tracks weight energy per source particle per source particle source o 0 0 re 365317 1 82498 01 2 1995E 02 nucl interaction 316017 1 5801F 01 3 2136F 02 energy citait 0 0 0 particle decay 0 o 0 time cutoff o 0 o weight window o By o weight window o 0 0 cell importance o 0 0 cell importance oo 0 0 weight cutoff 0 o 0 weight cutoff o 0 0 194 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 energy importance 0 QO 0 energy importance 0 0 0 dxtran 0 0s 0 dxtran 0 Os 0 forced collisions 0 0 0 forced collisions 0 0 0 exp transform 0 QO QO exp transform 0 QO 0 upscattering 0 0 0 downscattering 0 O s 9 8498E 00 tabular sampling 0 QO QO capture 0 1 4266E 02 7 6455E 02 n xn 78320 3 9123E 00 1 8804E 01 loss to n xn 25352 1 2660E 00 4 8878E 01 fission 0 0 0 loss to fission 0 0 0 photonuclear 0 0 0 nucl interaction 3668 1 8340E 01 6 1409E 01 tabular boundary 0 QO QO tabular boundary 0 QO QO
167. 8389 0 11261 2 19703 x 10 pipe symbol MCNPX User s Manual 65 MCNPX User s Manual E Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium Table 5 1 Particles in MCNPX EEE Mean Lifetime Low Kinetic seconds IPT Name of Particle Symbol Mass MeV Energy Cutoff MeV decayed on production 4 anti muor u 105 658389 0 11261 2 19703 x 10 5 tau T z 1777 1 1 894 2 92 x 10 6 electron neutrino Ve u 0 0 0 0 huge 6 anti electron neutrino u 0 0 0 0 huge 7 muon neutrino Vm v 0 0 0 0 huge 8 tau neutrino v w 0 0 0 0 huge Baryons 9 proton p h 938 27231 1 0 huge 9 anti proton p h 938 27231 1 0 huge lower case L 11 sigma 2 1189 37 1 2676 7 99 x108 12 sigma 1197 436 1 2676 1 479 x 1072 13 cascade x 1314 9 1 0 29x102 14 cascade E y 1321 32 1 4082 1 639 x102 15 omega Q o 1672 45 1 7825 8 22 x 103 16 lambda A Cc 2285 0 2 4353 2 06 x 10 i 17 cascade 67 2465 1 2 6273 3 5 x 10 i 18 cascade 2470 3 1 0 9 8 x 106 19 lambda A r 5641 1 0 1 07 x 104 Mesons 20 pion 0 139 56995 0 14875 2 6033 x 108 66 MCNPX User s Manual MCNPxX User s Manual Version 2 3 0 April 2002 xz LA UR 02 2607 Accelerator Production of Tritium
168. 96E 01 3 3488E 02 banked 316635 source particle 1 7300E 01 per source particle 2 3611E 01 472212 MCNPX User s Manual neutron loss tracks weight energy per source particle escape 313015 1 5635E 01 2 1374E 02 energy cutoff 0 0 QO time cutoff 0 O 0 weight window 0 0 0 cell importance 0 0 0 weight cutoff 0 EA 0 energy importance 0 QO QO dxtran 0 0 0 forced collisions 0 0 D exp transform 0 0 QO downscattering 0 0 7 3438E 00 capture 0 1 3051E 02 8 5469E 02 loss to n xn 29374 1 4667E 00 5 7124E 01 loss to fission 0 0 0 2 nucl interaction 3619 1 8095E 01 5 6576E 01 tabular boundary 1 5 0000E 05 7 4680E 03 particle decay 0 0 0 total 346009 1 7296E 01 3 3488E 02 average time of shakes cutoffs escape 5 7337E 00 tco 1 0000E 34 capture 4 7022E 01 eco 0 0000E 00 capture or escape 5 7293E 00 wcl 5 0000E 01 201 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 net multiplication 0 0000E 00 0000 any termination 5 1842E 00 wc2 2 5000E 01 Net neutron production for this case is 15 648 n p which is 14 3 than the base case value Note also that CEM took twice as long to run as the base case Both of these factors are well known and CEM improvements is a very active project in the MCNPX program The increase in time is understood and will be corrected in future versions through algorithm optimization The lower n p values are also being extensively benchmarked and improvements involving the tra
169. 99 1 5415E 01 3 2024E 02 energy cutoff particle decay 0 0 0 time cutoff weight window 0 s 0 weight window cell importance 0 0 0 cell importance weight cutoff 0 0 0 weight cutoff energy importance 0 QO 0 energy importance dxtran 0 0 0 dxtran forced collisions 0 0 0 forced collisions exp transform 0 QO 0 exp transform upscattering 0 0 0 downscattering tabular sampling 7166 3 5830E 01 1 8289E 00 capture n xn 78791 3 9358E 00 1 9090E 01 loss to n xn fission 0 0 0 loss to fission photonuclear 0 0 0 nucl interaction tabular boundary 0 QO 0 tabular boundary gamma xn 0 0 Gs particle decay adjoint splitting 0 s 0 gt total 394256 1 9709E 01 3 4116E 02 total number of neutrons banked 368932 average time of shake neutron tracks per source particle 1 9713E 01 escape 5 neutron collisions per source particle 2 7817E 01 capture 4 total neutron collisions 556332 capture or escape 5 net multiplication 0 0000E 00 0000 any termination Dis tracks 365199 25324 3733 394256 s 7563E 00 6071E 01 7522E 00 3292E 00 weight energy per source particle 1 8244E 01 2 1884E 02 QO QO QO 0 QO 0 0 0 QO QO QO 0 QO QO QO 0 QO QO QO 9 8423E 00 1 4179E 02 7 6277E 02 1 2646E 00 4 9542E 01 QO QO 1 8665E 01 6 2865E 01 QO 0 0 QO 1 9709E 01 3 4116E 02 cutoffs tco 1 0000E 34 eco 0 0000E 00 wcl 5 0000E 01 wc2 2 5000E 01 Net neutron production f
170. 9E 01 0 0 3 4016E 02 127 MCNPxX User s Manual E Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium neutron tracks per source particle 1 9717E 01 escape 5 7458E 00 tco 1 0000E 34 neutron collisions per source particle 2 7874E 01 capture 4 6648E 01 eco 0 0000E 00 total neutron collisions 557485 capture or escape 5 7417E 00 wcl 5 0000E 01 net multiplication 0 0000E 00 0000 any termination 5 3201E 00 wc2 2 5000E 01 The two methods for calculating total neutron production give the following results net nuclear interactions net n xn 15 801 0 1834 3 9123 1 2660 18 263 n p escapes captures 18 249 0 014226 18 263 n p Both methods give the same answer Since escapes captures is easier to calculate this is the method typically used A reasonable upper limit on the relative uncertainty of n p is 20 000 0 7 Case 1 The first variation considered is the impact of the extension of the evaluated neutron cross sections to 150 MeV on total neutron production To evaluate this impact we set the tran sition energy between LAHET physics and neutron transport using evaluated nuclear data given by the third value on the phys n card to 20 MeV Base Case phys n 1000 3 150 Case 1 phys n 1000 j 20 In this case neutron transport is done in the same manner as was done traditionally with LAHET and HMCNP The neutron problem summary for this case is shown below sample pr
171. A similar approach is taken to calculate net n xn production Net neutron production may also be calculated by realizing that the only loss mechanisms for neutrons are escape and capture The sum of the weights in the neutron loss column under escape and capture is thus equal to the net neutron production The values listed in the problem summary are collision estimators meaning they are tal lied when a collision occurs during transport Uncertainties are not calculated by MNCPX for these collision estimated quantities A reasonable upper limit on the relative uncer tainty would be given by the inverse square root of the number of source particles launched We provide here five different variations for the calculation of net neutron production for this simple target geometry In the base case we transport protons neutrons and charged pions The transition energy between LAHET physics and neutron transport MCNPX User s Manual 125 MCNPX User s Manual E Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium using tabular nuclear data is set at 150 MeV and the LA150 library is used All protons are transported using LAHET physics Nucleon and pion interactions simulated by LAHET physics use the Bertini intranuclear cascade model Variations from this base case are outlined in Table A 1 below For each case 20 000 source protons were transported Table A 1 Neutron Problem Summaries
172. CCC 715 MCNPX 2 4 0 OAK RIDGE NATIONAL LABORATORY managed by UT BATTELLE LLC for the U S DEPARTMENT OF ENERGY RSICC COMPUTER CODE COLLECTION MCNPX 2 4 0 Monte Carlo N Particle Transport Code System for Multiparticle and High Energy Applications Contributed by Los Alamos National Laboratory Los Alamos New Mexico RADIATION SAFETY INFORMATION COMPUTATIONAL CENTER Legal Notice This material was prepared as an account of Government sponsored work and describes a code system or data library which is one of a series collected by the Radiation Safety Information Computational Center RSICC These codes data were developed by various Government and private organizations who contributed them to RSICC for distribution they did not normally originate at RSICC RSICC is informed that each code system has been tested by the contributor and if practical sample problems have been run by RSICC Neither the United States Government nor the Department of Energy nor UT BATTELLE LLC nor any person acting on behalf of the Department of Energy or UT BATTELLE LLC makes any warranty expressed or implied or assumes any legal liability or responsibility for the accuracy completeness usefulness or functioning of any information code data and related material or represents that its use would not infringe privately owned rights Reference herein to any specific commercial product process or service by trade name trademark manufactur
173. CNPX which uses standard F90 allocation schemes for dynamic variables on all platforms RSICC tested this release on the following systems 1 AIX 4 3 3 IBM 43P 260 with XL C C 4 4 XL Fortran 6 1 2 Dell PowerEdge6400 running RedHat Linux 7 0 with PGF90 4 0 2 and gcc iv 3 Intel Pentium running RedHat Linux 6 1 with PGF90 3 3 2 and pgcc 4 Sun UltraSparc 60 under SunOS5 6 with F90 2 0 and C 5 0 The LANL developers ran MCNPX 2 4 0 on the following systems Their executables are included in the distribution Installation may fail with different compilers Sun Solaris WorkShop Fortran Compilers 6 update 2 Fortran 95 6 2 SGI IRIX MIPSpro Compilers Version 7 30 under 64 bit IRIX and 32 bit IRIX HP HPUX HP F90 v2 4 10 IBM AIX xlf90 Version 7 Release 1 DEC Alpha Tru64 running OSF1 V5 0 with Compaq Fortran V5 3 915 Intel Linux 7 with The Portland Group Fortran Group Inc f90 3 2 3 Windows2000 on Pentium IV Compaq Visual Studio 6 6 and Microsoft C 6 0 Note that Compaq Visual Studio 6 5 fails to compile the code but 6 1 works 10 REFERENCES a included in documentation MCNPX User s Manual Version 2 4 0 LA CP 02 408 September 2002 L S Waters ed MCNPX User s Manual Version 2 3 0 LA UR 02 2607 April 2002 b background references J F Briesmeister Ed MCNP A General Monte Carlo N Particle Transport Code Version 4C LA 13709 M April 2000 M B Chadwick P G Young S Chiba S
174. Contribution Card Variable Description n tally number P probability of contribution to detector n from cell i Default P 1 162 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 98 Detector Contribution Card Variable Description I number of cells in the problem Use Optional Consider also using the DD card Section 5 8 11 5 8 13 DXT DXTRAN Form DXT n X1 Y1 Z1 RI RO X2 Yo Z2 RI RO gt 2 wan DWC DWC DPWT Use DXTRAN deterministic transport method At each source or collision point a particle is put on the outermost DXTRAN sphere RO by the next event estimator The particles are then transported inside the DXTRAN sphere Table 5 99 DXTRAN Card Variable Description n particle type coordinates of the point at the center of the it pair X Z el of spheres radius of the i inner sphere in cm RI NOTE The inner sphere is only used to aim 80 of the DXTRN particles All particles start on the outer sphere RO radius of the i outer sphere in cm DWC upper weight cutoff in the spheres DWC lower weight cutoff in the spheres minimum photon weight Entered on DXT N card DPWT only Defaults Zero for DWC DWC and DPWT Use Optional Consider using the DXC N DXC P or DD cards when using DXTRAN 5 8 14 DXC DXTRAN Contribution Form DXCm nP P2
175. Creation Card Variable Description unit no 1 99 filename name of the file access sequential or direct form formatted or unformatted record length record length in direct access file Default None none sequential formatted if sequential unformatted if direct not required if sequential no default if direct Use When a user modified version of MCNP needs files whose characteristics may vary from run to run Not legal in a continue run Example FILES 21 ANDY S F 0 22 MIKE D U 512 If the filename is DUMN1 or DUMN2 the user can optionally use the execution line message to designate a file wnose name might be different from run to run for instance in a continue run Example FILES 17 DUMN1 MCNPX INP TEST3 DUMN1 POST3 5 10 SUMMARY OF MCNPX INPUT CARDS The following table lists the various input cards and when they are required Two kinds of defaults are involved in the following table 1 if a particular entry on a given card has a default value that value is listed in the appropriate location on the card and 2 the omission of a card from the input file sometimes has a default meaning and if so the default description is preceded by an asterisk Table 5 113 Summary of MCNPX Input Cards Use Card and Defaults Page General Categories optional Message block plus blank terminator 34 required Problem title card 34 174 MCNPX User s Manual MCNPX User s Manual Version 2 4
176. ET MCNP Code Merger X Division Research Note XTM RN U 97 012 LA UR 97 4891 Los Alamos National Laboratory April 1997 htto www xdiv lanl gov XT M hughes LA UR 97 4891 cover html 3 J F Briesmeister editor MCNP A General Monte Carlo N Particle Transport Code Los Alamos National Laboratory report LA 12625 M March 1997 http www xdiv lanl gov XCI PROJECTS MCNP manual html 4 J Linhard V Nielsen and M Scharff Kgl Dan Vidensk Selsk Mat Fys Medd 36 5 152 No 10 1968 M Robinson The Dependence of Radiation Effects on Primary Recoil Energy Radi ation Induced Voids in Metals AEC Symp Ser 26 p 397 US Atomic Energy Commission 1971 MCNPX User s Manual MCNPxX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium Appendix C Using XSEX3 with MCNP 1 Introduction XSEX3 is the code which analyzes a history file produced by LAHET3 or MCNPX and gen erates double differential particle production cross sections for primary beam interactions Cross section plots may also be generated by creating a file to be plotted by MCNP It is necessary to execute either code in a specific mode described below to achieve the desired cross section calculation The execution of XSEX3 assumes that the LAHET run was made using the option N1COL 1 Under this option the incident particle interacts directly in the specified material in which the source is
177. ISTP and all records on HISTX indiscriminately MCNPX User s Manual 217 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Surface crossing records appearing on a SSW written file are not distinguished as to whether they correspond to an internal surface crossing or to escape into the external void Therefore for use with MCNPX the original intent of this option may most easily be achieved by defining the external importance 0 leakage region as the exterior of a sphere containing the complete geometry then only specifying the defining spherical surface on the SSW card that controls the contents of the surface crossing file Energy binning is specified by the usual methods The number of energy bins is given by NERG The number of particle types for which surface crossing data are to be tallied is given by NTYPE and must be gt 0 The polar angle bins representing lines of latitude are defined by entering the NFPRM cosine values in the FPARM array Binning in the azimuthal angle corresponding to lines of longitude is determined by the value of NPARM which defines NPARM equal azimuthal angle bins from a lower bound of 0 on the first bin to an upper bound of 360 on the last bin The value of KOPT determines the orientation used to define the angles as shown in Figure D 1 The allowed options are as follows KOPT 1 the z axis defines the polar angle and is measured counter clockwise from the x direction K
178. If you are more familiar with csh you will need to adjust things appropriately NOTE Com ments about the shell commands start with the character Also don t be alarmed by the generous amount of output from the configure and make scripts They work hard so you don t have to 18 MCNPX User s Manual MCNPxX User s Manual Ap Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium go to your user home directory cd home me unpack the distribution that was copied from the net or a CDROM This creates home me mcnpx_2 3 0 gzip dc mcnpx_2 3 0 tar gz tar xf go into the unpacked distribution cd menpx_2 3 0 execute the configure script the prefix tells where to put the executables and libraries configure prefix home me Make the executable mcnpx program the bertin and pht libraries and run the regression tests make all make tests now install the executable mcnpx program and the bertin and pht libraries in nhome me bin and home me lib mcnpx make install 3 1 3 4 Individual Private Installation Done Better For a more flexible version of our second example we will look at the same single non privileged user Me on acomputer loading and building a private copy of the code This time however the user will use a second directory away from the mcnpx source code in which to do the build This can be done several times in different build directories with dif ferent
179. Interpolate Multiply and Jump amp Log Shortcuts 35 4 1 9 Vertical Input Format 2 00 c eee eee 36 4 1 10 Particle Designators 2 00 eee 38 vi MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 4 1 11 Default Values 000 eens 41 4 2 Input Error MessageS 00 cece eee eee eee eee eee 41 4 3 Geometry Errors eee eee eee a ee eee ee ee ee ee 41 4 4 Storage Limitations 00 0 eee 43 By POTENT NG ise ied i satan ae aan ca lyn ie a Va I 45 5 1 The Interactive Geometry Plotter 00 00 c eee ee eee 45 5 2 Tallies amp Cross sectionS 0c eee eee 47 5 2 1 Input for MCPLOT and Execution Line Options 47 5 2 2 Plot Conventions and Command Syntax 00 0 0200e00s 49 5 2 222 DIOL sts Tna sa Grae tebe ans ek tie art nae a a ene eas 49 5222 COMOUP plot rroiak ma Sea we ene cet be Pela d Selah eden oe 49 5 2 2 3 Command syntax a nus saaa aee 49 Plot Commands Grouped by Function 00cc cece ee eee 50 3 GOOMOLIY ce Sit tea nO ec i E E mu eden wean E E n e 58 53T Cell Le a on eS E a NA aa i a aei 58 53 2 Surface air aa eee ieee a a a a a a eae 60 5 3 2 1 Surfaces Defined by Equations 00 eee 60 5 3 2 2 Axisymmetric Surfaces Defined by Points 62 5 3 2 3 General Plane Defined by Three Points
180. LTEST Parameter Meaning NERG Defines the number of energy or momentum bins for which cross sections will be calculated For NERG GT 0 an energy momentum boundary record is required For NERG 0 only energy integrated cross sections will be generated The default is 0 NANG Defines the number of cosine bins for which cross sections will be calculated For NANG not equal to 0 a angular bound ary record is required For NANG 0 only angle integrated cross sections will be generated Positive values of NANG indi cate cosine bin boundaries will be defined negative values indicate angle bin boundaries in degrees will be specified The default is 0 FNORM An overall multiplicative normalization factor to be applied to all cross sections The default is 1 0 To convert to millibarns use FNORM 1000 0 obtain macroscopic cross sections use an atom density KPLOT A plot control flag the default is 0 Any nonzero value will cause the output to be written to a file XSTAL in the format of an MCNP MCTAL file for subsequent plotting see below MCNPX User s Manual 155 MCNPxX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium Parameter Meaning IMOM Chooses energy or momentum to be used in cross section def inition IMOM 0 cross sections are tabulated by energy MeV and differential cross sections are calculated per unit energy per Me
181. MCNP practice The input file default name INT for HTAPE3X has the following structure 1 Two records of title information 80 columns each 2 An option control record MCNPX User s Manual 205 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 3 Additional input as required by the chosen option Items 2 and 3 above are written as list directed input 1 Repeat counts are allowed including repeat counts for commas to take default values i e 4 expands to Multiple cases may be processed for each case the above structure applies Slashes are allowed only in the first pair of title cards unless each title card containing one or more slashes has an S in column 1 The option control record defines the options to be used and the additional input information that must be specified for the problem The structure of this record is IOPT NERG NTIM NTYPE KOPT NPARM NFPRM FNORM KPLOT IXOUT IRS IMERGE ITCONV IRSP ITMULT Some of the parameters in this record may optionally be preceded by a minus sign whose meaning is defined below Thus if NTIM is specified by inserting 3 in the option control record it is interpreted as NTIM 3 with a minus sign flag attached In the discussion which follows input control parameters are treated as positive or zero quantities even though the flag may be present Table B 1 Applicability of Input Control Parameters IOPT
182. MIPSpro Compilers Ver terminates with errors in random sion 7 30 places in the code HP HPUX HP F90 v2 4 10 terminates with errors in random places in the code IBM AIX the one that came lots of syntax errors with AIX 3 4 Alpha Tru64 OSF1 V5 0 Compaq works Fortran V5 3 915 Alpha Linux Compag Fortran works BUT behavior depends on the V1 1 0 1534 Compag Fortran Com file suffix piler V1 1 0 1534 46B31 F gt FORTRAN 77 and F90 gt For tran 90 Intel Linux pgf90 3 2 3 works 3 1 10 In the End Each subdirectory of the MCNPX distribution contains a different utility with its own install target The top level directory also has an install target that moves into the src subdirectory and executes the install target which covers all of the subdirectory install targets The ulti mate destination for the binary executables and associated library files depends upon what parameters were given when running the configure script If prefix VALUE was given to the configure script then the path represented by VALUE is the directory where two subdirectories shown in the table below will be created and populated If no prefix parameter was specified for the configure then a default directory of usr local is used In both cases the bin and lib subdirectories are created and populated 3 2 Libraries and Where to Find Them Several types of data libraries are used by MCNPX including the XSDIR pointer file to nuclea
183. Meson Transport Code with Fission Oak Ridge National Laboratory Report ORNL TM 7882 July 1981 BAR94 V S Barashenkov A Polanski Electronic Guide for Nuclear Cross Sections Comm JINR E2 94 417 Dubna 1994 BER63 WM J Berger Monte Carlo Calculation of Penetration and Diffusion of Fast Charged Particles in Methods in Computational Physics Vol 1 edited by B Alder S Fernbach and M Rotenberg Academic Press New York 1963 p 135 BER70 M J Berger and S M Seltzer Bremsstrahlung and Photoneutrons from Thick Target and Tantalum Targets Phys Rev C2 1970 621 BER63a H W Bertini Phys Rev 131 1963 1801 BER69 H W Bertini Phys Rev 188 1969 1711 MCNPX User s Manual 117 MCNPX User s Manual Ap Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium BET34 H A Bethe and W Heitler On Stopping of Fast Particles and on the Creation of Positive Electrons Proc Roy Soc London A146 1934 p 83 BEV69 Phillip R Bevington Data Reduction and Error Analysis for the Physical Sci ences McGraw Hill Book Company 1221 Avenue of the Americas New York NY 10020 1969 BLU50 O Blunck and S Leisegang Zum Energieverlust schneller Elektronen in dun nen Schichten Z Physik 128 1950 500 BLU51 O Blunck and R Westphal Zum Energieverlust energiereicher Elektronen in dunnen Schichten Z Physik 130 1951 641 BRE81 D J Brenner R E
184. N 0 N 0 N N N N N 105 N N N R N 0 N N N N N 8 N N N 0 N 0 0 N N N N 108 N N N R N 0 0 N N N N 9 109 O O R R O N N O O O O 10 110 O O R R N N N O O O O 11 111 O N R R O N N O N N N 12 112 O N R R O N N O N N N 13 O O R O O N N O O O O 14 N N N O N N N N N N O 136 MCNPX User s Manual MCNPX User s Manual i Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium Table B 1 Applicability of Input Control Parameters Continued IOPT NERG NTIM NTYPE NPARM NFPRM KPLOT IXOUT IMERGE ITCONV IRSP ITMULT 114 N N N R N N N N N N O 15 N N N O N O O N N N N 115 N N N R N O O N N N N 16 O N N 0 N O N N N N N 116 O N N R N O N N N N N R required O optional N not used IRS is optional with any value of IOPT IOPT defines the editing option to be applied as defined below For all but IOPT 13 100 may added to the basic option type to indicate that the tally over a list of cell surface or material numbers will be combined in a single tally Prefixing IOPT by a minus sign when allowed indicates an option dependent modification to the tally NERG when applicable defines the number of energy bins for the tally the maximum is 2000 The default is 0 implying that only a total over energy will be produced If NERG is gt and is preceded by a minus sign the tally in each energy bin will be divided b
185. NPX User s Manual Ap Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium 5 1 Reaction Probability Calculation Interaction physics in MCNPX is determined in two ways through table based data and through on line calculation with physics models The physics models are used wherever the lower energy tabular data are missing MCNPxX version 2 3 0 can be used with any of the existing libraries now available for use with the MCNP4B code These can be obtained through the RSICC facility at Oak Ridge National Laboratory or through the NEA outside the United States One set of libraries however is distributed directly with MCNPX the new LA150 compendium see Section 4 3 1 In the physics module energy regime the time tracking is governed by several cutoffs The actual interaction chosen is the minimum in time of the following e Particle decay time see Table 5 1 last column Time to the next interaction as determined by the computed cross section e Low energy cutoff see Table 5 1 fifth column Note that minimum energy cutoffs may be set by the user to 001 MeV for most particles Neutrons neutral pions and neutri nos are an exception where a 0 0 cutoff can be set However unless there is tabular data or a specially implemented low energy physics model no interactions of these particles will occur between below the minimum recommended in Table 5 1 e User specified time cutoff 5 2 Collisional Stopp
186. NPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 1 7552E 01 2 225 7E 02 QO QO QO QO 0 Dis QO QO QO QO QO 0 QO QO QO QO 0 QO QO 9 3603E 00 1 3946E 02 7 4771E 02 1 2545E 00 4 9306E 01 0 QO 1 9115E 01 6 4394E 01 5 0000E 05 7 4505E 03 QO 0 1 9011E 01 3 4571E 02 cutoffs tco 1 0000E 34 eco 0 0000E 00 wcl 5 0000E 01 wc2 2 5000E 01 Note the net neutron production calculated with the ISABEL INC model is 17 569 which is 3 8 below the value predicted by the Bertini INC model This is consistent with other studies that reveal slightly lower neutron production resulting from ISABEL as compared to Bertini Case 4 In the next variation from the base case we use the new evaluated proton libraries for transporting protons below 150 MeV replacing the Bertini model used at all proton energies in the base case We invoke transport of protons with energies less than 150 MeV by including a phys h card to specify the transition energy between LAHET physics and data evaluations for proton transport Base Case phys h 1000 j 0 MCNPX User s Manual 199 Case 4 phys h 1000 150 The neutron summary table for this case is shown below MCNPxX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Sample problem spallation target Case 4 neutron creation tracks weight energy neutron loss per source particle source 0 QO QO escape nucl interaction 3082
187. NPX User s Manual 119 MCNPX User s Manual E Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium and Safety International Centre for Theoretical Physics Miramare Trieste Italy 15 April 17 May 1996 proceedings published by World Scientific A Gandini G Reffo eds Vol 2 p 424 532 1998 FIR96 R B Firestone and V S Shirley Table of Isotopes 8th Edition John Wiley New York 1996 GOU40 S Goudsmit and J L Saunderson Multiple Scattering of Electrons Phys Rev 57 1940 24 GUD75 K K Gudima G A Osokov and V D Toneev Model for Pre Equilibrium Decay of Excited Nuclei Yad Fiz 21 1975 260 Sov J Nucl Phys 21 1975 138 GUD83 K K Gudima S G Mashnik and V D Toneev Cascade Exciton Model of Nuclear Reactions Nucl Phys A 401 1983 329 HAL88 J Halbleib Structure and Operation of the ITS Code System in Monte Carlo Transport of Electrons and Photons edited by Theodore M Jenkins Walter R Nelson and Alessandro Rindi Plenum Press New York 1988 153 HOW81 R J Howerton ENSL and CDRL Evaluated nuclear Structure Libraries UCRL 50400 Vol 23 Lawrence Livermore National Laboratory February 1981 HUG95 H G Hughes and L S Waters Energy Straggling Module Prototype Los Alamos National Laboratory Memorandum XTM 95 305 U November 29 1995 HUG97 H G Hughes R E Prael R C Little MCNPX The LAHET MCNP Code
188. NPX accepts all standard MCNP input cards with additional card options that take advantage of the multiparticle capabilities of MCNPX Modifications to standard MCNP inputs are described in Section 5 4 and following Section 5 5 7 describes new cards added to control the model physics options MCNPX uses when table based data are not available Use of high energy proton and photonuclear data library capabilities has already been described Accelerator simulation applications have a need for specialized source input to describe an incident particle beam Usually this takes the form of a directed beam of particles monoenergetic with a transverse gaussian profile To facilitate this a new source option has been added to MCNPX and is described in Section 5 6 7 4 1 INP FILE The INP file can have two forms initiate run and continue run 4 1 1 Initiate Run This form is used to set up a Monte Carlo problem describe geometry materials tallies etc and run if message block is present The initiate run file has the following form MCNPX User s Manual 31 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Message Block Optional Blank Line Delimiter Title Card Cell Cards Blank Line Delimiter Surface Cards Blank Line Delimiter Data Cards Blank Line Terminator Optional but recommended Anything Else Optional MCNPX interprets blank lines as the end of preceding information type MCNPX will stop reading the i
189. OPT 2 the z axis defines the polar angle and is measured counter clockwise from the y direction KOPT 3 the x axis defines the polar angle and is measured counter clockwise from the y direction KOPT 4 the x axis defines the polar angle and is measured count er clockwise from the z direction KOPT 5 the y axis defines the polar angle and is measured counter clockwise from the z direction KOPT 6 the y axis defines the polar angle and is measured counter clockwise from the x direction A value of KOPT 0 defaults to KOPT 1 For NPARM 1 a null record must be supplied in place of the LPARM array NPARM 0 defaults to NPARM 1 but the null record need not be supplied If a null record is supplied for the FPARM array NFPRM equal cosine bins from 1 0 to 1 0 are supplied The following is an example of the input for using option 13 Title 1 Option 13 Example Title 2 100 Equal Solid Angle Bins 218 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 13 10 1 1 10 10 0 5 800 1 In this case the energy is binned in 10 equal lethargy intervals of half decade width below 800 MeV and normalized per MeV No time binning is done Only neutrons are edited The z axis determines the polar angle and the azimuthal angle is measured from the x axis Ten azimuthal angle bins are used and 10 equal polar angle cosine bins are defined by takin
190. PARM equal azimuthal angle bins from a lower bound of 0 on the first bin to an upper bound of 360 on the last bin The value of KOPT determines the orientation used to define the angles as shown in Figure D 1 The allowed options are as follows KOPT 1 the z axis defines the polar angle and is measured counter clock wise from the x direction KOPT 2 the z axis defines the polar angle and is measured counter clock wise from the y direction KOPT 3 the x axis defines the polar angle and is measured counter clock wise from the y direction KOPT 4 the x axis defines the polar angle and is measured count er clock wise from the z direction MCNPX User s Manual 147 MCNPX User s Manual Ap Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium KOPT 5 the y axis defines the polar angle and is measured counter clock wise from the z direction KOPT 6 the y axis defines the polar angle and is measured counter clock wise from the x direction A value of KOPT 0 defaults to KOPT 1 For NPARM 1 a null record must be sup plied in place of the LPARM array NPARM 0 defaults to NPARM 1 but the null record need not be supplied If a null record is supplied for the FPARM array NFPRM equal cosine bins from 1 0 to 1 0 are supplied The following is an example of the input for using option 13 Title 1 Option 13 Example Title 2 100 Equal Solid Angle Bi
191. PRINT MPLOT PTRAC PERT 5 9 1 PRDMP Print and Dump Cycle Form PRDMP NDP NDM MCT NDMP DMMP Table 5 105 Print amp Dump Cycle Card Variable Description NDP increment for printing tallies NDM increment for dumping to RUNTPE file MCT gt 0 write MCTAL file but delete all timing information from MCTAL and OUTP NDMP maximum number of dumps on RUNTPE file TFC entries and rendezvous every lt 0 1000 particles DMMP 0 1000 particles or if multiprocessing 10 total during the run gt 0 DMMP particles Increment gt 0 histories or KCODE cycles lt 0 running time in minutes 166 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Default Print only after the calculation has successfully ended Dump every 15 minutes and at the end of the problem Do not write a MCTAL file Write all dumps to the RUNTPE file DMMP 0 see table above Use Recommended especially for complex problems 5 9 2 PRINT Output Print Tables Form PRINT x Table 5 106 Output Print Tables Variable Description table numbers to be included in the output file blank gives the basic output print X1 X92 prints basic output plus the tables specified by Xx the table numbers x4 Xo x1 Xp prints full output except the tables specified by X4 Xp Default No PRINT card in the INP file or no PRINT option on the execution li
192. Production of Tritium e Additional cards needed for photonuclear interactions are described in Section 6 1 e Discussion of PHYS CUT_N and CUT_H and EMAX revised in Section 6 1 7 A list of new MCNPX specific cards was added to section 6 1 10 e Default parameter settings for LCA LCB LEA and LEB were corrected e Nontracking Change The default setting for IPREQ on the LCA card has been changed to 1 Use pre equilibrium model after intranuclear cascade In 2 1 5 the default had been 0 No pre equilibrium model will be used This change was made at the overwhelming request of the users e Additional NOACT options added to table 6 3 for attenuation and cross section mode e Examples in Section 6 3 were reformatted for greater clarity A note regarding the dif ference between the a value defined in the manual and that shown in Table 10 of the MCNPX output file was included Chapter 7 e Mistype in SPABI corrected Chapter 8 e Inversion 2 1 5 cylindrical mesh tally grids must have an inner radius starting ata value greater than 0 0 This restriction has been removed in version 2 3 0 e Spherical Mesh Tally option is added e Clarification on normalization of Mesh Tallies is added e GNUPLOT has been added to the supported gridconv graphics options Appendix C which reproduces part of gridconv has been removed e Nontracking Change The form of the two radiography cards has been changed Input decks are back
193. ST not equal to 0 suppresses date and timing on the conventional output file OUTXS the default is 0 LTEST is used to produce output for compari son during MCNPX installation and should not be used gen erally At most two additional records may be required depending on the values specified for NERG and NANG For NERG gt 0 a record defining NERG upper energy bin boundaries from low to high defined as the array ERGB I I 1 NERG The first lower bin boundary is implicitly always 0 0 The definition may be done in four different ways 1 The energy boundary array may be fully entered as ERGB I 1 NERG in increas ing order If two or more but less than NERG elements are given with the record terminated by a slash the array is completed using the spacing between energy boundaries obtained from the last two entries If only one entry is given it is used as the first upper energy boundary and also as a constant spacing between all the boundaries If only two entries are given with the first negative and the second positive the second entry is used as the uppermost energy boundary ERGB NERG and the first entry is interpreted as the lethargy spacing between bin boundaries Thus the record bf 0 1 800 will specify ten equal lethargy bins per decade from 800 MeV down For NANG gt 0 arecord is required to define the NANG upper cosine bin boundaries They should be entered from low to high with the last
194. STP for all particles causing collisions If IOPT is pre ceded by a minus sign the edit is performed only for events initiated by the primary source particles For KOPT 0 or 1 separate edits are performed for cascade and evap oration phase production In addition total nucleon production from either phase is edited For KOPT 2 or 3 only the cascade production is edited For KOPT 4 or 5 only the evaporation phase production is edited For KOPT 6 or 7 only the total particle produc tion is edited For KOPT 8 or 9 only the pre fission evaporation production is edited For KOPT 10 or 11 only the post fission evaporation production is edited If KOPT is even the edit is over cell numbers if KOPT is odd the edit is over material numbers If NPARM is zero the edit is over the entire system The parameters NTYPE and NFPRM are not used If KPLOT 1 a plot is made of each edit table With KOPT 0 or 1 the cascade production for neutrons and protons is simultaneously plotted as a dotted line with the total production Unless otherwise modified tally option 3 or 103 represents the weight of particles emitted in a given bin per source particle As such it is a dimensionless quantity 6 Edit Option lIOPT 4 or 104 Track Length Estimate for Neutron Flux Option 4 is not available in this version use a standard F4 flux tally 7 Edit Option IOPT 5 or 105 Residual Masses and Average Excitation Option 5 provides an edi
195. Surface and Cell Tallies tally types 1 2 4 6 and 7 114 5 7 1 2 Repeated Structures Tallies 0000 c eae eee 116 5 7 1 2 1 Multiple bin format 0 0 eee 117 50222 Brackets iaar et ahead ud ata digs wena at ox 118 5 7 1 2 3 Universe format 2 0 eee 118 5 7 1 2 4 Use of SDn card for repeated structures tallies 119 5 7 1 3 Detector Tallies tally type 5 2 2 eee 120 5 7 1 4 Pulse height Tallies tally type 8 2 0008 121 5 7 2 FCn Tally Comment 00 00 eee eee 121 Sa En Tally Energy lt 2 3c 3 aceite oe Pe ee Se he ee ie 122 5 7 4 Tn Tally Time sasni eee cette vie eee eee ee ae 122 5 7 5 Cn Cosine Card tally type 1 and 2 0 20c cess 122 5 7 6 FQn Print Hierarchy 000 e eee eee 123 5 7 7 FMn Tally Multiplier 2000 eee 124 5 7 8 DEn and DFn Dose Energy and Dose Function 126 5 7 9 EMn Energy Multiplier 00 0 c eee eee 128 5 7 10 TMn Time Multiplier 000 eee 128 5 7 11 CMn Cosine Multiplier tally type 1 only 128 5 7 12 CFn Cell Flagging tally types 1 2 4 6 7 2 0000 129 5 7 13 SFn Surface Flagging tally types 1 2 4 6 7 129 5 7 14 FSn Tally Segment tally types 1 2 4 6 7 0055 130 5 7 15 SDn Segment Divisor tally types 1 2 4 6 7 131 5 7 16 FUn Special Tally or TALLYX Input 2 200000ee 131
196. T make a working space that reminds you it s a debug version mkdir mcnpx debug cd mcnpx debug execute the configure script request debug for the Makefiles also specify where to put the installed code and which compilers to use MCNPX_DIST configure with F C f90 with C C cc with LD usr ccs bin Id with DEBUG prefix home me libdir usr mcnpx data now make the executable mcnpx program We will omit the regression tests this time although it would be a good idea to run them again if different compiler optimization values are used make install That s all there is to it There are many other options available with this new version of mcnpx Please read the User s Notes or the Programmer s Notes for more details MCNPX User s Manual 17 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 3 1 4 Directory Reorganization In order to accommodate the use of the autoconf utility to generate the Makefiles it became necessary to arrange the source code and regression test directories a bit We also added a config directory to hold autoconf related code The new directory structure is depicted in Figure 3 1 Each of the levels contains a collection of autoconf files and links Removal of any of these files will break the automated configure and make capabilities First Level Data contains data used with the bertin phtlib makexs targets Docs contains files describing this mcnpx distribution Test
197. The x axis of the grid is defined as the cross product of a unit vector in the t direction and a unit vector in the reference direction S Cosine bins Image Grid A Scatter Contribution Ne Source Point Pinhole s SANS easter Ref Point sensors ccc aeghen NANA Transport eee ET h eed ee Fa N Segment bins an aT a mune x2 2 22 Source Geometry Figure 8 2 Pinhole image projection 8 2 2 Transmitted Image Projection In the transmitted image projection case the grid acts like a film pack in an X ray type image or transmitted image projection The diagram in figure 8 3 shows how the planar grid type of image capability is set up In MCNPX 2 3 0 additional capability has been added to allow the user to set up a cylindrical grid for generating an image In both cases for every source or scatter event a ray trace contribution is made to every bin in the detec tor grid This eliminates statistical fluctuations across the grid that would occur if the grid location of the contribution from each event were to be picked randomly as would be the case if one used a DXTRAN sphere and a segmented surface tally For each event source or scatter the direction to each of the grid points is determined and an attenuated ray trace contribution is made As in pinhole image projection there are no restrictions as to location or type of source used These tallies automatically bin in a
198. Theoretical Electron Atom Elastic Scattering Cross Sections Selected Elements 1 keV to 256 KeV Atom Data and Nucl Data Tables 15 1975 443 RUT11 E Rutherford The Scattering of a and b Particles by Matter and the Structure of the Atom Philos Mag 21 1911 669 SCH82 P Schwandt et al Phys Rev C 26 55 1982 SEL88 S M Seltzer An Overview of ETRAN Monte Carlo Methods in Monte Carlo Transport of Electrons and Photons edited by T M Jenkins W R Nelson and A Rindi Plenum Press New York 1988 p 153 SEL91 S M Seltzer Electron Photon Monte Carlo Calculations The ETRAN Code Appl Radiat Isot Vol 42 No 10 1991 pp 917 941 SNO96 E C Snow Radiography Image Detector Patch for MCNP private communication SNO98 E C Snow Mesh Tallies and Radiography Images for MCNPX Proceedings of the Fourth Workshop on Simulating Accelerator Radiation Environments SARE4 Tony A Gabriel ed 1998 113 STE71 R M Sternheimer and R F Peierls Phys Rev B 3 no 11 June 1 1971 3681 TRI97a R K Tripathi F A Cucinotta J W Wilson Universal Parameterization of Absorption Cross Sections NASA Technical Paper 3621 January 1997 TRI97b R K Tripathi J W Wilson and f A Cucinotta New Parameterization of neutron Absorption Cross Sections NASA Technical Paper 3656 June 1997 VAV57 P V Vavilov lonization Losses of High Energy Heavy Particles Sovie
199. User s Manual 105 MCNPxX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 53 Surface Source Read Card Keyword Description rzbze 0 lt zb lt ze Allparticles with acceptable polar angles relative to the surface normal are started so that they will pass through a cylindrical window of radius r starting at zb from the center of the source sphere and BCW ending at ze from the center The axis of the cylinder is parallel to the z axis of the auxiliary original coordinate system and contains the center of the source sphere The weight of each source particle is adjusted to compensate for this biasing of position and direction Default no cylindrical window Use Required for surface source problems Example 1 Original run SSW 1 2 3 Current run SSROLD 3 2 NEW 6 7 12 13 TRD5 COL 1 SISL 4 5 SP5 4 6 SB5 3 7 Particles starting on surface 1 in the original run will not be started in the current run because 1 is absent from the list of OLD surface numbers Particles recorded on surface 2 in the original run will be started on surfaces 7 and 13 and particles recorded on surface 3 in the original run will be started on surfaces 6 and 12 as prescribed by the mapping from the OLD to the NEW surface numbers The COL keyword causes only particles that crossed surfaces 2 and 3 in the original problem after having undergone collisions to be started in the current problem The TR entry indicates that dist
200. V IMOM not equal 0 cross sections are tabulated by momentum MeV c and differential cross sections are estimated per unit momentum per MeV c IYIELD not equal to 0 estimates differential yields or multiplicities for nonelastic and elastic reactions rather than cross sections The integral over energy and angle for each particle type will be the multiplicity per nonelastic reaction or unity for the elastic scat tering of the incident particle if it is included in the calculation LTEST not equal to 0 suppresses date and timing on the conventional output file OUTXS the default is 0 LTEST is used to produce output for compari son during MCNPX installation and should not be used gener ally At most two additional records may be required depending on the values specified for NERG and NANG For NERG gt 0 a record defining NERG upper energy bin boundaries from low to high defined as the array ERGB I lI 1 NERG The first lower bin boundary is implicitly always 0 0 The definition may be done in four different ways 1 The energy boundary array may be fully entered as ERGB I 1 NERG in increas ing order If two or more but less than NERG elements are given with the record terminated by a slash the array is completed using the spacing between energy boundaries obtained from the last two entries If only one entry is given it is used as the first upper energy boundary and also as a constant spac
201. Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium If you make changes to any of the input files or macros it will be necessary to regenerate the configure script so it can pick up all of the changes you have made to the component files To regenerate the configure scripts use the following command from the Top Level directory autoreconf localdir config f This forces regeneration of the configure scripts that live at each directory level of the distribution The localdir config parameter lets autoconf know where to find the macros that are called in the various configure in files it encounters 3 1 7 2 How to add a new hardware OS compiler Example 1 Add the Portland Group compiler to the Linux OS on all Intel platforms yes i s already there but we will step through it For hardware and operating system study the case statements in the menpx_2 3 0 con fig config guess and mcnpx_2 3 0 config config sub files You may need to insert a new case to handle your variation of hardware and operating system versions Luckily most of the current platforms are already specified therefore it is unlikely that you would have to edit either of these files For the most recent version of autoconf check with the lt http Awww gnu org software soft ware html HowToGetSoftware gt GNU autoconf distributions There may be a more recent version of autoconf s config guess config sub scripts that cover your conf
202. Version 2 4 0 September 2002 LA CP 02 408 Defaults If A or B is missing LOG is chosen for that table Example DES5 E E E E4 Ek DF5 LIN Fy Fo F3 F3 Fr This example will cause a point detector tally to be modified according to the dose function F E using logarithmic interpolation on the energy table and linear interpolation on the dose function table Table 5 66 Standard Dose Functions value of ic Meaning Neutron Dose Function 10 ICRP 21 1971 20 NCRP 38 1971 ANSI ANS 6 1 1 1977 31 ANSI ANS 6 1 1 1991 AP anterior posterior 32 ANSI ANS 6 1 1 1991 PA posterior anterior 33 ANSI ANS 6 1 1 1991 LAT side exposure 34 ANSI ANS 6 1 1 1991 ROT normal to length amp rotationally symmetric 40 ICRP 74 1996 ambient dose equivalent Photon Dose Function 10 ICRP 21 1971 20 Claiborne amp Trubey ANSI ANS 6 1 1 1977 31 ANSI ANS 6 1 1 1991 AP anterior posterior 32 ANSI ANS 6 1 1 1991 PA posterior anterior 33 ANSI ANS 6 1 1 1991 LAT side exposure 34 ANSI ANS 6 1 1 1991 ROT normal to length amp rotationally symmetric 35 ISO isotropic Default ic 10 Example DF4 DFO ic 40 iu 1 lin fac 123 4 DF1 iu 2 fac 2 log ic 34 Use optional MCNPX User s Manual 127 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 5 7 9 EMn Energy Multiplier Form EMn M My Table 5 67 Energy Multiplier Card V
203. X require three special libraries BERTIN containing the elemental cross section data needed by the Bertini model PHTLIB containing nuclear structure data needed to generate de excitation photons BARPOL DAT containing new high energy total reaction and elastic cross sections They are unpacked with the rest of the code and if make install is executed placed in the lib directory There are basically 2 ways that the code tries to find these files 1 MCNPX tries to open the files named bertin and phtlib in the current directory If the user wants to keep these file in another directory a symbolic link should be made from whatever directory you are in when running the code The following unix com mand can be used to do this In s home me lib bertin 2 A default pathname is coded in the fortran data statements in the file src Ics inbd F This can be changed by the user but you must remember to recompile the code Look for the variable currently holding the string usr local xcodes3 Icsdir ber tin and the similar variable referencing a location for phtlib Change them to reflect the appropriate location of the two data files on your system and re make the code A typical location for these two files might be usr local lib mcnpx This would be the preferable method when a community of users is accessing one copy of the code ona single system As suggested above we recommend making a s
204. X3 Analyzes a HISTP history file and generates double differential particle production cross sections for primary beam interactions RELATED DATA LIBRARIES Libraries specific to the LAHET Bertini model are included in a file called BERTIN Gamma production cross sections from spallation products are included in a file called PHTLIB A new version of PHTLIB is available for MCNPX 2 4 0 including improved data and also metastable state information High energy total reaction and elastic cross sections are contained in a file called BARPOL DAT MCNPXxX includes a test library of cross sections for running the sample problems but the test library is not suitable for real problems Running the code requires continuous energy cross section data included in the DOO20S5ALLCP03 MCNPXDATA package or equivalent data To receive the data from RSICC users must include MCNPXDATA on their request license and Export Control form The DO0205ALLCP03 MCNPXDATA package is comprised of DLC 200 MCNPDATA which was released for use with MCNP4C plus the LA150N library of 42 high energy neutron data tables LA150U photonuclear data for 12 isotopes and LA150H proton data tables for 41 isotopes In LA150N the neutron energy is extended to 150 MeV except for Be 9 which only goes to 100 MeV This library typically extends ENDF B VI data from 20 MeV to 150 MeV therefore charged particle and recoil nuclei data will sometimes not be available below 20 MeV Exceptions are
205. a eE E oven eating ieee hate balk Sette seat E RE oh 152 Appendix C Using XSEX3 with MCNPX 200 eee e eee 153 MCNPX User s Manual ix Accelerator Production of Tritium Figures Figure 3 1 Figure 4 1 Figure 8 1 Figure 8 2 Figure 8 3 Figure 8 4 Figure 8 5 Figure A 1 Figure B 1 MCNPX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 Directory Organization Structure c ceccccececeeeeeeeeeeeceeeeeeeeeseeaeeeseaeeseeeeeenaeeees 22 interaction processes oo eeeeceeeeeeeeeeeeeeeeeeeeeeaeeeeeeeaaeeeeeeeaaaeeeeeeeaaeeeeseeeaeeeeeeeaaes 39 Mesh Tally depiction of a sample spallation target neutron fluence 93 Pinhole image Projection eeccceeeeeeceeeeeeeeneeeeeeeenaeeeeeeeaeeeeeeeeaeeeeeeeaaeeeeeeeaees 104 Transmitted image Projection eeeeeeeeeeeeeeeeeeeeeeeeeeeeaeeeeeeeeaeeeeeeeaaeeeeneeaees 108 Effect of too fine binning ON energy spectra ooo eee eee eeeseeeeeeeenteeeeeeenaeeeeeeeaas 108 Energy spectra for neutrons produced from a proton beam on tungsten 111 Neutron production from a spallation target eeeceeeeeeeseeeeeeeenaeeeeeeennaeeeeeeeaas 125 Use of the KOPT Parameter for HTAPE3X Option 13 essere 149 MCNPX User s Manual Accelerator Production of Tritium Tables Table 3 1 Table 3 2 Table 3 3 Table 4 1 Table 4 2 Table 4 3 Table 4 4 Table 4 5 Table 5 1 Table 6 1 Table 6 1 Table 6 3 Table 6 4 Table 6 5 Ta
206. a is available at http t2 lanl gov data photonuclear html These tables are based on IAEA Photonuclear Data Library http iaeand iaea or at photonu clear and as of this writing are available for MCNPX use on a test basis only Forty two neutron evaluations have been completed for the LA150N library The neutron evaluations are a combination of existing ENDF B VI Release 5 neutron evaluations up to 20 MeV and new evaluated data from 20 150 MeV For the mercury isotopes the data below 20 MeV are from recent JENDL evaluations Note the Beryllium 9 neutron library is based on work completed 10 years ago and only goes to 100 MeV Proton evaluations to 150 MeV have been completed for the same materials except that 12C and 4 Ca are available rather than elemental C and Ca In contrast to the neutron eval uations the proton work is entirely new as no previous ENDF B VI low energy evaluations existed upon which to build The minimum energy of the LA150 proton evalu ations ranges from 1 keV to 3 MeV 150 MeV proton data libraries will be first distributed concurrent with the release of MCNPX version 2 3 0 48 MCNPX User s Manual MCNPxX User s Manual i Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium The gt 20 MeV neutron and all proton evaluations include e production cross sections for light particles e production cross sections for gammas e production cross sections for heavy recoil particles
207. a test team Code configuration management is 1 MCNPX MCNP MCNP4B LAHET and LAHET Code System LCS are trademarks of the Regents of the University of California Los Alamos National Laboratory MCNPX User s Manual xiii MCNPxX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium involves the CVS system and methods of assessing code development progress are being implemented Training courses are held regularly This manual has been developed to support the latest MCNPX version 2 3 0 RSICC release as an updated of the previously released document for version 2 1 5 Geometry basic tally and graphical capabilities of MCNPX do not fundamentally differ from the standard MCNP4B code as released to RSICC in March 1997 The MCNPX manual should be used as a Supplement to the MCNP4B manual although some additional remarks are made on basic concepts where they might need clarification for the high energy community The primary purpose of the MCNPX manual is to describe the exten sions and additional features incorporated that directly address the high energy multiparticle environment envisioned in these applications Except where noted in Chapter 2 all of the original capabilities of MCNP are intact and MCNPX is intended to be back ward compatible with standard MCNP input files MCNPX code development team is now testing a version of the code fully updated to the capabilities of MCNP4C We are
208. ags m4 file is included into the aclocal m4 file via the m4 include macro Because autoconf covers redefines the m4 include behavior the m4 built in macro is used to call the m4 version of include Within flags m4 the ARCH SYSTEM FCOMP CCOMP variables are used in various case statements to define needed symbols Check to see if your arch system fcomp and ccomp combination appear in this large case statement You may need to add your combination For our example we are looking for usages of intel linux pgf77 and gcc Around line 21 there is a case statement that depends on the value of the SYSTEM variable We must have case label for the linux operating system If linux did not occur we would add it as a case and define the needed symbols that our scripts will use later when generating the various Makefile files MCNPX User s Manual 29 MCNPxX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium Around line 70 we see a case statement that depends on the value of the ARCH vari able We must have a case label for the intel hardware architecture There is an i 86 label The is a wildcard character and will match a variety of intel machines i286 i386 i486 If i 86 did not appear we would add it as a case and define the needed symbols that our scripts will use later when generating the various Makefile files Throughout the rest of the flags m4 file we find a varie
209. al MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 preted as the lethargy spacing between bin boundaries Thus the record 0 1 800 will specify ten equal lethargy bins per decade from 800 MeV down e For NTIM gt 0 a record specifying NTIM upper time bin boundaries from low to high defined as the array TIMB I l 1 NTIM The first lower time boundary is always 0 0 The same four methods that are allowed for defining the energy boundaries may also be used to define the time bin boundaries Table B 4 Order of HTAPESX Input Records IOPT option control record always required ERGB I 1 NERG upper energy bin limits TIMB l l 1 NTIM upper time bin limits ITIP I l 1 NTYPE particle type identifiers LPARM l l 1 NPARM surface cell or material identifiers FPARM I l 1 NFPRM upper cosine bin boundaries DNPARM l l 1 NPARM 1 normalization divisors original source definition record for RESOURCE option new source definition record for RESOURCE option ITOPT TWIT TPEAK TWIT parameters for TIME CONVOLUTION ERESP I l 1 NRESP energy grid for RESPONSE FUNCTION FRESP I l 1 NRESP 1 function values for RESPONSE FUNCTION IRESP 1 l 1 NRESP 1 interpolation scheme for RESPONSE FUNCTION segment definition record or window definition record CN I I 1 3 arbitrary direction vector for defining cosine binning
210. al Fission Form TOTNUNO or blank Default If the TOTNU card is absent prompt is used for non KCODE calculations and total v is used for KCODE calculations Use All steady state neutron problems with fission should use this card 5 4 5 NONU Fission Turnoff Form NONUa do dj Amya or blank Table 5 27 Fission Turnoff Argument Description 0 fission in cell i treated as capture gammas produced 1 fission in cell i treated as real gammas produced di 2 fission in cell i treated as capture gammas not pro duced mxa number of cells in the problem MCNPX User s Manual 77 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Default If the NONU card is absent fission is treated as real fission Use Needed with SSR with fissioning neutron problems only Example NONU When fission is already modeled in the source such as SSR it should not be duplicated in transport and should be turned off with NONU 5 4 6 AWTAB Atomic Weight Form AWTAB ZAID AW ZAID 5 AW c i Table 5 28 Atomic Weight Argument Description ZAID ZAID used on the Mm material card excluding the X for class of data specification AW atomic weight ratios Default If the AWTAB card is absent the atomic weight ratios from the cross section directory file XSDIR and cross section tables are used Use Discouraged Occasionally useful when XS card introduces rare isoto
211. alf maximum of the observed energy broadening in a physical radiation detector fwhm a b4E cE where E is the energy of the particle The units of a b and c are MeV MeV and none respectively The energy actually scored is sampled from the Gaussian with that fwhm See Chapter 2 TMC ab All particles should be started at time zero The tally scores are made as if the source was actually a square pulse starting at time a and ending at time b INC No parameters follow the keyword but an FUn card is required Its bin boundaries are the number of collisions that have occurred in the track since the creation of the current type of particle whether at the source or at a collision where some other type of particle created it If the INC special treatment is in effect the call to TALLYX that the presence of the FUn card would normally trigger does not occur Instead IBU is set by calling JBIN with the number of collisions as the argument ICD MCNPX User s Manual 133 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 No parameters follow the keyword but an FUn card is required Its bins are the names of some or all of the cells in the problem If the cell from which a detector score is about to be made is not in the list on the FUn card the score is not made TALLYX is not called The selection of the user bin is done in TALLYD SCX k The parameter k is the name of one of the source distributions and is the k
212. allow the user control of physics options A summary of the cards follows The options controlling the Bertini and ISABEL physics modules are taken from the User Guide to LCS PRA8Q The user is referred to that document for further information CEM allows neutrons and protons up to 5 GeV and pions up to 2 5 Gev to initiate nuclear reactions Valid targets are nuclei with a charge number greater than 5 and a mass number greater than 11 The light nuclei are passed to the Bertini ISABEL models that use the Fermi Breakup model in this regime CEM consists of an intranuclear cascade model followed by a pre equilibrium model and an evaporation model Possible fission events are initiated in the equilibrium stage for compound nuclei with a charge number greater than 70 The fragmentation of the fission event is handled by modules from the RAL fission model Fission fragments undergo an evaporation stage that depends on their excitation energy After evaporation a de excitation of the residual nuclei follows generating gammas using the PHT data Future developments of MCNPX will allow greater freedom in the selection of physics options INC pre equilibrium evaporation fission etc so the user may compare the effect of varying one parameter at a time Presently CEM is still relatively self contained All of the input values on the four cards have defaults which will be taken in the absence of the cards or with the use of the MCNP style J input opt
213. ally type 3 Keyword Descriptions Continued Keyword Description mfact Can have from one to four numerical entries following it e The value of the first entry is in reference to an energy dependent response function given on a MSHMFn card no default e The second entry is 1 default 1 for linear interpolation and 2 for loga rithmic interpolation e Ifthe third entry is zero default 0 the response is a function of energy deposited otherwise the response is a function of the current particle energy e The fourth entry is a constant multiplier and is the only floating point entry allowed default 1 0 If any of the last three entries are used the entries preceding it must be present so that the order of the entries is preserved Only one mfact keyword may be used per tally nterg Allows one to record in a separate mesh array the local energy deposition only due to particles otherwise not considered or tracked in this problem This allows the user to ascertain the potential error in the problem caused by allow ing energy from non tracked particles to be deposited locally This can be a serious problem in neglecting the tracking of high energy photons or electrons trans Must be followed by a single reference to a TR card that can be used to trans late and or rotate the entire mesh Only one TR card is permitted with a mesh card MCNPX User s Manual 99 MCNPX User s Manual E Version 2 3 0 Ap
214. als Engineering Development and Demonstration ED amp D project A code develop ment team under the leadership of Dr H Grady Hughes was formed Because the Los Alamos accelerator community has long supported the work of Dr Richard Prael in the development of the LAHET Code System it was decided to build on this base by com bining the capabilities of LAHET and MCNP into one code This involved extending the capabilities of MCNP4B to all particles and all energies and including the use of physics models in the code to compute interaction probabilities where table based data are not available Additional development has been provided by the theoretical efforts of the T 16 group at Los Alamos particularly in the areas of nuclear data evaluation and expansion of physics based models A program of experimental activities was also undertaken including mea surement of various cross sections and development of more complex benchmarks specific to the APT and AAA projects Our commitment to modern software management and quality assurance methods in the development of MCNPX is very strong The code is used for the design of high intensity accelerator category 2 nuclear facilities and has already been used to design a major cat egory 3 activity at the LANSCE high power beamstop MCNPX development is guided by a set of requirements design and functional specification documents Code testing is per formed on a large scale by a volunteer bet
215. also assessing the implications of Fortran 90 conversion on all parts of the code We anticipate release of that code version later in 2002 The MCNPX team is actively exploring code modularity in a component architecture for mat which will enable the simple addition of new routines into the code and also allow the code to communicate with related software applications It will also give original authors full control of their contributions We anticipate that this advanced version of MCNPX will be available in 2003 It is hoped that MCNPX will be of use to the Monte Carlo radiation transport community in general The development of the modular approach in future versions of the code will facil itate the addition of new capabilities to the base code and make this tool a flexible reliable aid in the exploration of both traditional and new mixed energy multiparticle applications Laurie Waters Deputy Group Leader D 10 Nuclear Systems Design Los Alamos National Laboratory April 2002 xiv MCNPX User s Manual MCNPX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium 1 Introduction The MCNPX program represents a major extension of the MCNP code putting in place the ability to track all particles at all energies MCNPX version 2 3 0 is built on MCNP4B as released to RSICC in 1997 LAHET version 2 8 plus extensions developed in LAHET version 3 0 Additional development of the CEM code was
216. am hcnv to translate LAHET text input to binary input phtlib builds and executes a program trx to translate LAHET text input to binary input gridconv converts output files generated by mesh tally and mctal files into a variety of different graphics formats htape3x reads the history tapes optionally generated by mcnpx and performs post processing on them makexs a cross section library man agement tool that converts type 1 cross sections to type 2 cross sections and vice versa xsex3 a utility associated with the new cross section generation mode for mcnpx which allows tabulation of cross section sets based on physics models include contains include files shared across directories and include files localized in subdirectories mcnpx the organizing root directory for the mcnpx program Third Level cem dedx etc directories that organize the Fortran77 and C source code files that are related to different aspects of the MCNPX program MCNPX User s Manual 21 MCNPxX User s Manual Version 2 3 0 April 2002 ry LA UR 02 2607 Accelerator Production of Tritium Fourth Level individual Fortran77 and C source code files for a particular aspect of MCNPX Figure 3 1 Directory Organization Structure menpx_2 2 0 iscell f autoconf files and links configure in config install sh Makefile Readme autoconf files and links fluka89 autoconf files and links mene menpx main dedx gvaviv 77
217. ange emax Tabl J Unused be sure to put the J s in the keyword string Charged particle straggling control ISTRG 0 Vavilov model best fe 1 continuous slowing down approximation 1 old MCNPX_2 2 4 and earlier J see above MCNPX User s Manual 85 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 34 Proton Physics Options Keyword Description Light ion recoil control Number of light ions h d t s a to be created at each proton elastic scatter off H D T 3 He 4 He CUT n 2J 0 is usually needed for n h d t s a RECL Note that protons having elastic scatter with hydrogen pro duce more protons which may produce an overwhelming number of protons 0 lt RECL lt 1 Default PHYS h 100 OO0OJOJO Use Optional Example PHYS h 800 10 150 J 0j 2 5 5 2 5 Other Particles Form PHYS lt pl gt EMAX J J J ISTRG Table 5 35 Other Charged Particle Physics Options Keyword Description EMAX Upper limit for particle energy MeV JJJ Unused be sure to put the J s in the keyword string Charged particle straggling control 0 Vavilov model best 1 continuous slowing down approximation 1 old MCNPX_2 2 4 and earlier ISTRG Use Optional Default PHYS n 100 3J 0 Example PHYS d 800 3J 1 5 5 3 TMP Free Gas Thermal Temperature Form TMPn Typ Ton Tin Tin 86 MCNPX User s Manual MCNPX User s Manual Version
218. anual Version 2 4 0 September 2002 LA CP 02 408 Figure 3 1 Directory Organization Structure menpx_2 2 0 Data Docs src miscellany autoconf files and links configure script configure in config install sh Makefile Readme autoconf files and links flukas9 ies autocont files and links 3 1 5 User s Notes Do not edit the Makefiles generated by the configure script In order to change the contents of the generated Makefiles you must alter the contents of several input files that the configure script uses Please read the Programmer s Notes in the next subsection for instructions Table 3 1 contains options which are available for use as parameters to the configure script for mcnpx 2 4 0 MCNPX User s Manual 19 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 3 1 Configure Script Parameters Option Syntax Effect on the generated Makefile if requested Effect on the generated makefile if NOT requested compile step for the gener ated Makefiles with STATIC linking of the compiled files STATIC is the default cannot results in a static archive be used at the same time as mcnpx a SHARED with SHARED linking of the compiled files STATIC is used this option is results in a dynamically exploratory for future linked executable releases of MCNPX mcnpx so with DEBUG a debug switch appears inthe no debug switch appears in the compile step for t
219. ard see Section 5 7 14 and use the SDn card see Section 5 7 15 to enter the appropriate values You can also redefine the geometry as another solution to the problem The detector total is restricted to 20 The tally total is limited to 100 Note that a single type 5 tally may create more than one detector MCNPX User s Manual 113 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 5 7 1 1 Surface and Cell Tallies tally types 1 2 4 6 and 7 Simple Form Fn pl Sj Sk General Form Fn pl S4 So S3 Sq S5 Sg S7 T Table 5 54 Surface and Cell Tallies Variable Description n tally number pl particle designator S problem number of surface or cell for tallying T total over specified surfaces or cells Only surfaces bounding cells and listed in the cell card description can be used on F1 and F2 tallies Tally 6 does not allow E Tally 7 allows N only In the simple form above MCNP creates k surface or cell bins for the requested tally listing the results separately for each surface or cell In the more general form a bin is created for each surface or cell listed separately and for each collection of surfaces or cells enclosed within a set of parentheses Entries within parentheses also can appear separately or in other combinations Parentheses indicate that the tally is for the union of the items within the parentheses For unnormalized tallies tally type 1 the union
220. ariable Description n tally number M multiplier to be applied to the energy bin Default None Use Requires En card Tally comment recommended 5 7 10 TMn Time Multiplier Form TMn M My Table 5 68 Time Multiplier Card Variable Description n tally number M multiplier to be applied to the time bin Default None Use Requires Tn card Tally comment recommended 5 7 11 CMn Cosine Multiplier tally type 1 only Form CMn M Mx Table 5 69 Cosine Multiplier Card Variable Description n tally number Mi multiplier to be applied to the cosine bin 128 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Default None Use Tally type 1 Requires Cn card Tally comment recommended 5 7 12 CFn Cell Flagging tally types 1 2 4 6 7 Form CFn C Ck Table 5 70 Cell Flagging Card Variable Description n tally number C problem cell numbers whose tally contributions are to be flagged Default None Use Not with detectors or pulse height tallies Consider FQn card Example F4 N 6 10 13 CF4 3 4 In this example the flag is turned on when a neutron leaves cell 3 or 4 The print of Tally 4 is doubled The first print is the total track length tally in cells 6 10 and 13 The second print is the tally in these cells for only those neutrons that have left cell 3 or 4 at some time
221. aries used with MCNP4B model such effects in detail therefore we usually see a discontinuity in predictions in going from library upper limits to INC physics At energies around the pion threshold the simpler INC physics can ade quately model reaction probabilities Starting in 1996 the APT project undertook the extension of standard nuclear data evalu ations to 150 MeV for a number of elements of interest to the plant design At the same time proton evaluations were also developed and a program of photonuclear library devel 1 The pion production threshold is 290 MeV for nucleons interacting with nucleons at rest For a nucleon interacting with nucleons in a nucleus additional Fermi energy in the nucleus lowers the threshold to 200 MeV 46 MCNPX User s Manual MCNPxX User s Manual i Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium opment is underway Since that time other programs have contributed funding for other elements For example the Spallation Neutron Source SNS program has funded the development of mercury evaluations in order to design their liquid mercury target Pro grams involved in accelerator transmutation are working on actinide libraries In contrast to previous versions MCNPX version 2 3 0 can take full advantage of all features of the extended neutron libraries and has added proton and photonuclear libraries In addition work is underway to produce libraries of certain ligh
222. ary 1 5 0000E 05 7 4680E 03 tabular boundary 1 5 0000E 05 7 4680E 03 gamma xn 0 0 0 particle decay 0 0 0 adjoint splitting 0 0 Ox total 346009 1 7296E 01 3 3488E 02 total 346009 1 7296E 01 3 3488E 02 number of neutrons banked 316635 average time of shakes cutoffs neutron tracks per source particle 1 7300E 01 escape 5 7337E 00 tco 1 0000E 34 neutron collisions per source particle 2 3611E 01 capture 4 7022E 01 eco 0 0000E 00 total neutron collisions 472212 capture or escape 5 7293E 00 wcl 5 0000E 01 net multiplication 0 0000E 00 0000 any termination 5 1842E 00 wc2 2 5000E 01 Net neutron production for this case is 15 648 n p which is 14 3 than the base case value Note also that CEM took twice as long to run as the base case Both of these factors are well known and CEM improvements is a very active project in the MCNPX program The increase in time is understood and will be corrected in future versions through algo rithm optimization The lower n p values are also being extensively benchmarked and improvements involving the transitions from INC to Preequilibrium and Preequilibrium to evaporation have been developed Until the new version is available the user should be cautious in using the CEM model for production calculations Summary Results compiled for each case of this example are shown in Table A 2 Note the run time for the case where the ISABEL INC model is used is about 15 greater than the base case using the Bertini
223. ase Icaj j j Case 3 lca jJ 2 This changes the value of the variable IEXISA third value on the Ica card from its default value of 1 to 2 The neutron problem summary for this case follows sample problem spallation target Case 3 neutron creation tracks weight energy neutron loss tracks weight energy per source particle per source particle 198 MCNPX User s Manual source nucl interaction particle decay weight window cell importance weight cutoff energy importance dxtran forced collisions exp transform upscattering tabular sampling n xn fission photonuclear tabular boundary gamma xn adjoint splitting total 0 302047 0 0 0 0 380298 1 5102E 01 3 9089E 00 5 0000E 05 1 9011E 01 3 2679E 02 1 8916E 01 7 4505E 03 3 4571E 02 escape energy cutoff time cutoff weight window cell importance weight cutoff energy importance dxtran forced collisions exp transform downscattering capture loss to n xn loss to fission nucl interaction tabular boundary particle decay total 351353 25121 3823 380298 number of neutrons banked 355177 neutron tracks per source particle 1 9015E 01 neutron collisions per source particle 2 6865E 01 total neutron collisions 531297 net multiplication 0 0000E 00 0000 average time of escape capture shakes 5 7572E 00 4 9166E 01 capture or escape 5 7530E 00 any termination 5 3162E 00 MC
224. asic MCNP4B capability as well as extensions to the higher energy modules unique to MCNPX A full PVM version based on MCNPX 2 1 5 has already been prepared at Oak Ridge National Laboratory and that version is forming the basis for formal implementation into later versions of the code For those wishing to run with PVM we recommend the following compile with option with FFLAGS DMULTP DPVM Unfortunately there is no with FLIB option for the configure script therefore a small amount of editing must be done in Makefile h FLIB should be defined as L path lfpvm lpvm The user is warned that multiprocessing in 2 3 0 has not yet been extended to the higher energy physics region This is an area of active progress in code development 3 1 7 Programmer s Notes Autoconf is not new it has been available as a configuration management tool for several years We have just recently adopted its use to simplify the build process for the mcnpx end user community to allow the flexibility to build and keep multiple versions of mcnpx and to improve our software development process 26 MCNPX User s Manual MCNPxX User s Manual i Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium 3 1 7 1 Where it all starts the relevant files what s in them Refer to the diagram and related description given in the figure 3 1 Table 3 2 Config Directory File Name Purpose local m4 a file containing al
225. at f BLOCK DATA subrou tine This can be edited to change the directory but the code must be recompiled D pl Oy Ol BOO The actual coding in MCNP4B for this is a bit complex Upon detailed examination the MCNPX team has come up with the following slightly modified set of directions In the following cases if the desired file is found exit the list with the success 1 Look in the current working directory for the file 2 Look at the DATAPATHE input directive or the DATAPATH environment variable 2a If there is a DATAPATH directive in the input file look there for the file 2b If there was no DATAPATHz directive then examine the DATAPATH environment variable for a value 2b 1 If there is an environment value use that value as a directory to search for the file 2b 2 If there is no value environment variable not set then look for the file again in the current working directory 3 Look in a default place 3a If there was a DATAPATH directive then the default place is either the value of the DATAPATH environment variable if there was one or value of the pre processor symbol LIBPREFIX from the autoconfiguration process typically usr local lib mcnpx 3b If there was not a DATAPATHE directive in the input file then the default is just the LIBPREFIX pre processor symbol 4 If the file is not found by now then it is a fatal error The MCNPX teams plans to try and clarify this in the code for a future version It is rec
226. ater Once CVF and MSVC are installed simply open a Command Prompt window enter the MCNPX BLD directory and execute GNU Make C gt Make Be sure to execute the SETUPX BAT file as explained above so GNU Make can be found it is provided as an executable in MCNPX BIN Also be sure that your PATH environment variable is less than 255 characters as this version of GNU Make has a problem if this is exceeded A MAKEPATH BAT file is provided in MCNPX BIN as an example of how to reduce your PATH variable to a minimum set of directories note this assumes CVF and MSVC are installed on the C drive The X11 library and include files are provided in MCNPX LIB and should not be moved from here As on a Unix platform you can build any subcomponent of MCNPX by entering that directory and executing Make All the source files are in the MCNPX SRC directory and one should take care in modifying any of these files Patches to MCNPX can be developed as done for MCNP however one should contact us for the needed script file and instructions to apply such a patch If a stack overflow error is generated this is NOT an MCNPX bug A stack limit must be specified upon linking The included executable has a stack limit of 32 MBytes This can be increased by editing the Makfile h file in the MCNPX BLD SRC MCNPxX directory line 66 and rebuilding MCNPX 3 3 LIBRARIES AND WHERE TO FIND THEM Several types of data libraries are used by MCNPX including the XSDIR poin
227. ation specifies a density change to 0 03 atoms cm in cell 1 This change is applied to both neutron and photon interactions Example 2 3 1 1 12 34 56 mat 1 at 1 g cm 121 1 7 8 9 10 11 12 mat1 at 1 g cm C M1 material is semiheavy water M11001 334 1002 333 8016 333 C M8 material is heavy water M8 1002 667 8016 333 PERT2 n CELL 3 12 MAT 8 RHO 1 2 This perturbation changes the material composition of cells 3 and 12 from material 1 to material 8 The MAT keyword on the PERT card specifies the perturbation material The MCNPX User s Manual 141 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 material density was also changed from 1 0 to 1 2 g cm to change from water to heavy water Example 3 PERT3 n p CELL 110i12 RHO 0 METHOD 1 This perturbation makes cells 1 through 12 void for both neutrons and photons The estimated changes will be added to the unperturbed tallies Example 4 60 13 2 34 105 106 74 73 mat 13 at 2 34 g cm M13 1001 2 8016 2 13027 2 26000 2 29000 2 M15 1001 2 8016 2 13027 2 26000 2 29000 4 PERT1 p CELL 60 MAT 15 RHO 2 808 RXN 51 9i 61 91 ERG 1 20 PERT2 p CELL 60 RHO 4 68 RXN 2 This example illustrates sensitivity analysis The first PERT card generates estimated changes in tallies caused by a 100 increase in the Cu n n cross section ENDF B reaction types 51 61 and 91 above 1 MeV To effect a 100 increase double the composition fra
228. ave not changed the name of the new PHTLIB library but we do recommend that you call it PHTLIB_SPEC1 and make a symbolic link in your current directory such as In s nome user data PHTLIB_SPEC1 PHTLIB Further information on the contents of the new library can be found in PRA98c PRAOOb Although the new libraries do contain updated nuclear structure data termination at 1 msec metastable states may cause confusion in interpretation of results Careful thought should go into the decision to switch to the new library In the future we hope to produce a method whereby the user can designate the termination time in the code 4 3 2 Photoelectric Interactions No change in the standard MCNP4B capability to track photoatomic interactions and elec tron transport has been made in MCNPX Below we summarize part of the discussion presented in the MCNP4B manual with comments of interest to those using these capa bilities for high energy applications In particular the user should be aware that the upper limit for interactions by photons is 100 GeV and for electrons 1 GeV Cross sections for all photon and electron interactions are taken from the ENDF library Part of the future work for MCNPX will be to investigate the use of the LLNL Evaluated Photon and Electron librar ies which will also raise these energy limitations 54 MCNPX User s Manual Accelerator Production of Tritium 4 3 2 1 Photon Interactions MCNPX User s Manual Vers
229. between the factored cross section and the more accurate partial wave cross sections of Riley A discussion of the extension of this theory to heavy charged particles is found in Section 5 4 Electron Bremsstrahlung MCNP and MCNP4B use the Bethe Heitler BET34 Born approximation to sample bremsstrahlung photons Formulas and approximations relevant to the present level of theory in MCNP4B and MCNPX can be found in the paper of Koch and Motz KOC59 Particular prescriptions appropriate to Monte Carlo calculations have been developed by Berger and Seltzer BER70 These data have been converted to tables including bremsstrahlung production probabilities photon energy distributions and photon angular distributions and are used directly in MCNP4B MCNPX An alternative to the use of tabular data is a simple material independent probability distribution 2 1 pdu Pau 2 1 Bu MCNPX User s Manual 61 MCNPX User s Manual E Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium where u cos and B v c is used to sample for the angle of the photon relative to the direc tion of the electron according to the formula _ 2s 1 f w p1p where is a random number This method is used for detectors and DXTRAN spheres where a set of source contributions p u consistent with the tabular data is not available On should note that although bremsstrahlung for heavy charge particles is a valid physical ph
230. ble The amp continuation symbol is considered as part of the comment not as a continuation command Default No comment 5 6 2 KCODE Criticality Source Form KCODENSRCK RKK IKZ KCT MSRK KNRM MRKP KC8 ALPHA Table 5 50 KCODE Card Variable Description NSRCK number of source histories per cycle RKK initial guess for Ker IKZ number of cycles to be skipped before beginning tally accumulation KCT number of cycles to be done MSRK number of source points to allocate storage for KNR normalize tallies by O weight 1 histories maximum number of cycle values on MCTAL or MBISE RUNTPE summary and tally information averaged over KC8 0 all cycles 1 active cycles only Defaults NSRCK 1000 RKK 1 0 IKZ 30 KCT IKZ 100 MSRK 4500 or 2 NSRCK KNRM 0 MRKP 6500 KC8 1 Use Required for criticality calculations 5 6 3 KSRC Source Points for KCODE Calculation Form KSRCx 9121 X27232 102 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 51 KSRC Card Variable Description Xi Yi Zi location of initial source points Default None If this card is absent an SRCTP source file or SDEF card must be supplied to provide initial source points for a criticality calculation Use Optional card for use with criticality calculations 5 6 4 SSW Surface Source Write Form SSW SS2 C C S3S keyword values The
231. ble 6 6 Table 7 1 Table 8 1 Table 8 2 Table 8 3 Table 8 4 Table 8 5 Table 8 6 Table 8 7 Table 8 8 Table 8 9 Table A 1 Table A 2 Table B 1 Table B 2 Table B 3 Table B 4 MCNPX User s Manual Configure Script Parameters cccsccceeeceseeceeeeeeeeeeaaeeseneeeeaaeseeaeeeseaeeseaaeeeneas 23 Contig DifeCtory i 02 nae et See ate ee ee eee ee ee 27 Fortran 90 Compilers 0c ccccscecceee cesses a a a a i aaNet 33 Summary of Physics in Intermediate Energy Models cccccsceeeteeeeesteeteees 40 Intermediate Energy Model Recommended Ranges cesee eeceeeeeeeeeeeeees 41 Summary of LA150 Libraries 0 0 eeeececeeeeeeeeeeeeeeeeeeeaeeeeeeeeeeaaaeeeeneeeseaeeeseaeeeeeas 47 Charged Particle Production Thresholds for Low Energy Neutron Libraries 49 Summary of Photon Physics Options 0 ccccececcceseeeeeeeeeeeeeaeeeeeeeeseeeeesnaeeeeaes 55 Particles in MONPX stein ea ie tie ee eed 65 Setting Upper Limits for Neutron amp Proton Tabular Data ccccccccsssseeeeee 74 Turning on Photonuclear InteractiOnS c ccceeeeceeeeeeeeeeeeeeeeeeeeeeeeeeneaeeeeeeeneaees 75 LCA Keyword Descriptions ccccceeeeeeeeeeeneeeeeeeeeceaeeeeeaeeeeeaeeeesaaeeseneeeesaeesenees 77 LOB Keyword DeSCriptions ccccceeeeceeeeeeeeeeeeeeceaeeseeeeeeseaeeeeenaeesseeeeeeaeeeninees 80 LEA Keyword Descriptions 0 eececceeeeeneeeeeeenceeeeeeeaeeeeeeenaaeeeeeeeaaeeeeeee
232. ble 7 1 Secondary Particle Biasing Argument Descriptions Argument Description p Secondary Particle Type see Table 5 1 XXX List of primary particles to be considered For example nphe represents reactions of neutrons pho tons protons and electrons No spaces are allowed e Ifall particles are to be considered the entry should be all En Upper energy bin limit The lower bin limit is considered to be zero Sn Use Splitting if Sn gt 1 Splitting Use Roulette if 0 lt Sn lt 1 As many SPABI cards as needed can be used to cover any number of secondary particle types and there is no limit on the number of En Sn pairs Every time an interaction takes place in MCNPX which results in secondary particles gen eration the code checks to see if secondary particle biasing is turned on If so the particle causing the interaction is compared with the list of primary particles on the SPABI card to see if these secondary particles are to be considered or not If the primary particle is in the list the secondary particle energy is used to determine the energy bin and subsequent splitting or roulette to be played before the particles are banked An adjustment is then made to the number of particles resulting from this type interaction scored in the summary tables It should be noted that all of the split particles coming out of the bank are identical There fore if there is little or no scattering media between t
233. but an FUn card is required Its bins are a list of atomic weights in units of MeV of particles masquerading as neutrons in a multigroup data library The scores for tally n are then binned according to the particle type as differentiated from the masses in the multigroup data library For example 511 0 would be for electrons and photons masquerading as neutrons ELC c The single parameter c of ELC specifies how the charge on an electron is to affect the scoring of an F1 tally Normally an electron F1 tally gives particle current without regard for the charges of the particles There are 3 possible values for c c 1 to cause negative electrons to make negative scores c 2 to put positrons and negative electrons into separate user bins c 3 for the effect of both c 1 and c 2 134 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 If c 2 or 3 three user bins positrons electrons and total are created 5 7 18 Subroutine TALLYX User supplied Subroutine Use Called for tally n only if an FUn card is in the INP file See discussion in Appendix 5 7 19 TFn Tally Fluctuation Form TFn Le Ip ly Is lm ic Te Ir This card specifies the bin of the tally fluctuation chart statistical information and weight window generator Table 5 76 Tally Fluctuation Card Variable Description n non zero tally number IF of first cell surface or detector on Fn card I
234. cards have defaults which will be taken in the absence of the cards or with the use of the MCNP style J input option LCA IELAS IPREQ IEXISA ICHOIC JCOUL NEXITE NPIDK NOACT ICEM LCA is used to select the Bertini ISABEL or CEM models as well as set certain parame ters used in Bertini and ISABEL CEM is a self contained package with no internal options presently defined Table 6 3 LCA Keyword Descriptions Keyword Description IELAS 0 No nucleon elastic scattering 1 elastic scattering for neutrons only 2 elastic scattering for neutrons and protons default IPREQ 0 No pre equilibrium model will be used 1 Use pre equilibrium model after intranuclear cascade default 2 Use IPREQ 1 and IPREQ 3 randomly with an energy dependent proba bility that goes to IPREQ 3 at low energies and to IPREQ 1 at high incident energies 3 Use pre equilibrium model instead of the intranuclear cascade Note options IPREQ 2 and IPREQ 3 apply only when using the Bertini intranuclear cascade model IEXISA 0 when using the ISABEL model these options default to IPREQ 1 IEXISA 0 Do not use ISABEL intranuclear cascade model for any particle 1 Use Bertini model for nucleons and pions with ISABEL model for other particle types default 2 Use ISABEL model for all incident particle types Note The ISABEL INC model requires a much greater execution time In addi tion incident particle energies should be less than 1 GeV
235. case the entries on the FSn card are the distances along the sym MCNPX User s Manual 105 MCNPX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium metry axis of the cylinder and the entries on the Cn card are the angles in degrees as measured counterclockwise from the positive t axis Image Grid Sk s Cosine bins Source Contribution Object Geometry Source Point Ref Point x1 y1 z1 Center of Gnd f Segment bins Scatter Contribution x2 y2 22 Scatter Source Geometry Figure 8 3 Transmitted image projection When this type of detector is being used in a problem if a contribution is required from a source or scatter event an attenuated contribution is made to each and every detector grid bin Since for some types of source distributions very few histories are required to image the direct or source contributions an additional entry has been added to the NPS card to eliminate unwanted duplication of information from the source The new NPS card now becomes NPS NPP NPSMG Table 8 7 NPS Keyword Descriptions Keyword Description NPP Total number of histories to be run in the problem NPSMG Number of histories for which source contributions are to be made to the detec tor grid When the number of source histories exceeds NPSMG the time consuming process of determining the attenuati
236. cation to the tally NERG when applicable defines the number of energy bins for the tally the maximum is 2000 The default is 0 implying that only a total over energy will be produced If NERG is gt and is preceded by a minus sign the tally in each energy bin will be divided by the bin width to normalize per MeV The total over energy will be unnormalized Table B 2 Applicability of Minus Sign Flags on Input Control Parameters IOPT IOPT NERG NTIM NTYPE NPARM NFPRM 1 101 O O O O O 2 102 O O O O N 3 103 O O O O N MCNPX User s Manual 207 MCNPxX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table B 2 Applicability of Minus Sign Flags on Input Control Parameters Continued IOPT IOPT NERG NTIM NTYPE NPARM NFPRM 5 105 O N N N O N 8 108 O N N N O N 9 109 O O O N O O 10 110 O O O N O N 11 111 N O N N O O 12 112 N O N N O O 13 O O O N N N 14 114 N N N N O N 15 115 O N N N O N 116 O O N N O N O optional N not used NTIM defines the number of time bins for the tally when applicable the maximum is 100 The default is 0 implying that only a total over time will be produced If NTIM is gt 1 and is preceded by a minus sign the tally in each time bin will be divided by the bin width to normalize per nanosecond the total over time will be unnormalized NTYPE defines the number of parti
237. ce of cell 1 is 1 cell 2 is 2 cell 3 is 4 cell 4 is 0 and cells 5 through 25 is 1 A track will be split 2 for 1 going from cell 2 into cell 3 each new track having half the weight of the original track before splitting A track moving in the opposite direction will be terminated in about half that is probability 0 5 the cases but followed in the remaining cases with twice the weight Weight Window Cards See discussion in appendix 5 8 2 WWG Weight Window Generator Form WWG I I Wg JI J J Ig Table 5 87 Weight Window Generator Variable Description problem tally number n of the Fn card The particular L tally bin for which the weight window generator is opti mized is defined by the TFn card invokes cell or mesh based weight window generator gt 0 cell based weight window generator with as the Ic reference cell typically a source cell 0 mesh based weight window generator MESH card required value of the generated lower weight window bound for cell J or for the reference mesh see MESH card Wg 0 means lower bound will be half the average source weight J unused 154 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 87 Weight Window Generator Variable Description toggles energy or time dependent weight windows le 0 means interpret WWGE card as energy bins 1 means interpret WWGE card a
238. ceeding This software engi neering project fully recognizes that some elements of MCNPX are older well tested programs developed outside of the core MCNPX team and may even be written in different languages We also see a very strong future in building the capability to interface effectively with these and even other types of codes such as geometry builders transmutation and thermal hydraulics packages The MCNPX build system is the first step in this process and work on a formal software definition interface language is underway Geometry basic tally and graphical capabilities of MCNPX do not fundamentally differ from the standard MCNP4C code as released in 2000 Input cards have rarely been modified however a number of new cards have been added to control the physics model options set parameters for new particles and control new tally and variance reduction features The present MCNPX 2 4 0 manual differs fundamentally from those released for code ver sions in the past 2 1 5 2 3 0 We are now starting to build a more comprehensive description of the code which eventually will be issued in three parts Vol will cover phys ics and appropriate Monte Carlo methodology Vol II will be the practical user guide for the code Vol Ill will cover items of interest to code developers The present work is equivalent to Volume II and also integrates much more fundamental material than present in the pre vious manuals We are also seriously rethin
239. ces e Medical physics especially proton and neutron therapy e Investigations of cosmic ray radiation backgrounds and shielding for high altitude air craft and spacecraft e Accelerator based imaging technology such as neutron and proton radiography e Design of shielding in accelerator facilities e Activation of accelerator components and surrounding groundwater and air MCNPX User s Manual 1 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 e Investigation of fully coupled neutron charged particle transport for lower energy applications e High energy dosimetry and neutron detection e Design of neutrino experiments e Comparison of physics based and table based data e Charged particle tracking in plasmas e Charged particle propulsion concepts for spaceflight Single event upset in semiconductors from cosmic rays in spacecraft or from the neutron component on the earth s surface e Detection technology using charged particles i e abandoned landmines e Nuclear Safeguards e Nuclear criticality safety e Radiation protection and shielding e Oil well logging In addition to the activities of the beta test team the development of MCNPX is governed by several documents including MCNPX Software Management Plan e MCNPX Requirements MCNPX Design MCNPX Functional Specifications Configuration management of the code is done through CVS which allows us to conveniently track issues
240. cle types for which the edit is to be performed for those options where it is applicable the particle type is that of the particle causing the event which is recorded on the history tape The default is 0 however some options require that a value be supplied KOPT defines a sub option for tally option IOPT The default is 0 NPARM usually defines the number of cells materials or surfaces over which the tally is to be performed when applicable the maximum is 400 If NPARM is preceded by a minus sign NPARM I normalization divisors will be read in as described below The default is 0 however some options require that a value be supplied NFPRM at present is used only to define the number of cosine bin boundaries to read in for particle current tallies the maximum is 400 If NFPRM is preceded by a minus sign cosine bin tallies will be normalized per steradian the total over cosine bins will remain unnormalized i e angle integrated The default is 0 208 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 Table B 3 Particle Type Identification in HTAPE3X Type LAHET Usage MCNPX Usage 0 proton proton p 1 neutron neutron n 2 tt mt m 3 9 ne 4 w 5 ut u u u 7 deuteron deuteron 8 triton triton 9 3He 3He 10 alpha alpha 11 photon photon 12 Kt Kt K 13 Klong Klong 14 K short K short 15 K 16 p 17 h 18 electron electro
241. cnpx s use a default value of usr local lib is used as the full path name for the install step permissions of the destina tion may prohibit success of installation This value overrides the library portion of the prefix if both are given with no_paw or with no_paw yes this means that the symbol NO_ PAW will be defined for compilation and actions are taken in the source to omit PAW capabilities when com piling if omitted the default behav ior is system dependent if the detected hardware soft ware platform can handle PAW it is included MCNPX User s Manual 21 22 MCNPX User s Manual Version 2 4 0 September 2002 Table 3 1 Configure Script Parameters LA CP 02 408 Option Syntax Effect on the generated Makefile if requested Effect on the generated makefile if NOT requested with FFLAGS value There is a separate variable that is used for optimization switches See with FOPT in this table If in doubt run the con figure script and examine the system default or system computed values that appear in the gener ated Makefile h You may want to include the defaults in the string you specify for FFLAGS with this mechanism when configure is run again substitute a quoted or double quoted string for value that represents allowable com piler switch settings these settings will override the system default or system computed values
242. configure make install This method of installation works with MCNPX However the development team recommends a slightly different method so as not to clutter the original source tree with all the products of compiling and building More complex packages The GNU C compiler suite gcc comes to mind warn that the simple build procedure given above is a dangerous practice as it clutters the original source tree with generated Makefiles and compiled objects and makes it difficult to support multiple builds with different options They suggest using a different initially empty directory to be the target of the configure process gzip dc PACKAGE tar gz tar xf mkdir Build cd Build PATH_OF_PACKAGE SOURCE configure make install MCNPX User s Manual 11 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 The MCNPX team also makes this suggestion Please use an empty directory somewhere other than the source distribution s location as the target of the build It keeps the source tree clean and allows multiple builds with different options Even if you think that you will never need additional builds it costs nothing to have the flexibility in the future 3 1 3 Build Examples We will illustrate the new configure and make procedure with two primary examples A system manager installing the MCNPX release for a system with several users and an individual user installing the MCNPX release for their own use A few variatio
243. ction 2 to 4 and multiply the ratio of this increase by the original cell density RHO 1 2 1 0 2 34 2 808 g cm where the composition fraction for material 13 is 1 0 and that for material 15 is 1 2 A change must be made to RHO to maintain the other nuclides in their original amounts Otherwise after MCNP normalizes the M15 card it would be as follows which is different from the composition of the original material M13 M15 1001 167 8016 167 13027 167 26000 167 29000 333 The second PERT card PERT2 p gives the estimated tally change for a 100 increase in the elastic RXN 2 cross section of material 13 RHO 2 34 2 4 68 g cm Example 5 M4 6000 60C 5 6000 50C 5 M6 6000 60C 1 M8 6000 50C 1 PERT1 n CELL 3 MAT 6 METHOD 1 PERT2 n CELL 3 MAT 8 METHOD 1 The perturbation capability can be used to determine the difference between one cross section evaluation and another The difference between these perturbation tallies will give an estimate of the effect of using different cross section evaluations Example 6 1 1 0 05 1 2 3 mat1 at 0 05 x 1074 atoms cm 142 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 M1 1001 1 8016 2 92235 7 M9 1001 1 8016 22 92235 7 F14 n 1 FM14 1 1 6 7 Kos estimator for cell 1 PERT1 n CELL 1 MAT 9 RHO 0 051 METHOD 1 PERT2 n CELL 1 MAT 9 RHO 0 051 METHOD 1 These perturbations involve a 10 increase in the oxyg
244. ctory for the file 2 Look at the DATAPATH input directive or the DATAPATH environment variable 2a If there is a DATAPATH directive in the input file look there for the file 2b If there was no DATAPATH directive then examine the DATAPATH environment variable for a value 2b 1 If there is an environment value use that value as a directory to search for the file 2b 2 If there is no value environment variable not set then look for the file again in the current working directory 3 Look in a default place value 28 3a If there was a DATAPATH directive then the default place is either the of the DATAPATH environment variable if there was one or value of the pre processor symbol LIBPREFIX from the autoconfiguration process typically usr local lib mcnpx MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 3b If there was not a DATAPATH directive in the input file then the default is just the LIBPREFIX pre processor symbol 4 If the file is not found by now then it is a fatal error It is required that MCNPX be run with 64 bit libraries Earlier versions of the code could use 32 bit libraries however studies of long problems have shown that erroneous answer can result with the lesser accuracy data Conversion of Type 1 libraries to 64 bit binaries can be done with the MAKXSF routine described in Appendix C of the MCNP manual The LAHET physics modules in MCNP
245. d Argument Description arbitrary universe number Integer to which cell is n assigned 0 lt n lt 10 default 0 real world universe ii ki kk lattice element parameters for the upper and lower BJE bounds in the i j k directions fully specified fill universe numbers corresponding to cells in order of cells in the cell card section NOTE There must be a universe number for each cell in n4 N the problem The jump feature can be used for cells not J assigned a universe number universe numbers corresponding to each existing lattice element for fully specified fill Use Required for repeated structures Example FILL 0 2 1 2 0 1 442 i 0 1 2 for j 1 amp k 0 040 i 0 1 2 for j 2 amp k 0 033 i 0 1 2 for j 1 amp k 1 440 i 0 1 2 for j 2 amp k 1 Only eight elements of this lattice exist Elements 0 1 0 1 1 0 1 2 0 0 2 1 and 1 2 1 are filled with universe 4 Element 2 1 0 is filled with universe 2 Elements 1 1 1 and 2 1 1 are filled with universe 3 5 3 3 5 TRCL Cell Transformation Form TRCL n or TRCL 010203 XX YX ZX XY YY ZY XZ YZ ZZ M MCNPX User s Manual 71 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 23 Cell Transformation Card Statement Description number of the transformation J lt n lt 999 TRn means that the X Y Z are angles in degrees rather than being the c
246. d 0 lt n lt 10 universe numbers corresponding to cells in order of cells in the cell card section n4 n NOTE There must be a universe number for each cell in the problem The jump feature can be used for cells not assigned a universe number Use Required for repeated structures Examples MCNPX User s Manual 69 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 1 2 3 4 5 6 fill 1 7 1 3 8 u 1 fill 2 lat 1 11 u 2 11 u 2 1 2 3 4 5 6 a FF WN O O O O O px 0 px 50 py 10 py 10 pz 5 pz 5 NO on fF WON px 10 8 py 0 10 py 10 11 s 5504 Cell 1 is filled with cell 2 which is designated universe 1 Cell 2 is filled with cells 3 and 4 universe 2 It is also a square lattice cell to be discussed later Cell 3 is designated universe 2 indicating it is not truncated by the sides of the cell it fills This negative notation of untruncated cells can save computational time The above example can be described with macrobodies as follows 1 0 20 fill 4 2 0 30 u 1 fill 2 lat 1 3 0 11 u 2 4 0 11 u 2 5 0 20 20 rpp 0 50 10 10 5 5 30 rpp 0 10 0 10 11 s 5 5 0 4 5 3 3 4 FILL Fill Form fill n cell card entry or fill i i j j k k n4 ng n3 fully specified fill cell card entry 70 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 or fill n4 nz nz nj data card Table 5 22 Fill Car
247. d on each surface and the tally made within and without the window The window is defined by the intersection of a rectangular or circular tube parallel to the x y or z axis with the tally surface A window definition record appears in place of the segmenting record of option 1 For KOPT 0 1 2 3 or 4 the window is formed by the rectangular tube the window record has the following allowed forms parallel to x axis 1 y min y max z min z max parallel to y axis 2 z min z max x min x max parallel to z axis 3 x min x max y min y max For KOPT 5 6 7 8 or 9 the window is formed by a circular tube cylinder the window record has the following allowed forms parallel to x axis 1 y center z center radius parallel to y axis 2 z center x center radius parallel to z axis 3 x center y center radius 216 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 12 Edit Option IOPT 10 or 110 Surface Flux with Collimating Window Option 10 is identical to option 2 except that the edit is performed inside and outside a window defined as in option 9 Instead of the segmenting record of option 1 a window definition record appears whose form is described in option 9 For KOPT 0 the rectangular form is used and for KOPT 1 the circular form is used Parameter NFPRM is unused 13 Edit Option lIOPT 11 or 111 Pulse Shape of Surface Current For each define
248. d Particle Production Thresholds for Low Energy Neutron Libraries MeV Fe 57 26057 24c 1 943 20 0 20 0 0 8 Ni 58 28058 24c 0 5 20 0 20 0 0 5 Ni 60 28060 24c 2 076 20 0 20 0 2 021E 8 Ni 61 28061 24c 0 549 20 0 20 0 0 07 Ni 62 28062 24c 4 532 20 0 20 0 0 445 Ni 64 28064 24c 6 627 20 0 20 0 2 481 Cu 63 29063 24c 0 9 20 0 20 0 1 742 Cu 65 29065 24c 1 375 20 0 20 0 0 112 Ni 93 41093 24c 20 0 20 0 20 0 20 0 W 182 74182 24c 20 0 20 0 20 0 20 0 W 183 74183 24c 20 0 20 0 20 0 20 0 W 184 74184 24c 20 0 20 0 20 0 20 0 W 186 74186 24c 20 0 20 0 20 0 20 0 Hg 196 80196 24c 20 0 20 0 20 0 20 0 Hg 198 80198 24c 20 0 20 0 20 0 20 0 Hg 199 80199 24c 20 0 20 0 20 0 20 0 Hg 200 80200 24c 20 0 20 0 20 0 20 0 Hg 201 80201 24c 20 0 20 0 20 0 20 0 Hg 202 80202 24c 20 0 20 0 20 0 20 0 Hg 204 80204 24c 20 0 20 0 20 0 20 0 Pb 206 82206 24c 20 0 20 0 20 0 20 0 Pb 207 82207 24c 20 0 20 0 20 0 20 0 Pb 208 82208 25c 4 236 5 816 6 403 1 0e 11 Bi 209 83209 24c 20 0 20 0 20 0 20 0 Note no Helium 3 information or light ion production with Z gt 2 is currently available in the LA150N neutron libraries below 20 MeV Both the LA150 neutron and proton evaluations have also been accepted for incorporation into ENDF B VI as part of Release 6 A compendium CHA99b of neutron and proton data 50 MCNPX User s Manual MCNPX User s Manual i Version 2 3 0 April 2002 LA UR 02 2607
249. d all fission frag ments The default is 1 5 Zero and negative values are an error condition see YZERE above Note Applies only for ILVDEN 1 96 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 44 LEB Keyword Descriptions Continued Keyword Description BZERO The BO parameter in the level density formula for Z 71 and all fission frag ments The default is 10 0 for IEVAP 0 and is also 10 0 for IEVAP 1 Zero and negative values are an error condition see YZERE above Note Applies only for ILVDEN 1 5 6 SOURCE SPECIFICATION SDEF Sin SPn SBn DSn SCn KCODE KSRC SSW SSR SOURCE SRCDX 5 6 1 SDEF General Source Definition Form SDEFsource variable specification Use Required for problems using the general source Optional for problems using the criticality source Table 5 45 General Source Variables Variable Description Default Spades e explicit value none peti distribution function of another variable explicit e g cel 1 an explicit value is given for the variable specified value eds e g cel d1 a specification for a number of cells will be on the information card SI in distribution this case SI1 function of cel fpos d1 cell specification will depend on position specified in appropriate SI cards CEL Cell Determined from XXX YYY ZZZ and pos sibly UUU
250. d bin option 11 provides an edit of the current crossing a surface in an energy and angle bin the mean time t of crossing in the bin the standard deviation o of t given by Fs Y the figure of merit FOM1 given by current o and the figure of merit FOM2 given by current o Unless otherwise modified the current tally is dimensionless The units of t and are nanoseconds while FOM1 is in ns and FOM2 is in ns The parameter FNORM is used to adjust the units of the time variable which are nanoseconds in LAHET3 and does not modify the surface current edit Thus to convert from nanoseconds to microseconds use FNORM 0 001 The bin definition is identical to option 1 including surface segmenting except that NTIM is unused 14 Edit Option IOPT 12 or 112 Pulse Shape of Surface Current with Window Option 12 provides the same edits as option 11 with the same bin definition as option 9 using a collimating window The input is identical to option 9 with the exception that NTIM is unused 15 Edit Option IOPT 13 Global Emission Spectrum The original definition I of option 13 was given by Option 13 tallies the number of particles per unit solid angle entering the external void region with direction cosine falling within a segment of solid angle as such it represents the angular distribution of the emitted particles at a very large distance from the interaction region The option uses any NCOL 4 leakage records on H
251. dary particle biasing and new tallies have been created specific to the intermediate and high energy physics ranges The mesh and radiography tallies were included for 2 and 3 dimensional imaging purposes Energy deposition received a substantial reworking based on the demands of charged particle high energy physics An auxiliary program GRIDCONV converts the mesh and radiography tally as well as standard mctal file results for viewing by independent graphics packages The code may be run in parallel at all energies via PVM Information about MCNPX development can be found on the web site http mcnpx lanl gov Information about the MCNPX beta test program may be obtained from Laurie Waters at LANL A listserver is available for beta test participants 5 METHOD OF SOLUTION All capabilities of MCNP4C3 have been retained Consult the MCNPX User s Manual for applicability to high energy applications MCNPX 2 4 0 has been rewritten in Fortran 90 6 RESTRICTIONS OR LIMITATIONS All standard MCNP neutron libraries over their stated ranges Neutrons in the LA150 library from 0 0 150 0 MeV in tabular range for 42 isotopes except for 9Be to 100 MeV Neutrons from 1 0 MeV in physics model regime Protons from 1 0 to 150 0 MeV in tabular range for 41 isotopes Protons from 1 0 MeV in physics model regime Pions muons and kaons are treated only by physics models Photons from 1 keV 100 GeV Electrons from 1 keV 1 GeV Neutrons
252. de and the assorted data libraries that support it manually In particular the methods that the code used for locating cross section files and the binary data files used by the LCS portions of the code were different from each other and poorly documented Users had to resort to manually editing the Fortran source to customize default directories and to making symbolic links from place to place to support finding all the different sorts of data files Also in past MCNPX releases there was only one Build directory that was hard wired into the distribution s make procedure This build directory held all of the compilation and link ing results This inflexibility made it difficult to build different versions of the code in one place with variations of options debugging vs non debugging or comparing different compilers Sun f77 vs GNU g77 on Solaris or g77 vs the Portland Group pgf77 on Linux It was determined that it would be a great advantage to users if the configuration and build ing process of the software could better determine the hardware platform operating system needed libraries and compilers that were present and perform a more complete customization A utility is available the GNU Autoconf utility that makes this possible A special autoconf generated configure script distributed with MCNPX version 2 3 will examine your computing environment adjust the necessary parameters then generate all Makefiles in your chosen build di
253. detector techniques and is extensively described in SNO96 and SNO98 In essence the radiography focal plane grid is an array of point detectors 5 7 20 1 Pinhole Image Projection In the pinhole image projection case a point is defined in space that acts much like the hole in a pinhole camera and is used to focus an image onto a grid which acts much like the photographic film The pinhole is actually a point detector and is used to define the direction cosines of the contribution that is to be made to the grid The pinhole position relative to the grid is also used to define the element of the grid into which this contribution is scored Once the direction is established a ray trace contribution is made to the grid bin with attenuation being determined for the material regions along that path The source need not be within the object being imaged nor does it need to produce the same type of particles that the detector grid has been programmed to score The grid and pinhole will image either source or scattered events produced within the object see NOTRN card in Section 5 7 20 3 for either photons or neutrons These event type contributions can be binned within the grid tallies by binning as source only total or by using special binning relative to the number of collisions contributing cells etc The pinhole image projection is set up as follows Pin P X1 Y1 Z1 RO X2 Y2 Z2 F1 F2 F3 n is the tally number and must be a multiple of 5 since this i
254. dows platform this distribution is not the correct one for your needs Please request a separate Windows distribution Until an automated build system for Windows is created binary images will be distributed 3 1 2 Automated Building The process used when building mcnpx varies greatly depending upon the following e hardware platform e g SPARC ALPHA 1386 operating system e g Solaris Linux HP UX e available compilers e g f77 cc g77 gcc pgf77 gcc mcnpx program options e g the default path of cross sections and other data files In past versions of MCNPX coping with this complex set of build options required a top level Makefile that determined the architecture and propagated these decisions to lower level Makefiles It was not possible to go to some lower level makefile Build Ics Build mcenpf and do a make It was also difficult to cope with different user level options such as the desire to include mesh tallies or to exclude mesh tallies or to compile with or without debugging 14 MCNPX User s Manual MCNPX User s Manual a Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium A problem in the MCNPX 2 1 series with the various locations of X libraries on different systems added to the desire for a more complete and dynamic build system As more plat forms operating systems options and compilers were added the complexity skyrocketed Users of MCNPX have had to install the co
255. e The FATAL option on the MCNPX execution line instructs MCNPxX to ignore fatal errors and run particles but the user should be extremely cautious about doing this Most MCNPX error messages are warnings and are not fatal The user should not ignore these messages but should understand their significance before making important calculations In addition to FATAL and WARNING messages MCNPX issues BAD TROUBLE messages immediately before any impending catastrophe such as a divide by zero which would otherwise cause the program to crash MCNPX terminates as soon as the BAD TROUBLE message is issued User input errors in the INP file are the most common reason for issuing a BAD TROUBLE message These error messages indicate what corrective action is required 4 3 GEOMETRY ERRORS There is one important kind of input error that MCNPX will not detect while processing data from the INP file MCNPX cannot detect overlapping cells or gaps between cells until a particle track actually gets lost Even then the precise nature of the error may remain unclear However there is much that you can and should do to check your geometry before starting a long computer run MCNPX User s Manual 41 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Use the geometry plotting feature of MCNPX to look at the system from several directions and at various scales Be sure that what you see is what you intend Any gaps or overlaps in the geo
256. e products of compiling and building More complex packages The GNU C compiler suite gcc comes to mind warn that the simple build procedure given above is a dangerous practice as it clutters the original source tree with generated Makefiles and compiled objects and makes it difficult to sup port multiple builds with different options They suggest using a different initially empty directory to be the target of the configure process gzip dc PACKAGE tar gz tar xf mkdir Build cd Build PATH_OF_PACKAGE SOURCE configure make install The MCNPX team also makes this suggestion Please use an empty directory somewhere other than the source distribution s location as the target of the build It keeps the source tree clean and allows multiple builds with different options Even if you think that you will never need additional builds it costs nothing to have the flexibility in the future 3 1 3 MCNPX Build Examples We will illustrate the new configure and make procedure with two primary examples A sys tem manager installing the MCNPX release for a system with several users and an individual user installing the MCNPX release for their own use A few variations on these themes are given 3 1 3 1 System Wide Installation For purposes of the first illustration we will assume that the MCNPX 2 3 distribution has been unloaded from cdrom or fetched from the net and is in the file usr local src menpx_2 3 0 tar gz The system manager lo
257. e ForNTYPE gt 0 a record containing NTYPE particle types in any order defined as the array ITIP I I l LNTYPE In the present MCNPxX the contents of a surface source file WSSA are insufficient to distinguish between a particle and its antiparticle it is to be expected that this condition will be remedied in future releases of MCNPX For NPARM gt 0 a record containing NPARM user defined cell material or surface numbers integers in any order for which one wishes a tally to be made these are defined as the array LPARM l I 1 NPARM If a null record is supplied with NPARM gt 0 it is treated as 1 2 3 NPARM Note a different meaning for NPARM is used for IOPT 13 MCNPX User s Manual 211 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 For NFPRM gt 0 a record containing NFPRM upper cosine bin boundaries defined as the array FPARM I I 1 NFPRM The first lower cosine boundary is always 1 0 If a null record is supplied equal cosine bin boundaries from 1 0 to 1 0 will be defined by default e If NPARM is preceded by a minus sign a record containing NPARM or NPARM 1 nor malization divisors these are defined in HTAPE3X as the DNPARM array The NPARM values are in a one to one correspondence with the LPARM array The last NPARM 1 entry applies to a total over the NPARM entities where applicable if omit ted it defaults to 1 0 Through this feature it is possible to input a list of vo
258. e VOV can be printed for all bins in a tally by using the DBCN card 5 7 1 Fna Tally Seven basic neutron tally types six basic photon tally types and four basic electron tally types are available in MCNP as standard tallies All are normalized to be per source particle unless changed by the user with a TALLYX subroutine or normed by weight in a criticality KCODE calculation Mnemonic Tally Description Fn units Fn units F1 N or F1 P or F1 E Current integrated over a surface particles MeV F2 N or F2 P or F2 E Flux averaged over a surface particles cm MeV cm 112 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 F4 N or F4 P or F4 E Flux averaged over a cell particles em MeV cm F5a N or F5a P Flux at a point or ring detector particles cm MeV cm F6 N or F6 N P or F6 P Energy deposition averaged over MeV g jerks g a cell F7 N Fission energy deposition MeV g jerks g averaged over a cell F8 P or F8 E or F8 P E Energy distribution of pulses pulses MeV created in a detector F8 E Charge deposition charge N A The tallies are identified by tally type and particle type as follows Tallies are given the numbers 1 2 4 5 6 7 8 or increments of 10 thereof and are given the particle designator N P or E or N P only in the case of tally type 6 or P E only in the case of tally type 8 Thus you may have as many of any basic tally as you need each with different energy bins or flaggin
259. e effects of charged particle scat tering in these semi deterministic methods We have begun research to solve this long standing problem and will implement solutions in upcoming versions of the code 7 Certain Weight Window optimizations have not been fully implemented for high energy particles 8 The Mix and Match feature has yet to be implemented MCNPX version 2 3 0 will not switch between table based and physics based data where a number of tables with differing upper energies are present The switch between physics models and tabular data is made at one energy for all materials in the problem This energy is set on the PHYS card by the user see section 6 1 7 Therefore it is desirable that one use a Set of libraries all with the same upper energy limits Correctly implementing this feature involves a major rewrite of data structures in MCNPX and will be released in a future version 9 Charged particle reaction products are not included for some neutron reac tions below 20 MeV in the LA150N library In calculating total particle production cross sections the library processing routines include only those reactions where complete angular and energy information is given for secondary products The new 150 MeV evaluations are built on top of existing ENDF and JENDL evaluations which typically go to 20 MeV Although the 150 MeV evaluations do include the detailed sec ondary information in the 20 150 MeV range the lt 20 MeV data
260. e error only one material should be defined Note with N1COL 1 MCNPX will override the source specification and construct the source as a pencil beam in the z direction as required by XSEX3 Other MCNPX options may be used to suppress either nuclear elastic or nonelastic reactions 1 To create a HISTP file to be edited by XSEX3 include a HISTP card in the INP file 2 Define a volume parallel beam source in the z direction vec 0 0 1 which is com pletely contained inside a cell with the material for which the cross sections are to be calculated 3 Specify the incident particle type and kinetic energy on the SDEF card 4 Use NOACT 1 the 8th parameter on the LCA card The user may wish to suppress nuclear elastic scattering in the calculation by using IELAS 0 on the LCA card An AWTAB card may need to be supplied if the target isotope has no mass in XSDIR the value supplied is not used and is arbitrary As an example the following is a sample MCNPX input for a cross section calculation MCNPX User s Manual 225 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 MCNPxX standard cross section generation format for XSEX3 use c 1000 MeV protons on Sn121 an isotope not in MCNPX library c and for which no atomic weight is specified in XSDIR c Minimal geometric specification for this purpose 1 1 1 0 1 2 0 1 m1 501211 not in MCNPX libraries awtab 50121 119 864 need value but arbitrar
261. e fundamentals of intermediate and high energy physics are likely unfamiliar to most traditional users of MCNP This chapter gives a brief explanation of the physics options offered by MCNPX in these energy regions and also describes improvements in nuclear data libraries at lower energies An excellent discussion of these concepts is given in FER98 4 1 Intermediate Interaction Physics This section gives a brief overview of the basic elements common to most intermediate energy Monte Carlo physics packages MCNPX offers options based on three physics packages the Bertini and ISABEL models taken from the LAHET Code System and the CEM package which has been specially adapted by the author for the MCNPX work Below we describe the standard components of a physics based package in the energy MCNPX User s Manual 37 MCNPX User s Manual Version 2 3 0 April 2002 E LA UR 02 2607 Accelerator Production of Tritium regime of 150 MeV to a few GeV In future versions of MCNPX it will be possible to run the code with any combination of these options however in version 2 3 0 the components belonging to the three packages should be kept intact Figure 4 1 illustrates the major elements in pictorial form In the first stage a particle inci dent on a nucleus interacts with individual nucleons via particle particle cross sections in a potential which describes the density of the nucleus as a function of radius Intranuclear Cascade INC and
262. e output of MCNPX The results of these activi ties will be published separately and the code development team will strive to make available results from other projects We also solicit your input for potential code 2 1 Warnings and Known Bugs 1 Parallel processing in MCNPX version 2 3 0 has yet to be extended to all high energy code additions See Section 3 1 6 for further discussion 2 Pertubation methods used in MCNP have not yet been extended to the non tab ular models present in MCNPX In MCNPX version 2 3 0 there is a bug that can cause the code to crash if run for problems that invoke the pertubation capabilities of MCNPX4B This will be fixed in a future version 3 Not all plotting features have been verified for all possible outputs Since no changes have been made in geometry features the geometry plotting code works well However we have not yet been able to check out all the many features of mcplot The user should do reasonableness checks when using this feature For example cross section plotting for tables other than neutrons photons and electrons is not yet implemented MCNPX User s Manual 5 MCNPX User s Manual E Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium 4 KCODE criticality calculations have not been extended to include high energy neutrons Accelerator Transmutation applications should keep criticality limitations in mind when using this feature to include high energy neutro
263. e surfaces 1 3 and 6 and one which is the average of the flux across all three of the surfaces Example 2 F1 P 1 2 3 45 6 This card provides three photon current tallies one for the sum over surfaces 1 and 2 one for the sum over surfaces 3 4 and 5 and one for surface 6 alone Example 3 F371 N 123 14 7 This card provides three neutron current tallies one for the sum over surfaces 1 2 and 3 one for the sum over surfaces 1 and 4 and one for the sum over surfaces 1 2 3 and 4 The point of this example is that the 7 bin is not confused by the repetition of surface 1 5 7 1 2 Repeated Structures Tallies Simple Form Fn pl S4 Sk General Form Fn pls S2 ees S3 S4 S5 lt Ci Cal ees I lt U lt C3 C4 Cs more bi Table 5 55 Repeated Structure Tallies Variable Description n tally number pl particle designator Sj problem number of a surface or cell for tallying 116 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 55 Repeated Structure Tallies Variable Description Ci problem number of a cell filled with a universe T Total over specified surfaces or cells U problem number of a universe used on a fill card index data for a lattice cell element with three possible formats always in brackets I Indicating the 744 lattice element of cell Cs as defined by the FILL array l I I T3 I4 Is
264. e utility that is provided by the vendor On UNIX and Linux you must use the GNU make utility and it must be version 3 76 or later Sometimes the GNU make utility is installed in an executable file called gmake Sometimes system administrators make symbolic links called make that when resolved invoke the gmake utility You can make your own symbolic links in directories that you own and control so that when you execute the make command you will be executing the make you intend to use You can also establish an alias in the shell runtime control file whereby any make command you issue actually executes gmake You can also substitute the gmake com mand everywhere you see the make command in the examples that follow The important point of this discussion is to know your make and use the right one oth erwise this automated build system can fail If no make or gmake is found you either have a PATH value problem or you need some help from your system administrator to install GNU make If both make and gmake exist query each of them to see what version you have make v gmake v Some vendor supplied make utilities do not understand the v option that requests that the version number be printed If you see an error or usage message then your make is one of the vendor supplied variety Make sure you have GNU make version 3 76 or later installed and that it is found in your search path first If you work on a Win
265. eV For pions the Bertini INC model will be used below this value FLENB4 Kinetic Energy Default 2500 MeV For pions the FLUKA high energy generator will be used above this value See Notes under FLENB2 FLENB5 Kinetic Energy Default 800 MeV For nucleons the ISABEL INC model will be used below this value FLENB6 Kinetic Energy Default 800 MeV For nucleons an appropriate model will be used above this value for IEXISA 2 it applies to all particle types for IEXISA 1 it applies to all particles except nucleons and pions for IEXISA 0 it is immaterial See the example following this table for further explanation CTOFE The cutoff kinetic energy MeV for particle escape during the INC when using the Bertini model The cutoff energy prevents low energy nucleons from escaping the nucleus during the INC for protons the actual cutoff is the max imum of CTOFE and a Coulomb barrier CTOFE gt 0 CTOFE will be used as the cutoff energy CTOFE lt 0 a random cutoff energy uniformly distributed from zero to twice the mean binding energy of a nucleon will be sampled for each projectile target interaction and separately for neutrons and protons In this case the Coulomb barrier for protons is also randomized The randomized cutoff energy is the default CTOFE 1 0 For the ISABEL INC the randomized cutoff energy is always used FLIMO The maximum correction allowed for mass energy balancing in the cascade stage used w
266. each of several multiline cell parameter cards For source distributions corresponding SI SP and SB values are side by side Source options other than defaults are on the next line and must all be entered explicitly The amp continuation symbol is not needed and if present is ignored In column format card names are put side by side on one input line and the data values are listed in columns under the card names A is put somewhere in columns 1 5 on the line with the card names The card names must be all cell parameters all surface parameters or all something else If a card name appears on a card there must not be a regular horizontal card by that name in the same input file If there are more entries on data value lines than card names on the line the first data entry is a cell or surface 36 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 number If any cell names are entered all must be entered If cell names are entered the cells don t have to be in the same order as they are in the cell cards block If cell names are omitted the default order is the order of the cells in the cell card block The same rules apply to surface parameters but because we presently have only one surface parameter AREA column input of surface parameters is less useful There can be more than one block of column data in an input file Typically there would be one block for cell parameters and one
267. eaeeeeeeeaaas 82 LEB Keyword Descriptions 0 eececceeeeeneeeeeeenceeeeeeeaeeeeeeeaaaeeeeeeeaaeeeeeeeeaeeeeeeenaas 83 Secondary Particle Biasing Argument Descriptions cceceeseeeeeteeeeetteeeees 90 Track Averaged Mesh Tally type 1 Keyword Descriptions eeeeee 95 Source Mesh Tally type 2 Keyword Descriptions 0 ccccceeeesteeeeeeessteeeeeeees 97 Energy Deposition Mesh Tally type 3 Keyword Descriptions cceee 98 DXTRAN Mesh Tally type 4 Keyword Descriptions 0 eceeeeeeenteeeeeenaes 100 Pinhole Radiography Argument Descriptions ceeeeseeceeeeeeeeeeeeeenaeeeeeeeenaes 103 Transmitted Image Projection Argument Description 0 0 0 eeeeeeeeesteeeeeeeeees 105 NPS Keyword Descriptions ccceceeeeeeeeeeeneeeeeeeeesaeeeeeneeesaeeeseaaeeseeeeetaeessaes 106 Energy Deposition Card Argument Descriptions ccccceseeeeeeereeeeteeeeeeees 112 DFACT Argument Descriptions merianiae tarasie aaa aaiae ia aeie 114 Neutron Problem Summaries cccceecceeeeeeeeeeeeeeceaeeeeeneeeseaeeeeeaaeeseaeeeetaeeesaes 126 Results Compiled for Summary CaS S ccccecceeeeeeseeeeeeeeeeseaeeeeeeeeteaeeeennees 133 Applicability of Input Control Parameters ccccceceeeeeeceeeeeeeeeeeeeeeeeseaeeteneees 136 Applicability of Minus Sign Flags on Input Control Parameters 0 c 137 Particle Type Identification in HTAPE3X o ceeeeeseeeeeeeneeeeeeeenaeeee
268. eans determine the 150 MeV cross section value Note that one can specific different values for CUT_N and CUT_H For example specify ing CUT_H 0 will tell the code not to use any proton libraries only physics models PHYS n EMAX EMCNF CUT_N PHYS h EMAX EMCNF CUT_H j ISTRAG PHYS e EMAX IDES IPHOT IBAD ISTRG BNUM XNUM RNOK ENUM see MCNP4B manual for electron definitions PHYS all other charged particles EMAX j j j ISTRAG Table 6 1 Setting upper limits for neutron amp proton tabular data Keyword Description EMAX Upper limit for neutron or proton energy MeV EMCNF Energy boundary MeV above which neutrons are treated with implicit capture and below which they are treated with analog capture This variable is not read in for protons CUT_N Energy MeV below which table based data are used and CUT_H above which physics modules are used Neutron default is 20 0 MeV proton default is 0 0 MeV unused ISTRG 0 improved approach to Vavilov straggling default 1 continuous slowing down approximation 1 old Vavilov treatment from 2 1 5 ISTRG was placed in the 5th position of the PHYS card for heavy charged particles in order to be consistent with the cor responding entry on a PHYS e card ISTRG is not used for neutrons Photons After the maximum energies for all other particles have been set photons are considered If photons are being transported a photon maximum energy is set as the low
269. econdary particle 62 MCNPX User s Manual MCNPX User s Manual i Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium is determined by momentum conservation This angular deflection is used for the subse quent transport of the secondary electron However neither the energy nor the direction of the primary electron is altered by the sampling of the secondary particle On the aver age both the energy loss and the angular deflection of the primary electron have been taken into account by the multiple scattering theories Note the concept of knock on electrons from heavy charged particles is valid however is not implemented in MCNPX version 2 3 0 MCNPX User s Manual 63 Accelerator Production of Tritium 64 MCNPX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 MCNPX User s Manual MCNPX User s Manual i Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium 5 Multiparticle Extensions and General Tracking MCNPX has expanded the capability of MCNP4B to track 34 particles although in version 2 3 0 not all are fully transported Those which are not transported typically have very short halflives and are decayed immediately upon production these are marked by a in the mean lifetime column of Table 5 1 Decay of secondary particles continues until a set of transportable particles is obtained Table 5 1 lists all particles currently defined in MCNPxX version
270. ectors or DXTRAN MCNPX will also require a SRCDX routine See Appendix 5 6 7 Extended Source Options MCNPX extends the MCNP standard source SDEF in several ways which are now summarized 1 Spontaneous fission PAR SF 2 Character particle types PAR h is equivalent to PAR 9 3 The gaussian distribution source function 41 may be used for more than time SPn 41ab See the example below for specifying an accelerator beam source 4 Surface transformations and distributions of surface transformations are allowed SDEF TR n_ or SDEF TR Dn The transformation is applied to the particle after its coordinates and direction cosines have been determined See the example below for specifying a accelerator beam source An additional feature has been added through the specification of a general transformation on the SDEF card in one of two forms TR n or TR Dn In either case a general transformation is applied to a source particle after its coordinates and direction cosines have been determined using the other parameters on the SDEF card Particle coordinates are modified by both rotation and translation direction cosines are modified by rotation only This allows the user to rotate the direction of the beam or move the entire beam of particles in space The TR Dn card is particularly powerful since it allows the specification of more than one beam ata time An example of specifying a Gaussian beam MCNPX User s Manual 107 MCNPX
271. ectory prefix is usr local usr local src mcnpx_2 4 0 configure now make the executable mcnpx program and supporting LCS libraries make all run the regression tests for your architecture make tests install the executables and libraries in usr local make install clean up The build products are no longer needed cd tmp rm rf mcnpx 3 1 3 2 System Wide Installation With Existing Directories The previous example might typically be used when a new installation of MCNPX is performed on a system that has no pre existing mcnpx with which to be compatible If a user already has mcnpx then it may be desired to use the existing locations for the data files and cross sections Two options to the configure process can be used to customize the locations where mcnpx and its data will be installed and the default locations where MCNPX will find those files When the user wants to use the normal mcnpx directory layout of bin for executables and lib for data files MCNPX User s Manual 13 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 but does not wish to use the default directory usr local then the previous example can be adjusted with additional options In the previous example the configure script could be given the option usr local src mcnpx_2 4 0 configure prefix usr mcnpx and the make install process would install the mcnpx binary in usr mcnpx bin and the data files in usr mcnpx lib
272. ectron Collisional Stopping Power Berger BER63 gives the restricted electron collisional stopping power i e the energy loss per unit path length to collisions resulting in fractional energy transfers less than an arbitrary maximum value m in the form 56 MCNPX User s Manual MCNPX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium dE ae 5 NZC In E Nf r em 5 where 2 F 2 2 Pesje ap eae tel T Here and represent energy transfers as fractions of the electron kinetic energy E l is the mean ionization potential in the same units as E B is v c t is the electron kinetic energy in units of the electron rest mass is the density effect correction related to the polarization of the medium Z is the average atomic number of the medium N is the atom density of the medium in cm and the coefficient C is given by Ea 2ne 2 mv where m e and v are the rest mass charge and speed of the electron respectively The ETRAN codes and MCNP MCNPX do not make use of restricted stopping powers but rather treat all collisional events in an uncorrelated probabilistic way Thus only the total energy loss to collisions is needed and the above equations can be evaluated for the spe cial value of 1 2 The reason for the 1 2 is the indistinguishability of the two outgoing electrons The electron with the larger energy is by definition the p
273. eeenaeeeeeeeeaaes 138 Order of HTAPESX Input RecordS eccceeeeceeeeeeeeeneeeeeeeeeeeaaeeeeaeeeseaeeeeneeeees 141 MCNPX User s Manual xi MCNPX User s Manual Version 2 3 0 April 2002 F LA UR 02 2607 Accelerator Production of Tritium xii MCNPX User s Manual Zz MCNPX User s Manual Accelerator Version 2 3 0 April 2002 Erocmeton LA U R 02 2607 Preface Work on the MCNPX code has been sponsored by both the Accelerator Production of Tritium APT and Advanced Accelerator Applications AAA projects in response to requests from the facility designers Originally MCNPX was one part of the APT effort to provide a validated set of computer simulation tools to use in design of the APT spallation target surrounding lead blanket and associated shielding Other elements of this program included the production of new nuclear data evaluations from 20 to 150 MeV for neutrons and from 1 to 150 MeV for proton and photonuclear interactions Additional work was undertaken to provide improved total reaction and elastic cross section tables above 150 MeV and to improve the physics involved with the intermediate and high energy physics models through the CEM program Currently the requirements of the Accelerator Transmu tation of Waste program which is part of AAA are directed toward improvements in fission physics and actinide data Responsibility for the development of MCNPX was given to the APT Target Blanket and Materi
274. efault PHYS n 100 00 1 2000 Example PHYS n 800 100 3 20 1 1 5 5 2 2 Photons Form PHYS p EMCPF IDES NOCOH PNB Table 5 32 Photon Physics Options Keyword Description EMCPF Upper energy limit in MeV for detailed photon phys ics treatment IDES 0 photons will produce electrons in MODE E prob lems or bremsstrahlung photons with the thick target bremsstrahlung model 1 photons will not produce electrons as above NOCOH 0 coherent scattering occurs 1 coherent scattering will not occur MCNPX User s Manual 83 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 32 Photon Physics Options Keyword Description PNB 1 Analog photonuclear particle production 0 No photonuclear particle production 1 Biased phtonuclear particle production Default PHYS p 100000 Use Optional 5 5 2 3 Electrons Form PHYS E EMAX IDES IPHOT IBAD ISTRG Table 5 33 Electron Physics Options Keyword Description EMAX upper limit for electron energy in MeV IDES 0 1 photons will will not produce electrons IPHOT 0 1 electrons will will not produce photons 0 full bremsstrahlung tabular angular distribution IBAD 1 simple bremsstrahlung angular distribution approxima tion ISTRG 0 sampled straggling for electron energy loss 1 expected value straggling for electron energy loss lt 0 only applicable for el03 evaluation See below for
275. eger Array 0 20 cece 173 5 9 9 RDUM Floating Point Array 20 2 0 c eee eee eee 173 5 9 10 FILES File Creation 00 00sec eee eee 173 5 10 SUMMARY OF MCNPX INPUT CARDS 20 20 c eee eee eee 174 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 6 References 2 2 2 cis oie etec aed ee tee ew ee Cees eee eee ee 181 Appendix A Examples oi cos riny tma ered Cee eae ened Cee es 191 Appendix B HTAPESX for use with MCNPX 0000e 205 Appendix C Using XSEX3 with MCNPX 0000 eee eee 225 MCNPX User s Manual xi MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 xii MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Preface Work on the MCNPX code has been primarily sponsored by both the Accelerator Pro duction of Tritium APT and Advanced Accelerator Applications AAA projects in response to requests from the facility designers Originally MCNPX was one part of the APT effort to provide a validated set of computer simulation tools to use in design of the APT spallation target surrounding lead blanket and associated shielding Other elements of this program included the production of new nuclear data evaluations from 20 to 150 MeV for neutrons and from 1 to 150 MeV for proton and photonuclear interactions Addi tional work was undertaken to provide improved total reaction and elas
276. egmenting For basic option types 9 10 or 12 it is the collimating window definition Also for basic option types 1 9 11 or 12 an arbitrary vector for angular binning may be input 3 Edit Option IOPT 1 or 101 Surface Current Option 1 tallies the particle current across the NPARM designated surfaces it is analogous to the MCNP F1 tally If IOPT is preceded by a minus sign the weight binned 212 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 is multiplied by the particle energy The number of energy bins is given by NERG The number of particle types for which surface crossing data is to be tallied is given by NTYPE and must be gt 0 Current will be tallied on NPARM surfaces a total over surfaces is not performed Any of the above particle types may be specified Binning into NFPRM cosine bins is defined by the value of KOPT For KOPT 0 or 5 the cosine is taken with respect to the normal to the surface at the crossing point For KOPT 1 or 6 the cosine is taken with respect to the x axis For KOPT 2 or 7 the cosine is taken with respect to the y axis For KOPT 3 or 8 the cosine is taken with respect to the z axis For KOPT 4 or 9 the cosine is taken with respect to an arbitrary vector to be read in If KOPT 5 6 7 8 or 9 the current tallies are binned according to a slicing of each surface into NSEG 1 segments by NSEG planes In this case all additional record of the
277. ells Use Required for lattices Example 1 0 20 fill 4 2 0 30 u 1 fill 2 lat 1 3 0 11 2 4 0 11 u 2 5 0 20 20 rpp 0 50 10 10 5 5 30 rpp O0 10 0 10 11 s 5 5 0 4 Cell 2 is the base 0 0 0 element of a square lattice described by surface 30 a right parallelepiped with Xmin 0 Xmax 10 Ymin 0 Ymax 0 and infinite in the Z direction It is filled with Universe 2 cells 3 amp 4 and it is assigned to universe 1 which fills and is bounded by cell 1 an RPP with Xmin 0 Xmax 50 Ymin 10 Ymax 10 Zmin 5 and Zmax 5 In this case the lattice elements i j k would be 0 4 1 0 and 0 0 5 3 3 7 TRn Coordinate Transformation Form TRn 010203 XX YX ZX XY YY ZY XZ YZ ZZ M Table 5 25 Coordinate Transformation Card Statement Description number of the transformation J lt n lt 999 TRn means that the B are angles in degrees rather than being the cosines of the angles O O O3 displacement vector of the transformation B to Bg rotation matrix of the transformation MCNPX User s Manual 73 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 25 Coordinate Transformation Card Statement Description 1 the default means that the displacement vector is the location of the origin of the auxiliary coordi nate system defined in the main system 1 means that the displacement vector is the loca tion
278. empty working space cd mcnpx execute the configure script the prefix tells where to put the executables and libraries mcnpx_2 3 0 configure prefix home me now make the executable mcnpx program and the bertin and pht libraries run the tests and install in home me bin and home me lib make all tests install 3 1 3 5 Individual Private Installation special compilers and debugging As a final example suppose you want basically the same thing as the previous example but you would like to have the debug option turned on during compilation The compiled code will go into a private local library nhome me bin but you wish to use the cross section files and LCS data files already on your system We will assume that these data files already exist in the directory usr mcnpx data We will assume that the source distribution has already been unpacked by a system administrator into usr local src mcnpx_2 3 0 To add a bit more complexity assume for this example that we are building and running on a Sun Solaris system that has both the GNU g77 Fortran compiler and the vendor s commercial Fortran and C compilers installed Systems such as Sun s Solaris and HP s HP UX normally do not include development compilers These compilers are usually pur chased as additional items Versions of the GNU compilers are available on the net for such systems Thus such systems may have the GNU compilers the Vendor s commer cial compilers or
279. en atom fraction of material 1 RHO 0 05 x 1 02 1 0 0 051 The effect of this perturbation on tally 14 which is a track length estimate of ky will be provided as a differential change PERT 1 as well as with this change added to the unperturbed estimate of kez PERT2 Note if the RHO keyword is omitted from the PERT cards the 2 5U composition will be perturbed which can produce invalid results see Caution 4 Example 7 1 1 1 5 1 2 3 4 5 6 mat1 at 1 5 g cm M1 1001 4333 6000 2000 8016 3667 half water half plastic M2 1001 6666 8016 3334 water M3 1001 2000 6000 4000 8016 4000 plastic PERT1 n CELL 1 MAT 2 RHO 1 0 METHOD 1 PERT2 n CELL 1 MAT 3 RHO 2 0 METHOD 1 This example demonstrates how to make significant composition changes e g changing a region from water to plastic The unperturbed material is made from a combination of the two desired materials typically half of each PERT1 gives the predicted tally as if cell 1 were filled with water and PERT2 gives the predicted tally as if cell 1 were filled with plastic The difference between these perturbation tallies is an estimate of the effect of changing cell 1 from water to plastic 5 7 22 TMESH_ The Mesh Tally The Mesh Tally is a method of graphically displaying particle flux dose or other quantities on a rectangular cylindrical or spherical grid overlaid on top of the standard problem geometry Particles are tracked through the indepe
280. ence coordinates that establish the reference direction cosines for the outward normal to the detector grid plane as from X2 Y2 Z2 to X1 Y1 Z1 This is used as the outward normal to the detector grid plane for the TIR case and as the centerline of the cylinder for the TIC case F1 e F1 0 Both the source and scattered contributions will be scored at the grid F1 lt 0 Only the scatter contributions will be scored F1 gt 0 is not allowed in this application F2 F2 must be less than 0 to turn on this type of image application in 2 1 5 This restriction has been removed in 2 3 0 Do not make F2 0 as this will result in a fatal error plane grid case Radial restriction relative to the center of the grid for contributions to be made It defines a radial field of view on the grid cylindrical case Radius of the cylinder on which the grid is to be established F3 F3 0 All contributions are directed to the center of each grid bin F3 lt 0 Contributions are made with a random offset from the center of the grid bin This offset remains fixed and is used as the offset for contri butions toll of the grid bins for this event The grid itself is established with the use of FSn and Cn cards in the same manner as described for the pinhole case in Section 8 2 1 However X1 Y1 Z1 are now the coordi nates of the intersection of the reference direction and the grid plane as shown in Fig 8 3 In the cylindrical grid
281. enomena it is not currently implemented in MCNPX version 2 3 0 In intermediate energy physics applications this source is small however the user should be warned that at very high energies it could become a non negligible component Knock on Electrons The Moller cross section for scattering of an electron by an electron is P eee ET 1 8 do _ C 1 EEI E 2 2t 1 1 de a fe D where is the energy transfer as a fraction of electron kinetic energy E and tis the electron kinetic energy in units of the electron rest mass When sampling for transportable second ary particles one wants the probability of energy transfers greater than some cutoff energy below which particles will not be followed This probability can be written 1 2 do o T E The reason for the upper limit of 1 2 is the same as that given for collisional stopping power Explicit integration of this equation gives ate FE 1 at T Wee eee Eo Lees t 1 2 Gabe Eo Then the normalized probability distribution for the generation of secondary electrons with E gt E IS given by 1_do o e de g e de At each electron substep MCNP MCNPX uses o to determine randomly whether knock on electrons will be generated If so the distribution of o will be used to sample the energy of each secondary electron Once an energy has been sampled the angle between the primary direction and the direction of the newly generated s
282. ensity formulation See the LEB card for details on parameter inputs 0 Use Gilbert Cameron Cook Ignatyuk level density model PRA88 default 1 Use the Julich level density parameterization as a function of mass number CLO83 IEVAP 0 The RAL evaporation fission model ATC80 will be used default 1 The ORNL evaporation fission model BAR81 will be used Note The ORNL model allows fission only for isotopes with Z 91 NOFIS 1 Allow fission default 0 Suppress fission 5 5 7 4 LEB Form LEB YZERE BZERE YZERO BZERO This card controls level density input options for the original HETC implementation Table 5 44 LEB Keyword Descriptions Keyword Description YZERE The YO parameter in the level density formula for Z lt 70 The default is 1 5 zero or negative is an error condition For target nuclei with Z lt 70 the parameters BZERE and YZERE are used to compute level densities the default values are those used in LAHET before installation of the ORNL fission model For target nuclei with Z 71 the BZERO and YZERO parameters are used to compute level densities for the target nucleus and fission fragments Note Applies only for ILVDEN 1 BZERE The BO parameter level density formula for Z lt 70 The default is 8 0 zero or negative is an error condition see YZERE above Note Applies only for ILVDEN 1 YZERO The YO parameter in the level density formula for Z 71 an
283. ent for secondary generation lt 0 MCNP4A treatment of electron angles at secondary generation sites MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 110 Debug Information Card Variable Description X 0 default MCNP style energy indexing algorithm a 1 ITS style energy indexing algorithm X20 track previous version Use Optional 5 9 7 LOST Lost Particle Form LOST LOST 1 LOST 2 Table 5 111 Lost Particle Card Variable Description number of particles which can be lost before the job LOST 1 terminates with BAD TROUBLE maximum number of debug prints that will be made for Porte lost particles Defaults 10 lost particles and 10 debug prints Use Discouraged Losing more than 10 particles is rarely justifiable 5 9 8 IDUM Integer Array Form IDUM 1 J 1sns50 Default All array values zero Use Useful only in user modified versions of MCNP 5 9 9 RDUM Floating Point Array Form RDUM R R 1sns50 Default All array values zero Use Useful only in user modified versions of MCNP Entries up to 50 fill the RDUM array with floating point numbers 5 9 10 FILES File Creation Form FILES unit no filename access form record length MCNPX User s Manual 173 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 112 File
284. er s terminal has no graphics capability Use file aaaa as the source of plot requests When an EOF is read control is transferred to the terminal In a produc tion or batch situation end the file with an END command COM aaaa to prevent transfer of control Never end the COM file with a blank line If COM is absent the terminal is used as the source of plot requests Read file aaaa as the source of MCNP tally data The default is RUNTPE if it exists If the default RUNTPE file does not exist the user will be prompted for an RMCTAL or RUNTPE command RUNTPE aaaa Name the graphics metafile aaaa The default name is PLOTM For some systems this metafile is a stan PLOTM aaaa dard postscript file and is named PLOTM PS When CGS is being used there can be no more than six characters in aaaa Write all plot requests to file aaaa The default name is COMOUT MCPLOT writes the COMOUT file in order to give the user the opportunity to do the same plotting at some later time using all or part of the old COMOUT file as the COM file in the second run Unique names for the output files PLOTM and COMOUT will be chosen by MCNPX to avoid over writing existing files COMOUT aaaa Plot requests are normally entered from the keyboard of a terminal but alternatively can be entered from a file A plot is requested by entering a sequence of plot commands following a prompt character The request is terminated by a carriage
285. er EVAP A Fortran Program for Calculating the Evaporation of Vari ous Particles from Excited Compound Nuclei Oak Ridge National Laboratory Report ORNL TM 7882 July 1981 EVA55 R D Evans The Atomic Nucleus Robert E Krieger Publishing Co 1955 FAS94a A Fasso A Ferrari J Ranft P R Sala G R Stevenson and J M Zazula FLUKA92 Proceedings of the Workshop on Simulating Accelerator Radiation Environ ments SARE1 Santa Fe New Mexico January 11 15 1993 A Palounek ed Los Alamos LA 12835 C p 134 144 1994 FAS94b A Fasso A Ferrari J Ranft and P R Sala FLUKA Present Status and Future Developments Proceedings of the IV International Conference on Calorimetry in High Energy Physics La Biodola Elba September 19 25 1993 A Menzione and A Scribano eds World Scientific P 394 502 1994 FAS97 A Fasso A Ferrari J Ranft and P R Sala An Update about FLUKA Pro ceedings of the 2nd Workshop on Simulating Accelerator Radiation Environments SARE2 CERN Geneva October 9 11 1995 CERN Divisional Report CERN TIS RP 97 05 p 158 170 1997 FAV99 J A Favorite and K Adams Tracking Charged Particles Through a Magnetic Field Using MCNPX U X Division Research Note XCI RN U 99 002 February 5 1999 FER98 A Ferrari and P R Sala The Physics of High Energy Reactions Lecture given at the Workshop on Nuclear Reaction Data and Nuclear Reactors Physics Design MC
286. er or otherwise does not necessarily constitute or imply its endorsement recommendation or favoring by the United States Government the Department of Energy UT BATTELLE LLC nor any person acting on behalf of the Department of Energy or UT BATTELLE LLC Distribution Notice This code data package is a part of the collections of the Radiation Safety Information Computational Center RSICC developed by various government and private organizations and contributed to RSICC for distribution Any further distribution by any holder unless otherwise specifically provided for is prohibited by the U S Department of Energy without the approval of RSICC P O Box 2008 Oak Ridge TN 37831 6362 Documentation for CCC 715 MCNPX 2 4 0 Code Package PAGE RSICC Computer Code Abstract 2 nee eens ii MCNPX User s Manual Version 2 4 0 LA CP 02 408 September 2002 Section 1 L S Waters ed MCNPX User s Manual Version 2 3 0 LA UR 02 2607 April 2002 Section 2 September 2002 RSICC CODE PACKAGE CCC 715 1 NAME AND TITLE MCNPX Version 2 4 0 Monte Carlo N Particle Transport Code System for Multiparticle and High Energy Applications AUXILIARY PROGRAMS GRIDCONV Converts output of mesh and radiography tallies to input for external graphics programs HTAPE3X Postprocessor for MCNPX HISTP output MAKXSF Prepares MCNPX Cross Section Libraries HCNV and TRX Convert LAHET ASCII data to binary XSE
287. er 1983 MOL48 G Moliere Theorie der Streuung schneller geladener Teilchen II Mehrfa chund Vielfachstreuung Z Naturforsch 3a 1948 78 MOT29 N F Mott The Scattering of Fast Electrons by Atomic Nuclei Proc Roy Soc London A125 1929 425 MCNPX User s Manual 121 MCNPX User s Manual Ap Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium PRA88 R E Prael and M Bozoian Adaptation of the Multistage Pre equilibrium Model for the Monte Carlo Method I Los Alamos National Laboratory Report LA UR 88 3238 September 1998 PRA89 R E Prael and H Lichtenstein User Guide to LCS The LAHET Code System Los Alamos National Laboratory Report LA UR 89 3014 Revised September 15 1989 http www xdiv lanl gov XCl PROJECTS LCS lahet doc html PRA94 R E Prael A Review of Physics Models in the LAHETTM Code LA UR 94 1817 Los Alamos National Laboratory PRA95 R E Prael and D G Madland LAHET Code System Modifications for LAHET 2 8 Los Alamos National Laboratory Report LA UR 95 3605 September 1995 PRA96 R E Prael D G Madland A Nucleon Nucleus Elastic Scattering Model for LAHET in Proceedings of the 1996 Topical Meeting on Radiation Protection and Shield ing April 21 25 1996 No Falmouth Mass American Nuclear Society 1996 pp 251 257 PRA98a R E Prael A Ferrari R K Tripathi A Polanski comparison of Nucleon Cross Section Parameterization M
288. er blocks printed for each energy 5 7 7 FMn Tally Multiplier Form FMn bin set 1 bin set 2 T Table 5 64 Tally Multiplier Card Variable Description n tally number bin set i multiplier set 1 multiplier set 2 attenuator set attenuator set C 1 m px mp pxo multiplier set i C m reaction list 1 reaction list 2 special multiplier set i C k C multiplicative constant 1 flag indicating attenuator rather than multiplier set m material number identified on an Mm card density times thickness of attenuating material atom density if PA positive mass density if negative k special multiplier option sums and products of ENDF or special reaction numbers reaction list described in Appendix Example 1 FMn Cm R Ro R R3 Example 2 FMn Cm R R2 R3 These two examples reiterate that parentheses cannot be used for algebraic hierarchy within a reaction list The first example produces a single bin with the product of reaction R with the sum of reactions Rand R3 The second case creates two bins the first of which is reaction R alone the second is the sum of Rp and R3 without reference to R4 Example 3 F2 N1 2 3 4 124 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 FM2 C C C3 C4 T Example 4 F12 Ni2 3 4 FM12 C Example 5 F22 N 1 2 3 4T FM22 C 1 C2 C3 C4 These three exam
289. ergies or Times 155 5 8 4WWP Weight Window Parameter 20000eeeeee 155 5 8 5 WWN Cell Based Weight Window Bounds 156 5 8 6 WWE Weight Window Energies or Times 2 2 5 157 5 8 7 MESH Mesh Based Weight Window Generator 158 5 8 8 EXT Exponential Transform 200s eee eee ee eee 159 5 8 9 VECT Vector Input 20 00 cee eee 160 5 8 10 FCL Forced Collision 0 0 c eee eee eee 160 5 8 11 DDn Detector Diagnostics 2 00 eee 161 5 8 12 PDn Detector Contribution 00 cee eee eee 162 5 8 13 DXT lt DXTRAN i 2c oie Cite i et hee ae aed 163 5 8 14DXC DXTRAN Contribution 200 c eee eee 163 5 8 15 BBREM Bremsstrahlung Biasing 220 2eeeeeees 164 5 8 16 SPABI Secondary Particle Biasing 2 02 0000e 164 5 8 17 ESPLT Energy Splitting and Roulette 05 165 5 8 18 PWT Photon Weight 20000 cece 166 5 9 Output Control 2 025222 e bee ee ee Rie eee Seed eee aie 166 5 9 1PRDMP Print and Dump Cycle 0 200 cece eee 166 5 9 2 PRINT Output Print Tables 000 eee eee eee 167 5 9 3 MPLOT Plot tally while problem is running 169 5 9 4 PTRAC Particle Track Output 2000s 169 5 9 5 HISTP and HTAPESX 000 c eee eee 171 5 9 6 DBCN Debug Information 0 00 e eee eee 171 5 9 7LOST Lost Particle 00 cee eee 173 5 9 8IDUM Int
290. ersion 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium Items 2 and 3 above are written as list directed input 1 Repeat counts are allowed including repeat counts for commas to take default values i e 4 expands to Mul tiple cases may be processed for each case the above structure applies Slashes are allowed only in the first pair of title cards unless each title card containing one or more slashes has an S in column 1 The option control record defines the options to be used and the additional input informa tion that must be specified for the problem The structure of this record is IOPT NERG NTIM NTYPE KOPT NPARM NFPRM FNORM KPLOT IXOUT IRS IMERGE ITCONV IRSP ITMULT Some of the parameters in this record may optionally be preceded by a minus sign whose meaning is defined below see Table D2 for applicability Thus if NTIM is specified by inserting 3 in the option control record it is interpreted as NTIM 3 with a minus sign flag attached In the discussion which follows input control parameters are treated as pos itive or zero quantities even though the flag may be present Table B 1 Applicability of Input Control Parameters IOPT NERG NTIM NTYPE NPARM NFPRM KPLOT IXOUT IMERGE ITCONV IRSP_ ITMULT 1 O O R R O N N O O O O 101 O O R R O N N O O O O 2 102 O O R R N N N O O O O 3 O O N 0 N 0 N N 0 N N 103 O O N R N 0 N N 0 N N 5 N N
291. ersion 2 4 0 September 2002 LA CP 02 408 Use Required if mesh based weight windows are used or generated Example GEOM cyl REF 1e 6 1e 7 0 ORIGIN 1 2 3 IMESH 2 55 66 34 INTS 2 15 2 fine bins from 0 to 2 55 15 from 2 55 to 66 34 JMESH 33 1 42 1 53 4 139 7 JINTS 6 3 4 13 KMESH 5 1 KINTS 5 5 Example GEOM rec REF 1e 6 1e 7 0 ORIGIN 66 34 38 11 60 IMESH 16 5 3 8 53 66 IINTS 10 3 8 10 fine bins from 66 34 to 16 5 etc 5 8 8 EXT Exponential Transform Form EXT n 4142 4 A Table 5 93 Exponential Transform Card Descriptor Description n any particle designator or IPT number in Table 4 1 entry for cell i i Each entry 4 is of the form 4 OVm where Q describes the amount of stretching and Vm defines the stretching direction number of cells in the problem Default No transform 4 0 Use Optional Use cautiously Weight windows strongly recommended Example EXT N00 7V2 S SV2 6V9 0 5V9 SZ 4X VECT v9 000V2 111 MCNPX User s Manual 159 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 The 10 entries are for the 10 cells in this problem Path length stretching is not turned on for photons or for cells 1 2 and 7 Following is a summary of path length stretching in the other cells Table 5 94 stretching cell Aj Q Vm parameter direction 3 7V2 7 V2 p 7 toward point 1 1 1 4 S S p particle direction 5 SV2 S V
292. es and energy deposition will be handled in the regular process of tracking those particles Where there are no libraries available de dx nuclear recoil and the energies of some non tracked secondary particles are added to the F6 collision estimator A secondary particle can be produced either by collision or by particle decay In MCNPX version 2 3 0 the energies of neutral particles will never be added to the collision estimator this includes neutrons photons neutrinos pi0 and neutral Kaons This is not consistent with the library heating factor treatment and will be reconsidered in future versions of the code There fore it is especially important for the user to include all possible secondary particles on the MODE card especially photons and neutrinos in order to get the most accurate energy deposition tally Figure 8 5 illustrates the difference in an energy spectra for neutrons ona tungsten target when photons are not 8 5a or are 8 5b included in the MODE card The difference made by tracking the photons is substantial Figure 8 4 Energy Deposition Spectra for Neutrons produced by an 800 MeV proton beam on Tungsten a MODE hn dtsau file runtpe tally 36 10 9 10 8 tally nev particle 10 10 Tm orm i i i1 1 Energies of particles which fall below minimum energy cutoffs will also be deposited locally The user must be certain that the value of these cutoff energies will not cause the results of the
293. espect to the y axis For KOPT 3 or 8 the cosine is taken with respect to the z axis For KOPT 4 or 9 the cosine is taken with respect to an arbitrary vector to be read in If KOPT 5 6 7 8 or 9 the current tallies are binned according to a slicing of each sur face into NSEG 1 segments by NSEG planes In this case all additional record of the following form is required IFSEG NSEG FSEG 1 FSEG NSEG For IFSEG 1 the segmenting planes are perpendicular to the x axis for IFSEG 2 the y axis and for IFSEG 3 the z axis The FSEG I are the coordinates of the NSEG planes in increasing order Segmenting may also be accomplished by using segmenting cylinders The input has the same format as segmenting by planes however IFSEG negative designates cylindrical segmenting IFSEG 1 indicates that the segmenting cylinders are concentric with the x axis IFSEG 2 indicates that the segmenting cylinders are concentric with the y axis IFSEG 3 indicates that the segmenting cylinders are concentric with the z axis The val ues of the FSEG array are the radii of nested concentric cylinders and must be in increasing order Segmenting cylinders are concentric with an axis not just parallel For KOPT 4 or 9 an additional record must be supplied with the direction cosines of the arbitrary vector with which cosine binning is to be made The form of this record is CN 1 CN 2 CN 8 where the parameters input are the direction c
294. ess error bars The default is to include error bars THICK x Set the thickness of the plot curves to the value x The legal values lie in the range from 0 01 to 0 10 The default value of THICK is 0 02 THIN Set the thickness of the plot curves to the legal mini mum of 0 01 LEGEND x y Include or omit the legend according to the values of optional parameters x and y no x and no y put the legend in its normal place the default x 0 and no y omit the legend x and y defined for 2D plots only put most of the legend in its usual place but put the part that labels the plot lines at location x y CONTOUR cmin cmax cstep Commands that specify the form of contour plots Define cmin cmax and cstep as the minimum maximum and step values for contours If the optional symbol is included the first three parameters are interpreted as percentages of the minimum and maximum values of the dependent variable The default values are 5 95 10 available with COPLOT 5 3 GEOMETRY CELL SURFACE BOX RPP SPH RCC RHP HEX REC TRC ELL WED ARB VOL AREA U FILL TRCL LAT TRn 5 3 1 Cell Form j MCNPX User s Manual md geom params 58 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 or j LIKE n BUT list Table 5 4 Cell Cards Keyword Description cell number 1 lt j lt 99999 j If cell has transformation 1 lt j s 999 See Section 0
295. est of the set of maximum energies found among photon tables in the problem If electrons are being transported or only photons but with consideration of secondary electron thick target bremsstrahlung then the photon maximum energy is adjusted to be no higher than the electron maximum energy 74 MCNPX User s Manual MCNPX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium In order to turn on photonuclear interactions a fourth entry PNINT has been added to the PHYS card when used with the p designator WHI00 PHYS p EMCPF IDES NOCOH PNINT Table 6 2 Turning on Photonuclear Interactions Keyword Description EMCPF Upper energy limit in MeV for detailed photon phys ics treatment IDES 0 photons will produce electrons in MODE E prob lems or bremsstrahlung photons with the thick tar get bremsstrahlung model 1 photons will not produce electrons as above NOCOH 0 coherent scattering occurs 1 coherent scattering will not occur PNINT 1 Analog photonuclear interactions turned on 0 Photonuclear interactions turned off default 1 Biased phtonuclear interactions turned on No changes have been made to the TMP THTME or MTm cards 6 1 8 Problem Cutoffs Cards CUT ELPT NPS CTME The CUT and ELPT cards can now designate any particle symbol NPS can now have two arguments related to the radiography tally capability These are discussed i
296. ethods for Medium and High Energies in Proceedings of the Fourth Workshop on Simulating Accelerator Radiation Environments SARE4 Sep tember 14 16 1998 Knoxville Tn ed by Tony A Gabriel ORNL pp 171 181 PRA98b R E Prael Upgrading Physics Packages for LAHET MCNPX Proceedings of the American Nuclear Society Topical Meeting on Nuclear Applications of Accelerator Technology Gatlinburg TN Sept 20 23 1998 PRA98c R E Prael and W B Wilson Nuclear Structure Libraries for LAHET and MCNPX Proceedings of the Fourth workshop on simulating Accelerator Radiation Envi ronments SARE4 September 14 16 1998 Knoxville Tn ed by tony A Gabriel ORNL pp 183 PRA99 R E Prael Primary Beam Transport methods in LAHET Transactions of the June ANS Meeting Boston June 6 10 1999 PRAOOa R E Prael Proposed Modification to the Charged Hadron Tracking Algorithm in MCNPX Los Alamos Research Note X 5 RN U August 23 2000 LA UR 00 4027 PRAOOb R E Prael A New Nuclear Structure Library for MCNPX and LAHET3 Pro ceedings of the Fourth International topical Meeting on Nuclear Applications of Accelerator Technology Nov 12 15 2000 Washington DC pp 350 352 RAD77 Radiation Shielding Information Center HETC Monte Carlo High Energy Nucleon Meson Transport Code Report CCC 178 Oak Ridge National Laboratory August 1977 122 MCNPX User s Manual MCNPX User s Manual Ap Version 2 3 0 April 2
297. etic Energy Default 2500 MeV For pions the Bertini INC model will be used below this value FLENB4 Kinetic Energy Default 2500 MeV For pions the FLUKA high energy generator will be used above this value See Notes under FLENB2 FLENB5 Kinetic Energy Default 800 MeV For nucleons the ISABEL INC model will be used below this value FLENB6 Kinetic Energy Default 800 MeV For nucleons an appropriate model will be used above this value for IEXISA 2 it applies to all particle types for IEXISA 1 it applies to all particles except nucleons and pions for IEXISA 0 it is immaterial See the example following this table for further explanation 80 MCNPX User s Manual MCNPX User s Manual i Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium Table 6 4 LCB Keyword Descriptions Continued Keyword Description CTOFE The cutoff kinetic energy MeV for particle escape during the INC when using the Bertini model The cutoff energy prevents low energy nucleons from escap ing the nucleus during the INC for protons the actual cutoff is the maximum of CTOFE and a Coulomb barrier CTOFE gt 0 CTOFE will be used as the cutoff energy CTOFE lt 0 a random cutoff energy uniformly distributed from zero to twice the mean binding energy of a nucleon will be sampled for each projectile target interaction and separately for neutrons and protons In this case the Coulomb barrier for
298. ets the lower limit of the first bin and the other entries set the upper limit of each of the bins These limits are set relative to the intersection of the reference direction An example is discussed below FSn 20 99i 20 Cn 20 99i 20 These two cards set up a 100 x 100 grid that extends from 20 cm to 20 cm in both directions and has 10 000 equal size bins These bins need not be equal in size nor do they need to be symmetric about the reference direction The directions of the t axis and s axis of the grid are set up such that if the reference direction the outward normal to the grid plane is not parallel to the z axis of the geometry the t axis of the grid is defined by the intersection of the grid plane and plane formed by the z axis and the point where the reference direction would intersect the grid plane If the reference direction is parallel to the z axis of the geometry then the t axis of the grid is defined to be parallel to the y axis of the geometry The x axis of the grid is defined as the cross product of a unit vector in the t direction and a unit vector in the reference direction MCNPX User s Manual 137 MCNPxX User s Manual Version 2 4 0 September 2002 LA CP 02 408 5 7 20 2 Transmitted Image Projection In the transmitted image projection case the grid acts like a film pack in an X ray type image or transmitted image projection There is a cylindrical grid for generating an image In both cases for
299. f IRESP I outside the range 1 5 is treated as 1 i e constant over the interval The energy range for the specified response function need not span all possible particle energies in the problem If a particle energy falls below ERESP 1 then FRESP 1 is used as the value of the response function Similarly if a particle energy exceeds ERESP NRESP then FRESP NRESP is used as the value of the response function 23 Executing HTAPE3X The default file name for the input is INT the default file name for the output is OUTT the default file name for the history file is HISTP and the default file name for the surface crossing file is HISTX for input into HTAPESX The latter is written by MCNPX with the default file name WSSA If option 8 is requested the data file PHTLIB must be in the user s file space if option 16 is requested the data file BERTIN must be in the user s file space All these file names may be defined by file replacement on the execute line HTAPESX INT my_input OUTT my_output HISTP file1 HISTX file2 MCNPX User s Manual 151 MCNPX User s Manual Ap Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium References 1 2 R E Prael and H Lichtenstein User Guide to LCS The LAHET Code System LA UR 89 3014 Los Alamos National Laboratory September 1989 http www xdiv lanl gov XCI PROJECTS LCS lahet doc htm H G Hughes R E Prael and R C Little MCNPX The LAH
300. f results by batch size 180 Weight window generator bookkeeping summary controlled by WWG 7 not print card 190 basic Weight window generator summary 198 Weight windows from multigroup fluxes 200 basic Weight window generated windows MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Example PRINT 110 40 150 The output file will contain the basic tables plus tables 40 110 and 150 not 160 161 162 the default tables and the shortened version of 175 Example PRINT 170 70 110 The output file will contain all the basic tables all the default tables the long version of table 175 and all the optional tables except tables 70 110 and 170 applicable to your problem 5 9 3 MPLOT Plot tally while problem is running Form MPLOT MCPLOT keyword parameter Default None Use Optional This card specifies an intermediate tally results plot of that is to be produced periodically during the run The entries are MCPLOT commands for one picture The sign is optional During the run as determined by the FREQ n entry MCRUN will call MCPLOT to display the current status of one or more of the tallies in the problem If a FREQ n command is not included on the MPLOT card n will be set to 5000 The following commands can not appear on the MPLOT card RMCTAL RUNTPE DUMP and END All of the commands on the MPLOT card are executed for each displayed picture so coplot
301. f the new surface is a plane you must specify the portion to be used by means of POS and RAD or possibly X Y and Z source distributions Because there are no collisions a short run will generate a great many tracks through your system If there are any geometry errors they should cause some of the particles to get lost When a particle first gets lost whether in a special run with the VOID card or in a regular production run the history is rerun to produce some special output on the OUTP file Event log printing is turned on during the rerun The event log will show all surface crossings and will tell you the path the particle took to the bad spot in the geometry When the particle again gets lost a description of the situation at that point is printed You can usually deduce the cause of the lost particle from this output It is not possible to rerun lost particles in a multitasking run MCNPX User s Manual 42 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 If the cause of the lost particle is still obscure try plotting the geometry with the origin of the plot at the point where the particle got lost and with the horizontal axis of the plot plane along the direction the particle was moving The cause of the trouble is likely to appear as a dashed line somewhere in the plot or as some discrepancy between the plot and your idea of what it should look like 4 4 STORAGE LIMITATIONS Table 4 4 summarizes some of the
302. fission evaporation particle production spectrum e post fission evaporation particle production spectrum e fission precursor mass edit The CEM reaction model is of limited use when light reaction targets interact with high energy incident particles The Fermi Breakup model which usually han dles the reaction dynamics of light nuclei is not implemented into CEM in MCNPX version 2 3 0 This means that at sufficiently high energies CEM can boil off all neu trons from a nucleus and hands over an unphysical highly excited nucleus to the gamma deexitation module PHT For Sodium such events have been identified already at 500 MeV incident energy For heavier nuclei this limit is shifted to higher energies This will be corrected in a future version Specifying different densities for the same material is a fatal error In running a neutron only problem one can specify cells with the same material but different densi ties The scaling for such situations is always linear and adjustments are straightfor ward No so for charged particles there is a density correction in energy deposition which is not a strict linear function In MCNP4B which is the basis for the currently released MCNPX 2 3 0 the procedure is to search through all cells and find the first one with the material in question and use that density for the correction factor for all cells using that material The effect is small so this is an adequate procedure how ever MCNP doe
303. for each source distribution If a lot of cell parameter options are being used additional blocks of column data would be needed We strongly suggest keeping columns reasonably neat for user readability The column format is intended for input data that naturally fit into columns of equal length but less tidy data are not prohibited If a longer column is to the right of a shorter column the shorter column must be filled with enough J entries to eliminate any ambiguity about which columns the data items are in Special syntax items R M I Log and J are not as appropriate in column format as they are on horizontal lines but they are not prohibited They are of course interpreted vertically instead of horizontally Multiple special syntax items such as 9R are not allowed if cell or surface names are present The form of a column input block is E 8 So a Ss K D Dig Dim K Dz Dap Dam Kn Dn Dpn2 Dam The is somewhere in columns 1 5 2 Each line can be only 80 columns wide 3 Each column S through D where may be less than n represents a regular input card 4 The S must be valid MCNPX card names They must be all cell parameters all surface parameters or all something else 5 D through D must be valid entries for an S card except that D 1 through Dpi may be some J s possibly followed by some blanks 6 If D is non blank D must also be nonblank A J may be used if necessary to make D
304. function IRSP lt 0 indicates that the tally will be divided by a user supplied response function The default is 0 For a discussion see Sec tion 22 below ITMULT is the TIME MULTIPLIER flag ITMULT gt 0 indicates that the weights tallied will be multiplied by the event time This option applies only when the basic option type is 1 2 4 9 10 or 13 The standard definitions for these input variables may not apply for some options The applicability of the option control parameters is summarized in Table D1 According to the parameters specified on the option record the following records are required in the order specified e For NERG gt 0 a record defining NERG upper energy bin boundaries from low to high defined as the array ERGB I I 1 NERG The first lower bin boundary is implic itly always 0 0 The definition may be done in four different ways First the energy boundary array may be fully entered as ERGB I l 1 NERG Second if two or more but less than NERG elements are given with the record terminated by a slash the array is completed using the spacing between energy boundaries obtained from the last two entries Third if only one entry is given it is used as the first upper energy boundary and as a constant spacing between all the boundaries Fourth if only two entries are given with the first negative and the second positive the second entry is used as the uppermost energy boundary ERGB NERG and the first entr
305. g or anything else F4 N F14 N F104 N and F234 N are all legitimate neutron cell flux tallies they could all be for the same cell s but with different energy or multiplier bins for example Similarly F5 P F15 P and F305 P are all photon point detector tallies Having both an F1 N card and an F1 P card in the same INP file is not allowed The tally number may not exceed three digits Tally types 1 2 4 and 5 are normally weight tallies particles in the above table however if the Fn card is flagged with an asterisk for example F1 N energy times weight will be tallied The asterisk flagging can also be used on tally types 6 and 7 to change the units from MeV g to jerks g 1 jerk 1 GJ 109 J The asterisk on a tally type 8 converts from a pulse height tally to an energy deposition tally All of the units are shown in the above table Tally type 8 can also be flagged with a plus to convert it from an energy deposition tally flagged with an asterisk to a charge deposition tally The tally is the negative particle weight for electrons and the positive weight for positrons The F8 tally can be checked against an F1 E type surface tally Only the F2 surface flux tally requires the surface area The area calculated is the total area of the surface that may bound several cells not a portion of the surface that bounds only a particular cell If you need only the segment of a surface you might segment the full surface with the FSn c
306. g the default Note that the last four records could be written on one line as 0 5 800 V Tally option 13 may be considered as the time integrated particle current integrated over a sphere in a void at a very large distance for the interaction region Since it is normalized per unit solid angle the units are dimensionless being sr per source particle 16 Edit Option IOPT 14 or 114 Gas Production Option 14 provides an edit of hydrogen and helium gas production by isotope by element and total Unless modified by FNORM the units of gas production are atoms per source particle If KOPT 0 the edit is by cell number if KOPT 1 the edit is by material NERG NTIM and NTYPE are unused The estimate is made by tallying all H and He ions stopped in a cell or material including source particles 17 Edit Option IOPT 15 or 115 Isotopic Collision Rate Option 15 has been added to provide a collision rate edit by target isotope The input has the same meaning as for IOPT 8 with the following exceptions KOPT 0 or 1 tabulates all collisions KOPT 2 or 3 tabulates elastic scattering only KOPT 4 or 5 tabulates nonelastic events only If KOPT is even the edit is by cell number if KOPT is odd the edit is by material number A CINDER removal rate input file will produced for IXOUT gt 0 The default CINDER file name is OPT15A MCNPX User s Manual 219 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 40
307. gged is as root will unload the distribution into usr local src mcnpx_2 3 0 will build the system in tmp menpx will install the mcnpx executable in usr local bin and will install the libraries end eventually the mcnp 16 MCNPX User s Manual MCNPX User s Manual i Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium cross sections into usr local lib Naturally the specific name of the mcnpx distribution archive will vary depending on the version you have acquired The following example uses bourne shell commands that follow accomplish this task If you are more familiar with csh you will need to adjust things appropriately NOTE Com ments about the shell commands start with the character Also don t be alarmed by the generous amount of output from the configure and make scripts They work hard so you don t have to go to the installation directory cd usr local src Unpack the distribution This creates the directory mcnpx_2 3 0 gzip dc mcnpx_2 3 0 tar gz tar xf go to tmp and make the build directory cd tmp mkdir mcnpx go into that working space cd mcnpx execute the configure script no special option requests for the Makefiles the default directory prefix is usr local usr local src mcnpx_2 3 0 configure now make the executable mcnpx program and supporting LCS libraries make all run the regression tests for your architecture make tests install the exec
308. gram The capabilities of gridconv have recently been expanded so that any and all tallies writ ten to mctal can be processed The code is still interactive but now shows all tallies in the problem from which any may be selected The user has the option of generating one or two dimensional output The user is then told about the bin structure so the one or two free variables may be selected The energy is the default independent variable in the one dimensional case There is no default for the two dimensional case The order in which the two dimensional bin variables are selected does not make any difference to the output in that the order of the processing will be as it appears on the metal file Gridconv will work with mctal files produced both by MCNPX and MCNP MCNPX User s Manual 101 MCNPX User s Manual Ap Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium 8 2 The Radiography Tally A capability has been added to MCNPxX to allow the code to generate simulated radiogra phy images as one would expect to see from an X ray or pinhole projection of an object containing the particle source This allows the recording of both the direct source image as well as that due to background scatter This tool is an invaluable aid to the problem of image enhancement or extracting the source image from a background of clutter MCNPX includes two types of image capability the pinhole image projection and the transm
309. h2 h3 for a z hex with height h h1 h2 h3 00 h vector from the axis to the middle of the first facet for a pitch 2p facet normal to y axis r1 r2 r3 0p0 rm r2 r3 s1 s2 s3 vector to center of the 2nd facet t1 t2 t3 vector to center of the 3rd facet Example RHP 00 4 008 020 a hexagonal prism about the z axis whose base plane is at z 4 with a height of 8 cm and whose first facet is normal to the y axis at y 2 5 3 2 4 6 REC Right Elliptical Cylinder Form REC VxVyVz HxHyHz V1xV1yV1iz V2xV2y V2z Table 5 14 Macrobody Right Elliptical Cylinder Argument Description Vx Vy Vz X y z coordinates of cylinder bottom Hx Hy Hz cylinder axis height vector Vix Viy V1z ellipse major axis vector normal to Hx Hy Hz V2x V2y V2z ellipse minor axis vector orthogonal to Hx Hy Hz NOTE If there are 10 entries instead of 12 the 10th entry is the minor axis radius where the direction is determined from the cross product of H and v1 MCNPX User s Manual 65 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Example REC 0 50 0100 400 2 a 10 cm high elliptical cylinder about the y axis with the center of the base at x y z 0 5 0 and with major radius 4 in the x direction and minor radius 2 in the z direction 5 3 2 4 7 TRC Truncated Right Angle Cone Form TRC VxVyVz HxHyHz R1 R2 Table 5 15 Macrobody Truncated Right Angle Cone Argument Descrip
310. hat is required such as file type file names etc In most cases the default value is used and a return is all that is necessary 152 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Once the header information from mdata has been read from the file gridconv can either produce an ASCII file from a binary or generate the required graphics input files as requested by the user Note that the ASCII file contains raw data not normalized to the number of source particles The reason for the option to write an ASCII file is that sometimes users will want to look at the numbers in the mdata file before doing any plotting or check the numerical results for a test case The ASCII option is also very useful for porting the mdata file to another computer platform and for reading the data into graphics packages not currently supported by gridconv Gridconv is currently set up to generate one two or three dimensional graphics input files with any combination of binning choices Once the input file has been generated gridconv gives the user the options of producing another file from the currently selected mesh tally selecting a different mesh tally available on this mdata file or reading information from a different file Of course there is always the option to exit the program The capabilities of gridconv have recently been expanded so that any and all tallies written to mctal can be processed The code is s
311. hat portion of the captured particle 4 3 2 2 Electron Interactions Electron transport is described in detail in Part E of Chapter 2 of the MCNP4B manual Most users familiar with Monte Carlo techniques know that the very large number of inter actions in electron transport greatly slows computational time Therefore much work has been done to develop techniques which take advantage of the statistical nature of electron transport assuming that the energy loss with each individual interaction is small compared to the particle s kinetic energy In particular energy loss and angular deflection of electrons over short steps can be sampled from probability distributions This condensed history method of transport was first developed by Berger in 1963 BER63 Based on those tech niques Berger and Seltzer developed the ETRAN series of electron photon transport codes SEL88 John Halbleib and collaborators at Sandia National Laboratory used ETRAN as the basis of the Integrated TIGER series of electron photon transport codes HAL88 The electron physics in MCNP4B and MCNPx is essentially that of the Integrated TIGER series A brief discussion of the major physics models used in electron transport is given below We present this detail since these or modifications of these methods are also used in heavier charged particle transport as described in Chapter 5 This discussion is adapted from that given in Chapter 2 of the MCNP4B manual BRI97 El
312. he Standard Sun linker value will be used to link object code Unlike the with FC and with CC options whose names are used for more than just find ing the executable The value can be a full path to the location of the desired Id program as well as being a single name like Id configure will search for a linker and use the first one it finds This is typically needed on systems with both a vendor supplied compiler set and the GNU tool set In such cases there may be two versions of Id that must be differentiated this option can be used in combination with other options such as with DEBUG and with FC prefix value substitute a full path name for the value placeholder e g home team mcnpx the path given should be different from the working directory where the build is taking place value will be used in the install step to create bin and lib data directories for mcnpx s use a default value of usr local is used as the full path name for the install step Executa bles then go to usr local bin and data files go to usr local lib permissions of the destination may prohibit success of installation libdir value substitute a full path name for the value placeholder e g home team mcnpx the path given should be different from the working directory where the build is taking place value will be used in the install step to create a library data directory for m
313. he generated Makefiles this option can be used in com bination with other options such as with FC and with CC with FC value substitute the desired Fortran90 compiler name for the value placeholder e g with FC fort to use the fort compiler value will be used to compile Fortran source code loca tion of binary directory con taining value must be in your PATH environment variable configure will search for a Fortran90 compiler and use the first one it finds this option can be used in com bination with other options such as with DEBUG and with CC with CC value sub stitute the desired C compiler name for the value placeholder e g with CC gcc to use the gcc com piler value will be used to compile C source code location of binary directory containing value must be in your PATH environment vari able configure will search for a C compiler and use the first one it finds this option can be used in combination with other options such as with DEBUG and with FC 20 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 3 1 Configure Script Parameters Option Syntax Effect on the generated Makefile if requested Effect on the generated makefile if NOT requested with LD value Sub stitute the desired link editor for the value placeholder e g with LD usr ccs bin Id to use t
314. he Dual Parton Model event generators HADEVT and NUCEVT RAN85 for hadron hadron and hadron nucleus collisions as implemented in the form of EVENTQ in the FLUKA 87 hadron cascade code AAR86 AAR87 Some improvements mainly bug corrections were made by Ferrari and Sala in the 1989 1990 period Since 1987 three more FLUKA event generators have been released a Release contained in GEANT versions 3 16 to 3 21 and which was contained in the offi cial FLUKA code until Spring 1993 FAS94a FAS94b b Release contained in the official FLUKA code until Spring 1997 FAS97 FER98 c The release contained in the present version of FLUKA at this time COLOO 4 3 Nuclear Data Tables Tabular data is needed by MCNPxX in two ways For low energy neutrons the usual capa bility of MCNP4B to use tabular data has been retained In MCNPX this capability has been expanded to also use proton libraries and a program is now in place to develop pho tonuclear capability The collection of enhanced libraries is described in Section 4 3 1 For interactions above library cutoff energies additional tabular data are needed Total reaction and elastic cross section data are included in MCNPX in tabular format and sup plement the high energy physics capabilities This work is described in Section 4 3 2 4 3 1 Nuclear Data Libraries It has long been known that the intranuclear cascade physics includes no nuclear structure effects Standard nuclear data libr
315. he FLUKA code improvements added since that time See Section 5 5 7 for further information The FLUKA code module will be upgraded in a future version of MCNPX The contents of the HISTP file arising from interactions processed by the CEM mod ule do not distinguish among evaporation particles emitted before or after fission All are labeled as pre fission Therefore the HTAPE edits that depend on this distinction will not produce the intended output pre fission evaporation particle production spectrum epost fission evaporation particle production spectrum fission precursor mass edit MCNPX User s Manual 12 13 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 The CEM reaction model is of limited use when light reaction targets interact with high energy incident particles The Fermi Breakup model which usually handles the reac tion dynamics of light nuclei is not implemented into CEM in this version of MCNPX This means that at sufficiently high energies CEM can boil off all neutrons from a nucleus and hands over an unphysical highly excited nucleus to the gamma deexita tion module PHT For Sodium such events have been identified already at 500 MeV incident energy For heavier nuclei this limit is shifted to higher energies This will be corrected in a future version Specifying different densities for the same material produces a warning For charged particles there is a density
316. he banked particles and the tallies this type of splitting could be a total waste of time Roulette on the other hand eliminates the need to transport and tally a large number of insignificant particle tallies As with any splitting or roulette game the weights of the banked particles have to be adjusted to make the tallies correct In order to insure that the weight cutoff game does not have an adverse effect on the particles because of this type of weight reduction a splitting roulette factor is generated and banked with the particle When the weight cutoff game is played this factor is used to adjust the weight much in the same way as the adjustment made for cell splitting and roulette This factor could probably be used to correct a weight cutoff problem encoun tered with the energy splitting option currently in the code 90 MCNPX User s Manual MCNPX User s Manual Pr Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium 8 New and Improved Tallies and Data Analysis No fundamental changes have been made to the format of any output table as currently found in MCNP4b however additional lines have been added for information on the new particles These should be obvious and will not be described in detail MCNPX includes several new tally capabilities section 8 1 8 2 and 8 3 as well as mod ifications to the Energy Deposition scoring capabilities section 8 4 In addition the MCNPX distribution includes
317. he format of an MCNPX MCTAL file Plotting of XSTAL is performed by MCNPX using the execution option mcnpx z followed by the required instructions rmctal xstal nonorm The latter is essential since the data are normalized in XSEX3 Each case in XSEX3 is expanded in the XSTAL file for each particle type produced The tallies are identified by the numbering scheme 100 case number particle type 230 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 the latter defined in the table below The last in the sequence corresponds to the elastic scattering distribution of the incident particle When plotting XSEX3 output the appropriate y axis labels are barns MeV steradian barns MeV or barns steradian If the yield multiplicity option was used in XSEX3 the appropriate y axis labels are particles MeV steradian etc The energy axis may be either energy MeV or momentum MeV c according to the XSEX3 option employed Table 9 2 Type Particle 1 proton 2 neutron 3 pi 4 pid 5 pi 6 deuteron 7 triton 8 He 3 9 alpha 10 photon prompt gamma from residual 11 K 12 K all neutrals 13 K 14 antiproton 15 antineutron 16 elastic scattered projectile An example of a COMOUT file produced when plotting XSTAL is shown on the next page rmctal xstala MCNPX User s Manual 231
318. he target isotope has no mass in XSDIR the value supplied is not used and is arbitrary As an example the following is a sample MCNPX input for a cross section calculation MCNPX User s Manual 153 MCNPX User s Manual E Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium MCNPX standard cross section generation format for XSEX3 use c 1000 MeV protons on Sn121 an isotope not in MCNP library c and for which no atomic weight is specified in XSDIR c Minimal geometric specification for this purpose mi 50121 1 not in MCNP libraries awtab 50121 119 864 need value but arbitrary lea 2j0 imp h 10 phys h 1000 mode h print nps 1000 prdmp 2j 1 154 MCNPX User s Manual MCNPX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium 3 Input for XSEX3 The input file for XSEX default name INXS has the following structure 1 Two records of title information 80 columns each 2 An option control record list directed format 3 Additional records as required by the chosen options list directed format Multiple cases may be processed for each case the above input structure applies When multiple cases are processed input quantities default to the preceding case If the title records of the second and subsequent cases contain the record must begin with a S The option control record has the structure NERG NANG FNORM KPLOT IMOM IYIELD
319. he transformation on the TRn card which must be present in the INP file of the current problem Dn Distribution number for a set of SIn SPn and SBn cards If the surface source is transformed into several locations the SIn card lists the transformation numbers and the SPn and SBn cards give the probabilities and bias of each transformation Default no transforma tion TR c a nonnegative constant that is used in an approxima tion to the PSC evaluation for the probability of the sur PSC ae are c face source emitting a particle into a specified angle relative to the surface normal The following four keywords are used only with spherically symmetric sur face sources that is sources generated with SYM 1 on the SSW card uvw Direction cosines that define an axis through the AXS center of the surface sphere in the auxiliary original coordinate system This is the reference vector for EXT Default No axis Dn nis the number of a distribution SIn SPn and SBn cards that will bias the sampling of the cosine of the EXT angle between the direction AXS and the vector from the center of the sphere to the starting point on the sphere surface Default No position bias c Particles with a polar angle cosine relative to the POA source surface normal that falls between 1 and c will be accepted for transport All others are disregarded and no weight adjustment is made Default c 0 MCNPX
320. i breakup model only for A lt 5 ILVDEN 1 Use original HETC level density formulation See the LEB card for details on parameter inputs 0 Use Gilbert Cameron Cook Ignatyuk level density model PRA88 default 1 Use the Julich level density parameterization as a function of mass number CLO83 IEVAP 0 The RAL evaporation fission model ATC80 will be used default 1 The ORNL evaporation fission model BAR81 will be used Note The ORNL model allows fission only for isotopes with Z 91 NOFIS 1 Allow fission default 0 Suppress fission 82 MCNPX User s Manual Accelerator Production of Tritium MCNPX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 LEB YZERE BZERE YZERO BZERO This card controls level density input options for the original HETC implementation Table 6 6 LEB Keyword Descriptions Keyword Description YZERE The YO parameter in the level density formula for Z lt 70 The default is 1 5 zero or negative is an error condition For target nuclei with Z lt 70 the parameters BZERE and YZERE are used to compute level densities the default values are those used in LAHET before installation of the ORNL fission model For target nuclei with Z 71 the BZERO and YZERO parameters are used to compute level densities for the target nucleus and fission fragments Note Applies only for ILVDEN 1 BZERE The BO parameter level density formula for Z
321. ies between 20 and 150 MeV Case 2 In the second variation we transport not only nucleons denoted by the symbols n and h on the mode card and charged pions but also light ions deuterons tritons 3He and alphas denoted by d t s and a respectively The only differences between the two input decks are the two cards Base Case mode n h imp n h 1 1r 0 Case 2 mode nh dtsa imp n h d t s a 1 1r 0 Note that nuclear interactions by light ions are simulated using the ISABEL INC model The problem summary for this case is shown below sample problem spallation target Case 2 neutron creation tracks weight energy neutron loss tracks weight energy per source particle per source particle source 0 0 QO escape 366756 1 8321E 01 2 1938E 02 nucl interaction 316952 1 5848E 01 3 2187E 02 energy cutoff 0 Q 0 particle decay 0 0 0 time cutoff 0 0 us weight window 0 0 0 weight window 0 Ox 0 cell importance 0 QO QO cell importance 0 Os QO weight cutoff 0 0 0 weight cutoff 0 0 0 energy importance 0 QO QO energy importance 0 QO 0 dxtran 0 0 0 dxtran 0 0 0 MCNPX User s Manual 197 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 forced collisions 0 0 0 forced collisions 0 0 0 exp transform 0 QO D exp transform 0 QO QO upscattering 0 0 0 downscattering 0 0 9 8368E 00 tabular sampling 0 0 0 capture 0 1 4534E 02 7 7278E 02 n xn 79010 3 9467E 00 1 9031E 01 loss to
322. ies in the ionization implementation for heavy charged particles we do not recommend that the MCNPX user define more than one density for the same material Different Mn cards should be included for different densities see Section 6 1 6 5 3 Energy Straggling for Heavy Charged Particles MCNPxX like MCNP4B uses a sophisticated implementation of the Landau theory for electrons see Section 4 3 3 2 For heavy charged particles the assumptions of the Lan dau theory break down and the more complex Vavilov theory VAV57 must be used At low energies and large step sizes the Vavilov distribution approaches a Gaussian At very high energies or small step sizes and for electrons in almost all circumstances the Vavilov distribution approaches a Landau distribution The module implemented in MCNPxX to represent the Vavilov model does attempt to account for the Gaussian and Lan dau limits when step sizes and energies are appropriate for heavy charged particles This will be incorporated in future versions of the code An improved detailed logic for the use of the Vavilov straggling model for heavy charged particles is available and is now the default in Version 2 3 0 Previously the Vavilov model was used to establish a straggled energy loss rate closely tied to the step lengths of the major energy steps of the transport The smaller angular substeps and partial sub steps to boundaries or to potential interactions relied on a simple interpolati
323. if the cell is a void m material number if the cell is not a void This indicates that the cell is tocontain material m which is specified on the Mm card absent if the cell is a void d cell material density A positive entry is interpreted as the atomic density in units of 10 atoms cm A negative entry is interpreted as the mass density in units of g cm specification of the geometry of the cell It consists of signed surface numbers and Boolean operators that specify how the eom g regions bounded by the surfaces are to be combined Sains optional specification of cell parameters by entries in the key j word value form n name of another cell list set of keyword value specifications that define the attributes that differ between cell n and j Example 3 0 1 2 4 definition of cell 3 3 equivalent to next line 1 2 4 Example 2 3 37 1 IMP N 2 IMP P 4 3 LIKE 2 BUT TRCL 1 IMP N 10 This says that cell 3 is the same as cell 2 in every respect except that cell 3 has a different location TRCL 1 and a different neutron importance The material in cell 3 the density and the definition are the same as cell 2 and the photon importance is the same MCNPX User s Manual 59 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 5 3 2 Surface 5 3 2 1 Surfaces Defined by Equations Form j n a list Table 5 5 Surfaces Defined by Equations Keyword Description
324. ign the edit is performed only for events initiated by primary source particles If KOPT 0 or 1 the edit is of the final residual masses including elastic collisions If KOPT 2 or 3 the edit is of the residuals after the cascade phase and before evaporation If KOPT 4 or 5 the edit is of masses immediately preced ing fission If KOPT is even the edit is by cell number if KOPT is odd the edit is by material number If KPLOT 1 plots will be produced for each edit table Parameters NERG NTYPE and NFPRM are unused If IXOUT 1 an auxiliary output file appropriate for input to the CINDER program will be written the default file name is OPT8A Unless otherwise modified tally units are dimensionless weight of a residual nuclide per source particle An additional tabulation is produced which shows the estimated metastable state produc tion as a fraction of the total isotopic production As illustrated in the example here a state is identified by its excitation energy and half life the estimated fraction of total isotope pro duction associated with the particular metastable state is shown with the estimated relative standard deviation z a elev t half fraction 47 110 0 11770 2 17730D 07 4 00000D 01 0 3465 47 111 0 05990 6 50000D 01 8 00000D 01 0 2001 47 116 0 08100 1 05000D 01 S 00000D 01 0 5001 48 113 0 26370 4 41500D 08 2 85714D 01 0 3195 48 115 0 17340 3 87070D 06 5 00000D 01 0 3536 48
325. iguration if it is not present in the MCNPX distribution Check the config guess file to see if all Intel hardware platforms running Linux are spec ified Several uname commands at the beginning of the script set up four recognition factors that are concatenated with between them much like the setting of the PATH environment variable in some shell scripts This concatenation of the machine release system and version variables is then used in a long case statement when detecting com puting platforms Around line 336 in the copy current as this is being written the Linux case recog nizes any hardware platform not already recognized by previous cases that run the Linux OS Thus no modifications are needed to config guess Check the config sub file to see if all Intel hardware platforms running Linux are handled in the various case statement that handle the pieces of interest This script tries to con struct and return a string that is the concatenation of cpu type manufacturer and operating system with the character between them Again it is unlikely that you would have to modify this file as most current combinations are handled Check each of the case 28 MCNPX User s Manual MCNPxX User s Manual i Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium statements that use i 3 4 5 6 and linux to see if you have something different than what is specified For specifying a compi
326. il 2002 E LA UR 02 2607 Accelerator Production of Tritium Appendix A Examples Example 1 Neutron production from a spallation target One of the fundamental quantities of interest in most spallation target applications is the number of neutrons produced per beam particle incident on target For targets fed by pro ton accelerators this quantity is typically denoted as n p Here we demonstrate how one goes about calculating this quantity for a simple target geometry using MCNPX The geometry consists of a simple right circular cylinder of lead 10 cm in diameter by 30 cm long A beam of 1 GeV protons is launched onto the target The beam has a spot size of 7 cm diameter with a parabolic spatial profile see Fig A 1 Figure A 1 Neutron production from a spallation target In MCNPX net neutron production is tallied implicitly and is provided by default in the prob lem summary for neutrons The problem summary shows net neutron production resulting from nuclear interactions this is the component that accounts for neutron production by all particles transported using INC Preequilibrium Evaporation physics and net production by n xn reactions these are neutrons created in inelastic nuclear interactions by neu trons below the transition energy using evaluated nuclear data Net production from nuclear interactions is given by the difference of the neutron weights in the neutron cre ation and neutron loss columns
327. indows WWN and WWP cards are required unless importances IMP card or mesh based windows are used Example 1 WWE N E5 E3 WWN1 N wy77W72W73W 714 WWN2 N woywoowo3Wo4 WWN3 N w37W30W33 W34 These cards define three energy or time intervals and the weight window bounds for a four cell neutron problem Example 2 WWN1 Pwy w 2w73 This card without an accompanying WWE card defines an energy or time independent photon weight window for a three cell problem 5 8 6 WWE Weight Window Energies or Times Form WWE n E Ep Ej Ep j 99 Table 5 91 Variable Description n particle designator Ej upper energy or time bound of i window Ei lower energy or time bound of it window Eo 0 by definition Default One weight window energy MCNPX User s Manual 157 Use 5 8 7 MCNPX User s Manual Version 2 4 0 September 2002 Optional LA CP 02 408 MESH Mesh Based Weight Window Generator Form MESH mesh variable specification Table 5 92 Superimposed Mesh Variables Variable Meaning Default Mesh geometry either Cartesian xyz or rec or cylindrical rzt or xyz GEOM ie ee cyl x y and z coordinates of the reference point None variable REF must be present x y and z coordinates in MCNP cell geometry of the origin bottom 0 0 0 ORIGIN center for cylindrical or bottom left rear for rectang
328. ing but with diminished validity MCNPX User s Manual 81 MCNPxX User s Manual E Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium LEA IPHT ICC NOBALC NOBALE IFBRK ILVDEN IEVAP NOFIS LEA controls evaporation fermi breakup level density parameters and fission models All of these are external to the particular intranuclear cascade pre equilibrium model chosen Bertini ISABEL or CEM and may be used with any of these choices Table 6 5 LEA Keyword Descriptions Keyword Description IPHT 0 Do not generate photons in the evaporation stage 1 Generate de excitation photons default ICC Defines the level of physics to be applied for the PHT physics 0 The continuum model 1 Troubetzkoy E1 model 2 Intermediate model hybrid between 1 and 2 3 The spin dependent model 4 The full model with experimental branching ratios default NOBALC 0 Use mass energy balancing in the cascade phase 1 Turn off mass energy balancing in the cascade phase default Note A forced energy balance may distort the intent of any intranuclear cas cade model Energy balancing for the INC is controlled by the input parameter FLIMO NOBALE 0 Use mass energy balancing in the evaporation stage default 1 Turn off mass energy balancing in the evaporation stage IFBRK 1 Fermi breakup model for A lt 13 and for 14 lt A lt 20 with excitation below 44 MeV default 0 Use Ferm
329. ing Power for Heavy Charged Particles An improved collisional energy loss model has been added to MCNPX by modifying the stopping power module of LCS in several ways The ionization potentials have been enhanced to the values and interpolation procedures recommended in ICRU Report 37 ICR84 bringing the model into closer ICRU compliance The density effect correction now uses the parameterization of Sternheimer and Peierls STE71 Additional improve ments to the density effect calculation recommended in ICRU Report 37 will be incorporated in a future release For high energy protons and other light charged projectiles the approximate SPAR model ARM73 has been replaced with a full implementation of the maximum kinetic energy transfer For intermediate energies the shell corrections to the stopping power have been adapted from Janni JAN82 Finally a continuous transition in the stopping power between the ranges 1 31 MeV AMU Atomic Mass Unit for the high energy model and 5 24 MeV AMU the low energy SPAR model is achieved with a linear interpolation between the two models 68 MCNPX User s Manual MCNPX User s Manual i Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium These new procedures provide a small but significant improvement over LAHET practice above 1 MeV AMU while offering a smoother transition to the low energy model A more detailed discussion can be found in PRA98b Due to nonlinearit
330. ing between all the boundaries If only two entries are given with the first negative and the second positive the second entry is used as the uppermost energy boundary ERGB NERG and the first entry is interpreted as the lethargy spacing between bin boundaries Thus the record bf 0 1 800 will specify ten equal lethargy bins per decade from 800 MeV down For NANG gt 0 a record is required to define the NANG upper cosine bin boundaries They should be entered from low to high with the last upper boundary equal to 1 0 the lower limit of the first bin is always 1 0 If a null record is present only a then the range 1 1 is divided into NANG equal intervals 156 MCNPX User s Manual MCNPX User s Manual Ap Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium For NANG lt 0 a record is required to define the BAR NANG BAR lower degree bin boundaries They should be entered from low to high with the last lower boundary equal to 0 0 the upper limit of the first bin is always 180 degrees If a null record is present only a then the range 180 0 is divided into BAR NANG BAR equal intervals 4 Executing XSEX3 An input file and a history file are the only required input files The default file name for the input is INXS the default file name for the output is OUTXS and the default file name for the history file is HISTP A value of KPLOT NE 0 will result in the creation of a
331. ing the tracking of high energy photons or electrons Must be followed by a single reference to a TR card that can be used to trans trans late and or rotate the entire mesh Only one TR card is permitted with a mesh card 5 7 22 5 DXTRAN Mesh Tally Type 4 The fourth type of mesh tally scores the tracks contributing to all detectors defined in the input file for the P particle type If this mesh card is preceded by an asterisk tracks contributing to DXTRAN spheres are recorded Obviously a point detector or DXTRAN sphere must already be defined in the problem and the tally will record tracks corresponding to all such defined items in the problem The user should limit the MCNPX User s Manual 149 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 geometrical boundaries of the grid to focus on a specific detector or DXTRAN sphere in order to prevent confusion with multiple detectors although the convergence of the particle tracks should help in the interpretation This tally is an analytical tool useful in determining the behavior of detectors and how they may be effectively placed in the problem R C S MESHn P trans n 4 14 24 34 note number must not duplicate one used for an F4 tally P is a particle type neutron or photon There is no default see Table 4 1 Table 5 84 DXTRAN Mesh Tally type 4 Keyword Descriptions Keyword Description trans Must be followed by a si
332. ion 90 MCNPX User s Manual 5 5 7 1 Form LCA MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 LCA IELAS IPREQ IEXISA ICHOIC JCOUL NEXITE NPIDK NOACT ICEM LCA is used to select the Bertini ISABEL or CEM models as well as set certain parameters used in Bertini and ISABEL CEM is a self contained package with no internal options presently defined Table 5 41 LCA Keyword Descriptions Keyword Description IELAS 0 No nucleon elastic scattering 1 elastic scattering for neutrons only 2 elastic scattering for neutrons and protons default IPREQ 0 No pre equilibrium model will be used 1 Use pre equilibrium model after intranuclear cascade default 2 Use IPREQ 1 and IPREQ 3 randomly with an energy dependent proba bility that goes to IPREQ 3 at low energies and to IPREQ 1 at high incident energies 3 Use pre equilibrium model instead of the intranuclear cascade Note options IPREQ 2 and IPREQ 3 apply only when using the Bertini intranuclear cascade model IEXISA 0 when using the ISABEL model these options default to IPREQ 1 IEXISA 0 Do not use ISABEL intranuclear cascade model for any particle 1 Use Bertini model for nucleons and pions with ISABEL model for other particle types default 2 Use ISABEL model for all incident particle types Note The ISABEL INC model requires a much greater execution time In addi tion incident particle energies should be less than 1
333. ion 2 3 0 April 2002 LA UR 02 2607 Two physics treatments are offered the simple and detailed as described in Part D of Chapter 2 of the MCNP4B manual Table 4 5 summarizes the physics offered by these two options The simple physics treatment is intended for high energy photons where little coherent scattering occurs It is inadequate for high Z nuclides or deep penetration prob lems which the user should keep in mind when performing high energy accelerator applications Table 4 5 Summary of Photon Physics Options Process Simple used above energy EMCPF on the PHYS P card default 100 MeV Detailed used below energy EMCPF on the PHYS P card default 100 MeV Capture method analog capture used if WC1 0 on CUT P card oth erwise implicit capture used analog Fluorescence Not included accounted for after photoelec tric absorption Photoelectric Effect regarded as pure absorption by implicit capture Non cap tured weight undergoes either pair production or compton scattering Capture weight is either deposited locally or becomes a photoelectron for transport Incident photon is absorbed and 0 to 2 fluorescent pho tons emitted An orbital elec tron is ejected or excited Pair Production Considered only in the field of a nucleus threshold 1 022 MeV Same as detailed treatment Mode P E e and are cre ated photon terminates Mode P with TTB e
334. ions 13 Positrons may not be used as source particles in 2 3 0 Correcting this involves a change in the way the particle identification numbering system is handled for elec trons and positrons Historically this has not been treated in the same way as the method used for neutrons in MCNP4B which forms the basis for the multiparticle extension of MCNPx This will be corrected in a future MCNPX version 14 Beware of the results of an F6 p tally in small cells when running a photon or photon electron problem Photon heating numbers include the energy deposited by electrons generated during photon collisions but assume that the electron energy is deposited locally In a cell where the majority of the electrons lose all of their energy before exiting that cell this is a good approximation However if the cell is thin and or a large number of electrons are created near the cell boundary these electrons can carry significant energy into the neighboring cell which can result in the F6 p tally for this cell being too large This is a known problem in MCNP4B where the user is cau tioned that all energy transferred to electrons is assumed to be deposited locally MCNP4b manual page 2 73 In MCNPX the problem can be magnified because of the high energy nature of many applications and also because the F6 formalism is used in the type 3 Mesh Tally We are investigating this issue The user is also encour aged to carefully investigate the F8 tally
335. iped NOTE RPP surfaces will only be normal to X Y Z axes Form RPP Xmin Xmax YminYmax Zmin Zmax MCNPX User s Manual 63 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 10 Macrobody Rectangular Parallelepiped Argument Description Xmin Xmax termini of box sides normal to X Ymin Ymax termini of box sides normal to Y Zmin Zmax termini of box sides normal to Z Example RPP 1 1 1 1 1 1 equivalent to BOX above 5 3 2 4 3 SPH Sphere Form SPH VxVyVz R Table 5 11 Macrobody Sphere Argument Description Vx Vy Vz x y z coordinates of center R Radius incm 5 3 2 4 4 RCC Right Circular Cylinder Can Form RCC Vx Vy Vz Hx Hy Hz R Table 5 12 Macrobody Right Circular Cylinder Argument Description Vx Vy Vz x y z coordinates of center of base Hx Hy Hz cylinder axis vector R Radius incm Example RCC 0 5 0 0100 4 a 10 cm high can about the y axis base plane at y 5 with radius of 4 cm 64 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 5 3 2 4 5 RHP or HEX Right Hexagonal Prism NOTE Differs from ITS ACCEPT format Form RHP v1v2v3 h2h2h3 r1 r2r3 s1s2s3 t1t2t3 Table 5 13 Macrobody Right Hexagonal Prism HEX Descriptor Description v1 v2 v3 x y z coordinates of the bottom of the hex vector from the bottom to the top h1
336. is the particle type being tallied which may be absent depending on the type of mesh tally Up to 10 keywords are permitted depending on mesh type In MCNPxX there are four general types of mesh tally cards each with a different set of keywords 5 7 22 2 Track Averaged Mesh Tally Type 1 The first mesh type scores track averaged data flux fluence or current The values can be weighted by an MSHMF card through the DFACT dose conversion coefficient function or for energy deposition Form R C S MESHn Ptraks flux dose popul pedep mfact trans n 1 11 21 31 note number must not duplicate one used for an F1 tally P is a particle type There is no default see Table 4 1 1 The user should be warned that the mesh tally number must be different from any other tally in the prob lem For example an fl n tally will conflict with a RMESH1 n tally MCNPX User s Manual 145 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 81 Track Averaged Mesh Tally type 1 Keyword Descriptions Keyword Description traks The number of tracks through each mesh volume The average fluence is particle weight times track length divided by volume in flux units of number cm2 If the source is considered to be steady state in particles per second then the value becomes flux in number cm second default Causes the average flux to be modified by an energy dependent dose function The dose keyw
337. ission 1971 MCNPX User s Manual 223 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 224 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 9 Appendix C Using XSEX3 with MCNPX 1 Introduction XSEX3 is the code which analyzes a history file produced by LAHET3 or MCNPX and generates double differential particle production cross sections for primary beam interactions Cross section plots may also be generated by creating a file to be plotted by MCNP It is necessary to execute either code in a specific mode described below to achieve the desired cross section calculation The execution of XSEX3 assumes that the LAHET run was made using the option N1COL 1 Under this option the incident particle interacts directly in the specified material in which the source is located without any transport the only possible outcomes are a nuclear interaction or no interaction The procedure may be used to calculate double differential particle production cross sections from any of the interaction models in the code Bertini ISABEL CEM etc the procedure has no meaning if such a model is not allowed for the specified particle type at the specified energy 2 Input for MCNPX Since there is no way to avoid the MCNPX geometry input the user should define a region containing the material for which the cross sections are desired and locate the source in that region To avoid possibl
338. ith NOBAL 1 and NOBAL 3 FLIMO gt 0 Kinetic energies of secondary particles will be reduced by no more than a fraction of FLIMO in attempting to obtain a non negative excitation of the residual nucleus and a consistent mass energy balance A cascade will be re sampled if the correction exceeds FLIMO FLIMO 0 No correction will be attempted and a cascade will be re sampled if a negative excitation is produced FLIMO lt 0 default 1 0 The maximum correction is 0 02 for incident energy above 250 MeV 0 05 for incident energy below 100 MeV and is set equal to 5 incident energy between those limits As an example consider LCB 94 3000 3000 2000 2000 1000 1000 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 For IEXISAQ 1 the default nucleons will switch to the BERTINI model from the FLUKA model below 3 GeV and Pions would switch below 2 GeV Kaons and anti nucleons would switch to the ISABEL model from the FLUKA model below 1 GeV lons use only the ISABEL model and muons have no nuclear interactions For IEXISA 2 nucleons and pions would also switch to the ISABEL model below 1 GeV Note that the nominal upper energy limit for the ISABEL model is about 1 GeV nucleon it may actually execute at higher energies without crashing but with diminished validity 5 5 7 3 LEA Form LEA IPHT ICC NOBALC NOBALE IFBRK ILVDEN IEVAP NOFIS LEA controls evaporation fermi breakup
339. ition it provides surface flux and current edits which supplement the standard MCNP tallies HTAPE3X is an adaptation of the LAHET Code System HTAPE code Details may be found in User Guide to LCS PRA89 and the man ual as written for use in MCNPX is reproduced in Appendix B of this document The user should note the following comments since HTAPE3X does not contain any pro vision for many of the termination options allowed by MCNPX which affect the content of the HISTP file The user must be aware of the possible implications on normalization of outputs HTAPE3X will correctly process HISTP for the following cases 1 Normal completion after NPS histories N NPS is used for the degrees of freedom in the statistical analysis to compute means and variances 2 Termination is by 4c k or system crash HISTP lacks a final record N is taken to be the highest observed history number this is a good approximation if N is large and most histories contribute to the HISTP file However other modes of termination of the MCNPX produce the following results 3 Termination by 4c q with NPS input record present The correct N is unknown to HTAPES3X and NPS is used The user may normalize the HTAPE3X output by the ratio NPSIN but the calculated variances will not reflect this correction 4 Termination on time using CTME when NPS input record is present See comment 3 above 5 If an NPS record is absent HTAPE3X will crash in the termination
340. itle tritons MeV file free c linlog xlims 1 0 1 0 ytitle tritons steradian file tally 108 free e loglog xlims 0 1 1000 ytitle He 3 MeV file free c linlog xlims 1 0 1 0 ytitle He 3 steradian file tally 109 free e loglog xlims 0 1 1000 ytitle alphas MeV file free c linlog xlims 1 0 1 0 ytitle alphas steradian file tally 110 free e loglog xlims 0 1 100 ytitle photons MeV file free c linlog xlims 1 0 1 0 ytitle photons steradian file end MCNPX User s Manual 159
341. itons 3He and alphas denoted by d t s and a respectively The only differences between the two input decks are the two cards Base Case mode nh imp n h 1 1r 0 Case 2 mode nh dtsa imp n h d t s a 1 1r 0 Note that nuclear interactions by light ions are simulated using the ISABEL INC model The problem summary for this case is shown below sample problem spallation target Case 2 neutron creation tracks weight energy neutron loss tracks weight energy per source particle per source particle source 0 0 QO escape 366756 1 8321E 01 2 1938E 02 nucl interaction 316952 1 5848E 01 3 2187E 02 energy cutoff 0 0 0 particle decay 0 Ox 0 time cutoff 0 0 0 weight window 0 0 0 weight window 0 0 0 cell importance 0 0 0 cell importance 0 0 0 weight cutoff 0 0 0 weight cutoff 0 0 0 energy importance 0 0 0 energy importance 0 0 0 dxtran 0 0 0 dxtran 0 0 0 forced collisions 0 Qs 0 forced collisions 0 0 0 exp transform 0 QO QO exp transform 0 0 QO upscattering 0 0 0 downscattering 0 0 9 8368E 00 tabular sampling 0 QO QO capture 0 1 4534E 02 7 7278E 02 n xn 79010 3 9467E 00 1 9031E 01 loss to n xn 25539 1 2753E 00 4 9548E 01 fission 0 0 0 loss to fission 0 0 0 photonuclear 0 0 nucl interaction 3667 1 8335E 01 6 2061E 01 MCNPX User s Manual 129 MCNPX User s Manual E Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium tabular boundary 0 QO QO tabular boundary
342. itted image projection The radiography capability is based on point detector techniques and is extensively described in SNO96 and SNO98 In essence the radiography focal plane grid is an array of point detectors 8 2 1 Pinhole Image Projection In the pinhole image projection case a point is defined in space that acts much like the hole in a pinhole camera and is used to focus an image onto a grid which acts much like the photographic film The pinhole is actually a point detector and is used to define the direction cosines of the contribution that is to be made to the grid The pinhole position rel ative to the grid is also used to define the element of the grid into which this contribution is scored Once the direction is established a ray trace contribution is made to the grid bin with attenuation being determined for the material regions along that path The source need not be within the object being imaged nor does it need to produce the same type of particles that the detector grid has been programmed to score The grid and pinhole will image either source or scattered events produced within the object see NOTRN card in Section 8 2 3 for either photons or neutrons These event type contributions can be binned within the grid tallies by binning as source only total or by using special binning relative to the number of collisions contributing cells etc The pinhole image projection is set up as follows in version 2 1 5 Fin P X1 Y1
343. ium limit is 1 GeV per nucleon Running time is generally 5 10 times greater per collision than with the Bertini model A new third model is now offered in MCNPX version 2 3 0 the CEM code We do note that running times for this code are long however a new version will be issued in a future ver sion which substantially speeds up the code 4 1 2 Multistage Pre equilibrium Models MPM Subsequent de excitation of the residual nucleus after the INC phase may optionally employ a multistage multistep preequilibrium exciton model or MPM PRA88 The MPM is invoked at the completion of the INC with an initial particle hole configuration and exci tation energy determined by the outcome of the cascade At each stage in the MPM the excited nucleus may emit a neutron proton deuteron triton He or alpha alternatively the nuclear configuration may evolve toward an equilibrium exciton number by increasing the exciton number by one particle hole pair The MPM terminates upon reaching the equi librium exciton number at which point an evaporation or Fermi Breakup model is then applied to the residual nucleus with the remaining excitation energy In the LAHET Bertini model the inverse reaction cross sections are represented by the parameterization of Chatterjee The potentials from which the inverse reaction cross sec tions are obtained are those selected by Kalbach KAL85 for the PRECO D2 code When the ISABEL intranuclear cascade model is invo
344. ium Project APT The work involved a formal extension of MCNP to all particles and all energies improvement of physics simulation models extension of neutron proton and photonuclear libraries to 150 MeV and the formulation of new variance reduction and data analysis techniques The program also included cross section measurements benchmark experiments deterministic code development and improvements in transmutation code and library tools through the CINDER 90 project Since the closure of the APT project work on the code has continued under the sponsorship of the Advanced Accelerator Applications AAA and other programs Since the initial release of MCNPX version 2 1 on October 23 1997 an extensive beta test team has been formed to test the code versions prior to official release Approximately 900 users in approximately 200 institutions worldwide have had an opportunity to try the improvements in this version and to provide feedback to the developers This process is invaluable and we express our deepest appreciation to the participants in the beta test program Applications for the code among the beta test team are quite broad and constantly developing Examples include e Design of accelerator spallation targets particularly for neutron scattering facilities e Investigations for accelerator isotope production and destruction programs including the transmutation of nuclear waste e Research into accelerator driven energy sour
345. ive errors are all written by MCNPX to an unformatted binary file named mdata This file is overwritten each time a dump is written to the runtpe file Because of this overwrite in doing a restart of MCNPX with a mesh tally one must always use the last complete dump on the runtpe file The gridconv program is a post processing code used with the mdata output file It can also be used with the mctal output file from the radiography tally as described in Section 8 2 Gridconv converts the data arrays in mdata to forms compatible with various external graphics packages Those supported in MCNPX version 2 3 0 are PAW PAW Physics Analysis Workstation is distributed through the CERN Program Library http Avwwinfo cern ch asd paw index html 100 MCNPX User s Manual MCNPxX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium e IDL IDL Interactive Data Language is a product of Research Systems Inc 4990 Pearl East Circle Boulder Co 80301 http www rsinc com idl index cfm e Tecplot Tecplot is a product of Amtec Engineering Inc 13920 SE Eastgate Way Ste 220 Bellevue Wa 98005 http Awww amtec com e GNUPLOT Freeware http www gnuplot info Only 1 and 2 d plots supported Like MCNPX gridconv will compile on several platforms However currently the PAW part of the code will not compile on the Linux operating system since some of the PAW subroutines needed by the code a
346. ked it is possible to determine explicitly the particle hole state of the residual nucleus since a count of the valid excitations from the Fermi sea and the filling of existing holes is provided To define the initial con ditions for the MPM the number of particle hole pairs is reduced by one for each intranuclear collision for which both exiting nucleons are below the top of the nuclear potential well This method is the only option implemented in MCNX to link the MPM with the ISABEL INC In adapting the MPM to the Bertini INC it has not been possible yet to extract the same detailed information from the intranuclear cascade history Consequently the algorithm which defines the interface between the Bertini INC and the MPM is a rather crude approx imation intended to permit initial evaluation of the MPM but open to further improvement In this case the initial condition for the MPM is one particle hole pair beyond the minimum particle hole configuration allowed by the outcome of the INC The adaptive algorithm used with ISABEL is quite effective However given the initial condition algorithm used with the Bertini INC the user has a choice of invoking the MPM in one of three optional modes or not at all 3 The MPM continues from the final state of the INC with the initial condition defined as above normal MPM 42 MCNPX User s Manual MCNPX User s Manual i Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Productio
347. king information presentation and will soon issue a revision which incorporates a dictionary type lookup system for card definitions Until the complete set of manuals is issued we recommend using this document in tandem with the MCNP4C manual and the previously issued MCNPX 2 3 0 User s Manual It is hoped that MCNPX will be of use to the Monte Carlo radiation transport community in general and we are already seeing major applications in medical and space science fields also in areas where tracking of low energy charged particles is important The develop ment of the modular approach in future versions of the code will facilitate the addition of new capabilities to the base code and make this tool a flexible reliable aid in the explora tion of both traditional and new mixed energy multiparticle applications Laurie Waters Deputy Group Leader D 10 Nuclear Systems Design Los Alamos National Laboratory September 2002 xiv MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 1 Introduction MCNP xX is a general purpose Monte Carlo radiation transport code that tracks all particles at all energies It is the next generation in the series of Monte Carlo transport codes that began at Los Alamos fifty years ago MCNPX 2 4 0 is a superset of MCNP4C3 and MCNPX 2 3 0 LAHET 2 8 and CEM The MCNPX program began in 1994 as an extension of MCNP and LAHET in support of the Accelerator Production of Trit
348. l of the macro defi nitions found in the configure in files flags m4ac a file that is included in aclocal m4 it localizes the setting of flags for differ ent combinations of architecture operating system compilers checks m4 a file that is included in aclocal m4 it checks for the required version of gnu make and exits with instructions if not found configure generic in a shared file used to generate config ure scripts for the last level of the file tree install sh a shared header file template for the Makefile that all of the levels will use Makefile h in a shared header file template for the Makefile that all of the levels will use config guess a script that aids recognition of com puting environments when configure is run config sub a script that aids validation and canonicalization of a computing envi ronments when configure is run First Second and Third Level directories all contain special configure in files that prop agate the automated configuration down to the next levels The Fourth Level directories each share the configure generic in file in the config directory because there is no further propagation Each of the levels 1 4 also contain a special Makefile in and Makefile h in files When the configure script runs Makefile h is generated then the Makefile is generated The first line of each Makefile includes Makefile h MCNPX User s Manual 27 MCNPX User s Manual E
349. lanes perpendicular to the z axis Bins do not have to be equally spaced In the case of the cylindrical mesh the middle coordinate CORBn is the untransformed z axis which is the symmetry axis of the cylinder with radial meshes defined in the CORAn input line The first smallest radius may be equal to zero The values following CORBn define planes perpendicular to the untransformed z axis The values following CORCn are positive angles relative to a counter clockwise rotation about the untrans formed z axis These angles in degrees are measured from the positive x axis and must have at least one entry of 360 which is also required to be the last entry The lower limit of zero degrees is implicit and never appears on the CORCn card In the case of spherical meshes scoring will happen within a spherical volume and can also be further defined to fall within a conical section defined by a polar angle relative to the z axis and azimuthal angle CORAn is the radius of the sphere CORBn is the polar angle and CORCn is the same as in the cylindrical case It is helpful in setting up spherical problems to think of the longitude latitude coordinates on a globe The original capability of MCNP involving the i option is retained allowing a large number of regularly spaced mesh points to be defined with a minimum of entries on the coordinate lines All of the coordinate entries must be monotonically increasing for the tally mesh fea tures to
350. le named mdata This file is overwritten each time a dump is written to the runtpe file Because of this overwrite in doing a restart of MCNPX with a mesh tally one must always use the last complete dump on the runtpe file The gridconv program is a post processing code used with the mdata output file It can also be used with the mctal output file from the radiography tally as described in Section 8 2 Gridconv converts the data arrays in mdata to forms compatible with various external graphics packages Those supported in MCNPX are PAW PAW Physics Analysis Workstation is distributed through the CERN Program Library http Awwwinfo cern ch asd paw index html IDL IDL Interactive Data Language is a product of Research Systems Inc 4990 Pearl East Circle Boulder Co 80301 http Awww rsinc com idl index cfm Tecplot Tecplot is a product of Amtec Engineering Inc 13920 SE Eastgate Way Ste 220 Bellevue Wa 98005 http www amtec com GNUPLOT Freeware http www gnuplot info Only 1 and 2 d plots supported Like MCNPX gridconv will compile on several platforms However currently the PAW part of the code will not compile on the Linux operating system since some of the PAW subroutines needed by the code are not Linux compatible Gridconv may be compiled with a nopaw option Once gridconv is compiled one need type only the word gridconv to execute the code The code will then prompt the user for information t
351. le type P In contrast to the 3rd type of Mesh Tally energy dep osition can be obtained in this option for any particular particle This option allows one to score the equivalent of an F6 P see Section 8 3 heating tally for the particle type P Note the mesh is independent of problem geometry and a mesh cell may cover regions of several different masses Therefore the normalization of the pedep option is per mesh cell volume not per unit mass MCNPX User s Manual 95 Accelerator Production of Tritium Table 8 1 MCNPxX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 Track Averaged Mesh Tally type 1 Keyword Descriptions Continued Keyword Description mfact Can have from one to four numerical entries following it The value of the first entry is in reference to an energy dependent response function given on a MSHMFn card no default e The second entry is 1 default 1 for linear interpolation and 2 for logarith mic interpolation Ifthe third entry is zero default 0 the response is a function of energy deposited otherwise the response is a function of the current particle energy e The fourth entry is a constant multiplier and is the only floating point entry allowed default 1 0 If any of the last three entries are used the entries preceding it must be present so that the order of the entries is preserved Only one mfact keyword may be used per tally trans
352. lear interactions of source particles only transport and slowing down are turned off This option is for use in computing double differ ential particle production cross sections with the XSEX code See Appendix C ICEM 0 Use the Bertini or ISABEL model determined by the IEXISA parameter default 1 Use the CEM model MCNPX User s Manual 79 Accelerator Production of Tritium MCNPX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 LCB FLENB1 BLENB2 FLENB3 FLENB4 FLENB5 FLENB6 CTOFE FLIMO LCB controls which physics module is used for particle interactions depending on the kinetic energy of the particle Table 6 4 LCB Keyword Descriptions Keyword Description FLENB1 Kinetic Energy Default 3500 MeV For nucleons the Bertini INC model will be used below this value FLENB2 Kinetic Energy Default 3500 MeV For nucleons the FLUKA high energy generator will be used above this value Note The probability for selecting the interaction model is interpolated linearly between FLENB1 and FLBEN2 Note The version of FLUKA used in MCNPX version 2 3 0 should not be used below 500 MeV c momentum Note For nucleons the Bertini model switches to a scaling procedure above 3 495 GeV wherein results are scaled from an interaction at 3 495 GeV Although both models will execute to arbitrarily high energies a plausible upper limit for the Bertini scaling law is 10 GeV FLENB3 Kin
353. lent to 001 01 1 1 10 100 MCNPX User s Manual 35 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 These features apply to both integer and floating point quantities If n an integer is omitted in the constructs nR nl nLog and nJ then n is assumed to be 1 If x integer or floating point is omitted in xM it is a fatal error The rules for dealing with adjacent special input items are as follows nR must be preceded by a number or by an item created by R or M 2 nlandnLOG must be preceded by a number or by an item created by R or M and must be followed by a number 3 xM must be preceded by a number or by an item created by R or M 4 nJ may be preceded by anything except and may begin the card input list Examples 13M 2R 1333 1 3M I 4 133 54 1 3M 3M 1 39 1 2R 212 5 1111 52 02 5 1 R2M 1 12 1RR 111 1 214 3M 1 23412 1 214 21 10 123468 10 3J AR is illegal 1 41 3M is illegal 1 4l Jisillegal 4 1 9 Vertical Inout Format Column input is particularly useful for cell parameters and source distributions Cell importances or volumes strung out on horizontal input lines are not very readable and often cause errors when users add or delete cells In vertical format all the cell parameters for one cell can be on a single line labeled with the name of the cell If a cell is deleted the user deletes just one line of cell parameters instead of hunting for the data item that belongs to the cell in
354. ler in use the mcenpx_2 3 0 config aclocal m4 and the mecnpx_2 3 0 config flags m4 macro definition files and the various configure in files will be needed The configure in files determine the order in which the macros in the aclocal m4 file are activated The order of the macro calls is very important as some macros assume that prior work has been done There are configure in files in the following directories e configure in in the menpx_2 3 0 directory e configure in in the menpx_2 3 0 sre directory e configure in in the menpx_2 3 0 src menpx directory e configure generic in in the menpx_2 3 0 config directory All of the configure in files contain the same order of macro invocation The arch and system variables are set by a call to AC_SET_ ARCH from configure in The macro definition of AC_SET_ARCH in aclocal m4 uses AC_CANONICAL_SYSTEM which in turn uses config guess and or config sub to do its work to set our ARCH and SYSTEM variables These variables are then used in combination with the FCOMP vari able that specifies which Fortran compiler to use WARNING Assumptions are made that an expected compatible C compiler to match the Fortran compiler has been used You will receive warnings if the Fortran C combination is questionable Find the AC_ FLAGS BY ARCH SYS COMP macro call in the aclocal m4 file The cor responding definition for the AC_FLAGS_ BY _ARCH_SYS_ COMP macro is contained in its own file called flags m4 The fl
355. lerator Production of Tritium 22 The Response Function Option Any non zero value of the IRSP parameter allows the user to apply an energy dependent response function F E where E is the particle energy to the current and flux tallies given by edit option types 1 2 4 9 10 and 13 The user supplies a tabulation of the function F E by the pairs of values FRESP I ERESP I which are input as the arrays ERESP l l 1 NRESP and FRESP I l 1 NRESP described in Section 2 above The element IRESP I of the third input array then specifies an interpolation scheme for com puting the response function value within the interval ERESP I lt E lt ERESP I 1 For IRSP gt 0 the interpolated response function value multiplies the tally increment for IRSP lt 07 it divides the tally increment There are five interpolation schemes that may be specified individually for each energy interval in the response function tabulation using the following values for IRESP I 1 Constant the response function value is the value at the lower energy of the interval 2 Linear linear the response function is interpolated linearly in energy 3 Linear log the response function is interpolated linearly in the logarithm of the energy 4 Log linear the logarithm of the response function is interpolated linearly in energy 5 Log log the logarithm of the response function is interpolated linearly in the log arithm of the energy Any value o
356. light nuclei BRE89 4 1 4 Evaporation Model MCNPX when used with the Bertini or ISABEL options employs the Dresner evaporation model based on work originally due to Weisskopf After the INC MPM stage residual nuclei are in highly excited states and energy is dissipated by evaporation of n p d t 3He and a particles The probability p e that an excited nucleus will emit a particle x with kinetic energy e is proportional to Where S and m are the spin and mass of particle x scx is the cross section for formation of the compound nucleus in the inverse reaction bombarding the residual nucleus with particles of energy e E is the excitation of the residual nucleus and w E is the density of levels of the residual nucleus at excitation E A discussion of level density options is given in section 4 1 5 below Although the Dresner model can emit 19 different particles from a nucleus only those with Z up to 2 are implemented in MCNPX The probability of emission of a particle is given by MCNPX User s Manual 43 MCNPX User s Manual E Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium U Q 6 R 28 1 m JESU Q e de kV Q is the binding energy of the particle in the nucleus and kx are taken from inverse cross section parameterizations for each particle V is the Coulomb barrier and U is the initial excitation energy These integrals have been solved analytically for different particles
357. limited 140 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Notes Use MAT only if the perturbation changes the material from one cell material to another Use with caution especially if more than one nuclide in the material is changed New nuclide can not be added in the new material card RHO gt 0 perturbed atom density lt 0 perturbed gram density METHOD gt 0 print change in tally lt O print perturbed tally 1 2 3 1st amp 2nd order 1st order 2nd order perturbation calculation Limitations 1 Large gt 30 perturbations may be wrong if the 2nd order Taylor Series expansion is insufficient Try looking at 1st and 2nd order terms separately for large perturba tions SILENT no warning error message 2 Nuclide fraction changes MAT option are assumed independent Differential cross terms are ignored SILENT 3 FM tallies in perturbed cells can be wrong Surface tallies and tallies in perturbed cells are safe WARNING 4 Detectors and pulse height tallies fail zero perturbation 5 DXTRAN fails fatal error 6 Cannot unvoid a region fatal error 7 Cannot introduce a new nuclide into the perturbation fatal error 8 Perturbations increase running time 10 20 each 9 Some perturbations converge slowly small and ones 10 Limited to n p problems Examples of the PERT Card Example 1 PERT1 n p CELL 1 RHO 0 03 This perturb
358. linear y axis LINLOG Use linear x axis and logarithmic y axis This is the default LOGLIN Use logarithmic x axis and linear y axis LOGLOG Use logarithmic x axis and logarithmic y axis nsteps nsteps XLIMS min max YLIMS min max Define the lower limit upper limit and number of subdivisions on the x or y axis nsteps is optional for a linear axis and is ineffective for a logarithmic axis In the absence of any specification by the user the values of min max and nsteps are defined by an algorithm in MCNPX HIST SCALES n Put scales on the plots according to the value of n 0 no scales on the edges and no grid 1 scales on the edges the default 2 scales on the edges and a grid on the plot Make histogram plots This is the default if the independent variable is cosine energy or time PLINEAR Make piece wise linear plots This is the default if the independent variable is not cosine energy or time SPLINE x Use spline curves in the plots If the parameter x is included rational splines of tension x are plotted Otherwise Stinem and cubic splines are plotted Rational splines are available only with the DISSPLA graphics system BAR MCNPX User s Manual Make bar plots 57 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 3 MPLOT amp MCPLOT Commands Command Description NOERRBAR Suppr
359. list Table 5 7 argument meaning surface number 1 lt j lt 99999 If surface defines a cell that is transformed with TRCL 1 lt j lt 999 See Section absent for no coordinate transformation or number of TRn card a the letter X Y or Z list one to three coordinate pairs 5 3 2 3 General Plane Defined by Three Points Form j n P X7V4Z4X2V2Z0X3V3Z3 MCNPX User s Manual 62 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 8 General Plane Defined by Three Points argument meaning surface number 1 lt j lt 99999 or lt 999 if repeated structure absent or 0 for no coordinate transformation n gt 0 specifies number of a TRn card lt 0 specifies surface j is periodic with surface n P indicates this is a plane X Y Z coordinates of points to define the plane 5 3 2 4 Surfaces Defined by Macrobodies 5 3 2 4 1 BOX Arbitrarily oriented orthogonal box Note all corners are 90 Form BOX VxVyVz A1xAtyA1z A2xA2yA2z A3x A3y A3z Table 5 9 Macrobody BOX Argument Description Vx Vy Vz X y Z coordinates of corner Aix Aly A1z vector of 1st side A2x A2y A2z vector of 2nd side A2x A3y A3z vector of 3rd side Example BOX 1 1 1 200 020 002 a cube centered at the origin 2 cm ona side sides parallel to the major axes 5 3 2 4 2 RPP Rectangular Parallelep
360. lo Simulation methods Tabular data whose evaluation contains a careful con sideration of nuclear structure effects forms a convenient area of low energy phenomena In the intermediate range above the nuclear structure region 150 MeV in MCNPX to a few GeV the most common modeling methods include intranuclear pre equilibrium evaporation models Above the natural limitations of INC physics other meth ods involving quantum effects are used and MCNPX version 2 3 0 contains an early version of the FLUKA code to handle high energy interactions Although our knowledge of particle physics increases constantly in sophistication it is notable that a number of long used techniques are still employed in the intermediate and high energy regions Their speed of execution is the primary factor for retention of these models There is however a small but growing trend to use the more complex models to extend tabular data to high energy regimes In addition to improvements in computational time an additional benefit of extended tabular data is to facilitate the use of certain vari ance reduction techniques at all energies However much research still needs to be done to validate high energy data to the accuracy that low energy regimes can now achieve MCNPX will be able to handle appropriately processed tabular data as it increases in upper energy limit however we will also retain the option to use intermediate and high energy physics modules Th
361. loyed MCNPX User s Manual 157 Accelerator Production of Tritium MCNPxX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 Type Particle 1 proton 2 neutron 3 pi 4 pid 5 pi 6 deuteron 7 triton 8 He 3 9 alpha 10 photon prompt gamma from residual 11 K 12 K all neutrals 13 K 14 antiproton 15 antineutron 16 elastic scattered projectile An example of a COMOUT file produced when plotting XSTAL is shown on the next page 158 MCNPX User s Manual MCNPxX User s Manual i Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium rmctal xstala nonorm tally 101 free e loglog xlims 0 1 1000 ytitle protons MeV file free c linlog xlims 1 0 1 0 ytitle protons steradian file tally 102 free e loglog xlims 0 1 1000 ytitle neutrons MeV file free c linlog xlims 1 0 1 0 ytitle neutrons steradian file tally 103 free e loglog xlims 0 1 1000 ytitle pi MeV file free c linlog xlims 1 0 1 0 ytitle pi steradian file tally 104 free e loglog xlims 0 1 1000 ytitle pi0 MeV file free c linlog xlims 1 0 1 0 ytitle pi0 steradian file tally 105 free e loglog xlims 0 1 1000 ytitle pi MeV file free c linlog xlims 1 0 1 0 ytitle pi steradian file tally 106 free e loglog xlims 0 1 1000 ytitle deuterons MeV file free c linlog xlims 1 0 1 0 ytitle deuterons steradian file tally 107 free e loglog xlims 0 1 1000 yt
362. lumes areas or masses as appropriate obtained from a MCNP calculation When IOPT gt 100 the NPARM cell surface or material identifiers are treated as a single entity in constructing a tally edit In this case the NPARM normalization divisors are summed to a single divisor Consequently one may supply the full list of divisors if appropriate or just supply one value for the common tally For IRS gt 0 the original source definition record in LAHET format as described in Section 2 4 of reference 1 followed by the new source definition record also in LAHET format ForlTCONV 0 a LAHET source time distribution record as described in Section 2 4 of reference 1 e For IRSP 0 three records defining the user supplied response function ERESP 1 1 NRESP a monotonically increasing energy grid on which the value of the response function is tabulated FRESP I l 1 NRESP the values of the response function at the above energies IRESP I l 1 NRESP 1 interpolation scheme indicators where IRESP I indicates the interpolation scheme to be used for the response function in the I th energy interval The length NRESP lt 200 is obtained from the array ERESP input terminated by a The user must maintain the proper correspondence among the arrays see Section 22 below e Any additional input required for the particular option For basic option types 1 2 or 11 this may be the specification of surface s
363. ly intended to provide an analysis of the outcome of collisions in the medium and high energy range where the interaction physics is obtained from LAHET However all appropriate features have been retained even when they duplicate existing MCNP flux and current tallies 3 The latter features relate to editing a surface source write SSW file default name WSSA For experienced LAHET users they do provide some options not available with standard MCNP F1 and F2 tallies Note that the information written to HISTP comes only from interactions processed by the medium and high energy modules in MCNPX low energy neutron and proton and any photon electron collisions which utilize MCNP library data do not contribute to the collision information on the history file and will not contribute to edits by HTAPESX of collision data Surface crossing edits from data on the file WSSA will apply to all particle types and all energies 2 Input for HTAPE3X The input structure is largely unchanged from the description in reference 1 In general energy units are MeV time units are nanoseconds and length units are centimeters Note the difference in the time scale from MCNP practice The input file default name INT for HTAPE3X has the following structure 1 Two records of title information 80 columns each 2 An option control record 3 Additional input as required by the chosen option MCNPX User s Manual 135 MCNPX User s Manual E V
364. m bered electron tally Plot a perturbation associated with a tally where n is a PERT n number on a PERTn card PERT 0 will reset PERT n Suppress bin normalization The default in a 2D plot is to divide the tallies by the bin widths if the independent vari NONORM able is cosine energy or time However also see the description of the MCTAL file Bin normalization is not done in 3D or contour plots Multiply the data for axis a by the factor f and then add the term s a is x y Or Z FACTOR afs s is optional If s is omitted it is set to zero For the initial curve of a 2D plot reset the axis limits XLIMS or YLIMS to the default values FACTOR affects only the current curve or plot MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 3 MPLOT amp MCPLOT Commands Command Description Reset the parameters of command aa to their default values aa can be a parameter setting command COPLOT or ALL If aa is ALL the parameters of all RESET aa parameter setting commands are reset to their default values After a COPLOT command only COPLOT ALL or any of the parameter setting commands that are marked with an in this list may be reset Resetting COPLOT or ALL while COPLOT is in effect causes the next plot to be an initial plot Titling commands The double quotes are required Use aa as line n of the main title at the to
365. m will be the same as the 3rd entry on the SP cards The parameter a in Table 10 differs from the parameter a above by a factor of the square root of 2 This is a legacy item from the conversion of the 41 function from time to space and will be corrected in a future version The user generally does not want the beam Gaussian to extend infinitely in x and y there fore a cookie cutter option has been included to keep the distribution to a reasonable size CCC ccc tells MCNPX to look at the card labeled ccc ccc is a user specified cell num ber to define the cutoff volume The first entry on the ccc card is 0 which indicates a void cell The second number nnn nnn again is a user specified number indicates a surface card within which to accept particles In the example this is a SQ surface a 2 sheet hyper boloid is defined as follows MCNPX User s Manual 85 MCNPX User s Manual E Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium ROEG Any particle generated within this cell is accepted any outside of the cell is rejected Any well defined surface may be selected and it is common to use a simple cylinder to repre sent the extent of a beampipe In this example a source is generated in an x y coordinate system with the distribution centered at the origin and the particles travelling in the z direction The particle coordi nates can be modified to an x y coordinate system by
366. main histp 3 1 5 User s Notes Do not edit the Makefiles generated by the configure script In order to change the contents of the generated Makefiles you must alter the contents of several input files that the con figure script uses Please read the Programmer s Notes in the next subsection for instructions Table 3 1 contains options which are available for use as parameters to the configure script for mcnpx 2 3 0 22 MCNPX User s Manual Accelerator Production of Tritium Option Syntax Effect on the generated Makefile if requested MCNPX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 Table 3 1 Configure Script Parameters Effect on the generated makefile if NOT requested with FC value substitute the desired Fortran77 compiler name for the value placeholder e g with FC fort to use the fort compiler compile step for the gener ated Makefiles value will be used to compile Fortran source code location of binary directory containing value must be in your PATH environment variable with STATIC linking of the compiled files STATIC is the default cannot results in a static archive be used at the same time as mcnpx a SHARED with SHARED linking of the compiled files STATIC is used this option is results in a dynamically linked exploratory for future releases executable mcnpx so of MCNPX with DEBUG a debug switch appears inthe no debug switch appears in
367. meterization of neu tron Absorption Cross Sections NASA Technical Paper 3656 June 1997 VAV57 P V Vavilov lonization Losses of High Energy Heavy Particles Soviet Phys ics JETP 5 No 5 1957 749 WHI99 M C White R C Little and M B Chadwick Photonuclear Physics in MCNPX X Proceedings of the ANS meeting on Nuclear Applications of Accelerator Technology Long Beach California November 14 18 1999 WHIO0O M C White User Interface for Photonuclear Physics in MCNP X X5 MCW 00 88 U Los Alamos National Laboratory July 26 2000 and March 21 2001 revised MCNPX User s Manual 123 MCNPX User s Manual Ap Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium WIL97 W B wilson et al CINDER 90 code for Transmutation Calculations Pro ceedings of the International Conference on Nuclear Data for Science and Technology Trieste 19 24 May 1997 Italian Physical Society Bologna p 1454 1997 YAR79 Y Yariv and Z Fraenkel Phys Rev C 20 1979 2227 YAR81 Y Yariv and Z Fraenkel Phys Rev C 24 1981 488 YOU98 G Young E D Arthur and M B Chadwick Comprehensive Nuclear Model Calculations Theory and Use of the GNASH Code Proceedings of the IAEA Workshop on Nuclear Reaction Data and Nuclear Reactors Physics Design and Safety Trieste Italy April 15 May 17 1996 124 MCNPX User s Manual MCNPxX User s Manual Version 2 3 0 Apr
368. metry will probably show up as dashed lines The intersection of a surface with the plot plane is drawn as a dashed line if there is not exactly one cell on each side of the surface at each point Dashed lines can also appear if the plot plane happens to coincide with a plane of the problem if there are any cookie cutter cells in the source or if there are DXTRAN spheres in the problem Set up and run a short problem in which your system is flooded with particle tracks from an external source The necessary changes in the INP file are as follows 1 Add a VOID card to override some of the other specifications in the problem and make all the cells voids turn heating tallies into flux tallies and turn off any FM cards 2 Add another cell and a large spherical surface to the problem such that the surface surrounds the system and the old outside world cell is split by the new surface into two cells the space between the system and the new surface which is the new cell and the space outside the new surface which is now the outside world cell Be sure that the new cell has nonzero importance Actually it is best to make all nonzero importances equal If the system is infinite in one or two dimensions use one or more planes instead of a sphere 3 Replace the source specifications by an inward directed surface source to flood the geometry with particles SDEFSUR mNRM 1 where m is the number of the new spherical surface added in Step 2 I
369. mmands Command Description Specifies the interval between calls to MCPLOT to be every n histories In KCODE calculation interval is every n FREQ n cycles If n is negative the interval is in CPU minutes If n 0 MCPLOT is not called while MCNP is running histo ries The default is n 0 If MCPLOT was called by MCNPX while running histories or by PLOT while doing geometry plotting control returns RETURN to the calling subroutine Otherwise RETURN has no effect PLOT Call or return to the PLOT geometry plotter Use with COM aaaa option Hold each picture for n sec PAUSE n onds If no n value is provided each picture remains until the return key is pressed END Terminate execution of MCPLOT Inquiry Commands When one of these commands is encountered the requested display is made and then MCPLOT waits for the user to enter another line which can be just a carriage return before resuming The same thing will happen if MCPLOT sends any kind of warning or com ment to the user as it prepares the data for a plot OPTIONS or or Display a list of the MCPLOT command keywords HELP STATUS Display the current values of the plotting parameters PRINTAL Display the numbers of the tallies in the current RUNTPE or MCTAL file Display the IPTAL array for the current tally This array tells IPTAL how many elements are in each dimension of the current 8 dimensional tally Display the
370. model Case 2 also runs slower since the light ion interactions are pro 132 MCNPX User s Manual MCNPX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium vided by the ISABEL model Invoking the 150 MeV proton libraries slows execution by about 11 in this example Table A 2 Results Compiled for Summary Cases ar Runtime Case Variation from base case n p minutes base n a 27 66 18 263 1 LAHET transport for 20 150 MeV 28 44 18 364 neutrons 2 light ion transport amp nuclear inter 33 55 18 335 action 3 ISABEL INC for nucleons and 31 91 17 569 pions 4 evaluated data used for protons 30 66 18 285 below 150 MeV 5 CEM INC for nucleons and pions 60 14 15 638 a Cases were run on an IBM AIX box This example demonstrates how to calculate neutron production from a spallation target Use of the new LA150 library that extends evaluated nuclear data up to 150 MeV gives the most accurate results particularly if the new proton evaluations are used in addition to the neutron evaluations When the quantity of interest depends only on neutrons and one starts with a proton beam there is no need to transport any particles other than protons neutrons and charged pions as neutron production by other particles is negligible com pared to production by these three particle types Use of the various LAHET physics model options such as the ISABEL and CEM INC m
371. n 17 average col abs trk len k effand one estimated standard deviation by cycle skipped Can not plot fewer than 10 active cycles 18 average col abs trk len k eff figure of merit 19 average col abs trk len k eff relative error Commands for cross section plotting XSm Plot a cross section according to the value of m Mn a material card in the INP file Example XS M15 The available materials will be listed if a material is requested that does not exist in the INP file Z a nuclide ZAID Example XS 92235 50C The full ZAID must be provided The available nuclides will be listed if a nuclide is requested that does not exist in the INP file Print out a cross section plotting primer MTn Plot reaction n of material XS m The default is the total cross section The available reaction numbers can be caused to list by entering a reaction number that doesn t exist e g 999 MCNPX User s Manual 56 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 3 MPLOT amp MCPLOT Commands Command Description PAR p Plot the data for particle type p where p can be n p eorh of material Mn The default is the source particle type for XS Mn For XS z the particle type is determined from the data library type For example 92000 01g defines PAR p Must be first entry on line LINLIN Commands that specify the form of 2D plots Use linear x axis and
372. n However if the cell is thin and or a large number of electrons are created near the cell boundary these electrons can carry sig nificant energy into the neighboring cell which can result in the F6 p tally for this cell being too large This is a known problem in MCNP where the user is cautioned that all energy transferred to electrons is assumed to be deposited locally In MCNPX the problem can be magnified because of the high energy nature of many applica tions and also because the F6 formalism is used in the type 3 Mesh Tally The user is also encouraged to carefully investigate the F8 tally which attempts to score energy deposition by following individual particles Continue runs that include mesh tallies must use the last available complete restart dump The output file for mesh tallies is not integrated into the restart dump file RUNTPE However they are written at each dump cycle Since the mesh tally file is overwritten at each dump care must be taken to ensure that the files used to continue a run were generated at the same dump cycle and that the last complete dump on the RUNTPE file is used An old version of FLUKA is implemented in this version of MCNPX The version of FLUKA now in MCNPX is taken directly from the LAHET version 2 8 code and is known as FLUKA87 Only the high energy portion of FLUKA is present to handle interactions above the INC region This is not the latest version of FLUKA and does not contain any of t
373. n fraction fraction gt weight fraction of constituent i in the material Keyword Value flag for density effect correction to electron stopping power m 0 default calculation appropriate for material in the GAS m condensed solid or liquid state used m 1 calculation appropriate for material in the gaseous state used causes the number of electron substeps per energy step to be increased to n for the material If n is smaller than the ESTEP n built in default found for this material the entry is ignored Both the default value and the ESTEP value actually used are printed in Table 85 default internally set changes the default neutron table identifier to the string id NLIB id The neutron default is a blank string which selects the first matching entry in XSDIR PLIB id changes the default photon table identifier to id default first match in XSDIR PNLIB id changes the default photonuclear table identifier to id default first match in XSDIR MCNPX User s Manual 75 MCNPxX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 26 Material Card Argument Description ELIB id changes the default electron table identifier to id default first match in XSDIR changes the default proton table identifier to id eae default first match in XSDIR sets conduction state of a material only for el03 evaluation lt 0 nonconductor COND 0 default nonconductor if at leas
374. n of Tritium 4 The INC is used only to determine that an interaction has occurred and the MPM pro ceeds from the compound nucleus formed by the absorption of the incident particle pure MPM 5 Arandom selection is made of one of the above modes at each collision with a proba bility P min E4 E 1 0 of choosing the pure MPM mode where E is the incident energy and E 25 MeV hybrid MPM An examination of the effect of these various options can be found in PRA94 4 1 3 Fermi Breakup Model The Fermi Breakup model BRE81 replaces the evaporation model for the disintegration of light nuclei It treats the deexcitation process as a sequence of simultaneous breakups of the excited nucleus into two or more products each of which may be a stable or unsta ble nucleus or nucleon Any unstable product nucleus is subject to subsequent breakup The probability for a given breakup channel is primarily determined from the available phase space with probabilities for two body channels modified by Coulomb barrier angu lar momentum and isospin factors The model is applied only for residual nuclei with A lt 17 replacing the evaporation model for these nuclei In the LAHET MCNPX imple mentation only two and three body breakup channels are considered it is an abbreviated form of a more extensive implementation of the Fermi Breakup model with up to 7 body simultaneous breakup used previously for cross section calculations on
375. n positron 19 positron 20 neutrino neutrino antineutrino 21 antineutrino MCNPX User s Manual LA CP 02 408 209 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 FNORM may be used to apply an overall multiplicative normalization to all bins except for IOPT 11 111 12 or 112 For these cases FNORM multiplies the time variable e g use FNORM 0 001 to convert from nanoseconds to microseconds The default is 1 0 KPLOT is a plot control flag plotting is available for some options provided it has been installed with the code using the LANL CGS and CGSHIGH Common Graphics System libraries Using a 0 indicates that no PLOT file will be produced and is the default IXOUT is a flag to indicate that the tally will be written to a formatted auxiliary output file for post processing The details and the file name are option dependent however a 0 indicates that no such file will be written and is the default IRS is the RESOURCE option flag A non zero value indicates that the option will be turned on 0 is the default see Section 19 below IMERGE is not used in HTAPE3X see Section 20 below ITCONV is the TIME CONVOLUTION option flag A non zero value indicates that the option will be turned on 0 is the default see Section 21 below IRSP is the RESPONSE FUNCTION option flag IRSP gt 0 indicates that the tally will be multiplied by a user supplied response function IRSP lt 0 indicates tha
376. n DXTRAN spheres 2 for photon DXTRAN spheres Let A be the average score to a DXTRAN sphere ora detector n Then if k lt 0 DXTRAN or detector scores lt A k are rouletted Ki k gt 0 DXTRAN detector scores lt k A are rouletted k 0 no Russian roulette is played on small DXTRAN detector scores Mi Criterion for printing large contributions A diagnostic print is made at the first 600 source or collision points where a DXTRAN detector score is greater than m T where T k or T k A Defaults If k is not specified on a DDn card k on the DD card is used If that is not MCNPX User s Manual 161 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 specified k on the DD card is used If that is not specified k 0 1 is used A similar sequence of defaults defines m with a final default of m 1000 Use Optional Remember that Russian roulette will be played for detectors and DXTRAN unless specifically turned off by use of the DD card Consider also using the PDn or DXC cards Example DXT N x y z RI RO x2 y2 z2 RIz RO2 x3 y3 z3 RI RO3 DXT P x4 ya z4 RI4 RO 3 F15X P a r Ry ag r Ro DD 2 100 15 2000 DD1 1 1E25 3000 J J J 3000 DD15 4 10 Detector sphere k m sphere 1 1 1E25 3000 sphere 2 15 2000 sphere 3 2 3000 sphere 4 2 100 detector 1 4 10 detector 2 15 2000 5 8 12 PDn Detector Contribution Form PDn P Po Pi P Table 5 98 Detector
377. n MCNPX 03e tables will not work The 0le tables are included in DLC200 Features related to probability tables in delayed neutrons will be ignored in MCNPX 3 Special 150 MeV libraries described in Section 4 3 of this manual have been pro duced for use with MCNPX The neutron library is called LA150n The proton and photonuclear libraries are called la150h and la150u respectively The LA150N library is the same as DLC200 with the addition of 150 MeV evaluations above the DLC200 energy limits and eliminating the 03e electron tables so that 01e ZAIDs are the default Once the proton and photonuclear components are added the entire library will be reissued under the name DLC200X 4 Anumber of users are requesting secondary particle and recoil nuclei information for the lower energy portions of the libraries typically below 20 MeV Note that some information is available in the lower energy tables per table 4 4 in this manual but it is far from complete A proper fix to the problem will involve full re evaluations of the lower energy libraries which is a time consuming and often difficult task Nonethe less progress is being made and the user should look for improved library releases in the future The LANL group that formats libraries for MCNP MCNP xX is currently providing 64 bit type 2 binary files and MCNPX 2 3 0 will only accept these Therefore the user will find that older versions of 32 bit binary libraries won t w
378. n Section 8 2 2 Note A default set of low energy cutoffs is in place see Table 5 1 Energies for particles other than neutrons neutrinos and photons can be set to a minimum of 1 keV the excep tions can be set to 0 0 MeV However no interaction physics is present between 1 keV and the default minimum Note Care must be taken for non standard code terminations when using the HTAPE3X program Normalization may not be what NPS indicates See Section 8 5 for details MCNPX User s Manual 75 MCNPX User s Manual Ap Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium 6 1 9 Peripheral Cards PRDMP LOST DBCN FILES PRINT MPLOT PTRAC PERT No changes have been made in peripheral cards 6 1 10 New Cards Specific to MCNPX The following cards are new to MCNPx Detailed explanations can be found in the asso ciation manual sections LCA LCB LEA LEB These cards control physics parameters for the BERTINI ISABEL CEM and FLUKA options See Section 6 2 HISTP This card will turn on the production of the LAHET compatible HISTP file See Section 8 5 SPABI Secondary particle biasing variance reduction See Section 7 1 TMESH R C S MESHn CORAn CORBn CORCn ENDMD ERGSH MSHMF Mesh Tally Cards See section 8 1 Fin Pin FSn Cn TI R C n TIR TIC NOTRN TALNP Radiography tally cards See section 8 2 6 2 Physics Module Options Four new MCNPX input cards have been defined to allow the user control of physics
379. nal Conference on Calorimetry in High Energy Physics La Biodola Elba September 19 25 1993 A Menzione and A Scribano eds World Scientific P 394 502 1994 MCNPX User s Manual 183 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 FAS97 A Fasso A Ferrari J Ranft and P R Sala An Update about FLUKA Proceedings of the 2nd Workshop on Simulating Accelerator Radiation Environments SARE2 CERN Geneva October 9 11 1995 CERN Divisional Report CERN TIS RP 97 05 p 158 170 1997 FAV99 J A Favorite and K Adams Tracking Charged Particles Through a Magnetic Field Using MCNPX U X Division Research Note XCI RN U 99 002 February 5 1999 FER98 A Ferrari and P R Sala The Physics of High Energy Reactions Lecture given at the Workshop on Nuclear Reaction Data and Nuclear Reactors Physics Design and Safety International Centre for Theoretical Physics Miramare Trieste Italy 15 April 17 May 1996 proceedings published by World Scientific A Gandini G Reffo eds Vol 2 p 424 532 1998 FIR96 R B Firestone and V S Shirley Table of Isotopes 8th Edition John Wiley New York 1996 GOU40 S Goudsmit and J L Saunderson Multiple Scattering of Electrons Phys Rev 57 1940 24 GUD75 K K Gudima G A Osokov and V D Toneev Model for Pre Equilibrium Decay of Excited Nuclei Yad Fiz 21 1975 260 Sov J Nucl Phys 21 1975 138 GUD83 K
380. ndent mesh as part of the regular transport problem and the contents of each mesh cell written to a file at the end of the problem This file can be converted into a number of standard formats suitable for reading by various graphical analysis packages The conversion program gridconv is supplied as part of the overall MCNPX package section 5 7 22 7 Analysis of this data is limited only by the capabilities of the graphical program being used MCNPX User s Manual 143 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 5 7 22 1 Setting up the Mesh in the INP File A mesh tally is defined by several cards which are described below All of the control cards for mesh tallies must be in a block preceded by a card containing the word tmesh in the first five columns and terminated by a card containing the word endmad in the first five col umns For each mesh tally card the following set of cards must be present which give details on the mesh characteristics CORAn corra n 1 corra n 2 corra n N CORBn corrb n 1 corrb n 2 corrb n N CORCn corre n 1 corre n 2 corrce n N where the CORAn CORBn and CORCn cards are used to describe the three coordi nates as defined by the mesh type rectangular cylindrical or spherical prior to any trans transformation In the case of rectangular meshes CORAn represent planes perpendicular to the x axis CORBn are planes perpendicular to the y axis and CORCn are p
381. ne will result in a reduced output print Use Optional Table 5 107 MCNPX Output Tables Table Number Type Table Description 10 Source coefficients and distribution 20 Weight window information 30 Tally description 35 Coincident detectors 40 Material composition 50 Cell volumes and masses surface areas 60 basic Cell importances 62 basic Forced collision and exponential transform MCNPX User s Manual 167 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 107 MCNPX Output Tables 70 Surface coefficients 72 basic Cell temperatures 85 Electron range and straggling tables multigroup flux values for biasing adjoint calcs 86 Electron bremsstrahlung and secondary production 90 KCODE source data 98 Physical constants and compile options 100 basic Cross section tables 102 Assignment of S a B data to nuclides 110 First 50 starting histories 120 Analysis of the quality of your importance function 126 basic Particle activity in each cell 128 Universe map 130 Neutron photon electron weight balance 140 Neutron photon nuclide activity 150 DXTRAN diagnostics 160 default TFC bin tally analysis 161 default fx tally density plot 162 default Cumulative f x and tally density plot 170 Source distribution frequency tables surface source 175 shorten Estimated ke results by cycle 178 Estimated kef
382. ne vere Pee Wed coe eee ee PrelaCG i inea gi ten Cte Rota niin a aoe 1 Introduction i tari cake ieee es 2 Warnings Known bugs and Revision Notes 2 1 Warnings and Known Bugs 000 2 2 Release noteS 0 0 cece 3 MCNPX Installation 000e eee eee 3 1 MCNPX Build System 00000 ee eee 3 1 1 Inthe Beginning 0000 3 1 2 Automated Building 3 1 3 MCNPX Build Examples 3 1 4 Directory Reorganization 3 1 5 Users Notes 0 e eee eee 3 1 6 Multiprocessing 0000 3 1 7 Programmers notes 00005 3 1 8 Additional Software Requirements 3 1 9 Fortran 90 Compilers 3 1 10 Inthe End 0 0 00 cee eee 3 2 Libraries and Where to Find Them 4 Physics and Data 00c eee eee 4 1 Intermediate Interaction Physics 4 1 1 Intranuclear Cascade Models 4 1 2 Multistage Pre equilibrium Models MPM 4 1 3 Fermi Breakup Model 4 1 4 Evaporation Model MCNPX User s Manual MCNPxX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 vii MCNPxX User s Manual Version 2 3 0 April 2002 x LA UR 02 2607 Aesaueion of Tritium 4 1 5 Level Densities 0 0 0 0 44 4 1 6 High Energy Fission 2 sac0 249 pete Meena edie tena beet ed 45 4 2 High Ene
383. never modeling 3He coincidence 5 5 6 Problem Cutoff Cards 5 5 6 1 CUT Cutoffs Form CUT in T E WCl WC2 SWIM Table 5 38 CUT Card Keyword Description n particle type designator T time cutoff in shakes 1 shake 10 sec E lower energy cutoff in MeV WCI weight cutoff survival weight weight cutoff If weight goes below WC1 roulette is played to restore weight to WC2 Negative values make Me WC1 and WC2 relative to importances Setting WC1 WC2 0 invokes analog capture SWTM minimum source weight Use Optional as needed Neutron default 7 very large E 0 0 MeV WC7 0 50 WC2 0 25 SWTM minimum source weight if the general source is used 5 5 6 2 ELPT Cell by cell Energy Cutoff Form ELPT n x7 X9 Xj 0X 88 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 39 Cell by cell Energy Cutoff Keyword Description n particle type Xx lower energy cutoff of cell i I number of cells in the problem Default Cutoff from Cut n Use Optional A separate lower energy cutoff can be specified for each cell in the problem The higher of either the value on the ELPT n card or the global value E on the CUT n card applies 5 5 6 3 NPS History Cutoff Form NPS N NPP NPSMG Default Infinite Use As needed to terminate the calculation In a criticality calculation the NPS card has no meaning and a
384. ngle reference to a TR card that can be used to trans late and or rotate the entire mesh Only one TR card is permitted with a mesh card 5 7 22 6 Dose Conversion Coefficients MCNPxX contains a number of standard dose conversion coefficients This feature is accessed through the dose keyword of the Type 1 Mesh Tally See section 5 7 22 2 150 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 function DFACT id ic en it iu acr Table 5 85 DFACT Argument Descriptions ARGUMENT DESCRIPTION Particle identification number 1 neutron 2 photon Choice of conversion coefficient Note The 10 and 20 options are Dose Equivalent H i e absorbed dose ata point in tissue weighted by a distribution of quality factors Q related to the LET distribution of radiation at that point The 30 s options are Equivalent Dose Hy based on an average absorbed dose in the tissue or organ Dy weighted by the radiation weighting factor w summed over all component radiations neutrons 10 ICRP 21 1971 20 NCRP 38 1971 ANSI ANS 6 1 1 1977 31 ANSI ANS 6 1 1 1991 AP anterior posterior 32 ANSI ANS 6 1 1 1991 PA posterior anterior 33 ANSI ANS 6 1 1 1991 LAT side exposure 34 ANSI ANS 6 1 1 1991 ROT normal to length amp rotationally symmetric 40 ICRP 74 1996 ambient dose equivalent photons 10 ICRP 21 1971 20 Claiborne amp Trubey
385. nhole F3 The distance from the pinhole at X1 Y1 Z1 to the detector grid along the direction established from X2 Y2 Z2 to X1 Y1 Z1 and perpendicular to this reference vector The grid dimensions are established from entries on FS and C cards In this use the first entry sets the lower limit of the first bin and the other entries set the upper limit of each of the bins These limits are set relative to the intersection of the reference direction and the grid plane as shown in Figure 8 2 An example is discussed below FSn 20 99i 20 Cn 20 99i 20 These two cards set up a 100 x 100 grid that extends from 20 cm to 20 cm in both direc tions and has 10 000 equal size bins These bins need not be equal in size nor do they need to be symmetric about the reference direction The directions of the t axis and s axis of the grid are set up such that if the reference direc tion the outward normal to the grid plane is not parallel to the z axis of the geometry the t axis of the grid is defined by the intersection of the grid plane and plane formed by the z axis and the point where the reference direction would intersect the grid plane If the ref erence direction is parallel to the z axis of the geometry then the t axis of the grid is MCNPX User s Manual 103 MCNPxX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium defined to be parallel to the y axis of the geometry
386. nhole case in Section 5 7 20 1 However X1 Y1 Z1 are now the coordinates of the intersection of the reference direction and the grid plane In the cylindrical grid case the entries on the FSn card are the distances along the symmetry axis of the cylinder and the entries on the Cn card are the angles in degrees as measured counterclockwise from the positive t axis When this type of detector is being used in a problem if a contribution is required from a source or scatter event an attenuated contribution is made to each and every detector grid bin Since for some types of source distributions very few histories are required to image the direct or source contributions an additional entry has been added to the NPS card to eliminate unwanted duplication of information from the source See Section 5 5 6 3 5 7 20 3 Additional Radiography Input Cards A NOTRN card is added as an additional possible input When this card appears in the INP file no transport of the source particles takes place and only the direct or source contributions are made to the detector grid This is especially useful for checking the problem setup or doing a fast calculation to generate the direct source image This option works with either the pinhole or transmitted image options The option is also available to turn off the printing of all of the values in each of the grid bins in the OUTP file The card TALNP with no arguments turns off the bin print for all tallies in
387. ni model the n p value for this case should be considered more accurate than the value calculated in the base case MCNPX User s Manual 131 MCNPxX User s Manual E Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium Case 5 In the final variation from the base case we use the CEM model for neutron protons and pions CEM is turned on by setting the 9th entry of the LCA card to 1 Base Case ICA jjj Case 4 Rea a ae ah o ae Sy The neutron summary table for this case is shown below sample problem spallation target Case 5 n creation tracks weight energy neutron loss tracks weight energy per source particle per source particle source 0 0 Or escape 313015 1 5635E 01 2 1374E 02 nucl interaction 254437 1 2722E 01 3 1302E 02 energy cutoff 0 0 0 particle decay 0 0 0 time cutoff 0 0 0 weight window 0 0 0 weight window 0 0 0 cell importance 0 0 0 cell importance 0 0 0 weight cutoff 0 0 0 weight cutoff 0 0 0 energy importance 0 QO 0 energy importance 0 QO 0 dxtran 0 0 0 dxtran 0 0 0 forced collisions 0 0 0 forced collisions 0 0 0 exp transform 0 QO QO exp transform 0 QO QO upscattering 0 0 0 downscattering 0 0 7 3438E 00 tabular sampling 0 QO 0 capture 0 1 3051E 02 8 5469E 02 n xn 9157 1 4 5738E 00 2 1850E 01 loss to n xn 29374 1 4667E 00 5 7124E 01 fission 0 0 0 loss to fission 0 0 0 photonuclear 0 0 G nucl interaction 3619 1 8095E 01 5 6576E 01 tabular bound
388. nm 3 Tektronix 4014E using CGS This is the default 4115 Tektronix using GKS and UNICOS This is the default 1 Tektronix using the AIX PHIGS GKS library This is the default Check with your vendor for the proper terminal type if you are using a GKS library m specifies the baud rate of the terminal The default value is 9600 Send or don t send plots to the graphics metafile PLOTM PS according to the value of the parameter aa The graphics metafile is not created until the first FILE command is entered FILE has no effect in the NOTEK or TERM 0 cases FILE aa The allowed values of aa are blank only the current plot is sent to the graphics metafile ALL the current plot and all subsequent plots are sent to the metafile until another FILE command is entered NONE the current plot is not sent to the metafile nor are any subsequent plots until another FILE command is entered General Commands g Continue reading commands for the current plot from the next input line The amp must be the last thing on the line Plot a curve according to the commands entered so far and keep the plot open for co plotting one or more COPLOT additional curves COPLOT is effective for 2D plots only If COPLOT is the last command on a line it functions as if it were followed by an amp 50 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 3 MPLOT amp MCPLOT Co
389. noted in the MCNPX User s Manual All standard neutron libraries used with MCNP4B originally distributed in DLC 189 and now included in DLC 205 can be used with MCNPX however they will not contain emission data for charged particles or recoil nuclei therefore these products will not be produced and tracked All neutron photon and electron libraries developed for use with MCNP4C will work with MCNPX2 4 0 2 CONTRIBUTOR Advanced Accelerator Applications Los Alamos National Laboratory Los Alamos New Mexico 3 CODING LANGUAGE AND COMPUTER Fortran 90 and C IBM RS 6000 DEC Alpha SGI HP HP UX Sun Intel Linux Windows PC C00715MNYCP00 4 NATURE OF PROBLEM SOLVED The official release date of MCNPX 2 4 0 is August 1 2002 MCNPX extends the MCMNP4C3 code to higher energies and more particle types Photonuclear capability in the tabular range is included in this release Neutron tabular data are used as in MCNP4C3 above the table energy limits physics modules are used Current physics modules include the Bertini and ISABEL models taken from the LAHET Code System LCS and CEM An old version of FLUKA is available for calculations above the range of INC physics applicability MCNPX eliminates the need now present in LCS to transfer large files between separate codes MCNPX lt is released with libraries for neutrons iii photons electrons protons and photonuclear interactions In addition variance reduction schemes such as secon
390. nput file after the blank line terminator The space following the blank line terminator is a good place for problem documentation at the user s discretion 4 1 2 Continue Run Continue run allows the user to pick up a previously terminated job where it left off For example a job run for 2 hours may be continued run an additional amount of time The user can also reconstruct the output of a previous run A continue run must contain C or CN in the MCNPX execution line or message block to indicate a continue run It will start with the last dump or with the mth dump with the Cm option In addition to the C or CN option on the MCNPX execution line two files can be important for this procedure 1 the binary restart file default name RUNTPE and 2 an optional continue run input file default name INP The restart file generated by MCNPX in the initiate run sequence contains the geometry cross sections problem parameters tallies and all other information necessary to restart the job In addition the problem results at various stages of the run are recorded in a series of dumps See the PRDMP card Section 5 9 1 for a discussion of the selection of dump times As discussed below the run may be restarted from any of these dumps The CN execution message option differs from the C option only in that the dumps produced during the continue run are written immediately after the fixed data portion of the RUNTPE file rather than after the dump f
391. ns 13 10 1 1 10 10 0 5 800 1 In this case the energy is binned in 10 equal lethargy intervals of half decade width below 800 MeV and normalized per MeV No time binning is done Only neutrons are edited The z axis determines the polar angle and the azimuthal angle is measured from the x axis Ten azimuthal angle bins are used and 10 equal polar angle cosine bins are defined by taking the default Note that the last four records could be written on one line as 0 5 800 1 Tally option 13 may be considered as the time integrated particle current integrated over a sphere in a void at a very large distance for the interaction region Since it is normalized per unit solid angle the units are dimensionless being sr per source particle 16 Edit Option IOPT 14 or 114 Gas Production Option 14 provides an edit of hydrogen and helium gas production by isotope by element and total Unless modified by FNORM the units of gas production are atoms per source particle If KOPT 0 the edit is by cell number if KOPT 1 the edit is by material NERG NTIM and NTYPE are unused The estimate is made by tallying all H and He ions stopped in a cell or material including source particles 17 Edit Option IOPT 15 or 115 Isotopic Collision Rate Option 15 has been added to provide a collision rate edit by target isotope The input has the same meaning as for lOPT 8 with the following exceptions KOPT 0 or 1 tabula
392. ns in the physics based energy region We do not anticipate problems since criticality issues are concerned with very low energy neutron transport however the user should carefully check the answers for reasonableness when using this feature 5 Next Event Estimators i e point and ring detectors DXTRAN and radiography tally options sometimes underpredict the true answer in MCNPX These tallies rely on the angular distribution data for particles produced in an interaction to predict the next event Information on these distributions is available in tabular form in the libraries This information is not easily available in the required form from physics models used to produce secondary particles above the tabular region therefore no next event contributions are made If the user is certain that all particles in the prob lem will be produced from collisions within the tabular energy limits next event esti mators will work well However next event estimates even at energies within the tabular region are not accounted for properly if the source or collision particle is above the tabular region Thus the answer will be underestimated Correcting this problem is a major area of investigation for the MCNPX code developers 6 Next Event Estimators i e point and ring detectors DXTRAN and radiogra phy tally options will not work for charged particles in any energy region This is due to lack of proper algorithms which can treat th
393. ns on these themes are given 3 1 3 1 System Wide Installation For purposes of the first illustration we will assume that the MCNPX distribution has been unloaded from CDROM or fetched from the net and is in the file usr local src mcnpx_2 4 0 tar gz The system manager logged is as root will unload the distribution into usr local src mcnpx_2 4 0 will build the system in tmp mcnpx will install the mcnpx executable in usr local bin and will install the libraries end eventually the mcnp cross sections into usr local lib Naturally the specific name of the mcnpx distribution archive will vary depending on the version you have acquired The following example uses ell shell commands to accomplish this task If you are more familiar with csh you will need to adjust things appropriately NOTE Comments about the shell commands start with the character Also don t be alarmed by the generous amount of output from the configure and make scripts They work hard so you don t have to go to the installation directory cd usr local src Unpack the distribution This creates the directory mcnpx_2 4 0 gzip dc mcnpx_2 4 0 tar gz tar xf go to tmp and make the build directory cd tmp mkdir mcnpx go into that working space cd mcnpx 12 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 execute the configure script no special option requests for the Makefiles the default dir
394. nsitions from INC to Preequilibrium and Preequilibrium to evaporation have been developed Until the new version is available the user should be cautious in using the CEM model for production calculations Summary Results compiled for each case of this example are shown in A 2 Note the run time for the case where the ISABEL INC model is used is about 15 greater than the base case using the Bertini model Case 2 also runs slower since the light ion interactions are provided by the ISABEL model Invoking the 150 MeV proton libraries slows execution by about 11 in this example Table A 2 Results Compiled for Summary Cases ae Runtime Case Variation from base case s n p minutes base n a 27 66 18 263 1 LAHET transport for 20 150 28 44 18 364 MeV neutrons 2 light ion transport amp nuclear 33 55 18 335 interaction 3 ISABEL INC for nucleons and 31 91 17 569 pions 4 evaluated data used for pro 30 66 18 285 tons below 150 MeV 5 CEM INC for nucleons and 60 14 15 638 pions a Cases were run on an IBM AIX box This example demonstrates how to calculate neutron production from a spallation target Use of the new LA150 library that extends evaluated nuclear data up to 150 MeV gives the most accurate results particularly if the new proton evaluations are used in addition to the 202 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 neutron evaluations When
395. nt Form FCn info Table 5 59 Tally Comment Card Variable Description n Tally number amp type info provides title for tally in output and MCTAL file Default No comment MCNPX User s Manual 121 MCNPxX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Use Encouraged 5 7 3 En Tally Energy Form En E Ex Table 5 60 Tally Energy Card Variable Description n tally number Ej upper bound in MeV of the i energy bin for tally n Default If the En card is absent there will be one bin over all energies unless this default has been changed by an EO card Use Required if EMn card is used 5 7 4 Tn Tally Time Form Tn T1 Tk Table 5 61 Tally Time Card Variable Description n tally number Ty Tk upper bound in shakes of the i time bin for tally n Default If the Tn card is absent there will be one bin over all times unless this default has been changed by a TO card Use Required if TMn card is used Consider FQn card Example T2 1 1 1 0 37 NT This will separate an F2 flux surface tally into three time bins 1 from to 1 0 shake 2 from 1 0 shake to 1 0 shake and 3 from 1 0 shake to 1 0e37 shakes effectively infinity No total bin will be printed in this example 5 7 5 Cn Cosine Card tally type 1 and 2 Form Cn C4 Ck 122 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 20
396. nteger 0 Unlimited VALUE Real Integer Unlimited Table 5 109 Mnemonic Values for the FILTER Keyword Mnemonic MCNP Variable Description X XXX X coordinate of particle position cm Y YYY Y coordinate of particle position cm Z ZZZ Z coordinate of particle position cm U UUU Particle X axis direction cosine V VVV Particle Y axis direction cosine W WWW Particle Z axis direction cosine ERG ERG Particle energy MeV 170 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 109 Mnemonic Values for the FILTER Keyword WGT WGT Particle weight TME TME Time at the particle position shakes VEL VEL Speed of the particle cm shake IMP1 FIML 1 Neutron cell importance IMP2 FIML 2 Photon cell importance IMP3 FIML 3 Electron cell importance SPARE1 SPARE 1 Spare banked variable SPARE2 SPARE 2 Spare banked variable SPARE3 SPARE 3 Spare banked variable ICL ICL Problem number of current cell JSU JSU Problem number of current surface IDX IDX Number of current DXTRAN sphere NCP NCP Count of collisions for current branch LEV LEV Geometry level of particle location III Il 1st lattice index of particle location JJJ JJJ 2nd lattice index of particle location KKK KKK 3rd lattice index of particle location 5 9 5 HISTP and HTAPE3X In order to produce the LAHET compatible HISTP files the following card must be
397. oblem spallation target Case 1 neutron creation tracks weight energy neutron loss tracks weight energy per source particle per source particle source 0 0 Ga escape 367324 1 8351E 01 2 1946E 02 nucl interaction 376685 1 8834E 01 3 3940E 02 energy cutoff 0 QO 0 particle decay 0 0 0 time cutoff 0 0 0 weight window 0 0 0 weight window 0 0 0 cell importance 0 0 0 cell importance 0 0 0 weight cutoff 0 0 0 weight cutoff 0 0 0 energy importance 0 QO 0 energy importance 0 QO 0 dxtran 0 0 0 dxtran 0 0 0 forced collisions 0 0 0 forced collisions 0 0 0 exp transform 0 QO QO exp transform 0 QO Ox upscattering 0 0 0 downscattering 0 0 9 8003E 00 tabular sampling 0 QO 0 capture 0 1 3626E 02 5 7541E 02 n xn 20323 1 0137E 00 1 5895E 00 loss to n xn 9964 4 9705E 01 6 8449E 00 fission 0 0 0 loss to fission 0 0 0 photonuclear 0 0 Oz nucl interaction 19720 9 8600E 01 1 0482E 02 tabular boundary 11 5 5000E 04 1 0972E 02 tabular boundary 11 5 5000E 04 1 0972E 02 gamma xn 0 0 0 particle decay 0 0 0 adjoint splitting 0 0 0 total 397019 1 9848E 01 3 4100E 02 total 397019 1 9848E 01 3 4100E 02 number of neutrons banked 387055 average time of shakes cutoffs neutron tracks per source particle 1 9851E 01 escape 5 8655E 00 tco 1 0000E 34 neutron collisions per source particle 2 8027E 01 capture 4 8948E 01 eco 0 0000E 00 total neutron collisions 560536 capture or escape 5 8615E 00 wcl 5 0000E 01 net multiplication 0 0000E 0
398. oblem summary for this case is shown below MCNPX User s Manual 195 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 sample problem spallation target Case 1 neutron creation tracks weight energy per source particle source 0 0 0 nucl interaction 376685 1 8834E 01 3 3940E 02 particle decay 0 0 0 weight window 0 0 0 cell importance 0 0 0 weight cutoff 0 0 0 energy importance 0 0 QO dxtran 0 0 0 forced collisions 0 0 0 exp transform 0 0 0 upscattering 0 0 0 tabular sampling 0 0 0 n xn 20323 1 0137E 00 1 5895E 00 fission 0 0 0 photonuclear 0 0 QO tabular boundary 11 5 5000E 04 1 0972E 02 gamma xn 0 0 0 adjoint splitting 0 0 0 total 397019 1 9848E 01 3 4100E 02 number of neutrons banked 387055 neutron tracks per source particle 1 9851E 01 neutron collisions per source particle 2 8027E 01 total neutron collisions 560536 net multiplication 0 0000E 00 0000 neutron loss tracks escape 367324 energy cutoff 0 time cutoff 0 weight window 0 cell importance 0 weight cutoff 0 energy importance 0 dxtran 0 forced collisions 0 exp transform 0 downscattering 0 capture 0 loss to n xn 9964 loss to fission 0 nucl interaction 19720 tabular boundary 11 particle decay 0 total 397019 average time of shakes escape 5 8655E 00 capture 4 8948E 01 capture or escape 5 8615E 00 any termination 5 4273E 00 weight energy per source particle 1 8351E 01 2 1946E 02 QO QO
399. odules within MCNPX is encour aged this provides the user with the ability to test the sensitivity of the quantity of interest to the different physics models If significant differences are observed the user should evaluate which physics model is most appropriate for his or her particular application For example total neutron production from actinide targets is known to be more accurate if the multi step preequilibrium model MPM is turned off which is not its default setting 1 All particles should be included for energy deposition calculations as discussed in Section 8 3 MCNPX User s Manual 133 MCNPX User s Manual Version 2 3 0 April 2002 F LA UR 02 2607 Accelerator Production of Tritium 134 MCNPX User s Manual MCNPX User s Manual Ap Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium Appendix B HTAPESX for use with MCNPX This appendix is reprinted from HTAPE3X for Use with MCNPX Richard E Prael Los Alamos Report LA UR 99 1992 April 16 1999 Abstract HTAPESxX is a code for processing medium and high energy collision data written to a his tory file by MCNPX In addition it provides surface flux and current edits which supplement the standard MCNP tallies 1 The HTAPE3X Code HTAPESxX is a modification of the HTAPE code from the LAHET Code System 1 designed to provide analysis of the history file HISTP optionally written by MCNPX 2 It is primari
400. of the origin of the main coordinate system defined in the auxiliary system Use Convenient for many geometries Default TRn 000 100 010 001 1 Example 174 RCC 000 0120 5 TR4 2000 45 4590 1354590 90900 Surface 17 is transformed via transformation 4 resulting in it s being displaced to x y z 20 0 0 and rotated as in the example on the TRCL card above Other Data Cards All MCNPX input cards other than those for cells and surfaces are entered after the blank card delimiter following the surface card block The mnemonic must begin within the first five columns No data card can be used more than once with the same number or particle type designations For example M1 and M2 are acceptable as are CUT N and CUT P but two M1 cards or two CUT N cards are disallowed 5 4 MATERIALS Mm DRXS TOTNU NONU AWTAB XSn VOID PIKMT MGOPT 5 41 Mm Material Form Mm ZAID fraction ZAID fraction keyword value 74 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 26 Material Card Argument Description arbitrary material number match with material number on m cell cards either a full ZZZAAA nnX or partial ZZZAAA element or ZAID ol ee f nuclide identifier for constituent i ZZZ atomic number ve gt 0 atomic mass 0 naturally occurring element nn the library identifier X the class of data fraction gt atomic fractio
401. oint sources of arbitrary intensity The use of a distribution of transformations is invoked by specifying TR Dn on the SDEF card The cards SI SP and optionally SB are used as specified for the SSR card on page 3 57 of the MCNP User s Guide SInL Ik SPn optionP P SBn optionB I The L option on the SI card is required new input checking has been implemented to ensure this usage for both the SDEF and SSR applications The option on the SP and SB cards may be blank D or C The values 1 l identify k transformations which must be supplied The content of the SP and SB cards then follows the general MCNP rules The following example shows a case of three intersection Gaussian parallel beams each defined with the parameters a 0 2cm b 0 1cm and c 2 in the notation previously discussed For each the beam is normal to the plane of definition e Beam 1 is centered at 0 0 2 with the major axis of the beam distribution along the x axis emitted in the z direction with relative intensity 1 e Beam 2 is centered at 2 0 0 with the major axis of the beam distribution along the y axis emitted in the x direction with relative intensity 2 e Beam 3 is centered at 0 2 0 with the major axis of the beam distribution along the line x z emitted in the y direction with relative intensity 3 110 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 The card SBn is used to provide
402. ommended that MCNPX be run with 64 bit libraries Earlier versions of the code could use 32 bit libraries however studies of long problems have shown that erroneous answer can result with the lesser accuracy data Conversion of Type 1 libraries to 64 bit MCNPX User s Manual 35 MCNPX User s Manual E Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium binaries can be done with the MAKXSF routine described in Appendix C of the MCNP4B manual The LAHET physics modules in MCNPX require three special libraries BERTIN containing the elemental cross section data needed by the Bertini model PHTLIB containing nuclear structure data needed to generate de excitation photons BARPOL DAT containing new high energy total reaction and elastic cross sections They are unpacked with the rest of the code and if make install is executed placed in the lib directory There are basically 2 ways that the code tries to find these files 1 MCNPxX tries to open the files named bertin and phtlib in the current directory If the user wants to keep these file in another directory a symbolic link should be made from whatever directory you are in when running the code The following unix com mand can be used to do this In s home me lib bertin 2 A default pathname is coded in the fortran data statements in the file src Ics inbd F This can be changed by the user but you must remember to
403. on in centimeters if Ro is R entered as positive in mean free paths if entered as neg ative Negative entry illegal in a void Form for ring detectors Fna pl agr R Table 5 57 Ring Detector Variable Description n tally number a the letter X Y or Z pl N for neutrons or P for photons ao distance along axis a where the ring plane intersects the axis r radius of the ring in centimeters 120 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 57 Ring Detector Variable Description Same meaning as for point detectors but describes a sphere about the point selected on the ring R Default None 5 7 1 4 Pulse Height Tally tally type 8 Simple Form Fn plS Sk General Form Fn plS So 83 S4 55 Sg 7 Table 5 58 Pulse Height Tally Variable Description n tally number pl particle designator S problem number of cell for tallying or 7 Note Variance reduction is not allowed for problems with regular pulse height tallies It is allowed for energy pulse height tallies F8 if there are no energy bins The energy bins in the pulse height tally are different than for all other tallies Rather than tally the particle energy at the time of scoring the numbers of pulses depositing energy within the bins are tallied 5 7 2 FCn Tally Comme
404. on at the beginning of the substep P is the I Legendre polynomial and G is i do G 2nNf ggl Pilw du in terms of the microscopic cross section do dQ and the atom density N of the medium For electrons with energies below 0 256 MeV the microscopic cross section is taken from numerical tabulations developed from the work of Riley RIL75 For higher energy elec trons the microscopic cross section is approximated as a combination of the Mott MOT29 and Rutherford cross sections RUT11 with a screening correction Seltzer pre sents this factored cross section in the form 60 MCNPX User s Manual MCNPxX User s Manual Ap Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium i do do 2 7 e dQMott dQ py u 2 do dQRutherford where e p and v are the charge momentum and speed of the electron respectively The screening correction n was originally given by Moliere MOL48 as H amc 5 2 n 3 Dr Z 113 3 76 05 where a is the fine structure constant m is the rest mass of the electron and B v c MCNP MCNPxX now follows the recommendation of Seltzer SEL88 and the implemen tation in the Integrated TIGER Series by using the slightly modified form 1 2 Sad Va 2 Mg A z 113 3 76 a5 5 where Tt is the electron energy in units of electron rest mass The multiplicative factor in the final term is an empirical correction which improves the agreement at low energies
405. on for an estimate of energy loss In addition departure from the initial energy group during the sub step was ignored The new logic applies the Vavilov algorithm to each substep and to each partial substep and makes a better estimate of the continuous slowing down energy loss mean energy loss across energy group boundaries The new treatment leads to considerably improved results in a variety of physically interesting calculations such as the range of heavy charged particles A full description of the algorithm and some exam ples of the results can be found in a recent Los Alamos Research Note PRAOOa 5 4 Multiple Scattering for Heavy Charged Particles The full Goudsmit Saunderson model of multiple scattering for electrons as implemented in MCNP4B MCNPX was described in Section 4 3 3 2 In MCNPX version 2 3 0 a small angle Coulomb scattering treatment has been imple mented for heavy charged particles We use a Gaussian model based on a theory presented by Rossi In the original theory both angular deflections and small spatial dis placements were accounted for Since the complex geometric system of MCNPX does not yet accommodate transverse displacements in charged particle substeps we use only the MCNPX User s Manual 69 MCNPxX User s Manual Version 2 3 0 April 2002 E LA UR 02 2607 Accelerator Production of Tritium part of the theory that addresses the angular deflection In several test cases this slight approxima
406. on from nuclear interactions is given by the difference of the neutron weights in the neutron creation and neutron loss columns A similar approach is taken to calculate net n xn production Net neutron production may also be calculated by realizing that the only loss mechanisms for neutrons are escape and capture The sum of the weights in the neutron loss column under escape and capture is thus equal to the net neutron production The values listed in the problem summary are collision estimators meaning they are tallied when a collision occurs during transport Uncertainties are not calculated by MNCPX for these collision estimated quantities A reasonable upper limit on the relative uncertainty would be given by the inverse square root of the number of source particles launched We provide here five different variations for the calculation of net neutron production for this simple target geometry In the base case we transport protons neutrons and MCNPX User s Manual 191 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 charged pions The transition energy between LAHET physics and neutron transport using tabular nuclear data is set at 150 MeV and the LA150 library is used All protons are transported using LAHET physics Nucleon and pion interactions simulated by LAHET physics use the Bertini intranuclear cascade model Variations from this base case are outlined in A 1 below For each case 20 000 source
407. on of the direct contribution is avoided by adding the average of the previous direct contributions into each of the appropriate tally bins Depending on the time required for a particular problem this can save from a few seconds to upward of ten minutes per history in some cases As described above for a monoenergetic isotropic point source or amonoenergetic monodirectional surface source NPSMG 1 is adequate 106 MCNPX User s Manual MCNPX User s Manual i Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium 8 2 3 Additional Radiography Input Cards ANOTRN card is added as an additional possible input When this card appears in the INP file no transport of the source particles takes place and only the direct or source contri butions are made to the detector grid This is especially useful for checking the problem setup or doing a fast calculation to generate the direct source image This option works with either the pinhole or transmitted image options The option is also available to turn off the printing of all of the values in each of the grid bins in the OUTP file The card TALNP with no arguments turns off the bin print for all tal lies in the problem If there are entries it turns off the bin print for the tally numbers that are listed If after the run is completed one would like to see these numbers the printing of the bin values can be restored with the TALNP card in an INP file used in a continue
408. onization de dx energy is deposited uniformly along the track length which is important to keep in mind when doing a mesh tally All other energy deposition is calculated at the time of a nuclear interaction The energies of secondary particles if they are not to be tracked i e not included on the MODE card will be deposited at the point of the interaction Nuclear recoil energy will always be deposited at the point of interaction In order to obtain the most accurate energy deposition tallies possible the user must include all potential secondary particles on the MODE card Electrons can be omitted provided the user fully understands how energy deposition for photons is done The handling of energy deposition for non tracked secondary particles differs for the energies where libraries and physics models are used This procedure is under review and will likely be changed in future versions of the code Energies of all secondary particles except photons are added into the heating KERMA factors for the neutron and proton libraries This photon treatment was implemented in the MCNP libraries well before the development of the MCNPX code However since MCNP does not track charged particles standard practice was to include the energies of all other particles in the heating numbers for the evaluated libraries We are increasingly finding that local deposition of secondary particle energies causes difficulties particularly when the energies of
409. options such as debugging non debugging versions or different compiler types The local user building the private copy is again username me whose home directory is the directory home me The user has fetched the distribution from CDROM or from the net and has it in the file home me mcnpx_2 3 0 tar gz The user will unload the distribu tion package into home me menpx_2 3 0 With this method the source can be anywhere as long as the user has the pathname to it The user will build the system in the local directory home me menpx install the binary executable in home me bin and install the binary data files and eventually the mcnp cross sections in home me lib The following example uses bourne shell commands that follow accomplish this task If you are more familiar with csh you will need to adjust things appropriately NOTE Com ments about the shell commands start with the character Also don t be alarmed by the generous amount of output from the configure and make scripts They work hard so you don t have to go to your user home directory cd home me unpack the distribution that was copied from the net or a CDROM This creates home me mcnpx_2 3 0 gzip dc mcnpx_2 3 0 tar gz tar xf make a local directory for a build directory Call it mcnpx MCNPX User s Manual 19 MCNPX User s Manual E Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium mkdir mcnpx go into that new
410. or this case is 18 285 n p which is 0 1 greater than the base case value Thus as for neutrons the new 150 MeV proton evaluations for lead predict higher neutron production by protons in the energy range 20 to 150 MeV than does the Bertini INC model Since the proton evaluations are considered to be more accurate than 200 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 the Bertini model the n p value for this case should be considered more accurate than the value calculated in the base case Case 5 In the final variation from the base case we use the CEM model for neutron protons and pions CEM is turned on by setting the 9th entry of the LCA card to 1 Base Case LCA j j j Case 4 Ica jjjjjjjj1 sample problem spallation target Case 5 n creation source nucl interaction particle decay weight window cell importance weight cutoff energy importance dxtran forced collisions exp transform upscattering tabular sampling n xn fission photonuclear tabular boundary gamma xn adjoint splitting total tracks number of neutrons neutron tracks per neutron collisions total neutron collisions weight energy per source particle 0 On QO 254437 1 2722E 01 3 1302E 02 0 QO 0 0 QO 0 0 QO 0 0 0 0 0 QO 0 0 QO 0 0 QO 0 0 0 0 0 QO 0 0 QO 0 91571 4 5738E 00 2 1850E 01 0 QO 0 0 QO 0 1 5 0000E 05 7 4680E 03 0 0 0 0 QO QO 346009 1 72
411. ord may be followed by up to four entries where e Ifthe first entry is 1 to 9 an energy dependent dose function must be sup plied by the user on a MSHMF card e If the first entry is 10 20 31 35 or 40 the dose function comes from the dose function dfact See section 5 7 22 6 for details The next three entries define the input needed by that function the four needed entries corre spond to DFACT arguments ic it iu and acr Also see section 5 7 8 DFn Card e If no entries follow the dose keyword the default entries are 10 1 1 and 1 0 which form inputs into the dfact function Results are in rem hour Causes the population to be scored in each volume which is equivalent to the popul weight times the track length Scores the average energy deposition per unit volume MeV cm source parti cle for the particle type P In contrast to the 3rd type of Mesh Tally energy deposition can be obtained in this option for any particular particle pedep This option allows one to score the equivalent of an F6 P see Section 5 7 1 heating tally for the particle type P Note the mesh is independent of problem geometry and a mesh cell may cover regions of several different masses Therefore the normalization of the pedep option is per mesh cell volume not per unit mass 146 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 81 Track Averaged Mesh Tally
412. ork with the 2 3 0 The program MAKXS is provided with the MCNPX distribution to do the reformatting and details can be found in Appendix C of the MCNP4B manual An alternative is to use type 1 formatted sequential access libraries The XSDIR file tells the code all the information it needs to known on where to find individ ual data tables MCNPX uses the same procedure as MCNP4B to find the nuclear data libraries as described in Appendix F of the MCNP4B manual If XSDIR is not in your cur rent directory MCNPX will search the following places for both the libraries and XSDIR file 34 MCNPX User s Manual MCNPxX User s Manual i Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium in order starting from 1 We repeat that portion of the MCNP4B manual here with annotations 1 xsdir datapath on the MCNPX execution line note datapath is truncated to 8 characters which means that it is really the name of a file not a path It is easiest to assign a name via a symbolic link e g In s hnome me lib data xsdir xsdir1 Then you can say mcnpx xsdir xsdir1 DATAPATH datapath in the INP file message block this version of datapath can be a full description the current directory the DATAPATH entry on the first line of the XSDIR file the UNIX environmental variable setenv DATAPATH datapath the individual data table line in the XSDIR file the directory specified at MCNPX compile time in the blkd
413. orresponding recoil or damage energy MCNPX User s Manual 149 MCNPX User s Manual Ap Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium At any collision the damage energy Egis obtained from the recoil energy E of nucleus A Z by the relation of Linhard 4 Ey E L E using the formulation of Robinson 5 p 0 188745 22 92 A Avy 2 i i 3 3 4 9 gs a 7213 E eea tuk Dae A Ai Z Z Zr ZP glei ei 0 40244e7 3 4008 n fi Ess 3 1 kg e i 1 Ei where the summation is over the components of the material with atom fractions f 19 The Resource Option The RESOURCE option allows the user to edit the data available on a history file while altering the assumed spatial distribution of the source from that used in the original calcu lation For its application see reference 1 20 The Merge Option Not used in HTAPESX For any tally either the HISTP file or the HISTX file is edited but not both 21 The Time Convolution Option Assume that an initial calculation has been made with the default source time distribution i e all histories start at t O A time dependent tally for any of the allowed LAHET source time distributions may then be made with HTAPESxX without rerunning the transport calcu lation For details see reference 1 150 MCNPX User s Manual MCNPX User s Manual i Version 2 3 0 April 2002 LA UR 02 2607 Acce
414. ortance 0 0 0 weight cutoff 0 0 0 weight cutoff 0 0 0 energy importance 0 QO 0 energy importance 0 0 0 dxtran 0 0 0 dxtran 0 0 0 forced collisions 0 0 0 forced collisions 0 0 0 exp transform 0 QO QO exp transform 0 0 QO upscattering 0 0 0 downscattering 0 0 9 8423E 00 tabular sampling 7166 3 5830E 01 1 8289E 00 capture 0 1 4179E 02 7 6277E 02 n xn 78791 3 9358E 00 1 9090E 01 loss to n xn 25324 1 2646E 00 4 9542E 01 fission 0 0 0 loss to fission 0 0 0 photonuclear 0 0 nucl interaction 3433 1 8665E 01 6 2865E 01 tabular boundary 0 QO 0 tabular boundary 0 QO 0 gamma xn 0 0 0 particle decay 0 0 0 adjoint splitting 0 0 0 total 394256 1 9709E 01 3 4116E 02 total 394256 1 9709E 01 3 4116E 02 number of neutrons banked 368932 average time of shakes cutoffs neutron tracks per source particle 1 9713E 01 escape 5 7563E 00 tco 1 0000E 34 neutron collisions per source particle 2 7817E 01 capture 4 6071E 01 eco 0 0000E 00 total neutron collisions 556332 capture or escape 5 7522E 00 wcl 5 0000E 01 net multiplication 0 0000E 00 0000 any termination 5 3292E 00 we2 2 5000E 01 Net neutron production for this case is 18 285 n p which is 0 1 greater than the base case value Thus as for neutrons the new 150 MeV proton evaluations for lead predict higher neutron production by protons in the energy range 20 to 150 MeV than does the Bertini INC model Since the proton evaluations are considered to be more accurate than the Berti
415. ory Cutoff 0 00000 eee 89 5 5 6 3 CTME Computer Time Cutoff annaa aaa aaaea 89 5 5 7 Physics Models 2000 e eee eee eee eee eee eee 89 507 LE GAM Sanna eae o a n e tno thai wan Bre a hoe nth Suge ter a EE 90 BO fed GB tee tee aA area eee nt eaten cae A a A Maen Ie 92 55 FOELEN ya dss gerund Graveney tink a a a T beh ee antadl a me a vaste anabes Ae 94 DO MALEB a eo RR ee ee a 95 5 6 Source Specification 00 00 e eens 97 5 6 1SDEF General Source Definition 000005 97 5 6 1 1 SIn Source Information 0 0 naana anaana 99 5 6 1 2SPn Source Probability 000000 e eee eee 99 5 6 1 3SBn Source Bias 0 0 cc ees 100 5 6 1 4DSn Dependent Source Distribution 101 5 6 1 5 SCn Source Comment 0 00 0 cc eee eee ee 102 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 5 6 2 KCODE Criticality Source 000 cece eee 102 5 6 3 KSRC Source Points for KCODE Calculation 102 5 6 4 SSW Surface Source Write 20 20 e eee 103 5 6 5SSR Surface Source Read 0 0e cece eee eee eee 104 5 6 6 Subroutines SOURCE and SRCDX 0c eee e eee eens 107 5 6 7 Extended Source OptionS 0002 c eee eee 107 5 7 Tally Specitication 2 005 5 0 15 natin Ss aak wie dd ee eee ae Si ee 111 SA1 Pinas Tally cessed oecis ated parce aie are epee eee DE te eee 112 5 7 1 1
416. osines of the angles O4 O2 O3 displacement vector of the transformation XX YX ZX XY YY ZY XZ YZ ZZ rotation matrix of the transformation 1 the default means that the displacement vector is the location of the origin of the auxiliary coordi nate system defined in the main system 1 means that the displacement vector is the loca tion of the origin of the main coordinate system defined in the auxiliary system Use Convenient for many geometries Example 10 1 fill 1 rcc can 2 2 7 8 2 u 1 30 2 u 1 21 like 1 but trcl 200 0 45 45 90 1354590 9090 0 fill 2 Cell 21 is like cell 1 but is translated to x y z 20 0 0 and rotated 45 counter clockwise with respect to x and y If if the rotational matrix is left incomplete MCNPX will calculate what it should be but completeness is the only way to be sure you get what you want and get error messages if you are wrong 5 3 3 6 LAT Lattice Form LAT n on cell card LAT n1n2n3 nx data card Table 5 24 Lattice Card Argument Description 1 cell describes a rectangular square lattice 2 cell describes a hexagonal triangular lattice 72 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 24 Lattice Card Argument Description lattice type for corresponding cell 1 x use jump feature n1 nx to pass over cells which are not lattice c
417. osines of the arbi trary vector with respect to the x y and z axes The vector need not be normalized The surface current tally represents the time integrated current integrated over a surface area and an element of solid angle Unless otherwise normalized it is the weight of parti cles crossing a surface within a given bin per source particle As such it is a dimensionless quantity 4 Edit Option IOPT 2 or 102 Surface Flux The surface flux tally is analogous to an MCNP F2 tally All particle types listed above may be specified The number of energy bins is given by NERG The number of particle types for which surface flux data is to be tallied is given by NTYPE and must be gt 0 NFPRM is unused If KOPT 1 surface segmenting is performed as in option above the same input record to designate the segmenting planes or cylinders must be included as in option 1 If IOPT is preceded by a minus sign the particle weight is multiplied by its energy before tallying The surface flux tally represents the time integrated flux integrated over surface areas Unless otherwise modified it is a dimensionless quantity MCNPX User s Manual 143 MCNPX User s Manual E Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium 5 Edit Option IOPT 3 or 103 Particle Production Spectra Option 3 may be used to tally the spectra of particles produced in nuclear interactions It accesses all collision records on HI
418. ositron 19 positron 20 neutrino neutrino antineutrino 21 antineutrino LA UR 02 2607 FNORM may be used to apply an overall multiplicative normalization to all bins except for IOPT 11 111 12 or 112 For these cases FNORM multiplies the time variable e g use FNORM 0 001 to convert from nanoseconds to microseconds The default is 1 0 KPLOT is a plot control flag plotting is available for some options provided it has been installed with the code using the LANL CGS and CGSHIGH Common Graphics System libraries Using a 0 indicates that no PLOT file will be produced and is the default IXOUT is a flag to indicate that the tally will be written to a formatted auxiliary output file for post processing The details and the file name are option dependent however a 0 indicates that no such file will be written and is the default MCNPX User s Manual 139 MCNPX User s Manual E Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium IRS is the RESOURCE option flag A non zero value indicates that the option will be turned on 0 is the default see Section 19 below IMERGE is not used in HTAPE3X see Section 20 below ITCONV is the TIME CONVOLUTION option flag A non zero value indicates that the option will be turned on 0 is the default see Section 21 below IRSP is the RESPONSE FUNCTION option flag IRSP gt 0 indicates that the tally will be multiplied by a user supplied response
419. p total rather than flagged or uncollided flux ly of last user bin Is of last segment bin Iu of first multiplier bin on FMn card Ic of last cosine bin Ig of last energy bin Ir of last time bin 4 Use Whenever one or more tally bins are more important than the default bin Particularly useful in conjunction with the weight window generator Example Suppose an F2 tally has four surface entries is segmented into two segments the segment plus everything else by one segmenting surface and has eight energy bins By default one chart will be produced for the first surface listed for the part outside the segment and totaled over energy If we wish a chart for the fifth energy bin of the third surface in the first segment we would use TF2 3 2J 1 2J 5 MCNPX User s Manual 135 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 5 7 20 TIRn The Radiography Tally MCNPX can generate simulated radiography images as one would expect to see from an X ray or pinhole projection of an object containing the particle source This allows the recording of both the direct source image as well as that due to background scatter This tool is an invaluable aid to the problem of image enhancement or extracting the source image from a background of clutter MCNPX includes two types of image capability the pinhole image projection and the transmitted image projection The radiography capability is based on point
420. p mfact trans n 1 11 21 31 note number must not duplicate one used for an F1 tally P is a particle type There is no default see Table 5 1 Table 8 1 Track Averaged Mesh Tally type 1 Keyword Descriptions Keyword Description traks The number of tracks through each mesh volume flux The average fluence is particle weight times track length divided by volume in units of number cm If the source is considered to be steady state in particles per second then the value becomes flux in number cm second default dose Causes the average flux to be modified by an energy dependent dose function The dose keyword may be followed by up to four entries where e Ifthe first entry is 1 to 9 an energy dependent dose function must be sup plied by the user on a MSHMF card e Ifthe first entry is 10 20 31 35 or 40 the dose function comes from the function dfact see Section 8 4 for details The next three entries define the input needed by that function the four needed entries correspond to DFACT arguments ic it iu and acr e If no entries follow the dose keyword the default entries are 10 1 1 and 1 0 which form inputs into the dfact function Results are in rem hour popul Causes the population to be scored in each volume which is equivalent to the weight times the track length pedep Scores the average energy deposition per unit volume MeV cm source parti cle for the partic
421. p of the plot The allowed values of n are 1 and 2 The maximum length of aa is 40 characters The default is the TITLE n aa comment on the FC card for the current tally if any Otherwise it is the name of the current RUNTPE or MCTAL file plus the name of the tally KCODE plots have their own special default title Put the title below the plot instead of above it BELOW PELAN has no effect on 3D plots Write subtitle aa at location x y which can be SUBTITLE x y aa anywhere on the plot including in the margins between the axes and the limits of the screen Use aa as the title for the x axis The default is the AMEE pa name of the variable represented by the x axis n Use aa as the title for the y axis The default is the YTITLE aa name of the variable represented by the y axis ZTITLE aa Use aa as the title for the z axis in 3D plots The default is the name of the variable represented by the z axis MCNPX User s Manual 53 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 3 MPLOT amp MCPLOT Commands Command Description Use aa as the label for the current curve It is printed in the legend beside a short piece of the kind of line used to plot the curve The value of LABEL reverts to LABEL aa its default value blank after the current curve is plotted If LABEL is blank the name of the RUNTPE or MCTAL file being plotted is prin
422. particle set 2 KCODE criticality calculations have not been extended to include 20 150 MeV neu trons Accelerator transmutation applications should keep criticality limitations in mind when using this feature to include high energy neutrons in the physics based energy region Below 20 MeV MCNPX criticality calculations match MCNP 3 Certain weight window optimizations have not been fully implemented for high energy particles 4 The Mix and Match feature has yet to be implemented This version of MCNPX will not switch between table based and physics based data where a number of tables with differing upper energies are present The switch between physics models and tabular data is made at one energy for all materials in the problem This energy is set on the PHYS card by the user see section 5 5 2 Therefore it is desirable that one use a set of libraries all with the same upper energy limits Correctly implementing this feature involves a major rewrite of data structures in MCNPX and will be released in a future version 5 Charged particle reaction products are not included for some neutron reactions below 20 MeV in the LA150N library In calculating total particle production cross sections the library processing routines include only those reactions where complete angular and energy information is given for secondary products The new 150 MeV evalua tions are built on top of existing ENDF and JENDL evaluations which typically go
423. pes 5 4 7 XSn Cross Section File n 1to 999 Form XSn ZAID nnx AW Use XSDIR file entry for nuclide s not in XSDIR file 5 4 8 VOID Material Void Form VOID no entries or VOID Cy Co Ci C cell number Default Use problem materials Use Debugging geometry and calculating volumes 78 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 5 4 9 PIKMT Photon Production Bias Form PIKMT Z IPIK MT PMT MTy ipik PMT ipik Z IPIK MT PMTy7 MT y ipik PMT ipik Table 5 29 Photon Production Bias Argument Description Z the ZAID of the it entry Full or partial ZAIDs can be specified that is 29000 is equivalent to 29000 50 0 no biasing for ZAID photons from ZAID are produced with the normal sampling technique 1 no photons are produced from ZAID IPIK gt 0 there is biasing for ZAID The value of IPIK is the number of partial photon production reactions to be sam pled MT identifiers for the partial photon production reac J tions to be sampled only used if IPIK gt 0 PMT control to a certain extent the frequency with which oJ the specified MTs are sampled only used if IPIK gt 0 Default If the PIKMT card is absent there is no biasing of neutron induced photons If PIKMT is present any ZAID not listed has a default value of IPIK 1 Use Only useful for biasing photon production
424. ples illustrate the syntax when only the constant multiplier feature is used All parentheses are required in these examples Tally 2 creates 20 bins the flux across each of surfaces 1 2 3 and 4 with each multiplied by each constant C4 Co C3 C4 and the sum of the four constants Tally 12 creates 4 bins the flux across each of surfaces 1 2 3 and 4 with each multiplied by the constant C Tally 22 creates 12 bins the flux across surface 1 plus surface 2 plus surface 3 the flux across surface 4 and the flux across all four surfaces with each multiplied by each constant Cy Co C3 and C4 An FQn card with an entry of F M or M F would print these bins of the tallies in an easy to read table rather than strung out vertically down the output page Example 6 F4 p 1 FM4 1 2 5 6 SD4 1 F6 p 1 SD6 Multiplying the photon flux by volume SD4 1 times the atom density 1 for material 2 times the photon total cross section 5 times the photon heating number 6 is the same as the F6 p photon heating tally multiplied by mass SD6 1 namely the total energy deposition Note that the positive reaction numbers are photonuclear reactions Example 7 F4 n 1 FM4 1 3 6 7 SD4 1 Multiplying the neutron flux by volume SD4 1 times the atom density 1 for material 3 times the fission multiplicity lt nu gt 7 times the fission cross section 6 gives the track length estimate of criticality for cell 1 The FM card basically multi
425. plies by any tallied quantity flux current by any cross section to give nearly all reaction rates plus heating criticality etc Some common reaction numbers are MCNPX User s Manual 125 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Neutrons Photons Protons Photonuclear 1 5 1 1 Total cross section 4 6 4 4 Heating number 6 Fission A more comprehensive list is in Appendix G of the MCNP4C manual Several more examples of the FMn card are in Chapter 4 The DEMO example in Chapter 5 also illustrates the general form of the card 5 7 8 DEn and DFn Dose Energy and Dose Function Form DEn A E Ek DFn B F Fk DF n iu j fac F int ic i Table 5 65 User Specified Dose Energy amp Dose Function Cards Variable Description n tally number E an energy in MeV F the corresponding value of the dose function A LOG or LIN interpolation method for energy table B LOG or LIN interpolation method for dose function table Keyword Value 1 US units rem hr ia 2 international units sieverts hr default normalization factor for dose default 1 0 fac 1 ICRP60 1990 normalization 2 LANSCE albatross response function energy interpolation dose interpolation always linear int log loglin interpolation default lin _ linlin interpolation ic standard dose function 126 MCNPX User s Manual MCNPX User s Manual
426. protons were transported Table A 1 Neutron Problem Summaries Neutron Proton caso momode cece Mansiton mansion MeV MeV base Bertini nh 150 0 1 Bertini nh 20 0 2 Bertini nh dtsa 150 0 3 ISABEL nh 150 0 4 Bertini nh 150 150 5 CEM nh 150 0 For the sake of brevity we reproduce here just the neutron problem summaries from the MCNPxX output decks Base Case sample problem spallation target c neutron production with 20 MeV neutron transition energy c EJ Pitcher 1 Nov 99 c c cell cards c c Pb target 11 11 4 1 2 3 192 MCNPX User s Manual c bounding sphere 20 1 2 3 4 c outside universe 30 4 c surface cards 1 pz 0 0 2 pz 30 0 3 cz 5 0 4 so 90 0 c material cards c c Material 1 Pb without Pb 204 m1 82206 24c 0 255 82207 24c 0 221 82208 24c 0 524 c c data cards mode nh imp n h 1 1r0 phys n 1000 j 150 phys h 1000 j 0 Ica j jj MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 193 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 nps 20000 prdmp j 30j 1 c c source definition c 1 GeV proton beam 7 cm diam parabolic spatial profile sdef sur 1 erg 1000 dir 1 vec 0 0 1 rad d1 pos 0 0 0 par 9 sil a 0 0 0 1 0 2 0 3 0 4 0 5 0 6 0 7 0 8 0 9 1 0 1 1 1 2 1 3 1 41 5 1 6 1 7 1 8 1 9 2 0 2 1 2 2 2 3 2 4 2 5 2 6 2 7 2
427. provides surface flux and current edits which supplement the standard MCNP tallies 1 The HTAPE3X Code HTAPE3X is a modification of the HTAPE code from the LAHET Code System 1 designed to provide analysis of the history file HISTP optionally written by MCNPX 2 It is primarily intended to provide an analysis of the outcome of collisions in the medium and high energy range where the interaction physics is obtained from LAHET However all appropriate features have been retained even when they duplicate existing MCNP flux and current tallies 3 The latter features relate to editing a surface source write SSW file default name WSSA For experienced LAHET users they do provide some options not available with standard MCNP F1 and F2 tallies Note that the information written to HISTP comes only from interactions processed by the medium and high energy modules in MCNPX low energy neutron and proton and any photon electron collisions which utilize MCNP library data do not contribute to the collision information on the history file and will not contribute to edits by HTAPE3X of collision data Surface crossing edits from data on the file WSSA will apply to all particle types and all energies 2 Input for HTAPE3X The input structure is largely unchanged from the description in reference 1 In general energy units are MeV time units are nanoseconds and length units are centimeters Note the difference in the time scale from
428. ption Type of energy deposition scored e total energy deposited from any source default total e de dx ionization from charged particles de dx recol f f eS tlest delct recol energy transferred to recoil nuclei above tabular limits tlest track length folded with tabular heating numbers e delct non tracked particles assumed to deposit energy locally Can have from one to four numerical entries following it e The value of the first entry is in reference to an energy dependent response function given on a MSHMFn card no default e The second entry is 1 default 1 for linear interpolation and 2 for loga rithmic interpolation e Ifthe third entry is zero default 0 the response is a function of energy mfact deposited otherwise the response is a function of the current particle energy e The fourth entry is a constant multiplier and is the only floating point entry allowed default 1 0 If any of the last three entries are used the entries preceding it must be present so that the order of the entries is preserved Only one mfact keyword may be used per tally Allows one to record in a separate mesh array the local energy deposition only due to particles otherwise not considered or tracked in this problem This allows the user to ascertain the potential error in the problem caused by nterg allowing energy from non tracked particles to be deposited locally This can be a serious problem in neglect
429. ption to track energy deposition from one type of particle alone in a problem is included in the first Mesh Tally type see Table 8 1 keyword pedep The Energy Deposition Mesh Tally described here will give results for all particles tracked in the problem and has no option to specify a particular particle The request to track energy deposition by specific particle was received after this tally was developed and therefore was included in the more convenient Mesh Tally type 1 pedep keyword Note since the mesh is independent of problem geometry a mesh cell may cover regions of several different masses Therefore the normalization of the output is per mesh cell vol ume MeV cm source particle not per unit mass R C S MESHn total de dx recol tlest delct mfact nterg trans n 3 13 23 33 Table 8 3 Energy Deposition Mesh Tally type 3 Keyword Descriptions Keyword Description total Type of energy deposition scored de dx e total energy deposited from any source default recol tlest gy CoP y delct e de dx ionization from charged particles e recol energy transferred to recoil nuclei above tabular limits tlest track length folded with tabular heating numbers e delct non tracked particles assumed to deposit energy locally 98 MCNPX User s Manual MCNPX User s Manual i Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium Table 8 3 Energy Deposition Mesh T
430. r Production of Tritium Julich model A second model is the mass dependent model developed for the Julich ver sion of HETC DRE81 In MCNPX it is applied as originally formulated independent of energy but could be used as the low excitation limit in the Ignatyuk model HETC model The third option is the mass and isospin dependent model originally used in the evaporation model of HETC DRE81 1 yA 2Z PEA Ae ONE bo where the default values b 8 0 and y 1 5 may be changed by the user 4 1 6 High Energy Fission Two models for fission induced by high energy interactions are included in MCNPX e The ORNL Model BAR81 e The Rutherford Appleton Laboratory RAL model ATC80 The RAL model allows fission for Z 71 and is the default in MCNPxX It is actually two models one for actinide and one for subactinide fission The ORNL model covers fission only for actinides The subactinide fission routines of the RAL model produce cross sections which tend to be low compared to the most recent data and use of pre equilibrium models further reduce these values This is strong indication that improvements in subactnide fission models are warranted MCNPX User s Manual 45 MCNPX User s Manual Ap Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium 4 2 High Energy Interactions MCNPxX version 2 3 0 contains an early version of the FLUKA high energy code Formally this consists of t
431. r be rouletted more harshly than one in MXSPLN MASPEN MXSPLN 2 in zero window cells or meshes Required MXSPLN gt 1 decides where to check a particle s weight MWHERE 1 check the weight at collisions only 0 check the weight at surfaces and collisions 1 check the weight at surfaces only decides where to get the lower weight window bounds lt 0 get them from an external WWINP file SWITCHN 0 get them from WWNi cards gt 0 set the lower weight window bounds equal to SWITCHN divided by the cell importances from the IMP card 0 energy dependent windows WWE card MTIME 1 time dependent windows WWE card gt 1 multiplicative constant for all lower weight bounds on WWNI n cards or MOET WWNP file mesh based windows of particle type n 5 8 5 WWN Cell Based Weight Window Bounds Form WWNi n Wig Wi2 e Wij e Wig Table 5 90 Cell based Weight Window Bounds Variable Description n particle designator 156 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 90 Cell based Weight Window Bounds Variable Description lower weight bound in cell j and energy or time interval 1 lt E lt E E Eg 0 as defined on the WWE card If no Wij WWE card i 1 0 no weight window game 1 zero importance cell J number of cells in the problem Default None Use Weight w
432. r data tables for neutron proton and photonuclear reactions cross sections for the Bertini model BERTIN gamma emission data for decaying nuclei PHTLIB photon and electron interaction libraries and others Numerous questions in the beta test phase of MCNPX User s Manual 33 MCNPX User s Manual Ap Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium MCNPX have arisen concerning where these libraries should be kept and this section of the manual has been added for clarification The following set of nuclear data libraries may be used with MCNPX 2 3 0 1 All standard neutron libraries used with MCNP4B DLC189 can be used with MCNPxX however they will not contain emission data for charged particles or recoil nuclei these were processed only in the LA150N library Therefore charged second aries and recoil nuclei will not be produced or tracked in MCNPX within the tabular energy ranges 2 MCNP4C DLC200 libraries are the same as the MCNP4B DLC1839 set with certain new features These include unresolved resonances delayed neutrons new electron libraries ZAIDs end in 03e ENDL92 data and multi temperature U Np tables DLC200 tables may be used with MCNPX with the following cautions None of the DLC200 tables have charged particle or recoil data therefore these will not be produced or tracked in MCNPX Only the DLC200 electron tables with ZAID numbers ending in 0le will work prop erly i
433. r this case follows sample problem neutron creation source nucl particle decay 0 weight window 0 cell importance 0 weight cutoff 0 energy importance 0 dxtran 0 forced collisions 0 exp transform 0 upscattering 0 0 0 0 0 1 0 0 8 interaction tabular sampling n xn fission photonuclear tabular boundary gamma xn adjoint splitting total number of neutrons banked neutron tracks per source particle neutron collisions per source particle total neutron collisions net multiplication tracks spallation target Case 3 weight energy neutron loss tracks per source particle QO escape 351353 5102E 01 3 2679E 02 energy cutoff 0 0 time cutoff 0 0 weight window 0 0 cell importance 0 0 weight cutoff 0 0 energy importance 0 0 dxtran 0 0 forced collisions 0 QO exp transform 0 0 downscattering 0 QO capture 0 9089E 00 1 8916E 01 loss to n xn 25121 0 loss to fission 0 0 nucl interaction 3823 0000E 05 7 4505E 03 tabular boundary 1 0 particle decay 0 S Q 9011E 01 3 4571E 02 total 380298 355177 average time of shakes 1 9015E 01 escape 5 7572E 00 2 6865E 01 capture 4 9166E 01 537297 capture or escape 5 7530E 00 0 0000E 00 0000 any termination 5 3162E 00 ja 0 0 0 0 0 0 0 0 0 0 1 1 0 y 5 0 weight energy per source particle 7552E 01 2 225 7E 02 0 0 0 0 0 0 QO QO QO 9 3603E 00 3946E 02 7 4771E 02 2545E 00 4 9306E 01 0 91
434. r use in the transport pro cess Anew cross section treatment PRA98a provides a defined explicit reaction cross section as well as a defined nuclear elastic cross section previously utilized in the absence of data libraries these defined cross sections determine the transport process and constrain the corresponding reaction rates The new cross section treatment has been implemented including an interpolation table for neutron elastic and reaction cross sections derived from the new 150 MeV MCNPX neutron libraries CHA99a and some older 100 MeV libraries Elastic scattering for pro tons is as implemented in LAHET2 8 PRA96 Proton reaction cross sections are obtained by the methods of Barashenkov and Polanski BAR94 with Madland s optical model calculations MAD88 used where available augmented by the coding of Tripathi TRI97a TRI97b below 1 GeV and by the methods from FLUKA89 Moehring formulation MOH83 above 1 GeV Beyond the range of the new tabular data neutron reaction cross sections are similarly obtained Elastic and reaction cross sections for pions are derived from the methods of Barashenkov and Polanski and of FLUKA89 For antinucleons and kaons there are no elastic cross sections available and the reaction cross sections are obtained only from the FLUKA89 methods 4 3 1 4 Atomic Mass Tables MCNPX 2 3 0 includes a new atomic mass data base PRA98a and the code to access it this is used by all the physics packages
435. ransport step that they will skip over some energy bins set up in a tally causing a picket fence structure in energy spectra Figure 8 4 illustrates this effect which will show up in any spectra plotted as a function of energy for an 800 MeV proton beam hitting a tungsten target Figure 8 4 Effect of too fine binning on energy spectra a Proton Energy deposition spectra with 100 bins Note the picket fence effect at high energies file runtpe tally 6 10 6 tally mev particle o 100 200 300 40 500 600 700 800 energy mev MCNPX User s Manual MCNPX User s Manual i Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium b Proton energy deposition spectra with 50 bins Picket Fence effect has disappeared file runtpe tally 26 10 5 tally nev particle 10 6 10 7 o 100 200 300 400 500 600 700 aao energy mev The exact treatment of energy deposition depends on particle type Photons In a photon only problem the photon heating numbers are used to estimate the energy deposition as a function of track length in the cell In cells where the electrons that would be produced cannot travel very far this is a reasonably good approximation since the use of heating numbers assumes that the energy from these would be secondary particles is deposited locally However if the cells are thin to electron transport this becomes a poor approximation and one should use
436. rate from the lower energy evaluated library cross sections The Monte Carlo methodology was adapted from the HERMES code CLO88 with a rewritten sampling algorithm for the center of mass scattering angle Elastic cross section data below 400 MeV uses a global medium energy nucleon nucleus phenomenological optical model potential This is an intermediate step in the effort to provide a library of both elastic and non elastic cross sections from a global optical model potential for MCNPX usage up to 2 GeV incident energy The tabulated elastic scattering cross sections were generated with an interim global medium energy nucleon nucleus phenomenological optical model potential MAD88 The potential is based upon a relativistic Schrodinger representation and is applicable to neu tron and proton incident energies in the range 50 500 MeV and a target mass of 20 220 The starting point for this work was the proton optical potential of Schwandt et al SCH82 for the range 80 180 MeV The potential was modified to optimally reproduce experimental proton total reaction cross sections as a function of energy while allowing only minimal deterioration in the fits to other elastic proton scattering observables Further modifications in the absorptive poten tial were found necessary to extrapolate the modified potential to higher energies At this point explicit isospin was introduced and the potential was converted to a neutron nucleus potential by use of
437. rce Probability Functions ERG 6 ab Muir velocity Gaussian fusion spectrum ERG 7 ab Spare DIR RAD or EXT 21 a Power law p x c x DIR or EXT 31 a Exponential p w ce TME 41 ab Gaussian distribution of time 5 6 1 4 DSn Dependent Source Distribution Form DSn optionJ Jk or DSn T l Ji wee I Jk or DSn Q Vi Sy ai VkSk Table 5 49 Dependent Source Distribution Card Variable Description n distribution 1 999 Determines how J s are interpreted Allowed values are blank or H source variable values in continuous distribu tion for scalar variables only option L discrete source variable values follow p S distribution numbers follow T values of the dependent variable follow values of the independent variable which must be a discrete scalar variable Ii values of the dependent variable Q distribution numbers follow values of the independent variable which must be a scalar variable y monotonically increasing set of values of the inde f pendent variable Si distribution numbers for the dependent variable Default DSn H J Jk MCNPX User s Manual 101 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 5 6 1 5 SCn Source Comment Form SCn comment n distribution number n 1 999 The comment is printed as part of the header of distribution n in the source distribution table and in the source distribution frequency ta
438. rd The general rule in MCNP is that the particle with the lowest IPT value see Table 5 1 specified on the MODE card will be the source particle Thus MODE n h would result in a neutron source A modification has been made to the built in function for source probability F 41 The gaussian distribution in time has been converted to a gaussian distribution in space in order to model accelerator particle beams This modification is discussed in Section 6 3 6 1 5 Tally Specification Cards F1 p F2 p F4 p F5a p F6 p F7 n F8 p Any card with a particle designator can accept all new particle types The F6 energy dep osition options have been changed to accommodate the larger particle list A new F6 tally has been added to tally energy deposition from individual particle types see Section 8 3 New Mesh Tally and Radiography tally capabilities have been added See Sections 8 1 and 8 2 6 1 6 Material Specification Cards Mm DRXS TOTNU NONU AWTAB XSn VOID PIKMT MGOPT No changes have been made to any material specification cards for neutron problems We have made the designation of materials with more than one density a fatal error due to non linear density scaling effects for charged particle transport We recommend defin ing materials with more than one density should this case be encountered The fatal error can be overridden by setting the 19th entry of the DBCN card to a non zero value This will disable all fatal errors so the user should
439. re not Linux compatible Gridconv may be compiled with a nopaw option see table 3 1 Once gridconv is compiled one need type only the word gridconv to execute the code The code will then prompt the user for information that is required such as file type file names etc In most cases the default value is used and a return is all that is necessary Once the header information from mdata has been read from the file gridconv can either produce an ASCII file from a binary or generate the required graphics input files as requested by the user Note that the ASCII file contains raw data not normalized to the number of source particles The reason for the option to write an ASCII file is that some times users will want to look at the numbers in the mdata file before doing any plotting or check the numerical results for a test case The ASCII option is also very useful for porting the mdata file to another computer platform and for reading the data into graphics pack ages not currently supported by gridconv Gridconv is currently set up to generate one two or three dimensional graphics input files with any combination of binning choices Once the input file has been generated gridconv gives the user the options of producing another file from the currently selected mesh tally selecting a different mesh tally available on this mdata file or reading informa tion from a different file Of course there is always the option to exit the pro
440. recompile the code Look for the variable currently holding the string usr local xcodes3 lcsdir ber tin and the similar variable referencing a location for phtlib Change them to reflect the appropriate location of the two data files on your system and re make the code A typical location for these two files might be usr local lib mcnpx This would be the preferable method when a community of users is accessing one copy of the code ona single system As suggested above we recommend making a symlink to the bertin and phtlib files in your working directory If you have more than just one person running the code from a server then it is probably worthwhile to edit src Ics inbd F to point to a specific location on your system where everyone can get the files as in method 2 above In the future we will build in the ability to look for all libraries using the same method now used for the nuclear data table libraries 36 MCNPX User s Manual MCNPX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium 4 Physics and Data The definitions of low intermediate and high energy physics are greatly influenced by the background of the user of a simulation code In reactor physics a 14 MeV neutron is considered high energy but to a particle physicist such an energy is extremely low There is however a basis for division for these categories that can be made in the context of Monte Car
441. rectory so that they all match your particular computing environment The full structure is now in place to allow a graceful migration to individual feature tests during the autoconfiguration process in the future The autoconf generated configure script will search for GNU compilers first before attempting to locate any other compiler present on your computing environment Please be aware of exactly how many Fortran and C compilers exist in your comput ing environment It may be necessary to specify which Fortran and C compiler should be used You have that power via options given to the configure script See the with FC and with CC options later in this document Rather than having the one Build directory of past distributions one is now free to create as many build directories as desired anywhere one wants named anything one wants Through the use of options supplied to the configure script one can vary the resulting gen erated Makefiles to match a desired configuration MCNPX User s Manual 15 MCNPX User s Manual Ap Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium Most software packages that use autoconf have a basic build procedure that looks like gzip dc PACKAGE tar gz tar xf cd PACKAGE configure make install This method of installation works with MCNPX However the development team recom mends a slightly different method so as not to clutter the original source tree with all th
442. rgy Interactions ei sociau eai a a eee 46 4 3 Nuclear Data Tables 0 00 cece eee 46 4 3 1 Nuclear Data Libraries 0 0000 c cc eee 46 4 3 2 Photoelectric Interactions 20 00 cee 54 5 Multiparticle Extensions and General Tracking 65 5 1 Reaction Probability Calculation 0 000 ccs 68 5 2 Collisional Stopping Power for Heavy Charged Particles 68 5 3 Energy Straggling for Heavy Charged Particles 0 0 00 ee aee 69 5 4 Multiple Scattering for Heavy Charged Particles 000 00a 69 6 MCNPX Input Files 05 0 eae ete sud ne Seda we eines va 71 6 1 MCNP Card Modifications and Additions 0 0 00 0 cc eee eee 71 6 1 1 Problem Type Card 0 000 cee 71 6 1 2 Geometry CardS 0 0 tees 71 6 1 3 Variance Reduction Cards 000 cece eee 71 6 1 4 Source Specification Cards 20 0 ee 72 6 1 5 Tally Specification Cards 0 0c teens 72 6 1 6 Material Specification Cards 00 0 c eee 72 6 1 7 Energy and Thermal Treatment Cards 0 000000 eee 73 6 1 8 Problem Cutoffs Cards 0 c cc etna 75 6 1 9 Peripheral Cards 0 0 0 ccc tees 76 6 1 10 New Cards Specific to MCNPX 0 0 0 cece eee 76 6 2 Physics Module Options 00 60 c cece eee eee eee 76 6 3 Extended Source Options 0 0000 c eects 84 7 New Variance Reduction Techniques 2
443. ribution function 5 describes the required surface transformations According to the SI5 card surfaces 6 and 7 are related to surfaces 3 and 2 respectively by transformation TR4 surfaces 12 and 13 are related to 3 and 2 by TR5 The physical probability of starting on surfaces 6 and 7 is 40 according to the SP5 card and the physical probability of starting on surfaces 12 and 13 is 60 The SB5 card causes the particles from surfaces 3 and 2 to be started on surfaces 6 and 7 30 of the time with weight multiplier 4 3 and to be started on surfaces 12 and 13 70 of the time with weight multiplier 6 7 Example 2 Original run SSW3 SYM 1 Current run SSRAXS 0 0 1 EXT D99 S199 1 5 1 SP99C 75 1 106 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 SB99 5 5 All particles written to surface 3 in the original problem will be started on surface 3 in the new problem which must be exactly the same because no OLD NEW COL or TR keywords are present Because this is a spherically symmetric problem indicated by the SYM 1 flag in the original run the position on the sphere can be biased It is biased in the z direction with a cone bias described by distribution 99 5 6 6 Subroutines SOURCE and SRCDX Users may write their own source subroutine source to bypass the standard source capa bilities If there is no SDEF SSR or KCODE card then MCNPX will look for a subroutine SOURCE and if there are det
444. rical or spherical grid overlaid on top of the standard problem geometry Particles are tracked through the independent mesh as part of the regular trans port problem and the contents of each mesh cell written to a file at the end of the problem This file can be converted into a number of standard formats suitable for reading by various graphical analysis packages The conversion program gridconv is supplied as part of the overall MCNPX package section 8 1 2 An example of a mesh tally plot is shown in Fig MCNPX User s Manual 91 MCNPX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium 8 1 This represents a plan view of neutron fluence in a spallation target system Analysis of this data is limited only by the capabilities of the graphical program being used 8 1 1 Setting up the Mesh in the INP File A mesh tally is defined by several cards which are described below All of the control cards for mesh tallies must be in a block preceded by a card containing the word tmesh in the first five columns and terminated by a card containing the word endmad in the first five col umns For each mesh tally card the following set of cards must be present which give details on the mesh characteristics CORAn corra n 1 corra n 2 corra n N CORBn corrb n 1 corrb n 2 corrb n N CORCn corre n 1 corre n 2 corrce n N where the CORAn CORBn and CORCn cards are used to describe
445. ril 2002 LA UR 02 2607 Accelerator Production of Tritium DXTRAN Mesh Tally type 4 The fourth type of mesh tally scores the tracks contributing to all detectors defined in the input file for the P particle type If this mesh card is preceded by an asterisk tracks con tributing to DXTRAN spheres are recorded Obviously a point detector or DXTRAN sphere must already be defined in the problem and the tally will record tracks correspond ing to all such defined items in the problem The user should limit the geometrical boundaries of the grid to focus on a specific detector or DXTRAN sphere in order to pre vent confusion with multiple detectors although the convergence of the particle tracks should help in the interpretation This tally is an analytical tool useful in determining the behavior of detectors and how they may be effectively placed in the problem R C S MESHn P trans n 4 14 24 34 note number must not duplicate one used for an F4 tally P is a particle type neutron or photon There is no default see table 5 1 Table 8 4 DXTRAN Mesh Tally type 4 Keyword Descriptions Keyword Description trans Must be followed by a single reference to a TR card that can be used to trans late and or rotate the entire mesh Only one TR card is permitted with a mesh card 8 1 2 Processing the Mesh Tally Results The values of the coordinates the tally quantity within each mesh bin and the relat
446. rimary Therefore only the range e lt 1 2 is of interest With 1 2 the equation for f becomes ae ee p 1 1n2 5 m 5 MCNPX User s Manual 57 MCNPX User s Manual Ap Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium On the right side we can express both E and in units of the electron rest mass Then E can be replaced by t on the right side of the equation We also introduce supplementary constants C2 In 21 C3 1 1n2 C4 5 In2 so the stopping power becomes aa Bainter 2y C2 C3 caf This is the collisional energy loss rate in MeV cm in a particular medium In MCNP MCNPxX we are actually interested in the energy loss rate in units of MeV barns so that different cells containing the same material need not have the same density Therefore we divide this equation by N and multiply by the conversion factor 1074 barns cm We also use the definition of the fine structure constant c _ 2me he where h is Planck s constant to eliminate the electronic charge e from the equation The result is as follows q5 3 10 07h tt 1 B This is the form actually used in MCNP MCNPX to present collisional stopping powers at the energy boundaries of the major energy steps A discussion of how collisional stopping power is implemented for heavy charged particles is found in Section 5 2 fint I Oop Zy A of 2mmc Electron Energy Straggling Becau
447. rom the initial run is in your current directory The complete continue run execution line option is C m or CN m where m specifies which dump from the restart file to pick up with If m is not specified the last dump is taken by default If the initial run producing the restart file was stopped because of particle cutoff NPS card Section 5 5 6 3 NPS must be increased for a continue run via a continue run file CTME in a continue run is the number of minutes more to run not cumulative total time To run more KCODE cycles only the fourth entry KCT may be changed Like NPS KCT refers to total cycles to be run including previous ones In a continue run a negative number entered on the NPS card produces a print output file at the time of the requested dump No more histories will be run This can be useful when the printed output has been lost or you want to alter the content of the output with the PRINT or FQ cards Be cautious if you use a FILES card in the initial run See Section 5 9 10 MCNPX User s Manual 33 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 4 1 3 Message Block In computer environments where there are no execution line messages the message block is the only means for giving MCNPX an execution message Optionally is a convenient way to avoid retyping an often repeated message The message block starts with the string MESSAGE The message block ends with a blank line delimiter before the title
448. rom which the continue run started The new dumps overwrite the old dumps providing a way for the user to prevent unmanageable 32 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 growth of RUNTPE files RUNTPE growth also can be controlled by the NDMP entry on the PRDMP card The optional continue run input file must have the word CONTINUE as the first entry on the first line title card or after the optional Message Block and its blank line delimiter Alphabetic characters can be upper lower or mixed case This file has the following form Message Block Optional Blank Line Delimiter pagna CONTINUE Data Cards Blank Line Terminator Recommended Anything else Optional The data cards allowed in the continue run input file are a subset of the data cards available for an initiate run file The allowed continue run data cards are FQ DD NPS CTME IDUM RDUM PRDMP LOST DBCN PRINT KCODE MPLOT ZA ZB and ZC If none of the above items is to be changed and if the computing environment allows execution line messages the continue run input file is not required only the run file RUNTPE and the C option on the MCNPX execution line are necessary For example the command line sequence MCNPX C or MCNPX CN will pick up the job where it stopped and continue until another time limit or particle cutoff is reached or until you stop it manually This example assumes that a default restart filename f
449. ross sections will be generated Positive values of NANG indicate cosine bin boundaries will be defined negative val ues indicate angle bin boundaries in degrees will be speci fied The default is 0 FNORM An overall multiplicative normalization factor to be applied to all cross sections The default is 1 0 To convert to millibarns use FNORM 1000 0 obtain macroscopic cross sections use an atom density KPLOT A plot control flag the default is 0 Any nonzero value will cause the output to be written to a file XSTAL in the format of an MCNP MCTAL file for subsequent plotting see below 228 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 9 1 Parameter Meaning IMOM Chooses energy or momentum to be used in cross section def inition IMOM 0 cross sections are tabulated by energy MeV and differential cross sections are calculated per unit energy per MeV IMOM not equal 0 cross sections are tabulated by momentum MeV c and differential cross sections are estimated per unit momentum per MeV c IYIELD not equal to 0 estimates differential yields or multiplicities for nonelastic and elastic reactions rather than cross sections The integral over energy and angle for each particle type will be the multiplicity per nonelastic reaction or unity for the elastic scattering of the incident particle if it is included in the calculation LTE
450. s specific to the APT and AAA projects Our commitment to modern software management and quality assurance methods in the development of MCNPX is very strong The code is used for the design of high intensity accelerator category 2 nuclear facilities and has already been used to design a major cat egory 3 activity at the LANSCE high power beamstop MCNPX development is guided by a set of requirements design and functional specification documents Code testing is per formed on a large scale by a volunteer beta test team Code configuration management is involves the CVS system and methods of assessing code development progress are being implemented One of these involves nightly regression testing on a computer farm of over 20 hardware software platforms Training courses are held regularly 1 MCNPX MCNP MCNP4B LAHET and LAHET Code System LCS are trademarks of the Regents of the University of California Los Alamos National Laboratory MCNPX User s Manual xiii MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 We have also developed a unique autoconfiguration build system which allows a variety of compilation options to be easily executed on a large number of platforms MCNPX 2 4 0 extends the previous set of supported platforms to Windows PC This version of the code has also been rewritten in Fortran 90 and many of the code elements recast as modules Work on our component architecture approach is also pro
451. s a detector type tally P is the particle type for the tally Only neutrons or photons are allowed In MCNPX 2 x this card was called Fln P old input files are backward compatible 136 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 77 Pinhole Radiography Argument Descriptions Argument Description X1 Y1 Z1 The coordinates of the pinhole Always 0 zero for this application RO Note neither the pinhole nor the grid should be located within a highly scatter ing media The reference coordinates that establish the reference direction cosines for the X2 Y2 Z2 normal to the detector grid This direction is defined as being from X2 Y2 Z2 to the pinhole at X1 Y1 Z1 If F1 gt 0 the radius of a cylindrical collimator centered on and parallel to the ed reference direction which establishes a radial field of view through the object The radius of the pinhole perpendicular to the reference direction F2 F2 0 represents a perfect pinhole F2 gt 0 the point through which the particle contribution will pass is picked randomly This simulates a less than perfect pinhole The distance from the pinhole at X1 Y1 Z1 to the detector grid along the F3 direction established from X2 Y2 Z2 to X1 Y1 Z1 and perpendicular to this reference vector The grid dimensions are established from entries on FS and C cards In this use the first entry s
452. s give a warning message when you encounter such situations In MCNPX with more charged particles and greatly expanded energy range this for merly small correction now becomes increasingly important and the usual way of handling it is not sufficient We have therefore decided to make using the same mate rial with more than one density a fatal error If you want to run the problem anyway overriding the termination the usual MCNP4B process will be followed but we advise against it Instead we recommend that different materials be defined for areas of dif ferent densities 2 2 Release Notes Several corrections and improvements have been made to MCNPX version 2 3 0 new features have been added to the User s Manual These are summarized below Chapter 2 Warnings Caveats and Revision Notes Caveat regarding overprediction of heating values with th 150 MeV neutron libraries has been removed MCNPX User s Manual MCNPX User s Manual Pr Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium e Caveats regarding KCODE and energy straggling interpolations have been removed e Several known bugs and warnings have been added Chapter 3 MCNPX Installation e MCNP xX installation discussion has been revised to incorporate automated build sys tem Section 3 1 e The Cray computer platform is no longer supported Contact the code developers if you need to use a Cray e Notes on multiprocessing have been
453. s of more than one bin or tally are possible No output is sent to COMOUT MCPLOT will not take plot requests from the terminal and returns to MCRUN after each plot is displayed See Appendix B for a complete list of MCPLOT commands available Another way to plot intermediate tally results is to use the TTY interrupt lt ctrl c gt IMCPLOT or lt ctrl c gt IM that allows interactive plotting during the run At the end of the history that is running when the interrupt occurs MCRUN will call MCPLOT which will take plot requests from the terminal No output is sent to the COMOUT file The following commands can not be used RMCTAL RUNTPE DUMP and END 5 94 PTRAC Particle Track Output Form PTRAC keyword parameter s keyword parameter s Default See Table 5 108 Use Optional Table 5 108 PTRAC Keywords Parameter Values and Defaults Keyword Parameter Values Default Entries MCNPX User s Manual 169 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 OUTPUT CONTROL KEYWORDS BUFFER Integer gt 0 100 1 FILE asc bin bin 1 MAX Integer 0 10000 1 MEPH Integer gt 0 1 WRITE pos all pos 1 EVENT FILTER KEYWORDS EVENT src bnk sur col ter 1 5 FILTER Real Integer Mnemonic 2 72 TYPE n p e 1 3 HISTORY FILTER KEYWORDS NPS Integer gt 0 1 2 CELL Integer gt 0 Unlimited SURFACE Integer gt 0 Unlimited TALLY I
454. s of our work Several visitors have provided invaluable help to the nuclear data team with evaluations notably Dr Satoshi Chiba JAERI and Dr Arjan Koning ECN Petten We would also like to thank members of the Los Alamos Export Controls Office particu larly Sarah Jane W Maynard Crystal Johnson and Steve H Remde for their outstanding help in dealing with the export issues for our foreign beta test team members Publishing Team Finally we wish to thank Berylene Rogers for copyediting and preparing the final docu ment and Patty Montoya Barbara Olguin Arlene Lopez and Jean Harlow for their help in reproducing and assembling the manual iv MCNPX User s Manual Zz MCNPX User s Manual Accelerator Version 2 3 0 April 2002 Production LA UR 02 2607 of Tritium Dedication We dedicate this code to the memory of our respected colleague Dr Russell B Kidman Russ was an invaluable member of the APT Target Blanket design team and a computer simulations expert for many projects at Los Alamos His tragic and premature death has left us all with a deep sense of loss MCNPX User s Manual Accelerator Production of Tritium vi MCNPX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 MCNPX User s Manual Accelerator Production of Tritium Contents Acknowledgments 0000s eee cece eee eeeee Dedication cis3 wii te inser eee wae FIQUICS i iittala as Ghee tind waa a ees Tabl s rie te wd viweke wa
455. s points 166 optional PRINT short output 167 optional MPLOT none 169 optional PTRAC none 169 optional HISTP amp HTAPE3X 171 optional DBCN none 171 optional LOST 1010 173 optional IDUM 0 173 optional RDUM 0 173 optional FILES none none sequential formatted 173 This describes the effect of not using this particular card MCNPX User s Manual 179 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 180 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 6 References ARM73 T W Armstrong and K C Chandler SPAR A FORTRAN Program for Computing Stopping Powers and Ranges for Muons Charged Pions and Heavy lons ORNL 4869 Oak Ridge National Laboratory May 1973 AAR86 P A Aarnio A Fasso H J Moehring J Ranft and G R Stevenson CERN TIS RP 186 1986 FLUKA 86 users guide AAR87 P A Aarnio J Lindgren A Fasso J Ranft and G R Stevenson CERN TIS RP 190 1987 FLUKA 87 AAR90 P A Aarnio e al FLUKA89 Consiel Europeene Organisation pour La Recherche Nucleaire informal report January 2 1990 ART88 E D Arthur The GNASH Preequilibrium Statistical Model Code LA UR 88 382 Los Alamos National Laboratory February 1988 ATC80 F Atchison Spallation and Fission in Heavy Metal Nuclei under Medium Energy Proton Bombardment in Targets for Neutron Beam Spallation Sources Jul Conf 34 Kernforschungsanlage Julich GmbH Jan
456. s time bins Default No weight window values are generated Use Optional 5 8 3 WWGE Weight Window Generation Energies or Times Form WWGE rnE Ep Ei Ex j 15 Table 5 88 Weight Window Generation Energies or Times Variable Description n particle designator upper energy or time bound for weight window group to be E generated E gt E Default If this card is omitted and the weight window is used a single energy or time interval will be established corresponding to the energy time limits of the problem being run If the card is present but has no entries ten energy time bins will be generated with energies times of E 10 8 MeV shake and j 0 Both the single time energy and the energy time dependent windows are generated Use Optional 5 8 4 WWP Weight Window Parameter Form WWP n WUPNWSURVN MXSPLN MWHERE SWITCHN MTIME MULT MCNPX User s Manual 155 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 n particle designator Table 5 89 WWP Keyword Descriptions Keyword Description If the particle weight goes above WUPN times the lower weight bound WUPN the particle will be split Required WUPN 2 If the particle survives the Russian roulette game its weight becomes WSURVN MIN WSURVN times the lower weight bound WGT MXSPLN Required 1 lt WSURVN lt WUPN No particle will ever be split more than MXSPLN for one o
457. same as l 2 but using a 25 MeV potential well for pions 6 The same as l 2 but using a 25 MeV potential well for pions Note Not all the options for the ISABEL INC model have been thoroughly debugged JCOUL 1 Use Coulomb barrier on incident charged particle interactions default 0 No Coulomb barrier for incident charged particles NEXITE 1 Subtract nuclear recoil energy to obtain nuclear excitation energy default 2 Do not subtract nuclear recoil energy NPIDK 1 Force m to terminate by decay at the pion cutoff energy 0 Force 7r to interact by nuclear capture INC when cutoff is reached default Note The capture probability for any isotope in a material is proportional to the product of the number fraction and the charge of the isotope However capture on 1H leads to decay rather than interaction 92 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 41 LCA Keyword Descriptions Continued Keyword Description NOACT Note The use of the NOACT option other than the default is intended as a diagnostic tool allowing other processes to be more easily observed PRA99 2 Attenuation mode transport primary source particles without nonelastic reactions 1 Do not turn off nonelastic reactions default 0 Turn off all nonelastic reactions 1 Compute nuclear interactions of source particles only transport and slowing down are turned off This op
458. se an energy step represents the cumulative effect of many individual random col lisions fluctuations in the energy loss rate will occur Thus the energy loss will not be a simple average rather there will be a probability distribution f s A dA from which the energy loss A for the step of length s can be sampled Landau LAN44 studied this situa tion under the simplifying assumptions The mean energy loss for a step is small compared with the electron s energy 58 MCNPX User s Manual MCNPxX User s Manual Ap Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium e The energy parameter amp defined below is large compared with the mean excitation energy of the medium e The energy loss can be adequately computed from the Rutherford cross section RUT11 The formal upper limit of energy loss can be extended to infinity With these simplifications Landau found that the energy loss distribution can be expressed as f s A dA o A dA in terms of A a universal function of a single scaled variable 2 22 5 dep Here m and v are the mass and speed of the electron is the density effect correction B v c is the mean excitation energy of the medium and y is Eulers constant g 0 5772157 The parameter amp is defined by 2 A a ered hearin 2ne NZ g Ss mv where e is the charge of the electron and NZ is the number density of atomic electrons and the universal function is
459. section 6 1 7 the photon interaction cross section will be the sum of the photoatomic and the photonuclear cross sections Full compatibility with existing MCNPX features such as tallies will be ensured New summary table data are provided with relevant information about photonuclear absorption and secondary particle production Because photonuclear interactions are rare events some form of biasing is useful to enable photonuclear simulations to run in a reasonable time The concept currently imple mented is similar in nature to the forced collision biassing In forced collisions a particle traversing a cell is split in tow one particle is forced to undergo a collision in the material and the other is transported to the cell boundary Both have their weights updated accord ing to the probability that the photon would have undergone a collision before reaching the boundary Biased photonuclear collisions borrow from this model and split the colliding photon in two one particle undergoing a photoatomic collision the other particle undergo ing a photonuclear collisions and both having their weights updated appropriately The initial challenge in making LA150 photonuclear data available for MCNPX lay in pro viding an interface between the data and the code Photoatomic data tables already exist for MCNPX one option was to append the photonuclear data to the photoatomic tables However photoatomic data are determined by interactions with atomic
460. sefulness of this method involves locating the source of particles entering a certain volume or crossing a certain surface The user asks the question If particles of a certain type are present where did they originally come from In shielding problems the user can then try to shield the particles at their source Refinements in this tally will be forthcoming in further versions of MCNPX as user feedback is received This mesh tally is normalized as number per SDEF source particle R C S MESHn P 1 P 2 P 3 P 4 trans n 2 12 22 32 note number must not duplicate one used for an F2 tally 1 In MCNPX version 2 1 5 there was no option to chose individual particles The type 2 Mesh Tally produced source points for all particles in the problem in one plot MCNPX User s Manual 147 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 82 Source Mesh Tally type 2 Keyword Descriptions Keyword Description Particle type i e n p e etc up to 10 particle types see Table 4 1 Source particles are considered to be those that come directly from the source defined by the user and those new particles created during nuclear P i interactions One should be aware that storage requirements can get very large very fast depending on the dimensions of the mesh since a separate histogram is created for each particle chosen If there are no entries on this card the information for ne
461. ser defined response functions for dosimetry monitoring devices function DFACT id ic en it iu acr MCNPX User s Manual 113 Accelerator Production of Tritium ARGUMENT MCNPX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 Table 8 9 DFACT Argument Descriptions DESCRIPTION id Particle identification number 1 neutron 2 photon Choice of conversion coefficient Note The 10 and 20 options are Dose Equivalent H i e absorbed dose at a point in tissue weighted by a distribution of quality factors Q related to the LET distribution of radiation at that point The 30 s options are Equivalent Dose Hy based on an average absorbed dose in the tissue or organ Dy weighted by the radiation weighting factor w summed over all component radiations neutrons 10 ICRP 21 1971 20 NCRP 38 1971 ANSI ANS 6 1 1 1977 31 ANSI ANS 6 1 1 1991 AP anterior posterior 32 ANSI ANS 6 1 1 1991 PA posterior anterior 33 ANSI ANS 6 1 1 1991 LAT side exposure 34 ANSI ANS 6 1 1 1991 ROT normal to length amp rotationally symmetric 40 ICRP 74 1996 ambient dose equivalent photons 10 ICRP 21 1971 20 Claiborne amp Trubey ANSI ANS 6 1 1 1997 31 ANSI ANS 6 1 1 1991 AP anterior posterior 32 ANSI ANS 6 1 1 1991 PA posterior anterior 33 ANSI ANS 6 1 1 1991 LAT side exposure 34 ANSI ANS 6 1 1 1991 ROT normal to length amp rotationally symmetric
462. shared by LAHET and MCNPX MCNPX User s Manual 53 MCNPX User s Manual E Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium 4 3 1 5 Nuclear Structure Data Library PHTLIB The PHTLIB file contains nuclear structure data used in the gamma emission process fol lowing the termination of nucleon and ion emission in residual nuclei Two versions of PHTLIB are now available for use with MCNPX In the original PHTLIB released with MCNPX 2 1 5 all gamma emitting states are allowed to decay to ground Data was generated from CRDL structure data HOW81 This is a valid procedure for calculations where a source terminates and enough time has passed so that no metastable states remain However with new applications for transmutation of wastes it is essential that metastable state information for residual nuclei be calculated in MCNPX for subsequent input into codes such as CINDER 90 WIL97 A new version of PHTLIB is now available which not only updates the gamma emission data and also terminates the emission process for nuclear levels with t 2 gt 1 nsec The budapest_levels dat file compiled by G Molnar et al was obtained from the RIPL project library CHA9Q8 to provide the basis for the new library Data were compared with levels in the CINDER 90 libraries and most discrepancies resolved by reference to Firestone and Shirley FIR96 Improved information about low lying levels was also added We h
463. sion 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium 2 Warnings Known Bugs and Revision Notes Although considerable effort has gone into making MCNPX compatible with MCNP a number of features have not yet been included in MCNPX version 2 3 0 or have not yet been adequately checked out Many of these are works in progress to be released in future versions Currently inoperable features are listed as warnings below In addition the user must be aware of various limitations in certain code features in order to properly use these tools Some of these involve long outstanding problems yet to be resolved in the simulation community particularly involving the extension of variance reduction techniques to charged particles Others involve known features in MCNP which have now become more important in the high energy charged particle environment These are listed as caveats below All of the items listed here form a basis for future work on MCNPX All computer simulation codes must be validated for specific uses and the needs of one project may not overlap completely with the needs of other projects It is the responsibility of the user to ensure that his or her needs are adequately identified and that benchmark ing activities are performed to ascertain how accurately the code will perform The benchmarking process for the Accelerator Production of Tritium project is extensive yet does not cover the entire range of possibl
464. sists of a command keyword in most cases followed by some parameters Keywords and parameters are entered blank delimited no more than 80 characters per line Commas and equal signs are interpreted as blanks A plot request can be continued onto another line by typing an amp before the carriage return but each command the keyword and its parameters must be complete on one line Command keywords but not parameters can be abbreviated to any degree not resulting in ambiguity but must be correctly spelled The term current plot means the plot that is being defined by the commands currently being typed in which might not be the plot that is showing on the screen Only those commands marked with an in the list in section C can be used after the first COPLOT command in a plot request because the others all affect the framework of the plot or are for contour or 3D plots only MCNPX User s Manual 49 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 5 2 3 Plot Commands Grouped by Function Table 5 3 MPLOT amp MCPLOT Commands Command Description Device control Commands default is user s terminal n specifies device type 0 for a terminal with no graphics capability No plots will be drawn on the terminal and all plots will be sent to the graphics metafile TERM 0 is equivalent to putting NOTEK on MCNP s execute line 1 Tektronix 4010 using CGS 2 Tektronix 4014 using CGS TERM
465. sity as a func tion of radius and Fermi motion of the nucleons is taken into account in modeling the interactions In some models the quantum effects of Pauli blocking are taken into account however using this feature usually adds considerably to the computational time MCNPxX offers three choices of INC models the Bertini BER63a BER69 ISABEL and CEM MAS 74 packages The Bertini model is incorporated into MCNPX through the LAHET implementation of the HETC Monte Carlo code developed at Oak Ridge National Laboratory RAD77 An alternative INC model was also adapted for the LAHET code from the ISABEL code YAR78 YAR81 which allows hydrogen helium and antiprotons as projectiles ISABEL is derived from the VEGAS INC code CHE68 It has the capability of treating nucleus nucleus interactions as well as particle nucleus interactions although this capability has not been yet fully tested in LAHET or MCNP It allows for interactions between particles both of which are excited above the Fermi sea The nuclear density is represented by up to 16 density steps rather than the three of the Bertini INC It also allows antiproton anni hilation with emission of kaons and pions As presently implemented only projectiles with A lt 4are allowed and antiproton annihilation is not functional The upper incident energy MCNPX User s Manual 41 MCNPxX User s Manual E Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Trit
466. ssary tree hierarchy and generate Make files at all levels After successful configuration you can now make mcnpx using your new compiler with the following command from the top level of your working directory make mcnpx 3 1 8 Additional Software Requirements If you are a casual user and do not perform any software development for MCNPX capa bilities you must have the GNU make utility version 3 76 or greater See your system administrator if GNU make does not exist on your computing platform If you are a software developer for MCNPX capabilities or you wish to alter the way the autoconf generation of the configure script works you will need the following software GNU make version 3 76 or higher GNU m4 preferably version 1 4 GNU autoconf preferably version 2 13 GNU find preferably version 4 1 makedepend an X Windows routine preferably X Version 11 Release 6 3 1 9 Fortran 90 Compilers We have tried several Fortran 90 compilers with the default static construction method on several systems The following table shows what works and what doesn t This will change frequently so it is best to contact the code developers for the latest results 32 MCNPX User s Manual MCNPX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium Table 3 3 Fortran 90 Compilers Platform Compiler Result Sun Solaris WorkShop Compilers core dumps 5 0 FORT90 RAN 90 2 0 SGI IRIX
467. stem default COPT should be CFLAGS in this table or system computed values If in doubt run the configure script and examine the system default or system computed values that appear in the gener ated Makefile h You may want to include the defaults in the string you specify for COPT with this mechanism COPT settings are always appended to CFLAGS settings when configure is run again 3 1 6 Multiprocessing If you want to create the parallel PVM version of MCNPX use the following configure option with PVMLIB L path to pvm libraries Ifpvm3 lpvm3 It is recommended that you first install PVM as the configure scripts use various PVM environment variables to locate the PVM libraries One can alternatively give the path and library names following the PVMLIB option 3 1 7 Programmer s Notes Autoconf is not new it has been available as a configuration management tool for several years We have just recently adopted its use to simplify the build process for the MCNPX end user community to allow the flexibility to build and keep multiple versions of MCNPX and to improve our software development process MCNPX User s Manual 25 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 3 2 WINDOWS BUILD SYSTEM If you wish to modify the source or recreate the executables you will need the Compaq Visual Fortran CVF compiler version 6 1 or later and the MSVC compiler version 5 0 or l
468. t Physics JETP 5 No 5 1957 749 WAT02 L S Waters J S Hendricks H G Hughes G W McKinney E C Snow Medical Applications of the MCNPX Code 12th Biennial Radiation Protection and Shielding Division Topical Meeting Santa Fe NM American Nuclear Society ISBN 8 89448 667 5 ANS Order No 700293 April 14 18 2002 WHI99 M C White R C Little and M B Chadwick Photonuclear Physics in MCNPX X Proceedings of the ANS meeting on Nuclear Applications of Accelerator Technology Long Beach California November 14 18 1999 188 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 WHIO00 M C White User Interface for Photonuclear Physics in MCNP X X5 MCW 00 88 U Los Alamos National Laboratory July 26 2000 and March 21 2001 revised WIL97 W B wilson et al CINDER 90 code for Transmutation Calculations Proceedings of the International Conference on Nuclear Data for Science and Technology Trieste 19 24 May 1997 Italian Physical Society Bologna p 1454 1997 YAR79 Y Yariv and Z Fraenkel Phys Rev C 20 1979 2227 YAR81 Y Yariv and Z Fraenkel Phys Rev C 24 1981 488 YOU98 G Young E D Arthur and M B Chadwick Comprehensive Nuclear Model Calculations Theory and Use of the GNASH Code Proceedings of the IAEA Workshop on Nuclear Reaction Data and Nuclear Reactors Physics Design and Safety Trieste Italy April 15 May 17
469. t in the tally specification To tally within lattice elements of a real world level zero lattice cell use the special syntax that follows Cell 3 contains material 1 and is bounded by four surfaces The F4 card specifies a tally only in lattice element 0 0 0 This syntax is required because brackets can only follow a lt 3 1 1 0 1 2 3 4 lat 1 F4 N 3 lt 3 000 5 7 1 2 3 Universe format The universe format U is a shorthand method of including all cells and lattice elements filled by universe This format can be used in any level of the tally chain The following example illustrates valid shorthand U descriptions in the left column The right column shows the tally after the shorthand has been expanded Cells 4 and 5 are filled with universe 1 shorthand expanded F4 N u 1 45 u 1 4 5 u 1 lt 2 lt 3 45 lt 2 lt 3 118 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 u 1 lt 2 lt 3 45 lt 2 lt 3 1 lt u 1 lt 2 lt 3 1 lt 45 lt 2 lt 3 1 lt u 1 lt 2 lt 3 1 lt 45 lt 2 lt 3 In complex geometries the U format should be used sparingly especially with the multiple bin format If 100 cells are filled by universe 1 and 10 cells are filled by universe 2 then the tally F4 N u 1 lt u 2 will create 1000 output tally bins However F4 N u 1 lt u 2 will create only one output tally bin 5 7 1 2 4 Use of SDn card for repeated structures
470. t be dif ferentiated this option can be used in combination with other options such as with DEBUG and with FC 23 Accelerator Production of Tritium 24 MCNPxX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 Table 3 1 Configure Script Parameters Option Syntax Effect on the generated Makefile if requested Effect on the generated makefile if NOT requested prefix value substitute a full path name for the value placeholder e g home team mcnpx the path given should be different from the working directory where the build is tak ing place value will be used in the install step to create bin and lib data directories for mcnpx s use a default value of usr local is used as the full path name for the install step Executables then go to usr local bin and data files go to usr local lib permissions of the destina tion may prohibit success of installation libdir value substitute a full path name for the value placeholder e g home team mcnpx the path given should be different from the working directory where the build is tak ing place value will be used in the install step to create a library data directory for mcnpx s use a default value of usr local lib is used as the full path name for the install step permis sions of the destination may prohibit success of installa tion This value overrides the library portion of the prefix if bo
471. t by mass number A of the calculated residual masses and the average excitation energy for each mass Only nonelastic interactions are included The option accesses the records on HISTP for all interacting particle types The edit is per formed for both the final residual masses and the residuals after the cascade phase If IOPT is preceded by a minus sign the edit is performed for events initiated by primary source particles only For KOPT 0 the edit is by cell numbers if KOPT 1 the edit is by material numbers If NPARM 0 the edit is over the entire system The parameters NTIM NTYPE and NFPRM are immaterial KPLOT 1 will produce plots of each edit table Tally option 5 or 105 represents the particle weight producing a given nuclide per source particle as such it is a dimensionless quantity The mean excitation is in units of MeV 144 MCNPX User s Manual MCNPX User s Manual i Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium 8 Edit Option IOPT 6 or 106 Energy Deposition Option 6 is not available in this version 9 Edit Option lIOPT 7 Mass and Energy Balance Option 7 is not available in this version 10 Edit Option IOPT 8 or 108 Detailed Residual Mass Edit Option 8 provides a detailed edit of residual masses by Z andN by Z only by N only and by mass number A The option accesses the records on HISTP for all interacting particle types If IOPT is preceded by a minus s
472. t ion reactions The 150 MeV libraries are released with MCNPX version 2 3 0 under the name LA150N the proton libraries under the name LA150H and the photonuclear libraries under LA150U The method for evaluating neutron proton and photon induced cross sections uses a combination of measured cross section data and nuclear model calculations with the GNASH code The work has been described in detail elsewhere CHA99 The NJOY nuclear data processing system MAC94 is used to convert the nuclear data evaluations into a form that can be used by MCNPX New NJOY capabilities e g neutron induced charged particle data incident charged particle libraries and photonuclear libraries have been developed within the context of NJOY99 The full coupling of high energy physics modules and low energy tabular data in MCNPX is still in development The capability to use libraries which may each have different upper energy limits in one problem is referred to as the Mix and Match question In versions 2 3 0 the switch between neutron physics models and neutron tabular data is made at one user specified energy for all materials in the problem Therefore it is recommended that one use a set of libraries which all have upper energy limits above the user specified value The full coupling which can handle the trade off between libraries with different high energy limits and physics modules will be released in MCNPX 3 0 The formal solution of the Mix
473. t of problem surface numbers a subset of the surfaces on the SSW card that created the file WSSA OLD now called RSSA Macrobody surfaces are not allowed Default All surfaces in original run Cy Co Cy like OLD but for cells in which KCODE fission CEL neutrons or photons were written Default All cells in original run Sa1 Sa2 San Sb1 Spo Sbn problem surface numbers upon which the surface source is to start particles in this run The n entries may be repeated to start the surface source in a b transformed locations Default surfaces in the OLD list NEW 104 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 53 Surface Source Read Card Keyword Description m Collision option flag 1 start from the surface source file only those parti cles that came directly from the source without a collision 1 start from the surface source file only those particles that had collisions before crossing the recording surface 0 start particles without regard to collisions default COL WGT x Each particle weight is multiplied by the constant x as it is accepted for transport Default WGT 1 n transformation number Track positions and velocities are transformed from the auxiliary coordinate system the coordinate system of the problem that wrote the surface source file into the coordinate system of the current prob lem using t
474. t one nonconducting component otherwise a conductor gt 0 conductor if at least one conducting component Use Example Required if you want materials in cells M1 NLIB 50D 1001 2 8016 50C 1 6012 1 This material consists of three isotopes Hydrogen 1001 and carbon 6012 are not fully specified and will use the default neutron table that has been defined by the NLIB entry to be 50D the discrete reaction library Oxygen 8016 50C is fully specified and will use the continuous energy library The same default override hierarchy applies to photon and electron specifications 5 4 2 MTm S a p Material 76 Form Default Use Examples MTm X Xo X S a B identifier corresponding to a particular component on the Mm card None Essential for problems with thermal neutron scatter M1 1001 28016 1 light water MT1 LWTR 07 M14 1001 26012 1 polyethylene MT14 POLY 03 M8 6012 1 graphite MT8 GRPH O1 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 5 4 3 MPNm_ Photonuclear Material Form MPNm ZApny ZApnp The MPNm card allows different photonuclear ZAIDs than specified on the Mn card Use Generally needed for photonuclear problems See Phys P card on page 83 Example M23 1001 60c 2 8016 60c 9 8017 60c 1 MPN23 0 8016 8016 0 means produce no photonuclear particles from hydrogen use 8016 for 8016 and use 8016 for 8017 5 4 4 TOTNU Tot
475. t the tally will be divided by a user supplied response function The default is 0 For a discussion see Section 22 below ITMULLT is the TIME MULTIPLIER flag ITMULT gt 0 indicates that the weights tallied will be multiplied by the event time This option applies only when the basic option type is 1 2 4 9 10 or 13 The standard definitions for these input variables may not apply for some options The applicability of the option control parameters is summarized in Fable D According to the parameters specified on the option record the following records are required in the order specified For NERG gt 0 a record defining NERG upper energy bin boundaries from low to high defined as the array ERGB I I 1 NERG The first lower bin boundary is implic itly always 0 0 The definition may be done in four different ways First the energy boundary array may be fully entered as ERGB I 1 NERG Second if two or more but less than NERG elements are given with the record terminated by a slash the array is completed using the spacing between energy boundaries obtained from the last two entries Third if only one entry is given it is used as the first upper energy boundary and as a constant spacing between all the boundaries Fourth if only two entries are given with the first negative and the second positive the second entry is used as the uppermost energy boundary ERGB NERG and the first entry is inter 210 MCNPX User s Manu
476. tal Unix e SGI IRIX 32 and 64 bit e HP HP UX version 10 Sun Solaris e Intel 1386 Linux New hardware operating systems are being added check with the MCNPX team to get the latest status The code distribution contains full source code for the MCNPX 2 3 0 system and test sets for each of the supported architectures The CDROM also contains a recent source distri bution of the GNU make utility needed to properly build the system 3 1 MCNPX Build System 3 1 1 In the Beginning Remember that your PATH environment variable governs the search order for finding util ities You should be aware of the value of your PATH environment variable by issuing the following command echo PATH You may find it useful to set your PATH environment variable to a strategic search order so that the utilities that are found first are the ones you intend to use Setting of environ ment variables is done differently depending upon what shell you use Please consult the appropriate manuals for your shell Most systems have more than one shell Any system can have more than one version of any utility You must know your utilities If you work on a UNIX or Linux operating system you can use the following inquiry com mands to learn if you have more than one make utility MCNPX User s Manual 13 MCNPxX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium which make which gmake Many systems come with a mak
477. tallies When making tallies in repeated structure and lattice geometries often a volume or area is required and MCNP will be unable to calculate it Possibly the geometry causes the calculation to fail A universe can be repeated a different number of times in different cells and the code has no way to determine this There are two distinct options for entries on the SDn card relating to repeated structures and they cannot be mixed within a single tally The first option is to enter a value for each first level entry on the related F card If the entry on the F card is the union of cells the SD card value will be the volume of the union of the cells The following examples illustrate Fn card tally descriptions in the left column The right column shows the SDn card entries F4 N 1 lt 456 lt 78 SD4 V 123 lt 456 lt 78 Va Ne Va 123 lt 45 6 lt 7 8 Vi Vo V3 123 lt 456 lt 78 Vi V volume of cell i and V 53 volume of the union of cells 1 2 and 3 Even though the first line creates six tally bins only one SD value is entered This divisor is applied to all bins generated by the input tally bin You do not need to know the number of bins generated by each input tally bin in order to use the SD card The last line is the union of cells 1 2 and 3 and only one divisor is entered on the SD card The second option is to enter a value for each bin generated by the Fn card F4 N 1 lt 456 lt 78 SD4V v V3 v Vv v 123 lt 45
478. ted as the label for the curve Commands that specify what is to be plotted Use variable x y blank or variables x and y as the inde pendent variable or variables in the plot If only x is spec ified 2D plots are made If both x and y are specified either contour or 3D plots are made depending on whether 3D is in effect See keyword FIXED for the list of FREE xy the symbols that can be used for x and y The default value of xy is E and gives a 2D plot in which the indepen dent variable is energy The FREE command resets XTITLE YTITLE ZTITLE XLIMS YLIMS HIST BAR PLINEAR and SPLINE to their defaults Set n as the bin number for fixed variable q The symbols that can be used for q and the kinds of bins they repre sent are F cell surface or detector D total vs direct or flagged vs unflagged x FIXED qn U user defined S segment M multiplier C cosine E energy T time MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 3 MPLOT amp MCPLOT Commands Command Description SETfdusmcet Define which variables are free and define the bin numbers of the fixed variables SET does the job of the FREE and several FIXED commands in one compact command The value of each parameter can be a bin number the corresponding variable is then a fixed variable or an the corresponding variable is then a free variable If there is onl
479. ted configuration We placed it after the call that checks for the with DEBUG option The first parameter to AC_ARG_WITH is the feature you are looking for in this case OLDXS Next a descriptive string can be placed inside the quote symbols The third parameter is the name of the macro to be executed if with OLDXS is given when the con figure script is called There could be fourth parameter as in the check for the Fortran and C compilers which is the name of the macro to be executed if the option is not given We don t want to do anything if the with OLDXS option is not specified so we don t need to supply the fourth parameter Go to each of the remaining configure in files and place the AC_ARG_WITH call for han dling with OLDXS in the same place as you did in the first configure in file Now we need to define the macro that gets executed when the check for with OLDXS is made We called our macro AC_SET_OLDXS It is important to know that where we check for the presence of the parameter and where we eventually act on the notice of its pres ence could be anywhere in the macros found throughout the aclocal m4 file In this case we would like to have a local variable set indicating that the option is present then later act on that knowledge In aclocal m4 our macro definition of AC_SET_OLDXS uses the special variable with val that was set by the AC_ARG_WITH check for the presence of the option If the option is present
480. ter file to nuclear data tables for neutron proton and photonuclear reactions cross sections for the Bertini model BERTIN gamma emission data for decaying nuclei PHTLIB photon and electron interaction libraries and others Numerous questions in the beta test phase of MCNPX have arisen concerning where these libraries should be kept and this section of the manual has been added for clarification The following set of nuclear data libraries may be used with MCNPX 2 4 0 1 All standard neutron libraries used with MCNP4B DLC189 can be used with MCNPX however they will not contain emission data for charged particles or recoil nuclei these were processed only in the LA150N library Therefore charged second aries and recoil nuclei will not be produced or tracked in MCNPX within the tabular energy ranges 2 MCNP4C DLC200 libraries are the same as the MCNP4B DLC189 set with certain new features These include unresolved resonances delayed neutrons new electron 26 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 libraries ZAIDs end in 03e ENDL92 data and multi temperature U Np tables DLC200 tables may be used with MCNPX with the following cautions None of the DLC200 tables have charged particle or recoil data therefore these will not be produced or tracked in MCNPX Only the DLC200 electron tables with ZAID numbers ending in 03e will work properly in MCNPX 3 Special 15
481. tes 148 MCNPX User s Manual MCNPX User s Manual Ap Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium all collisions KOPT 2 or 3 tabulates elastic scattering only KOPT 4 or 5 tabulates non elastic events only If KOPT is even the edit is by cell number if KOPT is odd the edit is by material number A CINDER removal rate input file will produced for IXOUT gt 0 The default CINDER file name is OPT15A KOPT 1 KOPT 5 KOPT 6 a 6 Figure B 1 Use of the KOPT Parameter for HTAPE3X Option 13 18 Edit Option IOPT 16 or 116 Recoil Energy and Damage Energy Spectra Option 16 provides an edit of the spectra of total recoil energy elastic recoil energy total damage energy and elastic damage energy Also estimated are the mean weight of recoil ing fragments per history mean weight of recoil or damage energy per history and the mean energy per fragment the ratio of the previous two estimates NERG specifies the number of energy bins for the spectra a minus sign on NERG will have the tabulation normed per MeV recommended to produce a true spectrum Input variables NTIM NTYP NFPRM IXOUT IRS IMERGE ITCONV and IRSP are unused KOPT 0 indi cates tally by cell KOPT 1 indicates tally by material NPARM is the number of cells or materials to be read in for the tally If a minus sign flag is used with IOPT IOPT 16 the weights tallied for the spectra will be multiplied by c
482. th FOPT value substitute a quoted or double if omitted the default behav quoted string for value that ior is system dependent There is a separate represents allowable com the detected hardware soft variable that is used piler optimization switch set ware platform and compilers for non optimization tings these settings will determine what the default switches See with override the system default FOPT should be FFLAGS in this table or system computed values If in doubt run the configure script and examine the system default or system computed values that appear in the gener ated Makefile h You may want to include the defaults in the string you specify for FOPT with this mech anism FOPT settings are always appended to FFLAGS settings when configure is run again MCNPX User s Manual MCNPxX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 3 1 Configure Script Parameters Option Syntax Effect on the generated Effect on the generated p y Makefile if requested makefile if NOT requested with COPT value substitute a quoted or double if omitted the default behav quoted string for value that ior is system dependent There is a separate represents allowable com the detected hardware soft variable that is used piler optimization switch set ware platform and compilers for non optimization tings these settings will determine what the default switches See with override the sy
483. th are given with OLDXS the symbol OLDM is defined that is passed as DOLDM to the compile step of mcnpx in order to activate the old cross section capabilities nothing is done new cross section capabilities are used with no_paw or with no_paw yes this means that the symbol NO_PAW will be defined for compilation and actions are taken in the source to omit PAW capabilities when com piling if omitted the default behav ior is system dependent if the detected hardware soft ware platform can handle PAW it is included MCNPX User s Manual Accelerator Production of Tritium MCNPX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 Table 3 1 Configure Script Parameters Option Syntax Effect on the generated Makefile if requested Effect on the generated makefile if NOT requested with FFLAGS value There is a separate variable that is used for optimization switches See with FOPT in this table If in doubt run the configure script and examine the system default or system computed values that appear in the generated Make file h You may want to include the defaults in the string you specify for FFLAGS with this mechanism when configure is run again substitute a quoted or double quoted string for value that represents allowable compiler switch settings these set tings will override the system default or system computed values if omitted
484. that appears on the SIk card One user bin is created for each bin of source distribution k plus a total bin The scores for tally n are then binned according to which bin of source distribution k the source particle came from The score of the total bin is the score you would see for tally n without the special treatment if source distribution k is not a dependent distribution CAUTION For a dependent distribution the score in the total bin is the subtotal portion of the score from dependent distribution k SCD No parameters follow the keyword but an FUn card is required Its bins are a list of source distribution numbers from SIk cards The scores for tally n are then binned according to which distribution listed on the FUn card was sampled This feature might be used to identify which of several source nuclides emitted the source particle In this case the source distributions listed on the FUn card would presumably be energy distributions Each energy distribution is the correct energy distribution for some nuclide known to the user and the probability of that distribution being sampled from is proportional to the activity of that nuclide in the source The user might want to include an FCn card that tells to what nuclide each energy distribution number corresponds CAUTION If more than one of the source distributions listed on the FU card is used for a given history only the first one used will score PTT No parameters follow the keyword
485. the HTAPE3X code for backward compatibility with the LAHET Code System section 8 5 The new visual tallies Mesh and Radiography tallies are provided with an interpretation program gridconv sections 8 1 2 and 8 2 4 This stand alone program converts the out put of the tallies into a format consistent with several currently available graphics packages In MCNPX 2 3 0 gridconv will also convert the results of any tally contained in a MCTAL file This capability is described in the general gridconv discussion of section 8 1 2 Parallel processing is not yet implemented in MCNPX this is a major development which will be integrated into new data structures to be added in MCNPX version 3 0 We fully realize that applications in high energy regimes are computationally intensive and it has been long established practice to run Monte Carlo codes on many machines adding the final results together Notes for the user on this practice are given in section 8 6 8 1 The Mesh Tally The technique which has become known as the Mesh Tally has become very widely used in many applications The development of this method grew out of research with codes such as LCS GEANT FLUKA CALOR and MARS at the Superconducting Super Collider in 1993 Some form of this method is currently in standard use in most high energy Monte Carlo codes The Mesh Tally is a method of graphically displaying particle flux dose or other quantities on a rectangular cylind
486. the center of the cylinder on which the grid is established RO Always 0 zero in this application as in the pinhole case The reference coordinates that establish the reference direction cosines for the X2 Y2 22 outward normal to the detector grid plane as from X2 Y2 Z2 to X1 Y1 Z1 This is used as the outward normal to the detector grid plane for the TIR case and as the centerline of the cylinder for the TIC case 138 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 78 Transmitted Image Projection Argument Description Continued Argument Description e F1 0 Both the source and scattered contributions will be scored at the grid F1 lt 0 Only the scatter contributions will be scored F150 is not allowed in this application F1 plane grid case Radial restriction relative to the center of the grid for contribu tions to be made It defines a radial field of view on the grid cylindrical case Radius of the cylinder on which the grid is to be established F2 F3 0 All contributions are directed to the center of each grid bin F3 F3 lt 0 Contributions are made with a random offset from the center of the grid bin This offset remains fixed and is used as the offset for contributions to all of the grid bins for this event The grid itself is established with the use of FSn and Cn cards in the same manner as described for the pi
487. the data path itself can be customized like this usr local src mcnpx_2 3 0 configure libdir usr mcnpx which will leave the default executable location as usr local bin and set the location for the data files to usr mcnpx Finally both the prefix and the libdir options can be used together with the libdir options taking precedence over the library directory implied by the prefix These options should remove the need to edit paths in the source code In fact with sup port for these options there are no longer any paths in the code to edit 3 1 3 3 Individual Private Installation For the purpose of the second illustration we will look at a single non privileged user Me on a computer loading and building a private copy of the code The local user build ing the private copy is username me whose home directory is the directory home me The user has fetched the distribution from CDROM or from the net and has it in the file home me mcnpx_2 3 0 tar gz The user will unload the distribution package into home me menpx_ 2 3 0 The user will build the system in the same directory as the source install the binary executable in home me bin and install the binary data files and eventually the mcnp cross sections in home me lib This method makes it hard to make multiple ver sions with different options A better example will follow this one The following example uses bourne shell commands that follow accomplish this task
488. the default behav ior is system dependent the detected hardware software platform and compilers deter mine what the default FFLAGS should be with CFLAGS value There is a separate variable that is used for optimization switches See with COPT in this table If in doubt run the configure script and examine the system default or system computed values that appear in the generated Make file h You may want to include the defaults in the string you specify for CFLAGS with this mechanism when configure is run again substitute a quoted or double quoted string for value that represents allowable compiler switch settings these set tings will override the system default or system computed values if omitted the default behav ior is system dependent the detected hardware software platform and compilers deter mine what the default CFLAGS should be with FOPT value There is a separate variable that is used for non optimization switches See with FFLAGS in this table If in doubt run the con figure script and examine the system default or system com puted values that appear in the generated Makefile h You may want to include the defaults in the string you specify for FOPT with this mechanism FOPT settings are always appended to FFLAGS settings when configure is run again substitute a quoted or double quoted string for value that represents allowable compiler optimization switch settings
489. the problem EMAX is set to 100 MeV for all particles A third argument has been added to the PHYS n and PHYS h cards to accommodate the extended 150 MeV neutron and proton libraries Set the CUT_N or CUT_H value to the maximum energy to which table based data will be used for neutrons MCNPX version 2 1 5 and later and for protons MCNPX version 2 3 0 The CUT parameter must be used with caution MCNPX 2 3 0 cannot yet combine libraries with different upper energy limits however it is not a fatal error to call for a combination of such libraries Several examples can illustrate the potential problem 20 and 150 MeV libraries are our most commonly available tables however the user should be aware that other upper limits might be present e if CUT is set to 20 0 and all libraries have upper energies of 20 0 then libraries will be used to 20 MeV and physics models above that energy e if CUT is set to 20 0 and all libraries have upper energies of 150 0 then libraries will be used to 20 MeV and physics models above that energy e if CUT is set to 150 0 and any library has an upper energy of 20 0 then the code will use the cross section values found at 20 MeV in that library from 20 to 150 MeV No MCNPX User s Manual 73 MCNPxX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium attempt at extrapolation of the 20 MeV value to a value at 150 MeV will be made since there is currently no m
490. the problem If there are entries it turns off the bin print for the tally numbers that are listed If after the run is completed one would like to see these numbers the printing MCNPX User s Manual 139 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 of the bin values can be restored with the TALNP card in an INP file used in a continue run The tally numbers are entered on the TALNP card as negative numbers 5 7 20 4 Reading the Radiography Tally Output The output of the two radiography tally options is contained in the mctal file It can be formatted for use with external graphics programs with the gridconv routine The user is referred to Section 5 7 22 7 for information on how to use gridconv 5 7 21 PERTn Perturbation Form PERTn pl keyword parameter s keyword parameter s Implement the 2nd order differential operator perturbation method Table 5 79 Variable Description n unique arbitrary perturbation number pl N P or N P Not available for other particles keyword See Table Default Some keywords are required See Table Use Optional Table 5 80 PERT Keywords Parameter Values and Defaults Keyword Parameter Values Default Entries BASIC KEYWORDS CELL Integer gt 0 Required Unlimited MAT Integer gt 0 1 RHO Real integer 1 ADVANCED KEYWORDS METHOD 1 2 3 1 1 ERG Real Integer gt 0 All Energies 2 RXN Reaction number 1 Un
491. the quantity of interest depends only on neutrons and one starts with a proton beam there is no need to transport any particles other than protons neutrons and charged pions as neutron production by other particles is negligible compared to production by these three particle types Use of the various LAHET physics model options such as the ISABEL and CEM INC modules within MCNPX is encouraged this provides the user with the ability to test the sensitivity of the quantity of interest to the different physics models If significant differences are observed the user should evaluate which physics model is most appropriate for his or her particular application For example total neutron production from actinide targets is known to be more accurate if the multi step preequilibrium model MPM is turned off which is not its default setting 1 All particles should be included for energy deposition calculations as discussed in Section 8 3 MCNPX User s Manual 203 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 204 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 8 Appendix B HTAPESX for use with MCNPX This appendix is reprinted from HTAPE3X for Use with MCNPX Richard E Prael Los Alamos Report LA UR 99 1992 April 16 1999 Abstract HTAPE3xX is a code for processing medium and high energy collision data written to a history file by MCNPX In addition it
492. the three coordi nates as defined by the mesh type rectangular cylindrical or spherical prior to any trans transformation In the case of rectangular meshes CORAn represent planes perpendicular to the x axis CORBn are planes perpendicular to the y axis and CORCn are planes perpendicular to the z axis Bins do not have to be equally spaced In the case of the cylindrical mesh the middle coordinate CORBn is the untransformed z axis which is the symmetry axis of the cylinder with radial meshes defined in the CORAn input line The first smallest radius may be equal to zero The values following CORBn define planes perpendicular to the untransformed z axis The values following CORCn are positive angles relative to a counter clockwise rotation about the untrans formed z axis These angles in degrees are measured from the positive x axis and must have at least one entry of 360 which is also required to be the last entry The lower limit of zero degrees is implicit and never appears on the CORCn card In the case of spherical meshes scoring will happen within a spherical volume and can also be further defined to fall within a conical section defined by a polar angle relative to the z axis and azimuthal angle CORAn is the radius of the sohere CORBn is the polar angle and CORCn is the same as in the cylindrical case It is helpful in setting up spherical problems to think of the longitude latitude coordinates on a globe The original
493. tic cross section tables above 150 MeV and to improve the physics involved with the intermediate and high energy physics models through the CEM program Currently the requirements of the Accelerator Transmutation of Waste program which is part of AAA are directed toward improvements in fission physics and actinide data Responsibility for the development of MCNPX was given to the APT Target Blanket and Materials Engineering Development and Demonstration ED amp D project A code develop ment team under the leadership of Dr H Grady Hughes was formed Because the Los Alamos accelerator community has long supported the work of Dr Richard Prael in the development of the LAHET Code System it was decided to build on this base by com bining the capabilities of LAHET and MCNP into one code This was accomplished by extending the capabilities of MCNP4B to all particles and all energies and including the use of physics models in the code to compute interaction probabilities where table based data are not available In the present version MCNPX 2 4 0 the code has also incorpo rated all features of MCNP4C3 Additional development has been provided by the theoretical efforts of the T 16 group at Los Alamos particularly in the areas of nuclear data evaluation and expansion of physics based models A program of experimental activities was also undertaken including mea surement of various cross sections and development of more complex benchmark
494. ticle type and number 1 gt no lines N 0 gt MESH off 1 gt WW MESH WWMESH appears only if WWINP file is read in 5 2 TALLIES amp CROSS SECTIONS 5 2 1 Input for MCPLOT and Execution Line Options To run only MCPLOT and plot tallies after termination of MCNPX enter the following command menp z options where z invokes MCPLOT Options is explained in the next paragraph Cross section data cannot be plotted by this method The execute line command mcnpx inp filename ixrz options causes MCNPX to run the problem specified in filename and then the prompt mcplot gt appears for MCPLOT commands Both cross section data and tallies can be plotted Cross section data cannot be plotted after a TTY interrupt or by use of the MPLOT card The execute line command menpx inp filename ixz options is the most common way to plot cross section data The problem cross sections are read in but no transport occurs The following commands cannot be used 3D BAR CONTOUR DUMP FREQ HIST PLOT RETURN RMCTAL RUNTPE SPLINE VIEW and WMCTAL MCNPX User s Manual 47 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 The following options can be entered on the execution line Table 5 2 MPLOT Execution Line Options Keyword Description Suppress plotting at the terminal and send all plots to the graphics metafile PLOTM NOTEK is for production and NOTER batch situations and for when the us
495. till interactive but now shows all tallies in the problem from which any may be selected The user has the option of generating one or two dimensional output The user is then told about the bin structure so the one or two free variables may be selected The energy is the default independent variable in the one dimensional case There is no default for the two dimensional case The order in which the two dimensional bin variables are selected does not make any difference to the output in that the order of the processing will be as it appears on the mctal file Gridconv will work with mctal files produced both by MCNPX and MCNP 5 8 VARIANCE REDUCTION IMP WWG WWGE WWP WWN WWE MESH EXT VECT FCL DDn PDn DXT DXC BBREM SPABI ESPLT PWT 5 8 1 IMP Cell Importance Form IMP n X4 Xo Xi XI Table 5 86 Cell Importance Card Descriptor Description n any particle symbol or IPT number from Table Xi importance for cell i number of cells in the problem MCNPX User s Manual 153 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Default The default importance for all particles listed on the MODE card is unity If a cell importance is set to zero for any particle all importances for that cell will be set to zero unless specified otherwise Use An IMP n card is required with an entry for every cell unless a WWN weight window bound card is used Example IMP N12 2M0 120R The neutron importan
496. tion Vx Vy Vz X y Z coordinates of cone bottom Hx Hy Hz cone axis height vector R1 radius of lower cone base R2 radius of upper cone base Example TRC 500 1000 4 2 a 10 cm high truncated cone abut the x axis with the center of the 4 cm radius base at x y z 5 0 0 and with the 2 cm radius top at x y z 5 0 0 5 3 2 4 8 ELL Ellipsoid Form ELL VixViyV1z V2xV2yV2z Rm Table 5 16 Macrobody Ellipsoid Argument Description if Rm gt 0 1st foci coordinate Vix Vy V1z if Rm lt 0 center of ellipsoid if Rm gt 0 2nd foci coordinate ee VAZ if Rm lt 0 major axis vector length major radius if Rm gt 0 length of major axis If Rm gt 0 a if Rm lt 0 minor radius length Example ELL 00 2 002 6 66 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 ELL 000 003 2 an ellipsoid at the origin with major axis of length 6 in the z direction and minor axis radius of length 4 normal to the z axis 5 3 2 4 9 WED Wedge NOTE A right angel wedge has a right triangle for a base defined by V1 and V2 anda height V3 The vectors V1 V2 and V3 are orthogonal to each other Form WED VxVyVz V1xViyV1z V2xV2yV2z V3x V3y V3z Table 5 17 Macrobody Wedge Argument Description VxVyVz X y Z coordinates of wedge vertex Vix V1y V1z vector of 1st side of triangular base V2x V2y V2z vec
497. tion has been found to have negligible effect on the results 70 MCNPX User s Manual MCNPX User s Manual Ap Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium 6 MCNPX Input Files Standard MCNP input cards are all accepted in MCNPX however additional card options are now available to take advantage of the multiparticle capabilities Modifications to stan dard MCNP inputs are described in Section 6 1 Section 6 2 describes new cards added to control the Bertini ISABEL and CEM physics options which are used when table based data are not available Use of new high energy proton and photonuclear data library capabilities has already been described in Section 4 3 Accelerator simulation applications have a need for specialized source input to describe an incident particle beam Usually this takes the form of a directed beam of particles monoenergetic with a transverse gaussian profile To facilitate this a new source option has been added to MCNPX and is described in Section 6 3 6 1 MCNP Card Modifications and Additions 6 1 1 Problem Type Card MODE The MODE card can now take any argument listed in the Symbol column of Table 5 1 in any order It must list all particles that will be transported If a particle is designated the anti particle will also be transported For example MODE nh e will transport neutrons and anti neutrons protons and anti protons u and uw electrons and posi
498. tion is for use in computing double differ ential particle production cross sections with the XSEX code See Appendix C ICEM 0 Use the Bertini or ISABEL model determined by the IEXISA parameter default 1 Use the CEM model 5 5 7 2 LCB Form LCB FLENB1 FLENB2 FLENB3 FLENB4 FLENBS FLENB6 CTOFE FLIMO LCB controls which physics module is used for particle interactions depending on the kinetic energy of the particle Table 5 42 LCB Keyword Descriptions Keyword Description FLENB1 Kinetic Energy Default 3500 MeV For nucleons the Bertini INC model will be used below this value FLENB2 Kinetic Energy Default 3500 MeV For nucleons the FLUKA high energy generator will be used above this value Note The probability for selecting the interaction model is interpolated linearly between FLENB1 and FLBEN2 Note The version of FLUKA used in MCNPX should not be used below 500 MeV c momentum Note For nucleons the Bertini model switches to a scaling procedure above 3 495 GeV wherein results are scaled from an interaction at 3 495 GeV Although both models will execute to arbitrarily high energies a plausible upper limit for the Bertini scaling law is 10 GeV MCNPX User s Manual 93 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 42 LCB Keyword Descriptions Continued Keyword Description FLENB3 Kinetic Energy Default 2500 M
499. tion of 2 1 3 6 bins as follows MCNPX User s Manual 117 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 S4 lt Cy Coll Lal lt C3 Ss lt Cy Coll IA lt C3 S4 lt Cy Coll 1A lt Ca Sp lt Cy Co U IA lt Ca S4 lt C1 Col 1A lt Cs Ss lt C1 Co U IA lt Cs The repeated structure lattice input tally bin format with levels that have multiple entries automatically creates multiple output tally bins The total number of bins generated is the product of the number of entries at each level If parentheses enclose all entries at a level the number of entries at that level is one and results in the union of all those entries For unnormalized tallies type 1 8 the union is a sum For normalized tallies type 2 4 6 7 the union is an average A symbol T on the tally line creates an additional tally bin that is the union or total of all the other tally bins 5 7 1 2 2 Brackets Brackets enclose index data for lattice cell elements Brackets make it possible to tally on a cell or surface only when it is within the specified lattice elements The brackets must immediately follow a filled lattice cell Listing a lattice cell without brackets will produce a tally when the tally cell or surface is in any element of the lattice provided the tally cell or surface fills an entry at all other levels in the chain The use of brackets is limited to levels after the first l
500. tions LA 10248 MS Los Alamos National Laboratory 1985 KOC59 H W Koch and J W Motz Bremsstrahlung Cross Section Formulas and Related Data Rev Mod Phys 31 1959 920 LAN44 L Landau On the Energy Loss of Fast Particles by lonization J Phys USSR 8 1944 201 MAC94 R E MacFarlane and D W Muir The NJOY Nuclear Data Processing System version 91 Los Alamos National Laboratory Report LA 12740 M October 1994 MAD88 D G Madland Recent Results in the Development of a Global Medium Energy Nucleon Nucleus Optical Model Potential in Proceedings of a Specialist s Meet ing on Preequilibrium Reactions Semmering Austria February 10 12 1988 Edited by B Strohmaier OECD lt Paris 1988 p 103 116 MAS74 S G Mashnik and V D Toneev MODEX the Program for Calculation of the Energy Spectra of Particles Emitted in the Reactions of Pre Equilibrium and Equilibrium Statistical Decays in Russian Communication JINR P4 8417 Dubna 1974 25 pp MAS98 S G Mashnik A J Sierk O Bersillon and T A Gabriel Cascade Exciton Model Detailed Analysis of Proton Spallation at Energies from 10 MeV to 5 GeV Nucl Instr Meth A414 1998 68 Los Alamos National Laboratory Report LA UR 97 2905 http t2 lanl gov publications publications html MOH83 H J Moehring Hadron nucleus Inelastic Cross sections for Use in Hadron cascade Calculations at high Energies CERN report TIS RP 116 Octob
501. tis a dimensionless quantity 4 Edit Option IOPT 2 or 102 Surface Flux The surface flux tally is analogous to an MCNP F2 tally All particle types listed above may be specified The number of energy bins is given by NERG The number of particle types for which surface flux data is to be tallied is given by NTYPE and must be gt 0 NFPRM is unused If KOPT 1 surface segmenting is performed as in option above the same input record to designate the segmenting planes or cylinders must be included as in option 1 If IOPT is preceded by a minus sign the particle weight is multiplied by its energy before tallying MCNPX User s Manual 213 MCNPxX User s Manual Version 2 4 0 September 2002 LA CP 02 408 The surface flux tally represents the time integrated flux integrated over surface areas Unless otherwise modified it is a dimensionless quantity 5 Edit Option IOPT 3 or 103 Particle Production Spectra Option 3 may be used to tally the spectra of particles produced in nuclear interactions It accesses all collision records on HISTP for all particles causing collisions If IOPT is preceded by a minus sign the edit is performed only for events initiated by the primary source particles For KOPT 0 or 1 separate edits are performed for cascade and evaporation phase production In addition total nucleon production from either phase is edited For KOPT 2 or 3 only the cascade production is edited For KOPT 4 or 5 only
502. tor of 2nd side of triangular base V3x V3y V3z height vector Example WED 00 6 400 030 0012 a 12 cm high wedge with vertex at x y z 0 0 6 The triangular base and top are a right triangle with sides of length 4 in the x direction and 3 in the y direction and hypotenuse of length 5 5 3 2 4 10 ARB Arbitrary Polyhedron Form ARB axaaz bxbybz hxhyhz N1 N2 N3 N4 N5 N6 Table 5 18 Macrobody Arbitrary Polyhedron Argument Description X y z coordinates of 1st corner of polyhedron There must ax ay az be eight x y z triplets to describe the eight corners of the polyhedron MCNPX User s Manual 67 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 18 Macrobody Arbitrary Polyhedron Argument Description four digit number describing a side of the polyhedron in N1 N6 terms of it s corresponding two corners e g N1 1278 is a plane side bounded by corners 1 2 7 amp 8 a b g h NOTE Thirty entries are required to complete the argument of the card For polyhedrons of fewer than six sides zero entries must be supplied Example ARB 5 10 5 5 15 5 10 5 5 105 0120 000 000 000 1234 1250 1350 2450 3450 0 a 5 sided polyhedron with corners at x y z 5 10 5 5 10 5 5 10 5 5 10 5 0 12 0 and planar facets constructed from corners 1234 etc note the zero entry for the 6th facet 5 3 3 Geometry Data 5 3 3 1 VOL Cell Volume Form
503. trons 6 1 2 Geometry Cards VOL AREA U TRCL LAT FILL TR No modifications have been made to any cell or surface card 6 1 3 Variance Reduction Cards IMP ESPLT PWT EXT VECT FCL WWE WWN WWP WWG WWGE PDn DXC BBREM Any card with a particle designator can accept any particle symbol from Table 5 1 A new type of biasing Secondary particle biasing has been added and is described in Section 7 1 Note Detector variance reduction techniques will not work outside library energy limits Detector variance reduction techniques will also not work for charged particles MCNPX User s Manual 71 MCNPX User s Manual Ap Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium 6 1 4 Source Specification Cards SDEF Sin SPn SBn DSn SCn SSW SSR KCODE KSRC All source capabilities of MCNP are intact and additional features have been added to describe typical accelerator beams see Section 6 3 The argument PAR can be set to any IPT value in Table 5 1 Since particles and antipar ticles have the same IPT values an antiparticle source is designated with a minus sign For example PAR 9 will generate antiprotons in an SDEF card Note In MCNPX version 2 3 0 one cannot use positrons as a source PAR 3 This will be implemented in a near term future version As in MCNP only one source particle can be designated at any one time When PAR is absent the source particle generated depends on the arguments of the MODE ca
504. ts will be produced for each edit table Parameters NERG NTYPE and NFPRM are unused If IXOUT 1 an auxiliary output file appropriate for input to the CINDER program will be written the default file name is OPT8A Unless otherwise modified tally units are dimensionless weight of a residual nuclide per source particle An additional tabulation is produced which shows the estimated metastable state production as a fraction of the total isotopic production As illustrated in the example here a State is identified by its excitation energy and half life the estimated fraction of total isotope production associated with the particular metastable state is shown with the estimated relative standard deviation Table 8 1 z a elev t half fraction 47 110 0 11770 2 17730D 07 4 00000D 01 0 3465 MCNPX User s Manual 215 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 8 1 z a elev t half fraction 47 111 0 05990 6 50000D 01 8 00000D 01 0 2001 47 116 0 08100 1 05000D 01 S 00000D 01 0 5001 48 113 0 26370 4 41500D 08 2 85714D 01 0 3195 48 115 0 17340 3 87070D 06 5 00000D 01 0 3536 48 117 0 13000 1 22400D 04 2 50000D 01 0 4331 48 119 0 14640 1 62000D 02 6 00000D 01 0 2329 11 Edit Option IOPT 9 or 109 Surface Current with Collimating Window Option 9 is identical to option 1 except that a rectangular or circular window is impose
505. ty of case statements that depend on the value of the TFC and TCC variables in combination with ARCH and SYS TEM Some of these case statements are for compiler flag settings some of these case statements are for linking the output of the compiler into executables static and dynamic linking These flag and option setting vary by compiler vendor and hardware platform We must check each case statement to see if we need to add flags or options to the compile or link steps Make sure the pgf77 or pg and gcc compilers appear as case labels in these case statements and are set to your desired values If you add a new case label new statements to an existing case label or change the value of any setting you must regenerate the configure scripts at all the different levels of the file tree hierarchy by executing the following command from within the menpx_2 3 0 directory force regeneration of configure scripts at all levels autoreconf localdir config f Once the configure scripts at the various levels have been generated you can execute configure with the desired options that were added For our example we would execute the following to get our new pgf77 compiler when we make mcnpx from the top level of your build directory configure and request that pgf77 be used to compile Fortran usr local src mcnpx_2 3 0 configure with FC pgf77 The configure will recursively descend the necessary tree hierarchy and generate Make
506. type 1 Keyword Descriptions Continued Keyword Description Can have from one to four numerical entries following it e The value of the first entry is in reference to an energy dependent response function given on a MSHMFn card no default e The second entry is 1 default 1 for linear interpolation and 2 for logarith mic interpolation Ifthe third entry is zero default 0 the response is a function of energy mfact deposited otherwise the response is a function of the current particle energy e The fourth entry is a constant multiplier and is the only floating point entry allowed default 1 0 If any of the last three entries are used the entries preceding it must be present so that the order of the entries is preserved Only one mfact keyword may be used per tally Must be followed by a single reference to a TR card that can be used to trans trans late and or rotate the entire mesh Only one TR card is permitted with a mesh card 5 7 22 3 Source Mesh Tally Type 2 The second type of Mesh Tally scores source point data in which the weight of the source particles P 1 P 2 P 3 are scored in mesh arrays 1 2 3 therefore a separate mesh tally grid will be produced for each particle chosen Currently it is not possible to choose more than one particle type in a type 2 Mesh Tally However some graphics programs will enable the user to add separate histograms together offline The u
507. uary 1980 BAR73 V S Barashenkov A S lljinov N M Sobolevskii and V D Toneev Interaction of Particles and Nuclei of High and Ultrahighy Energy with Nuclei Usp Fiz Nauk 109 1973 91 Sov Phys Usp 16 1973 31 BAR81 J Barish T A Gabriel F S Alsmiller and R G Alsmiller Jr HETFIS High Energy Nucleon Meson Transport Code with Fission Oak Ridge National Laboratory Report ORNL TM 7882 July 1981 BAR94 V S Barashenkov A Polanski Electronic Guide for Nuclear Cross Sections Comm JINR E2 94 417 Dubna 1994 BER63 M J Berger Monte Carlo Calculation of Penetration and Diffusion of Fast Charged Particles in Methods in Computational Physics Vol 1 edited by B Alder S Fernbach and M Rotenberg Academic Press New York 1963 p 135 BER70 M J Berger and S M Seltzer Bremsstrahlung and Photoneutrons from Thick Target and Tantalum Targets Phys Rev C2 1970 621 BER63a H W Bertini Phys Rev 131 1963 1801 BER69 H W Bertini Phys Rev 188 1969 1711 MCNPX User s Manual 181 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 BET34 H A Bethe and W Heitler On Stopping of Fast Particles and on the Creation of Positive Electrons Proc Roy Soc London A146 1934 p 83 BEV69 Phillip R Bevington Data Reduction and Error Analysis for the Physical Sciences McGraw Hill Book Company 1221 Avenue of the Americas New York NY 1002
508. ue to particle interactions in the upstream accelerator If needed such fine detail must be specified with standard MCNP source specification methodology An additional feature has been added through the specification of a general transformation on the SDEF card in one of two forms TR n or TR Dn In either case a general trans formation is applied to a source particle after its coordinates and direction cosines have been determined using the other parameters on the SDEF card Particle coordinates are modified by both rotation and translation direction cosines are modified by rotation only This allows the user to rotate the direction of the beam or move the entire beam of particles in space The TR Dn card is particularly powerful since it allows the specification of more than one beam at a time An example of specifying a Gaussian beam is given below and may be modified at the user s need Title c Cell cards ccc 0 nnn cookie cutter cell c Surface Cards nnn SQ al b 0000 c2000 cookie cutter surface c Control Cards SDEF DIR 1 VEC 001 X D1 Y D2 Z 0 CCC ccc TR n SP1 4 fk 0 SP2 41 fy 0 TRn Xo YoZo cosp sinb 0 sinp cos 0 001 The SDEF card sets up an initial beam of particles travelling along the Z axis DIR 1 VEC 0 0 1 Information on the x and y coordinates of particle position is detailed in the two SP cards X D1 Y D2 indicating that the code must look for distributions 1 and 2 84 MCNPX User s Manual
509. ular of the super imposed mesh AXS vector giving the direction of the axis of the cylindrical mesh 0 0 1 VEC vector defining along with AXS the plane for O 0 1 0 0 locations of the coarse meshes in the x direction for rectangular geome 1 course mesh IMESH Bee Rea ny try or in the r direction for cylindrical geometry per direction number of fine meshes within corresponding coarse meshes in the x 1 in each coarse IINTS direction for rectangular geometry or in the r direction for cylindrical mesh geometry locations of the coarse meshes in the y direction for rectangular geome 1 course mesh JMESH aay oes ey try or in the z direction for cylindrical geometry per direction number of fine meshes within corresponding coarse meshes in the y 1 in each coarse JINTS direction for rectangular geometry or in the z direction for cylindrical mesh geometry locations of the coarse meshes in the z direction for rectangular geome 1 course mesh try or in the O direction for cylindrical geometry per direction number of fine meshes within corresponding coarse meshes in the z 1 in each coarse KINTS direction for rectangular geometry or in the 9 direction for cylindrical mesh geometry Note in th xyz rec mesh the IMESH JMESH and KMESH are the actual 158 x y z coordinates In the RZT CYL mesh IMESH radius and JMESH height are relative to ORIGIN and KMESH theta is relative to VEC MCNPX User s Manual MCNPX User s Manual V
510. und Vielfachstreuung Z Naturforsch 3a 1948 78 MOT29 N F Mott The Scattering of Fast Electrons by Atomic Nuclei Proc Roy Soc London A125 1929 425 PRA88 R E Prael and M Bozoian Adaptation of the Multistage Pre equilibrium Model for the Monte Carlo Method I Los Alamos National Laboratory Report LA UR 88 3238 September 1998 186 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 PRA89 R E Prael and H Lichtenstein User Guide to LCS The LAHET Code System Los Alamos National Laboratory Report LA UR 89 3014 Revised September 15 1989 http www xdiv lanl gov XCI PROJECTS LCS lahet doc html PRA94 R E Prael A Review of Physics Models in the LAHETTM Code LA UR 94 1817 Los Alamos National Laboratory PRAQ5 R E Prael and D G Madland LAHET Code System Modifications for LAHET 2 8 Los Alamos National Laboratory Report LA UR 95 3605 September 1995 PRA96 R E Prael D G Madland A Nucleon Nucleus Elastic Scattering Model for LAHET in Proceedings of the 1996 Topical Meeting on Radiation Protection and Shielding April 21 25 1996 No Falmouth Mass American Nuclear Society 1996 pp 251 257 PRA98a R E Prael A Ferrari R K Tripathi A Polanski comparison of Nucleon Cross Section Parameterization Methods for Medium and High Energies in Proceedings of the Fourth Workshop on Simulating Accelerator Radiation Environments SARE
511. upper boundary equal to 1 0 the lower MCNPX User s Manual 229 MCNPxX User s Manual Version 2 4 0 September 2002 LA CP 02 408 limit of the first bin is always 1 0 If a null record is present only a then the range 1 1 is divided into NANG equal intervals For NANG lt 0 a record is required to define the BAR NANG BAR lower degree bin boundaries They should be entered from low to high with the last lower boundary equal to 0 0 the upper limit of the first bin is always 180 degrees If a null record is present only a then the range 180 0 is divided into BAR NANG BAR equal intervals 4 Executing XSEX3 An input file and a history file are the only required input files The default file name for the input is INXS the default file name for the output is OUTXS and the default file name for the history file is HISTP A value of KPLOT NE 0 will result in the creation of a MCTAL format plot file with default name XSTAL These file names may be changed by file replacement The most general execute line has the format XSEX3 INXS OUTXS HISTP XSTAL 5 Plotting Output from XSEX3 The source code for XSEX3 contains a plotting package using the LANL Common Graphics System the latter is not generally available outside of Los Alamos National Laboratory A new feature has been added for this release whereby a nonzero value for the input quantity KPLOT will cause the writing of a file XSTAL in t
512. utables and libraries in usr local make install clean up The build products are no longer needed cd tmp rm rf mcnpx 3 1 3 2 System Wide Installation With Existing Directories The previous example might typically be used when a new installation of MCNPX is per formed on a system that has no pre existing mcnpx with which to be compatible If a user already has mcnpx then it may be desired to use the existing locations for the data files and cross sections Two options to the configure process can be used to customize the locations where mcnpx and its data will be installed and the default locations where MCNPX will find those files When the user wants to use the normal mcnpx directory layout of MCNPX User s Manual 17 MCNPX User s Manual Version 2 3 0 April 2002 E LA UR 02 2607 Accelerator Production of Tritium bin for executables and lib for data files but does not wish to use the default directory usr local then the previous example can be adjusted with additional options In the previous example the configure script could be given the option usr local src mcnpx_2 3 0 configure prefix usr mcnpx and the make install process would install the mcnpx binary in usr mcenpx bin and the data files in usr mcenpx lib The code will use usr mcnpx lib as its default location for find ing the data files When the user has an existing directory layout that does not follow the mcnpx default then
513. utrons is scored by default Must be followed by a single reference to a TR card that can be used to trans trans late and or rotate the entire mesh Only one TR card is permitted with a mesh card 5 7 22 4 Energy Deposition Mesh Tally Type 3 The third type of Mesh Tally scores energy deposition data in which the energy deposited per unit volume from all particles is included This can be due to the slowing of a charged particle the recoil of a nuclei energy deposited locally for particles born but not tracked etc The results are similar to the scoring of an F6 np tally as described in Section 8 3 Note that in MCNPX the option to track energy deposition from one type of particle alone in a problem is included in the first Mesh Tally type see Table 5 81 Keyword pedep The Energy Deposition Mesh Tally described here will give results for all particles tracked in the problem and has no option to specify a particular particle Note since the mesh is independent of problem geometry a mesh cell may cover regions of several different masses Therefore the normalization of the output is per mesh cell volume MeV cm source particle not per unit mass R C S MESHn total de dx recol tlest delct mfact nterg trans n 3 13 23 33 148 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 83 Energy Deposition Mesh Tally type 3 Keyword Descriptions Keyword Descri
514. ward compatible if the name of the card only is changed e A cylindrical mesh has been added for the transmitted image option e Section 8 3 on Energy Deposition has been extensively rewritten to clarify normaliza tion and to discuss handling of local energy deposition e Nontracking Change F6 no longer needs the n p designator e Option ic 40 ICRP 74 1996 ambient dose equivalent has been added for neutrons in table 8 9 DFACT e Section 8 5 adds comments on the use of the histp card Appendices e Added the base case input deck to Appendix A The table in Appendix B was incorporated into table 5 1 Appendix B is now the HTAPESxX discussion 10 MCNPX User s Manual Accelerator Production of Tritium Appendix C was changed to discuss the use of the XSEX3 MCNPX User s Manual MCNPX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 11 Accelerator Production of Tritium 12 MCNPX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 MCNPX User s Manual MCNPX User s Manual Ap Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium 3 MCNPX Installation This chapter describes how to build MCNPX on a system The system will need a C and FORTRAN 77 compiler MCNPxX installs and runs on a variety of common Unix workstations The Cray system is no longer supported as of version 2 3 0 Some of our supported systems include e IBM RS 6000 AIX e DEC Alpha Digi
515. warning error message is issued if it is used NPS NPP NPSMG Table 5 40 NPS Keyword Descriptions Keyword Description N number of particle histories NPP Total number of histories to be run in the problem NPSMG Number of histories for which source contributions are to be made to the detec tor grid See Section 5 7 20 2 When the number of source histories exceeds NPSMG the time consuming process of determining the attenuation of the direct contribution is avoided by adding the average of the previous direct contributions into each of the appropriate tally bins Depending on the time required for a particular problem this can save from a few seconds to upward of ten MCNPX User s Manual 89 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 minutes per history in some cases As described above for a monoenergetic isotropic point source or a monoenergetic monodirectional surface source NPSMG 1 is adequate 5 5 6 4 CTME Computer Time Cutoff Form CTME x x maximum amount of computer time in minutes to be spent in the Monte Carlo calculation Default infinite Use As needed For a continue run job the time on the CTME card is the time relative to the start of the continue run it is not cumulative 5 5 7 Physics Models LCA LCB LEA LEB These cards control physics parameters for the BERTINI ISABEL CEM and FLUKA options These MCNPxX input cards have been defined to
516. with FSn card Can be used without FSn card Example F4 N 1237 SD4 1111 Note that the SDn card can be used to define tally divisors even if the tally is not segmented In this example the tally calculates the flux in the three cells plus the union of the three cells The VOL card can be used to set the volume divisor of the three cells to unity for example but it cannot do anything about the divisor for the union Its divisor is the sum of the volumes whether MCNP calculated or user entered of the three cells But the divisors for all four of the cell bins can be set to unity by means of the SDn card These entries override entries on the VOL and AREA cards See Section 5 7 1 2 4 for use with repeated structure tallies 5 7 16 FUn Special Tally or TALLYX Input Form FUn Ke oa DG or FUn blank Table 5 74 TALLYX Input Card Variable Description n tally number MCNPX User s Manual 131 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 74 TALLYX Input Card Variable Description Xj input parameter establishing user bin Default If the FU card is absent subroutine TALLYX is not called Use Used with a user supplied TALLYX subroutine or FTn card 5 7 17 FTn Special Treatments for Tallies Form FTn ID P41 P32 P13 ID Poy Poo P23 Table 5 75 FTn Card Special Treatment for Tallies
517. word Description P i Particle type i e n p e etc up to 10 particle types see Table 5 1 Source particles are considered to be those that come directly from the source defined by the user and those new particles created during nuclear interactions One should be aware that storage requirements can get very large very fast depending on the dimensions of the mesh since a separate histogram is created for each particle chosen If there are no entries on this card the information for neutrons is scored by default trans Must be followed by a single reference to a TR card that can be used to trans late and or rotate the entire mesh Only one TR card is permitted with a mesh card 1 In MCNPX version 2 1 5 there was no option to chose individual particles The type 2 Mesh Tally produced source points for all particles in the problem in one plot MCNPX User s Manual 97 MCNPX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium Energy Deposition Mesh Tally type 3 The third type of Mesh Tally scores energy deposition data in which the energy deposited per unit volume from all particles is included This can be due to the slowing of a charged particle the recoil of a nuclei energy deposited locally for particles born but not tracked etc The results are similar to the scoring of an F6 np tally as described in Section 8 3 Note that in MCNPX version 2 3 0 the o
518. work properly but do not need to be equally spaced It should be noted that the size of these meshes scales with the product of the number of entries for the three coor dinates Machine memory could become a problem for very large meshes with fine spacing Additional cards which can be used with Mesh Tallies are ERGSHn E1 E2 144 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 MSHMFn E1 F1 E2 F2 FMn Where E1 is the lower energy limit for information to be stored to the mesh n and E2 is the upper energy limit as they appear on the ERGSH card The default is to consider all energies The entries on the MSHMF card are pairs of energies and the corresponding response functions as many pairs can be designated as needed The FM card is the same as described in the MCNP users manual Since it must be read and stored by the MCNP subroutines it must not appear within the mesh data block between the tmesh and endmd cards The structure of the mesh as well as what quantities that are to be written to it are defined on two control cards in the MCNPX INP file The general forms of the two mesh cards are RMESHn P keyword i i 1 10 CMESHn P keyword i i 1 10 SMESHn P keyword i i 1 10 RMESH is a rectangular mesh CMESH is a cylindrical mesh and SMESH is a spherical mesh The n is a user defined mesh number The last digit of n defines the type of infor mation to be stored in the mesh P
519. y 226 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 c Source 0 radius beam z direction 1 GeV proton Ica 06 1 lea 2j0 imp h 1 0 phys h 1000 mode h print nps 1000 prdmp 2j 1 MCNPX User s Manual 227 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 3 Input for XSEX3 The input file for XSEX default name INXS has the following structure 1 Two records of title information 80 columns each 2 An option control record list directed format 3 Additional records as required by the chosen options list directed format Multiple cases may be processed for each case the above input structure applies When multiple cases are processed input quantities default to the preceding case If the title records of the second and subsequent cases contain the record must begin with a Sg The option control record has the structure NERG NANG FNORM KPLOT IMOM IYIELD LTEST Table 9 1 Parameter Meaning NERG Defines the number of energy or momentum bins for which cross sections will be calculated For NERG GT 0 an energy momentum boundary record is required For NERG 0 only energy integrated cross sections will be generated The default is 0 NANG Defines the number of cosine bins for which cross sections will be calculated For NANG not equal to 0 a angular bound ary record is required For NANG 0 only angle integrated c
520. y source particles only For KOPT 0 the edit is by cell numbers if KOPT 1 the edit is by material numbers If NPARM 0 the edit is over the entire system The parameters NTIM NTYPE and NFPRM are immaterial KPLOT 1 will produce plots of each edit table 214 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Tally option 5 or 105 represents the particle weight producing a given nuclide per source particle as such it is a dimensionless quantity The mean excitation is in units of MeV 8 Edit Option IOPT 6 or 106 Energy Deposition Option 6 is not available in this version 9 Edit Option IOPT 7 Mass and Energy Balance Option 7 is not available in this version 10 Edit Option IOPT 8 or 108 Detailed Residual Mass Edit Option 8 provides a detailed edit of residual masses by Z and N by Z only by N only and by mass number A The option accesses the records on HISTP for all interacting particle types If IOPT is preceded by a minus sign the edit is performed only for events initiated by primary source particles If KOPT 0 or 1 the edit is of the final residual masses including elastic collisions If KOPT 2 or 3 the edit is of the residuals after the cascade phase and before evaporation If KOPT 4 or 5 the edit is of masses immediately preceding fission If KOPT is even the edit is by cell number if KOPT is odd the edit is by material number If KPLOT 1 plo
521. y is inter preted as the lethargy spacing between bin boundaries Thus the record 0 1 800 will specify ten equal lethargy bins per decade from 800 MeV down e For NTIM gt 0 a record specifying NTIM upper time bin boundaries from low to high defined as the array TIMB l I 1 NTIM The first lower time boundary is always 0 0 The same four methods that are allowed for defining the energy boundaries may also be used to define the time bin boundaries 140 MCNPX User s Manual MCNPX User s Manual i Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium Table B 4 Order of HTAPESX Input Records IOPT option control record always required ERGB I l 1 NERG upper energy bin limits TIMB I l 1 NTIM upper time bin limits ITIP I l 1 NTYPE particle type identifiers LPARM l l 1 NPARM surface cell or material identifiers FPARM I l 1 NFPRM upper cosine bin boundaries DNPARM l l 1 NPARM 1 normalization divisors original source definition record for RESOURCE option new source definition record for RESOURCE option ITOPT TWIT TREAK TWIT parameters for TIME CONVOLUTION ERESP I l 1 NRESP energy grid for RESPONSE FUNCTION FRESP l l 1 NRESP 1 function values for RESPONSE FUNCTION IRESP l l 1 NRESP 1 interpolation scheme for RESPONSE FUNCTION segment definition record or window definition record arbitrary direction vector for defining cosine binning e
522. y one 2D plots are made If there are two contour or 3D plots are made SET does the same resetting of parameters that FREE does TFC x Plot the tally fluctuation chart of the current tally The inde pendent variable is NPS Allowed values of x are M mean E relative error F figure of merit L 201 largest tallies vs x NONORM for frequency vs x N cumulative number fraction of f x vs x P probability f x vs x NONORM for number frequency vs x S SLOPE of the high tallies as a function of NPS T cumulative tally fraction of f x vs x V VOV as a function of NPS 1 8 1 to 8 moments of f x x 1to8 vs x NONORM for f x A x X 1t08 VS xX 1c 8c 1 to 8 cumulative moments of f x x 1tos VS X MCNPX User s Manual 55 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 Table 5 3 MPLOT amp MCPLOT Commands Command Description KCODE 1 The independent variable is the KCODE cycle The individ ual estimator plots start with cycle one The average col abs trk len plots start with the fourth active cycle Plot k eff or removal lifetime according to the value of i 1 k collision 2 k absorption 3 k track 4 prompt removal lifetime collision 5 prompt removal lifetime absorption 11 15 the quantity corresponding to i 10 averaged over the cycles so far in the problem 16 average col abs trk len k effand one estimated standard deviatio
523. y particles differs for the energies where libraries and physics models are used This procedure is under review and will likely be changed in future versions of the code Energies of all secondary particles except photons are added into the heating KERMA fac tors for the neutron and proton libraries This photon treatment was implemented in the MCNP libraries well before the development of the MCNPX code However since MCNP4B does not track charged particles standard practice was to include the energies of all other particles in the heating numbers for the evaluated libraries We are increasingly finding that local deposition of secondary particle energies causes difficulties particularly when the energies of the secondaries are high or when the user is simulating thin vol umes When secondary particles are indicated on the MODE card MCNPX will subtract 1 In MCNPX version 2 3 0 residual nuclei cannot be tracked This is usually not a problem for heavy residu als however for light residuals such as a scattered hydrogen nucleus errors in energy deposition in small volumes can occur This has caused some users problems when tracking in small volumes where it is unlikely that the recoil hydrogen nucleus will not stop We will modify this practice in an upcoming release 110 MCNPX User s Manual MCNPxX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium their energies from the heating valu
524. y related to this is the fact that no adequate algorithm yet exists for charged particle detectors MCNPX has an active program of improvement in high energy and charged particle vari ance reduction techniques and features will be added in future versions as they are developed MCNPX version 2 3 0 has currently implemented one special technique Sec ondary Particle biasing described in Section 7 1 7 1 Secondary Particle Biasing Secondary particle biasing has been introduced into the MCNPX code for two main reasons e It allows splitting of secondary particles from high energy cascades in the energy range of interest e It allows the user to roulette the large number of particles in energy ranges that are of no interest to the problem This technique is especially useful in deep penetration problems starting with very high energy particles where the very large number of low energy secondary particles have little or no chance of contributing to the answer On the other hand one needs all of the high energy particles that one can get MCNPxX version 2 3 0 has been upgraded to allow the user to control the numbers of sec ondary particles as a function of energy and primary particle interaction To this end anew card has been added to the INP file as shown below MCNPX User s Manual 89 MCNPX User s Manual E Version 2 3 0 April 2002 LA UR 02 2607 Accelerator Production of Tritium SPABI p xxx E1 S1 E2 S2 Ta
525. y the bin width to normalize per MeV The total over energy will be unnormalized Table B 2 Applicability of Minus Sign Flags on Input Control Parameters IOPT IOPT NERG NTIM NTYPE NPARM NFPRM 1 101 O O O Z O O 2 102 3 103 5 105 8 108 9 109 10 110 11 111 12 112 13 14 114 zIoiIiz z oo olojoJloO zZzIolojololoO Z ZJ OJO Z O 2 2 0 0 Z Z 0 0 ZI Z Z Z Z Z Z Z Z Zz O Z O0 0 OJ O O CO OC O ziz O O Z O Z Z Z 2 MCNPX User s Manual 137 Accelerator Production of Tritium MCNPX User s Manual Version 2 3 0 April 2002 LA UR 02 2607 Table B 2 Applicability of Minus Sign Flags on Input Control Parameters Continued IOPT IOPT NERG NTIM NTYPE NPARM NFPRM 15 115 O N N N O N 116 O O N N O N O optional N not used NTIM defines the number of time bins for the tally when applicable the maximum is 100 The default is 0 implying that only a total over time will be produced If NTIM is gt 1 and is preceded by a minus sign the tally in each time bin will be divided by the bin width to nor malize per nanosecond the total over time will be unnormalized NTYPE defines the number of particle types for which the edit is to be performed for those options where it is applicable the particle type is that of the particle causing the event which
526. ymlink to the bertin and phitlib files in your working directory If you have more than just one person running the code from a server then it is probably worthwhile to edit src Ics inbd F to point to a specific location on your system where everyone can get the files as in method 2 above In the future we will build in the ability to look for all libraries using the same method now used for the nuclear data table libraries MCNPX User s Manual 29 30 MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 MCNPX User s Manual MCNPX User s Manual Version 2 4 0 September 2002 LA CP 02 408 4 Input Files Input to MCNPX consists of a number of files They can be part of the code package generated by problem runs or user supplied This section focuses on the user supplied INP the default name file which describes the problem to be run Input cards are summarized by card type in Section 5 10 The user will provide only a small subset of all available input cards in a given problem The word card describes a single line of input up to 80 characters MCNPxX input item limitations are summarized in Section 4 4 Modification of these values is accomplished by altering the source code and recompiling All features of MCNPX should be used with caution and knowledge This is especially true of detectors and variance reduction schemes Read and understand the relevant sections of the manual before using them MC

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