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fispact-2007: User manual - Culham Centre for Fusion Energy

1. User Manuat Issue 1 Feb 2007 UKAEA Fusion 150 FISPACT Test6 NOHEAD AINP FISPACT IRRADIATION OF TI EEF FW 1 0 MW M2 FUEL 5 46 1 00619E24 TI47 9 18148E23 TI48 9 28210E24 TI49 6 91755E23 TI50 6 79178E23 MIND 1 E5 FLUX 4 27701E14 ATOMS LEVEL 10000 1 TIME 2 5 YEARS PATH 3 46 R TI45 D 5 45 R SC44M PATH 1 46 R 5 46 PATH 5 TI50 R TI51 D V51 R V52 D CR52 R CR51 Test IRRADIATION OF TI EEF FW 1 0 MW M2 DENSITY 4 54 00619E24 18148E23 28210E24 91755E23 79178 23 E5 27701E14 ATOMS 148 SC48 149 SC48 SC48 ERROR 2 TI48 SC48 0 5 49 SC49 0 2 IME 2 5 YEARS ATOMS END END Test8 NOHEAD AINP FISPACT 1 PPM OF CO IN FE FW 1 0 MW M2 IRON MASS 1 0 2 FE 99 9999 CO 0 0001 MIND 1 E5 WALL 1 ATOMS EVEL 100 1 TIME 2 5 YEARS UNCERT 3 ATOMS EVEL 20 1 FLUX 0 ZERO TIME 0 1 YEARS ATOMS TIME 0 9 YEARS ATOMS END END UKAEA Fusion User Manuat Issue 1 Feb 2007 FISPACT 151 Test9 SPACT IRRADIATION OF BE EEF FW 1 0 MW M2 S EVEL 100 1 TIME 0 5 YEARS UNCERT 3 ATOMS EVEL 20 1 FLUX 0 TIME 0 083 YEARS ATOMS WALL 1 0 EVEL 100 1 TIME 0 5 YEARS UNCERT 0 ATOMS EVEL 20 1 FLUX 0 TIME 0 083 VEARS ATOMS WALL 1 0 EVEL 100 1 TIME 0 5 YEARS ATOMS EVEL 20 1 FLUX 0 TIME 0 083 VEARS ATOMS WALL 1 0 EVEL 100 1 TI
2. DENSITY 4 54 FUEL 5 FLUX 4 27701E14 ATOMS EVEL 100 1 TIME 2 5 YEARS DOSE 1 ATOMS EVEL 20 1 FLUX 0 ZERO TIME 1 MINS ATOMS TIME 1 HOURS ATOMS TIME 1 DAYS ATOMS TIME 7 DAYS ATOMS WALL 1 0 EVEL 100 1 DOSE 2 1 GROUP 1 NOSTAB TIME 0 5 YEARS ATOMS EVEL 20 1 FLUX 0 ZERO TIME 1 YEARS SPECTRUM END Test3 RRADIATION OF TI EEF FW 1 0 MW M2 DENSITY 19 254 1 5 1 00 5 1 100 1 2 0 YEARS ATOMS 1 20 1 0 5 1 5 EVEL 100 1 FLUX 4 27701E14 TIME 0 5 YEARS ATOMS EVEL 20 1 FLUX 0 ZERO NOCOMP NOSTAB TIME 1 022 VEARS ATOMS END END UKAEA Fusion User Manuat Issue 1 Feb 2007 FISPACT 149 Test4 SPACT IRRADIATION OF TI EEF FW 1 0 MW M2 S TI 100 0 DENSITY 19 254 MIND 1 E5 WALL 1 00 ATOMS HALF HAZA 1 22 2 23 TAB3 24 4 25 CONV 10 1 2 1 2 BREM 4 AR39 AR42 42 CL38 UNCERT 4 0 98 0 01 5 3 12 5 E8 3 EVEL 100 1 TIME 2 5 YEARS ATOMS EVEL 20 1 FLUX 0 ZERO NOT1 NOT2 NOT3 NOT4 TIME 1 022 YEARS ATOMS END END Test5 IRRADIATION OF TI EEF FW 1 0 MW M2 DENSITY 4 54 5 00619E24 18148E23 28210E24 91755E23 79178 23 E5 27701E14 l OMS EVEL 10000 1 TIME 2 5 YEARS ROUTES 46 5 44 5 1 13 0 ROUTES 1146 45 3 1 18 1 RESULT 2 5 44 1 01843E15 3 73063E20
3. A9 7 The error on the number of atoms AN is given by equation A9 8 SE LAND apasanka qaa sss A9 8 j Each pathway is a series of reactions and decays and the total number of atoms formed is the product of the individual cross sections as shown by equation A9 9 NIEDA ia A9 9 The error on the number of atoms formed bv a particular pathway is given by equation 9 10 Z J 9 10 ij k Using 49 5 to rewrite A9 3 gives equation 49 11 AQ z 9 11 N 1 1 Using 9 8 to rewrite A9 11 gives equation 49 12 2 4 2 AQ gt AN UOCE A9 12 pug Using A9 10 to rewrite A9 12 gives equation A9 13 Y yy 40 pe A9 13 N Using A92 to rewrite A9 13 gives equation A9 14 AQ 89 nde nei A9 14 i Using A9 7 to rewrite 49 14 gives equation A9 15 ka 2 S na A9 15 kN O Equation 9 15 shows how the error of a radiological quantity depends on the individual cross section errors This formula 18 User Manuat Issue 1 Feb 2007 FISPACT 117 correct in cases where there are no fission reactions on actinides but requires modification if actinides are included in the input materials In the derivation above it has been assumed that all errors are completely uncorrelated however this assumption is no longer valid if fission is included A particular fi
4. ATOMS LEVEL 20 8 TIME 30 4375 DAYS 1 41 ATWO ATOMS TIME 60 875 DAYS ATOMS TIME 91 3125 DAYS ATOMS TIME 182 625 DAYS ATOMS TIME 182 625 DAYS ATOMS TIME 182 625 DAYS ATOMS LEVEL 20 1 FLUX 0 ZERO TIME 60 ATOMS TIME 1 DAYS ATOMS TIME 29 4375 DAYS ATOMS TIME 152 1875 DAYS ATOMS TIME 182 625 DAYS ATOMS TIME 2 YEARS ATOMS TIME 2 YEARS ATOMS TIME 5 YEARS ATOMS Test201 Test205 Identical to Test181 Test185 except that NPROJ 3 UKAEA Fusion replaces NPROJ 2 Note that the various testcases are divided into several sets that use cross section data in the various group structures Table A14 1 shows the details of the group structures used User Manuat Issue 1 Feb 2007 FISPACT 175 Table A14 1 Details of energy groups for testcases Energy structure Tests 69 group 21 24 60 61 100 groups 1 10 172 groups WIMS 31 32 51 52 172 groups Vitamin J 4 45 175 groups 11 20 25 26 211 groups neutrons 81 90 95 96 211 groups deuterons 181 185 211 groups protons 201 205 315 groups 70 74 351 groups 100 104 Timings To give some idea of the relative speeds of the code on various platforms the running times for the test cases can be compared Table A14 2 shows running times for four platforms Table A14 2 Running times seconds on various platforms asoma 1 6 GHz a 3 4 GHz b 3 2 GH2 c 2 6
5. EAFV 4 AINP FISPACT DECAY OF HE 6 FUEL 1 HE6 1 0E20 DENSITY 1 785E 4 MIND 1 E5 GRAPH 10 11 UNCERT 0 LEVEL 100 1 FLUX 0 0 ZERO TIME 0 01 ATOMS TIME 0 04 ATOMS TIME 0 05 ATOMS TIME 0 4 ATOMS TIME 0 5 ATOMS TIME 4 0 ATOMS TIME 5 0 ATOMS TIME 5 0 ATOMS END END User Manuat Issue 1 Feb 2007 UKAEA Fusion 218 FISPACT test d3 test d4 UKAEA Fusion YEARS ATOMS YEARS ATOMS YEARS ATOMS YEARS ATOMS YEARS ATOMS YEARS ATOMS YEARS ATOMS YEARS ATOMS YEARS ATOMS NOHEAD MONITOR 1 EAFV 4 AINP FISPACT DECAY OF BE 10 FUEL 1 BE10 1 0E20 DENSITY 1 848 MIND 1 E5 GRAPH 10 11 UNCERT 0 LEVEL 100 1 FLUX 0 0 ZERO TIME 5 0 5 TIME 5 0 5 TIME 5 0E5 TIME 5 0 5 TIME 3 0E6 TIME 5 0 6 TIME 1 0 7 TIME 1 0 7 TIME 2 0 7 END END NOHEAD MONITOR 1 EAFV 4 AINP FISPACT DECAY OF C 11 FUEL 1 C11 1 0E20 DENSITY 1 9 MIND 1 0 GRAPH 10 11 UNCERT 0 LEVEL 100 1 FLUX 0 0 ZERO TIME 5 0 MINS TIME 5 0 MINS TIME 10 0 MINS TIME 10 0 MINS TIME 30 0 MINS TIME 1 0 HOURS TIME 3 0 HOURS TIME 5 0 HOURS TIME 5 0 HOURS TIME 5 0 HOURS END END ATOMS ATOMS ATOMS ATOMS ATOMS ATOMS ATOMS ATOMS ATOMS ATOMS User Manuat Issue 1 Feb 2007 FISPACT 219 test d5 NOHEAD MONITOR 1 FISPACT DECAY OF AL 26M FUEL 1 AL26M 1 0E20 DENSITY 2 6989 MIND 1 E5 GRAPH 10 11 UNCERT 0 LEVEL 100 1 FLUX 0 0 ZERO TIME 25 TIME
6. 9 30 PN pri EX eX abi P Ditti 9 31 UNS N p 1 Equation A9 31 can be applied repeatedly until 1 yielding equation A9 32 User Manuat Issue 1 Feb 2007 FISPACT 121 Ilo ney u Su Sasu A9 32 lo A The solution for Ni Is given In equation A9 20 the Laplace transform of this is given in equation A9 33 An andi ie a masa k h NE A9 33 Combining equations A9 32 and A9 33 yields equation A9 34 the final expression for the transform DANCER A9 34 In order to obtain the expression for it is necessary to use the inverse Laplace transform that is given in equation A9 35 where the variable p has been written as z to emphasise that the integral is defined in the complex plane Ctico 1 5 0 7 9 35 value of c can be set to zero since all poles in the transform are for Real z lt 0 this corresponds to all decay constants and cross sections being positive Given the form of the transform shown in equation A9 34 it can be seen that completing the path of integration by a semicircle at infinity in the negative half plane will contribute nothing to the integral and it is therefore possible to replace it with a contour enclosing all the poles of the transform The value of the contour integral is given by 2 times the sum of the residues at the
7. seen 120 Pathways containing 2 decays eene eere 123 Pathways containing an arbitrary number of decays seen 124 Limits in arbitrary pathways aan 124 Pathways in which the final nuclide reacts and decays sees 125 Summary of factors for each type of pathway 127 uncertainties is c REGN EEG 127 Collapsing uncertainty data oett 129 Appendix 10 y group structures 131 Appendix 11 Error messages 132 Appendix 12 Sequential charged particle reactions 143 Appendix 13 Platform differences 145 Personal computer ace qase ade teuer o ved naan 145 UNIX qum TEE 146 Appendix 14 Standard test cases 147 DCN 175 Appendix 15 EASY User Interface 177 2 e onte e otio edm mf 177 files ost 179 Graph plotting tas cue 180 Summary Ol output TIES d uu uu uuu aktar Br spa abbinat 181 Kumm 182 a au 183 EAF group cross section data te etu dues vede etin 184 Neutron REM dE 185 Thee
8. 99 ER Version 20012 aman E E RENE NS 200 el ae cit aed e Saca ed estt Version 2009 eases A tii Ba Version tC h hu tani Use of FISPACT Input Output streams and files Preliminary input ii ii ii DOMINANT ADOM IR umn a asme aa nn na agama s 24 DOSENDOSE SISTA ua quna Ra testis ees 24 TA 25 BAL DT Pusu TE TE TAI Bl 25 ERROR ADATOT anuon aN A itam sana ghas M CAM DUE 25 FISCHOOSE NCHO 2 FISCHO I I 1 NCHO 0238 239 26 FISYIELD NYLD SYMB I I 1 NYLD gt tette 27 PEUS 71 28 A NA 29 GENERIQIGENER Ruas naa 29 GRAPH NUMG GRSHOW GUNCRT eet tte ettet tentent 30 GROW PIG ANGE TOY s erae e a a eet Een 31 GRPCONVERT NESTRC NDSTRC ettet ttt tette 33 ee etd ic er E T Le a DU N EE 34 TAZ RIS de p Lu Edi UNE I dE M 34 RON ONE NONE 35 DENEL O wanana EDAM I mom Mp M IE OS M Md 35 LOOPS TOO RE OE 36 use aba E d 37 MIND S atus un ee ME e RE pA TED 38 MONITOR MON estiz kera ain ta cca 39 NEWFILE JSTRM NEWNAM e ia cesta vs dalle sext am anaq ua gr g 39 INO COMP Macs Ri
9. EVEL 20 1 FLUX 0 0 TIME 30 DAYS SPECTRUM ENDPULSE gt gt FLUX 1 02292 22 EVEL 100 1 TIME 1 0E 9 ATOMS EVEL 20 1 FLUX 0 ZERO UNCERT 2 NOSTABLE TIME 1 0E 9 ATOMS TIME 0 5 ATOMS TIME 0 5 ATOMS TIME 1 MINS ATOMS TIME 1 HOURS ATOMS TIME 5 HOURS ATOMS TIME 0 75 DAYS ATOMS TIME 1 0 DAYS ATOMS TIME 1 DAYS ATOMS TIME 2 DAYS ATOMS TIME 2 DAYS ATOMS TIME 1 DAYS ATOMS TIME 1 DAYS ATOMS TIME 1 DAYS ATOMS TIME 1 DAYS ATOMS TIME 1 DAYS ATOMS TIME 5 DAYS ATOMS TIME 10 DAYS ATOMS TIME 10 DAYS ATOMS TIME 10 DAYS ATOMS TIME 10 DAYS ATOMS TIME 10 DAYS ATOMS TIME 50 DAYS ATOMS TIME 100 DAYS ATOMS TIME 252 DAYS ATOMS TIME 0 76923 YEARS ATOMS TIME 1 YEARS ATOMS TIME 3 YEARS ATOMS TIME 25 YEARS ATOMS END END UKAEA Fusion User Manuat Issue 1 Feb 2007 FISPACT 161 Test26 NOHEAD AINP FISPACT Aluminium MASS 1 57E 05 1 AL 100 0 MIND 1 SPLIT 1 UNCERT 2 UNCTYPE 3 DOMINANT 100 0 HAZA HALF CLEAR EVEL 100 1 FLUX 4 725 08 TIME 600 ATOMS EVEL 20 1 FLUX 0 0 ZERO TIME 123 ATOMS TIME 29 ATOMS TIME 154 ATOMS TIME 30 ATOMS TIME 269 ATOMS TIME 30 ATOMS TIME 271 ATOMS TIME 30 ATOMS Test31 AINP FISPACT PWR FUEL 3 1 U235 POY Paluel DENSITY 10 1 FUEL 2 0235 7 948 22 0238 2 453 24 MIND 1 5 GRAPH 51112345 FLUX 3 25E 14 ATOMS LEVEL 20 8 TIME 30 4375 DAYS 1 41 ATWO DOSE 1 l OMS ME 60 875 DAYS l OMS ME
10. Test13 RRADIATION OF FE U EEF 175 FW 1 0 MW M2 MASS 1 0 2 FE 99 9999 0 0001 D 1 E5 PH51112345 L 1 00 l OMS LEVEL 100 1 TIME 2 5 YEARS HAZA HALF ATWO UNCERT 2 MINS ATOMS HOURS ATOMS DAYS ATOMS DAYS ATOMS YEARS ATOMS 000 YEARS ATOMS User Manuat Issue 1 Feb 2007 UKAEA Fusion 154 FISPACT Test14 NOHEAD AINP FISPACT IRRADIATION OF Ti EEF 175 FW 1 0 MW M2 MASS 1 0 1 TI 100 0 MIND 1 E5 WALL 1 00 ATOMS LEVEL 100 1 TIME 2 5 YEARS HAZA HALF ATWO DOMINANT 80 0 UNCERT 3 DOSE 1 ATOMS LEVEL 20 1 1 MINS ATOMS 1 HOURS ATOMS TIME 1 DAYS ATOMS 7 1 DAYS ATOMS YEARS ATOMS Test15 NOHEAD AINP FISPACT IRRADIATION OF Ti EEF 175 FW 1 0 MW M2 MASS 1 0 1 100 0 ND 1 E5 WALL 1 00 ATOMS LEVEL 100 1 TIME 2 5 YEARS HAZA HALF ATWO lt lt Test case for comment gt gt GENERIC 0 UNCERT 3 DOSE 1 ATOMS LEVEL 20 1 IME 1 MINS ATOMS ME 1 HOURS ATOMS lt lt Test case for comment gt gt TIME 1 DAYS ATOMS TIME 7 DAYS ATOMS TIME 1 YEARS ATOMS END END UKAEA Fusion User Manuat Issue 1 Feb 2007 FISPACT 155 Test16 SPACT IRRADIATION OF Ti EEF 175 FW 1 0 MW M2 S MIND 1 E5 GRAPH 311123 WALL 1 00 ATOMS LEVEL 100 1 SEQU 1 1 TIME 2 5 YEARS HAZA HALF ATWO UNCERT 3 DOSE 1 ATOMS LEVEL 20 1 FLUX 0 ZERO TIME TIME TIME TIME TIME END END MINS ATOMS HO
11. END UKAEA Fusion User Manuat Issue 1 Feb 2007 FISPACT 221 test d9 NOHEAD MONITOR 1 EAFV 4 AINP FISPACT DECAY OF FE 53M FUEL 1 FE53M 1 0E20 DENSITY 7 874 MIND 1 0 OVER MN53 ALAM 1 16680E14 1 GRAPH 10 11 UNCERT 0 LEVEL 100 1 FLUX 0 0 ZERO TIME 1 0 MINS ATOMS TIME 1 0 MINS ATOMS TIME 1 0 MINS ATOMS TIME 2 0 MINS ATOMS TIME 5 0 MINS ATOMS TIME 5 0 MINS ATOMS TIME 5 0 MINS ATOMS TIME 5 0 MINS ATOMS TIME 5 0 MINS ATOMS TIME 10 0 MINS ATOMS TIME 10 0 MINS ATOMS TIME 10 0 MINS ATOMS TIME 1 0 HOURS ATOMS TIME 1 0 HOURS ATOMS TIME 3 0 HOURS ATOMS TIME 6 0 HOURS ATOMS TIME 12 0 HOURS ATOMS TIME 1 0 DAYS ATOMS END END test 410 NOHEA MONIT EAFV AINP FISPA D OR 4 CT DECAY OF TH 230 FUEL TH2 DENSI MIND GRAPH UNCER LEVEL FLUX ZERO TIME TIME END END il 30 1 0E20 TX 11 72 1 E3 1011 100 1 0 0 User Manuat Issue 1 Feb 2007 UKAEA Fusion 222 FISPACT test dil NOHEAD MONITOR 1 EAFV 4 AINP FISPACT DECAY OF CU 64 FUEL 1 CU64 1 0E20 DENSITY 8 96 MIND 1 0 GRAPH 10 11 UNCERT 0 LEVEL 100 1 HOURS ATOMS HOURS ATOMS HOURS ATOMS HOURS ATOMS HOURS ATOMS DAYS ATOMS DAYS ATOMS DAYS ATOMS DAYS ATOMS ti N Ul 2 OO Ul Ul Ul Ul test 412
12. FISPACT radionuclides while an earlier report reference 33 gives a formula that allows values for other nuclides to be calculated Data from these references for the nuclides listed are transferred to EAF CLEAR with the default prescription used for all radionuclides not explicitly listed Reference 30 documents the EAF CLEAR 2007 library FISPACT can use these data to show the clearance index for individual nuclides and for the irradiated material EAF STOP 2007 EAF STOP is one of the data libraries assembled by UKAEA Culham to enable the effect of sequential charged particle reactions SCPR to be investigated The data were generated using the code SRIM 2003 7 Details of this effect are given in Appendix 14 but the FZK report detailing the data used for previous EAF versions reference 35 should be consulted for further details EAF STOP contains the differential ranges for p d h t and in all the elements from H to U The term differential range at a particular energy defines the distance travelled by the particle in the material in loosing 1 MeV of energy Data are given from 0 to 60 MeV EAF SPEC 2007 EAF XN 2007 UKAEA Fusion EAF XS contains the cross section data for n x reactions while EAF SPEC describes the energy distribution of the charged particles emitted in these reactions This is the second of the libraries required for calculations with SCPR The data are calculated by a theoretical m
13. interspersed as required In this manual recommendations are made about suitable values of parameters These recommendations refer to typical applications in fusion devices and reflect the experience gained by the author in running FISPACT The user should however give some thought to the values of parameters for his particular application especially if this represents a neutron irradiation under very different conditions FISPACT is able to calculate the effects of irradiation of a wide range of actinides This facility was introduced primarily to allow the naturally occurring actinides U and Th to be included in trace amounts in fusion relevant materials The EAF libraries contain infinitely dilute cross sections and so if actinides constitute a significant proportion of the input material then the data will not be physically representative since self shielding and burn up effects are not included Thus whilst it is possible to perform irradiations on pure actinides either to model fuel in a fission power station or in a transmutation device the results will not be as exact as if a specialised fission reactor code such as FISPIN were used If knowledge of the changes in the neutron spectrum during irradiation is available then by using the NEWFILE code word some modelling of fuel burn up is possible Users are warned that FISPACT and EAF will only give approximate results for the irradiation of large amounts of actinides It
14. interval This code word causes data on the clearance data of radionuclides to be input for the calculations to include these data and for the results for individual nuclides and summed clearance indices to be output Appendix 6 contains more information on these data values Note that both ATWO and CLEAR MUST not be used in a particular case CONV MAXXT 10 CONV 2 0 10 CONVS 2 0 10 UKAEA Fusion This code word allows the user to alter the convergence limits used in the integrating routines The number of iterations MAXXT can be set in the range 1 10 instead of using the default value of 10 The convergence limit CONV against which each nuclide is tested in the integrating routines for normal inventory calculations be specified The convergence limit for the integrating routines involving pathways is set by CONVS Note if a nuclide has not converged then it is flagged on the printed output by a It should be noted that in the majority of cases the nuclides that are flagged as not converged are of little practical importance The output values for nuclides that User Manual Issue 1 Feb 2007 FISPACT 23 CULTAB have not converged do not have the accuracy of the rest of the output and thus should be used with caution An example of the use of this code word follows 5 0 005 0 01 This would allow a maximum of 5 iterations and the inventory calculations attempt to get the
15. 22 elements total 100 0000 Density 78710 g cm3 Figure A15 9 The View materials window Help file Windows allows the user to view information on the application by means of Help The user can view an index jump between topics see pop up definitions of terms and use context sensitive help The present version of the interface contains help on the FISPACT code words and error messages Information on the Interface e g pictures of the dialog boxes is also given Note that when viewing a dialog in the application help is available by pressing the Fl key The present document gives only a very brief introduction to the usage of the application for full details the Help file should be used Figure A15 10 shows a typical help screen for a code word note the underlined terms e g HOURS which enables a jump to another topic Any dotted underlined terms can be clicked and will give a definition of the term in a pop up window Note that under the Vista operating system the Help file will not work User Manuat Issue 1 Feb 2007 UKAEA Fusion 188 FISPACT EASY 2007 User Interface Help File Edi Bookmark Options Help TIME T This code word allows the input of the irradiation or cooling time interval 7 seconds The time may be input in other units apart from seconds by following the value with one of the following code words specifying the time unit MINS HOURS DAYS or YEARS Note it
16. 3 4 IT l gt Nb 1 10 B7 96 6 B7 The branching in this sertes of decays means that three Bateman calculations were carried out UKAEA Fusion 214 Decay times 8 NA Zr 95 N Zr 95 N Nb 95 Ne Nb 95 N Nb 95m FISPACT N Nb 95m 5 00000E 04 1 00000E 05 5 00000E 05 1 00000E 06 2 00000E 06 3 00000E 06 4 00000E 06 5 00000E 06 6 00000E 06 8 00000E 06 1 00000E 07 5 00000E 07 1 00000E 08 9 93755 19 9 87549 19 9 39275 19 8 82238 19 7 78344 19 6 86685E 19 6 05819E 19 5 34476E 19 4 71536 19 3 67017 19 2 85665 19 1 90234 17 3 61888 14 9 93755 19 9 87549 19 9 39275 19 8 82238 19 7 78344 19 6 86685 19 6 05819 19 5 34477 19 4 71536 19 3 67017 19 2 85665 19 1 90234 17 3 61889E 14 6 14462E 17 1 21874E 18 5 69544E 18 1 04522E 19 1 75687E 19 2 21402E 19 2 48127E 19 2 60884E 19 2 63534E 19 2 49620E 19 2 22418E 19 2 27795E 17 4 35724E 14 6 14462E 17 1 21874E 18 5 69544 18 1 04522E 19 1 75687E 19 2 21402E 19 2 48128E 19 2 60884E 19 2 63534E 19 2 49621E 19 2 22419E 19 2 27797 17 4 35726E 14 6 50115E 15 1 22777E 16 4 00890E 16 5 08443E 16 5 03606E 16 4 50258E 16 3 97879E 16 3 51094E 16 3 09756E 16 2 41097E 16 1 87656E 16 1 24967E 14 2 37729E 11 6 50115E 15 1 22777E 16 4 00891E 16 5 08444E 16 5 03615E 16 4 50267E 16 3 97887E 16 3 51101E 16 3 09762E 16 2 41100E 16 1 87659E 16 1 25025E 14 2 377
17. 4 97862 05 26 4 97871E 05 76 4 5050 05 78 4 50492E 05 36 4 50484 05 27 4 50492E 05 77 4 0760 05 79 14 07622 05 37 4 07615 05 28 4 07622E 05 78 3 8770E 05 80 3 87742 05 79 3 6880 05 81 3 68832E 05 38 3 68825 05 80 3 3370 05 82 3 33733E 05 39 3 33727E 05 81 3 0200 05 83 3 01974E 05 40 13 01968 05 29 3 01974E 05 8 3 02500E 05 82 2 9850 05 84 2 98491E 05 83 2 9720E 05 85 2 97211E 05 84 2 9450 05 86 2 94518 05 85 2 8730 05 87 2 87246 05 86 2 7320 05 88 2 73237 05 41 2 73232 05 30 2 73237 05 87 2 4720 05 89 2 47235 05 42 2 47231 05 31 2 47235 05 88 2 3520 05 90 2 35177E 05 89 2 2370 05 91 2 23708E 05 43 2 23704E 05 90 2 1280E 05 92 2 12797E 05 91 2 0240 05 93 12 02419 05 44 12 02415 05 92 1 9250 05 94 1 92547E 05 93 1 8320 05 95 1 83156 05 45 1 83153 05 32 1 83156 05 9 1 83000 05 06 5 1 35300 06 05 UKAEA Fusion User Manuat Issue 1 Feb 2007 FISPACT 85 1 7420 05 1 74224E 05 ku 1 6570E 05 2 1 65727E 05 46 1 65724E 05 96 1 5760E 05 98 1 57644E 05 97 1 5000E 05 99 1 49956E 05 47 1 49953E 05 98 1 4260E 05 100 1 42642E 05 99 1 3570E 05 101 1 35686E 05 48 1 35683E 05 100 1 2910E 05 102 1 29068E 05 101 1 2280E 05 103 1 22773E 05 49 1 22771 0
18. DECIN Too many input errors An error has occurred in one of the DECIN functions responsible for the processing ofthe INPUT file check the input syntax File HALFUNC is not connected to stream 38 UNCTYPE Check that the file FILES contains a valid file name on a line starting with stream 38 Fractional error required for ERMAT ERROR There is no uncertainty data in the cross section library so the fractional error value MUST be specified for ERMAT FUEL and MASS both used FUEL Only one of these two code words can be used per case FUEL and MASS both used MASS Only one of these two code words can be used per case Group structures in GRPC and COLL incompatible COLLAPSE The output group structure defined by GRPCONVERT and COLLAPSE MUST be compatible IGENER can takes values 0 or 1 GENERIC The generic output is either on or off Incompatible input group structure GRPCONVERT The output group structure defined by GRPCONVERT and COLLAPSE MUST be the same IPCWRT can takes values 0 or 1 SEQUENTIAL The pseudo cross section output 15 either on or off User Manuat Issue 1 Feb 2007 FISPACT 135 ISEQUE can takes values 0 or 1 SEQUENTIAL Sequential charged particle reactions are either considered or not ISPLIT can takes values 0 or 1 SPLIT Additional summary table is either considered or not Isomer appears stable lt CHAINP gt One of the isomers that is to be included in the calculation by LOOP
19. dN j Dx A N 01 N AN A9 19 The solution is given in equation A9 20 NOSNE esi dit t A9 20 The differential equation satisfied by nuclide 2 is given in equation A9 21 dN ES 0 o N A N o N RS A9 21 Using a standard integrating factor the solution is given in equation A9 22 NOSE a ee y scuta tete A9 22 Evaluating the integral in equation A9 22 assuming that 0 0 yields equation A9 23 2 A tes A9 23 The differential equation satisfied by nuclide 3 is given in equation A9 24 rs 49 24 The solution obtained by integrating equation 49 23 is given in equation A9 25 0 09 2 N t 1 6 4 1 7 l A9 25 We can consider two limiting cases of equation A9 25 termed long lived and short lived Consider typical values for the quantities and the irradiation time T 10 em s 107 cm and 108 s When a nuclide has a half life of 1 s then 0 693 gt gt and gt gt When nuclide has a half life of 1000 y then 2 196 10 lt lt o lt lt T If both nuclides 1 and 2 are long lived then At lt lt 1 and the exponential can be expanded keeping terms up to 2 This limit is given in equation A9 26 User Man
20. dominant nuclides by specifying XDOM which is the cumulative percentage contribution above which pathways are not calculated The default value has been set at 98 An example of the use of this code word follows DOMINANT 90 0 In this case pathways are calculated for each dominant nuclide until the contribution made by the dominant nuclides to each of the radiological quantities is no more than 90 0 Other dominant nuclides that contribute to the remaining 10 have no pathway information DOSE NDOSE 1 lt DIST gt 0 UKAEA Fusion In versions prior to 3 0 all dose rates were calculated for a semi infinite slab of the material This is still the default if the code word is not used or if NDOSE 1 but if NDOSE 2 then the calculations are done for a point source of 1 gm of material at a distance of DIST metres DIST is not used for the semi infinite slab as the contact dose rate is always assumed The minimum distance is 0 3 m if a smaller value is specified then DIST is set to 0 3 m and a message to this effect is printed Appendix 3 gives more details of the method of calculation for the two options described above An example of the use of this code word follows DOSE 2 1 0 User Manual Issue 1 Feb 2007 FISPACT 25 In this case the dose due to a point source 1 g of the irradiated material at a distance of 1 m is calculated END TITLE This code word terminates the input of data for a
21. output may be written to external data files FISPACT is a part of the European Activation System EASY and the current report should be viewed in parallel with the complete document set for EASY this is discussed in detail in Appendix 18 It should be noted that the report on validation which in releases prior to 2003 was a separate document is now included as Appendix 19 of the current report This contains a set of examples that confirm the correct processing of input data and cases where FISPACT results can be compared with analytical calculations These tests can be run for each new version as part of the quality assurance procedure User Manual Issue 1 Feb 2007 FISPACT 3 Version summary Version 1 Version 1 of FISPACT enabled inventories to be calculated but it was not possible to correctly follow the production of the gases H H and He These nuclides were only correctly counted if they arose as products via reactions such as He n p H ie the gas producing nuclide is written as the daughter nuclide They can also arise as the outgoing projectile i e the p in the previous reaction is actually but for these cases the gas producing nuclide was assumed to be lost from the system Version 2 Version 2 addressed the issue of gas production by storing both the cross section for the standard reaction e g X n p Y and for the reaction in the form X n Y H For the latter reaction the effect
22. 02 16 8 04049E 03 20 1948 145158E 02 20 9 8 174596401 Table 7 A selection of FISPACT ID numbers for the EAF 97 library ID Nuclide Table 8 A selection of reactions from the EAF 97 library section b In order to compare Tables 6 and 8 it is necessary to combine the values for various reactions given in Table 8 to produce the User Manuat Issue 1 Feb 2007 UKAEA Fusion 206 FISPACT values used in FISPACT COLLAPX file Table 6 Table 9 gives these calculations and column 4 can be directly compared with column 3 of Table 6 showing agreement Table 9 Combining effective cross sections to give values in FISPACT form Daughter Reaction Expression Result b ID combinations 1 8 12 52861E 02 2 52861E 02 1 06530E 02 1 06530E 02 8 04049E 03 8 04049E 03 1918 Gn n p 2 75172E 03 3 72359E 03 1 45158E 02 n 2n p 8 04049E 03 n d nap 5 98257E 03 2 75172E 03 8 73429E 03 1919 n n d n d 1 56047E 02 5 98257E 03 2 158727 02 14 nt 0 094 2 62831E 02 1 56047E 02 4 561139E 02 n 2n p 3 72359E 03 1920 2 62831E 02 2 62831E 02 8 17455E 01 8 17455E 01 1922 2 52861 02 8 1770786E1 0 8 17455 01 Decay tests A series of simple decays from a single starting nuclide is considered In these cases there is an analytical solution which ca
23. 2007 FISPACT 159 Test23 Test24 NOHEAD MONITOR 1 AINP FISPACT PWR FUEL 3 1 U235 PWRDEAN DENSITY 10 1 FUEL 2 U235 7 948E22 U238 2 453E24 MIND 1 E5 FISYIELD 0 HAZA HALF FLUX 3 25E 14 EVEL 20 50 TIME 730 5 DAYS UNCERT 0 ATOMS NOSORT EVEL 20 1 FLUX 0 ZERO TIME 60 ATOMS TIME 1 DAYS ATOMS TIME 29 4375 DAYS ATOMS TIME 152 1875 DAYS ATOMS TIME 182 625 DAYS ATOMS TIME 2 YEARS ATOMS TIME 2 YEARS ATOMS TIME 5 YEARS ATOMS END END NOHEAD MONITOR 1 AINP FISPACT PWR FUEL 3 1 U235 PWRDEAN DENSITY 10 1 FUEL 2 0235 7 948 22 0238 2 453 24 MIND 1 5 FISYIELD 2 0235 PU239 HAZA HALF FLUX 3 25E 14 EVEL 20 50 TIME 730 5 DAYS UNCERT 0 ATOMS NOSORT EVEL 20 1 FLUX 0 ZERO ME 60 l OMS IME 1 DAYS ATOMS IME 29 4375 DAYS ATOMS IME 152 1875 DAYS ATOMS IME 182 625 DAYS ATOMS IME 2 YEARS ATOMS IME 2 YEARS ATOMS IME 5 YEARS ATOMS END END User Manuat Issue 1 Feb 2007 UKAEA Fusion 160 FISPACT Test25 NOHEAD MONITOR 1 AINP FISPACT IRRADIATION LMJ FW imm thick DENSITY 2 7 MASS 848 23 2 78 57 C 21 43 SEQUENTIAL 1 O0 4 44 MIND 1 5 lt lt IRRADIATION HISTORY 1 YEAR 12 SHOTS gt gt HAZA HALF DOSE 1 SPECTRUM lt lt lt lt gt gt PULSE 11 FLUX 1 02292 22 EVEL 100 1 TIME 1 0E 9 SPECTRUM
24. 286 287 288 289 290 291 292 293 294 295 296 297 298 299 300 301 302 303 304 305 306 UKAEA Fusion 1 4400 00 1 3700 00 1 3050 00 1 2350 00 1 1700 00 1 1250 00 1 1100 00 1 0900 00 1 0800 00 1 0700 00 1 0350 00 1 0100 00 9 8600 01 9 3000 01 8 7640E 01 8 6000E 01 7 9000E 01 7 0500E 01 6 8260E 01 6 2500E 01 5 4000E 01 5 3160E 01 4 8500E 01 4 3300E 01 4 1400E 01 3 9100E 01 3 5200E 01 3 1450E 01 2 8250E 01 2 4800E 01 2 2000E 01 1 8900E 01 1 6000E 01 1 3400E 01 1 1500E 01 1 0000E 01 9 5000E 02 7 7000 02 5 9000 02 4 3000E 02 171 172 173 174 175 1 12535 00 8 76425E 01 6 82560E 01 5 31579E 01 4 13994E 01 1 00001E 01 97 98 99 100 1 12533 00 8 76410 01 6 82549 01 5 31570 01 4 13987 01 112 113 114 115 116 117 118 119 120 121 122 123 124 125 126 127 128 129 130 131 132 133 134 135 136 137 138 139 140 141 142 143 144 145 146 147 148 149 150 151 152 153 154 155 156 157 158 159 160 161 162 163 164 1 37000 00 1 33750 00 1 30000 00 1 23500 00 1 17000 00 1 15000 00 1 12535 00 1 11000 00 1 09700 00 1 07100 00 1 04500 00 1 03500 00 1 02000 00 9 96000 01 9 86000 01 9 72000 01 9 50000 01 9 30000 01 9 10000 01 8 60000 01 8 50000 01 7 900
25. Issue 1 Feb 2007 FISPACT Introduction FISPACT is an inventory code that has been developed for neutron deuteron and proton induced activation calculations for materials in fusion devices It is a powerful code that can answer the basic questions about the numbers of atoms and the activity in a material following neutron or charged particle irradiation and can also give details of the pathways by which these nuclides are formed It can treat trace amounts of actinides that are able to fission and includes the effects of sequential charged particle reactions only for neutron irradiation This manual describes version 2007 that represents the outcome of developments of the code during the last twenty one years FISPACT was developed from the FISPIN inventory code that was designed for fission reactor calculations and dealt in greater detail with inventories arising from the irradiated fuel in reactor FISPACT 15 complementary to FISPIN and has been designed for activation calculations however it can be used with any type of neutron spectrum and is not restricted to only fusion applications FISPACT is now used by many groups throughout Europe and has been adopted by the ITER project as the reference activation code It is available on two computer platforms UNIX workstations and personal computers running a Windows operating system On the latter it can be used as part of the EASY User Interface which gives a user frien
26. NOHEAD MONITOR 1 EAFV 4 AINP FISPACT DECAY OF ZR 95 FUEL 1 ZR95 1 0E20 DENSITY 6 506 MIND 1 0E5 GRAPH 10 11 UNCERT 0 LEVEL 120 1 FLUX 0 0 ZERO TIME 5 0E4 ATOMS TIME 5 0E4 ATOMS TIME 4 0E5 ATOMS TIME 5 0E5 ATOMS TIME 1 0 6 ATOMS TIME 1 0 6 ATOMS TIME 1 0 6 ATOMS TIME 1 0 6 ATOMS TIME 1 0 6 ATOMS TIME 2 0E6 ATOMS TIME 2 0E6 ATOMS TIME 4 0E7 ATOMS TIME 5 0E7 ATOMS UKAEA Fusion User Manuat Issue 1 Feb 2007 FISPACT 223 test r1 NOHEAD MONITOR 1 EAFV 4 AINP FISPACT Simplified reactions on O 16 FUEL 1 016 1 0E20 DENSITY 0 001429 lt lt 1918 1 gt gt lt lt 1919 2 gt gt lt lt 1920 3 gt gt lt lt 1922 4 gt gt OVER 016 ACROSS 017 ACROSS C13 ACROSS C12 ACROSS N16 ACROSS N15 ACROSS 1918 0 0 ACROSS 1919 0 0 ACROSS 1922 0 0 OVER O17 ACROSS O18 ACROSS O16 ACROSS C13 ACROSS N17 ACROSS N16 ACROSS N15 ACROSS C14 ACROSS 1918 ACROSS 1919 ACROSS 1920 ACROSS 1922 OVER O18 ACROSS O19 ACROSS O17 ACROSS O16 ACROSS C14 ACROSS N18 ACROSS N16 ACROSS C15 ACROSS 1918 0 0 ACROSS 1920 0 0 ACROSS 1922 0 0 OVER O19 ALAM 0 5 5 MIND 1 0 HALF DOMINANT 100 0 UNCERT 3 LEVEL 100 1 FLUX 1 0E15 TIME 1 0 YEARS ATOMS END END 60189 5 0 0 0 0 18499 4 oF 0 0 0 0 0 0 0 0 0 0 CUO QC 42613E 5 O QOO U
27. This saves time in setting up the input file time in calculation and in the size of the output However there is one important area in which the modelling of the irradiation is crucial when calculating pathways and uncertainties The method used by FISPACT in the routine calculation of uncertainties stores the values of the inventory at User Manuat Issue 1 Feb 2007 FISPACT 229 the end of the first irradiation period These values are then used at latter times when long lived nuclides become dominant and so contribute in the uncertainty calculation It is therefore important that the first irradiation period should contain the bulk of the fluence so that an accurate calculation of uncertainties is possible Using a short on pulse at the beginning of the irradiation would mean that completely wrong uncertainty estimates would be calculated This reinforces the conclusion of the previous section Use a steady flux for all but the last few 5 10 on off intervals Conserve total irradiation time and fluence User Manuat Issue 1 Feb 2007 UKAEA Fusion 230 FISPACT References UKAEA Fusion Burstall FISPIN A computer code for nuclide inventory calculations ND R 328 R 1979 2 Forrest FISPACT 99 Validation EDS 2a 1998 3 J Kopecky H Gruppelaar and R A Forrest European Activation File for Fusion S M Qaim editor Int Conf Nuc Data Sci Tech J lich Germa
28. UKAEA Fusion 2 reactions and 2 decays A diagram of this pathway is shown in Figure 8 1 60m Sink nuclide reaction gt decay Figure A8 1 Diagram of a pathway from to Ni Several points in this definition require further discussion The pathway is linear because any side paths either in or out can be considered as part of some other linear pathway Also shown are the reactions or decays to a sink nuclide this is a fictitious nuclide which is not followed by FISPACT the reaction to it is actually the sum of all physical reactions on the target which do not lead to the next nuclide on the pathway In the case of a radionuclide on the pathway connected to the next nuclide by a reaction then a decay link is shown to the sink nuclide The nuclides are all distinct as written down although this conceals an important improvement in the calculation of pathways since version 3 0 As can be seen in Figure A8 1 a reaction arrow is shown between Ni and Ni this means that the backward reaction Ni n y Ni is included with the other forward reactions in the calculation This loop involving n y and n 2n reactions allows the effects of burn up of the parent in high fluxes to be included correctly Note that in the first step of the pathway there is no loop as the half life of is too short to have reaction data in the cross section library Physically the inclusion of these loops m
29. but now the source term S is replaced by T defined in equation A5 3 pe A5 3 Thus in addition to calculating in inventory equations a similar method is used to calculate the sensitivity coefficients defined in equations A5 4 S N x TN e A5 4 User Manuat Issue 1 Feb 2007 UKAEA Fusion 98 FISPACT Appendix 6 Data libraries FISPACT requires connection to several data libraries before it can be used to calculate inventories While any libraries in the correct format could be used the development of FISPACT over the last few years has run in parallel with the development of the European Activation File and this library is the recommended source of cross section data Together FISPACT and EAF make up the European Activation System EASY which is a complete package tailored for fusion applications The following libraries are required Cross section data for neutron induced reactions e Cross section data for deuteron induced reactions e Cross section data for proton induced reactions e Uncertainty data for neutron induced reactions e Decay data e Fission yield data for neutron induced reactions e Fission yield data for deuteron induced reactions e Fission yield data for proton induced reactions e Biological hazard data e Legal transport data e Clearance data e Gamma absorption data e Charged particle ranges in materials e E
30. group format was missing data subroutine so all 22 groups from the second group output 13 6 95 The output y spectrum lists energy This feature added 565 572 574 per group It would also be useful 577 to show number of ys per group These data should also be available in TAB4 12 7 95 The warning message that Ratio of An error was noted in CALC in 578 580 Fission Products Fissions differs the fission source term This was from 2 is seen more often than corrected Also the warning expected message is not appropriate if very small amounts of actinides are input If actinides lt 0 1 of input atoms then no warning 17 7 95 In multiple irradiations of actinides Variables are being re initialised 581 600 the reported burnups and number in CALC This must not be done of fissions are not correct the initial number of fissionable nuclides must be stored in a common UKAEA Fusion User Manual Issue 1 Feb 2007 Fusion User Manuat Issue 1 Feb 2007 FISPACT 193 Date Problem Solution Modification numbers 17 7 95 Warning about having multiple Modify test for warning 601 subintervals with actinides should only be given in an irradiation step 17 10 95 Index file now contains Fm Format statements in COL069 602 614 isotopes for these the ZA value COL100 COL172 COL175 and requires an I7 format not an I6 ENDFPR changed 18 10 95 Group wise files must contain the Change made on COL069 615 623 62
31. respect to decay constant are calculated If XSENS SIGMA then the sensitivity coefficients with respect to cross section are calculated However only one of these options can be specified for a case the code word MUST not be input twice The cut off value XNSENT is the value below which sensitivity coefficients are not printed a typical value might be 1 10 10 For each of the INSEN4 nuclides specified the sensitivity of that nuclide to each of the NSENS cross sections or decay constants is calculated The maximum value of both INSEN3 and INSEN4 is 50 If either INSEN3 or INSEN4 are set to 1 then the calculations are done for all cross sections or all nuclides It is not recommended that INSEN3 be set to 1 as with a large library the computing time would be prohibitive Setting INSEN4 to 1 gives a large amount of output but does not require much more time than typical to run User Manuat Issue 1 Feb 2007 FISPACT 51 Sensitivity calculations will be performed only for one time interval so it is possible to follow the irradiation with cooling steps if these are needed See Appendix 5 for further details of the sensitivity method Examples of the use of this code word follow SENSITIVITY SIGMA 1 E 10 2 3 C12 C13 C13 C14 13 C14 BE10 The sensitivity coefficients of the amounts of the three nuclides B C HC and Be to the values of the cross sections for the two reactions 2
32. the library with the neutron spectrum The physical theory and the numerical approximations employed to solve the set of differential equations are described in Appendix 2 FISPACT 1s recommended to be used as part of the package of data and codes referred to as EASY European Activation System EASY has been developed as a self consistent system of data libraries and code with this the user 15 assured that data will be in the correct format and that data in the cross section and decay libraries are consistent with each other The point wise EAF library covers the energy range from thermal User Manuat Issue 1 Feb 2007 UKAEA Fusion 12 FISPACT UKAEA Fusion 60 MeV but the upper energy of the group structures are lower the values are shown in brackets following each structure WIMS 10 MeV GAM II 14 9 MeV XMAS 19 6 MeV VITAMIN J 19 6 MeV TRIPOLI 19 6 MeV VITAMIN J 55 MeV and TRIPOLI 55 MeV Appendix 6 describes the data libraries e g cross section data decay data potential biological hazards clearance data and legal transport data of EASY in more detail The input file constructed by a user consists of a series of code words that fall into two categories The first series preliminary are concerned with library specification and the second main give details of the materials and irradiation history A separate file containing a list of file names and the various data streams un
33. 0 MIND 1 E5 HAZA LEAR a 5 ATOMS FLUX 1 0E12 EVEL 100 1 PECTRUM ULSE 150 EVEL 20 1 FLUX 0 TIME 1 0 HOURS SPECTRUM EVEL 100 1 LUX 1 0E15 IME 1 0 HOURS SPECTRUM ENDPULSE FLUX 0 ZERO TIME 1 DAYS ATOMS TIME 9 DAYS ATOMS TIME 90 DAYS ATOMS TIME 265 25 DAYS ATOMS TIME 9 YEARS ATOMS TIME 90 YEARS ATOMS TIME 900 YEARS ATOMSEND END OF MULTIPLE RUN i E d Test72 NOHEAD MONITOR 1 AINP FISPACT PURE IRON DENSITV 7 874 MASS 1 0 1 FE 100 0 MIND 1 E5 1 1 0E12 VEL 100 1 ME 1 0 YEARS ECTRUM LSE 10 PULSE 10 PULSE 5 EVEL 20 1 FLUX 0 TIME 1 0 HOURS SPECI EVEL 100 1 LUX 1 0E15 IME 1 0 HOURS SPECT ENDPULSE ENDPULSE ENDPULSE FLUX 0 ZERO TIME 1 YEARS ATOMS END END OF COLLAPSE d UKAEA Fusion User Manual Issue 1 Feb 2007 FISPACT 169 Test73 NOHEAD MONITOR 1 AINP FISPACT IRRADIATION LMJ FW imm thick DENSITY 2 7 MASS 848 23 2 B 78 57 C 21 43 TAB4 44 MIND 1 E5 lt lt IRRADIATION HISTORY 1 YEAR 12 SHOTS gt gt HAZA HALF DOSE 1 SPECTRUM lt lt lt gt gt PULSE 11 FLUX 1 02292 22 EVEL 100 1 TIME 1 0E 9 SPECTRUM EVEL 20 1 FLUX 0 0 TIME 30 DAYS SPECTRUM ENDPULSE La I ka CC gt gt FLUX 1 02292E 22 EVEL 100 1 TIME 1 0E 9 ATOMS EVEL 20 1 FLUX 0
34. 08 d Neaf dataNeaf 2007Neaf n asscfy 20070 09 d eaf_data eaf_2007 eaf_n fis 20070 10 aye On each line the stream number is followed by the full pathname of the file on the user s system that is to be connected to the stream The file names shown in column 4 of Table 2 are generic names that will be used in the text of this manual Actual file names can be chosen as desired on a particular system It is important to note that for printing via stream 6 the length of the line should be set to 165 characters This is necessary to enable as much information as practical to be output on a line so making the output compact and free of repetition The FILES file MUST be located in the directory that contains the FISPACT executable If the user requires streams for external files e g specified by the TABn code words then these should have values greater than 43 The name of the file is allocated automatically the user only has to define the stream number The TABn files are located in the directory that contains the FISPACT executable Note that the gamma absorption data ABSORP on stream 39 is required when FISPACT 2007 is run Versions previous to FISPACT 97 contained an earlier set of data internally Thus even if a version of EAF prior to EAF 97 or a non EAF library is used as input then stream 39 MUST be connected to ABSORP The files COLLAPX and ARRAYX MUST exist even if they are empty at the locations stated in FILES Other files
35. 1 43242E 17 2 05183E 14 2 93908 11 4 21001 08 9 27949 18 1 63431 19 2 17200 19 2 58129 19 3 60291E 19 3 88043E 19 3 88599E 19 3 88600E 19 3 88600E 19 9 27949E 18 1 63431E 19 2 17199E 19 2 58129E 19 3 60301E 19 3 88057 19 3 88502 19 3 88502 19 3 88502 19 1 45998E 19 2 57133E 19 3 41729E 19 4 06125E 19 5 66860E 19 6 10524E 19 6 11399E 19 6 11400E 19 6 11400E 19 1 45998E 19 2 57132E 19 3 41729E 19 4 06124E 19 5 66876E 19 6 10546E 19 6 11246E 19 6 11246E 19 6 11246E 19 Max diff 7 000071 002522 0 02519 Decay times 8 Ax Total Ar Total Max diff 0 00075 1 80000 04 3 60000 04 5 40000 04 7 20000 04 1 72800 05 4 32000 05 8 64000 05 1 29600 06 1 72800 06 1 15386 15 8 78328 14 6 68590 14 5 08935 14 1 10428 14 2 17131 12 3 11022 09 4 45513 06 6 38161E 03 1 153862E 15 8 783278 14 6 685895E 14 5 089352E 14 1 104280E 14 2 171308E 12 3 110227E 09 4 455155E 06 6 381658E 03 This test considers the treatment of decay branching by FISPACT The branching ratios are taken from the decay data library There is excellent agreement with the analytical calculations two Bateman calculations 57r decay test d12 User Manuat Issue 1 Feb 2007 957r 5 53219 10 s 3 02184 10 s stable 98 9096 B l gt 5 Np 3 11760 10 s 1 10 B7
36. 16 This is the recommended method of inputting materials except if special isotopic compositions are required Note that both FUEL and MASS MUST not be used in the same case in the input file Note it is not essential that the total of all elements is exactly 100 however if the total was say 80 and 1 kg was specified for TOTM then in fact only 800 g of material would be considered in the calculation It is recommended to ensure that the total percentage of all elements equals 100 User Manuat Issue 1 Feb 2007 UKAEA Fusion 38 FISPACT An example of the use of this code word follows In this case the composition of a stainless steel ignoring impurities and minor elements is specified 1 kg of the steel containing the seven listed elements is to be irradiated MIND MIND 1 UKAEA Fusion This code word allows the input of a parameter indicating the minimum number of atoms which is not set to zero during the integrations It is usually not important to consider a few atoms of a nuclide The default value is 1 but this means that an inventory with an extremely large number of unimportant nuclides will be generated and it is recommended that a value such as 1 105 be used for the MIND parameter It is possible to use a parameter value less than 1 if information on a wide range of nuclides is required Note that the value of MIND corresponds to the amount of material specified it does not refer to numb
37. 1660E 01 67 9 16609E 01 210 8 5280E 01 211 7 8890 01 153 7 88932E 01 79 7 88919E 01 68 7 56736E 01 23 7 55000E 01 212 7 0790E 01 213 6 7900E 01 69 6 79041E 01 UKAEA Fusion User Manual Issue 1 Feb 2007 FISPACT 87 214 6 3100E 01 215 6 1440 01 154 6 14421E 01 80 6 14411E 01 216 5 5590E 01 70 5 55951E 01 71 5 15780 01 217 5 0120E 01 72 4 82516 01 24 4 80500E 01 218 4 7850 01 155 4 78512 01 81 4 78503E 01 219 4 5520E 01 73 14 55174E 401 220 3 9810E 01 74 4 01690E 01 221 3 7270 01 156 3 72665E 01 82 3 72659 01 75 3 72665E 01 222 3 3890E 01 76 3 37201E 01 223 3 0510E 01 77 3 05113E 01 224 2 9200 01 157 2 90232E 01 83 2 90227E 01 225 2 7920E 01 78 2 76077E 01 25 2 77000 01 226 2 4980 01 79 2 49805 01 227 2 2600E 01 158 2 26033 01 84 2 26029 01 80 2 26033E 01 228 2 0450E 01 229 1 9030 01 81 1 94548 01 230 1 7600 01 159 1 76035E 01 85 1 76031 01 231 1 6740E 01 82 1 59283E 01 26 1 59700E 01 232 1 5230E 01 233 1 3710E 01 160 1 37096E 01 86 1 37093 01 83 1 37096E 01 234 1 2590E 01 235 1 1220E 01 84 1 12245E 01 236 1 0680E 01 161 1 06770E 01 87 1 06768E 01 237 1 0000E 01 85 9 90555E 00 27 9 87700E 00 238 9 1900E 00 86 9 18981E 00 239 8 9130E 00 240 8 3150E 00 162 8 31529E 00 88 8 31515 00 87
38. 17 7 08158E 17 1 71837E 18 9 55568E 18 1 84802E 19 3 37744E 19 6 44928E 19 8 74358E 19 9 99965E 19 9 99992E 19 1 41885E 16 5 16878E 16 1 06405E 17 7 08158E 17 1 71838E 18 9 55363E 18 1 84768E 19 3 37685E 19 6 44823E 19 8 74217E 19 9 99814E 19 9 99811E 19 User Manual Issue 1 Feb 2007 FISPACT 211 Fe decay test 49 6 00000 01 1 20000 02 1 80000 02 3 00000 02 6 00000 02 9 00000E 02 1 20000E 03 1 50000E 03 1 80000E 03 2 40000E 03 3 00000E 03 3 60000E 03 7 20000E 03 1 08000E 04 2 16000E 04 4 32000E 04 8 64000E 04 1 72800E 05 Decay times s 5 00000E 00 1 00000E 01 1 50000E 01 5 00000E 01 1 00000E 02 5 00000E 02 1 00000 03 2 00000 03 5 00000 03 1 00000 04 5 00000 04 1 00000 05 Nx K 39 1 95399 06 1 47216E 07 4 65072E 07 1 15401E 09 6 09647E 09 1 91968E 11 7 67422E 11 2 92239 12 1 53464E 13 4 71565 13 3 68830 14 7 77088 14 39 1 97855 06 1 47479E 07 4 65182E 07 1 15403E 09 6 09649E 09 1 91910E 11 7 67251 11 2 92183 12 1 53439 13 4 71488E 13 3 68780E 14 7 11031E 14 Total 4 46448E 18 3 30824E 18 2 45285E 18 3 15617E 17 3 49077E 16 1 87923E 16 1 69380E 16 1 37602E 16 7 37160E 15 2 61055E 15 6 49757E 11 8 18493E 09 4 464480 18 3 308244 18 2 452848 18 3 156171 17 3 490771 16 1 878938 16 1 693535 16 1 375807E 16 7 370447 15 2 610142 15 6 496562 11 8
39. 18 7 03579E 18 1 58540E 19 3 94100E 19 5 84803E 19 7 20568E 19 8 13229E 19 8 75495E 19 9 44811E 19 9 75556E 19 9 89174E 19 9 99918E 19 9 99999E 19 1 00000E 20 1 00000E 20 1 00000E 20 1 00000E 20 9 75146E 17 3 48717E 18 7 03580E 18 1 58540E 19 3 94100E 19 5 84814E 19 7 20582E 19 8 13243E 19 8 75499E 19 9 44817E 19 9 75557E 19 9 89174E 19 9 99948E 19 9 99983E 19 9 99983E 19 9 99983E 19 9 99983E 19 9 99983E 19 Max diff 0 00000 0 00527 0 00300 User Manual Issue 1 Feb 2007 UKAEA Fusion 212 6 00000 01 1 20000 02 1 80000 02 3 00000 02 6 00000 02 9 00000 02 1 20000 03 1 50000 03 1 80000 03 2 40000 03 3 00000 03 3 60000 03 7 20000 03 1 08000 04 2 16000 04 4 32000 04 8 04000 04 1 72800 05 Max diff 1 11022 05 8 88178 05 2 73115 06 1 07914 07 6 03517E 07 1 48437E 08 2 65499E 08 4 02700E 08 5 53546E 08 8 79619E 08 1 22260E 09 1 57308E 09 3 70698E 09 5 84556E 09 1 22614E 10 2 50930E 10 5 07564E 10 1 02083E 11 1 19244E 05 8 77086E 05 2 72715E 06 1 07933E 07 6 03401E 07 1 48415E 08 2 65477E 08 4 02692E 08 5 53529E 08 8 79593E 08 1 22249E 09 1 57293E 09 3 70662E 09 5 84555E 09 1 22613E 10 2 50927E 10 5 07556E 10 1 02081E 11 3 72935E 17 3 13333E 17 2 65562E 17 1 95660E 17 1 03504E 17 6 19100E 16 3 93808E 16 2 57322E 16 1 70003E 16 7 49873E 15 3 31882E 15 1 46962E 15 1 10861E 13 8
40. 2 24867E 03 52 2 24867E 03 18 2 23900E 03 177 2 1130E 03 178 2 0350E 03 140 2 03468 03 66 2 03465 03 53 2 03468E 03 179 1 7960E 03 180 1 5850 03 141 11 58461E403 67 1 58458E 03 181 1 5070 03 54 1 50733E 03 55 1 43382E 03 19 1 42500E 03 182 1 3640E 03 183 1 2340E 03 142 1 23410E 03 68 1 23407E 03 56 1 23410E 03 184 1 1170E 03 185 1 0100E 03 57 1 01039E 03 186 9 6110E 02 143 9 61117E 02 69 9 61100E 02 58 9 14242E 02 20 9 06900 02 187 8 4820 02 188 7 4850E 02 144 7 48518 02 70 7 48505 02 59 7 48518 02 189 7 0790 02 190 6 7730E 02 60 6 77287E 02 191 6 3100E 02 192 5 8300E 02 145 5 82947E3 193 5 1450E 02 194 4 5400 02 146 4 53999E 02 72 4 53991 02 61 4 53999 02 195 3 9810 02 02 71 5 82937 02 62 3 71703E 02 21 3 67300 02 196 3 5360E 02 147 3 53575 02 73 3 53569E 02 197 3 0430E 02 63 3 04325E 02 198 2 7540E 02 148 2 75364E 02 74 2 75359E 02 199 2 4300E 02 200 2 1450 02 149 2 14454E 02 75 2 14450E 02 201 2 0400E 02 64 2 03995 02 202 1 7780E 02 203 1 6700E 02 150 1 67017E3 204 1 5850E 02 02 76 1 67014E 02 65 1 48625E 02 22 1 48700E 02 205 1 3670E 02 66 1 36742E 02 206 1 3010E 02 151 1 30073E 02 77 1 30070E 02 207 1 1220E 02 208 1 0130E 02 152 1 01301E 02 78 1 01299E 02 209 9
41. 33 GRPCONVERT NESTRC NDSTRC This code word allows the user to read a neutron spectrum in an arbitrary number of groups 5 and instruct FISPACT to convert it into one of the seven standard structures NDSTRC must therefore be 69 100 172 175 211 315 or 351 using any other value will result in an error message The user must prepare a file containing the following data and connect it to stream 3 in the FILES file e NESTRC 1 values representing the arbitrary energy boundaries starting with the highest energy NESTRC values representing the flux values n cm 787 in each group starting with the high energy group e First wall loading MWm Text string maximum of 100 characters identifying spectrum Note that each of the above groups of items should start on a new line in the file but there should be no blank lines separating them The OUTPUT file will contain information about the conversion what fraction of the input groups are included in each output group and details of the input and the output spectra The converted spectrum is written to the file connected to stream 20 this contains the standard information for a FLUXES file NDSTRC values representing the flux values em s in each group starting with the high energy group e First wall loading MWm Text string maximum of 100 characters identifying the spectrum Note that although the text string can contain
42. 34865 19 9 95817 19 9 99516 19 9 99522 19 9 99522 19 8 78024E 18 6 32668E 18 2 12197E 18 7 11709 17 4 63672 16 1 96801E 14 8 35305 11 1 50480 07 8 780241 18 6 326675E 18 2 121969E 18 7 117093E 17 4 636727E 16 1 968019E 14 8 353091 E 11 1 504814E 07 Max diff 000073 0 04780 0 00093 smy decay test 46 1 60500 10 s X stable IT Decay times s Na Y 89m 89 89 NF X 89 A Total 5 00000E 00 1 00000E 01 1 50000E 01 2 00000E 01 5 00000E 01 1 00000E 02 2 00000E 02 3 00000E 02 8 05789E 19 6 49295E 19 5 23195E 19 4 21585E 19 1 15402E 19 1 33175E 18 1 77357E 16 2 36195E 14 8 05789E 19 6 49296E 19 5 23195E 19 4 21585E 19 1 15402E 19 1 33175E 18 1 77357E 16 2 36196E 14 1 94211E 19 3 50705E 19 4 76805E 19 5 78415E 19 8 84598E 19 9 86682E 19 9 99823E 19 9 99998E 19 1 94211E 19 3 50704E 19 4 76805E 19 5 78415E 19 8 84615E 19 9 86736E 19 9 99908E 19 9 99997E 19 3 47994E 18 2 80410E 18 2 25951E 18 1 82069E 18 4 98382E 17 5 75141E 16 7 65945E 14 1 02005E 13 3 479938E 18 2 804095E 18 2 259508E 18 1 820686E 18 4 983821E 17 5 751412E 16 7 659470E 14 1 020053E 13 Max diff gt 000042 000850 0 00029 User Manuat Issue 1 Feb 2007 UKAEA Fusion 210 FISPACT 55 Co decay test 47 7 20000 03 1 80000 04 3 60000 04 8 64000 04 8 64000 05 1 7
43. 4 3 4 Test 184 117 43 35 45 Test 185 44 12 10 12 Test 201 3 1 1 2 Test 202 2 1 1 1 Test 203 7 3 2 2 Test 204 74 27 22 30 Test 205 38 11 9 11 Notes a SUN Studio 10 FORTRAN running under Solaris 2 10 b Salford FTN77 for Win32 FORTRAN V4 02 running in a Windows NT4 0 command window under Virtual PC 2004 in the foreground c Salford FTN77 for Win32 FORTRAN V4 02 running in a Windows XP command window in the foreground d Salford FTN77 FORTRAN V4 02 running in a Windows 2000 command window in the background e 100 group library f Using TAPA option g Using ARRAY option h With PRINT parameter set to 2 All runs performed with FISPACT 2007 and EAF 2007 UKAEA Fusion User Manuat Issue 1 Feb 2007 FISPACT 177 Appendix 15 EASY User Interface Introduction The EASY User Interface an interactive application was originally developed it was called FISPACT Windows Interface to help users to prepare input files It is now a fully featured 32 bit Windows application that makes all aspects of running FISPACT easier Note that even if the platform chosen to do the actual FISPACT runs is a UNIX workstation some users may wish to use a PC to analyse the output The EASY User Interface is a standard MDI Multiple Document Interface having the following features User Manuat Issue 1 Input files can be viewed edited and saved New input files appropriate for various types of run can b
44. 40 91 224 6 506 90 51 45 11 22 17 15 0 17 38 0 2 8 41 92 90638 8 57 93 100 0 42 95 94 10 22 92 14 84 0 9 25 15 92 16 68 9 55 24 13 0 9 63 43 0 11 50 44 101 07 12 41 96 5 54 0 1 87 12 76 12 6 17 06 31 55 0 18 62 45 102 9055 12 41 103 100 0 46 106 42 12 02 102 1 02 0 11 14 22 33 27 33 0 26 46 0 11 72 47 107 8682 10 50 107 51 839 0 48 161 48 112 411 8 65 106 1 25 0 0 89 0 12 49 12 80 24 13 12 22 28 73 0 7 49 49 114 818 7 31 113 4 29 0 95 71 50 118 710 5 75 112 0 97 0 0 66 0 34 14 54 7 68 24 22 8 59 32 58 0 4 63 0 5 79 51 121 76 6 691 121 57 21 0 42 79 52 127 60 6 24 120 0 09 0 2 55 0 89 4 74 7 07 18 84 0 31 74 0 34 08 53 126 90447 4 93 127 100 0 54 131 293 3 0589 124 0 095 0 0 089 0 1 91 26 4 4 071 21 232 26 909 0 10 436 0 8 857 55 132 90545 1 873 133 100 0 56 137 327 3 5 130 0 106 0 0 101 0 2 417 6 592 7 854 11 232 71 698 57 138 9055 6 145 138 0 09 99 91 58 140 115 6 770 136 0 185 0 0 251 0 88 45 0 11 114 59 140 90765 6 773 141 100 0 60 144 24 7 008 142 27 2 12 2 23 8 8 3 17 2 0 5 7 0 5 6 61 0 7 264 62 150 36 7 520 144 3 07 0 0 14 99 11 24 13 82 7 38 0 26 75 0 22 75 63 151 965 5 244 151 47 81 0 52 19 64 157 25 7 901 152 0 2 0 2 18 14 8 20 47 15 65 24 84 0 21 86 65 158 92534 8 230 159 100 0 66 162 50 8 551 156 0 06 0 0 1 0 2 3
45. FISPACT code Version available for Macintosh Power PC version developed New in version 97 the Macintosh This platform is not supported in platform versions 99 or 2001 Half life The estimation of uncertainties of New in version 97 uncertainties radiological quantities now includes the contribution of half life uncertainties in addition to cross section uncertainties Redefinition of file The file names defined in FILES for New in version 97 names during the streams 12 17 and 20 can be changed course of a run during the course of a run Useful if the neutron spectrum varies during the irradiation Calculation of Clearance data for radionuclides are New in version 99 clearance index available in EAF 99 and are read on stream 40 The clearance index for each nuclide and the total inventory are calculated if required Dates made Y2K Years in dates shown at start and end of New in version 99 compliant run are printed using 4 digits instead of 2 1999 not 99 Dominant nuclides The top 20 nuclides for clearance index New in version 2001 for additional and gamma and beta heats are listed for quantities each time interval Uncertainties for Uncertainty estimates for clearance index New in version 2001 additional quantities and gamma and beta heats can be calculated for each time interval New summary table A summary table containing data on beta New in version 2001 at end of case and gamma heat and mean energy can be d
46. FISPACT runs and the preliminary inputs are rather different 1 Formation of a collapsed library from the full group cross section libraries and a specified neutron spectrum 2 Reading and processing of the decay data and a collapsed library to produce a condensed library an ARRAYX file which can optionally be followed by an inventory run 3 Reading a condensed library and using this to perform an inventory run a sensitivity analysis or an evaluation of pathways Examples are given for each type below COLLAPSE 175 FISPACT Collapse of EAF 2007 175 Zone 12 END END OF COLLAPSE In this case a cross section library EAF 2007 in 175 group format is collapsed with a neutron spectrum identified as Zone 12 SPEK ENFA EAF DEC 07 EAF 2007 100 Zone 13 TAPA FISPACT Write to arrayx file END END OF RUN In this case there was no existing ARRAYX file produced from the current decay data DEC 2007 so the TAPA option was used By specifying SPEK any nuclides with no y spectral data had this synthesised approximately this 15 User Manuat Issue 1 Feb 2007 UKAEA Fusion 60 FISPACT UKAEA Fusion recommended The collapsed cross section file COLLAPX in this case for EAF 2007 in Zone 13 is read and added to ARRAYX SPEK ENFA EAF DEC 07 EAF 2007 175 Zone 15 ARRAY FISPACT Write to arrayx file END END of library run In this case there was an existing ARR
47. Features of FISPACT Feature Details Comments Cross section data 100 GAM ID format Neutron spectrum Arbitrary group structure can be used for New in version 4 neutron spectrum internally converted to one of the standard structures changes to be easily made H and He isotopes produced by e g n p reactions properly included Dose rate Surface contact dose from infinite slab calculation Dose at arbitrary distance from a point New in version 3 source Bremsstrahlung The contribution of y rays produced by New in version 2 energetic B particles can be included in the dose rate Graphical output Data produced in suitable format for pow 2 New in version 2 Modification of Particular cross sections or decay data library data can be modified for a run Uncertainty data can be modified New in version 3 Dominant nuclides The top 20 nuclides for activity y dose New in version 2 rate heating and biological hazards modified in version 4 listed for each time interval Sensitivity method Option to calculate effect of a change of New in version 2 Time cross section or decay constant on the consuming method to production of a nuclide identify important reactions Pathway Method Calculates the amount of a nuclide that is New in version 2 A fast produced by a particular pathway or method to identify calculates all pathways between a parent important reactions can and daughter be used rou
48. K Dickens Gatlinburg 854 858 1994 UKAEA Fusion User Manuat Issue 1 Feb 2007
49. MAIN gt Density MUST be specified if FUEL 15 used to specify the input material User Manuat Issue 1 Feb 2007 FISPACT 143 Appendix 12 Sequential charged particle reactions This appendix gives only a very brief summary of the theory developed by the group at KfK Karlsruhe for the treatment of sequential charged particle reactions SCPR in inventory calculations Full details are given in reference 40 SCPR is two step process in which charged particles x are created in primary neutron induced reactions A n x followed by a charged particle induced reaction B x n C producing residual nucleus C In general B z A since the initial material may contain many different nuclides or B may be formed by transmutation of A If this process is included it is then possible to form nuclides with atomic number Z 1 and 742 from a nuclide with atomic number Z Note that with neutron induced reactions products with atomic numbers of Z Z 1 and Z 2 can be formed directly Neutron induced reactions can form nuclides with atomic number Z 1 only by B decays SCPR therefore make it possible to form nuclides that are not formed or only in very small quantities by neutron induced reactions and can therefore significantly alter the activation properties of a material Reference 40 shows how an expression for a pseudo cross section can be derived which is formally identical to the effective cross section used by FISPACT This is sh
50. Note that under the Vista operating system the Help file will not work Figure A15 1 shows a screen shot of the Interface with an icon for one of the various child windows that can be opened an INPUT file is shown Behind the summary window a part of a graph window is visible As with other Windows applications there is a menubar and toolbar containing sixteen buttons giving the user quick access to the most important features at the top of the window Details of the various features are given below UKAEA Fusion User Manual Issue 1 Feb 2007 FISPACT 179 EASY 2007 User Interface File View Options Automate Window Help Summary of OUTPUT Biel ES gt lni x Time s gt gt 26230E 9 26230E 9 6 00000 1 3 00000 2 3 50000 3 Heat Kika 9 06085 6 ose rate Svih 1 04787 1 dose Swika 2 67725 5 inh dose 5 5 78145 5 Clearance index 0 00000E 0 Beta heat Kika 3 63834E 6 heat 5 42245E 6 OUTPUT file selected Quantity Total activity Heat output Include Brem Ingestion dose v Inhalation dose Effective 2 Clearance v Beta heat output F INPUT Input files Gamma dose rate E Gamma heat output 1 03903 11 7 89578E 10 3 01733E 10 7 55776 9 4 89647 5 3 69903E 5 1 34884E 5 2 92031E 6 5 65255 1 4 27526 1 1 57221 1 3 57193 0 2 62086 5 2 65015
51. Numeric value required for GUNCRT GRAPH A numeric value MUST follow the code word Numeric value required for IA TABI A numeric value representing the stream to connected the TAB1 file MUST follow the code word Numeric value required for IARG BREM A numeric value MUST follow the code word Numeric value required for IB TAB2 A numeric value representing the stream to connected the TAB2 file MUST follow the code word Numeric value required for IC TAB3 A numeric value representing the stream to connected the TAB3 file MUST follow the code word Numeric value required for ID TAB4 A numeric value representing the stream to connected the TAB4 file MUST follow the code word Numeric value required for IGAMGP GROUP A numeric value MUST follow the code word Numeric value required for IGENER GENERIC A numeric value MUST follow the code word Numeric value required for INDX2 MASS A numeric value MUST follow the code word User Manuat Issue 1 Feb 2007 UKAEA Fusion 138 FISPACT Numeric value required for INSEN3 SENSITIVITY A numeric value MUST follow the code word Numeric value required for INSEN4 SENSITIVITY A numeric value MUST follow the code word Numeric value required for IPCWRT SEQUENTIAL A numeric value MUST follow the code word Numeric value required for IPRPA ROUTES A numeric value MUST follow the code word Numeric value required for ISEQUE SEQUENTIAL A nu
52. The chemical symbol MUST represent one of the elements H Fm and be entered in upper case Chemical symbol required lt RENUCL gt Specify the nuclide identifier as 108 not 108AG Chemical symbol required MASS A chemical symbol e g MUST follow the code word Code word PULSE has not been used ENDPULSE The code words PULSE and ENDPULSE MUST occur in a pair cannot have ENDPULSE without a matching PULSE Code word required ENFA A second code word MUST follow the code word Contribution must be in range 0 to 100 DOMINANT A value between 0 and 100 MUST be used D or R required to specify link PATH When specifying a pathway use an R if the link is a reaction or a D if it is a decay Daughter isotope not recognised lt OVERID gt The daughter isotope specified after the OVER code word is not in the index of nuclides check that the isotope has been correctly entered Daughter nuclide not in library PATH The daughter nuclide specified in a particular pathway is not present in the decay or cross section libraries Daughter nuclide of reaction not in library lt COL069 gt The daughter nuclide of a reaction in the cross section library is not present in the decay library has the correct decay library been used Daughter nuclide of reaction not in library lt COL100 gt The daughter nuclide of a reaction in the cross section library is not present in the decay library has the correc
53. This notation is also convenient for displaying in the FISPACT output Table A9 4 also shows the reduced factor for each type of link and final nuclide This factor is obtained from the factors for links summarised in Table A9 3 but with the terms containing flux and numerical constants removed This reduced factor is convenient since for a particular calculation both the irradiation time and the flux are constant for all pathways The final column defines the number of each type of link in the pathway Thus corresponding to equation A9 9 equation A9 60 can be written to express the dependence of the number of atoms on the character of the pathway Nptn n r 1 np N Ti a s S podus A9 60 k 1 I 1 where 1 if the final nuclide is short lived and A 44 if the final nuclide is long lived Thus equation A9 10 is modified as shown in equation A9 61 to reflect the changes due to the decay constant uncertainties Note that the limit on the second summation contains both short lived reactions and long lived decays since the error terms for X and 1 A are identical User Manuat Issue 1 Feb 2007 FISPACT 129 2 na n 2 n np 3 ay 44 b A9 61 ij k l 1 where 1 if the final nuclide is short lived and 0 if the final nuclide is long lived In equations A9 15 and 9 18 one group collapsed cross sections are used It is necessary for FISPACT to a
54. UKAEA Fusion 9 41 Performing the integrals and using the relationship in equation 9 42 yields equation A9 43 1 RR ANE A9 42 User Manuat Issue 1 Feb 2007 FISPACT 123 Nya 7 CD Nye Ta a e a fT a j l i l A9 43 Comparing equations A9 38 and A9 43 it can be seen that the first n terms of the sum are already correctly given by the first sum in equation A9 43 If the final term j n 1 is also to be correct then the identity shown in equation A9 44 must be true E y sU essen A9 44 j l Pathways containing 2 decays only Consider the 2 link pathway both decays shown in Figure A9 2 where it can be seen that the quantity linking two nuclides is rather than o as in Figure A9 1 Figure A9 2 A 2 link pathway consisting of decays only The solution of the set of differential equations for N is formally the same as above if is replaced by The solution for N3 is given in equation A9 45 AA Ny 0 A A 1 6 4 1 7 9 45 Considering the limit of equation A9 45 in both the long and short lived cases yields equations A9 46 and A9 47 respectively G A9 46 T Nf EE E A9 47 Thus the factor that can be deduced for the long lived nuclide decaying is and for a short lived nuclide it is 1 User Manuat Issue 1 Feb 2007 UKAEA Fusion 12
55. UKAEA Fusion User Manual Issue 1 Feb 2007 FISPACT 195 Date Problem Solution Modification numbers precision and checks made so that 2 7 97 Error due to undefined variable 13 10 98 Need to be able to read clearance Subroutine CLINP added 820 61 13 10 98 Modifications due to CLINP Commons modified variables 821 860 added and output formats changed Need to be able to read 315 group Subroutine COL315 added data files 15 10 98 Modifications due to COL315 Commons modified variables 862 876 added and output formats changed 16 10 98 Error if more than 200 time Arrays storing summary data now intervals in case act as buffers with earlier data discarded so that only data for the most recent 200 intervals shown 20 10 98 Modifications due to COL315 Defining F1 F2 F3 for 315 878 879 groups 20 10 98 Group conversions need to include Change array dimensions add 880 884 315 as a standard new code in GRPCON Variables in some commons not 885 886 typed correctly 2 11 98 Ensure no Y2K issues Increase size of date string and 887 946 make changes to system clock calls for all versions 5 11 98 Error if collapse with a zero Additional test to trap cases where 947 952 spectrum total flux is zero 10 12 98 Error when using PRINTLIB with Correction made in COL315 953 54 315 group spectrum 10 12 98 Warning from UNIX compiler on Extra added to FORMAT 9 FORMAT statement in GRPCON statement 10 12 98 Error
56. absorption Coefficients from I keV to 20 MeV Int J Appli Radiat Isot 33 1269 1982 13 A Khursheed Neutron induced activation of materials for the first wall of conceptual fusion reactors Ph D Thesis Imperial College 1989 14 M F James Data for Decay Heat Predictions NEANDC 245 289 1987 15 R E MacFarlane D W Muir and R M Boicourt The NJOY Nuclear Data Processing System Volume 1 User s Manual LA 9393 M 1982 16 R A Forrest J Kopecky and J Ch Sublet The European Activation File EAF 2007 neutron induced cross section library UKAEA FUS 535 2007 17 J Ch Sublet R A Forrest J A Simpson J Kopecky and D Nierop The European Activation File EAF 97 Cross section library n reactions UKAEA FUS 352 1997 18 J Ch Sublet J Kopecky R A Forrest and D Nierop The European Activation File EAF 99 REPORT file EDS 3a 1998 19 R A Forrest The European Activation File EAF 2007 deuteron and proton induced cross section libraries UKAEA FUS 536 2007 20 J Kopecky D Nierop and R A Forrest Uncertainties in the European Activation File EAF 3 1 Subfile EAF UN 3 1 ECN C 94 015 1994 21 The JEFF 3 1 Nuclear Data Library JEFF Report 21 OECD NEA 2006 22 JEF 2 2 Radioactive Decay Data JEF Report 13 OECD NEA 1994 User Manuat Issue 1 Feb 2007 UKAEA Fusion 232 FISPACT UKAEA Fusion 23 E Brown and R
57. agreement of each nuclide amount between iterations to within 0 596 while for the pathway calculations an agreement of 1 is satisfactory This code word adds additional lines at the beginning and end of the TAB files so that the files can be more easily processed by other computer programs The data written are unchanged by the use of this code word which is retained for consistency with previous versions DENSITY DENSTY This code word enables the input of the density of the material undergoing neutron irradiation with the parameter DENSTY density g cm If this code word is used then the total activity will also be output in units of Cicm in addition to the standard output in Bq kg If FUEL is used to specify the input material for a run in which an inventory is calculated then the density MUST be specified An example of the use of this code word follows DENSITY 8 96 The density of the material specified by MASS or FUEL is 8 96 g User Manual Issue 1 Feb 2007 UKAEA Fusion 24 FISPACT DOMINANT 98 0 In versions of FISPACT prior to 3 1 pathways were calculated for all the dominant nuclides at a particular time interval With some materials this meant that the calculational time was excessive and it was decided to reduce the number calculated by default and to give the user the ability to vary this It is now possible to ignore the pathway information for some of the less important
58. and PC n y are calculated SENSITIVITY LAMBDA 1 E 10 1 2 C14 N14 N14 N15 The sensitivity coefficients of the amounts of the two nuclides N and PN to the value of the half life of the decay of to N is calculated In both cases the coefficients are only printed if they are larger than 1 0 107 SEQNUMBER LNNM 175 This code word allows the user to change the number of energy groups used for an inventory calculation including sequential charged particle reactions SCPR By default data in 175 groups is assumed but for a spectrum with neutrons with energies above 20 MeV LNNM can be set to 211 More details about SCPR are given in Appendix 12 An example of the use of this code word follows SEONUMBER 211 User Manuat Issue 1 Feb 2007 UKAEA Fusion 52 FISPACT In this case 211 group data must be used for the inventory calculation SEQUENTIAL SEQUE 0 IPCWRT 0 SPECTRUM UKAEA Fusion This code word allows the user to include the effect of sequential charged particle reactions SCPR in the inventory calculations By default SCPR are not considered but if ISEQUE is set to 1 then the additional data libraries are read and the pseudo cross sections calculated These pseudo cross sections are included with the standard collapsed cross sections in inventory calculations The values of the pseudo cross sections can be seen if PCWART is set to 1 as this causes them to be included in the OUTPU
59. as a convenient divider to separate the library input from the irradiation details PROJECTILE NPROJ 1 SPEK UKAEA Fusion This code word defines the incoming particle for the activation calculations This code word MUST be used if a deuteron or proton library is used and it MUST come before the ENFA or COLLAPSE code words For a deuteron library NPROJ should be set to 2 for a proton library NPROJ should be set to 3 A neutron library uses the default value of 1 The code word NOERROR MUST be used for deuteron and proton libraries An example of the use of this code word in the collapse of a deuteron library follows MONITOR 1 PROJ 2 NOERROR COLLAPSE 211 FISPACT COLLAPSE EAF 20070 WITH IFMIF END END OF RUN This code word calculates an approximate y spectrum for nuclides in the decay library which have no spectral data Details about the approximate spectrum are given in Appendix 4 These nuclides are flagged by amp in the standard output and in the output of library data produced in a run with the code word PRINTLIB User Manuat Issue 1 Feb 2007 FISPACT 21 Main input ATOMS ATWO This section follows the code word FISPACT and contains information about the particular material elemental or isotopic composition and mass and the irradiation history times and flux values Code words specifying options such as pathways sensitivity coefficients or graphs are described
60. at the end of the time interval The time 18 shown in a title at the top followed by the inventory TIME INTERVAL 1 TIME IS 7 8894 07 SECS OR 2 5000 00 YEARS The inventory contains up to 11 columns of data excluding the nuclide identifier and flags and contains the number of atoms of the isotope the mass of the isotope in grams the activity in the B energy in kW the amp energy in kW the y energy kW the y dose rate in Sv h the ingestion dose in Sv the inhalation dose in Sv the ratio of the activity to the A value or the clearance index and the half life in seconds The output of the last four columns depends on whether HAZARDS ATWO or CLEAR and HALF are present in the input file NUCLIDE ATOMS GRAMS Bq H 1 2 80723E 20 4 662 04 0 000E 00 H 3 1 35830 20 6 767 04 2 420 11 b Energy a Energy g Energy DOSE RATE kW kW kW Sv hr 0 000E 00 0 00E 00 0 000E 00 0 000E 00 A AAA 07 0 00 00 0 000 00 0 000 00 INGESTION INHALATION Bq A2 HALF LIFE DOSE Sv DOSE Sv Ratio seconds 0 000E 00 0 000E 00 0 000E 00 Stable 4 113E 00 4 113E 00 6 052E 03 3 891E 08 When data for all the nuclides has been printed the total number of nuclides and the total number of nuclides not converged are output Summary and elemental inventory The totals of the activity in Curies and Becquerels for the irradiated material and the split of the activity between B and y decays Note that the value for y decays
61. ate ERE massa 186 Help Hle a 187 Appendix 16 Density and abundance data 189 Appendix 17 FISPACT modifications 192 Appendix 18 EASY documentation set 199 Appendix 19 Validation 200 Introd DOT 200 Data library processing E fada 200 Decay library processing ioi lea ie E D e ED ene ete n ees 200 Collapsing Cross Sections u noi irte ie rite Te 203 Decay testem es an e coc A ai fe 206 UT deca testi oeni f e 208 OCC RN E U2 as Qa D ID anus 208 ou LD Sh A A 208 m E TT 209 aa E 209 xod decay ICSE dO asua led ae 209 decay test_d7 210 TS decay test TAO t 210 RECOGE esto 211 Th decay TOS P sO KO CEU SEE ELA E DE CIS ELA CU 212 Gu decay qahaq ua C LR 213 ma eA 213 R action y n punusha d Eta dU er i bu qi ed oodd aids 214 Oxygen reactions F1 deett cie esee e Ma ett eee button 215 Sulphur reactions test 72 215 Chromium reactions test 216 Annex FISPACT input files circ tac a
62. be called several times during an irradiation if required JA specifies the nuclide that is to have data changed The identifier can be specified either using the format TE129M or by the material number Note the material number is the identification given to the nuclide internally by FISPACT its value can be seen in the decay data part of the PRINTLIB output if required It is followed by one of three code word options ACROSS JB SIG N 1 NGR JB is the daughter of the reaction and S G N is the new cross section barns for the N th energy group For all existing EAF libraries NGR 1 NGR is used to retain compatibility with FISPIN input JB is specified in the same manner as JA above Note that if a fission reaction is specified then JB must be 0 ALAM THALF INDX THALF is the new half life of the nuclide and NDX specifies the time unit seconds minutes hours days na RC years ERROR JB ERRFCX JB is the daughter of the reaction and 15 the new error factor for the cross section User Manuat Issue 1 Feb 2007 UKAEA Fusion 44 FISPACT Examples of the uses of this code word follow OVER 9 ACROSS HE6 1 05490E 2 Here the 1 group cross section for the reaction Be n o 6He 15 given the value 10 549 mb for all subsequent calculations in the OVER C14 ALAM 3000 0 5 P Here the half life of 4C is given the value 3000 0 years for all subsequent calculat
63. code word Numeric value required for TOTM MASS A numeric value MUST follow the code word Numeric value required for WALL WALL A numeric value MUST follow the code word Numeric value required for XDOM DOMINANT A numeric value MUST follow the code word User Manuat Issue 1 Feb 2007 UKAEA Fusion 140 FISPACT UKAEA Fusion Numeric value required for XNSENI SENSITIVITY A numeric value MUST follow the code word Numeric value required for XP MASS A numeric value MUST follow the code word Numeric value required for XRESU RESULT A numeric value MUST follow the code word Numeric value required for ZZZLVL UNCERTAINTY A numeric value MUST follow the code word Only 5 levels of nesting allowed PULSE When using the PULSE ENDPULSE construction it is only possible to nest the construction 5 levels deep Parent nuclide of reaction not in library lt COL069 gt The parent nuclide of a reaction in the cross section library is not present in the decay library has the correct decay library been used Parent nuclide of reaction not in library lt COL100 gt The parent nuclide of a reaction in the cross section library is not present in the decay library has the correct decay library been used Parent nuclide of reaction not in library lt COL172 gt The parent nuclide of a reaction in the cross section library is not present in the decay library has the correct decay library been used P
64. data window shown in Figure A15 7 The type of incoming particle is selected from the Type menu item by default neutrons 1s selected the choice is indicated in the caption The required database is selected one of the eleven EAF 2007 multi group libraries any neutron spectra of the correct structure are shown and selecting one displays the Reaction Rate graph at the bottom right of the window A target nuclide is entered in the UKAEA Fusion User Manuat Issue 1 Feb 2007 FISPACT 185 Target nuclide text box and a reaction chosen and the summary values are displayed at the bottom left of the window EAF 2007 group cross section data neutrons Oe x File Edit Type Graph Options er amp 83 5s 5 E uc Database 211 groups flat weighting Spectrum JIFMIF test Target nuclide 59 g 108m format Reaction Graph type Final state Combination Single final state g ground Nuclide target All final states C Total for all states Sum reactions to same daughter nuclide Co 59 n g Co 60 Source JEF 22 MERGE 113 4 6 Reaction Rate Thermal cross section 1 64443E 01 bams m 1stisomer Resonance Integral 3 26616E 01 bams 14 5 MeV cross section 4 56870E 04 barns 14 5 Me systematic 6 66858E 04 barns Effective cross section 3 99203 03 00 Quality score 0 1 E 5 ev 0 53ev 111 20 MeV 55 MeV Databa
65. for cooling steps C 20 should be used Short lived nuclides User Manuat Issue 1 Feb 2007 UKAEA Fusion 36 FISPACT that are in equilibrium are calculated by an approximate method Further details are given in Appendix 2 Note that very large values of C MUST be given in scientific notation Wherever possible the number of subintervals should be set to 1 to reduce computing time However if a particular integration gives the message Case not converged then N can be increased However since the non convergent nuclides are flagged in the output it is easy to judge if the nuclides that have not converged are unimportant and whether the message can therefore be ignored For sensitivity calculations a value N 5 is recommended If actinides are amongst the starting nuclides then a value N 5 or 10 should be used since the rate of fission of actinides is only updated at the end of each subinterval An example of the use of this code word follows LEVEL 100 1 TIME 2 5 YEARS In this run the time interval of 2 5 years is specified this is not split into subintervals and all nuclides with half lives 6 3 days will be considered in equilibrium LOOPS TL UKAEA Fusion In calculations of pathways the possibility of excitation of an isomer of a ground state nuclide on the pathway is not automatically considered However if the isomer half life 18 short and it decays by an isomeric transition back
66. fusion but for example from accelerator beam target interactions e g IFMIF or experimental devices Such libraries also allow group wise data to be plotted without weighting It is the user s responsibility to select the appropriate group wise library depending on the type of activation calculations that will be made The micro flux weighting process can have significant impact on the cross sections particularly for reactions with high thresholds User Manuat Issue 1 Feb 2007 FISPACT 91 Appendix 2 Solution of the differential equations The core task of FISPACT is the solution of a set of differential equations that describe the amounts of atoms of various nuclides present following the irradiation of a given material in a neutron field The set of differential equations is given in equation A2 1 BNA eo DNA 5 N 01 OY k where N is the amount of nuclide i at time t is the decay constant of nuclide i s A is the decay constant of nuclide j producing i s G is the total cross section for reactions on i cm G is the reaction cross section for reactions on j producing i cm dis the fission cross section for reactions on actinide k cm is the neutron flux n cm 51 9 is the source of nuclide i from fission Y is the yield of nuclide i from the fission of nuclide k The final term is only required if actinides are included in the initial material It
67. group boundaries of the VITAMIN J 211 formats are listed UKAEA Fusion User Manuat Issue 1 Feb 2007 FISPACT 101 in Appendix 1 Only a flat weighting library is available which is suitable for accelerator applications EAF UN 2007 A unique feature among activation libraries is the inclusion of an uncertainty file EAF UN containing data for all neutron induced cross sections Reference 20 describes the uncertainty data for EAF 3 1 while reference 16 describes the modifications made for EAF 2007 The uncertainty data are very simplified with no covariance information however the file enables FISPACT to give broad brush estimates of uncertainties for fusion applications EAF DEC 2007 In addition to cross sections the other basic quantities required by an inventory code are information on the decay properties such as half life of all the nuclides considered These data are available in the various evaluated decay data libraries FISPACT is able to read the data directly in ENDF B V or VI format it requires no pre processing to be done EAF DEC 2007 is based primarily on the JEFF 3 17 and JEF 2 22 radioactive decay data libraries with additional data from recent UK evaluations However not all of the 2231 nuclides that are needed are included in such sources For these nuclides data are taken from sources such as Brown and Firestone and ENDF B VI format files are constructed Reference 24 documents the DEC 2007
68. in the three energy regions in the reverse order 16 12 96 FISYIELD and UNCTYP code Added the missing GO TO 100 738 739 words used code belonging to statements in MAIN following code word 3 9 1 97 SCPR not included if ROUTES Using ROUTES now causes the used without a full inventory pseudo cross sections to be calculated pathnames for entries in FILES CHARACTER 80 20 1 97 The variable WOR not defined in WOR defined as MAIN 4 20 1 97 Usage of the character causes Replace by 764 770 problems when printing on some platforms 21 1 97 New data available for calculating Data read from stream 39 rather 771 787 dose rates than taken from DATA statements New algorithm used for point source printed correctly in Mac version variables 3 2 97 UNIX version had problems using NEWFILE code word section library after collapsing 4 2 97 UNIX version required change of Changes in MAIN CLOCK and 803 805 name and addition of additional CLOKK comment lines 5 2 97 Errors when reading spontaneous fission data that decay is spontaneous fission 5 2 97 Label changed to reflect new compiler version 1 7 97 Error when irradiating H or He Additional test added to identify 808 case where only nuclides 1 5 present 1 7 97 Variables not initialised DELB array correctly initialised 2 7 97 Inconsistencies between platforms Some variables made double 810 818
69. is necessary to use an efficient method of solution of the set of equations 2 1 since the total number of nuclides considered is over 1900 The method used in FISPACT and in FISPIN from which it was developed is that of Sidell This method is an extension of the Euler first order Taylor series which uses an exponential function of the step length Equation A2 2 shows the standard Euler solution and A2 3 the Sidell solution for the step time h N t h N t S 22 t User Manual Issue 1 Feb 2007 UKAEA Fusion 92 FISPACT UKAEA Fusion e 1 dN STN ya gt AD i t h N t ETA A2 3 where A A The error in using A2 3 is lower than A2 2 but for stability of the solution it is still necessary that the time step be related to the reciprocal of the largest eigenvalue For this reason a restriction is placed on the largest eigenvalue considered some nuclides are considered in equilibrium The number of steps in the computational solution of the Sidell method is greater than the Euler method but not sufficiently to outweigh the advantages The procedure is to split the irradiation time into two steps perform the calculation test the convergence of all the nuclides and if the test fails then repeat with double the number of time steps This procedure is continued until sufficient accuracy is achieved The results at each stage are corrected using the results from the previ
70. is required then LAMBDA MUST follow the code word else SIGMA MUST follow the code word User Manuat Issue 1 Feb 2007 UKAEA Fusion 136 FISPACT UKAEA Fusion LINA TAPA or ARRAY required ENFA The code word MUST be one of these three options LNNM can take values of 175 or 211 SEQNUMBER Data with either 175 or 211 groups MUST be used M or N required for isomer FUEL Specify isomer by for first or N for second MONIT can takes values 0 or 1 MONITOR The output of code words is either on or off NDSTRC can only be 69 100 172 175 211 315 or 351 GRPCONVERT The output group structure MUST be one of the 7 standard types NEAFVN can takes values 2 3 4 5 or 6 EAFVERSION Only versions 2 3 4 5 or 6 of EAF are considered NESTRC can takes values 2 to 400 GRPCONVERT The input group structure MUST have between 2 and 400 groups No neutron spectrum available lt ENDFPR gt In order to process the fission yield data a FLUXES file MUST be available check FILES to ensure that the name is correct No space before isomer label OVER There MUST be no space between the atomic mass and the isomer label in a nuclide identifier No value for density if FUEL used MASSIN Density MUST be specified if FUEL is used it can be calculated only if MASS is used No wall loading or ID in input There is either no data on the wall loading or no a text string describing the spect
71. library Care has been taken to ensure that EAF XS and EAF DEC are compatible All nuclides including isomeric states that can be formed from the various reactions in EAF XS are included so long as their half lives are greater than 1 second Some nuclides with shorter half lives are included where it is felt that they are of particular importance Short lived 1 s isomers which return to the ground state by an isomeric transition have no impact on activation calculations and most of these have been ignored User Manuat Issue 1 Feb 2007 UKAEA Fusion 102 FISPACT EAF N FIS 2007 FISPACT requires fission yield data if actinides are included in the input materials EAF N FIS is taken completely from the JEFF 3 1 fission yield library and FISPACT reads the file in ENDF B VI format with no pre processing Only 19 of the 102 nuclides in EAF N XS which have fission cross sections have any fission yield data in JEFF 3 1 at relevant energies For the remainder a neighbouring fission yield is used The file connected to stream 8 see Table 2 contains these associations EAF D 15 2007 EAF D FIS is taken completely from the UKFY 4 0 fission yield library and FISPACT reads the file in ENDF B VI format with no pre processing Only 19 of the 90 nuclides in EAF D XS which have fission cross sections have any fission yield data in UKFY 4 0 at relevant energies For the remainder a neighbouring fission yield is used The file
72. more details of the calculation of uncertainties Examples of the use of this code word follow UNCERT 2 UNCTYPE 2 Uncertainty calculations will be done but only using the half life uncertainties Cross sections are assumed to have no uncertainties Such a calculation is useful to isolate the contribution generally small of half life uncertainties UNCERT 2 3 Uncertainty calculations will be done but using both the cross section and half life uncertainties This code word allows the input of the total neutron first wall loading WALL in units of MW m for a fusion device This is converted to a flux value by using data read from the neutron spectrum file The neutron spectrum file FLUXES contains a value of the first wall loading e g 4 15 MW m The energy integrated flux e g 1 80 10 n em s which is the sum of neutrons in all the groups 15 calculated and equated to the wall loading during library processing User Manuat Issue 1 Feb 2007 UKAEA Fusion 58 FISPACT ZERO Note that it is the user s responsibility to ensure that this wall loading is correct when the spectrum file is constructed If a wall loading of 2 0 MW m was input then a flux value of 2 0 4 15 1 80 10 n cm s would be used in the calculations It is a convenient alternative to using FLUX for irradiation of first wall materials but great care must be exercised if used for irradiations in other than first wall spe
73. pairs of reactions such as n y and 2 mean that it is not possible to start at the initial nuclide and solve the differential equations for each nuclide in the chain sequentially Indeed if it were easy to make analytical solutions then there would be no need for a numerical code such as FISPACT However by using the FISPACT facility to overwrite data for a particular run it is possible to consider UKAEA Fusion User Manuat Issue 1 Feb 2007 FISPACT 215 some simplified pathways that can be solved analytically Three cases are considered and one of these is the pathway that was considered in the IAEA Benchmark Oxygen reactions test r1 160 stable n y O stable n y O stable n y O 1 57788 10 s Cross section libraries contain many reactions on the oxygen isotopes other than the capture reactions For this simplified pathway all these cross sections are set to zero by means of the OVER code word in the input file Note that the correct half life of O is 26 91 s but with such a short half life all the O atoms will have already decayed by the end of the 1 y irradiation to avoid this the half life is altered to 0 5 y The values of the cross sections used for the calculations are included in the table below Cross section b diff 6 60189E 05 9 99998E 19 9 99998E 19 0 00000 1 18499E 04 2 08340 14 2 08339 14 0 00048 O 18 5 42613 05 3 89555 08 3 89546 0
74. that when specifying time intervals it is necessary to remember that the interval times are cumulative Specifying times of 1 year 5 years and 10 years will actually cause inventories to be calculated at 1 year 6 years 1 5 and 16 years 1 5 10 Forgetting this point is responsible for many of the problems experienced by users User Manuat Issue 1 Feb 2007 FISPACT 15 Input Output streams and files The FILES file shows the locations of all the files that FISPACT requires for input of data and the conditions for a particular run and for output FISPACT reserves streams 1 43 for this input output Table 2 gives the functions of the reserved streams with the direction of data flow indicated by I input or O output Table 2 Standard streams and generic file names Energy group structure neutron spectrum and wall ARB FLUX I I I I I I I I I I I I I I I I I loading for arbitrary group structure List of all files and their stream numbers System input System output Cross section uncertainties library Links between fissionable nuclides and fission yields Fission yield library Graphical data output A data Collapsed cross section library Condensed library data Biological hazard data Summary of nuclides and isomers Decay data Collapsed cross section library Index of materials in library Cross section library Neutron spectrum and wall loading Range of proton in each element Ra
75. the running time by a few percent User Manual Issue 1 Feb 2007 UKAEA Fusion 42 FISPACT NOSTAB NOT1 NOT2 NOT3 4 Note removing the dominant nuclide list also disables the output of pathways and uncertainty estimates that might have been requested by the UNCERTAINTY code word Use of this code word inhibits the printing of any stable nuclides in the inventory and is useful when the inventory is large and it is required to save space This code words switch off the output to the external files that was switched on by the TAB1 code word Both TAB1 and NOTI can be used several times during a case to restrict the output as required This code words switch off the output to the external files that was switched on by the TAB2 code word Both TAB2 and can be used several times during a case to restrict the output as required This code words switch off the output to the external files that was switched on by the TAB3 code word Both TAB3 and NOT3 can be used several times during a case to restrict the output as required This code words switch off the output to the external files that was switched on by the TAB4 code word Both 4 and NOTA can be used several times during a case to restrict the output as required User Manuat Issue 1 Feb 2007 FISPACT 43 OVER JA This code word enables library data to be modified for a particular case It can
76. this code word can still be used so long as only pathway data 15 required User Manuat Issue 1 Feb 2007 FISPACT 41 NOFISS NOHEAD NOSORT Note that if this code word is used with the ERROR code word then the user MUST supply values of the fractional error If an output of the data libraries is requested with the PRINTLIB code word and no uncertainty data exist then NOERROR MUST be used In all cases the code word MUST come near the top of the input file before the code words COLLAPSE UNCERTAINTY PRINTLIB or ERROR This code word stops the fission yield data from being input and processed during the preparation of the ARRAYX file Note it should only be used if the cross section library contains no actinide fission cross sections This code word stops the printing of the header and user information at the beginning of the output and is useful if it is required to reduce the printed output Note that if this code word is used it MUST be at the beginning of the input file and must not be proceeded by any characters not even blanks The default output includes a sorted list of the dominant nuclides where a maximum of 20 nuclides is shown The nuclides are sorted by activity heat y dose rate ingestion dose inhalation dose B heat y heat and clearance index The list can be removed by the use of this code word to reduce running time although including the list only typically increases
77. to remove bugs and add features in response to feedback from users Appendix 17 gives details of changes in FISPACT made in response to particular problems User Manuat Issue 1 Feb 2007 UKAEA Fusion ii FISPACT Acknowledgements Disclaimer The development of FISPACT and the production of this documentation have been supported by the United Kingdom Engineering and Sciences Research Council and the European Communities under the contract of Association between EURATOM and UKAEA and were carried out within the framework of the European Fusion Development Agreement The views and opinions expressed herein do not necessarily reflect those of the European Commission FISPACT has been developed over the last twenty one years at Harwell and Culham The efforts of D A J Endacott A Khursheed and J Ch Sublet and the advice of the late M G Sowerby in the early development are acknowledged Neither the authors nor UKAEA accept responsibility for consequences arising from any errors either in the present documentation or in the FISPACT code Contact person UKAEA Fusion Feedback on the use of FISPACT is welcomed Please contact R A Forrest with comments or in case of problems Dr R A Forrest EURATOM UKAEA Fusion Association D3 1 92 Culham Abingdon Oxfordshire OX14 3DB Tel 44 1235 466586 Fax 44 1235 466435 e mail robin forrest a jukaea org uk Internet www fusion org uk easy2007 User Manuat
78. will be created if they do not already exist User Manual Issue 1 Feb 2007 FISPACT 17 Preliminary input This section of the input file deals with the input and processing of library data and ends with the code word FISPACT An important part of library processing is the collapsing of the cross section library with a neutron spectrum The user must construct a file generic name FLUXES containing the spectrum data The format is not rigid and the numbers can be stored in the file as say a single column or six columns as convenient Note that the group flux values in the file MUST be entered starting with the highest energy group Following the spectrum a single number is required the wall loading MW m corresponding to the particular neutron spectrum If the WALL code word is used in the INPUT file then this value is used to convert the wall loading parameter to a flux value If the FLUX code word rather than WALL is always used then a dummy value such as 1 0 can be used Following this a line of text up to 22 characters identifying the spectrum is input This text string is used as part of the library information for each subsequent inventory run Diagrams of the basic types of FISPACT runs are shown in Figures 2 4 In Figure 2 the cross section library CROSSEC is collapsed with the neutron spectrum FLUXES to produce the collapsed cross section library COLLAPX In Figure 3 the collapsed cross sec
79. 0 targets are held in a modified ENDF B format In the case of capture and fission cross sections the point wise file has been processed from an evaluated file using NJOY P to reconstruct the resonance region from resonance parameters No self shielding is included and the temperature for Doppler broadening is 300K This library is available to users but before it can be used by FISPACT it is necessary to process it into a particular group cross section format Reference 19 documents the EAF 2007 deuteron induced cross section library EAF P XS 2007 EAF P XS is the point wise proton induced cross section library Data on 67 925 cross sections on 803 targets are held in a modified ENDF B format In the case of capture and fission cross sections the point wise file has been processed from an evaluated file using NJOY P to reconstruct the resonance region from resonance parameters No self shielding is included and the temperature for Doppler broadening is 300K This library is available to users but before it can be used by FISPACT it is necessary to process it into a particular group cross section User Manuat Issue 1 Feb 2007 UKAEA Fusion 100 FISPACT format Reference 19 documents the EAF 2007 proton induced cross section library EAF N GXS 2007 Eleven group cross section libraries are available for the neutron induced library that can be used as input to FISPACT The group boundaries of the WIMS 69 GAM II 100 XMAS
80. 00000 03 314 3 0000 03 172 3 00000E 03 315 1 1000E 04 316 1 0000E 05 176 1 00000 05 101 1 00000E 05 173 1 00000E 05 70 1 00000 05 Table 1 2 Energy group boundaries for the two high energy standard structures VITAMIN J 1 5 5000E 07 1 5 5000E 07 2 5 4000E 07 2 5 4000E 07 3 5 3000E 07 3 5 3000E 07 4 5 2000E 07 4 5 2000E 07 5 5 1000E 07 5 5 1000E 07 6 5 0000E 07 6 5 0000E 07 7 4 9000E 07 7 4 9000E 07 8 4 8000E 07 8 4 8000E 07 9 4 71000E 07 9 4 71000E 07 10 4 6000 07 10 4 6000 07 11 4 5000 07 11 4 5000 07 12 4 4000 07 12 4 4000 07 13 4 3000 07 13 4 3000 07 14 4 2000 07 14 4 2000 07 15 4 1000 07 15 4 1000 07 16 4 0000 07 16 4 0000 07 17 3 9000 07 17 3 9000 07 18 3 8000 07 18 3 8000E 07 19 3 7000E 07 19 3 7000 07 20 3 6000 07 20 3 6000 07 21 3 5000 07 21 3 5000 07 22 3 4000 07 22 3 4000 07 23 3 3000 07 23 3 3000 07 24 3 2000 07 24 3 2000 07 25 3 1000 07 25 3 1000 07 26 3 0000 07 26 3 0000 07 27 2 9000 07 27 2 9000 07 28 2 8000 07 28 2 8000 07 29 2 7000 07 29 2 7000 07 30 2 6000 07 30 2 6000 07 31 2 5000 07 31 2 5000 07 32 2 4000 07 32 2 4000 07 33 2 3000 07 33 2 3000 07 34 2 2000 07 34 2 2000 07 35 2 1000 07 35 2 1000 07 36 2 0000 07 36 2 0000 07 37 1 9640 07 1 1 9640E 07 37 1 9640E 07 1 1 9640E 07 38 1 7330E 07 2 1 7330E 07 38 1 7330E 07 2 1 7330E 07 n 36 lt n s n 36 n 5 338
81. 0000E 9 99983E 1 000008 1 00000 07 9 5120E 9 51229E 9 0480E 9 04837E 9 04822E 8 6070E 8 60708E 8 1870E 8 18731E 8 18717 4 8 18731E 06 7 7880 7 78801E 7 4080 7 40818E 7 40806 7 0470E 7 04688E 6 7030E 6 70320E 6 70309E 6 70320E 6 5920 6 59241E 6 3760E 6 37628E 6 0650E 6 06531E 6 06520E 6 0653 1E 6 06600E 06 5 7690E 5 76950E 5 4880 5 48812 5 48802E 5 48812 5 2200 5 22046E 4 9660E 4 96585E 4 96577E NN 3 gt Ut User Manuat Issue 1 Feb 2007 UKAEA Fusion 84 FISPACT GRP TRIPOLI VITAMIN GRP XMAS wms 4 7240 06 4 72367 06 4 4930E 06 4 49329 06 13 4 49321 06 11 4 49329E 06 34 4 0660 06 35 4 06570E 06 14 4 06562E 06 35 3 6790E 06 36 3 67879E 06 15 3 67873 06 12 3 67879E 06 3 13 67900E 06 36 3 3290 06 37 3 32871 06 16 3 32865E 06 37 3 1660E 06 38 3 16637E 06 38 3 0120E 06 39 3 01194E 06 17 3 01189 06 13 3 01194E 06 39 2 8650E 06 40 2 86505E 06 40 2 7250E 06 41 2 72532E 06 18 2 72527E 06 41 2 5920 06 42 2 59240E 06 42 2 4660E 06 43 2 46597E 06 19 2 46592 06 14 2 46597 06 43 2 3850 06 44 2 38513E 06 44 2 3650 06 45 2 36533E 06 45 2 3460 06 46 2 34570E 06 46 2 3070E 06 47 2 30693E 06 47 2 2310 06 48 2 23130E 06 20 2 23126 06 15
82. 00E 01 7 80000E 01 7 05000E 01 6 25000E 01 5 40000E 01 5 00000E 01 4 85000E 01 4 33000E 01 4 00000E 01 3 91000E 01 3 50000E 01 3 20000E 01 3 14500E 01 3 00000E 01 2 80000E 01 2 48000E 01 2 20000E 01 1 89000E 01 1 80000E 01 1 60000E 01 1 40000E 01 1 34000E 01 1 15000E 01 1 00001E 01 9 50000E 02 8 00000E 02 7 70000E 02 6 70000E 02 5 80000E 02 5 00000E 02 4 20000E 02 3 50000E 02 33 34 35 36 37 38 39 40 41 42 43 44 45 46 47 48 49 50 51 52 53 54 55 56 57 58 59 60 61 62 63 1 30000 00 1 15000 00 1 12300 00 1 09700 00 1 07100 00 1 04500 00 1 02000 00 9 96000 01 9 72000E 01 9 50000E 01 9 10000E 01 8 50000E 01 7 80000E 01 6 25000E 01 5 00000E 01 4 00000E 01 3 50000E 01 3 20000E 01 3 00000E 01 2 80000E 01 2 50000E 01 2 20000E 01 1 80000E 01 1 40000E 01 1 00000E 01 8 00000E 02 6 70000E 02 5 80000E 02 5 00000E 02 4 20000E 02 3 50000E 02 User Manuat Issue 1 Feb 2007 FISPACT 89 TRIPOLI VITAMIN J GAM II XMAS WIMS 307 3 2380E 02 308 3 2000E 02 309 3 0000E 02 165 3 00000E 02 64 13 00000 02 166 2 50000E 02 65 2 50000 02 310 2 0000E 02 167 2 00000 02 66 12 00000 02 311 1 5000 02 168 1 50000E 02 67 1 50000E 02 312 1 0000 02 169 1 00000E 02 68 1 00000E 02 170 6 90000 03 313 5 5000E 03 171 5 00000E 03 69 5
83. 04 124 2 18749E 04 142 2 1130E 04 143 2 0540 04 144 1 9950E 04 145 1 9310E 04 125 1 93045E 04 57 1 93042E 04 146 1 7780E 04 147 1 6620E 04 43 1 66156E 04 148 1 5850E 04 149 1 5030E 04 126 1 50344E 150 1 3830E 04 151 1 2730E 04 152 1 1710 04 127 1 17088 04 59 1 17086E 04 153 1 1140E 04 45 1 11378E 04 128 1 05946E 04 04 04 04 04 58 1 50341 04 44 1 50344E4 04 14 1 50300E 04 154 1 0080E 04 155 9 1190 03 129 9 11882E 03 60 9 11866E 03 46 9 11882E 03 15 9 11800E 03 User Manual Issue 1 Feb 2007 UKAEA Fusion 86 FISPACT 156 8 2510E 03 157 7 4660E 03 47 7 46586E 03 158 7 1020E 03 130 7 10174E 03 61 7 10162E 03 159 6 2670E 03 160 5 5310E 03 131 5 53084E 03 62 5 53075 03 48 5 53084 03 16 5 53000 03 161 5 0040E 03 49 5 00451 03 162 4 6430E 03 163 4 3070E 03 132 14 30742E403 63 4 30735E 03 164 3 9810E 03 165 3 7070E 03 133 3 70744E 03 166 3 5480E 03 50 3 52662E 03 17 3 51900E 03 167 3 3550E 03 134 3 35463E 03 64 3 35457 03 51 3 35463 03 168 oo 169 3 0350 03 135 3 03539E 03 170 2 8180E 03 171 2 7470E 03 136 2 74654E 03 172 2 6610E 03 173 2 6130E 03 137 2 61259E 03 65 2 61254E 03 174 2 4850E 03 138 2 48517E 03 175 2 3710E 03 176 2 2490E 03 139
84. 07 19 1 69018 18 3 71390 15 1 37931E 11 5 12262E 06 1 90249E 02 1 22310 10 s stable 8 43653E 19 7 11750E 19 5 06589E 19 3 60565E 19 1 30007E 19 1 69018E 18 3 71392E 15 1 37932E 11 5 12266E 06 1 90251E 02 1 56347E 19 2 88250E 19 4 93411E 19 6 39435E 19 8 69993E 19 9 83098E 19 9 99963E 19 1 00000E 20 1 00000E 20 1 00000E 20 1 56347E 19 2 88250E 19 4 93410E 19 6 39434E 19 8 69997E 19 9 83135E 19 1 00002E 20 1 00002E 20 1 00002E 20 1 00002E 20 4 78109E 16 4 03358E 16 2 87091E 16 2 04337E 16 7 36766E 15 9 57846E 14 2 10472E 12 7 81673E 07 2 90306E 03 1 07817E 01 4 781094E 16 4 033584E 16 2 870905E 16 2 043368E 16 7 367663E 15 9 578471E 14 2 104725E 12 7 816770E 07 2 903082E 03 1 078180E 01 Max diff 0 00105 0 00570 0 00093 26m A decay test 45 26 6 345 s gt stable Decay times s Na Al 26m N Al 26m Na Mg 26 26 Aq Total Ax Total 2 00000 00 5 00000 00 1 50000E 01 2 50000E 01 5 00000E 01 1 00000E 02 1 50000E 02 2 50000E 02 8 03735 19 5 79138 19 1 94243 19 6 51491E 18 4 24441 17 1 80150 15 7 64630 12 1 37748 08 8 03735 19 5 79138 19 1 94243E 19 6 51492E 18 4 24441E 17 1 80151E 15 7 64634E 12 1 37749E 08 1 96265E 19 4 20862E 19 8 05757 19 9 34851 19 9 95756 19 9 99982 19 1 00000 20 1 00000 20 1 96265 19 4 20861 19 8 05767 19 9
85. 1 0000E 01 302 1 0000E 01 211 1 0000E 01 175 1 0000E 01 nt36 n 351 1 1000E 04 315 1 1000E 04 352 1 0000E 05 316 1 0000E 05 212 1 0000E 05 176 1 0000E 05 User Manuat Issue 1 Feb 2007 UKAEA Fusion 90 FISPACT Weighting spectra UKAEA Fusion Different weighting spectra are used depending upon which group structure 15 required and for which application fusion or fission the calculation needs to be performed They are all generated at a temperature of 300 K The weighting spectra used to generate fission relevant libraries in the WIMS XMAS and TRIPOLI group format from EAF point wise data are as follows Energy range Micro flux weighting 1 010 0 2 eV Maxwellian T 0 0253 eV 0 2 eV 0 82085 MeV 0 82085 MeV Ema Maxwellian fission spectrum T 1 3539 MeV The weighting spectra used to generate fusion relevant libraries in the VITAMIN J GAM II and TRIPOLI group format from EAF point wise data are as follows Micro flux weighting 1 0 10 0 414 eV Maxwellian T 0 0253 eV 0 414 eV 12 52 MeV VE 12 52 15 68 MeV Velocity exponential fusion peak 14 07 MeV kT 0 025 MeV 15 68 19 64 MeV A flat weighting spectrum is used to generate special purpose libraries in the XMAS VITAMIN J VITAMIN J TRIPOLI and TRIPOLI group format from EAF point wise data Such libraries should be used to model cases where the neutron field is not produced by fission or DT
86. 100 characters only the first 22 will be used as the spectrum identifier so these should provide an unambiguous description User Manuat Issue 1 Feb 2007 UKAEA Fusion 34 FISPACT HALF HAZARDS UKAEA Fusion The conversion is done on an equal flux per unit lethargy basis e g if one of the input groups is split into two or more groups in the converted spectrum then the fraction of neutrons in each output group is determined by the ratio of each lethargy interval of the output structure to the total lethargy interval of the input structure There is a restriction on the number of arbitrary energy groups this MUST be greater than 2 and no more than 400 An example of the use of this code word follows GRPCONVERT 99 172 In this case a spectrum in 99 groups is converted into the XMAS 172 group structure This code word causes the half life of each nuclide to be printed in the output The units are seconds but if the nuclide is stable then the word Stable is printed If this code word is not used then an indication of the stable nuclides in the output can be seen in the flags section to the right of the nuclide identifier This code word causes data on potential ingestion and inhalation doses to be input and the dose due to individual nuclides to be printed in the output See Appendix 6 for more details of the data stored in the library User Manual Issue 1 Feb 2007 FISPACT 35 IR
87. 172 VITAMIN J 175 VITAMIN J 211 TRIPOLI 315 and TRIPOLF 351 formats are listed in Appendix 1 where details of the micro flux weighting spectra are also given Note that three choices of weighting spectra are available for the TRIPOLI format This is necessary because of the very different neutron spectra found in fission and fusion applications in addition a flat weighting library is available for other applications Two choices of weighting spectra are available for the XMAS format WIMS weighting and a flat weighting Two choices of weighting spectra are available for the VITAMIN J format standard VITAMIN J weighting and a flat weighting A single weighting spectrum is available for WIMS VITAMIN J TRIPOLI WIMS XMAS and TRIPOLI are appropriate for fission applications GAM II TRIPOLI and VITAMIN J are appropriate for fusion applications while VITAMIN J and TRIPOLIH are appropriate for IFMIF For other applications such as neutron sources flat weighting should be used EAF D GXS 2007 One group cross section library is available for the deuteron induced library that can be used as input to FISPACT The group boundaries of the VITAMIN J 211 formats are listed in Appendix 1 Only a flat weighting library is available which is suitable for accelerator applications EAF P GXS 2007 One group cross section library is available for the proton induced library that can be used as input to FISPACT The
88. 18 FISPACT 49 meyax 42 xe Exe jeS K m jeS NO 9 18 The above derivation 15 correct so long as only cross section uncertainties are considered In FISPACT 97 the facility to consider half life uncertainties was included The derivation therefore needs to be extended to do this requires additional justification of the whole of the pathway methodology this is included in the current Appendix for completeness Note that this extends the original theoretical development given in Appendix 1 of reference 39 Pathways containing 2 reactions only UKAEA Fusion Consider the 2 link pathway both reactions shown in Figure A9 1 where it is assumed that the final nuclide neither reacts nor decays This constraint is removed in the treatment of a later section Note that it is assumed that there is no cross section for a parent nuclide to be reformed from its daughter this is true in the low burn up limit Ni N3 e Figure A9 1 A 2 link pathway consisting of reactions only The symbols used in Figure 1 are defined below the label i can be used for nuclides 1 2 or 3 N Number of atoms of nuclide Decay constant s o Cross section of a pathway reaction cm Sum of all cross sections of a target excluding the pathway reaction cm The differential equation satisfied by nuclide 1 is given in equation A9 19 User Manuat Issue 1 Feb 2007 2 FISPACT 119
89. 183448E 09 Max diff 12562 0 01811 This case shows a significant difference 1 26 between the FISPACT and analytical calculations for at 5 s decay time However note that the number of atoms is very small and that with increasing decay time the differences get less significant 5 Fe 1 5480 10 s gt 5 1060 10 s gt Mn 1 16680 10 s stable IT g Decay times s N Fe s3m N Fe 53m N Fe 53 N Fe 53 Na Mn 53 N Mn 53 7 04402 19 5 84310 19 4 46648 19 2 60981 19 6 81110 18 1 77757 18 4 63910 17 1 21072 17 3 15974 16 2 15213 15 1 46584E 14 9 98395 12 9 96793 05 0 00000 00 0 00000 00 0 00000 00 0 00000 00 0 00000 00 7 64402E 19 5 84310E 19 4 46648E 19 2 60981E 19 6 81110E 18 1 77757E 18 4 63911E 17 1 21072E 17 3 15975E 16 2 15213E 15 1 46584E 14 9 98399E 12 9 96801E 05 0 00000E 00 0 00000E 00 0 00000E 00 0 00000E 00 0 00000E 00 2 25847E 19 3 80818E 19 4 82994E 19 5 80480E 19 5 37789E 19 3 97422E 19 2 74793E 19 1 85560E 19 1 24189E 19 5 51677E 18 2 44430E 18 1 08255E 18 8 16646E 15 6 16046E 13 2 64455E 07 0 00000E 00 0 00000E 00 0 00000E 00 2 25847E 19 3 80818E 19 4 82994E 19 5 80480E 19 5 37789E 19 3 97420E 19 2 74791E 19 1 85559E 19 1 24191E 19 5 51686E 18 2 44438E 18 1 08259E 18 8 16603E 15 6 16014E 13 2 64442E 07 0 00000E 00 0 00000E 00 0 00000E 00 9 75146E 17 3 48717E
90. 2 23130E 06 4 12 23100 06 48 2 1220 06 49 2 12248E 06 49 2 0190 06 50 12 01897 06 21 12 01893 061 16 2 01897 06 50 1 9210 06 51 1 92050E 06 51 1 8270E 06 52 1 82684 06 22 1 82680E 06 52 1 7380E 06 53 1 73774E 06 53 1 6530 06 54 1 65299E 06 23 1 65296 06 17 1 65299E 06 54 1 5720E 06 55 1 57237E 06 55 1 4960E 06 56 1 49569E 06 24 1 49566E 06 56 1 4230E 06 57 1 42274E 06 57 1 3530E 06 58 1 1 35335E 06 25 1 35333 06 18 1 35335E4 58 1 2870E 06 59 1 28735E 06 59 1 2250 06 60 1 22456E 06 26 1 22454 06 19 1 22456E 06 60 1 1650 06 61 1 16484E 06 61 1 1080E 06 62 1 10803E 06 27 1 10801 06 20 1 10803E 06 62 1 0030E 06 63 1 00259E 06 28 1 00257E 06 21 1 00259E 06 63 9 6160E 05 64 9 61672E 05 64 9 0720 05 65 9 07180E 05 29 9 07164 05 22 9 07180E4 65 8 6290 05 66 8 62936E 05 66 8 2090 05 67 8 20850E 05 30 8 20836 05 23 8 20850E 05 6 8 21000 05 67 7 8080 05 68 7 80817 05 68 7 4270 05 69 7 42736 05 31 7 42723E 05 69 7 0650 05 70 7 06512E 05 70 6 7210E 05 71 6 72055E 05 32 6 72044E 05 71 6 3930 05 72 6 39279E 05 72 6 0810 05 73 6 08101E 05 33 6 08090 05 24 6 08101E 05 73 5 7840 05 74 5 78443 05 74 5 5020 05 75 5 50232E 05 34 5 50223 05 25 5 50232 05 7 5 00000E 05 75 5 2340 05 76 5 23397 05 77 14 97871 05 35
91. 20 90 48 0 27 9 25 11 22 98977 0 971 23 100 0 12 24 305 1 738 24 78 99 10 0 11 01 13 26 981539 2 6989 27 100 0 14 28 0855 2 33 28 92 23 4 683 3 087 15 30 973761 1 82 31 100 0 16 32 065 2 07 32 95 02 0 75 4 21 0 0 02 17 35 453 1 8956 35 75 77 0 24 23 18 39 948 1 6504 36 0 3365 0 0 0632 0 99 6003 19 39 0983 0 862 39 93 2581 0 0117 6 7302 20 40 078 1 55 40 96 94 0 0 647 0 135 2 09 0 0 004 0 0 187 21 44 95591 2 989 45 100 0 22 47 867 4 54 46 8 25 7 44 73 72 5 41 5 18 23 50 9415 6 11 50 0 250 99 750 24 51 9961 7 19 50 4 345 0 83 789 9 501 2 365 25 54 938049 7 44 55 100 0 26 55 845 7 874 54 5 845 0 91 754 2 119 0 282 27 58 9332 8 9 59 100 0 28 58 6934 8 902 58 68 077 0 26 223 1 14 3 634 0 0 926 29 63 546 8 96 63 69 17 0 30 83 30 65 39 7 133 64 48 63 0 27 9 4 1 18 75 0 0 62 31 69 723 5 904 69 60 108 0 39 892 32 72 64 5 323 70 20 37 0 27 31 7 76 36 73 0 7 83 33 74 9216 5 73 75 100 0 34 78 96 4 79 74 0 89 0 9 37 7 63 23 77 0 49 61 0 8 73 User Manuat Issue 1 Feb 2007 UKAEA Fusion 190 FISPACT Atomic Atomic Density Mass of first Abundance number weight g stable isotope 35 79 904 3 12 79 50 69 0 49 31 36 83 80 2 6021 78 0 35 0 2 28 0 11 58 11 49 57 0 0 17 3 37 85 4678 1 532 85 72 17 0 27 83 38 87 62 2 54 84 0 56 0 9 86 7 0 82 58 39 88 90585 4 469 89 100 0
92. 2800 06 3 45600 06 5 18400 06 6 91200 06 8 64000 06 1 72800 07 2 59200 07 4 32000 07 58mCo 3 21840 10 s Co 6 12230 10 s gt Fe stable IT 8 56358E 19 6 78639E 19 4 60551 19 1 55548E 19 8 29197 11 6 87568 03 0 00000 00 0 00000 00 0 00000 00 0 00000 00 0 00000 00 0 00000 00 0 00000 00 8 56358 19 6 78639E 19 4 60551 19 1 55549 19 8 29201 11 6 87575 03 0 00000 00 0 00000 00 0 00000 00 0 00000 00 0 00000 00 0 00000 00 0 00000 00 1 43582 19 3 21013 19 5 38211 19 8 39128 19 9 11605 19 8 26655 19 6 79766 19 5 58978 19 4 59653 19 3 77977 19 1 42115 19 5 34340 18 7 55387 17 1 43582 19 3 21013 19 5 38211 19 8 39129 19 9 11608 19 8 26058 19 6 79769E 19 5 58980E 19 4 59655 19 3 77978 19 1 42116 19 5 34342 18 7 55391 17 6 00415 15 3 48311 16 1 23822 17 5 32307 17 8 83951E 18 1 73345 19 3 20234 19 4 41022 19 5 40347 19 6 22023 19 8 57885 19 9 46566E 19 9 92446E 19 6 00415E 15 3 48312E 16 1 23822E 17 5 32308E 17 8 83964E 18 1 73347E 19 3 20236E 19 4 41024E 19 5 40350E 19 6 22026E 19 8 57893E 19 9 46576E 19 9 92468E 19 Max diff Go 000102 0 00070 000222 Decay times s Ax Total Ar Total 7 20000 03 1 80000 04 3 60000 04 8 64000E 04 8 64000E 05 1 84597E 15 1 46522E 15 9 97982E 14 3 44505
93. 33 TIME 10 TIME 10 TIME 25 TIME 50 TIME 50 TIME 100 END END ATOMS ATOMS ATOMS ATOMS ATOMS ATOMS ATOMS ATOMS test 16 NOHEAD MONITOR 1 EAFV 4 AINP FISPACT DECAY OF Y 89M FUEL 1 Y89M 1 0E20 DENSITY 4 469 MIND 1 0 GRAPH 10 11 UNCERT 0 LEVEL 100 1 FLUX 0 0 ZERO TIME 5 0 ATOMS TIME 5 0 ATOMS TIME 5 0 ATOMS TIME 5 0 ATOMS TIME 30 0 ATOMS TIME 50 0 ATOMS TIME 100 0 ATOMS TIME 100 0 ATOMS END END User Manuat Issue 1 Feb 2007 UKAEA Fusion 220 FISPACT test d7 NOHEAD MONITOR 1 EAFV 4 AINP FISPACT DECAY OF CO 58M FUEL 1 CO58M 1 0E20 DENSITY 8 90 MIND 1 0 GRAPH 10 11 UNCERT 0 LEVEL 100 1 FLUX 0 0 ZERO TIME 2 0 HOURS ATOMS TIME 3 0 HOURS ATOMS TIME 5 0 HOURS ATOMS TIME 14 0 HOURS ATOMS TIME 9 0 DAYS ATOMS TIME 10 0 DAYS ATOMS TIME 20 0 DAYS ATOMS TIME 20 0 DAYS ATOMS TIME 20 0 DAYS ATOMS TIME 20 0 DAYS ATOMS TIME 100 0 DAYS ATOMS TIME 100 0 DAYS ATOMS TIME 200 0 DAYS ATOMS END END test d8 NOHEAD MONITOR 1 EAFV 4 AINP FISPACT DECAY 5 39 FUEL 1 539 1 0 20 DENSITY 2 07 MIND 1 0 GRAPH 10 11 UNCERT 0 LEVEL 100 1 FLUX 0 0 ZERO TIME 5 0 ATOMS TIME 5 0 ATOMS TIME 5 0 ATOMS TIME 35 0 ATOMS TIME 50 0 ATOMS TIME 400 0 ATOMS TIME 500 0 ATOMS TIME 1000 0 ATOMS TIME 3000 0 ATOMS TIME 5000 0 ATOMS TIME 40000 0 ATOMS TIME 50000 0 ATOMS END
94. 36297E 10 6 29958E 05 5 94058E 05 5 94058E 05 5 94058E 05 7 40574 Decay times 5 NA Cr 53 Ni Cr 53 AA Total Ar Total 3 729348 17 3 133328 17 2 655624 17 1 956602E 17 1 035036E 17 6 190978E 16 3 938055 16 2 573208 16 1 700059 16 7 498858 15 3 318944 15 1 469682 15 1 108551 13 8 362547 10 6 299463 05 5 940479 05 5 940479 05 5 940479 05 0 00532 FISPACT This case shows a significant difference 7 41 90 between the FISPACT and analytical calculations for Cr at 60 s decay time However note that the number of atoms is very small and that with increasing decay time the differences get less significant 230 ph decay test 410 20Th 2 37944 101 s gt 2Ra 5 04922 10 s gt 3 30480 10 s 2 1 8300 102 s 0 Decay times s Na Th 230 230 Na Ra 226 N amp Ra 226 Na Rn 222 Ni Rn222 1 00000 12 2 00000 12 3 00000 12 4 00000E 12 5 00000E 12 6 00000E 12 1 00000E 13 5 00000E 13 7 47286E 19 5 58437E 19 4 17312E 19 3 11852E 19 2 33043E 19 1 74149E 19 5 43088E 18 4 72446E 13 7 47286E 19 5 58437E 19 4 17312E 19 3 11852E 19 2 33043E 19 1 74150E 19 5 43089E 18 4 72448E 13 1 62013E 18 1 21071E 18 9 04744E 17 6 76103E 17 5 05242E 17 3 77561E 17 1 17743E 17 1 02427E 12 1 62013E 18 1 21071E 18 9 04744E 17 6 76103E 17 5 05243E 17 3 77561E 17 1 17743E 17 1 02428E 12 1 06041E 13 7 9242
95. 39E 11 Max diff 7 000028 000046 004641 Decay times s NA Mo 95 95 Total 5 00000 04 3 54592 15 3 54591 15 1 26065E 13 1 260649E 13 1 00000 05 1 41038 16 1 41038 16 1 26802 13 1 268019 13 5 00000 05 3 36946E 17 3 36946 17 1 31641 13 1 316405E 13 1 00000 06 1 27317E 18 1 27318E 18 1 35644E 13 1 356442E 13 2 00000E 06 4 54657E 18 4 54656E 18 1 38940E 13 1 389400E 13 3 00000E 06 9 14630E 18 9 14628E 18 1 37823E 13 1 378232E 13 4 00000 06 1 45655 19 1 45655 19 1 33705 13 1 337051E 13 5 00000 06 2 04289 19 2 04288 19 1 27588E 13 1 275883 13 6 00000 06 2 64621 19 2 64621 19 1 20219 13 1 202182 13 8 00000 06 3 83121 19 3 83121 19 1 03779 13 1 037787E 13 1 00000 07 4 91729 19 4 91729E 19 8 72274 12 8 722740 12 5 00000 07 9 95818 19 9 95820 19 7 63643E 10 7 636481E 10 1 00000 08 9 99992 19 9 99994 19 1 45816E 08 1 458174E 08 Max diff 5 00009 0 00096 The agreement between the FISPACT and the analytical calculations for this demonstrates the validity of the method used in FISPACT reasonably complex decay chain Reaction tests It is much more difficult to make direct comparison of FISPACT calculations with analytical calculations for cases that model a neutron irradiation This is because the non zero cross sections for
96. 4 FISPACT Pathways containing an arbitrary number of decays The general case of n 1 decay links can be solved by replacing by in equation A9 39 This is shown in equation A9 48 n l No 4 Sev D A 7 p A9 48 Limits in arbitrary pathways UKAEA Fusion In the cases of two reactions or two decays the long lived limits are shown in equations A9 26 and A9 46 and the short lived limits in equations A9 27 and A9 47 In the general cases it should be possible using the same approach to find the limiting forms of equations A9 39 and A9 48 However the algebra is very difficult and a set of identities similar to equation A9 44 is required It is much simpler to apply the limits to the original equation A9 28 and solve this In the long lived limit Ajt lt lt 1 and equation A9 28 can be simplified to give equation A9 49 dN 1 ES ON e TIN Ras Ee as esl atti A9 49 E This can be integrated step by step starting with the limiting form of equation A9 20 namely Ni f To find the general solution a proof by induction can be used Suppose that the form of N t is given by equation A9 50 Then integrating right hand side of equation A9 49 gives M as shown in equation 9 51 Comparing this with the form of equation A9 50 it can be seen that M 1 2 Because this form is true for n 2 see equation A9 26 this argument shows it will be true for all n2 2
97. 4 18 91 25 51 24 9 28 18 67 164 93032 8 795 165 100 0 68 167 259 9 066 162 0 139 0 1 601 0 33 503 22 869 26 978 0 14 91 69 168 93421 9 321 169 100 0 UKAEA Fusion User Manual Issue 1 Feb 2007 FISPACT 191 Atomic Atomic Density Mass of first Abundance number weight g stable isotope 70 173 04 6 903 168 0 13 0 3 04 14 28 21 83 16 13 31 83 0 12 76 71 174 967 9 841 175 97 41 2 59 72 178 49 13 31 174 0 16 0 5 26 18 6 27 28 13 62 35 08 73 180 9479 16 654 180 0 012 99 988 74 183 84 19 3 180 0 12 0 26 5 14 31 30 64 0 28 43 75 186 207 21 02 185 37 4 0 62 6 76 190 23 22 57 184 0 02 0 1 59 1 6 13 29 16 21 26 36 0 40 93 77 192 217 22 42 191 37 3 0 62 7 78 195 08 21 45 190 0 014 0 0 782 0 32 967 33 832 25 242 0 7 163 79 196 96655 19 3 197 100 0 80 200 59 13 546 196 0 15 0 9 97 16 87 23 1 13 18 29 86 0 6 87 81 204 3833 11 85 203 29 524 0 70 476 82 207 2 11 35 204 1 4 0 24 1 22 1 52 4 83 208 98038 9 747 209 100 0 84 0 9 32 85 0 5 0 86 0 5 0 87 0 5 0 88 0 5 0 89 0 10 07 90 232 03805 11 72 232 100 0 91 0 15 37 92 238 02891 18 95 234 0 0054 0 7204 0 0 99 2742 User Manual Issue 1 Feb 2007 UKAEA Fusion 192 FISPACT Appendix 17 FISPACT modifications Changes made to the FISPACT source code in response to problems or to add new features following version 4 0 a
98. 4 and 22 group formats Group Energy range MeV Group Energy range MeV number 24 groups number 22 groups HN BW L 12 NO N N N NO RR A U Ne DO WAN CO Q Q 0 00 0 01 0 01 0 02 0 02 0 05 0 05 0 10 0 10 0 20 0 20 0 30 0 30 0 40 0 40 0 60 0 60 0 80 0 80 1 00 1 00 1 22 1 22 1 44 1 44 1 66 1 66 2 00 2 00 2 50 2 50 3 00 3 00 4 00 4 00 5 00 5 00 6 50 6 50 8 00 8 00 10 00 10 00 12 00 12 00 14 00 14 00 User Manual Issue 1 Feb 2007 00 10 Q BW NO N OO O tan 0 00 0 01 0 01 0 10 0 10 0 20 0 20 0 40 0 40 1 00 1 00 1 50 1 50 2 00 2 00 2 50 2 50 3 00 3 00 3 50 3 50 4 00 4 00 4 50 4 50 5 00 5 00 5 50 5 50 6 00 6 00 6 50 6 50 7 00 7 00 7 50 7 50 8 00 8 00 10 00 10 00 12 00 12 00 14 00 UKAEA Fusion 132 FISPACT Appendix 11 Error messages UKAEA Fusion During the course of a FISPACT run the program can terminate prematurely if mistakes are made in the syntax of the code words in the INPUT file If this happens then the OUTPUT file will contain one of the following error messages If the message ends with a code word in square brackets then the user should consult the earlier sections to check the syntax and the allowed values of the parame
99. 49 0 438 0 445 0 399 0 398 0 343 0 106 0 107 0 007 1 402 3 13 808 0 002 0 103 0 323 0 023 Quantity Display Activity Clearance Value C Beta heat Nuclide labels included copy Percentage of total C Dose rate Gamma heat Inventory uncertainty Copy to Clipboard Refresh data Clear data Close Figure A15 3 The summary of dominant nuclides window Running FISPACT UKAEA Fusion FISPACT can be run directly from the EASY User Interface either in a visible or minimised command window This is a very convenient way to carry out the calculation prior to analysing the output or plotting graphs If there is an error message similar to the one shown in Figure A154 then the FISPACT run has terminated abnormally This is usually due to an incorrect file being specified in the FILES file To aid in correcting this fault the Troubleshoot FISPACT item on the Run menu should be clicked This brings up the dialog shown in Figure A15 5 User Manuat Issue 1 Feb 2007 FISPACT 183 e fispact y_2007 fisp20070_exe x Error File does not exist 1c0a527c do_io_erorlintint lt ptr char int B0 00b1 1c0a533d error handler intintint ptr char HBO 0125 1 0 5472 10 error enum amp E rrorMessage 40024 1 0 8486 10 open bcf 00401000 MAINS 03ae D047a677 SALFStart 06b2 Figure A15 4 An error b
100. 5 18 1 615115 16 2 216282E 14 Max diff 7 0 00735 9000610 0 00713 Be decay test_d3 Be 5 04922 10 s B stable B7 Decay times s NA Be 10 N Be 10 N B 10 N B 10 A Total 1 57788E 13 3 15576E 13 4 73364E 13 6 31152E 13 1 57788E 14 3 15576 14 6 31152E 14 9 46728E 14 1 57788E 15 8 05245E 19 6 48420E 19 5 22137 19 4 20449 19 1 14626 19 1 31391 18 1 72635 16 2 26826 14 3 91580 10 8 05245 19 6 48420 19 5 22137 19 4 20449 19 1 14626 19 1 31391 18 1 72635 16 2 26826 14 3 91582E 10 1 94755 19 3 51580 19 4 77863 19 5 79551 19 8 85374 19 9 86861 19 9 99827 19 9 99998 19 1 00000E 20 1 94755 19 3 51580 19 4 77862 19 5 79551 19 8 85391 19 9 86915 19 9 99913 19 9 99999 19 9 99999 19 1 10543 06 8 90138E 05 7 16780 05 5 77184 05 1 57356 05 1 80370 04 2 36990E 02 3 11382E 00 5 37553E 04 1 105425E 06 8 901383E 05 7 167798E 05 5 771836E 05 1 573560E 05 1 803705E 04 2 369900E 02 3 113828E 00 5 375559 04 Max diff 00051 2006 000054 UKAEA Fusion User Manuat Issue 1 Feb 2007 FISPACT 209 decay test 94 3 00000 02 6 00000 02 1 20000 03 1 80000 03 3 60000 03 7 20000 03 1 80000 04 3 60000 04 5 40000 04 7 20000 04 8 43653 19 7 11750 19 5 06589 19 3 60565 19 1 300
101. 5 33 1 22773E 05 102 1 1680E 05 104 1 16786E 05 103 1 1110E 05 105 1 11090E 05 50 1 11088 05 34 1 11090 05 10 1 11000E 05 104 9 8040E 04 106 9 80365E 04 105 8 6520E 04 107 8 65170E 04 51 8 65155E 04 106 fa 108 pore 107 8 2300E 04 35 8 22975E 04 108 7 9500E 04 109 7 94987E 04 109 7 4990 04 110 7 2000E 04 110 7 20245E 04 111 6 7380E 04 111 6 73795E 04 52 6 73783E 04 36 6 73795E 04 11 6 73400E 04 112 6 1730E 04 113 5 6560E 04 112 5 65622E 04 114 5 5170 04 37 5 51656E 04 115 5 2480E 04 113 15 24752E404 53 15 24743E 04 116 4 9390 04 117 4 6310E 04 114 4 63092E 04 118 4 3590E 04 119 4 0870 04 115 4 08677 04 54 4 08670 04 38 4 08677E 04 12 4 08500E 04 120 3 6980E 04 39 3 69786E 04 121 3 4310E 04 116 3 43067E 04 122 3 1830E 04 117 3 18278E 04 55 3 18272E404 123 3 1620E 04 124 3 0730E 04 125 2 9850 04 126 2 0010E 04 40 2 92830E 04 127 2 8500E 04 118 2 85011E 04 128 2 8180E 04 129 2 7380E 04 41 2 73944E 04 130 2 7000E 04 119 2 70001E 04 131 2 6610E 04 132 2 6060E 04 120 2 60584E 133 2 5850E 04 134 2 5120 04 135 2 4790E 04 121 2 47875 04 56 2 47871 04 42 2 47875 04 13 2 47800 04 136 2 4410E 04 137 2 4180E 04 122 2 41755 4 138 2 3580 04 123 2 35786E 139 2 3040E 04 140 2 2390E 04 141 2 1870E
102. 5 ZAI value in a coded form for Fm COL100 COL172 and COL175 isotopes 18 10 95 Format change from I6 to I7 Changes made in A2INP 616 618 required when reading A2 and HAZINP and ENDFPR Hazard files containing Fm isotopes and when outputting error data 18 10 95 Need to initialise XSECT array Change made in COL069 619 622 COL100 COL172 and COL175 23 10 95 Array overflow in INTERP More bremsstrahlung candidates 626 increase array dimensions from 400 to 450 particle too small constants redefined 14 11 95 For problems involving pulsed This feature added by introducing 637 645 irradiation it would be useful to the new code words PULSE and include a loop facility in the ENDPULSE FISPACT input file 12 2 96 For problems involving actinides it This feature added by introducing 646 648 653 would be useful to be able to the new code word FISYIELD 655 switch off fission product production from specified actinides 12 2 96 Problem found when considering Change a constant from 20 to 30 649 652 neutrons with only low energy e g in COL069 COL100 COL172 D D spectrum Cross section and COL 175 uncertainty files get out of step because so many of the collapsed cross sections are zero 656 15 3 96 Problem with pathways for some Incorrect use of the variable actinide targets LIMITI in CHAINP printed in the PRINTLIB output UTP 23 5 96 Printout of pseudo cross sections Changes so that str
103. 5 2 64867 5 2 64900 5 5 73356 5 5 78807E 5 5 78648 5 5 78709 5 0 00000 0 0 00000 0 0 00000 0 0 00000 0 1 97377 5 1 48326 5 5 20231E 6 9 28193 7 2 92266 5 2 21577 5 8 28611 1 99212E 6 Use ext as material ident Data shown are for Totals Time values included in copy Total Nuclide to Clipboard data Clear data Database Dominant nuclides Close Pathways 1 Figure 15 1 main window of the EASY User Interface A FISPACT run is specified by means of the INPUT file The EASY User Interface allows existing input files to be opened and edited using the cut copy and paste tools and saved For the various categories of FISPACT runs listed below input files can be prepared easily by entering data in a series of dialog boxes e Collapse cross section library e Process decay data and prepare an ARRAY file e Output a readable form of the nuclear data libraries PRINTLIB e Inventory run e Generate pathways either using the code words PATHS or ROUTES User Manuat Issue 1 Feb 2007 UKAEA Fusion 180 FISPACT Connecting the various input and output streams to external files by means of the FILES file is also simplified by a dialog box Graph plotting UKAEA Fusion The code word GRAPH in a FISPACT input file does not physically plot a graph it only writes the relevant data required
104. 6 GHz d Collapse e 2 2 Write f Write g Printlib h Test 1 Test 2 Test 3 Test 4 Test 5 Test 6 Test 7 Test 8 Test 9 Test 10 Test 11 Test 12 Test 13 Test 14 Test 15 Test 16 Test 17 Test 18 Test 19 Test 20 Test 21 Test 22 Test 23 Test 24 Test 25 Test 26 Test 31 Q Q D HAN Q KWH p O ON 1Ut 1 1 gt Q NK HN ON L 1 NWA L WN Q NR NRK ON ONNA ON N RRB Be NF WRK IR User Manual Issue 1 Feb 2007 UKAEA Fusion 176 FISPACT ENEL pa XP acl 1 6 GHz a 3 4 GHz b 3 2 GHz c 2 66 GHz d 51 18 Test 32 15 15 Test 41 4 2 2 3 Test 42 12 3 3 4 Test 43 14 4 4 5 Test 44 4 2 2 1 Test 45 5 2 1 2 Test 51 25 8 7 9 Test 52 51 16 14 17 Test 60 7 4 3 4 Test 61 23 9 7 11 Test 70 4 2 2 2 Test 71 108 46 42 49 Test 72 380 171 154 183 Test 73 13 6 5 6 Test 74 12 5 5 6 Test 81 22 8 6 8 Test 82 15 4 5 5 Test 83 20 6 5 7 Test 84 5 2 1 2 Test 85 21 8 7 8 Test 86 53 32 19 35 Test 87 51 23 18 31 Test 88 14 7 6 7 Test 89 4 2 1 2 Test 90 21 8 7 7 Test 95 155 106 73 198 Test 96 5 2 1 2 Test 100 3 3 2 2 Test 101 94 98 79 98 Test 102 63 66 53 66 Test 103 13 8 7 8 Test 104 13 24 20 25 Test 181 3 2 1 1 Test 182 3 2 1 2 Test 183 10
105. 7 9 18148E23 48 9 28210E24 TI49 6 91755E23 TI50 6 79178E23 MIND 1 E5 FLUX 4 27701E14 ATOMS LEVEL 100 5 SENSITIVITY SIGMA 1E 10 2 1 48 SCA8 TI49 SCA8 5 48 ERROR 2 2 148 5 48 1 I49 SC49 1 IME 2 5 YEARS ATOMS END END 3 3 Test43 RRADIATION OF FE U EEF 172 FW 1 0 MW M2 MASS 1 0 2 FE 99 9999 0 0001 D 1 E5 PH51112345 L 1 00 l OMS LEVEL 100 10 TIME 2 5 YEARS HAZA HALF ATWO UNCERT 2 MINS ATOMS HOURS ATOMS DAYS ATOMS DAYS ATOMS YEARS ATOMS 000 YEARS ATOMS User Manuat Issue 1 Feb 2007 UKAEA Fusion 164 FISPACT Test44 NOHEAD AINP FISPACT IRRADIATION OF Ti EEF 172 FW 1 0 MW M2 MASS 1 0 1 TI 100 0 MIND 1 E5 WALL 1 00 ATOMS LEVEL 100 1 TIME 2 5 YEARS HAZA HALF ATWO DOMINANT 80 0 UNCERT 3 DOSE 1 ATOMS LEVEL 20 1 1 MINS ATOMS 1 HOURS ATOMS TIME 1 DAYS ATOMS 7 T DAYS ATOMS YEARS ATOMS Test45 NOHEAD AINP FISPACT IRRADIATION OF Ti EEF 172 FW 1 0 MW M2 MASS 1 0 1 TI 100 0 MIND 1 E5 WALL 1 00 ATOMS LEVEL 100 1 TIME 2 5 YEARS HAZA HALF ATWO lt lt Test case for comment gt gt GENERIC 0 UNCERT 3 DOSE 1 ATOMS LEVEL 20 1 FLUX 0 ZERO TIME 1 MINS ATOMS TIME 1 HOURS ATOMS lt lt Test case for comment gt gt 1 DAYS ATOMS IME 7 DAYS ATOMS IME 1 YEARS ATOMS END END UKAEA Fusion User Man
106. 8 0 00205 ee 1 62633E 02 1 62549E 02 0 00052 This pathway is particularly simple since all the nuclides apart from the final one are stable Only a single Bateman calculation is required and agreement even for the final nuclide is impressive Sulphur reactions test r2 34S stable n y 255 7 5600 10 s n y S stable n y 775 1 57788 10 s stable n 2n CI 3 155760 10 s B7 The correct half lives of 775 and Cl are 4 99 m and 1 256 s respectively but to ensure sensible numbers of atoms after the 1 y irradiation these half lives are set to 0 5 and 10 y respectively The values of the cross sections used for the calculations are included in the table below User Manuat Issue 1 Feb 2007 UKAEA Fusion 216 FISPACT Nuclide Cross section b 6 58663E 03 9 99794 19 9 99792E 19 0 00016 5 64830E 03 6 78468E 15 6 78464E 15 0 00056 5 83017E 03 8 62295E 11 8 62290E 11 0 00058 37 4 44595E 07 4 44598E 07 0 00067 CI 34 5 02308 10 5 02304 10 0 00080 CI 35 2 92143E 04 1 39982E 16 1 39981E 16 0 00071 This pathway is more complicated since 5 is radioactive and therefore the reactions and decays compete giving a branched pathway that requires two Bateman calculations Chromium reactions test r3 Cr n y Cr 2 39360 10 s n y O Cr n y 2 lL SV g In the abo
107. 8 31529E 00 241 7 9430E 00 242 7 5240E 00 88 7 52398E 00 243 7 0790E 00 244 6 4760E 00 163 6 47595E 00 89 6 47584E 00 245 6 1600 00 89 6 16012E 00 246 5 6230E 00 90 5 34643E 00 247 5 0430 00 164 5 04348E 00 90 5 04339 00 91 5 04348E 00 248 4 6700E 00 249 4 4700E 00 250 4 1290E 00 92 4 12925E 00 93 4 00000 00 28 4 00000E 00 251 3 9280E 00 165 3 92786E 00 9 3 92779E 00 252 3 3810E 00 94 3 38075E 00 95 3 30000E 00 29 3 30000E 00 253 3 0590E 00 166 3 05902E 00 92 3 05897E 00 254 2 7680E 00 96 2 76792E 00 97 2 72000E 00 98 2 60000 00 30 2 60000E 00 99 2 55000E 00 255 2 3720E 00 167 2 38237E 00 93 2 38233E 00 256 2 3600E 00 100 2 36000E 00 257 2 1300E 00 101 2 13000E 00 102 2 10000 00 31 2 10000E 00 258 2 0200E 00 103 2 02000E 00 259 1 9300 00 104 1 93000E 00 260 1 8550E 00 168 1 85539E 00 94 1 85536E 00 261 1 8400E 00 105 1 84000E 00 262 1 7550 00 106 1 75500E 00 263 1 6700 00 107 1 67000E 00 264 1 5900 00 108 1 59000 00 265 1 5100 00 109 1 50000 00 32 1 50000E 00 110 1 47500 00 266 1 4450E 00 169 1 44498E 00 95 1 44495 00 111 1 44498E 00 User Manuat Issue 1 Feb 2007 UKAEA Fusion 88 FISPACT 267 268 269 270 271 272 273 274 215 276 271 278 279 280 281 282 283 284 285
108. 8E 12 5 92170E 12 4 42521E 12 3 30690E 12 2 47120E 12 7 70648E 11 6 70405E 06 1 06041E 13 7 92433E 12 5 92174E 12 4 42524E 12 3 30692E 12 2 47122 12 7 70653E 11 6 70408E 06 Max diff 0007 00 000081 1 00000E 12 2 00000E 12 3 00000E 12 4 00000E 12 5 00000E 12 6 00000E 12 1 00000E 13 5 00000E 13 UKAEA Fusion 5 87189 09 4 38799 09 3 27908 09 2 45041 09 1 83116 09 1 36840 09 4 26738 08 3 71230 03 5 88199 09 4 39554 09 3 28472 09 2 45043 09 1 83117 09 1 36841 09 4 26741 08 3 71232E 03 8 84916E 07 6 61287E 07 4 94170E 07 3 69287E 07 2 75963E 07 2 06223E 07 6 43111E 06 5 59458E 01 Decay times s 218 218 Ax Total Ar Total 8 8530 07 6 6160E 07 4 9440E 07 3 6927E 07 2 7597 07 2 0622 07 6 4300E 06 5 5940E 01 Max diff 01706 0 04733 User Manuat Issue 1 Feb 2007 FISPACT 213 Cu decay test_d11 4 57272 10 s Zn stable 38 86 I Ni stable 61 14 Decay times s Na Cu 64 64 Na Zn 64 Ni Zn 64 NA Ni 64 Ni Ni 64 1 80000E 04 3 60000E 04 5 40000E 04 7 20000E 04 1 72800E 05 4 32000E 05 8 64000E 05 1 29600E 06 1 72800E 06 7 61207E 19 5 79436E 19 4 41071E 19 3 35746E 19 7 28497E 18 1 43242E 17 2 05182E 14 2 93907 11 4 20998 08 7 61207 19 5 79437 19 4 41071 19 3 35747 19 7 28498 18
109. 9 1 85168E 19 1 05541E 19 6 01558E 18 3 61873E 17 5 46654E 18 2 45033E 19 4 30025E 19 5 69688E 19 6 75129E 19 8 14832E 19 8 94458E 19 9 39844E 19 9 96381E 19 5 46654E 18 2 45033E 19 4 30025E 19 5 69687E 19 6 75128E 19 8 14827E 19 8 94451E 19 9 39835E 19 9 96421E 19 1 68401E 11 1 34489E 11 1 01535E 11 7 66553E 10 5 78722E 10 3 29857E 10 1 88011E 10 1 07161E 10 6 44635E 08 1 684008E 11 1 344889E 11 1 015346E 11 7 665525E 10 5 787216E 10 3 298567E 10 1 880099E 10 1 071609E 10 6 446356E 08 Max diff 7 00005 000401 1 2000059 decay test 42 Decay times s 1 00000E 02 5 00000E 02 1 00000E 01 5 00000E 01 1 00000 00 5 00000 00 1 00000E 01 1 50000E 01 Na He 6 9 91459E 19 9 58019E 19 9 17801E 19 6 51243E 19 4 24118E 19 1 37225E 18 1 88306E 16 2 58403E 14 He 8 0810 107 s Li stable Nr He 6 9 91459 19 9 58019 19 9 17801 19 6 51242 19 4 24116 19 1 37221 18 1 88297 16 2 58384 14 NA Li 6 8 54076E 17 4 19806E 18 8 21987 18 3 48757 19 5 75882 19 9 86278 19 9 99812 19 9 99997 19 14 6 8 54081 17 4 19808 18 8 21992E 18 3 48758E 19 5 75885E 19 9 86306E 19 9 99873E 19 9 99967E 19 A Total 8 50418E 19 8 21735 19 7 87239 19 5 58600 19 3 63785 19 1 17704 18 1 61519 16 2 21644 14 Ar Total 8 504232 19 8 217401 19 7 872428 19 5 586019 19 3 637848 19 1 17701
110. 91 3125 DAYS l OMS IME 182 625 DAYS l OMS IME 182 625 DAYS l OMS ME 182 625 DAYS l OMS LEVEL 20 1 FLUX 0 NOSTABLE ZERO TIME 60 l OMS IME 1 DAYS ATOMS IME 29 4375 DAYS ATOMS IME 152 1875 DAYS ATOMS IME 182 625 DAYS ATOMS 2 YEARS ATOMS 2 YEARS ATOMS 5 YEARS ATOMS END END User Manuat Issue 1 Feb 2007 UKAEA Fusion 162 FISPACT Test32 NOHEAD EAFV 6 MONITOR 1 AINP FISPACT PWR FUEL 3 1 U235 POY Paluel DENSITY 10 1 FUEL 2 U235 7 948E22 U238 2 453E24 MIND 1 E5 FISCHOOSE 5 U235 U238 PU239 PU240 PU242 HAZA HALF GRAPH 51112345 FLUX 3 25E 14 ATOMS LEVEL 20 50 TIME 730 5 DAYS UNCERT 2 TAB1 41 ATWO DOSE 1 ATOMS LEVEL 20 1 FLUX 0 NOSTABLE ZERO TIME 60 ATOMS TIME 1 DAYS ATOMS TIME 29 4375 DAYS ATOMS TIME 152 1875 DAYS ATOMS TIME 182 625 DAYS ATOMS TIME 2 YEARS ATOMS TIME 2 YEARS ATOMS TIME 5 YEARS ATOMS END END Test41 NOHEAD AINP FISPACT IRRADIATION OF TI EEF 172 FW 1 0 MW M2 MASS 1 0 1 100 0 ND 1 E5 51112345 WALL 1 00 ATOMS LEVEL 100 1 TIME 2 5 YEARS HAZA HALF ATWO UNCERT 3 MINS ATOMS HOURS ATOMS DAYS ATOMS DAYS ATOMS YEARS ATOMS UKAEA Fusion User Manual Issue 1 Feb 2007 FISPACT 163 Test42 NOHEAD AINP FISPACT IRRADIATION OF TI EEF FW 1 0 MW M2 DENSITY 4 54 FUEL 5 46 1 00619E24 TI4
111. 95 Appendix 4 Approximate y spectral data 96 Appendix 5 Sensitivity equations 97 Appendix 6 Data libraries 98 5570 IT 98 BARD OS ON o ud u 99 EAF P _XS 2007 e meet leet 99 MUS Dior EL LM AEn S Aa Een 100 PAR DIO uet 100 enc 100 au cried ae aba a 101 JDE CU 2007 aco et ees e aa 101 EXP NERISOUU ues 102 PAR 200 sen io Bsa col bata 102 BAR P BIS Z ren ee ana Re 102 EAF HAZ 2 S u usu sanus Endo RA OD S ER MCI aS as 102 EAF A220077 e tte teet itae Ned A Sek a gy 103 EAFP CUEXR 2007 aci pe qe eus 103 LOR 2001 cxi qana aa i LOL LM LU 104 EAF SPEC 2007 5 te ut at diti t au ted 104 EAF 2007 nus ias ideoque cei quas 104 EAE ZXBS 200 T oes ein i al ap da tue ec 105 Appendix 7 Bremsstrahlung corrections 106 Appendix 8 Pathways 109 Appendix 9 Uncertainties 113 UN2007 led hm te fe wa 113 FISPACT uncertainty estimation ja a a 114 Pathways containing 2 reactions 118 Pathways containing arbitrary number of reactions
112. ACOK 0 98 FRACWT 0 005 NMAXB 3 NMAXR 3 NMAXC 12 ZZZLVL 50 0 JIUNCER gt 54 UNCIYPBJUNGTY o eee ds ea esed SS 57 HAL uet N as Mt ua 57 ZERO aa a r a Nuqa Gu Sha ss 58 58 Examples of preliminary input 59 Examples of main input 61 Interpretation of FISPACT output 66 Header and user information asss 66 Library am formatlofk eco e ei NN eode cii asua E 66 Nu clde tunt te adeat uya DNA au ean waka es 67 Summary and elemental 68 Gamma au la Rua in 69 SSS TU SH CTV YOU a shes TTE 70 Uncertainty estimates 70 Bremsstrahlung corrections 7 Pathway cde s doe aere tae a a DAE acer ERO ied a 72 End of Case uA i de etas iie q S u ms Sus ee mt 73 PRIN BLT Ut A aetas 74 Appendix 1 Cross section group structures 83 WEST OI ITN spectra siete s ENDE KNEE 90 Appendix 2 Solution of the differential equations 91 Appendix 3 y dose rate 93 Contact y dose cedi E u bu e 93 y dose rate from point source eese de Un ga
113. AYX file produced from the current decay data EAF DEC 2007 so the ARRAY option was used By specifying SPEK any nuclides with no y spectral data had this synthesised approximately this is recommended The collapsed cross section file COLLAPX in this case for EAF 2007 in Zone 15 is read and substituted into the existing ARRAYX file AINPUT FISPACT An inventory run This is the standard start to an inventory calculation see the following section for complete examples User Manual Issue 1 Feb 2007 FISPACT 61 Examples of main input Six examples of input data are given in each case it is assumed that a condensed library an ARRAYX file is being used see previous section More examples are given in Appendix 14 NOHEAD AINPUT FISPACT Produce PRINTLIB PRINTLIB O0 END END of run This run produces an output of the data libraries and it is recommended that this type of run be done to produce a reference document for a particular decay data library which can be used for subsequent work MONITOR 1 AINPUT FISPACT Irradiation of 1 ppm of K in FE MASS 1 0 2 K 1 4 ER 99 9999 BREM 1 10 MIND 1 5 HAZA ATWO GRAPH 2 0 0 1 3 WALL 5 0 LEVEL 100 1 TIME 2 5 YEARS ATOMS LEVEL 20 1 FLUX 0 0 ZERO TIME 0 1 YEARS ATOMS TIME 0 9 YEARS ATOMS TIME 49 0 YEARS ATOMS END END of K irradiation This case models the irradiation of 1 kg of iron containing 1 ppm of p
114. B Firestone able of Radioactive Isotopes John Wiley and Sons 1986 24 R A Forrest The European Activation File EAF 2007 decay data library UKAEA FUS 537 2007 25 R W Mills An initial study of providing energy dependent fission product yields JEFF Doc 1157 2006 26 International Commission on Radiological Protection Dose Coefficients for Intakes of Radionuclides by Workers ICRP Publication 68 1995 Pergamon Press Oxford and Age dependent Doses to Members of the Public from Intake of Radionuclides Part 5 ICRP Publication 72 1996 Pergamon Press Oxford 27 International Commission on Radiological Protection Age dependent doses to members of the public from intake of radionuclides Part 5 Compilation of ingestion and inhalation dose coefficients ICRP Publication 72 1996 28 A W Phipps G M Kendall J W Stather and T P Fell Committed Equivalent Organ Doses and Committed Effective Doses from Intakes of Radionuclides NRPB R245 1991 and A W Phipps and T J Silk Dosimetric Data for Fusion Applications NRPB M589 1995 29 R A Forrest Dosimetric data for FISPACT2 AEA FUS 182 1992 30 R A Forrest The European Activation File EAF 2007 biological clearance and transport libraries UKAEA FUS 538 2007 31 Regulations for the safe transport of radioactive material 1985 edition and supplement 1988 Safety Series No 6 IAEA Vienna 32 A
115. E 14 1 03209E 13 1 845965E 15 1 465219E 15 9 9798 19E 14 3 445054E 14 1 032095E 13 1 72800E 06 9 35912E 12 9 359153 12 3 45600E 06 7 69609E 12 7 696120 12 5 18400E 06 6 32857E 12 6 328591E 12 6 91200E 06 5 20404E 12 5 204060E 12 8 64000E 06 4 27933E 12 4 279347E 12 1 72800E 07 1 60899E 12 1 608992E 12 2 59200E 07 6 04962 11 6 049647E 11 4 32000 07 8 55225 10 8 552292E 10 Max diff 0 00049 396 decay test 48 395 1 1500 10 s gt CI 3 336 10 s 8 488990 10 s gt K stable By B7 5 00000E 00 1 00000E 01 1 50000E 01 5 00000E 01 1 00000E 02 5 00000E 02 1 00000E 03 2 00000E 03 5 00000E 03 1 00000E 04 5 00000E 04 1 00000E 05 Max diff 000086 gt 001570 002145 7 39805 19 5 47312 19 4 04904 19 4 91105 18 2 41184 17 8 16092 06 0 00000 00 0 00000 00 0 00000 00 0 00000 00 0 00000 00 0 00000 00 UKAEA Fusion 7 39805E 19 5 47312E 19 4 04904E 19 4 91105E 18 2 41184E 17 8 16099E 06 0 00000E 00 0 00000E 00 0 00000E 00 0 00000E 00 0 00000E 00 0 00000E 00 2 60053E 19 4 52171E 19 5 94032E 19 9 43808E 19 9 80404 19 9 04443 19 8 15198 19 6 62256E 19 3 55071E 19 1 25641E 19 3 08787E 15 9 50208E 10 2 60053E 19 4 52171E 19 5 94032E 19 9 43808E 19 9 80406E 19 9 04301E 19 8 15070E 19 6 62152E 19 3 55016E 19 1 25621E 19 3 08740E 15 9 50064E 10 1 41885E 16 5 16878E 16 1 06405E
116. E 20 FLUXES 02 ILE 17 COLLAPX4 02 PSE 69 20 FLUXES 03 LE 17 COLLAPX4 03 PSE 69 END OF CO UKAEA Fusion User Manuat Issue 1 Feb 2007 FISPACT 167 Test61 NOHEAD MONITOR 1 SPEK ENFA EAFDEC97 EAF97 spec 01 TAPA FISPACT THREE SPECTRA DENSITY 9 838 FUEL 6 0235 7 9991E20 0238 2 1149 22 016 4 3795 22 017 1 6682 19 018 8 7798 19 PU239 1 0 4 HALF MIND 1 0E8 UNCERT 0 FLUX 2 59032E 14 ATOMS LEVEL 50 10 TIME 6 109E 06 DAYS SPECTRUM NEWFILE 12 COLLAPX4 02 SPEK ENFA EAFDEC97 EAF97 spec 02 ARRAY FLUX 2 64634E414 TIME 6 108994E0 DAYS SPECTRUM NEWFILE 12 COLLAPX4 03 SPEK ENFA EAFDEC97 EAF97 spec 03 ARRAY FLUX 2 66930E414 TIME 2 44410E 01 DAYS ATOMS LEVEL 50 1 FLUX 0 0 ZERO TIME 1 00E2 ATOMS TIME 9 00E2 ATOMS TIME 9 00E3 ATOMS TIME 9 00E4 ATOMS TIME 9 00E5 ATOMS TIME 9 00E6 ATOMS END END OF MULTIPLE RUN Test70 NOHEAD MONITOR 1 AINP FISPACT PURE IRON DENSITY 7 874 MASS 1 0 1 FE 100 0 MIND 1 E5 1 0 15 1 100 1 1 0 YEARS ATOMS 1 20 1 LUX 0 ZERO ME 1 DAYS ATOMS ME 9 DAYS ATOMS IME 90 DAYS ATOMS IME 265 25 DAYS ATOMS IME 9 YEARS ATOMS END END OF COLLAPSE User Manual Issue 1 Feb 2007 UKAEA Fusion 168 FISPACT Test71 NOHEAD MONITOR 1 AINP FISPACT PURE IRON DENSITY 7 874 MASS 1 0 1 FE 100
117. EARS ATOMS END END test r3 NOHEAD MONITOR 1 EAFV 4 AINP FISPACT Simplified reactions on Cr 50 FUEL 1 CR50 1 0E25 DENSITY 7 19 lt lt 1918 1 gt gt lt lt 1919 2 gt gt lt lt 1920 H 3 gt gt lt lt 1921 3 gt gt lt lt 1922 4 gt gt OVER CR50 ACROSS CR49 0 0 ACROSS TI46 0 0 ACROSS 47 0 0 ACROSS 48 0 0 ACROSS 49 0 0 ACROSS CR51 1 36187E 2 ACROSS V50 3 37120E 2 ACROSS 49 3 28852 2 ACROSS 1918 6 52733E 2 ACROSS 1919 1 32379E 3 ACROSS 1921 0 0 ACROSS 1922 0 0 OVER CR51 ACROSS CR50 0 0 ACROSS 47 0 0 ACROSS TI50 0 0 ACROSS CR52 2 51242 2 ACROSS V51 3 11597E 2 ACROSS V50 3 80524 3 ACROSS V49 1 29794 4 ACROSS TI49 2 36955E 7 ACROSS TI48 8 43151E 3 ACROSS 1918 3 38729E 2 ACROSS 1919 1 09204E 3 ACROSS 1920 1 29794E 4 ACROSS 1921 2 36955E 7 ACROSS 1922 8 43151E 3 ALAM 2 3936 6 1 OVER CR52 ACROSS CR51 0 0 ACROSS 48 0 0 ACROSS V52 0 0 ACROSS TI50 0 0 ACROSS TI51 0 0 ACROSS CR53 5 26653E 3 ACROSS V51 8 62371 3 ACROSS TI49 3 94947E 3 ACROSS 1918 7 71701E 3 ACROSS 1919 9 06704E 4 ACROSS 1921 0 0 ACROSS 1922 3 94947 3 OVER CR53 ACROSS CR52 0 0 ACROSS TI49 0 0 ACROSS V53 0 0 ACROSS V52 0 0 ACROSS TI51 0 0 ACROSS TI50 0 0 ACROSS TI52 0 0 ACROSS CR54 1 49953E 2 ACROSS V51 6 37435 5 ACROSS 1918 0 0 ACROSS 1919 0 0 ACROSS 1920 6 37435E 5 ACROSS 1921 0 0 ACROSS 1922 0 0 U
118. HOURS ATOMS TIME 1 DAYS ATOMS TIME 7 DAYS ATOMS TIME 1 YEARS ATOMS NOHEAD AINP FISPACT PWR FUEL 3 1 U235 PWRDEAN DENSITY 10 1 FUEL 2 U235 7 948E22 U238 2 453E24 1 5 5 1112345 FLUX 3 25E 14 ATOMS LEVEL 20 8 TIME 30 4375 DAYS 1 41 ATWO DOSE 1 ATOMS 60 875 DAYS l OMS IME 91 3125 DAYS l OMS ME 182 625 DAYS l OMS IME 182 625 DAYS l OMS IME 182 625 DAYS l OMS LEVEL 20 1 FLUX 0 NOSTABLE ZERO 60 5 1 DAYS ATOMS IME 29 4375 DAYS ATOMS IME 152 1875 DAYS ATOMS IME 182 625 DAYS ATOMS IME 2 YEARS ATOMS IME 2 YEARS ATOMS IME 5 YEARS ATOMS END END User Manuat Issue 1 Feb 2007 IRRADIATION OF Ti EEF 175 FW 1 0 MW M2 UKAEA Fusion 158 FISPACT Test22 UKAEA Fusion NOHEAD EAFV 6 MONITOR 1 AIN FIS PW MIN HAL GRA R FUEL 3 1 DENSITY 10 1 FUEL 2 U235 7 948E22 U238 2 453E24 D 1 E5 NOSTABLE ZERO TIME 60 ATOMS TIME TIME TIME TIME TIME 2 YEARS AT TIME 2 YEARS AT TIME 5 YEARS AT END END U235 5 DAYS 1 DAYS ATOMS 29 4375 DAYS AT 152 1875 DAYS ATOMS 182 625 DAYS AT l OMS l OMS l OMS PWRDEAN FISCHOOSE 5 U235 U238 PU239 PU240 PU242 HAZA PH51112345 FLUX 3 25E 14 ATOMS LEVEL 20 50 TIME 730 UNCERT 2 TAB1 41 ATWO DOSE 1 ATOMS LEVEL 20 1 FLUX 0 TOMS TOMS User Manual Issue 1 Feb
119. L asus u e B a Ka 217 Appendix 20 Non steadv irradiation 227 Introductio nue Kat ita mat an 227 Modelling an Hrradiation 227 Implications for FISPACT REN ER EEAIME 228 References 230 FISPACT Document modifications The following procedure will be followed if the FISPACT 2007 user manual requires updating Each page carries the issue number and date if a page does not show Issue 1 then there has been a modification which will be shown by strikethrough of the old prior to the new underlined text and by a sidebar in the margin e g This is just a demonstration of the changes how the modifications are made In some cases it will be necessary to add new pages and the page numbering will be changed by adding a b after the page number e g Originally the page numbers were 26 27 28 If a new sheet is inserted after page 27 then the numbers will become 26 27 27a 27b 28 Users will be sent modified or additional pages and these should be inserted based on the Instructions which are given in the Modification List Modification List Issue Date Modified Pages Removed Pages The current issue of the manual is Issue 1 The current version of FISPACT is 07 0 0 build 23 Users are asked to notify UKAEA of any problems with FISPACT New versions will attempt
120. ME 0 5 YEARS EVEL 20 1 FLUX 0 TIME 0 083 VEARS ATOMS WALL 1 0 EVEL 100 1 TIME 0 5 YEARS User Manual Issue 1 Feb 2007 UKAEA Fusion 152 FISPACT Test10 NOHEAD AINP FISPACT IRRADIATION OF BE EEF FW 1 0 MW M2 MASS 1 0 1 BE 100 0 MIND 1 E5 WALL 1 00 ATOMS OVER BE9 ACROSS BE10 1 4195E 4 OVER H3 ALAM 1 9455E8 1 HALF DOSE 1 EVEL 100 1 TIME 0 5 YEARS UNCERT 3 ATOMS EVEL 20 1 FLUX 0 UNCERT 0 ZERO TIME 0 083 YEARS ATOMS PARTITION 2 H 0 HE 0 TIME 0 083 YEARS ATOMS END END Test11 NOHEAD INP ISPACT IRRADIATION OF TI EEF 175 FW 1 0 MW M2 MASS 1 0 1 TI 100 0 MIND 1 E5 GRAPH 51112345 WALL 1 00 ATOMS LEVEL 100 1 TIME 2 5 YEARS HAZA HALF ATWO UNCERT 3 DOSE 1 ATOMS LEVEL 20 1 FLUX 0 ZERO TIME 1 MINS ATOMS TIME 1 HOURS ATOMS TIME 1 l OMS 7 1 TIME DAYS ATOMS TIME YEARS ATOMS END END UKAEA Fusion User Manuat Issue 1 Feb 2007 FISPACT 153 Test12 NOHEAD AINP FISPACT IRRADIATION OF TI EEF FW 1 0 MW M2 DENSITY 4 54 FUEL 5 46 1 00619E24 TI47 9 18148E23 48 9 28210E24 TI49 6 91755E23 TI50 6 79178E23 MIND 1 E5 FLUX 4 27701E14 ATOMS LEVEL 100 5 SENSITIVITY SIGMA 1E 10 2 1 48 SCA8 TI49 SCA8 5 48 ERROR 2 2 148 5 48 1 I49 SC49 1 IME 2 5 YEARS ATOMS END END 3 3
121. N t s NE resti aee A9 50 1 n l n ay 9 Nun 1 nt 49 50 M t 1 a The exact equation for the production of 0 is the same as the limiting form for the production of N 1 f as can be seen by comparing equations A9 24 and A9 49 Thus trivially the long User Manual Issue 1 Feb 2007 FISPACT 125 lived limit for the final nuclide in an n 1 link pathway containing reactions is given in equation A9 52 1 Npa O gt 9 A9 52 i 1 In the short lived limit then gt gt and the differential can set to zero Equation A9 28 can be simplified to give equation A9 53 A A9 53 This is easilv solved to give equation A9 54 n 2 N t I A9 54 i 1 1 1 The number of atoms of final nuclide in pathway stable with no reactions is calculated by integrating the equation equivalent to equation A9 24 This is shown in equation A9 55 ae 0 4 N tdt 5 Nu 9 55 0 i l i Equations A9 52 and A9 55 are the general limiting forms of which equations A9 26 and A9 27 are special cases For the general case of a pathway containing decays the corresponding limits of equation A9 48 are given in equations A9 56 and A9 57 n l A T A Ne SED heane A9 56 jsl INER E cT E A9 57 Equations A9 56 and A9 57 are the general limiting
122. NVERT A numeric value MUST follow the code word Numeric value required for NESTRC GRPCONVERT A numeric value MUST follow the code word Numeric value required for NLINK PATH A numeric value MUST follow the code word Numeric value required for NMAX ROUTES A numeric value MUST follow the code word Numeric value required for NMAXB UNCERTAINTY A numeric value MUST follow the code word Numeric value required for NMAXC UNCERTAINTY A numeric value MUST follow the code word Numeric value required for NMAXR UNCERTAINTY A numeric value MUST follow the code word Numeric value required for NOPT GRAPH A numeric value MUST follow the code word Numeric value required for NPART PARTITION A numeric value MUST follow the code word Numeric value required for NPROJ PROJECTILE A numeric value MUST follow the code word Numeric value required for NPULSE PULSE A numeric value MUST follow the code word Numeric value required for NRESU RESULT A numeric value MUST follow the code word Numeric value required for NUMG GRAPH A numeric value MUST follow the code word Numeric value required for NYLD FISYIELD A numeric value MUST follow the code word Numeric value required for PMIN ROUTES A numeric value MUST follow the code word Numeric value required for T TIME A numeric value MUST follow the code word Numeric value required for TLOOP LOOPS A numeric value MUST follow the
123. ON LEVEL C N This code word should only be used for calculations where small quantities of impurities in an iron matrix are to be irradiated In a run without this code word the activity of the impurities would probably be masked by the activity of the iron In order to remove the background this code word causes the iron matrix to be replaced by a matrix of Fe a stable pseudo isotope with no neutron reactions so that the printed inventories and dose rates refer only to the impurities An example of the use of this code word follows IRON MASS 1 0 2 EH 99 9999 AG 1 0E 4 In this run corresponding to the irradiation of 1 ppm of silver impurity in iron the output will be due only to the reactions on silver However the y dose rate will represent decays of silver isotopes in an iron matrix rather than in solid silver This code word allows the input of the two parameters which determine the nuclides that are in equilibrium C and the number of time subintervals N into which the irradiation time is divided This code word MUST be used for the first irradiation but only needs to be used again for the remaining time intervals if the parameters need to be changed If the time interval is seconds long then the nuclides with decay constant 1 51 will be in equilibrium if AT gt C These nuclides are flagged with an in the printed output For typical irradiations a value of C 100 is recommended while
124. PACT 11 Use of FISPACT Details of the installation of FISPACT and the data libraries and differences in running FISPACT on different platforms are discussed in Appendix 13 This section concentrates on the information that a user must assemble prior to using FISPACT and on introducing the code words that instruct FISPACT to carry out various types of calculation Note that in the following text the word neutron is used for simplicity it must be remembered that in most cases deuteron or proton could also be used The user will require details of the material to be irradiated by neutrons the times of irradiation and cooling and most importantly details of the neutron spectrum that is to be used The spectrum must be available either in one of the standard energy structures used by FISPACT or in an arbitrary energy structure in which case the user must supply details of the energy boundaries The standard energy structures are WIMS XMAS VITAMIN J VITAMIN J TRIPOLI and TRIPOLI Cross section data in these seven energy structures are available for the EAF neutron induced cross section library and in the VITAMIN J structure for the deuteron and proton induced reaction library Details of the energy boundaries of the seven standard structures are given in Appendix 1 One of the seven group format libraries is used to form the l group effective cross sections that FISPACT requires by collapsing
125. PROJECTILE NPROJ 1 This code word defines the incoming particle for the activation calculations This code word MUST come before the AINP code word file if a deuteron or proton library is used For a deuteron library NPROJ should be set to 2 for a proton library NPROJ should be set to 3 A neutron library uses the default value of 1 The code word NOERROR MUST be used in a run with deuteron or proton data An example of the use of this code word for an activation calculation with a proton library follows MONITOR 1 PROJ 3 NOERROR AINP FISPACT IRRADIATE MATERIAL IN IFMIF PULSE NPULSE This code word is used to start the construct in the INPUT file NPULSE is the number of times that the code words between PULSE and ENDPULSE are repeated It is possible to nest this pair of code words up to 5 deep and the maximum value of NPULSE for any loop is 500 User Manuat Issue 1 Feb 2007 UKAEA Fusion 48 FISPACT This facility is included to that a series of identical pulses off time on time can be represented easily in the input file An example of the use of this code word follows PULSE 5 FLUX 0 0 TIME 1 0 HOURS SPEC TRUM FLUX 1 0 15 TIME 1 0 HOURS SPEC TRUM ENDPULSE FLUX 0 0 TIME 1 0 HOURS SPEC TRUM FLUX 1 0 15 TIME 1 0 HOURS ATOMS At the end of the irradiation it is wished to include six hour long pulses Five of these are specified in the loop using SPECTRUM so that no d
126. RS HAZA HALF ATOMS LEVEL 20 1 FLUX 0 ZERO TIME 1 MINS ATOMS TIME 1 HOURS ATOMS TIME 1 DAYS ATOMS TIME 7 DAYS ATOMS 1 Test182 YEARS ATOMS MONITOR 1 PROJ 2 NOERROR Al FI NP SPACT IRRADIATION OF TI by d IFMIF DENSITY 4 54 5 00619E24 18148E23 28210E24 91755E23 79178 23 JB GRAPH 300123 FLUX 5 0E14 1 100 1 2 5 YEARS 1 20 1 MINS ATOMS HOURS ATOMS DAYS ATOMS DAYS ATOMS UKAEA Fusion User Manual Issue 1 Feb 2007 FISPACT 173 Test183 Test184 NOHEAD MONITOR 1 PROJ 2 NOERROR AINP FISPACT IRRADIATION OF TI by d IFMIF MASS 1 0 1 TI 100 0 DENSITY 19 254 MIND 1 E5 FLUX 1 0E14 UNCERT 3 EVEL 100 1 TIME 1 0 YEARS HALF ATOMS EVEL 20 1 FLUX 0 ZERO TIME 10 0 YEARS ATOMS END END NOHEAD MONITOR 1 PROJ 2 NOERROR AINP FISPACT IRRADIATION OF FE U by d IFMIF MASS 1 0 2 FE 99 9999 U 0 0001 MIND 1 E5 GRAPH 50112345 FLUX 1 0E14 ATOMS UNCERT 3 LEVEL 100 10 TIME 2 5 YEARS MINS ATOMS HOURS ATOMS DAYS ATOMS DAYS ATOMS YEARS ATOMS 000 YEARS ATOMS User Manuat Issue 1 Feb 2007 UKAEA Fusion 174 FISPACT Test185 NOHEAD MONITOR 1 PROJ 2 NOERROR AINP FISPACT IRRADIATION OF PWR FUEL by d IFMIF DENSITY 10 1 FUEL 2 0235 7 948 22 0238 2 453 24 MIND 1 5 HAZA HALF GRAPH 50112345 FLUX 3 25 14
127. S appears to be stable has the TLOOP parameter been sensibly set Isomer does not decay to gs lt CHAINP gt One of the isomers that is to be included in the calculation by LOOPS does not decay back to the ground state nuclide on the pathway so no loop is formed Isomer not in library lt RENUCL gt The nuclide specified does not have an isomeric state in the current library has the correct decay library been used Isomer not in library FUEL The isomer specified is not in the decay library Isomer symbol not recognised lt CNVTXT gt isomer symbol MUST be either m or n and be in lower case Isomer symbol not recognised lt RENUCL gt The isomer symbol MUST be either m or n in upper lower case Isotope and daughter not recognised lt OVERID gt The isotope specified after the OVER code word is not in the index of nuclides check that the isotope has been correctly entered Isotope not recognised lt OVERID gt The isotope specified after the OVER code word is not in the index of nuclides check that the isotope has been correctly entered ITDEC Invalid argument An argument of one of the DECIN functions responsible for the processing of the INPUT file is invalid UKAEA should be contacted for advice JSTRM can only be 12 17 or 20 NEWFILE Only files connected to streams 12 17 or 20 can be redefined LAMBDA or SIGMA required SENSITIVITY If the sensitivity with respect to half life
128. SPACT 97 to prepare COLLAPX files for each spectrum and during the run to change the COLLAPX file that Is used as the run progresses The main new features of FISPACT 99 were the ability to input the EAF 99 library use of a new energy group structure the TRIPOLI 315 group structure and the use of clearance data to calculate a clearance index for the disposal of radioactive material Several bugs were dealt with the most important was to ensure that the summary at the end of the output could be printed for large numbers of time intervals This change means that pulsed irradiations with large numbers of time intervals could be run successfully User Manuat Issue 1 Feb 2007 UKAEA Fusion FISPACT Version 2001 Version 2003 Version 2005 UKAEA Fusion FISPACT does not use dates in the calculation of inventories but does print out various times and dates in the output file FISPACT 99 was designed for Y2K compliance and there have been no Y2K problems reported Changes were made to the format of printing dates these show 4 digits instead of 2 1999 not 99 So long as the operating system and computer used to run FISPACT are Y2K compliant then FISPACT should introduce no Y2K issues The main new features of FISPACT 2001 are the ability to input the EAF 2001 library calculation of dominant nuclides and uncertainties for three additional quantities additional information output and the updating of physical constants and
129. T file In previous versions it is assumed that a neutron spectrum in 175 groups is available for calculations of the pseudo cross sections In the current version a neutron spectrum in either 175 or 211 groups can be used ARRAYX and COLLAPX files prepared with a neutron spectrum in 175 or 211 groups MUST be used with this code word More details about SCPR are given in Appendix 12 An example of the use of this code word follows SEQUENTIAL 1 O0 In this case sequential charged particle reactions will be included in the calculations but the values of the pseudo cross sections are not displayed in the output This code word is an alternative to ATOMS It suppresses the inventory so that only the y spectrum and total values printed for the time interval It is useful if summary information is required for many time intervals but the details of the individual nuclide contributions are not needed User Manuat Issue 1 Feb 2007 FISPACT 53 SPLIT SPLIT 0 1 2 TAB3 C TABA D This code word allows the display of an additional summary table at the end of the run This summary table contains separate information on the heat production by beta and gamma radiation at each time interval to be output after the existing summary table By default this new summary table is not printed but it can be displayed if SPLIT is set to 1 Note that if the new summary table is required then the cod
130. This is followed by the current user information where any changes since the publication of this manual will be given Library information In response to the code word AINPUT another box giving the library information is printed FR Ao R S TENA FR 7 RAM Fa FISSION YIELD DATA FROM JEFF 31 WEIGHTED BY ABOVE NEUTRON SPECTRUM ARRAYX produced from EAF 2007 and above COLLAPX on 13 02 07 using FISPACT 07 0 0 OUTPUT produced by FISPACT 07 0 0 Build 0023 dated 12 02 07 COLLAPX from EAF 2007 0 211G and IFMIF test Vit J on 13 02 07 using FISPACT 07 0 0 This shows the name of the cross section library that was used for the collapsing the date that it was produced and the version of FISPACT that was used The origin of the fission yield data and the name of the decay data library and the date of production of the ARRAYX file follow The actual build number is shown as an additional check of the version used to produce the output UKAEA Fusion User Manuat Issue 1 Feb 2007 FISPACT 67 Nuclide inventory Following the code word FISPACT the remainder of the input file is read and each code word and its parameters are printed out before the processing of the code word Following ATOMS the number of iterations required for convergence is given If convergence has not been reached then the following message is shown CASE NOT PROPERLY CONVERGED but if no flags set then converge
131. UKAEA FUS 534 EURATOM UKAEA Fusion FISPACT 2007 User manual R A Forrest March 2007 UKAEA EURATOM UKAEA Fusion Association Culham Science Centre Abingdon Oxfordshire OX14 3DB United Kingdom Telephone 44 1235 466586 Facsimile 44 1235 466435 UKAEA EASY Documentation Series UKAEA FUS 534 FISPACT 2007 User manual R A Forrest EURATOM UKAEA Fusion Association Culham Science Centre Abingdon Oxfordshire OX14 3DB UK ISO 14001 Abstract FISPACT 15 the inventory code included in the European Activation System EASY A new version of FISPACT FISPACT 2007 has been developed and this report is the User manual for the code It explains the use of all the code words used in the input file to specify a FISPACT run and describes how all the data files are connected A series of appendices covers the working of the code and the physical and mathematical details Background information on the data files and extensive examples of input files suitable for various applications are included Contents Document modifications Acknowledgements Disclaimer Contact person Introduction Version summary Version huc RNA Shae aya WAGISION RR CNN HIER Version IT a alte dese UG
132. URS ATOMS l OMS DAYS ATOMS YEARS ATOMS D m gt Test17 UNCERT 3 DOSE 1 ATOMS LEVEL 20 1 MINS ATOMS HOURS ATOMS DAYS ATOMS DAYS ATOMS YEARS ATOMS User Manuat Issue 1 Feb 2007 UKAEA Fusion 156 FISPACT Test18 NOHEAD MONITOR 1 AINP FISPACT IRRADIATION OF Fe EEF 175 FW 1 0 MW M2 MASS 1 0 1 TIME 2 5 YEARS HALF ATWO UNCERT 2 SPECTRUM PULSE 5 EVEL 20 1 FLUX 0 TIME 1 0 HOUR SPECTRUM EVEL 100 1 WALL 1 0 TIME 1 0 HOUR SPECTRUM ENDPULSE EVEL 20 1 FLUX 0 TIME 1 0 HOUR SPECTRUM EVEL 100 1 WALL 1 0 TIME 1 0 HOUR ATOMS FLUX 0 0 ZERO TIME 1 MINS ATOMS TIME 1 HOURS ATOMS TIME 1 DAYS ATOMS TIME 7 DAYS ATOMS TIME 1 YEARS ATOMS END END Test19 NOHEAD MONITOR 1 AINP FISPACT IRRADIATION OF Sc 45 EEF FW DENSITY 2 989 FUEL 1 5 45 1 0 25 MIND 1 5 FLUX 4 27701E14 LEVEL 10000 1 TIME 2 5 YEARS LOOPS 20 PATH 2 5 45 R 5 46 R 5 47 PATH 3 5 45 R 42 D 42 R AR39 PATH 3 5 45 R K42 D 42 R CA41 ATOMS END END UKAEA Fusion User Manuat Issue 1 Feb 2007 FISPACT 157 Test20 Test21 NOHEAD MONITOR 1 AINP FISPACT MASS 1 0 1 TI 100 0 MIND 1 E5 WALL 1 00 ATOMS LEVEL 100 1 TIME 2 5 YEARS HAZA HALF ATWO UNCERT 2 UNCTYPE 2 ATOMS LEVEL 20 1 FLUX 0 ZERO TIME 1 MINS ATOMS TIME 1
133. ZERO UNCERT 2 NOSTABLE TIME 1 0E 9 ATOMS TIME 0 5 ATOMS TIME 0 5 ATOMS TIME 1 MINS ATOMS TIME 1 HOURS ATOMS TIME 5 HOURS ATOMS TIME 0 75 DAYS ATOMS TIME 1 0 DAYS ATOMS TIME 1 DAYS ATOMS TIME 2 DAYS ATOMS TIME 2 DAYS ATOMS TIME 1 DAYS ATOMS TIME 1 DAYS ATOMS TIME 1 DAYS ATOMS TIME 1 DAYS ATOMS TIME 1 DAYS ATOMS TIME 5 DAYS ATOMS TIME 10 DAYS ATOMS TIME 10 DAYS ATOMS TIME 10 DAYS ATOMS TIME 10 DAYS ATOMS TIME 10 DAYS ATOMS TIME 50 DAYS ATOMS TIME 100 DAYS ATOMS TIME 252 DAYS ATOMS TIME 0 76923 YEARS ATOMS TIME 1 YEARS ATOMS TIME 3 YEARS ATOMS TIME 25 YEARS ATOMS END END User Manuat Issue 1 Feb 2007 UKAEA Fusion 170 FISPACT Test74 NOHEAD MONITOR 1 AINP FISPACT IRRADIATION LMJ CONCRETE DENSITY 2 30 FUEL 3 K39 2 60886E 24 K40 3 27303E 20 K41 1 88275E 23 TAB1 41 TAB4 44 MIND 1 E5 LEVEL 100 1 PULSE 5 LUX 2 11598E 19 IME 1 0E 9 ATOMS LUX 0 0 ME 30 DAYS ATOMS NDPULSE 21 A BH d HE LUX 2 11598 19 ATOMS TIME 1 0E 9 HAZA HALF UNCERT 2 DOSE 1 TIME 1 0E 9 ATOMS TIME 1 DAYS ATOMS TIME 10 DAYS ATOMS TIME 10 DAYS ATOMS TIME 100 DAYS ATOMS Test81 Test86 Identical to Test11 Test16 Test87 UNCERT 3 DOSE 1 MINS ATOMS HOURS ATOMS DAYS ATOMS DAYS ATOMS YEARS ATOMS UKAEA Fusion User Manuat Issue 1 Feb 2007 FISPACT 171 Test88 Test90 Identical to Test18 Test20 Tes
134. ainty alone or both together could be defined in the input file Note that the decay data library used should contain these uncertainty User Manuat Issue 1 Feb 2007 FISPACT 5 Version 99 data for all nuclides as do EAF DEC 97 EAF DEC 2005 if this new option is to be used The data on photon absorption used to calculate dose rates was updated These are read from a file rather than being stored in the FISPACT code as in previous versions To aid the preparation of input files for cases with pulsed irradiation a simple loop construct was added This enables all the code words between the start and finish of the loop to be repeated as many times as required In cases where actinides are relevant it is possible to specify which of these will produce fission products when they undergo fission Such a facility allows all fission product production to be switched off or to be allowed for a particular set of actinides for detailed investigations of the contributions of individual nuclides Data on the higher actinides Bk Fm were included in EAF 97 and FISPACT 15 able to treat these actinides in the same ways as the other actinides At Cm In most work modelling irradiations it is possible to assume that the neutron spectrum remains constant However in some cases it is required to allow for a change in the shape of the neutron spectrum Rather than having to do a series of separate runs it was possible in FI
135. al details The installation can be tested by carrying out the following steps 1 Open the EASY 2007 User Interface by clicking the Start Programs EASY 2007 User Interface menu item 2 Click the FISPACT item on the Run menu and FISPACT will carry out the calculation shown in the INPUT file Note that all entries in FILES have been tailored to the user s system during installation User Manuat Issue 1 Feb 2007 UKAEA Fusion 146 FISPACT UNIX UKAEA Fusion 3 Use the Help file for more details The EASY 2007 code package including the inventory code FISPACT 2007 and the EAF 2007 libraries is available on a DVD R disc UNIX executables fisp07 are available for various systems IBM AIX SUN Sparc and Intel Red Hat LINUX Dec ALPHA and MacIntosh EASY 2007 requires a minimum of 650 Mbytes of free disk space on the UNIX machine on which it will be installed The FISPACT program and the associated EAF libraries under UNIX need to be placed in a particular fixed directory structure since the program will look for files in definite places in the directory tree The user is advised to check the compatibility of the directory tree layout with his own system setup and to edit correct the links as required in the script fisplink and the file named FILES A version of the script fisplink and file FILES in eaf2007 are provided to properly setup the EAF 2007 data sets They need to be edited for a particular ins
136. al values Note that much of the EAF 97 data are modified as described in the following to enable analytical results to be obtained Data library processing Prior to any activation calculations FISPACT has to read in the nuclear data libraries and output the processed data in a format that can be used for future calculations This section considers the various types of processing and compares the samples of the input data with those reported in the PRINTLIB output that contains the nuclear data in readable form Decay library processing Be decay data UKAEA Fusion The EAF 97 decay data library contains the data for Be shown in Table 1 The EAF 97 fission yield library JEF 2 2 contains data for shown in Table 2 FISPACT stores the nuclear User Manuat Issue 1 Feb 2007 FISPACT 201 data in internal arrays but in order to view the data used for a particular case a run using the code word PRINTLIB can be used to produce output in a format suitable for viewing The part of the PRINTLIB output for is shown in Table 3 When FISPACT reads the gamma spectrum data it converts the discrete data into binned data that conforms to the internal energy structure The calculations for the data are given in Table 4 FISPACT uses a weighted fission yield this is calculated by taking the yield data at the various neutron energies typically 0 0253 eV 400 keV and 14 MeV and multiplying by the fraction of neu
137. alues of ZZZLVL MUST be given in scientific notation Note the code word UNCERTAINTY MUST immediately proceed the code word ATOMS Note that if no uncertainty data exists in the cross section library then the valid values of UNCER are only 0 3 or 4 Examples of the use of this code word follow I s Omitting the code word will ensure that only inventory calculations are carried out and should be the case if a fast scoping run or multiple irradiation periods are required UNCERT 2 This will ensure that in addition to the inventory calculations the pathways to form the dominant nuclides and the uncertainty estimates are output This is the standard use of the code word for a full investigation of activation UNCERT 4 0 95 0 01 4 6 10 100 0 2 This resets the default values and then carries out a full calculation User Manuat Issue 1 Feb 2007 FISPACT 27 UNCTYPE UNCTY 1 WALL WALL This code word allows the user to specify the type of uncertainty contributions to include when calculating the uncertainties of the radiological quantities If JUNCTY is set to 1 or if the code word 15 not used then only the cross section uncertainties are used in the calculation of uncertainties If IUNCTY 2 then only the half life uncertainties taken from the decay data library are used in the calculation of uncertainties If JUNCTY 3 then both cross section and half life uncertainties are used Appendix 9 gives
138. arent nuclide of reaction not in library lt COL175 gt The parent nuclide of a reaction in the cross section library is not present in the decay library has the correct decay library been used Parent nuclide of reaction not in library lt COL211 gt The parent nuclide of a reaction in the cross section library is not present in the decay library has the correct decay library been used Parent nuclide of reaction not in library lt COL315 gt The parent nuclide of a reaction in the cross section library is not present in the decay library has the correct decay library been used Parent nuclide of reaction not in library lt COL351 gt The parent nuclide of a reaction in the cross section library is not present in the decay library has the correct decay library been used Parent nuclide not in library PATH The parent nuclide of a reaction in a pathway is not present in the decay library User Manual Issue 1 Feb 2007 FISPACT 141 Reaction for new uncertainty data not in library lt INTERP gt The reaction specified is not in the cross section library Check it has been entered correctly in INPUT file Reaction for new uncertainty data not in library lt OUTERR gt The reaction specified is not in the cross section library Check it has been entered correctly in INPUT file Reaction for new uncertainty data not in library lt OUTPUT gt The reaction specified is not in the cross section library Check it has been ent
139. at Issue 1 Feb 2007 FISPACT 95 where f mass of element total mass The value of 5 is calculated using equation A3 3 S SE N A a A3 3 where E energy of yray MeV N number of quanta per decay A t activity of material at time t Bq kg y dose rate from point source Equation A3 4 shows the standard formula taken from reference 8 for calculation of the dose rate from a point source in air In this section it is assumed that 1 g of material is present in the source 8 D 5 76 100 A3 4 gt Am 1000 9 where r distance from source m U E energy attenuation coefficient of air m The other symbols are as in equation A3 1 Both equations A3 1 and A3 4 are approximations suitable for FISPACT calculations but it is noted that they are not adequate for specific health physics problems User Manuat Issue 1 Feb 2007 UKAEA Fusion 96 FISPACT Appendix 4 Approximate y spectral data Wherever possible decay data from evaluated files such as JEFF 3 1 have been used to construct the decay data library EAF DEC 2007 used with FISPACT However for 254 unstable nuclides the file contains only the average Y energy no data for the y spectrum are available Without the y spectrum FISPACT is unable to calculate the y dose rate contribution for these nuclides In order to check if any of these nuclides are likely to significantly cont
140. atibility with FISPIN it can be ignored The average energies for vy decays shown as ALPHA lt BETA gt and lt GAMMA gt in MeV and the y energy MeV in each of the 24 groups follow The independent fission yield from each of the fissionable nuclides is given for the first 1696 nuclides 1669 in EAF 2005 1436 in EAF 2003 2001 99 97 1434 in EAF4 1 and 1190 in EAF3 1 The details of the neutron spectrum used to weight the fission yields are given with the fraction of the neutrons gt 5 MeV lt 5 MeV and gt 200 keV and lt 200 keV printed 52 8 000E 01 Fe 52m IT Fe 52 2 000 01 The second section gives the percentage branching ratio for each decay mode of the radionuclides The parent and daughter User Manual Issue 1 Feb 2007 FISPACT 75 nuclides are given with a code representing the decay between them These codes are b or electron capture b decay p B followed by neutron emission B followed by neutron emission B followed by proton emission a decay n neutron emission SF spontaneous fission b SF B followed by spontaneous fission double B decay b a followed by emission followed by emission p proton emission pp double proton emission IT isomeric transition n 2n Be 8 4 328E 03 4 0E 00 The third section gives the cross section in barns the effective cross s
141. ble to read and do New code word PROJECTILE 1695 1715 calculations with deuteron induced addition of NPROJ to SPCTRM reaction data common new code to deal with new data 31 1 06 Need to be able to read and do NPROJ has value 3 for proton 1716 1726 calculations with proton induced new code to deal with new data reaction data Output and PRINTLIB show correct particle in FORMAT statements neutron library magic numbers decay data files need to read STA 5 9 06 Improve output Change FORMATS so that neutron replaced by deuteron or proton in all places 17 1 07 More fissionable nuclides in Need to be able to read number as 1739 1743 ASSCFY I3 not D change initialisation code 18 1 07 Crash if input file contains FISCHO only initialised in AINP 1744 actinides COLL and ENFA routine added after call to User Manuat Issue 1 Feb 2007 UKAEA Fusion 198 FISPACT Problem Solution Modification mE Ps Crash 3 JEFF 3 1 fission JEF 2 2 contains SF data for uem 1746 yields Cf 252 Not present in JEFF 3 1 code did not detect end of file Output needs to report JEFF 3 1 or Change FORMAT statements UKFY4 0 FY data used depending on version and NPROJ 23 1 07 Position of A2 value in output Change FORMAT statements so 1748 depends on whether DENSITY that position is consistent code word used Cases problem in User Interface Users with problems are asked to supply the following information when rep
142. by FISPACT Its value can be seen in the decay data part of the PRINTLIB output if required The specification of nuclides is essential if materials are to be irradiated which do not have the natural isotopic abundance Appendix 16 gives the natural abundance of isotopes for each element used by FISPACT If different values are required then FUEL should be used The total mass of input material is calculated from the amounts of the nuclides input Note that both FUEL and MASS MUST not be used in a particular case If FUEL is used for a run in which an inventory is calculated then the density of the material MUST be specified using DENSITY An example of the use of this code word follows FUEL 2 LI6 8 5E24 LI7 1 5E24 In this case lithium highly enriched in the isotope is to be irradiated GENERIC GENER 1 In addition to the normal output of pathway data there is a section showing generic pathway data A generic pathway is one in which all instances of a link of type Nuclide isomer User Manuat Issue 1 Feb 2007 UKAEA Fusion 30 FISPACT state m n IT Nuclide state g is replaced by Nuclidefstate pathways that when simplified in this fashion have the same form belong to the same generic pathway and the contribution of all the pathways are added to give the contribution of the generic pathway The default is to always print the generic information but it can be switched off by s
143. cal quantities activity heat y dose rate and potential ingestion and inhalation hazards Secondly EAF 3 contained cross sections for nuclides in the atomic number range 85 96 actinides including the fission channel This version was able to use these data in conjunction with fission yield data taken from JEF 2 to calculate the inventory of fission products if actinides were initially present as trace elements in the material Version 4 includes the option of including sequential charged particle reactions as an additional mechanism for the production of activity Details of this mechanism are given in Appendix 12 Two additional energy group structures WIMS and XMAS are treated and in addition it is now possible to input the neutron spectrum in an arbitrary format and then convert to one of the 4 standard structures maximum of twenty rather than 10 dominant nuclides can now be listed and the presentation of the pathway information has been improved The decay data and fission yield data are based on JEF 2 2 Version 4 was designed to use EAF 4 data libraries reference 4 Starting with FISPACT 97 the version number indicates the year of release of the new version The main new features of FISPACT 97 were the ability to input the EAF 97 library and the use of half life uncertainties in addition to cross section uncertainties to calculate the uncertainties on radiological quantities Options to consider either source of uncert
144. clide from the parent Table A7 1 Possible bremsstrahlung nuclides Half life y Branching ratio Be 1 60 10 gt p 132 0 3 01 10 9 AT 269 0 K 1 26 10 Ar 33 00 K 33 00 CET 10 75 Sr 0 14 Sr 28 79 28 79 0 16 Zr 6 00 107 mND 0 18 Rh 1 02 108A g 418 0 UO Ag 0 68 aes Sl 14 1 BE 0 12 Bir 0 14 Ty 4 41 10 15 0 12 123 0 35 Te 0 30 2 00 1014 PCs 30 04 Ce 7 00 10 V Ce 5 00 10 p 0 78 148 0 11 Ig 1 40 10 Yb 1 30 104 Lu 0 44 Tm 0 35 186 2 00 10 0 19 Os 5 60 10 241 0 Ps 5 99 Hg 22 2 3 79 anor 3 00 10 22 2 25 21 77 SAW 14 33 TT 8 00 107 BITh 7 04 108 26Np 1 52 10 32Q Si 90 Sr 957t 106Pu 108m4 g NOMA g 14mm imp 100 0 100 0 100 0 100 0 100 0 100 0 1 9 10 100 0 100 0 2 5 10 99 9 100 0 User Manuat Issue 1 Feb 2007 FISPACT 109 Np 7365 0 Am 100 0 2 1 60 107 Cm 100 0 22Am 141 0 252m Agi 99 5 25Am 0 88 BK 1 4 10 GE 898 0 Appendix 8 Pathways FISPACT calculates the inventory of nuclides after irradiation with no reference to the actual paths that are followed in the production of the various nuclides The code can be used to carry out a sensitivity calculation to determine by how much the amount of a nuclide will vary if a particular cross section or half life is varied However although very valuable for i
145. connected to stream 8 see Table 2 contains these associations EAF P FIS 2007 EAF P FIS is taken completely from the UKFY 4 0 fission yield library and FISPACT reads the file in ENDF B VI format with no pre processing Only 19 of the 90 nuclides in EAF P XS which have fission cross sections have any fission yield data in UKFY 4 0 at relevant energies For the remainder a neighbouring fission yield is used The file connected to stream 8 see Table 2 contains these associations EAF HAZ 2007 UKAEA Fusion Activity is one quantity used to judge the potential hazard of an irradiated material However activity takes no account of the biological impact on human beings To enable FISPACT to give some indication of the potential biological hazard of irradiated materials a library of dose coefficients has been assembled which determine the dose received by a man over his User Manuat Issue 1 Feb 2007 FISPACT 103 EAF A2 2007 lifetime 50 years following the ingestion or inhalation of 1 Bq of activity of a particular radionuclide The basic sources for these data are reports published by 927 and the However these sources primarily cover radionuclides generated by the fission power producing community and consequently only cover some of the nuclides that can arise in fusion applications In order to extend the range of nuclides to all those in EAF DEC it has been necessary to use an approximate
146. ction values are stored with a negative sign so that it is possible to avoid double counting when summing the cross sections only sum positive values but when considering gas production the absolute value of the cross section is used The output of this processing is stored in the COLLAPX file which contains parent daughter and cross section values To validate the collapsing of the cross section library it is necessary to do several calculations and compare with values from a COLLAPX file Appendix 14 describes the COLLAPX files for each type of group library the 100 group neutron spectrum is used here for the validation In the validation report for FISPACT 99 a listing of the procedure that was used to calculate the collapsed values was given This is not repeated for this Appendix The collapsed cross section O is calculated as indicated by equation 1 where o are the group averaged cross sections and the 1 corresponding group flux 0 2 06 2 0 D 1 Table 6 presents a small part of the COLLAPX file for EAF 97 in the 100 group spectrum The nuclides corresponding to these ID numbers are shown in Table 7 Table 8 presents the relevant reactions present in the EAF 97 library with the calculated collapsed User Manuat Issue 1 Feb 2007 FISPACT 205 Table 6 A portion of the COLLAPX file for 100 group neutron spectrum Parent ID Dauehter ID Cross section n b L 20 8 20 21 106530
147. ctra in these cases the flux specified for the region must be that which would be present if the first wall loading shown in the file was present on the first wall It is recommended that FLUX is always used in preference to WALL unless the user has a run that makes its use essential This code word is used to reset the time value to zero after an irradiation After ZERO the output will show COOLING TIME rather than TIME in the title for the interval The flux MUST be set to zero by the code word FLUX with parameter 0 0 prior to the use of ZERO This code word MUST be used after an irradiation if the code word GRAPH is also used the input file comment UKAEA Fusion In versions of FISPACT prior to 3 0 it was not possible to include any comment lines in the input file This can now be done by enclosing the comment in double angle brackets lt lt gt gt Such a comment can be included anywhere in the input file that a code word would normally be used however it MUST not occur in the middle of a code word parameter combination Examples of the use of this construction follow FLUX 1 2E14 This is the first wall flux LEVEL 50 1 This is a correct usage of a comment User Manuat Issue 1 Feb 2007 FISPACT 59 FLUX This is the first wall flux gt gt 1 2 14 This is incorrect usage of a comment Examples of preliminary input There are basically three types of
148. dentifying important reactions and in the calculation of errors this method requires a great deal of computer time and the results are not always easy to understand To overcome these problems the method of pathway analysis has been implemented in FISPACT This technique enables the percentage of the final nuclide produced by a particular pathway to be calculated Although the concept of pathway is intuitive it requires careful definition In this context a pathway refers to a series of nuclides all of which are distinct joined in a linear fashion by links which either represent reactions or decays There are no additional entry points on the pathway and all the depletion modes for each nuclide can be assumed to go to a sink and play no further part in the pathway Note that because of the way that the first five nuclides in the decay library gas nuclides are labelled it is possible for these nuclides to be repeated in a pathway This inconsistency is of little practical importance but is noted as an area for improvement in a future version of FISPACT An example of a pathway is the production of Ni from gt 9 Co IT 9Co B 99Ni n2n Ni This is the most important pathway in the first wall of the EEF fusion device contributing see reference 5 54 of all the Ni formed from This pathway contains 4 links consisting of User Manuat Issue 1 Feb 2007 UKAEA Fusion 110 FISPACT 59
149. dge energies are included and identified and values of Up and are given just above and below each discontinuity to facilitate accurate interpolation Somewhat different values for the atomic photoeffect cross section have been used for Z 2 54 For compounds and mixtures values for can now obtained by simple addition i e combining values for the elements according to their proportions by weight Radiative losses are now included The total cross section per atom which is related to can be written as the sum over contributions from the principal photon interactions Otot Ope Ocoh Oincoh Opair i Otrip U Ophn where Ope is the atomic photoeffect cross section Ocoh and Oincon are the coherent Rayleigh and incoherent Compton scattering cross sections respectively Opair and are the cross sections for electron positron production in the fields of the nucleus and the atomic electrons respectively and Ophn is the photonuclear cross section However the latter contribution has been neglected as well as other less probable photon atom interactions The library EAF ABS 2007 contains cm g for all elements in increasing Z order u m and Uen P cm g lfor air and the mean energies of the 24 group structure The value of u the material is calculated from elemental values 4 u using equation A3 2 mj Les 2 o p EN A3 2 User Manu
150. dly front end to the main task of number crunching FISPACT uses external libraries of reaction cross sections and decay data for all relevant nuclides to calculate an inventory of nuclides produced as a result of the irradiation of a starting material with a flux of neutrons The actual output quantities include the amount number of atoms and grams the activity o B and y energies kW ydose rate Sv h the 1 To stop the text becoming unwieldy neutrons actually means neutrons protons and deuterons in the this report unless otherwise stated User Manual Issue 1 Feb 2007 UKAEA Fusion FISPACT UKAEA Fusion potential ingestion and inhalation doses Sv the legal transport limit A value the clearance index and the half life for each nuclide Amounts and heat outputs are also given for the elements and the y ray spectrum for the material is listed as well as various summed quantities such as total activity and total dose rate At the end of each time interval the dominant nuclides in terms of activity heat ydose rate potential biological hazards and clearance index and the pathway data for the production of these nuclides can be shown The uncertainties in eight total radiological quantities can be calculated and output As options data files can be produced for subsequent use by other programs to plot graphs of the total responses as functions of elapsed time and selected blocks of
151. e constructed by filling in a series of dialog boxes The connections to the input output streams can be set up simply by a dialog box An output file of arbitrary size can be viewed and searched A summary of an output file showing any of the total quantities e g activity or dose rate or values for a particular nuclide at each of the time intervals can be produced The summary of total quantities can be written to a database file The lists of dominant nuclides can be summarised Pathway information can be summarised Summary information can be placed on the clipboard for pasting to another application such as an Excel spreadsheet A log log plot can be displayed of any of the five possible FISPACT graphs activity heat y dose rate and ingestion or inhalation dose as functions of time Feb 2007 UKAEA Fusion 178 FISPACT e The graph can be printed on any Windows compatible printer with various options The PC version of FISPACT can be run EASY decay data can be viewed interactively e EASY multi group cross section data for neutrons deuterons and protons can be viewed interactively Neutron deuteron and proton spectra used by FISPACT be stored and plotted Elemental or isotopic compositions for a range of materials can be stored viewed and written to an INPUT file e A Windows Help file covering the application and containing much of the present manual is available
152. e summary values are shown for the total mass of material not for a unit mass maximum number of time intervals 200 can be printed in the summary In cases where more than 200 time User Manuat Issue 1 Feb 2007 UKAEA Fusion 74 FISPACT intervals are considered only the final 150 200 time intervals are displayed the earlier ones usually not important for pulsing scenarios are discarded A feature added in FISPACT 2001 is a second summary at the end of the case The intervals are listed as irradiation steps or cooling steps in both the most appropriate unit sec min days and in years Four columns present Beta Heat kW Gamma Heat kW Mean Beta Energy MeV and Mean Gamma Energy MeV For all quantities the estimated uncertainty is also given Note that this summary table is only displayed if the SPLIT code word is used with its parameter set to 1 PRINTLIB output Fe 52m D UKAEA Fusion Mn The PRINTLIB output contains five sections Firstly a summary of the decay data for each nuclide is given with thirteen nuclides listed per page For each nuclide its internal identifier number the decay constant A s and the half life in appropriate units for stable nuclides or Dee Py is printed are given followed by the number of spontaneous fission neutrons per second and the number of neutrons from oun reactions the latter value is always 0 0 and is included for comp
153. e 7 Notes documents in italics are not UKAEA reports There is no report on the EAF 2007 n y reactions however reference 16 contains data for EAF 97 There is no EAF 2007 Report file however reference 18 contains data for EAF 99 The first three processing reports in EDS 4 are of historical interest only for EAF 2007 The FISPACT 99 validation report reference 2 was available as a separate document User Manual Issue 1 Feb 2007 UKAEA Fusion 200 FISPACT Appendix 19 Validation Introduction A series of test cases for FISPACT is discussed in Appendix 14 These test cases cover the various code words and by comparing with the outputs from previous versions ensure that a new version of FISPACT is behaving correctly However to fully test a new version it is necessary that the validity of the code is checked by comparing particular outputs with the original input data libraries to show that these data are input and processed correctly The numerical method employed by FISPACT can also be validated by comparing with cases where analytical results are available This Appendix details a set of tests to provide this validation it consists of two main parts the first that considers the data processing and the second that considers the comparison with analytical calculations To test FISPACT 2007 the data of EAF 97 can be used there is no need to update the input data to EAF 2007 since the comparison is against analytic
154. e not present in the cross section library have the correct libraries been used Uncertainty data not consistent with cross section data lt COL100 gt There are reactions in the uncertainty library that are not present in the cross section library have the correct libraries been used User Manual Issue 1 Feb 2007 UKAEA Fusion 142 FISPACT UKAEA Fusion Uncertainty data not consistent with cross section data lt COL172 gt There are reactions in the uncertainty library that are not present in the cross section library have the correct libraries been used Uncertainty data not consistent with cross section data lt COL175 gt There are reactions in the uncertainty library that are not present in the cross section library have the correct libraries been used Uncertainty data not consistent with cross section data lt COL211 gt There are reactions in the uncertainty library that are not present in the cross section library have the correct libraries been used Uncertainty data not consistent with cross section data lt COL315 gt There are reactions in the uncertainty library that are not present in the cross section library have the correct libraries been used Uncertainty data not consistent with cross section data lt COL351 gt There are reactions in the uncertainty library that are not present in the cross section library have the correct libraries been used Value for density must be given if FUEL used lt
155. e word HAZARDS MUST be used to ensure that uncertainties are correctly printed This code word causes the inventory data in columns 1 and 2 number of atoms and grams of each nuclide to be written to an external file TAB1 via stream A Note that stream numbers greater than 43 MUST be used for all TAB files This code word causes the inventory data in columns 3 and 7 activity and dose rate Sv h of each nuclide to be written to an external file TAB2 via stream B Note that stream numbers greater than 43 MUST be used for all TAB files This code word causes the inventory data in columns 8 and 9 ingestion and inhalation dose Sv of each nuclide to be written to an external file TAB3 via stream C Note that stream numbers greater than 43 MUST be used for all TAB files This code word causes gamma ray spectrum in MeV 5 in the 24 energy group format or 22 group format if the User Manuat Issue 1 Feb 2007 UKAEA Fusion 54 FISPACT TIME T GROUP parameter is 1 to be written to an external file 4 via stream ID In addition a second column showing the number of gammas per group is also given in 4 Note that stream numbers greater than 43 MUST be used for all TAB files This code word allows the input of the irradiation or cooling time interval T seconds The time may be input in units other than seconds by following the value with one of the following code words specifying the
156. eam connected 660 662 was not done properly if more than to LINS is properly closed 1 subinterval defined 23 10 96 A Macintosh version of FISPACT Modifications so that the Master 663 688 Is required file contains modifications for a Macintosh version 24 10 96 Need to add the facility to include Add this feature 689 712 half life as well as cross section uncertainties to the uncertainty estimation 29 10 96 Need to include the Build number Add this feature 713 717 721 and date into the code and the 727 728 output when compiling a new version User Manuat Issue 1 Feb 2007 UKAEA Fusion 194 FISPACT Date Problem Solution Modification numbers 6 11 96 Noted that the ENDF codes for Changes made in ENDFP 718 720 722 some of the more exotic decays are 726 not treated correctly 12 11 96 For runs where the neutron This feature added by introducing 729 734 spectrum changes it would be the code word NEWFILE which useful to be able to change the allows the name of the file on a name of the ARRAYX stream given in FILES to be and FLUXES files during the redefined course of a run 14 11 96 When using the OVER code word Corrected this in OVERID 735 736 it was only possible to specify an isomer n when it is a parent for the daughter the FISPACT identification number is still required 14 11 96 PRINTLIB output contains the Corrected this in INTERP 737 neutron fractions
157. eans that in addition to the standard pathway all the others of the form shown below are also considered Co n y CoAT Co B 9 Ni n2n Ni n y Ni n2n Ni User Manuat Issue 1 Feb 2007 FISPACT 111 Co TIT Co B 99Ni n2n Ni n y 9 Ni n2n Ni n y 9 Ni n2n Ni It is important to note that these parallel pathways MUST not be asked for explicitly the code will include them automatically whenever they are required isomer ground state Figure 8 2 Detail of a diagram for a pathway containing an isomer loop There is a further type of loop the inclusion of which is at the discretion of the user If one of the nuclides on the pathway has an isomeric state of short half life which decays back to the nuclide via an isomeric transition IT then a loop of the form X nann X IT X could also be included in the pathway replacing the X alone An example of this is shown in Figure 8 2 The user is able to decide which isomers should be considered by means of the LOOPS code word and this is often essential for the correct calculation of the pathway contribution if short lived isomers of any of the nuclides in the pathway exist Since version 3 1 a change has been made during the routine calculation of pathways at the end of each time interval By default the LOOPS code word is used with the time parameter set to the larger of 1 second or time interval 1000 For pathway calculations made using PATHS o
158. ection obtained by collapsing with the neutron spectrum followed by the percentage error Note that if there are no uncertainty data the library then the code word NOERROR switches the output in this section to include only the cross section The parent and daughter nuclides are given with a code representing the reaction between them These codes are User Manuat Issue 1 Feb 2007 UKAEA Fusion 3 n 5n 7 n d n n p n He n n 2p n pd n n t H n 2nd H n 3np n 3nt n 4nd n 5np n Snt n 6nd n 7np n 2p n n a n 2nh n dt n n pt n 3n2p n 3nq n 4nh n 5n2p n 5no 7 n po H n dh H n n ph n ho n n ta 4 n 3npo 20 n 2n20 n 4n20 2 42 n 30 n fission The fourth section FISPACT n 2n 4 8 n t n n d n 2np n o n n h n tp n n pd n 2n2p n 2nt n 3nd n 4np n 4nt n 5nd n 6np n 6nt n 3p n 2no n 3nh n n dt n 4n2p 4 n 6n0 n ph 4 n n 3p n do H n n pa H n n dh n to nn do 4 n n th 2 4 20 3 2 n 3n20 20 120 the nuclides that produce UKAEA Fusion bremsstrahlung radiation from energetic B particles The user may choose nuclides from here for the input file although the most im
159. els the parent nuclide and n label the decays a b It is assumed that at time 0 there are Noo atoms of nuclide 0 present This series of decays is equivalent to Figure A19 2 where the starting numbers of atoms of nuclide 0 for each of the parallel pathways is given Each pathway is a decay chain that can be solved by the Bateman method and the total number of atoms of each type then found by suitable additions of the values for each pathway The calculation for each nuclide contains a table showing the comparison between the analytical results labelled by and the FISPACT results labelled by The number of atoms of each nuclide N and the total activity for all relevant nuclides A are given at a series of decay times For each nuclide the User Manuat Issue 1 Feb 2007 UKAEA Fusion 208 FISPACT H decav test di FISPACT column contains the value of Na Na which has the largest absolute value over the considered times and the time where the maximum value occurred is indicated by the shading The FISPACT input files for all the tests are given in the Annex 3 89105 10 s stable B7 3 15576E 07 1 57788E 08 3 15576E 08 4 73364E 08 6 31152E 08 9 46728E 08 1 26230E 09 1 57788E 09 3 15576E 09 9 45335E 19 7 54967E 19 5 69975E 19 4 30312E 19 3 24871E 19 1 85168E 19 1 05542E 19 6 01558E 18 3 61872E 17 9 45335E 19 7 54967E 19 5 69975E 19 4 30312E 19 3 24871E 1
160. en to a file in standard format suitable for a processing by the EASY 2007 User Interface with uncertainty data included from this file three graphs can be subsequently plotted GROUP GAMGP 0 This code word allows details of the y spectrum to be input The default IGAMGP 0 means that y spectrum data are output in a 24 energy group structure This structure is also used when processing the decay data and in the internal calculations However if IGAMGP 1 then the output is in the 22 group Steiner energy structure The values of the energy groups are shown in Appendix 10 Note that the structure determined by GAMGP is also used when TABA is specified to produce a file of the y spectrum data An example of the use of this code word follows GROUP 1 In this case data will be output in 22 energy groups User Manuat Issue 1 Feb 2007 UKAEA Fusion FISPACT 32 55316 ITER 3D Div207w 10 10 MN Mee 55 v 52 UN 108 Mo9 N _ Tc 99m 2 Cu 64 vd N Co60 gt N 5 Ag Ign lt 10 ti Mat Imp Impurities Pure material i 105 Ni 59 Sec Min Hour Day Mh Year x Nb 91 93 10 10 10 Time after irradiation years Figure 5 Graphical output produced using the PV WAVE UKAEA Fusion visualisation package Visual Numerics User Manual Issue 1 Feb 2007 FISPACT
161. equilibrium quickly after a change in flux Thus the inventory at the end of the irradiation is very insensitive to changes in flux at the start If there are changes in flux near the end of the irradiation this will affect levels of the short lived nuclides dramatically having an off period followed by a 10 times higher pulse at the end of irradiation will give approximately 10 times more nuclide than a steady flux that lasts 10 times as long fluence conserved Thus modelling the irradiation carefully at the end of the irradiation is crucial while careful modelling at the start is unimportant Thus the recommendation about a non steady irradiation is to model in detail the final time before shut down for example by using the correct on off details with variable fluxes For the rest of the irradiation the bulk of it after the start use a continuous flux such that the irradiation time and the fluence are conserved Further discussion of this is given in references 51 and 52 Implications for FISPACT UKAEA Fusion FISPACT is able to model non steady irradiations of arbitrary complexity Indeed the loop construct PULSE and ENDPULSE was specifically added to aid such calculations However bearing in mind the conclusions of the previous section there would be little to be gained from using all of say 10 000 equal pulses it would be preferable to use a steady irradiation for 9 990 of them and consider only the final 10 in detail
162. er of atoms for a unit mass An example of the use of this code word follows MIND 1 0 5 In this case all nuclides with numbers of atoms 1 10 are reset to zero during the calculation User Manuat Issue 1 Feb 2007 FISPACT 39 MONITOR MONIT 0j The progress of a FISPACT run can be monitored by printing the various code words as they are read in the input file to the standard output The default is not to print this information but it can be switched on by setting MONIT to 1 MONITOR should appear near the start of the input file but after NOHEAD if that code word is used An example of the use of this code word follows MONITOR 1 In this case the code words in the input are echoed to standard output NEWFILE JSTRM NEWNAM This code word allows a new file name to be specified for files connected to particular streams thus overriding the name defined in FILES JSTRM is the number of the stream only streams 12 17 or 20 can be chosen NEWNAM is the new file name a maximum of 12 characters can be used Thus either the FLUXES or COLLAPX files may be redefined Note that the new file MUST be in the same directory as the original file given in FILES This facility is included so that cases where the neutron spectrum changes significantly during the course of an irradiation can be modelled If the neutron spectra at a series of irradiation times are known then it 15 possible to prepare correspondin
163. ered correctly in INPUT file Sub library flagged as other than decay data lt ENDFP gt The decay library being used is not in ENDF B V or VI format has the correct decay library been used This graph type not defined GRAPH Five graph types 1 5 can be specified This version of FISPACT cannot handle half life uncertainties UNCTYPE If a version of FISPACT prior to 97 is used then including the UNCTYPE code word will generate this error message Too many alpha decays lt OUTPUT gt More than five 0 decay modes found for a nuclide has the correct decay library been used Too many beta decays lt OUTPUT gt More than five B decay modes found for a nuclide has the correct decay library been used Too many dominant nuclides JNUMB gt 200 lt OUTERR gt The run requires a total of more than 200 dominant nuclides Either simplify the run or contact UKAEA for assistance Too many input nuclides lt OUTERR gt The code words MASS or FUEL have been used to specify the material to be irradiated Only 300 nuclides may be input if MASS is used then there may be too many naturally occurring isotopes for the input elements Reduce the number of input elements or nuclides Unable to open FILES from MAIN The file FILES contains the names of all the other files required by the system it was not available Uncertainty data not consistent with cross section data lt COL069 gt There are reactions in the uncertainty library that ar
164. error of a quantity is defined as the maximum estimate best value thus o f o The relative or fractional error A is the error best value thus A 6 and f A If a cross section was known to 20 then A 0 2 f 1 2 and 0 2 o The value of one of the radiological quantities at a particular time Q is given by equation A9 1 0 4 9 1 where q is the value of the quantity for nuclide 7 and the sum is over all dominant nuclides The fractional contribution c of each dominant nuclide is given by equation 49 2 The error on the quantity Q AQ is given by equation A9 3 2 AO E A A9 3 i The radiological quantities are linearly dependent on the number of atoms present as shown by equation A9 4 where is the number of atoms of nuclide The error on the quantity g is given by equation A9 5 AN Aq 2 Quot n aa TU E A9 5 1 dominant nuclide can be produced by a set of parallel pathways the total number of atoms of i 15 given by equation 9 6 N 2 A 9 J where is the number of atoms of i formed by pathway j and the sum is over all pathways The fractional contribution d of each pathway to i is given by equation A9 7 User Manuat Issue 1 Feb 2007 UKAEA Fusion 116 FISPACT UKAEA Fusion QN
165. etailed inventory is produced The final pulse end of irradiation has a detailed inventory since ATOMS is used RESULT NRESU UKAEA Fusion SVM I X I l 1 NRESU This code word is used when calculating pathways The pathway output includes the percentage of the total amount of the daughter nuclide produced by a particular pathway One way to obtain this total amount is to do an inventory prior to the pathway calculation However it is much easier to be able to do the inventory in a separate run and then manually to use results from that inventory and input them into the pathway calculation NRESU nuclides are specified and for each the identifier SVM I e g TEI29M and the number of atoms X I are specified If ATOMS or SPECTRUM is not present then RESULT is necessary to start the pathway calculation and so MUST follow the code word PATH or ROUTES User Manuat Issue 1 Feb 2007 FISPACT 49 An example of the use of this code word follows 356E19 560E17 568 12 The number of atoms of C N and N produced in a standard inventory run are specified ROUTES PAR DAU NMAX PMIN IPRPA In addition to specifying a particular pathway with the code word PATH the code word ROUTES can be used This will search for all pathways from the parent nuclide PAR to the daughter nuclide DAU with a maximum of NMAX links reactions or decays The contribution of each pathway is calculated and if the nu
166. eti e ei abes 40 NOERBOR nc Eheu NIE C A aaa 40 NORB SSU Ape T x A A A 4l MOMPADA S Lone Me Ee 41 NOBONI a 41 M E 42 NOTI II 42 ta EET 42 IO Nel A Me qr 42 M MEE 42 81025 m Daal 43 ARTUR NPA setae 44 45 DRUMS LAM CAS A a CLER 46 PROC DEE NEROS SIL SSL Su tros tas ro indian tana 47 PU SEM M s Met ta i LM assia a 47 RESULT NIESULS I iecit pio te e eo ita E td te 48 ROUTES PAR DAU NMAX PMIN IPRPA ene etre a ei 49 SENSITIVITY XSENS XNSENI INSEN3 INSENA cese 50 SEOQNUMBER ENNM TITO e n i can s tta dera ut ias 51 SEQUENTIAL JSEQUE 0 IPCWRT 0 52 SPEC TRUM e Lots MSIE tu ea mM ME mut AG ah B 52 cana hamana Q aa B a q 53 TABIC Duc oec ab eec ads 35 TAB2 IB Reet EE 53 TAD eost ea n e Rune ea ent eM 53 ai 53 PME eno DE 54 UNCERTAINTY UNCER 0 lt FR
167. etting GENER to 0 More information on pathways is given in Appendix 8 GRAPH NUMG GRSHOW GUNCRT UKAEA Fusion NOPT I l 1 This code word specifies what information is stored in the file GRAPH for subsequent post processing The number of graphs required NUMG is input and for each graph an option number NOPT I is read Allowable values for the options are Total Activity Total y dose rate Total heat output Ingestion dose Inhalation dose nan RC Tr The parameter GRSHOW allows slightly different versions of the data file to be constructed If GRSHOW 0 then an output suitable for PC post processing is obtained if GRSHOW 1 then the output might be more suitable for other platforms The recommended method of producing graphs on a PC is to use the EASY User Interface Appendix 15 gives details of the Interface and gives a screen shot of a typical graph An example of a total dose rate output graph produced by a plotting package on a UNIX workstation is shown in Figure 5 The third parameter GUNCRT allows the user to specify if uncertainty data should be 1 or not be 0 written to the GRAPH file If the uncertainty data are written then the plotting routines can display the uncertainties on all five types of plots User Manuat Issue 1 Feb 2007 FISPACT 31 An example of the use of this code word follows GRAPH 3 0 1 In this case data on activity Y dose rate and ingestion dose are writt
168. for plotting up to five graphs to a file This file then has to be processed by a separate package to produce hard copy This process is extremely dependent on both the hardware and software available to the user In order to make standard FISPACT graph plotting more accessible to users the EASY User Interface includes the capability to display graphs and to print them on Windows supported printers Options are available on the menu to alter the appearance of the graph adding error bars or an uncertainty band if uncertainty data are included in the GRAPH file adding gridlines and changing colours and sizes For the y dose rate graph the option to include exclude the bremsstrahlung contribution is available as is the option of adding lines to indicate recycling 10 mSvh and hands on 25 uSvh limits Similar limits the ILW LLW and IAEA non radioactive limits can also be added to the activity plot Figure A15 2 shows a typical graph with activity for Eurofer steel plotted The uncertainty band and common times options have been selected User Manuat Issue 1 Feb 2007 FISPACT 181 EASY 2007 User Interface Graph E Fispact v_2005 ifmif_Eur g a Ele Edit Type Options View Window Help 8 x D Ig Es Fe Activity Bgkg 1 4 1 3 1 2 1E 1 1E0 1 1 2 1E3 1E4 Time after irradiation years E 1j Figure A15 2 The graph window of the EASY User Inte
169. forms of which equations A9 46 and A9 47 are special cases Pathways in which the final nuclide reacts and decays In the derivations so far it has been assumed that the final nuclide in the pathway neither reacts nor decays In general this will not be true and it is necessary to determine what additional User Manuat Issue 1 Feb 2007 UKAEA Fusion 126 FISPACT UKAEA Fusion factor would be introduced in both limits by allowing the constraint on the final nuclide to be removed Firstly assume that all the nuclides are long lived The limit of this case is trivially obtained from equation A9 50 there is no effect on the limit of the final long lived nuclide Secondly assume that all the nuclides are short lived As can be seen from equation A9 38 setting the exponential of all terms including the final one to zero means that the limit for the number of atoms of the final nuclide is also zero as expected physically since they have all decayed However this is not very helpful and a physical argument is required The pathway for the production of a final nuclide will only need to be considered if the final nuclide is long enough lived to contribute significantly to radiological quantities Thus the nuclide may be short lived compared to the irradiation time but it is not so short lived that all its atoms have decayed Thus the number of atoms of the final nuclide must be in secular equilibrium with the previous
170. g COLLAPX files prior to considering the irradiation At suitable times during the total irradiation NEWFILE can be used with one of the calculated COLLAPX files Thus if two different neutron fluxes exist and the collapsed files are collapx l and collapx 2 then collapx 1 is connected to stream 12 in FILES and the relevant portion of the INPUT file is as follows User Manuat Issue 1 Feb 2007 UKAEA Fusion 40 FISPACT NOCOMP NOERROR UKAEA Fusion ENFA EAF 2007 with spectrum 1 ARRAY FISPACT 1st part of irradiation FLUX 1 3E15 LEVEL 50 1 TIME 1 0 YEARS SPECTRUM NEWFILE 12 COLLAPX 2 ENFA EAF 2007 with spectrum 2 ARRAY FISPACT 2nd part of irradiation FLUX 1 4E15 TIME 1 0 YEARS ATOMS In this example the formation of the ARRAYX file is followed by an irradiation for 1 year The SPECTRUM code word is used so that the detailed inventory is not produced The second COLLAPX file is specified by NEWFILE and the ARRAYX file is recalculated The irradiation is then continued for a further 1 year note that the flux value has also changed and a detailed inventory produced by ATOMS This code word causes the table of elemental compositions to be omitted from the inventory printout This code word stops uncertainty information from being used It should be used if a cross section library with no uncertainty component is being collapsed or if such a collapsed library 18 used with the UNCERTAINTY code word
171. g the code word NOSORT but this is not usually desirable as it removes the ability to calculate uncertainties or pathways If the first UNCERTAINTY parameter is 1 or 3 then the uncertainty estimates for the eight total radiological quantities are printed next The title for this section identifies what types of uncertainties contribute to the estimates The output for the activity is shown below Total Activity is 5 62518E 12 1 16 12 Error is 2 06E401 of ise Eel e The uncertainty is given both absolutely and as a percentage of the total Note that if no clearance data have been calculated then only seven sets of nuclides are output As noted above there are probably more than twenty top ten nuclides in total and these are listed with the uncertainties in the following fashion the number of atoms the error in the number of atoms shown as E Atoms in the output similar columns for the other four total quantities and then a column giving the percentage error Note that numbers of atoms for clearance index and beta and gamma heats are not shown Bremsstrahlung corrections If the code word BREMSSTRAHLUNG is included in the input then the contributions to the dose rate from the nuclides specified are given here First details of the configuration infinite plane or point source are given and then the data in the following form Bremsstrahlung dose rate from 10 is 7 97779E 13 Sv h 7 97779 11 Rems
172. h This is 6 84908 08 of the total dose rate The dose rate from the particular nuclide is given in units of both Sv h and Rems h and then the percentage contribution to the total dose rate is given Note this total does not include User Manuat Issue 1 Feb 2007 UKAEA Fusion 72 FISPACT the bremsstrahlung corrections but in the following line a total of the y dose rate and all the bremsstrahlung corrections is given Pathway analysis If the first parameter following the UNCERTAINTY code word is either 2 or 3 or the code words PATHS or ROUTES are used then pathway analysis information is output A short key describing the way links are shown is given first The paths show REACTIONS as eg n p REACTIONS where projectiles are followed as p n type n X REACTIONS of sequential charged particles as eg DECAYS as eg b and DECAYS forming He4 as X Pathway shows type of link R r D d and if final nuclide is L ong or S hort lived Note that even if pathways for a deuteron or proton library are being printed this key is unchanged For standard reactions such as 5Ti n p Sc the use of n p is obvious although because of the lack of a Greek font a is used for and is used for y If the projectile is followed in order to calculate gas production then the above reaction would be shown as n X meaning Ti n 5 Sc H Decays are shown as b b and IT
173. h one isotope only since we require the pathways from a parent isotope not an element irradiated in a first wall flux of 5 MW m for 2 5 years The LEVEL parameters are set to 100 and 1 All the pathways with a maximum of 6 links between and and 6Zn and SNI and in addition the two specified pathways between Zn and User Manuat Issue 1 Feb 2007 UKAEA Fusion 64 FISPACT UKAEA Fusion Fe and 97 and Ga will be calculated The RESULT code word enables the amounts of and Ga that were calculated in a previous inventory run to be input The run is finished by END AINPUT FISPACT Irradiation of C EEF Zone 12 MASS 1 0 1 C 100 0 DENSITY 2 0 DOSE 2 5 0 GROUP 1 MIND 1 0E5 OVER 9 ACROSS HE6 1 05490E 2 FLUX 1 5E15 LEVEL 50 1 TIME 2 5 YEARS UNCERT 2 5 END END of over run This case models the irradiation of 1 kg of carbon for 2 5 years in a flux of 1 5 10 n cm si with the cross section of the reaction Be n o He set to 10 549 mb by the OVER code word The density of carbon is input 2 0 g by the DENSITY code word and the y dose rate output is requested for a point source by following DOSE with 2 at a distance of 5 m the spectral data is requested by following GROUP with 1 in the 22 group format The pathways and uncertainty estimates are requested by following UNCERT with 2 after the inventory ATOMS causes the inventory to be output a
174. he use of this code word follows PATH 5 48 R CA45 D SC45 R 42 D CA42 R CA41 The percentage of the daughter nuclide that is formed by the pathway Sc n o K B 1 21 is given in output PRINTLIB PRINT UKAEA Fusion This code word causes the output of the data libraries in a readable form The output consists of five blocks of data the contents of each are 1 decay data including fission yields if appropriate for each nuclide 2 the branching ratios of decays for each radionuclide 3 the cross section data including uncertainties for each reaction in the specified neutron spectrum 4 nuclides which will give a bremsstrahlung contribution to the y dose rate 5 the neutron spectrum used to collapse the cross section library The value of the parameter PRINT determines which blocks are output All five blocks Block 1 only Blocks 2 3 4 and 5 Block 5 only on User Manual Issue 1 Feb 2007 FISPACT 47 Note that if no uncertainty data exists in the library then the code word NOERROR MUST be used before PRINTLIB Note it is recommended that a separate FISPACT run giving a library output and no inventory be done for each decay data library and kept for reference An example of the use of this code word follows PRINTLIB 1 The library data for decays half lives average energies V spectrum and fission yield are output
175. ies of total radiological quantities would be too large An alternative approach followed in FISPACT involves a simplified procedure that nevertheless gives an estimate adequate both for the quality of the cross section uncertainty data and for fusion applications The main steps in the simplified procedure are 1 At the end of each time interval FISPACT lists the 20 most important nuclides for each of the eight radiological quantities activity heating y dose rate potential ingestion hazard potential inhalation hazard clearance index and beta and gamma heat outputs and the percentage contribution that each makes 2 For each of these dominant nuclides the pathways and fractional contributions are calculated 3 The uncertainty for each pathway is calculated using the sum of squares of the errors of each reaction 4 The errors for all parallel pathways contributing to a particular dominant nuclide are calculated 5 The errors for each dominant nuclide contributing to a radiological quantity are calculated In order to describe each step mathematically several terms need to be defined When deriving systematics of 14 5 MeV cross sections it was noted that the quantities log Gexpt were approximately normally distributed and that the best representation of error limits on Oca Were Ocak f and Ocak f User Manual Issue 1 Feb 2007 FISPACT 115 where fis termed the error factor The
176. ime without fluctuation The simplest approach to modelling such a situation is to use a continuous flux of neutrons for the total time of irradiation 7 such that the fluence in the model is the same as the measured fluence This simple approach conserves both the fluence and the irradiation time This will be adequate in many situations e g if waste management issues are being studied especially if the times of interest for the inventories are long compared to 7 However if the inventories at short times after the end of the irradiation are needed as for safety studies then the amounts of short lived radionuclides need to be accurately calculated Consider the cases of short and long lived radionuclides time scales are always considered relative to 7 formed by a reaction on a target that is initially present The long lived nuclide will build up in a linear fashion during the irradiation as the long half life means that little will have decayed In contrast the short lived nuclide will initially rise linearly but then saturate as equilibrium between formation and decay secular equilibrium is reached Thus at the end of the irradiation assuming that it is steady prior to shut down the effect of User Manual Issue 1 Feb 2007 UKAEA Fusion 228 FISPACT fluctuations of flux near the start will have little effect the long lived nuclide will build up when ever the flux is present while the short lived nuclide will reach
177. in this section This code word starts the calculation of the inventory equations over the time interval specified and causes the results isotopic elemental spectral pathway and uncertainty to be output It is the standard method of producing output other options are SPECTRUM and RESULT This code word causes data on the legal limits of activity for transport of radioactive material to be input for the calculations to include these data and for the results for individual nuclides and summed values to be output Appendix 6 contains more information on these data values Note that both ATWO and CLEAR MUST not be used in a particular case BREMSSTRAHLUNG ARG NUCLB JJ JJ 1 IARG This code word allows the input of the number ARG of nuclides and the identifiers NUCLB JJ for each of the nuclides The identifier should be specified using the format TE129M When the output is printed this code word causes the bremsstrahlung dose rate of each specified nuclide to be printed at the end of each time interval User Manuat Issue 1 Feb 2007 UKAEA Fusion FISPACT CLEAR Note a maximum of 25 nuclides can be specified and possible candidates if EAF 2007 is the decay library used are listed in Appendix 7 An example of the use of this code word follows BREM 4 CL36 AR39 AR42 K42 In this case the bremsstrahlung contributions of a OF 39 Ar Ar and are calculated and output at the end of each time
178. includes only the isomeric transitions follow If DENSITY has been input then the activity is given in both kg and Ci The total heat production in kW is also split into and y components and the totals for potential ingestion and inhalation hazards are output For all these quantities the values excluding tritium are UKAEA Fusion User Manuat Issue 1 Feb 2007 FISPACT 69 also given If ATWO was input then the effective A value for the material is output If CLEAR was input then the clearance index for the material is output The number of fissions and the burn up of actinides are output these are only non zero if actinides are initially present The initial mass in kg and the neutron flux in the interval are also included The composition of material by element is given next the column headings for this are number of atoms of the element number of gram atoms number of grams heat output Curie MeV and kW y heat output Curie MeV and kW and a heat output Curie MeV and kW Gamma spectrum DOSE RATE PLANE In this section the total energies MeV 51 from and yradiations and the total number of spontaneous fission neutrons are listed followed by two columns giving the y spectrum MeV s per group and number of gammas group 787 in either 24 or 22 group form depending on the parameter used for GROUP The total dose rate is then given in one of two forms dependan
179. individual pathways the sum of all in this case four of the pathways is given If the parameters in the code words have been set up correctly then the total of the listed pathways should be close to 100 Following the pathways information on the generic pathways is listed unless the GENERIC code word is used to switch it off The information in the sections above is repeated for each time interval but note that some of the above are only applicable for the first irradiation period A difference between irradiation and cooling intervals is that the title of all intervals where the flux has been set to 0 0 and the code word ZERO has been used will contain COOLING TIME rather than TIME Following the end of the output for the various time intervals the code word END triggers the end of the case with details of the CPU time used and then the run ends End of case summary A new feature added in FISPACT 4 1 is a summary at the end of the case containing the total values for each time interval The intervals are listed as irradiation steps or cooling steps in both the most appropriate unit sec min days and in years Six columns present Activity Bq Dose rate Sv h Heat output kW Ingestion dose Sv Inhalation dose Sv and Tritium activity Bq For all except the latter the estimated uncertainty is also given Following this the mass of input material kg and the density g are shown Note that th
180. input Size of ERR increased test on and fission is included value of JNUMB Some rare reactions did not New MT numbers defined and 1066 1109 contribute to gas production effect on gas production made consistent 20 6 01 Differences between PC and UNIX Inconsistent treatment of variable 1110 1280 outputs types in PC version Many variables made Double precision by default 25 6 01 Gamma Becquerel value in All decay modes made consistent 1281 1283 p OUTPUT is not consistent for all decays pathways arrays increased unstable nuclide could cause crash 4 12 01 Generic pathways resulting from ISIGN2 introduced 1292 1293 n IT m IT g links not correctly handled Need to be able to read 211 group Subroutine COL211 added 1294 1464 data files Subroutine COL351 added 1295 1465 data files 4 9 03 Larger number of nuclides in index Fe 48 now the fictional nuclide 1296 1297 1452 1461 mode 1463 2 10 03 Add definitions for these reactions 1435 2 10 03 Gas production caused by new Add these reactions 1436 1451 MU 24 10 03 Changes required in existing Make the changes 1466 1509 En oe 122271 51 groups fission products new decay files causes a crash energy release data causes a crash Exotic decay modes in decay file Define pp SF modes 1528 1543 cause a crash UKAEA Fusion User Manuat Issue 1 Feb 2007 FISPACT 197 Date Problem Solution Modification numbers 3 11 04 Need to include effect of SCPR f
181. ion of new nuclear data libraries for the treatment of sequential x n reactions in fusion material activation calculations Fus Tech 24 277 1993 4 S Ravndal P Oblozinsky S Kelzenberg and S Cierjacks User Manual for the code PCROSS KfK 4873 1991 42 N E Holden CRC Handbook of Chemistry and Physics 71st edition 1990 43 J K Tuli Nuclear Wallet Cards 6 Ed NNDC 2000 User Manuat Issue 1 Feb 2007 UKAEA Fusion 234 FISPACT 44 R A Forrest The European Activation System EASY 2007 Overview UKAEA FUS 533 2007 45 J A Simpson J Ch Sublet and D Nierop SYMPAL User guide UKAEA FUS 356 1997 46 J A Simpson and J Ch Sublet SYMPAL Utilities guide UKAEA FUS 357 1997 47 R A Forrest and J A Simpson SAFEPAQ User manual UKAEA FUS 355 1997 48 R A Forrest SAFEPAQ II User manual UKAEA FUS 454 Issue 7 2007 49 H Bateman Solution of a system of differential equations occurring in the theory of radio active transformations Proc Camb Phil Soc 16 423 1910 50 E T Cheng R A Forrest and A B Pashchenko Report on the second international activation calculational benchmark comparison study INDC NDS 300 1994 5 R A Forrest EASY a tool for activation calculations Fus Eng Design 37 167 174 1997 52 R A Forrest Nuclear data for fusion calculations Int Conf Nuc Data for Sci and Tech Ed J
182. ions in the run OVER C14 ERROR C13 1 10 Here the error factor for the C n 2n PC reaction is set to 1 10 for all subsequent calculations in the run Note that the ARRAYX and COLLAPX files are not altered so that in subsequent runs the cross section half life or error factor will revert to its original value Note a comment should not follow directly after OVER or ACROSS ALAM or ERROR ensure that at least one other code word is used before using a comment PARTITION NPART UKAEA Fusion SYM N XPART N N21 NPART This code word allows the material to be split or partitioned into two streams during an irradiation or cooling The part that continues to be considered by the code consists of all elements not specified NPART elements are specified and the fractions XPART N of the specified elements SYM N The stream containing the remainder is lost and cannot be followed any further by the code Typical uses of this code word might be to model recycling of irradiated material or the loss by diffusion of tritium from a User Manuat Issue 1 Feb 2007 FISPACT 45 PATH NLINK material In the first case PARTITION would be used after irradiation and cooling and would model the loss of volatile elements during re fabrication In the second case the irradiation might be split into several intervals PARTITION used in each interval to model the loss of tritium An example of the use of this code word foll
183. is important to stress that FISPACT is controlled by means of a series of code words that are collected together in an INPUT file These code words describe the type of the run the material that is to be considered and what type of output is required Many data files are required by FISPACT and the location of these files on the users system must be defined this is done by means of the FILES file described in the following section The results of the FISPACT run are written to the OUTPUT file User Manuat Issue 1 Feb 2007 UKAEA Fusion 14 FISPACT UKAEA Fusion This general structure can be represented by the diagram shown in Figure 1 HH Figure 1 Overview of files used by FISPACT Users of FISPACT are recommended to build up a collection of input files for performing typical tasks such as collapsing producing a library printout and standard runs under varying conditions and to alter the FILES file so that the required INPUT file is used Similarly several FILES files can be stored for various libraries or tasks and then these renamed so that the one to be used is named FILES and is located in the directory where the FISPACT executable resides For example the files FILES 1 FILES 2 FILES 3 and FILES exist in the folder containing the FISPACT executable isp20070 exe If data held in FILES 1 are required for a run then rename FILES to FILES 4 and FILES 1 to FILES before running FISPACT Users are reminded
184. is important when inputting times that it is the interval time not the total time that is specified Thus for cooling steps the time printed on the inventory is the sum of all the previous cooling time intervals after the code word ZERO Examples of the use of this code word follows ZERO TIME 2 5 YEARS ATOMS TIME 7 5 YEARS ATOMS Following irradiation the start of cooling is specified by the code word ZERO Inventories at the times of 2 5 and 10 years are output Figure A15 10 A Help topic ofthe EASY User Interface UKAEA Fusion User Manuat Issue 1 Feb 2007 FISPACT 189 Appendix 16 Density and abundance data All data on decay properties and cross sections are read in by FISPACT from external libraries However some basic physical quantities are held internally in FISPACT primarily to enable the numbers of atoms of isotopes to be calculated when amounts of elements are input These values are given in Table A16 1 The sources of the data are from references 42 and 43 Table A16 1 Data held internally in FISPACT Atomic Atomic Density Mass of first Abundance number weight g cm stable isotope 1 00794 0 0708 99 985 0 015 4 002602 0 1221 0 000137 99 999863 3 6 941 0 534 7 59 92 41 4 9 012182 1 848 9 100 0 5 10 811 2 34 10 19 8 80 2 6 12 011 2 1 12 98 89 1 11 7 14 0067 0 808 14 99 634 0 366 8 15 9994 1 14 16 99 762 0 038 0 2 9 18 998403 1 111 19 100 0 10 20 1797 1 2015
185. is long lived 1 Note there is an additional numeric factor of 1 n where number of long lived links n total number of links FISPACT uncertainties In the case that only uncertainties in the reaction cross sections are considered then the analysis above gives justification for the form of equation A9 9 If uncertainties in decay constants are also included then it is necessary to specify the type of each nuclide in the pathway since the final uncertainty will depend on whether each nuclide is long or short lived The notation L used in reference 39 for a reaction on a long lived nuclide in a pathway would be inconvenient to use in computer output Therefore the notation shown in Table A9 4 is proposed to describe pathways User Manuat Issue 1 Feb 2007 UKAEA Fusion 128 FISPACT Table A9 4 Notation for pathway description Symbol Description Reduced Number factor of links Link is reaction from a long lived nuclide EE Link is reaction from a short lived nuclide D Linkisdeayfromalonglivedmwlide amp m d Linkis decay from a short lived nuclide 1 m L Fimlmoideislongtived 1 S Final nuclide is short lived a UKAEA Fusion The notation for one of the 4 link pathways with a final nuclide long lived is written as RRrD L where the symbols for the links are enclosed by while the symbol for the final nuclide is enclosed by
186. isplayed at the end of case Updated physical The values of the amu the Avogadro New in version 2001 constants constant abundances and atomic masses have been updated New reaction types n 3no npo n 2n20 n d20 New in version 2003 included n t2a n nt20 n 30 n n3o reactions defined 211 VITAMIN J format New in version 2005 351 TRIPOLI format New in version 2005 Neutron spectrum Arbitrary group structure can be New in version 2005 converted to the new 211 or 351 group structures New decay modes n pp B B SF modes defined New in version 2005 User Manuat Issue 1 Feb 2007 UKAEA Fusion 10 FISPACT Feature Details Comments New reaction types n 5n n 6n n 7n n 8n n 3p New in version 2005 included n do nto nho n ph 4 nj5no n 7no n nto n 2nt n 3nt n 4nt n 5nt n 6nt n 3n20 n 4n20 4 reactions defined New charged p 2n h 2n amp 2n reactions defined New in version 2005 ee eee eG Sequential charged Sequential charged particle reactions can New in version 2005 particle reactions at be included for runs using 175 and 211 high energies group data Deuteron and Type of library determines the form of New in version 2007 proton induced many text string such as n 2n and reaction data Neutron UKAEA Fusion User Manual Issue 1 Feb 2007 FIS
187. its to which the input and output files are connected is also required The code words in both categories are described in the following sections in alphabetical order The code words are shown in BOLD type with details of the various parameters that accompany some of them shown in an TALIC font If a parameter is optional or only applicable under certain circumstances it is enclosed in angle brackets lt gt If default values are assumed if the code word is not used then these are shown after the parameters in curly brackets Many of the code words consist of many characters of which only the first four are read and so abbreviations may be used if wished AII the input is free format the user may include as many blanks white space between the words and parameters as desired so that the file is readable and easily understood So long as code words are input in the two categories there are usually no restrictions on the order or on repetition where a certain form or order must the used the reader is warned in the text by the word MUST Other typographical conventions that are used throughout the manual are the use of a bold COURIER font for file names and a COURIER font for actual computer commands The code words in the input file tend to follow chronologically the course of the neutron irradiation and User Manual Issue 1 Feb 2007 FISPACT 13 subsequent cooling with various options such as file dumps
188. ive cross section is stored as a negative value in order to identify this type of reaction and during the calculation this is converted back to a positive value before use This meant that the array A in which all the nuclear data are stored had to substantially increased in size and the convention has been adopted that the various files used for version 2 onwards have an X appended to them for identification There was no need to alter existing data libraries when moving to version 2 but a new index file INDEX had to be used which repeated the five gas nuclides at the end but with the ZA identifier indicating that the nuclide is a 3rd isomer Thus is labelled as 10010 at the start and as 10013 when repeated at the end of the index Gas production is also correctly followed in the pathway analysis so that the routes for the production of tritium can be calculated Dominant nuclides were given in a sorted list at the end of each time interval and the pathways for their production could be shown if required User Manuat Issue 1 Feb 2007 UKAEA Fusion 4 FISPACT Version 3 Version 4 Version 97 UKAEA Fusion Version 3 had two major new facilities Firstly it could use the uncertainty data for each reaction which became available as part of the European Activation File EAF 3 and the pathway data routinely generated at the end of each time interval to calculate the uncertainty in the total radiologi
189. lly performed by a simplified method The maximum value that can be specified for NERROR is 50 Examples of the use of this code word follow ERROR 2 LIT 118 1 0 BE9 HE6 1 0 Line 2 specifies that the reaction Li n y 8Li is to be considered Line 3 specifies that the reaction 0 6 is to be considered The uncertainty for both reactions is obtained from the uncertainty file ERROR 2 LI7 LI8 0 25 BE9 HE6 0 6 Line 2 specifies that the reaction Li n y 8Li is to be considered Line 3 specifies that the reaction is to be considered The uncertainty for the first reaction is set at 25 and for the second 60 The values in the uncertainty file are not used FISCHOOSE NCHO 2 FISCHO I 1 NCHO 0238 UKAEA Fusion PU239 When actinides are included as trace elements in a material then dominant nuclides that can be formed as a result of the fission of an actinide will be considered in the calculation of pathway User Manuat Issue 1 Feb 2007 FISPACT 27 information Although uranium thorium have been the only actinides input neutron induced reactions will create many other fissionable actinides and the user may wish to specify which of these actinides are considered as possible parents when calculating the pathways By default only U and Pu are considered but by increasing NCHO and specifying the identifiers of actinides e g AM242M then
190. lso process the uncertainty data given in the EAF uncertainty file before it can be used This process is described in the next section Collapsing uncertainty data The EAF uncertainty file contains values of the energy boundaries e g Ev Eu and 20 MeV and values of A for each energy range In the case of threshold reactions with threshold energy gt 20 MeV only one energy range is considered threshold 60 MeV and so the shape of the neutron flux 1s unimportant But for other threshold reactions two values are given while for non threshold reactions four values are given and these must be combined taking into account the shape of the neutron spectrum The effective cross section used by FISPACT is defined by equation A9 62 2 69 12 9 62 where is the cross sections in group i is the neutron flux in group i and the sum is over all energy groups If A is the relative error of the cross section in a particular energy group then the error of a particular weighted group cross section is given by equation A9 63 dm A9 63 Ye The following two assumptions are made User Manuat Issue 1 Feb 2007 UKAEA Fusion 130 FISPACT UKAEA Fusion 1 Errors in all the groups of a particular energy range are 100 correlated 2 errors in the three energy ranges are 0 correlated The first means that the error used for a particular energy range determi
191. mber of daughter atoms is greater than PMIN the pathway and the contribution will be printed in the output If IPRPA gt 0 then each pathway found irrespective of its contribution to the number of daughter atoms will be listed prior to the pathway output This option is only recommended if the output is not understood and the user wishes to check all the pathways calculated Note that ROUTES MUST come before the RESULT code word Nofe the maximum number of links that can be specified is 8 See Appendix 8 for more details of pathways An example of the use of this code word follows ROUTES AL27 NA22 3 1 E10 O0 RESULT 1 NA22 1 0E15 In a spectrum corresponding to the first wall of a fusion device the output would typically be as shown below more information on typical pathways is available in references 5 6 and 7 User Manuat Issue 1 Feb 2007 UKAEA Fusion 50 FISPACT 27 Al n2n Al n n o Na 70 Al n n a Na n2n Na 29 The PMIN value was set so that it was 1 10 times the total amount of Na formed in the irradiation given after the RESULT code word and with set to 0 none of the many pathways tested each with negligible contributions were printed SENSITIVITY XSENS XNSEN 1 INSENS INSEN4 UKAEA Fusion Parent l Daughter l l 1 INSEN3 Nuclide J J 1 INSEN4 This code word allows sensitivitv calculations to be performed If XSENS LAMBDA then the sensitivity coefficients with
192. meaning B or and Isomeric Transition respectively If it is required to follow the oto calculate the production of He then the decay is shown as X Sequential charged particle reactions are indicated in square brackets p n means that a proton from say a n p reaction reacts with the nuclide emitting a neutron For each of the daughter nuclides specified by the code words or calculated as dominant nuclides the following typical output is given C12 n a This pat This pat Cis mi te This pat C12 n a This pat 9 n d 118 h contributes 7 657 of total amount of Li 8 pathway type is RR S 11 118 h contributes 9 531 of total amount of Li 8 pathway type is RR S 11 118 h contributes 79 429 of total amount of Li 8 pathway type is RR S Be9 117 9 118 h contributes 3 294 of total amount of Li 8 pathway type is RRR S 4 paths which contribute SHS 912 5 Gie Fusion User Manual Issue 1 Feb 2007 UKAEA FISPACT 73 The pathway is given with the contribution that it makes to the production of the nuclide and the pathway type descriptor The type descriptor shows whether each link is a reaction R r or decay D d and whether the parent is long lived R D or short lived r d After the the final nuclide is shown as long lived L or short lived S Following the
193. meric value MUST follow the code word Numeric value required for ISPLIT SPLIT A numeric value MUST follow the code word Numeric value required for IUNCER UNCERTAINTY A numeric value MUST follow the code word Numeric value required for IUNCTY UNCTYPE A numeric value MUST follow the code word Numeric value required for JSTRM NEWFILE A numeric value MUST follow the code word Numeric value required for LNNM SEQNUMBER A numeric value MUST follow the code word Numeric value required for MAXXT CONV A numeric value MUST follow the code word Numeric value required for MIND MIND A numeric value MUST follow the code word Numeric value required for MONIT MONITOR A numeric value MUST follow the code word Numeric value required for N LEVEL A numeric value MUST follow the code word Numeric value required for N1 FUEL A numeric value MUST follow the code word Numeric value required for N2COLL COLLAPSE A numeric value MUST follow the code word Numeric value required for NCHO FISCHOOSE A numeric value MUST follow the code word Numeric value required for NDOSE DOSE A numeric value MUST follow the code word Numeric value required for NEAFVN EAFVERSION A numeric value MUST follow the code word Numeric value required for NERROR ERROR A numeric value MUST follow the code word UKAEA Fusion User Manuat Issue 1 Feb 2007 FISPACT 139 Numeric value required for NDSTRC GRPCO
194. method Reference 29 describes how available data for an element are used with decay data for a nuclide to derive Committed Effective Doses per unit uptake for ingestion and inhalation for the nuclides with no data In total 1209 nuclides have had data calculated approximately Reference 30 documents the EAF HAZ 2007 library Transport of radioactive material from place to place is governed by regulations set up by the IAEA Reference 31 gives details of A values for certain radionuclides Using these values it is possible to work out how much of a particular mixture of radioactive materials can be packed into a type of container and safely transported Data from this reference for the nuclides listed are transferred to EAF 2 with the default prescription given in reference 31 used for all radionuclides not explicitly listed Reference 30 documents the EAF A2 2007 library FISPACT can use these data to show the A limit for individual nuclides and the effective A value for the irradiated material EAF CLEAR 2007 Disposal of radioactive material in special repositories is expensive Regulations exist which determine activity levels for nuclides such that materials can be cleared or disposed of as if they are not radioactive Clearance data is being investigated by the IAEA and recommendations are available Reference 32 gives details of suggested clearance values for certain User Manuat Issue 1 Feb 2007 UKAEA Fusion 104
195. mines the number of nuclides treated as fission products If version n m of EAF is used then NEAFV should be set to the value n Note that for EAF 97 and EAF 99 the value 4 should also be used For EAF 2001 and EAF 2003 the value of 5 should be used For EAF 2005 the value of 6 should be used For EAF 2007 the default value of 7 should be used An example of the use of this code word follows EAFVERSION 5 AINPUT FISPACT Irradiation of SS316 steel The cross section library EAF 2003 or an ARRAYX file produced using this library is to be input TITLE This code word reads a title beginning with an and containing a maximum of 72 characters In versions prior to 3 0 the title had to contain relevant identifying data about the libraries used FISPACT now automatically takes text strings from the CROSSEC FLUXES DECAY and COLLAPX files and uses these to construct the identification information that is stored in ARRAYX The user MUST include at least a single character but it 15 still useful to put a sensible title to aid legibility of the output This code word causes the decay library and the collapsed 1 group library to be input and processed and the condensed library data to be written to an external file User Manuat Issue 1 Feb 2007 FISPACT 21 FISPACT One of the following code words MUST follow the title TAPA writes a summary file of nuclides and isomers ARRAY uses decay data fr
196. mitted particle spectral data from neutron induced reactions e Charged particle cross section data The last three libraries are only required if sequential charged particle effects are included All the current libraries are described in more detail below EAF N XS 2007 UKAEA Fusion EAF N XS is the point wise neutron induced cross section library Data on 65 565 cross sections on 816 targets are held in a modified ENDF B format The basic criterion used to decide which nuclides to include as targets is that all nuclides with a half life of greater than 6 hours have cross section data In the User Manuat Issue 1 Feb 2007 FISPACT 99 case of capture and fission cross sections the point wise file has been processed from an evaluated file using NJOY to reconstruct the resonance region from resonance parameters No self shielding is included and the temperature for Doppler broadening is 300K This library is available to users but before it can be used by FISPACT it is necessary to process it into a particular group cross section format Reference 16 documents the EAF 2007 neutron induced cross section library There is no printed report on the capture cross sections for EAF 2007 however reference 17 gives data for EAF 97 while reference 18 gives the complete listing of the EAF 99 REPORT file EAF D XS 2007 EAF D XS is the point wise deuteron induced cross section library Data on 66 864 cross sections on 81
197. n a matrix of charge Z is given by equation A7 1 dN az 2 dB terza i xama da ME A7 1 where dN number of y rays with energy E keV E energy of electron keV 2 7610 Consider a group structure where e n is the upper limit of the n th energy group in units of E 0 1 MeV Only energies greater than 100 keV are used so that only 19 of the 24 energy groups are required Integrating equation A7 1 over the n th group amp n E E mA E N n aZ HAZE ft 0 A7 2 n l E where a Eo E t a e n 1 e n e n 1 amp 0 21 Equation A7 2 shows that the y spectrum N n depends on the matrix through Z and nuclide through a If equation A7 2 is used to calculate the correction due to particles with low energies then the discrete nature of the group structure can cause problems Integrating over the first group 2 e gives term Je in units of 0 1 MeV and as is e 1 reduced this integral becomes zero This occurs when User Manuat Issue 1 Feb 2007 FISPACT 107 a 1 In2 145 It is assumed in FISPACT that such low energy bremsstrahlung corrections can be ignored and thus if the energy of the particle is less than 0 145 MeV then correction is set to zero The above discussion is valid only for mono energetic electrons but it is assumed that the same expressions are valid for the emission of B particles which ha
198. n be calculated using the standard solutions of the Bateman equations A procedure was written see reference 2 which uses as input a file which contains the half lives of all relevant nuclides a starting number of atoms and a series of times at which numbers of atoms and the total activity are to be calculated It was coded so that numbers of atoms less than 1 0 were written in the output as 0 0 The procedure used double precision variables and was tested against a hand calculator for various cases Data on half lives were extracted from the decay data library EAF DEC 97 and compared against the values shown in the FISPACT PRINTLIB output Decays were chosen to cover a wide range of nuclide half lives to have either a single daughter or a chain of daughters and to have different decay UKAEA Fusion User Manuat Issue 1 Feb 2007 FISPACT 207 modes Isomeric states either as the initial nuclide or as daughters and decay branching were also investigated The method for dealing with branched decays is as follows Figure A19 1 A general series of decays that includes branching 0 1 2 Poa Noo e N De N 1 3 fi Noo Su Se 4 5 Jo Noo Se Figure A19 2 The series of linear pathways equivalent to Figure 1 particular branched decay is shown in Figure A19 1 where the nuclides are labelled by numbers and the branching fractions are indicated by the symbols f where i lab
199. nce achieved for ALL printed isotopes IN FOLLOWING TABLES MEANS ISOTOPE IS CALCULATED BY APPROXIMATE METHOD 1 MEANS CONVERGENCE NOT REACHED FOR NUCLIDE MEANS GAMMA SPECTRUM IS APPROXIMATELY CALCULATED NUCLIDE IS STABLE MEANS NUCLIDE WAS PRESENT BEFORE IRRADIATION HER U E 0 This key lists five single character flags that are printed immediately following each nuclide identifier indicates that the nuclide has been calculated as if it were in equilibrium increasing the first parameter following LEVEL will mean that fewer nuclides are labelled by this flag is the convergence flag any nuclide labelled by this has not been calculated accurately enough decreasing the second parameter in CONV can remove the flag but in most cases the nuclides are of no practical importance and any error can be ignored 6 indicates that no y spectral data were present in the decay data library and that the code word SPEK was used to approximately calculate a spectrum If most of the y dose rate is produced from nuclides with this flag then the result should be treated with great caution indicates that the nuclide is stable and gt indicates that this nuclide was present in the material input specified by the MASS or FUEL code words User Manual Issue 1 Feb 2007 UKAEA Fusion 68 FISPACT The main output of this section follows this is the details of the nuclides present
200. nd the run is ended by END AINPUT FISPACT Pathways of Sc45 FUEL 1 5 45 1 0 25 FLUX 1 8E15 LEVEL 100 1 CONV 10 2 E 3 1 2 TIME 2 5 YEARS User Manuat Issue 1 Feb 2007 FISPACT 65 LOOPS 20 0 PATHS 2 SC45 R SC46 R 5 47 RESULT 1 SC47 1 62032 19 END END of pathways run This case models 1 10 atoms of Sc being irradiated for 2 5 years in a flux of 1 8 10 n cm s The convergence limit for the pathway calculation is reduced to 1 by the code word CONV and LOOPS is used to include any isomers with half lives less than 20 seconds in the calculation The pathway Sci 5Sc n y Sc specified by PATHS contains 2 reactions the user must give the identifiers of three nuclides and the total amount of the daughter nuclide Sc calculated in a previous run is given by the code word RESULT User Manuat Issue 1 Feb 2007 UKAEA Fusion 66 FISPACT Interpretation of FISPACT output The output from FISPACT consists of several distinct blocks that are described in detail below Header and user information The header gives a banner version of the program name and the date of the version this is followed by a box giving the version number and platform on which the run was done FISPACT VERSION 07 0 0 FEBRUARY 2007 PC Salford FTN77 32 Wow H Depending on the platform being used the end part of the printed text will be PC Salford FTN77 32 32 bit PC UNIX Various UNIX operating systems
201. nes the position of the cross section curve its shape is correct in that range but its absolute position is uncertain While the second implies that there are independent measurements in the various energy ranges Using assumption 1 the error in an energy range is given by equation A9 64 x A9 64 1 where Low Medium High Extended S is the set of groups in the various energy ranges Assumption 2 means that the total error is given by equation 9 65 EE 9 65 The corresponding total relative error A is given by equation 9 66 oM A9 66 The weighted cross section G in one of the energy regions is defined by equation A9 67 G 30 0 2 a A9 67 ieS Combining equations A9 64 A9 67 the total relative error is shown by equation A9 68 X 2 5 2 utentes Sues A9 68 Equation A9 68 is used by FISPACT to collapse the uncertainty data for a particular neutron spectrum User Manuat Issue 1 Feb 2007 FISPACT 131 Appendix 10 y group structures There are two gamma energy group structures used in FISPACT The 24 group structure is the default while the 22 group must be requested by the code word GROUP Table A10 1 shows the values of the group boundaries for both structures Table A10 1 Energy group structure for the 2
202. nge of deuteron in each element Range of alpha particle in each element Range of triton in each element Range of He 3 in each element Cross section data for p n reactions Cross section data for dn reactions Cross section data for reactions Cross section data for t n reactions Cross section data for h n reactions Cross section data for d 2n reactions Cross section data for t 2n reactions Energy spectrum of emitted particles part 1 Energy spectrum of emitted particles part 2 Energy spectrum of emitted particles part 3 Energy spectrum of emitted particles part 4 Energy spectrum of emitted particles part 5 Relative uncertainty values for nuclide half lives Gamma absorption data Clearance data Cross section data for p 2n reactions Cross section data for 2n reactions Cross section data for h 2n reactions User Manuat Issue 1 Feb 2007 FILES INPUT OUTPUT CROSSUNC ASSCFY FISSYLD GRAPH A2DATA COLLAPX ARRAYX HAZARDS SUMMARYX DECAY COLLAPX INDEX CROSSEC FLUXES STOP PRO STOP DEU STOP ALP STOP TRI STOP HE3 XN PN XN DN XN AN XN TN XN HN XN D2N XN T2N SPEC 1 SPEC 2 SPEC 3 SPEC 4 SPEC 5 HALFUNC ABSORP CLEAR XN P2N XN A2N XN H2N UKAEA Fusion 16 FISPACT UKAEA Fusion The beginning of a typical FILES file is shown below 03 d fispact tests v_2007 spectra 05 d fispact tests ss316 2 1 06 d fispact tests ss316 2 0 07 d Neaf dataNeaf 2007Neaf un 20070
203. nite slab or dose at a given distance from a point Source Contact y dose rate Equation A3 1 shows the formula used to calculate the y dose rate at the surface of a semi infinite slab of material it is taken from J aeger D 5 76 10 2 SE eee A A3 1 surface y dose rate Sv h D E mean energy of the i th energy group mass energy absorption coefficient Uen p of air m kg u mass energy attenuation coefficient of the material m kg build up factor 2 S rate of y emission MeV kg s The photon mass attenuation coefficient u p and the mass energy absorption coefficient Uen p for all elements with Z 1 100 have been produced using the XGAM program from the National Institute of Standards and Technology The new data base covers energies of photons X ray Y ray bremsstrahlung from 1 keV to 100 GeV and has been processed into a 24 group structure 1 keV 20 MeV identical to the FISPACT group structure The present compilation 15 an extension of the recent calculations of Seltzer and is intended to replace the values given in Hubble which were User Manuat Issue 1 Feb 2007 UKAEA Fusion 94 FISPACT UKAEA Fusion used in previous FISPACT versions The present data differ from the Hubble set in the following respects l The first 100 elements are included compared to the 40 selected elements previously covered All e
204. ns it MUST be remembered that only a single irradiation step should be considered If it is necessary to consider many irradiations say for pulsed operation then two possible solutions should be considered e Carry out pulsed calculations for the detailed inventory calculations but in a separate run use an average flux over the total irradiation time to calculate pathways Model the irradiation history so that the large majority of the fluence is in the initial irradiation step Pathways will be calculated for this interval and the contribution of the final few pulses will be ignored In most cases the second option will be most useful as uncertainty estimates can also be given based on these approximate pathways See Appendix 20 for more discussion of non steady irradiations User Manuat Issue 1 Feb 2007 FISPACT 113 Appendix 9 Uncertainties EAF3 was the first activation library to contain uncertainty information Details of this uncertainty file for EAF3 1 are given in Reference 20 The file has been improved for EAF 2007 and some details of these changes are given below The use of the uncertainty and half life uncertainties by FISPACT to calculate uncertainty estimates of the radiological quantities e g activity is then explained EAF UN 2007 EAF XS 2007 contains data up to an energy of 60 MeV all non threshold reactions capture fission and some n p reactions have four uncertaint
205. nuclide and so not all the previous nuclides can be short lived Assume that all the nuclides in the pathway are long lived except the final one Thus in secular equilibrium the number of atoms of the final nuclide 15 given by equation A9 58 nE s A9 58 In the long lived limit NM is given by equation A9 51 Substituting this value into equation A9 58 and rearranging yields equation A9 59 n 2 1 eat In D ee A9 59 i l Equation A9 59 shows that in the short lived limit for the final nuclide there is an additional factor of n 1 _ t due to the n 2 final nuclide being short lived rather than long lived User Manual Issue 1 Feb 2007 FISPACT 127 Summary of factors for each type of pathway link The previous two sections show that for both long and short lived limits of the number of atoms of the final nuclide in the pathway these are calculated by forming a product of factors one for each link in the pathway and multiplying this by the initial number of atoms of the first nuclide Table A9 3 lists these factors and also includes the additional factor that is determined by the half life of the final nuclide Table A9 3 Factors for pathway links Type of link Link is reaction from a long lived nuclide Link is reaction from a short lived nuclide Link is decay from a long lived nuclide M Link is decay from a short lived nuclide 1 Final nuclide
206. ny May 1991 p 828 Springer Verlag 1992 4 J Kopecky and D Nierop The European Activation File EAF 4 Summary Documentation ECN C 95 072 1995 5 M R Gilbert and R A Forrest Handbook of Activation Data calculated using EASY 2003 UKAEA FUS 509 2004 6C B A Forty R A Forrest D J Compton and C Raynor Handbook of Fusion Activation Data Part 1 Elements Hydrogen to Zirconium AEA FUS 180 1992 7 C B A Forty R A Forrest D J Compton and C Raynor Handbook of Fusion Activation Data Part 2 Elements Niobium to Bismuth AEA FUS 232 1993 8J Sidell EXTRA A digital computer program for the solution of stiff sets of ordinary value first order differential equations AEEW R 799 1972 9 R G Jaeger Editor in Chief Engineering Compendium on Radiation Shielding Vol 1 Springer Verlag 1968 10 NIST X ray and gamma ray attenuation coefficients and cross section database U S Department of Commerce National Institute of Standards and Technology Standard Reference Data Program Gaithersburg Maryland 20899 1995 11 J H Hubble and S M Seltzer Tables of X ray mass attenuation coefficients and mass energy absorption coefficients 1 keV to 20 MeV for elements Z 1 to 92 and 48 User Manuat Issue 1 Feb 2007 FISPACT 231 additional substances of dosimetric interest NISTIR 5632 U S Department of Commerce 1995 12 J H Hubble Photon Mass Attenuation and Energv
207. odel code TALYS due to the lack of experimental data Thirty nine input neutron energies are used and the outgoing charged particles are in 1 MeV bins covering the energy range 0 60 MeV EAF XN describes the cross sections of p d h t and a particles on 775 targets Ten reactions types are considered User Manuat Issue 1 Feb 2007 FISPACT 105 p n t n hn n p 2n d 2n t 2n h 2n and 0 2n This is the third of the libraries required for calculations with SCPR The data are calculated by a theoretical model code TALVS due to the lack of experimental data The data are given at 1 MeV intervals over the energy range 0 60 MeV EAF ABS 2007 EAF ABS 2007 contains the photon mass energy attenuation coefficient u p for all the elements Z 1 100 in increasing Z order The attenuation coefficient u and energy absorption coefficient LL p for air are also listed All data are stored in the same 24 group energy structure as described in Appendix 10 User Manuat Issue 1 Feb 2007 UKAEA Fusion 106 FISPACT Appendix 7 Bremsstrahlung corrections UKAEA Fusion The contribution of high energy particle bremsstrahlung to the total y dose rate can be significant in cases where the emission is small FISPACT uses a similar approach to Jarvis 37 who considers Y emission from a mono energetic electron The energy distribution of Y rays emitted by a mono energetic electron i
208. om an existing ARRAYX file LINA reads an existing summary file this option is included only for compatibility with FISPIN it is not recommended Users preparing an ARRAYX file for a new decay data library MUST use TAPA For subsequent runs when only the cross section library is different then the code word ARRAY can be used If this is done then the decay data is taken from the existing ARRAYX file rather than processed afresh from the DECAY file Note that there is no error checking that the ARRAYX file that is read is consistent with the cross sections and consequently an inappropriate use of this code word e g using an ARRAYX corresponding to a different decay data library would lead to a new ARRAYX file containing rubbish However when used carefully this option will make a large time saving see Table A14 2 in the preparation of ARRAYX files An example of the use of this code word follows SPEK ENFA EAF 2007 100 Group zone 13 TAPA FISPACT Write to ARRAYX file TITLE This code word reads a 72 character title beginning with an containing information about the particular run This title is also used to label the graphs but for the graph title only the first User Manuat Issue 1 Feb 2007 UKAEA Fusion 22 FISPACT 40 characters are used Note that the text used for TITLE here is different from that text following ENFA Note that the code word does not trigger any action it is only used
209. on UNIX when run SEQU Stream 19 not always closed 955 956 case with multiple irradiations error seen under UNIX Corrected Dominant nuclides for three 957 970 973 additional quantities printing 981 Final page of PRINTLIB does not 971 contain 627 printing 23 8 00 Point dose rate too large by a Factor of 10 needed to correct 972 982 984 factor 10 following a change of units in FISPACT 97 25 8 00 Incorrect MT value used for MT corrected but to ensure back 985 997 n 2np reactions because of error compatibility it was necessary to in SYMPAL introduce new EAFV value 3 10 00 All atomic masses held as integer AWR values from decay data read 998 1002 1004 values and held in floating point 1006 1015 variables 1017 1022 1024 1028 5 10 00 Physical constants require updating Avogadro constant changed from 1003 1005 6 02204459E 23 to 1016 1023 6 0221367 23 amu changed from 1 66056559E 27 to 1 6605402 27 4 1 01 Additional summary table showing Table implemented 1029 1037 User Manual Issue 1 Feb 2007 UKAEA Fusion 196 FISPACT 1043 1054 numbers F gamma contributions output incorrectly shown as per kg 5 1 01 FISPACT 2001 not recognised as a Check on VSIDNT changed now recent version when reading all versions since FISPACT 2 uncertainty options correctly recognised abundances need to be updated 24 1 01 Clock routines for UNIX CLOCK and CLOKK routines 1057 1064 ej Error when all elements are
210. or New code word SEQNUMBER 1544 1571 high energy data and additions to existing routines 4 11 04 New types of charged particle data Dimension of FILES increased to 1572 1576 43 4 11 04 Include new charged particle Define these new reactions 1577 1584 reactions in pathways 5 11 04 Charged particle libraries contain Allow for 7 digits in INDEX file 1585 data for Fm to describe nuclide 8 11 04 Printing of charged particle data Change P N to p n etc 1594 uH LN NN 8 11 04 Increased number of target Adjust ICPN array and add IYCP nuclides array 9 11 04 Charged particle data includes New INCPNT array 1614 1637 isomers 7 12 04 Problems using 175 data with a New array EXTRA track 1638 1664 high energy library carefully number of energy groups used 9 12 04 Use of PULSE code word withan Line of code added in error many 1665 old version of EAF causes a crash years ago but no effect except under these conditions Removed 15 12 04 Problem found when using Increase constants in all COLnnn 1666 1679 spectrum with only low energy routines neutrons in PRINTLIB nuclides per line Problems when collapsing with a Dimensions of arrays increased to 1686 1691 351 group spectrum same values used in COL211 22 12 04 Fe 49 not required for EAF 2005 Remove changes introduced in 1692 1693 so it can be fictional nuclide 1296 and 1297 and in other COLnnn routines names in output increase array dimensions 10 5 05 Need to be a
211. orting a bug e Version number of FISPACT and EAF libraries used Computer platform used Copy of input file Copy of neutron spectrum Details of error message given UKAEA Fusion User Manuat Issue 1 Feb 2007 FISPACT 199 Appendix 18 EASY documentation set The FISPACT 2007 User Manual forms part of the EASY documentation set The complete set shown in Table A18 1 15 available to commercial users of FISPACT most of it is also available as the individual UKAEA reports shown in Table 18 2 in some cases these only contain a sample of the complete data where these are very extensive Table A18 1 Parts of the EASY Documentation Series EDS Report Title EDS 0 EASY 2007 Overview EDS 1 FISPACT 2007 User manual EDS 2 FISPACT 2003 Qualification report EDS 3 EAF 2007 Data libraries EDS 4 EASY 2007 Processing system The Qualification report has not been produced for FISPACT 2007 Table A18 2 The composition of the EASY Documentation Series EDS Report UKAEA report number EDS 0 5 2007 Overview EDS 1 FISPACT 2007 User manual EDS 2 Selection of reports and papers m EDS 3 EAF 2007 Neutron induced cross section library 16 EAF 2007 Deuteron and proton induced cross 19 section libraries EAF 2007 Biological clearance and transport libraries SYMPAL User guide SYMPAL Utilities guide SAFEPAQ User manual SAFEPAQ II User manual Issu
212. ose JET D T spectrum calculated by Jim Edwards for region 461 Dae 23 09 98 Reference Validation of the MCNP FISPACT model for JET activation and predictions for future D T scenarios Edwards JEG Forrest RA F PLAwPA14 3b JEGE 1 01 09 98 Figure A15 8 The Neutron spectra window All FISPACT inventory runs require information on the material to be irradiated The EASY User Interface enables material details to be stored and viewed Clicking on the twelfth toolbar button in the main window Figure A15 1 displays the View materials window shown in Figure A15 9 A material can be selected in the list box and details for it are then displayed in the window A material composition can be defined in terms of the weight of the elements present or by the number of atoms of the various isotopes A new material can be defined by clicking on the Add button The material composition can be written to a file by clicking the Write file button and this can User Manuat Issue 1 Feb 2007 FISPACT 187 then be used in an INPUT file The materials database thus acts as a useful repository of materials ensuring that consistent compositions are used in various FISPACT runs View materials Description dd Target values for Eurofer steel Delete write file Reference didi Weight Draft amp 3 518E Eurofer Steel F Tavassoli Interim Structural Design Close El Summary
213. otassium in a first wall flux of 5 MW m for 2 5 years Progress of the run is included by using MONITOR with the User Manual Issue 1 Feb 2007 UKAEA Fusion 62 FISPACT UKAEA Fusion parameter 1 The MIND parameter is set to 10 biological hazards and transport limits are required and a bremsstrahlung correction is output for Be Two graphs total activity and total heat in PC format with no uncertainty data are requested and the LEVEL parameters are set to 100 and 1 1 subinterval and nuclides with half life 6 3 days are in equilibrium After the irradiation the inventory is printed by calling ATOMS and then the first LEVEL parameter is reduced to 20 The neutron flux is reduced to 0 the value of time reset to 0 by ZERO three cooling times follow and inventories are printed for 0 1 years year and 50 years The run is finished by END AINPUT FISPACT C sensitivity calculations MASS 10 0 1 C 100 FLUX 2 0E15 1 45 LEVEL 100 5 SENS SIGMA 1 10 3 2 C12 C13 C13 C14 C13 BE10 C14 BE10 ERROR 3 C12 C13 1 C13 C14 1 C13 BE10 0 5 TIME 5 0 YEARS ATOMS LEVEL 20 1 FLUX 0 0 NOT1 ZERO TIME 5 0 YEARS SPECTRUM TIME 95 0 YEARS ATOMS END END of sensitivity calculations This case models the irradiation of 10 kg of carbon in a flux of 2 0 101 n cm s for 5 years The MIND parameter is set to 10 and half lives are to be printed A dump to an external file of the nuclide am
214. other nuclides can be included Note that including many minor actinides will substantially increase the running time In most cases minor actinides are unlikely to have significant impact on the total radiological quantities and so are unlikely to be part of the important pathways Also this code word only affects the calculation of pathways all actinides are considered during the calculation of inventories An example of the use of this code word follows FISCHOOSE 4 0238 PU239 PU240 PU242 In this case any pathways containing a fission reaction can only 23877 239p 240 242 have one of the four actinides U Pu and Pu as parent FISYIELD NYLD lt SYMB I I 1 NYLD When actinides are included in the list of input elements then by default all the actinides will produce fission products when they fission If NYLD 0 then no fission products are produced from any of the actinides If NYLD is a positive integer then only the actinides that are specified in the list of identifiers SYMB e g 242 produce fission products If NYLD is a negative integer then all actinides except those that are specified in the list of identifiers SYMB e g AM242M produce fission products This facility is included so that information on the irradiated actinides alone can be obtained Also when investigating the User Manuat Issue 1 Feb 2007 UKAEA Fusion 28 FISPACT FLUX FLUX2 UKAEA Fusion p
215. ounts is required using data stream 45 The LEVEL parameters are set to 100 and 5 5 subintervals and User Manuat Issue 1 Feb 2007 FISPACT 63 nuclides with half life 12 6 days are in equilibrium The sensitivity of the nuclides and to the cross sections P C n y PC n y and PC n o are required with a cut off value of 1 1019 The fractional errors in the first two reactions are taken from the uncertainty file while the third is input directly These uncertainties are used to calculate the errors in the two nuclides previously specified After the irradiation the inventory is printed by calling ATOMS and then the LEVEL parameters are reduced to 20 and 1 respectively and the value of time reset to 0 by ZERO The neutron flux is reduced to 0 and the dump of data to the external file is stopped Two cooling times of 5 and 100 years follow the first only has summary information since SPECTRUM is used meaning that the full inventory is omitted and the run is finished by END AINPUT FISPACT Pathways of Zn66 5MW m2 FUEL 1 ZN66 2 56943E24 WALL 5 LEVEL 100 1 TIME 2 5 YEARS ROUTES ZN66 C060 6 1 11 0 ROUTES ZN66 NI63 6 1 16 0 PATHS 2 ZN66 R NI63 FEGO PATHS 5 ZN66 R ZN67 R ZN68 R ZN69 D GA69 R GA70 RESULT 4 C060 1 62032E16 NI63 1 78082 21 FE60 2 02345 19 GA70 4 45689E12 END END of pathways run This case has 2 56943 102 atoms of Zn Note that it is important to start wit
216. ous stage extrapolation to improve the convergence of the solution Experience with this solution method in both FISPACT and FISPIN shows that it is both rapid to converge and able to give sufficient accuracy The code implements a maximum number of iteration stages 10 but if convergence has not been achieved by then it is usually only for a very few unimportant nuclides The output flags these nuclides thus enabling the worth of the particular non converged run to be judged It was mentioned above that there is a limit on the largest eigenvalue considered in the solution of the equations This means that physically only nuclides with sufficiently long half lives are calculated by the above method The remainder are assumed to be in equilibrium and thus their values can be written down immediately as shown in equation A2 4 4 40 0 5 ja a acest A2 4 j The half life at which nuclides are considered to be in equilibrium is under the control of the user This is done by choosing the time interval code word TIME and the parameters following the LEVEL code word User Manuat Issue 1 Feb 2007 FISPACT 93 Appendix 3 y dose rate In addition to the activity of irradiated materials another measure of acceptability is the dose rate from emitted y rays FISPACT uses two approximate estimates of the y dose rate due to irradiation by neutrons contact dose from the surface of a semi infi
217. own in equation A12 1 1 24 175 24 E n 2 0 E gt gt NO E E gt f E E JAE AR E n k l A i l j k TERTA A12 1 where is the neutron flux in i th energy group 175 o 0 15 integrated neutron flux l N is the number of atoms of the nuclide A is k th charged particle energy step Ox is the production cross section of charged particle x in the th energy group User Manuat Issue 1 Feb 2007 UKAEA Fusion 144 FISPACT UKAEA Fusion Ex is the production cross section of nuclide C in the k th energy group da E is the normalised charged particle spectrum for neutron energy in 7 th energy group and in the k th outgoing energy step AR E is the differential thickness of the surrounding material for charged particle of starting energy E Note that in equation A12 1 sums are over 175 groups since version 2005 211 group data can be used as well as 175 In the original work of a separate code PCROSS was written to calculate the pseudo cross sections for a particular material in a specified neutron spectrum These pseudo cross sections were then merged with the collapsed cross sections and this new library used with FISPACT This process has been simplified by building the PCROSS subroutines into FISPACT and giving the user the option to include SCPR by means of a code word FISPACT calculates the pseudo cro
218. ows PARTITION 2 AR 0 01 K 0 20 In this case all elements except Ar and K remain unmodified all Ar Isotopes are reduced by a factor of 100 and all K isotopes are reduced to a fifth of their values before the code word was used INDXP I 1 NLINK 1 This code word allows a particular pathway consisting of NLINK reactions and decays to be specified The NLINK 1 nuclides in the pathway are input using the identifier e g 129 and in order to specify whether each link is a reaction or a decay an R or a D is placed between the identifiers This code word is only necessary if a special investigation of pathway information is needed Pathway data can be generated automatically for all the dominant nuclides by using the UNCERTAINTY code word PATH might be used for a particularly complicated pathway not generated automatically or to investigate nuclides only formed in small amounts Note when using this code word it is recommended that a standard inventory be produced first and the number of atoms of the daughter are specified in subsequent runs using the User Manual Issue 1 Feb 2007 UKAEA Fusion 46 FISPACT RESULT code word No inventory then needs to be calculated for the runs investigating the pathways making efficient interactive studies possible Note the maximum number of links that can be specified in a pathway is 20 See Appendix 8 for more details of pathways An example of t
219. ox displayed when a FISPACT run crashes Troubleshoot FISPACT FILES file INPUT file OUTPUT FLUXES files ARRAYS Data libraries E Fispact y_2007 FILES e fispact testcases vers_0O7_O7 input e ispact testcases vers_07_O7 output e fispact testcases vers_O7_O Mfluxes e Mispact testcases vers_O7_O7 arrayx Unknownl The following files specified in FILES are not present on the disks Files on streams 38919 Press Check duplicate dll to search disk to find all copies of the dll required to run FISPACT Figure A15 5 The Troubleshoot FISPACT dialog The Troubleshoot FISPACT dialog shows the path and names of some of the specified files and checks whether any of the files specified in FILES are missing or have zero size In Figure A15 5 the files connected to streams 3 arbitrary spectrum 19 cross section library and 20 neutron spectrum are missing If files such as the INPUT file are missing then FISPACT will terminated abnormally EAF decay data A new feature in the EASY User Interface since version 2001 is the ability to view EAF decay data Clicking on the ninth toolbar button in the main window Figure A15 1 displays the EAF 2007 decay data window shown in Figure A15 6 The User Manuat Issue 1 Feb 2007 UKAEA Fusion 184 FISPACT required nuclide is entered in the Nuclide text box and all the decay data from EAF DEC 2007 are displayed as well as biological hazard trans
220. parison between Table 1 row 4 and Table 3 row 2 shows agreement for energy e Comparison between Table 1 row 5 and Table 3 row 5 shows agreement for mode 1 e Comparison between Table 1 row 6 and Table 3 row 6 shows agreement for mode 2 e Comparison between Table 4 and Table 3 rows 7 15 shows agreement for the y spectrum e Comparison between Table 5 and Table 3 rows 16 17 shows agreement for the fission yields Collapsing cross sections FISPACT reads neutron induced cross section libraries in one of five energy group structures The folding of this energy dependant cross section with the neutron spectrum for a particular situation generates an effective or collapsed cross section for each reaction FISPACT also needs to use a single cross section value for the reaction from a parent to a daughter while the data libraries often contain more than one cross section entry for a particular pair of parent and daughter nuclides This is because for example the A n n pB and A n d B reactions both produce the same daughter nuclide The library keeps these reactions separate because the first produces H from the proton while the second produces H from the User Manuat Issue 1 Feb 2007 UKAEA Fusion 204 FISPACT UKAEA Fusion deuteron The method that FISPACT uses to keep track of these secondary gas products is to include further cross section values in the COLLAPX file for A n X H and A n X H These cross se
221. particular case Another case can then be input by specifying more input data unchanged data do not have to be input again The text used in TITLE is arbitrary except that at the end of the last case the first three characters after the MUST be END An example of the use of this code word follows END END of Fe run In this case a line of text is entered after END ENDPULSE This code word terminates the construct that was started by PULSE All code words between PULSE and ENDPULSE are repeated NPULSE times NPULSE is the parameter following PULSE ERROR NERROR Parent I Daughter l ERMAT l 1 1 NERROR Inputs the number VERROR of reactions and the identifiers of the parent and daughter of each reaction and optionally the fractional error of the reaction cross section In versions prior to 3 0 the user had to input a value of the fractional error but this is now available from the EAF uncertainty file If data from the file are to be used then ERMAT MUST be set to 1 User Manuat Issue 1 Feb 2007 UKAEA Fusion 26 FISPACT Note that if no uncertainty data exists in the library then the fractional error MUST be input using 1 will cause an error message to be printed This code word should only be used following the code word SENSITIVITY to give the error in the number of atoms of a nuclide due to the specified reactions for routine calculations the uncertainty calculations are automatica
222. poles Equation A9 34 shows that all the poles are simple and for the pole at A the residue is given by equation A9 36 The prime on the product sign in all equations in this section indicates that the term with i j is excluded User Manuat Issue 1 Feb 2007 UKAEA Fusion 122 FISPACT enu R 7 DA MA acta A9 36 n l i l There are n 1 simple poles and the sum of the residues is given in equation A9 37 n gt 9 37 i 1 j l i l Combining equations A9 34 A9 35 and A9 37 and introducing the factor 1 since the factor in the product in the denominator has the labels 7 and j interchanged yields equation A9 38 the final form of the solution of equation A9 28 ntl n n l N 0 7 CD 9 38 i l j l i l Corresponding to equation A9 25 for the final nuclide in the 2 link case the number of atoms of the final nuclide in the n 1 link case is given by equation A9 39 n l N CD po Sev D A j l m A9 39 An identity can be derived by using equation A9 38 and solving equation A9 28 by direct integration The solution of equation A9 28 using a standard integrating factor is given in equation 9 40 M A9 40 ntl Using equation A9 38 and rearranging yields equation A9 41 Nyailt D No Io e i 1 j l i l
223. port and clearance data Note that the units of half life and energy can be changed to give convenient numbers Clicking on the Photon lines tab enables the y and X ray lines to be displayed and these can be plotted using the fourth toolbar button The many options and facilities of this window are fully explained in the Help file A EAF 2007 decay data File Edit Graphs Search Database x Nuclide Aa 108m format Bi Stable SERATA palit 5 Source jeff 31 E Beta Mass 59 33387 amu Decay data Photon ines Matter ines Half life 5 2709E 00 9 9998E 04 Biological hazard coefficients 5 7 LI Alpha Ingestion 3 4000E 03 IT Alpha 0 0000 00 0 0000 00 31000E 08 Beta 9 6773E 04 2 4225E 02 Source ICRP72 sr Gamma 2 5038 06 3 5219 02 A Proton Number of decay modes 1 2 coefficient 0 40000 T Decay mode s Branching ratio EE Clearance level Clearance 1 00000E 02 Bq kg Get data Source IAEA G 1 7 Close Figure 15 6 The EAF 2007 decay data window EAF group cross section data Another new feature in the EASY User Interface since version 2001 is the ability to view EAF group cross section data Clicking on the tenth toolbar button in the main window Figure A15 1 displays the EAF 2007 group cross section
224. portant nuclides when EAF 2007 is used are summarised in Appendix 7 The fifth section lists the neutron spectrum used to collapse the Cross section library User Manuat Issue 1 Feb 2007 FISPACT 83 Appendix 1 Cross section group structures Seven standard group structures are used for the European Activation File and data in all these structures can be read into FISPACT Table 1 1 lists the group structures for the five low energy cases WIMS 69 GAM II 100 XMAS 172 VITAMIN J 175 and TRIPOLI 315 This method of presentation makes it clear in which energy ranges particular structures have most groups and will therefore give a good representation of the cross sections Table 1 2 lists part of the two high energy structures VITAMIN J 211 and TRIPOLI 351 showing how these join to the low energy ones Users are advised to prepare neutron spectra for a particular application in one of these structures as appropriate Table A1 1 Energy group boundaries for the five low energy standard structures 1 9640E 1 96403E 1 96403E 1 733081 1 73325 1 73325 1 6910E 1 69046E 1 6490E 1 64872E 1 5680 1 56831E 1 4920E 1 49182E 1 49180 07 1 49182 1 4550 1 45499 1 4190E 1 41907E 1 3840E 1 38403E 1 38403E 1 3500E 1 34986E 1 34983E 1 2840 1 28403E 1 25232 1 2210E 1 22140 4 1 22138 4 1 1620E 1 16183E 1 16183E 1 1050E 1 10517E 1 10515E 1 0510E 1 05127E 1 0000E 1 0
225. pplication of the Concepts of Exclusion Exemption and Clearance 2004 Safety Guide IAEA Safety Standards Series No RS G 1 7 IAEA Vienna User Manual Issue 1 Feb 2007 FISPACT 233 33 Clearance levels for radionuclides in solid materials application of exemption principles 1994 Draft Safety Guide IAEA Safety Series No 111 G 1 5 IAEA Vienna 34 SRIM 2003 available from NEA Data Bank Details given in J F Ziegler J P Biersack and U Littmark The Stopping and Range of Ions in Solids Pergamon Press New York 2003 35 S Cierjacks P Oblozinsky and B Rzehorz Nuclear data libraries for the treatment of sequential x n reactions in fusion materials activation calculations KfK 4867 1991 36 J Koning S Hilaire and M C Duijvestijn TAL YS Comprehensive nuclear reaction modelling Proceedings of the International Conference on Nuclear Data for Science and Technology Santa Fe NM USA September 26 October 1 2004 to be published 37 O N Jarvis Low activity materials reuse and disposal AERE R 10860 1983 38 R A Forrest Systematics of neutron induced threshold reactions with charged products at about 14 5 MeV AERE R 12419 1986 39 R A Forrest and D A J Endacott Activation data for some elements relevant to fusion reactors AERE R 13402 1989 40 S Cierjacks P Oblozinsky S Kelzenberg and B Rzehorz Development of a novel algorithm and product
226. properties Several bugs have been dealt with the most important was to correct the gamma dose rate from a point source which was a factor of 10 too large in both FISPACT 97 and 99 Also the nuclides masses from the decay data file are now used leading to more consistent mass calculations The main new features of FISPACT 2003 are the ability to input the EAF 2003 library and the inclusion of new reaction types mostly in readiness for a future extension of the energy range of the data libraries The main new feature of FISPACT 2005 is the ability to input the EAF 2005 library which in group form extends to an energy of 55 MeV This has involved most of the FISPACT arrays being significantly larger as the EAF 2005 library contains 62 637 reactions compared to only 12 617 in EAF 2003 There are many new reaction types typically with thresholds above 20 MeV and four new decay modes have been defined The data for sequential charged particles also extends above 20 MeV and can be used for runs using both 175 and 211 group libraries Three additional charged particle User Manuat Issue 1 Feb 2007 FISPACT Version 2007 Cross section files are available which require additional entries in the FILES file A bug that caused only some of the sequential charged particle data to be used was fixed consequently if using EAF 2003 data the contribution from SCPR will be larger than when using FISPACT 2003 The calculation of the prod
227. r ROUTES it is still necessary to switch on this feature with LOOPS by default it is not used Since version 3 0 the output of generic pathways has been possible This was introduced because if many of the nuclides on the path have isomeric states then a large number of separate User Manuat Issue 1 Feb 2007 UKAEA Fusion 112 FISPACT UKAEA Fusion pathways are identified all with the same basic structure and only differing by the presence of X IT X links Thus the two following pathways have the same generic pathway 50 9 21 MO lt ny Os B 2 Posta 2 Os B Generic By default generic pathways are listed but their output can be switched off by using the GENERIC code word FISPACT calculates the amount of the final daughter formed by a particular pathway in exactly the same way as for a full inventory except that the number of nuclides is very much smaller Only the nuclides in the pathway and a fictitious nuclide which acts as the sink for all the depletion modes of the nuclides and any isomers specified by the LOOPS code word are considered The fictitious nuclide is Fe which is assumed stable with zero reaction cross section This nuclide is used in other calculations impurities in an unreactive iron matrix so it is convenient to also use it for this purpose When using the pathway option either routinely or in special ru
228. re described in Table A17 1 Note that a modification number is used for each change made to the source code Table A17 1 List of FISPACT modifications Problem Solution Modification numbers 18 5 95 Inconsistent values of tritium in The specific tritium activity was 555 summary at end of each time stored and added to the total interval Only seen when masses z activity not specific So store the 1 kg used total activity for tritium 19 5 95 Titles of final summary at end of Change titles to show total values run say specific values The total values are shown and this is probably most useful 19 5 95 Would be useful to show total mass Add this feature 557 558 of material in final summary 24 5 95 Output of neutron spectrum in Changes made in COL172 559 560 Printlib is incorrect for XMAS subroutine because only first 69 172 group structures values of neutron spectrum written to COLLAPX file 24 5 95 The printing of the different group This feature added 561 562 563 structures does not distinguish between 172 and 175 groups 24 5 95 When an array created by ENFA Changes made in the ENDFPR ARRAY the new spectral data was subroutine so SPECN is input but if the original data was in reinitialised to 1 for those values more groups then the old data where data are not read in values remain This causes a problem in the Printlib output 12 6 95 The output y spectrum in the 22 Changes made to the OUTPUT
229. re then listed below WITH RESPECT TO CROSS SECTIONS NUCLIDE dN dx dN dx x N SENSITIVITY OF He 6 4 710 30 1 348E 09 Li 8 4 646 34 2 711 02 i He 6 1 609E 35 2 450E 08 Li 8 2 228E 38 6 914E 02 ERRORS IN NUCLIDE AMOUNTS He 6 2 2765E 12 1 0984E 12 4 8248E 01 Li 8 1 11170E 09 4 7476E 08 4 2502E 01 C 14 5 2430E 18 6 5638 13 1 2519 03 Be 10 1 1034 21 3 9654E 20 3 5937E 01 Similar output is obtained if the sensitivity with respect to the half life is requested Uncertainty estimates UKAEA Fusion At the end of each time interval the nuclides that contribute most to the activity heat output dose rate potential ingestion and inhalation doses clearance index and the beta and gamma heat outputs are calculated and a twenty given For each of the eight quantities the top twenty nuclides may be different so that more than twenty dominant nuclides in total will be calculated The residual after subtracting the contribution from the top ten is given as Rest so that the user can judge if there are actually more than twenty significant nuclides The contribution of each nuclide is given both absolutely and in User Manuat Issue 1 Feb 2007 FISPACT 71 percentage terms If there less than twenty radionuclides then only data for these is given The output of this list of dominant nuclides can be switched off by usin
230. rface Summary of output files The OUTPUT file corresponding to say the irradiation of an alloy with impurities followed by a series of cooling times can be large 0 2 2 MB and extracting information can be time consuming using the OUTPUT file viewer The EASY User Interface also allows the user to summarise the output displaying the required total quantities e g activity or y dose rate for each time interval in tabular form This can then be copied to the clipboard and pasted into another application such as a spreadsheet or written to a database file The Microsoft Access database format mdb is used for the database files the EASY User Interface can create open view the structure of and add data to a database file The information about the dominant nuclides and the pathways responsible for their production for each time interval can also User Manuat Issue 1 Feb 2007 UKAEA Fusion 182 FISPACT be summarised and presented in tabular form This can then be copied to the clipboard for further use Figure A15 1 shows a typical summary window while Figure 15 3 shows summary of the dominant nuclides for activity Summary of dominant nuclides pes e 486 14 827 16 424 8 583 0 441 0 368 0 352 0 208 633 0 036 19 344 9 991 0 518 0 433 0 411 0 219 Ingestion dose Inhalation dose 889 z 354 z 32 19 521 19 780 17 44 8 706 5 222 0 151 0 512 0 478 0 2
231. ribute to the total dose rate the following method is used to calculate an approximate spectrum The maximum energies for decays assumed in the method are given in Table A4 1 Table A4 1 Maximum y energies for various decay modes Bd KB The intensity in the i th group 7 is given by equation A4 1 a y e I aa SE a um MA A4 1 E l l a e where 14 arbitrary constant 7 E E UKAEA Fusion User Manual Issue 1 Feb 2007 FISPACT 97 Appendix 5 Sensitivity equations An important feature of FISPACT is the ability to calculate sensitivity coefficients of nuclide quantities to either reaction cross section or decay constant This development was done by Khursheed and was based on the work of James who implemented a similar facility in FISPIN This method relies on the quasi linearity of the inventory equations see Appendix 2 These are written in matrix form in equation A5 1 where N is a vector of nuclide quantities S Isa vector of source of nuclides due to fissions A is matrix of terms involving cross sections and half lives If x represents either 4 or o then differentiating equation 5 1 with respect to x and assuming that the order of differentiation can be exchanged then equation A5 2 is obtained d OA oS 2 1 21 2 x 572 Equation 5 2 has the same form and the same matrix as equation A5 1
232. roperties of various actinides it may be useful to be able to restrict which of these produce fission products Note that for a complete inventory this code word should not be used Examples of the use of this code word follow FISYIELD O0 None of the actinides will produce any fission products when fissioned EISYIERLD 2 0235 PU239 Only U and Pu will produce any fission products when they undergo fission FISYIELD 2 U238 AM241 238 All actinides except 280 and Am will produce fission products when they undergo fission This code word enables the total energy integrated neutron flux n cms to be specified for a particular time interval Note if several consecutive time intervals require the same flux value then it need only be entered once for these intervals Setting the total flux to zero gives a decay time step The flux MUST be set to 0 0 before using the code word ZERO An example of the use of this code word follows FLUX 1 5 15 For this particular time interval a total flux of 1 5 10 n cms will be used User Manuat Issue 1 Feb 2007 FISPACT 29 FUEL N1 IS J ATOMS J J 1 N1 Inputs the number 7 of nuclides and the identifier S J and the number of atoms ATOMS J for each nuclide The identifier can be specified either using the format TE129M or by the material number Note the material number is the identification given to the nuclide internally
233. rum NPULSE can take values 2 to 500 PULSE The value specified for the number of times to loop is invalid it must be in the range 2 500 Nuclide on pathway has different decay mode lt CHAINP gt One of the pathway nuclides does not decay to the next nuclide in the pathway Nuclide on pathway has no reactions lt CHAINP gt One of the pathway nuclides is followed by an R however the nuclide is so short lived that it has no cross section data in the library Nuclide on pathway is stable lt CHAINP gt One of the pathway nuclides is followed by a D however the nuclide is stable User Manuat Issue 1 Feb 2007 FISPACT 137 Number of fissionable parents 20 FISCHOOSE Can only specify a maximum of 20 fissionable parents Numeric value required for C LEVEL A numeric value MUST follow the code word Numeric value required for CONV CONV A numeric value MUST follow the code word Numeric value required for CONVS CONV A numeric value MUST follow the code word Numeric value required for DENSTY DENSITY A numeric value MUST follow the code word Numeric value required for FLUX2 FLUX A numeric value MUST follow the code word Numeric value required for FRACOK UNCERTAINTY A numeric value MUST follow the code word Numeric value required for FRACWT UNCERTAINTY A numeric value MUST follow the code word Numeric value required for GRSHOW GRAPH A numeric value MUST follow the code word
234. se summary gt gt Energy Figure A15 7 The EAF 2007 group cross section data window The cross section data can be plotted using the seventh toolbar button The many options and facilities of this window are fully explained in the Help file Neutron spectra Neutron spectra are central to all FISPACT calculations The EASY User Interface enables neutron deuteron and proton spectra to be stored referenced and plotted Clicking on the eleventh toolbar button in the main window Figure A15 1 displays the Neutron spectra window shown in Figure A15 8 Other spectra can be displayed by clicking on the Spectra menu item and selecting the required type This displays the spectra that are stored in the database and gives information about the User Manuat Issue 1 Feb 2007 UKAEA Fusion 186 FISPACT Materials UKAEA Fusion one selected in the list box This spectrum can be plotted and further spectra can be added to the database standard FISPACT FLUXES files are read A spectrum can be written out from the database in standard FISPACT format in order to carry out a FISPACT run The spectra database thus acts as a useful repository of spectra ensuring that they don t get lost Neutron spectra Oe XI FISPACT testcases 172v FISPACT testcases 172w JET AFP position 1 JET AFP position 2 SEAFP 2 PM2 IFW Type WITAMIN 175 groups 1 Write fles Description Cl
235. ser Manuat Issue 1 Feb 2007 UKAEA Fusion 224 FISPACT test r2 NOHEAD MONITOR 1 EAFV 4 AINP FISPACT Simplified reactions on S 34 FUEL 1 34 1 0E20 DENSITY 2 07 lt lt 1918 1 gt gt lt lt 1919 2 gt gt lt lt 1920 H 3 x lt lt 1921 3 gt gt lt lt 1922 4 gt gt lt lt 121 Cl 34m gt gt OVER 534 ACROSS 535 6 58663E 3 ACROSS 533 0 0 ACROSS 5130 0 0 ACROSS P34 0 0 ACROSS P33 0 0 ACROSS P32 0 0 ACROSS S132 ACROSS 5131 ACROSS 1918 ACROSS 1919 ACROSS 1920 ACROSS 1921 ACROSS 1922 OVER 35 ACROSS 536 5 64830E 3 ACROSS 534 0 0 ACROSS 5131 0 0 ACROSS P35 0 0 ACROSS P34 0 0 ACROSS P33 0 0 ACROSS S132 ACROSS 1918 ACROSS 1919 ACROSS 1920 ACROSS 1922 OVER 536 ACROSS 537 5 83017E 3 ACROSS 535 0 0 ACROSS 5132 0 0 ACROSS P36 0 0 ACROSS P35 0 0 ACROSS P34 0 0 ACROSS 5133 ACROSS 1918 ACROSS 1919 ACROSS 1920 ACROSS 1922 a ooo 0 mo Ol O OVER 37 ALAM 0 5 5 OVER CL35 ACROSS CL34 2 92143E 4 ACROSS 121 0 0 ACROSS P31 0 0 ACROSS CL36 0 0 ACROSS 535 0 0 ACROSS 534 0 0 ACROSS 533 0 0 ACROSS P33 0 0 ACROSS P32 0 0 ACROSS P34 0 0 ACROSS 1918 ACROSS 1919 ACROSS 1920 ACROSS 1921 ACROSS 1922 OVER CL34 ALAM 10 0 5 MIND 1 0 HALF DOMINANT 100 0 UNCERT 3 UKAEA Fusion User Manuat Issue 1 Feb 2007 FISPACT 225 LEVEL 100 1 FLUX 1 0E15 TIME 1 0 Y
236. ser Manuat Issue 1 Feb 2007 UKAEA Fusion 226 FISPACT UKAEA Fusion OVER CR54 ACROSS CR53 ACROSS TI50 ACROSS CR55 ACROSS V54 ACROSS V53 ACROSS V52 ACROSS TI51 ACROSS 1918 ACROSS 1919 ACROSS 1920 ACROSS 1922 OVER V52 ALAM 1 0E25 1 OVER V53 ALAM 1 0E25 1 OVER V54 ALAM 1 0E25 1 OVER MN53 ALAM 1 0E25 1 OVER MN54 ACROSS CR54 O0 ALAM 1 0E25 1 OVER MN55 ACROSS CR54 O0 OVER FE55 ACROSS CR54 O0 OVER FE57 ACROSS CR54 O0 OVER FE58 ACROSS CR54 O0 MIND 1 0 HALF DOMINANT 100 0 UNCERT 3 LEVEL 100 1 FLUX 3 64099E15 TIME 1 0 YEARS ATOMS END END User Manuat Issue 1 Feb 2007 FISPACT 227 Appendix 20 Non steady irradiation Introduction Modelling the irradiation history for a FISPACT calculation requires considerable understanding of the physics of the situation and the way in which FISPACT works Although several recommendations about non steady pulsed irradiations are made in the body of this document they are gathered together here with some additional background information Modelling an irradiation Most irradiation histories that correspond to realistic devices will be non steady This is because devices must be switched on and off for maintenance and because of unforeseen circumstances Also neutron sources cannot be expected to operate for long periods of t
237. sing the cross section library and writing the array file in separate runs A subsequent inventory run would then always require this code word It is also possible to put all the above steps together in a single input file when this code word would not be required This code word is thus typically used in all inventory runs an example of its use in an INPUT files follows AINPUT FISPACT Irradiation of SS316 steel COLLAPSE N2COLL This code word causes FISPACT to read the cross section library in 69 100 172 175 211 315 or 351 group format determined by N2COLL being 69 100 172 175 211 315 or 351 and combines this with the neutron spectrum to produce a l group effective cross section library which is used directly in subsequent runs Note that if no uncertainty data exists in the library as for deuteron and proton libraries then the code word NOERROR MUST be used before the code word COLLAPSE An example of the use of this code word follows COLLAPSE 211 FISPACT Collapsing EAF 2007 The cross section library in 211 groups will be used with the neutron spectrum read from stream 20 User Manual Issue 1 Feb 2007 UKAEA Fusion 20 FISPACT EAFVERSION NEAFV 7 ENFA UKAEA Fusion This code word MUST be used if a version of the European Activation File EAF cross section library prior to version 2007 is used It MUST be used before the other code words AINP or ENFA as it deter
238. ss sections and inserts these in the correct order into the correct place in the internal data storage space the array The modified array is not written to a file so that there is no permanent effect on the data libraries As the composition of a material changes during a run more nuclides will become available to act as targets for the charged particles User Manuat Issue 1 Feb 2007 FISPACT 145 Appendix 13 Platform differences This appendix explains how to install FISPACT and the EAF libraries on various computer systems FISPACT is currently available on two computer platforms e Personal Computer Intel or AMD processor running Windows 98 Windows NT 4 0 Windows 2000 Windows XP or Windows Vista UNIX workstation The details for each system are given below Personal computer FISPACT is supplied to users on a DVD FISPACT 2007 will run under Windows 98 Windows NT4 0 Windows 2000 Windows XP or Windows Vista The DVD contains the PC version of FISPACT a run time file and the EAF library The run time library is supplied by Salford Software To install 1 Insert DVD R disk assume drive D 2 Install the EASY User Interface see Appendix 15 by running d easy07ui setup 3 Install the Install EASY application by running d install easyNsetup 4 Run Install EASY by clicking the Start Programs Install EASY menu item Follow the onscreen instructions including the input of person
239. ssion product that is also a dominant nuclide can be produced by pathways that either include a fission reaction or do not In the latter case the formalism derived above is correct but if the pathway includes fission then it must be extended If fission occurs then this can involve one of several actinide nuclides This occurs because the input actinide is transmuted by say capture reactions and many fissionable actinides are therefore present All pathways from a particular actinide to the dominant nuclide contain the same fission cross section and thus errors in these pathways are completely correlated This means that the errors are added linearly rather than by the sum of squares procedure The set of pathways must therefore be partitioned into subsets labelled by the actinide from which it is produced This is shown by equation A9 16 Lp hel A9 16 where P indicates the j th pathway forming nuclide 7 the index 0 indicates pathways with no fission and the index m indicates that the fission was on actinide m To each of these subsets there corresponds a subset of j values S such that choosing a subset of j values defines a subset of pathways Equation A9 8 can then be rewritten as equation A9 17 xy X Zan PP RUN A9 17 jeSo m NjESm If A9 17 is used in the derivation then the final equation A9 15 is modified as shown in equation A9 18 User Manual Issue 1 Feb 2007 UKAEA Fusion 1
240. t decay library been used Daughter nuclide of reaction not in library lt COL172 gt The daughter nuclide of a reaction in the cross section library is not present in the decay library has the correct decay library been used Daughter nuclide of reaction not in library lt COL175 gt The daughter nuclide of a reaction in the cross section library is not present in the decay library has the correct decay library been used User Manual Issue 1 Feb 2007 UKAEA Fusion 134 FISPACT UKAEA Fusion Daughter nuclide of reaction not in library lt COL211 gt The daughter nuclide of a reaction in the cross section library is not present in the decay library has the correct decay library been used Daughter nuclide of reaction not in library lt COL315 gt The daughter nuclide of a reaction in the cross section library is not present in the decay library has the correct decay library been used Daughter nuclide of reaction not in library lt COL351 gt The daughter nuclide of a reaction in the cross section library is not present in the decay library has the correct decay library been used Decay library and index file not consistent lt ENDFPR gt A nuclide appears in the decay library which is not present in the INDEX file has the correct decay library been used Decay mode not allowed in library lt ENDFP gt A decay mode unknown to FISPACT has been found in the decay library has the correct decay library been used
241. t on the DOSE parameter these two outputs are for contact dose from a semi infinite slab of the material SOURCE FROM GAMMAS WITH ENERGY 0 20 MeV IS 2 67503 05 Sieverts hour 2 67503 07 Rems hour and for the dose from a point source at a specified distance DOSE RATE 1 g POINT SOURCE 1 0 FROM GAMMAS WITH ENERGY 0 20MeV IS WARNING 1 16480E 03 Sieverts hour 1 16480E 01 Rems hour If most of the dose rate is produced by nuclides with approximate Yy spectra then the following warning message will be given gt 20 OF DOSE FROM NUCLIDES WITH NO SPECTRAL DATA TREAT DOSE AND GAMMA SPECTRUM WITH CAUTION User Manuat Issue 1 Feb 2007 UKAEA Fusion 70 FISPACT Sensitivity output Li Li Be Be NUMBER NUMBER NUMBER NUMBER REACTION gt Li gt Li gt Li gt Li OF ATOMS OF OF ATOMS OF OF ATOMS OF OF ATOMS OF OO OO If the code word SENSITIVITY is used in the input then the sensitivity output is given at this point In the case of sensitivity with respect to cross section a part of the output is shown below In the first column is given the reaction e g Li n y Li that is varied the second column gives nuclide that 15 considered column three gives ON do and column four the sensitivity coefficient N go o N Using these sensitivity coefficients and uncertainty data for the cross sections the errors in the nuclides specified by ERROR a
242. t95 NOHEAD MONITOR 1 AINP FISPACT IRRADIATION LMJ FW imm thick DENSITY 2 7 MASS 848 23 2 B 78 57 C 21 43 SEON 211 SEQUENTIAL 1 O0 TAB4 44 MIND 1 E5 lt lt IRRADIATION HISTORY 1 YEAR 12 SHOTS gt gt HAZA HALF DOSE 1 SPECTRUM E gt gt PULSE 11 FLUX 1 02292 22 EVEL 100 1 TIME 1 0E 9 SPECTRUM EVEL 20 1 FLUX 0 0 TIME 30 DAYS SPECTRUM ENDPULSE lt lt gt gt FLUX 1 02292 22 EVEL 100 1 TIME 1 0E 9 ATOMS EVEL 20 1 FLUX 0 ZERO UNCERT 2 NOSTABLE TIME 1 0E 9 ATOMS ME 0 5 ATOMS ME 0 5 ATOMS ME 1 MINS ATOMS ME 1 HOURS ATOMS ME 5 HOURS ATOMS ME 0 75 DAYS ATOMS ME 1 0 DAYS ATOMS ME 1 DAYS ATOMS ME 2 DAYS ATOMS ME 2 DAYS ATOMS ME 1 DAYS ATOMS ME 1 DAYS ATOMS ME 1 DAYS ATOMS ME 1 DAYS ATOMS ME 1 DAYS ATOMS ME 5 DAYS ATOMS ME 10 DAYS ATOMS ME 10 DAYS ATOMS ME 10 DAYS ATOMS ME 10 DAYS ATOMS ME 10 DAYS ATOMS ME 50 DAYS ATOMS ME 100 DAYS ATOMS ME 252 DAYS ATOMS ME 0 76923 YEARS ATOMS ME 1 YEARS ATOMS IME 3 YEARS ATOMS 25 YEARS ATOMS JHHHHHHHHHHHHHHHHHHHHHHHHH Test96 Identical to Test26 Test100 Test104 dentical to Test70 Test74 User Manuat Issue 1 Feb 2007 UKAEA Fusion 172 FISPACT Test181 MONITOR 1 PROJ 2 NOERROR 11 FI NP SPACT IRRADIATION OF TI IFMIF MASS 1 0 1 TI MI 100 0 ND 1 E5 GRAPH 50112345 FLUX 1 0 13 ATOMS LEVEL 100 1 TIME 1 0 YEA
243. tallation A README file is supplied that details the installation and the QA procedures as well as the necessary adaptations required for use on the various UNIX systems User Manuat Issue 1 Feb 2007 FISPACT 147 Appendix 14 Standard test cases The following input files constitute the set of standard test cases This set covers all the code words and is supplied to users to enable them to confirm that a new installation is working correctly COLLAPSE typical case for 100 groups COLLAPSE 100 FISPACT COLLAPSE EAF4 100 WITH FW EEF END END OF RUN WRITE typical case for 175 groups using TAPA option SPEK ENFA EAF DECA4 00X EAF4 175 DATA EEF121M GPJ TAPA FISPACT WRITE DATA TO ARRAY FILE END END OF RUN PRINTLIB typical case with option to print only cross sections AINP FISPACT PRINTLIB OF FW EEF 0 END END OF PRINTLIB Test1 NOHEAD AINP FISPACT IRRADIATION OF TI EEF FW 1 0 MW M2 MASS 1 0 1 TI 100 0 MIND 1 E5 GRAPH 51112345 WALL 1 00 ATOMS LEVEL 100 1 TIME 2 5 YEARS HAZA ATOMS 1 MINS ATOMS 1 HOURS ATOMS TIME 1 DAYS ATOMS 1 7 1 DAYS ATOMS YEARS ATOMS User Manuat Issue 1 Feb 2007 UKAEA Fusion 148 FISPACT Test2 NOHEAD AINP FISPACT IRRADIATION OF TI EEF FW 1 0 MW M2 46 1 00619E24 TI47 9 18148E23 48 9 28210E24 TI49 6 91755E23 TI50 6 79178E23 MIND 1 E5 GRAPH 300123
244. ters If the message ends with a subroutine name in angle brackets lt gt then the error has occurred in the named subroutine UK AEA should be contacted if a solution to the problem cannot be found Error Messages 1 or 2 required DOSE Only 22 or 24 gamma groups can be specified 69 100 172 175 211 315 or 351 required for NZ2COLL COLLAPSE The neutron spectrum MUST be in 69 100 172 175 211 315 or 351 groups All nuclides must be fissionable FISYIELD Specified nuclides MUST be fissionable actinides ATWO and CLEAR both used ATWO Only one of these two code words can be used per case ATWO and CLEAR both used CLEAR Only one of these two code words can be used per case Cannot find uncertainty no library data UNCERTAINTY There is no uncertainty data in the cross section library so cannot work out error estimates only give pathway information IUNCER MUST be 0 3 or 4 Characters required for NEWNAM NEWFILE File name MUST consist of characters Chemical symbol not recognised lt CNVTXT gt The chemical symbol MUST represent one of the elements H Fm and be in normal form e g Chemical symbol not recognised lt RENUCL gt The chemical symbol MUST represent one of the elements H Fm Chemical symbol not recognised MASS The chemical symbol MUST represent a naturally occurring element User Manuat Issue 1 Feb 2007 FISPACT 133 Chemical symbol not recognised PARTITION
245. ther pathways with more links need to be considered FRACWT 0 005 is the fraction of the total contribution below which pathway contributions are not output NMAXB 3 is the maximum number of links allowed in pathways except where tritium is the final daughter NMAXB MUST be in the range 1 5 NMAXR 3 is the maximum number of links for tritium production NMAXR MUST be in the range 1 6 112 is the maximum number of links when only captures and decays are considered in a pathway MUST be in the range 1 12 ZZZLVL 50 0 is the first level parameter for calculation of the pathways IUNCER following all the other parameters allows values 0 1 2 or 3 to be input again so that after resetting the default values an actual calculation with the new values can be done Note if a time interval prior to the irradiation is specified then IUNCER MUST be set to 0 or UNCERTAINTY not used for this time interval User Manuat Issue 1 Feb 2007 UKAEA Fusion 56 FISPACT UKAEA Fusion Note if more than one irradiation time is considered in a run then UNCER MUST be set to 0 for all time intervals except the first If the uncertainty estimate 1s to be sensible for such a pulsed irradiation then the initial irradiation should contain the majority of the neutron fluence See Appendix 9 for more details on uncertainties and Appendix 20 for non steady irradiations Note that very large v
246. time unit MINS HOURS DAYS YEARS Note it is important when inputting times that it is the interval time not the total time that 15 specified Thus for cooling steps the time printed on the inventory is the sum of all the previous cooling time intervals after the code word ZERO Examples of the use of this code word follow ZERO TIME 2 5 YEARS ATOMS TIME 7 5 YEARS ATOMS Following irradiation the start of cooling is specified by the code word ZERO Inventories at the times of 2 5 and 10 years are output UNCERTAINTY UNCER 0 lt FRACOK 0 98 FRACWT UKAEA Fusion 0 005 NMAXB 3 NMAXR 3 NMAXC 12 ZZZLVL 50 0 IUNCER This code word allows user control of the uncertainty estimates and pathway information that is calculated and or output for each time interval This is primarily specified by the parameter IUNCER 0 The allowed values are User Manual Issue 1 Feb 2007 FISPACT 55 0 No pathways or estimates of uncertainty are calculated or output 1 Only estimates of uncertainty are output although all the pathway information has been calculated 2 Both estimates of uncertainty and the pathway information are output 3 Only the pathway information is output 4 Allows default values to be reset for a particular run Particular values can be specified by the following parameters only for this value of UNCER FRACOK 0 98 is the fraction of the daughter nuclide produced such that no fur
247. tinely Inclusion of Can treat actinide impurities including New in version 3 actinides fission in same detail e g pathways as other nuclides Uncertainties An uncertainty file and the pathway New in version 3 Routine method for the dominant nuclides production of uncertainties enables uncertainty on activity y dose is possible rate heating and biological hazards to be calculated for each time interval particle reactions included by inputting new data libraries file with the syntax lt lt comment Monitoring the The progress of a run can be monitored New in version 3 progress of a run by echoing the code words to the current window UKAEA Fusion User Manuat Issue 1 Feb 2007 FISPACT 9 Feature Details Comments Loop construct in If the irradiation consists of many New in version 97 the input file repeated steps the construction of the input file is made simpler by allowing some of the code words to be repeated Fission product Fission product production from New in version 97 production can be individual actinides can be controlled to switch off aid in understanding the various pathways involving fission Version available for Windows 95 and Windows NT4 0 New in version 97 32 bit PC operating versions developed systems New photon All dose rates are calculated using a set New in version 97 absorption data of photon absorption data A new set 18 read from a file rather than being stored in the
248. tion library COLLAPX is now used as input with the decay data library DECAY to produce the condensed library data ARRAYX Figure 4 shows a standard inventory calculation that used the condensed library data ARRAYX as input and produces and additional output the GRAPH file that will be used for plotting a graph of the decay curve Note that in all cases the FILES file is used to show which input and output files are connected and the location of all the other standard files and this MUST exist If activation due to deuteron induced reactions is to be calculated then the code word PROJECTILE must be used followed by the parameter 2 If activation due to proton induced reactions is be calculated then the code word User Manuat Issue 1 Feb 2007 UKAEA Fusion 18 FISPACT PROJECTILE must be used followed by the parameter 3 The NOERROR code word must also be used in both cases as there 15 no uncertainty file for the two charged particles Figure 2 Files used by FISPACT to produce collapsed library FISPACT ARRAYX d4 Figure 3 Files used by FISPACT to produce condensed library ARRAYX t Figure 4 Files used FISPACT for a standard run UKAEA Fusion User Manuat Issue 1 Feb 2007 FISPACT 19 AINPUT This code word causes condensed library data to be read The user is recommended to follow the standard procedure of collap
249. to the nuclide on the pathway then it is important to be able to also include the isomer in the calculation TLOOP specifies the maximum half life seconds of an isomer such that the reaction X n n X and subsequent decay X IT X will be included in the calculation even though X is not on the pathway User Manual Issue 1 Feb 2007 FISPACT 37 Pathways for the formation of the dominant nuclides that are calculated at the end of each time interval do include the isomers automatically If LOOPS has not been used then the value of TLOOP is set to T 1000 where T is the time interval specified by the TIME code word An example of the use of this code word follows LOOPS 20 0 In the case of the pathway Sc n y 6Sc n y Sc there is an isomeric state 9mSc with half life of 18 7 s Using the code word as shown would mean that this isomer was included in the calculations See Appendix 8 for more information on pathways MASS TOTM INDX2 SVM I XP I l 1 INDX2 This code word allows the input of the total mass TOTM kg and the number NDX2 of elements in the material to be irradiated For each element the chemical symbol SYM I e g FE and the percentage by weight XP I are then input This code word enables elements to be input with the number of atoms of each isotope calculated by FISPACT using natural abundance data that are stored internally The values used for natural abundance are given in Appendix
250. trons in the relevant energy range 200 keV 200 keV 5 MeV and gt 5 MeV and summing The PRINTLIB output for a particular neutron spectrum gives the fractions as 0 4120 0 3871 and 0 2009 respectively The calculation of the weighted yield is given in Table 5 Table 1 Data for in decay library 597 8 discrete lines User Manuat Issue 1 Feb 2007 UKAEA Fusion 202 FISPACT Table 2 Fission yield library data for Fissioning nuclide Energy eV Independent yield 253107 34639 107 4919 o0 _ i40 00 7 7526 10 Table 3 FISPACT PRINTLIB output data for y energy in GROUP 15 2 0 2 5 MeV 7 012 10 MeV y energy in GROUP 19 5 0 6 5 MeV 1 482 10 MeV Table 4 Binning of y spectrum data to FISPACT structure 7x0 6 45455 10 x 0 33 5 27273 10 x 0 33 UKAEA Fusion User Manuat Issue 1 Feb 2007 FISPACT 203 Table 5 Calculation of weighted fission yields Fissioning nuclide Expression Result MeV MM MET 3 4637 10 x 0 4120 x 100 0 1 4270 107 Pu 7 7526 10 x 0 4120 x 100 0 31941107 Looking at various parts of the previous tables it can be seen that e Comparison between Table 1 row 1 and Table 3 row 1 shows agreement for half life e Comparison between Table 1 row 2 and Table 3 row 3 shows agreement for B energy Comparison between Table 1 row 3 and Table 3 row 4 shows agreement for y energy e Com
251. uat Issue 1 Feb 2007 FISPACT 165 spectra GRPC 99 172 FISPACT SPECTRAL MODIFICATION 99 APOLLO 172 XMAS END END Test51 NOHEAD AINP FISPACT PWR FUEL 3 1 U235 POY Paluel DENSITY 10 1 FUEL 2 U235 7 948E22 U238 2 453E24 MIND 1 E5 GRAPH 51112345 FLUX 3 25E 14 ATOMS LEVEL 20 8 TIME 30 4375 DAYS 1 41 ATWO DOSE 1 ATOMS TIME 60 875 DAYS ATOMS TIME 91 3125 DAYS ATOMS TIME 182 625 DAYS ATOMS TIME 182 625 DAYS ATOMS TIME 182 625 DAYS ATOMS LEVEL 20 1 FLUX 0 NOSTABLE ZERO TIME 60 ATOMS TIME 1 DAYS ATOMS TIME 29 4375 DAYS ATOMS TIME 152 1875 DAYS ATOMS TIME 182 625 DAYS ATOMS TIME 2 YEARS ATOMS TIME 2 YEARS ATOMS TIME 5 YEARS ATOMS END END User Manual Issue 1 Feb 2007 UKAEA Fusion 166 FISPACT Test52 NOHEAD MONITOR 1 AINP FISPACT PWR FUEL 3 1 U235 POY Paluel DENSITY 10 1 FUEL 2 U235 7 948E22 U238 2 453E24 MIND 1 E5 FISCHOOSE 5 U235 U238 PU239 PU240 PU242 HAZA HALF GRAPH 51112345 FLUX 3 25E 14 LEVEL 20 50 TIME 730 5 DAYS UNCERT 2 TAB1 41 ATWO DOSE 1 ATOMS LEVEL 20 1 FLUX 0 NOSTABLE ZERO TIME 60 ATOMS TIME 1 DAYS ATOMS TIME 29 4375 DAYS ATOMS TIME 152 1875 DAYS ATOMS TIME 182 625 DAYS ATOMS TIME 2 YEARS ATOMS TIME 2 YEARS ATOMS TIME 5 YEARS ATOMS END END Test60 NOHEAD MONITOR 1 COLLAPSE 69 FISPACT THREE COLLAPSES NEWFIL
252. uat Issue 1 Feb 2007 UKAEA Fusion 120 FISPACT MN es oar NE en pati A9 26 If both nuclides 1 and 2 are short lived then gt gt 1 and the exponential can be set to zero This limit is given in equation A9 27 N 0 NA UA suse ease bete A9 27 Equations A9 26 and A9 27 suggest that in these two limits the number of atoms of the final nuclide in the pathway is obtained by multiplying the starting number of atoms by a factor for each link If the nuclide is long lived then the factor is oot while if the nuclide is short lived then the factor is 0 A In addition in the case of long lived nuclides there is a multiplicative constant factor Pathways containing arbitrary number of reactions UKAEA Fusion In general there can be an arbitrary number of links if there are 0 1 links then equation A9 28 expresses how the number of atoms is related to the number of atoms 1 dt na N ntl JUAN rre n aot sepatu A9 28 This is valid for all n gt 1 A solution can be found using the method of Laplace transforms The Laplace transform of t is R p which is defined in equation A9 29 je Noa SON A9 29 Using the standard result for the Laplace transform of a differential given by equation A9 30 it is possible to transform equation A9 28 as shown in equation 49 31 1 m 2
253. uction of fission products in previous versions was also found to be lacking changes have been made in FISPACT 2005 to correct this bug The main new feature of FISPACT 2007 is the ability to input the EAF 2007 library which contains data for deuteron and proton induced reactions in addition to the neutron induced data available in previous versions The data for the three incident particles in group form extend in energy to 55 MeV This extension has involved the definition of many new data strings so that pathway data can reflect the particular incident particle Many of the FISPACT arrays are larger as the EAF 2007 neutron induced library contains 65 565 reactions compared to 62 637 in EAF 2005 New fission yield data are available from the JEFF 3 1 library for neutrons while new data files on fission yields from deuteron and proton induced fission can also be used FISPACT 2007 has been tested running under the Windows 98 Windows NT4 0 Windows 2000 Windows XP and Windows Vista operating systems It runs in an MS DOS Command window and there is a 32 bit version of the EASY 2007 User Interface FISPACT 2007 can be run on a UNIX workstation running a variety of operating systems however there is no UNIX version of the EAS Y 2007 User Interface The main features of FISPACT and the version that first included it are summarised in Table 1 User Manual Issue 1 Feb 2007 UKAEA Fusion 8 FISPACT Table 1
254. ve a continuous energy distribution if the mean energy is used for The value of Z used in equation 7 2 is calculated from equation A7 3 LEX d d cei bati d A7 3 j where Z atomic number of the j th element n atomic fraction of the j th element number of atoms of j total number of atoms Only a subset of all the nuclides in the decay library needs to be considered for bremsstrahlung production The nuclides shown in Table A7 1 may make a contribution to the y dose rate because of bremsstrahlung emission from energetic B particles The bremsstrahlung correction can be estimated by including nuclides from the mass range of interest using the code word BREM The following criteria applied to the EAF DEC 2007 decay library give the nuclides shown in the Table The nuclide is radioactive with a half life gt 0 1 years or in the case of a short lived nuclide the half life of the parent 2 0 1 years nuclide is radioactive with a half life lt 5 0 1019 years The nuclide has an average B energy gt average The nuclide has an average B energy gt 0 145 MeV User Manuat Issue 1 Feb 2007 UKAEA Fusion 108 FISPACT UKAEA Fusion Column 1 in the Table specifies the nuclide giving the bremsstrahlung correction column 2 the half life of the nuclide or the parent column 3 the decay parent and column 4 the percentage branching ratio of the nu
255. ve diagram all the nuclides are stable except Cr In addition to the reactions shown other reactions on Cr Cr and are possible The values of the cross sections have been changed so that they agree with the values used in the benchmark calculation The values chosen can be seen in the FISPACT input file in the Annex In addition various half lives and cross sections have been changed so that the numbers of atoms of each chromium nuclide are determined solely by the pathway shown above The values of the cross sections used for the calculations are included in the table below 2 n n 5 26653E 3 4 37656E 18 4 37656E 18 0 00000 1 49953E 2 1 19852E 15 1 19852E 15 0 00000 E eel 6 29661E 11 6 29661E 11 0 00000 Two Bateman calculations are required in this case The agreement between FISPACT and the analytical solution demonstrates the validity of the methods used in FISPACT UKAEA Fusion User Manual Issue 1 Feb 2007 FISPACT 217 Annex FISPACT input files test di NOHEAD MONITOR 1 EAFV 4 AINP FISPACT DECAY OF H 3 FUEL 1 H3 1 0E20 DENSITY 8 988E 5 MIND 1 E5 GRAPH 10 11 UNCERT 0 LEVEL 100 1 FLUX 0 0 ZERO TIME 1 0 YEARS ATOMS TIME 4 0 YEARS ATOMS TIME 5 0 YEARS ATOMS TIME 5 0 YEARS ATOMS TIME 5 0 YEARS ATOMS TIME 10 0 YEARS ATOMS TIME 10 0 YEARS ATOMS TIME 10 0 YEARS ATOMS TIME 50 0 YEARS ATOMS END END test d2 NOHEAD MONITOR 1
256. y values in the file The energy regions are defined in Table A9 1 Threshold reactions contain two uncertainty values if the threshold is below 20 MeV Table A9 1 Definition of uncertainty energy regions Low energy region Medium energy High energy Extended region region energy region 1 0 105 eV Ev 20 20 60 MeV End of 1 v End of behaviour of the resolved resonance cross section region of the cross section The systematic values of the error factor used for capture and fission reactions are given in Table 49 2 Table A9 2 Systematic values of error factor f used for capture and fission reactions Reaction Low energy Medium energy High energy Extended I mp region Capture 7 S A AET TAW 2 0 2 0 In EAF 2007 the general principle that has been followed is that wherever experimental data are available these are used to User Manual Issue 1 Feb 2007 UKAEA Fusion 114 FISPACT estimate the error factors f for threshold reactions The remainder of the error estimates are based on systematics FISPACT uncertainty estimation UKAEA Fusion FISPACT can use the sensitivity coefficients defined in Appendix 5 to calculate uncertainties in the number of atoms of a particular nuclide due to an uncertainty in a cross section While practicable for particular cases the computer time involved in using this method for routine estimation of uncertaint

fispact-2007: User manual - Culham Centre for Fusion Energy

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