AMERICAN NUCLEAR SOCIETY
Contents
1. Nucl Technol 135 123 2001 14 I LUX J VEGH F ADORJAN and L BURGER Ex periences with the Upgraded VERONA u VVER 440 Core Monitoring System presented at Specialists Mtg Advanced Information Methods and AI in NPP Control Rooms Halden Norway September 1994 15 K A BARMSNES T JOHNSEN and C V SUNDLING Implementation of Graphical User Interfaces in Nuclear Ap plications presented at Topl Mtg I amp C of VVER Prague Czech Republic April 1997 NUCLEAR TECHNOLOGY VOL 139 AUG 2002 NT 8 01008 11 11 05 31 02 7 32 pm Page 11 V gh et al ON LINE PLANT INFORMATION SYSTEM FOR EMERGENCY RESPONSE 16 T AKERBAEK and M LOUKA The Software Bus an 17 J MIETTINEN and A HAMALAINEN SMABRE Code Object Oriented Data Exchange System HWR 466 Organi Manual Version 4 2 Vols 1 5 Technical Research Centre of zation for Economic Cooperation and Development Apr 1996 Finland Espoo Finland Oct 1992 Janos V gh MS physics 1979 Dr Univ nuclear science and engineering E tv s Lor nd University Budapest Hungary 1982 is a senior research scien tist at the Atomic Energy Research Institute of the Hungarian Academy of Sci ences working on developing on line process information and core surveillance systems for nuclear power plants His background includes development of real time expert systems critical safety functions monitoring software and safety parameter displays Csaba Majo2. al Ws Ss ss a S SG ISOLATION Y N i Ol 0 Cl Gl ce Gell Ga GE JOP Ce a A A A A HERMETIC AREA RELEASE Die oo a r e i iT Pd liao PRZ PRESSURE PRZ dP dt PRIMARY dT dt dT RPV TNDT EJECTOR ACT BOX PRESSURE PRZ LEVEL SG ACTIVITY bar bim ESERSE Ch EEEH C BRIT keq mbar mm kBq PRZ LEVEL PRZBTLYL P YA P YB ECCS TH FLOW ECCS TJ FLOW THS0 level TH60 level TH70 level TH80 level PRZ SAFETY VLV mm mm MPSS bar YI ih ERT th mm mm mm CEERSEPS mm BOX PRESSURE BOX TEMPERATURE BOX dP dt BOX ACTIVITY SPRINKLER FLOW GAS ACTIVITY IODINE ACTIVITY AEROSOL ACT mbar 2G BEER mbm EATE keg Te th kBq kBq Bq CSF Unit Detailed Archive Terminal Unit Screen Journal Event Simulator Overview Overview Help Replay window Selection Hardcopy Preparation Lists Display Fig 2 Main format of the CERTA VITA SPD showing a simulated transient six subwindows The detailed analysis of important plant variables can be performed on the second level Here the analysis is supported by predefined trend groups see Fig 3 plant pictures parameter summary tables logs and p T diagrams Predefined trend groups and pictures can be called easily from the main SPD screen by dedi cated pushbuttons see Fig 2 An important service of the VITA user interface is a convenient on line and his torical trending facility which is similar to the trend pack age applied in the VERONA core monit
3. 4 available time until anticipated core uncovery when RPV level is the same as the top of the fuel Figure 3 shows the results of a break calculation for a simulated transient pressurizer relief valve opening as shown on the VITA SPD screen During the development of the break parameter es timation module a large number of trial runs were per formed by a stand alone version of the program The results of these calculations were adjusted to the results NUCLEAR TECHNOLOGY VOL 139 AUG 2002 NT 8 01008 9 11 05 31 02 7 32 pm V gh et al 10 Cold leg LOCA Paks full scope simulator _ Calculation z T a 5 A oO a D o 0 500 1000 1500 2000 2500 3000 3500 4000 Time s 10 Cold leg LOCA 1600 7 T r i Paks full scope simulator 1400 k l Calculation w 1200 FA 1000 i MEN ENN w 2 lo x 600 pep L mo 400 p 200 il o LOUD The i i i i i 1500 2000 2500 3000 3500 4000 Time s 0 500 1000 Fig 4 Estimated primary pressure and break flow compared to simulated values 10 cold leg LOCA obtained from the Paks full scope simulator The simu lator uses the SMABRE code system for the thermal hydraulic modeling of LOCA transients The adjustment resulted in the a coefficients used in Eq 1 and the flow resistance constants required by the junction and break flow calculations The basic requirement during this adjustment was to achieve an
4. HTTP link Data link DECNet Data link Serial line Plant Computer System X terminal1 VITA computer Unit 2 is the same as Unit 1 Unit 4 is the same as Unit 3 Fig 1 Architecture of the communication between the CERTA and the Paks NPP parameters Plant satus monitoring is organized accord integrity primary coolant inventory hermetic area integ ing to seven critical safety functions CSFs reactivity rity and radioactive release The main parameters char core cooling secondary heat removal primary circuit acterizing the status of the seven CSFs are displayed in 4 NUCLEAR TECHNOLOGY VOL 139 AUG 2002 NT 8 01008 5 11 05 31 02 7 32 pm V gh et al Page 5 ON LINE PLANT INFORMATION SYSTEM FOR EMERGENCY RESPONSE 25x 4 THX THY THW 8 JAN 2001 17 46 48 1 HA Z HA 3 HA 4 Q Dw E IC POWER RANGE K IC INTR RANGE F IC SRC RANGE K IC PERIOD F IC PERIOD CRD GROUP VI BORON CONC Cys CT sec MERE sec ERI cm gikg J CORE Tout AVG BOILING MARGIN HOT LEG TEMP COLD LEG TEMP REACTOR dP PRZ LEVEL RPV LEVEL Py 2c 26 66 C 2C 2C bar EEE mm mm SG LEVEL AVG SG PRESS AVG SG STM FLOW AVG SG FEEDWTR AVG T YA T YB mm bar th th 2 AEFWTR FLOW SUM EDD th EF WTR FLOW SUM th TIME FROM SCRAM ALL CRD DOWN ESSE min MCP ON OFF LOOP OP CL NAT CIRC Y N PRIMARY CIRCUIT INTEGRITY PRIMARY COOLANT INVENTORY
5. NSD Simulated scenarios can be recorded and stored in data files at the simulator computer and 5 NT 8 01008 6 11 05 31 02 7 32 pm V gh et al Page 6 ON LINE PLANT INFORMATION SYSTEM FOR EMERGENCY RESPONSE 8 JAN 2001 1628 18 1 a 210 2 oo ia 7 1 4E 04 0 PRZ level mm 50 1 2E 04 1 0E 04 mm 000 0 RPV water level mm 6000 0 1 1E 04 1000 0 750 25 0 Break flow t h 500 375 t h 250 125 0 ECCS TH flow t h 211 5 CSF Detailed Overview Help Terminal window Archive Replay Primary pressure bar Simulator Display Journal Preparation Unit Screen Selection Hardcopy Fig 3 Detailed CSF analysis trend group showing the results given by the break parameter estimator module for a simulated transient replayed later at the VITA main computer This feature is utilized extensively for training purposes In recent years a large library of simulated transients has been com piled containing 50 important VVER 440 transients When one of the selected scenarios is replayed at the CERTA the NSD personnel can follow the changes of the main process parameters on the SPD screens and on trends in the same manner as during a real on line emergency Simulator data are treated as data coming from the fifth Paks unit by the VITA algorithms the performed calculations and the available display formats are ex actly the s
6. gateway can act as a general data server for the CERTA because it can collect any required measured or calculated signal through the plant network and transfer it to the NSD At present the CERTA gateway collects data from the plant central information system and from the new process computers of units 1 2 and 3 These upgraded process computer systems PCSs consist of several re dundant servers including the duplicated COMP com putational server pair which among others is dedicated to serving external data requests By using this ap proach the CERTA gateway can collect all the required data directly from the PCS Data from unit 4 are still accessed through the central plant information system which collects data from the old plant computers from the VVER on line analysis VERONA core monitoring systems and also from common plant dosimetry and radiation level monitoring systems Along with the progress of the ongoing plant computer upgrading project the CERTA gateway will gradually switch over to the new process computers at all units The scheme of the aforementioned communication architecture is illus trated in Fig 1 In the actual version of the CERTA VITA system 400 main process parameters are transferred to the emergency response center for each reactor unit To ensure data integrity and to avoid illegal user activi ties network connections are protected by firewall com puters on both sides The CERTA internal netw
7. optimal agreement with the simulated results for the primary pressure the break flow and the primary coolant mass Figure 4 shows the comparison of estimated and simulated curves for a 10 cold leg LOCA case while in Fig 5 results obtained for a SGTR case are plotted A large number of valida tion runs were then performed with the off line program the module was tested for all important VVER 440 LOCA and SGTR cases and generally a good agreement was achieved for the primary pressure and break flow curves After the stand alone break model was tuned the mod ule was integrated into the on line VITA system and extensive on line testing was performed to check the performance and reliability of the break calculation module AUG 2002 NUCLEAR TECHNOLOGY VOL 139 Page 9 ON LINE PLANT INFORMATION SYSTEM FOR EMERGENCY RESPONSE 1 Steam generator tube rupture T T F oat T i Paks full scope simulator Calculation T lt a 2 DE AES We eres patel aai a EEA a ehaieed Tamme paced RES 5 a oO a D o EEIN INOI POTEN 0 i 0 500 1000 1500 2000 2500 3000 3500 4000 Time s 1 Steam generator tube rupture 40 gt Paks full scope simulator 35 decks _ Calculation y nN ua Break flowrate kg s a 8 i i i i i i i 0 500 1000 1500 2000 2500 3000 3500 4000 Time s Fig 5 Estimated primary pressure and break flow compared to simulated values S
8. standard HTML browser on their workstations In the remote version of PLASMA the same information is available as in the control room The users can display mnemoschemes and CSF status trees and can perform EOP browsing The default cycle time of information updating on the remote screens is 10 s to keep the load of the 64 kilobyte s communication line at a reasonable level The remote version has been tested with simulated transients and response times at the CERTA and the reliability of the HTTP connection were found to be acceptable Based on these positive experiences the remote connection was installed on units 1 2 and 3 in 2001 providing an additional diverse tool for the continuous remote inspection of the NPP Unit 4 will have this connection installed during 2002 IV CONCLUSIONS Main system and human machine interface charac teristics of the CERTA VITA system are outlined in this paper The system was built during 1997 2001 at the Emergency Response Center of the Hungarian Nuclear Safety Directorate and performs on line remote inspec tion of the Hungarian Paks NPP Furher developments were carried out after the initial installation of the sys tem the construction of a communication gateway com puter at the NPP transferring simulated data from the Paks training simulator to the CERTA feeding accident prediction analysis and management codes with on line input data presentation of Picasso 3 mnemoschemes on the operat
9. AMERICAN NUCLEAR SOCIETY 555 North Kensington Avenue Tel 708 352 6611 La Grange Park Illinois E Mail NUCLEUS ans org 60526 5592 USA http www ans org Fax 708 352 0499 AUTHOR PROOF INSTRUCTIONS Your paper is scheduled for publication in a forthcoming issue of Nuclear Technology You are receiving your author proofs electronically You will not receive anything in the regular mail Please follow these procedures 1 Download your author proof print it out and review it 2 Proofread your paper very carefully This will be your final reading before publication This material has been copyedited and typeset from the disk electronic file that you provided with your manuscript or from the hard copy that you provided and it has been transmitted to you through the Internet This process produces excellent results but please check the following especially carefully translation of Greek letters accented letters special characters superscripts and subscripts and mathematical symbols typesetting of mathematics and formatting of tables Also note that your manuscript has been copyedited according to ANS technical journal style please check that technical meanings have been unaffected by the copyediting Please limit your corrections to those that are absolutely necessary because changes at this stage of processing are both time consuming and expensive 3 Please note that the typesetter has used low resolution electronically s
10. GTR one tube I11 G Remote Critical Safety Functions Monitoring System The process computer reconstruction is an ongoing project at the Paks NPP Units 1 2 and 3 are already running with the new PCS while unit 4 will switch to the new system in 2002 The new PCS includes an on line plant safety monitoring and assessment PLASMA system which is dedicated to critical safety functions monitoring CSFM and provides operator support dur ing the execution of the symptom based EOPs The sys tem has been built primarily for control room use during emergencies but a Hypertext Markup Language HTML based remote version was developed as well This ver sion was basically designed to support safety engineers working outside the control room These external users are served through the so called WEB server node of the new plant computer see Fig 1 The Nuclear Safety Directorate showed considerable interest in having a re mote access at the CERTA center to the information pre sented by the PLASMA system therefore a pilot version 9 NT 8 01008 10 11 05 31 02 7 32 pm V gh et al was installed and tested at the Paks full scope simulator in November 2000 The PLASMA runs in the PCS con figuration connected to the simulator and experts work ing in the CERTA can connect to the WEB server node of the PCS HTTP links are established through the ded icated digital communication line see Fig 1 hence external users only need a
11. Idaho National Engineering and En vironmental Laboratory Jan 1995 5 SCDAP RELAPS5 Manuals NUREG CR 6150 U S Nu clear Regulatory Commission 6 K K MURATA User s Manual for CONTAIN1 1 Sandia National Laboratories Nov 1989 7 J U VALENTE and J W YANG MAAP 3 0B Code Evaluation FIN L 1499 Brookhaven National Laboratory Oct 1992 8 MELCOR Computer Code Manuals Primer and User s Guides Version 1 8 3 NUREG CR 6119 SAND93 2185 Sandia National Laboratories Mar 1995 9 L KOBLINGER I NEMETH P P SZABO N FULOP and A KEREKES Simulator Software for Interactive Mod eling of Environmental Consequences of Nuclear Accidents presented at Int Symp Environmental Impact of Radioactive Releases Vienna Austria 1995 10 Generic Assessment Procedures for Determining Protec tive Actions During a Reactor Accident IAEA TECDOC 955 International Atomic Energy Agency Aug 1997 11 PSA Based Regulatory Applications Institute for Elec tric Power Research VEIKI Budapest Hungary 2000 12 GY GYENES A Severe Accident Simulator for the Paks Nuclear Power Plant IAEA TECDOC 762 International Atomic Energy Agency Sep 1994 13 A HORNAES J E HULSUND J VEGH CS MAJOR CS HORVATH S LIPCSEI and GY KAPOCS The EOP Visualization Module Integrated into the PLASMA On Line Nuclear Power Plant Safety Monitoring and Assessment Sys tem
12. RVATH and ZOLTAN HOZER KFKI Atomic Energy Research Institute H 1525 Budapest P O Box 49 Hungary FERENC ADORJAN IVAN LUX and KRISTOF HORVATH Hungarian Atomic Energy Authority Nuclear Safety Directorate H 1539 Budapest P O Box 676 Hungary Received February 1 2001 Accepted for Publication March 14 2002 The main design features services and human machine interface characteristics are described of the CERTA VITA on line plant information system devel oped and installed by KFKI AEKI at the Nuclear Safety Directorate NSD of the Hungarian Atomic Energy Au thority HAEA in cooperation with experts from the NSD The Center for Emergency Response Training and Analy sis CERTA located at the headquarters of NSD Buda pest Hungary was established in 1997 The center supports the NSD installation radiological monitoring and advisory team in case of nuclear emergencies with appropriate hardware and software for communication diagnosis prognosis and prediction The vital informa tion transfer and analysis VITA system represents an I INTRODUCTION In 1996 the Hungarian Atomic Energy Authority Nuclear Safety Directorate HAEA NSD launched an ambitious project to install a modern Center for Emer gency Response Training and Analysis CERTA The project has been financially supported by the Hungar ian and British governments by the International Atomic Energy Aency IAEA and by the European Commu nity The m
13. ain function of the CERTA is to support the monitoring analysis and advisory team of NSD in E mail vej sunserv kfki hu NUCLEAR TECHNOLOGY VOL 139 AUG 2002 important part of the CERTA as it provides for the con tinuous remote inspection of the four VVER 440 V213 units of the Hungarian Paks nuclear power plant NPP The on line information system maintains a continuous data link with the NPP through a managed leased line that connects CERTA to a gateway computer located at the Paks NPP The present scope of the system is a result of a 4 yr development project In addition to the basic safety parameter display functions the VITA system now includes an on line break parameter estimation modul an extensive training package based on simulated tran sients and on line data transfer capabilities to feed ac cident diagnosis analysis codes case of a nuclear emergency with appropriate hardware and software tools for communication diagnosis prog nosis and prediction The CERTA includes an on line vital information transfer and analysis VITA system transferring important process data from the Paks nu clear power plant NPP to the NSD s crisis center The first version of the VITA system was installed in 1997 but it has been continuously upgraded in sub sequent years to incorporate new services New func tions were either required by the safety authority or were demanded by changes in plant technology and by modifications
14. ame as for the real units 1 through 4 This feature has been extensively used for system validation and for testing the user interface in various emergency situations Besides verification and validation tests sim ulated data serve as a unique training tool and provide a very good method for conducting general emergency pre paredness exercises at the NSD premises in each year 6 I D Data Transfer to Accident Diagnosis and Analysis Codes The personnel of the NSD utilize several off line accident diagnosis and prognosis codes at the CERTA The SESAME is a best estimate code system original ly developed by Commissariat l Energie Atomique Institute for Protection and Nuclear Safety IPSN France which is able to diagnose a wide range of acci dents in pressurized water reactors The code system can be used for prognosis as well the escalation path of the accident and the possible consequences can be deter mined assuming a variety of possible accident manage ment measures The VVER version of SESAME was developed in the framework of a PHARE project by the IPSN and Hungarian Czech and Slovak institutes The system has an on line data acquisition module which is able to read input data 130 signals cyclically from a relational database In the CERTA the SESAME VVER NUCLEAR TECHNOLOGY VOL 139 AUG 2002 NT 8 01008 7 11 05 31 02 7 32 pm V gh et al runs on a PC and input data are stored in Microsoft Access data
15. base format Upon user s request on line or archived process signals characterizing the selected reactor unit or the simulator are transferred to the SESAME PC from the main VITA computer through network in the appropriate format The transferred data files are handled by a client program on the PC writing the data to the records of the Access database Then the SESAME acquisition module automatically reads in the new values The data transfer can be continuous one packet per minute or single step when archived data are fed to the SESAME for the selected time interval only The SESAME then performs its task according to the user s requirements Source term break size primary coolant mass and hydrogen risk estimation modules can be activated as in the stand alone SESAME A very similar on line and archive data transfer ar chitecture was elaborated for the ADAM program The PC based accident diagnostics analysis and manage ment ADAM is a tool for the analysis of accident con ditions based on measured plant data it was adopted for the Paks VVER 440 V213 units in 2000 The version operated at the NSD can work in two modes 1 On line accident diagnostics and monitoring mode the code calculates the state of the plant and assesses its accident status by using 110 measured process signals 2 Accident management analysis mode The code simulates different accident scenarios to assess the effec tiveness of the various accident
16. canned versions of your figures as FPOs for position only These will be replaced with high resolution versions at the printing stage 4 Necessary corrections should be sent as follows a a list of corrections sent by electronic mail and or b your manuscript with your corrections marked on it sent by fax your fax should include a list of your corrections If you have been sent a list of queries we would greatly appreciate your reply to these queries either by electronic mail or by fax 5 Please return your corrections within a week of receipt of your author proofs Note that delay in transmitting your corrections may delay publication of your paper or may require publication without your corrections 6 Return proofs using the following name and electronic mail address or fax number Betty Drucker E mail bdrucker ans org Fax 708 352 6464 Thank you for your prompt attention to these proofs If you have any questions please feel free to contact me Betty Drucker Staff Editor Leaders in the development dissemination and application of nuclear science and technology to benefit humanity NT 8 01008 1 11 05 31 02 7 32 pm BUILDING UP AN ON LINE PLANT INFORMATION SYSTEM FOR THE EMERGENCY RESPONSE CENTER OF THE HUNGARIAN NUCLEAR SAFETY DIRECTORATE NT 8 01008 NUCLEAR PLANT OPERATIONS AND CONTROL KEYWORDS accident diagno sis safety parameter display emergency preparedness JANOS VEGH CSABA MAJOR CSABA HO
17. e due to density increase and the coeffi cient is calculated according to ef Pz a 05 2 3 where p is the density determined from the primary pressure and average enthalpy and pz is the ratio of the total primary coolant mass and the total primary volume The break estimator module divides the primary cir cuit into two parts The pressurizer itself represents one single node while the rest of the primary circuit is com bined into another node i e this node contains the re actor vessel the cold and hot legs of the six loops and the primary sides of the steam generators Note that in Eq 1 there are no heat source terms i e it is assumed that the heat generated in the core is always compen sated for by the secondary heat removal This is obvi ously not true in certain phases of a LOCA scenario but 8 Page 8 ON LINE PLANT INFORMATION SYSTEM FOR EMERGENCY RESPONSE the results of test runs justified this approximation as providing acceptable results for the overall LOCA pro cess It is also assumed that in the pressurizer the two phase state is always present therefore the module describes correctly the emptying of the pressurizer only and does not handle refill processes In the on line VITA system the break estimator mod ule is called in every cycle for all units including the simulator After an initializing phase when internal vari ables are determined from the actual reactor state the progra
18. edure contains several decision points that cannot be handled precisely by software e g other in dications of imminent or actual core damage therefore we concentrated on those items that can be judged un ambiguously on the basis of measured and calculated process parameters The module works automatically without user intervention it is called in every calcula tion cycle and determines its output values for all units including the simulator The accident classification mod ule deals with the following parts of TECDOC 955 1 Procedure A1 accident classification during op erating standby hot shutdown cold shutdown and re fueling reactor modes see Tables A1 and A2 in Ref 10 The symptom groups that determine the actual accident state of the plant are calculated from simple logic expres sions that use critical safety function states and addi tional measured or calculated process parameters as input Reactor pressure vessel RPV water level plays an im portant role in the classification its value is taken from the break parameter estimator module the units of the Paks NPP have no RPV level measurements for accident conditions Other required parameters time elapsed from scram duration of negative subcooling duration of core uncovered state minimum required flow to compensate the coolant loss due to core decay heat etc are deter mined by the accident classification module 2 Procedure A2 assessment of core dama
19. ervice can be applied for consistent off line analysis of abnormal plant events During such a replay session pre viously recorded measurement snapshots taken from the periodic archive are loaded into the VITA database The contents of a snapshot are processed in the same way as in the on line system Events are generated trends SPD screens are refreshed etc The archive playback can be performed on the VITA computer parallel to the normal operation of the on line system as well Besides the standard signal processing the VITA system performs the following calculations for each of the reactor units 1 evaluation of critical safety function CSF alarm states according to simple logic diagrams called CSF status trees 2 determination of the operational state of the most important safety and auxiliary systems e g high and low pressure emergency core coolant system ECCS injection sprinkler hydroaccumulators and diesels 3 evaluation of logic diagrams to assess the sever ity of an accident state 4 continuous check for the presence of any primary leak and automatic determination of the main break parameters 5 calculation of miscellaneous parameters to be dis played on the SPD screens e g boiling margin loop natural circulation states reactor operation mode margin to the nil ductility temperature of the pressure vessel and change rate of the most important process parameters Since all the calculated signals a
20. ge spent fuel damage is not treated because the VITA system has no instrument readings from the spent fuel pool area Core damage estimation is based on the length of esti mated core uncovery time and on coolant isotope con centrations The installation of a suitable containment radiation monitor is in progress at the plant until then the core damage estimation procedure based on contain ment radiation levels is not treated by the system The severity of the accident state is determined ac cording to TECDOC 955 A normal state plus alert site and general emergency states are distinguished The emer gency state variable drives an alarm which is always visible in the header part of the VITA SPD screen see the icon labeled SES site emergency state in Fig 2 I F Break Parameter Estimator Module As previously mentioned the Paks VVER 440 units as most of the VVER 440 reactors do not have RPV level measurements supplying information about the core covered uncovered state in accident circumstances Therefore one must use reliable estimations to predict anticipated core uncovery time and other important ac cident parameters An analysis of the available VITA 7 NT 8 01008 8 11 05 31 02 7 32 pm V gh et al input signals has shown that the system has sufficient input measurements to feed an appropriate break param eter estimator algorithm The break size estimation algorithm is based on the homogeneous equilibrium mode
21. in the architecture of the plant informa tion system NT 8 01008 2 11 05 31 02 7 32 pm V gh et al Il MAIN FUNCTIONS AND TOOLS OF CERTA CERTA has the following tasks during nuclear emergencies 1 diagnosing the severity of an emergency includ ing source term estimation 2 estimating the consequences and possible escala tion of an accident 3 assessing the effectiveness of the accident man agement and mitigation measures 4 estimating radiological dose and proposing pro tective measures 5 communicating with the NPP and with other na tional crisis centers In addition to emergency handling tasks during nor mal plant operation CERTA is used for the following general purposes 1 on line data collection and archiving from the Paks NPP by use of the VITA system 2 periodic training of nuclear regulatory personnel 3 reception and review of licensee event reports on abnormal plant events 4 evaluation of licensee safety analysis reports on requested changes in plant technology 5 utilization of a severe accident simulator for beyond design basis accident analysis Accomplishment of the preceding tasks is supported by the following dedicated software subsystems or stand alone code systems 1 CERTA VITA an on line plant information system 2 ADAM accident diagnostics analysis and man agement code 3 SESAME a fast accident diagnosis and progno sis code 4 NPA a RELAPS5 based nuclear
22. l It is assumed that in the given volume the single or two phase medium is in a thermodynamic equilibrium state This yields a very fast mixing in case of emergency core cooling injec tions but numerical instabilities are avoided which is very important in an on line system The model distin guishes two basic states of the coolant subcooled liquid and saturated two phase mixture Pressure derivatives according to time are obtained from the energy and mass conservation equations assuming constant volume The AP pressure change in a given volume during a At calculation step is given by AP gt 4 G 1 where G is the flow through the j th junction connected to the given node these values are determined from the pressure differences and the a coefficients have differ ent values for the single and two phase cases The coefficient a is constant for the break and for the pressurizer surge line while for the ECCS injection lines it depends on the parameters of the injected cold water In case a two phase mixture is present in the given node the incoming cold water causes pressure decrease due to condensation therefore the coefficient corre lates with the difference between the ECCS enthalpy and the average primary circuit enthalpy in the following form a dj heccs primary 2 where aj is a fitted constant and h indicates specific enthalpy In single phase cases ECCS injection causes pres sure increas
23. m periodically checks the pressurizer level When ever a significant level change is detected the module tries to identify a possible leak in the primary circuit by using the following method To match the actual situa tion with predefined LOCA and SGTR classes the pro gram performs calculations for the following eight break classes the time step of the calculations is 1 s 1 Classes 1 and 2 small break LOCA 0 1 or 1 0 2 Class 3 medium break LOCA 10 3 Classes 4 and 5 large break LOCA 50 or 200 4 Classes 6 and 7 SGTR a single tube or several tubes 5 Class 8 opening of steam collector cover In the calculations ECCS injection flows are taken from the process measurements The selection of the most characteristic break class is based on the behavior of primary pressure The actual break is ordered into the class that provides the minimum deviation between the measured and calculated primary pressure history curves Those situations when the pressurizer level change is caused by leak free reactor transients e g by a scram or by a pump trip are also identified and handled by the program Once the presence of a break and the appropri ate break class have been identified the module deter mines the following break parameters 1 break type primary or primary to secondary leak 2 break flow equivalent diameter and cross section 3 RPV level determined from the total primary coolant mass
24. management strategies On line data transfer from the main VITA computer to the ADAM PC is applied basically for the accident diagnostics mode Measurement snapshots are copied ei ther cyclically one file per minute or in a large file containing measurements corresponding to a full acci dent scenario these latter data are extracted from the change sensitive archive The input files are prepared by a server program on the AlphaServer of the VITA system then copied to the ADAM PC The ADAM han dles its input files in a flexible manner therefore no client program is required on the PC The code system simply detects that a new input file arrived then auto matically reads in its contents The aforementioned on line data transfer capabili ties of the VITA system are utilized during emergency handling exercises and for training purposes IILE Accident Classification According to IAEA TECDOC 955 The IAEA TECDOC 955 Ref 10 provides simple but systematic procedures applicable for the assessment of a reactor accident The document has been adopted by KFKI AEKI for the conditions of the Paks VVER 440 V213 units in 1998 This created the possibility to pro NUCLEAR TECHNOLOGY VOL 139 AUG 2002 Page 7 ON LINE PLANT INFORMATION SYSTEM FOR EMERGENCY RESPONSE vide a programmed tool within the VITA system to support the experts using these IAEA procedures with on line information Obviously a generic accident as sessment proc
25. oring system Plant safety status overview pictures and detailed plant subsystem pictures are presented by using the Picasso 3 user interface management system gt devel oped by the Organization for Economic Cooperation and Development OECD Halden Reactor Project HRP Picasso 3 P3 runs on two Windows NT workstations actual plant data are transferred from the VITA computer to the graphic workstations by using the SoftwareBus communication system also developed by OECD HRP The SoftwareBus system has been ported to OpenVMS NUCLEAR TECHNOLOGY VOL 139 AUG 2002 for the sake of this task Picasso 3 pictures are used mainly for training purposes when simulated transients are re played at the CERTA It has to be noted that P3 pictures are not available for the real Paks units because the num ber of transferred signals must be limited due to the low capacity of the data communication line III C Connection to the Paks Full Scope Simulator The CERTA gateway is able to transfer data from the Paks full scope training simulator to the CERTA where simulated data can be used for training purposes In autumn 1998 this service was successfully used dur ing the INEX 2 HUN international accident manage ment exercise The accident scenario has been simulated at the Paks NPP and by using the CERTA gateway con figuration the simulated transient could be followed and analyzed on line at the CERTA crisis center by the ex perts of the
26. ork itself is separated from the office network of the NSD by an additional firewall The gateway has been configured in such a manner that it is able to run the entire VITA software without modification it contains exactly the same software as the main VITA computer and it is able to perform the same functions This created the possibility for the NSD inspectors resident at the plant site to have access through remote PC based X terminals to the same information as the NSD experts at the Budapest office I B User Interface Two X terminals standard Windows NT 4 0 work stations with 21 in color displays having 1280 X 1024 pixel resolution represent the primary human machine interface HMI of the VITA system Additional PC based remote X terminals can be connected to the sys tem as well The format of the SPD screens see Fig 2 was constructed in a hierarchical way On the first level concise critical safety function overview pictures are presented together with the most important reactor 3 NT 8 01008 4 11 05 31 02 7 32 pm Page 4 V gh et al ON LINE PLANT INFORMATION SYSTEM FOR EMERGENCY RESPONSE Simulator Simulator WEB COMP VERONA Server Computer Server Server Paks NPP Informatics Network Plant Information Genter Firewall analogue backup line 28 8 kbps digital telephone line Router l Paks NPP Technology Network Legend Data link TCP
27. ors workstations installation of a break size and break flow estimator module and the pilot version of a web based remote CSFM system The development of the VITA continues during 2002 as well Unit 4 will be coupled to the CERTA gateway and the final version of the remote CSFM system will be installed for all Paks units An increase of the managed leased line s capacity to 2 megabytes s is being considered to cope with in creased data transmission demands The higher band width will create the possibility to access further plant data such as radiological and meteorological measure ments Introduction of a containment damage assess ment module is planned for the future as well REFERENCES 1 L VOROSS Installation of Centre for Emergency Re sponse Training and Analysis at the Nuclear Safety Authori 10 Page 10 ON LINE PLANT INFORMATION SYSTEM FOR EMERGENCY RESPONSE ty Hungarian Atomic Energy Authority Nuclear Safety Inspectorate Budapest Hungary Jan 1996 2 M KATHIB RAHBAR ADAM An Accident Diag nostics Analysis and Management System Energy Research Inc Rockville Maryland 1999 3 B RAGUE L JANOT and A JOUZIER Evaluation and Prediction of Changes in a Accident Involving a Pressur ised Water Reactor Method General SEAC 93 295 Insti tute for Protection and Nuclear Safety France May 1993 4 Nuclear Plant Analyzer Manuals Vol 1 4 NUREG CR 6291 INEL 94 0123
28. plant analyzer 5 SCDAP RELAPS severe accident calculations gt 6 CONTAIN calculation of containment behavior 7 MAAP4 VVER severe accident calculations accident management 8 MELCOR severe accident calculations stand alone program version 9 SINAC simulator for environmental transport of radioactive releases 10 InterRAS a source term estimation program 11 a probabilistic safety assessment program for core damage probability estimation in severe accident conditions Page 2 ON LINE PLANT INFORMATION SYSTEM FOR EMERGENCY RESPONSE 12 SUBA a VVER 440 severe accident simula tor based on the MELCOR code Severe accident calculation and simulation codes are running on a high capacity IBM RISC 6000 computer under IBM AIX The SESAME and ADAM accident di agnostics and analysis programs and the risk monitor program run on personal computers PCs The CERTA VITA is hosted by a powerful AlphaServer 2100 com puter with open VMS operating system The listed code systems were installed at the CERTA by the Hungarian VEIKI Institute for Electric Power Research and by KFKI AEKI Il BASIC FUNCTIONS OF THE CERTA VITA SYSTEM The most important basic functions of the VITA sys tem are as follows 1 maintaining an on line data link with the Paks NPP 2 processing plant data secondary validity checks and limit violation checks 3 continuous data archiving 4 comprehensive archive processing ser
29. r MS computer science E tv s Lor nd University Budapest Hungary 1996 is a senior software developer at the Atomic Energy Research Institute of the Hungarian Academy of Sciences working on developing on line process information and data communication systems for nuclear power plants Csaba Horvath BS engineering informatics Kand K lm n Technical High School Budapest Hungary 1997 is a computer engineer at the Atomic Energy Research Institute of the Hungarian Academy of Sciences working on developing graphical user interfaces various intranet applications and relational database management tools Zoltan Hozer MS nuclear engineering Moscow Power Engineering Insti tute Moscow Russia 1984 Dr Univ fluid mechanics Technical University Budapest Hungary 1988 is a senior research scientist at the Atomic Energy Research Institute of the Hungarian Academy of Sciences working on fuel and reactor materials experimental and analytical studies His background includes two phase flow modeling simulator software development fuel behavior and severe accident analysis Ferenc Adorjan MS physics 1973 Dr Univ nuclear science and engi neering E tv s Lor nd University Budapest Hungary 1976 was a senior re search scientist at the Atomic Energy Research Institute of the Hungarian Academy of Sciences He is a senior staff member with the Nuclear Safety Directorate of the Hungarian Atomic Energy Authority HAEA supervising re
30. re stored in the tran sient archive they can be displayed on historical trends for further analysis at any time I A Architecture of the Communication with the NPP The cycle time of data communication between the CERTA and the NPP is 10 s Cyclic communication is performed continuously through a virtual local area net work LAN over a leased 64 kilobytes s digital telephone line according to TCP IP protocol In its initial configu ration the VITA system was connected to the plant cen tral information system which collects only a limited scope of information from the individual process com puters In 1998 the Paks NPP started the gradual replace ment of the old process computers by new state of the art systems see Ref 13 for details This created a new situation in which modern communication and program ming techniques could be utilized to a great extent To cope with the new architecture a CERTA gateway com NUCLEAR TECHNOLOGY VOL 139 AUG 2002 Page 3 ON LINE PLANT INFORMATION SYSTEM FOR EMERGENCY RESPONSE puter was installed at the NPP The gateway has access to the plant technology network which incorporates the unit process computers as well as to the general plant informatics network The communication line connects the AlphaServer 2100 VITA computer of the CERTA to an AlphaServer 1000 gateway computer located at the plant The main task of this computer is to collect data from different plant data sources This
31. search and development R amp D projects performed by technical support organizations His background includes development and application of on line core surveillance systems measuring techniques and evaluation statistics and computer program development Ivan Lux MS physics E tv s L r nd University Budapest Hungary 1969 mathematics 1973 candidate of physical sciences 1981 Dr of Hungarian Acad emy of Sciences 1993 was a senior scientist at the Atomic Energy Research Institute of the Hungarian Academy of Sciences working in the area of transport Monte Carlo methods later on project management in developing on line pro cess information and core surveillance systems for nuclear power plants Among others he supervises the R amp D related activities of the Hungarian nuclear regu latory body Krist f Horvath MS engineering physics Technical University Buda pest Hungary 1997 is a nuclear safety inspector at the Nuclear Safety Direc torate of the HAEA working in the area of nuclear emergency preparedness He is the HAEA host of analytical tools installed at CERTA as well the CERTA VITA system He is also responsible for root cause analysis of events reported by the Paks nuclear power plant NUCLEAR TECHNOLOGY VOL 139 AUG 2002 11
32. vices event localization and listing 5 presenting a safety parameter display SPD plant pictures and p T diagrams 6 presenting on line parameter trends and data trends from the archives 7 archive playback off line analysis of archived plant events by a fully consistent replay 8 maintaining an occasional on line data link with the full scope simulator of the Paks NPP 9 simulator playback replaying transients re corded at the full scope simulator 10 transferring on line or archived data to accident diagnosis and prognosis codes 11 classifying plant emergency states according to IAEA TECDOC 955 Ref 10 12 on line estimating break parameters in case of loss of coolant accident LOCA and steam gen erator tube rupture SGTR 13 presenting the web based version of the Paks NPP on line critical safety functions monitoring system with emergency operating procedure EOP browsing The system is designed for continuous operation thus plant data are continuously archived by the VITA system The time span of the archives is sufficient to cover a whole fuel cycle for each reactor unit Two dif ferent types of archives are maintained by the system NUCLEAR TECHNOLOGY VOL 139 AUG 2002 NT 8 01008 3 11 05 31 02 7 32 pm V gh et al The periodic archive stores only the measured signals while the transient change sensitive archive stores the significant changes for all signals The archive playback s
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