1. Nuclear Research Institute Řež plc 1 DEVELOPMENT OF RELAP5-3D MODEL FOR VVER-440 REACTOR 2010 RELAP5 International User’s Seminar West Yellowstone, Montana September 20-23, 2010 Marek Benčík, Jan Hádek

2. Nuclear Research Institute Řež plc 2010 RELAP5 International User’s Seminar 2 CONTENTS Introduction of Safety Analyses Department VVER-440 description Development of 3D model for VVER-440/213 Model validation Conclusion

3. Nuclear Research Institute Řež plc 2010 RELAP5 International User’s Seminar 3 1. NRI INTRODUCTION Main activities: –Thermal hydraulic and neutron kinetics calculations –Development of advanced analytical methods –Expert missions NPP Type –VVER 440/213 Dukovany –VVER 1000 Temelín –PWR

4. Nuclear Research Institute Řež plc 2010 RELAP5 International User’s Seminar 4 Thermal hydraulic and neutron kinetics calculations for: –Safety Analysis Report (SAR) –Pressurized Thermal Shock (PTS) analyses –Equipment qualification –Computer codes validation (e.g. under umbrella of OECD) –Verification of Emergency Operation Procedures –Accidents at low power and shutdown –Probabilistic Safety Assessment

5. Nuclear Research Institute Řež plc 2010 RELAP5 International User’s Seminar 5 Development of advanced analytical methods –Best estimate analyses with considered uncertainty of BE computer codes models, correlations and input data –Prediction of CHF use of CFD codes. New model of CHF calculation base on microstructure of process. –Two phase CFD computer codes. Expert missions include: –Support to the Czech nuclear power plants –Support to the regulatory bodies (Czech, Ukraine) –Consideration of new nuclear facilities –Participation in the development of advanced nuclear power plants (for example Generation IV)

6. Nuclear Research Institute Řež plc 2010 RELAP5 International User’s Seminar 6 2.VVER-440 (NPP DUKOVANY) VVER-440 description: Thermal power: 1444 MW Primary pressure: 12,36 MPa Number of fuel assemblies: 349 Loops: 6 Horizontal steam generators

7. Nuclear Research Institute Řež plc 2010 RELAP5 International User’s Seminar 7 The RELAP5/MOD3 input model for VVER-440 created in NRI Řež (P.Král, L. Krhounková) was used as a base for the RELAP5-3D input deck. Main characteristics of input model are following: The reactor vessel is described by two 3D multid object and bundle of 1D channels (core). All 6 loops of reactor coolant system are fully modeled, as well as the pressurizer system, ECCS system, main steam system (MSS) and feed water (FW) system lines, all relevant heat structures, control and protection systems. 3.DEVELOPMENT OF MODEL

8. Nuclear Research Institute Řež plc 2010 RELAP5 International User’s Seminar 8 Fig. 1 : Nodalization of reactor coolant system

9. Nuclear Research Institute Řež plc 2010 RELAP5 International User’s Seminar 9 Fig. 2: Main steam system nodalization

10. Nuclear Research Institute Řež plc 2010 RELAP5 International User’s Seminar 10 Fig. 3: Feed water system nodalization

11. Nuclear Research Institute Řež plc 2010 RELAP5 International User’s Seminar 11 Fig. 4: Reactor pressure vessel nodalization Fig. 5: Fuel assembly 125 288 258 228 UP 088 CL 041 CORE 438 418 HA 188 158 077 HL 408 428 HA DC + LP 035 100 130 160 200 230 260 7 8 9 3 14 15 16 2 3 4 5 6 7 4 8 9 10 1 1 2 6

12. Nuclear Research Institute Řež plc 2010 RELAP5 International User’s Seminar 12 Fig. 6: Neutronic data preparation

13. Nuclear Research Institute Řež plc 2010 RELAP5 International User’s Seminar 13 Fig. 7: Core model 349 fuel assemblies 31 TH channels 6 reflector channels

14. Nuclear Research Institute Řež plc 2010 RELAP5 International User’s Seminar 14 4. ANALYSED SCENARIOS RELAP5-3D with 3D TH and NK is particularly suitable for analyses of cases with non uniform power generation in core. For the time being the following cases were calculated for VVER 440: Steam line break (full power, HZP) Malfunction of feed water system (HZP) Boron dilution (HZP) AER 6 benchmark  Correct prediction of mixing in the pressure vessel is essential

15. Nuclear Research Institute Řež plc 2010 RELAP5 International User’s Seminar 15 Fig. 8: Malfunction of feed water system, HZP

16. Nuclear Research Institute Řež plc 2010 RELAP5 International User’s Seminar 16 5.MODEL VALIDATION Mixing of coolant in reactor during asymmetrical cool down Initial conditions: Reactor operated on HZP 6 MCP in operation Average primary temperature 260°C Secondary pressure 4.56 MPa

17. Nuclear Research Institute Řež plc 2010 RELAP5 International User’s Seminar 17 Scenario description: The selected cold leg is cooled down (~5°C) by steam reduction station Temperatures in core, cold and hot legs are measured Test is repeated for all the loops RELAP5-3D model: Only pressure vessel is modeled

18. Nuclear Research Institute Řež plc 2010 RELAP5 International User’s Seminar 18 Fig. 9 : Temperatures in core channels 2 and corresponding fuel assemblies

19. Nuclear Research Institute Řež plc 2010 RELAP5 International User’s Seminar 19 Fig. 10 : Temperatures in core channels 6 and corresponding fuel assemblies

20. Nuclear Research Institute Řež plc 2010 RELAP5 International User’s Seminar 20 Fig. 11 : Temperatures in core channels 13 and corresponding fuel assemblies

21. Nuclear Research Institute Řež plc 2010 RELAP5 International User’s Seminar 21 Fig. 12 : Temperatures in hot legs

22. Nuclear Research Institute Řež plc 2010 RELAP5 International User’s Seminar 22 6. CONCLUSIONS AND FUTURE WORK 3D TH and NK model of VVER 440/213 has been developed and improved during last decade in NRI Řež The paper presents our last validation effort focused on mixing in reactor vessel. The results of 3D calculation are in good agreement with measured data The original complex model of VVER 440 is too complicated, unsuitable for sensitivity and uncertainty studies Optimalization and simplification are needed

23. Nuclear Research Institute Řež plc 2010 RELAP5 International User’s Seminar 23 REFERENCES Král P.: Assessment of RELAP5 and Verification of Modelling Methods for VVER-Type Reactor Analysis. Paper for the RELAP5 International Users Seminar. Boston. July 1993. Macek J., Muhlbauer P., Krhounková J., Král P., Malačka M.: Thermal Hydraulic Analyses of NPPs with VVER-440/213 for the PTS Condition Evaluation. NURETH-8. 1997. Hádek J., Král P.: Final Results of the Sixth Three-Dimensional AER Dynamic Benchmark Problem Calculation. Solution of Problem with DYN3D and RELAP5- 3D Codes. 13th Symposium of AER on VVER Reactor Physics and Reactor Safety, Dresden, September 2003.