DEVELOPMENT OF RELAP5-3D MODEL FOR VVER-440 REACTOR

1. DEVELOPMENTOF RELAP5-3D MODELFOR VVER-440 REACTOR Marek Benčík, Jan Hádek 2010 RELAP5 International User’s Seminar West Yellowstone, MontanaSeptember 20-23, 2010

2. CONTENTS • Introduction of Safety Analyses Department • VVER-440 description • Development of 3D model for VVER-440/213 • Model validation • Conclusion

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. 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. 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. 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. DEVELOPMENT OF MODEL 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.

8. Fig. 1 : Nodalization of reactor coolant system Fig. 1 : Nodalization of reactor coolant system

9. Fig . 2: Main steam system nodalization

10. Fig . 3: Feed water system nodalization

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

12. S S n S S Neutronic Library: D, , , , a f f s KASSETA In kinetics node HELIOS Subroutine userxs calling subroutines: ã Program RELAP5 - 3D DYLIE Core model param rods record ã RELAP5 - 3D Library: r CSLIBR T , , T , c m m f b f idrcrd, crdfr, valusr, CSLIBR time, dt, mode Fig. 6: Neutronic data preparation

13. 349 fuel assemblies 31 TH channels 6 reflector channels Fig. 7: Core model

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. Fig. 8: Malfunction of feed water system, HZP

16. 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. 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. Fig. 9 : Temperatures in core channels 2 and corresponding fuel assemblies

19. Fig. 10 : Temperatures in core channels 6 and corresponding fuel assemblies

20. Fig. 11 : Temperatures in core channels 13 and corresponding fuel assemblies

21. Fig. 12 : Temperatures in hot legs

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. 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.