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CSPACE for a simulation of core damage progression during severe accidents

  • Received : 2020.12.29
  • Accepted : 2021.06.27
  • Published : 2021.12.25

Abstract

CSPACE (Core meltdown, Safety and Performance Analysis CodE for nuclear power plants) for a simulation of severe accident progression in a Pressurized Water Reactor (PWR) is developed by coupling of verified system thermal hydraulic code of SPACE (Safety and Performance Analysis CodE for nuclear power plants) and core damage progression code of COMPASS (Core Meltdown Progression Accident Simulation Software). SPACE is responsible for the description of fluid state in nuclear system nodes, while COMPASS is responsible for the prediction of thermal and mechanical responses of core fuels and reactor vessel heat structures. New heat transfer models to each phase of the fluid, flow blockage, corium behavior in the lower head are added to COMPASS. Then, an interface module for the data transfer between two codes was developed to enable coupling. An implicit coupling scheme of wall heat transfer was applied to prevent fluid temperature oscillation. To validate the performance of newly developed code CSPACE, we analyzed typical severe accident scenarios for OPR1000 (Optimized Power Reactor 1000), which were initiated from large break loss of coolant accident, small break loss of coolant accident, and station black out accident. The results including thermal hydraulic behavior of RCS, core damage progression, hydrogen generation, corium behavior in the lower head, reactor vessel failure were reasonable and consistent. We demonstrate that CSPACE provides a good platform for the prediction of severe accident progression by detailed review of analysis results and a qualitative comparison with the results of previous MELCOR analysis.

Keywords

Acknowledgement

This work was supported by the Korea Institute of Energy Technology Evaluation and Planning (KETEP) grant funded by the Korean government (Ministry of Trade, Industry and Energy) (No.KETEP-20193110100050).

References

  1. Michael Z. Podowski, Raf M. Podowski, Ha Kim Dong, Jun Ho Bae, Gun Son Dong, Compass - new modeling and simulation approach to PWR invessel accident progression, Nuclear Engineering and Technology 51 (2019) 1916-1938. https://doi.org/10.1016/j.net.2019.06.008
  2. J.H. Bae, et al., Core degradation simulation of the PHEBUS FPT3 experiment using COMPASS code, Nucl. Eng. Des. 320 (2017) 258-268. https://doi.org/10.1016/j.nucengdes.2017.05.030
  3. S.J. Ha, Development of the SPACE code for nuclear power plants, Nucl. Eng. Technol. 43 (5) (2011) 44-62.
  4. Dongkyu Lee, HeeCheon No, Bokyung Kim, Investigation of a best oxidation model and thermal margin analysis at high temperature under design extension conditions using SPACE, Nuclear Engineering and Technology 52 (Issue 4) (April 2020) 742-775. https://doi.org/10.1016/j.net.2019.09.011
  5. Yongjae Lee, Wonjun Choi, Joong Kim Sung, Efficacy assessment of independent severe accident mitigation measures in OPR1000 using MELCOR code, J. Nucl. Sci. Technol. 54 (1) (2017) 89-100. https://doi.org/10.1080/00223131.2016.1213670
  6. Wonjun Choi, Taeseok Kim, Joongoo Jeon, Kyung Kim Nam, Joong Kim Sung, Effect of molten corium behavior uncertainty on the severe accident progress, Science and Technology of Nuclear Installations (2018) 9, https://doi.org/10.1155/2018/5706409. Article ID 5706409.
  7. Kwang Soon Ha, Il Kim Sung, Hyung Seok Kang, Ha Kim Dong, SIRIUS: a code on fission product behavior under severe accident, Transactions of the Korean Nuclear Society Spring Meeting Jeju, Korea, May 18-19, 2017.
  8. SAND2018-13560 O, MELCOR computer code manuals, in: Reference Manual Version 2.2, vol. 2, November 2018, p. 11932.
  9. I.L. Pioro, et al., Nucleate pool-boiling heat transfer. II: assessment of prediction methods/, Int. J. Heat Mass Tran. 47 (2004) 5045-5057. https://doi.org/10.1016/j.ijheatmasstransfer.2004.06.020
  10. NUREG/CR-5535, INEL-95/0174, RELAP5/MOD3 Code Manual - Models and Correlations, vol. 4, Idaho National Engineering Laboratory, 1995 August.
  11. J.V. Cathcart, R.E. Pawel, R.A. McKee, R.E. Druscel, G.J. Yurek, J.J. Cambell, S.H. Jury, Zirconium metal-water oxidation kinetics IV. Reaction rate studies 17, ORNL/NUREG-, Aug. 1977.
  12. L. Baker, L.C. Just, Studies of metal-water reactions at high temperatures; III. Experimental and theoretical studies of the zirconium-water reaction, ANL-6548 (May 1962).
  13. Dong-gun Son, et al., Software Design Description for Severe In-Vessel Melt Progression in Lower Plenum Environment (SIMPLE) Program, KAERI/TR-6816/2017, 3. 17, 2017.
  14. IAEA, The Fukushima Daiichi Accident, vols. 1 - 4, 2015.
  15. T.H. Vo, D.H. Kim, J.H. Song, An analysis of radiological releases during a station black out accident for the APR1400, Nucl. Eng. Des. 332 (2018) 22-30. https://doi.org/10.1016/j.nucengdes.2018.03.016