Journal of the Korean Society of Safety (한국안전학회지)

The Korean Society of Safety (한국안전학회)
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Development and Validation of MARS-KS Input Model for SBLOCA Using PHWR Test Facility

중수로 실증 실험설비를 이용한 소형냉각재상실사고의 MARS-KS 입력모델 개발 및 검증계산

  • 백경록 (한국원자력안전기술원 규제검증평가실) ;
  • 유선오 (한국원자력안전기술원 규제검증평가실)
  • Received : 2020.11.20
  • Accepted : 2021.03.26
  • Published : 2021.04.30
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Abstract

Multi-dimensional analysis of reactor safety-KINS standard (MARS-KS) is a thermal-hydraulic code to simulate multiple design basis accidents in reactors. The code has been essential to assess nuclear safety, but has mainly focused on light water reactors, which are in the majority in South Korea. Few previous studies considered pressurized heavy water reactor (PHWR) applications. To verify the code applicability for PHWRs, it is necessary to develop MARS-KS input decks under various transient conditions. This study proposes an input model to simulate small-break loss of coolant accidents for PHWRs. The input model includes major equipment and experimental conditions for test B9802. Calculation results for selected variables during steady-state closely follow test data within ±4%. We adopted the Henry-Fauske model to simulate break flow, with coefficients having similar trends to integrated break mass and trip time for the power supply. Transient calculation results for major thermal-hydraulic factors showed good agreement with experimental data, but further study is required to analyze heat transfer and void condensation inside steam generator u-tubes.

Keywords

Acknowledgement

This work was supported by the Nuclear Safety Research Program through the Korea Foundation Of Nuclear Safety(KoFONS) using the financial resource granted by the Nuclear Safety and Security Commission(NSSC) of the Republic of Korea. (No. 1805003)

References

  1. Korea Institute of Nuclear Safety (KINS), "MARS-KS Code Manual Volume I: Theory Manual", KINS/RR-1822 Vol. 1, 2018.
  2. International Atomic Energy Agency (IAEA), "Intercomparison and Validation of Computer Codes for Thermalhydraulics Safety Analysis of Heavy Water Reactor," TECDOC-1395, 2004.
  3. International Atomic Energy Agency (IAEA), "Comparison of Heavy Water Reactor Thermalhydraulic Code Predictions with Small Break LOCA Experimental Data," IAEATECDOC-1688, 2012.
  4. A. Oussoren and A. Delja, "Assessment of Channel Coolant Voiding in RD-14M Test Facility using TRACE," NUREG/IA-0446, 2014.
  5. S. O. Yu, M. K. Cho and K. L. Baek, "Numerical Assessment of LOCA Experiments of RD-14M Using MARS-KS Code," International Journal of Advanced Nuclear Reactor Design and Technology, Vol. 2, pp. 1-9, 2020. https://doi.org/10.1016/j.jandt.2020.04.001
  6. H. T. Kim, B. W. Rhee and J. H. Park, "A Blind Simulation of RD-14M Small-Break LOCA Experiments Using CATHENA Code", Annals of Nuclear Energy, Vol. 38, pp. 389-403, 2011. https://doi.org/10.1016/j.anucene.2010.10.002
  7. H. T. Kim, "An Open Calculation of RD-14M Small-Break LOCA Experiments Using CATHENA Code," Annals of Nuclear Energy, Vol. 46, pp. 63-75, 2012. https://doi.org/10.1016/j.anucene.2012.03.006
  8. A. Dixit, S. K. Yadav, N. Kumar, T. A. Khan, S. Hajela and M. Singhal, "Validation of Computer Code 'ATMIKA' against RD-14M Small Break LOCA Experiments," Nuclear Engineering and Design, Vol. 323, pp. 427-433, 2017. https://doi.org/10.1016/j.nucengdes.2017.07.042
  9. Applied Programming Technology, Inc., "Symbolic Nuclear Analysis Package (SNAP) User's Manual (Version 2.6.3)", 2018.