Journal of the Korean Society of Safety (한국안전학회지)

The Korean Society of Safety (한국안전학회)
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A Study on the Final Probabilistic Safety Assessment for the Jordan Research and Training Reactor

JRTR 연구용원자로에 대한 최종 확률론적 안전성평가

  • Received : 2019.10.04
  • Accepted : 2020.01.03
  • Published : 2020.06.30
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Abstract

This paper describes the work and the results of the final Probabilistic Safety Assessment (PSA) for the Jordan Research and Training Reactor (JRTR). This final PSA was undertaken to assess the level of safety for the design of a research reactor and to evaluate whether it is probabilistically safe to operate and reliable to use. The scope of the PSA described here is a Level 1 PSA, which addresses the risks associated with core damage. After reviewing the documents and its conceptual design, nine typical initiating events were selected regarding internal events during the normal operation of the reactor. AIMS-PSA (Version 1.2c) was used for the accident quantification, and FTREX was used as the quantification engine. 1.0E-15/yr of the cutoff value was used to deliminate the non-effective Minimal Cut Sets (MCSs) when quantifying the JRTR PSA model. As a result, the final result indicates a point estimate of 2.02E-07/yr for the overall Core Damage Frequency (CDF) attributable to internal initiating events in the core damage state for the JRTR. A Loss of Primary Cooling System Flow (LOPCS) is the dominant contributor to the total CDF by a single initiating event (9.96E-08/yr), and provides 49.4% of the CDF. General Transients (GTRNs) are the second largest contributor, and provide 32.9% (6.65E-08/yr) of the CDF.

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References

  1. IAEA, Safety Series N0 35-G1.
  2. IAEA, Procedures for conducting Probabilistic Safety Assessments of Nuclear power Plants (Level 1), IAEA Safety Series No. 50-P-4.
  3. USNRC, PRA Procedures Guide, NUREG/CR-2300, 1982.
  4. Jordan Research and Training Reactor Final Safety Analysis Report, KAERI, 2016.
  5. S. H. Han, H. G. Lim and J. E. Yang, AIMS-PSA: A Software for Integrating Various Types of PSA, PSAM9, 2008.
  6. W. S. Jung et. al., "A Fast BDD Algorithm for Large Coherent Fault Trees Analysis", Journal of Reliability Engineering and System Safety, Vol. 83, pp. 369-374, 2007.
  7. IAEA, Generic Component Reliability Data for Research Reactor PSA, IAEA-TECDOC-930, 1997.
  8. ALWR PRA Key Assumptions and Ground Rules, Rev.7, EPRI, December 1997.
  9. KOREA Multipurpose Research Reactor Technical Report, KM-033-RT-K066, KAERI, 1994.
  10. Guidelines for System Modeling: Common Cause Failures (Rev.1), KAERI/TR-2916/2005, KAERI, 2005.
  11. U.S. NRC, Accident Sequence Evaluation Program Human Reliability Analysis (ASEP HRA) Procedure, NUREG/CR-4772, USNRC, USA, 1987.
  12. Y. H. Lee, "Aspects of Preliminary Probabilistic Safety Assessment for a Research Reactor in the Conceptual Design Phase", J. Korean Soc. Saf., Vol. 34, No. 3, 2019.
  13. Y. H. Lee, "Design Improvement to a Research Reactor for Safety Enhancement Using PSA", J. Korean Soc. Saf., Vol. 33, No. 5, 2018.