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Highlights of Preliminary Level 1 PSA Results for a Research Reactor

[+] Author Affiliations
Yoon-Hwan Lee, Won-Dea Jung, Su-Ki Park

Korea Atomic Energy Research Institute, Daejon, Korea

Jin-Hong Lee

Chungnam National University, Daejon, Korea

Paper No. ICONE20-POWER2012-55171, pp. 525-529; 5 pages
  • 2012 20th International Conference on Nuclear Engineering and the ASME 2012 Power Conference
  • Volume 2: Plant Systems, Structures, and Components; Safety and Security; Next Generation Systems; Heat Exchangers and Cooling Systems
  • Anaheim, California, USA, July 30–August 3, 2012
  • Conference Sponsors: Nuclear Engineering Division, Power Division
  • ISBN: 978-0-7918-4496-0
  • Copyright © 2012 by ASME


This preliminary PSA (Probabilistic Safety Assessment) 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 reported here is a Level 1 PSA that addresses the risks associated with core damage. It includes an evaluation of the types of accidents that could lead to core damage, and an assessment of their frequencies. After reviewing the documents and the conceptual design, 8 typical initiating events are selected regarding internal events during the normal operation of the reactor.

Simple fault tree models for the PSA are developed instead of the detailed model at this conceptual design stage. The failures of the major components and dependencies between systems have been considered for the fault tree analysis. Normal operating trains were assumed to have a pump, a check valve and a manual valve. The failures of pumps and supporting systems such as the electrical power are modeled, and the failure of the check valve or manual valve is also modeled for the train. Of course, the Common Cause Failure (CCF) and operator error are modeled.

The criterion for inclusion was all sequences with a point estimate frequency greater than a truncation value of 1.0E−13/yr. LOCA-I is the dominant contribution to the total CDF by a single initiating event. The CDF from the internal events of the research reactor is estimated to be 7.38E−07/year. The CDF for the representative initiating events is less than 1.0E−6/year even though conservative assumptions are used in the reliability data. The conceptual design of the research reactor is designed to be sufficiently safe from the viewpoint of safety.

Copyright © 2012 by ASME



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