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An Improvement of Estimation Method of Source Term to the Environment for Interfacing System LOCA for Typical PWR Using MELCOR code

  • Han, Seok-Jung (Risk and Environmental Safety Research Division, Korea Atomic Energy Research Institute) ;
  • Kim, Tae-Woon (Risk and Environmental Safety Research Division, Korea Atomic Energy Research Institute) ;
  • Ahn, Kwang-Il (Risk and Environmental Safety Research Division, Korea Atomic Energy Research Institute)
  • Received : 2015.07.17
  • Accepted : 2017.04.04
  • Published : 2017.06.30

Abstract

Background: Interfacing-system loss-of-coolant-accident (ISLOCA) has been identified as the most hazardous accident scenario in the typical PWR plants. The present study as an effort to improve the knowledge of the source term to the environment during ISLOCA focuses on an improvement of the estimation method. Materials and Methods: The improvement was performed to take into account an effect of broken pipeline and auxiliary building structures relevant to ISLOCA. An estimation of the source term to the environment was for the OPR-1000 plants by MELOCR code version 1.8.6. Results and Discussion: The key features of the source term showed that the massive amount of fission products departed from the beginning of core degradation to the vessel breach. Conclusion: The release amount of fission products may be affected by the broken pipeline and the auxiliary building structure associated with release pathway.

Acknowledgement

Supported by : National Research Foundation

References

  1. Galyean WJ, Kelly DL, Schroeder JA, Auflick LJ, Blackman HS, Gertman DI, Hanley LN. Intersystem LOCA risk assessment: methodology and results. Nucl. Eng. Des. 1994;152(1-3):159-174. https://doi.org/10.1016/0029-5493(94)90082-5
  2. U.S. Nuclear Regulatory Commission. Interfacing systems LOCA: Pressurized water reactors. NUREG/CR-5102. 1989;1.
  3. Lee M, Ko YC. Quantification of severe accident source terms of a Westinghouse 3-loop plant. Nucl. Eng. Des. 2008; 238(4):1080-1092. https://doi.org/10.1016/j.nucengdes.2007.09.003
  4. Korea Power Engineering Corporation. AOT/STI relaxation study for Korean standard nuclear power plants. The Ministry of Industry and Resources. 2004;15-156.
  5. U.S. Nuclear Regulatory Commission. State-of-the-art reactor consequence analyses (SOARCA) report. NUREG-1935. 2012;73.
  6. Kim DS, Kim HC, Sung KY. Comparison of MELCOR and maap calculation results in evaluating risk of fearly fatality for OPR-1000 plant. The Korean Nuclear Society Autumn Meeting. Gyengju Korea. October 29-30, 2009.
  7. U.S. Nuclear Regulatory Commission. MELCOR computer code manuals. NUREG/CR-6119. 2005;3-6.

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  1. A Conceptual Approach to Eliminate Bypass Release of Fission Products by In-Containment Relief Valve under SGTR Accident vol.2018, pp.1687-6083, 2018, https://doi.org/10.1155/2018/5936214