Study on EOPs for PORV-Break LOCA in Generic PWR Simulator

References

• J. PARK et al., “The Step Complexity Measure for Emergency Operating Procedures—Comparing with Simulation Data,” Reliab. Eng. Syst. Saf., 74, 1, 63 (2001); https://doi.org/10.1016/S0951-8320(01)00063-1.
   Web of Science ®Google Scholar
• “TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations,” NUREG-0578, U.S. Nuclear Regulatory Commission (1979); https://doi.org/10.2172/6116869.
   Google Scholar
• “NRC Action Plan Developed as a Result of the TMI-2 Accident,” NUREG-0660-V-1, U.S. Nuclear Regulatory Commission (1980).
   Google Scholar
• G. S. GALLETTI and A. B. SUTTHOFF, “Lessons Learned from the Special Inspection Program for Emergency Operating Procedures,” NUREG-1358, U.S. Nuclear Regulatory Commission (1992).
   Google Scholar
• “Clarification of TMI Action Plan Requirements,” NUREG-0737, U.S. Nuclear Regulatory Commission (1980).
   Google Scholar
• T. J. LIU, C. H. LEE, and C. Y. CHANG, “Power-Operated Relief Valve Stuck-Open Accident and Recovery Scenarios in the Institute of Nuclear Energy Research Integral System Test Facility,” Nucl. Eng. Des., 186, 1–2, 149 (1998); https://doi.org/10.1016/S0029-5493(98)00221-0.
   Web of Science ®Google Scholar
• J. LIM, “RELAP5 Analysis of Hydrodynamic Loads in Response to PORV Actuation,” Ann. Nucl. Energy, 115, 459 (2018); https://doi.org/10.1016/j.anucene.2018.02.023.
   Web of Science ®Google Scholar
• B. S. YOUN et al., “Analysis the PORV Accident for the Kori# 2 Using SPACE Code,” Transactions of the Korean Nuclear Society Spring Meeting, Taebaek, Korea, May 26–27, 2011.
   Google Scholar
• C. J. HSU, K. PERKINS, and R. YOUNGBLOOD, “Estimation of Risk Reduction from Improved PORV (Power Operated Relief Values) Reliability in PWRs,” NUREG/CR-4999; BNL-NUREG-52101. U.S. Nuclear Regulatory Commission, Division of Reactor Accident Analysis; Brookhaven National Laboratory (1988).
   Google Scholar
• J. KIM, D. S. KIM, and S. H. KIM, “Development of Basic Regulatory Framework for APR1400 Level 1 PSA Model,” KAERI/RR–4383/2018, Korea Atomic Energy Research Institute (2018) (in Korean).
   Google Scholar
• M. ZUBAIR, A. ABABNEH, and A. ISHAG, “Station Black Out Concurrent with PORV Failure Using a Generic Pressurized Water Reactor Simulator,” Ann. Nucl. Energy, 110, 1081 (2017); https://doi.org/10.1016/j.anucene.2017.08.023.
   Google Scholar
• D. W. JONES, D. K. BAER, and C. C. FRANCIS, “Nuclear Power Plant Simulators: Their Use in Operator Training and Requalification,” NUREG/CR-1482, ORNL/NUREG/TM-395, Oak Ridge National Laboratory (1980).
   Google Scholar
• W. JUNG, J. PARK, and J. KIM, “A Simulator Study and Its Application to an HRA in NPPs,” Proc. IEEE 8th Human Factors and Power Plants and HPRCT 13th Annual Mtg., Monterey, California, August 26–31, 2007, p. 205, IEEE (2007).
   Google Scholar
• R. L. BORING, “Using Nuclear Power Plant Training Simulators for Operator Performance and Human Reliability Research,” Proc. 6th Int. Topl. Mtg. Nuclear Plant Instrumentation, Control, and Human-Machine Interface Technologies, Knoxville, Tennessee, April 5–9, 2009, American Nucler Society (2009).
   Google Scholar
• S. M. SLOAN, R. R. SCHULTZ, and G. E. WILSON, “RELAP5/MOD3 Code Manual,” NUREG/CR-5535, U.S. Nuclear Regulatory Commission (1992).
   Google Scholar
• J. J. JEONG et al., “Development of a Multi-Dimensional Thermal-Hydraulic System Code, MARS 1.3.1,” Ann. Nucl. Energy, 26, 18, 1611 (1999); https://doi.org/10.1016/S0306-4549(99)00039-0.
   Web of Science ®Google Scholar
• R. M. SUMMERS et al., “MELCOR 1.8.0: A Computer Code for Nuclear Reactor Severe Accident Source Term and Risk Assessment Analyses,”Sandia National Laboratories (1991).
   Google Scholar
• FAUSKE AND ASSOCIATES, “MAAP4—Modular Accident Analysis Program for LWR Power Plants,” in “Code Structure and Theory,” Vol. 2, Part 1, Electric Power Research Institute (1994).
   Google Scholar
• D. R. DION, “The Use of Engineering Computer Codes in the Validation of a Nuclear Power Plant Training Simulator,” Trans. Am. Nucl. Soc., 54 (1987).
   Google Scholar
• “Guidelines for the Preparation of Emergency Operating Procedures,” NUREG-0899, Rev. 1, U.S. Nuclear Regulatory Commission (1982).
   Google Scholar
• “Generic PWR Simulator—Training Guide,” Western Services Corporation (2015).
   Google Scholar
• R. M. AL-CHALABI et al., “NESTLE: A Nodal Kinetics Code,” Trans. Am. Nucl. Soc., 68, 432 (1993).
   Google Scholar
• S. S. LEE, S. H. KIM, and K. Y. SUH, “The Design Features of the Advanced Power Reactor 1400,” Nucl. Eng. Technol., 41, 8, 995 (2009); https://doi.org/10.5516/NET.2009.41.8.995.
   Web of Science ®Google Scholar
• ANSI/ANS-3.5-2009, Nuclear Power Plant Simulators for Use in Operator Training and Examination, American Nuclear Society, La Grange Park, Illinois (2009).
   Google Scholar
• Y. KIM et al., “An HRA Method for Digital Main Control Rooms―Part I: Estimating the Failure Probability of Timely Performance,” KAERI/TR-7607/2019, Korea Atomic Energy Research Institute (2019).
   Google Scholar
• “Evaluation of Potential ERG and EPG Changes to Address NRC Bulletin 2003-01 Recommendations (PA-SEE-0085), Volume 1 – Engineering Evaluations and Analyses Report,” WCAP-16204, Rev. 1, Westinghouse Electric Company (2004).
   Google Scholar
• “APR 1400 Design Control Document Tier 2 Chapter 13: Conduct of Operations,” APR1400-K-X-FS-14002-NP, Rev. 3, Korea Electric Power Corporation – Korea Hydro & Nuclear Power Company (2018).
   Google Scholar
• J. KIM, W. JUNG, and J. PARK, “AGAPE-ET: Advanced Guidelines for Human Reliability Analysis of Emergency Tasks,” Proc. IEEE 8th Human Factors and Power Plants and HPRCT 13th Annual Mtg., Monterey, California, August 26–31, 2007, IEEE (2007).
   Google Scholar
• K. KAWANISHI et al., “Experimental Study on PORV Break LOCA in PWR Plants,” J. Nucl. Sci. Technol., 27, 2, 133 (1990); https://doi.org/10.1080/18811248.1990.9731162.
   Google Scholar
• K. KAWANISHI et al., “Experimental Study on Heat Removal During Cold Leg Small Break LOCAs in PWRs,” J. Nucl. Sci. Technol., 28, 6, 555 (1991); https://doi.org/10.1080/18811248.1991.9731395.
   Web of Science ®Google Scholar
• “APR 1400 Design Control Document Tier 2 Chapter 19: Probabilistic Risk Assessment and Severe Accident Evaluation,” APR1400-K-X-FS-14002-NP, Rev. 3, Korea Electric Power Corporation – Korea Hydro & Nuclear Power Company (2018).
   Google Scholar