Assessment study about the risk of Wilks’ formula for uncertainty quantification of design extension condition scenarios in prototype Gen-IV sodium fast reactor

References

• U.S. Nuclear Regulatory Commission. USNRC CFR Title 10, U.S., USNRC. 2011;1. Section 1.11, Protect public health and safety.
   Google Scholar
• U.S. Nuclear Regulatory Commission. USNRC CFR Title 10, U.S., USNRC. 2011. Part 50, Domestic licensing of production and utilization facilities.
   Google Scholar
• U.S. Nuclear Regulatory Commission. USNRC 10 CFR Title 10, U.S., USNRC. 2011. Part 100, Reactor site criteria.
   Google Scholar
• Wilson GE. Historical insights in the development of best estimate plus uncertainty safety analysis. Ann Nucl Energy. 2013;52:2–9.
   Web of Science ®Google Scholar
• Perez M, Reventos F, Batet L, et al. Uncertainty and sensitivity analysis of a LBLOCA in a PWR nuclear powr plant: results of the phase V of the BEMUSE programme. Nucl Eng Des. 2011;241:4206–4222.
   Web of Science ®Google Scholar
• Young SB, Byung-Gil H, Ae-ju C, et al. Experience in KINS on best estimate calculation with uncertainty evaluation, Paris, Francer: Nuclear Energy Agency; 2013. ( Report no. NEA/CSNI/R(2013)8/PART2).
   Google Scholar
• D'Auria F, Bousbia-Salah A, Petruzzi A, et al. State of the art in using best estimate calculation tools in nuclear technology. Nucl Eng Technol. 2006;38:11–31.
   Google Scholar
• Glaeser H. GRS analysis for CSNI uncertainty methods study (UMS), Paris, France: Nuclear Energy Agency; 1998. ( Report no. OECD NEA/CSNI/R(97)35 Vol. 1).
   Google Scholar
• Glaeser H. GRS method for uncertainty and sensitivity evaluation of code results and applications. Sci Technol Nucl Ins. 2008;2008:1–7. Article ID 798901.
   Google Scholar
• Glaeser H, Nuclear Energy Agency. Summary of existing uncertainty methods. OECD/CSNI Workshop on Best Estimate Methods and Uncertainty Evaluations; 17 Dec. 2013; Paris, France.
   Google Scholar
• Ortiz MG, Ghan LS. Uncertainty analysis of minimum vessel liquid inventory during a small-break LOCA in a B&W plant – an application of the CSAU methodology using the RELAP5/MOD3 computer code, Idaho Falls, U.S.: Idaho National Engineering Laboratory; 1992. ( Report no. NUREG/CR-5818, EGG-2665).
   Google Scholar
• Boyack BE. Quantifying reactor safety margins part 1: an overview of the code scaling, applicability, and uncertainty evaluation methodology. Nucl Eng Des. 1990;119:1–15.
   Web of Science ®Google Scholar
• Kwang-Won S, Dong Gu K, Byung Gil H, et al. Audit calculations of LBLOCA for Ulchin unit 1,2 power uprate, Daejeon, Republic of Korea: Korea Institute of Nuclear Safety; 2013. ( Report no. KINS/RR-1043).
   Google Scholar
• In SH, Deog YO, In GK. Generic application of Wilks’ tolerance limit evaluation approach to nuclear safety, Paris, France: Nuclear Energy Agency; 2013. ( Report no. NEA/CSNI/R(2013)8/PART2).
   Google Scholar
• Wilks SS. Determination of sample sizes for setting tolerance limits. Ann Math Stat. 1941;12:91–96.
   Google Scholar
• Wilks SS. Statistical prediction with special reference to the problem of tolerance limits. Ann Math Stat. 1942;13:400–409.
   Google Scholar
• Seung WL, Bub DC, Young-Seok B, et al. Analysis of uncertainty quantification method by comparing Monte-Carlo method and Wilks’ formula. Nucl Eng Technol. 2014;46:481–488.
   Web of Science ®Google Scholar
• Horie H, Miyagi K, Nakahara K, et al. Safety performance of the 4S reactor on the ATWS events – statistical estimation of uncertainty. Prog Nucl Energy. 2008;50:179–184.
   Web of Science ®Google Scholar
• Chiwoong C, Kwiseok H, Hae-Yong J. Analyses of UTOP events for the design of control rod stop system in PGSFR using MARS-LMR. Ann Nucl Energy. 2016;96:422–431.
   Web of Science ®Google Scholar
• Bub-Dong C, Jae-Jun J, Moon-Kyu H, Won Jae L, Young Jin L, Seung Wook L, Kyung-Doo K, Sung-won B. MARS code manual volume IV: Developmental Assessment Report. Daejeon, Republic of Korea: Korea Atomic Energy Research Institute; 2009. ( Report no. KAERI/TR-3042/2005).
   Google Scholar
• Kwi-Seok H, Hae-Young J, Young-Min K, et al. Liquid metal properties using a soft-sphere model. Proceeding of Korean Nuclear Society; 2006 May 25–26; Chuncheon(Republic of Korea).
   Google Scholar
• Seok JK, Jae SS, Doo HK. PIRT support and best-estimate MARS-LMR input development for SFR, Daejeon, Republic of Korea: Korea Atomic Energy Research Institute; 2016. ( Report no. KAERI/TR-6468/2016).
   Google Scholar
• Jaeseok H, Sarah K, Sung WB, et al. Uncertainty and sensitivity analysis for PGSFR under DEC scenario, Daejeon, Republic of Korea: Korea Atomic Energy Research Institute; 2016. ( Report no. KAERI/TR-6467/2016).
   Google Scholar
• Hoon S, Young-In K, Sang-Ji K, et al. Analysis of BFS-73-1 experiment, Daejeon, Republic of Korea: Korea Atomic Energy Research Institute; 1998. ( Report no. KAERI/TR-1133/98).
   Google Scholar
• Kazimi MS, Carelli MD. Heat transfer correlation for analysis of CRBRP assemblies. Westinghouse Electric Corporation; 1976. ( Report no. CRBRP-ARD-0034).
   Google Scholar
• Aoki S. Current liquid-metal heat transfer research in Japan. Prog Heat Mass Transfer. 1973;7:569–587.
   Google Scholar
• Graber H, Rieger M. Experimental study of heat transfer to liquid metals flowing in-line through tube bundles. Prog Heat Mass Transfer. 1973;7:151–166.
   Google Scholar
• Siefken LJ, Coryell EW, Harvego EA, et al. SCDAP/RELAP5/MOD3.3 code manual, Idaho Falls, U.S.: Idaho National Engineering and Environmental Laboratory; 2001. ( Report no. NUREG/CR-6150, Vol. 4, Rev. 2).
   Google Scholar
• Cheng SK, Todreas NE. Hydrodynamic models and correlations for bare and wire-wrapped hexagonal rod bundles – bundle friction factors, subchannel friction factors and mixing parameters. Nucl Eng Des. 1986;92:227–251.
   Web of Science ®Google Scholar
• Chen SK, Todreas NE, Nguyen NT. Evaluation of existing correlations for the prediction of pressure drop in wire-wrapped hexagonal array pin bundles. Nucl Eng Des. 2014;267:109–131.
   Web of Science ®Google Scholar
• Glaeser H, Hofer E, Kloos M, et al. Report of the uncertainty methods study for advanced best estimate thermal hydraulic code applications, Paris, France: Nuclear Energy Agency; 1998, (Report no. OECD NEA/CSNI/R(97)35 Vol. 2).
   Google Scholar
• Jaeseok H, Kyung DK. PAPIRUS, a parallel computing framework for sensitivity analysis, uncertainty propagation, and estimation of parameter distribution. Nucl Eng Des. 2015;292:237–247.
   Web of Science ®Google Scholar