Development of good practice guidance for quantification of thermal-hydraulic code model input uncertainty

被引:26
作者
Baccou, Jean [1 ]
Zhang, Jinzhao [2 ]
Fillion, Philippe [3 ]
Damblin, Guillaume [3 ]
Petruzzi, Alessandro [4 ]
Mendizabal, Rafael [5 ]
Reventos, Francesc [6 ]
Skorek, Tomasz [7 ]
Couplet, Mathieu [8 ]
Iooss, Bertrand [8 ]
Oh, Deog-Yeon [9 ]
Takeda, Takeshi [10 ]
机构
[1] IRSN, PSN Res SEMIA, Ctr Cadarache, F-13115 St Paul Les Durance, France
[2] Tractebel ENGIE, Blvd Simon Bolivar 34-36, B-1000 Brussels, Belgium
[3] Univ Paris Saclay, CEA, DEN DM2S STMF LMES, F-91191 Gif Sur Yvette, France
[4] NINE Nucl & INd Engn Srl, Via Chiesa 32,759, I-55100 Lucca, Italy
[5] CSN, Pedro Justo Dorado Dellmans 11, Madrid 28040, Spain
[6] UPC, Avda Diagonal 647, Barcelona 08028, Spain
[7] Gesell Anlagen & Reaktorsicherheit GRS GmbH, Forschungszentrum, D-85748 Garchin, Germany
[8] EDF R&D, 6 Quai Watier, F-78401 Chatou, France
[9] Korea Inst Nucl Safety, 62 Gwahak Ro, Daejeon 34142, South Korea
[10] Nucl Regulat Author, Minato Ku, 1-9-9 Roppongi, Tokyo 1068450, Japan
关键词
Good practice guidance; Model input uncertainty quantification; System approach; Thermal hydraulic code; Validation; Inverse quantification of uncertainty; SYSTEM CODES; VALIDATION; METHODOLOGY;
D O I
10.1016/j.nucengdes.2019.110173
中图分类号
TL [原子能技术]; O571 [原子核物理学];
学科分类号
0827 ; 082701 ;
摘要
Taking into account uncertainties is a key issue in nuclear power plant safety analysis using best estimate plus uncertainty methodologies. It involves two main types of treatment depending on the variables of interest input parameters or system response quantity. The OECD/NEA PREMIUM project devoted to the first type of variables has shown that inverse methods for input uncertainty quantification can exhibit strong user-effect. One of the main reasons was the lack of a clear guidance to perform a reliable analysis. This work is precisely devoted to the development of a first good practice guidance document for quantification of thermal-hydraulic code model input uncertainty. The developments have been done in the framework of the OECD/NEA SAPIUM project (January 2017-September 2019). This paper provides a summary of the main project outcome. Recommendations and open issues for future developments are also given.
引用
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页数:13
相关论文
共 57 条
[1]   Thermal-hydraulic phenomena for water cooled nuclear reactors [J].
Aksan, N. ;
D'Auria, F. ;
Glaeser, H. .
NUCLEAR ENGINEERING AND DESIGN, 2018, 330 :166-186
[2]  
Aksan N., 1994, SEPARATE EFFECTS TES, V1
[3]  
American National Standards Institute, 1988, N182 ANSIANS
[4]  
American National Standards Institute, 1973, N182 ANSIANS
[5]  
Annuziato A, 1996, CSNI INTEGRAL TEST F
[6]  
[Anonymous], 2012, Guidance Document JCGM 200
[7]  
[Anonymous], 2005, TRANS ACC AN METH
[8]  
[Anonymous], BEPU C LUCC IT
[9]  
Baccou J, 2017, NURETH 17 C XIAN CHI
[10]  
Cacuci D.G., 2019, BERRU Predictive Modeling: Best Estimate Results with Reduced Uncertainties