Sensitivity and uncertainty analysis for ULOF of PGSFR using PAPIRUS

被引:3
作者
Kang, Sarah [1 ]
Choi, ChiWoong [2 ]
Ha, Kwi-Seok [2 ]
Heo, Jaeseok [1 ]
机构
[1] Thermal Hydraul Safety Res Div, Daedeok Daero 989 Beon Gil, Daejeon, South Korea
[2] Korea Atom Energy Res Inst, Sodium Cooled Fast Reactor Design Div, 111 Daedeok Daero 989 Beon gil, Daejeon, South Korea
基金
新加坡国家研究基金会;
关键词
PGSFR; MARS-LMR; PAPIRUS; Sensitivity analysis; Uncertainty propagation; ULOF; SAFETY ANALYSIS; CODE; BUNDLES;
D O I
10.1016/j.anucene.2017.08.041
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
In this research, sensitivity and uncertainty analyses for 23 parameters were performed for unprotected loss of flow (ULOF) for the prototype Gen-IV sodium-cooled fast reactor (PGSFR) by using the parallel computing platform integrated for uncertainty and sensitivity analysis (PAPIRUS). Based on the development of the phenomena and model identification and ranking table (PIRT), the relative importance of the parameters was confirmed through the sensitivity analysis. The objective of the global uncertainty analysis is to evaluate all safety parameters of the system in the combined phase space formed by the parameters and dependent variables. The uncertainty propagation was performed by mapping the uncertainty bands of the model parameters through the MARS-LMR to determine the distributions for the fuel centerline, cladding, and coolant temperatures. The results show that the uncertainty bands of the temperatures are below the melting point. (C) 2017 Elsevier Ltd. All rights reserved.
引用
收藏
页码:1232 / 1241
页数:10
相关论文
共 17 条
  • [1] Aoki S., 1973, 1971 International Seminar on Heat Transfer in Liquid Metals, P569
  • [2] Cacuci D.G, 2010, Handbook of Nuclear Engineering
  • [3] Evaluation of existing correlations for the prediction of pressure drop in wire-wrapped hexagonal array pin bundles
    Chen, S. K.
    Todreas, N. E.
    Nguyen, N. T.
    [J]. NUCLEAR ENGINEERING AND DESIGN, 2014, 267 : 109 - 131
  • [4] HYDRODYNAMIC MODELS AND CORRELATIONS FOR BARE AND WIRE-WRAPPED HEXAGONAL ROD BUNDLES - BUNDLE FRICTION FACTORS, SUBCHANNEL FRICTION FACTORS AND MIXING PARAMETERS
    CHENG, SK
    TODREAS, NE
    [J]. NUCLEAR ENGINEERING AND DESIGN, 1986, 92 (02) : 227 - 251
  • [5] CRBRP-ARD-0034, 1976, CRBRPARD0034
  • [6] Graber H., 1973, 1971 International Seminar on Heat Transfer in Liquid Metals, P151
  • [7] Ha K. S., 2014, WORKSH SFR INH SAF
  • [8] SIMULATION OF THE EBR-II LOSS-OF-FLOW TESTS USING THE MARS CODE
    Ha, Kwi-Seok
    Jeong, Hae-Yong
    Cho, Chungho
    Kwon, Young-Min
    Lee, Yong-Bum
    Han, Dohee
    [J]. NUCLEAR TECHNOLOGY, 2010, 169 (02) : 134 - 142
  • [9] PAPIRUS, a parallel computing framework for sensitivity analysis, uncertainty propagation, and estimation of parameter distribution
    Heo, Jaeseok
    Kim, Kyung Doo
    [J]. NUCLEAR ENGINEERING AND DESIGN, 2015, 292 : 237 - 247
  • [10] Development of a multi-dimensional thermal-hydraulic system code, MARS 1.3.1
    Jeong, JJ
    Ha, KS
    Chung, BD
    Lee, WJ
    [J]. ANNALS OF NUCLEAR ENERGY, 1999, 26 (18) : 1611 - 1642