Uncertainty analysis of HPR-1000 LOCA with probabilistic and deterministic methods

被引：10

作者：

Chen, Wei ^{[1
]}

Xiong, Qingwen ^{[1
]}

Wu, Dan ^{[1
]}

Ding, Shuhua ^{[1
]}

Qian, Libo ^{[1
]}

Wu, Qing ^{[1
]}

机构：

[1] Nucl Power Inst China, Sci & Technol Reactor Syst Design Technol Lab, Chengdu 610213, Peoples R China

来源：

PROGRESS IN NUCLEAR ENERGY | 2022年 / 146卷

关键词：

HPR-1000; Uncertainty analysis; Deterministic method; Probabilistic method; LOCA; PRESSURIZED-WATER REACTOR; QUANTIFICATION; SENSITIVITY; SIZES;

D O I：

10.1016/j.pnucene.2022.104174

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：

The HPR-1000 is an advanced Gen III pressurized water reactor (PWR) in China. For the purpose of carrying out uncertainty analysis of the LOCA scenario, two different types of best estimate plus uncertainty (BEPU) method were utilized, namely the DRM of the deterministic method and a typical input-driven probabilistic method. The DRM is carried out by introducing a penalization mode to envelop the input uncertainties, while the probabilistic input-driven method is implemented by performing the uncertainty propagation after the uncertainty sources are identified and quantified. The peak cladding temperature (PCT) was selected to be the figure of merit, and the 95% confidence limit as well as the tolerance limit of the PCT in the LOCA scenario were obtained. Results show that even though there are some phenomenal differences exist, the uncertainty limit obtained through the two methods are similar and both fall below the acceptance criterion.

引用

页数：8

共 29 条

[1] THE CATHARE CODE STRATEGY AND ASSESSMENT
BARRE, F
BERNARD, M
[J]. NUCLEAR ENGINEERING AND DESIGN, 1990, 124 (03) : 257 - 284
[2] QUANTIFYING REACTOR SAFETY MARGINS .1. AN OVERVIEW OF THE CODE SCALING, APPLICABILITY, AND UNCERTAINTY EVALUATION METHODOLOGY
BOYACK, BE
CATTON, I
DUFFEY, RB
GRIFFITH, P
KATSMA, KR
LELLOUCHE, GS
LEVY, S
ROHATGI, US
WILSON, GE
WULFF, W
ZUBER, N
[J]. NUCLEAR ENGINEERING AND DESIGN, 1990, 119 (01) : 1 - 15
[3] A comparative review of sensitivity and uncertainty analysis of large-scale systems - II: Statistical methods
Cacuci, DG
Ionescu-Bujor, M
[J]. NUCLEAR SCIENCE AND ENGINEERING, 2004, 147 (03) : 204 - 217
[4] Chang Y., 2020, INT J ADV NUCL REACT, V2, P34, DOI [10.1016/j.jandt.2020.07.002, DOI 10.1016/J.JANDT.2020.07.002]
[5] DAuria F., 2008, IAEA SAFETY REPORT S, V52
[6] The development of ARSAC for modeling nuclear power plant system
Deng, Jian
Ding, Shuhua
Li, Zhongchun
Huang, Tao
Wu, Dan
Wang, Jie
Ma, Haifu
[J]. PROGRESS IN NUCLEAR ENERGY, 2021, 140
[7] UNCERTAINTY AND SENSITIVITY ANALYSIS OF A POST-EXPERIMENT CALCULATION IN THERMAL-HYDRAULICS
GLAESER, H
HOFER, E
KLOOS, M
SKOREK, T
[J]. RELIABILITY ENGINEERING & SYSTEM SAFETY, 1994, 45 (1-2) : 19 - 33
[8] IAEA, 2003, SAF MARG OP REACT AN
[9] Myers R., 1995, TECHNOMETRICS, V38, P285, DOI DOI 10.2307/1270613
[10] Nissley M.E, 2003, WESTINGHOUSE PROPRIE

← 1 2 3 →