Best estimate plus uncertainty analysis of a small‐break LOCA on an advanced Generation‐III pressurized water reactor

2020 ◽  
Vol 45 (8) ◽  
pp. 11916-11929
Author(s):  
Ye Yang ◽  
Chengcheng Deng ◽  
Jun Yang
Keyword(s):  
Uncertainty Analysis ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Best Estimate ◽  
Advanced Generation

Prospect on the LOCA Analysis Method on the Advanced Pressurized Water Reactor in China

2010 ◽  
Author(s):  
Yan Wang ◽  
Xie Heng
Keyword(s):  
Uncertainty Analysis ◽  
System Analysis ◽  
High Efficiency ◽  
Current System ◽  
National Laboratory ◽  
Pressurized Water Reactor ◽  
Economic Competition ◽  
Pressurized Water ◽  
Water Reactor ◽  
Best Estimate

The LOCA analysis for the advanced pressurized water reactor (PWR) is very important and the methods on it are developing. There are two basic approaches for LOCA (loss of coolant accident) licensing at current. One is based on the conservative requirement of Appendix K of 10CFR50.46 of USNRC, and another is the best estimate (BE) analysis methodology which needs strict sensitivity and uncertainty analysis. The results achieved by the best estimate analysis are closer to the reality than those achieved by the conservative methodology, and the realistic BELOCA analysis in nuclear realm becomes an international trend currently although its development still meet lots of challenges. The research and design on AP1000 to be built in China and larger advanced pressurized water reactor (CAP1400 or CAP1700) as one of Chinese national science & technology major project is in progress. The reliable licensing LOCA analysis as one of the most important accident safety analysis is absolutely necessary. There are three ways to get the code applied in licensing accident analysis: the first way is developing code based on the best estimated methodology with strict uncertainty analysis, the second way is to develop new analysis code based on the conservative Appendix K, and the third way is improving the current system analysis code, which had been verified and validated by many cases, to satisfy the requirements of Appendix K. The last one may be the most feasible way for the AP1000 design with high efficiency and economic competition. Some code like RELAP5 has been used for LOCA analysis, and its results showed good agreement with the test data. RELAP is the transient thermal-hydraulic system analysis code developed by Idaho National Laboratory, in which some model and correlations are not consistent with the conservative requirements of Appendix K, so it can not be applied for licensing LOCA analysis and evaluation directly. In this paper the way to develop analysis code for LOCA license is discussed, and some areas in RELAP code needed to be modified for according with Appendix K are also described, which will be helpful for the advanced PWR design and development in China.

scholarly journals Systematic and quantitative uncertainty analysis for rod ejection accident of pressurized water reactor

2017 ◽  
Vol 127 ◽  
pp. 369-376 ◽  
Author(s):  
Xinyi Pan ◽  
Bin Jia ◽  
Jingru Han ◽  
Jianping Jing ◽  
Chunming Zhang
Keyword(s):  
Uncertainty Analysis ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor

scholarly journals Uncertainty Analyses Applied to the UAM/TMI-1 Lattice Calculations Using the DRAGON (Version 4.05) Code and Based on JENDL-4 and ENDF/B-VII.1 Covariance Data

2013 ◽  
Vol 2013 ◽  
pp. 1-21 ◽  
Author(s):  
Augusto Hernández-Solís ◽  
Christophe Demazière ◽  
Christian Ekberg
Keyword(s):  
Uncertainty Analysis ◽  
Covariance Matrix ◽  
Cross Sections ◽  
Nuclear Data ◽  
Uncertainty Assessment ◽  
Test Case ◽  
Pressurized Water Reactor ◽  
Cross Sectional ◽  
Pressurized Water ◽  
Water Reactor

The OECD/NEA Uncertainty Analysis in Modeling (UAM) expert group organized and launched the UAM benchmark. Its main objective is to perform uncertainty analysis in light water reactor (LWR) predictions at all modeling stages. In this paper, multigroup microscopic cross-sectional uncertainties are propagated through the DRAGON (version 4.05) lattice code in order to perform uncertainty analysis on and 2-group homogenized macroscopic cross-sections. The chosen test case corresponds to the Three Mile Island-1 (TMI-1) lattice, which is a 15 15 pressurized water reactor (PWR) fuel assembly segment with poison and at full power conditions. A statistical methodology is employed for the uncertainty assessment, where cross-sections of certain isotopes of various elements belonging to the 172-group DRAGLIB library format are considered as normal random variables. Two libraries were created for such purposes, one based on JENDL-4 data and the other one based on the recently released ENDF/B-VII.1 data. Therefore, multigroup uncertainties based on both nuclear data libraries needed to be computed for the different isotopic reactions by means of ERRORJ. The uncertainty assessment performed on and macroscopic cross-sections, that is based on JENDL-4 data, was much higher than the assessment based on ENDF/B-VII.1 data. It was found that the computed Uranium 235 fission covariance matrix based on JENDL-4 is much larger at the thermal and resonant regions than, for instance, the covariance matrix based on ENDF/B-VII.1 data. This can be the main cause of significant discrepancies between different uncertainty assessments.

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