Fracture Analysis of Vessels. Oak Ridge FAVOR, v06.1, Computer Code: Theory and Implementation of Algorithms, Methods, and Correlations

Williams, P. T.¹  Dickson, T. L.¹  Yin, S.¹ 

[1] Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

关键词: pressurized thermal shock;    probabilistic fracture mechanics;    reactor pressure vessels;    FAVOR;    FAVOR v06.1;    code;    Heavy Section Steel Technology;   

DOI  :  10.2172/1154649
RP-ID  :  ORNL/TM--2007/026
PID  :  OSTI ID: 1154649
Others  :  R&D Project: 401001060, 41WT77802

学科分类：化学（综合）

美国|英语

来源: SciTech Connect

【 摘 要 】

The current regulations to insure that nuclear reactor pressure vessels (RPVs) maintain their structural integrity when subjected to transients such as pressurized thermal shock (PTS) events were derived from computational models developed in the early-to-mid 1980s. Since that time, advancements and refinements in relevant technologies that impact RPV integrity assessment have led to an effort by the NRC to re-evaluate its PTS regulations. Updated computational methodologies have been developed through interactions between experts in the relevant disciplines of thermal hydraulics, probabilistic risk assessment, materials embrittlement, fracture mechanics, and inspection (flaw characterization). Contributors to the development of these methodologies include the NRC staff, their contractors, and representatives from the nuclear industry. These updated methodologies have been integrated into the Fracture Analysis of Vessels -- Oak Ridge (FAVOR, v06.1) computer code developed for the NRC by the Heavy Section Steel Technology (HSST) program at Oak Ridge National Laboratory (ORNL). The FAVOR, v04.1, code represents the baseline NRC-selected applications tool for re-assessing the current PTS regulations. This report is intended to document the technical bases for the assumptions, algorithms, methods, and correlations employed in the development of the FAVOR, v06.1, code.