Application of the French Codes to the Pressurized Thermal Shocks Assessment

被引：10

作者：

Chen, Mingya ^{[1
]}

Qian, Guian ^{[2
]}

Shi, Jinhua ^{[3
]}

Wang, Rongshan ^{[1
]}

Yu, Weiwei ^{[1
]}

Lu, Feng ^{[1
]}

Zhang, Guodong ^{[1
]}

Xue, Fei ^{[1
]}

Chen, Zhilin ^{[1
]}

机构：

[1] Life Management Ctr, Suzhou Nucl Power Res Inst, Xihuan Rd, Suzhou 215004, Jiangsu, Peoples R China

[2] Paul Scherrer Inst, Nucl Energy & Safety Dept, Lab Nucl Mat, OHSA 06, CH-5232 Villigen, Switzerland

[3] Amec Foster Wheeler, Clean Energy Dept, 19B Brighouse Court,Barnett Way, Gloucester GL2 4NF, England

来源：

NUCLEAR ENGINEERING AND TECHNOLOGY | 2016年 / 48卷 / 06期

基金：

中国国家自然科学基金;

关键词：

Pressurized Thermal Shock; RCC-M; Reactor Pressure Vessel; RSE-M; Structural Integrity; STRUCTURAL INTEGRITY ASSESSMENT; VESSELS; PTS; RPV;

D O I：

10.1016/j.net.2016.06.009

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 integrity of a reactor pressure vessel (RPV) related to pressurized thermal shocks (PTSs) has been extensively studied. This paper introduces an integrity assessment of an RPV subjected to a PTS transient based on the French codes. In the USA, the "screening criterion" for maximum allowable embrittlement of RPV material is developed based on the probabilistic fracture mechanics. However, in the French RCC-M and RSE-M codes, which are developed based on the deterministic fracture mechanics, there is no "screening criterion". In this paper, the methodology in the RCC-M and RSE-M codes, which are used for PTS analysis, are firstly discussed. The bases of the French codes are compared with ASME and FAVOR codes. A case study is also presented. The results show that the method in the RCC-M code that accounts for the influence of cladding on the stress intensity factor (SIF) may be nonconservative. The SIF almost doubles if the weld residual stress is considered. The approaches included in the codes differ in many aspects, which may result in significant differences in the assessment results. Therefore, homogenization of the codes in the long time operation of nuclear power plants is needed. Copyright (C) 2016, Published by Elsevier Korea LLC on behalf of Korean Nuclear Society.

引用

页码：1423 / 1432

页数：10

共 23 条

[11] Lu F., 2010, P 18 INT C NUCL ENG
[12] Methodology for the pressurized thermal shock evaluation: recent improvements in French RPV PTS assessment
Moinereau, D
Bezdikian, G
Faidy, C
[J]. INTERNATIONAL JOURNAL OF PRESSURE VESSELS AND PIPING, 2001, 78 (2-3) : 69 - 83
[13] Oak Ridge National Laboratory, 2006, FRACT AN VESS OAK RI
[14] Structural integrity assessment of aging nuclear reactor pressure vessels
Pennell, WE
Malik, SNM
[J]. NUCLEAR ENGINEERING AND DESIGN, 1997, 172 (1-2) : 27 - 47
[15] Investigation on Constraint Effect of a Reactor Pressure Vessel Subjected to Pressurized Thermal Shocks
Qian, Guian
Niffenegger, Markus
[J]. JOURNAL OF PRESSURE VESSEL TECHNOLOGY-TRANSACTIONS OF THE ASME, 2015, 137 (01):
[16] Procedures, methods and computer codes for the probabilistic assessment of reactor pressure vessels subjected to pressurized thermal shocks
Qian, Guian
Niffenegger, Markus
[J]. NUCLEAR ENGINEERING AND DESIGN, 2013, 258 : 35 - 50
[17] RCC-M, 2010, DES CONSTR MECH COMP
[18] RSE-M, 2007, IN SERV INSP RUL MEC
[19] Transient computational fluid dynamics analysis of emergency core cooling injection at natural circulation conditions
Scheuerer, Martina
Weis, Johannes
[J]. NUCLEAR ENGINEERING AND DESIGN, 2012, 253 : 343 - 350
[20] The American Society of Mechanical Engineers, 2013, ASME BOIL PRESS VESS

← 1 2 3 →