Application of the French Codes to the Pressurized Thermal Shocks Assessment

Nuclear Engineering and Technology

Jinhua Shi¹  Guian Qian²  Guodong Zhang³  Mingya Chen³  Fei Xue³  Zhilin Chen³  Feng Lu³  Weiwei Yu³  Rongshan Wang³ 

[1] Amec Foster Wheeler, Clean Energy Department, 19B Brighouse Court, Barnett Way, Gloucester GL2 4NF, UK;Paul Scherrer Institute, Nuclear Energy and Safety Department, Laboratory for Nuclear Materials, OHSA/06, 5232, Villigen PSI, Switzerland;Suzhou Nuclear Power Research Institute, Life Management Center, Xihuan Road, 215004, Suzhou, Jiangsu Province, PR China;

关键词: Pressurized Thermal Shock;    RCC-M;    Reactor Pressure Vessel;    RSE-M;    Structural Integrity;   

DOI  :  10.1016/j.net.2016.06.009

来源: DOAJ

【 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.

【 授权许可】

Unknown