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Effects of ion irradiation on microstructure and properties of zirconium alloysdA review
Chunguang Yan,Rongshan Wang,Yanli Wang,Xitao Wang,Guanghai Bai 한국원자력학회 2015 Nuclear Engineering and Technology Vol.47 No.3
Zirconium alloys are widely used in nuclear reactors as structural materials. During the operation, they are exposed to fast neutrons. Ion irradiation is used to simulate the damage introduced by neutron irradiation. In this article, we briefly review the neutron irradiation damage of zirconium alloys, then summarize the effect of ion irradiation on microstructural evolution, mechanical and corrosion properties, and their relationships. The microstructure components consist of dislocation loops, second phase precipitates, and gas bubbles. The microstructure parameters are also included such as domain size and microstrain determined by X-ray diffraction and the S-parameter determined by positron annihilation. Understanding the relationships of microstructure and properties is necessary for developing new advanced materials with higher irradiation tolerance.
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Application of the French Codes to the Pressurized Thermal Shocks Assessment
Mingya Chen,Guian Qian,Jinhua Shi,Rongshan Wang,Weiwei Yu,Feng Lu,Guodong Zhang,Fei Xue,Zhilin Chen 한국원자력학회 2016 Nuclear Engineering and Technology Vol.48 No.6
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 RPVsubjected 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 theprobabilistic fracture mechanics. However, in the French RCC-M and RSE-M codes, whichare 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 forPTS analysis, are firstly discussed. The bases of the French codes are compared with ASMEand FAVOR codes. A case study is also presented. The results show that the method in theRCC-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 significantdifferences in the assessment results. Therefore, homogenization of the codes inthe long time operation of nuclear power plants is needed.
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