摘要
核石墨可用作裂变核能反应堆如气冷堆和熔盐堆的慢化剂材料,还可用作为冷却剂和控制棒提供通道的结构部件.为了保证反应堆的寿期安全性,石墨堆芯不仅需要保持完整,还要避免过度变形,从而保证在工作状态和事故环境下堆芯冷却剂不会受阻,也不会妨碍控制棒的移动.因此,核石墨构件的结构完整性评估是反应堆设计的基本要素之一.在反应堆环境下石墨构件的应力分析,除了通常的弹性应变和热应变,由于中子辐照引起的额外应变也是考虑因素之一.因此,需要定义辐照环境下核石墨应力和应变相关的本构方程.本文介绍了一种用于辐照环境下核石墨材料应力分析的材料模型,并应用此模型对核石墨砖进行了应力分析,以期了解由辐照环境引起的应变对石墨砖应力的影响,相应的计算结果对堆芯核石墨砖的设计具有理论参考意义.
Nuclear graphite is used as a solid moderator in many fission nuclear reactors such as gas-cooled reactors and molten salt reactors. The role of graphite is to slow down neutrons emitted by fission, thus enhance their probabilities of interacting to produce further fission reactions. Graphite also acts as a major structural component in nuclear reactors, providing channels for the coolant and control rods. Graphite undergoes significant changes in dimensions and material properties due to the effects of irradiation damages. These irradiation-induced changes can lead to build up of significant stresses and deformation in the graphite components. Throughout reactor life it is essential that the graphite core structure remains sufficiently strong and undistorted in order to maintain fuel cooling, permit loading and unloading of the fuel and allow the necessary movements of control rods, in both normal and fault conditions. To perform structural integrity assessments, the definition of the constitutive equation relating stress and strain in the irradiated graphite material is required. Apart from the elastic and thermal strain, graphite in nuclear reactors also experiences additional strains due to fast neutron irradiation. Irradiation creep and irradiation induced dimensional changes are two most important of these irradiation strains. Nuclear graphite component structural integrity assessments are usually carried out using the finite element method. To model the dimensional change, irradiation creep and material properties a complex material model is required. In this paper a simple mathematical material model for isotropic graphite under irradiation condition is derived here. The model largely relies on empirical graphite property data obtained from material test reactor experiments. The model can be used to assess the nuclear graphite component structural integrity throughout the life of reactors, taking account of increasing irradiation fast neutron fluence and operating temperature. Also the model is a useful t
作者
曾广礼
TSANG D K L(Shanghai Institute of Applied Physics,Chinese Academy of Sciences,Shanghai 201800,China)
出处
《中国科学：物理学、力学、天文学》
CSCD
北大核心
2019年第11期103-112,共10页
Scientia Sinica Physica,Mechanica & Astronomica
基金
中国科学院先导科技专项(编号:XDA02040100)资助项目
关键词
核石墨
应力分析
核反应堆
nuclear graphite
stress analysis
nuclear reactor
