Transient analysis of nuclear graphite oxidation for high temperature gas cooled reactor
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- 作者：Wei Xu ; wxu12@mails.tsinghua.edu.cn" class="auth_mail" title="E-mail the corresponding author ; Lei Shi ; Yanhua Zheng
- 刊名：Nuclear Engineering and Design
- 出版年：2016
- 出版时间：September 2016
- 年：2016
- 卷：306
- 期：Complete
- 页码：138-144
- 全文大小：1691 K
文摘
Graphite is widely used as moderator, reflector and structural materials in the high temperature gas-cooled reactor pebble-bed modular (HTR-PM). In normal operating conditions or water/air ingress accident, the nuclear graphite in the reactor may be oxidized by air or steam. Oxidation behavior of nuclear graphite IG-110 which is used as the structural materials and reflector of HTR-PM is mainly researched in this paper. To investigate the penetration depth of oxygen in IG-110, this paper developed the one dimensional spherical oxidation model. In the oxidation model, the equations considered graphite porosity variation with the graphite weight loss. The effect of weight loss on the effective diffusion coefficient and the oxidation rate was also considered in this model. Based on this theoretical model, this paper obtained the relative concentration and local weight loss ratio profile in graphite. In addition, the local effective diffusion coefficient and oxidation rate in the graphite were also investigated.