Thermo-mechanical analysis of the prismatic fuel assembly of VHTR in normal operational condition
- 作者：Ji-Ho Kang ; jhkang@kaeri.re.kr ; Nam-il Tak ; Min-Hwan Kim
- 关键词：Prismatic Modular Reactor (PMR) ; Very High Temperature Reactor (VHTR) ; High Temperature Gas-cooled Reactor (HTGR) ; Fuel assembly ; Thermal stress ; Structural integrity
- 刊名：Annals of Nuclear Energy
- 出版年：2012
- 期刊代码：1_03064549
- 类别：pw
- 出版时间：June, 2012
- 卷：44
- 期：Complete
- 页码：76-86
- 文件大小：2985 K
摘要
A Very High Temperature Reactor (VHTR) is designed to generate high temperature which is typically around 950 掳C at the outlet. The high temperature results in thermal stress with dimensional changes in addition to the mechanical stress caused by the external loads. It is necessary to assess the structural integrity of the graphite blocks and fuel compacts for thermal stress and dimensional changes; however, precise assessment has not been conducted because of the complexity of the fuel assembly geometry and absence of accurate temperature distribution. In the present study, a finite element analysis (FEA) of each fuel element in a 10-story fuel assembly was carried out by using the temperature distribution which was calculated from a previous study of computational fluid dynamics (CFD) analysis on the fuel assembly. The results showed that the maximum stress level in the fuel elements reaches 32%of the tensile strength. The maximum stress point was at the circumference of the bottom surface of a graphite plug on which the thermal expansion of the fuel compacts pushes up the graphite plug. Mitigation of the maximum stress by placing a small gap between the graphite plug and the fuel compact was investigated and the effect of the gap was confirmed. Stress evaluation was performed according to a draft of standard for nuclear graphite. All stress components lied within the limits and the structural integrity was regarded to be attained.