Heat transfer simulations of the UO2 particle-graphite system in TREAT fuel

被引：6

作者：

Mo, Kun ^{[1
]}

Yun, Di ^{[1
]}

Yacout, Abdellatif M. ^{[1
]}

Wright, Arthur E. ^{[1
]}

机构：

[1] Argonne Natl Lab, Nucl Engn Div, Lemont, IL 60439 USA

来源：

NUCLEAR ENGINEERING AND DESIGN | 2015年 / 293卷

关键词：

THERMAL-CONDUCTIVITY DEGRADATION; ACCIDENT-TOLERANT FUELS; THERMOPHYSICAL PROPERTIES; TEMPERATURE; MODEL; PERFORMANCE; OXIDATION;

D O I：

10.1016/j.nucengdes.2015.08.009

中图分类号：

TL [原子能技术]; O571 [原子核物理学];

学科分类号：

0827 ; 082701 ;

摘要：

In this study, a heat transfer simulation of a UO2 particle-graphite system in highly enriched nuclear fuel at the Transient Reactor Test Facility (TREAT) was performed using the finite element method. Different factors that can impact fuel performance were modeled and implemented in the simulated micro-scale UO2 particle-graphite system. The fission fragments caused an irradiation-induced degradation of the thermal conductivity of the graphite, which added major heat resistance to the irradiated system. The effect of graphite quality and irradiation on the UO2 particles has also been evaluated, but neither has an impact as pronounced as the fission fragment damage to the graphite. By combining these factors, the dynamic temperature profiles were obtained, and the limitations on particle size in the irradiated and unirradiated UO2 particle-graphite systems have been determined. (C) 2015 Published by Elsevier B.V.

引用

页码：313 / 322

页数：10

共 55 条

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