双冷环形燃料球团的相场断裂模拟

IF 0.4 Q4 NUCLEAR SCIENCE & TECHNOLOGY Journal of Nuclear Fuel Cycle and Waste Technology Pub Date : 2022-08-08 DOI:10.1115/icone29-92230
Wei Li
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引用次数: 0

摘要

双冷环形核燃料是一种先进的设计,与传统的圆柱形燃料销相比,即使在高线性功率密度下,也有望大大降低燃料温度。虽然燃料温度可以低得多,但环形球团也会受到高得多的中子通量,这可能在正常运行时引起严重的开裂。本文研究了中子辐射作用下双冷环形UO2球团的准静态开裂问题。分析基于相场断裂模型,并结合氧扩散模型、热传导模型和力学平衡模型。所考虑的核燃料的热力学性能和辐照行为都与温度和辐照有关。特别地,由于氧的再分配,燃料蠕变的加速被包括在内。裂缝由内聚相场断裂方法控制的标量相场变量表示。这些模型在多物理场耦合仿真框架MOOSE中进行了数值实现。首次将扩散-热-力耦合断裂模型应用于反应堆启动、功率斜坡和停堆过程中双冷环形UO2燃料球团的断裂。初步发现UO2辐照蠕变对燃料球团破碎起重要作用。所开发的能力支持实验数据的解释,并可指导先进陶瓷核燃料的材料设计。
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Phase-Field Fracture Simulation of Dual-Cooled Annular Fuel Pellet
The dual-cooled annular nuclear fuel is an advanced design that is expected to greatly lower fuel temperature even under high linear power density, as compared to traditional cylindrical fuel pin. Although fuel temperature can be much lower, the annular pellet also receives much higher neutron fluence, which may induce severe cracking during normal operation. This work deals with quasi-static cracking of dual-cooled annular UO2 pellet under neutron radiation. The analysis is based on the phase-field fracture model coupled with an oxygen diffusion model, heat conduction model and mechanical equilibrium model. The considered thermo-mechanical properties and irradiation behaviors of the nuclear fuel are both temperature and irradiation dependent. Especially, the acceleration of fuel creep due to oxygen redistribution is included. The fracture is represented by a scalar phase-field variable governed by a cohesive phase-field fracture method. These models are numerically implemented in the multi-physics coupling simulation framework MOOSE. For the first time, the diffusion-thermo-mechanical coupled fracture model is applied to the dual-cooled annular UO2 fuel pellet cracking during reactor startup, power ramp and reactor shutdown. Preliminarily, UO2 irradiation creep is found to play an important role on the fuel pellet fragmentation. The developed capability supports interpretation of experimental data and can guide material design of advanced ceramic nuclear fuel.
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来源期刊
CiteScore
0.80
自引率
25.00%
发文量
35
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