5-23. Nuclear fuel cladding performance assessment through materials genome engineering

Weijia Gong ¹，Pavel Trtik ²

1.Northwestern Polytechnical University, Xi’an 710072

2.Paul Scherrer Institut (PSI), Villigen 5232, Switzerland

Abstract: Zirconium alloys are widely used as nuclear fuel cladding materials because of their low neutron absorption cross-section. During service, zirconium cladding tubes take up a part of the hydrogen produced by the oxidation reaction between the reactor coolant water and the hot rod surface. Hydrogen distribution can often be non-uniform given the general tendency for hydrogen to diffuse down thermal and concentration gradients, and up stress gradients. Hydride platelets precipitate when hydrogen content exceeds the terminal solubility. Hydrides are predominantly oriented in the circumferential direction relative to the present α-Zr texture, and can be re-oriented in the radial direction when a high pressure-induced cladding tensile hoop stress is encountered, which originates mainly from fuel pellet-cladding interaction and fission gas pressure. These hydrides can raise the risk for the fuel rod integrity by decreasing cladding fracture toughness and tensile ductility. Radial hydrides may lead to microcrack propagation along the hydride/matrix interface and eventually resulting in radial fracture through the cladding wall. Therefore, hydrogen diffusion and precipitation are essential to evaluate cladding performance behavior in service and assess integrity safety during spent fuel storage.

Neutron penetrating most of metals, is able to reveal microstructure and composition in bulk materials. Neutron imaging provides a high-throughput technique to the determination of hydrogen concentration given a certain spatial resolution, due to an order of magnitude higher in the neutron absorption scattering cross-section of hydrogen than zirconium. In the study, the state-of-the-art detector of the PSI Neutron Microscope was employed to quantitatively investigate hydrogen diffusion and hydrides precipitation, where inhomogeneous hydrogen concentration field was efficiently quantified. In association with multi-physics computation, strengths of various diffusion driving force were evaluated. The presented work demonstrates the approach of materials genome engineering including high-resolution neutron imaging and high-throughput finite element modeling for the assessment of hydrogen up-take and performance in zirconium nuclear fuel cladding.

基于材料基因工程技术的核燃料包壳服役行为评价

公维佳¹，Pavel Trtik ²

1.西北工业大学，西安 710072

2.Paul Scherrer Institut (PSI), Villigen 5232, Switzerland

摘要：锆合金因中子吸收截面小，而被广泛应用于轻水堆核电站的燃料包壳材料。在堆内服役中，燃料包壳在冷却水一侧吸收锆水腐蚀产生的氢，固溶态氢在温度、浓度和应力梯度作用下扩散、迁移，从而通常呈非均匀分布。当包壳吸氢量超过溶解极限时，氢析出形成沿包壳管周向堆垛的氢化物，然而在燃料肿胀等因素导致的包壳管环向拉应力作用下，氢化物宏观取向由周向变为径向。研究氢的扩散及氢化物析出，是研究氢致力学性能降低等锆合金服役行为的基础，并对乏燃料贮存过程中包壳材料的结构完整性评价具有重要的指导意义。

中子穿透大部分金属材料，能够反映体材料的内部结构与成分信息，同时氢的中子截面比锆高一个数量级，所以中子成像技术是测定核燃料包壳锆合金中氢分布的一种有力的高通量表征手段。本工作借助目前中子成像领域分辨率最高的谱仪PSI Neutron Microscope，高效、定量研究锆合金中多因素协同作用下氢迁移及氢化物析出行为，实现非均匀氢浓度场数据的高通量表征与获取，然后通过有限元多物理场模拟，评估不同扩散驱动力的强度，从而实现以高通量模拟计算结合中子成像高通量表征的材料基因工程手段，对核燃料包壳锆合金吸氢与服役行为进行安全评价、理论预测。

关键词：高通量表征；中子成像；有限元多物理场模拟；锆合金；氢