Chromia-doped UO2 fuel: An engineering model for chromium solubility and fission gas diffusivity

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY Journal of Nuclear Materials Pub Date : 2024-07-26 DOI:10.1016/j.jnucmat.2024.155301

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Abstract

Increasing the average grain size of fuel pellets by doping them with chromium oxide is one strategy to improve oxide nuclear fuels performance. The promoted fission gas retention is thought to improve the performance of the fuel at high burnup. In this work, we review models for the solubility of chromium in UO₂, and the evolution of the chromium phases in the fuel matrix during irradiation. These models are implemented in SCIANTIX, an open-source mesoscale code describing inert gas behaviour in nuclear fuel. We adjusted the chromium solubility model keeping each parameter within its range of compatibility with experimental data, targeting a better representation of available electron probe microanalysis data of chromium content in fuel after irradiation. As for fission gas behaviour, we considered a physics-based description of the chromium impact on the fission gas diffusivity in fuel grains. The expression for the fission gas diffusivity in standard non-doped uranium oxide has been extended by introducing the impact of the concentration of defects introduced by interstitial oxygen excess representing the effect of chromium content in the fuel itself. A preliminary integral assessment of the proposed models has been carried out against the available experimental data.

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掺铬二氧化铀燃料：铬溶解度和裂变气体扩散的工程模型

通过掺杂氧化铬来增加燃料颗粒的平均粒度，是改善氧化物核燃料性能的一种策略。人们认为，提高裂变气体截留率可改善燃料在高燃耗下的性能。在这项工作中，我们回顾了铬在二氧化铀中的溶解度模型，以及辐照过程中燃料基质中铬相的演变模型。这些模型在 SCIANTIX 中实现，SCIANTIX 是描述核燃料中惰性气体行为的开源中尺度代码。我们调整了铬溶解度模型，使每个参数保持在与实验数据兼容的范围内，目的是更好地反映辐照后燃料中铬含量的电子探针显微分析数据。至于裂变气体行为，我们考虑用物理学方法来描述铬对裂变气体在燃料晶粒中扩散性的影响。标准无掺杂氧化铀中裂变气体扩散率的表达式已通过引入间隙氧过量引入的缺陷浓度的影响（代表燃料本身铬含量的影响）而得到扩展。根据现有的实验数据，对提出的模型进行了初步的整体评估。

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来源期刊

Journal of Nuclear Materials 工程技术-材料科学：综合

CiteScore

5.70

自引率

25.80%

发文量

601

审稿时长

63 days

期刊介绍： The Journal of Nuclear Materials publishes high quality papers in materials research for nuclear applications, primarily fission reactors, fusion reactors, and similar environments including radiation areas of charged particle accelerators. Both original research and critical review papers covering experimental, theoretical, and computational aspects of either fundamental or applied nature are welcome. The breadth of the field is such that a wide range of processes and properties in the field of materials science and engineering is of interest to the readership, spanning atom-scale processes, microstructures, thermodynamics, mechanical properties, physical properties, and corrosion, for example. Topics covered by JNM Fission reactor materials, including fuels, cladding, core structures, pressure vessels, coolant interactions with materials, moderator and control components, fission product behavior. Materials aspects of the entire fuel cycle. Materials aspects of the actinides and their compounds. Performance of nuclear waste materials; materials aspects of the immobilization of wastes. Fusion reactor materials, including first walls, blankets, insulators and magnets. Neutron and charged particle radiation effects in materials, including defects, transmutations, microstructures, phase changes and macroscopic properties. Interaction of plasmas, ion beams, electron beams and electromagnetic radiation with materials relevant to nuclear systems.