Engineered Zircaloy Cladding Modifications for Improved Accident Tolerance of LWR Fuel

该综合研究项目的目的是评估修改后的锆合金轻水堆包壳在正常的沸水堆/压水堆运行和非正常事件下的性能。计算和实验方案的组合将被用来设计和评估修改锆合金包层的腐蚀和加速氧化物生长，前者与正常运行，后者与蒸汽暴露在冷却剂损失事故（LOCA）和低压堆芯回灌。还将研究改进铀芯块以提高热导率的方法，目的是降低芯块温度梯度、相关的芯块膨胀和芯块-包壳相互作用（PCI）以及裂变产物释放。此外，Pb-Bi共晶液态金属间隙填料将被研究，以促进丸包传热。包壳和芯块性能评估都将纳入中子学和热工水力学的反应堆系统建模工作，从而为事故容忍核燃料提供一种整体方法。拟议项目汇集了与正常运行和非正常情况下改性核燃料和轻水堆性能研究相关的广泛技术领域的人员，设施和能力。拟议的项目利用现有的能源部NEUP支持，并将合作活动从美国学术界扩展到美国工业界，美国国家实验室和英国学术界。预期的交付成果将是：i）在正常沸水堆和压水堆运行条件下以及在非正常LOCA条件下，改进包壳和芯块性能的实验数据库; ii）改进燃料性能代码的预测能力; iii）改进中子学和热工水力学性能代码的预测能力。两种途径的事故容忍轻水堆燃料的设想，都基于现有的锆合金包层的修改。第一个是通过应用涂层对包层表面进行改性，该涂层旨在将M + O至MO反应从高温蒸汽暴露期间的氧化物生长转移开。第二个是本体包层组合物的改性，以促进制造期间次要相的沉淀。这些沉淀物在正常操作下将是稳定的，但在温度偏移期间溶解;溶质元素向自由表面的迁移将使反应远离氧化物形成。改进的芯块热导率将通过降低裂变气体间隙压力和PCI来限制包壳环向应力。预期燃料棒的协同响应，其中脆性放热氧化物形成的组合缓解、降低的包壳温度和降低的包壳应力导致关于包壳失效的事故容限。

This integrated research project aims to evaluate the modified Zircaloy LWR cladding performance under normal BWR/PWR operation and off-normal events. A combination of computational and experimental protocols will be employed to design and evaluate modified Zircaloy cladding with respect to corrosion and accelerated oxide growth, the former associated with normal operation, the latter associated with steam exposure during loss of coolant accidents (LOCAs) and low-pressure core refloods. Urania pellet modifications to improve thermal conductivity will be investigated as well, with the goal of reducing pellet temperature gradients, associated pellet swelling and pellet-cladding interaction (PCI), and fission product release. In addition, Pb-Bi eutectic liquid metal gap fillers will be investigated to promote pellet-cladding heat transfer. Both the cladding and pellet performance evaluations will be incorporated into a reactor system modelling effort of neutronics and thermal hydraulics, thereby providing a holistic approach to accident tolerant nuclear fuel. The proposed project brings together personnel, facilities, and capabilities across a wide range of technical areas relevant to the study of modified nuclear fuel and LWR performance during normal operation and off-normal scenarios. The proposed project leverages existing DOE NEUP support and extends collaborative activities from U.S. academia to U.S. industry, U.S. national laboratories, and to UK academia. Anticipated deliverables will be i) an experimental data base of modified cladding and pellet performance under normal BWR and PWR operational conditions and under off-normal LOCA conditions, ii) improved predictive capability of fuel performance codes, and iii) improved predictive capability of neutronics and thermal hydraulics performance codes. Two pathways toward accident tolerant LWR fuel are envisioned, both based on the modification of existing Zircaloy cladding. The first is the modification of the cladding surface by the application of a coating layer designed to shift the M+O to MO reaction away from oxide growth during steam exposure at elevated temperature. The second is the modification of the bulk cladding composition to promote precipitation of minor phase(s) during fabrication. These precipitates will be stable under normal operation, but dissolve during the temperature excursions; the migration of solute elements to the free surface would then shift the reaction away from oxide formation. Improved pellet thermal conductivity will act to limit cladding hoop stress via reduced fission gas gap pressure and PCI. A synergistic response of the fuel rod is anticipated in which the combined mitigation of brittle exothermic oxide formation, reduced cladding temperature, and reduced cladding stress lead to accident tolerance with respect to cladding failure.