Abstract Graphite moderators have an extensive historical performance record, but also feature inherent challenges for modular High Temperature Gas-Cooled Reactors (mHTGRs). Challenges with graphite include non-uniform expansion and contraction under irradiation and build-up of potential energy during the bombardment of high energy neutrons that results in a large energy release under annealing. These challenges have led to the investigation and development of alternative moderators to be utilized in mHTGRs, including beryllium- and hydride-based concepts with compositions selected for favorable moderating power and the potential for improved in-service lifetime as compared to graphite. The proposed moderators are fabricated as two-phase composites with magnesium oxide, MgO, as the radiation-stable host matrix and beryllium metal, Be, beryllium oxide, BeO, or zirconium hydride, ZrHx=1 (to account for hydrogen loss from the hydride phase during processing), as the entrained moderating phase. Here, we evaluate the reactor performance and safety characteristics of these moderator concepts relative to a graphite reference using a Ft. Saint Vrain-style fuel block. We assessed the cycle length, discharge burnup, natural resource utilization, neutron flux spectra, moderating power, moderating ratio, critical size, moderator and fuel temperature feedback, fuel cycle cost, spent nuclear fuel and high level waste radioactivity per unit energy generated, and environmental impact per unit energy generated. The results demonstrate that the advanced moderators have the potential for comparable or enhanced cycle performance to that of the graphite reference case with significantly improved performance for an optimized moderator-to-fuel ratio design. These advanced moderators are also assessed from a reactor safety standpoint for Design Basis Accidents (DBAs) including Pressurized Loss of Forced Cooling and Depressurized Loss of Forced Cooling accidents for a 350 megawatt thermal prismatic-type mHTGR. The full core thermohydraulic analysis of DBAs show that the high volumetric heat capacity of the beryllium-based moderator grants them a greater margin to fuel failure in these analyses than a conventional graphite moderated system, but the lower thermal conductivity of the beryllium-based moderators leads to longer times at elevated temperatures.

中文翻译：

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模块化高温气冷堆两相复合慢化剂概念的反应堆性能和安全特性

摘要 石墨慢化剂具有广泛的历史性能记录，但也面临模块化高温气冷反应堆 (mHTGR) 的固有挑战。石墨面临的挑战包括辐照下的不均匀膨胀和收缩，以及在高能中子轰击过程中势能的积累，导致退火时释放大量能量。这些挑战导致了对用于 mHTGR 的替代减速剂的研究和开发，包括基于铍和氢化物的概念，其成分选择为有利的减速能力和与石墨相比具有更长使用寿命的潜力。所提出的慢化剂制成两相复合材料，其中氧化镁 MgO 作为辐射稳定的基质和铍金属 Be，氧化铍、BeO 或氢化锆，ZrHx=1（考虑到加工过程中氢化物相的氢损失），作为夹带的慢化相。在这里，我们使用 Ft 评估这些慢化剂概念相对于石墨参考的反应堆性能和安全特性。圣维兰式燃料块。我们评估了循环长度、放电燃耗、自然资源利用、中子通量谱、慢化功率、慢化比、临界尺寸、慢化剂和燃料温度反馈、燃料循环成本、乏核燃料和产生的每单位能量的高放废物放射性，以及单位能源产生的环境影响。结果表明，先进慢化剂具有与石墨参考案例相当或增强的循环性能的潜力，并显着提高了优化的慢化剂燃料比设计的性能。这些先进的慢化剂还从反应堆安全角度进行评估，以应对 350 兆瓦热棱柱型 mHTGR 的设计基础事故 (DBA)，包括强制冷却的加压损失和强制冷却的减压损失事故。DBA 的完整核心热工水力分析表明，在这些分析中，铍基慢化剂的高体积热容量使它们比传统的石墨慢化系统具有更大的燃料失效裕度，但铍基慢化剂的较低热导率导致在高温下延长时间。