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首页> 外文期刊>Annals of nuclear energy >A fuel depletion analysis of the MSRE and three conceptual small molten-salt reactors for Mo-99 production
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A fuel depletion analysis of the MSRE and three conceptual small molten-salt reactors for Mo-99 production

机译：MSRE和三个概念性小型熔融盐反应堆用于Mo-99生产的燃料耗竭分析

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摘要

Producing the medical isotope Mo-99 in molten-salt fluoride fuel by using fission is an attractive option, in particular for global-scale supply. Theoretically, 1 MW of U-235 fission produces approximately 2000 TBq of Mo-99 in saturation. Molten-salt reactors (MSRs) are distinguished by the circulation of liquid fuel in and out of their cores, which provides unique advantages, such as online fuel addition and isotope extraction. However, these features also complicate reactor physics analysis compared with solid-fuel designs. A special sequence based on stepwise SCALE6/TRITON calculations was developed to automate MSR depletion analysis. Mo-99 production and accumulation were investigated by analyzing the Molten Salt Reactor Experiment (MSRE). Three conceptual MSR core configurations (HomoType, RingType and HeteType) were then examined and compared regarding their neutronic characteristics, including critical dimension, neutron spectrum, fuel loading, and the radiotoxicity of the remaining spent fuel. The HeteType core model, which offered superior fuel utilization and radiotoxicity minimization, was considered the most promising design.
机译：通过裂变在熔融盐氟化物燃料中生产医学同位素Mo-99是一个有吸引力的选择，特别是对于全球规模的供应而言。从理论上讲，1兆瓦的U-235裂变会产生约2000 TBq的Mo-99饱和。熔融盐反应堆（MSR）的特点是液体燃料在堆芯内外循环，这具有独特的优势，例如在线燃料添加和同位素提取。但是，与固体燃料设计相比，这些功能还使反应堆物理分析复杂化。开发了基于逐步SCALE6 / TRITON计算的特殊序列来自动进行MSR消耗分析。通过分析熔融盐反应堆实验（MSRE）研究了Mo-99的产生和积累。然后检查了三种概念性MSR堆芯配置（同型，环型和异型），并比较了它们的中子特性，包括临界尺寸，中子谱，燃料负载和剩余乏燃料的放射性。 HeteType核心模型可提供出色的燃料利用率和最小的放射毒性，被认为是最有前途的设计。

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  • 来源
    《Annals of nuclear energy》 |2014年第9期|111-117|共7页
  • 作者

    R.J. Sheu; C.C. Chao; O. Feynberg; Y.-W.H. Liu;

  • 作者单位

    Institute of Nuclear Engineering and Science, National Tsing-Hua University, Hsinchu 30013, Taiwan, ROC,Department of Engineering and System Science, National Tsing-Hua University, Hsinchu 30013, Taiwan, ROC;

    Institute of Nuclear Engineering and Science, National Tsing-Hua University, Hsinchu 30013, Taiwan, ROC;

    National Research Centre (NRC 'Kurchatov Institute'), Moscow, Russian Federation;

    Institute of Nuclear Engineering and Science, National Tsing-Hua University, Hsinchu 30013, Taiwan, ROC,Department of Engineering and System Science, National Tsing-Hua University, Hsinchu 30013, Taiwan, ROC;

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  • 正文语种 eng
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  • 关键词

    Molten salt reactor; Mo-99 production; SCALE6/TRITON; Fuel depletion;

    机译：熔融盐反应器;Mo-99生产;SCALE6 / TRITON;燃油消耗;

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