Data and modeling sensitivity analysis for molten salt fast reactor benchmark – Static calculations

IF 3.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY Progress in Nuclear Energy Pub Date : 2024-09-16 DOI:10.1016/j.pnucene.2024.105446

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Abstract

The Molten Salt Reactor (MSR) idea is increasingly being recognized in the nuclear field due to its potential safety, sustainability, and economic efficiency advantages. The Molten Salt Fast Reactor (MSFR) benchmark, introduced in 2019, highlighted variations in results tied to different neutron cross-section libraries. This study investigates the impact of utilizing the ENDF/B-VIII.0 and JEFF-3.3 cross-section libraries for MSFR benchmark assessment compared to the ENDF/B-VII.1 database. Monte Carlo based open source code OpenMC is used for the analyses. Rigorous sensitivity analyses assess the influence of individual components, including the cross-section database, resonance elastic scattering, and Thermal Scattering Law (TSL). Beyond the criticality assessments, parameters such as delayed neutron fraction, temperature coefficient of reactivity, and neutron spectrum are compared for different cross-section libraries. Our analyses reveal that incorporating new evaluations for ²³³U (n,γ) and fission cross-sections in ENDF/B-VIII.0 significantly alters criticality results, i.e., more than 1700 pcm difference is seen between libraries. Similarly, critical concentration using ENDF/B-VII.1 and JEFF-3.3 is over-predicted by approximately 3%. The variations in Thermal Scattering Law (TSL) files do not yield substantial differences in outcomes due to the fast spectrum of the reactor. In some cases, the treatment of resonance elastic scattering leads to reactivity differences greater than 50 pcm. The benchmark compares ²³³U-started and Minor Actinide (MA)-started core. From the reactor physics point of view, the MA-started core leads to a 29% higher (n, γ) reaction rate than the ²³³U-started core. A 3–4% smaller value of thermal reactivity coefficient is obtained using the ENDF/B-VIII.0 library compared to the ENDF/B-VII.1 value. Using the ENDF/B-VIII.0 for the MSFR benchmark signifies using newer and better data for the GEN-IV reactors neutron physics calculations.

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熔盐反应堆（MSR）的理念因其潜在的安全性、可持续性和经济效益优势而日益得到核领域的认可。2019 年推出的熔盐快堆（MSFR）基准凸显了与不同中子截面库相关的结果差异。与ENDF/B-VII.1数据库相比，本研究调查了利用ENDF/B-VIII.0和JEFF-3.3截面库进行MSFR基准评估的影响。分析使用了基于蒙特卡罗的开放源代码 OpenMC。严格的敏感性分析评估了各个组成部分的影响，包括截面数据库、共振弹性散射和热散射定律（TSL）。除了临界评估，我们还比较了不同截面库的延迟中子分数、反应性温度系数和中子谱等参数。我们的分析表明，在ENDF/B-VIII.0中加入对233U（n,γ）和裂变截面的新评估极大地改变了临界结果，即不同截面库之间的差异超过1700 pcm。同样，ENDF/B-VII.1 和 JEFF-3.3 对临界浓度的预测高出约 3%。由于反应器的快速频谱，热散射定律（TSL）文件的变化不会导致结果出现重大差异。在某些情况下，共振弹性散射的处理导致反应性差异超过 50 pcm。该基准比较了 233U 启动的堆芯和小锕系元素（MA）启动的堆芯。从反应堆物理学的角度来看，MA 启动的堆芯比 233U 启动的堆芯的（n，γ）反应速率高 29%。使用ENDF/B-VIII.0库得到的热反应系数比ENDF/B-VII.1的值小3-4%。在 MSFR 基准中使用 ENDF/B-VIII.0 意味着在 GEN-IV 反应堆中子物理计算中使用更新、更好的数据。

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来源期刊

Progress in Nuclear Energy 工程技术-核科学技术

CiteScore

5.30

自引率

14.80%

发文量

331

审稿时长

3.5 months

期刊介绍： Progress in Nuclear Energy is an international review journal covering all aspects of nuclear science and engineering. In keeping with the maturity of nuclear power, articles on safety, siting and environmental problems are encouraged, as are those associated with economics and fuel management. However, basic physics and engineering will remain an important aspect of the editorial policy. Articles published are either of a review nature or present new material in more depth. They are aimed at researchers and technically-oriented managers working in the nuclear energy field. Please note the following: 1) PNE seeks high quality research papers which are medium to long in length. Short research papers should be submitted to the journal Annals in Nuclear Energy. 2) PNE reserves the right to reject papers which are based solely on routine application of computer codes used to produce reactor designs or explain existing reactor phenomena. Such papers, although worthy, are best left as laboratory reports whereas Progress in Nuclear Energy seeks papers of originality, which are archival in nature, in the fields of mathematical and experimental nuclear technology, including fission, fusion (blanket physics, radiation damage), safety, materials aspects, economics, etc. 3) Review papers, which may occasionally be invited, are particularly sought by the journal in these fields.