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首页> 外文期刊>Annals of nuclear energy >Monte Carlo neutronics benchmarks on nuclear fuel depletion: A review
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Monte Carlo neutronics benchmarks on nuclear fuel depletion: A review

机译:核燃料耗尽的Monte Carlo Neutronics基准:审查

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摘要

Monte Carlo (MC) neutronics codes are used widely for academic and industrial needs. Several schemes of coupling MC neutronics code with isotope generation and depletion code exist, which are used for performing nuclear fuel depletion simulations. These simulations can estimate the inventory of isotopes in neutron irradiated nuclear reactor fuel. However, the accuracy of these simulations shall be validated through experiments. MC codes are seldom validated by isotopic benchmarks compared to criticality benchmarks. This work compiles and analyzes the fuel depletion benchmarks and validations used to analyze the performance of MC-based fuel depletion neutronics codes. Analyses of these benchmarks and validations showed that the computed concentrations of (CS)-C-133, (CS)-C-135, (CS)-C-137, Nd-148, (PU)-P-239, (PU)-P-240, and Pu-241 in the irradiated fuel by the depletion codes agreed with the measured values within 10% error. However, the computed concentrations of Sb-125 , Cm-242, Cm-243, Cm-244, Cm-245, and Cm-246 had errors more than 15% compared to the measured values. Ventina depletion code showed the most accurate predictions for the greatest number of isotope concentrations compared to ORIGEN2 and CINDER90. (C) 2021 Elsevier Ltd. All rights reserved.
机译:Monte Carlo(MC)中型码广泛用于学术和工业需求。存在具有同位素生成和耗尽码的耦合MC中光电码的几个方案,用于执行核燃料耗尽模拟。这些模拟可以估计中子辐照核反应堆燃料中同位素的库存。然而,这些模拟的准确性应通过实验进行验证。与关键性基准相比,MC代码很少通过同位素基准验证。该工作编译并分析了用于分析基于MC的燃料耗尽中子码的性能的燃料耗尽基准和验证。这些基准和验证的分析表明,(CS)-C-133,(CS)-C-135,(CS)-C-137,ND-148,(PU)-P-239,(PU)的计算浓度通过耗尽码在辐照燃料中的-P-240和PU-241通过10%误差内的测量值商定。然而,与测量值相比,Sb-125,CM-242,CM-243,CM-244,CM-245和CM-246的计算浓度具有大于15%的误差。 Ventina Fepletion代码显示与OROREN2和CINDER90相比最多的同位素浓度最准确的预测。 (c)2021 elestvier有限公司保留所有权利。

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