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SCALE 6.2.4 Validation for Light Water Reactor Decay Heat Analysis
Nuclear Technology ( IF 1.5 ) Pub Date : 2021-08-18 , DOI: 10.1080/00295450.2021.1935165
Germina Ilas 1 , Joseph R. Burns 1
Affiliation  

Abstract

Energy release from the decay of radionuclides in nuclear fuel after its discharge from reactor is a critical parameter for design, safety, and licensing analyses of used nuclear fuel storage, transportation, and repository systems. Well-validated computational tools and nuclear data are essential for decay heat prediction. This paper summarizes the validation of the SCALE nuclear analysis code system version 6.2.4, used with ENDF/B-VII.1 libraries, for decay heat analysis of light water reactor used fuel. The experimental data used for validation include full-assembly decay heat measurements that cover assembly burnups of 5 to 51 GWd/tonne U, cooling times after discharge in the 2- to 27-year range, and initial fuel enrichments up to 4 wt% 235U. The comparison between calculated (C) and experimental (E) decay heat showed very good agreement, with an average C/E over all considered measurements of 1.006 (σ = 0.016) for pressurized water reactor and 0.984 (σ = 0.077) for boiling water reactor assembly measurements. The effect of using assembly-average versus axially varying modeling data on the calculated decay heat, important to thermal analyses for used fuel transportation and storage systems, is discussed.



中文翻译:

SCALE 6.2.4 轻水反应堆衰变热分析的验证

摘要

核燃料从反应堆排出后,放射性核素衰变释放的能量是废核燃料储存、运输和储存系统设计、安全和许可分析的关键参数。经过充分验证的计算工具和核数据对于衰变热预测至关重要。本文总结了 SCALE 核分析代码系统版本 6.2.4 的验证,与 ENDF/B-VII.1 库一起使用,用于轻水堆使用燃料的衰变热分析。用于验证的实验数据包括全组件衰变热测量,涵盖 5 至 51 GWd/吨 U 的组件燃耗、2 至 27 年范围内排放后的冷却时间以及高达 4 wt% 的初始燃料浓缩235U. 计算的 (C) 和实验 (E) 衰变热之间的比较显示出非常好的一致性,压水反应堆的所有考虑测量值的平均 C/E 为 1.006 (σ = 0.016),对于压水反应堆的平均 C/E 为 0.984 (σ = 0.077)沸水反应堆组件测量。讨论了使用装配平均与轴向变化的建模数据对计算的衰变热的影响,这对于使用过的燃料运输和储存系统的热分析很重要。

更新日期:2021-08-18
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