...
  • 主题
高级搜索  >
历史搜索 清空历史
首页> 外文期刊>EPJ Nuclear Sciences & Technologies >Estimation of the radionuclide inventory in LWR spent fuel assembly structural materials for long-term safety analysis
【24h】

Estimation of the radionuclide inventory in LWR spent fuel assembly structural materials for long-term safety analysis

机译：估算轻水堆乏燃料组件结构材料中的放射性核素库存以进行长期安全分析

获取原文
           
页面导航
  • 摘要
  • 著录项
  • 相似文献
  • 相关主题

摘要

The radionuclide inventory of materials irradiated in a reactor depends on the initial material composition, irradiation history and on the magnitude and spectrum of the neutron flux. The material composition of a fuel assembly structure includes various alloys of Zircaloy, Inconel and stainless steel. The existing impurities in these materials are very important for accurate determination of the activation of all nuclides with a view to assessing the radiological consequences of their geological disposal. In fact, the safety assessments of geological repositories require the average and maximum (in the sense of very conservative) inventories of the very long-lived nuclides as input. The purpose of the present work is to describe the methodology applied for determining the activation of these nuclides in fuel assembly structural materials by means of coupled depletion/activation calculations and also to crosscheck the results obtained from two approaches. UO2 and MOX PWR fuels have been simulated using SCALE/TRITON, simultaneously irradiating the fuel region in POWER mode and the cladding region in FLUX mode and aiming to produce binary macro cross-section libraries by applying accurate local neutron spectra in the cladding region as a function of irradiation history that are suitable for activation calculations. The developed activation libraries have been re-employed in a second run using the ORIGEN-S program for a dedicated activation calculation. The axial variation of the neutron flux along the fuel assembly length has also been considered. The SCALE calculations were performed using a 238-group cross-section library, according to the ENDF/B-VII. The results obtained with the ORIGEN-S activation calculations are compared with the results obtained from TRITON via direct irradiation of the cladding, as allowed by the FLUX mode. It is shown that an agreement on the total calculated activities can be found within 55% for MOX and within 22% for UO2, whereas the latter is reduced to 9% when more accurate irradiation data are used (core-follow flux data instead of life-average flux data).
机译：在反应堆中辐照的放射性核素物质清单取决于初始物质组成，辐照历史以及中子通量的大小和光谱。燃料组件结构的材料成分包括Zircaloy，Inconel和不锈钢的各种合金。这些材料中存在的杂质对于准确确定所有核素的活化非常重要，以评估其地质处置的放射学后果。实际上，对地质储藏库的安全评估需要输入非常长寿的核素的平均和最大（在非常保守的意义上）库存。本工作的目的是描述用于通过耦合耗竭/活化计算确定燃料组件结构材料中这些核素活化的方法，并交叉核对从两种方法获得的结果。 UO 2 和MOX PWR燃料已使用SCALE / TRITON进行了模拟，同时以POWER模式照射燃料区域和以FLUX模式照射包层区域，并旨在通过应用精确局部来生成二进制宏截面库包层区域中的中子光谱是辐照历史的函数，适合激活计算。在第二次运行中，已使用ORIGEN-S程序重新开发了开发的激活库，以进行专门的激活计算。还考虑了中子通量沿燃料组件长度的轴向变化。根据ENDF / B-VII，使用238组截面库进行SCALE计算。在FLUX模式允许的情况下，将通过ORIGEN-S激活计算获得的结果与通过直接照射包层从TRITON获得的结果进行比较。结果表明，对于MOX，在总计算活性上的一致性可以在55％之内，而对于UO 2 ，则可以在22％之内;而当使用更精确的辐照数据时，后者减少到9％（核心跟踪磁通数据，而不是寿命平均磁通数据）。

著录项

相似文献

  • 外文文献
  • 中文文献
  • 专利
获取原文

客服邮箱：kefu@zhangqiaokeyan.com

京公网安备：11010802029741号 ICP备案号：京ICP备15016152号-6 六维联合信息科技 (北京) 有限公司©版权所有

  • 客服微信

  • 服务号