基于Mathematica 7.0建立了熔盐堆(Molten Salt Reactor,MSR)主回路系统衰变热流动模型,并与参考程序ORIGENS在静态燃耗下的计算结果以及熔盐实验堆(Molten Salt Reactor Experiment,MSRE)衰变热结果进行了初步验证,相对偏差分别在±...基于Mathematica 7.0建立了熔盐堆(Molten Salt Reactor,MSR)主回路系统衰变热流动模型,并与参考程序ORIGENS在静态燃耗下的计算结果以及熔盐实验堆(Molten Salt Reactor Experiment,MSRE)衰变热结果进行了初步验证,相对偏差分别在±4%和±2.76%的范围内符合较好。对2 MW液态燃料钍基熔盐实验堆(Thorium Molten Salt Reactor-Liquid Fuel 1,TMSR-LF1)正常运行工况下主回路系统管道及设备内的衰变热分布进行了定量分析。结果表明:启堆达到满功率和设定流量后约20 s各区域衰变热快速积累,随后便开始平缓上升并趋平衡。平衡时堆芯活性区衰变热占总衰变热的46.7%,上腔室、热管段#1、主泵、热管段#2、换热器、冷管段及下腔室区域分别占比31.8%、1.21%、14.6%、0.89%、2.21%、1.67%和0.94%。所建立的分析方法及结论可为熔盐堆主回路系统的热工水力安全分析、余热排出系统设计、反应堆功率调节与安全控制提供重要参考。展开更多
Calculation of the decay heat from the decay/buildup of radionuclides generated after nuclear fission is one of the highest priorities in the nuclear industry. These calculations become more important if they are made...Calculation of the decay heat from the decay/buildup of radionuclides generated after nuclear fission is one of the highest priorities in the nuclear industry. These calculations become more important if they are made together with the analysis of the most important isotopes affecting the decay heat. They are useful in designing the necessary nuclear safety for spent fuels, and their importance cannot be overlooked in the designs of transporting fuel storage containers as well as in the management of the radioactive waste generated. In this paper, by using MATLAB, the decay heat after the thermal fission of a U-235 nucleus was numerically calculated by solving linear differential equations for all the buildups/decays of the fission products. Also, the most contribution of radioactive isotopes to the decay heat was analyzed by using Microsoft Excel. The most influential isotopes were deduced in two ways;either by calculating the most influential isotopes at specific times, or by determining the largest influences in a cumulative manner. All required nuclear data such as decay constants their branching ratios, independent fission yield, and average α-, β-, and γ-energies released per disintegration of any nuclide, have been extracted from the latest version of the Evaluated Nuclear Data Files (ENDF) database ENDF/B-VIII.0. The two different methods used showed a difference in the contributing isotopes, which is logical for the difference in the method of calculation. The first method is suitable for instantaneous data while the second method is more suitable when there is a need to know the cumulative calculations. In sum, we can say that both methods complement each other, and neither of them can be dispensed with in the accurate calculations related to transportation and storage of spent fuel.展开更多
文摘基于Mathematica 7.0建立了熔盐堆(Molten Salt Reactor,MSR)主回路系统衰变热流动模型,并与参考程序ORIGENS在静态燃耗下的计算结果以及熔盐实验堆(Molten Salt Reactor Experiment,MSRE)衰变热结果进行了初步验证,相对偏差分别在±4%和±2.76%的范围内符合较好。对2 MW液态燃料钍基熔盐实验堆(Thorium Molten Salt Reactor-Liquid Fuel 1,TMSR-LF1)正常运行工况下主回路系统管道及设备内的衰变热分布进行了定量分析。结果表明:启堆达到满功率和设定流量后约20 s各区域衰变热快速积累,随后便开始平缓上升并趋平衡。平衡时堆芯活性区衰变热占总衰变热的46.7%,上腔室、热管段#1、主泵、热管段#2、换热器、冷管段及下腔室区域分别占比31.8%、1.21%、14.6%、0.89%、2.21%、1.67%和0.94%。所建立的分析方法及结论可为熔盐堆主回路系统的热工水力安全分析、余热排出系统设计、反应堆功率调节与安全控制提供重要参考。
文摘Calculation of the decay heat from the decay/buildup of radionuclides generated after nuclear fission is one of the highest priorities in the nuclear industry. These calculations become more important if they are made together with the analysis of the most important isotopes affecting the decay heat. They are useful in designing the necessary nuclear safety for spent fuels, and their importance cannot be overlooked in the designs of transporting fuel storage containers as well as in the management of the radioactive waste generated. In this paper, by using MATLAB, the decay heat after the thermal fission of a U-235 nucleus was numerically calculated by solving linear differential equations for all the buildups/decays of the fission products. Also, the most contribution of radioactive isotopes to the decay heat was analyzed by using Microsoft Excel. The most influential isotopes were deduced in two ways;either by calculating the most influential isotopes at specific times, or by determining the largest influences in a cumulative manner. All required nuclear data such as decay constants their branching ratios, independent fission yield, and average α-, β-, and γ-energies released per disintegration of any nuclide, have been extracted from the latest version of the Evaluated Nuclear Data Files (ENDF) database ENDF/B-VIII.0. The two different methods used showed a difference in the contributing isotopes, which is logical for the difference in the method of calculation. The first method is suitable for instantaneous data while the second method is more suitable when there is a need to know the cumulative calculations. In sum, we can say that both methods complement each other, and neither of them can be dispensed with in the accurate calculations related to transportation and storage of spent fuel.
