The purpose of the paper is to study the performance of the CANDU(Canada Deuterium Uranium)reactor when the reactor core is loaded with thorium fuel mixed with plutonium isotopes with ratio 3 and 5%.A three dimensiona...The purpose of the paper is to study the performance of the CANDU(Canada Deuterium Uranium)reactor when the reactor core is loaded with thorium fuel mixed with plutonium isotopes with ratio 3 and 5%.A three dimensional model is designed for the core of CANDU reactor.The computer code MCNPX(Monte Carlo N–Particle Transport)is used to calculate the processes in its core.The results are compared with natural UO2 case which is the typical fuel of the reactor.The results show that the multiplication factor of the reactor is higher even in the case of thorium fuel mixed with 3%plutonium isotopes,which indicates longer neutron life cycle length and more economic utilization of the reactor.展开更多
The molten salt reactor(MSR), as one of the Generation Ⅳ advanced nuclear systems, has attracted a worldwide interest due to its excellent performances in safety, economics, sustainability, and proliferation resistan...The molten salt reactor(MSR), as one of the Generation Ⅳ advanced nuclear systems, has attracted a worldwide interest due to its excellent performances in safety, economics, sustainability, and proliferation resistance. The aim of this work is to provide and evaluate possible solutions to fissile 233 U production and further the fuel transition to thorium fuel cycle in a thermal MSR by using plutonium partitioned from light water reactors spent fuel. By using an in-house developed tool, a breeding and burning(B&B) scenario is first introduced and analyzed from the aspects of the evolution of main nuclides, net 233 U production, spectrum shift, and temperature feedback coefficient. It can be concluded that such a Th/Pu to Th/^(233)U transition can be accomplished by employing a relatively fast fuel reprocessing with a cycle time less than 60 days. At the equilibrium state, the reactor can achieve a conversion ratio of about 0.996 for the 60-day reprocessing period(RP) case and about 1.047 for the 10-day RP case.The results also show that it is difficult to accomplish such a fuel transition with limited reprocessing(RP is 180 days),and the reactor operates as a converter and burns the plutonium with the help of thorium. Meanwhile, a prebreeding and burning(PB&B) scenario is also analyzed briefly with respect to the net 233 U production and evolution of main nuclides. One can find that it is more efficient to produce 233 U under this scenario, resulting in a double time varying from about 1.96 years for the 10-day RP case to about 6.15 years for the 180-day RP case.展开更多
This work presents the results of computer simulation of neutronic processes in a high-temperature gas-cooled thorium reactor for 30 different options of core loading.To guarantee stable and long-term reactor operatio...This work presents the results of computer simulation of neutronic processes in a high-temperature gas-cooled thorium reactor for 30 different options of core loading.To guarantee stable and long-term reactor operation(7-10 years),the quantity of fuel compact dispersion phase and starting fuel composition was selected.It is demonstrated that it is possible in principle to substitute the near-axial recirculation zone of the reactor core by a long magnetic trap with a high-temperature plasma column for generating thermonuclear neutrons.The distribution of neutron yield along the length of the plasma source is also presented.Such a thorium reactor,with a near-axial source of extra neutrons,can be applied for researching thermophysical and neutronic characteristics of dispersion thorium fuel to improve its properties.The results of the work are of great interest from the perspective of future advancement of the thermonuclear power industry,by means of creation of a hybrid installation based on a thorium reactor with a long plasma column as a source of additional neutrons.展开更多
Study on the behavior of thorium based fuel in a fuel bundle is the aim of this Simulation.check the spectrum flux in theoretical sample Shown that(Th,U)and(Th,Pu)cycle can work in one fuel bundle.
The molten salt fast reactor(MSFR) shows great promise with high breeding ratio(BR),large negative temperature coefficient of reactivity,high thermal-electric conversion efficiency,inherent safety,and online reprocess...The molten salt fast reactor(MSFR) shows great promise with high breeding ratio(BR),large negative temperature coefficient of reactivity,high thermal-electric conversion efficiency,inherent safety,and online reprocessing.Based on an improved MSFR optimized by adding axial fertile salt and a graphite reflector,the influences of ~7Li enrichment on Th-U breeding are investigated,aiming to provide a feasible selection for the molten salt with high fissile breeding and a relatively low technology requirement for ~7Li concentration.With the self-developed molten salt reactor reprocessing sequence based on SCALE6.1,the burn-up calculations with online reprocessing are carried out.Parameters are explored including BR,^(233)U production,double time(DT),spectrum,~6Li inventory,neutron absorption,and the tritium production.The results show that the Li enrichment of 99.95% is appropriate in the fast fission reactor.In this case,BR above 1.10 can be achieved for a long time,corresponding to the ^(233)U production of130 kg per year and DT of 36 years.After 80 years' operation,the tritium production for 99.5% is only about 7kg,and there is no obvious increase compared to that for 99.9995%.展开更多
为了促进激光诱导击穿光谱技术在核工业领域中的应用与发展,利用飞秒激光对高纯石墨中的钍(Th)元素开展了定量分析研究。采用标准加样法制备了钍含量在0.35%~35.15%范围内的9个分析样品,以类比钍基核燃料中的钍含量。通过改变光谱采集...为了促进激光诱导击穿光谱技术在核工业领域中的应用与发展,利用飞秒激光对高纯石墨中的钍(Th)元素开展了定量分析研究。采用标准加样法制备了钍含量在0.35%~35.15%范围内的9个分析样品,以类比钍基核燃料中的钍含量。通过改变光谱采集方式、延时条件及调节飞秒激光脉冲能量对实验条件进行优化。在优化的实验条件下,对所有样品进行激发以采集等离子体光谱信息用于定量分析研究。得出以下结果:对比定点激发采集光谱结果,采用靶面连续移动式的光谱重复性好,钍原子(Th I 396.21 nm)谱线强度获得大约2倍的增强,重复测量的相对标准偏差由20.4%降至5.7%;高含量区间内钍元素谱线存在明显的自吸收效应,采用指数函数对整个含量区间与分析线(Th I 394.42 nm、396.21 nm和766.53 nm)强度进行非线性拟合,可以有效获取分析线的饱和阈值;基本定标法适用于饱和阈值以下的含量区间,分析线对较低含量的未知样品的预测分析具有较高的精确度;采用内标法(以C I 247.85 nm线为内标线),可以实现积分强度和峰值强度与整个区间含量的线性拟合,其中,基于高饱和阈值分析线(766.53 nm)的积分强度能够较好地实现高含量未知样品的含量预测。实验结果说明:飞秒激光诱导击穿光谱技术具有钍基核燃料循环过程中钍含量监测分析的潜力。展开更多
文摘The purpose of the paper is to study the performance of the CANDU(Canada Deuterium Uranium)reactor when the reactor core is loaded with thorium fuel mixed with plutonium isotopes with ratio 3 and 5%.A three dimensional model is designed for the core of CANDU reactor.The computer code MCNPX(Monte Carlo N–Particle Transport)is used to calculate the processes in its core.The results are compared with natural UO2 case which is the typical fuel of the reactor.The results show that the multiplication factor of the reactor is higher even in the case of thorium fuel mixed with 3%plutonium isotopes,which indicates longer neutron life cycle length and more economic utilization of the reactor.
基金supported by the Chinese TMSR Strategic Pioneer Science and Technology Project(No.XDA02010000)the National Natural Science Foundation of China(No.91326201)the Frontier Science Key Program of the Chinese Academy of Sciences(No.QYZDY-SSW-JSC016)
文摘The molten salt reactor(MSR), as one of the Generation Ⅳ advanced nuclear systems, has attracted a worldwide interest due to its excellent performances in safety, economics, sustainability, and proliferation resistance. The aim of this work is to provide and evaluate possible solutions to fissile 233 U production and further the fuel transition to thorium fuel cycle in a thermal MSR by using plutonium partitioned from light water reactors spent fuel. By using an in-house developed tool, a breeding and burning(B&B) scenario is first introduced and analyzed from the aspects of the evolution of main nuclides, net 233 U production, spectrum shift, and temperature feedback coefficient. It can be concluded that such a Th/Pu to Th/^(233)U transition can be accomplished by employing a relatively fast fuel reprocessing with a cycle time less than 60 days. At the equilibrium state, the reactor can achieve a conversion ratio of about 0.996 for the 60-day reprocessing period(RP) case and about 1.047 for the 10-day RP case.The results also show that it is difficult to accomplish such a fuel transition with limited reprocessing(RP is 180 days),and the reactor operates as a converter and burns the plutonium with the help of thorium. Meanwhile, a prebreeding and burning(PB&B) scenario is also analyzed briefly with respect to the net 233 U production and evolution of main nuclides. One can find that it is more efficient to produce 233 U under this scenario, resulting in a double time varying from about 1.96 years for the 10-day RP case to about 6.15 years for the 180-day RP case.
文摘This work presents the results of computer simulation of neutronic processes in a high-temperature gas-cooled thorium reactor for 30 different options of core loading.To guarantee stable and long-term reactor operation(7-10 years),the quantity of fuel compact dispersion phase and starting fuel composition was selected.It is demonstrated that it is possible in principle to substitute the near-axial recirculation zone of the reactor core by a long magnetic trap with a high-temperature plasma column for generating thermonuclear neutrons.The distribution of neutron yield along the length of the plasma source is also presented.Such a thorium reactor,with a near-axial source of extra neutrons,can be applied for researching thermophysical and neutronic characteristics of dispersion thorium fuel to improve its properties.The results of the work are of great interest from the perspective of future advancement of the thermonuclear power industry,by means of creation of a hybrid installation based on a thorium reactor with a long plasma column as a source of additional neutrons.
文摘Study on the behavior of thorium based fuel in a fuel bundle is the aim of this Simulation.check the spectrum flux in theoretical sample Shown that(Th,U)and(Th,Pu)cycle can work in one fuel bundle.
基金supported by the Chinese TMSR Strategic Pioneer Science and Technology Project(No.XDA02010000)the National Natural Science Foundation of China(No.91326201)the Frontier Science Key Program of the Chinese Academy of Sciences(No.QYZDY-SSW-JSC016)
文摘The molten salt fast reactor(MSFR) shows great promise with high breeding ratio(BR),large negative temperature coefficient of reactivity,high thermal-electric conversion efficiency,inherent safety,and online reprocessing.Based on an improved MSFR optimized by adding axial fertile salt and a graphite reflector,the influences of ~7Li enrichment on Th-U breeding are investigated,aiming to provide a feasible selection for the molten salt with high fissile breeding and a relatively low technology requirement for ~7Li concentration.With the self-developed molten salt reactor reprocessing sequence based on SCALE6.1,the burn-up calculations with online reprocessing are carried out.Parameters are explored including BR,^(233)U production,double time(DT),spectrum,~6Li inventory,neutron absorption,and the tritium production.The results show that the Li enrichment of 99.95% is appropriate in the fast fission reactor.In this case,BR above 1.10 can be achieved for a long time,corresponding to the ^(233)U production of130 kg per year and DT of 36 years.After 80 years' operation,the tritium production for 99.5% is only about 7kg,and there is no obvious increase compared to that for 99.9995%.
文摘为了促进激光诱导击穿光谱技术在核工业领域中的应用与发展,利用飞秒激光对高纯石墨中的钍(Th)元素开展了定量分析研究。采用标准加样法制备了钍含量在0.35%~35.15%范围内的9个分析样品,以类比钍基核燃料中的钍含量。通过改变光谱采集方式、延时条件及调节飞秒激光脉冲能量对实验条件进行优化。在优化的实验条件下,对所有样品进行激发以采集等离子体光谱信息用于定量分析研究。得出以下结果:对比定点激发采集光谱结果,采用靶面连续移动式的光谱重复性好,钍原子(Th I 396.21 nm)谱线强度获得大约2倍的增强,重复测量的相对标准偏差由20.4%降至5.7%;高含量区间内钍元素谱线存在明显的自吸收效应,采用指数函数对整个含量区间与分析线(Th I 394.42 nm、396.21 nm和766.53 nm)强度进行非线性拟合,可以有效获取分析线的饱和阈值;基本定标法适用于饱和阈值以下的含量区间,分析线对较低含量的未知样品的预测分析具有较高的精确度;采用内标法(以C I 247.85 nm线为内标线),可以实现积分强度和峰值强度与整个区间含量的线性拟合,其中,基于高饱和阈值分析线(766.53 nm)的积分强度能够较好地实现高含量未知样品的含量预测。实验结果说明:飞秒激光诱导击穿光谱技术具有钍基核燃料循环过程中钍含量监测分析的潜力。
