The advantages of once-through molten salt reactors include readily available fuel,low nuclear proliferation risk,and low technical difficulty.It is potentially the most easily commercialized fuel cycle mode for molte...The advantages of once-through molten salt reactors include readily available fuel,low nuclear proliferation risk,and low technical difficulty.It is potentially the most easily commercialized fuel cycle mode for molten salt reactors.However,there are some problems in the parameter selection of once-through molten salt reactors,and the relevant burnup optimization work requires further analysis.This study examined once-through graphitemoderated molten salt reactor using enriched uranium and thorium.The fuel volume fraction(VF),initial heavy nuclei concentration(HN_(0)),feeding uranium enrichment(E_(FU)),volume of the reactor core,and fuel type were changed to obtain the optimal conditions for burnup.We found an optimal region for VF and HN_(0) in each scheme,and the location and size of the optimal region changed with the degree of E_(FU),core volume,and fuel type.The recommended core schemes provide a reference for the core design of a once-through molten salt reactor.展开更多
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
与氟盐堆相比,氯盐快堆具有超铀核素(Transuranics,TRU)溶解度更高、中子能谱更硬、熔点更低等方面的优势。基于熔盐嬗变堆(Molten Salt Actinide Recycler and Transmuter,MOSART)的堆芯结构,采用熔盐堆在线添料和后处理程序MSR-RS(Mol...与氟盐堆相比,氯盐快堆具有超铀核素(Transuranics,TRU)溶解度更高、中子能谱更硬、熔点更低等方面的优势。基于熔盐嬗变堆(Molten Salt Actinide Recycler and Transmuter,MOSART)的堆芯结构,采用熔盐堆在线添料和后处理程序MSR-RS(Molten Salt Reactor Reprocessing Sequence)进行分析,针对氯盐快堆的熔盐组成、后处理方式等方面进行了优化,以利于提升其增殖及嬗变性能。首先分析了不同载体盐和启动燃料对燃耗性能的影响,提出了熔盐成分优化方案;然后引入离线批处理和在线连续处理两种后方式来提升燃耗性能。结果表明:在氯盐快堆中,高重金属溶解度的Na Cl更适合作为载体盐;TRU中的次锕系核素(Minor Actinides,MA)有助于提升增殖性能;采用离线批处理能够达到较好的燃耗性能,降低对后处理系统的要求。优化后的堆芯燃耗时间延长到31 a,相应的燃耗深度提高至210 GW·d·t^(-1)左右,233U的积累量达到8 300 kg,并且最终消耗了约12 000 kg的TRU,嬗变率为62.1%。展开更多
基于熔盐嬗变堆(Molten Salt Actinide Recycler and Transmuter,简称MOSART)堆芯结构对氯盐快堆(Molten Chloride Salt Fast Reactor,简称MCFR)进行了优化,分析了熔盐成分和后处理方式的影响,使其燃耗性能得到明显的提升,但是相比熔盐...基于熔盐嬗变堆(Molten Salt Actinide Recycler and Transmuter,简称MOSART)堆芯结构对氯盐快堆(Molten Chloride Salt Fast Reactor,简称MCFR)进行了优化,分析了熔盐成分和后处理方式的影响,使其燃耗性能得到明显的提升,但是相比熔盐快堆(Molten Salt Fast Reactor,简称MSFR)的增殖及嬗变性能仍有一定差距。基于在线连续添料与后处理方式,采用SCALE6.1程序和熔盐堆在线添料和后处理程序(Molten Salt Reactor Reprocessing Sequence,简称MSR-RS)分析了堆芯结构、^(37)Cl富集度对增殖比(Breeding Ratio,简称BR)、核素吸收率、燃耗等方面的影响,提出了双区氯盐快堆的设计,进一步提升了增殖嬗变性能和钍基燃料的利用率,倍增时间缩短到20年左右,超铀核素(Transuranics,简称TRU)嬗变率达到68%左右。展开更多
采用液态燃料及重水慢化剂的重水慢化熔盐堆(Heavy Water moderated Molten Salt Reactor,HWMSR)具有高中子经济性,但堆芯出口温差较大,将会导致堆芯顶部管道构件热疲劳。本文旨在优化HWMSR堆芯设计,降低堆芯出口温差。采用中子学-热工...采用液态燃料及重水慢化剂的重水慢化熔盐堆(Heavy Water moderated Molten Salt Reactor,HWMSR)具有高中子经济性,但堆芯出口温差较大,将会导致堆芯顶部管道构件热疲劳。本文旨在优化HWMSR堆芯设计,降低堆芯出口温差。采用中子学-热工耦合程序以及堆芯临界搜索程序,深入分析了具有不同熔盐通道半径堆芯的功率分布、熔盐出口温度分布、初始易裂变核素233U装载量及钍铀增殖等性能。结果表明:增大堆芯内区熔盐通道尺寸将降低燃料熔盐功率密度峰值及最大出口温度,而对钍铀增殖比及^(233)U初始装载量影响非常有限。本研究为优化重水慢化熔盐堆堆芯设计提供参考。展开更多
India is one of the few countries committed to expansion of nuclear power. In view of the abundance of thorium relative to uranium, thorium cycle is under serious development and implementation. Both ThO2 and (U,Th)O2...India is one of the few countries committed to expansion of nuclear power. In view of the abundance of thorium relative to uranium, thorium cycle is under serious development and implementation. Both ThO2 and (U,Th)O2 are used. Fine powders of the same are mostly prepared through the aqueous chemical route, pressed and sintered. Extrusion and hot impact densification are also being used. Sol-gel method and other alternatives are also being pursued with the advantage of automation and remote operation. Relevant papers on the thorium cycle with emphasis on processing methods and related aspects are reviewed here.展开更多
基金supported by the Shanghai Sailing Program(No.19YF1457900)Chinese TMSR Strategic Pioneer Science and Technology Project(No.XDA02010000)+1 种基金National Natural Science Foundation of China(No.12005290)Youth Innovation Promotion Association of the Chinese Academy of Sciences(No.2020261)。
文摘The advantages of once-through molten salt reactors include readily available fuel,low nuclear proliferation risk,and low technical difficulty.It is potentially the most easily commercialized fuel cycle mode for molten salt reactors.However,there are some problems in the parameter selection of once-through molten salt reactors,and the relevant burnup optimization work requires further analysis.This study examined once-through graphitemoderated molten salt reactor using enriched uranium and thorium.The fuel volume fraction(VF),initial heavy nuclei concentration(HN_(0)),feeding uranium enrichment(E_(FU)),volume of the reactor core,and fuel type were changed to obtain the optimal conditions for burnup.We found an optimal region for VF and HN_(0) in each scheme,and the location and size of the optimal region changed with the degree of E_(FU),core volume,and fuel type.The recommended core schemes provide a reference for the core design of a once-through molten salt reactor.
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.
文摘基于熔盐嬗变堆(Molten Salt Actinide Recycler and Transmuter,简称MOSART)堆芯结构对氯盐快堆(Molten Chloride Salt Fast Reactor,简称MCFR)进行了优化,分析了熔盐成分和后处理方式的影响,使其燃耗性能得到明显的提升,但是相比熔盐快堆(Molten Salt Fast Reactor,简称MSFR)的增殖及嬗变性能仍有一定差距。基于在线连续添料与后处理方式,采用SCALE6.1程序和熔盐堆在线添料和后处理程序(Molten Salt Reactor Reprocessing Sequence,简称MSR-RS)分析了堆芯结构、^(37)Cl富集度对增殖比(Breeding Ratio,简称BR)、核素吸收率、燃耗等方面的影响,提出了双区氯盐快堆的设计,进一步提升了增殖嬗变性能和钍基燃料的利用率,倍增时间缩短到20年左右,超铀核素(Transuranics,简称TRU)嬗变率达到68%左右。
文摘采用液态燃料及重水慢化剂的重水慢化熔盐堆(Heavy Water moderated Molten Salt Reactor,HWMSR)具有高中子经济性,但堆芯出口温差较大,将会导致堆芯顶部管道构件热疲劳。本文旨在优化HWMSR堆芯设计,降低堆芯出口温差。采用中子学-热工耦合程序以及堆芯临界搜索程序,深入分析了具有不同熔盐通道半径堆芯的功率分布、熔盐出口温度分布、初始易裂变核素233U装载量及钍铀增殖等性能。结果表明:增大堆芯内区熔盐通道尺寸将降低燃料熔盐功率密度峰值及最大出口温度,而对钍铀增殖比及^(233)U初始装载量影响非常有限。本研究为优化重水慢化熔盐堆堆芯设计提供参考。
文摘India is one of the few countries committed to expansion of nuclear power. In view of the abundance of thorium relative to uranium, thorium cycle is under serious development and implementation. Both ThO2 and (U,Th)O2 are used. Fine powders of the same are mostly prepared through the aqueous chemical route, pressed and sintered. Extrusion and hot impact densification are also being used. Sol-gel method and other alternatives are also being pursued with the advantage of automation and remote operation. Relevant papers on the thorium cycle with emphasis on processing methods and related aspects are reviewed here.
