A thermal–hydraulic model was developed to analyze the three-dimensional(3D)temperature field of a graphite-moderated channel-type molten salt reactor(GMC-MSR).This model solves the temperature distribution of both t...A thermal–hydraulic model was developed to analyze the three-dimensional(3D)temperature field of a graphite-moderated channel-type molten salt reactor(GMC-MSR).This model solves the temperature distribution of both the graphite moderator and fuel salt using a single convection–diffusion equation.Heat transfer at the interface between the fuel salt and graphite was addressed by introducing an additional thermal resistance component at the interface and modifying the anisotropic thermal conductivity of the fuel salt.The mass flow distribution in different flow passages was determined by adjusting the mass flow rate until a uniform pressure drop was achieved across all fuel channels.This thermal–hydraulic model,constructed on COMSOL Multiphysics,was verified by comparing its temperature results with those from the RELAP5 code across two demonstration cases.A steady-state thermal–hydraulic simulation of this model was performed to evaluate the conceptual design of a 2-MW experimental molten salt reactor(2MW-MSR).In addition,detailed discussions of the 3D temperature field,heat flux,and mass flow distribution of the 2MW-MSR were presented.This model allows for a comprehensive 3D thermal–hydraulic analysis of the GMC-MSR.Moreover,it only requires the solution of a single convection–diffusion equation,which makes it invaluable for GMC-MSR design.展开更多
A molten salt reactor(MSR)has outstanding features considering the application of thorium fuel,inherent safety,sustainability,and resistance to proliferation.However,fissile material^(233)U is significantly rare at th...A molten salt reactor(MSR)has outstanding features considering the application of thorium fuel,inherent safety,sustainability,and resistance to proliferation.However,fissile material^(233)U is significantly rare at the current stage,thus it is difficult for MSR to achieve a pure thorium-uranium fuel cycle.Therefore,using plutonium or enriched uranium as the initial fuel for MSR is more practical.In this study,we aim to verify the feasibility of a small modular MSR that utilizes plutonium as the starting fuel(SM-MSR-Pu),and highlight its advantages and disadvantages.First,the structural design and fuel management scheme of the SM-MSR-Pu were presented.Second,the neutronic characteristics,such as the graphite-irradiation lifetime,burn-up performance,and coefficient of temperature reactivity were calculated to analyze the physical characteristics of the SM-MSR-Pu.The results indicate that plutonium is a feasible and advantageous starting fuel for a SM-MSR;however,there are certain shortcomings that need to be solved.In a 250 MWth SM-MSR-Pu,approximately 288.64 kg^(233)U of plutonium with a purity of greater than 90% is produced while 978.00 kg is burned every ten years.The temperature reactivity coefficient decreases from -4.0 to -6.5 pcm K^(-1) over the 50-year operating time,which ensures a long-term safe operation.However,the amount of plutonium and accumulation of minor actinides(MAs)would increase as the burn-up time increases,and the annual production and purity of^(233)U will decrease.To achieve an optimal burn-up performance,setting the entire operation time to 30 years is advisable.Regardless,more than 3600 kg of plutonium eventually accumulate in the core.Further research is required to effectively utilize this accumulated plutonium.展开更多
The heavy water-moderated molten salt reactor(HWMSR)is a newly proposed reactor concept,in which heavy water is adopted as the moderator and molten salt dissolved with fissile and fertile elements is used as the fuel....The heavy water-moderated molten salt reactor(HWMSR)is a newly proposed reactor concept,in which heavy water is adopted as the moderator and molten salt dissolved with fissile and fertile elements is used as the fuel.Issues arising from graphite in traditional molten salt reactors,including the positive temperature coefficient and management of highly radio-active spent graphite waste,can be addressed using the HWMSR.Until now,research on the HWMSR has been centered on the core design and nuclear fuel cycle to explore the viability of the HWMSR and its advantages in fuel utilization.However,the core safety of the HWMSR has not been extensively studied.Therefore,we evaluate typical accidents in a small modular HWMSR,including fuel salt inlet temperature overcooling and overheating accidents,fuel salt inlet flow rate decrease,heavy water inlet temperature overcooling accidents,and heavy water inlet mass flow rate decrease accidents,based on a neutronics and thermal-hydraulics coupled code.The results demonstrated that the core maintained safety during the investigated accidents.展开更多
Following publication of the original article,the authors observed that both Fig.5 and Fig.4 depict the same image.Figure 5 was inaccurately referenced and displayed.The correct Fig.5 is copied below:The original arti...Following publication of the original article,the authors observed that both Fig.5 and Fig.4 depict the same image.Figure 5 was inaccurately referenced and displayed.The correct Fig.5 is copied below:The original article has been updated.展开更多
钍基熔盐堆(TMSR)核能系统项目是中科院未来10年先导研究专项之一,其研究目标是研发第四代裂变反应堆核能系统,计划至2020年之前建成2MW钍基熔盐实验堆,形成支撑未来TMSR核能系统发展的若干技术研发能力,并解决钍铀燃料循环和钍基熔盐...钍基熔盐堆(TMSR)核能系统项目是中科院未来10年先导研究专项之一,其研究目标是研发第四代裂变反应堆核能系统,计划至2020年之前建成2MW钍基熔盐实验堆,形成支撑未来TMSR核能系统发展的若干技术研发能力,并解决钍铀燃料循环和钍基熔盐堆相关重大技术挑战,研制出工业示范级钍基熔盐堆,实现钍资源的有效使用和核能的综合利用。钍基核燃料具有232Th/233U转换效率高、在热中子堆中也能增殖、产生较少的高毒性放射性核素、有利于防核扩散等优点,但也面临燃料制备困难、232U衰变子核的强γ辐射给乏燃料处理和燃料再加工带来的困难、钍铀转换反应链中间核233Pa会吸收堆内中子从而影响233U产量。核燃料利用的工作模式有开环模式、改进的开环模式和闭环模式。熔盐堆是第四代反应堆的6个候选堆型之一,非常适合用作钍铀燃料循环,熔盐堆加上干法在线分离技术有可能实现完全的钍铀燃料闭式循环。本世纪初提出的氟盐冷却高温堆(Fluoride salt-cooled High temperature Reactors,FHRs),用氟化熔盐作为冷却剂,采用TRISO燃料颗粒作为核燃料,其中球床型氟盐冷却高温堆可以在改进的开环模式实现钍铀燃料循环。熔盐堆良好的高温特性使其成为核能非电应用主要候选者之一,反应堆产生的高温热可直接用于页岩油开采和高温制氢等工业领域。展开更多
作为四代堆6种候选堆型中唯一的液态燃料反应堆,熔盐堆对未来核能和钍资源利用具有重要意义,特别是熔盐快堆(Molten Salt Fast Reactor,MSFR)还具有较大的增殖比和较好的温度负反馈。由于启动新的熔盐快堆需要较高的燃料装载量,若能改善...作为四代堆6种候选堆型中唯一的液态燃料反应堆,熔盐堆对未来核能和钍资源利用具有重要意义,特别是熔盐快堆(Molten Salt Fast Reactor,MSFR)还具有较大的增殖比和较好的温度负反馈。由于启动新的熔盐快堆需要较高的燃料装载量,若能改善MSFR的增殖性能,则有利于提高233U产量并缩短燃料倍增时间。首先应用SCALE6.1针对MSFR的径向增殖盐、新增轴向增殖盐和新增石墨反射层这三方面分析了初始增殖比,同时从核素吸收率角度说明增殖比变化的原因和MSFR的设计不足并对其进行了优化;然后应用基于SCALE6.1开发的熔盐堆在线处理模块(Molten Salt Reactor Reprocessing Sequence,MSR-RS)进行燃耗分析。结果表明,新增轴向增殖盐可以进一步提高增殖性能;新增石墨反射层可以节省增殖盐装载量。改进后的堆型运行时增殖比可以维持在1.1以上,233U年产量提高至133 kg,倍增时间缩短至36 a,并且堆芯在整个运行寿期都能保持足够的温度负反馈。展开更多
数据存档系统是控制棒驱动机构(Control Rod Drive Mechanism,CRDM)样机控制系统的重要组成部分,主要用于存储控制棒的实时、报警信息及其他设备信息等数据,一方面能够使运行人员对熔盐堆CRDM样机的运行、调试和维修等工况进行分析和处...数据存档系统是控制棒驱动机构(Control Rod Drive Mechanism,CRDM)样机控制系统的重要组成部分,主要用于存储控制棒的实时、报警信息及其他设备信息等数据,一方面能够使运行人员对熔盐堆CRDM样机的运行、调试和维修等工况进行分析和处理,另一方面为今后基于数据分析的预警和诊断技术发展积累数据资源。本文基于开源、数据类型和接口丰富、扩展功能强大的关系型数据库Postgre SQL,设计实现了熔盐堆CRDM样机数据存档系统的数据库。为提高数据库的性能并确保安全可靠的存储数据,在数据采集上采用了双机热备技术,同时使用Java语言开发了控制棒棒位检索工具并连接数据存档系统。该系统在熔盐堆CRDM样机中的成功应用证明其是一种先进的、低成本的、稳定的数据存档系统。展开更多
基金supported by the National Natural Science Foundation of China(No.12075169)。
文摘A thermal–hydraulic model was developed to analyze the three-dimensional(3D)temperature field of a graphite-moderated channel-type molten salt reactor(GMC-MSR).This model solves the temperature distribution of both the graphite moderator and fuel salt using a single convection–diffusion equation.Heat transfer at the interface between the fuel salt and graphite was addressed by introducing an additional thermal resistance component at the interface and modifying the anisotropic thermal conductivity of the fuel salt.The mass flow distribution in different flow passages was determined by adjusting the mass flow rate until a uniform pressure drop was achieved across all fuel channels.This thermal–hydraulic model,constructed on COMSOL Multiphysics,was verified by comparing its temperature results with those from the RELAP5 code across two demonstration cases.A steady-state thermal–hydraulic simulation of this model was performed to evaluate the conceptual design of a 2-MW experimental molten salt reactor(2MW-MSR).In addition,detailed discussions of the 3D temperature field,heat flux,and mass flow distribution of the 2MW-MSR were presented.This model allows for a comprehensive 3D thermal–hydraulic analysis of the GMC-MSR.Moreover,it only requires the solution of a single convection–diffusion equation,which makes it invaluable for GMC-MSR design.
基金supported by the Chinese TMSR Strategic Pioneer Science and Technology Project(No.XDA02010000)Chinese Academy of Sciences Talent Introduction Youth Program(No.SINAP-YCJH-202303)Chinese Academy of Sciences Special Research Assistant Funding Project and Shanghai Pilot Program for Basic Research-Chinese Academy of Science,Shanghai Branch(JCYJ-SHFY-2021-003)。
文摘A molten salt reactor(MSR)has outstanding features considering the application of thorium fuel,inherent safety,sustainability,and resistance to proliferation.However,fissile material^(233)U is significantly rare at the current stage,thus it is difficult for MSR to achieve a pure thorium-uranium fuel cycle.Therefore,using plutonium or enriched uranium as the initial fuel for MSR is more practical.In this study,we aim to verify the feasibility of a small modular MSR that utilizes plutonium as the starting fuel(SM-MSR-Pu),and highlight its advantages and disadvantages.First,the structural design and fuel management scheme of the SM-MSR-Pu were presented.Second,the neutronic characteristics,such as the graphite-irradiation lifetime,burn-up performance,and coefficient of temperature reactivity were calculated to analyze the physical characteristics of the SM-MSR-Pu.The results indicate that plutonium is a feasible and advantageous starting fuel for a SM-MSR;however,there are certain shortcomings that need to be solved.In a 250 MWth SM-MSR-Pu,approximately 288.64 kg^(233)U of plutonium with a purity of greater than 90% is produced while 978.00 kg is burned every ten years.The temperature reactivity coefficient decreases from -4.0 to -6.5 pcm K^(-1) over the 50-year operating time,which ensures a long-term safe operation.However,the amount of plutonium and accumulation of minor actinides(MAs)would increase as the burn-up time increases,and the annual production and purity of^(233)U will decrease.To achieve an optimal burn-up performance,setting the entire operation time to 30 years is advisable.Regardless,more than 3600 kg of plutonium eventually accumulate in the core.Further research is required to effectively utilize this accumulated plutonium.
基金the National Natural Science Foundation of China(No.11905285)the Shanghai Natural Science Foundation(No.20ZR1468700)the Youth Innovation Promotion Association CAS(No.2022258).
文摘The heavy water-moderated molten salt reactor(HWMSR)is a newly proposed reactor concept,in which heavy water is adopted as the moderator and molten salt dissolved with fissile and fertile elements is used as the fuel.Issues arising from graphite in traditional molten salt reactors,including the positive temperature coefficient and management of highly radio-active spent graphite waste,can be addressed using the HWMSR.Until now,research on the HWMSR has been centered on the core design and nuclear fuel cycle to explore the viability of the HWMSR and its advantages in fuel utilization.However,the core safety of the HWMSR has not been extensively studied.Therefore,we evaluate typical accidents in a small modular HWMSR,including fuel salt inlet temperature overcooling and overheating accidents,fuel salt inlet flow rate decrease,heavy water inlet temperature overcooling accidents,and heavy water inlet mass flow rate decrease accidents,based on a neutronics and thermal-hydraulics coupled code.The results demonstrated that the core maintained safety during the investigated accidents.
文摘Following publication of the original article,the authors observed that both Fig.5 and Fig.4 depict the same image.Figure 5 was inaccurately referenced and displayed.The correct Fig.5 is copied below:The original article has been updated.
文摘钍基熔盐堆(TMSR)核能系统项目是中科院未来10年先导研究专项之一,其研究目标是研发第四代裂变反应堆核能系统,计划至2020年之前建成2MW钍基熔盐实验堆,形成支撑未来TMSR核能系统发展的若干技术研发能力,并解决钍铀燃料循环和钍基熔盐堆相关重大技术挑战,研制出工业示范级钍基熔盐堆,实现钍资源的有效使用和核能的综合利用。钍基核燃料具有232Th/233U转换效率高、在热中子堆中也能增殖、产生较少的高毒性放射性核素、有利于防核扩散等优点,但也面临燃料制备困难、232U衰变子核的强γ辐射给乏燃料处理和燃料再加工带来的困难、钍铀转换反应链中间核233Pa会吸收堆内中子从而影响233U产量。核燃料利用的工作模式有开环模式、改进的开环模式和闭环模式。熔盐堆是第四代反应堆的6个候选堆型之一,非常适合用作钍铀燃料循环,熔盐堆加上干法在线分离技术有可能实现完全的钍铀燃料闭式循环。本世纪初提出的氟盐冷却高温堆(Fluoride salt-cooled High temperature Reactors,FHRs),用氟化熔盐作为冷却剂,采用TRISO燃料颗粒作为核燃料,其中球床型氟盐冷却高温堆可以在改进的开环模式实现钍铀燃料循环。熔盐堆良好的高温特性使其成为核能非电应用主要候选者之一,反应堆产生的高温热可直接用于页岩油开采和高温制氢等工业领域。
文摘作为四代堆6种候选堆型中唯一的液态燃料反应堆,熔盐堆对未来核能和钍资源利用具有重要意义,特别是熔盐快堆(Molten Salt Fast Reactor,MSFR)还具有较大的增殖比和较好的温度负反馈。由于启动新的熔盐快堆需要较高的燃料装载量,若能改善MSFR的增殖性能,则有利于提高233U产量并缩短燃料倍增时间。首先应用SCALE6.1针对MSFR的径向增殖盐、新增轴向增殖盐和新增石墨反射层这三方面分析了初始增殖比,同时从核素吸收率角度说明增殖比变化的原因和MSFR的设计不足并对其进行了优化;然后应用基于SCALE6.1开发的熔盐堆在线处理模块(Molten Salt Reactor Reprocessing Sequence,MSR-RS)进行燃耗分析。结果表明,新增轴向增殖盐可以进一步提高增殖性能;新增石墨反射层可以节省增殖盐装载量。改进后的堆型运行时增殖比可以维持在1.1以上,233U年产量提高至133 kg,倍增时间缩短至36 a,并且堆芯在整个运行寿期都能保持足够的温度负反馈。
文摘数据存档系统是控制棒驱动机构(Control Rod Drive Mechanism,CRDM)样机控制系统的重要组成部分,主要用于存储控制棒的实时、报警信息及其他设备信息等数据,一方面能够使运行人员对熔盐堆CRDM样机的运行、调试和维修等工况进行分析和处理,另一方面为今后基于数据分析的预警和诊断技术发展积累数据资源。本文基于开源、数据类型和接口丰富、扩展功能强大的关系型数据库Postgre SQL,设计实现了熔盐堆CRDM样机数据存档系统的数据库。为提高数据库的性能并确保安全可靠的存储数据,在数据采集上采用了双机热备技术,同时使用Java语言开发了控制棒棒位检索工具并连接数据存档系统。该系统在熔盐堆CRDM样机中的成功应用证明其是一种先进的、低成本的、稳定的数据存档系统。
