General Fission(GEF)模型是描述裂变过程的一种半经验模型,它运用了量子力学和统计力学中的物理概念,并结合经验信息调整出了一组适用于不同裂变系统的参数,可以对大量原子核的裂变可观测量给出可靠的预测.本工作使用GEF模型计算了233...General Fission(GEF)模型是描述裂变过程的一种半经验模型,它运用了量子力学和统计力学中的物理概念,并结合经验信息调整出了一组适用于不同裂变系统的参数,可以对大量原子核的裂变可观测量给出可靠的预测.本工作使用GEF模型计算了233U中子诱发裂变产额的质量分布.结果表明,随着入射中子能量增加,对称裂变贡献逐渐增大,非对称裂变贡献逐渐减小;考虑了多次机会裂变,(n,f)裂变道相对贡献随着入射中子能量增加逐渐降低,而其他裂变道贡献增加;裂变产物的产额能量变化趋势与其在质量分布上的不同位置有关.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
In a thorium-based molten salt reactor(TMSR),it is difficult to achieve the pure 232Th–^(233)U fuel cycle without sufficient^(233)U fuel supply.Therefore,the original molten salt reactor was designed to use enriched ...In a thorium-based molten salt reactor(TMSR),it is difficult to achieve the pure 232Th–^(233)U fuel cycle without sufficient^(233)U fuel supply.Therefore,the original molten salt reactor was designed to use enriched uranium or plutonium as the starting fuel.By exploiting plutonium as the starting fuel and thorium as the fertile fuel,the high-purity^(233)U produced can be separated from the spent fuel by fluorination volatilization.Therefore,the molten salt reactor started with plutonium can be designed as a^(233)U breeder with the burning plutonium extracted from a pressurized water reactor(PWR).Combining these advantages,the study of the physical properties of plutonium-activated salt reactors is attractive.This study mainly focused on the burnup performance and temperature reactivity coefficient of a small modular molten-salt reactor started with plutonium(SM-MSR-Pu).The neutron spectra,^(233)U production,plutonium incineration,minor actinide(MA)residues,and temperature reactivity coefficients for different fuel salt volume fractions(VF)and hexagon pitch(P)sizes were calculated to analyze the burnup behavior in the SM-SMR-Pu.Based on the comparative analysis results of the burn-up calculation,a lower VF and larger P size are more beneficial for improving the burnup performance.However,from a passive safety perspective,a higher fuel volume fraction and smaller hexagon pitch size are necessary to achieve a deep negative feedback coefficient.Therefore,an excellent burnup performance and a deep negative temperature feedback coefficient are incompatible,and the optimal design range is relatively narrow in the optimized design of an SM-MSR-Pu.In a comprehensive consideration,P=20 cm and VF=20%are considered to be relatively balanced design parameters.Based on the fuel off-line batching scheme,a 250 MWth SM-MSR-Pu can produce approximately 29.83 kg of ^(233)U,incinerate 98.29 kg of plutonium,and accumulate 14.70 kg of MAs per year,and the temperature reactivity coefficient can always be lower than−4.0pcm/K.展开更多
An attempt is made to improve the evaluation of the prompt fission neutron emis- sion from 233U（n, f） reaction for incident neutron energies below 6 MeV. The multi-modal fission approach is applied to the improved v...An attempt is made to improve the evaluation of the prompt fission neutron emis- sion from 233U（n, f） reaction for incident neutron energies below 6 MeV. The multi-modal fission approach is applied to the improved version of Los Alamos model and the point by point model. The prompt fission neutron spectra and the prompt fission neutron as a function of fragment mass （usually named ＂sawtooth＂ data） v（A） are calculated independently for the three most dominant fission modes （standard I, standard II and superlong）, and the total spectra and v（A） are syn- thesized. The multi-modal parameters are determined on the basis of experimental data of fission fragment mass distributions. The present calculation results can describe the experimental data very well, and the proposed treatment is thus a useful tool for prompt fission neutron emission prediction.展开更多
The 14 MeV neutrons produced in the D-T fusion reactions have the potential of breeding Uranium-233 fissile fuel from fertile material Thorium-232. In order to estimate the amount of U-233 produced, experiments are ca...The 14 MeV neutrons produced in the D-T fusion reactions have the potential of breeding Uranium-233 fissile fuel from fertile material Thorium-232. In order to estimate the amount of U-233 produced, experiments are carried out by irradiating thorium dioxide pellets with neutrons produced from a 14 MeV neutron generator. The objective of the present work is to measure the reaction rates of 232Th ＋ in --＊ 233Th --＊ 2a^Pa --＊ 2a3U in different pellet thicknesses to study the self-shielding effects and adopt a procedure for correction. An appropriate assembly consisting of high-density polyethylene is designed and fabricated to slow down the high-energy neutrons, in which Thorium pellets are irradiated. The amount of fissile fuel （~3~U） produced is estimated by measuring the 312 keV gammas emitted by Protactinium-233 （half-life of 27 days）. A calibrated High Purity Germanium （HPGe） detector is used to measure the gamma ray spectrum. The amount of 233U produced by Th232 （n, ~） is calculated using MCNP code. The self-shielding effect is evaluated by calculating the reaction rates for different foil thickness. MCNP calculation results are compared with the experimental values and appropriate correction factors are estimated for self-shielding of neutrons and absorption of gamma rays.展开更多
Accelerator-driven systems based on molten salt fuel have several unique advantages and features for advanced nuclear fuel utilization.The aim of this work was to study the Th-U breeding capability in such systems,kno...Accelerator-driven systems based on molten salt fuel have several unique advantages and features for advanced nuclear fuel utilization.The aim of this work was to study the Th-U breeding capability in such systems,known as‘‘accelerator-driven subcritical molten salt reactors’’(ADS–MSRs).Breeding capacities including conversion ratio and net^(233)U production for various subcriticalities and different minor actinides(MA)loadings were analyzed for an ADS–MSR.The results show that the subcriticality of the core has a considerable effect on the Th-U breeding.A high subcriticality is favorable to improving the conversion ratio,increasing the net^(233)U production,and reducing the doubling time.Specifically,the doubling time for k_(eff)of 0.99 is larger than 80 years,while the counterpart for k_(eff)of 0.93 is only approximately22 years.Nevertheless,in an ADS–MSR with a high initial MA loading,MA results in a non-negligible^(233)U depletion in the first two decades,while increasing the net^(233)U production compared to reactors without MA loading.During the 50 years of operation,for the subcritical reactor(k_(eff)0:97)with MA fraction increasing from 1 to 14%,the net^(233)U production increases from 3.94 to 8.24 t.展开更多
Thorium as a suitable fertile with higher natural resources in comparison with uranium resources has been remarkably considered by different nuclear energy user countries in the last decades. Its prominent features su...Thorium as a suitable fertile with higher natural resources in comparison with uranium resources has been remarkably considered by different nuclear energy user countries in the last decades. Its prominent features such as suitable possibility for power flattening of a nuclear reactor, applicable breeder blanket to produce^(233)U fissile as well as neutron leakage prevention from a nuclear core has caused its application as power flatter, breeder material or other aimed utilizations be evaluated by the researches. In the present study, neutronics of a modeled CANDU 6loaded with Th O_2 and UO_2fuel rods have been computationally studied. The study aimed at reprocessing of burned Th O_2 seeds at CANDU 6 reactor to recover the total produced uranium, which is to be going under another compound fuel cycle. The obtained results showed all the core reactivity coefficients are sufficiently negative. The modeled core 949 GWd burn-up concluding in 99.99 %depletion of^(235)U initial loads. 18.38 kg of^(233) U was produced in the burnt Th O_2 fuel after 1-year burn-up time. In addition, 31.84 kg of^(239) Pu was produced in the UO_2 spent fuel rods after the burn-up time. After a proposed cooling time, about 50.01 kg of^(233)U will be available in the spent Th O_2 fuel.展开更多
钍铀燃料循环以其优异的物理和化学特性,受到核能界的广泛关注。本文利用单群的点燃耗计算程序ORIGEN,分别研究了钍燃料在沸水堆(Boiling Water Reactor,BWR)、压水堆(Pressurized Water Reactor,PWR)和加拿大重水铀反应堆(Canada Deute...钍铀燃料循环以其优异的物理和化学特性,受到核能界的广泛关注。本文利用单群的点燃耗计算程序ORIGEN,分别研究了钍燃料在沸水堆(Boiling Water Reactor,BWR)、压水堆(Pressurized Water Reactor,PWR)和加拿大重水铀反应堆(Canada Deuterium Oxide Uranium,CANDU,又称坎杜堆)能谱中辐照时,232Th、233Th、233Pa、233U等核素生成量随中子注量率和中子能谱的变化规律,并探索了多次"辐照-冷却"循环对钍铀转化率的影响。计算结果表明,能谱相同时,233Th和233Pa存量的最大值与注量率有关;233U存量的最大值与注量率无关,大概在注量(注量率×时间)为4×1016 n·cm-2左右;注量率相同时,能谱越硬,233U存量的最大值越大。采取循环"辐照-冷却"可以提高233Th-233U的转化率,对于相同的总辐照时间,每次循环周期内的辐照时间越短,相对于总辐照时间相同的单次辐照,转化率增量提高越明显;当总辐照时间超过两个月时,循环辐照对转化率增量的作用较小,与单次辐照不冷却相比,转化率相对增量不超过1倍。展开更多
A complete set of n＋233U neutron reaction data from 10-6 eV-20 MeV is updated and revised based on the evaluated experimental data and the feedback information of various benchmark tests, The main revised quantities ...A complete set of n＋233U neutron reaction data from 10-6 eV-20 MeV is updated and revised based on the evaluated experimental data and the feedback information of various benchmark tests, The main revised quantities are nubars, cross sections as well as angular distributions, etc. The benchmark tests indicate that the present evaluated data achieve very promising results.展开更多
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supported by the Chinese TMSR Strategic Pioneer Science and Technology Project(No.XDA02010000)Shanghai Pilot Program for Basic Research-Chinese Academy of Science,Shanghai Branch(No.JCYJ-SHFY-2021-003)the Chinese Academy of Sciences Special Research Assistant Funding Project.
文摘In a thorium-based molten salt reactor(TMSR),it is difficult to achieve the pure 232Th–^(233)U fuel cycle without sufficient^(233)U fuel supply.Therefore,the original molten salt reactor was designed to use enriched uranium or plutonium as the starting fuel.By exploiting plutonium as the starting fuel and thorium as the fertile fuel,the high-purity^(233)U produced can be separated from the spent fuel by fluorination volatilization.Therefore,the molten salt reactor started with plutonium can be designed as a^(233)U breeder with the burning plutonium extracted from a pressurized water reactor(PWR).Combining these advantages,the study of the physical properties of plutonium-activated salt reactors is attractive.This study mainly focused on the burnup performance and temperature reactivity coefficient of a small modular molten-salt reactor started with plutonium(SM-MSR-Pu).The neutron spectra,^(233)U production,plutonium incineration,minor actinide(MA)residues,and temperature reactivity coefficients for different fuel salt volume fractions(VF)and hexagon pitch(P)sizes were calculated to analyze the burnup behavior in the SM-SMR-Pu.Based on the comparative analysis results of the burn-up calculation,a lower VF and larger P size are more beneficial for improving the burnup performance.However,from a passive safety perspective,a higher fuel volume fraction and smaller hexagon pitch size are necessary to achieve a deep negative feedback coefficient.Therefore,an excellent burnup performance and a deep negative temperature feedback coefficient are incompatible,and the optimal design range is relatively narrow in the optimized design of an SM-MSR-Pu.In a comprehensive consideration,P=20 cm and VF=20%are considered to be relatively balanced design parameters.Based on the fuel off-line batching scheme,a 250 MWth SM-MSR-Pu can produce approximately 29.83 kg of ^(233)U,incinerate 98.29 kg of plutonium,and accumulate 14.70 kg of MAs per year,and the temperature reactivity coefficient can always be lower than−4.0pcm/K.
基金supported by the State Key Development Program for Basic Research of China (Nos. 2008CB717803, 2009GB107001, and2007CB209903)the Research Fund for the Doctoral Program of Higher Education of China (No. 200610011023)
文摘An attempt is made to improve the evaluation of the prompt fission neutron emis- sion from 233U（n, f） reaction for incident neutron energies below 6 MeV. The multi-modal fission approach is applied to the improved version of Los Alamos model and the point by point model. The prompt fission neutron spectra and the prompt fission neutron as a function of fragment mass （usually named ＂sawtooth＂ data） v（A） are calculated independently for the three most dominant fission modes （standard I, standard II and superlong）, and the total spectra and v（A） are syn- thesized. The multi-modal parameters are determined on the basis of experimental data of fission fragment mass distributions. The present calculation results can describe the experimental data very well, and the proposed treatment is thus a useful tool for prompt fission neutron emission prediction.
文摘The 14 MeV neutrons produced in the D-T fusion reactions have the potential of breeding Uranium-233 fissile fuel from fertile material Thorium-232. In order to estimate the amount of U-233 produced, experiments are carried out by irradiating thorium dioxide pellets with neutrons produced from a 14 MeV neutron generator. The objective of the present work is to measure the reaction rates of 232Th ＋ in --＊ 233Th --＊ 2a^Pa --＊ 2a3U in different pellet thicknesses to study the self-shielding effects and adopt a procedure for correction. An appropriate assembly consisting of high-density polyethylene is designed and fabricated to slow down the high-energy neutrons, in which Thorium pellets are irradiated. The amount of fissile fuel （~3~U） produced is estimated by measuring the 312 keV gammas emitted by Protactinium-233 （half-life of 27 days）. A calibrated High Purity Germanium （HPGe） detector is used to measure the gamma ray spectrum. The amount of 233U produced by Th232 （n, ~） is calculated using MCNP code. The self-shielding effect is evaluated by calculating the reaction rates for different foil thickness. MCNP calculation results are compared with the experimental values and appropriate correction factors are estimated for self-shielding of neutrons and absorption of gamma rays.
基金supported by the Chinese TMSR Strategic Pioneer Science and Technology Project(No.XDA02010000)the Frontier Science Key Program of the Chinese Academy of Sciences(No.QYZDY-SSW-JSC016)
文摘Accelerator-driven systems based on molten salt fuel have several unique advantages and features for advanced nuclear fuel utilization.The aim of this work was to study the Th-U breeding capability in such systems,known as‘‘accelerator-driven subcritical molten salt reactors’’(ADS–MSRs).Breeding capacities including conversion ratio and net^(233)U production for various subcriticalities and different minor actinides(MA)loadings were analyzed for an ADS–MSR.The results show that the subcriticality of the core has a considerable effect on the Th-U breeding.A high subcriticality is favorable to improving the conversion ratio,increasing the net^(233)U production,and reducing the doubling time.Specifically,the doubling time for k_(eff)of 0.99 is larger than 80 years,while the counterpart for k_(eff)of 0.93 is only approximately22 years.Nevertheless,in an ADS–MSR with a high initial MA loading,MA results in a non-negligible^(233)U depletion in the first two decades,while increasing the net^(233)U production compared to reactors without MA loading.During the 50 years of operation,for the subcritical reactor(k_(eff)0:97)with MA fraction increasing from 1 to 14%,the net^(233)U production increases from 3.94 to 8.24 t.
文摘Thorium as a suitable fertile with higher natural resources in comparison with uranium resources has been remarkably considered by different nuclear energy user countries in the last decades. Its prominent features such as suitable possibility for power flattening of a nuclear reactor, applicable breeder blanket to produce^(233)U fissile as well as neutron leakage prevention from a nuclear core has caused its application as power flatter, breeder material or other aimed utilizations be evaluated by the researches. In the present study, neutronics of a modeled CANDU 6loaded with Th O_2 and UO_2fuel rods have been computationally studied. The study aimed at reprocessing of burned Th O_2 seeds at CANDU 6 reactor to recover the total produced uranium, which is to be going under another compound fuel cycle. The obtained results showed all the core reactivity coefficients are sufficiently negative. The modeled core 949 GWd burn-up concluding in 99.99 %depletion of^(235)U initial loads. 18.38 kg of^(233) U was produced in the burnt Th O_2 fuel after 1-year burn-up time. In addition, 31.84 kg of^(239) Pu was produced in the UO_2 spent fuel rods after the burn-up time. After a proposed cooling time, about 50.01 kg of^(233)U will be available in the spent Th O_2 fuel.
文摘钍铀燃料循环以其优异的物理和化学特性,受到核能界的广泛关注。本文利用单群的点燃耗计算程序ORIGEN,分别研究了钍燃料在沸水堆(Boiling Water Reactor,BWR)、压水堆(Pressurized Water Reactor,PWR)和加拿大重水铀反应堆(Canada Deuterium Oxide Uranium,CANDU,又称坎杜堆)能谱中辐照时,232Th、233Th、233Pa、233U等核素生成量随中子注量率和中子能谱的变化规律,并探索了多次"辐照-冷却"循环对钍铀转化率的影响。计算结果表明,能谱相同时,233Th和233Pa存量的最大值与注量率有关;233U存量的最大值与注量率无关,大概在注量(注量率×时间)为4×1016 n·cm-2左右;注量率相同时,能谱越硬,233U存量的最大值越大。采取循环"辐照-冷却"可以提高233Th-233U的转化率,对于相同的总辐照时间,每次循环周期内的辐照时间越短,相对于总辐照时间相同的单次辐照,转化率增量提高越明显;当总辐照时间超过两个月时,循环辐照对转化率增量的作用较小,与单次辐照不冷却相比,转化率相对增量不超过1倍。
文摘A complete set of n＋233U neutron reaction data from 10-6 eV-20 MeV is updated and revised based on the evaluated experimental data and the feedback information of various benchmark tests, The main revised quantities are nubars, cross sections as well as angular distributions, etc. The benchmark tests indicate that the present evaluated data achieve very promising results.
