Burnup measurement is crucial for the management and disposal of spent fuel.The conventional approach indirectly estimates burnup by examining the fission product or actinide content.Compared to the first two methods,...Burnup measurement is crucial for the management and disposal of spent fuel.The conventional approach indirectly estimates burnup by examining the fission product or actinide content.Compared to the first two methods,the active neutron method exhibits a lower dependence on the irradiation history and initial enrichment degree of the spent fuel.In addition,it can be used to directly determine the content of fissile nuclides in spent fuel.This study proposed the design of a burnup measurement equipment specifically crafted for plate segments by utilizing a compact D-D neutron generator.The equipment initiates the fission of fissile nuclides within the spent fuel plate segment through thermal neutrons provided by the moderators.Subsequently,the burnup is determined by analyzing the transmitted thermal neutrons and counting the fission fast neutrons.The Monte Carlo program Geant4 was used to simulate the relationship between spent fuel plate segment assembly burnup and the detector count of 10 MW material test reactor designed by the International Atomic Energy Agency.Consequently,the feasibility of the method and rationality of the detector design were verified.展开更多
Depletion calculation is important for studying the transmutation efficiency of minor actinides and longlife fission products in accelerator-driven subcritical reactor system(ADS). Herein the Python language is used t...Depletion calculation is important for studying the transmutation efficiency of minor actinides and longlife fission products in accelerator-driven subcritical reactor system(ADS). Herein the Python language is used to develop a burnup code system called IMPC-Burnup by coupling FLUKA, OpenMC, and ORIGEN2. The program is preliminarily verified by OECD-NEA pin cell and IAEAADS benchmarking by comparison with experimental values and calculated results from other studies. Moreover,the physics design scheme of the CIADS subcritical core is utilized to test the feasibility of IMPC-Burnup program in the burnup calculation of ADS system. Reference results are given by the COUPLE3.0 program. The results of IMPC-Burnup show good agreement with those of COUPLE3.0. In addition, since the upper limit of the neutron transport energy for OpenMC is 20 MeV, neutrons with energies greater than 20 MeV in the CIADS subcritical core cannot be transported; thus, an equivalent flux method has been proposed to consider neutrons above 20 MeV in the OpenMC transport calculation. The results are compared to those that do not include neutrons greater than 20 MeV. The conclusion is that the accuracy of the actinide nuclide mass in the burnup calculation is improved when the equivalent flux method is used. Therefore, the IMPC-Burnup code is suitable for burnup analysis of the ADS system.展开更多
To benefit from recent advances in modeling and computational algorithms,as well as the availability of new covariance data,sensitivity and uncertainty analyses are needed to quantify the impact of uncertain sources o...To benefit from recent advances in modeling and computational algorithms,as well as the availability of new covariance data,sensitivity and uncertainty analyses are needed to quantify the impact of uncertain sources on the design parameters of small prismatic high-temperature gascooled reactors(HTGRs).In particular,the contribution of nuclear data to the k_(eff)uncertainty is an important part of the uncertainty analysis of small-sized HTGR physical calculations.In this study,a small-sized HTGR designed by China Nuclear Power Engineering Co.,Ltd.was selected for k_(eff)uncertainty analysis during full lifetime burnup calculations.Models of the cold zero power(CZP)condition and full lifetime burnup process were constructed using the Reactor Monte Carlo Code RMC for neutron transport calculation,depletion calculation,and sensitivity and uncertainty analysis.For the sensitivity analysis,the Contribution-Linked eigenvalue sensitivity/Uncertainty estimation via Track length importance Characterization(CLUTCH)method was applied to obtain sensitive information,and the "sandwich" method was used to quantify the k_(eff)uncertainty.We also compared the k_(eff)uncertainties to other typical reactors.Our results show that ^(235)U is the largest contributor to k_(eff)uncertainty for both the CZP and depletion conditions,while the contribution of ^(239)Pu is not very significant because of the design of low discharge burnup.It is worth noting that the radioactive capture reaction of ^(28)Si significantly contributes to the k_(eff)uncertainty owing to its specific fuel design.However,the k_(eff)uncertainty during the full lifetime depletion process was relatively stable,only increasing by 1.12%owing to the low discharge burnup design of small-sized HTGRs.These numerical results are beneficial for neutronics design and core parameters optimization in further uncertainty propagation and quantification study for small-sized HTGR.展开更多
Monte Carlo transport simulations of a full-core reactor with a high-fidelity structure have been made possible by modern-day computing capabilities. Performing transport–burnup calculations of a full-core model typi...Monte Carlo transport simulations of a full-core reactor with a high-fidelity structure have been made possible by modern-day computing capabilities. Performing transport–burnup calculations of a full-core model typically includes millions of burnup areas requiring hundreds of gigabytes of memory for burnup-related tallies. This paper presents the study of a parallel computing method for full-core Monte Carlo transport–burnup calculations and the development of a thread-level data decomposition method. The proposed method decomposes tally accumulators into different threads and improves the parallel communication pattern and memory access efficiency. A typical pressurized water reactor burnup assembly along with the benchmark for evaluation and validation of reactor simulations model was used to test the proposed method.The result indicates that the method effectively reduces memory consumption and maintains high parallel efficiency.展开更多
The analysis of the fuel depletion behavior is critical for maintaining the safety of accelerator-driven subcritical systems(ADSs). The code COUPLE2.0 coupling 3-D neutron transport and point burnup calculation was de...The analysis of the fuel depletion behavior is critical for maintaining the safety of accelerator-driven subcritical systems(ADSs). The code COUPLE2.0 coupling 3-D neutron transport and point burnup calculation was developed by Tsinghua University. A Monte Carlo method is used for the neutron transport analysis, and the burnup calculation is based on a deterministic method. The code can be used for the analysis of targets coupled with a reactor in ADSs. In response to additional ADS analysis requirements at the Institute of Modern Physics at the Chinese Academy of Sciences, the COUPLE3.0 version was developed to include the new functions of(1) a module for the calculation of proton irradiation for the analysis of cumulative behavior using the residual radionuclide operating history,(2) a fixed-flux radiation module for hazard assessment and analysis of the burnable poison, and(3) a module for multi-kernel parallel calculation, which improves the radionuclide replacement for the burnup analysis to balance the precision level and computational efficiency of the program. This paper introduces thevalidation of the COUPLE3.0 code using a fast reactor benchmark and ADS benchmark calculations. Moreover,the proton irradiation module was verified by a comparison with the analytic method of calculating the210 Po accumulation results. The results demonstrate that COUPLE3.0 is suitable for the analysis of neutron transport and the burnup of nuclides for ADSs.展开更多
The Molten Salt Reactor(MSR) is one of the six advanced reactor nuclear energy systems for further research and development selected by Generation IV International Forum(GIF),which is distinguished by its core in whic...The Molten Salt Reactor(MSR) is one of the six advanced reactor nuclear energy systems for further research and development selected by Generation IV International Forum(GIF),which is distinguished by its core in which the fuel is dissolved in molten fluoride salt.Because fuel flow in the primary loop,the depletion of MSR is different from that of solid-fuel reactors.In this paper,an MCNP5 and ORIGEN2 Coupled Burnup(MOCBurn) code for MSR is developed under the MATLAB platform.Some new methods and novel arrangements are used to make it suitable for fuel flow in the MSR.To consider the fuel convection and diffusion in the primary loop of MSR,fuel mixing calculation is carried out after each burnup time step.Modeling function for geometry with repeat structures is implicated for reactor analysis with complex structures.Calculation for a high-burnup reactor pin cell benchmark is performed using the MOCBurn code.Results of depletion study show that the MOCBurn code is suitable for the traditional solid-fuel reactors.A preliminary study of the fuel mixture effect in MSR is also carried out.展开更多
This paper presents the determination of the fuel burnup distribution of the Dalat nuclear research reactor(DNRR) using a method of measurements at subcritical conditions. The method is based on the assumption of line...This paper presents the determination of the fuel burnup distribution of the Dalat nuclear research reactor(DNRR) using a method of measurements at subcritical conditions. The method is based on the assumption of linear dependence of the reactivity on the burnup of fuel bundles and the measurements at subcritical conditions.The measurements were taken for seven selected fuel bundles in two different measuring sequences. The measured burnup values have also been compared with the calculations for verifying the method and the measurement procedure. The results obtained with the three detectors have a good agreement with each other with a discrepancy less than 1.0%. The errors of the measured burnup values are within 6%. Comparison between the calculated and measured burnup values shows that the discrepancy of the C/E ratio is within 9% compared to unity. The results indicate that the method of measurements at subcritical conditions could be well applied to determine the relative burnup distribution of the DNRR.展开更多
活化-燃耗计算是反应堆分析的重要组成部分,需要耦合临界输运程序和点燃耗程序来迭代求解。LightAB(Light Activation and Burnup)是一个新的轻量化的通用活化-燃耗分析程序,可以用来处理活化-燃耗系统。LightAB采用了基于燃耗计算程序O...活化-燃耗计算是反应堆分析的重要组成部分,需要耦合临界输运程序和点燃耗程序来迭代求解。LightAB(Light Activation and Burnup)是一个新的轻量化的通用活化-燃耗分析程序,可以用来处理活化-燃耗系统。LightAB采用了基于燃耗计算程序ORIGEN-2和ORIGEN-S的燃耗数据库,并使用处理刚性燃耗系统的切比雪夫有理逼近算法(Chebyshev Rational Approximation Method,CRAM)。LightAB支持衰变模式,定通量模式和定功率模式的活化-燃耗计算。为了验证LightAB程序的正确性,计算了237Np衰变问题和锆(Zr)合金的固定通量辐照问题,结果与燃耗程序ORIGEN-2.1一致。实现了LightAB和RMC程序的耦合,并用于计算压水堆(Pressurized Water Reactor,PWR)栅元燃耗基准题、PWR组件燃耗问题、经合组织/核能机构(Organisation for Economic Co-operation and Development/Nuclear Energy Agency,OECD/NEA)快堆燃耗基准问题,计算结果与RMC的燃耗计算模块一致。在超钚同位素的辐照生产与RMC对比模拟计算中,LightAB展现出良好的应用前景。展开更多
基金supported by the National Natural Science Foundation of China(No.12075105)the Major Science and Technology Projects of Gansu Province(No.22ZD6GB020)+1 种基金the NSFC-Nuclear Technology Innovation Joint Fund(No.U2167203)the Fundamental Research Funds for the Central Universities(lzujbky-2023-stlt01,lzujbky-2024-jdzx10)。
文摘Burnup measurement is crucial for the management and disposal of spent fuel.The conventional approach indirectly estimates burnup by examining the fission product or actinide content.Compared to the first two methods,the active neutron method exhibits a lower dependence on the irradiation history and initial enrichment degree of the spent fuel.In addition,it can be used to directly determine the content of fissile nuclides in spent fuel.This study proposed the design of a burnup measurement equipment specifically crafted for plate segments by utilizing a compact D-D neutron generator.The equipment initiates the fission of fissile nuclides within the spent fuel plate segment through thermal neutrons provided by the moderators.Subsequently,the burnup is determined by analyzing the transmitted thermal neutrons and counting the fission fast neutrons.The Monte Carlo program Geant4 was used to simulate the relationship between spent fuel plate segment assembly burnup and the detector count of 10 MW material test reactor designed by the International Atomic Energy Agency.Consequently,the feasibility of the method and rationality of the detector design were verified.
基金supported by the ‘‘Strategic Priority Research Program’’ of Chinese Academy of Sciences(No.XDA03030102)
文摘Depletion calculation is important for studying the transmutation efficiency of minor actinides and longlife fission products in accelerator-driven subcritical reactor system(ADS). Herein the Python language is used to develop a burnup code system called IMPC-Burnup by coupling FLUKA, OpenMC, and ORIGEN2. The program is preliminarily verified by OECD-NEA pin cell and IAEAADS benchmarking by comparison with experimental values and calculated results from other studies. Moreover,the physics design scheme of the CIADS subcritical core is utilized to test the feasibility of IMPC-Burnup program in the burnup calculation of ADS system. Reference results are given by the COUPLE3.0 program. The results of IMPC-Burnup show good agreement with those of COUPLE3.0. In addition, since the upper limit of the neutron transport energy for OpenMC is 20 MeV, neutrons with energies greater than 20 MeV in the CIADS subcritical core cannot be transported; thus, an equivalent flux method has been proposed to consider neutrons above 20 MeV in the OpenMC transport calculation. The results are compared to those that do not include neutrons greater than 20 MeV. The conclusion is that the accuracy of the actinide nuclide mass in the burnup calculation is improved when the equivalent flux method is used. Therefore, the IMPC-Burnup code is suitable for burnup analysis of the ADS system.
基金supported by the National Natural Science Foundation of China(No.12075067)the National Key R&D Program of China(No.2018YFE0180900)。
文摘To benefit from recent advances in modeling and computational algorithms,as well as the availability of new covariance data,sensitivity and uncertainty analyses are needed to quantify the impact of uncertain sources on the design parameters of small prismatic high-temperature gascooled reactors(HTGRs).In particular,the contribution of nuclear data to the k_(eff)uncertainty is an important part of the uncertainty analysis of small-sized HTGR physical calculations.In this study,a small-sized HTGR designed by China Nuclear Power Engineering Co.,Ltd.was selected for k_(eff)uncertainty analysis during full lifetime burnup calculations.Models of the cold zero power(CZP)condition and full lifetime burnup process were constructed using the Reactor Monte Carlo Code RMC for neutron transport calculation,depletion calculation,and sensitivity and uncertainty analysis.For the sensitivity analysis,the Contribution-Linked eigenvalue sensitivity/Uncertainty estimation via Track length importance Characterization(CLUTCH)method was applied to obtain sensitive information,and the "sandwich" method was used to quantify the k_(eff)uncertainty.We also compared the k_(eff)uncertainties to other typical reactors.Our results show that ^(235)U is the largest contributor to k_(eff)uncertainty for both the CZP and depletion conditions,while the contribution of ^(239)Pu is not very significant because of the design of low discharge burnup.It is worth noting that the radioactive capture reaction of ^(28)Si significantly contributes to the k_(eff)uncertainty owing to its specific fuel design.However,the k_(eff)uncertainty during the full lifetime depletion process was relatively stable,only increasing by 1.12%owing to the low discharge burnup design of small-sized HTGRs.These numerical results are beneficial for neutronics design and core parameters optimization in further uncertainty propagation and quantification study for small-sized HTGR.
基金supported by the Innovation Foundation of the Chinese Academy of Sciences(No.CXJJ-16Q231)the National Natural Science Foundation of China(No.11305203)+3 种基金the Special Program for Informatization of the Chinese Academy of Sciences(No.XXH12504-1-09)the Anhui Provincial Special Project for High Technology Industrythe Special Project of Youth Innovation Promotion Association of Chinese Academy of Sciencesthe Industrialization Fund
文摘Monte Carlo transport simulations of a full-core reactor with a high-fidelity structure have been made possible by modern-day computing capabilities. Performing transport–burnup calculations of a full-core model typically includes millions of burnup areas requiring hundreds of gigabytes of memory for burnup-related tallies. This paper presents the study of a parallel computing method for full-core Monte Carlo transport–burnup calculations and the development of a thread-level data decomposition method. The proposed method decomposes tally accumulators into different threads and improves the parallel communication pattern and memory access efficiency. A typical pressurized water reactor burnup assembly along with the benchmark for evaluation and validation of reactor simulations model was used to test the proposed method.The result indicates that the method effectively reduces memory consumption and maintains high parallel efficiency.
基金supported by the Strategic Priority Research Program of the Chinese Academy of Sciences(No.XDA03030102)
文摘The analysis of the fuel depletion behavior is critical for maintaining the safety of accelerator-driven subcritical systems(ADSs). The code COUPLE2.0 coupling 3-D neutron transport and point burnup calculation was developed by Tsinghua University. A Monte Carlo method is used for the neutron transport analysis, and the burnup calculation is based on a deterministic method. The code can be used for the analysis of targets coupled with a reactor in ADSs. In response to additional ADS analysis requirements at the Institute of Modern Physics at the Chinese Academy of Sciences, the COUPLE3.0 version was developed to include the new functions of(1) a module for the calculation of proton irradiation for the analysis of cumulative behavior using the residual radionuclide operating history,(2) a fixed-flux radiation module for hazard assessment and analysis of the burnable poison, and(3) a module for multi-kernel parallel calculation, which improves the radionuclide replacement for the burnup analysis to balance the precision level and computational efficiency of the program. This paper introduces thevalidation of the COUPLE3.0 code using a fast reactor benchmark and ADS benchmark calculations. Moreover,the proton irradiation module was verified by a comparison with the analytic method of calculating the210 Po accumulation results. The results demonstrate that COUPLE3.0 is suitable for the analysis of neutron transport and the burnup of nuclides for ADSs.
基金supported by the‘‘Strategic Priority Research Program’’of the Chinese Academy of Science(No.XDA02010000)
文摘The Molten Salt Reactor(MSR) is one of the six advanced reactor nuclear energy systems for further research and development selected by Generation IV International Forum(GIF),which is distinguished by its core in which the fuel is dissolved in molten fluoride salt.Because fuel flow in the primary loop,the depletion of MSR is different from that of solid-fuel reactors.In this paper,an MCNP5 and ORIGEN2 Coupled Burnup(MOCBurn) code for MSR is developed under the MATLAB platform.Some new methods and novel arrangements are used to make it suitable for fuel flow in the MSR.To consider the fuel convection and diffusion in the primary loop of MSR,fuel mixing calculation is carried out after each burnup time step.Modeling function for geometry with repeat structures is implicated for reactor analysis with complex structures.Calculation for a high-burnup reactor pin cell benchmark is performed using the MOCBurn code.Results of depletion study show that the MOCBurn code is suitable for the traditional solid-fuel reactors.A preliminary study of the fuel mixture effect in MSR is also carried out.
基金supported by National Foundation for Science and Technology Development(NAFOSTED)of Vietnam under Grant103.04-2016.30
文摘This paper presents the determination of the fuel burnup distribution of the Dalat nuclear research reactor(DNRR) using a method of measurements at subcritical conditions. The method is based on the assumption of linear dependence of the reactivity on the burnup of fuel bundles and the measurements at subcritical conditions.The measurements were taken for seven selected fuel bundles in two different measuring sequences. The measured burnup values have also been compared with the calculations for verifying the method and the measurement procedure. The results obtained with the three detectors have a good agreement with each other with a discrepancy less than 1.0%. The errors of the measured burnup values are within 6%. Comparison between the calculated and measured burnup values shows that the discrepancy of the C/E ratio is within 9% compared to unity. The results indicate that the method of measurements at subcritical conditions could be well applied to determine the relative burnup distribution of the DNRR.
文摘活化-燃耗计算是反应堆分析的重要组成部分,需要耦合临界输运程序和点燃耗程序来迭代求解。LightAB(Light Activation and Burnup)是一个新的轻量化的通用活化-燃耗分析程序,可以用来处理活化-燃耗系统。LightAB采用了基于燃耗计算程序ORIGEN-2和ORIGEN-S的燃耗数据库,并使用处理刚性燃耗系统的切比雪夫有理逼近算法(Chebyshev Rational Approximation Method,CRAM)。LightAB支持衰变模式,定通量模式和定功率模式的活化-燃耗计算。为了验证LightAB程序的正确性,计算了237Np衰变问题和锆(Zr)合金的固定通量辐照问题,结果与燃耗程序ORIGEN-2.1一致。实现了LightAB和RMC程序的耦合,并用于计算压水堆(Pressurized Water Reactor,PWR)栅元燃耗基准题、PWR组件燃耗问题、经合组织/核能机构(Organisation for Economic Co-operation and Development/Nuclear Energy Agency,OECD/NEA)快堆燃耗基准问题,计算结果与RMC的燃耗计算模块一致。在超钚同位素的辐照生产与RMC对比模拟计算中,LightAB展现出良好的应用前景。
