The study of accelerator-driven subcritical reactor systems(ADSs) has been an important research topic in the field of nuclear energy for years. The main code applied in ADS research is MCNPX, which was developed by L...The study of accelerator-driven subcritical reactor systems(ADSs) has been an important research topic in the field of nuclear energy for years. The main code applied in ADS research is MCNPX, which was developed by Los Alamos National Laboratory. We studied the application of the open-source Monte Carlo codes FLUKA and OpenMC to a coupled ADS calculation. The FLUKA code was used to simulate the reaction of highenergy protons with the nucleus of the target material in the ADS, which produces spallation neutrons. Information on the spallation neutrons, such as their energy, position,direction, and weight, can be recorded by a user-defined routine called FLUSCW provided by FLUKA. Then, the information was stored in an external neutron source file in HDF5 format by using a conversion code, as required by the OpenMC calculation. Finally, the fixed-source calculation function of OpenMC was applied to simulate the transport of spallation neutrons and obtain the distribution of the neutron flux in the core region. In the coupled calculation, the high-energy cross-section library JENDL4.0/HE in ACE format produced by NJOY2016 was applied in the OpenMC transport simulation. The OECD–ADS benchmark problem was calculated, and the results were compared with those obtained using MCNPX. It was found that the flux calculations performed by FLUKA–OpenMC and MCNPX were in agreement, so the coupling calculation method for ADS is reasonable and feasible.展开更多
开源蒙特卡罗程序OpenMC(OpenMonte Carlo code)只提供源代码而没有执行码,在编译OpenMC的过程中发现不同版本的辅助程序与之存在兼容性问题。本文通过分析OpenMPI、Mpich及HDF5各版本辅助程序,对0.6.2版本OpenMC源代码的支持情况进行研...开源蒙特卡罗程序OpenMC(OpenMonte Carlo code)只提供源代码而没有执行码,在编译OpenMC的过程中发现不同版本的辅助程序与之存在兼容性问题。本文通过分析OpenMPI、Mpich及HDF5各版本辅助程序,对0.6.2版本OpenMC源代码的支持情况进行研究,为正确编译OpenMC执行码给出了直接参考。为进一步验证OpenMC执行码计算临界问题的正确性,选择国际临界安全基准评价实验手册(The International Criticality Safety Benchmark Evaluation Project,ICSBEP)中的96道代表性例题进行基准校验,与通用蒙特卡罗程序的计算结果进行对比并以实验值作为参考。结果表明,OpenMC计算值与实验值及其他程序计算值吻合较好,验证了OpenMC临界计算的可行性和正确性,上述结论将为程序以后的实际应用及完善奠定基础。展开更多
聚变装置工程模型极其复杂,使得中子学分析的建模十分繁琐和耗时。开源蒙特卡罗程序OpenMC通过集成DAGMC(Direct Accelerated Geometry Monte Carlo),可以直接基于CAD模型进行粒子输运模拟计算,该特性可显著提高复杂工程模型的建模与分...聚变装置工程模型极其复杂,使得中子学分析的建模十分繁琐和耗时。开源蒙特卡罗程序OpenMC通过集成DAGMC(Direct Accelerated Geometry Monte Carlo),可以直接基于CAD模型进行粒子输运模拟计算,该特性可显著提高复杂工程模型的建模与分析效率。以中国聚变工程试验堆(China Fusion Engineerging Test Reactor,CFETR)为对象,开展OpenMC在聚变中子学分析中的应用研究。基于CFETR一维柱壳模型验证OpenMC与MCNP计数结果的一致性。根据等离子体空间分布特点,基于源扩展接口自定义源类和源函数准确描述复杂聚变中子源。利用DAG-OpenMC的CAD几何功能成功建立了CFETR的三维模型,并计算获得了中子壁负载分布、氚增殖率和核热沉积等物理量。结果表明:DAG-OpenMC与MCNP的计算结果具有极好的一致性。在建立复杂的聚变堆工程模型时,基于CAD几何功能极大地提高了建模效率。DAG-OpenMC在聚变中子学应用中关键问题的验证表明了其处理复杂工程结构条件下聚变中子学问题的可行性。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
The reactivity of a nuclear reactor is the most important safety and operating parameter. Due to short reactor period, the Light Water Reactor(LWR) designs require the compensations of rapid unfavorable reactivity inc...The reactivity of a nuclear reactor is the most important safety and operating parameter. Due to short reactor period, the Light Water Reactor(LWR) designs require the compensations of rapid unfavorable reactivity increases. The increase in fuel or moderator temperature leads to compensate the reactivity jumps as inherent safety characteristics. The safe and reliable reactor operation requires the accurate assessment of these reactivity changes. This paper highlights the improvements in the methodology to determine the feedback reactivity changes in IAEA MTR benchmark. This method incorporates the reactivity effects of fuel temperature in moderator regions and vice versa. For this purpose, a detailed 3D model of the IAEA 10 MW MTR benchmark reactor is developed employing OpenMC computer code. OpenMC is a probabilistic computer code for neutronic calculations. This work uses temperature-dependent JEFF 3.2 cross-sectional library. The model is validated against the reference results of eigenvalues for control rods(inserted and in fully withdrawn position), control rod reactivity worth, averaged thermal flux in the central flux trap, and power fraction for each fuel element at beginning of life. The validated model is applied to simulate the feedback reactivity coefficients against the conventional reference results. In order to improve the methodology, the effect of the moderator temperature and void on fuel is incorporated to obtain a more realistic value of the fuel temperature coefficient.Similarly, the moderator temperature coefficient and void coefficient are improved by incorporating the coupling effects of fuel temperature on moderator. This methodology can be applied to improve the LWR designs.展开更多
The physical quantities calculated by nuclear reactor Monte Carlo simulations are typically recorded on a grid of two or three spatial dimensions and one dimension of neutron energy.Because of this,increasing the reso...The physical quantities calculated by nuclear reactor Monte Carlo simulations are typically recorded on a grid of two or three spatial dimensions and one dimension of neutron energy.Because of this,increasing the resolution of the calculated quantities can have a significant impact on the memory and CPU time required to run a simulation.Convolutional neural networks have been shown to accurately upsample coarse-resolution photo-graphic images to resolutions multiple times finer than the originals.Here we show that a convolutional neural network can accurately upsample flux tallies in a Monte Carlo neutron transport simulation by a factor of two along the spatial and energy dimensions.Neutron flux tallies in pressurized water reactor assemblies were calculated using OpenMC at a 64×64 pixel spatial resolution and 8 neutron energy groups for input to the neural network.The network upsamples the low-resolution neutron flux to 128×128 pixel spatial resolution and 16 neutron energy groups.High-resolution neutron flux tallies and their uncertainties were also calculated with OpenMC and compared with the network’s predictions.The upsampled data and the high-resolution tally results agree to within the statistical uncertainty calculated by OpenMC.展开更多
无慢化罐式堆芯结构的熔盐快堆(Molten Salt Fast Reactor,MSFR)中存在中子物理与热工水力的强耦合。应用耦合蒙特卡罗粒子输运程序OpenMC与计算流体力学软件OpenFOAM,建立了一套适用于熔盐快堆的三维稳态核热耦合计算程序。该程序基于p...无慢化罐式堆芯结构的熔盐快堆(Molten Salt Fast Reactor,MSFR)中存在中子物理与热工水力的强耦合。应用耦合蒙特卡罗粒子输运程序OpenMC与计算流体力学软件OpenFOAM,建立了一套适用于熔盐快堆的三维稳态核热耦合计算程序。该程序基于python编程语言实现了OpenMC和OpenFOAM二者间的功率、燃料盐温度和密度分布等数据交互,可以获得堆芯内三维功率分布、中子通量密度分布、三维速度场和温度场分布。采用该耦合程序,建立了熔盐快堆的基准模型,研究了中子学区域划分数目和初始条件对keff、燃料盐速度和温度分布的影响。根据研究结果,推荐了一套合理的中子学区域划分方法与数目,表明了耦合程序设定的不同初始条件对keff结果无影响。最后,通过与熔盐快堆基准结果的对比验证了耦合程序的正确性,表明该程序适用于熔盐快堆的稳态核热耦合分析。展开更多
基金supported by the ‘‘Strategic Priority Research Program’’ of the Chinese Academy of Sciences(No.XDA03030102)
文摘The study of accelerator-driven subcritical reactor systems(ADSs) has been an important research topic in the field of nuclear energy for years. The main code applied in ADS research is MCNPX, which was developed by Los Alamos National Laboratory. We studied the application of the open-source Monte Carlo codes FLUKA and OpenMC to a coupled ADS calculation. The FLUKA code was used to simulate the reaction of highenergy protons with the nucleus of the target material in the ADS, which produces spallation neutrons. Information on the spallation neutrons, such as their energy, position,direction, and weight, can be recorded by a user-defined routine called FLUSCW provided by FLUKA. Then, the information was stored in an external neutron source file in HDF5 format by using a conversion code, as required by the OpenMC calculation. Finally, the fixed-source calculation function of OpenMC was applied to simulate the transport of spallation neutrons and obtain the distribution of the neutron flux in the core region. In the coupled calculation, the high-energy cross-section library JENDL4.0/HE in ACE format produced by NJOY2016 was applied in the OpenMC transport simulation. The OECD–ADS benchmark problem was calculated, and the results were compared with those obtained using MCNPX. It was found that the flux calculations performed by FLUKA–OpenMC and MCNPX were in agreement, so the coupling calculation method for ADS is reasonable and feasible.
文摘开源蒙特卡罗程序OpenMC(OpenMonte Carlo code)只提供源代码而没有执行码,在编译OpenMC的过程中发现不同版本的辅助程序与之存在兼容性问题。本文通过分析OpenMPI、Mpich及HDF5各版本辅助程序,对0.6.2版本OpenMC源代码的支持情况进行研究,为正确编译OpenMC执行码给出了直接参考。为进一步验证OpenMC执行码计算临界问题的正确性,选择国际临界安全基准评价实验手册(The International Criticality Safety Benchmark Evaluation Project,ICSBEP)中的96道代表性例题进行基准校验,与通用蒙特卡罗程序的计算结果进行对比并以实验值作为参考。结果表明,OpenMC计算值与实验值及其他程序计算值吻合较好,验证了OpenMC临界计算的可行性和正确性,上述结论将为程序以后的实际应用及完善奠定基础。
文摘聚变装置工程模型极其复杂,使得中子学分析的建模十分繁琐和耗时。开源蒙特卡罗程序OpenMC通过集成DAGMC(Direct Accelerated Geometry Monte Carlo),可以直接基于CAD模型进行粒子输运模拟计算,该特性可显著提高复杂工程模型的建模与分析效率。以中国聚变工程试验堆(China Fusion Engineerging Test Reactor,CFETR)为对象,开展OpenMC在聚变中子学分析中的应用研究。基于CFETR一维柱壳模型验证OpenMC与MCNP计数结果的一致性。根据等离子体空间分布特点,基于源扩展接口自定义源类和源函数准确描述复杂聚变中子源。利用DAG-OpenMC的CAD几何功能成功建立了CFETR的三维模型,并计算获得了中子壁负载分布、氚增殖率和核热沉积等物理量。结果表明:DAG-OpenMC与MCNP的计算结果具有极好的一致性。在建立复杂的聚变堆工程模型时,基于CAD几何功能极大地提高了建模效率。DAG-OpenMC在聚变中子学应用中关键问题的验证表明了其处理复杂工程结构条件下聚变中子学问题的可行性。
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.
文摘The reactivity of a nuclear reactor is the most important safety and operating parameter. Due to short reactor period, the Light Water Reactor(LWR) designs require the compensations of rapid unfavorable reactivity increases. The increase in fuel or moderator temperature leads to compensate the reactivity jumps as inherent safety characteristics. The safe and reliable reactor operation requires the accurate assessment of these reactivity changes. This paper highlights the improvements in the methodology to determine the feedback reactivity changes in IAEA MTR benchmark. This method incorporates the reactivity effects of fuel temperature in moderator regions and vice versa. For this purpose, a detailed 3D model of the IAEA 10 MW MTR benchmark reactor is developed employing OpenMC computer code. OpenMC is a probabilistic computer code for neutronic calculations. This work uses temperature-dependent JEFF 3.2 cross-sectional library. The model is validated against the reference results of eigenvalues for control rods(inserted and in fully withdrawn position), control rod reactivity worth, averaged thermal flux in the central flux trap, and power fraction for each fuel element at beginning of life. The validated model is applied to simulate the feedback reactivity coefficients against the conventional reference results. In order to improve the methodology, the effect of the moderator temperature and void on fuel is incorporated to obtain a more realistic value of the fuel temperature coefficient.Similarly, the moderator temperature coefficient and void coefficient are improved by incorporating the coupling effects of fuel temperature on moderator. This methodology can be applied to improve the LWR designs.
基金This research was supported by the Exascale Computing Project(17-SC-20-SC),a collaborative effort of the U.S.
文摘The physical quantities calculated by nuclear reactor Monte Carlo simulations are typically recorded on a grid of two or three spatial dimensions and one dimension of neutron energy.Because of this,increasing the resolution of the calculated quantities can have a significant impact on the memory and CPU time required to run a simulation.Convolutional neural networks have been shown to accurately upsample coarse-resolution photo-graphic images to resolutions multiple times finer than the originals.Here we show that a convolutional neural network can accurately upsample flux tallies in a Monte Carlo neutron transport simulation by a factor of two along the spatial and energy dimensions.Neutron flux tallies in pressurized water reactor assemblies were calculated using OpenMC at a 64×64 pixel spatial resolution and 8 neutron energy groups for input to the neural network.The network upsamples the low-resolution neutron flux to 128×128 pixel spatial resolution and 16 neutron energy groups.High-resolution neutron flux tallies and their uncertainties were also calculated with OpenMC and compared with the network’s predictions.The upsampled data and the high-resolution tally results agree to within the statistical uncertainty calculated by OpenMC.
文摘无慢化罐式堆芯结构的熔盐快堆(Molten Salt Fast Reactor,MSFR)中存在中子物理与热工水力的强耦合。应用耦合蒙特卡罗粒子输运程序OpenMC与计算流体力学软件OpenFOAM,建立了一套适用于熔盐快堆的三维稳态核热耦合计算程序。该程序基于python编程语言实现了OpenMC和OpenFOAM二者间的功率、燃料盐温度和密度分布等数据交互,可以获得堆芯内三维功率分布、中子通量密度分布、三维速度场和温度场分布。采用该耦合程序,建立了熔盐快堆的基准模型,研究了中子学区域划分数目和初始条件对keff、燃料盐速度和温度分布的影响。根据研究结果,推荐了一套合理的中子学区域划分方法与数目,表明了耦合程序设定的不同初始条件对keff结果无影响。最后,通过与熔盐快堆基准结果的对比验证了耦合程序的正确性,表明该程序适用于熔盐快堆的稳态核热耦合分析。
