The heterogeneous variational nodal method(HVNM)has emerged as a potential approach for solving high-fidelity neutron transport problems.However,achieving accurate results with HVNM in large-scale problems using high-...The heterogeneous variational nodal method(HVNM)has emerged as a potential approach for solving high-fidelity neutron transport problems.However,achieving accurate results with HVNM in large-scale problems using high-fidelity models has been challenging due to the prohibitive computational costs.This paper presents an efficient parallel algorithm tailored for HVNM based on the Message Passing Interface standard.The algorithm evenly distributes the response matrix sets among processors during the matrix formation process,thus enabling independent construction without communication.Once the formation tasks are completed,a collective operation merges and shares the matrix sets among the processors.For the solution process,the problem domain is decomposed into subdomains assigned to specific processors,and the red-black Gauss-Seidel iteration is employed within each subdomain to solve the response matrix equation.Point-to-point communication is conducted between adjacent subdomains to exchange data along the boundaries.The accuracy and efficiency of the parallel algorithm are verified using the KAIST and JRR-3 test cases.Numerical results obtained with multiple processors agree well with those obtained from Monte Carlo calculations.The parallelization of HVNM results in eigenvalue errors of 31 pcm/-90 pcm and fission rate RMS errors of 1.22%/0.66%,respectively,for the 3D KAIST problem and the 3D JRR-3 problem.In addition,the parallel algorithm significantly reduces computation time,with an efficiency of 68.51% using 36 processors in the KAIST problem and 77.14% using 144 processors in the JRR-3 problem.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 construction of China Spallation Neutron Source （CSNS） has been initiated in Dongguan, Guangdong, China. Thus a detailed radiation transport analysis of the shutter neutron beam stop is of vital importance. The ...The construction of China Spallation Neutron Source （CSNS） has been initiated in Dongguan, Guangdong, China. Thus a detailed radiation transport analysis of the shutter neutron beam stop is of vital importance. The analyses are performed using the coupled Monte Carlo and multi-dimensional discrete ordinates method. The target of calculations is to optimize the neutron beamline shielding design to guarantee personal safety and minimize cost. Successful elimination of the primary ray effects via the two-dimensional uncollided flux and the first collision source methodology is also illustrated. Two-dimensional dose distribution is calculated. The dose at the end of the neutron beam line is less than 2.5 μSv/h. The models have ensured that the doses received by the hall staff members are below the standard limit required.展开更多
基金supported by the National Key Research and Development Program of China(No.2020YFB1901900)the National Natural Science Foundation of China(Nos.U20B2011,12175138)the Shanghai Rising-Star Program。
文摘The heterogeneous variational nodal method(HVNM)has emerged as a potential approach for solving high-fidelity neutron transport problems.However,achieving accurate results with HVNM in large-scale problems using high-fidelity models has been challenging due to the prohibitive computational costs.This paper presents an efficient parallel algorithm tailored for HVNM based on the Message Passing Interface standard.The algorithm evenly distributes the response matrix sets among processors during the matrix formation process,thus enabling independent construction without communication.Once the formation tasks are completed,a collective operation merges and shares the matrix sets among the processors.For the solution process,the problem domain is decomposed into subdomains assigned to specific processors,and the red-black Gauss-Seidel iteration is employed within each subdomain to solve the response matrix equation.Point-to-point communication is conducted between adjacent subdomains to exchange data along the boundaries.The accuracy and efficiency of the parallel algorithm are verified using the KAIST and JRR-3 test cases.Numerical results obtained with multiple processors agree well with those obtained from Monte Carlo calculations.The parallelization of HVNM results in eigenvalue errors of 31 pcm/-90 pcm and fission rate RMS errors of 1.22%/0.66%,respectively,for the 3D KAIST problem and the 3D JRR-3 problem.In addition,the parallel algorithm significantly reduces computation time,with an efficiency of 68.51% using 36 processors in the KAIST problem and 77.14% using 144 processors in the JRR-3 problem.
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 National Natural Science Foundation of China (10875042,10705011)Program for Changjiang Scholars and Innovative Research Team at University (IRT0720)Beijing Science New Star Plan Project (2007B058)
文摘The construction of China Spallation Neutron Source （CSNS） has been initiated in Dongguan, Guangdong, China. Thus a detailed radiation transport analysis of the shutter neutron beam stop is of vital importance. The analyses are performed using the coupled Monte Carlo and multi-dimensional discrete ordinates method. The target of calculations is to optimize the neutron beamline shielding design to guarantee personal safety and minimize cost. Successful elimination of the primary ray effects via the two-dimensional uncollided flux and the first collision source methodology is also illustrated. Two-dimensional dose distribution is calculated. The dose at the end of the neutron beam line is less than 2.5 μSv/h. The models have ensured that the doses received by the hall staff members are below the standard limit required.
