燃耗数据库的精准性是影响燃耗计算结果精准性的主要因素。本文开发了燃耗库加工程序PSTL,并基于评价核数据库ENDF/B-VIII.0、EAF-2010和TENDL-2019制作了一个用于ORIGEN-S程序的压水堆燃耗数据库。该数据库包含两个子库:中子截面-裂变...燃耗数据库的精准性是影响燃耗计算结果精准性的主要因素。本文开发了燃耗库加工程序PSTL,并基于评价核数据库ENDF/B-VIII.0、EAF-2010和TENDL-2019制作了一个用于ORIGEN-S程序的压水堆燃耗数据库。该数据库包含两个子库:中子截面-裂变产额库和衰变库。中子截面数据首先应用NJOY2016程序将ENDF/B-VIII.0等评价库中的点连续截面处理为172群截面,然后采用指定燃耗深度下的组件或栅元中子能谱将172群截面并为三群截面;裂变产额数据取自MF8/MT454和MT459;衰变数据取自MF8/MT457。最后采用OECD/NEA机构的压水堆核素积存量基准题进行验证分析,结果表明:中子能谱对燃耗库以及核素积存量有显著影响,对于238Pu、237Np和147Sm等核素,由组件中子能谱所加工的燃耗库的计算结果更接近实验值,提高了ORIGEN-S的计算精度。The accuracy of burnup library is the main factor affecting the accuracy of burnup calculation results. In this paper, the library processing program PSTL is developed, and a PWR burnup library for ORIGEN-S program is made based on the evaluated nuclear libraries ENDF/B-VIII.0, EAF-2010 and TENDL-2019. The library contains two sublibraries, namely neutron cross section-fission yield library and decay library. First, the point continuous cross sections in ENDF/B-VIII.0 evaluated library and so on are processed into 172-group cross sections by NJOY2016 program. Then the 172-group cross section is further merged into three-group cross sections by using the neutron spectrum of the assembly or cell at the specified burnup depth. Fission yield data are taken from MF8/MT454 and MT459. Decay data are taken from MF8/MT457. Finally, the benchmark of nuclide inventory in PWR of OECD/NEA is used to verify and analyze. The results show that the neutron energy spectrum has a significant effect on burnup library and nuclide inventory. For such nuclides as 238Pu, 237Np and 147Sm, the calculated results of the burnup library processed by the assembly neutron spectrum are closer to the experimental values, which improves the calculation accuracy of ORIGEN-S.展开更多
An accelerator-driven subcritical system(ADS)is driven by an external spallation neutron source, which is generated from a heavy metal spallation target to maintain stable operation of the subcritical core, where the ...An accelerator-driven subcritical system(ADS)is driven by an external spallation neutron source, which is generated from a heavy metal spallation target to maintain stable operation of the subcritical core, where the energy of the spallation neutrons can reach several hundred megaelectron volts. However, the upper neutron energy limit of nuclear cross-section databases, which are widely used in critical reactor physics calculations, is generally 20 MeV.This is not suitable for simulating the transport of highenergy spallation neutrons in the ADS. We combine the Japanese JENDL-4.0/HE high-energy evaluation database and the ADS-HE and ADS 2.0 libraries from the International Atomic Energy Agency and process all the data files for nuclides with energies greater than 20 MeV. We use the continuous pointwise cross-section program NJOY2016 to generate the ACE-formatted cross-section data library IMPC-ADS at multiple temperature points. Using the IMPC-ADS library, we calculate 10 critical benchmarks of the International Criticality Safety Benchmark Evaluation Project manual, the 14-MeV fixed-source problem of the Godiva sphere, and the neutron flux of the ADS subcritical core by MCNPX. To verify the correctness of the IMPCADS, the results were compared with those calculated using the ENDF/B-VII.0 library. The results showed thatthe IMPC-ADS is reliable in effective multiplication factor and neutron flux calculations, and it can be applied to physical analysis of the ADS subcritical reactor core.展开更多
核数据库是中子输运计算的基础。基于ENDF/B-Ⅶ.1评价库,采用NJOY制作了用于MCNP(Monte Carlo N-Particle Transport Code)程序的AHD1.0(Advanced hybrid database1.0)库,并从国际临界核安全手册(International Handbook of Evaluated C...核数据库是中子输运计算的基础。基于ENDF/B-Ⅶ.1评价库,采用NJOY制作了用于MCNP(Monte Carlo N-Particle Transport Code)程序的AHD1.0(Advanced hybrid database1.0)库,并从国际临界核安全手册(International Handbook of Evaluated Criticality Safety Benchmark Experiments,ICSBEP)中分别选取了高浓铀、中浓铀、低浓铀的快谱、中间谱及热谱的部分基准装置,用MCNP程序调用该数据库进行了临界基准验证,验证结果显示:调用该库的计算值与实验值符合较好,误差在0.5%以内,具有较高的精确度,满足核设计对数据库精度的要求。但对于部分含有W、Fe、Gd等结构材料、吸收材料的基准检验中,存在较大的偏差,造成这些偏差的主要原因是计算过程中核素的处理及评价数据库的来源,需要进一步的研究验证。展开更多
文摘燃耗数据库的精准性是影响燃耗计算结果精准性的主要因素。本文开发了燃耗库加工程序PSTL,并基于评价核数据库ENDF/B-VIII.0、EAF-2010和TENDL-2019制作了一个用于ORIGEN-S程序的压水堆燃耗数据库。该数据库包含两个子库:中子截面-裂变产额库和衰变库。中子截面数据首先应用NJOY2016程序将ENDF/B-VIII.0等评价库中的点连续截面处理为172群截面,然后采用指定燃耗深度下的组件或栅元中子能谱将172群截面并为三群截面;裂变产额数据取自MF8/MT454和MT459;衰变数据取自MF8/MT457。最后采用OECD/NEA机构的压水堆核素积存量基准题进行验证分析,结果表明:中子能谱对燃耗库以及核素积存量有显著影响,对于238Pu、237Np和147Sm等核素,由组件中子能谱所加工的燃耗库的计算结果更接近实验值,提高了ORIGEN-S的计算精度。The accuracy of burnup library is the main factor affecting the accuracy of burnup calculation results. In this paper, the library processing program PSTL is developed, and a PWR burnup library for ORIGEN-S program is made based on the evaluated nuclear libraries ENDF/B-VIII.0, EAF-2010 and TENDL-2019. The library contains two sublibraries, namely neutron cross section-fission yield library and decay library. First, the point continuous cross sections in ENDF/B-VIII.0 evaluated library and so on are processed into 172-group cross sections by NJOY2016 program. Then the 172-group cross section is further merged into three-group cross sections by using the neutron spectrum of the assembly or cell at the specified burnup depth. Fission yield data are taken from MF8/MT454 and MT459. Decay data are taken from MF8/MT457. Finally, the benchmark of nuclide inventory in PWR of OECD/NEA is used to verify and analyze. The results show that the neutron energy spectrum has a significant effect on burnup library and nuclide inventory. For such nuclides as 238Pu, 237Np and 147Sm, the calculated results of the burnup library processed by the assembly neutron spectrum are closer to the experimental values, which improves the calculation accuracy of ORIGEN-S.
基金supported by the Strategic Priority Research Program of the Chinese Academy of Sciences(No.XDA03030102)
文摘An accelerator-driven subcritical system(ADS)is driven by an external spallation neutron source, which is generated from a heavy metal spallation target to maintain stable operation of the subcritical core, where the energy of the spallation neutrons can reach several hundred megaelectron volts. However, the upper neutron energy limit of nuclear cross-section databases, which are widely used in critical reactor physics calculations, is generally 20 MeV.This is not suitable for simulating the transport of highenergy spallation neutrons in the ADS. We combine the Japanese JENDL-4.0/HE high-energy evaluation database and the ADS-HE and ADS 2.0 libraries from the International Atomic Energy Agency and process all the data files for nuclides with energies greater than 20 MeV. We use the continuous pointwise cross-section program NJOY2016 to generate the ACE-formatted cross-section data library IMPC-ADS at multiple temperature points. Using the IMPC-ADS library, we calculate 10 critical benchmarks of the International Criticality Safety Benchmark Evaluation Project manual, the 14-MeV fixed-source problem of the Godiva sphere, and the neutron flux of the ADS subcritical core by MCNPX. To verify the correctness of the IMPCADS, the results were compared with those calculated using the ENDF/B-VII.0 library. The results showed thatthe IMPC-ADS is reliable in effective multiplication factor and neutron flux calculations, and it can be applied to physical analysis of the ADS subcritical reactor core.
文摘核数据库是中子输运计算的基础。基于ENDF/B-Ⅶ.1评价库,采用NJOY制作了用于MCNP(Monte Carlo N-Particle Transport Code)程序的AHD1.0(Advanced hybrid database1.0)库,并从国际临界核安全手册(International Handbook of Evaluated Criticality Safety Benchmark Experiments,ICSBEP)中分别选取了高浓铀、中浓铀、低浓铀的快谱、中间谱及热谱的部分基准装置,用MCNP程序调用该数据库进行了临界基准验证,验证结果显示:调用该库的计算值与实验值符合较好,误差在0.5%以内,具有较高的精确度,满足核设计对数据库精度的要求。但对于部分含有W、Fe、Gd等结构材料、吸收材料的基准检验中,存在较大的偏差,造成这些偏差的主要原因是计算过程中核素的处理及评价数据库的来源,需要进一步的研究验证。
