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为了研究利用西安脉冲堆(XAPR)热中子开展99Tc、129I嬗变的可行性,对乏燃料中长寿命裂变产物(LLFP)99Tc和129I核素的热中子嬗变计算方法进行理论与实验研究。利用NJOY程序,以ENDF/B VII.0库为基础,制作99Tc和129I在XAPR堆芯辐照温度下的蒙特卡罗程序(MCNP)截面库,并分析不同参数对截面数据的影响。采用ACE(A Compact ENDF)格式截面库和燃耗程序CINDER’90自带的63群活化截面,利用MCNP程序对ORIGEN2数据库中99Tc和129I的辐射俘获截面进行修正,用ORIGEN2程序分析一定规格的99Tc和129I靶件在XAPR内辐照后的嬗变情况。与实验结果值进行比较,截面数据的差异主要来自中子注量率测量值与实际值的误差,结果证明利用XAPR开展99Tc和129I嬗变是可行的。 相似文献
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为满足加速器驱动次临界系统ADS等高能核装置的蒙特卡罗输运计算需求,通过对相关核数据处理模块的改进,建立了一套基于我国自主的群常数制作软件Ruler与国际公认的核数据处理系统NJOY耦合的中高能评价核数据处理方法及程序系统。通过该方法基于日本高能评价核数据库JENDL-HE-2007制作了ACE格式连续点截面库,并通过一系列绘图及简单问题的蒙特卡罗输运计算,验证了该库是完整的、合理的,可用于蒙特卡罗输运计算,证明了Ruler与NJOY功能模块耦合的方法可用于高能评价核数据的处理。 相似文献
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MCNP5程序可以用于电子加速器驱动次临界系统的建模运算,其调用的截面数据库缺乏部分核素的光核数据。利用NJOY程序将ENDF/B-VII中的7种核素原始光核数据制作成MCNP5可利用的ACE格式的光核数据,并用MCNP5建立模型计算出所加工核素的光中子反应微观截面,得到光中子反应微观截面随光子能量变化曲线,并与IAEA编写的技术文档中各核数据中心的截面曲线进行对比。结果表明,所有核素光核反应数据制作过程正确,但52Cr、58Ni和91Zr的光中子反应微观截面在不同数据库中有一定的差别。 相似文献
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随着各国新版本评价核数据库的发布,不同版本评价核数据库对反应性的影响并不完全一致,为选取高精度的评价核数据库,以几何尺寸、核素种类较简单的压水堆包壳材料为研究对象,基于新版本的评价核数据库ENDF/B-VII.0、JEFF-3.3、JENDL-4.0和CENDL-3.1,采用NJOY2016程序制作压水堆常见包壳材料(不锈钢包壳、铝包壳和锆包壳)的截面数据。通过组件程序DRAGON5.0.1挂载不同评价核数据库版本得到包壳材料的多群截面库,计算WIMS库更新计划(WLUP)系列临界基准题,并将计算结果与实验值进行比较。结果表明,52Cr、56Fe、90Zr、91Zr、92Zr和94Zr这6个核素在不同评价核数据库版本中对反应性影响均较大;采用CENDL-3.1和JENDL-4.0这2个版本评价核数据库制作的压水堆包壳材料,其计算结果与实验值较为接近。 相似文献
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采用国际公认的群常数制作理论方法,包括共振重造方法、多普勒展宽方法、热散射率处理方法、群截面和散射矩阵计算方法、共振自屏处理方法等,研发了包括主驱动程序、评价数据输入输出模块、公共数学模块、系统公共子程序模块、进制转换模块、截面线性化和共振重造模块、截面温度展宽模块、不可分辨共振自屏模块、热散射截面计算模块、中子多群常数计算模块、WIMS-D格式接口模块等11个模块的群常数制作软件Ruler。采用与国际通用核数据处理程序NJOY99比较的方式对Ruler进行了验证,包括群常数比较和基准检验结果比较。验证结果表明,Ruler的计算精度与NJOY99相当,其计算速度、可维护性、可扩展性优于NJOY99。 相似文献
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OpenMC是麻省理工大学计算反应堆物理组开发的开源蒙特卡罗程序,能够方便地制作适用于特定堆芯中子能谱分布的多群反应截面及高阶勒让德散射截面以用于离散坐标输运程序ANISN的计算。本文基于ENDF/B-Ⅶ.1和CENDL-3.1评价数据库,利用OpenMC计算制作了ANSIN格式的多群截面并通过基准题的计算验证计算结果的准确性。通过截面转换程序的编写,将OpenMC给出的堆芯各阶勒让德散射分量,堆芯中子能谱分布,散射、吸收反应率以及裂变中子产生速率等信息转换为ANISN程序可读取的截面库格式。采用制作的截面库利用ANINS计算有效中子增殖因子及堆芯中子通量分布。结果表明,ANISN确定论的计算结果与OpenMC给出的蒙特卡罗计算结果相吻合,验证了这种方法可有效地为ANISN提供截面数据,将来可推广应用于二维、三维确定论中子输运计算。 相似文献
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NECP-Atlas是西安交通大学自主研制的核数据处理软件,具有丰富的功能,可将评价核数据制作为后续核设计所需的应用核数据库,本文在NECP Atlas中建立了光子相关数据的计算方法,可计算产生中子核反应释放的瞬发光子产生截面、光子与原子的反应截面,裂变产物衰变释放的缓发光子多群产生矩阵,以及光子辐照损伤截面等数据。数值结果显示,如果不考虑缓发光子,钠冷快堆中控制组件、反射层组件的光子功率与参考解的最大偏差可达3258%、2041%,采用NECP Atlas计算的多群缓发光子产生矩阵后两类组件偏差降为093%以下。采用文献结果对Fe的光子辐照损伤截面进行了验证,计算结果与参考解吻合良好。 相似文献
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R.E. MacFarlane 《Nuclear Data Sheets》2010,111(12):2739-2890
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Hatun Korkut Turgay Korkut Ayhan Kara Mustafa Yiğit Eyyup Tel 《Journal of Fusion Energy》2016,35(3):591-596
Advanced fusion structural materials (FSMs) have impact role in terms of efficiency of nuclear energy production. Besides engineering and design of fusion reactors, radiation durability of FSMs is another valuable issue that cannot be ignored. 17.9–22.3 MeV proton irradiation of bcc-Zirconium Fusion Structural Material was evaluated by using Monte Carlo based simulation tools. Total binary reaction cross sections were respectively calculated as 1167.6 and 1273.92 mb for 17.3- and 22.3 MeV proton energies via TALYS-1.6 version. Additionally, residual production cross sections and total particle production cross sections were obtained and analyzed by the TALYS code. Radiation damage parameters as Displacement Per Atom (DPA) and Stopping Power (SP) were studied by SRIM-2013 version. FLUKA 2011.1 used for only DPA calculations and making a complete comparison with the other calculation results. SP and Number of Secondaries were found by using GEANT4.10.p.04 version simulations. Natural Zr(p,x) reactions were studied in the given energy values in the plane of reaction probability and radiation damage calculations. 相似文献
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为弥补国内在由评价核数据出发计算中子比释动能(KERMA)系数和辐射损伤截面的程序方面的空白,建立了中子KERMA系数和辐射损伤截面计算方法,并基于FORTRAN-90程序语言开发了具有自主知识产权的中子KERMA系数和辐射损伤截面计算程序KDC。另外,针对能量平衡检查过程中发现的能量不平衡问题,提出了一种对不合理KERMA系数进行直接修订的方法,即用运动学上限替代不合理KERMA系数,并在KDC程序中实现了这一修订功能。通过将KDC程序与国际上广泛应用的核数据处理程序系统NJOY中的HEATR模块的计算结果进行比对,验证了KDC程序在计算结果和功能上的可靠性。 相似文献
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为满足聚变 裂变次临界混合堆设计和其他相关研究的需要 ,以世界几个主要基本评价核数据库为数据来源 ,通过优选建立了名为HENDL1 .0 /E的多用途核数据库 ,采用国际通行的核数据库处理程序系统NJOY和TRANSX等程序制作了相应的工作数据库 ,其中包括多能群输运截面库HENDL1 .0 /MG、连续能量点状输运截面库HENDL1 .0 /MC、燃耗数据库HENDL1 .0 /BU和响应函数库HENDL1 .0 /RF ,利用世界上流行的中子输运程序对已有的一系列基准检验实验进行模拟计算和比较分析以检验混合库HENDL1 .0的正确性和有效性。 相似文献
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This work was focused on the neutronic calculation of the nuclear parameters (neutron spectrum, displacement per atom (DPA), gas production, tritium breeding ratio (TBR), nuclear heating) for structural materials in the first wall (FW) and fuel clad (made of ferritic/martensitic steels, vanadium alloy, silicon carbide, copper alloy, and stainless steel) of an experimental hybrid reactor using the most current Monte Carlo Neutron-Particle Transport code MCNP5 1.4. Neutronic calculations were performed using a (DT) fusion driver hybrid reactor under a neutron wall loud of 2.25 MW/m2 by full reactor power for one year. Obtained results were compared with three different data libraries (ENDF/B-V, ENDF/B-VI and CLAW-IV). TBR values in the reactor blanket for all investigated materials became greater than the minimum requirement (TBR > 1.05). Nuclear parameters like DPA, He-production and nuclear heating were considered as radiation damage limits for structural materials, copper alloy (Cu0.5Cr0.3Zr) showed better performance than all investigated materials. 相似文献
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The aim of this study is to analyze the neutronic parameters of TRIGA Mark-II research reactor using the chain of NJOY-WIMS-CITATION computer codes based on evaluated nuclear data libraries CENDL-2.2 and JEFF-3.1.1. The nuclear data processing code NJOY99.0 has been employed to generate the 69 group WIMS library for the isotopes of TRIGA core. The cell code WIMSD-5B was used to generate the cross sections in CITATION format and then 3-dimensional diffusion code CITTATION was used to calculate the neutronic parameters of the TRIGA Mark-II research reactor. All the analyses were performed using the 7-group macroscopic cross section library. The CITATION test-runs using different cross section sets based on different models applied in WIMS calculations have shown a strong influence of those models on the final integral parameters. Some of the cells were specially treated with PRIZE options available in WIMSD-5B to take into account the fine structure of the flux gradient in the fuel-reflector interface region. It was observed that two basic parameters, the effective multiplication factor, keff and the thermal neutron flux, were in good agreement among the calculated results with each other as well as the measured values. The maximum power densities at the hot spot were 1.0446E02 W/cc and 1.0426E02 W/cc for the libraries CENDL-2.2 and JEFF-3.1.1 respectively. The calculated total peaking factors 5.793 and 5.745 were compared to the original SAR value of 5.6325 as well as MCNP result. Consequently, this analysis will be helpful to enhance the neutronic calculations and also be used for the further thermal–hydraulics study of the TRIGA core. 相似文献
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《Journal of Nuclear Science and Technology》2013,50(9):1131-1141
New evaluation of neutron-induced nuclear data for five stable isotopes of zinc (mass numbers A = 64, 66, 67, 68, and 70) was consistently carried out in the incident neutron energy range from 10?5 eV up to 20MeV. In the low energy region up to about 100keV, the resonance parameters were evaluated by taking account of the available measured data. In the fast neutron region, the comprehensive calculations with nuclear reaction models, in which compound, preequilibrium, and direct processes are taken into account, were performed to estimate cross sections for various reactions and double differential cross sections of emitted neutrons and γ-rays. The comparisons of the evaluated cross sections with the experimental data and existing evaluated nuclear data libraries are made and show a good agreement with the measurements. 相似文献