多群蒙特卡罗程序MCMG的开发与基准校验

基于连续能量蒙特卡罗程序MCNP开发了多群蒙特卡罗程序MCMG.利用由栅元程序WIMS产生的随燃耗变化的多群宏观均匀化截面取代连续能量点截面,大大提高了程序的计算速度,同时也解决了蒙特卡罗程序不能进行燃耗计算等问题.针对输运修正引起的自散射截面导致的负概率抽样现象,提出了一种非负修正方法,并用基准计算验证了该方法的正确性.     

MCMG蒙特卡罗多群-连续截面耦合中子输运计算

本文针对多群蒙特卡罗计算省时但共振自屏处理存在缺陷,以及连续截面蒙特卡罗输运计算精度高但计算费时的问题,发展了一种多群-连续截面耦合计算方法。该方法在自主研发的三维中子-光子耦合输运蒙特卡罗程序MCMG中得到应用,通过多个模型的计算验证了方法的有效性。MCMG耦合计算取得了与连续点截面MCNP程序一致的结果,其计算速度较MCNP的提高了1倍左右。     

三维多群P5中子输运蒙特卡罗程序MCMG及检验

本工作介绍了自主开发研制的三维多群P₅中子输运蒙特卡罗程序MCMG及从ENDF/B-Ⅶ库制作的47群P₅中子截面库G47B7P5N。MCMG程序发展了针对物质的碰撞机制,适合ANISN格式和非标准ANISN格式的多群中子截面。程序计算了6个临界基准模型和2个外源问题,模拟取得了与实验和连续截面MCNP程序一致的结果,计算速度较MCNP程序提高3倍以上。     

Bamboo-Lattice程序多群截面数据库的改进

本文基于ENDF/B-Ⅶ.0核评价数据库,利用核数据加工处理程序NJOY及LATTICE_PRE为Bamboo-Lattice程序研制了一套改进后的多群截面数据库NECL2.0。基于基准题和数值分析的结果表明：采用NECL2.0数据库计算得到的燃料组件的k_inf、裂变率分布、少群均匀化截面与参考解均吻合很好;考虑银铟镉共振对k_inf的计算精度可提高近1000 pcm,与参考解相比最大裂变率相对偏差从-0.97%降低到-0.53%;考虑包壳锆的共振对k_inf的计算精度可提高约60 pcm。     

ENDF/B-Ⅶ.0多群截面库MUSE1.0在压力容器基准实验分析中的初步应用

基于ENDF/B-Ⅶ.0评价核数据库开发了172n×42g多群截面数据库MUSE1.0,利用二维离散纵标法程序DORT,针对美国H．B．Robinson-2号机组压力容器基准实验,对辐照监督管处中子能谱、核反应率及比活度等参数进行了详细的计算分析,并与基于ENDF/B-Ⅵ的BUGLE-96多群参数库计算结果及实验测量值进行了比较分析。结果表明：MUSE1.0比活度计算值与实验测量值之比（C/M）平均为0.98±0.04,较BUGLE-96计算结果（平均C/M为0.90±0.04）精度有较大提高,满足压水堆压力容器快中子注量计算精度要求。     

基于CENDL-3.2的多群截面库含Be快临界基准分析

中国评价核数据库最新版CENDL-3.2(Chinese Evaluated Nuclear Data Library)已于2020年6月发布,对包括核工程计算中常用的235U、238U、239Pu、56Fe等134个核素的中子反应数据重新进行了评价和计算,与CENDL-3.1相比,CENDL-3.2数据种类和数据质量均有大幅提高。Be由于其散射截面大、吸收截面小,常被用作熔盐堆燃料载体盐成分之一,其反应截面数据的准确性在熔盐堆设计中不容忽视。基于CENDL-3.2评价核数据库,采用NJOY制作了199群中子、42群光子的MATXS格式多群截面库,挑选了35个含Be快临界基准对其进行检验分析,并与基于ENDF/B-7.1和JENDL-4.0的多群截面库计算结果进行对比。分析表明：基于CENDL-3.2多群截面库计算的26个基准题（74.29%）的结果与实验值偏差在0.5%以内,整体上优于ENDF/B-7.1和JENDL-4.0。表明CENDL-3.2中的Be数据和基于CENDL-3.2的多群截面库及其制作方法是可靠的,能够用于熔盐堆相关设计计算。     

厚活化箔群截面的加工制作方法

在利用活化法进行中子能谱测量中,活化箔的多群截面是必须的输入参数之一.本工作研究利用原始的评价核数据库数据进行群截面加工制作技术,提出对不同厚度的活化箔,特别是对于吸收和散射截面较大的核反应,需对活化箔无限稀释的群截面作相应的修正,不能盲目套用文献上已有或别人解谱中用到的群截面.本工作综合考虑Cd及活化箔本身对中子的吸收和散射作用,首次提出采用MCNP程序对包Cd后不同厚度活化箔的群截面进行加工.在实验基础上,对加工得到的群截面加工方法及准确度进行了检验.     

基于OpenMC的多群截面库制作及有效性验证

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。     

多群常数库NuDa-SN主库的研制

为给中物院高性能数值模拟软件中心自主研发的三维离散纵标（S_N）输运软件JSNT-S制作配套的多群常数库NuDa-SN,制定了主库与工作库配合的基本路线。其中,主库采用NJOY2016与CMiler程序基于ENDF/B-Ⅶ.1评价数据库制作。主库包含455个核素,中子能群数为199,光子能群数为42。为验证主库的可靠性,分别对主库进行了单核素测试、临界基准检验以及屏蔽基准检验,所有测试表明主库在临界计算和屏蔽计算方面具有相当的可靠性。     

三维多群中子输运蒙特卡罗程序MCMG-Ⅱ基准检验

三维多群P₃中子输运蒙特卡罗程序MCMG通过版本更新和功能扩充,能够配备各种多群微观、宏观截面库,截面输入文件进一步简化,版本从Ⅰ发展到Ⅱ,参数更新到ENDF/B-Ⅶ库。程序发展了针对物质的碰撞机制,具有并行计算功能。对于12个临界基准问题和1个外源问题,MCMG-Ⅱ计算取得了与实验和连续截面MCNP 5程序一致的结果,计算速度较MCNP-5提高了3~6倍。     

几种特殊栅元群截面库的近似处理

本文以后处理厂溶解槽、含拉西环溶液为例，通过分析其栅元特性，采取一系列合理近似，在现有的ＡＭＰＸ群截面库计算系统的基础上，寻找较为理想的栅元共振自屏及截面权重平均的近似处理法，为最终迁移计算（或蒙特－卡洛模拟计算）提供较为满意的群截面库。从而为这类非常规栅元系统的临界安全分析开辟一条可靠途径。     

核数据处理程序NECP-Atlas的开发与验证

本文开发了自主化的核数据处理程序NECP-Atlas,该程序将不同的核数据处理功能封装为不同的程序模块,可将评价核数据经过共振重构及线性化、多普勒展宽计算、不可分辨共振区处理、热中子散射计算、多群截面计算等过程,处理为WIMS-D/E格式多群数据库。采用WLUP（WIMSD library update project）基准题、国际临界安全基准题ICSBEP（international criticality safety benchmark evaluation project）等对NECP-Atlas加工产生的核数据进行验证,结果显示NECP-Atlas和NJOY-2016程序精度相当。     

基于改进缺陷因子方法的FCM燃料有效多群截面计算方法研究

全陶瓷微胶囊封装（FCM）燃料是重要的候选事故容错燃料,与传统燃料相比,FCM燃料的双重非均匀性使得其有效多群截面计算面临较大的挑战。本文提出一种改进的缺陷因子方法来处理FCM燃料在共振能区和非共振能区的自屏效应,实现FCM燃料的等效均匀化。通过颗粒丹可夫因子守恒来构建新的等效模型以克服传统的体积权重等效模型无法考虑燃料棒间自屏的影响;在共振能量段,基于新的等效一维球模型求解超细群慢化方程获得共振能量段的超细群缺陷因子;在非共振能量段,利用新等效模型的特征值计算获得快群和热群的多群缺陷因子;在此基础上实现FCM燃料棒的等效均匀化。本方法已在高保真中子学程序NECP-X上实现,并在一系列工况下进行了测试,与蒙特卡罗程序的比较表明,本方法能处理不同情况下的双重非均匀性,并可获得准确的有效自屏截面。     

Adjustment of Group Cross Sections by Means of Integral Data, (II)

Adjustment of the ABBN set is performed using various integral data obtained from fast critical experiments; through this adjustment, some practical properties are also examined in detail. The correlation coefficient between group cross sections is numerically obtained by assuming that the compound nucleus formation cross section can be described by a statistical model. The effect brought by this correlation to the adjustment of group cross sections is also studied. When the differential and integral data used in the adjustment contain a systematic error, the normalized sum of squares of residuals has a non-central Chi-square distribution. This is numerically examined by generating artificial differential and integral data with the aid of random numbers. The ABBN set collapsed into 15 energy groups is adjusted, and the principal results are compared with the measurements. The reliability of the adjusted integral data and group cross sections is also studied.     

Expression of Few-Group Effective Microscopic Cross Sections in BWR Burnup Calculations

A practical method is proposed to express few-group effective microscopic cross sections for BWR burnup analysis. A set of few-group cross sections is prepared for an infinite square lattice of fuel rods as a function of the ratios of number density of nuclides such as ²³⁵ U, ²³⁸U and ²³⁹Pu, and the water quantity around a fuel rod. Spatial variation of few-group cross sections in the fuel assembly is taken into account by adjusting the water quantity around a fuel rod.

Numerical studies show that the present method can evaluate effective few-group cross sections within the accuracy of 3% in comparison with a two-dimensional integral transport calculation.     

Interference Effect of Strong Scattering Resonances on Elastic Removal Cross Sections

The interference effect from the strong scattering resonance on the elastic removal cross sections was investigated for the core composition of fast reactors. To find the relation of the group elastic removal cross sections of background nuclides to the scattering property of the mixture, they were numerically calculated in the energy ranges of typical resonances. It was shown that the interference effect could be taken into account in the conventional group constant set by using two different methods.

In the first method, the shielding foctors of the background nuclides were characterized by the parameters ξ and σ₀ for the resonance nuclide. The interpolation scheme was similar to that adopted in the conventional set. In the second method, the shielding factors of the imaginary background nuclide were linearly fitted to those of the resonance nuclide. The composition dependence of macroscopic cross sections could be easily obtained with use of the linear relation of which coefficients were determined, beforehand, for the typical core composition. The accuracy of the second method was examined by comparing with the exact values. The present method could predict the macroscopic elastic removal cross sections within the errors of several percents.     

Benchmark Analysis of Criticality Experiments in the TRIGA Mark II Using a Continuous Energy Monte Carlo Code MCNP

The criticality analysis of the TRIGA-II benchmark experiment at the Musashi Institute of Technology Research Reactor (MuITR, 100kW) was performed by the three-dimensional continuous-energy Monte Carlo code (MCNP4A). To minimize errors due to an inexact geometry model, all fresh fuels and control rods as well as vicinity of the core were precisely modeled. Effective multiplication factors (k_eff) in the initial core critical experiment and in the excess reactivity adjustment for the several fuel-loading patterns as well as the fuel element reactivity worth distributions were used in the validation process of the physical model and neutron cross section data from the ENDF/B-V evaluation. The calculated k_eff overestimated the experimental data by about 1.0Δk/k for both the initial core and the several fuel-loading arrangements (fuels or graphite elements were added only to the outer-ring), but the discrepancy increased to 1.8Δk/k for the some fuel-loading patterns (graphite elements were inserted into the inner-ring). The comparison result of the fuel element worth distribution showed above tendency. All in all, the agreement between the MCNP predictions and the experimentally determined values is good, which indicates that the Monte Carlo model is enough to simulate criticality of the TRIGA-II reactor.     

Measurement of Effective Capture Cross Section of Americium-243 for Thermal Neutrons

The effective capture cross section of ²⁴³Am for thermal neutrons was measured with an activation method. A sample of ²⁴³Am was irradiated for 10 hrs at Kyoto University Reactor, KUR. After the irradiation, the sample was cooled for one month. In the cooling time, ^244mAm and ^244gAm produced by the irradiation decayed out to ²⁴⁴Cm. The α rays emitted from ²⁴³Am and ²⁴⁴Cm were measured with a silicon surface barrier detector. The neutron flux at the irradiation position was monitored using Au/Al and Co/Al wires. The effective capture cross section was deduced as 174.5±5.3b from the α-ray counts and the neutron flux. The quantity r√T/T₀ in Westcott's convention was 0.037±0.004. The present result was compared with the effective capture cross sections from the JENDL-3.3 and the Mughabghab evaluations.