深穿透屏蔽计算中MCNP减方差技巧应用及比较

在中子输运的深穿透屏蔽计算中,采用有效的MCNP减方差技巧可以有效缩减计算时间提高计算效率。以一维聚变反应堆(CFETR)平板模型作为深穿透屏蔽计算实例,分别应用了几何分裂与轮盘赌、源偏倚、强迫碰撞、权窗、指数变换等减方差技巧,对各种减方差技巧应用过程中遇到的如中子平分布问题,全零权窗问题等问题,进行了分析总结,并在此基础上提出了针对深穿透问题的新的解决思路。通过不同方法分析和组合,获得了权窗和指数变换的最佳组合,使计算效率加速了162倍。     

基于CADIS方法的蒙特卡罗深穿透屏蔽计算方差降低方法研究

正深穿透问题是指由于屏蔽层过厚、源强较弱或探测器体积较小导致粒子输运计算结果偏低的问题。针对于蒙特卡罗屏蔽计算中的深穿透问题,利用一致共轭驱动重要性抽样(CADIS)方法的相应理论,实现了蒙特卡罗软件MCNP减小计算方差,可通过蒙特卡罗方法的共轭计算来得到共轭通量,经过CADIS方法推导出的公式计算得到权窗参数和源偏倚参数。     

蒙特卡罗输运计算中的散射方向偏倚

利用蒙特卡罗法不能直接计算复杂结构系统的中子输运,如屏蔽层的穿透问题,这是因为输运程序不可能预知中子穿透屏蔽层的适当路线。本研究工作对输运计算中的探测器贡献分布进行估计,并应用这一分布进行散射方向偏倚,使探测器能有效地探测到中子。在一些简单模型中采用该偏倚程序进行测试计算,证实了该程序的基本功能。本文所采用的偏倚程序是一个很强的偏倚工具,能用来计算厚而复杂的结构,如漏流效应、天空回散照射等等。     

基于三维MC-SN耦合方法的PWR压力容器快中子注量计算基准分析

蒙特卡罗(MC)-离散纵标(SN)耦合方法是解决同时具有复杂几何和深穿透特点的核装置屏蔽问题的有效方法。本文首次将三维MC-SN耦合方法应用于压水堆屏蔽计算。针对NUREG/CR-6115压水堆基准模型,选取热屏蔽内表面为公共交界面,将其分为几何复杂的MC模拟区和具有深穿透特点的SN模拟区。三维MC程序用于精确描述堆芯到热屏蔽精细模型,并记录穿过热屏蔽内表面的中子径迹信息。接口程序将中子径迹转换为SN计算所需的边界源,提供给三维SN程序进行热屏蔽到压力容器的计算。计算结果包括压力容器内表面、1/4壁厚处及焊缝处快中子注量(E>1.0 MeV)圆周方向分布。三维耦合方法计算结果与基准报告提供的MCNP、DORT结果符合良好,验证了该方法处理圆柱坐标系屏蔽问题的有效性和程序使用的正确性。     

高性能计算分析技术在大型钠冷快堆屏蔽设计中的应用

钠冷快堆堆容器是一体化的池式结构,由众多堆内构件组成且结构复杂,堆芯到生物屏蔽外中子输运过程中各向异性明显且深穿透问题严重,大尺度范围下三维S_N方法计算是制约快堆屏蔽设计的瓶颈。通过将三维SN程序与高性能计算技术相结合,采用并行计算方法可解决快堆堆本体内各向异性的三维深穿透屏蔽问题。本文以中国示范快堆（CFR600）堆本体为研究对象,采用JSNT-CFR程序详细计算了堆本体内的中子注量率、光子注量率、剂量率,并将计算结果与已有的二维程序设计结果进行比较。结果表明,将传统屏蔽计算方法与高性能计算相结合,能满足CFR600堆本体屏蔽计算精度要求,获得更为全面的三维展示效果,在计算模型复杂、粒子穿透深度等复杂问题的屏蔽计算上具有较明显的优势,为大型钠冷快堆屏蔽设计提供有力支撑。     

蒙特卡罗粒子输运权窗减方差方法研究及在ITER屏蔽分析中应用

屏蔽问题是聚变中子学研究密切关注的问题,蒙特卡罗粒子输运中的权窗减方差技术是处理屏蔽计算深穿透难题的重要方法。本文对权窗减方差方法开展研究,并在超级蒙特卡罗计算软件SuperMC中实现。先采用混凝土基准例题进行校验,再将该方法应用于国际热核聚变实验堆ITER的屏蔽分析,计算结果和MCNP进行对比,符合偏差要求,通过权窗减方差的应用,计算效率也有明显提升,证明了方法和程序的正确性和有效性。     

10—in单球rem计测量中能重离子反应中子剂量当量的理论修正

中能重离子反应出射的中子具有较复杂的能谱,在穿过混凝土屏蔽层后,其能谱发生显著的变化。考虑到中子rem计的能量响应,在中能重离子反应出射中子理论计算能谱和角分布的基础上,估算了屏蔽层外中子能谱的变化和用10-in单球rem计在屏蔽层外测量中能重离子反应中子剂量当量时的理论修正系数。     

一种基于MCNP程序的深穿透问题处理方法

辐射屏蔽计算是核工程设计的核心内容之一。MCNP是最常用到的屏蔽计算软件,但MCNP在处理深穿透问题上存在一定困难,计算误差较大。本文以简单的深穿透屏蔽计算为例,介绍了MCNP的减方差技巧在深穿透屏蔽计算中的应用效果,并提出了一种应用减方差技巧进行深穿透计算的思路。     

小波尺度函数展开在连续能量中子输运方程计算中的应用

小波展开能够很好地拟合剧烈变化的函数,近年来已被应用于模拟中子角注量率随角度剧烈变化的问题,并取得了令人满意的结果.中子能谱在共振区具有剧烈震荡的特性,本文介绍了利用能群与小波尺度函数展开相耦合来离散连续能量中子输运方程中能量自变量的方法.对中子注量率在共振区关于能量用小波尺度函数进行拟合,而在快中子区和热中子区利用分群计算的方法.初步的数值结果表明,该方法使有效增殖系数计算精确,并能够得到中子注量率在共振区随能量的精细分布,对共振自屏蔽的精确计算具有重要意义.     

蒙特卡罗临界计算能量偏倚的最佳源偏倚方法

使用堆用蒙特卡罗程序RMC进行临界计算时采用了传统的裂变源迭代法,即每代源中子按照真实物理过程产生、存库和再抽样。传统裂变源迭代法的计算代之间存在较大的相关性,导致了方差低估计现象,同时总体方差未实现最优化。为实现源中子在空间和能量上的最佳分布,并消除方差低估计现象,提出了能量偏倚的最佳源偏倚方法。该方法基于最佳分层抽样法,结合香农熵诊断和组统计方法,对源中子的空间和能量进行偏倚,实现了全局减方差的计算效果。在RMC程序中开发了能量偏倚的最佳源偏倚方法,并对典型压水堆组件进行测试,计算结果证明了该方法的正确性和有效性。     

Application of High Performance Computing Analysis Technology in Shielding Design of Large Sodium-cooled Fast Reactor

The sodium-cooled fast reactor container is an integrated pool structure composed of numerous internal components and complex structure. The anisotropy is obvious and the deep penetration problem is serious in the process of neutron transport from core to biological shielding. The calculation of three-dimensional S_N method in large scale is the bottleneck restricting in the design of fast reactor shielding. By combining with high performance computing technology, the parallel computing scheme is used to solve the anisotropic three-dimensional deep penetration shielding calculation in the fast reactor. In this paper, the China Demonstration Fast Reactor (CFR600) reactor block was taken as the research object. Using JSNT-CFR code, the neutron flux rate, photon flux rate, and dose rate in the reactor block were calculated in detail. The calculation results were compared with those of the existing two-dimensional code. The results show that combining the traditional shielding calculation method with high performance computing can meet the requirements of CFR600 reactor block shielding calculation accuracy, and obtain a more comprehensive three-dimensional display effect. It can solve the problem of shielding calculation of complex problems such as complex model and particle penetration depth. It has obvious advantages and provides strong support for the large sodium-cooled fast reactor shielding design.     

Physical and engineering problems of small shield design

This article is a survey of the design problems associated with small shields for nuclear reactors, intended for transportation installations. The physical problems that arise in the design of small shields are discussed (radiation from the reactor, penetration of-rays and neutrons through the shield, the origin and penetration of capture-rays, shadow shielding), as are the engineering problems (arrangement of the shielding, selection of materials, the most effective sequence of shielding layers).     

基于正向蒙特卡罗计算的自动源偏倚方法在屏蔽计算中的应用

反应堆屏蔽计算是粒子输运数值计算的难点问题之一。由于仅有少量处于堆芯外围组件的高能中子能到达屏蔽层外,如果对源粒子采用无偏抽样,大量的计算时间用于模拟无用的源粒子,计算效率很低。偏倚抽样是提升蒙特卡罗模拟计算效率的重要途径,包含源偏倚、输运偏倚和碰撞偏倚等。MCNP程序的权窗发生器可为输运偏倚和碰撞偏倚提供参数,但不包含源偏倚。本文利用正向蒙特卡罗计算权窗发生器产生的重要性函数,生成源偏倚参数以及与之匹配的权窗系数,在屏蔽计算中取得了很好的效果。本文的方法与MCNP的权窗功能完全兼容,使用方便。     

Implementation of a low energy neutron transport module into a Monte Carlo hadronic shower code and its applications for accelerator shielding problems

The high energy particle cascade code FLUKA (version 1990 of CERN) was extended down below the usual 50 MeV cutoff to account for emission and transport of the lower energy neutrons. This was accomplished by adopting the evaporation module extracted from the HETC code and the multigroup neutron collision subroutine extracted from the MORSE code to FLUKA program environment, and by slightly modifying sampling of the excitation energy and of the intranuclear cascade energy in the EVENTQ inelastic interaction generator, following the experimental data and phenomenological suggestions published elsewhere. The resulting FLUNEV code is briefly described; common and distinct features of the CALOR and HERMES systems which can be used alternatively for simulation of the neutrons originated from hadronic cascades are also mentioned. Recent applications of our program to shielding problems at high energy proton accelerators are reviewed and new comparisons are presented with measurements of low energy neutron spectra in collider tunnels.     

基于RMC程序的用康普顿轮廓进行束缚态电子多普勒展宽修正研究

堆用蒙特卡罗程序RMC具备中子、光子、电子耦合输运能力，能完成精确的屏蔽计算，其中光子输运过程采用光子数据库进行了康普顿散射模拟。本文对康普顿散射物理原理及多普勒展宽方法进行分析，使用康普顿轮廓数据对束缚态电子进行多普勒展宽修正，实现了RMC程序对自由电子和束缚态电子的选择性处理。通过核素算例测试，观察到了多普勒能谱展宽的效应，证明了该方法的正确性。通过对典型压水堆组件的计算和对比，验证了用康普顿轮廓进行束缚态电子多普勒展宽修正的必要性和正确性。     

Virtual sampling in variational processing of Monte Carlo simulation in a deep neutron penetration problem

In a further study of virtually processed Monte Carlo estimates in neutron transport, a shielding problem has been studied. The use of virtual sampling to estimate the importance function at a certain point in the phase space depends on the presence of neutrons from the real source at that point. But in deep penetration problems, not many neutrons will reach regions far away from the source. In order to overcome this problem, two suggestions are considered: (1) virtual sampling is used as far as the real neutrons can reach, then fictitious sampling is introduced for the remaining regions, distributed in all the regions, or (2) only one fictitious source is placed where the real neutrons almost terminate and then virtual sampling is used in the same way as for the real source. Variational processing is again found to improve the Monte Carlo estimates, being best when using one fictitious source in the far regions with virtual sampling (option 2). When fictitious sources are used to estimate the importances in regions far away from the source, some optimization has to be performed for the proportion of fictitious to real sources, weighted against accuracy and computational costs. It has been found in this study that the optimum number of cells to be treated by fictitious sampling is problem dependent, but as a rule of thumb, fictitious sampling should be employed in regions where the number of neutrons from the real source fall below a specified limit for good statistics.     

The effect of carbon powder addition on the properties of hematite radiation shielding concrete

Carbon powder was added to shielding concrete made of Hematite aggregates to investigate its effects on shielding properties. The powder was added in different percentages, and the mechanical and radiation attenuation properties of the prepared concretes were determined.It was found that, the addition of carbon powder by 6% (by wt.) of the concrete could increase the strength on concrete by about 15%. The shielding effectiveness decreased for both gamma and neutrons with the increasing carbon powder percentage. But the loss in shielding effectiveness for both gamma rays and neutrons were within measurements error range for carbon powder addition of 6%.Simulation for the experimental measurements was carried out using Monte Carlo MCNP code, to understand the effect of carbon powder on the shielding effectiveness against neutrons. The results of the simulation were in good agreement with the experimental results.     

中子残余应力谱仪静态屏蔽体表面辐射剂量率初算

中子残余应力谱仪静态屏蔽体主要用于对谱仪装置的附加闸门、中子导管等组件的辐射剂量的屏蔽,使装置操作人员可以安全地在装置周围活动。通过MCNP5程序对谱仪装置静态屏蔽体的屏蔽能力进行了计算,可为该方案的改进、优化提供依据,以便最终制造出满足辐射剂量要求的屏蔽体。     

中子输运蒙特卡罗模拟的区域分解方法研究

蒙特卡罗模拟方法在反应堆物理分析中的发展和应用受到计算机内存不足的限制,区域分解是一种解决方法。本文将区域分解方法应用于中子输运蒙特卡罗模拟过程,基于反应堆蒙特卡罗程序RMC,实现区域分解基本功能,并测试、分析了其并行性能。结果表明,负载均衡与通信性能是影响区域分解方法的关键因素。对区域分解方法的结果可重复性问题进行了研究,提出对源中子进行排序的实现方法。