全局减方差方法在大空间γ辐射场计算中的应用

为了提高大空间γ辐射场全局计算效率,开展了全局减方差(Global Variance Reduction,GVR)方法在大空间γ辐射场计算中的应用研究。针对计数栅元/网格体积差异造成的过度分裂问题,引入体积修正因子修改全空间权窗参数。体积修正后的基于通量的GVR方法计算的全局品质因子(FOM_G)比直接模拟提高约39倍。针对非计数区计算耗时问题,提出了非计数区修正方法,使得FOM_G因子进一步提高40%。在引入体积和非计数区修正的基础上,在大空间γ辐射场计算中与基于粒子误差、权重、径迹、数目、能量、碰撞和通量的7种GVR方法进行对比。结果表明：7种GVR方法计算的FOM_G因子比直接模拟提高2～3个量级,基于误差的标准差σ降低2～3个量级;而基于权重的GVR方法计算的FOM_G因子比直接模拟提高2 304倍,在所有GVR方法中减方差效果最好。在基于通量的GVR方法中引入光滑因子SI后,模拟计算的权窗下限随SI增加而减小,FOM_G因子随SI的增加先增加后减小。当SI=0.8时,该方法计算...     

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。     

基于权窗的全局减方差方法中“长历史”问题的研究与改进

在具有全局特性的蒙特卡罗粒子输运计算中,常采用基于权窗的全局减方差方法,由此引起的"长历史"问题使计算效率大大降低。为了提高蒙特卡罗计算效率,减小计算资源的浪费,本文对"长历史"的成因及解决方案进行了研究。通过研究中国聚变工程试验反应堆(Chinese Fusion Engineering Testing Reactor,CFETR)中子输运计算中"长历史"的事件记录文件,提出了"长历史"的成因与粒子连续在轮盘赌中存活有关,针对此成因提出了通过粒子在轮盘赌中存活的次数来限制粒子分裂的解决方法,并通过CFETR模型对这个方法进行了测试。结果显示,次数限取1-4时,并行效率达到80%,无"长历史"出现,品质因子(pFoMG)达到不采用次数限时的约3倍。综上,采用这种方法可以有效克服"长历史"问题,使全局减方差方法的计算效率明显提升。     

全局减方差方法在乏燃料干式贮存容器屏蔽计算中的应用

开发了基于离散纵标(S_N)方法的蒙特卡罗(MC)全局减方差方法，针对乏燃料干式贮存容器，分别建立了中子源及光子源MC直接计算模型、S_N计算模型及全局减方差方法计算模型，并进行了计算精度和效率的比较。数值结果表明，全局减方差方法计算结果与无偏的MC及S_N计算结果相比吻合良好。其中S_N计算的次级光子剂量率与全局减方差方法计算的偏差较大，这主要是由于MC计算和S_N计算时的数据库差异导致的。和无偏的MC结果相比，全局减方差方法计算的中子及次级光子输运计算收敛效率提高了近2个数量级，初级光子输运计算收敛效率提高了1~2个数量级。     

全局减方差方法在乏燃料干式贮存容器屏蔽计算中的应用

开发了基于离散纵标(S_N)方法的蒙特卡罗(MC)全局减方差方法,针对乏燃料干式贮存容器,分别建立了中子源及光子源MC直接计算模型、S_N计算模型及全局减方差方法计算模型,并进行了计算精度和效率的比较。数值结果表明,全局减方差方法计算结果与无偏的MC及S_N计算结果相比吻合良好。其中S_N计算的次级光子剂量率与全局减方差方法计算的偏差较大,这主要是由于MC计算和S_N计算时的数据库差异导致的。和无偏的MC结果相比,全局减方差方法计算的中子及次级光子输运计算收敛效率提高了近2个数量级,初级光子输运计算收敛效率提高了1～2个数量级。     

全局减方差方法的HBR-2基准题应用

对于深穿透类型的屏蔽问题，在合理的时间内计算得到可信的结果对于蒙特卡罗(MC)方法是一个巨大的挑战。基于离散纵标(S_N)方法的局部和全局减方差方法能有效降低MC计算深穿透问题的计数误差。本文基于HBR-2基准题比较了全局减方差方法和局部减方差方法的计算效率。结果表明，对于HBR-2基准题，局部和全局减方差方法均取得了较好的结果。全局减方差方法1次计算即可同时优化辐照监督管和堆外探测器的计数，因此实际应用更加方便和高效。     

基于自动重要抽样方法的减方差技巧体系构建与验证

屏蔽计算问题根据求解目标不同一般可分为源-探测器问题、区域问题和全局问题。MCShield研究团队针对3类问题中存在的深穿透问题提出了相应的减方差技巧,本文以此为基础构建了基于自动重要抽样(AIS)方法的减方差技巧体系,并开展了验证工作。针对源-探测器问题,采用NUREG/CR-6115 PWR压力容器计算基准题对小探测器自动重要抽样(SDAIS)方法进行验证。结果表明,SDAIS方法的计算效率约为AIS方法的7倍。此外还提出并验证了基于AIS伴随蒙特卡罗的耦合减方差(AIS-CADIS)方法,将AIS方法引入到蒙特卡罗伴随计算中,取得了良好的效果。针对全局问题,提出网格化-AIS方法并使用简化反应堆屏蔽计算算例进行验证,结果表明,网格化-AIS方法的计算效率是AIS方法的12倍左右,是直接蒙特卡罗方法的290倍左右。     

CFETR氦冷陶瓷增殖包层中子学分析

为满足中国聚变工程实验堆(CFETR)包层的应用要求,本文提出氦冷陶瓷增殖(HCCB)包层方案。为验证HCCB包层设计方案的合理性与可行性,采用三维蒙特卡罗粒子输运程序MCNP,计算和分析了HCCB包层方案的氚增殖比、中子壁负载、中子通量密度、核热、辐照损伤等中子学特性。结果表明,HCCB包层方案满足氚自持要求,中子通量密度和核热分布合理,屏蔽性能良好,基本满足设计要求。     

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。     

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。     

Global variance reduction method for global Monte Carlo particle transport simulations of CFETR

It can be difficult to calculate some under-sampled regions in global Monte Carlo radiation transport calculations. The global variance reduction(GVR) method is a useful solution to the problem of variance reduction everywhere in a phase space. In this research, a GVR procedure was developed and applied to the Chinese Fusion Engineering Testing Reactor(CFETR). A cylindrical CFETR model was utilized for comparing various implementations of the GVR method to find the optimum.It was found that the flux-based GVR method could ensure more reliable statistical results, achieving an efficiency being 7.43 times that of the analog case. A mesh tally of the scalar neutron flux was chosen for the GVR method to simulate global neutron transport in the CFETR model.Particles distributed uniformly in the system were sampled adequately through ten iterations of GVR weight window.All voxels were scored, and the average relative error was 2.4% in the ultimate step of the GVR iteration.     

SP3-coupled global variance reduction method based on RMC code

A global variance reduction(GVR)method based on the SPN method is proposed.First,the global multi-group cross-sections are obtained by Monte Carlo(MC)global homogenization.Then,the SP3 equation is solved to obtain the global flux distribution.Finally,the global weight windows are approximated by the global flux distribution,and the GVR simulation is performed.This GVR method is implemented as an automatic process in the RMC code.The SP3-coupled GVR method was tested on a modified version of C5G7 benchmark with a thick-ened water shield.The results show that the SP3-coupled GVR method can improve the efficiency of MC criticality calculation.     

CFETR第一壁及赤道面外包层中子辐照损伤初步分析

中国聚变工程实验堆(Chinese Fusion Engineering Testing Reactor,CFETR)的包层和偏滤器第一壁面向堆芯等离子体,第一壁辐照损伤分析对于托克马克安全运行至关重要。赤道面外包层较其它包层距离堆芯等离子体中心更近,其结构材料承受中子辐照大。因此,进行中子辐照损伤评估十分必要。基于此目的,采用计算机辅助设计(Computer Aided Design,CAD)模型和蒙特卡罗中子学建模转换接口Mc CAD完成中子学建模,并用蒙特卡罗方法的粒子输运程序计算第一壁和氦冷固态外包层结构材料辐照损伤。此外,对比了铍和钨作为面向等离子体材料两种情况下第一壁的受损情况。计算结果表明,氦冷固态包层模型下结构材料可以满足CFETR一期的运行要求。     

Shielding Design and Analysis of Fusion–Fission Hybrid Energy Reactor Blanket

A preliminary design of fusion–fission hybrid energy reactor (FFHER) has been proposed by Institute of Nuclear Physics and Chemistry based on current fusion science and well-developed fission technology. In FFHER, shield blocks provide nuclear shielding and thermal shielding for internal and external blanket components. The hybrid of fusion core and fission blanket makes the spectra rather complex. Therefore, it is necessary to make detail shielding design and carry out radiation analysis according to the blanket structure and material property. In this study, a shielding design of combining several different material shield blocks has been proposed. The shielding analysis is performed by Monte Carlo (MC) method. For the radiation deep-penetration problem, the flux and statistical relative error of forward MC estimate are applied to get an optimal weight window for global variance reduction (GVR). The spatial distribution of neutron and gamma flux have been assessed along the shield block depth. Power deposited and dose rate distributions have also been simulated and analysed. Neutron irradiation damage has been studied to evaluate the material damage. Based on the configuration analysis, nuclear analysis and GVR method, an optimal FFHER blanket shielding design has been obtained.     

Novel hybrid Monte Carlo/deterministic technique for shutdown dose rate analyses of fusion energy systems

The rigorous 2-step (R2S) computational system uses three-dimensional Monte Carlo transport simulations to calculate the shutdown dose rate (SDDR) in fusion reactors. Accurate full-scale R2S calculations are impractical in fusion reactors because they require calculating space- and energy-dependent neutron fluxes everywhere inside the reactor. The use of global Monte Carlo variance reduction techniques was suggested for accelerating the R2S neutron transport calculation. However, the prohibitive computational costs of these approaches, which increase with the problem size and amount of shielding materials, inhibit their ability to accurately predict the SDDR in fusion energy systems using full-scale modeling of an entire fusion plant. This paper describes a novel hybrid Monte Carlo/deterministic methodology that uses the Consistent Adjoint Driven Importance Sampling (CADIS) method but focuses on multi-step shielding calculations. The Multi-Step CADIS (MS-CADIS) methodology speeds up the R2S neutron Monte Carlo calculation using an importance function that represents the neutron importance to the final SDDR. Using a simplified example, preliminary results showed that the use of MS-CADIS enhanced the efficiency of the neutron Monte Carlo simulation of an SDDR calculation by a factor of 550 compared to standard global variance reduction techniques, and that the efficiency enhancement compared to analog Monte Carlo is higher than a factor of 10,000.     

Application of local Monte Carlo method in neutronics calculation of EAST radial neutron camera

The Local Monte Carlo (LMC) method is used to solve the problems of deep penetration and long time in the neutronics calculation of the radial neutron camera (RNC) diagnostic system on the experimental advanced superconducting tokamak (EAST), and the radiation distribution of the RNC and the neutron flux at the detector positions of each channel are obtained. Compared with the results calculated by the global variance reduction method, it is shown that the LMC calculation is reliable within the reasonable error range. The calculation process of LMC is analyzed in detail, and the transport process of radiation particles is simulated in two steps. In the first step, an integrated neutronics model considering the complex window environment and a neutron source model based on EAST plasma shape are used to support the calculation. The particle information on the equivalent surface is analyzed to evaluate the rationality of settings of equivalent surface source and boundary. Based on the characteristic that only a local geometric model is needed in the second step, it is shown that the LMC is an advantageous calculation method for the nuclear shielding design of tokamak diagnostic systems.     

托卡马克聚变中子源的数值模型研究

托卡马克(Tokamak)聚变装置中子学分析中,聚变中子源描述是重要的输入参数,其准确性直接影响分析结果的可靠性。通过分析ITER和欧洲聚变示范堆(EU DEMO)中子学分析中所采用的聚变中子源模型,提出了一种完整描述Tokamak中L-mode、H-mode等离子体的D-D、D-T聚变中子源的数值模型。在中国聚变工程实验堆(CFETR)的工程集成设计平台上,编写了基于蒙特卡罗算法的程序SCG求解该数值模型,实现了读取(零维)等离子体参数、输出可供典型中子学软件MCNP直接读取的中子源定义文件的功能。以CFETR氦冷球床包层的中子学分析模型为基准,在相同的L-mode等离子体D-T聚变工况下,相较于采用EU DEMO源子程序,采用本模型计算得到的中子壁负载差异最大值为2.02%,包层氚增殖率差异为0.18%,全堆能量增益因子的差异为0.23%。结果表明,本模型与其他源描述的差异较小,可应用于CFETR的中子学分析。     

RMC均匀裂变源方法及其在堆用蒙特卡罗计算不对称性问题中的应用

堆用蒙特卡罗程序(RMC)不断被研发以适应于反应堆大规模精细模拟计算。在临界计算中,反应堆中子注量率、功率密度等分布不均匀导致其统计偏差的波动性增大,进而导致反应堆物理对称区域计算结果出现不对称性。蒙特卡罗堆芯计算不对称性主要由统计方差造成,减小方差波动可降低堆芯计算不对称程度。本工作在RMC临界计算中增加均匀裂变源方法,并对Hoogenboom-Martin基准题进行计算验证。结果表明,改进的方法在降低堆芯方差波动性与低功率区域方差方面效果显著。