核聚变装置停机剂量率分析计算的严格两步(R2S)法

在三维蒙特卡罗粒子输运程序MCNP的基础上，发展了一种用于几何结构复杂的核聚变装置如托卡马克装置停机剂量率的精确计算方法——严格两步法(R2S)。首先对R2S方法进行描述，然后在ITER停机剂量率实验T426的基础上进行校核计算，并与“直接一步法”(DIS)进行了比较分析，结果显示，R2S方法与实验吻合得非常好，最大误差大约为15％，而且，相对于DIS方法(最大误差为25％)而言更准确。     

核聚变装置HT-7U环境辐射剂量蒙特卡罗模拟计算与分析

在聚变评价数据库FENDL／2的基础上，采用三维蒙特卡罗输运程序MCNP／4C，对核聚变装置HT07U运行时的环境辐射剂量率进行了模拟计算及相关分析，着重研究了HT-7U硼水屏蔽层不同设计方案对环境剂量率的影响。计算结果表明，硼水层有效地减小了装置运行时周围环境的辐射剂量率，而浓缩硼水方案相对天然硼水方案，屏蔽效果提高不明显。同时本文也对聚变装置环境辐射剂量计算中的天空反(散)射效应进行了初步的研究，计算结果显示，当屏蔽大厅顶厚减至0．5m以下时，装置运行时大厅外辐射剂量分布呈现明显的阴影效应。     

核聚变实验装置HT-7U及大厅活化分析

使用一维ＳＮ离散坐标法输运程序ＡＮＩＳＮ、活化计算程序ＡＦＤＫＲ并使用一维球几何模型对ＨＴ－７Ｕ装置米要部件及混凝土屏蔽墙的活化进行了计算和分析,给出了中子能谱、γ能谱、大厅内外剂量率空间分布及放射性水平的时间分布,对ＨＴ－７Ｕ装置的主要中子学参数及周围居民接收剂量水平给出了定量的分析。     

核聚变实验装置HT-7U一维及二维辐射防护设计研究

主要介绍一维、二维中子输运程序ANISN,DOT3.5在核聚变实验装置HT-7U辐射屏蔽物理设计中的应用。计算和分析了该装置实验大厅内外中子注量/能谱、γ注量/能谱、中子剂量率、γ剂量率的空间分布,对屏蔽材料的选取及屏蔽层厚度进行了优化设计,为HT-7U装置的辐射屏蔽物理设计提供了建议性意见及理论依据。     

基于网格计数的停机剂量率计算方法研究

本文根据严格二步法的计算理论,研究了基于网格计数的停机剂量率计算方法,设计并实现了基于网格计数的停机剂量率计算程序。该程序能够支持圆柱坐标下的网格计算。本文使用源子程序进行复杂源描述。为了加快计算速度,本文采用了多节点和多线程等技术。本文利用国际热核聚变实验堆(ITER)停堆剂量基准实验ITER-T426进行测试,计算结果与实验值吻合良好,证明了该方法和程序的正确性和可用性。     

MCNP源子程序在停机剂量率计算中的应用

在核聚变装置的停堆剂量率的计算中,通常采用MCNP程序来实现光子的输运计算,但由于聚变装置几何和材料的高度复杂性使得栅元数量庞大,同时由于标准MCNP在进行光子输运计算时,SDEF通用源卡只支持1 000个以下的栅元描述,因此直接采用SDEF通用源卡的方法无法实现聚变堆的停堆剂量率精确计算与分析。本论文采用MCNP内置源子程序方法直接对衰变光子源进行抽样,解决了SDEF通用源卡受限的问题。以国际热核聚变实验堆ITER最新发布的停机剂量率基准例题以及ITER-T426基准实验例题对源子程序进行了校验,结果表明了该方法的可用性与正确性。     

基于严格两步法的停堆剂量率计算程序的基准分析

基于蒙卡程序cosRMC的网格计数功能,开发了以严格两步法为核心的停堆剂量率计算程序,通过耦合粒子输运计算和活化分析计算,精确求解停堆剂量场。其中,采用ALARA程序开展活化分析计算,将程序应用于ITER诊断窗口计算基准题上,开展了充分的计算分析,并与其他严格两步法程序计算得到的停堆剂量率结果有较好的一致性。另外,由于聚变装置几何十分复杂,结构网格难以准确描述几何结构,往往一个网格包含多个栅元,此时网格的通量平均对停堆剂量率的精度会带来不好的影响,而非结构网格具有良好的几何适应性,因此,基于非结构网格对停堆剂量率程序作了进一步开发,并在基准题上开展计算分析,验证了程序基于非结构网格计算停堆剂量率的可靠性。     

一维中子-光子输运程序ANISN在核聚变实验装置HT-7U周围环境辐射剂量计算中的应用

在一维几何模型的基础上采用ＮＡＩＳＮ程序,计算并分析了ＨＴ－７Ｕ超导Ｔｏｋａｍａｋ在Ｄ－Ｄ放电时周围环境辐射剂量当量率的变化规律,计算与分析的结果可供Ｔｏｄａｍａｋ核聚变实验装置作环境评价和防护设计参考。     

用蒙特卡罗方法模拟计算高气压电离室对 60Co和 137Cs源的空气吸收剂量率因子

用蒙特卡罗方法模拟计算高气压电离室对⁶⁰Co和¹³⁷Cs源的空气吸收剂量率因子，并在标准参考辐射场中进行对应刻度。计算和刻度结果表明：对¹³⁷Cs点源，高气压电离室空气吸收剂量率因子的计算值与刻度值间的相对偏差为0.65%；对⁶⁰Co点源，两者之间的相对偏差为－5.5%。计算值与刻度值在不确定度内一致。     

三维停堆剂量率计算程序研发及其在EAST上的应用

为了精确分析核装置停机后周围空间的三维辐射剂量场分布情况,本文基于停堆剂量计算的严格二步法思想,发展了基于蒙特卡罗输运计算程序MCNP和欧洲活化计算程序FlSPACT的耦合三维停堆剂量计算程序,实现了中子输运、材料活化和光子剂量计算的自动耦合.将该程序初步应用于EAST装置停堆剂量计算,得到了托卡马克装置停堆后周围空间...     

Implementation and application of PyNE sub-voxel R2S for shutdown dose rate analysis

PyNE R2S is a mesh-based R2S implementation with the capability of performing shutdown dose rate (SDR) analysis directly on CAD geometry with Cartesian or tetrahedral meshes. It supports advanced variance reduction for fusion energy systems. However, the assumption of homogenized materials of PyNE R2S with a Cartesian mesh throughout a mesh voxel introduces an approximation in the case where a voxel covers multiple non-void cells. This work implements a sub-voxel method to add fidelity to PyNE R2S with a Cartesian mesh during the process of activation and photon source sampling by performing independent inventory calculations for each cell within a mesh voxel and using the results of those independent calculations to sample the photon source more precisely. PyNE sub-voxel R2S has been verified with the Frascati Neutron Generator (FNG)-ITER and ITER computational shutdown dose rate benchmark problems. The results for sub-voxel R2S show satisfactory agreement with the experimental values or reference results. PyNE sub-voxel R2S has been applied to the shutdown dose rate calculation of the Chinese Fusion Engineering Testing Reactor (CFETR). In conclusion, sub-voxel R2S is a reliable tool for SDR calculation and obtains more accurate results with the same voxel size than voxel R2S.     

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.     

Monte Carlo calculations of the incineration of plutonium and minor actinides of laser fusion inertial confinement fusion fission energy(LIFE) engine

In this paper, neutronic analysis in a laser fusion inertial confinement fusion fission energy(LIFE) engine fuelled plutonium and minor actinides using a MCNP codes was investigated.LIFE engine fuel zone contained 10 vol% TRISO particles and 90 vol% natural lithium coolant mixture. TRISO fuel compositions have Mod(1): reactor grade plutonium(RG-Pu), Mod(2):weapon grade plutonium(WG-Pu) and Mod(3): minor actinides(MAs). Tritium breeding ratios(TBR) were computed as 1.52, 1.62 and 1.46 for Mod(1), Mod(2) and Mod(3), respectively. The operation period was computed as ～21 years when the reference TBR??1.05 for a selfsustained reactor for all investigated cases. Blanket energy multiplication values(M) were calculated as 4.18, 4.95 and 3.75 for Mod(1), Mod(2) and Mod(3), respectively. The burnup(BU)values were obtained as ～1230, ～1550 and ～1060 GWd tM~(-1), respectively. As a result, the higher BU were provided with using TRISO particles for all cases in LIFE engine.     

X射线对Kovar封装材料的剂量增强效应的Monte-Carlo计算

用 Monte- Carlo光子 -电子耦合输运程序计算了真实半导体封装 Kovar结构对不同能量 X射线在硅中的剂量增强因子 ,并与内层不涂金的 Kovar结构进行比较 ,计算了界面处两种结构进入硅的净电子数 ,结果证实了界面处产生的剂量增强主要来自界面处高 Z材料二次电子的贡献 ,该计算方法和结果为研究射线剂量增强效应提供了一种可靠的理论评估手段     

Development of an effective dose calculation method for the downstairs room of a linac facility

ABSTRACT

An effective dose calculation method is important in the design of efficient shields in radiation facilities. Some analytical methods have been shown to provide a simple and quick design analysis; however, no suitable method exists that can be applied to a room located directly under an X-ray irradiation room. We propose a new analytical method that uses the multiple reflection ratio predetermined by a Monte Carlo simulation and the differential dose albedo given by the Chilton–Huddleston semi-empirical equation. Our method is verified by comparison with the Monte Carlo simulation, performed for the case of an electron linac facility with an accelerated energy of 10 MeV, where the shielding floor has a thickness of 1.6–2.0 m and the downstairs room has a height of 0.5–1.5 m. The difference between the effective X-ray doses in the downstairs room calculated via the proposed analytical method and the Monte Carlo simulation is less than 25% when the horizontal distance from the X-ray beam to the evaluation point exceeds 3 m and the evaluation point is set at half of the height of the room. The new analytical method can be efficiently and accurately applied to the calculation of the effective dose.