14 MeV快中子照相准直屏蔽系统的设计与优化

基于强流氘氚中子源科学装置HINEG设计了一套快中子照相准直屏蔽系统。采用中子输运设计与安全评价软件系统SuperMC和ENDF/B-Ⅶ.0数据库计算了准直中子束的中子能谱及注量率、γ射线能谱及注量率、直射中子注量率与γ射线注量率比值(φ_d/φ_γ)、直射与散射中子注量率比值(φ_d/φ_s)、准直束中子注量率的不均匀度等特性参数,并采用MCNP5程序进行了对比验证。研究了准直屏蔽系统的内衬材料、尺寸等对特性参数的影响规律,并通过优化获取了最优设计方案。计算结果显示,在同等计算条件下,SuperMC计算结果与MCNP计算结果相对偏差小于1%,准直屏蔽系统的φ_d/φ_γ为50.1,φ_d/φ_s为5.7,在?30 cm视野范围内的中子注量率为4.80×10~7 cm~(-2)·s~(-1),其中直射中子注量率为4.09×10~7 cm~(-2)·s~(-1),中子注量率不均匀度为5.8%,满足快中子照相对准直束特性参数的要求。     

HT-7 超导托卡马克上辐射剂量测量与分析

HT-7超导托卡马克是我国正在运行中的中型准稳态磁约束聚变实验研究装置,运行时产生一个复杂的瞬发混合辐射场。本文介绍了该装置上中子辐射剂量率的测量方法以及在欧姆、波辅助加热、长脉冲等放电模式下的剂量率特点,并且给出了X与7射线的总辐射剂量测量方法和分析结果。在严格遵守安全操作规程的前提下,目前的等离子体物理实验对人员产生的总辐射剂量低于国家辐射安全标准的规定。     

加速器热中子照相装置CCD芯片屏蔽的模拟计算

建立了研究加速器中子源热中子照相装置CCD芯片屏蔽效果的蒙特卡罗模拟方法,对γ与中子吸收剂量率的模拟计算结果与实验相符.进行了基于9Be(d,n)反应的热中子照相装置屏蔽系统的优化设计,在复杂几何条件下用蒙特卡罗模拟分别计算了CCD芯片在中子、γ混合场中的吸收剂量率和快中子注量率,对CCD相机在辐射场中安全性能进行了评估.     

反应堆快中子实验装置辐射场参数测量

利用多箔活化法测量了设计的反应堆快中子实验装置的中子能谱及中子注量,并采用Monte Carlo方法分析了能谱的不确定度.用热释光剂量片法测量了装置的γ剂量.装置各参数测量结果均达到了预期的设计指标.     

14 MeV快中子照相准直屏蔽系统的设计与优化

基于强流氘氚中子源科学装置HINEG设计了一套快中子照相准直屏蔽系统。采用中子输运设计与安全评价软件系统SuperMC和ENDF/B-Ⅶ.0数据库计算了准直中子束的中子能谱及注量率、γ射线能谱及注量率、直射中子注量率与γ射线注量率比值（φ_d/φ_γ）、直射与散射中子注量率比值（φ_d/φ_s）、准直束中子注量率的不均匀度等特性参数,并采用MCNP5程序进行了对比验证。研究了准直屏蔽系统的内衬材料、尺寸等对特性参数的影响规律,并通过优化获取了最优设计方案。计算结果显示,在同等计算条件下,SuperMC计算结果与MCNP计算结果相对偏差小于1%,准直屏蔽系统的φ_d/φ_γ为50.1,φ_d/φ_s为5.7,在Φ30 cm视野范围内的中子注量率为4.80×10⁷ cm^-2•s^-1,其中直射中子注量率为4.09×10⁷ cm^-2•s^-1,中子注量率不均匀度为5.8%,满足快中子照相对准直束特性参数的要求。     

D-T快中子照相准直屏蔽体设计及中子束特性的模拟研究

设计一个用于氘氚(D-T)快中子照相的准直屏蔽体系统,对D-T中子发生器快中子在准直屏蔽体材料中输运的MCNP模拟研究,给出准直中子束的中子能谱、注量率及均匀性、γ射线能谱和γ射线注量率等重要参数.模拟结果显示,用D-T中子发生器中子源和合理的准直屏蔽体系统可得到快中子照相所需的准直快中子束.     

241Am-Be中子源紧凑型屏蔽装置及辐照器设计

²⁴¹Am-Be中子源被广泛用于实验研究,为保护实验人员免受中子及γ射线照射,需要设计适当的屏蔽。利用蒙特卡罗方法计算中子透射不同材料后的能谱分布与剂量,优选各层屏蔽材料种类与厚度,设计一套²⁴¹Am-Be中子源紧凑型屏蔽装置。装置由内而外采用钨+聚乙烯+含硼聚乙烯+不锈钢进行防护,外表面周围剂量当量率H^*(10)低于10μSv/h,满足辐射防护要求。同时对装置内部热中子、超热中子和快中子注量分布进行研究,确定装置快中子和热中子输出通道最佳位置。在辐照装置同时开放快中子和热中子通道进行实验测试时,需要设置距离大于130 cm的控制区,以保障操作人员安全。     

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。     

γ辐照装置屏蔽方案评价

依据《γ辐照装置的辐射防护与安全规范》,针对典型γ辐照装置升源状态不同辐射照射途经进行计算分析，以验证屏蔽设计的可靠性；对不同照射途经的辐射剂量率进行比较，提出优化屏蔽计算及设计建议。结果表明：(1)该γ辐照装置屏蔽设计方案满足辐射屏蔽要求；(2)屏蔽计算过程中，屏蔽体外天空反散射剂量率贡献大于直射辐射剂量率，迷道入口散射辐射剂量率贡献大于直射辐射剂量率；(3)考虑γ辐照装置工作负荷较大，并遵循辐射防护最优化原则：在屏蔽设计过程中应考虑一次散射照射剂量率贡献，必要时进行局部加厚处理，对于迷道散射设计次数应在5次以上，楼顶区域不建议布置长期人员居留场所。     

FLUKA在永久磁铁辐射退磁实验模拟中的应用

利用FLUKA程序,对韩国浦项加速器实验室的2.5 GeV电子直线加速器辐射退磁实验系统进行蒙特卡罗模拟,得到了永久磁铁处各种粒子(包括光子、电子、中子及质子)的能谱以及注量的空间分布.根据实验中的束流强度,估算出永久磁铁所受的吸收剂量率和1 MeV等效中子注量率分别为18.7 Gy/s和6.37×107 cm-2·s-1.     

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。     

SN2D程序在C5G7基准例题上的验证与分析

C5G7基准例题因其具有强烈非均匀性、组件能谱差异大等特点,常用来检验软件计算有效性。在对栅元进行不同的几何等效和对原有的栅元等效均匀化方法进行改进的情况下,采用二维离散纵标法输运计算程序SN2D对C5G7基准例题进行了分析计算,给出了各种计算条件下k_eff和功率分布的计算误差。结果表明,SN2D程序可应用于C5G7基准例题的求解。计算结果可为求解类似问题时计算程序及条件的选择提供直接参考。     

Measurement and Analysis of Cadmium Ratios of Activation Foils in Thermal Critical Assembly (KUCA B 3/8″P36EU-NU-EU)

Abstract

To examine the applicability of foil activation technique for the estimation of neutron spectrum in a thermal reactor, Cd ratios of 8 activation foils (Au, Th, Dy, In, Mn, W, D.U. and E. U.) were measured in the void at the core center of KUCA B3/8″P36 EU-NU-EU assembly. The Cd ratios were analyzed with SRAC code system using 107 group cross sections based on ENDF/B4. To make the correction for polyethylene plates facing to the void to the calculated spectrum with 2-dimensional (r-z) diffusion model, softening factor calculated with 1-dimensional infinite slab model was introduced. This model gave almost same neutron spectrum as that without this correction. For the model which distributes atoms of Al sheath and support cylinder homogeneously into simulating materials, and using pointwise (fine group) cross sections for Au, Th, W and D.U., the calculated values except for W and D.U. almost agreed with the experimental ones. For W and D.U. C/E values were–1.1. Since Cd ratios are sensitive to the change in neutron spectrum except for D.U., this method is useful to judge the appropriateness of calculated neutron spectrum.     

Remote Control for the HT-7 Distributed Data Acquisition System

HT-7 is the first superconducting tokamak device for fusion research in China. Many experiments have been done in the machine since 1994, and lots of satisfactory results have been achieved in the fusion research field on HT-7 tokamak^[1]. With the development of fusion research, remote control of experiment becomes more and more important to improve experimental efficiency and expand research results. This paper will describe a RCS (Remote Control System),the combined model of Browser/Server and Client/Server, based on Internet of HT-7 distributed data acquisition system (HT7DAS). By means of RCS, authorized users all over the world can control and configure HT7DAS remotely. The RCS is designed to improve the flexibility, opening, reliability and efficiency of HT7DAS. In the paper, the whole process of design along with implementation of the system and some key items are discussed in detail. The System has been successfully operated during HT-7 experiment in 2002 campaign period.     

Reconstruction of the background spectrum for processing airborne gamma survey data

The reconstruction of the background spectrum is based on a decomposition of the initial spectrum into components which are due to the radiation from naturally occurring radionuclides on the Earth’s surface, cosmic rays, and radiation from the spectrometer itself. For the spectra of the naturally occurring radionuclides, the response functions of the detector are calculated for ²³⁸U, ²³²Th, and ⁴⁰ K in the soil; the instrumental and cosmic-ray background is determined by means of aircraft flights at a large altitude. The method has been tested under real conditions and has given satisfactory results for the evaluation of the soil content of naturally occurring radionuclides and ¹³⁷Cs from global fallout. The method will increase considerably the sensitivity and accuracy of the aerial gamma-ray surveys for the detection of technogenic sources of radiation, whose activity is comparable to that of naturally occurring radionuclides and which have no distinct isolated photopeaks in the observed spectrum.     

基于Delaunay三角化方法的三维辐射场可视化研究

根据三维空间立体剖分特点,采用逐点插入法对空间离散点的三角剖分进行改进;对理论计算或实测产生的辐射场离散数据进行重构,得到三维空间内连续分布的辐射场数据;最后结合VR仿真技术,构建某核设施的三维仿真场景,并利用可视化映射技术,实现虚拟环境中的三维辐射场的可视化.结果表明,VR技术与Delaunay三角化方法结合使用可显...     

国产高Z闪烁晶体辐照损伤研究

建立了两套测量准确度分别达到0.5％及1％的测量系统。考察了国产BGO、CsI(Tl)、CsI(Na)和BaF_2晶体在5kGy吸收剂量下的辐照损伤效应。测量显示出两种不同掺杂的CsI晶体显著不同的辐照损伤及其自然恢复性能。发现辐照后BaF_2晶体存在逐渐衰减的强噪声,并且经天然光或日光灯光再照射后强噪声重新出现的新奇现象。     

Release behavior of tritium from graphite heavily irradiated by neutrons

The release behavior of tritium formed in graphite has been investigated as a function of radiation damage by means of isochronal annealing of samples heavily irradiated by neutrons. The lithium impurities in graphite were estimated as the source of tritium formation. The main chemical form of released tritium was hydrogen accompanied by a small quantity of methane. No other hydrocarbons could be detected. Tritiated water was always measured, but the formation mechanism was experimentally confirmed as the secondary oxidation of released HT molecule. The release spectrum of tritium in isochronal annealing was shifted to a higher heating temperature with the increase of the neutron fluence received by the graphite crystal. A relationship was established between the amount of tritium released up to a certain temperature and the degree of graphitization of the sample.     

Analysis of Novo-Voronezh NPP unit 3 radiation lifetime of the WWER-440 reactor pressure vessel given samplings

The analysis of the residual radiation lifetime of the Novo-Voronezh NPP Unit 3 reactor pressure vessel which had spherical samplings after annealing was performed for the spectrum of the ‘worst’ modes of the emergency situation category. For the residual radiation lifetime estimation within the given study, two approaches to determine stress intensity factors, K_I have been used simultaneously. The first approach included a direct numeric modelling of postulated cracks in the cut-out zone with the use of the 3D finite element method. The second approach included K₁ calculation using 3D weight functions calculated with the use of the boundary element method. For K_I, calculation flaws have been postulated as surface longitudinal semielliptical flaws located in the deepest point of a cut-out. The results of K_I calculations obtained using different methods were practically the same. The allowable critical brittleness temperature was determined as 175°C that permitted the extension of the radiation lifetime by up to 6 years after annealing.     

Electron Capture and Loss Processes in Heavy Ion and Molecular Ion Collisions with Solids

Angular (?) and velocity (ve) distributions of convoy electrons have been measured with monoatomic and molecular projectile ions (H+, D+, H+2, D+2, H+2, He+, He++ at 1.7 MeV/U and U+44, Ti+14 at 1.4 MeV/U) as a function of the projectile dwell time tD in carbon foils (2 to 50 ?g/cm2). The spectra with light projectiles from carbon foils are compared with those of gaseous CH4 targets. For the light and molecular projectiles the shapes and yield of convoy electrons depend strongly on the dwell time. A narrow, strongly ?-dependent contribution at Ve = vp can be distinguished from a ?-independent part at Ve < vp indicating different production mechanisms. A relation is presented which connects the yield of loss electrons from projectiles with the transmission of projectiles and electrons through solids. The yield of electrons in the spectrum for H+2 at ve = vp and under zero degrees is found to be more than about 20 times as large as is expected from the contribution of loss electrons in last layer collisions.