硼化钨材料中子屏蔽性能及次级γ剂量产生的模拟研究

对辐射防护材料硼化钨的中子吸收和次级γ射线屏蔽性能进行分析。采用Geant4程序,对材料厚度0～2 cm、能量为热中子～20 MeV的入射中子进行模拟分析。研究结果表明：(1)硼化钨材料主要作用于热中子～10^-2 MeV中子的吸收屏蔽。由不同材料对应的中子宏观分出截面和材料密度可知,厚度一定时,W₂B₅的中子吸收性能最优,质量一定时,WB₄中子吸收性能最优。以热中子为例,W₂B₅材料的中子宏观分出截面约为B203材料的8.67倍,是PB202屏蔽材料的40.59倍;(2)相比于传统中子吸收材料,W-B系化合物在低能中子吸收方面优势更为显著;(3)随着入射中子能量的增大,次级γ剂量对总剂量的贡献呈下降趋势;随着硼化钨材料厚度的增加,次级γ剂量对总剂量的贡献不断升高。为明确硼化钨应用场景及优势,实现中子源屏蔽装置的优化设计提供数据参考,具有实际的工程指导价值。     

基于蒙特卡罗方法的航空机组人员辐射剂量评估北大核心CSCD

采用辐射仿真人体面元模型和数字化飞机模型,基于蒙特卡罗方法开展了航空机组人员的航空辐射剂量研究。选取昆明至北京航线为例,评估了航空机组人员在该航线飞行时受到的航空辐射有效剂量率,分析了人体参数对航空辐射有效剂量率的影响,并探索了航线参数变化时机组人员受到的航空辐射有效剂量率的变化。结果表明,机组人员在昆明至北京航线受到的航空辐射有效剂量率为2.114μSv/h,基于中国人参考生理特征的体模和高加索人体参数体模的有效剂量率评估结果差异为25.3%;航线参数中航线的高度是最主要的影响因素,14 km飞行高度的航空辐射有效剂量率达到10 km高度时的1.8倍,同时,航线纬度升高时,机组人员受到的航空辐射有效剂量率也会产生显著的提升。该研究对航空机组人员的辐射剂量评估具有一定的参考和指导意义。     

蒙特卡罗方法模拟反符合屏蔽γ谱仪

本文用蒙特卡罗方法模拟反符合屏蔽γ谱仪中γ射线、电子及其二者级联簇射,给出了探测效率、沉积能谱、响应函数和能量分辨。计算结果同实验相符。可为设计反符合屏蔽γ谱仪提供理论数据。     

屏蔽材料对γ射线屏蔽情况的蒙特卡罗模拟

屏蔽防护在反应堆、一般放射性测量以及低水平放射性测量中都起着极为重要的作用.运用蒙特卡罗MCNP4C程序,模拟了几种常用的屏蔽材料对γ射线屏蔽情况.通过对模拟结果的分析,给出这些屏蔽材料对γ射线的衰减系数、有效衰减系数,并对能量发散情况进行定量分析,同时补充了一些辐射防护手册中没有提供而在实际工作中有用的屏蔽材料的屏蔽...     

高能α粒子辐射屏蔽的蒙特卡罗模拟

本文应用蒙特卡罗方法模拟α粒子在不同材料(铝、聚乙烯、水、液氢)中的输运.根据得到的射程与能量关系及二次粒子种类和数量,评价不同材料对α粒子的屏蔽效果.得出如下结论:在空间辐射防护中,铝并不是最佳的防护材料;在含氢物质中,随着氢含量的增加,辐射防护能力有所提高;并且α粒子随入射深度的能量沉积曲线中出现了典型的布拉格峰.     

新型核辐射屏蔽材料的优化设计

为研制兼具质地轻、体积小、耐高温等特点的新型中子、γ混合辐射屏蔽材料,用遗传算法建立了屏蔽材料优化设计方法。对应用该方法设计出的3种材料进行MCNP软件检验,并与铁等材料进行比较。结果表明:在达到一定厚度后,这些材料对裂变中子、γ的总屏蔽效果均超过铁,且相同屏蔽效果下的铁质量当量显著下降。     

工程试验混合堆γ屏蔽的蒙特卡罗法研究

本文给出了用蒙特卡罗和环坐标法研究工程试验混合堆γ屏蔽的结果。结果表明:γ屏蔽对于聚变-裂变混合堆的设计是个非常重要的问题。发现在各种γ射线源中,非弹性散射γ射线源是重要的。混合堆的这个特点与裂变堆是不同的。     

先进压水堆核电站辐射屏蔽优化设计对蒙特卡罗方法的要求

在分析M310堆型核电站辐射屏蔽设计中由于工具限制存在的问题以及"华龙一号"堆型核电站辐射屏蔽设计提出的要求的基础上,从程序界面、输入接口、计算功能和辐射场应用扩展4个方面提出先进压水堆核电站辐射屏蔽优化设计对于蒙特卡罗(MC)方法的要求。MC方法在"华龙一号"辐射屏蔽优化设计的应用实践表明,基于MC方法的计算程序在程序界面、输入接口和辐射场应用扩展方面进一步提升之后,可在先进压水堆核电站辐射屏蔽优化设计方面发挥巨大的作用,显著提升核电站辐射屏蔽优化设计的水平。     

蒙特卡罗方法用于LiFTLD自屏蔽因子的模拟计算

自屏蔽因子是用LiFTLD探测器高精度测量n-γ混合场的重要参数.本文采用MCNP程序对文献中给定尺寸、密度和6Li含量的LiFTLD的自屏蔽因子进行了计算,并与推导的自屏蔽公式、Horowitz等的计算结果进行了比较.同时,对计算结果还专门设计了实验进行验证,证实MCNP计算自屏蔽因子所采用的模型和方法是可靠的.对于文献中没有给出自屏蔽因子的现有的LiFTLD,通过选取合理的模型,用MCNP计算了它们的自屏蔽因子.该研究结果对于n-γ混合场区分测量以及TLD的LET效应的研究都具有重要意义.     

MC法模拟地层元素测井仪优化中子屏蔽体

利用MC法模拟了D-T中子源发出的粒子通过地层元素测井仪内部屏蔽体结构的过程,获得了不同粒子通过不同材料的屏蔽体后能量和核反应截面的分布,从而得出不同材料在不同厚度下的粒子屏蔽效果。模拟结果表明:采用17 cm厚的三层复合屏蔽体结构,所用材料第一层为10 cm厚的钨镍合金,第二层为5 cm厚含20%碳化硼的聚乙烯,第三层为2 cm厚的铅。三层结构对中子的屏蔽率达到98.47%,对γ光子的屏蔽率达到97.68%。可有效降低仪器内部元素干扰,提高分辨率与精确度。     

Am-Be中子源屏蔽优化设计的蒙特卡罗模拟研究

用低密度富氢材料作为²⁴¹Am-Be中子源防护罐屏蔽材料,防护罐尺寸大,屏蔽效率低,不利于现场测井作业。利用蒙特卡罗模拟方法,分别计算多种屏蔽材料对中子的慢化效果,优化设计了中子屏蔽效果好、相对轻便的防护罐。模拟结果得到:针对石油测井常用的18 Ci ²⁴¹Am-Be中子源屏蔽罐,内层选用钨作为高能快中子的慢化层,厚度取13 cm;外层选用硼聚乙烯作为较低能量快中子慢化和热中子吸收层,厚度取18 cm。防护罐整体尺寸为φ62 cm×62 cm,体积0.187 m³,质量430 kg,比传统石蜡罐直径和重量约小一半,屏蔽罐外辐射剂量率小于0.025 mSv·h^-1,符合辐射防护标准要求。     

Estimation method of systematic uncertainties in Monte Carlo particle transport simulation based on analysis of variance

In the present work, the authors propose a standard analytical method to estimate systematic uncertainties in the Monte Carlo particle transport simulation based on analysis of variance (ANOVA). As sample problems, two sets of neutron-shielding calculations for evaluating effective dose were performed using the Particle and Heavy Ion Transport code System coupled with ANOVA. In the simulation, we selected the total cross section and the water density in material as unclear physical quantities to induce systematic uncertainties. Systematic and statistical uncertainties were evaluated by changing the number of simulation conditions and trials per condition, where each condition is associated with one set of unclear quantities. Then, we analyzed the dependence of convergence of the systematic uncertainties on the number of conditions and trials. The analysis results show that simulations with only three conditions are adequate in the case where variations in the conditions have a monotonic effect on the calculation results, while more than 100 conditions are required in the other cases. In addition, we propose a useful criterion to determine the appropriate number of trials for converging systematic uncertainties.     

Method for converting in-situ gamma ray spectra of a portable Ge detector to an incident photon flux energy distribution based on Monte Carlo simulation

A matrix stripping method for the conversion of in-situ gamma ray spectrum, obtained with portable Ge detector, to photon flux energy distribution is proposed. The detector response is fully described by its stripping matrix and full absorption efficiency curve. A charge collection efficiency function is introduced in the simulation to take into account the existence of a transition zone of increasing charge collection after the inactive Ge layer. Good agreement is obtained between simulated and experimental full absorption efficiencies. The characteristic stripping matrix is determined by Monte Carlo simulation for different incident photon energies using the Geant4 toolkit system. The photon flux energy distribution is deduced by stripping the measured spectrum of the partial absorption and cosmic ray events and then applying the full absorption efficiency curve. The stripping method is applied to a measured in-situ spectrum. The value of the absorbed dose rate in air deduced from the corresponding flux energy distribution agrees well with the value measured directly in-situ.     

BF3中子探测器阵列探测效率的蒙特卡罗计算

在用241Am-Be中子源对BF3中子探测器阵列探测效率标定的基础上,用蒙特卡罗方法对其探测效率进行了模拟计算,获得了比较满意的结果.然后用蒙特卡罗方法对BF3中子探测器阵列的探测效率进行了研究.研究结果表明,焦面探测器具有较好的探测效率.     

基于离散纵标法与蒙特卡罗方法的三维耦合程序开发

辐射屏蔽设计是核装置工程设计的核心内容之一。单一的离散纵标法(比如SN)或蒙特卡罗方法(MC)在大型核装置屏蔽计算分析方面均存在一定限制。为了满足大型复杂核装置精确辐射屏蔽计算要求,本文实现了三维SN-MC耦合方法,并发展了相应的三维耦合程序系统。该程序结合了SN方法解决深穿透问题的优势和MC方法模拟复杂几何的长处,克服两种方法的缺点,为保证屏蔽系统优化设计的质量提供有力的技术支持。采用接口程序和MC自定义源抽样程序将SN计算得到的粒子角注量率转换为MC计算所需的源粒子信息,为下一步MC计算提供源项,实现三维SN-MC耦合输运计算。采用MC、SN、SN-MC耦合三种方法对直角坐标系和圆柱坐标系下的测试例题进行了计算比较分析。计算结果吻合良好,初步证明了所开发的三维SN-MC耦合程序的正确性。     

Developing architecture for upgrading I&C systems of an operating nuclear power plant using a quality attribute-driven design method

This paper presents the architecture for upgrading the instrumentation and control (I&C) systems of a Korean standard nuclear power plant (KSNP) as an operating nuclear power plant. This paper uses the analysis results of KSNP's I&C systems performed in a previous study. This paper proposes a Preparation–Decision–Design–Assessment (PDDA) process that focuses on quality oriented development, as a cyclical process to develop the architecture. The PDDA was motivated from the practice of architecture-based development used in software engineering fields. In the preparation step of the PDDA, the architecture of digital-based I&C systems was setup for an architectural goal. Single failure criterion and determinism were setup for architectural drivers. In the decision step, defense-in-depth, diversity, redundancy, and independence were determined as architectural tactics to satisfy the single failure criterion, and sequential execution was determined as a tactic to satisfy the determinism. After determining the tactics, the primitive digital-based I&C architecture was determined. In the design step, 17 systems were selected from the KSNP's I&C systems for the upgrade and functionally grouped based on the primitive architecture. The overall architecture was developed to show the deployment of the systems. The detailed architecture of the safety systems was developed by applying a 2-out-of-3 voting logic, and the detailed architecture of the non-safety systems was developed by hot-standby redundancy. While developing the detailed architecture, three ways of signal transmission were determined with proper rationales: hardwire, datalink, and network. In the assessment step, the required network performance, considering the worst-case of data transmission was calculated: the datalink was required by 120 kbps, the safety network by 5 Mbps, and the non-safety network by 60 Mbps. The architecture covered 17 systems out of 22 KSNP's I&C systems. The architecture is implementable with the equipment developed in South Korea. The architecture can be used as a model to upgrade the existing I&C systems in a planned, large-scale, and one-shot manner. A more detailed architecture down to software level will be developed in the future.