热中子透射计测管道油垢方法的初步研究

叙述了利用热中子透射计测石油管道内油垢厚度的原理、实验和结论。实验模拟装置是针对细管道设计的,主要由热中子束发射装置和热中子束探测装置组成。实验结果表明：若用热中子束测量油垢厚度,则很灵敏,测量准确度为1．6mm。若用总中子束测量,则方便但不很灵敏,准确度为4．4mm。     

脉冲中子-裂变中子铀矿测井技术的蒙特卡罗模拟

脉冲中子-裂变中子铀矿测井方法（PNFN）是采用脉冲式中子源,利用-3He管中子探测器记录瞬发裂变超热中子或缓发裂变热中子,得到地层中铀矿含量信息的测井方法。利用MCNP程序模拟了不同铀含量、不同地层孔隙度地层条件下PNFN的响应,分析了瞬发裂变超热中子和缓发裂变热中子与地层铀含量和孔隙度的关系。结果表明,地层孔隙度对利用PNFN确定地层铀含量有影响,孔隙度越大,利用裂变中子直接计算得到的地层铀含量比真实含量越小。利用瞬发裂变超热中子或热中子时间衰减谱计算得到地层宏观俘获截面,对裂变中子进行校正,可以有效提高地层铀含量计算结果的准确度。     

医院中子照射器物理启动

医院中子照射器是专门用于硼中子俘获治疗的核装置。在堆芯相对两侧,设有热中子束流和超热中子束流用于治疗,另外,在热中子束流内引出1条热中子束流用于病人血硼浓度测量。本文介绍其物理启动的6个实验,实验结果表明：满功率最大运行时间为12 h,最终后备反应性为4.2 mk,满功率运行时各工艺房间辐射水平满足设计辐射分区要求,4.2 mk反应性释放实验证明医院中子照射器具有固有安全特性。     

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。     

基于加速器~7Li(p,n)反应的硼中子俘获治疗中子源的优化设计

7Li(p,n)反应以中子产额大、反应阈能低等优点成为硼中子俘获治疗加速器驱动中子源所用中子反应的候选类型之一。本文重点研究了该中子产生反应作为加速器驱动中子源的中子产额及其能谱特性,并对产生的高能中子束流进行慢化,使其满足BNCT治疗要求。首先采用蒙特卡罗程序MCNPX2.5.0模拟加速器7Li(p,n)反应过程,得到1.9 3.0 MeV能量入射质子的中子产额及其能谱,并详细研究了质子入射能量为2.5 MeV的最佳条件下产生的中子束流特性;进而提出中子束流的慢化设计方案,并对慢化所得超热中子束品质进行分析研究。模拟计算结果表明,10 mA流量的2.5 MeV能量入射质子所产生的中子束经过慢化处理后,可以很好地满足硼中子俘获治疗的中子束流要求。     

中子诱发^235U裂变时^95Zr、^140Ba和^147Nd裂变产额随中子能量的变化

本厂作用直接γ能谱法测量了热中子和3.0、5.0、5.5、8.0和14.8 MeV中子诱发235U裂变时的裂变产物产额。本工作的目的是测量95Zr、140Ba、147Nd等核素的产额随入射中子能量的变化关系,进而研究其规律性。其中,140Ba是作为监督核测量的,它位于产额双驼峰曲线重峰的峰区。在本测量工作中,对样品中裂变产物的测量使用了直接γ能谱法。最后给出的测量结果除热中子是绝对测量值外,其余能点各核素的裂变产额值均是通过“快热比”方法给出的,即是以热中子诱发235U裂变产额为基准的相对产额值。采用“快热比”方法测量     

中子活化R矩阵元与中子能谱

实验中发现用由(n,γ)反应产生的一些放射性核素的生成率B_i~0和有关的热中子俘获截面文献值σ_i计算得出的表观中子通量I_i并不相等。其原因,除了放射性测量、γ探测效率、称重、同位素丰度,γ发射率和反应截面等的误差外,还有照射处中子能谱的影响。一般,照射处的中子能谱并非热中子能谱,超热中子的百分数含量随着与反应堆中心之间的距离的减小而增加(一般中子通量也随着增加)。因此,I_i与产生核素i的反应的共振积分S_i和σ_i之     

快中子散射法检测输油管道油垢的进一步研究

利用由中子束发射器、中子探测器和BH1224型微机多道谱仪组成实验装置,研究了采用同一轴线法检测油垢(聚乙烯模拟)厚度时不同散射角对散射中子计数的响应关系,另外利用MCNP/4C程序模拟了采用同一横截面法中不同散射角时中子计数与油垢厚度响应关系.结果显示,采用同一横截面法检测油垢厚度优于同一轴线法,大散射角优于小散射角...     

基于C8051F060的脉冲中子-热中子采集电路设计

针对脉冲中子综合测井仪的需求,设计了基于C8051F060型单片机的脉冲中子-热中子采集电路,用于测量地层中热中子数量随时间的变化趋势.采用集成模块化设计,使电路成本降低,体积缩小.现场试验表明,电路工作正常,达到了设计要求.     

基于硅酸盐钆掺杂的紧凑热中子准直器仿真模拟及性能研究

在核安保管理中,基于中子发生器的热中子成像技术可对被屏蔽核材料进行无损检测成像。为提高成像质量,需对慢化后的中子进行准直。本文使用蒙特卡罗软件Geant4对一种基于硅酸盐钆掺杂的紧凑中子准直器进行了理论建模,对该准直器的热中子透过率和准直率进行了模拟计算,计算结果将用于指导后续的核材料中子成像系统构建。     

~(115)In中子非弹性散射截面的实验测量及蒙特卡罗修正

~(115)In是一种重要的活化材料,准确测量它的中子非弹性散射截面数据对中子注量监测具有重要意义。在四川大学原子核科学技术研究所2.5 MV静电质子加速器上,利用核反应D(d,n)~3He产生的单能中子,以~(197)Au作为标准,采用活化法测量了2.95 Me V、3.94 Me V、5.24 Me V能点的~(115)In中子非弹性散射截面。用Monte Carlo程序MCNPX(Monte Carlo N-Particle eXtended)对靶头材料、冷却水层和样品的包层材料等引起的多次散射效应及注量率衰减效应等进行了修正计算,得到最终结果与Loevestam的计算值符合较好,并且实验中可通过减小靶管、靶底衬、水层及样品的包层材料等厚度来减小多次散射效应和自屏蔽效应的影响。     

基于特征线法计算的超细群慢化方程求解方法

为实现对复杂几何、复杂能谱组件的精细计算,提出了一种基于特征线的超细群慢化方程求解方法。通过耦合特征线法中的固定源计算,在共振能量范围内建立超细群慢化方程,通过精细能谱获得复杂结构下的共振自屏截面。对典型压水堆栅元问题、带有温度分布的栅元问题、燃料内部存在不均匀性的栅元问题以及板状燃料组件问题进行了计算。结果表明,基于特征线的超细群慢化方程求解方法可精确计算复杂几何、复杂能谱问题,为共振计算提供基准。     

Thermal neutron self-shielding correction factors for large sample instrumental neutron activation analysis using the MCNP code

Thermal neutron self-shielding within large samples was studied using the Monte Carlo neutron transport code MCNP. The code enabled a three-dimensional modeling of the actual source and geometry configuration including reactor core, graphite pile and sample. Neutron flux self-shielding correction factors derived for a set of materials of interest for large sample neutron activation analysis are presented and evaluated. Simulations were experimentally verified by measurements performed using activation foils. The results of this study can be applied in order to determine neutron self-shielding factors of unknown samples from the thermal neutron fluxes measured at the surface of the sample.     

Calculation of thermal neutron self-shielding correction factors for aqueous bulk sample prompt gamma neutron activation analysis using the MCNP code

In this work thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing materials is studied using bulk sample prompt gamma neutron activation analysis (BSPGNAA) with the MCNP code. The code was used to perform three dimensional simulations of a neutron source, neutron detector and sample of various material compositions. The MCNP model was validated against experimental measurements of the neutron flux performed using a BF₃ detector. Simulations were performed to predict thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing solutes. In practice, the MCNP calculations are combined with experimental measurements of the relative thermal neutron flux over the sample’s surface, with respect to a reference water sample, to derive the thermal neutron self-shielding within the sample. The proposed methodology can be used for the determination of the elemental concentration of unknown aqueous samples by BSPGNAA where knowledge of the average thermal neutron flux within the sample volume is required.     

Effective Cross Section of 238Samples for Analyzing Doppler Effect Measurements in Fast Critical Assembly

A new effective cross section of ²³⁸U sample has been derived to accurately analyze the Doppler effect measurements in a fast critical assembly. The neutron spectrum in the sample is determined by considering the interference effect of neutron spectrum between the sample and the surrounding fuel region, and is used to obtain the effective cross section of the sample. The new effective cross section can be calculated using the conventional self-shielding factors. The use of the present cross section has increased the Doppler reactivity worth in ZPPR-9 by about 4% compared to the result calculated by the conventional self-shielding factor method without the spectrum interference effect. The physical meaning of this increment is discussed.     

The pseudo-resonant-nuclide subgroup method based global–local self-shielding calculation scheme

The pseudo-resonant-nuclide subgroup method (PRNSM) based global–local self-shielding calculation scheme is proposed to simultaneously resolve the local self-shielding effects (including spatial self-shielding effect and the resonance interference effect) for large-scale problems in reactor physics calculations. This method splits self-shielding calculation into global calculations and local calculations. The global calculations obtain the Dancoff correction factor for each pin cell by neutron current method. Then an equivalent one-dimensional (1D) cylindrical problem for each pin cell is isolated from the lattice system by preserving Dancoff correction factor. The local calculation is to perform self-shielding calculations of the equivalent 1D cylindrical problem by the PRNSM. The numerical results show that PRNSM obtains accurate spatial dependent self-shielded cross sections and improves the accuracy of dealing with the resonance interference over the conventional Bondarenko iteration method and the resonance interference factor method. Furthermore, because both global and local calculation is linearly proportional to the size of problems, the global–local calculation scheme could be applied to large-scale problems.     

用金活化法测定加速器中子源中子注量及MCNP4C修正

在四川大学720所2.5MeV静电质子加速器上,由核反应7Li(p,n)7Be,T(p,n)3He产生中子,对中国工程物理研究院研制的新型中子探测器进行效率刻度实验中,需要知道探测器位置处的中子绝对注量,为此我们测量了0.165、0.352、0.576、1.400MeV四个能点的中子注量。测量方法采用的是金活化法,在实验测量中,由靶头材料、冷却水层和样品的包层材料等引起的多次散射效应及中子在样品中的自屏蔽效应等均对实验结果产生影响。这些因素在实验中不可避免,也难以通过实验方法扣除,因此用Monte Carlo程序MCNP4C对上述效应进行了修正计算。     

基于等价理论的压水堆包壳材料共振计算方法研究

共振自屏效应的处理是影响压水堆组件程序反应性精度的主要因素之一,压水堆锆包壳材料同样具有共振自屏效应,忽略其影响会对反应性造成100~300 pcm（1 pcm=10-5）的偏差。目前,主要通过提供经验上的参考稀释截面与包壳等价理论处理包壳材料的共振自屏效应,但并未对其适用性及精度进行完整的分析。因此,本文采用DRAGON程序,通过一系列压水堆算例对这2种方法进行测试,确定包壳共振自屏效应的主要影响因素以及这2种方法的适用性。结果表明,包壳材料的共振自屏效应仅仅与包壳区的原子核密度、厚度、慢化区的水铀比有关,并且参考稀释截面方法可以满足大部分典型压水堆系统的计算精度,但是对于包壳区尺寸、原子核密度、慢化区水铀比变化较大的系统计算精度较差,而包壳等价理论计算精度和普适性强,可用于不同类型压水堆系统包壳材料的共振自屏计算。      

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.     

Determination of boron over a large dynamic range by prompt-gamma activation analysis

An evaluation of the PGAA method for the determination of boron across a wide dynamic range of concentrations was performed for trace levels up to 5 wt.% boron. This range encompasses a transition from neutron transparency to significant self- shielding conditions. To account for self-shielding, several PGAA techniques were employed. First, a calibration curve was developed in which a set of boron standards was tested and the count rate to boron mass curve was determined. This set of boron measurements was compared with an internal standard self-shielding correction method and with a method for determining composition using PGAA peak ratios. The advantages and disadvantages of each method are analyzed. The boron concentrations of several laboratory-grade chemicals and standard reference materials were measured with each method and compared. The evaluation of the boron content of nanocrystalline transition metals prepared with a boron-containing reducing agent was also performed with each of the methods tested. Finally, the k₀ method was used for non-destructive measurement of boron in catalyst materials for the characterization of new non-platinum fuel cell catalysts.