An empirical formula for scattered neutron components in fast neutron radiography

Scattering neutrons are one of the key factors that may affect the images of fast neutron radiography. In this paper, a mathematical model for scattered neutrons is developed on a cylinder sample, and an empirical formula for scattered neutrons is obtained. According to the results given by Monte Carlo methods, the parameters in the empirical formula are obtained with curve fitting, which confirms the logicality of the empirical formula. The curve-fitted parameters of common materials such as ⁶LiD are given.     

中子照相数值模拟与实验的对比

 介绍了热中子照相的MCNP数值模拟方法，模拟了300＃反应堆Maxwell谱热中子束穿透样品射到转换屏的过程，给出了热中子对铝、铅、铁、铜的穿透能力，分析了引起模拟与实验结果不同的原因。通过模拟得到了清晰的图像，对比数值模拟与在300#反应堆中子照相装置上的实验结果，数值模拟结果图像与实验结果图像非常相似；散射中子对图像的影响也符合相同的规律，随着样品与转换屏之间的距离的增大，散射中子对结果图像的影响越来越小，当样品与转换屏之间的距离为样品尺寸的2倍时可以忽略散射中子的影响。     

Investigation of fast neutron interaction with235U

The angular distribution of neutrons emitted by elastic, inelastic and fission processes on²³⁵U were measured at the incident neutron energies of 1.5, 1.9, 2.3, 4.0, 4.5, 5.0 and 5.5 MeV using nanosecond time-of-flight technique. The differential elastic scattering cross sections and their angular distributions at all the seven energies are presented. The total elastic scattering cross sections, angle and energy integrated cross sections for the inelastically scattered neutrons in energy bands of 200 keV, fission cross sections and the angular distributions of fission neutrons were extracted at 1.5, 1.9 and 2.3 MeV incident neutron energies. The energy distributions of the prompt fission neutrons and of the inelastically scattered neutrons are given at the incoming neutron energies of 1.5, 1.9 and 2.3 MeV; and the average fission neutron energies and the inelastic neutron evaporation temperatures were also evaluated at these energies.     

14MeV中子照相中散射中子对成像影响的Monte Carlo模拟

依据实验参数，建立了14MeV快中子照相的物理模型，并利用Monte Carlo方法对照相过程进行了模拟. 分析了经聚乙烯样品散射的中子对快中子图像的影响随样品与探测器间距及样品参数的变化. 计算结果表明，样品与探测器的距离d<5cm时，样品中的散射中子对图像的影响强烈依赖于d，而当d>20cm时，样品散射中子对图像的影响可忽略；当样品密度为3—5g/cm³时散射中子对图像的影响相对最大；样品宽度越大，图像中的散射成分越多，当宽度在3cm以上时散射成分的强度趋于饱和. 关键词： 14MeV中子 快中子照相 散射中子 Monte Carlo模拟     

数值模拟高能中子照相

 模拟了14 MeV中子在穿透样品后与闪烁体光纤的作用。对每根光纤中的能量沉积进行了计算，并转换成可见光(496 nm)光子数。在模拟实验中，分析了影响图像质量的因素。计算了散射中子本底与闪烁体和样品(聚乙烯)间距的关系。当间距为cm量级时，散射中子本底对图像的影响很小。计算表明系统对样品的甄别厚度与入射中子总数有关，在一定范围内近似与中子总数的对数成线性关系。通过模拟结果给出了理想平行中子束入射情况下系统的平面分辨率。     

241Am-Be中子参考辐射的蒙特卡罗模拟研究

研究用于校准场所中子剂量监测仪表的241Am-Be中子参考辐射场计量特性。采用蒙特卡罗方法模拟了空气自由中子参考辐射(FRNR),GB/T 14055规定的最小尺寸中子参考辐射(SRNR)和实际中子参考辐射(ARNR)中不同检验点处中子周围剂量当量率、散射中子占比和能谱分布特征。研究结果表明,空气对FRNR中的剂量率和能谱分布影响小,近似为理想中子参考辐射;采用5%含硼聚乙烯作屏蔽的最小尺寸SRNR可减少热中子,降低散射中子占比,影锥法不适用于小尺寸中子参考辐射中对散射中子的修正;ARNR中的散射中子更少、占比更低,影锥法所得散射中子占比与理论值基本一致。     

Gamma Dose Distribution Studies in Ducted Shield Material Having Different Configurations

In the present work the radial distribution of total gamma doses arising from the interaction of reactor neutrons and gamma radiations inside ilmenite heavy concrete shield are measured. The ilmenite concrete is pierced with neutral beam injector ducts having different diameters and lengths. The results obtained show a relative increase in radial gamma doses at a distance of 10 cm from the neutral beam end. An empirical formula was obtained, fitting the measured gamma doses at the peak position in good agreement within a factor of 1.27. Moreover the variation of the total gamma radiation was measured along straight filled ducted with different diameters and length. The experimental data obtained show a reasonable agreement with calculated ones. The scattered as well as unscattered components of the gamma radiation transmitted through a straight duct were investigated. An empirical formula describing this behaviour was obtained.     

Time-of-flight neutron spectrometer for micro sample studies under high pressure

Abstract

A new neutron spectrometer for investigations of elastic and inelastic neutron scattering on polycrystal microsamples under high pressure in diamond and sapphire anvils cells is described. The spectrometer is operating at the IBR-2 pulsed reactor at JINR. The time-of-flight method and ring-shaped multicounter detector are used to register the scattered neutrons. Parameters and methodical peculiarities of the device and the examples of experimental studies are given.     

中子半影成像的散射中子对点扩展函数的影响

基于ICF中子半影成像分辨力的要求,利用蒙特卡罗方法,模拟了4种系统模型的点扩展函数,分析半影孔装置、靶室和中子束流通道的钢管、高密度内爆靶丸的散射中子对点扩展函数的影响;模拟了平面中子源在4种模型中的半影成像,并用维纳滤波方法重建,分析了散射中子对中子半影成像诊断的影响。模拟结果表明:半影孔装置、靶室等的散射中子不会造成点扩展函数的变形,仅仅增加本底,对诊断的影响可以忽略;高密度内爆靶丸的散射中子会造成点扩展函数的平滑,但对诊断的影响有限。     

On polarized neutron scattering from magnetized samples and the spin echo effect

The scattering of polarized neutrons from magnetized samples is considered in the general case of arbitrary orientation between the incident polarization, the magnetic field of the sample and the polarization of the scattered neutrons. Including both nuclear and magnetic scattering the general expressions for the cross section and the polarization of the scattered beam are derived in the case of small-angle scattering. It is shown that under certain experimental conditions an effect analogous to the Neutron Spin Echo should be observed, which could be used to measure small energy transfers. The contribution arising from the scattering by spin waves is examined in detail.     

2.45 MeV小型D-D中子管准直屏蔽结构设计

为获取小角度出射的单能中子源,采用蒙特卡罗软件对小型D-D中子管产生的能量为2.45 MeV的4立体角中子源进行了准直屏蔽结构设计。准直屏蔽结构分为准直器和捕获穴,准直器采用铁+含硼聚乙烯+铅的三层过滤结构,用于屏蔽照射野外杂散中子,捕获穴主要功能是增加反方向中子的弹性散射次数,从而降低出射束低能散射中子的比例。通过MCNP模拟得到了准直器各层材料的最佳厚度和出射孔尺寸以及捕获穴最佳结构。经验证,中子照射野处2.45 MeV的中子通量比照射野外大三个量级,中子照射野处低能中子通量比2.45 MeV的中子通量低一个量级,墙壁外总剂量率（中子+）在2.5 Gy/h以下。该研究对于小角度单能中子源的快速获取具有一定的实用价值,获取的中子源可用于中子剂量仪器有效性检验、中子监测仪性能测试等方面的研究。     

On the question of possible experimental observation of Anderson localization of the neutron

A possible experiment is discussed, for the observation of Anderson localization of the neutron. The localized state may be formed in the process of inelastic downscattering of thermal or cold neutrons in a highly disordered substance with low neutron capture and upscattering cross sections. The lifetime of trapped (localized) neutrons in the sample is measured by counting the upscattered neutrons with a neutron counter surrounding the sample. Estimations of experimental parameters relevant to such an experiment are given. Received: 12 May 1997 / Revised: 12 September 1997 / Accepted: 16 September 1997     

Is neutron evaporation from highly excited nuclei a poisson random process?

It is suggested that neutron emission from highly excited nuclei follows a Poisson random process. The continuous variable of the process is the excitation energy excess over the binding energy of the emitted neutrons and the discrete variable is the number of emitted neutrons. Cross sections for (HI, x n) reactions are analyzed using a formula containing a Poisson distribution function. The post- and pre-equilibrium components of the cross section are treated separately. The agreement between the predictions of this formula and the experimental results is very good.     

Multiquantum incoherent neutron scattering from molecular solids

General expressions are obtained for the scattering of neutrons from molecular solids with the exchange of m-quanta. Explicit expressions are obtained for the total cross section, when both the phonon and the torson modes are given by the Einstein distribution functions. The expressions can be readily utilized to understand neutron scattering from a molecular crystal.     

中国散裂中子源反角白光中子束流参数的初步测量

中国散裂中子源(CSNS)已于2018年5月建设完工,随后进行了试运行.其中的反角白光中子束线(Back-n)可用于中子核数据测量、中子物理研究和核技术应用等多方面的实验.本文报道对该中子束的品质参数测量实验过程以及最终实验结果.实验主要采用中子飞行时间法,利用~(235)U,~(238)U裂变室和~6Li-Si探测器测量了中子能谱和中子注量率,又利用闪烁体-互补金属氧化物半导体探测系统测量了中子束斑的剖面,得到了该束线的初步实验测量结果.其中白光中子的全能谱测量范围eV—100 MeV,给出了不确定度分析;给出了中子注量率两个实验厅位置的满功率值;给出了白光中子在直径60 mm情况下的全能区束斑.通过与模拟结果的比较探讨了以上结果的合理性,并提出了改进计划.这些实验结果为以后该束线的核数据测量和探测器标定实验奠定了基础.     

A semi-empirical formula for the neutron total cross section and the interpretation of the empirical formula forA ⩾ 40

Weaknesses in Angeli and Csikai’s interpretation of their empirical formula for the neutron total cross section (σ _T) are pointed out. Using the Fourier-Bessel re-presentation of the scattering amplitude a semi-empirical formula forσ _T is obtained which has greater applicability and also explains the success of the empirical formula for mass numberA ⩾ 40 in terms of the established trends in neutron optical potential parameters.     

Scission neutrons in the general systematics of the yields of light particles from ternary nuclear fission

An interpolation formula that was previously obtained on the basis of a statistical approach is applied to experimental data quoted in the literature over the past few years for light-particle yields from the ternary fission of nuclei. Arguments are adduced in favor of this approach, and an analysis of errors inherent in it is given. Tables are presented in which experimental light-particle yields for eight cases of fission are given along with those that were obtained from the interpolation formula. The applicability of this formula to estimating the yields of so-called scission neutrons is analyzed.     

On the Use of Depleted Uranium Dioxide as a Neutron Filter

A simple additive formula is given which, for neutron energies in the range 10^-4 < E<10 eV, permits calculation of the nuclear capture, thermal diffuse and Bragg scattering cross-sections of a crystalline material as a function of its crystal constants and temperature. Computer codes PUO and SUO were developed to calculate the total attenuation of reactor neutrons through poly- and mono-crystalline samples, respectively. The codes were applied to calculate the total cross-section of polycrystalline uranium and thorium dioxides in the energy range from 4 meV to 1 eV. The obtained agreement between the calculated values and available experimental values justifies the applicability of the used formula and the computer codes. A feasibility study on using depleted polycrystalline uranium dioxide as a cold neutron filter and the single crystal as a thermal neutron filter is given. The optimum crystal parameters and thickness for efficiently transmitting the thermal reactor neutrons while strongly attenuating both fast neutrons and -rays accompanying the thermal ones is also given.     

基于遗传算法构建巴耳末公式

基于遗传算法构建了巴耳末公式．通过建立合适的拟合数学模型，用计算机对数学模型进行优化，使之尽可能反映氢原子实际谱线分布，并求出经验常量和拟合公式，最后分析了所得结果．