Determination of thermal neutron cross-section and resonance integral for the As (n, γ)As reaction by activation method using Mn (n, γ)Mn as a monitor

Nuclear constants for use in reactor activation analysis especially (n, γ) cross-sections and absolute gamma intensities, are known to show a rather large scatter in literature. Thermal and resonance cross-sections for the ⁷⁵As (n, γ)⁷⁶As reaction is determined by the method of foil activation using ⁵⁵Mn (n, γ)⁵⁶Mn as a reference reaction. The experimental sample with and without a cadmium cover of 1-mm wall thickness was irradiated in the isotropic neutron field of the outer irradiation sites 7 of Ghana Research Reactor-1 facility which is a miniature neutron source reactor designed by the Chinese. The irradiation channel used has a neutron spectral parameter (α) found to be (0.037 ± 0.001). The induced activity in the sample was measured by gamma ray spectrometry with a high purity germanium detector. A standard solution of Arsenic was used for the analysis. The necessary correction for gamma attenuation, thermal neutrons and resonance neutron self-shielding effects were not taken into account during the experimental analysis because they were negligible. By defining cadmium cut-off energy of 0.55 eV, the result for ⁷⁵As (n, γ)⁷⁶As reaction was found to be: thermal neutron cross-section σ₀ = (4.28 ± 0.19) b and resonance integral I₀ = (61.88 ± 1.07) b.     

Improved formation density measurement using controllable D-D neutron source and its lithological correction for porosity prediction

Controllable D-D neutron sources have a long service life,low cost,and non-radioactivity.There are favorable prospects for its application in geophysical well logging,since traditional chemical radioactive sources used for well logging pose potential threats to the safety of the human body and environment.This paper presents an improved method to measure formation density that employs a D-D neutron source.In addition,the lithological effect on the measured density was removed to better estimate ...     

Monte Carlo Simulation of Gadolinium Neutron Gamma Spectroscopy Tracer Logging Method

The thermal neutron count and capture gamma ray spectroscopy before and after gadolinium seeping into sandstone and limestone formation are simulated by Monte Carlo Method, and the determination method of oil saturation is studied by using the thermal neutron count rate ratio and the n(Gd)/n(H) of capture gamma spectrum before and after gadolinium seeping. In addition, the relations of ratio with influence factors such as porosity, salinity of gadolinium seeping, saturation and lithology are given.      

Non-destructive interrogation of uranium using PGAA

Prompt gamma neutron activation analysis is proposed as an instant, non-destructive method for the assay of uranium and also for the determination of ²³⁵U-enrichment. Measurements were performed in the thermal and the cold neutron beams at the Budapest Research Reactor. A beam chopper was used to collect the delayed decay gamma radiation from short-lived nuclides separately. Partial gamma ray production cross-sections have been determined for a set of capture prompt, fission prompt and decay gamma lines and compared to those from the literature. The proposed method is also compared to NAA and the passive gamma spectroscopy of natural uranium.     

Application of BGO Scintillators to Absolute Measurement of Neutron Capture Cross Sections between 0.01 eV and 10 eV

Applying a total energy absorption γ-ray detector composed of 12 bricks (5 × 5 cm², 7.5 cm thick) of BGO scintillators, the absolute measurement of capture cross sections for Au and Sb has been made in an energy region between 0.01 and 10eV using the linac time-of-flight method. Incident thermal neutron flux was absolutely determined by using the BGO detection system with a Sm sample. To extend the neutron flux measurement from the thermal neutron region to higher neutron energies, the ¹⁰B(n, αγ) reaction was applied. Absolute capture yield for the relevant capture sample was obtained by the saturated capture yield at a large resonance of the sample.

Gold was selected to investigate the application of the BGO detection system to the absolute measurement of the capture cross sections, since the ¹⁹⁷Au(n, γ)¹⁹⁸ Au reaction cross section is a well known standard one. The result of the ¹⁹⁷Au(n, γ)¹⁹⁸ Au reaction cross section showed good agreement with the evaluated data in JENDL Dosimetry File and ENDF/B-VI. Then, the detection system was applied to the Sb(n, γ) cross section measurement. Antimony has a large scattering-to-capture cross section ratio comparing to that of gold. The result showed good agreement with the evaluated data in JENDL-3.2 and ENDF/B-VI.     

热中子俘获瞬发γ射线强度计算

简要介绍了热中子俘获发γ射线强度的计算方法,并以实例进行说明,还给出了有关物理自洽检验的方法及其讨论。     

Neutron Transport Benchmark Calculations, (I)

Five benchmark problems were selected as standards for the evaluation of calculations on neutron penetration. For three of these problems there were available reliable experimental results obtained in absolute intensities of the neutron angular spectrum in graphite and in water and of the neutron reaction rate in a two-dimensional water shield. The two remaining problems were based on PALLAS transport calculations in graphite and in water in plane geometry. For the first three problems, comparisons were made between the results of PALLAS calculations using ENDF/B-I cross sections and those obtained by experiment. It is concluded that the PALLAS calculations result in fairly good agreement with experiment, except at certain spatial meshes in two-dimensional geometry, due to ray effect. Comparisons are also made of the angular flux spectrum in graphite between calculations using ENDF/B-I and B-III data, as well of the scalar flux spectrum in water between those by PALLAS and by ANISN codes.     

Measurement of thermal neutron cross-section and resonance integral for 186W(n,γ)187W reaction by the activation method using a single monitor

The thermal neutron cross-section (σ₀) and the resonance integral (I₀) of the reaction ¹⁸⁶W(n,γ)¹⁸⁷W were measured by the activation method using ⁵⁵Mn as a single comparator. The diluted MnO₂ and WO₃ samples within and without a cylindrical Cd shield case were irradiated in an isotropic neutron field of the ²⁴¹Am–Be neutron source. The γ-ray spectra from the irradiated samples were measured by high resolution γ-ray spectrometry with a calibrated high purity Ge detector. The necessary correction factors for gamma ray attenuation, thermal and resonance neutron self-shielding effects, and the shape factor (α) for epithermal neutron spectrum were taken into account in the determinations. The thermal neutron cross-section for ¹⁸⁶W(n,γ)¹⁸⁷W reaction has been determined to be 39.5±2.3 b at 0.025 eV. This result has been obtained relative to the reference thermal neutron cross-section value of 13.3±0.1 b for the ⁵⁵Mn(n,γ)⁵⁶Mn reaction. The present value of 39.5±2.3 b for ¹⁸⁶W(n,γ)¹⁸⁷W, in general is in good agreement with most of experimental data and evaluated ones in ENDF/B-VI and JENDL-3.2 within the limits of error. The resonance integral has also been measured relative to the reference value of 14.0±0.3 b for the ⁵⁵Mn(n,γ)⁵⁶Mn monitor reaction using a 1/E^1+α epithermal neutron spectrum of the ²⁴¹Am–Be neutron source. By defining Cd cut-off energy 0.55 eV, the resonance integral obtained was 493±40 b. The existing experimental and evaluated data for the resonance integral are distributed from 290 to 534 b. The present resonance integral value agrees with some previously reported values.     

基于中子衍射的六方相Er1-xYxMnO3结构和磁学性能研究

为研究六方钙钛矿氧化物中磁结构和稀土离子对磁性能的影响,以六方相ErMnO3为母体,通过非磁性Y3+离子对磁性稀土Er3+离子的替代,利用X射线衍射、中子粉末衍射和磁性测量等表征手段,对六方化合物Er1-xYxMnO3的晶体结构和磁性进行系统研究。研究结果表明,Y3+离子的掺杂并未引起晶体结构和磁结构的明显改变,同时发现六方晶体结构随温度的增加呈各向异性热膨胀,在ab面内膨胀而沿着c轴收缩。中子衍射实验结果表明,Er04Y06MnO3样品的有序温度约为79 K,低温下系统的基态为反铁磁性,结构为Γ2（P63c′m′）或Γ4（P63′c′m）组态,与Er08Y02MnO3相同。磁化测试结果表明,非磁性Y3+离子掺杂量的增加削弱了磁化强度,2 K时磁性稀土离子Er3+的自旋有序在外磁场下表现出铁磁信号。该研究阐明了Er1-xYxMnO3的磁性本质和稀土磁性调控特性,为其潜在应用提供了有力指导。     

北京谱仪Ⅲ内冷却液用聚氨酯管的耐辐射性能研究

北京谱仪（BESⅢ）工作环境内存在以γ和中子为主的辐射,聚氨酯管作为BESⅢ束流管及加速器SCQ（超导四极铁）真空盒的冷却液导管,在BESⅢ工作环境下的耐辐射老化性能至关重要。本工作模拟聚氨酯管在BESⅢ环境内使用10a所接受的剂量,对聚酯型聚氨酯管和聚醚型聚氨酯管进行γ辐射和中子辐射。采用耐压测试、傅里叶红外光谱分析和热分析方法,研究辐射前后两种聚氨酯管的性能。对比结果表明,聚酯型聚氨酯管的耐辐射性能以及热稳定性好于聚醚型聚氨酯管,在接受γ辐射和中子辐射后,聚酯型聚氨酯管正常耐压强度基本未变,但聚醚型聚氨酯管在接受中子辐射后产生老化断裂。     

瞬发伽玛活化分析中3种探测器性能比较

利用中国先进研究堆(CARR)热中子束流孔道首次开展了瞬发伽玛中子活化分析(PGNAA)实验。对NH4Cl、Si、Fe、Al等4种样品进行了辐照,同时采用HPGe、LaBr_3、BGO 3种探测器对样品进行实时测量,在瞬发伽玛射线的能量为0.002~10 MeV范围内研究了3种探测器在宽能区的能量线性、能量分辨率、探测效率等性能。     

基于白光中子源的169Tm辐射俘获截面测量和共振参数分析

中子辐射俘获截面及共振参数在核工程设计、核天体物理等研究领域中有重要的应用价值。在中国散裂中子源（CSNS）反角白光中子束线（Back n）上,使用C6D6测量系统开展了169Tm辐射俘获反应测量。通过脉冲高度权重技术、共振吸收法和饱和归一法得到169Tm辐射俘获反应的产额。利用SAMMY程序拟合169Tm的产额数据,得到169Tm在1~100 eV能量区间的共振能量、中子宽度、辐射俘获宽度等共振参数。使用实验测得的共振参数和Reich Moore近似计算了169Tm在1~100 eV能量区间的辐射俘获截面。实验测量结果与ENDF/B Ⅷ.0数据库的推荐值总体符合较好,部分共振参数和截面存在一定的差异。产生这些差异的原因与Back n的源中子能谱结构、能量分辨率、实验本底的精度有关。     

Analyses of Data Acquired From High-Pressure$hbox Xe+ ^3!hbox He$Gas Ionization Chamber

It is shown that pulse-height spectra acquired from a high-pressure$hbox Xe+ ^3!hbox He$gas ionization chamber exposed to mixed gamma-neutron radiation fields can be unfolded into separate gamma ray and thermal neutron spectra. The procedure takes advantage of the unique shape of the recently discovered spectral response of thermal neutrons in this high pressure$hbox Xe+ ^3!hbox He$mixture. A template spectrum formed from only the pulse-height distribution of neutron signals is subtracted from the combined gamma and neutron spectrum to provide an estimate for the count of thermal neutrons. This procedure leaves a pure gamma spectrum for standard gamma ray spectrum analyses and isotope identification.     

Distribution of temperature rise in biological shield due to thermal neutrons

The aim of this work is the evaluation and calculation of the heat generation and temperature rise in local ordinary concrete due to capture of thermal neutrons. The total thermal and reactor thermal neutron fluxes were measured, and the results were used to evaluate the heat generation and temperature rise in ordinary concrete. In addition, the computer code ANISN (VAX version) and neutron multigroup cross-section library EURLiB-4 were used to calculate the total thermal neutron fluxes and the heat generation and hence the temperature rise. The results were displayed in curves to show the distribution of thermal neutron fluxes and the heat generation as well as the temperature rise in the shield thickness. The results showed a good agreement between the measured and calculated values. The results showed also that the heat generation and the temperature rise have their maximum values in the first layers of the shield thickness.     

Comparison of measured and calculated neutron and gamma ray energy spectra behind an in-line shielded duct

Integral experiments that measure the transport of 14 MeV neutrons through a 0.30-m-diameter duct having a length-to-diameter ratio of 2.83 that is partially plugged with a 0.15 m diameter, 0.51 m long shield comprised of alternating layers of stainless steel type 304 and borated polyethylene have been carried out at the Oak Ridge National Laboratory. Measured and calculated neutron and gamma ray energy spectra are compared at several locations relative to the mouth of the duct. The measured spectra were obtained using an NE-213 liquid scintillator detector with pulse shape discrimination methods used to simultaneously resolve neutron and gamma ray events. The calculated spectra were obtained using a computer code network that incorporates two radiation transport methods: discrete ordinates (with P₃ multigroup cross sections) and Monte Carlo (with continuous point cross sections). The two radiation transport methods are required to account for neutrons that singly scatter from the duct to the detectors. The calculated and measured neutron energy spectra above 850 keV agree within 5–50% depending on detector location and neutron energy. The calculated and measured gamma ray energy spectra above 750 keV are also in favorable agreement, 5–50%, depending on detector location and gamma ray energy.     

瞬发γ射线中子活化分析的现状与发展

概述了瞬发γ射线中子活化分析(PGNAA)的现状与发展，重点介绍了PGNAA的基本原理，国际上已建的PGNAA装置概况、PGNAA方法学中的单比较器k0法和高能γ射线的效率刻度技术，以及PGNAA在新材料、环境样品、生物和药物以及矿物中H、B、N、Cl等轻元素分析测定方面的日益发展的应用状况。此外，还介绍了在中国原子能科学研究院重水研究堆上建立的热中子束PGNAA实验装置。     

油砂地层中子-伽玛能谱Monte-Carlo模拟

本工作应用自行研制开发的中子 伽玛能谱测井的中子 伽玛射线随空间、能量、时间分布的蒙特卡罗模拟软件包计算了裸眼井、井眼里无下井仪器、油饱和砂岩地层、井眼注油条件下沿井轴和井壁的中子 伽玛射线随能量、时间分布及沿井轴和沿井壁的中子 伽玛射线能谱 ,研究了源距对中子 伽玛射线随能量、时间分布以及井轴与井壁中子 伽玛射线能谱的影响。     

Characterization of an Am-Be PGNAA set-up developed for in situ liquid analysis: Application to domestic waste water and industrial liquid effluents analysis

A prompt gamma neutron activation analysis (PGNAA) set-up with an Am-Be source developed for in situ analysis of liquid samples is described. The linearity of its response was tested for chlorine and cadmium dissolved in water. Prompt gamma efficiency of the system has been determined experimentally using prompt gamma of chlorine dissolved in water and detection limits for different elements have been derived for domestic waste water. A methodology to analyze any kind of liquid is then proposed. This methodology consists mainly on using standards with water as bulk or in the case of absolute method, to use gamma efficiency determined with prompt gammas emitted by chlorine dissolved in water. To take into account the thermal neutron flux variations inside the samples, flux monitoring was carried out using a He-3 neutron detector placed at the external sample container surface. Finally, to correct for the differences in gamma attenuation, average gamma attenuations factors were calculated using MCNP5 code. This method was then checked successfully by determining cadmium in industrial phosphoric acid and our result was in good agreement with that obtained with inductively coupled plasma (ICP) method.     

Improved Los Alamos model applied to the neutron induced fission of 235U and 237Np

The Los Alamos model, with multiple fission chances, was upgraded to include the linear relation between the average prompt gamma ray energy and the average prompt neutron multiplicity. A global, model-free parameterization of this relation versus the charge and mass number of fissioning nuclei is obtained. The average fission fragment kinetic energy dependence on the incident neutron energy is also taken into account. Employing this model for the n+²³⁵U and n+²³⁷Np reactions, the prompt fission neutron spectra, the average prompt neutron multiplicity and the average prompt gamma ray energy are obtained in very good agreement with the experimental data, proving a better predictive power.     

Evaluation of Neutron Capture Gamma-Ray Spectra in Hafnium and Tantalum

A method has been developed for evaluation of neutron capture γ-ray spectrum. It couples measured intensities of primary and secondary discrete—-ray with a γ-ray cascade model to calculate the unresolved part of the capture spectrum, and adds the discrete part and the unresolved part to obtain the whole spectrum. The cascade model uses the level density formula proposed by Gilbert & Cameron and the Brink & Axel form of El γ-ray profile function with a modification. This method was applied to thermal neutron capture spectra in six hafnium isotopes and ¹⁸¹Ta and was extended also to non-thermal capture spectra in ¹⁸¹Ta for 0.25 and 0.5 MeV neutrons with empirical assumptions. The calculated results were compared with experiments and agreement was good not only in terms of the gross structure, but also in terms of the fine structure which appears at high and low γ-ray energies.