CR-39测量中子能谱响应函数的计算

在一定的化学和电化学蚀刻条件下，计算了由聚乙烯、铅（金）和CR－39组成的叠层探测器的响应函数（1－20MeV）。计算结果表明，叠层探测器的响应函数分布近似为高斯分布，由铅组成的探测器的响应函数的半高宽比由金组成的探测器的响应函数的半高宽窄，由前者组成的探讨器测量中子能谱分辨率高于后者。     

CR-39测量中子能谱响应函数的计算

计算了在一定的化学和电化学蚀刻条件下,由聚乙稀、铅(金)和CR-39组成的叠层探测器的响应函数(1～20 MeV)。计算结果表明,叠层探测器的响应函数分布近似为高斯分布,由铅组成的探测器的响应函数的半高宽比由金组成的探测器的响应函数的半高宽窄,由前者组成的探测器测量中子能谱分辨率高于后者。     

中子能谱快速测量

为了实现中子能谱的快速获取,准确测量中子辐射场的剂量率,设计了一款一体化多球中子能谱仪。该系统能进行中子能谱的在线测量,实时显示中子剂量率。谱仪在Am-Be参考辐射场进行了验证,测量能谱与标准谱符合较好,转换后剂量率测量值与真值偏差<±7%。     

用阈探测器测量热中子引起的U~(235)裂变中子能谱

通过用阈探测器测量热中子引起的~(235)U裂变中子能谱,研究了SAND-Ⅱ和RDMM方法的解谱能力。并用蒙特卡罗误差分析程序分析了两种解谱方法的输入数据误差。对SAND-Ⅱ方法中解的唯一性误差和RDMM方法中权重函数的影响进行了讨论。两种方法对未知谱中结构的灵敏程度也作了简单的检验。结果表明,SAND-Ⅱ和RDMM是解快中子光滑变化能谱的较好方法。当未知谱中有简单结构时,用SAND-Ⅱ方法更为合适。     

核动力堆安全壳内外中子能谱和剂量测量

利用自制的多球谱仪测量了某核动力反应堆安全壳内外的中子能谱和剂量当量率。对安全壳外测量，中心探测器为球形^3He正比计数管；对安全壳内测量，中心探测器为球形金箔。系统的响应函数用MCNP程序计算，解谱程序为MIEKEB。为验证系统响应函数计算的准确性，进行了一些实验测量，并与理论计算结果进行了比较。结果表明，测量结果与计算结果在不确定度范围内相吻合。     

用阈探测器测量热中子引起的U~(235)裂变中子能谱

通过用阈探测器测量热中子引起的~(235)U裂变中子能谱,研究了SAND-Ⅱ和RDMM方法的解谱能力。并用蒙特卡罗误差分析程序分析了两种解谱方法的输入数据误差。对SAND—Ⅱ方法中解的唯一性误差和RDMM方法中权重函数的影响进行了讨论。两种方法对未知谱中结构的灵敏程度也作了简单的检验。结果表明,SAND-Ⅱ和RDMM是解快中子光滑变化能谱的较好方法。当未知谱中有简单结构时,用SAND—Ⅱ方法更为合适。     

西藏地区天然中子能谱测量

本文采用多球中子谱仪和中子周围剂量当量（率）仪分别对西藏地区的天然中子能谱和周围剂量当量率进行了测量,得到了西藏地区不同海拔处的室外天然中子能谱和周围剂量当量率。研究结果表明：该地区室外天然中子的能谱形状基本保持不变,其各能区的中子注量率随海拔的增加而增大,天然中子的总注量率和有效剂量率及周围剂量当量率均随海拔的增加呈指数规律增大;此外,天然中子的有效剂量率可用中子周围剂量当量（率）仪的测量结果乘以能谱 有效剂量转换因子得到。     

低本底β计数-α能谱测量装置

本文介绍了一种为远洋现场调查而研制的低本底β计数-α能谱测量装置。该测量装置能同时进行低活度的β测量和α能谱分析,克服了用两套仪器分别测量β计数和α能谱的麻烦,并减小了测量误差。     

同位素中子源参考辐射能谱测量及模拟中子谱分析

不同成分中子参考辐射场是对中子防护材料性能测试的基本条件。为获得较理想的测试中子防护材料的中子参考辐射场,构建基于³He正比计数管的多球中子能谱测量系统,对基于²⁴¹Am-Be和²⁵²Cf中子源的中子辐射场慢化情况进行MCNP模拟和能谱实验测量。为准确判断实验室校准参考位置的中子散射情况,基于多球中子能谱测量系统,开展了中子散射成分测量。采用ICRP推荐的注量-剂量转换系数对能谱进行了剂量率转化,与使用长计数器的测量结果进行比较,发现在校准参考位置的剂量率相对偏差≤20%。开展了聚乙烯慢化同位素中子源获得特定能谱的实验研究,进行了慢化谱的模拟计算和实验测量,两者结果符合较好。该工作为后续的中子计量测试和中子防护材料测试提供了可用的中子参考辐射场。     

傅里叶变换法扣除中子活化分析能谱本底

为了更加有效地扣除中子活化能谱散射本底,提高中子活化能谱分析的准确度,利用傅里叶变换本底扣除法,对MCNP程序包模拟14 Me V脉冲中子照射煤样获取的活化γ能谱进行了本底扣除。通过对全能峰面积解析误差的分析,表明该方法可有效提高特征峰面积的解析准确度。     

长硼计数器测量脉冲中子的漏计数问题探讨

根据BF3的探测原理,中子在聚乙烯的输运过程及管内10B（n,）7Li反应机理,通过大量的测量实验,模拟出脉冲测量中BF3漏计数概率很小的边界条件,该测量方法适合目前脉冲中子装置中的中子产额测量。     

Measurement Method of Ultra-low Fast Neutron Background

正Researches,such as measurement of doublebeta decay in nuclear physics,neutrino measurement in particle physics and dark matter directmeasurement in new physics,which need critical environments which radioactive background should be extremely low can be called ultra-low     

有源中子计数法测量铀总量方法研究

含铀物料中铀总量是核材料衡算与控制中最关注的信息之一,因非破坏性测量分析方法(NDA)不破坏样品,不产生二次废物,可对样品整体测量,因此常用作含铀物料中铀总量的测量分析.本研究基于有源中子计数法建立测量含铀物料中铀总量的新方法.对10组铀的标准样品进行测量分析,结果表明,测量误差为1.7％,小于5％,达到非破坏性分析技...     

反应性测量中本底与中子源修正计算方法研究

在核电厂反应堆的反应性测量中,γ本底电流、中子源引入的电流、电路噪声等干扰项在实测电流中占有显著份额,给反应性的测量计算带来很大影响,在动态刻棒等反应性测量工作中必须对其进行修正。本文分析了本底与中子源电流对反应性测量的影响,并自主开发了一种本底电流、中子源电流修正方法--双重拟合法。该方法基于点堆模型,通过对包含本底与中子源的逆动态方程的分析,计算得到准确的反应性以及本底电流、中子源电流。通过试验堆动态刻棒实验的数据对本方法进行了验证,得到了理想的结果。     

New 6Li Sandwich Counter for Measurement of Neutron Spectra in Fast Neutron Systems

A new form of ⁶Li sandwich counter has been developed for use in measurements of neutron spectra in fast critical assemblies. The counters have two large surface area diodes approached very close to each other, on one of which is evaporated the target ⁶Li. This arrangement proves to provide a ten-fold increase of detection efficiency, while the resolution is improved by approximately ½ compared with currently used counters. The new counters were first tested on mono-energetic neutrons, after which neutron spectrum measurements were performed on natural uranium blocks, in the fast source reactor “YAYOI”, in the FCA IV-l-P and in the FCA V-3 core. The resulting spectra are compared with the measurements made with other types of counter and with calculated data. Proof was obtained that the new counter can serve practically and usefully for in-core measurements of fast neutron energy spectrum between 0.8 and 5 MeV.     

锎源中子照射下硫的瞬发γ能谱测量

利用瞬发γ中子活化分析技术对不同质量的单质硫进行了定性测量,实验使用的是252Cf源和6″×6″NaI探测器,测量时间为1200s。实验结果表明：测量得到的硫元素的瞬发γ特征峰十分明显,元素特征得到了体现,证明该测量系统可对物料中的硫元素含量进行分析。     

Measurement of Double-Differential Neutron Emission Spectra from Uranium-238

We have performed the measurement of neutron emission spectra from ²³⁸U using a time-of-flight technique, and deduced the following data; (1) the prompt fission neutron spectra for 2 MeV incident neutrons at two emission angles of 90° and 135°, (2) the double-differential neutron emission cross sections at the incident energies of 1.2, 2.0, 4.2, 6.1 and 14.1 MeV. The emission spectra and the cross sections for scattering process were also deduced by subtracting the fission neutrons from the experimental spectra. The experimental results were compared with other experiments and the evaluations of JENDL-3 and ENDF/B-IV.

From the fission spectrum data ranging from 2 to 12 MeV, we have derived the best fit parameters for the Maxwellian and Watt type distribution functions. The experimental spectra are described with the Maxwellian spectrum with temperature of 1.24–1.26 MeV and are softer than both evaluations.

The spectra and cross sections for inelastic-scattering showed substantial disagreement with the evaluations concerning the discrete levels between 0.5 and 1.2 MeV, and continuum neutrons due to evaporation and pre-equilibrium processes. The secondary neutron angular distributions at 14 MeV incident energy were reproduced fairly well with the systematics.     

Measurement and Analysis of Neutron Spectra in Lithium Fluoride and Polytetrafluoroethylene Piles

For the assessment of neutron cross section data for fluorine, angular neutron spectra in the lithium fluoride (LiF) and polytetrafluoroethylene ((CF₂)n) piles were measured in the energy range from a few keV to a few MeV by the time-of-flight method with an electron linac, and the results were compared with those calculated by using nuclear data from JENDL-2 and ENDF/B-IV. Spatial distributions of neutron and X-ray fluxes were also measured in the test piles by the activation method, and the influence of photoneutrons generated in the sample material on the neutron spectrum in each pile was estimated. As a result, it was found that their influence on the neutron spectrum shape below 1 MeV was not so large as was necessary to be taken into account for the present assessment.

The calculated spectra using the JENDL-2 data and the ENDF/B-IV data show generally good agreement with those measured in both piles. However, both calculations underestimate the neutron fluxes around several 100 keV, and overestimate those below 100 keV, when they are normalized in the energy range of 10 keV～1 MeV. Large discrepancies are found between the shapes of the measured and calculated spectra around the resonances of fluorine cross section below 100 keV. The present measurements and analyses suggest that the reevaluations of the inelastic and elastic scattering cross sections below 1MeV and the resonance cross sections below 100 keV are necessary to reduce the observed discrepancies.     

D-T中子照射下含硼聚乙烯球泄漏γ能谱测量

为了更清楚地了解含硼聚乙烯的屏蔽性能,用反冲电子法测量了D T中子照射下的不同B4C含量的聚乙烯球的泄漏γ能谱,并用MCNP/4A程序和ENDF/BⅤ库数据进行模拟计算。实验测量值和计算值在误差范围内符合得较好。