复杂环境条件下脉冲中子散射本底测量方法

介绍了在复杂环境条件下脉冲中子散射本底测量方法和计算公式,并在实验室进行了实验测量,分析了影响实验结果的因素。     

中子监测器定度中散射本底的研究

测定了几都典型中子监测器在用~(124)Sb-Be、~(241)Am-Be 及14MeV 中子源定度时的散射本底响应,比较了常用的几种确定散射本底的方法,并给出了拟合公式。     

中子小角散射实验及原始数据的处理

中子小角散射技术是研究纳米尺度范围材料结构的有力工具。中子小角散射实验测量和原始数据的处理方法相对较复杂。为了获得样品的绝对中子小角散射强度数据,通常需要进行入射中子束强度、散射强度、样品的透射率、实验本底以及空样品盒的散射强度和透射率等多项实验测量。若要获取较宽散射矢量范围的实验数据,还要改变实验仪器设置,对同一样品进行几次测量。而对所测数据也需要进行多项处理,才可获得便于分析的小角散射强度曲线。本文简单介绍实验原理和测量方法,重点讨论原始数据的处理方法,其中详细讨论了各向同性散射数据的平均以及合并方法。     

单球多计数器的中子能量响应计算

根据球体内随深度变化时中子的慢化程度有所差异．按两两垂直的方法把三个位置灵敏正比计数器安装在一个慢化球体内。用MCNP4A程序计算了6种慢化球体和6种照射方向的能量响应,同时对球半径方向两种分区方法的计算结果进行分析和比较。     

高灵敏度环境中子剂量当量仪的优化设计与实验验证

本工作采用MCNP程序对探测器探头的主要参数进行了模拟计算和优化选择。提出了高灵敏度环境中子剂量当量仪的设计方案。最后通过实验验证，试验结果与计算结果基本吻合。     

在回旋加速器上用α束进行在束测量时的γ本底研究

随着加速器技术的发展以及高分辨、高效率半导体探测器的不断提高,在束γ谱学越来越受到人们的重视,因为它是研究原子核高自旋态的有力工具,可以为研究核结构提供更丰富的信息。在一定人射能量的核反应过程中,往往可以同时打开几个反应道,用Ge(Li)探测器进行在束测量时,就会记录到许多γ射线,使在束γ谱变得相当复杂,在这样测得的γ谱中,还附加了许多本底γ峰,它们主要来自Ge(Li)探测器中的非弹性中子散射和中子核反应所产     

MCNP程序在核弹头核查中子探测技术研究中的应用

为合理设计基于核弹头泄漏中子被动测量核查方法的实验方案，根据假想核弹头模型，用MCNP程序计算了其泄漏中子能谱。结果表明，核材料自发裂变中子在透射出弹体后被慢化，大部分中子成为慢中子，能量小于0.5MeV。分析了这一结果对核武器现场核查中子探测技术的意义。      

CR—39探测器测量中子剂量的实验研究

研究了应用ＣＲ－３９探测器测量中子剂量的方法,蚀刻条件,刻度和实际应用。结果表明最佳蚀刻条件为：６．５ｍｏｌ／Ｌ ＫＯＨ,７０℃,ｈ刻度系数为７．０×１０＾－３ｍＳｖ／（径迹数．ｃｍ＾－２）,最低可探测下限为０．０６ｍＳｖ,可用于一般中子－γ混合辐射场中中子剂量的测量。     

质子和中子的单粒子效应等效性实验研究

通过实验方法确定了高能质子和中子引起的单粒子效应等效关系，实验中采用金箔散射法降低了质子束流强度，并利用热释光剂量计（TLD）进行质子注量的监测，采用新研制的存储器长线实时监测系统，进行了64K位至4M位的SRAM器件单粒子效应实验，确定了两种粒子引起单粒子效应等效关系。     

SPRR-300反应堆混凝土屏蔽层内中子注量率分布研究

采用MCNP程序与ANISN程序结合的计算方案获取了SPRR-300反应堆混凝土屏蔽层内的中子注量率分布情况，同时采用固体核径迹探测器测量了混凝土屏蔽层外低水平中子注量率，两者吻合较好，说明了计算结果的可信性。上述结果为反应堆退役工作提供了放射性源项的计算依据。     

Si-PIN探测器中子直照灵敏度研究

介绍了利用K600中子发生器进行Si-PIN探测器灵敏度标定的实验方法,并在实验中测出了Si-PIN探测器对14MeV中子的直照灵敏度。同时,利用MCNP模拟程序对Si-PIN探测器不同能量的中子直照灵敏度进行了理论计算,实验灵敏度处理结果和理论计算值较为一致。     

我国的热中子散射工作现况和展望

我国已在中国原子能科学研究院建成了国内唯一的、面向全国的热中子散射实验研究基地,并已形成了一些稳定的研究方向,取得了一些有学科价值和应用前景的成果。今后如能继续得到重视和支持,可望开拓一些新的课题,取得新的成果,为以后的发展打下基础。     

Study on porosity of ceramic SiC using small angle neutron scattering

The mechanical properties of functional heat-resistant ceramics SiC are significantly influenced by the concentration and idmensions of pores.Small angle neutron scattering measurements for 3 SiC samples with different densities are performed on C1-2 SANS instrument of the University of Tokyo.Two groups of the neutron data are obtained using 8 and 16m of secondary flight path,1 and 0.7 nm of neutron wave lengths,respectively,After deduction of background measurement and transmission correction,both neutron data are linked up with each other,The patterns of neutron data of 3 samples with Q range from 0.028-0.5nm^-1 are almost with axial symmetry,showing that the shape of pores is almost spherical.Using Mellin transform,size distributions of pores in 3 samples are obtained.The average size (-19nm)of pores for hot-pressed SiC sample with higher density is smaller than the others (-21nm).It seems to be the reason why the density of hot-pressed SiC sample is higher than not hot-pressed sample.     

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。     

Benchmark experiment on titanium with DT neutron at JAEA/FNS

Titanium is contained in lithium titanate which is a tritium breeding material candidate. In the nuclear design, accurate nuclear data are needed. However, few benchmark experiments had been performed for titanium. We performed a benchmark experiment with a titanium assembly and a DT neutron source at JAEA/FNS. The titanium assembly was covered with Li₂O blocks in order to reduce background neutrons. Dosimetry reaction rates were measured with niobium, indium and gold foils inside the assembly. And fission rates of ²³⁵U were measured by using micro fission chambers. This experiment was analyzed by using the Monte Carlo neutron transport code MCNP5-1.40 with recent nuclear data libraries of ENDF/B-VII.0, ENDF/B-VII.1, JEFF-3.1.2, JENDL-4.0 and JENDL-4.0u1. The calculation results were compared with the measured one in order to validate the nuclear data libraries of titanium. The calculated results with ENDF/B-VII.1 agreed with the measured one the best because the (n,2n) and (n,n′_cont) reaction cross section data and resonance parameters were improved.     

Design of the thermal neutron beam for neutron radiography at the Syrian MNSR

Neutron beam design was studied at the Syrian reactor (MNSR, 30 kW) with a view to generating thermal neutron beam in the vertical irradiation sites for neutron radiography. The design of the neutron collimator was performed using MCNP4C and the ENDF/B-V cross-section library. Thermal, epithermal and fast neutron energy ranges were selected as <0.4 eV, 0.4 eV–10 keV, >10 keV, respectively. To produce a good neutron beam quality, bismuth was used as photon filter. In this design, the L/D ratio of this facility had the value of 125. The thermal neutron flux at the beam exit was about 2.548 × 10⁵ n/cm² s. If such neutron beam were built into the Syrian MNSR many scientific applications would be available using the neutron radiography.