反应堆中子注量率相对分布测量系统铅屏蔽计算分析

为批量化测量探测片活性,设计了多道中子注量率相对分布测量系统。每一个NaI探测器测量一个探测片,探测器之间采用铅屏蔽。为了消除其他通道探测片对本通道探测片测量结果的影响,通过设计保证邻近通道活化探测片对本通道干扰小于0.05%。本文采用点源积分方法计算(经蒙特卡罗模拟验证),并确定了满足屏蔽要求的铅屏蔽体尺寸。     

涂硼正比计数管慢化体对中子探测效率影响研究

涂硼正比计数管是一种常用的反应堆源量程探测器,对热中子测量有很高的探测效率,对于快中子反应堆则需要增加合适的慢化体,提高中子探测效率。本文利用蒙特卡罗程序MCNP,模拟计算涂硼正比计数管在不同慢化体厚度的情况下,对各能量单能中子的相对探测效率和绝对探测效率,得到在不同慢化体厚度下,计数管的相对探测效率和绝对探测效率与中子能量的关系。最后针对快中子反应堆的典型中子能谱,模拟计算涂硼正比计数管在不同的慢化体设计时的探测效率,得出了一种优化的慢化体设计方案,对快中子反应堆核测量系统设计具有一定指导意义。     

探测快中子的新技术

本文概述了近年来探测快中子的新技术，它们分别是：抑制γ射线、热中子和带电粒子的符合谱仪；高分辨宽能量范围正比计数器的＾３Ｈｅ夹心谱仪；含氢的纤维闪烁体用于抑制γ射线、中子位置分布和中子能谱测量；及含锂的纤维玻璃闪烁体用于长中子计数器测平均中子能量；中子的直接探测；用于中高能和重离子核物理的多元件阵列快中子探测器和极化仪；用于核核查的中子源影像探测器；高入射氘核能量和高能中子的伴随粒子技术等７个方面     

快中子能谱在线测量的反冲质子望远镜响应模拟研究

快中子能谱是基于散裂中子源开展大气中子单粒子效应研究的关键输入参数,在线测量宽能区快中子能谱在近散裂靶位置面临飞行时间法不确定度大、中子通量高、本底干扰强等问题。设计了反冲质子望远镜(RPT)系统,利用Geant4模拟了20～200 MeV中子轰击不同厚度聚乙烯转换靶产生的反冲质子产额、角分布以及能谱,为优化探测系统设计提供了指导依据。通过模拟硅探测器与新型快响应CLLB闪烁体组成的二重符合RPT系统对入射中子的响应,分析了影响探测系统探测效率和能量分辨率的因素,确定了聚乙烯转换靶厚度为1 mm、符合质子探测器摆放角度为26.6°和探测器尺寸等重要参数,得到了RPT系统的中子响应函数矩阵,并计算了其探测效率达10-5,对高中子通量和复杂本底干扰环境下的快中子能谱在线测量具有指导意义和参考价值。     

GdI3:Ce闪烁体中子探测性能测试与改进研究

为改善GdI₃:Ce闪烁体在探测中子过程中的γ抑制能力,使用Geant4和XCOM计算了其γ线性吸收系数,并通过模拟计算与实验测量研究了铅屏蔽法抑制γ的有效性。结果表明：GdI₃:Ce闪烁体在探测中子过程中易受低能γ射线的干扰;随着铅层厚度的增加,100 keV~1 MeV的γ射线对中子探测的干扰减小,而3~10 MeV的γ射线的干扰呈先增加后减小的趋势。对²⁵²Cf中子源的实验测试发现,在碘化钆闪烁体外围添加铅层后,中子峰得以显现;随着铅层厚度的增加,中子峰内净计数减小,而净计数与本底计数的比值上升。模拟和实验结果均表明,在使用GdI3:Ce闪烁体探测中子时,应根据中子探测效率和信噪比的优化确定γ屏蔽铅层的厚度。     

精确测定ST451型快中子探测器的有效中子阈和电子的相对闪烁响应

本实验用薄膜￣（２５２）Ｃｆ源的裂变中子测定ＳＴ４５１型快中子探测器的有效中子阈。文中叙述了用单能γ射线源刻度探测器对电子能量响应的方法，康普顿边的位置通过比较测量的脉冲幅度谱和ＭｏｎｔｅＣａｒｌｏ模拟的分布与探测器的脉冲幅度分辨进行折叠的理论谱被精确地确定。给出了中子能量在７Ｍｅｖ以下，ＳＴ４５１型快中子探测器的有效中子阈和电子的相对闪烁响应数据。     

精确测定ST451型快中子探测器的有效中子阀和电子的相…

本实验用薄膜＾２５２Ｃｆ源的裂变中子测定ＳＴ４５１型快中子探测器的有效中子阈。文中叙述了用单能ｒ射线源刻度探测器对电子能量响应的方法，康普顿边的位置通过比较测量的脉冲幅度谱和Ｍｏｎｔｅ Ｃａｒｌｏ模拟的分布与探测器的冲幅度分辨进行折叠的理论谱被精确地确定，。给出了中子能量在７ＭｅＶ以下，ＳＴ４５１型快中子探测器的有效中子阈和相对闪烁响应数据。     

聚碳酸酯和硝酸纤维素塑料的快中子反冲径迹响应特性

本文概述了聚碳酸酯和硝化纤维膜中的反冲径迹的蚀刻性能,径迹的直径和中子能量的关系,中子探测效率与能量的关系及其紫外线效应。研究表明,上述两种膜可作为快中子阈探测器,而聚碳酸酯还可作快中子的雷姆计数器,用来监测中子源、反应堆和加速器的快中子通量和剂量。测量剂量范围是1—1000拉德。如进一步采用电化学蚀刻技术扩大径迹,可测剂量下限为几个毫拉德的量级,这就可期望作常规的中子剂量测量。     

一个聚乙烯长计数管效率的刻度

长计数管是一种简便而可靠的快中子探测器。由于它有探测效率高,探测效率随中子能量的变化缓慢,以及易于甄别γ射线本底等优点而被广泛用做快中子通量测量的次级标准。     

岷江试验堆Q15孔道中子谱测量试验的研究

为了确定岷江试验堆(简称MJTR)堆芯铍反射层内中子能量E >1.0MeV的快中子注量率,采用一组中子探测片测量了岷江试验堆内第一层铍反射层外Q15孔道内的快中子注量率谱,并由测量谱求得各阈能探测片的快中子反应的谱平均截面和该孔道测量点位置的快中子注量率。结果表明,铍反射层内快中子(E>1.0MeV)注量率为1.985×1012 cm-2·s-1,MJTR每运行一段,铍反射层内快中子(E>1.0MeV)积分注量最大可达1.54×1018cm-2。通过本次试验研究,为更好的开发利用MJTR提供了试验依据。     

Some angular properties of the scattered photons for 1.43 MeV and 2.75 MeV source photons penetrating aluminium,steel and lead shields

Experimentally determined angular flux spectra for 1.43 and 2.75 MeV source photons from disc geometry penetrating shielding slabs of lead, steel and aluminium were analyzed. For both source energies, the proportion of photons scattered in a forward direction increases with increasing penetration thickness. Also, at any polar angle, the scattered photon energy decreases experimentally with increasing shield thickness. Changes in the scattered photon energy spectrum with polar angle and shield thickness are discussed. For the steel shields, angular contributions to the scalar build-up factor are presented.     

Calculation of photon shielding properties for some neutron shielding materials

The objective of the present study is to calculate photon shielding parameters for seven polyethylene-based neutron shielding materials. The parameters include the effective atomic number(Z_(eff)), the effective electron density(N_(eff)) for photon interaction and photon energy absorption,and gamma-ray kerma coefficient(kc). The calculations of Z_(eff)are presented as a single-valued and are energy dependent. While Z_(eff)values were calculated via simplistic powerlaw method, the energy-dependent Z_(eff)for photon interaction(Z_(PI-eff)) and photon energy absorption(Z_(PEA-eff)) are obtained via the direct method for energy ranges of 1 keV–100 GeV and 1 keV–20 Me V, respectively. The kccoefficients are calculated by summing the contributions of the major partial photon interactions for energy range of 1 keV–100 MeV. In most cases, data are presented relative to pure polyethylene to allow direct comparison over a range of energy. The results show that combination of polyethylene with other elements such as lithium and aluminum leads to neutron shielding material with more ability to absorb neutron and crays. Also, the kerma coefficient first increases with Z of the additive element at low photon energies and then converges with pure polyethylene at energies greater than 100 keV.     

Angular properties of 6.13 MeV source photons penetrating lead,steel and concrete

Experimentally determined angular flux spectra for 6.13 MeV source photons from disc geometry penetrating shielding slabs of lead, steel and concrete are analysed. For all three shield materials, the proportion of photons scattered in a forward direction increases with increasing penetration thickness. At any given polar angle, the scattered photon properties decrease exponentially with increasing shield thickness. Angular exposure dose build-up factor is defined and angular contributions to the scalar build-up factor are presented. Changes in the distribution of photon energies relating to shield penetration thickness and polar angle are discussed. The influence of bremsstrahlung and positron annihilation photon sources are considered.     

Shielding study on iron and concrete assemblies of bremsstrahlung photons and photoneutrons from copper target bombarded by 18, 28 and 38 MeV electrons

Following the angular distribution measurements of bremsstrahlung photons and photoneutrons, we measured the distributions of photon and neutron dose rates in the iron and concrete assemblies using a copper target bombarded by 18, 28 and 38 MeV electrons at the electron linear accelerator (linac) of Hokkaido University. In this experiment, seven types of shielding assemblies of iron and concrete layers were used and the photon and neutron dosemeters were inserted into the assemblies to get the depth–dose distribution. The measured results were compared with the results calculated using the Monte Carlo code MCNP5 to verify the calculated results. The calculated results of the ambient dose equivalent rates were in agreement with the measured results within 30% accuracy. Since no work on the radiation behavior in the shielding wall of medical linac room has ever been reported, this work gives valuable benchmark data for the detailed shielding design with high accuracy.     

Gamma-ray and neutron shielding efficiency of Pb-free gadolinium-based glasses

The radiation shielding efficiency of material depends upon photon attenuation, exposure buildup factors and neutron removal capacity. A newly developed Pb-free gadolinium-based glasses in compositions(80-x) B_2O_3-10 Si O_2-10 Ca O-x Gd_2O_3(where x = 15, 20, 25, 30 and35 mol%) had completely been investigated for their shielding efficiency with Geant4 simulation for mass attenuation coefficients and neutron total macroscopic cross section and by calculating exposure buildup factors.The exposure buildup factors for photon energy from 0.015 to 15 Me V had been calculated up to 40 mean free paths using five factors geometric progression method. The mass attenuation coefficients of the Pb-free glasses were simulated for energies from 223 to 2614 ke V and compared with the possible available experimental results. The neutron shielding efficiency of these glasses was discussed by calculating neutron total macroscopic cross section for energies from 1 e V to 14.1 Me V. Present investigations are found to be very useful for applications in nuclear engineering.     

10—in单球rem计测量中能重离子反应中子剂量当量的理论修正

中能重离子反应出射的中子具有较复杂的能谱,在穿过混凝土屏蔽层后,其能谱发生显著的变化。考虑到中子rem计的能量响应,在中能重离子反应出射中子理论计算能谱和角分布的基础上,估算了屏蔽层外中子能谱的变化和用10-in单球rem计在屏蔽层外测量中能重离子反应中子剂量当量时的理论修正系数。     

贫化铀用于核动力堆防护的屏蔽性能与放射性毒性计算分析

针对贫化铀的γ射线屏蔽进行了实验与模拟计算验证。构建了核动力压水堆屏蔽模型,模拟输出的屏蔽层内中子能谱与实际能谱分布较为一致。采用蒙特卡罗程序与燃耗计算程序相耦合的方法,模拟计算了贫化铀在不同位置处中子、γ混合辐射场中的综合屏蔽性能,并与铅作为屏蔽材料进行了对比分析。模拟计算了屏蔽层中子辐照贫化铀40 a后的活化和裂变产物,分析了材料辐照前后年摄入量限值（ALI）定义下的放射性毒性,结果表明,新增二次产物对放射性毒性影响不大。      

在特定实验条件下的散射中子本底研究

研究了d-T中子源与探测器距离较近时,扣除实验大厅散射中子本底的方法。实验上采用屏蔽法,用了铀裂变电离室。用MCNP/4A程序和FENDL2库数据计算了实验大厅散射中子本底曲线。采用实验和计算相结合的方法扣除了在特定实验条件下的散射中子本底,方法是可行的。     

硼铝复合材料热中子屏蔽性能分析

硼铝复合材料因制备工艺简单,力学性能良好,原材料价格低廉等诸多优点被广泛研究,并被用作诸多领域的热中子吸收材料。本文采用理论计算、MCNP软件模拟、实验测量等多种方法对硼铝复合材料的热中子屏蔽性能进行了评估分析。通过理论计算发现,对于相同配比的硼铝复合材料,从材料的热中子吸收性能方面,添加硼单质的效果优于添加碳化硼。通过MCNP程序模拟计算和实验测量发现,硼铝复合材料对能量低于10^-7 MeV的中子吸收效果比较显著。     

基于MCNP对γ辐射积累因子不同影响因素的研究

通过蒙特卡罗程序来模拟计算γ辐射积累因子,以找出不同条件下积累因子受各因素的影响,为屏蔽研究提供一定的数据参考。就γ辐射积累因子的影响因素:γ光子能量,源的几何尺寸,辐射角和屏蔽层厚度,通过MCNP程序进行了模拟计算。初步结论为:轻元素和中等元素构成的介质在厚度一定的情况下,积累因子随着γ光子初始能量的减小而增大;相对于轻材料,重材料的积累因子较小;随着源的线度增大而增大;随着准直角进一步增大而增大,源的各向同性程度增高会导致积累因子增加;随着源与探测器之间介质厚度的增加,积累因子增大,对于高能辐射源和具有中偏低原子序数Z的元素,积累因子增长速率接近于线性。