热释光中子个人剂量计蒙特卡罗设计

随着核工业、核电及中子辐射治疗的发展,对职业工作者的中子个人剂量监测显得越为重要。利用MCNP-3B蒙特卡罗程序对反照率热释光中子个人剂量计进行了设计计算,给出了光子及中子注量响应,并对中子的场所修正因子确定方法进行介绍。模拟计算表明,对光子和中子而言,适用的能量范围分别为33 keV～1.5 MeV和热中子-10 MeV,分别对应的能量响应偏差均小于30%和60%。     

CR-39塑料反冲径迹个人中子剂量计

本文叙述了一种用于测量中子剂量的“直接相互作用”型的CR-39塑料反冲径迹个人中子剂量计。该剂量计对~(241)Am-Be源的中子注量灵敏度约为(2.10±0.05)×10~(-4)径迹/n,可探测的中子注量下限约等于2.11×10~5n/cm~3,相应的中子剂量当量为7.81×10~(-5)Sv。总误差估计不超过±20％。     

一种中子注量率响应曲线的获取方法

针对一款常用的中子探测器,通过蒙特卡罗方法模拟得到中子注量率灵敏度—能量响应曲线,以单能加速器中子源的注量率响应验证曲线趋势,以同位素中子源对注量率响应做进一步修正,最终获得中子注量率灵敏度-能量响应曲线。     

中子个人剂量计少道解谱初探

针对中子辐射场所工作人员所受剂量评估结果偏差较大的问题,提出了用少道解谱方法来估算中子个人剂量当量。为检验该方法的有效性,本工作以已设计好的热释光个人剂量计（TLD）为研究对象,使用蒙特卡罗程序MCNP计算31个单能点的中子单位注量能量响应,通过Lagrange线性插值求得2×58响应矩阵,并利用MXD-FC31和GRV-FC31两种少道解谱软件分析了单能中子情况下的剂量估算问题。结果表明,少道解谱方法对中子个人剂量当量合理评价是可行的。     

中子气泡剂量计探测器用于混合中子场中子测量

气泡探测器具有中子实时剂量监测、宽能区中子能量响应稳定和探测限低的优势，因此，它在中子剂量计领域是一项有前景的技术。且被认为是唯一能满足国际放射防护委员会ICRP60要求的个人中子剂量计。目前，本实验室已经研制出具有自主知识产权的中子气泡剂量计。     

基于3层硅探测器的主动式中子个人剂量计研制

研制了一种具有能量分辨能力的中子个人剂量计。该个人剂量计由3层硅探测器组成，硅探测器周围装有⁶LiF、聚乙烯和含硼聚乙烯作为转化体、慢化体和吸收体。个人剂量计有直读和解谱两种工作模式。直读模式将中子能区划分为低能、中能和快中子3个能区，可实时测量。解谱模式可获得快中子能区的中子能量分布。利用GEANT4程序模拟了7 MeV γ射线的能量沉积谱，设置适当的甄别域以降低γ射线的影响。采用GEANT4程序模拟了个人剂量计对不同能量入射中子的个人剂量当量率响应。在加速器单能中子参考辐射场中完成了单能中子剂量响应的实验校准，对模拟计算的响应函数进行了实验修正，并得出了不同能区的平均中子个人剂量当量率响应。     

基于固体径迹探测的中子个人剂量计性能研究

CR-39固体径迹探测器作为一款被动式的中子个人剂量计,具有方便携带、价格低廉、抗干扰能力强、对γ、β射线不敏感、对快中子响应好等优点。针对新开发的一款中子个人剂量计,根据国际标准ISO 21909-1,对新型中子个人剂量计的CR-39固体径迹探测单元的相关性能开展了实验研究。实验结果表明,CR-39固体径迹探测器具备良好的测读重复性、批次均匀性、剂量线性、稳定性、参考辐射场响应以及对光子不敏感的属性,同时给出了CR-39测量中子剂量的剂量探测下限和不确定度,为国际标准引进提供了实验数据参考。     

受体对个人剂量计光子能量响应的影响

本文给出了 LiF-γ个人剂量计、LiF 反照中子剂量计和 Mg_2SiO_4-β手部剂量计在有、无受体情况下的光子能量响应。实验结果表明,有受体时光子能量响应明显地增高。     

基于CR39固体核径迹探测器的低能中子剂量测量方法初步研究

本文给出了一种采用固体核径迹探测器测量低能(0.5 eV以下能量)中子的双片径迹计数方法,利用蒙特卡罗模拟和实验,统计分析了低能中子引起的探测器单位面积径迹密度。结果显示,采用双片径迹计数方法,实现了CR39对于低能中子的响应,可以对低能中子引起的个人剂量进行估算。应用在中子职业个人剂量监测时,有效甄别了背景径迹的影响,因而降低了测量误差,对减小中子个人剂量监测的测量不确定度也是一个可行的方法。     

宽谱中子辐射场个人剂量评估方法的探讨

本论文基于反照率热释光个人剂量计设计,利用蒙卡程序计算了150多种宽谱中子辐射场的个人剂量计响应。鉴于两对探测元件具有不同响应的特点,通过二者的比值来推测职业受照人员所处工作环境中的中子平均能量方面信息。根据估计工作场所的平均能量,合理的引入了中子辐射场的修正因子,使个人剂量估算偏差进一步缩小。     

中子气泡探测器用于核测井中子个人剂量监测的研究

为研究国产中子气泡探测器在核测井中子个人剂量监测中的适用性,采用国内研制的中子气泡探测器对核测井运源车外表面、车内兼用储源仓周围等关注点的中子辐射水平进行监测,同时采用进口LB6411型中子周围剂量当量仪进行比对监测。实验结果表明,当运源车兼用储源仓内仅装载中子源时,中子气泡探测器与LB6411的测量结果无显著统计学差异,两者测量结果符合较好,中子气泡探测器的测量结果准确可信;当兼用储源仓分别装载中子源、中子-γ源时,两组中子气泡探测器的测量结果也无显著统计学差异,中子气泡探测器适用于中子-γ混合辐射场中子辐射剂量的测量。中子气泡探测器在运源车现场与在²⁴¹Am-Be源标准中子场中的剂量响应灵敏度因子间的相对偏差为7.4%,验证了其较好的能量响应特性,并显示了在核测井现场条件下用于中子个人剂量监测的适用性。     

热释光个人中子剂量计的刻度及性能试验

本文叙述了用作个人剂量监则的镉帽型热释光子中子剂量计的中子能响特性和刻度方法。用标准Ａｍ－Ｂｅ中子辐射场测量了剂量计的响应因子，检验了反平方律。给出了相对刻度的总不确定度为７．６％。对油田中子测井和使用中子水份测量仪等测量仪器的单位，可用该剂量计作个人中子防护剂监测。     

Flux and dose transmission through concrete of neutrons from proton induced reactions on various target elements

Simple empirical expressions for transmission of flux and dose through concrete are presented for neutrons from proton induced reactions. For this purpose the neutron emission from different targets in proton induced reactions in the energy range 25–200 MeV have been considered. The calculated effective dose outside a concrete shield shows overall good agreement with the effective dose estimated from measured neutron flux in the framework of the Moyer model. The calculated effective attenuation length shows a rising trend with incident proton energy and shield thickness.     

Radiation Dosage Due to Neutrons Produced in a Vehicle by Protons in the Trapped Radiation Belt

The rate of production of neutrons by geomagnetically trapped protons incident on a vehicle was measured by a neutron counting system carried into the trapped radiation belt by a pod flow piggy back on an Atlas rocket on December 19, 1961. The flux of neutrons produced by radiation belt protons incident on the pod was determined to be at least 700 neutrons/(cm2 sec); the actual value depends somewhat on the energy spectrum of the neutrons. This flux was estimated to be equivalent to a dose rate in tissue of 0.10 rems/hr. On the basis of proton flux measurements made in the radiation belt by Freden and White, a calculation was made of the tissue dosage which would have been received in the same environment directly from protons. These calculations were made by obtaining a numerical integration of the dE/dx times RBE times flux product over the entire energy spectrum. The total dose calculated amounted to 2.78 rems/ hr. Further calculations were made to estimate the dose rates which would have been received by tissue in the same environment but with varying amounts of shielding around the vehicle. The proton dose is, of course, reduced by the shield but the neutron dose actually increases as the shielding thickness is increased. It is seen that the neutron dose equals the proton dose at .3 rems/hr. when aluminum shielding of 2.6" surrounds the vehicle and it exceeds the proton dose with thicker shielding.     

Angular distribution of elastic and inelastic scattering of 2.34 Mev neutrons from chromium,iron, and lead

Spectra are obtained for neutrons with initial energy E₀=2.34 Mev scattered by chromium, iron, and lead nuclei, and the corresponding differential cross sections for elastic and inelastic scattering are measured for angles from 30 ° to 135 °. The neutrons were obtained from the D (d, n) He³ reaction with an initial deuteron energy of 1 Mev and a neutron emission angle of 110 °. Nuclear photographic emulsion was used as the detectorspectrometer and as the monitor of the incident neutron flux. The scatterers were 2.8 cm diameter spheres. In calculating the cross sections for elastic and inelastic scattering, corrections for self-absorption and multiple scattering were introduced to the incident neutron flux in the scatterer. The angular distribution of inelastically scattered neutrons from chromium and iron, as well as from lead when the 0.53 Mev level was excited, was found to be isotropic within the limits of the experiment. Inelastic scattering from lead when the 0.805 and 0.890 Mev levels are excited cannot be considered isotropic.     

测量热中子的CR－39 SSNTD

阐述了ＣＲ－３９ＳＳＮＴＤ实现热中子测量的机理，介绍了热中子ＣＲ－３９ＳＳＮＴＤ的制造方法及其在个人中子剂量计中的初步应用。     

Experimental Calibrations of a Silicon Rectifier Developed as a Fast Neutron Sub-Miniature Dosimeter

We can readily., summarize the significant points that have been elicited from these experiments. 1. The response of this type of dosimeter is fundamentally reproducible. It is felt that increased uniformity can be achieved by further optimization and control of the processing techniques. 2. The dosimeter can be given the desired, almost linear, sensitivity in the biological dose range of interest by selection of an appropriate base width (somewhere between 40 and 60 mils). 3. The response of the dosimeter is independent of the delivery rate of the neutron flux, up to a rate of at least 107 rads per second. 4. The dosimeter essentially does not respond to gamma radiation in the energr and dose ranges of interest. 5. The response of the dosimeter is essentially independent of its orientation with respect to the direction of the incident neutron flux. U. The dosimeter is subject to a significant amount (from 10 to 20 per cent) of room temperature anneal during the first few days after initial exposure. 7. The sensitivity of an exposed dosimeter changes when the device is anmealed at elevated temperatures. 8. The response of the dosimeter appears to be dependent on the energies of the incident neutrons. Some of these points refer to desirable features of the fast neutron dosimeter. The und-ersirable features, we feel, can be remedied or accounted for through further research and development efforts.     

基于单球多计数器、GM计数管同时测量n、γ剂量当量的装置研制

介绍了一款同时测量混合辐射场中n、γ剂量当量的装置研制,该装置通过单球多计数器和GM计数管相结合的方法构成,单球多计数器的探测器由三根两两垂直的3He正比计数管构成,通过解谱的方法得到中子剂量当量。在Am-Be和Cf-252中子参考辐射场中进行了测试,剂量当量相对误差30%;在能量范围为63keV-1.25MeV的入射光子下进行了能量响应试验,相对误差30%。