PGNAA系统中D-T中子管的慢化装置优化设计

为了提高PGNAA系统中D-T中子管的中子慢化效率,获得更高的热中子产额,借助蒙特卡罗模拟,确定了以铅为中子反射层、5个聚乙烯层和铅层相互交替作为中子慢化层、碳化硼含量为3%的含硼聚乙烯作为中子吸收层以及铅作为γ屏蔽层的中子慢化装置模型。针对中子产额为3×107 n/s 的D-T中子管,该慢化装置输出面低于5 eV中子通量可达5.28×106 n/s,占总中子通量的30.8%,有效提高了中子慢化效率。经过验证模拟结果能够满足实验要求。To improve the moderating efficiency of D-T Neutron Generator in PGNAA system, and get higher thermal neutron yield, the Monte Carlo code MCNP was used to optimize the moderation setup. The lead was selected as neutron reflector and gamma absorber, 5 polyethylene layers and 4 lead layers constituted the neutron moderator and 3% boron-doping polyethylene was selected as neutron absorber. For the yield of 3107 n/s D-T Neutron Generator, this moderation setup can provide a yield of lower than 5 eV of 5.28106 n/s, accounting for 30.8% of total neutron yield, dramatically improves the moderating efficiency. It is proved that the simulation results can satisfy the requirement of PGNAA system by preliminary experimental verification.     

基于D-T紧凑型中子源的快中子照相准直屏蔽体设计及中子束特性模拟研究

中子照相是一种重要的无损检测技术,它能用于火工产品、毒品和核燃料元件等的检测。基于紧凑型D-T中子发生器,完成了一个用于快中子照相的准直屏蔽体系统（BSA）的物理设计。根据D-T中子源的能谱和角分布建立了中子源模型,采用MCNP4C蒙特卡罗程序,模拟了准直屏蔽体系统中中子和γ射线的输运,准直中子束相对于单位源中子的中子注量可以达到9.30×10-6 cm-2,准直中子束中主要是能量大于10 MeV的快中子;在设置的样品平面直径14 cm的照射视野范围,准直束中子注量的不均匀度为4.30%,准直束中中子注量与γ注量的比值为17.20,中子通量和中子注量比值J/Φ为0.992,说明准直中子束有好的平行性;准直屏蔽体外的泄露中子注量率与准直束中子注量率相比降低了2个量级。所设计的准直屏蔽体能满足快中子照相的要求。Neutron radiography is an important nondestructive testing technique. It can be used to detect the explosive devices, drug and the nuclear fuel element, etc. A beam-shaping-assembly (BSA) based on a compact D-T neutron generator is designed for fast neutron radiography in this paper. D-T neutron source model is constructed based on the neutron energy spectrum and angular distribution data. The transportation of neutron and γ-ray in the BSA is simulated using MCNP4C code. The neutron fluence of the collimated neutron beam with respect to the neutron source of the unit source is 9.30×10-6 cm-2. The collimated neutron beams is mainly fast neutrons with energies greater than 10 MeV. In the irradiation field range with a diameter of 14 cm, the neutron fluence uniformity of the collimated beam is 4.3%, the ratio of the neutron fluence to the gamma fluence in the collimated beam is 17.20, and the neutron flux and the neutron fluence ratio (J/Φ) is 0.992 which indicates that the collimated neutron beam has good parallelism. The leakage neutron fluence in outside of BSA is two orders of magnitude lower than that of the collimated neutron beam. The designed BSA can meet the need of fast neutron radiography.     

Improvement of the determination of hydrogen content in a multicomponent sample by D-T generator

If a D-T generator is used as a neutron source to simultaneously measure the content of carbon, hydrogen and oxygen in a multicomponent sample by NIPGA (Neutron Induced Prompt Gamma-ray Analysis), the 14 MeV neutron flux can be regarded as a constant value. The relationship between the production of the hydrogen characteristic gamma-rays and its content is nonlinear. In this paper, we use MCNP (Monte Carlo N-Particle Transport code) to simulate the relationship and analyze it. In practical measurement of the characteristic gamma-ray, it's impossible to get the net count. Therefore, we use the experiment to obtain the relationship between the hydrogen content and the total count of its characteristic gamma-rays. If we use the relationship combined with the simulation result to calculate the hydrogen content, the metrical precision can be much increased. The deviation of hydrogen content between NIPGA and chemical analysis is less than 0.25%, which meets the requirement of coal industry.     

Improvement of the determination of hydrogen content in a multicomponent sample by D—T generator

If a D T generator is used as a neutron source to simultaneously measure the content of carbon, hydrogen and oxygen in a multicomponent sample by NIPGA （Neutron Induced Prompt Gamma-ray Analysis）, the 14 MeV neutron flux can be regarded as a constant value. The relationship between the production of the hydrogen characteristic gamma-rays and its content is nonlinear. In this paper, we use MCNP （Monte Carlo N-Particle Transport code） to simulate the relationship and analyze it. In practical measurement of the characteristic gamma-ray, it＇s impossible to get the net count. Therefore, we use the experiment to obtain the relationship between the hydrogen content and the total count of its characteristic gamma-rays. If we use the relationship combined with the simulation result to calculate the hydrogen content, the metrical precision can be much increased. The deviation of hydrogen content between NIPGA and chemical analysis is less than 0.25%, which meets the requirement of coal industry.     

D-T中子源n-γ密度测井规律的蒙特卡罗模拟

根据D-T反应中子的能谱和角分布数据,建立了中子源模型;根据石灰岩地层标准刻度井群数据,建立了井模型。采用MCNP程序模拟了井中中子和射线的输运,得到了不同地层密度、不同源距处NaI探测器中的混合能谱和非弹能谱。在混合能谱2.5~4.5 MeV能区开窗,混合射线相对计数随源距的变化曲线显示,源距应选择在20~80 cm,密度与混合射线计数之间呈现非线性关系。在非弹能谱1.0~8.0 MeV能区开窗,非弹射线相对计数随源距的变化数据显示,源距应选择在20~40 cm或80 cm附近,密度与非弹射线计数之间成近似线性关系。     

D-T中子源n-γ密度测井规律的蒙特卡罗模拟

根据D-T反应中子的能谱和角分布数据,建立了中子源模型;根据石灰岩地层标准刻度井群数据,建立了井模型。采用MCNP程序模拟了井中中子和射线的输运,得到了不同地层密度、不同源距处NaI探测器中的混合能谱和非弹能谱。在混合能谱2.5~4.5 MeV能区开窗,混合射线相对计数随源距的变化曲线显示,源距应选择在20~80 cm,密度与混合射线计数之间呈现非线性关系。在非弹能谱1.0~8.0 MeV能区开窗,非弹射线相对计数随源距的变化数据显示,源距应选择在20~40 cm或80 cm附近,密度与非弹射线计数之间成近似线性关系。     

Activation cross-sections of titanium isotopes at neutron energies of 13.5–14.8 MeV

The cross-sections for ~(46) Ti(n,2 n)~(45) Ti, ~(46) Ti(n,p)~(46 m+g) Sc+~(47) Ti(n,d*)~(46 m+g) Sc, ~(46)Ti(n,p)~(46 m+g) Sc, ~(47) Ti(n,p)~(47) Sc+~(48) Ti(n,d*)~(47) Sc, ~(47) Ti(n,p)~(47) Sc, ~(48) Ti(n,p)~(48) Sc+~(49) Ti(n,d*)~(48) Sc,~(48) Ti(n,p)~(48) Sc, and ~(50) Ti(n,α)~(47) Ca reactions were investigated around neutron energies of 13.5–14.8 Me V by means of the activation technique. Fast neutrons were produced by the~3 H(d,n)~4 He reaction. Neutron energies from different directions in the measurements were obtained in advance using the method of cross-section ratios for ~(90) Zr(n,2 n)~(89 m+g) Zr and ~(93) Nb(n,2 n)~(92 m) Nb reactions. The results obtained are analyzed and compared with the experimental data provided by the literature and verified nuclear data in the JEFF-3.3,CENDL-3.1, ENDF/B-VIII.0 libraries, as well as results calculated by Talys-1.9 code.     

D-T中子与平板型聚乙烯材料作用的次级中子角度谱实验研究

基于反冲质子法建立了一种测量D-T中子与平板型宏观样品作用的次级中子角度谱的实验方法.为保证探测器的能量线性并在较低的中子有效测量下阈(0.5 MeV)情况下获得好的中子-伽马射线甄别性能,采用高、低能段分别测量的方法.采用事件记录法,同时记录了次级中子和伴随伽马射线的脉冲形状甄别和脉冲幅度二维信息,利用基于ROOT数据分析平台编写的离线数据分析程序,完成了伴随伽马射线的挑选和扣除,以及高、低两能段反冲质子谱的拼接,并成功的将神经网络技术应用于中子能谱的解谱,获得了D-T中子与9和18 cm厚平板型聚乙烯材料作用的0.5-15 MeV的次级中子角度谱实验结果.实验模型的MC模拟由MCNP5完成,数据库采用ENDF-VI,实验结果和MC计算结果在实验不确定度范围内一致.关键词：D-T中子积分中子学反冲质子法次级中子能谱     

爆炸物检测中典型样品的瞬发γ谱测量

在中子检测爆炸物的研究中,利用14 MeV中子与原子序数大于5的原子核相互作用可产生特征射线的特性,采用伴随粒子法结合D-T中子飞行时间技术,使用尺寸为12.5 cm20 cm的大体积NaI(Tl)探测器,对爆炸物所含元素C,N,O以及一些模拟炸药样品进行了瞬发谱测量。获得了几种典型样品的特征谱,并对其进行了分析。实验结果与欧盟同期结果进行了比较,表明本实验研究达到了目前国际同类实验的水平,可以为中子检测爆炸物识别技术提供实验支持。     

散裂中子源耦合慢化器的中子学研究

利用蒙特－卡罗方法研究了散裂中子源中耦合慢化器的中子学特性。给出了冷中子与热中子慢化器的中子能谱。甲烷慢化器提供了性能非常好的冷中子,在低功率的散裂中子源中得到了应用。计算了慢中子引出的角分布。由于较低的氢密度,液态氢的漫化能力低于水与液态甲烷,但是可以通过增加预慢化器来弥补这一问题。2cm厚的水预慢化器层可以大约减少热量在低温慢化器中的热量沉积33％而不破坏中子特性。     

一维贫铀/聚乙烯交替系统中D-T中子诱发的~(232)Th(n,γ)反应率的测定与分析

开展了钍样品装置内钍核参数的积分中子学基础研究.参考混合堆概念设计搭建了内部放置了钍样品的一维贫铀/聚乙烯交替系统装置,采用加速器D-T中子源模拟聚变堆芯,利用前期开发的离线伽马测量方法测定了不同位置、不同中子谱情况下的232Th(n,γ)反应率,不确定度约为5%.结果显示,聚乙烯对14.1 MeV中子的慢化作用可有效提升钍俘获率,且贫铀对钍俘获率也有显著提升作用.实验结果与主流核数据库计算结果的对比显示,ENDF/B-VI.6和JENDL-3.3数据库的计算值比实验值平均约大6%,而较新的ENDF/B-VII.0数据库的计算值比实验值平均约大4%.因此,相比于之前数据库的钍核数据,ENDF/B-VII.0的计算值与实验结果匹配得较好,可作为相关概念设计的推荐核数据库.     

中子准直实验装置设计

设计了以中子发生器为源的中子准直实验装置,分别做了不同材料准直器的厚度、准直孔大小改变时,从准直孔和孔周围不同位置射出中子的实验,并将实验中得出的数据与理论值相比较,从而确定该实验装置的使用可行性.     

BNCT中子源用RFQ加速器

 分析了加速器作为硼中子俘获治疗(BNCT)中子源的优势,提出以射频四极场（RFQ）加速器作为BNCT用中子源的首选机型。对该RFQ的参数进行了选择,利用质子轰击锂靶近阈反应产生的前冲中子束能散低、散角小的优势,设定RFQ最终能量为1.9 MeV。采用“匹配均温”设计方法进行了此强流质子RFQ的束流动力学设计,并对设计方案进行了传输模拟,在入口归一化均方根发射度为0.25 mm·mrad、流强为100 mA时,束流传输效率为99.3％。选择合适厚度的锂靶,经过整形即可得到满足BNCT治疗需要的中子束。     

伴随粒子法γ能谱本底测量技术

为了解决伴随粒子法瞬发能谱实验中随机本底的影响,通过实验研究了标记中子和非标记中子与物质作用产生的瞬发信号和随机本底信号。对不同情况随机本底谱测量进行了分析,提出了小尺度样品和大尺度样品本底测量方式。研究表明:对小尺度样品,可以通过去掉样品,选择与有样品测量的谱相同的时间窗进行本底测量;对于大尺度样品,可以通过在随机本底范围选取时间窗进行本底测量。     

铝材料高能反射中子测量与计算

用~6~种阈探测器测得有无铝反射体时的绝对活化反应率, 获得了铝材料的高能中子反射率, 其值基本在1—1.14之间, 变化趋势比较一致. 结果表明: 高能反射中子场在小角度测量点比较弱, 而在大角度测量点, 尤其在120°点比较强. 高能中子反射率随阈能的增大而变小. 中子反射率的实验不确定度为3.4%—4.2%. 模拟计算结果与实验结果在大多数点基本相符.     

PGNAA煤质在线测量中同时提高C、O和其它元素测量精度的慢化体模拟

提出两种慢化体设计方案,第一种方案直接在原有慢化体中引入孔道,第二种方案在慢化体中增加一层铅之后再引入孔道.利用蒙特卡罗方法对两种方案进行研究.计算结果表明,采用第一种方案可以有效地提升C、O元素的测量精度,但同时会降低H、Si等元素的测量精度;采用第二种方案可提高C元素和O元素的测量精度,同时也可以提高H、Si等其它元素的测量精度.     

伴随粒子法检测隐藏爆炸物安检仪研发

为应对日益增长的恐怖活动, 研制了基于伴随粒子的中子飞行时间技术的隐藏爆炸物安全检测仪。介绍了该安检仪检测的工作原理,详细阐述了安检仪的伴随粒子探测器的DT中子管、屏蔽系统、支撑平台、阵列探测器、信号获取和处理系统、数据库、传输装置及外壳的研制过程。利用研制的安检仪对常用炸药和化学战剂进行了检测,结果表明该安检仪可以满足实际应用的需求。该仪器不仅可以用于安检通道内的爆炸物检测,通过训练也可以用于毒品和危化品的检测。     

贫铀球壳中D-T中子诱发的铀反应率的测量与分析

为校验次临界能源堆的概念设计,在R19.4/30.0 cm的贫铀球壳装置上采用活化法开展14 MeV中子学积分实验.布放6片贫铀活化片于球壳中与入射D离子束90°方向上的不同位置处活化,用HPGe探测器测量238U(n,γ)反应、238U(n,f)及235U(n,f)反应和238U(n,2n)各反应产物发射的特征γ射线,得到了相应的反应率.238U(n,γ)反应率的不确定度为3.6％-3.7％,238U(n,D和235U(n,f)反应率的不确定度为5.1％-5.9％,238U(n,2n)反应率的不确定为4.3％-4.7％.用MCNP5程序在ENDF66c数据库下进行模拟计算,238U(n,γ)反应率的计算值/实验值(C/E)为0.972-1.034,238U(n,f)和235U(n,f)反应率的C/E为0.983-1.058,238U(n,2n)反应率的C/E为0.979-1.019.     

伴随粒子法检测隐藏爆炸物安检仪研发

为应对日益增长的恐怖活动.研制了基于伴随α粒子的中子飞行时间技术的隐藏爆炸物安全检测仪。介绍了该安检仪检测的工作原理.详细阐述了安检仪的伴随α粒子探测器的DT中子管、屏蔽系统、支撑平台、阵列γ探测器、信号获取和处理系统、数据库、传输装置及外壳的研制过程。利用研制的安检仪对常用炸药和化学战剂进行了检测.结果表明该安检仪可以满足实际应用的需求。该仪器不仅可以用于安检通道内的爆炸物检测.通过训练也可以用于毒品和危化品的检测。     

Measurement of dose enhancement by neutron capture of boron in Cf interstitial brachytherapy

The effect of boron concentration in water on the gamma, fast and slow neutrons and alpha particles components at the central, forward and backward surfaces inside tumor phantom of 4.2 cm diameter and 4.4cm height, during brachytherapy by neutrons from ²⁵²Cf were investigated. The source was at the centre of a cubic shaped water phantom of 30 cm side. The study was carried for different concentrations of boron from H₃BO₃, Li₂B₄O₇ and H₃¹⁰BO₃. The effect of source to tumor distance on the different components of radiation was also measured. The results indicated that the use of ¹⁰B compounds enhances the damage and is recommended for successful boron neutron capture therapy (BNCT).