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.     

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.     

基于紧凑型D-D中子发生器的热中子照相慢化准直器设计

热中子照相是一种重要的无损检测技术,是X射线照相技术的重要补充,小型化热中子照相系统有重要研究应用前景。基于紧凑型D-D中子发生器,采用蒙特卡罗程序MCNP-4C,通过中子和γ射线的输运模拟,完成了热中子照相慢化准直器的模拟研究与设计。慢化准直器准直比约为3.58,模拟研究结果显示,在D-D中子发生器中子产额大于5×10⁸ n/s条件下,样品平面内热中子注量率可大于10³ n/(cm²·s),准直中子束中热中子占比可大于74%,在Φ70 mm的照射视野范围内,热中子注量的不均匀度约为7.3%,基本满足热中子照相的成像要求。     

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     

基于2.5 MeV质子加速器BNCT装置的中子慢化材料性能模拟研究

基于加速器中子源的硼中子俘获治疗(Boron Neutron Capture Therapy, BNCT)是新一代的放射治疗方法,束流整形体(Beam Shaping Assembly, BSA)作为硼中子俘获治疗装置的重要组成部分,其作用是将中子源中的快中子束流慢化至超热中子能区(0.5 eV~10 keV),并尽可能减少快中子、热中子以及$\gamma $射线的成分,使其满足BNCT用于治疗的中子束要求。本工作基于蒙特卡罗软件包Geant4(Geometry and Tracking),以2.5 MeV,10 mA质子流强的7Li(p, n)7Be中子源为对象,研究分析了AlF₃ 、Fluental、Al₂O₃、Al作为慢化体材料时,不同的厚度对束流出口处的超热中子注量率、超热中子注量与热中子注量比值、快中子成分、$ \gamma $成分所产生的影响。计算表明,当选用厚度为25 cm的AlF₃作为慢化体材料时,经过整形慢化后的超热中子束的束流参数,均满足国际原子能机构(International Atomic Energy Agency, IAEA)的中子束流参数推荐值。     

D-T中子^181Ta（n，2n）^180Ta^m反应截面的测量

Ta是裂变和聚变反应堆的重要结构材料。它具有高熔点和高中子倍增率等优点。^181Ta（n，2n）^180Ta^m付反应在监测En=9～20MeV中子能谱方面具有非常大的优越性，因为它有很大的反应截面，适合测量的半衰期和平滑的激发曲线，因此对^181Ta（n，2n）^180Ta^m反应截面的准确测量有重要的应用价值。     

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.     

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.     

2.45 MeV小型D-D中子管准直屏蔽结构设计

为获取小角度出射的单能中子源,采用蒙特卡罗软件对小型D-D中子管产生的能量为2.45 MeV的4立体角中子源进行了准直屏蔽结构设计。准直屏蔽结构分为准直器和捕获穴,准直器采用铁+含硼聚乙烯+铅的三层过滤结构,用于屏蔽照射野外杂散中子,捕获穴主要功能是增加反方向中子的弹性散射次数,从而降低出射束低能散射中子的比例。通过MCNP模拟得到了准直器各层材料的最佳厚度和出射孔尺寸以及捕获穴最佳结构。经验证,中子照射野处2.45 MeV的中子通量比照射野外大三个量级,中子照射野处低能中子通量比2.45 MeV的中子通量低一个量级,墙壁外总剂量率（中子+）在2.5 Gy/h以下。该研究对于小角度单能中子源的快速获取具有一定的实用价值,获取的中子源可用于中子剂量仪器有效性检验、中子监测仪性能测试等方面的研究。     

热中子分析探雷的中子源慢化装置设计与优化

与传统的地雷探测技术相比,热中子分析(Thermal Neutron Analysis,简称TNA) 探雷技术具有准确率高、虚警率低和对环境适应性强的特点,但探测速度较慢,制约了其广泛应用。为了提高地雷位置处的慢热中子通量,缩短探测时间,提出了一种基于252Cf 的中子源慢化装置设计构型,主要包含中子慢化层、中子反射层、本底 屏蔽层和侧向中子吸收层4 个部分。采用数值模拟的方法比较了4种常用中子慢化(反射) 材料的性能,优选高密度聚乙烯作为慢化材料,石墨作为反射材料。同时,为了满足辐射安全要求,对屏蔽材料的结构进行了优化计算。按照设计构型搭建了TNA 探雷实验平台。在104 n/s 中子源强下优化了慢化层和反射层的厚度,测试了装置慢化效能,在107 n/s 中子源强下评估了装置辐射安全性能。结果表明,采用该装置可使地雷位置处的慢热中子通量提升11 倍以上,并能有效保障辐射安全。Compared with the traditional landmine detection methods, Thermal Neutron Analysis (TNA) landmine detection has the advantages of high accuracy, low false alarm rate and strong adaptability to the environmental change.But the long detection time restrict the wide application of this technology. In order to shorten the detection time, one possible design of neutron moderation device based on 252Cf neutron source is proposed to enhance the moderated neutronflux in mine position. The device consists of four parts, the neutron moderator, the neutron reflector, the background shield and the useless neutron absorbing layer. Then, the performance of four widely used materials in neutronics was compared with MCNP5 code, and HDPE was chosen as the neutron moderator material, graphite as the neutron reflector material. The thickness of the useless neutron absorbing layer was optimized at the same time. Finally, an experimental platform of 252Cf neutron moderation device was assembled on the basis of simulation results, and a series of experiments were carried out to optimize the geometric dimensions and evaluate the dose equivalent with two different strengths neutron source, 104 and 107 n/s. The results indicate that this device can effectively enhance the thermal neutron flux at mine position by more than 11 times and ensure the radiation safety.     

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。     

Monte Carlo simulations of a D-T neutron generator shielding for landmine detection

Shielding for a D–T sealed neutron generator has been designed using the MCNP5 Monte Carlo radiation transport code. The neutron generator will be used in field for the detection of explosives, landmines, drugs and other ‘threat’ materials. The optimization of the detection of buried objects was started by studying the signal-to-noise ratio for different geometric conditions.     

RIBLL终端大面积中子探测器阵列研制方案的理论计算

介绍了一种用于丰中子核研究的RIBLL终端大面积中子探测器阵列的研制.利用Geant4软件包对该探测系统的中子探测效率、时间分辨和位置分布等进行了蒙特卡罗模拟,同时对探测器的cross?talk进行了研究.给出了探测器的设计方案.     

Neutron moderation in a bulk sample and its effects on PGNAA setup geometry

In a prompt gamma ray neutron activation analysis (PGNAA) setup, the neutron moderation in the bulk sample also plays a key role. This can even dominate the thermalization effects of the external moderator in some cases. In order to study the neutron moderation effect in the bulk sample, moderators with two different sizes of the sample were tested at the King Fahd University of Petroleum & Minerals (KFUPM) PGNAA facility. In these tests, the thermal neutron relative intensity and prompt gamma ray yield from the two moderators were measured using nuclear track detectors (NTDs) and NaI detector, respectively. As predicted by Monte Carlo simulations, the measured intensity of thermal neutron inside the large sample cavity due to the external moderator was smaller than that from the smaller sample cavity. Due to its larger size, additional thermalization of neutrons will take place in the larger sample. In spite of smaller thermal neutron yield from the external moderator at the large sample location, higher yield of the prompt gamma ray was observed as compared to that from the smaller sample. This confirms the significance of neutron moderation effects in the bulk sample and can thereby affect the PGNAA geometry size. This allows larger samples in conjunction with smaller moderators in the PGNAA setup.