基于3.5 MeV射频四极质子加速器硼中子俘获治疗装置的束流整形体设计

在硼中子俘获治疗(BNCT)装置中,束流整形体(BSA)的作用是将中子源产生的快中子束流慢化至超热中子能区(0.5 eVE10 keV),并尽可能减弱快中子、热中子和γ射线的成分,同时保证中子的方向性,其设计与优化是BNCT装置设计工作的核心内容之一.本文采用3.5 MeV,10 mA的质子束轰击锂靶,由核反应~7Li(p,n)~7Be产生的中子为源项,针对BSA的慢化体材料和结构、γ屏蔽层和热中子吸收层的厚度等参数进行蒙特卡罗模拟设计与优化.研究发现,采用Fluental和LiF两种慢化材料间隔2 cm层状堆叠的三明治BSA构型,在保证快中子剂量成分(D_f/φ_(epi)),γ剂量成分(D_γ/φ_(epi))和热中子比例φ_(th)/φ_(epi)满足IAEA-TECDOC-1223报告推荐要求的同时,在BSA出口处超热中子注量率优于单独使用Fluental和单独使用LiF的BSA设计.BSA出口处修正的Synder人头几何模型中的剂量分布计算结果显示:上述三明治构型的深度剂量分布与单独使用Fluental材料构型的结果基本相当,优于单独使用LiF构型,表明Fluental和LiF层状堆叠的三明治BSA构型是一种可行的BSA结构.     

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。     

基于加速器的多终端硼中子俘获治疗装置的束流整形组件设计

硼中子俘获治疗(Boron Neutron Capture Therapy, BNCT)是一种新型的精准放射治疗方法,束流整形组件(Beam Shaping Assembly, BSA)作为硼中子俘获治疗装置的重要组成部分,对于产生适用于BNCT的中子束至关重要。通过BSA可以将快中子慢化到适当的能量范围,并且减少其他不需要的束流成分,进而满足BNCT用于治疗的中子束要求。本文利用蒙特卡罗模拟软件MCNP,基于2.5 MeV/30 mA的质子加速器,设计出可将质子打Li靶产生的快中子束慢化到热中子能量范围(<0.5 eV)和超热中子能量范围(0.5 eV~10 keV)的多终端BSA方案。提出的热中子BSA方案使用D₂O作为慢化体,BeO作为反射体,Bi作为$\gamma $过滤体,而超热中子BSA方案使用MgF₂作为慢化体,Pb作为反射体,6LiF作为热中子过滤体。热/超热中子方案在BSA出口处的束流参数,均满足国际原子能机构(IAEA)对BNCT治疗提出的束流指标要求。     

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。     

400~1500 MeV质子轰击铅靶和钨靶的出射中子能谱的FLUKA和Geant4模拟研究

在ADS散裂靶系统的优化设计中,蒙特卡罗方法结合可靠的散裂反应理论模型进行中子学计算具有重要的作用。本工作利用Geant4程序中的INCLXX模型、BIC模型以及BERT模型和FLUKA程序分别模拟了597 MeV和1 500 MeV质子轰击薄铅靶不同出射角度的中子双微分截面,500,1 500 MeV质子轰击厚铅靶不同出射角度的中子双微分产额,以及400,600,800,1 000和1 200 MeV质子轰击厚钨靶在反角方向（175 °）的中子双微分产额,并与实验数据进行比较。研究表明,对于薄铅靶,Geant4程序的INCLXX模型和FLUKA程序模拟结果与实验符合得更好。能量在10～40 MeV范围内,BIC模型模拟结果明显高于实验数据,而BERT模型模拟结果略微低于实验数据。对于厚铅靶,在40 MeV左右所有的模拟结果都低于实验数据。对于厚钨靶,Geant4程序的BIC模型和FLUKA程序与实验数据符合得较好,INCLXX模型在能量高于60 MeV时模拟结果低于实验数据,BERT模型与实验数据差异较大。总体来看,Geant4程序的INCLXX模型和FLUKA程序进行ADS散裂靶相关的中子学的计算是合理和可靠的。The reliable Monte Carlo simulation codes coupled with nuclear reaction models play an important role in the neutronic calculation for the design and optimization of the ADS spallation target. In this work, the double differential cross sections at different angles produced from a thin lead target bombarded with 597 and 1 500 MeV protons, the neutron energy spectra at different angles produced from a thick lead target bombarded with 500 and 1 500 MeV protons, and the neutron energy spectra in the backward direction(175°) produced from a thick tungsten target bombarded with 400, 600, 800, 1 000 and 1 200 MeV protons are calculated with the Geant4 code coupled INCLXX, BIC and BERT models and the FLUKA code. The calculations are compared with the available experimental data. The results show that, for the thin lead target, the calculations with the Geant4 coupled INCLXX model and FLUKA code well reproduce the experimental results. In a energy range from 10 to 40 MeV, BIC model obviously overestimates the experimental results, and BERT model slightly underestimates the experimental results. For the thick lead target, all of the calculations underestimate the experimental results around 40MeV. For the thick tungsten target, the Geant4 coupled BIC model and FLUKA code well reproduce the experimental results. INCLXX model underestimates the experimental results above 60 MeV. BERT model bad reproduces the experimental results. Overall, the neutronic calculations with the Geant4 code coupled INCLXX model and FLUKA code for the ADS spallation target is reasonable and reliable.     

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建立了6Li D转换器中14Me V中子源强的计算模型,对转换器不带辐照样品和分别带2、3、4个辐照样品时的中子源强进行了计算,对转换器产生的中子和来自于堆芯的中子在样品内的能谱和中子注量率进行了计算。结果表明,辐照管内充水和氦气时,辐照样品内由转换器产生的能量大于13Me V的中子分别占能量在1Me V以上中子的25.7%、24.6%,辐照样品内由堆芯产生的能量大于13Me V的中子仅占能量在1Me V以上中子的10·5左右,样品内14Me V中子源强分别可达4.31×1013nT·s·1、3.34×1013 nT·s·1;中子注量率分别可达2.66×1010nT·cm·2·s·1、3.53×1010nT·cm·2·s·1。     

Developing a moderating block for the NG-400 neutron generator

The results from developing and determining the optimal characteristics of a thermal neutron source using the NG-400 neutron generator are presented. The primary aims of this study were to achieve the maximum conversion of fast neutrons (14 MeV) into thermal neutrons and to investigate the current density distribution of fast and thermal neutrons inside the moderator cavity to determine the zone with maximum density of the flux of thermal neutrons.     

A pulsed neutron source on the proton beam for the Moscow Meson Factory

A new pulsed neutron source based on a beam-blanking device has been under construction and improvement at the Moscow Meson Factory of the Institute for Nuclear Research of the Russian Academy of Sciences. Neutrons are generated in the course of the spallation process in a water-cooled tungsten target by a proton beam with an energy of 209 MeV. After water moderator (3 cm), neutrons are guided in three horizontal and one vertical channels with a length varying from 4 to 50 m. The standard duration of the proton pulse from the accelerator is 60 μs. At present, the average proton current is as high as 150 μA for a repetition rate of 50 Hz. The neutron fluence in the target is equal to 0.9 × 10¹⁵ neutrons/s, which corresponds to the requirements imposed on the intensity of modern pulsed neutron sources.     

Monte Carlo simulations of multiplexed electronic grade CVD diamond for neutron detection

A novel and evolutionary multiplexing technique is introduced in this work where electronic grade single crystal chemical vapor deposition (CVD) diamond plates are multiplexed together in both a series and parallel configuration, sending electronic signals from each diamond plate to a single electronic acquisition system. Modeling of this novel multiplexing technique consisted of MCNPX simulations and significant post processing. The model developed allowed for the characterization of charge collection efficiency corrections to the location of charge creation to determine the effect of increasing detection medium size with respect to charge collection direction on the measured pulse height spectrum. This work was conducted to show that this technique is theoretically capable of replacing a single crystal diamond plate of similar size for use in neutron detection without the immediate need of advancing CVD diamond growth technologies. Further, this work indicates the expected pulse height evolution from a singular large single crystal diamond if such a crystal is produced in the future. The results of this work indicate that a 14.1 MeV neutron induced energy pulse of 8.4 MeV (due to the ¹²C(n,α_o)⁹Be reaction) in the pulse height spectrum has its energy resolution broadened by a factor of two to a total value of 0.225 percent for a multiplexed array with a thickness from 0.05 to 1 cm and an intrinsic detection efficiency of 25.4 percent for a 1 cm thick diamond crystal. It is also qualitatively discussed that the number of secondary neutron interactions with the diamond detector array may be about 5 percent. The results of this work indicate the capability of multiplexing diamond plates together for spectroscopic neutron detection with a combined intrinsic detection efficiency and energy resolution greater than any other diamond-based neutron detection system reported to date.     

Ground-state, β and K = 11/2? γ bands in 163, 165Er

Neutron production by stopping 55 MeV deuterons in thick carbon and heavy-water targets has been measured by the activation method. The geometry was close to the one defined for the SPIRAL2 uranium-carbide target in the initial phase. A comparative method for obtaining the neutron flux has been used and is presented in detail. The neutron flux generated by 55 MeV deuterons on carbon is 2.3 times the flux at the deuteron energy of 40 MeV. The flux further increases by a factor 1.4 when using a heavy-water target. These results are discussed in the context of an energy upgrade of the SPIRAL2 driver accelerator.     

用于BNCT-SPECT的帕维亚TRIGA MARK II反应堆热柱优化设计

针对坐落于意大利帕维亚大学的TRIGA Mark II反应堆热柱结构进行优化设计,从而满足面向硼中子俘获治疗(BNCT)的单光子发射计算机断层成像(SPECT)研究要求。为提高计算效率并减小统计误差,对比分析使用SSW/SSR方法与直接使用反应堆为源项时热柱内照射位置处中子能谱,其结果基本一致,从而验证了SSW/SSR方法的可靠性。为在该反应堆开展BNCT中SPECT实验,热柱中子束需准直为笔形束。对比分析四种热柱优化方案下束流口处及探测器处热中子和光子通量:40 cm长石墨(射束口5 cm3 cm);0.5 cm厚硼包裹40 cm长石墨(射束口5 cm3 cm);30 cm长天然锂聚乙烯(射束口直径4 cm);30 cm长天然锂聚乙烯(20 cm长射束口直径5 cm,5 cm长射束口直径4 cm,5 cm长射束口直径2 cm)。结果显示,射束口处热中子通量分别为1.05108,2.52107,6.08107和5.10 107 #/(cm2s)。综合考虑中子准直效果及光子污染,方案三具有最优性能。为后续进行BNCT-SPECT理论和实验研究提供了基础,从而有效促进BNCT剂量准确评估方法的研究进程。     

Gamma全吸收型BaF2探测装置中子屏蔽体与吸收体的研究

中国原子能科学研究院已经建造完成了我国第一套全吸收型BaF₂探测装置，采用瞬发γ测量法，精确测量中子俘获反应截面。中子源是利用HI-13串列加速器产生的脉冲化质子束，通过7Li(p, n)7Be反应建立。为了有效降低周围环境材料和探测器产生的散射中子本底，约束中子束流的形状，使用MCNP程序模拟设计了屏蔽体，采用含硼聚乙烯(B₄C质量分数为5%)包裹5 cm铅的方案，以及准直器采用平行孔的方案。该设计使样品处的中子束斑平整均匀，直径约为2 cm，束斑外的中子注量降低5个数量级，γ注量降低3个数量级。同时设计了中子吸收体(外半径为10 cm，厚度为7 cm)用于吸收待测样品产生的散射中子。MCNP和GEANT4程序的模拟结果表明:选择含硼聚乙烯(10B₄C质量分数为10%)作为中子吸收体的加工材料，其中子吸收率达到了80%，并设置1 MeV的能量加和阈，能够满足在线测量中子俘获反应截面的实验要求。     

Application of 10B counter with moderator for neutron ambient dose equivalent measurement during radiation monitoring at joint institute for nuclear research

The results of an investigation into the possibility of applying a device based on a ¹⁰B neutron counter (CHM-14) with a polyethylene moderator as the dosimeter of neutron ambient dose equivalent H*(10) in radiation fields of nuclear physics installations at the Joint Institute for Nuclear Research (JINR) are presented. It is shown that the device can be used as the dosimeter of this quantity in the neutron energy range from 0.4 eV to 20 MeV with an error no larger than 30% due to the difference between the energy dependence of its response and the energy dependence of the neutron ambient dose equivalent. Applying the correction coefficients allows one to extend the energy range of neutron dose H*(10) measurement to hundreds MeV. The error due to the anisotropy of the device response does not exceed 35%.     

中子管非中子产物性质

采用两只经过标定的ST401闪烁探测器,测量了脉冲中子管的中子产额,在其中一只探测器前端增加铅板屏蔽,1 cm的铅屏蔽使探测器输出减少了18.20%,在加速器中子源上进行的类似实验表明,0.5 cm的铅使探测器的输出减少了2.90%。对两个中子源上测量的情况进行了蒙特卡罗模拟,加速器实验与模拟符合较好,脉冲中子管实验差别较大。对实验和模拟的情况进行了分析,结果表明:中子管除产生中子外,还会产生数量较多的轫致辐射X射线,这些X射线对准确测量中子管中子产额将造成不良影响。     

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。     

基于PGNAA的化学武器识别

未知化学武器弹药的定性识别在犀护社会安全方面是十分重要的,可指导化学武器的分类处理。瞬发伽马射线中子活化分析（PGNAA）技术利用分析活化产生的伽马射线能谱可以实现对物质中元素的无损,快速检测,在化学武器识别中具有独特的优势。因此,本研究基于PGNAA技术进行了化学武器弹药类型识别装置的设计,同时使用逻辑树判别方法对化学武器样品进行定性分析。首先,基于高纯锗（HPGe）探测器与Cf-252中子源,使用蒙特卡罗MCNP程序对装置结构进行设计优化,主要包括中子源容器尺寸、伽马屏蔽体厚度以及探测器相对位置等。为了最大化样品活化产生的特征伽马射线,需要提高样品位置处的热中子通量,采用聚乙烯作为慢化体,模拟结果显示聚乙烯厚度达到6 cm,宽度达到12 cm时,样品中热中子通量达到较高水平。为了降低周围材料活化噪声的干扰,选择铅作为屏蔽结构,模拟显示铅屏蔽厚度达到5 cm时,可满足屏蔽要求。同时,探测器与样品之间的距离也会影响对伽马射线的探测,最终模拟确定探测器与样品之间的距离为28 cm时,特征信号计数最高。根据优化结果搭建测量装置,使用分析纯试剂根据真实化学武器元素含量配制化学武器模拟样品,通过对5种化学武器模拟样品的测量获得伽马能谱。对能谱中的特征峰处理过程中,基于特征峰对元素进行分析,针对计数统计性较好的元素（如H,Cl,S）的特征峰,使用高斯及多项式拟合的方式对特征峰处的高能量康普顿平台进行扣除,获得特征伽马射线的全能峰信息。而对统计性较差的元素特征峰（如N元素的10.829 MeV）,采用能量区间加和法,对该能量下的全能峰至单逃逸峰之间的计数求和,进而可确定该元素在样品中的存在情况,最后利用建立的逻辑树判别方法根据元素存在信息对样品类型进行判别。实验结果表明,利用该优化的装置可以获得5种模拟样品的能谱,结合能谱分析方法可以得到化学武器模拟样品中的H,Cl,S和N等元素的存在信息,最后使用逻辑树判别方法可以对化学武器样品种类进行判别。     

Design of a graphite-moderated 241Am–Li neutron field to simulate reactor spectra

A neutron calibration field using ²⁴¹Am–Li sources and a moderator was designed to simulate the neutron fields found outside a reactor. The moderating assembly selected for the design calculation consists of a cube of graphite blocks with dimensions of 50 cm by 50 cm by 50 cm, in which the ²⁴¹Am–Li sources are placed. Monte Carlo calculations revealed the optimal depth of the source to be 15 cm. This moderated neutron source can be used to provide a test field that has a large number of intermediate energy neutrons with a small portion of MeV component.     

Neutron scattering from 208Pb

Elastic scattering of 7, 9, 11, 20 and 26 MeV neutrons from ²⁰⁸Pb has been measured with the Ohio University Tandem Van de Graaff accelerator. Standard pulsed beam time-of-flight techniques were employed. Measurements of the incident flux at 0° were used to normalize the differential cross sections. The measured cross sections were corrected for dead time, detector efficiency, flux attenuation, multiple scattering, finite geometry, neutron source anisotropy and compound elastic contribution. Relative uncertainties are estimated to be between 5%–10% and the uncertainty in the normalization is estimated to be less than 5 %. The data were used to obtain neutron optical potential parameters. A comparison with proton optical parameters is presented, and the (p, n) quasi-elastic cross section is calculated and compared with available data. Deformation parameters for the 3^? state (Q = ?2.615 MeV) and 5ā (Q = ?3.198 MeV) in ²⁰⁸Pb were obtained at incident energies of 11 and 26 MeV.     

Influence of Bonner sphere response functions above 20 MeV on unfolded neutron spectra and doses

Monte Carlo (MC) codes for neutron transport calculations such as MCNP, MCNPX, FLUKA, PHITS, and GEANT4, crucially rely on cross sections that describe the interaction of neutrons with nuclei. For neutron energies below 20 MeV, evaluated cross sections are available that are validated against experimental data. In contrast, for high energies (above 20 MeV) experimental data are scarce and, for this reason, every neutron transport code is based on theoretical nuclear models to describe interactions of neutrons with nuclei in matter. Here we report on the calculation of a complete set of response functions for a Bonner spheres spectrometer (BSS), by means of GEANT4 using the Bertini and Binary Intranuclear Cascade (INC) models for energies above 20 MeV. The recent results were compared with those calculated by MCNP/LAHET and MCNP/HADRON MC codes. It turns out that, whatever code used, the response functions were rather similar for neutron energies below 20 MeV, for all 16 detector/moderator combinations of the considered BSS system. For higher energies, however, differences of more than a factor of 2 were observed, depending on neutron energy, detector/moderator combination, MC code, and nuclear model used. These differences are discussed in terms of neutron fluence rates measured at the environmental research station (UFS), “Schneefernerhaus”, (Zugspitze mountain, Germany, 2650 m a.s.l.) for energies below 0.4 eV (thermal neutrons), between 0.4 eV and 100 keV (epithermal neutrons), between 100 keV and 20 MeV (evaporation neutrons), and above 20 MeV (cascade neutrons). In terms of total neutron fluence rates, relative differences of up to 4% were obtained when compared to the standard MCNP/LAHET results, while in terms of total ambient dose equivalent, relative differences of up to 8% were obtained. Both the GEANT4 Binary INC and Bertini INC gave somewhat larger fluence and dose rates in the epithermal region. Most relevant for dose, however, those response functions calculated with the GEANT4 Bertini INC model provided about 18% less neutrons in the cascade region, which led to a roughly 13% smaller contribution of these neutrons to ambient dose equivalent. It is concluded that doses from secondary neutrons from cosmic radiation as deduced from BSS measurements are uncertain by about 10%, simply because of some differences in nuclear models used by various neutron transport codes.     

Spectra of solar neutrons with energies of ~10–1000 MeV in the PAMELA experiment in the flare events of 2006–2015

The first results from measuring the spectra of solar neutrons with energies of ~10–1000 MeV in the solar flares of 2006–2015 observed by the PAMELA international space experiment are presented. The PAMELA neutron detector with ³He counters and a moderator with an area of 0.18 m² allows us to estimate the flux of solar neutrons during solar flares. Solar neutrons with energies of ~10–1000 MeV likely occurred in 21 out of the 24 analyzed flares of 2006–2015.