56,54Fe(n, α)53,51Cr, 56,54Fe(n,p)56,54Mn反应截面计算及普适性模型参数优化研究

基于中子与Fe发生反应产生“氢泡”,及“氦泡”对新型核能利用系统壁材料的影响,开展了中子诱发56,54Fe(n, α)53,51Cr, 56,54Fe(n, p)56,54Mn反应截面计算研究工作。本工作根据现有的56,54Fe(n, α)53,51Cr, 56,54Fe(n, p)56,54Mn反应实验数据、评价数据,对TALYS程序调用的物理模型(包括能级密度、对修正、核温度、光学势参数等)进行参数调校,得到了一组普适性强的模型参数。基于调校的参数,本工作采用核反应程序TALYS理论计算56,54Fe(n, α)53,51Cr, 56,54Fe(n, p)56,54Mn反应的截面、能量微分截面以及双微分截面,全部数据都能与实验数据、评价数据符合较好,且适用于较宽的中子能量区间0~175 MeV。本工作提出了56,54Fe(n, α)53,51Cr, 56,54Fe(n, p)56,54Mn反应的普适性模型参数,促进了核反应理论的发展,为核数据的评价奠定了基础。     

Study of an unfolding algorithm for D-T neutron energy spectra measurement using a recoil proton method

A proton recoil method for measuring D-T neutron energy spectra using polyethylene film and a Si(Au) surface barrier detector is presented. An iteration algorithm for unfolding the recoil proton energy spectrum to the neutron energy spectrum is investigated. The response matrices R of the polyethylene film at angles of 0° and 45° were obtained by simulating the recoil proton energy spectra from mono-energetic neutrons using the MCNPX code. With an assumed D-T neutron spectrum, the recoil proton spectra from the polyethylene film at angles of 0° and 45° were also simulated using the MCNPX code. Based on the response matrices R and the simulated recoil proton spectra at 0° and 45°, the respective unfolded neutron spectra were obtained using the iteration algorithm, and compared with the assumed neutron spectrum. The results show that the iteration algorithm method can be applied to unfold the recoil proton energy spectrum to the neutron energy spectrum for D-T neutron energy spectra measurement using the recoil proton method.     

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.     

120~430 MeV/u 碳离子的微剂量学特性研究(英文)

单粒子微剂量谱在放射治疗中是一个极其重要的参数,它可以用来评估辐射场的生物学效应。利用蒙特卡洛程序FLUKA模拟计算了由碳离子产生的混合辐射场能量沉积的微观模式。从已公开发表的文献中选取了实验测量300 MeV/u 碳离子的线能能谱,并与相同物理条件下模拟计算得到的线能能谱相比较,结果吻合得很好。此外,还计算了120~430 MeV/u 的碳离子的剂量平均线能能谱、频率平均线能和剂量平均线能。所得到的频率平均线能值为185~ 28.3 keV/m而剂量平均线能值则为272~ 64.1 keV/m。本文的结果对于制定碳离子放射治疗的治疗计划有着重要的意义.Microdosimetric single event spectrum is a significant parameter in radiotherapy, which can be used to evaluate the radiation biological effect. In this paper, microscopic patterns of energy deposition are simulated with Monte Carlo code FLUKA at mixed radiation fields during carbon ions therapy. The results are compared with experimental measured results at 300 MeV/u carbon ion and good agreement has been found. Meanwhile, dose-weighted lineal energy spectra, frequency averaged lineal energy values and dose averaged lineal energy values of carbon ion with energy from 120 to 430 MeV/u were calculated,too. The frequency averaged lineal energy values are from 185 to 28.3 keV/m while the dose averaged lineal energy values are from 272 to 64.1 keV/m. These studies are useful for treatment plan in carbon ion radiotherapy.     

用于GPPD谱仪的大面积闪烁体中子探测器性能测试

中国散裂中子源（Chinese Spallation Neutron Source,CSNS）的建成为中子散射技术的发展和应用奠定了基础。通用粉末衍射谱仪（general purpose powder diffractometer,GPPD）是CSNS一期建设的三条谱仪之一。研制了用于GPPD的大面积闪烁体探测器阵列,在CSNS的BL20中子束线上开展了探测器关键性能参数的实验测试。测试结果显示,对波长为1.4 Å和2.8 Å的中子,其探测效率分别为（38.5±1.7）%和（56.1±1.1）%,模拟与实验相结合,给出对2 Å的中子,其探测效率约为（46.1±2.0）%,有效面积内的探测效率不均匀度约为19.6%,经优化后可以降低到14.9%;探测器位置分辨可以达到4.0 mm×4.0 mm,最高计数率约79 kHz。探测器单元的整体性能指标能满足GPPD的物理需求。     

Calculation of the wide-angle neutron spectra from the ~9Be(d,xn) reaction in a thick beryllium target

The multi-layer computing model is developed to calculate wide-angle neutron spectra, in the range from0° to 180° with a 5° step, produced by bombarding a thick beryllium target with deuterons. The double-differential cross-sections(DDCSs) for the ~9 Be(d, xn) reaction are calculated using the TALYS-1.8 code. They are in agreement with the experimental data, and are much better than the PHITS-JQMD/GEM results at 15°, 30°, 45° and 60° neutron emission angles for deuteron energy of 10.0 MeV. In the TALYS-1.8 code, neutron contributions from direct reactions(break-up, stripping and knock-out reactions) are controlled by adjustable parameters, which describe the basic characteristics of typical direct reactions and control the relative intensity and the position of the ridgy hillock at the tail of DDCSs. It is found that the typical calculated wide-angle neutron spectra for different neutron emission angles and neutron angular distributions agree quite well with the experimental data for 13.5 MeV deuterons. The multi-layer computing model can reproduce the experimental data reasonably well by optimizing the adjustable parameters in the TALYS-1.8 code. Given the good agreement with the experimental data, the multi-layer computing model could provide better predictions of wide-angle neutron energy spectra, neutron angular distributions and neutron yields for the ~9 Be(d, xn) reaction neutron source.     

用于中子位置灵敏探测器的SiPM性能研究

为满足中国散裂中子源(China Spallation Neutron Source,CSNS)工程材料衍射谱仪的探测器需求,CSNS探测器组设计并研制了基于硅光电倍增管(Silicon Photomultiplier,SiPM)读出的闪烁体探测器。本工作针对该探测器,选取了Sensl MicroFJ-30035-TSV和Hamamatsu S13363-3050NE-16两种型号的SiPM,开展了其击穿电压、增益、温度特性、暗计数率等关键性能的测试。测试结果显示,两者单光子分辨能力,增益、暗计数率等性能均可满足当前闪烁体探测器需求,相同过偏压下,前者增益高于后者,且Hamamatsu SiPM增益对温度更敏感。测试了两SiPM的温度补偿系数分别为22.0 mV/℃(Sens1)和53.6 mV/℃(Hamamatsu),为后续SiPM温度补偿电路设计奠定了基础。利用研制的探测器工程样机,在CSNS BL09下测试了两种SiPM读出的探测器对2.8 ?中子探测效率分别为76%和68%,为目标探测器及同类型探测器的SiPM选型提供了参考。     

中子诱发232Th裂变初始碎片质量及动能分布Monte-Carlo研究

随着第四代反应堆以及先进核能利用系统的发展,对中子核数据提出了高精度、多核素、宽能区的新要求.目前,中国核数据评价库(CENDL库)中相关核裂变的数据较缺失,不足以满足当前核能发展的需求.因此,建立面向中子核数据需求的可靠计算方法和工具变得极为重要.本文基于Monte-Carlo方法建立了裂变碎片质量动能计算模型,研究了中低能中子诱发²³²Th(n,f)反应发射中子前裂变碎片的分布特性.对于裂变碎片质量分布,本模型计算结果与实验值最大偏差约1%,与GEF, TALYS程序计算结果 (与实验值最大偏差约2%)相比具有一定优势.对于发射中子前裂变碎片动能分布,本模型计算结果与实验数据一致.结果表明,所发展的计算模型能够较好地预测²³²Th(n,f)反应发射中子前裂变碎片数据,为中子诱发锕系核裂变反应计算提供一种新思路.     

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。     

分离扇回旋加速器的等时场在线优化

针对兰州重离子加速器冷却储存环（HIRFL-CSR）的分离扇回旋加速器（SSC）的等时场优化问题,依托HIRFL-CSR的主数据库系统Oracle,设计了系统控制界面,该界面不仅能够读取和给定电源电流值,而且能够实时更新束流相位信息。根据束流相位的实时数据,利用Kb-Kr方法以及Lagrange插值算法建立的等时场,实现了对SSC束流等时场的在线优化。优化后的等时场可以有效控制束流相位的漂移,提高束流品质。