D-T快中子治疗准直屏蔽体设计及中子特性模拟

设计了一个用于D-T快中子治疗的准直屏蔽体,通过D-T中子在准直屏蔽体中的MCNP模拟,计算了屏蔽体外透射中子和透射光子在水中的吸收剂量,由此评价了准直屏蔽体的屏蔽效果。利用MCNP程序,模拟了准直中子束及中子束中的γ射线在源皮距(SSD)100cm处的能谱,计算了γ射线与中子束在水中吸收剂量的比值,对准直中子束中γ射线的污染水平进行了评价。完成了准直中子束在人体组织等效水箱中输运的MCNP模拟,给出了吸收剂量深度分布、吸收剂量横向分布和吸收剂量等剂量曲线。     

14 MeV快中子照相准直屏蔽系统的设计与优化

基于强流氘氚中子源科学装置HINEG设计了一套快中子照相准直屏蔽系统。采用中子输运设计与安全评价软件系统SuperMC和ENDF/B-Ⅶ.0数据库计算了准直中子束的中子能谱及注量率、γ射线能谱及注量率、直射中子注量率与γ射线注量率比值(φ_d/φ_γ)、直射与散射中子注量率比值(φ_d/φ_s)、准直束中子注量率的不均匀度等特性参数,并采用MCNP5程序进行了对比验证。研究了准直屏蔽系统的内衬材料、尺寸等对特性参数的影响规律,并通过优化获取了最优设计方案。计算结果显示,在同等计算条件下,SuperMC计算结果与MCNP计算结果相对偏差小于1%,准直屏蔽系统的φ_d/φ_γ为50.1,φ_d/φ_s为5.7,在?30 cm视野范围内的中子注量率为4.80×10~7 cm~(-2)·s~(-1),其中直射中子注量率为4.09×10~7 cm~(-2)·s~(-1),中子注量率不均匀度为5.8%,满足快中子照相对准直束特性参数的要求。     

14 MeV快中子照相准直屏蔽系统的设计与优化

基于强流氘氚中子源科学装置HINEG设计了一套快中子照相准直屏蔽系统。采用中子输运设计与安全评价软件系统SuperMC和ENDF/B-Ⅶ.0数据库计算了准直中子束的中子能谱及注量率、γ射线能谱及注量率、直射中子注量率与γ射线注量率比值（φ_d/φ_γ）、直射与散射中子注量率比值（φ_d/φ_s）、准直束中子注量率的不均匀度等特性参数,并采用MCNP5程序进行了对比验证。研究了准直屏蔽系统的内衬材料、尺寸等对特性参数的影响规律,并通过优化获取了最优设计方案。计算结果显示,在同等计算条件下,SuperMC计算结果与MCNP计算结果相对偏差小于1%,准直屏蔽系统的φ_d/φ_γ为50.1,φ_d/φ_s为5.7,在Φ30 cm视野范围内的中子注量率为4.80×10⁷ cm^-2•s^-1,其中直射中子注量率为4.09×10⁷ cm^-2•s^-1,中子注量率不均匀度为5.8%,满足快中子照相对准直束特性参数的要求。     

T(d,n)4He反应快中子屏蔽体优化设计的蒙特卡罗模拟

在考虑T（d,n）^4He（D-T）反应快中子源能谱、角分布、面源结构及靶系统对中子的作用的基础上,利用MCNP程序模拟了（D-T）反应快中子在屏蔽材料中子的输运。通过屏蔽体外泄漏中子及γ射线的注量率、能谱及在水中的吸收剂量的分析,给出了满足T（d,n）^4He反应中子源快中子治疗屏蔽体的三种复合屏蔽方案。     

强流D-T中子发生器n-γ辐射场的MC模拟研究

根据D-T中子源260 keV氘束流能量下的中子能谱和角分布数据,建立了D-T中子源模型,在中子发生器各元件材料及实验大厅墙壁对快中子的慢化、散射和吸收的条件下,采用MCNP模拟程序对强流中子发生器n-γ辐射场进行了模拟研究,得到了中子能谱、中子和γ角分布、n-γ比等重要参数.模拟结果与有关文献实验数据的基本吻合,验证了所建立的MCNP模拟模型的可靠合理.     

D-T中子源快中子照相准直屏蔽系统设计

减少散射中子对成像质量的影响是快中子照相的关键技术之一。本文采用MCNP程序完成了D-T中子源快中子照相准直屏蔽系统的设计,该设计具有很好的中子散射屏蔽效果,能够将实验室墙壁的散射中子数量降低一个量级,大大减少了成像的对比度失真;同时,能够增加约20%的中子源强,提高了成像速度。     

小型中子源高能中子照相装置准直屏蔽系统设计

准直屏蔽系统是中子照相装置的必要设备。本文采用蒙特卡洛中子输运程序(MCNP)等软件对高能中子准直屏蔽系统进行理论设计,初步确定了其材料构成和外观尺寸,从理论上确定了装置包括成像处注量率、成像距离及相应视场等关键参数。     

加速器热中子照相装置CCD芯片屏蔽的模拟计算

建立了研究加速器中子源热中子照相装置CCD芯片屏蔽效果的蒙特卡罗模拟方法,对γ与中子吸收剂量率的模拟计算结果与实验相符.进行了基于9Be(d,n)反应的热中子照相装置屏蔽系统的优化设计,在复杂几何条件下用蒙特卡罗模拟分别计算了CCD芯片在中子、γ混合场中的吸收剂量率和快中子注量率,对CCD相机在辐射场中安全性能进行了评估.     

影响中子射线照相性能的主要因素

中子束注量率及其分布、镉比、准直比、信噪比、成像方式等是影响中子射线照相性能的主要因素,从实验和理论两方面针对这些因素进行了初步分析,并提出了应对技术措施。     

束流参数对热中子照相质量影响的模拟分析

中子束的准直比、镉比、散射中子及注量率空间分布等是影响中子射线照相性能的主要因素,本文通过蒙特卡罗模拟仿真计算来分析各因素对中子照相图像质量的影响,并提出相应的应对措施。     

Supermirror neutron guide system for neutron resonance spin echo spectrometers at a pulsed neutron source

A neutron guide system for neutron resonance spin echo spectrometers has been constructed at BL06 of the Japan Proton Accelerator Research Complex, Materials and Life Science Experimental Facility. The spectrometers consist of two types of neutron spin echo instruments, a modulated intensity by zero effort instrument (MIEZE) and a neutron resonance spin echo instrument (NRSE), to cover a wide energy range for various sample environments. A neutron beam from the moderator is deflected by supermirror neutron guides, split, and separately guided into the MIEZE and NRSE. The characteristic wavelengths of the neutron guide tube for the MIEZE and NRSE are 2.9 and 4.9 Å, respectively. The cross sections of the exit of the MIEZE and NRSE guides are 15 mm × 50 mm and 30 mm × 50 mm, respectively. The neutronics and shielding design were optimized by using the heavy ion transport code system (PHITS), and the absolute average neutron fluxes at the exits of the MIEZE and NRSE guides are estimated to be 2.7 × 10⁸ and 6.9 × 10⁸ n/cm²/s/MW, respectively. The measured fluxes of the MIEZE and NRSE neutron guides are 0.56 and 0.95 times the calculated values, respectively.     

Effects of neutron spectrum and external neutron source on neutron multiplication parameters in accelerator-driven system

The neutron multiplication parameters: neutron multiplication M, subcritical multiplication factor k_s, external source efficiency φ^*, play an important role for numerical assessment and reactor power evaluation of an accelerator-driven system (ADS). Those parameters can be evaluated by using the measured reaction rate distribution in the subcritical system. In this study, the experimental verification of this methodology is performed in various ADS cores; with high-energy (100 MeV) proton–tungsten source in hard and soft neutron spectra cores and 14 MeV D–T neutron source in soft spectrum core. The comparison between measured and calculated multiplication parameters reveals a maximum relative difference in the range of 6.6–13.7% that is attributed to the calculation nuclear libraries uncertainty and accuracy for energies higher than 20 MeV and also dependent on the reaction rate distribution position and count rates. The effects of different core neutron spectra and external neutron sources on the neutron multiplication parameters are discussed.     

基于球谱仪的中子剂量率仪设计

研制了一套基于球谱仪的中子剂量率仪。采用单个慢化球及位置灵敏计数器的探测结构,通过单次测量即可计算出中子剂量等信息。在基于电荷分配法的读出电路中,通过对电荷灵敏放大器及滤波成型电路的参数进行优化,使探测器系统的平均位置分辨达到6 mm。利用硬件寻峰等数字信号处理技术,使数据获取系统的最高计数率达到200 k·s-1,满足剂量率仪实时性的要求。     

裂变-聚变中子源

为适应聚变堆的发展和处理高放废物的需要,提出裂变-聚变中子源的概念,它是采用LiD组件放在高通量反应堆中或中国先进研究堆(CARR)重水区中,通过慢中子与6Li(n,a)反应产生2.739 MeV氚离子,它与LiD中的D发生聚变反应,产生聚变中子;随着LiD中氚的快速积累,14 MeV 中子产生的D反冲粒子流与氚发生聚变反应,增长聚变中子产额,使 14 MeV 中子注量率逐渐升高.当氚浓度接近0.5×1022时,D反冲粒子流与氚的聚变反应率的产额接近于1,聚变中子将成倍的增长,类似于连锁反应,使聚变中子产额达到饱和,即t时刻产生氚,都被用于产生聚变反应,形成裂变-聚变中子源.这时的通量非常高,必须在接近饱和前对设定的通量(如3.5×1014n/cm2·s)下逐步降低反应堆功率,如降低CARR 中子注量率,使其在设定的通量下达到饱和,适应聚变堆中子注量率的需求.论述了裂变-聚变中子源的原理,聚变中子产生率,氚的积累速率和浓度,D反冲粒子流和与氚的聚变反应速率,以及其影响因素.在均匀中子场下(即不考虑中子降抑的情况下)计算了外径180 mm、内径100 mm的LiD管道中聚变中子注量率.     

Delayed neutron precursors

Experimental data on delayed neutron precursors are compiled and evaluated. Where available, the following data are listed for each nuclide: half-life, neutron emission probability, total energy available for beta decay, neutron binding energy of the daughter nucleus, major peaks in the neutron spectrum, methods of identification and production. The literature survey ended in December 1972.     

Recent neutron sputtering results and the status of neutron sputtering

Extensive (d, t) and (d, Be) fast neutron sputtering results obtained by Batelle-Northwest are summarized for Nb and Au. The measured sputtering ratios are in the range of 10^?5 atoms/fast neutron. These results are compared to other available experimental results and to theoretical estimates. Some implications of fast neutron sputtering for future reactors are summarized.     

Monte Carlo simulation of neutron transport in intense neutron fields

Common Monte Carlo (MC) codes for neutron transport are usually applied to neutron fields of low density under the assumption that the isotopic composition of the structure materials will not be changed in neutron reactions. This assumption is no longer valid in intense neutron fields, where an appreciable number of nuclei of the structural material may get transformed into other isotopes. Due to this time-dependent isotope- transformation; however, neutrons injected later into the system will see a different cross-section profile than the neutrons injected earlier.     

Inelastic neutron scattering and prompt fission neutron spectra for Np

The secondary neutron spectra (inelastic, elastic, fission) for ²³⁷Np were measured by the neutron time of flight spectrometer of the IPPE at the incident energy range 1–2.5 MeV. The solid tritium target was used as a neutron source. The neptunium oxide (189 g) packed in the low mass stainless steel container was used as a scattering sample. The neutron background due to scattering on the target environment and tritium into the target backing was measured and was calculated with the appropriate model of the neutron source. The data were corrected for neutron background, the scattering on the oxygen and iron nuclei, and the effect of the finite sample size. The fission neutron spectra were measured, evaluated and subtracted from the emission neutron spectra to estimate inelastic neutron spectra and cross-sections. The experimental results were compared with ENDF/B-VI, BROND-2, JENDL-3 neutron data libraries.     

能量选择法中子照相技术

能量选择中子照相是近年来新兴的中子照相技术,具有广泛的潜在应用前景。介绍了能量选择中子照相技术的基本原理,比较分析了机械速度选择器法、双晶单色器法、飞行时间法以及超镜和晶体过滤器结合法4种获取单色中子束的方法。介绍了能量选择中子照相技术在增强图像对比度、应力测量、织构分析以及相变过程研究等方面体现的优异特性和应用潜力。