用基态特征γ射线计算同质异能态反应截面的公式及其应用

根据人工放射性的生长与衰变规律, 推导出一个在特殊情况下用基态特征γ射线计算同质异能态反应截面的公式, 并据此公式用活化法以⁹³Nb(n,2n)^92mNb反应截面为中子注量标准对¹²⁸Te(n,2n)^127mTe反应截面进行了测量. 由能量为(14.1±0.2)和(14.6±0.3)MeV的中子引发的¹²⁸Te(n,2n)^127mTe反应截面值分别为(737±69)和(853±82)mb. 单能中子用T(d,n)⁴He反应获得, 其能量用铌锆截面比法测定. 为减小热中子的(n, γ)反应的影响, 在辐照过程中对样品进行了包镉处理.     

大面积闪烁体探测器的中子探测效率刻度

介绍了一种利用同位素中子源(²⁴¹Am⁹Be)通过γn符合的中子飞行时间法对大面积探测器的中子探测效率进行刻度的方法,并用蒙特卡罗方法对该探测器的探测效率进行了模拟.实验与模拟结果基本一致.     

14MeV中子引起的Pb(n,x)203Hg, W(n,x)182Ta和W(n,x)183Ta的核反应截面测量

利用活化方法测量了14MeV中子引起的Pb(n,x)²⁰³Hg,W(n,x)¹⁸²Ta和W(n,x)¹⁸³Ta的反应截面.中子注量由监督反应⁹³Nb(n,2n)^92mNb给出,中子能量利用⁹⁰Zr(n,2n)^89m+gZr和⁹³Nb(n,2n)^92mNb反应的截面比来确定.     

轻丰中子核反应中子集团产生截面的理论研究

采用Boltzmann-Langevin方程研究了能量为35MeV/u的¹⁴Be, ⁸He,⁶He,¹¹Li,¹⁷B,¹¹Be,¹⁹C与¹²C靶的反应,计算了产生中子集团的截面, 发现^14,Be与¹²C靶反应产生⁴n的截面与实验值符合得很好. 通过这几个入射核与¹²C靶形成中子集团截面的对比, 发现核的晕中子越多产生中子集团的截面越大, 晕中子数相同时, 质量数越大产生中子集团的截面越大.中子集团可能主要来自晕核子.     

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。     

13.5—14.6MeV中子能区镓的同位素反应截面的测量

报道了在13.5—14.6MeV中子能区用活化法以⁹³Nb(n,2n)^92mNb反应截面为中子注量标准测得的⁶⁹Ga(n,2n)⁶⁸Ga,⁶⁹Ga(n,p)^69mZn,⁷¹Ga(n,p)^71mZn和⁷¹Ga(n,n′α)⁶⁷Cu的反应截面值.由(13.5±0.2),(14.1±0.1)和(14.6±0.2)MeV中子引起的⁶⁹Ga(n,2n)68Ga反应截面值分别为(794±31),(869±35)和(986±39)mb,⁷¹Ga(n,n′α)⁶⁷Cu反应截面值分别为(1.3±0.1),(1.7±0.1)和(2.5±0.1)mb.在中子能量为(14.1±0.1)MeV能量点,⁶⁹Ga(n,p)^69mZn反应截面值为(21.5±1.0)mb,⁷¹Ga(n,p)^71mZn反应截面值为(12.4±0.7)mb.单能中子由T(d,n)⁴He反应获得.文中还列举了尽可能收集到的数据以作比较.     

Ba(n,x)134Cs,134Ba(n,2n)133Ba,140Ce(n,2n)139Ce,142Ce(n,2n)141)Ce和23Na(n,2n)22Na反应截面的测量

用中子活化法相对于⁵⁴Fe(n,P)⁵⁴Mn反应,在13.50—14.80MeV中子能区测量了Ba(n,x)¹³⁴Cs,¹³⁴Ba(n,2n)¹³³Ba,¹⁴⁰Ce(n,2n)¹³⁹Ce,¹⁴²Ce(n,2n)¹⁴¹Ce和²³Na(n,2n)²²Na的反应截面.并将所测的结果和其他作者的结果进行了比较,中子能量是用⁹⁰Zr(n,2n)^89m+gZr反应和⁹³Nb(n,2n)^92mNb反应截面比法测定的.     

产生长寿命核的几个(n,2n)反应截面的测量

本文报道了E_n=13.5—14.8MeV中子能区几个长寿命生成核的(n,2n)反应截面的测量.测量方法是以⁹³Nb(n,2n)^92mNb反应截面为中子注量标准的相对活化法.测量的几个反应为:¹⁵¹Eu(n,2n)^150mEu、¹⁵³Eu(n,2n)¹⁵²Eu、¹⁵⁹Tb(n,2n)¹⁵⁸Tb和¹⁰⁹Ag(n,2n)^108mAg.中子能量是用铌锆截面比法测定的.本文的结果和已收集到的测量结果进行了比较.     

14MeV钕的同位素反应截面的测量

报道了在13.5—14.6MeV中子能区用活化法以⁹³Nb(n,2n)^92mNb反应截面为中子注量标准测得的¹⁵⁰Nd(n,2n)¹⁴⁹Nd,¹⁴⁸Nd(n,2n)¹⁴⁷Nd和¹⁴²Nd(n,2n)¹⁴¹Nd的反应截面值.由13.5±0.2,14.1±0.1和14.6±0.2MeV中子引起的¹⁵⁰Nd(n,2n)¹⁴⁹Nd反应截面值分别为2037±85,1737±68,1657±65mb,¹⁴⁸Nd(n,2n)¹⁴⁷Nd反应截面值分别为1394±58,1416±54,1956±76mb,^{142Nd(n,2n)141Nd反应截面值分别为1501±59,1623±62,1764±111mb.单能中子由T(d,n)4He反应获得.文中还收集了已发表的数据以作比较.}     

在轻核区核反应中的强非统计效应

本文研究²⁷Al(n,γ)和²⁸Si(n,γ)反应的非统计效应.考察的中子能量范围为热能至2MeV.所研究的非统计过程在低能区(热中子至第一共振前)包括位阱俘获、价俘获、以及两者的干涉;在高能区(共振区、研究对能量平均的截面)包括直接-半直接俘获,复弹性散射道及复非弹性散射道的俘获.计算结果表明,与中重质量核相比,这两个核的非统计效应在总的(n,γ)截面中所占的比例特别高:在热中子能量范围约为50%;在共振区区80%.本文指出,这一现象是与实验上观测到的关联系数相一致的.     

The dose comparison between the THOR and HFR epithermal neutron beams

This study is a part of the beam comparison campaign, inter-center dose comparison, between boron neutron capture therapy facilities at the Tsing Hua Open-pool Reactor and the High Flux Reactor. The clinical information exchange can improve the dosimetry uncertainty for medical physics in a mixed field. The method of paired Mg(Ar) and TE(TE) ionization chambers was used to determine the gamma-ray and neutron dose rates. Furthermore, activation foils, including gold, copper, and manganese, were employed to estimate the thermal and epithermal neutron fluxes. Measurements were performed free in air and also in a PMMA phantom. All the chambers were calibrated using a ⁶⁰Co primary standard source at the Institute of Nuclear Energy Research, Taiwan. Spectrum dependent neutron sensitivity of TE(TE) chamber is one of the important parameters to evaluate dose components. The requested neutron spectra were calculated by the Monte Carlo code MCNP. The measured thermal neutron fluxes, gamma-ray and neutron dose rates of the THOR beam in the phantom were 2.6, 2.2, and 2.1 times of the HFR beam at 2.5-cm depth, respectively. The higher thermal neutron flux and neutron and gamma-ray dose rates are due to the higher epithermal neutron beam intensity of the THOR.     

Microdosimetry for the characterization of the THOR epithermal neutron beam

The epithermal neutron beam of the Tsing Hua Open-pool Reactor (THOR) was constructed for the study of boron neutron capture therapy (BNCT). The THOR epithermal neutron beam was mainly composed of thermal neutrons, fast neutrons, and photons. For fast neutrons and photons, the absorbed dose and the relative biological effectiveness (RBE) were used to characterize radiation dose and radiation quality. The short-ranged alpha particles and lithium ions produced from ¹⁰B(n,α)⁷Li reactions in the BNCT required cellular- and micro-dosimetry characterizations. Due to the non-uniform microdistribution of boron in cells, these characterizations should depend on the source–target geometry. In this case, the geometry-dependent specific cellular dose and lineal energy could be used to describe radiation dose and radiation quality. In the present work, cellular- and micro-dosimetry were studied for the THOR epithermal neutron beam. The specific cellular dose and lineal energy were calculated for thermal neutron-induced α-particles and ⁷Li-ions with different source–target geometry and various cell sizes. Applying the linear energy dependent-biological weighting function, the geometry-dependent RBE of thermal neutron-induced heavy particles was determined. Finally, the effective RBE of the THOR epithermal neutron beam was estimated for tumors and normal tissues of specified ¹⁰B concentrations. This effective RBE should be multiplied by the total absorbed dose to determine the corresponding biological dose required in the treatment planning.     

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.     

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.     

(n,γ)反应实验研究中的中子屏蔽设计

用射线全吸收型装置(Gamma-ray Total Absorption Facility,GTAF),可以对中子俘获反应截面进行高精度测量。为了降低实验本底,实验中需要对源中子进行准直和屏蔽,还要对被样品散射的中子进行吸收以减少它们进入探测器后所形成的干扰。采用MCNP对中子的准直器、屏蔽体和中子吸收体进行了模拟设计,中子准直屏蔽体材料选用含硼聚乙烯(BC4 的质量分数为3%) 和铅。准直孔直径为13 mm,长度为500mm,经准直后样品处中子束斑坪顶直径为21 mm。中子吸收体材料选用聚乙烯和碳化硼,吸收体球壳内腔半径30 mm,聚乙烯壳层厚度60 mm,碳化硼壳层厚度10 mm,被样品散射的中子经吸收体后衰减93.7%。Neutron capture cross section can be measured by Gamma-ray Total Absorption Facility (GTAF) with high precision. To reduce the background of experiments, the neutron source must be collimated and shielded, and the neutrons scattered from the sample must be absorbed to minimise interference after they go into the detector. The shield, collimator and absorber were simulated and designed with MCNP code. Boron-ontainingpolyethylene with 3% BC4 and lead are used as the materials for the neutron collimator and shield. The diameter of the collimating aperture is 13 mm, and the length of the collimator is 500 mm. After being collimated, the diameter of neutron beam plateau at the sample position is 21 mm. The neutron absorber is made of polyethylene and BC4, and the thickness of polyethylene shell and BC4 shell are 60 and 10 mm, respectively. The simulated result shows that neutrons scattered from the sample can decay 93.7% through the neutron absorber.     

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.     

Fast-neutron dose evaluation in BNCT with Fricke gel layer detectors

Boron neutron capture therapy (BNCT) is a cancer radiotherapy that uses epithermal and thermal neutron beams. The determination of the absorbed dose in healthy tissue, separating the various dose contributions having different radiobiological effectiveness (RBE) is of great importance for therapy planning. However, a standard code of practice has not yet been established because suitable methods for dosimetry in BNCT are still in progress.A study about the characterization of the epithermal column of the LVR-15 research reactor in ?e? (CZ) has been performed, in particular concerning the fast-neutron dose. This dose is not negligible and its determination is important owing to its high RBE. Fast-neutron and photon dose distributions in a water phantom have been measured by means of Fricke gel layer dosimeters. Even if gel layer dosimetry is not yet standardized, it is presently the only method for obtaining images of each dose contribution in BNCT neutron fields.The results were compared with values measured with thermoluminescence detectors, twin ionization chambers data taken from literature and Monte Carlo simulations.     

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.     

氘-氚中子源辐照旋转靶剂量演化模拟计算

首次将蒙特卡罗自动建模系统(MCAM)、蒙特卡罗粒子输运程序(MCNP)及自主研发的活化程序BURNDOT相结合,实现了中子输运、材料活化、光子输运模拟计算的耦合。对14MeV氘-氚(D-T)中子源旋转靶室剂量演化分析,计算了氘-氚中子源辐照旋转靶室的瞬发中子、γ三维辐射剂量场分布及连续辐照8小时后缓发γ剂量变化情况,并考察了材质、栅元、主要核素对缓发γ剂量贡献的影响,得到了旋转靶的剂量时空演化规律,把计算结果与欧洲活化程序FISPACT-2007进行了对比。