基于加速器的硼中子俘获治疗装置束流整形体的设计及其临床参数研究

在硼中子俘获治疗(BNCT)中,束流整形体是BNCT装置产生高品质中子束的关键部件之一,其设计至关重要。本文基于2.5 MeV质子打锂靶产生中子的过程,对加速器驱动的BNCT中子源的束流整形体进行了可行性方案设计,研究了慢化体厚度差异对出口束流品质、头部模型中的剂量分布和临床参数等方面的影响。研究表明,可行性方案设计在30 mA质子束流驱动下,可达到IAEA对束流品质的要求;在本文3种慢化体厚度设计下,随着慢化体厚度的增加,出口超热中子束流强度减小,快中子份额减小,进一步导致优势深度变浅,正常组织最大剂量率减小,治疗时间变长。     

基于加速器的硼中子俘获治疗装置束流整形体的设计及其临床参数研究

在硼中子俘获治疗（BNCT）中,束流整形体是BNCT装置产生高品质中子束的关键部件之一,其设计至关重要。本文基于25 MeV质子打锂靶产生中子的过程,对加速器驱动的BNCT中子源的束流整形体进行了可行性方案设计,研究了慢化体厚度差异对出口束流品质、头部模型中的剂量分布和临床参数等方面的影响。研究表明,可行性方案设计在30 mA质子束流驱动下,可达到IAEA对束流品质的要求;在本文3种慢化体厚度设计下,随着慢化体厚度的增加,出口超热中子束流强度减小,快中子份额减小,进一步导致优势深度变浅,正常组织最大剂量率减小,治疗时间变长。     

BNCT人头体模内剂量分布计算

用修正的Synder人头体模几何模型和ICRU-46中的材料数据，用MCNP-4B程序对0．0253ev、1kev、2keV、10keV、100keV、1MeV单能中子束，0．2、0．5、1、2、5、10MeV单能光子束，以及与当前硼中子俘获治疗(BNCT)临床中使用的超热中子相似的超热中子束，计算了在人头体模中的剂量分布，计算结果与有关文献报道的结果一致，初步校验了我们正在编制的BNCT治疗计划软件。     

高额超热中子束流治疗孔道低浓化BNCT堆初步设计方案

根据硼中子俘获治疗(BNCT)中子源的要求,在高浓铀为燃料的微型反应堆(MNSR)的基础上,以富集度19.5%的UO2为燃料,将其堆芯低浓化并且添加水平超热中子束流治疗孔道,开展超热中子束流BNCT堆堆芯低浓化初步设计。计算BNCT堆的超热中子注量率、单位超热中子注量的快中子剂量率、单位超热中子注量的γ光子剂量率、超热中子注量与热中子的注量之比、中子束流能谱等关键参数。结果表明,该设计可以得到优良的超热中子束流。     

BNCT用质子直线加速器参数分析

正中子俘获治疗法是采用热中子束和亲肿瘤药物相结合的一种二元肿瘤治疗方法,其中亲肿瘤药物研究较多的是掺硼(~(10)B)的药物,其性能基本达到临床试治的要求,称为BNCT。对于用于BNCT治疗的中子源,IAEA所要求的射束强度为:中子为超     

基于14MeV质子回旋加速器的BNCT中子源实验终端设计

随着加速器技术的发展,基于加速器的硼中子俘获治疗装置越来越受到国内外关注。为了研究基于能量为14 MeV、流强为80μA的回旋质子加速器获得硼中子俘获治疗（Boron Neutron Capture Therapy,BNCT）中子源的可能性,利用Geant4软件对中子产生靶以及束流整形组件进行了优化设计,旨在获得理想的超热中子束实验终端。由于加速器的流强较低,增设了天然铀作为中子倍增器以提高中子注量。经过对铍靶、天然铀增殖层、AlF₃和TiF₃复合慢化体、热中子吸收层和γ屏蔽层等进行优化设计,在束流出口处能够获得超热中子占比高达95.6%,注量率可达6.26×10⁷n·cm^-2·s^-1的中子源终端。该方案可初步用于加速器BNCT中子源实验终端的技术验证。     

MCNP程序计算中国人特性头模BNCT剂量分布

对国际上普遍采用的Synder人体头部模型进行修正，建立了有中国人解剖模型特征的几何模型和用于BNCT的人体头部MCNP计算模型，建立了一套用于BNCT治疗计划的软件。在软件模型的基础上分别计算了不同能量的中子进行治疗时的肿瘤及正常组织剂量分布，并对计算结果进行了分析。     

基于重复结构卡的BNCT人体头部精细模型剂量分布计算

针对目前BNCT治疗脑胶质瘤所采用的头部模型的粗略性,用MCNP程序的Universe和Fill重复结构卡对人体头部的精细模型进行描述,分别使用热中子、超热中子和快中子对大脑在有、无载硼剂时的剂量分布进行了计算,得到了剂量率随深度分布的曲线.计算结果与有关文献的报道结果相吻合,本文所构建的人体头部精细模型是正确的.     

MCNP程序计算中国人特性头模BNCT剂量分布

本文对国际上普遍采用的Synder人体头部模型进行修正,建立了有中国人解剖模型特征的几何模型和用于BNCT的人体头部mcnp计算模型,建立了一套用于BNCT治疗计划的软件.在软件模型的基础上分别计算了不同能量的中子进行治疗时的肿瘤及正常组织剂量分布,并对计算结果进行了分析.     

241Am-Be中子源紧凑型屏蔽装置及辐照器设计

²⁴¹Am-Be中子源被广泛用于实验研究,为保护实验人员免受中子及γ射线照射,需要设计适当的屏蔽。利用蒙特卡罗方法计算中子透射不同材料后的能谱分布与剂量,优选各层屏蔽材料种类与厚度,设计一套²⁴¹Am-Be中子源紧凑型屏蔽装置。装置由内而外采用钨+聚乙烯+含硼聚乙烯+不锈钢进行防护,外表面周围剂量当量率H^*(10)低于10μSv/h,满足辐射防护要求。同时对装置内部热中子、超热中子和快中子注量分布进行研究,确定装置快中子和热中子输出通道最佳位置。在辐照装置同时开放快中子和热中子通道进行实验测试时,需要设置距离大于130 cm的控制区,以保障操作人员安全。     

IHNI-1中子束快中子注量率测量

为检验和确定用于硼中子俘获治疗(BNCT)的医院中子照射器(IHNI-1)的快中子污染源项,设计了用于快中子注量率测量的包硼~(235)U裂变电离室。利用MCNP程序对电离室的注量响应进行优化设计,计算包裹不同厚度硼壳时电离室的注量响应曲线,最终选择35mm厚B4C壳作为低能中子屏蔽层。利用该电离室测量IHNI-1热中子和超热中子束的快中子注量率,并与模拟计算值比较。结果显示,实测的中子束比模拟计算结果具有更多的快中子成分,低于国际原子能机构(IAEA)推荐的目标值。     

Characteristics comparison between a cyclotron-based neutron source and KUR-HWNIF for boron neutron capture therapy

At Kyoto University Research Reactor Institute (KURRI), 275 clinical trials of boron neutron capture therapy (BNCT) have been performed as of March 2006, and the effectiveness of BNCT has been revealed. In order to further develop BNCT, it is desirable to supply accelerator-based epithermal-neutron sources that can be installed near the hospital. We proposed the method of filtering and moderating fast neutrons, which are emitted from the reaction between a beryllium target and 30-MeV protons accelerated by a cyclotron accelerator, using an optimum moderator system composed of iron, lead, aluminum and calcium fluoride. At present, an epithermal-neutron source is under construction from June 2008. This system consists of a cyclotron accelerator, beam transport system, neutron-yielding target, filter, moderator and irradiation bed.In this article, an overview of this system and the properties of the treatment neutron beam optimized by the MCNPX Monte Carlo neutron transport code are presented. The distribution of biological effect weighted dose in a head phantom compared with that of Kyoto University Research Reactor (KUR) is shown. It is confirmed that for the accelerator, the biological effect weighted dose for a deeply situated tumor in the phantom is 18% larger than that for KUR, when the limit dose of the normal brain is 10 Gy-eq. The therapeutic time of the cyclotron-based neutron sources are nearly one-quarter of that of KUR. The cyclotron-based epithermal-neutron source is a promising alternative to reactor-based neutron sources for treatments by BNCT.     

Monte Carlo optimization of an epithermal neutron flux monitor for BNCT

An epithermal neutron (0.5 eV < E_n < 10 keV) flux monitor developed for boron neutron capture therapy (BNCT) was optimized by Monte Carlo simulations. Based on this optimization study, the optimization results for each component of the epithermal neutron flux monitor were obtained. The simulation results indicated that the epithermal neutron flux monitor with optimal configuration was more efficiently applicable to precisely measure the epithermal neutron fluxes of BNCT neutron sources.     

Radiation Field for Fission Neutron Enhanced Boron Neutron Capture Therapy Employing Fast Neutron Source Reactor “YAYOI”

Boron Neutron Capture Therapy (BNCT) of a localized tumor needs a sufficient thermal neutron flux at the tumor. A surgical operation including ennucleation of the main part of tumor is required for the case of thermal neutron beam from a thermal reactor because of the rapid decrease of the neutron flux in the tissue. Intermediate neutrons with little fast neutron component are only produced by a specifically designed reactor which awaits to be build.

In the present paper, a positive use of fast neutron beams in addition to BNCT is proposed for treatment of some kind of localized tumors employing a fission fast neutrons from a fast neutron source reactor “YAYOI” of University of Tokyo which is licenced as such. Dose distributions in a water phantom located at a proposed position for two collimator cases were measured and its availability was confirmed as a possible port for therapy.     

MCNP程序几何描述能力扩展及应用测试

为使MCNP程序能模拟数百万规模的反应堆“pin-by-pin”问题和医学体素模型，本文对MCNP程序进行了改进，使几何块、几何面数量可扩展。改进后的程序对硼中子俘获治疗（BNCT）的人体大脑进行几何建模，栅元数量达百万量级；计算了大脑的中子、光子吸收剂量率随深度的变化，为大脑BNCT提供理论支持。此外，对百万规模的“Like n But”重复结构模型进行了串、并行测试，验证了几何规模扩展后程序计算的正确性。     

Influences of Neutron Source Spectrum and Thermal Neutron Scattering Law Data on the MCNPX Simulation of a Cyclotron-Based Neutron Field for Boron Neutron Capture Therapy

In our previous study, the simulation of a cyclotron-based neutron field for boron neutron capture therapy (BNCT) using a (p,n) spallation source with the MCNPX code was validated through measurements of the neutron energy spectrum behind the moderator assembly and the thermal neutron distribution in an acrylic phantom using reaction rates of ¹⁹⁸Au. These validations showed that the simulation generally well reproduced the measurements. However, some discrepancies between the measurements and the calculation remained for clinical trials. In this paper, we investigated the influences of neutron source spectrum and thermal neutron scattering law data in the simulation to resolve those discrepancies. We also compared measured and calculated neutron doses behind the moderator assembly with results obtained using a tissue equivalent proportional counter. We clarified that the neutron source spectrum calculated using the LA150 data led to the overestimation of high-energy neutrons in a phantom, but this overestimation did not significantly affect the neutron dose distribution in a phantom, because a dominant part of the absorbed dose is due to neutrons of energies below 1MeV. The study of the influence of neutron scattering law data in a phantom also indicated that the use of selected S(α,β) data led to an improvement in the simulation of thermal neutron behavior.     

On the Cf primary and secondary gamma rays and epithermal neutron flux for BNCT

Monte Carlo simulation has been used to calculate the different components of neutrons and secondary gamma rays originated by ²⁵²Cf fission and also the primary gamma rays emitted directly by the ²⁵²Cf source at the exit face of a compact system designed for the BNCT. The system consists of a ²⁵²Cf source and a moderator/reflector/filter assembly. To study the material properties and configuration possibilities, the MCNP code has been used. The moderator/reflector/filter arrangement is optimised to moderate neutrons to epithermal energy and, as far as possible, to get rid of fast and thermal neutrons and photons from the therapeutic beam. To reduce the total gamma contamination and to have a sufficiently high epithermal neutron flux we have used different photon filters of different thickness. Our analysis showed that the use of an appropriate filter leads to a gamma ray flux reduction without affecting the epithermal neutron beam quality at the exit face of the system.     

医院中子照射器中子束流出口处热中子注量率的测量

医院中子照射器是基于微型反应堆而设计的专门用于硼中子俘获治疗（BNCT）的核反应堆装置,其额定功率为30 kW。在堆芯相对两侧分别设有一条热中子束流和超热中子束流用于病人照射,在热中子束流内引出一条实验用热中子束流,用于瞬发γ法测量病人血硼浓度。本工作利用²³⁵U裂变靶和白云母探测片测量了热、超热和实验用热中子束流出口处的热中子绝对注量率。结果显示,在30 kW额定功率运行时,热、超热和实验用热中子束流出口处的热中子注量率分别为1.67×10⁹、2.44×10⁷和3.03×10⁶ cm^-2•s^-1。以上结果达到了BNCT设计要求,并能满足瞬发γ测量血硼浓度的要求。