硼中子俘获治疗头颈部肿瘤临床试验进展

硼中子俘获治疗（boron neutron capture therapy,BNCT）是结合靶向治疗和重离子治疗的先进二元放疗技术。其原理是利用含有¹⁰B同位素的硼药在肿瘤细胞中靶向聚集,随后中子束流外部照射肿瘤部位,发生¹⁰B（n,α）⁷Li核反应,释放出杀伤范围为一个细胞大小（5~9 μm）的高传能线密度α粒子和⁷Li粒子杀死肿瘤细胞。BNCT具有精准的肿瘤靶向性,对正常组织损伤小,分割次数（1~3次）少于传统放疗（30次）等优点。BNCT使用的中子由反应堆或加速器产生,临床使用的硼药包括BPA和BSH两种。本文介绍国内外开展的头颈部肿瘤BNCT临床试验及取得的重要进展。BNCT对于头颈部肿瘤治疗具有良好疗效。随着加速器中子源的推广应用和新型硼药的研发,BNCT将会在临床放射治疗领域发挥更大的作用。     

用于硼中子俘获治疗的超热中子束理论设计

目的 设计用于硼中子俘获治疗(BNCT)的超热中子束理论方案。方法 基于清华大学试验核反应堆,以其1号孔道为材料布放孔道,设计了由慢化材料、热中子吸收材料、γ屏蔽材料组成,但材料布放位置具有差异的5种理论方案;利用蒙特卡罗(MC)模拟方法,分别计算5种方案束出口处的中子注量率、剂量率及γ剂量率值,通过与BNCT技术指标对比,从5种方案中选择一种合适的方案。结果 得到了一个符合BNCT各项技术指标的超热中子束理论方案,其慢化材料厚度为53.5 cm、热中子吸收材料厚度为2 mm、γ屏蔽材料厚度为9 cm。结论 本研究给出的超热中子束理论方案为基于反应堆实现BNCT提供一定的理论参考。     

肿瘤硼中子俘获治疗的理论基础与近期研究进展

硼中子俘获治疗(BNCT)是一种新型肿瘤精准治疗方法 ,通过肿瘤细胞内的10B俘获热中子发生核裂变反应产生a粒子和反冲7Li核选择性地杀死肿瘤细胞.将足量的10B选择性递送到肿瘤细胞内部是BNCT成功的关键.本文简要介绍了BNCT治疗肿瘤的理论基础,综述了BNCT所用的中子源和硼递送剂的近期研究进展,简述了BNCT临床...     

宽能中子辐射个人防护材料研究

目的 研究可用于制作宽能中子辐射个人防护用品的柔性屏蔽材料。方法 依据理论计算结果配比加工实验用材料样品,利用²⁵²Cf裂变中子和加速器单能中子开展中子屏蔽效率实验。结果 给出了1.5 cm厚含10％碳化硼(B₄C)的乙丙橡胶板对144 keV等6个单能中子和²⁵²Cf裂变中子的屏蔽效率,其中对²⁵²Cf裂变中子的屏蔽效率为31.02%,对144 keV中子的屏蔽效率可达到76.9％。结论 以乙丙橡胶为基材掺入B₄C研制的个人中子防护柔性材料对不同能量中子辐射均有一定屏蔽效果,可用于制作宽能中子个人防护用品。     

2017—2020年北京地区大气气溶胶中7 Be和210 Pb放射性活度浓度监测与分析

目的 监测与分析2017—2020年北京地区大气气溶胶中⁷Be和²¹⁰Pb放射性活度浓度变化情况,为有效防治空气污染提供科学依据。方法 利用大流量空气气溶胶采样器（SnowWhite）采集气溶胶样品1 074份,其中春季、夏季、秋季和冬季分别采集275、266、262和271份。使用低本底高纯锗伽马谱仪（ORTEC）分析气溶胶样品中⁷Be和²¹⁰Pb放射性活度浓度。结果 2017—2020年北京地区大气气溶胶中⁷Be放射性活度浓度的变化范围为0.56~14.84 mBq/m³,平均值为6.84 mBq/m³。²¹⁰Pb放射性活度浓度的变化范围为0.01~9.37 mBq/m³,平均值为3.19 mBq/m³。2017—2020年北京大气气溶胶⁷Be和²¹⁰Pb放射性活度浓度在春、夏、秋、冬四季中差异均有统计学意义（F=32.66、93.93,P<0.05）,其中⁷Be放射性活度浓度春季最高,秋季次之,夏季和冬季最低,²¹⁰Pb放射性活度浓度由高到低分别为冬季、秋季、春季、夏季。结论 2017—2020年北京地区大气气溶胶中⁷Be和²¹⁰Pb放射性活度浓度处于正常涨落水平范围内。     

硼中子俘获治疗的辐射场特点与剂量估算

BNCT(硼中子俘获治疗)基于这样一种思想:10B的载体化合物会优先选择癌细胞作为靶而后与热中子反应,进而产生高能、短射程裂变产物α粒子和Li粒子。它是一种双重的靶向治疗方法。BNCT的治疗效果依赖于两个主要因素:源于10B(n,α)7Li核反应的高LET(传能线密度)粒子的生物效应和在靶细胞及其特异区域内的硼沉积。本文总结和探讨了BNCT的发展概况、辐射场的特点以及吸收剂量的计算方法。     

硼中子俘获治疗的辐射场特点与剂量估算

BNCT（硼中子俘获治疗）基于这样一种思想：^10B的载体化合物会优先选择癌细胞作为靶而后与热中子反应，进而产生高能、短射程裂变产物α粒子和Li粒子。它是一种双重的靶向治疗方法。BNCT的治疗效果依赖于两个主要因素：源于^10B（n,α）^7Li核反应的高LET（传能线密度）粒子的生物效应和在靶细胞及其特异区域内的硼沉积。本总结和探讨了BNCT的发展概况、辐射场的特点以及吸收剂量的计算方法。     

医用电子直线加速器治疗室的中子防护屏蔽设计

目的 研究治疗室内中子来源和中子散射的实质,分析计算加速器机房治疗室及迷路内中子剂量的变化规律,进行屏蔽设计。方法 本文基于对一台运行中的15MeV加速器治疗平面上关注点的中子通量测量结果,参考NCRP 79号报告,进行屏蔽防护设计和计算。结果 医用加速器产生X射线达到一定能量时,光核反应中子是医用加速器机房中子污染的主要来源。在加速器治疗室出入口处,主要的辐射防护是通过散射进入迷路内侧入口的杂散中子及其俘获γ辐射。结论 中子污染与治疗作用无关,但却给相关人员增加了额外剂量负担,一定条件下,也可能产生中子照射危害。所以,应重视医用加速器产生的中子外照射危害,对医用加速器治疗室内中子污染需进行防护设计和评价。     

硼中子俘获疗法诱导SHG44胶质瘤细胞凋亡

目的 探讨硼中子俘获疗法(BNCT)是否抑制人脑胶质瘤细胞SHG44增殖及其作用机制。方法 BNCT作用后,应用四甲基偶氮唑蓝比色法检测SHG44细胞的增殖抑制,采用光镜、电镜、荧光显微镜观察细胞的形态学变化。应用流式细胞仪检测SHG44细胞的凋亡率,以Western blot检测细胞表达Bcl-2、Bax蛋白的变化。结果 BNCT对SHG44细胞的增殖抑制作用呈剂量依赖性。BNCT 4和8 Gy后48 h流式细胞仪检测凋亡率分别为63.2%和88.3%。BNCT作用后,Bax蛋白表达增高,Bcl-2蛋白表达下降。结论 BNCT对胶质瘤细胞SHG44具有明显的增殖抑制及诱导凋亡作用,并使Bax蛋白表达上调、Bcl-2蛋白表达下调。     

252Cf中子后装治疗机步进精度测量及其剂量偏差研究

目的 研究²⁵²Cf中子后装治疗机步进精度以及步进偏差引起的剂量偏差。方法 利用EBT3胶片测量²⁵²Cf中子源的步进,然后通过Image J软件测量光密度值找到每步²⁵²Cf中子源的中心,计算出每步之间的间隔;利用双电离室法测量²⁵²Cf中子源位置偏差引起的剂量偏差。结果 EBT3胶片测量²⁵²Cf中子源步进精度可达0.01 mm;²⁵²Cf中子源位置偏差<3 mm时,剂量偏差<2.5%。结论 对²⁵²Cf中子源步进精度以及位置偏差的研究可为制订²⁵²Cf中子后装治疗机质量控制标准提供参考。     

Optimization of an accelerator-based epithermal neutron source for neutron capture therapy.

A modeling investigation was performed to choose moderator material and size for creating optimal epithermal neutron beams for BNCT based on a proton accelerator and the (7)Li(p,n)(7)Be reaction as a neutrons source. An optimal configuration is suggested for the beam shaping assembly made from polytetrafluoroethylene and magnesium fluorine to be placed on high current IPPE proton accelerator KG-2.5. Results of calculation were experimentally tested and are in good agreement with measurements.     

High-power electron beam tests of a liquid-lithium target and characterization study of 7Li(p,n) near-threshold neutrons for accelerator-based boron neutron capture therapy

A compact Liquid-Lithium Target (LiLiT) was built and tested with a high-power electron gun at Soreq Nuclear Research Center (SNRC). The target is intended to demonstrate liquid-lithium target capabilities to constitute an accelerator-based intense neutron source for Boron Neutron Capture Therapy (BNCT) in hospitals. The lithium target will produce neutrons through the ⁷Li(p,n)⁷Be reaction and it will overcome the major problem of removing the thermal power >5 kW generated by high-intensity proton beams, necessary for sufficient therapeutic neutron flux.In preliminary experiments liquid lithium was flown through the target loop and generated a stable jet on the concave supporting wall. Electron beam irradiation demonstrated that the liquid-lithium target can dissipate electron power densities of more than 4 kW/cm² and volumetric power density around 2 MW/cm³ at a lithium flow of ~4 m/s, while maintaining stable temperature and vacuum conditions. These power densities correspond to a narrow (σ=~2 mm) 1.91 MeV, 3 mA proton beam. A high-intensity proton beam irradiation (1.91–2.5 MeV, 2 mA) is being commissioned at the SARAF (Soreq Applied Research Accelerator Facility) superconducting linear accelerator.In order to determine the conditions of LiLiT proton irradiation for BNCT and to tailor the neutron energy spectrum, a characterization of near threshold (~1.91 MeV) ⁷Li(p,n) neutrons is in progress based on Monte-Carlo (MCNP and Geant4) simulation and on low-intensity experiments with solid LiF targets. In-phantom dosimetry measurements are performed using special designed dosimeters based on CR-39 track detectors.     

Lithium neutron producing target for BINP accelerator-based neutron source.

Pilot innovative accelerator-based neutron source for neutron capture therapy is under construction now at the Budker Institute of Nuclear Physics, Novosibirsk, Russia. One of the main elements of the facility is lithium target, that produces neutrons via threshold (7)Li(p,n)(7)Be reaction at 25 kW proton beam with energies 1.915 or 2.5 MeV. In the present report, the results of experiments on neutron producing target prototype are presented, the results of calculations of hydraulic resistance for heat carrier flow and lithium layer temperature are shown. Calculation showed that the lithium target could run up to 10 mA proton beam before melting. Choice of target variant is substantiated. Program of immediate necessary experiments is described. Target design for neutron source constructed at BINP is presented. Manufacturing the neutron producing target up to the end of 2004 and obtaining a neutron beam on BINP accelerator-based neutron source are planned during 2005.     

High-power liquid-lithium target prototype for accelerator-based boron neutron capture therapy

A prototype of a compact Liquid-Lithium Target (LiLiT), which will possibly constitute an accelerator-based intense neutron source for Boron Neutron Capture Therapy (BNCT) in hospitals, was built. The LiLiT setup is presently being commissioned at Soreq Nuclear Research Center (SNRC). The liquid-lithium target will produce neutrons through the ⁷Li(p,n)⁷Be reaction and it will overcome the major problem of removing the thermal power generated using a high-intensity proton beam (>10 kW), necessary for sufficient neutron flux. In off-line circulation tests, the liquid-lithium loop generated a stable lithium jet at high velocity, on a concave supporting wall; the concept will first be tested using a high-power electron beam impinging on the lithium jet. High intensity proton beam irradiation (1.91–2.5 MeV, 2–4 mA) will take place at Soreq Applied Research Accelerator Facility (SARAF) superconducting linear accelerator currently in construction at SNRC. Radiological risks due to the ⁷Be produced in the reaction were studied and will be handled through a proper design, including a cold trap and appropriate shielding. A moderator/reflector assembly is planned according to a Monte Carlo simulation, to create a neutron spectrum and intensity maximally effective to the treatment and to reduce prompt gamma radiation dose risks.     

Beam shaping assembly optimization for 7Li(p,n)7Be accelerator based BNCT

Within the framework of accelerator-based BNCT, a project to develop a folded Tandem-ElectroStatic-Quadrupole accelerator is under way at the Atomic Energy Commission of Argentina. The proposed accelerator is conceived to deliver a proton beam of 30 mA at about 2.5 MeV. In this work we explore a Beam Shaping Assembly (BSA) design based on the ⁷Li(p,n)⁷Be neutron production reaction to obtain neutron beams to treat deep seated tumors.     

An Am/Be neutron source and its use in integral tests of differential neutron reaction cross-section data

An Am/Be neutron source, installed recently at the Rajshahi University, is described. Neutron flux mapping was done using the nuclear reactions ¹⁹⁷Au(n,γ)¹⁹⁸Au, ¹¹³In(n,γ)^114mIn, ¹¹⁵In(n,n′γ)^115mIn and ⁵⁸Ni(n,p)⁵⁸Co. An approximate validation of the neutron spectral shape was done using five neutron threshold detectors and the iterative unfolding code SULSA. Integral cross sections of the reactions ⁵⁴Fe(n,p)⁵⁴Mn, ⁵⁹Co(n,p)⁵⁹Fe and ⁹²Mo(n,p)^92mNb were measured with fast neutrons (E_n>1.5 MeV) and compared with data calculated using the neutron spectral distribution and the excitation function of each reaction given in data libraries: an agreement within±6% was found.     

Optimization parameters for BDE in BNCT using near threshold 7Li(p,n)7Be direct neutrons.

The dose contribution of (10)B(n,alpha)(7)Li reaction in BNCT using near threshold (7)Li(p,n)(7)Be direct neutrons can be increased through the use of materials referred to as boron-dose enhancers (BDE). In this paper, possible BDE optimization criteria were determined from the characteristics of candidate BDE materials namely (C(2)H(4))(n), (C(2)H(3)F)(n), (C(2)H(2)F(2))(n), (C(2)HF(3))(n), (C(2)D(4))(n), (C(2)F(4))(n), beryllium metal, graphite, D(2)O and (7)LiF. The treatable protocol depth (TPD) was used as the assessment index for evaluating the effect of these materials on the dose distribution in a medium undergoing BNCT using near threshold (7)Li(p,n)(7)Be direct neutrons. The maximum TPD (TPD(max)) did not exhibit an explicit dependence on material type as evidenced by its small range and arbitrary variations. The dependence of TPD on BDE thickness was influenced by the BDE material used as indicated by the sharply peaked TPD versus BDE thickness curves for materials with hydrogen compared to the broader curves obtained for those without hydrogen. The BDE thickness required to achieve TPD(max) (BDE(TPD(max))) were also found to be thinner for materials with hydrogen. The TPD(max), the dependence of TPD on BDE thickness, and the BDE(TPD(max)) were established as appropriate BDE optimization parameters. Based on these criteria and other practical considerations, the suitable choice as BDE among the candidate materials considered in this study for treatments involving tumors located at shallow depths would be (C(2)H(4))(n) while beryllium metal was judged as more appropriate for treatment of deep-seated tumors.     

Characteristics of BDE dependent on 10B concentration for accelerator-based BNCT using near-threshold 7Li(p,n)7Be direct neutrons.

The characteristics boron-dose enhancer (BDE) was evaluated as to the dependence on the (10)B concentration for BNCT using near-threshold (7)Li(p,n)(7)Be direct neutrons. The treatable protocol depth (TPD) was utilized as an evaluation index. MCNP-4B calculations were performed for near-threshold (7)Li(p,n)(7)Be at a proton energy of 1.900MeV and for a polyethylene BDE. Consequently, the TPD was increased by increasing T/N ratio, i.e., the ratio of the (10)B concentration in the tumor ((10)B(Tumor)) to that in the normal tissue ((10)B(Normal)), and by increasing (10)B(Tumor) and (10)B(Normal) for constant T/N ratio. It has been found that the BDE becomes unnecessary from the viewpoint of increasing the TPD, when (10)B(Tumor) is over a certain level.     

Calculations of the thermal and fast neutron fluxes in the Syrian miniature neutron source reactor using the MCNP-4C code

The MCNP-4C code, based on the probabilistic approach, was used to model the 3D configuration of the core of the Syrian miniature neutron source reactor (MNSR). The continuous energy neutron cross sections from the ENDF/B-VI library were used to calculate the thermal and fast neutron fluxes in the inner and outer irradiation sites of MNSR. The thermal fluxes in the MNSR inner irradiation sites were also measured experimentally by the multiple foil activation method (¹⁹⁷Au (n, γ) ¹⁹⁸Au and ⁵⁹Co (n, γ) ⁶⁰Co). The foils were irradiated simultaneously in each of the five MNSR inner irradiation sites to measure the thermal neutron flux and the epithermal index in each site. The calculated and measured results agree well.     

Analysis of epithermal neutron production by near-threshold (p,n) reactions

A recent advance in portable accelerator neutron source development was research on production of epithermal neutrons by near-threshold charged-particle reactions. When the projectile energy is accurately controlled at an energy close to the reaction threshold, the neutrons produced will have energies less than or around 100 keV and can be used with little or no moderation or filtration in neutron capture therapy. Although the total neutron yield is lower than at higher proton energies, the epithermal neutron flux may be sufficiently intense because of the softer energy spectrum and the requirement for less neutron moderation. This paper presents an analysis of the main characteristics of epithermal neutron production by this method using the Li (p,n) reaction as an example. The energy, yield and angular characteristics of neutron emission are discussed. The achievable epithermal fluxes are computed from experimental data. The results are used to assess the feasibility of near-threshold production of epithermal neutrons for neutron capture therapy with compact accelerators such as a RFQ proton acceelerator. The results indicated that, using a Li₃N target, 1 mA of 2 MeV protons will produce 10⁹ n/cm²/s with an average energy of 83 keV while 5.6 mA of 1.91 MeV protons can produce 10⁹ n/cm²/s with an average energy of 45 keV.