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。     

裂变中子场的比释动能测定

用自行研制的IRM80S-TE1和IRM81T-TE1两种组织等效电离室与用Mullard G-M计数管组装的γ剂量计,组成双探测器,建立了裂变中子-γ混合场比释动能测定方法。测量结果总的估计不确定度为±5.2％。并与用美国FWT IC-17A组织等效电离室所测得的数据进行了比较,三种电离室在裂变中子场中的总响应及中子比释动能的测量结果在±0.4％之内符合。本文所用的测量方法对其它快中子场测量也是适用的。     

硼中子俘获治疗的加速器中子源7Li(p,n)7Be的中子特性研究

目的 通过研究质子加速器 ⁷Li(p,n)⁷Be 反应的中子特性,为研究和制作适用于硼中子俘获治疗(BNCT)的加速器中子源提供基础数据。方法 加速质子使其轰击Li靶后产生中子;通过金属箔活化法,测量中子与In箔发生阈值反应后放出的γ射线;然后计算出In箔的放射性活度、加速器反应后放出中子的注量和反应的微分截面。结果 质子加速轰击Li靶后,在不同方向产生不同能量和注量的中子。加速器电压分别为3.0、2.8和2.6 MV,出射中子与入射质子束的方向一致时, ⁷Li(p,n)⁷Be 反应的微分截面约为50 mb/mr;夹角为60°时,反应的微分截面减小到30 mb/mr左右。由于部分中子与其他金属原子等发生弹性散射而射向后方,提高了这一范围内In箔的比放射性活度,影响了其微分截面的准确性。结论 用金属箔活化法测定中子简便易行,可同时测得多个方向的中子分布,但需对中子与其他金属弹性散射产生的影响进行进一步的研究; ⁷Li(p,n)⁷Be 反应后发射出的中子经慢化后,能得到适于BNCT治疗的热中子和超热中子;若作为BNCT的中子源,加速器的质子束流需达到10 mA。     

硼中子俘获治疗技术的研究现状

对硼中子俘获治疗技术的原理、硼中子俘获治疗系统需要研究的相关内容及研究现状作了主要介绍;对加速器或反应堆产生超热中子束的方法及与硼中子俘获治疗相关的硼化合物作了较详细的讨论。     

清华大学反应堆大动物照射的剂量测定

清华大学游泳池式轻水反应堆1号孔道是国内目前唯一的裂变中子生物辐照场所,通过把不同厚度的铅过滤板横插于入射辐射束中,获得了四种不同中子-γ比释动能比率的混合中子-γ辐射条件,对狗进厅单侧照射。狗体模由TE液体组成,用双电离室方法测量吸收剂量.上述四种混合辐射场的中子-γ比值分别为l 0.6,2.06、1.12和0.16。确定了不同中子-γ比条件下沿水平中心线的深度剂量分布.给出了中心剂量、入射剂量和出射剂量值。按照lCRU10_e报告建议的划分照射均匀程度的指标,四种条件下的大动物(狗)照射都属于非均匀照射.     

p（35）Be快中子辐射场生物剂量参数的测定

报道了ｐ（３５）Ｂｅ快中子、γ混合辐射场生物剂量参数，说明了采用双电离室方法测量ｐ（３５）Ｂｅ快中子时，剂量计算中有关因子和参数的确定原则和估算结果；给出了辐射场的中子／γ比，照射野内组织比释动能率分布，以及由比释动能率计算小鼠全身与局部照射和血液样品照射时的吸收剂量转换系数等剂量常参数；最后还对剂量测量的误差问题做了分析。     

CR39中子个人剂量监测的不确定度评定

目的 通过对CR39中子个人剂量监测的不确定度评定,定量分析中子个人监测结果的可靠性,提高中子个人剂量监测的质量水平。方法 建立适当数学模型,通过对个人剂量计各个环节因素引入不确定度的计算,评定CR39中子个人剂量监测的扩展不确定度。结果 中子个人监测的A类不确定度主要来自测量值和本底,测量值为A类不确定度的最大分量,为11.68%。引入B类不确定度主要为能量响应为7.85%、剂量线性检验为4.0%、距离响应为1.68%和蚀刻条件为3.0%等。相对扩展不确定度约为32%,k=2。结论 放射性同位素场所不确定度评定中,对于某一较小能量范围的中子（即中子的能量分群）,应采用符合其能量响应特性的中子探测器。     

15MV医用直线加速器光中子蒙特卡罗模拟

目的 研究高能医用直线加速器运行过程中因光核反应所形成的光中子辐射场。方法 利用蒙特卡罗(MC)程序模拟Clinic 2300CD型医用电子加速器15 MV X射线模式下光中子污染,掌握机头内不同位置光中子能谱和不同照射野下等中心处中子周围剂量当量变化,分析光中子在等中心平面内剂量分布和水模体中剂量衰减。结果 准直器关闭时,加速器机头内靶、主准直器、均整器和多叶准直器下表面的光中子平均能量分别为1.08、1.20、0.35、0.30MeV;等中心处中子周围剂量当量随着照射野的增大先增大后减少,在30 cm × 30 cm照射野下达到最大;随着测点在水模体中的深度增加,中子通量先增加后减小,而中子剂量却在逐渐减小;不同照射野下,光中子剂量率在水模体深度20 cm处,基本都接近本底。结论 探究高能医用直线加速器机头光中子谱和剂量分布特点,以及光中子在水模体内剂量沉积规律,能为进一步研究高能医用直线加速器光中子污染对患者产生的附加剂量提供支持。     

无源效率刻度方法在中子剂量估算中的应用

目的在意外中子照射情况下,快速给出探测器的探测效率,利用24Na能谱分析法估算中子剂量。方法使用Genius-2000 GeomComposer软件实现对探测器的效率刻度。结果可快速给出探测器对3 ml血样中1.368 MeV能量的γ光子的探测效率,其数值为4.05271×10-2,软件给出的误差为4.0%。对中子照射样品的剂量估算值为1.94～2.82 Gy,算术平均值为2.38 Gy,剂量值的不确定度约为20.07%。结论将无源效率刻度方法应用在人体血液24Na能谱分析法中,可提高估算速度。     

术中放疗加速器中子剂量当量率的测量研究

目的 测量并分析术中放疗加速器9和12 MeV电子线在手术室内引起的中子剂量当量率,与西门子Primus加速器相同电子线能量档产生的中子污染进行比较,为放射治疗引起的二次致癌风险提供数据参考。方法 利用中子探测仪测量术中放疗加速器在9和12 MeV电子线于机头两端、限光筒底端、患者治疗平面,以及其他关键位置产生的中子剂量当量率。取相似的位置在西门子Primus加速器上进行相同方法的测量。分析测量结果,并将两种加速器产生的中子进行比较。结果 手术过程中使用9和12 MeV的电子线治疗时会产生中子,对患者以及工作人员产生潜在的健康隐患。9 MeV时,术中放疗加速器机头两端以及限光筒底端两侧的中子剂量当量率分别为（51.8±3.1）、（45.2±1.5）、（70.5±4.9）和（68.2±3.3）μSv/h,比12 MeV产生的中子分别低5.9%、5.4%、17.8%和21.5%。手术室门内侧在9和12 MeV时产生的中子剂量当量率极低,可以忽略。西门子加速器出束9 MeV时,在相似测量点处产生的中子剂量当量率为（277.3±1.2）、（285.1±1.6）、（185.1±1.8）、（182.8±2.4）μSv/h,比12 MeV的分别低48.8%、47.6%、48.7%、和52.2%。能量达到12 MeV时,西门子Primus加速器产生的中子剂量当量率是术中加速器的10倍以上。结论 两种医用加速器12 MeV电子线产生的中子剂量当量率远高于9 MeV产生的中子,增加了患者第二原发癌的风险;传统医用加速器在相同能量档产生的中子剂量当量率远高于术中电子加速器,应采取适当的屏蔽防护。     

Experimental verification of beam characteristics for cyclotron-based epithermal neutron source (C-BENS)

A cyclotron-based epithermal neutron source has been developed for boron neutron capture therapy. This system consists of a cyclotron accelerator producing 1.1-mA proton beams with an energy of 30 MeV, a beam transport system coupled with a beryllium neutron production target, and a beam-shaping assembly (BSA) with a neutron collimator. In our previous work, the BSA was optimized to obtain sufficient epithermal neutron fluxes of using a Monte Carlo simulation code. In order to validate the simulation results, irradiation tests using multi-foil activation at the surface of a gamma-ray shield located behind the collimator and water phantom experiments using a collimated epithermal neutron beam were performed. It was confirmed experimentally that the intensity of the epithermal neutrons was 1.2×10⁹ cm⁻² s⁻¹.     

Evaluation of thermal neutron irradiation field using a cyclotron-based neutron source for alpha autoradiography

It is important to measure the microdistribution of ¹⁰B in a cell to predict the cell-killing effect of new boron compounds in the field of boron neutron capture therapy. Alpha autoradiography has generally been used to detect the microdistribution of ¹⁰B in a cell. Although it has been performed using a reactor-based neutron source, the realization of an accelerator-based thermal neutron irradiation field is anticipated because of its easy installation at any location and stable operation. Therefore, we propose a method using a cyclotron-based epithermal neutron source in combination with a water phantom to produce a thermal neutron irradiation field for alpha autoradiography. This system can supply a uniform thermal neutron field with an intensity of 1.7×10⁹ (cm⁻² s⁻¹) and an area of 40 mm in diameter. In this paper, we give an overview of our proposed system and describe a demonstration test using a mouse liver sample injected with 500 mg/kg of boronophenyl-alanine.     

Neutron fluence depth profiles in water phantom on epithermal beam of LVR-15 research reactor

Horizontal channel with epithermal neutron beam at the LVR-15 research reactor is used mainly for boron neutron capture therapy. Neutron fluence depth profiles in a water phantom characterise beam properties. The neutron fluence (approximated by reaction rates) depth profiles were measured with six different types of activation detectors. The profiles were determined for thermal, epithermal and fast neutrons.     

Spatial and spectral characteristics of a compact system neutron beam designed for BNCT facility

The development of suitable neutron sources and neutron beam is critical to the success of Boron Neutron Capture Therapy (BNCT). In this work a compact system designed for BNCT is presented. The system consists of ²⁵²Cf fission neutron source and a moderator/reflector/filter/shield assembly. The moderator/reflector/filter arrangement has been optimized to maximize the epithermal neutron component which is useful for BNCT treatment of deep seated tumors with the suitably low level of beam contamination. The MCMP5 code has been used to calculate the different components of neutrons, secondary gamma rays originating from ²⁵²Cf source and the primary gamma rays emitted directly by this source at the exit face of the compact system. The fluence rate distributions of such particles were also computed along the central axis of a human head phantom.     

Epithermal neutron beam interference with cardiac pacemakers

In this paper, a phantom study was performed to evaluate the effect of an epithermal neutron beam irradiation on the cardiac pacemaker function. Severe malfunction occurred in the pacemakers after substantially lower dose from epithermal neutron irradiation than reported in the fast neutron or photon beams at the same dose rate level. In addition the pacemakers got activated, resulting in nuclides with half-lives from 25 min to 115 d. We suggest that BNCT should be administrated only after removal of the pacemaker from the vicinity of the tumor.     

Optimization of the target of an accelerator-driven neutron source through Monte Carlo numerical simulation of neutron and gamma transport by the PRIZMA code

At Budker Institute of Nuclear Physics, epithermal neutron source for neutron-capture therapy was built and neutron generation was realized. Source is based on tandem accelerator and uses near-threshold neutron generation from the reaction ⁷Li(p,n)⁷Be. The paper describes target optimization through the numerical simulation of proton, neutron and gamma transport by Monte Carlo method (PRIZMA code). It is shown that the near-threshold mode attractive due low activation provides high efficiency of the dose and acceptable therapeutic ratio and advantage depth.     

252Cf裂变中子源在组织等效模体中的中子和γ辐射剂量分布的计算

目的：计算^252Cf裂变中子源的的中子和γ辐射在组织等模体内的剂量分布,为使用^252Cf裂变中子源进行中子放疗提供有用的剂量学参数。方法：建立^252Cf源和组织等效模体的三维几何计算模型,利用蒙特卡罗方法进行中子和γ辐射联合输运计算。结果：计算了两种医用^252Cf裂变中子源在水、血液、肌肉、皮肤、骨骼和肺组织等效材料构成的模体中距源不同距离点处的中子和γ辐射吸收剂量。结论：蒙特卡罗计算结果与文献数据以及使用双电离室实验测量的结果符合得较好。对^252Cf裂变中子源在5种组织材料构成的模体中中子和γ辐射的剂量分布进行了比较,使用水作为组织等效材料对^252Cf裂变中子源在在以肌肉、血液和皮肤构成的局部组织内的剂量分布进行模拟计算,可取得较可靠的结果。     

Comparison of neutron spectrum measurement methods used for the epithermal beam of the LVR-15 research reactor

The LVR-15 research reactor's horizontal channel with its epithermal neutron beam is used mainly for boron neutron capture therapy. Neutrons from the reactor core pass through a special filter before the collimator and the beam outlet. Neutron fluence and spectrum are the basic characteristics of an epithermal neutron beam. Three methods used to measure the beam's neutron spectrum are described: the activation method, a Bonner sphere spectrometer with gold activation detectors and a Bonner sphere spectrometer with LiI(Eu) scintillation detector. Examples of results are compared and discussed.