碳离子在不同材料叶片的多叶光栅上产生的次级粒子研究

在碳离子放射治疗中，碳离子束与治疗头设备和患者身体相互作用产生的次级粒子可以到达患者体内的许多区域，在产生的次级粒子中以中子和$\gamma $射线的产额为最大。在不影响束流配送功能的情况下，减少碳离子放疗中产生的次级中子和$\gamma $射线对于降低放疗后出现的正常组织并发症及二次肿瘤风险有着非常重要的意义。本文利用蒙特卡罗(Monte Carlo)方法模拟计算了被动式束流配送系统下，400 MeV/u碳离子束照射到由不同材料叶片构成的多叶光栅(MLC)形成典型的10 cm×10 cm方形射野时，在水模体中产生的次级中子和$\gamma $射线所沉积的剂量及空间分布等。模拟结果显示:碳离子束通过MLC形成射野后在水模体中产生的次级中子主要分布在水模体的入射端，次级$\gamma $射线较为均匀的分布在整个水模体内，且较多分布在具有展宽Bragg峰(SOBP)射野在水模体中贯穿时的坪区。对于MLC叶片材料的选择，则需根据实际情况对叶片厚度以及次级粒子当量剂量的要求来确定。本文通过模拟研究不同MLC叶片材料产生次级粒子的情况，为被动式束流配送系统中MLC叶片及其他元件的材料选择提供了科学依据。     

利用气泡探测器测量激光快中子

在利用超强激光驱动中子源的研究和应用研究中,中子源的产额及其角分布至关重要.我们在星光Ⅲ号激光装置上采用气泡探测器对强激光驱动的中子源的产额及其角分布进行了测量.利用超强皮秒激光与碳氘薄膜靶相互作用产生高能氘离子束撞击次级碳氘靶,通过氘-氘核反应产生准单能快中子.实验发现中子束的发射具有一定的方向性,在入射氘离子的传输方向上中子束具有更高的强度,测量得到的中子束最大强度为5.13×10⁷ n/sr.利用实验测量的氘离子能谱和角分布对中子束角分布进行了理论计算,结果与实验测量基本一致.     

50MeV/u~(12)C离子实验靶区出射中子角分布实验测量

本文用氟和碳阈探测器对50MeV/u~(12)C离子引起的重离子反应实验靶区E_n>11MeV和E_n>20MeV中子的角分布进行了实验测量,结果显示出明显的前冲分布,较高能量的中子具有更强的前冲趋势.     

50MeV/u 18O离子轰击Be、Cu和Au厚靶的中子产额和中子发射率

用阈探测器中子活化法测量了50MeV/u ¹⁸O离子轰击Be.Cu和Au厚靶的总中子产额和前向中子发射率. 实验结果表明,中子产额对靶核原子序数有一定的依赖性,较轻靶子的中子产额和中子发射率高于较重的靶子;当靶核(Cu)和入射离子单核能(50MeV)相同时,¹⁸O离子的中子产额约为¹²C离子的4倍.     

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实验研究了超强激光与平面薄膜Cu靶的相互作用中,在靶背法线方向产生的离子能量分布.实验中采用固体核径迹探测器CR39和Thomson谱仪相结合测量离子能谱.实验中同时测量到了质子和C4+、C3+、C2+和C1+离子.实验结果表明:离子沿着靶背法线方向发射,离子在一定能量处出现截断;在固体核径迹探测器CR39上可以看出碳离子信号比质子信号弱,质子的产额远高于碳离子,质子对碳离子有屏蔽作用;在相同发次中C4+、C3+、C2+和C1+离子随着荷质比的增大,相应离子的截止能增加.     

50MeV/u 12C离子实验靶区出射中子角分布实验测量

本文用氟和碳阈探测器对50MeV/u ¹²C离子引起的重离子反应实验靶区E_n>11MeV和E_n>20MeV中子的角分布进行了实验测量,结果显示出明显的前冲分布,较高能量的中子具有更强的前冲趋势.     

氢离子与碳膜相互作用的实验研究

此文报道了H+、H+2和H+3离子通过碳膜后,在其中产生的各种微量产物的测量结果.讨论了离子与碳膜作用中的电荷交换过程.分析了这些微量产物的形成机理.实验证明在产物的形成中电荷交换过程起关键作用.研究了离子与碳膜作用中的团簇效应和尾流效应.     

台面激光核聚变中子源研究进展

通过超短脉冲激光与合适的介质相互作用,在普通光学实验室中可以产生聚变中子,经过特殊设计的固体靶结构和特殊安排的级联过程,亚ＴＷ激光即可引发光核反应产生中子,通过台面超强激光与固体靶相互作用利用对离子的自导通道径向加速机制,可获得较多的高能离子和较多的中子产率,由于超强激光辐照下大的团簇爆炸时可获得大量高能离子,团簇靶实验中实现的中子产率比相似激光能量的固体靶实验高两个量级,文章讨论了上述三种产生中     

基于中子飞行时间法的ICF内爆热斑离子温度诊断技术北大核心CSCD

在惯性约束聚变研究中,内爆热斑离子温度反映了热斑能量的高低,对内爆对称性和内爆速度等物理量十分敏感,是理解内爆物理过程不可或缺的重要参数。介绍了一种基于中子飞行时间法的ICF内爆热斑离子温度诊断技术。建立了一种采用塑料闪烁探测器作为中子测量器件的快时间响应中子飞行时间谱仪。谱仪输出时间波形的半高全宽小于1.1ns,上升时间约为0.5ns。描述了基于反卷积运算和低通滤波的飞行时间谱解谱方法。在神光Ⅲ原型装置较低的中子产额和离子温度条件下通过这种诊断技术成功获得了内爆热斑离子温度。     

超强激光与等离子体相互作用产生中子的计算

对超强激光与等离子体相互作用过程中发生的氘-氘反应的中子能谱进行了计算.并将计算结 果与实验结果相比较.采用麦克斯韦能量分布和高斯形式的角分布对实验结果进行拟合，从 而确定了入射氘离子的温度和角分布，为研究离子的加速机制提供了依据. 关键词： 中子能谱 激光与等离子体相互作用     

Comparison of measured and calculated in-air secondary neutrons in passive carbon-ion radiotherapy

Undesired radiation exposure in normal tissues around a treatment volume in proton and carbon-ion radiotherapies is less than that in the conventional radiotherapies due to physical and/or biological properties of charged particles. Such exposure is always considered in a treatment planning, however, undesired exposure in normal tissues far from the treatment volume cannot be considered in the treatment planning, because it is caused by secondary radiation as well as leakage primary particles. Though this exposure is considerably lower than that near the treatment volume, it may be not negligible to estimate the risk of secondary cancer especially for the young patients. In particular, the assessment of the secondary neutrons that inevitably produced within the patient and beam line devices is very important due to the potency of their biological effect. The distributions of the absorbed dose and the biological effectiveness in phantom/patient are required to assess the risk, and Monte Carlo calculation plays a key role due to a difficulty of the measurements. In this study, comparison of measured and calculated in-air neutrons at the patient position in the Heavy Ion Medical Accelerator in Chiba (HIMAC) treatment room are performed to verify the accuracy of the Monte Carlo code, PHITS. Our calculations underestimated epithermal neutrons measured by Bonner sphere system. This discrepancy may be caused by an insufficiency of the calculational geometry modeling, consequently an underestimation of neutrons scattered and moderated by the beam line devices. However, it is unlikely that the underestimation significantly contribute to the dose estimation in phantom. On the other hand, the calculation reproduced the measured ambient dose equivalents well because they were dominated by neutrons above 0.1 MeV. This result showed that the PHITS code has a potential ability to evaluate the neutron exposure of the patient in passive carbon-ion radiotherapy.     

The influence of field size and off-axis distance on photoneutron spectra of the 18 MV Siemens Oncor linear accelerator beam

At present, high energy electron linear accelerators (LINACs) producing photons with energies higher than 10 MeV have a wide use in radiotherapy (RT). However, in these beams fast neutrons could be generated, which results in undesired contamination of the therapeutic beams. These neutrons affect the shielding requirements in RT rooms and also increase the out-of-field radiation dose to patients. The neutron flux becomes even more important when high numbers of monitor units are used, as in the intensity modulated radiotherapy. Herein, to evaluate the exposure of patients and medical personnel, it is important to determine the full radiation field correctly. A model of the dual photon beam medical LINAC, Siemens ONCOR, used at the University Hospital Centre of Osijek was built using the MCNP611 code. We tuned the model according to measured photon percentage depth dose curves and profiles. Only 18 MV photon beams were modeled. The dependence of neutron dose equivalent and energy spectrum on field size and off-axis distance in the patient plane was analyzed. The neutron source strength (Q) defined as a number of neutrons coming from the head of the treatment unit per x-ray dose (Gy) delivered at the isocenter was calculated and found to be 1.12 × 10¹² neutrons per photon Gy at isocenter. The simulation showed that the neutron flux increases with increasing field size but field size has almost no effect on the shape of neutron dose profiles. The calculated neutron dose equivalent of different field sizes was between 1 and 3 mSv per photon Gy at isocenter. The mean energy changed from 0.21 MeV to 0.63 MeV with collimator opening from 0 × 0 cm² to 40 × 40 cm². At the 50 cm off-axis the change was less pronounced. According to the results, it is reasonable to conclude that the neutron dose equivalent to the patient is proportional to the photon beam-on time as suggested before. Since the beam-on time is much higher when advanced radiotherapy techniques are used to fulfill high conformity demands, this makes the neutron flux determination even more important. We also showed that the neutron energy in the patient plane significantly changes with field size. This can introduce significant uncertainty in dosimetry of neutrons due to strong dependence of the neutron detector response on the neutron energy in the interval 0.1–5 MeV.     

2.45 MeV小型D-D中子管准直屏蔽结构设计

为获取小角度出射的单能中子源,采用蒙特卡罗软件对小型D-D中子管产生的能量为2.45 MeV的4立体角中子源进行了准直屏蔽结构设计。准直屏蔽结构分为准直器和捕获穴,准直器采用铁+含硼聚乙烯+铅的三层过滤结构,用于屏蔽照射野外杂散中子,捕获穴主要功能是增加反方向中子的弹性散射次数,从而降低出射束低能散射中子的比例。通过MCNP模拟得到了准直器各层材料的最佳厚度和出射孔尺寸以及捕获穴最佳结构。经验证,中子照射野处2.45 MeV的中子通量比照射野外大三个量级,中子照射野处低能中子通量比2.45 MeV的中子通量低一个量级,墙壁外总剂量率（中子+）在2.5 Gy/h以下。该研究对于小角度单能中子源的快速获取具有一定的实用价值,获取的中子源可用于中子剂量仪器有效性检验、中子监测仪性能测试等方面的研究。     

(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.     

Neutron spectrometry and determination of neutron ambient dose equivalents in different LINAC radiotherapy rooms

A project has been set up to study the effect on a radiotherapy patient of the neutrons produced around the LINAC accelerator head by photonuclear reactions induced by photons above ～8 MeV. These neutrons may reach directly the patient, or they may interact with the surrounding materials until they become thermalised, scattering all over the treatment room and affecting the patient as well, contributing to peripheral dose. Spectrometry was performed with a calibrated and validated set of Bonner spheres at a point located at 50 cm from the isocenter, as well as at the place where a digital device for measuring neutrons, based on the upset of SRAM memories induced by thermal neutrons, is located inside the treatment room. Exposures have taken place in six LINAC accelerators with different energies (from 15 to 23 MV) with the aim of relating the spectrometer measurements with the readings of the digital device under various exposure and room geometry conditions. The final purpose of the project is to be able to relate, under any given treatment condition and room geometry, the readings of this digital device to patient neutron effective dose and peripheral dose in organs of interest. This would allow inferring the probability of developing second malignancies as a consequence of the treatment. Results indicate that unit neutron fluence spectra at 50 cm from the isocenter do not depend on accelerator characteristics, while spectra at the place of the digital device are strongly influenced by the treatment room geometry.     

Measurement of photoneutron doses in and out of high-energy X-ray beam of a SATURNE-20 medical linear accelerator by ECE polycarbonate detectors

Photoneutron contaminations in and out of high energy X-ray beams of the medical linear accelerator SATURNE 20 (CGR) of the Radiotherapy Department of Omeed Hospital in Isfahan, Iran, have been determined using 250 μm polycarbonate (PC) dosimeters, in strips or in sheets, processed by electrochemical etching (ECE) using specially designed ECE chambers to etch larger sheets. A two dimensional or topographical distribution of neutron contamination was also determined in a full size beam. The neutron dose equivalents (Hn) in the beam of 18 MV X-rays at 80 cm FSD were determined to be linear functions of X-ray dose equivalents (Hx) up to 1400 cSv. The distribution of the Hn at different X-ray doses showed bell-shape profiles with maxima at the isocenter. The ratios of dose equivalents of neutrons to those of X-rays increased as the field size increased having values of 0.22%, 0.28%, 0.31% and 0.37% for field sizes of 10×10, 20×20, 30×30, and 40×40 cm² respectively. Although such neutron dose equivalents can be corrected for patient treatment, it can cause radiation protection problems for workers where the design of the facility is not well planned.     

Study of neutron dose equivalent at the HIRFL deep tumor therapy terminal

The secondary neutron fields at the deep tumor therapy terminal at HIRFL(Heavy Ion Research Facility in Lanzhou) were investigated. The distributions of neutron ambient dose equivalent were measured with a FHT762Wendi-II neutron ambient dose equivalent meter as ~(12)C ions with energies of 165, 207, 270, and 350 Me V/u were bombarded on thick tissue-like targets. The thickness of targets used in the experiments was larger than the range of the carbon ions. The neutron spectra and dose equivalent were simulated by using FLUKA code, and the results agree well with the experimental data. The experiment results showed that the neutron dose produced by fragmentation reactions in tissue can be neglected in carbon-ion therapy, even considering their enhanced biological effectiveness.These results are also valuable for radiation protection, especially in the shielding design of high energy heavy ion medical machines.     

Evaluation of neutron radiation field in carbon ion therapy

Carbon ions have significant advantages in tumor therapy because of their physical and biological properties. In view of the radiation protection, the safety of patients is the most important issue in therapy processes.Therefore, the effects of the secondary particles produced by the carbon ions in the tumor therapy should be carefully considered, especially for the neutrons. In the present work, the neutron radiation field induced by carbon ions was evaluated by using the FLUKA code. The simulated results of neutron energy spectra and neutron dose was found to be in good agreement with the experiment data. In addition, energy deposition of carbon ions and neutrons in tissue-like media was studied, it is found that the secondary neutron energy deposition is not expected to exceed 1% of the carbon ion energy deposition in a typical treatment.     

Neutron measurements onboard the space shuttle

The radiation environment inside a shielded volume is highly complex, consisting of both charged and neutral particles. Since the inception of human space flights, the charged particle component has received virtually all of the attention. There is however, a significant production of secondary neutrons, particularly from the aluminum structure in low earth orbiting spacecrafts. The interactions of galactic cosmic rays (GCR), and solar energetic particles with the earth's atmosphere produce a non-isotropic distribution of albedo neutrons. Inside any reasonable habitable module, the average radiation quality factor of neutrons is about 4-5 times larger than the corresponding average quality factor of charged particles. The measurement of neutrons and their energy spectra is a difficult problem due the intense sources of charged particles. This paper reviews the results of Shuttle flight experiments (made during both solar maximum and solar minimum) to measure the contribution of neutrons to the dose equivalent, as well as theoretical calculations to estimate the appropriate range of neutron energies that contribute most to the dose equivalent.