0.5eV到数百keV能区中子通量密度监测器的理论设计与实验研究

发布时间：2018-03-07 03:30

  本文选题：超热中子　切入点：十到数百keV能区中子　出处：《兰州大学》2016年博士论文　论文类型：学位论文

【摘要】：硼中子俘获治疗(born neutron capture therapy,BNCT)是一种非常有前途的癌症治疗技术。中子源是BNCT能否取得成功的关键因素之一。对于现代的BNCT中子源来说,超热中子(0.5 eVEn10 keV)通量密度是其基本物理特性之一。近年来,虽然BNCT中子源中能量较高的快中子(En10 keV)已经被尽可能适当地慢化,但十到数百keV能区内的快中子依然存在,很难被彻底去除。这种能量稍高于超热中子能量的快中子具有较大的相对生物学效应,会对人体的正常组织造成不必要的伤害。所以,为了评估BNCT中子源的品质,估计癌症病人在治疗过程中受到的侵害性中子辐照剂量,精确测量BNCT中子源中超热中子和十到数百keV能区中子的通量密度就显得十分必要了。然而,考虑到中子能谱的形状,想要直接而又精确地测量BNCT中子源中上述两类中子的通量密度还是相当困难的,因为截止到目前为止,还没有合适的谱仪可以用来直接测量中子能谱。因此,本论文工作基于活化法测量中子通量密度的基本原理,利用71Ga(n,?)72Ga活化反应,采用蒙特卡罗模拟的方法设计了以氮化镓(GaN)基片为活化材料的超热中子通量密度监测器和十到数百keV能区中子通量密度监测器,以分别用来精确测量BNCT中子源中这两类中子的通量密度。在当前设计的超热中子通量密度监测器中,活化材料,即GaN基片,被放置在聚乙烯球(中子慢化材料)的几何中心位置,聚乙烯球的外面由热中子吸收材料镉箔覆盖。本论文工作利用MCNP5程序模拟计算了这个监测器在中子能量0.01 eVEn10 MeV范围内的灵敏度。模拟结果表明,该监测器对超热中子十分灵敏且具有在超热中子能区内平坦的灵敏度曲线,而它对热中子(En0.5 eV)以及能量较高的快中子的灵敏度却明显较低。本论文工作设计的十到数百keV能区中子通量密度监测器由两个监测器组成。这两个监测器在外形上几乎完全相同,由外至内具有吸收剂/慢化剂/吸收剂/GaN基片的结构材料安排。这两个监测器之间的区别主要是所使用材料的种类、中子吸收材料的厚度以及中子慢化材料的直径等。利用MCNP5程序,本论文工作分别模拟计算了这两个监测器在中子能量0.01 eVEn10 MeV范围内的灵敏度。模拟结果表明,通过对这两个监测器的灵敏度作差,热中子、超热中子以及能量较高的快中子对监测器灵敏度的贡献几乎被完全地去除了,同时成功地得到了在十到数百keV能区内平坦的监测器灵敏度曲线。利用日本大阪大学的强流氘氚中子源装置OKTAVIAN,本论文工作对所设计的这两种不同类型中子通量密度监测器的性能分别进行了实验测试。实验结果表明,这两种不同类型的监测器可以分别用来精确测量BNCT中子源中超热中子和十到数百keV能区中子的通量密度,估计它们的测量精确度分别小于5%和10%。本论文工作设计的这两种不同类型的中子通量密度监测器均可以在不用测量中子能谱的情况下精确测量BNCT中子源中宽能量范围中子的通量密度,而这恰是本论文工作的创新性所在。
[Abstract]:Boron neutron capture therapy (born neutron capture therapy, BNCT) is a very promising technology of cancer treatment. Whether the BNCT neutron source is one of the key factors of success. For the modern BNCT neutron source, epithermal neutron flux density (0.5 eVEn10 keV) is one of the physical properties of the medium. In recent years, although the fast neutron neutron source in high energy BNCT (En10 keV) has been slow as well as possible, but ten to hundreds of keV fast neutron in the region still exist, it is difficult to be completely removed. This energy is slightly higher than that of epithermal neutron energy fast neutron has a relatively large effect on the biological. The body's normal tissue causing unnecessary harm. So, in order to evaluate the quality of BNCT neutron source, estimated by cancer patients during the treatment of invasive neutron irradiation dose, accurate measurement of BNCT neutron source in thermal neutron and ten to hundreds of K EV energy neutron flux density is necessary. However, considering the shape of the neutron spectrum, to direct and accurate measurement of BNCT neutron source in the above two kinds of neutron flux density is quite difficult, because so far, no suitable spectrometer can be used to directly measure the neutron spectrum. Therefore, the thesis based on the basic principle of the neutron flux density measurement with activation method, using 71Ga (n?) 72Ga activation reaction, using Monte Carlo method to design the gallium nitride (GaN) substrate for epithermal neutron flux density monitor of the active material and ten to hundreds of keV neutron flux density monitor. In the two kinds of neutron were used to accurate measurement of BNCT neutron source flux density. The activation material on the thermal neutron flux density monitor the current design, namely, the GaN substrate is placed in polyethylene ball (neutron Moderator material) geometric center, outside the ball by polyethylene thermal neutron absorbing material cadmium foil covered. This paper using MCNP5 program the monitor in the neutron energy sensitivity 0.01 eVEn10 MeV within the scope of the simulation. The simulation results show that the monitor of epithermal neutron is very sensitive and is in the super thermal neutron sensitivity the curve flat area, and its thermal neutron (En0.5 eV) and the sensitivity of fast neutron energy is higher the lower. This paper designed ten to hundreds of keV neutron flux density monitor from two monitoring device. The two monitors are almost identical in appearance, from the outside to the with the arrangement of absorbent / moderator / absorbent /GaN substrate material. The difference between the two monitor is the main type of materials used, neutron absorbing material thickness and neutron moderator material The diameter of the material. By using the MCNP5 program, the sensitivity of the two monitors were simulated in neutron energy of 0.01 eVEn10 in the range of MeV were calculated. The simulation results show that the sensitivity of the two monitors the difference of thermal neutron, fast neutron epithermal neutron energy and higher contribution to monitor the sensitivity of almost is completely removed, and successfully obtained in ten to hundreds of keV can monitor the sensitivity curve flat area. The Japanese Osaka University deuterium tritium neutron source device OKTAVIAN, performance of the work on the design of the two different types of neutron flux density monitor were tested. The experimental results this shows that the two types can be used to monitor the accurate measurement of BNCT neutron source in thermal neutron and ten to hundreds of keV neutron flux density estimation, their The measurement accuracy was less than 5% and 10%. the design of these two different types of neutron flux density monitor can accurate measurement of BNCT neutron source wide energy range neutron neutron spectrum measurement in no case of the flux density, which is the innovation of the thesis work.

【学位授予单位】：兰州大学
【学位级别】：博士
【学位授予年份】：2016
【分类号】：R730.5;O571.53
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本文编号：1577775