本文選題：超熱中子　切入點：十到數百keV能區(qū)中子　出處：《蘭州大學》2016年博士論文　論文類型：學位論文

【摘要】：硼中子俘獲治療(born neutron capture therapy,BNCT)是一種非常有前途的癌癥治療技術。中子源是BNCT能否取得成功的關鍵因素之一。對于現代的BNCT中子源來說,超熱中子(0.5 eVEn10 keV)通量密度是其基本物理特性之一。近年來,雖然BNCT中子源中能量較高的快中子(En10 keV)已經被盡可能適當地慢化,但十到數百keV能區(qū)內的快中子依然存在,很難被徹底去除。這種能量稍高于超熱中子能量的快中子具有較大的相對生物學效應,會對人體的正常組織造成不必要的傷害。所以,為了評估BNCT中子源的品質,估計癌癥病人在治療過程中受到的侵害性中子輻照劑量,精確測量BNCT中子源中超熱中子和十到數百keV能區(qū)中子的通量密度就顯得十分必要了。然而,考慮到中子能譜的形狀,想要直接而又精確地測量BNCT中子源中上述兩類中子的通量密度還是相當困難的,因為截止到目前為止,還沒有合適的譜儀可以用來直接測量中子能譜。因此,本論文工作基于活化法測量中子通量密度的基本原理,利用71Ga(n,?)72Ga活化反應,采用蒙特卡羅模擬的方法設計了以氮化鎵(GaN)基片為活化材料的超熱中子通量密度監(jiān)測器和十到數百keV能區(qū)中子通量密度監(jiān)測器,以分別用來精確測量BNCT中子源中這兩類中子的通量密度。在當前設計的超熱中子通量密度監(jiān)測器中,活化材料,即GaN基片,被放置在聚乙烯球(中子慢化材料)的幾何中心位置,聚乙烯球的外面由熱中子吸收材料鎘箔覆蓋。本論文工作利用MCNP5程序模擬計算了這個監(jiān)測器在中子能量0.01 eVEn10 MeV范圍內的靈敏度。模擬結果表明,該監(jiān)測器對超熱中子十分靈敏且具有在超熱中子能區(qū)內平坦的靈敏度曲線,而它對熱中子(En0.5 eV)以及能量較高的快中子的靈敏度卻明顯較低。本論文工作設計的十到數百keV能區(qū)中子通量密度監(jiān)測器由兩個監(jiān)測器組成。這兩個監(jiān)測器在外形上幾乎完全相同,由外至內具有吸收劑/慢化劑/吸收劑/GaN基片的結構材料安排。這兩個監(jiān)測器之間的區(qū)別主要是所使用材料的種類、中子吸收材料的厚度以及中子慢化材料的直徑等。利用MCNP5程序,本論文工作分別模擬計算了這兩個監(jiān)測器在中子能量0.01 eVEn10 MeV范圍內的靈敏度。模擬結果表明,通過對這兩個監(jiān)測器的靈敏度作差,熱中子、超熱中子以及能量較高的快中子對監(jiān)測器靈敏度的貢獻幾乎被完全地去除了,同時成功地得到了在十到數百keV能區(qū)內平坦的監(jiān)測器靈敏度曲線。利用日本大阪大學的強流氘氚中子源裝置OKTAVIAN,本論文工作對所設計的這兩種不同類型中子通量密度監(jiān)測器的性能分別進行了實驗測試。實驗結果表明,這兩種不同類型的監(jiān)測器可以分別用來精確測量BNCT中子源中超熱中子和十到數百keV能區(qū)中子的通量密度,估計它們的測量精確度分別小于5%和10%。本論文工作設計的這兩種不同類型的中子通量密度監(jiān)測器均可以在不用測量中子能譜的情況下精確測量BNCT中子源中寬能量范圍中子的通量密度,而這恰是本論文工作的創(chuàng)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.

【學位授予單位】：蘭州大學
【學位級別】：博士
【學位授予年份】：2016
【分類號】：R730.5;O571.53
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本文編號：1577775