合肥工业大学学报自科版

合肥工业大学学报(自然科学版)

JOURNAL OF HEFEI UNIVERSITY OF TECHNOLOGY (NATURAL SCIENCE)

主管：中华人民共和国教育部

主办：合肥工业大学

ISSN 1003-5060, CN 34-1083/N

基于 OpenMC 的 BNCT 剂量计算研究

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Study on OpenMC-based BNCT dose calculation

期刊信息

合肥工业大学（自然科学版），2026年7月，第49卷第7期：951-956

DOI: 10.3969/j.issn.1003-5060.2026.07.013

作者信息

郭鑫 $ ^{1} $，林辉 $ ^{1} $，卢棚 $ ^{2} $，伍秋染 $ ^{3} $，梁立振 $ ^{3} $

（1. 合肥工业大学物理学院，安徽合肥 230601；2. 合肥综合性国家科学中心能源研究院（安徽省能源实验室）中子技术应用研究中心，安徽合肥 230031；3. 中国科学院合肥物质科学研究院等离子体物理研究所，安徽合肥 230031）

摘要和关键词

摘要: 硼中子俘获治疗(boron neutron capture therapy, BNCT)是通过中子与 $ ^{10} $B反应产生α、 $ ^{7} $Li重离子杀死癌细胞的二元疗法,其剂量成分较为复杂,包含快中子、热中子、α与 $ ^{7} $Li、γ射线,各成分的相对生物学效应不尽相同,开展BNCT剂量计算方法研究对临床照射剂量的评估具有重要应用价值。文章基于OpenMC蒙特卡洛粒子输运程序进行剂量计算方法研究,并验证其在BNCT剂量计算中应用的可行性和有效性。建立Snyder人体头部的解析几何模型;利用质子与锂反应获得中子能谱、强度和角度等中子源项数据,提供OpenMC源项建模;基于OpenMC网格计数统计BNCT中子剂量,并与MCNPX和FLUKA程序进行比较。利用OpenMC程序计算BNCT的剂量分布结果,与MCNPX模拟计算结果相比,粒子通量误差为0.59%;与FLUKA计算结果相比,粒子通量误差为9.18%,硼剂量最大误差为2.30%。尽管与MCNPX和FLUKA在仿真参数设置及数据后处理上存在差异,但OpenMC在BNCT剂量计算方面是可行且有效的,可为未来发展的BNCT治疗计划系统剂量的精确计算与治疗策略的优化选取提供可靠的软件工具。

关键词: OpenMC 程序；硼中子俘获治疗；剂量计算；蒙特卡洛

Authors

GUO Xin $ ^{1} $, LIN Hui $ ^{1} $, LU Peng $ ^{2} $, WU Qiuran $ ^{3} $, LIANG Lizhen $ ^{3} $

(1. School of Physics, Hefei University of Technology, Hefei 230601, China; 2. Neutron Technology Application Research Center, Institute of Energy, Hefei Comprehensive National Science Center (Anhui Energy Laboratory), Hefei 230031, China; 3. Institute of Plasma Physics, Hefei Institutes of Physical Science, Chinese Academy of Sciences, Hefei 230031, China)

Abstract and Keywords

Abstract: Boron neutron capture therapy (BNCT) is a binary therapy that kills cancer cells through the reaction of neutrons with $ {}^{10} $B which produces $ \alpha $ and $ {}^{7} $Li heavy ions. Its dose components are complex, including fast neutrons, thermal neutrons, $ \alpha $ and $ {}^{7} $Li particles, and $ \gamma $-rays. The relative biological effectiveness of each component differs. Therefore, research on BNCT dose calculation methods holds significant application value for the evaluation of clinical irradiation doses. This paper investigates the dose calculation method based on the OpenMC Monte Carlo particle transport code and validates its feasibility and effectiveness for BNCT dose calculations. An analytical geometric model of the Snyder human head phantom was established. Neutron source term data, including neutron energy spectrum, intensity, and angular distribution, were obtained using the proton-lithium reaction to provide source term modeling for OpenMC. BNCT neutron doses were tallied using OpenMC's mesh-based counting and compared with those from the MCNPX and FLUKA codes. Compared with the MCNPX simulation. tion results, the particle flux error of the BNCT dose distribution calculated by the OpenMC code is 0.59%; compared with the FLUKA results, the particle flux error is 9.18%, and the maximum error of the boron dose is 2.30%. Despite differences in simulation parameter settings and data post-processing relative to MCNPX and FLUKA, OpenMC proves to be feasible and effective for BNCT dose calculations. It can provide a reliable software tool for accurate dose calculation in future BNCT treatment planning systems and for the optimization of treatment strategies.

Keywords: OpenMC; boron neutron capture therapy(BNCT); dose calculation; Monte Carlo

基金信息

安徽省自然科学基金资助项目（2308085MA22）；合肥综合性国家科学中心能源研究院（安徽省能源实验室）资助项目（21KZS202）

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