131Cs,125I和103Pd近距离治疗源的径向剂量函数研究

临床使用新粒子源近距离治疗之前,需要对粒子源的剂量学参数进行严格确定.本文依照美国医学物理家协会TG43U1推荐的剂量计算公式来研究131Cs,125I和103Pd等近距离治疗粒子源的径向剂量函数.计算中,采用MCNP和EGSnrc两种蒙特卡罗方法;并采用液态水和固体水(Solid Water,WT1)两种材料,粒子源径向研究范围为0.5cm～8cm.在同种粒子源计算结果与前研究的比较中,本研究中在液态和固体水中计算的结果与报道中采用蒙特卡罗计算和TLD测量的结果相当一致.     

6 MV Truebeam治疗头的MC模拟与特性研究

用MC程序EGSnrc准确快速地模拟6 MV Varian Truebeam医用直线加速器治疗头,分析其能谱特性,并验证模拟结果的准确性,为后续更精确的剂量计算建立基础。以Varian公司提供的出束窗口位置处的IAEA相空间文件作为输入源,模拟计算源皮距为100 cm,射野大小为10 cm×10 cm的治疗野平面上的粒子相空间分布,并以此粒子相空间数据为基础进一步模拟计算了边长为30 cm的标准水体模中的剂量分布。通过对数据分析得到了治疗野平面上的粒子能谱、角分布、平均能量、粒子注量等信息,以及均匀水体模中光子的百分深度剂量和离轴比,所得结果与文献报道符合较好。结果表明EGSnrc程序能够准确快速地模拟加速器治疗头,其剂量的模拟计算结果可为临床放射治疗供很好参考。     

不同蒙特卡罗程序高气压电离室能量响应特性差异的研究

蒙特卡罗程序常被用于高气压电离室能量响应特性的研究。但在相同辐射场条件下,用不同蒙特卡罗程序计算的高气压电离室能量响应的结果各不相同,与实验相比也存有差异。本次研究通过使用MCNP5、EGSnrc和FLUKA计算得到了低空气比释动能系列X射线和~(137)Cs能量光子的高气压电离室的灵敏度因子,并与实验结果比较。结果表明这三种MC程序计算的灵敏度因子基本一致,在~(137)Cs光子能量时与实验结果相差均小于25%。通过比较各MC程序计算所用时间,发现EGSnrc计算效率最高。最后分析得到因为这三种MC程序对电子输运的处理方法的不同导致了计算结果和计算效率的差异。     

~(60)Coγ辐照加工产品剂量分布的蒙特卡罗模拟

对上海金鹏源(SHJPY)~(60)Coγ辐照装置进行数学建模,运用蒙特卡罗方法(MCNP),模拟计算辐照装置在装载0.1g/cm3的均匀产品情况下的剂量分布,软件模拟计算结果的统计误差控制在5%以内,模拟计算结果与实际0.1g/cm3产品剂量分布测试结果比较,发现偏差的绝对值在15%(多数在8%以下)以内,模拟计算与测量数据基本吻合,计算结果可以反应产品吸收剂量的分布规律。     

MC法对移动式工业γ探伤机周边辐射模拟

为准确划定工业γ探伤机的控制区与监督区,应用MCNP5程序构建了移动式γ射线机计算模型。该模型充分考虑了工业γ探伤机散射对周围剂量的贡献,可有效地应用于工业γ探伤机控制区与监督区的划定与环境影响评价。实验结果表明:MCNP程序计算结果与经验公式结果基本一致,MC模拟方法对移动式工业γ射线机周边辐射剂量估算是可行的。     

一种高灵敏度环境中子剂量当量仪的能量响应计算与实验验证

利用MCNP4C程序计算了一种高灵敏度环境中子剂量当量仪的响应曲线.计算结果在感兴趣能量区间与ICRP建议书中的H*(10)曲线符合较好,通过实验验证,计算结果与实验数据相对偏差在-21.6%以内(对中子要求偏差在±50%以内),实验表明,用MCNP程序优化设计探测器可提高设计效率,并可同时为实验验证提供参考数据.     

高灵敏度环境中子剂量当量仪的能量响应计算

利用MCNP4C程序计算1种高灵敏度环境中子剂量当量仪的响应曲线.计算结果在感兴趣能量区间与ICRP建议书中的H*(10)曲线符合较好.对MCNP程序计算结果的合理性进行了计算验证.计算结果表明,用MCNP程序优化设计探测器可提高设计效率,并可同时为实验验证提供参考数据.     

~(137)Cs稳谱源的研制

通过对~(137)Csγ射线穿过陶瓷体源芯、不锈钢源壳和钨钢壳后衰减规律的研究,参考用户提供的~(137)Cs稳谱源的表面γ剂量率,确定~(137)Cs稳谱源的活度。对陶瓷体源芯吸附性能进行研究,以提高~(137)Cs稳谱源γ射线输出率的一致性,最终确定~(137)Cs稳谱源的制备工艺。经使用证明,该源的γ射线输出率准确、产品一致性好,已应用于岩性密度测井仪测井,可实现~(137)Cs稳谱源的国产化。     

60Co辐照装置剂量场均匀性的优化设计

采用蒙特卡罗方法和MATLAB(Matrix Laboratory)程序对由576根钴源构成的辐照装置中剂量场均匀性进行优化设计.首先将程序计算结果与实验测量进行比较,接着对程序设置3200个计算点计算了辐照装置的剂量场分布;剂量场均匀性计算中,不断地改变钴源棒位置并不断计算空间平面剂量场.计算结果发现,程序计算值与实验测量值较为一致,经调整后装置的空间剂量均匀度显著提高,使得辐照装置的能量使用率大大提高.     

计算X射线通量的几种方法的比较

综述了目前计算X射线在材料中通量的几种方法:利用质量衰减系数直接计算的理论方程和描述粒子运动统计规律的蒙特卡罗方法.分别采用理论方程、蒙特卡罗软件MCNP4C和EGSnrc三种途径计算了黑体谱X射线穿过几种常用复合材料的透射率和透射谱,根据计算过程和计算结果,结合相关参考文献,分析比较了理论方程与蒙特卡罗软件之间、两种蒙特卡罗软件MCNP4C和EGSnrc之间的差异.     

MCNP5模拟钴-60辐照装置产品吸收剂量的研究

根据某单位钴-60工业辐照装置的特点，以双板钴源源架中心为坐标原点，采用蒙特卡罗软件MCNP5建立更接近实际的辐照模型。模拟空气参考面的吸收剂量分布以验证排源方案。模拟计算不同对辐照箱（8对、16对、24对、48对）在不同悬挂链速度情况下单个辐照箱内产品的平均吸收剂量，并与实验测量值进行对比。结果表明，模拟值与实验测量值的平均相对偏差小于10%。MCNP5建立的辐照模型以及模拟计算结果具有很好的符合性并且可信，能反映辐照箱中产品的剂量分布情况，对实际工作和研究具有指导意义。     

The influence of calculated CT X-ray spectra on organ absorbed dose to a human phantom

We have evaluated the utilization of five X-ray spectra codes for Monte Carlo (MC) simulations of computed tomography (CT) examinations. Four codes (Xcomp5r, X-raytbc, X-rayb&m and Srs-78) are semi-empiricals and one is based on MC methods (EGSnrc/BEAM Monte Carlo code). The X-ray spectra calculated by the semi-empirical codes were compared with the X-ray spectrum calculated by the EGSnrc/BEAM MC code. The absorbed doses to each organ or tissue were also compared. The calculated doses, and its respective organs, for which occurs the greatest disagreement, as well as the calculated doses for the testes and red bone marrow (two important organs used for calculating effective dose) were presented. The results obtained in this work are in good agreement with those obtained by Ay [M.R. Ay, S. Sarkar, M. Shahriari, D. Sardari, H. Zaidi, Assessment of different computational models for generation of X-ray spectra in diagnostic radiology and mammography, Med. Phys. 32 (2005) 1660], mainly for the bremsstrahlung distribution. Also, it was noted that the total characteristic X-rays produced by the EGSnrc/BEAM MC code increases with the increase of voltage more intensely than with the Xcomp5r, X-raytbc and Srs-78 codes. Comparison between the absorbed dose to each organ or tissue showed that, for X-ray spectra with additional filtration, the code based on Tucker et al. is in agreement with EGSnrc/BEAM MC code. But, for X-ray spectra without additional filtration the code based on Tucker et al. model presented the strong disagreement with EGSnrc/BEAM MC code.     

MCNP模拟钴-60辐照装置空间剂量场

为模拟辐照室中辐照工位外的周围空间剂量场分布,采用蒙特卡罗粒子输运程序MCNP建立钴-60辐照装置模型。以单板源架中心点为坐标原点的笛卡尔坐标系,考虑钴-60源的γ射线非自吸收和自吸收两种情况,研究坐标轴方向上每隔10 cm间距的空气平面的剂量率和坐标轴上剂量率的变化规律。结果表明,辐照室中辐照产品占满辐照工位的情况下,周围空间剂量场空气面剂量率整体较小;单板源架中心坐标轴上的剂量率变化规律更符合二项式拟合函数。在钴-60源γ射线自吸收情况下,单板源架端面坐标轴附近的空气面剂量率明显偏小,且随着空气面远离单板源架,空气面上的高剂量率区域向两侧移动;在钴-60源γ射线非自吸收情况下,单板源架端面处的空气面高剂量区域始终位于坐标轴附近。MCNP理论模拟计算分析对于利用钴-60辐照装置辐照工位外的周围空间剂量场具有重要的实际指导意义。     

On-Board Dose Measurement and its Monte Carlo Analysis in a Low Level Waste Shipping Vessel

Abstract

On-board dose measurements were made in a shipping vessel for low level radioactive wastes, the Seiei Maru. The measured values are much smaller than the regulation values both on the hatch covers and in the accommodation area. The dose equivalent rates on the hatch cover are analysed by using a continuous energy Monte Carlo code, MCNP 4B, with two kinds of calculational models. One is the detailed model with the geometry of containers and LLW drums, and an asymmetrical source distribution. The results of the detailed calculation approached the shape of the measured dose rate distribution graphs. The other is the simplified model that mixes source volume uniformly. The calculated values obtained with the simplified model are twice as large as those calculated with the detailed model.     

Parametric Study of Shielding Codes used for Packaging

Abstract

It is very important to be able to predict the dose rates external to a flask package. Currently in Japan several shielding calculation codes are used to evaluate the dose rate around a package. It is, however, generally appreciated that there are differences between the results obtained when using different calculation codes for the same shielding calculation problems. The differences appear to be particularly important for gamma ray shielding calculations when using the point-kernel method on a package with a multi-layer wall because of the build-up factor. In this paper the calculation accuracy of some codes for gamma ray shielding calculations using the codes QAD and MARMER (a point-kernel method), the ANISN code (a discrete ordinate (Sn) method) and the MCNP code (a Monte Carlo method) are examined and compared using the benchmark problems. The calculation results are then compared with the results from a simulated actual flask package body wall. The results presented here show that the calculation results using MARNER and MCNP (which have recently been introduced into Japan) agree with the experimental measurements. These codes can therefore be used for future gamma shielding calculations and the calculation conditions of those codes in Japan.     

低能注量X射线辐照下圆柱腔内辐射环境及IEMP的模拟研究

通过MCNP程序模拟计算了低能注量X射线辐照圆柱腔外端面并透射进入腔体时,腔内各作用面的综合辐照环境及电子发射参数。结合3维粒子模拟（PIC）程序,对多发射面作用下圆柱腔内电磁场和粒子分布进行了模拟计算,并与仅存在上端面电子发射时的电磁场结果进行对比。结果表明,在实际情况下,圆柱侧壁和下端面会发射大量电子,能将上端面中心处的轴向电场强度增大至仅上端面发射电子时的2倍,而由于此时腔内多发射面作用下电流方向的复杂化,磁场强度则略微减小。同时,比较了由MCNP计算得到的前向散射电子能谱和部分文献采用的近似能谱,分别提供电子发射速度时内电磁脉冲（IEMP）的电场强度波形。结果表明,能谱的改变会对电磁场带来极大的改变,故建议通过相关蒙特卡罗（MC）程序计算IEMP电子发射能谱。     

基于RMC程序的k_(eff)对核数据的敏感性分析

适用于连续能量蒙特卡罗程序的敏感性分析方法是当前的研究热点。本文建立了5种不同反应类型的敏感性系数的计算公式,对当前应用广泛的反复裂变几率法的理论基础及算法进行了分析。分别使用RMC程序和MCNP6程序计算了keff对核数据的敏感性系数,计算结果吻合良好。本文结果表明RMC程序初步具备了敏感性分析的功能。     

Monte Carlo Simulation of the Electron Transport Through Air Slabs: A Comparative Study of penelope, geant3, geant4 and egsnrc Monte Carlo Codes

The Monte Carlo simulation of the electron transport through air slabs is studied with four codes: PENELOPE, GEANT3, Geant4 and EGSnrc. Monoenergetic electron beams with energies 6, 12 and 18 MeV are considered to impinge on air slabs with thicknesses ranging from 10 to 100 cm. The angular and radial distributions of the transmitted electrons are used to make a comparison between the codes. Non-negligible differences are observed in the radial distributions. These differences produce worth effects in the macroscopical dose distribution absorbed in a water phantom situated behind the slabs.     

MCNP simulation of neutron energy spectra for a TN-32 dry shielded container

This paper presents a numerical analysis of neutron energy spectra for a TN-32 spent fuel dry storage cask using Monte Carlo simulation. The analysis results were compared with experimental measurements to determine the suitability of using such codes for neutron flux calculations in soft-spectrum neutron environments. Complete spent fuel compositions were generated using Scale 4.4a. Variations in source definition and geometry determined that geometric and source simplifications in the computational model have negligible effect on final neutron energy distribution. Variations between experimental and computed spectra at energies above 1 MeV and below 100 keV demonstrated the shortfalls of various detection instruments used to collect the experimental neutron energy spectra data principally because these instruments were calibrated based on high neutron energy spectra. The MCNP calculations were generally in agree with the experimental data, but predicted that the detectors would over-respond to the neutron spectra around a spent fuel dry shielded container. Computed neutron energy spectra were always conservative when compared to experimental spectra.