热释光探测器测量n-γ混合辐射场中子剂量

介绍了热释光探测器在n-γ混合辐射场中测量中子剂量的原理及测量方法,选择¹³⁷Cs参考辐射场进行刻度,²⁵²Cf n-γ混合辐射场进行验证。验证结果显示,热释光探测器对n-γ混合辐射场中子剂量测量误差最大为-8.92%。     

基于SOI微剂量实验测量技术的研究现状与展望

对目前微剂量测量中常用的组织等效正比计数器(TEPC)和基于绝缘体上硅(SOI)技术的微剂量计的特点和研究现状进行了分析,对比了二者的优缺点,指出TEPC在微剂量实验测量方面存在的缺陷。重点分析了SOI微剂量计的研究现状,对五代SOI微剂量计的物理设计结构作了详细说明,并对其在中子、质子和重离子微剂量测量方面以及组织等效转换方面的研究现状进行了详细分析,指出该技术目前存在的问题,并在此基础上对其发展前景进行了展望。     

中子气泡探测器用于核测井中子个人剂量监测的研究

为研究国产中子气泡探测器在核测井中子个人剂量监测中的适用性,采用国内研制的中子气泡探测器对核测井运源车外表面、车内兼用储源仓周围等关注点的中子辐射水平进行监测,同时采用进口LB6411型中子周围剂量当量仪进行比对监测。实验结果表明,当运源车兼用储源仓内仅装载中子源时,中子气泡探测器与LB6411的测量结果无显著统计学差异,两者测量结果符合较好,中子气泡探测器的测量结果准确可信;当兼用储源仓分别装载中子源、中子-γ源时,两组中子气泡探测器的测量结果也无显著统计学差异,中子气泡探测器适用于中子-γ混合辐射场中子辐射剂量的测量。中子气泡探测器在运源车现场与在²⁴¹Am-Be源标准中子场中的剂量响应灵敏度因子间的相对偏差为7.4%,验证了其较好的能量响应特性,并显示了在核测井现场条件下用于中子个人剂量监测的适用性。     

液体闪烁体探测器对PuC源中子能谱的测量

目前,正在进行的大亚湾中微子实验中,中国原子能科学研究院承担了与中微子探测器刻度相关的研究工作。其中,238Pu-13C同位素中子源所产生的6．13MeV伽马射线和中子是中微子探测器重要的能量刻度源。为了能对能量刻度工作进行详细的模拟和研究,需要了解PuC源的中子能谱。     

闪烁体耦合SiPM测量γ能谱的模拟研究

硅光电倍增管(Silicon photomultiplier,SiPM)是一种新型的光电探测器件,由工作在盖革模式下的雪崩二极管阵列组成。利用GEANT4蒙特卡罗软件包对LaBr_3:10%Ce~(3+)、Na I(TI)闪烁晶体耦合SiPM测量γ射线能谱进行了细致的模拟,通过单色光LED光源照射SiPM,得到SiPM自身暗电流噪声经电子学放大后,与集成的盖革雪崩二极管之间的间隙引起吸收光子涨落对能谱的展宽。对模拟得到的662 keVγ射线能量分辨率进行修正,最后与实验结果对比能够很好地符合,还得到了一组对应闪烁晶体本征能量分辨率的程序参数Pr,s。结果验证了模拟程序设置的闪烁晶体与封装材料光学参数的合理性与可靠性,为闪烁体探测器设计提供了一套开发工具。     

基于Si-PIN的手部定向剂量当量率实时测量技术研究

研发了一款佩戴在手指上用于实时测量人员手部剂量率和总剂量的皮肤定向剂量当量率仪,实现了对能量范围8~250 keV、剂量率范围1 μSv/h~1 mSv/h的X/γ射线的实时测量,同时还具有超剂量阈值报警功能。剂量率仪包括探头、中继器和主机部分,探头包括Si PIN探测器和电荷灵敏前置放大专用集成电路,主机包括控制电路、显示屏和报警器。通过实验和理论计算研究了基于Monte Carlo模拟的补偿片设计,实现了尺寸为2 cm×1 cm×05 cm的微型探头,能直接佩戴在手指上实现实时测量。在国防科技工业电离辐射一级计量站核工业放射性计量测试中心开展了校准和检验,实验结果表明能量响应、剂量率线性范围满足相关规程的要求。     

用于个人剂量测量的硅探测器信号处理电路的设计

介绍了一种用于个人γ剂量测量的微型硅探测器的信号处理电路的设计,讨论了PCB(印制电路板)设计中应注意的问题。该信号处理电路主要包括前置放大电路、滤波成形电路和极-零相消电路,前置放大电路采用了电荷灵敏前置放大器,滤波成形电路采用了CR-RC滤波成形网络。设计的信号处理电路PCB面积仅有10 cm2,对于0.662 Me V的γ射线,信号处理电路的输出信号的信噪比达到了50:1,输出脉冲幅度达到了1.5 V左右,输出信号之后没有明显的下冲现象,其性能可以满足用于个人剂量测量的要求。     

组织等效正比计数器对单能中子响应的测量及模拟

在辐射防护中,组织等效正比计数器（TEPC）可作为主动式剂量计用于中子辐射场监测。采用研制的球形TEPC对加速器单能中子进行测量,获取了0.36、0.65、0.8、1.0、1.3 MeV各单能中子的微剂量谱,根据测得的线能谱计算了吸收剂量和平均品质因子,进一步得到了剂量当量,并与利用注量-周围剂量当量转化因子计算的结果进行了对比。采用FLUKA蒙特卡罗软件模拟了TEPC对各单能中子的响应,并与实验进行了比较。结果发现,理论模拟和实验测量符合很好。实验与理论结果均表明研制的球形TEPC对上述各单能中子具有较好的剂量当量响应。     

现有雷姆仪对中子周围剂量当量测量适用性的探讨

本文讨论了现有巡测仪测量中子周围剂量当量(Ambient Dose Equivalent)H·(10)的适用性问题,并扼要阐述了微剂量计数器和 Bonner 球计数器在这方面的应用。     

微剂量测量装置和方法的研究

本文介绍了为模拟在棒状生物体中线能 y 分布的测定而设计、制造的圆柱形组织等效壁正比计数管和测试系统的结构及性能。模拟组织的尺寸范围为0.2—6μm,当模拟1.12μm 组织时,y 的可测下限为0.1keV/μm。文中给出了~(60)Co、~(137)Cs、~(226)Ra 三种γ射线和~(241)Am-Be中子、14MeV 中子的微观能量沉积谱和剂量平均线能 y_D 值。本文还介绍了一种用于高线能射线 y 分布测量的不同增益谱的连结技术。     

GEANT4 simulation of slow positron beam implantation profiles

The paper presents the positron implantation profiles, which are important for proper interpretation of data produced in slow-positron depth defect spectroscopy (VEPAS). In the paper, we compared the profiles reported in other publications and those obtained using the GEANT4 codes, which are used for the simulation of interaction of energetic particles with matter. The comparison shows that the GEANT4 codes produce profiles which match fairly well with those generated by other codes, which take into account more accurately processes at low energies when positrons interact with core electrons, valence electrons, plasmons etc. The profiles in different materials simulated for different implant energies were parameterized using two analytical formulas: the Makhovian profile and the profile proposed by Ghosh et al. [V.J. Ghosh, D.O. Welch, K.G. Lynn, in: E. Ottewite, A.H. WeissSlow (Eds.), Positron Beam Techniques for Solids and Surfaces, Jackson Hole, Wyoming, AIP Conference Proceedings, Vol. 303, New York, 1994, p. 37]. The adjustable parameters obtained are presented in Table 1 and Table 2. The total backscattering probability obtained from the GEANT4 simulations is in agreement with experimental data reported.     

GEANT4 simulation of photo-peak efficiency of small high purity germanium detectors for nuclear power plant applications

GEANT4 – based Monte Carlo simulations have been carried out for the determination of photo-peak efficiency of heavily shielded small high purity germanium detector (HPGe) used for monitoring radiation levels in nuclear power plants. The GEANT4 simulated values of HPGe detector efficiency for point as well as for disk sources, for two different values of collimator diameter, have been found in good agreement with the corresponding published results obtained by using the MCNP code. The work has been extended to study the effect of radial displacement of a source relative to a detector on photo-peak efficiency for both point and disk source, and at various values of γ-ray energies. Also the effect of disk source radius on photo-peak efficiency has been studied. Besides the results of different available physics models in GEANT4 have also been compared. The computed values of efficiency for point as well as for disk sources using the Penelope and Livermore physics models have been found correspondingly consistent for various values of γ-ray energies while some differences (e.g., Penelope model yields 6.3% higher values of photo-peak efficiency for Eγ = 1.332 MeV, 10 mm collimator diameter) have been observed in the corresponding valued obtained by using the Standard physics model.     

NaI:TI与LaBr_3:Ce~(3+)闪烁体探测器能量分辨的GEANT4模拟

闪烁体探测器γ能谱测量中,闪烁光子的产生、传输及光电器件收集、倍增均对能谱的分辨有巨大影响。为了精确模拟闪烁体探测器的能量分辨,采用蒙特卡洛模拟软件GEANT4建立了一套模型,分别对Na I(TI),LaBr_3:0.5%Ce~(3+)闪烁体探测器测量能量为662 ke V的γ射线能谱进行模拟,获得的结果与实验测试结果能够很好的符合,验证了建立模型合理性与可靠性,为闪烁体的设计提供了一套更精确的开发工具。     

GEANT4接续计算方法以及在GRH系统优化设计中的应用

遗传算法调用GEANT4优化设计方法应用于气体切伦科夫探测器GRH(Gamma Reaction History)系统设计过程可有效提高探测器指标,然而GEANT4程序计算耗时。本文研究了GEANT4程序的接续计算方法,采用函数拟合方式和文件读入的方式进行了接续计算源描述,采用文件读入的方式接续计算结果与直接计算结果吻合良好,效率相差1.4%,时间谱半高宽相差2.1%。采用接续计算方法使得整体优化时间下降50%以上,提高了计算效率。该接续计算方法也可应用于其他采用GEANT4直接模拟收敛困难的计算问题。     

单球多位置灵敏计数器的中子能量响应合理性初步探索

设计了一种单球3个位置灵敏计数器不同分区方法,借用蒙特卡罗模拟对5个计数区的中子能量响应进行了计算,并分析5个计数区的能量响应合理性.基于能量响应假设实验,利用少道解谱UMG软件包对诸如注量和周围剂量当量的剂量学量进行计算,并对计算的结果进行了讨论.设计了一种单球3个位置灵敏计数器不同分区方法,借用蒙特卡罗模拟对5个计数区的中子能量响应进行了计算,并分析5个计数区的能量响应合理性.基于能量响应假设实验,利用少道解谱UMG软件包对诸如注量和周围剂量当量的剂量学量进行计算,并对计算的结果进行了讨论.     

Optimization of GEANT4 settings for Proton Pencil Beam Scanning simulations using GATE

This study reports the investigation of different GEANT4 settings for proton therapy applications in the context of Treatment Planning System comparisons. The GEANT4.9.2 release was used through the GATE platform. We focused on the Pencil Beam Scanning delivery technique, which allows for intensity modulated proton therapy applications. The most relevant options and parameters (range cut, step size, database binning) for the simulation that influence the dose deposition were investigated, in order to determine a robust, accurate and efficient simulation environment. In this perspective, simulations of depth-dose profiles and transverse profiles at different depths and energies between 100 and 230 MeV have been assessed against reference measurements in water and PMMA. These measurements were performed in Essen, Germany, with the IBA dedicated Pencil Beam Scanning system, using Bragg-peak chambers and radiochromic films. GEANT4 simulations were also compared to the PHITS.2.14 and MCNPX.2.5.0 Monte Carlo codes. Depth-dose simulations reached 0.3 mm range accuracy compared to NIST CSDA ranges, with a dose agreement of about 1% over a set of five different energies. The transverse profiles simulated using the different Monte Carlo codes showed discrepancies, with up to 15% difference in beam widening between GEANT4 and MCNPX in water. A 8% difference between the GEANT4 multiple scattering and single scattering algorithms was observed. The simulations showed the inability of reproducing the measured transverse dose spreading with depth in PMMA, corroborating the fact that GEANT4 underestimates the lateral dose spreading. GATE was found to be a very convenient simulation environment to perform this study. A reference physics-list and an optimized parameters-list have been proposed. Satisfactory agreement against depth-dose profiles measurements was obtained. The simulation of transverse profiles using different Monte Carlo codes showed significant deviations. This point is crucial for Pencil Beam Scanning delivery simulations and suggests that the GEANT4 multiple scattering algorithm should be revised.     

Irradiation performance of polytetrafluoroethylene (Teflon) in a mixed fast neutron and gamma radiation field

Samples of polytetrafluoroethylene have been irradiated with a mixed field of fast neutrons and gamma rays using the MIT Research Reactor. Dose levels from ∼0.3 to ∼50×10⁶ Gy for gamma and from ∼0.13 to 80×10⁴ Gy for fast neutrons were used. Weight loss, fluorine loss, and swelling were measured quantitatively. Subjective mechanical property tests were also performed to assess embrittlement. Aside from high levels of embrittlement, no large changes, ?1.5%, were observed in the properties which were assayed even at the highest doses.     

Comparison among Methods for Evaluating Dose Equivalent in Pulsed and Mixed Radiation Field

In order to ascertain the accuracy and applicability of the method which was developed in the previous paper for evaluating the dose equivalent in a single-burst and mixed radiation field, a comparison with two other ones has been carried out. One is based on the measurement of the neutron energy spectrum with TOF system and the separate measurement of the absorbed doses with twin chambers. In the other, a tissue-equivalent proportional counter is used for measuring the LET distribution of the absorbed dose. The average quality factor has been evaluated in a pulsed and mixed field generated by an electron linear accelerator. A good agreement has been found among these methods.

A problem about the experimental determination of the dose equivalent index has also been discussed. It was pointed out that the average quality factor as a function of the depth in tissue could be regarded as characteristic of the mixed radiation field.     

Method for converting in-situ gamma ray spectra of a portable Ge detector to an incident photon flux energy distribution based on Monte Carlo simulation

A matrix stripping method for the conversion of in-situ gamma ray spectrum, obtained with portable Ge detector, to photon flux energy distribution is proposed. The detector response is fully described by its stripping matrix and full absorption efficiency curve. A charge collection efficiency function is introduced in the simulation to take into account the existence of a transition zone of increasing charge collection after the inactive Ge layer. Good agreement is obtained between simulated and experimental full absorption efficiencies. The characteristic stripping matrix is determined by Monte Carlo simulation for different incident photon energies using the Geant4 toolkit system. The photon flux energy distribution is deduced by stripping the measured spectrum of the partial absorption and cosmic ray events and then applying the full absorption efficiency curve. The stripping method is applied to a measured in-situ spectrum. The value of the absorbed dose rate in air deduced from the corresponding flux energy distribution agrees well with the value measured directly in-situ.     

Calculations of Effective Dose and Ambient Dose Equivalent Conversion Coefficients for High Energy Photons

The conversion coefficients from photon fluence to ambient dose equivalent, H* (10) and effective doses were calculated for photons up to 10 GeV. A Monte Carlo code EGS4 was used for these calculations and secondary particle transports were considered. The calculated ambient dose equivalents were compared to the calculated effective doses. The comparison shows that the ambient dose equivalents at 1 cm depth, H* (10) underestimate the effective doses at the energy above 5MeV. H* (10) is not suitable operational quantity since it does not provide reasonable estimation of effective dose. It is difficult to define the operational quantity which can be consistently used for photons from low energy to high energy above 10 MeV. Instead of operational quantities, the maximum effective dose in various irradiation geometries can be used for shielding design calculations.