~3He正比计数器探测效率模拟及灵敏度刻度

利用Monte Carlo程序研究了3He正比计数器慢化体厚度与探测效率的关系,单能中子及Hess谱中子模拟给出5 cm的聚乙烯慢化体优化厚度。利用天然中子本底对探测器装置进行测试,并对探测器灵敏度进行了刻度。     

Calculation of photon shielding properties for some neutron shielding materials

The objective of the present study is to calculate photon shielding parameters for seven polyethylene-based neutron shielding materials. The parameters include the effective atomic number(Z_(eff)), the effective electron density(N_(eff)) for photon interaction and photon energy absorption,and gamma-ray kerma coefficient(kc). The calculations of Z_(eff)are presented as a single-valued and are energy dependent. While Z_(eff)values were calculated via simplistic powerlaw method, the energy-dependent Z_(eff)for photon interaction(Z_(PI-eff)) and photon energy absorption(Z_(PEA-eff)) are obtained via the direct method for energy ranges of 1 keV–100 GeV and 1 keV–20 Me V, respectively. The kccoefficients are calculated by summing the contributions of the major partial photon interactions for energy range of 1 keV–100 MeV. In most cases, data are presented relative to pure polyethylene to allow direct comparison over a range of energy. The results show that combination of polyethylene with other elements such as lithium and aluminum leads to neutron shielding material with more ability to absorb neutron and crays. Also, the kerma coefficient first increases with Z of the additive element at low photon energies and then converges with pure polyethylene at energies greater than 100 keV.     

Gamma-ray shielding and structural properties of PbO-SiO2 glasses

Gamma-ray attenuation coefficients have been determined experimentally using a narrow beam transmission method for the xPbO(1−x)SiO₂ (x = 0.45-0.70) glass system at 662, 1173 and 1332 keV photon energies. These values have also been obtained theoretically using the ‘mixture rule’ and the ‘XCOM’ computer software. The results have been used to calculate half value layer parameters. Gamma-ray shielding properties of PbO-SiO₂ glass samples have been compared with standard radiation shielding concretes. The molar volume, FTIR and acoustic investigations have been used to study the structural properties of the prepared glass system.     

不同材料中子反射与屏蔽效应研究

利用Monte Carlo粒子输运计算程序Super MC对厚度1-5 cm的多种材料进行中子反射和屏蔽性能分析计算。这些材料包括金属材料铍、铅、铜、含硼钢以及~(238)U和非金属材料聚乙烯、氢化锂、混凝土,中子能段选取10~(-5) e V-20 MeV。结果显示,中子反射能力和屏蔽性能都会随着材料厚度而增加,但增加的幅度逐渐减小。铍和聚乙烯在中子反射和屏蔽方面性能优越,而常用来屏蔽γ射线的铅在这两方面性能都是8种材料中最差的。~(238)U只在材料厚度很小时性能卓著,随着材料厚度增加,其性能便远不如大部分材料。考虑到聚乙烯的力学性能较差,在屏蔽材料的选择上有很大的限制,所以在8种材料中,铍的综合性能相对较好。     

中子辐射屏蔽材料PVA/PEO水凝胶的制备及其作用研究

为研究一种新型中子辐射屏蔽材料水凝胶的制备及其对中子辐射的防护作用,应用物理交联法制备不同厚度的单纯和含有金属离子的PVA/PEO水凝胶;利用基于Monte Carlo模拟的SHIELD程序计算不同组分水凝胶对中子输运的影响,以期在理论上证实PVA/PEO水凝胶材料对2.45MeV中子辐射的屏蔽作用;采用BF3中子辐射探测器测量了K-400型高压倍加器发射的2.45MeV中子经过不同水凝胶后的中子通量变化。模拟计算结果显示,随着水凝胶厚度的增加,中子通量和能量逐渐减少;与单纯组比较,相同厚度含金属组中子数和能量减少更明显。BF3探测器测量结果显示,厚度为6—10cm的含金属组的中子通量计数减少的百分率显著高于单纯水凝胶组,辐射屏蔽效率与水凝胶厚度符合线性方程y=-4.51x+86.23,10m厚的含金属离子水凝胶中子通量计数的百分率可减低61.3?。结果表明,高分子聚合物PVA/PEO水凝胶对快中子辐射具有良好的屏蔽作用,含金属组的中子屏蔽效果明显优于单纯组。     

高能~(56)Fe离子入射屏蔽材料的次级粒子模拟分析

空间辐射环境下的高能重离子入射屏蔽材料会产生大量次级粒子,为研究屏蔽材料产生的次级粒子对太空舱内辐射环境的影响,本文使用基于蒙特卡罗方法的Geant4软件模拟空间高能~(56)Fe离子入射铝、碳、聚乙烯、水4种屏蔽材料,分析透射屏蔽体的初级粒子及由屏蔽材料产生的次级电子、次级中子、次级质子和次级γ的能谱以及水吸收体中的能量沉积和深度剂量分布。分析产生的次级重粒子类型和能量,比较4种屏蔽材料对高能Fe离子的屏蔽性能。结果表明,聚乙烯材料对高能重离子的屏蔽性能最好,但同时产生的次级重粒子的能量最大,约为铝材料产生次级重粒子能量的4倍。屏蔽体产生所有次级粒子中,次级质子和原子序数为22-26的次级重粒子贡献最大。     

高纯锗谱仪探测效率的实验刻度与MC模拟

利用低本底高纯锗谱仪测量了标准源(~(133)Ba、~(137)Cs和~(60)Co)的各特征γ射线,得到了各特征能量下的探测效率,并通过拟合给出了高能端的探测效率曲线。同时,利用MC模拟工具包Geant4,模拟了各单能γ射线在高纯锗探测器组件中的输运过程,得到了不同能量下的模拟探测效率。比对结果发现:实验值与模拟值能很好地符合,可为开展相关产品的设计和制造提供参考。     

Simulation and calibration of the response function of multi-sphere neutron spectrometer

In order to realize the on-line real-time measurement of neutron specmnn of ITER fusion, this paper presents a multi-sphere spectrometer system which consists of eight thermal neutron detectors, namely SP9 3He proportional counter, embedded in eight different diameter polyethylene spheres. The response function of eight polyethylene spheres of rnulti-sphere neutron spectrometer was calculated after the simulation of the neutron transport processes in multi-sphere spectrometer by adopting software Geant4. The peak of the response function is in the low energy region for smaller diameter polyethylene sphere. As the polyethylene sphere diameter increased, the peak of the response function moves to the high energy region. The experimental calibration adopts 241Am-Be neutron source. The relative error between normalized data of experiment 47t solid angle counts and normalized data of simulated detection efficiency of 4in to 8in polyethylene sphere is from 1.152% to 12.222%.The experimental results verify the response function of the simulation. All these results provide a theoretical and experimental basis for solving the on-line real-time neutron spectrum of ITER fusion.     

The behavior of HfB2 under neutron irradiation

Due to the interesting properties of hafnium diboride(HfB_2) as a ceramic, it has drawn considerabl attention from several researchers. To understand th radiation properties of HfB_2 that may be used in the nuclea industry, the interaction of this composite under neutro irradiation was studied. It is obvious that, because HfB_2 ha Boron, this composite has the potential to absorb neutron particularly in the range of thermal energy. The ~(10)B(n, a7 Li interaction is dominant among other interactions tha produce alpha particles. The results revealed that alph particles have a high fluctuation diagram following a saw tooth spectrum for thermal neutron energy, which has to b carefully analyzed. If HfB_2 is intended to be used in fusio facilities, for example in fusion reactors, its interactio with high neutrons(14 Me V) should be studied. Th results of this case showed that, in the wide range of alph energy, the amount of alpha particles is almost constant, s this continuum spectrum is almost flat and there is just small peak at 2.31 Me V that belongs to the very famou interaction [~(10)B(n, a)~7Li].     

Verification of a fusion neutron diagnostic Bonner sphere spectrometer on measurement of a ~(241)Am–Be neutron source

A Bonner sphere spectrometer(BSS) was developed for neutron diagnostic on HL_2A Tokamak.It contains eight polyethylene spheres embedded with SP9~3 He proportional counter.Before setting up on the Tokamak experimental hall,a verification experiment was arranged on a~(241)Am–Be neutron source to test its spectrometry capability.The neutron response functions were calculated by Monte Carlo code Geant4 to simulate the real measurement environments.By least square method,the neutron spectrum was finally unfolded on log domain from0.1 e V to 11 Me V.It has a remarkable consistency to the ISO 8529-1 standard~(241)Am–Be neutron spectrum.This shows that the BSS is effective and reliable for neutron spectrum determination.     

Monte-Carlo simulation to determine detector efficiency of plastic scintillating fiber

1 Introduction The plastic scintillating fiber (PSF) has many applications in different fields such as particle trackingin high-energy particle physics, calorimeters, etc., buusing PSF as an imaging detector is the main goal othis research. Therefore, we selected 10~400 keVmono-energy electromagnetic radiation that is used inmedical and non-medical imaging (e.g. non destructive testing). Some scintillation characteristics of plastic scintillating optic fiber have been discover…     

On the neutron diffraction in crystals in the field of a standing sound wave

The diffraction of neutrons is considered in crystals under the influence of a standing sound wave. The scattering probability is calculated for the elastic neutron-crystal interaction, whereas the neutron-standing sound wave interaction can be either elastic and inelastic. The possibility of short-wave (high-energy) neutrons diffraction is illustrated. It is shown that the Debye-Waller factor can be changed and tuned. The analysis of conservation laws are adduced both for thermal and short-wave neutrons. The formation of a “sublattice” is shown in the process of neutrons elastic diffraction with respect to standing sound wave. The analogous to the Kapitza-Dirac effect is considered for neutrons.The problem is solved within the frame of non-stationary S-matrix theory, where the neutron-phonon interaction is described by the Fermi pseudopotential, which is considered as a perturbation.     

闪烁光纤在快中子辐照下部分特性的蒙特卡罗模拟

采用蒙特卡罗方法对闪烁光纤在快中子辐照下的特性进行了研究。利用Geant4模拟计算得到了快中子照射下,闪烁光纤中的能量沉积效率与光纤长度、半径及入射中子能量的变化曲线,同时对量子效率随上述参数的变化关系进行了模拟分析。研究结果表明选择合适尺寸的闪烁光纤和入射中子能量对于快中子成像技术具有重要的意义,并为实际测量工作中的参数选择提供了的理论指导。     

Neutron response functions and detection efficiency of a spherical proton recoil proportional counter

<正>The neutron response function and detection efficiency of a spherical proton recoil proportional counter (SP) play key roles in precise measurement of neutron spectra of the interior materials.In this paper,the response functions and detection efficiency of three SPs developed at CAEP are simulated by Geant4.The simulated spectra are compared with pulse-height spectra measured at 0.165,0.575,1.4,and 14.1 MeV of incident neutrons.And the calculated detector efficiencies agree within 5%with the data obtained by neutron activation.     

Stopping power and ranges of electrons, protons and alpha particles in liquid water using the Geant4-DNA package

This paper presents stopping power and ranges of electrons, protons, and alpha particles in liquid water, calculated using the latest Geant4-DNA processes implemented in the Geant4 Monte Carlo simulation toolkit. Inelastic cross sections are obtained using the first Born approximation and semi-empirical formulas like Rudd’s model for ionisation and the Miller and Green formula for excitation. Elastic collisions and vibrational excitations are considered for tracking electrons until complete thermalisation (0.025 eV). A speed scaling procedure with an effective charge screening term was used to compute alpha particle and heavy ion cross sections. Geant4-DNA simulations were carried out using thin liquid water volumes to determine the linear energy loss (dE/dX), while larger volumes were used to obtain the particle range. While results converge for highly energetic particles, differences are observed for low energies when the applied theoretical models begin to diverge from each other. Results show a good agreement between the analytical calculations obtained from the models, the Geant4-DNA Monte Carlo simulation predictions and the data published in the ICRU reports. Geant4-DNA processes apply to the following energy ranges: 0.025 eV-1 MeV for electrons, 100 eV-100 MeV for protons and 1 keV-400 MeV for alpha particles in liquid water, however since experimental data for very low energies is scarce and very difficult to obtain these processes could not be thoroughly validated so they are recommended for energies above 1 eV for electrons, 1 keV for protons and 10 keV for alpha particles. Relativistic, highly charged ions were implemented in our own “house” version of the code and will be available in future releases of Geant4.     

Comparative study of silicate glasses containing Bi2O3, PbO and BaO: Radiation shielding and optical properties

The radiation shielding and optical properties of xBi₂O₃:(100-x)SiO₂, xPbO:(100-x)SiO₂ and xBaO:(100-x)SiO₂ glass systems (where 30 ? x ? 70 is the composition by weight%) have been investigated. Total mass attenuation coefficients (μ_m) of glasses at 662 keV were improved by increasing their Bi₂O₃ and PbO content, which raised the photoelectric absorption in glass matrices. Raising the BaO content to the same fraction range, however, brought no significant change to μ_m. These results indicate that photon is strongly attenuated in Bi₂O₃ and PbO containing glasses, and but not in BaO containing glass. The results from the optical absorption spectra show an edge that was not sharply defined; clearly indicating the amorphous nature of glass samples. It is observed that the cutoff wavelength for Bi₂O₃ containing glass was longer than PbO and BaO containing glasses.     

Development of paraffin and paraffin/bitumen composites with additions of B2O3 for thermal neutron shielding applications

In this work, paraffin and paraffin/bitumen composites with additions of boron oxide (B₂O₃) were prepared to evaluate the viscosity, flexural, and thermal neutron shielding properties for uses as thermal neutron shielding materials. The results showed that the addition of 3 wt% or 9 wt% bitumen to paraffin increased the overall flexural properties with the content of 9 wt% bitumen having the highest values. The improvement in flexural properties made the composites less brittle, stiffer, and longer-lasting. Furthermore, different contents of B₂O₃ (0, 7, 14, 21, 28, and 35 wt%) were added to paraffin and paraffin/bitumen composites to investigate the effects of the B₂O₃ contents. The results indicated that an increase in B₂O₃ contents improved the shielding properties but slightly reduced the flexural properties. Specifically for 5-mm paraffin and 5-mm paraffin/bitumen samples with 35 wt% of B₂O₃, both samples could reduce neutron flux by more than 70%. The overall results suggested that the paraffin and paraffin/bitumen composites with additions of B₂O₃ showed improved properties for utilization as effective thermal neutron shielding materials.     

Determination and calculation of gamma and neutron shielding characteristics of concretes containing different hematite proportions

To obtain gamma and neutron attenuation characteristics experimental measurements and Monte Carlo simulations were carried out for hematite-loaded concrete samples. Also, the mechanical strength of samples was studied. However, the results for neutron attenuation reveal that there is no effect of hematite inclusion in concrete with respect to the neutron absorption capability. Gamma-ray attenuation capability and the mechanical strength both increased with increasing the hematite percent.     

Effects of neutron spectrum and external neutron source on neutron multiplication parameters in accelerator-driven system

The neutron multiplication parameters: neutron multiplication M, subcritical multiplication factor k_s, external source efficiency φ^*, play an important role for numerical assessment and reactor power evaluation of an accelerator-driven system (ADS). Those parameters can be evaluated by using the measured reaction rate distribution in the subcritical system. In this study, the experimental verification of this methodology is performed in various ADS cores; with high-energy (100 MeV) proton–tungsten source in hard and soft neutron spectra cores and 14 MeV D–T neutron source in soft spectrum core. The comparison between measured and calculated multiplication parameters reveals a maximum relative difference in the range of 6.6–13.7% that is attributed to the calculation nuclear libraries uncertainty and accuracy for energies higher than 20 MeV and also dependent on the reaction rate distribution position and count rates. The effects of different core neutron spectra and external neutron sources on the neutron multiplication parameters are discussed.     

Shielding design and analyses of KIPT neutron source facility

Argonne National Laboratory (ANL) of USA and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the conceptual design development of a neutron source facility. An electron accelerator drives a sub-critical facility (ADS) is used for generating the neutron source. The facility will be utilized for performing basic and applied nuclear researches, producing medical isotopes, and training young nuclear specialists. Monte Carlo code MCNPX has been utilized as the major design tool for the design, due to its capability to transport electrons, photons, and neutrons at high energies. However the ADS shielding calculations with MCNPX need enormous computational resources and the small neutron yield per electron makes sampling difficulty for the Monte Carlo calculations. The high energy electrons (E > 100 MeV) generate very high energy neutrons and these neutrons dominant the total radiation dose outside the shield. The radiation dose caused by high energy neutrons is ∼3-4 orders of magnitude higher than that of the photons. However, the high energy neutron fraction within the total generated neutrons is very small, which increases the sampling difficulty and the required computational time. To solve these difficulties, the user subroutines of MCNPX are utilized to generate a neutron source file, which record the generated neutrons from the photonuclear reactions caused by electrons. This neutron source file is utilized many times in the following MCNPX calculations for weight windows (importance function) generation and radiation dose calculations. In addition, the neutron source file can be sampled multiple times to improve the statistics of the calculated results. In this way the expensive electron transport calculations can be performed once with good statistics for the different ADS shielding problems. This paper presents the method of generating and utilizing the neutron source file by MCNPX for the ADS shielding calculation and similar accelerator facilities, and the accurate radiation dose analyses outside the shield using modest computational resources.