基于D-T紧凑型中子源的快中子照相准直屏蔽体设计及中子束特性模拟研究

中子照相是一种重要的无损检测技术,它能用于火工产品、毒品和核燃料元件等的检测。基于紧凑型D-T中子发生器,完成了一个用于快中子照相的准直屏蔽体系统（BSA）的物理设计。根据D-T中子源的能谱和角分布建立了中子源模型,采用MCNP4C蒙特卡罗程序,模拟了准直屏蔽体系统中中子和γ射线的输运,准直中子束相对于单位源中子的中子注量可以达到9.30×10-6 cm-2,准直中子束中主要是能量大于10 MeV的快中子;在设置的样品平面直径14 cm的照射视野范围,准直束中子注量的不均匀度为4.30%,准直束中中子注量与γ注量的比值为17.20,中子通量和中子注量比值J/Φ为0.992,说明准直中子束有好的平行性;准直屏蔽体外的泄露中子注量率与准直束中子注量率相比降低了2个量级。所设计的准直屏蔽体能满足快中子照相的要求。Neutron radiography is an important nondestructive testing technique. It can be used to detect the explosive devices, drug and the nuclear fuel element, etc. A beam-shaping-assembly (BSA) based on a compact D-T neutron generator is designed for fast neutron radiography in this paper. D-T neutron source model is constructed based on the neutron energy spectrum and angular distribution data. The transportation of neutron and γ-ray in the BSA is simulated using MCNP4C code. The neutron fluence of the collimated neutron beam with respect to the neutron source of the unit source is 9.30×10-6 cm-2. The collimated neutron beams is mainly fast neutrons with energies greater than 10 MeV. In the irradiation field range with a diameter of 14 cm, the neutron fluence uniformity of the collimated beam is 4.3%, the ratio of the neutron fluence to the gamma fluence in the collimated beam is 17.20, and the neutron flux and the neutron fluence ratio (J/Φ) is 0.992 which indicates that the collimated neutron beam has good parallelism. The leakage neutron fluence in outside of BSA is two orders of magnitude lower than that of the collimated neutron beam. The designed BSA can meet the need of fast neutron radiography.     

中子散射对快中子照相质量影响的分析

通过一内空的圆柱体模型，对在快中子照相时由样品引起的散射中子强度与样品形状和探测距离之间的关系进行模拟，并用^241Am—Be作中子源对散射中子的影响进行实验验证。结果表明，在快中子照相时，由样品引起的散射中子的强度与探测距离以及样品形状有关。对于同一样品，探测距离增加，散射中子的影响则降低。The relationships between intensities of scattered neutrons by specimens and their shapes and detecting distances have been simulated using a hollow cylinder model, and the results were validated by experiments of fast neutron imaging using ^241Am-Be neutron source. The results showed that the intensities of scattered neutrons are closely related to the detecting distances and sample＇ s shapes. The influences of scattered neutronns in fast neutron imaging will be reduced while detecting distances increased.     

An empirical formula for scattered neutron components in fast neutron radiography

Scattering neutrons are one of the key factors that may affect the images of fast neutron radiography. In this paper, a mathematical model for scattered neutrons is developed on a cylinder sample, and an empirical formula for scattered neutrons is obtained. According to the results given by Monte Carlo methods, the parameters in the empirical formula are obtained with curve fitting, which confirms the logicality of the empirical formula. The curve-fitted parameters of common materials such as ⁶LiD are given.     

PGNAA系统中D-T中子管的慢化装置优化设计

为了提高PGNAA系统中D-T中子管的中子慢化效率,获得更高的热中子产额,借助蒙特卡罗模拟,确定了以铅为中子反射层、5个聚乙烯层和铅层相互交替作为中子慢化层、碳化硼含量为3%的含硼聚乙烯作为中子吸收层以及铅作为γ屏蔽层的中子慢化装置模型。针对中子产额为3×107 n/s 的D-T中子管,该慢化装置输出面低于5 eV中子通量可达5.28×106 n/s,占总中子通量的30.8%,有效提高了中子慢化效率。经过验证模拟结果能够满足实验要求。To improve the moderating efficiency of D-T Neutron Generator in PGNAA system, and get higher thermal neutron yield, the Monte Carlo code MCNP was used to optimize the moderation setup. The lead was selected as neutron reflector and gamma absorber, 5 polyethylene layers and 4 lead layers constituted the neutron moderator and 3% boron-doping polyethylene was selected as neutron absorber. For the yield of 3107 n/s D-T Neutron Generator, this moderation setup can provide a yield of lower than 5 eV of 5.28106 n/s, accounting for 30.8% of total neutron yield, dramatically improves the moderating efficiency. It is proved that the simulation results can satisfy the requirement of PGNAA system by preliminary experimental verification.     

强中子本底辐射屏蔽体优化设计技术研究

在国内首次采用高强度窄脉冲DPF中子源,采用直照法测量零功率堆瞬发中子时间常数α。由于外中子源本底太强,导致直照中子和散射中子产生的干扰信号比测量信号高三个量级。为有效地抑制外本底,针对不同能量的干扰中子,采用不同材料进行屏蔽。通过数值模拟的方法优化辐射屏蔽体设计,在屏蔽中子的同时也对散射γ 进行了有效屏蔽,使测量信噪比达到了7.5:1,并与实验结果相符合,实验中所采用的新型无机晶体也有效抑制了中子本底。The value of prompt neutron multiplication, α, is measured under the condition of using a denser plasma focus(DPF) neutron-source irradiating zero power assembly for the first time in China. The acquired signal is lower three orders of magnitudes than that of the noise caused by direct and scattered neutrons from the extra-high-intensity neutron-source. Using different kinds of material to decrease the noise caused by neutron with different kinds of energies, an optimized design for radiation shielding is developed by the method of numerical simulation to suppress noise signal. Both neutron and γ-ray are shielded simultaneously. The Signal/Noise Ratio (SNR) with the optimal design was up to 7.5:1 and was consistent with the experimental results. The noise of neutron is decreased effectively by the new kind of unorganic crystals used.     

Corrections on energy spectrum and scatterings for fast neutron radiography at NECTAR facility

Distortions caused by the neutron spectrum and scattered neutrons are major problems in fast neutron radiography and should be considered for improving the image quality. This paper puts emphasis on the removal of these image distortions and deviations for fast neutron radiography performed at the NECTAR facility of the research reactor FRM-II in Technische Universit/it Mfinchen （TUM）, Germany. The NECTAR energy spectrum is analyzed and established to modify the influence caused by the neutron spectrum, and the Point Scattered Function （PSeF） simulated by the Monte-Carlo program MCNPX is used to evaluate scattering effects from the object and improve image quality. Good analysis results prove the sound effects of the above two corrections.     

Measurement of the neutron spectrum of a Pu-C source with a liquid scintillator

The neutron response function for a BC501A liquid scintillator （LS） has been measured using a series of monoenergetic neutrons produced by the p-T reaction. The proton energies were chosen such as to produce neutrons in the energy range of 1 to 20 MeV. The principles of the technique of unfolding a neutron energy spectrum by using the measured neutron response function and the measured Pulse Height （PH） spectrum is briefly described. The PH spectrum of neutrons from the Pu-C source, which will be used for the calibration of the reactor antineutrino detectors for the Daya Bay neutrino experiment, was measured and analyzed to get the neutron energy spectrum. Simultaneously the neutron energy spectrum of an Am-Be source was measured and compared with other measurements as a check of the result for the Pu-C source. Finally, an error analysis and a discussion of the results are given.     

γ探测器中子灵敏度标定技术

叙述了γ探测器的14.1 MeV中子灵敏度标定原理和屏蔽方法,利用MCNP程序建模, 优化设计了对γ和中子高屏蔽性能的屏蔽体。 实验研究表明, 其信噪比最高达10∶1, 比测量点本身的信噪比1.35∶1提高了7倍, 从而实现了γ探测器的中子灵敏度标定。 It is introduced that a method and principle for calibrating 14.1 MeV neutron sensitivity of γ ray detector. A shield system for scattering neutrons and γ rays has been optimized by MCNP code. The experimental results show that the signal to noise ratio of the system is about 10:1, 7 times higher than the value of 1.35〖KG-*4〗∶〖KG-*4〗1 without shield system. Calibration of neutron sensitivity of γ ray detector is then accomplished.     

中子源有源法核查技术研究

用有源 (主动 )方法研究了贫化铀组合系统的中子诱发裂变缓发中子探测技术 .在不同屏蔽和组合等条件下测量和比较了贫化铀系统的缓发裂变中子分布 ,进一步研究了实验系统的可核查性 .探讨了区分核与非核系统的方法. The technique for detecting the delayed neutrons from neutron induced fission in uranium systems was studied by using an active method with 3He proportional counting tube array and a 14 MeV D T neutron source. Under the conditions of different shielding and combination, the distributions of delayed fission neutrons from depleted uranium systems were measured and the reliability of the systems was studied. The method to distinguish a nuclear system from a non nuclear one was discussed.     

低能α粒子与中子耦合输运模拟

开发了低能α粒子与中子耦合输运程序，可以计算低能α粒子在各种介质中的输运、含氧介质中的（α，n）中子产额，跟踪次级中子。采用蒙特卡罗方法中带权重跟踪的技巧，通过大量小权重事件的模拟成功解决了低能（α，n）反应中子产额太低、无法直接模拟的困难。α粒子的射程与阻止本领用SRIM程序计算，（α，n）截面取自最新的JENDL带电粒子库，次级中子跟踪使用MCNP程序。A new code package of alpha-neutron coupled transportation is developed. It can be used to simulate the transportation of alpha particle in media, to compute the neutron yield in media contains oxygen and to track the history of the secondary neutrons. The secondary neutrons are simulated efficiently by tracking lots of neutrons with small weight, which is determined by the alpha-neutron yield. The stopping power and range of alpha particle in media are given by SRIM code, the alpha-neutron cross section is from charged particle library in JENDL, and the neutrons are treated by MCNP code.     

Neutron collimator design of neutron radiography based on the BNCT facility

For the research of CCD neutron radiography,a neutron collimator was designed based on the exit of thermal neutron of the Boron Neutron Capture Therapy(BNCT)reactor.Based on the Geant4 simulations,the preliminary choice of the size of the collimator was determined.The materials were selected according to the literature data.Then,a collimator was constructed and tested on site.The results of experiment and simulation show that the thermal neutron flux at the end of the neutron collimator is greater than 1.0×106n/cm2/s,the maximum collimation ratio(L/D)is 58,the Cd-ratio(Mn)is 160 and the diameter of collimator end is 10 cm.This neutron collimator is considered to be applicable for neutron radiography.     

Study on spatial resolution of micromegas as a neutron detector under condition of high neutron flux and γ ray background

In this paper Micromegas has been designed to detect neutrons. The simulation of the spatial resolution of Micromegas as neutron detector is carried out by GEANT4 toolkit. The neutron track reconstruction method based on the time coincidence technology is employed in the present work. The influence of the flux of incident 14 MeV neutron and high gamma background on the spatial resolution is carefully studied. Our results show that the spatial resolution of the detector is sensitive to the neutron flux, but insensitive to the intensity of γ background if the neutron track reconstruction method proposed by our group is used. The γ insensitivity makes it possible for us to use the Micromegas detector under condition which has high γ-rays background.     

Calculation of prompt fission neutron spectrum for 233U（n,f） reaction by the semi-empirical method

The prompt fission neutron spectra for the neutron-induced fission of 233U for low energy neutrons （below 6 MeV） are calculated using nuclear evaporation theory with a semi-empirical method, in which the partition of the total excitation energy between the fission fragments for the nth＋233U fission reactions is determined by the available experimental and evaluation data. The calculated prompt fission neutron spectra agree well with the experimental data. The proportions of high-energy neutrons of prompt fission neutron spectrum versus incident neutron energies are investigated with the theoretical spectra, and the results are consistent with the systematics. The semi-empirical method could be a useful tool for the prompt evaluation of fission neutron spectra.     

Measurement of the response function and the detection effciency of an organic liquid scintillator for neutrons between 1 and 30 MeV

The light output function of a φ50.8 mm×50.8 mm BC501A scintillation detector was measured in the neutron energy region of 1 to 30 MeV by fitting the pulse height （PH） spectra for neutrons with the simulations from the NRESP code at the edge range. Using the new light output function, the neutron detection effciency was determined with two Monte-Carlo codes, NEFF and SCINFUL. The calculated effciency was corrected by comparing the simulated PH spectra with the measured ones. The determined effciency was verified at the near threshold region and normalized with a Proton-Recoil-Telescope （PRT） at the 8-14 MeV energy region.     

Neutron beam optimization based on a ^7Li（p,n）^7Be reaction for treatment of deep-seated brain tumors by BNCT

Neutron beam optimization for accelerator-based Boron Neutron Capture Therapy（BNCT） is investigated using a ^7Li（p,n）^7Be reaction. Design and optimization have been carried out for the target, cooling system,moderator, filter, reflector, and collimator to achieve a high flux of epithermal neutron and satisfy the IAEA criteria.Also, the performance of the designed beam in tissue is assessed by using a simulated Snyder head phantom. The results show that the optimization of the collimator and reflector is critical to finding the best neutron beam based on the ^7Li（p,n）^7Be reaction. Our designed beam has 2.49×109n/cm^2 s epithermal neutron flux and is suitable for BNCT of deep-seated brain tumors.     

中子、γ射线在稀土-高分子材料中的输运(英文)

为研究新型复合屏蔽材料的最佳厚度与各种成分最佳配比, 用MCNP计算了中子、 γ射线在稀土 高分子与重金属复合材料中的通量。 对中子、 γ射线在屏蔽体中变化规律进行了深入探索, 同传统复合屏蔽材料的屏蔽性能进行了对比。 结果表明, 中子和γ射线通过屏蔽体时, 其强度遵循指数衰减规律。 新型屏蔽材料对中子的屏蔽效果均优于铅硼聚乙烯, 对γ射线的屏蔽效果均劣于W Ni合金, 且并非稀土含量越高, 材料对中子辐射屏蔽能力越强。 A series of shielding analyses have been performed to estimate the material composition and optimum thickness required for a new radiation shield with various rare earth doped polymer and heavy metal mixtures. The neutron and γ photon fluxes have been calculated by Monte Carlo N Particle(MCNP) transport code. The results indicate that the relative fluxes of γ photon and neutron in both traditional and new composite materials follow an exponential decay rule with the distance of penetration. It can be seen that the composite material consisting of rare earth doped polymer and heavy metal has stronger neutron shielding performance than lead boron polyethylene, but weaker γ shielding effectiveness than W Ni alloy. It is also found that materials with more components of rare earth elements don’t always provide better neutron shielding performance.     

基于Micromegas探测技术的中子CT可行性研究(英文)

在培育新品种过程中能够实时掌握了解土壤中植物根系的形态以及其生长情况将为植物学家提供很多不可或缺的信息。对基于 micromegas 探测技术的中子 CT 在这一新应用的试验环境进行了仿真模拟, 通过三维图像重建后,得到了令人满意的预期结果。首先通过 Geant4 和 Garfield 模拟计算利用聚乙烯薄膜作为中子转换层的 micromegas 中子探测器, 得到了非常理想的位置分辨, 说明基于micromegas 探测技术建立中子 CT 照相系统的可行性。 然后利用MCNP仿真模拟14 MeV 中子CT 的实验环境, 最后由Matlab程序进行图像重建。In this paper, the possibility using micromegas(Micro Mesh Gaseous Structure) as neutron detector in 14 MeV neutron computed tomography(CT) has been simulated. The results show that the micromegas neutron detector has high spatial resolution and is a good candidate for neutron radiography. The three dimensional images of plant roots in soil are successfully and clearly obtained by the 14 MeV neutron CT with micromegas as a neutron detector. In the present simulation, MCNP is employed for 14 MeV neutron transport in the sample and Matlab for the 3 D photograph reconstruction.     

Influence of Hyperon on 1S0 Superfluidity of Nucleons in Neutron Star Matter

We investigate the property of ^1S0 superfluidity of neutrons and protons in the neutron star matter. On the basis of the result, we study the effects of hyperons on the ^1S0 pairing gaps of the two species of the particles. The parameter sets we use are for the Hartree approximation of the relativistic σ - ω model and the mean field approximation of the Walecka model, respectively. The predicted domain of superfluidity is very close to other works, whereas differences appear in the predicted value of the maximum gap. It is found that ∑^-, A and ∑^- have some influences, the scales of which depend on hyperon-meson coupling constants we use besides other factors, on the ^1S0 superfluidity of protons in some density range, and do not have influence on superfluidity of neutrons. Other hyperons have no influence on the ^1S0 superfluidity of neutron and proton due to these hyperons appearing after ^1S0 neutron and proton pairs disappear.     

2.8MeV中子诱发238U裂变的瞬发中子谱高能段实验测量

在中国原子能科学研究院的高压倍加器上利用中子飞行时间方法测量了2.8 MeV中子引起238U裂变的瞬发中子能谱,通过增大样品量和设计合适的屏蔽体提高了效应/本底比,使得测量数据的不确定度满足预期目标（在5.5～14 MeV能区内,能量间隔为0.5 MeV条件下能谱的不确定度小于10%）。将实验几何、中子源能量分布及角分布、探测效率、束流的时间结构等实验条件输入到MCNP程序里,模拟了出射的中子谱,模拟结果与测量结果在不确定度范围内一致,验证了在入射中子能量较低时238>U的裂变瞬发中子谱评价数据的可靠性。The Prompt Fission Neutron Spectrum (PFNS) of 238U induced by 2.8 MeV neutron was measured using the Cockcroft-Walton accelerator in China Institute of Atomic Energy (CIAE). The signal/background ratio was improved by increasing the amount of sample mass and using an appropriate shielding system. The final uncertainty of neutron energy spectrum in 5.5～14 MeV region is less than 10% with a bin size of 0.5 MeV which has reached this project's anticipation. The experimental geometry, the angular distribution and energy distribution of neutron source, the detection efficiency and time structure of deuteron beam were inputted into the MCNP code to simulate the outgoing neutron spectrum. The simulated results agree with the experimental ones within the uncertainty. The result indicates that the evaluated PFNS for 238U at low neutron energy is reliable.     

Characteristics of the sampling paddle modules of neutron wall from the cosmic ray test

In the construction of a large area neutron detector （neutron wall） that is used to detect neutrons at GeV energies, the performances of all the sampling paddle modules prepared for the neutron wall are investigated with a specially designed test bench. Tested by cosmic rays, an average intrinsic time resolution of 222.5 ps is achieved at the center of the modules. The light attenuation length and the effective speed of the light in the module are also investigated.