一种宽能谱多球中子谱仪的研制

利用蒙特卡罗程序MCNPX模拟计算了纯聚乙烯球和加入辅助材料的聚乙烯球对不同能量中子的响应函数曲线,使用计算出来的响应函数作为U-M-G软件解谱所需输入文件。研发了一套专门为此多球谱仪进行数据采集的放大甄别一体化电路,该电路可为SP9管提供900 V的工作高压,甄别阈设为0.5 V,总的放大倍数为200倍。使用研制的Bonner球谱仪对已知源强的Pu-Be中子源进行能谱测量,测量结果显示解出的能谱数据与实际Pu-Be源中子能谱较为符合,实验结果验证了该套多球谱仪可用于测量Pu-Be能区的中子谱。Neutron response of Bonner spheres which include pure polyethylene and polyethylene with auxiliary material was calculated with Monte Carlo code MCNPx, the calculated response was used as the input le of U-M-G code for neutronspectrum unfolding. A special screening of ampli cation integrated circuits was developed, which can provides high voltage 900 V working for the SP9 tube, the screening threshold is set to 0.5 V and total magni cation is 200. Neutron energy spectrum of a Pu-Be source were measured with the developed Bonner spheres spectrometer,good agreement was found in the measured result of the spectrum datasolutionand the real spectrum, which indicated that the multi-sphere spectrometer was reliable in the neutron measurement at energy region of Pu-Be neutron source.     

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建立了6Li D转换器中14Me V中子源强的计算模型,对转换器不带辐照样品和分别带2、3、4个辐照样品时的中子源强进行了计算,对转换器产生的中子和来自于堆芯的中子在样品内的能谱和中子注量率进行了计算。结果表明,辐照管内充水和氦气时,辐照样品内由转换器产生的能量大于13Me V的中子分别占能量在1Me V以上中子的25.7%、24.6%,辐照样品内由堆芯产生的能量大于13Me V的中子仅占能量在1Me V以上中子的10·5左右,样品内14Me V中子源强分别可达4.31×1013nT·s·1、3.34×1013 nT·s·1;中子注量率分别可达2.66×1010nT·cm·2·s·1、3.53×1010nT·cm·2·s·1。     

一个宽范围能响平坦的长中子计数器

由于对γ射线灵敏度低，而且可在很宽的范围内中子能量响应比较平坦，长计数管在中子产额的测量中得到了广泛的应用。为了提高探测效率，一般用BF3或^3He正比管外包围一定厚度的石蜡或聚乙烯慢化体来构成长计数管探测系统。建立的长计数器主要是针对中子管产生的DD（2．4MeV）或DT（14．1MeV）或两种能量混合的脉冲中子进行测量。为了达到探测系统设计要求，首先详细模拟了慢化体尺寸及结构对探测效率的影响，以便对探测器系统的加工制作提供依据。根据模拟结果建立了探测器，从实验上对探测器的性能进行了测量。     

用MCNPx程序计算宽能谱中子雷姆仪的响应曲线

利用MCNPx程序计算了宽能谱中子雷姆仪的响应曲线。计算表明,增加铅层对低能中子的响应没有明显的影响,但在高能区(几百MeV以上)宽能谱中子雷姆仪的响应与铅层的厚度有关。铅层厚度为0．6cm时响应比普通雷姆仪提高约3倍,当铅层厚度增加到1．2cm时响应高约5倍。虽然计算结果与ICRP建议书中的H^*(10)曲线相比还有一定的差别,但改变慢化体的结构对提高高能中子的探测效率是有明显效果的。The responses of two extended neutron rein counters as function of neutron energy was ca1cuated Monte Carlo code,MCNPx,was applied in the calculations．Isotropic neutron incidence was employed．The results show that the neutron response is independent on thickness of lead 1aver at low energies,and it is clearly increased with thickness of lead layer at high energies．Although tt1e shape of the response curve does not completely agree with the H (10) curve in ICRP 74 report ,the resuhs obtained give good bases for the practical use of the new instrument in high—energy neutron fields     

介质型脉冲高能中子探测器

首次报道一种基于反冲质子法和高速带电粒子在物质表面引起二次电子发射的高能中子探测方法,研制成功的新一代强流脉冲高能中子探测器,即介质型脉冲高能中子探测器.与传统的强流脉冲中子探测器相比,该探测器对中子的探测及信号的传输过程是在介质中完成的,因而不需要真空和高压也可以正常工作,并具有多项优异的物理性能.该探测器的输出信号来自于高能中子在聚乙烯转换靶产生的高能反冲质子和这些质子穿越电荷收集极表面时产生的部分二次电子.选择适当的卡阈吸收片和电荷收集极材料,二次电子在探测灵敏度中的份额及其随能量的变化可由实验测定.     

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.     

核爆发电问题中核燃料循环的中子学研究

研究核爆聚变电站中的一类中子学问题——如何利用核爆炸产生的大量中子生产核燃料.首先,根据核爆中子场的特点,确定可以利用的中子能量范围.用慢化理论对慢化材料的厚度进行估计,并用MCNP程序进行数值计算.研究如何利用中子反照效应减少造材料层的厚度,最后提出一个较合理的造材料技术路线.     

GaAs光电阴极光谱响应曲线形状的变化

利用光谱响应测试仪测试了反射式GaAs光电阴极在激活过程中以及激活后衰减过程中的光谱响应曲线,测试结果显示在这两个过程中光谱响应曲线形状都在不断发生变化。在激活过程中随着GaAs表面双偶极层的形成,阴极表面有效电子亲和势不断降低,光谱响应则不断提高,但长波响应提高得更快。在激活结束后,位于激活室中受白光照射的GaAs光电阴极由于Cs的脱附影响了双偶极层结构,阴极表面有效电子亲和势不断升高,光谱响应则不断下降,但长波响应下降得更快。上述现象无法用常用的反射式阴极量子效率公式进行解释,它们与阴极高能光电子的逸出有关。由于反射式阴极发射电子能量分布随着入射光子能量的升高而向高能端偏移,同时阴极表面势垒形状的变化对低能电子比对高能电子的影响更大,从而导致了光谱响应曲线形状的变化。     

煤质检测系统D-T中子源屏蔽的MC模拟研究

构建了煤质检测系统D-T中子源采用的3层屏蔽结构:第1层采用铁、不锈钢、钨等不同慢化材料,第2层采用含不同碳化硼质量分数的聚乙烯,第3层采用铅.利用MCNP5程序计算了不同屏蔽结构的中子透射率和中子占比.模拟结果表明:第1层为25cm厚钨,第2层采用含碳化硼质量分数为20%的15cm厚聚乙烯材料,第3层采用5cm厚铅,屏蔽效果最佳.     

HL-2A中子相机屏蔽系统模拟

利用蒙特卡罗模拟程序,建立了HL-2A中子相机蒙特卡罗粒子输运(MCNP)物理模型,对D-D聚变中子和γ射线的屏蔽进行了模拟计算。对石蜡碳酸锂混合物、聚乙烯、铅和316L不锈钢4种常用中子慢化吸,收剂组成的屏蔽层材料的屏蔽效果进行了对比。计算结果表明,石蜡碳酸锂混合物和铅组合是中子相机的最佳屏蔽层材料,其中石蜡碳酸锂混合物用于慢化吸收中子,铅用于屏蔽中子和γ射线。此外,利用MCNP模拟计算得到了屏蔽中子和γ射线所需的屏蔽厚度,以及准直管的中子散射率。     

Application of 10B counter with moderator for neutron ambient dose equivalent measurement during radiation monitoring at joint institute for nuclear research

The results of an investigation into the possibility of applying a device based on a ¹⁰B neutron counter (CHM-14) with a polyethylene moderator as the dosimeter of neutron ambient dose equivalent H*(10) in radiation fields of nuclear physics installations at the Joint Institute for Nuclear Research (JINR) are presented. It is shown that the device can be used as the dosimeter of this quantity in the neutron energy range from 0.4 eV to 20 MeV with an error no larger than 30% due to the difference between the energy dependence of its response and the energy dependence of the neutron ambient dose equivalent. Applying the correction coefficients allows one to extend the energy range of neutron dose H*(10) measurement to hundreds MeV. The error due to the anisotropy of the device response does not exceed 35%.     

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.     

锂玻璃探测器中子探测效率的刻度

为了精确测量keV能区的中子俘获截面,中国原子能科学研究院正在建造一台4π 全吸收型γ 探测装置---GTAF,锂玻璃探测器将会作为中子束流监视器测量中子能谱。利用5SDH-2 加速器刻度了锂玻璃探测器在两个入射中子单能点(250 和565 keV) 的探测效率,并使用EANT4 和MCNP 程序模拟计算了锂玻璃探测器的相对探测效率。通过归一化实验数据和模拟结果,得到了锂玻璃探测器在10keV~1 MeV 能区的中子探测效率曲线。对于把锂玻璃探测器测量得到的飞行时间谱转化为中子束流能谱,是一项非常重要的工作,同时为探测器效率刻度提供了新方法。In order to accurately measure the neutron capture cross section in the energy range of keVMeV, 4 πgamma-ray total absorption facility (GTAF) is being constructed at China Institute of Atomic Energy (CIAE). The lithium glass detector will be used as a neutron beam monitor for GTAF. The detection efficiency of the lithium glass detector at two incident neutron energy points (250, 565 keV) was calibrated in 5SDH-2 accelerator, and the relative detection efficiency was simulated by GEANT4 and MCNP code. By the normalization of the experimental data and simulation result, the neutron detection efficiency curve of the lithium glass detector between 10 keV and 1 MeV was obtained. This work will be important to convert the Time-of-flight spectrum that be measured by Li-glass detector to the energy spectrum of neutron beam, and provide the new method for calibration of detection efficiency.     

利用GEANT4和TALYS研究中子与铁作用的次级中子双微分截面

铁的次级中子双微分截面对核装置的设计、运行与维护具有重要作用。相关实验数据缺乏,且评价数据不完善,需要使用可靠的核理论模型进行计算。本工作利用GEANT4程序结合不同的中子评价数据计算了8.17, 11.5, 14.1和18 MeV等入射能量下中子轰击薄铁靶不同出射角度的次级中子双微分截面;同时利用TALYS程序和GEANT4程序结合BIC、BERT和INCLXX模型计算了25.7, 65, 100和150 MeV等入射能量下中子轰击薄铁靶不同出射角度的次级中子双微分截面,并与实验数据进行对比。研究表明,在20 MeV以下能区,ENDF/B-VIII.0库的计算结果与实验数据符合较好,BROND-3.1、CENDL-3.1、JENDL-4.0u和JEFF-3.3库的计算结果与实验数据存在差异。在20~150 MeV能区,GEANT4程序的BERT模型和TALYS程序的计算结果与实验数据符合较好,INCLXX模型和BIC模型的计算结果与实验数据存在分歧。整体来看,需要对铁的中子评价数据和核反应理论模型做进一步研究。     

Method for Monitoring of Neutron Fields near High-Energy Accelerators

The monitoring of neutron radiation from high-energy accelerators cannot fully rely on the standard dosimeters and radiometers manufactured in Russia, since these are sensitive only to neutrons with energies below some 10 MeV. This is because neutrons of higher energies can significantly contribute to the personnel doses both close to the accelerator shield and in the neutron multiscattered field around the shield. In this paper, we propose to measure the ambient neutron dose in energy range 10^–2 MeV to 1 GeV with a device consisting of two polyethylene balls with diameters of 3 and 10 in. housing slow-neutron detectors. The larger ball also comprises a lead converter (10'' + Pb). This device can be implemented in zonal radiation monitoring in the near-accelerator area.     

Monte Carlo calculations and experimental calibrations of Bonner sphere systems with a new cylindrical helium-3 proportional counter

The response functions to mono-energetic neutrons from thermal energy to 20 MeV of eight polyethylene moderating spheres with a cylindrical proportional counter were calculated with the MCNP-4B code for each moderator and confirmed by measurements in wide neutron fields (ISO), as Am–Be and ²⁵²Cf, and with the CANEL and SIGMA IPSN Cadarache (France) facilities, which simulate realistic field and thermal neutron response. With the described spectrometry system, we have carried out neutron measurements at different points from the annular zone inside the containment building of the Vandellòs II Nuclear Power Plant (Tarragona, Spain). The aim of this work is to present the results obtained during the sphere spectrometer development, with reference to the experimental and theoretical aspects and to show that our calculated spectra correspond to points of measurement compared with those obtained using the MORELPA code.     

基于2.5 MeV质子加速器BNCT装置的中子慢化材料性能模拟研究

基于加速器中子源的硼中子俘获治疗(Boron Neutron Capture Therapy, BNCT)是新一代的放射治疗方法,束流整形体(Beam Shaping Assembly, BSA)作为硼中子俘获治疗装置的重要组成部分,其作用是将中子源中的快中子束流慢化至超热中子能区(0.5 eV~10 keV),并尽可能减少快中子、热中子以及$\gamma $射线的成分,使其满足BNCT用于治疗的中子束要求。本工作基于蒙特卡罗软件包Geant4(Geometry and Tracking),以2.5 MeV,10 mA质子流强的7Li(p, n)7Be中子源为对象,研究分析了AlF₃ 、Fluental、Al₂O₃、Al作为慢化体材料时,不同的厚度对束流出口处的超热中子注量率、超热中子注量与热中子注量比值、快中子成分、$ \gamma $成分所产生的影响。计算表明,当选用厚度为25 cm的AlF₃作为慢化体材料时,经过整形慢化后的超热中子束的束流参数,均满足国际原子能机构(International Atomic Energy Agency, IAEA)的中子束流参数推荐值。     

Neutron dose measurement in carbon ion radiation therapy at HIRFL （IMP）

For radiation protection purposes, the neutron dose in carbon ion radiation therapy at the HIRFL （Heavy Ion Research Facility in Lanzhou） was investigated. The neutron dose from primary ^12C ions with a specific energy of 100 MeV/u delivered from SSC was roughly measured with a standard Anderson-Broun rem-meter using a polyethylene target at various distances. The result shows that a maximum neutron dose contribution of 19 mSv in a typically surface tumor treatment was obtained, which is less than 1% of the planed heavy ion dose and is in reasonable agreement with other reports. Also the T-ray dose was measured in this experiment using a thermo luminescent detector.