用于中子测量的俘获探测器研究

为了在弱中子场和有限小空间内测量绝对裂变率, 制作了俘获探测器, 研究了俘获探测器的性能。 介绍了用于中子测量的俘获探测器和铅屏蔽室, 以及该探测器系统在特定条件下测量裂变反应率的结果, 并与裂变室测量结果进行了比较。 探讨了铅屏蔽室大小对测量结果的影响。 To detect the absolute neutron flux in a weak neutron field and restricted space, the fission fragment trapping detector was fabricated and the properties of the detector were studied. In this paper, the detector and shielding chamber used in neutron detection were described and the experimental measurements of the fission rate in specific condition were performed with the detection system and the result has been compared with that obtained by fission chamber. The influence of the shielding chamber on the measured results was analyzed.     

Investigating Transmission Efficiency of Light Guide by Monte Carlo Simulation

A large area neutron detector to detect the energy of about 1 GeV neutron by time-of flight method will be installed at RIBLL II of CSR. To obtain good energy resolution, the time resolution of the detector is a crucial parameter. For this purpose, the transmission efficiency of the light guide to transport the photons from detec-tor unit to light sensitive detector has been investigated by Monte-Carlo simulation. Here, the simulations were done mainly with two types of the light guides, namely type A and type B as shown in Figs.1 and 2 respectively.     

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.     

LS301快中子探测器的效率刻度

主要介绍了利用252Cf标准裂变中子能谱刻度快脉冲LS301型液闪中子探测器的探测效率的实验方法和结果, 简要介绍了实验数据的处理过程, 得到了阈值分别为0.5, 0.7, 1.0和1.6 MeV, 以及中子能量在10 MeV以下的探测器效率, 并对测量结果进行了误差分析。同时为了检验实验结果的准确性, 将实验结果与理论计算结果进行比较, 两者在不确定度范围内符合得很好。Neutron detection efficiency of LS301 fast neutron detector was calibrated by measuring the neutron energy spectrum of 252Cf source, which has a standard fission neutron spectrum. A low mass, fast ionization chamber is used as the fission fragments detector in the time of flight(TOF) spectrometer and afforded the start signal of neutron flight. The stop signal was offered by the anode of LS301. A measured TOF spectrum was turned to the neutron energy spectrum which will be compared with the standard one. Consequently, the fast neutron detection efficiency of LS301 was obtained. Calibration procedures of experimental and data processing was given. Relative detection efficiencies were obtained separately for threshold settings of 0.5, 0.7, 1.0 and 1.6 MeV for neutron energies under 10 MeV. Experimental results accorded with theoretical efficiency curves which were calculated with the Monte Carlo code NEFF50.     

应用于热中子的nTHGEM探测器性能研究

随着脉冲强流中子源的发展,对高性能中子探测器提出了更大的挑战,3He气体资源严重短缺和高计数率中子探测器的迫切需求,已开始制约着中子源应用技术的发展。中国科学院高能物理研究所针对中子的特殊性,专门研发了一种陶瓷基材的nTHGEM（neutron Thick Gaseous Electron Multiplier）探测器用于中子探测。基于nTHGEM的中子探测器具有高计数率、高位置与时间分辨能力、增益大、制作工艺简单,且便于大面积制作的特点,是目前国际上发展替代3He探测技术的重要方向之一。为了详细研究nTHGEM探测器的本身性能,本工作使用55Fe放射源研究了nTHGEM探测器的增益、计数率稳定性、能量分辨率等关键参数与nTHGEM膜间电压、收集场强、漂移场强之间的关系,优化了nTHGEM探测器在不同工作气体中的工作参数,为后续进一步优化nTHGEM探测器设计和工艺奠定了基础。实验结果表明,单层nTHGEM探测器在Ar（90%）+CO₂（10%）混合气体中增益能达到103,探测器计数率稳定性良好。另外,还在中国原子能科学研究院的CARR反应堆（China Advanced Research Reactor）上进行了中子束流实验,通过狭缝测量到探测器位置分辨率为（3.01±0.03）mm（FWHM）,已经接近高气压3He MWPC中子探测器水平。With the development of pulsed intense neutron source, the high-performance neutron detector poses more challenges. The severe shortage of 3He gas resources and the urgent need of neutron detector with high counting rate have begun to restrict the neutron source application technology development. In response to the particularity of neutrons, the Institute of High Energy Physics of CAS developed a nTHGEM(neutron Thick Gaseous Electron Multiplier)of ceramic substrate for neutron detection. The neutron detector based on nTHGEM is one of the most important directions for the development of alternative 3He detection technology in the world at present because of its high counting rate, high position and time resolution, large gain, simple fabrication process and large area production. In order to study the properties of nTHGEM detector in detail, this paper studied the relationship between nTHGEM detector's gain, counting rate stability, energy resolution and other key parameters and nTHGEM film voltage, collection field strength and drift field strength using 55Fe radioactive source, Optimized the working parameters of nTHGEM detector in different working gases, which laid the foundation for further optimization of nTHGEM detector design and process. The experimental results show that the single-layer nTHGEM detector has a gain of 103 in a Ar(90%)+CO₂(10%) mixed gas with good counting rate stability. In addition, a neutron beam experiment was performed on the China Advanced Research Reactor at the China Institute of Atomic Energy, and the position resolution of the detector was (3.01±0.03) mm (FWHM) measured by slits. Its performance is close to the high pressure 3He MWPC neutron detector level.     

Study on gamma response function of E J301 organic liquid scintillator with GEANT4 and FLUKA

The gamma response function is required for energy calibration of EJ301 （5 cm in diameter and 20 cm in height） organic liquid scintillator detector by means of gamma sources. The GEANT4 and FLUKA Monte Carlo simulation packages were used to simulate the response function of the detector for standard 22Na, 60Co, 137Cs gamma sources. The simulated results showed a good agreement with experimental data by incorporating the energy resolution function to simulation codes. The energy resolution and the position of the maximum Compton electron energy were obtained by comparing measured light output distribution with simulated one. The energy resolution of the detector varied from 21.2% to 12.4% for electrons in the energy region from 0.341 MeV to 1.12 MeV. The accurate position of the maximum Compton electron energy was determined at the position 81% of maximum height of Compton edges distribution. In addition, the relation of the electron energy calibration and the effective neutron detection thresholds were described in detail. The present results indicated that both packages were suited for studying the gamma response function of EJ301 detector.     

D-D 中子发生器n-γ 密度测井的蒙特卡罗模拟研究

根据D-D 反应中子的能谱和角分布数据,建立了中子源模型;根据石灰岩地层标准刻度井群数据,建立了井模型。采用MCNP 程序模拟了井中中子和γ 射线的输运,得到了不同地层密度和不同源距处NaI 探测器中的混合γ 射线能谱和非弹γ 射线能谱。在混合γ 射线能谱2.5~5.0 MeV 能区开窗,研究了开窗区混合γ 射线相对计数随源距的变化关系,确定源距应选择在30~80 cm 范围,给出了密度与混合γ 射线计数之间的非线性关系。研究表明,可以利用D-D中子源的混合γ 射线能谱来实现n-γ 密度测井。A D-D neutron source model was developed according to the neutron spectrum and the neutron angular distribution of D-D reaction. A standard calibration well model was built. The transports of the D-D neutrons and γ-rays in the well were simulated using MCNP code. The mixed γ-spectra and inelastic γ-spectra in the NaI detector were obtained for both different distances from a neutron source and different densities. Mixed γ count in the energy range of 2.5 to 5.0 MeV as a function of distance shows that the NaI detector should be located at a distance of 3080 cm from the neutron source. The nonlinear relationships between the density and the mixed γ count were presented in this paper.It was demonstrated that the energy spectrum of mixed γ rays can realize the n-γ density logging.     

Effects due to a Pu-C source on a HPGe detector and the corresponding neutron shielding

A gamma spectrum of a Pu-C source is measured using a p-type HPGe detector, whose three peaks (full energy, single-escape and double-escape peak) can be used as a calibration source for the beam energy measurement system of BEPCII. The effect of fast neutron damage on the energy resolution of the HPGe detector is studied, which indicates that the energy resolution begins to deteriorate when the detector is subject to 2×10⁷ n/cm² fastneutrons. The neutron damage mechanism and detector repair methods are reviewed. The Monte Carlo simulation technique is utilized to study the shielding of the HPGe detector from the fast neutron radiation damage, which is of great significance for the future commissioning of the beam energy measurement system.     

Study on the novel neutron-to-proton convertor for improving the detection efficiency of a triple GEM based fast neutron detector

A high-efficiency fast neutron detector prototype based on a triple Gas Electron Multiplier(GEM) detector, which, coupled with a novel multi-layered high-density polyethylene(HDPE) as a neutron-to-proton converter for improving the neutron detection efficiency, is introduced and tested with the Am-Be neutron source in the Institute of Modern Physics(IMP) at Lanzhou in the present work. First, the developed triple GEM detector is tested by measuring its effective gain and energy resolution with55 Fe X-ray source to ensure that it has a good performance.The effective gain and obtained energy resolution is 5.0×104and around 19.2%, respectively. Secondly, the novel multi-layered HDPE converter is coupled with the cathode of the triple GEM detector making it a high-efficiency fast neutron detector. Its effective neutron response is four times higher than that of the traditional single-layered conversion technique when the converter layer number is 38.     

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.     

Neutron measurements using bubble detectors--terrestrial and space

Like all other radiation monitors currently used in the space program, the bubble detector (which has flown on several missions) was adapted from a technology that was developed for terrestrial radiation. Bubble detectors are the most recent technology for applications in personal neutron dosimetry. They are now regarded as a mature technology and are used in many countries as a neutron dosimeter of record with approval from the respective regulatory authorities. Extensive type testing and QA of bubble detectors has been done by numerous groups, many of these to show that bubble detectors meet national radiation protection requirements prior to their acceptance as a dosimeter of record. In fact, it has been stated "Only bubble detectors achieve a sufficiently low detection threshold (to meet ICRP-60 recommendations)..." (Portal and Dietze, 1992).     

Study of bubble damage detectors for neutron detection

Bubble damage detectors have been prepared by using polyacrylamide as detector solid and freon as detector liquid. Tests show that the prepared detectors are sensitive to fast neutrons and have proportionality between bubble number and neutron fluence within a certain range of neutron fluence. Therefore, it can be used as a fast neutron detector and a dosimeter. Our experiments also show that the sensitivity of the detector increases with temperature.     

Effect of gamma energy, temperature and shelf life on the response of gamma and BD-100R neutron bubble detectors

The use of commercially available bubble detectors for gamma and neutron detection has been reported by several groups. We have carried out systematic investigations to study the response of gamma bubble detector as a function of energy, the effect of waiting time on sensitivity of BD-100R neutron bubble detector and the effect of shelf life on both types of bubble detectors. Our results show that the response of the gamma bubble detector is energy dependent. Waiting time of up to 260 hours for BD-100R neutron detector has no significant effect. Both the detectors have a limited shelf life. These factors must be taken into account while using these detectors.     

Study of the improvement of TLD cards for personal neutron dosimetry

In this work, personal thermoluminescence dosimeter (TLD) cards type of GN-6770 (holder type 8806) from Harshaw were used for personal neutron dosimetry. The response of the dosimeters has been determined in terms of the personal absorbed dose and personal dose equivalent for different neutron energy components, based on the recommendations of ICRP-60 and ICRU-49. Neutron irradiation was performed using a 5 mCi Am–Be neutron source. The TLD reader, type Harshaw 6600, was installed and calibrated for accurate neutron doses equivalent to gamma-ray doses. It was found that fast neutron doses measured by TLD (badges or cards) are in agreement with those measured by neutron TE (tissue equivalent gas) ionization chambers and neutron monitors. Thermal neutron doses measured by TLD cards were overestimated when compared with those measured by neutron monitors. Additional Cd was used to reduce thermal neutron doses to be in agreement with actual thermal doses. Other configurations for TLD crystals are also suggested for accurate thermal neutron dose measurements.     

An investigation into the accuracy of the albedo dosimeter DVGN-01 in measuring personnel irradiation doses in the fields of neutron radiation at nuclear power installations of the joint institute for nuclear research

The calculated results of research into the accuracy of an individual albedo dosimeter DVGN-01 as it corresponds to the personal equivalent dose for neutrons H _p (10) and to the effective dose for neutrons E _eff in the neutron fields at Joint Institute for Nuclear Research Nuclear Power Installations (JNPI) upon different geometries of irradiations are presented. It has been shown that correction coefficients are required for the specific estimation of doses by the dosimeter. These coefficients were calculated using the energy sensitivity curve of the dosimeter and the known neutron spectra at JNPI. By using the correction factors, the uncertainties of both doses will not exceed the limits given to the personnel according to the standards.     

An investigation into the sensitivity of various albedo neutron dosimeters aimed at correcting the readings

The results of an experimental determination of the sensitivity of three types of individual neutron albedo dosimeters in neutron reference fields on the basis of radionuclide sources and at the top concrete shielding of the U-70 accelerator are presented. The results show that the ratios between the responses of the albedo dosimeters designed earlier at the Joint Institute for Nuclear Research (the albedo dosimeter (AD) and the multicomponent dosimeter (MD)) and the currently used DVGN-01 dosimeter are constant within 25% in a wide range of neutron energy. This fact makes it possible to use the results of measuring the AD and MD responses obtained earlier in neutron fields of nuclear-physical installations at the Joint Institute for Nuclear Research (JINR) for the correction of DVGN-01 dosimeter measurement results to apply it to personal radiation monitoring (PRM) at these installations. The correction factors for DVGN-01 measurement results are found and recommended to be used in PRM for most JINR installations.     

A model for distinguishing between static and dynamic exposure of personal thermoluminescence dosimeters

To distinguish between static and dynamic (normal) exposure of personal TL dosimeters, a model of radiation deposition and of TL light transport in the TLD dosimeter is proposed. The RADOS dosimeter badge using MCP-N (LiF:Mg,Cu,P) TL detectors with standard filters replaced by special Pb and Cu filters with a pattern of holes or inserts was modelled. The photon radiation transport in the dosimeter and energy deposition in the TL detector were simulated by the Penelope Monte Carlo transport code. The model of TL light transport within the TL pellet takes into account the distribution of energy deposition in the TL detector, light self-absorption in the detector and reflection of TL light of the heating planchet. The shape and hole pattern of the filters were optimized with respect to best distinction between static and dynamic exposures. The results of calculations were verified experimentally by exposing RADOS badges with modified filters to beams of low energy X-rays directed at various angles.     

An automatic evaluation system for NTA film neutron dosimeters

At CERN, neutron personal monitoring for over 4000 collaborators is performed with Kodak NTA films, which have been shown to be the most suitable neutron dosimeter in the radiation environment around high-energy accelerators. To overcome the lengthy and strenuous manual scanning process with an optical microscope, an automatic analysis system has been developed. We report on the successful automatic scanning of NTA films irradiated with ²³⁸Pu---Be source neutrons, which results in densely ionised recoil tracks, as well as on the extension of the method to higher energy neutrons causing sparse and fragmentary tracks. The application of the method in routine personal monitoring is discussed.     

Studying the response of CR-39 and BD-100R detectors for neutron dosimetry in the light of ICRP-60 recommendations

The latest radiobiological findings show that the damage caused by the neutrons to human body is far greater than was previously thought. The International Commission on Radiological Protection, therefore, revised Q for neutrons and made recommendations in their Report No. ICRP-60 which have serious consequences for many operational neutron dosimetry services. The new recommendations include the reduction in dose limits from 50 mSv to 20 mSv per year, change in the quality factors and the dose estimation in some circumstances down to a few mSv per year. The overall effect of the ICRP-60 recommendations is to reduce the detection threshold in order to account for the reduction in the effective equivalent dose. We have studied the impact of ICRP-60 recommendations on the response of CR-39 based cubical neutron dosimeter and BD-100R bubble detector and the results obtained are presented in this article.     

Calibration factor for estimating personal dose equivalent with imaging plates

A personal imaging plate (IP) dosimeter is in the process of being developed for neutron fields using the BaFBr:Eu²⁺ phosphor. A configuration incorporating a polyethylene radiator placed before the IP detector is used to produce protons via (n,p) elastic scattering. For a dosimeter sensitive to thermal neutrons, a Nylon plate ( thick) is placed between the polyethylene (1.2 mm thick) radiator and the IP ( thick sensitive layer) detector to produce protons via the ¹⁴N(n,p)¹⁴C reaction. Dosimeters having these configurations have been exposed to neutrons from ²⁴¹Am–Be and ²⁵²Cf sources at the Institute for Radioprotection and Nuclear Safety of Cadarache at angles of 0 (normal incidence), 30 and 60 and several dose equivalents. The personal dose equivalent response in terms of H_p(10) is evaluated from the net measured photostimulated luminescence densities (). The calibration factor obtained for estimating the personal dose equivalent with this dosimeter is for ²⁴¹Am–Be and for ²⁵²Cf.