Study of neutron spectra in a water bath from a Pb target irradiated by 250 MeV protons

Spallation neutrons were produced by the irradiation of Pb with 250 MeV protons. The Pb target was surrounded by water which was used to slow down the emitted neutrons. The moderated neutrons in the water bath were measured by using the resonance detectors of Au, Mn and In with a cadmium (Cd) cover. According to the measured activities of the foils, the neutron flux at different resonance energies were deduced and the epithermal neutron spectra were proposed. Corresponding results calculated with the Monte Carlo code MCNPX were compared with the experimental data to check the validity of the code. The comparison showed that the simulation could give a good prediction for the neutron spectra above 50 eV, while the finite thickness of the foils greatly effected the experimental data in low energy. It was also found that the resonance detectors themselves had great impact on the simulated energy spectra.     

Study of neutron activation yields in spallation reaction of 400 MeV/u carbon on a thick lead target

The spallation-neutron yield was studied experimentally by bombarding a thick lead target with 400 MeV/u carbon beam. The data were obtained with the activation analysis method using foils of Au, Mn, Al, Fe and In. The yields of produced isotopes were deduced by analyzing the measured γ spectra of the irradiated foils. According to the isotopes yields, the spatial and energy distributions of the neutron field were discussed. The experimental results were compared with Monte Carlo simulations performed by the GEANT4+FLUKA code.     

MCNP Dose Calculations in a CT Phantom for Therapeutic External Photon Beam

In this paper, we have addressed the problem of the radiation transport with the Monte Carlo N particle(MCNP) code. This is a general purpose Monte Carlo tool designed to transport neutron, photon and electron in three dimensional geometries. To examine the performance of MCNP5 code in the field of external radiotherapy, we performed the modeling of an Electron Density phantom (EDP) irradiated by photons from 60Co source. The model was used to calculate the Percent Depth Dose (PDD) at different depths in an EDP. One field size for PDD has been examined. A 60Co photons source placed at 80 cm source to surface distance (SSD). The results of calculations were compared to TPS data obtained at National Institute of Oncology of Rabat.     

15N（p,n)15O反应中子源0°角激发函数测量

利用中国原子能科学研究院HI-13串列加速器上的多探测器快中子飞行时间谱仪, 测量了38个不同入射质子能量点下15N（p, n)15O反应0°角方向的激发函数。 测量数据用蒙特卡罗方法进行了模拟, 以进行中子注量衰减和入射窗厚度的不确定度修正。 实验在入射质子能量位于6.029—8.056 MeV之间时发现了3个共振峰, 这一点与DROSG 2000评价数据及PTB数据相符合, 但三家的截面数值存在差异, 对这些差异需要作进一步深入探讨。 The excitation function of 15N（p, n)15O reaction at 0 degree was measured at 38 energy points using the fast neutron Time of Flight (TOF) spectrometer at the HI 13 Tandem Accelerator in the China Institute of Atomic Energy（CIAE）. The measured data were analyzed by Monte Carlo simulation for the corrections of neutron flux attenuation and uncertainty of the thickness of the entrance foil. Three resonance peaks were observed in this experiment in the energy range from 6.029 to 8.056 MeV,which is comparable with the DROSG 2000 evaluated data and the PTB data. However,more experimental studies are needed since the cross sections deviate with each other.     

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.     

Measurement of tritium production rate distribution for a fusion-fission hybrid conceptual reactor

A fusion-fission hybrid conceptual reactor is established. It consists of a DT neutron source and a spherical shell of depleted uranium and hydrogen lithium. The tritium production rate (TPR) distribution in the conceptual reactor was measured by DT neutrons using two sets of lithium glass detectors with different thicknesses in the hole in the vertical direction with respect to the D~+ beam of the Cockcroft-Walton neutron generator in direct current mode. The measured TPR distribution is compared with the calculated results obtained by the threedimensional Monte Carlo code MCNP5 and the ENDF/B-Ⅵ data file. The discrepancy between the measured and calculated values can be attributed to the neutron data library of the hydrogen lithium lack S (α,β) thermal scattering model, so we show that a special database of low-energy and thermal neutrons should be established in the physics design of fusion-fission hybrid reactors.     

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.     

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.     

Calculation of the wide-angle neutron spectra from the ~9Be(d,xn) reaction in a thick beryllium target

The multi-layer computing model is developed to calculate wide-angle neutron spectra, in the range from0° to 180° with a 5° step, produced by bombarding a thick beryllium target with deuterons. The double-differential cross-sections(DDCSs) for the ~9 Be(d, xn) reaction are calculated using the TALYS-1.8 code. They are in agreement with the experimental data, and are much better than the PHITS-JQMD/GEM results at 15°, 30°, 45° and 60° neutron emission angles for deuteron energy of 10.0 MeV. In the TALYS-1.8 code, neutron contributions from direct reactions(break-up, stripping and knock-out reactions) are controlled by adjustable parameters, which describe the basic characteristics of typical direct reactions and control the relative intensity and the position of the ridgy hillock at the tail of DDCSs. It is found that the typical calculated wide-angle neutron spectra for different neutron emission angles and neutron angular distributions agree quite well with the experimental data for 13.5 MeV deuterons. The multi-layer computing model can reproduce the experimental data reasonably well by optimizing the adjustable parameters in the TALYS-1.8 code. Given the good agreement with the experimental data, the multi-layer computing model could provide better predictions of wide-angle neutron energy spectra, neutron angular distributions and neutron yields for the ~9 Be(d, xn) reaction neutron source.     

高压倍加器靶单元组件对源中子能谱影响研究

靶单元组件对中子有散射和吸收作用,与入射粒子也会发生反应,因此高压倍加器中子场在空间的分布除了由反应本身的角分布决定外,还受到靶单元组件的影响。实验利用飞行时间法测量了高压倍加器T(d,n)4He 中子源产生的实际中子场,测量角度为0°~95°,共8 个角度,采用伴随 粒子法对入射束流进行归一。借助蒙特卡罗方法模拟实验过程和靶单元组件等对中子场分布的影响,将实验结果同模拟计算得到的结果进行了比较与分析,为高压倍加器相关实验布局的合理布置和靶单元组件的优化提供了一定的依据。The spatial distribution of neutron field for the High Voltage Multiple Accelerator (HVMA) is not only determined by the angular distribution of the reaction itself, but also affected by scattering and absorption of the target unit. Meanwhile, undesired nuclear reactions of the incoming ions with the target and self-target build-up may contribute to this neutron background, which disturbs the primary neutron field. The spatial distribution of actual neutron field generated by T(d, n)4He on HVMA was measured using the flight-time method. The neutron emission energy spectra were measured at 8 different angles from 0° to 95°. The results were monitored by the accompanying-particle method. TARGET and MCNP (Monte Carlo simulation process) programs were used to simulate the experiment. The results of simulated calculation were compared with the experimental data. The analyzed results will give some qualitative and quantitative conclusions for target unit optimizing and provide some foundational works for physical measurement in HVMA.     

启明星Ⅱ次临界非稳态热中子注量率测量

中子注量率及分布是反应堆的重要参数,本工作通过核数守恒在非稳态情况下的推导和求解,从理论上论证了次临界反应堆非稳态情况下中子注量率测量的可行性。将活化法和固体径迹法有机结合,利用固体径迹探测器标定活化片的测量数据,测量了启明星Ⅱ零功率装置的He-3管实验孔道内及反应堆外壁的中子注量率的分布,并与模拟计算结果进行了比较,利用MCNPX程序得到的模拟计算结果与实验结果的趋势一致,证明了该测量方法可以测量低通量的中子注量率,可实现反应堆不同时刻、不同位置的中子注量率测量,为CiADS技术的研发提供了实验数据与技术支撑。Neutron flux measurements were carried out at VENUS-Ⅱ lead-based zero power reactor by neutron activation method combined with solid-state nuclear track detectors (SSNTD). This experimental method was proposed based on the principle of nuclear number conservation when a foil was irradiated in an unsteady-state neutron field. By this method, thermal neutron flux distributions inside the He-3 duct were measured when VENUS-Ⅱ was operated under unsteady-state. The neutron flux distributions were also calculated with MCNPX code and were consistent with the experimental data. In addition, the neutron fluxes in the outer layer of VENUSⅡ were measured under steady-state. These results would benefit the further study of experimental methods for neutron flux measurement and provide important support for the design of CiADS.     

Theoretical analysis of double-differential neutron emission cross sections for n+56Fe reactions at incident energies of 7-13 MeV

The double-differential neutron emission cross sections for n+56Fe reactions at incident energies of 7-13 MeV at different angles are calculated by the UNF(abbreviation for unified,2009 Version)code,which is based on the unified Hauser-Feshbach and exciton model.The results indicate that the higher the incident energies,the better the results,although there are some discrepancies between the calculated results and the measured data for natural iron.These discrepancies are analyzed in detail in this paper.In addition,the calculated results are also compared with the evaluated results of ENDF/B VII.O and JEFF-3.1.1 near the angle of 90°at incident energies of 8.17 and 11.5 MeV,respectively.     

Measurement of differential cross sections of neutron-induced deuteron production reactions on carbon from 25 to 52 MeV

The angle-differential cross sections of neutron-induced deuteron production from carbon were measured at six neutron energies from 25 to 52 MeV relative to those of n-p elastic scattering at the China Spallation Neutron Source(CSNS) Back-n white neutron source.By employing the ΔE-E telescopes of the Light-charged Particle Detector Array(LPD A) system at 15.1° to 55.0° in the laboratory system,ratios of the angle-differential cross sections of the ~(12)C(n,xd) reactions to those of the n-p scattering were measured,and then,the angle-differential cross sections of the ~(12)C(n,xd) reactions were obtained using the angle-differential cross sections of the n-p elastic scattering from the JENDL-4.0/HE-2015 library as the standard.The obtained results are compared with data from previous measurements,all of which are based on mono-energic neutrons,the evaluated data from the JENDL-4.0/HE-2015 library and the ENDF-B/Ⅷ.0 library,and those from theoretical calculations based on INCA code and Talys-1.9 code.Being the first white-neutron-source-based systematic measurement of the angle-differential cross sections of neutron-induced deuteron production reactions on carbon in several tens of MeV,the present work can provide a reference to the data library considering the lack of experimental data.     

75As中子辐射俘获截面的实验测量及蒙特卡罗修正

在四川大学2.5 MV静电加速器上， 用活化法测量了29—1 100 keV 能区中子引起的75As(n， γ)76As反应截面， 并用MCNP 4C程序对实验中由中子的多次散射、 注量率衰减效应等引起的偏差进行了修正， 最后将所得结果与已有数据进行了比较。 We have measured the neutron radiative capture cross section of 75As in the energy range of En=29—1 100 keV using the activation technique at 2.5 MV electrostatic accelerator of Sichuan University. The experimental data were then corrected for multiple scattering and self shielding effects with MCNP 4C code. We also compared our results， with the existing measurement， calculation and evaluated data in ENDF/B Ⅵ.     

D-T中子与平板型聚乙烯材料作用的次级中子角度谱实验研究

基于反冲质子法建立了一种测量D-T中子与平板型宏观样品作用的次级中子角度谱的实验方法.为保证探测器的能量线性并在较低的中子有效测量下阈(0.5 MeV)情况下获得好的中子-伽马射线甄别性能,采用高、低能段分别测量的方法.采用事件记录法,同时记录了次级中子和伴随伽马射线的脉冲形状甄别和脉冲幅度二维信息,利用基于ROOT数据分析平台编写的离线数据分析程序,完成了伴随伽马射线的挑选和扣除,以及高、低两能段反冲质子谱的拼接,并成功的将神经网络技术应用于中子能谱的解谱,获得了D-T中子与9和18 cm厚平板型聚乙烯材料作用的0.5-15 MeV的次级中子角度谱实验结果.实验模型的MC模拟由MCNP5完成,数据库采用ENDF-VI,实验结果和MC计算结果在实验不确定度范围内一致. 关键词： D-T中子 积分中子学 反冲质子法 次级中子能谱     

贫化铀/聚乙烯球壳交替系统中铀-238中子俘获率的测量与分析

为校验次临界能源堆的概念设计,建立了贫化铀/聚乙烯球壳交替系统, 采用活化法测量²³⁸U的中子俘获率. 贫化铀片置于系统内与入射D离子束成90^o的方向上活化 ,用HPGe探测器测量²³⁸U俘获中子衰变产生的²³⁹Np 衰变产生的277.6 keV特征γ射线计数,实验修正了贫铀片对277.6 keV γ射线的自吸收, 得到了交替系统中²³⁸U (n, γ)反应率的径向分布,反应率的相对不确定度为3.5%-3.7%, 并计算得到系统上整个贫铀区中²³⁸U的总中子俘获率为2.24 ± 0.09. 用MCNP5程序在常用ENDF库下进行了模拟计算, ²³⁸U (n, γ)反应率分布计算与实验一般在5%以内符合, 总俘获率在1%以内符合.     

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。     

(n,γ)反应实验研究中的中子屏蔽设计

用射线全吸收型装置(Gamma-ray Total Absorption Facility,GTAF),可以对中子俘获反应截面进行高精度测量。为了降低实验本底,实验中需要对源中子进行准直和屏蔽,还要对被样品散射的中子进行吸收以减少它们进入探测器后所形成的干扰。采用MCNP对中子的准直器、屏蔽体和中子吸收体进行了模拟设计,中子准直屏蔽体材料选用含硼聚乙烯(BC4 的质量分数为3%) 和铅。准直孔直径为13 mm,长度为500mm,经准直后样品处中子束斑坪顶直径为21 mm。中子吸收体材料选用聚乙烯和碳化硼,吸收体球壳内腔半径30 mm,聚乙烯壳层厚度60 mm,碳化硼壳层厚度10 mm,被样品散射的中子经吸收体后衰减93.7%。Neutron capture cross section can be measured by Gamma-ray Total Absorption Facility (GTAF) with high precision. To reduce the background of experiments, the neutron source must be collimated and shielded, and the neutrons scattered from the sample must be absorbed to minimise interference after they go into the detector. The shield, collimator and absorber were simulated and designed with MCNP code. Boron-ontainingpolyethylene with 3% BC4 and lead are used as the materials for the neutron collimator and shield. The diameter of the collimating aperture is 13 mm, and the length of the collimator is 500 mm. After being collimated, the diameter of neutron beam plateau at the sample position is 21 mm. The neutron absorber is made of polyethylene and BC4, and the thickness of polyethylene shell and BC4 shell are 60 and 10 mm, respectively. The simulated result shows that neutrons scattered from the sample can decay 93.7% through the neutron absorber.