用阈探测器活化法测量50MeV/u重离子实验靶区的中子注量率、能谱和中子产额

用阈探测器中子活化法测量了50MeV/u ~(12)C离子实验靶区的次级中子平均注量率、角分布、粗略能谱,并估算了重离子反应的中子产额。     

用阈探测器活化法测量50MeV／u重离子实验靶区的中子注量率,…

用阈探测器中子活化法测定了５０ＭｅＶ／ｕ＾１２Ｃ离子实验靶区的次级中子平均注量率、角分布、粗略能谱，并估算了重离子反应的中子产额。     

Geant4在中子辐射效应中的应用

中子辐射效应是半导体器件在辐射环境中损伤的重要因素。本文建立了中子在半导体材料中的电离和非电离能量沉积、原子空位密度的Geant4模拟方法。电离能量沉积可用于分析电离总剂量效应,非电离能量沉积可用于分析位移损伤效应。电离kerma因子的模拟结果定量解释了中子辐照在CMOS工艺单片机中引起的电离增强效应。通过原子空位密度计算了中子引入的附加陷阱密度,分析了位移损伤对电离效应的增强作用。实验和模拟结果表明,中子的电离能量沉积加剧了CMOS工艺单片机的退化。     

利用LR 115 2B研究7.4GeV质子束轰击铅靶的散裂中子源外表面中子分布

利用LR 115 2B探测器研究了8cm× 2 0cm的铅靶在 7.4GeV质子的轰击下 ,散裂中子在实验装置外表面的分布 ,对实验结果进行了模拟解析 ,同时利用中子活化法验证了LR 115 2B探测结果及理论模拟结果利用LR 115 2B探测器研究了8cm× 2 0cm的铅靶在 7.4GeV质子的轰击下 ,散裂中子在实验装置外表面的分布 ,对实验结果进行了模拟解析 ,同时利用中子活化法验证了LR 115 2B探测结果及理论模拟结果     

利用次级中子诊断燃料面密度

介绍了一种用于测量在低中子产额下的压缩燃料面密度诊断方法,并且这种方法在实验中得到了实现。它是利用塑料闪烁体探测器对辐射驱动充纯D的内爆靶所产生的初级中子和次级中子进行测量,在中子产额较低的情况下,实现了利用次级中子和初级中子的产额比值法来诊断燃料面密度〈ρR〉,并获得了实验数据。这个工作对我们以后开展燃料面密度的诊断奠定了基础。     

keV能区单能中子参考辐射场核反应靶设计

质子轰击中等质量靶核是产生keV能区单能中子的一种常用方法。选择⁴⁵Sc(p,n)⁴⁵Ti反应中子源作为keV能区单能中子参考辐射场的中子源,利用FloWizard软件模拟大束流条件下核反应靶的温度分布,分析了影响靶温度分布的主要因素。利用Target程序模拟核反应靶出射中子能谱,分析了不同材料的散射中子本底。同时精细调节5SDH 2加速器端电压,测量了薄靶（Sc）的激发曲线,测量结果与NPL和PTB的接近。     

“核-光转换”中子探测系统物理结构的模拟研究

"核-光转换"中子探测器是以惰性气体为介质将裂变碎片能量转换为光辐射的裂变室,拥有电离探测器所没有的优点:不需要供电电源;信号传输方式采用光导或光纤,而不是绝缘电缆;对伽马辐射极不灵敏;输出信号较大,可以避免在探测器附近使用前置放大器。根据"核-光转换"中子测量系统的特点,采用Geant4模拟了铀裂变靶厚度、惰性气体成分、腔体材料等对到达惰性气体的裂变碎片和可见光的影响,给出了NOC结构设计的最佳参数和中子能量响应。     

合肥国家同步辐射实验室核物理实验区的辐射屏蔽设计和防护评价

本文主要介绍合肥国家同步辐射实验室(HNSRL)核物理实验区辐射屏蔽计算的方法及结果。对能量为400MeV、束流功率为6 kW 的电子束轰击靶时产生的轫致辐射和次级中子进行屏蔽计算;估算了靶室顶部次级中子的天空反射对周围辐射水平的影响;对靶室内的感生放射性气体(主要是~(13)N)的浓度进行了估计,确定了排放烟囱的高度。文中对靶室和核物理实验大厅提出了设置局部屏蔽和防护门等防护措施。     

~(12)C与~(209)Bi反应中Au同位素的产额与Q_(gg)的关系

用47MeV/u12C离子轰击天然铋靶,通过炮弹和靶核之间的核子转移反应产生Au同位素。使用放射化学方法从大量Bi和复杂反应产物中分离、纯化Au,并制备Au的γ射线测量源。使用HPGe探测器测量放射性Au同位素的γ活性。根据照射结束时Au同位素的活度和其他相关数据,确定每个Au同位素的产生截面。分析发现,缺中子Au同位素的产生截面与Qgg值之间不遵从指数依赖关系,这可用重离子碰撞中的次级过程加以解释。     

基于FMEA的密封氘氚中子管可靠性分析

中子管是可控中子发生器的关键核心部件。国产中子管在可靠性、稳定性等方面与国外先进水平还有一定差距。为了得出影响中子管可靠性的关键因素,以预制靶密封氘氚中子管为例,采用风险优先数法(Risk Priority Number, RPN)开展了故障模式及影响分析(Failure Mode and Effect Analysis, FMEA),按不同约定层次对所有可能的故障模式、故障原因、故障影响、后果严酷度、发生概率及危害程度进行了详细分析。根据FMEA结果,讨论了危害性较大的故障模式,并给出了提高中子管可靠性的设计改进方向。研究表明:离子源和靶是影响中子管可靠性的关键部件,离子源阴极/阳极短路导致的电离异常、靶膜氧化或脱落导致的吸附氘氚气体能力丧失等故障模式的危害性较大。     

Monte Carlo simulations of response functions for gas filled and scintillator detectors with MCNPX code

Monte Carlo simulation of a detector response function presents a very challenging problem. The detector response functions have been calculated for different neutron and gamma detectors: ³He gas filled proportional counter, NE213 organic scintillator, BrillanCe 350 or LaCl₃(Tl), and an ionization chamber with mixed gas composition. MCNPX code was used for simulations. The simulations were done with different neutron and gamma energies. The effects of neutron scattering, wall effects, recoil continua and contribution from charged particles have been included. The detector response function for the NE213 organic scintillator was obtained with consideration of light output curves of different products of neutron reactions with materials of the scintillator. The simulated data has been compared with experiments.     

New integral experiments for large angle scattering cross section data benchmarking with DT neutron beam at JAEA/FNS

A new integral experiment with a deuteron–triton fusion (DT) neutron beam started in order to validate scattering cross section data. First the DT neutron beam was constructed with a collimator. The performance of the collimator system and the characteristics of the DT neutron beam were measured. Second a new integral experiment for type 316 stainless steel (SS316) was carried out with this DT neutron beam. The DT neutron beam of 3.5 cm in diameter was injected to the front surface center of an SS316 cylindrical assembly. Reaction rates of the ⁹³Nb(n,2n)^92mNb reaction in the assembly were measured with the activation foil method and were calculated with the Monte Carlo transport calculation code. The measurement points were located at three positions, on the center of the beam axis and at 15 cm and 30 cm apart from the axis. The ratio of calculation to experiment of the ⁹³Nb(n,2n)^92mNb reaction rate became smaller than 1 with the distance from the beam axis. Then, the dependency of each reaction rate on scattering angle was calculated. It was proved that at off-axis positions, where C/E is smaller than 1, 90° scattering contribute relatively larger than at on-axis positions and backward scattering made little contribution to the results in this experiment. The reasons of the discrepancy between the measured and calculated data will be investigated.     

Evaluation of Power History during Power Burst Experiments in TRACY by Combination of Gamma-Ray and Thermal Neutron Detectors

A combination method using γ-ray and thermal neutron detectors was newly applied to the accurate evaluation of power histories during reactivity-initiated power burst experiments in the Transient Experiment Critical Facility (TRACY). During an initial power burst, the power history was determined using a fast response γ-ray ionization chamber, which was used because of its ability to exactly trace the power history within a short duration of the initial burst. After the initial burst, a micro fission chamber containing highly enriched uranium was used for the determination of the power history because the γ-ray ionization chamber could not be applied due to the contribution of delayed γrays from fission products. By the present method, the power histories were evaluated for the experiments in the range of 1.50 to 2.93$ of the reactivity insertion. It was found that the peak power and integrated power as determined by the previous method using only the micro fission chamber were underestimated to be 40% and 30% in maximum, respectively, in comparison with the results from the present evaluation. The numerical simulation performed by using the Monte Carlo method indicated that the underestimation could be comprehended by considering the time delay of thermal neutron detection of the fission chamber, which arose from the flight-time of neutrons from the TRACY core to the fission chamber.     

反冲质子探测系统中子灵敏度的Monte Carlo算法

在强流脉冲中子测量工作中,常利用PIN探测器测量中子与聚乙烯中的H核发生弹性散射产生的反冲质子.本文利用Monte Carlo技术进行快速计算的反冲质子探测系统中子灵敏度表达式,给出了算法和计算流程.     

The design of “multiple energies from one source” in D-T neutron tube

To obtain multiple monoenergetic neutron sources and realize the on-site calibration of radiation monitoring equipment for nuclear-involved places,the structural characteristics and neutron source features of D-T neutron tube were analyzed;Monte Carlo method was adopted to simulate the effect of interaction between typical materials and different energy neutrons;multilayered shielding materials were combined and optimized to acquire the optimal scheme to shield the neutron sources from the neutron tube.On the base,a tapered alignment filtration construction was designed and Monte Carlo method was employed to simulate the effect of alignment construction.The result showed that the tapered alignment filtration construction can create monoenergetic neutrons including14.1 MeV,0.18 MeV and thermal neutrons and demonstrated good monochrome performance which provides multiple monoenergetic sources for the on-site calibration.     

Improvement of Monte Carlo Lattice Burnup Calculation Performance with the Correlated Sampling Method

For next generation reactor designs, which are attempting wide variations of assembly configurations, the flexibility Monte Carlo method holds is attractive, but still costly for repetitive design study works. This paper presents an advanced correlated sampling (ACS) method which was developed to speed up Monte Carlo lattice burnup calculations. The ACS method is the combination of the correlated sampling method and a pseudo-scattering technique. All burnup steps are considered as consecutive perturbed problems using the same neutron collision history, which is pre-calculated based on a selected unperturbed problem. Since neutron weights can be adjusted on every collision point, rather than along paths between them, the perturbed calculation is very fast and the neutron collision history is light enough to be stored in memory or physical storage, which is an indispensable feature for consecutive perturbed calculations. The presented theory shows that the ACS method has good potential to work for a wide range of neutron absorption variations, the dominant perturbation in the lattice burnup. In an example calculation on a BWR lattice, the ACS calculation results of 600,000 neutrons/step agree well with the independent Monte Carlo runs of 20,000,000 neutrons/step within 0.1%dk/k in terms of k_? throughout 95 steps (~50GWd/t). Average calculation time of neutron tracking with the former method is 3.4 s/step with 600,000 neutron histories on a single processor of an Alpha21164-600 MHz, and the speed-up factor against the Monte Carlo calculation turns out to be about 100.     

T(d,n)4He强流中子发生器的中子能谱和角分布模拟计算

采用将厚靶分割成薄靶的方法对厚氚钛靶、260keV氘束流能量条件下T(d,n)⁴He反应中子源的能谱和角分布进行计算。以分割法计算得到的能谱和角分布数据为基础,建立了D-T中子源Monte-Carlo模拟抽样模型,在考虑中子发生器各元件材料及实验大厅墙壁对快中子的慢化、散射和吸收的条件下,采用MCNP程序对兰州大学3×10¹²s^-1强流中子发生器260keV氘束流能量下的中子能谱和角分布进行了模拟,给出了模拟结果。为检验模拟结果的可靠性,与实验测量能谱进行了比较,Monte-Carlo模拟谱和实验测量谱基本符合。     

中子背散射探雷技术的分析与评估

本文分析了中子背散射探雷技术的原理,通过数值模拟对中子背散射探雷技术进行研究,并对该技术进行分析评估,为进一步开展中子背散射探雷技术的工程研究提供理论基础。     

Calculation and analysis of p + Ca reaction cross sections at incident energies from threshold to 250 MeV

All cross sections, angular distributions and energy spectra of neutron, proton, deuteron, triton, helium, alpha particle emission for p + ^{40,42,43,44,46,48,nat}Ca reactions have been calculated and analyzed at incident proton energies from threshold to 250 MeV by nuclear theoretical models. The theoretical calculated results are in good agreement with existing experimental data.     

Novel hybrid Monte Carlo/deterministic technique for shutdown dose rate analyses of fusion energy systems

The rigorous 2-step (R2S) computational system uses three-dimensional Monte Carlo transport simulations to calculate the shutdown dose rate (SDDR) in fusion reactors. Accurate full-scale R2S calculations are impractical in fusion reactors because they require calculating space- and energy-dependent neutron fluxes everywhere inside the reactor. The use of global Monte Carlo variance reduction techniques was suggested for accelerating the R2S neutron transport calculation. However, the prohibitive computational costs of these approaches, which increase with the problem size and amount of shielding materials, inhibit their ability to accurately predict the SDDR in fusion energy systems using full-scale modeling of an entire fusion plant. This paper describes a novel hybrid Monte Carlo/deterministic methodology that uses the Consistent Adjoint Driven Importance Sampling (CADIS) method but focuses on multi-step shielding calculations. The Multi-Step CADIS (MS-CADIS) methodology speeds up the R2S neutron Monte Carlo calculation using an importance function that represents the neutron importance to the final SDDR. Using a simplified example, preliminary results showed that the use of MS-CADIS enhanced the efficiency of the neutron Monte Carlo simulation of an SDDR calculation by a factor of 550 compared to standard global variance reduction techniques, and that the efficiency enhancement compared to analog Monte Carlo is higher than a factor of 10,000.