多核DSP在空间中子探测仪中的应用

空间任务要求电子系统在很少甚至没有维护的情况下稳定工作,但空间电子系统易受到粒子辐射的影响而导致功能紊乱。为了解决空间中子探测仪中的通用多核DSP受粒子辐射影响所导致的程序安全性下降的问题,需要对DSP进行抗辐射加固。本文将介绍中子探测仪中的多核DSP所采用的多种抗SEU软件防护措施。故障注入实验表明,这些防护方法能够有效地减少SEU导致的软件错误。     

蒙特卡罗方法在中子活化在线分析系统设计中的应用

选取重水、石墨、聚乙烯等6种慢化材料，利用MCNP程序对不同的慢化材料进行模拟计算分析。计算结果表明，中子活化在线分析系统的最优化慢化材料为聚乙烯。实验测定了以聚乙烯为慢化材料的中子活化分析系统的热中子注量率随源到引出孔之间的距离以及探测器处于不同位置时的分布关系，为下一步进行中子活化在线分析研究提供了依据。     

慢化球探测器组合法测试中子屏蔽性能应用

随着核技术在各领域的应用推广,中子屏蔽材料得到越来越广泛的应用,而目前国内外还没有测试中子屏蔽性能的统一标准。为了探索一种简单可行、能够在较宽中子能量范围内测试材料屏蔽性能的方法,本文对3He正比计数管、计数管外包镉及计数管外包不同直径的聚乙烯(Polyethylene,PE)慢化球共12个模型进行了MCNP(Monte Carlo N Particle Transport Code)模拟计算,得到一种慢化球探测器组合测试方法,使测试能够在1×10-5-1.25 Me V能量范围内有较一致的响应。利用这种方法测试了2 cm和4 cm厚PE对252Cf中子的透射率,与多球谱仪解谱法得到的结果在±1.0%内相吻合,对几种材料的测试结果也符合不同类型材料对中子的屏蔽规律,证实了这种简易组合测试方法的可行性。     

中子吸收剂~6LiH和B(CH_2)_x性质的理论研究

本文用慢化理论及蒙特卡罗方法计算了中子吸收剂~6LiH及B(CH_2)_x的中子学特性。组合逃脱共振几率P、平均慢化时间T及平均慢化长度L_s,我们引进一个新的量│p=p,它能很好地表示中子吸收剂的性质。对二种吸收剂作了比较。我们发现对含硼聚乙烯吸收剂存在一个最佳配比。     

中子吸收剂~6LiH和B(CH_2)_x性质的理论研究

本文用慢化理论及蒙特卡罗方法计算了中子吸收剂~6LiH及B(CH_2)_x的中子学特性。组合逃脱共振几率P、平均慢化时间T及平均慢化长度L_s,我们引进一个新的量|P=P,它能很好地表示中子吸收剂的性质。对二种吸收剂作了比较。我们发现对含硼聚乙烯吸收剂存在一个最佳配比。     

形态滤波在快中子图像降噪处理中的应用

在基于CCD耦合闪烁光纤阵列的快中子照相系统中,快中子图像受噪声污染较为严重.针对快中子图像中的椒盐噪声和泊松噪声,研究了形态滤波技术在降噪方面的应用.结果表明在对快中子图像处理过程中,基于二维多方向结构元素的形态滤波在滤除噪声和保持图像细节等方面效果较佳.     

核燃料棒235U富集度均匀性扫描装置中的中子慢化系统计算

对于核燃料棒235U富集度均匀性扫描装置,为了合理有效地利用252Cf中子源,提高检测灵敏度,需要合理选择中子慢化材料,优化中子慢化过程。本文利用Monte Carlo方法对中子慢化系统进行了优化计算,在保证慢化中子(En<1MeV)通量密度较高和235U与238U的裂变反应几率比R5/8也较高的前提下,给出了几种慢化材料及其组合的结果。     

~(241)Am-Be源在煤中所形成中子场的参数拟合

研究了7种煤中主要元素对241Am-Be中子源在煤中形成中子场的影响,给出了描述中子场中快中子和热中子数量变化曲线的经验公式和拟合参数。     

兰州大学的中子发生器研制及应用展望

1988年兰州大学成功研制了3×10¹² s^-1的ZF-300强流中子发生器,主要用于核数据测量、材料辐照损伤等研究。为进一步开展活化法中子核数据测量、裂变物理等研究,兰州大学启动了基于倍压加速器的ZF-400强流中子发生器研制工程,该中子发生器的设计指标为D束流能量400 keV、D束流强度大于30 mA、D-D中子产额大于5×10¹⁰ s^-1,D-T中子产额大于5×10¹² s^-1。在裂变物理研究方面,已成功发展了描述裂变核断点裂变势的势驱动模型(potential-driving model),并开展了中子诱发典型锕系核素裂变发射中子前裂变产物的质量分布计算研究;将potential-driving model植入Geant4程序,发展了用于裂变发射中子后裂变产物质量分布、动能分布、裂变中子能谱等模拟的蒙特卡罗方法,并开展了可靠性评估研究;研制了一套用于裂变产物实验测量的双屏栅电离室(TFGIC),并完成了初步实验测试。在中子应用技术方面,为满足小型...     

Evaluation of Neutron Capture Gamma-Ray Spectra in Hafnium and Tantalum

A method has been developed for evaluation of neutron capture γ-ray spectrum. It couples measured intensities of primary and secondary discrete—-ray with a γ-ray cascade model to calculate the unresolved part of the capture spectrum, and adds the discrete part and the unresolved part to obtain the whole spectrum. The cascade model uses the level density formula proposed by Gilbert & Cameron and the Brink & Axel form of El γ-ray profile function with a modification. This method was applied to thermal neutron capture spectra in six hafnium isotopes and ¹⁸¹Ta and was extended also to non-thermal capture spectra in ¹⁸¹Ta for 0.25 and 0.5 MeV neutrons with empirical assumptions. The calculated results were compared with experiments and agreement was good not only in terms of the gross structure, but also in terms of the fine structure which appears at high and low γ-ray energies.     

Construction of Neutron Guide Tubes in Upgraded JRR-3

Five neutron guide tubes have been installed in the upgraded JRR-3 (Japan Research Reactor No. 3). Two of them are for thermal neutrons and the other three are for cold ones. The characteristic wavelength of the thermal neutron guide tubes is 2 Å, and those of the cold neutron guide tubes are 4 and 6 Å. The longest guide tube is 59.9 m long and the total length of guide tubes is 232.1 m.

The beam sizes are 2 cm × 20 cm for the thermal neutron beams and 2 cm × 12 cm for the cold neutron beams. A curved part of the neutron guide is assembled by a polygonal approximation with use of 85 cm long straight units. The neutron mirrors of these units are made of natural Ni deposited borosilicate glasses. The Ni layer is about 2,000 Å in thickness.

The mean fabrication error of guide tube units is 4 μm. The mean installation errors are 8 μm for the positional abutment error and 5 × 10^?6 rad for the angular error. The neutron losses by these errors will be about 5%, and the neutron fluxes at the exits of the neutron guides are estimated to be about 2 × 10⁸ n/cm²·s.     

中子照相多路准直器优化设计与应用

设计了一种多路准直器用于消除中子照相中的散射中子,利用MCNP5对准直器的中子吸收材料、长度进行了优化设计,利用该准直器对不同厚度的水层样品在不同样品探测器距离下进行了中子照相的MC模拟计算。计算结果表明选用25μm的Gd作为准直器的中子吸收涂层,准直器长度为1cm时可消除98%的散射中子,使用该准直器可以有效提高中子照相定量分析的精度。     

Neutron Economy of Transmutation of TRU in Thermal and Fast Neutron Fields

Neutron economy of the transmutation of TRU was examined in well thermalized, thermal and fast neutron fields. Burn-up chains of ²³⁷Np, ²⁴¹Am and ²⁴³Am, which are the main TRU nuclides in the high level waste, were calculated in the flux region from 10¹⁴ to 10¹⁷ n/cm².s. Numbers of neutrons absorbed and produced of each chain were calculated using JENDL-3. The net number of neutron produced n _net, which was obtained by the difference of the two numbers, largely varied with the neutron fields, the nuclides and the flux levels. The n _net value in the fast neutron field was positive (0.0–1.0) for ²³⁷Np, ²⁴¹Am, ²⁴³Am and TRU with the nuclide composition in the high-level waste generated by the conventional PWR. The transmutation of TRU by fission can be performed with producing neutrons in the fast neutron field. On the other hand, the n _net value was negative for the well thermalized and thermal neutron fields. For TRU in the high-level waste, the values in those fields were —1.0 at 10¹⁴ n/cm².s and 0.0 at 10¹⁶ n/cm².s. In the high flux region of 10¹⁶ n/cm².s, TRU in the high-level waste can be transmuted by fission without consuming additional neutrons. In the flux region about 10¹⁴ n/cm².s, the transmutation of TRU in the high-level waste by fission requires about one neutron.     

Research and Improvement of Pu Sample Property Measurement Based on Fast Neutron Multiplicity Counter

Fast neutron multiplicity counting (FNMC) analysis method can effectively measure the properties of samples. Based on the fourth-order FNMC analytical equations, a set of three-layer fast neutron multiplicity counters with six liquid scintillators per layer was constructed for Geant4 simulation, and the values of related parameters were determined. Metal Pu sample with 1 cm iron, aluminum, carbon, and stainless steel packaging material was externally simulated, and the sample satisfied the assumption by the equation adaptive analysis. The measurement parameters such as detection efficiency and multiplicity counting rate were simulated. When the mass of Pu sample is less than 500 g, the increase of sample solution mass deviation is less than 1.20% with carbon as packaging material, and the influence of iron material and stainless steel material is less. According to the measurement results, the self-multiplication factor was corrected for the sample without shell, and the third-order polynomial fitting equation was obtained and the goodness of fit is 0.933. The corrected solution mass deviation of sample with mass less than 1 kg is less than 6.00%. The results show that the medium-heavy metal with thickness of 1 cm has little effect on the measurement of Pu samples. The combination of the fast neutron multiplicity counter and the coefficient correction method can achieve more accurate measurement of the sample properties.     

中子探测土壤水分的信号采集处理过程

本文以ＬＮＷ－５０Ｃ型中子仪在红壤实验地田间标定为例，详细阐述中子探测土壤水分的信号采集处理过程。     

Assessment of Neutron Cross Sections for Titanium with Neutron Spectrum in Titanium Pile

The measurement of angular neutron spectrum in a quasi-spherical pile of Ti was carried out by the linac time-of-flight method for the assessment of neutron cross sections for Ti in the energy range from a few keV to a few MeV. The measured spectrum in the pile is generally in good agreement with the calculated one from ENDF/B-IV (MAT = 1,286 for Ti) except in the energy range from about 60 keV to a few 100 keV, where the calculation gives considerably lower neutron flux than the measurement.

In order to investigate the cause of this discrepancy between the measured and calculated spectra, the total cross sections for Ti were measured by the transmission method. The results give larger values of total cross sections for Ti by about 30% than ENDF/B-IV below 200 keV, and smaller values by about 10% above 200 keV. These results were ensured at 55 and 147 keV by the measurement using a Si-filtered neutron beam. The calculation based on the measured cross sections shows better agreement with the measured spectrum than that based on ENDF/B-IV. The discrepancy is still observed around 100 keV.

The sensitivity analysis shows the importance of cross sections above 1 MeV and elastic cross sections in the resonance energy region to solve the disagreement between the measured spectrum and the calculated one.     

Time Delay of Thermal Neutron Detection during Power Burst in TRACY

Time delay of thermal neutron detection at a micro fission chamber has been experimentally confirmed during an initial power burst caused by reactivity addition in the Transient Experiment Critical Facility (TRACY). In the experiment, the power was measured by using a fast response γ-ray ionization chamber, which was newly designed exclusively for the TRACY experiment, in addition to the fission chamber. As results of the experiment, the transient power measured by the fission chamber was observed about 4 ms later than that by the γ-ray chamber during the power increasing, and then the delay time expanded after the power peak. The time delay must be taken into account to evaluate accurately the transient power. To comprehend the observed time delay in detail, time-dependent analyses were performed using the MCNP4B code. In the analyses, time response of neutron detection at the fission chamber was calculated and used for the simulation of the transient power measured by the fission chamber. From results of the analyses, it was confirmed that the simulated power well agreed with the measured power and the observed time delay was caused from the thermalization of neutrons emitted at the TRACY core and the following flight to the fission chamber.