Detection quantum efficiency and spatial resolution of neutron CCD-detectors

A method has been developed to calculate detection quantum efficiency and spatial resolution of CCD-detectors for thermal neutrons. It is based on a physical model and a Monte-Carlo code developed to describe the processes which take place by visualising a spatial distribution of the neutron flux with a CCD-detector. Recommendations for the application of existing detectors of different design were deduced from the results of carried out calculations.     

^252Cf中子分布条件下的中子探测效率测量

利用BC-501A探测器测量了252Cf自发裂变中子源,得到不同中子阈能的252Cf中子分布条件下的中子探测效率,实验得到了252Cf中子源的核温度T=1.42±0.03MeV.     

中子注量率密度时空分布数据采集软件开发

介绍基于Windows98平台的中子注量率密度时空分布32位数据采集软件。软件采用多线程技术进行数据采集和处理。采用数据曲线和数据表格两种形式显示、打印数据．采用多文档界面技术同时处理多个中子注量率密度时空分布数据谱。     

Calculation of the pulse height distribution and counting efficiency of a fast neutron scintillation detector

An efficient method was developed for calculating the pulse height distribution from a scintillation detector irradiated by monoenergetic neutrons. The method was based on the separation of the pulse height distribution conlributions from single and multiple neutron scattering by nuclei of the scintillator material. The contribution from single scattering was computed analytically, and that from multiple scattering by the Monte-Carlo method, The new method is considerably more precise than the usual Monte-Carlo method.Consideration was given to a number of effects which were neglected in previous work.The computations performed were mainly designed to obtain information about fast neutron spectra from experimental data.Translated from Atomnaya Énergiya, Vol. 15, No. 3, pp. 194–200, September, 1963     

The effect of temperature and control rod position on the spatial neutron flux distribution in the Syrian miniature neutron source reactor

The effect on the spatial neutron flux distribution for both of water and fuel temperature increase as well as the change in the control rod position are presented in the Syrian miniature neutron source reactor (MNSR). The cross-sections of all the reactor components at different temperatures are generated using the WIMSD4 code. These group constants are used then in the CITATION code to calculate the spatial neutron flux distribution at different water and fuel temperatures and different control rod positions using four energy groups. This work shows that the increase in water and fuel temperatures during the reactor daily operating time does not affect the spatial neutron flux distribution in the reactor. The change in the control rod position does not affect as well the spatial neutron flux distribution in the reactor except in the region around the control rod position.     

Economic efficiency of radionuclide neutron sources

Conclusion The method considered here enables us to make comparative economic estimates at the stages of development and application of RNS in the solution of the following problems: determining the limits of competitiveness of RNS based on different radionuclides, and also of RNS and other types of NS, determining the efficiency of utilization of specific RNS in comparison with conventional methods of performing the work, justifying the nomenclature of RNS manufacture with specific radionuclides, justifying the unit power and the number of simultaneously used RNS, estimating the efficiency of utilization of different combinations of RNS that have been used for a specified time and of new RNS, determining the efficiency of the successive use of RNS in different regions as their power is depleted, justifying the desirability of chemical reprocessing of the RNS used, etc.Today the method is used in practice for different purposes. In particular, an estimate has been prepared for the economic efficiency of the chemical reprocessing of californium RNS that have been taken out of service. The method has also been used for determining the limit of RNS power above which chemical reprocessing is effective. It has been shown that the chemical reprocessing makes it possible to increase the output of californium and reduce the cost of its manufacture. Researchers have also determined the economically advantageous interval of reprocessing with due regard for the decay of californium and the desirability of increasing the batches reprocessed.Translated from Atomnaya Énergiya, Vol. 58, No. 5, pp. 355–359, May, 1985.     

中子计数器探测效率的校准

设计制作了一个 BF3长硼中子计数器 ,采用伴随粒子法标定了其中子探测效率 ,约为 3.17×10 -4 ( 1± 18% )。利用这个中子探测器测量了中子管脉冲中子产额 ( 10 7/ pulse)     

Comparison of different plasma chambers in microwave ion source for the intense neutron tube

The microwave absorption efficiency, which is relevant to magnet field and its distribution, is a major parameter of the microwave ion source (MWIS) for the intense neutron tube. Based on previous work, the relations between microwave absorption efficiency and plasma chamber structure and thickness of the microwave introduction window are studied. The microwave absorption efficiency reaches to 100% when plasma chamber is 100mm long and the window thickness is 30mm. The microwave absorption efficiency as a function of pressure is also presented.     

中子发生器产额稳定性

中子发生器的应用是很广泛的,这是因为中子发生器具有独特的优点:如可关断性和单能性.但是,中子发生器的产额稳定性和开机重复率差,达不到元素分析的领域的要求,电子学自动控制是一种简便易行效果显著的好方法.     

中子照相图像分辨率计算方法

中子照相技术属无损检测技术中的一种,通过中子透射成像获取物体内部的结构、缺陷等信息,已广泛应用于基础研究和工业应用。中子照相图像的分辨率是评价中子照相设备性能好坏最重要的参数之一。描述和计算分辨率的方法主要有点分布函数、线分布函数、刀口分布函数以及调制解调函数等。本文从基本原理出发,详细说明了各函数的物理概念和数学推导,并对各计算方法的优缺点进行了分析比较。     

Gamma-ray and neutron shielding efficiency of Pb-free gadolinium-based glasses

The radiation shielding efficiency of material depends upon photon attenuation, exposure buildup factors and neutron removal capacity. A newly developed Pb-free gadolinium-based glasses in compositions(80-x) B_2O_3-10 Si O_2-10 Ca O-x Gd_2O_3(where x = 15, 20, 25, 30 and35 mol%) had completely been investigated for their shielding efficiency with Geant4 simulation for mass attenuation coefficients and neutron total macroscopic cross section and by calculating exposure buildup factors.The exposure buildup factors for photon energy from 0.015 to 15 Me V had been calculated up to 40 mean free paths using five factors geometric progression method. The mass attenuation coefficients of the Pb-free glasses were simulated for energies from 223 to 2614 ke V and compared with the possible available experimental results. The neutron shielding efficiency of these glasses was discussed by calculating neutron total macroscopic cross section for energies from 1 e V to 14.1 Me V. Present investigations are found to be very useful for applications in nuclear engineering.     

由中子注量率空间分布形状测量确定反应堆次临界度的方法研究

从堆物理的基础理论出发,提出了通过堆内中子注量空间分布的测量来确定反应堆次临界度的一种新方法,并通过对我国启明星1 号次临界实验装置的数值模拟,初步说明了该方法的可行性.     

能谱变化对239Pu裂变室测量快中子注量的影响

热中子和共振区的中子在快中子临界装置中所占的份额很小,但是由于其相对大的截面,在慢化物存在的情况下,热中子和共振中子份额的微小变化,对^239Pu裂变室测量中子注量的结果影响很大。通过测量^239Pu裂变电离室在包镉和包硼、周围有无慢化物等情况下的反应率,Au、In活化片的镉比,S活化片在能谱变化下与^239。Pu的反应率比等,分析了快中子临界装置中热中子和共振区中子的分布,讨论了中子能谱变化对^239Pu裂变室测量快中子注量的影响及解决办法。