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。     

零功率反应堆内中子通量相对分布实时测量

在堆物理实验中,经常需要进行堆内中子通量相对分布的测量,以便获得有关的参数,如全堆平均热中子通量及功率不利因子、控制棒对中子通量分布的影响等等。为了要获得这些数据,有时不得不进行几千个测点的测量,才能求得结果。以往一般都采用经典的活化法。这种方法的最大缺点是测量工作量大,花费的人力多,不能很快地得到所需要的结果。为此,我们利用一种微型的中子探头,配以适当的电子仪器和机械设备,在轻水零功率反应堆内进     

加速器中子源的中子注量测量方法

在用静电加速器中子源标定探测器的中子灵敏度实验中,采用“BF3长计数管 定标器”系统过渡,用^197Au中子活化分析方法达到了对中子注量在线、绝对监测的目的。这种方法给出与加速器束流不同角度、不同距离处的中子注量。介绍了这种中子注量测量方法。     

用于CSNS反角白光中子能谱及注量率测量的快裂变室

即将建成的中国散裂中子源(China Spallation Neutron Source,CSNS)反角白光中子束线可为核数据测量提供高注量率的脉冲白光中子束流,填补我国核数据测量用白光中子源的空白,提高我国核数据测量水平,满足核能、核技术及基础核物理研究对核数据的需求。该束线建成后,其中子能谱及注量率的精确测量将是开展其它物理实验的基础,快裂变电离室因其独特优点被选为中子能谱和注量率测量探测器。通过实验研究了快裂变电离室的粒子分辨性能、时间分辨性能;确定阴、阳极的合理间距为10 mm,据此测得电离室的时间分辨约15 ns;利用235U样品量计算的探测效率与利用伴随粒子法给出的探测效率在不确定度范围内符合,因此可以标定快裂变室的探测效率。通过这些工作,完成了满足反角白光中子束能谱及注量率测量需求的快裂变室的物理设计。     

The study of aeroball system for measuring 3D neutron flux distribution in reactor core

Aeroball system is attractive in several aspects because it can easily transport the map of neutron flux distribution to be measured from incore to outside of a reactor vessel.However,before the aeroball system is put to practical use in the heating reactor.there are four topics that have to be further studied.They are the stability of the activated positions,enhancement of signal/noise(S/N)ratio,distributed control and data-acquisition system and on-lin nbeutron flux distribution reconstruction.Besides describing the rasons for them,this paper gives out the theory,concept and solution about the first two topics and it is helptul to give the possibility to enhance the reactor-power.     

软件仿真核电站堆芯核测量系统的设计与实现

针对堆芯核测量系统设计制造周期长、关键机械设备采购难、控制系统调试困难等问题,设计了仿真堆芯核测量系统。基于RSView32的仿真测量系统仿真电气设备和机械设备的功能,并配备了测量系统的状态监控界面。本文介绍了仿真测量系统的原理和实现方法。通过仿真测量系统与真实控制软件的联调试验,证实在系统设计过程中,仿真测量系统完全可以代替电气和机械设备,辅助测量系统的控制程序进行调试运行。该仿真测量系统的使用可显著减少机械磨损,缩短控制系统的调试周期。     

Direct measurement of mechanical momentum of a neutron flux

Translated from Atomnaya Énergiya, Vol. 66, No. 1, pp. 48–49, January, 1989.     

利用金属钆测量热中子通量的实验研究

研究了一种测量热中子通量的新方法,利用金属钆与热中子反应产生的次级γ射线来确定热中子通量。使用两个NaI探测器进行符合测量,设计了对伽玛射线和散射中子的良好屏蔽体。经过对本底和钆样品的多次测量,对钆与中子反应产生的伽玛能谱进行了分析,计算出了中子与钆反应产生的计数率(样品净计数率),其最大值达到2.74 Hz,从而证实了这种方法的可行性。     

The fast neutron energy spectrum and the thermal neutron flux distribution in the experimental channel of the VVR reactor

Using threshold detectors of aluminum, indium, iron, titanium, zinc, silicon, magnesium, mercury, and sulfur, we have measured the fast-neutron energy distribution in the experimental vertical channel of the VVR-2 reactor, traveling in the immediate vicinity of the active zone. We have obtained the spectral composition of the fast neutrons in the center of the channel at the level of the middle of the active zone, the thermal neutron flux and its distribution in the channel. The activity of all detectors except the sulfur and copper detectors was determined from the -radiation. The obtained experimental data are compared with the results of the theoretical calculation.     

固体径迹探测器测量束流装置内的中子通量密度

在微型反应堆零功率装置上搭建了硼中子俘获治疗拟采用的热中子束流装置。利用固体径迹探测器(SSNTD)测量了束流装置中心轴线上不同位置处的中子通量密度。结果显示,在束流装置入口处中子通量密度为5.39×107cm–2·s–1时,出口处热中子通量密度为5.63×104cm–2·s–1,热中子通量密度衰减到入口处的1/957。而利用热释光(TLD)方法和MCNP/4B程序测量和计算结果分别为1/1032和1/972。     

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.     

移动式堆芯中子注量率测量系统概述

堆芯中子注量率测量系统是压水堆核电站核测量系统的主要组成部分,用于测量反应堆堆芯的中子注量率水平,从而提供反应堆的功率分布情况。文章介绍了中核（北京）核仪器厂国产化的移动式堆芯中子注量率测量系统,并对测量系统的概况、系统组成、工作原理及功能等进行了描述。     

Intra-pellet neutron flux distribution measurements in LWR critical lattices

We have developed inexpensive and easy-handling measurement methods on intra-pellet neutron flux. A foil activation method with metallic foils, which were fabricated by punching out technique and etching technique to reduce fabrication error and positioning error, was used for the intra-pellet neutron flux distribution measurement. The developed method was applied to measure intra-pellet neutron flux distributions in a reduced–moderation light water reactor (LWR) lattices, and uncertainty of the distributions was estimated to be 1% to 2%. Measured values were analyzed with a continuous energy Monte Carlo code. Comparison of measurements and analyses revealed that the developed method is useful for the validation of an advanced fuel design method considering neutron behavior in fuel pellets.     

Development of an innovative neutron flux mapping system

An innovative in-core flux mapping system has been developed and applied successfully for service in commercial pressurized water reactors. With the benefit of double indexing path selector mechanism, the reliability of the detector drive system has been improved five times higher than that of a conventional system. The problems caused by strong friction generated between the detector cable and guide tubes have been resolved by the double indexing path selector architecture enabling the detector guide tubes to have smaller curvature and shorter length in nature. The system normally exerts two-thirds of force to make the detector reach the top of the fuel position, compared with conventional ones. In addition, simple and fast maintenance is realized by optimizing the number of components in the detector drive system and providing easy access to the components, thereby guaranteeing minimum radiation exposure to maintenance personnel. The programmable logic controller-based digital controller with Windows^®-based operator console provides fully automated and user-friendly operation and maintenance support means. The developed in-core flux mapping systems have been deployed at the Kori nuclear units 1–4.     

Fast neutron flux measurement and spectroscopy using silicon semiconductor detectors

A method is developed for measuring fast neutron fluxes and spectra using small silicon semiconductor detectors in the presence of relatively high gamma radiation levels. A general formula is derived for calculating the averaged cross section and the effective energy. Good agreement is found between the calculation and experimental data. The sensitivity of the method is much higher than that of any other threshold neutron detector with comparable size.     

The effect of a partially inserted absorbing control rod on the neutron flux density distribution

This paper presents a method of calculating the effectiveness of a partially inserted absorbing, cylindrical control rod in a reactor without a reflector. The method makes it possible to give a comparatively simple estimate of the distortion of the neutron flux density near the control rod. The calculations are made using the one group equation + B² = 0, with special boundary conditions on the surface of the rod. As an example of the use of the method, the values of B² and the corresponding flux distributions have been calculated for several positions of the rod in the reactor.We wish to thank R. Zezula, I. Svatosh, and M. Prazhskaya for carrying out the numerical calculations. particularly great initiative was shown by R. Zezula, who also calculated the tables of the function Zk'(a. B²)/Zk(a, B²) on the electronic BÉSM setup in the Calculational Center of the Academy of Sciences of the USSR.     

靶件内中子注量率分布计算方法研究

选择ANISN作为实验靶件内中子注量率分布计算的程序，编制辅助程序输入混合材料截面。计算得到延时水箱附近的中子注量率，与测量数据作对比。计算得到靶片自屏因子，并与2000年实验数据对比。确认计算方法可行后，计算得到实验靶件内热中子注量率分布数据。     

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.