Systematic errors using a multicomponent activation detector to determine the neutron flux of252Cf fission

Translated from Atomnaya Énergiya, Vol. 69, No. 5, pp. 329–330, November, 1990.     

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。     

能量选择法中子照相技术

能量选择中子照相是近年来新兴的中子照相技术,具有广泛的潜在应用前景。介绍了能量选择中子照相技术的基本原理,比较分析了机械速度选择器法、双晶单色器法、飞行时间法以及超镜和晶体过滤器结合法4种获取单色中子束的方法。介绍了能量选择中子照相技术在增强图像对比度、应力测量、织构分析以及相变过程研究等方面体现的优异特性和应用潜力。     

Eu活化过渡法测量绝对热中子注量率

在相同条件下,以不同时间辐照Ａｕ箔和Ｅｕ箔,用γ谱仪测量＾１９８Ａｕ的４１１ｋｅＶγ射面积和＾１５２Ｅｕ的４１１ｋｅＶγ峰面积,再通过４πβ－γ符合装置测得金箔的绝对中子注量率,根据两者的峰面积之比和测得的Ａｕ箔绝对中子注量率,求得＾１５３Ｅｕ的４１１ｋｅＶγ峰面积对应的中子注量率。     

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.     

The design of the J-type neutron flux distribution measurement system

This paper introduces the design of a J-type aeroball system that the tube penetrates the lateral wall of reactor pressurized vessel (RPV), then immediately goes down to the vessel bottom and then goes up through the lower core support plate into the reactor core. Some experimental results related to gas flowing within a thin tube are presented in the paper, such as the gas friction drag coefficient on the ball’s way and etc. From theoretical and experimental viewpoints, the feasibility of the system is proved in pneumatic holding-up and β measurement aspects. In order to ensure an enough ratio of signal to noise, the maximum distance between two measuring points in water reactors is given. The paper gives out the measuring number per two-assemblies width N_i=Int(n_i/2+0.51), which is the accuracy relation between the number of fuel assembly and the minimum one of measuring point to only reconstruct 2D neutron flux distribution completely by the measured data.     

HT-7超导托卡马克装置上中子通量分布的实验研究

本文介绍了HT-7超导托卡马克装置上中子通量分布的测量实验.在实验中,使用了两种有较强抗伽马能力的中子探测系统:BF3和3He正比计数管及其相应的电子学与数据采集.在2005年春季实验中,在相同实验条件下,细致测量并比较研究了真空室外不同水平位置、垂直位置的中子通量.实验研究了真空室窗口和变压器铁芯的存在对测量结果的影响,验证了测量结果与三维数值计算的一致性.本项工作为更深入的物理研究奠定了基础.     

用钴活化法测定反应堆中热中子积分通量

本文叙述了用钴活化法测定高通量堆中热中子积分通量的方法。测得的热中子积分通量值与计算值作了比较。本法适于测定在高通量堆中长期辐照的较高热中子积分通量。     

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。     

Three-dimensional visualization of void fraction distribution in steady two-phase flow by thermal neutron radiography

Three-dimensional void fraction distributions of a steady air–water two-phase flow in a 4×4 rod-bundle with circular ferrule type spacers were measured by neutron radiography using a CT method. The high flux thermal neutron radiography system at JRR-3M in JAERI was used. Two-phase flow was visualized with a SIT tube camera and time-averaged one-dimensional cross sectional averaged void fraction distributions were calculated. Visualization with high spatial resolution up to 0.18 mm was carried out by using a cooled CCD camera. Projections in 250 directions were obtained and were reconstructed by a filtered back projection method after using some image processing techniques. Animations were made to show the three-dimensional distributions. One-dimensional and three-dimensional void fraction distributions of the steady state two-phase flow in the rod bundle near the spacer were clearly visualized.     

Determination of critical mass and neutron flux distribution by means of physical models

The design of nuclear reactors, especially new reactors, requires experimental measurements in order to obtain accurate values of the pertinent parameters. In the present paper we present a new method for the preliminary determination of the critical mass of a reactor and the neutron flux distribution; this method is based on the use of physical models. In carrying out these experiments use is made of a model of the reactor which does not contain fissionable material. The working channels in the model are filled with a neutron absorber whose cross section simulates the absorption cross section for neutrons in the fissionable material. The production of fast fission neutrons is simulated by means of a neutron source which is moved along the channels. The distribution of thermal neutrons is measured by means of detectors which are sensitive to thermal neutrons. If the source strength and the absolute value of the neutron flux are known, it is possible to find the critical mass of the reactor.This method has been checked in a reactor with uranium hexafluoride. The value of the critical mass found experimentally was found to be in good agreement with the value obtained when the reactor was started up.The proposed method can also be useful in preliminary investigations of reactor designs, the choice of optimum lattice parameters, etc. The technique is extremely simple and does not require fissionable material or high neutron fluxes.     

SPRR-300反应堆混凝土屏蔽层内中子注量率分布研究

采用MCNP程序与ANISN程序结合的计算方案获取了SPRR-300反应堆混凝土屏蔽层内的中子注量率分布情况，同时采用固体核径迹探测器测量了混凝土屏蔽层外低水平中子注量率，两者吻合较好，说明了计算结果的可信性。上述结果为反应堆退役工作提供了放射性源项的计算依据。     

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。     

Application of global variance reduction method to calculate a high-resolution fast neutron flux distribution for TMSR-SF1

The solid fuel thorium molten salt reactor(TMSR-SF1) is a 10-MWth fluoride-cooled pebble bed reactor. As a new reactor concept, one of the major limiting factors to reactor lifetime is radiation-induced material damage. The fast neutron flux(E 0.1 MeV) can be used to assess possible radiation damage. Hence, a method for calculating high-resolution fast neutron flux distribution of the full-scale TMSR-SF1 reactor is required. In this study,a two-step subsection approach based on MCNP5 involving a global variance reduction method, referred to as forward-weighted consistent adjoint-driven importance sampling, was implemented to provide fast neutron flux distribution throughout the TMSR-SF1 facility. In addition,instead of using the general source specification cards, the user-provided SOURCE subroutine in MCNP5 source code was employed to implement a source biasing technique specialized for TMSR-SF1. In contrast to the one-step analog approach, the two-step subsection approach eliminates zero-scored mesh tally cells and obtains tally results with extremely uniform and low relative uncertainties.Furthermore, the maximum fast neutron fluxes of the main components in TMSR-SF1 are provided, which can be used for radiation damage assessment of the structural materials.     

Quantitative estimations in neutron capture radiography

Theoretical models are proposed for the calculation of the local concentrations, in a specimen, of nuclides such as ⁶Li, ¹⁰B, ¹⁴N or ¹⁷O, from the corresponding local densities of tracks on the detectors. Several different cases have been studied: i) one or several types of particle involved, ii) initial energy of the particles below, or above the “upper threshold of detection”, iii) possible nonnegligible abrasion of the detectors during etching and iv) contribution of background tracks originating from the detectors themselves. The theoretical equations have been applied to the interpretation of experimental data.