物料动态在线水分测量装置和测量方法

本发明涉及一种利用中子及了射线技术进行物料动态在线水分测量的装置和方法。装置上由于增加了了射线电离室探测器，使原来对皮带输送机中部物料一点测量变为对物料整个横断面测量，更有利于矫正物料形状对水分测量的影响，同时将慢中子探测器与快中子发生器放在同一平面上，变已有的透射式为散射式，提高探测区域慢中子通量，提高水分测量的准确度及灵敏度；在测量方法上合理的探测器布局及数学模型可使得水分测量不受物料密度、厚度、形状变化的影响，提高测量精确度。检测水分误差小于0．5％。利用检测到的水分值去控制生产设备，实现生产过程中物料水分的自动调节，可以保证产品质量，提高生产效率，给用户带来显著的经济效益。     

放射性固体废物压缩减容技术研究

本工作对以压缩作为处理放射性固体废物的方法进行了实验研究,分析了压缩过程中影响各类物料压缩减容倍数的因素。实验结果表明,高压压缩处理物料的范围广,对物料的分类要求不严格;在压力为600kN压缩下,依据物料的性质和装填密度不同,各种物料的减容倍数在2.3~11之间,压缩后各类物料压缩的平均物料密度在1 200~3 900kg/m3之间;对于某些软物料、有孔硬物料和弹性物料,压缩后物料的平均密度大于物料的物理密度;不同装填密度的废物,高压压缩后物料的平均密度基本相同,物料的压缩减容倍数与物料本身的性质、装填密度等有关。弹性物料压缩后反弹明显,甚至破坏压饼,其他物料反弹在20%以内。     

TGS装置的算法研究

正层析γ扫描(TGS)技术是当前核材料非破坏性分析(NDA)技术中重要的分析技术之一。它专用于准确测量中低密度非均匀分布介质中的放射性核素及其含量,是核设施中可回收物料及核废物测量分析的主要方法之一。     

有源中子计数法测量铀总量方法研究

含铀物料中铀总量是核材料衡算与控制中最关注的信息之一,因非破坏性测量分析方法(NDA)不破坏样品,不产生二次废物,可对样品整体测量,因此常用作含铀物料中铀总量的测量分析.本研究基于有源中子计数法建立测量含铀物料中铀总量的新方法.对10组铀的标准样品进行测量分析,结果表明,测量误差为1.7％,小于5％,达到非破坏性分析技...     

中子水分仪测量方式的改进研究

中子水分仪在工业现场应用时,有时会因被测量物料中水分不均匀而引起测量误差。本工作利用蒙特卡罗方法模拟了中子水分仪应用中由于物料中水分不均匀所带来的测量误差。为了减小由于水分不均匀带来的测量误差,提出了一种新型的测量方式,即透射加散射测量方式。实验证明,透射加散射测量方式能够有效降低中子水分仪在被测量物料水分分布不均匀情况下的测量误差。该方法用于水分分布复杂条件下的现场测量还需作进一步的分析和研究。     

有源符合中子法测量材料中的高浓铀

研究了有源符合中子法测量不同试样中高浓铀的方法和技术。将１个高计数率符合中子计数器改装为有源井型符合中子计数器装置，测量由Ａｍ－Ｌｉ中子源发射的中子诱发物料中的２３５Ｕ裂变时产生的裂变中子，以确定物料中２３５Ｕ含量。用Ｕ３Ｏ８粉末工作样品进行了探测效率刻度，在５００ｇＵ３Ｏ８粉末范围内，以快中子方式测量的刻度曲线近似线性。实测了含铀粉末不同形态、氧化程度各异的多种试样，给出了试样的形态、密度、碳含量及轴向和径向分布不均匀性等因素引起的各种误差，并进行了实验校正。实验结果表明：测量误差在４％—１１％之间，大部分数据与常规重量法在９％内符合。试样中碳含量对测量结果影响较大。     

核径迹热释中子探测器测量高中子通量密度

本文提出一种用于高中子通量密度测量的方法,即使用核径迹热释中子探测器测量中子通量密度,该方法在低中子通量密度测量方面已成功在微型中子源反应堆上得到验证。为了测试其在高中子通量密度测量方面的适用性,在中国先进研究堆辐照孔道内进行了应用研究。结果表明:孔道内中子通量密度相对分布总体趋势与MCNP的计算结果符合较好,此种方法测量高中子通量密度有效可行。     

基于LabVIEW的核子秤计量系统

核子秤利用低能γ射线透过物料时被物料吸收衰减的规律来测量物料重量,是一种非接触式的连续计量装置,具有结构简单、运行可靠等优点.LabVIEW作为一种图形化的编程语言,是一种标准的数据采集和仪器控制软件,论文针对核子秤测量原理和特点,利用LabVIEW编程环境设计了一种核子秤计量系统监控软件,软件构架主要由监控软件、数据库和Web服务三大基本模块组成,实现了对核子秤计量数据采集、状态监控、数据管理及远程监控的功能.     

一个动态不连续物料的在线中子测水系统

介绍了一个用γ密度计作密度补偿的用于测量在传送带上处于运动状态的一块一块不连续物料中水分的在线中子测水系统。来自中子、γ射线及电子秤探测器的实测数据均通过一个计数卡输入到ＰＣ机中，由微机作数据处理、记录和显示，并用表格和图形给出各时刻的瞬时湿度及某段时间间隔内通过传送带的物料的总重量、干料量和水分量。     

蓄电池电解液密度测量仪的研制

研究了采用低能γ射线穿透方法检测蓄电池液密度的可行性.设计和制作了样机.密度的测量范围为1.000～1.300g/cm3,在线测量的绝对值误差好于±0.005g/cm3.     

Theory of heterogeneous reactors with cylindrical lumps of a finite radius

This article presents a consistent theory of heterogeneous reactors with cylindrical lumps of a finite, but small, radius. The density of thermal neutrons inside a lump and in the moderator is described by taking into account the azimuthal dependence. The diffusion theory is applied to the entire reactor volume. Simple expressions for the diffusion length in the parallel and perpendicular directions with an accuracy to the first power of the lump surface to cell surface ratio are obtained. It is shown that the average scattering length for the perpendicular direction depends on the lump shape. A simple method for determining the diffusion tensor for the case of weak absorption and large spacing between the lumps is considered in section 4.The author is indebted to B. I. Il'ichev, N. I. Laletin, and Ya. V. Shevelev for the discussion of this paper.     

利用多道幅度分析器测量线性变化电压的非线性系数

本文描述一个线性变化电压非线性系数的测量方法。利用采样脉冲对线性变化电压进行采样,测量采样后脉冲的幅度分布密度,经过最小二乘法拟合,便可求出非线性系数。文中还提出了决定采样脉冲频率的方法。     

Calculations of positron absorption in targets

Using the multiple-step method for calculation of transmission of charged particles through matter, the density distributions of thermalized positrons in a number of materials (6 ? Z ? 42) have been obtained under irradiation by monoenergetic positrons and positrons from ²²Na and ⁵⁸Ga sources. Fluctuations of energy losses, absorber boundaries, annihilation of positrons have been taken into account. Elastic scattering of the positrons was calculated according to Mott's theory. One can use the obtained data when constructing similar theoretical distributions and when determining the positron annihilation probabilities for a wide range of materials involving compounds.     

Prediction Method in Numerical Simulation of Dynamics Based on Prompt and Delayed Neutrons in Multiplying Systems

The basic elements of a method of numerical simulation of processes based on prompt and delayed neutrons in multiplying systems are presented. The method is based on predicting the contribution of the instantaneous state of the system to its state at subsequent definite times and summing the predicted contribution as a systematic transition is made to a new moment in time. The key element of the method is determining the initiation functions – the probabilities that a neutron emitted by a source in a prescribed volume of the system initiates a prompt fission neutron in a definite volume of the system sometime after emission. A procedure is proposed for determining the initiation functions; this procedure is based on the first-collision probability method and uses the standard stationary computer codes. The material presented in this paper comprises the basic results of the first stage in the development of the numerical model for full-scale simulation of the dynamics of the subcritical blanket of an experimental accelerator-driven subcritical system under construction at the Institute of Theoretical and Experimental Physics. It is noted that the model being developed can be used to analyze many important processes in other types of multiplying systems.     

SHTV,as a technique for core calculation using spatial homogenization and temperature variation

The accuracy of static neutronic parameters in the nuclear reactors depends upon the determination of group constants of the diffusion equation in the desired geometry. Although several methods have been proposed for calculating these parameters, there is still the need for more reliable methods. In this paper a powerful and innovative method based on Spatial Homogenization and Temperature Variation (SHTV) of physical properties of a WWER-1000 nuclear reactor core for calculating the relative power distribution of Fuel Assemblies (FA) and the hot fuel rod, is presented. The method is based on replacing the heterogeneous lattices of materials with different properties by an equivalent homogeneous mixture of these material for determining the few group constants, while the effect of temperature variation in the fuel and coolant density along the axial core direction is considered. All calculations are performed using WIMS and CITATION codes. The obtained results are compared with the results of Final Safety Analysis Report (FSAR) prepared by the designer, and good agreement between the two results is shown.     

HTGR燃料芯核及包覆颗粒的物性测试

本文介绍用图象分析技术对高温气冷堆燃料芯核的直径、密度和球形度,以及包覆燃料颗粒的疏松碳层,致密碳层和碳化硅层的涂层厚度和涂层密度的定量测试方法。本方法快速、准确,测量相对标准偏差为1%。     

Development of a finite volume inter-cell polynomial expansion method for the neutron diffusion equation

Heterogeneous nuclear reactors require numerical methods to solve the neutron diffusion equation (NDE) to obtain the neutron flux distribution inside them, by discretizing the heterogeneous geometry in a set of homogeneous regions. This discretization requires additional equations at the inner faces of two adjacent cells: neutron flux and current continuity, which imply an excess of equations. The finite volume method (FVM) is suitable to be applied to NDE, because it can be easily applied to any mesh and it is typically used in the transport equations due to the conservation of the transported quantity within the volume. However, the gradient and face-averaged values in the FVM are typically calculated as a function of the cell-averaged values of adjacent cells. So, if the materials of the adjacent cells are different, the neutron current condition could not be accomplished. Therefore, a polynomial expansion of the neutron flux is developed in each cell for assuring the accomplishment of the flux and current continuity and calculating both analytically. In this polynomial expansion, the polynomial terms for each cell were assigned previously and the constant coefficients are determined by solving the eigenvalue problem with SLEPc. A sensitivity analysis for determining the best set of polynomial terms is performed.     

Real-time measurement of implanted deuterons by using the nuclear reaction H(d,p)H

We propose a method for real-time measurement of the density of implanted deuterons in solid materials during deuterium implantation. In this method, the density of implanted deuterons is measured by detecting the emitted protons from the deuteron fusion reaction between the incident deuterons and the implanted deuterons. A negative deuterium beam with an energy of 90 keV is used both for the implanted atoms and a probe beam to measure the implanted deuterons. The density of the implanted deuterium atoms was observed to be drastically different between the single crystal and poly-crystal in Ta.