中子通量密度

根据中子通量密度的对数功率测量装置

Logarithmic power measuring assembly based on the neutron flux density

核电站反应堆中子通量密度的一种预测控制方法

A predictive control method of the neutron flux density for a nuclear reactor

该方法在时间方面，中子通量密度按时间二阶展开（QEM）；在空间方面，采用Green函数节块法（NGFM）。

In the Temporal Quadratic Expansion Nodal Green 's Function Method ( TQE / NGFM ), the Quadratic Expansion Method ( QEM ) is used for the temporal solution with the Nodal Green 's Function Method ( NGFM ) employed for the spatial solution .

反应堆功率运行时，燃耗变化会引起堆外中子通量密度变化，造成RPN核功率测量系统测得的反应堆功率与实际功率出现偏差。

When the reactor is operating , the variation of the burn-up of the fuel will cause vari-ation of the neutron flux , which can make the core power measured by the RPN nuclear power system deviate from the actual power .

六角形轻水堆组件中子通量密度分布的计算

Calculation for the Flux Distribution in Hexagonal fuel Assembly of LWR

固体径迹探测器测量束流装置内的中子通量密度

Neutron flux measurement in neutron beam equipment by SSNTD

界面流法计算反应堆六角形燃料组件中子通量密度分布

Calculation of Flux Distribution in Hexagonal LWR Fuel Assembly by Interface Current Method

低中子通量密度试验性反应堆

Dielectric ( flux ) density low intensity test reactor

14MeV中子通量密度的测量

Measurement of 14 MeV neutron fluence dencity

反应堆中子通量密度仿真研究

Simulation Research of Reactor Neutron Flux Density

中子通量密度扫描装置

Neutron flux density scanning assembly

用该方法对压水堆中子通量密度进行求解，取得了令人满意的结果，并已在核电厂仿真中得到应用。

The result is satisfied and this method has been applied in nuclear power plant simulation system .

中子通量密度指示器

Neutron flux density indicator

在东风3号上开展加速器驱动次临界系统有效增殖系数和中子通量密度分布研究

Research of Effective Neutron Multiplication Factor and Neutron Flux Distribution for Accelerator Driven Sub-Critical System on DF-3 Facility

利用界面流法计算两维六角形轻水堆燃料组件内中子通量密度分布。

This paper describes the model of calculating the flux distribution in two dimensional hexagonal geometry light water reactor fuel assembly by interface current method .

同时，测量了对应功率下反应堆内辐照座的热中子通量密度，得到单位功率的热中子通量密度，即额定中子通量密度下的运行功率。

Meanwhile , the thermal neutron flux density at certain point is measured and then the thermal neutron flux density per unite power is given .

论文分为六章，第一章为中子通量密度时空分布与中子动力学研究及相关研究工作；第二章介绍了微机化多时空通道中子通量密度数据采集系统的总体设计方案；

The thesis is divided into six chapters : Chapter 1 introduces the research work of neutron flux temporal and spatial distribution and neutron dynamics ;

利用金箔活化法测量了特殊束流装置中心轴线上不同位置处的热中子通量密度比、金镉比、锰镉比以及出口处γ/n的剂量比。

The neutron flux density ratio , Au-Cd ratio , Mn-Cd ratio at various positions in special neutron beam equipment are measured by gold foil method .

六角形节块内的中子通量密度分布采用高次多项式近似表示，最后导出通量矩方程及偏流的响应矩阵方程。

The nodal equations and response matrix equations are derived using higher order polynomial approximations to the spatial dependence of the flux within the hexagonal-z node .

针对核反应堆点堆动态非线性模型，提出了一种非线性状态反馈的中子通量密度恒值控制的新方法。

According to the nonlinear dynamic model of a nuclear reactor , a new constant neutron flux density control method based on nonlinear state feedback is presented .

中子通量密度采用固体核径迹探测器测量，γ辐射剂量率用热释光探测器测定。

The thermal neutron flux density is measured with Soild State Nuclear Track Detector ( SSNTD ), and the γ radiation dose rate is detected with Thermoluminescence Dosimeter ( TLD ) .

本论文是研制微机化多时空通道中子通量密度数据采集系统的技术基础，将为最终实现微机化多时空通道中子通量密度数据采集系统提供技术支持。

This thesis is the technology elements for development of the data acquisition system of neutron flux with multiple space-time channels based on a microcomputer , that provide technology sustain to realize the data acquisition system of neutron flux with multiple space-time channels based on a microcomputer .

中毒法测量微堆堆芯热中子绝对通量密度

Measurement of absolute thermal neutron flux density in the core of miniature neutron source reactor

小型密封式中子发生器中子通量密度的水平方向分布

Neutron Flux Density of Horizental Direction Distribution for Sealed Small-Sized Neutron Generator

用MCNP程序计算核燃料废包壳缓发裂变中子形成的热中子通量密度

Computing with MCNP the thermal flux density due to the delayed fission neutron generated in processing the clads of burned-up fuel roads

文中提出：在用芯内中子探测器读数重构出测量的全堆热群中子通量密度分布的基础上，约束堆芯截面参数，并采用节块格林函数法对其进行校准的一套完整方法。

Based on the " measured " neutron flux density distribution reconstructed by using harmonics synthesis method ( HSM ) from the readings of in-core neutron detectors and some necessary restrictions of core parameters , a procedure is proposed to calibrate the core parameters after burn-up .