PWR

Real Time Simulation and Safety Analysis Code for the PWR Core

压水堆堆芯安全分析程序实时仿真的分析研究

Sealing Analysis and Test for Nuclear Vessel of PWR

压水堆核容器的密封分析与试验

It can be applied to the design and core reloading calculation of PWR.

表明所提出的理论模型是正确的,软件包的研制是成功的,可以应用于PWR的设计、换料计算.

The load following of pressurized water reactor ( PWR) is studied. A three-dimensional fuzzy control system of boron concentration is designed.

对压水堆负荷跟踪运行进行了研究，提出将一种三维模糊控制系统应用于硼浓度自动调节的设计方案。

There are four types for PWR pressure-relieving devices: direct acting safety valve, pilot-operated pressure relief valve, power-operated pressure relief valve and safety valve with auxiliaries.

PWR的压力释放装置主要有四种类型：直接作用式安全阀、先导操作压力释放阀、动力作用式压力释放阀和具有辅助装置的安全阀。

The drive is different from electric-magnetic drive for PWR or hydraulic drive for BWR.

这种传动不同于压水堆上使用的电磁 - 机械传动,也不同于沸水堆上使用的液压传动.

Southwest Center of Reactor Engineering has made the conceptual design of a 600 MWe advanced PWR nuclear power plant ( AC-600).

西南反应堆工程研究设计院对600MWe先进压水堆核电厂（AC-600）进行了概念设计。

New reactor for nineties: on improvement of inherent safety for PWR

90年代新堆型：关于提高压水堆固有安全性的问题

The major aging problems in a PWR plant and their assessment technologies were summarized as the radiation embrittlement of reactor pressure vessel materials, thermal aging of cast stainless steel, stress corrosion cracking of steam generator tubing and metal fatigue in primary circuit of a PWR plant.

主要的老化机理有反应堆压力容器的辐照脆化、铸造不锈钢的热老化、蒸汽发生器传热管的应力腐蚀开裂和金属疲劳。因而可能是一种潜在的辐射防护剂。

The hybrid breeders can provide plenty of fissile material for PWR or FBR.

聚变-裂变混合堆可以为压水堆或快堆提供充足的核燃料.

Control rods are used in a PWR typically to change the power level and also to shut down the reactor.

控制棒是在一个加压水冷却反应堆里使用的，用来改变能量的级别，也可以用来停止反应堆。

There exists probability of fuel assembly damage in PWR during refueling period.

在压水堆核电站换料的过程中存在燃料棒破损的可能.

In-service Inspection of Primary Loop's Equipment of PWR Nuclear Power Plant in Franch

法国核电厂压水堆型核锅炉一回路设备的在役检查

Presented is a model for the multi region and variable node PWR core.

建立了多区变节点压水堆堆芯的数学物理模型.

Experimental Study of the Natural Vibration of a Model Core Barrel with Thermal Shield of PWR

压水堆吊篮、热屏结构模型在空气中的固有振动试验分析

A real time model of simulation of secondary system for PWR nuclear power unit is presented.

介绍了核电机组二回路主蒸汽与给水回热系统实时仿真.

Influence of Structure Improvement of Guide Tubes and Bundles in Pressurized Water Reactor ( PWR) on Drop of Control Rods

压水堆控制棒导向筒结构改进对落棒的影响

Casting duplex stainless steels for primary coolant loop in PWR is safe during the service life of 40 years.

保守估计核电一回路管道用铸造双相不锈钢在预定的服役时间40年内是安全的。

The modeling problem of PWR nuclear power station core is expound, critical reactor neutron dynamics model, temperature effect neutron dynamics model and core transfer of heat model is built up apply to microcomputer artificial nuclear power station.

阐述了PWR核电站堆芯的模型化问题，建立了适用于微机仿真的核电站的临界堆中子动力学模型、温度效应中子动力学模型和堆芯热传递模型。

Specific Radioactivity Produced by the Secondary Charged Particle Nuclear Reaction in the Primary Circuit Water of PWR

压水反应堆一回路水中二次带电粒子核反应产物的比活度

Preliminary study of physical properties for hexagon lattice PWR Core Using thorium

压水堆加钍（ThO2）六角形组件堆芯物理特性的初步探讨

Study on the Bootstrap PWR AMP Working in Common Emitter

对自举式功率放大器共射工作的分析研究

Theoretical and experimental stress analysis of pressure vessel of PWR

30万千瓦核电站压水堆压力容器的应力分析和试验

Equation of electric price in connection-grid of PWR nuclear power plant wes proposed, moreover, several values of fixed investment, annual interest rate and fund recycle period were assumed.

本文提出了压水堆核电厂上网电价方程，并对2×600MW核电机组的固定投资、年利息率和资金回收期假定了几种不同数值，在此基础上进行了上网电价计算。

This paper studied the modeling and controlling of PWR and developed simulation software as well.

本文以大型压水堆核电站为对象,研究了核电站的建模控制以及仿真软件的开发.

Operation Characteristics Study of Modular High Temperature Gas-Cooled Reactor Nuclear Power Plant Optimization of Leakage Calculation in RCP System of PWR

模块式高温气冷堆核电站的运行特性研究压水堆核电厂一回路冷却剂泄漏率计算的优化

The comparison of the computational results for reactor core of PWR and BWR are also presented.

文中还给出了压水堆和沸水堆堆芯计算实例的比较.

A Simulated Test of Physical Starting and Reactor Physics on Zero Power Facility of PWR

压水堆零功率装置上的堆物理和物理启动模拟试验

Steam generator ( SG) is one of main equipments in the pressurized water reactor ( PWR) nuclear power plant, but also it is a pivotal equipment which brings heat from reactor core to the secondary loop coolant by primary loop water.

蒸汽发生器是压水堆核动力装置中的主要设备之一，也是把一回路冷却剂从反应堆堆芯带出的热量传给二回路水的关键设备。

A study on some problems of thermal design in PWR of nuclear power plant

核电站压水反应堆热工设计中若干问题的探讨