失水事故

从大破口失水事故分析方法的发展过程，阐述了法国大破口失水事故分析方法DRM。

From the developing of Large Break LOCA analysis method , this paper introduces the French DRM ( Deterministic Realistic Methodology ) used for Large Break LOCA .

核反应堆地震和失水事故下结构动力分析研究

Research on Reactor Dynamic Analysis under Seismic and LOCA Conditions

ESDFT综合分析法在船用堆失水事故分析中的应用

ESD_FT Method in the Analysis of Marine Reactor Water-loss Accident

大破口失水事故的DRM分析方法介绍

Introduction of DRM for Large Break LOCA Analysis

PWR冷管段1%小破口失水事故实验研究

Experimental investigation on 1 % small break loss of coolant accident of pressurized water reactor cold leg

回流冷凝是压水堆发生失水事故（LOCA）时的重要传热方式之一。

Counterflow condensation is one of the important core cooling mechanisms during a loss of coolant accident ( LOCA ) of PWR .

3）参考的两环路核电厂严重事故的支配性初因依次为失水事故（LOCA）和丧失热阱。

3 ) the loss of coolant accident ( LOCA ) and loss of heat sink are the first and second initiating events respectively which could cause serious accidents .

ACR-700核电厂小破口失水事故分析

Analysis of Small Break LOCA for ACR-700 NPP

失水事故工况（LOCA）下反应堆下降环腔内的流动和传热研究，对反应堆压力容器（RPV）的安全具有重要的意义。

The study on flow and heat transfer in downcomer under loss of coolant accident ( LOCA ) is important to the safety of reactor pressure vessel ( RPV ) .

当核反应堆处于失水事故时，利用堆芯补水箱（CMT）内含硼水的重力特性对堆芯进行非能动注水，实现应急堆芯冷却。

When the nuclear reactor is in the state of leaking , the boric water 's gravity characteristics can be used to supply water to core passively , which can cool the core urgently .

对小破口失水事故（SBLOCA）及其研究状况进行了综述。

The Small Break Loss of Coolant Accident ( SBLOCA ) and its research art-of-work are reviewed in this paper .

应用GO-FLOW法分析在失水事故过程中紧急堆芯冷却系统的可靠性。

Based on GO-FLOW methodology , the reliability of emergency core cooling system ( ECCS ) in lose of coolant accident ( LOCA ) was analyzed .

在失水事故分析中首次提出了一种新的、先进的计算模型（真实堆芯裸露模型）和计算机程序（MCRLOCA）。

A new advanced calculation model ( True Core Uncovered Model ) of LOCA of PRC - II Reactor and the relevant code ( MCRLOCA ) are first put forward .

在高压综合实验装置（HPITF）上进行了压水堆冷管段2%小破口失水事故实验（NSB－6），破口方向为冷管段底部，破口面积为2%。

A Small Break Loss of Coolant Accident ( SBLOCA ) experiment NSB 6 was carried out at the High Pressure Integral Test Facility ( HPITF ) . The break was oriented at the bottom of the horizontal cold leg with its area of 2 % .

小破口失水事故喷放阶段瞬态热工水力特性的模拟试验研究

Transient thermohydraulic characteristics under blowdown conditions simulating a small break loca

重点介绍了大破口失水事故分析，给出了分析计算的主要假设条件和分析计算结果。

Main assumptions and calculation results are provided in detail .

60万千瓦核电站大破口失水事故中的喷射力

Blowdown Force During the Large LOCA for 600 MW Nuclear Power Plant

小破口失水事故研究综述

Review on research of small break loss of coolant accident

上空腔小破口失水事故模拟实验

Simulation Experiment of Small Break LOCA in Upper Plenum

核电厂主设备在地震加失水事故下的结构反应分析研究

Structural Response Research for Primary Components under Earthquake and LOCA Accidents on NPP

脉冲反应堆失水事故分析

Analysis of Loss of Coolant Accident for Pulsed Reactor

利用可选择源项分析大破口失水事故的放射性后果

Radiological Consequences in Large Break Loss of Coolant Accident With Alternative Source Term

基于应急运行的核动力装置一回路小破口失水事故

On small-break loss-of-coolant accident in marine nuclear power plant based on emergency operation

核动力装置一回路辅助系统失水事故的仿真研究

Simulation of LOCA in the auxiliary primary loop system of nuclear power plant

大亚湾核电站18个月换料非失水事故分析

GNPS 18-Months Fuel Cycles Non - LOCA Accident Analyses

破口面积对小破口失水事故瞬态热工水力特性的影响

The influence of break area on transient thermohydraulic characteristics of a small break loca

小破口失水事故处理中人员的行为可靠性分析

Operator Reliability Analysis during NPP Small Break LOCA

新型转换堆失水事故模拟实验的验算和分析

The analysis and checking calculation of LOCA simulating test in an advanced thermal reactor

反应堆失水事故条件下热管段内的两相流逆向流动研究

Research on Two-phase Countercurrent Flow in Reactor Hot Leg at the Loss of Coolant Accident

三峡升船机船厢失水事故动态特性研究

Study on dynamic properties of ship-lift during water-losing from ship chamber of Three Gorges Project