位置：成果数据库 > 期刊 > 期刊详情页
Effects of power level on thermal-hydraulic characteristics of steam generator
  • ISSN号：0149-1970
  • 期刊名称：Progress in Nuclear Energy
  • 时间：2015.5
  • 页码：245-253
  • 相关项目：先进核能系统热工、物理及其耦合
作者： Zhang, Rui|Tian, Wenxi|Qiu, Suizheng|Su, Guanghui|
同期刊论文项目
先进核能系统热工、物理及其耦合
期刊论文 55 会议论文 24 著作 2
同项目期刊论文
An improved mechanistic critical heat flux model and its application to motion conditions
Development of a thermal-hydraulic analysis software for a passive residual heat removal system
Development of VTSAS 1.0 and application to an IPWR
Preliminary transient thermal-hydraulic analysis for new coated UN and UC fuel options in SCWR
Study of traveling wave reactor (TWR) and CANDLE strategy: A review work
Three-dimensional study on steady thermohydraulics characteristics in secondary side of steam genera
Entrainment at T-junction: A review work
Development of a thermal hydraulic code for an integral reactor
Analysis of Westinghouse MB2 test using the Steam-generator Thermohydraulics Analysis code STAF
Development of CHF models for inner and outer RPV gaps in a meltdown severe accident
Comparative study on effect of air-water and steam-water mediums on liquid entrainment through ADS-4
Prediction of CHF in vertical heated tubes based on CFD methodology
The development of a zirconium oxidation calculating program module for Module In-vessel Degraded An
Development of TSACO and application to Chinese HCCB TBM cooling system
Study on secondary side flow of steam generator with coupled heat transfer from primary to secondary
Severe accident analysis for a typical PWR using the MELCOR code
Research on the leak-rate characteristics of leak-before-break (LBB) in pressurized water reactor (P
Development of a transient thermal-hydraulic code for analysis of China Demonstration Fast Reactor
Preliminary design study of a board type radial fuel shuffling sodium cooled breed and burn reactor
CFD analysis on subcooled boiling phenomena in PWR coolant channel
Minor actinide transmutation in a board type sodium cooled breed and burn reactor core
Preliminary safety analysis of the PWR with accident-tolerant fuels during severe accident condition
Effects of turbulence models on forced convection subcooled boiling in vertical pipe
COPRA experiments on natural convection heat transfer in a volumetrically heated slice pool with hig
钠冷行波堆TP-1瞬态安全分析
Analysis of PWR RPV lower head SBLOCA scenarios with the failure of high-pressure injection system u
Preliminary study of coupling CFD code FLUENT and system code RELAP5
Comparison of CORA & MELCOR core degradation simulation and the MELCOR oxidation model
Analysis of the loss of pool cooling accident in a PWR spent fuel pool with MAAP5
Depressurization study of supercritical fluid blowdown from simple vessel
Thermal-hydraulic design of water-cooled pressure tube blanket for a fusion driven subcritical react
Experimental study on spray characteristics of pressure-swirl nozzles in pressurizer
Analysis of accidental loss of pool coolant due to leakage in a PWR SFP
Development of a MCNP-ORIGEN burn-up calculation code system and its accuracy assessment
A theoretical CHF model for downward facing surfaces and gaps under saturated boiling
基于多孔介质模型的行波堆TP-1堆芯稳态温度场与流场数值模拟
径向和轴向倒料行波堆概念设计研究
TOPAZ-Ⅱ改进型热管辐射换热器传热单元数值研究
硬壳对熔融池换热特性影响试验研究
遗传神经网络对水平通道流动沸腾传热系数的预测
压水堆蒸汽发生器一、二次侧稳态流场耦合分析
蒸汽发生器传热管束过冷沸腾区两相流动数值模拟
耦合一、二次侧换热的蒸汽发生器二次侧流场分析
基于多孔介质模型的压水堆堆芯温场数值模拟
压水堆燃料棒束通道内过冷沸腾分析