F210 Development of Thermal Hydraulic Computer Code for Steam-Water Flow in Steam Generator of Fast Breeder Reactor

2008 ◽  
Vol 2008.13 (0) ◽  
pp. 495-496
Author(s):  
Ryuji YOSHIKAWA ◽  
Hirotsugu HAMADA ◽  
Hiroyuki OHSHIMA ◽  
Hideki YANAGISAWA
Keyword(s):  
Water Flow ◽  
Steam Generator ◽  
Computer Code ◽  
Fast Breeder Reactor

Ultrasonic Inspection of Liquid-Metal Fast Breeder Reactor Steam Generator Duplex Tubing

1980 ◽  
Vol 47 (2) ◽  
pp. 257-267
Author(s):  
K. J. Longua ◽  
G. K. Whitham ◽  
C. C. Allen
Keyword(s):  
Liquid Metal ◽  
Steam Generator ◽  
Ultrasonic Inspection ◽  
Fast Breeder Reactor

Development of a Thermal-Hydraulic Analysis Code for Helical Coiled Once Through Steam Generator

2018 ◽  
Author(s):  
Jun Huang ◽  
Junli Gou ◽  
Haifu Ma ◽  
Jie Fan ◽  
Jianqiang Shan
Keyword(s):  
Heat Transfer ◽  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
High Efficiency ◽  
Fluid Model ◽  
Computer Code ◽  
Medium Size ◽  
Hydraulic Characteristics ◽  
Pressurized Water

Due to their advantages, such as compactness and high efficiency in heat transfer, helically coiled heat exchangers have been widely used by different type of nuclear power plants, especially by small and medium size reactors (SMRs). In order to analyze the thermal-hydraulic characteristics of a helical coiled once through steam generator (OTSG) for a small integral pressurized water reactor, a computer code is developed in this paper. The code is based on two-fluid model. The constitutive correlations are recommended based on the assessments with the compiled databases from the reviewed literatures. NUSOL SG is validated and verified against heat transfer in helical coiled tubes, and the calculation results agree well with the experiment data. The present study could provide references for the investigators to perform further investigations on the thermal hydraulic characteristics of helical coiled OTSGs.

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