A Two-Fluid Two-Phase Model for Thermal-Hydraulic Analysis of a U-Tube Steam Generator

1993 ◽  
Vol 101 (2) ◽  
pp. 227-236
Author(s):  
Huan-Jen Hung ◽  
Ching-Chang Chieng ◽  
Bau-Shei Pei ◽  
Song-Feng Wang
Keyword(s):  
Steam Generator ◽  
Phase Model ◽  
Hydraulic Analysis ◽  
Two Phase ◽  
Thermal Hydraulic Analysis ◽  
Two Fluid

Computational Analysis of Downcomer Boiling Phenomena Using a Component Thermal Hydraulic Analysis Code CUPID

2010 ◽  
Author(s):  
H. K. Cho ◽  
B. J. Yun ◽  
I. K. Park ◽  
J. J. Jeong
Keyword(s):  
Nuclear Reactor ◽  
Reactor Vessel ◽  
Physical Models ◽  
Injection System ◽  
Hydraulic Analysis ◽  
Two Phase ◽  
Rate Condition ◽  
The Core ◽  
Benchmark Calculation ◽  
Thermal Hydraulic Analysis

A component scale thermal hydraulic analysis code, CUPID (Component Unstructured Program for Interfacial Dynamics), is being developed for the analyses of components of a nuclear reactor, such as reactor vessel, steam generator, containment, etc. It adopts three-dimensional, transient, two-phase and three-field model, and includes various physical models and correlations of the interfacial mass, momentum and energy transfer for the closure relations of the two-fluid model. In the present paper, the two-phase models were assessed against the DOBO (DOwncomer BOiling) experiment, which was constructed to simulate the downcomer boiling phenomenon. It may happen in the downcomer of a nuclear reactor vessel during the reflood phase of a postulated loss of coolant accident. The stored energy release from the reactor vessel to the liquid inside the downcomer causes the boiling on the wall, and it can reduce the hydraulic head of the accumulated water, which is the driving force of water reflooding to the core. This phenomenon has been considered as a crucial safety issue of an advanced power reactor because it is concerned with the core cooling capability of the safety injection system. In this paper, the physical models and correlations that were incorporated into the CUPID code were introduced and the validation results against the experiment were reported. The benchmark calculation results concluded that the CUPID code can appropriately predict the boiling phenomena under a low pressure and low flow rate condition with modification of the bubble size correlation.

Thermal-hydraulic analysis code development for sodium heated once-through steam generator

2019 ◽  
Vol 127 ◽  
pp. 385-394 ◽  
Author(s):  
Rongshuan Xu ◽  
Dalin Zhang ◽  
Wenxi Tian ◽  
Suizheng Qiu ◽  
G.H. Su
Keyword(s):  
Steam Generator ◽  
Hydraulic Analysis ◽  
Thermal Hydraulic Analysis ◽  
Code Development

Development and Application of a UTSG Thermal-Hydraulic Analysis Code

2014 ◽  
Author(s):  
Tenglong Cong ◽  
Guanghui Su ◽  
Wenxi Tian ◽  
Suizheng Qiu
Keyword(s):  
Steam Generator ◽  
Structural Integrity ◽  
Tube Bundle ◽  
Flow Boiling ◽  
Three Dimensional ◽  
Hydraulic Parameters ◽  
Two Phase ◽  
Drift Flux ◽  
Thermal Hydraulic Analysis ◽  
Secondary Side

Structural integrity of steam generator should be maintained during operation, since it performs as the pressure and heat transfer boundary of primary side coolant. Localized thermal-hydraulic parameters of secondary side are essential for the analysis of tube wastage, fatigue and failure. In this paper, a three-dimensional thermohydraulics analysis code, named STAF, is developed based on FLUENT. With STAF code, three-dimensional thermohydraulics of secondary side of AP1000 steam generator are generated. This code is developed based on the porous media theory. In this code, the drift flux two-phase model coupled with a simplified flow boiling model is utilized to present two-phase flow among the U-tube bundle. Downcomer, tube bundle, support plates and primary separators in steam generator are considered in STAF code. The calculated results are compared with a general steam generator thermohydraulic analysis code ATHOS, which is developed by EPRI steam generator group. The comparison indicates that STAF code performs well in evaluating thermal-hydraulic parameters in steam generator. The results show that the flow field varies significantly at different position in AP1000 steam generator. Flow vapor quality at the inlet of primary separators varies significantly, which is a severe challenge to the capacity design of separators.

Thermal hydraulic analysis of two phase helium flow through hydro formed cryopanel

2016 ◽  
Vol 41 (1) ◽  
pp. 124
Author(s):  
Reena Sayani ◽  
Samiran Shanti Mukherjee ◽  
Ranjana Gangradey
Keyword(s):  
Hydraulic Analysis ◽  
Helium Flow ◽  
Two Phase ◽  
Flow Through ◽  
Thermal Hydraulic Analysis

Two-Phase Thermal-Hydraulic Analysis Using Parity Simulation

1990 ◽  
Vol 104 (2) ◽  
pp. 169-182
Author(s):  
Eduardo V. Depiante ◽  
John E. Meyer
Keyword(s):  
Hydraulic Analysis ◽  
Two Phase ◽  
Thermal Hydraulic Analysis
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