Two-phase flow mechanism model and dynamic simulation of steam generator in nuclear power plant

2018 ◽  
Author(s):  
Yongling Li ◽  
Lihui Zhao ◽  
Shujun Ma
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Dynamic Simulation ◽  
Steam Generator ◽  
Nuclear Power ◽  
Two Phase Flow ◽  
Phase Flow ◽  
Flow Mechanism ◽  
Two Phase ◽  
Mechanism Model

CFD Investigating the Changes of FAC Wear Sites Due to the Power Uprate of Nuclear Power Plant

2009 ◽  
Author(s):  
Tsun Fu Hung ◽  
Yuh Ming Ferng ◽  
Bing Hong Lin ◽  
Chunkuan Shih ◽  
Bau-Shei Pei
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Two Phase Flow ◽  
Control Volume ◽  
Flow Characteristics ◽  
Piping System ◽  
Phase Flow ◽  
Two Phase ◽  
Simulation Results

Changes of flow-accelerated corrosion (FAC) wear sites on the piping due to the power uprate of nuclear power plant are investigated by way of computational fluid dynamics (CFD) models. These models proposed in this paper include the three-dimensional two-phase flow models and appropriate FAC models. The computations are performed using commercial code Fluent 6.2 which is control-volume-based. A boiling water reactor (BWR), located at Taiwan, is selected in the present analytical works. Simulation results clearly reveal that the present model can precisely capture the two-phase phenomena within the piping system. Coupled with the calculated two-phase flow characteristics, the appropriate FAC indictors can predict the local distributions of severe FAC sites. These predicted results show reasonable agreement with the plant measurements. Therefore, the impacts of power uprate on the changes of wear sites can be confidently investigated by the present CFD model. Through the comparisons of predictions for the selected BWRs under 100%, 105%, and 110% power levels, the simulation results clearly reveal that the power uprate does not significantly change the characteristics of FAC wear sites.

VVER-440 Steam Generator’s Two-Phase Flow Analysis

2016 ◽  
Vol 821 ◽  
pp. 57-62
Author(s):  
Lukas Joch ◽  
Roman Krautschneider
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Steam Generator ◽  
Nuclear Power ◽  
Flow Analysis ◽  
Three Dimensional ◽  
Reactor Power ◽  
Physical Models ◽  
Two Phase ◽  
Ansys Fluent

The subject of this report is creation of three-dimensional thermal hydraulic model of horizontal steam generator for Dukovany nuclear power plant. A procedure is presented for simulation and analysis of secondary side of PGV-440 steam generator for nominal and increased reactor power. A two-fluid approach is applied for modeling physical processes inside the steam generator. Physical models were implemented in ANSYS Fluent CFD environment using User Defined Functions (UDFs). Results from this thermal hydraulic numerical model can be used for various other subsequent nuclear power plant operations and safety analysis.

Two Phase Flow Analysis for Secondary Side in Steam Generator

2002 ◽  
Author(s):  
Y. Yagi ◽  
M. Murase ◽  
A. Nakamura ◽  
Y. Fujii
Keyword(s):  
Heat Transfer ◽  
Steam Generator ◽  
Nuclear Power ◽  
Two Phase Flow ◽  
Flow Analysis ◽  
Phase Flow ◽  
Two Phase ◽  
Pressurized Water ◽  
Friction Coefficients ◽  
Void Fractions

The steam generator (SG) in a pressurized water reactor (PWR) is an important component as boundary between the primary loop and the secondary loop. In this study, we performed two-phase flow analysis of SG reliability tests conducted by the Nuclear Power Engineering Corporation (NUPEC) using the two-fluid models of a thermal-hydraulic computer code PHOENICS. It was difficult to calculate the location of the boiling initiation accurately because the location was greatly affected by the friction coefficients (i.e. velocity distributions) and the heat transfer distributions. However, the friction coefficients and the heat transfer distributions did not greatly affect the void fractions in the upper region of the U-bent tubes and the calculated average void fractions agreed with the measured within 4%.

Time Domain Simulation of the Vibration of a Steam Generator Tube Subjected to Fluidelastic Forces Induced by Two-Phase Cross-Flow

2013 ◽  
Vol 135 (3) ◽  
Author(s):  
Téguewindé Sawadogo ◽  
Njuki Mureithi
Keyword(s):  
Steam Generator ◽  
Two Phase Flow ◽  
Work Rate ◽  
Fretting Wear ◽  
Phase Flow ◽  
Reference Case ◽  
Two Phase ◽  
Steam Generator Tube ◽  
Wide Range ◽  
Instability Regime

Having previously verified the quasi-steady model under two-phase flow laboratory conditions, the present work investigates the feasibility of practical application of the model to a prototypical steam generator (SG) tube subjected to a nonuniform two-phase flow. The SG tube vibration response and normal work-rate induced by tube-support interaction are computed for a range of flow conditions. Similar computations are performed using the Connors model as a reference case. In the quasi-steady model, the fluid forces are expressed in terms of the quasi-static drag and lift force coefficients and their derivatives. These forces have been measured in two-phase flow over a wide range of void fractions making it possible to model the effect of void fraction variation along the tube span. A full steam generator tube subjected to a nonuniform two-phase flow was considered in the simulations. The nonuniform flow distribution corresponds to that along a prototypical steam-generator tube based on thermal-hydraulic computations. Computation results show significant and important differences between the Connors model and the two-phase flow based quasi-steady model. While both models predict the occurrence of fluidelastic instability, the predicted pre-instability and post instability behavior is very different in the two models. The Connors model underestimates the flow-induced negative damping in the pre-instability regime and vastly overestimates it in the post instability velocity range. As a result the Connors model is found to underestimate the work-rate used in the fretting wear assessment at normal operating velocities, rendering the model potentially nonconservative under these practically important conditions. Above the critical velocity, this model largely overestimates the work-rate. The quasi-steady model on the other hand predicts a more moderately increasing work-rate with the flow velocity. The work-rates predicted by the model are found to be within the range of experimental results, giving further confidence to the predictive ability of the model. Finally, the two-phase flow based quasi-steady model shows that fluidelastic forces may reduce the effective tube damping in the pre-instability regime, leading to higher than expected work-rates at prototypical operating velocities.

scholarly journals Research Method and Two-Phase Flow Stability of the Steam Generator of HTR-10

2001 ◽  
Vol 38 (9) ◽  
pp. 739-744 ◽  
Author(s):  
Huaiming JU ◽  
Yuanhui XU ◽  
Zhiyong HUANG ◽  
Yu YU
Keyword(s):  
Steam Generator ◽  
Research Method ◽  
Two Phase Flow ◽  
Flow Stability ◽  
Phase Flow ◽  
Two Phase

Leak behavior of SCC degraded steam generator tubings of nuclear power plant

2005 ◽  
Vol 235 (23) ◽  
pp. 2477-2484 ◽  
Author(s):  
Seong Sik Hwang ◽  
Hong Pyo Kim ◽  
Joung Soo Kim ◽  
Kenneth E. Kasza ◽  
Jangyul Park ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Steam Generator ◽  
Nuclear Power

scholarly journals Aplicación de la Teoría de Perturbación – Método Diferencial- al Análisis de Sensibilidad en Generadores de Vapor de Centrales Nucleares PWR-Caso Angra I

2019 ◽  
pp. 119-126
Keyword(s):  
Perturbation Theory ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Steam Generator ◽  
Nuclear Power ◽  
Homogeneous Model ◽  
Differential Method ◽  
Differential Approach ◽  
Good Potential ◽  
Direct Calculations

Aplicación de la Teoría de Perturbación – Método Diferencial- al Análisis de Sensibilidad en Generadores de Vapor de Centrales Nucleares PWR-Caso Angra I Aplication of the Perturbation Theory- Differential Methodto Sensibility Análisis in PWR Nuclear Power Plant Steam Generator- Angra I Giol Sanders R, Andrade de Lima F, Marques A, Gallardo A, Bruna M, Zúñiga A Institución Peruano de Energía Nuclear Universidad Federal de Rio De Janeiro-Brasil DOI: https://doi.org/10.33017/RevECIPeru2011.0033/ RESUMEN En este trabajo basado en la tesis del Magíster Roberto Giol S. [1] presenta una aplicación del formalismo diferencial de la teoría de perturbación a un modelo termohidráulico homogéneo de simulación del comportamiento estacionario de uno de los generadores de vapor de la Central Nuclear tipo PWR Angra I del Brasil. Se desarrolla un programa de cálculo PERGEVAP tomando como base el código GEVAP de Souza[2]. El programa PERGEVAP permite realizar cálculos de sensibilidad de funcionales lineales (temperatura media del primario)y no lineales (flujo de calor medio a través de las paredes de los tubos del generador) con relación a las variaciones de ciertos parámetros termo-hidráulicos(flujo másico del primario, calor específico, etc), Los resultados obtenidos con este formalismo son luego comparados con los obtenidos del cálculo directo con el propio código GEVAP, pudiéndose verificar una excelente concordancia. Este método se muestra promisorio para efectuar cálculos repetitivos asociados al diseño y análisis de Seguridad de los componentes de las Centrales Nucleares. Descriptores: teoría de perturbación, método diferencial, sensibilidad, generador de vapor, central nuclear PWR. ABSTRACT This report presents an application of the differential approach of the perturbation theory to an homogeneous model of a PWR steam generator in the Angra 1 Nuclear Power Plan in Brazil under steady-state conditions. Program PERGEVAP was built fom the code GEVAP developed by Souza and allows sensitivity calculations of linear (average primary loop temperature) and non-linear (average heat flux) functionals due to variations in some thermo-hydraulics parameters (flow rate, specific heat, , etc). Results obtained with this approach are then compared with direct calculations performed using the GEVAP code, with excellent agreements. The method has good potential to treat repeated calculations needed in the design and safety analysis of the Nuclear Plant components. Keywords: perturbation theory, differential method, steam generator, PWR nuclear Power Plant.

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