Particle-laden flow around an obstacle in a square pipe: experiments and modeling

Particle trapping and deposition around an obstacle occur in many natural and industrial situations and in particular in the nuclear industry. In the steam generator of a nuclear power plant, the progressive obstruction of the flow due to particle deposition reduces the efficiency and can induce tube cracking leading to breaking and damage. The steam generator then loses its role as a safety barrier of the nuclear power plant. From a fundamental standpoint, dilute and concentrated particulate flows have received a growing attention in the last decade. In this study, we investigate the transport of solid particles around obstacles in a confined flow. Experiments were performed in a simplified configuration by considering a laminar flow in a vertical tube. An obstacle was inserted at the middle height of the tube and neutrally-buoyant particles were injected at different locations along the tube. We have investigated first the trajectories of individual particles using particle tracking (PT). Then, the particle trajectories were modeled by using the Boussinesq-Basset-Oseen equation with a flow velocity field either measured using particle image velocimetry (PIV) or calculated by the Code_Saturne software in order to account for the three-dimensional (3D) character of the obstacle wake. This paper presents a comparison between the experimental observations and the predictions of the modeling for an obstacle consisting of a rectangular step at a Reynolds number of ≈100 and evidences the importance of accounting for the 3D complex nature of the flow.

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

Applied Mechanics and Materials ◽

10.4028/www.scientific.net/amm.821.57 ◽

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.

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Assessment of Limit Loads for Circumferential Cracks in Steam Generator Tube Considering Constraining Effect of Tube Support

Volume 3: Design and Analysis ◽

10.1115/pvp2005-71308 ◽

2005 ◽

Author(s):

Hyun Su Kim ◽

Jong Sung Kim ◽

Tae Eun Jin ◽

Hong Deok Kim ◽

Han Sub Chung

Keyword(s):

Power Plant ◽

Nuclear Power Plant ◽

Steam Generator ◽

Nuclear Power ◽

Three Dimensional ◽

Fission Products ◽

Limit Load ◽

Steam Generator Tube ◽

Integrity Assessment ◽

Steam Generator Tubes

The steam generator in a nuclear power plant is a large heat exchanger that uses heat from reactor to generate steam to drive the turbine generators. Rupture of a steam generator tube can result in release of fission products to environment. Therefore, an accurate integrity assessment of the steam generator tubes with cracks is of great importance for maintaining safety of a nuclear power plant. The steam generator tubes are supported at periodic intervals by support plates and rotations of the tubes are constrained. Although it was reported that the limit load for a circumferential crack was significantly affected by boundary condition of the tube, existing limit load solutions do not include the constraining effect of tube supports. This paper provides detailed limit load solutions for circumferential cracks in steam generator tubes considering the actual boundary conditions to simulate the constraining effect of the tube supports. Such solutions are developed based on three dimensional (3D) finite element analyses. The resulting limit load solutions are given in a polynomial form, and thus can be simply used in practical integrity assessment of the steam generator tubes.

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Study on 3-D Flow Field of Concrete Volute Conduit of Circulating Water Pump in Nuclear Power Plant

Applied Mechanics and Materials ◽

10.4028/www.scientific.net/amm.99-100.350 ◽

2011 ◽

Vol 99-100 ◽

pp. 350-353

Author(s):

Xiao Bing Sun ◽

Xu Bin Qiao

Keyword(s):

Power Plant ◽

Nuclear Power Plant ◽

Nuclear Power ◽

Stokes Equations ◽

Three Dimensional ◽

Gradual Change ◽

Water Pump ◽

Pump System ◽

Circulating Water ◽

Operating Modes

As the largest unit capacity of nuclear power plant at present, the flow conduit of circulating water pump in EPR1750 nuclear power plant is a volute conduit, which is a cast-in-situ conceret structure with complexly gradual change cavity. Therefore, the hydraulic efficiency of circulating water pump is not only related with the design of pump leaves, but also closely related to the design of volute and the complicated spatial type of intake and outtake conduits. With the pump leaves and the intake and outtake conduits of conceret volute as the research model, based on computational fluid dynamics (CFD)and the three dimensional Reynolds averaged Navier-Stokes equations, an analytic model suitable for computation is established to simulate the three-dimensional steady flow in the whole pumping system under different operating modes. By use of the commercial fluid-computation softer ANSYS, the distribution of basic physic quantities in the fluid field inside the pump and the conduits is obtained. The analysis and prediction of the performance of pump system are made, and the spatial type design of intake and outtake conduits is evaluated. The calculation results can be referenced to improve the design of pump systems in the similar projects.

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

Nuclear Engineering and Design ◽

10.1016/j.nucengdes.2005.06.013 ◽

2005 ◽

Vol 235 (23) ◽

pp. 2477-2484 ◽

Cited By ~ 10

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

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

Revista ECIPeru ◽

10.33017/reveciperu2011.0033/ ◽

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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