Integrity of a Curved Divider Plate in Steam Generators of Pressurized Water Reactors Due to Dynamic Pressure Loading Considering Fluid Structure Interaction

2002 ◽  
Author(s):  
Konrad Schramm
Keyword(s):  
Fluid Structure Interaction ◽  
Differential Pressure ◽  
Pressurized Water Reactors ◽  
Plastic Behavior ◽  
Pressure Loading ◽  
Steam Generators ◽  
Pressurized Water ◽  
Fluid Structure ◽  
Structure Interaction ◽  
Water Reactors

A non-linear analysis procedure has been developed to investigate the plastic behavior of the divider (or partition) plate in Steam Generators (SG) of Pressurized Water Reactors (PWRs). The integrity of the divider plate in the primary channel head of a Steam Generator of a Pressurized Water Reactor has to be verified for a postulated loss of coolant accident (LOCA). Primary stresses due to differential pressure Δp have to be analyzed for faulted conditions. A static elastic-plastic finite element analysis for the divider plate was performed showing unrealistically high plastic deformation due to the high differential pressure loading resulting from the assumption of rigid walls in the fluid analysis. Therefore the LOCA is analyzed in a Fluid Structure Interaction Analysis (FSIA), representing the divider plate as a single degree of freedom system (SDOFS) acting like a piston. The present paper explains how the Δp across the plate was correlated with an equivalent displacement ve and how the mass of the plate is considered. In the FSIA a realistic dynamic response of the divider plate is considered now, with a significant reduction of differential pressure Δp. The analysis results proved the resistance of the divider plate with sufficient margin.

Identification of Non-Linear Damping of Nuclear Reactor Components in Case of One-to-One Internal Resonance

2016 ◽  
Author(s):  
Joachim Delannoy ◽  
Marco Amabili ◽  
Brett Matthews ◽  
Brian Painter ◽  
Kostas Karazis
Keyword(s):  
Fuel Assembly ◽  
Internal Resonance ◽  
Fluid Structure Interaction ◽  
Fundamental Parameter ◽  
Pressurized Water ◽  
Fluid Structure ◽  
Structure Interaction ◽  
Safety Margins ◽  
Non Linear ◽  
One To One

In Pressurized Water Reactors (PWR) assemblies are exposed to challenging thermal, mechanical, and irradiation loads during operation. Global core and local fuel assembly flow fields coupled with seismic excitation result in fuel assembly and fuel rod vibrations. The fact that vibrations may become excessive in certain conditions has consequences on operational safety margins in fuel assemblies designs. In order to understand how the fuel assembly responds dynamically to an external excitation, it is important to identify the main characteristics of the structures. Among them, the fuel assembly system damping is a fundamental parameter that is usually identified by a number of experiments involving fluid-structure interaction. Recent studies have shown that the damping ratio increases with the excitation force when the structure is entering large-amplitude vibrations, in which case the geometric non-linearities have to be taken into account. The present paper presents an advanced identification procedure developed to identify the system characteristics from experimental non-linear response curves obtained from forced vibration tests, accounting for fluid-structure interaction, at different excitation levels. Furthermore, the numerical tool developed in this analysis is capable of working with systems presenting one-to-one internal resonance, i.e. systems with symmetry such as circular tubes and circular cylindrical shells. The method relies on a harmonic decomposition of the displacement to cope with the data usually available by vibration measurements.

Evaluation of single heated channel and subchannel modeling of a nuclear once through steam generator (OTSG)

Kerntechnik ◽  
2020 ◽  
Vol 85 (1) ◽  
pp. 54-67
Author(s):  
A. Hamedani ◽  
O. Noori-Kalkhoran ◽  
R. Ahangari ◽  
M. Gei
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Axial Direction ◽  
Pressurized Water Reactors ◽  
Steam Generators ◽  
Pressurized Water ◽  
Heated Channel ◽  
Water Reactors ◽  
Secondary Side

Abstract Steam generators are one of the most important components of pressurized-water reactors. This component plays the role of heat transfer and pressure boundary between primary and secondary side fluids. The Once Through Steam Generator (OTSG) is an essential component of the integrated nuclear power system. In this paper, steady-state analysis of primary and secondary fluids in the Integral Economizer Once Through Steam Generator (IEOTSG) have been presented by Single Heated Channel (SHC) and subchannel modelling. Models have been programmed by MATLAB and FORTRAN. First, SHC model has been used for this purpose (changes are considered only in the axial direction in this model). Second, the subchannel approach that considers changes in the axial and also radial directions has been applied. Results have been compared with Babcock and Wilcox (B&W) 19- tube once through steam generator experimental data. Thermal- hydraulic profiles have been presented for steam generator using both of models. Accuracy and simplicity of SHC model and importance of localization of thermal-hydraulic profiles in subchannel approach have been proved.

scholarly journals A review on clogging of recirculating steam generators in Pressurized-Water Reactors

2017 ◽  
Vol 97 ◽  
pp. 182-196 ◽  
Author(s):  
Guangze Yang ◽  
Véronique Pointeau ◽  
Etienne Tevissen ◽  
Alexandre Chagnes
Keyword(s):  
Pressurized Water Reactors ◽  
Steam Generators ◽  
Pressurized Water ◽  
Water Reactors
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