Risks for Flow Induced Vibrations (FIV) at EPU

2008 ◽  
Author(s):  
Per Nilsson ◽  
Eric Lillberg
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Cross Flow ◽  
Flow Induced Vibration ◽  
Resonance Vibrations ◽  
Flow Induced Vibrations ◽  
Risk Areas ◽  
Tube Banks ◽  
Semi Empirical ◽  
Simple Search

This work deals with risk areas for flow induced vibration at extended power uprate, EPU. The focus is on the mechanisms of excitation in one phase relevant for Swedish BWRs and PWRs. FIV-events that have occurred in nuclear power plants over the world have been collected and categorized. The most relevant events for EPU are summarized to: vibrations in steam systems due to turbulence or vortex shedding and resonance, vibrations of internal parts and also thermal mixers and sleeves or in valves and vibrations of tube banks in partial or full cross flow. Based on the collected events and some semi-empirical methods, a simple search list for FIV by power uprate has been developed. In principle these changes lead to increased risks: changed flow velocity, decreased water temperature and increased steam temperature and decreased structural damping, mass or stiffness. In addition to that, the typical collected events should be regarded.

Measurement of Flow-Induced Forces: A Single Flexible Tube in a Rigid Triangular-Pitch Bundle Subjected to Two-Phase Cross-Flow

2015 ◽  
Author(s):  
Enrico Deri ◽  
Joël Nibas ◽  
Olivier Ries ◽  
André Adobes
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Tube Bundle ◽  
Fluid Dynamic ◽  
Cross Flow ◽  
Maintenance Policy ◽  
Excitation Spectra ◽  
Two Phase ◽  
Force Coefficients ◽  
Out Of Plane

Flow-induced vibrations of Steam Generator tube bundles are a major concern for the operators of nuclear power plants. In order to predict damages due to such vibrations, EDF has developed the numerical tool GeViBus, which allows one to asses risk and thereafter to optimize the SG maintenance policy. The software is based on a semi analytical model of fluid-dynamic forces and dimensionless fluid force coefficients which need to be assessed by experiment. The database of dimensionless coefficients is updated in order to cover all existing tube bundle configurations. Within this framework, a new test rig was presented in a previous conference with the aim of assessing parallel triangular tube arrangement submitted to a two-phase cross-flow. This paper presents the result of the first phase of the associated experiments in terms of force coefficients and two-phase flow excitation spectra for both in-plane and out-of-plane vibration.

Direct Numerical Simulation of a Turbulent Flow on an Oscillating Wall Using Cut-Cell Based Immersed Boundary Method

2016 ◽  
Author(s):  
Chiaki Kino
Keyword(s):  
Turbulent Flow ◽  
Nuclear Power ◽  
Power Plants ◽  
Separated Flow ◽  
Immersed Boundary ◽  
Turbulent Structures ◽  
Flow Induced Vibration ◽  
Production Term ◽  
Large Eddy Simulation Model ◽  
Wide Range

The flow-induced vibration of a pipe is an important issue in various engineering fields, and this phenomenon is widely observed in nuclear power plants. Although turbulent structures play important roles in the velocity and pressure fields in a pipe, only a few studies have been conducted on the turbulent flow on an oscillating wall. In this study, direct numerical simulations were conducted to establish a large eddy simulation model for a turbulent flow on an oscillating wall and scrutinize the energy transfer between the grid scale (GS) and sub-grid scale (SGS). Although energy is generally transferred from the GS to SGS (forward scatter), it is likely that energy is transferred from the SGS to GS (backward scatter) under specific conditions. The present numerical results indicated that backward scatter exists in the production term in the case of a static wavy wall. On the other hand, such backward scatter could not be observed in the case of an oscillating wall. It is well known that separated flows and backward flows are generated behind the crest. Stronger backward flows accelerate the main flow and enhance the velocity gradients in a wide range behind the crest. In the case of an oscillating wall, the development of separated flow is immature because the shape of the wall is not fixed. Eventually, the backward scatter is deemed to be suppressed.

Damping of Heat Exchanger Tubes in Two-Phase Flow: Review and Design Guidelines

2004 ◽  
Vol 126 (4) ◽  
pp. 523-533 ◽  
Author(s):  
M. J. Pettigrew ◽  
C. E. Taylor
Keyword(s):  
Heat Exchanger ◽  
Two Phase Flow ◽  
Cross Flow ◽  
Design Guidelines ◽  
Phase Flow ◽  
Flow Induced Vibration ◽  
Two Phase ◽  
Design Guideline ◽  
Fluid Properties ◽  
Semi Empirical

Two-phase flow exists in many shell-and-tube heat exchangers such as condensers, evaporators, and nuclear steam generators. Some knowledge on tube damping mechanisms is required to avoid flow-induced vibration problems. This paper outlines the development of a semi-empirical model to formulate damping of heat exchanger tube bundles in two-phase cross flow. The formulation is based on information available in the literature and on the results of recently completed experiments. The compilation of a database and the formulation of a design guideline are outlined in this paper. The effects of several parameters such as flow velocity, void fraction, confinement, flow regime and fluid properties are discussed. These parameters are taken into consideration in the formulation of a practical design guideline.

Design and Maintenance in an IAEA Technical Guidelines on Fluid-Structure Interaction

2014 ◽  
Author(s):  
Fumio Inada ◽  
Tomomichi Nakamura ◽  
Takashi Nishihara ◽  
Shigehiko Kaneko ◽  
Manwoong Kim ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Structural Integrity ◽  
General Description ◽  
Fluid Structure Interactions ◽  
Flow Induced Vibration ◽  
Energy Agency ◽  
Fluid Structure ◽  
Component Systems

In nuclear power plants, fluid structure interactions (FSI) occurring in component systems can cause excessive forces or stresses to the structures resulting in mechanical damages that may eventually threaten the structural integrity. FSI in the guidelines includes flow-induced vibration, water hammer, and pipewhip. It can also include movement, deformation, or fracture of equipments by tsunami etc. They can be issues of design and maintenance. Authors cannot find complete guidelines to correspond to the FSI phenomena which can be important in the design and maintenance of nuclear power plants. Based on the background, International Atomic Energy Agency (IAEA) has drafted guidelines on FSI. This paper summarizes general description of FSI as well as design and maintenance against FSI.

Fluid-Structural Interaction in a Slip Joint of a Jet Pump Assembly of a BWR-5

2020 ◽  
Vol 399 ◽  
pp. 105-114
Author(s):  
Juan Cruz Castro ◽  
Yunuén López Grijalba ◽  
Luis Héctor Hernández Gómez ◽  
Israel Abraham Alarcón Sánchez ◽  
Pablo Ruiz López ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Mechanical Damage ◽  
Steel Plates ◽  
Jet Pump ◽  
Flow Induced Vibration ◽  
Structural Interaction ◽  
Pump Assembly ◽  
Volume Method

Flow-induced vibrations occur in some of the internal components of a nuclear reactor. When specific conditions are present, these vibrations may result in excessive deformations or fatigue that can generate mechanical damage. Several boiling water reactor (BWR) of nuclear power plants (NPP) have experienced failures in the jet pump assembly due to flow-induced vibration (FIV) which could be caused by acoustic pulsations derived from recirculation pumps, vibration induced by turbulence and vibration induced by leakage at the slip joint. The purpose of this paper is to establish a viable numerical methodology to evaluate the fluid-structural interaction at the slip joint of a jet pump. In this analysis, the fluid-structural interaction was evaluated with the finite element method and finite volume method with ANSYS® code in the case of two steel plates with a divergent gap. Results show that a critical velocity could cause fluidelastic instability, if only one flow in a two-way fluid-structural interaction was considered. This is one of the phenomena that could take place at the slip joint of a jet pump assembly.

Flow-Induced Vibration Analysis of an Integral Economizer Once-Through Steam Generator

1989 ◽  
Vol 111 (4) ◽  
pp. 501-506
Author(s):  
M. K. Au-Yang ◽  
B. Brenneman
Keyword(s):  
Fuel Assembly ◽  
Steam Generator ◽  
Flow Rate ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
First Generation ◽  
Second Generation ◽  
Flow Induced Vibration ◽  
Operational Test

The integral economizer once-through steam generator is a second-generation steam generator used in B&W’s 205-fuel assembly nuclear power plants. Besides having an integral economizer, this steam generator differs from the first generation units, sixteen of which have been operating with B&W’s 177 fuel assembly nuclear power plants for more than ten years, in having a much higher flow rate. This higher flow rate induces a correspondingly higher fluid-dynamic load on all of the steam generator internal components, particularly the tube bundle. This paper describes the flow-induced vibration design analysis of this second-generation nuclear steam generator. The three most commonly known flow-induced vibration phenomena were considered: fluid-elastic instability, turbulence-induced vibration and vortex-induced vibration. To minimize uncertainties in the many experimentally determined input parameters such as damping ratios, Connors’ constant and the dynamic pressure power spectral densities, a parallel analysis was carried out on the operating first-generation steam generator, and the results compared. The analytical results were verified by the recent start-up of B&W’s first 205-fuel assembly nuclear plant. No vibration problems were encountered during either the pre-operational test or in several months of full power operations.

Investigation of Rod Vibrations in Droplet Two-Phase Flows

2011 ◽  
Author(s):  
Yoshiteru Komuro ◽  
Zensaku Kawara ◽  
Tomoaki Kunugi
Keyword(s):  
Numerical Simulation ◽  
High Speed ◽  
Nuclear Power ◽  
Power Plants ◽  
Motion Tracking ◽  
High Speed Camera ◽  
Two Phase Flows ◽  
Two Phase ◽  
Wet Condition ◽  
Flow Induced Vibrations

Flow-induced vibrations are important problems in nuclear power plants from the view point of reactor safety. In the investigations of these vibrations especially those induced by two-phase flows, a numerical simulation plays a significant role, so it is necessary to obtain the experimental datasets that can validate the results of the numerical simulation. This paper deals with the experimental data of one-end-supported rod vibration, and focuses on the differences between the rod vibrations induced by single-phase air flows and those induced by droplet two-phase flows. In the experiments, the displacement of the non-supported end of the test rod was visualized by the high speed camera with high spatial and temporal resolutions, namely 9.5 μm and 500 μsec. Using an image analyzing software, the rod vibration displacements were measured by the motion tracking method. The curved surface of the rod was observed by another high speed camera and the relationship between the rod vibrations and the wet condition on the surface of the rod was investigated. In addition, the vibrations measured by the strain gages and those by the high speed camera were compared to discuss the differences in these two ways of the measurements.

A Numerical Characterization of Flow-Induced Vibration and Fretting Wear Potential in Nuclear Steam Generator Tube Bundles

2011 ◽  
Author(s):  
Marwan Hassan ◽  
Atef Mohany
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Complex Dynamics ◽  
Fretting Wear ◽  
Contact Ratio ◽  
Flow Induced Vibration ◽  
Numerical Characterization ◽  
Wear Damage ◽  
Effect Of Support ◽  
Tube Failure

Nuclear power plants have experienced problems related to tube failures in steam generators. While many of these failures have been attributed to corrosion, it has been recognized that flow-induced vibrations contribute significantly to tube failure. In order to avoid these excessive vibrations, tubes are stiffened by placing supports along their length. Various tube/support geometries have been used, but the majority are either support plates (plates with drilled or broached holes) or flat bars. Unfortunately, clearance is often considered necessary between the tubes and their supports to facilitate tube/support assembly and to allow for thermal expansion of the tubes. A combination of flow-induced turbulence and fluidelastic forces may then lead to unacceptable tube fretting-wear at the supports. The fretting wear damage could ultimately cause tube failure. Such failures may require shut downs resulting in production losses, and pose potential threats to human safety and the environment. Therefore, it is imperative to predict the nonlinear tube response and the associated fretting wear damage to tubes due to fluid excitations. Tubes in loose flat-bar supports have complex dynamics due to the possible combinations of geometry. The understanding of tube dynamics in the presence of this type of support and the associated fretting wear is still incomplete. These issues are addressed in this paper through simulations of the dynamics of tubes subjected to crossflow turbulence and fluidelastic instability forces. The finite element method is utilized to model the vibrations and impact dynamics. The tube model simulates a U-tube supported by 16 flat bars with clearances and axial offset. Results are presented and comparisons are made for the parameters influencing the fretting-wear damage such as contact ratio, impact forces and normal work rate. The effect of support clearance and support axial offset are investigated.

Predictive Study of CAP1400 Core Barrel Vibration: Part 2 — Structural Responses

2017 ◽  
Author(s):  
Li Yuan ◽  
Zhang Wei ◽  
Zhang Ming ◽  
Yu Qing
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Structural Response ◽  
Mass Effect ◽  
Scale Model ◽  
Pressurized Water Reactor ◽  
Flow Induced Vibration ◽  
Pressurized Water ◽  
Structural Responses ◽  
Predictive Study

As described in Part 1 of this paper, [1], CAP1400 is a 1400 MWe pressurized water reactor developed by SNERDI to be the next series of nuclear power plants in China. As a part of the feasibility study, a 1/6 scale model test, conducted in Chengdo, China, of the CAP1400 pressure vessel and its internals was carried out to study the flow induced vibration (FIV) characteristics. This paper describes the predictive study on the structural responses of the core barrel vibration with particular emphasis on using the “Fourier node” method in modeling the hydrodynamic mass effect. It is noted that this is the second part of a two-part series and is focusing on the structural response calculation using the forcing functions described in Part 1, [1], and the comparison with the measured data.

Research on the Law of Thermal Effect of Floating Nuclear Power Plants Thermal Discharge

2015 ◽  
Vol 799-800 ◽  
pp. 734-738
Author(s):  
Tian Qi Dai ◽  
Shi Wei Yao ◽  
Zhi Guo Wei
Keyword(s):  
Thermal Effect ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Water Discharge ◽  
Cooling Water ◽  
Cross Flow ◽  
Pollution Prevention ◽  
Thermal Pollution ◽  
Thermal Discharge

The waste heat emissions of thermal discharge from floating nuclear power plants may have a negative thermal effect on the environment. Study on the dilution and diffusion of cooling water plays an important role in thermal pollution prevention. The cooling water discharge process can be condensed into the thermal jet in cross flow. According to the theory of computational fluid dynamics, the mathematical model of round horizontal thermal jets in cross flow is established. The 3D numerical simulation of thermal jets based on finite volume method is achieved by using the Realizable k-ε turbulence model and the Semi-implicit method for pressure linked equations, and the three-dimensional trajectory of thermal jet are obtained. The rationality of analysis method is approved by comparing calculation value with experimental value. The temperature distributions in thermal jets are studied through the numerical experiments conducted under different cross-flow velocity and different emission angle. As a result, the impacts of these conditions on thermal pollution area are found, and the theoretic bases are provided for the design of the cooling water discharge pipe.

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