Flow-Induced Vibration Analysis of Tube Bundles—A Proposed Section III Appendix N Nonmandatory Code

1991 ◽  
Vol 113 (2) ◽  
pp. 257-267 ◽  
Author(s):  
M. K. Au-Yang ◽  
R. D. Blevins ◽  
T. M. Mulcahy
Keyword(s):  
Vibration Analysis ◽  
Steam Generators ◽  
Flow Induced Vibration ◽  
Nuclear Plants ◽  
Tube Bundles ◽  
Operations And Maintenance ◽  
Asme Code ◽  
Excitation Mechanisms ◽  
American Society ◽  
Tube Banks

This paper presents guidelines for flow-induced vibration analysis of tubes and tube bundles such as those commonly encountered in steam generators, heat exchangers, condensers and nuclear fuel bundles. It was proposed as a nonmandatory code to be included in Section III Appendix N (N-1300 series) of the American Society of Mechanical Engineers (ASME) Boiler Code. In preparing this code, the authors tried to limit themselves to the better-defined flow excitation mechanisms—vortex-induced vibration, fluid-elastic instability and turbulence-induced vibration—and include only the more-established methods. References are, however, given for other methods whenever justified. This guideline covers only design analysis. A companion guideline on the testing and data analysis of heat exchanger tube banks was proposed as part of the ASME Code on Operations and Maintenance of Nuclear Plants. The latter is not included in this paper.

Nonproprietary Flow-Induced Vibration Analysis of San Onofre Nuclear Generating Station Replacement Steam Generators to ASME Code Section III Appendix N

2018 ◽  
Vol 140 (3) ◽  
Author(s):  
R. D. Blevins
Keyword(s):  
Vibration Analysis ◽  
Elastic Stability ◽  
Nuclear Regulatory Commission ◽  
Stability Diagram ◽  
Design Flow ◽  
Mode Shapes ◽  
Steam Generators ◽  
Flow Induced Vibration ◽  
Public Data ◽  
Asme Code

Flow-induced vibration analysis of the San Onofre Nuclear Generating Station (SONGS) replacement steam generators (RSG) is made using nonproprietary public data for these steam generators on the Nuclear Regulatory Commission public web site (www.NRC.gov). The analysis uses the methodology of Appendix N Section III of the ASME Boiler and Pressure Vessel Code, Subarticle N-1300 Flow-Induced Vibration of Tubes and Tube Banks. First, the tube geometry is assembled, and overall flow and performance parameters are developed at 100% design flow; then, the analysis is made to determine the flow velocity in the gap between tubes and tube natural frequencies and mode shapes. Finally, the mass damping and reduced velocity for tubes on the U bend are assembled and plotted on the ASME code Figure N-11331-4 fluid elastic stability diagram.

An Experimental Investigation of the Dynamics of a Loosely Supported Tube Array

2017 ◽  
Author(s):  
Amro Elhelaly ◽  
Marwan Hassan ◽  
Atef Mohany ◽  
Soha Moussa
Keyword(s):  
Experimental Investigation ◽  
Impact Force ◽  
Cross Flow ◽  
Open Loop ◽  
Diameter Ratio ◽  
Force Level ◽  
Steam Generators ◽  
Flow Induced Vibration ◽  
System Integrity ◽  
Tube Bundles

The integrity of tube bundles is very important especially when dealing with high-risk applications such as nuclear steam generators. A major issue to system integrity is the flow-induced vibration (FIV). FIV is manifested through several mechanisms including the most severe mechanism; fluidelastic instability (FEI). Tube vibration can be constrained by using tube supports. However, clearances between the tube and their support are required to allow for thermal expansion and for other manufacturing considerations. The clearance between tubes may allow frequent impact and friction between tube and support. This in turn may cause fatigue and wear at support and potential for catastrophic tube failure. This study aims to investigate the dynamics of loosely supported tube array subjected to cross-flow. The work is performed experimentally in an open-loop wind tunnel to address this issue. A loosely-supported single flexible tube in both triangle and square arrays subjected to cross-flow with a pitch-to-diameter ratio of 1.5 and 1.733, respectively were considered. The effect of the flow approach angle, as well as the support clearance on the tube response, are investigated. In addition, the parameters that affect tube wear such as impact force level are presented.

A Life-Cycle Risk-Informed Systems Structured Nuclear Code

2002 ◽  
Author(s):  
Ralph S. Hill
Keyword(s):  
Life Cycle ◽  
American Concrete Institute ◽  
Design Code ◽  
Replacement Decisions ◽  
Plant Operations ◽  
Operations And Maintenance ◽  
Asme Code ◽  
American Society ◽  
Construction Operation ◽  
Design Construction

Current American Society of Mechanical Engineers (ASME) nuclear codes and standards rely primarily on deterministic and mechanistic approaches to design. The design code is a separate volume from the code for inservice inspections and both are separate from the standards for operations and maintenance. The ASME code for inservice inspections and code for nuclear plant operations and maintenance have adopted risk-informed methodologies for inservice inspection, preventive maintenance, and repair and replacement decisions. The American Institute of Steel Construction and the American Concrete Institute have incorporated risk-informed probabilistic methodologies into their design codes. It is proposed that the ASME nuclear code should undergo a planned evolution that integrates the various nuclear codes and standards and adopts a risk-informed approach across a facility life-cycle — encompassing design, construction, operation, maintenance and closure.

Non-Proprietary Application of ASME Code Section III, Appendix N, to SONGS Replacement Steam Generators

2017 ◽  
Author(s):  
R. D. Blevins
Keyword(s):  
Elastic Stability ◽  
Nuclear Regulatory Commission ◽  
Stability Diagram ◽  
Design Flow ◽  
Mode Shapes ◽  
Steam Generators ◽  
Flow Induced Vibration ◽  
Public Data ◽  
Asme Code ◽  
Regulatory Commission

Flow-induced vibration analysis of the San Onofre Nuclear Generating Station (SONGS) Replacement Steam Generators is made using non-proprietary public data for these steam generators on the Nuclear Regulatory Commission public web site, www.NRC.com. The analysis uses the methodology of Appendix N Section III of the ASME Boiler and Pressure Vessel Code, Subarticle N-1300 Flow-Induced Vibration of Tubes and Tube Banks. First the tube geometry is assembled and overall flow and performance parameters are developed at 100% design flow, then analysis is made to determine the flow velocity in the gap between tubes and tube natural frequencies and mode shapes. Finally, the mass damping and reduced velocity for tubes on the U bend are assembled and plotted on the ASME code Figure N-11331-4 fluid elastic stability diagram.

Vibration Analysis of Steam Generators and Heat Exchangers: An Overview — Part II: Vibration, Response, Fretting-Wear, Guidelines

2002 ◽  
Author(s):  
Michel J. Pettigrew ◽  
Colette E. Taylor
Keyword(s):  
Heat Exchangers ◽  
Vibration Analysis ◽  
Response Prediction ◽  
Design Guidelines ◽  
Vibration Response ◽  
Fretting Wear ◽  
Steam Generators ◽  
Flow Induced Vibration ◽  
Two Phase ◽  
Service Water

Design guidelines were developed to prevent tube failures due to excessive flow-induced vibration in shell-and-tube heat exchangers. An overview of vibration analysis procedures and recommended design guidelines is presented in this paper. This paper pertains to liquid, gas and two-phase heat exchangers such as nuclear steam generators, reboilers, coolers, service water heat exchangers, condensers, and moisture-separator-reheaters. Part 2 of this paper covers forced vibration excitation mechanisms, vibration response prediction, resulting damage assessment, and acceptance criteria.

scholarly journals A novel methodology for design optimization of heat recovery steam generators with flow-induced vibration analysis

Energy ◽  
2021 ◽  
Vol 226 ◽  
pp. 120325
Author(s):  
Han Deng ◽  
Geir Skaugen ◽  
Erling Næss ◽  
Mingjie Zhang ◽  
Ole A. Øiseth
Keyword(s):  
Design Optimization ◽  
Vibration Analysis ◽  
Heat Recovery ◽  
Steam Generators ◽  
Flow Induced Vibration

Modeling and Simulation of Cross-Flow Induced Vibrations in Tube Bundles Using Bondgraph Approach

2006 ◽  
Author(s):  
M. Afzaal Malik ◽  
Badar Rashid ◽  
M. Anwar Khan ◽  
Khawaja Sajid Bashir ◽  
Shahab Khushnood
Keyword(s):  
Research Work ◽  
Cross Flow ◽  
Acoustic Resonance ◽  
Elastic Instability ◽  
Steam Generators ◽  
Two Phase ◽  
Considerable Research ◽  
Tube Bundles ◽  
Flow Induced Vibrations ◽  
Excitation Mechanisms

A considerable research has been carried out in the field of Cross-Flow Induced Vibrations (CFIV) in tube bundles of process exchangers and nuclear steam generators. Various excitation mechanisms such as vortex shedding, turbulent buffeting, fluid-elastic instability and acoustic resonance and other parameters like natural frequencies, damping, wear work rates at the loose tube supports and various geometric tube arrangements have been the focus in single and two-phase cross-flow. In the current research work, CFIV has been studied by using Bondgraph approach. The Bondgraph models have been subjected to simulation using the software (20-SIM). Results obtained have shown a strong usefulness of Bondgraph approach to complex CFIV systems.

Fully Plastic Failure Stresses and Allowable Crack Sizes for Circumferentially Surface Cracked Pipes Subjected to Tensile Loading

2021 ◽  
Author(s):  
Kunio Hasegawa ◽  
David Dvorak ◽  
Vratislav Mares ◽  
Bohumir Strnadel ◽  
Yinsheng Li
Keyword(s):  
Experimental Data ◽  
Bending Moment ◽  
Limit Load ◽  
Tensile Loading ◽  
Crack Size ◽  
Plastic Failure ◽  
Circumferential Crack ◽  
Asme Code ◽  
American Society ◽  
Better Than

Abstract Fully plastic failure stresses for circumferentially surface cracked pipes subjected to tensile loading can be estimated by means of limit load criteria based on the net-section stress approach. Limit load criteria of the first type (labelled LLC-1) were derived from the balance of uniaxial forces. Limit load criteria of the second type are given in Section XI of the ASME (American Society of Mechanical Engineering) Code, and were derived from the balance of bending moment and axial force. These are labelled LLC-2. Fully plastic failure stresses estimated by using LLC-1 and LLC-2 were compared. The stresses estimated by LLC-1 are always larger than those estimated by LLC-2. From the literature survey of experimental data, failure stresses obtained by both types of LLC were compared with the experimental data. It can be stated that failure stresses calculated by LLC-1 are better than those calculated by LLC-2 for shallow cracks. On the contrary, for deep cracks, LLC-2 predictions of failure stresses are fairly close to the experimental data. Furthermore, allowable circumferential crack sizes obtained by LLC-1 were compared with the sizes given in Section XI of the ASME Code. The allowable crack sizes obtained by LLC-1 are larger than those obtained by LLC-2. It can be stated that the allowable crack size for tensile stress depends on the condition of constraint of the pipe, and the allowable cracks given in Section XI of the ASME Code are conservative.

Flow-Induced Vibration in a Single Row of Cylinders With p/D = 1.26

2021 ◽  
Author(s):  
Roberta F. Neumeister ◽  
Adriane P. Petry ◽  
Sergio V. Möller
Keyword(s):  
Hot Wire ◽  
Flow Induced Vibration ◽  
Single Row ◽  
Reference Flow ◽  
Wake Interaction ◽  
Hot Wire Anemometry ◽  
Tube Banks ◽  
Comparison Of The Results ◽  
The Impact ◽  
Space Ratio

Abstract Crossflow over a row of cylinders with a close space ratio presents an asymmetric configuration with large and narrow wakes behind the cylinders. The wake interaction can impact the vibration response of the cylinders. In tube banks, the impact results in damages to the equipment. The present experimental study aims to analyze the influence of close space observed in a single row of cylinders on the flow-induced vibration. The study compares a single row with fixed cylinders and a single row with one cylinder free to vibrate. The cylinder free to vibrate is tested in four configurations. The study was conducted with an aerodynamic channel with a cross-section of 0.193 × 0.146 m and smooth cylinders with a diameter of 25.1 mm, space ratio is 1.26. The measurements are executed with hot-wire anemometry and accelerometers, for the cases with one cylinder free to vibrate and with hot-wire anemometry and microphones for the case with all fixed cylinders. The Reynolds number ranges between 1.0 × 104 and 4.5 × 104, obtained with the reference flow velocity, measured with a Pitot tube, and the cylinder diameter. The comparison between the wake response for single row fixed and single row and free to vibrate are executed using Fourier transform and Wavelet Transform. The comparison of the results with the models presented in the literature to predict the elastic instability of the fluid in a single row of cylinders is performed.

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