Experimental Damping of a Heat Exchange Tube With a Large Number of Supports in Air and Water

2017 ◽  
Author(s):  
Tae-Jung Park ◽  
Chang-Hoon Ha ◽  
Min-Ki Cho ◽  
Heung Seok Kang ◽  
Kang Hee Lee
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Large Scale ◽  
Damping Ratio ◽  
Water Environment ◽  
Random Force ◽  
Flow Induced Vibration ◽  
Heat Exchange Tube ◽  
Half Power ◽  
Support Grid

The flow induced vibration occurs frequently in a steam generator in the nuclear power plant. The large-scale steam generator has a large number of tube supports whose cell has rhombus-type shape, and there is a tiny clearance between tube and its support grid. The damping is very complex because of non-linearity and randomness. The experiment for damping was performed to investigate it with a number of 13 support spans both in air and water environment. The lower part of multi-span fixture was excited by root-mean-square random force with the range of 1∼10 newton to get the frequency response function. The half-power bandwidth method was applied to obtain the damping ratio. The sensitivity of a number of spans was investigated in the range of 9 ∼ 13. In addition, the damping was reviewed from a comparison with Pettigrew [1∼4] and ASME B&PV Code [5].

CFD Analysis of the Primary Side in a Helical-Coil Once-Through Steam Generator

2017 ◽  
Author(s):  
Kai Ye ◽  
Yaoli Zhang ◽  
Jianshu Lin ◽  
Ning Li ◽  
Yinglin Yang ◽  
...  
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Rational Design ◽  
Large Scale ◽  
Three Dimensional ◽  
Helical Coil ◽  
Complicated Structure ◽  
Ansys Fluent ◽  
Flow Parameters ◽  
Computational Fluid Dynamics Cfd

The helical-coil once-through steam generator (OTSG) is usually used in the nuclear power plant when the compactness of equipment was taken into consideration. The investigation of flow parameters in the primary side is valuable for the optimization of the OTSG. The purpose of this research is to obtain a further understanding of fluid behaviors in the primary side of the OTSG to achieve a more rational design. Using ANSYS ICEM and ANSYS FLUENT, a three-dimensional (3D) computational fluid dynamics (CFD) model was created and analyzed. Through a series of cases, the velocity profiles and pressure drop through the primary side of the helical-coil OTSG have been calculated, and the influences of different structure designs on the coolant flow parameters have also been tested. Ultimately some pertinent suggestions for improvements were proposed, and insight is obtained into the importance of various modeling considerations in such a model with a complicated structure and large-scale grids.

Flow-Induced Vibration Responses of U-Tube Bundle in Air-Water Flow

2007 ◽  
Author(s):  
In-Cheol Chu ◽  
Heung June Chung ◽  
Chang Hee Lee ◽  
Hyung Hyun Byun ◽  
Moo Yong Kim
Keyword(s):  
Steam Generator ◽  
Two Phase Flow ◽  
Tube Bundle ◽  
Damping Ratio ◽  
Geometrical Parameters ◽  
Phase Flow ◽  
Elastic Instability ◽  
Flow Induced Vibration ◽  
Two Phase ◽  
Vibration Responses

In the present study, a series of experiments have been performed to investigate a fluid-elastic instability of a nuclear steam generator U-tube bundle in an air-water two-phase flow condition. A total of 39 U-tubes are arranged in a rotated square array with a pitch-to-diameter ratio of 1.633. The diameter and other geometrical parameters of U-bend region are the same to those of an actual steam generator, but the vertical length of U-tubes are reduced to 2-span in contrast to 9-span of an actual steam generator. The following parameters were experimentally measured to evaluate a fluid-elastic instability of U-tube bundles in a two-phase flow: a general tube vibration response, a critical gap velocity, a damping ratio and a hydrodynamic mass. Based on the experimental measurements, the instability factor, K, of Connors’ relation was preliminary assessed with some assumptions on the velocity and density profiles of the two-phase flow.

Damping of Steam Generator Tubes

2003 ◽  
Author(s):  
Cordelia Kaye Chandler
Keyword(s):  
Steam Generator ◽  
Damping Ratio ◽  
Critical Factor ◽  
Nuclear Industry ◽  
Original Design ◽  
Flow Induced Vibration ◽  
Steam Generator Tube ◽  
Term Operation ◽  
Structural Support ◽  
Root Cause

The flow-induced vibration (FIV) analysis is one of the most critical evaluations needed to ensure long-term operation of steam generator. Original design hardware is usually based on experience and testing. Performance changes, such as an increase in power with increased feedwater flow, or tube repairs including plugs, stabilizers, and sleeves are best evaluated analytically. A critical factor in an FIV analysis is damping. The effective damping for a steam generator tube is a function of the structural support configuration, vibration amplitude, and the void fraction of the shell-side fluid. Numerous tests have been performed in the nuclear industry to determine appropriate damping ratios for the analysis of steam generator tube bundles. As part of the root-cause analysis for a recent plugged tube sever in a Once-Through Steam Generator (OTSG), Framatome ANP performed damping tests on tubes that were swollen due to internal pressure. Damping tests were performed on virgin tubes. The tests were repeated after swelling the tube into the tube supports and again after the swollen tube was severed and stabilized. Tests were performed with air or water inside the tubes. The tests results showed a damping ratio of 2% to 5% for both a virgin tube and a tube that was swollen and locked into the tubes supports. Stabilization increased the damping ratio to greater than 8%.

DEVELOPMENT OF THE HIGH TEMPERATURE FRETTING WEAR SIMULATOR FOR STEAM GENERATOR

2010 ◽  
Vol 24 (15n16) ◽  
pp. 2603-2608
Author(s):  
CHOON YEOL LEE ◽  
JOONG HO KIM ◽  
JOON WOO BAE ◽  
YOUNG SUCK CHAI
Keyword(s):  
High Temperature ◽  
Steam Generator ◽  
Nuclear Power ◽  
Contact Region ◽  
Water Environment ◽  
Control Unit ◽  
Fretting Wear ◽  
Test Rig ◽  
Realistic Condition ◽  
Wear Simulator

In nuclear power plant, fretting wear due to a combination of impact and sliding motions of the U-tubes against the supports and/or foreign objects caused by flow induced vibration, can make a serious problem in steam generator. A test rig, fretting wear simulator, is developed to elucidate fretting wear mechanism qualitatively and quantitatively. The realistic condition of steam generator of high temperature up to 320°C, high pressure up to 15 MPa, and water environment could be achieved by a test rig. The fretting wear simulator consists of main frame, water loop system, and control unit. Actual contact region under a realistic condition of steam generator was isolated using autoclave. Effects of various parameters such as the amounts of impact and sliding motions, applied loads and initial gaps and so forth are considered in this research. After the experiment, wear damage was measured by a three-dimensional profiler and the surface was also studied by SEM microscopically. Initial results were also presented.

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.

Flow-Induced Vibration Characteristics of Steam Generator U-Tubes With Defect

2003 ◽  
Author(s):  
Jong-Chull Jo ◽  
Myung Jo Jhung ◽  
Woong-Sik Kim ◽  
Hho-Jung Kim
Keyword(s):  
Flow Field ◽  
Steam Generator ◽  
Damping Ratio ◽  
Three Dimensional ◽  
Added Mass ◽  
Shell Side ◽  
Flow Induced Vibration ◽  
Stability Ratio ◽  
Fluidelastic Instability ◽  
Side Flow

This paper investigates the fluidelastic instability characteristics of steam generator (SG) U-tubes with defect. The operating SG shell-side flow field conditions for determining the fluidelastic instability parameters such as damping ratio and added mass are obtained from three-dimensional SG flow calculation. Modal analyses are performed for the U-tubes either with axial or circumferential flaw with different sizes. Special emphases are on the effects of flaw orientation and size on the modal and instability characteristics of tubes, which are expressed in terms of the natural frequency, corresponding mode shape and stability ratio.

Assessment of Choking Flow Models in RELAP5 for Subcooled Choking Flow Through a Small Axial Crack of a Steam Generator Tube

2017 ◽  
Author(s):  
Mark A. Brown ◽  
Hung Nguyen ◽  
Shripad T. Revankar ◽  
Jovica Riznic
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Large Scale ◽  
Maximum Flow ◽  
Channel Length ◽  
Steam Generator Tube ◽  
Flow Rates ◽  
Flow Models ◽  
Engineered Safeguards ◽  
Channel Lengths

Choking flow plays an integral part not only in the engineered safeguards of a nuclear power plant (NPP), but also to everyday operation. Current NPP steam generators operate on the leak-before-break approach. The ability to predict and estimate a leak rate through a steam generator tube crack is an important safety parameter. Knowledge of the maximum flow rate through a crack in the steam generator tube allows the coolant inventory to be designed accordingly while limiting losses during loss of coolant accidents. Here an assessment of the choking flow models in thermal-hydraulics code RELAP5/MOD3.3 is performed and its suitability to predict choking flow rates through small axial cracks of the steam generator tubes is evaluated based on previously collected experimental data. Three sets of the data were studied in this work which corresponds to steam generator tube crack sample 1, 2, and 3. Each sample has a wall thickness, channel length (L), of 1.285 mm to 1.3 mm. Exit areas of these samples are 5.22 mm2, 9.05 mm2, and 1.72 mm2 respectively. Samples 1 and 2 have the same flow channel length to hydraulics diameter ratio (L/D) of 2.9 whereas sample 3 has a L/D of 6.5. A pressure differential of 6.8 MPa was applied across the samples with a range of subcooling from 5 °C to 60 °C. Flow rates through these samples were modeled using the thermal-hydraulic system code RELAP5/MOD3.3. Simulation’s results are compared to experimental values and modeling techniques are discussed. It is found that both the Henry-Fauske (H-F) and Ransom-Trapp (R-T) models better predict choking mass flux for longer channels. As the channel length decreases both models’ predictions diverge from each other. While RELAP5/MOD3.3 has been shown to predict choking flow in large scale geometries, further investigation of data sets need to be done to determine if it is suited well for small channel lengths.

Vibration Damping of Stabilized Steam-Generator Tubes

2005 ◽  
Author(s):  
V. P. Janzen ◽  
Y. Han ◽  
B. A. W. Smith ◽  
S. M. Fluit
Keyword(s):  
Steam Generator ◽  
Time Domain ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Damping ◽  
Flow Induced Vibration ◽  
Steam Generator Tubes ◽  
Free Vibration Response ◽  
Spectral Responses

The integrity of steam-generator tubes is an important aspect of the long-term reliable operation of nuclear power plants. In situations where a tube is judged to be at risk, it must be either plugged, or removed, or reliably stabilized in some manner to avoid excessive motion of the tube due to flow-induced vibration. The present work describes measurements of the effect of an internal cable-type stabilizer on the structural damping of steam-generator tubes. The free-vibration response of unstabilized and stabilized tubes was analyzed to provide damping ratios from frequency-domain spectral responses, time-domain logarithmic decrement ratios and time-domain vibration decay-curves. The structural damping ratios typically increased from approximately 1.6% to approximately 4.3% with the addition of the stabilizer. This last value is somewhat less than recently published values for stabilized tubes from a different type of steam generator, suggesting that tube stabilization, while effective, has limitations that need to be conservatively assessed.

Sliding Wear Behaviors of Steam Generator Tubes in Various Environments

2006 ◽  
Vol 510-511 ◽  
pp. 566-569
Author(s):  
Gyung Guk Kim ◽  
Seung Dae Noh ◽  
Gi Sung Park ◽  
Seon Jin Kim ◽  
Deok Hyun Lee ◽  
...  
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Sliding Wear ◽  
Power Plants ◽  
Wear Behavior ◽  
Water Environment ◽  
Wear Volume ◽  
Wear Properties ◽  
Inconel 600 ◽  
Steam Generator Tubes

Wear damage of steam generator tubes for nuclear power plants can cause the leakage of radioactive substances. Therefore, the evaluation of integrity and safety for tubes is very important from the viewpoint of nuclear ecocide. In the present study, to investigate the wear properties of Inconel 600 and 690 steam generator tube materials mated with 409 stainless steel commonly used as support plate, sliding wear tests were performed with increasing sliding distance in air and in elevated temperature water environment, respectively. The wear volume of tube materials was less than those of supports under all conditions. There were no significant differences in the wear behavior for the Inconel 600 and 690 tubes, independently of the testing environment.

Seismic Proving Test of Heavy Component With Energy Absorbing Support: Proving Test Results

2002 ◽  
Author(s):  
T. Iwatsubo ◽  
M. Konno ◽  
H. Abe ◽  
K. Kuroda ◽  
K. Tai ◽  
...  
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
High Performance ◽  
Large Scale ◽  
Shaking Table ◽  
Scale Model ◽  
Piping System ◽  
Heavy Component ◽  
Design Earthquake ◽  
Energy Absorbing

The Seismic Proving Test of Heavy Component with Energy Absorbing Support has been conducted to prove the reliability of advanced seismic technology, supporting heavy component such as PWR Steam Generator with large capacity energy absorbing supports. This paper introduces the result of the proving test, using 1/2.5 scale model of PWR Steam Generator supported by Lead Extrusion Damper (LED), conducted on 2000–2001. This project has been conducted by the Nuclear Power Engineering Corporation (NUPEC) since 1995, using a large-scale, high-performance shaking table at Tadotsu Engineering Laboratory under the sponsorship of the Ministry of Economy Trade and Industry. Up to 2.9 S2 (extreme design earthquake), LED showed stable energy absorbing performance and Steam Generator- primary piping system indicate no meaningful damage.

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