Study on the Stream-Wise Fluidelastic Instability of Rotated Square Arrays of Circular Cylinders Subjected on Cross-Flow

2017 ◽  
Author(s):  
Tomomichi Nakamura ◽  
Tomoki Tsujita
Keyword(s):  
Nuclear Power ◽  
Cross Flow ◽  
Critical Factor ◽  
Circular Cylinders ◽  
Test Facility ◽  
Test Results ◽  
Steam Generators ◽  
Triangular Arrays ◽  
Specific Direction ◽  
Fluidelastic Instability

Stream-wise fluidelastic instability has recently been in the spotlight due to a practical event in steam generators at a nuclear power plant [1]. The instability has been reported to occur easily in rotated triangular arrays, but at all hardly in normal triangular arrays [2][3]. In square arrays, no instability has been reported in the authors’ test facility [4]. This paper presents the test results on rotated square arrays of the pitch-diameter ratios from 1.2 to 1.5. The tests have been conducted with a wind tunnel, where cylinders have been constrained to move in only one specific direction, either stream-wise or transverse to the flow. The results indicate that fluidelastic vibrations occur only in the stream-wise direction. The critical factor corresponds to the previous data reported in some guidelines [5], but here the factor is related mainly to stream-wise instability.

Investigation of In-Flow Fluidelastic Instability of Square Tube Arrays Subjected to Air Cross Flow

2015 ◽  
Author(s):  
Tomomichi Nakamura ◽  
Shinichiro Hagiwara ◽  
Joji Yamada ◽  
Kenji Usuki
Keyword(s):  
Nuclear Power ◽  
Flow Instability ◽  
Flow Direction ◽  
Cross Flow ◽  
Diameter Ratio ◽  
Nuclear Industry ◽  
Flexible Cylinder ◽  
Triangular Arrays ◽  
Tube Arrays ◽  
Fluidelastic Instability

In-flow instability of tube arrays is a recent major issue in heat exchanger design since the event at a nuclear power plant in California [1]. In our previous tests [2], the effect of the pitch-to-diameter ratio on fluidelastic instability in triangular arrays is reported. This is one of the present major issues in the nuclear industry. However, tube arrays in some heat exchangers are arranged as a square array configuration. Then, it is important to study the in-flow instability on the case of square arrays. The in-flow fluidelastic instability of square arrays is investigated in this report. It was easy to observe the in-flow instability of triangular arrays, but not for square arrays. The pitch-to-diameter ratio, P/D, is changed from 1.2 to 1.5. In-flow fluidelastic instability was not observed in the in-flow direction. Contrarily, the transverse instability is observed in all cases including the case of a single flexible cylinder. The test results are finally reported including the comparison with the triangular arrays.

Experimental Investigation of Subcooled Choking Flow in a Steam Generator Tube Crack

2016 ◽  
Author(s):  
Hung Nguyen ◽  
Mark Brown ◽  
Shripad T. Revankar ◽  
Jovica Riznic
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Steam Generator ◽  
Nuclear Power ◽  
Test Facility ◽  
Experimental Program ◽  
Steam Generators ◽  
Steam Generator Tube ◽  
Loss Of Coolant ◽  
Steam Generator Tubes

Steam generator tubes have a history of small cracks and even ruptures, which lead to a loss of coolant from the primary side to the secondary side. These tubes have an important role in reactor safety since they serve as one of the barriers between radioactive and non-radioactive materials of a nuclear power plant. A rupture then signifies the loss of the integrity of the tube itself. Therefore, choking flow plays an integral part not only in the engineered safeguards of a nuclear power plant, but also to everyday operation. There is limited data on actual steam generators tube wall cracks. Here experiments were conducted on choked flow of subcooled water through two samples of axial cracks of steam generator tubes taken from US PWR steam generators. The purpose of the experimental program was to develop database on critical flow through actual steam generator tube cracks with subcooled liquid flow at the entrance. The knowledge of this maximum flow rate through a crack in the steam generator tubes of a pressurized water nuclear reactor will allow designers to calculate leak rates and design inventory levels accordingly while limiting losses during loss of coolant accidents. The test facility design is modular so that various steam generator tube cracks can be studied. Two sets of PWR steam generators tubes were studied whose wall thickness is 1.285 mm. Tests were carried out at stagnation pressure up to 6.89 MPa and range of subcoolings 16.2–59°C. Based on these new choking flow data, the applicability of analytical models to highlight the importance of non-equilibrium effects was examined.

Effect of Angle of Attack on Fluidelastic Instability of a Rotated Triangular Tube Bundle

2007 ◽  
Author(s):  
A. Khalvatti ◽  
N. W. Mureithi ◽  
M. J. Pettigrew
Keyword(s):  
Hydraulic Resistance ◽  
Tube Bundle ◽  
Angle Of Attack ◽  
Cross Flow ◽  
Point Of View ◽  
Steam Generators ◽  
Fluidelastic Instability ◽  
Shell And Tube ◽  
The Stability ◽  
Resistance Forces

In the operation of shell-and-tube heat exchangers, vibration of the tubes can be induced by fluid flowing over the tube array in cross flow. The region of concern in Steam Generators (SG) is the upper most U-bend region where the flow crosses a large number of tubes which can cause significant hydraulic resistance. This hydraulic resistance forces the flow to change direction. From a fluidelastic instability point of view, the tube bundle is excited by oblique cross flow. The purpose of this paper is to examine the instability phenomena in a rotated triangular tube bundle subjected to oblique single phase cross flow. In this present work tests are conducted in a wind tunnel on a rotated triangle tube array. Fluidelastic instability results are in agreement with what was expected. The results show that fluidelastic instability is strongly dependent on the angle of attack. The results also show that, generally, the elimination of bundle flexibility in the direction transverse to the flow, greatly affects the stability behavior of the array.

Electroslag Strip Cladding of Steam Generators With Alloy 690

2006 ◽  
Author(s):  
M. Consonni ◽  
F. Maggioni ◽  
F. Brioschi
Keyword(s):  
Nuclear Power ◽  
Filler Metal ◽  
Hot Cracking ◽  
Test Results ◽  
Steam Generators ◽  
High Productivity ◽  
Cladding Layer ◽  
Power Stations ◽  
Alloy 690 ◽  
Submerged Arc

The present paper details the results of electroslag cladding and tube-to-tubesheet welding qualification tests conducted by Ansaldo-Camozzi ESC with Alloy 690 (Alloy 52 filler metal) on steel for nuclear power stations’ steam generators shell, tubesheet and head; the possibility of submerged arc cladding on first layer was also considered. Test results, in terms of chemical analysis, mechanical properties and microstructure are reproducible and confidently applicable to production cladding and show that electroslag process can be used for Alloy 52 cladding with exceptionally stable and regular operation and high productivity. The application of submerged arc cladding process to the first layer leads to a higher base metal dilution, which should be avoided. Moreover, though the heat affected zone is deeper with electroslag cladding, in both cases no coarsened grain zone is found due to recrystallisation effect of second cladding layer. Finally, the application of electroslag process to cladding of Alloy 52 with modified chemical composition, was proved to be highly beneficial as it strongly reduces hot cracking sensitivity, which is typical of submerged arc cladded Alloy 52, both during tube-to-tubesheet welding and first re-welding.

Vibration and Work-Rate Measurements of Steam-Generator U-Tubes in Air-Water Cross-Flow

2002 ◽  
Author(s):  
Victor P. Janzen ◽  
Erik G. Hagberg ◽  
James N. F. Patrick ◽  
Michel J. Pettigrew ◽  
Colette E. Taylor ◽  
...  
Keyword(s):  
Cross Flow ◽  
Work Rate ◽  
Vibration Response ◽  
Fretting Wear ◽  
Steam Generators ◽  
Two Phase ◽  
Void Fractions ◽  
Vibration Properties ◽  
Wide Range ◽  
Fluidelastic Instability

In nuclear power plant steam generators, the vibration response of tubes in two-phase cross-flow is a general concern that in some cases has become a very real long-term wear problem. This paper summarizes the results of the most recent U-bend vibration-response tests in a program designed to address this issue. The tests involved a simplified U-tube bundle with a set of flat-bar supports at the apex, subjected to two-phase air-water cross-flow over the mid-span region of the U-bend. Tube vibration properties and tube-to-support interaction in the form of work-rates were measured over a wide range of flow velocities for homogeneous void fractions from zero to 90%, with three different tube-to-support clearances. The measured vibration properties and work-rates could be characterized by the relative influence of the two most important flow-induced excitation mechanisms at work, fluidelastic instability and random-turbulence excitation. As in previous similar tests, strong effects of fluidelastic instability were observed at zero and 25% void fraction for pitch velocities greater than approximately 0.5 m/s, whereas random turbulence dominated the tube vibration and work-rate response at higher void fractions. In both cases, a link between vibration properties and the effect of the flat-bar supports could be established by comparing the vibration crossing frequency, extracted from time-domain vibration signals, to the participation of the lowest few vibration modes and to the measured work-rate. This approach may be useful when fluidelastic instability, random turbulence and loose supports all combine to result in high work-rates. Such a combination of factors is thought to be responsible for excessive U-tube fretting-wear in certain types of operating steam generators.

Investigation on In-Flow Fluidelastic Instability of an Array of Tubes

2013 ◽  
Author(s):  
Kazuo Hirota ◽  
Hideyuki Morita ◽  
Jun Hirai ◽  
Akihisa Iwasaki ◽  
Seiho Utsumi ◽  
...  
Keyword(s):  
Flow Velocity ◽  
Flow Direction ◽  
Cross Flow ◽  
Critical Flow ◽  
Steam Generators ◽  
Critical Flow Velocity ◽  
Fluidelastic Instability ◽  
Vibration Mechanism ◽  
The Cross ◽  
The Relationship

Fluidelastic instability (FEI) remains the most important vibration mechanism in steam generators. Fluidelastic instability of an array of tubes thought to be mainly occurred in the cross-flow direction. In the present day, some researchers reported possibility of occurrence of fluidelastic instability in the in-flow direction. However, the phenomenon of the in-plane FEI has not been well recognized compared to the transverse FEI. In this study, air flow tests using cantilevered straight cylinder array of tubes in triangular configuration were conducted. It is confirmed that the in-flow FEI could be occurred and the critical flow velocity in the in-flow direction is larger than that of in the cross flow direction. Furthermore, the relationship between P/D of an array of tubes and the critical flow velocity in the in-flow direction was also investigated.

The Effect of Angle of Attack on the Fluidelastic Instability of Tube Bundle Subjected to Two-Phase Cross-Flow

2009 ◽  
Author(s):  
T. F. Joly ◽  
N. W. Mureithi ◽  
M. J. Pettigrew
Keyword(s):  
Tube Bundle ◽  
Angle Of Attack ◽  
Cross Flow ◽  
Diameter Ratio ◽  
Test Results ◽  
Two Phase ◽  
Void Fractions ◽  
Fluidelastic Instability ◽  
Flexible Tubes ◽  
Existing Data

Tests were done to study the effect of angle of attack on the fluidelastic instability of a fully flexible tube bundle subjected to two-phase (Air-Water) cross-flow. A test array having nineteen flexible tubes in a rotated triangular configuration with a pitch-to-diameter ratio of 1.5 was tested. Four different angles of attack ranging for 0 degree (inline flexibility) through 30 and 60 degrees to 90 degrees (transverse flexibility) were studied. For each angle of attack several homogeneous void fractions have been tested (70%, 80%, 90%, and 95%). Stability test results show that the angle of attack strongly affect the tube bundle dynamic behavior. The different mechanisms underlying the fluidelastic instability are highlighted and the results compared to existing data on fluidelastic instability.

Investigations of In-Plane Fluidelastic Instability in a Multi-Span U-Bend Test Rig: Tests in Two-Phase Flow

2019 ◽  
Author(s):  
Paul Feenstra ◽  
Teguewinde Sawadogo ◽  
Bruce Smith ◽  
Victor Janzen ◽  
Anne McLellan ◽  
...  
Keyword(s):  
Steam Generator ◽  
Cross Flow ◽  
Test Results ◽  
Test Rig ◽  
Two Phase ◽  
Practical Concern ◽  
Fluidelastic Instability ◽  
Phase Liquid ◽  
R 134A ◽  
Steam Generator Tubes

Abstract Tests to study FluidElastic Instability (FEI) in an array of U-bend tubes were recently completed in the Multi-Span U-Bend (MSUB) test rig at Canadian Nuclear Laboratories (CNL). These tests were sponsored by the Electric Power Research Institute (EPRI) and were designed to study In-Plane (IP or streamwise) FEI of steam generator tubes in two-phase cross flow. This instability mechanism was first observed in previous research tests by Atomic Energy of Canada Limited (AECL). Although AECL planned additional research into IP FEI, this mechanism was not thought to be a serious practical concern until it recently caused severe damage to tubes in a new replacement steam-generator. The MSUB tests were conducted both with flows of air and two-phase liquid/vapour Refrigerant 134a. With 22 flexible U-bend tubes supported by a configurable flat-bar arrangement, the tests were focused on the effects of support geometry and tube-to-support interaction. Data was recorded from 33 dynamic signals from accelerometers, displacement probes, force transducers, and void-fraction probes. The paper describes the experimental test setup and reviews some of the initial test results and their implications for steam-generator users and researchers. Tests with two-phase Freon refrigerant (R-134a) are presented here.

Streamwise Fluidelastic Instability of Tube Arrays in Two-Phase Cross Flow

2014 ◽  
Author(s):  
Stephen Olala ◽  
Njuki W. Mureithi
Keyword(s):  
Nuclear Power ◽  
Single Phase ◽  
Cross Flow ◽  
Force Measurements ◽  
Two Phase Flows ◽  
Two Phase ◽  
Fluid Force ◽  
Void Fractions ◽  
Tube Arrays ◽  
Fluidelastic Instability

In-plane instability of tube arrays has not been a major concern to steam generator designers until recently following observations of streamwise tube failure in a nuclear power plant in U.S.A. However, modeling of fluidelastic instability in two-phase flows still remains a challenge. In the present work, detailed steady fluid force measurements for a kernel of an array of tubes in a rotated triangular tube array of P/D=1.5 subjected to air-water two-phase flows for a series of void fractions and a Reynolds number (based on the pitch velocity), Re = 7.2 × 104 has been conducted. The measured steady fluid force coefficients and their derivatives, with respect to streamwise static displacements of the central tube, are employed in the quasi-steady model [1, 2], originally developed for single phase flows, to analyze in-plane fluidelastic instability of multiple flexible arrays in two-phase flows. The results are consistent with dynamic stability tests [3].

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