Influence of thermal expansion bend and tubesheet geometry on guided wave inspection of steam generator tubes of a fast breeder reactor

2021 ◽  
pp. 147592172098352
Author(s):  
MM Narayanan ◽  
V Arjun ◽  
Anish Kumar
Keyword(s):  
Thermal Expansion ◽  
Steam Generator ◽  
Mode Conversion ◽  
Guided Wave ◽  
Test Facility ◽  
Discrete Wavelet ◽  
Correlation Technique ◽  
Element Analysis ◽  
Flexural Mode ◽  
Steam Generator Tubes

Periodic assessment of steam generator tubes of a sodium-cooled nuclear reactor is very crucial for smooth operation of steam generators. To examine the integrity, an in-bore magnetostrictive transducer capable of launching and receiving longitudinal ultrasonic guided waves (L(0,2) mode) from the inner diameter side of a steam generator tube developed in-house is used. Preliminary tests conducted on defective steam generator tubes with thermal expansion bends (three successive bends) of the mockup steam generator test facility yield a good sensitivity of 20% wall thickness deep flaw (0.46-mm deep and 1-mm wide half-circumferential groove) and the location accuracy of 10 mm. In order to remove high noise, wavelet-based denoising using discrete wavelet transform is used which improves the signal-to-noise ratio by 5–10 dB. In addition, cross-correlation technique is also used to denoise and unambiguously identify the defect echoes amid noise and multiple reflections between the defects. Furthermore, influence of the thermal expansion bend and tubesheet–spigot structure on L(0,2) mode is studied using the finite element analysis. It is observed that in the thermal expansion (multiple) bend, axisymmetric L(0,2) mode becomes non-axisymmetric (maximum and minimum amplitudes at extrados and intrados, respectively) and undergoes mode conversion to a weak flexural mode F(1,3). The results are validated experimentally. In the tubesheet–spigot structure, L(0,2) mode is found to have ∼10% reflection from spigot–tubesheet transitions, and it is seen to mode convert to bulk waves in the tubesheet. In conclusion, thicker tubesheets are found to be better from the perspective of inspection.

A Study on the Behavior of Ultrasonic Guided Wave Mode in a Pipe Using a Comb Transducer

2005 ◽  
Vol 297-300 ◽  
pp. 2182-2186
Author(s):  
Ik Keun Park ◽  
Yong Kwon Kim ◽  
Youn Ho Cho ◽  
Won Joon Song ◽  
Yeon Shik Ahn ◽  
...  
Keyword(s):  
Frequency Analysis ◽  
Mode Conversion ◽  
Group Velocity Dispersion ◽  
Longitudinal Mode ◽  
Wave Mode ◽  
Guided Wave ◽  
Flexural Mode ◽  
Time Frequency Analysis ◽  
Time Frequency ◽  
Ultrasonic Guided Wave

A preliminary study of the behavior of ultrasonic guided wave mode in a pipe using a comb transducer for maintenance inspection of power plant facilities has been verified experimentally. Guided wave mode identification is carried out in a pipe using time-frequency analysis methods such as wavelet transform (WT) and short time Fourier transform (STFT), compared with theoretically calculated group velocity dispersion curves for longitudinal and flexural mode. The results are in good agreement with analytical predictions and show the effectiveness of using the time-frequency analysis method to identify the individual guided wave modes. And, It was found out that longitudinal mode (0, 1) is affected by mode conversion less than the other modes. Therefore, L (0, 1) is selected as a optimal mode for evaluating location of the surface defect in a pipe.

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.

Coalescence Criterion of Part-Through Wall Cracks in Steam Generator Tubes of Nuclear Power Plants

2004 ◽  
Author(s):  
Jeries Abou-Hanna ◽  
Timothy McGreevy ◽  
Saurin Majumdar ◽  
Amit J. Trivedi ◽  
Ashraf Al-Hayek
Keyword(s):  
Finite Element ◽  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Local Instability ◽  
Element Analysis ◽  
Failure Pressure ◽  
Ligament Failure ◽  
Steam Generator Tubes

In scheduling inspection and repair of nuclear power plants, it is important to predict failure pressure of cracked steam generator tubes. Nondestructive evaluation (NDE) of cracks often reveals two neighboring cracks. If two neighboring part-through cracks interact, the tube pressure, under which the ligament between the two cracks fails, could be much different than the critical burst pressure of an individual equivalent part-through crack. The ability to accurately predict the ligament failure pressure, called “coalescence pressure,” is important. The coalescence criterion, established earlier for 100% through cracks using nonlinear finite element analyses [1–3], was extended to two part-through-wall axial collinear and offset cracks cases. The ligament failure is caused by local instability of the radial and axial ligaments. As a result of this local instability, the thickness of both radial and axial ligaments decreases abruptly at a certain tube pressure. Good correlation of finite element analysis with experiments (at Argonne National Laboratory’s Energy Technology Division) was obtained. Correlation revealed that nonlinear FEM analyses are capable of predicting the coalescence pressure accurately for part-through-wall cracks. This failure criterion and FEA work have been extended to axial cracks of varying ligament width, crack length, and cases where cracks are offset by axial or circumferential ligaments. The study revealed that rupture of the radial ligament occurs at a pressure equal to the coalescence pressure in the case of axial ligament with collinear cracks. However, rupture pressure of the radial ligament is different from coalescence pressure in the case of circumferential ligament, and it depends on the length of the ligament relative to crack dimension.

Derivation of Transverse Tensile Properties of Alloy 690 Steam Generator Tubes Using Ring-Tensile Specimen and Finite Element Analysis

2018 ◽  
Author(s):  
Jongmin Kim ◽  
Minchul Kim
Keyword(s):  
Finite Element Analysis ◽  
Finite Element ◽  
Tensile Properties ◽  
Steam Generator ◽  
Tensile Specimen ◽  
Optimization Process ◽  
Element Analysis ◽  
Transverse Tensile ◽  
Alloy 690 ◽  
Steam Generator Tubes

Transverse tensile properties of the Alloy 690 steam generator tube used in Korean PWR(pressurized water reactor) nuclear power plant is one of the main properties to analyze creep rupture caused by the hoop stress in the severe accident conditions. Transverse tensile properties of steam generator tubes cannot be obtained directly machined from the tubes and it is also difficult to make miniaturized flat tensile specimens due to their geometry limitations. The ring-tensile specimens are widely used to measure the tensile properties in the circumferential direction, but membrane and bending stresses are mixed in the ring-tensile specimen during the test due to the shape of ring-tensile specimens. Therefore, obtaining the transverse tensile properties need the inverse analysis using finite element analysis and optimization process. In this study, ring-tensile tests were carried out using 3-piece loading mandrel. The stress-strain curve of ring-tensile specimen in transverse direction is derived by comparing the load-displacement curves of the finite element analysis with those of test. The load-displacement curves from tests agree well with the finite element results. Use of finite element analysis and optimization process as used in this work is effective to predict the transverse tensile properties of ring-tensile specimen.

Development of ultrasonic guided wave inspection methodology for steam generator tubes of prototype fast breeder reactor

Ultrasonics ◽  
2019 ◽  
Vol 93 ◽  
pp. 112-121 ◽  
Author(s):  
M.M. Narayanan ◽  
Anish Kumar ◽  
S. Thirunavukkarasu ◽  
C.K. Mukhopadhyay
Keyword(s):  
Steam Generator ◽  
Guided Wave ◽  
Fast Breeder Reactor ◽  
Ultrasonic Guided Wave ◽  
Steam Generator Tubes

Ultrasonic Flexural Torsional Guided Wave Pipe Inspection Potential

2003 ◽  
Author(s):  
Zongqi Sun ◽  
Li Zhang ◽  
Brian Gavigan ◽  
Takahiro Hayashi ◽  
Joseph L. Rose
Keyword(s):  
Mode Conversion ◽  
Three Dimensional ◽  
High Sensitivity ◽  
Guided Wave ◽  
Defect Size ◽  
Propagation Characteristics ◽  
Pipe Inspection ◽  
Flexural Mode ◽  
Axial Defects

The excitation and propagation characteristics of guided wave torsional flexural mode are exploited in this paper. Theoretical computations and experiments are carried out to show the angular profiles of the torsional flexural modes propagation characteristics and the subsequent natural focusing effects. Because of such inherent advantages as less mode conversion and high sensitivity to axial defects, torsional modes and focusing possibilities have great potential in pipe inspection. By combining longitudinal and torsional modes, defect characterizations including defect size, shape etc. can be determined by truly three-dimensional guided wave pipe inspection.

Experimental Research on Damages Detection of Pipeline Structure by Using PZT-Based Waves

2011 ◽  
Author(s):  
Shi Yan ◽  
Binbin He ◽  
Naizhi Zhao
Keyword(s):  
Damage Detection ◽  
Lead Zirconate Titanate ◽  
Smart Materials ◽  
Arrival Time ◽  
Mode Conversion ◽  
Guided Wave ◽  
The Other ◽  
Damage Location ◽  
The Waves ◽  
Peak Value

Pipeline structure may generate damages during its service life due to the influence of environment or accidental loading. The damages need to be detected and repaired if they are severe enough to influence the transportation work. Non-destructive detection using smart materials combined with suitable diagonal algorithms are widely used in the field of structural health monitoring (SHM). Piezoelectric ceramics (such as Lead Zirconate Titanate, PZT) is one of the smart materials to be applied in the SHM due to the piezoelectric effect. So far, the PZT-based wave method is widely used for damage detection of structures, in particular, pipeline structures. A series of piezoelectric patches are bonded on the surface of the pipeline structure to monitor the damages such as local crack or effective area reduction due to corrosion by using diagonal waves. The damage of the pipeline structure can be detected by analysis of the received diagonal waves which peak value, phase, and arriving time can be deferent from the health ones. The response of the diagonal wave is not only correlated to the damage location through estimation of the arrival time of the wave peak, but also associated with the peak value of the wave for the reduction of wave energy as the guided wave passing through the damages. Therefore, the presence of damages in the pipeline structure can be detected by investigating the parameter change of the guided waves. The change of the wave parameters represents the attenuation, deflection and mode conversion of the waves due to the damages. In addition, the guided wave has the ability of quick detecting the damage of the pipeline structure and the simplicity of generating and receiving detection waves by using PZT patches. To verify the proposed method, an experiment is designed and tested by using a steel pipe bonded the PZT patches on the surface of it. The PZT patches consist of an array to estimate the location and level of the damage which is simulated by an artificial notch on the surface of the structure. The several locations and deep heights of the notches are considered during the test. A pair of the PZT patches are used at the same time as one is used as an actuator and the other as a sensor, respectively. A tone burst of 5 cycles of wave shape is used during the experiment. A wave generator is applied to create the proposed waves, and the waves are amplified by an amplifier to actuate the PZT patch to emit the diagonal waves with appropriately enough energy. Meanwhile, the other PZT patch is used as a sensor to receive the diagonal signals which contain the information of the damages for processing. For data processing, an index of root mean square deviation (RMSD) of the received data is used to estimate the damage level by compare of the data between the damaged and the health peak valves of the received signals. The time reversal method which aimed at increasing the efficiency of the detection is also used to detect the damage location by estimating the arrival time of the reflected wave passing with a certain velocity. The proposed method experimentally validates that it is effective for application in damage detection of pipeline structure.

scholarly journals Longitudinal Guided Wave Inspection of Small-Diameter Elbowed Steel Tubes with a Novel Squirrel-Cage Magnetostrictive Sensor

2019 ◽  
Vol 2019 ◽  
pp. 1-9
Author(s):  
Yao Liu ◽  
Xiucheng Liu ◽  
Chehua Yang ◽  
Wenxin Guo ◽  
Bin Wu ◽  
...  
Keyword(s):  
Finite Element ◽  
Guided Waves ◽  
Mode Conversion ◽  
Small Diameter ◽  
Longitudinal Mode ◽  
Guided Wave ◽  
Wall Defect ◽  
Squirrel Cage ◽  
Simulation Results ◽  
Magnetostrictive Sensor

In the study, ultrasonic longitudinal mode guided waves were employed to detect defects in elbowed tubes (without welds) with a diameter of 10 mm. Finite element simulation results highlighted that the emitted L(0,1) mode guided waves experienced strong reflection and mode conversion at the elbow region to generate F(1,1) mode, followed by slow and weak F(2,1) mode. The guided wave reflected from the elbow with a through-wall defect was manifested as two overlapped wave packets, which were good indicators of a defective elbow. To conduct L(0,1) mode guided waves inspection on the small-diameter elbowed tubes, a novel tailored squirrel-cage magnetostrictive sensor was employed in the experiment. The new sensor employed the configuration of segmental iron-cobalt strips and small-size permanent magnet arrays. The entire sensor is composed of two identical C-shaped sensor elements and can be recycled and installed conveniently. Experimental results obtained from healthy and defective tubes were consistent with the conclusions obtained from finite element simulations. An artificial through-wall defect at the elbow and a notch defect at the straight part of the tube could be simultaneously detected by L(0,1) mode guided waves through comparing experimental signals with simulation results.

Parametric Study of External Pressure on Steam Generator Tube Bends

2012 ◽  
Author(s):  
Mitch Hokazono ◽  
Clayton T. Smith
Keyword(s):  
Steam Generator ◽  
Parametric Study ◽  
External Pressure ◽  
Light Water Reactor ◽  
Material Strength ◽  
Steam Generators ◽  
Elliptical Cross ◽  
Light Water ◽  
Water Reactor ◽  
Steam Generator Tubes

Integral light-water reactor designs propose the use of steam generators located within the reactor vessel. Steam generator tubes in these designs must withstand external pressure loadings to prevent buckling, which is affected by material strength, fabrication techniques, chemical environment and tube geometry. Experience with fired tube boilers has shown that buckling in boiler tubes is greatly alleviated by controlling ovality in bends when the tubes are fabricated. Light water reactor steam generator pressures will not cause a buckling problem in steam generators with reasonable fabrication limits on tube ovality and wall thinning. Utilizing existing Code rules, there is a significant design margin, even for the maximum differential pressure case. With reasonable bend design and fabrication limits the helical steam generator thermodynamic advantages can be realized without a buckling concern. This paper describes a theoretical methodology for determining allowable external pressure for steam generator tubes subject to tube ovality based on ASME Section III Code Case N-759-2 rules. A parametric study of the results of this methodology applied to an elliptical cross section with varying wall thicknesses, tube diameters, and ovality values is also presented.

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