Optimisation of helical geometry of inspection probe for steam generator tubes of the PFBR

The steam generator (SG) tubes of the prototype fast breeder reactor (PFBR) located in Kalpakkam, India, need to be periodically inspected using the remote field eddy current (RFEC) technique. During the pre-service inspection of the SG tubes, it was found that the RFEC probes experienced frequent mechanical breakages. To avoid these failures, changes in the existing structural design of the RFEC probe were required. A helical groove design was proposed to obtain a smooth transition in the variation of stress across the probe during the inspection. It was difficult to calculate the flexural stiffness of the proposed helical geometry probe due to the varying cross-section along its length. In this paper, the smearing approach adopted to calculate the stiffness of the RFEC probe and the sensitivity analysis carried out to determine the optimal design of the probe are discussed. A probe was fabricated based on the helical groove design and tested to qualify its suitability for the SG inspection. The RFEC probe with helical grooves was employed for the pre-service inspection of the SG tubes of the PFBR. More than 200 tubes have been inspected using the proposed design and no mechanical failure of the probe has been observed.

Evaluation of the Creep Rupture Behavior of Alloy 690 Steam Generator Tubes Considering the Pressure Ramp Rate

The materials used previously for steam generator tubes around the world have been replaced and will be replaced by Alloy 690 given its improved corrosion resistance relative to that of Alloy 600. However, studies of the high- temperature creep and creep-rupture characteristics of steam generator tubes made of Alloy 690 are insufficient compared to those focusing on Alloy 600. In this study, several creep tests were conducted using half tube shape specimens of the Alloy 690 material at temperatures ranging from 650 to 850C and stresses in the range of 30 to 350 MPa, with failure times to creep rupture ranging from 3 to 870 hours. Based on the creep test results, creep life predictions were then made using the well-known Larson Miller Parameter method. Steam generator tube rupture tests were also conducted under the conditions of a constant temperature and pressure ramp using steam generator tube specimens. The rupture test equipment was designed and manufactured to simulate the transient state (rapid temperature and pressure changes) in the event of a severe accident condition. After the rupture test, the damage to the steam generator tubes was predicted using a creep rupture model and a flow stress model. A modified creep rupture model for Alloy 690 steam generator tube material is proposed based on the experimental results. A correction factor of 1.7 in the modified creep rupture model was derived for the Alloy 690 material. The predicted failure pressure was in good agreement with the experimental failure pressure.

A Bayesian updating of crack distributions in steam generator tubes

The structural failure of steam generator tubes is a common problem that can a ect the availability and safety of nuclear power plants. To minimize the probability of occurrence of failure, it is needed to implement maintenance strategies such as periodic nondestructive inspections of tubes. Thus, a tube is repaired or plugged whenever it has detected a crack which a threshold size is overtaken. In general, uncertainties and errors in crack sizes are associated with the nondestructive inspections. These uncertainties and errors should be appropriately characterized to estimate the actual crack distribution. This work proposes a Bayesian approach for updating crack distributions, which in turn allows computing the failure probability of steam generator tubes at current and future times. The failure criterion is based on plastic collapse phenomenon, and the failure probability is computed by using the Monte-Carlo simulation. The failure probability at current and future times is in good agreement with the ones presented in the literature.

Numerical Study of the Deposition of Radioactive Nucleoids (Aerosol) in PWR Steam Generator

This research presents numerical study of deposition efficiency and decontamination factor of radioactive nuclide in steam generator tubes of a typical 325 MWe PWR. To find out the deposition of aerosol, the discrete phase model (DPM) has been used. The flow has been characterized as compressible, adiabatic, turbulent and wall bounded. When steam generator tube gets ruptured, the radioactive nuclides can escape from primary side and create a radioactive field in the secondary side. This can be harmful for the personnel working at the plant. Therefore, in order to ensure the safety of the plant and personnel, it is important to study the particles deposition on the wall of steam generator tubes. In the present study, a CFD methodology has been first developed and validated with the published results. After methodology validation, it has been applied to the U-tube of a typical PWR steam generator. It has been observed that due to the action of centrifugal force near the bent, the velocity magnitude is high towards the inner wall and the flow separates at the bent entrance. Furthermore, the flow inside the tube is rotational with vortices throughout the domain due to the presence of the bent. Finally, the deposition efficiency and decontamination factor have been calculated and it has been observed that both increase with the increase in particle size due to inertial effects.

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

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.

Analysis of Eddy-Current Probe Signals in Steam Generator U-Bend Tubes Using the Finite Element Method

To ensure the integrity and safety of steam generator tubes in nuclear power plants, eddy-current testing is periodically employed. Because steam generators are equipped with thousands of thin-walled tubes, the eddy current is tested using a bobbin probe that can be used at high speed. Steam generator heat pipes in nuclear power plants have different sizes and shapes depending on their row number. In particular, heat pipes in the first row are located inside the steam generator and are of the U-bend type because the radius of the curved pipe is the smallest. A steam generator heat pipe has a thickness of about 1 mm, so a geometrical cross-sectional area change may occur due to residual stress when manufacturing the curved pipe. It is difficult to determine an exact shape because the change in cross-sectional area generated during the manufacturing process varies depending on the position of the pipe and the distortion rate. During eddy-current testing (ECT), to ensure the integrity and safety of the steam generator tubes, the shape change of the bend may cause a noise signal, making it difficult to evaluate defects in the pipe. However, the noise signals generated in this situation are difficult to analyze in a real measurement environment, and difficult to verify by producing a mock-up, which complicates distinguishing a noise signal from a defective signal. To solve this problem, various noise signals were obtained using the electromagnetic analysis method of COMSOL Multiphysics, a commercial program based on numerical analysis of the finite element method, to simulate the environment, including the change in cross-sectional area of the heat pipe. When compared to the actual measurement signal, the accuracy of the simulations improved, and various types of noise signals were detected, which may be helpful for accurate evaluations of defects during actual inspections.