• Journal of the Korean Society for Nondestructive Testing
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• v.29 no.2
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• pp.98-103
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• 2009

The nuclear power plants have been safely operated by plugging the steam generator tubes which have the crack indications. Tube rupture events can occur if analysts fail to detect crack indications during in-service inspection. There are various types of crack indication in steam generator tubes and they have been detected by the eddy current test. The integrity assessment should be performed using the crack-sizing results from eddy current data when the crack indication is detected. However, it is not easy to evaluate the crack-depth precisely and consistently due to the complexity of the methods. The current crack-sizing methods were reviewed in this paper and the suitable ones were selected through the laboratory tests. The retired steam generators of Kori Unit 1 were used for this study. The round robin tests by the domestic qualified analysts were carried out and the statistical models were introduced to establish the appropriate depth-sizing techniques. It is expected that the proposed techniques in this study can be utilized in the Steam Generator Management Program.

• Nuclear Engineering and Technology
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• v.50 no.7
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• pp.1160-1167
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• 2018

Detectability of the loose parts (LPs) in steam generator (SG) was studied with eddy current testing technique such as X-probe, bobbin and rotating coils ($MRPC^{(R)}$) as a function of LP size and spacing between LP and tube or between LP and support structures. SG mockup simulating SG tube and support structures with LP was fabricated. The X-probe showed slightly better detectability than $MRPC^{(R)}$ for LP of ferrous (F-LP) material and vice versa for LP of nonferrous (NF-LP) material. In terms of feasibility, inspection rate and other predictable features of the SG tubing inspections, X-probe can be used reliably for monitoring the LPs and the flaws formed by LPs on SG tubes.

• Archives of Metallurgy and Materials
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• v.64 no.2
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• pp.543-545
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• 2019

The safe and reliable operation of pressurized water reactors (PWRs) depends on the integrity of structural material. In particular, the failure of steam generator (SG) tubes on the secondary side is one of the major concerns of operating nuclear power plants. To establish remediation techniques and manage damage, it is necessary to articulate the mechanism through which various impurities affect the SG tubes. This research aims to understand the effect of impurities (e.g., S, Pb, and Cl) on the stress corrosion cracking of Alloy 600 and 690.

• Journal of the Korean Society for Nondestructive Testing
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• v.33 no.5
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• pp.465-471
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• 2013

A steam generator in nuclear power plant is a heatexchager which is used to convert water into steam from heat produced in a nuclear reactor core, and the steam produced in steam generator is delivered to the turbine to generate electricity. Because of damage to steam generator tubing may impair its ability to adequately perform required safety functions in terms of both structural integrity and leakage integrity, eddy current testing is periodically performed to evaluate the integrity of tubes in steam generator. This assessment is normally performed during a reactor refueling outage. Currently, the eddy current testing for steam generator of nuclear power plant in Korea is performed in accordance with KEPIC & ASME Code requirements, the eddy current testing system is consists of remote data acquisition unit and data analysis program to evaluate the acquired data. The KEPIC & ASME Code require that the electrical properties of remote data acquisition unit, such as total harmonic distortion, input & output impedance, amplifier linearity & stability, phase linearity, bandwidth & demodulation filter response, analog-to-digital conversion, and channel crosstalk shall be measured in accordance with the KEPIC & ASME Code requirements. In this paper, the measurement requirements of electrical properties for eddy current testing instrument described in KEPIC & ASME Code are presented, and the measurement results of newly developed eddy current testing instrument by KHNP(Korea Hydro & Nuclear Power Co., LTD) are presented.

• Journal of the Korean Society for Nondestructive Testing
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• v.28 no.2
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• pp.184-190
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• 2008

Various ECT techniques have been applied basically to assess the integrity of steam generator tithing in nuclear power plant. Among these techniques, the bobbin probe technique is applied generally to examine the volumetric flaws such as a crack-like defect and wear which is generally occurred on steam generator tubing, and additionally MRPC probe is used to examine closely tile top of tubesheet and bending regions due to the high possibility of cracking. Dent and bulge also may be formed on tube during installation process and operation of steam generator, but the dent and bulge indications greater than specific size criteria are recorded on examination report because these indications are not considered as flaw. These indications can be easily detected with bobbin probe and approximately sized with profile bobbin probe, but the size and shape can not be accurately verified. Accordingly, in this study, the $8{\times}1$ multi-array EC probe was designed to increase the measurement accuracy of the sectional profiling EC testing of tube. As a result, we would like to propose the application of $8{\times}1$ multi-array EC probe for the measurement of size and shape of profile change on steam generator tube in OPR-1000 nuclear power plant.

• Journal of the Korean Society for Nondestructive Testing
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• v.20 no.1
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• pp.1-9
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• 2000

This study shows the defect detection and sizing capability of ultrasonic guided waves in the nondestructive inspection of heat exchanger and steam generator tubing. Phase and group velocity dispersion curves for the longitudinal and flexural modes of a sample Inconel tube were presented for the theoretical analysis. EDM(Electric Discharge Machining) wears in tubing under a tube support plate and circumferential laser notches in tubing were detected by an axisymmetric and a non-axisymmetric transducer set up, respectively. EDM wears were detected with L(0, 2), L(0, 3) and L(0, 4) modes and among them L(0, 4) mode was found to be the most sensitive. It was also found that the flexural modes around L(0, 1) mode could be used for the detection and sizing of laser notches in the tubing.

• Nuclear Engineering and Technology
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• v.45 no.2
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• pp.219-222
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• 2013

Alloy 690 has been selected as a steam generator tubing material for SMART owing to a near immunity to primary water stress corrosion cracking. The steam generators of SMART are faced with a neutron flux due to the integrated arrangement inside a reactor vessel, and thus it is important to know the irradiation effects of the thermal conductivity of Alloy 690. Alloy 690 was irradiated at HANARO to fluences of (0.7-28) ${\times}10^{19}n/cm^2$ (E>0.1MeV) at $250^{\circ}C$, and its thermal conductivity was measured using the laser-flash equipment in the IMEF. The thermal conductivity of Alloy 690 was dependent on temperature, and it was a good fit to the Smith-Palmer equation, which modified the Wiedemann-Franz law. The irradiation at $250^{\circ}C$ did not degrade the thermal conductivity of Alloy 690, and even showed a small increase (1%) at fluences of (0.7~28) ${\times}10^{19}n/cm^2$ (E>0.1MeV).

• Corrosion Science and Technology
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• v.7 no.3
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• pp.187-191
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• 2008

Transgranular stress corrosion cracking of nickel base alloys was reported by Copson and Dean in 1965. Study to establish this cracking mechanism needs to be carried out. Laboratory stress corrosion tests were performed for mill annealed(MA) or thermally treated(TT) steam generator tubing materials in a high temperature water containing lead. An electrochemical interaction of lead with the alloying elements of SG tubings was also investigated. Alloy 690 TT showed a transgranular stress corrosion cracking in a 40% NaOH solution with 5000 ppm of lead, while intergranular stress corrosion racking was observed in a 10% NaOH solution with 100 ppm lead. Lead seems to enhance the disruption of passive film and anodic dissolution of alloy 600 and alloy 690. Crack tip blunting at grain boundary carbides plays a role for the transgranular stress corrosion cracking.

• Proceedings of the KSME Conference
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• 2004.04a
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• pp.1120-1125
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• 2004

Welding characteristics of STS304 stainless steel and Inconel 600 using a continuous wave Nd:YAG laser are experimentally investigated. Alloy 600 being used in steam generator tubing of pressurized water reactor(PWR) exposed to some corrosion environment, stress corrosion cracking can occur on this material. Presented here are the results from a series of experiments in which dissimilar metal welds were made using the gas tungsten arc welding process with pure argon shielding gas. But It is well known that solidification cracking susceptibility of austenitic alloys depends on the solidification temperature range and amount/distribution of solute rich liquid that exists at the terminal stages of solidification. An experimental study was conducted to determine effects of welding parameters and to optimize those parameters that have the most influence on eliminating or reducing the extent welding zone formation at dissimilar metal welds.

• Journal of the Korean Society for Precision Engineering
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• v.21 no.11
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• pp.53-60
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• 2004

Laser welding techniques have been characterised for various materials. In this paper, the laser weldability of STS304 stainless steel and Inconel 600 at dissimilar metal welds using a continuous wave Nd:YAG laser are experimentally investigated. Inconel 600 is being used in a steam generator tubing of pressurized water reactor(PWR) exposed to some corrosion. Stress corrosion cracking can occur on this material. An experimental study was conducted to determine effects of welding parameters, on eliminating or reducing the extent welding zone formation at dissimilar metal welds and to optimize those parameters that have the most influence parameters such as focus length, power, beam speed, shielding gas, and wave length of laser were tested.