• Journal of the Korea Academia-Industrial cooperation Society
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• v.9 no.6
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• pp.1569-1573
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• 2008

This paper describes a development of Ultrasonic NDE software to analyze steam generator of nuclear power plant. The developed software includes classical analysis method such as A, B, C and D-scan images. And it can analyze the detected internal cracks using 3D image processing method. To do such, we obtain raw data from specimens of real pipeline of power plants, and get the envelope signal using Empirical Mode Decomposition from obtained ultrasonic 1-dimensional data. The reconstructed 3D crack images offer useful information about the location, shape and size of cracks, even if there is no special 2D image analysis technique. The developed analysis software is applied to specimens containing various cracks with known dimensions. The results of application showed that the developed software provided accurate and enhanced 2D images and reconstructed 3D image of cracks.

• Journal of the Korean Society for Nondestructive Testing
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• v.24 no.4
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• pp.384-389
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• 2004

As the increase of the operation year of nuclear power plants, the probabilities of the degradation of the major facilities and materials in the nuclear power plants are increased. The integrity of those facilities shall be monitored and verified by the non-destructive examination methods with the regulation codes, so called inservice inspection(ISI). The ISI of Yonggwang unit 6 was performed in four different parts, 1) non-destructive examinations for the components, piping weldments and structures, 2) automated ultrasonic examinations for pressure vessels, 3) visual examinations for the interior structures of the reactor, 4) eddy current examinations for the steam generator tubes. As the results, there was no severe indication and all detected indications were evaluated as non-relavent. Especially for the examinations of the piping weldments, PD(Performance Demonstration) was applied as a W examination method defined in the 1995 edition of ASME Code Sec. XI. The implementation of the PD for the piping weld results in an improvement of the reliability of the UT examinations.

• Journal of the Korean Society for Nondestructive Testing
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• v.35 no.1
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• pp.61-67
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• 2015

Recently, the initiation of outside diameter stress corrosion cracking (ODSCC) at the tube support plate region of domestic steam generators (SG) with Alloy600 HTMA tubes has been increasing. As a result, SGs with Alloy600 HTMA tubes must be replaced early or are scheduled to be replaced prior to their designed lifetime. ODSCC is one of the biggest threats to the integrity of SG tubes. Therefore, the accurate evaluation of tube integrity to determine ODSCC is needed. Eddy current testing (ECT) is conducted periodically, and its results could be input as parameters for evaluating the integrity of SG tubes. The reliability of an ECT inspection system depends on the performance of the inspection technique and abilty of the analyst. The detection probability and ECT sizing error of degradation are considered to be the performance indices of a nondestructive evaluation (NDE) system. This paper introduces an optimized evaluation method for ECT, as well as the sizing error, including the analyst performance. This study was based on the results of a round robin program in which 10 inspection analysts from 5 different companies participated. The analysis of ECT sizing results was performed using a linear regression model relating the true defect size data to the measured ECT size data.

• Journal of the Korean Society for Nondestructive Testing
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• v.32 no.4
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• pp.410-417
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• 2012

Currently domestic thermal and nuclear power plants are comprised of many type's condenser and steam generator tubes to produce the electricity of good quality. There are some methods to inspect these tubes in the event that several defects were discovered in these facilities. Among many non-destructive methods, we used guided wave to inspect the soundness of tubes, because this method is very fast to detect the defect and very simple to install the equipment and also, can inspect up to the long range at a fixed point. Also, this method has a drawback that does not detect a very small size defect. So, we made an effort to overcome this drawback through the experimentation and signal analysis according to the size and shape of the defect through the manufacture of various artificial cracks capable to generate within the small size tube in the study and we anticipate that these detect limits can be overcome along with the development of the signal processing and manufacturing technology of the sensor for the inspection.

• Journal of the Korean Society for Nondestructive Testing
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• v.26 no.2
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• pp.90-98
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• 2006

Signals captured from a Combo calibration standard tube paly a crucial role in the evaluation of motorized rotating pancake coil (MRPC) probe signals from steam generator (SG) tubes in nuclear power plants (NPPs). Therefore, the Combo tube signals should be consistent and accurate. However, MRPC probe signals are very easily affected by various factors around the tubes so that they can be distorted in their amplitudes and phase angles which are the values specifically used in the evaluation. To overcome this problem, in this study, we explored possibility of simulation to be used as a practical calibration tool far the evaluation of real field signals. For this purpose, we investigated the characteristics of a MRPC probe and a Combo tube. And then using commercial software (VIC-3D) we simulated a set of calibration signals and compared to the experimental signals. From this comparison, we verified the accuracy of the simulated signals. Finally, we evaluated two defects using the simulated Combo tube signals, and the results were compared with those obtained using the actual field calibration signals.

• Journal of Radiation Protection and Research
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• v.38 no.4
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• pp.228-233
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• 2013

It has been continuously generated the requirement for the replacement of the main components such as a steam generator due to the deterioration of the nuclear power plant all around the world. Also, a large amount of radioactive metal was generated during the decommissioning in a short period. It is required to make an accurate measurement of the residual radioactivity for recycling the metal waste for releasing from regulatory control. In planning the measurement procedures, the influence of geometry, self-absorption, density and other relevant factors on the representativeness of the measurements should be considered for the decommissioning metal waste. In this study, the method for measurement procedures, the source term evaluation, the ways to secure representative samples, the measurement device for wide area and the self-absorption correction factors for different density were evaluated. The metal samples for measurement were prepared for securing the simple geometry and representative by melting process. The developed correction method for measuring the radioactivity a variety density of metal waste could improve the reliability of the evaluation results for clearance.

• Journal of Radiation Protection and Research
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• v.29 no.3
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• pp.187-196
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• 2004

The health effects resulting from severe accidents of typical 1,000MWe KSNP(Korea Standard Nuclear Plant) PWR and typical 600MWe CANDU(CANada Deuterium Uranium) plants were estimated and compared. The population distribution of the site extending to 80km for both site were considered. The releaese fraction for various source term categories(STC) and core inventories were used in the estimation of the health effects risks by using the MACCS2(MELCOR Accident Consequence Code System2) code. Individuals are assumed to evacuate beyond 16km from the site. The health effects considered in this comparative study are early and cancer fatality risk, and the results are presented as CCDF(Complementary Cumulative Distribution Function) curves considering the occurrence probability of each STC's. According to the results, the early and cancer fatality risks of PHWR plants we lower than those of PWR plants. This is attributed the fact that the amount of radioactive mateials that released to the atmosphere resulting from the postulated severe accidents of PHWR plants are smaller than that of PWR plants. And, the dominating initiating event of STC that shows maximum early and cancer fatality risk is SGTR(Steam Generator Tube Rupture) for both plants. Therefore, the appropriated actions must be taken to reduce the occurrence probability and the amounts of radioactive materials released to the environment in order to protect the public for both PWR and PHWR plants.