• Journal of the Korean Society for Nondestructive Testing
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• v.25 no.1
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• pp.27-35
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• 2005

Ultrasonic inspection methods are widely used for detecting flaws in materials. The signal analysis step plays a crucial part in the data interpretation process. A number of signal processing methods have been proposed to classify ultrasonic flaw signals. One of the more popular methods involves the extraction of an appropriate set of features followed by the use of a neural network for the classification of the signals in the feature spare. This paper describes an alternative approach which uses the least mean square (LMS) method and exportation maximization (EM) algorithm with the model based deconvolution which is employed for classifying nondestructive evaluation (NDE) signals from steam generator tubes in a nuclear power plant. The signals due to cracks and deposits are not significantly different. These signals must be discriminated to prevent from happening a huge disaster such as contamination of water or explosion. A model based deconvolution has been described to facilitate comparison of classification results. The method uses the space alternating generalized expectation maximiBation (SAGE) algorithm ill conjunction with the Newton-Raphson method which uses the Hessian parameter resulting in fast convergence to estimate the time of flight and the distance between the tube wall and the ultrasonic sensor. Results using these schemes for the classification of ultrasonic signals from cracks and deposits within steam generator tubes are presented and showed a reasonable performances.

• 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.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.36 no.3
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• pp.313-318
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• 2012

In CANDU, feeder pipes supply heavy water to pressure tube and steam generator. Under service conditions, Flow-Accelerated Corrosion (FAC) produces local wall-thinning in the feeder pipes. The wall-thinning in these pipes affects the integrity of the piping system, as verified in previous research. This paper provides limit loads for wallthinned feeder pipes with $45^{\circ}$ and $60^{\circ}$ bend angles, and proposes an equation that predicts the limit loads for wallthinned feeder pipes with arbitrary bend angles. On the basis of finite element limit analyses, limit loads are obtained for wall-thinned feeder pipes under in-plane bending and internal pressure. There are two cases of in-plane bending: the in-plane closing direction and the in-plane opening direction. The material is considered the effect of the large deformation, so an elastic-perfectly-plastic material is assumed in the calculations.

• Journal of the Korean Society of Safety
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• v.21 no.4 s.76
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• pp.42-48
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• 2006

The strategies of maintenance and operation are usually established based on the number of flaws and their size distribution obtained from nondestructive inspection in order to preserve safety of the plant. But non destructive inspection results are different from the physical flaws which really exist in the equipments. In case of a single inspection, it is easy to estimate the number of physical flaws using the POD curve. However, we may be faced with some difficulties in obtaining the number of physical flaws from the periodic in-service non destructive inspection data. In this study a simple method for estimating the number of physical flaws from periodic in-service nondestructive inspection data was proposed. In order to obtain the flaw growth history, the flaw growth was simulated using the Monte Carlo method and the flaw size and the corresponding POD value were obtained for each flaw at each periodic inspection time. The flaw growth rate used in the simulation was statistically calculated from the in-service inspection data. By repeating the simulation numerous flaw growth data could be generated and the effective POD curve was obtained as a function of flaw size. From the effective POD curve the number of physical flaws was obtained. The usefulness and convenience of the proposed method was evaluated from several applications and satisfactory results were obtained.

• Corrosion Science and Technology
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• v.3 no.3
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• pp.112-117
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• 2004

This work investigated the material properties of Ni-P-B alloy electrodeposits obtained from a Ni sulfamate bath as a function of the contents of the P and B sources($H_3PO_3$ and dimethyl amine borane complex(DMAB), respectively) with/without additives. Chemical composition, residual stress, microstructure and micro hardness were investigated using ICP(inductively coupled plasma) mass spectrometer, flexible strip, XRD, TEM and micro Vickers hardness tester, respectively. From the results of the compositional analysis, it was observed that P and B are incorporated competitively during the electrodeposition and the sulfur from the additive is codeposited into the electrodeposit. The measured residual stress value increased in the order of Ni, Ni-P, Ni-B and Ni-P-B electrodeposits indicating that boron affects the residual tensile stress greater than phosphorus. As the contents of the alloying element sources of P and B increased, crystallinity and the grain size of the electrodeposit decreased. The effect of boron on crystallinity and grain size was also relatively larger than the phosphorus. It can be explained that the boron with a smaller atomic radius contributes to the increase of residual stress in the tensile direction and the larger restraining force against the grain growth more significantly than the phosphorus with a larger atomic radius. Introduction of an additive into the bath retarded crystallization and grain growth, which may be attributed to the change of the grain growth kinetics induced by the additive adsorbed on the substrate and electrodeposit surfaces during electrodeposition.

• Nuclear Engineering and Technology
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• v.47 no.2
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• pp.204-211
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• 2015

Probabilistic safety assessment (PSA) has had a significant role in quantitative decision-making by finding design and operational vulnerabilities and evaluating cost-benefit in improving such weak points. In particular, it has been widely used as the core methodology for risk-informed applications (RIAs). Even though the nature of PSA seeks realistic results, there are still "conservative" aspects. One of the sources for the conservatism is the assumptions of safety analysis and the estimation of failure frequency. Surveillance, diagnosis, and prognosis (SDP), utilizing massive databases and information technology, is worth highlighting in terms of its capability for alleviating the conservatism in conventional PSA. This article provides enabling techniques to solidify a method to provide time- and condition-dependent risks by integrating a conventional PSA model with condition monitoring and prognostics techniques. We will discuss how to integrate the results with frequency of initiating events (IEs) and probability of basic events (BEs). Two illustrative examples will be introduced: (1) how the failure probability of a passive system can be evaluated under different plant conditions and (2) how the IE frequency for a steam generator tube rupture (SGTR) can be updated in terms of operating time. We expect that the proposed model can take a role of annunciator to show the variation of core damage frequency (CDF) depending on operational conditions.

• Proceedings of the KWS Conference
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• 2010.05a
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• pp.43-46
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• 2010

As the nuclear power plant has been constructed continuously for several decades in Korea, the welding technology for components manufacturing and installation has been improved largely. Standardization for weld test and qualification was also established systematically according to the concerned code. The welding for the main components requires the high reliability to keep the constant quality level, which means the repeatability of weld quality. Therefore the weld process qualified by thorough test and evaluation is able to be applied for manufacturing. Narrow gap SAW and GTAW process are usually applied for girth seam welding of pressure vessel like Reactor vessel, steam generator, and etc. For the surface cladding with stainless steel and Inconel material, strip welding process is mainly used. Inside cladding of nozzles is additionally applied with Hot wire GTAW and semi-auto welding process. Especially the weld joint having elliptical weld line on curved surface needs a specialized weld system which is automatically rotating with adjusting position of the head torch. The small sized pipe, tube, and internal parts of reactor vessel requests precise weld processes like an automatic GTAW and electron beam welding. Welding of dissimilar materials including Inconel690 material has high possibility of weld defects like a lack of fusion, various types of crack. To avoid these kinds of problem, optimum weld parameters and sequence should be set up through the many tests. As the life extension of nuclear power plant is general trend, weld technologies having higher reliability is required gradually. More development of specialized welding systems, weld part analysis and evaluation, and life prediction for main components should be taken into a consideration extensively.

• Corrosion Science and Technology
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• v.11 no.3
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• pp.96-102
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• 2012

Outer diameter stress corrosion cracking (ODSCC) has occurred for Alloy 600 (Ni 75 wt%, Cr 15 wt%, Fe 10 wt%) as a heat exchanger tube of the steam generator (SG) in nuclear power plants (NPP) during long term operation. Among many causes for SCC, lead (Pb) is known to be one of the most deleterious species in the secondary system. In the present work, the oxide formed on Alloy 600 was characterized as a function of the PbO content in 0.1 M NaOH at $315^{\circ}C$ by using an electrochemical impedance spectroscopy (EIS), a transmission electron microscopy (TEM), equipped with an energy dispersive x-ray spectroscopy (EDS). The oxide property was analyzed in view of SCC susceptibility.

• Plant Journal
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• v.9 no.4
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• pp.26-30
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• 2013

Based on case that HRSG low pressure steam generator tube was damaged by FAC in 500 MW A CCPP. This case analyzed the effect of application about the block of hydrazine water treatment which is applied for increasing dissolved oxygen. And also try to deduce the major factor of FAC Which is caused by lacking of dissolved oxygen of boiler feed system. After 1 year of water treatment, the figure of dissolved oxygen in the boiler feed water has increased from 0.15 ppb to 3~5 ppb and the figure of oxidation reduction potential has increased from -245 mV to 170 mV. And Iron content, the corrosion products by FAC has decreased from 18.5 ppb to 5~7 ppb. According to the result of experiment, we could able to confirm that the interception of hydrazine of water treatment is effective to reduce FAC.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.6 no.2
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• pp.20-27
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• 2010

In order to prevent the corrosion of component contacted primary water designed alloy 600 material in the nuclear power plant. But the primary water stress corrosion cracking(PWSCC) of alloy 600 and weld area occurs continuously due to the residual stress. The leakage accident resulted from PWSCC in the drain nozzle of the steam generator of domestic power plants. Heater sleeves of the pressurizer are welded with alloy 600 weld material and therefore exposed to the primary water environment. PWSCC occurred in heater sleeve material and weld area of many foreign power plants. The current issue of domestic nuclear power plants are consequently concentrated to PWSCC of similar material. In order to improve the detection and the sizing of the PWSCC in the welding sleeve of the pressurizer, the automatic UT system and multi-directions probe sets have been developed. The experimental studies have been performed using the mock-up block containing artificial reflectors(ID connected EDM notch) and semi-artificial cracks made from thermal fatigue. The automatic UT System is applied in the detection and the length sizing of the ID/OD on the tube and the J-groove weld area of the artificial reflectors and results of the detection and the sizing are compared respectively. Also, the developed automatic UT system is successfully accomplished to inspect the heater sleeve and the J-groove weld area on the pressurizer for the detection of PWSCC.