• Transactions of the Korean Society of Mechanical Engineers A
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• v.21 no.11
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• pp.1796-1808
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• 1997

In order to perform leak-before-break design of nuclear piping systems and integrity evaluation of reactor vessels, full stress-strain curves and fracture resistance(J-R) curves are required. However it is time-consuming and expensive to obtain J-R curves experimentally. To resolve these problems, three different methods for predicting J-R curves from tensile data were proposed by the authors previously. The objective of this paper is to develop a computer program based on those J-R curve prediction methods. The program consists of two major parts ; the main program part for the J-R curve prediction and the database part. Several case studies were performed to verify the program, and it was shown that the predicted results were, in general, in good agreement with the experimental ones.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.10 no.1
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• pp.75-81
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• 2014

A simulation of eddy current testing has been utilized for predicting the signal characteristics to the various defects and developing the probes. Especially, CIVA which is a simulation tool dedicated to nondestructive testing has a good accuracy and speed, and provides a three-dimensional graphical user interface for improved visualization and familiar data displays consistent with NDE technique. Although internal validations have been performed by the CIVA software development specialists, an independent validation study is necessary for the accuracy assessment of the software prior to practical use. For this purpose, in this study, eddy current testing signals of ASME FBH calibration standard tube for bobbin probe were simulated using CIVA and the results were compared to the experimental inspected signals based on the relationship between each flaw signal in terms of amplitude and phase, and the shape of the Lissajous curve. And then we verified the accuracy of the simulated signals and the possible range for simulation. Overall, there is a good qualitative agreement between the CIVA simulated and experimental results in the absolute and differential modes at the two inspection frequencies.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.10 no.1
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• pp.113-121
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• 2014

In this work, an integrated model including molecular dynamics and chemical rate theory was implemented to calculate the growth of point defect clusters(PDC) and copper-rich precipitates(CRP) which could change the mechanical properties of reactor pressure vessel(RPV) steels in a nuclear power plant. A number of time-dependent differential equations were established and numerically integrated to estimate the evolution of irradiation defects. The calculation showed that the concentration of the vacancies was higher than that of the self-interstitial atoms. The higher concentration of vacancies induced a formation of the CRPs in the later stage. The size of the CRPs was used to estimate the mechanical property changes in RPV steels, as is the same case with the PDCs. The calculation results were compared with the measured values of yield strength change and Charpy V-notch transition temperature shift, which were obtained from the surveillance test data of Korean light water reactors(LWRs). The estimated values were in fair agreement with the experimental results in spite of the uncertainty of the modeling parameters.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.9 no.1
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• pp.1-7
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• 2013

In around 2000, the U.S. NPPs have developed the various advanced engineering processes based on the INPO AP-913(Equipment Reliability Process Description) and showed the high performance in availability. With these benchmarking cases, the Korean NPPs have introduced the advanced engineering technology since 2005. The first step of the advanced engineering is to analyze and determine component importance for all components of a plant. This process is called Functional Importance Determination(FID). These results are basically utilized to determine the priority with limited resources in various areas. However, because the consistency of FID results is insufficient despite applying the same criteria in the existing operating NPPs, the degree of application is low. Therefore, this paper presents the improved methodology for FID interfacing system functions of Maintenance Rule Program and results of Single Point Vulnerability(SPV). This improved methodology is expected to contribute to enhance the reliability of FID data.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.21 no.8
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• pp.733-740
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• 2011

This paper deals with the process to identify the transient exciting force generated from a rotary compressor. The compressor was assumed to be a rigid body. The equation of motion of a rigid compressor supported by three mounts was derived with 6 degree-of-freedom. The exciting forces at the center of mass of the compressor were estimated from the acceleration data measured at compressor shell. Compressor-pipe system was modeled numerically. The accelerations of compressor and pipe were predicted numerically by using the estimated exciting force. A new shape of pipe model was proposed to reduce the vibration. In the prediction by the method in this paper, the maximum acceleration of the pipe could be reduced by 53.7 % at the steady-state and by 12 % at the transient process. In the real experiments, the maximum acceleration of the pipe was reduced by 54.2 % at steady-state and 14.7 % at the transient process. It was verified that the numerical results showed good agreement with experimental results.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.11 no.2
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• pp.20-24
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• 2015

During refueling Outage, In-service inspections(ISIs) for the Nuclear Power Plant components are mandatory requirement in accordance with ASME Code Sec. XI. Especially, in current ultrasonic testing is one of the most important NDT techniques that are used for volumetric examination methods for nuclear power plant components, and accurate sizing of flaw indication by UT is essential to assure the integrity of the components. However, ASME code specifies minimum requirement for vessel examination procedure, and so far many different flaw sizing approaches have been tried to apply. Through the Round Robin Test(RRT), the accuracy of ultrasonic flaw sizing using DAC techniques was measured with the mock-ups simulating typical pressure vessel welds. These mock-ups contain artificially introduced flaws of known size and location. This paper shows experimental comparison data on the accuracy of techniques using such as 6dB drop, 50%DAC, 20%DAC and 20%DAC with beam spread correction, and also shows that diverse DAC techniques can be effectively applied to the assessment of the flaw sizing for pressure vessel welds in the stage of welding and fabrication.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.11 no.2
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• pp.13-19
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• 2015

Steam generator is one of the most important component of nuclear power plant, and it's integrity and reliability are to be assured to high level by pre-service inspection and in-service inspection. To improve the reliability of steam generator heat exchanger tubes and to secure the management of nuclear power plant safely, KHNP CRI recently has developed eddy current testing system for steam generator. KHNP CRI have performed a series of experimental verification and field application to confirm the performance of the developed ECT system in accordance with ASME Code requirements. The ECT system consists of a remote data acquisition unit, an ECT signal acquisition and analysis software, a water chamber robot controller and a probe push-puller. In this paper, we will details of the developed ECT system and the software and their experimental performance. And also we will report the field applying performance and the issues for further steps.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.4
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• pp.19-25
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• 2011

Steam Generator(SG) tube is an important component of Nuclear Power Plant(NPP), which is the pressure boundary between the primary and secondary systems. The integrity of SG tube has been confirmed by the eddy current test every outage. The eddy current technique adopting bobbin probe is currently the primary technique for the steam generator tubing integrity assesment. The bobbin probe is one of the essential components which consist of the whole ECT examination system and provides us a decisive data for the evaluation of tube integrity. Until now, all of the ECT bobbin probes in Korea which is necessary to carry out inspection are imported from overseas. However, KHNP has recently developed the bobbin probe design technology and transferred it to domestic manufacturers to fabricate the probes. This study has been conducted to establish technical requirements applicable to the steam generator tube inspection using the bobbin probes fabricated by the domestic manufactures. The results have been compared with the results obtained by using foreign probe to identify the availability to the steam generator tube inspection. As a result, it is confirmed that the domestic bobbin probe is generally applicable to SG tube inspection in the NPPs.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.15 no.2
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• pp.31-39
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• 2019

Cracked component analysis is needed for structural integrity analysis under seismic loading. Under large amplitude cyclic loading conditions, the change in material properties can be complex, depending on the magnitude of plastic strain. Therefore the cracked component analysis under cyclic loading should consider appropriate cyclic hardening model. This study introduces a procedure for determining an appropriate cyclic hardening model for cracked component analysis. The test material was nuclear-grade TP316 stainless steel. The material cyclic hardening was simulated using the Chaboche combined hardening model. The kinematic hardening model was determined from standard tensile test to cover the high and wide strain range. The isotropic hardening model was determined by simulating C(T) test under cyclic loading using ABAQUS debonding analysis. The suitability of the material hardening model was verified by comparing load-displacement curves of cyclic C(T) tests under different load ratios.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.15 no.2
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• pp.50-57
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• 2019

This paper investigates deformation characteristics of gas pipeline using the in-plane bending experiment and finite element analysis of a pipe bend. The effect of the bending angle and internal pressure on the deformation characteristics is analyzed. The pipe bend used in this study is API 5L X65 (out diameter: 20 inch) material with the thickness of 11.9 mm. The maximum load, displacement at maximum load, angle and local strain of 90° pipe bend are obtained from the in-plane bending experiment. Comparison between FE results and experimental data shows overall good agreements. In addition, the deformation characteristics of 22.5° and 45° pipe bend are calculated using the finite element analysis. As a result, the effect of the bend angle on the deformation characteristics is discussed.