• Journal of the Korean Society of Manufacturing Process Engineers
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• v.19 no.1
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• pp.43-48
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• 2020

At present, 24 nuclear power plants are in operation nationwide as the main power source responsible for about 27% of Korea's electricity, and five nuclear power plants are currently under construction. Issues of nuclear safety and reliability have always existed, but after the Fukushima accident, ensuring reliability has become an even more important issue for safety. Compared to other kinds of accidents, the initial response after a nuclear accident is more important than any other accident. Prior to accidents, it is important to be able to predict and judge the accident in advance for the sake of prevention. In this research, non-destructive inspection methods for existing pipe welds include radiographic, ultrasonic, magnetic particle practice, and liquid penetration testing. For this experiment, carbon steel pipes like that of the material used in nuclear pipes were adopted, and specimen welded to the flange (Flange) were manufactured. After testing, the weld specimen were not damaged through the infrared thermography (IRT) experiment. This study attempted to improve the safety of carbon steel pipes through a comparative analysis of finite element analysis.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.404-410
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• 1998

In Yonggwang nuclear units 3 and 4 currently under commercial operation, the cold leg temperature was very close to the technical specification limit of 298$^{\circ}C$ during initial startup testing, which was caused by the higher-than-expected reactor coolant system flow. Accordingly, the reference temperature (Tref) program needed to be revised to allow more flexibility for plant operations. In this study, the method of a specific test performed at Yonggwang nuclear unit 4 to revise the Tref program was described and the test results were discussed. In addition, the modified Tref program was evaluated on its potential impacts on system performance and safety. The methods of changing the Tref program and the associated pressurizer level setpoint program were also explained. Finally, for Ulchin nuclear unit 3 and 4 currently under initial startup testing, the effects of reactor coolant system flow rate on the coolant temperature were evaluated from the thermal hydraulic standpoint and an optimum Tref program was recommended.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.3
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• pp.245-253
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• 2006

A test of the 1/3 scale model was conducted to verify the tip-over analysis of a dry. concrete storage cask under a hypothetical accident condition. The tip-over analysis was executed using the velocity at each point as the initial conditions of the model just before the impact. The initial velocity was determined from the initial angular velocity, which would make the equivalent kinetic energy to the potential energy. To confirm the structural integrity of the canister, the visual testing and the non-detective testings such as Liquid Penetrant testing and Ultrasonic Testing were conducted. The lid of a storage cask was plastically deformed near the impact point. The structural integrity of storage cask was maintained. To verify the tip-over analysis the strains and the accelerations acquired by the tip-over test were compared with those by the analyses. The results of the analysis were larger than the test results about two times.

• Journal of the Korean Society for Nondestructive Testing
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• v.30 no.4
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• pp.317-323
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• 2010

The manual ultrasonic examination for the nuclear power plant piping welds has been demonstrated by using KPD(Korean Performance Demonstration) generic procedure. For automated ultrasonic examination, there is no generic procedure and it should be qualified by using applicable automated equipment. Until now, most of qualified procedures used pulse-echo technique and there is no qualified procedure using phased array technique. In this study, data acquisition and analysis software were developed and phased-array transducer and wedge were designed to implement phased array technique for nuclear power plant in-service inspection. The developed procedure are qualified for performance demonstration for the flaw detection, length sizing and depth sizing. The qualified procedure will be applied for the field examination in the nuclear power plant piping weld inspection.

• Journal of the Korean Society for Nondestructive Testing
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• v.11 no.1
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• pp.52-60
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• 1991

The importance and role of preservice and inservice inspection(PSI/ISI) for nuclear power plant components are intimately related to plant design, safety, reliability and operation etc.. The Korea Atomic Energy Research Institute(KAERI) has been performing PSI/ISI in Korea since the PSI of Kori nuclear power plant, unit 1 had been performed in 1977. KAERI has localized PSI/ISI technology and has done much experience in ultrasonic flaw detection, evaluation and its excavation on the weldments of large pressure vessels. The results of flaw estimation using ultrasonic examination are compared with the actual flaw sizes revealed by field excavation. KAERI's experience regarding PSI/ISI was described and some discussions were added.

• Journal of the Korean Society for Nondestructive Testing
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• v.18 no.4
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• pp.313-320
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• 1998

To record and evaluate the flaws which were found during pre-service/in-service inspection performance of nuclear power plants in Korea, the center line beam method described in ASME code and 6 dB drop method stated in CSA code were used. The measured through wall dimensions and lengths by these methods were compared and analyzed in this report. With the measured data analysis, the ekact understanding and use of these methods improves the reliability of flaw sizing and assures the integrity of nuclear power plant components.

• Proceedings of the KIEE Conference
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• 2005.10c
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• pp.53-55
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• 2005

The signals of the eddy current testing(ECT) for the examination of the steam generator(SG) tubes in the nuclear power plant(NPP) determine the existence, size, and kind of defects using the variation of impedance signals when a testing coil, driven by alternating current, passes through the SG tube contains defects. The aim of this paper is building a database of the RPC probe signals on the basis of the sizes variation of defects and frequency variation of probe. In this paper 3-D numerical analysis of the ECT signals using the finite element method is performed. Through this study, it is shown variation of magnitude and phase of impedance according to variation of defect size and frequency. From the result of this paper, we can obtain the information which is useful in defect discrimination of SG tube in nuclear power plant.

• Journal of the Korean Society for Nondestructive Testing
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• v.12 no.2
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• pp.14-20
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• 1992

Stress analysis has been performed using computer program ANSYS in the pressurizer surge line in accordance with ASME Sec. III in order to predict possibility of fatigue failure due to thermal stratification phenomena in pipes connected to reactor coolant system of nuclear power plants. Highly vulnerable area to crack generation has been chosen by the analysis of fatigue due to thermal stress in pressurizer surge line. This kind of result can be helpful to choose the location requiring intensive care during inservice inspection of nuclear power plants.

• Nuclear Engineering and Technology
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• v.42 no.2
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• pp.203-210
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• 2010

The $\underline{H}igh$ flux $\underline{A}dvanced$ $\underline{N}eutron$ $\underline{A}pplication$ $\underline{R}eact\underline{O}r$ (HANARO), an open-tank-in-pool type reactor, is one of the multi-purpose research reactors in the world. Since the commencement of HANARO's operations in 1995, a significant number of experimental facilities have been developed and installed at HANARO, and continued efforts to develop more facilities are in progress. Owing to the stable operation of the reactor and its frequent utilization, more experimental facilities are being continuously added to satisfy various fields of study and diverse applications. The irradiation testing equipment for nuclear fuels and materials at HANARO can be classified into capsules and the Fuel Test Loop (FTL). Capsules for irradiation tests of nuclear fuels in HANARO have been developed for use under the dry conditions of the coolant and materials at HANARO and are now successfully utilized to perform irradiation tests. The FTL can be used to conduct irradiation testing of a nuclear fuel under the operating conditions of commercial nuclear power plants. During irradiation tests conducted using these capsules in HANARO, instruments such as the thermocouple, Linear Variable Differential Transformer (LVDT), small heater, Fluence Monitor (F/M) and Self-Powered Neutron Detector (SPND) are used to measure various characteristics of the nuclear fuel and irradiated material. This paper describes not only the status of HANARO and the status and perspective of irradiation devices and instrumentation for carrying out nuclear fuel and material tests in HANARO but also some results from instrumentation during irradiation tests.

• Nuclear Engineering and Technology
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• v.55 no.9
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• pp.3260-3267
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• 2023

Zr-Nb alloy have been widely used as fuel rods in nuclear power plants. However, from the Fukushima nuclear accident, the weakness of the rod was revealed under harsh conditions, and research on the safety of these types of rods was conducted after the disaster. The method of depositing chromium onto the existing Zr-Nb alloy fuel rods is being considered as a means by which to compensate for the weakness of Zr-Nb alloy rods because chromium is strong against oxidation at high temperatures and has high strength. In order to secure these advantages, it is important to maintain the Cr thickness of the rods and properly inspect the rods before and during their use in power generation. Eddy current testing is a typical means of evaluating the thickness of thin metals and detecting surface defects. Depending on the size and shape of the inspected object, various eddy current sensors can be applied. In particular, because pancake sensors can be manufactured in very small sizes, they can be used for inspections even in narrow spaces, such as a nuclear fuel assembly. In this study, an eddy current technique was developed to confirm the feasibility of Cr coating thickness evaluations. After determining the design parameters of the pancake sensor by means of a FEM simulation, a FPCB pancake sensor was manufactured and the optimal frequency was selected by measuring minute changes in the Cr-coating thickness using the developed sensor.