• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.6 no.1
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• pp.9-15
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• 2010

This paper describes the Visual Tool for automatic ultrasonic examination that is under developing as a part of the project for development of automatic ultrasonic wave acquisition and analysis program. This tool that is supported by various image processing techniques will be adopted to detect the flaws in the component and piping welds in NPP. Visual Tool will enhance the integrity of nuclear power plant. The object of this paper is to address the Visual Tool which is developing for automatic ultrasonic inspection of welds in NPP.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.10 no.1
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• pp.20-24
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• 2014

Recently, it is getting more and more important ensuring the integrity for the equipment degradation according to the increase of nuclear power plant operating period. In many equipment of the nuclear power plant, snubbers mainly installed in reactor coolant pumps, steam generators and piping protected the equipment and piping from the occurrence of transient dynamic loads such as the earthquake, thermal load during the plant operation. This report describes the function, regulation, inspection requirements and management status of the snubbers installed in domestic nuclear power plants.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.11 no.2
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• pp.25-30
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• 2015

In the present study, complex leakage probabilities of nuclear pipes due to fatigue and stress corrosion cracking are evaluated by using the PINTIN(Piping INTegrity INner flaws) that is developed based on the existing PRAISE(Piping Reliability Analysis Including Seismic Events) program. With regard to the aging and crack instability, small leak and big leak probabilities are calculated for several pipes in a reactor coolant system of domestic nuclear plant. Moreover, sensitivity analysis is also performed to find out the effect of parameters for the leakage of pipes, which shows the coolant temperature is the most influencing parameter.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.17 no.1
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• pp.56-63
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• 2021

The feedwater ring is an assembly in steam generator internal piping, which distributes feedwater into the secondary side of the steam generator. It consists of an assembly of carbon steel piping, pipe fittings and J-nozzles which are inserted into the top of the feedwater ring and welded to the diameter of the ring. The feedwater ring at the attachment region of the J-nozzle may be susceptible to flow accelerated corrosion (FAC) due to flow turbulence which increases local fluid velocities. If a J-nozzle becomes a loose part, it can cause damage to tubing near the tube sheet. In this paper, the structural stress analysis for a wall thinned feedwater ring and integrity evaluations under assumed loading conditions are carried out in compliance with ASME B&PV SecIII, NB-3200.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.222-222
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• 1999

• 유체기계공업학회:학술대회논문집
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• 2002.12a
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• pp.419-424
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• 2002

The HANARO, a multi-purpose research reactor of 30 MWth open-tank-in-pool type, has been under normal operation since its initial criticality In February, 1995. Many experiments should be safely performed to activate the utilization of the HANARO. A flow simulation facility is being developed for the endurance test of reactivity control units for extended life times and the verification of structural integrity of those experimental facilities prior to loading in the HANARO. This test facility is composed of three major parts; a half-core structure assembly, flow circulation system and support system. The flow circulation system is composed of a circulation pump, a core flow pipe, a core bypass flow pipe and instruments. The system is to be filled with de-mineralized water and the flow should be met the design flow to simulate similar flow characteristics in the core channel of the half-core test facility to the HANARO. This paper, therefore, describes an analytical analysis to study the flow behavior of the system. The computational flow analysis has been performed for the verification of system pressure variation through the three-dimensional analysis program with standard k-$\epsilon$ turbulence model and for the verification of the structural piping integrity through the finite element method. The results of the analysis are satisfied the design requirements and structural piping integrity of flow circulation system.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.9 no.1
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• pp.56-61
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• 2013

This paper presents the results of integrity assessment for the cracks happened in reactor vessel upper head penetration nozzles. The crack morphology for a boat sample from crack area was analyzed through microscope. The stress condition including weld residual stress around crack was analyzed using finite element analysis. From the results of crack morphology and stress condition, the crack was concluded as primary water stress corrosion cracking. The integrity of the cracked nozzle was assessed by the methodology provided in ASME Section XI. According to the assessment results, the remaining life of the cracked nozzle was 1.43 yrs. and the plant decided to repair it.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.1
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• pp.11-16
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• 2016

Fuel assembly's mechanical characterization test facility (FAMeCT) in KAERI was constructed with upgraded functional features such as increased loading capacity, underwater vibration testing and severe earthquake simulation for extended fuel design guideline. This facility is designed and developed to provide out-pile fuel data for accident analysis model and fuel licensing. Functional tests of FAMeCT were performed to confirm functionality, structural integrity, and validity of newly-built fuel assembly mechanical test facility. Test program includes signal check of data acquisition system, load delivering capacity using real-sized fuel assemblies and a standard loading cylindrical rigid specimen. Fuel assembly's lateral bending test was carried out up to 30 mm of pull-out displacement. Limit case axial compression loading test up to 33 kN was performed to check structural integrity of UCPS (Upper Core Plate Simulator) support frame. Test results show that all test equipment and measurement system have acceptable range of alignment, signal to noise ratio, load carrying capacity limit without loss of integrity. This paper introduces newly constructed fuel assembly's mechanical test facility and summarizes results of functional test for the mechanical test equipment and data acquisition system.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.5 no.2
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• pp.13-19
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• 2009

As the operating time in nuclear power plants (NPPs) increases, the integrity of nuclear components may be continually degraded due to aging effects of systems, structures and components. Recently, a number of NPPs are being operated beyond their design life to produce more electricity without shutting down. The critical issue in extending a lifetime is to maintain the level of safety during the extended operation period while satisfying the international regulatory standards. Therefore, it is beneficial to build a monitoring system to measure an aging status. In this paper, the Aging Monitor (AM) based on lots of aging database obtained from the operating plants and research results on the aging effects was developed to monitor, manage and evaluate the aging phenomena systematically and effectively in NPPs. The AM for the CANDU is divided into 6 modules: (1) Aging Alarm/Coloring Monitor, (2) Aging Database, (3) Aging Document, (4) Real-time Integrity Monitor, (5) Surveillance and Inspection Management System, and (6) Continued Operation and Periodic Safety Review (PSR) Safety Evaluation. The proposed system is expected to provide the integrity assessment for the major mechanical components of an NPP under concurrent working environments.

• 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.