• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.8 no.2
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• pp.33-41
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• 2012

The component cooling water heat exchangers are critical components in a nuclear power plant. As the operation years of the heat exchanger go by, the maintenance costs required for continuous operation also increase. Most heat exchangers have carbon steel shells, tube support plates and flow baffles. The titanium tube is susceptible to flow induced vibration. The damage on carbon steel tube support rod and titanium tube around cooling water entrance area is inevitable. Therefore, analysis of internal flow around the component cooling water entrance and tube channel is a good opportunity to seek for failure prevention practice and maintenance method. The numerical study was carried out by FLUENT code to find out the causes of tube failure and its location.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.8 no.2
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• pp.21-25
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• 2012

Functional Equipment Group(FEG) of Nuclear Power Plant(NPP) is the bundling of multiple equipment subcomponents around a major equipment with a common system isolation characteristic or out-of-service consideration. The main purpose of FEG is to reduce out-of-service time to enhance maintenance effectiveness. KHNP(Korea hydro & Nuclear Power Co.) needs to develop these FEGs to successfully perform On-line Maintenance(OLM) which will be expected to institute in the future. Recently, there has been growing interest in OLM in KHNP. The aim of this paper is to suggest the way of FEG development and to build the FEG of MCR(Main Control Room) HVAC(Heat, Ventilation and Air Conditioning) system. The results of this study might be make good use of OLM and be helpful for equipment maintenance for MCR HVAC system.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.8 no.2
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• pp.1-6
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• 2012

A PHE (Process Heat Exchanger) in a nuclear hydrogen system is a key component required to transfer heat energy of $950^{\circ}C$ generated in a VHTR (Very High Temperature gas cooled Reactor) to the chemical reaction that yields a large quantity of hydrogen. A small-scale PHE prototype made of Hastelloy-X is being tested in a small-scale gas loop at Korea Atomic Energy Research Institute. Previous research on the high-temperature structural analysis of the small-scale PHE prototype had been performed only using parent material properties. In this study, high-temperature structural analysis using weld properties in weld zone was performed and the analysis results compared with the previous research.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.1
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• pp.41-47
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• 2011

Residual stress is the key parameter for reliability and lifetime assessment because it can reduce the fatigue strength and fracture properties of industrial structures. Recently, instrumented indentation testing (IIT) has been widely used for evaluating it, since it does not need specific specimen and time-consuming procedure. However, conventional Oliver-Pharr method, which is used for calibrating contact depth to analyze indentation load-depth curve, cannot estimate plastic pile-up between indenter and surface of specimen. Here, we introduce f parameter which is the ratio of contact depth and maximum depth, to consider pile-up height. And, its application for evaluating residual stress of weldment is introduced.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.15 no.2
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• pp.9-18
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• 2019

The objective of this study is to investigate the feasibility of simulating the mechanical properties of irradiatied austenitic stainless steels by cold-working. In this study, the tensile properties, cyclic hardening behaviors and fracture toughness of cold-worked TP316L stainless steel were compared with those of austenitic stainless steels irradiated by neutrons. It showed that cold-working can properly simulate the increase in strength and the decrease in ductility and fracture resistance of austenitic stainless steels by neutron irradiation, even though it could not perfectly simulate the microstructures of irradiated austenitic stainless steels. Also, cold-working can appropriately simulate the hardening behaviors of neutron irradiated austenitic stainless steels under monotonic and cyclic loading conditions.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.15 no.2
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• pp.40-49
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• 2019

In this paper, failure probabilities of the OPR1000 reactor vessel under pressurized thermal shock (PTS) were estimated using the probabilistic fracture mechanics code, R-PIE. Input variables of initial crack distribution, crack size, copper contents, and upper shelf toughness were selected for the sensitivity analyses. A wide range of the input data were considered. Through-wall cracking frequencies determined by the product of the vessel failure probability and the corresponding occurrence frequency of the transient were also compared to the acceptance criterion. The results showed that transient history had the most significant impact on the vessel failure probability. Moreover, conservative assumptions resulted in extremely high through-wall cracking frequencies.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.4 no.2
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• pp.29-33
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• 2008

As the measures of nuclear power plant utilities and manufacturers to reduce the defects of tube expansion region during manufacturing steam generators, many types of NDEs(Non-Destructive Examinations) are conducted to inspect the expansion region. The expansion region of tube is subject to degrade because of stress concentration induced by tube expansion, sludge pile and high temperature. So the inspections for tube expansion region have been reinforced. Liquid penetrant test, helium leak test, Bobbin profile test and hydraulic test are performed to confirm the integrity of tube expanded by hydraulic expansion method. Liquid penetrant test and helium leak test are used to inspect seal weld region on tubesheet end part. Bobbin Profile test is used to inspect fully the expanded region of steam generator tube. Hydraulic test finally verifies the integrity of seal weld region on tubesheet end part.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.4 no.2
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• pp.46-49
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• 2008

NDE(Nondestructive examination) detects a flaw or discontinuity in materials. Flaws detected by the pre-service or in-service examinations shall be sized for the purpose of analysis and repair. A flaw that is initiated from the surface is difficult to determine its depth by NDE. The depth of the surface flaw can be measured using an ultrasonic diffracted wave. To find the optimum standard for ultrasonic parameter(For example, frequency & size of transducer), a mock-up test and simulation were established and studied. This inspection technology may show the depth sizing possibility of the flaw down to nearly two(2) mm.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.4 no.2
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• pp.57-61
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• 2008

During the period of retubing work for the licensing renewal, the fuel channels, calandria tubes and feeders of CANDU Reactors will be replaced, and calandria visual examination will be performed. This period is a unique opportunity to inspect the inside of the calandria. The visual inspection for the calandria vessel and its internals of Wolsong NPP is scheduled for confirming the calandria integrity. The first visual inspection for the calandria is planned in Pt. Lepreau led by AECL. The visual inspection for Wolsong NPP, led by NETEC(Nuclear Engineering & Technology Institute) of KHNP, will employ 3D laser scanner and 3D CAD Mock-up for the first time in the world, in addition to a conventional video camera. The inspection system is composed of a robot with the 3D laser scanner, a video camera and a hardness meter.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.13 no.2
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• pp.19-30
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• 2017

Austenitic Stainless Steels (ASSs) are widely used as structural materials in the pressurized water reactors (PWRs) because of their superior mechanical properties and corrosion resistance. However, it is well known that ASSs are susceptible to the environmental assisted cracking (EAC) such as environmental assisted fatigue (EAF) during the long term operation. There have been extensive tests and researches to understand the extent and the mechanisms of environmental effects. In this paper, the world-wide EAF test results of ASSs are introduced including those of Korean test programs. The suggested EAF mechanisms of ASSs are also discussed. Finally, the areas of further research to resolve the issue of EAF are suggested.