• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.3
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• pp.1-7
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• 2011

The analysis shows that the vibration is one of the main reasons of turbine failure. Especially, the problems caused by vibration occur right after retrofit of the turbine-generator and restarting the turbine. Through the case study of high vibration caused by after the turbine trip and restart, turbine vibration was identified to be influenced by startup condition. Turbine startup at high casing temperature right after unscheduled turbine trip cause radial expansion in rotor by contraction in axial direction, while casing continues to contract by steam flowing into casing. Consequently, gap between rotor and casing decrease until to metal contact to cause high vibration. Through the case study of high vibration of turbine-generator system after generator retrofit, it was identified that generator replacement could cause high vibration in turbine-generator system if the influence of generator replacement on entire system was not considered properly. To prevent startup delay caused by high vibration, it is important to keep the gaps at the design standard and start the turbine after thermal equilibrium.

• KEPCO Journal on Electric Power and Energy
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• v.5 no.4
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• pp.323-330
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• 2019

The 500 MW Korean standard coal-fired power plant is the largest standardized power plant in Korea and has played a pivotal role in domestic power generation for over 20 years. In addition to the aging degradation due to long term operation, the probability of failure of power generation facilities is increasing due to frequent startup and stop caused by the lower utilization rate due to air pollution problem caused by coal-fired power plants. Among them, steam piping plays an important role in transferring high-temperature & pressure steam produced in a boiler to turbine for power generation. In recent years, failure of steam piping of large coal-fired power plant has frequently occurred. Therefore, in this study, failure analysis of high pressure piping weld was conducted. We identify the damage caused by high stress due to abnormal supporting structure of the piping and suggest improved supporting structure to eliminate high stress through microstructure analysis and piping stress analysis to prevent the occurrence of the similar failure of other power plant in the case of repetitive damage to the main steam piping system of the 500 MW Korean standard coal-fired power plant.

• Nuclear Engineering and Technology
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• v.49 no.7
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• pp.1463-1471
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• 2017

In order to monitor the pipe thinning caused by flow-accelerated corrosion (FAC) that occurs in coolant piping systems, a shear horizontal ultrasonic pitch-catch waveguide technique was developed for accurate pipe wall thickness monitoring. A clamping device for dry coupling contact between the end of the waveguide and pipe surface was designed and fabricated. A computer program for multi-channel on-line monitoring of the pipe thickness at high temperature was also developed. Both a four-channel buffer rod pulse-echo type and a shear horizontal ultrasonic waveguide type for high-temperature thickness monitoring system were successfully installed to the test section of the FAC proof test facility. The overall measurement error can be estimated as ${\pm}10{\mu}m$ during a cycle from room temperature to $200^{\circ}C$.

• Nuclear Engineering and Technology
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• v.41 no.8
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• pp.1115-1126
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• 2009

As one of the ways to enhance the economical features in sodium-cooled fast reactor development, the concept of an integrated steam generator and pump system (ISGPS) is proposed from a structural point of view. And the related intermediate heat transfer system (IHTS) piping layout compatible with the ISGPS is described in detail. To assure the creep design lifetime of 60 years, the structural integrity is investigated through high temperature structural evaluation procedures by the SIE ASME-NH computer code, which implements the ASME-NH design rules. From the results of this study, it is found that the proposed ISGPS concept is feasible and applicable to a commercial SFR design.

• Journal of Ocean Engineering and Technology
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• v.21 no.5
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• pp.55-60
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• 2007

Bellows are mechanical components which prevent the damage of system by absorption of the vibration and the displacement of axle and radial direction. Thermal piping system is expanded by the fluid of the high temperature from the heat engine inside. At this time, bellows prevent the damage of the piping due to the thermal expansion. Recently, design of bellows is required to fit some other operational environments which are not suggested in the E.J.M.A code book. And it is difficult to produce and to maintain bellows of high temperature and high pressure bemuse of its complicated shape and this causes the manufacturing cost to rise. The objective of this study is to determine optimum shape of bellows which can endure in the high temperature and high pressure. The maximum stress has an effect on the fatigue life of bellows, therefore it needs to be minimized. This study attempts to find a shape which minimizes the stress occurring in the bellows by the design of experiment. The model used in this study is not presented in the E.J.M.A code book, therefore, from the result of design of experiment we find the factors which give effects on the characteristic value and we presents the recession model using the RSM, which can predict the characteristic values depending on the change of factor values.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.19 no.1
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• pp.27-35
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• 2023

Pressure tubes are the main components of PHWR core and serve as the pressure boundary of the primary heat transport system. However, because pressure tubes have changed their geometrical dimensions under the severe operating conditions of high temperature, high pressure and neutron irradiation according to the increase of operation time, all dimensional changes should be predicted to ensure that dimensions remain within the allowable design ranges during the operation. Among the deformations, the diameter expansion due to creep leads to the increase of bypass flow which may not contribute to the fuel cooling, the decrease of critical channel power and finally the deration of the power to maintain the operational safety margin. This study is focused on the modeling of the expansion of the pressure tube diameter based on the operating conditions and measured diameter data. The pressure tube diameter expansion was modeled using the neutron flux and temperature distributions of each fuel channel and each fuel bundle as well as the measured diameter data. Although the basic concept of the current modeling approach is simple, the diameter prediction results using the developed methodology showed very good agreement with the real data, compared to the existing methodology.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.33 no.2
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• pp.179-184
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• 2009

The Fuel Test Loop(FTL) which is capable of an irradiation testing under a similar operating condition to those of PWR(Pressurized Water Reactor) and CANDU(CANadian Deuterium Uranium reactor) nuclear power plants has been developed and installed in HANARO, KAERI(Korea Atomic Energy Research Institute). It consists of In-Pile Section(IPS) and Out-of Pile System(OPS). The IPS, which is located inside the pool is divided into 3-parts; the in-pool pipes, the IVA(IPS Vessel Assembly) and the support structures. The test fuel is loaded inside a double wall, inner pressure vessel and outer pressure vessel, to keep the functionality of the reactor coolant pressure boundary. The IVA is manufactured by local company and the functional test and verification were done through pressure drop, vibration, hydraulic and leakage tests. The brazing technique for the instrument lines has been checked for its functionality and performance. An IVA has been manufactured by local technique and have finally tested under high temperature and high pressure. The IVA and piping did not experience leakage, as we have checked the piping, flanges, assembly parts. We have obtained good data during the three cycle test which includes a pressure test, pressure and temperature cycling, and constant temperature.

• Nuclear Engineering and Technology
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• v.41 no.10
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• pp.1323-1332
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• 2009

The IHX (Intermediate Heat eXchanger) for a pool-type SFR (Sodium-cooled Fast Reactor) system transfers heat from the primary high temperature sodium to the intermediate cold temperature sodium. The upper structure of the IHX is a coaxial structure designed to form a flow path for both the secondary high temperature and low temperature sodium. The coaxial structure of the IHX consists of a central downcomer and riser for the incoming and outgoing intermediate sodium, respectively. The IHX of a pool-type SFR is supported at the upper surface of the reactor head with an IHX support structure that connects the IHX riser cylinder to the reactor head. The reactor head is generally maintained at the low temperature regime, but the riser cylinder is exposed in the elevated temperature region. The resultant complicated temperature distribution of the co-axial structure including the IHX support structure may induce a severe thermal stress distribution. In this study, the structural feasibility of the current upper support structure concept is investigated through a preliminary stress analysis and an alternative design concept to accommodate the IHTS (Intermediate Heat Transport System) piping expansion loads and severe thermal stress is proposed. Through the structural analysis it is found that the alternative design concept is effective in reducing the thermal stress and acquiring structural integrity.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2000.06a
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• pp.1192-1197
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• 2000

Piping systems of the power generating stations have been generally protected by hydraulic and mechanical snubbers which can allow large displacements arising from temperature change while those can reduce or absorb stresses due to vibrations. However these snubbers require amounts of budget for maintenance or replacement because of the leakage, lubrication and finally short life cycle. Recently the snubbers consisted of wire rope have been proved to reduce vibrations of piping systems. While the wire rope snubbers are free of maintenance such as leakage and lubrication, imported price are so high. Now it is necessary to design, manufacture and certificate these wire rope snubbers.

• Nuclear Engineering and Technology
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• v.30 no.5
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• pp.435-443
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• 1998

Leak-before-break(LBB) approach has been shown to be both cost effective and risk reductive when applied to high energy Piping in nuclear Power Plants. For the Korean Next Generation Reactor (KNGR) development, LBB application is considered for the Main Steam Line(MSL) piping inside containment. Unlike the primary system leakages, the MSL leak detection systems must be based on principles other than radioactivity measurements. Among humidity, heat and acoustic noise currently being considered as indicators of leakage, we explored humidity as an effective one and developed ceramic-based humidity sensor which can be qualified for LBB applications. The ceramic material, sintered and annealed MgCr$_2$O$_4$-TiO$_2$, is shown to increase its electrical conductivity drastically upon water vapor adsorption over the entire temperature range of interest. With this ceramic sensor specimen, we suggested installation-inside-the-piping method by which we can detect leakage more rapidly and sensitively. In this paper, we describe the progress in the development and characterization of ceramic humidity sensor for the LBB application to the MSL of KNGR.