• Journal of Korean Society of Industrial and Systems Engineering
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• v.41 no.3
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• pp.186-193
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• 2018

Modern manufacturing industries is to produce both precise and robust mechanical parts without failure while they are in service. In order to prevent a part failure for its lifetime, a mechanical design for a part should be examined on a basis of mechanical simulation. A nozzle plate, being a key part in steam engines, changes flow directions of steam in a turbine used in power plant. This paper is to the design and test for part safety and durability. Currently, nozzle plates are fabricated by welding nozzles to their plates. Welding causes some defects on the used materials while they are being manufactured. Another major defect is un-even pitches between welded nozzles. Welding causes phase changes because of high melting temperature of metal. This leads to decay on the welding spots, which weakens their structural strength and then, may lead to early damages on mechanical structures. This research proposes assembly-typed nozzle plate without welding. From the beginning, nozzle and plate are designed for insertion-typed assembly. Nozzle head and foot are designed in accordance with the grooves on outer ring and inner ring of a plate to make mating surfaces. Then the nozzle plate should be proved for structural and fatigue safety before they are put in manufacturing. This research adopts commercial softwares for modeling and mechanical simulation. The test result shows that the design with smaller mating area and deeper insertion produces higher safety in terms of structure and durability. From the conclusion, this paper proposes the assembly-typed nozzle plate to replace the welding typed.

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.1974-1987
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• 2023

The load-following operation of small modular reactors (SMRs) requires accurate prediction of transient behaviors that can occur in the balance of plants (BOP) and the nuclear steam supply system (NSSS). However, 1-D thermal-hydraulics analysis codes developed for safety and performance analysis have conventionally excluded the BOP from the simulation by assuming ideal boundary conditions for the main steam and feed water (MS/FW) systems, i.e., an open loop. In this study, we introduced a lumped model of BOP fluid system and coupled it with NSSS without any ideal boundary conditions, i.e., in a closed loop. Various methods for coupling boundary conditions at MS/FW were tested to validate their combination in terms of minimizing numerical instability, which mainly arises from the coupled boundaries. The method exhibiting the best performance was selected and applied to a transient simulation of an integrated NSSS and BOP system of a SMART. For a transient event with core power change of 100-20-100%, the simulation exhibited numerical stability throughout the system without any significant perturbation of thermal-hydraulic parameters. Thus, the introduced boundary-condition coupling method and BOP fluid system model can expectedly be employed for the transient simulation and performance analysis of SMRs requiring daily load-following operations.

• Proceedings of the KIPE Conference
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• 2001.10a
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• pp.56-59
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• 2001

An effective method of detecting inter-turn short circuits on round rotor windings is described. Shorted-turns can have significant effects on a generator and its performance. A method of detecting inter-turn short circuits on rotor windings is described. The approach used is to measure the rate of change of the air-gap flux density wave when the rotor is at operating speed and excitation is applied to the field winding. The inter-turn short circuits sensor for synchronous generator's field winding has been developed. The sensor, installed in the generator air-gap, senses the slot leakage flux of field winding and produces a voltage waveform proportional to the rate of change of the flux. For identification of reliability for sensor, a inter-turn short circuits test was performed at the West-Inchon combined cycle power plant on gas turbine generator and steam turbine generator. This sensor will be used as a detecting of shorted-turn for field winding of synchronous generator. The purpose of this paper is to describe the design and operation of a sensitive inter-turn short circuits detector. In this paper, development of inter-turn short circuits sensor for field winding of synchronous generator and application in a field.

• Journal of Power System Engineering
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• v.13 no.5
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• pp.95-102
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• 2009

A steam generator (boiler) in thermal power plants, consisting of more than 30,000 parts and components, can lead to the plant shutdown with damage to even the small part of the components; esp., like weld failures on boiler tubes. Consequently it is greatly demanded to improve the quality of the weld on the boiler tube for the stable operation of the power plants. Because of the feature of the welding, which is done past by melting the work pieces and adding a filler material that cools to become a strong coalescence, there is a great possibility that weld failures take place. As a result, it is regulated to make a non-destructive testing, like radiography test, to detect defects and flaws in the weld. The current film radiography test provides a lower image quality exceeding 2.0% of a basic quality level for a penetrameter, it is very likely to fail to detect micro defect. As a result, the prevention for the boiler tube failure has not been made effectively. In this study, computed radiography technology has been applied as a digital radiography test to the boiler tube weld, and Se-75 radiation source was used to improve the image quality, instead of Ir-192 source. As a result of this study, it is proven to save the time and cost for test and to enhance the quality level of penetrameter penetrating image, which enables to upgrade the quality of radiography test to the boiler tube weld.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.6 no.2
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• pp.20-27
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• 2010

In order to prevent the corrosion of component contacted primary water designed alloy 600 material in the nuclear power plant. But the primary water stress corrosion cracking(PWSCC) of alloy 600 and weld area occurs continuously due to the residual stress. The leakage accident resulted from PWSCC in the drain nozzle of the steam generator of domestic power plants. Heater sleeves of the pressurizer are welded with alloy 600 weld material and therefore exposed to the primary water environment. PWSCC occurred in heater sleeve material and weld area of many foreign power plants. The current issue of domestic nuclear power plants are consequently concentrated to PWSCC of similar material. In order to improve the detection and the sizing of the PWSCC in the welding sleeve of the pressurizer, the automatic UT system and multi-directions probe sets have been developed. The experimental studies have been performed using the mock-up block containing artificial reflectors(ID connected EDM notch) and semi-artificial cracks made from thermal fatigue. The automatic UT System is applied in the detection and the length sizing of the ID/OD on the tube and the J-groove weld area of the artificial reflectors and results of the detection and the sizing are compared respectively. Also, the developed automatic UT system is successfully accomplished to inspect the heater sleeve and the J-groove weld area on the pressurizer for the detection of PWSCC.

• The KSFM Journal of Fluid Machinery
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• v.16 no.4
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• pp.35-43
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• 2013

Most of the nuclear power plants(NPPs) adopts pressure difference type flow meters such as venturi and orifice meters for the measurement of feedwater flow rates to calculate reactor thermal power. However, corrosion products in the feedwater deposits on the flow meter by fouling as operating time goes. These effects lead to severe errors in the flow indication and then determination of reactor thermal power. The averaging BDFT, which has developed by Yun et al., has a potentiality to minimize this problem thanks to its inherent measurement principle. Therefore, it is expected that the averaging BDFT can replace the venturi meter for the feedwater pipe of steam generator of NPPs. The present work compares the amplification factor, K, based on CFD calculation against the K obtained from experiments in order to confirm whether a commercial CFD code can be applicable to the evaluation of characteristic for the averaging BDFT. In addition to this, the simulations to take into account of fouling effect are also carried out by rough wall option. The results show that the averaging BDFT is a promising flow meter for the accurate measurement of flow rates in the fouling condition of the NPPs.

• Journal of the Korean Society for Heat Treatment
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• v.35 no.3
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• pp.121-129
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• 2022

Recently the growth of the renewable energy production has caused the flexible operation in LNG combined cycle power plant. Due to the rapid start and stop operations, large CrMoV castings used for turbine casings and valve bodies could be distorted and lead to replacement or welding repair. This study was performed to find out the characteristics of the repair welding for a damaged CrMoV casting steel. A typical field repair method (arc & TIG welding) was applied to making specimens. The degraded N2 packing head sample from the steam turbine was used. The evaluations of weldments were carried out in terms of microstructural characterization, microhardness measurements, tensile, creep-rupture and fatigue tests. Color etching was also applied for better understanding of welding microstructures. As the boundary between HAZ and base material was deteriorated by welding, it caused microstructural changes formed during PWHT and the shortening of the remaining residual life. By comparing the properties according to repair welding method, it was possible to derive what important welding factors were. As a result, arc welding method is more suitable for repair welding on CrMoV castings.

• Nuclear Engineering and Technology
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• v.25 no.2
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• pp.276-284
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• 1993

The ABB-CE System-80 reactor power cutback system(RPCS) is designed to enable continuous operation of the reactor without trip in the events of the loss of one of the two main feedwater pumps and loss of load, and thus improves plant availability in a cost effective manner. In this study expansion of RPCS has been investigated for continuous reactor operation without trip in the event of an inward control element assembly(CEA) deviation including a single rod drop. Under the expanded function of RPCS the control system will provide a rapid core power reduction on demand by releasing CEAs to drop into the core and reduce the turbine power, if necessary, to follow the reactor power variation. This design feature which is included as the new design features to be incorporated in the ABB-CE System-80+ meets the EPRI advanced light water reactor(ALWR) requirements. For this study core analysis models of System-80+ have been developed to simulate the nuclear steam supply system(NSSS) response as well as the RPCS initiation of rapid CEA insertion. The results of this study demonstrate that the reactor trip can be avoided in the event of inward CEA deviation including a single rod drop by the RPCS initiation and thus the plant availability and capacity factor would be increased.

• Nuclear Engineering and Technology
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• v.18 no.1
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• pp.9-16
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• 1986

Simulation of the system thermal-hydraulic parameters was carried out following the KNUl (Korea Nuclear Unit-1) loss of normal feedwater transient sequence occurred on November 14, 1984. Results were compared with the plant transient data, and good agreements were obtained. Some deviations were found in the parameters such as the steam flowrate and the RCS (Reactor Coolant System) average temperature, around the time of reactor trip. It can be expected since the thermal-hydraulic parameters encounter rapid transitions due to the large reduction of the reactor thermal power in a short period of time and, thereby, the plant data involve transient uncertainties. The analysis was performed using the RELAP5/MOD1/NSC developed through some modifications of the interphase drag and the wall heat transfer modeling routines of the RELAP5/MOD1/CY018.

• KSCE Journal of Civil and Environmental Engineering Research
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• v.32 no.2A
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• pp.123-130
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• 2012

A cooling system is indispensable for the fossil and nuclear power plants which produce electricity by rotating the turbines with hot steam. A cycle of the typical cooling system includes pumping of seawater at the intake pump house, exchange of heat at the condenser, and discharge of hot water to the sea. The cooling type of the nuclear power plants in Korea recently evolves from the conventional surface intake/discharge systems to the submerged intake/discharge systems that minimize effectively an intake temperature rise of the existing plants and that are beneficial to the marine environment by reducing the high temperature region with an intensive dilution due to a high velocity jet and density differential at the mixing zone. It is highly anticipated that the future nuclear power plants in Korea will accommodate the submerged cooling system in credit of supplying the lower temperature water in the summer season. This study investigates the approach flow patterns at the velocity caps and discharge flow patterns from diffusers using the 3-D computational fluid dynamics code of $FLOW-3D^{(R)}$. The approach flow test has been conducted at the velocity caps with and without a cap. The discharge flow from the diffuser was simulated for the single-port diffuser and multi-ports diffuser. The flow characteristics to the velocity cap with a cap demonstrate that fish entrainment can significantly be minimized on account of the low vertical flow component around the cap. The flow pattern around the diffuser is well agreed with the schematic diagram by Jirka and Harleman.