• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.622-627
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• 1996

Detailed analysis of Feedwater Line Break (FLB) event for the fuel failure point of view are lack because the event was characterized as the increase in reactor coolant system (RCS) pressure. Up to now, the potential of the rapid system heatup case has been emphasized and comprehensively studied. The cooldown effects of FLB event is considered to be bounded by the Steam Line Break (SLB) event since the cooldown effect of SLB event is larger than that of the FLB event. This analysis provides a new possible path which can cause the fuel failure. The new path means that the fuel failure can occur under the heatup scenario because the Pressurizer Safety Valves (PSVs) open before the reactor trips. The 1000 MWe typical C-E plant FLB event assuming Loss of Offsite Power (LOOP) at the turbine trip has been analyzed as an example and the results show less than 1% of the fuel failure. The result is well within the acceptance criteria. In addition to that, a study was accomplished to prevent the fuel failure for the heatup scenario case as an example. It is found that giving the proper pressure gap between High Pressurizer Pressure Trip (HPPT) analysis setpoint and the minimum PSV opening pressure could prevent the fuel failure.

• Proceedings of the Korean Society for Technology of Plasticity Conference
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• 1999.08a
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• pp.390-398
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• 1999

The thermomechanical control process(in hot rolling strip mill) was employed to produce 2.25Cr-1Mo steel, which is to be construction material for the steam generator for power plant. Although the Conventional processes has been the primary means of producing the 2.25Cr-1Mo steel, an alternative method was used to meet the specification of ASTM heat treatment for A387-22-Classl using autotempering after coiling in hot rolling strip mill. The microstructures, tensile properties at various temperatures, and creep-rupture properties have been investigated to compare the properties with those of materials produced by the conventional process and to certify the application of the thermomechanical control process to an actual process of manufacturing 2.25-Cr-1Mo steel, this in turn, will reduce the cost of the process. About 14 to 34% glanular bainite (remainder proetectoid ferrite) formed in a coil, and this variety of volume fraction stems from the different cooling rates, which varies with position of the coil after coiling. Tensile testing from room temperature to 700$^{\circ}C$ indicated that strength increases with test temperature showing peaks at around 600$^{\circ}C$. Creep-rupture properties have been being investigated at the temperature of 500$^{\circ}C$ with 27.5, 32kg/$\textrm{mm}^2$ loads and have showed no rupture for over 1000 hours.

• Proceedings of the KSME Conference
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• 2000.04a
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• pp.7-12
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• 2000

The automatic turbine startup system provides turbine control based on thermal stress. During the startup, control system monitors and evaluates main components of turbine using damage mechanism and life assessment. In case of valve chest, the temperature of inner/outer wall is measured by thermo-couples and the safety of these values are evaluated by using allowable ${\Delta}T$ limit curve during the startup. Because allowable ${\Delta}T$ limit curve includes life assessment, it is possible to apply this curve to turbine control system. In this paper, low cycle fatigue damage and combined rupture and low cycle fatigue damage criterion proposed for yielding the allowable ${\Delta}T$ limit curve of CV(control valve) chest. To calculate low cycle fatigue damage, the stress analysis of valve chest has peformed using FEM. Automatic turbine startup to assure service life of CV was achieved using allowable ${\Delta}T$ limit curve.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.26 no.1
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• pp.1-6
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• 2002

The automatic turbine startup system provides turbine control based on thermal stress. During the startup, control system monitors and evaluates main components of turbine using damage mechanism and life assessment. In case of valve chest, the temperature of inner/outer wall is measured by thermo-couples and the safety of these values are evaluated by using allowable △T limit currie during the startup. Because allowable ΔT limit curve includes life assessment, it is possible to apply this curve to turbine control system. In this paper, low cycle fatigue damage, combined rupture and low cycle fatigue damage criterion were proposed for yielding the allowable ΔTf limit curve of CV(control valve) chest. To calculate low cycle fatigue damage, the stress analysis of valve chest has been performed using FEM. Automatic turbine startup to assure service life of CV was achieved using allowable ΔT limit curve.

• Nuclear Engineering and Technology
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• v.29 no.3
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• pp.229-239
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• 1997

To investigate the realistic behavior of mass and energy release and resultant containment response during large break Loss of Coolant accident (LOCA), analyses are performed for Yonggwang (YGN) 3&4 nuclear power plants by using a merged version of RELAP5/CONTEMPT4 computer code. Comparative analyses by using conservative design computer codes are also peformed. The break types analyzed are the double-ended guillotine breaks at the cold leg and hot leg. The design analysis resulted in containment peak pressure during post-blowdown phase for the cold leg break. However, the RELAP5/CONTEMPT4 analyses show that the containment pressure has a peak during blowdown phase, thereafter it decreases monotonously without the second port-blowdown peak. For the hot leg break, revised design analysis shows much lower pressure than that reported in YGN 3&4 final safety analysis report. The RELAP5/CONTEMPT4 analysis shoos similar trend and confirmed that the bypass flow through the broken loop steam generator during post-blowdown is negligibly small compared to that of cold leg break. The low pressure and temperature predicted tv realistic analysis presented in this paper suggest that the design analysis methodology contains substantial margin and it can be improved to provide benefit in investment protection, such as, relaxing plant technical specifications and reducing containment design pressure.

• Journal of Welding and Joining
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• v.17 no.1
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• pp.82-87
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• 1999

The basic remote sleeve repair-welding technology by the pulsed Nd:YAG laser for increasing the lifetime of the steam generator tube in a nuclear power plant has been developed. The relationship between the connection width and welding parameters have been investigated for the fundamental research to apply the sleeve-repair-welding technique to the nuclear industry. The Inconel 600 tube and Inconel 690 sleeve used for high temperature and high pressure service were welded as round lap welding by Nd:YAG laser. It was observed that the tensile shear strength, 340MPa of the welded specimen is equivalent to about 60% of that of the base metal (Inconel 600), 550MPa. The difference between the hardness of the base metal and that of the laser welds was about 10%. Ductile fracture was partly occurred in the weld but the cracking has not been observed. In spite of absence of the crack, the strength of welds was not sufficient in terms of the tensile shear strength.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.469-474
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• 1997

To provide tile information ell severe accident progression is very important for advanced or new type of nuclear power plant (NPP) design. A parametric study, therefore was performed to investigate the effect of thermal hydraulic design parameters ell severe accident progression of pressurized water reactors (PWRs), Nine parameters, which are considered important in NPP design or severe accident progression, were selected among the various thermal hydraulic design parameters. The backpropagation neural network (BPN) was used to determine parameters, which might more strongly affect the severe accident progression, among mile parameters. For training. different input patterns were generated by the latin hypercube sampling (LHS) technique and then different target patterns that contain core uncovery time and vessel failure time were obtained for Young Gwang Nuclear (YGN) Units 3&4 using modular accident analysis program (MAAP) 3.0B code. Three different severe accident scenarios, such as two loss of coolant accidents (LOCAs) and station blackout(SBO), were considered in this analysis. Results indicated that design parameters related to refueling water storage tank (RWST), accumulator and steam generator (S/G) have more dominant effects on the progression of severe accidents investigated, compared to tile other six parameters.

• Proceedings of the KSME Conference
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• 2004.04a
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• pp.2053-2058
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• 2004

The pressurized light water cooled, medium power (330 MWt) SMART (System-integrated Modular Advanced ReacTor) has been under development at KAERI for a dual purpose : seawater desalination and electricity generation. The SMART design verification phase was followed to conduct various separate effects tests and comprehensive integral effect tests. The high temperature / high pressure thermal-hydraulic test facility, VISTA(Experimental Verification by Integral Simulation of Transient and Accidents) has been constructed to simulate the SMART-P (the one fifth scaled pilot plant) by KAERI. Experimental tests have been performed to investigate the thermal-hydraulic dynamic characteristics of the primary and the secondary systems. Heat transfer characteristics and natural circulation performance of the PRHRS (Passive Residual Heat Removal System) of SMART-P were also investigated using the VISTA facility. The coolant flows steadily in the natural circulation loop which is composed of the steam generator (SG) primary side, the secondary system, and the PRHRS. The heat transfers through the PRHRS heat exchanger and ECT are sufficient enough to enable the natural circulation of the coolant.

• Corrosion Science and Technology
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• v.5 no.3
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• pp.85-89
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• 2006

KAERI(Korea Atomic Energy Research Institute) has developed a repassivation rate test system which can be operated at $300^{\circ}C$. It consists of an autoclave, three electrodes for an electrochemical test and a diamond scratch tip. All the electrodes are electrically insulated from the autoclave by using high temperature fittings. Reproducible repassivation curves of alloy 600 at 300 C were obtained. Repassivation rate of alloy 600 at pH 13 was slower than that of pH 10. Stress corrosion cracking test was carried as a function of the pH at a high temperature. At pH 10, alloy 600 showed a severe stress corrosion cracking(SCC), whereas it did not show a SCC at pH 7. From the viewpoint of a relationship between the current density and the charge density, a big difference was observed in the two solutions; the slope of pH 13 was steeper than that of pH 10. So the stress corrosion susceptibility at pH 13 seems to be higher than that of pH 10. The system would be a good tool to evaluate the SCC susceptibility of alloy 600 at a high temperature.

• Journal of the Korean Professional Engineers Association
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• v.44 no.6
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• pp.40-44
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• 2011

Nuclear energy in Korea began in 1958, when the Korea's atomic energy act was formulated and the relevant organizations were founded. Since then, notwithstanding the two catastrophe like TMI and Chernobyl accident, Korea made a wise decision to expand the peaceful uses of the nuclear energy as well as to localize the essential nuclear design technology of fuel and nuclear steam supply system. This decision resulted in the success of export of nuclear power plants as well as research reactor in 2010s. The Korea's nuclear policy, which well utilized 'international crisis in nuclear business' as 'opportunity of Korea to get. nuclear technology', is believed nice policy as a role model of nuclear new-comer countries. Based upon the success story of localization of nuclear technology, Korea had an eye for a niche market, which was a basis of development of SMART, Korea-made integral PWR. The operation of a SMART plant can sufficiently provide not only electricity but also fresh water for 100,000 residents. Last two years, Korea's nuclear industry team led by the Korea Atomic Energy Research Institute completed the standard design of SMART and applied to the Korea's regulatory body for standard design approval. Now the Korea's licensing authority is reviewing the design with the relevant documents, and the design team is doing its best to realize its hope to get the approval by the end of this year. From next year, the SMART business including construction and export will be explored by the KEPCO consortium.