• Journal of the Korean Society of Safety
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• v.12 no.4
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• pp.201-208
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• 1997

In order to improve the safety of nuclear power plant, we performed a human factors review for the CFMS(Critical Function Monitoring system) design of nuclear power plant. Three works were performed in this study. In first work, we developed human factors engineering program plan(HFEPP) and human factors engineering verification and validation plan (HFE-V & V plan) to effectively perform CFMS design and review. In second work, we identified human engineering discrepancies(HEDs) for CFMS design through human factors design review and proposed those resolutions. In the third work, we developed the evaluation and management methodology for identified KEDs. Methodology developed in this study can be used in other complex system as well as in CFMS design review.

• 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.

• Journal of Energy Engineering
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• v.24 no.2
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• pp.17-21
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• 2015

Restriction orifices in the feed and bleed circuit of nuclear power plant are designed using computer program capable of handling multiple hole cascade orifice assembly. Single hole stages of orifice assembly are alternated with multihole stages where necessary. The distance between stages is such that it allows full pressure recovery. The minimum static pressure is higher than vapor pressure at the operating temperature so that cavitation does not occur. Piping sizes are reviewed and increased if necessary to improve rigidity.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2014.10a
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• pp.839-841
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• 2014

The Nuclear Steam Supply System(NSSS) is designed to provide an integrated approach that includes areas of monitoring relevant to the integrity of the NSSS. LPMS is designed to function as an alarm system by providing sensor channel alarms for the associated subsystems. LPMS is equipped to provide analysis tools for new alarm events, historical events and for historical periodically stored channel data (e.g. waveforms) for most channels. This paper is intended to introduce the diagnosis principle and abnormal symptom of loose parts monitoring system as a monitoring tool in Nuclear Steam Supply System. And also, we are going to introduce signal analysis program in order to perform the actual diagnosis in power plants.

• Nuclear Engineering and Technology
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• v.28 no.6
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• pp.532-541
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• 1996

Theoretical and numerical study on heat transfer and evaporation in the vertical channel has been carried out and basic correlations have been derived for the heat transfer evaluation of PCCS. Analysis program was developed with low-Reynolds-number k-$\varepsilon$ model and surface transfer rates were calculated for the turbulent natural convection in the vertical channel. In relation to dry cooling by buoyancy-driven air, first, the system parameters which govern overall heat transfer rate are determined through the adequate nondimensionalization procedure. After comparison with existing experimental data, numerical results are used to derive heat transfer correlation by sensitivity calculations. In relation to wet cooling by falling water film, numerical analysis are carried out for evaporation process with real film surface conditions and evaporation correlation is derived through analogy concept and correction factors.

• Nuclear Engineering and Technology
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• v.47 no.6
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• pp.719-728
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• 2015

Following the Fukushima accident, a special safety inspection was conducted in Korea. The inspection results show that Korean nuclear power plants have no imminent risk for expected maximum potential earthquake or coastal flooding. However long- and short-term safety improvements do need to be implemented. One of the measures to increase the mitigation capability during a prolonged station blackout (SBO) accident is installing injection flow paths to provide emergency cooling water of external sources using fire engines to the steam generators or reactor cooling systems. This paper illustrates an evaluation of the effectiveness of external cooling water injection strategies using fire trucks during a potential extended SBO accident in a 1,000 MWe pressurized water reactor. With regard to the effectiveness of external cooling water injection strategies using fire engines, the strategies are judged to be very feasible for a long-term SBO, but are not likely to be effective for a short-term SBO.

• Journal of the Korean Society for Nondestructive Testing
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• v.12 no.2
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• pp.14-20
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• 1992

Stress analysis has been performed using computer program ANSYS in the pressurizer surge line in accordance with ASME Sec. III in order to predict possibility of fatigue failure due to thermal stratification phenomena in pipes connected to reactor coolant system of nuclear power plants. Highly vulnerable area to crack generation has been chosen by the analysis of fatigue due to thermal stress in pressurizer surge line. This kind of result can be helpful to choose the location requiring intensive care during inservice inspection of nuclear power plants.

• Nuclear Engineering and Technology
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• v.15 no.2
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• pp.98-109
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• 1983

A two-step alternating direction explicit method is developed to solve the space-dependent reactor kinetics equations in two space dimensions. As a special case in the general class of alternating direction implicit methods, this method is analysed for accuracy and stability. To test the validity of this method it is compared with the implicit-difference method used in the TWIGL program. It is shown that the two methods are closely related. The time dependent neutron fluxes of the pressurized water reactor (PWR), during control rod insertion, and, of the CANDU-PHW reactor, in case of postulated loss of coolant accident, are obtained from the numerical calculation results.

• Nuclear Engineering and Technology
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• v.40 no.4
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• pp.277-284
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• 2008

Recently, the long-term operation of a nuclear power plant beyond its licensed term has become a worldwide trend as long as the safety of the plant is maintained in the extended period. Kori Unit 1, the oldest PWR in Korea, is the foremost example of this type of long-term operation in Korea. Comprehensive technical evaluation of the long-term operation of this plant was completed to confirm the overall safety of the plant. The technical evaluation included a review of PSR results, an assessment on aging management programs and time limited aging analyses, and a statement of radiological impact on the environment. Based on all of the results of the technical evaluation activities, Kori Unit 1 was approved to operate for an additional 10 years beyond its original design life of 30 years.

• Nuclear Engineering and Technology
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• v.45 no.4
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• pp.553-564
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• 2013

This study evaluates the local damage of a turbine in an auxiliary building of a nuclear power plant due to an external impact by using the LS-DYNA finite element program. The wall of the auxiliary building is SC structure and the material of the SC wall plate is high manganese steel, which has superior ductility and energy absorbance compared to the ordinary steel used for other SC wall plates. The effects of the material of the wall, collision speed, and angle on the magnitude of the local damage were evaluated by local collision analysis. The analysis revealed that the SC wall made of manganese steel had significantly less damage than the SC wall made of ordinary steel. In conclusion, an SC wall made of manganese steel can have higher effective resistance than an SC wall made of ordinary steel against the local collision of an airplane engine or against a turbine impact.