• Transactions of the Korean hydrogen and new energy society
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• v.23 no.1
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• pp.56-64
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• 2012

The ITER fuel cycle is designed for DT operation in equimolar ratio. It involves not only a group of fuelling system and torus cryo-pumping system of the exhaust gases through the divertor from the torus in tokamak plant, but also from the exhaust gas processing of the fusion effluent gas mixture connected to the hydrogen isotope separation in cryogenic distillation to the final safe storage & delivery of the hydrogen isotopes in tritium plant. Tritium plant system supplies deuterium and tritium from external sources and treats all tritiated fluids in ITER operation. Every operation and affairs is focused on the tritium inventory accountancy and the confinement. This paper describes the major fuel cycle processes and interfaces in the tritium plant in aspects of upcoming technologies for future hydrogen and/or hydrogen isotope utilization.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.4
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• pp.12-18
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• 2011

The renewed global interest in nuclear power has arisen from the need to reduce greenhouse gas emissions and to provide sufficient electricity for a growing global population. Many countries with nuclear power plants (NPPs) are still implementing license extensions of 10~20 years, and even consideration is being given to the concept of life-beyond-60, a further period of license extension from 60 to 80 years. To solve the materials aging problem is integral to its success. A foundation for effective aging management of nuclear power plants is that aging is properly taken into account at each stage of a plant's lifetime, i.e. in design, manufacture, construction and operation including long term operation and decommissioning. To evaluate the plant aging phenomena, a lot of background information such as materials and environment of the parts of the reactor and plant systems is needed by the experts. Information on degradation mechanisms is also used. In this paper, a regulation on the integrated materials aging management for nuclear facilities is proposed. The proposed regulation identifies key elements of effective aging management for nuclear power plants and provides the requirements on aging management for nuclear facilities throughout all stages of the lifetime of the plant.

• Corrosion Science and Technology
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• v.3 no.5
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• pp.216-221
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• 2004

The operating experience showed that the fatigue is one of the major piping failure mechanisms in nuclear power plants (NPPs). The pressure and/or temperature loading transients, the vibration, and the mechanical cyclic loading during the plant operation may induce the fatigue failure in the nuclear piping. Recently, many fatigue piping failure occurred at the socket weld area have been widely reported. Many failure cases showed that the gap requirement between the pipe and fitting in the socket weld was not satisfied though the ASME Code Sec. III requires 1/16 inch gap in the socket weld. The ASME Code OM also limits the vibration level of the piping system, but some failure cases showed the limitation was not satisfied during the plant operation. In this paper, the fatigue behavior of the socket weld in the nuclear piping was estimated by using the three dimensional finite element method. The results are as follows. (1) The socket weld is susceptible to the vibration if the vibration levels exceed the requirement in the ASME Code OM. (2) The effect of the pressure or temperature transient load on the socket weld in NPPs is not significant because of the very low frequency of the transient during the plant lifetime operation. (3) 'No gap' is very risky to the socket weld integrity for the specific systems having the vibration condition to exceed the requirement in the ASME OM Code and/or the transient loading condition. (4) The reduction of the weld leg size from $1.09*t_1$ to $0.75*t_1$ can affect severely on the socket weld integrity.

• KEPCO Journal on Electric Power and Energy
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• v.2 no.1
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• pp.109-113
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• 2016

Techno-economic evaluation of Non-Capture $CO_2$ Utilization (NCCU) technology for the production of high-value-added products using greenhouse gas ($CO_2$) was performed. The general scheme of NCCU process is composed of $CO_2$ carbonation and brine electrolysis process. Through a carbonation reaction with sodium hydroxide that is generated from brine electrolysis and $CO_2$ of the flue gas, it is possible to get high-value-added products such as sodium bicarbonate, sodium hydroxide, hydrogen & chloride and also to reduce the $CO_2$ emission simultaneously. For the techno-economic study on NCCU technology, continuous operation of bench-scale facility which could treat $2kgCO_2/day$ was performed. and based on the key performance data evaluated, the economic evaluation analysis targeted on the commercial chemical plant, which could treat 6 tons $CO_2$ per day, was performed using the net present value (NPV) metrics. The results showed that the net profit obtained during the whole plant operation was about 7,890 mKRW (million Korean Won) on NPV metrics and annual $CO_2$ reduction was estimated as about $2,000tCO_2$. Also it was found that the energy consumption of brine electrolysis is one of the key factors which affect the plant operation cost (ex. electricity consumption) and the net profit of the plant. Based on these results, it could be deduced that NCCU technology of this study could be one of the cost-effective $CO_2$ utilization technology options.

• JOURNAL OF ELECTRICAL WORLD
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• no.8 s.116
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• pp.30-36
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• 1986

• Proceedings of the Membrane Society of Korea Conference
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• 2001.07a
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• pp.95-132
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• 2001

• 방재와보험
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• s.76
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• pp.13-23
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• 1998

• JOURNAL OF ELECTRICAL WORLD
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• no.9 s.117
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• pp.31-37
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• 1986

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2018.05a
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• pp.251-252
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• 2018

• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.2859-2865
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• 2021

One of the important issues raised from the Fukushima-Daiichi accident is the safety of multi-unit sites when simultaneous accidents occur at the site and recently a multi-unit PSA methodology is being developed worldwide. Since all operation modes of the plant should be considered in the multi-unit PSA, the accident analysis needs to be performed for shutdown operation modes, too. In this study, a station blackout during the mid-loop operation is selected as a reference scenario. The overall accident progression for the mid-loop operation is slower than that for the full-power operation because the residual heat per mass of coolant is about 6 times lower than that in the mid-loop scenario. Though the fractions of Cs released from the core to the RCS in both operation modes are almost the same, the amount of Cs delivered to the containment atmosphere is quite different due to the chemisorption in the RCS. While 45.5% of the initial inventory is chemisorbed on the RCS surfaces during the full-power operation, only 2.2% during the mid-loop operation. The containment remains intact during the mid-loop operation, though 83.9% of Cs is delivered to the containment.