• Nuclear Engineering and Technology
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• v.28 no.3
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• pp.339-348
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• 1996

Plant monitoring algorithm developments seem to be saturated which means that display system to show the results from the algorithm should be the well-defined and interactive tools for operator's diagnosing, controlling, restoring the abnormal plant situations. On the other hand, very little generalized display design concepts and evaluations are available. Events that are unfamiliar to operators and that has not been anticipated by designers may cause great threats to the nuclear power plant system safety operation. The abstraction hierarchy, considered most popular display design methodology but not generalized for nuclear power plant design space, has ken proposed as a representation frame work that can be adopted to design interfaces and supports operators in diagnosing overlooked events that should have been considered to operate plant safely. However most practical plant display systems do not fully stick to this design concept but partially rely on their philosophy from design experiences. Abstraction hierarchy display design concept will be do scribed and the trend of Advanced Control Room(ACR) CRT design will also be presented with the conventional display for the several type of plants. Consequently this complementary material should be of interest to designer and regulators concerned with nuclear power plant.

• Proceedings of the KSRS Conference
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• 2002.10a
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• pp.323-329
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• 2002

The thermal effluent discharged from nuclear power plants can affect the offshore ecosystem change. The ability of measuring sea surface temperature in high resolution with Airborne MSS thermal spectral band(8.5 $\mu$m ~ 12.5 $\mu$m) and Landsat ETM+(10.4$\mu$m ~ 12.5 $\mu$m) gives us an information of spread range of thermal effluent. This information can be used as one of major factors fur analyzing the impact of the fish farm damage around the nuclear power plants. Every season from November 1999, this research has been conducted to investigate the extent of diffusion of thermal effluent discharged from KoRi, UlJin and WolSung Nuclear Power Plant located at the coastline of the East Sea of Korea.

• Proceedings of the KSRS Conference
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• 2004.10a
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• pp.331-335
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• 2004

This study is focused to analyze the movement of thermal effluent dischargeed from nuclear power plant by season, ebb and flow, and before and after foundation of tide embankment using thermal infrared band image of 28 scenes observed from Landsat from 1987 to 2004, which is the early stage of operation of young-kwang nuclear power plant. In diffusion of thermal effluent discharge by seasons, spring and summer is spreading further than autumn and winter. It is considered to distribute widely mixed with thermal effluent discharge and hot water, which is distributed naturally along the seaside. It is known the fact that tidal currents control the direction of diffusion of thermal effluent discharge by the change of ebb and flow. Namely, it is distributed widely on the Southwest direction along the seaside by tidal currents when ebb and, it is moved widely on the Northeast direction along the seaside by tidal current when flood. However, in the early stage of flood current, the mainstream of thermal effluent discharge is spread on Southwest direction and, the direction is changed on North­east way when the latter period of flood current. Similarly, in the early stage of ebb current, the mainstream of thermal effluent discharge is spread on Northeast direction and, the direction is changed on Southwest direction when the latter period of ebb current. As the result of comparing to the diffusion pattern of thermal effluent discharge before and after the foundation of seawall, discharged thermal effluent from the drain of plant by the foundation of dike is shown as curved circle pattern on Northeast to West direction from the ending portion of the seawall.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2019.05a
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• pp.261-262
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• 2019

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.551-556
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• 1997

A present study addresses a loss-of-RHR event as an initiating event under specific low power or shutdown conditions. Two typical plant configurations, cold leg opening case with water-filled steam generators and pressurizer opening case with emptied steam generators, were evaluated using the RELAP5/ MOD3.2 code. The calculation was compared with the experiment conducted at ROSA-IV/LSTF in Japan. As a result, the code was capable of simulating the system transient behavior following the event. Especially, thermal hydraulic transport processes including non-condensable gas behavior were reasonably predicted with an appropriate time step and CPU time. However, there were some code deficiencies such as too large system mass errors and severe flow oscillations in core region.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.458-463
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• 1997

A transient natural circulation behavior during a LOCA at hot-standby operation is evaluated for YGN Units 3/4. The plant initial condition is determined within the EOP limitation as suitable to hot-standby mode and the transient scenario is prepared as relevant to evaluation of transient natural circulation. A 0.4% cold leg break with loss of off-site power is calculated with RELAP5/MOD3.2, whose predictability has been verified for SBLOCA natural circulation test, S-NC-8B. Through one hour transient analysis, it is found that the plant has its own decay heat removal capability by natural circulation following a LOCA, at hot-standby mode. Additional calculation is performed to investigate an effect of HPSI flow on natural circulation.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 2006.05a
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• pp.384-389
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• 2006

• 제어로봇시스템학회:학술대회논문집
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• 2002.10a
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• pp.96.5-96
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• 2002

This paper describes extraction procedures for the center coordinates of steam generator tubes of Youngkwang nuclear power plant No. 6 unit. The centering coordinates of tubes are needed for monitoring whether ECT probe is exactly inserted into tube or not. However, The tube image tends to have poor contrast because steam generator bowl is sealed. The centering coordinates extraction procedure consists of two steps. The first step is to process the region with high contrast in entire image of steam generator tubes. Using the center points extracted in the first step and the geometry of tubes lined up in regular triangle patterns the centering coordinates of the rest region with low contrast...

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.1
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• pp.13-22
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• 2022

The decommissioning of a nuclear power plant generates large amounts of radioactive waste, which is of several types. Radioactive concrete powder is classified as low-level waste, which can be disposed of in a landfill. However, its safe disposal in a landfill requires that it be immobilized by solidification using cement. Herein, a safety assessment on the disposal of solidified radioactive concrete powder waste in a conceptual landfill site is performed using RESRAD. Furthermore, sensitivity analyses of certain selected input parameters are conducted to investigate their impact on exposure doses. The exposure doses are estimated, and the relative impact of each pathway on them during the disposal of this waste is assessed. The results of this study can be used to obtain information for designing a landfill site for the safe disposal of low-level radioactive waste generated from the decommissioning of a nuclear power plant.

• Nuclear Engineering and Technology
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• v.33 no.1
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• pp.12-24
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• 2001

The post-LOCA long term cooling (LTC) performance of the Korean Standard Nuclear Power Plant (KSNPP) is analyzed for both small break loss-of-coolant accidents (LOCA) and large break LOCA at cold leg. The RELAP5/MOD3.2.2 beta code is used to calculate the LTC sequences based on the LTC plan of the Korean Standard Nuclear Power Plants (KSNPP). A standard input model is developed such that LOCA and the followed LTC sequence can be calculated in a single run for both small break LOCA and large break LOCA. A spectrum of small break LOCA ranging from \ulcorner.02 to 0.5 k2 of break area and a double-ended guillotine break are analyzed. Through the code calculations, the thermal-hydraulic behavior and the boron behavior are evaluated and the effect of the important action including the safety injection tank (SIT isolation and the simultaneous injection in LTC procedure is investigated. As a result, it is found that the sufficient margin is available in avoiding the boron precipitation in the core. It is also found that a further specific condition for the SIT isolation action need to be setup and it is recommended that the early initiation of the simultaneous injection be taken for larger break LTC sequences.