• Proceedings of the KIEE Conference
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• 2000.07c
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• pp.1859-1861
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• 2000

The high temperature thermal plasma technology applied to waste treatment has undoubtedly gained high importance owing to its outstanding properties such as flexibility, compact reactor. and clean treatment as the environmental problem goes to a main issue in public talks, because the thermal plasma with temperature of around 10,000K or little less is particularly suitable for waste treatment. Since the thermal plasma is, in general, governed by a number of parameters, some complicated and elaborate controls might be mandatory. The high maintenance cost caused by big input power has been a main obstacle to the growth of the waste treatment plant based on thermal plasma technology, but the recent R&D on the waste-to-energy shows that the problem could be solved soon. In this paper, the authors introduce the current R&D activity related to three-phase ac plasma torch in KERI.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.33 no.7
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• pp.664-672
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• 2009

Nozzle in nuclear power plant is connected to pipe using safe end. Dissimilar metal weld between nozzle and safe end is followed by similar metal weld between safe end and pipe. And thus residual stress in dissimilar metal weld can be affected by similar metal weld. Similar metal weld impose bending stress on dissimilar metal weld, which is according to the length of safe end. In this study, simple nozzle model which covers various radius to thickness ratios was proposed to quantify residual stress in dissimilar metal weld based on finite element analyses. As a result, short length of safe end was proved to be more effective to mitigate residual stress in dissimilar metal weld and critical effective length of safe end is provided according to the radius to thickness ratio.

• Korean Journal of Environment and Ecology
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• v.7 no.2
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• pp.172-180
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• 1994

The forest of Choksangsansong area was studied to investigate vegetational structure with twenty-five plots(20m$\times$25m). Relative importance values, DBH class distribution, species, diversity indices, DCA ordination and TWINSPAN classification were used for vegetational structure analysis. Quercus mongolica, Carpinus cordata, Quercus serrata and Cornus controversa were appeared to be dominant species in each plot and dominant species in the Choksangsansong area was Quercus mongolica. The forest of Choksangsansong area was classified into four groups and showed seral stage from Quercus mongolica to Carpinus cordata. Vegetational succession in the Choksangsansong area shall be accelerated to moist and shade tolerant species due to cool and humid climatic condition by upper dam construction of the Mujuyangspalchonso (pumping up power station).

• Journal of the Korean Society of Safety
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• v.28 no.3
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• pp.123-128
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• 2013

The Advanced Power Reactor Plus(APR+), which is a GEN III+ reactor based on the APR1400, is being developed in Korea. In order to enhance the safety of the APR+, a passive auxiliary feedwater system(PAFS) has been adopted in the APR+. The PAFS replaces the conventional active auxiliary feedwater system(AFWS) by introducing a natural driving force mechanism while maintaining the system function of cooling the primary side and removing the decay heat. As the PAFS completely replaces the conventional AFWS, it is required to verify the cooling capacity of PAFS for the core damage frequency(CDF) evaluation. For this reason, this paper discusses the cooling performance of the PAFS during transient accidents. The test case and scenarios were picked from the result of the sensitivity analysis in APR+ Probabilistic Safety Assessment(PSA). The analysis was performed by the best estimate thermal-hydraulic code, RELAP5/.MOD3.3. This study shows that the plant maintains the stable state without the core damages under the given test scenarios. The results of PSA considering this analysis' results shows that the CDF values are decreased. The analysis results can be used for more realistic and accurate performance of a PSA.

• Journal of the Korea institute for structural maintenance and inspection
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• v.15 no.1
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• pp.180-188
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• 2011

Recently a number of researches about mechanical splice have been studied to apply on a large diameter reinforcing steel bars of spliced sleeves. In this study the structural performance of large diameter reinforcing bars with spliced sleeves was evaluated. For the application of nuclear power plant structures, two different types of existing splices with #11, 14, 18 rebars were fabricated and static and dynamic test were performed on the basis of ASME SEC III DIV.2CC-4330.

• Journal of the Korean Society of Combustion
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• v.16 no.2
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• pp.1-8
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• 2011

As one of the main technologies for carbon capture and storage in power generation, oxy-coal combustion is being developed for field demonstration in Korea. This study presents the results of numerical simulation for combustion in a single-wall-fired 100 $MW_e$-scale boiler proposed for the initial design of the demonstration plant. Using a commercial CFD code, the detailed combustion, flow and heat transfer characteristics were assessed both for air-mode and oxy-mode combustion. The results show that stable combustion can be achieved in the dual mode operation with the current boiler configuration. However, the differences in the flow pattern and heat transfer between the two combustion modes need to be considered in the design and operation which is mainly due to the larger density and specific heat of $CO_2$ compared to $N_2$. Further development of the boiler design is required using improved numerical modeling for radiative heat transfer and combustion.

• Proceedings of the Korean Institute of Information and Commucation Sciences Conference
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• 2017.10a
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• pp.714-715
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• 2017

Much of the interest in ocean radiation detection has been heightened as a lot of radioactivity has leaked to the ocean due to the accident at the Fukushima nuclear power plant in Japan. In the study, MCNP simulation for radiation detection in the ocean was performed. Unlike in the air, the marine environment must ensure the stability of the sensor from water depth, temperature, pressure, and salinity. In the marine environment, too much radiation is shielded. Therefore, it is an object to select a housing with a low radiation shielding ratio.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.14 no.1
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• pp.33-38
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• 2013

The general purposed control system for driving a motion of many different typed robot end-effector, which consists of a controller based on ARM Cotex M3-11017 MCU and an application software for generating a motion of end-effector, was developed. Experimental results show that a positioning error is nearly negligible and a repeatability error is 0.04%. Accordingly the developed control system can be applied practically to actuate a robot end-effector for inspection and maintenance of steam generator heat pipe in nuclear power plant.

• Nuclear Engineering and Technology
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• v.32 no.5
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• pp.465-476
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• 2000

In a nuclear power plant, steam generator tubes cover a major portion of the primary pressure-retaining boundary. Thus very conservative approaches have been taken in the light of steam generator tube integrity According to the present criteria, tubes wall-thinned in excess of 40% should be plugged whatever causes are. However, many analytical and experimental results have shown that no safety problems exist even with thickness reductions greater than 40%. The present criterion was developed about twenty years ago when wear and pitting were dominant causes for steam generator tube degradation. And it is based on tubes with single cracks regardless of the fact that the appearance of multiple cracks is more common in general. The objective of this study is to review the conservatism of the present plugging criteria of steam generator tubes and to propose a new coalescence model for two adjacent through-wall cracks existing in steam generator tubes. Using the existing failure models and experimental results, we reviewed the conservatism of the present plugging criteria. In order to verify the usefulness of the proposed new coalescence model, we performed finite element analysis and some parametric studies. Then, we developed a coalescence evaluation diagram.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.18 no.11
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• pp.1143-1149
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• 2008

In a nuclear power plant, loose part monitoring and its diagnostic technique is one of the major issues for ensuring the structural integrity of the reactor system. Typically, accelerometers are mounted on the surface of a reactor vessel to localize impact location cavsed by the impact of metallic substances on the reactor system. However, in some cases, the number of the accelerometers is not enough to estimate the impact location precisely. In such a case, one of alternative plan is to utilize another type sensors that can measure the vibration of the reactor structure even though the measuring frequency ranges are different from each others. The AE sensors installed on the reactor structure can be utilized as additional sensors for loose part monitoring. In this paper, we proposed a new method to estimate impact location by using both accelerometer signal and AE signal, simultaneously. The feasibility of the proposed method is verified by an experiment. The experimental results demonstrate that we can enhance the reliability and precision of the loose part monitoring.