• 대한기계학회논문집B
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• 제25권11호
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• pp.1616-1623
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• 2001

A basic experimental study has been carried out to find out the design parameters of fuel 2-staging atomizers in order to reduce nitrogen oxides(NOx) rate emitted from the steam boilers used the liquid fuel. The heavy fuel oil(B-Coil) and fuel 2-staging Y-jet twin-fluid atomizers were adopted in this study. The results of this paper were obtained from the real as well as the model scale atomizers. In the case of model atomizers test, NOx reduction rate was strongly dependent on the staged fuel rate, but it was weakly dependent on the injection hole arrangement and air swirl conditions. The real scale atomizers was designed and manufactured on the base of these test results, and those was mounted and operated in the real boiler generates 185 ton steam per an hour. The reduction rate of the model and real plant was reached 10∼30% of base NOx by atomizers. but dust was sharply increased in the low O$_2$combustion region of the real plant.

• 품질경영학회지
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• 제47권3호
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• pp.425-435
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• 2019

Purpose: The purpose of this study was to resolve the Naval ship's Local Operation Panel(LOP) malfunction problems which caused by overheating in summer season and dispatching to equatorial regions. Methods: Instead of using dual type heat insulation materials(consist with ceramic wool and glass wool), aerogel heat insulation materials were used for decreasing heat emissions from gas-turbine heat waste steam pipes passing water-jet- room. Experiment and Computational analysis of heat flow were conducted to analyze the internal room temperature changes. Results: The results of this study are as follows; The aerogel heat insulation materials suppress heat emission more efficiently than dual type insulation materials. The cold surface temperature of insulation was far more decreased and internal room, LOP surface temperature also showed significant results too. Conclusion: The substituted heat insulation materials appeared remarkable performance in decreasing room temperature that it could be used for suppressing the LOP overheatings and malfunctions.

• 에너지공학
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• 제14권1호
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• pp.62-71
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• 2005

B&C(Blowdown and Condensation)장치를 이용하여 APR1400 실규모 I-Sparger의 증기제트 응축에 의한 수조내 열혼합 현상에 대한 실험이 수행되었다. 한정된 가압기 용량으로 인하여 과도상태 실험이 수행되었으며, 실험을 통해 수조내에 배치된 열전대를 사용하여 열혼합 자료를 얻었다. 측정된 열혼합 자료를 바탕으로 지역별 온도 변화의 경향과 수조 수직-단면상의 온도 윤곽도를 작성하였으며 이를 바탕으로 I-Sparger의 열혼합 특성을 파악하였다. 실험결과에서 I-Sparger에 의한 열혼합 특성은 I-Sparger 설계특성이 나타나는 열혼합 경향을 보이고 있음을 확인하였다.

• Nuclear Engineering and Technology
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• 제38권4호
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• pp.391-398
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• 2006

It is recommended to clean steam generators of nuclear power plants during plant outages. Under normal operations, sludge is created and constantly accumulates in the steam generators. The constituents of this sludge are different depending on each power plant characteristics. The sludge of the Kori Unit 1 steam generator, far example, was found to be composed of 93% ferrous oxide, 3% carbon and 1% of silica oxide and nickel oxide each. The research to develop a lancing system that would remove sludge deposits from the tubesheet of a steam generator was started in 1998 by the Korea Electric Power Research Institute (KEPRI) of the Korea Electric Power Corporation (KEPCO). The first commercial domestic lancing system in Korea, the $KALANS^(R)-I$ Lancing System, was completed in 2000 for Kori Unit 1 for cleaning the tubesheet of its Westinghouse Delta-60 steam generator. Thereafter, the success of the development and site implementation of the $KALANS^(R)-I$ lancing system for YGN Units 1&2 and Ulchin Units 3&4 was also realized in 2004 for sludge removal at those sites. The upper bundle cleaning system for Westinghouse model F steam generators is now under development.

• Nuclear Engineering and Technology
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• 제54권7호
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• pp.2702-2713
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• 2022

A steam generator tube rupture accompanying a core damage may cause the fission product to be released to environment bypassing the containment. In such an accident, the steam generator is the major path of the radioactive aerosol release. AEOLUS facility, the scaled-down model of Korean type steam generator, was built to examine the aerosol removal in the steam generator during the steam generator tube rupture accident. Integral and separate effect tests were performed with the facility for the dry and flooded conditions, and the decontamination factors were presented for different tube configurations and submergences. The dry test results were compared with the existing test results and with the analyses to investigate the aerosol retention physics by the tube bundle, with respect to the particle size and the bundle geometry. In the flooded tests, the effect of submergence were shown and the retention in the jet injection region were presented with respect to the Stokes number. The test results are planned to be used to constitute the aerosol retention model, specifically applicable for the analysis of the steam generator tube rupture accident in Korean nuclear power plants to evaluate realistic fission product behavior.

• 한국마린엔지니어링학회:학술대회논문집
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• 한국마린엔지니어링학회 2005년도 전기학술대회논문집
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• pp.1063-1068
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• 2005

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 추계학술발표회요약집
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• pp.136-136
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• 2002

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 추계학술발표회요약집
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• pp.155.1-155
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• 2002

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2003년도 추계학술발표대회 요약집
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• pp.240.1-240.1
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• 2003

• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 2009년 추계학술대회논문집
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• pp.121-128
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• 2009

During a hypothetical high-pressure accident in a nuclear power plant (NPP), molten corium can be ejected through a breach of a reactor pressure vessel (RPV) and dispersed by a following jet of a high-pressure steam in the RPV. The dispersed corium is fragmented into smaller droplets in a reactor cavity of the NPP by the steam jet and released into other compartments of the NPP by a overpressure in the cavity. The fragments of the corium transfer thermal energy to the ambient air in the containment or interact chemically with steam and generate hydrogen which may be burnt in the containment. The thermal loads from the ejected molten corium on the containment which is called direct containment heating (DCH) can threaten the integrity of the containment. DCH in a NPP containment is related to many physical phenomena such as multi-phase hydrodynamics, thermodynamics and chemical process. In the evaluation of the DCH load, the melt dispersion rates depending on the RPV pressure are the most important parameter. Mostly, DCH was evaluated by using lumped-analysis codes with some correlations obtained from experiments for the dispersion rates. In this study, MC3D code was used to evaluate the dispersion rates in the APR1400 NPP during the high-pressure accidents. MC3D is a two-phase analysis code based on Eulerian four-fields for melt jet, melt droplets, gas and water. The dispersion rates of the corium melt depending on the RPV pressure were obtained from the MC3D analyses and the values specific to the APR1400 cavity geometry were compared to a currently available correlation.