• Transactions of the Korean Society of Mechanical Engineers B
• /
• v.25 no.11
• /
• pp.1616-1623
• /
• 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.

• Journal of Korean Society for Quality Management
• /
• v.47 no.3
• /
• pp.425-435
• /
• 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.

• Journal of Energy Engineering
• /
• v.14 no.1
• /
• pp.62-71
• /
• 2005

An experimental study on thermal mixing of steam jet condensation through the I-Sparger of APR1400 design using B&C (Blowdown and Condensation) test facility. Due to the limit of the steam supply capability of the pressurizer, transient thermal mixing experiments were conducted. Temperature distributions in the quench tank were measured using thermocouples located at various positions. From the experimental data, local temperature variations for various locations and vertically cross-sectional temperature distributions for several times were depicted and presented. The result shows the characteristics of thermal mixing of the I-Sparger depending on the design features of the I-Sparger.

• Nuclear Engineering and Technology
• /
• v.38 no.4
• /
• pp.391-398
• /
• 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
• /
• v.54 no.7
• /
• pp.2702-2713
• /
• 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.

• Proceedings of the Korean Society of Marine Engineers Conference
• /
• 2005.06a
• /
• pp.1063-1068
• /
• 2005

• Proceedings of the Korean Nuclear Society Conference
• /
• 2002.10a
• /
• pp.136-136
• /
• 2002

• Proceedings of the Korean Nuclear Society Conference
• /
• 2002.10a
• /
• pp.155.1-155
• /
• 2002

• Proceedings of the Korean Nuclear Society Conference
• /
• 2003.10a
• /
• pp.240.1-240.1
• /
• 2003

• 한국전산유체공학회:학술대회논문집
• /
• 2009.11a
• /
• pp.121-128
• /
• 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.