• Nuclear Engineering and Technology
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• 제56권1호
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• pp.275-282
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• 2024

Lead bismuth eutectic (LBE) is used as coolant for MicroURANUS, a small marine nuclear power plant, and this coolant is transported in the plant by an electromagnetic pump. Given the considerable heat generated by the electromagnetic pump, the cooling of the pump is essential. This study compared air cooling and water-cooling methods and found that the maximum temperatures during air and water cooling were 640 K and 372 K, respectively. These findings were utilized to design an electromagnetic pump with water-cooling. The maximum temperature of the pump was lower than the boiling point of water; thus, the pump did not require a separate pressurization. Consequently, the resistance problem of the coil and the deformation problem of the material caused by generated heat can be solved through water-cooling.

• Nuclear Engineering and Technology
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• 제44권7호
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• pp.709-718
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• 2012

The pressurized water reactor APR1400 adopts DVI (Direct Vessel Injection) for the emergency cooling water in the upper downcomer annulus. The International Standard Problem number 50 (ISP-50) was launched with the aim to investigate thermal hydraulic phenomena during a 50% DVI line break scenario with best estimate codes making use of the experimental data available from the ATLAS facility located at KAERI. The present work describes the calculation results obtained for the ISP-50 using the RELAP5/MOD3.3 system code. The work aims at validation and assessment of the code to reproduce the observed phenomena and investigate about its limitations to predict complicated mixing phenomena between the subcooled emergency cooling water and the two-phase flow in the downcomer. The obtained results show that the overall trends of the main test variables are well reproduced by the calculations. In particular, the pressure in the primary system show excellent agreement with the experiment. The loop seal clearance phenomenon was observed in the calculation and it was found to have an important influence on the transient progression. Moreover, the collapsed water levels in the core are accurately reproduced in the simulations. However, the drop in the downcomer level before the activation of the DVI from safety injection tanks was underestimated due to multi-dimensional phenomena in the downcomer that are not properly captured by one-dimensional simulations.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2011년도 춘계학술대회 논문집
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• pp.77-82
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• 2011

Indoor noise is very important related to the morale and fighting power of the crew as well as environmental condition for them in a navel vessel. Especially, the crew members working in CIC(Combat Information Center) and sonar equipment room require much more quiet environment condition in order to increase the performance of their combat system. One of the most serious noise source in CIC is the cooling fan noise installed in the console of the combat system. Therefore, in this paper, the design of the plenum chamber is studied to reduce the noise of the fan from these consoles by numerical analysis.

• 한국소음진동공학회논문집
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• 제21권6호
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• pp.529-535
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• 2011

Indoor noise is very important related to the morale and fighting power of the crew as well as environmental condition for them in a navel vessel. Especially, the crew members working in CIC (combat information center) and sonar equipment room require much more quiet environment condition. One of the most serious noise source in CIC is the cooling fan noise installed in the console of the combat system. Therefore, in this paper, the design of the plenum chamber is studied to reduce the noise of the fan from these consoles by numerical analysis and experiment.

• Nuclear Engineering and Technology
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• 제46권3호
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• pp.373-380
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• 2014

Safety-related parameters are very important for confirming the status of a nuclear power plant. In particular, the reactor vessel water level has a direct impact on the safety fortress by confirming reactor core cooling. In this study, the reactor vessel water level under the condition of a severe accident, where the water level could not be measured, was predicted using a fuzzy neural network (FNN). The prediction model was developed using training data, and validated using independent test data. The data was generated from simulations of the optimized power reactor 1000 (OPR1000) using MAAP4 code. The informative data for training the FNN model was selected using the subtractive clustering method. The prediction performance of the reactor vessel water level was quite satisfactory, but a few large errors were occasionally observed. To check the effect of instrument errors, the prediction model was verified using data containing artificially added errors. The developed FNN model was sufficiently accurate to be used to predict the reactor vessel water level in severe accident situations where the integrity of the reactor vessel water level sensor is compromised. Furthermore, if the developed FNN model can be optimized using a variety of data, it should be possible to predict the reactor vessel water level precisely.

• Nuclear Engineering and Technology
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• 제48권6호
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• pp.1330-1337
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• 2016

A visualization test of the fuel-coolant interaction in the Test for Real cOrium Interaction with water (TROI) test facility was carried out. To experimentally simulate the In-Vessel corium Retention (IVR)- External Reactor Vessel Cooling (ERVC) conditions, prototypic corium was released directly into the coolant water without a free fall in a gas phase before making contact with the coolant. Corium (34.39 kg) consisting of uranium oxide and zirconium oxide with a weight ratio of 8:2 was superheated, and 22.54 kg of the 34.39 kg corium was passed through water contained in a transparent interaction vessel. An image of the corium jet behavior in the coolant was taken by a high-speed camera every millisecond. Thermocouple junctions installed in the vertical direction of the coolant were cut sequentially by the falling corium jet. It was clearly observed that the visualization image of the corium jet taken during the fuel-coolant interaction corresponded with the temperature variations in the direction of the falling melt. The corium penetrated through the coolant, and the jet leading edge velocity was 2.0 m/s. Debris smaller than 1 mm was 15% of the total weight of the debris collected after a fuel-coolant interaction test, and the mass median diameter was 2.9 mm.

• 수산해양교육연구
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• 제27권6호
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• pp.1734-1744
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• 2015

Majority of marine accidents that occur on fishing vessels are engine accidents. This comprises more than 26 % of the total annual fishing vessel marine accident cases. Large numbers of engine accidents happen in the cooling water system, which are mostly caused by negligence on regular check-up and repair. Notably, small-sized ships have higher engine accidents occurrence rate compared to medium-and large-sized ships. Based on the Report of the Korea Ship Safety Technology Authority, engine accident cases reached 3,032 out of the total 3,081 cases. This study researches on the differences between the small-sized ship pilot, an operator of a vessel engine of less than 200 tons, and a 6th level marine engineer, in terms of the relationship between management forms and what causes the marine accidents in association with the cooling water system. It also studies and analyzes the differences in frequency of the accident occurrence between the two groups. ${\chi}^2$ qualification was imposed through the SPSS statistical analysis program and it got qualified at the significance level of 5%. The research shall be utilized as one of the base line data for the reduction of marine accidents.

• 한국전산유체공학회지
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• 제13권3호
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• pp.21-27
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• 2008

The reactor head of a sodium-cooled fast reactor KALIMER-600 should be cooled during the reactor operation in order to maintain the integrity of sealing material and to prevent a creep fatigue. Analyzing turbulent natural convection flow in the cover gas region of reactor vessel with the commercial CFD code CFX10.0, the cooling requirement for the reactor head and the performance of the insulation plate were assessed. The results showed that the high temperature region around reactor vessel was caused by the convective heat transfer of Helium gas flow ascending the gap between the insulation plate and the reactor vessel inner wall. The insulation plate was shown to sufficiently block the radiative heat transfer from pool surface to reactor head to a satisfactory degree. More than $32.5m^3$/sec of cooling air flow rate was predicted to maintain the required temperature of reactor head.

• Nuclear Engineering and Technology
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• 제45권4호
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• pp.469-476
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• 2013

In a postulated severe core damage accident in a PHWR, multiple failures of core cooling systems may lead to the collapse of pressure tubes and calandria tubes, which may ultimately relocate inside the calandria vessel forming a terminal debris bed. The debris bed, which may reach high temperatures due to the decay heat, is cooled by the moderator in the calandria. With time, the moderator is evaporated and after some time, a hot dry debris bed is formed. The debris bed transfers heat to the calandria vault water which acts as the ultimate heat sink. However, the questions remain: how long would the vault water be an ultimate heat sink, and what would be the failure mode of the calandria vessel if the heat sink capability of the reactor vault water is lost? In the present study, a numerical analysis is performed to evaluate the thermal loads and the stresses in the calandria vessel following the above accident scenario. The heat transfer from the molten corium pool to the surrounding is assumed to be by a combination of radiation, conduction, and convection from the calandria vessel wall to the vault water. From the temperature distribution in the vessel wall, the transient thermal loads have been evaluated. The strain rate and the vessel failure have been evaluated for the above scenario.

• 대한기계학회논문집A
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• 제38권4호
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• pp.355-361
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• 2014

신규 원전에서 추진중인 외벽침수냉각 방식의 적용이 실패할 경우 노심용융물과 원자로공동 내유체의 상호작용으로 인해 증기폭발이 발생하며, 이는 격납건물 및 관통부 배관을 포함한 각 구조물의 건전성을 위협할 수 있다. 본 논문에서는 선행연구 분석결과를 토대로 증기폭발 현상을 모사할 수 있는 개선된 해석기법을 도출하고 알루미나 실험 모사를 통해 타당성을 확인하였다. 또한 동일한 기법을 원자로공동 해석에 적용하여 가상 파손위치에 따른 증기폭발 압력이력을 예측하였으며, 측면파손에 의한 최대압력 값이 하부파손에 의한 것보다 최대 70% 정도 높음을 보였다.