• 설비공학논문집
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• 제13권5호
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• pp.398-410
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• 2001

The flow and heat transfer characteristics for three-dimensional mixed convection flows in a radiator flat tube with U--shaped grooves are analyzed numerically. The flow and temperature fields are calculated by using the modified SIMPLE algorithm for irregular geometry. One tube specification among the various flat tube exchangers is recommended by considering the heat transfer and pressure drop. The effects of variation of coolant flow conditions and external air conditions on the flow and the thermal characteristics for the selected tube are investigated. the results show that inlet velocity of coolant flow is the very important factor in heat transfer and pressure drop, and top side is better position than the others as fin cleave to tube.

• Nuclear Engineering and Technology
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• 제48권4호
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• pp.975-981
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• 2016

The containment response during the first 24 hours of a low-pressure severe accident scenario in a nuclear power plant with a two-loop Westinghouse-type pressurized water reactor was simulated with the CONTAIN 2.0 computer code. The accident considered in this study is a large-break loss-of-coolant accident, which is not successfully mitigated by the action of safety systems. The analysis includes pressure and temperature responses, as well as investigation into the influence of spray on the retention of fission products and the prevention of hydrogen combustion in the containment.

• 한국압력기기공학회 논문집
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• 제11권2호
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• pp.25-30
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• 2015

In the present study, complex leakage probabilities of nuclear pipes due to fatigue and stress corrosion cracking are evaluated by using the PINTIN(Piping INTegrity INner flaws) that is developed based on the existing PRAISE(Piping Reliability Analysis Including Seismic Events) program. With regard to the aging and crack instability, small leak and big leak probabilities are calculated for several pipes in a reactor coolant system of domestic nuclear plant. Moreover, sensitivity analysis is also performed to find out the effect of parameters for the leakage of pipes, which shows the coolant temperature is the most influencing parameter.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.669-674
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• 1996

The analysis for the coolability of the reactor cavity in typical Korean 1000 MWe Nuclear Unit under severe accidents is performed using MELCOR 1.8.3 code. The key parameters molten core-concrete interaction(MCCI) such as melt temperature, concrete ablation history and gas generation are investigated. Total twenty cases are selected according to ejected debris fraction and coolant mass, The ablation rate of concrete decreases as mass of the melt decreases and coolant mass increases. Heat loss from molten pool to coolant is comparable to total decay heat, so concrete ablation is delayed until water is absent and crust begins to remove. Also, overpressurization due to non-condensible gases generated during corium and concrete interacts can cause to additional risk of containment failure. It is concluded that flooded reactor cavity condition is very important to minimize the cavity ablation and pressure load by non-condensible gases on containment.

• 한국전산구조공학회:학술대회논문집
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• 한국전산구조공학회 1995년도 가을 학술발표회 논문집
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• pp.61-67
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• 1995

The steam generator, one of the major components in the reactor coolant system, plays an important role in transferring the thermal energy made in the reactor during normal operation to the secondary side and producing steam to drive turbine. A hydraulic snubber system is used in order to protect the steam generator under the dynamic loading condition and to absorb the thermal expansion transmitted by the reactor coolant piping due to high temperature and pressure during normal operation. In this study, the model for a geometrical linkage system is presented to analyze the snubber stroke of the steam generator and the parameters in the snubber stroke analysis are investigated. A method to analyze lever ratio of the linkage system which is required in the process of determining the snubber stiffness value is also presented. To discuss the validation of the suggested analysis, the analysis results are compared with the measured data during the hot functional test for the standardized 1000 Mwe pressurized water reactor plant under the construction.

• Nuclear Engineering and Technology
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• 제50권6호
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• pp.829-841
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• 2018

An experiment was conducted for the OECD/NEA ROSA-2 Project using the large-scale test facility (LSTF), which simulated a 17% hot leg intermediate-break loss-of-coolant accident in a pressurized water reactor (PWR). In the LSTF test, core uncovery started simultaneously with liquid level drop in crossover leg downflow-side before loop seal clearing, and water remaining occurred on the upper core plate in the upper plenum. Results of the uncertainty analysis with RELAP5/MOD3.3 code clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges. For studying the scaling problems to extrapolate thermal-hydraulic phenomena observed in scaled-down facilities, an experiment was performed for the OECD/NEA PKL-3 Project with the Primarkreislaufe Versuchsanlage (PKL), as a counterpart to a previous LSTF test. The LSTF test simulated a PWR 1% hot leg small-break loss-of-coolant accident with steam generator secondary-side depressurization as an accident management measure and nitrogen gas inflow. Some discrepancies appeared between the LSTF and PKL test results for the primary pressure, the core collapsed liquid level, and the cladding surface temperature probably due to effects of differences between the LSTF and the PKL in configuration, geometry, and volumetric size.

• Nuclear Engineering and Technology
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• 제49권5호
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• pp.928-940
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• 2017

An experiment using the $Prim{\ddot{a}}rkreisl{\ddot{a}}ufe$ Versuchsanlage (PKL) was performed for the OECD/NEA PKL-3 Project as a counterpart to a previous test with the large-scale test facility (LSTF) on a cold leg smallbreak loss-of-coolant accident with an accident management (AM) measure in a pressurized water reactor. Concerning the AM measure, the rate of steam generator (SG) secondary-side depressurization was controlled to achieve a primary depressurization rate of 200 K/h as a common test condition; however, the onset timings of the SG depressurization were different from each other. In both tests, rapid recovery started in the core collapsed liquid level after loop seal clearing, which caused whole core quench. Some discrepancies appeared between the LSTF and PKL test results for the core collapsed liquid level, the cladding surface temperature, and the primary pressure. The RELAP5/MOD3.3 code predicted the overall trends of the major thermal-hydraulic responses observed in the LSTF test well, and indicated a remaining problem in the prediction of primary coolant distribution. Results of uncertainty analysis for the LSTF test clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges.

• 한국압력기기공학회 논문집
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• 제12권1호
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• pp.107-114
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• 2016

Four cylindrically shaped IHXs(Intermediate Heat Exchangers) are installed in the PHTS(Primary Heat Transfer System) of the PGSFR(Prototype Gen IV Sodium cooled Fast Reactor). As for the IHX, the temperature difference of structure is inevitable result caused by heat transfer between primary coolant sodium and IHTS(Intermediate Heat Transport System) sodium. It is necessary to evaluate the high temperature structural integrity of IHXs which operate at the elevated temperature condition over the creep temperature. In this paper, the high temperature structural integrity of IHX under assumed loading conditions has been reviewed according to ASME code.

• 한국압력기기공학회 논문집
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• 제19권2호
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• pp.109-116
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• 2023

This study presents a CRUD modelling to simulate the thermal resistance behavior of CRUD, deposited on the surface of a cladding tube of a fuel assembly. When heat produced from fuels transfers to a coolant through a cladding tube, the CRUD acting as an additional thermal resistance is expressed as two layers, i.e., a solid oxide layer and an imaginary fluid layer, which are added to the experimental tube's heat structure of the MARS-KS input data. The validation calculation for the experiments performed in UNIST-DISNY facility showed that the center and surface temperatures of the cladding tube increased as the porosity and the steam amount inside pores of the CRUD got higher. In addition, the temperature gradient in the imaginary fluid layer was calculated to be larger than that in the solid oxide part, indicating that the steam amount inside the layer acted more largely as thermal resistance. It was also evaluated through sensitivity calculations that the cladding tube temperature was more sensitive to the CRUD porosity and the steam amount in pores than to the inlet flow rate of the coolant.

• 한국추진공학회:학술대회논문집
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• 한국추진공학회 2003년도 제20회 춘계학술대회 논문집
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• pp.9-12
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• 2003

KSR-III 축소형 엔진을 원형으로 하는 8채널형 칼로리미터의 냉각성능해석을 수행하였다. 축대칭 압축성 해석에 의한 연소실 벽으로의 열유속을 이용하여 3차원 냉각유로 내부의 열전달 해석을 수행하였다. 연소실벽으로의 열유속은 문헌에서 제시하는 수준으로 확인되었으며 열전달 해석을 통하여 칼로리미터 개발과 운용에 필요한 냉각수의 압력강하, 냉각수 온도상승 및 연소실벽의 최고온도를 제시하였다.