• Proceedings of the Korean Nuclear Society Conference
• /
• 2003.05a
• /
• pp.180-180
• /
• 2003

• Proceedings of the Korean Nuclear Society Conference
• /
• 2002.05a
• /
• pp.113.1-113
• /
• 2002

• Proceedings of the Korean Nuclear Society Conference
• /
• 1999.10a
• /
• pp.130.2-130
• /
• 1999

• Nuclear Engineering and Technology
• /
• v.23 no.3
• /
• pp.340-351
• /
• 1991

Experimental studies were conducted on a l/5.03 scale reactor flow model of the Yong-gwang Nuclear Units 3 and 4. The purpose of the flow model test was to estimate the hydraulic effect in the reactor vessel due to the relative size difference between the ABB-CE's System 80 and the YGN 3&4 reactors. The flow model was designed according to the principle of similarity. Obtained from the test were the core inlet flow distribution, the core exit pressure deviations, and the segmental and overall pressure losses across the flow path from the reactor vessel inlet to outlet nozzle. These data will be used to provide input data for the core thermal margin analysis and to verify the analytical hydraulic design method.

• Proceedings of the Korean Institute of Information and Commucation Sciences Conference
• /
• 2002.05a
• /
• pp.314-317
• /
• 2002

This paper present a methods for performance evaluation of digital governor and we treat the way to reduce the trial and error when the developed digital governing system is applied to station. The parameters of the model are estimated using input-output data, the model adjustment technique and a genetic algorithm(GA). To verify validity of the completed model, the comparision model and real output signals have been carried out. The developed digital governing system is applied to Sumjingang hydro-power plant, Korea Hydro Nuclear Power Corporation, it has replaced mechanical governing system. Test was performed in order to demonstrate the performance evaluation of the digital governor parameters.

• Journal of the Korean Society of Propulsion Engineers
• /
• v.11 no.5
• /
• pp.37-50
• /
• 2007

This study investigated the characteristics of spray generated by a liquid coaxial swirl injector used in a combustor of the liquid rocket engine. The linear stability analysis considered long and short wave was introduced in liquid sheet breakup. Through the hydrodynamic analysis the initial liquid sheet thickness spray angle and injection velocity were predicted. To evaluate the effect of turbulence model standard $k-{\varepsilon}$ and RNC $k-{\varepsilon}$ model were applied to numerical calculation and it was known that RNC $k-{\varepsilon}$ model was more applicable to predict spray characteristics. On the basis of this evaluation validation of the developed model was performed with swirl injector installed in LPRE and the predicted results of breakup length, spray angle, and SMD agreed well with experiments qualitatively and quantitatively.

• Nuclear Engineering and Technology
• /
• v.18 no.3
• /
• pp.175-182
• /
• 1986

The KNUI (Korea Nuclear Unit 1) loss of offsite power transient as a design-base accident has been simulated using the RELAP5/MOD1/NSC computer code. The analysis is carried out using the best-estimate methodology, but the sequence and its assumptions are based on the evaluation methodology th at emphasizes conservatism. Important thermal-hydraulic parameters such as average temperature, steam generator level and pressurizer water volume are compared with the results in the KNU1 Final Safety Analysis Report (FSAR). The present analysis gives much lower RCS average temperature and pressurizer water volume, and much higher S/G water volume at the turnaround point, which may be considered to be additional improved safety margins. This is expected since the present analysis deals with the best-estimate thermal-hydraulic models as well as the initial conditions on a best-estimate basis. These additional safety margins may contribute to further validate the safety of the KNU1 in this type of accidents(Decrease in Heat Removal by the Secondary System).

• Proceedings of the Korean Nuclear Society Conference
• /
• 1999.05a
• /
• pp.133-133
• /
• 1999

• Proceedings of the Korean Nuclear Society Conference
• /
• 2004.05a
• /
• pp.163.2-163.2
• /
• 2004

• Proceedings of the Korean Nuclear Society Conference
• /
• 1996.11a
• /
• pp.363-368
• /
• 1996

경수로형 원자로의 중대사고 진행시 압력용기내 노심의 용융현상 및 재배치과정 등에 대한 MELCOR 코드내 노심손상모델의 예측 및 분석능력을 검증하고자 하였다. 이를 위하여 노심손상 모의실험중 하나인 독일의 KfK에서 실시된 CORA-13 실험을 선정한 후 이 실험을 MELCOR 1.8.2 코드를 이용하여 계산하였다. 실험결과와 계산결과를 비교분석하고 또한, MELCOR 코드에 대한 민감도분석을 수행함으로써 MELCOR 코드내 손상된 노심의 거동에 대한 열수력모델들을 검증하였다.