• Proceedings of the Korean Society of Propulsion Engineers Conference
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• 한국추진공학회 2011년도 제37회 추계학술대회논문집
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• pp.228-236
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• 2011

The analysis of the performance and the internal flow according to various numerical models used for the interior ballistics has been conducted. The internal flow has been mainly affected by the drag model. As results, oscillations of the pressure differences between the breech and the shot base has been reduced with a deceased drag of the propellant. The major performance of the interior ballistics has no relation to the models using Nusselt number for heat transfer coefficient. The negative pressure difference without the heat transfer of the propellant has not been shown.

• Nuclear Engineering and Technology
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• 제26권1호
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• pp.126-139
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• 1994

Best-estimate thermal-hydraulic codes, CATHARE2 V1.2 and RELAP5/MOD3, hate been assessed against the BETHSY 6.2 tc six-inch cold leg break loss-of-coolant accident (LOCA) test. Main objective is to analyze the overall capabilities of the two codes on physical phenomena of concern during the small break LOCA i.e. two-phase critical flow, depressurization, core water level de-pression, loop seal clearing, liquid holdup, etc. The calculation results show that the too codes predict well both in the occurrences and trends of major two-phase flow phenomena observed. Especially, the CATHARE2 calculations show better agreements with the experimental data. However, the two codes, in common, show some deviations in the predictions of loop seal clearing, collapsed core water level after the loop seal clearing, and accumulator injection behaviors. The discrepancies found from the comprision with the experimental data are larger in the RELAP5 results than in the CATHARE2. To analyze the deviations of the two code predictions in detail, several sensitivity calculations have been performed. In addition to the change of two-phase discharge coefficients for the break junction, fine nodalization and some corrections of the interphase drag term are made. For CATHARE2, the change of interphase drag force improves the mass distribution in the primary side. And the prediction of SG pressure is improved by the modification of boundary conditions. For RELAP5, any single input change doesn't improve the whole result and it is found that the interphase drag model has still large uncertainties.

• Journal of computational fluids engineering
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• 제16권2호
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• pp.62-65
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• 2011

Characteristics of the pressure drop in an expanded bed have been compared to those in a packed bed for numerical study of the interphase drag in gas-particle flows. A numerical analysis of the pressure drop by the particle drag has been conducted according to the tube-to-particle diameter ratios and Reynolds numbers for comparison. As the tube-to-particle diameter ratios increase at the same Reynolds number, the pressure drop tends to converge. It has been confirmed that characteristics of the pressure drop in the expanded bed are similar to those in the packed bed.

• Journal of the Korean Society of Propulsion Engineers
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• 제17권4호
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• pp.16-24
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• 2013

The analysis of performance and internal flow according to various numerical models for interior ballistics has been conducted. The initial flow has been mainly affected by the drag model of propellants and their drag degradation reduces oscillations of differential pressure between the breech and the shot base. Models of Nusselt number haven't influenced the major performance of interior ballistics. The negative differential pressure isn't generated in the case without the heat transfer of propellants.

• Nuclear Engineering and Technology
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• 제18권1호
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• pp.9-16
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• 1986

Simulation of the system thermal-hydraulic parameters was carried out following the KNUl (Korea Nuclear Unit-1) loss of normal feedwater transient sequence occurred on November 14, 1984. Results were compared with the plant transient data, and good agreements were obtained. Some deviations were found in the parameters such as the steam flowrate and the RCS (Reactor Coolant System) average temperature, around the time of reactor trip. It can be expected since the thermal-hydraulic parameters encounter rapid transitions due to the large reduction of the reactor thermal power in a short period of time and, thereby, the plant data involve transient uncertainties. The analysis was performed using the RELAP5/MOD1/NSC developed through some modifications of the interphase drag and the wall heat transfer modeling routines of the RELAP5/MOD1/CY018.