• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.23-28
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• 1996

The potential for the RELAP5/MOD3.2 was assessed for the loss-of-RHR event during the mid-loop operation and the predictability of major thermal-hydraulic phenomena was also evaluated for the long term transient. The analysis results of the typical two cases(cold leg opening case and pressurizer opening case) were compared with experimental data which was conducted at ROSA-IV/LSTF in Japan. As a result, it was shown that tile code was capable of simulating the thermal-hydraulic transport process with appropriate time step during the reduced inventory operation with the loss-of- RHR system.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.551-556
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• 1997

A present study addresses a loss-of-RHR event as an initiating event under specific low power or shutdown conditions. Two typical plant configurations, cold leg opening case with water-filled steam generators and pressurizer opening case with emptied steam generators, were evaluated using the RELAP5/ MOD3.2 code. The calculation was compared with the experiment conducted at ROSA-IV/LSTF in Japan. As a result, the code was capable of simulating the system transient behavior following the event. Especially, thermal hydraulic transport processes including non-condensable gas behavior were reasonably predicted with an appropriate time step and CPU time. However, there were some code deficiencies such as too large system mass errors and severe flow oscillations in core region.

• Korean Chemical Engineering Research
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• v.57 no.1
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• pp.11-16
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• 2019

$La_{0.6}Sr_{0.4}Ti_{0.3}Fe_{0.7}O_{3-{\delta}}$(LSTF) coated $Ba_{0.5}Sr_{0.5}Co_{0.8}Fe_{0.2}O_{3-{\delta}}$(BSCF) membranes which has properties of high oxygen permeability and stability to $CO_2$ were applied to a bench scale apparatus to conduct oxygen permeation experiments. Also, the membranes of the laboratory and the bench scale device were divided into three regions according to the temperature gradient in the membrane reactor for comparative analysis. While oxygen permeation experiment were conducted up to $900^{\circ}C$, temperature dependence of Cr deposition was investigated. As a result, it was confirmed that the oxygen permeability was $2.37ml/min{\cdot}cm^2$, which was significantly lower than $3.79ml/min{\cdot}cm^2$ measured in the laboratory apparatus. It was found through XRD and SEM/EDS analysis that the decrease in oxygen permeability was originated from the deposition of gaseous Cr on the membrane surface released from the alloy material of the housing. In particular, a large amount of Cr was found in the medium temperature region.

• Nuclear Engineering and Technology
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• v.26 no.1
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• pp.41-53
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• 1994

A SGTR accident postulated at Kori unit 1 is simulated with Mihama unit experience, which occurred on February 1991, to evaluate the capability of plant to cope with the transient. The system design and plant conditions of Kori Unit 1 are much similar with those of Mihama Unit 2. Therefore, special concern has been given to evaluate the sequence and the resulting consequence of the postulated SGTR accident at the Kori unit 1 An analysis is peformed as realistically as possible, with following the EOP of Kori unit 1. The result indicates that the leak through tube break terminates within about forty minutes, and the Kori unit 1 may be sufficient to cope with SGTR accident with same type of sequence. However, the reconsideration may be required for the design of Kori unit 1 which disconnects non-safety AC power from off-site power on SI signal generation. It may be pointed out that the content of EOP for SGTR accident is not enough to require operator's proper judgements. An analysis of SGTR accident tested in the LSTF which simulated the SGTR accident at the Mihama Unit 2 is peformed using the RELAP5/MOD3. The results indicates that the code yields in general good agreement with the test, except the break flowrate at the early stage of the event.

• Nuclear Engineering and Technology
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• v.35 no.3
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• pp.179-190
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• 2003

The multiple steam generator tube rupture (SGTR) event scenario with available safety systems was experimentally and analytically evaluated. The experiment was conducted on the large scaled test facility to simulate the multiple SGTR event and investigate the effectiveness of operator actions. As a result, it indicated that the opening of pressurizer power operated relief valve was significantly effective in quickly terminating the primary-to-secondary break flow even for the 6.5 tubes rupture. In the analysis, the recent version of RELAP5 code was assessed with the test data. It indicated that the calculations agreed well with the measured data and that the plant responses such as the water level and relief valve cycling in the damaged steam generator were reasonably predicted. Finally, sensitivity study on the number of ruptured tubes up to 10 tubes was performed to investigate the coolant release into atmosphere. It indicated that the integrated steam mass released was not significantly varied with the number of ruptured tubes although the damaged steam generator was overfilled for more than 3 tubes rupture. These findings are expected to provide useful information in understanding and evaluating the plant ability to mitigate the consequence of multiple SGTR event.

• International Journal of Naval Architecture and Ocean Engineering
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• v.3 no.3
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• pp.216-224
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• 2011

In This study develops a quasi-three dimensional numerical model of wave driven coastal currents with accounting the effects of the wave-current interaction and the surface rollers. In the wave model, the current effects on wave breaking and energy dissipation are taken into account as well as the wave diffraction effect. The surface roller associated with wave breaking was modeled based on a modification of the equations by Dally and Brown (1995) and Larson and Kraus (2002). Furthermore, the quasi-three dimensional model, which based on Navier-Stokes equations, was modified in association with the surface roller effect, and solved using frictional step method. The model was validated by data sets obtained during experiments on the Large Scale Sediment Transport Facility (LSTF) basin and the Hazaki Oceanographical Research Station (HORS). Then, a model test against detached breakwater was carried out to investigate the performance of the model around coastal structures. Finally, the model was applied to Akasaki port to verify the hydrodynamics around coastal structures. Good agreements between computations and measurements were obtained with regard to the cross-shore variation in waves and currents in nearshore and surf zone.

• Journal of Korean Society of Coastal and Ocean Engineers
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• v.29 no.6
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• pp.350-362
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• 2017

The Eulerian nearshore current model is more advantageous than the Lagrangian model in the way that numerical results from the Eulerian model can be directly compared with the measurements by the stationary equipment. It is because the wave mass flux is not included in the computed mass flux of Euleran nearshore current model. In addition, the Eulerian model can simulate the longshore currents with depth varying parabolic profile. However, the numerical models proposed by different researcher have different forms of the wave stress terms. For example, wave stresses in Newberger and Allen's (2007) model is constant over the depth, while those of Chun (2012) are vertically distributed. In the present study, these wave stress terms were compared against Hamilton et al.'s (2001) laboratory experiments to see the effects of different wave stress terms performed on the computation of nearshore currents.

• Nuclear Engineering and Technology
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• v.24 no.3
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• pp.243-251
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• 1992

Behavior of loop seal formation and clearing during small break loss of coolant accident is investigated using the RELAP5/MOD 2 and /MOD3 codes with the test of SB-CL-18 of the LSIF (Large Scale Test Facility）. The present study examines the thermal-hydraulic mechanisms responsible for early core uncovery including the manometric effect due to an asymmetric coolant holdup in the steam generator upflow and downflow side. The analysis with the RELAP5/MOD2 demonstrates the main phenomena occuring in the depressurization transient including the loop seal formation and clearing with sufficient accuracy. Nevertheless, several differences regarding the evolution of phenomena and their timing have been pointed out in かe base calculations. The RELAP5/MOD3 predicts overall phenomena, particularly the steam generator liquid holdup better than the RELAP5/MOD2. The nodalization study in the components of the steam generator U-tubes and the cross-over legs wiか the RELAP5/MOD3 results in good prediction of the loop seal clearing phenomena and their timing.

• Membrane Journal
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• v.22 no.1
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• pp.8-15
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• 2012

Tubular $Ba_{0.5}Sr_{0.5}Co_{0.8}Fe_{0.2}O_{3-{\delta}}$ membranes with $La_{0.6}Sr_{0.4}Ti_{0.3}Fe_{0.7}O_{3-{\delta}}$ porous coating layer were prepared by extrusion and dip coating technique. XRD and SEM result showed the tubular membrane possessed the perovskite structure and porouscoating layer (thickness= about $2{\mu}m$) in surface. The oxygen permeation test was measured at condition of ambient air (feed side) and vacuum (permeate side) in the temperature range from 750 to $950^{\circ}C$. The oxygen permeation flux of $Ba_{0.5}Sr_{0.5}Co_{0.8}Fe_{0.2}O_{3-{\delta}}$ tubular membrane with $La_{0.6}Sr_{0.4}Ti_{0.3}Fe_{0.7}O_{3-{\delta}}$ porous coating layer reached maximum $3.2mL/min{\cdot}cm^2$ at $950^{\circ}C$ and was higher than non-coated $Ba_{0.5}Sr_{0.5}Co_{0.8}Fe_{0.2}O_{3-{\delta}}$ tubular membrane. Long-term stability test result indicated that the oxygen permeation flux was quite stable during the 11 day.