• Proceedings of the Korean Nuclear Society Conference
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• 1998.10a
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• pp.158-158
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• 1998

• Advances in aircraft and spacecraft science
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• v.7 no.3
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• pp.229-249
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• 2020

This paper is the logical follow-up of four papers by the author on the subject "aerodynamics in Mars atmosphere". The aim of the papers was to evaluate the influence of two Mars atmosphere models (NASA Glenn and GRAM-2001) on aerodynamics of a capsule (Pathfinder) entering the Mars atmosphere and also to verify the feasibility of evaluating experimentally the ambient density and the ambient pressure by means of the methods by McLaughlin and Cassanto respectively, therefore to correct the values provided by the models. The study was carried out computationally by means of: i) a code integrating the equations of dynamics of an entry capsule for the computation of the trajectories, ii) two Direct Simulation Monte Carlo (DSMC) codes for the solution of the 2-D, axial-symmetric and 3-D flow fields around the capsule in the altitude interval 50-100 km. The computations verified that the entry trajectories of Pathfinder from the two models, in terms of the Mach, Reynolds and Knudsen numbers, were very different. The aim of the present paper is to continue this study, considering other aerodynamic problems and then to provide a contribution to a long series of papers on the subject "aerodynamics in Mars atmosphere". More specifically, the present paper evaluated and quantified the effects from the two models of: i) chemical reactions on aerodynamic quantities in the shock layer, ii) surface temperature, therefore of the contribution of the re-emitted molecules, on local (pressure, skin friction, etc.) and on global (drag) quantities, iii) surface recombination reactions (catalyticity) on heat flux. The results verified that the models heavily influence the flow field (as per the shock wave structure) but, apart from the surface recombination reactions, the effects of the different conditions on aerodynamics of the capsule are negligible for both models and confirmed what already found in the previous paper that, because of the higher values of density from the NASA Glenn model, the effects on aerodynamics of a entry capsule are stronger than those computed by the GRAM-2001 model.

• Journal of the Korean Society of Safety
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• v.36 no.3
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• pp.74-80
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• 2021

Korea Atomic Energy Research Institute (KAERI) has been operating an integral effects test facility, the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS), according to APR1400 for transient experimental and design basis accident simulation. Moreover, based on the experimental data, the domestic standard problem (DSP) program has been conducted in Korea to validate system codes. Recently, through DSP-05, the performance of the passive auxiliary feedwater system (PAFS) in the event of multiple steam generator tube rupture (MSGTR) has been analyzed. However, some errors exist in the reference input model distributed for DSP-05. Furthermore, the calculation results of the heat loss correlation for the secondary system presented in the technical report of the reference indicate that a large difference is present in heat loss from the target value. Thus, in this study, the reference model is corrected using the geometric information from the design report and drawings of ATLAS. Additionally, a new heat loss correlation is suggested by fitting the results of the heat loss tests. Herein, MSGTR-PAFS accident analysis is performed using MARS-KS 1.5 with the improved model. The steady-state calculation results do not significantly differ from the experimental values, and the overall physical behavior of the transient state is properly predicted. Particularly, the predicted operating time of PAFS is similar to the experimental results obtained by the modified model. Furthermore, the operating time of PAFS varies according to the heat loss of the secondary system, and the sensitivity analysis results for the heat loss of the secondary system are presented.

• Journal of Energy Engineering
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• v.28 no.4
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• pp.103-110
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• 2019

The domestic innovative power reactor named iPOWER will employ the passive molten corium cooling system(PMCCS) to cool down and stabilize the core melt in the severe accident. The final design concept of the PMCCS is yet to be determined, but the in-vessel retention through external reactor vessel cooling has been also considered as a viable strategy to cope with the severe accident. In this study, the two-phase natural circulation flow established between the reactor vessel and the insulation was simulated using a thermal-hydraulic system code, MARS-KS. The flow path of cooling water was modeled with one-dimensional nodes, and the boundary condition of the heat load from the molten core was defined to estimate the naturally-driven flow rate. The evolution of major thermal-hydraulic parameters were also evaluated, including the temperature and the level of cooling water, the void fraction around the lower head of the reactor vessel, and the heat transfer mode on its external surface.

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.155.2-155
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• 2002

A visual environment for system analysis codes (hereinafter called "ViSA") has been developed to support code users in their input preparations, code executions, and output interpretations. ViSA provides a more convenient way for base input data generation and modification on a user-friendly basis. It also provides on-line graphical displays to give an in-depth understanding of transient thermal-hydraulic behaviors in nuclear power plants. This paper presents the main features of ViSA.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.565-566
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.305-306
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.17-18
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.500-501
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• 2005

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1698-1711
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• 2022

A combined safety approach, which uses a best-estimate computer code and adopts conservative assumptions for safety systems availability, is developed and applied to the safety margin evaluation for the Loss of Condenser Vacuum (LOCV) of the 1000 MWe Korean Nuclear Power Plant. The Multi-dimensional Analysis of Reactor Safety-KINS standard (MARS-KS) code is selected as a best-estimate code and the PAPIRUS program is used to obtain different initial operational conditions through random sampling of control variables. During an LOCV event, fuel integrity is not threatened by the increase in Departure from Nuclear Boiling Ratio (DNBR). However, the high pressure in the primary coolant system and the secondary system might affect the system integrity. Thus, the peak pressure becomes a major safety concern. Transient analyses are performed for 124 cases of different initial conditions and the most conservative case, which results in the highest system pressure is selected. It is found the suggested methodology gives similar peak pressures when compared to those predicted from existing methodologies. The proposed approach is expected to minimize the time and efforts required to identify the conservative plant conditions in the existing conservative safety methodologies.