• Nuclear Engineering and Technology
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• v.55 no.12
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• pp.4335-4349
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• 2023

Nuclear marine propulsion has been emerging as a next generation carbon-free power source, for which proper passive residual heat removal systems (PRHRSs) are needed for long-term safety. In particular, the characteristics of unlimited operation time and compact design are crucial in maritime applications due to the difficulties of safety aids and limited space. Accordingly, a compact and long-term-operable PRHRS has been proposed with the key design concept of using both air cooling and seawater cooling in tandem. To confirm its feasibility, this study conducted system design and a transient analysis in an accident scenario. Design results indicate that seawater cooling can considerably reduce the overall system size, and thus the compact and long-term-operable PRHRS can be realized. Regarding the transient analysis, the Multi-dimensional Analysis of Reactor Safety (MARS-KS) code was used to analyze the system behavior under a station blackout condition. Results show that the proposed design can satisfy the design requirements with a sufficient margin: the coolant temperature reached the safe shutdown condition within 36 h, and the maximum cooling rate did not exceed 40 ℃/h. Lastly, it was assessed that both air cooling and seawater cooling are necessary for achieving long-term operation and compact design.

• Nuclear Engineering and Technology
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• v.34 no.5
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• pp.421-432
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• 2002

As one of the advanced design features of the APR1400, direct vessel injection (DVI) system is being considered instead of conventional cold leg injection (CLI) system. It is known that the DVI system greatly enhances the reliability of the emergency core cooling (ECC) system. However, there is still a dispute on its performance in terms of water delivery to the reactor core during the reflood phase of a large-break loss-of-coolant accident (LOCA). Thus, experimental validation is under progress. In this paper, test results of direct ECC bypass performed in the steam-water test facility tailed MIDAS (Multi-dimensional Investigation in Downcomer Annulus Simulation) are presented. The test condition is determined, based on the preliminary analysis of TRAC code, by applying the ‘modified linear scaling method’with the l/4.93 length scale . From the tests, ECC direct bypass fraction, steam condensation rate and information on the flow distribution in the upper annulus downcomer region are obtained.

• Journal of the Korean Society of Safety
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• v.33 no.6
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• pp.144-156
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• 2018

Level 3 Probabilistic Safety Assessment (PSA) is performed for the risk assessment that calculates radioactive material dispersion to the environment. This risk assessment is performed with a tool of MELCOR Accident Consequence Code System (MACCS2 or WinMACCS). For the off-site consequence analysis of multi-unit nuclear power plant (NPP) accident, the single location (Center Of Mass, COM) method has been usually adopted with the assumption that all the NPPs in the nuclear site are located at the same COM point. It was well known that this COM calculation can lead to underestimated or overestimated radionuclide concentration. In order to overcome this underestimation or overestimation of radionuclide concentrations in the COM method, Multiple Location (ML) method was developed in this study. The radionuclide concentrations for the individual NPPs are separately calculated, and they are summed at every location in the nuclear site by the post-processing of radionuclide concentrations that is based on two-dimensional Gaussian Plume equations. In order to demonstrate the efficiency of the ML method, radionuclide concentrations were calculated for the six-unit NPP site, radionuclide concentrations of the ML method were compared with those by COM method. This comparison was performed for conditions of constant weather, yearly weather in Korea, and four seasons, and the results were discussed. This new ML method (1) improves accuracy of radionuclide concentrations when multi-unit NPP accident occurs, (2) calculates realistic atmospheric dispersion of radionuclides under various weather conditions, and finally (3) supports off-site emergency plan optimization. It is recommended that this new method be applied to the risk assessment of multi-unit NPP accident. This new method drastically improves the accuracy of radionuclide concentrations at the locations adjacent to or very close to NPPs. This ML method has a great strength over the COM method when people live near nuclear site, since it provides accurate radionuclide concentrations or radiation doses.

• Transactions of The Korea Fluid Power Systems Society
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• v.6 no.2
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• pp.1-6
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• 2009

The APU(Auxiliary Power Unit) needs a set of complex pipeline for the fuel supply system where some of the main valves controlling the flow rate consist of the servo valve worked with a torque motor. The input electric current produces an induced magnetic field almost perpendicular to the background magnetic filed generated by fixed permanent magnets. The induced torque deforms the tubular bushing, and directly rotates an armature, which can open and close the valve. In this study, we start from a basic analytic model using a simple electro-magneto-statics, and expand our model to the three-dimensional one computationally applying a commercial code named COMSOL. The result is compared with each other, and reasonable numerical data are obtained for the dynamic behavior and multi-physics system.

• Journal of computational fluids engineering
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• v.21 no.2
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• pp.112-119
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• 2016

NEXTSat-1 is the next-generation small-size artificial satellite system planed by the Satellite Technology Research Center(SatTReC) in Korea Advanced Institute of Science and Technology(KAIST). For the control of attitude and transition of the orbit, the system has adopted a RHM(Resisto-jet Head Module), which has a very simple geometry with a reasonable efficiency. An axisymmetric model is devised with two coil-resistance heaters using xenon(Xe) gas, and the minimum required specific impulse is 60 seconds under the thrust more than 30 milli-Newton. To design the module, seven basic parameters should be decided: the nozzle shape, the power distribution of heater, the pressure drop of filter, the diameter of nozzle throat, the slant length and the angle of nozzle, and the size of reservoir, etc. After quasi one-dimensional analysis, a theoretical value of specific impulse is calculated, and the optima of parameters are found out from the baseline with a series of multi-physical numerical simulations based on the compressible Navier-Stokes equations for gas and the heat conduction energy equation for solid. A commercial code, COMSOL Multiphysics is used for the computation with a FEM (finite element method) based numerical scheme. The final values of design parameters indicate 5.8% better performance than those of baseline design after the verification with all the tuned parameters. The present method should be effective to reduce the time cost of trial and error in the development of RHM, the thruster of NEXTSat-1.

• Nuclear Engineering and Technology
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• v.45 no.6
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• pp.759-766
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• 2013

As passive safety features for nuclear power plants receive increasing attention, various studies have been conducted to develop safety systems for 3rd-generation (GEN-III) nuclear power plants that are driven by passive systems. The Passive Auxiliary Feedwater System (PAFS) is one of several passive safety systems being designed for the Advanced Power Reactor Plus (APR+), and extensive studies are being conducted to complete its design and to verify its feasibility. Because the PAFS removes decay heat from the reactor core under transient and accident conditions, it is necessary to evaluate the heat removal capability of the PAFS under hypothetical accident conditions. The heat removal capability of the PAFS is strongly dependent on the heat transfer at the condensate tube in Passive Condensation Heat Exchanger (PCHX). To evaluate the model of heat transfer coefficient for condensation, the Multi-dimensional Analysis of Reactor Safety (MARS) code is used to simulate the experimental results from PAFS Condensing Heat Removal Assessment Loop (PASCAL). The Shah model, a default model for condensation heat transfer coefficient in the MARS code, under-predicts the experimental data from the PASCAL. To improve the calculation result, The Thome model and the new version of the Shah model are implemented and compared with the experimental data.

• Nuclear Engineering and Technology
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• v.52 no.4
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• pp.755-763
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• 2020

The critical flow phenomenon has been studied because of its significant effect for design basis accidents in nuclear power plants. Transition points from thermal non-equilibrium to equilibrium are different according to the geometric effect on the critical flow. This study evaluates the uncertainty parameters of the critical flow model for analysis of DBA (Design Basis Accident) with the MARS-KS (Multi-dimensional Analysis for Reactor Safety-KINS Standard) code used as an independent regulatory assessment. The uncertainty of the critical flow model is represented by three parameters including the thermal non-equilibrium factor, discharge coefficient, and length to diameter (L/D) ratio, and their ranges are determined using large-scale Marviken test data. The uncertainty range of the thermal non-equilibrium factor is updated by the MCDA (Model Calibration through Data Assimilation) method. The updated uncertainty range is confirmed using an LBLOCA (Large Break Loss of Coolant Accident) experiment in the LOFT (Loss of Fluid Test) facility. The uncertainty ranges are also used to calculate an LBLOCA of the APR (Advanced Power Reactor) 1400 NPP (Nuclear Power Plants), focusing on the effect of the PCT (Peak Cladding Temperature). The results reveal that break flow is strongly dependent on the degree of the thermal non-equilibrium state in a ruptured pipe with a small L/D ratio. Moreover, this study provides the method to handle the thermal non-equilibrium factor, discharge coefficient, and length to diameter (L/D) ratio in the system code.

• Journal of Advanced Marine Engineering and Technology
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• v.25 no.6
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• pp.1341-1352
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• 2001

Vacuum freeze drying process by which frozen water in a drying material is removed sublimation under vacuum condition, is now applied to various industrial field such as the manufacturing and packaging of pharmaceuticals in pharmaceutical industry, the drying of bio- products in bio-technology industry, the treatment of various quality food stuff in food technology, and so on. The Knowledge about the heat and mass transfer characteristics related with the vacuum freeze drying process is crucial to improve the efficiency of the process as well as the quality of dried products. In spite of increasing needs for understanding of the process, the research efforts in this fields are still insufficient. In this paper, a numerical code that can predict primary drying in a vial is developed based on the finite volume method with a moving grid system. The calculation program can handle the axis- symmetric and multi-dimensional characteristics of heat and mass transfer of the vial freeze drying process. To demonstrated the usefulness of the present analysis, a practical freeze drying of skim Milk solution in a vial is simulated and various calculation results are presented.

• Ocean Systems Engineering
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• v.8 no.3
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• pp.329-343
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• 2018

The inclination of seabed profile (sloped seabed) is one of the known topographic features which can be observed at different seabed level in the large offshore basin. A mooring system connected between the platform and global seabed is an integral part of the floating structure which tries to keep the floating platform settled in its own position against hostile sea environment. This paper deals with an investigation of the motion responses of an FPSO platform moored on the sloped seabed under the combined action of wave, wind and current loads. A three-dimensional panel discretization method has been used to model the floating body. To introduce the connection of multi-segmented non-linear elastic catenary mooring cables with the sloped seabed, a quasi-static composite catenary model is employed. The model and analysis have been completed by using hydrodynamic diffraction code AQWA. Validation of the numerical model has been successfully carried out with an experimental work published in the latest literature. The analysis procedure in this study has been followed time domain analysis. The study involves an objective oriented investigation on platform motions, in order to identify the effects of the slopped seabed, the action of the wave, wind and current loads and the presence of riser system. In the end, an effective analysis has been performed to identify a stable mooring model in demand of reducing structural responses of the FPSO.

• Journal of the Korean Society for Aeronautical & Space Sciences
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• v.36 no.4
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• pp.321-327
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• 2008

The shock tunnel as a hypersonic ground test facility was designed, constructed and its performance test was conducted to reproduce the high speed flow which the hypersonic propulsion system is encountered. The design points were understood and the conceptual design was completed using the quasi one dimensional operation analysis code. After that, the specific performance and compartment design were completed using CFD simulation as the part analysis. The facility was then constructed according to those design results and the performance test was conducted for various operation conditions. In this paper, we suggested the design method of hypersonic shock tunnel including the conceptual and performance design using theoretical analysis and the quasi 1D Multi-species computational fluid dynamics code.