• Nuclear Engineering and Technology
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• v.17 no.4
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• pp.279-289
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• 1985

The purpose of this paper is to investigate the density-wave oscillation type instability in the recirculating loop of U-tube steam generator (UTSG). The perturbed and nodalized conservations equations based on the drift-flux model have been derived to obtain the single-and two-phase pressure drop perturbations, by taking into account the slip between phases, nonuniform heat flux and heated wall dynamics. To assess the stability, the frequency domain technique with the Nyquist criterion has been used under the constant pressure drop boundary condition through the loop. The computer implementation of this model, SASG, was used for the parametric study of the steam generator in Kori-Unit 1. The results of the parametric study revealed important factors influencing UTSG stability margin.

• Nuclear Engineering and Technology
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• v.54 no.12
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• pp.4776-4797
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• 2022

In the one-dimensional thermal-hydraulic (TH) codes, a subcooled boiling model to predict the void fraction profiles in a vertical channel consists of wall heat flux partitioning, the vapor condensation rate, the bubbly-to-slug flow transition criterion, and drift-flux models. Model performance has been investigated in detail, and necessary refinements have been incorporated into the Safety and Performance Analysis Code (SPACE) developed by the Korean nuclear industry for the safety analysis of pressurized water reactors (PWRs). The necessary refinements to models related to pumping factor, net vapor generation (NVG), vapor condensation, and drift-flux velocity were investigated in this study. In particular, a new NVG empirical correlation was also developed using artificial neural network (ANN) techniques. Simulations of a series of subcooled flow boiling experiments at pressures ranging from 1 to 149.9 bar were performed with the refined SPACE code, and reasonable agreement with the experimental data for the void fraction in the vertical channel was obtained. From the root-mean-square (RMS) error analysis for the predicted void fraction in the subcooled boiling region, the results with the refined SPACE code produce the best predictions for the entire pressure range compared to those using the original SPACE and RELAP5 codes.

• Nuclear Engineering and Technology
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• v.26 no.4
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• pp.562-569
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• 1994

In this study improvement of transient analysis model, KOTRAC, for the passive reactor has been performed. In the KOTRAC, mixture drift flux model is adopted to simulate thermal hydraulic behavior, which can simulate ECCS injection in the passive plant. However, there is a difficulty to handle complete phase separation phenomena due to the near-zero density, which may occur in the pressurizer surge line or horizontal flow paths. In this study, a couple of model changes to over-come Courant limit feilure has been examined. One of key features is to substitute flow distribution parameters with Ishii's correlation. Corrected results are nil compared to those of RELAP/MOD3 analysis.

• Journal of Astronomy and Space Sciences
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• v.39 no.4
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• pp.145-158
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• 2022

The dynamics of the outer zone radiation belt has received a lot of attention mainly due to the correlation between the occurrence of enhancing relativistic electron flux and spacecraft operation anomalies or even failures (e.g., Baker et al. 1994). Relativistic electron events are often observed during great storms associated with ultra low frequency (ULF) waves. For example, a large buildup of relativistic electrons was observed during the great storm of March 24, 1991 (e.g., Li et al. 1993; Hudson et al. 1995; Mann et al. 2013). However, the dominant processes which accelerate magnetospheric radiation belt electrons to MeV energies are not well understood. In this paper, we present observations of Pc5 ULF waves in the recovery phase of the Bastille day storm of July 16, 2000 and electron and proton flux simultaneously oscillating with the same frequencies as the waves. The mechanism for the observed electron and proton flux modulations is examined using ground-based and satellite observations. During this storm time, multiple packets of discrete frequency Pc5 ULF waves appeared associated with energetic particle flux oscillations. We model the drift paths of electrons and protons to determine if the particles drift through the ULF wave to understand why some particle fluxes are modulated by the ULF waves and others are not. We also analyze the flux oscillations of electrons and protons as a function of energy to determine if the particle modulations are caused by a ULF wave drift resonance or advection of a particle density gradient. We suggest that the energetic electron and proton modulations by Pc5 ULF waves provide further evidence in support of the important role that ULF waves play in outer radiation belt dyanamics during storm times.

• Nuclear Engineering and Technology
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• v.19 no.3
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• pp.198-204
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• 1987

A plant simulation code, MCSIM (Micro-Computer SIMulator), has been developed to simulate plant transient accidents for pressurized water reactors. Reactor coolant system is modeled using decoupled energy and momentum equations, drift flux two-phase flow model and integral momentum equation. A two-fluid pressurizer model is used to simulate the pressurizer dynamics. Pot Boiler model is used for steam generator, steady-state decoupled energy and momentum equations for secondary side system, and point kinetics equations for nuclear power calculation. For test of the present version of MCSIM, complete loss of flow and RCCA withdrawal accidents are calculated with MCSIM. The results are compared with those in FSAR of KNU 5 & 6.

• Nuclear Engineering and Technology
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• v.45 no.4
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• pp.439-458
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• 2013

REX-10 is a fully-passive small modular reactor in which the coolant flow is driven by natural circulation, the RCS is pressurized by a steam-gas pressurizer, and the decay heat is removed by the PRHRS. To confirm design decisions and analyze the transient responses of an integral PWR such as REX-10, a thermal-hydraulic system code named TAPINS (Thermal-hydraulic Analysis Program for INtegral reactor System) is developed in this study. Based on a one-dimensional four-equation drift-flux model, TAPINS incorporates mathematical models for the core, the helical-coil steam generator, and the steam-gas pressurizer. The system of difference equations derived from the semi-implicit finite-difference scheme is numerically solved by the Newton Block Gauss Seidel (NBGS) method. TAPINS is characterized by applicability to transients with non-equilibrium effects, better prediction of the transient behavior of a pressurizer containing non-condensable gas, and code assessment by using the experimental data from the autonomous integral effect tests in the RTF (REX-10 Test Facility). Details on the hydrodynamic models as well as a part of validation results that reveal the features of TAPINS are presented in this paper.

• Nuclear Engineering and Technology
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• v.56 no.1
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• pp.19-33
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• 2024

In nuclear thermal-hydraulic system codes, most correlations used for vertical pipes, under downward two-phase flow, have been developed considering small pipes or pool systems. This suggests that there could be uncertainties in applying the correlations to accident scenarios involving large vertical pipes owing to the difference in the characteristics of two-phase flows, or flow conditions, between large and small pipes. In this study, we modified the Multi-dimensional Analysis of Reactor Safety KINS Standard (MARS-KS) code using correlations, such as the drift-flux model and two-phase multiplier, developed in a plant-scale air-inflow experiment conducted for a pipe of diameter 600 mm under downward two-phase flow. The results were then analyzed and compared with those based on previous correlations developed for small pipes and pool conditions. The modified code indicated a good estimation performance in two plant-scale experiments with large pipes. For the siphon-breaking experiment, the maximum errors in water flow for modified and original codes were 2.2% and 30.3%, respectively. For the air-inflow accident experiment, the original code could not predict the trend of frictional pressure gradient in two-phase flow as / increased, while the modified MARS-KS code showed a good estimation performance of the gradient with maximum error of 3.5%.

• Journal of the Society of Naval Architects of Korea
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• v.38 no.3
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• pp.1-13
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• 2001

RANS solver has been developed to solve the flows past a ship with a drift angle. The solver employs a finite volume method for the spatial discretization and Euler implicit method for the time integration. Turbulent flows are simulated by Spalart-Allmaras one-equation model. Developed solver is applied to analyze the hydrodynamic forces and flows of two tankers with a same forebody but different afterbodies. The computed flows and hydrodynamic forces are compared with the measured flows and captive model test data. The computed results show good agreements with experimental data and show clearly the effects of stern hull form on the hydrodynamic forces and the flows.

• Transactions of the Korean Society of Mechanical Engineers
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• v.14 no.6
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• pp.1629-1638
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• 1990

Using a 1/10 scale model pump, which is designed and manufactured to simulate two phase performance of reactor coolant pump of Y.G.N

• Nuclear Engineering and Technology
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• v.22 no.1
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• pp.66-74
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• 1990

The thermal-hydraulic characteristics of the 17$\times$17 OFA (Optimized Fuel Assembly) used in the KNU 7&8 are analyzed and compared with that of the 17$\times$17 SFA (Standard Fuel Assembly) loaded in the KNU 5&6. The thermal-hydraulic characteristics analyzed are minimum DNBR, fuel centerline temperature and exit void fraction at normal operation and design over power transient. Additionally, local linear rod power, which will cause fuel centerline melting, is calculated. The DNBR sensitivity calculations are performed with respect to the reactor operating parameters. COBRA-IV-I code is used for these calculations. The modified W-3 correltion and the drift-flux model are applied for the critical heat flux calculation and the void fraction calculation, respectively. From the calculated results, it has been found that the possibility of DNB occurrence is higher in the OFA than in the SFA. The other hand, the local linear power resulting in fuel centerline moiling of the OFA is nearly equal to that of the SFA.