• Transactions of the Korean Society of Mechanical Engineers B
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• v.29 no.3 s.234
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• pp.340-348
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• 2005

Repeated ribs are used on heat exchange surfaces to promote turbulence and enhance convective heat transfer. Applications include fuel rods of gas-cooled nuclear reactors, inside cavities of turbine blades, and internal surfaces pipes used in heat exchangers. Despite the great number of literature papers, only few experimental data concern detailed distributions of friction factors and heat transfer coefficients in square channels varying the number of rough walls. This issue is tackled by investigating effects of different number of ribbed walls on heat transfer and friction characteristics in square channel. The rough wall have a $45{\circ}C$ inclined square rib. Uniform heat flux is maintained on whole inner heat transfer channel area. The heat transfer coefficient and friction factor values increase with increasing the number of rough walls.

• Nuclear Engineering and Technology
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• v.27 no.4
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• pp.518-531
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• 1995

A study has been Peformed to investigate the thermal margin increase by replacing the single-channel analysis model with a multichannel analysis model. h new critical heat flux(CHF) correlation, which is applicable to a 17$\times$17 Korean Fuel Assembly(KOFA)-loaded core, was developed on the basis of the local conditions predicted by the subchannel analysis code, TORC. The hot sub-channel analysis was carried out by using one-stage analysis methodology with a prescribed nodal layout of the core. The result of the analysis shooed that more than 5％ of the thermal margin can be recovered by introducing the TORC/KRB-1 system(multichannel analysis model) instead of the PUMA/ERB-2 system(single-channel anal)sis model). The thermal margin increase was attributed not only to the effect of the local thermal hydraulic conditions in the hot subchannel predicted by the code, but also to the effect of the characteristics of the CHF correlation.

• Nuclear Engineering and Technology
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• v.48 no.4
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• pp.847-858
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• 2016

An overview is given on the work of the Laboratory of Nuclear Energy Systems at ETH, Zurich (ETHZ) and of the Laboratory of Thermal Hydraulics at Paul Scherrer Institute (PSI), Switzerland on tight-lattice bundles. Two-phase flow in subchannels of a tight triangular lattice was studied experimentally and by computational fluid dynamics simulations. Two adiabatic facilities were used: (1) a vertical channel modeling a pair of neighboring sub-channels; and (2) an arrangement of four subchannels with one subchannel in the center. The first geometry was equipped with two electrical film sensors placed on opposing rod surfaces forming the subchannel gap. They recorded 2D liquid film thickness distributions on a domain of $16{\times}64$ measuring points each, with a time resolution of 10 kHz. In the bubbly and slug flow regime, information on the bubble size, shape, and velocity and the residual liquid film thickness underneath the bubbles were obtained. The second channel was investigated using cold neutron tomography, which allowed the measurement of average liquid film profiles showing the effect of spacer grids with vanes. The results were reproduced by large eddy simulation + volume of fluid. In the outlook, a novel nonadiabatic subchannel experiment is introduced that can be driven to steady-state dryout. A refrigerant is heated by a heavy water circuit, which allows the application of cold neutron tomography.

• Nuclear Engineering and Technology
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• v.52 no.4
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• pp.661-673
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• 2020

The quarter-core simulation of BEAVRS Cycle 2 depletion benchmark has been conducted using the MCS/CTF coupling system. MCS/CTF is a cycle-wise Picard iteration based inner-coupling code system, which couples sub-channel T/H (thermal/hydraulic) code CTF as a T/H solver in Monte Carlo neutron transport code MCS. This coupling code system has been previously applied in the BEAVRS benchmark Cycle 1 full-core simulation. The Cycle 2 depletion has been performed with T/H feedback based on the spent fuel materials composition pre-generated by the Cycle 1 depletion simulation using refueling capability of MCS code. Meanwhile, the MCS internal one-dimension T/H solver (MCS/TH1D) has been also applied in the simulation as the reference. In this paper, an analysis of the detailed criticality boron concentration and the axially integrated assembly-wise detector signals will be presented and compared with measured data based on the real operating physical conditions. Moreover, the MCS/CTF simulated results for neutronics and T/H parameters will be also compared to MCS/TH1D to figure out their difference, which proves the practical application of MCS into the BEAVRS benchmark two-cycle depletion simulations.

• Nuclear Engineering and Technology
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• v.23 no.3
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• pp.328-339
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• 1991

A multi-purpose research reactor called KMRR has been developed by Korea Atomic Energy Research Institute(KAERI) to generate a maximum thermal output of 30 MW. As a part of thermal hydraulics study, pressure drop characteristics of the longitudinally finned fuel rod bundles were experimentally investigated in a recirculating water test loop. The present study is focused on the investigation of fin effects on pressure drop and the development of pressure drop correlation for the finned rod bundles in a wide range of flow conditions. Friction factor correlations for each design of the finned rod bundles are developed. The value of friction factor for the finned rod bundles was higher than the analytical solution (64/Re) of laminar circular channel new but became lower than the Blasius equation as Reynolds number was increased.

• Nuclear Engineering and Technology
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• v.36 no.1
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• pp.36-46
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• 2004

The ISAAC fission product release calculation is based on detailed FPRAT models developed by Jaycor. For volatile fission product release calculations, either the Cubicciotti steam oxidation correlation or the NUREG-0772 correlation is used. In this study, evaluation is carried out for these volatile fission product release models. As a result, in the case of early release, the IDCOR model with an in-vessel Te release option shows the most conservative results and for the late release case, the NUREG-0772 model shows the most conservative results. Considering both early and late release, the IDCOR model with an in-vessel Te bound option is evaluated to show mitigated conservative results. In addition, a sensitivity study on detailed core nodalization is performed. In the study, 380 horizontal fuel channels in the Wolsong plant are nodalized into 12 (6 channels per loop, $3{\times}3$ Core Pass) representative channels and detailed by 16/20/24 channels. For reference accidents, LOAH and large LOCA are selected as representing high and low pressure sequences, respectively. According to the results, the original 12 channel approach with $3{\times}3$ core passes is evaluated to be sufficient as an optimal scheme.

• Journal of the Korean Society for Nondestructive Testing
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• v.22 no.1
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• pp.59-64
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• 2002

A FEM model of the remote-field eddy current effect is presented for zirconium-2.5 percent niobium(Zr-2.5%Nb) nuclear reactor pressure tubes to demonstrate the important electromagnetic field phenomena. This model is applied to evaluate the optimal operating frequency and detector position. There are many ambiguous experimental results connected with this technique. Finite element calculations can be used in the interpretation of these experimental results even though the electromagnetic fields measured in the remote-field technique are very small.

• Proceedings of the KIEE Conference
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• 2001.07b
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• pp.690-692
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• 2001

A FEM model of the remote-field eddy current effect is presented for zirconium-2.5percent niobium(Zr-2.5%Nb) nuclear reactor pressure tubes to demonstrate the important electromagnetic field. Phenomena that describe this effect. This model is applied to evaluate the optimal operating frequency and detector position. There are many ambiguous experimental results connected with this technique. Finite element calculations can be used in the interpretation of these experimental results even though the electromagnetic fields measured in the remote-field technique are very small.

• Bulletin of the Korean Chemical Society
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• v.34 no.2
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• pp.510-514
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• 2013

The structural change of Nafion polymer electrolyte membrane (PEM) induced by hot-pressing, which is one of the representative procedures for preparing membrane-electrode-assembly for low temperature fuel cells, was investigated by $^2H$ nuclear magnetic resonance (NMR) spectroscopy. The hydrophilic channels were asymmetrically flattened and more aligned in the membrane plane than along the hot-pressing direction. The average O-$^2H$ director of $^2H_2O$ in polymer electrolyte membrane was employed to extract the structural information from the $^2H$ NMR peak splitting data. The dependence of $^2H$ NMR data on water contents was systematically analyzed for the first time. The approach presented here can be used to understand the chemicals' behavior in nano-spaces, especially those reshaping and functioning interactively with the chemicals in the wet and/or mixed state.

• Journal of Radiation Protection and Research
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• v.39 no.4
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• pp.176-181
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• 2014

The object of this paper is to evaluate the fission product inventories and radiological doses in a non-LOCA event, based on the U.S. NRC's regulatory methodologies recommended by the TID-14844 and the RG 1.195. For choosing a non-LOCA event, one fuel assembly was assumed to be melted by a channel blockage accident. The Hanul nuclear power reactor unit 6 and the CE $16{\times}16$ fuel assembly were selected as the computational models. The burnup cross section library for depletion calculations was produced using the TRITON module in the SCALE6.1 computer code system. Based on the recently licensed values for fuel enrichment and burnup, the source term calculation was performed using the ORIGEN-ARP module. The fission product inventories released into the environment were obtained with the assumptions of the TID-14844 and the RG 1.195. With two kinds of source terms, the radiological doses of public in normal environment reflecting realistic circumstances were evaluated by applying the average condition of meteorology, inhalation rate, and shielding factor. The statistical analysis was first carried out using consecutive three year-meteorological data measured at the Hanul site. The annual-averaged atmospheric dispersion factors were evaluated at the shortest representative distance of 1,000 m, where the residents are actually able to live from the reactor core, according to the methodology recommended by the RG 1.111. The Korean characteristic-inhalation rate and shielding factor of a building were considered for a series of dose calculations.