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REFERENCE LINKING PLATFORM OF KOREA S&T JOURNALS
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Nuclear Engineering and Technology
Journal Basic Information
Journal DOI :
Korean Nuclear Society
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Volume & Issues
Volume 30, Issue 6 - Dec 1998
Volume 30, Issue 5 - Oct 1998
Volume 30, Issue 4 - Aug 1998
Volume 30, Issue 3 - Jun 1998
Volume 30, Issue 2 - Apr 1998
Volume 30, Issue 1 - Feb 1998
Selecting the target year
A Feasibility Study of Seismic Isolation for Wolsong Reactor Building
Kim, Kang-Soo ; Kim, Tae-Wan ; Lee, Jeong-Yoon ;
Nuclear Engineering and Technology, volume 30, issue 2, 1998, Pages 83~90
To predict effects of seismic isolation, seismic isolation bearings were applied to the Wolsong reactor building and the analytical study was performed. For this study, the Wolsong reactor building was modeled using lumped masses and beam elements. Design Basis Earthquake with a ground acceleration of 0.2g was applied. And then, the behavior of the isolated structure was compared with that of the unisolated structure. The horizontal response acceleration at the top of the unisolated reactor building was 0.99g, while that of the isolated one was 0.14g(15% damping) and the acceleration response along the height of the structure was constant. The maximum displacement of the unisolated structure was 8.3mm, while that of the isolated structure was 66mm. The application of isolation bearings on the reactor building reduces seismic loads but increases the displacement of the structure on a large scale. Therefore, when using isolation bearings, the reactor building and BOP should be located on a common mat to cover the large displcement.
Three-Dimensional Seismic Analysis for Spent Fuel Storage Rack
Lee, Gyu-Mahn ; Kim, Kang-Soo ; Park, Keun-Bae ; Park, Jong-Kyun ;
Nuclear Engineering and Technology, volume 30, issue 2, 1998, Pages 91~98
Time history analysis is usually performed to characterize the nonlinear seismic behavior of a spent fuel storage rack(SFSR). In the past, the seismic analyses of the SFSR were performed with two-dimensional planar models, which could not account for torsional response and simultaneous multi-directional seismic input In this study, three-dimensional seismic analysis methodology is developed for the single SFSR using the ANSYS code. The 3D- Model can be used to determine the nonlinear behavior of the rack, i.e., sliding, uplifting, and impact evaluation between the fuel assembly and rack, and rack and the pool wall, This paper also reviews the 3-D modeling of the SFSR and the adequacy of the ANSYS for the seismic analysis. AS a result of the adquacy study, the method of ANSYS transient analysis with acceleration time history is suitable for the seismic analysis of highly nonlinear structure such as an SFSR but it isn't appropriate to use displacement time history of seismic input.
Microstructural Properties of the Insoluble Residue in a Simulated Spent Fuel
Kim, J. S. ; B. C. Song ; K. Y. Jee ; Kim, J. G. ; K. S. Chun ;
Nuclear Engineering and Technology, volume 30, issue 2, 1998, Pages 99~111
Chemical composition of the insoluble residue in a simulated spent PWR fuel(SIMRJEL) were studied. SIMFUELS were prepared by adding calculated amount of FP(fission product) elements with a burnup of 3.6% FIMA(fission per initial metal atom) to uranium in nitrate solution, evaporating the mixed solution to dryness, calcining at 90
in a stream of 4% H
+ 96% He, and heating the pellet at 140
under high and low oxygen potentials. Insoluble residue was obtained from the dissolution of the SIMFUEL with HNO
(1 : 1). The chemical composition of the SIMFUELs and the insoluble residues was determined by EPMA(electron probe microanalysis), XPS(X-ray photoelectron spectroscopy) and by XRD (X-ray diffraction) measurements. All of the insoluble residues suspended and precipitated were composed mainly of Mo, Ru with a small amount of Zr, Rh, Pd and Cd. The amount of insoluble residue(<1 wt.%) and a Mo/Ru ratio decreased with increasing oxygen potential. Formation of the zirconium molybdate precipitate, ZrMo
, was observed in the residues. The possible role of Mo on the phase formation was discussed in regard to oxygen potential.l.
Analysis of Reflux Cooling in the SG U-Tubes Under Loss of RHRS During Midloop Operation with Primary System Partly Open
Son, Young-Seok ; Kim, Won-Seok ; Kim, Kyung-Doo ; Chung, Young-Jong ; Chang, Won-Pyo ;
Nuclear Engineering and Technology, volume 30, issue 2, 1998, Pages 112~127
The present study is to assess the applicability of the best-estimate thermal-hydraulic codes, RELAP5/MOD3.2 and CATHARE2V1.3U, to the analysis of thermal-hydraulic behavior in PWRs during midloop operation following the loss of RHRS. The codes simulate an integral test, BETHSY 6.94, which was conducted in the large scale test facility of BETHSY in France. The test represents the accident where the loss of RHRS occurs during midloop operation with the pressurizer and upper head vents open and the sight level indicator broken. Besides, the hot legs are half filled with water and the upper parts of the primary cooling system are filled with nitrogen, with a letdown line open and only one SG available. The purposes of this study are to understand the physical phenomena associated with reflux cooling in the 5G U-tubes when noncondensable gas is present under low pressure and to assess the applicability of the codes to simulate the loss of RHRS event by comparing the predictions with the test results. The results of the study may contribute to actual applications for plant safety evaluation and description of the emergency operating procedure.
Powder Property and Oxygen Potential on Sintering Characteristics of
Song, Kun-Woo ; Kim, Keon-Sik ; Yoo, Ho-Sik ; Jung, Youn-Ho ;
Nuclear Engineering and Technology, volume 30, issue 2, 1998, Pages 128~139
The effect of UO
powder property and oxygen potential on characteristics of sintered UO
fuel pellets has been investigated. Two types of powder, mixture of AUC-UO
powders (type I) and mixture of ADU-UO
powders (type II), have been prepared, pressed, and sintered at 168
for 4 hours. Four sintering atmospheres with different mixing ratios of
gas ranging from 0 to 0.3 have been used. UO
fuel has lower sintered density than UO
fuel, and the density drop is larger for powder type I than for powder type II. As the oxygen potential increases, the sintered density of UO
pellets increases but that of UO
pellets decreases. It is found that pores are newly formed in UO
pellets in accordance with the decrease in density. The grain size of UO
fuel increases and a short range G4 distribution becomes homogeneous as the oxygen potential increases. A long range ed distribution and grain structure are inhomogeneous for powder type II. The lattice parameter of (U,Gd)O
solid solution decreases linearly with Gd
content. The dependence of UO
fuel characteristics on powder type and sintering atmosphere have been discussed.
Conceptual Study for the Moderator Selection of the Cold Neutron Source Facility for HANARO
Cho, Young-Sik ; Jonghwa Chang ; Park, Chang-Oong ;
Nuclear Engineering and Technology, volume 30, issue 2, 1998, Pages 140~147
Basic concept of a cold neutron source for a 30 MW heavy water moderated reactor (HANARO) is developed. The source is a cold bottle located in a vertical hole near the reactor core. Since the bottle does not have sufficient volume for cooling, the optimum liquid mixture ratio is studied between liquid hydrogen and liquid deuterium. We also studied the variation of the gain depending on the volume of the bottle. The calculation is performed by a coupled MCNP model and by a semi-analytic approach. For the current geometry, 80% liquid deuterium mixture with liquid hydrogen gives the highest gain at 10 A neutron wave.
Flow Induced Material Degradation In Power Plant Secondary Systems-A Review
Kim, I. S. ; M. Van Der Helm ; R. G. Ballinger ;
Nuclear Engineering and Technology, volume 30, issue 2, 1998, Pages 148~163
Flow Induced Material Degradation (FIMD) is reviewed focusing on Flow Accelerated Corrosion (FAC) models. Several examples of FAC related incidents are described, which include nuclear and fossile power plants. Lastly, mitigation techniques such as inspection, material selection, water chemistry, temperature, and hydrodynamic factor are discussed.
Modelling of RV Ledge Region for Dynamic Analysis of Coupled Reactor Vessel Internals and Core
Jhung, Myung J. ;
Nuclear Engineering and Technology, volume 30, issue 2, 1998, Pages 164~172
This paper presents the detailed modelling of reactor vessel ledge region for the dynamic analysis of the coupled internals and core model. The dynamic responses due to earthquake and pipe break are calculated using the input motions of reactor vessel taken from Ulchin nuclear power plant units 3 and 4. Two different representations for detailed and simplified models of the RV ledge region are made. The dynamic responses of the reactor internals components are compared between them. Response characteristics are reported and simplified model is suggested for earthquake and pipe break analysis for the future design of the reactor internals.
Analysis of Corrosion Behavior of KOFA Zircaloy-4 Cladding
Lee, Chan-Bock ; Kim, Ki-Hang ;
Nuclear Engineering and Technology, volume 30, issue 2, 1998, Pages 173~179
The corrosion behavior of KOFA cladding which is a standard Zircaloy-4 manufactured by Westinghouse Specialty Metal Plant according to the Siemens/KWU's HCW (Highly Cold Worked) standard Zircaloy-4 specification was analyzed using the oxide measurement data of KOFA fuel irradiated in Kori-2 nuclear power plant. Analysis of the measured KOFA cladding oxidation showed that oxidation of KOFA cladding was lower than the design prediction based upon Siemens/KWU's HCW standard Zircaloy-4 cladding. Although the measured fuel rods have relatively low burnup and oxidation and the amount of the measured data are small, analysis of manufacturing and in-reactor operation conditions of KOFA cladding indicates that the differences in the manufacturing processes and chemical composition of the Siemens/KWU's HCW (Highly Cold Worked) standard Zircaloy-4 and KOFA cladding may have somewhat contributed to lower corrosion of KOFA cladding than the expected.