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REFERENCE LINKING PLATFORM OF KOREA S&T JOURNALS
> Journal Vol & Issue
Nuclear Engineering and Technology
Journal Basic Information
Journal DOI :
Korean Nuclear Society
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Volume & Issues
Volume 25, Issue 4 - Dec 1993
Volume 25, Issue 3 - Sep 1993
Volume 25, Issue 2 - Jun 1993
Volume 25, Issue 1 - Mar 1993
Selecting the target year
Area Effect on Galvanic Corrosion of Condenser Materials with Titanium Tubes in Nuclear Power Plants
Hwang, Seong-Sik ; Kim, Joung-Soo ; Kim, Uh-Chul ;
Nuclear Engineering and Technology, volume 25, issue 4, 1993, Pages 507~514
Titanium tubes have recently been used in condensers of nuclear power plants since titanium has very good corrosion resistance to seawater. However, when it is connected to Cu alloys as tube sheet materials and these Cu alloys are connected to carbon steels as water box materials, it makes significant galvanic corrosion on connected materials. It is expected from electrochemical tests that the corrosion rate of carbon steel will increase when it is galvanically coupled with Ti or Cu in sea water and the corrosion rate of Cu will increase when it is coupled with Ti, if this couple is exposed to sea water for a long time. It is also expected that the surface area ratios, R
(surface area of carbon steel/surface area of Ti) and R
(surface area of carbon steel/surface area of Cu) are very important for the galvanic corrosion of carbon steel and that these should not be kept to low values in order to minimize the galvanic corrosion on the carbon steel of the water box. Immersed galvanic corrosion tests show that the corrosion rate of carbon steel is 4.4 mpy when the ratio of surface area of Fe/ surface area of Al Brass is 1 while it is 570 mpy when this ratio is 10
. The galvanic corrosion rate of this carbon steel is increased from 4.4 mpy to 13 mpy at this area ratio, 1, when this connected galvanic specimen is galvanically coupled with a Ti tube. This can be rationalized by the combined effects of R
on the polarization curve.
Design Enhancements of Automatic Depressurization System in a Passive PWR
Yu, Sung-Sik ; Seong, Poong-Hyun ;
Nuclear Engineering and Technology, volume 25, issue 4, 1993, Pages 515~528
In a Passive PWR, the successful actuation of Automatic Depressurization System (ADS) is essentially required so that no core damage is occurred following small LOCA. But it has been shown in the previous studies that Core Damage Frequency (CDF) from small LOCA is significantly caused by unavailability of ADS. In this study, the design vulnerabilities impacting the ADS unavailability have been identified and the design improvement items have been proposed through the system reliability assessment using the fault tree methodology The impacts on CDF according to the change of system unavailability have also been analyzed. In addition, small LOCA simulation using RELAP5/MOD3 code has been performed to show the thermal-hydraulic feasibility of the suggested design enhancements.
A Nuclide Transport Model in the Fractured Rock Medium Using a Continuous Time Markov Process
Lee, Y.M. ; C.H. Kang ; P.S. Hahn ; Park, H.H. ; Lee, K.J. ;
Nuclear Engineering and Technology, volume 25, issue 4, 1993, Pages 529~538
A stochastic way using continuous time Markov process is presented to model the one-dimensional nuclide transport in fractured rock matrix as an extended study for previous work ［1］. A nuclide migration model by the continuous time Markov process for single planar fractured rock matrix, which is considered as a transient system where a process by which the nuclide is diffused into the rock matrix from the fracture may be no more time homogeneous, is compared with a conventional deterministic analytical solution. The primary desired quantities from a stochastic model are the expected values and variance of the state variables as a function of time. The time-dependent probability distributions of nuclides are presented for each discretized compartment of the medium given intensities of transition. Since this model is discrete in medium space, parameters which affect nuclide transport could be easily incorporated for such heterogeneous media as the fractured rock matrix and the layered porous media. Even though the model developed in this study was shown to be sensitive to the number of discretized compartment showing numerical dispersion as the number of compartments are decreased, with small compensating of dispersion coefficient, the model agrees well to analytical solution.
Continuous Time Markov Process Model for Nuclide Decay Chain Transport in the Fractured Rock Medium
Lee, Y.M. ; C.H. Kang ; P.S. Hahn ; Park, H.H. ; Lee, K.J. ;
Nuclear Engineering and Technology, volume 25, issue 4, 1993, Pages 539~547
A stochastic approach using continuous time Markov process is presented to model the one-dimensional nuclide transport in fractured rock media as a further extension for previous works［1-3］. Nuclide transport of decay chain of arbitrary length in the single planar fractured rock media in the vicinity of the radioactive waste repository is modeled using a continuous time Markov process. While most of analytical solutions for nuclide transport of decay chain deal with the limited length of decay chain, do not consider the case of having rock matrix diffusion, and have very complicated solution form, the present model offers rather a simplified solution in the form of expectance and its variance resulted from a stochastic modeling. As another deterministic way, even numerical models of decay chain transport, in most cases, show very complicated procedure to get the solution and large discrepancy for the exact solution as opposed to the stochastic model developed in this study. To demonstrate the use of the present model and to verify the model by comparing with the deterministic model, a specific illustration was made for the transport of a chain of three member in single fractured rock medium with constant groundwater flow rate in the fracture, which ignores the rock matrix diffusion and shows good capability to model the fractured media around the repository.
Uranium Fluorescence Analysis in the Raffinate Solution of Nuclear Fuel Conversion Process Using Time-resolved Laser-induced Fluorimetry
Lee, Sang-Mock ; Kim, Duk-Hyeon ; Shin, Jang-Soo ;
Nuclear Engineering and Technology, volume 25, issue 4, 1993, Pages 548~551
A simple and new uranium analysis technique for raffinate solution of nuclear fuel conversion process was developed using a time-resolved laser-induced fluorimetry. The addition of 4 M-phosphoric acid more than 10 times in volume to the raffinate sample was found to be efficient for obtaining stable uranium fluorescence signal which was not influenced by many fluorescence quenchers. A calibration curve of a good linearity for the fluorescence intensity vs. the uranium concentration was obtained at the range of 3.0
in the raffinate samples.
A Study on the Fuel Assembly Stress Analysis for Seismic and Blowdown Events
Kim, Il-Kon ;
Nuclear Engineering and Technology, volume 25, issue 4, 1993, Pages 552~560
In this study, the detailed fuel assembly stress analysis model to evaluate the structural integrity for seismic and blowdown accidents is developed. For this purpose, as the first step, the program MAIN which identifies the worst bending mode shaped fuel assembly(FA) in core model is made. And the finite element model for stress calculation of FA components is developed. In the model the fuel rods (FRs) and the guide thimbles are modelled by 3-dimensional beam elements, and the spacer grid spring is modelled by a linear and relational spring. The constraints come from the results of the program MAIN. The stress analysis of the 16
16 type FA under arbitary seismic load is performed using the developed program and modelling technique as an example. The developed stress model is helpful for the stress calculation of FA components for seismic and blowdown loads to evaluate the structural integrity of FA.
Radiation Shielding Calculation on Shield System of CANDU 6 Plant Using the Coupled DOT4.2 and QAD-CG Codes
Kim, Kyo-Youn ; Kim, Jong-Kyung ;
Nuclear Engineering and Technology, volume 25, issue 4, 1993, Pages 561~569
DOT4.2-QAD-CG coupling method was used to analyze the dose rates outside the side and the bottom shield system of CANDU 6 plant. The average dose rates at the main airlock and the new fuel loading area are approximately 6
Sv/h as it is required. The calculated dose rates have a good agreement with the measurements at the operating CANDU 6 plant. The method used in this paper can be applied to the radiation shielding analysis of Wolsong 2, 3, and 4 CANDU 6 type plants which will be constructed in the near future.
Internal Hydriding of Defected Zircaloy Cladding Fuel Rods : A Review
Kim, Yongsoo ; Donald R. Olander ; Wonmok Jae ;
Nuclear Engineering and Technology, volume 25, issue 4, 1993, Pages 570~587
Recently a number of severe fuel degradation events, seemingly due to internal secondary hydriding, have been reported. This paper reviews internal hydriding of defected zircaloy cladding. First, the history of zircaloy cladding development and the environment of the zircaloys in service in the nuclear reactor are introduced. Fundamental aspects of zircaloy hydriding, such as hydrogen permeability in zirconium oxide, terminal solubility and precipitation in zirconium and its alloys, and the deleterious effect of hydrides are reviewed. The mechanism of massive internal hydriding in defected zircaloy fuel rods is qualitatively described based on the observed phenomena. Significant factors affecting the hydriding process are discussed. A quantitative model for the massive hydriding as a part of an effort to mitigate fuel degradation is briefly mentioned and necessary information and recommended future work for improvement of the model are outlined.