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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 29, Issue 6 - Dec 1997
Volume 29, Issue 5 - Oct 1997
Volume 29, Issue 4 - Aug 1997
Volume 29, Issue 3 - Jun 1997
Volume 29, Issue 2 - Apr 1997
Volume 29, Issue 1 - Feb 1997
Selecting the target year
On the Sampling and Transport of Radioactive Aerosols from Waste Thermal Process
Yang, Hee-Chul ; Kim, Joon-Hyung ; Yong Kang ;
Nuclear Engineering and Technology, volume 29, issue 4, 1997, Pages 269~279
The errors associated with incorrect sampling and transport of radioactive aerosol from radwaste thermal process off-gas are analyzed and the conditions of representative sampling and correct transport of radioactive aerosol for off-gas system evaluation are discussed. An estimation method of sampling errors for individual radionuclides is proposed and applied to simulated vitrification melter aerosols. Prediction methods for particle deposition in sample transport tube under laminar as well as turbulent flow conditions are also described by example calculations with simulated incinerator off-gas From the results of example calculations and plots, instrumental and operational conditions of radioactive aerosol sampling system with minimized errors and correction methods for nonideal sampling and transport are recommended.
Robust Controller Design for the Nuclear Reactor Power Control System
Lee, Yoon-Joon ; Park, Jung-In ;
Nuclear Engineering and Technology, volume 29, issue 4, 1997, Pages 280~290
The robust controller for the nuclear reactor power control system is designed. The nuclear reactor is modeled by use of the point kinetics equations and the singly lumped energy balance equations, Since the model is not exact, the controller which can make the actual system robust is necessary. The perturbed plant is investigated by employing the uncertainties of the initial power level and the physical properties, and by introducing the delay into the modeled plant The overall system is configured into the two port model and the H
controller is designed. In designing the H
controller, two factors of the loop shaping and the permissible magnitude of control input are taken into account The designed controller provides the sufficient margins for the robustness, and the transients of the system output power and the control input satisfy their associated requirement.
A Study of the Evaporation Heat Transfer in Advanced Reactor Containment
Y. M. Kang ; Park, G. C. ;
Nuclear Engineering and Technology, volume 29, issue 4, 1997, Pages 291~298
In advanced nuclear reactors, the passive containment cooling has been suggested to enhance the safety. The passive cooling has two mechanisms, air natural convection and oater cooling with evaporation. To confirm the coolability of PCCS, many works have been performed experimentally and numerically. In this study, the water cooling test was performed to obtain the evaporative heat transfer coefficients in a scaled don segment type PCCS facility which have same configuration with AP600 prototype containment. Air-steam mixture temperature and velocity, relative humidity and well heat flux are measured. The local steam mass flow rates through the vertical plate part of the facility are calculated from the measured data to obtain evaporative heat transfer coefficients. The measured evaporative heat transfer coefficients are compared with an analytical model which use a mass transfer coefficients. From the comparison, the predicted coefficients show good agreement with experimental data however, some discrepancies exist when the effect of wave motion is not considered. Finally, a new correlation on evaporative heat transfer coefficients are developed using the experimental values.
Feasibility Study on the Utilization of Mixed Oxide Fuel in Korean 900MWe PWR Core Through Conceptual Core Nuclear Design and Analysis
Joo, Hyung-Kook ; Kim, Young-Jin ; Jung, Hyung-Guk ; Sohn, Dong-Seong ;
Nuclear Engineering and Technology, volume 29, issue 4, 1997, Pages 299~309
The neutronic feasibility of typical Korean three-loop 900MWe class PWR core loaded with mixed oxide fuels for both annual and 18-month cycle strategies has been investigated as a means for spent fuel management. For this study, a method of determining equivalent plutonium content was developed under the equivalence concept which gives the same cycle length as uranium fuel. Optimal plutonium zoning within the MOX assembly was also designed with the aim of minimizing the peak md power. Conceptual core designs hate hen developed for equilibrium cycle with the following variations: annual and 18-month cycle, 1/3 and full MOX loading schemes, and typical and high moderation lattice. The analysis of key core physics parameters shows that in all cases considered satisfactory core designs seem to be feasible, though addition of control rod system and change in Technical Specification for soluble boron concentration are required for full MOX loading in order to meet the current design requirements.
Stress Index Development for Piping with Trunnion Attachment Under Pressure and Moment Loadings
Lee, Dae-hee ; Kim, Jong-Min ; Park, Sung-ho ;
Nuclear Engineering and Technology, volume 29, issue 4, 1997, Pages 310~319
A finite element analysis of a trunnion pipe anchor is presented. The structure is analyzed for the case of internal pressure and moment loadings. The stress results are categorized into the average (membrane) stress, the linearly varying (bending) stress and the peak stress through the thickness. The resulting stresses are interpreted per Section III of the ASME Boiler and Pressure Vessel Code from which the Primary(B
) and Peak(K
) stress indices for pressure, the Primary (B
) and Peak(K
) stress indices for moment are developed. Based on the comparison between stress value by stress indices derived in this paper and stress value represented by the ASME Code Case N-391-1, the empirical equations for stress indices are effectively used in the piping stress analysis. Therefore, the use of empirical equations can simplify the procedure of evaluating the local stress in the piping design stage.
A Nuclide Decay Chain Transport Model by the Method of Characteristics
Lee, Youn-Myoung ; Kang, Chul-Hyung ; Hahn, Pil-Soo ; Chun, Kwan-Sik ;
Nuclear Engineering and Technology, volume 29, issue 4, 1997, Pages 320~326
The nuclide transport in the one-dimensional porous medium is considered as a first step in developing a decay chain transport in multidimensional inhomogeneous media. A method of solving conventional advection-dispersion equation with decay chain of arbitrary length by using the method of characteristics (MOC) is introduced. In specific cases where the advection are dominant rather than dispersion, the method is known to be useful : one of the most distinctive advantages in applying the model is that the MU minimizes the numerical dispersion, which is distinguished in such common numerical schemes as finite element method and finite difference method. The suggested model is considered to be effective through several illustrations for the case that decay chain of arbitrary length is involved during transport which is difficult to solve by standard numerical solutions if the medium becomes more complicated.
Analysis of the Irradiated Nuclear Fuel Using the Heavy Atom and Neodynium Isotope Correlations with Burnup
Kim, Jung-Suk ;
Nuclear Engineering and Technology, volume 29, issue 4, 1997, Pages 327~335
The correlation of isotope composition of uranium, plutonium and neodymium with the burnup in M uranium dioxide fuel has been investigated experimentally. The total and fractional(
U) burnup were determined by Nd-148 and, U and Pu mass spectrometric method respectively. The isotope compositions of these elements, after their separation from the fuel samples were measured by mass spectrometric. The content of the elements in the irradiated fuel ore determined by isotope dilution mass spectrometric method using
Nd as spikes. The content of plutonium in the irradiated fuel was expressed by the correlation with uranium isotopes. The correlations between isotope compositions themselves and the total and fractional burnup ore compared with those calculated from ORIGEN2 code.
An Experimental Study on the Transient Interaction Between High Temperature Thermite Melt and Concrete
Nho, Ki-Man ; Kim, Jong-Hwan ; Kim, Sang-Baik ; Shin, Ki-Yeol ; Mo Chung ;
Nuclear Engineering and Technology, volume 29, issue 4, 1997, Pages 336~347
During postulated severe accidents in Light water Reactors, molten corium which was ejected from the reactor vessel bottom, may erode the concrete basemat of the containment and there by threaten the containment integrity. This study experimentally examines the molten core-concrete interaction (MCC) using 20kg of thermite melt (Fe +
) and the concrete, used in Yonggwang Nuclear Power Plant Units 3 and 4 (YGN 3 & 4) in Korea. The measured data are the downward heat fluxes, concrete erosion rate, gases and particle generation rates during MCCI. Transient results ore compared with those of TURCIT experiment conducted by SNL in USA. The peak downward heat flux to the concrete was measured to be about 2.1㎿/
. The initial concrete erosion rate was 175cm per hour, decreasing to 30cm per hour. It was shown from the post-test that the erosion was progressed downward up to 18mm in the concrete slug.
Coolant Options and Critical Heat Flux Issues in Fusion Reactor Divertor Design
Baek, Won-Pil ; Chang, Soon-Heung ;
Nuclear Engineering and Technology, volume 29, issue 4, 1997, Pages 348~359
This paper reviews cooling aspects of the diverter system in Tokamak fusion devices with primary emphasis on the critical heat flux (CHF) issues for oater-cooled designs. General characteristics of four (4) coolant options for diverter cooling gases, oater, liquid metal, and organic liquid - are discussed first, focusing on the comparison of advantages and disadvantages of those options. Then results of recent studies on the high-heat-flux CHF of water at subcooled high-velocity conditions are reviewed to provide a general idea on the feasibility of the water-cooled diverter concept for future Tokamak fusion reactors. Water is assessed to be the most viable and practical coolant option for diverters of future experimental Tokamaks.