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REFERENCE LINKING PLATFORM OF KOREA S&T JOURNALS
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Nuclear Engineering and Technology
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Journal DOI :
Korean Nuclear Society
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Volume & Issues
Volume 28, Issue 6 - Dec 1996
Volume 28, Issue 5 - Oct 1996
Volume 28, Issue 4 - Aug 1996
Volume 28, Issue 3 - Jun 1996
Volume 28, Issue 2 - Apr 1996
Volume 28, Issue 1 - Feb 1996
Selecting the target year
Experimental Investigation of the Combined Effects of Heat Exchanger Geometries on Nucleate Pool Boiling Heat Transfer in a Scaled IRWST
Kang, Myeong-Gie ; Chun, Moon-Hyun ;
Nuclear Engineering and Technology, volume 28, issue 1, 1996, Pages 1~16
In an effort to determine the combined effects of major parameters of heat exchanger tubes on the nucleate pool boiling heat transfer in the scaled in-containment refueling water storage tank (IRWST), a total of 1,966 data for q'quot; versus
has been obtained using various combinations of tube diameters, surface roughness, and tube orientations. The experimental results show that (1) increased surface roughness enhances heat transfer for both horizontal and vertical tubes, (2) the two heat transfer mechanisms, i.e.,enhanced heat transfer for both horizontal and vertical tubes, (2) the two heat transfer mechanisms, i.e., enhanced heat transfer due to liquid agitation by bubbles generated and reduced heat transfer by the formation of large vapor slugs and bubble coalescence are different in two regions of low heat fluxes (q'quot;
and high heat fluxes (q'quot;
depending on the orientation of tubes and the degree of surface roughness, and (3) the heat transfer rate decreases as the tube diameter is increased for both horizontal and vertical tubes, but the effect of tube diameter on the nucleate pool boiling heat transfer for vertical tubes is greater than that for horizontal tubes. Two empirical heat transfer correlations for q'quot;, one for horizontal tubes and the other for vertical tubes, are obtained in terms of surface roughness
and tube diameter (D). In addition, a simple empirical correlation for nucleate pool boiling heat transfer coefficient
is obtained as a function of heat flux (q'quot;) only.ucleate pool boiling heat transfer coefficient
is obtained as a function of heat flux (q'quot;) only.
Development of A Methodology for In-Reactor Fuel Rod Supporting Condition Prediction
Kim, K. T. ; Kim, H. K. ; K. H. Yoon ;
Nuclear Engineering and Technology, volume 28, issue 1, 1996, Pages 17~26
The in-reactor fuel rod support conditions against the fretting wear-induced damage can be evaluated by residual spacer grid spring deflection or rod-to-grid gap. In order to evaluate the impact of fuel design parameters on the fretting wear-induced damage, a simulation methodology of the in-reactor fuel rod supporting conditions as a function of burnup has been developed and implemented in the GRIDFORCE program. The simulation methodology takes into account cladding creep rate, initial spring deflection, initial spring force, and spring force relaxation rate as the key fuel design parameters affecting the in-reactor fuel rod supporting conditions. Based on the parametric studies on these key parameters, it is found that the initial spring deflection, the spring force relaxation rate and cladding creepdown rate are in the order of the impact on the in-reactor fuel rod supporting conditions. Application of this simulation methodology to the fretting wear-induced failure experienced in a commercial plant indicates that this methodology can be utilized as an effective tool in evaluating the capability of newly developed cladding materials and/or new spacer grid designs against the fretting wear-induced damage.
The Unsteady 2-D Numerical Analysis in a Horizontal Pipe with Thermal Stratification Phenomena
Youm, Hag-Ki ; Park, Man-Heung ; Kim, Sang-Nung ;
Nuclear Engineering and Technology, volume 28, issue 1, 1996, Pages 27~35
In this paper, an unsteady analytical model for the thermal stratification in the pressurizer surge line of PWR plant has been proposed to investigate the temperature profile, flow characteristics, and thermal stress in the pipe. In this model, the interface level, between hot and cold fluid, is assumed to be a function of time while the other models had developed for time independent or steady state. The dimensionless governing equations are solved by using a SIMPLE (Semi-Implicit Method for Pressure Linked Equations) algorithm. The analysis result for an example shows that the maximum dimensionless temperature difference is about 0.78 between hot and cold sections of pipe wall and the maximum thermal stress by thermal stratification is calculated about 276 MPa at the dimensionless time 27.0 under given conditions.
Development of a Real-Time Thermal Performance Diagnostic Monitoring System Using Self-Organizing Neural Network for KORI-2 Nuclear Power Unit
Kang, Hyun-Gook ; Seong, Poong-Hyun ;
Nuclear Engineering and Technology, volume 28, issue 1, 1996, Pages 36~43
In this work, a PC-based thermal performance monitoring system is developed for the nuclear power plants. The system performs real-time thermal performance monitoring and diagnosis during plant operation. Specifically, a prototype for the KORI-2 nuclear power unit is developed and examined in this work. The analysis and the fault identification of the thermal cycle of a nuclear power plant is very difficult because the system structure is highly complex and the components are very much inter-related. In this study, some major diagnostic performance parameters are selected in order to represent the thermal cycle effectively and to reduce the computing time. The Fuzzy ARTMAP, a self-organizing neural network, is used to recognize the characteristic pattern change of the performance parameters in abnormal situation. By examination, this algorithm is shown to be able to detect abnormality and to identify the fault component or the change of system operation condition successfully. For the convenience of operators, a graphical user interface is also constructed in this work.
Adsorption Characteristics of Elemental Iodine and Methyl Iodide on Base and TEDA Impregnated Carbon
Lee, Hoo-Kun ; Park, Geun-Il ;
Nuclear Engineering and Technology, volume 28, issue 1, 1996, Pages 44~55
For the purpose of controlling the release of radioiodine to the environment in nuclear power plants, adsorption characteristics of elemental iodine and methyl iodide on the base carbon and 2%, 5% TEDA impregnated carbons were studied. The amounts of adsorption of elemental iodine and methyl iodide on the carbons were compared with Langmuir, Freundlich, Sips and Dubinin-Astakhov(DA) isotherm equations. Adsorption data were well correlated by the DA equation based on the potential theory. Adsorption energy distributions were obtained from the parameters of the DA equation derived from the condensation approach method. For the adsorption of methyl iodide and elemental iodine-carbon system, the DA equation can be well expressed by the degree of heterogeneity of the micropore system because the surface is nonuniform when its potential energy is unequal. The adsorption energy distribution wes investigated to find a surface heterogeneity on the carbon. The surface heterogeneity for iodine-carbon system is highly affected by the adsorbate-adsorbent interaction as well as the pore structure. The surface heterogeneity increases as a content of TEDA impregnated increases. The adsorption nature of methyl iodide on carbon turned out to be more heterogeneous than that of elemental iodine.
Spacer Grid Effects on Turbulent Flow in Rod Bundles
Yang, Sun-Kyu ; Chung, Moon-Ki ;
Nuclear Engineering and Technology, volume 28, issue 1, 1996, Pages 56~71
The local hydrulic characteristics in subchannels of 5
5 nuclear fuel bundles with spacer grids were measured at upstream and downstream of the spacer grid for the investigation of the spacer grid effects on turbulent flow structure by using an LDV(Laser Doppler Velocimeter). The measured parameters are axial velocity and turbulent intensity, skewness factor, and flatness factor. Pressure drops were also measured to evaluate the loss coefficient for the spacer grid and the friction factor for rod bundles. From these data, it was found that the turbulent mixing and forced mixing occur up to
and 20 from the spacer grid, respectively. The turbulence decay behind spacer grid behaves in the similar decay rate as turbulent flow through mesh grids or screens. Mixing factors useful in subchannel analysis code were correlated from the data and show the highest value near spacer grid and then have a stable values.
A Regression Program COVAFIT Accounting for Variance-Covariances in Experimental Nuclear Data
Oh, Soo-Youl ; Jonghwa Chang ;
Nuclear Engineering and Technology, volume 28, issue 1, 1996, Pages 72~78
A computer program COVAFIT has been developed and applied to the evaluation of experimental cross sections for MeV energy incident particles. The program utilizes weighted least-square linear regression method with high-order polynomials derived in this study. Meeting the growing demand for the treatment of covariances in nuclear data, it deals with the variance and covariance data provided along with experimental cross sections and yields those for the evaluated ones. The evaluated results on two sets of neutron total cross section of oxygen and three sets of proton cross section for
production reactions confirm the methodology formulated in and the applicability of the program.
Assessment of Gas Generation in Underground Repository of Low-Level Waste
Cho, Chan-Hee ; Kim, Chang-Lak ; Lee, Myung-Chan ; Park, Heui-Joo ;
Nuclear Engineering and Technology, volume 28, issue 1, 1996, Pages 79~92
In a repository containing low-level waste, gas generation will occur principally by the coupled processes of metal corrosion and microbial degradation of cellulosic waste. This paper describes a mathematical model designed to address gas generation by these mechanisms and assesses the potential effects of gas generation on the performance of a radioactive waste repository. The metal corrosion model incorporates a three-stage process encompassing aerobic and anaerobic corrosion regimes ; the microbial degradation model simulates the activities of eight different microbial populations, which are maintained as functions both of pH and of the concentrations of particular chemical species. A prediction is made for gas concentrations and generation rates over an assessment period of ten thousand years in a radioactive waste repository. The results suggest that H
will be the principal gas generated within the radioactive waste cavern.
기체 크로마토그래피법을 이용한 수소동위원소 분리기술
Nuclear Engineering and Technology, volume 28, issue 1, 1996, Pages 93~101