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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
Editor in Chief :
Volume & Issues
Volume 14, Issue 4 - Dec 1982
Volume 14, Issue 3 - Sep 1982
Volume 14, Issue 2 - Jun 1982
Volume 14, Issue 1 - Mar 1982
Selecting the target year
Generation and Benchmark Test of 26-group Constant Set for Fast Reactor Calculations
Jung-Do Kim ; Jong-Tai Lee ;
Nuclear Engineering and Technology, volume 14, issue 4, 1982, Pages 163~171
An ABBN-type 26-group constant set, KAERI-26G, which can be reliably applicable to fast reactor calculations has been generated using the nuclear data of ENDF/B-IV or ENDL-78 and a processing code ETOX-K4. The KAERI-26G set was evaluated by analysing measured integral quantities such as effective multiplication factor, central reaction-rate ratio, and central reactivity coefficient for a variety of critical assemblies. All these calculated quantities were compared with results from other workers using similar-type sets.
States of Am in Aqueous Solution
Won Mok Jae ;
Nuclear Engineering and Technology, volume 14, issue 4, 1982, Pages 172~177
The state of Am in the concentration of 10
M has been studied in the pH range of 5 to 10 by filtration and centrifugation method. By the experiments, we could estimate possible Am hydrolysis products and solubility constants. If the solubility of Am (OH)
estimated by Baes and Mesmer is increased about a factor of 10, i.e. changing logk=-l8.7 to logk=-17 5 it was found that the calculated curve of Am concentration versus pH agreed completely with experimental values.
A Finite Element Solution to the Group Diffusion Problems with Albedo-Type Boundary Conditions
Kun Joong Yoo ; Chang Hyo Kim ; Chang Hyun Chung ;
Nuclear Engineering and Technology, volume 14, issue 4, 1982, Pages 178~185
Albedo-type boundary condition is incorporated into the finite element formulation of the cubic Hermite polynomials for the two-dimensional solution of the two-group diffusion problem. Two modifications are introduced with respect to the conventional expression for the weak form of the group diffusion equation with the zero flux or zero current boundary condition and the cubic element functions over the boundary nodes. The finite element formulations obtained from those modifications are tested with the two-dimensional ZION problem. The numerical effectiveness of the modifications are examined.
Study on the Steam Line Break Accident for Kori Unit-1
Tae Woon Kim ; Jung In Choi ; Un Chul Lee ; Ki In Han ;
Nuclear Engineering and Technology, volume 14, issue 4, 1982, Pages 186~195
The steam line break accident for Kori Unit 1 is analyzed by a code SYSRAN which calculates nuclear power and heat flux using the point kinetics equation and the lumped-parameter model and calculates system transient using the mass and energy balance equation with the assumption of uniform reactor coolant system pressure. The 1.4 f
steam line break accident is analyzed at EOL (End of Life), hot shutdown condition in which case the accident would be most severe. The steam discharge rate is assumed to follow the Moody critical flow model. The results reveal the peak heat flux of 38% of nominal full power value at 60 second after the accident initiates, which is higher than the FSAR result of 26%. Trends for the transient are in good agreement with FSAR results. A sensitivity study shows that this accident is most sensitive to the moderator density coefficient and the lower plenum mixing factor. The DNBR calculation under the assumption of
=3.66, which is used in the FSAR with all the control and the shutdown assemblies inserted except one B bank assembly and of Fz=1.55 shows that minimum DNBR reaches 1.62 at 60 second, indicating that the fuel failure is not anticipated to occur. The point kinetics equation, the lumped-parameter model and the system transient model which uses the mass and energy balance equation are verified to be effective to follow the system transient phenomena of the nuclear power plants.lear power plants.
Design Study of A Spent Fuel Shipping Cask for Korea Nuclear Unit-1
Moo Han Kim ; Chang Sun Kang ;
Nuclear Engineering and Technology, volume 14, issue 4, 1982, Pages 196~203
To transport the spent fuel assemblies of Korea Nuclear Unit 1, which is a Westinghouse type two loop pressurized water reactor, it has been found that steel is the most appropriate material for the design of a shipping cask in comparison with lead and depleted uranium. The proposed shipping cask will transport nine fuel assemblies at the same time and is well within the weight limit of transportation by unrestricted rail car. The cask requires 33cm thick steel shield and 27cm thick water region to satisfy the 3 feet apart dose rate limit set forth in 10 CFR 71, and 1.27cm thick steel boron fuel basket to hold the fuel elements inside the cask and control the effective multiplication factor. As a safety analysis, the fuel cladding and centerline temperatures were calculated under the accident condition of complete loss of water coolant, and it was found that the temperature was much lower than the limit of the melting point. k
was calculated with fresh fuel assemblies, which was found to be well lower than 0.95. For shielding computation, the multipurpose Monte Carlo code MORSE-CG and one dimensional discrete ordinates transport code ANISN were used, and the Monte Carlo codes KENO and MORSE-CG were used for criticality calculation. The radiation source terms were calculated using ORIGEN-79.9.
Description and Discussion of the Current State of the Knowledge about the Leidenfrost Phenomenon
Moon Ki Chung ; Young Whan LEE ;
Nuclear Engineering and Technology, volume 14, issue 4, 1982, Pages 204~218
The purpose of this report is to describe and discuss the current state of the knowledge about the Leidenfrost phenomenon which is a heavily studied subject in the field of boiling heat transfer. The strong interest is due to reactor safety considerations since it is desirable to obtain a better understanding of the physical mechanisms involved in the rewetting of high temperature surface after a loss of coolant accident. Brief survey of the theoretical and experimental results from available open literatures indicates that considerable discrepancy exists in the prediction of the Leidenfrost temperature at the elevated pressures and more investigations are needed in this area.
일본에서의 발전로연료안전성에 관한 연구
Nuclear Engineering and Technology, volume 14, issue 4, 1982, Pages 219~227
일본에서의 경수로핵연료의 연구개발
Nuclear Engineering and Technology, volume 14, issue 4, 1982, Pages 228~234