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REFERENCE LINKING PLATFORM OF KOREA S&T JOURNALS
> Journal Vol & Issue
Nuclear Engineering and Technology
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Journal DOI :
Korean Nuclear Society
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Volume & Issues
Volume 22, Issue 4 - Dec 1990
Volume 22, Issue 2 - Jun 1990
Volume 22, Issue 1 - Mar 1990
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Computer Program Development for D
O Upgrader Performance Management
Ahn, Do-Hee ; Kim, Kwang-Rag ; Chung, Hong-Suck ; Kim, Yong-Eak ; Jeong, Ill-Seok ; Hon, Sung-Yull ; Ko, Jae-Wook ;
Nuclear Engineering and Technology, volume 22, issue 1, 1990, Pages 1~11
Heavy water is used as a moderator and a coolant in the pressurized heavy water reactor Because of the high cost of heavy water, downgraded heavy water generated in the reactor system is recycled to the reactor after being concentrated up to 99.8% or more in heavy water upgraders. This study investigates the process of upgraders and then suggests a theoretical model. The relations between process variables are derived from tower packing characteristics, vapour-liquid equilibria, and mass-heat balance equations at a steady state operation of the upgrader h computer program UPGR is developed, using the algorithm that solves the nonlinear equations step by step. It shows that the results of computer simulation are in good agreement with the operating data of the Wolsung upgrader. Thus, this computer code offers the optimum operating guide and is now applied to manage the performance of upgraders for the effective operation of the heavy water upgraders.
Assessment of RELAP5/MOD2 Code using Loss of Offsite Power Transient of Kori Unit 1
Chung, Bub-Dong ; Kim, Hho-Jung ; Lee, Young-Jin ;
Nuclear Engineering and Technology, volume 22, issue 1, 1990, Pages 12~19
The Loss of Offsite Power Transient at 77.5% power which occurred on June 9, 1981 at the Kori Unit 1 PWR (Pressurized Water Reactor) is simulated using the RELAP5/MOD2 system thermal-hydraulics computer code. Major thermal-hydraulic parameters are compared with the available plant data. The comparison of the analysis results with the plant data demonstrates that the RELAP5/MOD2 code has the capability to simulate the thermal-hydraulic behaviour of PWRs under accident conditions of this type with accuracy, except the pressurizer pressure and level. The pressurizer pressure increase is sensitive to the in surge now it is believed that the interracial heat transfer in a horizontal stratified flow regime may be estimated low and the compression effect due to insurge flow may be high. In the nodalization sensitivity study it is found that S/G noding with junctions between bypass plenum and steam dome is preferred to simulate the S/G water level decreasing and avoid the spurious level peak at trubine trip
A Study on the Migration Characteristics of Cs-137 in a Packed Column
Lee, Jae-Owan ; Cho, Won-Jin ; Han, Kyung-Won ; Park, Hun-Hwee ;
Nuclear Engineering and Technology, volume 22, issue 1, 1990, Pages 20~28
In this study the migration experiment using packed column with crushed tuff was conducted as a basic research to develop migration model of radionuclides through geologic media. The main emphasis was put on evaluating the validity of migration models. For this, two models were introduced: one is the model which is based on the assumption of instantaneous equilibrium reaction and the other the model based on kinetic process such as intraparticle diffusion. The coefficient of hydrodynamic dispersion in packed column was determined using iodine as nonsorbing tracer. The hydrodynamic dispersion coefficient, D
was shown to be 0.11
/min under the condition of the column porosity of 0.483 and the average water velocity of 0.915
cm/min. The distribution coefficient, Kd of Cs-137 on crushed tuff was 11.3 cc/g at the concentration of 2
M and the temperature of 2
. The breakthrough curve of Cs-137 through packed column was shown to have an asymmetric curve in which long trailing tail appears at the end part of the curve. The results obtained from the comparison of introduced models with experimental data indicated that the mass transfer model with intraparticle diffusion as rate-controlling step simulated the behaviors of Cs-137 migration more adequately, when compared with the bulk reaction model in which the assumption of instantaneous equilibrium reaction was maded. Consequently, the intraparticle diffusion was found to be an important factor in the migration of Cs-137 through packed column.n.
Geochemical Modeling of U Solubility in Groundwater Conditions
Cho, Young-Hwan ; Han, Kyung-Won ; Suh, In-Suk ;
Nuclear Engineering and Technology, volume 22, issue 1, 1990, Pages 29~35
Uranium solubilities have been calculated for a range of conditions expected in a nuclear waste disposal repository. Variables taken into consideration include the pH and Eh range expected for deep groundeaters, the effect of the composition of groundwater. The model used in these calculations is based on the assumption of chemical equilibrium. Calculations show that the major variables influencing uranium solubility under the repository conditions are pH and Eh. The results of this study can be applied to an assessment of the nuclear waste disposal.
Development of the Safety Assessment Code (CALM) for the Disposal of Low-and Intermediate-Level Radioactive Waste
Han, Kyong-Won ; Cho, Won-Jin ; Lee, Han-Soo ; Lee, Youn-Myoung ; Park, Hee-Sung ; Suh, Kyung-Suk ; Park, Heu-Joo- ; Park, Hun-Hwee ;
Nuclear Engineering and Technology, volume 22, issue 1, 1990, Pages 36~44
A safety assessment computer code CALM (Computer program of Assessment for LILW Management) is developed for the theoretical prediction of long-term safety of low-and intermediate-level radioactive waste disposal. CALM is composed of three submodels, which are the resaturation model, the geosphere migration model, and the radiation dose model. For the verification of its usefulness, the safety assessment of an assumed waste repository is performed. The results show that the computer code, CALM developed through this study can be a useful tool for the safety assessment of low- and intermediate-level radioactive waste repository.
A Steady-State Margin Comparison between Analog and Digital Protection Systems
Auh, Geun-Sun ; Hwang, Dae-Hyun ; Kim, Si-Hwan ;
Nuclear Engineering and Technology, volume 22, issue 1, 1990, Pages 45~57
A steady-state margin comparison study was performed between analog and digital protection systems. The systems compared are the thermal overpower and overtemperature delta T system of Westinghouse, and Core Protection Calculator System of Combustion Engineering, Inc. No dynamic offset was considered to eliminate the margin differences by different safety analysis methodologies. The result shows that the digital protection system has about 30% more rated power margin than the analog system in protecting against the fuel rod centerline melting. The digital protection system is shown to have almost same margin with the analog protection system in preventing the DNB at EOC (End of Cycle) even if the digital protection system has about 10% more margin at BOC(Beginning of Cycle).
PC-Based Random Neutron Process Measurement in a Thermal Reactor
Jun, Byung-Jin ; Park, Sang-Jun ; Hong, Kwang-Pyo ; Lee, Chung-Sung ;
Nuclear Engineering and Technology, volume 22, issue 1, 1990, Pages 58~65
A PC-based system for measuring and analysing random neutron process in the thermal reactor is developed and applied to TRIGA Mark-II reactor at KAERI. It is confirmed that this system has several advantages compared to conventional methods. So far, two techniques, autocorrelation and variance to mean ratio (VTMR), have been applied for analysing the count data collected from the single detector by using this system. The results of the two techniques agree within acceptable difference, but VTMR's results show much superior statistical reliability than those of autocorrelation especially when it is near critical. The
/Λ of TRIGA Mark-II reactor is measured to be about 125/sec when the reactivity is within -3＄ and about 150/sec when it is below -4＄.
Comparison of the Thermal-Hydraulic Characteristics of Optimised Fuel Assembly with That of Standard Fuel Assembly
Paik, Hyun-Jong ; Rim, Chang-Saeng ; Park, Goon-Cherl ;
Nuclear Engineering and Technology, volume 22, issue 1, 1990, Pages 66~74
The thermal-hydraulic characteristics of the 17
17 OFA (Optimized Fuel Assembly) used in the KNU 7&8 are analyzed and compared with that of the 17
17 SFA (Standard Fuel Assembly) loaded in the KNU 5&6. The thermal-hydraulic characteristics analyzed are minimum DNBR, fuel centerline temperature and exit void fraction at normal operation and design over power transient. Additionally, local linear rod power, which will cause fuel centerline melting, is calculated. The DNBR sensitivity calculations are performed with respect to the reactor operating parameters. COBRA-IV-I code is used for these calculations. The modified W-3 correltion and the drift-flux model are applied for the critical heat flux calculation and the void fraction calculation, respectively. From the calculated results, it has been found that the possibility of DNB occurrence is higher in the OFA than in the SFA. The other hand, the local linear power resulting in fuel centerline moiling of the OFA is nearly equal to that of the SFA.
Measurement of The Thermal Contact Conductance in Nuclear Fuel Element
Sung-Deok Hong ; ; Goon-Cherl Park ;
Nuclear Engineering and Technology, volume 22, issue 1, 1990, Pages 75~81
Experiments to predict the thermal contact conductance between the fuel pellet and cladding have been performed, which is important to determine the temperature distibution within the fuel rod. UO
and Zircaloy-2 are used in these experiments. The measuring apparatus is composed of a presser which controls the contact pressure, a thermometer with 5.5 sheathed thermocouples, a vacuum pump, pellet and cladding rods, and two heating devices, etc. The thermal contact conductances were measured with varying the contact pressure and surface roughnesses of UO
and Zircaloy-2 bars. The results show that an increase in the contact pressure and a decrease of surface roughness resulted in increase of the thermal contact conductance. Finally, a fitting correlation has been established and compared with widely-used correlations.