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REFERENCE LINKING PLATFORM OF KOREA S&T JOURNALS
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Nuclear Engineering and Technology
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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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Analysis of Water Purification Capability of the Spent Fuel Storage Pool Using Consolidated Fuel Storage in Uljin 1&2
Lim, Chae-Joon ; Park, Goon-Cherl ; Chung, Chang-Hyun ;
Nuclear Engineering and Technology, volume 22, issue 2, 1990, Pages 83~94
The radioactivity in the spent fuel storage pool is calculated to ensure to maintain its concentration below the permissible limit, when the storage capacity of Uljin nuclear power plant unit 1&2 is extended from 9/3 to 32/3 core using consolidated fuels in maximum density rack (MDR). For this evalulation, two models to calculate the spent fuel pool activities on the continuous and intermittent operating its purification system are developed and these results compared, The results of above two cases show that the current water purification system can not guarantee the radioactivity concentration below the design limit, 5
Ci/ml, for the extention to 32/3 core. Therefore, it has been concluded that a modification of the current purification system is necessary to extend the spent fuel storage capacity with the above method. The alternative way suggested in this study is to increase the number of cation bed demineralizers.
Analysis of Hydraulic Lift Force of a Fuel Assembly
Sim, Yoon-Sub ; Oh, Dong-Seok ; Hong, Soung-Dug ; Kwon, Hyuk-Sung ;
Nuclear Engineering and Technology, volume 22, issue 2, 1990, Pages 95~100
The exact expression for the 1151 force on a fuel assembly in a reactor core is derived in terms of calculable hydraulic parameters. The relation for the lift force. pressure drop, buoyancy force, viscous force. and fuel assembly weight is discussed. Based on the derived exact expression. error analysis is made for a simple expression applying COBRA IV-i to a typical PWR fuel assembly. The error analysis revealed that the error of the simple expression consists of four terms and the overall error depends on the flow rate change direction, and its magnitude is about 1%.
A Study on PIXE Spectrum Analysis for the Determination of Elemental Contents
Jong-Seok OH ; ; Hae-ILL Bak ;
Nuclear Engineering and Technology, volume 22, issue 2, 1990, Pages 101~107
The PIXE (Proton Induced X-ray Emission) method is applied to the quantitative analysis of trace elements in tap water, red wine, urine and old black powder samples. Sample irradiations are performed with a 1.202 MeV proton beam from the SNU 1.5-MV Tandem Van de Graaff accelerator, and measurements of X-ray spectra are made by the Si(Li) spectrometer To increase the sensitivity of analysis tap water is preconcentrated by evaporation method. As an internal standard, Ni powder is mixed with black powder sample and yttrium solution is added to the other samples. The analyses of the PIXE spectra are carried out by using the AXIL (Analytical X-ray Analysis by Iterative Least-squares) computer code, in which the routine for least-squares method is based on the Marquardt algorithm. The elements such as Mg, Al, Si, Ti, Fe and Zn are analyzed at sub-ppm levels in the tap water sample. In the red wine sample prepared without preconcentration. the element Ti is detected in the amount of 3ppm. In conclusion, the PIXE method is proved to be appropriate for the analysis of liquid samples by relative measurements using the internal standard. and is expected to be improved by the use of evaluated X-ray production cross-sections and the development of sample preparation techniques.
Measurements of X-Ray Production Cross-Sections for 0.5¡1.2-MeV Proton Beam
Hae-ill BAK ; Jun-Gyo BAK ;
Nuclear Engineering and Technology, volume 22, issue 2, 1990, Pages 108~115
The measurements of X-ray production cross-sections for 0.5~1.2-MeV proton beam are carried out on Cu and Au. For this experiment, the proton tram generated from the SNU 1.5-MV Tandem Van do Graaff accelerator is Incident on the target. The X-rays and the backscattered protons from the irradiated target are detected simultaneously by the Si(Li) X-ray detector and the SSB (Silicone Surface Barrier) charged particle detector The measured values of X-ray production cross-sections are compared with other experimental values and theoretical values such as the PWBA (Plane Wave Born Approximation) and the ECPSSR(Perturbed Stationary State corrected Energy loss, Coulomb deflection, Relativistic effects) values. For measured cross-sections near 1.0- MeV proton energy, the ECPSSR (D.D. Cohenet al., 1985) shows better agreement than the PWBA. Particularly, that of Au for 1.2 MeV proton beam is 9.69
0.39 barns which deviates from the ECPSSR by less than 5%. and the experimental data for 0.5~1.2- MeV proton agree with most of other experimental values within 30%.
A Study on Effects of Axial Gas Flow in the Gap and Fuel Cracking on Fission Gas Release under Power Ramping
Han, Jin-Kyu ; Yoon, Young-Ku ;
Nuclear Engineering and Technology, volume 22, issue 2, 1990, Pages 116~127
The fission gas release model used In the SPEAR-BETA fuel performance code was modified by use of effective thermal conductivity for cracked fuel and by laking Into account axial fission-gas mixing between the fuel-clad gap and the plenum. With use of this modified model the fission gas release was analyzed under various power ramping conditions of P
.fP. Effective fuel thermal conductivity that accounts for the effect of fuel tracking was used in calculation of the fuel temperature distribution and the Internal gas pressure under power ramping conditions. Mixing and dilution effects due to axial gas flow were also considered in computing the width and the thermal conductivity of the gap. The effect of axial gas flow w3s solved by the Crank-Nicholson method. The finite difference method was used to save running time in the calculation. The present modified fission-gas release model was validated by comparing its predicted results with experimental data from various lamping tests In the literature and calculated results with use of the models used In the SPEAR-BETA and FEMAXI-IV codes. Results obtained with use of the present modified model showed better agreement with experimental data reported in the literature than those results with use of the latter codes. The fuel centerline temperature calculated with introduction of effective thermal conductivity for centerline temperature calculated with Introduction of effective thermal conductivity for cracked fuel was 200 higher fission gas release predicted with use of the modified model was nearly 6% larger on the average than that calculated by use of the unmodified model used in the SPEAR-BETA code.e SPEAR-BETA code.e.
Development of an Expert System (ESRCP) for Failure Diagnosis of Reactor Coolant Pumps
Cheon, Se-Woo ; Chang, Soon-Heung ;
Nuclear Engineering and Technology, volume 22, issue 2, 1990, Pages 128~138
This paper presents a prototype expert system (ESRCP) for Reactor Coolant Pumps. The purpose of this system is to diagnose RCP failures and to offer corrective operational guides to plant operators. The first symptoms for the diagnosis are the alarms which are related to the RCP domain. Alarm processing is required to find a primary causal alarm when multiple alarms occur. The system performs the alarm processing by rule-based deduction or priority factor operation. To diagnose the RCP failure, the system performs rule-based deduction or Bayesian inference. Various sensor readings are required as symptoms to infer a root cause. When the symptoms are insufficient or uncertain to diagnose accurately, Bayesian inference is performed.
An Observer-Theoretic Approach to Estimating Neutron Flux and Precursor Spatial Distributions
Park, Young-Ho ; Cho, Nam-Zin ;
Nuclear Engineering and Technology, volume 22, issue 2, 1990, Pages 139~150
This paper describes a method for estimating the flux and precursor spatial distributions using only limited flux measurements. It is based on the Luenberger observer in control theory, extended to the distributed parameter systems such as the space-time reactor dynamics equation. The results of the application of the method to simple reactor models showed that the flux distribution could be estimated by the observer very efficiently using information from only a few sensors.
Effectiveness of the Discrete Elements Method for the Slab-Geometry Neutron Transport Equation
Na, Byung-Chan ; Kim, ong-Kyung ;
Nuclear Engineering and Technology, volume 22, issue 2, 1990, Pages 151~158
The new discrete elements method (DEM) is applied to the one-group neutron transport equation in one-dimensional slab geometry. The fixed source and the criticality problems are treated and three spatial differencing schemes (the DD, the SC, -and the LC schemes) are tested to determine the most computationally efficient in the DEM. In all cases, the accuracy of the results obtained from the DEM shows an improvement over that obtained from the standard discrete ordinates calculations. And the LC scheme gives the most accurate results in the DEM.
신형원자로 개발현황 및 전망
Nuclear Engineering and Technology, volume 22, issue 2, 1990, Pages 160~172
신형 원자로의 안전 설계개념 검토
Nuclear Engineering and Technology, volume 22, issue 2, 1990, Pages 173~185
신형원자로의 열수력학적 특성에 관한 고찰
Nuclear Engineering and Technology, volume 22, issue 2, 1990, Pages 186~195
개량형 경수로의 현황과 특성 비교
Nuclear Engineering and Technology, volume 22, issue 2, 1990, Pages 196~216