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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 19, Issue 4 - Dec 1987
Volume 19, Issue 3 - Sep 1987
Volume 19, Issue 2 - Jun 1987
Volume 19, Issue 1 - Mar 1987
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A Study on the Method of Combining Empirical Data and Deterministic Model for Fuel Failure Prediction
Cho, Byeong-Ho ; Yoon, Young-Ku ; Chang, Soon-Heung ;
Nuclear Engineering and Technology, volume 19, issue 4, 1987, Pages 233~241
Difficulties are encountered when the behavior of complex systems (i.e., fuel failure probability) that have unreliable deterministic models is predicted. For more realistic prediction of the behavior of complex systems with limited observational data, the present study was undertaken to devise an approach of combining predictions from the deterministic model and actual observational data. Predictions by this method of combining are inferred to be of higher reliability than separate predictions made by either model taken independently. A systematic method of hierarchical pattern discovery based on the method developed in the SPEAR was used for systematic search of weighting factors and pattern boundaries for the present method. A sample calculation was performed for prediction of CANDU fuel failures that had occurred due to power ramp during refuelling process. It was demonstrated by this sample calculation that there exists a region of feature space in which fuel failure probability from the PROFIT model nearly agree with that from observational data.
Trace Analysis of Uranium in Aqueous Samples by Laser-induced Fluorescence Spectroscopy
Jung, Kwang-Woo ; Kim, Jeong-Moog ; Kim, Cheol-Jung ; Lee, Jong-Min ;
Nuclear Engineering and Technology, volume 19, issue 4, 1987, Pages 242~248
A sensitive, direct method for the determination of trace amounts of uranium in solution has been developed utilizing laser-induced fluorescence spectroscopy and a fluorescence enhancing reagent 'Fluran.' Standard addition technique is incorporated into the analysis to eliminate sample matrix effects. Analytical data show that a detection limit of 0.1 ppb (part per billion) uranium has been achieved and the precision of the analysis is in the range of 5% relative standard deviation. Results using the laser fluorescence method on many sets of unknown samples have been compared against corresponding values determined by other methods.
Development of Statistical Package for Uncertainty and Sensitivity Analysis(SPUSA) and Application to High Level Waste Repostitory System
Kim, Tae-Woon ; Cho, Won-Jin ; Chang, Soon-Heung ; Le, Byung-Ho ;
Nuclear Engineering and Technology, volume 19, issue 4, 1987, Pages 249~265
For the probabilistic risk assessment of the high level radioactive waste repository, some methods have been proposed up to now. Since the system has highly uncertain input parameters, the evaluated risk for some input parameter values has high uncertainty. In this paper, methods of uncertainty and sensitivity analysis are devised to analyse systematically these factors and applied to a probabilistic risk assessment model of the high level waste repository, The statistical package SPUSA developed through this study can be used for any other fields, e.g., statistical thermal margin analysis, source term uncertainty analysis, etc.
On Setting Low-level Performance Criteria and Uncertainty Characterization for a Nuclear Power Plant
Jo, Nam-Jin ;
Nuclear Engineering and Technology, volume 19, issue 4, 1987, Pages 266~278
This paper addresses the issues in setting performance criteria for safety regulation of nuclear power plants. Since setting criteria at the low level is a much more difficult task than it is at the top level, the low-level performance criteria should be derived consistently from the more easily determinable top-level performance criteria. The paper also proposes several approaches to characterizing uncertainties in performance criteria, by extending the reliability allocation methodology that is based on the mean-to-mean mapping to a stochastic multi-objective optimization problem where the state variables are uncertain.
Reliability Analysis of the Reactor Protection System Using Markov Processes
Jo, Nam-Jin ;
Nuclear Engineering and Technology, volume 19, issue 4, 1987, Pages 279~291
The event tree/fault tree techniques used in the current probabilistic risk assessment (PRA) of nuclear power plants are based on the binary and static description of the components and the system. While these techniques Bay be adequate in most of the safety studies, more advanced techniques, e.g., the Markov reliability analysis, are required to accurately study such problems as the plant availability assessments and technical specifications evaluations that are becoming increasingly important. This paper describes a Markov model for the Reactor Protection System of a pressurized water reactor and presents results of model evaluations for two testing policies in technical specifications.
A Large Dry PWR Containment Response Analysis for Postulated Severe Accidents
Chun, Moon-Hyun ;
Nuclear Engineering and Technology, volume 19, issue 4, 1987, Pages 292~309
A large dry PWR containment response analysis for postulated severe accidents was performed as part of the Zion Risk Rebaselining study for input to the U.S. NRC's "Reactor Risk Reference Document," NUREG-1150. The Methodologies used in the present work were developed as part of the Severe Accident Risk Reduction Program (SARRP) at Sandia National Laboratory specifically for the Surry Plant, but they were extrapolated to Zion. Major steps of the quantification of risk from a nuclear power plant are first outlined. Then, the methodologies of containment response analysis for severe accidents used for Zion are described in detail: major features of the containment event tree (CET) analysis codes and CET quantification procedures are summarized. In addition, plant specific features important to containment response analysis are presented along with the containment loading and performance issues included in the present uncertainty analysis. Finally, a brief summary of the results of deterministic and statistical containment event tree analysis is presented to provide a perspective on the large dry PWR containment response for postulated severe accidents.accidents.
Numerical Calculation of λ-Mode of the Diffusion Equation
T.W. Noh ; S.K. Oh ; Kim, S.Y. ; Kim, C.H. ;
Nuclear Engineering and Technology, volume 19, issue 4, 1987, Pages 310~316
A successive iteration method to calculate the λ-modes of the diffusion equation was developed. The 2-group, 3-dimensional computer code MOGEN was developed to implement this method, The accuracy of the method was demonstrated using 2-dimensional bare homogeneous rectangular reactor. The numerical solution shows good agreement with the analytic solution in terms of eigenvalue and eigenfunction As for the standard CANDU-600 reactor, the 2-dimensional modes were generated and these represent the conventional mode characteristics well. Finally, application of theλ-mode in reactor engineering problems is described briefly.
A Study on Spatial Neutron Kinetics of a Pressurized Water Reactor
Kim, Chang-Hyo ;
Nuclear Engineering and Technology, volume 19, issue 4, 1987, Pages 317~324
The purpose of this work is to present a spatial neutron kinetics computational scheme for the analysis of space-dependent transients like rod ejection accident of pressurized water reactors. In this work modified Borresen's 1.5 group coarse mesh scheme was formulated for the neutronic computation of the space-dependent transients and applied to the analysis of hypothetical rod ejection accident of KNU no. 1 PWR core at BOC, HZP. The computational accuracy of the modified Borresen's scheme is examined by comparing calculations for core power and control rod worths with startup core physics test results. Effects of such parameters as ejected rod worths and number of delayed neution group ell transient results as well as computational efficiency are also examined. OB this basis it is suggested that the modified Borresen's method is a useful scheme for the analysis of spatial neutron transients of PWR's.