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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 37, Issue 6 - Dec 2005
Volume 37, Issue 5 - Oct 2005
Volume 37, Issue 4 - Aug 2005
Volume 37, Issue 3 - Jun 2005
Volume 37, Issue 2 - Apr 2005
Volume 37, Issue 1 - Feb 2005
Selecting the target year
COMPUTATIONAL INTELLIGENCE IN NUCLEAR ENGINEERING
UHRIG ROBERT E. ; HINES J. WESLEY ;
Nuclear Engineering and Technology, volume 37, issue 2, 2005, Pages 127~138
Approaches to several recent issues in the operation of nuclear power plants using computational intelligence are discussed. These issues include 1) noise analysis techniques, 2) on-line monitoring and sensor validation, 3) regularization of ill-posed surveillance and diagnostic measurements, 4) transient identification, 5) artificial intelligence-based core monitoring and diagnostic system, 6) continuous efficiency improvement of nuclear power plants, and 7) autonomous anticipatory control and intelligent-agents. Several changes to the focus of Computational Intelligence in Nuclear Engineering have occurred in the past few years. With earlier activities focusing on the development of condition monitoring and diagnostic techniques for current nuclear power plants, recent activities have focused on the implementation of those methods and the development of methods for next generation plants and space reactors. These advanced techniques are expected to become increasingly important as current generation nuclear power plants have their licenses extended to 60 years and next generation reactors are being designed to operate for extended fuel cycles (up to 25 years), with less operator oversight, and especially for nuclear plants operating in severe environments such as space or ice-bound locations.
DESIGN AND VALIDATION OF ROBUST AND AUTONOMOUS CONTROL FOR NUCLEAR REACTORS
SHAFFER ROMAN A. ; EDWARDS ROBERT M. ; LEE KWANG Y. ;
Nuclear Engineering and Technology, volume 37, issue 2, 2005, Pages 139~150
A robust control design procedure for a nuclear reactor has been developed and experimentally validated on the Penn State TRIGA research reactor. The utilization of the robust controller as a component of an autonomous control system is also demonstrated. Two methods of specifying a low order (fourth-order) nominal-plant model for a robust control design were evaluated: 1) by approximation based on the 'physics' of the process and 2) by an optimal Hankel approximation of a higher order plant model. The uncertainty between the nominal plant models and the higher order plant model is supplied as a specification to the ,u-synthesis robust control design procedure. Two methods of quantifying uncertainty were evaluated: 1) a combination of additive and multiplicative uncertainty and 2) multiplicative uncertainty alone. The conclusions are that the optimal Hankel approximation and a combination of additive and multiplicative uncertainty are the best approach to design robust control for this application. The results from nonlinear simulation testing and the physical experiments are consistent and thus help to confirm the correctness of the robust control design procedures and conclusions.
HUMAN-MACHINE INTERACTION IN NUCLEAR POWER PLANTS
YOSHIKAWA HIDEKAZU ;
Nuclear Engineering and Technology, volume 37, issue 2, 2005, Pages 151~158
Advanced nuclear power plants are generally large complex systems automated by computers. Whenever a rare plant emergency occurs the plant operators must cope with the emergency under severe mental stress without committing any fatal errors. Furthermore, The operators must train to improve and maintain their ability to cope with every conceivable situation, though it is almost impossible to be fully prepared for an infinite variety of situations. In view of the limited capability of operators in emergency situations, there has been a new approach to preventing the human error caused by improper human-machine interaction. The new approach has been triggered by the introduction of advanced information systems that help operators recognize and counteract plant emergencies. In this paper, the adverse effect of automation in human-machine systems is explained. The discussion then focuses on how to configure a joint human-machine system for ideal human-machine interaction. Finally, there is a new proposal on how to organize technologies that recognize the different states of such a joint human-machine system.
HUMAN RELIABILITY ASSESSMENT IN CONTEXT
HOLLNAGEL ERIK ;
Nuclear Engineering and Technology, volume 37, issue 2, 2005, Pages 159~166
Human reliability assessment (HRA) is conducted on the unspoken premise that 'human error' is a meaningful concept and that it can be associated with individual actions. The basis for this assumption it found in the origin of HRA, as a necessary extension of PSA to account for the impact of failures emanating from human actions. Although it was natural to model HRA on PSA, a large number of studies have shown that the premises are wrong, specifically that human and technological functions cannot be decomposed in the same manner. The general experience from accident studies also indicates that action failures are a function of the context, and that it is the variability of the context rather than the 'human error probability' that is the much sought for signal. Accepting this will have significant consequences for the way in which HRA, and ultimately also PSA, should be pursued.
PERUPS (PERFORMANCE UPGRADE SYSTEM) FOR ON-LINE PERFORMANCE ANALYSIS OF A NUCLEAR POWER PLANT TURBINE CYCLE
KIM SEONGKUN ; CHOI KWANGHEE ;
Nuclear Engineering and Technology, volume 37, issue 2, 2005, Pages 167~176
We developed the PERUPS system to aid the on-line performance analysis for the turbine cycle of the YongGwang 3 and 4 nuclear power plants. Procedure of measurement validation is included in the performance calculation to obtain heat balance. Precision of on-line performance calculation is increased via practical modifications of standard calculation algorithms based on the PTC (Performance Test Code). The proposed system also provides useful Web-based aids for performance analysis, including performance data management, a graphic viewer for heat balance and turbine expansion lines, and synthesized reports of performance.
MSET PERFORMANCE OPTIMIZATION THROUGH REGULARIZATION
HINES J. WESLEY ; USYNIN ALEXANDER ;
Nuclear Engineering and Technology, volume 37, issue 2, 2005, Pages 177~184
The Multivariate State Estimation Technique (MSET) is being used in Nuclear Power Plants for sensor and equipment condition monitoring. This paper presents the use of regularization methods for optimizing MSET's predictive performance. The techniques are applied to a simulated data set and a data set obtained from a nuclear power plant currently implementing empirical, on-line, equipment condition monitoring techniques. The results show that regularization greatly enhances the predictive performance. Additionally, the selection of prototype vectors is investigated and a local modeling method is presented that can be applied when computational speed is desired.
PRELIMINARY RESULTS OF THE BEAM CONTROL AND DETECTION OF THE KIRAMS ELECTRON MICROBEAM SYSTEM
SUN G.M. ; KIM E.H. ; SONG K.B. ; JEONG J.W. ; CHOI H.D. ;
Nuclear Engineering and Technology, volume 37, issue 2, 2005, Pages 185~190
The Korea Institute of Radiological and Medical Sciences (KIRAMS) electron microbeam system has been built with its prototype components. The system is composed of an electron gun, a beam transport chamber, and a cell image acquisition and positioning stage. Each component has been upgraded through repetitive performance tests for various parametric arrangements. This paper presents the preliminary results of the performance test on the beam control and detection parts of the system.
SHAKING TABLE TEST OF STEEL FRAME STRUCTURES SUBJECTED TO SCENARIO EARTHQUAKES
CHOI IN-KlL ; KIM MIN KYU ; CHOUN YOUNG-SUN ; SEO JEONG-MOON ;
Nuclear Engineering and Technology, volume 37, issue 2, 2005, Pages 191~200
Shaking table tests of the seismic behavior of a steel frame structure model were performed. The purpose of these tests was to estimate the effects of a near-fault ground motion and a scenario earthquake based on a probabilistic seismic hazard analysis for nuclear power plant structures. Three representative kinds of earthquake ground motions were used for the input motions: the design earthquake ground motion for the Korean nuclear power plants, the scenario earthquakes for Korean nuclear power plant sites, and the near-fault earthquake record from the Chi-Chi earthquake. The probability-based scenario earthquakes were developed for the Korean nuclear power plant sites using the PSHA data. A 4-story steel frame structure was fabricated to perform the tests. Test results showed that the high frequency ground motions of the scenario earthquake did not damage the structure at the nuclear power plant site; however, the ground motions had a serious effect on the equipment installed on the high floors of the building. This shows that the design earthquake is not conservative enough to demonstrate the actual danger to safety related nuclear power plant equipment.
THEORETICAL ANALYSIS FOR STUDYING THE FRETTING WEAR PROBLEM OF STEAM GENERATOR TUBES IN A NUCLEAR POWER PLANT
LEE CROON YEOL ; CHAI YOUNG SUCK ; BAE JOON WOO ;
Nuclear Engineering and Technology, volume 37, issue 2, 2005, Pages 201~206
Fretting, which is a special type of wear, is defined as small amplitude relative motion along the contacting interface between two materials. The structural integrity of steam generators in nuclear power plants is very much dependent upon the fretting wear characteristics of Inconel 690 U-tubes. In this study, a finite element model that can simulate fretting wear on the secondary side of the steam generator was developed and used for a quantitative investigation of the fretting wear phenomenon. Finite element modeling of elastic contact wear problems was performed to demonstrate the feasibility of applying the finite element method to fretting wear problems. The elastic beam problem, with existing solutions, is treated as a numerical example. By introducing a control parameter s, which scaled up the wear constant and scaled down the cycle numbers, the algorithm was shown to greatly reduce the time required for the analysis. The work rate model was adopted in the wear model. In the three-dimensional finite element analysis, a quarterly symmetric model was used to simulate cross tubes contacting at right angles. The wear constant of Inconel 690 in the work rate model was taken as
from experimental data obtained using a fretting wear test rig with a piezoelectric actuator. The analyses revealed donut-shaped wear along the contacting boundary, which is a typical feature of fretting wear.