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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 40, Issue 7 - Dec 2008
Volume 40, Issue 6 - Oct 2008
Volume 40, Issue 5 - Aug 2008
Volume 40, Issue 4 - Jun 2008
Volume 40, Issue 3 - Apr 2008
Volume 40, Issue 2 - Mar 2008
Volume 40, Issue 1 - Feb 2008
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RISK-INFORMED REGULATION: HANDLING UNCERTAINTY FOR A RATIONAL MANAGEMENT OF SAFETY
Zio, Enrico ;
Nuclear Engineering and Technology, volume 40, issue 5, 2008, Pages 327~348
DOI : 10.5516/NET.2008.40.5.327
A risk-informed regulatory approach implies that risk insights be used as supplement of deterministic information for safety decision-making purposes. In this view, the use of risk assessment techniques is expected to lead to improved safety and a more rational allocation of the limited resources available. On the other hand, it is recognized that uncertainties affect both the deterministic safety analyses and the risk assessments. In order for the risk-informed decision making process to be effective, the adequate representation and treatment of such uncertainties is mandatory. In this paper, the risk-informed regulatory framework is considered under the focus of the uncertainty issue. Traditionally, probability theory has provided the language and mathematics for the representation and treatment of uncertainty. More recently, other mathematical structures have been introduced. In particular, the Dempster-Shafer theory of evidence is here illustrated as a generalized framework encompassing probability theory and possibility theory. The special case of probability theory is only addressed as term of comparison, given that it is a well known subject. On the other hand, the special case of possibility theory is amply illustrated. An example of the combination of probability and possibility for treating the uncertainty in the parameters of an event tree is illustrated.
PRA RESEARCH AND THE DEVELOPMENT OF RISK-INFORMED REGULATION AT THE U.S. NUCLEAR REGULATORY COMMISSION
Siu, Nathan ; Collins, Dorothy ;
Nuclear Engineering and Technology, volume 40, issue 5, 2008, Pages 349~364
DOI : 10.5516/NET.2008.40.5.349
Over the years, probabilistic risk assessment (PRA) research activities conducted at the U.S. Nuclear Regulatory Commission (NRC) have played an essential role in support of the agency's move towards risk-informed regulation. These research activities have provided the technical basis for NRC's regulatory activities in key areas; provided PRA methods, tools, and data enabling the agency to meet future challenges; supported the implementation of NRC's 1995 PRA Policy Statement by assessing key sources of risk; and supported the development of necessary technical and human resources supporting NRC's risk-informed activities. PRA research aimed at improving the NRC's understanding of risk can positively affect the agency's regulatory activities, as evidenced by three case studies involving research on fire PRA, human reliability analysis (HRA), and pressurized thermal shock (PTS) PRA. These case studies also show that such research can take a considerable amount of time, and that the incorporation of research results into regulatory practice can take even longer. The need for sustained effort and appropriate lead time is an important consideration in the development of a PRA research program aimed at helping the agency address key sources of risk for current and potential future facilities.
ANALYZING DYNAMIC FAULT TREES DERIVED FROM MODEL-BASED SYSTEM ARCHITECTURES
Dehlinger, Josh ; Dugan, Joanne Bechta ;
Nuclear Engineering and Technology, volume 40, issue 5, 2008, Pages 365~374
DOI : 10.5516/NET.2008.40.5.365
Dependability-critical systems, such as digital instrumentation and control systems in nuclear power plants, necessitate engineering techniques and tools to provide assurances of their safety and reliability. Determining system reliability at the architectural design phase is important since it may guide design decisions and provide crucial information for trade-off analysis and estimating system cost. Despite this, reliability and system engineering remain separate disciplines and engineering processes by which the dependability analysis results may not represent the designed system. In this article we provide an overview and application of our approach to build architecture-based, dynamic system models for dependability-critical systems and then automatically generate dynamic fault trees (DFT) for comprehensive, tool-supported reliability analysis. Specifically, we use the Architectural Analysis and Design Language (AADL) to model the structural, behavioral and failure aspects of the system in a composite architecture model. From the AADL model, we seek to derive the DFT(s) and use Galileo's automated reliability analyses to estimate system reliability. This approach alleviates the dependability engineering - systems engineering knowledge expertise gap, integrates the dependability and system engineering design and development processes and enables a more formal, automated and consistent DFT construction. We illustrate this work using an example based on a dynamic digital feed-water control system for a nuclear reactor.
REVIEW OF VARIOUS DYNAMIC MODELING METHODS AND DEVELOPMENT OF AN INTUITIVE MODELING METHOD FOR DYNAMIC SYSTEMS
Shin, Seung-Ki ; Seong, Poong-Hyun ;
Nuclear Engineering and Technology, volume 40, issue 5, 2008, Pages 375~386
DOI : 10.5516/NET.2008.40.5.375
Conventional static reliability analysis methods are inadequate for modeling dynamic interactions between components of a system. Various techniques such as dynamic fault tree, dynamic Bayesian networks, and dynamic reliability block diagrams have been proposed for modeling dynamic systems based on improvement of the conventional modeling methods. In this paper, we review these methods briefly and introduce dynamic nodes to the existing reliability graph with general gates (RGGG) as an intuitive modeling method to model dynamic systems. For a quantitative analysis, we use a discrete-time method to convert an RGGG to an equivalent Bayesian network and develop a software tool for generation of probability tables.
OPTIMIZED NUMERICAL ANNULAR FLOW DRYOUT MODEL USING THE DRIFT-FLUX MODEL IN TUBE GEOMETRY
Chun, Ji-Han ; Lee, Un-Chul ;
Nuclear Engineering and Technology, volume 40, issue 5, 2008, Pages 387~396
DOI : 10.5516/NET.2008.40.5.387
Many experimental analyses for annular film dryouts, which is one of the Critical Heat Flux (CHF) mechanisms, have been performed because of their importance. Numerical approaches must also be developed in order to assess the results from experiments and to perform pre-tests before experiments. Various thermal-hydraulic codes, such as RELAP, COBRATF, MARS, etc., have been used in the assessment of the results of dryout experiments and in experimental pre-tests. These thermal-hydraulic codes are general tools intended for the analysis of various phenomena that could appear in nuclear power plants, and many models applying these codes are unnecessarily complex for the focused analysis of dryout phenomena alone. In this study, a numerical model was developed for annular film dryout using the drift-flux model from uniform heated tube geometry. Several candidates of models that strongly affect dryout, such as the entrainment model, deposition model, and the criterion for the dryout point model, were tested as candidates for inclusion in an optimized annular film dryout model. The optimized model was developed by adopting the best combination of these candidate models, as determined through comparison with experimental data. This optimized model showed reasonable results, which were better than those of MARS code.
FAULT TREE ANALYSIS OF KNICS RPS SOFTWARE
Park, Gee-Yong ; Koh, Kwang-Yong ; Jee, Eunk-Young ; Seong, Poong-Hyun ; Kwon, Kee-Choon ; Lee, Dae-Hyung ;
Nuclear Engineering and Technology, volume 40, issue 5, 2008, Pages 397~408
DOI : 10.5516/NET.2008.40.5.397
This paper describes the application of a software fault tree analysis (FTA) as one of the analysis techniques for a software safety analysis (SSA) at the design phase and its analysis results for the safety-critical software of a digital reactor protection system, which is called the KNICS RPS, being developed in the KNICS (Korea Nuclear Instrumentation & Control Systems) project. The software modules in the design description were represented by function blocks (FBs), and the software FTA was performed based on the well-defined fault tree templates for the FBs. The SSA, which is part of the verification and validation (V&V) activities, was activated at each phase of the software lifecycle for the KNICS RPS. At the design phase, the software HAZOP (Hazard and Operability) and the software FTA were employed in the SSA in such a way that the software HAZOP was performed first and then the software FTA was applied. The software FTA was applied to some critical modules selected from the software HAZOP analysis.
PERFORMANCE EVALUATION OF U-Mo/Al DISPERSION FUEL BY CONSIDERING A FUEL-MATRIX INTERACTION
Ryu, Ho-Jin ; Kim, Yeon-Soo ; Park, Jong-Man ; Chae, Hee-Taek ; Kim, Chang-Kyu ;
Nuclear Engineering and Technology, volume 40, issue 5, 2008, Pages 409~418
DOI : 10.5516/NET.2008.40.5.409
Because the interaction layers that form between U-Mo particles and the Al matrix degrade the thermal properties of U-Mo/Al dispersion fuel, an investigation was undertaken of the undesirable feedback effect between an interaction layer growth and a centerline temperature increase for dispersion fuel. The radial temperature distribution due to interaction layer growth during irradiation was calculated iteratively in relation to changes in the volume fractions, the thermal conductivities of the constituents, and the oxide thickness with the burnup. The interaction layer growth, which is estimated on the basis of the temperature calculations, showed a reasonable agreement with the post-irradiation examination results of the U-Mo/Al dispersion fuel rods irradiated at the HANARO reactor. The U-Mo particle size was found to be a dominant factor that determined the fuel temperature during irradiation. Dispersion fuel with larger U-Mo particles revealed lower levels of both the interaction layer formation and the fuel temperature increase. The results confirm that the use of large U-Mo particles appears to be an effective way of mitigating the thermal degradation of U-Mo/Al dispersion fuel.
VISUAL MEASUREMENT METHOD USING A CIRCULAR GROOVE IMAGE FOR MEASURING INTERNAL DEFECTS OF PIPES IN NUCLEAR POWER PLANT
Choi, Young-Soo ; Jeong, Kyung-Min ; Lee, Sung-Uk ; Jung, Seung-Ho ;
Nuclear Engineering and Technology, volume 40, issue 5, 2008, Pages 419~428
DOI : 10.5516/NET.2008.40.5.419
During the overhaul period of nuclear power plants in Korea, an ROV(Remotely Operated Vehicle) enters the cold-leg pipes connected with the reactor to examine the state of the thermal sleeves and their positions in the safety injection nozzles. To measure the positions of the thermal sleeves or scratches with video images recorded during the examination, time-varying camera parameters should be known, such as the focal length and principal points used for the capturing each video image. In this paper, we propose a camera calibration and measurement scheme by using a single image containing two circular grooves of a cylindrical nozzle whose radius and distance are known.
DEPTH AND LAYOUT OPTIMIZATIONS OF A RADIOACTIVE WASTE REPOSITORY IN A DISCONTINUOUS ROCK MASS BASED ON A THERMOMECHANICAL MODEL
Kim, Jhin-Wung ; Koh, Yong-Kwon ; Bae, Dae-Seok ; Choi, Jong-Won ;
Nuclear Engineering and Technology, volume 40, issue 5, 2008, Pages 429~438
DOI : 10.5516/NET.2008.40.5.429
The objective of the present study is the depth and layout optimizations of a single layer, high level radioactive waste repository in a discontinuous rock mass with special joint set arrangements. A single layer repository model, considering variations in the repository depths, pitches, and tunnel spacings, is used to analyze the thermomechanical interaction behavior. It is assumed that the repository is constructed in saturated granite with joints; the PWR spent fuel in a disposal canister is installed in a deposition drift which is then sealed with compacted bentonite; and the backfill material is filled in the repository tunnel. The decay heat generated by the high level radioactive wastes governs the thermomechanical behavior of the near field rock mass of the repository. The temperature and displacement behavior of the repository is influenced more by the pitch variations than the tunnel spacing and repository depth. However, the stress behavior is influenced more by the repository depth variations than the pitch and tunnel spacing. For the final selection of the tunnel spacing, pitch, and repository depth, other aspects such as the nuclide migration through a groundwater flow path, construction costs, operation costs, and so on should be considered.