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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 44, Issue 8 - Dec 2012
Volume 44, Issue 7 - Oct 2012
Volume 44, Issue 6 - Aug 2012
Volume 44, Issue 5 - Jun 2012
Volume 44, Issue 4 - May 2012
Volume 44, Issue 3 - Apr 2012
Volume 44, Issue 2 - Mar 2012
Volume 44, Issue 1 - Feb 2012
Selecting the target year
A SUMMARY OF 50
OECD/NEA/CSNI INTERNATIONAL STANDARD PROBLEM EXERCISE (ISP-50)
Choi, Ki-Yong ; Baek, Won-Pil ; Kang, Kyoung-Ho ; Park, Hyun-Sik ; Cho, Seok ; Kim, Yeon-Sik ;
Nuclear Engineering and Technology, volume 44, issue 6, 2012, Pages 561~586
DOI : 10.5516/NET.02.2012.708
This paper describes a summary of final prediction results by system-scale safety analysis codes during the OECD/NEA/CSNI ISP-50 exercise, targeting a 50% Direct Vessel Injection (DVI) line break integral effect test performed with the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS). This ISP-50 exercise has been performed in two consecutive phases: "blind" and "open" phases. Quantitative comparisons were performed using the Fast Fourier Transform Based Method (FFTBM) to compare the overall accuracy of the collected calculations. Great user effects resulting from the combination of the possible reasons were found in the blind phase, confirming that user effect is still one of the major issues in connection with the system thermal-hydraulic code application. Open calculations showed better prediction accuracy than the blind calculations in terms of average amplitude (AA) value. A total of nineteen organizations from eleven countries participated in this ISP-50 program and eight leading thermal-hydraulic system analysis codes were used: APROS, ATHLET, CATHARE, KORSAR, MARS-KS, RELAP5/MOD3.3, TECH-M-97, and TRACE.
INTERNATIONAL STANDARD PROBLEM 50: THE UNIVERSITY OF PISA CONTRIBUTION
Cherubini, Marco ; Lazzerini, Davide ; Giannotti, Walter ; D'auria, Francesco ;
Nuclear Engineering and Technology, volume 44, issue 6, 2012, Pages 587~596
DOI : 10.5516/NET.02.2012.709
The present paper deals with the participation of the University of Pisa in the last International Standard Problem (ISP) focused on system thermal hydraulic, which was led by the Korean Atomic Energy Research Institution (KAERI). The selected test was a Direct Vessel Injection (DVI) line break carried out at the ATLAS facility. University of Pisa participated, together with other eighteen institutions, in both blind and open phase of the analytical exercise pursuing its methodology for developing and qualifying a nodalization. Qualitative and quantitative analysis of the code results have been performed for both ISP-50 phases, the latter adopting the Fast Fourier Transfer Based Method (FFTBM). The experiment has been characterized by three-dimensional behavior in downcomer and core region. Even though an attempt to reproduce these phenomena, by developing a fictitious three-dimensional nodalization has been realized, the obtained results were generally acceptable but not fully satisfactory in replicating 3D behavior.
SEPARATE AND INTEGRAL EFFECT TESTS FOR VALIDATION OF COOLING AND OPERATIONAL PERFORMANCE OF THE APR+ PASSIVE AUXILIARY FEEDWATER SYSTEM
Kang, Kyoung-Ho ; Kim, Seok ; Bae, Byoung-Uhn ; Cho, Yun-Je ; Park, Yu-Sun ; Yun, Byoung-Jo ;
Nuclear Engineering and Technology, volume 44, issue 6, 2012, Pages 597~610
DOI : 10.5516/NET.02.2012.710
The passive auxiliary feedwater system (PAFS) is one of the advanced safety features adopted in the APR+, which is intended to completely replace the conventional active auxiliary feedwater system. With an aim of validating the cooling and operational performance of PAFS, an experimental program is in progress at KAERI, which is composed of two kinds of tests; the separate effect test and the integral effect test. The separate effect test, PASCAL (
ondensing Heat Removal
oop), is being performed to experimentally investigate the condensation heat transfer and natural convection phenomena in PAFS. A single, nearly-horizontal U-tube, whose dimensions are the same as the prototypic U-tube of the APR+ PAFS, is simulated in the PASCAL test. The PASCAL experimental result showed that the present design of PAFS satisfied the heat removal requirement for cooling down the reactor core during the anticipated accident transients. The integral effect test is in progress to confirm the operational performance of PAFS, coupled with the reactor coolant systems using the ATLAS facility. As the first integral effect test, an FLB (feedwater line break) accident was simulated for the APR+. From the integral effect test result, it could be concluded that the APR+ has the capability of coping with the hypothetical FLB accident by adopting PAFS and proper set-points of its operation.
SCALING ANALYSIS IN BEPU LICENSING OF LWR
D'auria, Francesco ; Lanfredini, Marco ; Muellner, Nikolaus ;
Nuclear Engineering and Technology, volume 44, issue 6, 2012, Pages 611~622
DOI : 10.5516/NET.02.2012.711
"Scaling" plays an important role for safety analyses in the licensing of water cooled nuclear power reactors. Accident analyses, a sub set of safety analyses, is mostly based on nuclear reactor system thermal hydraulics, and therefore based on an adequate experimental data base, and in recent licensing applications, on best estimate computer code calculations. In the field of nuclear reactor technology, only a small set of the needed experiments can be executed at a nuclear power plant; the major part of experiments, either because of economics or because of safety concerns, has to be executed at reduced scale facilities. How to address the scaling issue has been the subject of numerous investigations in the past few decades (a lot of work has been performed in the 80thies and 90thies of the last century), and is still the focus of many scientific studies. The present paper proposes a "roadmap" to scaling. Key elements are the "scaling-pyramid", related "scaling bridges" and a logical path across scaling achievements (which constitute the "scaling puzzle"). The objective is addressing the scaling issue when demonstrating the applicability of the system codes, the "key-to-scaling", in the licensing process of a nuclear power plant. The proposed "road map to scaling" aims at solving the "scaling puzzle", by introducing a unified approach to the problem.
A COMPARATIVE STUDY OF LATTICE BOLTZMANN AND VOLUME OF FLUID METHOD FOR TWO-DIMENSIONAL MULTIPHASE FLOWS
Ryu, Seung-Yeob ; Ko, Sung-Ho ;
Nuclear Engineering and Technology, volume 44, issue 6, 2012, Pages 623~638
DOI : 10.5516/NET.02.2011.025
The volume of fluid (VOF) model of FLUENT and the lattice Boltzmann method (LBM) are used to simulate two-phase flows. Both methods are validated for static and dynamic bubble test cases and then compared to experimental results. The VOF method does not reduce the spurious currents of the static droplet test and does not satisfy the Laplace law for small droplets at the acceptable level, as compared with the LBM. For single bubble flows, simulations are executed for various Eotvos numbers, Morton numbers and Reynolds numbers, and the results of both methods agree well with the experiments in the case of low Eotvos numbers. For high Eotvos numbers, the VOF results deviated from the experiments. For multiple bubbles, the bubble flow characteristics are related by the wake of the leading bubble. The coaxial and oblique coalescence of the bubbles are simulated successfully and the subsequent results are presented. In conclusion, the LBM performs better than the VOF method.
MODELING THE HYDRAULIC CHARACTERISTICS OF A FRACTURED ROCK MASS WITH CORRELATED FRACTURE LENGTH AND APERTURE: APPLICATION IN THE UNDERGROUND RESEARCH TUNNEL AT KAERI
Bang, Sang-Hyuk ; Jeon, Seok-Won ; Kwon, Sang-Ki ;
Nuclear Engineering and Technology, volume 44, issue 6, 2012, Pages 639~652
DOI : 10.5516/NET.02.2011.026
A three-dimensional discrete fracture network model was developed in order to simulate the hydraulic characteristics of a granitic rock mass at Korea Atomic Energy Research Institute (KAERI) Underground Research Tunnel (KURT). The model used a three-dimensional discrete fracture network (DFN), assuming a correlation between the length and aperture of the fractures, and a trapezoid flow path in the fractures. These assumptions that previous studies have not considered could make the developed model more practical and reasonable. The geologic and hydraulic data of the fractures were obtained in the rock mass at the KURT. Then, these data were applied to the developed fracture discrete network model. The model was applied in estimating the representative elementary volume (REV), the equivalent hydraulic conductivity tensors, and the amount of groundwater inflow into the tunnel. The developed discrete fracture network model can determine the REV size for the rock mass with respect to the hydraulic behavior and estimate the groundwater flow into the tunnel at the KURT. Therefore, the assumptions that the fracture length is correlated to the fracture aperture and the flow in a fracture occurs in a trapezoid shape appear to be effective in the DFN analysis used to estimate the hydraulic behavior of the fractured rock mass.
EVALUATION OF SAMG EFFECTIVENESS IN VIEW OF GROUP DECISION
Huh, Chang-Wook ; Suh, Nam-Duk ; Park, Goon-Cherl ;
Nuclear Engineering and Technology, volume 44, issue 6, 2012, Pages 653~662
DOI : 10.5516/NET.03.2010.045
We evaluate the technical and organizational aspects of the severe accident management guideline (SAMG), focusing on the decision-making process in the technical support center (TSC). From the technical aspects, we conclude that the present SAMG is a good tool that can assist the TSC in efficiently managing probable severe accidents. However, we suggest that the clear separation of the emergency operating procedure (EOP) and SAMG, which shifts plant control from the main control room (MCR) to the TSC, might not be an effective framework from an organizational perspective. Studies on organizational behavior demonstrate that a group decision made under a risky situation might be polarized in either a risky or cautious way. We recognize that we cannot be free from the polarization effect since the current SAMG recommends that the TSC evaluate the advantages and disadvantages of strategies to be implemented and choose the best one based on a group decision process. Illustrative examples of accident management under risky conditions are recapitulated from previous studies of the authors and we propose that the SAMG should be more proceduralized to remove this polarization from the decision-making process.
RISKY MODULE PREDICTION FOR NUCLEAR I&C SOFTWARE
Kim, Young-Mi ; Kim, Hyeon-Soo ;
Nuclear Engineering and Technology, volume 44, issue 6, 2012, Pages 663~672
DOI : 10.5516/NET.04.2011.023
As software based digital I&C (Instrumentation and Control) systems are used more prevalently in nuclear plants, enhancement of software dependability has become an important issue in the area of nuclear I&C systems. Critical attributes of software dependability are safety and reliability. These attributes are tightly related to software failures caused by faults. Software testing and V&V (Verification and Validation) activities are hence important for enhancing software dependability. If the risky modules of safety-critical software can be predicted, it will be possible to focus on testing and V&V activities more efficiently and effectively. It should also make it possible to better allocate resources for regulation activities. We propose a prediction technique to estimate risky software modules by adopting machine learning models based on software complexity metrics. An empirical study with various machine learning algorithms was executed for comparing the prediction performance. Experimental results show SVMs (Support Vector Machines) perform as well or better than the other methods.
THREE DIMENSIONAL ATOM PROBE STUDY OF NI-BASE ALLOY/LOW ALLOY STEEL DISSIMILAR METAL WELD INTERFACES
Choi, Kyoung-Joon ; Shin, Sang-Hun ; Kim, Jong-Jin ; Jung, Ju-Ang ; Kim, Ji-Hyun ;
Nuclear Engineering and Technology, volume 44, issue 6, 2012, Pages 673~682
DOI : 10.5516/NET.07.2012.018
Three dimensional atom probe tomography (3D APT) is applied to characterize the dissimilar metal joint which was welded between the Ni-based alloy, Alloy 690 and the low alloy steel, A533 Gr. B, with Alloy 152 filler metal. While there is some difficulty in preparing the specimen for the analysis, the 3D APT has a truly quantitative analytical capability to characterize nanometer scale particles in metallic materials, thus its application to the microstructural analysis in multi-component metallic materials provides critical information on the mechanism of nanoscale microstructural evolution. In this study, the procedure for 3D APT specimen preparation was established, and those for dissimilar metal weld interface were prepared near the fusion boundary by a focused ion beam. The result of the analysis in this study showed the precipitation of chromium carbides near the fusion boundary between A533 Gr. B and Alloy 152.
FATIGUE ANALYSIS OF A REACTOR PRESSURE VESSEL FOR SMART
Jhung, Myung-Jo ;
Nuclear Engineering and Technology, volume 44, issue 6, 2012, Pages 683~688
DOI : 10.5516/NET.09.2011.031
The structural integrity of mechanical components during several transients should be assured in the design stage. This requires a fatigue analysis including thermal and stress analyses. As an example, this study performs a fatigue analysis of the reactor pressure vessel of SMART during arbitrary transients. Using heat transfer coefficients determined based on the operating environments, a transient thermal analysis is performed and the results are applied to a finite element model along with the pressure to calculate the stresses. The total stress intensity range and cumulative fatigue usage factor are investigated to determine the adequacy of the design.
DESIGN SCOPE AND LEVEL FOR STANDARD DESIGN CERTIFICATION UNDER A TWO STEP LICENSING PROCESS
Suh, Nam-Duk ; Huh, Chang-Wook ;
Nuclear Engineering and Technology, volume 44, issue 6, 2012, Pages 689~696
DOI : 10.5516/NET.03.2011.030
A small integral reactor SMART (System Integrated Modular Advanced ReacTor), being developed in Korea since late 1990s and targeted to obtaining a standard design approval by the end of 2011, is introduced. The design scope and level for design certification (DC) is well described in the U.S. NRC SECY documents published the early 1990s. However, the documents are valid for a one-step licensing process called a combined operating license (COL) by the U.S. NRC, while Korea still uses a two-step licensing process. Thus, referencing the concept of the SECY documents, we have established the design scope and level for the SMART DC using the contexts of the standard review plan (SRP). Some examples of the results and issues raised during our review are briefly presented in this paper. The same methodology will be applied to other types of reactor under development in Korea, such as future VHTR reactors.
DEVELOPMENT OF RPS TRIP LOGIC BASED ON PLD TECHNOLOGY
Choi, Jong-Gyun ; Lee, Dong-Young ;
Nuclear Engineering and Technology, volume 44, issue 6, 2012, Pages 697~708
DOI : 10.5516/NET.04.2011.004
The majority of instrumentation and control (I&C) systems in today's nuclear power plants (NPPs) are based on analog technology. Thus, most existing I&C systems now face obsolescence problems. Existing NPPs have difficulty in repairing and replacing devices and boards during maintenance because manufacturers no longer produce the analog devices and boards used in the implemented I&C systems. Therefore, existing NPPs are replacing the obsolete analog I&C systems with advanced digital systems. New NPPs are also adopting digital I&C systems because the economic efficiencies and usability of the systems are higher than the analog I&C systems. Digital I&C systems are based on two technologies: a microprocessor based system in which software programs manage the required functions and a programmable logic device (PLD) based system in which programmable logic devices, such as field programmable gate arrays, manage the required functions. PLD based systems provide higher levels of performance compared with microprocessor based systems because PLD systems can process the data in parallel while microprocessor based systems process the data sequentially. In this research, a bistable trip logic in a reactor protection system (RPS) was developed using very high speed integrated circuits hardware description language (VHDL), which is a hardware description language used in electronic design to describe the behavior of the digital system. Functional verifications were also performed in order to verify that the bistable trip logic was designed correctly and satisfied the required specifications. For the functional verification, a random testing technique was adopted to generate test inputs for the bistable trip logic.