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REFERENCE LINKING PLATFORM OF KOREA S&T JOURNALS
> Journal Vol & Issue
Nuclear Engineering and Technology
Journal Basic Information
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
PRESENT DAY EOPS AND SAMG - WHERE DO WE GO FROM HERE?
Vayssier, George ;
Nuclear Engineering and Technology, volume 44, issue 3, 2012, Pages 225~236
DOI : 10.5516/NET.03.2012.700
The Fukushima-Daiichi accident shook the world, as a well-known plant design, the General Electric BWR Mark I, was heavily damaged in the tsunami, which followed the Great Japanese Earthquake of 11 March 2011. Plant safety functions were lost and, as both AC and DC failed, manoeuvrability of the plants at the site virtually came to a full stop. The traditional system of Emergency Operating Procedures (EOPs) and Severe Accident Management Guidelines (SAMG) failed to protect core and containment, and severe core damage resulted, followed by devastating hydrogen explosions and, finally, considerable radioactive releases. The root cause may not only have been that the design against tsunamis was incorrect, but that the defence against accidents in most power plants is based on traditional assumptions, such as Large Break LOCA as the limiting event, whereas there is no engineered design against severe accidents in most plants. Accidents beyond the licensed design basis have hardly been considered in the various designs, and if they were included, they often were not classified for their safety role, as most system safety classifications considered only design basis accidents. It is, hence, time to again consider the Design Basis Accident, and ask ourselves whether the time has not come to consider engineered safety functions to mitigate core damage accidents. Associated is a proper classification of those systems that do the job. Also associated are safety criteria, which so far are only related to 'public health and safety'; in reality, nuclear accidents cause few casualties, but create immense economical and societal effects-for which there are no criteria to be met. Severe accidents create an environment far surpassing the imagination of those who developed EOPs and SAMG, most of which was developed after Three Mile Island - an accident where all was still in place, except the insight in the event was lost. It requires fundamental changes in our present safety approach and safety thinking and, hence, also in our EOPs and SAMG, in order to prevent future 'Fukushimas'.
CORIUM BEHAVIOR IN THE LOWER PLENUM OF THE REACTOR VESSEL UNDER IVR-ERVC CONDITION: TECHNICAL ISSUES
Park, Rae-Joon ; Kang, Kyoung-Ho ; Hong, Seong-Wan ; Kim, Sang-Baik ; Song, Jin-Ho ;
Nuclear Engineering and Technology, volume 44, issue 3, 2012, Pages 237~248
DOI : 10.5516/NET.03.2012.701
Corium behavior in the lower plenum of the reactor vessel during a severe accident is very important, as this affects a failure mechanism of the lower head vessel and a thermal load to the outer reactor vessel under the IVR-ERVC (In-Vessel corium Retention through External Reactor Vessel Cooling) condition. This paper discusses the state of the art and technical issues on corium behavior in the lower plenum, such as initial corium pool formation characteristics and its transient behavior, natural convection heat transfer in various geometries, natural convection heat transfer with a phase change of melting and solidification, and corium interaction with a lower head vessel including penetrations of the ICI (In-Core Instrumentation) nozzle are discussed. It is recommended that more detailed analysis and experiments are necessary to solve the uncertainties of corium behavior in the lower plenum of the reactor vessel.
COMBINED ANALYTICAL AND EXPERIMENTAL INVESTIGATIONS FOR LWR CONTAINMENT PHENOMENA
Allelein, Hans-Josef ; Reinecke, Ernst-Arndt ; Belt, Alexander ; Broxtermann, Philipp ; Kelm, Stephan ;
Nuclear Engineering and Technology, volume 44, issue 3, 2012, Pages 249~260
DOI : 10.5516/NET.03.2012.702
Main focus of the combined nuclear research activities at Aachen University (RWTH) and the Research Center J
LICH) is the experimental and analytical investigation of containment phenomena and processes. We are deeply convinced that reliable simulations for operation, design basis and beyond-design basis accidents of nuclear power plants need the application of so-called lumped-parameter (LP) based codes as well as computational fluid dynamics (CFD) codes in an indispensable manner. The LP code being used at our institutions is the GRS code COCOSYS and the CFD tool is ANSYS CFX mostly used in German nuclear research. Both codes are applied for safety analyses especially of beyond design accidents. Focal point of the work is containment thermal-hydraulics, but source term relevant investigations for aerosol and iodine behavior are performed as well. To increase the capability of COCOSYS and CFX detailed models for specific features, e.g. recombiner behavior including chimney effect, building condenser, and wall condensation are developed and validated against facilities at different scales. The close connection between analytical and experimental activities is notable and identifying feature of the RWTH/J
CONTRIBUTIONS OF THE VULCANO EXPERIMENTAL PROGRAMME TO THE UNDERSTANDING OF MCCI PHENOMENA
Christophe, Journeau ; Piluso, Pascal ; Correggio, Patricia ; Ferry, Lionel ; Fritz, Gerald ; Haquet, Jean Francois ; Monerris, Jose ; Ruggieri, Jean-Michel ; Sanchez-Brusset, Mathieu ; Parga, Clemente ;
Nuclear Engineering and Technology, volume 44, issue 3, 2012, Pages 261~272
DOI : 10.5516/NET.03.2012.703
Molten Core Concrete Interaction (MCCI) is a complex process characterized by concrete ablation and volatile generation; Thermal and solutal convection in a bubble-agitated melt; Physico-chemical evolution of the corium pool with a wide solidification range (of the order of 1000 K). Twelve experiments have been carried out in the VULCANO facility with prototypic corium and sustained heating. The dry oxidic corium tests have contributed to show that silica-rich concrete experience an anisotropic ablation. This unexpected ablation pattern is quite reproducible and can be recalculated, provided an empirical anisotropy factor is assumed. Dry tests with oxide and metal liquid phases have also yielded unexpected results: a larger than expected steel oxidation and unexpected topology of the metallic phase (at the bottom of the cavity and also on the vertical concrete walls). Finally, VULCANO has proved its interest for the study of mitigation solutions such as the COMET bottom flooding core catcher.
APPLICATION OF A GENETIC ALGORITHM FOR THE OPTIMIZATION OF ENRICHMENT ZONING AND GADOLINIA FUEL (UO
) ROD DESIGNS IN OPR1000s
Kwon, Tae-Je ; Kim, Jong-Kyung ;
Nuclear Engineering and Technology, volume 44, issue 3, 2012, Pages 273~282
DOI : 10.5516/NET.01.2011.047
A new effective methodology for optimizing the enrichment of low-enriched zones as well as gadolinia fuel (
) rod designs in PLUS7 fuel assemblies was developed to minimize the maximum peak power in the core and to maximize the cycle lifetime. An automated link code was developed to integrate the genetic algorithm (GA) and the core design code package of ALPHA/PHOENIX-P/ANC and to generate and evaluate the candidates to be optimized efficiently through the integrated code package. This study introduces an optimization technique for the optimization of gadolinia fuel rod designs in order to effectively reduce the peak powers for a few hot assemblies simultaneously during the cycle. Coupled with the gadolinia optimization, the optimum enrichments were determined using the same automated code package. Applying this technique to the reference core of Ulchin Unit 4 Cycle 11, the gadolinia fuel rods in each hot assembly were optimized to different numbers and positions from their original designs, and the maximum peak power was decreased by 2.5%, while the independent optimization technique showed a decrease of 1.6% for the same fuel assembly. The lower enrichments at the fuel rods adjacent to the corner gap (CG), guide tube (GT), and instrumentation tube (IT) were optimized from the current 4.1, 4.1, 4.1 w/o to 4.65, 4.2, 4.2 w/o. The increase in the cycle lifetime achieved through this methodology was 5 effective full-power days (EFPD) on an ideal equilibrium cycle basis while keeping the peak power as low as 2.3% compared with the original design.
NUMERICAL ANALYSIS FOR PRANDTL NUMBER DEPENDENCY ON NATURAL CONVECTION IN AN ENCLOSURE HAVING A VERTICAL THERMAL GRADIENT WITH A SQUARE INSULATOR INSIDE
Lee, Jae-Ryong ; Park, Il-Seouk ;
Nuclear Engineering and Technology, volume 44, issue 3, 2012, Pages 283~296
DOI : 10.5516/NET.02.2011.027
The natural convection in a horizontal enclosure heated from the bottom wall, cooled at the top wall, and having a square adiabatic body in the center is studied. Three different Prandtl numbers (0.01, 0.7 and 7) are considered for the investigation of the effect of the Prandtl number on natural convection. Adiabatic boundary conditions are employed for the side walls. A two-dimensional solution for unsteady natural convection is obtained, using an accurate and efficient Chebyshev spectral methodology for different Rayleigh numbers varying over the range of
. It had been experimentally reported that the heat transfer mode becomes oscillatory when Pr is out of a specific Pr band beyond the critical Ra. In this study, we reproduced this phenomenon numerically. It was found that when Ra=
, only the case for intermediate Pr (=0.7) reached a non-changing steady state and the low and high Pr number cases (Pr=0.01 and 7) showed a periodically oscillatory fashion hydrodynamically and thermally. The variation of time- and surface-averaged Nusselt numbers on the hot and cold walls for different Rayleigh numbers and Prandtl numbers are presented to show the overall heat transfer characteristics in the system. Further, the isotherms and streamline distributions are presented in detail to compare the physics related to their thermal behavior.
IDENTIFICATION AND ASSESSMENT OF AGING-RELATED DEGRADATION OCCURRENCES IN NUCLEAR POWER PLANTS
Choi, In-Kil ; Choun, Young-Sun ; Kim, Min-Kyu ; Nie, Jinsuo ; Braverman, Joseph I. ; Hofmayer, Charles H. ;
Nuclear Engineering and Technology, volume 44, issue 3, 2012, Pages 297~310
DOI : 10.5516/NET.03.2010.048
Aging-related degradation of nuclear power plant components is an important aspect to consider in securing the long term safety of the plant, especially the seismic safety, since the degradation of the components affects not only their seismic capacity but their response. This can cause a change in the seismic margin of a component and the overall seismic safety of a system. To better understand the status and characteristics of degradation of components in Nuclear Power Plants (NPPs), the degradation occurrences of components in the U.S. NPPs were identified by reviewing recent publicly available information sources and the characteristics of these occurrences were evaluated and compared to observations from the past. Ten categories of components that are of high risk significance in Korean NPPs were identified, comprising anchorage, concrete, containment, exchanger, filter, piping systems, reactor pressure vessels, structural steel, tanks, and vessels. Software tools were developed to expedite the review process. Results from this review effort were compared to previous data in the literature to characterize the overall degradation trends.
COMPARATIVE ANALYSIS OF STATION BLACKOUT ACCIDENT PROGRESSION IN TYPICAL PWR, BWR, AND PHWR
Park, Soo-Yong ; Ahn, Kwang-Il ;
Nuclear Engineering and Technology, volume 44, issue 3, 2012, Pages 311~322
DOI : 10.5516/NET.03.2011.046
Since the crisis at the Fukushima plants, severe accident progression during a station blackout accident in nuclear power plants is recognized as a very important area for accident management and emergency planning. The purpose of this study is to investigate the comparative characteristics of anticipated severe accident progression among the three typical types of nuclear reactors. A station blackout scenario, where all off-site power is lost and the diesel generators fail, is simulated as an initiating event of a severe accident sequence. In this study a comparative analysis was performed for typical pressurized water reactor (PWR), boiling water reactor (BWR), and pressurized heavy water reactor (PHWR). The study includes the summarization of design differences that would impact severe accident progressions, thermal hydraulic/severe accident phenomenological analysis during a station blackout initiated-severe accident; and an investigation of the core damage process, both within the reactor vessel before it fails and in the containment afterwards, and the resultant impact on the containment.
UNCERTAINTY ANALYSIS OF DATA-BASED MODELS FOR ESTIMATING COLLAPSE MOMENTS OF WALL-THINNED PIPE BENDS AND ELBOWS
Kim, Dong-Su ; Kim, Ju-Hyun ; Na, Man-Gyun ; Kim, Jin-Weon ;
Nuclear Engineering and Technology, volume 44, issue 3, 2012, Pages 323~330
DOI : 10.5516/NET.09.2011.032
The development of data-based models requires uncertainty analysis to explain the accuracy of their predictions. In this paper, an uncertainty analysis of the support vector regression (SVR) model, which is a data-based model, was performed because previous research showed that the SVR method accurately estimates the collapse moments of wall-thinned pipe bends and elbows. The uncertainty analysis method used in this study was an analytic uncertainty analysis method, and estimates with a 95% confidence interval were obtained for 370 test data points. From the results, the prediction interval (PI) was very narrow, which means that the predicted values are quite accurate. Therefore, the proposed SVR method can be used effectively to assess and validate the integrity of the wall-thinned pipe bends and elbows.
ASSESSMENT OF PROPERTIES AND DURABILITY OF FLY ASH CONCRETE USED IN KOREAN NUCLEAR POWER PLANTS
Cho, Myung-Sug ; Noh, Jae-Myoung ;
Nuclear Engineering and Technology, volume 44, issue 3, 2012, Pages 331~342
DOI : 10.5516/NET.09.2010.031
Since the opening of the Shin-Kori #1,2 in 2005, fly ash mixed concrete has been used for NPP concrete structures under construction in Korea with the aim of preventing aging and improving durability. In this paper, the quality suitability of fly ash manufactured in Korea is assessed and the basic physical properties of fly ash mixed concrete and its durability against primary causes of aging are verified through experimental methods. Because of the internal structure filling effect from the pozzolanic reaction of fly ash and the resulting improvements in mechanical performance in such areas as strength and salt damage resistance, the durability of fly ash mixed concrete is shown to be superior. It is judged that this result can be applied in measures not only for improving the safety of NPP structures in operation in Korea but also for implementing effective structure life management should extending the life of structures be needed in the future.
ASSESSMENT OF POSSIBILITY OF PRIMARY WATER STRESS CORROSION CRACKING OCCURRENCE BASED ON RESIDUAL STRESS ANALYSIS IN PRESSURIZER SAFETY NOZZLE OF NUCLEAR POWER PLANT
Lee, Kyoung-Soo ; Kim, W. ; Lee, Jeong-Geun ;
Nuclear Engineering and Technology, volume 44, issue 3, 2012, Pages 343~354
DOI : 10.5516/NET.09.2010.066
Primary water stress corrosion cracking (PWSCC) is a major safety concern in the nuclear power industry worldwide. PWSCC is known to initiate only in the condition in which sufficiently high tensile stress is applied to alloy 600 tube material or alloy 82/182 weld material in pressurized water reactor operating environments. However, it is still uncertain how much tensile stress is re-quired to generate PWSCC or what causes such high tensile stress. This study was performed to pre-dict the magnitude of weld residual stress and operating stress and compare it with previous experi-mental results for PWSCC initiation. For the study, a pressurizer safety nozzle was selected because it is reported to be vulnerable to PWSCC in overseas plants. The assessment was conducted by nu-merical analysis. Before performing stress analysis for plant conditions, a preliminary mock-up ana-lysis was done. The result of the preliminary analysis was validated by residual stress measurement in the mock-up. After verification of the analysis methodology, an analysis under plant conditions was conducted. The analysis results show that the stress level is not high enough to initiate PWSCC. If a plant is properly welded and operated, PWSCC is not likely to occur in the pressurizer safety nozzle.