Park H.,Korea Institute of Nuclear Safety |
Cho J.-Y.,Seoul National University
ACI Materials Journal | Year: 2017
This paper presents the effect of high-strength strands on the design of prestressed concrete (PSC) members; in particular, the tension-controlled and compression-controlled strain limits that are provided in the current ACI 318 code are discussed. The modification of the strain limits is based on a ductility analysis for which various ductility and deformability models were used so that the characteristics of the high-strength strands and the effect of the prestressing force could be considered. The ductility and deformability values were obtained from the moment-curvature relationships and the load-displacement relationships of the simply supported PSC beams. The material model of the Grade 1860 (270 ksi) strand was modified to represent the actual stress-strain behaviors of the high-strength strands. The effects of several design parameters such as cross-section type, compressive strength of the concrete, tensile strength and area of the strand, and level of the prestressing force are presented in detail. Copyright © 2017, American Concrete Institute. All rights reserved.
Kang D.G.,Korea Institute of Nuclear Safety
Annals of Nuclear Energy | Year: 2017
In best estimate (BE) calculation, the definition of system nodalization is important step influencing the prediction accuracy for specific thermal-hydraulic phenomena. The upper region of reactor is defined as the region of the upper guide structure (UGS) and upper dome. It has been assumed that the temperature of upper region is close to average temperature in most large break loss of coolant accident (LBLOCA) analysis cases. However, it was recently found that the temperature of upper region of APR-1400 reactor might be little lower than or similar to hot leg temperature through the review of detailed design data. In this study, the nodalization of APR-1400 was modified to reflect the characteristic of upper region temperature, and the effect of nodalization and temperature of reactor upper region on LBLOCA consequence was evaluated by sensitivity analysis including best estimate plus uncertainty (BEPU) calculation. In basecase calculation, in case of modified version, the peak cladding temperature (PCT) in blowdown phase became higher and the blowdown quenching (or cooling) was significantly deteriorated as compared to original case, and as a result, the cladding temperature in reflood phase became higher and the final quenching was also delayed. In addition, thermal-hydraulic parameters were compared and analyzed to investigate the effect of change of upper region on cladding temperature. In BEPU analysis, the 95 percentile PCT used in current regulatory practice was increased due to the modification of upper region nodalization, and it occurred in the reflood phase unlike original case. © 2016 Elsevier Ltd
Agency: European Commission | Branch: FP7 | Program: NoE | Phase: Fission-2008-2.1.1 | Award Amount: 39.59M | Year: 2009
Most of the actors involved in severe accident research in Europe, plus Canada, Korea and the United States (41 partners), will network in SARNET2 (Severe Accident Research NETwork of Excellence - Phase 2) their capacities of research in order to resolve important pending issues on postulated severe accidents of existing and future Nuclear Power Plants (NPPs). The project has been defined in order to optimise the use of the available means and to constitute a sustainable consortium in which common research programmes and a common computer tool to predict NPP behaviour during a postulated severe accident (ASTEC integral code) are developed. With this aim, the SARNET2 partners contribute to a Joint Programme of Activities, which consists of: - Maintaining and improving an advanced communication tool (developed during SARNET Phase 1) for accessing all project information, fostering exchange of information, and managing documents; - Harmonizing and re-orienting the research programmes, and defining new ones; - Performing experimental programmes on high priority issues, defined during SARNET Phase 1; - Analyzing experimental results in order to elaborate a common understanding of relevant phenomena; - Developing the ASTEC code (including its applicability to all types of European NPPs), which capitalizes in terms of physical models the knowledge produced within SARNET2; - Developing Scientific Databases, in which all the results of research programmes are stored in a common format (DATANET); - Developing education courses on severe accidents for students and researchers, and training courses for specialists; - Promoting personnel mobility amongst various European organizations; - Organizing yearly a large international conference on Severe Accident research (ERMSAR). After the first phase (2004-2008), and the four-year proposed second phase, co-funded by the EC, the network will evolve toward self-sustainability: a legal entity will be created.
Kang D.G.,Korea Institute of Nuclear Safety |
Ahn S.-H.,Korea Institute of Nuclear Safety
Annals of Nuclear Energy | Year: 2015
The thermal-hydraulic analysis using MARS-KS code was performed for 6-inch cold leg break test of ATLAS (Advanced Thermal-Hydraulic Test Loop for Accident Simulation), which was the second domestic standard problem. The calculation results were compared with experimental data to assess the code capability to simulate the transient thermal-hydraulic behavior for small break loss of coolant accident (SBLOCA). The sequence of events, except for the location of loop seal clearing (LSC) and safety injection tank (SIT) injection time was predicted well. The loop seals of 1A and 2B intermediate legs were cleared at 398 s in the experiment, while that of 1A was only cleared in the calculation at the same time. The prediction showed good agreement with the experimental data for pressurizer pressure and break mass flow rate. The sudden decrease and increase of water level at the LSC time were predicted qualitatively. After LSC, there was significant water level dip at SIT injection time which was not seen in the experiment. In addition, sensitivity study to investigate the cause of core level dip at SIT injection time was performed and discussions were made for it. In conclusion, MARS-KS code has good capabilities to simulate cold leg break SBLOCA, however, including interfacial heat and mass transfer, especially condensation model needs to be improved to predict more accurate results. © 2014 Elsevier Ltd. All rights reserved.
Jo J.C.,Korea Institute of Nuclear Safety
American Society of Mechanical Engineers, Pressure Vessels and Piping Division (Publication) PVP | Year: 2015
This study addresses a numerical analysis of the thermal-hydraulic response of the secondary side of a steam generator (SG) model with an internal structure to a main steam line break (MSLB) at a pressurized water reactor (PWR) plant. The analysis model is comprised of the SG upper space where steam occupies and the part of the main steam pipe between the SG outlet nozzle and the broken pipe end upstream of the main steam isolation valve. To investigate the effects of the presence of the SG internal structure on the thermal-hydraulic response to the MSLB, the numerical calculation results for the analysis model having a perforated horizontal plate as the SG internal structure are compared to those obtained for a simple analysis model having no SG internal structure. Both analysis models have the same physical dimensions except for the internal structure. The initial operating conditions for both SG models are identical to those for an actual operating plant. To simplify the analyses, it is assumed that steam is constantly generated from the bottom of the SG secondary side space during the blowdown process. As the results, it has been found that the pressure wave significantly attenuates as it passes through the perforated internal structure and as time elapses. This leads to reduction in instantaneous hydraulic load on the internal structure including tubing. However, it is seen that the presence of the internal structure does not affect the transient velocities of steam passing through the SG tube bundle during the blowdown, which are 2 to 8 times the velocities during the normal reactor operation as in the case for the empty SG. Consequently, the present findings should be considered for the design of the steam generator to ensure the reactor safety as such elevated high steam velocities can cause fluidelastic instability of tubes which results in high cycle fatigue failure of the tubes. Copyright © 2015 by ASME.
Lee S.-H.,Korea Institute of Nuclear Safety |
Choo Y.-W.,Korea Advanced Institute of Science and Technology |
Kim D.-S.,Korea Advanced Institute of Science and Technology
Soil Dynamics and Earthquake Engineering | Year: 2013
In dynamic centrifuge tests, appropriate boundary conditions are required in order to simulate the seismic semi-infinite soil layer responses within the confines of a finite size model container. An ESB (equivalent shear beam) model container first designed at the University of Cambridge was built with a stack of light-weight aluminum frames separated by rubber to experimentally achieve this goal. In this paper, a significant number of dynamic centrifuge tests and the corresponding seismic response analyses were performed to evaluate the dynamic performance of a newly constructed ESB model container and to shed light on the range of testable soil conditions. In the set of conducted tests, it appears that the end walls of the ESB model container behave in accordance with the dynamic response of the soil deposit, despite a difference in the natural period depending on the relative density of the sand deposit. This is attributed to the differences in mass and stiffness of the end walls compared to those of the contained soil model. For partially filled model container, significant differences in seismic responses are observed in the end walls and in the soil deposit due to seismic interaction caused by the upper unfilled frames of the container. These findings suggest that dynamic model tests using this ESB model container should be conducted with the container completely filled. In addition, on the basis of a comparison with the seismic soil behavior inside a rigid-walled model container, it is clear that the ESB model container can provide a more representative lateral boundary configuration for dynamic site response studies. © 2012 Elsevier Ltd.
Na C.,Korea Advanced Institute of Science and Technology |
Kim S.-P.,Korea Institute of Nuclear Safety |
Kwak H.-G.,Korea Advanced Institute of Science and Technology
Journal of Sound and Vibration | Year: 2011
The detection and identification of structural damage is important in monitoring of structural systems during their lifetime. Many researchers have proposed a variety of damage evaluation methods based on structural monitoring. The stiffness matrix is used in some conventional damage detection methods; however, it leads to inevitable error due to the lack of data provided by structural monitoring. To overcome this problem, this study introduces a new damage evaluation method that identifies the structural damage in a shear building based on a genetic algorithm using the structural flexibility matrix with dynamic analyses. The proposed method enables the deduction of the extent and location of structural damage, even when there is insufficient data on the dynamic characteristics and insufficient accurate measurements of the structural stiffness and mass. The validity of the proposed damage evaluation method is demonstrated through numerical analyses using OpenSees. Crown Copyright © 2011 Published by Elsevier Ltd. All rights reserved.
Kang D.G.,Korea Institute of Nuclear Safety
Annals of Nuclear Energy | Year: 2016
The best estimate (BE) calculation with uncertainty evaluation of large break loss of coolant accident (LBLOCA) has been increasingly applied to the licensing applications of nuclear power plants (NPPs). The KINS-realistic evaluation methodology (KINS-REM) was developed for the independent audit calculation based on the best estimate plus uncertainty (BEPU) method, and the code accuracy and statistical method have been improved. Power uprate has been implemented at a number of NPPs in many countries. It is generally categorized based on the increment of power and the method of increasing power. In this study, assuming the 4.5% stretch power uprate of typical three-loop nuclear power plant and corresponding design modifications, the LBLOCA was analyzed by applying the KINS-REM. The MARS-KS code was used as a frozen BE code, and 18 uncertainty parameters were considered in the analysis. The nodalization of the three-loop plant was defined by implementing specific separate and integral effect test nodalization scheme, and the thermal-hydraulic behavior in the system during the LBLOCA was analyzed through the basecase calculation. The 95 percentile peak cladding temperature (PCT) with 95% confidence level was determined by 124 calculations based on Wilks' formula of the non-parametric statistics, and additional biases for emergency core coolant (ECC) bypass and steam binding were added to account for untreatable phenomena and models. It was confirmed that the analysis results of the LBLOCA for the three-loop plant power uprate met the PCT acceptance criteria. © 2015 Elsevier Ltd. All rights reserved.
Jhung M.J.,Korea Institute of Nuclear Safety
Nuclear Engineering and Technology | Year: 2012
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.
Hong T.-K.,Yonsei University |
Choi H.,Korea Institute of Nuclear Safety
Tectonophysics | Year: 2012
The Korean Peninsula, eastern China and the Yellow Sea comprise the eastern Eurasian plate, and are believed to share considerable tectonic evolution history. The tectonic structures in the Yellow Sea are poorly understood, raising difficulty in reconstruction of tectonic evolution history in the eastern Eurasian plate. The tectonic structures in the Yellow Sea are constrained by seismicity and fault-plane solutions of earthquakes. The fault-plane solutions are determined by waveform inversions and seismic phase polarity analyses. The ambient stress fields are deduced from the fault-plane solutions. The primary stress field around the Yellow Sea is composed of ENE-WSW directional compression and NNW-SSE directional tension. Normal-faulting earthquakes with ENE-WSW directional strikes are observed in the central Yellow Sea between the Shandong Peninsula and the central Korean Peninsula. The normal-faulting region is interpreted to be a northern margin of collision belt between the North and South China blocks. The normal-faulting system suggests post-collisional lithospheric delamination, causing reverse activation of paleo-thrustal faults that were developed by the collision between the North and South China blocks in the early Jurassic period. © 2012 Elsevier B.V.