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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 46, Issue 6 - Dec 2014
Volume 46, Issue 5 - Oct 2014
Volume 46, Issue 4 - Aug 2014
Volume 46, Issue 3 - Jun 2014
Volume 46, Issue 2 - Apr 2014
Volume 46, Issue 1 - Feb 2014
Selecting the target year
SEISMIC ISOLATION OF NUCLEAR POWER PLANTS
Whittaker, Andrew S. ; Kumar, Manish ; Kumar, Manish ;
Nuclear Engineering and Technology, volume 46, issue 5, 2014, Pages 569~580
DOI : 10.5516/NET.09.2014.715
Seismic isolation is a viable strategy for protecting safety-related nuclear structures from the effects of moderate to severe earthquake shaking. Although seismic isolation has been deployed in nuclear structures in France and South Africa, it has not seen widespread use because of limited new build nuclear construction in the past 30 years and a lack of guidelines, codes and standards for the analysis, design and construction of isolation systems specific to nuclear structures. The funding by the United States Nuclear Regulatory Commission of a research project to the Lawrence Berkeley National Laboratory and MCEER/University at Buffalo facilitated the writing of a soon-to-be-published NUREG on seismic isolation. Funding of MCEER by the National Science Foundation led to research products that provide the technical basis for a new section in ASCE Standard 4 on the seismic isolation of safety-related nuclear facilities. The performance expectations identified in the NUREG and ASCE 4 for seismic isolation systems, and superstructures and substructures are described in the paper. Robust numerical models capable of capturing isolator behaviors under extreme loadings, which have been verified and validated following ASME protocols, and implemented in the open source code OpenSees, are introduced.
A SEISMIC DESIGN OF NUCLEAR REACTOR BUILDING STRUCTURES APPLYING SEISMIC ISOLATION SYSTEM IN A HIGH SEISMICITY REGION -A FEASIBILITY CASE STUDY IN JAPAN
Kubo, Tetsuo ; Yamamoto, Tomofumi ; Sato, Kunihiko ; Jimbo, Masakazu ; Imaoka, Tetsuo ; Umeki, Yoshito ;
Nuclear Engineering and Technology, volume 46, issue 5, 2014, Pages 581~594
DOI : 10.5516/NET.09.2014.716
A feasibility study on the seismic design of nuclear reactor buildings with application of a seismic isolation system is introduced. After the Hyogo-ken Nanbu earthquake in Japan of 1995, seismic isolation technologies have been widely employed for commercial buildings. Having become a mature technology, seismic isolation systems can be applied to NPP facilities in areas of high seismicity. Two reactor buildings are discussed, representing the PWR and BWR buildings in Japan, and the application of seismic isolation systems is discussed. The isolation system employing rubber bearings with a lead plug positioned (LRB) is examined. Through a series of seismic response analyses using the so-named standard design earthquake motions covering the design basis earthquake motions obtained for NPP sites in Japan, the responses of the seismic isolated reactor buildings are evaluated. It is revealed that for the building structures examined herein: (1) the responses of both isolated buildings and isolating LRBs fulfill the specified design criteria; (2) the responses obtained for the isolating LRBs first reach the ultimate condition when intensity of motion is 2.0 to 2.5 times as large as that of the design-basis; and (3) the responses of isolated reactor building fall below the range of the prescribed criteria.
SEISMIC ISOLATION OF LEAD-COOLED REACTORS: THE EUROPEAN PROJECT SILER
Forni, Massimo ; Poggianti, Alessandro ; Scipinotti, Riccardo ; Dusi, Alberto ; Manzoni, Elena ;
Nuclear Engineering and Technology, volume 46, issue 5, 2014, Pages 595~604
DOI : 10.5516/NET.09.2014.717
SILER (Seismic-Initiated event risk mitigation in LEad-cooled Reactors) is a Collaborative Project, partially funded by the European Commission in the
Framework Programme, aimed at studying the risk associated to seismic-initiated events in Generation IV Heavy Liquid Metal reactors, and developing adequate protection measures. The project started in October 2011, and will run for a duration of three years. The attention of SILER is focused on the evaluation of the effects of earthquakes, with particular regards to beyond-design seismic events, and to the identification of mitigation strategies, acting both on structures and components design. Special efforts are devoted to the development of seismic isolation devices and related interface components. Two reference designs, at the state of development available at the beginning of the project and coming from the
Framework Programme, have been considered: ELSY (European Lead Fast Reactor) for the Lead Fast Reactors (LFR), and MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) for the Accelerator-Driven Systems (ADS). This paper describes the main activities and results obtained so far, paying particular attention to the development of seismic isolators, and the interface components which must be installed between the isolated reactor building and the non-isolated parts of the plant, such as the pipe expansion joints and the joint-cover of the seismic gap.
EFFECTS OF MECHANICAL PROPERTY VARIABILITY IN LEAD RUBBER BEARINGS ON THE RESPONSE OF SEISMIC ISOLATION SYSTEM FOR DIFFERENT GROUND MOTIONS
Choun, Young-sun ; Park, Junhee ; Choi, In-Kil ;
Nuclear Engineering and Technology, volume 46, issue 5, 2014, Pages 605~618
DOI : 10.5516/NET.09.2014.718
The effects of variability of the mechanical properties of lead rubber bearings on the response of a seismic isolation system are investigated. Material variability in manufacturing, aging, and operation temperature is assumed, and two variation models of an isolation system are considered. To evaluate the effect of ground motion characteristics on the response, 27 earthquake record sets with different peak A/V ratios were selected, and three components of ground motions were used for a seismic response analysis. The response in an isolation system and a superstructure increases significantly for ground motions with low A/V ratios. The variation in the mechanical properties of isolators results in a significant influence on the shear strains of the isolators and the acceleration response of the superstructure. The variation provisions in the ASCE-4 are reasonable, but more strict variation limits should be given to isolation systems subjected to ground motions having low A/V ratios. For application of seismic isolation systems to safety-related nuclear structures, the variation in the material and mechanical properties of the isolation system should be properly controlled during the manufacturing and aging processes. In addition, special consideration should be given to minimize the accidental torsion caused by the dissimilarity in the stiffness variations of the isolators.
IMPLEMENTATION OF DATA ASSIMILATION METHODOLOGY FOR PHYSICAL MODEL UNCERTAINTY EVALUATION USING POST-CHF EXPERIMENTAL DATA
Heo, Jaeseok ; Lee, Seung-Wook ; Kim, Kyung Doo ;
Nuclear Engineering and Technology, volume 46, issue 5, 2014, Pages 619~632
DOI : 10.5516/NET.02.2013.083
The Best Estimate Plus Uncertainty (BEPU) method has been widely used to evaluate the uncertainty of a best-estimate thermal hydraulic system code against a figure of merit. This uncertainty is typically evaluated based on the physical model's uncertainties determined by expert judgment. This paper introduces the application of data assimilation methodology to determine the uncertainty bands of the physical models, e.g., the mean value and standard deviation of the parameters, based upon the statistical approach rather than expert judgment. Data assimilation suggests a mathematical methodology for the best estimate bias and the uncertainties of the physical models which optimize the system response following the calibration of model parameters and responses. The mathematical approaches include deterministic and probabilistic methods of data assimilation to solve both linear and nonlinear problems with the a posteriori distribution of parameters derived based on Bayes' theorem. The inverse problem was solved analytically to obtain the mean value and standard deviation of the parameters assuming Gaussian distributions for the parameters and responses, and a sampling method was utilized to illustrate the non-Gaussian a posteriori distributions of parameters. SPACE is used to demonstrate the data assimilation method by determining the bias and the uncertainty bands of the physical models employing Bennett's heated tube test data and Becker's post critical heat flux experimental data. Based on the results of the data assimilation process, the major sources of the modeling uncertainties were identified for further model development.
FLUID-STRUCTURE INTERACTION IN A U-TUBE WITH SURFACE ROUGHNESS AND PRESSURE DROP
Gim, Gyun-Ho ; Chang, Se-Myoung ; Lee, Sinyoung ; Jang, Gangwon ;
Nuclear Engineering and Technology, volume 46, issue 5, 2014, Pages 633~640
DOI : 10.5516/NET.02.2014.001
In this research, the surface roughness affecting the pressure drop in a pipe used as the steam generator of a PWR was studied. Based on the CFD (Computational Fluid Dynamics) technique using a commercial code named ANSYS-FLUENT, a straight pipe was modeled to obtain the Darcy frictional coefficient, changed with a range of various surface roughness ratios as well as Reynolds numbers. The result is validated by the comparison with a Moody chart to set the appropriate size of grids at the wall for the correct consideration of surface roughness. The pressure drop in a full-scale U-shaped pipe is measured with the same code, correlated with the surface roughness ratio. In the next stage, we studied a reduced scale model of a U-shaped heat pipe with experiment and analysis of the investigation into fluid-structure interaction (FSI). The material of the pipe was cut from the real heat pipe of a material named Inconel 690 alloy, now used in steam generators. The accelerations at the fixed stations on the outer surface of the pipe model are measured in the series of time history, and Fourier transformed to the frequency domain. The natural frequency of three leading modes were traced from the FFT data, and compared with the result of a numerical analysis for unsteady, incompressible flow. The corresponding mode shapes and maximum displacement are obtained numerically from the FSI simulation with the coupling of the commercial codes, ANSYS-FLUENT and TRANSIENT_STRUCTURAL. The primary frequencies for the model system consist of three parts: structural vibration, BPF(blade pass frequency) of pump, and fluid-structure interaction.
DEVELOPMENT OF A CORE THERMO-FLUID ANALYSIS CODE FOR PRISMATIC GAS COOLED REACTORS
Tak, Nam-Il ; Lee, Sung Nam ; Kim, Min-Hwan ; Lim, Hong Sik ; Noh, Jae Man ;
Nuclear Engineering and Technology, volume 46, issue 5, 2014, Pages 641~654
DOI : 10.5516/NET.02.2014.020
A new computer code, named CORONA (Core Reliable Optimization and thermo-fluid Network Analysis), was developed for the core thermo-fluid analysis of a prismatic gas cooled reactor. The CORONA code is targeted for whole-core thermo-fluid analysis of a prismatic gas cooled reactor, with fast computation and reasonable accuracy. In order to achieve this target, the development of CORONA focused on (1) an efficient numerical method, (2) efficient grid generation, and (3) parallel computation. The key idea for the efficient numerical method of CORONA is to solve a three-dimensional solid heat conduction equation combined with one-dimensional fluid flow network equations. The typical difficulties in generating computational grids for a whole core analysis were overcome by using a basic unit cell concept. A fast calculation was finally achieved by a block-wise parallel computation method. The objective of the present paper is to summarize the motivation and strategy, numerical approaches, verification and validation, parallel computation, and perspective of the CORONA code.
RECENT IMPROVEMENTS IN THE CUPID CODE FOR A MULTI-DIMENSIONAL TWO-PHASE FLOW ANALYSIS OF NUCLEAR REACTOR COMPONENTS
Yoon, Han Young ; Lee, Jae Ryong ; Kim, Hyungrae ; Park, Ik Kyu ; Song, Chul-Hwa ; Cho, Hyoung Kyu ; Jeong, Jae Jun ;
Nuclear Engineering and Technology, volume 46, issue 5, 2014, Pages 655~666
DOI : 10.5516/NET.02.2014.023
The CUPID code has been developed at KAERI for a transient, three-dimensional analysis of a two-phase flow in light water nuclear reactor components. It can provide both a component-scale and a CFD-scale simulation by using a porous media or an open media model for a two-phase flow. In this paper, recent advances in the CUPID code are presented in three sections. First, the domain decomposition parallel method implemented in the CUPID code is described with the parallel efficiency test for multiple processors. Then, the coupling of CUPID-MARS via heat structure is introduced, where CUPID has been coupled with a system-scale thermal-hydraulics code, MARS, through the heat structure. The coupled code has been applied to a multi-scale thermal-hydraulic analysis of a pool mixing test. Finally, CUPID-SG is developed for analyzing two-phase flows in PWR steam generators. Physical models and validation results of CUPID-SG are discussed.
KTM TOKAMAK OPERATION SCENARIOS SOFTWARE INFRASTRUCTURE
Pavlov, V. ; Baystrukov, K. ; Golobokov, Yu. ; Ovchinnikov, A. ; Mezentsev, A. ; Merkulov, S. ; Lee, A. ; Tazhibayeva, I. ; Shapovalov, G. ;
Nuclear Engineering and Technology, volume 46, issue 5, 2014, Pages 667~674
DOI : 10.5516/NET.04.2014.012
One of the largest problems for tokamak devices such as Kazakhstan Tokamak for Material Testing (KTM) is the operation scenarios' development and execution. Operation scenarios may be varied often, so a convenient hardware and software solution is required for scenario management and execution. Dozens of diagnostic and control subsystems with numerous configuration settings may be used in an experiment, so it is required to automate the subsystem configuration process to coordinate changes of the related settings and to prevent errors. Most of the diagnostic and control subsystems software at KTM was unified using an extra software layer, describing the hardware abstraction interface. The experiment sequence was described using a command language. The whole infrastructure was brought together by a universal communication protocol supporting various media, including Ethernet and serial links. The operation sequence execution infrastructure was used at KTM to carry out plasma experiments.
PARTICLE SIZE-DEPENDENT PULVERIZATION OF B
C AND GENERATION OF B
C/STS NANOPARTICLES USED FOR NEUTRON ABSORBING COMPOSITES
Kim, Jaewoo ; Jun, Jiheon ; Lee, Min-Ku ;
Nuclear Engineering and Technology, volume 46, issue 5, 2014, Pages 675~680
DOI : 10.5516/NET.06.2014.015
Pulverization of two different sized micro-
) was investigated using a STS based high energy ball milling system. Shapes, generation of the impurities, and reduction of the particle size dependent on milling time and initial particle size were investigated using various analytic tools including SEM-EDX, XRD, and ICP-MS. Most of impurity was produced during the early stage of milling, and impurity content became independent on the milling time after the saturation. The degree of particle size reduction was also dependent on the initial
size. It was found that the STS nanoparticles produced from milling is strongly bounded with the
particles forming the
/STS composite particles that can be used as a neutron absorbing nanocomposite. Based on the morphological evolution of the milled particles, a schematic pulverization model for the
particles was constructed.
HEAT-UP AND COOL-DOWN TEMPERATURE-DEPENDENT HYDRIDE REORIENTATION BEHAVIORS IN ZIRCONIUM ALLOY CLADDING TUBES
Won, Ju-Jin ; Kim, Myeong-Su ; Kim, Kyu-Tae ;
Nuclear Engineering and Technology, volume 46, issue 5, 2014, Pages 681~688
DOI : 10.5516/NET.07.2014.052
Hydride reorientation behaviors of PWR cladding tubes under typical interim dry storage conditions were investigated with the use of as-received 250 and 485ppm hydrogen-charged Zr-Nb alloy cladding tubes. In order to evaluate the effect of typical cool-down processes on the radial hydride precipitation, two terminal heat-up temperatures of 300 and
, as well as two terminal cool-down temperatures of 200 and
, were considered. In addition, two cooling rates of 2.5 and
during the cool-down processes were taken into account along with zero stress or a tensile hoop stress of 150MPa. It was found that the 250ppm hydrogen-charged specimen experiencing the higher terminal heat-up temperature and the lower terminal cool-down temperature generated the highest number of radial hydrides during the cool-down process under 150MPa hoop tensile stress, which may be explained by terminal solid hydrogen solubilities for precipitation, and dissolution and remaining circumferential hydrides at the terminal heat-up temperatures. In addition, the slower cool-down rate generates the larger number of radial hydrides due to a cooling rate-dependent, longer residence time at a relatively high temperature that can accelerate the radial hydride nucleation and growth.
MODAL CHARACTERISTIC ANALYSIS OF THE APR1400 NUCLEAR REACTOR INTERNALS FOR SEISMIC ANALYSIS
Park, Jong-Beom ; Choi, Youngin ; Lee, Sang-Jeong ; Park, No-Cheol ; Park, Kyoung-Su ; Park, Young-Pil ; Park, Chan-Il ;
Nuclear Engineering and Technology, volume 46, issue 5, 2014, Pages 689~698
DOI : 10.5516/NET.09.2014.017
Reactor internals are sensitive to dynamic loads such as earthquakes and flow induced vibration. Thus, it is essential to identify the dynamic characteristics to evaluate the seismic integrity of the structures. However, a full-sized system is too large to perform modal experiments, making it difficult to extract data on its modal characteristics. In this research, we constructed a finite element model of the APR1400 reactor internals to identify their modal characteristics. The commercial reactor was selected to reflect the actual boundary conditions. Our FE model was constructed based on scale-similarity analysis and fluid-structure interaction investigations using a fabricated scaled-down model.
PROBABILISTIC SEISMIC ASSESSMENT OF BASE-ISOLATED NPPS SUBJECTED TO STRONG GROUND MOTIONS OF TOHOKU EARTHQUAKE
Ali, Ahmer ; Hayah, Nadin Abu ; Kim, Dookie ; Cho, Ung Gook ;
Nuclear Engineering and Technology, volume 46, issue 5, 2014, Pages 699~706
DOI : 10.5516/NET.09.2014.030
The probabilistic seismic performance of a standard Korean nuclear power plant (NPP) with an idealized isolation is investigated in the present work. A probabilistic seismic hazard analysis (PSHA) of the Wolsong site on the Korean peninsula is performed by considering peak ground acceleration (PGA) as an earthquake intensity measure. A procedure is reported on the categorization and selection of two sets of ground motions of the Tohoku earthquake, i.e. long-period and common as Set A and Set B respectively, for the nonlinear time history response analysis of the base-isolated NPP. Limit state values as multiples of the displacement responses of the NPP base isolation are considered for the fragility estimation. The seismic risk of the NPP is further assessed by incorporation of the rate of frequency exceedance and conditional failure probability curves. Furthermore, this framework attempts to show the unacceptable performance of the isolated NPP in terms of the probabilistic distribution and annual probability of limit states. The comparative results for long and common ground motions are discussed to contribute to the future safety of nuclear facilities against drastic events like Tohoku.
KOREAN STUDENTS' BEHAVIORAL CHANGE TOWARD NUCLEAR POWER GENERATION THROUGH EDUCATION
Han, Eun Ok ; Kim, Jae Rok ; Choi, Yoon Seok ;
Nuclear Engineering and Technology, volume 46, issue 5, 2014, Pages 707~718
DOI : 10.5516/NET.10.2014.033
As a result of conducting a 45 minute-long seminar on the principles, state of use, advantages, and disadvantages of nuclear power generation for Korean elementary, middle, and high school students, the levels of perception including the necessity (p<0.017), safety (p<0.000), information acquisition (p<0.000), and subjective knowledge (p<0.000), objective knowledge (p<0.000), attitude (p<0.000), and behavior (p<0.000) were all significantly higher. This indicates that education can be effective in promoting widespread social acceptance of nuclear power and its continued use. In order to induce behavior change toward positive judgments on nuclear power generation, it is necessary to focus on attitude improvement while providing the information in all areas related to the perception, knowledge, attitude, and behavior. Here, the positive message on the convenience and the safety of nuclear power generation should be highlighted.
DEVELOPMENT OF A COMPUTER PROGRAM TO SUPPORT AN EFFICIENT NON-REGRESSION TEST OF A THERMAL-HYDRAULIC SYSTEM CODE
Lee, Jun Yeob ; Suh, Jaeseung ; Kim, Kyung Doo ; Jeong, Jae Jun ;
Nuclear Engineering and Technology, volume 46, issue 5, 2014, Pages 719~724
DOI : 10.5516/NET.02.2014.004
During the development process of a thermal-hydraulic system code, a non-regression test (NRT) must be performed repeatedly in order to prevent software regression. The NRT process, however, is time-consuming and labor-intensive. Thus, automation of this process is an ideal solution. In this study, we have developed a program to support an efficient NRT for the SPACE code and demonstrated its usability. This results in a high degree of efficiency for code development. The program was developed using the Visual Basic for Applications and designed so that it can be easily customized for the NRT of other computer codes.