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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 41, Issue 10 - Dec 2009
Volume 41, Issue 9 - Nov 2009
Volume 41, Issue 8 - Oct 2009
Volume 41, Issue 7 - Sep 2009
Volume 41, Issue 6 - Aug 2009
Volume 41, Issue 5 - Jun 2009
Volume 41, Issue 4 - May 2009
Volume 41, Issue 3 - Apr 2009
Volume 41, Issue 2 - Mar 2009
Volume 41, Issue 1 - Feb 2009
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CURRENT STATUS AND IMPORTANT ISSUES ON SEISMIC HAZARD EVALUATION METHODOLOGY IN JAPAN
Ebisawa, Katsumi ;
Nuclear Engineering and Technology, volume 41, issue 10, 2009, Pages 1223~1234
DOI : 10.5516/NET.2009.41.10.1223
The outlines of seismic PSA implementation standards and seismic hazard evaluation procedure were shown. An overview of the cause investigation of seismic motion amplification on the Niigata-ken Chuetsu-oki (NCO) earthquake was also shown. Then, the contents for improving the seismic hazard evaluation methodology based on the lessons learned from the NCO earthquake were described. (1) It is very important to recognize the effectiveness of a fault model on the detail seismic hazard evaluation for the near seismic source through the cause investigation of the NCO earthquake. (2) In order to perform and proceed with a seismic hazard evaluation, the Japan Nuclear Energy Safety Organization has proposed the framework of the open deliberation rule regarding the treatment of uncertainty which was made so as to be able to utilize a logic tree. (3) The b-value evaluation on the "Stress concentrating zone," which is a high seismic activity around the NCO hypocenter area, should be modified based on the Gutenberg-Richter equation.
ISSUES IN PROBABILISTIC SEISMIC HAZARD ANALYSIS FOR NUCLEAR FACILITIES IN THE US
Mcguire, Robin K. ;
Nuclear Engineering and Technology, volume 41, issue 10, 2009, Pages 1235~1242
DOI : 10.5516/NET.2009.41.10.1235
Probabilistic seismic hazard analysis (PSHA) is routinely conducted in the US for nuclear plants, for the determination of appropriate seismic design levels. These analyses incorporate uncertainties in earthquake characteristics in stable continental regions (where direct observations of large earthquakes are rare), in estimates of rock motions, in site effects on strong shaking, and in the damage potential of seismic shaking for engineered facilities. Performance goals related to the inelastic deformation of individual components, and related to overall seismic core damage frequency, are used to determine design levels. PSHA has the ability to quantify and document the important uncertainties that affect seismic design levels, and future work can be guided toward reducing those uncertainties.
PROBABILISTIC SEISMIC HAZARD ANALYSIS FOR NUCLEAR POWER PLANTS - CURRENT PRACTICE FROM A EUROPEAN PERSPECTIVE
Klugel, Jens-Uwe ;
Nuclear Engineering and Technology, volume 41, issue 10, 2009, Pages 1243~1254
DOI : 10.5516/NET.2009.41.10.1243
The paper discusses the methodology and the use of probabilistic seismic hazard analysis (PSHA) for nuclear power plants from a European perspective. The increasing importance of risk-informed approaches in the nuclear oversight process observed in many countries has contributed to increasing attention to PSHA methods. Nevertheless significant differences with respect to the methodology of PSHA are observed in Europe. The paper gives an overview on actual projects and discusses the differences in the PSHA-methodology applied in different European countries. These differences are largely related to different approaches used for the treatment of uncertainties and to the use of experts. The development of a probabilistic scenario-based approach is identified as a meaningful alternative to the development of uniform hazard spectra or uniform confidence spectra.
STATUS OF THE PSHA IN KOREA FOR NUCLEAR POWER PLANT SITES
Seo, Jeong-Moon ; Noh, Myung-Hyun ; Chang, Chun-Joong ; Yun, Kwan-Hee ;
Nuclear Engineering and Technology, volume 41, issue 10, 2009, Pages 1255~1262
DOI : 10.5516/NET.2009.41.10.1255
This paper introduces the status of and issues related to the PSHA (Probabilistic Seismic Hazard Analysis) of Korean Nuclear Power Plant sites. PSHA was first introduced to the nuclear industry in the mid-1980s. The Korean PSHA is based on Cornell and accommodates the modem approach for eliciting expertise and statistical treatment. Due to the low seismicity in Korea, large uncertainties exist in the PSHA database including seismic source maps, seismicity parameters of seismic sources, and attenuation formulae. Though research in seismology, geology, and earthquake engineering since the mid-1990s has significantly reduced uncertainties, a considerable amount still exists. Considering the low seismicity of the Korean Peninsula, especially the lack of strong motion data, further reduction will take several decades.
CORE THERMAL HYDRAULIC BEHAVIOR DURING THE REFLOOD PHASE OF COLD-LEG LBLOCA EXPERIMENTS USING THE ATLAS TEST FACILITY
Cho, Seok ; Park, Hyun-Sik ; Choi, Ki-Yong ; Kang, Kyoung-Ho ; Baek, Won-Pil ; Kim, Yeon-Sik ;
Nuclear Engineering and Technology, volume 41, issue 10, 2009, Pages 1263~1274
DOI : 10.5516/NET.2009.41.10.1263
Several experimental tests to simulate a reflood phase of a cold-leg LBLOCA of the APR1400 have been performed using the ATLAS facility. This paper describes the related experimental results with respect to the thermal-hydraulic behavior in the core and the system-core interactions during the reflood phase of the cold-leg LBLOCA conditions. The present descriptions will be focused on the LB-CL-09, LB-CL-11, LB-CL-14, and LB-CL-15 tests performed using the ATLAS. The LB-CL-09 is an integral effect test with conservative boundary condition; the LB-CL-11 and -14 are integral effect tests with realistic boundary conditions, and the LB-CL-15 is a separated effect test. The objectives of these tests are to investigate the thermal-hydraulic behavior during an entire reflood phase and to provide reliable experimental data for validating the LBLOCA analysis methodology for the APR1400. The initial and boundary conditions were obtained by applying scaling ratios to the MARS simulation results for the LBLOCA scenario of the APR1400. The ECC water flow rate from the safety injection tanks and the decay heat were simulated from the start of the reflood phase. The simulated core power was controlled to be 1.2 times that of the ANS-73 decay heat curve for LB-CL-09 and 1.02 times that of the ANS-79 decay curve for LB-CL-11, -14, and -15. The simulated ECC water flow rate from the high pressure safety injection pump was 0.32 kg/s. The present experimental data showed that the cladding temperature behavior is closely related to the collapsed water level in the core and the downcomer.
NUMERICAL ANALYSIS OF A SO
PACKED COLUMN DECOMPOSITION REACTOR WITH ALLOY RA 330 STRUCTURAL MATERIAL FOR NUCLEAR HYDROGEN PRODUCTION USING THE SULFUR- IODINE PROCESS
Choi, Jae-Hyuk ; Tak, Nam-Il ; Shin, Young-Joon ; Kim, Chan-Soo ; Lee, Ki-Young ;
Nuclear Engineering and Technology, volume 41, issue 10, 2009, Pages 1275~1284
DOI : 10.5516/NET.2009.41.10.1275
A directly heated
decomposer for the sulfur-iodine and hybrid-sulfur processes has been introduced and analyzed using the computational fluid dynamics (CFD) code CFX 11. The use of a directly heated decomposition reactor in conjunction with a very high temperature reactor (VHTR) allows for higher decomposition reactor operating temperatures. However, the high temperatures and strongly corrosive operating conditions associated with
decomposition present challenges for the structural materials of decomposition reactors. In order to resolve these problems, we have designed a directly heated
decomposer using RA330 alloy as a structural material and have performed a CFD analysis of the design based on the finite rate chemistry model. The CFD results show the maximum temperature of the structural material could be maintained sufficiently below 1073 K, which is considered the target temperature for RA 330. The CFD simulations also indicated good performance in terms of
decomposition for the design parameters of the present study.
POOL BOILING HEAT TRANSFER IN A VERTICAL ANNULUS WITH A NARROWER UPSIDE GAP
Kang, Myeong-Gie ;
Nuclear Engineering and Technology, volume 41, issue 10, 2009, Pages 1285~1292
DOI : 10.5516/NET.2009.41.10.1285
The effects of the narrowed upside gap on nucleate pool boiling heat transfer in a vertical annulus were investigated experimentally. For the study, a stainless steel tube with a diameter of 25.4 mm and saturated water that kept an atmospheric condition were used. The ratio between the gaps measured at the upper and the lower regions of the annulus ranged from 0.18 to 1. Two different lengths of the modified gap also were investigated. The change in heat transfer due to the modified gap became evident as the gap ratio decreased and the length of the gap increased. As the gap ratio became less than 0.51, a significant decrease in heat transfer was observed compared to the plain annulus. The longer gap size resulted in an additional decrease in heat transfer. The major cause for the tendency was attributed to the formation of lumped bubbles around the upper region of the annulus followed by the increased flow friction between the fluid and the surface around the modified gap.
MODELLING THE DYNAMICS OF THE LEAD BISMUTH EUTECTIC EXPERIMENTAL ACCELERATOR DRIVEN SYSTEM BY AN INFINITE IMPULSE RESPONSE LOCALLY RECURRENT NEURAL NETWORK
Zio, Enrico ; Pedroni, Nicola ; Broggi, Matteo ; Golea, Lucia Roxana ;
Nuclear Engineering and Technology, volume 41, issue 10, 2009, Pages 1293~1306
DOI : 10.5516/NET.2009.41.10.1293
In this paper, an infinite impulse response locally recurrent neural network (IIR-LRNN) is employed for modelling the dynamics of the Lead Bismuth Eutectic eXperimental Accelerator Driven System (LBE-XADS). The network is trained by recursive back-propagation (RBP) and its ability in estimating transients is tested under various conditions. The results demonstrate the robustness of the locally recurrent scheme in the reconstruction of complex nonlinear dynamic relationships.
DETERMINATION OF BURNUP AND PU/U RATIO OF PWR SPENT FUELS BY GAMMA-RAY SPECTROMETRY
Park, Kwang-June ; Ju, June-Sik ; Kim, Jung-Suk ; Shin, Hee-Sung ; Chun, Yong-Bum ; Kim, Ho-Dong ;
Nuclear Engineering and Technology, volume 41, issue 10, 2009, Pages 1307~1314
DOI : 10.5516/NET.2009.41.10.1307
The isotope ratio of
in a spent PWR fuel sample was obtained with a newly developed gamma/neutron combined measuring system at KAERI. Burnup and Pu/U ratio of the spent fuel sample were determined by using the measured isotope ratio and the burnup-isotope ratio correlation equations calculated from the ORIGEN-ARP computer code. The results were compared and evaluated with the chemically determined burnup and Pu/U ratio. As a result of the comparative evaluation, the nondestructively determined burnup and Pu/U ratio values showed a good agreement with the chemically obtained results to within a 4.5% and 0.8% difference, respectively.
INTERPRETATION OF ELECTROCHEMICAL NOISE PARAMETERS AS INDICATORS OF INITIATION AND PROPAGATION OF SCC OF AN ALLOY 600 SG TUBE AT HIGH TEMPERATURES
Kim, Sung-Woo ; Kim, Hong-Pyo ;
Nuclear Engineering and Technology, volume 41, issue 10, 2009, Pages 1315~1322
DOI : 10.5516/NET.2009.41.10.1315
The present article is concerned with the application of an electrochemical noise (EN) monitoring technique to analyze the initiation and propagation of Pb-assisted stress corrosion cracking (SCC) of an Alloy 600 material in a simulated environment of a steam generator (SG) sludge pile at high temperatures. A typical increase of electrochemical current noise (ECN) and electrochemical potential noise (EPN) was frequently recorded from the EN measurement in a caustic solution with such impurities as PbO and CuO, indicating that there are localized corrosion events occurring. With the aid of microscopic and spectral analyses, the EN data involving information on such stochastic processes as uniform corrosion and the initiation and propagation of SCC, were analyzed based on a stochastic theory.
DESIGN STUDY OF AN IHX SUPPORT STRUCTURE FOR A POOL-TYPE SODIUM-COOLED FAST REACTOR
Park, Chang-Gyu ; Kim, Jong-Bum ; Lee, Jae-Han ;
Nuclear Engineering and Technology, volume 41, issue 10, 2009, Pages 1323~1332
DOI : 10.5516/NET.2009.41.10.1323
The IHX (Intermediate Heat eXchanger) for a pool-type SFR (Sodium-cooled Fast Reactor) system transfers heat from the primary high temperature sodium to the intermediate cold temperature sodium. The upper structure of the IHX is a coaxial structure designed to form a flow path for both the secondary high temperature and low temperature sodium. The coaxial structure of the IHX consists of a central downcomer and riser for the incoming and outgoing intermediate sodium, respectively. The IHX of a pool-type SFR is supported at the upper surface of the reactor head with an IHX support structure that connects the IHX riser cylinder to the reactor head. The reactor head is generally maintained at the low temperature regime, but the riser cylinder is exposed in the elevated temperature region. The resultant complicated temperature distribution of the co-axial structure including the IHX support structure may induce a severe thermal stress distribution. In this study, the structural feasibility of the current upper support structure concept is investigated through a preliminary stress analysis and an alternative design concept to accommodate the IHTS (Intermediate Heat Transport System) piping expansion loads and severe thermal stress is proposed. Through the structural analysis it is found that the alternative design concept is effective in reducing the thermal stress and acquiring structural integrity.
DYNAMIC CHARACTERISTICS OF CYLINDRICAL SHELLS CONSIDERING FLUID-STRUCTURE INTERACTION
Jhung, Myung-Jo ; Kim, Wal-Tae ; Ryu, Yong-Ho ;
Nuclear Engineering and Technology, volume 41, issue 10, 2009, Pages 1333~1346
DOI : 10.5516/NET.2009.41.10.1333
To assure the reliability of cylinders or shells with fluid-filled annulus, it is necessary to investigate the modal characteristics considering fluid-structure interaction effect. In this study, theoretical background and several finite element models are developed for cylindrical shells with fluid-filled annulus considering fluid-structure interaction. The effect of the inclusion of the fluid-filled annulus on the natural frequencies is investigated, which frequencies are used for typical dynamic analyses such as responses spectrum, power spectral density and unit load excitation. Their response characteristics are addressed with respect to the various representations of the fluid-structure interaction effect.
DEVELOPMENT OF THE MULTI-DIMENSIONAL HYDRAULIC COMPONENT FOR THE BEST ESTIMATE SYSTEM ANALYSIS CODE MARS
Bae, Sung-Won ; Chung, Bub-Dong ;
Nuclear Engineering and Technology, volume 41, issue 10, 2009, Pages 1347~1360
DOI : 10.5516/NET.2009.41.10.1347
A multi-dimensional component for the thermal-hydraulic system analysis code, MARS, was developed for a more realistic three-dimensional analysis of nuclear systems. A three-dimensional and two-fluid model for a two-phase flow in Cartesian and cylindrical coordinates was employed. The governing equations and physical constitutive relationships were extended from those of a one-dimensional version. The numerical solution method adopted a semi-implicit and finite-difference method based on a staggered-grid mesh and a donor-cell scheme. The relevant length scale was very coarse compared to commercial computational fluid dynamics tools. Thus a simple Prandtl's mixing length turbulence model was applied to interpret the turbulent induced momentum and energy diffusivity. Non drag interfacial forces were not considered as in the general nuclear system codes. Several conceptual cases with analytic solutions were chosen and analyzed to assess the fundamental terms. RPI air-water and UPTF 7 tests were simulated and compared to the experimental data. The simulation results for the RPI air-water two-phase flow experiment showed good agreement with the measured void fraction. The simulation results for the UPTF downcomer test 7 were compared to the experiment data and the results from other multi-dimensional system codes for the ECC delivery flow.
HUMAN ERRORS DURING THE SIMULATIONS OF AN SGTR SCENARIO: APPLICATION OF THE HERA SYSTEM
Jung, Won-Dea ; Whaley, April M. ; Hallbert, Bruce P. ;
Nuclear Engineering and Technology, volume 41, issue 10, 2009, Pages 1361~1374
DOI : 10.5516/NET.2009.41.10.1361
Due to the need of data for a Human Reliability Analysis (HRA), a number of data collection efforts have been undertaken in several different organizations. As a part of this effort, a human error analysis that focused on a set of simulator records on a Steam Generator Tube Rupture (SGTR) scenario was performed by using the Human Event Repository and Analysis (HERA) system. This paper summarizes the process and results of the HERA analysis, including discussions about the usability of the HERA system for a human error analysis of simulator data. Five simulated records of an SGTR scenario were analyzed with the HERA analysis process in order to scrutinize the causes and mechanisms of the human related events. From this study, the authors confirmed that the HERA was a serviceable system that can analyze human performance qualitatively from simulator data. It was possible to identify the human related events in the simulator data that affected the system safety not only negatively but also positively. It was also possible to scrutinize the Performance Shaping Factors (PSFs) and the relevant contributory factors with regard to each identified human event.