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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 36, Issue 6 - Dec 2004
Volume 36, Issue 5 - Oct 2004
Volume 36, Issue 4 - Aug 2004
Volume 36, Issue 3 - Jun 2004
Volume 36, Issue 2 - Apr 2004
Volume 36, Issue 1 - Feb 2004
Selecting the target year
Transient Multicomponent Mixture Analysis Based On an ICE Numerical Technique for the Simulation of an Air Inggess Accident in an HTGR
Lim, Hong-Sik ; No, Hee-Cheon ;
Nuclear Engineering and Technology, volume 36, issue 5, 2004, Pages 375~387
This paper presents a transient multicomponent mixture analysis tool developed to analyze the molecular diffusion, natural convection, and chemical reactions related to air ingress phenomena that occur during a primary-pipe rupture of a high temperature gas-cooled reactor (HIGR). The present analysis tool solves the one-dimensional basic equations for continuity, momentum, energy of the gas mixture, and the mass of each gas species. In order to obtain numerically stable and fast computations, the implicit continuous Eulerian scheme is adopted to solve the governing equations in a strongly coupled manner. Two types of benchmark calculations were performed with the data of prerious Japanese inverse U-tube experiments. The analysis program, based on the ICE technique, runs about 36 times faster than the FLUENT6 for the simulation of the two experiments. The calculation results are within a 10% deviation from the experimental data regarding the concentrations of the gas species and the onset times of natural convection.
Phenomena Identification and Ranking Table for the APR-1400 Main Steam Line Break
Song, J.H. ; Chung, B.D. ; Jeong, J.J. ; Baek, W.P. ; Lee, S.Y. ; Choi, C.J. ; Lee, C.S. ; Lee, S.J. ; Um, K.S. ; Kim, H.G. ; Bang, Y.S. ;
Nuclear Engineering and Technology, volume 36, issue 5, 2004, Pages 388~402
A phenomena identification and ranking table(PIRT) was developed for a main steam line break (MSLB) event for the Advanced Power Reactor-1400 (APR-1400). The selectee event was a double-ended steam line break at full power, with the reactor coolant pump running. The developmental panel selected the fuel performance as the primary safety criterion during the ranking process. The plant design data, the results of the APR-1400 safety analysis, and the results of an additional best-estimate analysis by the MARS computer code were used in the development of the PIRT. The period of the transient was composed of three phases: pre-trip, rapid cool-down, and safety injection. Based on the relative importance to the primary evaluation criterion, the ranking of each system, component, and phenomenon/process was performed for each time phase. Finally, the knowledge-level for each important process for certain components was ranked in terms of existing knowledge. The PIRT can be used as a guide for planning cost-effective experimental programs and for code development efforts, especially for the quantification of those processes and/or phenomena that are highly important, but not well understood.
Heat Transfer Characteristics of an Internally-Heated Annulus Cooled with R-134a Near the Critical Pressure
Hong, Sung-Deok ; Chun, Se-Young ; Kim, Se-Yun ; Baek, Won-Pil ;
Nuclear Engineering and Technology, volume 36, issue 5, 2004, Pages 403~414
An experimental study of heat transfer characteristics near the critical pressure has been performed with an internally-heated vertical annular channel cooled by R-134a fluid. Two series of tests have been completed: (a) steady-state critical heat flux (CHF) tests, and (b) heat transfer tests for pressure reduction transients through the critical pressure. In the present experimental range, the steady-state CHF decreases with increase of the system pressure for fixed inlet mass flux and subcooling. The CHF falls sharply at about 3.8 MPa and shows a trend towards converging to zero as the pressure approaches the critical point of 4.059 MPa. The CHF phenomenon near the critical pressure does not lead to an abrupt temperature rise of the heated wall, because the CHF occurs at remarkably low power levels. In the pressure reduction transients, as soon as the pressure passes below the critical pressure from the supercritical pressure, the wall temperatures rise rapidly up to very high values due to the departure from nucleate boiling. The wall temperature reaches a maximum at the saturation point of the outlet temperature, and then tends to decrease gradually.
In-line Monitoring of an Oxide Ion in LiCI Molten Salt Using a YSZ Based Oxide Ion Selective Electrode
Cho, Young-Hwan ; Jeon, Jong-Seon ; Yeon, Jei-Won ; Choi, In-Kyu ; Kim, Won-Ho ;
Nuclear Engineering and Technology, volume 36, issue 5, 2004, Pages 415~419
The electrode potential characteristics of a YSZ based membrane metal oxide electrode have been studied in molten LiCL at
by the potentiometric method. The electrode exhibited a good potential response to log[
] and data reproducibility. The calibration plot (potential vs. log[
] was found to be linear, obeying the Nernst equation. The electrode potential showed a good reversibility corresponding to increase/decrease of the oxide ion present in the molten LiCl. The physical and chemical durability appeared to be sound after several repeated uses, resulting in reproducible results. However, "the proposed electrode" failed when metallic Li was present in the melt.
Estimation of the Nuclear Power Peaking Factor Using In-core Sensor Signals
Na, Man-Gyun ; Jung, Dong-Won ; Shin, Sun-Ho ; Lee, Ki-Bog ; Lee, Yoon-Joon ;
Nuclear Engineering and Technology, volume 36, issue 5, 2004, Pages 420~429
The local power density should be estimated accurately to prevent fuel rod melting. The local power density at the hottest part of a hot fuel rod, which is described by the power peaking factor, is more important information than the local power density at any other position in a reactor core. Therefore, in this work, the power peaking factor, which is defined as the highest local power density to the average power density in a reactor core, is estimated by fuzzy neural networks using numerous measured signals of the reactor coolant system. The fuzzy neural networks are trained using a training data set and are verified with another test data set. They are then applied to the first fuel cycle of Yonggwang nuclear power plant unit 3. The estimation accuracy of the power peaking factor is 0.45% based on the relative
error by using the fuzzy neural networks without the in-core neutron flux sensors signals input. A value of 0.23% is obtained with the in-core neutron flux sensors signals, which is sufficiently accurate for use in local power density monitoring.
A Quantitative Model of System-Man Interaction Based on Discrete Function Theory
Kim, Man-Cheol ; Seong, Poong-Hyun ;
Nuclear Engineering and Technology, volume 36, issue 5, 2004, Pages 430~449
A quantitative model for a control system that integrates human operators, systems, and their interactions is developed based on discrete functions. After identifying the major entities and the key factors that are important to each entity in the control system, a quantitative analysis to estimate the recovery failure probability from an abnormal state is performed. A numerical analysis based on assumed values of related variables shows that this model produces reasonable results. The concept of 'relative sensitivity' is introduced to identify the major factors affecting the reliability of the control system. The analysis shows that the hardware factor and the design factor of the instrumentation system have the highest relative sensitivities in this model. T도 probability of human operators performing incorrect actions, along with factors related to human operators, are also found to have high relative sensitivities. This model is applied to an analysis of the TMI-2 nuclear power plant accident and systematically explains how the accident took place.
Measurement of the Elemental Composition in Airborne Particulate Matter Using Instrumental Neutron Activation Analys
Chung, Yong-Sam ; Lim, Jong-Myoung ; Moon, Jong-Hwa ; Kim, Sun-Ha ; Cho, Hyun-Je ; Kim, Young-Jin ;
Nuclear Engineering and Technology, volume 36, issue 5, 2004, Pages 450~459
For the evaluation of emission sources by air sampling, airborne particulate matter for fine (<2.5
) and coarse partical (2.5-10
fractions were collected using a Gent stacked filter unit low volume sampler and two types of polycarbonate filters. Air samples were collected twice monthly at two regions in and around Daejeon city in the Republic of Korea from January to December 2002. Monthly mass concentration of
were measured and the concentrations of 10 marker elements (Al, Sc, Ti ; Na, Cl ; As, V. Sb, Br, Se) were determined by an instrumental neutron activation analysis. Analytical quality control was corried out using certified reference materials. Enrichment factors were also calculated from the monitoring data to classify the anthropogenic and crustal origins.
Effects of Pool Subcooling on Boiling Heat Transfer in an Annulus
Kang, Myeong-Gie ;
Nuclear Engineering and Technology, volume 36, issue 5, 2004, Pages 460~474
Effects of liquid subcooling on pool boiling heat transfer in an annulus with an open bottom have been investigated experimentally. A tube of 19.1mm diameter and the water at atmospheric pressure have been used for the fest. Up to
of liquid subcooling has been tested and experimental data of the annulus have been compared with the data of a single unrestricted tube. Temperatures on the heated tube surface fluctuate only slightly regardless of the heat flux in the annulus, whereas high variation is observed on the surface of the single tube. An increase in the degree of subcooling decreases heat transfer coefficients greatly both for the single tube and the annulus. Heat transfer coefficients increase suddenly at
and much greater change in heat transfer coefficients is observed at the annulus. To obtain effects of subcooling on heat transfer quantitatively, two new empirical equations have been suggested, and the correlations predict the empirical data within
error bound excluding some data at lower heat transfer coefficients.
Evaluation of Nuclear Plant Cable Aging Through Condition Monitoring
Kim, Jong-Seog ; Lee, Dong-Ju ;
Nuclear Engineering and Technology, volume 36, issue 5, 2004, Pages 475~484
Extending the lifetime of a nuclear power plant [(hereafter referred to simply as NPP)] is one of the most important concerns in the global nuclear industry. Cables are one of the long-life items that have not been considered for replacement during the design life of a NPP. To extend the cable life beyond the design life, it is first necessary to prove that the design life is too conservative compared with actual aging. Condition monitoring is useful means of evaluating the aging condition of cable. In order to simulate natural aging in a nuclear power plant. a study on accelerated aging must first be conducted. In this paper, evaluations of mechanical aging degradation for a neoprene cable jacket were performed after accelerated aging under tcontinuous and intermittent heating conditions. Contrary to general expectations, intermittent heating to the neoprene cable jacket showed low aging degradation, 50% break-elongation, and 60% indenter modulus, compared with continuous heating. With a plant maintenance period of 1 month after every 12 or 18 months operation, we can easily deduce that the life time of the cable jacket of neoprene can be extended much longer than extimated through the general EQ test. which adopts continuous accelerated aging for determining cable life. Therefore, a systematic approach that considers the actual environment conditions of the nuclear power plant is required for determining cable life.