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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 43, Issue 6 - Dec 2011
Volume 43, Issue 5 - Oct 2011
Volume 43, Issue 4 - Aug 2011
Volume 43, Issue 3 - Jun 2011
Volume 43, Issue 2 - Apr 2011
Volume 43, Issue 1 - Feb 2011
Selecting the target year
EFFECT OF SOLUBLE ADDITIVES, BORIC ACID (H
) AND SALT (NaCl), IN POOL BOILING HEAT TRANSFER
Kwark, Sang-M. ; Amaya, Miguel ; Moon, Hye-Jin ; You, Seung-M. ;
Nuclear Engineering and Technology, volume 43, issue 3, 2011, Pages 195~204
DOI : 10.5516/NET.2011.43.3.195
The effects on pool boiling heat transfer of aqueous solutions of boric acid (
) and sodium chloride (NaCl) as working fluids have been studied. Borated and NaCl water were prepared by dissolving 0.5~5% volume concentration of boric acid and NaCl in distilled-deionized water. The pool boiling tests were conducted using
flat heaters at 1 atm. The critical heat flux (CHF) dramatically increased compared to boiling pure water. At the end of boiling tests it was observed that particles of boric acid and NaCl had deposited and formed a coating on the heater surface. The CHF enhancement and surface modification during boiling tests were very similar to those obtained from boiling with nanofluids. Additional experiments were carried out to investigate the reliability of the additives deposition in pure water. The boric acid and NaCl coatings disappeared after repeated boiling tests on the same surface due to the soluble nature of the coatings, thus CHF enhancement no longer existed. These results demonstrate that not only insoluble nanoparticles but also soluble salts can be deposited during boiling process and the deposited layer is solely responsible for significant CHF enhancement.
THE EFFECT OF MICRO/NANOSCALE STRUCTURES ON CHF ENHANCEMENT
Ahn, Ho-Seon ; Kim, Moo-Hwan ;
Nuclear Engineering and Technology, volume 43, issue 3, 2011, Pages 205~216
DOI : 10.5516/NET.2011.43.3.205
Recently, many research studies have investigated the enormous critical heat flux (CHF) enhancement caused by nanofluids during pool boiling and flow boiling. One of the main reasons for this enhancement is nanoparticle deposition on the heated surface. However, in real applications, nanofluids create many problems when used as working fluids because of sedimentation and aggregation. Therefore, artificial surfaces on silicon and metal have been developed to create an effect similar to that of nanoparticle deposition. These modified surfaces have proved capable of greatly increasing the CHF during pool boiling, and good results have also been observed during flow boiling. In this study, we demonstrate that the wetting ability of a surface, i.e., wettability, and the liquid spreading ability (hydrophilic surface property), are key parameters for increasing the CHF during both pool and flow boiling. We also demonstrate that when the fuel surface in nuclear power plants is modified in a similar manner, it has the same effect, producing a large CHF enhancement.
NANOTECHNOLOGY FOR ADVANCED NUCLEAR THERMAL-HYDRAULICS AND SAFETY: BOILING AND CONDENSATION
Bang, In-Cheol ; Jeong, Ji-Hwan ;
Nuclear Engineering and Technology, volume 43, issue 3, 2011, Pages 217~242
DOI : 10.5516/NET.2011.43.3.217
A variety of Generation III/III+ water-cooled reactor designs featuring enhanced safety and improved economics are being proposed by nuclear power industries around the world in efforts to solve the future energy supply shortfall. Thermal-hydraulics is recognized as a key scientific subject in the development of innovative reactor systems. Phase change by boiling and condensation in the reverse process is a highly efficient heat transport mechanism that accommodates large heat fluxes with relatively small driving temperature differences. This mode of heat transfer is encountered in a wide spectrum of nuclear systems,and thus it is necessary to determine the thermal limit of water-cooled nuclear energy conversion in terms of economic and safety. Such applications are being advanced with the introduction of new technologies such as nanotechnology. Here, we investigated newly-introduced nanotechnologies relevant to boiling and condensation in general engineering applications. We also evaluated the potential linkage between such new advancements and nuclear applications in terms of advanced nuclear thermal-hydraulics.
ANALOG COMPUTING FOR A NEW NUCLEAR REACTOR DYNAMIC MODEL BASED ON A TIME-DEPENDENT SECOND ORDER FORM OF THE NEUTRON TRANSPORT EQUATION
Pirouzmand, Ahmad ; Hadad, Kamal ; Suh, Kune Y. ;
Nuclear Engineering and Technology, volume 43, issue 3, 2011, Pages 243~256
DOI : 10.5516/NET.2011.43.3.243
This paper considers the concept of analog computing based on a cellular neural network (CNN) paradigm to simulate nuclear reactor dynamics using a time-dependent second order form of the neutron transport equation. Instead of solving nuclear reactor dynamic equations numerically, which is time-consuming and suffers from such weaknesses as vulnerability to transient phenomena, accumulation of round-off errors and floating-point overflows, use is made of a new method based on a cellular neural network. The state-of-the-art shows the CNN as being an alternative solution to the conventional numerical computation method. Indeed CNN is an analog computing paradigm that performs ultra-fast calculations and provides accurate results. In this study use is made of the CNN model to simulate the space-time response of scalar flux distribution in steady state and transient conditions. The CNN model also is used to simulate step perturbation in the core. The accuracy and capability of the CNN model are examined in 2D Cartesian geometry for two fixed source problems, a mini-BWR assembly, and a TWIGL Seed/Blanket problem. We also use the CNN model concurrently for a typical small PWR assembly to simulate the effect of temperature feedback, poisons, and control rods on the scalar flux distribution.
MAJOR THERMAL-HYDRAULIC PHENOMENA FOUND DURING ATLAS LBLOCA REFLOOD TESTS FOR AN ADVANCED PRESSURIZED WATER REACTOR APR1400
Park, Hyun-Sik ; Choi, Ki-Yong ; Cho, Seok ; Kang, Kyoung-Ho ; Kim, Yeon-Sik ;
Nuclear Engineering and Technology, volume 43, issue 3, 2011, Pages 257~270
DOI : 10.5516/NET.2011.43.3.257
A set of reflood tests has been performed using ATLAS, which is a thermal-hydraulic integral effect test facility for the pressurized water reactors of APR1400 and OPR1000. Several important phenomena were observed during the ATLAS LBLOCA reflood tests, including core quenching, down-comer boiling, ECC bypass, and steam binding. The present paper discusses those four topics based on the LB-CL-11 test, which is a best-estimate simulation of the LBLOCA reflood phase for APR1400 using ATLAS. Both homogeneous bottom quenching and inhomogeneous top quenching were observed for a uniform radial power profile during the LB-CL-11 test. From the observation of the down-comer boiling phenomena during the LB-CL-11 test, it was found that the measured void fraction in the lower down-comer region was relatively smaller than that estimated from the RELAP5 code, which predicted an unrealistically higher void generation and magnified the downcomer boiling effect for APR1400. The direct ECC bypass was the dominant ECC bypass mechanism throughout the test even though sweep-out occurred during the earlier period. The ECC bypass fractions were between 0.2 and 0.6 during the later test period. The steam binding phenomena was observed, and its effect on the collapsed water levels of the core and down-comer was discussed.
FIRE PROPAGATION EQUATION FOR THE EXPLICIT IDENTIFICATION OF FIRE SCENARIOS IN A FIRE PSA
Lim, Ho-Gon ; Han, Sang-Hoon ; Moon, Joo-Hyun ;
Nuclear Engineering and Technology, volume 43, issue 3, 2011, Pages 271~278
DOI : 10.5516/NET.2011.43.3.271
When performing fire PSA in a nuclear power plant, an event mapping method, using an internal event PSA model, is widely used to reduce the resources used by fire PSA model development. Feasible initiating events and component failure events due to fire are identified to transform the fault tree (FT) for an internal event PSA into one for a fire PSA using the event mapping method. A surrogate event or damage term method is used to condition the FT of the internal PSA. The surrogate event or the damage term plays the role of flagging whether the system/component in a fire compartment is damaged or not, depending on the fire being initiated from a specified compartment. These methods usually require explicit states of all compartments to be modeled in a fire area. Fire event scenarios, when using explicit identification, such as surrogate or damage terms, have two problems: (1) there is no consideration of multiple fire propagation beyond a single propagation to an adjacent compartment, and (2) there is no consideration of simultaneous fire propagations in which an initiating fire event is propagated to multiple paths simultaneously. The present paper suggests a fire propagation equation to identify all possible fire event scenarios for an explicitly treated fire event scenario in the fire PSA. Also, a method for separating fire events was developed to make all fire events a set of mutually exclusive events, which can facilitate arithmetic summation in fire risk quantification. A simple example is given to confirm the applicability of the present method for a
rectangular fire area. Also, a feasible asymptotic approach is discussed to reduce the computational burden for fire risk quantification.
A STUDY ON THE AGING DEGRADATION OF ETHYLENE-PROPYLENE-DIENE MONOMER (EPDM) UNDER LOCA CONDITION
Seo, Yong-Dae ; Lee, Hyun-Seon ; Kim, Yong-Soo ; Song, Chi-Sung ;
Nuclear Engineering and Technology, volume 43, issue 3, 2011, Pages 279~286
DOI : 10.5516/NET.2011.43.3.279
The aging degradation and lifetime assessment of a domestic class 1E Ethylene-Propylene-Diene-Monomer (EPDM), which is a popular insulating elastomer for electrical cables in the nuclear power plants, were studied for equipment qualification verification under the Loss of Coolant Accident (LOCA) conditions. The specimens were acceleratively aged, underwent a LOCA environment, as well as tested mechanically, thermo-gravimetrically, and spectroscopically according to the American Society of the Testing of Materials (ASTM) procedures. The tensile test results revealed that the elongation at break gradually decreased with an increasing aging temperature. The lifetime of EPDM aged isothermally at
was 1,316 hours and reduced to 1,120 hours after experiencing the severe accident test. The activation energies of the elongation reduction were
eV before and after the LOCA condition, respectively. The TGA test results also showed that the activation energy of the aging decomposition decreased from 1.35 eV to 1.02 eV after undergoing the LOCA environment. Although the mechanical property changes were discernibly observed during the aging process, along with the LOCA simulation, the FT-IR analysis showed that the spectroscopic peaks and their intensities did not alter significantly. Therefore, it can be concluded that the degradation of the domestic class 1E EPDM due to aging can be tolerable, even in severe accident conditions such as LOCA, and thus it qualifies as a suitable insulating material for electrical cables in the nuclear power plants.
COUNTING STATISTICS MODIFIED BY TWO DEAD TIMES IN SERIES
Choi, H.D. ;
Nuclear Engineering and Technology, volume 43, issue 3, 2011, Pages 287~300
DOI : 10.5516/NET.2011.43.3.287
Counting statistics modified by introducing two dead times in series under a Poisson input distribution are discussed. A previous study examined the two cases of series combinations of nonextended-extended (NE-E) and extended-extended (EE) dead times. The present study investigated the remaining two cases of extended-nonextended (E-NE) and nonextended-nonextended (NE-NE) dead times. For the three time origins of the counting processes - ordinary, equilibrium, and shifted processes - a set of formulae was newly developed from a general formulation and presented for the event time interval densities, total densities, and exact mean and variance of the counts in a given counting duration. The asymptotic expressions for the mean and variance of the counts, which are most convenient for applications, were fully listed. The equilibrium mean count rates distorted by the three dead times in series were newly derived from the information obtained in these studies. An application of the derived formulae is briefly discussed.
DIAMETRAL CREEP PREDICTION OF THE PRESSURE TUBES IN CANDU REACTORS USING A BUNDLE POSITION-WISE LINEAR MODEL
Lee, Sung-Han ; Kim, Dong-Su ; Lee, Sim-Won ; No, Young-Gyu ; Na, Man-Gyun ; Lee, Jae-Yong ; Kim, Dong-Hoon ; Jang, Chang-Heui ;
Nuclear Engineering and Technology, volume 43, issue 3, 2011, Pages 301~308
DOI : 10.5516/NET.2011.43.3.301
The diametral creep of pressure tubes (PTs) in CANDU (CANada Deuterium Uranium) reactors is one of the principal aging mechanisms governing the heat transfer and hydraulic degradation of the heat transport system (HTS). PT diametral creep leads to diametral expansion, which affects the thermal hydraulic characteristics of the coolant channels and the critical heat flux (CHF). The CHF is a major parameter determining the critical channel power (CCP), which is used in the trip setpoint calculations of regional overpower protection (ROP) systems. Therefore, it is essential to predict PT diametral creep in CANDU reactors. PT diametral creep is caused mainly by fast neutron irradiation, temperature and applied stress. The objective of this study was to develop a bundle position-wise linear model (BPLM) to predict PT diametral creep employing previously measured PT diameters and HTS operating conditions. The linear model was optimized using a genetic algorithm and was devised based on a bundle position because it is expected that each bundle position in a PT channel has inherent characteristics. The proposed BPLM for predicting PT diametral creep was confirmed using the operating data of the Wolsung nuclear power plant in Korea. The linear model was able to predict PT diametral creep accurately.
BEHAVIORS OF MOLYBDENUM IN UO
Ha, Yeong-Keong ; Kim, Jong-Goo ; Park, Yang-Soon ; Park, Soon-Dal ; Song, Kyu-Seok ;
Nuclear Engineering and Technology, volume 43, issue 3, 2011, Pages 309~316
DOI : 10.5516/NET.2011.43.3.309
Molybdenum is the most abundant fission product since its fission yield is equivalent to that of xenon, and it has a very special role in the chemistry of nuclear fuel because it influences the oxygen potential of
fuel. In this study, the distribution of molybdenum in spent
fuel specimens with 33.3, 41.0 and 57.6 GWd/tU burnup was measured by a LA-ICP-MS system and the reproducibility of the measured data was obtained. The Mo distribution was almost constant along the radius of a fuel except an increase at the periphery of the fuel. It showed a drop in reproducibility with relatively high deviation of measured values for the highest burnup fuel. To explain this, the state of molybdenum in a
matrix and its effect on the oxidation behavior of
were investigated. The low reproducibility was explained by the segregation of molybdenum, and the inhibition of oxidation by the molybdenum was also observed.