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REFERENCE LINKING PLATFORM OF KOREA S&T JOURNALS
> Journal Vol & Issue
Nuclear Engineering and Technology
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Journal DOI :
Korean Nuclear Society
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Volume & Issues
Volume 30, Issue 6 - Dec 1998
Volume 30, Issue 5 - Oct 1998
Volume 30, Issue 4 - Aug 1998
Volume 30, Issue 3 - Jun 1998
Volume 30, Issue 2 - Apr 1998
Volume 30, Issue 1 - Feb 1998
Selecting the target year
Effects of Aperture Densitv Distribution on the Flow Through a Rock Fracture with Line-Source and Line-Collection
Park, Chung-Kyun ; Hahn, Pil-So ;
Nuclear Engineering and Technology, volume 30, issue 6, 1998, Pages 485~495
Migration characteristics of tracers in a rock fracture in a case of line-source and line-collection was studied. The fracture plane was discretized into a square mesh to which variable apertures were assigned. The spatially varying apertures of a fracture were generated using a geostatistical method, based on a given aperture probability density distribution and a specified spatial correlation length. The flow potential and pressure at each node were computed. Calculations showed that fluid flow occurs predominantly through a few preferred paths. Hence, the large range of apertures in the fracture gives rise to flow channeling. The solute transport was calculated using a particle tracking method. The migration plumes of tracer between injection line and withdrawal line are displayed in contour plots. The elution curves are shown to be controlled by the aperture density distribution and to be insensitive to statistical realization and spatial correlation length.
Turbulent Natural Convection in a Hemispherical Geometry Containing Internal Heat SourcesZ
Lee, Heedo ; Park, Goon-cherl ;
Nuclear Engineering and Technology, volume 30, issue 6, 1998, Pages 496~506
This paper deals with the computational modeling of buoyancy-driven turbulent heat transfer involving spatially uniform volumetric heat sources in semicircular geometry. The Launder & Sharma low-Reynolds number k-
turbulence model without any modifications and the SIMPLER computational algorithm were used for the numerical modeling, which was incorporated into the new computer code CORE-TNC. This computer code was subsequently benchmarked with the Mini-ACOPO experimental data in the modified Rayleigh number range of 2
. The general trends of the velocity and temperature fields were well predicted by the model used, and the calculated isotherm patterns were found to be very similiar to those observed in previous experimental investigations. The deviation between the Mini-ACOPO experimental data and the corresponding numerical results obtained with CORE-TNC for the average Nusselt number was less than 30% using fine grid in the near-wall region and the three-point difference formula for the wall temperature gradient. With isothermal pool boundaries, heat was convected predominantly to the upper and adjacent lateral surfaces, and the bottom surface received smaller heat fluxes.
A Calculation Model for Fuel Constituent Redistribution and Temperature Distribution on Metallic U-10Zr Fuel Slug of Liquid Metal Reactors
Nam, Cheol ; Hwang, Woan ;
Nuclear Engineering and Technology, volume 30, issue 6, 1998, Pages 507~517
Unlike conventional fuel types, fuel constituent redistribution and sodium intrusion into the fuel slug are the unique phenomena of the irradiated metallic fuel. A thermal calculation model on metallic U-10 wt.%Zr fuel rod for LMRs is developed with considerations given to these phenomena. The amount of constituent redistribution is estimated based on the thermotransport process. The temperature profile of fuel slug is predicted by taking into account of Zr redistribution, porosity formation and sodium logging effects. A sample calculation is performed and compared to experimental data in literature. As a result, the predicted redistribution and temperature profile are well agreed with experimental data, assuming that 15 times increment of ex-reactor diffusivity,
is -50 kJ/mole and sodium is infiltrated only outside of the fuel slug. Furthermore, the redistribution effects on fuel integrity and fuel temperature profile are discussed.d.
Chemical Methods Used in Petrological Analysis of Koongarra Uranium Ore Samples in ASSAR Natural Analogue Program
Park, Yong-Joon ; Pyo, Hyung-Ryul ; Kim, Ji-Young ; Kim, Won-Ho ;
Nuclear Engineering and Technology, volume 30, issue 6, 1998, Pages 518~530
A natural analogue study has been performed for the Koongarra uranium ore deposit in Australia as an international agreement of the Analogue Studies in the Alligator Rivers Region (ASARR). Rocks obtained from the Koongarra deposit, Northern Territory of Australia, were examined in order to understand uranium migration processes of primary and secondary ore-body in both weathered and unweathered zones. Total alpha activities of rock samples were measured to compare the relative amount of uranium in the sample. Uranium distributions have been investigated by means of both the alpha-autoradiography and the fission track registration technique after irradiation in a flux of thermal neutrons (~10
) for 2 minutes. The mineral phases corresponding to the registered alpha-tracks and fission tracks were identified by petrological observation with optical microscope as well as X-ray diffraction and electron microprobe analysis (EPMA). Uranium was found mostly inside of the fracture of the quartzite and its mineral phase was identified as sklodowskite. The mineral phase associated with high uranium concentration was found as illeminite by petrological observation with optical microscope as well as EPMA.
Implementation of DYLAM-3 to Core Uncovery Frequency Estimation in Mid-Loop Operation
Kim, Dohyoung ; Chang hyun Chung ; Moosung Jae ;
Nuclear Engineering and Technology, volume 30, issue 6, 1998, Pages 531~540
The DYLAM-3 code which overcomes the limitation of event tree/fault tree was applied to LOOP (Loss of Off-site Power) in the mid-loop operation employing HEPs (Human Error Probabilities) supplied by the ASEP (Accident Sequence Evaluation Program) and the SEPLOT (Systematic Evaluation Procedure for Low power/shutdown Operation Task) procedure in this study. Thus the time history of core uncovery frequency during the mid-loop operation was obtained. The sensitivity calculations in the operator's actions to prevent core uncovery under LOOP in the mid-loop operation were carried out. The analysis using the time dependent HEP was performed on the primary feed & bleed which has the most significant effect on core uncovery frequency. As the result, the increment of frequency is shown after 200 minutes duration of simulation conditions. This signifies the possibility of increment in risk after 200 minutes. The primary feed & bleed showed the greatest impact on core uncovery frequency and the recovery of the SCS (Shutdown Cooling System) showed the least impact. Therefore the efforts should be taken on the primary feed & bleed to reduce the core uncovery frequency in the mid-loop operation. And the capability of DYLAM-3 in applying to the time dependent concerns could be demonstrated.
COSMOS : A Computer Code for the Analysis of LWR
and MOX Fuel Rod
Koo, Yang-Hyun ; Lee, Byung-Ho ; Sohn, Dong-Seong ;
Nuclear Engineering and Technology, volume 30, issue 6, 1998, Pages 541~554
A computer code COSMOS has been developed based on the CARO-D5 for the thermal analysis of LWR UO
and MOX fuel rod under steady-state and transient operating conditions. The main purpose of the COSMOS, which considers high turnup characteristics such as thermal conductivity degradation with turnup and rim formation at the outer part of fuel pellet, is to calculate temperature profile across fuel pellet and fission gas release up to high burnup. A new mechanistic fission gas release model developed based on physical processes has been incorporated into the code. In addition, the features of MOX fuel such as change in themo-mechanical properties and the effect of microscopic heterogeneity on fission gas release have been also taken into account so that it can be applied to MOX fuel. Another important feature of the COSMOS is that it can analyze fuel segment refabricated from base irradiated fuel rods in commercial reactors. This feature makes it possible to analyze database obtained from international projects such as the MALDEN and RISO, many of which were collected from refabricated fuel segments. The capacity of the COSMOS has been tested with some number of experimental results obtained from the HALDEN, RISO and FIGARO programs. Comparison with the measured data indicates that, although the COSMOS gives reasonable agreement, the current models need to be improved. This work is being performed using database available from the OECD/NEA.
Flare Test Evaluation and Stress Prediction of PWR's Steam Generator Tubes
Woo-Gon Kim ; Chang Kyu Rhee ; Il-Hiun Kuk ;
Nuclear Engineering and Technology, volume 30, issue 6, 1998, Pages 555~567
Alloy 600 and 690 steam generator tubes fabricated in Korea were evaluated by flare tests according to ASTM standards. The stress acting in the tube elements during the tests was predicted. All the tubes, including alleys 600 and 690, satisfied the requirement of a 30% or 35% O.D expansion. Flow curves obtained from the flare test were found to be higher in alloy 690 tubes than in alloy 600 ones. The difference between alloy 600 and 690 tubes increased gradually with flaring percentage (F.P,%). An effective stress corresponding to mean yield stress was introduced and calculated. It showed that the prediction values were in good agreement with the measured ones for all the 690 and 600 alloy tubes. It became possible to predict the amount of acting stresses within tubes during expansion process.
An Experimental Study on Sodium-Concrete Reactions
Bae, Jae-Heum ; Shin, Min-Chul ; Min, Byong-Hun ; Kim, Su-Man ; Kim, Byong-Ho ; Kwon, Sang-Woon ; Hwang, Seong-Tae ;
Nuclear Engineering and Technology, volume 30, issue 6, 1998, Pages 568~580
A sodium-concrete reaction facility with a test chamber of 0.226㎥(
H0.8m) was constructed to carry out experiments of sodium-concrete reaction which might take place in sodium metal fast-breeder reactor Utilizing this facility, several experiments were conducted to closely examine the characteristics of sodium-concrete reactions under different conditions : Sodium mass : 100, 250g ; Sodium temperature : 450, 550,
; Concrete age = 30, 45, 50, 90days. Our experiments show that the amount of the H2 generated by sodium-concrete reaction has increased up to its flammable range as the amount of spilled sodium and its temperature have increased. The maximum hydrogen concentration was 31mo1% at the concrete age of 30days, sodium temperature : 55
, and sodium mass : 250g. The major components of sodium-concrete reaction products were also determined as Na
Evaluation of Thermal Stratification Effect in a Long Horizontal Pipeline with Turbulent Natural Convection
Park, Man-Heung ; Ahn, Jang-Sun ; Nam, Seung-Deog ;
Nuclear Engineering and Technology, volume 30, issue 6, 1998, Pages 581~591
Numerical analysis was peformed for the two-dimensional turbulent natural convection for a long horizontal line with different end temperatures. The turbulent model has been applied a standard k-
two equation model of turbulence similar to that the proposed by the Launder and Spalding. The dimensionless governing equations are solved by using SIMPLE (Semi-Implicit Method for Pressure Linked Equations) algorithm which is developed using control volumes and staggered grids. The numerical results are verified by comparison with the operating PWR test data. The analysis focuses on the effects of variation of the heat transfer rates at the pipe surface, the thermal conductivities of the pipe material and the thickness of the pipe wall on the thermal stratification. The results show that the heat transfer rate at the pipe surface is the controlling parameter for mitigating of thermal stratification in the long horizontal pipe. A significant reduction and disappearance of the thermal stratification phenomenon is observed at the Biot number of 4.82
. The results also show that the increment of the thermal conductivity and thickness of the wall weakens a little the thermal stratification and somewhat reduces temperature gradient of y-direction in the pipe wall. These effects are however minor, when compared with those due to the variation of the heat transfer rates at the surface of the pipe wall.
Several Problems in Reactor Coolant System Flow Rate Measurement
Ahn, Seung-Hoon ; Auh, Geun-Sun ; Suh, nam-Cuk ; Park, Jun-Sang ; Koo, Bon-Hyun ;
Nuclear Engineering and Technology, volume 30, issue 6, 1998, Pages 592~608
Inspection of RCS flow measurements for the domestic pressurized water reactors has been performed by the Korea Institute of Nuclear Safety (KINS) as one of the authorized periodical inspection activities. The inspection results for the Westinghouse-type plants reveal that 1) the RCS flow instrumentation has been calibrated by using the initial design and commissioning test result, without reflecting the cycle specific reference flow measurements, 2) the loop-to-loop now variation in the actual flow measurement which has not been considered in the safety analysis affects the asymmetric How transient results, and 3) the measured RCS flows in Kori 3 and 4, Yonggwang 1 and 2 do not support the definition of the best estimate RCS flow, approaching the RCS flow limit. In this study, the revealed problems were discussed with review of the design and the RCS flow measurement uncertainty evaluation, and the technical approaches and recommendations for resolving these problems were proposed.
Estimation of Thermal Aging Embrittlement of LWR Primary Pressure Boundary Components
Kim, Sunki ; Kim, Yongsoo ;
Nuclear Engineering and Technology, volume 30, issue 6, 1998, Pages 609~616
Cast duplex stainless steels are extensively used for primary pressure boundary components. These components are, however, embrittled due to the precipitation of
' phase by spinodal decomposition and other processes when exposed to reactor operating temperature for a design lifetime or life extension conditions. This report presents a procedure for estimating the current condition and the residual life of safety-related stainless steel components by using ANL database and correlations. The database of Charpy impact energy suggests that CF-8M grade is the most susceptible to thermal aging and CF-3 grade is the least. Thus, the integrity of CF-8M alleys may be degraded seriously and the degree of deterioration may exceed acceptance limit after several years of service in the nuclear reactors.
eXtended Statistical Combination of Uncertainties (XSCU) Method for Digital Nuclear Power Plants
In, Wang-Kee ; Hwang, Dae-Hyun ; Kim, Joon-Sung ; Auh, Geun-Sun ;
Nuclear Engineering and Technology, volume 30, issue 6, 1998, Pages 617~627
A technically more direct Statistical Combination of Uncertainties (SCU) method, extended SCU (XSCU), was developed to statistically combine the uncertainties associated with the DNBR alarm setpoint and the DNBR trip setpoint of digital nuclear power plants. The Modified SCU (MSCU) method is currently used as the USNRC approved design method to perform the same function. In this study, the MSCU and XSCU methods were compared in terms of the total uncertainties, and the thermal margins to the DNBR alarm and trip setpoints. The MSCU method resulted in small total uncertainties due to large negative biases which are unphysical. The XSCU method gives virtually unbiased total uncertainties which are physically meaningful in order to represent the actual magnitude of the total uncertainties associated with the DNBR alarm and trip setpoints. But the thermal margins to the DNBR alarm and trip setpoints by the MSCU method agree with those by the XSCU method within allowable statistical Variations.