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REFERENCE LINKING PLATFORM OF KOREA S&T JOURNALS
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Nuclear Engineering and Technology
Journal Basic Information
Journal DOI :
Korean Nuclear Society
Editor in Chief :
Volume & Issues
Volume 34, Issue 6 - Dec 2002
Volume 34, Issue 5 - Oct 2002
Volume 34, Issue 4 - Aug 2002
Volume 34, Issue 3 - Jun 2002
Volume 34, Issue 2 - Apr 2002
Volume 34, Issue 1 - Feb 2002
Selecting the target year
Influence of Radioactive Contamination to Agricultural Products Due to Rain During a Nuclear Accident
Won Tae Hwang ; Eun Han Kim ; Kyung Suk Suh ; Moon Hee Han ; Han Soo Lee ;
Nuclear Engineering and Technology, volume 34, issue 5, 2002, Pages 415~420
The previous dynamic food chain model was improved for the consideration of the influence of radioactive contamination to agricultural products due to rain during the environmental releases of radionuclides in a nuclear accident Wet interception coefficients for the agricultural plants were derived as a function of radionuclide and rainfall amount, and mathematical formulations of the previous model were modified. As a result, rain during accidental releases was influential in agricultural contamination. The contamination level of agricultural products decreased dramatically according to increasing rainfall amount. It means that predictive concentrations in agricultural products using the previous model, in which dry interception to the agricultural plants is only considered, can be overestimated. The influence of rainfall in agricultural contamination was the most sensitive for
I, and the least sensitive for
Sr among the radionuclides considered in this study.
Direct ECC Bypass Phenomena in the MIDAS Test Facility During LBLOCA Reflood Phase
B.J. Yun ; T.S. Kwon ; D.J. Euh ; I.C. Chu ; Park, W.M. ; C.H. Song ; Park, J.K. ;
Nuclear Engineering and Technology, volume 34, issue 5, 2002, Pages 421~432
As one of the advanced design features of the APR1400, direct vessel injection (DVI) system is being considered instead of conventional cold leg injection (CLI) system. It is known that the DVI system greatly enhances the reliability of the emergency core cooling (ECC) system. However, there is still a dispute on its performance in terms of water delivery to the reactor core during the reflood phase of a large-break loss-of-coolant accident (LOCA). Thus, experimental validation is under progress. In this paper, test results of direct ECC bypass performed in the steam-water test facility tailed MIDAS (Multi-dimensional Investigation in Downcomer Annulus Simulation) are presented. The test condition is determined, based on the preliminary analysis of TRAC code, by applying the ‘modified linear scaling method’with the l/4.93 length scale . From the tests, ECC direct bypass fraction, steam condensation rate and information on the flow distribution in the upper annulus downcomer region are obtained.
Relation Between Density and Porosity in Sintered
Sang Ho Na ; Si Hyung Kim ; Young-Woo Lee ; Myung June Yoo ;
Nuclear Engineering and Technology, volume 34, issue 5, 2002, Pages 433~435
The relation between sintered densities and porosities in UO
pellets is investigated. The open porosity decreases linearly up to about 95% T.D.,(theoretical density) as the sintered density increases whereas, above 96% T.D., sintered UO
pellets do not have any open pores. The fraction of open porosity to the total porosity also decreases linearly as the sintered density increases, though the slope is lower than that of open porosity and, above 95% T.D., the fraction decreases rapidly to approach a zero.
Robust Controller Design of Nuclear Power Reactor by Parametric Method
Yoon-Joon Lee ; Man-Gyun Na ;
Nuclear Engineering and Technology, volume 34, issue 5, 2002, Pages 436~444
The robust controller for the nuclear reactor power control system is designed. Since the reactor model is not exact, it is necessary to design the robust controller that can work in the real situations of perturbations. The reactor model is described in the form of transfer function and the bound of each coefficient is determined to set up the linear interval system. By the Kharitonov and the edge theorem, a frequency based design template is made and applied to the determination of the controller. The controller designed by this method is simpler than that obtained by the H
. Although the controller is designed with the basis of high power, it could be used even at low power.n at low power.
An Experimental Study on the Sorption of U(VI) onto Granite
Min-Hoon Baik ; Pil-Soo Hahn ;
Nuclear Engineering and Technology, volume 34, issue 5, 2002, Pages 445~454
The sorption of U(Vl) on a domestic granite is studied as a function of experimental conditions such as contact time, solution-solid ratio, ionic strength, and pH using a batch procedure. The distribution coefficients,
＇s, of U(VI) are about 1-100mL/g depending on the experimental conditions. The sorption of U(VI) onto granite particles is greatly dependent upon the contact time, solution-solid ratio, and pH, but very little is dependent on the ionic strength. It is noticed that an U(VI)-carbonate ternary surface complex can be formed in the neutral range of pH. In the alkaline range of pH above 7, U(VI) sorption onto granite particles is greatly decreased due to the formation of anionic U(VI)-carbonate aqueous complexes.s.
Modifications and Assessment of RELAP5/MOD3.2 for HANARO Thermal-Hydraulic Safety Analyses
Gee Yang Han ; Kwi Seok Ha ;
Nuclear Engineering and Technology, volume 34, issue 5, 2002, Pages 455~467
RELAP5/MOD3.2 was modified to perform the thermal-hydraulic safety analysis for HANARO transients. Several aspects of RELAP5/MOD3.2 were modified or replaced by new features to properly simulate the unique HANARO characteristics such as the finned fuel element, the cooling mechanisms by both plate type heat exchanger and the natural circulation. Especially, the heat transfer packages were modified to be more appropriate for the safety analysis and the heat transfer models were developed for the plate type heat exchanger as well as natural circulation through the pool water. This modified version of RELAP5/MOD3.2 is renamed as RELAP5/HANARO. The thermal-hydraulic simulations of the single fuel pin test and plate type heat exchanger were peformed to assess the realistic predicting capabilities of RELAP5/HANARO and compared with experimental results and manufacturer＇s data in this paper. In addition, the natural circulation experiment using the scaled bundle was simulated to validate the capability of RELAP5/HANARO. The simulation results show almost similar trend with experimental data. Therefore, it is proved that RELAP5/HANARO has a confidence to use for the safety analyses of HANARO.
FARE Device Operational Characteristics of Remote Controlled Fuelling Machine at Wolsong NPP
I. Namgung ; Lee, S.K. ; Kim, Y.B. ;
Nuclear Engineering and Technology, volume 34, issue 5, 2002, Pages 468~481
There are 4 CANDU6 type reactors operating at Wolsong site. For fuelling operation of certain fuel channels (with flow less than 21.5 kg/s) a FARE flow Assist Ram Extension) device is used. During the refuelling operation, two remote controlled F/Ms (Fuelling Machines) are attached to a designated fuel channel and carry out refuelling job. The upstream F/M inserts new fuel bundles into the fuel channel while the downstream F/M discharges spent fuel bundles. In order to assist fuelling operation of channels that has lower coolant How rate, the FARE device is used instead of F/M C-ram to push the fuel bundle string. The FARE device is essentially a How restricting element that produces enough drag force to push the fuel bundle string toward downstream F/M. Channels that require the use of FARE device for refuelling are located along the outside perimeter of reactor. This paper presents the FARE device design feature, steady state hydraulic and operational characteristics and behavior of the device when coupled with fuel bundle string during fuelling operation. The study showed that the steady state performance of FARE device meets the design objective that was confirmed by downstream F/M C-ram force to be positive.
A Thermal Conductivity Model for LWR MOX Fuel and Its Verification Using In-pile Data
Byung-Ho Lee ; Yang-Hyun Koo ; Jin-Silk Cheon ; Je-Yong Oh ; Hyung-Koo Joo ; Dong-Seong Sohn ;
Nuclear Engineering and Technology, volume 34, issue 5, 2002, Pages 482~493
The MOX fuel for LWR is fabricated either by direct mechanical blending of UO
or by two stage mixing. Hence Pu-rich particles, whose Pu concentrations are higher than pellet average one and whose size distribution depends on a specific fabrication method, are inevitably dispersed in MOX pellet. Due to the inhomogeneous microstructure of MOX fuel, the thermal conductivity of LWR MOX fuel scatters from 80 to 100 % of UO
fuel. This paper describes a mechanistic thermal conductivity model for MOX fuel by considering this inhomogeneous microstructure and presents an explanation for the wide scattering of measured MOX fuel＇s thermal conductivity. The developed model has been incorporated into a KAERI＇s fuel performance code, COSMOS, and then evaluated using the measured in-pile data for MOX fuel. The database used for verification consists of homogeneous MOX fuel at beginning-of-life and inhomogeneous MOX fuel at high turnup. The COSMOS code predicts the thermal behavior of MOX fuel well except for the irradiation test accompanying substantial fission gas release. The over-prediction with substantial fission gas release seems to suggest the need for the introduction of a recovery factor to a term that considers the burnup effect on thermal conductivity.
Development of Moving Alternating Magnetic Filter Using Permanent Magnet for Removal of Radioactive Corrosion Product from Nuclear Power Plant
M. C. Song ; Kim, S. I. ; Lee, K. J. ;
Nuclear Engineering and Technology, volume 34, issue 5, 2002, Pages 494~501
Radioactive Corrosion Products (CRUD) which are generated by the neutron activation of general corrosion products at the nuclear power plant are the major source of occupational radiation exposure. Most of the CRUD has a characteristic of showing strong ferrimagnetisms. Along with the new development and production of permanent magnet (rare earth magnet) which generates much stronger magnetic field than the conventional magnet, new type of magnetic filter that can separate CRUD efficiently and eventually reduce radiation exposure of personnel at nuclear power plant is suggested. This separator consists of inner and outer magnet assemblies, coolant channel and container surrounding the outer magnet assembly. The rotational motion of the inner and outer permanent magnet assemblies surrounding the coolant channel by driving motor system produces moving alternating magnetic fields in the coolant channel. The CRUD can be separated from the coolant by the moving alternating magnetic field. This study describes the results of preliminary experiment performed with the different flow rates of coolant and rotation velocities of magnet assemblies. This new magnetic filter shows better performance results of filtering the magnetite at coolant (water). How rates, rotating velocities of magnet assemblies and particle sizes turn out to be very important design parameters.
Analysis of Channel Flow Low During Fuelling Operation of Selected Fuel Channels at Wolsong NPP
I. Namgung ; Lee, S.K. ;
Nuclear Engineering and Technology, volume 34, issue 5, 2002, Pages 502~516
Wolsong NPP are CANDU6 type reactors and there are 4 CANDU6 type reactors in commercial operation. CANDU type reactors require on-power refuelling by two remote controlled F/Ms (Fuelling Machines). Most of channels, fuel bundles are float by channel coolant flow and move toward downstream, however in about 30% of channels the coolant flow are not sufficient enough to carry fuel bundles to downstream. For those channels a special device, FARE (Flow Assist Ram Extension) device, is used to create additional force to push fuel bundles. It has been showing that during fuelling operation of some channels the channel coolant flow rate is reduced below specified limit (80% of normal), and consequently trip alarm signal turns on. This phenomenon occurs on selected channels that are instrumented for the channel flow and required to use the FARE device for refuelling. Hence it is believed that the FARE device causes the problem. It is also suspected that other channels that do not use the FARE device for refuelling might also go into channel flow low state. The analysis revealed that the channel How low occurs as the FARE device is introduced into the core and disappears as the FARE device is removed from the core. This paper presented the FARE device behavior, detailed fuelling operation sequence with the FARE device and effect on channel flow low phenomena. The FARE device components design changes are also suggested, such as increasing the number or now holes in the tube and flow slots in the ring orifice.
Speciation and Solubility of Major Actinides Under the Deep Groundwater Conditions of Korea
Dong-Kwon Keum ; Min-Hoon Baik ; Pil-Soo Hahn ;
Nuclear Engineering and Technology, volume 34, issue 5, 2002, Pages 517~531
The speciation and solubility of Am, Np, Pu and U have been analyzed by means of the geochemical code MUGREM, under the chemical conditions of domestic deep groundwater, in order to support the preliminary safety assessment for a Korean HLW disposal concept. Under the conditions of groundwaters studied, the stable solid phase is AmOHC
(s) or Am(OH)
) or N
(am), and Pu(OH)
(am) for Am, U, Np, and Pu, respectively. The dominating aqueous species are as follows: the complexes of Am(III), Am(OH)
, the complexes of U(VI), U
, the complexes of Np(IV), Np(OH)
(aq) and Np(OH)
, and the complexes of Pu(IV), Pu(OH)
(aq) and Pu(OH)
. The calculated solubilities exist between 1.9E-10 and 1.3E-9 mol/L for Am, between 5.6E-6 and 1.2E-4 mol/L for U, between 3.1E-9 and 1.3E-8 mol/L for Np, and between 6.6E-10 and 2.4E-10 mol/L for Pu, depending on groundwater conditions. The present solubilities of each actinide agree well with the results of other studies obtained under similar conditions.s.