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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
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Volume & Issues
Volume 25, Issue 4 - Dec 1993
Volume 25, Issue 3 - Sep 1993
Volume 25, Issue 2 - Jun 1993
Volume 25, Issue 1 - Mar 1993
Selecting the target year
A Proposed Heuristic Methodology for Searching Reloading Pattern
Park, K.Y. ; Y.K. Yoon ;
Nuclear Engineering and Technology, volume 25, issue 2, 1993, Pages 193~203
A new heuristic method for loading pattern search has been developed to overcome short-comings of the algorithmic approach. To reduce the size of vast solution space, general shuffling rules, a regionwise shuffling method, and a pattern grouping method were introduced. The entropy theory was applied to classify possible loading patterns into groups with similarity between them. The pattern search program was implemented with use of the PROLOG language. A two-group nodal code MEDIUM-2D was used for analysis of power distribution in the core. The above mentioned methodology has been tested to show effectiveness in reducing of solution space down to a few hundred pattern groups. Burnable poison rods were then arranged in each pattern group in accordance with burnable poison distribution rules, which led to further reduction of the solution space to several scores of acceptable pattern groups. The method of maximizing cycle length(MCL) and minimizing power-peaking factor(MPF) were applied to search for specific useful loading patterns from the acceptable pattern groups. Thus, several specific loading patterns that have low power-peaking factor and large cycle length were successfully searched from the selected pattern groups.
The Effect of Tributary Pipe Breaks on the Core Support Barrel Shell Responses
Jhung, Myung-Jo ; Hwan, Won-Gul ;
Nuclear Engineering and Technology, volume 25, issue 2, 1993, Pages 204~214
Work on fracture mechanics has provided a technical basis for elimination of main coolant loop double ended guillotine breaks from the structural design basis of reactor coolant system. Without main coolant loop pipe breaks, the tributary pipe breaks must be considered as design bases until further fracture mechanics work could eliminate some of these breaks from design consideration. This paper determines the core support barrel shell responses for the 3 inch pressurizer spray line nozzle break which is expected to be the only inlet break remaining in the primary side after leak-before-break evaluation is extended to smaller size pipes in the near future. The responses are compared with those due to 14 inch safety injection nozzle break and main coolant loop pipe break. The results show that, when the leak-before-break concept is applied to the primary side piping systems with a diameter of 10 inches or over, the core support barrel shell responses due to pipe breaks in the primary side are negligible for the faulted condition design.
Analysis of Fuel/Coolant Mixing in Steam Explosion
Lee, Tae-Ho ; Jo, Seong-Youn ; Park, Goon-Cherl ;
Nuclear Engineering and Technology, volume 25, issue 2, 1993, Pages 215~221
A required initial condition for a steam explosion to occur following core meltdown accidents of a nuclear power plant is the formation of a coarse mixture of molten fuel and water. The extent of a premixing is the measure of efficiency of steam explosion that may follow. A simple one-dimensional, transient model and the flooding criteria have been applied to evaluate the fuel/coolant mixing limit. Also, both instant breakup and dynamic breakup models for the mixing process have been separately used here and compared each other. The results indicate that fuel temperature, ambient pressure, mixing diameter, water depth, and pouring diameter are the important parameters affecting the mixing behavior.
Fuzzy Algorithms to Generate Level Controllers for Nuclear Power Plant Steam Generators
Moon, Byung-Soo ; Park, Jae-Chang ; Kim, Dong-Hwa ; Kim, Byung-Koo ;
Nuclear Engineering and Technology, volume 25, issue 2, 1993, Pages 222~232
In this paper, we present two sets of fuzzy algorithms for the steam generator level control ; one for the high power operations where the flow error is available and the other for the low power operations where the flow error is not available. These are converted to a PID type controller for the high power case and to a quadratic function form of a controller for the low power case. These controllers are implemented on the Compact Nuclear Simulator at Korea Atomic Energy Research Institute and tested by a set of four simulation experiments for each. For both cases, the results show that the total variation of the level error and of the flow error are about 50% of those by the PI controllers with about one half of the control action. For the high power case, this is mainly due to the fact that a combination of two PD type controllers in the velocity algorithm form rather than a combination of two PI type controllers in the position algorithm form is used. For the low power case, the controller is essentially a PID type with a very small integral component where the average values for the derivative component input and for the controller output are used.
Interrelationship Between the Drift-flux Model and the Two-fluid Model
No, Hee-Cheon ;
Nuclear Engineering and Technology, volume 25, issue 2, 1993, Pages 233~236
For one-dimensional two-phase flow without phase change and without axially-temporally rapid change of pressure, the interrelationship between the drift-flux model and the two-fluid model is studied. It is derived on the basis of the fact that the vapor conservation equation is related to the momentum equation by the drift flux. Starting from the two-fluid model, we obtain the interfacial friction expressed in terms of drift-flux parameter. Also, by deriving the void propagation equation, the drift-flux is shown to have jnterrelationship with forces in the two-fluid model.
Cooling Time Determination of Spent Nuclear Fuel by Detection of Activity Ratio
Lee, Young-Gil ; Eom, Sung-Ho ; Ro, Seung-Gy ;
Nuclear Engineering and Technology, volume 25, issue 2, 1993, Pages 237~247
Activity ratio of two radioactive primary fission products which had sufficiently different half-lives was expressed as functions of cooling time and irradiation histories in which average burnup, irradiation time, cycle interval time and the dominant fissile material of the spent fuel were included. The gamma-ray spectra of 36 samples from 6 spent PWR fuel assemblies irradiated in Kori unit-1 reactor were obtained by a spectrometric system equipped with a high purity germanium gamma-ray detector. Activity ratio
Cs, analyzed from each spectrum, was used for the calculation of cooling time. The results show that the radioactive fission products
Cs are considered as useful monitors for cooling time determination because the estimated cooling time by detection of activity ratio
Cs agreed well with the operator declared cooling time within relative difference of
5 % despite the low counting rate of the gamma-ray of
Ce (about 10
count per second). For the samples with several different irradiation histories, the determined cooling time by modeled irradiation history showed good agreement with that by known irradiation history within time difference of
0.5 year. From this result, it would be expected to be possible to estimate reliably the cooling time of spent nuclear fuel without the exact information about irradiation history. The feasibility study on identification of and/or sorting out spent nuclear fuel by applying the technique for cooling time determination was also performed and the result shows that the detection of activity ratio
Cs by gamma-ray spectrometry would be usefully applicable to certify spent nuclear fuel for the purpose of safeguards and management in a facility in which the samples dismantled or cut from spent fuel assemblies are treated, such as the post irradiation examination facility.mination facility.
Thermal Analysis for Dry Transport of a Shipping Cask
Lee, J.C. ; H.Y. Kang ; J.H. Yoon ; S.H. Chung ; E.H. Kwack ;
Nuclear Engineering and Technology, volume 25, issue 2, 1993, Pages 248~254
The purpose of this study is to evaluate the thermal safety for dry transport of a shipping cask. Analysis condition was based on an ambient temperature of 38
for normal heat condition. The cask was designed to carry 4PWR spent fuel assemblies with a burnup of 38,000 MWD/MTU and 3 years of cooling time. Thermal analysis was carried out by using the COBRA-SFS code. The fuel cavity was considered to be filled with air, nitrogen or helium gas for dry transport. The results of analysis showed that the maximum temperatures of fuel rod cladding in air and helium cavity would be 277
, respectively, for 3 years of cooling time. These values were less than the specified temperature to maintain the thermal integrity of fuel assembly for dry transport.
An Improved Laser-Induced Fluorimetry for Assay of Uranium in Urine
Lee, Sang-Mok ; Shin, Jang-Soo ; Kim, Cheol-Jung ;
Nuclear Engineering and Technology, volume 25, issue 2, 1993, Pages 255~258
A method for analysis of trace uranium in urine sample was studied using a time-resolved
-laser-induced fluorimetry. The Fluran solution was found to be efficient to mask the chloride ions which are known to quench uranium fluorescence in the fluorimetric assay of uranium in urine. This improved method made the sample preparation much simpler than other conventional ones. The fluorescence intensities at 1% urine mixture with 10% Fluran aqueous solution showed good linearities in the concentration range of 10-500 ppb(before dilution).
Separation of Pu and Nd from Uranium Matrix by Equilibrated Cation Exchanger for Burnup Measurement of Irradiated Nuclear Fuel
Joe, Kih-Soo ; Kim, Jung-Suk ; Jeon, Young-Shin ; Han, Sun-Ho ; Eom, Tae-Yoon ;
Nuclear Engineering and Technology, volume 25, issue 2, 1993, Pages 259~264
Ion chromatographic method has been applied for burnup measurement of irradiated nuclear fuel by dynamic system using 1-octanesulfonate as a cation exchanger and
-hydroxyisobutyric acid as an eluant. A number of elution techniques were evaluated for the optimum separation of plutonium, uranium and neodymium. These elements were individually separated and collected by gradient elution between 0.05 M and 0.40 M of
-hydroxyisobutyric acid in a single column, and finally determined by isotope dilution mass spectrometry. The burnup data from this method were compared with those from conventional anion exchange method. The results showed a good agreement within 3.5 % of difference between two methods.
Thermodynamic Analysis of Vapor Explosion Phenomena
Bang, Kwang-Hyun ;
Nuclear Engineering and Technology, volume 25, issue 2, 1993, Pages 265~275
A vapor explosion has been a concern in nuclear reactor safety due to its potential for a destructive mechanical energy release. In order to properly assess the hazard of a vapor explosion, it is necessary to accurately estimate the conversion efficiency of the thermal energy to mechanical energy. In the absence of a complete model to determine the explosive energy yield, one may have to rely on a simpler upper bound estimate such as a thermodynamic model. This paper discusses various thermodynamic models and presents a clarification of each model in their mathematical formulation and the thermodynamic work conversion. It is shown that the work release in the shock adiabatic model of Board and Hall is essentially equal to that of Hicks-Menzies thermodynamic model. The effect of coolant void fraction on the explosion efficiency is also predicted based on these thermodynamic models. Finally, the Hicks-Menzies model is modified to account for the chemical reaction between a metallic fuel and water and the resultant effects on the explosion expansion work are discussed.
An Expanded Use of Reactor Power Cutback System to Avoid Reactor Trips in the Event of an Inward Control Element Assembly Deviation
Hwang, Hae-Ryong ; Ahn, Dawk-Hwan ;
Nuclear Engineering and Technology, volume 25, issue 2, 1993, Pages 276~284
The ABB-CE System-80 reactor power cutback system(RPCS) is designed to enable continuous operation of the reactor without trip in the events of the loss of one of the two main feedwater pumps and loss of load, and thus improves plant availability in a cost effective manner. In this study expansion of RPCS has been investigated for continuous reactor operation without trip in the event of an inward control element assembly(CEA) deviation including a single rod drop. Under the expanded function of RPCS the control system will provide a rapid core power reduction on demand by releasing CEAs to drop into the core and reduce the turbine power, if necessary, to follow the reactor power variation. This design feature which is included as the new design features to be incorporated in the ABB-CE System-80+ meets the EPRI advanced light water reactor(ALWR) requirements. For this study core analysis models of System-80+ have been developed to simulate the nuclear steam supply system(NSSS) response as well as the RPCS initiation of rapid CEA insertion. The results of this study demonstrate that the reactor trip can be avoided in the event of inward CEA deviation including a single rod drop by the RPCS initiation and thus the plant availability and capacity factor would be increased.
경수로핵연료 수리기술의 현황과 개발방향
Nuclear Engineering and Technology, volume 25, issue 2, 1993, Pages 285~291
핵연료봉 주위의 난류 유동장 특성에 대한 연구 현황과 검토
Nuclear Engineering and Technology, volume 25, issue 2, 1993, Pages 292~299
고속증식로의 나트륨 화재 특성연구
Nuclear Engineering and Technology, volume 25, issue 2, 1993, Pages 300~313
차세대 원자로 설계를 위한 실증시험
Nuclear Engineering and Technology, volume 25, issue 2, 1993, Pages 317~325
중대사고 실증실험에 대한 고찰
Nuclear Engineering and Technology, volume 25, issue 2, 1993, Pages 326~334
핵연료 개발을 위한 국내열 유동 시험기술 개발 현황
Nuclear Engineering and Technology, volume 25, issue 2, 1993, Pages 335~342
KMRR의 열수력학적 설계를 위한 실증실험
Nuclear Engineering and Technology, volume 25, issue 2, 1993, Pages 343~352