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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 26, Issue 4 - Dec 1994
Volume 26, Issue 3 - Jun 1994
Volume 26, Issue 2 - Jun 1994
Volume 26, Issue 1 - Mar 1994
Selecting the target year
An Application of the Enrichment Zoning Concept to
Kim, K.S. ; Kim, J.H. ; S.K. Zee ; J.W. Song ;
Nuclear Engineering and Technology, volume 26, issue 3, 1994, Pages 337~344
Enthalpy rise hot channel factor(
) is one of the most limiting constraints in determining the fuel loading pattern(LP) for PWR's. In order to enhance the LP design flexibility without any changes of not only basic fuel specifications but also Technical Specifications and Operation Procedures, we apply the enrichment zoning concept to Westinghouse designed PWR's to flatten the rod power distributions within the fuel assembly and thus to reduce
. Enrichment zoning is described that each assembly consists of two different enrichment fuels ; the lower enriched fuels are located in positions which are expected to have the higher rod power and vice versa for the higher enriched fuels. As a result of unit assembly calculations to flatten the rod power distribution within the assembly, the appropriate enrichment difference is found to be 0.3~0.4w/o. Through core depletion calculations for the 18-month cycle of Kori Unit 4, the
behavior in core with the enrichment zoning concept is investigated. A comparison with the reference case without the enrichment zoning results in a reduction in
of approximately 1.5%.TEX>H/
of approximately 1.5%.
Study of the Secondary Flow Effect on the Turbulent Flow Characteristics in Fuel Rod Bundles
Lee, Kye-Bock ; Jang, Ho-Cheol ; Lee, Sang-Keun ;
Nuclear Engineering and Technology, volume 26, issue 3, 1994, Pages 345~354
Numerical Predictions including secondary flows have been Performed for fully developed turbulent single-phase rod bundle flows. The k-
turbulence model(two equation model) for the isotropic eddy viscosity, together with an algebraic stress model for generating secondary velocities, enabled the prediction of mean axial velocities, secondary velocities, and turbulent kinetic energy and turbulent stresses. Comparisons with experiment hate shown that the influence of secondary motion on mean flow and turbulence is dearly evident. The convective transport effects of secondary flow on the velocity field have been identified.
Quantification of Reactor Safety Margins for Large Break LOCA with Application of Realistic Evaluation Methodology
B.D. Chung ; Lee, Y.J. ; T.S. Hwang ; Lee, W.J. ; Lee, S.Y. ;
Nuclear Engineering and Technology, volume 26, issue 3, 1994, Pages 355~366
The USNRC issued a revised ECCS rule that allows the use of best estimate computer codes for safety analysis. The rule also requires an estimation of uncertainty in calculated system response when applying the best estimate computer codes. A practical realistic evaluation methodology to evaluate the ECCS performance that satisfies the requirements of the ECCS rule has been developed and this paper describes the application of new realistic evaluation methodology to large break LOCA for, the demonstration of the new methodology. The computer code RELAP5/MOD3/KAERI, which was improved from RELAP5/MOD3.1, was used as the best estimate code in the application. The uncertainty of the code was evaluated by assessing several separate and integral effect tests, and for the application to actual plant Kori 3 & 4 was selected as the reference plant. Response surfaces for blowdown and reflood PCTs were generated from the results of the sensitivity analyses and probability distribution functions were established by random sampling or Monte-Carlo method for each response surface. Final uncertainties were quantified at 95% probability level and safety margins for large break LOCA were discussed
The Development of a Signal Validation Scheme for the Redundant Multi-Channel Measurement System
Hwang, In-Koo ; Na, Nan-Ju ; Kwon, Kee-Choon ; Ham, Chang-Shik ;
Nuclear Engineering and Technology, volume 26, issue 3, 1994, Pages 367~373
It is necessary to adopt a simple signal validation for avoiding the complexity of algorithm and verification in the design process of the instrumentation and control system in nuclear plants. This paper suggests a signal validation method developed on the basis of consistency checking for the multi-channel measurement system without any analytic process model. It includes a simplified algorithm for estimating the fixed bias error of each channel and a weighted averaging method. The weighting factor of each channel is updated according to its calculated bias error. The developed method has been tested to verify its performance through several input scenarios.
Simulation of Interlinkage of Grain Boundary Gas Bubbles to Free Surfaces by the Monte Carlo Technique
Koo, Yang-Hyun ; Park, Heui-Joo ; Sohn, Dong-Seong ; Yoon, Young-Ku ;
Nuclear Engineering and Technology, volume 26, issue 3, 1994, Pages 374~380
A method to simulate the extent of interlinkage of grain boundary gas bubbles to the free surfaces of fuel pellet was developed. With the shape of UO
gain treated as tetrakaidecahedron (TKD)), the interlinked fraction of fission gas bubbles to free surfaces at grain comers was calculated as a function of the radius of grain corner bubbles by the Monte Carlo technique. In spite of two dimensional analysis, the present method shooed reasonable agreement between predicted and measured fuel swelling at the moment that complete bubble interlinkage was achieved. However, for more realistic simulation of interlinkage, grain comer bubbles should be treated three dimensionally.
The Response Correction Function of TL Dosimeter for Shallow Dose Assessment in Tl-204 Beta Fields
Lee, Sang-Yoon ; Kim, Jang-Lyul ; Seo, Kyung-Won ;
Nuclear Engineering and Technology, volume 26, issue 3, 1994, Pages 381~388
Recently, the American National Standards Institute (ANSI) had made some changes in the radiation sources specified from those in the original performance test criteria ANSI N13. 11-1983. In case or beta category, in addition to the high-energy
Y beta source, the
Tl source was added because many workplaces have significant levels of lower energy betas. In this study, the performance or the Teledyne PB-3 personnel dosimetry system in the fields of
Y beta was investigated using the PTB beta secondary standard sources. The new beta correction function of PB-3 personnel dosimetry system for
Tl beta was also developed in this response experiment. The results show that the Teledyne PB-3 personnel dosimetry system is very effective for
Y beta dose assessment. In case of
Tl beta radiation, however, the results of simple performance test indicated that the use of beta correction factor(=2.088) which was recommanded by manufacturer may result in unexpectable overestimation of delivered dose by about 60%, while the use of developed beta correction function could measure the delivered doses in errors of 15%.
Uncertainty Quantification of RELAP5/MOD3/KAERI on Reflood Peak Cladding Temperature
Park, Chan-Eok ; Chung, Bub-Dong ; Lee, Young-Jin ; Lee, Guy-Hyung ; Lee, Sang-Yong ;
Nuclear Engineering and Technology, volume 26, issue 3, 1994, Pages 389~400
The predictability of KAERI version of RELAP5/MOD3 on reflood peak cladding temperature during large break loss-of-coolant accident is assessed against 18 test runs in FLECHT SEASET test data. The associated uncertainty is statistically quantified. The selected test runs include a gravity feed test and several forced feed tests with wide range of the parameters such as flooding rate, system pressure, initial clad temperature, rod bundle power. The results show that the code under-predicts the peak cladding temperature by 7.56 K on average. The upper limit of the associated uncertainty at 95% confidence level is evaluated to be about 99 K, It including the bias due to the under-prediction.
Study on Radionuclide Migration Modelling for a Single Fracture in Geologic Medium : Characteristics of Hydrodynamic Dispersion Diffusion Model and Channeling Dispersion Diffusion Model
D.K. Keum ; W.J. Cho ; P.S. Hahn ; Park, H.H. ;
Nuclear Engineering and Technology, volume 26, issue 3, 1994, Pages 401~410
Validation study of two radionuclide migration models for single fracture developed in geologic medium the hydrodynamic dispersion diffusion model(HDDM) and the channeling dispersion diffusion model(CDDM), was studied by migration experiment of tracers through an artificial granite fracture on the labolatory scale. The tracers used were Uranine and Sodium lignosulfonate know as nonsorbing material. The flow rate ranged 0.4 to 1.5 cc/min. Related parameters for the models were estimated by optimization technique. Theoretical breakthrough curves with experimental data were compared. In the experiment, it was deduced that the surface sorption for both tracers did not play an important role while the diffusion of Uranine into the rock matrix turned out to be an important mass transfer mechanism. The parameter characterizing the rock matrix diffusion of each model agreed well The simulated result showed that the amount of flow rate could not tell the CDDM from the HDDM quantitatively. On the other hand, the variation of fracture length gave influence on the two models in a different degree. The dispersivity of breakthrough curve of the CDDM was more amplified than that of the CDDM when the fracture length was increased. A good agreement between the models and experimental data gave a confirmation that both models were very useful in predicting the migration system through a single fracture.
The Probabilistic Analysis on the Containment Failure by Hydrogen Burning at Severe Accidents in Nuclear Power Plants
Park, I.K. ; J.H. Moon ; Park, G.C. ;
Nuclear Engineering and Technology, volume 26, issue 3, 1994, Pages 411~419
The containment failure probability due to hydrogen burning during severe accidents proceeding in a low pressure sequence is calculated using Monte Carlo method. The probability distribution functions for this Monte Carlo calculation is obtained from the statistical method. The calculations are performed for Kori unit 2, and the sensitivity studies on the input variables-the amount of hydrogen generated at SFD, cerium diameter, cerium length, oxidation rate at FCI, and the amount of hydrogen generated during MCCI-are also performed. It is revealed that SFD is the main factor in hydrogen generation, but the other sources also cannot be neglected. The containment failure probability due to the hydrogen burning lies within 6% in case of Kori unit 2.
A Study on Improvement of PWR Steam Generator Water Level Control at Low Power Operation
Yun, Jae-Hee ; Han, Jai-Bok ; Joon Lyou ;
Nuclear Engineering and Technology, volume 26, issue 3, 1994, Pages 420~424
This paper presents an improved water level control scheme for Pressurized Water Reactor(PWR) Steam Generator(S/G) at the low power operation and transient states. To reduce fluctuations of the water level by the swell and shrink phenomena, the scheme adds feedforward terms considering S/G pressure and the feedwater temperature into the conventional proportional-integral feedback controller. The simulation results using the Compact Nuclear Simulator show that smaller level errors and much faster settling time than those of the conventional scheme can be obtained. The proposed algorithm is easily implementable and has a potential for the real applications.
Study of the Effect of Hydrazine Form and Titanium Electrode Condition on Reduction of Uranium(VI) n Nitric Acid
Kim, K.W. ; Lee, E.H. ; Y.J. Shin ; J.H. Yoo ; Park, H.S. ; Kim, Jong-Duk ;
Nuclear Engineering and Technology, volume 26, issue 3, 1994, Pages 425~432
Voltammogram analysis of U(VI) reduction at electrochemically non-pretreated/pretreated Ti electrodes in nitric acid and hydrazine(
) media was done in order to determine the effect of hydrazine form and Ti electrode condition on the reduction of U(VI) in nitric acid. In the case of non-pretreated Ti electrode, the reduction in nitric acid and hydrazine mono-hydrate solution needed a high activation overpotential and was affected by the ratio of hydrazine to nitric acid rather than by only absolute amount of hydrazine because of the decrease of solution conductivity and increase of iR drop, which were caused by proton consumption in the solution by the hydrazine. In the case of pretreated Ti electrode in nitric acid and protonated hydrazine solution, the reduction current peaks of U(VI) were clearer and higher enough to perform a kinetic analysis, compared with the case with the non-pretreated Ti electrode at the same potential, and the behavior was strongly affected by nitric acid. The presence of hydrazine was important in the reduction of U(VI) at the pretreated Ti electrode for preventing the reoxidation of U(IV), but the concentration of protonated hydrazine was not.t.
원자로 내부구조물의 설계방법이 같은 경우 원자로의 상대적 크기 변화에 따른 노심에서의 열수력학적 특성에 대한 연구
Nuclear Engineering and Technology, volume 26, issue 3, 1994, Pages 433~439
Effect of Transverse Convex Curvature on Turbulent Fluid Flow in Fuel Channel
Lee, Yung ; Ahn, Seung-Hoon ; Kim, Hyong-Chol ;
Nuclear Engineering and Technology, volume 26, issue 3, 1994, Pages 440~452
Nuclear fuel bundles are designed such that the heat flux at a-fuel pin surface should not exceed the critical heat flux (CHF) during normal operation and anticipated transient. Therefore, evaluation of the CHF for fuel bundle is demanded in an exact and reliable manner. One of the major concerns with the current application of CHF correlations is that the CHF based on circular tubes is applied to the fuel bundle subchannel analysis, mainly in terms of the hydraulic diameter with correction factors which may result in a source of possibly large uncertainties in CHF prediction. The hydraulic diameter does not recognize the local properties of fluid nor such effect as the surface curvature; the turbulence action on the convex surface is much more pronounced than that on the concave surface. Even for the tube having concave curvature, the effect of tube diameter on CHF becomes important with decreasing diameter. These facts imply that the convex curvature effect is significant and crucial to the reliable CHF prediction. This paper reviews and discusses analytical and experimental aspects of effect of transverse convex curvature in incompressible turbulent flow and heat transfer, and on CHF. Flow models to quantify this effect are briefly mentioned and future works are recommended.