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REFERENCE LINKING PLATFORM OF KOREA S&T JOURNALS
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Nuclear Engineering and Technology
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Korean Nuclear Society
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Volume & Issues
Volume 27, Issue 6 - Dec 1995
Volume 27, Issue 5 - Aug 1995
Volume 27, Issue 4 - Aug 1995
Volume 27, Issue 3 - Jun 1995
Volume 27, Issue 2 - Apr 1995
Volume 27, Issue 1 - Feb 1995
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A Generalized Model for the Prediction of Thermally-Induced CANDU Fuel Element Bowing
H. C. Suk ; K-S. Sim ; Park, J. H. ;
Nuclear Engineering and Technology, volume 27, issue 6, 1995, Pages 811~824
The CANDU element bowing is attributed to actions of both the thermally induced bending moments and the bending moment due to hydraulic drag and mechanical loads, where the bowing is defined as the lateral deflection of an element from the axial centerline. This paper consider only the thermally-induced bending moments which are generated both within the sheath and the fuel and sheath by an asymmetric temperature distribution with respect to the axis of an element The generalized and explicit analytical formula for the thermally-induced bending is presented in con-sideration of 1) bending of an empty tube treated by neglecting the fuel/sheath mechanical interaction and 2) fuel/sheath interaction due to the pellet and sheath temperature variations, where in each case the temperature asymmetries in sheath are modelled to be caused by the combined effects of (i) non-uniform coolant temperature due to imperfect coolant mixing, (ii) variable sheath/coolant heat transfer coefficient, (iii) asymmetric heat generation due to neutron flux gradients across an element and so as to inclusively cover the uniform temperature distributions within the fuel and sheath with respect to the axial centerline. As the results of the sensitivity calculations of the element bowing with the variations of the parameters in the formula, it is found that the element bowing is greatly affected relatively with the variations or changes of element length, sheath inside diameter, average coolant temperature and its variation factor, pellet/sheath mechanical interaction factor, neutron flux depression factor, pellet thermal expansion coefficient, pellet/sheath heat transfer coefficient in comparison with those of other parameters such as sheath thickness, film heat transfer coefficient, sheath thermal expansion coefficient and sheath and pellet thermal conductivities.
Effect of Loading Variables and Temperature on Fatigue Crack Propagation in SA508 Cl.3 Nuclear Pressure Vessel Steel
Kim, B. S. ; Lee, B. H. ; Kim, I. S. ;
Nuclear Engineering and Technology, volume 27, issue 6, 1995, Pages 825~832
The effect of loading variables and temperature on fatigue crack growth rate in SA508 Cl.3 nuclear pressure vessel steel was investigated in air environment Crack growth rate tests on compact tension specimen of thickness 12mm were conducted by using sinusoidal waveform. The crack length was monitored by compliance method. Test conditions were at 0.1 and 0.5 of load ratio, at 1 and 10 Hz of loading frequency, and at room temperature to 40
. At the lower temperatures, the fatigue crack propagation was not affected by the frequency and temperature, while at the higher temperatures above 12
, fatigue crack growth rate increased with decreasing loading frequency and increasing temperature. This accelerated fatigue crack propagation was associated with the increase of oxidation rate at the ahead of crack tip. Fatigue crack growth rate increased with in-creasing the load ratio. The effect of load ratio was more significant at the lower temperature, while the dependence on load ratio decreased with increasing temperature. The sensitivity of load ratio to temperature can be explained by crack closure with the oxidation process.
Self-Tuning Predictive Control with Application to Steam Generator
Kim, Chang-Hwoi ; Sang Jeong lee ; Ham, Chang-Shik ;
Nuclear Engineering and Technology, volume 27, issue 6, 1995, Pages 833~844
In self-tuning predictive control algorithm for steam generator is presented. The control algorithm is derived by suitably modifying the generalized predictive control algorithm. The main feature of the unposed method relies on considering the measurable disturbance and a simple adaptive scheme for obtaining the controller gain when the parameters of the plant are unknown. This feature makes the proposed approach particularly appealing for water level control of steam generator when measurable disturbance is used. In order to evaluate the performance of the proposed algorithm, computer simulations are done for an PWR steam generator model. Simulation result show satisfactory performances against load variations and steam flow rate estimation errors. It can be also observed that the proposed algorithm exhibit better responses than a conventional PI controller.
Waveform Relaxation Method for Reactor Transient Analysis
Park, Keon-Woo ; Co, Nam-Zin ;
Nuclear Engineering and Technology, volume 27, issue 6, 1995, Pages 845~852
We investigate the concurrent solution of differential equations by the waveform relaxation (WR) method, an iterative method for analyzing linear and nonlinear dynamical systems in the time do-main. The method, at each iteration, decomposes the dynamical system into several subsystems, each of which is analyzed for the entire given time interval. The method, when efficiently implemented, results in algorithms with a highly parallelizable concurrent fraction. In this paper, the waveform relaxation method is introduced and applied to two types of reactor dynamics problems. It is concluded that the U method can be applied to reactor dynamics equations, but that its parallel performance on the KMRR dynamics is only modest.
Equivalent Pre- Xenon-Oscillation Method for Core Transient Simulation
J. S. Song ; Lee, C. K. ; Lee, C. C. ; C. S. Yoo ; Kim, Y. R. ;
Nuclear Engineering and Technology, volume 27, issue 6, 1995, Pages 853~858
The initial condition of a core transient should be consistent with real core state for the simulation of the core tansient. The initial xenon distribution, which can not be measured in the core, has a significant effect on the transient with xenon dynamics. In the simulation of the transient starting from non-equilibrium xenon state, the accurate initialization of the non-equilibrium xenon distribution is essential for the prediction of the core transient behavior. In this study, a xenon initialization method to predict the core transient more accurately was developed through the equivalent pre-xenon-oscillation which represents the tenon oscillation before the transient and verified by the application of the simulation for a startup test of Yonggwang Unit 3.
Analysis of the Vent Path Through the Pressurizer Manway Under the Loss of Residual Heat Removal(RHR) System During Mid-Loop Operation in PWR
G. S. Ha ; Kim, W. S. ; W. P. Chang ; K. J. Yoo ;
Nuclear Engineering and Technology, volume 27, issue 6, 1995, Pages 859~869
The present study is to understand the physical phenomena anticipated during the accident with RHR loss under mid-loop operation in a PWR and, at the same time, to examine the prediction capability of RELAP5/MOD3.1 on such an accident, by simulating an integral test relevant to this accident for reliable analysis in an actual PWR. The selected experiment, i.g. BETHSY Test 6.9a, represents the configuration with the pressurizer manway open and steam generators unavailable during the accident. Accordingly, the results of this ok are sure to contribute to understanding both the key events as well as the sensitive parameters, anticipated in the accident, for validity of the actual analysis. In the simulation result overall behavior as well as major phenomena observed in the experiment have been predicted reasonably by RELAP5/MOD3.1, however, the problem associated with enormous computing time .due to small time step size has been encountered. Besides, the code prediction of higher swollen level in the pressure vessel has given rise to overestimation of both pressurizer level and RCS pressure. Subsequently, overprediction of the break flow through the manway has led to earlier core uncovery than that in the experiment by about 400 seconds. As a whole, it is demonstrated from both the experiment and the analysis that gravity feed has not been sufficient to recover the core level and thus additional forced feed has been necessary in this configuration.
A Three-Dimensional Nodal Diffusion Code Based on the AFEN Methodology
Hong, Ser-Gi ; Cho, Nam-Zin ; Noh, Jae-Man ;
Nuclear Engineering and Technology, volume 27, issue 6, 1995, Pages 870~876
In this paper, a new three-dimensional nodal diffusion code which is based on the AFEN methodology is described and tested. The method expands the homogeneous flux within a node in ter-ms of eighteen analytic basis functions satisfying the diffusion equation at any point of the node. And the nodal coupling equations are derived such that nodal balance, current continuity and leakage balance within an infinitesimally small box around the edge are satisfied. To verify its accuracy, the code was applied to the well-known static LMW benchmark problem and a small core benchmark problem that has the same material properties as the three-dimensional IAEA benchmark problem and compared with two other codes (QUANDRY, VENTURE). The results show that the code provides good accuracy both in the power distribution and in the effective multiplication factor.
Analysis and Evaluation of CPC / COLSS Related Test Result During YGN 3 Initial Startup
S. G. Chi ; S. S. Yu ; W. K. In ; G. S. Auh ; J. Y. Doo ; Kim, D. K. ;
Nuclear Engineering and Technology, volume 27, issue 6, 1995, Pages 877~887
YGN 3 is the first nuclear power plant to use the Core Protection Calculator (CPC) as the core protection system and the Core Operating Limit Supervisory System (COLSS) as the core monitor-ing system in Korea. The CPC is designed to provide on-line calculations of Departure from Nucleate Boiling Ratio (DNBR) and Local Power Density (LPD) and to initiate reactor trip if the core conditions exceed the DNBR or LPD design limit. The COLSS is designed to assist the operator in implementing the Limiting Conditions for Operation (LCOs) in Technical Specifications for DNBR/Linear Heat Rate (LHR) margin, azimuthal tilt, and axial shape index and to provide alarm when the LCOs are reached. During YGN 3 initial startup testing, extensive CPC/COLSS related tests ore peformed to ver-ify the CPC/COLSS performance and to obtain optimum CPC/COLSS calibration constants at var, -ious core conditions. Most of test results met their specific acceptance criteria. In the case of missing the acceptance criteria, the test results ore analyzed, evaluated, and justified. Through the analysis and evaluation of each of the CPC/COLSS related test results, it can be concluded that the CPC/COLSS are successfully Implemented as designed at YGN 3.
Study on the Relationship Between Turbulent Normal Stresses in the Fully Developed Bare Rod Bundle Flow
Lee, Kye-Bock ; Lee, Byung-Jin ;
Nuclear Engineering and Technology, volume 27, issue 6, 1995, Pages 888~893
The turbulence structure for fully developed flow through the subchannels formed by the bare rod array depends on the pitch to rod diameter ratio. For fairly open spaced bare rod arrays, the distributions of the three components of the turbulent normal stresses are similar to those measured in circular pipe. However, for more closely spaced arrays, the turbulence structure, especially in the gap region, depart markedly from the pipe flow distribution. A linear relationship between turbulent normal stresses and turbulent kinetic energy for fully developed turbulent flow through regularly spaced bare rod arrays has been developed. This correlation can be used in connection with various theoretical analyses applied in turbulence research.
PC Based-Proto Type 고장검출 프로그램에 의한 중수로 안전계통(SDS#1, SDS #2)주기시험 시스템 개발
Nuclear Engineering and Technology, volume 27, issue 6, 1995, Pages 894~897
경수로핵연료 하단고정체 유로판의 두께 최적화
Nuclear Engineering and Technology, volume 27, issue 6, 1995, Pages 898~910
로듐 자기 기전력형 중성자 계측기의 수명 연장에 관한 연구
Nuclear Engineering and Technology, volume 27, issue 6, 1995, Pages 911~917
이론적 강제대류CHF 해석 모델의 연구 현황 및 성능 평가
Nuclear Engineering and Technology, volume 27, issue 6, 1995, Pages 918~931
CANDU형 원자로 주열수송 계통에 대한 Acoustic 해석
Nuclear Engineering and Technology, volume 27, issue 6, 1995, Pages 932~937
방사성 폐기물 처분동굴 내로 유입되는 지하수량 추정 및 처분동굴 폐쇄후 지하수 유동 경로 분석
Nuclear Engineering and Technology, volume 27, issue 6, 1995, Pages 938~943