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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
Proposed Concept of a Tube-Type Passive Water-Cooled Reactor Without Emergency Core Cooling System
Chang, Soon-Heung ; Baek, Won-Pil ; Lee, Goung-Jin ; Lee, Jae-Young ;
Nuclear Engineering and Technology, volume 26, issue 2, 1994, Pages 161~167
This paper presents a concept of a pressure tube-type water-cooled reactor without the emergency core cooling system. It adopts an innovative fuel channel design using metallic fuel matrix to improve heat transfer from fuel to moderator at loss of coolant cooling. The heat produced in the fuel is cooled by the coolant system during normal operation, but by the passive moderator system at loss of coolant cooling including the loss-of-coolant accident(LOCA). Simple analysis shows that the fuel channel temperature can be maintained within the permissible range for both normal operation and a complete LOCA.
On the Tools of Decision Trees and Influence Diagrams for Assessing Severe Accident Management Strategies
Moosung Jae ; Park, Chang-Kue ;
Nuclear Engineering and Technology, volume 26, issue 2, 1994, Pages 168~178
Accident Management involves all measures to prevent core damage and retain the core within the reactor vessel, maintain containment integrity and minimize off-site releases. The accident management approach includes : (1) advanced evaluation of candidate strategies, (2) development of procedures to execute appropriate actions efficiently, and (3) identification and provision for materials, tools, and possible modifications to the plant system that may be needed for such execution. When assessing accident management strategies it effectiveness, adverse effect and its feasibility, including information needs and compatibility with existing procedures, must be considered. The objective of this paper is to introduce analytical tools of decision trees and influence diagrams to develop a framework for modeling and assessing severe accident management strategies. The characteristics associated with these took are presented. Based on decision trees and influence diagrams, the framework is applied to a simple example associated with a single decision.
Prediction of Reactor Coolant Pump Performance Under Two-Phase Flow Conditions
Lee, S. ; Y.S. Bang ; Kim, H.J. ;
Nuclear Engineering and Technology, volume 26, issue 2, 1994, Pages 179~189
A performance of reactor coolant pump in two-phase flow is examined using the pump geometric conditions and the performance of the pump in single-phase flow. Wall friction loss of the reactor coolant pump in single-phase flow is prdicted using the Truckenbrodt boundary layer theory, and the head loss in two-phase flow is predicted with calculated well friction loss and separation loss coefficients. The analysis results are compared with the Combustion Engineering pump test data. The effect of two-phase multiplier on the peak clad temperature in Loss-of-Coolant Accident is also examined using the RELAP5 and the results indicate the importance of its accuracy.
Ball-milling Effect on the Sinterability of the
Kim, H.S. ; Park, C.H. ; Park, C.J. ; Park, C.B. ; S.H. Jung ; H.C. Suk ;
Nuclear Engineering and Technology, volume 26, issue 2, 1994, Pages 190~196
In order to investigate the ball-milling effect on the property changes of UO
ex-AUC powder, the sinterability of ball -milled powder was studied in terms of the ball -milling time. Spherical shape was found to be kept for ball-milled UO
powder and the particle size showed a bimodal distribution, which seems to have a higher packing ratio compared with those having monomodal gaussian distribution. The increase of sintered density of the ball -milled UO
powder is assumed to be mainly affected by the packing ratio, which increase with longer ball -milling time. It is confirmed that the sinterability of UO
ex-AUC powder is improved by the ball-milling process.
Development of A Computer Program for Drop Time and Impact Velocity of the Rod Cluster Control Assembly
Park, Ki-Seong ; Kim, Il-Kon ;
Nuclear Engineering and Technology, volume 26, issue 2, 1994, Pages 197~204
In a PWR rod cluster control assembly(RCCA) for shutdown is released upon action of control rod drive mechanism and falls down through the guide thimble by its weight. Drop time and impact velocity of the RCCA are two key parameters with respect to reactivity insertion time and the mechanical integrity of fuel assembly. Therefore, the precise control of drop time and impact velocity is prerequisite to modifying the existing design features of the RCCA and guide thimble or newly designing them. During its falling down into the core, the RCCA is retarded by various forces acting on it such as fluid resistance caused by the RCCA movement, buoyance and mechanical friction caused by contacting inner surface of the guide thimble, etc. However, complicated coupling of the various forces makes it difficult to derive an analytical dynamic equation for the drop time and impact velocity. This paper deals with the development of a computer program containing an analytical dynamic equation applicable to the Korean Fuel Assembly(KOFA). The computer program is benchmarked with an available single control rod drop tests. Since the predicted values are in good agreement with the test results, the computer program developed in this paper can be employed to modify the exiting design features of the RCCA and guide thimble and to develope their new design features for advanced nuclear reactors.
Digital Dynamic Compensation Methods of Rhodium Self-Powered Neutron Detector
Auh, Geun-Sun ;
Nuclear Engineering and Technology, volume 26, issue 2, 1994, Pages 205~211
The best method is selected among the 3 digital dynamic compensation methods which are developed or applied for the Rhodium self-powered neutron detector. The three digital dynamic compensation methods are the existing Dominant Pol Tustin method of the COLSS(Core Operating Limit Supervisory System), the Direct Inversion method and Kalman Filter method. The Direct Inversion method is an improved method of D. Hoppe and R. Maletti and the Kalman Filter method is developed using the Kalman Filter. Response times of the compensated signals to achieve 90% of a step input are 28.1, 17.2 and 6.5 seconds respectively for the same noise gain telling that the Kalman Filter method is the best amens the 3 methods.
Dynamic Stability of a Flexible Cylinder Subjected to Inviscid Flow in a Coaxial Cylindrical Duct Based on Spectral Method
Sim, Woo-Gun ; Bae, Yoon-Yeong ;
Nuclear Engineering and Technology, volume 26, issue 2, 1994, Pages 212~224
A numerical method has been developed for studying the dynamics of a flexible cylinder in a coaxial cylindrical duct, immersed in inviscid flow. The unsteady inviscid fluid-dynamic force acting on the oscillating cylinder has been estimated more rigorously by means of a spectral collocation method without simplification of governing equations. This numerical approach is applicable to the system haying wider annular gap and/or shorter length of cylinder as compared to existing potential theory. The governing equation of the unsteady flow was obtained from Laplace equation. The equation of cylinder motion coupled with the fluid motion was discretized by Galerkin's method, from which the dynamic behaviour of the system has been evaluated. The effect of the length of the cylinder and the annular gap on the critical flour velocity, where the system loses stability by buckling, was investigated. To validate the numerical method, the potential flow theory developed by Hobson based on thin film approximation has been improved. Typical results of the present numerical theory on the dynamics and stability of the system are compared with those of available existing theory and the present approximate results. Good agreement was found between the results. It was also found that a nondimensional critical flow velocity becomes larger as increasing the annular gap and decreasing the length of cylinder.
RELAP5/MOD3 Analysis for Hydraulic Load Calculation of the SEBIM POSRV Discharge Riping System
Han, Kee-Soo ; Song, Jin-Ho ;
Nuclear Engineering and Technology, volume 26, issue 2, 1994, Pages 225~236
The sudden discharge of the loop seal water, which is present upstream of the SEBIM POSRV, creates large momentum and inertia forces on the downstream of the discharge piping system. This study provides the procedures and results of analysis of the thermal-hydraulic transient in the SEBIM POSRV discharge piping during the valve opening. The analysis is peformed by RELAP5/MOD3. The appropriate modeling of the discharge piping system, SEBIM POSRV opening characteristics, and loop seal water discharge for the RELAP5/MOD3 analysis is suggested. Also performed is the sensitivity study for the selection of proper options for the junction and volume control. flags. The analysis results demonstrate the adequacy of the RELAP5/HOD3 for the thermal-hydraulic transient analysis of the loop seal water discharge of the SEBIM POSRV discharge piping system. From the sensitivity analysis results, it is shown that the smooth area change option with reasonable geometric pressure drop distribution, non-equilibrium option, and proper time step should be selected for loop seal water discharge analysis.
Development of a High Flow CHF Correlation for the KMRR Fuel
Park, Cheol ; Hwang, Dae-Hyun ; Yoo, Yeon-Jong ; Park, Jong-Ryul ;
Nuclear Engineering and Technology, volume 26, issue 2, 1994, Pages 237~246
A high flow critical heat flux (CHF) correlation, based on the single-pin CHF experimental data for finned and unfinned heated rods, was developed for the thermal-hydraulic design and safety analysis of the Korea Multi-purpose Research Reactor (KMRR) core. The correlation consists of dimensionless parameters such as Reynolds number, thermodynamic equilibrium quality, liquid-to-vapor density ratio, and hydraulic equivalent diameter ratio. The fin effect was taken into account in the correlation by a finned-to-unfinned heated perimeter ratio. The effects of a cold wall and non-uniform axial power distribution ore discussed to verify the applicability of the single-pin based correlation to the KMRR fuel bundle. The correlation limit departure from nucleate boiling ratio (DNBR) was determined as 1.44 from the statistical analysis of the CHF data.
An Adaptation of the SAV Standard Nuclide Chain for the CASMO3/MEDIUM3 Procedure
Lee, Chang-Ho ; Kim, Young-Jin ;
Nuclear Engineering and Technology, volume 26, issue 2, 1994, Pages 247~256
The nuclide chain model used in SAV90 has been modified for the CASMO3/MEDIUM3 procedure. Since the default nuclide chain in SAV90, using 21 nuclides, is not sufficient to reproduce the CASMO3 results in the MEDIUM3 calculation, the extended nuclide chain models have been investigated and verified with various types of fuel assemblies. Among the extended nuclide chain models proposed, the 22 nuclide chain model, which contains only Pu238 additionally to the 21 nuclide chain, is recommended in terms of both accuracy and computing efficiency. Using this model core follow calculations for YGN-1 have been performed. The results showed good performance when compared to plant measurements.
Experimental Investigation on the Vapor Explosions with Water/R22
Park, I.K. ; Park, G.C. ;
Nuclear Engineering and Technology, volume 26, issue 2, 1994, Pages 257~264
Experimental studies hate been peformed to investigate vapor explosion phenomena which may threaten the containment integrity during severe accidents in nuclear power plants. In this study, experimental equipment is constructed for vapor explosion experiments, and the vapor explosion experiments were conducted using water/R22. During the experiments, water/R22 interaction phenomena were observed using the high speed camera, and the explosion pressure and released mechanical energy were measured with pressure transducer and pressure relief tube. And the effects of some important parameters-hot liquid temperature, hot liquid injection velocity, hot liquid injection velocity, hot liquid injection time, and cold liquid depth-were investigated on the vapor explosion. Also, the experiment with grid was conducted to study reactor -vessel-lower-structure effect on fuel/coolant interaction. Water/R22 explosion conversion ratios were measured between 0.5∼1.6%.
Improvements to the RELAP5/MOD3 Reflood Model and Assessment
B.D. Chung ; Lee, Y.J. ; Park, C.E. ; Park, C.J. ; T.S. Hwang ;
Nuclear Engineering and Technology, volume 26, issue 2, 1994, Pages 265~276
Several improvements to the RELAP5/MOD3 reflood model hate been made. These improvement were made to correct deficiencies in the reflood model identified by the assessment of the RELAP5/MOD3 code against FLECHT-SEASET experiments. The improvements consist of modification of reflood wall heat transfer package and adjusting the droplet size in dispersed flow regime. The time smoothing of wall vaporization and level tracking of transition flow are also added to eliminate the pressure spikes and level oscillation during reflood process. Assessment of the improved model against FLECHT-SEASET experimental data and application of LBLOCA analysis for plant shows that the deficiencies have been corrected.
Near-Field Transport of Radionuclide Decay Chains
Kang, Chul-Hyung ;
Nuclear Engineering and Technology, volume 26, issue 2, 1994, Pages 277~284
Much attention has been given to predict the near-field mass transfer of a single radioactive species from a waste solid into surrounding porous medium. But only limited considerations have been given to predict the coupled mass transfer of species with a radioactive decay chain. In this study we present an analysis assuming that the members of a decay chain dissolve congruently with a solubility-limited matrix. We give general, non-recursive analytic solutions for the transport of a radioactive decay chain in a finite porous medium when nuclides are released congruently with the matrix. As an illustration we consider the decay chain
Ra from spent fuel. These solutions may be useful and potentially important in performance assessment of radioactive waste repositories.
Diffusion Characteristics of Iodide in a Domestic Bentonite of Korea
Lee, J.O. ; W.J. Cho ; P.S. Hahn ; Park, H.H. ;
Nuclear Engineering and Technology, volume 26, issue 2, 1994, Pages 285~293
The transport of radionuclides in a compacted bentonite is dominated by diffusion. Through-diffusion tests for iodide were performed to investigate the diffusion characteristics of anionic radionuclides in a domestic bentonite. The bentonite used was sampled from the southeastern area of Korea and the solution was synthetic groundwater spiked with a tracer of I -125(as Na
I). The dry densities of compacted bentonite were 1.2, 1.4, and 1.7 Mg/㎥. The apparent diffusion coefficients and the effective diffusion coefficients of the iodide decrease with increasing dry density. The values were from 3.80 to 7.12
/s for the apparent diffusion coefficients and from 1.25 to 7.97
/s for the effective diffusion coefficient, respectively. The experimental results also showed that the apparent diffusion coefficients depended on the pore structure of compacted bentonite and the effective diffusion coefficients were attributed to the pore structure and the effective porosity that represents the available pathway for the diffusional transport of iodide. The results obtained will be used as basic data for the safety assessment of a repository.
Margin Benefit Assessment of A Digital Monitoring System for Existing Analog Plants
Auh, Geun-Sun ; Yoon, Tae-Young ;
Nuclear Engineering and Technology, volume 26, issue 2, 1994, Pages 294~299
Margin benefits are quantatively assessed when a Digital Monitoring System(DMS) is assumed to be installed to an operating Westinghouse analog type plant. Applied plant and cycle is YongGwang Unit 1 Cycle 6. The referenced digital monitoring system is the COLSS (Core Operating Limit Supervisory System) of ABB-CE. Considered fuel design limits are DNBR and LDCA Fq. 2003-D Power distributions within the present CAOC (Constant Axial Offset Control) limits are calculated for the analysis. The most limiting DNB prevention event of CEA Withdrawal is analyzed with the ROPM (Required OverPower Margin) concept of ABB-CE. The result show that the DMS can bring around 7% more margins for both DNB and LOCA Fq standpoints of view. The DMS can also monitor the PCI (Pellet-Cladding Interaction) limits.
TRIGA원자로의 폐지 현황
Nuclear Engineering and Technology, volume 26, issue 2, 1994, Pages 300~305
경수로형 원전의 초음파 검사 기술현황
Nuclear Engineering and Technology, volume 26, issue 2, 1994, Pages 306~311
원전 2차계통 배관재의 침식-부식 손상
Nuclear Engineering and Technology, volume 26, issue 2, 1994, Pages 312~323
원전 1차계통 방사선량 감소를 위한 코발트 합금 대체기술 개발
Nuclear Engineering and Technology, volume 26, issue 2, 1994, Pages 324~336