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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 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
Selecting the target year
Heat Transfer in the Passive Containment Cooling System
Cha, Jong-Hee ; Jun, Hyung-Gil ; Chung, Moon-Ki ;
Nuclear Engineering and Technology, volume 27, issue 3, 1995, Pages 281~291
The objective of this work is to obtain the experimental data for the heat transfer processes occurring both on the inside and outside surfaces of containment steel wall with dry and wet outer surface conditions in the passive containment cooling system. The test model represented a 60
section of a containment vessel based on the AP 600 geometry. Major linear dimensions of the test model ore reduced tv a factor of ten. To simulate the decay heat a steam generator heated by electricity was placed in the test model. The maximum heat flux was 8.91 kW/
. Two types of tests were performed. The one was the tort on the natural convection of air without water film flow. The other was the evaporative heat transfer test with the falling water film flow and natural air draft. no test result shooed that the heat transfer capability by the natural convection from the containment to the air without oater film flow was limited at about 1.48 kW/
heat flux. It was found that the heat removal capability was remarkably enhanced in the tests with the waster film flow and air draft. The obtained heat transfer data ore compared with the existing correlations.
A New Dynamic HRA Method and Its Application
Jae, Moo-Sung ; Park, Chan-Kue ;
Nuclear Engineering and Technology, volume 27, issue 3, 1995, Pages 292~300
This paper present a new dynamic HRA (Human Reliability Analysis) method and its application for Quantifying the human error probabilities in implementing an accident management action. For comparisons of current HRA methods with the new method, the characteristics of THERP, HCR, and SLIM-MAUD, which are most frequently used methods in PSAs, are discussed. The action associated with the implementation of the cavity flooding during a station blackout sequence is considered for its application. This method is based on the concepts of the quantified correlation between the performance requirement and performance achievement. The MAAP 3.0B code and Latin Hypercube sampling technique are used to determine the uncertainty of the performance achievement parameter. Meanwhile, the value of the performance requirement parameter is obtained from interviews. Based on these stochastic distributions obtained, human error probabilities are calculated with respect to the various means and variances of the timings. It is shown that this method is very flexible in that it can be applied to any kind of the operator actions, including the actions associated with the implementation of accident management strategies.
Natural Convection in a Water Tank with a Heated Horizontal Plate Facing Downward
Yang, Sun-Kyu ; Chung, Moon-Ki ; Helmut Hoffmann ;
Nuclear Engineering and Technology, volume 27, issue 3, 1995, Pages 301~316
experimental and computational studies ore carried out to investigate the natural convection of the single phase flow in a tank with a heated horizontal plate facing downward. This is a simplified model for investigations of the influence of a core melt at the bottom of a reactor vessel on the thermal hydraulic behavior in a oater filled cavity surrounding the vessel. In this case the vessel is simulated by a hexahedron insulated box with a heated plate Horizontally mounted at the bottom of the box. The box with the heated plate is installed in a water filled hexahedron tank. Coolers are immersed in the U-type water volume between the box and the tank. Although the multicomponent flows exist more probably below the heated plate in reality, present study concentrates on the single phase flow in a first step prior to investigating the complicated multicomponent thermal hydraulic phenomena. In the present study, in order to get a better understanding for the natural convection characteristics below the heated plate, the velocity and temperature are measured by LDA(Laser Doppler Anemometry) and thermocouples, respectively. And How fields are visualized by taking pictures of the How region with suspended particles. The results show the occurrence of a very effective circulation of the fluid in the whole How area as the heater and coolers are put into operation. In the remote region below the heated plate the new is nearly stagnant, and a remarkable temperature stratification can be observed with very thin thermal boundary. Analytical predictions using the FLUTAN code show a reasonable matching of the measured velocity fields.
A Fuzzy Controller for the Steam Generator Water Level Control and Its Practical Self-Tuning Based on Performance
Na, Nan-Ju ; Bien, Zeun-Gnam ;
Nuclear Engineering and Technology, volume 27, issue 3, 1995, Pages 317~326
The oater level control system of the steam generator in a pressurized water reactor and its control Problems are analysed. In this work a stable control strategy Particularly during low Power operation based on the fuzzy control method is studied. The control strategy employs substitutional information using the bypass valve opening instead of incorrectly measured signal at the low How rate as the fuzzy variable of the flow rate during low power operation, and includes the flexible scale adjusting method for fast response at a large transient. A self-tuning algorithm based on the control performance and the descent method is also suggested for tuning the membership function scale. It gives a practical way to tune the controller under real operation. Simulation was carried out on the Compact Nuclear Simulator set up at Korea Atomic Energy Research Institute and its result showed the good performance of the controller and effectiveness of its tuning.
Nozzle Dam Design Improvement in Steam Generator
Kim, Tae-Ryong ; Park, Jin-Seok ; Jung, Seung-Ho ; Park, Jin-Ho ;
Nuclear Engineering and Technology, volume 27, issue 3, 1995, Pages 327~335
The normal shutdown and maintenance period of a nuclear power plant can be remarkably shortened when the examination and maintenance works in steam generator tubes are simultaneously carried out with refueling job. There are nozzle dams to Hock the coolant How from reactor to steam generator. Workers are reluctant to install nozzle dam because of the high radiation exposure and the limited working space in steam generator. Moreover, the heavy weight of present nozzle dam makes it installation and removal works much difficult. In this paper, a lighter KAERI nozzle dam with increased flexural rigidity-to-weight was designed and manufactured by changing the structure design of the present nozzle dam and by selecting new material, carbon fiber-reinforced plastic.
Application of Adaptive Control Theory to Nuclear Reactor Power Control
Ha, Man-Gyun ;
Nuclear Engineering and Technology, volume 27, issue 3, 1995, Pages 336~343
The Self Tuning Regulator(STR) method which is an approach of adaptive control theory, is ap-plied to design the fully automatic power controller of the nonlinear reactor model. The adaptive control represent a proper approach to design the suboptimal controller for nonlinear, time-varying stochastic systems. The control system is based on a thirdorder linear model with unknown, time-varying parameters. The updating of the parameter estimates is achieved by the recursive extended least square method with a variable forgetting factor. Based on the estimated parameters, the output (average coolant temperature) is predicted one-step ahead. And then, a weighted one-step ahead controller is designed so that the difference between the output and the desired output is minimized and the variation of the control rod position is small. Also, an integral action is added in order to remove the steadystate error. A nonlinear M plant model was used to simulate the proposed controller of reactor power which covers a wide operating range. From the simulation result, the performances of this controller for ramp input (increase or decrease) are proved to be successful. However, for step input this controller leaves something to be desired.
Evaluation of Gap Heat Transfer Model in ELESTRES for CANDU Fuel Element Under Normal Operating Conditions
Lee, Kang-Moon ; Ohn, Myung-Yong ; Lim, Hong-Sik ; Park, Jong-Ho ; Hwang, Son-Tae ;
Nuclear Engineering and Technology, volume 27, issue 3, 1995, Pages 344~357
The gap conductance between the fuel and the sheath depends strongly on the gap width and has a significant influence on the amount of initial stored energy. The modified Ross and Stoute gap conductance model in ELESTRES is based on a simplified thermal deformation model for steady-state fuel temperature calculations. A review on a series of experiments reveals that fuel pellets crack relocate, and are eccentrically positioned within the sheath rather than solid concentric cylinders. In this paper, the hue recently-proposed gap conductance models (offset gap model and relocated gap model) are described and are applied to calculate the fuel-sheath gap conductances under experimental conditions and normal operating conditions in CANDU reactors. The good agreement between the experimentally-inferred and calculated gap conductance values demonstrates that the modified Ross and Stoute model was implemented correctly in ELESTRES. The predictions of the modified Ross and Stoute model provide conservative values for gap heat transfer and fuel surface temperature compared to the offset gap and relocated gap models for a limiting power envelope.
CANDU-6 Heat Transport System Stability Analysis With Canflex Fuel Bundle
Shin, Jung-Cheol ; Park, Ju-Hwan ; Kim, Tae-Han ; Suk, Ho-Chun ;
Nuclear Engineering and Technology, volume 27, issue 3, 1995, Pages 358~373
The Heat Transport system loop stability of CANDU-6 reactor using the CANFLEX fuel bundle was studied. The Thermal-hydraulic behavior of CANFLEX fuel bundle is similar to the conventional 37-element fuel bundle since the reactor power and the frictional pressure drop through the fuel channel is almost the same each other, Mounter the CANFLEX fuel bundle gives higher critical channel power and more homogeneous enthalpy distributions in the subchannels than 37-element fuel bundle. The SOPHT modelling or the CANFLEX fuel bundle and the Reactor outlet Header(ROH) interconnection line was made and the stability analysis response of Wolsong-1 reactor with CANFLEX fuel bundle was obtained. Without the ROH interconnection line the Heat Transport system loop using 43-element fuel bundle is unstable like the current 37-element fuel bundle. With the ROH interconnection line, however, the Heat Transport system is stable within
1％ of nominal flow. In the Heat Transport system loop stability point of view for Wolsong-1 plant therefore, the CANFLEX fuel loading is considered to be acceptable.
Mathematical Adjoint Solution to Analytic Function Expansion Nodal (AFEN) Method
Cho, Nam-Zin ; Hong, Ser-Gi ;
Nuclear Engineering and Technology, volume 27, issue 3, 1995, Pages 374~384
The mathematical adjoint solution of the Analytic Function Expansion (AFEN) method is found by solving the transposed matrix equation of AFEN nodal equation with only minor modification to the forward solution code AFEN. The perturbation calculations are then performed to estimate the change of reactivity by using the mathematical adjoint The adjoint calculational scheme in this study does not require the knowledge of the physical adjoint or the eigenvalue of the forward equation. Using the adjoint solutions, the exact and first-order perturbation calculations are peformed for the well-known benchmark problems (i.e., IAEA-2D benchmark problem and EPRI-9R benchmark problem). The results show that the mathematical adjoint flux calculated in the code is the correct adjoint solution of the AFEN method.
Validation Testing of Safety-critical Software
Kim, Hang-Bae ; Han, Jai-Bok ;
Nuclear Engineering and Technology, volume 27, issue 3, 1995, Pages 385~392
A software engineering process has been developed for the design of safety critical software for Wolsong 2/3/4 project to satisfy the requirement of the regulatory body. Among the process, this paper described the detail process of validation testing peformed to ensure that the software with its hardware, developed by the design group, satisfies the requirements of the functional specification prepared by the independent functional group. To perform the test, test facility and test software ore developed and actual safety system computer was connected. Three kinds of test cases, i.e., functional test performance test and self-check test were programmed and run to verify each functional specifications. Test failures ore fedback to the design group to revise the software and test result were analyzed and documented in the report to submit to the regulatory body. The test methodology and procedure were very efficient and satisfactory to perform the systematic and automatic test. The test results were also acceptable and successful to verify the software acts as specified in the program functional specification. This methodology can be applied to the validation of other safety-critical software.
A Study On The Thermal Movement Of The Reactor Coolant System For PWR
Yoon, Ki-Seok ; Park, Taek sang ; Kim, Tae-Wan ; Jeon, Jang-Hwan ;
Nuclear Engineering and Technology, volume 27, issue 3, 1995, Pages 393~402
The structural analysis of the reactor coolant system mainly consist of too fields. The one is the static analysis considering the impact of pressure and temperature built up during normal operation. The other is the dynamic analysis to estimate the impact of postulated events such as the seismic loads or postulated branch line pipe breaks event. Since the most important goal of the RCS structural analysis is to prove the safety of the RCS during normal operation or postulated events, a widely proven theory having enough conservatism is adopted. The load occurring on the RCS during normal operation is considered as the basic design loading condition throughout whole plant life time. The most typical characteristic of the RCS during normal operation is the thermal expansion of the RCS caused by reactor coolant with high temperature and pressure. Therefore, the exact estimation on the thermal movement of the RCS is needed to get more clear understanding on the thermal movement behavior of the RCS. In this study, the general structural analysis concept and modeling method to evaluate the thermal movement of the RCS under the normal plant operation condition are presented. To discuss the validation of the suggested analysis, analysis results are compared with the measured data which ore referred from the standardized 1000 MWe PWR plant under construction.
Energy Economics of Nuclear and Coal Fired Power Plant
Lee, Gi-Won ; Cho, Joo-Hyun ; Kim, Seong-Rae ; Park, Hae-Yun ;
Nuclear Engineering and Technology, volume 27, issue 3, 1995, Pages 403~417
The upturn of Korean nuclear power program can be considered to have started in early 70's while future plants for the construction of new nuclear power plane virtually came to a halt in United States. It is projected that power plant systems from combination of nuclear and coal fired types might shift to all coal fired type, considering the current trend of construction on the new Plants in the United States. However, with the depletion of natural resources, it is desirable to understand the utilization of two competitive utility technologies in terms of invested energy. Presented in this paper is a comparison between two systems, nuclear power plant and coal fired steam power plant in terms of energy investment. The method of comparison is Net Energy Analysis (NEA). In doing so, Input-Output Analysis (IOA) among industries and commodities is done. Using these information, net energy ratios are calculated and compared. NEA is conducted for power plants in U. S. because the availability of necessary data are limited in Korea. Although NEA does not offer conclusive solution, this method can work as a screening process in decision making. When considering energy systems, results from such analysis can be used as a general guideline.
ANSYS 피로해석 모듈을 이용한 CANDU 6 핵연료채널 응력해석 및 ASME Code에 따른 해석절차 개발
Nuclear Engineering and Technology, volume 27, issue 3, 1995, Pages 418~426
AECL CANDU 중수로형 발전소에서의 컴퓨터 적용 기술
Nuclear Engineering and Technology, volume 27, issue 3, 1995, Pages 427~438
개량형 중수로 비상노심냉각계통의 단순화 및 피동화 방안
Nuclear Engineering and Technology, volume 27, issue 3, 1995, Pages 439~446
경.중수로 연계 핵연료 주기 (DUPIC)관련 핵물질 보장조치 (Safeguards)
Nuclear Engineering and Technology, volume 27, issue 3, 1995, Pages 447~452