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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
Development of a Prototype Expert System for Intelligent Operation Aids in Rod Consolidation Process
Kim, Ho-Dong ; Kim, Ki-Joo ; Yoon, Wan-Ki ;
Nuclear Engineering and Technology, volume 25, issue 1, 1993, Pages 1~7
This paper describes a prototype expert system to aid operation in rod consolidation process. The knowledgebase is composed of three database groups and 60 rules with production, and object oriented techniques that correlates database groups. The expert system is designed to track the transitions of nuclear materials through the operation areas of the rod consolidation process, to diagnose current status in any operating conditions, normal and off-normal, and to advise operators to properly recover off-normality. The expert system can give efficient management of nuclear material accountability and process operation in the rod consolidation.
A Study on the Local Boiling of the Consolidated Spent Fuel Storage Pool
Lee, Chang-Ju ; Lee, Kun-Jai ;
Nuclear Engineering and Technology, volume 25, issue 1, 1993, Pages 8~19
The natural convection model of the consolidated system has been developed to make sure the removal of decay heat generated in the spent fuel for the loss of forced cooling accident. The numerical technique employed was based on the ADI scheme. The calculation of heat generation rate in the spent fuel was peformed by the ANS-79 decay heat model, and the nonuniform surface heat flux is assumed with a chopped sine curve for the conservative decay heat generation input. The sensitivity study was performed to examine the possibility of the pool bulk boiling by varying the various parameters, i.e. inter-fuel spacing ratio, heat generation power, and radius of the fuel rod. The application results of this model show that the natural circulation flow through compacted spent fuel bundles enables the pool temperature to control in a safe and effective manner, after the required cooling time. The corresponding acceptance criteria of the cooling time for rearranging the spent fuel rods were also found.
Development of In-Core Fuel Management Scoping Tools for PWR
Kim, Chang-Hyo ; Kim, Teak-Kyum ;
Nuclear Engineering and Technology, volume 25, issue 1, 1993, Pages 20~27
This paper concerns with developing a simplified in-core fuel management scoping tool for PWR. For this purpose the point reactivity model is put into a fuel cycling decision code, FCYPRM. Modified Borresen's coarse-mesh diffusion theory and nodal expansion method are utilized to form a spatial neutron analysis code, CMSNAP. Numerical experiments are per- formed to determine a set of empirical shuffling rules for working out an automated fuel loading pattern search code, ALPS. The numerical examples are presented for verifying effectiveness and applicability of individual codes. By structuring and applying three codes for reload core design problem of a PWR, it is demonstrated that these codes provide an effective in-core fuel management scoping tool for PWR.
Study on the Seismic Analysis of the Reactor Vessel Internals
Jhung, Myung-Jo ; Park, Keun-Bae ; Hwang, Won-Gul ;
Nuclear Engineering and Technology, volume 25, issue 1, 1993, Pages 28~36
Much effort is being done to standardize the PWR-type nuclear power plant in Korea. This paper presents the development of seismic design criteria for the reactor internals as a part of the standardization program for nuclear power plant. The seismic design loads of the reactor internals are calculated using the reference input motions of reactor vessel taken from Yong-gwang Nuclear Power Plant Units 3 and 4. An overview of analysis related to the basic parameters and methodologies is presented. Also, the response of internal components for the reactor vessel motions is carefully investigated.
Evaluation of Total Loss of Feedwater Accident/Recovery Phase and Investigation of the Associated EOP
Bang, Young-Seok ; Seul, Kwang-Won ; Kim, Hho-Jung ;
Nuclear Engineering and Technology, volume 25, issue 1, 1993, Pages 37~50
To evaluate the sequence of event and the Thermohydraulic behavior during total loss of feedwater accident and recovery procedure, a RELAP5/MOD3 calculation is performed and compared with the LOFT L9-l/L3-3 experiment. Also, the predictability of the code for the major Thermohydraulic phenomena following the accident is assessed. As a result, it is found that a pressure control using the spray until the time the water level reaches the top of the pressurizer, an overpressure protection by pressurizer PORV, a recovery of the secondary heat removal capability by refilling steam generator, and an effective cooldown by the continued natural circulation can be performed without core uncovery. It is also found that the plant-specific evaluation is necessary to confirm the effectiveness of the current symptom-oriented emergency operating procedure, especially in an overpressure protection performance and steam generator recovery performance.
Development of Radiation Shielding Analysis Program Using Discrete Elements Method in X-Y Geometry
Park, Ho-Sin ; Kim, Jong-Kyung ;
Nuclear Engineering and Technology, volume 25, issue 1, 1993, Pages 51~62
A computational program ［TDET］ of the particle transport equation is developed on radiation shielding problem in two-dimensional cartesian geometry based on the discrete element method. Not like the ordinary discrete ordinates method, the quadrature set of angles is not fixed but steered by the spatially dependent angular fluxes. The angular dependence of the scattering source term in the particle transport equation is described by series expansion in spherical harmonics, and the energy dependence of the particles is considered as well. Three different benchmark tests are made for verification of TDET : For the ray effect analysis on a square absorber with a flat isotropic source, the results of TDET calculation are quite well conformed to those of MORSE-CG calculation while TDET ameliorates the ray effect more effectively than S
calculation. In the analysis of the streaming leakage through a narrow vacuum duct in a shield, TDET shows conspicuous and remarkable results of streaming leakage through the duct as well as MORSE-CG does, and quite better than S
calculation. In a realistic reactor shielding situation which treats in two cases of the isotropic scattering and of linearly anisotropic scattering with two groups of energy, TDET calculations show local ray effect between neighboring meshes compared with S
calculations in which the ray effect extends broadly over several meshes.eshes.
A Conservative Safety Study on Low-Level Radioactive Waste Repository Using Radionuclide Release Source Term Model
Kim, Chang-Lak ; Lee, Myung-Chan ; Cho, Chan-Hee ;
Nuclear Engineering and Technology, volume 25, issue 1, 1993, Pages 63~70
A simplified safety assessment is carried out on rock-cavern type disposal of LLW using the analytical repository source term (REPS) model. For reliable prediction of the leach rates for various radionuclides, degradation of concrete structures, corrosion rate of waste container, degree of corrosion on the container surface, and the characteristics of radionuclides are considered in the REPS model. The results of preliminary assessment show that Cs-137, Ni-63, and Sr-90 are dominant. For the parametric uncertainty and sensitivity analysis, Latin hypercube sampling technique and rank correlation technique are applied. The results of the potential public health impacts show that radiological dose to intruder in the worst case scenario will be negligible and that more attention should be given to near-field performance.
A Study on the Defect Annealing of Hafnium Metal By Positron Annihilation Techniques
Kang, Myung-Soo ; Jung, Sung-Hoon ; Yoon, Young-Ku ; Park, Yong-Ki ;
Nuclear Engineering and Technology, volume 25, issue 1, 1993, Pages 71~79
Positron annihilation characteristics and microhardness of 25% cold worked and isochronally annealed hafnium specimens were measured to study recovery and recrystallization stages of hafnium specimens. The annihilation lifetime of positrons in hafnium has been measured for the distinct cases of annihilation in the annealed lattice and annihilation after trapping at lattice defects generated by cold deformation at room temperature. The annihilation lifetime in the annealed lattice was 187
3.7 psec, whereas it was 217
4.2 psec for positrons trapped at deformation-induced defects (mostly dislocations). The changes in Doppler broadening and hardness showed similar trend in the recrystallization range, however, the measured value of Doppler broadening variation were quite sensitive to changes in the recovery region in which the variation in hardness value was completely insensitive. Recovery of cold worked hafnium initiated at about 623 K and recrystallization occurred at around 1023 K.
Generation of Gamma-Ray Streaming Kernels Through Cylindrical Ducts Via Monte Carlo Method
Kim, Dong-Su ; Cho, Nam-Zin ;
Nuclear Engineering and Technology, volume 25, issue 1, 1993, Pages 80~90
Radiation streaming through penetrations has been of great concern in radiation shielding design and analysis. This study developed a Monte Carlo method and constructed a data library of results calculated by the Monte Carlo method for radiation streaming through a straight cylindrical duct in concrete walls of a broad, mono-directional, mono-energetic gamma-ray beam of unit intensity. It was demonstrated that average dose rate due to an isotropic point source at arbitrary positions can be well approximated using the library with acceptable error. Thus, the library can be used for efficient analysis of radiation streaming due to arbitrary distributions of gamma-ray sources.
Measurements of Void Concentration Parameters in the Drift-Flux Model
B.J. Yun ; Park, G.C. ; C.H. Chung ;
Nuclear Engineering and Technology, volume 25, issue 1, 1993, Pages 91~101
To predict accurately the thermal hydraulic behavior of light water reactors during normal or abnormal operation, the accurate estimation of the void distribution is required. Up to date, many techniques for predicting void fraction of two-phase flow systems have been suggested. Among these techniques, the drift-flux model is widely used because of its exact calculation ability and simplicity. However, to get more accurate prediction of void fraction using drift-flux model, slip and flow regime effects must be considered more properly In the drift-flux method, these two effects are accounted for by two drift-flux parameters ;
and (equation omitted). At earlier stage,
is measured in a circular tube. In this study,
is experimentally determined by measuring local void fraction and vapor velocity distribution in a rectangular subchannel having 4 heating rods which simulates nuclear subchannels. The measurements are peformed with two-electrical conductivity probes which are known to be adequate for measuring local parameters. The experiments are performed at low flow rate and the system pressure less than 3 atmo spheric pressure. In this experiment, (equation omitted), is not measured, but quoted from well-known empirical correlation to formulate
is expressed as a function of channel averaged void fraction. fraction.
RADAP-A PC Program for Real-Time Prediction of Doses Following a Nuclear Accident
Park, Jae-Won ; Kang, Chang-Sun ;
Nuclear Engineering and Technology, volume 25, issue 1, 1993, Pages 102~109
A PC-computer program RADAP has been developed in this study to perform a quick real-time analysis of dose assessment following an accident in a nuclear facility. RADAP uses an interactive LKagrangian puff model in simulating the transport and diffusion of radioactive plume in the atmosphere. For real-time analysis, RADAP treats one or multiple puffs of ground-level releases, simultaneously. It is assumed to maintain a Gaussian distribution within the puff and the diffusion coefficients are computed using the USNRC's normal sigma curve method. The program, however, does not consider the spatial variations but the temporal variations in wind conditions. Whole body and thyroid doses for 3
31 grid are directed to output files, and they are also displayed through computer graphics on VGA or EGA color monitor. The results show that RADAP can be an excellent tool for quick estimation of accidental doses.
Nonlinear Model-Based Robust Control of a Nuclear Reactor Using Adaptive PIF Gains and Variable Structure Controller
Park, Moon-Ghu ; Cho, Nam-Zin ;
Nuclear Engineering and Technology, volume 25, issue 1, 1993, Pages 110~124
A Nonlinear model-based Hybrid Controller (NHC) is developed which consists of the adaptive proportional-integral-feedforward (PIF) gains and variable structure controller. The controller has the robustness against modeling uncertainty and is applied to the trajectory tracking control of single-input, single-output nonlinear systems. The essence of the scheme is to divide the control into four different terms. Namely, the adaptive P-I-F gains and variable structure controller are used to accomplish the specific control actions by each terms. The robustness of the controller is guaranteed by the feedback of estimated uncertainty and the performance specification given by the adaptation of PIF gains using the second method of Lyapunov. The variable structure controller is incorporated to regulate the initial peak of the tracking error during the parameter adaptation is not settled yet. The newly developed NHC method is applied to the power tracking control of a nuclear reactor and the simulation results show great improvement in tracking performance compared with the conventional model-based control methods.
Digitalization of the Nuclear Steam Generator Level Control System
Lee, Yoon-Joon ; Lee, Un-Chul ;
Nuclear Engineering and Technology, volume 25, issue 1, 1993, Pages 125~135
The safe and efficient operation of nuclear plants is recognized to be accomplished through the application of plant automation using digital technology, which is one of main targets of the next generation nuclear plants. For plant level automation, it is first required that each major subsystem be digitalized, and the steam generator water level control system is discussed in this study. The transfer functions between inputs and the level are derived by employing the thermal hydraulic model of the steam generator and are applied to the analysis of the current three-element control system. Since the control scheme in this study includes the steam generator itself as a process plant, the system order is high and the numerical instability arises in digitalizing. Together with this, the unreliability of the feedwater feedback signal at low power level leads to the proposal of a two-element control system with a proper digital controller. The digital PI controller developed for this system has the initial power adaptive gain and integration time constant. And it makes the overall system response satisfy the stability and other necessary control specifications simultaneously. Since the two-element control system using this controller depends on the initial power only, it is simple to define and it shows a similar level response behavior to that of its corresponding analog system.
Control of Outmost Poloidal Flux Surface of Tokamak Plasma in RTP
Lee, Kwang-Won ; Oh, Byung-Hoon ;
Nuclear Engineering and Technology, volume 25, issue 1, 1993, Pages 136~147
The paper describes : ⅰ) Mathematical modeling of poloidal flux to define and calculate the tokamak plasma position based on a property of the plasma boundary which is always a flux surface. Controlling the plasma boundary position is therefore equivalent to equalizing the flux value on several points belonging to a curve tangent to the limiter. ⅱ) Experimental method for determining the outmost poloidal isoflux surface by a linear combination of measurements of magnetic fluxes, fields and field gradients, without requiring knowledge of internal plasma parameters for the feedback control, i.e., with neither corrections for variation in the poloidal beta and the plasma current distribution, nor compensations for the induced currents in the vacuum vessel. ⅲ) Feedback control algorithm for the regulation of plasma boundary position and its electronics hardware based on the PID control theory. ⅳ) Experimental results obtained from the RTP tokamak experiments using the present plasma control system.
Evaluation of Axial Buckling Effect in On-Line Axial Power Shape Synthesis
In, Wang-Kee ; Kim, Joon-Sung ; Yoon, Tae-Young ; Auh, Geun-Sun ; Kim, Hee-Cheol ;
Nuclear Engineering and Technology, volume 25, issue 1, 1993, Pages 148~153
A fifth-order Fourier series technique is applied in Core Operating Limit Supervisory System (COLSS) to construct the on-line core average axial power shape from in-core detector signals because of its simplicity and fast computation. Such a synthesizing accuracy depends on number of Fourier series modes and axial boundary conditions. COLSS currently uses the five-mode Fourier series technique which utilizes the five axially located fixed in-core detector signals and a constant axial boundary condition. Therefore, the constant axial boundary condition should be appropriately chosen based on the evaluation of its effect on the accuracy of the on-line calculations. The four cases of axial buckling (0.75, 0.8, 0.9 and 1.0) were examined for Yonggwang Nuclear Units 3&4 as the axial boundary conditions in this paper. The core average axial power shapes and the operating margins were compared for each case to determine the optimal constant axial buckling. The axial buckling of 0.9 was found to be the optimal value.
A Stochastic Model for the Nuclide Migration in Geologic Media Using a Continuous Time Markov Process
Lee, Y.M. ; C.H. Kang ; P.S. Hahn ; Park, H.H. ; Lee, K.J. ;
Nuclear Engineering and Technology, volume 25, issue 1, 1993, Pages 154~165
A stochastic method using continuous time Markov process is presented to model the one-dimensional convective nuclide transport in geologic media, which have usually heterogeneous feature in physical/geochemical parameters such as velocity, dispersion coefficient, and retardation factor resulting poor description by conventional deterministic advection-dispersion model. The primary desired quantities from a stochastic model are the mean values and variance of the state variables as a function of time. The time-dependent probability distributions of nuclides are presented for each discretized compartment given the volumetric groundwater flux and the intensity of transition. Since this model is discrete in medium space, physical/geochemical parameters which affect nuclide transport can be easily incorporated for the heterogeneous media as well as remarkably layered media having spatially varied parameters. Even though the Markov process model developed in this study was shown to be sensitive to the number of discretized compartments showing numerical dispersion as the number of compartments are increased, this could be easily calibrated by comparing with the analytical deterministic model.
Parametric Effects of Ambient Conditions on Thermal Safety of Wolsong (CANDU) Unit 1 Spent Fuel Dry Storage Canister
Park, Jong-Woon ; Chun, Moon-Hyun ; Shon, Soon-Hwan ; Song, Myung-Jae ;
Nuclear Engineering and Technology, volume 25, issue 1, 1993, Pages 166~177
A simplified thermal analysis method to evaluate the maximum temperature of the CANDU 37-element fuel bundle within a fuel basket in a given spent fuel dry storage canister has been presented along with the results of sample analyses performed to examine the parametric effects of the ambient conditions on the maximum fuel temperature within a canister. To solve the multi-dimensional heat transfer problem of the complex geometry of rod bundles within a canister where three modes of heat transfer are superimposed, the CANDU spent fuel bundles stored in the dry storage canister are first replaced by equivalent concentric fuel cylinders. The simplified axi-symmetric two-dimensional multi-mode heat transfer problem of the equivalent fuel cylinders is then analyzed with an existing computer code, HEATING5, using additional input data and heat transfer correlations. A comparison between the predicted temperature profile and the mock-up test results shows that the agreement is quite satisfactory.
An Experimental Study on the Characteristics of Electromagnetic Filter
Lee, Geum-Yong ; Lim, Seung-Cheol ; Lee, Kun-Jai ;
Nuclear Engineering and Technology, volume 25, issue 1, 1993, Pages 178~191
The electromagnetic filter has been recognized as a technological replacement for the conventional filtration systems of the nuclear power plant coolant. But, as of now there are neither clear understandings of the phenomena occurring in the electromagnetic filter nor the general theoretical analyses. These facts make the application or the electromagnetic filter to the real systems a little risky, and therefore it has not been commercialized although it shows excellent performances in such situations as the plant abnormality, where the conventional filters usually fail. This experimental study of the low power electromagnetic filter aims at the clarification of the general characteristics under varying operational parameters. Since the detailed characteristics may differ from one electromagnetic filter to another, they are considered secondary. The impurities applied are the highly magnetic magnetite (Fe
) and the diamagnetic cuprous oxide (Cu
O). The empirical equations are derived from the experimental data by the regressional analyses. They are classified of three types : Efficiencies vs. Time, Efficiencies vs. Load, and Load vs. Time. The characteristics of the electromagnetic filter observed in this experiment agreed well with other related works in many aspects. Especially in this study, some assumptions and discussions including the physical deposition are combined for the explanations of the filter characteristics found in our and other experimental works.