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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 24, Issue 3 - Sep 1992
Volume 24, Issue 2 - Jun 1992
Volume 24, Issue 1 - Mar 1992
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Drying Experiment of Borate Waste and Characteristics of Dried Products
Kang, Mun-Ja ; Kim, Hwan-Young ; Kim, Joon-Hyung ;
Nuclear Engineering and Technology, volume 24, issue 3, 1992, Pages 219~227
An experiment was conducted to determine the reaction of boric acid with lime and the drying of its product using a reactor-dryer. no characteristics of dried products were observed. The major chemical species of dried Products was calcium borate of 2CaO.B
. From the particle size distribution of the dried products, it was found that quick lime was better than slaked lime as an additive. The Ca/B mole ratio of reaction was determined to be 3/4 considering the cohesion and agglomeration properties of dried products. The flowability of dried products up to 8 wt% of water content was acceptable for transport process and to reduce drying time.
Mass Transport of Soluble Species Through Backfill into Surrounding Rock
Kang, Chul-Hyung ; Park, Hun-Hwee ;
Nuclear Engineering and Technology, volume 24, issue 3, 1992, Pages 228~235
Some soluble species may not be solubility-limited or congruent-released with the matrix species. For example, during the operation of the nuclear reactor, the fission products can be accumulated in the fuel-cladding gap, voids, and grain boundaries of the fuel rods. In the waste package for spent-fuel placed in a geologic repository, the high solubility species of these fission products accumulated in the“gap”, e.g. cesium or iodine are expected to dissolve rapidly when ground water penetrates fuel rods. The time and space dependent mass transport for high solubility nuclides in the gap is analyzed, and its numerical illustrations are demonstrated. The approximate solution that is valid for all times is developed, and validated by comparison with an asymptotic solution and the solution obtained by the numerical inversion of Laplace transform covering the entire time span.
Optimization of Dynamic Terms in Core Overtemperature Delta-T Trip Function
Park, Jin-Ho ; Yoon, Han-Young ; Kim, Hee-Cheol ; Lee, Chong-Chul ;
Nuclear Engineering and Technology, volume 24, issue 3, 1992, Pages 236~242
The characteristics of dynamic terms in the core overtemperature Delta-T trip function are investigated for various time constants and the effects on the trip setpoint are studied for the uncontrolled RCCA bank withdrawal at power event by using the NLOOP and the PUMA code. Based on this study, a procedure determining the optimal dynamic term is suggested and accordingly the optimum time constants are determined for the KORI 3&4 transition core. It reveals that the vessel average temperature-lead-lag term is the most sensitive in DNB trip setpoint and the optimized time constants are 21 seconds for lead and 4 seconds for lag.
Prediction of Loop Seal Formation and Clearing During Small Break Loss of Coolant Accident
Lee, Sukho ; Kim, Hho-Jung ;
Nuclear Engineering and Technology, volume 24, issue 3, 1992, Pages 243~251
Behavior of loop seal formation and clearing during small break loss of coolant accident is investigated using the RELAP5/MOD 2 and /MOD3 codes with the test of SB-CL-18 of the LSIF (Large Scale Test Facility）. The present study examines the thermal-hydraulic mechanisms responsible for early core uncovery including the manometric effect due to an asymmetric coolant holdup in the steam generator upflow and downflow side. The analysis with the RELAP5/MOD2 demonstrates the main phenomena occuring in the depressurization transient including the loop seal formation and clearing with sufficient accuracy. Nevertheless, several differences regarding the evolution of phenomena and their timing have been pointed out in かe base calculations. The RELAP5/MOD3 predicts overall phenomena, particularly the steam generator liquid holdup better than the RELAP5/MOD2. The nodalization study in the components of the steam generator U-tubes and the cross-over legs wiか the RELAP5/MOD3 results in good prediction of the loop seal clearing phenomena and their timing.
A New Approach for the Solution of Multi-Dimensional Neutron Kinetics Equations in LWR's
Song, Jae-Woong ; Kim, Jong-Kyung ;
Nuclear Engineering and Technology, volume 24, issue 3, 1992, Pages 252~262
The intent of this study is to develop an efficient calculation method which can be used to analyze the heterogeneous time-dependent reactor problems. By using the nodal theory one can not only reduce the calculational efforts, but accurately determine the group dependent flux densities averaged over the entire homogeneous nodes. This method uses correction factors(called“discontinuity factors”) in a rigorous manner to obtain the relationship between the node-averaged flux and the surface-averaged fluxes and currents. The discontinuity factors are calculated from the node-averaged fluxes, diffusion coefficients, and the discontinuity factors of the previous time step. The test results for two benchmark problems demonstrate the accuracy and efficiency of the method developed for the transient application in which assembly-size nodes can be used.
Measurements of Turbulent How in
PWR Rod Bundles With Spacer Grids
Yang, Sun-Kyu ; Chung, Heung-June ; Chun, Se-Young ; Chung, Moon-Ki ;
Nuclear Engineering and Technology, volume 24, issue 3, 1992, Pages 263~273
The study on the velocity distribution and the pressure drop characteristic of the nuclear fuel assembly is of importance for the thermal hydraulic design and safety analysis. The purpose of this experimental study is to investigate the hydraulic mixing behind the different kinds of spacer grids in the now or rod bundles. In this study, the detailed hydraulic characteristics in subchannels of 5
5 PWR(Pressurized Water Reactor) rod bundles were measured using one-component He-Ne LDV(Laser Doppler Velocimeter). Measurements of the axial velocity, turbulent intensities and pressure drops were peformed Lateral velocity, turbulent intensities and Reynolds shear stress were also measured by adjust-ing LDV alignment. Friction factors in rod bundles and loss coefficients for spacer grids were evaluated from the measured pressure drops. Hydraulic mixing performance for different kinds of spacer grids could be investigated by estimating the turbulent cross-flow mixing rates between neighboring subchannels.
An Assessment of the Best Estimate Thermal-Hydraulic Analysis Code CATHARE on CREARE Downcomer Experiment
Chang, Won-Pyo ; Lee, Jae-Hoon ; Kim, Dong-Su ; Chae, Sung-Ki ;
Nuclear Engineering and Technology, volume 24, issue 3, 1992, Pages 274~284
A 1/15-scale CREARE experiment, which simulates the thermal-hydraulic behavior in the reactor pressure vessel of a PWR during a hypothetical Loss Of Coolant Accident, has been analyzed using CATHARE code for the associated model assessment to represent the phenomenon. The key parameters examined in the CREARE experiment were known as ECC water injection rate. ECC water subcooling, system pressure, and steam flow rate coming out from the core bottom. The present CATHARE simulation, however, has been mainly focused on qualitative analysis of a countercurrent flow in the downcomer. The discrepancy of the simulation results with the experimental data is considered arising primarily from an inadequate numerical representation as well as an interfacial friction model. Accordingly it is suggested from the sensitivity studies that either multidimensional approach or further examination of momentum equations at a junction near a volume element in CATHARE be necessary in order to represent the phenomenon more realistically.
Conceptual Development of the Plant Operations Regulator for Nuclear Power Plant Operating Flexibility
Park, Jung-In ; Lee, Myeong-Hoon ; Song, In-Ho ; Oh, Soo-Youl ; Hah, Yung-Joon ;
Nuclear Engineering and Technology, volume 24, issue 3, 1992, Pages 285~296
The conceptual design of the Plant Operations Regulator (POR) is presented for the pressurized water reactor plants. The POR is a digital supervisory limitation control system The POR assures that the plant does not exceed the operating limits by regulating the plant operations through monitoring the operating margins of the critical parameters. The POR is aimed at increasing the operating flexibility which allows the nuclear plant to meet the grid demand in very efficient manner. It responds to the grid demand without penalizing plant availability by limiting the load demand or by modifying the plant control schemes when the operating limits are approached or violated. The POR design concepts were tested using simulation responses of the 1000 MWe pressurized water reactors, Yonggwang Units 3 & 4. The simulation results illustrate that the POR can be used to improve operating flexibility.
Assessments of FLECHT SEASET Unblocked Forced Reflood Tests Using RELAP5/MOD3
Baek, Joo-Seok ; Lee, Won-Jae ; Lee, Sang-Yong ; Kuh, Jung-Eui ;
Nuclear Engineering and Technology, volume 24, issue 3, 1992, Pages 297~310
FLECHT SEASET unblocked forced reflood tests are assessed using Apollo version of RELAP5/MOD3 5M5. The main purpose of the study is to examine the code predictability under forced reflood conditions having different initial power levels and flooding rates. Among various test matrices, the assessment calculations are performed for the test numbers 31701 31302, 31203, 31805, 34524, 31021, 34006 and 35807 These have been selected because they have similar initial conditions but different initial peak rod powers or flooding rates. In addition, various sensitivity calculations are performed for test number 31203 on the improved models of RELAP5/MOD3. Those are for the effect of Counter Current Flow Limit (CCFL) option at the outlet junction of the test section, for the effect of grid modelling on the interfacial drag calculations as well as on the heat structure calculations, and for the effect of nodalization and the time step size. The results of sensitivity studies show that the improved models of RELAP5/MOD3 enhance the code predictability. The assessment results show that the RELAP5/MOD3 has a tendency to underpredict the turn around temperature and the turn around time. But RELAP5/MOD3 silghtly overpredicts the turn around temperature for high flooding rate. The results also show that the calculated quenching by RELAP5/MOD3 is delayed with the increase of the rod power or the decrease of the flooding rate.
A Study on the Free Surface Vortex in the Pipe System
Kim, Sang-Nyung ; Jang, Wan-Ho ;
Nuclear Engineering and Technology, volume 24, issue 3, 1992, Pages 311~318
During mid-loop operation of Nuclear Power Plant, to prevent the Decay Heat Removal System (DHRS) from failure due to air entrainment of free surface vortex in the piping system, a set of simulating experiments was performed. Through these experiments, a relation between the non-dimensionalized numbers, such as H/d, Froude number, Reynolds number, was found. It was also found that the perturbation of the system by the disturbance such as pump start, valve operation, etc., has a strong effect on the free surface vortex. Furthermore, from viewpoint of reactor safety, a modified inlet device which is reducer type is strongly recommended for the prevention of air entrainment into DHRS
Finite Element Analysis of the Neutron Transport Equation in Spherical Geometry
Kim, Yong-Ill ; Kim, Jong-Kyung ; Suk, Soo-Dong ;
Nuclear Engineering and Technology, volume 24, issue 3, 1992, Pages 319~328
The Galerkin formulation of the finite element method is applied to the integral law of the first-order form of the one-group neutron transport equation in one-dimensional spherical geometry. Piecewise linear or quadratic Lagrange polynomials are utilized in the integral law for the angular flux to establish a set of linear algebraic equations. Numerical analyses are performed for the scalar flux distribution in a heterogeneous sphere as well as for the criticality problem in a uniform sphere. For the criticality problems in the uniform sphere, the results of the finite element method, with the use of continuous finite elements in space and angle, are compared with the exact solutions. In the heterogeneous problem, the scalar flux distribution obtained by using discontinuous angular and spatical finite elements is in good agreement with that from the ANISN code calculation.
방사성 폐기물 소멸처리 기술의 현황과 전망
Nuclear Engineering and Technology, volume 24, issue 3, 1992, Pages 329~335
방사선화학의 공업적 이용
Nuclear Engineering and Technology, volume 24, issue 3, 1992, Pages 336~342