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REFERENCE LINKING PLATFORM OF KOREA S&T JOURNALS
> Journal Vol & Issue
Nuclear Engineering and Technology
Journal Basic Information
Journal DOI :
Korean Nuclear Society
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Volume & Issues
Volume 20, Issue 4 - Dec 1988
Volume 20, Issue 3 - Sep 1988
Volume 20, Issue 2 - Jun 1988
Volume 20, Issue 1 - Mar 1988
Selecting the target year
A Study on Determination of Boron Makeup Flow Rate During the Load Follow Operation
Song, Yong-Mann ; Lee, Un-Chul ; Chung, Chang-Hyun ;
Nuclear Engineering and Technology, volume 20, issue 1, 1988, Pages 1~8
During power plant operation, the flow rate from the CVCS makeup system is estimated using the continuity equation and mass balance equation, when the primary loop boron concentration change is required due to the power transient. For this purpose, primary loop, pressurizer and VCT(volume control tank)(in CVCS) are modeled by three control volumes which contain each mass and boron concentration. Connecting pipes between primary loop, pressurizer and CVCS are also modeled by time delay. Calculation for 14-2-6-2 (power 100-50-100) load follow case (at EOL, for KNU 7) is made using these models.
Development of a Method for Optimal Fuel Distribution in 1-D Cylindrical Geometry
Kim, Yun-Ho ; Oh, Soo-Youl ; Kim, Jung-Hwan ; Hong, Seung-Ryong ; Lee, Un-Chul ;
Nuclear Engineering and Technology, volume 20, issue 1, 1988, Pages 9~18
Previously determining the fuel loading pattern is based on the trial and error method. For a candidate pattern, the core analysis is performed and the pattern is examined whether it satisfies the imposed constraints such as the power peaking or not. The pattern, then, is revised by the shuffling of assemblies and the revision is repeated until all of the conditions are met. This method unavoidably requires many iterative diffusion calculations, computing times and accumulated experiences. To overcome these disadvantages, a new method which is called backward diffusion calculation is introduced. If the most desirable power distribution is already known, the optimal loading pattern can be obtained by solving the backward diffusion equation with simple calculation. In this study, the basic equation for the backward diffusion calculation is derived and the optimal power and fuel distributions are searched in one-dimensional cylindrical geometry by using the proposed method. In addition, the basis to determine the optimal power and fuel distributions is suggested for the real core geometry.
Analysis of Burnable Poison Effect on Power Distribution using Power Sensitivity Coefficient Concept
Yi, Yu-Han ; Oh, Soo-Youl ; Seong, Seung-Hwan ; Lee, Un-Chul ;
Nuclear Engineering and Technology, volume 20, issue 1, 1988, Pages 19~26
The low leakage leading pattern has features as the placement of some fresh fuel assemblies in the core interior to reduce the neutron fluence on the pressure vessel and to enhance the neutron economics. But as fresh fuel assemblies are loaded in the core interior, the local power tends to exceed safety limit due to the high reactivity of the fresh assemblies. Therefore, a large number of burnable poisons must be utilized in a low leakage scheme to suppress the high assembly power as well as the excess reactivity. In this study the effects of burnable poisons are treated as a perturbation on the power distribution, and the 'Power Sensitivity Coefficient' concept is adopted. An application study is performed for cycle 1 of the Korea Nuclear Unit-7 (KNU-7) to justify the usefulness of the reverse depletion method coupled with the above concept. To obtain the optimal burnable poision distribution at the given burnup step, the linear programming technique is adopted. The result shows maximum 4.5% error in the amount of burnable poisons between the calculated and the reference values. It is concluded that the design methodology which consists of the reverse depletion, the power sensitivity coefficient concept, and the linear programming technique can be used to find the optimal turnable poison distribution.
The Criticality Analysis of Spent Fuel Pool with Consolidated Fuel in KNU 9 & 10
Jae, Moo-Sung ; Park, Goon-Cherl ; Chung, Chang-Hyun ; Jang, Jong-Hwa ;
Nuclear Engineering and Technology, volume 20, issue 1, 1988, Pages 27~34
Since the lack of the spent fuel storage capcity has been expected for all Korean nuclear power plants in the mid-1990s, the maximum density rack (MDR) with consolidated fuels can be proposed to overcome the shortage of the storage capacity in KNU 9 & 10 which have most limited capacities. To ensure the safety when the alternatives are applied in the KNU 9 & 10, the multiplication factor are calculated with varying the rack pitch and the thickness of consolidated storage box by the AMPX-KENO IV codes. The computing system is verified by the benchmark calculation with criticality experiments for arrays of consolidated fuel modules, which was reported by B & W in 1981. Also an abnormal condition, i.e. malposition accident, is simulated. The results indicate that the KNU 9 & 10 storage pools with consolidated fuel are safe in the view of the criticality. Thus the storage capacity can be expanded from 9/3 cores into 27/3 cores even with considering equipments and cooling spaces.
Analysis of Anisotropic Turbulent Heat Transfer in Nuclear Fuel Bundles
Kim, Sin ; Park, Goon-Cherl ;
Nuclear Engineering and Technology, volume 20, issue 1, 1988, Pages 35~46
The prediction of clad surface temperatures is important to the design and the safety anlaysis of nuclear reactor. The accurate prediction requires the detailed knowledge of the flow structure and heat transfer, which is complicate due to anisotropic turbulent phenomena. A two-equation model including anisotropic eddy viscosity model is applied to forecast the velocity distribution. And the temperature field is calculated with uniform wall heat flux. The Galerkin's weighted residual finite element method has been used to calculate the turbulent quantities right up to the wall. The numerical results show good agreement with available data and that turbulence anisotropy strongly affects on the mean flow and thus the temperature field. And Nu-P/D correlation is established for sodium coolant in close-packed equilateral triangular bundle in the P/D range of 1.05 to 1.30.
A Reliability Analysis of HHSIS of KNU 5,6,7 and 8 Following the Removal of s-signal from Charging/safety Injection Pump Mini-flow Line Valves
Chung, Dae-Wook ; Chung, Chang-Hyun ; Kang, Chang-Soon ;
Nuclear Engineering and Technology, volume 20, issue 1, 1988, Pages 47~53
The objective of this study is to evaluate the reliability of the High Head Safety Injection System (HHIS) of KNU 5, 6, 7 and 8 following the removal of safety injection signal (s-signal) from the mini-flow bypass line valves of charging/safety injection pumps. The unavailability of HHSIS and the rupture probability of a charging/safety injection pump have been computed for two different cases; with s-signal on and removed. The results show that when the s-signal is removed from the mini-flow bypass line valves, the unavailability of HHSIS slightly increases while the rupture probability of a charging/safety injection pump is significantly reduced. Hence, based upon the results of this study we conclude that it is more reasonable to remove the s-signal from the mini-flow bypass line valves of KNU 5, 6, 7 and 8 in the normal plant operation. And to improve the availability of HHSIS, the modification of operational procedures and the emphasis on operator training are recommended.
Assessment of SCDAP Using the Full-Length High-Temperature FLHT-2 Test
Park, Choon-Kyung ; Park, Jong-Hwa ; Yoo, Kun-Jung ; Chae, Sung-Ki ;
Nuclear Engineering and Technology, volume 20, issue 1, 1988, Pages 54~64
This paper assesses the models in the SCDAP code using the results of the FLHT-2 test. Calculations show that the SCDAP correctly predicts Ire temperatures, oxidation front movement, overall hydrogen generation and peak generation rate, internal fuel rod pressure, and cladding rupture due to ballooning. A comparison of the calculated results with measured data shows that two phase level is underpredicted, and that radiation heat transfer and auto-catalytic reaction temperature of zircaloy are overpredicted. These models are recommended to be modified. The analysis also shows that the simulation of the gap in a fuel rod improves the code prediction on core damage progression.
Shielding Design of Shipping Cask for 4 PWR Spent Fuel Assemblies
Kang, Hee-Yung ; Yoon, Jung-Hyoun ; Seo, Ki-Seog ; Ro, Seung-Gy ; Park, Byung-Il ;
Nuclear Engineering and Technology, volume 20, issue 1, 1988, Pages 65~70
A Shielding analysis of the shipping cask designed conceptually, of which shielding material are lead and resin, for containing 4 PWR spent fuel assemblies, has been made with the help of a computer code, ANISN. The shielding materials being used in the cask have been selected and arranged to minimize cask weight while maintaining an overall shielding effectiveness. Radiation source terms have been calculated by means of ORIGIN-2 code under the assumptions of 38,000 MWD/MTU burnup and 3-year cooling time. A calculation of gamma-ray and neutron dose rates on the cask surface and 1m from the surface has been done. It is revealed that the total dose rates under the normal transport and hypothetical accident conditions meet the standards specified.