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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 15, Issue 4 - Dec 1983
Volume 15, Issue 3 - Sep 1983
Volume 15, Issue 2 - Jun 1983
Volume 15, Issue 1 - Mar 1983
Selecting the target year
Proposed Method to Predict Core Inventory history and Operator Time Margin during Small Break Accident
Hee Cheon No ;
Nuclear Engineering and Technology, volume 15, issue 4, 1983, Pages 219~228
The blowdown history of the TMI-2 accident up to the isolation of the relief valve associated with a small break LOCA is reviewed briefly. An analysis is made to determine what instruments should be added in the core in order to prevent core damage in the case of the TMI-2 accident. With the added instruments a procedure is presented on how to predict the uncovered level of the core and how to calculate operator time margin. Sample calculations are done for the TMI-2 accident to determine the uncovered level and operator time margin. Finally, the map to show the uncovered level of the core and operator time margin is drawn with measurable parameters by the above methods.
The Common Mode Failures Analysis of The Redundant System with Dependent Human Error
Myung Ki Kim ; Soon Heung Chang ;
Nuclear Engineering and Technology, volume 15, issue 4, 1983, Pages 229~235
Common Mode Failures (CMFs) have been a serious concern in the nuclear power plant. There is a broad category of the failure mechanisms that can cause common mode failures. This paper is a theoretical investigation of the CMFs on the unavailability of the redundant system. It is assumed that the total CMFs consist of the potential CMFs and the dependent human error CMFs. As the human error dependence is higher, the total CMFs are more effected by the dependent human error. If the human error dependence is lower, the system unavailability strongly depends on the potential CMFs, rather than the mechanical failure or the dependent human error. And it is shown that the total CMFs are dominant factor to the unavailability of the redundant system.
DNBR Sensitivities to Variations in PWR Operating Parameters
Hyun Koon Kim ; Ki In Han ;
Nuclear Engineering and Technology, volume 15, issue 4, 1983, Pages 236~247
Analyzed are the the DNBR(Departure from Nucleate Boiling Ratio) sensitivities to variations in various PWR operating parameters utilizing the Korea Nuclear Unit 1(KNU-1) design and operating data. Studied parameters in the analysis are core power level, system pressure, core inlet flow rate, core inlet temperature, enthalpy rise hot channel factor, and axial power peaking factor and axial offset. The calculations are performed using the steady state and transient thermal-hydraulics computer program, COBRA-IV-K, which is the revised version of COBRA-IV-i that has been adapted, partially modified and verified at KAERI. A reference case is established based on the design and operating condition of the KNU-1 reactor core, and this provides a basis for the subsequent sensitivity analysis. From the calculation results it is concluded that the most sensitive parameter in the DNBR thermal design is the coolant core inlet temperature while the axial power peaking factor is the least sensitive.
Thermal Analysis on the Spent Fuel Shipping Cask for a PWR Fuel Assembly
Hee Yung Kang ; Eun Ho Kwack ; Byung Jin Son ;
Nuclear Engineering and Technology, volume 15, issue 4, 1983, Pages 248~255
The thermal analysis on the spent fuel shipping cask for a PWR fuel assembly is performed. Under the normal and fire-accident conditions the temperature distribution through a multilayer cask calculated in compliance with 10 CFR Part 71. A KNU 5&6 spent fuel assembly is assumed to be the decay heat source, which has the maximum discharge turnup of 45, 000MWD/MTU and has been stored in the spent fuel storage pool for 300 days. As a result of thermal analysis, the maximum cladding temperature in case of dry cavity under fire-accident conditions is calculated to be 455
. This value is much less than the limiting value specified in 10 CFR Part 50.46. It indicates that no fuel rod cladding rupture could occur under fire-accident conditions. It was also found that no melting of lead would take place in the major shield region.
An Evaluation of Cooling of Core Debris and Impact on Containment Transient Pressure under Severe Accident Conditions
Jong In Lee ; Jin Soo Kim ; Byung Hun Lee ;
Nuclear Engineering and Technology, volume 15, issue 4, 1983, Pages 256~266
An evaluation of containment transient pressure due to the particulate debris/water/concrete interaction under severe accident conditions is presented for a pressurized water reactor with a large dry containment building. A particulate debris/water/concrete model is developed and incorporated into the MARCH computer code. Comparisons with the existing MARCH molten debris/concrete model were performed for the TMLB' and S
D sequences. The results yield a much slower concrete decomposition rate and release less gases into the containment atmosphere. Contrary to the molten debris model, the particulate debris model exhibits a strong interaction with water and causes a higher containment pressure. The effect of gas influx on the debris bed heat transfer was found to be insignificant.
Calculation of Power Distributions on Uranium- and Plutonium-Loaded Cores Moderated by Light Water
Sang Keun Lee ; Kap Suk Moon ; Jong-Hwa Jang ; Ji Bok Lee ; Chang Kun Lee ;
Nuclear Engineering and Technology, volume 15, issue 4, 1983, Pages 267~279
An analytical system has been established for scrutinizing both uranium- and plutonium-fueled lattices moderated by light water. This system consists of two primary codes. One is a unit cell program called KICC, which has theoretical foundation on the models of GAM and THERMOS incorporated with appropriate approximate treatments for various phenomena, whereas the other is a multi-dimensional diffusion-depletion program entitled KIDD. The adequacy of this system is verified by performing extensive benchmark calculations on a variety of critical experiments. The average value of effective multiplication factors for the selected nineteen UO
critical experiments of heterogeneous lattice structure is calculated to be 1.0006 with a standard deviation of 0.0039. Power distributions have also been calculated for some critical experiments fueled with both uranium and plutonium of varying concentrations. The maximum percentage difference between the measured and calculated power distributions appears to be less than 5%. This result, together with the previously reported result, illustrates that the KICC/KIDD system is a very effective tool for the analysis of a light water reactor core.
Effects of Fission Neutron Spectra in Reactor Calculations
Nuclear Engineering and Technology, volume 15, issue 4, 1983, Pages 280~285
Effects of fission neutron spectra in the reactor calculations have been analysed through applications of several cases of spectra in the criticality calculations of fast critical assemblies. They were the application of Maxwellian or Watt-Cranberg type formulae, of region dependent spectrum, of composition dependent spectrum, of fission transfer matrix, and the effects due to the selection of nuclear temperature in Maxwellian formula.
법정판결의 당위성과 리스크
Nuclear Engineering and Technology, volume 15, issue 4, 1983, Pages 286~289