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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 16, Issue 4 - Dec 1984
Volume 16, Issue 3 - Sep 1984
Volume 16, Issue 2 - Jun 1984
Volume 16, Issue 1 - Mar 1984
Selecting the target year
Study of the RBTRAN Code for Upper Plenum Analysis in Very Small LOCA
Hee Cheon No ;
Nuclear Engineering and Technology, volume 16, issue 3, 1984, Pages 125~130
In the application of the RETRAN code to the analysis of very small LOCA one of main concerns is placed on use of the bubble rise model in the upper plenum, because the bubble rise model nay cause a numerical divergence problem and coefficients used to describe it are based on experimental results of large LOCA. In order to solve this problem, a method, which enables us to predict the mixture level in the upper plenum without use of the bubble rise model, was proposed. For this method the local void distribution in the core and upper plenum was derived by using a simplified slip model. It was shown that results predicted from the derived equation are in excellent agreement with experimental data. Additionally it was found that local void in the upper plenum has a uniform distribution unlike a linear distribution in large LOCA. Communication between the upper plenum and upper head was investigated. By introducing the concept of Taylor instability, it was proved that counter-current Hon between them is possible.
Determination of Sr-90 in the Vetebrae of Reference Korean
Yung Jin Kim ; Gook Hyun Chung ;
Nuclear Engineering and Technology, volume 16, issue 3, 1984, Pages 131~135
The determination of Sr-90 in 93 Korean vertebrae was carried out using modified method of tri-n-butyl phosphate extraction. As a result, the average content of Sr-90 in Korean vertebrae was 2.29 pCi/g Ca in the female and 1.73 pCi/g Ca in the male and the average level of both sexes was 2.01 pCi/g Ca, which was slightly higher than those of other countries. On the other hand, the Sr-90 injected into intraperitonal cavity of rat was accumulated in bones mostly and distributed evenly to various types of hones. The rate of accumulation and removal was not dependent on the amount of Sr-90 injected and over one half Sr-90 injected was accumulated in bones within one day and then it was removed gradually after two days from the injection.
An Off-Site Consequence Modeling for Accident Using Monte Carlo Method
Chang Sun Kang ; Sae Yul Lee ;
Nuclear Engineering and Technology, volume 16, issue 3, 1984, Pages 136~140
A new medal is presented in order to evaluate the risk from a nuclear facility following accidents directly combining the on-site meteorological data using the Monte Carlo Method. To estimate the radiological detriment to the surrounding population-at-large (collective dose equivalent), in this study the probability distribution of each meteorological element based upon on-site data is analyzed to generate atmospheric dispersion conditions. The random sampling is used to select the dispersion conditions at any given time of effluent releases. In this study it is considered that the meteorological conditions such as wind direction, speed and stability are mutually independent and each condition satisfies the Markov condition. As a sample study, the risk of KNU-1 following the large LOCA was calculated, The calculated collective dose equivalent in the 50 mile region population from the large LOCA with 50 percent confidence level is 2.0
Severe Accident Sequence Analysis - Part 1: Analysis of Postulated Core Meltdown Accident Initiated by Small Break LOCA in Kori-1 PWR Dry Containment
Jong In Lee ; Seung Hyuk Lee ; Jin Soo Kim ; Byung Hun Lee ;
Nuclear Engineering and Technology, volume 16, issue 3, 1984, Pages 141~154
An analysis is presented of key phenomena and scenario which imply some general trends for beyond design-basis-accident in Kori-1 PWR dry containment. The study covers a wide range of severe accident sequences initiated by small break LOCA. The MARCH computer code, with KAERI modifications was used in this analysis. The major emphasis of the paper are two folds, 1) the phenomenologic understanding of severe accident and 2) a study of H2 combustion and debris/ water interactions in a specific small break LOCA for Kori-1 plant. The sensitivity studies for the specific plant data and thermal interaction modelings used in the SASA were performed. The results show that if hydrogen burning does occur at low concentration, the resulting peak pressure does not exceed the design value, while the lower concentration assumption results in repeated burning due to the continuing H
generation. For debris/water interaction, the particle size has no effect on the magnitude of peak pressure for the amount of water assumed to be in the reactor cavity. But, the occurrence of peak pressure is considerably delayed in case of using the dryout correlation. The peak containment pressure predicted from the hydrogen combustion and steam pressure spite during full core meltdown scenario does not present a severe threat to the containment integrity.
Thermal Margin Analysis of the Korea Nuclear Unit 1 Reactor Core Consisting of Standard or Optimized Fuel Assemblies
Hyun Koon Kim ; Ki In Han ;
Nuclear Engineering and Technology, volume 16, issue 3, 1984, Pages 155~160
Analyzed is the thermal margin of the Korea Nuclear Unit 1 (KNU-1) reactor core consisting of either 14 x 14 standard fuel assemblies (SFA) or optimized fuel assemblies (OFA). Employed for the analysis are two different thermal design methods; traditional and statistical thermal design method. Compared to the traditional design thermal method, the statistical thermal design method improves the core thermal margin utilizing best-estimate values for the core operating parameters combining their uncertainties in a statistical manner. Calculations are performed using a steady state and transient thermal-hydraulic analysis computer program, COBRA-IV-i. Calculated results show that the statistical thermal design method significantly improves the thermal margin and satisfies the core thermal design base of the KNU-1 SFA and OFA core. However, the thermal design base can not be met, if the traditional thermal design method is employed for the OFA role analysis.
A Study on the Removal Efficiency of a TEDA Impregnated Charcoal Bed for Methyl iodide under Humid Conditions
Won Jin Cho ; Soon Heung Chang ;
Nuclear Engineering and Technology, volume 16, issue 3, 1984, Pages 161~168
The adsorption model to predict the time dependent removal efficiency of methyl iodide by triethylenediamine (TEDA) impregnated charcoal bed under humid condition is proposed. Under humid conditions, the reduction of equilibrium adsorption capacity and effective pore diffusivity is considered. The predicted values are compared with the experimental results.
Evaluation of Fuel Cladding Failures from the Fission Product Activities in the Reactor Coolant
Ho Ju Moon ; Sung Ki Chae ;
Nuclear Engineering and Technology, volume 16, issue 3, 1984, Pages 169~179
An efficient procedure of evaluating the fuel cladding failures occurring in the normal operations of typical PWR's has been investigated through the analysis of fission product(FP) activities in the reactor coolant using an analytical model, FIPREL code. Performed by this code is an extensive study on the sensivities of FP activities to such physical parameters as enrichment, turnup, and operation temperature of failed fuel rod as well as the effective failure size quantified in terms of the magnitude of gap release coefficient. The results of study are generally in agreement with those by PROFIP method. In the presence of tramp uranium the portion of activities released from failed rod is separated by an iterative calculation based on the activity ratios of fission nuclides chemically more stable than iodines. Obtained are the linear power density and the number of failed rods, the effective failure size, and the mass of tramp uranium. The operation experiences of 4 cycles of Kori Unit 1 are analyzed and the results show that the model is highly reliable for the survey and evaluation of fuel rod conditions during reactor operations.