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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
Editor in Chief :
Volume & Issues
Volume 29, Issue 6 - Dec 1997
Volume 29, Issue 5 - Oct 1997
Volume 29, Issue 4 - Aug 1997
Volume 29, Issue 3 - Jun 1997
Volume 29, Issue 2 - Apr 1997
Volume 29, Issue 1 - Feb 1997
Selecting the target year
Emittance Measurements of the Ion Sources for Induction Linac Driven Heavy Ion Fusion
Lee, Heon-Ju ;
Nuclear Engineering and Technology, volume 29, issue 3, 1997, Pages 181~185
The ion sources for induction linac driven heavy ion fusion were fabricated and their omittance characteristics were investigated. For to kinds of ion sources, i. e. a carbon vacuum arc ion source and a cusp field rf ion source, the emittance was measured with a double slit beam scanner. The required normalized omittance of an ion source for heavy ion fusion is 10
m-rod, and the measured emittances of the ion beams from carbon vacuum arc ion source and cusp field rf ion source (Ne
) were 2
m-rad and 4
A Study on the Coolant Mixing Phenomena in the Reactor Lower Plenum
Park, Yong-Seog ; Park, Goon-Cherl ; Um, Kil-Sup ;
Nuclear Engineering and Technology, volume 29, issue 3, 1997, Pages 186~195
When asymmetric thermal-hydraulic conditions occur between cold legs, the core inlet temperature will be nonuniform if the coolant is not mixed perfectly in the lower plenum. These uneven core inlet conditions may induce the change in core power distribution. Thus realistic prediction of thermal mixing is important in such abnormal conditions. In this study, reactor internals, which are scaled down as to conserve the flow area ratio, are set up in the model of KORI Unit 1 with the scaling factor of 1/710 by volume and coolant temperatures are measured beneath the lower core plate. Based on experimental results, the ability of COMMIX-1B code to simulate the coolant mixing phenomena in the lower plenum is estimated. The results show that complete mixing never occurs in any conditions and the mixing pattern is characterized according to the plant type.
Powder Types on Sintered Density and Grain Size of the
Yoo, Ho-Sik ; Kim, Hyung-Soo ;
Nuclear Engineering and Technology, volume 29, issue 3, 1997, Pages 196~200
The variation of sintered density and fain size in ex-AUC, ex-ADU and granulated ex-ADU UO
pellets in which 0.1~1.0wt% Nb
were doped were examined. Pellets were sintered in an atmosphere of H
for 4h. All the specimens tested shooed more than 94% T.D.(Theoretical Density). Sintered density decreased with increasing the amount of Nb
. Powder types had little influence on the sintered density. Pore size distribution was shifted to the larger ones as Nb
was added. The increase of total pore volume and grain growth due to the addition of Nb
were thought to be the cause of the sintered density decrease. The largest grain size was seen in the 1. 0wt% Nb
doped ex-AUC UO
pellets. Their average size was 13.9
Modelling of Thermal Conductivity for High Burnup
Fuel Retaining Rim Region
Lee, Byung-Ho ; Koo, Yang-Hyun ; Sohn, Dong-Seong ;
Nuclear Engineering and Technology, volume 29, issue 3, 1997, Pages 201~210
A thermal conductivity correlation has been proposed which can be applied to high turnup fuel by considering both of thermal conductivity with turnup across fuel pellet and additional degradation at pellet rim due to very high porosity. In addition, a correlation has been developed that can estimate the porosity of rim region as a function of rim burnup under the assumptions that all the produced fission gases are retained in the in porosity and threshold pellet average burnup required for the formation of rim region is 40 MWD/㎏U. Rim width is correlated to rim burnup using measured data. For the RISO experimental data obtained at pellet average turnup of 43.5 MWD/㎏U for three linear heat generation rates of 30, 35 and 40 ㎾/m, radial temperature distributions ore calculated using the present correlation and compared with the measured ones. This comparison shows that the present correlation gives the best agreement with the measured data when it is combined with the HALDEN's correlation for thermal conductivity considering its degradation with burnup. Another comparison with the HALDEN's measured fuel centerline temperature as a function of burnup at 25 ㎾/m up to about 44 MWD/㎾U also suggest that the present correlation yields the best agreement when it is combined with the HALDEN's thermal conductivity.
Development of STI/AOT Optimization Methodology and an Application to the AFWPs with Adverse Effects
You, Young-Woo ; Yang, Hui-Chang ; Chung, Chang-Hyun ; Moosung Jae ;
Nuclear Engineering and Technology, volume 29, issue 3, 1997, Pages 211~217
Adverse effects caused by the surveillance test for the components of nuclear power plant involve plant transients, unnecessary wear, burden on licensee's time, and the radiation exposure to personnel along with the characteristics of each component. The optimization methodology of STI and AOT has been developed and applied to AFWPs of a reference plant. The approach proposed in this paper consist of the resole in minimal mean unavailability of the two-out-of-four system with adverse effects are analytically calculated for the example system. The surveillance testing strategy are given by the sequential test, the staggered test and the train staggered test which is a mined test scheme. In the system level, the sensitivity analyses for the STI and AOT, are performed for the measure of the system unavailability of the top event in the fault tree developed for the example system. This methodology may contribute to establishing the basis for the risk-based regulations.
Prediction of the Volume of Solid Radioactive Wastes to be Generated from Korean Next Generation Reactor
Cheong, Jae-Hak ; Lee, Kun-Jai ; Maeng, Sung-Jun ; Song, Myung-Jae ; Park, Kyu-Wan ;
Nuclear Engineering and Technology, volume 29, issue 3, 1997, Pages 218~228
Correlations between the amount of DAW (Dry Active Waste) generated from present Korean PWRs and their operating parameters were analyzed. As the result of multi-variable linear regressions, a model predicting the volume of DAW using the number of shutdowns (
) and total personnel exposure (
) was derived. Considering one standard error bound, the model could successfully simulate about 8575 of the real data. In order to predict the amount of DAW to be generated from a KNGR another model was derived by taking into account the additional volume reduction by supercompaction system. In addition, the volume of WAW (Wet Active Waste) to be generated from KNGR (Korean Next Generation Reactor) was calculated by considering conceptual design data and replacement effect of radwaste evaporator with selective ion exchangers. Finally, total volume of SRW (Solid Radioactive Waste) to be generated from KNGR was predicted by inserting design goal values of
into the model. The result showed that the expected amount of SRW to be generated from KNGR would be in the range of 33~44㎥.
. It was proved that the value would meet the operational target of KNGR proposed by KEPCO, that is, 50㎥.
Realistic Large Break Loss of Coolant Accident Mass and Energy Release and Containment Pressure and Temperature Analyses
Kwon, Young-Min ; Song, Jin-Ho ;
Nuclear Engineering and Technology, volume 29, issue 3, 1997, Pages 229~239
To investigate the realistic behavior of mass and energy release and resultant containment response during large break Loss of Coolant accident (LOCA), analyses are performed for Yonggwang (YGN) 3&4 nuclear power plants by using a merged version of RELAP5/CONTEMPT4 computer code. Comparative analyses by using conservative design computer codes are also peformed. The break types analyzed are the double-ended guillotine breaks at the cold leg and hot leg. The design analysis resulted in containment peak pressure during post-blowdown phase for the cold leg break. However, the RELAP5/CONTEMPT4 analyses show that the containment pressure has a peak during blowdown phase, thereafter it decreases monotonously without the second port-blowdown peak. For the hot leg break, revised design analysis shows much lower pressure than that reported in YGN 3&4 final safety analysis report. The RELAP5/CONTEMPT4 analysis shoos similar trend and confirmed that the bypass flow through the broken loop steam generator during post-blowdown is negligibly small compared to that of cold leg break. The low pressure and temperature predicted tv realistic analysis presented in this paper suggest that the design analysis methodology contains substantial margin and it can be improved to provide benefit in investment protection, such as, relaxing plant technical specifications and reducing containment design pressure.
A Study on the Development of Advanced Model to Predict the Sodium Pool Fire
Lee, Yong-Bum ; Park, Seok-Ki ;
Nuclear Engineering and Technology, volume 29, issue 3, 1997, Pages 240~250
Liquid sodium is widely used as a coolant of LMR(Liquid Metal Reactor) because of its physical and nuclear properties. However, the liquid sodium is very chemically reactive with oxygen and water so that the study on the sodium fire plays an important role in the LMR safety analysis. In this study, a sodium fire model is suggested to analyze the sodium pool fire where both the flame and the reaction products are considered. And also, sodium pool fire analysis computer code, SOPA, is developed. The sensitivity study on the experimental parameters such as the thermal radiation from flame to atmospheric gas, the vessel cooling and the duration of sodium spill was performed. The results showed good agreements with experimental data in the literature.
Helium-Air Exchange Flours Through Partitioned Opening and Two-Opening
Kang, Tae-il ;
Nuclear Engineering and Technology, volume 29, issue 3, 1997, Pages 251~259
This paper describes experimental investigations of helium-air exchange flows through partitioned opening and too-opening. Such exchange flows may occur following rupture accident of stand pipe in high temperature engineering test reactor. A test vessel with the too types of small opening on top of test cylinder is used for experiments. An estimation method of mass increment is developed to measure the exchange now rate. Upward flow of the helium and downward flow of the air in partitioned opening system interact out of entrance and exit of the opening. Therefore, an experiment with too-opening system is made to investigate effect of the fluids interaction of partitioned opening system. As a result of comparison of the exchange flow rates between too types of the opening system, it is demonstrated that the exchange flow rate of the two-opening system is larger than that of the partitioned opening system because of absence of the effect of fluids interaction.
Evaluation of Concrete Degradation Under Disposal Environment
Keum, D.K. ; Cho, W.J. ; Hahn, P.S. ;
Nuclear Engineering and Technology, volume 29, issue 3, 1997, Pages 260~268
The effects of three mechanisms, calcium depletion, sulphate and carbonate penetration, on the concrete degradation have been studied. The shrinking core model (SCM) and the HYDROGEOC. HEM (HGC) model have been applied to evaluate how fast the mechanisms proceed. The SCM is an analytical approximation model and the HGC is a numerical mass transport model coupled with chemical reaction. The SCM leads to more conservative results than the HGC, and turns out to be very useful in the viewpoint of simplicity and conservatism. During 300 years, calcium has been depleted within 10 cm from the concrete outer surface, and sulphate has penetrated less than 13.5 cm into the concrete. Carbonate has not penetrated own 7 cm into the concrete in contact with the bentonite, and, furthermore, its penetration into the concrete with the groundwater is negligible. Conclusively, the concrete is expected to maintain its integrity for at least 300 years that are regarded as institutional control period of intermediate and low-level radioactive waste repository.