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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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Development of a One-Group Cross Section Data Base of the ORIGEN2 Computer Code for Research Reactor Applications
Kim, Jung-Do ; Gil, Choong-Sub ; Lee, Jong-Tai ; Hwang, Won-Guk ;
Nuclear Engineering and Technology, volume 24, issue 1, 1992, Pages 1~13
A one-group cross section data base of the ORIGEN2 computer code was developed for research reactor applications. For this, ENDF/B-IV and -V data were processed using the NJOY code system into 69-group data. The burnup-dependent weighting spectra for KMRR were calculated with the WIMS-KAERI computer code, and then the 69-group data were collapsed to one-group using the spectra. The ORIGEN2-predicted burnup-dependent acti-nide compositions of KMRR spent fuel using the newly developed data base show a good agreement with the results of detailed multigroup transport calculation. In addition, the burnup characteristics of KMRR spent fuel was analyzed with the new data base.
Development of a Computer Code for Common Cause Failure Analysis
Park, Byung-Hyun ; Cho, Nam-Zin ;
Nuclear Engineering and Technology, volume 24, issue 1, 1992, Pages 14~29
COMCAF, a computer code for the common-cause failure analysis, is developed to treat the common-cause failures in nuclear power plants. In the treatment of common-cause failures, the minimal cut sets of the system are obtained first without changing the fault-tree structure. The occurrence probabilities of the minimal cut sets are then calculated accounting for the common-cause failures among components in the same minimal cut set or in different minimal cut sets. The basic parameter model is used to model the common-cause failures between similar or identical components. For dissimilar components, the assumption of symmetry used in the basic parameter model is applied to the basic events affecting two or more components. The top event probability is evaluated using the inclusion-exclusion method. In addition to the common-cause failures of components in the same minimal cut sets, failures of components in the different minimal cut sets are also easily accounted for by this method. This study applied this common-cause failure analysis to the PWR auxiliary feedwater system. The results in the top event probability for the system are compared with those of no common-cause failures.
Development of Integrated Boration and Dilution Model for Boron Concentration Behavior Analysis
Chi, Sung-Goo ; Park, Han-Kwon ; Kuh, Jung-Eui ;
Nuclear Engineering and Technology, volume 24, issue 1, 1992, Pages 30~39
In this study, an integrated boration and dilution (INBAD) model is proposed to predict the required makeup flowrate for RCS boron concentration change and to analyze the boron concentration behavior at each subsystem within the RCS including CVCS during boration and dilution operation. The INBAD model is constructed by integrating an existing neutronic code and a boration and dilution model. The boration and dilution model has been developed for our specific purpose using the one-cell model and multi-cell model. In addition, in order to assess the boron concentration behavior more realistically, two important features such as variable pressurizer heater output and optional makeup mode (either direct or indirect injection) are implemented in this model. In order to demonstrate the usefulness of this model, the boron concentration behavior analysis at each subsystem were performed for both direct and indirect injection mode using YGN 3 and 4 design data. Also, the effect of pressurizer heater output on the primary loop boron concentration was investigated. The results showed that the boron concentration changes can be predicted accurately at each subsystem during boration and dilution operation.
A Comprehensive Swelling Model of Silicide Dispersion Fuel for Research Reactor
Woan Hwang ; Suk, Ho-Chun ; Jae, Won-Mok ;
Nuclear Engineering and Technology, volume 24, issue 1, 1992, Pages 40~51
One of the important irradiation performance characteristics of the silicide dispersion fuel element in research reactors is the diameteral increase resulting from fuel swelling. This paper, will attempt to develop a physical model for the fuel swelling, DFSWELL, by analyzing the basic irradiation behaviours and some experimental evidences. From the experimental evidences, it was shown that the volume changes in irradiated U
Si-Al were strongly dependent on temperature and fission rate. The quantitative-amount of swelling for silicide fuel is estimated by considering temperature, fission rate, solid fission product build-up and gas bubble behavior. The swelling for the silicide fuel is comprised of three major components : i ) a volume change due to the formation of an interfacial layer between the fuel particle and matrix. ii ) a volume change due to the accumulation of gas bubble nucleation iii ) a volume change due to the accumulation of solid fission products The DFSWELL model which takes into account the above three major physical components predicts well the absolute magnitude of silicide fuel swelling in accordance with the power histories in comparison with the experimental data.
Evaluation of Primary Coolant pH Operation Methods for the Domestic PWRs
Paek, Seung-Woo ; Na, Jung-Won ; Kim, Yong-Eak ; Bae, Jae-Heum ;
Nuclear Engineering and Technology, volume 24, issue 1, 1992, Pages 52~62
Radioactive nuclides deposited on out-of-core surface after the radiation in the core by the transport of corrosion products (CRUD) through the primary coolant system in PWR which is the major plant type in Korea, are leading sources of radiation exposure to plant maintenance personnel. Thus, the optimal chemistry operation method is required for the reduction of radiation exposure by the corrosion products. This study analysed the actual water chemistry operation data of four operating domestic PWRs. And in order to evaluate the coolant chemistry operation data, a computer code which can calculate the activity buildup in the various chemistry conditions of PWR coolant was employed. Through the analysis of comparison between the activity buildup of actual water chemistry operation mode and that of assumed Elevated Li operation mode calculated by the computer code, it was found that the out-of core radioactivity can be reduced by diminishing the deposition of corrosion products on the core in case that the Elevated Li operation mode is applied to the coolant chemistry operation of PWR. And the higher coolant pH operation was shown to have the advantage of the reduction of out-of-core activity buildup if the integrity of system structural materials and fuel cladding is guaranteed.
Resonant Formation Rates of Muonic Molecular ion in Muon-Catalyzed Fusion
Im, Ki-Hak ; Hong, Sang-Hee ;
Nuclear Engineering and Technology, volume 24, issue 1, 1992, Pages 63~74
The resonant formation rates of muonic molecular ion dtr in the muon-catalyzed fusion are calculated in various fuel temperatures and densities. The elastic scattering cross sections between t
and deuterons are obtained by making use of the partial wave method. The transition property of the excited compound molecule [(dt
)dee]＊ derived by the impulse approximation in the form of a bound-state form factor. The radiative, Auger, and collisional deexcitations are considered as the deexcitation mechanisms of the excited dt
, and each deexcitation width is calculated as well as back decay width. The resultant reaction widths are used to calculate the formation cross sections of resonant dt
. The resonant formation rates for dt
-d and dt
-t collisions are computed as functions of fuel temperature and density. The calculations show that the resonant formation rates increase with fuel densities and have the maximum values at the particular temperatures where the relative collision energies are equal to the resonant ones.
Texture Transformations and Its Role on the Yield Strength of (
) Heat Treated Zircaloy-4
Yoo, Jong-Sung ; Kim, In-Sup ;
Nuclear Engineering and Technology, volume 24, issue 1, 1992, Pages 75~85
The texture changes and their effect on the 0.2% yield strength of Zircaloy-4 sheet were examined after quenched from the (
) phase temperature. When the prior (
) gram size was slightly larger than that of the
-annealed, the observed texture was similar to the
-annealed texture having an ideal orientation of the (0001) basal pole at 30
away from the normal direction toward the transverse direction. When the prior (
) grain size was twice as large as that of the
-annealed, the location of maximum basal pole intensity was distributed between the transverse and the rolling direction making an angle 15
from the normal direction, and the observed texture became isotropic. It was found that the Kearns texture parameter, fr, in the rolling direction increased steadily, and fr in the transverse direction increased slightly, while fr in the the normal direction decreased with increasing heat treatment time. With a small increase in fr, the 0.2% yield strength increased drastically. The influence of texture was analyzed by deriving the Schmid orientation factors and the resolved shear stresses for the deformation systems. It was found that the large increase in the 0.2% yield strength was attributed mainly to the microstructural changes and partly to the texture changes by the (
) heat treatment.
Evaluation of Direct Vessel Injection Design With Pressurized Thermal Shock Analysis
Cha, Jong-Hee ; Jun, Hyung-Gil ;
Nuclear Engineering and Technology, volume 24, issue 1, 1992, Pages 86~97
The purpose of this paper is to evaluate the direct vessel injection design from a pressurized thermal shock(PTS) viewpoint for the Combustion Engineering System 80+ A break of the main steam line from zero power and a 0.05 ft
small break loss-of-coolant accident (LOCA) from full power were selected as the potential PTS events. In order to investigate the stratification effects in the reactor downcomer region, the fluid mixing analysis was performed using the COMMIX-IB code for steam line break and using the REMIX code for 0.05 ft
small break LOCA. The stress distributions within the reactor vessel walls experiencing the pressure and the temperature transients were calculated using the OCA-P code for both events. The results of the analysis showed that a small break LOCA without decay heat presented the greatest challenge to the vessel, however, there is no crack initiation through end-of-life of the vessel with consideration of decay heat.
A Study on the Fatigue Failure Behavior of Cheon-Ho Mt. Limestone Under Cyclic Loading
Lee, Jong-Uk ; Rhee, Chan-Goo ; Kim, Il-Jung ; Kim, Yeong-Seok ;
Nuclear Engineering and Technology, volume 24, issue 1, 1992, Pages 98~109
In this study uniaxial cyclic loading tests were performed on Cheon-Ho Mt. Limestone specimens to investigate the fatigue failure behavior. The loading rate was kept constantly at 760kg/
/sec under cyclic loading. In order to reveal the fatigue behavior for each rock type, the test results were mutually compared with previous studies carried out on Indiana Limes-tone and Seong-Ju Sandstone. Fatigue data is presented in the form of S-N curves, which illustrate the relationship of maximum applied stress(S) to the number of cycles(N) required to produce failure. For the purpose of comparing the S-N curves for each rock type, the test data were formulated up to 10
cycles and the correlation coefficients(R) on Cheon-Ho Mt. Limestone and Seong-Ju Sandstone specimen are 0.886 and 0.983, respectively. All three rock specimens were found to have shorter fatigue life at higher applied stress levels. The fatigue life for each rock type was considered as no less than 81.5, 70 and 74.8%, for Cheon-Ho Mt. Limestone, Indiana Limestone and Seong-Ju Sandstone, respectively. The comparison in static strength for monotonic loaded specimens and specimens which did not fail even after 10
cycles indicated that the increasing rate of strength was about 6.18 and 10.96% , for Cheon-Ho Mt. Limestone and Indiana Limestone, respectively. Poisson's ratio and volumetric strain for Cheon-Ho Mt. Limestone and Seong-ju Sandstone, tended in all the cases to rapidly increase at higher stress levels and with an increase in number of cycles. This increasing trend becomes rapid and obvious just before failure. Also Poisson's ratio and volumetric strain for each stress level were compared and analyzed at the first cycle and the cycle prior to failure.