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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 33, Issue 6 - Dec 2001
Volume 33, Issue 5 - Oct 2001
Volume 33, Issue 4 - Aug 2001
Volume 33, Issue 3 - Jun 2001
Volume 33, Issue 2 - Apr 2001
Volume 33, Issue 1 - Feb 2001
Selecting the target year
The Estimation of Early Health Effects for Different Combinations of Release Parameters and Meteorological Data
Jeong, Jongtae ; Jung, Wondea ;
Nuclear Engineering and Technology, volume 33, issue 6, 2001, Pages 557~565
Variations in the number of early health effects resulting from the severe accidents of the YGN 3&4 nuclear power plants were examined for different combinations of release parameters and meteorological data . The release parameters and meteorological data were selected in combination to define a limited number of basic spectra characterized by release height, heat content, release time, warning time, wind speed, rainfall rate, and atmospheric stability class. Variant seasonal spectra were also defined in order to estimate the potential significance of seasonal variations as a factor determining the incidence or number of early health effects. The results show that there are large differences in consequences from spectrum to spectrum, although an equal amount and mix of radioactive material is released to the atmosphere in each case. Also, there are large differences in the estimated number of health effects from season to season due to distinct seasonal variations in meteorological combinations in Korea. Therefore, it is necessary to consider seasonal characteristics in developing optimum emergency response strategies.
Evaluation of Creep-Fatigue Damage of KALIMER Reactor Internals Using the Elastic Analysis Method in RCC-MR
Koo, Gyeong-Hoi ; Bong Yoo ;
Nuclear Engineering and Technology, volume 33, issue 6, 2001, Pages 566~584
In this paper, the progressive deformation and the creep-fatigue damage for the conceptually designed reactor internals of KALIMER(Korea Advanced Liquid MEtal Reactor) are carried out by using the elastic analysis method in the RCC-MR code for normal operating conditions including the thermal load, seismic load (OBE) and dead weight. The maximum operating temperature of this reactor is 53
and the total service lifetime is 30 years. Thus, the time- dependent creep and stress-rupture effects become quite important in the structural design. The effects of the thermal induced membrane stress on the creep-fatigue damage are investigated with the risk of the elastic follow-up. To calculate the thermal stress, detailed thermal analyses considering conduction, convection and radiation heat transfer mechanisms are carried out with the ANSYS program. Using the results of the elastic analysis, the progressive deformation and creep-fatigue damages are calculated step by step using the RCC-MR in detail. This paper ill be a very useful guide for an actual application of the high temperature structural design of the nuclear power plant accounting for the time-dependent creep and stress-rupture effects.
The Effect of Non-condensable Gas on Direct Contact Condensation of Steam/Air Mixture
Lee, Hanchoon ; Kim, Moohwan ; Park, Suki ;
Nuclear Engineering and Technology, volume 33, issue 6, 2001, Pages 585~595
A series of experiments have been carried out to investigate the effects of non-condensable gas on the direct contact film condensation of vapor mixture under an adiabatic wall condition. The average heat transfer coefficient of the direct contact condensation was obtained at the atmospheric pressure with four main parameters ; air-mass fraction, mixture velocity, film Reynolds number, and the degree of water film subcooling having an influence on the condensation heat transfer coefficient. With the analysis of 88 experiments, a correlation of the average Nusselt number for direct contact film condensation of steam/air mixture at an adiabatic vertical wall was proposed as functions of film Reynolds number, mixture Reynolds number, air mass fraction, and Jacob number. The average heat transfer coefficient for steam/air mixture condensation decreased significantly while air mass fraction increased. The average heat transfer coefficients also decreased as the Jacob number increased, and were scarcely affected by the film Reynolds number below a mixture Reynolds number of about 245,000.
A Quantitative Study on Important Factors of the PSA of Safety-Critical Digital Systems
Kang, Hyun-Gook ; Taeyong Sung ;
Nuclear Engineering and Technology, volume 33, issue 6, 2001, Pages 596~604
This paper quantitatively presents the effects of important factors of the probabilistic safety assessment (PSA) of safety-critical digital systems. The result which is quantified using fault tree analysis methodology shows that these factors remarkably affect the system safety. In this paper we list the factors which should be represented by the model for PSA. Based on the PSA experience, we select three important factors which are expected to dominate the system unavailability. They are the avoidance of common cause failure, the coverage of fault tolerant mechanisms and software failure probability. We Quantitatively demonstrate the effect of these three factors. The broader usage of digital equipment in nuclear power plants gives rise to the safety problems. Even though conventional PSA methods are immature for applying to microprocessor-based digital systems, practical needs force us to apply it because the result of PSA plays an important role in proving the safety of a designed system. We expect the analysis result to provide valuable feedback to the designers of digital safety- critical systems.
Three-Dimensional Modelling and Sensitivity Analysis for the Stability Assessment of Deep Underground Repository
Kwon, S. ; Park, J.H. ; Park, J.W. ; Kang, C.H. ;
Nuclear Engineering and Technology, volume 33, issue 6, 2001, Pages 605~618
For the mechanical stability assessment of a deep underground high-level waste repository. computer simulations using FLAC3D were carried out and important parameters including stress ratio, depth, tunnel size, joint spacing, and joint properties were chosen from sensitivity analysis. The main effect as well as the interaction effect between the important parameters could be investigated effectively using fractional factorial design . In order to analyze the stability of the disposal tunnel and deposition hole in a discontinuous rock mass, different modelings were performed under different conditions using 3DEC and the influence of joint distribution and properties, rock properties and stress ratio could be determined. From the three dimensional modelings, it was concluded that the conceptual repository design was mechanically stable even in a discontinuous rock mass.
FIV Analysis for a Rod Supported by Springs at Both Ends
H. S. Kang ; K. N. Song ; Kim, H. K. ; K. H. Yoon ;
Nuclear Engineering and Technology, volume 33, issue 6, 2001, Pages 619~625
An axial-flow-induced vibration model was proposed for a rod supported by two translational springs at both ends. For developing the model, a one-mode approximation was made based on the assumption that the first mode was dominant in vibration behavior of the single span rod. The first natural frequency and mode shape functions for the flow-induced vibration, called the FIV model were derived by using Lagrange＇s method. The vibration displacements at reactor conditions were calculated by the proposed model for the spring-supported rod and by the previous model for the simple-supported(55) rod. As a result, the vibration displacement for the spring-supported rod was larger than that of the 55 rod, and the discrepancy between both displacements became much larger as flow velocity increased. The vibration displacement for the spring-supported rod appeared to decrease with the increase of the spring constant. AS flow velocity increased, the increase rate of vibration displacement was calculated to go linearly up, and that of the rod having the short span length was larger than that of the rod having the long span length although the displacement value itself of the long span rod was larger than that of the short one.
Energy Response in Chemiluminescence Dosimetry with Sugar and Sorbite
Jun, J.S. ; Guggenberger, R. ; Dalheimer, A. ;
Nuclear Engineering and Technology, volume 33, issue 6, 2001, Pages 626~637
A series of study on energy dependence in chemiluminescence dosimetry with sugar and sorbite produced in two different countries was carried out administering a dose of 5 Gy to the samples at six different mean photon energies of 30, 50, 80, 130, 662 and 1250 keV. The results revealed distinct energy dependence of chemiluminescence(CL) output of sugar and sorbite. Although the energy dependence, in general, could be fitted by a polynomial of (os E, with I being radiation energy, up to cubic term, we reached a conclusion that the adoption of a fitting function,
+d, deduced from theoretical energy response curve calculated as the ratio of the mass energy absorption of the samples of interest to the soft tissue is more reasonable and rational. Herecoefficients
is CL intensity, and a, b, c and d are constants to be determined in the fitting process. Energy dependence of relative sensitivities of one sample to the other, discrepancy in sensitivities of the samples from the two countries, and prominent grain size effect in Sorbitol were also shown.shown.
Post Test Analysis to Natural Circulation Experiment on the BETHSY Facility Using the MARS 1.4 Code
Chung, Young-Jong ; Kim, Hee-Cheol ; Chang, Moon-Hee ;
Nuclear Engineering and Technology, volume 33, issue 6, 2001, Pages 638~651
The present study is to assess the applicability of the best-estimate thermal-hydraulic code, MARS 1.4, for the analysis of thermal-hydraulic behavior in PWRs during natural circulation conditions. The code simulates a natural circulation test, BETHSY test 4. la, which was conducted on the integral test facility of BETHSY. The test represented the cooling states of the primary cooling system under single-phase natural circulation, two-phase natural circulation and the reflux condensation mode with conditions corresponding to the residual power, 2% of the rated core power value and 6.8 MPa at the secondary system. Based on MARS 1.4 calculations, the major thermal-hydraulic behaviors during natural circulation are evaluated and the differences between the experimental data and calculated results are identified. The calculated results show generally good behavior with regard to the experimental results; the region of single-phase natural circulation is 100-92% of the initial mass inventory, two-phase natural circulation is 84-63 %, and the reflux condensation mode occurred below 58 %. U-tubes empty and the core uncovery are obtained at 39 % and 34 % of the initial mass inventory, respectively.