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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 32, Issue 6 - Dec 2000
Volume 32, Issue 5 - Oct 2000
Volume 32, Issue 4 - Aug 2000
Volume 32, Issue 3 - Jun 2000
Volume 32, Issue 2 - Apr 2000
Volume 32, Issue 1 - Feb 2000
Selecting the target year
Probabilistic Structural Integrity Assessment of a Reactor Vessel Under Pressurized Thermal Shock
Kim, Ji-Ho ; Kim, Yong-Wan ; Kim, Tae-Wan ; Hyung-Huh ; Kim, Jong-In ;
Nuclear Engineering and Technology, volume 32, issue 2, 2000, Pages 99~107
A probabilistic integrity analysis method is presented for a reactor vessel under pressurized thermal shock(PTS) based on Monte Carlo simulation. This method can be applied to the structural integrity assessment of a reactor vessel subjected to pressurized thermal shock where the coolant temperature transient cannot be expressed explicitly as a time function. An axially or circumferentially oriented infinite length surface crack is assumed to be in the beltline weld region of the rector vessel's inside surface. The random variables are the initial crack depth, neutron fluence on the vessel's inside surface, the copper and nickel content of the vessel materials, R
, and K/aub la/. The reliability of a sample reactor vessel under PTS is assessed quantitatively and the influence of the amount of neutron fluence is also examined by applying the present method.sent method.
An Experimental Study on the Mass and Energy Release for a Hot Leg Break LBLOCA During Post Blowdown
S.J. Hong ; Kim, J.H. ; Park, G.C. ;
Nuclear Engineering and Technology, volume 32, issue 2, 2000, Pages 108~127
Hot leg break LBLOCA(Large Break LOCA) had a potential to be a containment maximum pressure accident in YGN3&4, which was induced from excessive conservatism in the CE analysis methodology of mass and energy release. This study conducted mass and energy release experiment for the hot leg break LBLOCA during post blowdown with an integral test facility, SNUF(Seoul National University Facility). This facility simulated YGN 3&4 with volume ratio of 1/1140 based on Ishii's three level scaling. Experiment showed that SI(Safety Injection) water refilled cold leg first and core later. SI water was vaporized in the core, which resulted in the repressurization of reactor. This increase of pressure drove the water in cold leg to flow up half height of U tubes. However, since the water was drained back soon, the release through the SG side broken section by evaporation was negligibly small. This study also provided experimental assessment of RELAP5 results by KAERI for the release through the SG side broken section.
Critical Heat Flux in Uniformly Heated Vertical Annulus Under a Wide Range of Pressures 0.57 to 15.0 MPa
Chun, Se-Young ; Chung, Heung-June ; Hong, Sung-Deok ; Yang, Sun-Kyu ; Chung, Moon-Ki ;
Nuclear Engineering and Technology, volume 32, issue 2, 2000, Pages 128~141
The critical heat flux (CHF）experiments have been carried out in a wide range of pressures for an internally heated vertical annulus. The experimental conditions covered ranges of pressures from 0.57 to 15.01 MPa, mass fluxes of 0 kg/
s and from 200 to 650 kg/
s, and inlet subcoolings from 85 to 413 kJ/kg. The characteristics of the present data and the effect of pressure on CHF are discussed. Most of the CHFs were identified to dryout of the liquid film in the annular or annular-mist flow. For the mass flux of 200 kg/
s, there were the indications that the CHF occurred at the transition from annular to annular-mist How in the pressure range of 3~10 MPa. For the mass fluxes of 550 and 650 kg/
s, the CHFs had a maximum value at a pressure of 2~3 MPa, and the pressure at the maximum CHF values had a trend moving toward the pressure at the peak value of pool boiling CHF as the mass flux decreased. The CHF data under a zero mass flux condition indicate that both the effects of pressure and inlet subcooling on the CHF were smaller, compared with those on the CHF with net water upward flow.
Interfacial Condensation Heat Transfer for Countercurrent Steam-Water Stratified Flow in a Circular Pipe
Chu, In-Cheol ; Chung, Moon-Ki ; Yu, Seon-Oh ; Chun, Moon-Hyun ;
Nuclear Engineering and Technology, volume 32, issue 2, 2000, Pages 142~156
An experimental study of steam condensation on a subcooled thick water layer (0.018 ~0.032 m) in a countercurrent stratified flow has been performed using a nearly horizontal circular pipe. A total of 103 average interfacial condensation heat transfer coefficients were obtained and parametric effects of steam and water flow rates and the degree of subcooling on condensation heat transfer were examined. The measured local temperature and velocity distributions in the thick water layer revealed that there was a thermal stratification due to the lack of full turbulent thermal mixing in the lower region of the water layer Two empirical Nusselt number correlations, one in terms of average steam and water Reynolds numbers, and the water Prandtl number, and the other in terms of the Jakob number in place of the Prandtl number, which agree with most of the data within
25%, were developed based on the bulk flow properties. Comparisons of the present data with existing correlations showed that the present data were significantly lower than the values predicted by existing correlations.
The Level Control System Design of the Nuclear Steam Generator for Robustness and Performance
Lee, Yoon-Joon ; Lee, Heon-Ju ; Kim, Kyung-Yeon ;
Nuclear Engineering and Technology, volume 32, issue 2, 2000, Pages 157~168
The nuclear steam generator level control system is designed by robust control methods. The feedwater controller is designed by three methods of the H
, the mixed weight sensitivity and the structured singular value. Then the controller located on the feedback loop of the level control system is designed. For the system performance, the controller of simple PID whose coefficients vary with the power is selected. The simulations show that the system has a good performance with proper stability margins.
Numerical Analysis of Evolution of Thermal Stratification in a Curved Piping System
Park, Seok-Ki ; Nam, Ho-Yun ; Jo, Jong-Chull ;
Nuclear Engineering and Technology, volume 32, issue 2, 2000, Pages 169~179
A detailed numerical analysis of the evolution of thermal stratification in a curved piping system in a nuclear power plant is performed. A finite volume based thermal-hydraulic computer code has been developed employing a body-fitted, non-orthogonal curvilinear coordinate for this purpose. The cell-centered, non-staggered grid arrangement is adopted and the resulting checkerboard pressure oscillation is prevented by the application of momentum interpolation method. The SIMPLE algorithm is employed for the pressure and velocity coupling, and the convection terms are approximated by a higher-order bounded scheme. The thermal-hydraulic computer code developed in the present study has been applied to the analysis of thermal stratification in a curved duct and some of the predicted results are compared with the available experimental data. It is shown that the predicted results agree fairly well with the experimental measurements and the transient formation of thermal stratification in a curved duct is also well predicted.
Nonlinear Flexural Analysis of PSC Test Beams in CANDU Nuclear Power Plants
Bae, In-Hwan ; Choi, In-Kil ; Seo, Jeong-Moon ;
Nuclear Engineering and Technology, volume 32, issue 2, 2000, Pages 180~190
In this study, nonlinear analyses of prestressed concrete(PSC) test beams for inservice inspection of prestressed concrete containments for CANDU nuclear power plants are presented. In the analysis the material nonlinearities of concrete, rebar and prestressing steel are used. To reduce the numerical instability with respect to the used finite element mesh size, the tension stiffening effect has been considered. For concrete, the tensile stress-strain relationship derived from tests is modified and the stress-strain curve of rebar is assumed as a simple bilinear model. The stress-strain curve of prestressing steel is applied as a multilineal curve with the first straight line up to 0.8fpu. To prove the validity of the applied material models, the behavior and strength of the PSC test specimens tested to failure have been evaluated. A reasonable agreement between the experimental results and the predictions is obtained. Parametric studies on the tension stiffening effects, the impact of prestressing losses with time, and the compressive strength of concrete have been conducted.
Benefits of the S/F Cask Impact Limiter Weldment Imperfection
Ku, Jeong-Hoe ; Lee, Ju-Chan ; Kim, Jong-Hun ; Park, Seong-Won ; Park, Hyun-Soo ;
Nuclear Engineering and Technology, volume 32, issue 2, 2000, Pages 191~203
This paper describes the beneficial effect of weldment imperfection of the cask impact limiter, by applying intermittent-weld, for impact energy absorbing behavior. From the point of view of energy absorbing efficiency of an energy absorber, it is desirable to reduce the crush load resistance and increase the deformation of the energy absorber within certain limit. This paper presents the test results of intermittent-weldment and the analysis results of cask impacts and the discussions of the improvement of impact mitigating effect by the imperfect-weldment. The rupture of imperfect weldment of an impact limiter improves the energy-absorbing efficiency by reducing the crush load amplitude without loss of total energy absorption. The beneficial effect of weldment imperfection should be considered to the cask impact limiter design.