• Journal of the Korean Society of Safety
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• v.8 no.4
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• pp.27-40
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• 1993

High temperature tensile tests, steady state creep tests, Internal stress tests and creep rupture tests using A17075 alloy( $T_{6}$ ) were performed over the temperature range of 9$0^{\circ}C$~50$0^{\circ}C$ (0.4 $T_{m}$ ~0.85 $T_{m}$ ) and stress range of 0.64~17.2(kgf/$\textrm{mm}^2$). The main results obtained in this paper were as follows. (1) The activation energies for yielding at the temperature of 0.4 $T_{m}$ ~0.75 $T_{m}$ were calculated to be 25.7~36.5kcal/mol, which were nearly equal to the activation energies for creep. (2) At around the temperature of 9$0^{\circ}C$~12$0^{\circ}C$ and under the stress level of 10~17.2(kgf/$\textrm{mm}^2$), and at around the temperature of 200~41$0^{\circ}C$ and under the stress level of 1.53~9.55(kgf/$\textrm{mm}^2$) and again at around the temperature of 470~50$0^{\circ}C$ and under the stress level of 0.62~l.02(kgf/$\textrm{mm}^2$), the applied stress dependence of steady state creep rate $n_{measu}$ measured were, respectively, 3.15, 6.62 and 1.1, which were in good agreement the calculated stress dependence $n_{ealeu}$ obtained by the difference of the applied stress dependence of the Internal stress and the ratio of the internal stress to the applied stress. (3) At the temperature range of 0.4~0.43 $T_{m}$ , and at the temperature range of 0.52~0.75 $T_{m}$ and again at the temperature range of 0.82~0.85 $T_{m}$ , the activation energies $Q_{measu}$ obtained by steady state creep rate, respective, 26. 16, 34.9, 36.2 and 36.1kcal/mol, which were in good agreement with those obtained with the activation energies under constant effective stress and the temperature dependence of Internal stress. (4) At the temperature range of the 0.52~0.73 $T_{m}$ and under the stress level of 1.53~9.55(kgf/$\textrm{mm}^2$), the stress dependence of rupture life(n’) measured was 6.3~6.6, which was in good agreement with the stress dependence of steady state creep rate(n). And at the same condition the activation energy for rupture( $Q_{f}$ ) measured was 32.0~36.9kca1/mol, which was also in good agreement with the activation energy obtained by steady state creep rate ( $Q_{c}$ ). (5) The rupture life( $t_{f}$ ) might be represented by athermal process attributed to the difference of the applied stress dependence of the internal stress and the ratio of the internal stress to the applied stress, and the thermal activated process attributied to the temperature dependence of the internal stress as $t_{f}$ = A'$\sigma$$_{a}$ ｛n(1-d $\sigma$$_{i}$ /d $\sigma$$_{a}$ )/(1-$\sigma$$_{i}$ / $\sigma$$_{a}$ )｝.exp[｛ $Q_{c}$ $^{*}$-( $n_{o}$ R $T^2$/ $E_{(T)}$) (d $E_{(T)}$/dT) - ( $n_{0}$ R $T^2$/ $\sigma$$_{a}$ - $\sigma$$_{i}$ ) (d $\sigma$$_{i}$ /dT)｝/RT]. (6) The relationship betwween Larson-Miller rupture parameter and logarithmic stress was linearly decreased, so creep rupture life of Al 7075 alloy seemed to be predicted exactly with Larson-Miller parameter.meter.

• Journal of the Korean Society of Manufacturing Technology Engineers
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• v.7 no.6
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• pp.117-125
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• 1998

The paper attempts to estimate the incubation time of a cavity in the interface between a power law creep particle and an elastic matrix subjected to a uniaxial stress. Since the power law creep particle is time dependent, the stresses in the interface relax. The volume free energy associated with Helmholtz free energy includes strain energies caused by applied stress and dislocations piled up in interface(DPI). The energy due to DPI is found by modifying the result of Dundurs and Mura[4]. The volume free energies caused by both applied stress and DPI are a function of the cavity size(r) and elapsed time(t) and arise from stress relaxation in the interface. Critical radius $r^*$ and incubation time $t^*$ to maximise Helmholtz free energy is found in present analysis. Also, kinetics of cavity formation are investigated using the results obtained by Riede [7]. The incubation time is defined in the analysis as the time required to satisfy both the thermodynamic and kinetic conditions. Through the analysis it is found that 1) strain energy caused by the applied stress does not contribute significantly to the thermodynamic and kinetic conditions of a cavity formation, 2) in order to satisfy both thermodynamic and kinetic conditions, critical radius $r^*$ decreases or holds constant with increase of the time until the kinetic condition(eq. 2.3) is satisfied. there for the cavity may not grow right after it is formed, as postulated by Harris [15], and Ishida and Mclean [16], 3) the effects of strain rate exponent (m), material constant $\sigma$0, volume fraction of the particle to matrix(f)and particle size on the incubation time are estimated using material constants of the copper as matrix.

• Nuclear Engineering and Technology
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• v.37 no.5
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• pp.479-490
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• 2005

The thermalhydraulic performance of a CANDU-6 reactor loaded with various CANFLEX fuel bundles is evaluated by the NUCIRC code, which is incorporated with recent models of pressure drop and critical heat flux (CHF) predictions based on high-pressure steam-water tests for the CANFLEX bundle as well as a 37-element bundle. The distributions of channel flow rate, channel exit quality, critical channel power (CCP), and critical power ratio (CPR) for the CANFLEX bundles (with natural or recycled uranium fuel) in the CANDU-6 reactor fuel channel are calculated by the code. The effects of axial and radial heat flux on CCP are evaluated by assuming that the recycled uranium fuel (CANFLEX-RU) has the same geometric data as the natural uranium fuel bundle (CANFLEX-NU), but a different power distribution due to different fuel composition and refueling scheme. In addition, the effects of pressure tube creep and bearing-pad height are examined by comparing various results of uncrept, and $3.3\%\;and\;5.1\%$ crept channels loaded with CANFLEX bundles with 1.4 mm or 1.7 mm high bearing-pads with those of the 37-element bundle. The distributions of the channel flow rate and CCP for the CANFLEX-NU or -RU bundle show a typical trend for a CANDU-6 reactor channel, and the CPRs are maintained above at least 1.444 (NU) or 1.455 (RU) in the uncrept channel. The enhanced CHF of the CANFLEX bundle (particularly with 1.7mm height bearing-pads) produces a higher thermal margin and considerably less sensitivity to CCP reduction due to the pressure tube creep than the 37-element bundle. The CCP enhancement due to the raised bearing-pads is estimated to be about $3\%\~5\%$ for the CANFLEX-NU and $2\%\~6\%$ for the CANFLEX-RU bundle, respectively.

• Journal of the Korean Geographical Society
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• v.45 no.3
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• pp.375-389
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• 2010

In order to investigate the rates and factors of slow mass movement on a subalpine slope of Mount Halla, two painted stone lines were monitored in a bare patch at 1,710 m a.s.l. The mean movement of surface gravels is 58.2 cm, equivalent to 0.24 cm/day. However, the rates of movement vary with seasons. Compared with 0.05 cm/day of a non-frozen season, a frozen season shows 0.3 cm/day. It implies that the movement of surface gravels could be largely controlled by periglacial processes during a frozen season. In particular, frost creep including needle ice creep plays a main role in the movement of gravels under the thermal and soil conditions which are favorable for needle ice development. Since line II is located at a steeper slope than line I, the movement of line II was always larger than that of line I. However, slope gradient is not the most dominant factor contributing the movement of gravels, which can be interrupted by downslope big gravels and vegetation patches. The size and specific weight of gravels also can influence the movement of gravels. Porous and light scoriae result in relatively quick movement of gravels on the subalpine slope of Mount Halla.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.20 no.11
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• pp.3516-3524
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• 1996

Service conditions for structures at elevated temperatures in nuclear power plant involve transient thermal and mechanical load levels that are severe enough to caeuse inelastic deformations due to creep and plasticity. Therefore, a systematic mehtod of inelastic analysis is needed for the design of structural components in nuclear poser plants subjected to such loading conditions. In the present investigation, the Chabodhe model, one of the unified viscoplastic constitutive equations, was selected for systematic inelastic analysis. The material response was integrated based on GMR ( generallized mid-point rule) time integral scheme and provided to ABAQUS as a material subroutine, UMAT program. By comparing results obtaned from uniaxial analysis using the developed UMAT program with those from Runge-Kutta solutions and experimentaiton, the validity of the adopted Chaboche model and the numerical stability and accuracy of the developed UMAT program were verified. In addition, the developed material subroutine was applied for uniaxial creep and tension analyses for the plate with a hole in the center. The application further demonstrates usefulness of the developed program.

• Journal of the Microelectronics and Packaging Society
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• v.30 no.1
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• pp.17-29
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• 2023

In this paper, a method for evaluating the reliability of solder joints is introduced, as they play a crucial role in packaging technology due to the miniaturization and high-performance requirements of electronic device. Firstly, properties of solder based on various alloy compositions and solder types are described, followed by an analysis of solder joint structures in different packages. Next, the influence of solder alloy composition and microstructure on the thermal and mechanical properties of solder is analyzed, and solder creep behavior is briefly introduced. Subsequently, analytical techniques considering creep models and fatigue models for reliability evaluation are presented, and various ways to improve the reliability of solder joints are discussed. This study is expected to provide valuable information for evaluating and enhancing the reliability of solder joints in the semiconductor packaging technology field.

• Proceedings of the KSME Conference
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• 2001.06d
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• pp.834-839
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• 2001

Calandria tube wrapping each pressure tube is one of the key structural components of CANDU reactor(Calandria) which is consisted of many pressure tubes containing nuclear fuel assemblies. As the Calandria tube(made of zirconium alloy) is sagging due to its thermal and irradiation creep during the plant operation, it possibly contacts with liquid injection nozzle crossing beneath the Calandria tube, which subsequently results in difficulties on the safe operation. It is therefore necessary to check the gap for the confirmation of no contacts between the two tubes, Calandria tube and liquid injection tube, with a proper measure during the life of plant. In this study, an ultrasonic measurement method was selected among several methods investigated. The ultrasonic device being developed for the measurement of the gap was introduced and its preliminary performance test results were presented here. The gap between LIN and CT at site was measured using by this ultrasonic device at site.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.9 no.1
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• pp.48-55
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• 2013

In this paper, W-DCAP-HTR(Web-based Design Compatibility Assessment Program for High Temperature Reactor) which will be used to check the design criteria for high temperature reactor is newly proposed. To do this, the assessment procedure of the ASME Sec.III Div.5 such as time-dependent primary stress limit, accumulated inelastic strain, and creep-fatigue damage evaluation were investigated. Furthermore, the trend of candidate materials for high temperature reactor was also reviewed. Then, all assessment procedures for high temperature reactor have been computerized to enhance the efficiency and to reduce the possibility of human error during calculating procedure by hand calculation. It can be directly conducted by adopting the actual thermal and structural analysis results. The validation of W-DCAP-HTR has been demonstrated by benchmark analysis.

• Nuclear Engineering and Technology
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• v.28 no.2
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• pp.150-161
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• 1996

A deformation model was developed to calculate the deformation of the uranium-silicide dispersion fuel (U$_3$Si-Al) elements for research reactors. The model was based on the elasto-plasticity theory and power-law creep theory. Also, isotopic swelling was assumed for the fuel meat and isotropic thermal expansion for the fuel meat and dadding. The new model calculated successfully the deformation of the fuels of HANARO and NRU (in Canada). As the most important result, it was shown that the primary deformation mechanism in the fuel meat was swelling and that in the cladding was creep. For all cases simulated, the maximum hoop stress at cladding outer surface was lass than 5MPa, probably well below the yield stress of the dadding, and finally, the volume change was predicted to be less than 10% in the whole burnup range.

• Journal of the Korean Society for Precision Engineering
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• v.16 no.8
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• pp.46-52
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• 1999

Large valves for steam turbines of fossil power plants are exposed to a severe mechanical and thermal loading resulting from steam with high pressure and high temperature. Valve casings are designed to withstand such a loading. During the operation of a plant, temperatures at inner and outer surface of the casings are measured and steam flow is controlled so that the measured difference is lower than the maximum allowable value determined in the design stage. In this paper, a method is presented to calculate the maximum allowable temperature difference at the inner and outer surface of valve casings for steam turbines of fossil power plants. The finite element method is used to analyze distribution of temperature and stresses of a casing under the operating condition. Low cycle fatigue and creep rupture are taken into consideration to determine the maximum allowable temperature difference. The method can be usefully applied in the design stage of the large valves for the steam turbines, contributing to safe and reliable operation of the fossil power plants.