• 대한기계학회논문집
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• 제19권5호
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• pp.1182-1189
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• 1995

Several methods have been developed to predict the creep rupture time of the steam pipes in thermal power plant. However, existing creep life prediction methods give very conservative value at operating stress of power plant and creep rupture strain cannot be well estimated. Therefore, in this study, creep rupture time and strain prediction method accounting for material damage and grain boundary sliding is newly proposed and compared with the existing experimental data. The creep damage evolves by continuous cavity nucleation and constrained cavity growth. The results showed good correlation between the theoretically predicted creep rupture time and the experimental data. And creep rupture strain may be well estimated by using the proposed method.

• 대한기계학회논문집A
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• 제24권11호
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• pp.2812-2822
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• 2000

Several methods have been developed to predict the rupture time of the boiler tubes in thermal power plant. However, existing life prediction methods give very conservative value at operating stress of power plant and rupture strain cannot be well estimated. Therefore, in this study, rupture time and strain prediction method accounting for creep, corrosion and heat transfer is newly proposed and compared with the current research results. The creep damage evolves by continuous cavity nucleation and constrained cavity growth. The corrosion damage evolves by steam side and fire side corrosion. The results showed good correlation between the theoretically predicted rupture time and the current research results. And rupture strain may be well estimated by using the proposed method.

• 한국안전학회지
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• 제37권4호
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• pp.129-138
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• 2022

A new method for injecting cooling water into the Korean research reactor (KRR) in the event of beam tube rupture is proposed in this paper. Moreover, the research evaluates the risk to the reactor core in terms of core damage frequency (CDF). The proposed method maintains the cooling water in the chimney at a certain level in the tank to prevent nuclear fuel damage solely by gravitational coolant feeding from the emergency water supply system (EWSS). This technique does not require sump recirculation operations described in the current procedure for resolving beam tube accidents. The reduction in the risk to the core in the event of beam tube rupture that can be achieved by the proposed change in the cooling water injection design is quantified as follows. 1) The total CDF of the KRR for the proposed design change is approximately 4.17E-06/yr, which is 8.4% lower than the CDF of the current design (4.55E-06/yr). 2) The CDF for beam tube rupture is 7.10E-08/yr, which represents an 84.1% decrease compared with that of the current design (4.49E-07/yr). In addition to this quantitative reduction in risk, the modified cooling water injection design maintains a supply of pure coolant to the EWSS tank. This means that the reactor does not require decontamination after an accident. Thermal hydraulic analysis proves that the water level in the reactor pool does not cause damage to the nuclear fuel cladding after beam tube rupture. This is because the amount of water in the chimney can be regulated by the EWSS function. The EWSS supplies emergency water to the reactor core to compensate for the evaporation of coolant in the core, thus allowing water to cover the fuel assemblies in the reactor core over a sufficient amount of time.

• Journal of Mechanical Science and Technology
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• 제19권7호
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• pp.1432-1440
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• 2005

High-temperature rupture behavior of 5083-Al alloy was tested for failure at 548K under multiaxial stress conditions: uniaxial tension using smooth bar specimens, biaxial shearing using double shear bar specimens, and triaxial tension using notched bar specimens. Rupture times were compared for uniaxial, biaxial, and triaxial stress conditions with respect to the maximum principal stress, the von Mises effective stress, and the principal facet stress. The results indicate that the von Mises effective and principal facet stresses give good correlation for the material investigated, and these parameters can predict creep life data under the multiaxial stress states with the rupture data obtained from specimens under the uniaxial stress. The results suggest that the creep rupture of this alloy under the testing condition is controlled by cavitation coupled with highly localized deformation process, such as grain boundary sliding. It is also conceivable that strain softening controls the highly localized deformation modes which result in cavitation damage in controlling rupture time of this alloy.

• 한국압력기기공학회 논문집
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• 제16권1호
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• pp.49-55
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• 2020

Creep rupture data for Alloy 690 steam generator tubes in a pressurized water reactor are essentially needed to demonstrate a severe accident scenario on thermally-induced tube failures caused by hot gases in a damaged reactor core. The rupture data were obtained using the tube specimens under different applied-stress levels at 650℃, 700℃, 750℃, 800℃, and 850℃. Important creep constants were proposed using various creep laws in terms of Norton power law, Monkman-Grant (M-G) relation, damage tolerance factor (λ), and Zener-Hollomon parameter (Z). In addition, a creep activation energy (Q) value for Alloy 690 tube was reasonably determined using experimental data. Creep behaviors such as creep strength, creep rates, rupture elongation showed the results of temperature dependence well. Modified M-G plot improved a correlation of the creep rate and rupture life. Damage tolerance factor for Alloy 690 tubes was found to be λ =2.20 in an average value. Creep activation energy for Alloy 690 tube was optimized for Q=350 (kJ/mol). A plot of Z parameter obeyed a good linearity, and the same creep mechanism was inferred to be operative in the present test conditions.

• 한국지반공학회논문집
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• 제20권7호
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• pp.79-89
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• 2004

본 연구에서는 탄.소.점성 구성모델을 비교적 간단한 수학적 합성유도방식에 기초하여 제안하였다. 이를 위하여 비등방성 modified Cam-Clay model을 일반응력공간으로 확장시켰으며generalized viscous theory를 단순화하여 각각 소성 및 점성의 구성관계로 활용하였다. Damage 원리를 구성모델에 추가하였으며, 모든 식의 변형 및 개발은 모델정수의 수를 감소시키는 원칙에 입각하여 수행하였다. 개발된 구성모델을 활용하여 점성토의 크리프거동을 예측하였으며 이를 실험결과와 비교분석하였다. 예측된 결과는 크리프파괴의 경우를 포함한 실험결과와 비교적 양호하게 일치하는 결과를 보여주었다.

• Journal of Biosystems Engineering
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• 제14권3호
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• pp.207-214
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• 1989

In this study, impact force and angular displacement of the pendulum were measured by the load cell and potentiometer. Mechanical behavior of rough rice under impact loading was able to analyze precisely and efficiently, because measured data were accumulated and handled by the automatic data acquisition system making use of microcomputer system. Impact force and angular displacement were measured with a resolutiln of 1/1500 seconds in time. Mechanical behavior such as force and energy at rupture point of Japonica type and Indica type rough rice were measured with this system. After impact loading, the damage of rough rice was examined with the microphotograph and an allowable impact force was measured. The results obtained in this study are summarized as follows. 1. Machanical behavior of rough rice under impact loading was analyzed precisely and efficiently because measured data were accumulated and handled by this data acquisition system. 2. Rupture force and rupture energy of rough rice were appeared to be the lowest value in the range of 16 to 18 % moisture content, and rupture force and rupture energy of Japonica type were higher than those of Indica type in each level of moisture content. 3. From the result of the damage examined after the impact loading, allowable impact force was the lowest in the range of 16 to 18 % moisture content, and the value of the allowable impact force of Japonica type was higher than that of Indica type in each level of moisture content.

• International Journal of Precision Engineering and Manufacturing
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• 제8권4호
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• pp.33-37
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• 2007

Creep failure of 316LN stainless steel (SS) occurs due to the nucleation and growth of cracks. An investigation was performed to correlate the creep damage with ultrasonic wave speeds and angular frequencies using creep-tested 316LN SS specimens. Ultrasonic wave measurements were made in the direction of and perpendicular to the loading using contact probes with central frequencies of 10, 15, and 20 MHz. We found that the angular frequency and wave speed decreased with increasing creep time to rupture by analyzing the ultrasonic signals from the 15 and 20 MHz probes. Therefore, the creep damage was sensitive to the angular frequency and wave speed of ultrasonic waves.

• 대한기계학회논문집A
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• 제24권7호
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• pp.1833-1840
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• 2000

Commercially pure copper is tested to obtain creep curves at 2500C. Constitutive relations adopting continuum damage mechanics concept is found to be appropriate to model the creep defor mation up to the tertiary stage. Microscopic observation by SEM reveals that creep condition induces cavities and microcracks subsequently. The constitutive equations along with evaluated creep parameters are implemented into finite element analysis code. The analysis reproduces creep curves under step loading as well as constant loading with reasonable accuracy. Distribution and evolution of damage under creep loading are numerically simulated for two different types of notched specimen. Predicted creep life agrees quite well with rupture test results. The influence of mesh size at notch tip on rupture time prediction is studied, and a degree of refinement is suggested for the specific notched specimens.

• Nuclear Engineering and Technology
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• 제36권1호
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• pp.112-120
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• 2004

The purpose of this paper is to analyze the piping failure frequency for the main feedwater system in domestic nuclear power plants(NPPs) for the application to an in-service inspection(ISI), leak before break(LBB) concept, aging management program(AMP), and probabilistic safety analysis(PSA). First, a database was developed for piping failure events in domestic NPPs, and 23 domestic piping failure events were collected. Among the 23 events, 12 locations of wall thinning due to flow accelerated corrosion(FAC) were identified in the main feedwater system in 4 domestic WH 3-loop NPPs. Two types of the piping failure frequency such as the damage frequency and rupture frequency were considered in this study. The damage frequency was calculated from both the plant population data and damage(s) including crack, wall thinning, leak, and/or rupture, while the rupture frequency was estimated by using both the well-known Jeffreys method and a new method considering the degradation due to FAC. The results showed that the damage frequencies based on the number of the base metal piping susceptible to FAC ranged from $1.26{\times}10^{-3}/cr.yr\;to\;3.91{\times}10^{-3}/cr.yr$ for the main feedwater system of domestic WH 3-loop NPPs. The rupture frequencies obtained from the Jeffreys method for the main feedwater system were $1.01{\times}10^{-2}/cr.yr\;and\;4.54{\times}10^{-3}/cr.yr$ for the domestic WH 3-loop NPPs and all the other domestic PWR NPPs respectively, while those from the new method considering the degradation were higher than those from the Jeffreys method by about an order of one.