• 한국안전학회지
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• 제22권4호
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• pp.7-12
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• 2007

In order to evaluate the integrity of Zr-2.5Nb pressure tubes, probabilistic fracture mechanics(PFM) approach was employed. Failure assessment diagram(FAD), plastic collapses, and critical crack lengths(CCL) were used for evaluating the failure probability as failure criteria. The Kr-FAD as failure assessment diagram was used because fracture of pressure tubes occurred in brittle manner due to hydrogen embrittlement of material by deuterium fluence. The probabilistic integrity evaluation observed AECL procedures and used fracture toughness parameters of EPRI and recently announced theory. In conclusion, the probabilistic approach using the Kr-FAD made it possible to determine major failure criterion in the pressure tube integrity evaluation.

• Structural Engineering and Mechanics
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• 제19권3호
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• pp.347-359
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• 2005

This paper attempts to develop the analytical model of estimating the fatigue damage using a linear elastic fracture mechanics method. The stress history on a welding member, when a truck passed over a bridge, was defined as a block loading and the crack closure theory was used. These theories explain the influence of a load on a structure. This study undertook an analysis of the stress range frequency considering both dead load stress and crack opening stress. A probability method applied to stress range frequency distribution and the probability distribution parameters of it was obtained by Maximum likelihood Method and Determinant. Monte Carlo Simulation which generates a probability variants (stress range) output failure block loadings. The probability distribution of failure block loadings was acquired by Maximum likelihood Method and Determinant. This can calculate the fatigue reliability preventing the fatigue failure of a welding member. The failure block loading divided by the average daily truck traffic is a predictive remaining life by a day. Fatigue reliability analysis was carried out for the welding member of the bottom flange of a cross beam and the vertical stiffener of a steel box bridge by the proposed model. Results showed that the primary factor effecting failure time was crack opening stress. It was important to decide the crack opening stress for using the proposed model. Also according to the 50% reliability and 90%, 99.9% failure times were indicated.

• Nuclear Engineering and Technology
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• 제48권2호
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• pp.545-553
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• 2016

The failure probabilities of the reactor pressure vessel (RPV) for low temperature over-pressurization (LTOP) and cool-down transients are calculated in this study. For the cool-down transient, a pressure-temperature limit curve is generated in accordance with Section XI, Appendix G of the American Society of Mechanical Engineers (ASME) code, from which safety margin factors are deliberately removed for the probabilistic fracture mechanics analysis. Then, sensitivity analyses are conducted to understand the effects of some input parameters. For the LTOP transient, the failure of the RPV mostly occurs during the period of the abrupt pressure rise. For the cool-down transient, the decrease of the fracture toughness with temperature and time plays a main role in RPV failure at the end of the cool-down process. As expected, the failure probability increases with increasing fluence, Cu and Ni contents, and initial reference temperature-nil ductility transition ($RT_{NDT}$). The effect of warm prestressing on the vessel failure probability for LTOP is not significant because most of the failures happen before the stress intensity factor reaches the peak value while its effect reduces the failure probability by more than one order of magnitude for the cool-down transient.

• 한국정밀공학회지
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• 제22권10호
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• pp.137-143
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• 2005

The integrity of steam generator tubes in nuclear power plant should be maintained sufficiently during operation. For sake of this, complicated assessment procedures are required such as fracture mechanics analysis, etc. The integrity assessment of tubes has been performed by using conventional deterministic approaches while there are many uncertainties to carry out a rational evaluation. In this respect, probabilistic integrity assessment is considered as an alternative method for integrity assessment. The objectives of this study are to develop an integrity assessment system based on probabilistic fracture mechanics and to predict the failure probability of steam generator tubes containing an axial through-wall crack. The developed integrity assessment system consists of three evaluation modules, which apply first order reliability method, second order reliability method and Monte Carlo simulation method, respectively. The system has been applied to predict failure probability of steam generator tubes and the estimation results showed a promising applicability of the probabilistic integrity assessment system.

• 한국안전학회지
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• 제21권1호
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• pp.28-34
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• 2006

A procedure of estimating failure probability is demonstrated for a pressurized pipe of CrMo steel used at $538^{\circ}C$. Probabilistic fracture mechanics were employed considering variations of pressure loading, material properties and geometry. Probability density functions of major material variables were determined by statistical analyses of implemented data obtained by previous experiments. Distributions of the major variables were reflected in Monte Carlo simulation and failure probability as a function of operating time was determined. The creep crack growth life assessed by conventional deterministic approach was shown to be conservative compared with those obtained by probabilistic one. Sensitivity analysis for each input variable was also conducted to understand the most influencing variables to the residual life analysis. Internal pressure, creep crack growth coefficient and creep coefficient were more sensitive to failure probability than other variables.

• 대한기계학회논문집A
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• 제27권6호
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• pp.987-996
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• 2003

The structural integrity of the reactor vessel with the approaching end of life must be assured for pressurized thermal shock. The regulation specifies the screening criteria for this and requires that specific analysis be performed for the reactor vessel which is anticipated to exceed the screening criteria at the end of plant life. In case the screening criteria is exceeded by the deterministic analysis, probabilistic analysis must be performed to show that failure probability Is within the limit. In this study, probabilistic fracture mechanics analysis of the reactor vessel for pressurized thermal shock is performed and the effects of residual stress and master curve on the failure probability are investigated.

• 마이크로전자및패키징학회지
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• 제21권2호
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• pp.59-64
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• 2014

Transfer is the critical issue of producing high-quality and scalable graphene electronic devices. However, conventional transfer processes require the removal of an underlying metal layer by wet etching process, which induces significant economic and environmental problems. We propose the etching-free mechanical releasing of graphene using polymer adhesives. A fracture mechanics approach was introduced to understand the releasing mechanism and ensure highyield process. It is shown that the thickness of adhesive and target substrate affect the transferability of graphene. Based on experimental and fracture mechanics simulation results, we further observed that compliant adhesives can reduce the adhesive stress during the transfer, which also enhances the success probability of graphene transfer.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 추계학술대회논문집A
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• pp.155-160
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• 2001

This paper describes a probabilistic fracture mechanics (PFM) analysis based on Monte Carlo (MC) simulation. In the analysis of CANDU pressure tube, it is necessary to perform the PFM analyses based on statistical consideration of flaw generation time. A depth and an aspect ratio of initial semi-elliptical surface crack, a fracture toughness value, delayed hydride cracking (DHC) velocity, and flaw generation time are assumed to be probabilistic variables. In all the analyses, degradation of fracture toughness due to neutron irradiation is considered. Also, the failure criteria considered are plastic collapse, unstable fracture and crack penetration. For the crack growth by DHC, the failure probability was evaluated in due consideration of flaw generation time.

• Nuclear Engineering and Technology
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• 제44권8호
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• pp.961-970
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• 2012

Often, probabilistic fracture mechanics (PFM) approaches have been adopted to quantify the failure probabilities of Ni-base alloy components, especially due to primary water stress corrosion cracking (PWSCC), in a primary piping system of pressurized water reactors. In this paper, the key features of an advanced PFM code, PINEP-PWSCC (Probabilistic INtegrity Evaluation for nuclear Piping-PWSCC) for such purpose, are described. In developing the code, we adopted most recent research results and advanced models in calculation modules such as PWSCC crack initiation and growth models, a performance-based probability of detection (POD) model for Ni-base alloy welds, and so on. To verify the code, the failure probabilities for various Alloy 182 welds locations were evaluated and compared with field experience and other PFM codes. Finally, the effects of pre-existing crack, weld repair, and POD models on failure probability were evaluated to demonstrate the applicability of PINEP-PWSCC.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2008년도 추계학술대회A
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• pp.727-728
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• 2008

The objective of this study is to evaluate the integrity for a reactor pressure vessel under the pressurized thermal shock by applying the probability fracture mechanics. A semi-elliptical axial crack is assumed to be in the beltline region of the reactor pressure vessel. The selected random variables are the neutron fluence on the vessel inside surface, the content of copper, nickel, and phosphorus in the reactor pressure vessel material, and initial RTNDT. The probabilistic integrity analysis was performed using the Monte Carlo simulation.