Fracture Analysis of Vessels. Oak Ridge FAVOR, v06.1, Computer Code: Theory and Implementation of Algorithms, Methods, and Correlations

Williams, Paul T.; Dickson, T.L.; Yin, Shixiong

doi:10.2172/1154649

Cited by 26 publications

(48 citation statements)

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“…The other probabilistic fracture mechanics codes for piping were not developed for NRC, but by other organizations, and as such are not generally available in the public domain (Bell and Chapman 2003;Bishop 1997). Excluded from the discussion are other significant probabilistic fracture mechanics codes, such as FAVOR (Dickson 1994;Dickson et al 2004;Williams et al 2004) and VISA-II (Simonen et al 1986), which were specifically developed to predict failure probabilities for reactor pressure vessels that are subject to radiation embrittlement.…”

Section: Probabilistic Fracture Mechanics Codesmentioning

confidence: 99%

Probabilistic Fracture Mechanics Evaluation of Selected Passive Components ? Technical Letter Report

Simonen¹,

Doctor²,

Gosselin³

et al. 2007

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This report addresses the potential application of probabilistic fracture mechanics computer codes to support the Proactive Materials Degradation Assessment (PMDA) program as a method to predict component failure probabilities. The present report describes probabilistic fracture mechanics calculations that were performed for selected components using the PRO-LOCA and PRAISE computer codes. The calculations address the failure mechanisms of stress corrosion cracking, intergranular stress corrosion cracking, and fatigue for components and operating conditions that are known to make particular components susceptible to cracking. It was demonstrated that the two codes can predict essentially the same failure probabilities if both codes start with the same fracture mechanics model and the same inputs to the model. Comparisons with field experience showed that both codes predict relatively high failure probabilities for components under operating conditions that have resulted in field failures. It was found that modeling assumptions and inputs tended to give higher calculated failure probabilities than those derived from data on field failures. Sensitivity calculations were performed to show that uncertainties in the probabilistic calculations were sufficiently large to explain the differences between predicted failure probabilities and field experience.iii iv Executive SummaryThe U.S. Nuclear Regulatory Commission (NRC) has supported the research program Proactive Materials Degradation Assessment (PMDA). The objective of this program has been to predict future occurrences of materials degradation that may or may not have been observed in the field or in the laboratory. Evaluations have focused on materials degradation modes associated with operating environments for specific components, including stress corrosion cracking, fatigue, flow-accelerated corrosion, boric acid corrosion, thermal embrittlement, and radiation effects. A detailed review addressed over 2000 components in the primary, secondary, and tertiary systems of specific reactor designs. A group of experts provided their judgments to score individual components in terms of "degradation susceptibility" and the "extent of knowledge" available for developing mitigation actions.This report addresses the possible application of probabilistic fracture mechanics computer codes to support the PMDA program as a method to predict component failure probabilities. Probabilistic fracture mechanics calculations are described that were performed for selected components using the PRO-LOCA and PRAISE computer codes. The calculations address the failure mechanisms of stress corrosion cracking, intergranular stress corrosion cracking, and fatigue for components and operating conditions that are known to have failed components in the field. The calculations allowed the two computer codes to be benchmarked against each other and, more importantly, benchmarked against field experience.A review of the calculations showed how uncertainties and modeling assumptions can im...

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Section: Probabilistic Fracture Mechanics Codesmentioning

confidence: 99%

Probabilistic Fracture Mechanics Evaluation of Selected Passive Components ? Technical Letter Report

Simonen¹,

Doctor²,

Gosselin³

et al. 2007

View full text Add to dashboard Cite

show abstract

“…Related to the previous version of FAVOR code, the principal modifications of the advanced version are as follows [9,10]: (1) having the capability to model the flaw populations such as internal or external surface breaking flaws, thus permitting it to analyze either cool-down or heat-up transient effects on RPVs and (2) modifying the stress intensity factor influence coefficients (SIFICs), which consider various flaw orientations, geometries, categories, and aspect ratios to calculate the mode I stress intensity factor for each simulated flaw, to be applicable to / ratio of RPVs from 10 to 20 for either PWR or BWR geometry. In this work, FAVOR's two processors, FAVLoad and FAVPFM, were used.…”

Section: Evaluation Methodologymentioning

confidence: 99%

Probabilistic Structural Integrity Analysis of Boiling Water Reactor Pressure Vessel under Low Temperature Overpressure Event

Chou

Huang

2015

International Journal of Nuclear Energy

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The probabilistic structural integrity of a Taiwan domestic boiling water reactor pressure vessel has been evaluated by the probabilistic fracture mechanics analysis. First, the analysis model was built for the beltline region of the reactor pressure vessel considering the plant specific data. Meanwhile, the flaw models which comprehensively simulate all kinds of preexisting flaws along the vessel wall were employed here. The low temperature overpressure transient which has been concluded to be the severest accident for a boiling water reactor pressure vessel was considered as the loading condition. It is indicated that the fracture mostly happens near the fusion-line area of axial welds but with negligible failure risk. The calculated results indicate that the domestic reactor pressure vessel has sufficient structural integrity until doubling of the present end-of-license operation.

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“…The regulations and standards have been taken into account the PFM approach to evaluate structural integrity of RPV and RPV cladding including several cases (Vesely et al1978, U.S. Nuclear, 1987. Depending on the PFM approach several number of computer programs has been developed worldwide to assess the structural integrity of RPV cladding and piping (ASME, 1991, Muhammed, 2003, Williams et al 2004, Saarenheimo and Simola 2004, Onizawa et al 2004. Safety aspects of crack initiation in nuclear RPV is inconceivable and emergency conditions may arise due to its consequence.…”

Section: Introductionmentioning

confidence: 99%

Probabilistic Assessment of Crack Failure of Reactor Pressure Vessel (RPV) Cladding Material

Sohel

Hossain

Datta

et al. 2018

J Bangladesh Acad Sci

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To design a Reactor Pressure Vessel (RPV), material property like crack must be considered as it is an unavoidable property of materials. Presence of crack in materials must be kept within limit to prevent material's failure. So, crack propagation must be analyzed and observed. In this paper, crack propagation due to stress and materials fracture toughness of reactor pressure vessel cladding has been observed to estimate cumulative probability of crack failure using Probabilistic Fracture Mechanics (PFM). Average crack size is guessed as 3 mm and geometry factor is considered as 1.12 to analyze edge crack. Final crack analysis range has been found to be 1.8 mm with crack propagation rate of  30% of its average size. Variation of critical crack size and crack initiation point for several design stresses and fracture toughness has been investigated with probabilistic fracture mechanics technique. The observed crack propagation by calculating final crack size and the cumulative crack failure probability of the reactor pressure vessel materials are presented in this work.

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Fracture Analysis of Vessels. Oak Ridge FAVOR, v06.1, Computer Code: Theory and Implementation of Algorithms, Methods, and Correlations

Cited by 26 publications

References 0 publications

Probabilistic Fracture Mechanics Evaluation of Selected Passive Components ? Technical Letter Report

Probabilistic Fracture Mechanics Evaluation of Selected Passive Components ? Technical Letter Report

Probabilistic Structural Integrity Analysis of Boiling Water Reactor Pressure Vessel under Low Temperature Overpressure Event

Probabilistic Assessment of Crack Failure of Reactor Pressure Vessel (RPV) Cladding Material

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