TEM characterization of dislocation loops in irradiated bcc Fe-based steels

Yao, Bo; Edwards, Danny J.; Kurtz, Richard J.

doi:10.1016/j.jnucmat.2012.12.002

Cited by 167 publications

(59 citation statements)

References 24 publications

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“…Compared to the only Cu being the alloying element, introduction of other alloying elements such as Ni and Mn was found to delay the precipitation kinetics, in terms of both reduced density and average size of the precipitated. It has been shown in previous AKMC study that the presence of Ni enhances the nucleation but impedes the growth of Cu clusters [23]. Indeed, if we consider all Cu clusters with 2 or more Cu atoms, the density was higher for the Cu-Ni-Mn case than that for Cu only.…”

Section: Precipitation Kinetics From Akmc Simulationsmentioning

confidence: 72%

Modeling of Late Blooming Phases and Precipitation Kinetics in Aging Reactor Pressure Vessel (RPV) Steels

Zhang¹,

Chakraborty²,

Biner³

2013

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The reactor pressure vessel, RPV, houses the reactor core and because of this function it has direct safety significance. The RPV is considered to be irreplaceable or prohibitively expensive to replace, which means that if the RPV degrades sufficiently it could be the operational life-limiting feature of a nuclear plant.In general, there are three main contributors in service embrittlement:Damage evolution in the form of vacancy and self interstitial atom (SIA) and clusters as a result of displacement cascades.Segregation of alloying elements Cu, Ni, Mn and Si, forming 2-3 nanometer clusters/precipitates and possible segregation of P to the grain boundaries.Environmental variables, irradiation dose, and dose rate and temperature.Currently, the RPV integrity is assessed using a fracture mechanics approach based on:1. The development of a Master Fracture Toughness Curve (ASME Code Case N-631 Section III).2. The determination of the ductile-brittle transition temperature shift in the master fracture toughness curve during in service.The determination of this temperature shift is the prime importance in ensuring the structural integrity of the RPVs. This is accomplished through surveillance programs, accelerated irradiation experiments (PIA), and followed by the statistical modeling of the surveillance and PIA database by non-linear and least-square regression analysis, such as the EONY model, which are valid for only available data set.To extend these current approaches beyond the scope of the surveillance test program is a challenge because of the scarcity of the future data. Currently, there is not enough experimentally generated data to assess with a high degree of certainty that current embrittlement trend curves, developed from the nuclear power plants, can be extrapolated to beyond 40 years. The reasons for this are:There is a clear need to establish, whether or not the additional damage mechanisms which were not been observed so far, appear at higher influences. There are now indications of late segregations of Ni, Mn and Si (so called late blooming phases) near the end of 40 years in service.Most of the databank on RPVs embrittlement has been obtained from the specimens irradiated in test or experimental reactors, as opposed to irradiations performed as part of power reactor surveillance programs. Many of the test reactor specimens have been exposed to fluences far higher than anticipated, even at or beyond 80 years of operation. However, because of the accelerated exposure achieved in test reactors versus power reactors, the concerns persist that the different damage mechanisms may be active in these different reactor environments. If it is true, this definitely would make test reactor data an unreliable predictor for the power reactor embrittlement.There are inconsistencies with the master curve approach, where the measured and predicted behaviors sometimes do not agree. This is typically in situations where one or more of the basic assumptions (usually either the fracture mode or homogeneity) are...

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Section: Precipitation Kinetics From Akmc Simulationsmentioning

confidence: 72%

Modeling of Late Blooming Phases and Precipitation Kinetics in Aging Reactor Pressure Vessel (RPV) Steels

Zhang¹,

Chakraborty²,

Biner³

2013

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“…According to experimental observations, the loops may have either ½<111> or <100> burgers vector, residing on {111} or {100} crystallographic planes respectively [50]. In the following, only the results for {100} loops will be summarized.…”

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confidence: 99%

Light Water Reactor Sustainability Program Grizzly Year-End Progress Report

Spencer¹,

Zhang²,

Chakraborty³

et al. 2013

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The Grizzly software application is being developed under the Light Water Reactor Sustainability (LWRS) program to address aging and material degradation issues that could potentially become an obstacle to life extension of nuclear power plants beyond 60 years of operation. Grizzly is based on INL's MOOSE multiphysics simulation environment, and can simultaneously solve a variety of tightly coupled physics equations, and is thus a very powerful and flexible tool with a wide range of potential applications. Grizzly, the development of which was begun during fiscal year (FY) 2012, is intended to address degradation in a variety of critical structures.The reactor pressure vessel (RPV) was chosen for an initial application of this software. Because it fulfills the critical roles of housing the reactor core and providing a barrier to the release of coolant, the RPV is clearly one of the most safety-critical components of a nuclear power plant. In addition, because of its cost, size and location in the plant, replacement of this component would be prohibitively expensive, so failure of the RPV to meet acceptance criteria would likely result in the shutting down of a nuclear power plant.During service, the RPV is subjected to intense neutron flux. This flux, combined with thermal aging affects, has an embrittling effect on the RPV steel, which is manifested as an increase in the brittle/ductile transition temperature of that material. This means that at a given temperature, aging effects cause the steel to have a lower toughness and tend to fracture in a more brittle manner.The current practice used to perform engineering evaluations of the susceptibility of RPVs to fracture is to use the ASME Master Fracture Toughness Curve (ASME Code Case N-631 Section III), which provides the fracture toughness as a function of temperature. This is used in conjunction with empirically based models that describe the evolution of this curve due to embrittlement in terms of a transition temperature shift. These models are based on an extensive database of surveillance coupons that have been irradiated in operating nuclear power plants, but this data is limited to the lifetime of the current reactor fleet. This is an important limitation when considering life extension beyond 60 years. The currently available data cannot be extrapolated with confidence further out in time because there is a potential for additional damage mechanisms (i.e. late blooming phases) to become active later in life beyond the current operational experience.To improve our understanding of the degradation of RPV steels under the conditions that would be seen under extended service life, science-based predictive modeling at the atomistic scale is needed. This modeling, combined with the limited applicable experimental data in the regime of interest, can provide understanding of the fundamental degradation mechanisms, with the goal of eventually informing models that can be useful for engineering-scale evaluations of RPVs.

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“…For each zone axis imaged, dislocation loop orientation maps generated by Yao, et al [163] were used to determine that the loops commonly reside on the {111} or {001} habit planes. Following Fe 2+ irradiation to 1 dpa, dislocation loops have an average diameter of 8.5 ± 2.7 nm and number density of 2.8 ± 0. are also similar at 9.8 ± 3.4 nm and 1.9 ± 0.5 x 10 21 m -3 , respectively.…”

Section: Microstructure Results In Fe-9%cr Odsmentioning

confidence: 99%

“…Quantitative TEM microstructure results are provided in Table 5 For each zone axis imaged, dislocation loop orientation maps generated by Yao, et al [163] were used to determine that the loops commonly resided on the {111} or {001} habit planes [163]. Upon proton irradiation, dislocation loops were consistent in size following 1 dpa and 3 dpa, at 7.6 ± 2.1 nm 7.6 ± 2.4 nm, respectively, while number density increased from 1.6 ± 0.…”

Section: Microstructure Results In Hcm12amentioning

confidence: 99%

“…In some cases, dislocation loops will appear as round circles, while loops with different orientation may appear as "edge-on". For loops imaged close to an "edge-on" condition, the longer dimension observed is taken to be the loop diameter [163]. For each loop identified, the diameter is measured, and an overall average diameter of loops (dl), along with the standard deviation, is calculated.…”

Section: Stem Imagingmentioning

confidence: 99%

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The Mechanism of Radiation-Induced Nanocluster Evolution in Oxide Dispersion Strengthened and Ferritic-Martensitic Alloys

Swenson¹

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TEM characterization of dislocation loops in irradiated bcc Fe-based steels

Cited by 167 publications

References 24 publications

Modeling of Late Blooming Phases and Precipitation Kinetics in Aging Reactor Pressure Vessel (RPV) Steels

Modeling of Late Blooming Phases and Precipitation Kinetics in Aging Reactor Pressure Vessel (RPV) Steels

Light Water Reactor Sustainability Program Grizzly Year-End Progress Report

The Mechanism of Radiation-Induced Nanocluster Evolution in Oxide Dispersion Strengthened and Ferritic-Martensitic Alloys

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