Microstructural evolution in fast-neutron-irradiated austenitic stainless steels

Stoller, RE

doi:10.2172/5436209

Cited by 14 publications

(11 citation statements)

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“…Another contributor to this dose dependence is the accumulation of transmutation products that stabilize point defect clusters. In particular, helium is known to promote formation of black spot damage [78] and delay the annealing of damage structure [79]. In ferritic steels, damage microstructure is not easily visible so T da must be determined from disappearance of radiationinduced increases in yield strengths and Charpy impact transition temperatures.…”

Section: Discussion Of Phase 3 Findingsmentioning

confidence: 99%

MAPPING FLOW LOCALIZATION PROCESSES IN DEFORMATION OF IRRADIATED REACTOR STRUCTURAL ALLOYS - FINAL REPORT. Nuclear Energy Research Initiative Program No. MSF99-0072. Period: August 1999 through September 2002. (ORNL/TM-2003/63)

Farrell¹

2003

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Section: Discussion Of Phase 3 Findingsmentioning

confidence: 99%

MAPPING FLOW LOCALIZATION PROCESSES IN DEFORMATION OF IRRADIATED REACTOR STRUCTURAL ALLOYS - FINAL REPORT. Nuclear Energy Research Initiative Program No. MSF99-0072. Period: August 1999 through September 2002. (ORNL/TM-2003/63)

Farrell¹

2003

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“…The experimentally determined evolution of the precipitates is provided as input to the code, by parameterizing the precipitate radius and density as a function of dpa. Secondly, the dislocation microstructure is treated as either experimentally input to the code, or is allowed to evolve dynamically by a model for dislocation loop and network growth provided in RIME [24,25]. Finally, the M A N U S C R I P T…”

Section: Accepted Manuscriptmentioning

confidence: 99%

“…For the purpose of this paper, the helium treatment was neglected because helium pre-injection is expected to predominantly affect the nucleation regime. A full description of the model applied to austenitic stainless steels under light water reactor conditions is given in [22,23], which was developed from work by Stoller and Odette [24,25]. Basic physics will be briefly described here, with the focus on describing the additional treatments for precipitates, dislocations and carbon in solution.…”

Section: A N U S C R I P Tmentioning

confidence: 99%

The co-evolution of microstructure features in self-ion irradiated HT9 at very high damage levels

Getto

VanCoevering

Was

2017

Journal of Nuclear Materials

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Understanding the void swelling and phase evolution of reactor structural materials at very high damage levels is essential to maintaining safety and longevity of components in Gen IV fast reactors. A combination of ion irradiation and modeling was utilized to understand the microstructure evolution of ferritic-martensitic alloy HT9 at high dpa. Self-ion irradiation experiments were performed on alloy HT9to determine the co-evolution of voids, dislocations and precipitates up to 650 dpa at 460°C. Modeling of microstructure evolution was conducted using the modified Radiation Induced Microstructure Evolution (RIME) model, which utilizes a mean field rate theory approach with grouped cluster dynamics.Irradiations were performed with 5 MeV raster-scanned Fe 2+ ions on samples pre-implanted with 10 atom parts per million He. The swelling, dislocation and precipitate evolution at very high dpa was determined using Analytical Electron Microscopy in Scanning Transmission Electron Microscopy (STEM) mode.Experimental results were then interpreted using the RIME model. A microstructure consisting only of dislocations and voids is insufficient to account for the swelling evolution observed experimentally at high damage levels in a complicated microstructure such as irradiated alloy HT9. G phase was found to have a minimal effect on either void or dislocation evolution. M 2 X plays two roles; a variable biased sink for defects, and as a vehicle for removal of carbon from solution, thus promoting void growth. When accounting for all microstructure interactions, swelling at high damage levels is a dynamic process that continues to respond to other changes in the microstructure as long as they occur.

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“…Table 1 lists the nominal or baseline set of material and irradiation parameters employed in the results shown below. The nominal material parameters such as the dislocation bias are taken from previous work investigating the swelling of 316 stainless steel at higher temperatures [3,19], and the irradiation parameters are representative of the service environment of LWR internal components. Simulations have been carried out over a range of temperatures to account for the gamma heating contribution to temperature in the thicker components.…”

Section: Results Of Simulationsmentioning

confidence: 99%

Swelling in light water reactor internal components: Insights from computational modeling

Stoller

Barashev

Golubov

2015

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A modern cluster dynamics model has been used to investigate the materials and irradiation parameters that control microstructural evolution under the relatively low-temperature exposure conditions that are representative of the operating environment for in-core light water reactor components. The focus is on components fabricated from austenitic stainless steel. The model accounts for the synergistic interaction between radiation-produced vacancies and the helium that is produced by nuclear transmutation reactions. Cavity nucleation rates are shown to be relatively high in this temperature regime (275 to 325°C), but are sensitive to assumptions about the fine scale microstructure produced under low-temperature irradiation. The cavity nucleation rates observed run counter to the expectation that void swelling would not occur under these conditions. This expectation was based on previous research on void swelling in austenitic steels in fast reactors. This misleading impression arose primarily from an absence of relevant data. The results of the computational modeling are generally consistent with recent data obtained by examining ex-service components. However, it has been shown that the sensitivity of the model's predictions of low-temperature swelling behavior to assumptions about the primary damage source term and specification of the mean-field sink strengths is somewhat greater that that observed at higher temperatures. Further assessment of the mathematical model is underway to meet the long-term objective of this research, which is to provide a predictive model of void swelling at relevant lifetime exposures to support extended reactor operations.

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Microstructural evolution in fast-neutron-irradiated austenitic stainless steels

Cited by 14 publications

References 0 publications

MAPPING FLOW LOCALIZATION PROCESSES IN DEFORMATION OF IRRADIATED REACTOR STRUCTURAL ALLOYS - FINAL REPORT. Nuclear Energy Research Initiative Program No. MSF99-0072. Period: August 1999 through September 2002. (ORNL/TM-2003/63)

MAPPING FLOW LOCALIZATION PROCESSES IN DEFORMATION OF IRRADIATED REACTOR STRUCTURAL ALLOYS - FINAL REPORT. Nuclear Energy Research Initiative Program No. MSF99-0072. Period: August 1999 through September 2002. (ORNL/TM-2003/63)

The co-evolution of microstructure features in self-ion irradiated HT9 at very high damage levels

Swelling in light water reactor internal components: Insights from computational modeling

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