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Garner, F. A.

Publications and source records attributed to Garner, F. A..

15 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Corrigendum to “Accelerated materials evaluation for nuclear applications” [J. Nucl. Mater. 488 (2017) 46–62]

The original article addresses the opportunities and complexities of using materials test reactors with high neutron fluxes to perform accelerated studies of material aging in power reactors operating at lower neutron fluxes and with different neutron flux spectra. Radiation damage and gas production in different reactors have been compared using the code, SPECTER. This code provides a common standard from which to compare neutron damage data generated by different research groups using a variety of reactors. This Corrigendum identifies a few typographical errors. Tables 2 and 3 are included in revised form.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Effect of tube processing methods on microstructure, mechanical properties and irradiation response of 14YWT nanostructured ferritic alloys

In this research, innovative thermal spray deposition (Process I) and conventional hot extrusion processing (Process II) methods have been used to produce thin walled tubing (~0.5 mm wall thickness) out of 14YWT, a nanostructured ferritic alloy. The effects of processing methods on the microstructure, mechanical properties and irradiation response have been investigated by using scanning electron microscopy (SEM), transmission electron microscopy (TEM) and, micro- and nano-hardness techniques. It has been found that these two processes have a significant effect on the microstructure and mechanical properties of the as-fabricated 14YWT tubes. Even though both processing methods yield the formation of various size Y-Ti-O particles, the conventional hot extrusion method results in a microstructure with smaller, homogenously distributed nano-oxides (NOs, Y-Ti-O particles < 5 nm) with higher density. Therefore, Process II tubes exhibit twice the hardness of Process I tubes. It has also been found that these two tremendously different initial microstructures strongly affect irradiation response in these tubes under extremely high dose ion irradiations up to 1100 peak dpa at 450 °C. The finer, denser and homogenously distributed NOs in the Process II tube result in a reduction in swelling by two orders of magnitude. On the other hand, inhomogeneity of the initial microstructure in the Process I tube leads to large variations in both swelling and irradiation induced hardening. Moreover, hardening mechanisms before and after irradiation were measured and compared with detailed calculations. In conclusion, this study clearly indicates the crucial effect of initial microstructure on radiation response of 14YWT alloys.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Radiation instability of equal channel angular extruded T91 at ultra-high damage levels

Although several studies have shown that nano-grained alloys have higher radiation resistance than coarse-grained variants, with dense grain boundary structures acting as defect sinks to remove damage, the present study shows the failure of severe deformation to suppress void swelling in T91 alloy at ultra-high damage levels. Fe self-ion irradiation up to 1000 peak dpa produced significantly more swelling in extruded T91 than conventional T91 at 475 °C. Dramatic radiation-enhanced grain growth was observed in the extruded T91, manifested in two rapid grain growth stages. One stage occurs at a relatively low damage level, and the other at a higher damage level beyond 750 peak dpa. The second rapid growth stage is correlated with void swelling and starts at the end of the swelling incubation period. The study suggests that the defect sink properties of the growing grain boundaries under a coarsening process are different from that of stable boundary configurations. When void swelling becomes significant and voids act as a biased sink for vacancies, the growing grain boundaries act as a biased sink for interstitials, leading to rapid growth at higher damage levels.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Influence of injected interstitials on the void swelling in two structural variants of 304L stainless steel induced by self-ion irradiation at 500 °C

The two variants of AISI 304L stainless steel (SS) with different grain size distributions were ion irradiated at 500 °C to a peak dose of ~60 dpa. In the coarse-grained annealed variant, a peak swelling of ~12% was observed closer to the specimen surface rather than at the depth of peak displacement damage. The forward shift in depth between peak swelling and peak dose is proposed to be a consequence of suppression of void nucleation by injected interstitials. The swelling behavior in the front portion of the ion range mirrors that of neutron-induced swelling in this steel, exhibiting significant curvature with increasing dose as the swelling rate approaches the terminal swelling rate of 1%/dpa. Furthermore, an ultrafine grain variant of this steel produced by severely plastic deformation exhibits a similar suppression of void nucleation in the injected interstitial region, but also shows a significantly extended transient regime, not reaching the terminal swelling rate by 60 dpa.

304 stainless steel↗
DOE OSTI2017

Accelerated materials evaluation for nuclear applications

This report addresses the opportunities and complexities of using materials test reactors with high neutron fluxes to perform accelerated studies of material aging in power reactors operating at lower neutron fluxes and with different neutron flux spectra. Radiation damage and gas production in different reactors have been compared using the code, SPECTER. This code provides a common standard from which to compare neutron damage data generated by different research groups using a variety of reactors.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Stability of nanosized oxides in ferrite under extremely high dose self ion irradiations

We produced a nanostructured ferritic alloy (NFA), 14YWT, in the form of thin walled tubing. The stability of the nano-oxides (NOs) was determined under 3.5 MeV Fe +2 irradiations up to a dose of ~585 dpa at 450 °C. Transmission electron microscopy (TEM) and atom probe tomography (APT) show that severe ion irradiation results in a ~25% reduction in size between the unirradiated and irradiated case at 270 dpa while no further reduction within the experimental error was seen at higher doses. Conversely, number density increased by ~30% after irradiation. Moreover, this ‘inverse coarsening’ can be rationalized by the competition between radiation driven ballistic dissolution and diffusional NO reformation. There were no significant changes in the composition of the matrix or NOs observed after irradiation. Modeling the experimental results also indicated a dissolution of the particles.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Radiation response of alloy T91 at damage levels up to 1000 peak dpa

Ferritic/martensitic alloys are required for advanced reactor components to survive 500e600 neutroninduced dpa. Ion-induced void swelling of ferritic/martensitic alloy T91 in the quenched and tempered condition has been studied using a defocused, non-rastered 3.5 MeV Fe-ion beam at 475 C to produce damage levels up to 1000 peak displacements per atom (dpa). The high peak damage level of 1000 dpa is required to reach 500e600 dpa level due to injected interstitial suppression of void nucleation in the peak dpa region, requiring data extraction closer to the surface at lower dpa levels. At a relatively low peak damage level of 250 dpa, voids began to develop, appearing first in the near-surface region. With increasing ion fluence, swelling was observed deeper in the specimen, but remained completely suppressed in the back half of the ion range, even at 1000 peak dpa. The local differences in dpa rate in the front half of the ion range induce an “internal temperature shift” that strongly influences the onset of swelling, with shorter transient regimes resulting from lower dpa rates, in agreement not only with observations in neutron irradiation studies but also in various ion irradiations. Swelling was accompanied by radiation-induced precipitation of Cu-rich and Si, Ni, Mn-rich phases were observed by atom probe tomography, indicating concurrent microchemical evolution was in progress. In comparison to other ferritic/martensitic alloys during ion irradiation, T91 exhibits good swelling resistance with a swelling incubation period of about 400 local dpa.

Environmental Molecular Sciences Laboratory↗