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Chancey, Matthew

Publications and source records attributed to Chancey, Matthew.

6 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Microstructurally resolved electrochemical evolution of mechanical- and irradiation-induced damage in nuclear alloys

Abstract There is a need for high-throughput, scale-relevant, and direct electrochemical analysis to understand the corrosion behavior and sensitivity of nuclear materials that are exposed to extreme (high pressure, temperature, and radiation exposure) environments. We demonstrate the multi-scale, multi-modal application of scanning electrochemical cell microscopy (SECCM) to electrochemically profile corrosion alterations in nuclear alloys in a microstructurally resolved manner. Particularly, we identify that both mechanically deformed and irradiated microstructures show reduced charge-transfer resistance that leads to accelerated oxidation. We highlight that the effects of mechanical deformation and irradiation are synergistic, and may in fact, superimpose each other, with implications including general-, galvanic-, and/or irradiation-activated stress-corrosion cracking. Taken together, we highlight the ability of non-destructive, electrochemical interrogations to ascertain how microstructural alterations result in changes in the corrosion tendency of a nuclear alloy: knowledge which has implications to rank, qualify and examine alloys for use in nuclear construction applications.

Chen, Xin↗
DOE OSTI2022

The mechanism behind the high radiation tolerance of Fe-Cr alloys

With the great demand for high radiation tolerant materials for advanced nuclear energy technologies, Fe-Cr alloys are at the forefront with long standing validated performance. Yet, the real mechanism behind their high radiation resistance is in question and understanding the effect of varying Cr percentage is a grand challenge limiting further improvements. Here we applied depth resolved atomic scale probe of defects to uncover the real mechanism on how Cr improves radiation resistance and explain the controversial impact of increasing Cr percentage. By combining depth-resolved positron annihilation lifetime spectroscopy and Doppler broadening spectroscopy we investigated the effect of Cr alloying on the formation and evolution of atomic size clusters induced by ion irradiation in Fe. We also used atom probe tomography to investigate the possible presence of Cr clusters or a’ phase with high Cr composition. The study reveals that the well-known resistance to radiation in Fe-Cr alloys arises from the stabilization of vacancy clusters around Cr atoms which act as sinks for radiation-induced defects. Thus, Cr atoms do not provide a direct sink for interstitials; rather defect complexes for that consist of Cr atoms and vacancies in turn act as sinks for irradiation-induced vacancies and interstitials. Most importantly, we find that lower amounts of Cr create smaller, uniformly distributed defect clusters that act as efficient sinks for radiation damage, but larger quantities of Cr form a defect structure that is less homogenous in size and spatial distribution, resulting in less efficient damage recombination. No evidence of phase a’ was found before or after irradiation, which indicates that it does not contribute to the observed radiation tolerance.

36 MATERIALS SCIENCE↗
DOE OSTI2020

An exploratory study on helium mobility in amorphous and crystallized bulk metallic glasses

In this study, Zr 52.5 Cu 17.9 Ni 14.6 Al 10 Ti 5 and Cu 60 Zr 20 Hf 10 Ti 10 amorphous and crystallized bulk metallic glasses (BMGs) were implanted at room temperature by 150 keV 3 He + ions to fluences of 2 × 10 15 cm -2 and 5 × 10 15 cm -2 , producing peak implanted He concentrations of ~1,100 and 3,500 appm, respectively. Nuclear reaction analysis (NRA) performed on post-implantation annealed samples at various temperatures between 250 and 600 °C (~0.30 to 0.75 of the absolute melting temperature T m of the BMGs) revealed no appreciable He migration. Complementary thermal desorption spectroscopy (TDS) on as-implanted samples, which involved in situ annealing, showed appreciable He release only at temperatures higher than ~700 °C (~0.76-0.90 T m ) in all the samples. Half or more of the implanted He did not desorb up to the maximum investigated TDS annealing temperature of 770 °C (~0.85-0.97 T m ). The lack of pronounced He diffusion up to 600 °C as observed from the NRA and TDS data was attributed to He trapping in free volume sites present in the BMGs, which seems to play a similar role to vacancies in crystalline materials. It was also observed that in both the amorphous and crystallized forms of the Zr BMG, the sample implanted to the lower fluence released more He as compared to the sample implanted with a higher fluence. The crystalline forms of both the Cu and Zr BMGs released more He as compared to their amorphous counterparts.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Nitrogen effects on radiation response in 12Cr ferritic/martensitic alloys

Ferritic/martensitic steels are one of the best candidates for structural materials for high dose applications in next generation nuclear reactors. The composition of structural materials must be optimized for reliable service during irradiation. This study reports the effect of interstitial nitrogen on radiation response in 12Cr ferritic/martensitic HT9 steels having a controlled nitrogen concentration. Results show that a high amount of ‘free’ nitrogen in the matrix stabilizes the interstitial clusters which leads to (i) larger loop sizes (ii) lower loop density and (iii) slightly reduced radiation induced hardening. It also affects diffusion mechanism of Ni and formation of Ni/Si-rich precipitates.

36 MATERIALS SCIENCE↗