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Title: Lower Length Scale Characterization and Validation of Formation and Stability of Helium Bubbles in Nano-structured Ferritic Alloys under Irradiation

Abstract

In order to extend the operating license of current light water reactors (LWRs) in the United States and other countries to as many as 80 years or longer, it is demanding to identify potential materials for many of the internal structural components and fasteners. We proposed that 14YWT iron alloy can be adopted in such applications with its excellent material properties, such as high-temperature strength, low creep rate, and high irradiation resistance. Application with 14YWT would improve the void/helium swelling characteristics of the LWR fuels, extend the burn-up limits with the tolerant temperature up to 800oC and reduce the hydrogen production. One key feature of 14YWT material property enhancement is the ultrafine high density of 2-4nm Y-Ti-O enriched nanoclusters (NCs) within the 14YWT iron matrix. The NCs can effectively pin the ultra-fine grain boundaries and dislocations, which significantly enhance mechanical properties of the alloy. Moreover, these nanoclusters remain stable with no coarsening after a large dose of ion irradiation. After ion irradiation, the helium bubbles are observed extremely uniform in size (1nm) and quite homogeneously distributed within the 14YWT matrix, which indicates that the microstructure of 14YWT remains remarkably tolerance to radiation damage. However, there is a lack of understandingmore » of 14YWT both theoretically and experimentally in order to understand the mechanism behind the material property enhancement and to further develop and design a new generation of advanced structural material for current LWR applications and future fusion applications.« less

Authors:
; ;
Publication Date:
Research Org.:
Clemson University
Sponsoring Org.:
USDOE Office of Nuclear Energy (NE)
OSTI Identifier:
1419637
Report Number(s):
13-5408
13-5408
DOE Contract Number:
NE0000728
Resource Type:
Technical Report
Country of Publication:
United States
Language:
English

Citation Formats

Zhao, Huijuan, Yun, Di, and Hoelzer, David. Lower Length Scale Characterization and Validation of Formation and Stability of Helium Bubbles in Nano-structured Ferritic Alloys under Irradiation. United States: N. p., 2018. Web. doi:10.2172/1419637.
Zhao, Huijuan, Yun, Di, & Hoelzer, David. Lower Length Scale Characterization and Validation of Formation and Stability of Helium Bubbles in Nano-structured Ferritic Alloys under Irradiation. United States. doi:10.2172/1419637.
Zhao, Huijuan, Yun, Di, and Hoelzer, David. 2018. "Lower Length Scale Characterization and Validation of Formation and Stability of Helium Bubbles in Nano-structured Ferritic Alloys under Irradiation". United States. doi:10.2172/1419637. https://www.osti.gov/servlets/purl/1419637.
@article{osti_1419637,
title = {Lower Length Scale Characterization and Validation of Formation and Stability of Helium Bubbles in Nano-structured Ferritic Alloys under Irradiation},
author = {Zhao, Huijuan and Yun, Di and Hoelzer, David},
abstractNote = {In order to extend the operating license of current light water reactors (LWRs) in the United States and other countries to as many as 80 years or longer, it is demanding to identify potential materials for many of the internal structural components and fasteners. We proposed that 14YWT iron alloy can be adopted in such applications with its excellent material properties, such as high-temperature strength, low creep rate, and high irradiation resistance. Application with 14YWT would improve the void/helium swelling characteristics of the LWR fuels, extend the burn-up limits with the tolerant temperature up to 800oC and reduce the hydrogen production. One key feature of 14YWT material property enhancement is the ultrafine high density of 2-4nm Y-Ti-O enriched nanoclusters (NCs) within the 14YWT iron matrix. The NCs can effectively pin the ultra-fine grain boundaries and dislocations, which significantly enhance mechanical properties of the alloy. Moreover, these nanoclusters remain stable with no coarsening after a large dose of ion irradiation. After ion irradiation, the helium bubbles are observed extremely uniform in size (1nm) and quite homogeneously distributed within the 14YWT matrix, which indicates that the microstructure of 14YWT remains remarkably tolerance to radiation damage. However, there is a lack of understanding of 14YWT both theoretically and experimentally in order to understand the mechanism behind the material property enhancement and to further develop and design a new generation of advanced structural material for current LWR applications and future fusion applications.},
doi = {10.2172/1419637},
journal = {},
number = ,
volume = ,
place = {United States},
year = 2018,
month = 1
}

Technical Report:

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  • The authors report here a study of Kr ion implantation and the resultant bubble formation in Zr and Zr alloys, including Zircaloy-2 and Zircaloy-4. Implantations into thin foils were performed in the HVEM/Tandem facility at Argonne National Laboratory at temperatures between 300 to 800 C and to doses up to 2 x 10{sup 16} ion.cm{sup {minus}2}. Bulk specimens were implanted in an ion-beam chamber and then thinned for viewing by TEM. In thin foils, only small bubbles (30--100 {angstrom}) were formed at all temperatures with the exception of the Cr-rich V alloy where bubbles of 130 {angstrom} bubbles were formed.more » Bulk samples implanted at 300 C contained a bubble morphology similar to that observed after implantation into thin foils. However, at high temperatures (500--800 C) large faceted bubbles (up to 300 {angstrom}) were produced in bulk specimens. The results indicate that bubble formation and evolution below 500 C is controlled by gas concentration, while it is controlled by bubble mobility at high temperatures.« less
  • The alloys were three V-Ti binary alloys with Ti up to 21.0 at%, six V-Ti-Be ternary alloys and six V-Ti-Zr ternary alloys with Be and Zr up to 1.0 at.%. The alloys were preimplanted with helium to 80 appM, followed by Ni/sup 2 +/ ion irradiation to a damage dose of 25 dpa at 700/sup 0/C. TEM revealed the production of helium gas bubbles in all alloys. Neither the primary solute Ti nor the secondary solutes Be and Zr affected the bubble formation except the distribution mode. Two modes of distribution were observed: relatively uniform in mode I and highlymore » nonuniform in mode II. The particular distribution mode that predominated in a given alloy depended upon whether or not the matrix of the alloy was free of inclusion particles and second phase precipitates. The distribution in the V-5% Ti alloy was strictly in mode I because the alloy was relatively free of oxides, whereas in the other two binary alloys with higher Ti contents, thus containing more TiO inclusions, the mode varied from I to II in different regions. In all ternary alloys, especially those containing Zr, the distribution mode was predominantly II. The Ni/sup 2 +/ ion irradiations induced needle-like precipitates V/sub 2/Zr intermetallic with a preferential orientation in the <001> directions in the V-Ti-Zr alloys. The shape of bubbles was spherical in the as-irradiated state, but became cuboidal after a post-irradiation anneal at 1050/sup 0/C for 100 h.« less
  • A scientific question vitally important to the materials community is whether there exist self-assembled nanoclusters that are thermodynamically stable at elevated temperatures. Using in-situ neutron scattering, we characterized the structure and thermal stability of a nano-structured ferritic (NSF) alloy. Nanometer sized clusters were found to persist up to ~1400 C, providing direct evidence of a thermodynamically stable alloying state for the nanoclusters. Cluster formation requires the coexistence of Y, Ti, and O without the precipitation of oxide phases. The presence of thermally stable nanoclusters at grain boundaries limits the diffusion of Fe atoms, thereby stabilizing the microstructure of the ferriticmore » matrix at high temperatures. Our experimental results provide physical insights of the dramatically improved high-temperature mechanical properties in NSF alloy and point to a new direction in alloy design.« less
  • Ferritic/martensitic (FM) steels such as HT-9, T-91 and NF12 with chromium concentrations in the range of 9-12 at.% Cr and high Cr ferritic steels (oxide dispersion strengthened steels with 12-18% Cr) are receiving increasing attention for advanced nuclear applications, e.g. cladding and duct materials for sodium fast reactors, pressure vessels in Generation IV reactors and first wall structures in fusion reactors, thanks to their advantages over austenitic alloys. Predicting the behavior of these alloys under radiation is an essential step towards the use of these alloys. Several radiation-induced phenomena need to be taken into account, including phase separation, solute clustering,more » and radiation-induced segregation or depletion (RIS) to point defect sinks. RIS at grain boundaries has raised significant interest because of its role in irradiation assisted stress corrosion cracking (IASCC) and corrosion of structural materials. Numerous observations of RIS have been reported on austenitic stainless steels where it is generally found that Cr depletes at grain boundaries, consistently with Cr atoms being oversized in the fcc Fe matrix. While FM and ferritic steels are also subject to RIS at grain boundaries, unlike austenitic steels, the behavior of Cr is less clear with significant scatter and no clear dependency on irradiation condition or alloy type. In addition to the lack of conclusive experimental evidence regarding RIS in F-M alloys, there have been relatively few efforts at modeling RIS behavior in these alloys. The need for predictability of materials behavior and mitigation routes for IASCC requires elucidating the origin of the variable Cr behavior. A systematic detailed high-resolution structural and chemical characterization approach was applied to ion-implanted and neutron-irradiated model Fe-Cr alloys containing from 3 to 18 at.% Cr. Atom probe tomography analyses of the microstructures revealed slight Cr clustering and segregation to dislocations and grain boundaries in the ion-irradiated alloys. More significant segregation was observed in the neutron irradiated alloys. For the more concentrated alloys, irradiation did not affect existing Cr segregation to grain boundaries and segregation to dislocation loops was not observed perhaps due to a change in the dislocation loop structure with increasing Cr concentration. Precipitation of α’ was observed in the neutron irradiated alloys containing over 9 at.% Cr. However ion irradiation appears to suppress the precipitation process. Initial low dose ion irradiation experiments strongly suggest a cascade recoil effect. The systematic analysis of grain boundary orientation on Cr segregation was significantly challenged by carbon contamination during ion irradiation or by existing carbon and therefore carbide formation at grain boundaries (sensitization). The combination of the proposed systematic experimental approach with atomistic modeling of diffusion processes directly addresses the programmatic need for developing and benchmarking predictive models for material degradation taking into account atomistic kinetics parameters« less