Nanostructured Engineering Alloys for Nuclear Application
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Nanostructured Engineering Alloys for Nuclear Application Peter Hosemann1, Erich Stergar1, Andrew T. Nelson2, C. Vieh3, Stuart A. Maloy2, 1
Nuclear Engineering, University of California Berkeley, Berkeley, California; Material Science and Engineering, Los Alamos National Laboratory, Los Alamos, New Mexico. 3 Paul Scherrer Institute, Villigen, Switzerland 2
ABSTRACT In advanced nuclear applications, high temperature and a corrosive environment are present in addition to a high dose radiation field causing displacement damage in the material. In recent times it has been shown that Nanostructured Ferritic Alloys (NFA’s) such as advanced Oxide Dispersion Strengthened (ODS) steels are suitable for this environment as they tolerate high dose irradiation without significant changes in microstructure or relevant mechanical properties. Ion beam irradiation is a fast and cost effective way to induce radiation damage in materials but has limited penetration depth. Therefore, small scale mechanical testing such as nanoindentation and micro compression testing in combination with FIB based sample preparation for micro structural characterization has to be performed allowing a full assessment of the materials’ behavior under radiation environment. In this work two different ODS materials have been irradiated using proton and combined proton and He beams up to 1 dpa at different temperatures. Nanoindentation and LEAP measurements were performed in order to assess the changes in properties of these alloys due to irradiation. The same techniques were applied to intermetallic nanostructured alloys in order to investigate the effectiveness of the metal-intermetallic interface to provide defect sinks for He and radiation damage. It was found that irradiation can cause the formation of intermetallic particles even at room temperature while increasing the material strength significantly.
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INTRODUCTION Structural materials for nuclear applications suffer a wide range of basic property changes such as embrittlement, swelling etc. in an irradiation environment. The reason for the changes in properties can be due to the formation of gas in the material (n-alpha reaction) and/or to the creation of a large number of point defects which further evolve into larger defect clusters leading to a change in microstructure, dislocation density, void formation, or phase composition. The formation of cascades due to particle irradiation and the formation of a large number of point defects cannot be avoided initially. However, the recombination rate of the created defects can be enhanced by choosing materials with a small net bias and or by enhancing the recombination of defects by trapping. Recent studies have shown that the formation of large defects can be delayed or avoided by trapping the initial defects at a large number of defect sinks where the defects then annihilate, preventing their accumulation within the lattice [1, 2]. The most effective defect sinks and traps have been proven to be interfaces such as grain boundaries or phase bound
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