Hot Cell Tensile Testing of Neutron Irradiated Additively Manufactured Type 316L Stainless Steel

Abstract

Westinghouse began additively manufactured (AM) materials research activities in 2012 to support development of advanced fuel related components. The initial objective of this work has been the fabrication and delivery of a lead test component to a Westinghouse nuclear utility customer for in-reactor insertion for a limited number of fuel cycles. It is generally recognized that a key criterion for the implementation of AM components would be a thorough understanding of the material response to neutron irradiation. Several alloys were AM fabricated, heat treated, neutron irradiated, and extensively evaluated both in the as-printed condition and in the printed and irradiated condition. Irradiation of AM miniature tensile specimens was performed at the Massachusetts Institute of Technology (MIT) Nuclear Research Reactor to 0.8 dpa at 300 °C (572 °F). Although extensive laboratory testing was performed on these materials, this paper specifically summarizes the results from room and elevated temperature tensile testing of unirradiated and irradiated AM 316L. Testing of the miniature tensile specimens was performed inside a hot cell utilizing in-cell digital image correlation (DIC) and advanced video extensometry (AVE). Additional AM alloys are currently being irradiated in the MIT reactor to higher dpa values. The first of these samples will be shipped and subsequently tested in 2017.