The Irradiated Microstructure of Ferritic-Martensitic Steel T91 and 9Cr-ODS
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The Irradiated Microstructure of Ferritic-Martensitic Steel T91 and 9Cr-ODS J. Gan1, T. R. Allen2, J. I. Cole1, S. Ukai3, S. Shutthanandan4 and S. Thevuthasan4 1 Nuclear Technology Division, Argonne National Laboratory, USA 2 Department of Engineering Physics, The University of Wisconsin, USA 3 Oarai Engineering Center, Japan Nuclear Cycle Development Institute, Japan 4 Enviromental Molecular Sciences Laboratory, Pacific Northwest National Laboratory, USA ABSTRACT A ferritic steel T91 and an oxide dispersion strengthened (ODS) martensitic steel 9Cr-ODS were irradiated with 5 MeV Ni ions at 500ºC at a dose rate of 1.38x10-3 dpa/s to doses of 5, 50 and 150 dpa. Both alloys are iron-based with 9Cr and have been designed for use in higher temperature energy systems. However, the radiation effects on these two alloys are not well characterized. For T91, the irradiated microstructure was dominated by tangled dislocation and precipitates, similar to the unirradiated condition except the presence of large dislocation loops of type a. The microstructure of alloy 9Cr-ODS for both the unirradiated and irradiated cases was dominated by dense dislocations, precipitates and yttrium oxides particles and no dislocation loops were observed. The average size of yttrium oxides particles slightly decreased with dose from 11.8 nm for the unirradiated to 9.1 nm at 150 dpa. No voids were detected for both alloys up to a dose of 150 dpa.
INTRODUCTION Advanced nuclear energy systems proposed under the Generation IV initiative are aimed at making revolutionary improvements in economics, safety and reliability, and sustainability. To achieve these advancements, Generation IV systems will operate at much higher temperatures and in higher radiation fields than current reactors. Metallic alloy components will experience unprecedented microstructural and mechanical property evolution as they progress to higher doses. Of the candidate alloy systems that could be considered for improved radiation tolerance, ferritic-martensitic alloy systems are expected to play an important role as structural components in Generation IV systems that operate in the temperature range 350-700°C and to doses up to 200 dpa. Ferritic-martensitic steels are expected to be used as high dose components in sodium and lead reactors, low temperature components in gas-cooled reactors, and possibly components in supercritical water reactors. Ferritic-martensitic steels offer better swelling resistance but may suffer from grain boundary and/or matrix creep and loss of strength at temperatures above ~600°C and unacceptably low toughness at lower temperatures. However, the growing body of data on ferritic-martensitic steels, combined with specific, tailored microstructure modifications may be able to address the deficiencies at high temperature/dose. Heavy-ion irradiation provides a unique approach in this exploratory task to evaluate material tolerance to radiation up to very high doses. Alloy T91 (9Cr-1Mo) is a ferritic/martensitic steel developed for high temperature a
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