Enhancement of Nanostructured Ferritic Alloy 14YWT Properties via Heat Treatment for Post-Consolidation Processing

C. J. Rietema, T. A. Saleh, D. T. Hoelzer, B. P. Eftink, E. Aydogan, K. D. Clarke, A. J. Clarke, S. A. Maloy

Research output: Contribution to journalArticlepeer-review

2 Scopus citations

Abstract

The nanostructured ferritic alloy 14YWT is a promising candidate for in-core use in generation IV nuclear reactors, due to a dense dispersion of insoluble, ultrafine-scale Y-Ti-O nano-oxides, which provide a high degree of irradiation tolerance and thermal stability. This study investigates the effects of heat treatment on the workability of 14YWT, along with the effect of processing history on abnormal grain structures and radial microstructural uniformity. In this study, a 14YWT rod consolidated at 850 °C was heat treated in argon at 1100 °C, 1150 °C, 1200 °C, or 1250 °C for 1 or 8 hours and changes in mechanical properties and microstructure were examined using microhardness and electron backscattered diffraction (EBSD). Two distinct types of large abnormal grains were observed, each with unique processing origins, including one with high and the other with low strain energy. Consolidation via direct extrusion resulted in radial microstructural gradients, where the center of the rod was softer with larger grain sizes and lower strain energies. These gradients persisted and intensified throughout heat treatment. Based upon this work, the recommended heat treatment for increased workability with minimal microstructural change is 1150 °C for 1 hours.

Original languageEnglish
Pages (from-to)2821-2829
Number of pages9
JournalMetallurgical and Materials Transactions A: Physical Metallurgy and Materials Science
Volume52
Issue number7
DOIs
StatePublished - Jul 2021

Funding

This work was supported by the US Department of Energy (DOE), Office of Nuclear Energy’s Fuel Cycle Research and Development (FCRD) program, Advanced Fuel Campaign. Portions of this work were conducted as a part of the US DOE—Japanese Atomic Energy Agency Civilian Nuclear Working Group (DOE-JAEA CNWG). AJC and KDC acknowledge support from the US DOE Office of Nuclear Energy’s Nuclear Energy University Programs (NEUP) under funding opportunity announcement DE-FOA-0001515 and the Advanced Steel Processing and Products Research Center (ASPPRC) at Colorado School of Mines during the preparation of this manuscript. The authors have no competing interests. This work was supported by the US Department of Energy (DOE), Office of Nuclear Energy’s Fuel Cycle Research and Development (FCRD) program, Advanced Fuel Campaign. Portions of this work were conducted as a part of the US DOE—Japanese Atomic Energy Agency Civilian Nuclear Working Group (DOE-JAEA CNWG). AJC and KDC acknowledge support from the US DOE Office of Nuclear Energy’s Nuclear Energy University Programs (NEUP) under funding opportunity announcement DE-FOA-0001515 and the Advanced Steel Processing and Products Research Center (ASPPRC) at Colorado School of Mines during the preparation of this manuscript.

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