Gary Was

Professor

gsw@umich.edu

1921 Cooley

T: (734) 763-4675

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Design of Radiation-Tolerant Alloys for Generation IV

Collaborators: S. Bruemmer, PNNL, and T. Allen, ANL
Sponsor: U.S. Department of Energy, Nuclear Energy Research Initiative (NERI)
Under the Generation IV Reactor initiative, revolutionary improvements in nuclear energy system design in the areas of sustainability, economics, and safety and reliability are being pursued. To meet these goals, advanced nuclear energy systems demand materials that minimize resource use, minimize waste impact, improve proliferation resistance, extend component lifetime, and reduce uncertainty in component performance, all while potentially operating in higher temperature environments, to greater radiation dose, and in unique corrosion environments compared to previous generations of nuclear energy systems. The irradiation performance of structural materials will likely be the limiting factor in successful nuclear energy system development. Based on experience, materials not tailored for irradiation performance generally experience profound changes in virtually all important engineering and physical properties because of fundamental changes in structure caused by radiation damage.This project will develop and characterize the radiation performance of materials with improved radiation resistance. Material classes will be chosen that are expected to be critical in multiple Generation IV technologies. The material design strategies to be tested fall into three main categories: (1) alloying, by adding oversized elements to the matrix; (2) engineering grain boundaries; and (3) microstructural/nanostructural design, such as adding matrix precipitates.The materials to be examined include both austenitic and ferritic-martensitic steels, both classes of which are expected to be key structural materials in many Generation IV concepts. The irradiation program will consist of scoping studies using proton and heavy-ion irradiations of key alloys and tailored alloy condition and examination of materials irradiated in BOR-60 to confirm charged particle results. Examinations will include microstructural characterization, mechanical properties evaluation using hardness and shear punch, and stress corrosion cracking.
Highlights (Click an image for more information)
  • Reduction on radiation induced segregation with addition of oversize solutes

    The use of oversized solute additions in 316 stainless steel has previously demonstrated an ability to reduce the effects of radiation-induced segregation (RIS). RIS has been implicated as one of several factors in enhancing stress corrosion cracking (SCC) under irradiation, so the use of oversized solutes could help to promote SCC resistance. Zr or Hf are added to 316 steels and then irradiated with 3 MeV protons at temperatures of 400 and 500°C to doses between 3 10 dpa. Samples are analyzed in HR-STEM to measure the grain boundary RIS for each dose and temperature. Zr additions are shown to reduce the amount of RIS at low doses, but the effect disappears by higher doses, while Hf appears to have only a small effect on RIS even at low dose. Resistance to RIS is explained through the use of kinetic rate-theory modeling which highlights the effectiveness of a vacancy trapping mechanism to reduce RIS. The loss of RIS resistance is explained by the loss of oversized solute in the matrix, which helps to explain the transient behavior of the oversized solute additions on reducing RIS. Combined with radiation hardening measurements, dislocation loop imaging, x-ray diffraction measurements and atom probe spectroscopy, the oversized solutes are shown to form precipitates and lose their ability with irradiation dose to trap point defects and enhance recombination necessary to eliminate RIS and other radiation damage effects. Nevertheless, the binding energy of the oversized solutes with vacancy defects is shown both computationally and experimentally to be sufficiently large to demonstrate a measurable change in microstructure morphology under irradiation in austenitic stainless steels.

  • Improvement in creep properties of Grain Boundary Engineered T91

    Gaurav Gupta, a grad student in the group has been working on this the improvement of creep of T91 by increasing the CSL fraction in the alloy. Gaurav has shown that creep properties of F-M alloy T91 can be improved by grain boundary engineering (GBE). The attached figures shows fraction of low angle boundaries in as-received (AR) and Coincident Site Lattice-Enhanced (CSLE) T91 and the comparitive results of creep experiments in argon on AR T91 and CSLE T91.