Radiation Effects in Innovative Structural Materials
- University of Tennessee, Knoxville, TN, 37966 (United States)
- Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States)
The extreme radiation environments and demanding structural conditions of next-generation fission and proposed fusion reactors will require high performance materials with very high radiation resistance. Although conventional structural materials such as ferritic/ martensitic steels offer several attractive features, there is value in exploring alternative materials systems that may offer unprecedented radiation stability while also offering good thermomechanical properties. As noted elsewhere, options for designing radiation resistance include (a) the traditional approach of introducing a very high sink strength for recombination of radiation-induced vacancies and interstitials (e.g., Ti-modified stainless steels or 'nano-cluster-strengthened ferritic alloys', (b) utilizing materials with immobile vacancies at the operating temperature (e.g., silicon carbide), and (c) utilizing certain classes of materials that inherently exhibit superior resistance to radiation defect retention and accumulation (e.g., ferritic steels). Three relatively new but promising structural materials systems were selected for investigation, based on their favorable unirradiated mechanical properties and potential for good stability under irradiation: High entropy alloys (HEAs), bulk metallic glass (BMG), and MAX phase ceramics. The high configurational entropy in HEAs might alter point defect (solute and radiation defects) diffusion processes, thus potentially enhancing the recombination of radiation-produced damage. Therefore, HEAs offer the potential of a new class of materials for radiation resistance category (c) mentioned previously. A variety of BMGs such as Zr{sub 52.5}Cu{sub 17.9}Ni{sub 14.6}Al{sub 10}Ti{sub 5} offer good mechanical properties including fracture toughness values that may exceed 100 MPa-m{sup 1/2} and near net shape fabrication potential. The lack of crystallinity in BMGs prohibits the formation of conventional Frenkel pair defects under irradiation and therefore offer the possibility of another new material system for radiation resistance category (c). MAX phase ceramics (e.g., Ti{sub 3}SiC{sub 2}, Ti{sub 3}AlC{sub 2}, Ti{sub 2}AlC) have good mechanical properties and contain a nano-laminate structure that could provide a very high point defect sink strength. Due to their refractory nature, many of the MAX phase ceramics may exhibit negligible vacancy mobility up to high temperatures. Therefore, MAX phase ceramics offer the potential to provide radiation resistance in either categories (a) and (b), depending on the operating temperature. For this study, a novel 27%Fe-27%Mn-28%Ni-18%Cr single phase face centered cubic high entropy alloy was synthesized that exhibited good strength and ductility over a wide temperature range. Zr{sub 52.5}Cu{sub 17.9}Ni{sub 14.6}Al{sub 10}Ti{sub 5} (BAM-11) was selected as a prototypic advanced BMG. Several MAX phase ceramics (e.g., Ti{sub 3}SiC{sub 2}, Ti{sub 3}AlC{sub 2}, Ti{sub 2}AlC) that have good unirradiated thermal and mechanical properties were also investigated. The irradiation damage behaviors of these three novel materials systems have been studied after fission neutron at ∼75 deg. C up to 1 dpa and Ni ion irradiation at 25- 700 deg. C up to 30 dpa (lower temperatures and doses for the BMGs). All three sets of materials generally exhibited good radiation stability within certain temperature regimes. For example, the BMG exhibited only minor (<10%) change in strength that appeared to saturate after a dose of ∼1 dpa. A key shortcoming of current BMGs are that they can not be used at temperatures >>300 deg. C due to glass to crystalline transition at elevated temperature. The MAX phase ceramics irradiated at 400-700 deg. C appeared to be in the 'low temperature' regime where vacancies are immobile and therefore the radiation hardening saturated at relatively low doses. The Ti{sub 3}SiC{sub 2} MAX phase exhibited the best overall radiation stability over a wide temperature range. The HEA exhibited fine scale dislocation loop formation, no observable cavity swelling and reduced amounts of Cr depletion compared to traditional austenitic Fe-Cr-Ni alloys after ion irradiation, which suggests suppressed point defect and solute diffusion behavior. (authors)
- OSTI ID:
- 22992142
- Journal Information:
- Transactions of the American Nuclear Society, Vol. 114, Issue 1; Conference: Annual Meeting of the American Nuclear Society. Embedded topical meeting 'Nuclear fuels and structural material for the next generation nuclear reactors', New Orleans, LA (United States), 12-16 Jun 2016; Other Information: Country of input: France; 1 ref.; Available from American Nuclear Society - ANS, 555 North Kensington Avenue, La Grange Park, IL 60526 United States; ISSN 0003-018X
- Country of Publication:
- United States
- Language:
- English
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