Report on fundamental modeling of irradiation-induced swelling and creep in FeCrAl alloys

Technical Report · Fri Sep 23 00:00:00 EDT 2016

DOI:https://doi.org/10.2172/1337855· OSTI ID:1337855

Kohnert, Aaron A. ^[1]; Dasgupta, Dwaipayan ^[2]; Wirth, Brian ^[1]; Linton, Kory D. ^[1]

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1. Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
2. Univ. of Tennessee, Knoxville, TN (United States)

In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromium-aluminum (FeCrAl) alloys due to much slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and allow the cladding to remain integral longer in the presence of high temperature steam, making accident mitigation more likely. As a continuation of the development for these alloys, the material response must be demonstrated to provide suitable radiation stability, in order to ensure that there will not be significant dimensional changes (e.g., swelling), as well as quantifying the radiation hardening and radiation creep behavior. In this report, we describe the use of cluster dynamics modeling to evaluate the defect physics and damage accumulation behavior of FeCrAl alloys subjected to neutron irradiation, with a particular focus on irradiation-induced swelling and defect fluxes to dislocations that are required to model irradiation creep behavior.

• Research Organization: Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
• Sponsoring Organization: USDOE Office of Nuclear Energy (NE)
• DOE Contract Number: AC05-00OR22725
• OSTI ID: 1337855
• Report Number(s): ORNL/TM-2016/569; AF5810000; NEAF278; TRN: US1701425
• Country of Publication: United States
• Language: English

Related Subjects

36 MATERIALS SCIENCE
CREEP
SWELLING
IRON BASE ALLOYS
CHROMIUM ALLOYS
ALUMINIUM ALLOYS
TERNARY ALLOY SYSTEMS
TEMPERATURE RANGE 0400-1000 K
FUEL CANS
WATER MODERATED REACTORS
PHYSICAL RADIATION EFFECTS
NEUTRONS
DISLOCATIONS
RADIATION HARDENING
COMPUTERIZED SIMULATION
WATER COOLED REACTORS
STABILITY
ACCIDENT-TOLERANT NUCLEAR FUELS