Fuel Performance Simulation of Iron-Chrome-Aluminum Cladding during Steady-State LWR Operation
- University of Tennessee, Knoxville, TN 37916 (United States)
During a severe accident scenario, such as a loss-of-coolant accident, a nuclear reactor may lose its capacity to cool its fuel. The fuel rod temperatures begin to increases, quickly reaching temperatures where the coolant begins to evaporate. This lowers the coolant level in the core and eventually uncovers the fuel. Without sufficient heat transfer from the fuel rods to the coolant, the temperature of the fuel and cladding will increase dramatically. In the case of traditional zirconium-based alloys, as the temperature of the cladding reaches ∼1200 deg. C the zirconium rapidly begins to oxidize with H{sub 2}O in the coolant and steam. This oxidation reaction causes both thinning of the Zr-alloy cladding as the metal reacts and releases large amounts of H{sub 2} gas into the reactor pressure vessel. In order to increase the safety margin of LWR fuel in severe accident scenarios, several alternative cladding materials have been proposed to replace the currently used zirconium-based alloys. Of these materials, there is a particular focus on select iron-chrome-aluminum (FeCrAl) alloys because of slower oxidation Kinetics in high temperature steam than zirconium-based alloys. This increased oxidation resistance may provide additional coping time to mitigate any further damage resulting from an accident. However, by changing the cladding material, the neutronic characteristics of the core will change. FeCrAl alloys exhibit thermal neutron absorption cross sections nearly ∼12-16 x greater than Zr-alloys. This decreases the reactivity of the core, subsequently decreasing the potential cycle length for operation. To counteract this, fissile material may be added in the form of increased enrichment or increased fuel mass or the cladding thickness might be decreased, effectively reducing the neutronic penalty from using the alloy. In this study, all three changes are considered. To assess the viability of FeCrAl as an alternative cladding to Zr-alloys, engineering-scale fuel performance analysis is used to monitor the cladding and integral fuel rod behavior through its operation. Ongoing work is being performed to identify key parameters essential to modeling FeCrAl alloys for LWR applications and properly model fuel compliance and stress relief mechanisms for UO{sub 2} fuel. To improve the predictive capability of fuel performance codes for FeCrAl cladding under normal operation, high-fidelity mechanistic/physics-based material models are needed. This information can be generated and improved by access to experimental data on targeted FeCrAl alloys. (authors)
- OSTI ID:
- 22992124
- 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; 8 refs.; 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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