Corrosion of aluminum clad spent nuclear fuel in the 70 ton cask during transfer from L area to H-canyon
Aluminum-clad spent nuclear fuel will be transported for processing in the 70-ton nuclear fuel element cask from L Basin to H-canyon. During transport these fuels would be expected to experience high temperature aqueous corrosion from the residual L Basin water that will be present in the cask. Cladding corrosion losses during transport were calculated for material test reactor (MTR) and high flux isotope reactors (HFIR) fuels using literature and site information on aqueous corrosion at a range of time/temperature conditions. Calculations of the cladding corrosion loss were based on Arrhenius relationships developed for aluminum alloys typical of cladding material with the primary assumption that an adherent passive film does not form to retard the initial corrosion rate. For MTR fuels a cladding thickness loss of 33% was found after 1 year in the cask with a maximum temperature of 263 °C. HFIR fuels showed a thickness loss of only 6% after 1 year at a maximum temperature of 180 °C. These losses are not expected to impact the overall confinement function of the aluminum cladding.
- Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)
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- Resource Type:
- Technical Report
- Research Org:
- Savannah River Site (SRS), Aiken, SC (United States)
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- Country of Publication:
- United States
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