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Assessment of Aging Mechanisms for Zirconium-Based High Burnup Fuel Cladding in Dry Storage Systems

Picture for Assessment of Aging Mechanisms for Zirconium-Based High Burnup Fuel Cladding in Dry Storage Systems
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This paper provides detailed technical bases for defining whether a variety of aging mechanisms are credible for high burnup cladding alloys in dry storage beyond 20 years, and up to 60 years.

 

Product Number: 51317--8939-SG
ISBN: 8939 2017 CP
Author: Pavan Shukla
Publication Date: 2017
Industry: Energy Generation
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$20.00

At independent spent fuel storage installations (ISFSIs) in the United States spent nuclear fuel (SNF) is stored in welded- or bolted-sealed dry storage systems (DSS). The safety analyses for the DSS rely on the fuel assemblies having a known configuration during the period of operation (e.g. geometric form cladding integrity). Sufficient experimental data has provided adequate confirmation that low burnup SNF (?45 GWd/MTU) will remain in its analyzed configuration during dry storage without deleterious effect on the cladding. This work expands to define credible degradation mechanisms for high burnup fuel cladding to support the technical bases for dry storage beyond 20 years. For burnups exceeding 45 GWd/MTU the fuel pellets and the zirconium-based cladding undergo distinct changes during reactor service including increased hydrogen absorption by the cladding swelling of the fuel pellets and increased cladding stresses due to helium and fission gas release and pellet-pellet and pellet-cladding interactions. The storage environment considered in the assessment was helium or an alternative inert cover gas in high radiation and temperature (both of which decrease over time). The expected amount of residual water (about 0.43 gram-mole) was also considered to be retained inside the cask/canister during fuel loading operations. The aging mechanisms considered for zirconium-based cladding include hydride-induced embrittlement delayed hydride cracking thermal and athermal (low-temperature) creep localized mechanical overload radiation embrittlement fatigue oxidation pitting corrosion galvanic corrosion and stress corrosion cracking. This paper provides detailed technical bases for defining if these aging mechanisms are credible for high burnup cladding alloys in dry storage beyond 20 years.This abstract is an independent product of the CNWRA and does not necessarily reflect the view or regulatory position of the NRC. The NRC staff views expressed herein are preliminary and do not constitute a final judgment or determination of the matters addressed or of the acceptability of any licensing action that may be under consideration at the NRC.Key words: zirconium-based alloys high burnup spent nuclear fuel and dry storage.

Key words: zirconium-based alloys, high burnup, spent nuclear fuel, and dry storage

At independent spent fuel storage installations (ISFSIs) in the United States spent nuclear fuel (SNF) is stored in welded- or bolted-sealed dry storage systems (DSS). The safety analyses for the DSS rely on the fuel assemblies having a known configuration during the period of operation (e.g. geometric form cladding integrity). Sufficient experimental data has provided adequate confirmation that low burnup SNF (?45 GWd/MTU) will remain in its analyzed configuration during dry storage without deleterious effect on the cladding. This work expands to define credible degradation mechanisms for high burnup fuel cladding to support the technical bases for dry storage beyond 20 years. For burnups exceeding 45 GWd/MTU the fuel pellets and the zirconium-based cladding undergo distinct changes during reactor service including increased hydrogen absorption by the cladding swelling of the fuel pellets and increased cladding stresses due to helium and fission gas release and pellet-pellet and pellet-cladding interactions. The storage environment considered in the assessment was helium or an alternative inert cover gas in high radiation and temperature (both of which decrease over time). The expected amount of residual water (about 0.43 gram-mole) was also considered to be retained inside the cask/canister during fuel loading operations. The aging mechanisms considered for zirconium-based cladding include hydride-induced embrittlement delayed hydride cracking thermal and athermal (low-temperature) creep localized mechanical overload radiation embrittlement fatigue oxidation pitting corrosion galvanic corrosion and stress corrosion cracking. This paper provides detailed technical bases for defining if these aging mechanisms are credible for high burnup cladding alloys in dry storage beyond 20 years.This abstract is an independent product of the CNWRA and does not necessarily reflect the view or regulatory position of the NRC. The NRC staff views expressed herein are preliminary and do not constitute a final judgment or determination of the matters addressed or of the acceptability of any licensing action that may be under consideration at the NRC.Key words: zirconium-based alloys high burnup spent nuclear fuel and dry storage.

Key words: zirconium-based alloys, high burnup, spent nuclear fuel, and dry storage

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