Consequence Analysis of Residual Water in a Storage Canister (Preliminary Report)

Shukla, Prashant Kumar; Sindelar, Robert L.; Lam, Poh-Sang

doi:10.2172/1570349

Cited by 4 publications

(3 citation statements)

References 26 publications

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“…Define the volume increase as Δ = − 0 , then Note that, if the pellet has no void space ( = 0) and if UO2 is fully oxidized to U3O8 ( = 1), then Δ = − 1 = 0.36443 which is the conversion limit or the maximum net volume expansion of the pellet (36% by volume) [30][31][32].…”

Section: Calculation Of Total Volume and Volume Changementioning

confidence: 99%

Crack Stability in Breached Fuel

LAM¹,

SINDELAR²

2021

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Additional improvements for evaluation of flaw tolerance under pellet swelling loading are suggested. Additional characterization of the fracture toughness of High-Burnup (HBU) cladding is also suggested. This report fulfills the M4 milestone M4SF-21SR010203045, "Crack Opening in Breached Fuel" under Work Package Number SF-21SR01020304.1 The design initial pellet/clad gap of 0.107 mm is assumed to be closed through fuel burnup and pellet expansion therewith. The details for evaluation of pellet swelling with oxidation are provided in Appendix A to this present report.2 Above the critical crack length (crack instability length) determined by the fracture toughness KIC, crack propagation would occur unless there is crack arrest. Crack arrest would occur if crack loading conditions result in a fall of the applied stress intensity to KIa, the crack-arrest fracture toughness of the cladding.

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Section: Calculation Of Total Volume and Volume Changementioning

confidence: 99%

Crack Stability in Breached Fuel

LAM¹,

SINDELAR²

2021

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“…Residual water (post-drying and sealing) in a spent nuclear fuel (SNF) canister can result in corrosion and radiolysis that could affect the spent nuclear fuel and canister internal structural materials [CNWRA 2013, Shukla et al, 2019. Recent work [Bryan et al, 2019a] showed that inadvertent free water (~100 ml) may remain following drying even with a dryness criterion of 3 torr pressure limit following a 30-minute hold after active drying is complete a and measured ~500 ppmv of H2 after 12 days post-dryout in the High Burnup (HBU) Demonstration Cask.…”

Section: Introductionmentioning

confidence: 99%

Measurement of radiolytic hydrogen generation from surrogate zr-based materials in a "mini-canister"

D'Entremont

VERST

Sindelar

2022

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This report describes a test campaign to investigate radiolytic hydrogen generation from surrogate materials resembling the (Zr-based) cladding of commercial spent nuclear fuel (SNF) with an inventory of residual water post-dryout. The radiolytic hydrogen generation rates from residual waters in the SNF-incanister system are important for predicting the in-canister environment for sealed SNF dry storage canisters over time and the impact to continued safe dry storage.Under gamma radiation, radiolytic breakdown of water remaining in a canister (free, physisorbed, or chemisorbed) causes the generation of hydrogen gas (H2). SNF assemblies, clad in zirconium alloys, provide a large surface area to host surface-adsorbed water (physisorbed water). Other components in the cask, including stainless steel and aluminum structural components and aluminum neutron absorbers, also contribute to the total water inventory and radiolytic H2 generation in an SNF canister.Water adsorbed to the ZrO2 on the cladding surface may experience accelerated radiolysis compared to free water due to energy exchange with the oxide. In addition, surface water decomposed by radiolysis may be replenished by water vapor in the canister gas, providing a mechanism for accelerated radiolysis of the free water as well. The previous simplified analysis [d'Entremont et al., 2020] predicted that radiolysis of physisorbed water on the Zr-alloy components would dominate the short-term H2 generation due to these factors. Therefore, the yield associated with water radiolysis on the Zr surfaces (physisorbed water, with potential replenishment from water vapor) is the focus of the current testing.The lab-scale testing is conducted on non-radioactive surrogates, namely assemblies of zirconium coupons, irradiated inside a miniature steel canister ("mini-canister") designed to allow intermittent, insitu sampling of the canister gas during irradiation. The mini-canister approach for irradiation with in-situ monitoring was used in recent and on-going studies [Verst and d'Entremont, 2021] to measure radiolytic hydrogen generation from aluminum SNF cladding surrogate materials. This proposed testing forms part of the overall materials performance evaluation of the commercial SNF-in-canister system and is part of the technical bases for their continued safe dry storage.The test plan draws on the previous radiolysis testing on aluminum as well as the previous analysis, "Evaluation of Hydrogen Generation in High Burnup Demonstration Dry Storage Cask" [d'Entremont et al., 2020], which provided a best-estimate evaluation of residual water content (post-dryout) in the High Burnup (HBU) LWR Spent Fuel Demonstration project TN-32 cask and evaluated the expected radiolysis of the residual water, including free, physisorbed, and chemisorbed water in the sealed cask, based on literature data and models.The experiments focus on the contribution associated with the Zr-alloy components, i.e., surface-adsorbed water as well as interactions with free water vapor present in the gas p...

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“…The study considered residual water amount as high as 55 moles (CNWRA, 2013). A report under the SFWST program prepared in 2019 updated and refined the corrosion/oxidation models used in the NRC work (Shukla, et al, 2019). Both of those studies did not explicitly look at the corrosion of non-fuel canister internals.…”

Section: Introductionmentioning

confidence: 99%

Effects of Residual Water on Storage Canister Internal Components

Shukla¹,

Sindelar²

2020

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This report describes an evaluation of the spent nuclear fuel cladding and canister internal materials' corrosion/oxidation due to (inadvertent) residual free water inside a dry storage canister post-dryout. Drying of spent nuclear fuel (SNF) and the impact of (inadvertent) residual water in a SNF canister is being addressed under the DOE-NE Spent Fuel and Waste Science and Technology (SFWST) program to ensure the safe extended dry storage and transportation of commercial SNF. A long-standing understanding has been that the amount of residual water in a storage canister after vacuum drying is not expected to be much more than trace amounts of 0.43 gram mole (NRC, 2010). However, recent findings from the High Burnup Demonstration project (Bryan et al., 2019a-c), and an Integrated Research Project (Knight, 2018) show that residual free water well above the amount of approximately 0.43 gm-moles that had been assumed for a 3 torr rebound pressure, may remain within an SNF canister following prototypic drying. In fact, an NRC-NMSS sponsored study considered residual water amount as high as 55 moles (CNWRA, 2013). Considering the recent findings and the NRC-sponsored work, this present work was conducted to study the effects of unspecified amount of residual water on oxidation and corrosion of canister internals, i.e., it was assumed that the residual water amount is not limiting. It was assumed that the spent nuclear fuel content in the canister is not exposed to the internal environment, i.e., the packaged canister does not contain any damaged (breached) fuel. The overall approach consisted of explicit accounting of both spatial and temporal variations of thermal conditions in a generic storage canister in the corrosion/oxidation of the materials. Radiolysis of the residual water is expected to result in the formation of hydrogen peroxide, both in liquid and vapor phase of the residual water. Literature information suggests that vapor phase hydrogen peroxide would decompose into oxygen and water; this reaction will be catalyzed by the cladding surface. On other hand, hydrogen peroxide in liquid water would persist. It was assumed that radiolysis would yield approximately 0.1 mole of hydrogen peroxide for each liter of residual water or 0.1 M concentration. Considering this, it was considered for the corrosion evaluation that hydrogen peroxide concentration in liquid phase residual water ranges between 0 to 2 M. Corrosion rates of various internal component materials were obtained as a function of temperature and hydrogen peroxide. Cladding materials' oxidation rates, reported in the previous study (Shukla, et al, 2019), were used. Additional analysis also indicated that cladding oxidation could only occur when temperatures are high enough for the residual water to be in vapor phase. Cladding oxidation rates and corrosion rates of the internal components were integrated with the spatial and temporal thermal conditions inside the canister for the storage period of 300 years. The specific components were those for the f...

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Consequence Analysis of Residual Water in a Storage Canister (Preliminary Report)

Cited by 4 publications

References 26 publications

Crack Stability in Breached Fuel

Crack Stability in Breached Fuel

Measurement of radiolytic hydrogen generation from surrogate zr-based materials in a "mini-canister"

Effects of Residual Water on Storage Canister Internal Components

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