Spent nuclear fuel integrity during dry storage - performance tests and demonstrations

McKinnon,; Doherty, A.L.

doi:10.2172/527885

Cited by 2 publications

(7 citation statements)

References 7 publications

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Section: A Storage Conditionsmentioning

confidence: 99%

“…During the following cask performance tests [2][3][4], the fuel rods were in He, N 2 , and vacuum atmospheres in horizontal and vertical configurations. The peak cladding temperature adjusted for the position of the assembly in the cask varied in each configuration [3] as indicated in Table I. It should be noted that the estimated cladding temperatures during the initial 2160-h test period in the Castor-V/21 cask ranged from ~25 to 100°C above a PWR cladding temperature during normal reactor operation.…”

Section: A Storage Conditionsmentioning

confidence: 99%

“…It should be noted that the estimated cladding temperatures during the initial 2160-h test period in the Castor-V/21 cask ranged from ~25 to 100°C above a PWR cladding temperature during normal reactor operation. At the completion of the cask performance tests, the fuel spent approximately 2880 h vertically in a 70% He-30% air atmosphere [3]. Since that time in March 1986, the fuel has been stored vertically in a He/< 1% air atmosphere.…”

Section: A Storage Conditionsmentioning

confidence: 99%

“…Cover gas samples were taken periodically throughout the performance-testing period and storage duration to look for cask leaks and fuel rod breaches [3]. In all cases, the intended atmospheres indicated in Table I were maintained.…”

Section: A Storage Conditionsmentioning

confidence: 99%

“…No leaks were found. The details of these tests have been reported in a number of documents [2][3][4]. Subsequently, the cask sat on the storage pad at the INEEL for ≈15 years with the fuel in an inert atmosphere (helium).…”

Section: Introductionmentioning

confidence: 99%

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Examination of Spent PWR Fuel Rods After 15 Years in Dry Storage

Einziger

Tsai

Billone

et al. 2002

10th International Conference on Nuclear Engineering, Volume 4

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b Argonne-West, Idaho Falls, ID, USA Virginia Power Surry Nuclear Station Pressurized Water Reactor (PWR) fuel was stored in a dry inert atmosphere Castor V/21 cask at the Idaho National Environmental and Engineering Laboratory (INEEL) for 15 years at peak cladding temperatures decreasing from about 350 to 150°C. Prior to the storage, the loaded cask was subjected to extensive thermal benchmark tests. The cask was opened to examine the fuel for degradation and to determine if it was suitable for extended storage. No rod breaches had occurred and no visible degradation or crud/oxide spallation were observed.Twelve rods were removed from the center of the T11 assembly and shipped from INEEL to the Argonne-West HFEF for profilometric scans. Four of these rods were punctured to determine the fission gas release from the fuel matrix and internal pressure in the rods. Three of the four rods were cut into five segments each, then shipped to the Argonne-East AGHCF for detailed examination. The test plan calls for metallographic examination of six samples from two of the rods, microhardness and hydrogen content measurements at or near the six metallographic sample locations, tensile testing of six samples from the two rods, and thermal creep testing of eight samples from the two rods to determine the extent of residual creep life. The results from the profilometry (12 rods), gas release measurements (4 rods), metallographic examinations (2 samples from 1 rod), and microhardness and hydrogen content characterization (2 samples from 1 rod) are reported here. The tensile and creep studies are just starting and will be reported at a later date, along with the additional characterization work to be performed.Although only limited prestorage characterization is available, a number of preliminary conclusions can be drawn based on comparison with characterization of Florida Power Turkey Point rods of a similar vintage. Based on this comparison, it appears that little or no cladding thermal creep and fission gas release from the fuel pellets occurred during the thermal benchmark tests or storage. Measurements of the cladding outer-diameter, oxide thickness and wall thickness are in the expected range for cladding of the Surry exposure. The measured hydrogen content is consistent with the oxide thickness. The volume of hydrides varies azimuthally around the cladding, but there is little variation across the thickness, of the cladding. It is most significant that all of the hydrides appear to have retained the circumferential orientation typical of prestorage PWR fuel rods.

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Section: A Storage Conditionsmentioning

confidence: 99%

Section: A Storage Conditionsmentioning

confidence: 99%

Section: A Storage Conditionsmentioning

confidence: 99%

Section: A Storage Conditionsmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

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Examination of Spent PWR Fuel Rods After 15 Years in Dry Storage

Einziger

Tsai

Billone

et al. 2002

10th International Conference on Nuclear Engineering, Volume 4

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Radiolysis Model Sensitivity Analysis for a Used Fuel Storage Canister

Wittman¹

2013

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! #' ()* ()+, ! #' ()* ()+, 1 2This report fulfills the M3 milestone (M3FT+13PN0810027) to report on a radiolysis computer model analysis that estimates the generation of radiolytic products for a storage canister. The analysis considers radiolysis outside storage canister walls and within the canister fill gas over a possible 300+year lifetime. Previous work relied on estimates based directly on a water radiolysis G+value. a This work also includes that effect with the addition of coupled kinetics for 111 reactions for 40 gas species to account for radiolytic+induced chemistry, which includes water recombination and reactions with air.The main results for radiolysis inside the canister fill gas are described as follows.Significant radiolysis of water vapor requires the presence of residual air to disable recombination.Reactions between residual water (1 L) and air in 4500 L of free space result in percent levels of H 2 , O 2 and HNO 3 at 300 years, and about one+half the maximum values reached in the first 16 years.Calculations indicate that a much greater volume of residual water (20 L) would be required to reach the 4% H 2 flammability limit in 16 years, and between 3 and 4 L of water would be required to reach the 4% H 2 flammability limit in 300 years.Increased residual air results in greater H 2 and HNO 3 concentrations, but also in the depletion of O 2 because it is more effectively removed by a radiolytically induced reaction with N 2 .For lower (0.1%) residual air, O 2 is initially depleted and then generated for storage times greater than 50 years, resulting in approximately 0.5% O 2 .For lower (0.1 L) residual water and 1% air, all radiolytic products are less than 1%.The main results for radiolysis outside the canister follow.The main radiolytic products formed in moist air are HNO 3 , N 2 O, NO 2 , CO, and small amounts of O 3 . Even for extremely long residence times, the highest concentrations are less than 50 ppm and are less than 1 ppm for more typical flow conditions. Dry air gives similar concentrations as moist air with the exception of increased NO 2 and the near absence of HNO 3 .

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Spent nuclear fuel integrity during dry storage - performance tests and demonstrations

Cited by 2 publications

References 7 publications

Examination of Spent PWR Fuel Rods After 15 Years in Dry Storage

Examination of Spent PWR Fuel Rods After 15 Years in Dry Storage

Radiolysis Model Sensitivity Analysis for a Used Fuel Storage Canister

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