ML20321A094

Appendix 8C - Fuel Oxidation and Cladding Splitting

ML20321A094
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Issue date:	04/30/2020
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APPENDIX 8C FUEL OXIDATION AND CLADDING SPLITTING Irradiated uranium dioxide (UO2) exposed to an oxidizing atmosphere will eventually oxidize to triuranium octoxide (U3O8). The time it takes to oxidize is a function of burnup and temperature.

At temperatures during dry storage system (DSS) fuel loading operations, this reaction can occur within a matter of hours.

The grain boundaries of irradiated fuel are highly populated with voids and gas bubbles. Initially, the grain boundaries are oxidized to U4O9, resulting in a slight matrix shrinkage and further opening of the pellet structure. Oxidation then proceeds into the grain until there is complete transformation of the grains to U4O9 (Einziger et al. 1992). The grains remain in this phase for a temperature-dependent duration until the fuel resumes oxidizing to the U3O8 state. The transformation to U3O8 occurs with about 33-percent lattice expansion that breaks the ceramic fragment structure into grain-sized particles. At higher temperatures, these two transformations occur so rapidly that they are difficult to distinguish. The mechanism of oxidation in irradiated fuel appears to be different than in unirradiated fuel where U3O7 is formed and oxidation proceeds from the fragment surface and not down the grain boundaries. This mechanistic change occurs between about 10 and 30 gigawatt days per metric ton of uranium (GWd/MTU).

When the UO2 is in the form of a fuel rod, the expansion of the fuel when it transforms to U3O8 induces a circumferential stress in the cladding. Because of the swelling of the fuel, the process is usually initially localized to the original cladding crack site. The cladding strains because of this stress range from 2 to 6 percent before the initial crack starts to propagate along the rod. The incubation time to initiate the propagation and the rate of propagation have an Arrhenius temperature dependence. Axial propagation, spiral propagation, and a combination of the modes that result in splitting have been observed in pressurized-water reactor (PWR) rods (Einziger and Strain 1986).

The database for oxidation was developed mostly in the 1980s in the United States, Canada, England, and Germany. The data usually appear in four forms: (1) O/M ratio (ratio of oxygen to metal content of the oxide) versus time, (2) time to the UO2.4 plateau versus time, (3) cladding splitting incubation versus time, and (4) cladding splitting rate versus time. Japanese researchers performed some later work on the effects of oxygen depletion (Nakamura 1995). French researchers are also working on similar questions (Ferry et al. 2005). Work on cladding splitting was done in the early 1980s by researchers in the United States (Einziger and Cook 1984; Einziger and Strain 1986; Johnson et al. 1984) and Canada (Novak and Hastings 1984; Boase and Vandergraaf 1977) and is limited. The Department of Energy (DOE) (Bechtel 2005) has issued an analysis of the oxidation issue in relationship to the handling of potentially breached fuel in a proposed handling facility at a repository. This analysis depends on variables such as the gap between the fuel and the cladding, and burnup in a manner that is currently under technical review. In total, this research has shown that there are a number of variables that can affect the rates at which the fuel oxidizes and the cladding splits: burnup, moisture content of the air, cladding material, and type of initial defect.

The DOE developed a model for fuel oxidation and cladding splitting (Bechtel 2005) for use during long durations at a disposal facility that tries to account for the fuel-to-cladding gap and burnup of the fuel. The gap is the as-measured cold gap and does not account for the closing of the gap as a result of differential thermal expansion of the cladding and fuel material, which could be calculated. There are inadequate data to verify the correctness of the DOE model. Plots in Einziger and Strain (1986) present actual data and comparisons with the data taken by other 8C-1

researchers at 30 GWd/MTU. The measurements of splitting implicitly account for the gap closure. However, no burnup effects can be inferred from these data.

No oxidation or cladding splitting studies have been conducted on fuel with burnup greater than 45 GWd/MTU. Data between 30 and 45 GWd/MTU show a decrease in the oxidation rate as a result of the presence of certain actinides and fission products that are burned into the fuel. There is no reason that this should not continue at higher burnups, but the strength of the effect may change with burnup. Higher burnup fuel (greater than 55 GWd/MTU) forms an external rim on the pellets that consists of very fine grains (1 micron). As indicated earlier, the oxidation process is a grain boundary effect. The fuel pellet should be divided into two regions for the purpose of oxidation analysis: the center of the pellet where the grains have grown slightly, and the rim.

While the rate of the oxidation may decrease with burnup, the total amount of fuel that is oxidized may increase because of a much greater intergranular surface area in the rim region. The DOE model (Bechtel 2005) uses a linear decrease in oxidation with burnup, but this has not been substantiated as of yet. A burnup effect is supported by Hansons analysis (Hanson 1998) of Einziger and Cooks data (1984) from the NRC whole-rod tests, in which defect propagation was observed to occur earlier at the defects at the lower end of the rod where the burnup was lower.

Studies using a low partial pressure of water vapor in air have not shown any dependence of the oxidation rate on the moisture content of the air (Ferry et al. 2005). On the other hand, some studies have shown a large increase in the oxidation rate when the moisture content is above 50 percent of the dew point. Oxidation in a 100-percent steam atmosphere is a different process.

Studies also indicate that the oxidation rate will decrease if the oxygen content in the atmosphere drops into the range of a few torr or less (Nakamura 1995). It does not appear that there is an effect of oxygen content at higher oxygen levels, but the data are sparse.

With few exceptions, oxidation studies on fuel have been conducted on light-water reactor fuel (Einziger and Strain 1986; Johnson et al. 1984). However, the UO2 matrix is essentially the same in both PWR and boiling-water reactor (BWR) fuel. At the higher burnups, oxidation behavior may vary slightly as the actinide and fission product burn-in varies. The effect of the process on the splitting of the cladding may vary considerably because of the difference in gap size between the cladding types, and the thicker cladding in BWR rods.

Limited cladding splitting studies have been conducted on Zircaloy-clad PWR (Einziger and Cook 1984; Einziger and Strain 1986; Johnson et al. 1984) and Canada Deuterium Uranium (CANDU) fuel. Defects were put in the fuel either by a stress-corrosion cracking process producing small, sharp holes, more typical of those found in reactor-initiated stress-corrosion cracking, and by drilling, which produced a larger, duller hole. Most of the defects used in the studies were of the latter type. No measurements were made in cladding above 30 GWd/MTU.

Very few data points were measured to determine the splitting rate; therefore, the time to start splitting has to be determined by interpolation. As a result, there is large uncertainty in both measurements. Further, the splitting of other alloy types (e.g., ZIRLO', M5) or at higher burnups should be assessed per the design-bases fuel contents. Fuel oxidation would introduce uncertainties for fuel performance and fuel retrievability.

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References Bechtel, Commercial Spent Nuclear Fuel Handling in Air Study, 000-30R-MGR0-00700-000000, March 2005.

Boase, D.G. and T.T. Vandergraaf, The Canadian Spent Fuel Storage Canister: Some Materials Aspects, Nuclear Technology, Vol. 32, pp. 60-71, 1977.

Einziger, R.E., and J.A. Cook, LWR Spent Fuel Dry Storage Behavior at 229 °C, HEDLTME 84-17, NUREG/CR-3708, Hanford Engineering Development Laboratory, August 1984.

Einziger, R.E., and R.V. Strain, Oxidation of Spent Fuel at Between 250° and 360°C, Electric Power Research Institute Report NP-4524, 1986.

Einziger, R.E., L.E. Thomas, H.V. Buchanan, and R.B. Stout, Oxidation of Spent Fuel in Air at 175 to 195 °C, J. Nucl. Mater., Vol. 190, p. 53, 1992.

Einziger, R.E., S. D. Atkin, D. E. Stellbrecht, and V. Pasupathi, High Temperature Postirradiation Materials Performance of Spent Pressurized Water Reactor Fuel Rods Under Dry Storage Conditions. Nuclear Technology, Vol. 57, p. 65. 1982.

Ferry, C, C. Poinssot, P. Lovera, A. Poulesquen, V. Broudic, C. Cappelaere, L. Desgranges, P. Garcia, C. Jegou, D. Roudil, P. Marimbeau, J. Gras, and P. Bouffioux, Synthesis on the Spent Fuel Long Term Evolution, Rapport CEA-R6084, 2005.

Hanson, B.D., The Burnup Dependence of Light Water Reactor Spent Fuel Oxidation, PNNL-11929, Richland, Washington, Pacific Northwest National Laboratory, TIC: 238459, 1998.

Johnson, A.B., E.R. Gilbert, D. Stahl, V. Pasupathi, and R. Kohli, Exposure of Breached BWR Fuel Rods at 325 °C to Air and Argon, Proc. NRC Workshop on Spent Fuel/Cladding Reaction During Dry Storage, Gaithersburg, Maryland, August 1983, NUREG/CR-0049, D.

Reisenweaver, Ed., U.S. Nuclear Regulatory Commission, 1984.

Nakamura, J., T. Otomo, T. Kikuchi, and S. Kawasaki, Oxidation of Fuel Rods under Dry Storage Condition, J Nuc. Sci. Tech., Vol. 32, No. 4, p. 321, April 1995.

Novak, J., and I.J. Hastings, Post-Irradiation Behavior of Defected UO2 in Air at 220-250 °C, Proc. NRC Workshop on Spent Fuel/Cladding Reaction During Dry Storage, Gaithersburg, Maryland, August 1983, NUREG/CR-0049, D. Reisenweaver, Ed., U.S. Nuclear Regulatory Commission, 1984 8C-3