United States Nuclear Regulatory Commission - Protecting People and the Environment

Zinc-Zircaloy Interaction in Dry Storage Casks (NUREG/CR-6732)

On this page:

Download complete document

Publication Information

Manuscript Completed: June 2001
Date Published: August 2001

Prepared by:
H. Tsai, Y. Yan

Argonne National Laboratory
Argonne, IL 60439

D. N. Kalinousky, NRC Project Manager

Prepared for:
Division of Engineering Technology
Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001

NRC Job Code Y6373

Availability Notice

Abstract

Due to limited storage capacity in spent-fuel pools, some nuclear power plants are storing spent-fuel assemblies in inert-atmosphere dry casks until long-term geological depositories are available. VSC-24 is one such type of cask being used by utilities. During spent-fuel loading, the VSC-24 casks are submerged in the borated pool water and then dried and sealed. To mitigate contamination of the pool water and corrosion of the cask, a zinc-based primer coating, CarboZinc11, is applied to the cask structure, such as the fuel assembly sleeves and the cask shell. A series of laboratory tests was conducted to evaluate possible metallurgical interactions between zinc from the CarboZinc11 primer and the Zircaloy-4 cladding. The postulated transport mechanism of zinc is vaporization. If such interaction occurs during spent-fuel storage, performance of the Zircaloy-4 cladding as the primary barrier for fuel and fission products could be degraded.

The tests were designed to simulate realistic cask loading and storage conditions, including borated water immersion, vacuum drying, helium backfill, and elevated temperature holds. Prototypical cladding and coating materials were used and the cladding was preoxidized to form a surface layer comparable to those on irradiated fuel rods at medium burnup. The test temperature was 300°C (which envelops the peak cladding temperature of 282°C at a realistic cask heat load of 12 kW) and the maximum hold time at temperature was 90 days. No zinc Zircaloy interaction was found in any of the tests. These results differ from those of earlier tests conducted under less prototypical and generally more aggressive conditions. The major influence on the test results apparently was the oxide layer on the cladding specimens in the present tests. The oxide layer appears to be effective in blocking the migration of zinc vapor to the Zircaloy metal substrate. Because an oxide layer is always formed on fuel rod cladding from in-reactor service, the cladding would be protected as long as the oxide layer remains intact and adherent to the Zircaloy metal.

Page Last Reviewed/Updated Wednesday, June 12, 2013