Corrosion Behavior of Zirconium Alloy Nuclear Fuel Cladding
- PDF / 1,398,970 Bytes
- 8 Pages / 420.48 x 639 pts Page_size
- 94 Downloads / 238 Views
CORROSION BEHAVIOR OF ZIRCONIUM ALLOY NUCLEAR FUEL CLADDING
ANNA C. FRAKER AND JONICE S. HARRIS Metallurgy Division, National Institute of Standards and Technology, Gaithersburg, MD 20899 ABSTRACT Both alloys are Zircaloy-2 and -4 are used as nuclear fuel cladding. more than ninety-eight percent zirconium and are corrosion resistant to various media. Electrochemical measurements using polarization techniques have been made on these alloys in aqueous media with a pH of 8.5 and varying ionic concentration (lX and 1OX) at temperatures of 22°C and 95°C. Results showed that under the test conditions of the study these alloys passivated and had negligible corrosion rates, but there were some variations in passivation due to surface preparation and some crevice corrosion was observed. Data are presented and discussed in terms of passivity, breakdown potential and susceptibility to localized corrosion. INTRODUCTION The purpose of this study was to provide data for use in evaluating corrosion behavior of the zirconium alloys, Zircaloy-2 and Zircaloy-4, and for use in determining whether long term credit can be claimed for the cladding in preventing radionuclide release to the environment. The U. S. Nuclear Regulatory Commission (NRC) requires that nuclear waste containment shall be substantially complete for a period of 300 to 1000 years and that thereafter, no more than one part in 105 of the inventory of radionuclides present at 1000 years after closure may be released annually from the engineered barrier system[l]. It is not known whether it would be necessary to take credit for the cladding to meet the release requirement. The cladding tube with a 11 to 12 mm outside diameter and a wall thickness of less than 1 mm surrounds the nuclear fuel, uranium dioxide pellets that have been sintered to 95% theoretical density, for the purpose of reducing coolant activity levels. Zircaloy-2, Zircaloy-4, other zirconium alloy compositions and the 300 series stainless steels have been used as cladding materials, but the bulk of the cladding in the United States is Zircaloy-2 and Zircaloy-4. Metallurgical aspects of zirconium alloys and information on corrosion behavior in various media have been discussed previously[2]. Essentially, Zircaloys-2 and -4 are ninety-eight percent zirconium and are free of hafnium. Zircaloys-2 and4 are highly corrosion resistant in various media and environmental conditions. Zirconium materials are highly reactive and obtain corrosion resistance by the formation of a protective film. Ions which penetrate or react with this film, or oxidation temperatures and conditions which change it would have a negative effect on the good corrosion properties of these zirconium alloys. The work reported here provides corrosion data and electrochemical measurements of Zircaloy-2 and -4 in aqueous media at 95°C with a pH of 8.5 and an ionic content representative of that found in the Nye County, Nevada in the J-13 well. These data can be used to characterize the corrosion behavior of Zircaloy under these conditions as i
Data Loading...
