Studies of Zirconium Alloy Corrosion and Hydrogen Uptake During Irradiation


Book Description

The in-reactor corrosion and hydrogen pickup of Zircaloy-2 and Zr-2.5Nb pressure tube materials are being studied in two test loops: a light water loop in the NRU research reactor, and a new heavy water loop in the Halden reactor. The complimentary test programs examine the corrosion behavior of small specimens as a function of fast neutron flux and fluence, temperature, water chemistry, and specimen pre-oxidation.




Radiation Effects on Corrosion of Zirconium Alloys


Book Description

From the wide use of zirconium alloys as components in nuclear reactors, has come clear evidence that reactor radiation is a major corrosion parameter. The evidence emerges from comparisons of zirconium alloy corrosion behavior in different reactor types, for example, BWRs versus PWRs and in corresponding reactor loop chemistries; also, oxidation rates differ with location along components such as fuel rods and reactor pressure tubes. In most respects, oxidation effects on power reactor components are paralleled by oxidation behavior on specimens exposed to radiation in reactor loops.