Published: Jan 1977
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The corrosion and hydriding performance of zirconium-base alloys under pressurized water reactor (PWR) and boiling water reactor (BWR) conditions, as gaged by a comprehensive review of the technical literature, has been evaluated. Starting with a brief historical description of the development of zirconium for cladding and structural material in nuclear reactors and the corrosion problems associated with the use of the pure metal, it is shown that the development of zirconium-base alloys proceeded down two major paths. One development involved the zirconium-tin system and led to the development of the Zircaloys, whereas the other concentrated upon zirconium-niobium materials and produced the two major alloys of this system in use today: Zr-1Nb and Zr-2.5Nb. The corrosion data generated for each system, both in- and ex-reactor, are evaluated, and the benefits and potential problems associated with each alloy are discussed for both PWR and BWR applications. Potential areas of concern for the Zircaloy alloys in both applications include exposure temperature limitations and the formation of nonuniform accelerated corrosion products in the oxygenated irradiation environment. The zirconium-niobium alloys are found to be very sensitive to oxygen in the coolant and to prior heat treatment in ex-reactor experiments but show either minimum or negative acceleration due to the presence of neutron irradiation. Alloys that combine these two additives (for example, Zr-3Nb-1Sn and Ozhennite-0.5) do not appear to show promise as possible replacements for the Zircaloys under present-day conditions. The lack of a uniting theory for describing the mechanisms involved in the corrosion of zirconium-base alloys may hamper seriously possible future applications under different design conditions.
zirconium alloys, irradiation, corrosion, reactors, hydrides
Fellow engineer, Westinghouse Electric Corporation, Bettis Atomic Power Laboratory, West MiffLin, Pa.
Paper ID: STP35573S