Corrosion of Zirconium-Base Alloys—An Overview

Abstract

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.

Hydrazine and Hydrogen Co-injection to Mitigate Stress Corrosion Cracking of Structural Materials in Boiling Water Reactors, (III) Effects of Adding Hydrazine on Zircaloy-2 Corrosion

“…(4) and (5), the activation energy was estimated to be 133 kJ/mol. The obtained activation energy agreed with the reported value 9) within the experimental error of 10%. Supercritical water at 723 K accelerated the corrosion rate by 8.3 times over that of water at 561 K. Figure 3 compares the weight gains of Zircaloy-2A and Zircaloy-2B in water at 561 K and in SCW at 723 K with -irradiation.…”

RETRACTED ARTICLE: Consideration on application of empirical corrosion rate model of Zircaloy at 280–400°C to deep underground temperature

“…The K c at 180uC falls on the linear interpolation between 80-120uC and 280-400uC. 2,5,14,15 4 Image (TEM), electron beam diffractions and EDX profiles of cross-section surface of Zircaloy specimen after 81 days corrosion at 180uC…”

Accommodation of tin in tetragonal ZrO2

“…Zirconium alloys are used in pressurised water reactors due to their low capture cross-section for thermal neutrons, high corrosion resistance and acceptable mechanical properties. Initially, corrosion follows an approximately cubic rate law [1]. After a few microns of oxide growth, there is a sudden increase in the corrosion rate as the protective layer breaks down; this process is generally referred to as transition, or break-away corrosion.…”