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.
Three original Zircaloy-2 clad blanket fuel bundles from the pressurized-water reactor (PWR) at the Shippingport Atomic Power Station were discharged after continuous exposure during Cores 1 and 2. Detailed visual examination of these components after ∼6300 calendar days of operation (51 140 effective full power hours) revealed only the anticipated uniform light gray (posttransition) corrosion products with no evidence of unexpected corrosion deterioration, fuel rod warpage, or other damage. All corrosion films were found to be tightly adherent to the underlying cladding. An extensive destructive examination of a selected fuel rod from each of three fuel bundles produced, as expected, significantly greater end-of-life rod average oxide film thicknesses when compared with corresponding values calculated from the time-temperature history of each component, employing a set of empirical equations generated from the out-of-pile (autoclave) testing of Zircaloy coupons. It is shown that the pretransition region and the time to transition are in adequate agreement with the empirical equations generated from the ex-reactor data but that there is a significant acceleration of the posttransition corrosion rate due to the irradiation exposure. Post-transition corrosion rate data were used to develop an empirical expression of the form R′=A(φ)n·R where R′ is the in-reactor posttransition corrosion rate for Zircaloy-2 tubing, A and n are experimentally determined constants, φ is the posttransition fast flux exposure, and R is the posttransition temperature-dependent corrosion rate generated from exreactor (autoclave) testing. The Multipurpose Extended Life Blanket Assembly cladding in the mid-sections of the rods (α microstructure) was found to have corroded at a rate that is approximately 25 percent greater than the corresponding welded end cap regions (β-quenched struture). It has been observed that the total hydrogen content of the cladding was less than that calculated from the ex-reactor expressions, using the measured oxide film thickness. It has also been found that the hydrogen pickup per unit weight gain is two to three times greater in the late posttransition period than that observed in the early posttransition region, possibly casting doubt on the validity of the application of the ex-reactor hydrogen pickup expression (that is, a constant ΔH/Δ0 ratio for the entire posttransition region) to in-reactor exposures.
Three original blanket fuel bundles from the Zircaloy-2 clad pressurized water reactor (PWR) at the Shippingport Atomic Power Station were discharged after continuous exposure for the entire first core, consisting of one original seed and three seed refuelings, and for the first seed life of the second core. Detailed visual examinations of these components after ∼4100 calendar days of operation (41 000 EFPH) revealed only the anticipated uniform gray-tan (post-transition) corrosion products with no evidence of a gross Zircaloy corrosion acceleration; all corrosion films were found to be tightly adherent to the underlying cladding. An extensive destructive evaluation of two fuel rods from each of the three fuel bundles produced greater oxide film thicknesses (7 to 12 μm or 0.28 to 0.47 mils) than the values predicted from the time-temperature history of the components (3 to 5 μm or 0.12 to 0.20 mils); an average corrosion acceleration factor of 2.4 can be calculated from the data. In slightly different terms, the observed corrosion acceleration was equivalent to raising the average surface temperature by ∼40°F. The accelerated corrosion attack was accompanied by a considerably lower total hydrogen pickup than that calculated from the observed film thicknesses. This combination Of an irradiation-induced corrosion acceleration and a less-than-predicted hydrogen absorption is suggestive of operation in an oxygenated environment although the Shippingport PWR coolant is maintained at a hydrogen concentration in the range 10 to 60 c3 H2/kg H2O.
Cyhnders of iron, zinc, and cadmmm were rotated in a highly cmrosive solution of acid contaimng nitrate ion as a depolarizer, as described prevmusly. Three kinds of mhlbitors were added to the solutmn (a) dichromate mn plus complexmg or chelating agents for metal ions; (b) a wetting or emulsifying agent whmh is strongly adsorbed; and (c) a reagent which forms a very insoluble precipitate w~th ferrous and ferric ions. Measurements of the effectweness of these mhibitors are g~ven
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