Zirconium alloy

Abstract: The motivation for exploring the potential development of accident tolerant fuels in light water reactors to replace existing Zr alloy clad monolithic (U, Pu) oxide fuel is outlined. The evaluation includes a brief review of core degradation processes under design-basis and beyond-design-basis transient conditions. Three general strategies for accident tolerant fuels are being explored: modification of current state-of-the-art zirconium alloy cladding to further improve oxidation resistance (including use of coatings), replacement of Zr alloy cladding with an alternative oxidation-resistant high-performance cladding, and replacement of the monolithic ceramic oxide fuel with alternative fuel forms.

Abstract: New bulk amorphous alloys exhibiting a wide supercooled liquid region before crystallization were found in Fe–(Co,Ni)–(Zr,Nb,Ta)–(Mo,W)–B systems. The Tg is as high as about 870 K and the supercooled liquid region reaches 88 K. The high thermal stability of the supercooled liquid enabled the production of bulk amorphous alloys with diameters up to 6 mm. These bulk amorphous alloys exhibit a high compressive strength of 3800 MPa, high Vickers hardness of 1360, and high corrosion resistance. Besides, the amorphous alloys exhibit a high magnetic-flux density of 0.74–0.96 T, low coercivity of 1.1–3.2 A/m, high permeability exceeding 1.2×104 at 1 kHz, and low magnetostriction of about 12×10−6.

Abstract: In recent years sufficient new information has accumulated to change current views of which mechanisms of corrosion are operating in water-cooled reactors. The total number of publications is now so enormous that it is impossible for a short review to be completely comprehensive. This review concentrates on those studies that have resulted in changed views of the importance of various mechanisms. There seems to be insufficient evidence to support an hypothesis that increased corrosion rates in reactor result directly from displacement damage to the oxide by fast neutron bombardment. Replacing this hypothesis are the observations that redistribution of Fe from second phase particles (SPPs) into the Zr matrix by fast neutron recoil reduces the corrosion resistance of Zircaloy type alloys both in-reactor and in laboratory tests. There is little support for the idea that an irradiation induced phase change from monoclinic to tetragonal (or cubic) zirconia is important in-reactor, since such transformed zirconias are unstable in ∼300 °C water and revert to the monoclinic phase; as do chemically stabilised zirconias. New alloys with improved corrosion resistance in PWRs are generally low in Fe (and Sn), or have Fe in the form of more radiation resistant SPPs than those in the Zircaloys. Similarly, the hypothesis that nodular corrosion in BWRs was directly related to an effect of irradiation produced radical species in the water is unsupported. However, local dissolution of the oxide film by radiation produced species such as H2O2 may be occurring, and the close mechanistic relationship between nodular corrosion and ‘shadow corrosion’ is very evident. Thus, galvanic potentials between large SPPs (or clusters of SPPs) and the Zr matrix, aided by greatly increased electronic conduction of zirconia in irradiated systems appears to offer an hypothesis that provides a rationale for the observed effects of SPP sizes and numbers. Irradiation induced redistribution of Fe from the SPPs into the Zr matrix eliminates nodular corrosion susceptibility in Zircaloys.