Institute of Metals Division - The Irradiation Stability of Low Wt Pct Uranium-Zirconium Alloys

In this paper the results of an exploratory study conducted by personnel of the Knolls Atomic Power Laboratory on 18.6, 22, and 40 wt pct uranium-zirconium alloy will be presented. LARGE power output and long life or endurance from small, compact nuclear reactors require the use of enriched uranium for a fuel material. However, the ability to remove heat from the element, the tolerable fission product concentration, or some physical limitation of the system can limit the allowable fuel enrichment, or the volume of the fuel. In such an event, the fuel-element designer must find a suitable diluent for the uranium. Zirconium is an attractive material for such a purpose as it has good nuclear properties and a low absorption cross section. Because it has a good corrosion resistance in water, zirconium is a suitable cladding and the uranium-zirconium alloy makes a good core material. The alloy has a high thermal conductivity, a low elastic modulus, and a low thermal-expansion coefficient, all of which tend to minimize the thermal stresses in the core. Furthermore, the unirradiated material is reasonably ductile and has adequate strength. DESCRIPTION OF SAMPLES Two types of cylindrical samples were used in this study 1) short, 1/2-to 1-in.-long samples to test the irradiation stability of the alloy and 2) 2-ft-long samples to investigate the resistance of an element to longitudinal deformation. Composite samples having 0.060-to 0.140-in. core diameters and clad with a 0.010-in. thick zirconium jacket were fabricated by coextrusion and by hydrostatic pressure bonding of the cladding to the preformed core material. Both fabrication processes produced the uniform metallurgical core-to-cladding bond which was felt necessary to minimize the uncertainty of the central core temperature. Both consumable-arc and induction-melting techniques were used to melt the uranium-sponge zirconium core alloy. Since a graphite crucible was used in induction-melting this alloy contained more carbon impurity, which resulted in a finer grain size than the arc-melted alloy. The core to cladding interface was uniform in elements coextruded from the fine-grain, induction-melted core material, while those elements fabricated using arc-melted core material had very irregular interfaces because of the coarser grain size. Both cold-worked and heat-treated fuel materials were irradiation tested. Cold-worked alloys were swaged to about a 5 pct final reduction in area without subsequent annealing. The heat-treated alloys were heated to 800°C, held at temperature for 15 min and furnace-cooled to about room temperature. This