Zircaloy irradiation effect

Fig. 3.1 -93 Irradiation effects on mechanical properties of a zircaloy [1.68], YS - yield stress UTS - ultimate tensile strength. Solid lines irradiation in the annealed state. Dashed lines irradiation in the 10% cold worked state...

I0.6.8.I Cladding failure in oxide fuel pins of nuclear reactors. The long-term operational performance of nuclear fuel pins is critically governed by the reactions that occur in the gap between the fuel and its cladding. Ball et al. (1989) examined this for the cases of (1) Zircaloy-clad pellets of U02+, in a pressurised water reactor (PWR) and (2) stainless-steel-clad pellets of (U, P)02+, in a liquid-metal-cooled fast-breeder reactor (LMFBR). In particular they were interested in the influence of O potential on Cs, I, Te and Mo and the effects of irradiation on the gaseous species within the fuel-clad gaps. 

To help improve the corrosion resistance of Zircaloy, several new zirconium alloys have been developed, such as Zirlo (Zr-1.0% Nb-1.0% Sn-0.1% Fe). Notwithstanding the progress so far, materials reliability does have a significant effect on the economics of nuclear power plants, and there is considerable incentive to develop a full understanding of the mechanisms of corrosion of zirconium alloys in reactors and to develop alloys that are resistant to both irradiation and corrosion in reactors [18]. 

An important consideration for zircaloy pressure tubes is the enhancement of the creep rate by exposure to neutron irradiation. This is illustrated in Fig. 5.6, which shows measurements of the transverse strain in a zircaloy-2 pressure tube which had been machined to different thicknesses along its length, so that it was possible to investigate the effects of stress variation as... 

Pig. 1 Effect of hydrogen and irradiation on the suh-size, V-notch impact transition temperature for Zircaloy-2... 

With regard to the zirconium-niobium alloy, it appears that its corrosion rate is not much affected by irradiation, but it is very dependent on the local chemistry and can be accelerated by oxidizing conditions. The corrosion rate is also dependent on heat treatment condition, and good resistance is shown by material aged for 72 hours at 500°C. With such material and with suitable chemical control, it is expected that, as with Zircaloy-2, there will be no problem due to loss of section and the important consideration is that of hydrogen absorption and its consequent effect on mechanical properties. 

Attention has been drawn to the accelerating effect of neutron irradiation on the creep of zirconium alloys. The present evidence indicates that this will not cause a problem with cold-worked Zlrcaloy-2 pressure tubes in SGHWR designs. Considerably more information on this aspect of Zircaloy-2 and heat-treated Zr-2- Nb alloy behaviour will arise from current experimental work in research reactors and in the operation of the Winfrith SGHWR. 

Hafnium has a hexagonal structure, this inducing anisotropic deformation under irradiation ( growth ) similar to that observed on the Zircaloy sheaths (Fig. 15.15). This is due to a combination of two effects ... 

The corrosion -irradiation program for Zircaloy 2-clad, uranium-niobium core alloy (3 and 6% niobium) plates described in ANLi-5260 was extended to include the I-1/2 % nipbium-5%zirconium alloy. The plates are to be exposed in the Argonne Water Loop at MTR both in and out of neutron flux in contact with 500F water. Deliberate, (drilled hole) defects will then be put into the clad and the plates will be further corrosion tested out of flux to determine the effect of irradiation on corrosion resistance., ... 

---