As outlined in Section 4.3.1, irradiation produces damage in the form of small dislocation loops (<a> component loops) which harden the material. The result is an increase in strength and decrease in ductility.

At reactor start-up, the tensile properties are the unirradiated properties reported by the fuel supplier. Mechanical properties begin changing imme­diately upon startup, and by an exposure of 5 MWd/KgU or a fluence of about 1 x 1025 n/m2 (E > 1 MeV) an increase in strength and decrease in ductility reach fluence-saturated values. Figure 4.20 illustrates this point for Zircaloy-4 irradiated and tested at 315°C (588K) (after Morize et al., 1987). Note also that the UTSs of cold worked stress relieved (CWSR) and recrys­tallized (RX) materials become similar at low exposures. This is a general trend which depends on the balance of hardening by pre-existing disloca­tions (cold work) and irradiation-produced defects.

Fuel cladding requires sufficient strength to prevent inward plastic defor­mation of the cladding at beginning-of-service conditions. PWR strength must be higher than for BWRs due to the higher water pressure needed to suppress boiling; therefore, PWR Zircaloy cladding has traditionally been in the cold work stress relieved annealed (SRA) condition. The discussion above points out that the difference in strength between SRA and RXA materials is short-lived under reactor conditions.

Properties of zirconium alloys and their applications in LWRs 177