A brief summary of hydride effects is given here to provide background for pellet-cladding mechanical interaction (PCMI) type failures. All zirconium alloy reactor components absorb hydrogen during reactor service through the corrosion reaction between zirconium and water. Basics of these phenomena are given in ZIRAT Special Topical Reports (Cox & Rudling, 2000; Adamson et al, 2006; Strasser et al, 2008). Hydrides tend to embrittle zirconium alloys and therefore their effects are important for in-reactor normal service, for ex-reactor handling operations and for accident and transient scenarios such as LOCA and RIA. It is thought that individual hydrides themselves are actu­ally brittle at all normal reactor temperatures (Simpson & Cann, 1979; Shi & Puls, 1999); and it is clear that high concentrations of hydrides (5000-16000 ppm) are very brittle, as in hydride blisters or rims.

Under normal conditions, hydride platelets form in the circumferential direction in fuel cladding illustrated in Fig. 4.27a, but under some circum­stances such as during long term storage or during power transients they

4.27 Hydride orientation in Zircaloy-4 (SRA) cladding: (a) circumferential, (b) radial (Chu et al., 2005).

can form in the radial direction (Fig. 4.27b). Because in high power rods a temperature gradient encourages hydrogen to diffuse to the colder outer clad surface, rims of hydrides can form, illustrated in Fig. 4.28a.

Hydrides effects are listed here, giving appropriate figures and references.

• The effect of hydrides is strongly dependent on testing temperature. Material at 300°C (573K) (reactor operating temperature regime) retains much more ductility than at 20°C. Figures 4.28b and 4.29 indicate the ductile-to-brittle transition for unirradiated material is less than 200°C, for circumferentially oriented hydrides. Figures 4.30 and 4.31 indicate that at 332°C the primary reduction in ductility comes from the irradia­tion effect, while at room temperature the effect on ductility of irradia­tion and hydrides is additive for uniformly distributed hydrides below about 1000 ppm. It is apparent that below 100°C ductility is very low.

• The distribution of hydrides is important. Dense layers of hydrides (for instance at fuel cladding surfaces) retain little ductility at any tempera­ture, and are susceptible to crack formation. Whether or not the crack will be arrested by the relatively ductile zirconium matrix depends on the layer thickness, as shown in Fig. 4.32.

• The strength of irradiated or unirradiated Zircaloy is insensitive to hydrogen content. See Fig. 4.33.

• Existence of radial hydrides can substantially reduce ductility, particu­larly at room temperature. Figure 4.34 shows the failure strains for the range of hydride orientations given in Fig. 4.35. When radial hydrides exist as in Fig. 4.35c failure strain is low. Figure 4.36 indicates that a high percentage of radial hydrides reduces the failure strain at room

As-received 200 ppm 400 ppm 500-650 ppm

800-950 ppm 1000-1300 ppm 1300-1450 ppm >1550 ppm

temperature but not at 300°C (573K). All specimens are unirradiated and are tested with applied stress normal to the hydride platelet. For similar materials having the applied stress parallel to the hydride plate­let, no hydride effect is seen (Yagnik et al, 2004).