Properties of zirconium alloys and their applications in light water reactors (LWRs) - Materials' Ageing and Degradation in Light Water Reactors - page 94

Table 4.5 Radiation damage: <c> loops in Zircaloy

Nature Vacancy

Size >100 nm (1000 )

Density 0.5 × 1018 m −2 (for Fig. 4.8)

Incubation fl uence 3 × 1025 n/m 2 ( E >1 MeV)

Thermal Stable to >560 ° C (833K)

Form at >200 ° C (475K)

Effect growth, creep?

Source: A.N.T. International (2011).

4.8 <c> type dislocations in Zircaloy-4 after a fl uence of 12 × 1025 n/m 2
at 561K. (Source: Adamson, 2000.)

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to high temperature (>833K). It is thought that <c> loops strongly infl u-
ence irradiation growth and creep behaviour and probably do not affect
mechanical properties. Figure 4.8 shows TEM images of a high density of
<c> loops in highly irradiated Zircaloy. Such <c> loops, unlike <a> loops,
do not appear to form in all zirconium alloys, particularly in those having
additions of Nb, or Nb and Fe (Shishov et al ., 2002), until high fl uences are
experienced.
As outlined in Tables 4.4 and 4.5, the formation kinetics of <a>- and <c>-
type loops differ. The density of <a> type dislocation builds up quickly and
saturates at a fl uence less than 1 × 10 25 n/m 2 , E > 1 MeV, as illustrated in
Fig. 4.9. It appears that a fl uence-incubation period exists before <c> type
loops begin to form at about 3 × 10 25 n/m 2 , E >1 MeV for typical reactor
temperatures, as illustrated in Fig. 4.10 .