Properties of zirconium alloys and their applications in light water reactors (LWRs) - Materials' Ageing and Degradation in Light Water Reactors - page 107

corrosion mechanisms. For Zircaloy Yang (1989), Kruger (1990) and Cheng
et al . (1994) report that post-irradiation annealing causes SPPs to recrystal-
lize, to regain Fe and Ni, and to form under specifi c conditions of time and
temperature. Minimal effects are observed for 316°C (589K) for 30 days,
but for 200 days signifi cant amounts of Fe diffuse back to the precipitates.
At 400°C (673K) Fe diffuses back to precipitates in less than 10 days, and
Fe-rich precipitates form at grain boundaries. At higher temperatures
>560°C (833K) amorphized SPPs recrystallize, Fe and Cr diffuse back to
SPPs, and re-precipitation occurs in the matrix and grain boundaries. Recent
studies by Vizcaino et al . (2010) tend to confi rm the earlier results.

4.4 Mechanical properties of zirconium alloys

By 'mechanical properties' we essentially mean strength and ductil-
ity. Strength is expressed in terms of hardness, tensile strength, burst
strength, fatigue strength, etc. Ductility is likewise expressed in terms of
strain-to-failure or strain-to-some limit for the various loading conditions.
Fracture toughness is a combination of strength and ductility which describes
the stress required to propagate a specifi c crack geometry under specifi c
loading conditions. In this section we discuss various mechanical properties
as affected by reactor neutron irradiation. In addition, we describe mecha-
nisms and parameters which are related to mechanical properties and which
affect reactor component behaviour. This section deals primarily with prop-
erties which can be determined by out-of-reactor (or post-irradiation) test-
ing. For instance, tensile properties (strength and ductility) of interest for
in-reactor performance are mainly dependent on fl uence and independent
of fl ux. However, if the rate at which strain is applied becomes very low
(<10 − 6 s − 1 ), the mechanism of deformation changes, and fl ux becomes an
important variable (Azzarto et al ., 1969). That phenomenon is dealt with
more as creep, in the section on dimensional stability.
As described in Section 4.3.1, neutron irradiation dramatically alters the
microstructure of zirconium alloys. Of importance for mechanical prop-
erties are creation of <a> dislocation loops, and to a lesser extent, disso-
lution of precipitates (SPPs). Irradiation increases strength and hardness,
and decreases ductility. The effect on fatigue life (or strength) is less clear
and depends on testing technique, but generally appears to be small,
with some reduction of fatigue life in the low cycle region (Wisner et al .,
1994). Fracture toughness is clearly reduced by irradiation in Zr-2.5Nb
(Davies et al ., 1994), with concurrent effects of trace elements like chlo-
rine (Coleman & Theaker, 2004), and there are indications of a smaller
reduction in Zircaloys, for example (Bertsch et al ., 2010 ). The combination
of irradiation and hydride effects is important; for uniformly distributed
hydrides the observed reduction in ductility is mainly an irradiation, rather

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