Overview of ageing and degradation issues in light water reactors (LWRs) - Materials' Ageing and Degradation in Light Water Reactors - page 377

on basal planes and to boundaries parallel to basal planes. This gives rise
to elongation in one direction and contraction in the other. The creep rate
can be controlled by suitable alloying additions and modifying the texture
of the zirconium matrix such that the dislocation glide rate is reduced. The
complex behaviour was modelled by Nichols 38 using dislocation climb-glide
processes as a function of stress and neutron radiation dose. There is an
extensive literature on radiation creep of different materials and the reader
is referred to the literature for more details.

1.3.4 Effect of radiation on fatigue

Since LCF and HCF are controlled by ductility and strength respectively,
and radiation results in hardening and embrittlement, we expect life in
HCF to be improved and that in LCF to be degraded. Murty and Holland 15
examined the fatigue characteristics of Type 304 SS from hexagonal cans
of EBR-II guide tubes before and after irradiation to a fast fl uence of ~8 ×

10 26 n/m 2 (Fig. 1.12a). Tests were performed in four-point bend mode at con-
stant displacements under symmetrical strain reversal fatigue at 0.1 cps and
strains were varied from ~1% to 2.4% with the number of cycles to failure
varying from 500 to 40 000. While a slight decrease in fatigue life is noted
at high strains or low cycles, the data clearly revealed improved fatigue life
at low strains or high cycles (Fig. 1.26) where the model predictions using
universal slopes are correlated with experimental results. According to their
model,

Type 304 stainless steel - 325°C

Model
Prediction

5

Expt.

Unirradiated
Irradiated

￿ ￿ ￿ ￿ ￿ ￿

1

5

10 3 10 4

5 5

N f

1.26 Strain amplitude versus number of cycles to failure at 325°C for
unirradiated and irradiated 304SS depicting improved fatigue life in
HCF and degradation in LCF. 15