Overview of ageing and degradation issues in light water reactors (LWRs) - Materials' Ageing and Degradation in Light Water Reactors - page 387

reaction of metal with water, hydrogen released by radiolysis of water and
hydrogen gas that is added in the coolant to keep the oxygen potential
low. 72 The defects present in the clad (such as manufacturing defect, PCI
crack, debris fretting, etc.) can aid the pick-up of hydrogen and the coolant
could surge in through these defects when they grow through-thickness
and form steam. The steam reacts with the fuel and hydrogen is released.
When the hydrogen-to-steam ratio crosses a critical value (steam starva-
tion), the growing oxide layer on the ID of the tube fi nally breaks down
and the hydrogen diffuses into the matrix of the tube. The hydrogen thus
picked up can reduce the toughness of the zirconium matrix in three ways:
(i) hydride reorientation, (ii) delayed hydrogen cracking and (iii) forma-
tion of a hydride blister.
The solubility limit of hydrogen in zirconium at the reactor operat-
ing temperature is about 100 ppm. When the temperature is reduced (for
instance, during reactor shutdown), the excess hydrogen precipitates in the
form of hydride. The hydride precipitates along the radial direction of the
tube owing to the texture and the hoop stress in the tube. The hoop stress
required for reorientation (in an unirradiated and recrystallized Zircaloy-2)
is about 80 MPa. 73 The differential temperature between ID and OD of the
clad wall (either during service or at wet repository) drives the hydrogen to
the OD side which is at a lower temperature. The concentration of hydrides
found in the tube after irradiation is higher near the water side than at the
fuel side which is attributed to the corrosion reaction between the clad
OD and the coolant. 74 The threshold stress for failure of irradiated and
hydride-reoriented spent fuel cladding is signifi cantly higher than the stress
due to the internal pressure of the fuel rod. The degree of oxidation and
hydriding in the more advanced fuel claddings commonly used these days in
LWRs, such as low-Sn Zircaloy-4 (Sn content around 1.3 wt.%), optimized
Zircaloy-4, Zirlo (a Zr-1Sn-1Nb-0.1Fe alloy), M5 (a Zr-1Nb alloy) and opti-
mized Zircaloy-2, is relatively low even at high burnup.
Frequently, a hydride blister is produced when a fuel rod that contains
spalled oxide is operated continuously to high burnup. During steam star-
vation, hydrogen ingress is faster than its diffusion into the tube matrix. This
leads to excess amounts of hydrogen getting localized at the inner wall of
the clad tube forming a large hydride called a blister. The hydrogen atoms,
diffusing down the temperature gradient, form radial hydrides in a sun-burst
pattern.
Delayed hydrogen cracking (DHC) is important for spent fuel in either
wet or dry repository and is a two-step process. Hydrogen migrates up the
stress gradient towards a stressed crack-tip and precipitates as hydride that
cracks and extends further. There is an incubation time for the hydrogen
to arrive and the concentration to build to the required level, so that the

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