Volume 6 Paper 14
Mechanisms of Zirconium Alloy Corrosion in Nuclear Reactors
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Mechanisms of Zirconium Alloy Corrosion in Nuclear Reactors
JCSE Volume 6 Paper 14
6 paper 14Submitted 6th July 2003
Mechanisms of Zirconium Alloy Corrosion in Nuclear
for Nuclear Engineering, University
of Toronto, Toronto,
ON, Canada, M5S 3E4
extent to which irradiation in a nuclear reactor core accelerates the corrosion
of zirconium alloys has always been a concern.�
It was expected that radiation induced defects in the protective
zirconia film would enhance diffusion controlled oxide growth, and lead to
early breakdown of the protective oxide.�
Studies of irradiated oxide films (>10yr. in reactor) have found no
evidence for irradiation induced defects in the critical region of the oxide
close to the metal/oxide interface.�
This suggests that recombination of the point defects produced must be
essentially complete.� Enhanced oxide
growth in-reactor must, therefore, involve other processes.� Early studies showed that electronic
conduction in the oxide, rather than oxygen diffusion was the rate determining
process.� In-reactor, therefore, the
enhanced conductivity of the oxide film should relax this limitation on the
oxide growth rate.� Under such
conditions galvanic currents between the zirconium alloy and a dissimilar metal
could then become significant as well.�
That these two effects can operate simultaneously and independently was
shown in experiments in the Advanced
Thermal Reactor (ATR) at low temperatures.� Specimens were contained in aluminum holders, and experienced
both cathodic hydriding in areas of proximity to Al (galvanic), and enhanced
anodic growth overall.� The cathodic
hydriding was eliminated by a change to Zircaloy specimen holders, but this had
no effect on the rate of anodic oxide growth.�
The latter suggests that the limiting oxide thickness determined by low
electronic conductivity of the oxide out-reactor had been relaxed.
observations allow an understanding of the "Shadow Corrosion"
phenomenon in Boiling Water Reactors (BWRs), and suggest that the mechanism of
the common nodular corrosion phenomenon in BWRs is a similar process operating
on a micro-scale, where the second-phase particles (SPPs) in the zirconium
matrix become the "dissimilar metal" in the galvanic couple.� This would explain the well known dependence
of nodular corrosion on SPP size (i.e. cathode to anode ratio).� In Pressurized Water Reactors (PWRs) the
galvanic effects appear to be absent because of the high dissolved hydrogen
content of the water (all surfaces at the reversible hydrogen potential) and
other processes causing degradation of the corrosion resistance of the
Zircaloys must be invoked.
since zirconium alloys were first developed for use as structural materials in
the cores of water cooled nuclear reactors  there has been concern that the
radiation fields in-reactor would accelerate the corrosion process.� Such in-reactor accelerations of corrosion
rates have been observed, but there has been little or no agreement on how
these accelerations were effected , and the number of published papers and
reports on the subjects has become legion.�
A few critical experiments have changed our view of what the important
processes are, and may have resulted in a viable hypothesis for these effects.
corrosion kinetics of zirconium alloys typically exhibit two stages, for alloys
such as the Zircaloys and others containing second-phase precipitates (SPPs) of
relatively insoluble transition metals such as Fe, Cr, Ni, V etc.� The initial stage is a diffusion controlled
growth of a thin protective oxide film that follows a kinetic law that is close
to cubic, rather than parabolic, at typical water cooled reactor
temperatures.� Once this oxide exceeds a
thickness of about 2:m this protective film breaks down to give approximately linear
kinetics, indicative of a roughly constant thickness residual barrier oxide
film.� It was expected that reactor
irradiation would accelerate both the diffusion controlled growth, and the
subsequent breakdown processes .� It
was therefore something of a surprise when results showed apparently no effect
on the pre-transition diffusion controlled processes, and significant
accelerations of the post-transition corrosion rates [4,5].
situation only pertained to behaviour in PWRs (Pressurised Water Reactors),
where large concentrations of dissolved hydrogen were present, and the oxide films
remained uniform in thickness until oxide spalling ensued at $100:m.� In BWRs (Boiling Water
Reactors) where little or no hydrogen was added to the water, and extensive
radiolysis of the coolant occurred, the oxide films usually exhibited severely
localised nodular corrosion .� This
behaviour could be controlled by careful attention to the size and distribution
of the SPPs .� This effect was
ascribed to a direct effect of oxidising radicals produced in the water by
radiolysis .� Although it was evident
from early in-reactor loop tests that there was a significant galvanic
corrosion component to the formation of nodules, which were observed only under
stainless steel grids, and not under Zircaloy grids, on the Zircaloy cladding
in a fuel assembly with alternate stainless steel and Zircaloy grids .� This aspect of in-reactor corrosion in BWRs
was largely ignored until recent severe "shadow corrosion" adjacent
to Inconel grids nearly led to fuel failures in the Leibstadt (KKL) BWR [8,9].
Zr alloys are usually covered by thick oxide films in-reactor it has always
seemed improbable that radiolytically produced oxidising radicals could do
anything beyond changing the surface potentials at the oxide film/water
surface.� Thus, in order to understand
this in-reactor behaviour a knowledge of how irradiation affected the transport
processes within the zirconia films on the metal surface is needed.� Early work , recently corroborated ,
showed that the cations in zirconia films were immobile and only the oxygen
anions migrated.� Charge balance was
maintained by electron conduction through what was normally a good electrical
information needed to formulate a mechanism for the in-reactor corrosion,
therefore, comprised a knowledge of which of the transport processes
(electronic or oxygen anionic conduction) was rate determining in the absence
of irradiation, and the relative effects that
in-reactor irradiation had on
the two processes.� This paper will summarise the evidence obtained on these effects.
of the rate controlling process during oxidation requires a measurement of the
potential difference that develops across the oxide film during its
growth.� The difficulty with such a
measurement is the method of applying a contact on the outer surface of the
oxide without affecting either of the conduction processes in the oxide
.� It was decided that this could
best be achieved by using an oxygen containing fused salt environment, rather
than either an aqueous solution or an outer metallic (or conducting oxide)
contact .� This allowed measurements
to be made at typical reactor temperatures (300-350�C) under conditions where
the oxidation rates of zirconium alloys were the same as in high temperature
water, but where polarisation of surface reactions in the environment did not
affect the results.
of effects of high in-reactor radiation fields on the separate anodic (oxide
growth) and cathodic (metal hydriding) processes were made in the ATR at Idaho
Falls (USA) during experiments to study the irradiation induced growth of
Zircaloys at low temperatures (~ 50�C) in water .� Specimens were irradiated initially in
aluminum specimen holders,
and were later changed to Zircaloy specimen holders.� The fast neutron flux (>1 MeV) was 3x1014 n.cm-2.s-1.� The oxide film thickness was determined from
an interference colour chart after photography through the hot-cell window with
a telephoto lens.� The specimen was
placed on a standard grey background with a colour strip adjacent to it.� The colour strip was included in the area
photographed and the effect of the lead-glass cell window (yellow) was
eliminated by rebalancing the colour of the negative so that the colour strip
in the photograph matched the original.��
Evidence for irradiation damage in the oxide films formed on Zr alloys
was obtained by studying the oxides on highly irradiated (10-12 years
in-reactor) Zr-2.5%Nb pressure tubes from CANDU reactors in a Philips CM30
transmission electron microscope .
potential across the oxide film formed isothermally on Zircaloy-2 at 300�C
always showed the metal to be negative with respect to the oxide/environment
surface (Figure 1).� The specimen took
several minutes to develop a potential of �1.1 V from an initial value close to
zero.� The origin is not visible in
figure 1 because of the logarithmic time scale (but see figure 4,�� ref.13), as the oxide thickened this
potential increased but remained negative.�
These potentials could be materially changed, and even shifted to
positive values (Figure 2) by a high temperature (445�C) pre-oxidation, which
would have redistributed the iron alloying addition within the oxide formed
ATR interference-coloured oxides, that increased almost linearly with fluence
were observed at ~50�C (Figure 3).� As
the oxide thickened the interference colours changed, but remained relatively
uniform over most of the specimen surfaces (Figure 4).� Only the oxide on the narrow rim of the
specimen that fitted in the slots in the aluminum specimen holders was slightly
thinner than on the specimen face.� The
magnitude of this reduction could not be accurately measured through the
hot-cell window because of the small width of this zone and the low magnification.�
metallographic cross-sections of specimens showed that a rim of hydride had
been formed within the region that had been in close proximity to the aluminum
specimen holder.� This caused a small
dimensional change in the specimen that interfered with the accurate
measurements of specimen dimensions needed for the irradiation growth
determinations (Figure 5).� The
hydriding of the specimens was eliminated by a change from aluminum to
Zircaloy-2 specimen holders for subsequent irradiation exposures.� This did not materially change the rate of
growth of the interference-coloured oxide films.
resolution TEM studies of highly irradiated oxide films have not shown any
evidence of irradiation induced point defect clusters or dislocation loops.� The oxides appear to have recrystallised in
the outer regions (older oxide), but not near the oxide/metal interface.� This process must be a solid state process
because nanopores are visible (Figure 6), and are thought to be small helium
bubbles resulting from the (n,") reaction on 16O .
observations of the potentials developed across the oxide during isothermal
oxidation experiments show that the electronic conduction through the oxide
film is more difficult than the oxygen diffusion that determines the rate of
oxide growth.� Thus, the negative
potential on the metal increases, and accelerates the electron migration
(concurrently retarding oxygen ion migration) until the two diffusion processes
become equal.� A steady potential then
persists unless changes in the intrinsic properties of the oxide occur as it
thickens.� These potentials are
determined by the kinetics of the two competing processes, rather than being
thermodynamically determined (i.e. they are not the open circuit potential of
an electrochemical cell), and can be modified by pre-treatments (such as high
temperature pre-oxidation) that alter the balance between the electronic and
anionic migration processes.
the absence of irradiation the electronic resistance is so great that the oxide
film is unable to grow thicker than ~2nm without an externally applied anodic
potential.� The anodic oxide produced
grows to a limiting thickness determined by this voltage and the electrolyte in
which the anodisation takes place (2.5 � 2.7 nm/V).� This is typical of electrolytes in which an impervious barrier
film is produced.� However, some
electrolytes form porous oxide films, which continue to grow in thickness
linearly with time .� The best known
examples of this are oxides grown in nitric acid or nitrate solutions
.� Under irradiation it appears that
the reduced resistivity of the oxide  allows the oxide to grow at low
temperatures (~50�C) under the potential resulting from the free energy of the
oxidation reaction (~2.34V) , and that the films that are formed are
porous, because the oxide growth continues essentially linearly without
saturating.� Regrettably, at the time
that these specimens were being examined no transmission electron microscopy
was carried out because of the very high radiation fields associated with these
reduced resistivity of the passive oxides on both zirconium and aluminum also
permits the cathodic hydriding of zirconium alloys in contact with aluminum in
reactors.� This effect has been observed
previously in the early severe hydriding of Zircaloy-2 flow tubes in the K-East
reactor (KER) at Hanford [21,22].� It
was found to be impossible to simulate this galvanic effect in the laboratory
in the absence of irradiation.� Surprisingly
thick interference-coloured oxides were also reported in the KER tubes ,
but the association with the hydriding, and the respective locations of the
interference-coloured oxides and the solid hydride were not noted.� This work is the first report of the
presence of relatively thick oxides and hydride layers at the same
location.� This indicates that the
cathodic hydriding and anodic oxide forming processes operated independently of
each other, and with relatively little interaction (as indicated by the
slightly thinner oxide on the Zr surfaces in the specimen holder slots).� It is well known that zirconia films cannot
be reduced electrolytically in neutral aqueous electrolytes, and the magnitude
of the oxide thickness reduction in the slots appears to be compatible with the
polarisation curves measured for oxide growth in fused salts .� The common facilitating factor for both
anodic and cathodic effects is the reduced resistivity of the oxide films in
Application to Light Water Reactor
observations allow both the nodular corrosion and "shadow corrosion"
mechanisms in BWRs to be understood.� In
PWRs the high dissolved hydrogen concentrations in the water eliminate the
galvanic potential differences that are necessary in addition to the
irradiation enhanced conductivity of the oxide films if enhanced oxide growth
is to be observed beyond the interference-coloured oxide region�  (Figure 7).� It appears that, in PWRs, all metals operate at the reversible
hydrogen potential  so that galvanic potentials between dissimilar metals
are absent.� In BWRs there is usually
insufficient hydrogen added to the water to achieve this situation so that
galvanic potential differences remain, and can result in large increases in
corrosion rate for Zr alloy components within ~ 5mm. of a dissimilar metal that
will anodically polarize them (e.g. stainless steel, nickel alloys).� The short range of the effect results from
the rather small increase in the water conductivity under irradiation.� This, rather than the increased oxide
conductivity then becomes the rate limiting step .� The "shadow corrosion" phenomena can be readily
explained by the above situation.� The
close similarity of the nodular and shadow corrosion phenomena requires a more
micro-mechanistic approach. Although some of the first observations of nodular
corrosion showed it to occur only in close proximity to stainless steel grids,
with material that was very susceptible to nodular corrosion  the nodules
were often fairly uniformly distributed over the Zircaloy-2 surfaces between
the grids.� The sensitivity to nodular
corrosion was heavily dependent on the precise fabrication route that had been
followed for the batch of cladding .�
This sensitivity to nodular corrosion could be correlated with the
second phase particle (SPP) size in the alloy.�
The number of nodules, however, was always far fewer than the number of
SPPs per unit area of surface.� Thus,
some other selection rule (in addition to just the particle size) was needed to
specify the locations at which a nodule would nucleate.� Many suggestions have been made for what
this selection rule might be , but the frequency of clusters of closely
spaced SPPs seems to be a preferred one .
close similarity between the galvanic factors in nodular and shadow corrosion
suggests that nodular corrosion is a micro-version of shadow corrosion.� In the initiation of a nodule
then, the area
of the " - Zr matrix around a large SPP or cluster of SPPs would be the
anode and the SPPs would be the cathode.�
For a given SPP composition then the galvanic potential difference with
the matrix would be set and the magnitude of the anodic current would be
determined by the cathode area (i.e. the SPP size, or cluster size).� Changes in fabrication route, which changed
the SPP composition, could change the galvanic potential available, and (for
otherwise similar SPP size and distribution) hence change the susceptibility to
nodular corrosion.� At present, there is
insufficient data on SPP/matrix potential differences to allow this hypothesis
to be thoroughly tested.� It does,
however, provide a rationalisation for many of the observations.� The observation that hydrogen uptake by
batches of cladding showing severe nodular corrosion was not significantly
different from that of well behaved batches, is matched by the absence of
enhanced hydrogen uptake in the area of severe shadow corrosion in KKL.� In both instances the cathodic hydrogen
evolution process was occurring at the cathodes and hydrogen was not entering
the Zr.� Other observations that are
compatible with this hypothesis include:-
§19 1. Absence of nodular corrosion, but enhanced uniform corrosion, for
material with very small SPPs .�
These SPPs dissolve rapidly by Fe recoil from fast neutron collisions
 and result in enhanced uniform corrosion in post-irradiation laboratory
tests, and are probably one cause of the accelerated corrosion in PWRs when
oxide thicknesses exceed ~ 10:m.
§20 2. Zr-Nb alloys (both 1%Nb and 2.5%Nb) that do not contain SPPs of the
type present in Zircaloys do not suffer from nodular corrosion unless the Fe
impurity level is very high.
§21 3. The absence of visible radiation damage in oxide films on zirconium
alloys implies no direct acceleration of in-reactor corrosion due to enhanced
diffusion in the oxide.� Thus, reduced
in-reactor corrosion rates become possible if the in-reactor annealing of the
metal structure results in improved corrosion resistance .
observations of enhanced oxide growth rates and simultaneous severe hydriding
of Zircaloy specimens show that the anodic and cathodic processes follow
independent mechanistic rules.� The key
to understanding these phenomena is the enhanced electrical conductivity of the
oxide films under irradiation, and the persistence of galvanic potentials
between dissimilar metals if the reactor water contains insufficient dissolved
hydrogen.� No evidence has been found
for irradiation damage in zirconia films resulting from fast neutron
collisions.� Only nanometric helium
bubbles resulting from the transmutation of 16O by the (n,") reaction are
visible.� These factors permit the
prevalent observation of "shadow corrosion" in BWR water
chemistries.� Severe examples of this
are fortunately infrequent.� Since SPPs
are galvanically dissimilar from the zirconium matrix it is concluded that
nodular corrosion is merely a microscopic variant of shadow corrosion.� In the absence of Fe-containing SPPs, or
after the irradiation induced dissolution of very small SPPs, nodular corrosion
should be absent.� However, irradiation
enhanced dissolution of SPPs should lead to enhanced uniform corrosion in BWRs
and PWRs where this phenomenon will degrade the corrosion resistance of the alloy.
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§24 §25 §26
§27 Fig 1. Potential/time curves for Zircaloy-2 in fused salt at
§29 Fig 2. Effect of pre-oxidation at 445�C on potential/time curves at
§32 Fig 3. Plot of oxide thickness vs fast neutron dose in ATR
§35 Fig 4. Visual appearance of "anodic" oxide formed at 50�C in
§37 Fig 5a. Typical hydride layer formed in ATR during exposure in Al specimen
§39 Fig 5b.
Thickest hydride layer observed formed in ATR during exposure in Al specimen
§43 Fig 6a and 6b. Under and over-focussed images of the equiaxed
grains near the surface regions of an oxide made from B1W19 (Zr-2.5%Nb) at the
5m location, showing nanopores and cracks (arrowed) at grain
§46 Fig 7. Oxidation of Zircaloy-2 in irradiated steam