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Effects of irradiation on the oxidation of zirconium alloys in high temperature aqueous environments
The ~~~rnent~ d&B on ths e&b& of&sion fr~~6~t~ neutron and gamma irradiation on the oxidation of zirconium ~&oyahave been rev&wed. It ie shown that both fast neutron i~~ti~~ and the presenoe of free oxygen or oxidiaing radiolytic species must bs simultaneously pnX4entbefor imy affect CthIL be Observed in high ~rn~~tnm watar. In water where the concentration of oxidising species is low, ap&mens preautocl~ved at 400 “C show e larger ~c~~~~~~tion of the oxidation rate than @cimene p~tr~~ted at lower t0mpi3rSture8, or in the pickled +3ondition. Hydrogen uptake is apparently largely ~~~~~ndent of th8 oxidation process, s;nd is also &h&%d by the presence of axidising species, but ta a different extent thranthe ~&G&ion process.. An h~~~~~~ie based on $r shift in the b&nce between the electronic and ionic transport processes thmngh the f&n during irmdiation {due to the different festivity ctf the two procemes to r&&ion snd ~h~rn~~~ conditions) hi been shown LOaxplain the observations qualitrstivoly,
~~~~o~tion~hyd~~~~~
e& ~pp&~~~~t ind&endante da processus d’oxydation et est au& affect& prarla ptisence d’esp&aesoxydantes mais 8, u.n degr6 diff&en$ de celui du prooessus d’oxydation. Une hypoth&e baa& sur un d6plaeement &ns la balance entre 18s proeessuf3 ds transport ~l~t~~q~ et ion&e & tmvem le film durant ~~~di~t~on (due Lp une asnsibilit$ diff&mte dea 2 proem via-&-vis de la mdia$ion et des conditions chimiqueafa.tstemontr6v s~c~~tible d’expliquer bs ob~~&t~~~ d’un point de vue qualitatif.
EB w0rdt3n~~~~~t~lle D&m i&r den ~~~ von S~ltp~~ten~ sowie Neutronen- und Gammabestra&mg anf die Oxydation von ~ir~~~l0~e~ge~ ~~mrn%nf~nd b0tr~~~tet, Dab& zoigt eioh, dasa sowohl Be&r&lung mit sohnellenKeutronen, al%such freier Sauemtoff o&r oxydierend w~r~~~d~ radiolytiadm-4 M&erial gl#~~~zeitigvorliegenmijseen, damit ein Xffekt in Wasser bai hohen Tern~r~t~n beobaohtet werden kann. In Wasser, wo die Konzentration oxydierender Stoffe gering ist, zeigen Proben, die &ch vorher bei 800 ‘C in einem Autoklaven Lee dorm&a exp&imentales EFLZP l’effet des fragmenti befanden, eine st&rkem Zunahme der ’ Qxydationsals diajenigen Proben, welehe bd de f&&on, de ~i~i~ti~~ par dea neutzwm et des ge~~~~digkeit oder im g&&ten Zustand rayons y sur l’oxydation des alliege~ de zirccsnium tiefertmTempn&umn sont psf&3s en revue. vorbehmdelt wurden. Die Wa~ser&off&hahme ist vom Oxy~~t~o~Bvorg~g; I1 est montr8 qu’B la foie l’irradi~tion par fes im w~~~tli~hen ~~b~~~~ neutrons rapide0 et la. pr&mae, ~oxyg$ne libre ou au& wird sic dumb oxydierende St&e beei~~st~ d’esp&errradiolytiques oxydantes doivent Btm simmaber Bnders als der Oxydationsproze~, Die ~eo~h~g~ri l-en s&h qu&t&iv denten, tai&ment pr&ents avant qu’nn effet pui@3e Qtre ottserve dans l&u & ha&e temg&&.zre. rtthne l’eau wenn eine Verachiebung des ~l~i~~e~~ht~ xwimhen air la co~~~tr~tion des esp&aesoxydantea est f&b&, dem elsktroniaohen und ionischen ~~~o~proz~~ des Bohantillonnsoumis Etuparrwrant B 1’EbuLoolave B bei B&r~hlung sngencnrnnen wird ; diesaVerschiebung 400 “C mont~~t une phrs gmnde acc&%ration de Itt beruht auf der ~ter~~~d~~hen Emp~dlichkeit der vitesse d’oxydation que 1~ ~~h~ntillonB~r~~~~i~ beiden Vorg&nge gegantiber chemisohen und Strahaux temp&&nrea inf&ieuma ah dans S&at &cap& ~hmg~bedingungen. 1
2
1.
B. COX
Introduction
Zirconium alloys have been extensively used as fuel cladding in water cooled reactors throughout the world 19a). More recently the use of thersealloys for pressure tubes in similar reactors has been increasing 3). The lifetimes generally expected for these components are 3-P years for fuel cladding at a temperat~e of 300-340 “C, and up to 30 years for pressure tubes at 250-300 “C. Extrapolations of laboratory data for oxidation and hydrogen uptake (a natural corollary of oxidation in a hydrogenous environment) suggest that, if neither process were accelerated by irradiation, no limitation on component life from either process is likely. The low failure rate of zirconium alloy clad fuel confirms this optimism 2) as far as current operating conditions are concerned ; however, the desire to operate under more rigorous conditions and to guarantee a 30 year life for pressure tubes requires us to make long extrapolations from the security of laboratory tests and current operating experience. SufEcient examples of greater than expected oxide film thicknesses and hydrogen contents have been reported to ensure that acceleration of both oxidation and hydrogen absorption processes can neither be ignored nor readily demonstrated to be within tolerable limits under all desired operating conditions for zirconium alloys. It is important, therefore, that we not only develop empirical rules for predicting the occurrence and magnitude of such effects, but that we gain an understanding of their cause, if the full potential of zirconium alloys in water-cooled reactors is to be exploited. The history of irradiation effects on zirconium alloys is now quite extensive, and it is useful to review this at the present time; firstly because the field has become sufficiently large that many of the experiments which would help towards an explanation tend to be forgotten or ignored, and secondly, because we may be able now to offer an explanation leading to predictions which can be tested experimentally. Thus, it is hoped that this review may lead the reader to devise critical experiments which will take
the proposed mechanisms out of the realms of speculation. It should be borne in mind that, although most of the experimental data refer to the oxidation process and our knowledge of this process is considerably better than our knowledge of the hydrogen absorption process, yet it is the latter process which is generally critical where the permissible operating limits of zirconium alloys are concerned. There are circumstances where the thickness of the oxide film formed is important (e.g. on heat transfer surfaces) or where metal loss may be critical (e.g. the loss in thickness of high strength zirconium alloy pressure tubes over 30 years), but in general the embrittlement of the component by hydrogen is the most feared result of the oxidation process. The oxidation of zirconium alloys in the absence of irradiation is usually considered to proceed initially according to an approximately cubic rate law (pretransition oxidation); this period is followed by a transition to a more rapid, approximately linear oxidation rate (post transition oxidation). The significance of these periods will be discussed in more detail in section 6. 2.
History
2.1.
ENHANCED OXIDATION
IN FISSILE
SOLUTIONS
The first observations of the acceleration of the oxidation rate of a zirconium alloy in an aqueous environment in-reactor go back to 1954 and the initial autoclave and loop tests carried out in uranyl sulphate solutions at ORNL * for the Homogeneous Reactor Project 4). At very much the same time workers at KAPL 5) and Bettis 8) were reporting no effect of reactor ideation on Zircaloy-2 after 2500 h in PII 11 water at 580 “F (305 “C) containing 500 cm3 hydrogenpitre. Thus, apparently contradictory experimental results were present from the earliest stages of the investigation. There was no doubt about the presence of * See glossary of names and initials at end.
EFFECTS
OF
IRRADIATION
ON
THE
OXIDATION
OF
ZIRCONIUM
ALLOYS
3
the
quite accurate for an autoclave without specimens, but needed some interpretation in the
absence of irradiation the oxidation of zirconium alloys in uranyl sulphate solutions (fig. 1) was virtually identical with that in high temperature
case of the more complicated loop tests. Oxygen pressure drop measurements during in-reactor autoclave experiments (fig. 2) showed
water 7). Effort was concentrated, therefore, on understanding the factors affecting the enhanced oxidation, and searching for alloys with im-
that the oxidation-kinetics were linear. A short incubation period was sometimes observed at
proved corrosion resistance. Experiments were carried out either in autoclaves 8) made of the alloy under test (with or without separate corrosion specimens supported within them) or in circulating loops 9) containing groups of test specimens. Whereas in most experiments re-
easily measured 4). This might
accelerated the
oxidation
size of the
in the ORNL
effect
tests since
was so large.
In
ported here oxidation was assessed from weight gain or oxide thickness, in these tests it was assessed : 1. From specimen weight loss * ; the oxide film was either lost during the experiment (loop tests) or could easily be removed from the dried specimen (autoclave tests) ; 2. from the rate of consumption of the oxygen overpressure (required for the stability of the uranyl sulphate solution). The latter technique gave an average value for the oxidation rate * which was generally *
Usually
equivalent
expressed as mils (0.001”) of metal lost;
to 590 mg/dm2 weight
used elsewhere.
An oxidation
gain in the units
rate of 1 mil/y (MPY)
is equivalent to w 1.62 mg/dmg.day
in the other units.
the beginning
of the experiment
but was not correspond
to
the pretransition part of the normal out-reactor oxidation kinetics, but generally the oxidation rate was so high in the ORNL tests that it was impossible to confirm this supposition. The first measurement in fig. 2a lies on the linear rate line so that if the period prior to this point represented pretransition oxidation the weight gain at transition must have been reduced to Q 15 mg/dmr from the normal out-reactor value of m 30 mg/dmz. Over a similar period of time parallel studies were being carried out at AERE 1% ii), where facilities did not permit operation at as high a fission power density as could be obtained in the ORNL tests. Since it was impossible to achieve power densities typical of homogeneous reactor operation at AERE, workers there concentrated on examining the very low power density region, where the change from normal out-reactor oxidation kinetics was occurring. Results from AERE (fig. 2b) showed that the oxidation kinetics in uranyl sulphate solutions were formally analogous to normal out-reactor behaviour. Thus, a pre-transition oxidation period with an exponent in the equation
-I
Aw=ktn+c,
UNIRRADIATED
Fig. lated
1.
A comparison
from
various
of Zircaloy-2
the decrease
in oxygen
corrosion calcuoverpressure
in
uranyl sulphate environments at 250 “C and 290 “C [ref. ‘)I. (1 mil = 25.4 pm).
of between 0.3 and 0.5 was followed by a period with n w 1. The effect of irradiation in fissile solutions was to accelerate both pre- and posttransition kinetics with little change in the exponent of the pre-transition kinetics and a reduction in the weight gain at which the transition took place. Because of oxide apalling, a problem in all uranyl sulphate experiments, weight gains could not be followed far beyond transition. It was concluded, however, that the
4
B.
00X
0.24
0.46
USING
27O’C
DATA
RADIATION
TJME (hr
01 3’Mwl
A typical measurement of corrosion by following the consumption of the oxygen overpressure; ORNL Expt. L52Z-136 [ref. I@)]. (1 mil G 25.4 pm; mpy 3 mil per yeer). are-transition exponent was closer to 0.5 under irradiation than in its absence, and that the
where E and L are constants. was based on a postulated
weight gain at transition was reduced from the expected value of m 30 mg/dm2 to a value of FW20 mg/dma. The ORNL results indicated that this weight gain was reduced even father at higher doses rate. Experiments at ORNL revealed a complicated dependence of the oxidation rate on the fission power density in solution. The oxidation rate was not a linear function of power density but tended to a satiation vaIue at high dose rates. Early efforts to fit this dependence gave a relationship between the corrosion rate (R) and the fission power density in solution (P)raised approximately to the 0.4 power 7912). This was elaborated to give a relationship
corrosion effects in which defects in the oxide lead to a loss of protective oxide. However, after further experimentation 1% 14)
R=EP[l-exp
(-L/&5)],
it was concluded that a better by an equation of the form
This expression model for the
fit was given
l/R = (&/(KfW} + (l/K), where for a given material, K is a constant for a given ~mper&ture and solution, Kl is a constant for a given temperature, and b: is the factor by which the effective power density is greater than the solution power density due to uranium absorption at or near the corroding surface. In this instance the accelerated oxidation was explained on the basis of radiation damage in the metal substrate. Plots of some
EFFECTS
Fig.
2b.
OF
IRRADIATION
ON
TEE
OXIDATION
OF
ALLOYS
5
Irradiation corrosion of Zircaioy-2 in 300 ‘,C nrenyl sulphate solutions followed by weight gain [ref. lo)]. f. IL III. IV. V. VI.
Unirradiated Neutron/Gamma, radiation only Power density 0.0125 Watt/ml. Power density 0.025 Watt/ml. Power density 0.0375 Watt/ml. Power density 0.075 W&ttjml.
28-
iI P
APPROXIMATE
FOR EXPERIMENTS
I
’
I
1
I
’
I
I
1
,
1
,
l
* 28O’C. NoACID
. .l I//__
-
-2o-
? I &SB 92-
CONDITIONS
Conditions: 300 ‘C : 200 psi 00 (cold) Solution: O.OSM KrSO4; 0.02M HaSOr; 0.005M CuS04+=5UOaS04 to required power density.
~~
-24-
-
R1 J)
a/t4/
fI=i*5p 25O*C,WITH
0
2
4
6
8
FlS6K&YA’Ek4DEN&Y 10
18
(W/ML)
Fig. 3%
ZIRCONIUM
Some ORNL
autoclave
(hi@“’ H&O,
R’,‘)
20
22
results for Zircaloy-2 in uranyl snlphate solution per year = 26.4 ,um/year).
24
26
[ref. la)]. (mpy = mil
EXPERIMENT
SOLUTION
Fig. 3b.
FISSION
POWER
DENSITY
A
o
L-2-19
B
,
C-4-16
8’
v
L-4-16
c
A
L-2-15
E
t
L-2-10
(ft
per
six)
b/ml1
Some ORNL in-reactor loop results for Zirceloy-2 in urenyl sulphate solution at 280 “C [ref.
le”)l.
0.60
0.45
0.40
0.35 h ;
0.30
.-_ 5 0.25
EXPERIMENT
0.20
(ff per sc
O.iS
o.*o
Q(H
0
t SOLUTMN
Fig. 4,
ORNL irradktion
F,SStON POYlER DENSITY
(ml/w)
corrosion data in umnyl sulphate solution at 280 “C fitted to eq. (2) [ref. la)], (I mil 3 25.4 pm).
EFFECTS
of the ORNL
OF
IRRADIATION
results according
ON
THE
oXIbA!cION
to the above
tests
OF
were
ZIRCONIUM
carried
out
7
ALLOYS
with
uranyl
sulphate
relationships are shown in figs. 3 and 4. A wide range of zirconium alloys was examined at ORNL in an attempt to find alloys less affected by irradiation in fissile solutions. However, the
solutions depleted in z%U, in sulphate solutions containing no uranium, and in non-fissile solutions with hydrogen instead of oxygen
range of variation from alloy to alloy was small (fig. 5), and results on unalloyed zirconium were
major source of energy deposition was from the fast neutron and y radiation from the test
comparable
reactor (MTR).
with those for Zircaloy-2
overpressures 437915). In these experiments
7). Thus,
the effect of irradiation in fissile solutions largely swamped the variations present from alloy to alloy out-reactor, and suggested that hopes of producing zirconium alloys resistant to this type of attack were likely to be remote. During the course of these studies autoclave
It was observed
rates in HzO, DzO and depleted
that oxidation UOzSO4 with
oxygen overpressures were identical with those in other tests when plotted as a function of the fast neutron power density (defined as the energy deposited per cm3 of the solution from fast neutron interactions).
ALL COUPONS DEFILMED IMPACT SPECIMENS NOT DEFILMED
“A-
900°C,
3hr.
WATER
QUENCHED
6 -900°C. 600°C,
3 hr. WATER OUENCHED 46 hr. AIR COOLED
C - 65OOC.
30 min
D-900’C. 600°C.
3hr. WATER 3 hr. WATER
;
OUENCHED: OUENCHED
0 0
2
3 POWER
the
DENSITY
Fig.
5e
4 (w/ml)
5
6
8
3. C~MER~~ALLY
a-2
0.2
ANNEALED
Zr-tS
Nb
9Oo.C‘
2-3
Zr-fS
Nk
SOCTC,
Z-3%,
WATER
QUENCHED;
4oO’C,
2 week,
AIR COOtED
Mb
9OQ*C,
2-Jhr,
WATER
QUENCHED;
4oO’C,
i week,
AIR COOLED
40Q”C,
1 week,
AIR
COOLED
SOO’C,
f wwk,
AIR
CCCLED
Zr-
“0
00x
(5
WATER
Ouf NCHED
Zr-fS
Nb-
9OcJ*c,
z-3flr.
WATER
QUENCHED;
Zr-f5
Nb-tMo
9QO°C,
2-3hr,
WATER
OUENCHED;
Zr-fS
Nb-fCu
S’OO’C,
2-3
hr,
WATER
QUENWED;
Zr-f5
Nb-f
9OO’C,
2-3
hr,
WATER
QUENCHED;
0.4
ZMo
hr,
Cu
0.6
0.8
f.0
, SOLUTION
FISSION
POWER
DENSITY
1.2 (*/ill
1.4
i.6
i’
Fig. Eib
Fig. 5% b,
ORNL results for corrosion of & munber of euperim%ntal zirconium alloys in m%myl f3uIpllata
App~nt exceptions to this were the experiments in water at 290 “C with a hydrogen overpressure, which showed oxidation rates little different from those during normal laboratory tests. The apparent discrepancy between the published oxidation rates in these tests and out-reactor oxidation rates arises from the ORNL assumption of linear kinetics throughout their tests and the calculation of an oxidation rate on this basis, If the experimental data are converted to a weight gain, they lie very close to the normal out-reactor oxidation curve (fig. 6). These tests, not well knowu to investigators working only in high ~m~~rature water, provided the first clue to the apparent dis-
crepancy between results obtained in water at different establishments.
Throughout the period 1956-1960, experiments in high temperature water loops 1**17) and examination of fuel cladding from pressurised water reactors 18) had failed to reveal any acceleration of zirconium alloy oxidation by reactor irradiation in (pH10 LiOH with M 30 cm311 hydrogen overpressure (although results from the ~xa~nation of fuel cladding did not become available until early 1962). The ORNL experiments with oxygen overpressure had ~ndio~~d, however, that at equivalent rates of energy deposition in the solution fast
EFFECTS
OF IRRADIATION
ON THE OXIDATION
OF ZIRCONIUM
ALLOYS
1000 1
1
I$ _ 00
VBWR
8
RECENT ORNL (OPEN
DRESDEN DRESDEN
FUEL
CLADDING
CLADDING
AUTOCLAVE TESTS + O2 .CLOSED +
IN MTR Hz1
BETTIS
DATA
FOR
, OYlOATlON /
.’
I
Ill
OUT --REAiiOR 290%
AT
.’
-L
1
IO00
TIME Fig.
6.
Oxidation
in neutral boiling water at m 300 “C of Zircaloy-2 compared
with ORNL
neutrons were as effective as fission fragments at accelerating the oxidation of zirconium alloys I4b). It was some time before the fact that all PWR studies were carried out with high hydrogen over-pressures was correlated with the low oxidation rates in ORNL experiments with a hydrogen overpressure and a possible explanation of the apparent discrepancy between results in water and uranyl sulphate suggested. Subsequent examination of fuel cladding from the Shippingport and Saxton reactors has failed to detect any increased oxidation during in-reactor exposures of up to 1157 days at full power 18119). 2.3.
(days)
EFFECTS OF VARIOUS BOMBARDING SPECIES
Concurrent experiments using other bombarding particles were carried out in an attempt to determine the relative efficiencies of the various types of radiation. No accelerating effect of electron irradiation on Zircaloy-2 was
in-reactor
fuel cladding from VBWR
and Dresden
data.
observed by workers at ORNL ro), but Glick at Bettis reported acceleration of the oxidation of zirconium by a factor of ten in some experiments with deuterons 20). The ORNL work was carried out in oxygenated solutions and appears to be good evidence of the negligible effect of electron (and presumably also y) irradiation in the absence of fission fragment or fast neutron irradiation. The Bettis work is more ambiguous; it was conducted in nominally degassed water at 315 “C, but since the critical effect of water chemistry was not appreciated at this time it is likely that conditions were relatively oxidising during bombardment. Thus deuterons, like neutrons, will accelerate the oxidation of zirconium under suitable experimental conditions. Calculations of the relative efficiency of various types of radiation in producing damage in ZrOz in the form of interstitials and vacancies 21) did not give good agreement with the experimental oxidation data. However, the ORNL results in fast neutron fluxes 15) gave
10
B. COX
hope that an empirical correlation might be used in which power densities for fission fragment and
lowered, but satisfactory post-transition data could not be obtained for specimens originally
fast neutron damage were equated for purposes
in the pickled condition
of calculating
without long in-reactor
the oxidation
rate. Such a calcu-
exposures
lation led to the prediction
that a fast neutron
cant) decision was made to establish the degree of acceleration of the post-transition oxidation
flux of 101anf/cmz. set should lead to an oxidation rate for Zircaloy-2 equivalent to a fission power density
of
0.015
W/ml
solution 21). Experience
in
uranyl
sulphate
had shown that such a
power density led to an increase of about a factor of two in the pretransition weight gain of Zircaloy-2 at a given time, with a correspondingly larger factorial increase ( x 8) in the post-transition rate lop ll). Although exposure of Zircaloy-2 to PWR conditions, at considerably higher fast neutron fluxes, had not given any evidence for such an acceleration in the oxidation of the Zircaloys, the experimental details of this work had yet to be published and only verbal statements of this observation were available. Accordingly it was decided at AERE to institute a deliberate search for such fast neutron effects 14121).
3. 3.1.
Enhanced oxidation of Zircaloy-2 in irradiated water and steam EXPERIMENTS AT AERE,
HARWELL
Small 1 atm steam circulating loops for insertion in the DID0 or PLUTO reactors were built to employ specimens which could be handled easily when active 2% 23). That fission fragments and therefore presumably also fast neutrons caused acceleration in other than the aqueous environments had already been shown by the results in air of Yee et al. 24). Initial experiments both at AERE and elsewhere employed predominantly Zircaloy-2 and it is with respect to this alloy that the ensuing discussion is directed. The experiments in a high fast neutron flux (w 3 x 1013 nrvt) in the DID0 reactor soon revealed an increase in oxidation under irradiation at 340 “C of about the predicted magnitude 2% 25). Pre-transition oxidation rates were accelerated by about a factor of two, the weight gain at transition was apparently
(fig. 7). Thus a further (very signifi-
rate of Zircaloy-2 from specimens taken beyond the transition point at high temperature (450 “C) and then continued
at the normal in-
reactor experimental temperature (340-350 “C) until the “memory effects” had largely disappeared. Experiments using this type of specimen (figs. 8, 9) showed accelerations of up to a factor of ten in fluxes of m 3 x lOi n&m2 . set se-2a). Using these techniques the irradiation enhanced oxidation was shown to disappear with increasing temperature. The acceleration was barely detectable, above the normal oxidation rate, at 400 “C, and was definitely absent at 450 “C! (table 1). Earlier tests in high pressure steam in the BEPO reactor, which apparently showed an effect at temperatures above 400 “C were shown to have resulted from faulty positioning of thermocouples 2% 25). At lower temperatures there did not appear to be a progressive increase in the acceleration factor with decreasing temperature in experiments at high fast neutron flux. At both 300 and 350 “C post-transition oxidation rates appeared to be increased by factors of 8-12 (table 1). The effects of neutron flux and neutron/ gamma flux ratio were investigated in several positions in the DID0 and PLUTO reactors (table 1). These showed that the increase in oxidation rate diminished with decreasing fast neutron flux, and that gamma flux was unimportant. The fast flux in the BEPO reactor was too low to produce any effect 28-28). Experiments in high pressure steam (500 psi) in a relatively pure gamma flux (DID0 reactor cooling pond) apparently showed a small effect (fig. 10) on the early stages of oxidation [interference colour region 25)]. This was ascribed to an effect on the electrical properties of the oxide which were thought to control this stage of the process. Subsequent work on the electrical
EFFECTS
OF IRRADIATION
ON TEE
OXIDATION
OF ZIRCONIUM
BASED REF.
SHOW
II
1
I
I
STANDARD
I
Fig.
7.
11
cg
125
DEVIATION
I
I
IO TIME
ON
ALLOYS
I 100
I
I
I
I 1000
(DAYS)
Effect of irradiation in the DID0 reactor (w 3 x 1013n&m2*sec* , w 5 x 108 R/h y) on the oxidation of Zircaloy-2 in atmospheric pressure steam at 340 “C [ref. 26)].
l undar
radiation
Fig. 8. Effect of irmdktion in the DID0 retitor on the oxid&ion of pre-oxidised Zircaloy-2 in atmospheric pressure steam st 300 “C [ref. 26)].
properties of ZrOz (section 5.6) showed that an effect of y radiation on these was unlikely, and the observations were discredited. However, the subsequent experiments, purporting to show no effect of y radiation under similar conditions 2*), were not comparable as they were performed in 1 atm steam rather than in high pressure steam. If as now seems likely there may be a chemical component to the irradiation effect, then radiolytic species capable of producing an acceleration in the oxidation would
be more readily available at the specimen surface in y-irradiated steam at high pressure, than at 1 atm (section 5.8). The AERE observations of an acceleration of Zircaloy-2 oxidation by a fast neutron flux in a relatively pure aqueous environment remained unique, and largely unbelieved, during the years (1960-1962). This period was devoted to demonstrating that errors in the knowledge of the specimen tempaerture could not account for the observed effect. However, two additional experimental programmes eventually provided information supporting the AERE work. These were the examination of fuel elements from the reactors VBWR and Dresden-l 29-31), and the examination of specimens from the Zircaloy-2 surveillance programme in the Hanford G-7 loop in the ETR reactor at Idaho Falls 32).
3.2.
EXAMINATION
OF FUEL CLADDINU FROM
BOILING WATER REACTORS
The examination of Zircaloy-2 fuel cladding from VBWR and subsequently from Dresden which had operated in neutral boiling water with radiolytic hydrogen and oxygen showed oxide films much thicker than anticipated (fig. 6). These increased approximately linearly
12
B.
COX
95 -
“; 900 E” I- as3 3” 80z TYPICIL
s ?I 75-
CONTROL
SPECIMENS
m : -
_________-----
_--_--
I___-__--
70-.
r_-_-_-_
___-- -60-
(FIGURES 1 125
I IS0 TIME
Fig. 9.
- -
Effect of irradiation in the DID0
IN
e-
-_-_=---=_-%
______-_------
IN ABSENCE OF RAD~AT,oN UNDER RADIATION PARENTHESIS
I 175 (EXCLUDING
ARE
I 200 PM-OXIDATION)
FOR
(f0)2)
I 225
DAYS
reactor on the oxidation of Zr-2.6 wt y0 Nb in atmospheric! pressure steam at 300 “C [ref. 28)].
Fig. 10. Effect of a gamma flux on the oxidation of Zircaloy-2 by 500 psig steam at 350” C [ref. 26)]. Billet Z 8000: l Gamma flux 105 r/h; A Gamma flux 7-9 x 105 r/h; H Gamma flux 2-5x 106 r/h.
EFFECTS
OF IRRADIATION
ON
THE
OXIDATION
OF ZIRCONIUM
ALLOYS
13
TABLE 1 AERE,
Harwell results for the factorial increase in oxidation rate of Zircaloy-2 under irradiation in atmospheric pressure steam Post-transition region
Pm-transition region
1 1.5
Fuel element rigs 2-3 x 1013 nfvt 5x108 R/h (y)
(up to 6) s-s* 8-12 *
1 1.5? *
2V rigs 2 X 1012 nfvt 3x10’3 R/h (y)
4-9
6-8 *
450
1
1
400
1t
1
350
1
1
300
* t
I
1
Ampoule rigs in BEPO 1 X 1011 nfvt 2 x 106 R/h (y)
I
For Zr-2.5 wt% Nb. Similar result obtained in 02 at 80 mm pressure using balance rig.
with exposure time in-reactor to values of M 700 mg/dm2 at 1200 days. Happily the hydrogen concentration in the fuel cladding at any given time was not significaqtly higher than was expected from an extrapolation of laboratory data. Although the amount of oxidation had increased by a factor of w 50, the fraction of the hydrogen released by the oxidation which was absorbed had decreased so as to approximately balance this. Thus, the main practical concern was about the
effect of the thick oxide film on the heat transport properties of the fuel cladding, and the possibility of runaway oxidation due to the rise in temperature at the metal/oxide interface 33-35). 3.3.
EXPERIMENTS LOOPS
IN PRESSURISED WATER
Examination of the specimens from the G-7 loop in the ETR reactor at Idaho Falls also showed thick oxide films3% 35-39,and in this
14
B.
1
a.
Zircaloy-2
COX
Lot AT
200
200 Time
at Temperature,
Days
ON
instance also high hydrogen concentrations, although nominally operating under PWR conditions of pHl0 LiOH but without added hydrogen. Weight gains and film thicknesses increased approximately linearly with time, at a given neutron flux (fig. 11) but were generally greater than the weight gain expected from AERE studies for oxidation under irradiation. The effect of small differences
in fast neutron
flux from experiment to experiment could be allowed for by plotting the results as a function of dose (fig. 12). The oxidation rate did not increase steadily with increasing fast neutron flux, however, but showed evidence of saturation between 1013and 1Ol4n&m2.sec (fig. 13). Thus plotting results as a function of dose only corrects for experiments at fluxes below the saturation level where weight gain is apparently a more nearly linear function of flux. During this period (1960-1963) experimental work at CRNL failed to give a clear indication of irradiation effects sgl 40), some tests apparently showing them and others not. The fast neutron flux in these experiments was known
1000 1 -
n
AT-50
.
HT-37
_ A CT-19 _
5
100
0
AT-50
0
HT-37
A
CT-19
Irradiated G-l
In
Loop
Unlrradiated Controls
7
2 k
10
1
100
Time
-
500
iDays
(b)
Typical results for oxidation of ZircaFig lla, b. lay-2 in water with LiOH or NHdOH additions at 280 “C a
a function ETR
of time in the G-7 loop in [ref. 469 *8)].
EF’FECTS
.
OF
AT-50
IRRADIATION
ON
TEE
Irradiated in O-7 LOOP
HT-37 CT-19 AT-50 0 HT-37
Unlrradirtsd Controls
c
Out-of-Rerctor Lot (5700 firs.)
2
5
2
10
E
Out-of-Rtrctor (500 Hrr.)
Loop
,* 1.0
0.1
Integrated
10
Exposure
- tnvt
40
x lO_zOf
Fig. 12. The results from fig. llb, obtained in @Xl0 LiOH in the G-7 loop in ETR, plotted as & function of fast neutron dose [ref. a*)].
OXIDATION
OB
ZIRCONIUM
ALLOYS
15
to be relstively low, however, so that not much weight was laid on these apparent disorepanoies. A few examples of oxide films thicker than expected were also observed by Bettis workers 41). These were mainly during loop tests, and predominantly when defected fuel, or fissile mate&l from earlier defects, had been present in the loop. This led some workers to propose that enhanced oxidation was only observed in the presence of fissile material in the coolant. They apparently arrived at this viewpoint largely by rejecting the results of oxidation tests where fissile material could not conceivably have been present, on the grounds that only tests in high temperature water were relevant. Most of the exceptions to this thesis [work at Harwell 22,23, 25, 26, 52), C1fJ.E 49-51)] were not in this environment. Nevertheless it appeared that most of the then available d&a could be explained if both i~adiation and a suitable environment were needed before enh&ncement of oxidation by irradiation could occur. The critical feature of
= 60 -
0
26
@
23
0
19
I
16
0
l
0 40 / -
// /
2o -1
0 x 1013
2
4 Neutron
Fig, 13.
6 Flux
(>
I MN),
8
1 x 1014
II”
Some oxidation results at 280 ‘C from the ETR loop for water (pH10 LiOH) containing m 1 ppm 0~ and < 0.1 ppm 08 plotted as a function of fast neutron flux [ref. s’)].
16
B.
the environment seemed to be the presence or absence of free oxygen, or oxygen containing radicals produced by radiolysis 42). Thus in normal
PWR
conditions
the
presence
of
hydrogen suppresses the radiolysis of the water, whereas in a BWR, taining
radicals
oxygen
are normally
and oxygen present
oon-
in oon-
siderable quantity al). Other experimental results could be rationalised on the same basis, with the exception of the G-7 loop results in ETR. Here the original information was that the water was degassed 43); however, subsequent investigation showed that oxygen was being admitted with the feed water, and that the average oxygen concentration was 0.8 - 1.0 ppm 3% 44). Although the diverse results then available could be correlated assuming that no aooeleration would be observed without the combined presence of fast neutron irradiation and suitably oxidising species in the environment, much work was needed to demonstrate that this correlation would hold for future work. Programmes were started to study the effect of water chemistry on the phenomenon, and the relative magnitudes of the effects of oxygen content and fast neutron flux. The question of whether radiolytio species alone could accelerate oxidation required testing as some of the initial experiments suggested that a “downstream” effect existed outside the neutron flux.
Etched
10ZO Integrated fastNeulron
1021 Flua.
> I MeV
Fig. 14. A comparison of in-reactor oxidation of Z&alloy-2 in the three G-7 loop environments [ref. 45)]. Note : In-flux weight gainsreported only for specimens prefilmed in water at 300 “C, except for etched specimens exposed in pH-IO NH40H pH-10 LiOH, N 1 ppm 02: x W. A. Burns 37) + A. B. Johnson, Jr. 45) A Johnson and Irvin 38,49 PH-IO LiOH, ~0.1
ppm 02: o W. A. Burns 37)
pH-10 NHaOH, < 0.05 ppm 02: q A. B. Johnson, Jr.
showed that the irradiation effect was virtually FACTORS AFFECTINQ THE ENHANCEMENT eliminated for specimens initially in the etched OF OXIDATION BY IRRADIATION condition (figs. 1416). Specimens which had A joint CRNL/Hanford programme for the been preautoolaved in steam at 400 “C before insertion, however, still showed an apparent use of the G-7 loop in ETR was instrumental effect of irradiation (figs. 15, 16). in demonstrating the significance of water Further investigation showed that the magnichemistry. Initial work by Hanford at 280 “C tude of the apparent irradiation effect was showed that addition of hydrogen to the dependent on the precise preautoolaving treatpH10 LiOH reduced the average oxygen ment (figs. 15, 16). Pre-autoolaving at 400 “C content from w 1 ppm to < 0.1 ppm, and gave a much larger effect that preautoolaving reduced the magnitude of the observed irradiat 300 “C, and at any temperature the size of ation effect (fig. 14) 37). Further improvements the effect was larger for thicker initial oxide to the water chemistry in the joint programme films. This observation has a profound effect involved rigorous degassing of the makeup on the interpretation of earlier AERE data, water, and ammonia additions for both PH and where most results had been obtained on oxygen control. The results 45, 46) of these tests 3.4.
EFFECTS
OF
IRRADIATION
ON
THE
OXIDATION
OPEN CLOSED
OF
ZIRCONIUM
ALLOYS
17
SPACES-OUTREACTOR SPACESIN -REACTOR
-PRETREATMENT
Fig. 15.
Effect of preautoclaving treatments on the oxidation of Zircaloy-2 in high and low oxygen water in the G-7 loop [ref. 45)].
specimens taken beyond transition at 450 “C and then further oxidised out-reactor at 340 “C. Failure to observe any effect of preautoclaving in the earlier Hanford experiments in oxygen containing water probably resulted from two factors. Firstly the high weight gains in most of the earlier experiments were accompanied by appreciable scatter in the results, which could have obscured the effect. Secondly, preautoclaving treatments in the earlier tests had generally been at 300 “C, a temperature subsequently shown to give only small increases in oxidation under irradiation. At CRNL loop tests were performed in boiling and pressurised water at M 280 “C to evaluate the effect of ammonia additions on the observed oxidation enhancement under irradiation ~*a*). Results supported the hypothesis that a reduction in oxidising species in the loop water reduced the observed enhance-
ment of oxidation (fig. 17). However, even loop tests where no radiolytic oxygen was detected in the loop water at the outlet from the test section of boiling experiments still showed some enhancement, when ammonia additions were only just sufficient to eliminate detectable oxygen. Higher ammonia concentrations were needed before the irradiation effect disappeared. Thus it appears that transient radiolytic species are at least as important as molecular oxygen in determining the effect. Following the Hanford observation of effects of pretreatment the same phenomenon was observed in the results of the CRNL tests 48). In these tests, however, the effects were all relatively small due to the generally low fast neutron flux, and pretreatment effects were therefore not immediately obvious. Early Hanford experiments 36)had apparently shown accelerated corrosion of some specimens
18
B.
COX
OPEN
SQUARES
OUT- REACTOR
CLOSED
SQUARES
IN - REACTOR
r
Fig.
16.
Effect of preautoclaving
t,reatments on the oxidation in the G-7 loop [ref.
of Zircdoy-4
in high and low oxygen water
45)].
zzzj
loo-
IN-REACTOR LOOP
go-
SO-
OUT-Of-FLUX WITH wITwuT
0X”SEN OXYSEN
OUT-REACTOR
LOW
IN-FLUX
0
0
Q
IJ
n
•p
70-
n
605040 -
Fig.
17.
Results
obtained
at CRNL
for the oxidation of Zircaloy-2 at 270-290 With various water chemistries [ref. 491.
“C in in-reactor
loop t,ests
EFFECTS
OF
IRRADIATION
ON
THE
downstream of the test section; however, these results were not confirmed in later experiments 45). At CRNL, specimens at the outlet to the test section also apparently showed evidence for increased oxidation in some experiments. The differences between these and other specimens were small, however, and it was not possible to state positively that they were outside experimental scatter. Results from the CISE loop in the Avogadro I reactor showed that downstream effects were absent for pickled specimens in a steam-water fog at 280 “C 49-51). This loop was used solely for oxidation specimens (no fuel specimens were present) and so they could be distributed throughout the inand out-of flux regions. (In most other loops the oxidation specimens had taken second priority to fuel tests). CISE results showed accelerated oxidation in a steam-water fog when no additions were made to control radiolytic species (fig. 18). The results at different neutron fluxes show a peak in the weight gain curves at approximately the peak neutron flux (fig+ 19). Hanford workers 46) have observed that specimens with thick oxide films from previous exposure in oxygen-containing water continued fo oxidise at the same high rate when added to a test where ammonia was added to suppress radiolysis. Fresh specimens in this test, showed low oxidation rates (fig. 20). At AERE, irradiated specimens soon showed a normal outreactor oxidation rate when transferred to
Fig. 18. CISE results for the oxidation of Zircaloy-2 in a water-steam fog at 280 “C in-reactor plotted as 5 function of time [ref. “11. Open points, out-reector; closed points, in-reactor.
OXIDATION
OF
ZIRCONIUM
19
ALLOYS
c 9.6 d
27.56
.. .
. \
*
.
.‘
I
~~ 10.5d
I
t
1
*
37.36
I
h
fl
I
I
1.
2
3
4
10
15
20
Neutron
flux
(x10’*)
Weight
rnsj&
gain
Fig. 19. CISE results for oxidation by a water-steam fog et 280 “C in the Avogadro I reactor plotted in comparison with the neutron flux [ref. 49)]. I, 9.6 d&ys; +, 18.5 days; a, 27 days; o, 37.3 days.
conditions 2s). These observations can be explained if the importanf region is that immediately adjacent to the interface at which the reaction is proceeding. For thick porous oxide films this region will be near the metal-oxide interface and changing the chemistry of the environment may have little effect on the production of radiolytic species within the oxide. That, the production of these species is important is shown by the rapid return to normal oxidation kinetics when irradiation is discontinued. Further additions to our knowledge from recent AERE experiments with oxygen and hydrogen additions have been slight, since it does not appear to be possible to control the chemistry of the steam in the atmospheric pressure steam loops or the water chemistry (at present) in the recently designed in-reactor autoclave 52). The experiments (fig. 9) which unirra~ated
showed no effect of adding oxygen and hydrogen to the atmosphere
pressure steam loops may
be unreliable z8). The gases
were added to the -
20
B. COX 1600
1400 N
E
_
.z t z 0) 5 3 2 E E
pH-10 LiOH
1200
z P
pH-10
NH40H
1000
800
600 s AT and
HT
400
200
0
4
8
12
16
Integrated
Fig. 20.
ETR
Out-of-Reactor
I
0
20 Fast
28
Neutron
Flux,
32 nvt
;x
I
36
40
I 44x1020
nvt
> 1 MeV
results showing the effect of changing the waker chemistry during the experiment differences between various batches of material [ref. 45)].
test section, but due to the slow rate of recombination in the gas phase, the oxygen content may not have changed significantly. In addition, these tests were carried out with pretreated specimens already bearing thick porous oxide films ; under these conditions Hanford results show that no change in oxidation rate would be expected on changing the chemistry. Variations in the effect of .neutron flux have been studied at AERE, but the number of points on the neutron flux distribution is insufficient to gain other than a general impression of the flux effect (table 1). 3.5.
24
MISCELLANEOUSEXPERIMENTS
A collaborative programme between AERE and EIR, Wiirenlingen, Switzerland, will utilize the high pressure steam circulating loop at the latter site. Initial experiments were in steam at 450 “C, a temperature at which no effect of irradiation would be expected (based on AERE and ORNL tests). It was therefore not surprising that none was observed 5s). Subsequent experiments at 350 “C have given contradictory results in steam containing relatively low oxygen
and
concentrations and a fast neutron flux of iu 5 x 1Ol2 n/cm2. Experiments in gaseous environments at ORNL 54) and Hanford 55) have shown that accelerated oxidation can be observed in almost any moist gaseous environment, whether the major component is helium, carbon dioxide, or doubts about the specimen air. However, temperature in the Hanford tests, and the recent knowledge about the importance of environment chemistry, render these experiments of little significance in the investigations of the irradiation effects on oxidation. Asher 53) has also reported irradiation enhanced oxidation of Zircaloy-2, but not of Zr-2.5 wt o/o Nb in oxygen at atmospheric pressure. Other experiments have been carried out by Metallgesellschaft AG (in the GKSS reactor near Hamburg) 569 57) and Saclay 58) in small high pressure autoclaves. Problems associated with experiments of this type, where water volume was small compared with exposed surface area (autoclave + specimens) and no control of the conditions inside the small autoclave was possible, render the results of
EFFECTS
these experiments autoclave
designs
OF
IRRADIATION
ON
THE
OXIDATION
OF
ZIRCONIUM
21
ALLOYS
of little value. The in-reactor at AERE
and
GKSS
are
being modified to permit control of the chemistry of the contents. The first experiment in the AERE autoclave (without controlled chemistry) showed large weight gains and unusually high zirconium concentrations in the residual water. This was apparently a colloidal solution zirconium produced during irradiation 52).
of
Many miscellaneous pieces of information have come from the examination of fuel cladding, reactor pressure tubes and other components. Generally the actual operating conditions are not known with any precision, so that it is not possible always to fit these results into any scheme. NPD fuel cladding exposed during the period of reactor operation when the hydrogen (deuterium) concentration of the loop water was above 1 cma/kg has generally shown black oxide films. Fuel removed more recently, since the deuterium concentration has been allowed to fall to its natural value of < 1 ems/kg from corrosion of the mild steel circuit, has shown a significant number of instances of white oxide spots 59). The areas covered with white spots were often adjacent to wear pads and/or wire wrap, where flow conditions may be disturbed, and local chemistry may be different from the bulk (fig. 21). Further examples of thicker than anticipated oxide flms have been found on fuel cladding, corrosion coupons and other components (driver rod housing, etc.) from the HWCTR 60) at Savannah River. The precise operating conditions of these components are not well documented, and it is possible that here too the local water chemistry was sufficiently different from the bulk to permit radiolysis (“oxidising conditions”). Thick oxide films have also been observed on pressure tubes and fuel cladding from the PRTR 61) reactor at Hanford. Here the film thickness was found to vary from end to end with a maximum near the flux centre line and minimum at the ends. Again precise water chemistry information is not available. One example of a thick oxide film on a pressure
Fig.
21.
NPD
Post-transition
fuel
cladding
white
(bundle
oxide
0988)
observed
after
during a period of low hydrogen concentration water
(pH10
LiOH,
on
exposure in the
SW280 “C). Magnification
15 X.
[ref. so)]. The horizontal bar is part of the wire wrap, with one spot weld showing.
tube from the KER-1 loop was also observed 62363). Subsequent examination. of Zircaloy-2 specimens which had been inserted in the primary coolant of the Swedish Agesta reactor 6% 65) (pressurised heavy water at 220 “C) has shown little difference from specimens prepared in the laboratory, either in weight gain, or in the appearance of the oxide film after the initial period in which high weight gains were observed. Examination of fuel from in-reactor loop tests (not containing corrosion specimens) at CRNL has produced some examples of thicker than expected oxide films, although again the precise operating conditions were not well characterized ‘33-6s). Studies in the USSR have been almost universally on a Zr-1 wt o/o Nb alloy, and as such will be reported in the next section. 3.6.
ENHANCED IN WATER
The
OXIDATION AND
alloys most
OF OTHER
ALLOYS
STEAM extensively studied, other
than the Zircaloys, are Zr-Nb
alloys.Work
on
a Zr-2.5 wt o/o Nb alloy at AERE 2*), ORNL 4s) and Hanford 45); on a Zr-3 wt oh Nb-1 wt o/o Sn alloy at Hanford 45) and Metallgesellschaft 57);
22
B. COX
and on Zr-1 wt y@ Nb and Zr-2.5 alloys (some with boron USSR s@+70) lead generally
wt y0 Nb
additions) in the to the same con-
elusions. These alloys are generally susceptible to increased oxidation in the presence of free oxygen in the environment in the absence of
irradiation. Thus, more pronounced
effects of
water chemistry in-reactor are observed than was the case with the Zircaloys (figs. 22, 23). The generation of oxygen-containing species in the test section of a loop can and does lead to enhanced oxidation of these alloys downst~am, whereas no such effect has been positively demonstrated with the Zircaloys. Other than the direct chemical effect of oxygen present, or generated, in the loop there appears to be no added effect of neutron i~adiation on Zr-2.5 wt OjoNb a~cimens. This is true whether pre-autoclaved or not, and irrespective of heat treatment, although the susceptibility to small amounts of oxygen in the water varies considerably with variations in metall~gic~ ~nditions. In practice a small, but definite, decrease in oxidation rate is observed in the test section when relatively
II
I
II
I
I
I
oxygen-free conditions are present (figs. 22, 23). The same situation may occur in the presence of oxygen, but it is obscured by the high weight gains and experimental scatter. Similar effects of molecular,
or radiolytic
oxygen in the environment have been observed in the USSR with both 1 wt y@ and 2.5 wt y0 Nb alloys. The magnitude of the effect is larger for the higher niobium alloy, as would be expected from out-reactor evidence on the effect of added oxygen on the oxidation of these alloys 71). The additional damage to the oxide resulting from a-particle bombardment when 0.1 wt o/o boron is added to these alloys does not apparently increase the oxidation rate. It cannot be assumed that no effect would result if the boron were added to Zircaloy-2, however, as there appears to be no direct effect of fast neutron damage on the oxidation of Zr-Nb alloys anyway. A few experiments at BNWL have included specimens of Zr-3 wt o/o Nb-1 wt y. Sn alloy. The sensitivity of this alloy to added oxygen in the absence of radiation has not been well characterized so that it is difficult to comment on the loop results (fig. 23). However, if it
11
11
11
1
}
11
1
~4D_H)2030405060708090100110t20130140150160m3180190_ EXI’DWRE TIME - DAYS
Zr-2’lr
% Nb
IDD“E P so-
Fig. 22.
Results
P a
SD-
z$
70-
F 3 G
5D-
s
40-
50-
obtained at CRNL for the oxidation of Zr-2.6 wt y0 Nb alloy in in-reactor loop tests (at F* 280 “C) with various water chemistries [ref. 491.
EFFECTS
OF
IRRADIATION
ON
THE
OXIDATION
OF
ZIRCONIUM
ALLOYS
23
OPEN SQUARES-OUT-REACTOR CLOSED
SQUARES -
IN - REACTOR
r -Zr-3Nb-ISn
-
+Zr-2!$
Nb-
B-QUENCH ; 30KCW 24hr 500°C Ann Fig.
23.
ETR
results showing
the effect of pretreatment and low oxygen
behaves similarly to the Zr-2.5 wt y0 Nb alloy with respect to molecular oxygen, then it appears to show a small increase in rate due to neutron irradiation (unlike the binary alloy) and a significant effect of pretreatment. Results obtained on this alloy by Metallgesellschaft AG are of little value due to the inadequacies of their experimental technique 57) which led to high oxidation rates both in and out of the reactor. The only other alloys to have received serious study in a high neutron flux are two experimental alloys (Valloy and Valloy-S) developed at Vallecitos. These alloys are respectively Zr-1.2 wt % Cr-0.08 wt y0 Fe and Zr-1.2 wt y0 Cu-0.28 wt y. Fe. In the G-7 loop in ETR (fig. 24) Valloy behaves comparably with Zircaloy-2 when originally in the etched condition, but gives very high weight gains when preautoclaved at 400 “C, even in loop tests
on the oxidation
of Zr-Nb
a;lloys at 280 “C in high
W&tm [ref. &)I.
containing little radiolytic oxygen. Valloy-S has generally shown the highest weight gains of any etched specimens, but the weight gains under irradiation have been little different from unirradiated specimens. Weight gains of prefilmed specimens have been lower than those of etched specimens with again little effect of irradiation. This alloy, therefore, seems to be little affected by either radiation or oxygen in the water; its general corrosion resistance in 280 “C water is poor, however. 4.
Hydrogen uptake
Most investigators have reported hydrogen contents for their specimens, with but little attempt at oo~elation. The results observed appear to fall into two groups. Firstly, there are the tests which have shown hydrogen contents close to those calculated (considering time and temperature only) despite the obser-
OPEN SQUARES - OUT- REACTOR CLOSED SQUARES - IN -REACTOR
+---Zr-Cr-Fe * VALLOY
___I
-Zr-Cu-Fe-----w “VALLOY -S ”
’
Fig. 24. ETR results for the oxidation of the “Vafloys” ia pHlO NR4UI-I ( < 0.05 ppm 0~) rat280 “C [ref. 4511.
DRESDEN FUEL
DEVELOPMENT
CYCLE
DRESDEN
REACTOR
SHlPPlNGPURT SAXTON
PROGRAMME
IN VBWR
REACTOR
REACTOR
500 EXb3SURE
Fig, 25,
TIME
tdaysl
absorbed by Zircaloy-2 fuel cladding in pressurkd w&+x and neutral boiling water at w 300 “C in VBWR, Dresden, Shippingport and Saxton maotore [ref. sl*)f.
Hydrogen
EFFECTS
OF
IRRADIATION
ON
THE
OXIDATION
vation of much thicker oxide films than were expected. This is the general experience with BWR fuel sQ-31)(fig. 25) and with specimens from the CRNL boiling water loops (fig. 26, points in the absence of ammonia additions) 40),
OB
ZIRCONIUNT
ALLOYS
25
and seems to be typical of situations where free oxygen is readily measured in the environment. Secondly, experiments at Hanford 36-s** 43-49, AERE 25-z*)and CISE 4%50) have generally shown percentage hydrogen uptakes (figs.
. ‘f 1.6-
IO 20 30 40 50 60 70 60 90 loo 110 120 I30 14oI60 160170 180190 EXPOSURE
TIME
- DAYS w
2.4 “E P g $.
I.8 -
z
1.6 -
c
l.4-
*
0
2.2 P.O-
0.8
-
0.6
-
0
0.40.2
0
0
-
PO
l1ll1ll*illlIlll
n
l
I
I
l
(b) Fig.
208,
b.
Hydrogen absorbed by Zircaloy-2 and Zr-2.5 wt oh Nb alloy specimensat w 280 ‘C in in-ma&or loops at CRNL [ref. 493.
26
B. COX
0
340%
100%
/
0
/
Y SO0
t-
0
,‘.
\ +/IN ABSENCEOF RADIATION
QNTERPOLATED FRDM FIGURES GIVEN IN RER 125) I I I 30 40 so 0 A0 mgdm’2 0
IO
/’
-
l
20
IN AhENCE RADIATION (c’ ref. 125)
OF
0 44
_C--
I
I
IO
20
I
30
I
I
40
50
I
I
I
I
60
70
80.
90
60 mgdm-2
Fig. 27. Hydrogen absorbed by Zircaloy-2 from 1 atm steam at 340 and 400 “C during AERE experiments [ref. ?I*
27, 28, 29) similar to those observed in the laboratory (i.e., 20-60 y0 of the available hydrogen), so that the total hydrogen in the specimens was increased in the same ratio as
Irradiated In e-7
tocp
10 1
100
10
1000
Time - IDayak
Fig. 28. Hydrogen absorbed by Zircalosr_tZ; in pH10 LiOH at 280 “C in the ETR G-7 loop operating with w 1 ppm Oa [ref. 88)].
the weight gain was increased by irradiation. Examples are also known 6%72)where hydrogen (or deute~um) uptake of fuel and st~ct~al components has been far higher than could be accounted for by the uptake of 100 y0 of the hydrogen theoretically released by the oxidation reaction (figs. 30,31). Thus the inference is that under some eon~tions {section 8) molecular hydrogen can be absorbed directly from the dissolved hydrogen present in the water. In general, although the percentage uptake in such instances has been high, the total amount of hydrogen absorbed has not been large because the low temperature of operation and the pretreatment of the components led to very small amounts of oxidation under PWR conditions. The enhanced uptake of hydrogen under aviation reported by Blanchet et al 58) appears as likely to be the result of their
EFFECTS
OF
IRRADIATION
ON
THE
OXIDATION
OF
ZIRCONIUM
A0
ALLOYS
27
mg/dm’
Fig. 29. Hydrogen absorption during in-reactor experiments in a water-steam fog at 280 “C in Avogadro I reactor [ref. 49)]. a, 400 “C steam, lab. autoclave; 0, w 280 “C, loop, out-reactor; 0, - 280 “C, loop, in-reactor.
I
I
I 2
I 3 YEARS
Fig. 30.
IN
I 4
I 5
REACTOR
Hydrogen uptake by in-reactor pressure tube from CR-V loop at CRNL [ref. e*)]. Numbers show elevation (El.) in feet.
28
B.
COX
i.EGE!@ CROSS HATCHED SOLID
t 200
AREA
LINE
BROKEN LINE
-
X
180
ZIRCALOY-2 - EARLY MANUFACTURE02 CONCENTRATION IN NPD COOLANT,
IRRADIATED
WITH HIGH
ZIRCALOY -2 - RECENT MANUFACTURE - IRRAQIATED D2 CONCENTRATION IN NPD COOLANT,
s
f
-
ZIRCALOY-2 IRAAOIATED
WITH HIGH
AND Ni - FREE ZIRCALOY-2 - kECENT MANUFA!CTUREWITH LOW D2 CONCENTRATION IN NPO COOLANT.
ZIRCALOY -2 - EARLY MANUFACTURE - IRRAOIATED 0, CONCENTRATION IN NPD COOLANT.
:: - 160
WITH LOW
3 s z z y
140 120
s 3
100
tt E
80
3 a
60 40 20
y______ I 100
__-I
200
I 300
i
I
I
I
I
I
400
500
600
700
800
900
DAYS IN HOT COOLANT
Pig. 3 1.
Deuterium absorbed from @Xl 0 LiOD at w 290 “C by NPD fuel cladding of various types at different periods during the reactor operation [ref. 73)].
experimental technique as a direct effect of irradiation. Gaseous hydrogen from both oxidation and radiolysis would have accumulated in their microautoclaves. Fuel cladding in PWR’s, which generally shows no acceleration of oxidation rate, usually shows hydrogen contents no higher than expected (fig. 25), within the accuracy of determination 18). However, the accuracy of such determinations (and predictions) may be relatively low. Thus, most fuel cladding was preautoclaved ; this increased by an uncertain amount the initial amount of hydrogen in the cladding. The initial amount of hydrogen is usually known only from a few spot analyses on a batch of cladding as received. Thus there are already two sources of error in estimating the starting hydrogen content. Even allowing for errors in measuring the final hydrogen content of radioactive material, there is little laboratory evidence from which to predict the
amount of hydrogen likely to be absorbed when preautoclaved material is subsequently exposed to hydrogenated water at a much lower temperature. Most laboratory hydrogen uptake d&a are for temperat~es of 316 “C and above, and for an initially pickled surface. There have been indications in L few laboratory tests that specimens preautoclaved at a higher temperature may absorb molecular hydrogen from the water for a short time when transferred to low temperature water 64). Thus, all that can be safely said in most instances is that the total hydrogen at the end of life was tolerable. However, this does not help predictions of the behaviour of fuel exposed under more stringent conditions. The examination of fuel cladding from NPD has many advantages over those available to other investigators. This material has been exposed only to heavy water following its initial autoclaving in light water. Thus, any
EFFECTS deuterium
found
OF IRRADIATION in the sheathing
ON
THE
must have
been absorbed during exposure in the reactor. Although metallography has failed to show any increase in oxide film thickness in-reactor (until the recent observation of white oxide spots,
section
3.5), appreciable
quantities
of
OXIDATION
OB ZIRCONIUM
29
ALLOYS
ZrOs, led to the proposal that a transformation to oubic under ideation
was
of the enhanced oxidation. that a transformation from
74) A tetra-
from monoclinic the cause suggestion
gonal to monoclinic transition
in the
could be the cause of the oxidation
kinetics
of
un-
deuterium have been found 72). These seem to have been absorbed during a relatively short
irradiated material had been proposed earlier 7J),
period after insertion in the reactor,
incorrect 76). Further studies showed that a transformation from monoclinic to cubic could be induced by irradiation 77-**), although there was some argument about whether fission fragment bombardment was essential 81, sa), or whether it was the uranium (added to generate the fission fragments) which was chemically stabilising the cubic phase in the presence of any form of heavy particle irradiation (fast neutron or fission fragment) 8% 84).
with little
subsequent absorption (fig. 31). The slow absorption seems to represent a reasonable fraction (M 60 %) of that liberated by the oxidation predieted from the laboratory oxidation data. The initial uptake is quite large, representing several hundred percent of any hydrogen likely to have been released by the oxidation process. It may therefore consist of deuterium absorbed directly from the deuterium overpressure. It is notable that the height of this initial step (fig. 31) has decreased with the decrease in the Da overpressure in NPD 73). The amount absorbed in this initial step appears also to be dependent on the metallurgical history of the cladding; however, even in the worst cases the total hydrogen plus deuterium at the end of life does not exceed about 100 ppm by weight of hydrogen. 5.
Proposed mechanisms for enhancement oxydation by irradiation
of
A large fraction of the experimental work reported so far has been devoted to demonstrating the presence or absence of an effect of irradiation on the oxidation behaviour. Little work has been directed towards studying the mechanisms of such processes. However, this has not prevented the postulation of a multitude of different mechanisms in the past. Before going on to study the present situation and to attempt any explanation it seems to be worthwhile to examine the past explanations in some detail. 5.1. PHASE TRANSFORMATION
IN THE OXIDE
The obse~ation that oxide films removed from zirconium specimens during the homogeneous reactor programme were largely cubic
and had not at that time been shown to be
Irrespective of the final conclusion about the precise cause of the irradiation induced transformation, which appears to lie in favour of the second explanation, measurements of the rate of this transformation showed that it was too slow to affect the oxidation rate over the short periods of time involved in some of the uranyl sulphate studies 1% 79~85). Thus the presence of oxide transformed to the cubic form by irradiation was an effect occurring subsequent to whatever other processes were affecting the oxidation rate. 5.2.
ENHANCED
DISSOLUTION OF Zr02
This was an early explanation proposed by workers at ORNL in the lightof their experimental observation of oxide loss from the specimen 12). In loop experiments it was observed that little of the oxide formed remained on the specimen, but was deposited generally in the unirradiated sections of the loop. These deposits were fairly adherent and did not give the impression of being transported merely in suspension, hence the supposition that the dissolution of Zr02 was being enhanced by fission fragment irradiation, Accurate meas~ements of the dissolutjon rate of ZrO2, after i~adiation to a high fast neutron dose, showed a small increase in dis-
30
B. COX
solution rate at short times 86). However,
this
failure process accentuated
by velocities
of up
was shown to be due to a small increase in the fraction of fine oxide particles as a result
to 40 ft/sec (13 m/set) in the in-reactor loops. Several routes by which the irradiation damage
of irradiation, and was not thought to represent
in the metal might
any increase in the intrinsic
have been proposed.
Thus,
although
demonstrate
this
dissolution
experiment
did
rate.
RADIATION
DAMAGE
IN THE
the oxidation
rate
not
that such an increase was absent
during irradiation, this explanation was not considered as a serious contender. Interest in the phenomenon was revived by the observations that the metallographic thickness of the oxide film on irradiated NPD fuel cladding was sometimes apparently less than that measured after the initial autoclave treatment, and that appreciable concentrations of zirconium in the water could apparently be observed in some loop experiments 40). A careful survey of NPD 87) and Shippingport 88) reactors, and the CRNL in-reactor loops failed to show any evidence for accumulations of zirconium in sufficient quantity to explain the irradiation corrosion phenomenon. Thus again interest lapsed, and loss of oxide by dissolution was discredited as a contributor to in-reactor oxidation tests. Recently we have again come full circle with the apparent observation of colloidal solutions of zirconcium in AERE in-reactor autoclave tests 52). The AERE observation could have resulted from spalling of the oxide on a fine scale, although the fact that the zirconium in the water was apparently not filterable makes this explanation improbable. At present, therefore, although it appears that there may be conditions under which formation of true, or colloidal, solutions of zirconium can take place under irradiation, dissolution of the oxide film is not regarded as a significant contributor to the observed weight changes in irradiated tests.
5.3.
affect
METAL
ORNL eventually discarded the idea of enhanced dissolution of the oxide in favour of a mechanism in which radiation damage in the metal was the predominant process 18~i4). The oxide was thought to be lost by a mechanical
5.3.1.
Enhanced diffusion of oxygen in zirconium
Experiments at ORNL, aimed at providing support for a “metal damage” hypothesis, have apparently produced evidence that the oxygen content of the surface layers of the metal may have a critical effect on the phenomenon 89). It has been proposed that enhanced diffusion of oxygen in the metal is the explanation of the acceleration of the oxidation rate under irradiation, although it was not clear how this hypothesis actually explained the practical observations. The suggestion that irradiation might cause the precipitation of oxide particles, which could lead to accelerated oxidation, is implausible. Irradiation does not normally achieve chemical reactions which were not thermodynamically possible in its absence, and oxygen in solution in zirconium is more stable thermodynamically than ZrOa. Metallography of irradiated specimens has not provided any evidence which would support such a supposition, and the experiments on which it is based can be criticised for lack of adequate controls 89). The complete absence of an oxide film from zirconium during a prolonged irradiation is exceedingly unlikely, irrespective of the method of preparation and irradiation. It seems likely that the effects being observed on nominally “oxide free” specimens were due to contamination of the specimens (e.g. with carbon or metal atoms sputtered during irradiation), whose surfaces were not reprepared after irradiation. This one omission, to reprepare the surface of irradiated and unirradiated controls side by side immediately prior to oxidation, has largely vitiated the results of this experiment. Irradiation of anodic oxide films with 65 keV neon
ions,
and
subsequent
oxidation
90) ap-
EFFECTS
OF
IRRADIATION
ON
THE
parently supported a view that damage to the metal was the important factor. The absence of i~adiated oxide was not demonstrated in any of these experiments, and hence the results cannot be uniquely attributed to metal damage. Insufficient control experiments (e.g. varying bombardment energy) were performed to demonstrate whether these results were or were not applicable to the case of neutron and fission fragment irradiation.
OXIDATION
ZIRCONIUM
ALLOYS
31
in the metal, the recent evidence that water chemistry and preoxidation treatments play a major part in the effect makes it much less likely that radiation damage in the metal can be an important parameter. In fact if anything the recent evidence suggests that a region at or near the oxide/environment interface, and not the oxide as a whole, is critical in determining the behaviour.
5.4.
Enhanced in-reactor creep of zirconium alloys has been observed 91). Since extension of the metal core (either by applied stress during oxidation, or by stresses generated by the oxidation of thin specimens) might lead to cracking of the oxide film and hence to accelerated oxidation this is a possible contributor to the observed phenomena. No oxidation rates on PWR fuel cladding, or pressure tubes, have been observed, however, which are significantly higher than those observed on unstressed corrosion coupons (e.g. fig. lo), so that there does not seem to be any experimental support for the importance of this effect. It seems unlikely that a small rate of growth of the metal should affect the growth of an oxide which, under normal circumstances, accommodates a 50 */* volume increasing during oxidation. The oxidation of thin specimens, which creep during oxidation under the influence of the stresses generated by the oxidation, is not significantly different from that of thicker specimens 92), and oxidation of specimens stressed beyond the yield point has shown no change prior to transition and only small increases in oxidation after transition $3~94). Thus it seems unlikely that this effect could be participating in the enhanced oxidation of Zircaloy-2 in-reactor for normally stressed specimens, and even for highly stressed material it could only account for a fraction of the observed increase in oxidation. In general, although during the early work there was no reason why damage to the oxide should be a preferable explanation to damage
OB
FISSILE
MATERIAL
The Bettis suggestion that instances of enhanced oxidation during radiation in water were due to contamination with fissile materiala) is hardly an explanation. It merely seeks to relate the phenomenon in water to another unexplained (but well established) phenomenon, enhancement of oxidation in fissioning many1 sulphate solutions. A broader view, which accepts evidence not obtained in high temperature water, shows that in many instances of enhanced oxidation under i~a~ation [AERE 2%2%25*26~ 52) and CISE 49-51) work], there could not conceivably have been contamination by fissile material, since fuel materials had never been in the equipment at any time. Thus as a general correlation of the phenomena this must fail. However, there may be far more similarity between the mechanisms in both water and fissile solutions than has been assumed in the past. Thus in both environments the phenomenon is affected by the presence or absence of oxygen 7*15~37). ~ilst undoub~~y fission fragment irradiation is a far more efficient means of damaging the oxide than fast neutrons, it may be that the vastly enhanced radiolytic decomposition of water in the presence of fission fragments is at least as important. Thus a relatively small amount of contamination by fissile material which might be insufficient to cause any major change in the damage rate in the oxide film, oould easily affect the local chemistry of the water by increasing the production rate of radiolytic species.
32
B. COX
5.5.
INDUCED UNDER
If
the
DIIwusfoN
IN THE
OXIDE
IRRADIATION
irradiation
of
the
oxide
produces
vacancies and interstitials, which are able to transport reactive species through the film, and the dose rate is sufficiently
high that the con-
centration of defects is increased, and if transport of such defects is the rate controlling process, then an acceleration
in the oxidation
process should be observed without any change in the kinetics law which is obeyed 95). However, it was realised from an early stage in these studies that if reasonable values for the rate of production and annealing of these defects were used in a calculation. then the effect should be below observable limits at the fast neutron damage rates available, and inadequate to explain the very high oxidation rates in fissile solutions. To overcome this problem it was suggested that permanent regions of damage must be produced which were large enough to act as easy paths for diffusion through the oxide IQ). In the homogeneous reactor case it was felt that fission fragment tracks in the oxide might fulfil this purpose, although there is no particular reason why any other radiation induced change in the oxide leading to enhanced diffusion (e.g. nucleation of smaller oxide crystallites under irradiation) could not explain the phenomenon
conductivity in thin ZrOQ films could lead to increased oxidation of Zircaloy-2 in a y-flux Q5), and the apparent observation of increased conductivity of ZrOQ in some in-reactor experiments IQQ),led Wanklyn et al. to an extensive study of the electrical properties of zirconia films under irradiation 96, 97, iQi-103). The results were based on the a.c. conductivity
measured
mainly on anodic oxide films at ambient temperat~e, and to that extent their applicability to oxidation processes (de conductivity) at high temperature may be in doubt. Some allowance for these differences can be made, on the assumption (doubtful) that we know the electron transport process in thin oxide films. The irradiation effects observed were so small, however, that it seems unlikely that the large difference between the experimental conditions and normal oxidation would completely alter the conclusions. The overall conclusion of these experiments was that no observable change in the conductivity of zirconia films is produced by a gamma flux and only small increases in conductivity resulted from fast neutron doses > lOQQnf/cmQ~sec. Neither fluxes nor doses, which have been observed to affect the oxidation rate, showed any effect on conductivity outside measurement errors. Theoretical calculations of the effect of y-irradiation tended to support the ~sumption that no effect would be expected ; however, some of the parameters involved in the theoretical treatment were not
equally well. However, the assumption that lattice diffusion of vacancies is the controlling process during oxidation in the absence of radiation has led investigators to reject an explanation of the irradiation enhanced oxidation on the basis of irradiation effects on the ionic transport process QQ,Q’), on the ground that production of vacancies and interstitials was inadequate, and that irradiation produced no increase in the mobility of these defects 97). This criticism is inapplicable if the ionic transport process is not controlled by a lattice effusion process in the oxide (as appears to be the case) Q*pQQ).
sufficiently well known for this to be said with certainty. Thus it would appear that any effect of gamma radiation on the oxidation process cannot be due to a direct effect on the conductivity of the oxide, and other explanations are needed for the Hanford observation of enhanced conductivity. A change in the environmental chemistry, under gamma irradiation, which could affect the growth of thin oxide films, is not ruled out.
0.0.
5.7.
The
ENHANCED
suggestion
ELECTRICAL
that
CONDUCTIVITY
enhanced
electrical
EMBRITTLEMENT
Having
eliminated
OF THE
electrical
OXIDE
effects
FILM
cxperi-
EFFECTS
OF IRRADIATION
ON THE OXIDATION
ALLOYS
33
of the energy absorbed
by the
OF ZIRCONIUM
and changes in ionic transport mentally, (implausible; see above), Wanklyn et al. have suggested that embrittlement of the oxide leading to earlier cracking of the film and acceleration of the subsequent oxidation rate could be an explanation of the irradiation
major
enhanced
194). Results
peroxide
(the
and thermal
produced
by the y-irradiation
oxidation
of Zircaloy-2
of bending
oxidation
phenomenon
pre-irradiated
anodic
films at room temperature round a mandrel have given results supporting this hypothesis. Such an hypothesis can operate only in the post-transition region, or could explain the reduction in the weight gain at transition. There appear to be ample experimental observations (e.g. at AERE, CISE, Hanford and CRNL) of enhanced oxidation by a parabolic or cubic rate law prior to transition (figs. 7, 15, 16, 18) to demonstrate that this cannot be the whole explanation. Thus embrittlement of the oxide may be a contributor to the later stages of the irradiation enhanced oxidation, but cannot be the primary cause of the phenomenon. The conditions of these experiments were appreciably different from normal oxidation conditions. It is known that cracking due to thermal stresses can occur in oxide films on cooling to room temperature 195), and bending tests performed under these conditions do not necessarily lead to valid conclusions about behaviour at the oxidation temperature. The cracks observed in the optical microscope 194) are probably not the important ones causing the normal rate transition 195).The important cracks or pores are much smaller (w 50 d dia.) than these and are difficult to observe even by electron microscope techniques. 5.8.
DIRECT CHEMICALEFFECTS
The idea that aggressive radiolytic species in the solution could be attacking the oxide film on Zircaloy-2 directly, without any direct contribution from radiation damage to the process, gained some credence when the effects of water chemistry first became well established. However, such an effect would be equally prominent in a gamma flux alone, since the
fraction
water arises from this source. Gamma flux effects are at most of only minor importance in the thin film region, and are completely absent when the oxide film becomes thicker. An examination of the effect of hydrogen most
likely
long-lived
species
of water) on the
showed no effect 99).
The insensitivity of the oxidation rate to environment chemistry in the laboratory 199) and the failure to obtain evidence that would substantiate any increase in oxidation rate downstream of the irradiated test section 49-51) argue further against any direct chemical effect. Only in the very early stages of oxidation in water and steam is there evidence that the chemical nature of the environment is important 197). Th ese effects influence the weight gain during pre-autoclaving treatments but disappear subsequently. It is in this region of weight gain (O-15 mg/dm9) that effects attributed to the gamma flux have been observed, and only in high pressure steam. Therefore any effect of gamma flux may be purely one of environment chemistry which would be most important under the observed conditions. The density of radiolytic species would be higher in water or high pressure steam than in low pressure atmospheres, and this might explain the failure to detect effects in tests at 1 atm. For Zr-Nb alloys the picture is different; these alloys are affected by the presence of oxygen-containing species in the water during laboratory tests 199), and hence the production of such species, even if short lived, in reactor can result in a direct chemical effect 59) both in flux and downstream of the irradiated region. The magnitude of these effects is very dependent on alloy composition and heat treatment. 5.9.
ENERCJY TRANSFER PROCESSES
Radiation chemists have recently become aware of the major part which energy transfer processes can play in radiolysis 199). Thus, in the presence of a suitable surface, energy absorbed in the solid phase can be transferred
34
B.
COX
to adsorbed molecules to cause their radiolysis.
to be no thicker
In this way radiolysis rates may be obtained which are far higher than those which would
based on the calculated effect of heat flux across the oxide film has been applied in
result
from
direct
interaction
between
the
radiation and the liquid or gas whose radiolysis is being studied. In effect the surface is acting as a catalyst for the radiolysis in a manner analogous
to that which
a chemical
catalyst
plays in a chemical reaction. Such a phenomenon might explain the necessity for both irradiation and suitable chemistry to be present simultaneously, since the active species which subsequently cause the enhanced oxidation are produced in large quantities on the surface sites of oxide where the probability of reaction is greatest. While no actual measurements have been made which would lend support to this hypothesis, the observation of accelerated oxidation under conditions where no oxygen or oxidising radiolytic species could be detected at the outlet to the loop test section, suggests that this type of process may be playing a major part in the overall phenomenon, and that radiolysis of adsorbed species rather than isotropic radiolysis within the environment may be the important chemical factor. which eliminated direct The arguments chemical reaction with radiolytic species as a contributor to other than the growth of thin
obtaining
than
the expected
expected 35) a factor
film thickness.
Experi-
mental verification of a heat flux effect in the absence of irradiation, which depends on certain assumptions
about
the oxidation
mechanism
which may not apply, is absent ilO--112).Measurements with electrically heated specimens 111) could be affected by the presence of even small ac potentials across the oxide. A rise in the surface temperature of fuel due to the deposition of crud on surfaces under heat flux might have contributed to the increased oxide thicknesses observed on BWR fuel cladding. However, since oxidation rates of fuel cladding are comparable with those on the isothermal G-7 loop specimens (fig. 6) this may be a minor contribution, and cannot be a general explanation of irradiation effects even if its contribution to the oxidation of BWR cladding were significant.
5.11.
QUALITY OF MATERIAL
Examination of fuel cladding from the VBWR and Dresden reactors showed that, in addition to the accelerated uniform oxidation, the Zircaloy-2 formed a large number of pustules of thicker oxide 299so). In measuring
films (lack of any effect under gamma radiation) would seem to rule this explanation out as the sole process responsible; some form of direct
the film thickness on fuel elements these areas of irregular oxide were ignored, and only uniform oxide was measured 39). Examination
damage process in the necessary as well.
of recent fuel from Dresden has shown no pustules and lower oxide thicknesses at equivalent times 31) than for the earlier cladding. It has been suggested that this represents the result of improvements in the quality of the Zircaloy-2. Whilst experiments elsewhere 46) have shown some variation from batch to batch of Zircaloy-2 the range of behaviour has been less than a factor of two. Thus the large difference observed with Dresden fuel may arise from some other change made in the operational details (e.g. change in preautoclaving). Insufficient details are available to judge this possibility.
5.10.
oxide
seems
to
be
EFFECT OF HEAT FLUX
Many of the examples of thicker than expected oxide films have been observed on fuel cladding from Boiling Water Reactors. The contribution to this thick film resulting from the rise in temperature at the metal/oxide interface due to the high heat fluxes prevailing across the oxide film has been discussed 33-35). In examinations of fuel cladding from Pressurised Water Reactors where the oxide film is stated
EFFECTS
6.
Oxidation
OF
IRRADIATION
ON
TEE
mechanisms
OXIDATION
OF
ZIRCONIUM
ALLOYS
35
alloy, and in the influence of grosser regions of disorder introduced
by the presence
of second
Before attempting an explanation of the phenomena observed under irradiation, it is essential to understand the oxidation process in the absence of irradiation. The present
phase particles in many of the common alloys. However, it would be unwise to assume that it can be inferred from the kinetic curves that
situation on the oxidation mechanism and effects of irradiation on transport processes as seen by the author are summarised here.
ionic transport is the rate controlling process ; electrons must also be transported through the oxide film in the reverse direction. The transport of electrons through insulators such as zirconia
6.1.
The commonly accepted mechanism for the pre-transition oxidation of all zirconium alloys has been that it was controlled by oxygen diffusion inwards via anion vacancies 11s). It has been suggested that the approximately cubic kinetics represented only a minor deviation from the norm and could still be explained on the basis of the Wagner-Hauffe theory 114). The fact that in the early stages the rate laws were often far from good fits to any readily analysed process has been dismissed also as deviation from the norm 115). Attempts to fit the pretransition oxidation kinetics to a sequence of parabolic and cubic periods of oxidation do not assist in explaining what is happening in the absence of studies of the individual transport processes llQ9l17). However, the oxide film formed on zirconium and its alloys consists of discrete crystallites (50-250 A dia.) which grow slowly in size as
has not been extensively studied. However, investigations of electron transport through glasses, and the effect of devitrification iQQ), show that whereas crystallite boundaries and amorphous material may represent easy paths for ionic transport, they may represent barriers to electron diffusion. The easy paths for electron transport in this case would be through the crystallites. Thus growth of oxide crystallitea in the oxide film, with increasing thickness, which would represent a decrease in the area of material available for easy transport of ions, could imply an increase in the ease of electron transport. Experiments show that electron transport through oxide films on zirconium alloys may be mainly at flaws, or the sites of intermetallic particles lQ1), and is probably not diffusion controlled, even for thick oxide films QQ). Measurements of the potential across the oxide film formed on zirconium and its alloys
the film thickens, and which may be embedded in a matrix of amorphous material (comprising a significant fraction of the bulk) in the thin
in a fused salt bath provide a clue to which of the two processes is the more difficult QQ). Thus, crystal bar zirconium oxidising at a
film region 1139119). Attempts to measure the lattice diffusion coefficient of oxygen in zirconia films have suggested that at the temperatures normally used in the practical applications of zirconium alloys the actual diffusion coefficient may be as much as five orders of magnitude higher than the lattice diffusion coefficient 98). Thus the conclusion that oxygen ion transport at = 300 “C is almost exclusively via crystallite boundaries or other easy paths seems inescapable. The primary differences between the ionic transport process in the oxides of zirconium and its alloys may be in the way the crystallite morphology develops from alloy to
temperature above 250 “C rapidly develops a potential of - 1.1 to - 1.4 V, which falls slowly throughout subsequent oxidation. Although the anodic and cathodic polarisation curves are not well established, the high negative potential observed indicates that the transport of electrons through the oxide film is the more difficult process, and that it may be occurring by a Schottky emission process over barriers associated with flaws in the oxide. The potential on an alloy such as Zircaloy-2 oxidising at 300 “C passes through a maximum of about - 1 V and then falls rapidly, as the oxide film thickens, to a fairly stable value of
IN
THE
ABSENCE
OF
IRRADIATION
36
B. COX
about -0.25
V. This suggests that the electron
transport process becomes relatively easier as the oxide thickens. It is impossible to say, at present, which is the most difficult process under conditions where the potential is M 0.25 V (metal negative) ; but since the ionic and electronic currents must be equal during oxidation the two processes may be equally important in controlling the oxidation rate. Transient changes in the conditions of oxidation produce changes in the balance between the ionic and electronic transport processes. Thus initial oxidation at R+ 400 “C, and transfer to m 300 “C, results in small positive potentials on the metal which persist for a long period (weeks) without annealing out. The presence of a small positive potential shows that the balance between electronic and ionic transport is shifted even further in the direction of easy electron flow. Since the oxidation rate (ionic current) is little affected by the pretreatment at m 400 “C (the memory effects were very small under the conditions used) this situation must have arisen because of the increasing ease of the electronic process, rather than a decrease in the ease of ionic flow (which would have resulted in a lower oxidation rate) 99). Similar experiments on other zirconium alloys have shown wide variations in behaviour, which may help to explain the different behaviour of these alloys under irradiation. Investigation of the formation of pores and cracks at the transition point has shown that the important defects in the oxide are a network of fine pores < 100 A in diameter 105). These have not been identified by electron microscopy but may be associated with the growth of the oxide crystallites to a certain critical size. The occurrence of larger cracks in the oxide at the oxidation temperature and the importance of the stresses produced by oxidation in causing these (and the small pores) has not been satisfactorily studied. 6.2.
EFFECT
OF IRRADIATION
ELECTRONIC
Wanklyn’s
ON IONIC AND
TRANSPORT
work
suggests
that
electronic
transport
through
oxide
films
on
zirconium
alloys is likely to be little affected by irradiation with fast neutrons, gamma rays or other species (section 5.6). Since for Zircaloy-2 the more difficult process, other than in the thin film oxidation region, may be ionic transport, it must be an effect on this process which leads to the observed acceleration of the pre-transition oxidation rate 25945,48-50). We have seen (section 5.5) that such an acceleration cannot be accounted for on the basis of the production of additional vacancies, nor would our interpretation of the out-reactor oxidation mechanism suggest that production of additional vacancies would be important even if it were large enough to affect the normal vacancy concentration. Permanent changes in the oxide film under irradiation are needed to change the oxidation rate. These could be in the nature of additional easy paths, represented, for example, by fission fragment tracks is), or the tracks of primary knock-on particles if these are of sutllcient size (i.e. they should span a significant fraction of the film thickness). While this gives us an argument that would support the importance of fast rather than thermal neutrons (in the absence of fission fragments), since only neutrons with energies greater than about 1 MeV would produce primary knock-ons with a range in ZrOz comparable to the film thickness, it seems unlikely that the density of such tracks would approach that of the crystallite boundaries when crystallite size is only 50-250 A. Perhaps a more likely manner in which irradiation could be affecting the ionic transport process is by increasing the nucleation frequency or decreasing the growth rate of ZrOa crystallites, thus leading to a film with a smaller average crystallite size. The only known investigation by transmission electron microscopy 6% 65) of oxide films formed in-reactor did not show any significant difference from films formed in the laboratory. However, the subtle changes involved in this hypothesis might well have escaped the rather cursory examination given to these films. A report that
EFFECTS
the crystallite
OF
IRRADIATION
ON
THE
size in the oxide film formed in
OXIDATION
appear
OF
ZIRCONIUM
that the electron
37
ALLOYS
transport
process
is
a fused nitrate bath at 400 “C was much smaller than for a similar film formed in air iss), would support the contention that such effects may be very important, since it is above about
the one most affected by water chemistry; again this seems a reasonable conclusion since laboratory data suggest that the most important region in controlling electron flow is a thin
400 “C that the oxidation
barrier, which may be at the oxide/environment
rate in a fused nitrate larger than that
interface QQ) ,
in air. For interpretive purposes, however, it may be the balance between the ionic and electronic transport processes which is important. Thus, if one of the two processes is being altered by irradiation more than the other, the oxidation rate will only be changed if the process most affected by irradiation (in this instance the ionic transport) is the more difficult process. The magnitude of the acceleration under irradiation will be limited to the point at which
Adsorption
medium becomes
significantly
control of the oxidation process swings over to the least affected process (electron transport). Thus the magnitude of the expected increase in oxidation rate under irradiation will be dictated by the relative difficulty of the ionic and electronic processes prior to irra~ation. So far, however, we have not introduced any factor that could be associated with water chemistry. 6.3.
EFFECT
OF ENVIRONMENT
OXIDATION
PROCESSES
CHEMISTRY
ON
easily
affect
of species at this interface electron
and hole injection
could into
this layer of the oxide. In this respect the work of Leach 123) which showed large changes in oxide conductivity during cathodic polarisation may be impo~ant. Although the inclusion of protons in the oxide, as required by Leach’s proposed mechanism, has never been demonstrated, these may be acting only at the sites of easy electron flow. 7.
Hypothetical explanation effects on oxidation
of radiation
We may have available, therefore, a basis for interpreting the irradiation effects. The supporting experimental data from laboratory tests or direct meas~ements under i~adiation is, in most instances, inadequate, however, and we are therefore drawing inferences on tenuous grounds. Nevertheless, in the interests of rationalising the in-reactor observations, and demonstrating the importance of certain types
of experimental data, the attempt seems to We have seen (section 5.8) that in the absence be worthwhile, even if it merely serves to of irradiation, environment chemistry has little stimulate experimental work to disprove it. To or no effect on the pre-transition oxidation rate this end we will treat the observations on the of Zircaloy-2, and that in this region potential alloys individually. measurements suggest that ionic transport is the more diflieult process (or at most that ionic 7.1. ZIRCALOY-2 AND ZIRCALOY-4 and electronic transport are about equally difficult). We can conclude, therefore, that the A summary of the important experimental ionic transport process is little affected by observations first seems to be in order: environment chemistry, as would appear to be (4 Acceleration of the oxidation rate of the reasonable if it is controlled by transport along Zircaloys is always observed in BWR crystallite boundaries in the oxide. conditions, or in pressurised water conEnvironment chemistry does have an effect taining oxygen. on oxidation rates in the thin film region (b) Little or no acceleration is observed in (O-7 mg/dmz) of Zircaloy-2 oxidation I*?), where PWR conditions for the Zircaloys in the potential measurements show electron transport pickled state, or when pretreated at 300 “C. to be the most difficult process. Thus, it would (4 When pretreated at 400 “C acceleration of
38
B.
the
oxidation
rate
of
the
Zircaloys
is
observed even in the absence of measurable oxygen contents, although the magnitude
(4
cox Starting with the premise that the conclusions drawn from potential measurements in fused salts in the laboratory may be applied to
of the increase is less than in the presence
aqueous oxidation,
of oxygen. Both pre-
specimen, over the region of film thickness 0.5-Z ,um, the ionic and electronic transport
and
post-transition
rates are accelerated,
oxidation
with a reduction
then, for a normal pickled
in
processes are about equally difficult (fig. 32a) 90).
the time to transition and the weight gain at transition.
Then if only the ionic transport process is speeded up under irradiation there will be little
The effect saturates at high neutron fluxes. All effects disappear at or above 400 “C. Appreciable differences in the magnitude of the effects are apparent from batch to batch of Zircaloy-2 and Zircaloy-4.
or no observable effect on the oxidation rate, but the control of the process will now be due to electron transport and the potential across the oxide should become more negative (fig. 32b). If there is no measurable concentration
0) HYPOTHETICAL
POLARIZATION AND
CURVES
FOR PICKLED
( SUB P)
AIJTOCLAVED
(SUE A 1 SPECIMENS
b) EFFECT
OF IRRADIATION
PWR CONDITIONS
PREIN LAB
CURVE
UNDER
; ONLY ANODIC
AFFECTED
PICKLED SPECIMENS
Eox /(
v
v
METAL POSITIVE
METAL POSITIVE
1’
PREAUTOCLAVED SPECIMENS
WITH OXYGEN I c) EFFECT
OF IRRADIATION
CONDITIONS
UNDER OXIDIZING
; BOTH CURVES
AFFECTED
d) SPECIMENS IN
(~J,WITH
Fig. 32.
Hypothetical
effecte of irradiation
PREAUTOCLAVED
RELATIVELY SMALL
AT 4033%
LOW OXYGEN WATER NH,
ADDITIONS)
on electronic and ionic transport
processes.
EFFECTS
OF
IRRADIATION
ON
THE
of oxygen or oxidising radicals in the reactor (i.e. PWR con~tions) which might affect the electron transport process then the net result would be little effect of irradiation on the oxidation rate. However, if in addition to neutron irradiation sufficient to swing the control over to the electronic process there is oxygen in the coolant (e.g. BWR’s or G-7 loop, 1 ppm Os), or a high rate of production of oxidising species (BWR conditions, inadequately doped with ammonia) then the oxidation rate will be increased by whatever factor the electron current is increased by the presence of an “oxi~sing” environment (fig. 320). This factor will always be limited by the possibility of a return to ionic control if the increase in the electronic conductivity (due to oxygen) is greater than the increase in the ionic conductivity (due to fast neutron irradiation). Thus the highest irradiation enhancements will be observed under conditions of both high oxygen content and high fast neutron flux (G-7 loop, or BWR with untreated water). However, it would not be expected that the oxidation rate under irradiation should increase indefinitely with increasing oxygen content or neutron flux ; both effects would be expected to ssturate, although it is not possible from present data to estimate any likely point where this saturation would occur. It is also clear that neither a high fast neutron flux in the presence of a minor amount of oxygen, nor a large amount of oxygen in a relatively low fast neutron flux, would lead to very large accelerations in the oxidation rate. In the absence of any numerical information on the sensitivity of the anodic and cathodic processes to fast neutrons and oxygen respectively it is not possible to predict quantitatively what the relative effects of the two contributors would be. Fission fragment bombardment will affect both the anodic and cathodic processes simultaneously via its damaging effect on the oxide and the very high rates of radiolytic decomposition of water it produces. Thus the presence
OXIDATION
OF
ZIRCONIUM
ALLOYS
39
of fissile material in the water should be the most effective way of producing conditions suitable for accelerated oxidation. Fission fragment i~adiation would also be expected to lead to the largest enhancement factors (for similar reasons), and in practice the largest observed acceleration of zirconium alloy oxidation rates was in homogeneous reactor solutions. Although the above arguments apply predominantly to pretransition oxidation processes, no major changes in the potential across the oxide film have been observed at the transition point Sa)* Thus, while embrittlement of the oxide by irradiation may result in a reduction in time to transition and weight gain at transition, it would appear that the same arguments as above would apply to the post-transition oxidation rate as well as to the pre-transition rate. The effect of specimen pretreatment can be interpreted from the observed effect of pretreatment on the potential developed across the oxide film during oxidation in fused salts. l?or pretreated specimens the electronic processes have already been made relatively easier by the pretreatment. Thus, even in the absence of oxidising species an acceleration of the oxidation rate would be expected (fig. 32d) ; however, it would probably not be very large without the presence of oxidising species to increase further the electronic conductivity. At high oxygen levels the large effect on the electronic conductivity would reduce the effect of pretreatment to a small pert~bation of the oxidation rate. ~etreatments at lower tem~rat~es than PO0“C should result in a lesser effect, but there is insufficient laboratory evidence on which to draw a positive conclusion. The arguments in the above paragraph will apply only to relatively thin films. If either the preformed oxide is very thick, or the film is allowed to become very thick in-reactor, then the oxidation rate becomes virtually independent of water chemistry, and is equivalent to that in high oxygen water (fig. 20). This probably arises because formation of radiolytic
40
B. COX
oxygen
within the thick porous oxide is inde-
pendent of the external water chemistry. To explain the disappearance of the irradi-
these batches under the worst conditions of flux and oxygen, since here the effect of oxygen on the electronic conductivity could swamp
ation effects at or above 400 “C requires the acceptance of two additional pieces of infor-
the other differences ; this does not appear to be borne out in practice.
mation. Firstly, that, even for the Zircaloys with their intermetallic particles available to
We have thus been able to explain, in general, the features of the irradiation enhanced oxidation of the Zircaloys. If a more detailed
give
paths
of easy
electron
flow,
the
more
difficult process above 400 “C is always electron transport. Potentials measured above 400 “C are usually more negative than at lower temperatures QQ). Secondly, that above 400 “C the electron transport process becomes insensitive to environmental effects. This latter is borne out by the lack of environmental effect for oxides thicker than m I. ,um in laboratory tests in 400 “C steam. This may result from a more rapid oxidation of the intermetallic particles at temperatures above 400 “C than below this temperature, which could lead to the electron current being controlled by diffusion rather than by emission over a barrier. Intermetallic particles in Zr-Al alloys have been observed to remain virtually unoxidised at 300 “C, whereas they oxidise progressively more rapidly (relative to the thickening of the whole oxide) with increasing temperature 124). If these two premises are accepted then no acceleration due to irradiation would be expected at the higher temperatures under any conditions of dose rate or environmental chemistry. The differences from batch to batch Zircaloy-2 and Zircaloy-4 may be related
of to
an effect of the distribution of the intermetallic particles on the electronic conductivity. This distribution appears to have little effect on the ionic conductivity since the pre-transition oxidation rate in the absence of irradiation varies little from batch to batch 125).Thus a variation in the balance between the anodic and cathodic processes in the absence of irradiation could be present from batch to batch. This would be reflected in a difference in their behaviour, especially in “oxygen-free” conditions for pretreated specimens, where these differences would be most signi~oant. It might be expected that little difference would be apparent between
quantitative explanation is to be made, then further experimental measurements must be made of the effect of the reactor environment on the controlling processes. Especially needed are measurements in-reactor of the potential developed across the growing oxide film which should demonstrate the swing from one controlling process to another as the conditions are changed. 7.2.
Zr-2.5 wt % Nb ALLOY
A similar summary of the important inreactor observations for these alloys shows the following : (a) Oxidation rate enhanced both in-flux and downstream in the presence of “oxidising” coolants. (b) Little difference between in-flux and downstream weight gains. Evidence generally shows slightly lower in-flux weight gains. (c) Little apparent effect of pretreatment. (d) Little evidence for flux dependence of oxidation rate, but oxidation rate apparently saturates at high oxygen levels. (e) Large effect of specimen heat-treatment. (f) Effect of oxygen does not disappear at 400 “C. Potential measurements on Zr-2.5 wt y0 Nb alloy specimens have generally shown negative values between 0.5 and 0.9 V. The highest potentials were for material quenched from the p or (a+@) phases, with a steady reduction in the potential as a result of annealing at 500 “C. However, it appears that, qualitatively, even after long annealing treatments the electron transport remains in control of the overall oxidation rate QQ). This observation, plus the fact that the oxidation of this alloy is sensitive to the presence
EFFECTS of oxygen
OF IRRADIATION
or oxidising
ment in the absence
ON THE OXIDATION
species in the environof radiation,
auf&es
to
those has
OF ZIRCONIUM
of the oxidation suggested
that
ALLOYS
process.
there
are
41
Recent
work
at least
two
explain most of the observed in-reactor behaviour. Thus no direct acceleration of the oxidation
processes normally operating in parallel lz*). 1. Direct access of molecular hydrogen via flaws in the oxide film. Possibly the same
rate by fast neutrons in the absence of oxygen would be expected. If oxidising species are ,produced in the reactor then both in-flux and
2. Diffusion of hydrogen (protons) through the bulk of the oxide film.
out-of-flux
oxidation
rates should
be affected
to about the same degree. The observation of slightly lower weight gains in-flux compared to those out-of-flux is difficult to explain on this basis, however. If control of the electron transport process lies near the outer surface of the oxide, then it is possible that a direct effect of radiation on this surface, or an energy transfer process leading to a specific adsorbed species, might limit injection into the conduction band. This remains pure speculation, however, as no experimental evidence is available in support of such an explanation. However, if the electron transfer process were affected in this way then one might expect the oxidation rate of Zr-Nb to be PH dependent in water. This pH dependence would need careful separation from the specific effect of oxygen and oxidising ions (e.g. nitrate) which is already well known. Other experiments which might lead to evidence on this point are a study of effect of y-radiation or U.V. illumination on the oxidation rate in air. Photoeffects remain virtually uninvestigated in zirconium or its alloys. 7.3.
OTHER ALLOYS
In general the background evidence available on the other alloys which have been tested is inadequate for any explanation of the phenomena. It seems reasonable to assume that similar experiments to those reported for Zircaloy-2 and Zr-2.5 wt y0 Nb would lead to a qualitative explanation of the behaviour of the Valloys and Zr-3 wt y0 Nb-1 wt y0 Sn. 8.
Radiation
effects on hydrogen
uptake
Investigations of the mechanism of hydrogen uptake have not been nearly so extensive as
sites at which the electronic
The
total
hydrogen
gaining
current flows.
access
to the
metal may be controlled by a barrier-type process at or near the oxide/metal interface (and possibly within the region of metal saturated with oxygen). The balance between the two transport processes through the film is probably set by the individual characteristics of the particular film and the environment outside it. Thus, the observation of the rapid uptake of small amounts of hydrogen, without a corresponding amount of oxidation, immediately after a specimen preautoclaved at 400 “C is transferred to < 300 “C water, may result from a breakdown of the initial oxide film leading to an acceleration of process (1) l@j). This would continue only until the defects had healed, and this healing might not reveal itself as accelerated oxidation over the same period. This could explain the observed deuterium uptake curves in NPD fuel for example 7% 73). In general it appears that the hydrogen uptake and oxidation processes are controlled by separate but parallel mechanisms. Thus, under irradiation it is possible for the two effects to be modified to different extents. It is known that the addition of relatively small quantities of oxygen to high temperature water or steam results in a significant reduction in the percentage hydrogen uptake 126). In the reactor environments containing much free oxygen (e.g. BWR’s), the reduction in percentage uptake resulting from the oxygen apparently almost balances the increase in oxidation rate. In other environments where the uptake behaviour is intermediate between BWR and PWR conditions (e.g. G-7 loop, AERE 1 atm steam loops) then there is apparently insu~cient oxygen to result in a reduction in the hydrogen uptake,
42
B. COX oxi-
of the oxide as well as just short-circuiting specimens. This type of experiment may not
It would appear that, if the environment cannot be maintained in a “reducing” condition
be carried out readily in the usual high resistivity water, where the oxide resistance is not large
although
the environment
dising to result in enhanced
equivalent
to normal
total hydrogen
PWR
is sufficiently oxidation,
operation,
and if
at the end of component
life
compared
with that of the water. The experi-
is the critical problem, then one might be better
ments would be easier in an electrolyte whose resistance was small compared with the oxide
operating with a large excess of oxygen, than with only a marginal amount. The oxidation
film resistance. Direct meas~ements
rate is not likely to be increased further under these conditions due to a saturation of the increase in oxidation rate with increasing oxygen content. Thus VBWR and Dresden fuel did not oxidise any faster than specimens in the G-7 loop (with oxygen) at equivalent fiuxes, but the total hydrogen uptake was much lower for BWR fuel than for specimens in the G-7 loop (with oxygen}. The relative efficiency of different enviro~ents for reducing hy~ogen uptake is difficult to ascertain from present evidence.
of the potential
across
the oxide film and the I-V characteristics of the oxide during oxidation in-reactor should prove to be interesting. Changes from more positive to more negative potentials on the metal should result from the application of a fast neutron flux, in the absence of “oxidising” conditions, and shifts in the polarisation curves such as those postulated in fig. 32 should be observed. The effect of increasing the neutron flux and additions of oxygen to the environment could also be investigated. A more detailed examination of oxide films stripped from specimens that have exhibited 9. Experimental testing of hypotheses enhanced oxidation rates would be interesting. Since detection of enhanced oxidation is not The value of any of the above suggestions readily achieved in the earliest stages of oxi(sections 6- 8) is only as great as their ability dation, this would require investigation of films to predict the behaviour in future experiments. much thicker (up to 1 pm) than those that The critical point in the explanation of can normally be viewed in an electron microenhanced oxidation is that only one (ionic scope. This should be possible with currently transport) of the two main processes is affected available 1 MeV microscopes. The in~resting by irradiation, and therefore control of the phenomena are likely to be differences in the oxidation rate can switch to the other (electron size of the small crystallites present, and other transport) process under irradiation. If the changes such as the fraction of the oxide volume oxide film were to be deliberately shortconsisting of resolvable crystallites. It would circuited during an irradiation experiment then be predicted that under i~adiation either the electron transport could never become the rate average crystallite size decreases and the volume controlling process and the oxidation rate of of the oxide in the form of resolvable crystallites all alloys should be enhanced by irradiation remains constant, or that volume of resolvable (even in strongly “reducing” conditions). Thus, crystallites diminishes with no change (or an under PWR conditions, short-circuited apeciincrease) in the average size. mens of both Zircaloy-2 and Zr-2.5 wt o/0 Nb Studies of irradiated post-transition oxides should show enhanced oxidation compared might concentrate on changes in pore structure with normal irradiated control specimens, and the extent of embrittlement and cracking although initial results on Zircaloy-2 in the with unirradiated laboratory do not confirm this expectation @9). of these films compared oxide films. In order to obtain a clear picture of the behaviour experiments on the pH dependence of the under irradiation, however, it would probably oxidation of Zr-Nb alloys, and y and photobe necessary to study the I-V characteristics
EFFECTS
OF
IRRADIATION
ON
TEE
effects on the same system have already been suggested (section 7.2). The e5Feot of reducing the oxygen content in the G-7 loop has been the only effect studied so far; it would be interesting to observe the effect of deliberately increasing the oxygen content. This would give information on the saturation level for the effect of oxygen on the oxidation rate, and some clue to the level of oxygen needed to reduce the percentage of hydrogen absorbed. 10.
OXIDATION
Acknowledgements The author would like to thank allthose who haye oont~buted to the preparation of this review by oritioising the manus~~pt, supplying copies of figures, and giving permission to use
ALLOYS
43
figures and to refer to work not otherwise available in unclassified form. Glossary AECL
Atomic Energy of Canada Limited, Chalk River Nuclear Laboratories (CRNL), Chalk River, Ontario, Canada (Research Reactors NRX, NRU, NPD *).
AERE
Atomic Energy Research Establishment, Harwell, Didcot, Berks., England (Research Reactors BEPO, DIDO, PLUTO).
Bettis
Westinghouse Electric Corporation, Bettis Atomic Power Laboratory, Pittsburgh, Pa., U.S.A.
CISE
Centro Informazioni Studi Esperienze, Segrate, Milan, Italy (Research Reactor Avogadro I at Saluggia).
EIR
Eidg. Institut fiir ~akto~o~oh~g, W~enlingen, Switzerland.
GKSS
Institut fiir Reaktorphysik der Gesellschaft fiir Kernenergiewertung in Schiffbau und Schiffahrt m..b.H., Reaktorstation, Geesthacht, W. Germany.
Hanford
Hanford Atomic Products Operation, Richland, Washington ; now Pacific Northwest Laboratories, Battelle North-West (B~L), (Reactor PRTR).
Idaho Falls
National Reactor Test Site, Idaho Falls, Idaho, U.S.A. (Research Reactors MTR, ETR).
KAPL
Knolls Atomic Power Laboratory, Schenectady, New York, U.S.A.
ORNL
Oak Ridge National Laboratory, Oak Ridge, Tennessee, U.S.A. (Research Reactor LITR, ORR).
Conclusion
It appears that by considering all the evidence currently available to us from in-reactor experiments, and laboratory studies of the oxidation mechanism, we are able to propose an hypothesis which will explain most of the phenomena on a qualitative basis. The value of this hypothesis will depend on its ability to predict the results of future experiments; several such predictions have been made, and it will be interesting to see whether they are borne out, if and when suitable ex~riments are performed. The addition to the body of data of the results of further purely kinetic experiments does not now seem to be warranted. The conditions under which the phenomena are observed, and their magnitude, now seem to be fairly well established. What are needed are carefully controlled experiments which include sufficient variables to permit some elucidation of the mechanism. Such experiments are difficult to carry out in loops devoted to the testing of protot~e reactor fuel in addition to corrosion experiments, since the operating conditions of the loops are generally set by the fuel development programme. There is a strong case for the provision of in-reactor loop facilities devoted solely to oxidation studies in which all the parameters can be controlled.
OS!'ZIRCONIUM
*
Owned and operated by Ontario Hydra.
44
B.
Saclay
Centre d’Etudes Nuclbaires de Saclay, Gif-sur-Yvette, Seine et Oise, France.
Vallecitos Vallecitos Atomic Laboratory, General Electric Co., Pleasanton, California, U.S.A. (Reactor VBWR).
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EFFECTS
OF
IRRADIATION
ON
TIIE
H. H. Klepfer and D. L. Douglass, ASTM Spew. Techn. Publ. 368 (1963) 118 94) I. H. Dyce, Nucl. Eng. 9 (1964) 253 35 a) K. C. Thomas and R. J. Allio, Trans. ASM 58 (1965) 668 b) L. A. Waldman and P. Cohen, USA Report, WAPD-MDM-8 (1954) W. A. Burns end H. P. M&f&, USA Reports HW-SA-3168 (1963), HW-76636 (1962) and Proc. Symp. Corrosionof Ziroonium Alloys (New York, Nov. 1963) ASTM-STP-368 9 W. A. Burns, USA Report, BNWL-88 (1965) 38 1 J. E. Irvin, El~trochem, Tech. 4 (1966) 240 8)) F. H. Krenz, 1st Intern. Congress on Met&c Corrosion (London, April, 1061) 4b) B. Cox, Atomic Energy of Canada Ltd. Report, AECL-2257 (1965) 4l) R. C. Daniel, M. L. Bleiberg, H. B. Meiemn and W. Yeniscavitch, USA Report, WAPD-263 (1062) 43 B. Cox; private communication to R. L. Dillon, 1 Hanford, (Oct. 1963) 43 W. A. Burns, H. P. Ma&i and R. A. Thiede, ) Quarterly Progress Report, Metallurgy Research Operation, USA Report, HW-76228, (1962) p* 51 44 W. A. Burns, Quarterly Progress Report, Metal) lurgy Research Operation, USA Report, m79766 (1063) p. 6.1 "5) A. B. Johnson, Jr., 23rd NACE Annual Conf. (Los Angeles, 1967) USA Report BNWL-SA-822 4% A. B. Johnson, Jr. and J. E. Irvin, USA Report, f BNWL-463 (1067) 47) J. E. LeSurf, P. E. C. Brysnt and M. C. Tanner, Atomic Energy of Canada Ltd. Report AECL2562 and Corrosion 23 (1067) 57 43) J. E. LeSurf and P. E. C. Bryant, 23*d Annual NACE Conf. (Los Angeles, March 1067) Studi Esperienze, Italy, "f Centro Info~~tio~ Report, CISE-RI36 (1964) E. Cerrai, Italy, Report, CISE-RI40 (1066) 9 61) E. Cerrai, to be published in En. Nucl. R. C. Asher, D. Davies, A. Hall, T. B. A. Kirstein 9 and P. F. White, Proc. Conf. Use of zirconium &Iloys in nuclear reactors (Marianske Lazne, Cz~hoslovaki&, Oct. 1066) p. 233 53 R. C. Asher, unpublished results (1967) 64 B. Cox, unpublished results f 55 D. W. Shannon and R. S. Hope, USA Report ) HW-74756 (classified) referred to in 36) W. Jung-Kiinig, H. Richter, W. Spalthoff and 9 E. Starke; Atomke~ener~e 11 (1066) 47 =) H. Richter, W. Ruckdeachel, W. Spalthoff and E. Starke, Use of zirconium alloys in nuclea;r reactors (Ma&make Lame, Czechoslovakia, Oct. 1966) See also Euratom Reports EURAEC-1725, and EUR-33350. 1796 119661
9
OXIDATION 68
f
9
60
1
81
1
9 R8
4 es ) 88 ) 87 1 68 1
89 1
70 1
71 1
'9
73 1
74) 75 )
OF
ZIRCONIUM
ALLOYS
45
J. Blanchet, H. Coriou, L. Grail, R. Monhot and M. P&as, ibid. a. J. E. LeSurf and P. E. C. Bryant, Atomic Energy of Canada Ltd. Report, MET-I-66 (July 1006) * b. R. T. Popple, G. H. Chalderand L. N. Herbert, Atomic Energy of CansadaLimited Report, EXPNPD-106 (1966) * D. F. Babcock and R. R. Hood, Heavy water moderated power reactor progressreports, Sevannab River, USA Reports DP-905, 915, 925 and 045 (1964) W. P. Maffei, Quarterly Progress Reports, Metallurgy Research Operation, USA Reports: HW82651, 84281 and 84573 (1964) L. J. Defferding, USA Report, HW-67040 Rev. (1962) R. C. Aungst, USA Report, HW-75052 (1962) G. &tberg and H. P. Myers, 36 Intern. Conf. Peaceful Uses of At. En. (Geneva,, 1064) A/Conf. 28/P/410 M. Gronnes, AB Atomenergi, Swedish Report s-351 (1066) A. S. Bain, J. Christie and A. R. Daniel, Atomic Energy of Canada Ltd. Report, AECL-1805 (1064) R. D. McDonald and G. W. Parry, Atomic Energy of Canada Ltd. Report, AECL-1052 (1964) G. W. Parry, Post-irmdiation examination of specimens from the CR-V Zircaloy-2 pressure tube, Atomic Energy of Canada Ltd. Report, M.ET-I-53 (1965) * A. D. Amaev, R. S. Amb~~s~yan, A. A. Kiselev, S. T. Konobeevsky, I. A. Anisimova, A. M. Glukov, L. M. Lebedev, V. A. Myshkin, V. V. Goncharov, E. G. Ivanov, M. S. Orlov, N. F. Pravdyuk, E. P. Ryazzantsevand V. V. Skvortsov, 3rd Intern. Conf. Peaceful Uses of Atomic Energy (Geneva, 1064) A/Conf. 28/P/342 A. D. Amsev, R. S. Ambartsumyan, V. V. Goncharov, A. M. Glukov, K. P. Dubrovsky, E. G. Ivanov, L. M. Lebedev, N. F. Pravdyuk and E. P. Ryazantsev, Kurohatov Institute, USSR Report-IAE-1181 (1966) A. D. Amaev and R. S. ~ba~symy&n, communication
private
A. S. Bain, Deuterium concentration in NPD she&thing; status in March 1966, Atomic Energy of Canada, Ltd., Report, RMI-14 (1966) * R. D. Page, IAEA Conf. Heavy W&or Reactors, Paper SM99-48 and Atomic Energy of c&n&d8 Ltd. Report, AECL-2949 (1967) J. R. Johnson, Trans. AIME 212 (1958) 13; and USA Report ORNL-2029 (1956) C. M. Schwartz, D. A. Vaughan end G. G. Cocks, USA Report, BMI-703 (1952)
46 76)
B.
B. COX, Progress in Nuclear Energy,
Series IV,
COX
loo) D. W. Shannon, USAEC
4 (1961) p. 166 77) J. H.
Crawford
and M. C. Wittels,
Conf. Peaceful Uses of Atomic
Development GEAP-4089
2* Intern.
Energy
(Geneva,
AERE-R4703 (1965) 155
78) M. C. Wittels and F. A. Sherrill, J. Appl. Phys. 27 (1956)
643 B. Cox, R. Murdoch
and R. G.
(1959) sl) M.
(Geneva,
1958) p. 46
193)
and B. Cox, J. Nucl.
Energy
(A)
11
3 (1959)
A.
Sherrill,
Phys.
Rev.
176
J. 0.
Energy
(A and B) 16 (1962)
237
Rogers,
J. Nucl.
Mat.
9
Report,
AERE-
Bettis,
J. Davis,
personal communication
personal communication
J. Electrochem.
Report,
Sot.
113
(1966)
b) J. J. Holmes,
Corrosion Sci-
J. A. Williams,
Spec. Techn. Publ.
Energy
unpublished
380
of Canada
Ltd.
and
Harwell,
and
R.
R. di Pietro, G. Masini,
Sesini,
3rd Intern.
Energy
Conf.
(Geneva,
1964)
95) J. V. Cathcart and F. W. Young, Jr., Corrosion 17 and J. N.
Congress on Metallic
Wanklyn,
(1967) 350 B. Cox and C. Roy,
Atomic
AECL-2350,
this work)
Energy
(1967)
1966)
J. Nucl. Mat. 22
Energy
of Canada
Electrochem.
121 and recent unpublished
99) B. Cox, Atomic
3rd Intern.
Corrosion (Moscow,
97) P. J. Harrop and J. N. Wanklyn,
AECL-2777
L.
G-l-8,
Round
UKAEA
Report,
1098
Grail,
Willermoz,
J.
Symp.
(Paris,
Hure,
M.
Pelras
Corrosion
October
1961)
in
CEA
and
Nuclear (France)
and Energie Nucleaire 4 (1962)
109 111) W. K. Anderson and M. J. McGoff, USA Report, KAPL-2203
113)
B.
(1962)
Lustman
conium Acta
NUCI. Sot. and F.
(McGraw
9 (1966)
Kerze,
Hill,
Met.
9 (1961)
31
Metallurgy
New York,
of Zir-
1966) J. S. Kirkaldy,
880
D. F. Fischer, J. Electrochem. on Zirconium 1962) USA 116) J. K.
Sot. 107 (1960) 506
Alloy
Report
Dawson,
Tech. 4
data
of Canada Ltd. Report,
(see also references cited in
U.
Development
Symp.
(Castlewood,
GEAP-4089 C. Baugh
Tech.
4 (1966)
and J. F. White, 137
117) J. K. Dawson, G. Long, W. E. Seddon and J. F. White,
(1961) 55t 96) P. J. Harrop
Report,
Committee
(1963)
Electrochem.
A/Conf.28/P/574
(1966)
Report,
b) R. D. Misch and C. van Drunen, USAEC
results
F. Barbesino, E. Brutto,
Ltd.
USA
115) a) H. A. Porte, J. G. Schnizlein, R. C. Vogel and
data
Peaceful Uses of Atomic
98)
Chirigos,
114) W. W. Smeltzer, R. R. Haeringand
Atomic
Perona
N.
ASTM
Trans. Amer.
D. H. Nyman
385
unpublished G.
J.
112) C. R. Johnson, D. W. Koch and T. M. Campbell,
93) C. F. Knights and R. Perkins, UKAEA 94)
AERE-R5294
(1962)
Report DM-1114
AERE-R5279;
and J. C. Tobin, ASTM
UKAEA,
of
Industry
124
91) a) V. Fidleris, J. Nucl. Mat. 26 (1968) 51
Cox,
and
Coriou,
H.
1967) 279
(1965)
Minutes
110) H.
ence 7 (1967) 289 and Proc. Brit. Ceram. Sot. 9
B.
Report
109) J. G. Rabe, Birgit Rabe and A. 0. Allen, J. Phys.
eo) P. J. Harrop, N. J. M. Wilkins and J. N. Wanklyn,
93)
81
J. Nucl. Mat. 21
Energy of Canada Ltd. Report,
Chem. 70 (1966) CRNL,
1222 and 114 (1967)
(July
Tech. 4 (1966)
103) B. Cox and Mrs. J. A. Read,
(1959)
UKAEA
Hillner
AERE-R4459
66) K. Alcock and B. Cox, UKAEA
*B) R.
J.
2 (1965/1966) D.
211
s*) E. Hillner,
and
Robin tests on autoclaving procedure for Zircaloy-
435 and M.
87) A. Bancroft,
(1964)
(1967)
WAPD-TM-307 197)
R3114
310 and UKAEA
AECL-2668 106) E.
543 17 (1963)
Electrochem.
105) B. Cox, Atomic Stiegler and F. A. Sherrill,
s4) J. Adam and B. Cox, J. Nucl. Energy (A and B) *5) J. Adam
Report,
(1966)
83) J. Adam and B. Cox, Phys. Rev. Letters 3 (1959)
(1963)
J. N. Wanklyn, (1967)
and F.
s2) M. C. Wittels, J. Nucl.
UKAEA
104) P. J. Harrop and J. N. Wanklyn,
31
C. Wittels
Letters
Report
(1964) and Brit. J. Appl. Phys. 16
UKAEA Report, AERE-R4779 Nucl. Mat. 16 (1965) 290
Sowden, 2* Intern. Conf. Peaceful Uses of Atomic Energy,
USA
102) P. J. Harrop, M. J. M. Wilkins and J. N. Wanklyn,
Dawson,
8o) J. Adam
1962)
rol) P. J. Harrop and J. N. Wanklyn,
1958) p 679
78) J. K.
Symp. Zirconium Alloy
(Castlewood,
J. Nucl.
Mat.
25 (1968)
179
118) R. A. Plot, Atomic Energy of Canada Ltd. Reports, AECL-2751 and AECL-2794 (1967) 119) P.
Gondi
and
EURAEC-1000 (1967) 223 12”) N. F. Mott,
F.
Missirolla,
(1964) Adv.
Phys.
121) N. Ramasubramanian, Ltd., 122) L.
Report
Tech.
4 (1966)
and 157
16 (1967)
Report
Cimento
28
49
Atomic Energy of Canada
AECL-3082
L. Anderson
Euratom
and Nuovo
(1968)
G. R.
Hill,
Electrochem.
EFFECTS
OF IRRADIATION
ON
THE
188) J. S. Llewelyn-Leach et al., J. Electroohem. Sot. 110 (1963) 680; lll(lQ64) 781 and 113 (1966) 105 124) B. Cox, Atomic Energy of Can& Ltd. Report, AECL-2776 (1967) 125) B. Cox, UKAEA Report, AERE-R4348 (1963) 186) B. Cox and C. Roy, Atomic Energy of Canada Ltd. Report, AECL-2519 (1965)
OXiDATrCIV
OF
ZIRCONIUM
ALLOYS
47
127) H. D. M&f& and D. W. Shannon, USA Reports, HW-67437 (1960) and HW-72266 (1962) * lnternsl reports; copies available on request from Atomic Energy of Canada Ltd., Chalk River, Ontario.
~~~~o~tion~hyd~~~~~
e& ~pp&~~~~t ind&endante da processus d’oxydation et est au& affect& prarla ptisence d’esp&aesoxydantes mais 8, u.n degr6 diff&en$ de celui du prooessus d’oxydation. Une hypoth&e baa& sur un d6plaeement &ns la balance entre 18s proeessuf3 ds transport ~l~t~~q~ et ion&e & tmvem le film durant ~~~di~t~on (due Lp une asnsibilit$ diff&mte dea 2 proem via-&-vis de la mdia$ion et des conditions chimiqueafa.tstemontr6v s~c~~tible d’expliquer bs ob~~&t~~~ d’un point de vue qualitatif.
EB w0rdt3n~~~~~t~lle D&m i&r den ~~~ von S~ltp~~ten~ sowie Neutronen- und Gammabestra&mg anf die Oxydation von ~ir~~~l0~e~ge~ ~~mrn%nf~nd b0tr~~~tet, Dab& zoigt eioh, dasa sowohl Be&r&lung mit sohnellenKeutronen, al%such freier Sauemtoff o&r oxydierend w~r~~~d~ radiolytiadm-4 M&erial gl#~~~zeitigvorliegenmijseen, damit ein Xffekt in Wasser bai hohen Tern~r~t~n beobaohtet werden kann. In Wasser, wo die Konzentration oxydierender Stoffe gering ist, zeigen Proben, die &ch vorher bei 800 ‘C in einem Autoklaven Lee dorm&a exp&imentales EFLZP l’effet des fragmenti befanden, eine st&rkem Zunahme der ’ Qxydationsals diajenigen Proben, welehe bd de f&&on, de ~i~i~ti~~ par dea neutzwm et des ge~~~~digkeit oder im g&&ten Zustand rayons y sur l’oxydation des alliege~ de zirccsnium tiefertmTempn&umn sont psf&3s en revue. vorbehmdelt wurden. Die Wa~ser&off&hahme ist vom Oxy~~t~o~Bvorg~g; I1 est montr8 qu’B la foie l’irradi~tion par fes im w~~~tli~hen ~~b~~~~ neutrons rapide0 et la. pr&mae, ~oxyg$ne libre ou au& wird sic dumb oxydierende St&e beei~~st~ d’esp&errradiolytiques oxydantes doivent Btm simmaber Bnders als der Oxydationsproze~, Die ~eo~h~g~ri l-en s&h qu&t&iv denten, tai&ment pr&ents avant qu’nn effet pui@3e Qtre ottserve dans l&u & ha&e temg&&.zre. rtthne l’eau wenn eine Verachiebung des ~l~i~~e~~ht~ xwimhen air la co~~~tr~tion des esp&aesoxydantea est f&b&, dem elsktroniaohen und ionischen ~~~o~proz~~ des Bohantillonnsoumis Etuparrwrant B 1’EbuLoolave B bei B&r~hlung sngencnrnnen wird ; diesaVerschiebung 400 “C mont~~t une phrs gmnde acc&%ration de Itt beruht auf der ~ter~~~d~~hen Emp~dlichkeit der vitesse d’oxydation que 1~ ~~h~ntillonB~r~~~~i~ beiden Vorg&nge gegantiber chemisohen und Strahaux temp&&nrea inf&ieuma ah dans S&at &cap& ~hmg~bedingungen. 1
2
1.
B. COX
Introduction
Zirconium alloys have been extensively used as fuel cladding in water cooled reactors throughout the world 19a). More recently the use of thersealloys for pressure tubes in similar reactors has been increasing 3). The lifetimes generally expected for these components are 3-P years for fuel cladding at a temperat~e of 300-340 “C, and up to 30 years for pressure tubes at 250-300 “C. Extrapolations of laboratory data for oxidation and hydrogen uptake (a natural corollary of oxidation in a hydrogenous environment) suggest that, if neither process were accelerated by irradiation, no limitation on component life from either process is likely. The low failure rate of zirconium alloy clad fuel confirms this optimism 2) as far as current operating conditions are concerned ; however, the desire to operate under more rigorous conditions and to guarantee a 30 year life for pressure tubes requires us to make long extrapolations from the security of laboratory tests and current operating experience. SufEcient examples of greater than expected oxide film thicknesses and hydrogen contents have been reported to ensure that acceleration of both oxidation and hydrogen absorption processes can neither be ignored nor readily demonstrated to be within tolerable limits under all desired operating conditions for zirconium alloys. It is important, therefore, that we not only develop empirical rules for predicting the occurrence and magnitude of such effects, but that we gain an understanding of their cause, if the full potential of zirconium alloys in water-cooled reactors is to be exploited. The history of irradiation effects on zirconium alloys is now quite extensive, and it is useful to review this at the present time; firstly because the field has become sufficiently large that many of the experiments which would help towards an explanation tend to be forgotten or ignored, and secondly, because we may be able now to offer an explanation leading to predictions which can be tested experimentally. Thus, it is hoped that this review may lead the reader to devise critical experiments which will take
the proposed mechanisms out of the realms of speculation. It should be borne in mind that, although most of the experimental data refer to the oxidation process and our knowledge of this process is considerably better than our knowledge of the hydrogen absorption process, yet it is the latter process which is generally critical where the permissible operating limits of zirconium alloys are concerned. There are circumstances where the thickness of the oxide film formed is important (e.g. on heat transfer surfaces) or where metal loss may be critical (e.g. the loss in thickness of high strength zirconium alloy pressure tubes over 30 years), but in general the embrittlement of the component by hydrogen is the most feared result of the oxidation process. The oxidation of zirconium alloys in the absence of irradiation is usually considered to proceed initially according to an approximately cubic rate law (pretransition oxidation); this period is followed by a transition to a more rapid, approximately linear oxidation rate (post transition oxidation). The significance of these periods will be discussed in more detail in section 6. 2.
History
2.1.
ENHANCED OXIDATION
IN FISSILE
SOLUTIONS
The first observations of the acceleration of the oxidation rate of a zirconium alloy in an aqueous environment in-reactor go back to 1954 and the initial autoclave and loop tests carried out in uranyl sulphate solutions at ORNL * for the Homogeneous Reactor Project 4). At very much the same time workers at KAPL 5) and Bettis 8) were reporting no effect of reactor ideation on Zircaloy-2 after 2500 h in PII 11 water at 580 “F (305 “C) containing 500 cm3 hydrogenpitre. Thus, apparently contradictory experimental results were present from the earliest stages of the investigation. There was no doubt about the presence of * See glossary of names and initials at end.
EFFECTS
OF
IRRADIATION
ON
THE
OXIDATION
OF
ZIRCONIUM
ALLOYS
3
the
quite accurate for an autoclave without specimens, but needed some interpretation in the
absence of irradiation the oxidation of zirconium alloys in uranyl sulphate solutions (fig. 1) was virtually identical with that in high temperature
case of the more complicated loop tests. Oxygen pressure drop measurements during in-reactor autoclave experiments (fig. 2) showed
water 7). Effort was concentrated, therefore, on understanding the factors affecting the enhanced oxidation, and searching for alloys with im-
that the oxidation-kinetics were linear. A short incubation period was sometimes observed at
proved corrosion resistance. Experiments were carried out either in autoclaves 8) made of the alloy under test (with or without separate corrosion specimens supported within them) or in circulating loops 9) containing groups of test specimens. Whereas in most experiments re-
easily measured 4). This might
accelerated the
oxidation
size of the
in the ORNL
effect
tests since
was so large.
In
ported here oxidation was assessed from weight gain or oxide thickness, in these tests it was assessed : 1. From specimen weight loss * ; the oxide film was either lost during the experiment (loop tests) or could easily be removed from the dried specimen (autoclave tests) ; 2. from the rate of consumption of the oxygen overpressure (required for the stability of the uranyl sulphate solution). The latter technique gave an average value for the oxidation rate * which was generally *
Usually
equivalent
expressed as mils (0.001”) of metal lost;
to 590 mg/dm2 weight
used elsewhere.
An oxidation
gain in the units
rate of 1 mil/y (MPY)
is equivalent to w 1.62 mg/dmg.day
in the other units.
the beginning
of the experiment
but was not correspond
to
the pretransition part of the normal out-reactor oxidation kinetics, but generally the oxidation rate was so high in the ORNL tests that it was impossible to confirm this supposition. The first measurement in fig. 2a lies on the linear rate line so that if the period prior to this point represented pretransition oxidation the weight gain at transition must have been reduced to Q 15 mg/dmr from the normal out-reactor value of m 30 mg/dmz. Over a similar period of time parallel studies were being carried out at AERE 1% ii), where facilities did not permit operation at as high a fission power density as could be obtained in the ORNL tests. Since it was impossible to achieve power densities typical of homogeneous reactor operation at AERE, workers there concentrated on examining the very low power density region, where the change from normal out-reactor oxidation kinetics was occurring. Results from AERE (fig. 2b) showed that the oxidation kinetics in uranyl sulphate solutions were formally analogous to normal out-reactor behaviour. Thus, a pre-transition oxidation period with an exponent in the equation
-I
Aw=ktn+c,
UNIRRADIATED
Fig. lated
1.
A comparison
from
various
of Zircaloy-2
the decrease
in oxygen
corrosion calcuoverpressure
in
uranyl sulphate environments at 250 “C and 290 “C [ref. ‘)I. (1 mil = 25.4 pm).
of between 0.3 and 0.5 was followed by a period with n w 1. The effect of irradiation in fissile solutions was to accelerate both pre- and posttransition kinetics with little change in the exponent of the pre-transition kinetics and a reduction in the weight gain at which the transition took place. Because of oxide apalling, a problem in all uranyl sulphate experiments, weight gains could not be followed far beyond transition. It was concluded, however, that the
4
B.
00X
0.24
0.46
USING
27O’C
DATA
RADIATION
TJME (hr
01 3’Mwl
A typical measurement of corrosion by following the consumption of the oxygen overpressure; ORNL Expt. L52Z-136 [ref. I@)]. (1 mil G 25.4 pm; mpy 3 mil per yeer). are-transition exponent was closer to 0.5 under irradiation than in its absence, and that the
where E and L are constants. was based on a postulated
weight gain at transition was reduced from the expected value of m 30 mg/dm2 to a value of FW20 mg/dma. The ORNL results indicated that this weight gain was reduced even father at higher doses rate. Experiments at ORNL revealed a complicated dependence of the oxidation rate on the fission power density in solution. The oxidation rate was not a linear function of power density but tended to a satiation vaIue at high dose rates. Early efforts to fit this dependence gave a relationship between the corrosion rate (R) and the fission power density in solution (P)raised approximately to the 0.4 power 7912). This was elaborated to give a relationship
corrosion effects in which defects in the oxide lead to a loss of protective oxide. However, after further experimentation 1% 14)
R=EP[l-exp
(-L/&5)],
it was concluded that a better by an equation of the form
This expression model for the
fit was given
l/R = (&/(KfW} + (l/K), where for a given material, K is a constant for a given ~mper&ture and solution, Kl is a constant for a given temperature, and b: is the factor by which the effective power density is greater than the solution power density due to uranium absorption at or near the corroding surface. In this instance the accelerated oxidation was explained on the basis of radiation damage in the metal substrate. Plots of some
EFFECTS
Fig.
2b.
OF
IRRADIATION
ON
TEE
OXIDATION
OF
ALLOYS
5
Irradiation corrosion of Zircaioy-2 in 300 ‘,C nrenyl sulphate solutions followed by weight gain [ref. lo)]. f. IL III. IV. V. VI.
Unirradiated Neutron/Gamma, radiation only Power density 0.0125 Watt/ml. Power density 0.025 Watt/ml. Power density 0.0375 Watt/ml. Power density 0.075 W&ttjml.
28-
iI P
APPROXIMATE
FOR EXPERIMENTS
I
’
I
1
I
’
I
I
1
,
1
,
l
* 28O’C. NoACID
. .l I//__
-
-2o-
? I &SB 92-
CONDITIONS
Conditions: 300 ‘C : 200 psi 00 (cold) Solution: O.OSM KrSO4; 0.02M HaSOr; 0.005M CuS04+=5UOaS04 to required power density.
~~
-24-
-
R1 J)
a/t4/
fI=i*5p 25O*C,WITH
0
2
4
6
8
FlS6K&YA’Ek4DEN&Y 10
18
(W/ML)
Fig. 3%
ZIRCONIUM
Some ORNL
autoclave
(hi@“’ H&O,
R’,‘)
20
22
results for Zircaloy-2 in uranyl snlphate solution per year = 26.4 ,um/year).
24
26
[ref. la)]. (mpy = mil
EXPERIMENT
SOLUTION
Fig. 3b.
FISSION
POWER
DENSITY
A
o
L-2-19
B
,
C-4-16
8’
v
L-4-16
c
A
L-2-15
E
t
L-2-10
(ft
per
six)
b/ml1
Some ORNL in-reactor loop results for Zirceloy-2 in urenyl sulphate solution at 280 “C [ref.
le”)l.
0.60
0.45
0.40
0.35 h ;
0.30
.-_ 5 0.25
EXPERIMENT
0.20
(ff per sc
O.iS
o.*o
Q(H
0
t SOLUTMN
Fig. 4,
ORNL irradktion
F,SStON POYlER DENSITY
(ml/w)
corrosion data in umnyl sulphate solution at 280 “C fitted to eq. (2) [ref. la)], (I mil 3 25.4 pm).
EFFECTS
of the ORNL
OF
IRRADIATION
results according
ON
THE
oXIbA!cION
to the above
tests
OF
were
ZIRCONIUM
carried
out
7
ALLOYS
with
uranyl
sulphate
relationships are shown in figs. 3 and 4. A wide range of zirconium alloys was examined at ORNL in an attempt to find alloys less affected by irradiation in fissile solutions. However, the
solutions depleted in z%U, in sulphate solutions containing no uranium, and in non-fissile solutions with hydrogen instead of oxygen
range of variation from alloy to alloy was small (fig. 5), and results on unalloyed zirconium were
major source of energy deposition was from the fast neutron and y radiation from the test
comparable
reactor (MTR).
with those for Zircaloy-2
overpressures 437915). In these experiments
7). Thus,
the effect of irradiation in fissile solutions largely swamped the variations present from alloy to alloy out-reactor, and suggested that hopes of producing zirconium alloys resistant to this type of attack were likely to be remote. During the course of these studies autoclave
It was observed
rates in HzO, DzO and depleted
that oxidation UOzSO4 with
oxygen overpressures were identical with those in other tests when plotted as a function of the fast neutron power density (defined as the energy deposited per cm3 of the solution from fast neutron interactions).
ALL COUPONS DEFILMED IMPACT SPECIMENS NOT DEFILMED
“A-
900°C,
3hr.
WATER
QUENCHED
6 -900°C. 600°C,
3 hr. WATER OUENCHED 46 hr. AIR COOLED
C - 65OOC.
30 min
D-900’C. 600°C.
3hr. WATER 3 hr. WATER
;
OUENCHED: OUENCHED
0 0
2
3 POWER
the
DENSITY
Fig.
5e
4 (w/ml)
5
6
8
3. C~MER~~ALLY
a-2
0.2
ANNEALED
Zr-tS
Nb
9Oo.C‘
2-3
Zr-fS
Nk
SOCTC,
Z-3%,
WATER
QUENCHED;
4oO’C,
2 week,
AIR COOtED
Mb
9OQ*C,
2-Jhr,
WATER
QUENCHED;
4oO’C,
i week,
AIR COOLED
40Q”C,
1 week,
AIR
COOLED
SOO’C,
f wwk,
AIR
CCCLED
Zr-
“0
00x
(5
WATER
Ouf NCHED
Zr-fS
Nb-
9OcJ*c,
z-3flr.
WATER
QUENCHED;
Zr-f5
Nb-tMo
9QO°C,
2-3hr,
WATER
OUENCHED;
Zr-fS
Nb-fCu
S’OO’C,
2-3
hr,
WATER
QUENWED;
Zr-f5
Nb-f
9OO’C,
2-3
hr,
WATER
QUENCHED;
0.4
ZMo
hr,
Cu
0.6
0.8
f.0
, SOLUTION
FISSION
POWER
DENSITY
1.2 (*/ill
1.4
i.6
i’
Fig. Eib
Fig. 5% b,
ORNL results for corrosion of & munber of euperim%ntal zirconium alloys in m%myl f3uIpllata
App~nt exceptions to this were the experiments in water at 290 “C with a hydrogen overpressure, which showed oxidation rates little different from those during normal laboratory tests. The apparent discrepancy between the published oxidation rates in these tests and out-reactor oxidation rates arises from the ORNL assumption of linear kinetics throughout their tests and the calculation of an oxidation rate on this basis, If the experimental data are converted to a weight gain, they lie very close to the normal out-reactor oxidation curve (fig. 6). These tests, not well knowu to investigators working only in high ~m~~rature water, provided the first clue to the apparent dis-
crepancy between results obtained in water at different establishments.
Throughout the period 1956-1960, experiments in high temperature water loops 1**17) and examination of fuel cladding from pressurised water reactors 18) had failed to reveal any acceleration of zirconium alloy oxidation by reactor irradiation in (pH10 LiOH with M 30 cm311 hydrogen overpressure (although results from the ~xa~nation of fuel cladding did not become available until early 1962). The ORNL experiments with oxygen overpressure had ~ndio~~d, however, that at equivalent rates of energy deposition in the solution fast
EFFECTS
OF IRRADIATION
ON THE OXIDATION
OF ZIRCONIUM
ALLOYS
1000 1
1
I$ _ 00
VBWR
8
RECENT ORNL (OPEN
DRESDEN DRESDEN
FUEL
CLADDING
CLADDING
AUTOCLAVE TESTS + O2 .CLOSED +
IN MTR Hz1
BETTIS
DATA
FOR
, OYlOATlON /
.’
I
Ill
OUT --REAiiOR 290%
AT
.’
-L
1
IO00
TIME Fig.
6.
Oxidation
in neutral boiling water at m 300 “C of Zircaloy-2 compared
with ORNL
neutrons were as effective as fission fragments at accelerating the oxidation of zirconium alloys I4b). It was some time before the fact that all PWR studies were carried out with high hydrogen over-pressures was correlated with the low oxidation rates in ORNL experiments with a hydrogen overpressure and a possible explanation of the apparent discrepancy between results in water and uranyl sulphate suggested. Subsequent examination of fuel cladding from the Shippingport and Saxton reactors has failed to detect any increased oxidation during in-reactor exposures of up to 1157 days at full power 18119). 2.3.
(days)
EFFECTS OF VARIOUS BOMBARDING SPECIES
Concurrent experiments using other bombarding particles were carried out in an attempt to determine the relative efficiencies of the various types of radiation. No accelerating effect of electron irradiation on Zircaloy-2 was
in-reactor
fuel cladding from VBWR
and Dresden
data.
observed by workers at ORNL ro), but Glick at Bettis reported acceleration of the oxidation of zirconium by a factor of ten in some experiments with deuterons 20). The ORNL work was carried out in oxygenated solutions and appears to be good evidence of the negligible effect of electron (and presumably also y) irradiation in the absence of fission fragment or fast neutron irradiation. The Bettis work is more ambiguous; it was conducted in nominally degassed water at 315 “C, but since the critical effect of water chemistry was not appreciated at this time it is likely that conditions were relatively oxidising during bombardment. Thus deuterons, like neutrons, will accelerate the oxidation of zirconium under suitable experimental conditions. Calculations of the relative efficiency of various types of radiation in producing damage in ZrOz in the form of interstitials and vacancies 21) did not give good agreement with the experimental oxidation data. However, the ORNL results in fast neutron fluxes 15) gave
10
B. COX
hope that an empirical correlation might be used in which power densities for fission fragment and
lowered, but satisfactory post-transition data could not be obtained for specimens originally
fast neutron damage were equated for purposes
in the pickled condition
of calculating
without long in-reactor
the oxidation
rate. Such a calcu-
exposures
lation led to the prediction
that a fast neutron
cant) decision was made to establish the degree of acceleration of the post-transition oxidation
flux of 101anf/cmz. set should lead to an oxidation rate for Zircaloy-2 equivalent to a fission power density
of
0.015
W/ml
solution 21). Experience
in
uranyl
sulphate
had shown that such a
power density led to an increase of about a factor of two in the pretransition weight gain of Zircaloy-2 at a given time, with a correspondingly larger factorial increase ( x 8) in the post-transition rate lop ll). Although exposure of Zircaloy-2 to PWR conditions, at considerably higher fast neutron fluxes, had not given any evidence for such an acceleration in the oxidation of the Zircaloys, the experimental details of this work had yet to be published and only verbal statements of this observation were available. Accordingly it was decided at AERE to institute a deliberate search for such fast neutron effects 14121).
3. 3.1.
Enhanced oxidation of Zircaloy-2 in irradiated water and steam EXPERIMENTS AT AERE,
HARWELL
Small 1 atm steam circulating loops for insertion in the DID0 or PLUTO reactors were built to employ specimens which could be handled easily when active 2% 23). That fission fragments and therefore presumably also fast neutrons caused acceleration in other than the aqueous environments had already been shown by the results in air of Yee et al. 24). Initial experiments both at AERE and elsewhere employed predominantly Zircaloy-2 and it is with respect to this alloy that the ensuing discussion is directed. The experiments in a high fast neutron flux (w 3 x 1013 nrvt) in the DID0 reactor soon revealed an increase in oxidation under irradiation at 340 “C of about the predicted magnitude 2% 25). Pre-transition oxidation rates were accelerated by about a factor of two, the weight gain at transition was apparently
(fig. 7). Thus a further (very signifi-
rate of Zircaloy-2 from specimens taken beyond the transition point at high temperature (450 “C) and then continued
at the normal in-
reactor experimental temperature (340-350 “C) until the “memory effects” had largely disappeared. Experiments using this type of specimen (figs. 8, 9) showed accelerations of up to a factor of ten in fluxes of m 3 x lOi n&m2 . set se-2a). Using these techniques the irradiation enhanced oxidation was shown to disappear with increasing temperature. The acceleration was barely detectable, above the normal oxidation rate, at 400 “C, and was definitely absent at 450 “C! (table 1). Earlier tests in high pressure steam in the BEPO reactor, which apparently showed an effect at temperatures above 400 “C were shown to have resulted from faulty positioning of thermocouples 2% 25). At lower temperatures there did not appear to be a progressive increase in the acceleration factor with decreasing temperature in experiments at high fast neutron flux. At both 300 and 350 “C post-transition oxidation rates appeared to be increased by factors of 8-12 (table 1). The effects of neutron flux and neutron/ gamma flux ratio were investigated in several positions in the DID0 and PLUTO reactors (table 1). These showed that the increase in oxidation rate diminished with decreasing fast neutron flux, and that gamma flux was unimportant. The fast flux in the BEPO reactor was too low to produce any effect 28-28). Experiments in high pressure steam (500 psi) in a relatively pure gamma flux (DID0 reactor cooling pond) apparently showed a small effect (fig. 10) on the early stages of oxidation [interference colour region 25)]. This was ascribed to an effect on the electrical properties of the oxide which were thought to control this stage of the process. Subsequent work on the electrical
EFFECTS
OF IRRADIATION
ON TEE
OXIDATION
OF ZIRCONIUM
BASED REF.
SHOW
II
1
I
I
STANDARD
I
Fig.
7.
11
cg
125
DEVIATION
I
I
IO TIME
ON
ALLOYS
I 100
I
I
I
I 1000
(DAYS)
Effect of irradiation in the DID0 reactor (w 3 x 1013n&m2*sec* , w 5 x 108 R/h y) on the oxidation of Zircaloy-2 in atmospheric pressure steam at 340 “C [ref. 26)].
l undar
radiation
Fig. 8. Effect of irmdktion in the DID0 retitor on the oxid&ion of pre-oxidised Zircaloy-2 in atmospheric pressure steam st 300 “C [ref. 26)].
properties of ZrOz (section 5.6) showed that an effect of y radiation on these was unlikely, and the observations were discredited. However, the subsequent experiments, purporting to show no effect of y radiation under similar conditions 2*), were not comparable as they were performed in 1 atm steam rather than in high pressure steam. If as now seems likely there may be a chemical component to the irradiation effect, then radiolytic species capable of producing an acceleration in the oxidation would
be more readily available at the specimen surface in y-irradiated steam at high pressure, than at 1 atm (section 5.8). The AERE observations of an acceleration of Zircaloy-2 oxidation by a fast neutron flux in a relatively pure aqueous environment remained unique, and largely unbelieved, during the years (1960-1962). This period was devoted to demonstrating that errors in the knowledge of the specimen tempaerture could not account for the observed effect. However, two additional experimental programmes eventually provided information supporting the AERE work. These were the examination of fuel elements from the reactors VBWR and Dresden-l 29-31), and the examination of specimens from the Zircaloy-2 surveillance programme in the Hanford G-7 loop in the ETR reactor at Idaho Falls 32).
3.2.
EXAMINATION
OF FUEL CLADDINU FROM
BOILING WATER REACTORS
The examination of Zircaloy-2 fuel cladding from VBWR and subsequently from Dresden which had operated in neutral boiling water with radiolytic hydrogen and oxygen showed oxide films much thicker than anticipated (fig. 6). These increased approximately linearly
12
B.
COX
95 -
“; 900 E” I- as3 3” 80z TYPICIL
s ?I 75-
CONTROL
SPECIMENS
m : -
_________-----
_--_--
I___-__--
70-.
r_-_-_-_
___-- -60-
(FIGURES 1 125
I IS0 TIME
Fig. 9.
- -
Effect of irradiation in the DID0
IN
e-
-_-_=---=_-%
______-_------
IN ABSENCE OF RAD~AT,oN UNDER RADIATION PARENTHESIS
I 175 (EXCLUDING
ARE
I 200 PM-OXIDATION)
FOR
(f0)2)
I 225
DAYS
reactor on the oxidation of Zr-2.6 wt y0 Nb in atmospheric! pressure steam at 300 “C [ref. 28)].
Fig. 10. Effect of a gamma flux on the oxidation of Zircaloy-2 by 500 psig steam at 350” C [ref. 26)]. Billet Z 8000: l Gamma flux 105 r/h; A Gamma flux 7-9 x 105 r/h; H Gamma flux 2-5x 106 r/h.
EFFECTS
OF IRRADIATION
ON
THE
OXIDATION
OF ZIRCONIUM
ALLOYS
13
TABLE 1 AERE,
Harwell results for the factorial increase in oxidation rate of Zircaloy-2 under irradiation in atmospheric pressure steam Post-transition region
Pm-transition region
1 1.5
Fuel element rigs 2-3 x 1013 nfvt 5x108 R/h (y)
(up to 6) s-s* 8-12 *
1 1.5? *
2V rigs 2 X 1012 nfvt 3x10’3 R/h (y)
4-9
6-8 *
450
1
1
400
1t
1
350
1
1
300
* t
I
1
Ampoule rigs in BEPO 1 X 1011 nfvt 2 x 106 R/h (y)
I
For Zr-2.5 wt% Nb. Similar result obtained in 02 at 80 mm pressure using balance rig.
with exposure time in-reactor to values of M 700 mg/dm2 at 1200 days. Happily the hydrogen concentration in the fuel cladding at any given time was not significaqtly higher than was expected from an extrapolation of laboratory data. Although the amount of oxidation had increased by a factor of w 50, the fraction of the hydrogen released by the oxidation which was absorbed had decreased so as to approximately balance this. Thus, the main practical concern was about the
effect of the thick oxide film on the heat transport properties of the fuel cladding, and the possibility of runaway oxidation due to the rise in temperature at the metal/oxide interface 33-35). 3.3.
EXPERIMENTS LOOPS
IN PRESSURISED WATER
Examination of the specimens from the G-7 loop in the ETR reactor at Idaho Falls also showed thick oxide films3% 35-39,and in this
14
B.
1
a.
Zircaloy-2
COX
Lot AT
200
200 Time
at Temperature,
Days
ON
instance also high hydrogen concentrations, although nominally operating under PWR conditions of pHl0 LiOH but without added hydrogen. Weight gains and film thicknesses increased approximately linearly with time, at a given neutron flux (fig. 11) but were generally greater than the weight gain expected from AERE studies for oxidation under irradiation. The effect of small differences
in fast neutron
flux from experiment to experiment could be allowed for by plotting the results as a function of dose (fig. 12). The oxidation rate did not increase steadily with increasing fast neutron flux, however, but showed evidence of saturation between 1013and 1Ol4n&m2.sec (fig. 13). Thus plotting results as a function of dose only corrects for experiments at fluxes below the saturation level where weight gain is apparently a more nearly linear function of flux. During this period (1960-1963) experimental work at CRNL failed to give a clear indication of irradiation effects sgl 40), some tests apparently showing them and others not. The fast neutron flux in these experiments was known
1000 1 -
n
AT-50
.
HT-37
_ A CT-19 _
5
100
0
AT-50
0
HT-37
A
CT-19
Irradiated G-l
In
Loop
Unlrradiated Controls
7
2 k
10
1
100
Time
-
500
iDays
(b)
Typical results for oxidation of ZircaFig lla, b. lay-2 in water with LiOH or NHdOH additions at 280 “C a
a function ETR
of time in the G-7 loop in [ref. 469 *8)].
EF’FECTS
.
OF
AT-50
IRRADIATION
ON
TEE
Irradiated in O-7 LOOP
HT-37 CT-19 AT-50 0 HT-37
Unlrradirtsd Controls
c
Out-of-Rerctor Lot (5700 firs.)
2
5
2
10
E
Out-of-Rtrctor (500 Hrr.)
Loop
,* 1.0
0.1
Integrated
10
Exposure
- tnvt
40
x lO_zOf
Fig. 12. The results from fig. llb, obtained in @Xl0 LiOH in the G-7 loop in ETR, plotted as & function of fast neutron dose [ref. a*)].
OXIDATION
OB
ZIRCONIUM
ALLOYS
15
to be relstively low, however, so that not much weight was laid on these apparent disorepanoies. A few examples of oxide films thicker than expected were also observed by Bettis workers 41). These were mainly during loop tests, and predominantly when defected fuel, or fissile mate&l from earlier defects, had been present in the loop. This led some workers to propose that enhanced oxidation was only observed in the presence of fissile material in the coolant. They apparently arrived at this viewpoint largely by rejecting the results of oxidation tests where fissile material could not conceivably have been present, on the grounds that only tests in high temperature water were relevant. Most of the exceptions to this thesis [work at Harwell 22,23, 25, 26, 52), C1fJ.E 49-51)] were not in this environment. Nevertheless it appeared that most of the then available d&a could be explained if both i~adiation and a suitable environment were needed before enh&ncement of oxidation by irradiation could occur. The critical feature of
= 60 -
0
26
@
23
0
19
I
16
0
l
0 40 / -
// /
2o -1
0 x 1013
2
4 Neutron
Fig, 13.
6 Flux
(>
I MN),
8
1 x 1014
II”
Some oxidation results at 280 ‘C from the ETR loop for water (pH10 LiOH) containing m 1 ppm 0~ and < 0.1 ppm 08 plotted as a function of fast neutron flux [ref. s’)].
16
B.
the environment seemed to be the presence or absence of free oxygen, or oxygen containing radicals produced by radiolysis 42). Thus in normal
PWR
conditions
the
presence
of
hydrogen suppresses the radiolysis of the water, whereas in a BWR, taining
radicals
oxygen
are normally
and oxygen present
oon-
in oon-
siderable quantity al). Other experimental results could be rationalised on the same basis, with the exception of the G-7 loop results in ETR. Here the original information was that the water was degassed 43); however, subsequent investigation showed that oxygen was being admitted with the feed water, and that the average oxygen concentration was 0.8 - 1.0 ppm 3% 44). Although the diverse results then available could be correlated assuming that no aooeleration would be observed without the combined presence of fast neutron irradiation and suitably oxidising species in the environment, much work was needed to demonstrate that this correlation would hold for future work. Programmes were started to study the effect of water chemistry on the phenomenon, and the relative magnitudes of the effects of oxygen content and fast neutron flux. The question of whether radiolytio species alone could accelerate oxidation required testing as some of the initial experiments suggested that a “downstream” effect existed outside the neutron flux.
Etched
10ZO Integrated fastNeulron
1021 Flua.
> I MeV
Fig. 14. A comparison of in-reactor oxidation of Z&alloy-2 in the three G-7 loop environments [ref. 45)]. Note : In-flux weight gainsreported only for specimens prefilmed in water at 300 “C, except for etched specimens exposed in pH-IO NH40H pH-10 LiOH, N 1 ppm 02: x W. A. Burns 37) + A. B. Johnson, Jr. 45) A Johnson and Irvin 38,49 PH-IO LiOH, ~0.1
ppm 02: o W. A. Burns 37)
pH-10 NHaOH, < 0.05 ppm 02: q A. B. Johnson, Jr.
showed that the irradiation effect was virtually FACTORS AFFECTINQ THE ENHANCEMENT eliminated for specimens initially in the etched OF OXIDATION BY IRRADIATION condition (figs. 1416). Specimens which had A joint CRNL/Hanford programme for the been preautoolaved in steam at 400 “C before insertion, however, still showed an apparent use of the G-7 loop in ETR was instrumental effect of irradiation (figs. 15, 16). in demonstrating the significance of water Further investigation showed that the magnichemistry. Initial work by Hanford at 280 “C tude of the apparent irradiation effect was showed that addition of hydrogen to the dependent on the precise preautoolaving treatpH10 LiOH reduced the average oxygen ment (figs. 15, 16). Pre-autoolaving at 400 “C content from w 1 ppm to < 0.1 ppm, and gave a much larger effect that preautoolaving reduced the magnitude of the observed irradiat 300 “C, and at any temperature the size of ation effect (fig. 14) 37). Further improvements the effect was larger for thicker initial oxide to the water chemistry in the joint programme films. This observation has a profound effect involved rigorous degassing of the makeup on the interpretation of earlier AERE data, water, and ammonia additions for both PH and where most results had been obtained on oxygen control. The results 45, 46) of these tests 3.4.
EFFECTS
OF
IRRADIATION
ON
THE
OXIDATION
OPEN CLOSED
OF
ZIRCONIUM
ALLOYS
17
SPACES-OUTREACTOR SPACESIN -REACTOR
-PRETREATMENT
Fig. 15.
Effect of preautoclaving treatments on the oxidation of Zircaloy-2 in high and low oxygen water in the G-7 loop [ref. 45)].
specimens taken beyond transition at 450 “C and then further oxidised out-reactor at 340 “C. Failure to observe any effect of preautoclaving in the earlier Hanford experiments in oxygen containing water probably resulted from two factors. Firstly the high weight gains in most of the earlier experiments were accompanied by appreciable scatter in the results, which could have obscured the effect. Secondly, preautoclaving treatments in the earlier tests had generally been at 300 “C, a temperature subsequently shown to give only small increases in oxidation under irradiation. At CRNL loop tests were performed in boiling and pressurised water at M 280 “C to evaluate the effect of ammonia additions on the observed oxidation enhancement under irradiation ~*a*). Results supported the hypothesis that a reduction in oxidising species in the loop water reduced the observed enhance-
ment of oxidation (fig. 17). However, even loop tests where no radiolytic oxygen was detected in the loop water at the outlet from the test section of boiling experiments still showed some enhancement, when ammonia additions were only just sufficient to eliminate detectable oxygen. Higher ammonia concentrations were needed before the irradiation effect disappeared. Thus it appears that transient radiolytic species are at least as important as molecular oxygen in determining the effect. Following the Hanford observation of effects of pretreatment the same phenomenon was observed in the results of the CRNL tests 48). In these tests, however, the effects were all relatively small due to the generally low fast neutron flux, and pretreatment effects were therefore not immediately obvious. Early Hanford experiments 36)had apparently shown accelerated corrosion of some specimens
18
B.
COX
OPEN
SQUARES
OUT- REACTOR
CLOSED
SQUARES
IN - REACTOR
r
Fig.
16.
Effect of preautoclaving
t,reatments on the oxidation in the G-7 loop [ref.
of Zircdoy-4
in high and low oxygen water
45)].
zzzj
loo-
IN-REACTOR LOOP
go-
SO-
OUT-Of-FLUX WITH wITwuT
0X”SEN OXYSEN
OUT-REACTOR
LOW
IN-FLUX
0
0
Q
IJ
n
•p
70-
n
605040 -
Fig.
17.
Results
obtained
at CRNL
for the oxidation of Zircaloy-2 at 270-290 With various water chemistries [ref. 491.
“C in in-reactor
loop t,ests
EFFECTS
OF
IRRADIATION
ON
THE
downstream of the test section; however, these results were not confirmed in later experiments 45). At CRNL, specimens at the outlet to the test section also apparently showed evidence for increased oxidation in some experiments. The differences between these and other specimens were small, however, and it was not possible to state positively that they were outside experimental scatter. Results from the CISE loop in the Avogadro I reactor showed that downstream effects were absent for pickled specimens in a steam-water fog at 280 “C 49-51). This loop was used solely for oxidation specimens (no fuel specimens were present) and so they could be distributed throughout the inand out-of flux regions. (In most other loops the oxidation specimens had taken second priority to fuel tests). CISE results showed accelerated oxidation in a steam-water fog when no additions were made to control radiolytic species (fig. 18). The results at different neutron fluxes show a peak in the weight gain curves at approximately the peak neutron flux (fig+ 19). Hanford workers 46) have observed that specimens with thick oxide films from previous exposure in oxygen-containing water continued fo oxidise at the same high rate when added to a test where ammonia was added to suppress radiolysis. Fresh specimens in this test, showed low oxidation rates (fig. 20). At AERE, irradiated specimens soon showed a normal outreactor oxidation rate when transferred to
Fig. 18. CISE results for the oxidation of Zircaloy-2 in a water-steam fog at 280 “C in-reactor plotted as 5 function of time [ref. “11. Open points, out-reector; closed points, in-reactor.
OXIDATION
OF
ZIRCONIUM
19
ALLOYS
c 9.6 d
27.56
.. .
. \
*
.
.‘
I
~~ 10.5d
I
t
1
*
37.36
I
h
fl
I
I
1.
2
3
4
10
15
20
Neutron
flux
(x10’*)
Weight
rnsj&
gain
Fig. 19. CISE results for oxidation by a water-steam fog et 280 “C in the Avogadro I reactor plotted in comparison with the neutron flux [ref. 49)]. I, 9.6 d&ys; +, 18.5 days; a, 27 days; o, 37.3 days.
conditions 2s). These observations can be explained if the importanf region is that immediately adjacent to the interface at which the reaction is proceeding. For thick porous oxide films this region will be near the metal-oxide interface and changing the chemistry of the environment may have little effect on the production of radiolytic species within the oxide. That, the production of these species is important is shown by the rapid return to normal oxidation kinetics when irradiation is discontinued. Further additions to our knowledge from recent AERE experiments with oxygen and hydrogen additions have been slight, since it does not appear to be possible to control the chemistry of the steam in the atmospheric pressure steam loops or the water chemistry (at present) in the recently designed in-reactor autoclave 52). The experiments (fig. 9) which unirra~ated
showed no effect of adding oxygen and hydrogen to the atmosphere
pressure steam loops may
be unreliable z8). The gases
were added to the -
20
B. COX 1600
1400 N
E
_
.z t z 0) 5 3 2 E E
pH-10 LiOH
1200
z P
pH-10
NH40H
1000
800
600 s AT and
HT
400
200
0
4
8
12
16
Integrated
Fig. 20.
ETR
Out-of-Reactor
I
0
20 Fast
28
Neutron
Flux,
32 nvt
;x
I
36
40
I 44x1020
nvt
> 1 MeV
results showing the effect of changing the waker chemistry during the experiment differences between various batches of material [ref. 45)].
test section, but due to the slow rate of recombination in the gas phase, the oxygen content may not have changed significantly. In addition, these tests were carried out with pretreated specimens already bearing thick porous oxide films ; under these conditions Hanford results show that no change in oxidation rate would be expected on changing the chemistry. Variations in the effect of .neutron flux have been studied at AERE, but the number of points on the neutron flux distribution is insufficient to gain other than a general impression of the flux effect (table 1). 3.5.
24
MISCELLANEOUSEXPERIMENTS
A collaborative programme between AERE and EIR, Wiirenlingen, Switzerland, will utilize the high pressure steam circulating loop at the latter site. Initial experiments were in steam at 450 “C, a temperature at which no effect of irradiation would be expected (based on AERE and ORNL tests). It was therefore not surprising that none was observed 5s). Subsequent experiments at 350 “C have given contradictory results in steam containing relatively low oxygen
and
concentrations and a fast neutron flux of iu 5 x 1Ol2 n/cm2. Experiments in gaseous environments at ORNL 54) and Hanford 55) have shown that accelerated oxidation can be observed in almost any moist gaseous environment, whether the major component is helium, carbon dioxide, or doubts about the specimen air. However, temperature in the Hanford tests, and the recent knowledge about the importance of environment chemistry, render these experiments of little significance in the investigations of the irradiation effects on oxidation. Asher 53) has also reported irradiation enhanced oxidation of Zircaloy-2, but not of Zr-2.5 wt o/o Nb in oxygen at atmospheric pressure. Other experiments have been carried out by Metallgesellschaft AG (in the GKSS reactor near Hamburg) 569 57) and Saclay 58) in small high pressure autoclaves. Problems associated with experiments of this type, where water volume was small compared with exposed surface area (autoclave + specimens) and no control of the conditions inside the small autoclave was possible, render the results of
EFFECTS
these experiments autoclave
designs
OF
IRRADIATION
ON
THE
OXIDATION
OF
ZIRCONIUM
21
ALLOYS
of little value. The in-reactor at AERE
and
GKSS
are
being modified to permit control of the chemistry of the contents. The first experiment in the AERE autoclave (without controlled chemistry) showed large weight gains and unusually high zirconium concentrations in the residual water. This was apparently a colloidal solution zirconium produced during irradiation 52).
of
Many miscellaneous pieces of information have come from the examination of fuel cladding, reactor pressure tubes and other components. Generally the actual operating conditions are not known with any precision, so that it is not possible always to fit these results into any scheme. NPD fuel cladding exposed during the period of reactor operation when the hydrogen (deuterium) concentration of the loop water was above 1 cma/kg has generally shown black oxide films. Fuel removed more recently, since the deuterium concentration has been allowed to fall to its natural value of < 1 ems/kg from corrosion of the mild steel circuit, has shown a significant number of instances of white oxide spots 59). The areas covered with white spots were often adjacent to wear pads and/or wire wrap, where flow conditions may be disturbed, and local chemistry may be different from the bulk (fig. 21). Further examples of thicker than anticipated oxide flms have been found on fuel cladding, corrosion coupons and other components (driver rod housing, etc.) from the HWCTR 60) at Savannah River. The precise operating conditions of these components are not well documented, and it is possible that here too the local water chemistry was sufficiently different from the bulk to permit radiolysis (“oxidising conditions”). Thick oxide films have also been observed on pressure tubes and fuel cladding from the PRTR 61) reactor at Hanford. Here the film thickness was found to vary from end to end with a maximum near the flux centre line and minimum at the ends. Again precise water chemistry information is not available. One example of a thick oxide film on a pressure
Fig.
21.
NPD
Post-transition
fuel
cladding
white
(bundle
oxide
0988)
observed
after
during a period of low hydrogen concentration water
(pH10
LiOH,
on
exposure in the
SW280 “C). Magnification
15 X.
[ref. so)]. The horizontal bar is part of the wire wrap, with one spot weld showing.
tube from the KER-1 loop was also observed 62363). Subsequent examination. of Zircaloy-2 specimens which had been inserted in the primary coolant of the Swedish Agesta reactor 6% 65) (pressurised heavy water at 220 “C) has shown little difference from specimens prepared in the laboratory, either in weight gain, or in the appearance of the oxide film after the initial period in which high weight gains were observed. Examination of fuel from in-reactor loop tests (not containing corrosion specimens) at CRNL has produced some examples of thicker than expected oxide films, although again the precise operating conditions were not well characterized ‘33-6s). Studies in the USSR have been almost universally on a Zr-1 wt o/o Nb alloy, and as such will be reported in the next section. 3.6.
ENHANCED IN WATER
The
OXIDATION AND
alloys most
OF OTHER
ALLOYS
STEAM extensively studied, other
than the Zircaloys, are Zr-Nb
alloys.Work
on
a Zr-2.5 wt o/o Nb alloy at AERE 2*), ORNL 4s) and Hanford 45); on a Zr-3 wt oh Nb-1 wt o/o Sn alloy at Hanford 45) and Metallgesellschaft 57);
22
B. COX
and on Zr-1 wt y@ Nb and Zr-2.5 alloys (some with boron USSR s@+70) lead generally
wt y0 Nb
additions) in the to the same con-
elusions. These alloys are generally susceptible to increased oxidation in the presence of free oxygen in the environment in the absence of
irradiation. Thus, more pronounced
effects of
water chemistry in-reactor are observed than was the case with the Zircaloys (figs. 22, 23). The generation of oxygen-containing species in the test section of a loop can and does lead to enhanced oxidation of these alloys downst~am, whereas no such effect has been positively demonstrated with the Zircaloys. Other than the direct chemical effect of oxygen present, or generated, in the loop there appears to be no added effect of neutron i~adiation on Zr-2.5 wt OjoNb a~cimens. This is true whether pre-autoclaved or not, and irrespective of heat treatment, although the susceptibility to small amounts of oxygen in the water varies considerably with variations in metall~gic~ ~nditions. In practice a small, but definite, decrease in oxidation rate is observed in the test section when relatively
II
I
II
I
I
I
oxygen-free conditions are present (figs. 22, 23). The same situation may occur in the presence of oxygen, but it is obscured by the high weight gains and experimental scatter. Similar effects of molecular,
or radiolytic
oxygen in the environment have been observed in the USSR with both 1 wt y@ and 2.5 wt y0 Nb alloys. The magnitude of the effect is larger for the higher niobium alloy, as would be expected from out-reactor evidence on the effect of added oxygen on the oxidation of these alloys 71). The additional damage to the oxide resulting from a-particle bombardment when 0.1 wt o/o boron is added to these alloys does not apparently increase the oxidation rate. It cannot be assumed that no effect would result if the boron were added to Zircaloy-2, however, as there appears to be no direct effect of fast neutron damage on the oxidation of Zr-Nb alloys anyway. A few experiments at BNWL have included specimens of Zr-3 wt o/o Nb-1 wt y. Sn alloy. The sensitivity of this alloy to added oxygen in the absence of radiation has not been well characterized so that it is difficult to comment on the loop results (fig. 23). However, if it
11
11
11
1
}
11
1
~4D_H)2030405060708090100110t20130140150160m3180190_ EXI’DWRE TIME - DAYS
Zr-2’lr
% Nb
IDD“E P so-
Fig. 22.
Results
P a
SD-
z$
70-
F 3 G
5D-
s
40-
50-
obtained at CRNL for the oxidation of Zr-2.6 wt y0 Nb alloy in in-reactor loop tests (at F* 280 “C) with various water chemistries [ref. 491.
EFFECTS
OF
IRRADIATION
ON
THE
OXIDATION
OF
ZIRCONIUM
ALLOYS
23
OPEN SQUARES-OUT-REACTOR CLOSED
SQUARES -
IN - REACTOR
r -Zr-3Nb-ISn
-
+Zr-2!$
Nb-
B-QUENCH ; 30KCW 24hr 500°C Ann Fig.
23.
ETR
results showing
the effect of pretreatment and low oxygen
behaves similarly to the Zr-2.5 wt y0 Nb alloy with respect to molecular oxygen, then it appears to show a small increase in rate due to neutron irradiation (unlike the binary alloy) and a significant effect of pretreatment. Results obtained on this alloy by Metallgesellschaft AG are of little value due to the inadequacies of their experimental technique 57) which led to high oxidation rates both in and out of the reactor. The only other alloys to have received serious study in a high neutron flux are two experimental alloys (Valloy and Valloy-S) developed at Vallecitos. These alloys are respectively Zr-1.2 wt % Cr-0.08 wt y0 Fe and Zr-1.2 wt y0 Cu-0.28 wt y. Fe. In the G-7 loop in ETR (fig. 24) Valloy behaves comparably with Zircaloy-2 when originally in the etched condition, but gives very high weight gains when preautoclaved at 400 “C, even in loop tests
on the oxidation
of Zr-Nb
a;lloys at 280 “C in high
W&tm [ref. &)I.
containing little radiolytic oxygen. Valloy-S has generally shown the highest weight gains of any etched specimens, but the weight gains under irradiation have been little different from unirradiated specimens. Weight gains of prefilmed specimens have been lower than those of etched specimens with again little effect of irradiation. This alloy, therefore, seems to be little affected by either radiation or oxygen in the water; its general corrosion resistance in 280 “C water is poor, however. 4.
Hydrogen uptake
Most investigators have reported hydrogen contents for their specimens, with but little attempt at oo~elation. The results observed appear to fall into two groups. Firstly, there are the tests which have shown hydrogen contents close to those calculated (considering time and temperature only) despite the obser-
OPEN SQUARES - OUT- REACTOR CLOSED SQUARES - IN -REACTOR
+---Zr-Cr-Fe * VALLOY
___I
-Zr-Cu-Fe-----w “VALLOY -S ”
’
Fig. 24. ETR results for the oxidation of the “Vafloys” ia pHlO NR4UI-I ( < 0.05 ppm 0~) rat280 “C [ref. 4511.
DRESDEN FUEL
DEVELOPMENT
CYCLE
DRESDEN
REACTOR
SHlPPlNGPURT SAXTON
PROGRAMME
IN VBWR
REACTOR
REACTOR
500 EXb3SURE
Fig, 25,
TIME
tdaysl
absorbed by Zircaloy-2 fuel cladding in pressurkd w&+x and neutral boiling water at w 300 “C in VBWR, Dresden, Shippingport and Saxton maotore [ref. sl*)f.
Hydrogen
EFFECTS
OF
IRRADIATION
ON
THE
OXIDATION
vation of much thicker oxide films than were expected. This is the general experience with BWR fuel sQ-31)(fig. 25) and with specimens from the CRNL boiling water loops (fig. 26, points in the absence of ammonia additions) 40),
OB
ZIRCONIUNT
ALLOYS
25
and seems to be typical of situations where free oxygen is readily measured in the environment. Secondly, experiments at Hanford 36-s** 43-49, AERE 25-z*)and CISE 4%50) have generally shown percentage hydrogen uptakes (figs.
. ‘f 1.6-
IO 20 30 40 50 60 70 60 90 loo 110 120 I30 14oI60 160170 180190 EXPOSURE
TIME
- DAYS w
2.4 “E P g $.
I.8 -
z
1.6 -
c
l.4-
*
0
2.2 P.O-
0.8
-
0.6
-
0
0.40.2
0
0
-
PO
l1ll1ll*illlIlll
n
l
I
I
l
(b) Fig.
208,
b.
Hydrogen absorbed by Zircaloy-2 and Zr-2.5 wt oh Nb alloy specimensat w 280 ‘C in in-ma&or loops at CRNL [ref. 493.
26
B. COX
0
340%
100%
/
0
/
Y SO0
t-
0
,‘.
\ +/IN ABSENCEOF RADIATION
QNTERPOLATED FRDM FIGURES GIVEN IN RER 125) I I I 30 40 so 0 A0 mgdm’2 0
IO
/’
-
l
20
IN AhENCE RADIATION (c’ ref. 125)
OF
0 44
_C--
I
I
IO
20
I
30
I
I
40
50
I
I
I
I
60
70
80.
90
60 mgdm-2
Fig. 27. Hydrogen absorbed by Zircaloy-2 from 1 atm steam at 340 and 400 “C during AERE experiments [ref. ?I*
27, 28, 29) similar to those observed in the laboratory (i.e., 20-60 y0 of the available hydrogen), so that the total hydrogen in the specimens was increased in the same ratio as
Irradiated In e-7
tocp
10 1
100
10
1000
Time - IDayak
Fig. 28. Hydrogen absorbed by Zircalosr_tZ; in pH10 LiOH at 280 “C in the ETR G-7 loop operating with w 1 ppm Oa [ref. 88)].
the weight gain was increased by irradiation. Examples are also known 6%72)where hydrogen (or deute~um) uptake of fuel and st~ct~al components has been far higher than could be accounted for by the uptake of 100 y0 of the hydrogen theoretically released by the oxidation reaction (figs. 30,31). Thus the inference is that under some eon~tions {section 8) molecular hydrogen can be absorbed directly from the dissolved hydrogen present in the water. In general, although the percentage uptake in such instances has been high, the total amount of hydrogen absorbed has not been large because the low temperature of operation and the pretreatment of the components led to very small amounts of oxidation under PWR conditions. The enhanced uptake of hydrogen under aviation reported by Blanchet et al 58) appears as likely to be the result of their
EFFECTS
OF
IRRADIATION
ON
THE
OXIDATION
OF
ZIRCONIUM
A0
ALLOYS
27
mg/dm’
Fig. 29. Hydrogen absorption during in-reactor experiments in a water-steam fog at 280 “C in Avogadro I reactor [ref. 49)]. a, 400 “C steam, lab. autoclave; 0, w 280 “C, loop, out-reactor; 0, - 280 “C, loop, in-reactor.
I
I
I 2
I 3 YEARS
Fig. 30.
IN
I 4
I 5
REACTOR
Hydrogen uptake by in-reactor pressure tube from CR-V loop at CRNL [ref. e*)]. Numbers show elevation (El.) in feet.
28
B.
COX
i.EGE!@ CROSS HATCHED SOLID
t 200
AREA
LINE
BROKEN LINE
-
X
180
ZIRCALOY-2 - EARLY MANUFACTURE02 CONCENTRATION IN NPD COOLANT,
IRRADIATED
WITH HIGH
ZIRCALOY -2 - RECENT MANUFACTURE - IRRAQIATED D2 CONCENTRATION IN NPD COOLANT,
s
f
-
ZIRCALOY-2 IRAAOIATED
WITH HIGH
AND Ni - FREE ZIRCALOY-2 - kECENT MANUFA!CTUREWITH LOW D2 CONCENTRATION IN NPO COOLANT.
ZIRCALOY -2 - EARLY MANUFACTURE - IRRAOIATED 0, CONCENTRATION IN NPD COOLANT.
:: - 160
WITH LOW
3 s z z y
140 120
s 3
100
tt E
80
3 a
60 40 20
y______ I 100
__-I
200
I 300
i
I
I
I
I
I
400
500
600
700
800
900
DAYS IN HOT COOLANT
Pig. 3 1.
Deuterium absorbed from @Xl 0 LiOD at w 290 “C by NPD fuel cladding of various types at different periods during the reactor operation [ref. 73)].
experimental technique as a direct effect of irradiation. Gaseous hydrogen from both oxidation and radiolysis would have accumulated in their microautoclaves. Fuel cladding in PWR’s, which generally shows no acceleration of oxidation rate, usually shows hydrogen contents no higher than expected (fig. 25), within the accuracy of determination 18). However, the accuracy of such determinations (and predictions) may be relatively low. Thus, most fuel cladding was preautoclaved ; this increased by an uncertain amount the initial amount of hydrogen in the cladding. The initial amount of hydrogen is usually known only from a few spot analyses on a batch of cladding as received. Thus there are already two sources of error in estimating the starting hydrogen content. Even allowing for errors in measuring the final hydrogen content of radioactive material, there is little laboratory evidence from which to predict the
amount of hydrogen likely to be absorbed when preautoclaved material is subsequently exposed to hydrogenated water at a much lower temperature. Most laboratory hydrogen uptake d&a are for temperat~es of 316 “C and above, and for an initially pickled surface. There have been indications in L few laboratory tests that specimens preautoclaved at a higher temperature may absorb molecular hydrogen from the water for a short time when transferred to low temperature water 64). Thus, all that can be safely said in most instances is that the total hydrogen at the end of life was tolerable. However, this does not help predictions of the behaviour of fuel exposed under more stringent conditions. The examination of fuel cladding from NPD has many advantages over those available to other investigators. This material has been exposed only to heavy water following its initial autoclaving in light water. Thus, any
EFFECTS deuterium
found
OF IRRADIATION in the sheathing
ON
THE
must have
been absorbed during exposure in the reactor. Although metallography has failed to show any increase in oxide film thickness in-reactor (until the recent observation of white oxide spots,
section
3.5), appreciable
quantities
of
OXIDATION
OB ZIRCONIUM
29
ALLOYS
ZrOs, led to the proposal that a transformation to oubic under ideation
was
of the enhanced oxidation. that a transformation from
74) A tetra-
from monoclinic the cause suggestion
gonal to monoclinic transition
in the
could be the cause of the oxidation
kinetics
of
un-
deuterium have been found 72). These seem to have been absorbed during a relatively short
irradiated material had been proposed earlier 7J),
period after insertion in the reactor,
incorrect 76). Further studies showed that a transformation from monoclinic to cubic could be induced by irradiation 77-**), although there was some argument about whether fission fragment bombardment was essential 81, sa), or whether it was the uranium (added to generate the fission fragments) which was chemically stabilising the cubic phase in the presence of any form of heavy particle irradiation (fast neutron or fission fragment) 8% 84).
with little
subsequent absorption (fig. 31). The slow absorption seems to represent a reasonable fraction (M 60 %) of that liberated by the oxidation predieted from the laboratory oxidation data. The initial uptake is quite large, representing several hundred percent of any hydrogen likely to have been released by the oxidation process. It may therefore consist of deuterium absorbed directly from the deuterium overpressure. It is notable that the height of this initial step (fig. 31) has decreased with the decrease in the Da overpressure in NPD 73). The amount absorbed in this initial step appears also to be dependent on the metallurgical history of the cladding; however, even in the worst cases the total hydrogen plus deuterium at the end of life does not exceed about 100 ppm by weight of hydrogen. 5.
Proposed mechanisms for enhancement oxydation by irradiation
of
A large fraction of the experimental work reported so far has been devoted to demonstrating the presence or absence of an effect of irradiation on the oxidation behaviour. Little work has been directed towards studying the mechanisms of such processes. However, this has not prevented the postulation of a multitude of different mechanisms in the past. Before going on to study the present situation and to attempt any explanation it seems to be worthwhile to examine the past explanations in some detail. 5.1. PHASE TRANSFORMATION
IN THE OXIDE
The obse~ation that oxide films removed from zirconium specimens during the homogeneous reactor programme were largely cubic
and had not at that time been shown to be
Irrespective of the final conclusion about the precise cause of the irradiation induced transformation, which appears to lie in favour of the second explanation, measurements of the rate of this transformation showed that it was too slow to affect the oxidation rate over the short periods of time involved in some of the uranyl sulphate studies 1% 79~85). Thus the presence of oxide transformed to the cubic form by irradiation was an effect occurring subsequent to whatever other processes were affecting the oxidation rate. 5.2.
ENHANCED
DISSOLUTION OF Zr02
This was an early explanation proposed by workers at ORNL in the lightof their experimental observation of oxide loss from the specimen 12). In loop experiments it was observed that little of the oxide formed remained on the specimen, but was deposited generally in the unirradiated sections of the loop. These deposits were fairly adherent and did not give the impression of being transported merely in suspension, hence the supposition that the dissolution of Zr02 was being enhanced by fission fragment irradiation, Accurate meas~ements of the dissolutjon rate of ZrO2, after i~adiation to a high fast neutron dose, showed a small increase in dis-
30
B. COX
solution rate at short times 86). However,
this
failure process accentuated
by velocities
of up
was shown to be due to a small increase in the fraction of fine oxide particles as a result
to 40 ft/sec (13 m/set) in the in-reactor loops. Several routes by which the irradiation damage
of irradiation, and was not thought to represent
in the metal might
any increase in the intrinsic
have been proposed.
Thus,
although
demonstrate
this
dissolution
experiment
did
rate.
RADIATION
DAMAGE
IN THE
the oxidation
rate
not
that such an increase was absent
during irradiation, this explanation was not considered as a serious contender. Interest in the phenomenon was revived by the observations that the metallographic thickness of the oxide film on irradiated NPD fuel cladding was sometimes apparently less than that measured after the initial autoclave treatment, and that appreciable concentrations of zirconium in the water could apparently be observed in some loop experiments 40). A careful survey of NPD 87) and Shippingport 88) reactors, and the CRNL in-reactor loops failed to show any evidence for accumulations of zirconium in sufficient quantity to explain the irradiation corrosion phenomenon. Thus again interest lapsed, and loss of oxide by dissolution was discredited as a contributor to in-reactor oxidation tests. Recently we have again come full circle with the apparent observation of colloidal solutions of zirconcium in AERE in-reactor autoclave tests 52). The AERE observation could have resulted from spalling of the oxide on a fine scale, although the fact that the zirconium in the water was apparently not filterable makes this explanation improbable. At present, therefore, although it appears that there may be conditions under which formation of true, or colloidal, solutions of zirconium can take place under irradiation, dissolution of the oxide film is not regarded as a significant contributor to the observed weight changes in irradiated tests.
5.3.
affect
METAL
ORNL eventually discarded the idea of enhanced dissolution of the oxide in favour of a mechanism in which radiation damage in the metal was the predominant process 18~i4). The oxide was thought to be lost by a mechanical
5.3.1.
Enhanced diffusion of oxygen in zirconium
Experiments at ORNL, aimed at providing support for a “metal damage” hypothesis, have apparently produced evidence that the oxygen content of the surface layers of the metal may have a critical effect on the phenomenon 89). It has been proposed that enhanced diffusion of oxygen in the metal is the explanation of the acceleration of the oxidation rate under irradiation, although it was not clear how this hypothesis actually explained the practical observations. The suggestion that irradiation might cause the precipitation of oxide particles, which could lead to accelerated oxidation, is implausible. Irradiation does not normally achieve chemical reactions which were not thermodynamically possible in its absence, and oxygen in solution in zirconium is more stable thermodynamically than ZrOa. Metallography of irradiated specimens has not provided any evidence which would support such a supposition, and the experiments on which it is based can be criticised for lack of adequate controls 89). The complete absence of an oxide film from zirconium during a prolonged irradiation is exceedingly unlikely, irrespective of the method of preparation and irradiation. It seems likely that the effects being observed on nominally “oxide free” specimens were due to contamination of the specimens (e.g. with carbon or metal atoms sputtered during irradiation), whose surfaces were not reprepared after irradiation. This one omission, to reprepare the surface of irradiated and unirradiated controls side by side immediately prior to oxidation, has largely vitiated the results of this experiment. Irradiation of anodic oxide films with 65 keV neon
ions,
and
subsequent
oxidation
90) ap-
EFFECTS
OF
IRRADIATION
ON
THE
parently supported a view that damage to the metal was the important factor. The absence of i~adiated oxide was not demonstrated in any of these experiments, and hence the results cannot be uniquely attributed to metal damage. Insufficient control experiments (e.g. varying bombardment energy) were performed to demonstrate whether these results were or were not applicable to the case of neutron and fission fragment irradiation.
OXIDATION
ZIRCONIUM
ALLOYS
31
in the metal, the recent evidence that water chemistry and preoxidation treatments play a major part in the effect makes it much less likely that radiation damage in the metal can be an important parameter. In fact if anything the recent evidence suggests that a region at or near the oxide/environment interface, and not the oxide as a whole, is critical in determining the behaviour.
5.4.
Enhanced in-reactor creep of zirconium alloys has been observed 91). Since extension of the metal core (either by applied stress during oxidation, or by stresses generated by the oxidation of thin specimens) might lead to cracking of the oxide film and hence to accelerated oxidation this is a possible contributor to the observed phenomena. No oxidation rates on PWR fuel cladding, or pressure tubes, have been observed, however, which are significantly higher than those observed on unstressed corrosion coupons (e.g. fig. lo), so that there does not seem to be any experimental support for the importance of this effect. It seems unlikely that a small rate of growth of the metal should affect the growth of an oxide which, under normal circumstances, accommodates a 50 */* volume increasing during oxidation. The oxidation of thin specimens, which creep during oxidation under the influence of the stresses generated by the oxidation, is not significantly different from that of thicker specimens 92), and oxidation of specimens stressed beyond the yield point has shown no change prior to transition and only small increases in oxidation after transition $3~94). Thus it seems unlikely that this effect could be participating in the enhanced oxidation of Zircaloy-2 in-reactor for normally stressed specimens, and even for highly stressed material it could only account for a fraction of the observed increase in oxidation. In general, although during the early work there was no reason why damage to the oxide should be a preferable explanation to damage
OB
FISSILE
MATERIAL
The Bettis suggestion that instances of enhanced oxidation during radiation in water were due to contamination with fissile materiala) is hardly an explanation. It merely seeks to relate the phenomenon in water to another unexplained (but well established) phenomenon, enhancement of oxidation in fissioning many1 sulphate solutions. A broader view, which accepts evidence not obtained in high temperature water, shows that in many instances of enhanced oxidation under i~a~ation [AERE 2%2%25*26~ 52) and CISE 49-51) work], there could not conceivably have been contamination by fissile material, since fuel materials had never been in the equipment at any time. Thus as a general correlation of the phenomena this must fail. However, there may be far more similarity between the mechanisms in both water and fissile solutions than has been assumed in the past. Thus in both environments the phenomenon is affected by the presence or absence of oxygen 7*15~37). ~ilst undoub~~y fission fragment irradiation is a far more efficient means of damaging the oxide than fast neutrons, it may be that the vastly enhanced radiolytic decomposition of water in the presence of fission fragments is at least as important. Thus a relatively small amount of contamination by fissile material which might be insufficient to cause any major change in the damage rate in the oxide film, oould easily affect the local chemistry of the water by increasing the production rate of radiolytic species.
32
B. COX
5.5.
INDUCED UNDER
If
the
DIIwusfoN
IN THE
OXIDE
IRRADIATION
irradiation
of
the
oxide
produces
vacancies and interstitials, which are able to transport reactive species through the film, and the dose rate is sufficiently
high that the con-
centration of defects is increased, and if transport of such defects is the rate controlling process, then an acceleration
in the oxidation
process should be observed without any change in the kinetics law which is obeyed 95). However, it was realised from an early stage in these studies that if reasonable values for the rate of production and annealing of these defects were used in a calculation. then the effect should be below observable limits at the fast neutron damage rates available, and inadequate to explain the very high oxidation rates in fissile solutions. To overcome this problem it was suggested that permanent regions of damage must be produced which were large enough to act as easy paths for diffusion through the oxide IQ). In the homogeneous reactor case it was felt that fission fragment tracks in the oxide might fulfil this purpose, although there is no particular reason why any other radiation induced change in the oxide leading to enhanced diffusion (e.g. nucleation of smaller oxide crystallites under irradiation) could not explain the phenomenon
conductivity in thin ZrOQ films could lead to increased oxidation of Zircaloy-2 in a y-flux Q5), and the apparent observation of increased conductivity of ZrOQ in some in-reactor experiments IQQ),led Wanklyn et al. to an extensive study of the electrical properties of zirconia films under irradiation 96, 97, iQi-103). The results were based on the a.c. conductivity
measured
mainly on anodic oxide films at ambient temperat~e, and to that extent their applicability to oxidation processes (de conductivity) at high temperature may be in doubt. Some allowance for these differences can be made, on the assumption (doubtful) that we know the electron transport process in thin oxide films. The irradiation effects observed were so small, however, that it seems unlikely that the large difference between the experimental conditions and normal oxidation would completely alter the conclusions. The overall conclusion of these experiments was that no observable change in the conductivity of zirconia films is produced by a gamma flux and only small increases in conductivity resulted from fast neutron doses > lOQQnf/cmQ~sec. Neither fluxes nor doses, which have been observed to affect the oxidation rate, showed any effect on conductivity outside measurement errors. Theoretical calculations of the effect of y-irradiation tended to support the ~sumption that no effect would be expected ; however, some of the parameters involved in the theoretical treatment were not
equally well. However, the assumption that lattice diffusion of vacancies is the controlling process during oxidation in the absence of radiation has led investigators to reject an explanation of the irradiation enhanced oxidation on the basis of irradiation effects on the ionic transport process QQ,Q’), on the ground that production of vacancies and interstitials was inadequate, and that irradiation produced no increase in the mobility of these defects 97). This criticism is inapplicable if the ionic transport process is not controlled by a lattice effusion process in the oxide (as appears to be the case) Q*pQQ).
sufficiently well known for this to be said with certainty. Thus it would appear that any effect of gamma radiation on the oxidation process cannot be due to a direct effect on the conductivity of the oxide, and other explanations are needed for the Hanford observation of enhanced conductivity. A change in the environmental chemistry, under gamma irradiation, which could affect the growth of thin oxide films, is not ruled out.
0.0.
5.7.
The
ENHANCED
suggestion
ELECTRICAL
that
CONDUCTIVITY
enhanced
electrical
EMBRITTLEMENT
Having
eliminated
OF THE
electrical
OXIDE
effects
FILM
cxperi-
EFFECTS
OF IRRADIATION
ON THE OXIDATION
ALLOYS
33
of the energy absorbed
by the
OF ZIRCONIUM
and changes in ionic transport mentally, (implausible; see above), Wanklyn et al. have suggested that embrittlement of the oxide leading to earlier cracking of the film and acceleration of the subsequent oxidation rate could be an explanation of the irradiation
major
enhanced
194). Results
peroxide
(the
and thermal
produced
by the y-irradiation
oxidation
of Zircaloy-2
of bending
oxidation
phenomenon
pre-irradiated
anodic
films at room temperature round a mandrel have given results supporting this hypothesis. Such an hypothesis can operate only in the post-transition region, or could explain the reduction in the weight gain at transition. There appear to be ample experimental observations (e.g. at AERE, CISE, Hanford and CRNL) of enhanced oxidation by a parabolic or cubic rate law prior to transition (figs. 7, 15, 16, 18) to demonstrate that this cannot be the whole explanation. Thus embrittlement of the oxide may be a contributor to the later stages of the irradiation enhanced oxidation, but cannot be the primary cause of the phenomenon. The conditions of these experiments were appreciably different from normal oxidation conditions. It is known that cracking due to thermal stresses can occur in oxide films on cooling to room temperature 195), and bending tests performed under these conditions do not necessarily lead to valid conclusions about behaviour at the oxidation temperature. The cracks observed in the optical microscope 194) are probably not the important ones causing the normal rate transition 195).The important cracks or pores are much smaller (w 50 d dia.) than these and are difficult to observe even by electron microscope techniques. 5.8.
DIRECT CHEMICALEFFECTS
The idea that aggressive radiolytic species in the solution could be attacking the oxide film on Zircaloy-2 directly, without any direct contribution from radiation damage to the process, gained some credence when the effects of water chemistry first became well established. However, such an effect would be equally prominent in a gamma flux alone, since the
fraction
water arises from this source. Gamma flux effects are at most of only minor importance in the thin film region, and are completely absent when the oxide film becomes thicker. An examination of the effect of hydrogen most
likely
long-lived
species
of water) on the
showed no effect 99).
The insensitivity of the oxidation rate to environment chemistry in the laboratory 199) and the failure to obtain evidence that would substantiate any increase in oxidation rate downstream of the irradiated test section 49-51) argue further against any direct chemical effect. Only in the very early stages of oxidation in water and steam is there evidence that the chemical nature of the environment is important 197). Th ese effects influence the weight gain during pre-autoclaving treatments but disappear subsequently. It is in this region of weight gain (O-15 mg/dm9) that effects attributed to the gamma flux have been observed, and only in high pressure steam. Therefore any effect of gamma flux may be purely one of environment chemistry which would be most important under the observed conditions. The density of radiolytic species would be higher in water or high pressure steam than in low pressure atmospheres, and this might explain the failure to detect effects in tests at 1 atm. For Zr-Nb alloys the picture is different; these alloys are affected by the presence of oxygen-containing species in the water during laboratory tests 199), and hence the production of such species, even if short lived, in reactor can result in a direct chemical effect 59) both in flux and downstream of the irradiated region. The magnitude of these effects is very dependent on alloy composition and heat treatment. 5.9.
ENERCJY TRANSFER PROCESSES
Radiation chemists have recently become aware of the major part which energy transfer processes can play in radiolysis 199). Thus, in the presence of a suitable surface, energy absorbed in the solid phase can be transferred
34
B.
COX
to adsorbed molecules to cause their radiolysis.
to be no thicker
In this way radiolysis rates may be obtained which are far higher than those which would
based on the calculated effect of heat flux across the oxide film has been applied in
result
from
direct
interaction
between
the
radiation and the liquid or gas whose radiolysis is being studied. In effect the surface is acting as a catalyst for the radiolysis in a manner analogous
to that which
a chemical
catalyst
plays in a chemical reaction. Such a phenomenon might explain the necessity for both irradiation and suitable chemistry to be present simultaneously, since the active species which subsequently cause the enhanced oxidation are produced in large quantities on the surface sites of oxide where the probability of reaction is greatest. While no actual measurements have been made which would lend support to this hypothesis, the observation of accelerated oxidation under conditions where no oxygen or oxidising radiolytic species could be detected at the outlet to the loop test section, suggests that this type of process may be playing a major part in the overall phenomenon, and that radiolysis of adsorbed species rather than isotropic radiolysis within the environment may be the important chemical factor. which eliminated direct The arguments chemical reaction with radiolytic species as a contributor to other than the growth of thin
obtaining
than
the expected
expected 35) a factor
film thickness.
Experi-
mental verification of a heat flux effect in the absence of irradiation, which depends on certain assumptions
about
the oxidation
mechanism
which may not apply, is absent ilO--112).Measurements with electrically heated specimens 111) could be affected by the presence of even small ac potentials across the oxide. A rise in the surface temperature of fuel due to the deposition of crud on surfaces under heat flux might have contributed to the increased oxide thicknesses observed on BWR fuel cladding. However, since oxidation rates of fuel cladding are comparable with those on the isothermal G-7 loop specimens (fig. 6) this may be a minor contribution, and cannot be a general explanation of irradiation effects even if its contribution to the oxidation of BWR cladding were significant.
5.11.
QUALITY OF MATERIAL
Examination of fuel cladding from the VBWR and Dresden reactors showed that, in addition to the accelerated uniform oxidation, the Zircaloy-2 formed a large number of pustules of thicker oxide 299so). In measuring
films (lack of any effect under gamma radiation) would seem to rule this explanation out as the sole process responsible; some form of direct
the film thickness on fuel elements these areas of irregular oxide were ignored, and only uniform oxide was measured 39). Examination
damage process in the necessary as well.
of recent fuel from Dresden has shown no pustules and lower oxide thicknesses at equivalent times 31) than for the earlier cladding. It has been suggested that this represents the result of improvements in the quality of the Zircaloy-2. Whilst experiments elsewhere 46) have shown some variation from batch to batch of Zircaloy-2 the range of behaviour has been less than a factor of two. Thus the large difference observed with Dresden fuel may arise from some other change made in the operational details (e.g. change in preautoclaving). Insufficient details are available to judge this possibility.
5.10.
oxide
seems
to
be
EFFECT OF HEAT FLUX
Many of the examples of thicker than expected oxide films have been observed on fuel cladding from Boiling Water Reactors. The contribution to this thick film resulting from the rise in temperature at the metal/oxide interface due to the high heat fluxes prevailing across the oxide film has been discussed 33-35). In examinations of fuel cladding from Pressurised Water Reactors where the oxide film is stated
EFFECTS
6.
Oxidation
OF
IRRADIATION
ON
TEE
mechanisms
OXIDATION
OF
ZIRCONIUM
ALLOYS
35
alloy, and in the influence of grosser regions of disorder introduced
by the presence
of second
Before attempting an explanation of the phenomena observed under irradiation, it is essential to understand the oxidation process in the absence of irradiation. The present
phase particles in many of the common alloys. However, it would be unwise to assume that it can be inferred from the kinetic curves that
situation on the oxidation mechanism and effects of irradiation on transport processes as seen by the author are summarised here.
ionic transport is the rate controlling process ; electrons must also be transported through the oxide film in the reverse direction. The transport of electrons through insulators such as zirconia
6.1.
The commonly accepted mechanism for the pre-transition oxidation of all zirconium alloys has been that it was controlled by oxygen diffusion inwards via anion vacancies 11s). It has been suggested that the approximately cubic kinetics represented only a minor deviation from the norm and could still be explained on the basis of the Wagner-Hauffe theory 114). The fact that in the early stages the rate laws were often far from good fits to any readily analysed process has been dismissed also as deviation from the norm 115). Attempts to fit the pretransition oxidation kinetics to a sequence of parabolic and cubic periods of oxidation do not assist in explaining what is happening in the absence of studies of the individual transport processes llQ9l17). However, the oxide film formed on zirconium and its alloys consists of discrete crystallites (50-250 A dia.) which grow slowly in size as
has not been extensively studied. However, investigations of electron transport through glasses, and the effect of devitrification iQQ), show that whereas crystallite boundaries and amorphous material may represent easy paths for ionic transport, they may represent barriers to electron diffusion. The easy paths for electron transport in this case would be through the crystallites. Thus growth of oxide crystallitea in the oxide film, with increasing thickness, which would represent a decrease in the area of material available for easy transport of ions, could imply an increase in the ease of electron transport. Experiments show that electron transport through oxide films on zirconium alloys may be mainly at flaws, or the sites of intermetallic particles lQ1), and is probably not diffusion controlled, even for thick oxide films QQ). Measurements of the potential across the oxide film formed on zirconium and its alloys
the film thickens, and which may be embedded in a matrix of amorphous material (comprising a significant fraction of the bulk) in the thin
in a fused salt bath provide a clue to which of the two processes is the more difficult QQ). Thus, crystal bar zirconium oxidising at a
film region 1139119). Attempts to measure the lattice diffusion coefficient of oxygen in zirconia films have suggested that at the temperatures normally used in the practical applications of zirconium alloys the actual diffusion coefficient may be as much as five orders of magnitude higher than the lattice diffusion coefficient 98). Thus the conclusion that oxygen ion transport at = 300 “C is almost exclusively via crystallite boundaries or other easy paths seems inescapable. The primary differences between the ionic transport process in the oxides of zirconium and its alloys may be in the way the crystallite morphology develops from alloy to
temperature above 250 “C rapidly develops a potential of - 1.1 to - 1.4 V, which falls slowly throughout subsequent oxidation. Although the anodic and cathodic polarisation curves are not well established, the high negative potential observed indicates that the transport of electrons through the oxide film is the more difficult process, and that it may be occurring by a Schottky emission process over barriers associated with flaws in the oxide. The potential on an alloy such as Zircaloy-2 oxidising at 300 “C passes through a maximum of about - 1 V and then falls rapidly, as the oxide film thickens, to a fairly stable value of
IN
THE
ABSENCE
OF
IRRADIATION
36
B. COX
about -0.25
V. This suggests that the electron
transport process becomes relatively easier as the oxide thickens. It is impossible to say, at present, which is the most difficult process under conditions where the potential is M 0.25 V (metal negative) ; but since the ionic and electronic currents must be equal during oxidation the two processes may be equally important in controlling the oxidation rate. Transient changes in the conditions of oxidation produce changes in the balance between the ionic and electronic transport processes. Thus initial oxidation at R+ 400 “C, and transfer to m 300 “C, results in small positive potentials on the metal which persist for a long period (weeks) without annealing out. The presence of a small positive potential shows that the balance between electronic and ionic transport is shifted even further in the direction of easy electron flow. Since the oxidation rate (ionic current) is little affected by the pretreatment at m 400 “C (the memory effects were very small under the conditions used) this situation must have arisen because of the increasing ease of the electronic process, rather than a decrease in the ease of ionic flow (which would have resulted in a lower oxidation rate) 99). Similar experiments on other zirconium alloys have shown wide variations in behaviour, which may help to explain the different behaviour of these alloys under irradiation. Investigation of the formation of pores and cracks at the transition point has shown that the important defects in the oxide are a network of fine pores < 100 A in diameter 105). These have not been identified by electron microscopy but may be associated with the growth of the oxide crystallites to a certain critical size. The occurrence of larger cracks in the oxide at the oxidation temperature and the importance of the stresses produced by oxidation in causing these (and the small pores) has not been satisfactorily studied. 6.2.
EFFECT
OF IRRADIATION
ELECTRONIC
Wanklyn’s
ON IONIC AND
TRANSPORT
work
suggests
that
electronic
transport
through
oxide
films
on
zirconium
alloys is likely to be little affected by irradiation with fast neutrons, gamma rays or other species (section 5.6). Since for Zircaloy-2 the more difficult process, other than in the thin film oxidation region, may be ionic transport, it must be an effect on this process which leads to the observed acceleration of the pre-transition oxidation rate 25945,48-50). We have seen (section 5.5) that such an acceleration cannot be accounted for on the basis of the production of additional vacancies, nor would our interpretation of the out-reactor oxidation mechanism suggest that production of additional vacancies would be important even if it were large enough to affect the normal vacancy concentration. Permanent changes in the oxide film under irradiation are needed to change the oxidation rate. These could be in the nature of additional easy paths, represented, for example, by fission fragment tracks is), or the tracks of primary knock-on particles if these are of sutllcient size (i.e. they should span a significant fraction of the film thickness). While this gives us an argument that would support the importance of fast rather than thermal neutrons (in the absence of fission fragments), since only neutrons with energies greater than about 1 MeV would produce primary knock-ons with a range in ZrOz comparable to the film thickness, it seems unlikely that the density of such tracks would approach that of the crystallite boundaries when crystallite size is only 50-250 A. Perhaps a more likely manner in which irradiation could be affecting the ionic transport process is by increasing the nucleation frequency or decreasing the growth rate of ZrOa crystallites, thus leading to a film with a smaller average crystallite size. The only known investigation by transmission electron microscopy 6% 65) of oxide films formed in-reactor did not show any significant difference from films formed in the laboratory. However, the subtle changes involved in this hypothesis might well have escaped the rather cursory examination given to these films. A report that
EFFECTS
the crystallite
OF
IRRADIATION
ON
THE
size in the oxide film formed in
OXIDATION
appear
OF
ZIRCONIUM
that the electron
37
ALLOYS
transport
process
is
a fused nitrate bath at 400 “C was much smaller than for a similar film formed in air iss), would support the contention that such effects may be very important, since it is above about
the one most affected by water chemistry; again this seems a reasonable conclusion since laboratory data suggest that the most important region in controlling electron flow is a thin
400 “C that the oxidation
barrier, which may be at the oxide/environment
rate in a fused nitrate larger than that
interface QQ) ,
in air. For interpretive purposes, however, it may be the balance between the ionic and electronic transport processes which is important. Thus, if one of the two processes is being altered by irradiation more than the other, the oxidation rate will only be changed if the process most affected by irradiation (in this instance the ionic transport) is the more difficult process. The magnitude of the acceleration under irradiation will be limited to the point at which
Adsorption
medium becomes
significantly
control of the oxidation process swings over to the least affected process (electron transport). Thus the magnitude of the expected increase in oxidation rate under irradiation will be dictated by the relative difficulty of the ionic and electronic processes prior to irra~ation. So far, however, we have not introduced any factor that could be associated with water chemistry. 6.3.
EFFECT
OF ENVIRONMENT
OXIDATION
PROCESSES
CHEMISTRY
ON
easily
affect
of species at this interface electron
and hole injection
could into
this layer of the oxide. In this respect the work of Leach 123) which showed large changes in oxide conductivity during cathodic polarisation may be impo~ant. Although the inclusion of protons in the oxide, as required by Leach’s proposed mechanism, has never been demonstrated, these may be acting only at the sites of easy electron flow. 7.
Hypothetical explanation effects on oxidation
of radiation
We may have available, therefore, a basis for interpreting the irradiation effects. The supporting experimental data from laboratory tests or direct meas~ements under i~adiation is, in most instances, inadequate, however, and we are therefore drawing inferences on tenuous grounds. Nevertheless, in the interests of rationalising the in-reactor observations, and demonstrating the importance of certain types
of experimental data, the attempt seems to We have seen (section 5.8) that in the absence be worthwhile, even if it merely serves to of irradiation, environment chemistry has little stimulate experimental work to disprove it. To or no effect on the pre-transition oxidation rate this end we will treat the observations on the of Zircaloy-2, and that in this region potential alloys individually. measurements suggest that ionic transport is the more diflieult process (or at most that ionic 7.1. ZIRCALOY-2 AND ZIRCALOY-4 and electronic transport are about equally difficult). We can conclude, therefore, that the A summary of the important experimental ionic transport process is little affected by observations first seems to be in order: environment chemistry, as would appear to be (4 Acceleration of the oxidation rate of the reasonable if it is controlled by transport along Zircaloys is always observed in BWR crystallite boundaries in the oxide. conditions, or in pressurised water conEnvironment chemistry does have an effect taining oxygen. on oxidation rates in the thin film region (b) Little or no acceleration is observed in (O-7 mg/dmz) of Zircaloy-2 oxidation I*?), where PWR conditions for the Zircaloys in the potential measurements show electron transport pickled state, or when pretreated at 300 “C. to be the most difficult process. Thus, it would (4 When pretreated at 400 “C acceleration of
38
B.
the
oxidation
rate
of
the
Zircaloys
is
observed even in the absence of measurable oxygen contents, although the magnitude
(4
cox Starting with the premise that the conclusions drawn from potential measurements in fused salts in the laboratory may be applied to
of the increase is less than in the presence
aqueous oxidation,
of oxygen. Both pre-
specimen, over the region of film thickness 0.5-Z ,um, the ionic and electronic transport
and
post-transition
rates are accelerated,
oxidation
with a reduction
then, for a normal pickled
in
processes are about equally difficult (fig. 32a) 90).
the time to transition and the weight gain at transition.
Then if only the ionic transport process is speeded up under irradiation there will be little
The effect saturates at high neutron fluxes. All effects disappear at or above 400 “C. Appreciable differences in the magnitude of the effects are apparent from batch to batch of Zircaloy-2 and Zircaloy-4.
or no observable effect on the oxidation rate, but the control of the process will now be due to electron transport and the potential across the oxide should become more negative (fig. 32b). If there is no measurable concentration
0) HYPOTHETICAL
POLARIZATION AND
CURVES
FOR PICKLED
( SUB P)
AIJTOCLAVED
(SUE A 1 SPECIMENS
b) EFFECT
OF IRRADIATION
PWR CONDITIONS
PREIN LAB
CURVE
UNDER
; ONLY ANODIC
AFFECTED
PICKLED SPECIMENS
Eox /(
v
v
METAL POSITIVE
METAL POSITIVE
1’
PREAUTOCLAVED SPECIMENS
WITH OXYGEN I c) EFFECT
OF IRRADIATION
CONDITIONS
UNDER OXIDIZING
; BOTH CURVES
AFFECTED
d) SPECIMENS IN
(~J,WITH
Fig. 32.
Hypothetical
effecte of irradiation
PREAUTOCLAVED
RELATIVELY SMALL
AT 4033%
LOW OXYGEN WATER NH,
ADDITIONS)
on electronic and ionic transport
processes.
EFFECTS
OF
IRRADIATION
ON
THE
of oxygen or oxidising radicals in the reactor (i.e. PWR con~tions) which might affect the electron transport process then the net result would be little effect of irradiation on the oxidation rate. However, if in addition to neutron irradiation sufficient to swing the control over to the electronic process there is oxygen in the coolant (e.g. BWR’s or G-7 loop, 1 ppm Os), or a high rate of production of oxidising species (BWR conditions, inadequately doped with ammonia) then the oxidation rate will be increased by whatever factor the electron current is increased by the presence of an “oxi~sing” environment (fig. 320). This factor will always be limited by the possibility of a return to ionic control if the increase in the electronic conductivity (due to oxygen) is greater than the increase in the ionic conductivity (due to fast neutron irradiation). Thus the highest irradiation enhancements will be observed under conditions of both high oxygen content and high fast neutron flux (G-7 loop, or BWR with untreated water). However, it would not be expected that the oxidation rate under irradiation should increase indefinitely with increasing oxygen content or neutron flux ; both effects would be expected to ssturate, although it is not possible from present data to estimate any likely point where this saturation would occur. It is also clear that neither a high fast neutron flux in the presence of a minor amount of oxygen, nor a large amount of oxygen in a relatively low fast neutron flux, would lead to very large accelerations in the oxidation rate. In the absence of any numerical information on the sensitivity of the anodic and cathodic processes to fast neutrons and oxygen respectively it is not possible to predict quantitatively what the relative effects of the two contributors would be. Fission fragment bombardment will affect both the anodic and cathodic processes simultaneously via its damaging effect on the oxide and the very high rates of radiolytic decomposition of water it produces. Thus the presence
OXIDATION
OF
ZIRCONIUM
ALLOYS
39
of fissile material in the water should be the most effective way of producing conditions suitable for accelerated oxidation. Fission fragment i~adiation would also be expected to lead to the largest enhancement factors (for similar reasons), and in practice the largest observed acceleration of zirconium alloy oxidation rates was in homogeneous reactor solutions. Although the above arguments apply predominantly to pretransition oxidation processes, no major changes in the potential across the oxide film have been observed at the transition point Sa)* Thus, while embrittlement of the oxide by irradiation may result in a reduction in time to transition and weight gain at transition, it would appear that the same arguments as above would apply to the post-transition oxidation rate as well as to the pre-transition rate. The effect of specimen pretreatment can be interpreted from the observed effect of pretreatment on the potential developed across the oxide film during oxidation in fused salts. l?or pretreated specimens the electronic processes have already been made relatively easier by the pretreatment. Thus, even in the absence of oxidising species an acceleration of the oxidation rate would be expected (fig. 32d) ; however, it would probably not be very large without the presence of oxidising species to increase further the electronic conductivity. At high oxygen levels the large effect on the electronic conductivity would reduce the effect of pretreatment to a small pert~bation of the oxidation rate. ~etreatments at lower tem~rat~es than PO0“C should result in a lesser effect, but there is insufficient laboratory evidence on which to draw a positive conclusion. The arguments in the above paragraph will apply only to relatively thin films. If either the preformed oxide is very thick, or the film is allowed to become very thick in-reactor, then the oxidation rate becomes virtually independent of water chemistry, and is equivalent to that in high oxygen water (fig. 20). This probably arises because formation of radiolytic
40
B. COX
oxygen
within the thick porous oxide is inde-
pendent of the external water chemistry. To explain the disappearance of the irradi-
these batches under the worst conditions of flux and oxygen, since here the effect of oxygen on the electronic conductivity could swamp
ation effects at or above 400 “C requires the acceptance of two additional pieces of infor-
the other differences ; this does not appear to be borne out in practice.
mation. Firstly, that, even for the Zircaloys with their intermetallic particles available to
We have thus been able to explain, in general, the features of the irradiation enhanced oxidation of the Zircaloys. If a more detailed
give
paths
of easy
electron
flow,
the
more
difficult process above 400 “C is always electron transport. Potentials measured above 400 “C are usually more negative than at lower temperatures QQ). Secondly, that above 400 “C the electron transport process becomes insensitive to environmental effects. This latter is borne out by the lack of environmental effect for oxides thicker than m I. ,um in laboratory tests in 400 “C steam. This may result from a more rapid oxidation of the intermetallic particles at temperatures above 400 “C than below this temperature, which could lead to the electron current being controlled by diffusion rather than by emission over a barrier. Intermetallic particles in Zr-Al alloys have been observed to remain virtually unoxidised at 300 “C, whereas they oxidise progressively more rapidly (relative to the thickening of the whole oxide) with increasing temperature 124). If these two premises are accepted then no acceleration due to irradiation would be expected at the higher temperatures under any conditions of dose rate or environmental chemistry. The differences from batch to batch Zircaloy-2 and Zircaloy-4 may be related
of to
an effect of the distribution of the intermetallic particles on the electronic conductivity. This distribution appears to have little effect on the ionic conductivity since the pre-transition oxidation rate in the absence of irradiation varies little from batch to batch 125).Thus a variation in the balance between the anodic and cathodic processes in the absence of irradiation could be present from batch to batch. This would be reflected in a difference in their behaviour, especially in “oxygen-free” conditions for pretreated specimens, where these differences would be most signi~oant. It might be expected that little difference would be apparent between
quantitative explanation is to be made, then further experimental measurements must be made of the effect of the reactor environment on the controlling processes. Especially needed are measurements in-reactor of the potential developed across the growing oxide film which should demonstrate the swing from one controlling process to another as the conditions are changed. 7.2.
Zr-2.5 wt % Nb ALLOY
A similar summary of the important inreactor observations for these alloys shows the following : (a) Oxidation rate enhanced both in-flux and downstream in the presence of “oxidising” coolants. (b) Little difference between in-flux and downstream weight gains. Evidence generally shows slightly lower in-flux weight gains. (c) Little apparent effect of pretreatment. (d) Little evidence for flux dependence of oxidation rate, but oxidation rate apparently saturates at high oxygen levels. (e) Large effect of specimen heat-treatment. (f) Effect of oxygen does not disappear at 400 “C. Potential measurements on Zr-2.5 wt y0 Nb alloy specimens have generally shown negative values between 0.5 and 0.9 V. The highest potentials were for material quenched from the p or (a+@) phases, with a steady reduction in the potential as a result of annealing at 500 “C. However, it appears that, qualitatively, even after long annealing treatments the electron transport remains in control of the overall oxidation rate QQ). This observation, plus the fact that the oxidation of this alloy is sensitive to the presence
EFFECTS of oxygen
OF IRRADIATION
or oxidising
ment in the absence
ON THE OXIDATION
species in the environof radiation,
auf&es
to
those has
OF ZIRCONIUM
of the oxidation suggested
that
ALLOYS
process.
there
are
41
Recent
work
at least
two
explain most of the observed in-reactor behaviour. Thus no direct acceleration of the oxidation
processes normally operating in parallel lz*). 1. Direct access of molecular hydrogen via flaws in the oxide film. Possibly the same
rate by fast neutrons in the absence of oxygen would be expected. If oxidising species are ,produced in the reactor then both in-flux and
2. Diffusion of hydrogen (protons) through the bulk of the oxide film.
out-of-flux
oxidation
rates should
be affected
to about the same degree. The observation of slightly lower weight gains in-flux compared to those out-of-flux is difficult to explain on this basis, however. If control of the electron transport process lies near the outer surface of the oxide, then it is possible that a direct effect of radiation on this surface, or an energy transfer process leading to a specific adsorbed species, might limit injection into the conduction band. This remains pure speculation, however, as no experimental evidence is available in support of such an explanation. However, if the electron transfer process were affected in this way then one might expect the oxidation rate of Zr-Nb to be PH dependent in water. This pH dependence would need careful separation from the specific effect of oxygen and oxidising ions (e.g. nitrate) which is already well known. Other experiments which might lead to evidence on this point are a study of effect of y-radiation or U.V. illumination on the oxidation rate in air. Photoeffects remain virtually uninvestigated in zirconium or its alloys. 7.3.
OTHER ALLOYS
In general the background evidence available on the other alloys which have been tested is inadequate for any explanation of the phenomena. It seems reasonable to assume that similar experiments to those reported for Zircaloy-2 and Zr-2.5 wt y0 Nb would lead to a qualitative explanation of the behaviour of the Valloys and Zr-3 wt y0 Nb-1 wt y0 Sn. 8.
Radiation
effects on hydrogen
uptake
Investigations of the mechanism of hydrogen uptake have not been nearly so extensive as
sites at which the electronic
The
total
hydrogen
gaining
current flows.
access
to the
metal may be controlled by a barrier-type process at or near the oxide/metal interface (and possibly within the region of metal saturated with oxygen). The balance between the two transport processes through the film is probably set by the individual characteristics of the particular film and the environment outside it. Thus, the observation of the rapid uptake of small amounts of hydrogen, without a corresponding amount of oxidation, immediately after a specimen preautoclaved at 400 “C is transferred to < 300 “C water, may result from a breakdown of the initial oxide film leading to an acceleration of process (1) l@j). This would continue only until the defects had healed, and this healing might not reveal itself as accelerated oxidation over the same period. This could explain the observed deuterium uptake curves in NPD fuel for example 7% 73). In general it appears that the hydrogen uptake and oxidation processes are controlled by separate but parallel mechanisms. Thus, under irradiation it is possible for the two effects to be modified to different extents. It is known that the addition of relatively small quantities of oxygen to high temperature water or steam results in a significant reduction in the percentage hydrogen uptake 126). In the reactor environments containing much free oxygen (e.g. BWR’s), the reduction in percentage uptake resulting from the oxygen apparently almost balances the increase in oxidation rate. In other environments where the uptake behaviour is intermediate between BWR and PWR conditions (e.g. G-7 loop, AERE 1 atm steam loops) then there is apparently insu~cient oxygen to result in a reduction in the hydrogen uptake,
42
B. COX oxi-
of the oxide as well as just short-circuiting specimens. This type of experiment may not
It would appear that, if the environment cannot be maintained in a “reducing” condition
be carried out readily in the usual high resistivity water, where the oxide resistance is not large
although
the environment
dising to result in enhanced
equivalent
to normal
total hydrogen
PWR
is sufficiently oxidation,
operation,
and if
at the end of component
life
compared
with that of the water. The experi-
is the critical problem, then one might be better
ments would be easier in an electrolyte whose resistance was small compared with the oxide
operating with a large excess of oxygen, than with only a marginal amount. The oxidation
film resistance. Direct meas~ements
rate is not likely to be increased further under these conditions due to a saturation of the increase in oxidation rate with increasing oxygen content. Thus VBWR and Dresden fuel did not oxidise any faster than specimens in the G-7 loop (with oxygen) at equivalent fiuxes, but the total hydrogen uptake was much lower for BWR fuel than for specimens in the G-7 loop (with oxygen}. The relative efficiency of different enviro~ents for reducing hy~ogen uptake is difficult to ascertain from present evidence.
of the potential
across
the oxide film and the I-V characteristics of the oxide during oxidation in-reactor should prove to be interesting. Changes from more positive to more negative potentials on the metal should result from the application of a fast neutron flux, in the absence of “oxidising” conditions, and shifts in the polarisation curves such as those postulated in fig. 32 should be observed. The effect of increasing the neutron flux and additions of oxygen to the environment could also be investigated. A more detailed examination of oxide films stripped from specimens that have exhibited 9. Experimental testing of hypotheses enhanced oxidation rates would be interesting. Since detection of enhanced oxidation is not The value of any of the above suggestions readily achieved in the earliest stages of oxi(sections 6- 8) is only as great as their ability dation, this would require investigation of films to predict the behaviour in future experiments. much thicker (up to 1 pm) than those that The critical point in the explanation of can normally be viewed in an electron microenhanced oxidation is that only one (ionic scope. This should be possible with currently transport) of the two main processes is affected available 1 MeV microscopes. The in~resting by irradiation, and therefore control of the phenomena are likely to be differences in the oxidation rate can switch to the other (electron size of the small crystallites present, and other transport) process under irradiation. If the changes such as the fraction of the oxide volume oxide film were to be deliberately shortconsisting of resolvable crystallites. It would circuited during an irradiation experiment then be predicted that under i~adiation either the electron transport could never become the rate average crystallite size decreases and the volume controlling process and the oxidation rate of of the oxide in the form of resolvable crystallites all alloys should be enhanced by irradiation remains constant, or that volume of resolvable (even in strongly “reducing” conditions). Thus, crystallites diminishes with no change (or an under PWR conditions, short-circuited apeciincrease) in the average size. mens of both Zircaloy-2 and Zr-2.5 wt o/0 Nb Studies of irradiated post-transition oxides should show enhanced oxidation compared might concentrate on changes in pore structure with normal irradiated control specimens, and the extent of embrittlement and cracking although initial results on Zircaloy-2 in the with unirradiated laboratory do not confirm this expectation @9). of these films compared oxide films. In order to obtain a clear picture of the behaviour experiments on the pH dependence of the under irradiation, however, it would probably oxidation of Zr-Nb alloys, and y and photobe necessary to study the I-V characteristics
EFFECTS
OF
IRRADIATION
ON
TEE
effects on the same system have already been suggested (section 7.2). The e5Feot of reducing the oxygen content in the G-7 loop has been the only effect studied so far; it would be interesting to observe the effect of deliberately increasing the oxygen content. This would give information on the saturation level for the effect of oxygen on the oxidation rate, and some clue to the level of oxygen needed to reduce the percentage of hydrogen absorbed. 10.
OXIDATION
Acknowledgements The author would like to thank allthose who haye oont~buted to the preparation of this review by oritioising the manus~~pt, supplying copies of figures, and giving permission to use
ALLOYS
43
figures and to refer to work not otherwise available in unclassified form. Glossary AECL
Atomic Energy of Canada Limited, Chalk River Nuclear Laboratories (CRNL), Chalk River, Ontario, Canada (Research Reactors NRX, NRU, NPD *).
AERE
Atomic Energy Research Establishment, Harwell, Didcot, Berks., England (Research Reactors BEPO, DIDO, PLUTO).
Bettis
Westinghouse Electric Corporation, Bettis Atomic Power Laboratory, Pittsburgh, Pa., U.S.A.
CISE
Centro Informazioni Studi Esperienze, Segrate, Milan, Italy (Research Reactor Avogadro I at Saluggia).
EIR
Eidg. Institut fiir ~akto~o~oh~g, W~enlingen, Switzerland.
GKSS
Institut fiir Reaktorphysik der Gesellschaft fiir Kernenergiewertung in Schiffbau und Schiffahrt m..b.H., Reaktorstation, Geesthacht, W. Germany.
Hanford
Hanford Atomic Products Operation, Richland, Washington ; now Pacific Northwest Laboratories, Battelle North-West (B~L), (Reactor PRTR).
Idaho Falls
National Reactor Test Site, Idaho Falls, Idaho, U.S.A. (Research Reactors MTR, ETR).
KAPL
Knolls Atomic Power Laboratory, Schenectady, New York, U.S.A.
ORNL
Oak Ridge National Laboratory, Oak Ridge, Tennessee, U.S.A. (Research Reactor LITR, ORR).
Conclusion
It appears that by considering all the evidence currently available to us from in-reactor experiments, and laboratory studies of the oxidation mechanism, we are able to propose an hypothesis which will explain most of the phenomena on a qualitative basis. The value of this hypothesis will depend on its ability to predict the results of future experiments; several such predictions have been made, and it will be interesting to see whether they are borne out, if and when suitable ex~riments are performed. The addition to the body of data of the results of further purely kinetic experiments does not now seem to be warranted. The conditions under which the phenomena are observed, and their magnitude, now seem to be fairly well established. What are needed are carefully controlled experiments which include sufficient variables to permit some elucidation of the mechanism. Such experiments are difficult to carry out in loops devoted to the testing of protot~e reactor fuel in addition to corrosion experiments, since the operating conditions of the loops are generally set by the fuel development programme. There is a strong case for the provision of in-reactor loop facilities devoted solely to oxidation studies in which all the parameters can be controlled.
OS!'ZIRCONIUM
*
Owned and operated by Ontario Hydra.
44
B.
Saclay
Centre d’Etudes Nuclbaires de Saclay, Gif-sur-Yvette, Seine et Oise, France.
Vallecitos Vallecitos Atomic Laboratory, General Electric Co., Pleasanton, California, U.S.A. (Reactor VBWR).
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OXIDATION 68
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