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Grain boundary sinks in neutron-irradiated Zr and Zr-alloys
Journal of Nuclear Materials 159 (1988) 405-416 North-Holland, Amsterdam
405
GRAIN BOUNDARY SINKS IN NEUTRON-IRRADIATED Zr AND %-ALLOYS M. GRIFFITH& R.W. GILBERT and C.E. COLEMAN ~e~ailur~icuI Engineering Branch, Chalk Riuer Nuclear Laboratories, Chalk River, Ontario, Kill IJO Canada
Samples of annealed sponge and crystal-bar Zr and Zircaloy-2 have been examined following irradiation in EBR-II at temperatures - 700 K. Loop analysis shows that there is selective denuding of interstitial loops near to some grain boundaries indicating that such boundaries are net sinks for interstitial point defects. Furthermore, in sponge Zr and Zircaloy-2, vacancy c-component loops are observed running into the grain boundaries showing that the grain boundaries are not preferred sinks for vacancies. Cavities are observed in all samples. In crystal-bar Zr and sponge Zr they are mostly observed adjacent to grain boundaries. They are also sometimes found within grains associated with precipitates. The cavities are more common in the crystal-bar Zr and this is probably because both the sponge Zr and Zircaloy-2 contain vacancy c-component loops which compete for vacancies (assuming that the cavities are vacancy sinks). Only some of the grain boundaries have cavities adjacent to them and this may be related to the orientation of the boundary.
1. Introduction 1.1. The importance growth
of grain boundaries
in irradiation
Since irradiation growth in annealed polycrystalhne Zr is generally about ten times greater than for single crystal material [1,2] and increases with decreasing grain size [2-41, grain boundaries have been considered as important sinks for point defects, especially in annealed Zr and Zr-alloys for which the only other type of sink is (a)-type dislocation loops. In single crystal material the growth quickly saturates for fluences - 0.1 X 102’ n mV2 (E > 1 MeV). In annealed polycrystalline material growth almost saturates at low fluences but there is a smaller long term component, proportional to the fluence, which indicates the important of grain boundaries. The various irradiation growth data for annealed Zr and Zr-alloys are summarized in fig. 1. Carpenter et al. showed that single crystals exhibited irradiation growth indicating that there is some component of growth which is not related to the absorption of point defects at grain boundaries. In the absence of any ~crost~ctural data, this growth was attributed to the presence of sub-microscopic c-component vacancy loops or elastic relaxation around vacancy point defects and vacancy clusters or a non-Hookeian contraction around small (a)-type loops.
Murgatroyd and Rogerson [2] presented evidence that growth of Zr is sensitive to and inversely dependent on grain size at 553 K but is relatively independent of grain size at 353 K for grains of 5 gm and 40 pm diameter. The growth of polycrystalhne iodide Zr at 353 K was an order of magnitude higher than at 553 K for material with 40 pm diameter grains and about double for the material with 5 pm diameter grains; growth was still continuing in the latter material whereas it had virtually saturated at about 0.1 X 10” n rn-’ in the former. This temperature dependence was reversed for Zircaloy-2 and was non-existent for single crystal Zr. The results were explained by the formation of vacancy (u)-loops at the higher temperature and the fact that grain boundaries compete more effectively for vacancies when the grain size is small thereby reducing the vacancy (“)-type loop population compared with larger grained material. The lower growth of the Zircaloy at 353 K was explained in terms of the trapping of vacancies by Sn atoms. Fidleris et al. [3] have shown that growth is inversely dependent on grain size in annealed polycrystalline Zr for grains of 23 pm and 225 gm diameter irradiated at 330 IL There was a similar dependence for Zircaloy-2 with grains of 12 pm and 31 pm diameter irradiated at about 680 K but this was reversed for irradiation at 330 K. Cann et al. [4] also showed that there is a marked grain size effect on the initial growth transient at tem-
M. Griffiths et al. / Grain boundary sinks in neutron-IrradiatedZr and Zr-alloys
406
0 14
P g.s.(Zr and Zr-2)
0.12 0 10
REFERENCE
0 08 0.06 0.04 0.02
cc--
SINGLE CRYSTAL
(Zr)
i
ia1
I 10 0
5.0 FLUENCE (~t)/lOzs n.mm2
REFERENCE 014 -
0 2 pm g.s.iZrt
FLUENCE #V10z5
-23 ----
[ll
n.m-*
Fig. 1. Irradiationgrowth measuredalong the a-axis (singlecrystal) or longitudinalaxis (polycrystal) for annealedZr and Zircaloy-2 during neutron irradiation at (a) 330-355 K and (b) 550-590 K. For poly~~s~~~ne specimensthe fraction of basal poles in the longitudiialdirectionis = 0.1 (i.e. between0.07-0.16). Grain sizediameters(g.s.) are indicatedfor each plot.
peratures between 566 K and 621 K with the smallest grain size material exhibiting the highest growth strain; the slower growth strain rates after the initial transient were similar for the different grain size materials. They also showed that swelling of both annealed and coldworked material occurred at temperatures between 703 K and 778 K with the peak at 741 K, this could be attributed in part to the presence of cavities (believed to be He gas bubbles) in this temperature range. Recent observations [S] show that the cavities are distributed non-uniformly in these samples and are either observed
on precipitates or in clusters near to grain or sub-grain boundaries. Vacancy c-component loops are also often found in the vicinity of the cavities. 1.2. Grain boundaries in models for irradiation growth To account for the net a-axis expansion and c-axis contraction observed during irradiation growth, Kelly [6] originally proposed a mechanism in which vacancy and interstitial point defects migrate a~sotropi~ly to grain boundaries. This anisotropic flow develops in
M. Griffiths et al. / Grain boundary sinks in neutron-irradiated
response to the internal stress generated within grains because of anisotropic thermal expansion. The stress would not be maintained once the strain from the thermal expansion is relieved by the point defect flux to the grain boundaries and the result would be a saturation of strain with increasing fluence. The net strain on cooling would be positive along the u-axes and negative along the c-axis. Unfortunately this model did not account for the growth of single crystals or the slower long term component of growth for the polycrystalline material, unless this was a result of thermal cycling when the samples were removed from the reactor for each dimensional measurement. Carpenter and Northwood 171 proposed a simpler model based on the intrinsic bias of dislocations resulting in a net flux of interstitials to (a)-type dislocations (network or loops) leaving residual vacancies to annihilate at grain boundaries which they considered to be neutral sinks. Whereas this may be valid for cold-worked materials, containing dislocations with mostly (a)-type Burgers vectors which are considered as net interstitial sinks because of the size-effect interaction [8-lo], it cannot easily account for growth of annealed materials unless all or most of the dislocation loops generated during irradiation are interstitial in nature. This was the basis for the explanation of growth in single crystal Zr [l]; the c-axis contraction was then attributed to either relaxation around vacancies, the formation of very small c-component vacancy loops or a e-axis contraction around (a)-type dislocation loops. There are no rigorous dislocation loop analyses for material irradiated at temperatures < 580 K, i.e. at temperatures relevant to nuclear reactor applications; most dislocation loop analyses are for material irradiated at > 580 K [11,12]. Analysis of the loops in crystal bar Zr [13] and Zircaloy-2 [14] following irradiation at - 575 K have so far revealed that both vacancy and interstitial (o)-type loops are formed. However, there is insufficient evidence to estimate the relative numbers of each or whether there is selective denuding adjacent to grain boundaries. For annealed Zircaloys, as the fluence is increased to - 5 X 10” n rnF2, e-component vacancy loops having Burgers vectors of b = i (2023) begin to form [15]. This loop formation corresponds with a state of accelerated growth 1161and introduces a new factor into the treatment of growth of annealed materials. It also suggests that the treatment for cold-worked materials should include a contribution from the [c] or (c + a) network dislocations even though they have a different (larger) Burgers vector compared with the c-component loops formed during irradiation. Current models can account
407
Zr and Zr-alloys
for the observed irradiation growth by assuming either anisotropic interstitial diffusion [17] or a modified interstitial bias such that (a)-type dislocations have a larger bias parameter compared with c-component dislocations [lSJ, The result is a net flux of interstitials to (a)-type dislocations and a net flux of vacancies to c-component dislocations. Grain boundaries cease to be neutral sinks when point defect diffusion is anisotropic 1171 and the sink strength is then dependent on their orientation. Herring et al. [19] reported that the zone denuded of interstitial loops close to grain boundaries is greater than that for vacancy loops in electron irradiated Zr and Zircaloy-4 for temperatures - 675 K. This implies that the grain boundaries may be net sinks for interstitial point defects. It was also indicated that there may be a reIation between the magnitude of the denuded zone for interstitial loops and the orientation of the thin foil i.e. the effect is more pronounced for foil normals close to [OOOl]. This may be related to the grain boundary orientation but is equally likely to be related to the foil orientation; the foil surfaces being an altemative sink. Zones denuded of dislocation loops at grain boundaries have been observed in Zr and its alloys irradiated with neutrons at temperature > 670 K 19,201. In this paper we describe the defect structures observed near to grain boundaries in neutron irradiated crystalbar Zr, sponge Zr and Zircaloy-2.
2. Experlrnental Samples of crystal-bar, sponge Zr and Zircaloy-2 were irradiated in EBR-II at temperatures between 665 Table 1 Ingot analyses for crystal-bar Zr, sponge Zr and Zircaloy-2 major impurity or ailoying elements (ppm by wt)
Sn Fe Cr Ni B c N 0 H Hardness in BHN
Crystal-bar Zr
Sponge Zr
Zircaloy-2
14750 1750 1050 600 < 0.2 115 35 1200 12
57
143
174
408
M.
Grifjiths et al. / Grain boundav sinks in neutron-irradiated
and 710 K for fluences up to 1.5 X 1O26 n m-’ (E > 1 MeV). The ingot analyses for the major impurity or alloying elements in these samples are given in table 1. Thin foils for transmission electron microscopy were prepared by twin-jet electropolishing using a solution of 10% per&lo& acid in ethanol at temperatures < 243 K and were analyzed using a Philips EM3OOG electron
Fig. 2. Distribution of vacancy and intersuuiu IUu,.r -‘rm *n grain boundaries in sponge Zr irradiated at 710 K to a fluence of 4.5x1025 ” m-2 (E > 1 MeV). Loops exhibiting inside contrast with a z25 diffracting vector are vacancy in nature; beam direction _ [ii23].
Zr and Zr-alloys
microscope operating at 100 kV. Dislocation loop analyses were performed using the Maher and Eyre [21] and Foll and Wilkens [22] techniques.
3. Results Analysis of the (u)-type dislocation loops near to grain boundaries in sponge Zr irradiated at 710 K to a fluence of 4.5 X 102’ n rnp2 is illustrated in fig. 2. There is a demarcation (dotted lines) between regions containing vacancy loops only, i.e. adjacent to the grain boundaries, and regions containing a mixture of interstitial and vacancy loops, i.e. in the centre of the grain. A similar demarcation is observed in crystal bar Zr irradiated at 700 K to fluences of 1.5 x lot6n rne2. In both cases the mid-grain region contains approximately equal numbers of vacancy and interstitial loops and also a dislocation network which developed from loop growth; it is not possible to say which type of loops contributed to the network. In some small grains (about 2-3 pm diameter} the interstitial loop population was estimated at about 25%. There was no qualitative difference between the distribution of loops close to high angle grain boundaries and sub-grain boundaries (consisting mostly of rows of edge (u)-type dislocations with spacings of about 0.01-0.02 pm on {lOiO> and {1Oi 1f planes). This indicates that the long range stress fields of both types are similar and that any difference has an insignificant effect on the loop distribution. Another common feature of the irradiation damage in the crystal bar Zr close to grain boundaries is the existence of cavities (fig. 3(a)). Similar cavity clusters are also observed in sponge Zr (fig. 3(b)) but are less common. They are sometimes observed nucleated on or close to precipitates (fig. 3(c)). They are mostly found close to grain boundaries but are occasionally observed in mid-grain regions and are invariably associated with precipitates (fig. 3(d)). Analysis shows that the zones containing cavities adjacent to grain boundaries are denuded of interstitial loops as illustrated in fig. 4. This figure also indicates that there is an apparent orientation dependence for the width of the interstitial denuded zone, i.e. grain boundaries parallel with (1120) directions have a narrower denuded zone compared with boundaries which are not. If the existence of cavities is taken as an indication of a depletion of interstitial point defects, this effect appears to be most marked for grain boundaries parallel or perpendicular to (0001) which contains all three (1120) directions. There is little evidence cavities adjacent to boundaries having normals
M. Griffiths et al. / Grain bounahty sinks in neutron-irradiated
Zr and Zr-alloys
Fig. 3. Cavities in crystal-bar Zr and sponge Zr (B) irradiated at 700 K to a fluence of 1.5 X lox n mm2 (E > 1 MeV). (A) Cluster adjacent to a grain boundary, (B) cluster adjacent to a grain boundary in sponge Zr. Note faeetting on (OOOl),{ liO1)and (lOi0) planes, (C) cluster adjacent to a grain boundary and associated with precipitates,(D) cavity nucleated on or close to a precipitate.
(N) for which N * (11~0) = 0, as illustrated in fig. 5. Dislocation loop analysis indicates that grain boundaries parallel with (0001) are still preferential interstitial sinks although the width of the denuded zone for interstitial as well as vacancy defects is smaller than for other grain boundary orientations. The width of the zone denuded of vacancy loops appears to be smaller than that for interstitial loops irrespective of the orientation of the boundary.
Post-irradiation annealing at 700 K for 24 h resulted in little discernible change in the number and size distribution of cavities. In both the as-irradiated and annealed samples there was a bimodal distribution of diameters with a large group peaking at about 30 nm and a smaller group peaking at about 60 nm. Annealing at 825 K for 16 h resulted in an apparent reduction in the number of cavities, although this may simply be the result of a random sampling error. Those remaining
410
hf. Griffithr et al. / Grain ~~~
sinks in neutron-herniated Zr and Zr-ailoys
were Iarge with an average diameter of about 80 nm. One reason for there being fewer cavities in sponge Zr compared with crystal bar Zr may be because of the stability of basal plane c-component loops in the former material. These loops are invariably vacancy in nature [15] and are often observed running into grain boundaries (fig. 6). Whereas there is a clear indication of an orientation dependence for the width of the interstitial denuded zone in crystal bar Zr, there is insuf~cient data to discern a grain boundary orientation effect in sponge Zr. There is even an indication, from one observation of cavities adjacent to a basal plane grain boundary in Zircaloy-2 (fig. 7(a)), that the same “rules” do not necessarily apply in alloyed Zr, at least for temperatures - 675 K. IIowever, cavity alignment in layers parallel with (0001) appears to be a feature of material containing large basal plane stacking faults [23], and the alignment in this case may be related to the local solute concenwation or even the alignment of precipitates as opposed to a denuding of interstitial point defects. This type of alignment (unrelated to a grain boundary) has been observed in Zircaloy-2 (fig. 7(b)) which also contains large basal plane stackingfaults and precipitates in layers parallel to (0001). Cavities and stacking-faults are often associated with precipitates (fig. 8).
4. Discupsion
Fig. 4. Micrographs illustrating the distribution of cavities (a) and dislocation loops(b) near to grain boundaries in crystal-bar Zrirradiatedat700Ktoafiuenceof1.5x10”nm-2(E>1 MeV). The boundaries are parallel with the electron beam direction at the orientation shown (B - [liol]); vacancy and interstitial loops are labeki v and i, respectively.
Models for irradiation growth which assume that there is a net fhtx of interstitials to (a)-type dislocations and a net flux of vacancies to grain boundaries may be valid for cold-worked materials because network dislocations will be biased interstitial sinks due to the sizeeffect interaction. However, such models cannot account for irradiation growth of annealed materials containing equal numbers of interstitial and vacancy (4)~type loops or cold-worked materials containing large number of c-component dislocations. It has become increasingly obvious that irradiation growth can only be consistently modelled if it is assumed that there is intrinsic anisotropic diffusion of interstitial point defects [17]. In such circumstances grain boundaries are no longer-neutral sinks and their effective sink strength is then determined by their orientation. The results presented here show that grain boundaries need not necessarily be considered as net vacancy sinks when the dislocation density is low (< 1013 mm2). A general orientation dependence is observed for the sink efficiency of various grain boundaries {as determined from the width of the denuded zones for interstitial or
hf. Griffhs et al. / Grain boundary sinks in neutron-irradiated Zr and Zr-alloys
411
Fig. 5. Micrographs illustrating cavities adjacent to grain boundaries of various orientations in crystal-bar Zr irradiated at 700 K to a fluence of 1.5 X 1O26 n m-* (E < 1 MeV). Those grain boundary orientations that have been determined are shown in the micrographs. Uniabelled arrows indicate projected (1120) directions and labelled amows indicate actual diffracting vectors.
vacancy loops) which supports tial anisotropic diffusion.
the concept
of intersti-
4. I. Grain boundary capture volume In crystal bar Zr irradiated at 700 K the width of the zone denuded of vacancy clusters adjacent to grain boundaries is small (< 0.05 pm) and the maximum width of the denuded zone for interstitial clusters is estimated at about 0.6 pm. This implies that at this temperature the boundaries of grains with diameters c 1.2 pm may be expected to be strongly biased interstitial sinks. Herring and Loretto [24] reported a denuded zone width of about 0.2 pm for interstitials
and 0.1 pm for vacancies in electron irradiated Zr at 675 K. Calculations (below), using the same expressions and parameters as those employed in ref. [24], indicate that the discrepancy is unlikely to be the result of the difference in temperature and suggest that the value reported in ref. [24] is not a maximum or that differences exist between bulk and thin foil measurements. Makin [25] has shown that the width (x) of an interstitials denuded zone is given by: x = 0.66h( Yi/Kz)o.25, where is the jump distance, X pi is the interstitial jump frequency,
(1)
412
M. GriJith et al. / Grain boundary sinks in neutron-irradiated Zr and Zr-alloys
Fig. 6. Distribution of basal plane e-component vacancy loops close to a grain boundary in sponge Zr irradiated at 700 K to a fluenceof1.5x1026nm-2(E>1MeV).
rate, is a the number of atoms in the vacancy/interstitial mutual recombination volume. Using this and the expression given by Holt [26] for interstitial jump frequencies: K
is the damage production
value of 1.8 x lo4 for Z which implies that mutual recombination occurs at a radius of about 5 nm at 700 K. This appears to be an unusually large recombination radius implying that Makin’s theory does not necessarily apply in this case. It is difficult to estimate the width of the denuded zone expected at temperatures relevant to nuclear reactor applications (- 575 K) because the temperature dependence of 2 is not known. However, the loop spacing is reduced by a factor - 3 (determined from measurements of loop densities at 700 K and 575 K [12]), and the width of the denuded zone should also be reduced by the same factor resulting in a maximum value of 0.2 pm for interstitial loops. If it is further assumed that the maximum binding energy with solute atoms is 0.1 eV for Ziicaloy or ZrNb alloys [27], this gives an approximate value for the denuded zone of x = 0.12 pm in these alloys at 573 K. The irradiation growth of annealed small-grained materials (- 1 pm diameter) would then be expected to be dominated by the grain boundary sinks and as such would be largely dependent on such factors as the grain shape and anisotropic interstitial diffusion. 4.2. Cavity formation
Z
pi = 1 X lOI exp( - E,/kT),
(2)
together with an interstitial migration energy (E,) given by Buckley and Manthorpe [27] of 0.27 eV, this gives a
Most reports of cavities in neutron irradiated Zr conclude that they are gas-filled bubbles either because there is a correlation between the number of gas atoms required to fill the cavities and the amount of He produced from (n, a) reactions with impurities such as boron or oxygen [13], or because they are stable during post-irradiation annealing [4]. However Jostsons et al.
Fig. 7. Alignment of cavities parallel with (Oool). (a) Zircaloy-2 irradiated at 665 K to a fluence of 1.6 X 1O26n me2 (E > 1 MeV). (b) Zircaloy-2 irradiated at 710 K to a fluence of 4.2 x 102’ n mT2 (E :, 1 MeV).
M. Griffithr et al. f Grain bow&~
(121 have observed cavities in one sample containing a particularly high B concentration (- 0.1 ppm by wt) compared with other samples irradiated under the same conditions and concluded that they were voids because they were absent at lower ( < 675 K) and higher ( - 723 K) irradiation temperatures. He (an insoluble gas) in this case is required to stabihse a cavity nucleus against collapse to a vacancy loop. Farrell 1281has reported that cavities are absent in neutron irradiated Zr pre-implanted with He and concluded that this is because vacancy loops (formed in collision cascades) are stable and cavities will not form in the presence of significant populations of vacancy loops. This is clearly not the case in the results of Jostsons et al. [12] as the voids are observed along with vacancy loops in the same samples and other factors, such as the overall impurity content of the material, may have to be considered. Voids have been reported in electron irradiated Zr by Carpenter [29] and Faulkner and Woo 1301 only in the presence of pre-implanted He. Buckley and Manthorpe [31] reported voids in specimens that were not pre-implanted with inert gas but also showed that the presence of such gases greatly increased the cavity yield. The voids in these cases appear to be the most stable form of vacancy cluster because vacancy loops were not reported. Cavities are a common feature of irradiated crystalbar Zr and are mostly found adjacent to grain boundaries. Observations indicate that they are faceted by (0001), (liol}and {lOiO}planes, (fig. 3(b)), which is consistent with what might be expected from a consideration of interfacial free energies of hexagonal close-packed metals reported by Miller et al. [32]. The nature of the cavities i.e. whether they are gas filled bubbles or essentially voids (containing some insoluble gas), is not certain at the present time. Nevertheless, it has been shown by Rowcliffe et al. [33] that even gas bubbles require a vacancy flux to grow and therefore they will grow preferentially in volumes containing the highest vacancy concentrations. Although the present circumstances are somewhat different from the results of [33] which were for stainless steels, the existence of cavities close to grain boundaries implies that there is a higher vacancy concentration in these regions or conversely that the interstitial concentration is lower than in the centre of the grains. This is consistent with the dislocation loop analyses indicating that there is a zone denuded of interstitial point defects close to the boundaries and therefore that they are net interstitial sinks. The fact that cavities are more common in crystal-bar Zr compared with sponge Zr or Zircaioy-2 is consistent with high ~n~ntration of vacancy c-compo-
sinks in neutron-irr~d~~t~dZr and Zr-atloys
413
nent loops (which compete for vacancy point defects) in the latter materials even though gas bubbles are more likely to be produced because of the increased probability of producing He from the higher solute/impurity concentration. 4.3. Grain boundaries and irradiation growth 4.3.1. Effect of grain boundaries on microstructure The non-u~form dist~bution of cavities is generally
consistent with the existence of an interstitial denuded zone (fig. 4), or the presence of precipitates acting either as nucleation sites or as sources of He from (n,o) reactions with impurities [28], (figs. 3 and 8). The correlation with the presence of precipitates also appears to be linked with the appearance of c-component loops (see stacking-faults in fig. 8(a)) and it is apparent that impurities influence both the formation of cavities and the formation of the c-component defects 1151. Heterogeneous dist~butio~ of cavities or cavity-free volumes have been observed in ion or neutron irradiated Fe [34], Ni [35] and stainless steels 133,363. The observations of cavity free volumes adjacent to gram boundaries in material containing an otherwise uniform distribution of cavities (voids) have generally been explained in terms of gram boundary migration [35,36]. The observation of isolated clusters of cavities either at grain boundaries [33] or within grams [34] has been attributed to the presence of impurities or precipitates at grain boundaries. Cavities form on the gram boundaries which then migrate leaving the cavity cluster behind. The same mechanism is unlikely to be occurring in the present circumstances because cavities are sometimes observed on both sides of the boundary (figs. 3(c) and 5(a)) and because there are no instances for which cavities are observed actually on the gram boundary. Recent work on neutron irradiated Ti [23] has shown that clusters of cavities parallel with (0001) are often associated with large faulted c-component loops and by implication with a high local impurity concentration. The existence of the cavities and c-component loops in the same volume has been associated with the local solute/precipitate concentration (this work) but is also found in some cases to be dependent on the proximity of grain boundaries [5]. This may be because of the presence of impurities at grain boundaries but is also likely to be because of the absence of interstitial point defects. Similar observations regarding the role of grain boundaries as preferential interstitial sinks have been made by Horsewell and Singh on neutron irradiated aluminum [ 371. Apart from the results of Herring et al. 1191and the
M. Grifjiths et al. / Grain boundary sinks,in neutron-irradiated
Zr
and Zr-alloys
Fig. 8. Cavities (arrowed) associated with precipitates or precipitate clusters. (a) Sponge Zr irradiated at 710 K to a fluence of 4.5 x 10z5 n me2 (E > 1 MeV). Note the stacking-fault (SF). (b) Zircaloy-2 irradiated at 675 K to a fluence of 6 X 1O25 n me2 (E > 1 MeV). Stacking-faults (not visible at this orientation) are also present in the field of view.
present work the only other report of grain boundary influence on microstructure was given by Jostsons et al. [12]. They reported a lower vacancy loop concentration in one sample of Zr which was attributed to the presence of low angle subgrain boundaries. This suggests that the sub-grain boundaries in that case were net vacancy sinks and is inconsistent with the present results. It is also inconsistent with other observations of increased cavity formation close to sub-grain boundaries [S] implying that sub-grain boundaries act as net interstitial sinks. The reason for this is not clear. One possible explanation is that the sub-grain boundaries referred to in ref. [12] may have had a different orientation to those referred to here (parallel with (lOi0) and
point defects, i.e. they may be diffusing faster than vacancies along [OOOl] as well as (1120) directions. The observation of cavities along the particular boundary shown in fig. 7(a) may be related to the presence of solutes or precipitates (which are apparent in the micrograph), or to cavity nucleation in a layer parallel with (0001) which is coincidental and not related to the adjacent grain boundary. It is also possible that the anisotropy of diffusion is affected by the increased solute concentration in the matrix during irradiation. There is some indication [38] that whereas there is a marked anisotropy in the diffusion of some species (either self-interstitials or interstitial Fe) during irradiation of Zircaloy-2 at about 580 K, this is not very apparent at 665 K and may even be reversed.
4.3.2. Orientation dependence In the absence of any impurity effects, the one common feature of the denuded zones close to grain boundaries is the dependence of width on the orientation of the grain boundary. This dependence is such that grain boundaries which are parallel with closepacked (1120) directions are the least efficient sinks for interstitial point defects. This is consistent with and supports the concept of anisotropic interstitial diffusion .[.17]. This criterion does not restrict cavity formation at unfavourably oriented grain .boundaries in certain circumstances, illustrated in fig. 7(a) for Zircaloy-2 irradiated at 665 K, because grain boundaries parallel with
4.3.3. Application to irradiation growth In deriving models to explain most of the growth data for polycrystalline and single crystal Zr [2,3], one can assume that grain boundaries are either net interstitial or vacancy sinks. The results of Jostsons et al. [12] indicate that during neutron irradiation at 673 K certain sub-grain boundaries in annealed Zr may be net vacancy sinks. The treatment of grain boundaries as net vacancy sinks in annealed materials not containing c-component vacancy loops requires that most of the (a)-type dislocation loops are interstitial in nature. Treating the grain boundaries as net vacancy sinks is mostly applicable to cold-worked materials if (a)-type network dislocations are considered as net interstitial sinks. The growth of
{ioil).
(0001)
may
still
be considered
as sinks
for interstitial
A4. Griffhs
et al. / Grain boundq
cold-worked materials requiring a net flux of interstitials to (a)-type dislocations and vacancies to c-component dislocations or grain boundaries can then best be explained in terms of anisotropic diffusion [17]. For annealed materials, treating the grain boundaries as interstitial sinks requires that growth is also explained in terms of anisotropic diffusion. However, grain shape effects (with more interstitials impinging on grain boundaries perpendicular to the direction of smallest aspect ratio) and the formation of c-component vacancy dislocation loops are important and may dominate in certain circumstances. One explanation of the single crystal data relies on the formation of sub-microscopic vacancy c-component loops [l] and, further to this, it has been proposed that such loops (formed by cascade collapse) provide a fixed contribution to the irradiation growth [39]. Whereas anisotropic diffusion appears to be a factor which will affect irradiation growth, it cannot account for the higher growth experienced by annealed polycrystalline material at 353 K compared with 553 K [3]. A combination of mechanisms can best explain the observations with the assumption that there are more interstitial (a)-type loops following irradiation at 353 K compared with 553 K [3], although there is no experimental data to support this. Long term growth after the dislocation structure has saturated may best be explained by anisotropic diffusion and will be inversely dependent on grain size. Irradiation growth will be further complicated by the development of intergranular stresses induced by the growth of individual grains. The present observations indicate that in certain circumstances, for annealed materials at least, grain boundaries can be considered as dominant interstitial sinks in the overall microstructure. This may be applicable to anomalous growth observed in new ZrNb pressure tube materials [40,41].
4.4. Summary
of results
(1) There is selective denuding
(2)
of interstitial point defects close to grain and sub-grain boundaries in neutron irradiated Zr and Zr-alloys at temperatures 1675 K. This is manifested by the appearance of cavities or the absence of interstitial dislocation loops. The width of the denuded zone is smallest for boundaries parallel with one or more (11~0) directions implying that diffusion of interstitial point defects is most rapid along (1120) directions.
sinks in neutron-irradiated Zr and Zr-alloys
415
5. Conclusions There are two main implications of the results presented here with respect to in-reactor deformation: (1) Grain boundaries can be net sinks for interstitials point defects. In annealed polycrystalline or single crystal material, unless there are more interstitial compared with vacancy (a)-type loops or unless there are vacancy c-component loops, irradiation growth can only be the result of the grain shape or anisotropic diffusion or both. (2) The orientation dependence for the width of the zone denuded of interstitial point defects, but which may contain cavities and vacancy loops, is consistent with and supports the concept of interstitial anisotropic diffusion in the basal plane.
Acknowledgements The authors would like to thank C.H. Woo of WNRE and B.A. Cheadle of CRNL for useful comments on the manuscript.
References PI G.J.C. Carpenter, R.A. Murgatroyd, A. Rogerson and J.F. Watters, J. Nucl. Mater. 101 (1981) 28. PI R.A. Murgatroyd and A. Rogerson, in: Proc. BNES Conf. on Dimensional Stability and Mechanical Behaviour of Irradiated Metals, Brighton, Vol. 2 (BNES, London, 1984) p. 93. 131 V. Fidleris, R.P. Tucker and R.B. Adamson, ASTM STP 939 (1987) 49. 141 C.D. Cann, D. Faulkner, K. Nuttall, R.C. Styles, A.J. Shillinglaw, C.K. Chow and A.J. Rogowski, AECL-8406 (1986). [51 M. Griffiths, C.D. Cann and R.C. Styles, unpublished work. 161 P.M. Kelly, in: Proc. Int. Conf. on Physical Metallurgy of Reactor Fuel Elements, Berkeley Nuclear Laboratories, UK (The Metals Society, London, 1973) p, 123. 171 G.J.C. Carpenter and D.O. Northwood, J. Nucl. Mater. 56 (1975) 260. 181 C.C. Dollins, J. Nucl. Mater. 59 (1976) 61. 191 R.A. Holt and E.F. Ibrahim, Acta Metall. 27 (1979) 1319. [101 S.R. MacEwen and G.J.C. Carpenter, J. Nucl. Mater. 90 (1980) 108. Vll D.O. Northwood, R.W. Gilbert, L.E. Bahen, P.M. Kelly, R.G. Blake, A. Jostsons, P.K. Madden, D. Faulkner, W. Bell and R.B. Adamson, J. Nucl. Mater. 79 (1979) 379. [12] A. Jostsons, P.M. Kelly, R.G. Blake and K. Farrell, ASTM STP 683 (1979) 46.
416
M. Griffith
et al. / Grain boundary sinks in neutron-irradiated
[13] R.W. Gilbert, K. Farrell and C.E. Coleman, J. Nucl. Mater. 84 (1979) 137. [14] M. Griffiths and R.W. Gilbert, unpublished work. [15] M. Griffiths and R.W. Gilbert, J. Nucl. Mater. 150 (1987) 169. [16] R.A. Holt and R.W. Gilbert, J. Nucl. Mater. 137 (1986) 185. 1171 C.H. Woo and U. Gosele, J. Nucl. Mater. 119 (1983) 219. [IS] A.R Causey, V. Fidleris and S.R. MacEwen, unpublished work. (191 R.A. Herring, M. Griffiths, M.H. Loretto and R.E. Smallman, in: Proc. Annual Conf. of the Electron Microscopical Society of America, Louisville, Kentucky (1985). [20] C.E. Ells, C.E. Coleman and CD. Williams, in: Proc. Symp. on Mechanical Behaviour of Materials, Kyoto (Society of Materials Science of Japan, 1974) p. 77. [21] D.M. Maher and B.L. Eyre, Philos. Mag. 23 (1971) 409. [22] H. Foll and M. Wilkens, Phys. Status Solidi A31 (1975) 519. [23] M. Griffiths, C.D. Camt and R.C. Styles, J. Nucl. Mater. 149 (1987) 200. [24] R.A. Herring and M.H. Loretto, Proc. 6th Annual Conf. of Canadian Nuclear Society Ottawa, Eds. P.M. French and G.J. Phillips (1985) p. 8.18. [25] M.J. Makin, Philos. Mag. 20 (1969) 1133. [26] R.A. Holt, J. Nucl. Mater. 90 (1980) 193. [27] S.N. Buckley and S.A. Manthorpe, J. Nucl. Mater. 90 (1980) 169.
[28] [29] [30] [31]
(321 [33] [34] [35] [36] [37] [38] [39]
[40] [41]
Zr and Zr-alloys
K. Farrell, Radiat Eff. 53 (1980) 175. G.J. Carpenter, Radiat. Eff. 19 (1973) 189. D. Faulkner and C.H. Woo, J. Nucl. Mater. 90 (1980) 307. S.N. Buckley and S.A. Manthorpe, in: Proc. Int. Conf. on Physical Metallurgy of Reactor Fuel Elements, Berkeley Nuclear Labs. (The Metals Society, London, 1973) p. 127. W.A. Miller, G.J.C. Carpenter and G.A. Chadwick, Philos. Mag. 19, No. 158 (1969) 305. A.F. Rowcliffe, G.J.C. Carpenter, H.F. Merrick and R.B. Nicholson, ASTM SIP 426 (1967) 161. K. Farrell and J.T. Houston, J. Nucl. Mater. 35 (1970) 352. C.W. Chen and R.W. Buttry, Radiat. Eff. 56 (1981) 219. W.V. Vaidya, J. Nucl. Mater. 113 (1983) 219. A. Horsewell and B.N. Singh, ASTM STP 870 (1985) 248. M. Griffiths, R.W. Gilbert and G.J.C. Carpenter, J. Nucl. Mater. 150 (1987) 53. W. Phythian and R.J. McElroy, paper presented at this Conference (Int. Conf. on Fundamental Mechanisms of Radiation-Induced Creep and Growth, Hecla Island, Manitoba, Canada, June 22-25,1987). J.D. Parker, V. Perovic, M. Leger and R.G. Fleck, ASTM STP 939 (1987) 86. R.G. Fleck, J. Tadros, V. Perovic and R.A. Holt, in: these Proc. J. Nucl. Mater. 159 (1988).
405
GRAIN BOUNDARY SINKS IN NEUTRON-IRRADIATED Zr AND %-ALLOYS M. GRIFFITH& R.W. GILBERT and C.E. COLEMAN ~e~ailur~icuI Engineering Branch, Chalk Riuer Nuclear Laboratories, Chalk River, Ontario, Kill IJO Canada
Samples of annealed sponge and crystal-bar Zr and Zircaloy-2 have been examined following irradiation in EBR-II at temperatures - 700 K. Loop analysis shows that there is selective denuding of interstitial loops near to some grain boundaries indicating that such boundaries are net sinks for interstitial point defects. Furthermore, in sponge Zr and Zircaloy-2, vacancy c-component loops are observed running into the grain boundaries showing that the grain boundaries are not preferred sinks for vacancies. Cavities are observed in all samples. In crystal-bar Zr and sponge Zr they are mostly observed adjacent to grain boundaries. They are also sometimes found within grains associated with precipitates. The cavities are more common in the crystal-bar Zr and this is probably because both the sponge Zr and Zircaloy-2 contain vacancy c-component loops which compete for vacancies (assuming that the cavities are vacancy sinks). Only some of the grain boundaries have cavities adjacent to them and this may be related to the orientation of the boundary.
1. Introduction 1.1. The importance growth
of grain boundaries
in irradiation
Since irradiation growth in annealed polycrystalhne Zr is generally about ten times greater than for single crystal material [1,2] and increases with decreasing grain size [2-41, grain boundaries have been considered as important sinks for point defects, especially in annealed Zr and Zr-alloys for which the only other type of sink is (a)-type dislocation loops. In single crystal material the growth quickly saturates for fluences - 0.1 X 102’ n mV2 (E > 1 MeV). In annealed polycrystalline material growth almost saturates at low fluences but there is a smaller long term component, proportional to the fluence, which indicates the important of grain boundaries. The various irradiation growth data for annealed Zr and Zr-alloys are summarized in fig. 1. Carpenter et al. showed that single crystals exhibited irradiation growth indicating that there is some component of growth which is not related to the absorption of point defects at grain boundaries. In the absence of any ~crost~ctural data, this growth was attributed to the presence of sub-microscopic c-component vacancy loops or elastic relaxation around vacancy point defects and vacancy clusters or a non-Hookeian contraction around small (a)-type loops.
Murgatroyd and Rogerson [2] presented evidence that growth of Zr is sensitive to and inversely dependent on grain size at 553 K but is relatively independent of grain size at 353 K for grains of 5 gm and 40 pm diameter. The growth of polycrystalhne iodide Zr at 353 K was an order of magnitude higher than at 553 K for material with 40 pm diameter grains and about double for the material with 5 pm diameter grains; growth was still continuing in the latter material whereas it had virtually saturated at about 0.1 X 10” n rn-’ in the former. This temperature dependence was reversed for Zircaloy-2 and was non-existent for single crystal Zr. The results were explained by the formation of vacancy (u)-loops at the higher temperature and the fact that grain boundaries compete more effectively for vacancies when the grain size is small thereby reducing the vacancy (“)-type loop population compared with larger grained material. The lower growth of the Zircaloy at 353 K was explained in terms of the trapping of vacancies by Sn atoms. Fidleris et al. [3] have shown that growth is inversely dependent on grain size in annealed polycrystalline Zr for grains of 23 pm and 225 gm diameter irradiated at 330 IL There was a similar dependence for Zircaloy-2 with grains of 12 pm and 31 pm diameter irradiated at about 680 K but this was reversed for irradiation at 330 K. Cann et al. [4] also showed that there is a marked grain size effect on the initial growth transient at tem-
M. Griffiths et al. / Grain boundary sinks in neutron-IrradiatedZr and Zr-alloys
406
0 14
P g.s.(Zr and Zr-2)
0.12 0 10
REFERENCE
0 08 0.06 0.04 0.02
cc--
SINGLE CRYSTAL
(Zr)
i
ia1
I 10 0
5.0 FLUENCE (~t)/lOzs n.mm2
REFERENCE 014 -
0 2 pm g.s.iZrt
FLUENCE #V10z5
-23 ----
[ll
n.m-*
Fig. 1. Irradiationgrowth measuredalong the a-axis (singlecrystal) or longitudinalaxis (polycrystal) for annealedZr and Zircaloy-2 during neutron irradiation at (a) 330-355 K and (b) 550-590 K. For poly~~s~~~ne specimensthe fraction of basal poles in the longitudiialdirectionis = 0.1 (i.e. between0.07-0.16). Grain sizediameters(g.s.) are indicatedfor each plot.
peratures between 566 K and 621 K with the smallest grain size material exhibiting the highest growth strain; the slower growth strain rates after the initial transient were similar for the different grain size materials. They also showed that swelling of both annealed and coldworked material occurred at temperatures between 703 K and 778 K with the peak at 741 K, this could be attributed in part to the presence of cavities (believed to be He gas bubbles) in this temperature range. Recent observations [S] show that the cavities are distributed non-uniformly in these samples and are either observed
on precipitates or in clusters near to grain or sub-grain boundaries. Vacancy c-component loops are also often found in the vicinity of the cavities. 1.2. Grain boundaries in models for irradiation growth To account for the net a-axis expansion and c-axis contraction observed during irradiation growth, Kelly [6] originally proposed a mechanism in which vacancy and interstitial point defects migrate a~sotropi~ly to grain boundaries. This anisotropic flow develops in
M. Griffiths et al. / Grain boundary sinks in neutron-irradiated
response to the internal stress generated within grains because of anisotropic thermal expansion. The stress would not be maintained once the strain from the thermal expansion is relieved by the point defect flux to the grain boundaries and the result would be a saturation of strain with increasing fluence. The net strain on cooling would be positive along the u-axes and negative along the c-axis. Unfortunately this model did not account for the growth of single crystals or the slower long term component of growth for the polycrystalline material, unless this was a result of thermal cycling when the samples were removed from the reactor for each dimensional measurement. Carpenter and Northwood 171 proposed a simpler model based on the intrinsic bias of dislocations resulting in a net flux of interstitials to (a)-type dislocations (network or loops) leaving residual vacancies to annihilate at grain boundaries which they considered to be neutral sinks. Whereas this may be valid for cold-worked materials, containing dislocations with mostly (a)-type Burgers vectors which are considered as net interstitial sinks because of the size-effect interaction [8-lo], it cannot easily account for growth of annealed materials unless all or most of the dislocation loops generated during irradiation are interstitial in nature. This was the basis for the explanation of growth in single crystal Zr [l]; the c-axis contraction was then attributed to either relaxation around vacancies, the formation of very small c-component vacancy loops or a e-axis contraction around (a)-type dislocation loops. There are no rigorous dislocation loop analyses for material irradiated at temperatures < 580 K, i.e. at temperatures relevant to nuclear reactor applications; most dislocation loop analyses are for material irradiated at > 580 K [11,12]. Analysis of the loops in crystal bar Zr [13] and Zircaloy-2 [14] following irradiation at - 575 K have so far revealed that both vacancy and interstitial (o)-type loops are formed. However, there is insufficient evidence to estimate the relative numbers of each or whether there is selective denuding adjacent to grain boundaries. For annealed Zircaloys, as the fluence is increased to - 5 X 10” n rnF2, e-component vacancy loops having Burgers vectors of b = i (2023) begin to form [15]. This loop formation corresponds with a state of accelerated growth 1161and introduces a new factor into the treatment of growth of annealed materials. It also suggests that the treatment for cold-worked materials should include a contribution from the [c] or (c + a) network dislocations even though they have a different (larger) Burgers vector compared with the c-component loops formed during irradiation. Current models can account
407
Zr and Zr-alloys
for the observed irradiation growth by assuming either anisotropic interstitial diffusion [17] or a modified interstitial bias such that (a)-type dislocations have a larger bias parameter compared with c-component dislocations [lSJ, The result is a net flux of interstitials to (a)-type dislocations and a net flux of vacancies to c-component dislocations. Grain boundaries cease to be neutral sinks when point defect diffusion is anisotropic 1171 and the sink strength is then dependent on their orientation. Herring et al. [19] reported that the zone denuded of interstitial loops close to grain boundaries is greater than that for vacancy loops in electron irradiated Zr and Zircaloy-4 for temperatures - 675 K. This implies that the grain boundaries may be net sinks for interstitial point defects. It was also indicated that there may be a reIation between the magnitude of the denuded zone for interstitial loops and the orientation of the thin foil i.e. the effect is more pronounced for foil normals close to [OOOl]. This may be related to the grain boundary orientation but is equally likely to be related to the foil orientation; the foil surfaces being an altemative sink. Zones denuded of dislocation loops at grain boundaries have been observed in Zr and its alloys irradiated with neutrons at temperature > 670 K 19,201. In this paper we describe the defect structures observed near to grain boundaries in neutron irradiated crystalbar Zr, sponge Zr and Zircaloy-2.
2. Experlrnental Samples of crystal-bar, sponge Zr and Zircaloy-2 were irradiated in EBR-II at temperatures between 665 Table 1 Ingot analyses for crystal-bar Zr, sponge Zr and Zircaloy-2 major impurity or ailoying elements (ppm by wt)
Sn Fe Cr Ni B c N 0 H Hardness in BHN
Crystal-bar Zr
Sponge Zr
Zircaloy-2
14750 1750 1050 600 < 0.2 115 35 1200 12
57
143
174
408
M.
Grifjiths et al. / Grain boundav sinks in neutron-irradiated
and 710 K for fluences up to 1.5 X 1O26 n m-’ (E > 1 MeV). The ingot analyses for the major impurity or alloying elements in these samples are given in table 1. Thin foils for transmission electron microscopy were prepared by twin-jet electropolishing using a solution of 10% per&lo& acid in ethanol at temperatures < 243 K and were analyzed using a Philips EM3OOG electron
Fig. 2. Distribution of vacancy and intersuuiu IUu,.r -‘rm *n grain boundaries in sponge Zr irradiated at 710 K to a fluence of 4.5x1025 ” m-2 (E > 1 MeV). Loops exhibiting inside contrast with a z25 diffracting vector are vacancy in nature; beam direction _ [ii23].
Zr and Zr-alloys
microscope operating at 100 kV. Dislocation loop analyses were performed using the Maher and Eyre [21] and Foll and Wilkens [22] techniques.
3. Results Analysis of the (u)-type dislocation loops near to grain boundaries in sponge Zr irradiated at 710 K to a fluence of 4.5 X 102’ n rnp2 is illustrated in fig. 2. There is a demarcation (dotted lines) between regions containing vacancy loops only, i.e. adjacent to the grain boundaries, and regions containing a mixture of interstitial and vacancy loops, i.e. in the centre of the grain. A similar demarcation is observed in crystal bar Zr irradiated at 700 K to fluences of 1.5 x lot6n rne2. In both cases the mid-grain region contains approximately equal numbers of vacancy and interstitial loops and also a dislocation network which developed from loop growth; it is not possible to say which type of loops contributed to the network. In some small grains (about 2-3 pm diameter} the interstitial loop population was estimated at about 25%. There was no qualitative difference between the distribution of loops close to high angle grain boundaries and sub-grain boundaries (consisting mostly of rows of edge (u)-type dislocations with spacings of about 0.01-0.02 pm on {lOiO> and {1Oi 1f planes). This indicates that the long range stress fields of both types are similar and that any difference has an insignificant effect on the loop distribution. Another common feature of the irradiation damage in the crystal bar Zr close to grain boundaries is the existence of cavities (fig. 3(a)). Similar cavity clusters are also observed in sponge Zr (fig. 3(b)) but are less common. They are sometimes observed nucleated on or close to precipitates (fig. 3(c)). They are mostly found close to grain boundaries but are occasionally observed in mid-grain regions and are invariably associated with precipitates (fig. 3(d)). Analysis shows that the zones containing cavities adjacent to grain boundaries are denuded of interstitial loops as illustrated in fig. 4. This figure also indicates that there is an apparent orientation dependence for the width of the interstitial denuded zone, i.e. grain boundaries parallel with (1120) directions have a narrower denuded zone compared with boundaries which are not. If the existence of cavities is taken as an indication of a depletion of interstitial point defects, this effect appears to be most marked for grain boundaries parallel or perpendicular to (0001) which contains all three (1120) directions. There is little evidence cavities adjacent to boundaries having normals
M. Griffiths et al. / Grain bounahty sinks in neutron-irradiated
Zr and Zr-alloys
Fig. 3. Cavities in crystal-bar Zr and sponge Zr (B) irradiated at 700 K to a fluence of 1.5 X lox n mm2 (E > 1 MeV). (A) Cluster adjacent to a grain boundary, (B) cluster adjacent to a grain boundary in sponge Zr. Note faeetting on (OOOl),{ liO1)and (lOi0) planes, (C) cluster adjacent to a grain boundary and associated with precipitates,(D) cavity nucleated on or close to a precipitate.
(N) for which N * (11~0) = 0, as illustrated in fig. 5. Dislocation loop analysis indicates that grain boundaries parallel with (0001) are still preferential interstitial sinks although the width of the denuded zone for interstitial as well as vacancy defects is smaller than for other grain boundary orientations. The width of the zone denuded of vacancy loops appears to be smaller than that for interstitial loops irrespective of the orientation of the boundary.
Post-irradiation annealing at 700 K for 24 h resulted in little discernible change in the number and size distribution of cavities. In both the as-irradiated and annealed samples there was a bimodal distribution of diameters with a large group peaking at about 30 nm and a smaller group peaking at about 60 nm. Annealing at 825 K for 16 h resulted in an apparent reduction in the number of cavities, although this may simply be the result of a random sampling error. Those remaining
410
hf. Griffithr et al. / Grain ~~~
sinks in neutron-herniated Zr and Zr-ailoys
were Iarge with an average diameter of about 80 nm. One reason for there being fewer cavities in sponge Zr compared with crystal bar Zr may be because of the stability of basal plane c-component loops in the former material. These loops are invariably vacancy in nature [15] and are often observed running into grain boundaries (fig. 6). Whereas there is a clear indication of an orientation dependence for the width of the interstitial denuded zone in crystal bar Zr, there is insuf~cient data to discern a grain boundary orientation effect in sponge Zr. There is even an indication, from one observation of cavities adjacent to a basal plane grain boundary in Zircaloy-2 (fig. 7(a)), that the same “rules” do not necessarily apply in alloyed Zr, at least for temperatures - 675 K. IIowever, cavity alignment in layers parallel with (0001) appears to be a feature of material containing large basal plane stacking faults [23], and the alignment in this case may be related to the local solute concenwation or even the alignment of precipitates as opposed to a denuding of interstitial point defects. This type of alignment (unrelated to a grain boundary) has been observed in Zircaloy-2 (fig. 7(b)) which also contains large basal plane stackingfaults and precipitates in layers parallel to (0001). Cavities and stacking-faults are often associated with precipitates (fig. 8).
4. Discupsion
Fig. 4. Micrographs illustrating the distribution of cavities (a) and dislocation loops(b) near to grain boundaries in crystal-bar Zrirradiatedat700Ktoafiuenceof1.5x10”nm-2(E>1 MeV). The boundaries are parallel with the electron beam direction at the orientation shown (B - [liol]); vacancy and interstitial loops are labeki v and i, respectively.
Models for irradiation growth which assume that there is a net fhtx of interstitials to (a)-type dislocations and a net flux of vacancies to grain boundaries may be valid for cold-worked materials because network dislocations will be biased interstitial sinks due to the sizeeffect interaction. However, such models cannot account for irradiation growth of annealed materials containing equal numbers of interstitial and vacancy (4)~type loops or cold-worked materials containing large number of c-component dislocations. It has become increasingly obvious that irradiation growth can only be consistently modelled if it is assumed that there is intrinsic anisotropic diffusion of interstitial point defects [17]. In such circumstances grain boundaries are no longer-neutral sinks and their effective sink strength is then determined by their orientation. The results presented here show that grain boundaries need not necessarily be considered as net vacancy sinks when the dislocation density is low (< 1013 mm2). A general orientation dependence is observed for the sink efficiency of various grain boundaries {as determined from the width of the denuded zones for interstitial or
hf. Griffhs et al. / Grain boundary sinks in neutron-irradiated Zr and Zr-alloys
411
Fig. 5. Micrographs illustrating cavities adjacent to grain boundaries of various orientations in crystal-bar Zr irradiated at 700 K to a fluence of 1.5 X 1O26 n m-* (E < 1 MeV). Those grain boundary orientations that have been determined are shown in the micrographs. Uniabelled arrows indicate projected (1120) directions and labelled amows indicate actual diffracting vectors.
vacancy loops) which supports tial anisotropic diffusion.
the concept
of intersti-
4. I. Grain boundary capture volume In crystal bar Zr irradiated at 700 K the width of the zone denuded of vacancy clusters adjacent to grain boundaries is small (< 0.05 pm) and the maximum width of the denuded zone for interstitial clusters is estimated at about 0.6 pm. This implies that at this temperature the boundaries of grains with diameters c 1.2 pm may be expected to be strongly biased interstitial sinks. Herring and Loretto [24] reported a denuded zone width of about 0.2 pm for interstitials
and 0.1 pm for vacancies in electron irradiated Zr at 675 K. Calculations (below), using the same expressions and parameters as those employed in ref. [24], indicate that the discrepancy is unlikely to be the result of the difference in temperature and suggest that the value reported in ref. [24] is not a maximum or that differences exist between bulk and thin foil measurements. Makin [25] has shown that the width (x) of an interstitials denuded zone is given by: x = 0.66h( Yi/Kz)o.25, where is the jump distance, X pi is the interstitial jump frequency,
(1)
412
M. GriJith et al. / Grain boundary sinks in neutron-irradiated Zr and Zr-alloys
Fig. 6. Distribution of basal plane e-component vacancy loops close to a grain boundary in sponge Zr irradiated at 700 K to a fluenceof1.5x1026nm-2(E>1MeV).
rate, is a the number of atoms in the vacancy/interstitial mutual recombination volume. Using this and the expression given by Holt [26] for interstitial jump frequencies: K
is the damage production
value of 1.8 x lo4 for Z which implies that mutual recombination occurs at a radius of about 5 nm at 700 K. This appears to be an unusually large recombination radius implying that Makin’s theory does not necessarily apply in this case. It is difficult to estimate the width of the denuded zone expected at temperatures relevant to nuclear reactor applications (- 575 K) because the temperature dependence of 2 is not known. However, the loop spacing is reduced by a factor - 3 (determined from measurements of loop densities at 700 K and 575 K [12]), and the width of the denuded zone should also be reduced by the same factor resulting in a maximum value of 0.2 pm for interstitial loops. If it is further assumed that the maximum binding energy with solute atoms is 0.1 eV for Ziicaloy or ZrNb alloys [27], this gives an approximate value for the denuded zone of x = 0.12 pm in these alloys at 573 K. The irradiation growth of annealed small-grained materials (- 1 pm diameter) would then be expected to be dominated by the grain boundary sinks and as such would be largely dependent on such factors as the grain shape and anisotropic interstitial diffusion. 4.2. Cavity formation
Z
pi = 1 X lOI exp( - E,/kT),
(2)
together with an interstitial migration energy (E,) given by Buckley and Manthorpe [27] of 0.27 eV, this gives a
Most reports of cavities in neutron irradiated Zr conclude that they are gas-filled bubbles either because there is a correlation between the number of gas atoms required to fill the cavities and the amount of He produced from (n, a) reactions with impurities such as boron or oxygen [13], or because they are stable during post-irradiation annealing [4]. However Jostsons et al.
Fig. 7. Alignment of cavities parallel with (Oool). (a) Zircaloy-2 irradiated at 665 K to a fluence of 1.6 X 1O26n me2 (E > 1 MeV). (b) Zircaloy-2 irradiated at 710 K to a fluence of 4.2 x 102’ n mT2 (E :, 1 MeV).
M. Griffithr et al. f Grain bow&~
(121 have observed cavities in one sample containing a particularly high B concentration (- 0.1 ppm by wt) compared with other samples irradiated under the same conditions and concluded that they were voids because they were absent at lower ( < 675 K) and higher ( - 723 K) irradiation temperatures. He (an insoluble gas) in this case is required to stabihse a cavity nucleus against collapse to a vacancy loop. Farrell 1281has reported that cavities are absent in neutron irradiated Zr pre-implanted with He and concluded that this is because vacancy loops (formed in collision cascades) are stable and cavities will not form in the presence of significant populations of vacancy loops. This is clearly not the case in the results of Jostsons et al. [12] as the voids are observed along with vacancy loops in the same samples and other factors, such as the overall impurity content of the material, may have to be considered. Voids have been reported in electron irradiated Zr by Carpenter [29] and Faulkner and Woo 1301 only in the presence of pre-implanted He. Buckley and Manthorpe [31] reported voids in specimens that were not pre-implanted with inert gas but also showed that the presence of such gases greatly increased the cavity yield. The voids in these cases appear to be the most stable form of vacancy cluster because vacancy loops were not reported. Cavities are a common feature of irradiated crystalbar Zr and are mostly found adjacent to grain boundaries. Observations indicate that they are faceted by (0001), (liol}and {lOiO}planes, (fig. 3(b)), which is consistent with what might be expected from a consideration of interfacial free energies of hexagonal close-packed metals reported by Miller et al. [32]. The nature of the cavities i.e. whether they are gas filled bubbles or essentially voids (containing some insoluble gas), is not certain at the present time. Nevertheless, it has been shown by Rowcliffe et al. [33] that even gas bubbles require a vacancy flux to grow and therefore they will grow preferentially in volumes containing the highest vacancy concentrations. Although the present circumstances are somewhat different from the results of [33] which were for stainless steels, the existence of cavities close to grain boundaries implies that there is a higher vacancy concentration in these regions or conversely that the interstitial concentration is lower than in the centre of the grains. This is consistent with the dislocation loop analyses indicating that there is a zone denuded of interstitial point defects close to the boundaries and therefore that they are net interstitial sinks. The fact that cavities are more common in crystal-bar Zr compared with sponge Zr or Zircaioy-2 is consistent with high ~n~ntration of vacancy c-compo-
sinks in neutron-irr~d~~t~dZr and Zr-atloys
413
nent loops (which compete for vacancy point defects) in the latter materials even though gas bubbles are more likely to be produced because of the increased probability of producing He from the higher solute/impurity concentration. 4.3. Grain boundaries and irradiation growth 4.3.1. Effect of grain boundaries on microstructure The non-u~form dist~bution of cavities is generally
consistent with the existence of an interstitial denuded zone (fig. 4), or the presence of precipitates acting either as nucleation sites or as sources of He from (n,o) reactions with impurities [28], (figs. 3 and 8). The correlation with the presence of precipitates also appears to be linked with the appearance of c-component loops (see stacking-faults in fig. 8(a)) and it is apparent that impurities influence both the formation of cavities and the formation of the c-component defects 1151. Heterogeneous dist~butio~ of cavities or cavity-free volumes have been observed in ion or neutron irradiated Fe [34], Ni [35] and stainless steels 133,363. The observations of cavity free volumes adjacent to gram boundaries in material containing an otherwise uniform distribution of cavities (voids) have generally been explained in terms of gram boundary migration [35,36]. The observation of isolated clusters of cavities either at grain boundaries [33] or within grams [34] has been attributed to the presence of impurities or precipitates at grain boundaries. Cavities form on the gram boundaries which then migrate leaving the cavity cluster behind. The same mechanism is unlikely to be occurring in the present circumstances because cavities are sometimes observed on both sides of the boundary (figs. 3(c) and 5(a)) and because there are no instances for which cavities are observed actually on the gram boundary. Recent work on neutron irradiated Ti [23] has shown that clusters of cavities parallel with (0001) are often associated with large faulted c-component loops and by implication with a high local impurity concentration. The existence of the cavities and c-component loops in the same volume has been associated with the local solute/precipitate concentration (this work) but is also found in some cases to be dependent on the proximity of grain boundaries [5]. This may be because of the presence of impurities at grain boundaries but is also likely to be because of the absence of interstitial point defects. Similar observations regarding the role of grain boundaries as preferential interstitial sinks have been made by Horsewell and Singh on neutron irradiated aluminum [ 371. Apart from the results of Herring et al. 1191and the
M. Grifjiths et al. / Grain boundary sinks,in neutron-irradiated
Zr
and Zr-alloys
Fig. 8. Cavities (arrowed) associated with precipitates or precipitate clusters. (a) Sponge Zr irradiated at 710 K to a fluence of 4.5 x 10z5 n me2 (E > 1 MeV). Note the stacking-fault (SF). (b) Zircaloy-2 irradiated at 675 K to a fluence of 6 X 1O25 n me2 (E > 1 MeV). Stacking-faults (not visible at this orientation) are also present in the field of view.
present work the only other report of grain boundary influence on microstructure was given by Jostsons et al. [12]. They reported a lower vacancy loop concentration in one sample of Zr which was attributed to the presence of low angle subgrain boundaries. This suggests that the sub-grain boundaries in that case were net vacancy sinks and is inconsistent with the present results. It is also inconsistent with other observations of increased cavity formation close to sub-grain boundaries [S] implying that sub-grain boundaries act as net interstitial sinks. The reason for this is not clear. One possible explanation is that the sub-grain boundaries referred to in ref. [12] may have had a different orientation to those referred to here (parallel with (lOi0) and
point defects, i.e. they may be diffusing faster than vacancies along [OOOl] as well as (1120) directions. The observation of cavities along the particular boundary shown in fig. 7(a) may be related to the presence of solutes or precipitates (which are apparent in the micrograph), or to cavity nucleation in a layer parallel with (0001) which is coincidental and not related to the adjacent grain boundary. It is also possible that the anisotropy of diffusion is affected by the increased solute concentration in the matrix during irradiation. There is some indication [38] that whereas there is a marked anisotropy in the diffusion of some species (either self-interstitials or interstitial Fe) during irradiation of Zircaloy-2 at about 580 K, this is not very apparent at 665 K and may even be reversed.
4.3.2. Orientation dependence In the absence of any impurity effects, the one common feature of the denuded zones close to grain boundaries is the dependence of width on the orientation of the grain boundary. This dependence is such that grain boundaries which are parallel with closepacked (1120) directions are the least efficient sinks for interstitial point defects. This is consistent with and supports the concept of anisotropic interstitial diffusion .[.17]. This criterion does not restrict cavity formation at unfavourably oriented grain .boundaries in certain circumstances, illustrated in fig. 7(a) for Zircaloy-2 irradiated at 665 K, because grain boundaries parallel with
4.3.3. Application to irradiation growth In deriving models to explain most of the growth data for polycrystalline and single crystal Zr [2,3], one can assume that grain boundaries are either net interstitial or vacancy sinks. The results of Jostsons et al. [12] indicate that during neutron irradiation at 673 K certain sub-grain boundaries in annealed Zr may be net vacancy sinks. The treatment of grain boundaries as net vacancy sinks in annealed materials not containing c-component vacancy loops requires that most of the (a)-type dislocation loops are interstitial in nature. Treating the grain boundaries as net vacancy sinks is mostly applicable to cold-worked materials if (a)-type network dislocations are considered as net interstitial sinks. The growth of
{ioil).
(0001)
may
still
be considered
as sinks
for interstitial
A4. Griffhs
et al. / Grain boundq
cold-worked materials requiring a net flux of interstitials to (a)-type dislocations and vacancies to c-component dislocations or grain boundaries can then best be explained in terms of anisotropic diffusion [17]. For annealed materials, treating the grain boundaries as interstitial sinks requires that growth is also explained in terms of anisotropic diffusion. However, grain shape effects (with more interstitials impinging on grain boundaries perpendicular to the direction of smallest aspect ratio) and the formation of c-component vacancy dislocation loops are important and may dominate in certain circumstances. One explanation of the single crystal data relies on the formation of sub-microscopic vacancy c-component loops [l] and, further to this, it has been proposed that such loops (formed by cascade collapse) provide a fixed contribution to the irradiation growth [39]. Whereas anisotropic diffusion appears to be a factor which will affect irradiation growth, it cannot account for the higher growth experienced by annealed polycrystalline material at 353 K compared with 553 K [3]. A combination of mechanisms can best explain the observations with the assumption that there are more interstitial (a)-type loops following irradiation at 353 K compared with 553 K [3], although there is no experimental data to support this. Long term growth after the dislocation structure has saturated may best be explained by anisotropic diffusion and will be inversely dependent on grain size. Irradiation growth will be further complicated by the development of intergranular stresses induced by the growth of individual grains. The present observations indicate that in certain circumstances, for annealed materials at least, grain boundaries can be considered as dominant interstitial sinks in the overall microstructure. This may be applicable to anomalous growth observed in new ZrNb pressure tube materials [40,41].
4.4. Summary
of results
(1) There is selective denuding
(2)
of interstitial point defects close to grain and sub-grain boundaries in neutron irradiated Zr and Zr-alloys at temperatures 1675 K. This is manifested by the appearance of cavities or the absence of interstitial dislocation loops. The width of the denuded zone is smallest for boundaries parallel with one or more (11~0) directions implying that diffusion of interstitial point defects is most rapid along (1120) directions.
sinks in neutron-irradiated Zr and Zr-alloys
415
5. Conclusions There are two main implications of the results presented here with respect to in-reactor deformation: (1) Grain boundaries can be net sinks for interstitials point defects. In annealed polycrystalline or single crystal material, unless there are more interstitial compared with vacancy (a)-type loops or unless there are vacancy c-component loops, irradiation growth can only be the result of the grain shape or anisotropic diffusion or both. (2) The orientation dependence for the width of the zone denuded of interstitial point defects, but which may contain cavities and vacancy loops, is consistent with and supports the concept of interstitial anisotropic diffusion in the basal plane.
Acknowledgements The authors would like to thank C.H. Woo of WNRE and B.A. Cheadle of CRNL for useful comments on the manuscript.
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(321 [33] [34] [35] [36] [37] [38] [39]
[40] [41]
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