Scanuk: A collaborative programme to develop new zirconium cladding alloys

Journal of Nuclear Materials 66 (1977) 163-186 Q North-Holland Publishing Company

SCANUK:ACOLLABORATIVEPROGRAMMETODEVELOPNEWZfRCONlUMCLADDINGALLOYS c. TYZACK l, P. HURST ’ , G.F. SLATTERY ‘, F.W. TROWSE 2,A.GARLICK 3,R.SUMERLING 3, A.STUTTARD 3, K. VIDEM 4, L. LUNDE 4, M. WARREN ‘, E. TOLKSDORF ‘, P. TARKPEA 6, J. FORSTEN 7 ’ UKAEA, REML, Risley, Warrington, England 2 UKAEA, RFL, Springfields, Salwick, Preston, England 3 UKAEA, RDL, Windscafe, Cumberland, England 4 Institutt for Atomenergi, Kjeller, Norway ’ Atomic Energy Commission, Rise, Roskilde, repark ’ Aktiebolaget Atomenergi, Studsvik, Sweden ’ Valtion Teknillinen Tutkimuskeskus, Helsinki, Finhd

Received 6 July 1976 The primary aim of the programme was to develop alloys with better performance than Zircaloy-2 under BWR/PWR conditions and specifically with improved resistance to short-term high temperature transients. Secondly the alloys were to be capable of full-term reactor service over a wider temperature range than usual (up to 450°C). For the fist objective a Zr-1 wt% Nb alloy was selected, and for the second, alloys were composed of small amounts of chromium and/or molybdenum added to a base composition of 0.5 or 1 wt% niobium in zirconium. This paper describes the test programme and results obtained on the physical metallurgy, mechanical properties and corrosion resistance of the alloys both before and after irradiation. Although the requirement for cladding to operate at elevated temperatures is no longer of prime importance, the development work has demonstrated that with some further optimisation some of the alloys might present a viable alternative to Zircaloy-2 for in-reactor operation at * 300°C in oxygenated coolants. Especially with regard to nodular oxidation resistance these alloys based on modest additions of niobium to zirconium tend to be better than Zircaloy-2 but their performance does not consistently approach that of Zr-2.5% Nb. L’objectif primaire du programme etait de developper des alliages de meilleures performances que le Zircaloy-2 dans les conditions de fonctionnement des rhacteurs BWR/PWR et en particulier possedant une &stance accrue aux elevations transitoires de temperature de courte duree. En second lieu, les alliages devaient Ctre capables dun fonctionnement en reacteur a plein temps sur un intervalle de temperatures plus large qu’habituellement (jusqu’ci 450°C). Pour le premier objectif un allige Zr-1% Nb (en poids) a 6th choisi, et pour le second les alliages &talent constituees de petites quantites de chrome et/au de molybdkne ajoutees $ un alliage de base Zr-Nb ii 0,5 ou 1% en poids de Nb. Cet article decrit le programme d’essai et passe en revue les resultats obtenus sur la metallurgic physique, les proprietes mecaniques et la resistance a la corrosion des alliages a la fois avant et apres irradiation. Bien que le fonctionnement du gainage $ temperature dlevie ne soit plus une exigence de premiere importance, les etudes en voie de dkveloppement ont demontrk que quelques-uns des alliages, avec une certalne optimi~tion ulterieure, pourraient p&enter une solution de rechange viable vis-&is du Z&oIoy-2 dans le fonctionnement en reacteurs i 300°C dans des fluides de refroidissement oxyg~n~s. En particulier, du point de vue de la resistance i l’oxydation nodulaire, ces alliages bases sur des additions modestes de niobium au zirconium tendent i &ire mcilleurs que le Zircaloy-2 mais leurs performances ne se rapprochent pas sensiblement de celles de l’alhage Zr-25% Nb. Das Hauptziel des Programms besteht in der Entwicklung von Legierungen mit einem besseren Verhalten als das von Zry-2 unter SWR- und DWR-Bedingungen, insbesondere mit einem hliheren Widerstand gegentiber kurzen Hochtemperaturtransienten. Zweitens sollen die Legierungen im Dauerbetrieb im Reaktor in einem breiteren Temperaturbereich als tiblich (bis zu 45O’C) einsatzfahig sein. Fiir das erste Ziel wurde eine Zr-I % Nb-Leg&rung ausgewahlt, ftir das zweite wurden Legierungen hergestellt, denen geringe Mengen Cr undloder MO zur ~r~lndzu~rnrnen~tzung von 0,s oder 1% Nb in Zr zugegeben wurden. In dieser Arbeit wird das Untersuchungsprogramm beschrieben und tiber die Ergebnisse auf den Gebieten der Metallkunde, der mechanischen Eigenschaften und des Korrosionswiderstands vor und nach der Bestrahlung berichtet. Obwohl die Anforderungen an die HBlle fiir einen Betrieb bei hoheren Temperaturen nicht mehr von vorrangiger Bedeutung sind, haben die Entwicklungsarbeiten gezeigt, dass einige der Legierungen bei weiterer Optimierung eine lebensfiihige Alternative zum Zry-2 unter Reaktorbetrieb bei etwa 300°C in oxidierenden Ktihlmitteln darstellen kiinnten. Insbesondere im Hinblick auf den Widerstand gegentiber knotenfiirmiger Oxidation neigen diese Legierungen, die auf geringen Nb-Zuschliigen zum Zr beruhen, zu einer Verbesserung gegentiber Zry-2, aber ihr Verhalten kommt dem von Zr-2,5% Nb keineswegs nahe. 163

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Tyzncket al. / SCANUK

1. Introduction About five years ago materials scientists from the UK and Scandinavia held discussions to review the possibility of developing improved zirconium alloys for service as cladding in water reactors. The intentions were twofold: (a) to specify an alloy that offered an improvement over Zircaloy-2 at operating temperatures x300°C and which, if possible, showed a better resistance to short-term high-temperature transients, and (b) to develop an alloy to withstand full-term reactor exposure at temperatures up to 45O”C, with tolerance for short term transients to even higher temperatures. On the basis of this review several alloy compositions were selected for evaluation. Now after five years of collaborative study at a fairly modest level of effort on corrosion resistance, physical metallurgy and mechanical properties, it is of Interest to assess the achievements in relation to the original aims and to other criteria that have only more recently assumed importance in cladding behaviour .

2. Experimental 2.1. Selection of alloying additions It was expected that improvements in corrosion resistance and mechanical prOperties would be obtained by alloying zirconium with elements selected from Group VA and VI A, and possibly Group VIII A, of the Periodic Table. Another alloying element of proven worth in the context of 300°C fuel element operition is tin. It is beneficial as a solid solution strengthener and partial age hardener, is known to be effective ln counteracting adverse effects of nitrogen in aqueous corrosion out-of-reactor at 300°C and may improve oxide film adherence. Tin appears to maintain a mildly beneficial influence on corrosion up to =350°C, but in 450°C steam and at higher temperatures, particularly at high pressures, its presence becomes significantly detrimental. The general attractiveness of Group VA and VI A elements as minor alloying additions is to be expected for several reasons. Application of Wagner-Hauffe ideas suggests that small additions of such elements might be beneficial from a corrosion viewpoint, if they be-

come incorporated in the zirconia film, assuming that the oxide behaves as an anion deficient (n-type) semiconductor. On the other hand, experiments in the UK [l] showed the aqueous corrosion rate of Zr-2.5% Nb under oxygenated conditions out-of-reactor was approximately proportional to the square-root of oxygen pressure, suggesting that the oxide had ceased to be deficient in anions, and that oxygen ions were present as interstitials over a large proportion of the film thickness, leading to p-type properties; a small layer of n-type oxide being present adjacent to the metal/oxide interface. However, this sensitivity of corrosion rate to oxygen content of the water disappeared when Zr-2.5% Nb was exposed in 310°C pressurised water under high neutron fluxes and the corrosion rates were lower than those observed out-of-reactor for comparable oxygen levels. In short, the in-reactor behaviour of the alloy was similar to that of Zircaloy-2 out-of-reactor, assumed to be typically n-type, while, in contrast, the oxidation rate of Zircaloy-2 inreactor was sensitive to the oxygen level in the water, the oxide apparently exhibiting p-type behaviour. Whatever the detailed explanation of this behaviour, the observed decreased corrosion rate in-reactor suggested Zr-Nb alloys should be suitable for further development. However, there were already indications that the whole process might be very finely balanced because Zr-2.5% Nb cladding exposed in the Halden reactor at 240°C suffered an increased corrosion rate. 2.2. Physical metallurgy and strength The binary alloy systems of zirconium with Group VA, VI A and VIII A elements are either of the form in which the p phase extends over the whole composition range with no intermetallic phases (e.g. Nb, Ta) or of the eutectic type including intermetallic phases (e.g. V, Cr, MO, W, Fe, Co, Ni, Cu). All these binary alloys exhibit a eutectoid reaction at low alloy contents and the a/P transformation temperature is decreased with increasing concentration. The solubility of these elements in ol-zirconium is very limited and the likelihood of appreciable solid-solution hardening appeared to be small. There remained, however, the possibility of dispersion strengthening by eutectoid decomposition since both tensile strength and creep resistance can be improved at temperatures

C. Tyzack et al / SCANUK


165

dioxide [2,3] and in steam at 500°C [4,5], but in general behaved poorly at x300°C [6] with extensive spalling. 2.4. Factors affecting the choice of alloy elements a) Iron and chromium Small additions of iron and chromium appear to be beneficial in promoting high temperature oxidation resistance although it is not clear why this is so [7-l 11. A fine dispersion of intermetalhcs gives the best corrosion resistance, while slow cooling from high temperature favours growth of coarse intermetallics with the associated local denudation of the matrix of iron and chromium. If ferric ions are present in the zirconium oxide film it is difficult to see what beneficial effect they can exert, except possibly in improving oxide plasticity. Similarly, it seemed unlikely that chromium would be present in the oxide film in the hexavalent state at low oxygen potential. Both of these elements appeared to be similarly beneficial, or at worst innocuous, when present in small amounts at lower temperatures. b) Molybdenum This was a potential addition for precipitation hardening through the formation of ZrMoa or martensitic hardening by stabilisation of the p phase. Evidence suggested that molybdenum improved oxidation resistance at elevated temperatures, particularly at somewhat higher oxidation potentials (wet CO* at 500°C [2,3]) and it appeared to be associated with low corrosion hydrogen absorption as for niobium. Its value as an alloying addition at ,lower temperatures, however, appeared to be equivocal. c) Niobium Niobium is attractive since it has the lowest parasitic capture cross-section of the Group V and VI A elements. Zr-1 wt% Nb is predominantly a solid solution alloy with a strength in the annealed state perhaps comparable with 20% cold-worked Zircaloy-2, and with good ductility even after extensive irradiation. At that stage there was no in-reactor corrosion experience with the 1% Nb alloy tn the West, although the Russians had extensive operational experience with it as fuel cladding and had commented favourably [12-141. Another alloy that has been studied in some detail in the UK [15-181, and in Canada and the US [19-

166

C. Tvzack et al. / SCANUK

231 is Zr-2.5% Nb. This can be used in an annealed cu+/3 (Nb) form OKsolution treated, preferably in the Q + 0 region, quenched to produce cyt 1~’martensite and subsequently annealed below the eutectoid temperature to precipitate the supersaturated niobium content of the martensite. It has been shown that this ageing process contributes the major part of the hardening. It was also clear from work in the UK that corrosion resistance, particularly in oxygenated water, was extremely sensitive to the amount of excess niobium over equilibrium remaining in the partially tempered martensite. For example, the corrosion rate of p-quenched material could be reduced by w 3 orders of magnitude by ageing for prolonged periods at 500°C (241. This sensitivity to heat treatment brings in its train problems with heat-affected zones of welds on fuel pins, and the view was taken that it would be easier to utihse the essentially solid solution alloy provided its corrosion resistance proved adequate. Another attractive feature of zirco~um-niobium alloys is the characteristically low corrosion hydrogen absorption, especially at temperatures up to 300°C. In general it does not exceed 10% of the theoretical, and is even less under conditions where the aqueous coolant is oxidising. Reasonable oxidation resistance is retained at 400°C and even at 500-55’0°C it does not appear to suffer the catastrophic oxidation observed with Zircaloy-2 in high pressure steam, although the precentage of corrosion hydrogen absorbed increased at high temperatures. 2.5. The coilab~rati~e programme As a result of the deliberations outlined above, a collaborative programme of work was initiated beTable 1 Scanuk alloy compositions Al10 y

Nb 0.91 0.93 1.12 0.52 0.49 0.58 0.014 0.100

_~_.__.

All compositions

in weight %

2.6. Frabrieation and heat treatment Material was fabricated in the form of tubing and sheet and some was retained as billets or semi-finished products to enable the effect of fabrication route to be studied. Ingots, 152 mm diameter, were cast and soaked at IOOO-1050°C for one hour after which they were upset to 203 mm diameter at &95O”C, and soaked at IOOO-105O’C for a further hour. The billets were then forged to 140 mm diameter, again at >95O”C, and further’annealed at 10~-1050°C for two hours before water quenching. Chemical and metallographic analyses were made before proceeding on to the appropriate hot or cold rolling processes. In the case of tubing, a 27 kg billet was machined, clad in

Fe

Stl

.___~ 1 2 3 4 5 6 VaIloy Ozhennite 0.5

tween Denmark, Norway, Sweden and the. UK in 1969. This followed a previous Scandinavian toll aboration on zirconium-c~omiunl-iron alloys. In 1971 the membership was extended to include Finland. Melts were made of the six selected alloy compositions and the ingots fabricated to sheet, rod and tubing as described in the following section. The compositions are summarised in table 1. The experimental programme was rationalised between the partners, the main topics investigated at the various laboratories being summarised in table 2. Each partner contributed equally towards the cost of alloy manufacture and information arising from the programme was freely shared. This paper does not set out to deal comprehensively with the work carried out under the collaborative programme but gives some selected examples and is intended to act as an introduction. More detailed papers on specific topics will be published later.

___. .I ---_ 0.026 0.038 0.045 0.036 0.037 0.044 0.101 0.059

cr

Ni

Oxygen

_

-

< 0.01

0.073 0.060 0.060 0.047 0.060 < 0.001 0.091

< 0.01 0.49 0.49 < 0.01 0.32 1.29 < 0.01

___-._-_-_..____

MO < 0.005 < 0.005 < 0.005 0.004 0.280 0.220 0.004 < 0.004

< 0.005 < 0.005 < 0.005 0.005 < 0.005 0.005 0.005 0.052

0.100 0.096 0.126 0.134 0.097 0.125 0.135 0.117

. ..-

C. Tyzacket al. / SCANUK Table 2 Topic

_____-

Out of reactor corrosion Determination of corrosion hydrogen In-reactor corrosion (coupons) a) In-reactor loops. RD3 Risd. Steam and BWR conditions b) Halden BWR. BWR conditions (24O’C). c) SGHWR. CO2 vault gas. Heat treatment: Effect on mechanical properties Effect on hardness and microstructure Mechanical properties. Room temperature and 250 to 400°C ‘Superplastic’ behaviour. 750 to 900°C. Strength and ductility of tubing-ring testing Determination of microstructure Dilatometric investigations Texture measurements on tubing Quench-ageing studies Irradiation of fuel pins in HBWR Irradiation of minipins in DR3. Irradiation of fuel pins in R2 Studsvik Temperature transient tests, a) oxidation b) mechanical properties/oxidation Stress-corrosion Fabrication of minipins for irradiation Post-irradiation examination of minipins Fabrication and irradiation of element in SGHWR Post-irradiation examination of element.

Establishment REML, RFL, AEK, IFA REML

AEK IFA REML RFL AEK AEK VI-T RFL, AE VTT VTT VTT, RFL AEK IFA AEK AE AE RFL IFA IFA, AEK AEK RFL RDL

REML - Risley Engineering and Materials Laboratory, UKAEA. RFL - Reactor Fuel Element Laboratory, Springfields, UKAEA. AEK - Danish Atomic Energy Commission. IFA - Institutt for Atomenergi, Norway. VTT - Valtion Tekmillinen Tutkimuskeskus, Finland. AE - Aktiebologet Atomenergi, Sweden. RDL - Reactor Development Laboratories, Windscale, UKAEA.

copper, soaked at 750°C and extruded. After removal of the cladding by acid pickling the billet was annealed at 675°C and tube-reduced to size with intermediate anneals at 675°C as required. The final tube reduction was aimed to give a 70% reduction of area, with no subsequent anneal. The tubing was finally linished outside and vaqua-blasted on the bore. In the experimental programme the only heat treatment regularly applied was annealing to eliminate cold

167

work. Tests showed that hardness started to level off aiter annealing at 450°C the final softening was achieved at 550°C. In view of this the standard recrystallisation annealing treatment was at 600°C for 4h. 2.7 Microstructure and dilatometry of alloys The purpose of the microstructural investigation was mainly to character& the occurrence and nature of the second phase particles and is presently limited to examination of features which could further the understanding of corrosion behaviour, mechanical property data and heat treatment processes. Precipitates were observed in all recrystallised material and were particularly numerous in alloys 3,4,5 and 6, i.e. those containing chromium. More uneven distributions were present in the other alloys and there was considerable variation in precipitate size. Increasing the niobium content from 0.5% to 1% led to coarser precipitation in alloys containing 0.5% chromium. The crystal structure of the precipitates in alloys 1 and 2 is niobium with a slightly increased lattice parameter suggesting the presence of some zirconium in solid solution. Some information on precipitation processes has been obtained in a dilatometric investigation on alloys 1 and 4, the tests being done at heating and cooling rates of 2 and 5°C per min on recrystallised bar material in an argon atmosphe e. On the dilation curves for alloys 1 and 4 (fig. 1) \the points marked A-F are inflection points indicating changes in temperature coefficient of expansion. Alloy 1 consists of an cu-zirconium matrix and precipitates assumed to be rich in bee niobium. In alloy 4, which contains $ niobium as well as chromium, the precipitates are obviously rich in chromium because of its very limited solubility in a-zirconium. The structures in this case might be expected to be hexagonal or fee. Point A in fig. l(a) was identified with the commencement of dissolution of 0 (Nb) precipitates. As the bee lattice of /!I(Nb) is less closely packed than that of OLzirconium there is a relative decrease in dilation between A and B. Similar dissolutidn of precipitates takes place in alloy 4 (fig. l(b)) but to a minor extent. Supporting optical and transmission electron microscopy (TEM) studies indicated that point C coincides with passage into the a! + fl phase field but the occurrence of a maximum point in the curve at D cannot

C Tyzack et ai. / SCANUK

168

lb)

‘2 II

SC4NUK 4

F

4 UNITS

d UNKS

3

Fig. 1. DiIatometer curves for Scanuk alloys at a heating and cooling rate of Z”C/min.

be explained from the phase diagram alone. The dilatational drop in the region D-E is greater at increased heating rates suggesting it may be related to factors other than simple phase transformation. It is assumed to be connected with so-called “super-plastic” behaviour in the (Y+ fi range where very high ductility, low strength and high strain rate sensitivity have been observed and attributed to the one-baled duplex struo ture formed in the a! t /3phase region. The net dilational change during a temperature cycle is very dependent upon heating and cooling rates. Ageing of cold-worked and p quenched samples of alloys I,3 and 4 in the range 600-l 040°C followed by air cooling or water quenching led to a variety of precipitates and these were examined by X-ray analysis. For example in alloy 4, annealed at 8SO°C, niobium precipitates were identified but some unknown phases were also present. Samples annealed at 1040°C and cooled consisted of martensitic (Y’as well as unidentified complex phase(s). TEM studies on deformed Scanuk alloys revealed dislocation networks and tangles as well as twins, observations comparable to those for Zircaloy-2.

that the (~02) pole lay within about 25” of the sheet normal. The texture of Scanuk alloy tubes was similar to that of commercial Zircaloy-2 tubes, the (0002) peak having two distinct intensity maxima located in the transverse plane (fig. 2). The intensity distribution was generally symmetrical about the axial direction and meas~ements near the tube bore indicated that the texture was more pronounced there than at the outer surface. The (0002) pole figure was not significantly different for tubing in the 6OO”C,4h recrystallised condition.

2.8. Texrure measurements X-ray measurements of texture on cold rolled sheets of the Scanuk alloys using a Siemens goniometer showed

Fig. 2. Texture of Scanuk alloy 1 tube.

C. Tyzack et al. J SCANUK

2.9. Mechanical properties The mechanical properties of the Scanuk alloys can be evaluated against the desiderata of improvements over Zircaloy-2 at cladding temperatures +280°C, with the possibility of full-term operation at temperatures up to 450°C and a greater resistance than Zircaloy-2 to short-term high temperature transients. The temperature dependence of mechanical properties was determined for fully recrystallised Scanuk alloys and compared with Zircaloy data. Tensile tests were made at two estab~~ents: (a) on flat s~cimens, gauge dimensions 15 mm X 4 mm X 0.75 mm, machined from the rolling direction of 60% cold-worked sheet and subsequently vacuum annealed for 4h at 600°C; (b) on ring specimens from 16 mm o.d. tubing with a gauge length 25% of the circumference and a comparable strain rate (=I .O X 10m3 set-‘) to that for the sheet specimens. Tests of type (a) allowed the yield, ultimate and fracture stresses to be determined along with reduction of area at fracture for temperatures in the range 25-400°C and typical values obtained for Scanuk alloys 1,4 and 6 are compared with Zircaloy-2 data in fig. 3. The UTS values obtained at 25°C on annealed sheet and tube at the two establishments are in good agreement, the maximum observed difference being 4% (table 3). In general, the mechanical properties of the Scanuk alloys compared favourably with those of Zircaloy-2 and in some respects represented an improvement. Thus Table 3 Comparison of UTS determined on annealed sheet and tube at different establishments ..__ -_I_ ____ Alloy

1 2 3 4 5 6 Zircaloy-2

UTS at 25’C (kg/mm2) AEK results from tensile tests on sheet (longitudinal)

RFL results from ring tensile tests on tube (transverse)

47 45 50 47 47 48 50

49 46.3 48 45.5 48 47 52

169

values determined for the fracture stress of Zircaloy showed much more scatter than those for the Scanuk alloys, nevertheless the fracture strengths of alloys 3, 4 and 6 were 96 to 145 MN/m2 greater. The Scanuk alloys had higher strengths than Zircaloy-2 in the temperature range 250-400°C and the decrease in strength with increasing temperature was much less severe, Alloys 3 and 4 were best in this respect in the temperature range 300-400°C their strength exceeded that of Zircaloy-2 by some l&25%. Various degrees of load discontinuity have been observed during yielding, Zircaloy-2 and alloys 1 and 2 e~ibiting load drops followed by Luders band extension for the whole temperature range studied; this might be associated with high nitrogen content. Load drops were observed only occasionally for alloys 3,4 and 6 and at 400°C no major yield discontinuities occurred, Values of the strain hardening exponent n, defined as d log u/d log E in the strain range 2 to 10% were determined at tem~ratures in the range 25-4OO’C (Table #4).Scanuk alloys exhibited a peak in strainhardening exponent whereas that for Zircaloy-2 showed a progressive increase with temperature. log (true stress) versus log (true strain) curves showed very little difference at 25°C between Scanuk alloys and Zircaloy, but at 400°C that for Zircaloy lay signi~cantly below those of the precipitation hardened alloys 3,4 and 6. These plots also gave some indication of the effect of yielding on subsequent strain hardening behaviour: for alloys 3,4 and 6 the strain associated with yield loses its influence beyond 0.6% (log e e -2.25) whereas in Zircaloy-2 the effect persists to a strain of 1.2% (log f =-I .9). In all aloys this was relatively inde~ndent of tem~rature. The ability of the alloys to strain harden is well demonstrated by the stress increment arising from a strain increase from 1 to 5% (table 4). Those Scanuk alloys with little precipitate (alloys 1 and 2) behaved comparably with Zircaloy-2, but alloys 3,4,5 and 6 had considerably higher strain hardening rates. It is noteworthy that alloy 5 which had the worst precipitate distribution also showed the least improvement over Zircaloy-2. Reduction of area at fracture was similar for Zircaloy-2 and Scanuk alloys. In another series of ring expansion tests the ductility of Scanuk alloy tubing was determined in the transverse direction at room temperature and 3OO”C,from measure-

C. Tyzack et al. / f$(XWX

170

SCANUK

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100

200

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ZIRCALOY

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TEST TE~vIFEF~ATLJRE<~C>

0

100

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Fig. 3. High temperature mechanical properties of fkanuk alloys and zircaloy in the recrystalised condition (6 = 1.1 X 10m3 per set). ment of specimen o.d. before and after testing. Strain was applied uniformly and fracture occurred in the range 38 to 5 1% at room tem~rature and 60 to 68% at 300°C.

lowed by alloy 1 and Zircaloy-2. In the Scanuk alloys, hydride precipitates were small and circumferentially orientated, but for Zircaloy-2 Fee values were about 0.3, compared to
Scanuk alloys 1 and 4 and Zircaloy-2 were examined after reaction with hydrogen at 6OO”C,alloy 4 being the most resistant to hydrogen absorption fol-

It might be argued that aqueous corrosion data obtamed out-of-reactor under degassed ~nditions are irrelevant to reactor exposure in basically an oxygenated

C. Tyzack et al. / SCANUK Table 4 Strain-hardening characteristics of Zircaloy-2 and the Scanuk alloys

171

--

----

Ahoy

Test temp. eo

1

2

3

4

5

6 --

21-2

-.Strain-hardeningexponent in range 2-I 0% strain 25 250 300 350 400

* * * 0.178 8;

0.151 * 0.204 * *

0.144 0.156 0.178 0.178 0.160

0.149 0.182 0.180 0.164 0.160

0.163 0.208 * 0.212 0.206

0.150 0.181 0.191 0.180 0.171

0.122 0.153 0.171 0.180 0.180

491.5 289.4 260.9 245.3 232.5

456.2 260.0 235.4 234.5 214.8

445.4 250.2 215.8 207.0 190.3

467.9 269.8 227.6 220.7 214.8

489.5 245.3 180.5 190.3 178.5

96.1 60.8 66.7 65.1 59.8

89.3 64.7 55.9 55.9 50.0

85.3 55.9 52.0 49.0 47.1

93.2 68.7 58.9 51.9 54.0

64.7 40.2 35.3 37.3 36.3

True stressat a strain of 5% (MN/&)

25 2.50 300 350 400

446.4

436.5

252.1 225.6 204.0 192.3

246.2 205 .o 199.1 189.3

True stress increase from 1% to 5% strain (MNhn2) 25 250 300 350 400

-

58.9 41.2 39.2 43.2 36.3

12.6 47.1 40.2 41.2 41.1

The table shows the effect of test temperature on 1) the strain hardening exponent ‘n’ defined as log true stress/log true piastic strain in the range 2-10% strain; 2) the true stress at a true plastic strain of 5%; 3) the stress increase arising from a strain increase from 1% to 5% as a measure of the strain-hardening rate. * log/log plot not obtained.

coolant. While this is probably correct at reactor operating tem~rat~es (around 300%‘) it is not the case at &lOO*C where there is no observed effect of radiation on corrosion rates and out-of-reactor data should be valid. Some work of this nature is desirable in order to establish a datum for comparison with the out-ofreactor behaviour of well established materials such as Zircaloy-2 and 4 on which there is a great deal of corrosion information. It is interesting therefore to compare the corrosion resistance of the Scanuk alloys in degassed water at 290°C and in steam at > 400°C with that of typical Zircaloy-2 sheet material. The pre-transition weight gains at 29O”C, 7.39 MN/m* under degassed conditions for sheet material in static tests are generally within =.50% of those for Zircaloy-2, which form a lower bound (fig. 4). Investi-

gators agree that alloy 4 generally shows least weight gain of the Scanuk alloys under these conditions. Values of pretransition rate constants (table 5) have been calculated on the assumption that the relationship (Ot - we)3 = kt applies. The rate constant k1j3 generally falls within a factor of 2 of that observed for Zircaloy-2. Static autoclave data now extend to around 21 000 h and there are indications that the Scanuk alloys are either in transition or starting to be sq, whereas Zircaloy-2 has not yet gone into transition. Exposure under dynamic conditions at 290°C appears to promote spalling on most of the Scanuk alloys, with the exception of alloy 4 which shows behaviour comparable to Zircaloy-2, retaining an adherent film. Corrosion hydrogen absorption for sheet specimens

C. Tyzack et al. / SCANUK

172

c

*

+ KANUK

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4

0.50 A m l

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SHEET C.W. Pals5 H.T. 24 h

.

500%

I

.a...1 1000

loo

TUBE

”

T. )

,

IO 000

HOURS

Fig. 4. Corrosion data at 290°C (7.39 MN/m’) in degassed water.

Table 5 Calculation of oxidation rate constants at 290°C and 400°C Alloy

290°c k1/3 (mddm2

400°C kz

zg/dm2)

(mg/dm2day)

day 113) 1 2 3 4 5 6 Zr-2 sheet Zr-2 sheet Zr-2.5 Nb sheet

3.58 3.85 4.25 2.57 3.27 2.77 1.93 1.99

ND Zr-2.5 Nb pressure tube (a) 4.27 Zr-2.5 Nb pressure tube(b) 4.5

ND = Not Determined

1.42 0.60 0.02 3.40 2.49 5.59 7..05 7.54 ND

0.67 0.58 0.99 0.56 0.67 0.66 0.53 0.80 ND

1.07

ND

-0.84

ND

exposed in 290°C degassed water up to 8000 h is summarked in table 6. Percentage absorption was
173

C. Tyzack et al. f SCANUK Table 6 Corrosion hydrogen absorption for sheet specimens exposed at 29O”C, 7.39 MN/m2 in degassed water Exposure 8004 h

Exposure 2856 h mg H2/dm2

mg Oz/dm*

%H2

mg 02/dm2

mg H2/dm2

24.9 26.2 28.0 20.2 24.5 24.5 21.7 23.4 33.4 28.7

0.15 ND 0.20 0.15 0.25 0.22 1.05 1.92 0.28 0.19

%H2 ~I_ 4.7 ND 5.7 5.9 8.9 7.0 38.8 65.8 6.6 5.4

0.52

14.8

I_._1 2 3 4 5 6 Zr-2 sheet 11 II zr-2.5 Nb sheet n-2.5 Nb pressure tube *

0.14 ND 0.23 0.20 0.28 0.29 0.88 1.70 0.38 0.30

17.5 18.1 19.3 14.8 19.4 19.9 16.9 19.4 21.5 21.7

6.3 ND 9.7 10.7 11.7 11.8 41.9 70.0 14.0 11.0

Exposure 7836 h

Exposure 2688 h 2~2.5 Nb pressure tube **

0,62

20.9

23.6

27.9

* Cold worked ** Heat treated: water quenched from 88O”C, tempered at 500°C for 24h.

f

k too

-0

I

0

Table 7 Corrosion hydrogen absorption for sheet specimens exposed at 400°C, 6.85 ~N/m2 in steam

2

SCANUK

A

‘8

30

*

4

D

-

50

”

b

A I

(Y

SCANUK

Lr- 2 SHEET tr - 2f 70 NB SHEET * C.W. PRESS

Alloy

Exposure 12 900 h mg 02/dm2

mg H2/dm2

% Hz

382.2 350.8 354.0 307.4 364.2 375.8 367.9 539.5 858.2

7.81- 9.79 8.22- 9.54 12.55-14.39 11.34-12.07 10.68-12.07 12.86-14.76 27.72-28.06 36.93-42.36 20.12-24.79

16.3-20.5 18.7-21.8 18.8-21.6 29.5-31.4 23.5-26.5 27.4-31.4 60.3-61.0 54.8-62.8 18.8-23.1

TUBE

”

E”

I

0

$

I

P

400

6

Exposure 12 828 h

5’ 100

0

AhI

1 2 3 4 5 6 Zr-2 sheet Zr-2 sheet Zr-2.5 Nb sheet Zr-2.5 Nb pressure tube *

‘

500

li

1000

Zr-2.5 Nb pressure tube **

1044.7

23.86-25.03

18.3-19.2

HOURS

Fig. 5. Corrosion data at 290°C (8.90 MN/m21 in water containing 7 ppm oxygen.

* Cold worked. ** Heat treated: water quenched from 88O”C, tempered at SOO”Cfor 24 h.

174

C. Tyzack et al. /SCjlNUK

linear rate constants (table 5) calculated by regression analysis for the period 1000-l 3 000 h were comparable with that for Zircaby-2 and in one cast marginally better (alloy 4). Hydrogen absorption after 13 000 h at 400°C was in the range 16-30% of theoretical (table 7), representing an increase over that observed at 290°C but still less than that for Zircaloy-2 (~60%) under comparable conditions. All the Scanuk alloys show markedly lower oxidation rates than Zircaloy-2 at higher temperatures (50055O’C). The sheet material tested in 5OO”C,6.85 MN/m2 steam for 1000 h had weight gains in the range 9001000 mg/dm2 for Zr-2.5% Nb, 600-800 mg/dm2 for Scanuk alloys, while the Zircaloy-2 specimen had disintegrated after 168 h exposure. Hydrogen absorption of the Scanuk alloys was in the range 37 to 57% of theoretical after 1 000 h. 2.12. In-reactor corrosion A part of the collaborative programme was to determine corrosion resistance of the alloys in neutral, oxy-

genated (BWR) environments. Corrosion coupons were irradiated in the Halden BWR in neutral boiling D20 at 240 + 2°C under a neutron flux of 10r2n/cm2 set (>I MeV) for 331 days (3970 MWd). The six Scanuk alloy coupons were cut from cold rolled sheet as were the Zircaloy-2,Zr-2.5% Nb and Ozhennite 0.5 specimens. Valley, a Zr-1% Cr 0.1% Fe alloy developed in the US for improved resistance to high temperature steam, was included for comparison purposes. This was in the form of split and flattened fuel tubing. All specimens were pickled in HF/HNO, before irradiation. After exposure, alloys 3 and 4 had weight gains similar to that for Zircaloy-2, while those for alloys 1 and 2 were slightly higher, comparable with Zr-2.5% Nb. The molybdenum containing alloys 5 and 6 showed a further increment in weight gain but considerably less than that for Ozhennite and Valley which were factors 2 to 3 higher than Zircaloy-2. A second series of tests was made in HBWR on specimens cut from fuel tubing of Scanuk compositions and these were exposed to a neutron flux of 5 X 1012

4 % l

% A

+ 3CANUK

I

0

A

3

V

.

4

5

0

*

6

.

L)

n

A

zr-2

8

Zr-2j?e

XANUK

2

s+f%ET N% SHEET

*

-

C.W. PRESS

l

*

H.T.

(800°E.0.

Fig. 6. Corrosion data at 400°C (6.85 MN/rn2~ in steam.

Uh

TUBE

* SOOW

l

1.)

C. Tyzack et a!. / SCANUK

TUDlNO

SNELTS

150 -

.

.I

100-

-z .

PO-

.

OlHLNNlTE

.

“ALLO”

0 0

“ALLOY 02WCNWITE

/

f 2

80

5

To-

/ /

so / 6e

/

J s

ls

TYPICAL

/

FOR N O*I

60

-

Zr-PiNb

0 am IO ::

.50-

40 :

73-l

IN

R

/ /

I /

:: : :,-2i

/

/

nr /

::‘-2

/ /

RANG2

Zr-2

/ .

/

/

/

:/I

, 200

200 EaPOsUR2

/

I 400

TIME ( OIIL

SO0

)

Fig. 7. Corrosion behaviour of Scanuk alloys compared to other alloys during exposure in Halden boiling water reactor.

n/cm* set (>l MeV) for 255 days (3068 MWd) under conditions otherwise similar to those in the first series. The tubular specimens had higher weight gains than the sheet material exposed earlier, and appeared relatively less good in relation to Zircaloy-2. This may be the result of differences in metallurgical condition between the two sets of materials, as the sheets were cold-worked and the tubes were fully recrystallised by heating at 600°C for 4 h. In this test oxidation of the Scanuk alloys 1,2,3 and 4 was = 40% higher than Zircaloy-2 while alloys 5 and 6 it was approximately double that for Zircaloy. The rates for Ozhennite and Valloy were again much higher. The data are summarised in fig. 7. 2.13. Nodular corrosion One of the more significant developments in the field of fuel cladding behaviour over the last few years has been the steady increase in the number of observations and reports of a localised form of corrosion on Zircaloy-2. This is generally known as ‘nodular’ corrosion and has been observed in boiling water envir-

11.5

onments in the presence of radiation where radiolytic oxygen-containing radicals are being produced in the aqueous phase. The term ‘nodule’ is used to describe a local thickening of the oxide film appearing as an eruption of white oxide, generally approximately circular and typically between 0.02 and 0.5 mm diameter. As the nodules multiply and spread, coalescence occurs to give complete coverage of larger areas with white oxide, described by workers in the US as ‘patch-type’ corrosion 1251. This phenomenon was observed on the high burn-up Dresden BWR (ex VBWR) fuel. Earlier reports [26-281 described distinct nodules on the fuel, but after 1200-1800 days only coalesced patches were present, with a maximum thickness of 160 m. It has also been observed on Zircaloy-2 clad fuel in the SGHWR Reactor from quite an early stage, the earliest reported nodules being on an element irradiated for 77 effective full power days (EFPD) which had uniform oxide 2 pm thick with occasional nodules up to 7 m [29]. The first element with extensive nodular corrosion had been exposed for 275 EFPD (5500 MWd/ te U mean irradiation) and this showed features which have since come to be accepted as typical, such as a preponderance of nodules opposite grid positions both on the tube and on the braze heat-affected zones at the spacer-pad positions. Thus nodule formation seems to be associated in some way with disturbances to coolant flow in the vicinity of grids under boiling oxygenated conditions. It is not generally present in PWR conditions when sufficiently high overpressures of hydrogen or deuterium are maintained. It has occasionally been observed [30] where subcooled boiling may have occurred and deuterium overpressures have been low, so that oxygen-bearing radical production has not been fully suppressed. When the Scanuk alloy programme was devised in 1969 the significance of nodular corrosion in boiling oxygenated environments was not appreciated in Europe and the alloys were not designed to counteract it. Inevitably, however, interest has subsequently been expressed in the relative susceptibility of these alloys and Zircaloy-2 to local&d attack of this type, so that to some extent the alloys are being judged against criteria for which they were not designed. The alloys have not so far exhibited nodular corrosion after exposure in Halden BWR but temperatures and fluxes are somewhat low in comparison with larger BWR’s. As will be seen later some nodular corro-

C. Tyzack et al. / SCANUK

176

sion was observed in the mi~ature pin irrad~tions in DR3 at RisQalthough at this stage there is no directly comparable information on Zircaloy-2 pins and the water chemistry regime is not well characterised although it is probably more severe than in power producing BWR’s.Perhaps the most representative information we have at present is that from irradiation of an element in SGHWR, described later in this paper, although once again the exposure is not really typical of BWR conditions as the exit steam quality was high. Nevertheless the results do allow comparisons to be made between Scanuk alloys 4 and 6 and Zircaloy-2 in two heat treatments.

Iodine stress-corrosion susceptibility of Scanuk alloy fuel tubing has been compared with that of Zircaloy-2 in tests under conditions of internal pressurisation with argon, the iodine concentration being

+

SCANUK

4

SCANUK

I : 0.2

--A-*SCMUK

4:

UT5

--k-SCANUK t \

I : UTS

\ \ \\

4:

0.2

4

SCANUK

3:

UTS

0

SC#&X

3 : 0.2

I

ZIRCALOY-

0

ZlRCMLlY

2:

YIELD

STRENGTH

YIELD

STRENGTH

YIELD

STREffil’H

UTS

- 2 : 0.2

YIELD

Sl%NGV

1 rng~~rn3at the test temperature of 340°C. The tubing was fully recrystallised. For tubes without a starter crack the stress had to exceed a certain threshold value in order to initiate strew-corro~on cracking. This was a284 MN/m’ for the Scanuk alloys, and the diametral strain at rupture was in the range 3 to 7%, also comparable with that for Zircaloy-2 annealed at the same temperature (6OO’Cfor 4 h), so it would ap pear that Scanuk alloys do not have any advantage over Ziicaloy in this respect. 2.15. Behaviour in temperature transients A study was made into effects of short-term temperature transients on the subsequent ductility and corrosion resistance of Scanuk alloys,,Zircaloy-2 and Valloy at normal reactor operational temperatures. The specimens were heated in steam from 350°C to temperatures in the range 700-l 100°C within a period of 3 set, held there for 7 set and then cooled to 350°C in 12-l 5 sec. Tests were of 1,5 or 10 cycles duration. Weight gain during the transients was similar for each material. The hydrogen absorption was small

0

SCANUK

I

0

SCAJWK

3

A

SCANUK

4

n

21-2

\ A “1 \

A--6.A

Pig. 8; Ultimate tensile strength and 0.2% yield strength of Scanuk alloys and Zircaloy-2 as a function of temperature.

:

\ \

Fig. 9. Dependence of eiongation upon temperature in tensile tests.

C. Tyzack et aL / SCANUK

for Valloy and moderate for the other valloys except at > 1000°C for Zircaloy-2, and > 1100°C for alloy 1 where there were indications of increased hydrogen absorption. The ductility, measured in ring tests at room temperature and 35O”C, decreased slowly with increasing maximum temperature in the transient up to about 9OO”C,where there was a pronounced drop. Oxygen penetration into the metal and possibly accelerated hydrogen absorption are suggested as the main causes of embrittlement. A comparison between the alloys with regard to embrittlement is complicated because they differed considerably in initial ductility. In order of post-test ductility, the alloys can be tentatively ranked as follows: 1 and 2 > 3 and 4 > 5 and 6. In subsequent autoclave tests there was further acceleration of the weight gain and this memory effect, the extent of which depended on the particular transient treatment, decreased after 3-4 weeks at 35O’C. The accelerated oxidation was accompanied by a rather rapid absorption of hydrogen, In support of the temperature transient studies, hot tensile tests on the Scanuk alloys 1,3 and 4 have been performed in the range 750 to 900°C (corresponding to the (Y+ fl region) under a vacuum of < 10m4 torr. Flow stress was generally independent of total strain, only Zircaloy-2 exhibiting any strain hardening at 75O*C. Strength was low and ducti~ty high for all the alloys (figs. 8 and 9). Reduction of area at fracture exceeded 90% and total elongation peaked above 200% for each alloy between 850 and 900°C so they are superplastic in this temperature range. This high ductility is associated with the 01t 0 structure and corresponds to the temperature at which the dilational drop was observed. The high temperature properties of the Scam& alloys were also compared with Zircaloy-2 in isothermal internal pressurisation tests with a steam environment at the tube exterior. At 75O’C and 1050°C their failure strains were similar to that for Zircaloy probably because all alloys were hardened by oxygen absorption. At 860°C the Scanuk alloys exhibited larger strains than Zircaloy-2 in accordance with the tensile test data in which the largest strains for Zircaloy-2 were observed at higher temperatures (9OO*C). 2.16. Fuel pin irradiations Before a new zirconium alloy can be considered as an alternative cladding to Zircaloy-2, the fuel manu-

171

facturers and utilities would require a comprehensive demonstration of improved performance based on Irradiation of fuel pins and eventually fuel elements. In-reactor tests on fuel are therefore considered to be an ~~rtant part of a project seeking to develop improved materials. Several fuel pin irradiations have been performed in various reactor facilities accessible to the participants in the collaboration. These included irradiations of minipins clad in Scanuk alloys in DR3 at Ris$, irradiations of pins in I-Ialden BWR and R2 at Studsvik, and a number of pins in an SGIiWR element exposed at high exit steam quality (28%). Details are given in table 8, The six Scanuk alloy-clad minipins were irradiated for 120 days under neutral sub-cooled boiling conditions in DR3. Sub-cooled boiling along the length of the pin was arranged without vertical convection by siting the pm within a vaned tube which guided the bubbles away from it almost IateralIy. The cladd~g temperature was = 287-29O”C, approximately 1O’C above the boiling point corresponding to the saturation pressure. The rig operated essentially under conditions of natural circulation, without gas separation. The measured oxygen levels were 0.43 and 4.3 ml/kg in the water and steam respectively which suggests an env~onment considerably richer in oxygen than for a typical BWR and SGHWR. The small pins were 170 mm overall length, 14.3 mm O.D., 12.7 mm I.D. and were irradiated to 8000 MWd/teU at an average heat rating of a:555 W/cm. The objeyt of the experiment was to study the effects of irradiation on the corrosion behaviour, hydrogen absorption and mechanical properties of the Scam& alloys. Unfortunately it was not possible to include comparable controls clad in Zircaloy-2, although its behaviour under similar conditions can be derived from other experiments. Further mini-pm experiments are in progress which will incorporate Zircaloy-2 controls. Visual inspection of the mini-pins after irradiation indicated nothing unusual except-for alloy 5 on which extensive oxide spalling had occurred. Met~o~ap~c examination of sections showed that on alloy 1 the external oxide was < 1.5 m thick on one side of the pin but 17 /.un thick and discontinuous on the other. Oxide on alloy 2 was of irregular thickness. On alloy 3 oxide was also continuous but on one side there

“SGHWR” 288’C High quality

SGHWR AEE Winfrith

Scanuk 4 & 6* Zircaloy-2 Ozhennite Zr-2.5% Nb Valloy

Zircaloy-2

Scanuk 1 a) pickled b) pickled & autoclaved

2.012% enriched OD 15.94 mm ID 14.6 mm WT 0.67 mm 36 element

1 to 5x10’3

1013

1 to 4x1013

Neutron flux (n/cm2) (>I MeV)

5.0

15.3

9.0

6.8-8.0

Irradiation @fWd/kgU)

* All cladding was in the recrystalhsed condition, with the exception of some Zircaloy-2 pins.

“BWR” 280°C

R2 Studsvik

Zircaloy-2

OD 16.08 mm ID 14.59 mm WT 0.74 mm ID 10.5-10.9 mm OD 12.12-12.58 mm WT 0.81-0.84 mm

8 “half pins”

,,

All alloys annealed

“BWR” 24O’C

HBWR Halden

I,

1.24% enriched . mini pins Length 170 mm OD 14.3 mm ID 12.7 mm WT 0.8 mm

All alloys annealed.

“BWR” 290°c

DR3 Risd

“PWR” 290°C

Pin dimensions

Cladding material

Environment

Reactor

Table 8 Programme of fuel pin irradiations

238 (EFPD)

300

500 (EFPD) Still in reactor

I38

Exposure time (days)

5.50

500

450

600

500-

W/cm peak

Heat flux

b) 50, 130

a) 130

W/cm2

Neutral 0.1 ppm oxygen 26.5-28% quality steam at exit

Neutral 20 ppm oxygen 15% quality steam (steam-water mixing channel entry)

Neutral oxygen

Neutral. Very high oxygen

Coolant characteristics

5

179

C, Tyzack et al. / SCANUK

were several locally thickened regions in the form of nodules, the axial extent of which varied from 50 to 300 Mm, maximum thickness being 29 pm. There was some evidence of fragmentation from the tops of nodules. On alloy 4 oxide was again thin with occasional nodules distributed randomly but these were significantly fewer and smaller than on alloy 3, the largest nodule being = 18 I.crnthick. Oxide was very thick and fragile on alloy 5, in accordance with the observations made during visual inspection, whereas on alloy 6 it was discontinuous and = 32 nrn thick on one side but thinner, discontinuous and patchy on the other. It was concluded from the limited experimental information that alloy 3 and 4 were superior to the others, alloy 4 being marginally better than alloy 3 with respect to nodular corrosion behaviour under these conditions. The post-irradiation mechanical properties of the Scanuk alloys measured in ring tensile tests compared favourably with Zircaloy-2. The strengths in the temperature range 250-400°C were comparable or slightly higher than that of cold-worked, stress-relieved Zircaloy, e.g. alloy 4 had equal strength at 250°C and was about 49 MN/m* stronger at 400°C. The size of ridges measured during profilometry indicated that, of the Scanuk alloys, alloy 4 had the greatest resistance to ridge formation. Comparing the post-irradiation strengths of the alloys exposed in the fully annealed condition, alloy 4 was 20-25% stronger than Zircaloy-2 at 250°C and 40-50% stronger at 400°C the improvements arising without any significant loss in ductility. 2.17. Full scale fuel pin irradiations An SGHWR element was made up of pins clad in Zircaloy-2 (in both the stress-relieved and the recrystallised conditions) and Ozhennite and Valloy, Zr-2.5% Nb and Scanuk alloys 4 and 6 all in the recrystallised condition. The tubing had a grit-blasted bore and a beltground outer surface; eight of the Zircaloy-2 clad pins were autoclaved on the outer surface, four in the stressrelieved and four in the recrystallised condition. This fuel element was irradiated in SGHWR for 238 EFPD (4980 MWd/teU mean). Element power varied between 3.4 and 3.8 MW, coolant flow rates were between 8.6 and 9.1 kg/set and exit steam qualities were in the range 26.5 to 28%. There was no marked difference in appearance between the various alloys after irradiation, apart from a very distinct white ring of

___i

SCANUK

.-I

6

Zr2hNb

1 SCANUK

-

4

ZIRCALOY

I

2

OZHEMJITE

i

0 PIN

0.1

0.2

EXTENSIONS

(MEAN)

VALLOY

0.3 %

Fig. 10. Pin extensions after irradiation in SGHWR.

corrosion on the top end welds of the Zr-2.5% Nb pins and, on one of the Valloy pins, some evidence of spalling in regions at grids 7-l 1. (The fuel pins are supported by a series of eleven grids fixed to the central sparge tube, and numbered from the top (exit end) of the channel downwards). The thin crud below grid 11 had a speckled appearance which may have been associated with spalling. A similar but less marked effect was seen on two other Valloy pins in the outer ring. The most striking result from pm length measurements (fig. IO) was the greater extension of the Valloy and Ozhennite pins compared with Zircaloy, Zr-2.5% Nb and Scanuk 4 and 6 pins tended to have lower extensions than Zircaloy but there was no significant difference between Zircaloy pins in the two heat treat-

VALLOY

t+ OZHENNITE

t--,

ZIRCALOY

2

c:Z.nYi

Nb

1-1

SCANUK

4

/___

0

STACK

SHRINKAGE

/PIN

EXTENCION

WTIO

Fig. 11. Stack shrinkage/pin extension ratios for intermediate ring pins from the SGHWR element.

180

C. Tyzack et al. / SC4NUK

ments. The relative order outlined above was similar in both intermediate and outer rings of the element. The length increases were compared with fuel stack length changes and it was apparent that the unexpectedly large length increases for Vahoy pins were associated with a lower than usual fuel stack shrinkage. Other stack length changes could not be graded satisfactorily and only the Valloy pins were consistent in showing the smallest effect. Earlier work at Windscale had shown that for Zircalay-2 pin length increases (AL) were related to stack length decreases (A!$ the ratio of the two (~~~) being approximately constant for pins from the same ring in the element. The @[AL ratio is plotted in fig. 11. In the intermediate ring the order of increasing M/&L ratio was similar to that for pin length changes, namely: - Valloy > Ozhennite > Zircaloy > Scanuk 4 and 6 and Zr-2.5% Nb. Assuming that the ratings of pins in each ring were comparable, the variable most affecting Iength change would be irradiation creep strength of the cladding so this implies that Valloy and Ozhennite have the lowest irradiation creep strength and Zr-2.5% Nb and Scanuk alloys 4 and 6 the highest.

-

SCANUK

6

1 SCANUK

4

ZlRCALOY

I

,

2

OZHENNITE

-,

2,2’/2Nb

I VALLOY

0

I

k

20

40

RELATIVE IAMETRAL (FROM ECT 9 APPROX

, bo RIDGE MEAN

[N

m)

HEIGHTS VALUES.

Fig. 12. Relative diametral ridge heights on intermediate ring pins from the SGHWR element.

In eddy current tests the largest ridges were seen on Valloy pins and the smallest on Scanuk alloy pins (fig. 12). Differences in ridging behaviour between recrystallised and stress-reiieved Zircaloy-2 did not appear to be significant. Results on Ozhennite and 2.5% Nb-Zr did not fit well into the length change, ridge height correlations, the ridges on Ozhennite being somewhat smaller

Fig. 13. Oxidation at spacer pad positions on pins from SGHWR element. (a) Valley, (b) Stress relieved Zircaloy-2. (c) Recrystallised Zircaloy-2. (d) Scanuk 6.

PAD

43

5

PA06

TO

PAD7

V-Vdloy 46Sconuk 4 ond 6 SR - strP% I-diPmd Zircdoy 2 R - ~ryst~d zifCO!O)’ 2 Nb -B-t%! lbNb

Fig. 14. Oxide thickness on titermediate

and those on 2.5% Nb-Zr larger than expected. Some pins were decrudded with a wire brush and wiped clean prior to visual examination. The Vailoy and Uzhennite pins were the worst from a corrosion st~dpoint (fig. 13a), a thick oxide layer being present at spacer pad positions particularly below pads 6. At

ring pins of element irradiated in SGHWR,

pads 7 to 11 spalling of oxide had occurred beneath the grid band. The stress-relieved Zircaloy pins were unusual in that oxide nodules, typically 0.2 to 0.5 mm in diameter were nucleated preferentia~y along circumferent~l and axial lines (fig. 13b). This may be typical for this material

4

PAD4

70

PA0

7

~

PA0

ALLOY Fig.

15.

IENIIFICATIONS

a

A5

PA0

FOR

io

FIG. 14

Oxide distribution on intermediate ring pins of element irradiated

in SGHWR,

182

C. Tyzack et al. / SCANUK

Fig. 16. Oxidation at outer surface of cladding from SGHWR element. (a) Zr-2.5% Nb. (b) Recrystallised Zircaloy-2. (c) Stress relieved Zircaloy-2. (d) Valloy.

with a ground surface and it is noteworthy that circumferential scratches on a Scanuk 4 pin did not show nodular oxidation. The stress-relieved Zircaloy-2 had a denser population of nodules in bands under the grids at spacer pad positions, grids 8 to 10 being worst affected. Nodular corrosion was also present between spacer pad positions and there was considerable circumferential variation in nodule concentration. Incontrast, nodules on recrystallised Zircaloy pins were smaller, typically 0.1 to 0.2 mm diameter, covering essentially the same areas as in the stress-relieved pins but the overall impression was of less extensive attack (fig. 13~). The two Scanuk alloys appeared similar in many respects to the recrystallised Zircaloy, nodular oxidation being on a still finer scale (fig. 13d). This was also true of the Zr-2.5% Nb alloy which showed the least oxidation of all, with the exception of the accelerated attack at welds. These observations were confirmed during metallographic examination of sections, the summary in figs. 14 and 15 being based on many measurements around the whole circumference of each section. Attention was concentrated on spacer pad positions (5,8 and 10)

and especially on the region that had been adjacent to a grid band during irradiation because nodules formed there preferentially. It was confirmed that there was little nodular oxidation on Zr-2.5% Nb pins (fig. 16a) the mean oxide thickness being 2 p. Nodules up to 12 pm thick were present at spacer pad 8 but coverage was less than i%. Of the remaining alloys Scanuk 6 was superior, both in terms of nodule thickness and coverage. Comparison of stress-relieved and recrystallised Zircaloy was complicated by differences in distribution and morphology of the oxide; on recrystallised Zircaloy there were squat compact nodules (fig. 16b) whereas those on stress-relieved material were lenticular (fig. 16c), but oxidation was generally less on the recrystallised alloy, sometimes by as much as 40%. Oxide on the Valloy cladding was up to 100 w thick (fig. 16d). Oxidation resistance at spacer pad positions was also assessed by analysis of eddy current signals on seven pins taken from the intermediate ring. This relied on measurement of residual wall thickness and led to a ranking broadly confirming that from metallography.

C. l)zack et al. / SCANUK

3. Discussion

One can only speculate on the circumstances that could lead to adoption of a replacement alloy for Zircaloy-2 in a fuel cladding or fuel element context. This might occur, for example, if some major improvement in alloy properties was found in the course of a research and development programme, on the other hand it seems much more likely that a change could be made only on the basis of some serious operational shortcoming in the properties of Zircaloy-2. On present showing this is perhaps more likely under BWR/ SGHWR than PWR conditions. Zircaloy-2 is an excellent alloy whose worth has been proved many times ‘over, particularly in pressurised reducing conditions. Under oxygenated boiling conditions, however, it appears to be susceptible to local&d accelerated attack particularly in regions of flow discontinuity such as grids, and it has been suggested by workers at Windscale that as the grids are stainless steel there may be galvanic coupling between these and the fuel pins accounting for the locally enhanced attack. If a galvanic mechanism is involved then it may be preferable to change the grid material rather than the alloy. However, nodular corrosion has been observed in circumstances where this type of galvanic coupling was absent. Examination of pins from the SGHWR element referred to earlier showed that Scanuk alloy 4 was comparable with recrystallised Zircaloy in respect of susceptibility to nodular corrosion while Scanuk alloy 6 was probably somewhat less susceptible but it was noteworthy that Zr-2.5% Nb was the most resistant. On this basis it is tempting to link susceptibility to nodular corrosion with the extent and size of iron-chromium intermetallic particles which tend to be more corrosion resistant than the matrix. These form very fine second phase dispersions in the Scanuk alloys 3 to 6 and might be associated with the finer scale nodules observed on these alloys. The Zr-2.5% Nb alloy in the quenched and tempered condition contains small amounts of niobium-rich phase, perhaps about 2 vol%, but this phase might be expected to be substantially less corrosion resistant than the matrix. There is still considerable uncertainty about the corrosion properties of zirconium-niobium alloys. In oxygenated water at 29O”C, out of reactor, alloys containing 0 .5,1 and 2.5% Nb showed progressively higher weight gains for a given exposure, although /I solution

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treatment followed by quenching and ageing at SOO’C decreased the susceptibility of Zr-2.5% Nb to that intermediatebetween the 0.5 and 1% Nb alloys, probably by reducing the amount of niobium remaining in solid solution below 1%. In-reactor, however, under oxygenated conditions, this relative order was not preserved, Scanuk alloys 5 and 6 having higher weight gains than alloys 1,2,3 and 4. The Scanuk alloy corrosion coupons with the lowest weight gains in both the Halden BWR and the Danish DR3 irradiations in boiling oxygenated water (not reported in this paper) were alloys 1,4 and 3, alloy 3 having the lowest followed closely by alloy 4, suggesting the overall beneficial effect of chromium. Thus while alloys 4, 5 and 6 seem to be somewhat better than the others under oxygenated conditions out of reactor, the reverse seems to be true when irradiated id oxygenated coolants, certainly for alloys 5 and 6, although alloy 4 seems relatively good in either environment. On the other hand it has to be recognised that the relatively poor behaviour of alloys 5 and 6 may be related to their molybdenum content. In retrospect it would have been instructive to include a Zr-O.5% Nb binary alloy among those selected for evaluation. The DR3 mini-pin irradiations, with oxygen levels in the water considerably higher than one might anticipate in BWR conditions, suggested that alloys 3 and 4 were most resistant to local&d corrosion, but alloys 1 and 2 did not appear to be immune and alloys 5 and 6 seemed generally less resistant. These findings are broadly in accord with the DR3 corrosion coupon irradiations in boiling, highly-oxygenated water. The SGHWR irradiation was perhaps more representative of BWR conditions with respect to flux and oxygen levels in the water but had a higher steam quality than usual. It was unfortunate that alloy 1 could not be included to check if it were less susceptible to nodular attack than alloys 4 and 6 as appears to be the case for Zr-2.5% Nb. The improved resistance to nodular attack of alloy 6 compared with alloy 4 might be related to the fact that it contains only 0.3% Cr and presumably fewer chromium-containing intermetallics. Further irradiations with Scanuk alloys 1,3,4 and 6 as fuel cladding seem desirable particularly at lower steam qualities more representative of BWR/SGHWR conditions. The decrease in strength with increasing temperature in the range 25O’C to 400°C is less severe in

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Scanuk alloys than for Zircaloy. Although the upper yield strength and UTS show the anticipated decreasing trend with increasing temperature it is less drastic in Scanuk alloys and they also start from a higher UTS baseline at 25O’C than Zircaloy. The alloys retain 5-15% more of their room temperature strength in the temperature ranges of interest than does Zircaloy, those containing precipitates, particularly Scanuk 3 and 4, being the best. These improvements in high temperature strength are maintained after irradiation and are not gained at the expense of reduced ductility which remains perfectly adequate. As noted previously, alloys 3,4,5 and 6, which contain precipitates, have better strain hardening properties than alloys 1 and 2 and Zircaloy, and similarly would not be expected to suffer so much from the disadvantages of strain ageing. Thus although iron-chromium bearing intermetallic particles may be suspected of playing an adverse role in nodule formation under oxygenated conditions, they have, nevertheless an important influence on high temperature mechanical properties of some of the Scanuk alloys. Thus one would need to balance the relative advantages and disadvantages of retaining them or strengthening the alloys in some other way if improved strength were judged to be important. Behaviour of the alloys in short-term temperature transients indicated a gradual decrease in ductility with maximum temperature of exposure up to 9OO’C where a pronounced drop was observed, which can reasonably be correlated with an increase in the kinetics of oxygen solution into metal. The high temperature tensile measurements in the range 750-900°C on Scanuk alloys 1,3 and 4 and Zircaloy are not directly app~cable to biaxial stressing of tubes but are useful to indicate the temperature ranges for ‘superplastic’ behaviovr associated with a fine-grained duplex structure, i.e. where elongation is large and the strength very low. If a criterion of >200% elongation is adopted, it is apparent that all the alloys are superplastic over part of the temperature range. This range was larger for Scam& alloy 1 than for alloy 4 or Zircaloy, and can be correlated with a wider OL + fl region in the former alloy, there being a rapid decrease in elongation whenthe temperature exceeds 85O’C due presumably td transformation to the &phase and grain growth. Zircaloy has similar superplastic behaviour with a oeak temnerature around 900°C, i.e. 30-50°C higher

than for the Scanuk alloys examined. On the basis of these tests large failure strains might be expected under loss of coolant accident conditions in the LY+ /3 re8ion. The fact that the region of maximum deformation is around 900°C for Zircaloy, i.e. where strengthening by oxygen solution starts, may be a point in its favour .

4. Conclusions Taking a broad view it is probably fair to say that the Scanuk alloys could, with further study and optimisation, present a viable alternative to Zircaloy as cladding material. This class of alloy, based on modest additions of niobium to zirconium, has a basically different behaviour in oxygenated aqueous environments, out-ofreactor, where the rate of oxidation is probably proportional to the amount of niobium in solution and to the square root of oxygen concentration in the water, whereas the corrosion behaviour of Zircaloy out-ofreactor appears to be essentially independent of oxygen concentration. The performance of Scanuk alloys under irradiation in oxygenated conditions is not altogether clear-cut but there are captions that while oxidation rates are tolerable in, for example, Halden BWR and SGHWR, with modest oxygen levels in the water, rates are much faster under the high flux, oxygen radical-rich conditions in DR3. It would be ~teresting to study the behaviour of a quenched and tempered Zr-2.5% Nb alloy in this latter environment as there is clear evidence that the corrosion rate of this alloy is suppressed in oxygenated conditions in-reactor (SG~R) compared with behaviour out of the reactor. It is possible that the corrosion resistance of Scanuk alloys 1 and 2 might be improved by a quench and tempering treatment of the type applied to Zr-2.5% Nb. It would also be instructive to know how the corrosion resistance of Z&.5% Nb compared with higher niobium content alloys, although it seems unlikely that such a binary alloy would have a strength comparable with that of Zircalo y .

A further complicating factor is that Scanuk alloys, 1 and 2 as-fabricated tubes contained more nitrogen (70 and;40 ppm) than that specified for the Russian Zr-1% I& alloy (30 ppm) and this may have had adverse effects in the present tests.

C. Tyzock et al. f SCANUK

The Scanuk alloys 3 and 4, containing respectively I and 0.5% niobium with 0.5% c~o~urn appear to be excellent from many points of view and certainly possess improved higher temperature properties in the range 250 to 400°C, due to better precipitate dispersion. They also have excellent oxidation resistance under BWR conditions with some slight tendency to nodular oxidation, which is probably associated with chromiumcontaining intermetallic particles. It is interesting to note that increasing the content from 0.5% to 1% in the 0.5% chromium alloy gave coarser precipitates so that, if there is a correlation between intermetallic particles and nodular corrosion, Scanuk alloy 4 is perhaps to be preferred to 3 from this point of view. This seems to be borne out in the results of the minipin irradiations. On balance the molybdenum~onta~ng alloys S and 6 seem to be not so good as alloys 3 and 4, alloy 5 in particular being character&d by thick oxide and spalling in a range of different conditions. The distribution of precipitates in this alloy was uneven and it showed the least improvement in high temperature mechanical properties over Zircaloy. It would almost certainly be discarded in any follow-up programme. Alloy 6 had somewhat uneven behaviour, generally among the least corrosion resistant in Halden BWR and DR3 tests on coupons and mini-pins, but surprisingly good in the SGHWR irradiation with a somewhat lower pro~nsity to nodular for~tion than Scanuk 4. This may be due to the lower chromium content. Probably the molybdenum addition would show its beneficial effect in promoting oxidation resistance only at higher tem~ratures. Although the alloy compositions were chosen with higher temperature service in mind, this is not the way reactor development has proceeded. All ideas of using nuclear superheated steam appear to have been abandoned and even the once-through concept with sub-cooled water at inlet and superheated steam at outlet with dryout on the cladding surface no longer finds favour. It is generally agreed that irradiation effects on oxidation behaviour of zirconium alloys fade out above 400°C consequently the oxidation data obtained out-of-reactor in superheated steam should be relevant to a first approximation if the effect of oxygen in the steam is ignored. At 400°C the Scanuk alloys are certainly comparable in oxidation resistance with Ziicaloy and hydrogen absorption is one third

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to one half that observed with Zircaloy. At 500°C and above, the Scanuk alloys were markedly superior to Zircaloy and have been shown to be immune to the high temperature nodular attack which leads to fairly rapid disintegration of Zircaloy in high pressure steam. The alloys, as expected, do not have as good oxidation resistance as Valloy at the highest steam temperatures, but on the other hand they are much better at 290°C and thus show excellent resistance across a broad range of tem~ratures extending higher than for Zircaloy. In terms of response to short-term high temperature transients there is probably not much to choose between Zircaloy and the Scanuk alloys but in relation to the so-called ‘superplastic’ behaviour in the cy+ p range the fact that this occurs in the niobium alloys at somewhat lower temperatures and for Scanuk alloys I and 2 over rather broader temperature ranges than Zircaloy needs more detailed consideration in the LGCA context, and calculations would probably be meaningful only in relation to a specific design.

Acknowledgements

Thanks are due to many people in the various organisations who have helped to make this collaboration a success. It is a pleasure to acknowledge the generosity of staff at AEK and IFA in sponsoring ~~-pin and other irradiations in DR3 at Ris$ and in the Halden Boiling Water Reactor, and the continued support and enthusiasm of Mr. D.O. Pickman which facilitated the ~radiation of a fulI-scale element containing Scanuk alloys in SGHW Reactor. Special thanks are also due to Dr. Niels Hansen for his continuing advocacy of the programme and to Dr. Karen Laursen AEK who joined the collaboration in its later stages. Ac~owledgements are made to Dr. R.C. Asher of AERE, Harwell for his participation in formulating the original programme, and to Dr. V.W. Eldred, RDL for facilitating the postirradiation examination of the SGHWR element. References [i] C.S. Campbelland C. Tyzack, Br. Corros. T. 5 (1970) 172 [2] British Patent Specification No. 829,688 [ 31 W.G. O’Driscoll, C. Tyzack and T. Raine. Second UN Intern. Conf. on Peaceful uses of atomic energy, paper no. 1450 (1958)

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[4] British Provisional Patent Specification No. 20056/59 [5] British Provisional Patent Specification No. 21338/59 161 D. Armitage, internal Report, AEI Power Group Research Laboratory, Manchester, (1965) [ 71 KM. Goldman and DE. Thomas, Properties of zircaloy 2, WAPD-T-43 (1953) [S] S. Kass, The corrosion behaviour of zirconium and its alloys in high temperature water and steam, Proc. Atomic Industrial Forum on Zirconium, (1955). [9] D.E. Thomas, Aqueous corrosion of zirconium and its alloys at elevated temperature, First UN Intern. Conf. on Peaceful uses of atomic energy, paper no. 537 (1955). (i 0] B. Lustman and F. Kerr-e, The metallurgy of zirconium (McGraw-HBl, New York, 1955). [ 111 S. Kass, Corrosion testing of two zirconium alloys, Corrosion 16,58lt-585t (1960). [ 121 A.D. Amaev et al,, Research and tests on experimental assemblies of fuel rods, USSR 1203, (1966). [ 131 A.D. Amaev et al., Third Intern. Conf. on Peaceful uses of atomic energy, Vol. II, (1969) p. 492. [ 14) A.D. Amaev et al., Investigation on the behaviour of zirconium alloys in water and steam-water mixtures in the MR reactor loops, USSR-1739 (1968). [IS] J. Wllton and R.A. Murgatroyd, The effect of variations in composition and heat treatment on the properties of Zr-Nb alloys, Symp. on Zirconium and its alloys, Fall Meeting of Electrochemical Society, Buffalo, New York, (1965) p 358. [ 161 G.F. Slattery, Some aspects of the p * LYtransformation of Zirconium 2.5% Nb alloy with variable cooling rate and soak temperature, ibid, p. 336. [ 171 D.S. Wood, J. Winton and B. Watkins, Effect of irradiation on the impact properties of hydrided Zircaloy-2 and zirconium-niobium alloy, ibid, p. 250.

[ 181 G.T. Higgins and E.E. Banks, The basic features of the isothermal promonotectoid transformation in a Zirconium 2.5% Nb alloy, ibid, p. 341. [ 19 ] C.E. Ells and V. Fidleris, Effect of neutron irradiation on the tensile properties of the zirconium 2.5% niobium alloy, ibid, p. 268. [20] B.A. Cheadle and C.E. Ells, The effect of heat treatment on the texture of fabricated zirconium-rich alloys, ibid. p, 329. [21] J. Boulton, Current knowledge of zirconium alloys for reactor usage, AECL 3365, (1969). 1221 A.B. Johnson, Jr., Corrosion and hydriding response and oxide properties of six irradiated zirconium alloys, BNWL1819 rev. (1968). [ 231 A.B. Johnson, J.E. LeSurf and R.A. Proebstle, A study of zirconium alloy corrosion parameters in the advanced test reactor, ASTM Symp. on Zircaloy in nuclear application, Portland (1973). [24] P. Hurst, Recent results on the out-of-pile corrosion of zirconium alloys at RML, UK/ Canada Zirconium Conf. Bl UK1 (1967). 1251 F.H. Megarth et al., Zircaloy-clad LJOz fuel rod evaluation program,GEAP-10371 (1971). [26f R.C. NelsonCEAP~O89 {Vol. Ii) 175-17-22 (1972). [27] R.N. Duncan et al., Corrosion of Zircaloy-2 fuel rod cladding in boiling water reactors, NACE meeting (1967). [28] F.H. Megarth, Zircaloy-clad UO2 fuel rod evaluation programme, Quarterly progress report No. 3, GEAP 5667, (1968). [29] F.W. Trowse, Some observations of nodular corrosion of zircaloy cladding in a BWR environment, Paper presented at the OECD Halden Reactor Project Enlarged HPG Meeting, Sanderstgen, Norway (1973). [30] A.S. Barn and J.E. LeSurf, Oxidation and hydriding of Zircaloy in NPD Reactor AECL-3065, (1969).