Journal of Nuclear Materials 429 (2012) 257–262

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Journal of Nuclear Materials

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Presented at ANM-2011 Conference Feb 9–11, Mumbai, India

Microstructural examination of high temperature creep failure of Zircaloy-2cladding in irradiated PHWR fuel pins

Prerna Mishra, D.N. Sah ⇑,1, Sunil Kumar, S. AnantharamanPost Irradiation Examination Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085, India

a r t i c l e i n f o a b s t r a c t

Article history:Available online 9 June 2012

0022-3115/$ - see front matter � 2012 Elsevier B.V. Ahttp://dx.doi.org/10.1016/j.jnucmat.2012.06.002

⇑ Corresponding author. Tel.: +91 22 25595009; faxE-mail address: dnsah@barc.gov.in (D.N. Sah).

1 Raja Ramanna Fellow.

Cladding samples taken from the ballooned region of the irradiated Zircaloy-2 cladded PHWR fuel pinswhich failed during isothermal heating tests carried out at 800–900 �C were examined using opticaland scanning electron microscopy. The examination of samples from the fuel pin tested at 900 �C showedan intergranular mode of failure in the cladding due to formation of cracks, cavities and zirconiumhydride precipitates on the grain boundaries in the cladding material. A thin hard a-Zr(O) layer wasobserved on outer surface due to dissolution of the oxide layer formed during reactor operation. Grainboundary sliding was identified to be the main mode of creep deformation of Zircaloy-2 at 900 �C. Exam-ination of the cladding tested at 800 �C showed absence of cracks or cavities in the deformed material andno localisation of hydrides was observed at the grain boundaries. The failure of the cladding occurredafter necking followed by extensive wall thinning of the cladding tube.

� 2012 Elsevier B.V. All rights reserved.

1. Introduction

Zircaloy is used as the cladding material for water reactor fuelpins due to its low neutron absorption, good mechanical strengthat operating temperature and corrosion resistance in aqueous envi-ronment. During postulated loss of coolant accident (LOCA) condi-tions, the Zircaloy cladding of the fuel pins is likely to undergofailure by excessive deformation and ballooning of cladding tubeas a result of high temperature (700–1200 �C) and high hoop stresspresent in the cladding. The deformation of the cladding underthese conditions occurs by creep. Several investigations of defor-mation and ballooning of Zircaloy cladding tubes have been re-ported in the literature [1–7], but the exact mechanism of creepand fracture of Zircaloy cladding at such high temperatures hasnot yet been established. Earlier investigations on creep and frac-ture have been carried out at lower temperatures of 500 �C and600 �C [8,9]. Troy et al. [8] reported formation of creep cavities nearfracture at 600 �C. The mode of fracture during creep was reportedto change from transgranular at 450 �C to intergranular mode at500 �C [9]. The objective of the present study is to examine themode and mechanism of creep failure of the irradiated Zircaloy-2cladding tube by examining samples from failure locations of theballooned and failed fuel pins. For this purpose samples selectedfrom our earlier experiment [10,11] in which fuel pins from irradi-ated PHWR fuel bundles having burnup of 7500 MWd/tU and

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15,000 MWd/tU and having internal fission gas pressure of0.55 MPa and 2.4 MPa were tested, have been examined The re-sults of microstructural examination of failure are presented in thispaper.

2. Experimental

2.1. Zircaloy-2 cladding tube

The diameter and wall thickness of the Zircaloy-2 cladding tubeused in the fuel pins was 1.52 cm and 0.04 cm respectively. Thechemical composition of Zircaloy-2 cladding tube is given in Table 1below. The cladding tubes are fabricated at Nuclear Fuel complexHyderabad by four stage pilgering with intermediate annealing.The final tubes are stress relieved.

2.2. Fuel pin heating

A closed cylindrical type horizontal electrical furnace developedfor heating a small part of the irradiated PHWR fuel pin inside thehot cells was used. The details of this furnace are given elsewhere[10]. The overall length of the furnace was 750 mm which alloweda full length fuel pin to be inserted in the furnace. A 10 cm length ofthe fuel pin towards one end was heated under argon atmospherein the temperature range 800–900 �C for 10–15 min. The furnacewas cooled and the fuel pin was removed from the furnace andsubjected to post test examination like visual examination, dimen-sion measurement, leak testing, optical and scanning electronmicroscopy, etc.

Portion of the fuel pin heated during the experiment.

Fig. 1. Appearance of irradiated PHWR fuel pins after ballooning test inside the hotcells.

Table 1Chemical composition of Zircaloy-2 cladding (wt%).

Sn Fe Cr Ni O Zr + impurities

1.41 0.15 0.10 0.06 0.12 Balance

Table 2Test conditions of the failed fuel pins.

Sl.no

Fuel pinburnup(MWd/tU)

Cold internalpressure in fuelpin (MPa)

Heatingtemperature(�C)

Holdtime(min)

Deformationat balloonedlocation (%)

1 15,000 2.4 800 10 38–412 15,000 2.4 900 106 7500 0.55 900 15

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2.3. Cutting and mounting of samples

The deformation and ballooning had occurred in the 10 cm por-tion of the fuel pin towards one end which was heated to high tem-perature. Transverse sections were cut from the failed fuel pinsfrom the location showing maximum ballooning and one sectionwas taken at a distance of about 40 cm from the ballooned location.The cut slices were defuelled and the cladding rings were mountedin cold setting resin. Mounted samples were prepared inside thehot cells using metallographic sample preparation system.

2.4. Metallographic examination

Metallographically prepared samples were examined first inas-polished condition. The samples were examined for claddingthickness and oxide layer thickness, presence of cracks, etc. Subse-quently, the samples were etched with a mixture of HF, HNO3 andglycerol to examine the hydride platelet distribution in thecladding.

2.5. Hydrogen measurement in the cladding

In order to determine the hydrogen content in the initial mate-rial before testing, hydrogen analysis was carried out on the clad-ding samples taken from a location about 40 cm away from theballooned region of the fuel pins, using differential scanning calo-rimetry (DSC). The hydrogen content of the cladding samples takenfrom the ballooned regions after the test was estimated from thearea fraction of hydrides in the photomicrographs of the samplesusing image analysis techniques.

2.6. Microhardness testing

Microhardness measurements were carried out on the claddingat 100 gm load using a microhardness tester with Vickers pyramidindenter attached to the microscope. Measurements were carriedout on the cladding from the ballooned and the unballooned regionof the fuel pin.

2.7. SEM examination

Metallographically prepared samples were examined under ascanning electron microscope in secondary electron mode and inback scattered electron mode.

3. Results

3.1. General observations

The results on the extent of cladding deformation and balloon-ing in the fuel pins tested in this experiment are presented else-where [10–12]. Typical appearance of the irradiated fuel pinsafter ballooning test is shown in Fig. 1.The ballooning and defor-mation of the cladding tube was near one end because only a10 cm length of the fuel pin at this end was subjected to heatingin the experiment. Since there was no visible burst opening or widesplit of cladding on the fuel pin surface, the failed fuel pins wereidentified by carrying out liquid nitrogen–alcohol leak test.

Table 2 shows the test conditions of the fuel pins which failedduring the test. The failed fuel pins included two pins with highinternal pressure which failed during 10 min heating test. Thepin with low internal pressure failed during heating for 15 min at900 �C. The maximum diametral deformation in the ballooned re-gion in failed fuel pins was in the range 38–41%.

3.2. Surface appearance of ballooned regions in the failed pins

A close examination of the ballooned region of the failed fuelpins was carried out using a wall mounted periscope in the hotcells to examine the surface features at the failure site. Fig. 2aand b show the appearance of the cladding surface in ballooned re-gion of two failed fuel pins tested at 800 �C and 900 �C, respec-tively. From visual examination, it was not possible to clearlylocate the failure since there was no wide split of the cladding ineither case as is usually seen in the cladding tube failure in bursttest or ballooning test. The failure in these pins was in the formof micro cracks. The ballooned surface of fuel pin tested at 800 �C(Fig. 2a) only shows presence of a small crack like cavity and sev-eral axial depression marks. The ballooned surface of fuel pintested at 900 �C (Fig. 2b) shows a region which appears unevenand contains depression marks, fine cracks and surface cavities,some of which appear to have joined together to form fine axialcracks on the surface.

3.3. Metallographic observations

Samples from the unballooned region (far away from the heat-ing location) of the cladding tube revealed a uniform oxide layer atthe outer surface of the cladding with an average oxide layer thick-ness of 3.2 lm (Fig. 3a), which was formed during reactor irradia-tion. However, no oxide layer was present in the sample takenfrom the ballooned region of fuel pin heated at 900 �C for 10 min(Fig. 3b). Transverse cracks were observed on the outer surface ofthe cladding samples. However, cladding sample taken from theballooned region of fuel pin heated at 800 �C for 10 min showed

Fuel pin heated at 800oC

Fuel pin heated at 900oC

Fig. 2. Surface appearance of failed fuel pins in the ballooned region.

Zircaloy cladding

Oxide layer

Fig. 4. Photomicrograph showing presence of original oxide layer on the outersurface even after heating at 800 �C for 10 min.

Thin oxide layer on ID surface

Hard layer on OD surface

Fig. 5. Thin oxide layer on the inside surface of the cladding after the test at 900 �C.

P. Mishra et al. / Journal of Nuclear Materials 429 (2012) 257–262 259

that the original oxide layer was still present on the cladding sur-face (Fig. 4).

Examination of the inside surface of cladding in the unbal-looned portion of the fuel pin heated at 900 �C, showed presenceof oxide layer having 3–5 lm thickness. A thin layer of oxide wasstill present at some locations on the inside surface on the bal-looned region of this fuel pin (Fig. 5). Oxide layer was missing atlocations where wall thinning was high.

The microstructure of cladding from ballooned region of fuelpin tested at 800 �C is shown in Fig. 6a and b in the as-polishedand etched condition respectively. The as-polished surface showsthat deformed cladding is free from cavities. The examination ofthe etched surface showed presence of circumferentially orientedhydride precipitates (Fig. 6b).

Zircaloy Cladding

Oxide layer

a b

Fig. 3. (a) Oxide on the outer surface in the cold region (b) absence of oxide layer on theduring heating.

Microstructure observed on etched surface of cladding samplestaken from the ballooned region of the fuel pins tested at 900 �C isshown in Fig. 7. The microstructure shows clearly demarcatedequiaxed grains in the cladding. The grain boundary was foundto be decorated with a second phase material. Since these featureswere observed after etching of the sample for hydrides, it is be-lieved that the black phase on grain boundaries are zirconium hy-dride precipitates. Average grain size was measured to be 18 lmand the area fraction of the grain boundary phase was 8.5%. A fea-tureless band (35 lm thick and having high hardness of 425 VHN)with no visible grain structure was noticed at the outer surface ofthe cladding in the samples taken from ballooned region of fuel pinheated at 900 �C. Transverse cracks emanating from outer surfacewere found to be present in this layer.

Sample taken from fuel pin tested at 800 �C for 10 min showedextensive thinning at the necked region (Fig. 8). The thinned clad-ding showed presence of hydride platelets in it. Necking withsomewhat lower deformation and presence of an intergranularcrack initiated at the inner surface and penetrating about 75% ofwall thickness was observed in the cladding sample from fuel pintested at 900 �C (Fig. 9a–b). A through wall crack without anyappreciable necking was also observed in the cladding (Fig. 10)in another sample. One of the cracks existing on the outer surface(see Fig. 3b) is seen to have propagated through the thickness con-necting the large cavities on the way.

3.4. Hydrogen content and hydride distribution

The result of hydrogen analysis of cladding samples is given inTable 3. Hydrogen concentration in initial material before testing,

Zircaloy Cladding

outer surface after heating at 900 �C for 10 min indicating dissolution of oxide layer

Hydrides

Etched (b)

As -polished (a)

Fig. 6. Microstructure of ballooned cladding in samples from fuel pin heated at800 �C: (a) as-polished condition and (b) etched condition.

High hardness

Fig. 7. Microstructure of ballooned cladding in the fuel pins heated at 900 �C.

260 P. Mishra et al. / Journal of Nuclear Materials 429 (2012) 257–262

obtained from differential scanning calorimetry of the claddingsamples taken from the cold region (unballooned portion) in thefuel pin, was found to be 49 ppm. The hydrogen content in thecladding from the ballooned region after testing at 800 �C and at900 �C, estimated by image analysis technique was found to be

As - polished

Fig. 8. Necking in the cladding i

35.4 ppm and 105 ppm respectively. There is a significant increasein the hydrogen content in the cladding tested at 900 �C due topresence of b-Zr phase in Zircaloy-2 at this temperature. Reductionin hydrogen content in sample tested at 800 �C may be due to dif-fusion of some hydrogen from this location to the cooler regiondown the temperature gradient.

The hydride distribution in the initial material before testingand in the cladding samples from ballooned locations are shownin the photomicrographs in Fig. 11a–c. The microstructure of initialmaterial shows circumferentially oriented fine hydride plateletsuniformly distributed in the cladding material (Fig. 11a). Hydridesin the cladding deformed at 800 �C are observed to have the shapeand orientation almost similar to the hydrides in the cladding be-fore testing. However, the spacing between the hydrides has chan-ged and some clustering of hydride platelets is also observedtowards inner and outer surface (Fig. 11b) as a result of deforma-tion. Fig. 11c shows the morphology and distribution of hydridesin the cladding tested at 900 �C. The size, shape, orientation anddistribution of the hydride platelets in the cladding is observedto be completely different from those observed in the initial mate-rial before testing or in the cladding tested at 800 �C. The hydridesin the sample tested at 900 �C are fine, irregular in shape, orientedrandomly and are partially covering the grain boundaries in thecladding microstructure. Since Zircaloy-2 has a two phase micro-structure at 900 �C consisting of a-Zr grains surrounded by b-Zron the grain boundaries, the hydrides observed on the grainboundaries in this sample are due to relocation of hydrogen inthe prior b-Zr phase in the alloy material.

3.5. Microhardness of cladding

Microhardness measurements carried out with 100 gm load onthe cladding sample taken from fuel pin heated at 900 �C revealedhigh hardness of 425 HV in the featureless band present towardsthe outer surface as compared to 179 HV at mid-thickness. Thehardness near inside surface was measured to be 194 VHN. Themeasured hardness in the sample of the cladding taken from fuelpin heated at 800 �C was uniform across the cladding wall thick-ness and was close to the hardness of starting material.

3.6. SEM observations

Scanning electron microscopy of cladding samples from bal-looned region of the fuel pin heated at 900 �C revealed presenceof cavities on the grain boundaries and wedge cracks at triplepoints (Fig. 12a and b). There were no cavities in the grain interior.Cavities on the grain boundaries had joined to form cracks on thegrain boundaries.

4. Discussion

Metallographic examination of the cladding sample from fuelpin tested at 900 �C showed formation of a 35 lm thick band to-wards outer surface having high hardness of 425 VHN. The oxide

As - etched

n fuel pin heated at 800 �C.

50 µµm

a b

Fig. 9. SEM photomicrograph showing details at the necking region in cladding sample from fuel pin tested at 900 �C: (a) low magnification view and (b) enlarged view.

Cavities

Inner surface

Outer surface

Fig. 10. A through wall crack in cladding in the fuel pin heated at 900 �C.

Fig. 11. Hydride platelet distribution in the cladding: (a) before ballooning (b) afterballooning test at 800 �C and (c) after ballooning test at 900 �C.

P. Mishra et al. / Journal of Nuclear Materials 429 (2012) 257–262 261

layer which was present on the cladding outer surface before theheating, was not present after the heating test. This indicates thatthe oxide layer had dissolved during the high temperature anneal-ing and ingress of this oxygen has formed an oxygen stabilized a-Zr(O) layer containing high oxygen and having high hardness. Maet al. [13] has studied the dissolution of oxide layer in Zircaloy-4at temperature 700–850 �C. He found dissolution of about 2.5 lmof oxide layer in 45 min at 850 �C but at 800 �C, more than onehour was required to dissolve 0.8 lm oxide layer. So it is expectedthat a 3.2 lm oxide layer will dissolve in 10 min at 900 �C but notat 800 �C as observed in the present experiment. From the mea-sured hardness value the oxygen content of hard zircaloy layercan also be estimated using Leistikov et al. correlation [14]. Accord-ing to this, a hardness value of 425 VHN will correspond to an aver-age concentration of about 1.56 wt% oxygen in the layer.

In the samples taken from the ballooned region of the fuel pintested at 900 �C, a network of cavities, cracks and zirconium hy-drides was observed at the grain boundaries. Garlick [15] also ob-served an intergranular network in zircaloy cladding samples attemperature of 860–880 �C in the fuel rods tested under simu-lated LOCA. It was attributed to the prior beta phase on the grainboundaries. The intergranular network observed in the samples inour study can also be related to the two phase (a + b) structure ofZircaloy at 900 �C. At this temperature Zircaloy-2 contains a largevolume fraction of a-Zr phase and a small volume fraction of b-Zr

Table 3Result of hydrogen analysis of cladding samples.

Cladding sample Location in the fuel pin Hydrogen by DSC (ppm) Hydrogen from image analysis

Area fraction of hydrides (%) Hydrogen (ppm)

Initial material Cold region 49 3.6 –After testing at 800 �C Ballooned region – 2.6 35.4After testing at 900 �C Ballooned region – 7.8 105

Stress direction

a b

Fig. 12. SEM photomicrographs of cladding deformed at 900 �C showing presence of cracks and cavities on the grain boundaries of the cladding material.

262 P. Mishra et al. / Journal of Nuclear Materials 429 (2012) 257–262

phase, which is present along the alpha grain boundaries. Most ofthe hydrogen present in the cladding may get concentrated in thebeta phase on the grain boundaries at this temperature. Aftercooling, this hydrogen will appear as hydride precipitates in theprior beta phase region at the grain boundaries. The hydrogencontent in the cladding of the irradiated fuel pin tested at900 �C was found to be 105 ppm which is about twice the valueof hydrogen in the initial material. Localisation of this hydrogenin the beta phase at grain boundaries may result in a localhydrogen concentration of more than 1000 ppm at the grainboundaries.

At 800 �C, Zircaloy exists as single phase a-Zr and this could bethe reason for extensive thinning of the cladding at necking loca-tion in the samples heated at 800 �C. The circumferential orienta-tion of hydrides observed in the sample is due to predominantlyradial basal pole texture in the cladding tube.

Cavities were observed on the grain boundaries and wedgecracks at the triple point in the microstructure of sample heatedat 900 �C. The grains were found to retain their equiaxed shapeeven though the deformation of the cladding was up to 40%. Noelongation of grains was observed in the stress direction. All theseobservations indicate that grain boundary sliding was the mode ofdeformation in the cladding at this temperature. Grain boundarysliding was reported to be a dominant mechanism of deformationin Zircaloy-2 at 850–900 �C by Garde et al. [16].

The failure of the cladding at 900 �C is found to be intergranularin nature. The crack propagated through the creep cavities in thematerial. Presence of b-Zr phase containing large hydrogen concen-tration also helped in intergranular failure. The failure in the clad-ding at 800 �C is observed to have occurred after considerableplastic deformation followed by extensive wall thinning.

5. Conclusions

Microstructural examination of Zircaloy-2 cladding samplesfrom the ballooned regions of irradiated PHWR fuel pins, whichfailed during isothermal heating test at 800 �C and 900 �C has beencarried out. The main findings of the examination are as follows:

1. The cladding that deformed at 900 �C showed presence ofcracks, cavities and zirconium hydride precipitates on the grainboundaries throughout the sample. The grain interior was freefrom cavities in this sample. The presence of zirconium hydrideprecipitate is attributed to concentration of hydrogen in theprior b-Zr phase present at the grain boundaries in Zircaloy-2cladding at 900 �C.

2. Failure of the cladding during heating at 900 �C occurred at theballooned location by intergranular crack propagation withoutany necking. Necking associated with incipient intergranularcrack initiated at inner surface was also observed.

3. The cladding tested at 800 �C did not show cracks or cavities inthe deformed material. Necking followed by extensive wallthinning of the cladding was observed at the failure location.Cladding failure occurred by pin hole formation.

4. Presence of creep cavities on grain boundaries, formation ofwedge cracks at triple points of grains and absence of grainelongation in the stress direction indicated that grain boundarysliding was the main mode of creep deformation of zircaloycladding at 900 �C.

Acknowledgement

The authors express their sincere thanks to Shri V.P.Jathar, ShriP.M.Satheesh and Shri S. Katwankar of Post Irradiation Examina-tion Division for their help and support in carrying out the exper-iments and sample preparation inside the hot cells.

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