Journal of Nuclear Materials 427 (2012) 396–400

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Journal of Nuclear Materials

journal homepage: www.elsevier .com/locate / jnucmat

Letter to the Editor

Oxidation of fuel cladding candidate materials in steam environmentsat high temperature and pressure q

Ting Cheng a,⇑, James R. Keiser a, Michael P. Brady a, Kurt A. Terrani b, Bruce A. Pint a

a Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN, United Statesb Fuel Cycle and Isotopes Division, Oak Ridge National Laboratory, Oak Ridge, TN, United States

a r t i c l e i n f o

Article history:Received 9 April 2012Accepted 1 May 2012Available online 11 May 2012

0022-3115/$ - see front matter Published by Elsevierhttp://dx.doi.org/10.1016/j.jnucmat.2012.05.007

q This manuscript has been authored by UT-BatteDE-AC05-00OR22725 with the US Department ofGovernment retains and the publisher, by acceptinacknowledges that the United States Government retairrevocable, world-wide license to publish or reprodumanuscript, or allow others to do so, for United State⇑ Corresponding author.

E-mail address: chengt@ornl.gov (T. Cheng).

a b s t r a c t

Under certain severe accident conditions, the fuel rods of nuclear power plants are exposed to high tem-perature/pressure steam environments in which the Zr alloy cladding is rapidly oxidized. As alternativecladdings, the oxidation resistances of SiC-based materials and stainless steels with high Cr and/or Aladditions have been examined from 800-1200 �C in high-pressure steam environments. Very low reac-tion kinetics were observed with alumina-forming FeCrAl alloys at 1200 �C while Fe-Cr alloys with only15-20% Cr were rapidly attacked.

Published by Elsevier B.V.

1. Introduction

The reaction kinetics of zirconium alloys with steam have beenextensively studied [1–3] and are well understood as are the acci-dent scenarios surrounding loss-of-coolant-accidents (LOCAs) infission power reactors. During a LOCA, clad temperature will in-crease due to fuel decay heat resulting in rapid steam-oxidationof zirconium alloy. This phenomenon becomes self-catalytic attemperatures greater than 1200 �C, with the magnitude of the en-thalpy of the zirconium reaction surpassing the heat production inthe fuel. Also of consequence, the zirconium–steam reaction pro-duces significant and dangerous levels of hydrogen that must bedealt with. One approach to achieve larger margins of safetyagainst severe accident scenarios is to replace zirconium alloy withmore oxidation-resistant cladding and/or cladding with less heat/hydrogen production. It is noted here that we assume the mostpractical approach under both normal operating and accident con-ditions is the formation of a protective oxide through reaction ofthe environment with the clad as compared to solutions such asa protective coatings applied to zirconium. As such, the nuclearindustry can draw upon the advanced in alloy development andtesting manifested under a number of research and industrial pro-grams supporting, as example, fossil energy production. From thisbroad area of study it is well understood that above �600 �C in air,

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lle, LLC, under Contract No.Energy. The United States

g the article for publication,ins a non-exclusive, paid-up,ce the published form of thiss Government purposes.

the best oxidation resistance is generally achieved through forma-tion of chromia (Cr2O3), alumina (Al2O3), or silica (SiO2) surfacelayers by Cr-, Al- and Si-containing structural materials or coatings[4]. In dry air conditions for relatively long lifetimes (>1000 h),chromia can potentially provide protection up to �1000 �C; Al2O3

up to �1400 �C; and SiO2 up to �1700 �C (exact upper temperaturelimit depends on multiple factors such as component thickness,exposure conditions, and desired lifetime). The environment insidea nuclear reactor core under severe accident conditions may becomprised of steam or steam–hydrogen mixtures. In high-temper-ature steam environments, Cr2O3 and SiO2 oxide scales, inparticular, form volatile hydroxide species which can reduce theirupper-temperature use limit from that of a dry air condition byhundreds of degrees [5]. Alumina forming alloys are generallyresistant to accelerated oxidation in water vapor conditions [6];but, depending on specific alloy grade, can have issues related tomanufacturability, joining, and cost. Therefore, while there isgeneral guidance on the performance of the various protectionschemes, the actual performance under fission reactor LOCAperformance cannot be determined a priori.

Gains in safety margins under nuclear reactor accident condi-tions require oxidation resistance of the cladding alloy to be main-tained for relatively short periods (minutes to days); as opposed tothe tens of thousands of hours lifetimes typically needed for theapplications where chromia-, alumina-, and silica-forming materi-als have been typically developed (energy conversion and produc-tion equipment, chemical processing, etc.). Therefore, althoughchromia- and silica-forming materials are generally not acceptablefor long-term use at 1200 �C in steam, they could potentially bebeneficial for the short lifetimes needed under LOCA scenarios innuclear reactors. The objective of this work was to screen theshort-term oxidation resistance of a range of candidate chromia-,alumina-, and silica-forming cladding materials (relative to

Table 1Analyzed compositions (wt.%) by inductively coupled plasma and combustion techniques.

Alloy Fe Zr Ni Cr Al Mo Sn Mn Si C O Y Ti

Zircaloy-2 0.14 98.13 0.05 0.10 – – 1.42 – 0.01 0.019 0.115 – –Zircaloy-4 0.22 98.23 – 0.11 – – 1.27 – 0.01 0.016 0.118 – –304L 70.63 – 8.27 18.78 0.01 0.27 0.01 0.73 0.42 0.028 – – –317L 64.15 – 11.86 18.88 – 3.13 – 0.62 0.38 0.023 – – –321L 69.93 – 9.47 17.69 0.03 0.24 0.01 1.14 0.45 0.033 – – –347L 68.88 – 9.97 17.52 – 0.16 – 1.80 0.66 0.04 – – –Fe–15Cr 85.14 – <0.01 14.81 <0.01 <0.01 – <0.01 <0.01 0.003 0.004 – –Fe–20Cr 80.29 – <0.01 19.65 <0.01 <0.01 – <0.01 0.01 0.002 0.003 – –Fe–25Cr 74.58 – – 25.34 – – – – 0.02 0.004 0.037 – –PM 2000 74.58 – 0.1 18.92 5.1 0.01 – 0.11 0.04 0.01 0.25 0.37 0.45NITE-SiC 0.32 – 0.04 0.09 0.38 – – – 67.80 27.74 – 0.44 –CVD-SiC – – 0.01 – – – – – 69.80 30.17 – –

T. Cheng et al. / Journal of Nuclear Materials 427 (2012) 396–400 397

Zircaloy) from 800 to 1200 �C in high-pressure steam environ-ments; conditions for which little literature data is available formany of these material systems.

2. Materials and experimental procedures

The candidate cladding materials evaluated in this study were1-2mm thick coupons of Rohm & Haas chemical-vapor-deposited(CVD) SiC, nano-infiltrated-transient-eutectic (NITE) SiC [7], com-mercial oxide dispersion strengthened (ODS) FeCrAl alloyPM2000 and wrought model binary Fe-xCr alloys (x = 15, 20, and25%) as well as tube forms of Zircaloy-2 (0.9 mm wall thickness),Zircaloy-4 (0.65mm) and stainless steel types 304L, 317L, 321Land 347 (�16-19 wt.% Cr) with wall thicknesses of 0.7-1.6 mm.Analyzed compositions are shown in Table 1. Exposures were con-ducted for 8 h in 100% steam at temperatures of 800 �C, 1000 �Cand 1200 �C and total pressures of 0.34 or 1 MPa (flow rates of94 cm/min and 29 cm/min, respectively, were used to achieve100% steam conditions at the targeted pressures). Details of thetest rig are presented elsewhere [8]. Test samples were analyzedby standard cross-section metallography. The polished cross-sec-tional specimens were characterized using light optical and scan-ning electron microscopy (SEM).

0

100

200

300

400

500

600

Pawel-Cathcart Zircaloy-2 Zircaloy-4

1 fa

ce m

ater

ial r

eces

sion

(µm

)

Entire c

Fig. 1. Material recession of zirconium alloys, 317L steel and NITE SiC after 8 h exposureCVD�) under that condition was approximated based on the results of CVD SiC specimensCathcart correlation [3] is provided for comparison. (Note that the materials loss valuembrittled regions.)

The extent of attack by oxidation was assessed by specific masschange and by thickness of intact remaining material measured insample cross-section by an optical measurement microscope. Be-cause of oxide scale spallation and/or volatility-driven oxide scaleloss, mass change values do not necessarily represent the true ex-tent of oxidation. For the intact thickness measurements, data ispresented as sample 1-face recession data (total material loss di-vided by two). The samples were assessed by dividing the cross-section into 3 regions (ignoring the first few millimeters at edge/corner regions) and visually selecting the location of greatest mate-rials loss from each region, with the average and standard devia-tion (error scale) of the 3 measurements rounded to the nearest1 lm presented.

3. Results and discussion

The 1-face material loss of the candidate cladding materialsafter 8 h exposure at 800, 1000, and 1200 �C in 1 MPa steam areshown in Fig. 1. Corresponding cross-sections of select materialsare shown in Figs. 2–4.

Significant oxidation was observed in the Zircaloy 2 and 4 after8 h exposure at 800 and 1000 �C in 1 MPa steam, with features con-sistent with literature reports of extensive Zr-base oxide formation

317L CVD NITE

1200

1000

800

1200 °C

1000 °C

800 °C

ladding reacted

at 800 �C, 1000 �C and 1200 �C in 1 MPa steam. Materials loss of CVD SiC (labeled asexposed in 0.34 MPa steam. Extent of zirconium alloy oxidation predicted by Pawel–es for the Zircaloys at 800 �C and 1000 �C do not include oxygen-stabilized a-Zr

Fig. 2. Light optical cross-section micrographs of Zircaloy 2 after exposure for 8 h in 1 MPa steam (a) 800 �C, (b) 1000 �C and (c) 1200 �C. Similar observations were made forZircaloy-4 specimen.

Fig. 3. Optical micrographs of cross sections of 317L stainless steel after exposure for 8 h in 1 MPa steam (a) 800 �C, (b) 1000 �C and (c) 1200 �C.

398 T. Cheng et al. / Journal of Nuclear Materials 427 (2012) 396–400

and brittle oxygen-stabilized a-Zr [2]. A minor effect of steam pres-sure (>�3 MPa) on zirconium alloy oxidation kinetics has been pre-viously reported [9] and is known to diminish with increasingtemperature. On this basis, comparison of current test results forzirconium alloys to prior atmospheric steam test data is appropri-ate. The Zircaloy 4 exhibited a modestly lower extent of materialsloss than did Zircaloy 2 (Fig. 1). At 1200 �C, both Zircaloy alloys 2and 4 were completely converted to oxide (Fig. 2c) (the greater va-

lue for materials loss for Zircaloy 2 in Fig. 1 was due to a greaterinitial tubing thickness).

The 300 series stainless steels (304L, 317L, 321L and 347) exhib-ited oxidation behavior similar to each other, consistent with theirsimilar alloy chemistries (Table 1). Data for the best performing ofthese alloys, 317L, is shown in Figs. 1 and 3. The 317L showed en-hanced corrosion resistance when compared to zirconium alloysfor 8 h at 800 and 1000 �C in 1 MPa steam. However, nearly half

Fig. 4. Scanning electron micrographs of cross sections of NITE SiC after exposure in 1 MPa steam for 8 h at (a) 800 �C, (b) 1000 �C and (c) 1200 �C.

1200 ºC, 0.34 MPa

0.1

1

10

100

1000

317L Fe-15Cr Fe-20Cr Fe-25Cr NITE CVD PM2000

1-fa

ce m

ater

ial r

eces

sion

(µ

m)

Fig. 5. Material recession of Fe–Cr, PM2000 alloys and NITE and CVD SiC afterexposure for 8 h at 1200 �C in 0.34 MPa steam.

T. Cheng et al. / Journal of Nuclear Materials 427 (2012) 396–400 399

of the 317L was consumed to oxide at 1200 �C (Fig. 3c). In contrast,the NITE SiC exhibited small levels of materials loss throughout the800–1200 �C temperature range in the pressurized steam. At 800and 1000 �C, material loss for the NITE SiC was insignificant (Figs. 1and 4a and b). More extensive attack of the NITE SiC was observedat 1200 �C, with a �100 lm thick porous silica scale (Fig. 4c) and a1-face materials loss of 29 lm. The CVD SiC showed less recession(Fig. 1), based on an extrapolation of data from exposure in0.34 MPa steam. The CVD data shown in the plot represents thegreatest recession among the extrapolation results based on themechanisms of parabolic oxidation controlled by diffusion and lin-ear recession loss of oxide from volatilization [10]:

Fig. 6. Backscatter mode SEM images of Fe–25Cr (a) and PM2

kP / x2 / ðPH2OÞn ð1Þ

kP / x / ðPH2OÞ3=2t1=2 ð2Þ

where kp is the parabolic oxidation rate constant, kl is the linearrecession rate constant, x is the 1-face material recession, PH2O isthe total stream pressure, n (0.5 6 n 6 1) is the power of the steampressure depending on different diffusion mechanism and t is thegas flow velocity. Here, we assume that the SiC linear recessiondue to silica volatilization is controlled by Si(OH)4 diffusion througha laminar gaseous boundary layer.

Based on the initial screening results indicating unacceptableoxidation resistance for the nominal 16–19 wt.% Cr range 300 ser-ies stainless steels at 1200 �C, a series of model binary Fe–xCr(x = 15, 20, and 25) alloys and PM2000 were selected for study.The materials recession data for these alloys after 8 h exposure at1200 �C in 0.34 MPa steam are shown in Fig. 5. The NITE andCVD SiC and 317L were also included for comparison.

At 1200 �C, the Fe–15Cr alloy was completely consumed by oxi-dation and Fe-20Cr alloy was heavily attacked, indicating thathigher Cr contents are needed under these conditions. Protectivebehavior was observed for the Fe-25Cr alloy specimen with 1-facematerials recession of 10 lm. Further, backscatter SEM imagingindicated an intact Cr-rich oxide scale of �17 lm thickness formedon the Fe-25Cr (Fig. 6a). As expected [4], the alumina-forming PM2000 alloy exhibited the best resistance of all materials studied,with 1-face materials loss of 1 lm and SEM backscatter mode

000 (b) after 8 h exposure at 1200 �C in 0.34 MPa steam.

400 T. Cheng et al. / Journal of Nuclear Materials 427 (2012) 396–400

imaging indicating an intact Al-rich oxide scale of only �3 lmthick consistent with alumina.

For comparison purposes, at 1200 �C–0.34 MPa, the materialloss for 317L and NITE SiC was 158 lm and 20 lm, respectivelyin 8 h. To compare, as the pressure is increased from 0.34 to1 MPa the material loss increases for the 317L and NITE SiC to276 lm and 29 lm, respectively (see Figs. 1 and 5) This result sug-gests enhanced oxidation kinetics with increasing pressure, whichis consistent with the parabolic relation mentioned previously forSiC (holds for Cr2O3-forming Fe-base alloys as well). The NITESiC under the 0.34 MPa steam condition exhibited a material lossof 20 lm as compared to that of CVD SiC (5 lm) (Fig. 5). Replicatesamples and exposure for a range of times are in progress to con-firm those speculations. Detailed study and comparison of thesevarious materials classes as a function of specific alloy/materialsgrade temperature, exposure time, steam-hydrogen mixture, andpressure are in progress to permit greater behavior differentiationand down select for subsequent assessment of uniform corrosionand stress corrosion cracking processes under typical fuel claddingoperating conditions (for irradiated and unirradiated conditions).Neutronics assessment of the iron-based alloys will also be made,as the gains in oxidation resistance over Zirconium alloys willcome at a neutronic penalty for the cladding.

4. Conclusions

A range of current zirconium and conventional 300-series stain-less steel materials were exposed to steam at 800�-1200�C in 0.34-1MPa steam. At 1200�C, these materials were all heavily attacked.

More oxidation resistant materials such as SiC and iron-based al-loys with high Cr (>20 wt.%) or Al that forms a slow-growing alu-mina surface oxide show promise for the highest temperatureaccident conditions for a nuclear reactor core. However, moreextensive evaluation of these candidates properties is needed.

Acknowledgements

The authors thank Dr. Y. Katoh, and Dr. P.F. Tortorelli for helpfulcomments on this manuscript. The work presented in this manu-script was supported under the Advanced Fuel Campaign of theFuel Cycle R&D program at Office of Nuclear Energy, US Depart-ment of Energy. Research supported in part by ORNL’s SharedResearch Equipment (ShaRE) User Facility, which is sponsored bythe Office of Basic Energy Sciences, US Department of Energy.

References

[1] M. Moalem, D.R. Olander, J. Novel, Mater 182 (1991) 170–194.[2] V.F. Urbanic, T.R. Heidrick, J. Nucl. Mater. 75 (1978) 251–261.[3] R.E. Pawel, J.V. Cathcart, R.A. McKee, J. Electrochem. Soc. 126 (1979) 1105–

1111.[4] M.P. Brady, I. Wright, B. Gleeson, JOM J. Miner. Met. Mater. Soc. 52 (2000) 16–

21.[5] E. Opila, N. Jacobson, D. Myers, E. Copland, JOM J. Miner. Met. Mater. Soc. 58

(2006) 22–28.[6] S.R.J. Saunders, M. Monteiro, F. Rizzo, Prog. Mater. Sci. 53 (2008) 775–837.[7] Y. Katoh, S.M. Dong, A. Kohyama, Fusion Eng. Des. 61–62 (2002) 723–731.[8] K.L. More, P.F. Tortorelli, M.K. Ferber, L.R. Walker, J.R. Keiser, N. Miriyala, W.D.

Brentnall, J.R. Price, J. Eng. Gas Turbines Power 122 (2000) 212–218.[9] R.E. Pawel, J.V. Cathcart, J.J. Campbell, J. Nucl. Mater. 82 (1979) 129–139.

[10] E.J. Opila, J. Am. Ceram. Soc. 82 (1999) 625–636.