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Nuclear Engineering and Design 240 (2010) 985–994

Contents lists available at ScienceDirect

Nuclear Engineering and Design

journa l homepage: www.e lsev ier .com/ locate /nucengdes

ffect of dissolved oxygen on oxidation and hydrogen pick upehaviour—Zircaloy vs Zr–Nb alloys

. Kiran Kumara,∗, Sachin Aggarwalb, Vivekanand Kaina, Timo Saarioc, Martin Bojinovd

Materials Science Division, Bhabha Atomic Research Center, Trombay, Mumbai 400085, IndiaDept. of Metallurgical Engineering, Punjab Engineering College, Chandigarh 160012, IndiaVTT Technical Research Centre of Finland, PO Box 1000, Kemistintie 3, FIN-02044 VTT, Espoo, FinlandDepartment of Physical Chemistry, University of Chemical Technology and Metallurgy, 1756 Sofia, Bulgaria

r t i c l e i n f o

rticle history:eceived 16 June 2009eceived in revised form 4 December 2009ccepted 16 December 2009

a b s t r a c t

Upcoming advanced nuclear power reactors contemplate partial boiling inside the core and it is importantto establish the effects of partial boiling on oxidation and hydrogen pick up for sufficiently long periods oftime. The effect of partial boiling on oxidation behaviour was assessed in this study in terms of change inthe level of dissolved oxygen (DO) in static autoclaves. The results clearly demonstrated that the oxidationbehaviour of Zircaloy-2 is quite immune to the level of dissolved oxygen, while Zr–Nb alloys showedvery different oxidation behaviour with change in the dissolved oxygen level. The weight gain due tooxidation in Zr–Nb alloys during long term exposure was almost doubled with increase in the DO level.Nevertheless, in case of all the three alloys, the corresponding hydrogen pick up was found to be lower formaterials oxidized at higher DO levels irrespective of the effect of DO level changes on weight gain. This

is attributed to the preferential reaction of the metal with free oxygen rather than water to produce theoxide when the DO levels are high. Further, the oxidation behaviour of Zr–1 wt%Nb alloy, as a functionof dissolved oxygen content in lithiated water, was monitored in situ by electrochemical impedancespectroscopic (EIS) measurements using controlled displacement electrochemistry (CDE) technique in arecirculation loop at reactor operation temperature. The oxidation behaviour (for upto 90 h duration at310 ◦C) in high DO content of water was in good correlation with the results obtained from accelerated

ring

oxidation experiments du

. Introduction

The industrial importance of zirconium based alloys is mainlyn the context of their nuclear applications. Owing to their loweutron absorption cross-section coupled with desirable mechani-al and electrochemical properties, zirconium alloys (both singlend two-phase alloys) are used as structural materials such asressure tube and as fuel cladding in nuclear reactors (Ganguly,002; Sundaram, 1986). Though these alloys have a better corrosionesistance compared to many other materials, uniform oxidation,s well as nodular corrosion (Shah et al., 2006) at high temper-ture and high pressure water/steam (for boiling water reactors,WRs) and hydrogen embrittlement (for pressurized water reac-

ors, PWRs/pressurized heavy water reactors, PHWRs) remain the

ain concerns and are considered limiting factors of the serviceife of components. Zircaloy-2 (UNS R60802) (B 353 ASTM, 2007nd B 811 ASTM, 2007) is currently used as a fuel cladding mate-

∗ Corresponding author.E-mail address: mkiran@barc.gov.in (M. Kiran Kumar).

029-5493/$ – see front matter © 2010 Elsevier B.V. All rights reserved.oi:10.1016/j.nucengdes.2009.12.021

short term exposure (up to 24 h at 400 ◦C) in static autoclaves.© 2010 Elsevier B.V. All rights reserved.

rial for BWR, in which conditions nodular corrosion (acceleratedlocalized oxidation) is the main concern. In case of PHWR, wherehydrogen embrittlement is the major concern, Zircaloy-4 (UNSR60804) (B 353 ASTM, 2007; B 811 ASTM, 2007), a lower-nickel ver-sion of Zircaloy-2 is being used (Ganguly, 2002; Sundaram, 1986).Upcoming advanced reactors are considering the use of similarwater chemistry as that of current PHWRs but with partial boil-ing being allowed inside the core to achieve a better power rating.The advanced heavy water reactor (AHWR) has light water as aprimary coolant and is based on the concept of partial boiling inthe core (Sinha and Kakodkar, 2002, 2006). Such conditions areexpected to change the water chemistry, in terms of the level ofdissolved oxygen (DO) present as well as other species like lithiumion concentration.

BWRs rely on boiling inside the reactor to directly produce steam(Adler and Schemker, 1993; IAEA Tecdoc-996, 1998; Lin et al., 1992;

Cox, 1968; Sava, 1998; Amaev et al., 1970). Radiolysis of water inthe core of the reactor leads to formation of oxidizing species andthe equivalent dissolved oxygen levels increase. Though hydrogeninjection is commonly used to suppress the oxidizing species, itbecomes less effective when boiling takes place. Therefore, a direct

986 M. Kiran Kumar et al. / Nuclear Engineering and Design 240 (2010) 985–994

Table 1Chemical composition (B 353 ASTM, 2007 and B 811 ASTM, 2007) of the three categories of zirconium based alloys used in the present investigation. Single figures indicatemaximum allowable values.

Material Chemical composition (wt%)

Zr Nb Sn Fe Cr Ni N C O

Zr-2 Base 0.01 1.2–1.7 0.07–0.20 0.05–0.15 0.03–0.08 0.008 0.027 0.09–0.16

2007

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aZr–1Nb Base 0.9–1.1 0.05 0.05Zr–2.5Nb Base 2.4–2.8 0.01 0.15

a Typical composition only, not incorporated in the ASTM standard (B 353 ASTM,

onsequence of boiling (or partial boiling) is an increase in the dis-olved oxygen levels (Adler and Schemker, 1993; IAEA Tecdoc-996,998, Lin et al., 1992; Cox, 1968; Sava, 1998; Amaev et al., 1970) inhe primary coolant.

Different alloys are expected to respond differently, in terms ofheir oxidation behaviour, to the level of dissolved oxygen. Suchifferences may arise from a difference in alloy composition andhases (IAEA Tecdoc-996, 1998) present. Certain phases/elementshat have a higher affinity for oxygen/oxidation potential will havehigher weight gain with increase in DO level. For example, an

mportant difference between the corrosion of Zr–Nb alloys and theircaloys in the laboratory tests (IAEA Tecdoc-996, 1998; Cox, 1968;maev et al., 1970) was that the former showed increased corrosionates in the presence of traces of oxygen in the water, whereas theatter did not. Therefore, it is important to establish the relativeesponse of various zirconium alloys by studying their oxidationehaviour at different DO levels. Initial tests had shown differences

n oxidation behaviour of Zircaloy-2 and Zr–1 wt%Nb/Zr–2.5 wt%Nblloys in 400 ◦C steam/3 days tests (G2 M ASTM, 2005; Kain et al.,005) with different levels of DO.

The present study aims at establishing comparative oxidationinetics of single-phase Zircaloy-2 and two-phase Zr–Nb alloysboth Zr–2.5 wt%Nb and Zr–1 wt%Nb) in the partial boiling condi-ions. The effect of partial boiling was assessed in terms of changen the level of dissolved oxygen in static autoclaves. Tests in thetatic autoclaves were carried out in the steam phase, where thepecimens were exposed to steam with or without dissolved oxy-en, though the exact quantification of the level of DO was notossible. Further, the oxidation behaviour of one of these materi-ls (Zr–1 wt%Nb) was monitored in situ through electrochemicalmpedance spectroscopy (EIS) using controlled distance electro-hemistry (CDE) in an autoclave attached to a recirculation loophere a better and quantitative control of DO level was possible.

uch results from in situ studies were then related to results fromccelerated oxidation (weight gain) tests with short term exposuresn static autoclaves. Scanning electron microscopy (SEM) was usedo study the morphology of the oxide layers. The effects of changesn DO levels on the oxidation and the corresponding hydrogenick-up behaviour are also discussed in the present paper. Mecha-istic aspects of the observed differences are also attempted to bexplained.

. Materials and methodology

.1. Materials

The materials that were used for the present study are

(a) Zircaloy-2 and Zr–2.5 wt%Nb pressure tube materials.b) Zr–1 wt%Nb fuel clad material.

The chemical composition of these alloys is shown in Table 1.The test specimens of approximately 20 × 20 mm were cut from

ircaloy-2 (82.4 mm ID and 91 mm OD) and Zr–2.5 wt%Nb pres-ure tubes (82.6 mm ID and 90 mm OD). In case of Zr–1 wt%Nb fuel

0.02 0.02 0.006 0.02 0.10.02 0.007 0.008 0.027 0.09–0.15

and B 811 ASTM, 2007).

cladding, tubes of dimensions of 4.8 mm ID and 5.8 mm OD andring pieces of 20 mm length were used as test specimens. The spec-imens for oxidation were mechanically polished up to 600 grit andthen etched chemically. The etchant used (G2 M ASTM, 2005) forZircaloy-2 was 3% HF + 39% HNO3 + 58% water; while a solution of30% HNO3 + 30% H2SO4 + 31% water + 9%HF was used for the etchingof Zr–Nb alloys.

2.2. Experimental methods

2.2.1. Oxidation in static autoclavesOxidation tests were carried out in static autoclaves (made of

Inconel) in steam at a temperature of 400 ◦C and a pressure of10 MPa according to the ASTM standard test (G2 M ASTM, 2005).This test is an accelerated test used for qualifying zirconium basedalloys for use in nuclear reactor applications. It specifies use ofwater with a deaeration practice to reduce dissolved oxygen to lowlevels. The deaeration was achieved by first purging the water withAr (97 vol%) and then venting the autoclave several times while thetemperature is increased. In the present study, other than the G-2M standard test, another version of the same test with no deaera-tion (venting) was also used (Kain et al., 2005). The purpose of sucha modification was to maintain a high level of dissolved oxygen.Though the exact determination of DO level in static autoclaves isnot possible, the level that is attained when performing the G2 Mtest as per the prescribed deaeration practice is much lower than 45ppb. In the variation of the test when no deaeration is done, the DOlevels are considerably higher (the solubility of oxygen in water atambient temperature being typically 6–8 ppm, although the exactquantification at test conditions were not possible). Further, con-sidering the consumption of DO with time during the reaction, theautoclave was periodically (after each 6 days) refreshed with waterfor long term tests. Such an environment is expected to simulatethe condition of partial boiling (i.e. its effect on DO). The tests withdifferent DO levels were carried out in two separate autoclaves.The procedure described in the standard test (G2 M ASTM, 2005)for measuring weight change was followed. The weight was mea-sured with an accuracy of 0.001 mg. The weight gain per unit areawas calculated from the weight gain and exposed area of the testspecimens for different durations ranging from 7 h to 90 days.

2.2.2. In situ controlled distance electrochemistryTests were performed in an AISI 316 autoclave connected to a re-

circulating loop. Two test runs were performed in normal reactorprimary water at 310 ◦C, 10 MPa (2.2 ppm Li as LiOH, 1126 ppm B asH3BO3), pH(310 ◦C) = 7.04, and under an overpressure of 1.5 bar H2in the feed water mixing tank (input level of DH ≈ 30 cm3/kg STP).Another set of two test runs were performed in otherwise similarconditions, but with the dissolved oxygen maintained at 300 ppb(±1 ppb). The main test method was electrochemical impedance

spectroscopy (EIS), using a counter electrode made of Ir (99%). Thespecific set-up used was the controlled distance electrochemistry(CDE)—device with a distance of about 10 �m between the speci-men and the counter electrode (Bojinov et al., 2000, 2005a,b, 2008).The measurements were performed in a two-electrode mode and

M. Kiran Kumar et al. / Nuclear Engineering and Design 240 (2010) 985–994 987

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ig. 1. Appearance of oxide layer grown in 400 ◦C steam at 10 MPa pressure for 60 daxygen levels.

he counter electrode also served as a low-impedance fast responseeference electrode. Electrochemical impedance spectra were mea-ured with a Solartron 1287/1260 system controlled by ZPlotoftware (Scribner Associates) in a frequency range of 10−3 to04 Hz at an AC amplitude of 20 mV. The spectra were analyzedsing the equivalent circuit approach of the ZView software (Scrib-er Associates) to arrive at estimates of parameters related to thexidation rate (especially the zero-frequency limit of the Warburgmpedance, i.e. the ion transport resistance).

.2.3. Other characterizationsThe hydrogen content in all the oxidized specimens after each

xposure was analyzed using a standard inert gas fusion (IGF) tech-ique (E1447 ASTM, 2005). Scanning electron microscopy (SEM)as applied on gold sputtered oxides to observe the oxidized sur-

ace morphology.

. Results

The results of the investigation are presented in four subsectionsealing with (a) the typical appearance of the oxidized specimens,b) the actual oxidation kinetics in static autoclaves demonstrat-ng the effect of dissolved oxygen level along with correspondingydrogen pick-up rates, (c) in situ electrochemical impedance spec-roscopic observations using CDE and (d) electron microscopicbservations as regard to the morphology of oxidation.

.1. Appearance of the oxidized surface

The physical appearance of oxide (ZrO2) grown on the alloyurface at various stages of oxidation is expected to provide a pre-iminary qualitative index of the protective nature of the oxide. Athe early stages of oxidation, the surface oxide layer, irrespectivef the type of material, was quite adherent and had a black andlossy (shiny) appearance. Such black lustrous appearance contin-ed to be observed till 30 days of exposure at 400 ◦C for all the threelloys. Clear differences in the appearance of the oxides on differentlloys started to appear after about 60 days of exposure, as shownn Fig. 1. The specimens of both Zircaloy-2 as well as the Zr–Nblloys were still lustrous black when tested in steam with very lowissolved oxygen content. The oxide formed on Zircaloy-2 speci-ens did have a similar black lustrous appearance both at high and

ow DO levels, even at exposures longer than 60 days. On the otherand, in case of Zr–2.5 wt%Nb and Zr–1 wt%Nb, there was a signifi-ant difference in appearance at low and high DO levels. The oxides

ormed at high DO levels had a non-uniform surface appearance—alack lustrous layer with specs of white oxide uniformly distributedver it. These white specs of oxide had a plate/needle-like (butrregular in shape) appearance in case of Zr–2.5 wt%Nb, while they

ere present as uniform spots on the surface of Zr–1 wt%Nb tubes.

(a) Zircaloy-2, (b) Zr–1 wt%Nb and (c) Zr–2.5 wt%Nb alloys at low and high dissolved

These observations give a preliminary indication that at high DOcontent and exposure periods longer than 60 days, both Zr–Nballoys had a tendency to oxidize at a faster rate and form somenon-uniform non-protective surface oxide in the process (IAEATecdoc-996, 1998; Dyce, 1964). On the other hand, oxidation ofZircaloy-2 appeared to remain unaffected by changes in the DOlevel.

3.2. Oxidation kinetics and hydrogen pick-up

3.2.1. Oxidation kineticsThe oxidation behaviour at each stage was quantified from the

weight gain per unit area (expressed in mg/dm2). Such values werecompared, in case of each alloy, for low and high level of dissolvedoxygen in the steam. The results obtained for all the three alloys(Zircaloy-2, Zr–1 wt%Nb and Zr–2.5 wt%Nb) are presented in Fig. 2together with the empirical oxidation rate (mg/dm2 d) calculatedby numerical differentiation of mass gain data. It is to be noted thatin case of Zircaloy-2 a weight gain of 14.7 mg/dm2 was consideredequivalent to an oxide thickness of 1 �m (IAEA Tecdoc-996, 1998;Hillner, 1977).

3.2.1.1. Oxidation behaviour of Zircaloy-2. The weight gain forZircaloy-2 remained low in the initial duration for both sets ofexposures (low DO and high DO). The weight gain remained nearlysimilar at both DO levels with no significant alterations. The weightgain was 20.8 mg/dm2 after 3 days exposure and 34.8 mg/dm2 after14 days exposure when the DO levels were kept low. These valueswere 17.7 and 35.2 mg/dm2 respectively when the DO values werekept high. Upto about 21 days of exposure, when the weight gainwas ∼41 mg/dm2, the oxidation kinetics followed a rate law thatis close to cubic, the data upto 21 days in Fig. 2a can best be fit-ted to Eq. (1). Beyond this period, there was a transition observedfrom near cubic to linear weight gain with a linear rate constantof 0.519, the data being fitted to Eq. (3). The kinetic parameters forboth pre- and post-transition oxidation estimated by curve fittingare listed in Table 2, along with the same parameters arrived atby other researchers (Dyce, 1964; Van der Linde, 1965; Dalgaard,1976; Hillner, 1977). It can be summarized that the effect of DO onweight gains was insignificant for Zircaloy-2.

The kinetic equation for the pre-transition oxidation of Zircaloy-2 is

log �W = 0.37 log t + 1.149 or �W2.84 = 1784.32t (1)

whereas a general cubic pre-transition kinetics is expressed by the

equation

�W3 = kct (2)

where �W = specimen weight gain in mg/dm2, t = exposure time indays and kc = pre-transition rate constant.

988 M. Kiran Kumar et al. / Nuclear Engineering and Design 240 (2010) 985–994

Fig. 2. Weight gain (left ordinate) and oxidation rate (right ordinate) as a function of exposure time, in 400 ◦C steam at 10 MPa pressure, of (a) Zircaloy-2, (b) Zr–2.5 wt%Nband (c) Zr–1 wt%Nb alloys at low and high dissolved oxygen levels.

Table 2Kinetic parameters for the oxidation of Zircaloy-2 in static autoclaves at 400 ◦C and 10 MPa pressure steam (for both the low DO and high DO conditions).

References Pre-transition, rate constant, kc (mg/dm2)2.84 d−1 Post-transition, rate constant, kL (mg/dm2 d−1) Weight gain at transition, �W (mg/dm)

Van der Linde (1965) 1569 1.18 3864219

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DO level was much more evident, see Fig. 2. Further, at low DO con-ditions both the Zr–Nb alloys followed an oxidation (weight gain)kinetic law that was in between cubic and parabolic (please see thedata being fitted to Eqs. (5) and (6)), whereas the kinetic law was

Table 3Weight gain (in mg/dm2) for all three Zr alloys studied at low and high DO conditionsin steam at 400 ◦C and 10 MPa pressure.

Exposure time (h) DO levels Zircaloy-2 Zr–2.5 wt%Nb Zr–1 wt%Nb

7 Low DO 8.94 12.87 11.69High DO 7.9 4.88 7.38

18 Low DO 12.29 14.33 12.57High DO 8.31 5.6 8.0

Dyce (1964) 1728 1.2Dalgaard (1976) – 1.1Hillner (1977) 1014 0.9Current study 1784.3 0.5

The kinetic equation for the post-transition oxidation ofircaloy-2 is

W = 0.519t + 31.59 (3)

On the other hand, a general linear post-transition kinetics isxpressed by the equation

W = kLt (4)

here kL is the linear post-transition rate constant.

.2.1.2. Oxidation behaviour of Zr–Nb alloys. The difference in oxi-ation behaviour upon changing the DO level was apparent forr–1 wt%Nb and Zr–2.5 wt%Nb alloys. It is important to note herehat there were distinct trends as regards to the effect of DO at initialshort) exposures of upto 18 h and at higher exposures (see Table 3or exact values). For the initial exposures of 7–18 h the weightain was lower when exposed to higher DO conditions for theame period. This kind of behaviour was also observed through in

itu electrochemical impedance spectroscopy measurements usingDE during exposures in recirculation loop; described in Section.2.2.

At longer exposures, the weight gain in steam always wentp with increased DO level. For Zr–2.5 wt%Nb at low DO level

40–

3341.51

(deaerated condition), the weight gain values were 21.5 and37.8 mg/dm2 after 3 and 14 days, respectively, increasing to 38.4and 60.4 mg/dm2 at high DO level. A similar trend was observedfor Zr–1 wt%Nb alloy (Table 3). At still longer periods of exposure,the difference in oxidation behaviour of Zr–Nb alloys with change in

24 Low DO 14.55 16.93 14.17High DO 13.79 17.04 19.99

72 Low DO 20.83 21.51 16.24High DO 17.78 38.36 27.72

M. Kiran Kumar et al. / Nuclear Engineering and Design 240 (2010) 985–994 989

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ig. 3. Hydrogen content as a function of exposure time, in 400 ◦C steam at 10 MPaissolved oxygen levels.

lose to parabolic at high DO conditions (please see the data beingtted to Eqs. (7) and (8)).

Low DO conditions:For Zr–2.5 wt%Nb,

og �W = 0.37 log t + 1.149 or �W2.7 = 1278.88t (5)

or Zr–1 wt%Nb,

og �W = 0.38 log t + 1.029 or �W2.63 = 512.83t (6)

High DO conditions:For Zr–2.5 wt%Nb,

og �W = 0.52 log t + 1.196 or �W1.92 = 199.68t (7)

or Zr–1 wt%Nb,

og �W = 0.56 log t + 1.02 or �W1.79 = 65.82t (8)

here �W is in mg/dm2 and t is in hours.As a conclusion, it can be mentioned that for both Zr–2.5 wt%Nb

nd Zr–1 wt%Nb alloys, the weight gain almost doubled withncrease in DO level, while the difference was insignificant forircaloy-2.

.2.2. Hydrogen pick upThe oxidized Zr alloys after each exposure time at 400 ◦C and

0 MPa steam were analyzed for the hydrogen content by IGF tech-ique (E1447 ASTM, 2005). The objective was to study the effect ofhange in the dissolved oxygen level of the steam on the hydrogenick up behaviour in the alloys studied. The values of the hydro-

en content after each exposure for all three alloys (Zircaloy-2 andoth Zr–Nb alloys) in both high and low DO conditions are plotted

n Fig. 3. The results can be used for comparison of the hydrogenptake between different alloys and also to assess the effect of DO

evel change for a specific alloy. It can be noted that among all the

sure, of (a) Zircaloy-2, (b) Zr–2.5 wt%Nb and (c) Zr–1 wt%Nb alloys at low and high

three alloys studied, Zr–2.5 wt%Nb did show the lowest hydrogenpick up which remained low even at longer exposures. On the otherhand, both Zircaloy-2 and Zr–1 wt%Nb showed higher hydrogencontent with increasing exposure time.

Regarding the effect of DO level change on the hydrogen pickup of a given alloy (Fig. 3), an interesting trend is observed in con-nection to the weight gain data (Fig. 2). For all the three alloysstudied, the hydrogen content was invariably lower when the dis-solved oxygen was kept higher for the same exposure time. Generalunderstanding is that higher the weight gain, higher is the hydro-gen pick up for a given alloy, as the hydrogen production duringoxidation reaction is more. However, hydrogen pick up is also afunction of alloy composition. In the present investigation, for theZr–Nb alloys, in spite of the weight gain being higher at high DOconditions, the hydrogen content was lower. On the other hand, incase of Zircaloy-2, though the change in DO level led to insignifi-cant alteration in the weight gain, the hydrogen content with DOchange followed a similar pattern as that of Zr–Nb alloys. Thiscan be attributed to the fact that oxygen, in high DO conditions,directly reacts with Zr to produce oxide, but there is no asso-ciated hydrogen pick-up as hydrogen is not a product in suchreaction. This point is further elaborated in the discussion segment,Section 4.

3.2.3. In situ electrochemical impedance spectroscopyIn order to have an estimation of electronic and ionic trans-

port properties at initial stages of oxidation and to establish theeffect of DO level changes on the oxidation behaviour, in situ EISmeasurements were carried out using CDE at high temperatureand pressure (Bojinov et al., 2005a,b, 2008). Such measurementswere done only on Zr–1 wt%Nb. The DO levels were quantitatively

990 M. Kiran Kumar et al. / Nuclear Engineering and Design 240 (2010) 985–994

Fig. 4. Comparison of the EIS test results for Zr–1 wt%Nb when hydrogen wasreplaced by 300 ppb O2. Data measured after 90 h of exposure.

Fp

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Bgoaf(celYeWtwTit

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ig. 5. The equivalent circuit used in the modelling phase to describe the oxidationrocess and to fit the experimental data.

ontrolled through the attached recirculation loop as described inection 2.2.2.

The impedance spectroscopy results are shown in Fig. 4 as aode plot. In the second test run with 300 ppb dissolved oxy-en there was extensive disturbance at the lower frequency endf the spectrum. Thus, data for this run is only shown down tobout 4 Hz. The equivalent circuit (Bojinov et al., 2005a,b) usedor the analysis is shown in Fig. 5. It comprises of a resistance R1the electrolyte resistance), a resistance R3 in combination with aonstant phase element CPE1 (the resistive and capacitive prop-rties of the presumably porous outer oxide layer and the doubleayer at the oxide/water interface), a charge transfer resistance R2,oung’s impedance DX1 (corresponding to the capacitive prop-rties of the dense inner oxide layer, i.e. the barrier layer) and aarburg impedance W1 (the impedance attributed to the mass

ransfer through the barrier layer). Fig. 6 shows two of the test runs

ith modelling result for the first test run with dissolved oxygen.

he fit is rather good, giving confidence in the calculated Warburgmpedance values. The modelling results shown in Table 4 indicatehat the zero-frequency limit of the Warburg impedance, i.e. the

able 4odelling results for the zero-frequency limit of the Warburg impedance, calculated

rom the CDE experiments.

Test run Zero-frequency limit of the Warburg impedance (ionictransport resistance)/� cm2

Zr1Nb H2-1 3890Zr1Nb H2-2 7120Zr1Nb O2-1 12,950Zr1Nb O2-2 13,160

Fig. 6. Comparison of the EIS test results for Zr–1 wt%Nb in normal PWR conditions(H2) and when hydrogen was replaced by 300 ppb O2. One test run is shown foreach condition for clarity. The red lines show the modelling result. Data measuredafter 90 h exposure.

ion transport resistance, which could be taken as inversely propor-tional to the rate of ion transfer through the barrier layer, is clearlyhigher in presence of 300 ppb dissolved oxygen than without it.Normally, higher the Warburg impedance, slower the mass trans-fer thus slower is the oxide growth. Thus, based on the preliminarymodelling of the EIS results it can be argued that replacing totallydeaerated condition by 300 ppb of dissolved oxygen in PWR waterat 310 ◦C slows down the ion transfer through the oxide layer forZr–1 wt%Nb alloy (for the initial stages of oxidation). A SEM inves-tigation was performed at three different locations (as an averageof more than 20 measurements) on both types of specimen thatexposed to PWR water with H2 and that exposed to PWR water with300 ppb O2. The average thickness was 0.54 �m for PWR water withH2 (exposure time 90 h, i.e. 6 nm/h) and 0.58 �m for PWR waterwith 300 ppb O2 (exposure time 190 h, i.e. 3 nm/h). Using a crudeestimation and assuming a linear growth rate these results indi-cate that the growth rate is a factor of two lower in presence of 300ppb O2. In reality the growth most probably follows a parabolic ora cubic law instead of a linear one. Similar differences (for shortterm exposure at 400 ◦C steam) in weight gain data is visible forZr–1 wt%Nb alloy with differences in DO level (Table 3).

3.2.4. Morphological observationsThe explanation for the above observations demonstrating dif-

ferent oxidation behaviour for Zircaloy-2 and Zr–Nb alloys withchanging dissolved oxygen levels, can be sought from microscopicobservations on the oxide layer. In Fig. 7, surface oxide morpholo-gies of Zr–Nb alloys, after an exposure of 60 days at high levelof dissolved oxygen, are presented. Fig. 7a shows the surface ofZr–1 wt%Nb alloy. The earlier observation (Section 3.1) of the whitespot-like specs on the Zr–1 wt%Nb surface is visible and the appear-ance indicates spallation of the whitish spots from the surface.On the other hand, the Zr–2.5 wt%Nb surface shows interestingmorphology after 60 days exposure at high DO levels. In Section3.1, the same surface had lustrous oxide with localized well dis-

tributed white needle-like specs of oxide over it. These needles,when examined under SEM (see Fig. 7b and c) were seen as thickwhite oxide and the associated darker features indicated spallationof the thick oxide. Further, Fig. 8 presents the oxide morphology

M. Kiran Kumar et al. / Nuclear Engineering and Design 240 (2010) 985–994 991

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ig. 7. Surface of the oxide layer grown on (a) Zr–1 wt%Nb, (b) and (c) Zr–2.5 wt%Nbf dissolved oxygen.

n Zr–2.5 wt%Nb alloy at higher magnifications. These micrographseveal clearly the pattern of localized oxidation and associated spal-ation of the thick oxide. The morphologies of the thick oxidesn both Zr–1 wt%Nb (spots) and Zr–2.5 wt%Nb (irregular needlehaped) appear close to the morphologies of second phase � inhe respective alloys (Kiran Kumar et al., 2004; Srivastava et al.,995). The localized accelerated oxidation in both the Zr–Nb alloysre interpreted to be caused due to accelerated oxidation aroundb-rich � phase and are discussed in more detail in Section 4.1.

. Discussion

.1. Mechanistic aspects

The perspective with which the above results are viewed ishreefold:

a) The oxidation kinetics, i.e. the change of weight gain as a func-tion of exposure time for each individual alloy and influence of

the level of dissolved oxygen (in the accelerated test in steam)on the above behaviour.

b) Comparative view of the effect of DO level on the oxidationbehaviour and hydrogen uptake behaviour among the three(Zircaloy-2, Zr–1 wt%Nb and Zr–2.5 wt%Nb) alloys.

in 400 ◦C steam at 10 MPa pressure for an exposure period of 60 days at high levels

(c) Correlation between kinetic parameters through in situ CDE-EIS measurements in lithiated water at 310 ◦C with acceleratedweight gain in static autoclaves at 400 ◦C steam.

In the case of single-phase Zircaloy-2 material, in the initial (lessthan 10 days exposure period) stages of oxidation, the weight gainof 20–30 mg/dm2 was quite low and same for both the DO levels(Table 3). At higher exposure durations, the oxidation kinetic fol-lows a similar path and the difference of weight gain is also minimalbetween high and low dissolved oxygen conditions (Fig. 2a). Therate constant Kc and KL (for both low and high DO conditions) cal-culated (Table 2) from the oxidation results obtained in this study(Fig. 2a) matches well with that reported by other researchers. Theweight gain that was obtained for 90 days exposure at a low DOlevel was 78 mg/dm2 while for the high DO condition, the weightgain was 64.8 mg/dm2 in 60 days.

On the other hand, both Zr–2.5 wt%Nb and Zr–1 wt%Nb hadquite a different kinetics with change in the DO levels. At lowDO levels, these two alloys showed, by and large, a similar trendin the change of weight gain (Fig. 2b and c). When the dissolved

oxygen content was increased, the weight gain almost doubledfor Zr–2.5 wt%Nb. In the case of Zr–1 wt%Nb alloy, the extent ofincrease in the weight gain with high DO level was even largercompared to Zr–2.5 wt%Nb. Nevertheless, the values of weight gainwere always higher in Zr–2.5 wt%Nb than in Zr–1 wt%Nb alloy.

992 M. Kiran Kumar et al. / Nuclear Engineering and Design 240 (2010) 985–994

F 0 MP

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ig. 8. Surface of the oxide layer grown on Zr–2.5 wt%Nb alloys in 400 ◦C steam at 1

hese observations are consistent with the results on the surfaceorphology of the oxide layers, demonstrating that Zr–Nb alloys

howed a non-uniform preferential oxidation for 60 days exposurend beyond. These are broad observations not considering thenfluence of different fabrication routes employed (Ganguly, 2002)or pressure tube (Zr–2.5 wt%Nb) and the fuel clad (Zr–1 wt%Nb).arlier, in 3 days exposure tests at 10 and 20 MPa, 400 ◦C steam,imilar results were obtained (Kain et al., 2005).

Oxidation at high temperature and high pressure water/steamf zirconium based alloys has been studied for many years (IAEAecdoc-996, 1998). The traditional approach to such studies haseen oxidation in static autoclaves followed by weight gain mea-urements. ASTM G-2 M procedure has been adopted and used as aualification test for structural components such as pressure tubend fuel clad materials. Such test has also been made part of the fab-ication schedule and the current alloys, which are expected to stayn the reactor for quite a long time, are serving the purpose well.owever, the direct correlation between the weight gain from theseualifications tests (3 and 14 days exposures as per G-2 M standardrocedure) with the in-reactor performance has not been clearlyemonstrated, as this test remain at best a test to assess compar-tive oxidation behaviour in deaerated condition. Advanced waterooled power reactors, worldwide envisage partial boiling of theater in the core while the chemistry of water phase remains the

ame (alkaline/demineralized). From the corrosion point of view,direct consequence of such an alteration is that at regions whereoiling takes place, the DO level increases. The oxidation of zirco-ium based alloys is viewed mechanistically as an electrochemicalrocess (Bojinov et al., 2005a,b). The zirconium oxide is an n-type

a pressure for an exposure period of 60 days at high levels of dissolved oxygen.

oxide involving transport of oxygen ions and probably also zirco-nium ions in the zirconium oxide matrix (Shah et al., 2006; IAEATecdoc-996, 1998; Bojinov et al., 2005a,b). Eqs. (9)–(11) representthe reactions involved in the oxidation. Zirconium oxidizes eitherby reacting with water or by direct combination with the dissolvedoxygen present.

At the alloy/oxide interface : Zrm → ZrZr + 2VO•• + 4e− (9)

At the oxide/water interface : 2VO•• + 2H2O + 4e− → 2OO+ 4H

(10)

2VO•• + O2 + 4e− → 2OO (11)

where Zrm represents a zirconium atom in the alloy,ZrZr and OO are respectively the zirconium and oxygen ion posi-

tions in the ZrO2 lattice, VO••

represents an oxygen vacancy in thelattice.

In case reaction (10) controls the oxidation, it’s invariablyinvolved with nascent hydrogen formation and hydrogen pick up.However, if reaction (11) dominates, the oxidation process will con-tinue with lower nascent hydrogen formation. Both oxidation and

hydrogen pick up and consequent hydrogen embrittlement is con-sidered life limiting factors for these components. So, in order toqualify any material for these modern reactor environments (par-tial boiling) it is required that they should be tested at varying levelsof dissolved oxygen.

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M. Kiran Kumar et al. / Nuclear Eng

The aim of the current study was to compare the effectf dissolved oxygen on oxidation behaviour and hydrogen pickp behaviour of various alloys (Zircaloy-2, Zr–1 wt%Nb andr–2.5 wt%Nb). The change in oxidation behaviour of Zircaloy-2as found to be insignificant with changing DO level. On the otherand, large differences in oxidation behaviour upon a change in DO

evel have been observed for both the Zr–Nb alloys under study.he weight gain, in these cases, increased almost twice at high DOevels.

The explanation for such an observation may lie in the contextf Nb-rich precipitates/phase such as the � phase in these alloys.hermodynamic data (Kubaschewski and Alcock, 1979) suggesthat niobium has a lower tendency towards oxidation than zir-onium. The Nb rich �-phase in Zr–1 wt%Nb and in Zr-2.5wt%Nbill have Nb depleted matrix around these microstructural fea-

ures. This Nb depleted matrix around the second phase is likelyo have more oxidation tendency than the remaining part of the

atrix that has a uniform distribution of Nb (Kubaschewski andlcock, 1979). However, this remains to be demonstrated experi-entally. As we can see in the Figs. 7 and 8, the thicker oxides are

xpected to form at the locations adjacent to the �-phase regions.or Zr–1 wt%Nb the fine � precipitates are reported (Neogy et al.,003) to be smaller than 0.1 �m size. However in Zr–2.5 wt%Nb, thephase is in the shape of a bamboo structure with elongated Nb

ich morphology (Kiran Kumar et al., 2004; Srivastava et al., 1995).he results shown in Fig. 1b and c show formation of thicker oxidesorresponding to these morphologies in Zr–1 wt%Nb (rounded size)nd Zr–2.5 wt%Nb (elongated shape). These observations indicateormation of thicker oxides directly near the locations of Nb richhase (i.e. above the Nb depleted matrix around these phases). TheEM micrographs in Figs. 7 and 8 confirm this observation, though ane to one correspondence of the shape and size of the � phase (andhe Nb depleted matrix around it) and the oxide formed over it wasot obtained. Almost all the black regions in Fig. 7b are associatedith regions where thick oxide had formed and oxide spallationad taken place. Such an observation can be attributed to the localevelopment of compressive residual stresses at the thicker oxidesnd subsequent spallation at adjacent regions, an aspect that haseen investigated in detail by other researchers (Bradhurst andeuer, 1970; Cox, 1971) earlier for general oxidation of zirconiumased alloys.

The hydrogen pick up behaviour, on the other hand showed annteresting trend (Fig. 3). When the steam had high dissolved oxy-en present, the hydrogen content in the oxidized specimens wasound to be lower than that in the presence of small amounts ofissolved oxygen. Zirconium can oxidize by either of the reactions10) or (11). In other words, it can react with either water or freexygen to produce oxide. Reaction (11) is expected to be the rateetermining step when excess DO is available in water/steam. Inuch a case, the generation of nascent hydrogen is less as comparedo reaction (11) with water. This does explain the lower hydrogenontent for oxidized material at higher DO conditions for all thehree alloys (Fig. 3). Further, the effect of change in DO levels onydrogen content of a given alloy did not depend on their weightain per se. For the Zr–Nb alloys, in spite of the weight gain beingigher at high DO conditions, the hydrogen content remained low.n the other hand, in case of Zircaloy-2, though the change in DO

evel led to insignificant alteration in the weight gain, the hydrogenontent with DO change followed a similar pattern as that of Zr–Nblloys.

The objectives of the in situ EIS measurements (for Zr–2.5 wt%Nb

lloy) leading to determination of kinetic parameters were theollowing. First, a better/quantitative DO control through theecirculation loop is possible. Secondly, estimation of parameterselated to the kinetics based on in situ measurements during thectual growth of oxide layer does give a better picture of oxida-

g and Design 240 (2010) 985–994 993

tion mechanism (Saario et al., 1998; Bojinov et al., 2000, 2001,2005a,b, 2008; Betova et al., 2001). Thirdly, to relate the trendsobtained in short term oxidation exposures in static autoclave withthe same obtained in terms of the kinetic parameters from theCDE-EIS measurements (as the CDE measurements were done fora maximum of 90 h and at lower temperature, 310 ◦C and in lithi-ated water). It can be noted here that lithiated water does enhancethe weight gain (IAEA Tecdoc-996, 1998; Zhou et al., 2008; Albinetet al., 2000; Oscarsson et al., 2001). As can be seen from Table 3for oxidation of both the Zr–Nb alloys, there is a distinct differ-ence in the effect of DO up to 18 h of exposure and beyond thatvalue. Increasing DO level had a reducing effect on the weight gainduring the initial exposures of at least 18 h. With regard to the insitu CDE-EIS measurements at 310 ◦C in alkaline water, the val-ues of the zero-frequency limit of the Warburg impedance (seeTable 4) accounting for mass transfer through the inner adher-ent oxide layer is higher with the presence of 300 ppb dissolvedoxygen. This indicates to a lower oxide growth rate at 300 ppboxygen level as compared to that at deaerated condition. Thus,the trend in oxidation growth rate in terms of EIS parameters(in 310 ◦C alkaline water) up to at least 18 h of exposure doescorrelate with the short term weight gain in static autoclavesat 400 ◦C steam. For exposure times longer than 24 h at 400 ◦C,an increase in DO level does instead lead to an increase of theweight gain for the same exposure time in the accelerated tests inautoclave when compared to the measurements in a recirculatingloop.

4.2. Implications of difference in oxidation behaviour due to DOlevels

The advanced nuclear power reactors mandate partial boil-ing inside the core (Sinha and Kakodkar, 2002, 2006; Kain et al.,2005) and as discussed earlier, this is expected to result in anincrease in dissolved oxygen content (IAEA Tecdoc-996, 1998; Linet al., 1992; Cox, 1968; Sava, 1998; Amaev et al., 1970). The cur-rent study shows a marked increase in oxidation (weight gain) ofZr–1 wt%Nb and Zr–2.5 wt%Nb alloys when the DO levels are high;whereas oxidation of Zircaloy-2 has shown insensitivity to the DOlevels.

It has been indicated by the results from this study (Figs. 4–6and Tables 3 and 4) that the weight gain for Zr–1 wt%Nb decreasewith DO content in the initial phase of oxidation (310 ◦C alkalinewater test from CDE-EIS tests upto 90 h and 400 ◦C steam test forthe initial 18 h of oxidation).

These results imply that selection of materials for componentsthat would remain in the nuclear reactor core for a long time (e.g.pressure tubes and other structural components) has to take intoaccount the influence of high DO levels (due to partial boiling) (IAEATecdoc-996, 1998; Lin et al., 1992; Cox, 1968; Sava, 1998; Amaev etal., 1970) on weight gain. The current results reported here indicatethat thick oxides form adjacent to Nb rich second phase Zr–1 wt%Nband Zr–2.5 wt%Nb alloys respectively when the DO levels are high.The exact enhancement of weight gain needs to be measured incontrolled test conditions in a recirculation test loop in which afixed level of DO can be maintained throughout the test (Bojinovet al., 2005a,b, 2008) as compared to the accelerated tests reportedin this study using static autoclaves. The DO levels in static auto-claves go on reducing with time as the available air reacts with hot

indicative of a broad trend of increased weight gain for Zr–1 wt%Nband Zr–2.5 wt%Nb alloys in high DO steam. Under low DO levels,Zr–1 wt%Nb alloy showed lowest weight gain after 90 days of expo-sure at 400 ◦C, 10 MPa, whereas the weight gain for Zr–2.5 wt%Nband Zircaloy-2 were comparable.

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94 M. Kiran Kumar et al. / Nuclear Eng

. Conclusions

The oxidation tests were carried out at high levels of dissolvedxygen to simulate the effect of partial boiling conditions, at 400 ◦C,0 MPa for durations up to 90 days. In addition controlled distancelectrochemistry (CDE) experiments at 310 ◦C primary (lithiated)ater in a recirculation loop with totally deaerated and at 300pb DO were done on Zr–1 wt%Nb alloy. The results of this studyighlight that

The oxidation behaviour of Zircaloy-2 is quite immune to thechange in dissolved oxygen level; while for both the Zr–Nb alloys,the weight gain increases significantly with increase in the levelof DO in long term accelerated tests.Though the studies on demonstrating the mechanism for suchdifferences are still in progress, the morphology of the thickeroxides indicate that preferential oxidation of Nb depleted matrixaround the Nb-rich second phases present in the Zr–Nb alloysmight be responsible for such a difference.Hydrogen content in alloys oxidized at high DO conditions wasinvariably lower than that obtained at low DO conditions, irre-spective of the effect of DO level change on weight gain for all thethree alloys. The preferential reaction of the metal directly withfree oxygen rather than water to produce oxide explains such anobservation.The effect of dissolved oxygen on the short term acceleratedweight gain in static autoclaves was correlated and matched wellwith the kinetic parameter (Warburg impedance) derived fromin-situ electrochemical impedance spectroscopy (EIS) throughCDE done in a recirculation loop at 310 ◦C with controlled levelsof DO and for duration upto 90 h.The high DO levels resulted in lower weight gain in the shortduration tests (upto 24 h in 400 ◦C steam as per G2 M ASTM) forZr–Nb alloys but the weight gain doubled in long duration tests(upto 60 days). This trend in oxidation with DO levels was notapparent in Zircaloy-2 in long term tests.

cknowledgements

The authors are grateful to Mr Thomas Abraham and Mr Vivekhaudhary of Materials Science Division, Bhabha Atomic Researchentre for their help in operation of the autoclaves used for thexperiments reported in this study. They also acknowledge theelp with hydrogen analysis from Mr V.D. Alur of Post Irradia-ion Examination Division, Bhabha Atomic Research Centre. Onef the authors wishes to gratefully acknowledge the funding fromhe Finnish Funding Agency for Technology and Innovation (TEKES)hrough the RITMO-project and The Finnish Research Programmen Nuclear Power Plant Safety 2007–2010 (SAFIR 2010) throughhe WATCHEM-project.

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