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Hydrogen uptake by oxidized zirconium alloys
Journal of
ALLOY5 AND COMPOUND5 ELSEVIER
Journal of Alloys and Compounds 231 (1995) 716-721
Hydrogen uptake by oxidized zirconium alloys M.B. Elmoselhi Ontario Hydro Technologies, Toronto, Ont. MSZ 5S4, Canada
Abstract Zirconium alloys are used in CANDU (Canadian deuterium uranium) reactor pressure tubes and nuclear reactor fuel claddings. Although the alloy is usually prefilmed with a protective thin oxide, undesirable hydrogen builds up in the bulk alloy over years of operation. Deuterium transported through zirconium oxide into the bulk metal has been determined by exposing thin ( - 1 mm thickness) prefilmed samples to a controlled gaseous environment at the relatively high temperature of 380 °C for 15 days. The exposures were followed by bulk alloy hydrogen analysis using differential scanning calorimetry, which is insensitive to hydrogen in the oxide. Amounts of deuterium transported through oxides grown on different substrates of zirconium and its alloys have been determined for comparison. They range from 1 to 4 i~gcm -2. Measurable amounts of deuterium were also transported through the oxide at the lower temperature range of 200 to 330 °C (which encompasses the pressure tube operating range of -250 to 315 °C) by exposure to a relatively high pressure of deuterium gas (-230 to 780 kPa) for 10 days. Measured fluxes of deuterium after such exposures ranged from 1 to 14 ~g cm 2. Through-oxide-thickness deuterium concentration profiles of the samples were obtained using secondary ion mass spectrometry. The profiles show features that may be correlated to changes in the nature of the oxide and its function as a barrier against deuterium uptake. Keywords: Zirconium alloys; Nuclear materials; Hydrogen diffusion; Secondary ion mass spectrometry
L Introduction Pressure tubes in C A N D U (Canadian deuterium uranium) nuclear reactors and fuel claddings in most nuclear reactors are made out of zirconium alloys. In pressure tubes, the alloy is stress relieved by steam autoclaving at 400 °C for 24 h. The autoclaving produces a thin oxide film ( - 0 . 5 to 1 txm) which protects the alloy against high hydrogen uptake and consequently delayed hydride cracking. However, over years of operation the hydrogen level in the bulk alloy increases, There are two possible routes for hydrogen uptake by the body of operating C A N D U pressure tubes (at temperatures from - 2 5 0 to - 3 1 0 °C): the inside and the outside surfaces of the tube. The inside tube surfaces are exposed to the primary heat transport system coolant, lithiated heavy water ( L i O H is used to maintain a p H of -10.5), at - 1 0 MPa. At this surface, hydrogen is mainly absorbed as part of the corrosion reaction. However, dissolved hydrogen gas in the water (usually added to suppress oxygen produced by radiolysis) may also permeate through the oxide to the 0925-8388/95/$09.50 © 1995 Elsevier Science S.A. All rights reserved SSDI 0925-8388(95)01759-3
bulk alloy. In contrast, the outside tube surfaces are exposed to dry CO 2 (dew point ranges from - - 1 0 to --40°C)at approximately atmospheric pressure with up to 130 Pa of deuterium, and at this surface the uptake may be mainly controlled by the permeation of deuterium from the gas phase through the oxide film. The latter is the focus of this study. The objectives of this work are to compare zirconium oxides grown under different conditions and on different zirconium alloys as substrates in terms of their relative permeability to hydrogen isotopes, and to improve understanding of the mechanism of hydrogen transport. In the following sections direct procedures are explained which were developed to determine amounts of transported deuterium through zirconium oxide into the bulk of the substrate.
2. Experimental procedure The experiments are simply based on exposing samples, with their surface totally covered with oxide,
M.B. Elmoselhi / Journal of Alloys and Compounds 231 (1995) 716-721
to hydrogen followed by measuring the increase in their bulk alloy hydrogen content due to the exposure, The main experimental difficulty is that with exposures simulating operating conditions the uptake is too slow to allow measurable hydrogen flux through the oxide within reasonable time in the laboratory. Therefore it is necessary to accelerate the uptake process in the experiments, without significantly changing the nature of the oxide under consideration. Such acceleration was achieved in the laboratory by two methods: (1) conducting the exposures at the relatively high temperature of 380 °C, at which temperature the microstructure changes of the substrate alloy are negligibly slow; (2) exposing the samples to relatively high pressures of deuterium (230 to 780 kPa) at temperatures between 200 and 330 °C, which includes the range of reactor operating temperatures, The average level of hydrogen in the bulk zirconium alloy was determined using differential scanning calorimetry (DSC). DSC determines the total hydrogen isotope content of the sample and is an appropriate technique for measuring accumulated hydrogen uptake in the metal during permeation experiments, since it is insensitive to hydrogen in the oxide. Thin samples of Zr, Zr-2.5wt.% Nb and Zircaloy-2 (zirconium samples were 8-10 mm diameter and i mm thick, while zirconium alloys were 20 mm x 20 mm x 0.5 to 1.0 mm thick) with machine finish (~<1 ixm) were used. The circular samples were used in pairs of consecutive slices, while the rectangular samples were cut into two parts (10 × 20 mm each). The pairs were autoclaved in s t e a m ( U 2 0 ) or oxidized together in 02 at 400 °C for 24 h. One part was used as a reference while the other part was exposed to the specified environment. The chosen dimensions of the sample (with high surface to volume ratio) helped reduce the edge effects and increase the sensitivity of detecting the uptake by the bulk alloy (in terms of ppm). From the measured increase in bulk alloy hydrogen content, the amount of hydrogen transported through the oxide was inferred. The selec,ted experimental parameters resulted in a measurable increase of hydrogen isotope
717
content in the bulk alloy of the samples ranging from 3 to 24 ppm following exposures to deuterium gas (D2) in the laboratory for 10 to 15 days. Through-oxide-thickness depth profiling was performed on oxides before and after gaseous exposures using secondary ion mass spectrometry (SIMS). The intensities of the ions 1H-, 2 D - and l ° 6 Z r O are typically measured through the oxide and into the base alloy (using a 10keV Cs + primary ion beam). Since ~refilming was performed with light water steam, the D - profile following exposure provides an indication of the penetration of this isotope into the oxide. The 1H- content is mostly associated with hydrogen present in the oxide formed by steam prefilming, which is the pressure tube as-received condition. By using ionimplanted standards of deuterium and an absolute technique such as nuclear reaction analysis (NRA) for measuring concentration, quantitative SIMS depth profiles for deuterium have been obtained. The l°6ZrO profile provides information on the location and width of the metal-oxide interface.
3. Results 3.1. Permeation at 380 °C
Exposures at the high temperature of 380 °C were conducted for the prefilmed samples listed in Table 1. The samples were exposed together to a steady pressure of - 1 0 -3 Pa of pure D 2 monitored by a mass spectrometer for 15 days. The post exposure increase of the bulk alloy hydrogen content was measured for each sample using DSC. Amounts of transported deuterium (in ~g cm -2, of the macroscopically measured oxide surface area) were then determined. The obtained values are shown in Table 1. Each value is the average of a set of at least three macroscopically identical samples processed together, while the indicated error range is the standard deviation of the set. It appears that oxides grown on zirconium are less permeable to hydrogen than those grown on zirconium
Table 1 Samples prefilmed and exposed to ~10 3 Pa of pure deuterium at 380 °C for 15 days Sample substrate
F'refilming environment
Hydrogen transported through the oxide (l~g cm 2)
Zr-2.5Nb Zr-2.5Nb Zr-2 Zr-2 Zr (polycrystal) Zr (polycrystal) Zr (single crystal)
Oxygen gas Steam autoclaving Oxygen gas Steam autoclaving Oxygen gas S team autoclaving Oxygen gas
3.2 _+ 1.4 1.4 _+ 1.1 3.8 -+ 1.7 3.7 _+2.0
M.B. Elmoselhi / Journal of Alloys and Compounds 231 (1995) 716-721
718
3.2. High pressure permeation
alloy substrates. Typical D through-oxide-thickness depth profiles are shown in Fig. 1. Fig. l(a) represents the case of high D uptake by the bulk alloy while Fig. l(b) shows the opposite case with undetectable increase of the bulk metal D content due to the exposure. Within statistical scatter the position of the oxide-metal interface (in other words, the thickness of the oxide) has been practically unaffected by the exposure conditions. Post-exposure D profiles appear to show some features relevant to the transport of D to the bulk alloy, especially near the oxide surface, near the oxide-metal interface and into the bulk metal. Such features will be discussed in Section 4.
10
As mentioned before, exposure to higher pressures of deuterium (230-780 kPa) was used as a method to accelerate hydrogen uptake in the laboratory at a temperature range which encompasses pressure tube operating temperatures. The experimental matrix was chosen to examine the relationship between pressure and uptake at a constant temperature and vise versa within the considered ranges. The samples were all macroscopically identical steam autoclaved Zr-2.5Nb coupons and all exposures were run for 10 days. The determined amount of transported deuterium due to
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Fig. 1. Typical through-oxide-thickness D concentration profiles for prefilmed samples before and after exposure (BE and A E ) to - 0 . 0 0 1 P a of deuterium at 380 ° C f o r 15 d a y s : (a) Z r - 2 a n d ( b ) Z r single crystal.
M.B. Elmoselhi / Journal of Alloys and Compounds 231 (1995) 716-721
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Fig. 2. A plot of deuterium transported through the oxide vs. the square root of exposure pressure at 330 °C. The solid line represents a least square fit.
exposures to the pressures 23, 440 and 780 kPa while maintaining the temperature at 330°C are plotted in Fig. 2. A typical through-oxide-thickness D depth profile of a high pressure exposure sample is shown in
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Fig. 3 in comparison with a profile with a high temperature exposure which resulted in approximately a similar amount of uptake by the bulk metal. Also, transported amounts of deuterium through the oxide resulting from exposures to a pressure of 440 kPa at the temperatures 200, 250 and 330 °C are shown in Fig. 4. Each plotted value represents the average of a set of at least three macroscopically identical samples prepared and exposed to the same conditions. The standard deviation from each set was within _+30%.
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S,mp,,O,p~~) Fig. 3. Deuterium through-oxide-thickness profles for a prefilmed Zr-2.5Nb sample after exposure to 440 kPa of D 2 at 330 °C for 10 days (curve 1). The dotted line (curve 2) is from Fig. l(a) (10 -3 Pa,
380°C, 15 days)for comparison,
Oxides grown on pure zirconium are shown to be less permeable than those grown on zirconium alloys (see Table 1). The results are qualitatively in agreement with the reported diffusion coefficients from implanted hydrogen ions, which are approximately three orders of magnitude higher in an oxide grown on Zr-2.5Nb than for those grown on polycrystalline zirconium [1]. The observed difference between oxides grown on zirconium vs. on its alloys suggests a critical role of the bulk microstructure and its influence on the nature of the oxide in terms of their permeability to hydrogen isotopes. For example, in the case of Z r the microstructure c o n s i s t s o f a-zirconium grains surrounded by fl-zirconium grain boundaries;
2.5Nb,
720
M.B. Elmoselhi / Journal of Alloys and Compounds 231 (1995) 716-721
the two-phases are known to be reflected in the oxide microstructure [2,3] and may lead to extra interfaces and consequently extra diffusional pathways for hydrogen transport. Also, the grain size in polycrystalline zirconium is known to be generally larger than in its alloys which results in fewer grain boundaries leading to lower permeability. The difference among oxides grown on different zirconium alloys and with different oxidation chemistry is unclear within the indicated statistical scatter, Exposure to a non-oxidizing environment (D2) is known to result in oxide dissolution into the metal at the interface depending mainly on the temperature and the duration of exposure [4]. At 380 °C the extent of dissolution was estimated (using the procedure discussed in Ref. [4]) as a loss of thickness in the order of -0.1 Ixm, while at 330 °C the estimated value was about 0.005 p~m. Effects on this scale would be undetectable in the depth profiles shown in Fig. 1. The indicated locations of the oxide-metal interfaces (i.e. the oxide thicknesses) seem to imply a non-uniform thickness with a scatter in the same order of magnitude as the estimated loss of thickness. However, the D profile corresponding to the oxide exposed to the higher temperature (Fig. l(a)) shows high D-concentration at the interface. This region, starting from the lowest D-concentration within the oxide to the oxidemetal interface (-50% of the oxide thickness) may represent a partially dissolved oxide, ('degraded') since the D-concentration distribution appears to be continuous into the bulk metal [5]. The profile in the metal near the interface represents hydrides formed upon exceeding the terminal solubility when the sampie was brought to room temperature before profiling, Neither the profile with low measured D content in the bulk metal (Fig. l(b)), nor that of the high pressure exposure at the lower temperature of 330 °C (Fig. 3, solid curve), show a 'degraded' region near the interface. In Fig. l(b) the lack of a degradation region at such a high temperature may be attributed to the nature of the substrate and may be linked to the observed low D uptake by the bulk metal. In Fig. 3, although the observed bulk uptake is high following the exposure to a high pressure, degradation is expected to be negligible at 330 °C. In view of Fig. 2, the transport of deuterium through the oxide at significantly higher pressures appears to follow a certain trend. Above a certain pressure threshold (in this case -200 kPa), the rate of uptake by the bulk alloy is roughly proportional to the square root of the gas pressure (P~). This is similar to the reported effect of pressure on the rate of hydrogen uptake by zirconium samples, which reflects the dissociation of hydrogen molecules prior to absorption by the metal [6]. A typical through-oxide-thickness D profile for a
high pressure exposure sample is depicted in Fig. 3 in comparison with a typical profile for a low pressure/ high temperature exposure sample. Two main features are worth noting: (1) no signs of oxide degradation near the interface for the high pressure/low temperature profile, as expected since degradation is a function of temperature; (2) the concentration of deuterium at the interface for the high pressure profile is significantly higher in support of the implied hypothesis of the oxide being filled with deuterium up to the interface due to exposure to high pressure. The initial pressure, or the threshold for detectable uptake by the bulk metal, also reflects the presence of the oxide and its resistivity to the gas pressure. This initial deviation from the linear relationship with P~ has been repeatedly reported in the literature and has often been explained in terms of the adsorption of hydrogen on the surface prior to its absorption by the bulk metal [6,7]. However, although the samples in such experiments were supposed to have bare metal surfaces, it is most likely, considering the available experimental techniques at the time of the experiments, that the samples had an air-formed oxide which would contribute to the observed initial pressure threshold. In Fig. 4 the amounts of deuterium transported through the oxide due to the high pressure exposures at different temperatures appear to approximately follow an Arrhenius fit. This implies that the uptake process is thermally activated with an inferred activation energy of - 2 6 kJ gmo1-1. Within the warranted statistical scatter, this is roughly comparable with the reported value of 34 kJ gmo1-1 for the temperature range of 225 to 421°C [8].
5. Summary and coneh~ions A procedure to test the quality of zirconium oxide in terms of its permeability to hydrogen isotopes has been established. The procedure uses the relatively high temperature of 380 °C to accelerate the uptake and ensure a measurable amount of transported hydrogen through the oxide within days in the laboratory. The increase in the bulk alloy hydrogen content due to gas exposure was measured using DSC. The test was used to compare zirconium oxides grown on different substrates and by different chemistries. It appears that, in general, oxides grown on alloyed zirconium are more permeable to deuterium (at least at 380 °C) than those grown on pure zirconium. This result reflects the effect of bulk alloy microstructure, such as grain boundaries and grain size on their oxide
M.B. Elmoselhi / Journal of Alloys and Compounds 231 (1995) 716-721
microstructure. The test does not seem to distinguish between single crystal ws. polycrystalline zirconium as substrates for oxides in terms of their permeability to deuterium. Also, the difference in permeability between oxides grown by steam autoclaving vs. exposure to oxygen gas on the same substrate was undetectable, Through-oxide-thickness D-depth profiles produced by SIMS show some features relevant to the mobility of D through the oxide such as the D concentration at the oxide surface, the driving gradient and the dissolution of the oxide near the interface with the metal. Owing to a concern that the D permeability of the oxide measured at high temperature (380 °C) may only be indirectly related to the permeability at the reactor operating temperatures (--250-315 °C), a different procedure was tested which u s e s a relatively high pressure of D 2 (230 to 780 kPa) at lower temperatures (250 to 330 °C) which spans the reactor operating range. At pressures higher than -200 kPa, deuterium penetrates the oxide and interacts with the metal at the interface. The rate of uptake is then proportional 1 to PL The temperature dependency at a constant pressure of 440 kPa produces an mrrhenius fit with a thermal activation energy of - 2 6 kJ gmo1-1, roughly comparable with a reported value of 34 kJ mo1-1 for the temperature range of 225 to 421 °C.
Acknowledgements Funding for this work was provided by the CANDU Owners Group (COG). Staff technologist L. Grant of Ontario Hydro Technologies (OHT) carried out much
721
of the sample preparation, exposure and DSC. SIMS analysis was performed by G. Mount at the University of Western Ontario. Valuable discussions with OHT staff, Dr. M. Leger, Dr. B.D. Warr and the corrosion group are greatly appreciated. Discussions with staff at the Atomic Energy of Canada Limited, Chalk River Nuclear Laboratories are also acknowledged.
References [1] D. Khatamian and D. Manchester, An ion beam study of hydrogen diffusion in oxides of Zr and Zr-Nb, J. Nucl. Mater.,
166 (1989)300-306. [2] B.D. Warr, M.B. Elmoselhi, A.B. Brennenstuhl, P.C. Lichtenberger, N.S. Mclntyre and S.B. Newcomb, Oxide characteristics and their relationship to hydrogen uptake in zirconium alloys, 9th Int. Symp. on Zirconium in the Nuclear Industry, Kobe, Japan, November 5-8, 1990, American Society for Testing and
Materials, ASTM, STP 1132. [3] Y.P. Lin, On the Microstructure of Steam Formed Oxides on a Zr-2.5 Nb Alloy, OHRD/COG-93-262, July 1993. [4] J.P. Pemsler, Diffusion of oxygen in zirconium and its relation to oxidation and corrosion, J. Electrochem. Soc., 105(6)(1958).
[5] M.B. Elmoselhi,B.D.Wart and N.S. Mclntyre, A study of the hydrogen uptake mechanism in zirconium alloys, lOth Int. Syrup. on Zirconium in the Nuclear Industry, Baltimore, MD, June 21-24, 1993, American Society for Testing and Materials,
ASTM, STP 1245.
[6] E.A. Gulbransen and K.F. Andrew, Kinetics of the reactions of zirconium with 02, N 2 and H2, Metals Trans., 185 (1949) 515. [7] C.J. Smithells and C.E. Ransley, The diffusion of gases through metals, Proc. Roy. Soc., 150A (1935) 172.
[8] T. Smith, Kinetics and mechanismof hydrogen permeation of oxide films on zirconium, J. Nuc. Mater., 18 (1966) 323-336.
ALLOY5 AND COMPOUND5 ELSEVIER
Journal of Alloys and Compounds 231 (1995) 716-721
Hydrogen uptake by oxidized zirconium alloys M.B. Elmoselhi Ontario Hydro Technologies, Toronto, Ont. MSZ 5S4, Canada
Abstract Zirconium alloys are used in CANDU (Canadian deuterium uranium) reactor pressure tubes and nuclear reactor fuel claddings. Although the alloy is usually prefilmed with a protective thin oxide, undesirable hydrogen builds up in the bulk alloy over years of operation. Deuterium transported through zirconium oxide into the bulk metal has been determined by exposing thin ( - 1 mm thickness) prefilmed samples to a controlled gaseous environment at the relatively high temperature of 380 °C for 15 days. The exposures were followed by bulk alloy hydrogen analysis using differential scanning calorimetry, which is insensitive to hydrogen in the oxide. Amounts of deuterium transported through oxides grown on different substrates of zirconium and its alloys have been determined for comparison. They range from 1 to 4 i~gcm -2. Measurable amounts of deuterium were also transported through the oxide at the lower temperature range of 200 to 330 °C (which encompasses the pressure tube operating range of -250 to 315 °C) by exposure to a relatively high pressure of deuterium gas (-230 to 780 kPa) for 10 days. Measured fluxes of deuterium after such exposures ranged from 1 to 14 ~g cm 2. Through-oxide-thickness deuterium concentration profiles of the samples were obtained using secondary ion mass spectrometry. The profiles show features that may be correlated to changes in the nature of the oxide and its function as a barrier against deuterium uptake. Keywords: Zirconium alloys; Nuclear materials; Hydrogen diffusion; Secondary ion mass spectrometry
L Introduction Pressure tubes in C A N D U (Canadian deuterium uranium) nuclear reactors and fuel claddings in most nuclear reactors are made out of zirconium alloys. In pressure tubes, the alloy is stress relieved by steam autoclaving at 400 °C for 24 h. The autoclaving produces a thin oxide film ( - 0 . 5 to 1 txm) which protects the alloy against high hydrogen uptake and consequently delayed hydride cracking. However, over years of operation the hydrogen level in the bulk alloy increases, There are two possible routes for hydrogen uptake by the body of operating C A N D U pressure tubes (at temperatures from - 2 5 0 to - 3 1 0 °C): the inside and the outside surfaces of the tube. The inside tube surfaces are exposed to the primary heat transport system coolant, lithiated heavy water ( L i O H is used to maintain a p H of -10.5), at - 1 0 MPa. At this surface, hydrogen is mainly absorbed as part of the corrosion reaction. However, dissolved hydrogen gas in the water (usually added to suppress oxygen produced by radiolysis) may also permeate through the oxide to the 0925-8388/95/$09.50 © 1995 Elsevier Science S.A. All rights reserved SSDI 0925-8388(95)01759-3
bulk alloy. In contrast, the outside tube surfaces are exposed to dry CO 2 (dew point ranges from - - 1 0 to --40°C)at approximately atmospheric pressure with up to 130 Pa of deuterium, and at this surface the uptake may be mainly controlled by the permeation of deuterium from the gas phase through the oxide film. The latter is the focus of this study. The objectives of this work are to compare zirconium oxides grown under different conditions and on different zirconium alloys as substrates in terms of their relative permeability to hydrogen isotopes, and to improve understanding of the mechanism of hydrogen transport. In the following sections direct procedures are explained which were developed to determine amounts of transported deuterium through zirconium oxide into the bulk of the substrate.
2. Experimental procedure The experiments are simply based on exposing samples, with their surface totally covered with oxide,
M.B. Elmoselhi / Journal of Alloys and Compounds 231 (1995) 716-721
to hydrogen followed by measuring the increase in their bulk alloy hydrogen content due to the exposure, The main experimental difficulty is that with exposures simulating operating conditions the uptake is too slow to allow measurable hydrogen flux through the oxide within reasonable time in the laboratory. Therefore it is necessary to accelerate the uptake process in the experiments, without significantly changing the nature of the oxide under consideration. Such acceleration was achieved in the laboratory by two methods: (1) conducting the exposures at the relatively high temperature of 380 °C, at which temperature the microstructure changes of the substrate alloy are negligibly slow; (2) exposing the samples to relatively high pressures of deuterium (230 to 780 kPa) at temperatures between 200 and 330 °C, which includes the range of reactor operating temperatures, The average level of hydrogen in the bulk zirconium alloy was determined using differential scanning calorimetry (DSC). DSC determines the total hydrogen isotope content of the sample and is an appropriate technique for measuring accumulated hydrogen uptake in the metal during permeation experiments, since it is insensitive to hydrogen in the oxide. Thin samples of Zr, Zr-2.5wt.% Nb and Zircaloy-2 (zirconium samples were 8-10 mm diameter and i mm thick, while zirconium alloys were 20 mm x 20 mm x 0.5 to 1.0 mm thick) with machine finish (~<1 ixm) were used. The circular samples were used in pairs of consecutive slices, while the rectangular samples were cut into two parts (10 × 20 mm each). The pairs were autoclaved in s t e a m ( U 2 0 ) or oxidized together in 02 at 400 °C for 24 h. One part was used as a reference while the other part was exposed to the specified environment. The chosen dimensions of the sample (with high surface to volume ratio) helped reduce the edge effects and increase the sensitivity of detecting the uptake by the bulk alloy (in terms of ppm). From the measured increase in bulk alloy hydrogen content, the amount of hydrogen transported through the oxide was inferred. The selec,ted experimental parameters resulted in a measurable increase of hydrogen isotope
717
content in the bulk alloy of the samples ranging from 3 to 24 ppm following exposures to deuterium gas (D2) in the laboratory for 10 to 15 days. Through-oxide-thickness depth profiling was performed on oxides before and after gaseous exposures using secondary ion mass spectrometry (SIMS). The intensities of the ions 1H-, 2 D - and l ° 6 Z r O are typically measured through the oxide and into the base alloy (using a 10keV Cs + primary ion beam). Since ~refilming was performed with light water steam, the D - profile following exposure provides an indication of the penetration of this isotope into the oxide. The 1H- content is mostly associated with hydrogen present in the oxide formed by steam prefilming, which is the pressure tube as-received condition. By using ionimplanted standards of deuterium and an absolute technique such as nuclear reaction analysis (NRA) for measuring concentration, quantitative SIMS depth profiles for deuterium have been obtained. The l°6ZrO profile provides information on the location and width of the metal-oxide interface.
3. Results 3.1. Permeation at 380 °C
Exposures at the high temperature of 380 °C were conducted for the prefilmed samples listed in Table 1. The samples were exposed together to a steady pressure of - 1 0 -3 Pa of pure D 2 monitored by a mass spectrometer for 15 days. The post exposure increase of the bulk alloy hydrogen content was measured for each sample using DSC. Amounts of transported deuterium (in ~g cm -2, of the macroscopically measured oxide surface area) were then determined. The obtained values are shown in Table 1. Each value is the average of a set of at least three macroscopically identical samples processed together, while the indicated error range is the standard deviation of the set. It appears that oxides grown on zirconium are less permeable to hydrogen than those grown on zirconium
Table 1 Samples prefilmed and exposed to ~10 3 Pa of pure deuterium at 380 °C for 15 days Sample substrate
F'refilming environment
Hydrogen transported through the oxide (l~g cm 2)
Zr-2.5Nb Zr-2.5Nb Zr-2 Zr-2 Zr (polycrystal) Zr (polycrystal) Zr (single crystal)
Oxygen gas Steam autoclaving Oxygen gas Steam autoclaving Oxygen gas S team autoclaving Oxygen gas
3.2 _+ 1.4 1.4 _+ 1.1 3.8 -+ 1.7 3.7 _+2.0
M.B. Elmoselhi / Journal of Alloys and Compounds 231 (1995) 716-721
718
3.2. High pressure permeation
alloy substrates. Typical D through-oxide-thickness depth profiles are shown in Fig. 1. Fig. l(a) represents the case of high D uptake by the bulk alloy while Fig. l(b) shows the opposite case with undetectable increase of the bulk metal D content due to the exposure. Within statistical scatter the position of the oxide-metal interface (in other words, the thickness of the oxide) has been practically unaffected by the exposure conditions. Post-exposure D profiles appear to show some features relevant to the transport of D to the bulk alloy, especially near the oxide surface, near the oxide-metal interface and into the bulk metal. Such features will be discussed in Section 4.
10
As mentioned before, exposure to higher pressures of deuterium (230-780 kPa) was used as a method to accelerate hydrogen uptake in the laboratory at a temperature range which encompasses pressure tube operating temperatures. The experimental matrix was chosen to examine the relationship between pressure and uptake at a constant temperature and vise versa within the considered ranges. The samples were all macroscopically identical steam autoclaved Zr-2.5Nb coupons and all exposures were run for 10 days. The determined amount of transported deuterium due to
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"degraded" region
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Fig. 1. Typical through-oxide-thickness D concentration profiles for prefilmed samples before and after exposure (BE and A E ) to - 0 . 0 0 1 P a of deuterium at 380 ° C f o r 15 d a y s : (a) Z r - 2 a n d ( b ) Z r single crystal.
M.B. Elmoselhi / Journal of Alloys and Compounds 231 (1995) 716-721
20 -
330~C
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I
200"C
I
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D I~ansported in p g / c m 2 = - 15.072 + 0.99775 P ^ l / 2 ( k P a ^ l / 2 ) 18" (D transported in pg/craA2) = 534.14 "Exp {- 26 (kJ/mole)/RT} 16'
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Square Root of Pressure (P^1/2- kPaAll2)
Fig. 2. A plot of deuterium transported through the oxide vs. the square root of exposure pressure at 330 °C. The solid line represents a least square fit.
exposures to the pressures 23, 440 and 780 kPa while maintaining the temperature at 330°C are plotted in Fig. 2. A typical through-oxide-thickness D depth profile of a high pressure exposure sample is shown in
x°°
Start of the
->'in,e,~efor ~
~
(1)
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/ 10 -2
Fig. 3 in comparison with a profile with a high temperature exposure which resulted in approximately a similar amount of uptake by the bulk metal. Also, transported amounts of deuterium through the oxide resulting from exposures to a pressure of 440 kPa at the temperatures 200, 250 and 330 °C are shown in Fig. 4. Each plotted value represents the average of a set of at least three macroscopically identical samples prepared and exposed to the same conditions. The standard deviation from each set was within _+30%.
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4. D i s c u s s i o n
'., "~ 1°3
~ k,,
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VT Fig. 4. A plot of deuterium transported through the oxide in 10 days under exposure to the same pressure of 440kPa at the different indicated temperatures (K).
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.
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.
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S,mp,,O,p~~) Fig. 3. Deuterium through-oxide-thickness profles for a prefilmed Zr-2.5Nb sample after exposure to 440 kPa of D 2 at 330 °C for 10 days (curve 1). The dotted line (curve 2) is from Fig. l(a) (10 -3 Pa,
380°C, 15 days)for comparison,
Oxides grown on pure zirconium are shown to be less permeable than those grown on zirconium alloys (see Table 1). The results are qualitatively in agreement with the reported diffusion coefficients from implanted hydrogen ions, which are approximately three orders of magnitude higher in an oxide grown on Zr-2.5Nb than for those grown on polycrystalline zirconium [1]. The observed difference between oxides grown on zirconium vs. on its alloys suggests a critical role of the bulk microstructure and its influence on the nature of the oxide in terms of their permeability to hydrogen isotopes. For example, in the case of Z r the microstructure c o n s i s t s o f a-zirconium grains surrounded by fl-zirconium grain boundaries;
2.5Nb,
720
M.B. Elmoselhi / Journal of Alloys and Compounds 231 (1995) 716-721
the two-phases are known to be reflected in the oxide microstructure [2,3] and may lead to extra interfaces and consequently extra diffusional pathways for hydrogen transport. Also, the grain size in polycrystalline zirconium is known to be generally larger than in its alloys which results in fewer grain boundaries leading to lower permeability. The difference among oxides grown on different zirconium alloys and with different oxidation chemistry is unclear within the indicated statistical scatter, Exposure to a non-oxidizing environment (D2) is known to result in oxide dissolution into the metal at the interface depending mainly on the temperature and the duration of exposure [4]. At 380 °C the extent of dissolution was estimated (using the procedure discussed in Ref. [4]) as a loss of thickness in the order of -0.1 Ixm, while at 330 °C the estimated value was about 0.005 p~m. Effects on this scale would be undetectable in the depth profiles shown in Fig. 1. The indicated locations of the oxide-metal interfaces (i.e. the oxide thicknesses) seem to imply a non-uniform thickness with a scatter in the same order of magnitude as the estimated loss of thickness. However, the D profile corresponding to the oxide exposed to the higher temperature (Fig. l(a)) shows high D-concentration at the interface. This region, starting from the lowest D-concentration within the oxide to the oxidemetal interface (-50% of the oxide thickness) may represent a partially dissolved oxide, ('degraded') since the D-concentration distribution appears to be continuous into the bulk metal [5]. The profile in the metal near the interface represents hydrides formed upon exceeding the terminal solubility when the sampie was brought to room temperature before profiling, Neither the profile with low measured D content in the bulk metal (Fig. l(b)), nor that of the high pressure exposure at the lower temperature of 330 °C (Fig. 3, solid curve), show a 'degraded' region near the interface. In Fig. l(b) the lack of a degradation region at such a high temperature may be attributed to the nature of the substrate and may be linked to the observed low D uptake by the bulk metal. In Fig. 3, although the observed bulk uptake is high following the exposure to a high pressure, degradation is expected to be negligible at 330 °C. In view of Fig. 2, the transport of deuterium through the oxide at significantly higher pressures appears to follow a certain trend. Above a certain pressure threshold (in this case -200 kPa), the rate of uptake by the bulk alloy is roughly proportional to the square root of the gas pressure (P~). This is similar to the reported effect of pressure on the rate of hydrogen uptake by zirconium samples, which reflects the dissociation of hydrogen molecules prior to absorption by the metal [6]. A typical through-oxide-thickness D profile for a
high pressure exposure sample is depicted in Fig. 3 in comparison with a typical profile for a low pressure/ high temperature exposure sample. Two main features are worth noting: (1) no signs of oxide degradation near the interface for the high pressure/low temperature profile, as expected since degradation is a function of temperature; (2) the concentration of deuterium at the interface for the high pressure profile is significantly higher in support of the implied hypothesis of the oxide being filled with deuterium up to the interface due to exposure to high pressure. The initial pressure, or the threshold for detectable uptake by the bulk metal, also reflects the presence of the oxide and its resistivity to the gas pressure. This initial deviation from the linear relationship with P~ has been repeatedly reported in the literature and has often been explained in terms of the adsorption of hydrogen on the surface prior to its absorption by the bulk metal [6,7]. However, although the samples in such experiments were supposed to have bare metal surfaces, it is most likely, considering the available experimental techniques at the time of the experiments, that the samples had an air-formed oxide which would contribute to the observed initial pressure threshold. In Fig. 4 the amounts of deuterium transported through the oxide due to the high pressure exposures at different temperatures appear to approximately follow an Arrhenius fit. This implies that the uptake process is thermally activated with an inferred activation energy of - 2 6 kJ gmo1-1. Within the warranted statistical scatter, this is roughly comparable with the reported value of 34 kJ gmo1-1 for the temperature range of 225 to 421°C [8].
5. Summary and coneh~ions A procedure to test the quality of zirconium oxide in terms of its permeability to hydrogen isotopes has been established. The procedure uses the relatively high temperature of 380 °C to accelerate the uptake and ensure a measurable amount of transported hydrogen through the oxide within days in the laboratory. The increase in the bulk alloy hydrogen content due to gas exposure was measured using DSC. The test was used to compare zirconium oxides grown on different substrates and by different chemistries. It appears that, in general, oxides grown on alloyed zirconium are more permeable to deuterium (at least at 380 °C) than those grown on pure zirconium. This result reflects the effect of bulk alloy microstructure, such as grain boundaries and grain size on their oxide
M.B. Elmoselhi / Journal of Alloys and Compounds 231 (1995) 716-721
microstructure. The test does not seem to distinguish between single crystal ws. polycrystalline zirconium as substrates for oxides in terms of their permeability to deuterium. Also, the difference in permeability between oxides grown by steam autoclaving vs. exposure to oxygen gas on the same substrate was undetectable, Through-oxide-thickness D-depth profiles produced by SIMS show some features relevant to the mobility of D through the oxide such as the D concentration at the oxide surface, the driving gradient and the dissolution of the oxide near the interface with the metal. Owing to a concern that the D permeability of the oxide measured at high temperature (380 °C) may only be indirectly related to the permeability at the reactor operating temperatures (--250-315 °C), a different procedure was tested which u s e s a relatively high pressure of D 2 (230 to 780 kPa) at lower temperatures (250 to 330 °C) which spans the reactor operating range. At pressures higher than -200 kPa, deuterium penetrates the oxide and interacts with the metal at the interface. The rate of uptake is then proportional 1 to PL The temperature dependency at a constant pressure of 440 kPa produces an mrrhenius fit with a thermal activation energy of - 2 6 kJ gmo1-1, roughly comparable with a reported value of 34 kJ mo1-1 for the temperature range of 225 to 421 °C.
Acknowledgements Funding for this work was provided by the CANDU Owners Group (COG). Staff technologist L. Grant of Ontario Hydro Technologies (OHT) carried out much
721
of the sample preparation, exposure and DSC. SIMS analysis was performed by G. Mount at the University of Western Ontario. Valuable discussions with OHT staff, Dr. M. Leger, Dr. B.D. Warr and the corrosion group are greatly appreciated. Discussions with staff at the Atomic Energy of Canada Limited, Chalk River Nuclear Laboratories are also acknowledged.
References [1] D. Khatamian and D. Manchester, An ion beam study of hydrogen diffusion in oxides of Zr and Zr-Nb, J. Nucl. Mater.,
166 (1989)300-306. [2] B.D. Warr, M.B. Elmoselhi, A.B. Brennenstuhl, P.C. Lichtenberger, N.S. Mclntyre and S.B. Newcomb, Oxide characteristics and their relationship to hydrogen uptake in zirconium alloys, 9th Int. Symp. on Zirconium in the Nuclear Industry, Kobe, Japan, November 5-8, 1990, American Society for Testing and
Materials, ASTM, STP 1132. [3] Y.P. Lin, On the Microstructure of Steam Formed Oxides on a Zr-2.5 Nb Alloy, OHRD/COG-93-262, July 1993. [4] J.P. Pemsler, Diffusion of oxygen in zirconium and its relation to oxidation and corrosion, J. Electrochem. Soc., 105(6)(1958).
[5] M.B. Elmoselhi,B.D.Wart and N.S. Mclntyre, A study of the hydrogen uptake mechanism in zirconium alloys, lOth Int. Syrup. on Zirconium in the Nuclear Industry, Baltimore, MD, June 21-24, 1993, American Society for Testing and Materials,
ASTM, STP 1245.
[6] E.A. Gulbransen and K.F. Andrew, Kinetics of the reactions of zirconium with 02, N 2 and H2, Metals Trans., 185 (1949) 515. [7] C.J. Smithells and C.E. Ransley, The diffusion of gases through metals, Proc. Roy. Soc., 150A (1935) 172.
[8] T. Smith, Kinetics and mechanismof hydrogen permeation of oxide films on zirconium, J. Nuc. Mater., 18 (1966) 323-336.