Prototypical experiments relating to air oxidation of Zircaloy-4 at high temperatures

The mechanism of the reaction between Zircaloy-4 and air at temperatures from 800 to 1500 °C was studied. Air attack under prototypical conditions with air ingress during a hypothetic severe nuclear reactor accident was investigated. Oxidation in air and in air and nitrogen-containing atmospheres leads to a major degradation of the cladding material. The main mechanism is the formation of zirconium nitride and its re-oxidation. Pre-oxidation in steam prevents air attack as long as the oxide scale is intact. Under steam/oxygen starvation conditions, the oxide scale is reduced and significant external nitride formation takes place. When modeling air ingress in severe accident computer codes, parabolic correlations for oxidation in air may be applied only for high temperatures (>1400 °C) and for pre-oxidized cladding (?1100 °C). Under all other conditions, faster, rather linear reaction kinetics should be applied.     

Oxidation kinetics of Zircaloy-4 and Zr-1Nb-1Sn-0.1Fe at temperatures of 700-1200 °C

The oxidation characteristics for the Zircaloy-4 and Zr-1.0Nb-1.0Sn-0.1Fe alloys were investigated in the temperature ranges of 700-1200 °C for 3600 s under steam supply conditions, using a modified thermo-gravimetric analyzer. The oxidation at these temperatures generally complied with the parabolic rate law for the examined duration up to 3600 s. However, the parabolic rate was not obeyed in the temperature ranges of 800-1050 °C. The oxidation kinetics were changed depending on the oxidation temperatures due to the phase transformations of the base metal and its oxide. The oxidation rate exponents of the Zr-1.0Nb-1.0Sn-0.1Fe alloy at all the temperatures were higher than those of Zircaloy-4. Considering the data controlled by the parabolic rates at 700, 1100, 1150, and 1200 °C, the oxidation rate constants were the same slopes as the Baker-Just relationship. The rate transition at 800 °C could have resulted from the phase transformation of the base metal and those at 1000 and 1050 °C could have resulted from the lateral cracks in the oxide due to the ZrO₂ phase transformation from a monoclinic structure to a tetragonal structure.     

Oxidation properties of Zr–Nb alloys at 973–1273 K in air

In order to investigate the oxidation behavior of LWR cladding materials under the condition of reactor accidents, e.g. LOCA, Zr–Nb alloys with 1–10 wt%Nb and Zircaloy-4 (0 wt%Nb) were oxidized at 973–1273 K in dry air. The weight gain due to oxidation increased with Nb content at 973 and 1073 K was the smallest for 2.5 wt%Nb at 1173 and 1273 K. The oxidation kinetics obeyed the parabolic rate law without a few cases, e.g. 6–10 wt%Nb and 1273 K. The parabolic rate constant at high temperatures had the somewhat low activation energy compared to that at low temperatures. These results implied that such oxidation behaviors of Zr–Nb alloys related to the lattice structures of oxide films as well as underlying metal during oxidation. Especially at high temperatures, 6ZrO₂–Nb₂O₅ compound might promote the oxidation of Zr–Nb alloys with high content of Nb.     

Steam oxidation of Zr-1.5Nb-0.4Sn-0.2Fe-0.1Cr and Zircaloy-4 at 900-1200 °C

The steam oxidation characteristics for the Zr-1.5Nb-0.4Sn-0.2Fe-0.1Cr (HANA-4) and Zircaloy-4 claddings were elucidated at LOCA temperatures of 900-1200 °C by using a modified thermo-gravimetric analyzer. After the oxidation tests, the oxidation behaviors, oxidation rates, surface appearances, and microstructures of the as-received, as-oxidized, and burn-up simulated claddings were evaluated in this study. The high-temperature oxidation resistance of the as-received HANA-4 cladding was superior to that of the Zircaloy-4. The superior oxidation resistance of the HANA-4 cladding could be attributed to the higher Nb and the lower Sn within its cladding. The pre-oxidized layer formed at the low temperatures below 500 °C could retard the oxidation rate at the high temperatures above 900 °C. And the soundness of the pre-oxidized layer formed at a lower temperature could influence the oxidation kinetics and the rate constants during a steam oxidation at LOCA temperatures from 900 to 1200 °C.     

High temperature oxidation of some zirconium alloys in flowing oxygen

The oxidation behavior of three zirconium alloys, Zr-2.2wt%Hf, Zr-2.5wt%Nb and Zr-3wt%Nb-1wt%Sn, has been studied in flowing oxygen in the temperature range 873–1173 K to 120 ks (2000 min). Zr-2.5Nb and Zr-3Nb-1Sn showed a transition to rapid linear kinetics after initial parabolic oxidation at all temperatures. Zr-2.2Hf, on the other hand, showed this transition at temperatures in the range 973–1173 K; no transition was observed at 873 K within the oxidation times reported. Zr-2.2Hf showed the smallest weight gains, followed by Zr-2.5Nb and Zr-3Nb-1Sn. Increased oxidation rates and shorter time-to-rate transition of Zr-2.5Nb and Zr-3Nb-1Sn as compared with Zr-2.2Hf are attributed to the presence of the alloying elements Nb, Sn and Hf. Based on the Nomura-Akutsu model, Hf should delay the rate transition, while Nb and Sn lead to shorter transition times. The scale on Zr-2.2Hf was identified as monoclinic zirconia, while the tetragonal phase, 6ZrO₂ · Nb₂O₅, was contained in the monoclinic zirconia scales on both other alloys.     

Synopsis and outcome of the QUENCH experimental program

The paper gives an overview of the main outcome of the QUENCH program launched in 1997 at the Karlsruhe Institute of Technology (KIT), formerly Karlsruhe Research Center (FZK). The research program comprises bundle experiments as well as complementary separate-effects tests. The focus of the experiments performed from 1997 to 2009 was on scenarios of severe accidents whereas that of the future test program will be on large-break loss-of-coolant accidents (LOCA) in the frame of design-basis accidents, and debris coolability, in the frame of severe accidents. The major objective of the program is to deliver experimental and analytical data to support the development and validation of quench and quench-related models as used in code systems that model severe accident progression in light water reactors.So far, 15 integral bundle QUENCH experiments with 21-31 electrically heated fuel rod simulators of 2.5 m length have been conducted. The following parameters and their influence on bundle degradation and reflood have been investigated: degree of pre-oxidation, temperature at initiation of reflood, flooding rate, influence of neutron absorber materials (B₄C, AgInCd), air ingress, and influence of the type of cladding alloy.In six tests, reflooding of the bundle led to a temporary temperature excursion driven by runaway oxidation of zirconium alloy components and resulting in release of a significant amount of hydrogen, typically two orders of magnitude greater than in those tests with “successful” quenching in which cool-down was rapidly achieved. Considerable formation, relocation, and oxidation of melt were observed in all tests with escalation. The temperature boundary between rapid cool-down and temperature escalation was typically in the range of 2100-2200 K in the “normal” quench tests, i.e. in tests without absorber and/or steam starvation. Tests with absorber and/or steam starvation were found to lead to temperature escalations at lower temperatures.All phenomena occurring in the bundle tests have been investigated additionally in parametric and more systematic separate-effects tests. Oxidation kinetics of various cladding alloys, including advanced ones, have been determined over a wide temperature range (873-1773 K) in different atmospheres (steam, oxygen, air, and their mixtures). Hydrogen absorption by different zirconium alloys was investigated in detail, recently also using neutron radiography as non-destructive method for determination of hydrogen distribution in claddings. Furthermore, degradation mechanisms of absorber rods including B₄C and AgInCd as well as the oxidation of the resulting low-temperature melts have been studied. Steam starvation was found to cause deterioration of the protective oxide scale by thinning and chemical reduction.The most recent topic of the QUENCH program has been investigation of the behavior of advanced cladding materials (ACM) in comparison with the classical Zircaloy-4. Although separate-effects tests have shown some differences in oxidation kinetics, the influence of the various cladding alloys on the integral bundle behavior during oxidation and reflooding was only limited.     

Air oxidation of Zircaloy-4 in the 600-1000 °C temperature range: Modeling for ASTEC code application

Progress in the treatment of air oxidation of zirconium in severe accident (SA) codes are required for a reliable analysis of severe accidents involving air ingress. Air oxidation of zirconium can actually lead to accelerated core degradation and increased fission product release, especially for the highly-radiotoxic ruthenium. This paper presents a model to simulate air oxidation kinetics of Zircaloy-4 in the 600-1000 °C temperature range. It is based on available experimental data, including separate-effect experiments performed at IRSN and at Forschungszentrum Karlsruhe. The kinetic transition, named “breakaway”, from a diffusion-controlled regime to an accelerated oxidation is taken into account in the modeling via a critical mass gain parameter. The progressive propagation of the locally initiated breakaway is modeled by a linear increase in oxidation rate with time. Finally, when breakaway propagation is completed, the oxidation rate stabilizes and the kinetics is modeled by a linear law. This new modeling is integrated in the severe accident code ASTEC, jointly developed by IRSN and GRS. Model predictions and experimental data from thermogravimetric results show good agreement for different air flow rates and for slow temperature transient conditions.     

Oxidation kinetics of Zry-4 fuel cladding in mixed steam–air atmospheres at temperatures of 1273–1473 K

This paper deals with the oxidation behavior of Zry-4 nuclear fuel cladding tubes in mixed steam–air atmospheres at temperatures of 1273 and 1473 K. The main goal is to study the oxidation kinetics of Zry-4 fuel cladding in dependence on the air fraction in steam in the range from 0% up to 100%. The purpose of this study is to provide experimental data suitable for an oxidation correlation, applicable for severe accident computer codes of nuclear power reactors. The influence of the air addition in steam on parameters of Zry-4 kinetic equation has been quantified using the results of weight gain measurements. At 1273 K, both pre-transition and post-transition regimes were treated. The results of weight gain measurements showed a strong dependence of the Zry-4 oxidation kinetics on the air fraction in steam, especially at 1473 and at 1273 K in the post-transition regime.     

Oxidation behavior of high-burnup advanced fuel cladding tubes in high-temperature steam

To evaluate the oxidation behavior of high-burnup advanced fuel cladding tubes in high-temperature steam, laboratory-scale isothermal oxidation tests were conducted using the following advanced fuel cladding tubes with burnups of up to 85 GWd/t: M-MDA^TM, low-tin ZIRLO^TM, M5®, and Zircaloy-2 (LK3). These oxidation tests were performed in steam-flowing conditions at temperatures ranging from 1173 to 1473 K for durations between 120 and 4000 s, and the oxidation kinetics was evaluated. The oxidation kinetics of the high-burnup advanced fuel cladding tube specimens estimated by assuming the parabolic rate law was comparable to or slower than that of the unirradiated Zircaloy-4 cladding tube specimens reported in a previous study. It is considered that the protective e?ect of the corrosion layer hindered oxidation. Furthermore, no increase in the oxidation kinetics because of the pre-hydriding was observed. The onset times of the breakaway oxidations of these cladding tube specimens were comparable to those of the unirradiated Zircaloy-4 cladding tubes reported in previous studies. Therefore, it is considered that the burnup extension up to 85 GWd/t and the use of the advanced fuel cladding tubes do not significantly increase the oxidation kinetics and do not significantly reduce the onset time of the breakaway oxidation.     

Microstructure evolution in Zr1Nb fuel cladding during high-temperature oxidation

This work is concerned with the microstructure evolution during high-temperature oxidation inside the cladding wall fabricated of Zr1Nb alloy. The kinetics of the α-Zr(O) growth in the temperature interval of 800-1200 °C has been assessed using LM, SEM, and an image analyzer. In addition, the results of nanohardness measurement were used to determine the thickness of the α-Zr(O) layer. Eventually, the experimental results were compared to analytical calculations. Very good agreement has been achieved between experimental and theoretical results, mainly for lower thicknesses. The paper also deals with the kinetics of oxide and (α + β)-Zr layers. The oxide layer growth kinetics changes depending on temperature. The kinetics of the (α + β)-Zr layer is independent of temperature and slightly faster when compared to the parabolic growth.     

Separate-effect tests on zirconium cladding degradation in air ingress situations

In the event of air ingress during a reactor or spent fuel pond low probability accident, the fuel rods will be exposed to air-containing atmospheres at high temperatures. In comparison with steam, the presence of air is expected to result in a more rapid escalation of the accident.A state-of-the-art review performed before SARNET started showed that the existing data on zirconium alloy oxidation in air were scarce. Moreover, the exact role of zirconium nitride on the cladding degradation process was poorly understood. Regarding the cladding behaviour in air + steam or nitrogen-enriched atmospheres (encountered in oxygen-starved conditions), almost no data were available.New experimental programmes comprising small-scale tests have therefore been launched at FZK, IRSN (MOZART programme in the frame of the International Source Term Program—ISTP) and INR. Zircaloy-4 cladding in PWR (FZK, IRSN) and in CANDU (INR) geometry are investigated. On-line kinetic data are obtained on centimetre size tube segments, by thermogravimetry (FZK, IRSN and INR) or by mass spectrometry (FZK). Plugged tubes 15 cm long (FZK) are also investigated. The samples are air-oxidised either in the “as-received” state, or after pre-oxidation in steam. “Analytical” tests at constant temperature and gas composition provide basic kinetic data, while more prototypical temperature transients and sequential gas compositions are also investigated. The temperature domains extend from 600 °C up to 1500 °C. Systematic post-test metallographic inspections are performed.The paper gives a synthesis of the results obtained, comparing them in terms of kinetics and oxide scale structure and composition. A comparative analysis is performed with results of the QUENCH-10 (Q-10) bundle test, which included an air ingress phase. It is shown how the data contribute to a better understanding of the cladding degradation process, especially regarding the role of nitrogen. For modelling of the oxide scale degradation under air exposure, important features that have to be taken into account are highlighted.     

Radiation-damage in molybdenum-rhenium alloys for space reactor applications

Various Mo-Re alloys are attractive candidates for use as fuel cladding and core structural materials in spacecraft reactor applications. Molybdenum alloys with rhenium contents of 41-47.5% (wt%), in particular, have good creep resistance and ductility in both base metal and weldments. However, irradiation-induced changes such as transmutation and radiation-induced segregation could lead to precipitation and, ultimately, radiation-induced embrittlement. The objective of this work is to evaluate the performance of Mo-41Re and Mo-47.5Re after irradiation at space reactor relevant temperatures. Tensile specimens of Mo-41Re and Mo-47.5Re alloys were irradiated to ∼0.7 displacements per atom (dpa) at 1073, 1223, and 1373 K and ∼1.4 dpa at 1073 K in the High Flux Isotope Reactor at Oak Ridge National Laboratory. Following irradiation, the specimens were strained to failure at a rate of 1 × 10⁻³ s⁻¹ in vacuum at the irradiation temperature. In addition, unirradiated specimens and specimens aged for 1100 h at each irradiation temperature were also tested. Fracture mode of the tensile specimens was determined. The tensile tests and fractography showed severe embrittlement and IG failure with increasing temperatures above 1100 K, even at the lowest fluence. This high temperature embrittlement is likely the result of irradiation-induced changes such as transmutation and radiation-induced segregation. These factors could lead to precipitation and, ultimately, radiation-induced embrittlement. The objective of this work is to examine the irradiation-induced degradation for these Mo-Re alloys under neutron irradiation.     

Hydride behavior in Zircaloy cladding tube during high-temperature transients

In order to study the hydride behavior in high burnup fuel cladding during temperature transients expected in anticipated operational occurrences and accidents, unirradiated hydrided Zircaloy-4 cladding tubes were rapidly heated to temperatures ranging from 673 to 1173 K and annealed for holding time ranging from 0 to 3600 s. Hydrides were localized in the peripheral region of the cladding tubes prior to the annealing, as observed in high burnup fuel cladding. The localized hydride layer (hydride rim) was annealed out, and the radial hydride distribution became uniform after the annealing at 873 K for 600 s, 973 K for 60 s, or 1173 K for 0 s. The annealing out of the hydride rim is caused by the phase transformation from the α + δ phase to the α + β or β phase in the hydride rim and the subsequent drastic increase in the solid solubility and diffusion of hydrogen in Zircaloy. On the other hand, the radial distribution and morphology of hydrides did not change at lower temperatures: Thus, the hydride remains almost intact below the phase transformation temperature for the short time ranges.     

Experimental and post-test calculation results of the integral reflood test QUENCH-12 with a VVER-type bundle

The QUENCH-12 experiment was carried out to investigate the effects of VVER materials (niobium-bearing alloys) and bundle geometry on core reflood, in comparison with test QUENCH-06 using western PWR materials (Zircaloy-4) and bundle geometry. The test protocol involved pre-oxidation to a maximum of about 150 μm oxide thickness at a temperature of about 1450 K, followed by a power ramp until a temperature of 2050 K was reached, then reflood with water at room temperature was initiated. The total hydrogen production was 58 g (QUENCH-06: 36 g), 24 g of which were released during reflood (QUENCH-06: 4 g). Reasons for the increased hydrogen production may be extensive damaging of the cladding surfaces due to the breakaway oxidation and local melt formation with subsequent melt oxidation. Post-test videoscope observations and metallographic investigations showed an influence of the breakaway oxidation with extensive spalling of oxide scales of rod claddings, shroud and auxiliary corner rods. The hydrogen content in the corner rods, withdrawn from the bundle during the test, reached more than 30 at% at the bundle elevations of 850 and 1100 mm. Post-test calculations were performed with local versions of SCDAP/RELAP5 following on from pre-test analyses with SCDAP/RELAP5 and SCDAPSIM.     

A review of refractory metal alloys and mechanically alloyed-oxide dispersion strengthened steels for space nuclear power systems

Mechanical and thermo-physical properties of refractory metal alloys and mechanically alloyed (MA)-oxide dispersion strengthened (ODS) steels are reviewed and their potential for use in space nuclear reactors is examined. Preferable refractory alloys for use in liquid metal and gas-cooled space reactors include Nb-1%Zr, PWC-11, Mo-TZM, Mo-xRe where x varies from 7% to 44.5%, T-111 and ASTAR-811C. These alloys are heavy, difficult to fabricate, and are not readily available. The advantages of the MA-ODS alloys are: (a) their strength at high temperatures (>1000 K), which decreases slower with temperature than those of niobium and molybdenum alloys; (b) relatively lightweight and less expensive; (c) low swelling and no embrittlement with exposure to high-energy neutrons (>0.1 MeV) up to 10²⁷ n/m²; and (d) high resistance to oxidation and nitration. The few data available on compatibility of MA-ODS alloys with alkali liquid metals up to 1100 K are encouraging, however, additional tests at typical temperatures (1000-1400 K) in space nuclear reactors are needed. The anisotropy of MA-ODS alloys when cold worked, and particularly rolled into tubes, should not hinder their use in space nuclear power systems, in which operation pressure is either near atmospheric or as high as 2 MPa, but joints weldability is an issue.     

High-pressure hydriding of Zircaloy cladding by the thermogravimetry and tube-burst techniques

Hydriding kinetics of modified Zircaloy claddings was studied by the thermogravimetric method at 400 °C and the tube-burst technique at 315 °C. Some specimens were prefilmed with a thin oxide layer by air oxidation on both the inner and outer surfaces which were either pickled or blasted. In the thermogravimetric test, the hydriding rates of bare cladding specimens (no oxide prefilm) were in the range 0.9-1.6 mg/cm² h with little effect of the surface treatment. Incubation times were less than 1 h or even zero. In the tube-burst test, immediate and extensive hydrogen uptake was observed for these non-coated specimens. On the other hand, the cladding specimens with oxide prefilm exhibited lower hydriding rates ranging from 0.01 to 0.05 mg/cm² h and incubation times increased to 42 h. In addition, no hydrogen uptake was observed for all oxide-coated specimens for 100-750 h.     

Response of nanoclusters in a 9Cr ODS steel to 1 dpa, 525 °C proton irradiation

Ferritic-martensitic (F/M) alloys are expected to play an important role as cladding or structural components in Generation IV and other advanced nuclear systems operating in the temperature range 350-700 °C and to doses up to 200 displacements per atom (dpa). Oxide dispersion strengthened (ODS) F/M steels have been developed to operate at higher temperatures than traditional F/M steels. These steels contain nanometer-sized Y-Ti-O nanoclusters for additional strengthening. A proton irradiation to 1 dpa at 525 °C has been performed on a 9Cr ODS steel to determine the nanocluster stability at low dose. The evolution of the nanocluster population and the composition at the nanocluster-matrix interface were studied using electron microscopy and atom probe tomography. The data from this study are contrasted to those from a previous study on heavy-ion irradiated 9Cr ODS steel.     

Oxidation kinetics and oxygen diffusion in low-tin Zircaloy-4 up to 1523 K

This paper deals with the study of oxidation kinetics and the identification of oxygen diffusion coefficients of low-tin Zy-4 alloy at intermediate (973 K ? T ? 1123 K) and high temperatures (T ? 1373 K). Two different cases were considered: dissolution of a pre-existing oxide layer in the temperature range 973 K ? T ? 1123 K and oxidation at T ? 1373 K. The results are the following ones: in the temperature range 973-1123 K, the oxygen diffusion coefficient in α_Zr phase can be expressed as D_α = 6.798 exp(−217.99 kJ/RT) cm²/s. In the temperature range 1373-1523 K, the oxygen diffusion coefficients in α_Zr, β_Zr and ZrO₂, were determined using an ‘inverse identification method’ from experimental high temperature oxidation data (i.e., ZrO₂, and α_Zr(O) layer thickness measurements); they can be expressed as follows: D_α = 1.543 exp(−201.55 kJ/ RT) cm²/s, D_β = 0.0068 exp(−102.62 kJ/ RT) cm²/s and D_ZrO2=0.115exp(−143.64kJ/RT)cm²/s. Finally an oxygen diffusion coefficient in α_Zr in the temperature range 973 K ? T ? 1523 K was determined, by combining the whole set of results: D_α = 4.604exp(−214.44 kJ/RT) cm²/s. In order to check these calculated diffusion coefficients, oxygen concentration profiles were determined by Electron Probe MicroAnalysis (EPMA) in pre-oxidized low-tin Zy4 alloys annealed under vacuum at three different temperatures 973, 1073 and 1123 K for different times, and compared to the calculated profiles. At last, in the framework of this study, it appeared also necessary to reassess the Zr-O binary phase diagram in order to take into account the existence of a composition range in the two zirconia phases, α_ZrO2 and β_ZrO2.     

Experimental and calculation results of the integral reflood test QUENCH-15 with ZIRLO™ cladding tubes in comparison with results of previous QUENCH tests

The QUENCH-15 experiment investigated the effect of ZIRLO™¹ cladding material on bundle oxidation and core reflood, in comparison with tests QUENCH-06 (standard Zircaloy-4), QUENCH-12 (VVER, E110), and QUENCH-14 (M5^®). The QUENCH-15 bundle cross-section corresponded to a Westinghouse PWR core design and consisted of 24 heated rods (internal tungsten heaters between 0 and 1024 mm axial elevation, cladding oxidised region between −470 and 1500 mm), six corner rods made of Zircaloy-4, two corner rods made of E110, and a Zirconium 702 shroud. The test was conducted in principle with the same protocol as QUENCH-06, -12 and -14, so that the effects of the change of cladding material and bundle geometry could be more easily observed. The test protocol involved pre-oxidation to a maximum of about 150 μm oxide thickness at a temperature of about 1473 K over a period of about 3000s. The power was then ramped at a rate of 0.25 W/s/rod to cause a temperature increase until the desired maximum bundle temperature of 2153 K was reached. The maximum oxide layer thickness observed was 380 μm. Then reflood with 1.3 g/s/rod water at room temperature was initiated. The electrical power was reduced to 175 W/rod during the reflood phase, approximating effective decay heat level. The post-test metallography of the bundle showed neither noticeable breakaway oxidation of the cladding nor melt release into space between rods. The average outer oxide layer thickness at hottest elevation of 950 mm was 620 μm (QUENCH-06: 630 μm). The molten cladding metal at hottest elevation was localised between the outer and inner oxide layers. The thickness of inner oxide layer reaches 20% of that of the outer oxide layer. The measured hydrogen release during the QUENCH-15 test was 41 g in the pre-oxidation and transient phases and 7 g in the quench phase which are comparable with those in QUENCH-06, i.e. 32 g and 4 g, respectively. Post-test calculations were performed using a version of SCDAP/RELAP5/MOD3.2. The calculation results support the heuristic observation that there was no major difference between the influence of Zircaloy-4, M5^® or ZIRLO™ for the beyond-design basis accident present conditions here studied.     

Oxidation kinetics of magnesium alloys treated by tantalum ions implantation

AZ31 magnesium alloys were implanted with tantalum ions with doses of 1 × 10¹⁶, 5 × 10¹⁶ and 1 × 10¹⁷ ions/cm², using a metal vapor vacuum arc (MEVVA) at an extraction voltage of 45 kV. Auger electron spectroscopy (AES) and X-ray photoelectron spectroscopy (XPS) analysis suggested that tantalum ions implantation promoted the formation of the pre-oxidation layer and a new Ta₂Al phase was formed in the implanted layer. Then, the oxidation kinetics of the implanted specimens was investigated by isothermal oxidation at 773 K in pure O₂ up to 90 min. The results showed that after implantation treatments the oxidation resistance of the specimens was significantly improved and the specimen with the highest dose had the best oxidation resistance. Finally, the mechanism of the anti-oxidation effects was also discussed.