Showing papers on "Zirconium alloy published in 1996"

Quinary metallic glass alloys

Abstract: At least quinary alloys form metallic glass upon cooling below the glass transition temperature at a rate less than 10 3 K/s. Such alloys comprise zirconium and/or hafnium in the range of 45 to 65 atomic percent, titanium and/or niobium in the range of 4 to 7.5 atomic percent, and aluminum and/or zinc in the range of 5 to 15 atomic percent. The balance of the alloy compositions comprise copper, iron, and cobalt and/or nickel. The composition is constrained such that the atomic percentage of iron is less than 10 percent. Further, the ratio of copper to nickel and/or cobalt is in the range of from 1:2 to 2:1. The alloy composition formula is: (Zr,Hf).sub.a (Al,Zn).sub.b (Ti,Nb).sub.c (Cu.sub.x Fe.sub.y (Ni,Co) z ) d wherein the constraints upon the formula are: a ranges from 45 to 65 atomic percent, b ranges from 5 to 15 atomic percent, c ranges from 4 to 7.5 atomic percent, d comprises the balance, d·y is less than 10 atomic percent, and x/z ranges from 0.5 to 2.

Corrosion resistance and corrosion fatigue strength of new titanium alloys for medical implants without V and Al

Abstract: The corrosion resistance and the corrosion fatigue strength of Ti-15Zr-4Nb-4Ta-0.2Pd-0.2O-0.05N and Ti-15Sn-4Nb-2Ta-0.2Pd-0.2O alloys were compared with those of Ti-6Al-4 V extra low interstitial (ELI), Ti-6Al-2Nb-1Ta, pure Ti grade 2 and β type Ti-15%Mo-5Zr-3Al alloys. Anodic polarization and corrosion fatigue testings were performed in various physiological saline solutions at 310 K. The corrosion fatigue test was carried out under the condition of a tension to tension mode with a sine wave at a stress ratio of 0.1 and at a frequency of 10 Hz. The tensile properties of these alloys were measured at room temperature. The change in current density was small up to passivity zone in 1 wt.% lactic acid, PBS(−), calf serum and eagle's MEM + fetal bovine serum solutions except 5 wt.% HCl. The current density of Ti-15Zr-4Nb-4Ta-0.2Pd-0.2O-0.05N alloy at potential up to 5 volt tend to be lower than that of Ti-6Al-4V ELI. Otherwise passive current density of the β type Ti-15Mo-SZr-3Al alloy was higher than that of α + β type alloys. The passive films formed on Ti-15Zr-4Nb-4Ta-0.2Pd-0.2O alloy in the calf serum consisted mainly of TiO2, ZrO2, Nb2O5, Ta2O5 and Pd or PdO as demonstrated using X-ray photoelectron spectroscopy. The cycle to failure of Ti-15Zr-4Nb-4Ta-0.2Pd-0.2O-0.05N and Ti-15Sn-4Nb-4Ta-0.2Pd-0.2O alloys annealed at 973 K for 7.2 ks increased with decreasing applied maximum stress. The fatigue strength at 108 cycles in those alloys was about 600 MPa. The fatigue strength of Ti-6Al-2Nb-1Ta alloy at 108 cycles was about 700 MPa. The fatigue strength of β type Ti-15Mo-5Zr-3Al alloy at 107 cycles was lower than that of α + β type alloys.

METALLIC GLASS ALLOYS OF Zr, Ti, Cu AND Ni

Abstract: At least quaternary alloys form metallic glass upon cooling below the glass transition temperature at a rate less than 103 K/s. One group of such alloys comprise titanium from 19 to 41 atomic percent, an early transition metal (ETM) from 4 to 21 atomic percent and copper plus a late transition metal (LTM) from 49 to 64 atomic percent. The ETM comprises zirconium and/or hafnium. The LTM comprises cobalt and/or nickel. The composition is further constrained such that the product of the copper plus LTM times the atomic proportion of LTM relative to the copper is from 2 to 14. The atomic percentage of ETM is less than 10 when the atomic percentage of titanium is as high as 41, and may be as large as 21 when the atomic percentage of titanium is as low as 24. Furthermore, when the total of copper and LTM are low, the amount of LTM present must be further limited. Another group of glass forming alloys has the formula: (ETM?1-x?Tix)aCub(Ni1-yCoy)c, wherein x is from 0.1 to 0.3, y?.?c is from 0 to 18, a is from 47 to 67, b is from 8 to 42, and c is from 4 to 37. This definition of the alloys has additional constraints on the range of copper content.

Stable Ti-based quasicrystal offers prospect for improved hydrogen storage

Abstract: The desorption of hydrogen from a novel material, a Ti45Zr38Ni17‐H quasicrystal, was observed using high‐temperature powder x‐ray diffraction, demonstrating the potential utility of Ti‐based quasicrystals in place of crystalline or amorphous hydrides for hydrogen storage applications. The maximum observed change in hydrogen concentration was from 61 at. %, corresponding to a hydrogen‐to‐metal ratio (H/M) of 1.54, at 91 °C to less than 2.5 at. % (H/M=0.025) at 620 °C. The onset temperature of desorption is below 350 °C. Surface oxidation was found to promote the formation of crystalline hydride phases. Highly oxidized samples transformed to a mixture of the C14 Laves and C15 Laves crystalline hydrides, and the Ti2Ni phase. When the oxidation was less severe, a reversible transformation between the quasicrystal and crystalline hydride phases was clearly observed, demonstrating the stability of the Ti45Zr38Ni17 quasicrystal at very low hydrogen concentrations, and temperatures as high as 661 °C. This is the ...

Hydrogen pickup and redistribution in alpha-annealed Zircaloy-4

Abstract: Zircaloy-4, which is widely used as a core structural material in Pressurized-Water Reactors (PWR), picks up hydrogen during service. Hydrogen solubility in Zircaloy-4 is low and hydrides precipitate after the Zircaloy-4 matrix becomes supersaturated with hydrogen. These hydrides embrittle the Zircaloy-4. To study hydrogen pickup and concentration, a postirradiation nondestructive radiographic technique for measuring hydrogen concentration was developed and qualified. Experiments on hydrogen pickup were conducted in the Advanced Test Reactor (ATR). Ex-reactor tests were conducted to determine the conditions for which hydrogen would dissolve, migrate, and precipitate. Finally, a phenomenological model for hydrogen diffusion was indexed to the data. This presentation describes the equipment and the model, presents the results of experiments, and compares the model predictions to experimental results.

Zirconium in the nuclear industry : eleventh international symposium

Abstract: Since their development in the 1950s and introduction into commercial nuclear power plants in the 1960s, the zirconium-based alloys Zircaloy-2 and -4, Zr-1Nb, and Zr-2.5Nb are the alloys currently used in the world`s reactors. However, with increasing fuel duty, the margins displayed by these alloys have eroded, and considerable research has been conducted to improve these materials and also to develop more advanced alloys. Optimization of alloying constituents and processing parameters, coupled with a more basic understanding of performance-limiting phenomena, are the primary themes of most of the papers contained herein. Fully half of the papers are directly concerned with the corrosion of Zr-based alloys. The detailed characterization of the effects of irradiation on the microstructure of irradiated Zircaloys has confirmed the loss of iron from second phase particles, with or without amorphization of the particles. Also, the correspondence between iron in solution in the matrix, the formation of -type dislocations, and the onset of accelerated irradiation induced growth has been verified. Unfortunately, the role of dissolved iron in the matrix in the nucleation of -type dislocations has not been established. One observation that has been verified, however, is the low irradiation growth in Zr-Nb-Sn-Fe alloys, also presumably due tomore » suppression of dislocation formation. Several papers in the symposium focused on fuel clad modeling, and although a schism still exists between fundamental material properties and fuel performance predictive codes, the modeling papers presented are attempts to link component response to the quantifiable material behavior in a manner consistent with qualitative structural observations. In summary, the data, analyses, hypotheses, and theories presented in this book represent the current state of zirconium technology as applied to nuclear power reactors. Separate abstracts were prepared for 43 papers in this volume.« less

Microstructure and phase identification in type 304 stainless steel-zirconium alloys

Abstract: Stainless steel-zirconium alloys have been developed at Argonne National Laboratory to contain radioactive metal isotopes isolated from spent nuclear fuel. This article discusses the various phases that are formed in as-cast alloys of type 304 stainless steel and zirconium that contain up to 92 wt pct Zr. Microstructural characterization was performed by scanning electron microscopy (SEM) and energy dispersive X-ray spectroscopy (EDS), and crystal structure information was obtained by X-ray diffraction. Type 304SS-Zr alloys with 5 and 10 wt pct Zr have a three-phase microstructure—austenite, ferrite, and the Laves intermetallic, Zr(Fe,Cr,Ni)2+x. whereas alloys with 15, 20, and 30 wt pct Zr contain only two phases—ferrite and Zr(Fe,Cr,Ni)2+x. Alloys with 45 to 67 wt pct Zr contain a mixture of Zr(Fe,Cr,Ni)2+x and Zr2(Ni,Fe), whereas alloys with 83 and 92 wt pct Zr contain three phases—α-Zr, Zr2(Ni,Fe), and Zr(Fe,Cr,Ni)2+x. Fe3Zr-type and Zr3Fe-type phases were not observed in the type 304SS-Zr alloys. The changes in alloy microstructure with zirconium content have been correlated to the Fe-Zr binary phase diagram.

Evolution of Microstructure in Zirconium Alloys During Irradiation

Abstract: X-ray diffraction (XRD) and transmission electron microscopy (TEM) have been used to characterize microstructural and microchemical changes produced by neutron irradiation in zirconium and zirconium alloys. Zircaloy-2, Zircaloy-4, and Zr-2.5Nb alloys with differing metallurgical states have been analyzed after irradiation for neutron fluences up to 25 {times} 10{sup 25} n.m{sup {minus}2} for a range of temperatures between 330 and 580 K. Irradiation modifies the dislocation structure through nucleation and growth of dislocation loops and, for cold-worked materials in particular, climb of existing network dislocations. in general, the a-type dislocation structure tends to saturate at low fluences. The c-component dislocation structure, however, may evolve over long periods of irradiation. The phase structure is also modified by irradiation. The common alloying/impurity elements, Fe, Cr, and Ni, are relatively insoluble in the {alpha}-phase during irradiation irrespective of the state of the phase initially containing these elements, i.e., metastable {beta}-phase or stable intermetallic precipitate. The stable intermetallic particles may undergo structural changes dependent on their composition and the temperature. For the metastable dual-phase {alpha}/{beta}-alloys (Zr-2.5Nb alloy), the {beta}-phase structure is modified during irradiation, but the change is complex, being a combination of thermal decomposition and radiation-induced mixing.

Role of Intermetallic Precipitates in Hydrogen Uptake of Zircaloy-2

Abstract: The hydrogen uptake behavior of two types of Zircaloy-2 specimens containing either fine intermetallic precipitates or coarse ones was studied at 623–723 Kin the pre-transition period of oxidation to clarify the role of the precipitates in the hydrogen transport through the oxide film. In the former case, the amount of hydrogen taken up was small and did not show the oxidation temperature dependence. In the latter case, the amount of hydrogen taken up was large and dependent markedly upon the oxidation temperature; it increased with decreasing oxidation temperature. These results were successfully explained with the model that the intermetallic precipitates remaining in a metallic state in the oxide film act as the fast transport route of hydrogen.

Zirconium Alloy E635 as a Material for Fuel Rod Cladding and Other Components of VVER and RBMK Cores

Abstract: Data are given on Zr alloy E635 (Zr-1.2Sn-1Nb-0.4Fe), developed in Russia as a fuel rod cladding and other component material for use in cores of VVER and RBMK types. The alloy is much superior to binary alloys with 1.0 and 2.5% Nb and Zircaloys in terms of its resistance to irradiation-induced creep and growth and nodular corrosion. The creep rate of the alloy is slightly dependent on irradiation temperature, stress, neutron fluence, and neutron density. The alloy is subject to substantial irradiation hardening while retaining its high-percent elongation. Corrosion, creep, and growth resistances are slightly dependent on the structure of components (alloy, final product). Based on the previously studied influence of impurities, structure, heat treatment, and working schedules, the technological processes were designed and mastered commercially for fabrication of tubes, bars, strips, and fuel rod claddings from this alloy. Components are produced commercially. Fuel assemblies with fuel rods clad in the E635 alloy were successfully tested in the RBMK reactor at the Leningrad NPP as well as in experimental reactors under VVER-1000 conditions. Today, the E635 alloy is recommended as a promising material for use in cores of VVER-1000 and VVER of new generations as well as RBMK-type reactors havingmore » a longer fuel cycle.« less

Temperature dependence of the mixing enthalpy and excess heat capacity in the liquid system nickel-zirconium

Abstract: The mixing enthalpyΔHm of the liquid system Ni-Zr was measured in the Ni-rich range at 1916 K up toxZr=0.34 at % and for the first time in the Zr-rich range at 2270 K up toxNi = 0.54 at %. Using the thermodynamic-dapted power series, a composition- and temperature-dependent description ofΔHm was given. Furthermore, the partial differentiation ofΔHm (x, T) by T yielded the excess heat capacityCpxs(x, T). The existence of chemical short-range order (associate) in the vicinity of Ni7Zr2 and NiZr was shown and was discussed with reference toΔHm(x, T) andCpxs(x, T) ( 1748 to 2270 K). With decreasing temperature, the influence of chemical short-range order tended toward the composition NiZr.

Microstructure of Oxide Films Formed during the Waterside Corrosion of the Zircaloy-4 Cladding in Lithiated Environment

Abstract: Zircaloy-4 cladding materials have been oxidized in a lithiated environment in autoclave and in out-of-pile loop tests. In such oxidation tests, a strong enhancement of the oxidation rate can occur depending on the water chemistry conditions and on the oxidation time. In this work, in order to improve the understanding of the detrimental effect of lithium on the corrosion behavior of the Zircaloy-4 cladding, the microstructure of oxide films has been characterized by TEM. Simultaneously, the lithium profiles and concentrations in the oxide layers have been determined using the SIMS technique, special attention being paid to the metal-oxide interface. These experimental results are described extensively and analyzed, and their contribution to the understanding of the influence of lithium on the oxidation process of Zircaloy-4 cladding material is discussed as a function of an existing thin inner barrier layer at the metal-oxide interface.

Mechanical and Thermal Properties of Two Cu-Cr-Nb Alloys and NARloy-Z

Abstract: A series of creep tests were conducted on Cu-8 Cr-4 Nb (Cu-8 at.% Cr-4 at.% Nb), Cu-4 Cr-2 Nb (Cu-4 at.% Cr-2 at% Nb), and NARloy-Z (Cu-3 wt.% Ag-0.5 wt.% Zr) samples to determine their creep properties. In addition, a limited number of low cycle fatigue and thermal conductivity tests were conducted. The Cu-Cr-Nb alloys showed a clear advantage in creep life and sustainable load over the currently used NARloy-Z. Increases in life at a given stress were between 100% and 250% greater for the Cu-Cr-Nb alloys depending on the stress and temperature. For a given life, the Cu-Cr-Nb alloys could support a stress between 60% and 160% greater than NARloy-Z. Low cycle fatigue lives of the Cu-8 Cr-4 Nb alloy were equivalent to NARloy-Z at room temperature. At elevated temperatures (538 C and 650 C), the fatigue lives were 50% to 200% longer than NARloy-Z samples tested at 538 C. The thermal conductivities of the Cu-Cr-Nb alloys remained high, but were lower than NARloy-Z and pure Cu. The Cu-Cr-Nb thermal conductivities were between 72% and 96% that of pure Cu with the Cu-4 Cr-2 Nb alloy having a significant advantage in thermal conductivity over Cu-8 Cr4 Nb. In comparison, stainless steels with equivalent strengths would have thermal conductivities less than 25% the thermal conductivity of pure Cu. The combined results indicate that the Cu-Cr-Nb alloys offer an attractive alternative to current high temperature Cu-based alloys such as NARloy-Z.

Influence of Processing Variables and Alloy Chemistry on the Corrosion Behavior of ZIRLO Nuclear Fuel Cladding

Abstract: Variations in the thermal heat treatments used during the fabrication of ZIRLO (Zr-1Nb-1Sn-0.1Fe) fuel clad tubing and in ZIRLO alloy chemistry were explored to develop a further understanding of the relationship between processing, microstructure, and cladding corrosion performance. Heat treatment variables included intermediate tube annealing temperatures as well as a beta-phase heat treatment during the latter stages of the tube reduction schedule. Chemistry variables included deviations in niobium and tin content from the nominal composition. The effects of both heat treatment and chemistry on corrosion behavior were assessed by autoclave tests in both pure and lithiated water and high-temperature steam. Analytical electron microscopy demonstrated that the best out-reactor corrosion performance is obtained for microstructures containing a fine distribution of beta-niobium and Zr-Nb-Fe particles. Deviations from this microstructure, such as the presence of beta-zirconium phase, tend to degrade corrosion resistance. ZIRLO fuel cladding was irradiated in four commercial reactors. In all cases, the microstructure in the cladding included beta-niobium and Zr-Nb-Fe particles. ZIRLO fuel cladding processed with a late-stage beta heat treatment to further refine the second-phase particle size exhibited in-reactor corrosion behavior that was similar to reference ZIRLO cladding. Variations of the in-reactor corrosion behavior of ZIRLO were correlated tomore » tin content, with higher oxide thickness observed in the ZIRLO cladding containing higher tin. The results of these studies indicate that optimum corrosion performance of ZIRLO is achieved by maintaining a uniform distribution of fine second-phase particles and controlled levels of tin.« less

Metal alloy coating for mitigation of stress corrosion cracking of metal components in high-temperature water

Abstract: A method for mitigating crack initiation and propagation on the surface of metal components in a water-cooled nuclear reactor. A metal coating having an electrically insulating outer layer is applied on the surfaces of IGSCC-susceptible reactor components. The preferred metal coating is a zirconium alloy with a zirconia outer layer. The presence of an electrically insulating layer on the surface of the metal components shifts the corrosion potential in the negative direction without the addition of hydrogen and in the absence of a noble metal catalyst. Corrosion potentials « -0.5 VSHE can be achieved even at high oxydant concentrations and in the absence of hydrogen.

Tube for a nuclear fuel assembly and method for making same

Abstract: A zirconium alloy tube for forming the whole or the outer portion of a nuclear fuel pencil housing or a nuclear fuel assembly guide tube. The zirconium alloy contains 0.8-1.8 wt.% of niobium, 0.2-0.6 wt.% of tin and 0.02-0.4 wt.% of iron, and has a carbon content of 30-180 ppm, a silicon content of 10-120 ppm and an oxygen content of 600-1800 ppm. The tube may be used when recrystallised or stress relieved.

Comparison of the quench experiments CORA-12, CORA-13, CORA-17

Abstract: The CORA quench experiments 12.13 (PWR) and 17 (BWR) are in agreement with the inpile tests LOFT LP-FP-2 and PBF SFD-STand the TMI accident: Flooding of hot Zircaloy clad fuel rods does not result in an immediate cooldown of the bundle, but produces a remarkable temporary temperature increase connected to a strong peak in hydrogen production. For the preparation of new quench bundle tests, necessaryfor the understanding of the mechanisms governing the quench process and support for validation of future quench models in SFD codes the three tests are compared to each other and to the relevant non-quench tests CORA-29 (PWR) and COW-1 6 (BWR). The PWRtests CORA-l2 and CORA-l3 are of the same geometrical arrangement and test conduct. An exception is the shorter time between power shutdown and quench initiation for CORA 13, resulting in a higher temperature of the bundle at start of quenching. The BWR test CORA-17 used B4C absorber and Zircaloy channel box walls, but was in respect to the delay time between power shutdown and start of quenching similar to test CORA-12. All three tests showed during the quench phase the temporary temperature increase correlated to a hydrogen peak. In test CORA-12 with a delay of 300 s between power shutdown and start of quenching, resulting in a cooldown of more than 100 K, a delay of about 50 s was registered between start of quenching and the initiation of the increase of temperature and hydrogen production. The water level at this time had already reached the elevation of about 200 mm. In test CORA-13 with a start of quenching 30 s before power shutdown, temperature and hydrogen production increase started immediately after start of quenching. Immediately after quenching BWR test bundle CORA-17 experiences a modest increase for 20 sand changed then in a steep increase resulting in the highest temperature and hydrogen peaks of the three tests. CORA-l7 also showed a temperature increase in the lower part of the bundle, in contrastto CORA-12 and CORA-13 with temperature increase only in the upper half of the bundle. We interpret this earlier starting and stronger reaction due to the influence of the boron carbide, the absorber material of the BWR test. B4C has a exothermic reaction energy 4 to 5 times larger than Zry and produces about 6 times more hydrogen. Probably the hot remained columns of B4C (seen in the non-quench test CORA-16) react early in the quench process with the FZKA 5679 Q U E N C H increased upcoming steam. The bundle temperature, raised by this reaction increases the reaction rate of the remained metallic Zry (exponential dependence). Due to the larger amount of Zry in the BWR bundle (channel box walls) and the smaller steam input during the heatup phase (2 gls instead of 6 91s) more metalliczry can have survived oxidation during the heatup phase. The different behavior of the three tests provides a good basic for the validation of further quench models.

Fracture toughness of Zircaloy cladding tubes

Abstract: The fracture toughness of Zircaloy-2 cladding has been estimated by means of the recently developed pin-loading (PL) tension test. Axially notched ring specimens, cut directly from different cladding (annealed, cold-worked, hydrided, and irradiated), have been tested in a way similar to that used for compact tension specimens. The results of the PL tension tests, performed at temperatures of 293 and 573 K, revealed for actual cladding all main phenomena observed earlier for Zircaloy materials. A threshold hydrogen content of 600 to 700 wtppm, above which only brittle fracture occurred at ambient temperature, was observed for unirradiated cladding. Existence of a continuous hydride network in the cladding facilitated brittle fracture. There was no obvious influence of the hydrogen content of about 900 wtppm on the fracture toughness at 573 K. The irradiation changed the fracture toughness in accordance with its known influence on the ductility and strength of the cladding. Intensive load serrations occurred for both irradiated and unirradiated cladding during plastic deformation of the notched ring specimens at 573 K. A maximum-load fracture toughness of about J{sub max} {approx} 100 kN/m, obtained for irradiated cladding at 573 K, is comparable with fracture toughness of the cold-worked unirradiated cladding and reasonablymore » agrees with published results for pressure tube materials. The PL tension test was shown to be an effective method for evaluating the fracture toughness of actual cladding after hydriding or irradiation.« less

Modification of oxide film electrical conductivity to maintain low corrosion potential in high-temperature water

Abstract: A method for mitigating general corrosion and crack initiation and growth on the surface of a metal components in a water-cooled nuclear reactor. A compound containing a non-noble metal such as zirconium or titanium is injected into the water of the reactor in the form of a solution or suspension. This compound decomposes under reactor thermal conditions to release ions or atoms of the non-noble metal which incorporate in the surfaces of the components, including the interior surfaces of any cracks formed therein. The preferred compounds are zirconium compounds such as zirconium acetylacetonate, zirconium nitrate and zirconyl nitrate. Zirconium incorporated in the oxided surface of a metal component will reduce the electrochemical corrosion potential at the surface to a level below the critical potential to protect against intergranular stress corrosion cracking without the addition of hydrogen.

Effect of in-PWR irradiation on size, structure, and composition of intermetallic precipitates of Zr alloys

Abstract: The corrosion behavior of Zr alloys depends on the kind, size, and distribution of the intermetallic second-phase particles. TEM examinations of Zr-Sn-Fe-Cr alloys irradiated in PWRs at temperatures between 300 and 370 C and fast fluences in the range of 5E21 to 1.3E22 cm{sup {minus}2} have been performed to study the irradiation-induced effects on the precipitates. The alloys contained different types of second-phase particles such as Zr(Fe,Cr){sub 2}, Zr{sub 2}(Fe,Si), and Zr{sub 3}Fe before irradiation. The influence of irradiation was found to depend on temperature and type of second-phase particles. In Fe-containing Sr alloys with little or no Cr, rather large Zr{sub 3}Fe precipitates are the most frequent particles. These particles are not dissolved by irradiation even at low temperatures. This was confirmed by annealing after irradiation. As a hypothesis, it was assumed that the different behavior of the various precipitates can be related to their melting or decomposition temperatures by using the homologous temperature (i.e., the temperature under consideration in K normalized to the melting or decomposition temperature in K). This interrelationship has been found to apply for irradiation-induced amorphization. The empirical approach to describe the thermal ripening behavior of second-phase particles before irradiation and to describe the transitionmore » from irradiation-induced dissolution to irradiation-induced growth by a normalized (homologous) temperature led to reasonable results.« less

Oxidation Behavior of Iron and Nickel in Zr2(Fe, Ni) Precipitates in Zircaloy-2

Abstract: The oxidation behavior of Zr2 (Fe, Ni) precipitate in Zircaloy-2 was examined by microprobe Auger electron analysis focussing attention on the oxidation behavior of iron and nickel. In the pre-transition period, the iron oxide was formed only on the top surface of zirconium oxide layer of the precipitate and the thickness of iron oxide layer increased with the oxidation time, and nickel was oxidized via dissolution in the matrix zirconium oxide near the top surface. In the post-transition period, the oxidation of iron and nickel proceeded also inside the oxide layer. Such variety of oxidation behavior of iron and nickel between the pre- and post-transition period was attributed to the situation whether the oxygen potential gradient existed or not in the matrix zirconium oxide layer.