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Zirconium alloys for supercritical water reactor applications: Challenges and possibilities
Abstract
The corrosion behavior of model Zr-based alloys at 500 °C is assessed by long term (up to 400 days) corrosion testing in an effort to evaluate their potential for use in the supercritical water reactor and to assess the influence of alloying elements on corrosion behavior. The corrosion weight gains from such systematic testing are seen to be a factor of five higher than those measured at 360 °C but the protectiveness ranking of the alloys is similar. Detailed characterization of the oxide layers to rationalize the differences in corrosion behavior was performed using synchrotron radiation and systematic differences are observed in protective and non-protective oxides, especially near the oxide–metal interface. The overall corrosion rate of the best Zr-based alloys compared favorably with those of other alloys being considered for use in the supercritical water reactor.
... The corrosion behavior of SPPs embedded in oxide films is greatly related to their physical and chemical properties. These differences in SPPs have a profound impact on corrosion behavior [22]. In terms of production process optimization, the microstructure of zirconium alloys can be well-regulated and excellent corrosion resistance can be obtained by applying a proper production process and heat treatment. ...
... Zr(Fe,Cr) 2 with an HCP structure appears in the oxide films of reprocessed plates, and defects appear along with the SPPs (Figure 9f,g). It has been reported that SPPs are embedded in the oxide film at the beginning of oxidation, and then they undergo delayed oxidation [22]. In this study, SPPs were mainly composed of Zr, Fe and Cr, and their P.B. ratios are large. ...
... t-ZrO 2 cannot exist at room temperature or in a PWR environment. Only high compressive stress can stabilize t-ZrO 2 [22], indicating that the oxide film with high t-ZrO 2 content usually has a large amount of compressive stress. Therefore, the delay of stress relaxation limits the initiation and expansion of micro-cracks and pores, ensuring the integrity of the oxide film. ...
To study the effect of reprocessing on the microstructure and corrosion resistance of Zr-Sn-Nb alloy, the original plates of Zr-Sn-Nb alloy were hot-rolled, cold-rolled and recrystallized to obtain the reprocessed plates. The microstructure of both plates was observed with a scanning electron microscope (SEM), a transmission electron microscope (TEM) and electron backscattering diffraction (EBSD). The original plates and reprocessed plates were put into a static autoclave for 300 days in 360 °C/18.6 MPa water. The relationship between the microstructure and corrosion resistance of the Zr-Sn-Nb alloy was discussed. The coarse deformation grains with twins and fine recrystallized grains were obtained, and grain sizes became smaller. The Ostwald ripening of second phase particles (SPPs) happened, and the average size of SPPs increased. Some SPPs changed from an HCP structure to an FCC structure. Reprocessing made the transition advance, which is related to the accelerated evolution of cracks in the oxide film and the increase in metal-oxide film interface roughness. The deterioration of corrosion resistance is closely related to the change of grain size, SPP size and SPP structure.
... 4,78 (8) 4,81 (9) 4,92 (16) 4,81 (17) 4,70 (12) 4,85 (17) 5,02 (25) Рис. 4. Некоторые результаты определения реактивности при сбросе стержней СУЗ в сборку РБМК. Измерено: 71 камера (▬), камера № 2 (▬), камера № 7 (▬); поправлено: камера № 8 (---), камера № 6 (---) ...
... Международные оценки размещения в объекте геологической изоляции ВАО варьируются в диапазоне от 3,4 до 13,6 млн руб./м 3 [23]. Другой пример -российский тариф на размещение отходов класса 4 на 2018 г. составляет 0,04 млн руб./м 3 , тогда как соответствующее значение для объекта компании-оператора LLW Repository (Великобритания) находится на уровне ~0,3 млн руб./м 3 [24], для объектов штата Техас (США) ~0,2 млн руб./м 3 [25], что составляет разницу в порядок. ...
... Изменение кинетики прироста массы оксидного слоя для циркония и его сплавов (прирост массы обусловлен взаимодействием с кислородом и следует общей кинетике коррозии) оказалось двухэтапным (рис. 4) [24,25]. ...
... In principle, Zr clad is subjected to corrosion from the coolant (water) and fuel sides, both by way of oxygen and hydrogen penetration. The oxidation and hydrogenation of zirconium fuel components in LWR may affect reactor safety and efficiency, which makes corrosion a critical design aspect of Zr materials response in nuclear environments [12][13][14][15][16][17][18]. ...
... Table 3, each element is 100-nm thick. Subsequent growth of these embryos occurs at a rate given by the combination of the rates of H-atom absorption (Equation (16)) and dissolution (Equation (17)), i.e., (k n − s n ), as shown in Figure 8. Rapid net growth is seen in the initial stages of hydridization close to the oxide/metal interface. However, these rates gradually abate both in time and with increasing depth until almost no net growth is observed, particularly at depths greater than 700 nm after 1.4 h of evolution. ...
... For example, hydride nucleation is simulated by considering the interplay between (i) aggregation, (ii) growth, and (iii) dissolution processes, which together determine the net nucleation and growth rates. Processes (i), (ii), and (iii) are embodied in Equations (14), (16) and (17), respectively. Each one of these processes is treated as a stochastic event sampled with the probabilities given by each respective rate. ...
The formation of elongated zirconium hydride platelets during corrosion of nuclear fuel clad is linked to its premature failure due to embrittlement and delayed hydride cracking. Despite their importance, however, most existing models of hydride nucleation and growth in Zr alloys are phenomenological and lack sufficient physical detail to become predictive under the variety of conditions found in nuclear reactors during operation. Moreover, most models ignore the dynamic nature of clad oxidation, which requires that hydrogen transport and precipitation be considered in a scenario where the oxide layer is continuously growing at the expense of the metal substrate. In this paper, we perform simulations of hydride formation in Zr clads with a moving oxide/metal boundary using a stochastic kinetic diffusion/reaction model parameterized with state-of-the-art defect and solute energetics. Our model uses the solutions of the hydrogen diffusion problem across an increasingly-coarse oxide layer to define boundary conditions for the kinetic simulations of hydrogen penetration, precipitation, and dissolution in the metal clad. Our method captures the spatial dependence of the problem by discretizing all spatial derivatives using a stochastic finite difference scheme. Our results include hydride number densities and size distributions along the radial coordinate of the clad for the first 1.6 h of evolution, providing a quantitative picture of hydride incipient nucleation and growth under clad service conditions.
... Concerning chemical reactivity, they can easily form alloys and metallic glasses. Due to all these properties, the technological applications and the economic importance of these elements, and their alloys, is immense [1][2][3][4]. As such, transition metals are among the most intensively studied family of elements, and in the past decades a big effort has been devoted to map and interpret systematic trends in their properties. ...
... We also note a good agreement between the -stability of the phase observed in this study and the -ℎ phase boundary reported in previous studies [30]. However, because of the minimal temperatures that can be generated and measured in LH-DAC experiments it was only 2 The observed -ℎ coexistence is a consequence of the alignment of the two lasers on the opposite surfaces of the samples and the successive rapid quench performed before the beginning of the actual heating ramp. ...
... Nuclear grade Zr-alloys like Zircaloy-4 alloy are subjected to neutron irradiation as well as severe conditions of high temperature and high-pressure flowing water [46]. The harsh, aggressive environment exacerbates the corrosion and hydrogen absorption, thus causing fast deterioration of the mechanical strength of structural components [47,48]. ...
... Zr alloys demonstrated low neutron absorption coefficients and high yield strength at higher temperatures, which led to their selection for nuclear applications, such as fuel tubes and coolant channel materials in water-cooled power reactors [10,46,61,62]. Furthermore, Zr alloys suffer from low wear resistance in comparison with other nuclear materials, such as stainless steels and nickel alloys [63], while the fretting wear on tubes is an important concern in pressurized water reactors [63,64]. ...
A plasma electrolytic oxidation (PEO) is an electrochemical and eco-friendly process where the surface features of the metal substrate are changed remarkably by electrochemical reactions accompanied by plasma micro-discharges. A stiff, adhesive, and conformal oxide layer on the Zr and Zr-alloy substrates can be formed by applying the PEO process. The review describes recent progress on various applications and functionality of PEO coatings in light of increasing industrial, medical, and optoelectronic demands for the production of advanced coatings. Besides, it explains how the PEO coating can address concerns about employing protective and long-lasting coatings with a remarkable biocompatibility and a broad excitation and absorption range of photoluminescence. A general overview of the process parameters of coatings is provided, accompanied by some information related to the biological conditions, under which, coatings are expected to function. The focus is to explain how the biocompatibility of coatings can be improved by tailoring the coating process. After that, corrosion and wear performance of PEO coatings are described in light of recognizing parameters that lead to the formation of coatings with outstanding performance in extreme loading conditions and corrosive environments. Finally, a future outlook and suggested research areas are outlined. The emerging applications derived from paramount features of the coating are considered in light of practical properties of coatings in areas including biocompatibility and bioactivity, corrosion and wear protection, and photoluminescence of coatings
... Transition metals are defined as those elements that have a partially filled d-electron sub-shell. The strong metallic bonding due to the delocalization of d-orbitals is responsible for a series of very interesting properties, such as high yield strength, corrosion and wear resistance, good ductility, easy alloy and metallic glass formation, paramagnetism, and high melting points and molar enthalpies of fusion, leading to a plethora of industrial applications [1][2][3][4][5][6]. ...
... As a monotonically increasing function, the Simon-Glatzel equation can mainly describe melting curves that rise with pressure. The tangent melting slope is defined by Equation (1) or by the derivative of Equation (2). For most metals, the density of the liquid is lower than that of the solid, while the entropy is higher, leading to mostly positive melting slopes (i.e., the melting temperature increases with pressure) [16,22,24,29,[35][36][37]39,40,63]. ...
The accurate determination of melting curves for transition metals is an intense topic within high pressure research, both because of the technical challenges included as well as the controversial data obtained from various experiments. This review presents the main static techniques that are used for melting studies, with a strong focus on the diamond anvil cell; it also explores the state of the art of melting detection methods and analyzes the major reasons for discrepancies in the determination of the melting curves of transition metals. The physics of the melting transition is also discussed.
... Fig. 1 Modeled geometry of the Canadian SCWR fuel channel [2] Zirconium alloys typically experience high corrosion rates at temperatures exceeding 400 °C in supercritical water (SCW),[AQ3] and they are known to experience delayed hydride cracking as a result of hydrogen pickup due to corrosion. Early studies on zirconium alloy corrosion in SCW showed some favorable results at 500 °C for Zr-Fe-Cr alloys and chromium-coated Zr-Nb alloys [3][4][5][6][7][8][9][10][11][12][13][14][15]; however, the rate of hydrogen uptake would seem to preclude a 5year fuel cycle. A more thorough study is needed to assess the suitability of zirconium alloys for the small SCWR, including a detailed examination of coating options. ...
... Among all the uncoated materials, the bare Zr-4 had the highest weight gain, while the Ti-grade 5 alloy had the lowest weight gain. The poor performance of Zr-4 was expected based on the results of previous studies, as it showed high oxidation in 500 °C SCW [5,6]. Table 3 shows the theoretical thickness of oxide calculated from weight gain using the fraction of oxygen in the oxide when assuming only ZrO 2 was formed on the Zr-based and only TiO 2 was formed on the Tibased alloys. ...
The 300 MWe small Canadian SCWR, which is a scaled-down version of the original 1200 MWe concept, has a smaller core, uses low enriched uranium fuel instead of a plutonium-thorium fuel, and features a lower (maximum) cladding temperature of 500 °C. The lower cladding temperature may permit the use of different alloys, including zirconium alloys, which had been ruled out as candidates for the Canadian SCWR, whose cladding temperature may reach 850 °C. The potential to use zirconium alloys is exciting because they have a low neutron cross section, which in turn means that fewer neutrons are lost, and the fuel can be used more efficiently. One advantage, for example, is that the fuel cycle can be lengthened. In this paper, we report on the results of corrosion experiments used to screen zirconium- and titanium-based alloys as well as corrosion resistant coating materials such as Cr and Al as potential candidates for fuel cladding in the small Canadian SCWR. These experiments were conducted in a refreshed autoclave in deaerated supercritical water at 500 °C and 23.5 MPa. After exposure, the weight gain was measured, and the oxide thickness and the oxide phases were examined. Of all materials, the coated and uncoated Ti-Grade 2 and Ti-Grade 5 alloys met our screening qualification criteria, however, Al/Cr-coated zirconium coupons showed notable improvement and will be explored further in future testing.
... Transition metals exhibit a series of interesting and useful properties originating from their partially filled d-band and the consequent delocalization of their d-orbitals. Consequently, they attract significant attention from a range of scientific communities [1][2][3][4]. As a result, an extensive amount of research efforts have been dedicated to this group of metals in recent decades. ...
... These pores would act as short circuit paths for the diffusion of oxygen, water molecule, OH-from SCW, and metallic atoms from the matrix [29]. Therefore, Fe elements would diffuse outward from the matrix to the interface of SCW and the outer oxide layer to form Fe-rich oxides [44]. Subsequently, Fe-rich oxides would react with Cr 2 O 3 to form FeCr 2 O 4 on the surface of the samples, as shown in Fig. 9. ...
The effect of alloying elements, Mn, Ti, and Mo, on the corrosion behavior of FeCoNiCr-based high entropy alloy (HEA) in supercritical water was investigated. The corrosion resistance of the HEA was enhanced by adding Mo while reduced with the addition of Ti and Mn. The segregation of Ti and Mo resulted in the formation of oxide films that were both structurally and chemically heterogeneous. Grain boundaries accelerated the diffusion rate of Mn, leading to the formation of MnCr2O4 oxide scales decorated with Mn2O3 particles. In addition, the governing mechanisms of alloying elements on the corrosion behavior of HEAs were also discussed.
... This work shows a path to connect mechanical property modeling with experimental observations to understand the material response of heterogeneous materials with an example showing the modeling of Zircaloy material with the heterogeneity of hydride phases. Zircaloy is widely used in the nuclear industry due to its low neutron absorption cross section, excellent corrosion resistance, and outstanding thermo-mechanical properties [24]. Zircaloy is used as a getter for the hydrogen isotope tritium (symbol 3 H) during its production in a nuclear reactor using a system called tritium producing burnable absorber rod (TPBAR). ...
This work developed a microstructure-based finite element model to predict the stress state of alloys with second phase inclusions. Quantitative microstructural details were extracted from scanning electron microscopy (SEM) images and were used to generate heterogeneous microstructures. Generation of digital microstructure was achieved through two steps of tessellations using software Neper and Matlab. This digital microstructure can be further used to perform multiple structural simulations as per the need. Such simulations are very useful in many applications as the manufacturing process can be altered to obtain the morphology resulting in desired overall material properties. The process is demonstrated using an example of Zircaloy material with secondary phases of hydrides dispersed within and stress-strain response of Zircaloy containing hydrides was predicted. Crystal plasticity based constitutive material model was used for Zircaloy whereas the hydrides were modeled as isotropic elasto-plastic material. A parametric study was conducted to understand the effect of volume fraction, and lamellae thickness of the hydride phase on the mechanical properties of the overall material. The introduction of the hydride phase was found to increase the yield stress of Zircaloy with an almost linear relationship between yield stress and hydride volume fraction. On the other hand, the width of hydride lamina was found to have a negligible effect on the yield stress of Zircaloy. These results can help designers to alter the manufacturing process to obtain the enhanced mechanical properties for components used in nuclear applications made by Zircaloy material.
... Zirconium alloys have been used as a main structural material of nuclear fuel claddings and spacer grits in light water reactors (LWR) [1] . Different alloying elements such as Sn, Nb, Fe, O and Cr are added to improve mechanical and functional properties of Zr alloys (E110, E110opt, E635, M5 Framatome , Zry-4, ZIRLO TM and others [2][3][4] ) that makes possible to use of Zr fuel assembly during longterm operation cycles in LWRs under normal operating conditions ( ∼360 °C, 18.6 MPa). ...
The influence of chromium coating on hydrogenation behavior of laser beam welds (LBW) of E110 zirconium alloy was studied at 360-900°C and hydrogen pressure of 2 bar. The deposition of Cr coating can slow down hydrogen absorption rate of the LBW samples as activation energy for hydrogen absorption (70 kJ/mol) is higher compared to uncoated ones (61 kJ/mol). The growth of hydrides depends on the hydrogenation conditions and original microstructure of welded alloy in different zones. Nanoindentation and three-point bending tests show that hydride precipitations can cause hardening of LBW E110 alloy samples and loss of its plasticity. Some additional aspects of applying Cr coatings for protection of LBW Zr alloy in hydrogen atmosphere are presented.
... Zirconium-based alloys, such as Zircaloy and Zr-1Nb, are widely used in nuclear reactor fuel cladding and structural components (grid spacers, guide tubes, nozzles/debris screens, and fuel channels) due to their low neutron absorption cross section, excellent corrosion resistance, and outstanding thermo-mechanical properties [1][2][3]. Zircaloy is a Zr-based alloy with about 1.5% Sn and smaller proportions of other elements. Zircaloy-4 tends to be used in pressurized water reactors (PWRs), while Zircaloy-2 tends to be used in boiling water reactors (BWRs). ...
In nuclear reactors, hydrides can form in fuel cladding due to hydrogen absorption in Zircaloy and cause embrittlement. This work presents a microstructure-based finite element model to predict the stress–strain response of Zircaloy containing hydrides. Microstructural details extracted from scanning electron microscopy (SEM) images were used to generate heterogeneous microstructures including the morphology and spatial distribution of hydrides. The constitutive material model for Zircaloy in this study is based on crystal plasticity theory which considers the hexagonal close-packed (HCP) atomic structure of Zircaloy material. The hydrides were modeled as brittle material along with a damage model. Hydride formation inside the Zircaloy matrix results in residual stress. This phenomenon is also captured in this model. A parametric study has been conducted to understand the effect of volume fraction, orientation, and lamellae width of the hydride phase on the mechanical properties of the overall material and the findings were validated against experimental results from the literature.
... This is because interfaces, between the closed packed planes of the matrix and the hydrides, have misorientation dislocation and resulting misfit strains. Hence, orientations relations (ORs) have to be physically represented to account for both coherent and semicoherent interfaces [18,19]. ...
A crystalline dislocation-density formulation that was incorporated with a non-linear finite-element (FE) method was utilized to understand and to predict the thermo-mechanical behavior of an hexagonal closest packed (h.c.p.) zircaloy system with hydrides with either face centered cubic (f.c.c.) or body centered cubic (b.c.c.) hydrides. This formulation was then used with a recently developed fracture methodology that is adapted for finite inelastic strains and multiphase crystalline systems to understand how different microstructurally-based fracture modes nucleate and propagate. The interrelated microstructural characteristics of the different crystalline hydride and matrix phases with the necessary orientation relationships (ORs) have been represented, such that a detailed physical understanding of fracture nucleation and propagation can be predicted for the simultaneous thermo-mechanical failure modes of hydride populations and the matrix. The effects of volume fraction, morphology, crystalline structure, and orientation and distribution of the hydrides on simultaneous and multiple fracture modes were investigated for radial, circumferential, and mixed distributions. Another key aspect was accounting for temperatures changes due to the effects of thermal conduction and dissipated plastic work and their collective effects on fracture. For hydrided aggregates subjected to high temperatures, thermal softening resulted in higher ductility due to increased dislocation-density activity, which led to higher shear strain accumulation and inhibited crack nucleation and growth. The predictions provide validated insights of why circumferential hydrides are more fracture resistant than radial hydrides for different volume fractions and thermo-mechanical loading conditions.
... For example, scanning electron microscopy, Xray analysis provides information about the surface of a very small area (500 µm) [13][14], while corrosion occurs at a much larger scale. In overcoming the scale factor can play a major role electrochemical methods developing in the direction of the electrochemical analysis of solid-phase materials [15][16][17][18]. Earlier it was found [19], that diagnosis of the protective ability of passivating films should be carried out in systematic combination with the results of the local electrochemical analysis allowing to estimate the composition of the surface layers of passivating films voltammetric evaluation of the fraction of free area, impedance measurements of electrical conductivity. ...
AННОТАЦИЯ Данная статья о системе электрохимического анализа свойств пассивирующих пленок на поверхности нагрева теплоэнергетического оборудования. Было обнаружено, что прогнозирование скорости коррозии низкоуглеродистой стали в теплоэнергетических системах возможно на основе уравнений множественной регрессии по количеству фаз оксида кремния и гидроксида железа в пленках, доле свободной поверхности и активной составляющей импеданса пленок в щелочном электролите и ртути. Построение уравнений регрессии необходимо проводить с предварительной классификацией данных по величинам активной составляющей импеданса пленок в ртути и щелочном электролите. ABSTRACT This article is about electrochemical analysis system of passive films corrosion properties of the thermal power equipment heating surfaces. It was found that forecasting the low carbon steel corrosion rate in thermal power systems is possible based on the multiple regression equations, which includes the amount of silicon oxide and iron hydroxide phases in the films, the fraction of free area and the active component of impedance of the films in the alkaline electrolyte and mercury. Construction of the regression equation should be carried out with the preliminary classification of data on the quantities of the active component of the film impedance of mercury and an alkaline electrolyte. Ключевые слова: скорость коррозии, пассивирующие пленки, теплоэнергетические системы, хронопотенциометрия, углеродистая сталь, доля свободной поверхности, импеданс.
... Because it is generally accepted that chromia-containing spinels are better permeation barriers to cations (metal) and anions (oxygen, OH, etc.) relative to iron oxides, 9 it is the inner oxide layer that can provide the best corrosion resistance in the SCW environment. This has been observed in recent work on steels under nuclear SCW conditions [10][11][12] in terms of increasing corrosion resistance with an increasing chromium content of the alloy. Fig. 2 shows that nickel-based alloys have the greatest resistance to oxidation in SCW, followed by austenitic stainless steels and then ferritic-martensitic alloys. ...
Commercial nuclear power plants in the United States are light water reactors (LWRs) that use water as a coolant, with temperatures ranging between 280°C and 320°C. Water purity is tightly controlled; nevertheless, the high temperatures expose components in the water circuit to degradation by corrosion by many different types of environmental attack. Uniform corrosion occurs across the entire surface of a material and is prevalent in engineering systems to some extent. Site-specific corrosion processes, such as crevice corrosion, intergranular attack, or galvanic corrosion, are common in complex engineering systems that consist of multiple materials joined by welds or other solid state joining processes. Even within a single component, if second phase strengthening is used or if surface defects are present, localized corrosion can occur.
... Due to the growing demand for implant materials with enhanced corrosion resistance, mechanical properties, and bioactivity, the number of research reports focused on the development of novel biomaterials for orthopedic and dental applications has soared in the last couple of decades [1][2][3][4][5][6][7]. Zirconium based materials are of great interest for biomedical applications owing to their high chemical stability, low thermal conductivity, high mechanical strength and fracture toughness, low elastic modulus as well as good biocompatibility [1,2,[4][5][6][7][8][9][10]. Although the formation of a very thin (2-5 nm), native oxide layer (ZrO 2 ) on the surface of Zr substrates renders these materials corrosion resistant, it deteriorates the bioactivity of Zr due to the bio-inert nature of ZrO 2 by hindering the formation of chemical bonds between bone tissues and Zr implant [11][12][13]. ...
Surface properties can affect cell attachment and growth ability on the biomaterials. In this work, the effect of the morphology and chemical composition of the surface coatings synthesized on zirconium (Zr) substrates on bioactivity was studied. The surface of pure Zr was modified via a plasma electrolytic oxidation process to obtain oxide-based coatings with different surface chemical compositions and pore sizes by changing the process parameters in the same electrolyte. The characterization of the coatings was conducted using X-ray diffraction (XRD), scanning electron microscopy (SEM), contact angle goniometry (CAG) and energy dispersive spectroscopy (EDS), and Fourier transform infrared spectroscopy (FTIR). The results showed that the coatings comprised t-ZrO2 and Ca0.134Zr0.86O1.86 phases. During the oxidation of Zr substrates, applying an anode current with different densities and frequencies altered the surface roughness Ra of the samples. The relationship between the surface morphology, chemical composition, and bioproperties was investigated by simulated body fluid (SBF) and cell culture tests. The coated samples were immersed into SBF solution and the formation of hydroxyapatite (HA) structure on the surface was detected after 7 days of immersion. A close relationship between the development of HA, the size of the porosity and chemical composition of the coating surface was observed. Additionally, the cell attachment and cell growth on the surface were found to be dependent on the chemistry and topography of coated surfaces. The cell culture test revealed that the growth of the cells was improved with increasing pore sizes on the coating surface.
... The presence of these phases can significantly improve the tensile strength and creep resistance of the alloy. Zr-based alloys have high corrosion resistance in high-temperature aqueous solutions, good mechanical properties, and resistance to radiation damage, which are widely used in aerospace [4], military [5], nuclear reaction [6], atomic energy [7], chemical engineering [8], marine engineering, medicine [8], biology and so on [9]. At present, many research scientists are shifting their focus to the mechanical properties of Zrbased alloys to expand further the application potential of Zr in the structural field [10][11][12][13]. ...
In this paper, the microstructure and wear resistance of Zr-17Nb alloy treated by high current pulsed electron beam were studied in detail. A phase change occurs after pulse treatments using X-Ray Diffraction (XRD) analysis, showing β (Nb) phase and α (Zr) phase transformed by a part of β (Zr, Nb) phase. Also, narrowing and shifting of β (Zr, Nb) diffraction peaks were found. Scanning Electron Microscope (SEM) and metallographic analysis results reveal that the microstructure of alloy surface before high current pulsed electron beam (HCPEB) treatment is composed of equiaxed crystals. But, after 15 and 30 pulse treatments, crater structures are significantly reduced. Besides, it was also found that the alloy surface has undergone eutectoid transformation after 30 pulse treatments, and the reaction of β (Zr, Nb) → αZr + βNb had occurred. Microhardness test results show that microhardness value presents a downward trend as the number of pulses increases, which is mainly due to the coarsening of the grains and the formation of a softer β (Nb) phase after phase transformation. The wear resistance test results show that the friction coefficient increases first, then decreases and then increases with the increase of pulse number.
... Zirconium alloys are structural materials for core elements of nuclear power reactors due to their low neutron capture cross-section, high melting temperature, high and stable corrosion resistance in water, steam and other aggressive media, good plasticity, and high strength characteristics [1][2][3]. However, the strength characteristics of these alloys are insufficient for operation upon exposure to radiation. ...
In this paper, comparison studies of the hydrogen effect on the structural and phase state, deformation behavior, and mechanical properties of the fine- (average grain size 4 µm) and ultrafine-grained (average element size 0.3 and 0.4 µm) Zr–1wt.%Nb (hereinafter Zr–1Nb) alloy under tension at temperatures in the range of 293–873 K were conducted. The formation of an ultrafine-grained structure is established to increase the strength characteristics of the Zr–1Nb alloy by a factor of 1.5–2 with a simultaneous reduction of its resistance to the localization of plastic deformation at the macro level and the value of deformation to failure. The presence of hydrogen in the Zr–1Nb alloy in the form of a solid solution and hydride precipitates increases its resistance to the localization of plastic deformation at the macro level if the alloy has an ultrafine-grained structure and decreases if the structure of the alloy is fine-grained. In the studied temperature range, the Zr–1Nb alloy in the ultrafine-grained state has a higher resistance to hydrogen embrittlement than the alloy in the fine-grained state.
... 32 Some work has been devoted to using ambient pressure (0.1 MPa to 10 MPa) steam as a surrogate for SCW to extend the testing conditions to higher temperatures and/or longer times. [33][34][35][36][37][38] The results of these studies have been mixed and it seems that the level of success in using ambient pressure steam as a reasonable surrogate for SCW depends on the alloy examined. ...
The possibility of surrogating high-pressure supercritical water (SCW) with an ambient pressure air-10% H2O (wet air) mixture for studying wet oxidation of Type 310S stainless steel (UNS S31008) at high temperatures was examined using gravimetric measurements coupled with electron microscopy techniques. A distinctly different wet oxidation kinetic rate law was observed in the two environments at 550 ºC: parabolic for the wet air mixture and para-linear (âbreakawayâ) for SCW. The differing rate law is attributed to differences in the mass transport kinetics parameters, which are not well defined for SCW, that likely control Cr volatilization from the oxide scale, starting with the presumed MnCr2O4 spinel outer layer. Although the wet air mixture is not considered to be a suitable surrogate environment for SCW at the temperature of interest, it may prove useful to help determine microstructure instability effects on wet oxidation at elevated temperatures.
... The alloying elements depletion region corrodes much faster than its neighboring normal region (as shown in Fig. 6d). Although the center part with agglomerations of SPPs may corrode slowly due to enough amount of SPPs [9,22], the center part will be wrapped quickly due to the accelerated corrosion of its surrounding metal (as shown in Fig. 6e). With the fast growth of the depletion region and the sustaining self-regulating of the oxide, the lens-like nodule is formed finally (as shown in Fig. 6f). ...
Zr–0.85Sn–0.16Nb–0.37Fe–0.18Cr alloy samples prepared by low-temperature process (LTP) and high-temperature process (HTP) show different corrosion behaviors in 500 °C steam. The former exhibits uniform corrosion, while the latter exhibits nodular corrosion casually. The occurrence of nodular corrosion on HTP samples is attributed to the formation of local alloying elements depletion region in the Zr matrix. During hot rolling at 800 °C, transformation of zirconium from α-phase to β-phase happens locally. Alloying elements migrate to β-phase from its neighboring α-phase. The local β-phase dissolves in the subsequent cold rolling and final annealing, leaving the microstructure with agglomeration of precipitate particles and neighboring alloying elements depletion regions in Zr matrix. And the undesirable microstructure becomes the causation of nodular corrosion.
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Supercritical water-cooled reactor (SCWR) is an innovative Generation IV reactor and merits further research and development with the intent of being pursued for implementation in the next 30 years. The SCWR is a high temperature and high pressure water-cooled reactor in which supercritical water (SCW) is used as the primary coolant for a direct once-through cycle to achieve noticeable advantages. For optimum thermal efficiency, the Canadian SCWR concept requires a fuel core outlet temperature of 625 °C at 25 MPa with a predicted peak temperature as high as 800 °C. As a result, material selection for the fuel cladding is one of the most challenging aspects for the realization of the SCWR. Detailed materials assessments based on public data have carried out by Canadian research groups and several nickel-based alloys, such as UNS N06625, UNS R20033 and UNS N07214 alloys, are selected for the fuel cladding. However, significant knowledge gaps exist in determining whether the alloys can be used for the fuel cladding in the SCWR. This paper introduces our most recent laboratory results on corrosion and stress corrosion cracking (SCC) of the nickel-based alloys in SCW.
The oxide layers formed on 9CrODS have been characterized using synchrotron radiation fluorescence and diffraction. This analysis showed a three-layer structure with an outer layer containing only Fe3O4, an inner layer containing a mixture of FeCr2O4 and Fe3O4, and a diffusion layer containing a mixture of metal grains and FeCr2O4 precipitates. A Cr2O3 ribbon formed at the diffusion layer-metal interface on the samples exposed to 600°C supercritical water for 4 and 6 weeks. Calculations of the oxidation behavior were undertaken to calculate the activation energy and the corrosion rate constant n of power law kinetics. These calculations showed that the oxidation behavior of this alloy could not be described by a power law because the oxide microstructure changed with exposure time and temperature. Additionally, the outward flow of iron was calculated and showed that not enough iron migrates outwards to be able to form the outer layer suggesting that other mechanisms might be at work. Finally, a qualitative description of the oxidation behavior of 9CrODS is displayed showing the importance of the role played by Cr2O3 in the corrosion process.
Exploring the effect of Fe addition and ion irradiation on surface hardness in Zr alloys not only elucidates the strengthening mechanisms of individual alloying elements but also facilitates the assessment of mechanical properties under varying damage levels. In this work, the microstructural variations and surface hardness of Zr and Zr-Fe alloys were examined through Ne+ and Au3+ irradiation experiments. Electron backscatter diffraction (EBSD) characterization suggested the potential of Fe addition in Zr-Fe alloys for grain refinement. Transmission Electron Microscopy (TEM) observations indicated the formation of dislocation loops in irradiated materials, accompanied by a transformation of Zr3Fe particles from a crystalline to an amorphous state. Furthermore, nano-indentation tests were employed to measure the depth-dependent hardness, revealing an augmentation in hardness with increasing Fe content and highlighting noticeable irradiation hardening in Zr alloys. A mechanical model was then developed to theoretically investigate the contribution of diverse hardening mechanisms in ion-irradiated Zr alloys, particularly addressing the impact of non-uniformly distributed defects. Theoretical analysis denoted that the irradiation-induced defects hardening at varying depths follows distinct laws, whereas the precipitation hardening in Zr-Fe alloys was attributed to high-density strengthening particles induced by Fe addition. Moreover, it was determined that the surface hardness is governed by geometrically necessary dislocations (GNDs) and irradiation-induced defects at shallow depths, whereas precipitates and statistically stored dislocations (SSDs) dominate at relatively larger depths.
Description
The latest in this popular series provides 44 peer-reviewed papers on the latest international research covering all aspects of zirconium alloy properties and performance relevant to the nuclear industry.
Topics address:
• All aspects of the fuel cycle
• Basic metallurgy
• Spent fuel
• Corrosion
• Mechanical properties
• Deformations mechanisms
• Failure mechanisms
• LOCA and Transients.
These papers provide examples of studies where experimental techniques are combined with analytical tools and calculations; reflect a stronger trend towards efforts that enable fundamental in-reactor measurements, which are essential for further development of experimental techniques as well as for the possibility to enhance the development of improved alloys; and examine an increasing number of alternative alloys being developed and verified.
One of the challenges of small modular reactors (SMRs) in comparison with large reactors is the greater difficulty in achieving high burnups in smaller cores. With greater neutron leakage through the periphery, a key factor is the neutron economy of the fuel cladding. However, all large supercritical water-cooled reactor (SCWR) concepts have employed neutron-absorbing stainless steels and nickel-based alloys in order to meet all the requirements in terms of corrosion and thermalhydraulics. In order to achieve higher burnups and extend the time between refueling in a SCW-SMR, the use of chromium-coated zirconium alloy as a potential fuel cladding candidate has been explored. Chromium coatings up to a few micrometers thick have shown improved oxidation resistance of zirconium-based claddings under operating conditions relevant to SCWR concepts. In this study, Zr-2.5Nb alloy (UNS R60904) from pressure tube samples was coated using a physical vapor-deposition (PVD) method. Oxidation tests were performed on coated samples at 500 °C and approximately 25 MPa in a refreshed autoclave. The effects of the oxide on heat transfer and hydraulic resistance are also discussed in this study. Last, but not least, this study evaluates the coating cost of the fuel cladding with chromium in a vacuum plasma spray process.
The corrosion characteristics of M5 alloy in lithiated/borated water and pure supercritical water were investigated. Oxide films that formed on the alloy were primarily made up of m-ZrO2. Promoting inward diffusion of O²⁻ into the matrix through defects, Li⁺ and OH⁻ shared a synergistic effect on the occurrence and acceleration of waterside corrosion. B-(OH) and zirconium borate produced by the reaction between boric acid and OH⁻ are conducive to alleviating LiOH ingress. In supercritical water, M5 alloy experienced weight loss. The microcrack density and porosity were significantly larger than those in the Li/B environment, and long transverse cracks parallel to the O/M (oxide/metal) interface existed. NiO, Ni2FeBO5 and NiMoO4 simultaneously deposited in lithiated/borated water, while large-sized spinel oxides were the main Ni-based deposits in supercritical water.
A new alumina-forming austenitic oxide dispersion strengthened steel was prepared, and the general corrosion behavior was studied. The results showed that the Y2O3 addition reduced the grain size. With small grains, the grain boundary density increased, which accelerated the formation of stable oxide scale in the corrosion process and obviously decreased the weight gain of AFA-ODS steel. A four-layer oxide scale was formed on the both sides of grain boundary in the inner layer, and a Cr-depletion and metallic Ni layer were formed at the oxide/matrix interface. The upper part of inner layer had many cavities. Al2O3 protective layer appeared in the inner oxide layer and the corrosion degree of the underlying material was reduced. Grain boundary migration was also induced by the element diffusion. The structure surrounding Y2O3 particles in the inner layer was affected slightly by Y2O3 particles. Y2O3 particles at the oxide/matrix interface could promote the corrosion and caused the sunken corrosion pits because the element could diffuse easily through the interface between Y2O3 particles and the material. However, with the addition of Al, Al could diffuse into the sunken pits and wrapped Y2O3 particles, which inhibited the further development of the sunken pits.
We have developed a new real-time framework for calculating the simultaneous accumulation of oxidation-induced and the internal/external fluid stresses during the corrosion of the zirconium metal, Zr. In order to track such interfacial stress when the zirconium metal turns oxide, we quantify the hypothetical real-time infiltration of the oxygen within the metal matrix in the curved boundary, leading to the augmentation in the volume, and we stoichiometrically compute the resulted equivalent oxide thickness. Subsequently, we calculate the accumulated compressive stress in real-time from both irreversible (plastic) and reversible (elastic) events which could be used for anticipation of the onset of failure. The developed analytical framework could quantify the design parameters for the safe operation of high-pressure pipes in corrosive environments.
Clay is a natural substance widely existing in the environment. Nanoclays have small particle size, large surface area, and high porosity. Due to their special characteristics, nanoclays can be used in many different industrial applications. There is also an emerging trend for the use of nanoclays in environmental applications. Nanoclays can be used as adsorbents for the removal of various pollutants from water and gas. The suitability of nanoclays for certain type of application will depend on the requirement for pollution control, as well as the specifications of nanoclays. This article provided a comprehensive review of the specific characteristics of different types of nanoclays. The industrial applications of nanoclays were summarized. The environmental applications of nanoclays for water and gas emission treatment, as well as their toxicity, were discussed. The challenges and recommendations for future study were also proposed.
The paper presents the results of numerical simulation of the distribution of thermal fields during the formation of Cr-Zr surface alloy using a pulsed low-energy high-current electron beam (LEHCEB). The melting thresholds of the Cr-Zr system for different thicknesses of Cr films were calculated. The melting threshold of the Cr-Zr system increases linearly with increasing Cr film thickness. A linear regression dependency model of the melting threshold on the film thickness is proposed. Evaporation thresholds of the Cr-Zr system for different thicknesses of Cr films were calculated. The evaporation threshold of the Cr-Zr system increases linearly with increasing Cr film thickness. A linear regression dependency model of the evaporation threshold on the film thickness is proposed. The value of the LEHCEB energy density at which the lifetime of the film and substrate are equal is calculated. This value is a maximum value for the effective formation of Cr-Zr. A model of the LEHCEB energy density, at which the lifetime of the film and the substrate are equal, in the form of a third-degree polynomial is proposed.
As a novel method to fabricate ZrC-reinforced Zr-based composites, pre-melted electron beam freeform fabrication was proposed to elevate the hardness and wear resistance of Zr alloys. After careful choice for material and precise shape design, a graphite diversion nozzle, which was used as a preliminary vessel for diffusion of C from graphite into liquid Zr under the oscillating effect of bombardment and stir by electron beam, played a predominately critical role in ZrC formation and the consequent ZrC/Zr composite fabrication, indicating a pre-melted process during electron beam freeform fabrication. The crystal structure of ZrC was characterized by ordered arrangement along <1 1 1>, and the interface between ZrC and Zr matrix was negative lattice misfits. The strain at the interface led to a deflection angle of about 4 ° between the (2 0 0)ZrC and (1 0 1)Zr. The lattice distortion at the interface and dispersion strengthening effect of ZrC were the main reasons for the increase in hardness and wear resistance of Zr matrix composites.
The objectives of this study were to investigate the incipient oxidation behavior of ZrO₂-coated Zircaloy-4 (Zry-4), and evaluate the feasibility of producing ZrO₂ protective layer by vacuum annealing of ZrN coating. ZrO₂ coatings with tetragonal (t) and monoclinic (m) dominant phases were deposited on Zry-4 using unbalanced magnetron sputtering. The wettability tests and the widths of interdiffusion zones at film/substrate interface indicated the ZrO₂ coatings having good intrinsic adhesion on the Zry-4 substrate. The oxidation behavior of the coated specimens and blank Zry-4 substrate were evaluated by thermogravimetric analysis performed at 700, 800 and 900°C in argon atmosphere. After oxidation, X-ray diffraction (XRD) patterns revealed that both thermally formed ZrO₂ and the deposited ZrO2 possessed an m-ZrO₂ dominant structure. The weight gain at 700°C indicated no significant difference in the oxidation behavior between uncoated and ZrO₂-coated Zry-4, and the oxidation kinetics obeyed the parabolic law. However, the oxidation kinetics above 800°C deviated from the parabolic law, where m-ZrO₂-coated Zry-4 showed the least weight gain at 800°C, and t-ZrO₂-coated Zry-4 displayed a better oxidation resistance at 900°C. Oxide nodules were observed on the uncoated Zry-4 specimens after oxidation; in contrast, the ZrO₂ coatings, both monoclinic and tetragonal phases, could act as a barrier to prevent the formation of oxide nodules. The azimuthal XRD cos²αsin²ψ technique was used to measure residual stress of t-ZrO₂ and m-ZrO₂ phases and reveal the distribution of each phase in the specimens. The results indicated that the oxide phases were not uniformly distributed in the specimens. The azimuthal XRD cos²αsin²ψ technique can be used as a convenient and nondestructive method to examine the uniformity of the phase distribution in a multi-phase oxide layer. The experimental results revealed that producing a ZrO₂ coating on Zry-4 by vacuum annealing of ZrN-coated Zry-4 was infeasible.
Due to increasing attentions to supercritical water reactor (SCWR), the corrosion response of materials in supercritical water (SCW) has attracted lots of research interests. This paper summarizes the current corrosion test results conducted in Shanghai Jiao Tong University (SJTU) and compares the data with literature data. The corrosion resistances of candidate materials (including ferritic/martensitic (F/M) steel, austenitic stainless steel, alumina-forming austenitic (AFA) stainless steel and oxide dispersion strengthened (ODS) steel) in supercritical water are discussed and analyzed. High Cr content austenitic stainless steel, proper Cr content AFA stainless steel and ODS steel are promising cladding materials for SCWR since they generally have satisfactory corrosion resistance even in 650 °C SCW.
A simple methodology was developed for the determination of trace metallic constituents (Ag, B, Ba, Bi, Ca, Cd, Ce, Ga, Gd, In, K, La, Li, Lu, Mg, Mn, Na, Co, Cr, Cu, Dy, Eu, Fe, Nd, Ni, Pb, Pr, Sm, Sr, Tl, and Zn) in Zr-Nb alloy. The systematic study of the spectral interference of Zr as well as Nb was carried out on these analytes. This includes identification of interference-free lines, tolerance level, and correction factors associated with each line. The analytical performance of these lines including detection limit, sensitivity, and linear dynamic range was also investigated. Based on these two factors, the best line of each analyte was chosen. Chemical separation of the major matrix followed by the analysis of the raffinate was the strategy adopted in the development of the method. To optimize the chemical separation procedure, TBP, TOPO, and DHOA were used for preferential separation of the major matrix without loss of the analytes even at trace levels. The optimized method was validated using a synthetic sample, which revealed that five contacts of 1.1 M TBP in dodecane was the ideal choice for separation and all of these analytes can be determined at the 1-mg L⁻¹ level in a Zr-Nb matrix with a RSD less than 5%.
The oxidation behavior and microstructure evolution of Fe-13Cr-6Al-2Mo-0.5Nb alloys with various yttrium contents ranging from 0 to 0.23 wt.% have been investigated at 1200 °C for 4 h in steam environment. The oxide weight gain decreased with minor addition of yttrium, but this tendency was lost with excess addition. The grain boundary migration was hindered due to high binding energy of yttrium atom and vacancy. The combination of yttrium with vacancies and a large number of fine Laves phases at grain boundaries suppressed the outward diffusion of Al³⁺, resulting in the change of oxide growth mechanism and oxide scale adhesion.
To better understand the role of dissolved oxygen (DO) in affecting corrosion behavior of zirconium alloys, the Zr–0.85Sn–0.16Nb–0.37Fe–0.18Cr (wt.%) alloy was corroded in super-heated steam at 500 °C and 10.3 MPa under 1×10⁻⁶ DO and deaeration conditions. The microstructure of the alloy and oxide films was investigated by SEM, TEM, EDS and EBSD. Results show that the corrosion is aggravated under 1×10⁻⁶ DO. Compared with the deaeration condition, the oxide film is looser, and has more micro-cracks and more uneven inner surface under DO condition. For the oxide film forming under deaeration condition, the selected area diffraction (SAED) spots of planes (002)m, and (101)t are strong, while those of the (001)m and are weak. However, for the oxide film forming under DO condition, the SAED spots of planes (111)m, (200)m and (101)t are strong, while those of the (100)m and (110)m are weak. The higher DO content in super-heated steam accelerates the growth of oxide films, thus decreasing the corrosion resistance of zirconium alloys.
Deformation twinning is a prevalent plastic deformation mode in hexagonal close-packed (HCP) materials, such as magnesium, titanium, and zirconium, and their alloys. Experimental observations indicate that these twins occur heterogeneously across the polycrystalline microstructure during deformation. Morphological and crystallographic distribution of twins in a deformed microstructure, or the so-called twinning microstructure, significantly controls material deformation behavior, ductility, formability, and failure response. Understanding the development of the twinning microstructure at the grain scale can benefit design efforts to optimize microstructures of HCP materials for specific high-performance structural applications. This article reviews recent research efforts that aim to relate the polycrystalline microstructure with the development of its twinning microstructure through knowledge of local stress fields, specifically local stresses produced by twins and at twin/grain–boundary intersections on the formation and thickening of twins, twin transmission across grain boundaries, twin–twin junction formation, and secondary twinning.
In this paper, the plasma electrolytic oxidation (PEO) and pulsed laser deposition (PLD) methods were combined for the preparation of ZrO2 and Cr/CrN/Cr2O3 composite film on Zr-4 alloy. The PEO process was carried out in the alkaline electrolyte containing KOH and Na3PO4. For the PLD process, a Cr2N target was decomposed to CrN and Cr, and the latter is partially oxidized under the background gas of oxygen to form a Cr/CrN/Cr2O3 film. The composite film exhibits excellent corrosion resistance at high temperature. This work provides a methodological guidance for the preparation of composite film on zirconium alloys, which are usually used as structural materials in the nuclear industry.
Using the method of in situ impedance spectroscopy, an optimization of plasma electrolytic oxidation (PEO) was performed, and the influence of the pulse frequency was evaluated for the coating of Zr-1Nb alloy. It was shown that the in situ impedance modulus decreased with increasing the pulse frequency in a pulsed unipolar mode. Also, the protective properties of the coating improved with frequency; this was attributed to the increase in the uniform distribution of the microdischarges over the sample surface. As a result, the optimal frequency range of PEO was justified based on the impedance spectroscopy studies.
In present paper, Zircaloy-4 surface alloying with copper (Cu) using a high-current pulsed electron beam (HCPEB) device was firstly studied. For investigating the changes in microstructures and corrosion resistance prior and after surface alloying, a series of characterization methods were adopted. Results revealed that, after surface alloying by HCPEB, a Cu-riched alloying layer with a thickness of about 3 μm was formed within the melted layer, inside which martensitic transformation was largely inhibited and large numbers of ultra-fine equiaxed β grains were generated. Most of the Cu elements in the alloying layer were uniformly distributed in the β-Zr supersaturated solid solution, others were existed in the form of fine dispersed ZrCu SPPs. Besides, Zr(Fe,Cr)2 SPPs were dissolved into the Zr-matrix within the HCPEB affect area. Compared with initial and directly irradiated specimens, the specimen after surface alloying exhibited extremely excellent anti-corrosion performance at 500 °C/10.3 MPa superheated steam, which was mainly attributed to those supersaturated Cu elements.
Several zirconium alloys with differing weight percentages of Cr, Fe, Cu, and Mo were exposed to flowing, pure supercritical water at 500°C for up to 150 days in an effort to determine their corrosion behavior for consideration in the supercritical water reactor. The weight gains of the alloys were measured, and oxides were characterized after various times. The test results showed a wide range of corrosion behavior depending on the alloy composition and process temperature. The alloys most resistant to corrosion were those containing Cr and Fe, three of which showed protective stable oxides, low corrosion rates, and no breakaway behavior. The ZrCr, ZrCu, ZrMo, and ZrCuMo alloys all exhibited high corrosion rates and non-protective oxides. Analysis of the oxide layer showed that the oxide consisted mostly of monoclinic zirconia (ZrO2). The structure of the oxide-metal interface in the five protective alloys exhibited characteristics that were also seen in protective oxides formed at low temperature, especially the presence of a suboxide layer and an intense (002)T peak at the interface, indicating the presence of a highly oriented tetragonal phase associated with the protective oxide. The change in corrosion kinetics from cubic to linear was directly linked to the size and density of cracks in the oxides.
The purpose of this study was to determine the corrosion behavior of austenitic and ferritic-martensitic (F-M) alloys in pure supercritical water and to gain a better understanding of the kinetics and thermodynamics of the oxidation mechanism. Austenitic stainless steel alloys 304L. 316L, 316, 800H, and D9, and F-M alloys T91, HT9, HCM12A and NF616, and nickel-base alloys 625 and 690 were exposed in supercritical water over the temperature range 400°C to 550°C for periods up to 1026 hr. The dissolved oxygen concentration ranged from <10ppb to 2000 ppb. Weight gains, which vary roughly according to alloy class, are dependent on both the temperature and the oxygen concentration in supercritical water. Overall, the kinetics of oxidation of the F-M alloys are approximately an order of magnitude faster than that of the austenitic alloys, while the weight gains in the Ni-base alloys are the smallest. The scatter in weight gain data is smaller for the F-M alloys than for austenitic alloys. For low oxygen (10-100 ppb) conditions, stable oxide growth roughly followed a parabolic rate law and the activation energies for oxide growth on each of the alloys were consistent with the values for cation diffusion through the oxide, indicating that oxide growth is probably controlled by solid state diffusion of cations through the oxide film. Exposure experiments in low oxygen SCW resulted in the formation of a two-layer oxide on the austenitc and F-M steels, in which the outer layer is porous magnetite and the inner layer is a denser iron-chromium spinel. The oxide on 625 was too thin to measure. Exposure in higher oxygen concentrations resulted in higher oxidation rates and, in some cases, a tendency towards exfoliation.
The diffusion coefficients of oxygen in zirconium were determined in the temperature range of 165–230°C. Thin oxide films were dissolved and the oxygen diffused by heating the samples at different temperatures. The evolution of film thicknesses was followed using XPS (X-ray photoelectron spectroscopy) techniques. A model was developed for film dissolution that related the evolution of film thicknesses with the diffusion coefficients of O in Zr. The equation D = 3.1 × 10-6 exp(-23750/RT) cm2/s is satisfied by the calculated coefficients. This expression is close to that obtained by Ritchie and Atrens [J. Nucl. Mater. 67 (1977) 254] after a regression analysis of several author's data and called RGB (grain boundary diffusion). The posibilities and limitations of XPS as a non-conventional technique for diffusion measurements are discussed.
The surface oxidation characteristics of oxygen saturated zirconium (Zr:Oss) have been investigated between room temperature and 1190 K using Auger spectroscopy and sputter-depth-profiling techniques. Zr:Oss substrates were prepared by absorbing oxygen from surface oxide layers into Zr foils at high temperature. Quantitative Auger analysis confirms that the composition of Zr:Oss is approximately ZrO0.44.Zr:Oss oxidizes slightly faster at room temperature than pure Zr during initial O2 exposure. The high temperature surface oxidation of Zr:Oss was studied in the absence of interference from the simultaneous bulk absorption of oxygen, which occurs with pure Zr substrates. The thickness of the ZrO2 surface layer formed on Zr:Oss increases with the extent of O2 exposure, exposure time and oxidation temperature up to 900 K. Above 900 K the surface oxidation is not complete to ZrO2.
The paper reports on the identification of different oxidation states of zirconium which result from the interaction of oxygen with a clean (Ar+-sputtered) polycrystalline zirconium surface in low oxygen pressures (PO2 = 10-6 mbar) at room temperature. The XPS spectra corresponding to the Zr (3d3/2,5/2)-core levels have been analyzed in detail after different oxygen exposures. The shape of the peaks was approximated by a mixed Gaussian-Lorentzian function. The spectra could not be correctly fitted unless the four oxidation states were taken into account. It is suggested that a series of different sub-oxides (ZR2O, ZrO and Zr2O3) form during the first steps of the oxidation process before ZrO2 starts to grow.
The initial oxidation of zirconium and Zircaloy-2 with oxygen and water vapor has been investigated at room temperature with Auger electron and X-ray photoelectron spectroscopies. Three suboxides of Zr2O, ZrO and Zr2O3 accompanied by the oxide of ZrO2 are formed on surfaces in both oxygen and water vapor atmospheres. Formation of ZrO2 starts at about 0.2–0.4 L (1 L ≡ 10−6 Torr s) oxygen exposure. The oxidation rate of zirconium with oxygen is slightly larger than that of Zircaloy-2. In the case of low pressure water vapor exposure (< 10 L), the amounts of suboxides on the zirconium are larger than on the Zircaloy-2, since dissociation of H2O molecules proceeds more easily on the former. The ZrO2 starts to form after a 1.0 L water vapor exposure, and its formation rate in water vapor ambience is much smaller than that in oxygen. An average composition of ZrO1.50–1.53 and film thickness of 2.3 nm are obtained for surfaces exposed to 30 L oxygen, and a 1.3 run thick film with composition Zr1.06 is obtained for 30 L of water vapor exposure.
To further advance the mechanistic understanding of microstructural evolution in zirconium alloys for high burnup applications, it is important to obtain a quantitative measurement of the volume fractions of second-phase precipitates present in the bulk alloys as a function of the heat treatment and irradiation fluence. In this work, X-ray diffraction from a synchrotron radiation source was used to identify and follow the growth kinetics of second-phase particles in zirconium alloys. The high energy flux, energy resolution and signal-to-noise ratio of this light source allowed us to study the very small (<0.2–0.4%) precipitate volume fractions in these alloys. A preliminary calculation of the precipitate volume fraction as a function of cumulative annealing parameter (CAP) showed the precipitate volume fraction increases rapidly above 10−19 h and saturates at about 10−17 h.
A transmission electron microscopy investigation was performed on oxides formed on three zirconium alloys (Zircaloy-4, ZIRLO and Zr 2.5Nb) in pure water and lithiated water environments. This research is part of a systematic study of oxide microstructures using various techniques to explain differences in corrosion rates of different zirconium alloys. In this work, cross-sectional transmission electron microscopy was used to determine the morphology of the oxide layers (grain size and shape, oxide phases, texture, cracks, and incorporation of precipitates). These characteristics were found to vary with the alloy chemistry, the corrosion environment, and the distance from the oxide/metal interface. These are discussed and used in conjunction with observations from other techniques to derive a mechanism of oxide growth in zirconium alloys.
In this thesis research, oxide layers formed on Zr-based alloys in high temperature water and steam autoclave environments are characterized in an effort to understand the oxide growth mechanism. The study included oxides formed on Zircaloy-4, ZIRLO, Zr-2.5Nb and Zr-2.5Nb-0.5Cu alloys in 360°C pure water, 360°C lithiated water and 400--426°C steam autoclave environments, which is pressurized to 2200 psi. The oxides are characterized using microbeam x-ray diffraction, microbeam x-ray fluorescence, transmission electron microscopy, transmitted light optical microscopy and scanning electron microscopy. The variation of the crystal structure and the chemical composition across the oxide layers are determined on the sub-micron (˜0.2 mum) scale by using the high flux micro-x-ray beam on the 2-IDD beamline at Advanced Photon Source at Argonne National Laboratory. This unique synchrotron radiation facility allowed us to examine the oxides in a more detailed manner than previously done as it combines high flux with spatial resolution to produce new and detailed information on oxide structure. A marked periodicity was observed in various oxide features (cracks, diffraction intensity, white/black bands etc.) as a function of distance from the oxide/metal interface was observed by the several techniques used in this study. The observed microstructural periodicity was linked to the periodicity observed in the corrosion data. The structure of the oxide/metal interface is different from that in the bulk of the oxide and does not change significantly as the oxide advances, indicating a restructuring of the oxide in the region behind the oxide front to maintain a self-similar structure at the oxide front as the oxide advances. A highly oriented tetragonal zirconia phase was observed at the oxide/metal interface region which is the precursor of the highly oriented monoclinic phase in the bulk of the oxide layer, which forms to minimize stresses. The measured tetragonal phase fraction higher near the oxide/metal interface region than in the bulk of the oxide. These tetragonal grains are oriented randomly and their size is smaller than the highly oriented tetragonal grains at the oxide/metal interface. The structure of the region in the metal near the oxide/metal interface (called the "suboxide" region) is different than the bulk of the oxide layer and the metal. A Zr3O suboxide phase with alpha-Zr and small equiaxed tetragonal ZrO2 grain was identified in this region and its region of existence corresponds well to a region where oxygen content is high as determined by transmission electron microscopy analysis. A comparison between the oxides formed in different alloys shows systematic differences: oxide that grows slowly exhibit well-oriented, and comparatively thicker columns of monoclinic oxide than the oxide that grows faster. Oxides formed in lithiated water are more porous than those formed in pure water. These observations helped us to propose a mechanism for the oxide growth which can be made to understand the influence of alloy microstructure on corrosion rate.
A model for predicting the texture development in monoclinic ZrO2 films grown on Zr alloys has been proposed. The model is based on assumptions of selective nucleation and anisotropic oxide grain growth. In this model during the stage of oxide nucleation, oxide and substrate lattice matching is the main factor influencing the orientation of nuclei. In the stage of oxide grain growth, the matching stress at oxide–metal interface plays a major role. The predicted oxide textures have been compared with the results from oxidation experiments performed on three samples obtained from different faces of the Zr–2.5% Nb pressure tube. A good agreement between the predicted oxide texture and the measured texture is observed. Texture analysis of the monoclinic oxide is based on the orientation distribution function (ODF).
This paper presents a compilation and an analysis of data relating to oxygen diffusion in alpha-zirconium. Internal friction and strain ageing measurements of the oxygen jump time extend the range of conventional data to the temperature range of technological importance. The data span the temperature range 290 to 1500°C and fourteen orders of magnitude in the diffusion coefficient. Two equations can be used to represent the volume diffusion of oxygen in α-zirconium over the whole temperature range. For temperatures in the range 290 to 650°C the volume diffusion coefficient is given by D = 0.0661 exp( -44000 RT), 290 < T ≤ 650°C, and is attributed to the jumps of oxygen interstitials in the basal plane. For temperature in the range 650 to 1500°C the volume diffusion coefficient is given by D = 16.5 exp( -54700 RT), 650 < T < 1500°C, and it is suggested that jumps of oxygen interstitials in the c-direction are rate controlling.
The structures of oxides formed in water and lithiated water on three Zr-based alloys with varied corrosion behavior were studied with micro-beam synchrotron radiation and optical microscopy. Micro-beam synchrotron radiation (0.2 μm spot) has a unique combination of high elemental sensitivity (ppm level) and fine spatial resolution that allowed the determination of various oxide characteristics such as phase content, texture, grain size, and composition as a function of distance from the oxide-metal interface.
Micro-beam X-ray fluorescence shows that the oxides formed in lithiated water have increased levels of Fe absorbed from the autoclave environment indicating greater oxide porosity in these oxides. The phase content, texture, and grain size of oxides were studied in detail using synchrotron radiation micro-beam diffraction for samples corroded in water and lithiated water. A remarkable periodicity was observed in the oxide structures using various techniques including X-ray peak intensities for both monoclinic and tetragonal zirconia, texture, and optical microscopy. The periods were similar to the transition period and were less visible in the oxides that behaved worse in lithiated water. These results are discussed in terms of models of oxide growth and of the differences between alloys.
The effects of texture, grain size and cold work on the precipitation of oriented hydrides in Zircaloy tubing and plate have been investigated. Hydride orientations were measured in samples cooled under no applied stress and also under a tensile stress of 20 000 psi. In the former samples, the orientations were found to correlate with crystallographic texture as characterized by inverse pole figure data. In the stressed samples, the stress produced a change in hydride orientation which was related more to grain size and degree of residual cold work.
Zirconium was oxidized at 700 °C for 1000 h and at 850 °C for 160 h. The oxygen penetration into the metal was measured with an electron microprobe and within the experimental accuracy it was found that the oxygen diffusion obeys the normal diffusion equation with concentration-independent diffusion constant. The hardness was measured in the diffusion zone and the non-linear relation between hardness and oxygen concentration was derived. A plateau at a composition corresponding to Zr6O was found. The presence of the ordered phase Zr3O in thin foils of oxidized zirconium was established by selected area diffraction. The dislocation density increased during the oxidation and the new dislocations had very little mobility.
A detailed study was undertaken of oxides formed in 360 °C water on four Zr-based alloys (Zircaloy-4, ZIRLO™,1 Zr–2.5%Nb and Zr–2.5%Nb–0.5%Cu) in an effort to relate oxide structure to corrosion performance. Micro-beam X-ray diffraction was used along with transmitted light optical microscopy to obtain information about the structure of these oxides as a function of distance from the oxide–metal interface. Optical microscopy revealed a layered oxide structure in which the average layer thickness was inversely proportional to the post-transition corrosion rate. The detailed diffraction studies showed an oxide that contained both tetragonal and monoclinic ZrO2, with a higher fraction of tetragonal oxide near the oxide–metal interface, in a region roughly corresponding to one oxide layer. Evidence was seen also of a cyclic variation of the tetragonal and monoclinic oxide across the oxide thickness with a period of the layer thickness. The results also indicate that the final grain size of the tetragonal phase is smaller than that of the monoclinic phase and the monoclinic grain size is smaller in Zircaloy-4 and ZIRLO than in the other two alloys. These results are discussed in terms of a model of oxide growth based on the periodic breakdown and reconstitution of a protective layer.
Behavior of Structural Materials for Fuel and Control Elements in Light Water Reactor Cooled Power Reactors, IAEA IAEA-STI/Pub 721
- F Garzarolli
- H Stehle