Nuclear fuel debris generated at the Fukushima Daiichi nuclear power plant during the loss of coolant accident in 2011, still resides within the reactor units, constantly cooled by water. Until it is retrieved, the fuel debris will corrode, releasing radioactive elements into the coolant water and the ground surrounding the reactors. To predict the corrosion behaviour of these materials, and to establish parameters for experiments with U-containing and real fuel debris, the corrosion of two surrogate fuel debris materials, with a composition of Ce(1-x)ZrxO2 (x = 0.2 and 0.4), was investigated. Materials were synthesised by a wet chemistry route and pellets were sintered at 1700°C in air atmosphere. Due to the slow corrosion kinetics, aggressive conditions were applied, and corrosion experiments were performed in 9 mol.L-1 HNO3 under static conditions. The incorporation of Zr into the structure of Ce reduced the normalised dissolution rate; from (3.75 ± 0.15) × 10-6 g.m-2.d-1 to (4.96 ± 0.28) × 10-6 g.m-2.d-1 for RL(Ce) of Ce0.8Zr0.2O2 and Ce0.6Zr0.4O2, respectively.

The fatigue behavior of a low-cost Zr52.1Ti5Cu17.9Ni14.6Al10Y0.4 (at%) (ZrCuNi-based) bulk-metallic glass (BMG) prepared by industrial-grade material was investigated under three-point bending loading modes. In order to obtain the fatigue stress-life (S-N) data, stress-controlled experiments were conducted using a computer-controlled material test system electrohydraulic testing machine at 60 Hz with a 0.1 R ratio in the air at room temperature. The fatigue limit (~174 MPa) in stress amplitude and fatigue ratio (~0.14) of this BMG is comparative to the similar BMG (Vit-105) prepared by high pure raw materials. The crack initiated from inclusions near the rectangular corners at the outer surface of the rectangular beam due to stress concentration. The striations and vein-like patterns were observed in the crack propagation region and fast fracture region, respectively.

The oxide films formed on spent fuel claddings are regarded as a potential source of the instantaneous release of radionuclides, such as 14C, after waste disposal. We investigated the instant release fraction using the irradiated oxide exfoliated from a Zircaloy-2 water rod, whose bundle burnup was 53.0 GWd/MTU. We performed a rapid leaching test in a dilute NaOH solution (pH of 12.5) for 10 min in an ultrasonic bath to ensure the release of radionuclides. The activity ratios of the leached amount to the total amount for 14C, 60Co, and 125Sb were extremely low at approximately 10-4 to 10-3, among which the maximum value was 2.65 × 10-3 for 125Sb. These ratios were higher than that predicted from the thermodynamic solubility of ZrO2, i.e., less than 10-6. However, given the low ratios, it is too conservative to regard the inventory of all radionuclides in the Zircaloy oxide as instantaneous release. A small part of the released 14C was found as volatile species.

The recent discovery of ferroelectric behavior in doped hafnia-based dielectrics, attributed to a non-centrosymmetric orthorhombic phase, has potential for use in attractive applications such as negative differential capacitance field-effect-transistors (NCFET) and ferroelectric random access memory devices (FeRAM). Alloying with similar oxides like ZrO2, doping with specific elements such as Si, novel processing methods, encapsulation and annealing schemes are also some of the techniques that are being explored to target structural modifications and stabilization of the non-centrosymmetric phase. In this study, we utilized synchrotron-based x-ray diffraction in the grazing incidence in plane geometry (GIIXRD) to determine the crystalline phases in hafnia-zirconia (HZO) compositional alloys deposited by atomic layer deposition (ALD). Here we compare and contrast the structural phases and ferroelectric properties of mechanically confined HZO films in metal-insulator-metal (MIM) and metal-insulator-semiconductor (MIS) structures. Both MIM and MIS structures reveals a host of reflections due to non-monoclinic phases in the d-spacing region between 1.75Å to 4Å. The non-monoclinic phases are believed to consist of tetragonal and orthorhombic phases. Compared to the MIS structures a suppression of the monoclinic phase in MIM structures with 50% zirconia or less was observed. The correlation of the electrical properties with the structural analysis obtained by GIIXRD highlights the importance of understanding the effects of the underlying substrate (metal vs. Si) for different target applications.

The construction of zirconium (Zr) budgets for metamorphic reactions in high-grade rocks provides new insight into zircon growth during metamorphism. In this study we target reactions involving garnet, as they enable zircon growth to be related to known pressure and temperature conditions. Two reactions involving the breakdown of Zr-bearing garnet from Rogaland, SW Norway have been investigated in detail, showing contrasting behaviour of Zr, with zircon formation being subject to the solubility of Zr in product phases. In the decompression reaction garnet + sillimanite + quartz → cordierite, Zr released during garnet breakdown cannot be incorporated into the cordierite structure, resulting in zircon nucleation and growth. In contrast, for the reaction garnet + biotite + sillimanite + quartz → osumilite + orthopyroxene + spinel + magnetite, no new zircon growth takes place, despite the garnet involved containing more than double the Zr concentration of the former reaction. In the latter case, all the Zr released by garnet breakdown can be detected in the product phases osumilite and orthopyroxene, thereby preventing growth of new metamorphic zircon. This study highlights the potential for high resolution geochronology in metamorphic rocks by relating zircon growth to specific metamorphic reactions.

Zirconium has attracted a lot of attention recently due to its distinctive properties that make it suitable for extensive applications in the nuclear power and chemical industry. Zirconium and its alloys are undergoing long-term development as promising materials for the nuclear industry and power engineering. Recently, advanced Zr-based alloys are aimed for service in more severe operating conditions such as higher burn-up, increased operation temperature, and high-PH operation. In this work we observe the temperature dependence of Zr50Nb50, Zr78Nb22, Zr78Nb19Co3 and Zr50Nb49Sn1. It was observed that ternary additions with small atomic percentages of Co and Sn have significant impact on Zr-Nb alloy; and their elastic properties showed a possible enhancement on high temperature applications and physical strength.

Understanding the precipitation of brittle hydride phases is crucial in establishing a failure criterion for various zirconium alloy nuclear fuel cladding. Accordingly, it is important to quantify the sensitivity of hydride precipitation to the component microstructure. This experimental investigation focuses on two microstructural characteristics and their role as hydride nucleation sites: The grain size and the alloy chemical composition. Samples of commercially pure zirconium (Zr-702) and Zircaloy-4, each with a wide range of grain sizes, were hydrided to 100 ppm and micrographs of the hydride distribution were optically analyzed for inter-granular and intra-granular precipitate sites. For most grain sizes, it was found that a significantly lower fraction of the precipitated hydrides nucleated at grain boundaries in Zircaloy-4 than in Zr-702, suggesting that a higher SPP content encourages the formation of intra-granular hydrides. Moreover, this effect became more prominent as the grain size increased; large-grain specimens contained a higher fraction of intra-granular hydrides than small-grain specimens of both Zr-702 and Zircaloy-4, highlighting the potency of grain boundaries as nucleation sites and how SPPs can influence the hydride distribution profile.

Bulk metallic glass and their composites are unique new materials which have superior mechanical and structural properties as compared to existing conventional materials. Owing to this, they are potential candidates for tomorrow’s structural applications. However, they suffer from disadvantages of poor ductility and little or no toughness which render them brittle and they manifest catastrophic failure on the application of force. Their behavior is dubious and requires extensive experimentation to draw conclusive results. In present study, an effort has been made to overcome this pitfall by simulation. A quantitative mathematical model based on KGT theory has been developed to describe nucleation and growth of second phase dendrites from melt in glassy matrix during solidification. It yields information about numerical parameters necessary to understand the behaviour of each individual element in multicomponent sluggish slurry and their effect on final microstructural evolution. Model is programmed and simulated in MATLAB®. Its validation is done by comparison with identical curves reported in literature previously for similar alloys. Results indicate that the effect of incorporating all heat transfer coefficients at macroscopic level and diffusion coefficients at microscopic level play a vital role in refining the model and bringing it closer to actual experimental observations. Two types of hypo and eutectic systems namely Zr65Cu15Al10Ni10 and Zr47.5Cu45.5Al5Co2 respectively were studied. Simulation results were found to be in good agreement with prior simulated and experimental values.

Understanding the ratcheting effect of hydrogen and hydride accumulation in response to thermal cycling is important in establishing a failure criterion for zirconium alloy nuclear fuel cladding. We propose a simple discrete dislocation plasticity model to study the evolution of the dislocation content that arises as a micro-hydride repeatedly precipitates and dissolves over a series of thermal cycles. With each progressive thermal cycle, we find a steady growth in the residual dislocation density in the vicinity of the hydride nucleation site; this corresponds to a gradual increase in the hydrogen concentration and, consequently, the hydride population. The simulated ratcheting in the dislocation density is consistent with experimental observations concerning the hysteresis in the terminal solid solubility of hydrogen in zirconium, which can be correlated to the plastic relaxation of hydrides.

Hydride re-orientation in high burn-up PWR fuel rods was induced by high pressure at high temperatures. The high-burnup specimens were sectioned from PWR rods taken from a 15×15 assembly of the H.B. Robinson (HBR) Unit 2 reactor. Out-of cell benchmark tests performed on unirradiated hydrided Zircaloy-4 specimens were conducted to determine the appropriate temperature, pressure, cooling rate, and number of cooling cycles for the reorientation of the irradiated in-cell specimens. The in-cell hydride reorientation tests were performed using highburnup fuel specimens under a hoop stress ≈145 MPa at 400ºC. The specimens were heated to the target temperature of 400ºC, held for 3 hours, cooled at 1ºC/min to 170ºC,and then heated at 1ºC/min to the target temperature again for five cycles. Post test metallographic examinations showed that a significant amount of radial hydrides were induced in the HBR fuel rods. The length of radial hydride was up to 60 µm. For unirradiated materials, the ductility of the radial hydride treated specimens is significantly reduced as compared to the as-hydrided specimens having the same hydrogen concentration (≈300 wppm in this work). The mechanical testing on irradiated fueled samples with hydride reorientation experiments have been performed, and will be reported separately in the near future.

Two-sided oxidation experiments were recently conducted at 1000-1200°C in flowing steam with samples of sponge-based Zr-1Nb alloy E110. Although the old electrolytic E110 tubing exhibited a high degree of susceptibility to nodular corrosion and experienced breakaway oxidation rates in relatively short time, the new sponge-based E110 has demonstrated steam oxidation behavior comparable to Zircaloy-4. The sponge-based E110 followed the parabolic law, and the derived oxidation rate constant is in good agreement with the Cathcart-Pawel (CP) correlation at 1100-1200°C. For 1000°C oxidation, the weight-gain of sponge-based E110 is much lower than Zircaloy-4. No breakaway oxidation was observed at 1000°C up to 8000 s. Ring compression tests were conducted to evaluate the residual ductility of oxidized samples at room temperature and at 135°C. All sponge-based E110 specimens were still ductile at 135°C after being oxidized up to 20% equivalent cladding reacted at 1000-1200°C. Metallographic examinations were performed on oxidized E110 specimens to correlate material performance with microstructure.

The growth of thin films of chemical vapour deposition (CVD) diamond onto flat samples of pure Zr has been studied using various CVD growth conditions in a hot filament reactor. We find that although growth is straightforward, adhesion of the diamond layer onto the Zr is poor, with the diamond layer often delaminating upon cooling. SIMS depth profiles show this to be due to the presence of a strongly-bonded native oxide on the Zr surface which is not removed in the reducing H2 atmosphere during CVD. This, plus the lack of any substantial carbide interfacial layer to ‘glue’ the diamond onto the surface, together with a poor thermal expansion mismatch between Zr and diamond, and the Zr hcp-to-bcc phase transition at ∼860°C, all lead to poor adhesion. Some of these difficulties can be reduced by depositing at lower temperature (<500°C) at the cost of poorer quality diamond.

Five well-known Zr-based alloys of the systems Zr–Cu–Al–(Ni–Nb, Ni–Ti, Ag) (Cu = 15.4–36 at.%) with the highest glass-forming ability were comparatively analyzed regarding their pitting corrosion resistance and repassivation ability in a chloride-containing solution. Potentiodynamic polarization measurements were conducted in the neutral 0.01 M Na2SO4 + 0.1 M NaCl electrolyte and local corrosion damages were subsequently investigated with high resolution scanning electron microscopy (HR-SEM) coupled with energy dispersive x-ray spectroscopy (EDX). Both pitting and repassivation potential correlate with the Cu concentration, i.e., those potentials decrease with increasing Cu content. Pit morphology is not composition dependent: while initially hemispherical pits then develop an irregular shape and a porous rim. Corrosion products are rich in Cu, O, and often Cl species. A combination of low Cu and high Nb or Ti contents is most beneficial for a high pitting resistance of Zr-based bulk metallic glasses. The bulk glassy Zr57Cu15.4Al10Ni12.6Nb5 (Vit 106) and Zr52.5Cu17.9Al10Ni14.6Ti5 (Vit 105) alloys exhibit the highest pitting resistance.

A non-destructive neutron scattering method was developed to precisely measure the uptake of total hydrogen in nuclear grade Zircaloy-4 cladding. The hydriding apparatus consists of a closed stainless steel vessel that contains Zircaloy-4 specimens and hydrogen gas. By controlling the initial hydrogen gas pressure in the vessel and the temperature profile, target hydrogen concentrations from tens of ppm to a few thousands of ppm have been successfully achieved. Following hydrogen charging, the hydrogen content of the hydrided specimens was measured using the vacuum hot extraction method (VHE), by which the samples with desired hydrogen concentration were selected for the neutron study. Small angle incoherent neutron scattering (SAINS) were performed in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). Our study indicates that a very small amount (≈ 20 ppm) hydrogen in commercial Zircaloy-4 cladding can be measured very accurately in minutes for a wide range of hydrogen concentration by a nondestructive method. The hydrogen distribution in a tube sample was obtained by scaling the neutron scattering rate with a factor, which is determined by calibration process with direct chemical analysis method on the specimen. This scale factor can be used for future test with unknown hydrogen concentration, thus provide a nondestructive method for absolute hydrogen concentration determination.

The carbon-14 generated in Zircaloy (Zry) hull waste is considered an important radionuclide in the TRU waste geological disposal concept in Japan. Given that the metal Zry is highly corrosion-resistant in the anaerobic and low-temperature conditions of the repository, and that the C-14 release rate is assumed to be controlled by the corrosion rate, a variety of corrosion and leaching tests have been performed. However, since the Zry corrosion rate is extremely slow, it is not possible to predict long-term corrosion behavior through low-temperature corrosion tests conducted in a reasonable time period. A vast amount of testing has been conducted in the higher-temperature range of 523 to 633 K, and corrosion correlations have been obtained from these tests. Corrosion correlations have been used to predict the corrosion rate of Zry in a tuff repository. Long-term Zry autoclave corrosion data have been analyzed to develop new corrosion correlations. Extrapolating these correlations to a lower temperature range requires verification that the mechanisms do not change over the range of testing and extrapolation. Factors that influence corrosion rates under geological disposal conditions, such as material and environmental factors, should also be examined. Corrosion correlations, factors influencing corrosion rates, the results of corrosion and leaching tests, and a preliminary evaluation are discussed.

Pre-storage drying-transfer operations and early stage storage expose cladding to higher temperatures and much higher pressure-induced tensile hoop stresses relative to normal operation in-reactor and pool storage under these conditions. Radial hydrides precipitate during cooling and could provide an additional embrittlement mechanism as the cladding temperature decreases below the ductile-to-brittle transition temperature. To simulate this behavior, unirradiated Zircaloy-4 samples were hydrided by a gas charging method to levels that encompass the range of hydrogen concentrations observed in current used fuel. Mechanical testing was carried out by the ring compression test (RCT) method at various temperatures to evaluate the sample’s ductility for both as-hydrided and post-hydride reorientation treated specimens. As-hydrided samples with higher hydrogen concentration (>800 ppm) resulted in lower strain before fracture and reduced maximum load. Increasing RCT temperatures resulted in increased ductility of the as-hydrided cladding. A systematic radial hydride treatment was conducted at various pressures and temperatures for the hydrided samples with H content around 200 ppm. Following the radial hydride treatment, RCTs on the hydride reoriented samples were conducted and exhibited lower ductility compared to as-hydrided samples.

Electrochemical Impedance Spectroscopy (EIS) and the Parallel Electrical Dielectric Response Analysis (PEDRA) application were used to describe the inner barrier oxide films on irradiated zirconium alloys. This is achieved with minimal surface preparation and without disturbing the outer porous oxide. These two distinguishable inner and outer oxide layers result from a growth-fracture oxidation mechanism. Key to success of the EIS technique in describing the barrier oxide layer are: 1) the model and procedure used to fit EIS spectra, 2) the validation of the fit, and 3) converting circuit parameters (R, C and n) into physical attributes of the barrier oxide.

The barrier oxide is defined as the inner-dense layer adjacent to the metal-oxide interface. The integrity of barrier oxide is thought to effect both oxidation (i.e. access of water to the interface), and hydrogen pickup (i.e. failure hydrogen to escape away from the interface). Using EIS and the PEDRA application, the barrier oxide is described in terms of multiple independent dielectric responses to yield a unique 'micro-macro' picture of the barrier oxide that can be used to explain observed H pickup behavior.

Corrosion behavior is a key issue in the assessment of disposal performance for activated waste such as spent fuel assemblies (i.e., hulls and end-pieces) because corrosion is expected to initiate radionuclide (e.g., C-14) leaching from such waste. Because the anticipated corrosion rate is extremely low, understanding and modeling Zircaloy (Zry) corrosion behavior under geological disposal conditions is important in predicting very long-term corrosion. Corrosion models applicable in the higher temperature ranges of nuclear reactors have been proposed based on considerable testing in the 523−633 K temperature range.

In this study, corrosion tests were carried out to confirm the applicability of such existing models to the low temperature range of geological disposal, and to examine the influence of material, environmental, and other factors on corrosion rates under geological disposal conditions. A characterization analysis of the generated oxide film was also performed.

To confirm applicability, the corrosion rate of Zry-4 in pure water with a temperature change from 303 K to 433 K was obtained using a hydrogen measuring technique, giving a corrosion rate for 180 days of 8 × 10-3 μm/y at 303 K.

To investigate the influence of various factors, corrosion tests were carried out. The corrosion rates for Zry-2 and Zry-4 were almost same, and increased with a temperature increase from 303 K to 353 K. The influence of pH (12.5) compared with pure water was about 1.4 at 180 days at 303 K.

This study investigated the microstructure and machining characteristics of a Zr38.5Ti16.5Cu15.25Ni9.75Be20 bulk metallic glass (Zr-BMG) alloy machined using electro-discharge machining (EDM). After EDM, the hardening effect near the outer surface of the electro-discharge machined (EDMed) Zr-BMG alloy originated from the surface carbides of the recast layer, ZrC and TiC. The thickness of the recast layer, crater size, and the surface roughness increased with greater pulse energy. Furthermore, the EDM can generate a porous recast layer and convert the Zr-BMG alloy surface into a carbide surface, which is a potential method to fabricate biomaterials. Experimental results also show that the material removal rate of this alloy in the EDM process was significantly related to the pulse current IP and pulse duration τP. Many electro-discharge craters and recast materials were observed on the surface of the EDMed Zr-BMG alloy. The surface roughness of the EDMed Zr-BMG alloy was found to obey the empirical equation of Ra = β(IP × τP)α.

Ce0.35Zr0.65−xRExO2 (RE = Y and La; x = 0 and 0.10) and Ce0.35Zr0.50Y0.075La0.075O2 were prepared by a coprecipitation method. The textures, structures, oxygen storage capacity (OSC), and redox properties of all samples were investigated using Brunauer–Emmett–Teller surface area characterization, x-ray diffraction (XRD), Raman spectra, temperature-programmed technique, and oxygen pulse reaction. The results showed that the fresh Ce0.35Zr0.65O2 has cubic phase, 434 μmol/g of OSC, 82 m2/g of surface area, and good redox properties; after aging at 1000 °C, Ce0.35Zr0.65O2 still has cubic phase, 418 μmol/g of OSC, and 50 m2/g of surface area; when Y3+ or La3+ is added to CeO2–ZrO2, the aged Ce0.35Zr0.65−xRExO2 (RE = Y and La; x = 0 and 0.10) still remains cubic phase, high OSC, and large surface area (47 m2/g); when Y3+ and La3+ are simultaneously added into CeO2–ZrO2, a stable solid solution with cubic phase is formed and has 459 μmol/g of OSC; and the aged Ce0.35Zr0.50Y0.075La0.075O2 reaches to 60 m2/g of surface area and has 390 μmol/g of OSC.