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In order to investigate the occurring processes during long-term dry storage of spent fuel assemblies, a joined project called SPIZWURZ, between the Karlsruhe Institute of Technology and the Gesellschaft für Anlagen-und Reaktorsicherheit (GRS), was started. Aim of the SPIZWURZ project is the determination and quantification of the influence of texture and elastic strain on diffusion and solubility of hydrogen in three different zirconium alloys used in western Europe during a long-term cooling transient (1 K/d) starting at 400 °C. The strain in the cladding of an irradiated spent fuel rod shall be measured. Models predicting the formation of radial oriented hydrides will be validated, improved, and implemented in the GRS fuel rod performance code TESPA-ROD. This paper describes the SPIZWURZ project and already obtained first results.

Chromium carbide (CrC) coatings were proposed as an accident-tolerant fuel complementary concept to provide enhanced protection for the inner side of nuclear fuel claddings, with preliminary results showing promising performance. To evaluate the neutronics performance of CrC coatings, a reactor physics–based analysis was performed. A single VVER-1200 fuel assembly was used as a model, and the Monte Carlo code MCNPX was used to perform the calculations. Results were compared to previous work on metallic chromium performance as inner-side coating material. Results showed that CrC coatings generally have less negative impacts on neutronics performance compared to chromium coatings. Neutron flux spectra showed slight reductions in the thermal energy region that reached up to −0.6% in a 40-µm CrC internally coated fuel assembly at an energy of 0.025 eV. The analysis of CrC internally coated fuel assembly nuclide inventories showed a relative increase in the isotopic concentration of some nuclides such as 239Pu and 241Pu, which was less than 1% for the cases considered. Comparing the calculated negative neutronics impacts, such as thermal neutron flux and fuel assembly operating time reductions, caused by CrC and Cr coating materials, the study revealed that the difference between these induced negative neutronics impacts is proportional to coating thickness. Therefore, CrC coatings will be most effective in terms of mitigating negative neutronics impacts when the specified coating thickness is large. [ABSTRACT FROM AUTHOR]

Zirconium alloys have outstanding nuclear properties and are irreplaceable key materials for the development of nuclear power systems. The preparation of coatings on the surface of zirconium alloys is one of the potential solutions of fuel cladding materials. In this paper, the research progress of the preparation and high-temperature oxidation resistance of metal coating for fuel cladding zirconium alloys is reviewed, which includes the preparation method, coating-type selection, high-temperature oxidation resistance, and high-temperature oxidation failure mechanism of different coatings. This review provides important references for the development of fuel cladding zirconium alloy surface coating technology. [ABSTRACT FROM AUTHOR]

The failure strain of a tube is a function of the biaxial strain ratio (axial strain/hoop strain) to which it is subjected. The relationship between failure strain and the strain ratio can be determined experimentally using expansion due to compression tests with a tensile load (EDCT), in which a ductile pellet placed inside the tube is compressed axially so it expands in diameter and imposes a hoop strain on the tube. At the same time, a tensile load on the ends of the tube creates an axial strain. This study investigates the capabilities and limitations of EDCT tests using two devices that allow experiments to be performed on a standard tensile testing machine. The first device applies an axial force on the ends of the sample, and the second device applies an axial displacement. Tests on zirconium alloy tubes confirmed that the failure strain is dependent on the strain ratio and the metallurgical state of the material. EDCT tests can produce a range of strain ratios, but there is an upper limit on the strain ratio that can be obtained, and it is dependent on the plastic behaviour of the sample and the friction conditions between the components. [ABSTRACT FROM AUTHOR]

The following work is motivated by the desire to devise an internal pressure test that can mimic a displacement-controlled loading scenario and demonstrate how to apply the multiaxial stress and strain data from the test to develop an elastic/plastic constitutive model for a thin-walled tubular component. This is achieved by conducting simultaneous measurements of tangential and axial strain during the pressure test and integrating these strain measures into a feedback loop with the pressure controller. It is shown how data from such a test can be used to develop a large mechanical property data set relevant to biaxial loading conditions. The data obtained have high confidence evidenced by their low variability and alignment with other literature studies. Additionally, data from these internal pressure tests combined with full-tube axial tensile tests allow for the derivation of the Hill anisotropic yield function. The developed Hill yield function is validated by comparing the plastic strain ratios from the full-tube tension tests and by comparing the predicted yield stress in the tangential direction with measured values from ring tension tests in a previous study. [ABSTRACT FROM AUTHOR]

Modern industry has become increasingly reliant on composite materials for a variety of applications. and the nuclear industry is no exception to this.Among the materials being researched as Enhanced Accident Tolerant Fuels, ceramic matrix composites such as SiC-fiber-reinforced SiC (Sicf/SiC) figure as some prime candidates due to their excellent high temperature performances. Sicf/SiC so far shows adequate nuclear, mechanical and chemical propertiex: still, the thermal properties need further investigation.The thermal behavior of a material ix an important factor for its performance as a nuclear fuel cladding, i.e. the first barrier encapsulating the fuel pellets. Many features determine the resulting properties of composite materials, such as matrix and fiber reinforcement properties and orientation, void fraction, and pore morphology. This study establishes a methodology to study the physical properties of composite materials and applies it to Sicf/SiC . A FEM model is uxed to characterize the thermal properties of a fundamental Sicf/SiC element. referred to as a "unit cell", with the objective of accurately predicting the thermal properties of this complex class of materials where experimental data is often difficult to obtain. The unit cell is built based on data acquired with high-resolution tomographic microscopy performed at the TOMCAT beamline of the Swiss Light Source. By using phase-retrieval prior to tomographic reconstruction, the pores, fibers and matrix that compose the material can be distinguished in the data analysis. The separated information is processed to obtain geometrical information about the individual pores and fibers, which is then used to parametrize them as cylindrical objects. This allows constructing a FEM model of a cubic unit cell that is used to extract the effective then mal properties of Sicf/SiC. The analysis scheme includes steady-state and dynamic thermal transport simulations, which yield directional effective thermal conductivity and diffusivity values, respectively. Both modes of analysis show isotropic thermal conductivity values in the range of 71 W/m/K at room temperature, more than three times that of currently employed nuclear cladding materials. Combining these results with the data on the larger structural features of the material will lead to realistic results on the macroscopic thermal properties. [ABSTRACT FROM AUTHOR]

An approach coupling experimental tests and numerical simulation of the pressure resistance welding (PRW) process is proposed for optimizing fuel cladding welds for the new generation of nuclear reactors. Several experimental welds were prepared by varying the dissipated energy, which accounts for the effect of electric current and welding time applied during the PRW process. A working zone, a function of both applied dissipated weld energy and plug-displacement, was then identified on the basis of the microscopy observations of the weld defects. In addition, the numerical approach, based on a 2D axisymmetric multi-physics finite element model, was developed to simulate the PRW process in a plug-tube configuration. The proposed model accounted for interactions between the electrical, thermal and mechanical phenomena and the electro-thermo-mechanical contact between the pieces and electrodes. Numerical simulations were first validated by comparison to experimental measurements, notably by comparing the plug-displacement and the size and position of the heat-affected zone (HAZ). They were then used to assess the effect of the applied parameters on the maximum temperature and cumulated plastic strain reached during welding and the effect of the welding force on the quality of the weld. According to the numerical computations, the maximum temperature reached in the weld remains well below the melting temperature. Changing the welding force implies also modifying the applied energy in order to maintain the quality of the welds. Applied to different plug and clad geometries, the proposed model was shown to be useful for optimizing the joint plane for such a welding configuration. [ABSTRACT FROM AUTHOR]

High-burnup Zr-based nuclear fuel claddings exhibit accelerated irradiation growth, corrosion and hydrogen pick-up, all correlated with the emergence of c-component dislocation loops. We made use of sub-nm-resolution atom probe tomography to characterize the nanoscale chemistry of c-loops in fuel cladding from boiling water reactor operation. We found segregation of Fe, Ni and Sn to dislocation lines and depletion of Sn and O inside the loops, resulting in nearly pure Zr islands. We also observed nucleation of suboxide inside one c-loop, pointing to a possible mechanism of accelerated in-reactor corrosion. Such Zr-islands might also promote hydride precipitation and associated degradation. [Display omitted] [ABSTRACT FROM AUTHOR]

Herein, a reliable diffusion bonding of Ti 3 Si(Al)C 2 ceramic is achieved by applying Au foil as an interlayer at 650 °C for 30 min with an axial pressure of 20 MPa. This novel method significantly decreases the bonding temperature, which is about 150 °C lower than the lowest bonding temperature from current research to the best of our knowledge. Maximum shear strength of 58 MPa is achieved at 650 °C among the bonding temperature range of 600 °C~800 °C. The microstructure evolution mechanism and the relationship between microstructure and mechanical property are discussed. The facile mutual diffusion of Au with de-intercalated Al and Si from Ti 3 Si(Al)C 2 is considered critical in achieving sound interfacial bonding. [ABSTRACT FROM AUTHOR]

Single-rod oxidation and quench experiments at very high temperatures in steam atmosphere were conducted with advanced, nuclear grade SiCf/SiC CMC cladding tube segments. A transient experiment was performed until severe local degradation of the sample at maximum temperature of approximately 1845 °C. The degradation was caused by complete consumption of the external CVD-SiC sealcoat, resulting in steam access to the fiber–matrix composite with less corrosion resistance. Approaching these very high temperatures was accompanied by accelerated gas release mainly of H2 and CO2, the formation of surface bubbles and white smoke. Three one-hour isothermal tests at 1700 °C in steam with final water flooding and one three-hour experiment with fast cool-down in Ar atmosphere were run under nominally identical conditions. All isothermally tested samples survived the tests without any macroscopic degradation. The mechanical performance of these quenched clad segments was not significantly affected, while maintaining a high capability to tolerate damages. Despite these harsh exposure conditions, load transfer between SiC fibers and matrix remained efficient, allowing the composites to accommodate deformation. [ABSTRACT FROM AUTHOR]

The pressurized water reactor (PWR) employs UO2 pellets as nuclear fuel, which are packed in zirconium alloy tubes called nuclear fuel cladding. In the manufacture of the nuclear fuel, machining scraps are generated which are not easily discarded as scraps because of its high cost. These zirconium nuclear alloys are very costly and are not produced in Brazil. In this work, novel methods to recycle Zircaloy scraps using vacuum induction melting and electron beam furnaces were used to obtain ingots. The cast ingots were subjected to thermal treatments and then chemically analyzed, followed by microstructural characterization, mechanical properties evaluation, and X-ray diffraction. The results indicated the feasibility of the processes for obtaining alloys for application in the nuclear area, chemical industry or materials for biological applications such as dental prostheses.

An approach coupling experimental tests and numerical simulation of the pressure resistance welding (PRW) process is proposed for optimizing fuel cladding welds for the new generation of nuclear reactors. Several experimental welds were prepared by varying the dissipated energy, which accounts for the effect of electric current and welding time applied during the PRW process. A working zone, a function of both applied dissipated weld energy and plug-displacement, was then identified on the basis of the microscopy observations of the weld defects. In addition, the numerical approach, based on a 2D axisymmetric multi-physics finite element model, was developed to simulate the PRW process in a plug-tube configuration. The proposed model accounted for interactions between the electrical, thermal and mechanical phenomena and the electro-thermo-mechanical contact between the pieces and electrodes. Numerical simulations were first validated by comparison to experimental measurements, notably by comparing the plug-displacement and the size and position of the heat-affected zone (HAZ). They were then used to assess the effect of the applied parameters on the maximum temperature and cumulated plastic strain reached during welding and the effect of the welding force on the quality of the weld. According to the numerical computations, the maximum temperature reached in the weld remains well below the melting temperature. Changing the welding force implies also modifying the applied energy in order to maintain the quality of the welds. Applied to different plug and clad geometries, the proposed model was shown to be useful for optimizing the joint plane for such a welding configuration.

The filament winding process is a competitive performing technology for nuclear fuel cladding tubes due to its high automation. The study of the yarn path on the mandrel surface is vital to design and produce the cladding tube with the desired mechanical properties, reducing manufacturing time and costs. The geodesic and semi-geodesic trajectories are used to create a 3D yarn path in this paper. A 3D yarn path optimization method based on the principle of minimum potential energy is proposed to simulate the overlap effect in accord with the real winding process. The finite element (FE) mesh based on the 3D yarn path has been used for the mechanical analysis of the cladding tube. The embedded region constraint is applied to define the interaction between the matrix mesh and the yarn mesh to model the meso-structure of the cladding tube. Based on the meso-scale FE model, the mechanical behavior of the wound SiCf/SiC nuclear fuel cladding tube is studied in detail. The results show that due to the neglect of the overlap effect, the conventional laminate model overestimates the cladding tube strength. The proposed meso-scale FE model can accurately predict the failure of the cladding tube. The results also confirm that the creation of a 3D yarn path and the derived meso-scale FE model, representing an accurate wound structure, are of importance to the prediction of the performance of the cladding tube.

A generalized multiscale (micro-macro) finite element (FE) model for SiC-fiber reinforced SiC-matrix ceramic (SiCf/SiC) nuclear fuel claddings is established. In the macro level, the solid mesh of braided preform, which can be tailored by machine settings (braid angle, yarn width, and so on), is generated based on the braiding process simulation using the dynamic FE-solver, hiring the contact constraints. The matrix mesh and the yarn mesh are integrated by the embedded region constraint, with which the meshing difficulties can be avoided. In the micro-UD model, the progressive damage of the ceramic matrix is modeled using the phase field method (PFM) and the fracture is captured by Mohr–Coulombs criterion, which are stable and efficient in the description of the brittle crack initiation, coalition, and branching. Based on this multiscale model, the mechanical behavior of the braided SiCf/SiC nuclear fuel cladding tube is studied in detail. The superiorities over the homogenized tube model are demonstrated, too.

Steady state flow boiling experiments were conducted on a technically smooth Inconel 625 tube with outer diameter 9.1 mm at inlet pressures 131, 220 and 323 kPa, inlet temperatures 62, 78 and 94 °C and approximately 400, 600 and 1000 kg/(m2.s) mass flow. Water of these parameters was entering into the vertically aligned annulus, where the uniformly heated tube was placed until the critical heat flux (CHF) appeared. The experimental data were compared to estimations of CHF by local PGT tube correlation and Groeneveld's look-up tables for tubes. The results imply that in the region of low pressure and low mass flux, the differences between calculations and experiments are substantial (more than 50% of CHF). The calculations further imply that look-up tables and tube correlations should be corrected to the annulus geometry. Here, the Doerffer's approach was chosen and led to a substantial enhancement of CHF estimation. Yet, a new correlation for the region of low pressure and flow is needed. [ABSTRACT FROM AUTHOR]

Zirconium alloy tubes used as nuclear fuel cladding are subject to oxidation and subsequent hydrogen pickup during their long service in commercial light water reactors. The hydrogen picked up in the cladding can precipitate as a brittle hydride rim feature on the cladding outer surface. To better understand the effect of hydride rims on the fracture behavior of stress-relieved zirconium alloy cladding tubes, a procedure to produce these rim-like structures has been developed and is described herein. Extensive characterization of the 'as hydrided' tubes is performed. The hydrogen charging apparatus consists of a tube furnace with a quartz chamber that is connected to a vacuum pump as well as a bottle of pure hydrogen gas. The charging station uses a static charge of hydrogen as opposed to a flowing gas. The chief advantage of this approach is the ability to monitor the pressure drop in the hydriding chamber and correlate this pressure drop to a known rate of hydrogen pickup in the cladding tube. It was found that the hydrogen partial pressure, metal temperature, and surface treatment all clearly played a role in whether a hydride rim was formed. Extensive characterization of the hydride rims shows they consist of needle like platelets of δ phase hydrides (ZrH 1.66) oriented in the circumferential direction with a radial spacing of several microns in a sandwich-like structure. [ABSTRACT FROM AUTHOR]

To understand the degradation behavior of light water reactor (LWR) fuel-cladding tubes under neutron irradiation, a detailed mechanism of hydrogen pickup related to the point defect formation (i.e., a-component and c-component dislocation loops) and to the dissolution of precipitates must be elucidated. In this study, 3.2 MeV Ni3+ ion irradiation was conducted on Zircaloy-2 samples at room temperature. Thermal desorption spectroscopy is used to evaluate the deuterium desorption with and without Ni3+ ion irradiation. A conventional transmission electron microscope and a spherical aberration-corrected high-resolution analytical electron microscope are used to observe the microstructure. The experimental results indicate that radiation-induced dislocation loops and hydrides form in Zircaloy-2 and act as major trapping sites at lower (400–600 °C) and higher (700–900 °C)-temperature regions, respectively. These results show that the detailed microstructural changes related to the hydrogen pickup at the defect sinks formed by irradiation are necessary for the degradation of LWR fuel-cladding tubes during operation.

The Reversed Field Pinch (RFP) configuration looks to be an attractive option for fusion-fission hybrid reactors: the toroidal magnetic systems would be made of copper coils instead of more expensive superconductive magnets; fusion conditions could be reached by ohmic heating only, therefore additional heating systems would not be required; the fission blanket could be located in the most external part of the torus thus facilitating maintenance operations. The paper aims at assessing the potentialities, such as fuel fertilization and/or nuclear waste transmutation and electricity production of a hybrid reactor with a RFP fusion core (R = 6 m, a = 1) whose conceptual design and plasma performances are based on RFX-mod, the largest RFP experiment currently in operation. Fusion conditions can be reached by heating a D-T plasma up to 9.6 keV by ohmic heating, generated by a 20 MA plasma current induced and sustained by flux swing only. The neutron flux (2.1 × 1013 fast neutron/cm2/s) is used to breed tritium in both the inner and outer blanket sections and induce fission reactions in dedicated areas in the external blanket section where Pu + MA (60%)-Zr (40%) rods are located. Both neutronic and safety analyses corroborate the viability of a FFH reactor with a RFP core. [ABSTRACT FROM AUTHOR]

Nickel-chromium coatings were deposited on Zr 1Nb alloy using magnetron sputtering systems with «hot» Ni and cooled Cr targets. The effect of coating composition on high-temperature oxidation resistance and hydrogen uptake of Zr 1Nb was studied. Hydrogen uptake of the alloy was measured in situ under gas-phase hydrogenation at 633 K. High-temperature oxidation was performed in air atmosphere at 1173–1373 K for 20 min. It was shown that the coating with high Ni content (83 at.%) drastically increases hydrogen uptake of the Zr 1Nb alloy and demonstrates low oxidation resistance even at 1173 K. The coatings with Cr content ≥45 at.% have low hydrogen permeability which reduces the rate of hydrogen uptake of the alloy. The oxidation resistance of the Ni Cr coatings increases with Cr content in the as-deposited coatings. The pure Cr coating exhibits the best oxidation resistance: only 8 μm-thick oxide layer was observed. There is also found the intensive diffusion of nickel into the alloy during high-temperature oxidation of the samples coated by Ni Cr films with 55 and 17 at.% Ni. The as-deposited Ni Cr coatings are less brittle than the pure Cr coating, but their mechanical properties degrade stronger than for the Cr coating after the oxidation test. • Ni Cr coatings were deposited on Zr 1Nb alloy by hot target sputtering. • Cr-rich coatings demonstrate higher oxidation resistance. • Ni diffuses into Zr at high temperatures when 55 and 17 at.% Ni in the coatings. • Pure Cr film has better mechanical properties after high-temperature oxidation. [ABSTRACT FROM AUTHOR]

Development and characterization of clad-to-plug resistance upset welds is presented for two oxide dispersion strengthened steels candidates for sodium-cooled fast reactors fuel cladding: 9Cr ferrito-martensitic steel and 14Cr ferritic steel. A comparative approach is adopted regarding these two materials implying weldability studies and weld mechanical strength tests. A special attention is paid on welding metallurgical effects on the specific microstructure of these materials. Among others, some major grain property modifications by dynamic recrystallization and a slight modification of precipitate sizes and volume fraction are found and discussed. [ABSTRACT FROM AUTHOR]

Abstract We explore the distribution, morphology and structure of zirconium hydrides formed using different cooling rates through the solid state Zr+[H] → Zr + hydride transus, in fine and blocky alpha Zircaloy-4. We observe that cooling rate and grain size control the phase and distribution of hydrides formed. The blocky alpha (coarse grain, > 200 μm) Zircaloy-4, has a smaller grain boundary area to grain volume ratio and this significantly affects nucleation and growth of hydrides as compared to fine grain size (∼11 μm) material. [ABSTRACT FROM AUTHOR]

Abstract Ring compression tests on un-irradiated nuclear fuel rod cladding were conducted to analyze its mechanical behavior under pinching loads using appropriate boundary conditions, such as the consideration of fuel pellet support. The tested specimens included as-fabricated and artificially hydrogen-charged Zircaloy-4 cladding samples with a hydrogen content in a range of 285–470 ppm. Part of the samples were subjected to radial hydride treatment including a peak cladding hoop stress of 79 MPa and a peak cladding temperature of 400 °C to simulate SNF vacuum drying and to investigate treatment effects on the cladding response. Half of the tested rings were loaded with 20 mm long stainless steel pellets to analyze the impact of fuel pellet presence on the cladding loading capacity. The pellet-cladding gap width ranged from 60 to 180 μm. The other half of the rings was tested in empty state. The load-displacement curves obtained from ring compression tests conducted at room temperature on as-fabricated cladding exhibited a highly ductile material behavior. The presence of hydrogen in the cladding significantly embrittled the material, but unexpectedly, radial hydride treatment increased the cladding ductility. The ring compression tests conducted under pellet presence did not induce cladding cracking, even under extreme pinching loads. The results indicate that hydride-related material embrittlement likely does not cause nuclear fuel cladding failure when subjected to pinching loads, under the premise that a fuel pellet provides sufficient support to the cladding. Highlights • Strain capacity reduces by one order of magnitude due to hydrogen charging. • Radial hydride treatment at 80 MPa does not cause significant radial hydrides. • Loading capacity reduces by about 20–30% due to hydrogen charging. • Fuel pellet support increases loading capacity by several orders of magnitude. [ABSTRACT FROM AUTHOR]

Micro-scale damage with a topographical contrast has been observed in cold-worked Nb-stabilised 20%Cr-25%Ni stainless steel, following irradiation with 2.2 MeV protons at 400 °C and a dose rate of ∼10 −5  dpa/s. After an irradiation dose of 3 and 5 dpa, microstructural changes were found to a depth of 22 μm below the irradiated surface, coincident with the estimated mean range of protons in the material as predicted by the TRIM code. Large variations of intragranular mis-orientations, in combination with a depletion of chromium at grain boundaries, were observed in the proton-irradiated layer. This was accompanied by an increase in nano-hardness, which was linked to a partial transformation of tangled dislocation networks into dislocation loops. Such visible microstructural changes have not been reported before in the absence of post-irradiation annealing treatments. [ABSTRACT FROM AUTHOR]

We report the development of a cone wedge ring expansion test method to determine the tensile hoop properties of irradiated nuclear fuel cladding in a hot cell. A closed-form solution was developed to explicitly relate the hoop stress σ θ to the axial piston loading using the scale factor χ and the shear stress correction factor α, where the χ and α factors are used to convert the applied piston load and the radial dilatation of the ring into the tensile hoop stress-strain curve. The applicability of this new test method was demonstrated by successfully evaluating ring specimens of Zr-4 alloy to failure. The ring specimens deformed uniformly in the axial and circumferential directions until onset of necking. The χ and α factors were successfully determined by using a finite-element stress analysis that benchmarked with applied test load and total work to account for frictional forces and the associated energies dissipation between the different sliding components of the testing system. The tensile hoop strength values obtained using the proposed cone-wedge-ring-expansion test method were 10% higher than values obtained from the tensile evaluation of the cladding material along the axial direction. The variability in results was found to be within the range reported in the literature. [ABSTRACT FROM AUTHOR]

Chemical vapor deposition (CVD) techniques have been evaluated for fabrication of free-standing 0.25 mm thick molybdenum tubes with the end goal of nuclear fuel cladding applications. In order to produce tubes with the wall thickness and microstructures desirable for this application, long deposition durations on the order of 50 h with slow deposition rates were employed. A standard CVD method, involving molybdenum pentachloride reduction by hydrogen, as well as a fluidized-bed CVD (FBCVD) method was applied towards these objectives. Characterization of the tubes produced in this manner revealed regions of material with fine grain microstructure and wall thickness suitable for fuel cladding applications, but lacking necessary uniformity across the length of the tubes. A path forward for the production of freestanding molybdenum tubes that possess the desired properties across their entire length has been identified and can be accomplished by future optimization of the deposition system. [ABSTRACT FROM AUTHOR]

In an effort to develop coatings to increase the oxidation resistance of nuclear fuel cladding during a loss-of-coolant-accident (LOCA), undoped and ytterbium-doped titanium aluminum nitride (Ti x Al 1-x N) coatings were deposited onto ZIRLO® and grade-5 titanium substrates using a hybrid coating technique incorporating both resistance evaporation and cathodic arc physical vapor deposition (CA-PVD). The coating systems consisted of a titanium bond layer, an undoped TiAlN layer, and an outer TiAlN layer doped with varying levels of ytterbium (0.44 to 33.24 at.%). The coated materials were subjected to high temperature atmospheric oxidation testing and differential scanning calorimetry (DSC) to evaluate the effect of different Yb dopant concentrations on the Ti x Al 1-x N coating oxidation resistance. The results showed that the coatings evaluated in this study protected the tubular ZIRLO® cladding material from oxidation when exposed to air at temperatures up to 1200 °C. DSC data showed that the corrosion resistance of coatings with Yb dopant concentrations of 0.44 and 0.64 at.% was better than that of undoped Ti x Al 1-x N coatings and of Ti x Al 1-x N coatings with Yb concentrations ≥ 4.78 at.%. Yb doping and the 0.44 and 0.64 at.% level also resulted in a lower rate of oxide scale growth and improved scale adherence, which further slowed oxidation kinetics. Peak oxidation resistance was observed in the TiAlN coating with a 0.44 at.% ytterbium dopant concentration. Ytterbium concentrations of 4.78 at.% and greater led to accelerated oxidation rates for the testing conditions studied compared to undoped TiAlN. These results show that ytterbium doped coatings are possible candidates for high temperature resistance of accident tolerant fuel coatings. [ABSTRACT FROM AUTHOR]

Optical properties of Zr and its oxide have been measured on the surface of nuclear fuel cladding tubes. It has been shown that ellipsometry with focusing can routinely be used to measure thin layers and surface properties on Zr tubes with a diameter as small as 9.1 mm. Multi-sample and depth profiling models have been used to determine reference dielectric function spectra for both the Zr substrate and its oxide. Temporal behavior of the oxide thickness has been measured for oxidation temperatures of 600 °C and 800 °C. A vertical inhomogeneity of the oxide properties has been found by the optical measurements as well as by depth-profiling X-ray photoelectron spectroscopy investigations that revealed the formation of sub-oxides at the interface region of Zr and its surface oxide. [ABSTRACT FROM AUTHOR]

In this work, the hoop fracture toughness of ZIRLO ® fuel cladding is calculated as a function of three parameters: hydrogen concentration, temperature and displacement rate. To this end, pre-hydrided samples with nominal hydrogen concentrations of 0 (as-received), 150, 250, 500, 1200 and 2000 ppm were prepared. Hydrogen was precipitated as zirconium hydrides in the shape of platelets oriented along the hoop direction. Ring Compression Tests (RCTs) were conducted at three temperatures (20, 135 and 300 °C) and two displacement rates (0.5 and 100 mm/min). A new method has been proposed in this paper which allows the determination of fracture toughness from ring compression tests. The proposed method combines the experimental results, the cohesive crack model, finite elements simulations, numerical calculations and non-linear optimization techniques. The parameters of the cohesive crack model were calculated by minimizing the difference between the experimental data and the numerical results. An almost perfect fitting of the experimental results is achieved by this method. In addition, an estimation of the error in the calculated fracture toughness is also provided. [ABSTRACT FROM AUTHOR]

The geometry of the nuclear fuel cladding (thin-walled tube) makes it difficult to obtain its hoop mechanical properties. A new procedure is devised to obtain the constitutive equation of nuclear fuel cladding along the hoop direction from ring compression tests. The method combines experimental results, finite element simulations and an original iterative algorithm to adjust the experimental data. The process is successfully applied to unirradiated pre-hydrided ZIRLO nuclear fuel cladding, tested at three temperatures (20, 135 and 300 °C) with hydrogen contents (0, 150, 250, 500, 1200 and 2000 ppm). The stress-strain curves were obtained for each configuration with an excellent agreement between the numerical results (based on back calculation of the obtained constitutive equation) and the experimental data. The stress-strain curves calculated show that the mechanical properties do not depend strongly on hydrogen concentration, only a small ductility decrease with the hydrogen concentration was observed. The cladding shows a light strain hardening which is similar for the samples tested at 20 and 135 °C and does not depend on the hydrogen concentration. However, at 300 °C, the samples with the highest hydrogen concentrations (1200 and 2000 ppm) present a behavior that is close to an elastic-perfectly plastic material. [ABSTRACT FROM AUTHOR]

It is well known that the presence of hydrides in nuclear fuel cladding may reduce its mechanical and fracture properties. This situation may be worsened as a consequence of the formation of hydride blisters. These blisters are zones with an extremely high hydrogen concentration and they are usually associated to the oxide spalling which may occur at the outer surface of the cladding. In this work, a method which allows us to reproduce, in a reliable way, hydride blisters in the laboratory has been devised. Depth-sensing indentation tests with a spherical indenter were conducted on a hydride blister produced in the laboratory with the aim of measuring its mechanical behaviour. The plastic stress-strain curve of the hydride blister was calculated for first time by combining depth-sensing indentation tests results with an iterative algorithm using finite element simulations. The algorithm employed reduces, in each iteration, the differences between the numerical and the experimental results by modifying the stress-strain curve. In this way, an almost perfect adjustment of the experimental data was achieved after several iterations. The calculation of the constitutive equation of the blister from nanoindentation tests, may involve a lack of uniqueness. To evaluate it, a method based on the optimization of parameters of analytical equations has been proposed in this paper. An estimation of the error which involves this method is also provided. [ABSTRACT FROM AUTHOR]

In an attempt to develop a nuclear fuel cladding that is more tolerant to loss-of-coolant-accidents (LOCA), ceramic coatings were deposited onto a ZIRLO™ 1 1 ZIRLO is a trademark of Westinghouse Electric Co. substrate by cathodic arc physical vapor deposition (CA-PVD). The coatings consisted of either Ti 1 – x Al x N or TiN ceramic monolithic layers with a titanium bond coating layer as the interlayer between the ceramic coating and the ZIRLO™ substrate to improve coating adhesion. Several coating deposition trials were performed investigating the effects of bond coating thickness (200–800 nm), ceramic coating thickness (4, 8 and 12 μm), substrate surface roughness prior to deposition, and select coating deposition processing parameters, such as nitrogen partial pressure and substrate bias, in order to optimize the coating durability in a corrosion environment. Corrosion tests were performed in static pure water at 360 °C and saturation pressure (18.7 MPa) for 3 days. The optimized nitride-based ceramic coatings survived the autoclave test exposure showing very low weight gain of 1–5 mg/dm 2 compared to the uncoated ZIRLO™ samples which showed an average weight gain of 14.4 mg/dm 2 . Post-corrosion exposure analytical characterization showed that aluminum depletion occurred in the TiAlN coated samples during the autoclave corrosion test, which led to the formation of the boehmite phase that degraded the corrosion durability of some of the TiAlN samples. However, by eliminating the aluminum content and depositing TiN, the boehmite phase was prevented from forming. Best results in TiAlN coated samples were obtained with 600 nm Ti bond coating thickness, 12 µm coating thickness and 0.25 µm substrate surface roughness (E14). Results are discussed in terms of the capability of TiN and Ti 1 – x Al x N coatings to improve the high temperature corrosion resistance and oxidation resistance of zirconium alloy cladding. [ABSTRACT FROM AUTHOR]

Nuclear grade silicon carbide fiber (SiC f ) reinforced silicon carbide matrix (SiC m ) composite is a promising candidate material for accident tolerance fuel (ATF) cladding. A major challenge is ensuring the mechanical robustness of the ceramic cladding under accident conditions. In this work the high temperature mechanical response of a SiC f –SiC m composite tubing is studied using a novel thermo-mechanical test method. A solid surrogate tube is placed within and bonded to the SiC f –SiC m sample tube using a ceramic adhesive. The bonded tube pair is heated from the center using a ceramic glower. During testing, the outer surface temperature of the SiC sample tube rises up to 1274 K, and a steep temperature gradient develops through the thickness of the tube pair. Due to CTE mismatch and the temperature gradient, the solid surrogate tube induces high tensile stress in the SiC sample. During testing, 3D digital image correlation (DIC) method is used to map the strains on the outer surface of the SiC-composite, and acoustic emissions (AE) are monitored to detect the onset and progress of material damage. The thermo-mechanical behavior of SiC-composite sample is compared with that of monolithic SiC samples. Finite element models are developed to estimate stress–strain distribution within the tube assembly. Model predicted surface strain matches the measured surface strain using the DIC method. AE activities indicated a progressive damage process for SiC f –SiC m composite samples. For the composites tested in this study, the threshold mechanical hoop strain for matrix micro-cracking to initiate in SiC f –SiC m sample is found to be ∼300 microstrain. [ABSTRACT FROM AUTHOR]

Abstract: The aim of the present work is to synthesize high strength monolithic SiC tubes to improve the imperviousness of a SiC/SiC composite structure. A few hundred micrometer-thick tubular coatings were produced by chemical vapor deposition (CVD) at atmospheric pressure from CH3SiHCl2/Ar/H2 mixtures. The CVD-SiC tubes were obtained by deposition on the inner walls of a SiO2-tube substrate, previously coated with a pyrocarbon interfacial layer to promote delamination. A continuous deposition process was developed to allow the realization of relatively long CVD-SiC tubes by sliding the heating system along the substrate. The chemical composition and the microstructure of the tubes were studied by electron probe microanalysis, Raman spectroscopy and scanning electron microscopy. The deposition rate, composition and microstructure of the CVD-SiC coatings were investigated as a function of the substrate temperature and the gas flow rates. A Fourier transformed infrared (FTIR) spectroscopy analysis was carried out at the reactor outlet to characterize the gas phase reactions. The FTIR analysis of pure species from the Si–C–Cl–H system as well as ab initio calculations at the density functional theory (DFT) level allowed the assignment of the main IR features in the experimental spectra and the quantitative analysis of the complex gas mixture. This study has led to the proposal of a simplified dichloromethylsilane decomposition scheme which is consistent with the influence of the CVD parameters on the nature of the gas phase and the coating. The deposition rate, the Si/C atomic ratio, the SiC crystalline state and the surface morphology are indeed strongly related to the CH3SiHCl2 decomposition rate and the further progress of homogeneous reactions. [Copyright &y& Elsevier]