Search

This study developed a probabilistic determination model with respect to cladding high-temperature burst conditions based on the Bayesian statistical method to reasonably evaluate fuel behaviors under loss-of-coolant accident conditions, including fuel fragmentation, relocation, and dispersal. The candidate models (correlations) were based on the widely accepted empirical model established based on nonirradiated fuel cladding data. Explanatory variables were added to improve the applicability of these models with respect to irradiated materials and generalization performance. The posterior predictive distribution of each candidate model was evaluated using Bayesian estimation comprising 238 sets of high-temperature burst test data. The generalization performance was evaluated using information criteria. The results of model evaluation showed improved predictive performance by considering the effect of hydrogen content. A comparison with burnup as an alternative explanatory variable confirmed that hydrogen content was the better parameter and other burnup-associated effects, such as irradiation hardening of the metal matrix and oxide growth (reduction of the metal matrix), were less dominant under burst conditions. [ABSTRACT FROM AUTHOR]

Hydrogen diffuses in zirconium alloys in response to gradients in hydrogen concentration, temperature, and stress. This essay discusses the results of several evaluations of the coefficient describing the effects of temperature gradients called the heat of transport. The values distinguish between when hydrogen is in solution, (25 ± 3) kJ/mol, and when stable hydrides are present, (116 ± 17) kJ/mol. [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]

Coatings of metal specimens are known to have an impact on hydrogen gettering (hydrogen absorption). The coating can have one or more functions, such as enhancing gettering, preventing gettering and/or preventing oxidation of the metal substrate. It is known that contaminants and surface texture can impact hydrogen gettering/absorption performance, but has not previously been thoroughly explored. This study evaluated the role of different post-plating heat treatments of nickel-plated zircaloy-4 getters (NPGs) and the role of the heat treatments on gettering rates, surface composition and texture. Nickel plating is applied to prevent oxidation of the Zircaloy-4 surface and also enhances gettering. The nickel plating must be heat treated before desirable gettering can occur. Our NPG getters with historically known satisfying performance were pre-heat treated in air followed by activation heat treatment in a vacuum at a higher temperature. In this study, we were interested in finding out if both heat treatment steps were necessary to obtain a desirable gettering performance, or if one step could be omitted. XPS analysis showed that if the nickel surface is not heat treated before bonding the nickel to the zirconium in the activation step, there will be carbon contaminants on the surface, which significantly reduces gettering. We studied the texture of Zircaloy-4 using SEM/EBSD to compare NPGs with both heat treatment steps with NPGs that had no post-plating heat treatment to learn if the degree of cold work could be impacted by the heat treatment steps. We did not observe any differences in texture between them. We measured gettering rates of both pretreated and activated NPGs and NPGs that had been activated without first being pre-heat treated. We found that the NPGs without the first post-plating heating step had up to a seven times slower gettering rate and obtained higher plateau pressures due to the contaminated surface. Thus, the pre-heat treatment in air before activation is necessary to avoid slower gettering rates and higher plateau pressures. [ABSTRACT FROM AUTHOR]

Carburization of cladding materials has long been a concern for the nuclear industry and has led to the restricted use of high-thermal conductivity fuels such as uranium carbides. With the rise of small modular reactors (SMRs) that frequently implement a graphite core-block, carburization of reactor components is once more in the foreground as a potential failure mechanism. To ensure commercial viability for SMRs, neutron-friendly cladding materials such as Zr-based alloys are required. In this work, the carburization kinetics of Zircaloy-4 (Zry-4), for the temperature range 1073–1673 K (covering typical operating temperatures and off-normal scenarios) are established. The following Arrhenius relationship for the parabolic constant describing ZrC growth is derived: Kp (in μm2/s) = 609.35 exp(−1.505 × 105/RT)). Overall, the ZrC growth is sluggish below 1473 K which is within the operational temperature range of SMRs. In all cases the ZrC that forms from solid state reaction is hypo-stoichiometric, as confirmed through XRD. The hardness and elastic modulus of carburized Zry-4 are also examined and it is shown that despite the formation of a ZrC layer, C ingress in the Zry-4 bulk does not impact the mechanical response after carburization at 1073 K and 1473 K for 96 h. [ABSTRACT FROM AUTHOR]

As a widely used nuclear structural material, zirconium (Zr) alloys are susceptible to brittle hydride-induced cracking. The intrinsic inability to arrest this cracking tendency in Zr poses a significant threat to the service safety of Zr cladding tubes. Here, we propose a microstructural design strategy, by introducing a hierarchical nanolayered duplex-phase structure in the Zr-2.5Nb alloy, to effectively inhibit the propagation of cracks. An unprecedented fracture toughness of K JIc ∼ 165 MPa·m1/2, corresponding to J Ic ∼ 256 kJ/m2, is achieved due to the uniform and dense plastic deformation that develops ahead of the crack tip, which is uncharacteristic of Zr-based materials. Our analysis reveals that the effective crack tip blunting occurs because the high density of multiply oriented α/β interfaces in the hierarchical Zr-2.5Nb alloy stimulates ample dislocations and facilitates deformation twin nucleation, both of which are challenging to activate in conventional Zr-based materials at room temperature. The exceptional fracture toughness enabled by the hierarchical nanolayered structure provides a new pathway to designing damage-tolerant hexagonal metals for safety-critical applications. [Display omitted] [ABSTRACT FROM AUTHOR]

In this study, an electrochemical method for decladding used nuclear fuel (UNF) was investigated by electrochemically oxidizing pieces of zircaloy in a LiF–NaF–ZrF 4 (26-37-37 mol%) salt. Electrochemically decladding UNF allows for zirconium (Zr) to be separated from fission products, such that Zr no longer needs to be stored as high-level waste (HLW). As zircaloy samples were oxidized, both the Zr deposited on the cathode and in the bulk salt were checked for the buildup of elements other than Zr to vet this process. The current efficiency of the system was also found to be over 90% and tested samples of zircaloy were oxidized entirely, showing the current used is not wasted on other reactions than oxidizing zircaloy and that the process works to completion. • Electrochemical oxidation of zircaloy is possible. • The current efficiency of oxidation is over 90%. • After dissolving the cladding ZrF 4 in the salt will react with the UO 2 pellets. • This is the first step in processing SNF in fluoride salts. [ABSTRACT FROM AUTHOR]

Second phase Zr(Fe,Cr) 2 precipitates (SPPs) in Zircaloys can undergo a number of structural changes in-reactor including amorphization and dissolution. These changes have a direct effect on the mechanical and corrosion properties of the alloy, so understanding them is critical to improving performance. This paper describes a classical model that predicts SPP evolution over the whole range of observed behaviour, from complete amorphization at low temperature through formation of an amorphous rim, iron loss, and dissolution at intermediate temperatures, and a return to stability at high temperatures. A single set of assumptions is used, with the different regimes of behaviour arising from a natural competition between mechanisms. The model suggests that there is significant overlap in regimes, and for the same temperatures formation of an amorphous rim and SPP dissolution can occur simultaneously. The most interesting behaviour is seen at intermediate temperatures (520–620 K), where there is a transition from irradiation to thermally dominated processes. The model qualitatively captures experimental observations, but for more accurate quantitative predictions a number of unknown parameters require further study. These include the effect of irradiation on the solubility and diffusion rates of alloying elements. [ABSTRACT FROM AUTHOR]

• PEO mimics breakaway corrosion in Zircaloy-2 on prolonged exposure to coolant. • At 300V- 400 V, Mg modified the oxide on Zircaloy-2 lowering its corrosion rate. • Presence of Mg2+ increased R ox , R pit , t-ZrO 2 content, lowered defect densities. • Mg may block the connected porosity in the oxide impeding oxygen ion transport. • Under a bias of 500 V, even in the presence of Mg, the oxide film was highly porous. General corrosion of nuclear reactor in core material like Zircaloy-2 and breakaway corrosion in particular are of great importance in ensuring its smooth long term operation. Metal ions like Mg2+ added to the coolant, to mitigate corrosion of other structural materials like Carbon steel, can get incorporated in the corrosion product oxide on Zircaloy-2 and alter its corrosion behavior. Plasma Electrolytic Oxidation (PEO) method was employed to investigate the effect of Mg2+ ions during breakaway corrosion of Zircaloy-2. The main objective was to understand the morphology, protectiveness, semiconducting properties of the Mg modified oxide films on Zircaloy-2 surface. DC potentials 300, 400 and 500 V were used to accelerate corrosion kinetics and form oxide mimicking the breakaway corrosion regime. Borate buffer (pH 9.8) was used as the electrolyte, and Mg acetate was added as Mg source for probing the effect of Mg2+ ions. Oxide morphology depended largely on the formation potentials. The presence of magnesium resulted in thinner oxides with lower defect densities. GIXRD of the oxide showed stabilization of tetragonal phase in presence of Mg at 300 and 400 V. Oxide resistance and charge transfer resistances measured by EIS were observed to be higher due to Mg incorporation at these potentials. Mg addition facilitated formation of t-ZrO 2. The electrochemical measurements suggested that the presence of Mg in ZrO 2 could reduce the connected porosity in the oxide thereby stifling the diffusion paths. These factors lead to improved corrosion protectiveness of the oxide formed in the presence of Mg. [ABSTRACT FROM AUTHOR]

We report results from atom probe tomography (APT) experiments capturing Zr(Fe,Cr) 2 second phase particles (SPPs) in Zircaloy-2-type fuel cladding after reactor operation. In light of recent reports of H trapping around SPPs, we assess the feasibility of H analysis in modern commercial atom probe instruments on this system. To this end we employed voltage and laser pulsing APT on specimens prepared by focused ion beam (FIB) at room and cryogenic temperature. Room temperature FIB caused transformation of the α-Zr matrix into δ-hydride, but left SPPs mostly unaffected. This indicates that α-Zr has a higher affinity for H than SPPs. However, even under optimized conditions, we were not able to find evidence for H trapping near SPPs located within the α-Zr matrix in cryogenically FIB sharpened specimens, where no hydride transformation occurs. [Display omitted] [ABSTRACT FROM AUTHOR]

This work evaluates the stress relaxation behavior of textured Zircaloy-4 cladding to understand how mechanical anisotropy influences pellet-cladding interactions. Uniaxial and biaxial stress relaxation tests are performed using full-tube axial tension and internal pressurization, respectively, aiming to achieve 0.25 %, 1 %, and 2 % equivalent strains in the cladding samples at a temperature of 300 °C. Internal pressure relaxation test results display enhanced stress relaxation compared to axial testing results, particularly for samples loaded beyond yield. Results of electron backscatter diffraction indicate increased deformation microstructure during loading and increased strain homogenization and recovery during relaxation for samples loaded via internal pressurization. Analysis indicates that the increased production and activity of basal 〈 a 〉 dislocations play a significant role in the enhanced relaxation measured in samples subjected to internal pressurization. [ABSTRACT FROM AUTHOR]

This work evaluates deformation activity in textured zircaloy-4 cladding with the aim of elucidating the mechanisms that govern mechanical behavior across a temperature range anticipated in reactor service. In particular, the dependence of mechanical anisotropy on temperature and stress state (ratio of applied stresses) is examined via uniaxial and biaxial tensile experiments at room temperature and 400 °C. The mechanical behavior is interpreted based on micro-texture and dominant slip system activity. Thermal activation of basal systems is found as a key mechanism affecting mechanical behavior across the temperature range. The biaxial stress states produce resolved shear stresses contrasting to uniaxial stress states. Mechanical testing and implications for reactor pellet-cladding mechanical interaction stress conditions are discussed. [Display omitted] [ABSTRACT FROM AUTHOR]

Presented here is the evolution of dynamic recrystallization (DRX) with strain in Zircaloy-2 deformed at 800 °C and 10–1 s−1. The flow stress showed a peak followed by softening to steady state. Specimens were deformed to predetermined strains and their microstructure was characterized using electron back scatter diffraction. DRX grains were identified using the criteria of grain orientation spread. The grain size decreased with strain up to a strain of 0.6, beyond which grain size distribution remained very similar. From grain reference orientation deviation and kernel average misorientation maps, it was seen that grain boundaries move into regions of higher dislocation density, suggesting that during hot deformation, Zircaloy-2 underwent discontinuous dynamic recrystallization. Compared to hot deformed Zr, it was observed that the final grain size of Zircaloy-2 was smaller. The final texture showed the basal poles to be aligned nearly parallel to the compression axis. [ABSTRACT FROM AUTHOR]

Diffusion bonding of stainless steel 304 L and zircaloy‐4 (Zr‐4) in solid state is conducted using Ni and Ta as the hybrid interlayer at the bonding temperature of 850−1000 °C for the bonding time of 30 min under the bonding pressure of 6 MPa in vacuum. The effect of the bonding temperature on the microstructure and the mechanical properties of the joints is investigated. The microstructure analysis is conducted by a scanning electron microscope, energy dispersive spectroscopy, and X‐ray diffraction. The mechanical properties of the joints are measured by a shear test and microhardness test. The results show that Ni3Ta intermetallic compounds are formed at the Ni/Ta interface, and the solid solution of Zr and Ta appears at the Ta/Zr‐4 interface. In the meantime, the maximum shear strength reaches 194 MPa and the highest hardness value across the interface of joints is 356 HV in the Ni3Ta layer. [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.

Coatings of metal specimens are known to have an impact on hydrogen gettering (hydrogen absorption). The coating can have one or more functions, such as enhancing gettering, preventing gettering and/or preventing oxidation of the metal substrate. It is known that contaminants and surface texture can impact hydrogen gettering/absorption performance, but has not previously been thoroughly explored. This study evaluated the role of different post-plating heat treatments of nickel-plated zircaloy-4 getters (NPGs) and the role of the heat treatments on gettering rates, surface composition and texture. Nickel plating is applied to prevent oxidation of the Zircaloy-4 surface and also enhances gettering. The nickel plating must be heat treated before desirable gettering can occur. Our NPG getters with historically known satisfying performance were pre-heat treated in air followed by activation heat treatment in a vacuum at a higher temperature. In this study, we were interested in finding out if both heat treatment steps were necessary to obtain a desirable gettering performance, or if one step could be omitted. XPS analysis showed that if the nickel surface is not heat treated before bonding the nickel to the zirconium in the activation step, there will be carbon contaminants on the surface, which significantly reduces gettering. We studied the texture of Zircaloy-4 using SEM/EBSD to compare NPGs with both heat treatment steps with NPGs that had no post-plating heat treatment to learn if the degree of cold work could be impacted by the heat treatment steps. We did not observe any differences in texture between them. We measured gettering rates of both pretreated and activated NPGs and NPGs that had been activated without first being pre-heat treated. We found that the NPGs without the first post-plating heating step had up to a seven times slower gettering rate and obtained higher plateau pressures due to the contaminated surface. Thus, the pre-heat treatment in air before activation is necessary to avoid slower gettering rates and higher plateau pressures.

Carburization of cladding materials has long been a concern for the nuclear industry and has led to the restricted use of high-thermal conductivity fuels such as uranium carbides. With the rise of small modular reactors (SMRs) that frequently implement a graphite core-block, carburization of reactor components is once more in the foreground as a potential failure mechanism. To ensure commercial viability for SMRs, neutron-friendly cladding materials such as Zr-based alloys are required. In this work, the carburization kinetics of Zircaloy-4 (Zry-4), for the temperature range 1073–1673 K (covering typical operating temperatures and off-normal scenarios) are established. The following Arrhenius relationship for the parabolic constant describing ZrC growth is derived: Kp (in μm2/s) = 609.35 exp(−1.505 × 105/RT)). Overall, the ZrC growth is sluggish below 1473 K which is within the operational temperature range of SMRs. In all cases the ZrC that forms from solid state reaction is hypo-stoichiometric, as confirmed through XRD. The hardness and elastic modulus of carburized Zry-4 are also examined and it is shown that despite the formation of a ZrC layer, C ingress in the Zry-4 bulk does not impact the mechanical response after carburization at 1073 K and 1473 K for 96 h.

Understanding fuel system behavior during postulated loss-of-coolant accidents is pertinent for continued safe and efficient operation of light water reactors, particularly as higher burnups are being pursued and safety margins re-evaluated. Conventional mechanical models for the incumbent Zr alloys typically rely on the assumption that steady-state creep is the dominant fuel cladding response during transient accident conditions. To investigate this assumption, simulated accident burst testing was performed on Zircaloy-4 claddings with balloon behavior measured in-situ. Two distinct loading conditions were utilized during burst testing: (1) constant-gas-inventory where pressure was allowed to increase with temperature and (2) constant pressure. In-situ strains and strain rates were measured via 2-dimensional digital image correlation techniques and synchronized with temperature to determine deformation dependencies. The temperature dependence of strain rate was characterized by a two segment Arrhenius relationship, with a distinct transition between the high and low temperature/strain regimes. The average activation energy of the lower temperature/strain regime was 328 ± 25 kJ/mol, in agreement with the ∼320 kJ/mol used for conventional LOCA models. However, the higher temperature/strain segment, which encompassed most of ballooning, showed increased activation energies as well as a dependence on whether the burst region was in view. For tests that burst away from the camera view, the average high temperature/strain segment activation energy was 635 ± 150 kJ/mol. For samples where the rupture opening formed in view, the average activation energy was 1015 ± 179 kJ/mol. This observed shift in temperature dependence indicates a transition in deformation mechanism at the end of life, possibly to time independent failure mechanisms, which has not yet been visualized in the literature for Zr alloys. Parameters at the transition points were analyzed to determine thresholds for this change in behavior, which occurred at an average hoop strain of 6.9 ± 2.1 %. [ABSTRACT FROM AUTHOR]

• 2D RZ mesh sensitivity study for clad balloon and burst times. • Inclusion of spacer grids and vanes. • Effect of spacer grid placement and size on burst behavior. • Effect of rod initial fill pressure on burst behavior. • Development of boundary conditions to approximate spacer grids in 3D. • 3D quarter rod simulations. This provides an initial scoping study on clad balloon and burst behavior for burnup extension of reactor fuel. The associated issues with burnup extension are fuel fragmentation, relocation, and dispersal in the event of cladding failure. The general finding of this work is that the structural features, spacer grids and mixing vanes, locally suppress cladding deformation but have little impact on the overall clad performance during loss-of-coolant accidents. The work detailed in a previous report by Capps et al. focused on core optimization via neutronics, thermal hydraulics and thermomechanical analysis for burnups beyond 62 GWD/tU and enrichments above 5%. Uncertainty of fuel fragmentation relocation and dispersal in high-burnup rods during accident conditions was also investigated. The dispersal aspect of fuel fragmentation depends on cladding rupture. Thus, assessing uncertainties in the rupture behavior is helpful in estimating the dispersal of high-burnup fuel. This study builds on the previous work by assessing the impact of assembly structural features on cladding balloon and burst behavior in a full-length fuel rod. In this work, the BISON fuel performance code was used to generate 2D radial and height meshes containing structural features commonly used in nuclear fuel assemblies. First, meshes were generated with spacer grids. Results were then compared to the cladding burst temperature and balloon strain results from the previous work. A mesh sensitivity study was performed to ensure that mixing vanes and spacer grid effects were appropriately considered, resulting in a more refined mesh than the previous study. The balloon deformation and burst times of the cladding were compared to the original case. Consideration was also given to the effect of rod initial pressure. 3D quarter rod simulations were also performed and found good agreement with the 2D simulations in clad deformation and reasonable agreement in burst times. [ABSTRACT FROM AUTHOR]

In the US, commercial spent nuclear fuel (SNF) is transferred to interim dry storage casks where it will be stored for decades awaiting transport to a consolidated interim storage facility or a geologic repository. Because the fuel rod cladding is the first barrier against any radioactive material release, understanding the behavior of SNF cladding, particularly at high burnup (HBU), in dry storage conditions is crucial to safely store and transport the spent fuel. In this study, a series of metallographic examinations and cladding hydrogen measurements were conducted on HBU SNF cladding at Oak Ridge National Laboratory as a part of the High Burnup Spent Fuel Data Project, which is sponsored by the US Department of Energy (DOE) Office of Nuclear Energy (NE). To investigate the effect of simulated drying conditions on the cladding, three as-received fuel rods with different cladding materials—M5, ZIRLO, and Zircaloy-4—were heated to 400°C and then slow-cooled to room temperature. The pellet and cladding were then qualitatively and quantitatively analyzed and compared in terms of pellet crack morphology, HBU rim, waterside oxide, cladding H, and cladding hydride morphologies. This paper presents and discusses the results of these analyses in detail. [ABSTRACT FROM AUTHOR]

NEURAP is a dedicated set-up at the Swiss neutron spallation source (SINQ) at the Paul Scherrer Institut (PSI), optionally implemented as a special configuration of the neutron-imaging station NEUTRA. It is one of very few instrumentations available worldwide enabling neutronimaging of highly radioactive samples to be performed routinely, with special precautions and following a specific procedure. Since the relevant objects are strong -sources, dedicated techniques are needed to handle the samples and to perform neutron-imaging despite the radiation background. Dysprosium (Dy)-loaded imaging plates, effectively made sensitive to neutrons only, are employed. Neutrons are captured by Dy during neutron irradiation. Then the imaging plate is erased removing gamma detections. A subsequent relatively long self-exposure by the radiation from the intrinsic neutron-activated Dy within the imaging plate yields the neutron-only radiograph that is finally read out. During more than 20 years of NEURAP operation, images have been obtained for two major applications: (a) highly radioactive SINQ target components were investigated after long-term operation life; and (b) spent fuel rods and their cladding from Swiss nuclear power plants were characterized. Quantitative analysis of the image data demonstrated the accumulation of spallation products in the lead filled "Cannelloni" Zircaloy tubes of the SINQ target and the aggregation of hydrogen at specific sites in used fuel pins of power plants and their cladding, respectively. These results continue to help understanding material degradation and optimizing the operational regimes, which might lead to extending the safe lifetimes of these components. [ABSTRACT FROM AUTHOR]

For several decades, extensive research has been performed on the recovery and purification of zirconium from spent nuclear fuel cladding using a variety of chlorination reaction processes. After the reaction between fuel cladding and chlorine gas, zirconium tetrachloride is separated from other chloride species based on their boiling/sublimation points; however, the presence of iron and niobium chloride impurities limits the efficiency of these processes. In this work, chlorination products of Zr, Fe, and Nb mixtures were analyzed by thermogravimetric analysis, and the results suggest that Fe impurities cannot be removed via chlorination alone. Purification of zirconium from Zircaloy-2, Zircaloy-4, and a Zr-Nb alloy was performed via hydrochlorination using a sealed tube reaction system. The purity of the final ZrCl4 products is higher than 99.99% after successful removal of Fe and Nb. [ABSTRACT FROM AUTHOR]

Hydrogen uptake by nuclear fuel claddings during normal operation as well as loss of coolant during design basis and severe accidents beyond design basis has a high safety relevance because hydrogen degrade the mechanical properties of the zirconium alloys applied as cladding material. Currently, claddings with enhanced accident tolerance are under development. One group of such accident tolerant fuel (ATF) claddings are zirconium alloys with surface coatings reducing corrosion and high-temperature oxidation rate, as well as the chemical heat and hydrogen release during hypothetical accidents. The hydrogen permeation through the coating is an important parameter ensuring material safety. In this work, the hydrogen permeation of Ti2AlC and Cr2AlC MAX phase coatings on Zircaloy-4 is investigated by means of neutron radiography. Both coatings are robust hydrogen diffusion barriers that effectively suppress hydrogen permeation into the matrix.

A two-step ion exchange process has been developed for decontamination of used pressure tube zircaloy, from pressurized heavy water reactors, with respect to 125Sb and 60Co. The steps are: anion exchange separation of Sb (III) and cation exchange separation of Co (II); from Zr (IV). Our earlier results on distribution coefficients, isotherms and kinetic separation of Co were the erecting pillars of the present integrated separation process. Oxidation state of Sb in the feed solution, need for reduction of Sb (V) to Sb (III), the nature and amount of reagents required were investigated. Effect of Zr concentration on uptake of Sb and Co, resin capacity, requirement, column dimensions, and flow rate was evaluated. [ABSTRACT FROM AUTHOR]

A diffused interface between a ceramic coating and a metallic substrate is expected to signify a more solid bonding. In addition to depth profiling, a novel 3D view and imaging approach based on ToF‐SIMS analysis was developed to investigate the diffused species around the interface. The diffusion of Al species in a Cr2AlC ceramic coating and Zircaloy substrate system was investigated in both the as‐deposited and postannealed states. In terms of the 3D view and imaging of CsAl+ after Gaussian convolution, Al species visibly diffused into the substrate after annealing at 800°C for only 5 minutes compared with that in the as‐deposited sample. [ABSTRACT FROM AUTHOR]