Download or read book Modeling Of Hydrogen And Hydride Formation In Zirconium Alloy Cladding Using High fidelity Multi physics Coupled Codes written by Michael Mankosa and published by . This book was released on 2015 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The reactor environment, in which nuclear fuel operates, requires improved multi-dimensional fuel and cladding simulation and analysis to accurately describe fuel behavior. The high-fidelity fuel performance code BISON was developed at Idaho National Laboratory (INL) to address this need. BISON is a three-dimensional finite-element based fuel performance code. In the high temperature environment of a reactor, the zirconium in the cladding undergoes waterside corrosion, releasing hydrogen in the process. Some of this hydrogen is absorbed by the cladding. Once hydrogen is absorbed in the cladding, its distribution is extremely sensitive to temperature, stress, and concentration gradients. Hydrogen migrates down temperature and concentration gradients and at a high enough concentration, precipitates as hydrides which can embrittle the cladding. Hydrogen distribution as a hydride precipitate in cladding has been identified as a possible ersatz for validating reactor simulation code temperature models. This thesis shows development efforts of using high-fidelity multi-physics codes to model temperature, hydrogen, and hydride distribution. Several multi-physics code couplings were used to model pressurized water reactor sub-assemblies. The first was the Penn State University (PSU) developed DeCART-CTF coupling. This thesis outlines a demonstration of this coupling's ability to predict hydrogen distribution in the two dimensional radial (r, [theta]) direction. The Consortium for the Advanced Simulation of Light Water Reactors coupled multi-physics code, Tiamat, was used to model a select sub-assembly with spacer grids. A section of a single fuel pin was then selected from this sub-assembly and modeled in a three-dimensional BISON problem to obtain three-dimensional hydrogen and hydride distributions at selected areas around spacer grids. The hydrogen and hydride model showed results and behavior that accurately produced the expected physics of hydrogen and hydrides in all three dimensions. The hydrogen model is under continuing development at PSU and INL.This thesis also presents a code-to-code comparison between CTF, BISON, and FRAPCON to compare the ability of CTF to predict fuel pin temperatures. The evaluations between the codes show good agreement on fuel pin temperature distribution. If CTF is well informed by BISON or FRAPCON with gap conductance, burnup, and radial power profile values, then CTF will accurately reproduce fuel temperatures. The largest difference reported in fuel centerline temperature between CTF and FRAPCON was 6.78 degrees Kelvin or 0.49%. The largest centerline difference between CTF and BISON was 4.39 degrees Kelvin or 0.30%.

Download or read book Development of a Phase Field Model of Hydride Morphology in Zirconium Alloy Nuclear Fuel Cladding written by Pierre Simon and published by . This book was released on 2017 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Zirconium alloys are widely used in the nuclear industry as fuel cladding due to their particular properties. During normal operation conditions, hydrogen enters the cladding and forms brittle hydride precipitates. The effect of the presence of hydrides on the deformation behavior of the cladding largely depends on the orientation and the morphology of the hydrides. Because of the zirconium texture and the thermo-mechanical conditions, hydrides usually precipitate circumferentially in the cladding. However, temperature cycling and the application of additional stress can lead to hydride reorientation in the radial direction, which eases crack propagation through the cladding, and thus threatens the integrity of the fuel rod. In an effort to understand the mechanisms governing the orientation and the morphology of the hydrides, two different phase field models were developed using the Multi-physics Object Oriented Simulation Environment MOOSE. The first model was first proposed by Wheeler, Boettinger, and McFadden and is known as the WBM model. The second model, called the grand potential model, has the advantage of allowing the definition of the interfacial thickness independently of the bulk free energy of the different phases of the system. It thus allows the use of thicker interfaces, which means coarser mesh, making the simulations computationally less expensive. Because of the importance of the mechanical contributions in the nucleation and growth of hydride precipitates, both phase field models have then been coupled with elastic schemes. The first scheme, called the Voight-Taylor scheme (VTS), was shown to strongly overestimate the elastic free energy contribution at the interface, while the Khachaturya's scheme (KHS) performed better with just a small underestimation of the elastic free energy at the interface. In the project presented in this thesis, the multi-phase models simulated the alpha phase of the zirconium as well as the zeta, the gamma, and the delta phase of the hydrides. The models are dimensional, use the Gibbs free energy of formation of the different phases and the mechanical properties found in the literature. In this study, the phase field models have been carefully verified, meaning that their implementations have been successfully tested by comparing their results to widely accepted solutions. Once the models were applied to the zirconium hydride system, the first steps towards the validation of the code were promising. Simulated hydrides grew preferentially in the direction of the basal plane of the zirconium matrix, thus reproducing experimental observations.

Download or read book Phase Field Modeling and Quantification of Zirconium Hydride Morphology written by Pierre Clement Simon and published by . This book was released on 2021 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: In light water nuclear reactors, waterside corrosion of the cladding material leads to the production of hydrogen, a fraction of which is picked up by the zirconium cladding. Once the hydrogen concentration reaches its solid solubility limit in zirconium, it precipitates into brittle hydride particles. These nanoscale hydride particles aggregate into mesoscale hydride clusters. Depending on the material's texture and the thermomechanical treatment imposed on the cladding, these mesoscale hydride clusters exhibit different morphologies. In particular, the principal orientation of the hydride platelets in the cladding tube can be circumferential or radial. Because hydrides are usually more brittle than the zirconium matrix, the morphology of the mesoscale hydride clusters can affect cladding integrity. This is in part because radial hydrides can ease crack propagation through the cladding thickness and because the concentration of hydrides in specific locations driven by temperature, hydrogen concentration, and stress gradients can create local weak points in the cladding. This dissertation work investigates the link between precipitation conditions, hydride morphology, and hydride embrittlement in zirconium cladding material. The first part focuses on understanding which physics and mechanisms govern the formation of specific hydride microstructures. A quantitative phase field model has been developed to predict the hydride morphology observed experimentally and identify which mechanisms are responsible for circumferential and radial hydride precipitation. The model accurately predicts the elongated nanoscale hydride shape and the stacking of hydrides along the basal plane of the hexagonal zirconium matrix. When investigating the role of applied stress on hydride morphology, the model challenges some of the mechanisms proposed in previous studies to explain hydride reorientation. Although hydride reorientation has been hypothesized to be caused by a change in nanoscale hydride shape and orientation, the current model shows that these mechanisms are unlikely. This study focuses on the precipitation of nanoscale hydrides in polycrystalline zirconium to understand the physics and mechanisms responsible for the change in hydride microstructure from circumferential to radial under applied stress. It proposes a new mechanism where the presence of an applied stress promotes hydride precipitation in grains with circumferentially aligned basal poles. Nanoscale hydrides, even though they still grow along the basal plane of the hexagonal matrix, now grow and stack radially, thus leading to radial mesoscale hydrides. This mechanism is consistent with experimental observations performed in other studies. The second part of this dissertation focuses on the link between hydride morphology and hydride embrittlement. Although hydride microstructure can significantly influence Zr alloy nuclear fuel cladding's ductility, quantifying hydride microstructure is challenging and several of the metrics currently being used have significant shortcomings. A new metric has been developed to quantify hydride microstructure in 2D micrographs and relate it to crack propagation. As cladding failure usually results from a hoop stress, this new metric, called the Radial Hydride Continuous Path (RHCP), is based on quantifying the continuity of brittle hydride particles along the radial direction of the cladding tube. Compared to previous metrics, this approach more closely relates to the propensity of a crack to propagate radially through the cladding tube thickness. The RHCP takes into account hydride length, orientation, and connectivity to choose the optimal path for crack propagation through the cladding thickness. The RHCP can therefore be more closely linked to hydride embrittlement of the Zr alloy material, thus creating a relationship between material structure, properties, and performance. The new definition, along with previously proposed metrics such as the Radial Hydride Fraction (RHF), the Hydride Continuity Coefficient (HCC), and the Radial Hydride Continuity Factor (RHCF), have been implemented and automated in MATLAB. These metrics were verified by comparing their predictions of hydride morphology against expected values in simple cases, and the implementation of the new metric was validated by comparing its predictions with manual measurements of hydride microstructure performed on ImageJ. The RHCP was also validated against experimental measurements of fracture behavior and it was shown to correlate with cladding failure better than previous metrics. The information provided by these metrics will help accurately assess cladding integrity during operation, transportation, and storage.

Download or read book Modeling Zirconium Hydride Precipitation and Dissolution in Zirconium Alloys written by Evrard Lacroix and published by . This book was released on 2019 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Nuclear fuel cladding undergoes waterside corrosion during normal operating conditions in pressurized water reactors, whereby the zirconium (Zr) in the fuel cladding reacts with the oxygen present in water, creating zirconia (ZrO) and releasing hydrogen. Part of the hydrogen created by the corrosion reaction can be absorbed into the fuel cladding. Once in the cladding, hydrogen redistributes by solid state diffusion in the metal, in response to gradients of concentration, temperature and stress. Once the local hydrogen solubility is exceeded, zirconium hydride precipitates are formed.The precipitation of hydrides may impact the integrity of zirconium-based nuclear fuel cladding, both during normal operation and during extended dry storage. It is important to model hydrogen behavior accurately, so as to assess cladding properties both in reactor and during dry storage. This is because the cladding is the first containment barrier, which prevents fission products to be released into the primary circuit. For this reason, this study aims to first understand hydride precipitation and dissolution and then implement this understanding into a hydride precipitation and dissolution model. To this end, differential scanning calorimetry (DSC) and in-situ synchrotron X-ray diffraction experiments were used to study the precipitation and dissolution of hydrides in Zircaloy-4 under different thermo-mechanical conditions.Results showed that when hydrided samples were cooled at cooling rates above 1C/min the hydrogen content in solid solution decreased, following the Terminal Solid Solubility for Precipitation (TSSP) curve. However, when the samples were held at a fixed temperature for a long anneal, the hydrogen content in solid solution continued to decrease below the TSSP and approached the Terminal Solid Solubility for Dissolution (TSSD). This result suggests that TSSP is a kinetic limit and that a unique solubility limit, i.e. TSSD governs the equilibrium hydrogen concentration in solid solution. DSC was used to perform isothermal precipitation experiments, from which the hydride precipitation rate and the degree of precipitation completion were quantified between 280 and 350C for the first time. The data obtained was used to generate a TTT diagram for hydride precipitation in Zircaloy-4 showing that hydride precipitation is diffusion-controlled at low temperatures and reaction-controlled at high temperatures. The experimental precipitation rate was fitted using the Johnson-Mehl-Avrami-Kolmogorov model to obtain a value of the Avrami parameter of 2.56 (2.5 is the theoretical value for the growth of platelet-shaped precipitates). It was also possible to derive the precipitation activation energy of for each process. Because it was possible to separate hydride nucleation and hydride growth, it was possible to ascertain that if the hydrogen content in solid solution is greater than TSSP, precipitation occurs by hydride nucleation. In contrast, precipitation occurs by hydride growth as long as hydride platelets are present and the hydrogen content in solid solution is above TSSD. Hydride dissolution will take place if hydrides are present and the hydrogen content in solid solution is below TSSP. Using this new understanding of hydrogen precipitation and dissolution mechanisms, experiments were conducted at the Advanced Photon Source (APS) using high temperature change rates to measure hydride nucleation and dissolution kinetics. These observations and measurements were combined to existing theory to a model, entitled Hydride Growth, Nucleation, and Dissolution model (HNGD model) that can accurately simulate hydrogen behavior in Zircaloy fuel cladding and that shows a significant improvement on the model used in BISON.The development of such a model is the first step towards obtaining a model for the impact of the development of hydride microstructure on nuclear fuel cladding mechanical properties during normal operation and to address concerns over fuel handling during dry storage. The use and benchmarking of such a code can be used to justify a safe burnup extension of nuclear fuel, which would reduce the cost of nuclear energy in an increasingly competitive market.

Download or read book The Effect of Hydrogen and Hydrides on the Integrity of Zirconium Alloy Components written by Manfred P. Puls and published by Springer Science & Business Media. This book was released on 2012-08-04 with total page 475 pages. Available in PDF, EPUB and Kindle. Book excerpt: By drawing together the current theoretical and experimental understanding of the phenomena of delayed hydride cracking (DHC) in zirconium alloys, The Effect of Hydrogen and Hydrides on the Integrity of Zirconium Alloy Components: Delayed Hydride Cracking provides a detailed explanation focusing on the properties of hydrogen and hydrides in these alloys. Whilst the emphasis lies on zirconium alloys, the combination of both the empirical and mechanistic approaches creates a solid understanding that can also be applied to other hydride forming metals. This up-to-date reference focuses on documented research surrounding DHC, including current methodologies for design and assessment of the results of periodic in-service inspections of pressure tubes in nuclear reactors. Emphasis is placed on showing how our understanding of DHC is supported by progress in general understanding of such broad fields as the study of hysteresis associated with first order phase transformations, phase relationships in coherent crystalline metallic solids, the physics of point and line defects, diffusion of substitutional and interstitial atoms in crystalline solids, and continuum fracture and solid mechanics. Furthermore, an account of current methodologies is given illustrating how such understanding of hydrogen, hydrides and DHC in zirconium alloys underpins these methodologies for assessments of real life cases in the Canadian nuclear industry. The all-encompassing approach makes The Effect of Hydrogen and Hydrides on the Integrity of Zirconium Alloy Component: Delayed Hydride Cracking an ideal reference source for students, researchers and industry professionals alike.

Download or read book A Solid Mechanics Approach to the Modeling of Hydride Formation and Cracking in Zirconium Alloys written by and published by . This book was released on 1998 with total page 55 pages. Available in PDF, EPUB and Kindle. Book excerpt: The authors use a solid mechanics approach to investigate hydride formation and cracking in zirconium-niobium alloys used in the pressure tubes of CANDU nuclear reactors. In this approach, the forming hydride is assumed to be purely elastic and its volume dilatation is accommodated by elasto- plastic deformation of the surrounding matrix material. The energetics of the hydride formation is revisited and the terminal solid solubility of hydrogen in solution is defined on the basis of the total elasto-plastic work done on the system by the forming hydride and the external loads. Hydrogen diffusion and probabilistic hydride formation coupled with the material deformation are modelled at a blunting crack tip under plane strain loading. A full transient finite element analysis allows for numerical monitoring of the development and expansion of the hydride zone as the externally applied loads increase. Using a Griffith fracture criterion for fracture initiation, the reduced fracture resistance of the alloy can be predicted and the factors affecting fracture toughness quantified.

Download or read book Hydrogen Migration and Mechanical Behavior of Hydrided Zirconium Alloys written by Soyoung Kang and published by . This book was released on 2023 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt: Zirconium alloys have been widely used for nuclear fuel cladding materials in light-water nuclear reactors. The cladding corrodes as a result of exposure to the coolant water and produces hydrogen as a result of the corrosion reaction. A fraction of this hydrogen can be picked up into the cladding material. Once the hydrogen content reaches the terminal solid solubility, zirconium hydride particles start to precipitate. The cladding suffers waterside corrosion in service, leading to hydrogen ingress, which can redistribute in the cladding and form hydrides. Because these zirconium hydrides are more brittle than the zirconium matrix, they can deteriorate the ductility of the cladding. Therefore, understanding hydrogen behavior in cladding is important to maintain cladding integrity. This study aims to investigate hydrogen migration under a temperature gradient and mechanical behavior of hydrided zirconium alloys. The hydrogen transport and hydride precipitation /dissolution model HNGD was implemented in the fuel performance code BISON to predict hydrogen behavior. The hydrogen is distributed inhomogeneously in the cladding as a result of Fick's law and Soret effect. The hydrogen tends to move from higher to lower concentration governed by Fick's law and higher to lower temperature based on the Soret effect. Hydrogen migration tests were designed to determine the heat of transport value (Q*) of hydrogen in Zr, a parameter needed to evaluate the Soret effect. Hydrided samples were subjected to a long annealing schedule in a temperature gradient to re-distribute the hydrogen. The annealed samples were cut into several pieces along the temperature gradient, and the hydrogen contents were analyzed using hot vacuum extraction. The hydrogen distribution along the temperature gradient was observed in this experiment, and from this data, the heat of transport value (Q*) was determined. Further, the mechanical behavior of zirconium alloys was assessed using ring compression tests. The zirconium alloy tubes were characterized by electron backscatter diffraction (EBSD) to identify the microstructure of materials. Stress relieved anneal ZIRLO (SRA) and low Sn Partially recrystallized anneal LT ZIRLO (PRXA) show different grain shapes and sizes. After characterization, the zirconium alloy tubes were hydrogen charged and cut into 8 mm length rings. The ring samples were subjected to compression at 12 o'clock following a specified thermomechanical cycle. This thermomechanical treatment caused partial precipitation of radial hydrides in certain positions of the ring samples. The radial hydride fractions were characterized and showed a difference between ZIRLO and LT ZIRLO because of their different microstructures. Finite element modeling conducted using ABAQUS could then determine the threshold stress for two materials by comparing simulation results (stress state) and hydride morphologies. In addition, the ring compression tests for assessing hydrided cladding ductility for various hydride morphologies were conducted at room temperature. Ring samples with different radial hydride continuity factors (RHCF) were tested to determine their load-displacement curves. The 1% permanent strain and 2 % offset strain criteria were chosen to assess the ductility of samples. The ductility degrades with increasing RHCF.

Download or read book Specific Zirconium Alloy Design Program Quarterly Progress Report written by and published by . This book was released on 1962-02 with total page 128 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Multiphase Field Modeling of the Formation Path of Delta Hydrides in Zirconium written by Jacob Luke Bair and published by . This book was released on 2016 with total page 121 pages. Available in PDF, EPUB and Kindle. Book excerpt: "Zirconium alloys are commonly used in nuclear fuel rod claddings due to their high ductility, good corrosion resistance, and low neutron absorption cross section. Among the most important weaknesses of zirconium alloys is their affinity for hydrogen, resulting in formation of hydrides in the cladding, and leading to embrittlement and mechanical failure. Despite numerous studies on hydride precipitation in zirconium alloys, the nucleation and formation path of stable [delta] hydrides in [alpha] zirconium matrix are not yet fully understood. In this Ph. D. research project, two novel quantitative phase-field models were developed and utilized to advance our understanding of mechanisms of formation and evolution of hydrides in zirconium alloys. First, a phase-field model for unstable [gamma] hydride precipitation was created to build on previous computational models by including the actual Gibbs free energy of formation of hydrides in the total free energy of the system. Results from isothermal simulations of seeded and random nucleation in single crystal [alpha]-zirconium matrix showed that the thickness of non-equilibrium hydrides varied with temperature during evolution, and the hydrides were more rod-like (thinner) at higher temperatures and thicker at lower temperatures. Quench simulations with random nucleation indicated that the majority of precipitation occurs at early stages of quenching, but the size and shape of hydrides change as the temperature decreases. The most detrimental phase of hydrides in claddings is the stable [delta] phase. A multiphase model including the two metastable phases ([zeta] and [gamma]) and the [delta] phase was created to determine the effects of the intermediate phases on the nucleation and morphology of [delta] hydrides. Results from simulations both with and without applied strains indicated that the intermediate phases are influential in the initial formation and evolution of [delta] hydrides"--Abstract, page iv.

Download or read book Internal Zirconium Hydride Formation in Zircaloy Fuel Element Cladding Under Irradiation written by J. M. Markowitz and published by . This book was released on 1963 with total page 52 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Hydrogen Entry in Zircaloy 4 Fuel Cladding written by Jennifer Anne Jarvis and published by . This book was released on 2015 with total page 318 pages. Available in PDF, EPUB and Kindle. Book excerpt: Corrosion and hydrogen pickup of zirconium alloy fuel cladding in water cooled nuclear reactors are life-limiting phenomena for fuel. This thesis studies the fate of hydrogen liberated by waterside corrosion of Zircaloy-4 fuel cladding in Pressurized Water Reactors (PWRs): are the adsorbed protons incorporated into the oxide and eventually the metal, or are they evolved into molecular hydrogen and released into the coolant? Water chemistry modeling was used to understand effects of radiolysis and CRUD. Density functional theory (DFT) was used to investigate the role of oxidized Zr(Fe,Cr)2 second phase particles. Chemical potentials and the electron chemical potential were used to connect these two modeling efforts. A radiolysis model was developed for the primary loop of a PWR. Dose profiles accounting for fuel burnup, boron addition, axial power profiles, and a CRUD layer were produced. Dose rates to the bulk coolant increased by 21-22% with 12.5-75 pim thick CRUD layers. Radially-averaged core chemistry was compared to single-channel chemistry at individual fuel rods. Calculations showed that local chemistry was more oxidizing at high-power fuel and fuel with CRUD. Local hydrogen peroxide concentrations were up to 2.5 ppb higher than average levels of 5-8 ppb. Radiolysis results were used to compute chemical potentials and the corrosion potential. Marcus theory was applied to compare the band energies of oxides associated with Zircaloy-4 and the energy levels for proton reduction in PWR conditions. Hydrogen interactions with Cr203 and Fe203, both found in oxidized precipitates, were studied with DFT. Atomic adsorption of hydrogen was modeled on the Cr and Feterminated (0001) surfaces. Climbing Image-Nudged Elastic Band calculations were used to model the competing pathways of hydrogen migration into the subsurface and molecular hydrogen formation. A two-step mechanism for hydrogen recombination was identified consisting of: reduction of an adsorbed proton (H+) to a hydride ion (H-) and H2 formation from an adjacent adsorbed proton and hydride ion. Overall, results suggest that neither surface will be an easy entrance point for hydrogen ingress and that Cr203 is more likely to be involved in hydrogen evolution than the Fe203.

Download or read book Effect of Hydrogen on Mechanical Behavior of a Zircaloy 4 Alloy written by and published by . This book was released on 2005 with total page 182 pages. Available in PDF, EPUB and Kindle. Book excerpt: Hydride formation is one of the main degradation mechanisms of zirconium alloys in hydrogen-rich environments. When sufficient hydrogen is present, zirconium- hydride precipitates can be formed. Cracking of the brittle hydrides near a crack tip can initiate the growth of a crack leading to the premature failure of the material. Hydride formation is believed to be enhanced by the presence of residual or applied stresses. Therefore, the increase in the stress field ahead of a crack tip may promote precipitation of additional hydrides. In order to verify these phenomena, the effect of internal stresses on the zirconium-hydride-precipitate formation, and in turn, the influence of the hydrides on the subsequest intergranular strain evolution in a hexagonal-close-packed zircaloy-4 alloy were investigated, using neutron and x-ray diffraction. First, the evolution of intergranular strains in a zircaloy-4 was investigated in-situ, using neutron diffraction, to understand the deformation behavior at the microscopic length scale. A series of uniaxial tensile loads up to 500 MPa was applied to a round-bar tensile specimen in the as-received condition and the intergranular (hkl-specific) strains, parallel and perpendicular to the loading direction, were studied. The results provide a fundamental understanding of the anisotropic elastic-plastic deformation of the zirconium alloy under applied stresses. Then the hydride formation was examined by conducting qualitative phase mapping across the diameter of two tensile specimens charged with hydrogen gas for 1/2 hour and 1 hour, respectively. It was observed that the zirconium hydrides ([delta]-ZrH2) form a layer, in a ring shape, near the surface with a thickness of approximately 400 [mu]m. The hydrogen-charging effects on intergranular strains were investigated and compared to the as-received specimen. Second, spatially-resolved internal-strain mapping was performed on a fatigue pre-cracked compact-tension (CT) specimen using in-situ neutron diffraction under applied loads of 667 and d4,444 newtons, to determine the in-plane (parallel to the loading direction) and through-thickness (perpendicular to the loading direction) lattice-strain profiles around the crack tip. An increase in elastic lattice strains near the crack tip was observed with the increase in the applied stresses. The effect of hydrogen charging was also investigated on CT specimens electrochemically charged with hydrogen. X-ray diffraction results clearly showed the presence of zircomium hydrides on the surfaces of the specimen.

Download or read book White Paper Summary of 2nd ASTM International Workshop on Hydrides in Zirconium Alloy Cladding written by and published by . This book was released on 2015 with total page 34 pages. Available in PDF, EPUB and Kindle. Book excerpt: This white paper recommends that ASTM International develop standards to address the potential impact of hydrides on the long term performance of irradiated zirconium alloys. The need for such standards was apparent during the 2nd ASTM International Workshop on Hydrides in Zirconium Alloy Cladding and Assembly Components, sponsored by ASTM International Committee C26.13 and held on June 10-12, 2014, in Jackson, Wyoming. The potentially adverse impacts of hydrogen and hydrides on the long term performance of irradiated zirconium-alloy cladding on used fuel were shown to depend on multiple factors such as alloy chemistry and processing, irradiation and post irradiation history, residual and applied stresses and stress states, and the service environment. These factors determine the hydrogen content and hydride morphology in the alloy, which, in turn, influence the response of the alloy to the thermo-mechanical conditions imposed (and anticipated) during storage, transport and disposal of used nuclear fuel. Workshop presentations and discussions showed that although hydrogen/hydride induced degradation of zirconium alloys may be of concern, the potential for occurrence and the extent of anticipated degradation vary throughout the nuclear industry because of the variations in hydrogen content, hydride morphology, alloy chemistry and irradiation conditions. The tools and techniques used to characterize hydrides and hydride morphologies and their impacts on material performance also vary. Such variations make site-to-site comparisons of test results and observations difficult. There is no consensus that a single material or system characteristic (e.g., reactor type, burnup, hydrogen content, end-of life stress, alloy type, drying temperature, etc.) is an effective predictor of material response during long term storage or of performance after long term storage. Multi-variable correlations made for one alloy may not represent the behavior of another alloy exposed to identical conditions and the material responses to thermo-mechanical exposures will be different depending on the materials and systems used. The discussions at the workshop showed several gaps in the standardization of processes and techniques necessary to assess the long term performance of irradiated zirconium alloy cladding during dry storage and transport. The development of, and adherence to, standards to help bridge these gaps will strengthen the technical basis for long term storage and post-storage operations, provide consistency across the nuclear industry, maximize the value of most observations, and enhance the understanding of behavioral differences among alloys. The need for, and potential benefits of, developing the recommended standards are illustrated in the various sections of this report.

Download or read book Hydride Reorientation in Zirconium Alloy Nuclear Fuel Cladding Studied Using Synchrotron Radiation written by Jennifer Anne Jarvis and published by . This book was released on 2010 with total page 63 pages. Available in PDF, EPUB and Kindle. Book excerpt: During reactor operation, waterside corrosion of Zirconium alloy fuel cladding leads to hydrogen pickup. Hydrides platelets will normally precipitate circumferentially in the cladding. However, under temperature and load, the hydrides can reorient in the radial direction. These brittle radial hydrides can drastically reduce the ductility and the fracture toughness of the cladding. This work contains an in-situ study of hydride reorientation. Using synchrotron radiation from the Advanced Photon Source (APS) at Argonne National Laboratory, x-ray diffraction in transmission geometry was used to study a hydrogen-charged sample under mechanical and thermal load. The sample was rolled Zircaloy-2 with a hydrogen content of 96 wt.ppm. It was studied under a loading cycle with temperature cycling from 25 to 400°C and with an applied tensile stress of 100 MPa. Under two cycles of loading, partial reorientation was achieved. This diffraction data was used to perform an evaluation of the intensity, peak broadening, and peak shift of hydride peaks, in order to characterize the kinetics of reorientation. Additionally, the dissolution and precipitation temperatures were studied. Optical microscopy was used to compare the microstructure and hydride orientation before and after the experiment.

Download or read book Effect of Hydride Distribution on the Mechanical Properties of Zirconium Alloy Fuel Cladding and Guide Tubes written by Suresh K. Yagnik and published by . This book was released on 2014 with total page 31 pages. Available in PDF, EPUB and Kindle. Book excerpt: Localization of hydride precipitates exacerbates the hydrogen embrittlement effects on the deformation and fracture properties of Zircaloy fuel cladding materials. Thus, at comparable hydrogen concentration levels, localized hydride precipitates are more detrimental from the standpoint of cladding integrity during service. Indeed, the hydride precipitates are often non-homogeneously distributed in fuel assembly components; for example, in irradiated fuel cladding, the hydride rim is formed near the outer oxide-metal interface because of the temperature gradient that exists during operation. With increasing fuel burnup, this hydride rim not only becomes denser but might be accompanied by gradients in local hydrogen and hydride concentrations through the rest of the cladding wall thickness. Whereas the importance of hydride spacing and their orientation, as well as the alloy matrix ligaments interspaced with the distributed hydride has been recognized in the literature, little work has been reported on the effects of hydride precipitate distribution on the mechanical properties of Zircaloy fuel assembly component materials. In this paper, we report on an extensive mechanical test program on low-tin Zircaloy-4 specimens from stress-relieved cladding and recrystallized guide tubes, charged with hydrogen to obtain uniform, rimmed, and layered hydride distributions. The hydrogen concentration (0-1200 ppm) and hydride rim thickness (10-90 ?m) were also varied. The strain rate was kept at 10-4/s to simulate in-service steady-state conditions and the tests were conducted both at room temperature and 300°C. All test specimens were of small-gauge-section, cut-outs from cladding, and guide tubes. The loading configurations included slotted-arc test (SAT) on half-ring-shaped specimens and uniaxial tension test (UTT) on dog-bone-shaped cut-outs. Further, prompted by the finite-element analysis of the gauge-section region, a unique geometry of internal slotted-arc specimens with parallel gauge section (ISATP) was chosen. Detailed stress-strain curves for all tests were measured, and post-test fractography and local hydrogen concentrations within the gauge sections were measured by hot extractions. Comparative data on the measured strengths and elongations for the three types of hydride distributions (i.e., uniform, rimmed, and layered) are presented. Quantification and analyses of these effects have provided a general constitutive stress-strain relationship for assessing margins to cladding or guide tube failures.

Download or read book Hydride Platelet Reorientation in Zircaloy Studied with Synchrotron Radiation Diffraction written by Arthur T. Motta and published by . This book was released on 2011 with total page 27 pages. Available in PDF, EPUB and Kindle. Book excerpt: Hydrogen ingress into zirconium alloy fuel cladding in light water reactors can degrade cladding performance as a result of the formation of brittle hydrides. In service, hydrides normally precipitate in the circumferential direction and are homogeneously distributed through the cladding thickness in ideal cases. However, temperature and stress gradients in the cladding can promote hydrogen redistribution. This hydrogen redistribution is responsible for the formation of hydride rims, dissolution, and reorientation of hydride precipitates and for the formation of brittle hydrides at stress concentration locations, all of which can reduce cladding resistance to failure. Thus, it is crucial to understand the kinetics of hydride dissolution and precipitation under load and at temperature. Studies of hydrogen behavior in zirconium alloys are normally performed post facto, which causes them to suffer both from a scarcity of data points and from the confounding effects of studying hydrides at room temperature that might be dissolved at higher temperature. In the current study, we have used synchrotron radiation diffraction to study the kinetics of hydride precipitation and dissolution in situ (under load and at temperature). Samples of hydrided Zircaloy-4 were examined in transmission by using 80 keV synchrotron radiation while undergoing heating and cooling in a furnace. Temperatures ranged from 20 to 550°C, and loads from 75 to 100 MPa were applied. The hydrides dissolved and reprecipitated in a different orientation when sufficiently high loads were applied. Through careful study of the intensities and full-width half maxima of the diffraction peaks as a function of time, load, and temperature, it was possible to identify the characteristic diffraction patterns for the reoriented hydrides so that the kinetics of dissolution, reprecipitation, and orientation of the hydrides could be followed. The analysis of the diffraction patterns allowed a detailed understanding of the kinetics of hydride evolution under temperature and stress, as presented in this work.

Download or read book Hydride Nucleation Growth Dissolution HNGD Model written by Florian Passelaigue and published by . This book was released on 2021 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: In nuclear reactors, waterside corrosion of the Zircaloy nuclear fuel cladding tube causes hydrogen pick-up into the material. This hydrogen can cause zirconium hydrides to precipitate within the cladding. Since these hydrides are usually more brittle than the alloy, they can decrease the ductility of the cladding. Previous efforts made to describe hydrogen behavior and hydride precipitation resulted in the Hydride Nucleation-Growth-Dissolution (HNGD) model. This model can predict the distribution of hydrogen and the partition between solid solution and hydrides in Zircaloy samples that are subjected to a thermal treatment. However, in some cases the HNGD model gives unphysical results. Notably, if the system is close to steady-state, or if the initial hydrogen distribution is significantly heterogeneous the model calculations diverge from experimental results, predicting either no hydride precipitation, or hydride precipitation in a single node of the simulation mesh. The study presented in this dissertation describes how the HNGD model was improved upon to address these shortcomings. This was done using two hypotheses described below. Chapter 1 introduces the issue of hydrogen in Zircaloy cladding, and the experimental data used to validate the HNGD model. In Chapter 2, we review the phenomena described by the HNGD model, the equations used, and associated parameters. We also expose the causes for the unphysical results previously mentioned. Chapter 3 focuses on the two hypotheses. The first hypothesis is based on the assumption that, given enough time, the hydrogen atoms will find the most favorable nucleation spots in the Zircaloy (matrix defects, dislocations, etc), resulting in a decrease of the nucleation barrier. This is translated with a decrease of $TSS_P$ during temperature holds. The second hypothesis postulates that the hydride particles deform the matrix in a way that impacts the hydrogen solubility. These two hypotheses together allow for hydride precipitation to be triggered more easily, and for the hydrides to stay stable. A complete analytical solution (i.e. for the hydrogen in solid solution and in hydrides) was derived for the steady state of the system. Using this tool and the large experimental data set from Kammenzind, the impact of the newly introduced parameters is studied. An extensive validation of the modified HNGD model is performed using Kammenzind's experiments, as well as the benchmark and validation cases used during the initial development of the HNGD model. We show that the modified HNGD is able to predict the thickness of the hydride peak at steady state, which is a significant improvement compared to the initial model. Finally, the implementation of the modified HNGD model into the nuclear fuel performance code Bison is described in Chapter 4. We describe how the quality of the code is ensured when implementing an update in Bison. This modified HNGD model yields physical results when modeling experiments that mimic reactor conditions in terms of hydrogen pick-up, and does not degrade the simulations of experiments that were accurately modeled using the initial HNGD model. This improved HNGD model represents an improvement of the capability to predict the hydrogen behavior in cladding tubes during operation and in spent nuclear fuel during storage and transport.