Hydride Precipitation Crack Propagation in Zircaloy Cladding During a Decreasing Temperature History

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Book Synopsis Hydride Precipitation Crack Propagation in Zircaloy Cladding During a Decreasing Temperature History by :

Download or read book Hydride Precipitation Crack Propagation in Zircaloy Cladding During a Decreasing Temperature History written by and published by . This book was released on 2000 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: An assessment of safety, design, and cost tradeoff issues for short (ten to fifty years) and longer (fifty to hundreds of years) interim dry storage of spent nuclear fuel in Zircaloy rods shall address potential failures of the Zircaloy cladding caused by the precipitation response of zirconium hydride platelets. If such assessment analyses are to be done rigorously, they will be necessarily complex because the precipitation response of zirconium hydride platelets is a stochastic functional of hydrogen concentration, temperature, stress, fabrication defect/texture structures, and flaw sizes of the cladding. Thus, there are, and probably always will be, zirhydride questions to analytically and experimentally resolve concerning the consistency, the completeness, and the certainty of models, data, the initial and the time-dependent boundary conditions. Some resolution of these questions will be required in order to have a defensible preference and tradeoffs decision analysis for assessing risks and consequences of the potential zirhydride induced cladding failures during dry storage time intervals. In the following brief discussion, one of these questions is posed as a consequence of an anomaly described in data reproducibility that was reported in the results of tests for hydrogen induced delayed cracking. The testing anomaly consisted of observing a significant differential in the measurable crack velocities (quasi-steady state at a prescribed load and temperature values) that depended on the approach direction, from above or from below, to the test temperature value. The testing method used was restricted to approaching a prescribed test temperature value from above. This anomaly illustrates the known thermodynamic non-equilibrium processes in the precipitation kinetics of zirhydride platelets that are dependent on temperature and stress histories. Detailed solubility limits of hydrogen in Zircaloy as a function of temperature, in terms of zirhydride precipitation and zirhydride dissolution solubility curves, were reported recently. In addition, other tests to evaluate the influence of an applied stress state on zirhydride precipitation kinetics have also been recently reported.

Delayed Hydride Cracking in Irradiated Zircaloy Cladding

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ISBN 13 :
Total Pages : 16 pages
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Book Synopsis Delayed Hydride Cracking in Irradiated Zircaloy Cladding by : K. Pettersson

Download or read book Delayed Hydride Cracking in Irradiated Zircaloy Cladding written by K. Pettersson and published by . This book was released on 2000 with total page 16 pages. Available in PDF, EPUB and Kindle. Book excerpt: Slow stable crack growth by a mechanism identified as a form of delayed hydride cracking has been studied on irradiated Zircaloy cladding. The background to the investigation was the formation of long axial cracks in defected fuel rods. Post-irradiation examination of such fuel rods has indicated that precipitation and subsequent cracking of hydrides at the tips of the long cracks has played an important role in the crack growth process. The present investigation conducted on irradiated cladding with hydrogen concentrations above about 500 ppm has demonstrated that a hydrogen-induced crack growth process can occur in such material. In the laboratory it was necessary to subject the samples to an overtemperature cycle in order to initiate crack growth after fatigue precracking. It was also observed that an incubation time on the order of 20 h was necessary before crack growth started. The crack growth rates were strongly dependent on the applied stress intensity factor K in a narrow range above a threshold value KIH, which was about 10 MPa?m, Stage I. The growth rate then reached a plateau value when it was independent of K, Stage II. This plateau value was about 10-6 m/s at 300°C and about 2 x 10-7 m/s at 200°C. This temperature dependence is consistent with a mechanism based on stress-induced diffusion of hydrogen at the stress concentration of the crack tip. Metallographic and fractographic observations suggest that the details of the mechanism can be best described as a localized reduction of fracture toughness due to reorientation of hydrides so that they become perpendicular to the applied stress in the region of the crack tip. This is somewhat in contrast to previous DHC mechanisms in which longer-range diffusion of hydrogen to one large hydride at the crack tip is usually modeled. The difference is that in the present case the hydride content is higher and therefore more hydrides are present.

Modeling Zirconium Hydride Precipitation and Dissolution in Zirconium Alloys

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Book Synopsis Modeling Zirconium Hydride Precipitation and Dissolution in Zirconium Alloys by : Evrard Lacroix

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.

The Effect of Hydrogen and Hydrides on the Integrity of Zirconium Alloy Components

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Publisher : Springer Science & Business Media
ISBN 13 : 1447141954
Total Pages : 475 pages
Book Rating : 4.4/5 (471 download)

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Book Synopsis The Effect of Hydrogen and Hydrides on the Integrity of Zirconium Alloy Components by : Manfred P. Puls

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.

Temperature and Hydrogen Concentration Limits for Delayed Hydride Cracking in Irradiated Zircaloy

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ISBN 13 :
Total Pages : 19 pages
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Book Synopsis Temperature and Hydrogen Concentration Limits for Delayed Hydride Cracking in Irradiated Zircaloy by : JS. Schofield

Download or read book Temperature and Hydrogen Concentration Limits for Delayed Hydride Cracking in Irradiated Zircaloy written by JS. Schofield and published by . This book was released on 2002 with total page 19 pages. Available in PDF, EPUB and Kindle. Book excerpt: Critical temperatures for delayed hydride cracking (DHC) initiation during cooling (TRIT) and crack arrest during heating (TDAT) have been measured in experiments on specimens from a Zircaloy-2 electron beam weld irradiated to a fluence of 3 to 5 > 1025 n • m-2 (E > 1 MeV) and then hydrided to concentrations of 35 and 55 ppm. The experimental observations are shown to be consistent with a simple model based on previously determined hydrogen terminal solid solubility data for dissolution and precipitation, which describes conditions for sustained hydride precipitation at the crack tip. When plotted against bulk hydrogen concentration in solution, both TRIT and TDAT fall below the dissolution solvus temperature and above the precipitation solvus temperature. A key assumption in the model is that, while the local crack tip stress concentration causes local enhancement of the hydrogen concentration in solution, the hydride precipitation solvus is unaffected by stress. The good agreement obtained between measured and predicted critical temperatures provides strong support for this assumption.

Phase Field Modeling and Quantification of Zirconium Hydride Morphology

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ISBN 13 :
Total Pages : pages
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Book Synopsis Phase Field Modeling and Quantification of Zirconium Hydride Morphology by : Pierre Clement Simon

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.

Nuclear Science Abstracts

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ISBN 13 :
Total Pages : 1256 pages
Book Rating : 4.3/5 (91 download)

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Book Synopsis Nuclear Science Abstracts by :

Download or read book Nuclear Science Abstracts written by and published by . This book was released on 1973 with total page 1256 pages. Available in PDF, EPUB and Kindle. Book excerpt:

The Influence of Temperature and Yield Strength on Delayed Hydride Cracking in Hydrided Zircaloy-2

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ISBN 13 :
Total Pages : 11 pages
Book Rating : 4.:/5 (125 download)

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Book Synopsis The Influence of Temperature and Yield Strength on Delayed Hydride Cracking in Hydrided Zircaloy-2 by : P. Efsing

Download or read book The Influence of Temperature and Yield Strength on Delayed Hydride Cracking in Hydrided Zircaloy-2 written by P. Efsing and published by . This book was released on 1996 with total page 11 pages. Available in PDF, EPUB and Kindle. Book excerpt: To determine if delayed hydride cracking (DHC) can be the cause of the long axial cracks occasionally found in BWR fuel cladding, a systematic study of DHC in Zircaloy cladding has begun. In the initial stage of the project, a test technique was developed and applied to unirradiated samples of Zircaloy. The present study includes an investigation of the influence of the yield strength and temperature on the crack growth rate and the threshold stress intensity that must be exceeded before cracking begins.

Effect of Local Hydride Accumulations on Zircaloy Cladding Mechanical Properties

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ISBN 13 :
Total Pages : 20 pages
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Book Synopsis Effect of Local Hydride Accumulations on Zircaloy Cladding Mechanical Properties by : Armin Hermann

Download or read book Effect of Local Hydride Accumulations on Zircaloy Cladding Mechanical Properties written by Armin Hermann and published by . This book was released on 2007 with total page 20 pages. Available in PDF, EPUB and Kindle. Book excerpt: Mechanical response of fuel cladding with local hydride accumulations is crucial in the assessment of cladding integrity at high burn-ups. We have performed high-temperature low-strain rate burst tests on irradiated cladding samples with and without hydride lenses or blisters to seek answers to the following questions: Does the presence of a hydride lens inevitably lead to rupture at a lower pressure? How does it mechanistically affect the crack initiation and propagation? The irradiated samples in our investigation were taken from the regions of the fuel cladding with oxide spallation. Subsequently, we used neutron radiography to further select samples covering a range of hydride blister sizes on which the burst testing was performed. Rupture pressure, hoop strength, and circumferential strain data will be reported. For each sample tested, detailed metallography and fractography were performed on 2-mm size sections containing the burst opening to provide insights into the mechanism of crack initiation and propagation. Local and mean hydrogen concentrations were measured. The paper will include and elucidate new details often not fully investigated by other burst test investigations reported in the open literature. In samples with multiple blisters, the crack initiates at the largest one, which also governs the fracture mode. Reduction in the rupture pressure can be simply correlated to the reduction in sample wall thickness excluding the blister (i.e., its remaining ligament). There is a lower bound on the blister size to have any influence on the rupture pressure. Further, local plastic circumferential strain at each blister can be correlated to relative hydride lens area, as projected onto the cladding surface.

Materials Ageing and Degradation in Light Water Reactors

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Publisher : Elsevier
ISBN 13 : 0857097458
Total Pages : 441 pages
Book Rating : 4.8/5 (57 download)

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Book Synopsis Materials Ageing and Degradation in Light Water Reactors by : K L Murty

Download or read book Materials Ageing and Degradation in Light Water Reactors written by K L Murty and published by Elsevier. This book was released on 2013-02-18 with total page 441 pages. Available in PDF, EPUB and Kindle. Book excerpt: Light water reactors (LWRs) are the predominant class of nuclear power reactors in operation today; however, ageing and degradation can influence both their performance and lifetime. Knowledge of these factors is therefore critical to safe, continuous operation. Materials ageing and degradation in light water reactors provides a comprehensive guide to prevalent deterioration mechanisms, and the approaches used to handle their effects.Part one introduces fundamental ageing issues and degradation mechanisms. Beginning with an overview of ageing and degradation issues in LWRs, the book goes on to discuss corrosion in pressurized water reactors and creep deformation of materials in LWRs. Part two then considers materials’ ageing and degradation in specific LWR components. Applications of zirconium alloys in LWRs are discussed, along with the ageing of electric cables. Materials management strategies for LWRs are then the focus of part three. Materials management strategies for pressurized water reactors and VVER reactors are considered before the book concludes with a discussion of materials-related problems faced by LWR operators and corresponding research needs.With its distinguished editor and international team of expert contributors, Materials ageing and degradation in light water reactors is an authoritative review for anyone requiring an understanding of the performance and durability of this type of nuclear power plant, including plant operators and managers, nuclear metallurgists, governmental and regulatory safety bodies, and researchers, scientists and academics working in this area. Introduces the fundamental ageing issues and degradation mechanisms associated with this class of nuclear power reactors Considers materials ageing and degradation in specific light water reactor components, including properties, performance and inspection Chapters also focus on material management strategies

Mechanisms of Hydride Reorientation in Zircaloy-4 Studied in Situ

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ISBN 13 :
Total Pages : 31 pages
Book Rating : 4.:/5 (125 download)

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Book Synopsis Mechanisms of Hydride Reorientation in Zircaloy-4 Studied in Situ by : Arthur Motta

Download or read book Mechanisms of Hydride Reorientation in Zircaloy-4 Studied in Situ written by Arthur Motta and published by . This book was released on 2014 with total page 31 pages. Available in PDF, EPUB and Kindle. Book excerpt: Zirconium hydride platelet reorientation in fuel cladding during dry storage and transportation of spent nuclear fuel is an important technological issue. Using an in situ x-ray synchrotron diffraction technique, the detailed kinetics of hydride precipitation and reorientation can be directly determined while the specimen is under stress and at temperature. Hydrided Zircaloy-4 dogbone sheet samples were submitted to various thermo-mechanical schedules, while x-ray diffraction data was continuously recorded. Post-test metallography showed that nearly full hydride reorientation was achieved when the applied stress was above 210 MPa. In general, repeated thermal cycling above the terminal solid solubility temperature increased both the reoriented hydride fraction and the connectivity of the reoriented hydrides. The dissolution and precipitation temperatures were determined directly from the hydride diffraction signal. The diffraction signature of reoriented hydrides is different than that of in-plane hydrides. During cooling under stress, the precipitation of reoriented hydrides occurs at lower temperatures than the precipitation of in-plane hydrides, suggesting that applied stress suppresses the precipitation of in-plane hydrides. The analysis of the elastic strains determined by the shift in position of hydride and zirconium diffraction peaks allowed following of the early stages of hydride precipitation. Hydride particles were observed to start to nucleate with highly compressive strain. These compressive strains quickly relax to smaller compressive strains within 30°C of the onset of precipitation. After about half of the overall hydride volume fraction is precipitated, hydride strains follow the thermal contraction of the zirconium matrix. In the case of hydrides precipitating under stress, the strains in the hydrides are different in direction and trend. Analyses performed on the broadening of hydride diffraction peaks yielded information on the distribution of strains in hydride population during precipitation and cooldown. These results are discussed in light of existing models and experiments on hydride reorientation.

Is Spent Nuclear Fuel Immune from Delayed Hydride Cracking During Dry Storage? An IAEA Coordinated Research Project

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ISBN 13 :
Total Pages : 28 pages
Book Rating : 4.:/5 (125 download)

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Book Synopsis Is Spent Nuclear Fuel Immune from Delayed Hydride Cracking During Dry Storage? An IAEA Coordinated Research Project by : Christopher E. Coleman

Download or read book Is Spent Nuclear Fuel Immune from Delayed Hydride Cracking During Dry Storage? An IAEA Coordinated Research Project written by Christopher E. Coleman and published by . This book was released on 2018 with total page 28 pages. Available in PDF, EPUB and Kindle. Book excerpt: Delayed hydride cracking (DHC) has been responsible for cracking in zirconium alloy pressure tubes and fuel cladding and is a concern for spent fuel storage. For cracking to start, sufficient hydrogen must be present for hydride to form at a flaw tip and the local tensile stress must be sufficiently large to crack the hydride (a crack will not extend if the threshold in the stress intensity factor, KIH, is not exceeded. A high-temperature limit exists when the yield stress of the cladding alloy becomes too low to crack the hydride. In this paper we describe measurements of KIH and the crack growth rate, V, in unirradiated Zircaloy-4 fuel cladding containing approximately 130 ppm hydrogen in the cold-worked stress-relieved condition representing pressurized water reactors (PWRs) and pressurized heavy-water (PHWR) reactors. Four methods are used to evaluate KIH. The test specimen and fixture used in these methods was the pin-loading tension configuration. The test temperature ranged from 227 to 315°C. The mean value of KIH below 280°C had little temperature dependence; it was about 5.5 MPa?m in the PWR cladding and slightly higher at 7 MPa?m in the PHWR material. At higher test temperatures, KIH increased dramatically to more than 12 MPa?m, whereas the crack growth rate declined toward zero. This behavior suggests that unirradiated Zircaloy-4 fuel cladding is immune from DHC above about 320°C; this temperature may be increased to 360°C by irradiation. The implications for spent fuel storage are that during early storage when the temperatures are high, any flaw will not extend by DHC, whereas at low temperatures, after many years of storage, flaws would have to be very large, approaching through wall, before being extended by DHC. To date, spent nuclear fuel is not known to have failed by DHC during storage, confirming the inference.

Experimental and Analytical Investigation of the Mechanical Behavior of High-Burnup Zircaloy-4 Fuel Cladding

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ISBN 13 :
Total Pages : 22 pages
Book Rating : 4.:/5 (125 download)

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Book Synopsis Experimental and Analytical Investigation of the Mechanical Behavior of High-Burnup Zircaloy-4 Fuel Cladding by : Robert S. Daum

Download or read book Experimental and Analytical Investigation of the Mechanical Behavior of High-Burnup Zircaloy-4 Fuel Cladding written by Robert S. Daum and published by . This book was released on 2008 with total page 22 pages. Available in PDF, EPUB and Kindle. Book excerpt: Sufficient mechanical ductility of high-burnup Zircaloy-4 fuel cladding is important to prevent large-opening ruptures and significant fuel dispersal during postulated in-reactor and spent-fuel processing accidents. The effect of irradiation, oxidation, and hydriding at high fuel burnup may degrade cladding ductility to the extent that such large ruptures are possible under severe loadings. To understand this susceptibility to failure, this study focused on mechanical testing coupled with detailed finite-element modeling and analyses. Under ring-compression-type loading at room temperature, tensile cracks form within the corrosion-induced oxide layer under elastic loading. The oxide crack then propagates into the cladding wall under additional loading with little to no measurable plastic strain, as confirmed by both experiment and analyses of plastic hoop strain in the ring. For cladding with the oxide removed prior to testing at ≤1 %/s, cracking of the underlying hydride rim comprised of circumferentially oriented hydrides occurs at low plastic hoop strain (≤3 %), whereas the finite-element analysis suggests that the base alloy with a relatively small amount of hydrides appears to fail at higher strain (>8 %). At even higher strain rates (≈400 %/s), cracking within the hydride rim occurs at near-zero ductility, but the base alloy continues to remain highly ductile. These room-temperature results indicate that the hydride rim is sensitive to strain rate, whereas the base alloy is relatively not. With the precipitation of ≈100 % radially oriented hydrides, the cladding exhibits near-zero ductility at room temperature and ≈0.1 %/s. This study suggests that the ring-compression test coupled with finite-element modeling and analysis may be used to estimate crack-initiation strains in irradiated cladding materials with susceptible microstructures and under various deformation rates.

In Situ Scanning Electron Microscope Observation and Finite Element Method Analysis of Delayed Hydride Cracking Propagation in Zircaloy-2 Fuel Cladding Tubes

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ISBN 13 :
Total Pages : 23 pages
Book Rating : 4.:/5 (125 download)

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Book Synopsis In Situ Scanning Electron Microscope Observation and Finite Element Method Analysis of Delayed Hydride Cracking Propagation in Zircaloy-2 Fuel Cladding Tubes by : Toshio Kubo

Download or read book In Situ Scanning Electron Microscope Observation and Finite Element Method Analysis of Delayed Hydride Cracking Propagation in Zircaloy-2 Fuel Cladding Tubes written by Toshio Kubo and published by . This book was released on 2011 with total page 23 pages. Available in PDF, EPUB and Kindle. Book excerpt: The objective of the present research is to build a modeling method for delayed hydride cracking (DHC) of zirconium alloys. DHC tests were carried out on Zircaloy-2 cladding tubes in the chamber of a scanning electron microscope to directly observe the crack propagation and measure the crack velocity in the radial direction. These in situ observations showed that a sharply tipped crack propagated at a relatively high rate, while the velocity decreased when the crack tip was blunted, supporting the occurrence of intermittent crack propagation that could be expected from the DHC mechanism. V-KI curves or diagrams of crack velocity, V, versus stress intensity factor at a crack tip, KI, were obtained as a function of 0.2 % offset yield stress, hydride orientation, and pre-crack depth. The steady state crack velocity and the threshold stress intensity factor for the onset of the crack propagation tended to increase or decrease, respectively, with an increase in the 0.2 % offset yield stress. Analyses of stress distribution and hydrogen diffusion around a crack tip were made using a finite element computer code. The analyses showed that a strong hydrostatic pressure field was generated concentrically around the crack tip and hydrogen diffused towards the crack tip according to the hydrostatic pressure gradient. The crack velocity was estimated from the calculated hydrogen flux rate assuming the critical hydrogen quantity for the crack propagation. There was good agreement between the experiments and the calculations regarding the crack velocity and its dependency on KI. Calculations showed that the increase in the 0.2 % offset yield stress would accelerate the crack propagation by increasing the hydrostatic pressure at the crack tip.

Effect of Hydride Distribution on the Mechanical Properties of Zirconium-Alloy Fuel Cladding and Guide Tubes

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ISBN 13 :
Total Pages : 31 pages
Book Rating : 4.:/5 (125 download)

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Book Synopsis Effect of Hydride Distribution on the Mechanical Properties of Zirconium-Alloy Fuel Cladding and Guide Tubes by : Suresh K. Yagnik

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.

Zirconium in the Nuclear Industry

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Author :
Publisher : ASTM International
ISBN 13 : 0803128959
Total Pages : 891 pages
Book Rating : 4.8/5 (31 download)

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Book Synopsis Zirconium in the Nuclear Industry by : Gerry D. Moan

Download or read book Zirconium in the Nuclear Industry written by Gerry D. Moan and published by ASTM International. This book was released on 2002 with total page 891 pages. Available in PDF, EPUB and Kindle. Book excerpt: Annotation The 41 papers of this proceedings volume were first presented at the 13th symposium on Zirconium in the Nuclear Industry held in Annecy, France in June of 2001. Many of the papers are devoted to material related issues, corrosion and hydriding behavior, in-reactor studies, and the behavior and properties of Zr alloys used in storing spent fuel. Some papers report on studies of second phase particles, irradiation creep and growth, and material performance during loss of coolant and reactivity initiated accidents. Annotation copyrighted by Book News, Inc., Portland, OR.

Fundamental Nuclear Energy Research

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Publisher :
ISBN 13 :
Total Pages : 434 pages
Book Rating : 4.3/5 (91 download)

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Book Synopsis Fundamental Nuclear Energy Research by : U.S. Atomic Energy Commission. Division of Plans and Reports

Download or read book Fundamental Nuclear Energy Research written by U.S. Atomic Energy Commission. Division of Plans and Reports and published by . This book was released on 1963 with total page 434 pages. Available in PDF, EPUB and Kindle. Book excerpt: