Download or read book Mechanical Properties of Zircaloy 4 PWR Fuel Cladding with Burnup 54 64MWd kgU and Implications for RIA Behavior written by J. Desquines and published by . This book was released on 2005 with total page 22 pages. Available in PDF, EPUB and Kindle. Book excerpt: The PROMETRA material testing program is a support program related to the study of high burnup fuel rod behavior under Reactivity Initiated Accidents (RIA) and to the interpretation of the CABRI REP-Na RIA test results. Hoop and axial tensile tests have been performed on fresh and irradiated Zircaloy-4 cladding alloy first at CEA Grenoble hot labs and now at CEA Saclay in order to assess the cladding mechanical behavior during RIA transients. Efforts have been continuously carried out in order to improve the prototipicallity of the tests for RIA studies involving new specimens and new testing techniques. The corrosion level of irradiated specimens reached up to 130 ?m of oxide layer thickness. The influence of in-pile oxide layer spallation has also been addressed. High strain-rate material properties of irradiated Zircaloy-4 and the consequences of hydride embrittlement can be derived from the PROMETRA program.

Download or read book Comprehensive Nuclear Materials written by and published by Elsevier. This book was released on 2020-07-22 with total page 4871 pages. Available in PDF, EPUB and Kindle. Book excerpt: Materials in a nuclear environment are exposed to extreme conditions of radiation, temperature and/or corrosion, and in many cases the combination of these makes the material behavior very different from conventional materials. This is evident for the four major technological challenges the nuclear technology domain is facing currently: (i) long-term operation of existing Generation II nuclear power plants, (ii) the design of the next generation reactors (Generation IV), (iii) the construction of the ITER fusion reactor in Cadarache (France), (iv) and the intermediate and final disposal of nuclear waste. In order to address these challenges, engineers and designers need to know the properties of a wide variety of materials under these conditions and to understand the underlying processes affecting changes in their behavior, in order to assess their performance and to determine the limits of operation. Comprehensive Nuclear Materials, Second Edition, Seven Volume Set provides broad ranging, validated summaries of all the major topics in the field of nuclear material research for fission as well as fusion reactor systems. Attention is given to the fundamental scientific aspects of nuclear materials: fuel and structural materials for fission reactors, waste materials, and materials for fusion reactors. The articles are written at a level that allows undergraduate students to understand the material, while providing active researchers with a ready reference resource of information. Most of the chapters from the first Edition have been revised and updated and a significant number of new topics are covered in completely new material. During the ten years between the two editions, the challenge for applications of nuclear materials has been significantly impacted by world events, public awareness, and technological innovation. Materials play a key role as enablers of new technologies, and we trust that this new edition of Comprehensive Nuclear Materials has captured the key recent developments. Critically reviews the major classes and functions of materials, supporting the selection, assessment, validation and engineering of materials in extreme nuclear environments Comprehensive resource for up-to-date and authoritative information which is not always available elsewhere, even in journals Provides an in-depth treatment of materials modeling and simulation, with a specific focus on nuclear issues Serves as an excellent entry point for students and researchers new to the field

Download or read book Oxidation and the Testing of Turbine Oils written by Cyril A. Migdal and published by ASTM International. This book was released on 2008 with total page 929 pages. Available in PDF, EPUB and Kindle. Book excerpt: This work presents papers from a December 2005 symposium held in Norfolk, Virginia, and sponsored by ASTM Committee D2 on Petroleum Products and Lubricants and its Subcommittees D02.09 on Oxidation and D02.C0 on Turbine Oils. Contributors include equipment manufacturers, end users, lubricant producers, lubricant additive suppliers, test equipment manufacturers, and standard test method developers. They share information on industry trends, evolving technologies, and changing equipment designs and operating conditions, with a focus on how these factors impact oxidation. Some topics covered include turbine oil performance limits, a new form of the rotating pressure vessel oxidation test, and degradation mechanisms leading to sludge and varnish in modern turbine oil formulations. B&w photos are included. There is no subject index. Migdal is affiliated with Chemtura Corporation.

Download or read book Handbook of Nuclear Engineering written by Dan Gabriel Cacuci and published by Springer Science & Business Media. This book was released on 2010-09-14 with total page 3701 pages. Available in PDF, EPUB and Kindle. Book excerpt: This is an authoritative compilation of information regarding methods and data used in all phases of nuclear engineering. Addressing nuclear engineers and scientists at all levels, this book provides a condensed reference on nuclear engineering since 1958.

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.

Download or read book Behavior of Irradiated Zircaloy 4 Fuel Cladding Under Simulated LOCA Conditions written by T. Takahashi and published by . This book was released on 2000 with total page 21 pages. Available in PDF, EPUB and Kindle. Book excerpt: High-temperature oxidation and mechanical strength under a loss-of-coolant accident (LOCA) were investigated using irradiated and unirradiated fuel cladding. Cladding from fuel rods irradiated up to 49 GWD/MTU in a Japanese commercial PWR and unirradiated cladding, most of which was preoxidized and precharged with hydrogen to simulate high burnup fuel, were subjected to the tests. High-temperature oxidation tests showed that the oxidation weight gain for irradiated cladding was equal to, or slightly lower than, that for unirradiated material. Preoxidized cladding showed less oxidation weight gain, and no effect of hydrogen absorption on oxidation behavior was observed. The mechanical tests (uniaxial strength and ductility) after a thermal sequence simulating a LOCA showed comparable behavior with that of unirradiated cladding, due to the recovery of the irradiated microstructure. These test results suggest: (1) there is no adverse effect of irradiation on the high-temperature oxidation behavior; and (2) radiation damage in cladding is eliminated during a LOCA condition.

Download or read book Deformation Characteristics of Cold Worked and Recrystallized Zircaloy 4 Cladding written by DL. Baty and published by . This book was released on 1984 with total page 34 pages. Available in PDF, EPUB and Kindle. Book excerpt: Thermomechanical processing of Zircaloy-4 cladding plays a major role in determining its deformation behavior. Crystallographic texture and the related anisotropy in mechanical properties of Zircaloy-4 have been shown to be affected by different processing paths. In this program, the deformation behavior of four Zircaloy-4 cladding types was evaluated in laboratory and in-reactor studies under typical pressurized-water reactor (PWR) conditions. In particular, the creep behavior, stress-free growth, and mechanical property changes of these materials were examined.

Download or read book Mechanical Properties of Irradiated Zirconium Zircaloy and Aluminum written by Richard E. Schreiber and published by . This book was released on 1961 with total page 112 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Preliminary Investigation of Zircaloy 4 as a Research Reactor Cladding Material written by and published by . This book was released on 2012 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: As part of a scoping study for the ATR fuel conversion project, an initial comparison of the material properties of Zircaloy-4 and Aluminum-6061 (T6 and O-temper) is performed to provide a preliminary evaluation of Zircaloy-4 for possible inclusion as a candidate cladding material for ATR fuel elements. The current fuel design for the ATR uses Aluminum 6061 (T6 and O temper) as a cladding and structural material in the fuel element and to date, no fuel failures have been reported. Based on this successful and longstanding operating history, Zircaloy-4 properties will be evaluated against the material properties for aluminum-6061. The preliminary investigation will focus on a comparison of density, oxidation rates, water chemistry requirements, mechanical properties, thermal properties, and neutronic properties.

Download or read book Effect of Irradiation at 588 K on Mechanical Properties and Deformation Behavior of Zirconium Alloy Strip written by J. Baicry and published by . This book was released on 1987 with total page 19 pages. Available in PDF, EPUB and Kindle. Book excerpt: Irradiation tests conducted for the Advanced Fuel Assembly (AFA) grid development allowed the measurement of the mechanical characteristics, growth, and relaxation of the following materials: stress-relieved, ?-annealed, and ?-quenched Zircaloy-4, and ?-annealed Zr-3Sn-1Mo. The mechanical characteristics of the zirconium alloys approached a limit value as early as 5 x 1024 n m-2, except for the ?-quenched Zircaloy-4 for which the limit value was reached later at about 5 x 1025 n m-2. Uniform elongation at 588 K showed a minimum at about 5 x 1024 n m-2; beyond this fluence, stress-relieved Zircaloy-4 presented the most marked tendency to increase. Longitudinal direction growth of annealed Zircaloy-4 and longitudinal and transverse direction growth of stress-relieved Zircaloy-4 may be expressed by the formula ? = A(?t)n where n = 0.67 (stress-relieved Zircaloy-4) and n = 0.4 (annealed Zircaloy-4). It is doubtful that the growth is associated with a density change. The stress-relaxation of the zirconium base materials is nearly complete for 4 x 1025 n m-2, with the exception of Zr-1Nb which has an intermediate behavior between those of Zircaloy-4 and Inconel 718.

Download or read book On the Embrittlement of Zircaloy 4 Under RIA Relevant Conditions written by AT. Motta and published by . This book was released on 2002 with total page 18 pages. Available in PDF, EPUB and Kindle. Book excerpt: The extended use of Zircaloy cladding in light water reactors degrades its mechanical properties by a combination of irradiation embrittlement, coolant-side oxidation, hydrogen pickup, and hydride formation. The hydrides are usually concentrated in the form of a dense layer or rim near the cooler outer surface of the cladding. Utilizing plane-strain ringstretch tests to approximate the loading path in a reactivity-initiated accident (RIA) transient, we examined the influence of a hydride rim on the fracture behavior of unirradiated Zircaloy-4 cladding at room temperature and 300°C. Failure is sensitive to hydride-rim thickness such that cladding tubes with a hydride-rim thickness >100 ?m (?700 wppm total hydrogen) exhibit brittle behavior, while those with a thickness

Download or read book On the Embrittlement of Zircaloy 4 Under RIA relevant Conditions written by and published by . This book was released on 2002 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The extended use of Zircaloy cladding in light water reactors degrades its mechanical properties by a combination of irradiation embrittlement, coolant-side oxidation, hydrogen pickup, and hydride formation. The hydrides are usually concentrated in the form of a dense layer or rim near the cooler outer surface of the cladding. Utilizing plane-strain ring-stretch tests to approximate the loading path in a reactivity-initiated accident (RIA) transient, we examined the influence of a hydride rim on the fracture behavior of unirradiated Zircaloy-4 cladding at room temperature and 300 C. Failure is sensitive to hydride-rim thickness such that cladding tubes with a hydride-rim thickness>100[micro]m ([approx]700 wppm total hydrogen) exhibit brittle behavior, while those with a thickness

Download or read book Zircaloy 4 Cladding Deformation During Power Reactor Irradiation written by DG. Franklin and published by . This book was released on 1982 with total page 33 pages. Available in PDF, EPUB and Kindle. Book excerpt: The four primary Zircaloy fuel cladding deformation phenomena--axial elongation, circumferential creep, ovalization, and ridging--have been investigated for fuel irradiated in four modern pressurized water reactors. The axial elongation of fueled and nonfueled rods is examined by a regression fit for dependence on fluence, clad texture, yield stress, applied stress and, for fuel rods, fuel pellet length to diameter ratio. For fueled rods, only fluence and stress are found to be important, although the range of texture data is small. For nonfueled rods, the texture is found to influence elongation.

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 Structure of High burnup fuel Zircaloy Cladding PWR BWR written by and published by . This book was released on 1983 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Zircaloy cladding from high-burnup (> 20 MWd/kg U) fuel rods in light-water reactors is characterized by a high density of irradiation-induced defects (RID), compositional changes (e.g., oxygen and hydrogen uptake) associated with in-service corrosion, and geometrical changes produced by creepdown, bowing, and irradiation-induced growth. During a reactor power transient, the cladding is subject to localized stress imposed by thermal expansion of the cracked fuel pellets and to mechanical constraints imposed by pellet-cladding friction. As part of a program to provide a better understanding of brittle-type failure of Zircaloy fuel cladding by pellet-cladding interaction (PCI) phenomenon, the stress-rupture properties and microstructural characteristics of high-burnup spent fuel cladding have been under investigation. This paper reports the results of the microstructural examinations by optical microscopy, scanning (SEM), 100-keV transmission (TEM), and 1 MeV high-voltage (HVEM) electron microscopies of the fractured spent fuel cladding with a specific empahsis on a correlation of the structural characteristics with the fracture behavior.

Download or read book Mechanical Behavior of Zircaloy 4 Cladding Under Simulated LOCA Conditions written by Jeremy K. Jones and published by . This book was released on 1978 with total page 168 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Corrosion Performance of Zircaloy 2 and Zircaloy 4 PWR Fuel Cladding written by T. Andersson and published by . This book was released on 1989 with total page 14 pages. Available in PDF, EPUB and Kindle. Book excerpt: A variety of Zircaloy-2 and Zircaloy-4 cladding tubes with different microstructures has been manufactured. Test rods have been included in two fuel assemblies, and the assemblies have been inserted into the Ringhals 3 pressurized water reactor. Samples from the same cladding tube variants have been subjected to steam autoclave testing in the range 400 to 500°C for different exposure times and have been characterized with respect to their microstructure.