Download or read book Effect of Pre hydriding on the Post Irradiation Ring Tensile Properties of Zircaloy 2 Cladding from Japanese Test Fuel Experiment IFA 209 written by Masaaki Uchida and published by . This book was released on 1976 with total page 48 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Effect of Hydrogen and Hydride Morphology on the Tensile Properties of Zircaloy 2 written by W. J. Babyak and published by . This book was released on 1963 with total page 54 pages. Available in PDF, EPUB and Kindle. Book excerpt:
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 Hydrogen Effects on Zircaloy 2 Tensile Properties written by H. H. Burton and published by . This book was released on 1959 with total page 108 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Statistical Analysis of Hydride Reorientation Properties in Irradiated Zircaloy 2 written by S. Valance and published by . This book was released on 2011 with total page 21 pages. Available in PDF, EPUB and Kindle. Book excerpt: The orientation of hydrides in fuel cladding determines the anisotropic fracture behavior of Zircaloy and the failure modes of cladding tubes. Approach coupling experiments using the cladding tube deformation test and finite element analysis have successfully led to the quantification of the stress influencing reorientation of hydrides in unirradiated samples. An improved version of this procedure was applied to six samples of irradiated Zircaloy-2 from two different rods with three classes of thermo-mechanical loading. It was found that at medium maximum temperature, when no more than half of the hydrides were dissolved, the mechanical loading showed no measurable effect. When most of the hydrides were dissolved, the orientation and location of the hydrides depended strongly on the mechanical loading: The hydrides spatial location followed the hoop tensile stress. When the number of loading cycles was raised, the fraction of radial hydrides increased even for very low hoop tensile stress. The inner side of the cladding showed a marked depletion of hydrides whatever the size of the hoop stress. Since our test setup involved a tri-axial stress state, the possible influence of the other components of the stress tensor was evaluated. Through the use of a classical nucleation law, it was shown that for our test setup, the hoop stress was the important mechanical quantity. Therefore, the inner side depletion of hydrides may be attributed to three other factors: Residual stress, a memory effect, and a pumping effect by the inner liner.
Download or read book The Effect of Neutron Irradiation on the Mechanical Properties of Zirconium Alloy Fuel Cladding in Uniaxial and Biaxial Tests written by DG. Hardy and published by . This book was released on 1970 with total page 43 pages. Available in PDF, EPUB and Kindle. Book excerpt: Short-time axial tension, transverse ring tension, and biaxial closed end burst tests were conducted on sections of fuel cladding from 17 different batches and heat treatments of Zircaloy 2, Zircaloy 4, and Zr-2.5Cb alloy. Specimens were irradiated at 2 to 3x1020 n/cm2, E > x MeV, at temperatures of 125 to 250 C and tested at temperatures of 20 and 300 C.
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 Effects of Hydride Precipitate Localization and Neutron Fluence on the Ductility of Irradiated Zircaloy 4 written by AM. Garde and published by . This book was released on 1996 with total page 24 pages. Available in PDF, EPUB and Kindle. Book excerpt: The ductility of highly irradiated Zircaloy-4 material was evaluated by conducting tube burst, tube tensile, and ring tensile tests on fuel cladding and guide tubes irradiated in two PWRs. The specimen fluence ranged between 9 and 12.3 x 1021 n/cm2 (E > 1 MeV), and test temperatures ranged from 313 to 673 K. The average thickness of the waterside oxide layer on the specimens ranged from 12 to 114 ?m. Specimens with an oxide thickness greater than about 100?m contained regions of spalling oxide and local areas of oxide significantly thicker than the specimen average. The corresponding average hydrogen contents ranged from 40 to 674 ppm for specimens without spalling oxide and estimated to be greater than 950 ppm with spalling. Non-uniform hydride distributions were observed in the specimens due to temperature gradients during operation.
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 Effect of Local Hydride Accumulations on Zircaloy Cladding Mechanical Properties written by Suresh K. Yagnik and published by . This book was released on 2009 with total page 22 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.
Download or read book High pressurization rate burst test of hydrided Zircaloy 4 fuel cladding at room temperature written by Fumihisa Nagase and published by . This book was released on 1998 with total page 25 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book High pressurization rate burst test of hydrided zircaloy 4 fuel cladding at 620 K written by Fumihisa Nagase and published by . This book was released on 2000 with total page 31 pages. Available in PDF, EPUB and Kindle. Book excerpt:
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 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 Impact of Irradiation Damage Recovery During Transportation on the Subsequent Room Temperature Tensile Behavior of Irradiated Zirconium Alloys written by B. Bourdiliau and published by . This book was released on 2010 with total page 25 pages. Available in PDF, EPUB and Kindle. Book excerpt: Zirconium alloys are commonly used in pressurized water reactor as fuel rod cladding tubes. After irradiation and cooling in pool, the spent nuclear fuel assemblies are either transported for wet storage to a devoted site or loaded in suitable casks for dry storage on a nuclear power plant site. During dry transportation or at the beginning of dry storage, at temperatures around 400°C, the cladding experiences a creep deformation under the hoop stress induced by the internal pressure of the fuel rod. During creep, a recovery of the radiation damage can occur that can affect the subsequent mechanical properties. The mechanical behavior of the cladding has been investigated in laboratory on two neutron irradiated cladding materials: fully recrystallized Zr-1 % Nb and stress-relieved Zircaloy-4. Creep tests under internal pressure were conducted at 400 and 420°C. After depressurization and cooling, ring tensile tests were carried out at room temperature. In addition, transmission electron microscopy observations have been performed after testing. The post-creep mechanical response exhibited a decrease of the strength compared to the as-irradiated material. This decrease is associated with a significant recovery of the ductility, which becomes close to the ductility of the unirradiated material. The transmission electron microscopy examinations, conducted on recrystallized Zr-1 % Nb ring samples, revealed that the radiation defects have been annealed. It was also observed that as for the unirradiated material, the deformation occurred homogeneously throughout the grains. No dislocation channeling was observed contrary to the as-irradiated material. These observations explain the recovery of the strength and of the ductility after post-irradiation creep that may also occur during dry transportation or at the beginning of dry storage.
Download or read book Effects of Cold Work and Neutron Irradiation on the Tensile Properties of Zircaloy 2 written by Arden Lee Bement (Jr.) and published by . This book was released on 1963 with total page 214 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Evaluating Strength and Ductility of Irradiated Zircaloy Task 5 written by and published by . This book was released on 1981 with total page 452 pages. Available in PDF, EPUB and Kindle. Book excerpt:
