Download or read book Characterization of the Microstructure in Recrystallized Zircaloy 2 Cladding Irradiated to a High Neutron Dose written by Kjell Pettersson and published by . This book was released on 2003 with total page 44 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Deformation and Fracture Properties of Neutron Irradiated Recrystallized Zircaloy 2 Cladding Under Uniaxial Tension written by T. Yasuda and published by . This book was released on 1987 with total page 14 pages. Available in PDF, EPUB and Kindle. Book excerpt: Sufficient evaluation of the changes in mechanical properties, such as elastic, plastic, and failure properties, due to neutron irradiation in service is required to precisely predict fuel performance. This paper presents the results of the uniaxial tensile tests performed for recrystallized (850 K, 2.5 h) Zircaloy-2 claddings irradiated in commercial BWRs to fluences of 5 x 1023 to 4 x 1025 n/m2 (E > 1 MeV). The material constants of irradiated Zircaloy-2 were obtained precisely, using a high temperature elongation detector in a hot-cell and computer analyses of digital stress-strain data.

Download or read book Irradiation Induced Growth and Microstructure of Recrystallized Cold Worked and Quenched Zircaloy 2 NSF and E635 Alloys written by D. W. White and published by . This book was released on 2008 with total page 19 pages. Available in PDF, EPUB and Kindle. Book excerpt: This paper is devoted to the study of the effect of the texture, phase composition, and microstructure on the irradiation-induced growth strain (GS) of zirconium-based alloys. GS measurements and TEM microstructural examinations were performed on Zry-2, NSF, and E635 samples in the recrystallized, beta quenched and cold-worked (CW) conditions. The samples were irradiated in the BOR-60 reactor in the temperature range of 315-325°C up to a neutron fluence level of 1.1 x 1026 n/m2 (E>1MeV), i.e., up to a damage dose of 23 dpa. Growth strains of NSF and E635 alloys in all states and in the longitudinal and transverse directions are lower as compared to those of Zry-2, and do not exceed 0.2 % even at the maximum fluence level. As for recrystallized Zry-2, the GS kinetics are characterized by the appearance of the accelerated growth stage. A combination of a certain amount of Nb, Fe, and Sn in the matrix content plays a key role in GS kinetics. The higher the degree of CW, the higher the irradiation growth but its rate of increase with increasing fluence is different for alloys of different compositions. The maximum GS, reaching 0.72 %, is observed in the 20 % CW Zry-2 samples. Texture, along with the alloy composition, is one of the main GS-determining factors. Irradiation growth of the transversal samples is lower as compared to the longitudinal ones because of texture. As for quenched alloys, the texture is practically isotropic and GS values are low, independent of the alloy composition. In CW materials, the density of ‹c›- dislocations greatly affects the irradiation growth strain. Particles of Zr(Fe,Cr)2 and Zr2(Fe,Ni) phases in Zry-2 as well as Zr(Nb,Fe)2 in NSF and E635 are depleted in iron under irradiation. The Fe goes into the matrix and modifies its properties. The HCP lattice structure in the Laves phases in NSF and E635 changes into BCC (?-Nb-type). FCC (Zr,Nb)2Fe precipitates preserve on the whole their composition and structure; no amorphization of the Nb-containing precipitates is observed. The Zr2(Fe,Ni) precipitates with a BCT lattice remain crystalline, and HCP Zr(Cr,Fe)2 precipitates undergo amorphization. The average particle size in the irradiated alloys is larger and the concentration is a little lower as compared to the unirradiated ones. Irradiation-induced fine dispersed precipitates about 3 nm in size, probably enriched in niobium, appear in NSF and E635. The observed changes of microhardness are discussed from the viewpoint of generation of radiation defects (clusters, dislocation loops), evolution of the initial dislocation structure, and matrix composition (enrichment in Fe, Cr, and, probably, Nb).

Download or read book Effects of High Neutron Fluences on Microstructure and Growth of Zircaloy 4 written by F. Garzarolli and published by . This book was released on 1989 with total page 17 pages. Available in PDF, EPUB and Kindle. Book excerpt: Irradiation of Zircaloy affects its microstructure and macroscopical properties, for example, influencing its irradiation growth. To gain more insight into these phenomena, experimental fuel rods and growth specimens with various fabrication parameters were irradiated in a pressurized water reactor (PWR) to high fluences. Some of the growth specimens were exposed to a fast neutron fluence of up to 2.3 x 1022 cm-2 (?0.82 MeV) over a period of 10 years. Following exposure, the irradiation-induced alterations of the microstructure and the intermetallic precipitates were studied by optical microscopy (OM), scanning electron microscopy (SEM), and transmission electron microscopy (TEM). At a temperature of 300°C during irradiation to fluences up to 7 x 1021 cm-2, growth increases with increasing yield strength. Recrystallized material, which has a low yield strength, exhibits an increased growth rate at very high fluences (?1 x 1022 cm-2). Postirradiation annealing studies indicate that the early irradiation growth of the recrystallized material can be recovered, whereas the later accelerated growth does not seem to be recoverable.

Download or read book Effects of Recrystallization and Neutron Irradiation on Creep Anisotropy of Zircaloy Cladding written by ST. Mahmood and published by . This book was released on 1991 with total page 20 pages. Available in PDF, EPUB and Kindle. Book excerpt: Zirconium alloys have hcp crystal structure with low c/a ratio at and below the reactor operating temperatures and exhibit preferred orientations or textures. These textures result in anisotropic mechanical properties, which in-turn affect their in-service behavior such as in-pile creep down of the cladding tubes. The purpose of the present study has been to investigate the effects of recrystallization and neutron irradiation on the anisotropic biaxial creep behavior of Zircaloy cladding tubes. The creep anisotropies of the cold-worked and recrystallized tubes were considered using results from the closed-end internal pressurization tests superimposed with axial loading. The in situ biaxial strain measurements were made using laser and linear variable differential transformer (LVDT) extensometers. Creep data were obtained at various stress ratios, and creep loci were constructed at constant energy dissipation for both cold-worked and recrystallized tubes. X-ray diffraction techniques were used to measure the textures which were then described quantitatively in terms of crystallite orientation distribution functions (CODFs). These CODFs were employed to predict the anisotropy parameters R and P, and the anisotropic creep behavior. The creep behavior of Zircaloy tubes changed with recrystallization. The effect of neutron irradiation on the recrystallized material is modeled by invoking secondary slip systems. The considerable amount of plastic anisotropy observed in the unirradiated recrystallized tubes shows a tendency to decrease and to become isotropic at high fluences. On the other hand, neutron irradiation does not produce any significant changes in the anisotropy of the cold-worked material when radiation growth is taken into consideration.

Download or read book Microstructure Evolutions and Iron Redistribution in Zircaloy Oxide Layers written by X. Iltis and published by . This book was released on 1996 with total page 23 pages. Available in PDF, EPUB and Kindle. Book excerpt: To understand the acceleration of the Zircaloy corrosion kinetics in PWR conditions, TEM microstructural characterizations of oxide layers grown in an autoclave or directly in-reactor have been performed. To separate the influence on the oxidation process of the irradiation damage in the alloy from the dynamic effect of neutron flux, oxide layers have also been grown in an autoclave on previously neutron-irradiated cladding. The comparative characterization of these oxide layers leads to the following results: the nucleation and growth process are observed to be similar on oxides formed in-autoclave and significantly different on oxides grown directly in-reactor, indicating that this process is essentially affected by neutron irradiation or, more generally, parameters specific to the reactor environment. Concerning grain growth phenomena, it appears that the high microstructural instability noticed in oxides formed in-reactor is also the consequence of parameters specific to the reactor environment such as neutron irradiation or the lithium concentration gradient. Finally, the iron distribution in the oxide is almost the direct image of the iron distribution in the metal.

Download or read book Effects of Neutron Irradiation on the Microstructure of Alpha Annealed Zircaloy 4 written by BF. Kammenzind and published by . This book was released on 2002 with total page 27 pages. Available in PDF, EPUB and Kindle. Book excerpt: Analytical electron microscopy (AEM) was used to study the separate effects of the irradiation parameters on the evolution of the microstructure in recrystallized alpha-annealed Zircaloy-4 under controlled irradiation conditions. The effects of fast neutron flux from ~4 x 1013 n/cm2-s to ~1.5 x 1014 n/cm2-s (E > 1 MeV)3 neutron fluence in the range of ~15 x 1020 n/cm2 to ~50 x 1020 n/cm2 and temperature from ~270 to ~330°C were studied. The completeness of the test matrix and the exposure in the controlled environment of the advanced test reactor permitted the separate effects of fast neutron flux, fluence, and irradiation temperature to be delineated for the first time. It was found that an increase in the neutron flux increases the degree of amorphization of the second-phase precipitates but retards the redistribution of iron out of the amorphous region (neutron fluence and irradiation temperature remaining the same), whereas increasing temperature (neutron flux and neutron fluence remaining the same) has a reverse effect. Overall, the rate of amorphization of the second-phase precipitates observed in this work was larger than that predicted by many existing literature models. Finally, neither segregation of alloying elements to grain boundaries nor precipitation of any new phases were encountered.

Download or read book Post Irradiation Characterization of Ultra High Fluence Zircaloy 2 Plate written by ST. Mahmood and published by . This book was released on 2000 with total page 31 pages. Available in PDF, EPUB and Kindle. Book excerpt: Zircaloy-2 plate with neutron exposure more than 3 x 1026 n/m2 (E > 1 MeV) has been tested to study the effects of ultra-high neutron fluence on the mechanical and physical behavior of Zircaloy-2.

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.

Download or read book High Fluence Irradiation Growth of Zirconium Alloys at 644 to 725 K written by RB. Adamson and published by . This book was released on 1984 with total page 23 pages. Available in PDF, EPUB and Kindle. Book excerpt: Irradiation growth behavior of zirconium, Zircaloy-2 and Zircaloy-4,Zr-2.5Nb, and Zr-3.5Sn-0.8Mo-0.8Nb (EXCEL) was studied on specimens irradiated in the Experimental Breeder Reactor II (EBR-II) to fluences of 1.2 to 16.9 x 1025 neutrons (n).m-2 (E > 1 MeV) in the temperature range 644 to 725 K. In Zircaloy, growth and growth rate were observed to increase continuously with fluence up to 16.9 x 1025 n.m-2 with no indication of saturation in either recrystallized or cold-worked materials. Positive growth strains of 1.5% and negative strains of approximately 2% to 2.5% were observed in both recrystallized and cold-worked Zircaloy. The formation of both a-type loops and c component dislocations is recrystallized Zircaloy under irradiation appears to be the basis in this material for growth strains similar in magnitude to those in cold-worked Zircaloy. Alloy additions to zirconium can increase growth by as much as an order of magnitude for a given texture at the higher irradiation temperatures and fluences. A sharp change to increasing growth rate with temperature occurs in Zircaloy at ~670 K, with a similar trend indicated for the other alloys. Although growth in all these alloys is a strong function of crystallographic texture, an exact (1-3f) type of dependence is not always apparent. In Zr-2.5Nb the dependence of growth on texture appears to be masked by the precipitation of betaniobium, with a transition to a well-defined texture dependence being a function of fluence and temperature. Significant differences in growth behavior were observed in nominally similar Zircaloys, apparently due to minor microstructural or chemical differences.

Download or read book Zirconium in the Nuclear Industry written by George P. Sabol and published by ASTM International. This book was released on 1996 with total page 907 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Nuclear Science Abstracts written by and published by . This book was released on 1975 with total page 1090 pages. Available in PDF, EPUB and Kindle. Book excerpt:

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 Evolution of Microstructure in Zirconium Alloys During Irradiation written by M. Griffiths and published by . This book was released on 1996 with total page 23 pages. Available in PDF, EPUB and Kindle. Book excerpt: X-ray diffraction (XRD) and transmission electron microscopy (TEM) have been used to characterize microstructural and microchemical changes produced by neutron irradiation in zirconium and zirconium alloys. Zircaloy-2, Zircaloy-4, and Zr-2.5Nb alloys with differing metallurgical states have been analyzed after irradiation for neutron fluences up to 25 x 1025 n.m-2 (E > 1 MeV) for a range of temperatures between 330 and 580 K.

Download or read book Characteristics of Autoclave and In reactor Nodular Corrosion of Zircaloys written by and published by . This book was released on 1990 with total page 54 pages. Available in PDF, EPUB and Kindle. Book excerpt: Nodular corrosion characteristics of recrystallized Zircaloy-4 were investigated in static autoclave tests at 500°C and 10.3 MPa. The roles of annealing temperature, cooling rate after beta-treating at 1050°C, cold work, and surface treatment in corrosion tests were correlated with the results of microstructural characterization by scanning and transmission electron microscopies. A good correlation was obtained between average size of intermetallic precipitates and weight gain, in contrast to nodule coverage and nodule number density. These results could be best explained by the hypothesis that nodules nucleate in local regions that are depleted of Fe and Cr alloying elements. Some observations were inconsistent with the premise that nodules nucleate on or near intermetallic precipitates. Nodular corrosion characteristics and microstructures of commercial Zircaloy-2 cladding of fuel and gadolinia rods, obtained from several BWRs after burnup of 11--30 MWd/kgU, were also examined. Partial amorphization of intermetallic precipitates in BWR Zircaloy-2, and virtual dissolution and in an extreme case spinodal- like fluctuations of dissolved alloying elements in PWR Zircaloy-4 cladding were observed. Occurrence of nodular oxidation of Zircaloy-2 in BWRs could best be correlated to average size of intermetallic precipitates before irradiation and to fuel cladding operating temperature. For an intermetallic size range of 250--700 nm, nodular oxides were observed at 288°C, but only thick uniform oxide was observed at 307°C. 53 refs., 14 figs., 1 tab.

Download or read book Microstructural Development in Neutron Irradiated Zircaloy 4 written by WJS Yang and published by . This book was released on 1990 with total page 15 pages. Available in PDF, EPUB and Kindle. Book excerpt: Zircaloy-4, a zirconium base alloy used extensively as cladding and core structural material in water cooled nuclear reactors, was examined by transmission electron microscopy after neutron irradiation and postirradiation annealing. Phase instabilities found during irradiation include the amorphous transformation and the dissolution of intermetallic precipitate Zr(Fe,Cr)2 in the ?-recrystallized matrix and the dissolution of the metastable precipitate Zr4(Fe,Cr) in the ?-quenched matrix. The alloy is driven toward a single phase solid solution during the irradiation. The presence of fast diffusion iron species in the matrix due to the precipitate dissolution may have caused the irradiation growth breakaway phenomenon. The microstructural evolution during irradiation consists of ̄c dislocation development and grain boundary migration. The presence of ̄c dislocations indicates permanent strain in the matrix. The postirradiation annealing at 833 K does not anneal out the ̄c dislocations. The ̄c dislocation is postulated to have developed due to the intergranular constraints under the continuous growth in the breakaway region.

Download or read book Localized Ductility of Irradiated Zircaloy 2 Cladding in Air and Iodine Environments written by DS. Tomalin and published by . This book was released on 1977 with total page 16 pages. Available in PDF, EPUB and Kindle. Book excerpt: A laboratory testing method which is capable of imposing plane-stress and localized plane-strain loading conditions on tubing specimens was used to examine effects of annealing temperature and test environment on the strength and ductility of neutron irradiated Zircaloy-2 tubing. Tests were performed at 350°C in both air and iodine environments on specimens previously irradiated at 327°C in a helium atmosphere to fluences ranging between 1.26 and 1.60 x 1021 neutrons (n)/cm2 (E > 1 MeV). The tensile ductility under plane-strain loading conditions depended on preirradiation annealing temperature, with recrystallized specimens showing the largest plastic strain at maximum load (6 to 10 percent) in air. By way of comparison, stress-relieved specimens showed the largest plastic strain at maximum load (4 to 5 percent) under conditions of plane-stress loading in air. An iodine environment reduced the load-carrying capacity of the cladding (plane-strain loading) and lowered the strain at maximum load to near the detection limit of the test technique (approximately 1 percent), regardless of prior heat treatment. A change in failure mode from dimpled ductile rupture to quasicleavage and fluting accompanied this loss of ductility.