Download or read book Effect of Alloying Elements Cold Work and Hydrogen on the Irradiation Induced Growth Behavior of Zirconium Alloy Variants written by Suresh Yagnik and published by . This book was released on 2018 with total page 48 pages. Available in PDF, EPUB and Kindle. Book excerpt: In-reactor dimensional changes in zirconium-based alloys result from a complex interplay of many factors, such as (1) alloy type and composition, including the addition of elements such as niobium, iron, and tin; (2) fabrication process, including cold work, texture, and residual stresses; (3) irradiation temperature; and (4) hydrogen levels. In many cases, the observed dimensional changes in light water reactor fuel-assembly components--especially at high exposures--cannot be fully explained based on current growth and creep models. Therefore, a systematic approach was taken in this multiyear (2005-2011) Nuclear Fuel Industry Research Program investigation. The objective was to measure stress-free irradiation-induced growth (IIG) of specially fabricated alloys through irradiation under controlled conditions in the BOR-60 fast-flux test reactor up to a high fluence of approximately 2 x 1026 m-2 (E > 1 MeV)--equivalent to maximum of approximately 37 dpa exposure--followed by postirradiation examinations (PIEs). Irradiation temperature was within a narrow temperature range (320 ± 10°C). The PIEs included dimensional-change and microhardness measurements, metallography and hydride etching, and scanning transmission electron microscopy (STEM) or transmission electron microscopy (TEM).

Download or read book Mechanical and Creep Behavior of Advanced Materials written by Indrajit Charit and published by Springer. This book was released on 2017-02-04 with total page 299 pages. Available in PDF, EPUB and Kindle. Book excerpt: This collection commemorates the occasion of the honorary symposium that celebrated the 75th birthday and lifelong contributions of Professor K.L. Murty. The topics cover the present status and recent advances in research areas in which he made seminal contributions. The volume includes articles on a variety of topics such as high-temperature deformation behaviors of materials (elevated temperature creep, tensile, fatigue, superplasticity) and their micromechanistic interpretation, understanding mechanical behavior of HCP metals/alloys using crystallographic texture, radiation effects on deformation and creep of materials, mechanical behavior of nanostructured materials, fracture and fracture mechanisms, development and application of small-volume mechanical testing techniques, and general structure-property correlations.

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 Proceedings of the USAEC Symposium on Zirconium Alloy Development Castlewood Pleasanton California November 12 14 1962 written by and published by . This book was released on 1962 with total page 428 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book The effect of cold work and stress relieving on the irradiation growth behaviour of zirconium alloys written by V. Fidleris and published by . This book was released on 1973 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt:

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 Irradiation Growth of Zirconium Alloys written by JY. Ren and published by . This book was released on 1994 with total page 8 pages. Available in PDF, EPUB and Kindle. Book excerpt: Experimental investigation of irradiation growth on annealed Zircaloy-4 and 20% to 50% cold-worked Zr-2.5wt%Nb specimens with stress relief has been carried out. The specimens are irradiated in a heavy water reactor at 610 K to 4.2 x 1024 n/m2 (E > 1.0 MeV). The growth strains increase linearly with fluence. The saturation of growth is not observed for all specimens. The difference of growth behavior between two kinds of Zircaloy-4 tube may be associated with the content of minor alloying elements and impurities that influence the microstructure evolution under irradiation.

Download or read book Specific Zirconium Alloy Design Program Quarterly Progress Report written by and published by . This book was released on 1962-07 with total page 68 pages. Available in PDF, EPUB and Kindle. Book excerpt:

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

Download or read book Effect of Hydrogen on Irradiation Creep and Growth for ZIRLO Alloy and Zr 1 0Nb written by John Paul Foster and published by . This book was released on 2018 with total page 23 pages. Available in PDF, EPUB and Kindle. Book excerpt: The impact of hydrogen on the irradiation growth and creep of stress-relief annealed (SRA) ZIRLO® alloy and recrystallized annealed (RXA) Zr-1.0Nb cladding tubes is evaluated in this paper. The samples were charged with hydrogen in the range of approximately 160-720 ppm using the gaseous method. Biaxial in-reactor creep tests were performed after Cycle 1, Cycle 2, Cycle 3, and Cycle 4, on both as-received and precharged hydrogen cladding tubes. Outside diameter and axial length measurements were performed on the samples. The results showed that hydrogen had no effect on the axial irradiation creep but a relatively large effect on the axial irradiation growth. Increasing hydrogen decreased the axial irradiation growth in RXA Zr-1.0 Nb, which was opposite from the behavior of SRA ZIRLO cladding. This unique hydrogen effect on the irradiation axial growth of RXA Zr-1.0Nb could be due to the different intergranular stress resulting from the fabrication process or the absence of alloying elements, including tin. The total axial strain of both Zr-1.0Nb and SRA ZIRLO cladding increased with increasing fast fluence, and Zr-1.0Nb increased at a faster rate relative to SRA ZIRLO cladding. In the diameter direction, hydrogen had a minimal effect on the total diameter strain and the diameter irradiation creep strain for both SRA ZIRLO samples and the RXA Zr-1.0Nb sample. This finding from in-reactor test is contrary to the out-reactor tests results from the literature that have shown that hydrogen significantly decreases thermal creep. The total diameter strain and diameter irradiation creep behavior for the SRA ZIRLO samples and RXA Zr-1.0Nb were similar.

Download or read book Understanding Irradiation Growth Through Atomistic Simulations written by Mikael Christensen and published by . This book was released on 2018 with total page 31 pages. Available in PDF, EPUB and Kindle. Book excerpt: Irradiation-induced structural changes of ?-zirconium alloys and in particular the effect of iron were investigated by molecular dynamics simulations using embedded atom potentials derived from first-principles calculations. The simulations revealed that at temperatures between 500 and 600 K self-interstitial atoms (SIAs) diffuse rapidly in a cooperative movement, preferably parallel to basal planes (a directions; a), forming nanoclusters with an extension in a and c. Vacancies diffuse more slowly than SIAs and remain isolated for a longer period of time. Nanoclusters associated with SIAs cause a pronounced overall expansion in a directions, as well as local strains. Under compressive strain in the c direction, vacancy diffusivity increases in the c direction. In contrast, the diffusivity of SIAs increases in the c direction under a tensile strain in the c direction. SIA nanoclusters are highly mobile within basal planes. Vacancy clusters grow by merging, leading to a contraction in the a direction, compensating for the expansion caused by SIA nanoclusters and possibly contributing to the plateau in growth after the initial rapid expansion. At the onset of breakaway growth, possibly due to stress buildup, the vacancy nanoclusters can condense into c loops, thereby diminishing the compensation effect. The alloying elements iron, nickel, chromium, and niobium liberated from secondary phase particles under irradiation or already in solution are attracted to vacancies and SIAs and are found inside vacancy and SIA loops. The interaction of alloying elements with defect clusters is discussed, with a particular focus on iron. Iron has been found to promote cluster formation in zirconium, and the structures of zirconium-iron clusters have been analyzed. Tin is repelled by SIA clusters and only weakly attracted by vacancies. Niobium impedes the diffusion of SIAs (and therefore may increase annihilation rates with nearby vacancies) and does not destabilize vacancy or SIA clusters. Ab initio calculations of the dimensional and elastic coefficients of the intermetallic phases occurring in secondary phase particles, such as Zr2Fe and Zr3Fe, are presented, allowing an assessment of local strains in a zirconium matrix. Thus, novel results from extended molecular dynamics simulations provide new insights and contribute to a deeper understanding of the complex mechanisms causing irradiation-induced dimensional changes and the breakaway growth of zirconium alloys.

Download or read book Scaling of Zirconium and Zirconium Alloys written by J. A. Burka and published by . This book was released on 1957 with total page 50 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Influence of Zirconium Alloy Chemical Composition on Microstructure Formation and Irradiation Induced Growth written by AV. Tselischev and published by . This book was released on 2002 with total page 22 pages. Available in PDF, EPUB and Kindle. Book excerpt: The studies of the dislocation structure, phase, and microchemical compositions of alloy Zr-1Nb-1.2Sn-0.35Fe (E635) and its modifications containing Fe from 0.15 to 0.65% were carried out before and after research reactor irradiation at ~350°C to maximal fluence of ~1027 m-2 (E > 0.1 MeV) and at ~60°C. The size and concentration of the a-type loops depend on the alloy composition and fluence and saturate even at low doses (1 dpa). The evolution of the c-component dislocation structure in recrystallized alloys of E365 type is determined by the chemical and phase compositions of alloys specifically, by the Fe/Nb ratio and the threshold dose, and is consistent with the irradiation growth strain acceleration. In E635 alloy containing 0.15%Fe the accelerated growth is observed after the dose of 15 dpa and is attended with the evolution of the c dislocation structure which is similar to Zr-1Nb (E110) alloy behavior. The irradiation induced growth of E635 type alloy containing 0.65% Fe is similar to that of E635 having the normal composition; no

Download or read book Nuclear Science Abstracts written by and published by . This book was released on 1967 with total page 978 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Peculiarities of Structural and Behavioral Changes of Some Zirconium Alloys at Damage Doses of Up to 50 Dpa written by VN. Shishov and published by . This book was released on 2004 with total page 14 pages. Available in PDF, EPUB and Kindle. Book excerpt: The irradiation-induced damage of zirconium alloys subjected to neutron irradiation up to dose levels of ~50 dpa was investigated. Specimens of unalloyed zirconium, Zr-1%Nb, Zr-2.5%Nb and Zr-1%Nb-1.3%Sn-0.4%Fe were irradiated in the BOR-60 reactor over the temperature range 320-420°C. The dose dependence of the irradiation growth strain increased sharply in zirconium and Zr-Nb irradiated at ~350°C at doses above ~10 dpa. In the case of Zr-1%Nb-1.3%Sn-0.4%Fe, it increased at doses of ~37 dpa. Upon increasing the irradiation temperature to 420°C, a sharp accelerated irradiation growth of the Zr-1%Nb alloy began shifting up to about 30 dpa. For the Zr- 1%Nb-1.3%Sn-0.4%Fe, no change of the irradiation growth rate was observed up to a dose of 55 dpa. The onset of increased irradiation growth in alloys correlates with the occurrence of c-component dislocation loops which coincides with a loss of coherence of finely-dispersed precipitates. Post-irradiation annealing experiments demonstrated that a delay in loop formation leads to displacement of the "break-away" beginning in the dose dependence of the irradiation growth in the direction of high doses. The a+c-type dislocation loops were also formed in Zr-1%Nb alloy at high doses, but their influence on the change of macroscopic properties was not observed.

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 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.