Download or read book In PWR Irradiation Performance of Dilute Tin Zirconium Advanced Alloys written by GP. Smith and published by . This book was released on 2002 with total page 17 pages. Available in PDF, EPUB and Kindle. Book excerpt: Zirconium alloys containing about 0.5% tin, which are classified as dilute tin alloys, possess excellent uniform waterside corrosion resistance necessary for the PWR fuel applications. Mechanical and irradiation growth properties of the dilute alloys can be adjusted for specific component application by controlling the additions of other alloying elements such as iron, chromium, niobium, and oxygen. Cladding alloys with such additions have been successfully irradiated to burnups up to 69 GWd/MTU, showing significant improvements in corrosion resistance and irradiation growth characteristics compared to low-tin Zircaloy-4, one of the current standard materials. The in-PWR creep resistance of such dilute alloys is comparable to that of low-tin Zircaloy-4. Another dilute alloy with predominantly iron-containing second-phase particles that are unstable under neutron irradiation (in a cold-worked microstructure, cold work introduced prior to irradiation) appears to be most suitable for the grid strip application. Cold-worked I-spring of this alloy in a transverse stamped grid provides excellent fuel rod support by inward motion of the spring within the grid cell due to irradiation growth. The hydrogen pickup fraction of several zirconium alloys, including Zircaloy-4 and dilute alloys, exhibits a well-behaved correlation with oxide thickness under non-heat flux conditions. A similar correlation is expected under heat flux conditions. Under heat flux conditions, the hydrogen pickup fraction for Zircaloy-4 approaches a constant value of about 15% for oxide thicknesses greater than 50 ?m. For the non heat-flux conditions, the pickup fraction is less than 5% for oxide thickness values greater than 50 ?m. Possible reasons for influence of oxide thickness and heat flux on the hydrogen pickup fraction are the porosity traps in thick oxide layers and atomic vibrations of oxide lattice under heat flux conditions. The in-PWR performance characteristics of the dilute alloys such as corrosion resistance, ductility, and dimensional stability can be controlled by optimization of the composition and fabrication process. These parameters influence the composition of the second-phase particles (SPP) in the alloy microstructure, which determines the radiation stability of the SPP. Irradiation stabilityof SPP has strong impact on the in-PWR performance characteristics of zirconium alloys.

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 Advanced Zirconium Alloy for PWR Application written by A. M. Garde and published by . This book was released on 2010 with total page 17 pages. Available in PDF, EPUB and Kindle. Book excerpt: Westinghouse is evaluating several advanced zirconium-based alloys, designated collectively as AXIOMTM, to achieve improved performance for more demanding fuel management schemes. There are five candidate AXIOM alloys currently being evaluated by Westinghouse. The in-pressurized water reactor (PWR) performance of one candidate alloy, X5A, is reviewed in this paper. The irradiation performance of X5A (previously identified as Alloy A) cladding that was fabricated using high-temperature processing (HTP) was published in 2002. Since then, the fabrication process of X5A was optimized by the use of a low-temperature process (LTP). Cladding tubes with the improved processing have been irradiated in two commercial PWRs (PWR A and PWR B) and in two test reactors (Test Reactor C and Test Reactor D). The irradiation performance of both versions (HTP X5A and LTP X5A) is reviewed in this paper with a primary emphasis on the current LTP cladding. After achieving an intermediate burnup in the range of 48-54 GWd/Metric Ton of Uranium (MTU) in PWR A, the average maximum oxide layer thickness for LTP X5A was about 23 ?m or about 27 % lower than the oxide thickness on ZIRLO® clad fuel rods. In addition, the fuel rod axial growth strain for LTP X5A was about 50 % of ZIRLO rod growth. Lead fuel rod irradiation of LTP X5A in PWR B with burnup in the range of 47-53 GWd/MTU showed about 30 % lower corrosion relative to ZIRLO rods and a 7 % lower axial rod growth strain than ZIRLO. An irradiation experiment in Test Reactor C was designed to study breakaway irradiation growth (in the absence of waterside oxidation) of several alloys. LTP X5A cladding showed a growth strain of about 20 % that of ZIRLO cladding at a fluence of 16 x 1025n/m2. In Test Reactor D, at a burnup of about 44 GWd/MTU, HTP X5A had the same oxide thickness as ZIRLO. However, the post-irradiation hydrogen pick-up was 35 % lower for HTP X5A compared to ZIRLO. In addition to the irradiation experience, the supplemental out-reactor autoclave evaluation of X5A welds indicates adequate weld corrosion resistance. While additional in-PWR exposures are required along with post-irradiation examination, the results to date demonstrate that X5A is a promising alloy for future PWR application.

Download or read book Zirconium in the Nuclear Industry written by George P. Sabol and published by ASTM International. This book was released on 2000 with total page 953 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Zirconium in the Nuclear Industry Tenth International Symposium written by A. M. Garde and published by ASTM International. This book was released on 1994 with total page 805 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Zirconium in the Nuclear Industry written by D. G. Franklin and published by ASTM International. This book was released on 1984 with total page 866 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Zirconium in the Nuclear Industry written by D. Franklin and published by ASTM International. This book was released on 1982 with total page 516 pages. Available in PDF, EPUB and Kindle. Book excerpt:

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 Zirconium in the Nuclear Industry written by and published by ASTM International. This book was released on with total page 627 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Zirconium in the Nuclear Industry written by Leo F. P. Van Swam and published by ASTM International. This book was released on 1989 with total page 412 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Zirconium in the Nuclear Industry written by Leo F. P. Van Swam and published by ASTM International. This book was released on 1989 with total page 781 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Microstructure Evolution in Zr Alloys During Irradiation written by M. Griffiths and published by . This book was released on 2010 with total page 10 pages. Available in PDF, EPUB and Kindle. Book excerpt: The performance of zirconium alloys in BWR, PWR, and PHWR nuclear reactors is dependent on the microstructure. Accordingly, the characterization of the microstructure is an integral part of any study conducted to develop models for in-reactor performance. Although the as-fabricated microstructure (texture, grain size, dislocation density, and phase or precipitate distribution) determines the basic physical properties of a given component, there are changes that occur during irradiation that can have a significant effect on these properties. Microstructures that illustrate specific features of the radiation damage that forms in Zr alloys will be illustrated and discussed in terms of the dose, dose rate, and impurity factors that are applicable. The original paper was published by ASTM International in the Journal of ASTM International, December 2007.

Download or read book Effect of In PWR Irradiation on Size Structure and Composition of Intermetallic Precipitates of Zr Alloys written by F. Garzarolli and published by . This book was released on 1996 with total page 16 pages. Available in PDF, EPUB and Kindle. Book excerpt: The corrosion behavior of Zr alloys depends on the kind, size, and distribution of the intermetallic second-phase particles. TEM examinations of Zr-Sn-Fe-Cr alloys irradiated in PWRs at temperatures between 300 and 370°C and fast fluences in the range of 5E21 to 1.3E22 cm-2 have been performed to study the irradiation-induced effects on the precipitates. The alloys contained different types of second-phase particles such as Zr(Fe,Cr)2, Zr2(Fe,Si), and Zr3Fe before irradiation. The influence of irradiation was found to depend on temperature and type of second-phase particles.

Download or read book Irradiation Creep and Growth Behavior and Microstructural Evolution of Advanced Zr Base Alloys written by A. Soniak and published by . This book was released on 2000 with total page 23 pages. Available in PDF, EPUB and Kindle. Book excerpt: This paper deals with the irradiation-induced changes in the microstructure of FRAMATOME advanced Zr-base alloys and the correlations with their irradiation creep and growth behavior. The first part is dedicated to experimental irradiations performed at 280 and 350°C in a CEA metallurgical test reactor (Siloé, 1 to 1018 n/m2 s, E > 1 MeV) on stress-relieved (SRA) and recrystallized (RXA) low-tin Zircaloy-4 and two advanced RXA materials (Alloy 4 (M4): Zr-SnFeV, and Alloy 5 (M5): Zr-NbO), which are proposed for fuel rod cladding applications in PWRs. The irradiation creep results confirm the improved behavior of RXA Zy-4, M4, and M5 in comparison to that of SRA Zy-4. Similar trends are observed for irradiation growth, the SRA Zy-4 exhibiting a quasi-linear behavior with increasing fluence while RXA alloys undergo an early saturation phenomenon. Among RXA materials, M5 has the higher irradiation growth resistance. These creep and growth results at moderate neutron fluences (

Download or read book Irradiation Effects on the Mechanical Properties of Zirconium and Dilute Zirconium Alloys written by and published by . This book was released on 1976 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The effects of fast flux (E greater than or equal to 1 MeV) neutrons on zirconium and dilute zirconium alloys are discussed. The effects on the elastic constants, strength and ductility, creep, fatigue and fracture, and irradiation growth are reviewed. (FS).

Download or read book Preliminary Irradiation Effect on Corrosion Resistance of Zirconium Alloys written by Vladimir Markelov and published by . This book was released on 2018 with total page 24 pages. Available in PDF, EPUB and Kindle. Book excerpt: In order to improve the composition and structure of zirconium alloys, it is important to know how changes in their structure-phase state under irradiation affect various properties, including corrosion. In this work, autoclave corrosion tests were carried out on fuel cladding samples made of zirconium-niobium and zirconium-niobium-tin-iron alloy systems. All samples were preliminarily irradiated in a BOR-60 nuclear reactor at a fluence of 2 x 1026 m-2 (E > 0.1 MeV). The autoclave tests of irradiated and unirradiated samples were performed at 350°C for 240 days in distilled water containing 10 ppm lithium and 1,600 ppm boron.

Download or read book Comprehensive Nuclear Materials written by Todd R Allen and published by Elsevier. This book was released on 2011-05-12 with total page 3552 pages. Available in PDF, EPUB and Kindle. Book excerpt: Comprehensive Nuclear Materials, Five Volume Set discusses the major classes of materials suitable for usage in nuclear fission, fusion reactors and high power accelerators, and for diverse functions in fuels, cladding, moderator and control materials, structural, functional, and waste materials. The work addresses the full panorama of contemporary international research in nuclear materials, from Actinides to Zirconium alloys, from the worlds' leading scientists and engineers. Critically reviews the major classes and functions of materials, supporting the selection, assessment, validation and engineering of materials in extreme nuclear environment Fully integrated with F-elements.net, a proprietary database containing useful cross-referenced property data on the lanthanides and actinides Details contemporary developments in numerical simulation, modelling, experimentation, and computational analysis, for effective implementation in labs and plants