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 Zirconium in the Nuclear Industry written by J. H. Schemel and published by ASTM International. This book was released on 1979 with total page 656 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book SiC CMC Zircaloy 4 Nuclear Fuel Cladding Performance During 4 Point Tubular Bend Testing written by and published by . This book was released on 2013 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The U.S. Department of Energy Office of Nuclear Energy (DOE NE) established the Light Water Reactor Sustainability (LWRS) program to develop technologies and other solutions to improve the reliability, sustain the safety, and extend the life of current reactors. The Advanced LWR Nuclear Fuel Development Pathway in the LWRS program encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. Recent investigations of potential options for "accident tolerant" nuclear fuel systems point to the potential benefits of silicon carbide (SiC) cladding. One of the proposed SiC-based fuel cladding designs being investigated incorporates a SiC ceramic matrix composite (CMC) as a structural material supplementing an internal Zircaloy-4 (Zr-4) liner tube, referred to as the hybrid clad design. Characterization of the advanced cladding designs will include a number of out-of-pile (nonnuclear) tests, followed by in-pile irradiation testing of the most promising designs. One of the out-of-pile characterization tests provides measurement of the mechanical properties of the cladding tube using four point bend testing. Although the material properties of the different subsystems (materials) will be determined separately, in this paper we present results of 4-point bending tests performed on fully assembled hybrid cladding tube mock-ups, an assembled Zr-4 cladding tube mock-up as a standard and initial testing results on bare SiC-CMC sleeves to assist in defining design parameters. The hybrid mock-up samples incorporated SiC-CMC sleeves fabricated with 7 polymer impregnation and pyrolysis (PIP) cycles. To provide comparative information; both 1- and 2-ply braided SiC-CMC sleeves were used in this development study. Preliminary stress simulations were performed using the BISON nuclear fuel performance code to show the stress distribution differences for varying lengths between loading points and clad configurations. The 2-ply sleeve samples show a higher bend momentum compared to those of the 1-ply sleeve samples. This is applicable to both the hybrid mock-up and bare SiC-CMC sleeve samples. Comparatively both the 1- and 2-ply hybrid mock-up samples showed a higher bend stiffness and strength compared with the standard Zr-4 mock-up sample. The characterization of the hybrid mock-up samples showed signs of distress and preliminary signs of fraying at the protective Zr-4 sleeve areas for the 1-ply SiC-CMC sleeve. In addition, the microstructure of the SiC matrix near the cracks at the region of highest compressive bending strain shows significant cracking and flaking. The 2-ply SiC-CMC sleeve samples showed a more bonded, cohesive SiC matrix structure. This cracking and fraying causes concern for increased fretting during the actual use of the design. Tomography was proven as a successful tool to identify open porosity during pre-test characterization. Although there is currently insufficient data to make conclusive statements regarding the overall merit of the hybrid cladding design, preliminary characterization of this novel design has been demonstrated.
Download or read book Zirconium in the Nuclear Industry written by ASTM Committee B-10 on Reactive and Refractory Metals and Alloys and published by ASTM International. This book was released on 1977 with total page 694 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book The Linear Reactivity Model for Nuclear Fuel Management written by M. J. Driscoll and published by . This book was released on 1990 with total page 264 pages. Available in PDF, EPUB and Kindle. Book excerpt: The Linear Reactivity Model (LRM) is a simple nuclear fuel management model that comes with a diskette containing three programs. Consisting of a collection of algorithms and methods, the LRM describes complex core behavior, but it is simpler than the complex programs developed for design calculations. This makes the LRM particularly useful as a teaching tool to explain the basic principles of nuclear fuel management. The LRM mainly focuses on the pressurized water reactor, but it is also directly applicable to the boiling water reactor. Application of the LRM to the CANDU reactor is also covered.
Download or read book Rupture Characteristics of Zircaloy 4 Cladding with Internal and External Simulation of Reactor Heating written by AR. Barber and published by . This book was released on 1977 with total page 14 pages. Available in PDF, EPUB and Kindle. Book excerpt: Rupture characteristics of internally heated Zircaloy-4 nuclear fuel cladding were investigated to develop rupture data that would provide a basis to bench-mark mathematical modeling of clad rupture. Single- and five-tube configurations of pressurized water reactor cladding were ruptured at constant pressures and heating rates with simulated reactor boundary conditions. The results showed that the ? + ? material property transformation region had a strong effect on the material ductility, and the clad rupture temperature varied linearly with hoop stress in the ? + ? region. Axial distribution of multitube ruptures was not coplanar. The rupture positions on all specimens were distributed normally about the point of maximum temperature. This paper emphasizes test techniques as well as the rupture characteristics of Zircaloy-4 clad.
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 Oxidation of Zirconium and Zirconium Alloys written by and published by . This book was released on 1959 with total page 48 pages. Available in PDF, EPUB and Kindle. Book excerpt: The oxidation rate was found to be relatively insensitive to various types of surface preparations in the temperature range 400 to 700 deg C. No dependence of reaction rate on oxygen pressure was observed. The cubic rate law also was obeyed by foil specimens at 700 deg C; however, the rate constants were slightly larger than values obtained from parallelepiped samples.
Download or read book Energy Research Abstracts written by and published by . This book was released on 1995 with total page 782 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Creep of Zirconium Alloys in Nuclear Reactors written by D. G. Franklin and published by ASTM International. This book was released on 1983 with total page 322 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Scientific and Technical Aerospace Reports written by and published by . This book was released on 1969 with total page 1508 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Nuclear Safety written by and published by . This book was released on 1969 with total page 1204 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 1976 with total page 612 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Review of Zircaloy 2 and Zircaloy 4 Properties Relevant to N S Savannah Reactor Design written by C. L. Whitmarsh and published by . This book was released on 1962 with total page 74 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Corrosion of Research Reactor Aluminium Clad Spent Fuel in Water written by International Atomic Energy Agency and published by . This book was released on 2003 with total page 236 pages. Available in PDF, EPUB and Kindle. Book excerpt: This report describes research performed in ten laboratories within the framework of the IAEA Co-ordinated Research Project on Corrosion of Research Reactor Aluminium Clad Spent Fuel in Water. The project consisted of exposure of standard racks of corrosion coupons in the spent fuel pools of the participating research reactor laboratories and evaluation of the coupons after predetermined exposure times, along with periodic monitoring of the storage water. A group of experts in the field contributed a state of the art review and provided technical supervision of the project. Localized corrosion mechanisms are notoriously difficult to understand, and it was clear from the outset that obtaining consistency in the results and their interpretation from laboratory to laboratory would depend on the development of an excellent set of experimental protocols. These experimental protocols are described in the report, together with guidelines for the maintenance of optimum water chemistry to minimize the corrosion of aluminium clad research reactor fuel in wet storage.
Download or read book ERDA Energy Research Abstracts written by United States. Energy Research and Development Administration and published by . This book was released on with total page 600 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book ERDA Energy Research Abstracts written by United States. Energy Research and Development Administration. Technical Information Center and published by . This book was released on 1976 with total page 588 pages. Available in PDF, EPUB and Kindle. Book excerpt:
