Fabrication technological development of the oxide dispersion strengthened alloy MA957 for fast reactor applications Metadata

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Title

  • Main Title Fabrication technological development of the oxide dispersion strengthened alloy MA957 for fast reactor applications

Creator

  • Author: Hamilton, ML
    Creator Type: Personal
  • Author: Gelles, DS
    Creator Type: Personal
  • Author: Lobsinger, RJ
    Creator Type: Personal
  • Author: Johnson, GD
    Creator Type: Personal
  • Author: Brown, WF
    Creator Type: Personal
  • Author: Paxton, MM
    Creator Type: Personal
  • Author: Puigh, RJ
    Creator Type: Personal
  • Author: Eiholzer, CR
    Creator Type: Personal
  • Author: Martinez, C
    Creator Type: Personal
  • Author: Blotter, MA
    Creator Type: Personal

Contributor

  • Sponsor: United States. Department of Energy.
    Contributor Type: Organization
    Contributor Info: US Department of Energy (United States)

Publisher

  • Name: Pacific Northwest National Laboratory (U.S.)
    Place of Publication: Richland, Washington
    Additional Info: Pacific Northwest National Lab., Richland, WA (United States)

Date

  • Creation: 2000-03-27

Language

  • English

Description

  • Content Description: A significant amount of effort has been devoted to determining the properties and understanding the behavior of the alloy MA957 to define its potential usefulness as a cladding material, in the fast breeder reactor program. The numerous characterization and fabrication studies that were conducted are documented in this report. The alloy is a ferritic stainless steel developed by International Nickel Company specifically for structural reactor applications. It is strengthened by a very fine, uniformly distributed yttria dispersoid. Its fabrication involves a mechanical alloying process and subsequent extrusion, which ultimately results in a highly elongated grain structure. While the presence of the dispersoid produces a material with excellent strength, the body centered cubic structure inherent to the material coupled with the high aspect ratio that results from processing operations produces some difficulties with ductility. The alloy is very sensitive to variations in a number of processing parameters, and if the high strength is once lost during fabrication, it cannot be recovered. The microstructural evolution of the alloy under irradiation falls into two regimes. Below about 550 C, dislocation development, {alpha}{prime} precipitation and void evolution in the matrix are observed, while above about 550 C damage appears to be restricted to cavity formation within oxide particles. The thermal expansion of the alloy is very similar to that of HT9 up to the temperature where HT9 undergoes a phase transition to austenitic. Pulse magnetic welding of end caps onto MA957 tubing can be accomplished in a manner similar to that in which it is performed on HT9, although the welding parameters appear to be very sensitive to variations in the tubing that result from small changes in fabrication conditions. The tensile and stress rupture behavior of the alloy are acceptable in the unirradiated condition, being comparable to HT9 below about 700 C and exceeding those of HT9 at higher temperatures. Neither tensile nor rupture strength appear to be degraded by irradiation to fast fluencies on the order of 8 x 10{sup 22} n/cm{sup 2} in the range of 370--760 C, although some loss of ductility has been observed. The impact resistance of the alloy is very poor in the unirradiated condition, and is significantly degraded by irradiation.
  • Physical Description: Medium: P; Size: 110 pages

Subject

  • Keyword: Fbr Type Reactors
  • Keyword: Mechanical Properties
  • Keyword: Physical Radiation Effects
  • STI Subject Categories: 36 Materials Science
  • Keyword: Ferritic Steels
  • Keyword: Yttrium Oxides
  • Keyword: Reactor Materials
  • Keyword: Microstructure
  • Keyword: Fabrication
  • STI Subject Categories: 21 Specific Nuclear Reactors And Associated Plants

Source

  • Other Information: PBD: 27 Mar 2000

Collection

  • Name: Office of Scientific & Technical Information Technical Reports
    Code: OSTI

Institution

  • Name: UNT Libraries Government Documents Department
    Code: UNTGD

Resource Type

  • Report

Format

  • Text

Identifier

  • Report No.: PNNL-13168
  • Report No.: AT6020000
  • Grant Number: AC06-76RL01830
  • DOI: 10.2172/752621
  • Office of Scientific & Technical Information Report Number: 752621
  • Archival Resource Key: ark:/67531/metadc708437

Note

  • Display Note: INIS; OSTI as DE00752621