Use of miniature and standard specimens to evaluate effects of irradiation temperature on pressure vessel steels Metadata

Metadata describes a digital item, providing (if known) such information as creator, publisher, contents, size, relationship to other resources, and more. Metadata may also contain "preservation" components that help us to maintain the integrity of digital files over time.

Title

  • Main Title Use of miniature and standard specimens to evaluate effects of irradiation temperature on pressure vessel steels

Creator

  • Author: Haggag, F.M.
    Creator Type: Personal
  • Author: Nanstad, R.K. (Oak Ridge National Lab., TN (United States))
    Creator Type: Personal
  • Author: Byrne, S.T. (ABB/Combustion Engineering, Inc., Windsor, CT (United States))
    Creator Type: Personal

Contributor

  • Sponsor: United States. Department of Energy.
    Contributor Type: Organization
    Contributor Info: DOE; USDOE, Washington, DC (United States)

Publisher

  • Name: Oak Ridge National Laboratory
    Place of Publication: Tennessee

Date

  • Creation: 1991-01-01

Language

  • English

Description

  • Content Description: The effects of neutron irradiation on the steel reactor vessel for the modular high-temperature gas-cooled reactor (MHTGR) are being investigated, primarily because the operating temperatures are low (121 to 210{degrees}C (250--410{degrees}F)) compared to those for commercial light-water reactors (LWRs) ({approximately}288{degrees}C (550{degrees}F)). The need for design data on the reference temperature shift necessitated the irradiation at different temperatures of A 533 grade B class 1 plate. A 508 class 3 forging, and welds used for the vessel shell, vessel closure head, the vessel flange. This paper presents results from the first four irradiation capsules of this program. The four capsules were irradiated in the University of Buffalo Reactor to an effective fast fluence of 1 {times}10{sup 18} neutron/cm{sup 2} (0.68 {times} 10{sup 18} neutron/cm{sup 2} (>1 MeV)) at temperatures of 288, 204, 163, and 121{degrees}C (550, 400, 325, and 250{degrees}F), respectively. The yield and ultimate strengths of both steel plate materials of the MHTGR Program increased with decreasing irradiation temperature. Similarly, the 41-J Charpy V-notch (CVN) transition temperature shift increased with decreasing irradiation temperature (in agreement with the increase in yield strength). The miniature tensile and automated ball indentation (ABI) test results (yield strength and flow properties) were in good agreement with those from standard tensile specimens. The miniature tensile and ABI test results were also used in a model that utilizes the changes in yield strength to estimate the CVN ductile-to-brittle transition temperature shift due to irradiation. The model predictions were compared with CVN test results obtained here and in earlier work. 5 refs., 11 figs., 6 tabs.
  • Physical Description: Pages: (7 p)

Subject

  • Keyword: Reactor Vessels
  • STI Subject Categories: 36 Materials Science
  • Keyword: Charpy Test
  • Keyword: Alloys
  • Keyword: Materials Testing
  • Keyword: Tensile Properties
  • Keyword: Materials
  • Keyword: Steels
  • Keyword: Welded Joints
  • Keyword: Graphite Moderated Reactors
  • Keyword: Radiation Effects
  • Keyword: Steel-Astm-A508
  • Keyword: Impact Tests
  • STI Subject Categories: 360103 -- Metals & Alloys-- Mechanical Properties
  • Keyword: Reactor Materials
  • Keyword: Reactors
  • STI Subject Categories: 21 Specific Nuclear Reactors And Associated Plants
  • Keyword: Iron Alloys
  • Keyword: Mechanical Properties
  • Keyword: Joints
  • Keyword: Mechanical Tests
  • Keyword: Gas Cooled Reactors
  • Keyword: Neutron Fluence
  • Keyword: Htgr Type Reactors
  • Keyword: Carbon Steels
  • Keyword: Steel-Astm-A533-B
  • STI Subject Categories: 360106 -- Metals & Alloys-- Radiation Effects
  • Keyword: Testing 210300* -- Power Reactors, Nonbreeding, Graphite Moderated
  • Keyword: Destructive Testing
  • Keyword: Iron Base Alloys
  • Keyword: Low Alloy Steels
  • Keyword: Containers

Source

  • Conference: 5. international symposium on environmental degradation on materials in nuclear power systems - water reactors, Monterey, CA (United States), 25-29 Aug 1991

Collection

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

Institution

  • Name: UNT Libraries Government Documents Department
    Code: UNTGD

Resource Type

  • Article

Format

  • Text

Identifier

  • Other: DE91018849
  • Report No.: CONF-910808-5
  • Grant Number: AC05-84OR21400
  • Office of Scientific & Technical Information Report Number: 5177041
  • Archival Resource Key: ark:/67531/metadc1057430

Note

  • Display Note: OSTI; NTIS; INIS; GPO Dep.