Stress corrosion cracking of candidate waste container materials; Final report Metadata

Title

• Main Title Stress corrosion cracking of candidate waste container materials; Final report

Creator

• Author: Park, J. Y.
  Creator Type: Personal
• Author: Maiya, P. S.
  Creator Type: Personal
• Author: Soppet, W. K.
  Creator Type: Personal
• Author: Diercks, D. R.
  Creator Type: Personal
• Author: Shack, W. J.
  Creator Type: Personal
• Author: Kassner, T. F.
  Creator Type: Personal
  Creator Info: Argonne National Lab., IL (United States)

Contributor

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

Publisher

• Name: Argonne National Laboratory
  Place of Publication: Illinois
• Name: Lawrence Livermore National Laboratory
  Place of Publication: California
  Additional Info: Lawrence Livermore National Lab., CA (United States)

Date

• Creation: 1992-06-01

Language

• English

Description

• Content Description: Six alloys have been selected as candidate container materials for the storage of high-level nuclear waste at the proposed Yucca mountain site in Nevada. These materials are Type 304L stainless steel (SS). Type 316L SS, Incoloy 825, phosphorus-deoxidized Cu, Cu-30%Ni, and Cu-7%Al. The present program has been initiated to determine whether any of these materials can survive for 300 years in the site environment without developing through-wall stress corrosion cracks. and to assess the relative resistance of these materials to stress corrosion cracking (SCC)- A series of slow-strain-rate tests (SSRTs) and fracture-mechanics crack-growth-rate (CGR) tests was performed at 93{degree}C and 1 atm of pressure in simulated J-13 well water. This water is representative, prior to the widespread availability of unsaturated-zone water, of the groundwater present at the Yucca Mountain site. Slow-strain-rate tests were conducted on 6.35-mm-diameter cylindrical specimens at strain rates of 10-{sup {minus}7} and 10{sup {minus}8} s{sup {minus}1} under crevice and noncrevice conditions. All tests were interrupted after nominal elongation strain of 1--4%. Scanning electron microscopy revealed some crack initiation in virtually all the materials, as well as weldments made from these materials. A stress- or strain-ratio cracking index ranks these materials, in order of increasing resistance to SCC, as follows: Type 304 SS < Type 316L SS < Incoloy 825 < Cu-30%Ni < Cu and Cu-7%Al. Fracture-mechanics CGR tests were conducted on 25.4-mm-thick compact tension specimens of Types 304L and 316L stainless steel (SS) and Incoloy 825. Crack-growth rates were measured under various load conditions: load ratios M of 0.5--1.0, frequencies of 10{sup {minus}3}-1 Hz, rise nines of 1--1000s, and peak stress intensities of 25--40 MPa{center_dot}m {sup l/2}.
• Physical Description: 76 p.

Subject

• Keyword: Tensile Properties
• Keyword: Radioactive Waste Disposal
• Keyword: Cracks
• Keyword: Crack Propagation
• Keyword: Fracture Mechanics
• STI Subject Categories: 36 Materials Science
• Keyword: Strain Rate
• STI Subject Categories: 05 Nuclear Fuels
• Keyword: Stainless Steel-316L
• Keyword: Progress Report
• Keyword: Stainless Steel-304L
• Keyword: High-Level Radioactive Wastes
• Keyword: Containers
• Keyword: Incoloy 825
• Keyword: Yucca Mountain Project
• Keyword: Stress Corrosion

Source

• Other Information: PBD: Jun 1992

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

• Other: DE92019239
• Report No.: ANL--92/28
• Grant Number: W-31109-ENG-38
• DOI: 10.2172/140794
• Office of Scientific & Technical Information Report Number: 140794
• Archival Resource Key: ark:/67531/metadc625520

Note

• Display Note: INIS; OSTI as DE92019239