Zircaloy cladding degradation under repository conditions
Conference
·
OSTI ID:137958
- Auburn Univ., AL (USA). Dept. of Materials Engineering
- Lawrence Livermore National Lab., CA (USA)
Creep, a potential degradation mechanism of Zircaloy cladding after repository disposal of spent nuclear fuel, has been investigated. The deformation and fracture map methodology has been used to predict maximum allowable initial storage temperatures to achieve a thousand year life without rupture as a function of spent-fuel history. Maximum allowable temperatures are 340{degree}C (613 K) for typically stressed rods (70--100 MPa) and 300{degree}C (573 K) for highly stressed rods (140--160 MPa). 10 refs., 2 figs.
- Research Organization:
- Lawrence Livermore National Lab., CA (United States)
- DOE Contract Number:
- W-7405-ENG-48
- OSTI ID:
- 137958
- Report Number(s):
- UCRL--100212; CONF-890820--6; ON: DE91006241
- Country of Publication:
- United States
- Language:
- English
Similar Records
Zircaloy cladding rupture during repository storage
Modeling of Zircaloy cladding degradation under repository conditions
Low-temperature rupture behavior of Zircaloy-clad pressurized water reactor spent fuel rods under dry storage conditions
Journal Article
·
Sat Feb 29 23:00:00 EST 1992
· Nuclear Technology; (United States)
·
OSTI ID:5433549
Modeling of Zircaloy cladding degradation under repository conditions
Conference
·
Sat Jul 01 00:00:00 EDT 1989
·
OSTI ID:137739
Low-temperature rupture behavior of Zircaloy-clad pressurized water reactor spent fuel rods under dry storage conditions
Journal Article
·
Mon Oct 01 00:00:00 EDT 1984
· Nucl. Technol.; (United States)
·
OSTI ID:5944989