Zircaloy cladding degradation under repository conditions

Conference · Sat Dec 01 00:00:00 EST 1990

OSTI ID:137958

Santanam, L; Raghavan, S; Chin, B A ^[1]; Shaw, H ^[2]

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1. Auburn Univ., AL (USA). Dept. of Materials Engineering
2. 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. (LLNL), Livermore, CA (United States)
• DOE Contract Number: W-7405-ENG-48
• OSTI ID: 137958
• Report Number(s): UCRL-100212; CONF-890820-6; ON: DE91006241; TRN: 91:002949
• Resource Relation: Conference: 4. international symposium on environmental degradation of materials in nuclear power systems: water reactors, Jekyll Island, GA (United States), 6-10 Aug 1989; Other Information: PBD: Dec 1990
• Country of Publication: United States
• Language: English

Related Subjects

05 NUCLEAR FUELS
36 MATERIALS SCIENCE
FUEL CANS
DEFORMATION
ZIRCALOY
CREEP
SPENT FUEL ELEMENTS
FRACTURES
STRESS CORROSION
RADIOACTIVE WASTE STORAGE
Yucca Mountain Project