Large-scale thermal-shock experiments with clad and unclad steel cylinders

Technical Report · Thu Dec 31 04:00:00 EST 1992

OSTI ID:10163357

Cheverton, R D

Flaw behavior trends associated with pressurized-thermal-shock (PTS) loading of pressurized-water-reactor pressure vessels have been under investigation at the Oak Ridge National Laboratory for nearly 20 years. During that time, twelve thermal-shock experiments with thick-walled (152 mm) steel cylinders were conducted as a part of the investigations. The first eight experiments were conducted with unclad cylinders initially containing shallow (8--19 mm) two-dimensional and semicircular inner-surface flaws. These experiments demonstrated, in good agreement with linear elastic fracture mechanics, crack initiation and arrest, a series of initiation/arrest events with deep penetration of the wall, long crack jumps, arrest with the stress intensity factor (K{sub I}) increasing with crack depth, extensive surface extension of an initially short and shallow (semicircular) flaw, and warm prestressing with K{sub I} {le} 0. The remaining four experiments were conducted with clad cylinders containing initially shallow (19--24 mm) semielliptical subclad and surface flaws at the inner surface. In the first of these experiments one of six equally spaced (60{degrees}) {open_quotes}identical{close_quotes} subclad flaws extended nearly the length of the cylinder (1,220 mm) beneath the cladding (no crack extension into the cladding) and nearly 50% of the wall, radially. For the final experiment, four of the semielliptical subclad flaws that had not propagated previously were converted to surface flaws, and they experienced extensive extension beneath the cladding with no cracking of the cladding. Information from this series of thermal-shock experiments is being used in the evaluation of the PTS issue.

• Research Organization: Oak Ridge National Lab., TN (United States)
• Sponsoring Organization: Nuclear Regulatory Commission, Washington, DC (United States)
• DOE Contract Number: AC05-84OR21400
• OSTI ID: 10163357
• Report Number(s): CONF-9210209--3; ON: DE93015447
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210200
CRACK PROPAGATION
CYLINDERS
DEFECTS
EXPERIMENTAL DATA
FRACTURE MECHANICS
FUEL CANS
ORNL
POWER REACTORS
NONBREEDING
LIGHT-WATER MODERATED
NONBOILING WATER COOLED
PRESSURE VESSELS
PWR TYPE REACTORS
REGULATIONS
STEELS
THERMAL SHOCK