The Effect Of Neutron Irradiation On The Mechanical Properties Of Welded Zircaloy-2

Technical Report · Sun Jul 15 00:00:00 EDT 1962

OSTI ID:4782148

Evans, D G

Zircaloy-2 tensile specimens, subsize impact bars and representative spigot welds were subjected to three NRX cycles in the X-5 loop. Average loop temperature was 260 deg C over the three cycles. One group of tensile specimens was heat-treated in vacuum at 900 deg C for 40 minutes, another group contained welded areas in the center of the gauge length and a third group was hydrided after welding. Notches of the impact specimens were located in the fusion zone of the weld. Spigot welds were made on autoclaved and unautoclaved simulated production assemblies. Neutron irradiation had no effect on the impact properties of welded Zircaloy-2. Welding decreased the uniform and total elongation at room temperature and at 260 deg C, and increased the 260 deg C PL, YS, and UTS. Hydriding to a nominal 100 ppm hydrogen had no effect on the unirradiated tensile properties at either test temperature. The heat treatment decreased the strength properties but did not affect the ductility. Neutron irradiation increased the YS of the welded and hydrided material by 20% and the heat treated YS by 40%. Irradiation also increased the 260 deg C strength properties of the as-welded material. The unautoclaved spigot welds had a generally higher tensile strength than the autoclaved and welded specimens. For specimens welded in either condition, the outer welds of the 19-element bundle had a lower average breaking load than the inner welds. Neutron irradiation had no effect on the tensile strength of these welds. It was also demonstrated that a cup-and-cone type of fracture could be produced in a bend test. The fractures were similar to those observed in irradiated fuel bundles which was damaged during transfer operations. (auth)

• Research Organization: Atomic Energy of Canada Ltd., Chalk River, Ontario (Canada)
• NSA Number: NSA-16-030898
• OSTI ID: 4782148
• Report Number(s): CRGM-1093; AECL-1566
• Country of Publication: Canada
• Language: English

Related Subjects

METALS, CERAMICS, AND OTHER MATERIALS
CHEMICAL REACTIONS
DEFORMATION
DUCTILITY
ELASTICITY
EXPANSION
FAILURES
HEAT TREATMENTS
HYDROGEN
IMPACT SHOCK
IN PILE LOOPS
IRRADIATION
MECHANICAL PROPERTIES
NEUTRON BEAMS
NRX
RADIATION EFFECTS
TEMPERATURE
TENSILE PROPERTIES
TESTING
VACUUM
WELDS
ZIRCALOY
NESDPS Office of Nuclear Energy Space and Defense Power Systems