POSTIRRADIATION EXAMINATION OF 17-4 PH STAINLESS STEEL CONTROL ROD DRIVE RACK FROM SM-1 REACTOR

Technical Report · Fri Dec 01 00:00:00 EST 1961

DOI:https://doi.org/10.2172/4829940· OSTI ID:4829940

Klindt, K. K.; Richt, A. E.; Thurber, W. C.

A portion of the control rod drive system in the SM-1 Reactor, fabricated from l7-4 PH stainless steel, was examined at the Oak Ridge National Laboratory (ORNL) hot cells after successful operation in a pressurized-water environment for approximately three years. Examination included visual inspection, magnetic-particle tests, fluorescent penetrant tests, and metallography. No evidence of stress-corrosion cracking or other damage was noted. (auth)

• Research Organization: Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
• DOE Contract Number: W-7405-ENG-26
• NSA Number: NSA-16-003890
• OSTI ID: 4829940
• Report Number(s): ORNL-3218
• Resource Relation: Other Information: Orig. Receipt Date: 31-DEC-62
• Country of Publication: United States
• Language: English

Related Subjects

REACTOR TECHNOLOGY
CONFIGURATION
CONTROL ELEMENTS
CORROSION
CRACKS
FAILURES
FLUORESCENCE
IRRADIATION
MACHINE PARTS
MAGNETIC MATERIALS
MATERIALS TESTING
METALLOGRAPHY
MOTORS
POWER PLANTS
PRESSURE
RADIATION EFFECTS
RADIATION PROTECTION
REACTOR CORE
REMOTE HANDLING
RODS
SERVOMECHANISMS
SM-1
STAINLESS STEELS
STRESSES
WATER COOLANT