Accession Number:

AD0671807

Title:

NOTCH DUCTILITY PROPERTIES OF SM-1A REACTOR PRESSURE VESSEL FOLLOWING THE IN-PLACE ANNEALING OPERATION.

Descriptive Note:

Final rept.,

Corporate Author:

NAVAL RESEARCH LAB WASHINGTON D C

Report Date:

1968-05-21

Pagination or Media Count:

31.0

Abstract:

The embrittlement condition of the Army SM-1A reactor pressure vessel, as modified by the recently completed in-place anneal, was assessed and an analysis was made of the reembrittlement behavior of the vessel steel with subsequent radiation service. Experimental results from the reactor surveillance program developed through one complete irradiation and annealing cycle are presented, together with a summary of experimental information on the annealing response of the vessel steel A350-LF1, Mod. from accelerated irradiation programs. These data indicate a 0 deg F maximum pressure vessel wall Charpy-V 30 ft-lb transition temperature after the in-place anneal versus a -80 deg F preservice transition temperature based on the notch-ductility properties of a duplicate ring forging. The maximum Charpy-V 30 ft-lb transition temperature of the pressure vessel before the annealing operation was estimated at 190 deg F. A projection of postanneal pressure vessel lifetime in terms of neutron fluence 0.5 Mev was derived from spectra calculations and the experimentally predicted reirradiation response of the pressure vessel steel. The maximum permissible vessel wall fluence is estimated at 5.5x10 to the 19th power nsq cm 0.5 Mev. This is comparable to 124.7 Megawatt years of reactor operation. Author

Subject Categories:

  • Containers and Packaging
  • Nuclear Power Plants and Fission Reactor Engineering
  • Radioactivity, Radioactive Wastes and Fission Products

Distribution Statement:

APPROVED FOR PUBLIC RELEASE