Notch Ductility Properties of SM-1A Reactor Pressure Vessel Following the In-Place Annealing Operation

Publisher Website
Google Scholar
Abstract
Embrittlement of the Army SM-1A reactor pressure vessel, as modified by the recently completed in-place anneal, was assessed, and an analysis made of its reembrittlement behavior with subsequent radiation service. Experimental results from a surveillance program covering 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, Modified) from accelerated irradiation programs. These data indicate a 0°F maximum pressure vessel wall Charpy- V 30-ft-lb transition temperature after the in-place anneal vs a −80°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° 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 5.5 × 1019 n/cm2 (>0.5 MeV). This is comparable to-124.7 MW-y of reactor operation.

This publication has 3 references indexed in Scilit: