A Study of the Embrittlement of Reactor Vessel Steel Supports

Abstract

Neutron transport analyses have been done on three standard USA design pressurized water reactors to determine the fluence levels at the vessel supports. The shift in the nil-ductility transition temperature (ΔNDTT) at the supports has been estimated directly with the calculated fluence levels in accordance with ASTM Standard E706. Values of the displacement per atom unit (dpa) are also provided as an alternate estimate of the ΔNDTT.

The results indicate that supports with relatively high copper contents experience a large transition temperature shift. Further, the use of dpa values with E>1.0 MeV predict a higher transition temperature shift than using fluence with E>1.0 Mev.