Item A-11: Reactor Vessel Materials Toughness (former USI)
DESCRIPTION
Because of the remote possibility of failure of nuclear reactor pressure vessels designed to the ASME Boiler and Pressure Vessel Code, the design of nucl
Item A-10: BWR Feedwater Nozzle Cracking (former USI)
DESCRIPTION
Inspections of operating BWRs conducted up to April 1978 revealed cracks in the feedwater nozzles of 20 reactor vessels.
Item A-9: ATWS (former USI)
DESCRIPTION
The technical report on ATWS for water-cooled reactors (WASH-1270)[1]discussed the probability of an ATWS event as well as a
Item A-8: Mark II Containment Pool Dynamic Loads Long-Term Program (former USI)
DESCRIPTION
As a result of the GE testing program for the MARK III pressure-suppression containment program, new containment loads associated with a postulated LOCA were identified in 1975 which h
Item A-7: Mark I Long-Term Program (former USI)
DESCRIPTION
During testing for an advanced BWR containment system design (MARK III), suppression pool hydrodynamic loads were identified which had not been considere
Item A-6: Mark I Short-Term Program (former USI)
DESCRIPTION
During the conduct of a large scale testing program for an advanced design BWR pressure suppression containment system (MARK III), new suppression pool hydrodynamic loads associated wi
Item A-7: Mark I Long-Term Program (former USI)
DESCRIPTION
This item was originally identified in NUREG-0371[1]and was later declared a USI in NUREG-0510.
Item A-4: CE Steam Generator Tube Integrity (former USI)
DESCRIPTION
This item was originally identified in NUREG-0371[1]and was later declared a USI in NUREG-0510.
Item A-3: Westinghouse Steam Generator Tube Integrity (former USI)
DESCRIPTION
Prior to 1978, operating experience with PWR steam generators was characterized by extensive corrosion and mechanically-induced degradation of the steam
Item A-2: Asymmetric Blowdown Loads on Reactor Primary Coolant Systems (former USI)
DISCUSSION
On May 7, 1975, the NRC was informed by VEPCO that an asymmetric loading on the reactor vessel supports resulting from a postulated reactor coolant pipe r