Item A-20: Impacts of the Coal Fuel Cycle
DESCRIPTION
At the time this issue was identified in NUREG-0371,[1]compliance with NEPA required that alternatives to a proposed Feder
Item A-19: Digital Computer Protection System
DESCRIPTION
At the time this issue was identified in NUREG-0371,[1]trends in the design of nuclear power plants showed an increase in
Item A-18: Pipe Rupture Design Criteria
DESCRIPTION
Historical Background
A major objective of this NUREG-0371[1]item was the development of consistent criteria for a
Item A-17: Systems Interactions in Nuclear Power Plants (former USI)
DESCRIPTION
Nuclear power plants contain many structures, systems, and components (SSCs), some of which are safety- related. Certain SSCs are designed to interact to perform their intended functions.
Item A-16: Steam Effects on BWR Core Spray Distribution
DESCRIPTION
Historical Background
Prior to May 1978, tests conducted by GE showed that the presence of steam and/or increased pressure in and above the upper core region of BWRs could adversely affect the distribution
Item A-15: Primary Coolant System Decontamination and Steam Generator Chemical Cleaning
DESCRIPTION
Historical Background
Operation of a LWR results in slow corrosion of the interior metal surfaces of the primary coolant system. The resulting corrosion products circulate through the reactor core and are
Item A-14: Flaw Detection
DESCRIPTION
Historical Background
After the 1970 issuance of inspection requirements in Section XI of the ASME Boiler and Pressure Vessel
Item A-13: Snubber Operability Assurance
DESCRIPTION
Historical Background
In May 1978, the ACRS and the staff expressed concern over the substantial number of LERs related to the malfunction of snubbers, the most frequent of w
Item A-12: Fracture Toughness of Steam Generator and Reactor Coolant Pump Supports (former USI)
DESCRIPTION
During the course of the licensing action for North Anna Units 1 and 2, a number of questions were raised as to the potential for lamellar tearing and low fracture toughness of the s
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 nuclear facilities does not provi