Section
Resolution Product Category
Facilities Type
Related Technical Area
Image
NUREG 0933
Applied Filters
Displaying 1 - 25 of 75
The objective of this task was to improve the reliability and capability of nuclear power plant systems for removing decay heat and achieving safe shutdown conditions following transients and under post-accident conditions. ITEM II.E.3.1: RELIABILITY OF …
1
DESCRIPTION Inspections of operating BWRs conducted up to April 1978 revealed cracks in the feedwater nozzles of 20 reactor vessels. Most of these BWRs contained 4 nozzles with diameters ranging from 10 in. to 12 in. Although most cracks ranged from 1/2 …
1
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 provide protection against reactor vessel failure. Prevention of …
1
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 which were: (1) seal leakage in hydraulic snubbers; and (2) high …
1
DESCRIPTION Historical Background The Class 1E power sources provide the electric power for the plant systems that are essential to reactor shutdown, containment isolation, reactor core cooling, containment heat removal or are otherwise essential in …
1
DESCRIPTION By letter dated June 18, 1975, licensees of operating reactor facilities were sent a preliminary copy of a staff paper, "Guidance for Proposed License Amendments Relating to Refueling," and "Refueling Information Request Form." The purpose was …
1
DESCRIPTION Historical Background This NUREG-0371 [1] item involves the development of consistent and formalized acceptance criteria regarding the conversion to higher density storage racks (increased storage capacity) in existing spent fuel storage …
1
DESCRIPTION Historical Background This issue is documented in NUREG-0371 [1] and addresses the adequacy of safety-related DC power supplies which was questioned by a nuclear consultant in a letter to the ACRS in April 1977 (NUREG-0305, [2] Attachment B). …
1
DESCRIPTION This issue was raised in NUREG-0371 [1] following a series of events at operating plants that involved the offsite power systems. These events indicated that the reliability of the preferred source of emergency power might have been less than …
1
DESCRIPTION At all nuclear plants, overhead cranes are used to lift heavy objects in the vicinity of spent fuel. If a heavy object such as a spent fuel shipping cask or shielding block were to fall onto spent fuel in the storage pool or reactor core …
1
DESCRIPTION Historical Background The AEC first established missile-protection requirements in 1967. GDC-2 and GDC-4 of 10 CFR Part 50, Appendix A, require in part that structures, systems, and components important to safety be designed to be able to …
1
DESCRIPTION Structures, systems, and components important to the safety of nuclear power plants are required to withstand the effects of natural phenomena such as earthquakes. Broad requirements for earthquake resistance are specified in 10 CFR Parts 50 …
1
DESCRIPTION Historical Background In a memorandum [1] dated June 7, 1976, NRR recommended that a study be initiated on the quantification of inherent seismic safety margins in NRR's seismic design requirements. This memo suggested the initiation of a …
1
DESCRIPTION The complete loss of AC electrical power to the essential and nonessential switchgear buses in a nuclear power plant is referred to as a "Station Blackout." Because many safety systems required for reactor core decay heat removal are dependent …
1
DESCRIPTION The design criteria and methods for the seismic qualification of mechanical and electrical equipment in nuclear power plants underwent significant changes during the course of the licensing of commercial nuclear power plants. Consequently, it …
1
DESCRIPTION Neutron irradiation of reactor pressure vessel weld and plate materials decreases the fracture toughness of the materials. The fracture toughness sensitivity to radiation-induced change is increased by the presence of certain materials such as …
1
This item has been divided into two parts which have been evaluated separately. PART I - Ductility of Two-Way Slabs and Shells DESCRIPTION Historical Background This issue was identified in NUREG-0471 [1] and involved concern over the lack of information …
1
DESCRIPTION Historical Background This issue was identified as a generic problem in NUREG-0471 [1] and concerns the design of pressure vessels and piping systems components which must be designed to accommodate individual and combined loads due to normal …
1
DESCRIPTION Some prototype electrical penetration failures have occurred to date. In addition, failures of low voltage penetration modules have occurred at a licensed facility. It was originally postulated that failures of the low voltage penetration …
1
DESCRIPTION Historical Background Individual reactor fuel rods sometimes fail during normal operations and many fuel rods are expected to fail during severe accidents. To ensure that these fuel failures do not result in unacceptable releases to the …
1
DESCRIPTION This NUREG-0471 [1] task covers a spectrum of technical efforts related to the staff review of LMFBR fuel designs. The efforts include: (1) evaluating licensing requirements for the behavior of stainless steel cladding and hexcans in an LMFBR …
1
DESCRIPTION The seismic qualification of complex electrical and mechanical components has undergone rapid changes in the past several years. Revisions in industry standards (IEEE 344) have resulted in the need to audit the qualification programs of NSSS …
1
DESCRIPTION Applicants are required to provide confirmation of the adequacy of computer programs used in the structural analysis and design of piping systems and components. At the time this issue was identified, this consisted of applicants providing, …
1
DESCRIPTION Historical Background As described in NUREG-0471, [1] this issue involves staff evaluations to assess the adequacy of specific containment penetration designs from the point of view of structural integrity, ISI requirements, and new …
1
DESCRIPTION Since the adoption by the ASME Code, Section III, of Subsection NF on component supports, technical review has been limited to conformance of the information provided in the application and commitment by the applicants to component support …
1
Page Last Reviewed/Updated 3/1/2026
Disclaimer: Some of the formatting in NUREG-0933 may not be correct. We are currently working on fixing the formatting.