Section
Resolution Product Category
Facilities Type
Related Technical Area
Image
NUREG 0933
Displaying 1 - 25 of 71
The objective of this task was to improve the reliability and capability of nuclear power plant containment structures to reduce the radiological consequences and risk to the public from design basis events and degraded-core and core-melt accidents. ITEM …
1
DESCRIPTION This item was originally identified in NUREG-0371 [1] and was later declared a USI in NUREG-0510. [2] (See Item A-3 for further details.) CONCLUSION This item was RESOLVED and requirements were established. (See Item A-3 for further details.) …
1
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 with a postulated LOCA were identified which had not been …
1
DESCRIPTION During testing for an advanced BWR containment system design (MARK III), suppression pool hydrodynamic loads were identified which had not been considered in the original design of the MARK I containment system. To address this issue, a MARK I …
1
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 had not been explicitly included in the original design 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 Since the issuance of Appendix J to 10 CFR Part 50 in February 1973, certain requirements of the appendix have been found to be conflicting, impractical for implementation, or subject to a variety of interpretations by …
1
DESCRIPTION Historical Background The safety concern of this NUREG-0371 [1] item deals with the consideration of alternatives to the basic design of nuclear power plants with the emphasis primarily on reduction of the vulnerability of reactors to …
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 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 Following a LOCA in an LWR, combustible gases, principally hydrogen, may accumulate inside the primary reactor containment as a result of: (1) metal-water reaction involving the fuel element cladding; (2) the radiolytic decomposition of the …
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 Historical Background The description of this item given in NUREG-0471 [1] is as follows: "This is an ACRS generic concern. Evaluation and approval is required of various aspects of the MARK III containment design which differs from the …
1
DESCRIPTION The calculations of differential pressure that occur in containment subcompartments from a loss-of-coolant event require a complex fluid dynamic analysis to assure that the subcompartment design pressures are not exceeded. To check the various …
1
DESCRIPTION The rationale for normal and postaccident containment cooling will be reviewed to determine the adequacy of the design requirements imposed on the containment ventilation systems. By reviewing typical designs, the staff will develop a basic …
1
DESCRIPTION Test data from the Marviken containment tests were obtained for the purpose of validating containment pressure codes used for performing independent calculations related to licensing reviews. The Marviken data are containment pressure …
1
DESCRIPTION The CONTEMPT computer code is used by the NRC staff to perform independent containment analyses. This NUREG-0471 [1] task involves the maintenance and revision of the CONTEMPT code to accommodate new containment designs or new problem areas as …
1
DESCRIPTION Historical Background SRP [1] Section 3.6, issued in 1975, addressed pipe breaks outside containment by combining limited design basis breaks for mechanistic protection with unlimited breaks for non-mechanistic protection. Prior to this, …
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.