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NUREG 0933
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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 …
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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 …
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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 …
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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 …
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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 …
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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 …
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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 …
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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 …
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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 …
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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 …
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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, …
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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 …
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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 …
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DESCRIPTION According to NUREG-0471, [1] the safety concern in this issue relates to results from inspections of various structural components in the torus support systems of operating BWRs. These inspections have revealed several inconsistencies between …
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DESCRIPTION Surface cracks have been discovered in control rod drive internal parts at some operating BWR plants. Although only observed to be localized in nature, this cracking, if propagated, could potentially affect the capability of the control rod …
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DESCRIPTION GDC-53, "Provisions for Containment Testing and Inspection," requires in part that the reactor containment be designed to permit: (1) periodic inspection of all important areas, and (2) an appropriate surveillance program. 10 CFR 50, Appendix …
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DESCRIPTION Historical Background This issue was identified at an NRC Operating Reactor Events meeting on January 7, 1982, [1] and addressed fire protection system (FPS) actuations that resulted in adverse interactions with safety-related equipment at …
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DESCRIPTION Historical Background This issue was initiated to address concerns raised by the Union of Concerned Scientists. (References [1] , [2] , and [3] .) The purposes for including this issue as a generic issue are to: (1) provide brief background …
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DESCRIPTION Historical Background This issue was raised [1] in March 1985 to address the staff's concern that there were no requirements for dynamic qualification testing or dynamic surveillance testing of large bore hydraulic snubbers (> 50 kips load …
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DESCRIPTION Historical Background This issue was identified in a RRAB memorandum [1] in March 1985 and addressed the possibility of relay contact chatter during a seismic event and its resulting effect upon safety and safety-related electrical control …
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DESCRIPTION In December 1984, the staff recommended in SECY-83-357B [1] that rulemaking with regard to H 2 control for LWRs with large, dry containments could be safely deferred due to the greater inherent capability of these containments to accommodate …
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DESCRIPTION Historical Background This issue was identified by NRR/EIB in February 1986 when it was suggested that Issue 106, "Piping and the Use of Highly Combustible Gases in Vital Areas," be expanded to include new safety concerns associated with the …
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DESCRIPTION Historical Background This issue was identified [1] by NRR when concerns were expressed that the seismic loading on equipment and pipe-mounted components may have been underestimated. These concerns could be divided into two sub- issues: …
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DESCRIPTION Historical Background This issue was raised in SECY-89-170 [1] and addressed the potential for control system vulnerabilities as a result of fire-induced alternate shutdown/control room panel interactions. Concern for these interactions arose …
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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.