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NUREG 0933
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The objective of this task was to provide instrumentation to monitor plant variables and systems during and following an accident. Indications of plant variables and status of systems important to safety are required by the plant operator (licensee) …
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The objective of this task was to increase the reliability and diversification of the electrical power supplies for certain safety-related equipment. ITEM II.G.1: POWER SUPPLIES FOR PRESSURIZER RELIEF VALVES, BLOCK VALVES, AND LEVEL INDICATORS DESCRIPTION …
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The objective of this task was to improve public radiation protection in the event of a nuclear power plant accident by improving (1) radioactive effluent monitoring, (2) the dose analysis for accidental releases of radioiodine, tritium, and carbon-14, …
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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 the time this issue was identified in NUREG-0371, [1] trends in the design of nuclear power plants showed an increase in the use of digital computer technology in safety-related instrumentation and control systems. The first application of …
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DESCRIPTION CP applicants for which SERs were issued after July 1, 1974, were required by the NRC to qualify all safety- related equipment to IEEE 323. [1] From the time this standard was originated, the industry developed methods that were used to …
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). …
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DESCRIPTION This NUREG-0371 [1] item was initiated in order to develop criteria and guidelines to be used by applicants, licensees, and staff reviewers to support implementation of Regulatory Guide 1.97, [2] Revision 1. Such criteria and guidelines were …
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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 …
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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 Nuclear power plant instrumentation and control systems are composed of safety-related protection systems and non-safety-related control systems. The safety-related protection systems are designed to satisfy the General Design Criteria …
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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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