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NUREG 0933
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Displaying 126 - 147 of 147
DESCRIPTION Historical Background As identified in NUREG-0471, [1] this issue involved staff evaluations of vendors' data and approaches for determining LOCA heat sources and developing staff positions as needed. The contributors to LOCA heat sources, …
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DESCRIPTION Combinations of fabrication, stress, and environmental conditions have resulted in isolated instances of stress corrosion cracking of low pressure Schedule 10 Type 304 stainless steel piping systems. Although these systems are not part of a …
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DESCRIPTION Historical Background Dose calculations by AAB/NRR in 1975 indicated that operation of the main steam isolation valve leakage control system (MSIVLCS) required for some BWRs could result in higher offsite accident doses than if the system were …
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DESCRIPTION Historical Background A BWR RHR system is designed for: (1) containment spray/suppression pool cooling, (2) fuel pool cooling augmentation, (3) low pressure coolant injection, and (4) bringing the reactor down to a cold shutdown condition. …
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DESCRIPTION This NUREG-0471 [1] task will respond to a concern of the ACRS about the effectiveness of various containment sprays to remove airborne radioactive materials which could be present within the containment following a LOCA. This concern has been …
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DESCRIPTION Historical Background The operating experience of nuclear power plants indicates that a number of valves, valve operators, and pumps fail to operate as specified in the technical specifications either under testing conditions or when they are …
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DESCRIPTION Historical Background Structural damage to the primary system, including the reactor pressure vessel and internals, associated piping and steam generator tubing in PWRs, can be caused by vibrations of sufficient magnitude. These vibrations can …
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DESCRIPTION This NUREG-0471 [1] item was an ACRS generic concern that initially addressed the common mode failure of identical components exposed to identical or nearly-identical conditions or environments. This concern was later expanded to include other …
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DESCRIPTION Licensees are required to estimate the design-basis flood levels for each nuclear power plant site consistent with the requirements in General Design Criterion 2, "Design Bases for Protection against Natural Phenomena," of Appendix A, "General …
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DESCRIPTION SRP [1] Section 15.7.3 requires an analysis of the consequences of failure of tanks containing radioactive liquids outside containment. This NUREG-0471 [2] task involves the development of a NUREG report that will describe a consistent and …
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DESCRIPTION Interpretations of NEPA require the environmental impact assessment include land use impacts and alternatives in nuclear power plant licensing cases. The staff has performed both economic and non-economic land resource assessments in …
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DESCRIPTION There are no current criteria for acceptability of solidification agents. This NUREG-0471 [1] task involves the development of criteria for acceptability of radwaste solidification agents to properly implement a process control program for the …
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DESCRIPTION Historical Background This issue is described in NUREG-0471 [1] and was raised by the ACRS who recommended that studies be made of the technique for seismic scram and the potential safety advantages and disadvantages of prompt reactor scram, …
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DESCRIPTION This issue was documented in NUREG-0471 [1] and stemmed from an ACRS recommendation that the staff explore diverse means of obtaining ECCS capability for future plants. The ACRS also recommended the staff explore the issue in the context of …
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DESCRIPTION This NUREG-0471 [1] item is an ACRS generic concern which involves assessing the uncertainties in calculations of the control rod drop accident, including the choice of negative reactivity insertion rate due to a scram and the potential …
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The following is an excerpt from Appendix VI, Section 2 of WASH-1400 (NUREG-75/104), "Reactor Safety Study: An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants," dated October 1975. On this page: Section 2 Releases from …
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The table below identifies those generic issues (GIs) that have resulted in a regulatory product: hence, they may be applicable to operating and future nuclear power plants. In accordance with Title 10 of the Code of Federal Regulations (10 CFR) …
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TABLE 1 RISK THRESHOLDS (a) The priority rank is always HIGH when any of the following risk (or risk related) thresholds are estimated to be exceeded (or when extraordinary uncertainty suggests that they may well be exceeded): (1) 11000 person-rem …
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This appendix documents those activities related to generic issues, i.e., related generic activities (RGA) that did not meet the criteria for designation as generic issues (GI) but were important enough to require the development of Action Plans by NRR to …
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This appendix documents those generic communication and compliance activities (GCCA) completed by NRR that did not meet the criteria for designation as generic issues (GI), but were important enough to require the issuance of Information Notices (IN) …
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This appendix documents those non-reactor GSIs identified, prioritized, and resolved by NMSS. As stated in SECY-98-001, [1] the prioritization procedure for these issues is contained in NMSS Policy and Procedures Letter 1-57, [2] "NMSS Generic Issues …
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BACKGROUND I. History On October 8, 1976, the Commission directed the staff to develop "a program plan for resolution of generic issues (GIs) and completion of technical projects." The Commission further requested that "this plan should include: task …
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Page Last Reviewed/Updated 3/1/2026
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