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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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TASK III.A.1: IMPROVE LICENSEE EMERGENCY PREPAREDNESS - SHORT-TERM The objectives of this task were to improve and upgrade licensee emergency preparedness by requiring improvements in facilities, plans, procedures, offsite support, technical assistance, …
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The objective of this task was to upgrade the emergency preparedness of nuclear power plants. Specific criteria to meet this objective were delineated in NUREG-0654. [1] ITEM III.A.2.1: AMEND 10 CFR 50 AND 10 CFR 50, APPENDIX E The four parts of this item …
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The objective of this task is to enable NRC, in the event of a nuclear accident at a licensed reactor facility, to: (1) monitor and evaluate the situation and potential hazards, (2) advise the licensee's operating staff as needed, and (3) in an extreme …
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The objective of this task is to upgrade the state of emergency preparedness of State and local governments affected by nuclear facilities. The Federal Emergency Management Agency was given the lead on this effort by the President on December 7, 1979. …
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The objectives of this task are: (1) to have information available for the news media and the public describing how nuclear plants operate, radiation and its health effects, and protective actions against radiation; and (2) to provide training for members …
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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 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 …
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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 …
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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 …
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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 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 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 …
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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 …
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DESCRIPTION The fuel assembly is a highly nonlinear structure which can be subjected to substantial loadings during seismic excitations and LOCA transients. The integrity of this assembly is critical for plant safety. Extensive work has been completed by …
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DESCRIPTION Present NRC actions taken in response to a serious incident are directed from an Incident Response Center (IRC). To implement an adequate response, it is necessary that the IRC be equipped with appropriate communications services, information …
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DESCRIPTION Historical Background The ACRS raised a concern that there is a need to develop criteria for vibration monitoring systems which could provide early warning of excessive vibration inside the reactor vessel. This item is documented in …
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DESCRIPTION Historical Background This NUREG-0471 [1] item was developed because of concerns regarding the long-term capability of hermetically- sealed instruments and equipment which must function in postaccident environments. Certain classes of …
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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 Background This issue was identified [1] after a review of LERs involving the failure or degraded performance of essential equipment that resulted from the failure of protective devices such as fuses and circuit breakers. The affected systems …
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DESCRIPTION Historical Background This issue is identified in Appendix D of NUREG-0572 [1] and is one of the key observations made after the ACRS requested its members and consultants to make comprehensive reviews of all LERs issued during the years 1976, …
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DESCRIPTION BWRs are operated in such a manner as to attempt to produce an axial power distribution with as little peaking as possible. When the reactor moderator is not boiling (e.g., during startup and heatup), the axial flux distribution is much more …
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Page Last Reviewed/Updated 3/1/2026
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