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NUREG 0933
Displaying 26 - 50 of 88
DESCRIPTION Inadvertent operation of containment sprays can result in a rapid depressurization of the containment building. Where containment external design pressure may be exceeded, many plants have been provided with vacuum breakers or control system …
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DESCRIPTION Various kinds of insulation are used on piping and components inside the containment of a nuclear power plant. The concern of this NUREG-0471 [1] item was the behavior of insulation under pipe break accident conditions where the potential …
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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 …
1
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 …
1
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 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 Licensee Event Reports (LERs) …
1
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 …
1
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, …
1
DESCRIPTION Historical Background This issue apparently originated as a DOR proposal and was discussed in SECY-80-325. [1] The issue as previously evaluated [2] is whether periodic replacement of the squib charges and circuit checks of the traveling …
1
DESCRIPTION This issue involves a potential deficiency in the ability to control leakage past the main steam isolation valves (MSIV) in BWR plants. Requirements for MSIV leakage control systems outlined in Regulatory Guide 1.96 [1] were developed as a …
1
DESCRIPTION Historical Background In the event of a LOCA in which offsite power remains available, the reactor operator may choose to shut down the emergency diesel generators shortly after the LOCA. If, after the diesels are shut down, offsite power is …
1
DESCRIPTION On November 10, 1979, an event occurred at the Oconee Power Station, Unit 3, that resulted in loss of power to a non-Class 1E 120-volt AC single-phase power panel that supplied power to the integrated control system (ICS) and the nonnuclear …
1
DESCRIPTION Historical Background This concern was raised [1] because of the potential for a high-altitude nuclear weapon detonation causing a large electromagnetic pulse (EMP) which subsequently could induce large currents and voltages in electrical …
1
DESCRIPTION This issue concerns the slow loss of control air pressure in the scram system of BWRs. [1] Air pressure dropping at a certain rate will first allow some of the CRD scram outlet valves to open slightly, thus filling the scram discharge volume …
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DESCRIPTION In a San Onofre Unit 1 Preliminary Notification issued in September 1980, it was reported that, during testing, the licensee had identified a problem with the design of the diesel generator sequencing circuitry. This problem occurred when a …
1
DESCRIPTION This issue was identified during plant design and emergency procedure reviews which raised questions as to whether certain safety actions have to be accomplished automatically or whether manual operator action is acceptable. CONCLUSION This …
1
DESCRIPTION Historical Background Prior to 1981, the number of bolting-related incidents reported by licensees was on the increase. A large number of these were related to primary pressure boundary applications and major component support structures. As a …
1
DESCRIPTION Historical Background This issue corresponds to AEOD recommendation 4 highlighted in an AEOD memorandum [1] to NRR in December 1980. The AEOD recommendation resulted from a natural circulation cooldown event from full power at St. Lucie Unit …
1
DESCRIPTION This issue was identified in an AEOD memorandum [1] to NRR in May 1980. The concern is related to a feature on some B&W reactors in which the failure of the Non-Nuclear Instrumentation and Integrated Control System (NNI/ICS) power supply …
1
DESCRIPTION Historical Background This issue was identified [1] after AEOD completed a study on internal appurtenances in LWRs. This study, AEOD/E101, [2] was initiated because of the relatively high number of LERs that described events in which internal …
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DESCRIPTION Historical Background This issue was identified [1] when AEOD expressed concerns about the use inside containment of a particular polymer coating that could flake off and fail when subjected to DBA conditions. In addition to the concern for …
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DESCRIPTION Concerns for inoperability of instrumentation due to extreme cold weather were raised [1] by AEOD after an event at Arkansas Nuclear One, Unit 2 (ANO-2) in which all four RWST instrumentation channels were lost when the level transmitters …
1
DESCRIPTION An AEOD memorandum [1] to NRR on July 15, 1980, identified the concern that some operating nuclear power plants do not have technical specifications or Administrative controls governing operational restrictions for Class 1E 120 VAC vital …
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DESCRIPTION Historical Background In January 1982, AEOD published a report (AEOD/C201 [1] ) on safety concerns associated with reactor vessel level instrumentation in BWRs. The report was forwared to NRR for further action. Safety Significance BWRs use …
1
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 …
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.