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NUREG 0933
Displaying 26 - 50 of 91
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 …
1
DESCRIPTION Historical Background This NUREG-0471 [1] item deals with two concerns regarding the ice condenser containment design. The first concern arises from an ACRS comment on the D. C. Cook Unit 1 review. The normal procedure used by the staff (CSB) …
1
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 …
1
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 …
1
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 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 Historical Background This concern was identified by the ACRS as a result of their review [1] of LERs covering a period from 1976 through 1978 and, subsequently, a comprehensive review of the problem was completed which covered the LERs and …
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 Historical Background This issue effects all PWR-type reactors (Westinghouse, CE, B&W). The issue as described, [1] , [2] , [3] concerns postulated accidents resulting from a steam line break which consequentially results in a steam generator …
1
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 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 …
1
DESCRIPTION In AEOD/CO05, [1] AEOD identified potential safety problems concerning steam generator overfill due to control system failures and combined primary and secondary blowdown. As a result of discussions with the Commissioners and the EDO, NRR …
1
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 …
1
DESCRIPTION AEOD issued a memorandum [1] in which a potential safety issue involving combined primary and secondary system LOCAs was raised. The issue was discussed at Commission meetings on October 16, 1980 and on November 10, 1980. NRR informed AEOD of …
1
DESCRIPTION Historical Background This issue was initiated in response to an immediate action memorandum [1] issued by AEOD in September 1981 regarding desiccant contamination of instrument air lines. NRR responded to the AEOD memorandum by establishing a …
1
DESCRIPTION Historical Background This proposed generic issue has its origin in a January 1981 event at Millstone Unit 2. [1] An operator inadvertently opened a 125V DC main feeder breaker, causing the loss of one of the two redundant emergency systems …
1
DESCRIPTION Historical Background On April 7, 1980, Arkansas Nuclear One (ANO) Units 1 and 2 experienced a significant event resulting from a loss of offsite power. Although both units were safely shut down, the analysis and evaluation of the event …
1
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 On April 8, 1981, while operating at 100% power, the Arkansas Nuclear One, Unit 1 experienced a moderate feedwater-overfill transient in one of its once-through steam generators. AEOD performed a case study of this event, concluded that the …
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.