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NUREG 0933
Displaying 1 - 25 of 37
The objective of this task was to evaluate whether existing requirements for valve testing provided adequate assurance of performance under design conditions. ITEM II.E.6.1: TEST ADEQUACY STUDY DESCRIPTION Historical Background The purpose of this TMI …
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DESCRIPTION The issue was raised after the occurrence of various incidents of water hammer that involved steam generator feedrings and piping, emergency core cooling systems, RHR systems, containment spray, service water, feedwater, and steam lines. The …
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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 Historical Background Many PWRs have no positive means of detecting boron dilution during cold shutdown. [1] Some operations carried out during outages (e.g., steam generator decontamination) reduce the RCS volume thereby speeding up dilution. …
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DESCRIPTION Background This issue addressed the high rate of reactor coolant pump (RCP) seal failures that challenge the makeup capacity of the ECCS in PWRs. At the time this issue was identified [1] in 1980, RCP seal failures in BWRs occurred at a …
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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 …
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DESCRIPTION Historical Background On June 19, 1981, AEOD issued a preliminary report [1] on the incident at Calvert Cliffs Unit 1 in which the plant lost both redundant trains of service water when the service water system became air-bound as a result of …
1
DESCRIPTION Historical Background On April 3, 1981, AEOD published draft NUREG-0785, "Safety Concerns Associated with Pipe Breaks in the BWR Scram System." [1] As a result of the development of these safety concerns and the findings presented in the …
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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 …
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DESCRIPTION Historical Background This issue was raised in a DL memorandum [1] to DST in March 1982 and addressed the subject of service water system (SWS) fouling at operating plants primarily by aquatic bivalves. Prior to and following this memorandum, …
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DESCRIPTION Historical Background Protection systems in nuclear power plants are required to meet the design criteria of IEEE-279, "Criteria for Protection Systems for Nuclear Power Generating Stations." [1] One of the criteria of IEEE-279 requires that …
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DESCRIPTION Historical Background Following the steam generator tube rupture (SGTR) event at Ginna in January 1982, [1] the staff proceeded to develop generic steam generator requirements which would help mitigate or reduce steam generator tube …
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DESCRIPTION Historical Background PORVs and block valves were originally designed as non-safety components in the reactor pressure control system for use only when plants are in operation. The block valves were installed because of expected leakage from …
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DESCRIPTION Historical Background During the period 1978 to 1980, there were reports of fatigue failure of thermal sleeve assemblies in the piping systems of both PWRs and BWRs. The BWR problem was addressed by GE in NEDO-21821 and was resolved with a …
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DESCRIPTION Historical Background On two occasions (February 22 and 25, 1983), Salem Unit 1 failed to scram automatically due to failure of both reactor trip breakers to open on receipt of an actuation signal. In both cases, the unit was successfully …
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DESCRIPTION Historical Background In March 1982, leaks were detected in the heat-affected zones of the safe-end-to-pipe welds in two of the 28 in. diameter recirculation loop safe ends at Nine Mile Point Unit 1. Subsequent UT revealed extensive cracking …
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DESCRIPTION Historical Background The HPCI steam supply line has two containment isolation valves in series: one inside and one outside of the containment. Both are normally open in most plants; however, two plants were found to operate with the HPCI …
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DESCRIPTION Historical Background This issue was identified [1] following a staff evaluation of allegations that improper consideration of "stiff" pipe clamps in Class 1 piping systems could result in unsafe plant operation. IE Information Notice No. …
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DESCRIPTION Historical Background On August 12, 1983, one of the three emergency diesel generators (EDG) at the Shoreham Plant failed during overload testing as a result of a fractured crankshaft. The failure occurred in EDG-102 and similar crankshaft …
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DESCRIPTION Historical Background This issue was recommended [1] for prioritization by DSI after a review of the AEOD engineering evaluation report (AEOD/E325) [2] on vapor binding of the AFW pumps at H.B.Robinson Unit 2. Further AEOD study of the event …
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DESCRIPTION Historical Background Low temperature overpressurization originally identified in NUREG 0371 [1] as item A-26. This issue later became USI A-26 and was resolved in September 1978 with a revision to SRP [2] Section 5.2. The resolution of USI …
1
DESCRIPTION Historical Background On April 17, 1984, a DSI memorandum [1] on the subject of RHR interlocks for W plants described staff concerns that the design basis for RHR interlocks had been misunderstood and that these concerns had not been …
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DESCRIPTION Historical Background The issue of using the most recent NOAA procedures for determining probable maximum precipitation (PMP) was raised [1] after an OL applicant disputed the NRC use of NOAA Hydrometeorological Report (HMR) Nos. 51 [2] and …
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DESCRIPTION Historical Background BWRs are equipped with SRVs to control primary system pressurization. Upon SRV actuation and following the clearing of air from the discharge lines, essentially pure steam is injected into the pool. Experiments indicate …
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DESCRIPTION Historical Background Indications of possible stress corrosion cracking (SCC) in the Indian Point Unit 3 (IP-3) steam generator prompted MTEB to review foreign and domestic operating experiences related to possible indications of SCC in …
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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.