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NUREG 0933
Displaying 1 - 25 of 31
DESCRIPTION Background Flow-induced vibrations caused by vortex shedding resulting from rapid area change, buffeting due to random flow turbulence, fluid structures interaction instability, leakage excitation, steady operation of positive displacement …
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 Historical Background A memorandum [1] from AEOD to NRR dated May 23, 1980 drew attention to the generic issue of BWR jet pump integrity. The concern that motivated the AEOD memo was a February 1980 jet pump failure at Dresden Unit 3, together …
1
DESCRIPTION Historical Background This is an ACRS concern raised by the Subcommittee on TMI-2 Implications in October 1979. This issue centers around the possibility of a breach in the reactor coolant system boundary caused by the failure of nonsafety …
1
DESCRIPTION Historical Background Cracking has occurred in PWR piping systems as a result of stress corrosion, vibratory and thermal fatigue, and dynamic loading. However, as of February 1981, no cracking had been experienced in the primary system piping …
1
DESCRIPTION Historical Background This issue addressed the potential problem of radiation embrittlement of reactor vessel support structures (RVSS). It was originally identified as a Candidate USI in NUREG-0705 [1] where it was recommended for further …
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 …
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DESCRIPTION Pressurized water reactors are susceptible to certain types of hypothetical accidents that, under circumstances such as operation of the reactor beyond a critical time in its life, could result in failure of the pressure vessel as a result of …
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 was raised by AEOD [1] , [2] and involved isolation of the reactor coolant system charging and letdown system following a spurious safety injection transient at H. B. Robinson on January 29, 1981. Following the …
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 In response to a 1967 ACRS concern relative to the potential of melting and subsequent disintegration of a portion of a fuel assembly due to inlet orifice flow blockage, GE submitted NEDO-10174 [1] in May 1970. As a …
1
DESCRIPTION This issue was raised in November 1981 by the Materials Engineering Branch (MTEB), DE/NRR, and was based on the concern that NRC provides no control regulations or guides for bolting other than for the reactor vessel head. CONCLUSION In …
1
DESCRIPTION Historical Background Cracks were found in the normal make-up high pressure injection (MU/HPI) nozzles of several B&W plants following an inspection of the 8 B&W plants licensed to operate. These cracks appeared to be directly related to loose …
1
DESCRIPTION Background This issue was raised in Board Notification No. 82-81 [1] and addressed the failure of control rod drive (CRD) guide tube support pins in W reactors. The first pin failure at a U.S. plant occurred at North Anna 1 in 1982 where a …
1
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 …
1
DESCRIPTION Historical Background STS for newer OLs require licensees to keep account of the number of transient occurrences to ensure that transient limits, based on design assumptions, are not exceeded. However, a number of older plants for which …
1
DESCRIPTION Historical Background On March 18, 1983, B&W expressed [1] concern for unanalyzed reactor vessel thermal stress that could occur during natural convection cooldown of PWRs. The concern emerged from a preliminary B&W evaluation of a voiding …
1
DESCRIPTION Historical Background This issue was identified by the ACRS in 1978 during the operating license reviews of some BWRs. The ACRS posed questions concerning the likelihood and effects of a LOCA which could cause interactions with the CRD …
1
DESCRIPTION This issue was raised by the ACRS in a memorandum [1] to the Commission in October 1983. Following the TMI-2 accident, the purpose and use of PORVs had been the subject of considerable analyses and discussion. The original purpose for which …
1
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 …
1
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 This issue was identified during the staff review of the Indian Point 2 and Zion PRAs [1] ; in both of these studies, the dominant interfacing systems LOCA events were estimated to be through the RHR suction valves. The …
1
DESCRIPTION In the prioritization of Issue 22, "Inadvertent Boron Dilution," it was found that inadvertent boron dilution events during cold shutdown operation do not constitute a significant risk to the public. Further work by DSI confirmed this …
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DESCRIPTION Historical Background Potential stress corrosion cracking (SCC) of reactor vessel (RV) closure studs was raised as a GSI in December 1984. [1] Concerns were expressed that two contributors to SCC had not been subject to careful research and …
1
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