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NUREG 0933
Displaying 1 - 25 of 31
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 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 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 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 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 Historical Background The containment flooding issue stems from a flooding event that occurred at the Indian Point 2 reactor in October 1980. [1] A large quantity of water leaked from fan coolers onto the containment building floor and …
1
DESCRIPTION Historical Background The SRVs of a BWR plant provide protection against overpressurization of the reactor primary system. During normal operation, the SRVs which are mounted in the main steam lines open on high pressure permitting steam to …
1
DESCRIPTION Historical Background In BWRs, SRVs are mounted on the main steam line inside the drywell. Each SRV discharge is piped through its own discharge line (tailpipe) to a point below the minimum water level in the primary containment suppression …
1
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 …
1
DESCRIPTION Historical Background This issue was initiated to address concerns raised by the Union of Concerned Scientists. (References [1] , [2] , and [3] .) The purposes for including this issue as a generic issue are to: (1) provide brief background …
1
DESCRIPTION Historical Background This concern was raised by an NRC resident inspector who questioned the practice of leaving the refueling canal drain valve in the closed position during operations at H. B. Robinson Unit 2. [1] A subsequent investigation …
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 Historical Background Combustible gases such as H 2 , propane, and acetylene are used during normal operations of nuclear power plants in limited quantities and for relatively short periods of time. H 2 , the most prevalent of these gases in …
1
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 …
1
DESCRIPTION Historical Background This issue was raised [1] in March 1985 to address the staff's concern that there were no requirements for dynamic qualification testing or dynamic surveillance testing of large bore hydraulic snubbers (> 50 kips load …
1
DESCRIPTION Historical Background This issue was identified in a RRAB memorandum [1] in March 1985 and addressed the possibility of relay contact chatter during a seismic event and its resulting effect upon safety and safety-related electrical control …
1
DESCRIPTION Historical Background On January 27, 1985, a dented and leaking tendon grease cap was found during inspections at Farley Unit 2 prior to the integrated leak rate test of the prestressed concrete containment structure. Subsequent detailed …
1
DESCRIPTION In December 1984, the staff recommended in SECY-83-357B [1] that rulemaking with regard to H 2 control for LWRs with large, dry containments could be safely deferred due to the greater inherent capability of these containments to accommodate …
1
DESCRIPTION Historical Background The issue of deinerting upon discovery of RCS leakage was identified [1] by DL/NRR, based on data collected by OIE. The related but separate concern of deinerting with one train of a safety system inoperable was also …
1
DESCRIPTION The issue, as originally proposed, [1] addressed the concern for fission product removal by the containment sprays and suppression pools in PWRs and BWRs, respectively; however, it was expanded to include PWR plants that use ice condenser …
1
DESCRIPTION Historical Background This issue was identified [1] by NRR when concerns were expressed that the seismic loading on equipment and pipe-mounted components may have been underestimated. These concerns could be divided into two sub- issues: …
1
DESCRIPTION Historical Background This issue was identified [1] by DSIR/RES and addressed the concern for overpressurization of containment piping penetrations following a containment isolation and subsequent heat-up. Containment isolation at all nuclear …
1
DESCRIPTION In 1977, the NRC initiated the Systematic Evaluation Program (SEP) to review the designs of 51 older, operating nuclear power plants. The SEP was divided into 2 phases. In Phase I, the staff defined 137 issues for which regulatory requirements …
1
The results of ongoing staff-sponsored research which culminated in the assessment of risk at five U.S. nuclear reactors (NUREG-1150 [1] ) indicated that, for the Peach Bottom MARK I containment, the core-melt probability was relatively low. However, it …
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.