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| Entered date | Event description | |
|---|---|---|
| ENS 41794 | 24 June 2005 13:35:00 | |
| ENS 41547 | 29 March 2005 19:17:00 | The following is a portion of a facsimile submitted by GE Energy-Nuclear: The defect is a calculation of an anticipated operational occurrence (AOO), which predicts that the Pressure Regulator Failure Maximum Demand (Open) (PRFO) transient will be terminated by a high water level trip as a result of level swell in the reactor. An improved (and approved) model predicts that MSIV closure will occur when steam line pressure reaches the low-pressure isolation setpoint (LPIS), rather than terminate due to a high water level trip. Depending upon the plant-specific response to a PRFO, including the value of the LPIS, reactor steam dome pressure could decrease to below 785 psig while thermal power exceeds 25% of rated, which would be a violation of SL 2.1.1.1. This constitutes a Defect as defined in 10 CFR 21.3, even though there is no safety hazard created. SL 2.1.1.1 was intended to protect fuel cladding integrity during startup conditions without the need to perform a Critical Power Ratio (CPR) calculation. The AOO of concern is a transient from normal operating conditions that causes CPR to increase, so the event produces additional margin to the Minimum Critical Power Ratio Safety Limit (SLMCPR) and does not threaten fuel cladding integrity. The LPIS should not be considered as a Limiting Safety System Setting (LSSS) for SL 2.1.1.1 since it does not provide a 'significant safety function' with regard to protecting fuel cladding integrity. This indicates that SL 2.1.1.1 is overly conservative because an event that causes CPR to increase and does not threaten fuel cladding integrity, may result in exceeding a reactor core SL. Notified Plants: AFFECTED: Nine Mile Point 2, Fermi 2, Pilgrim, Vermont Yankee, Limerick 1 & 2, Peach Bottom 2 & 3, Perry 1, and Hope Creek. POTENTIALLY AFFECTED: Clinton, Oyster Creek, Brunswick 1 & 2, Nine Mile Point 1, FitzPatrick, Grand Gulf, River Bend, Dresden 2 & 3, LaSalle 1 & 2, Quad Cities 1 & 2, Cooper, Duane Arnold, Monticello, Hatch 1 & 2, Browns Ferry 1, 2 & 3. |
| ENS 40982 | 24 August 2004 17:36:00 | Global Nuclear Fuel (GNF) and GE Nuclear Energy (GENE) have determined that the current GNF process for determination of the Safety Limit Minimum Critical Power Ratio (SLMCPR) can result in a non-conservative SLMCPR. GENE has historically used a non-conservative SLMCPR impact of 0.01 as the threshold for reportability under 10CFR21. A preliminary screening evaluation has been completed for all plants operating with a SLMCPR calculated by GNF to determine those that have a nonconservative impact of 0.01 or greater. Verification has been completed for those plants that the screen showed to have a non-conservative SLMCPR impact of 0.01 or greater. Verification has not been completed for the plants that the screen showed had an impact of less than 0.01 or were unaffected, thus requiring a 60-Day Interim Report notification pending verification completion. The plants for which GNF calculates the SLMCPR are identified (below). Those plants for which the preliminary screen indicated that the current SLMCPR is unaffected are identified as a 60-Day Interim Report. Upon completion of verification (assuming the results of the screen are confirmed) the status of these plants will be changed to Not Reportable. GENE will provide a follow-up report to the NRC by September 29, 2004. The plants for which the current SLMCPR is non-conservative by 0.01 or greater are identified as a Reportable Condition under 10CFR21.21(d). These plants will take action to address the Reportable Condition. GNF has notified all plants that have been confirmed to be affected. The plants that have a non-conservative SLMCPR for current plant operation will take action to mitigate the potential impact. Depending on the specific circumstances, mitigating actions to protect the SLMCPR may include increasing the OLMCPR to assure compliance with the low flow calculated SLMCPR. In some cases sufficient conservatism may exist in the OLMCPR at low flow to bound the increased SLMCPR. Each affected plant will notify the NRC and take appropriate action if their Technical Specifications are affected. There are no actions necessary for the plants that are unaffected pending completion of verification. If, in the course of verification, GNF determines that there is an impact, the affected utility will be notified immediately. GNF will complete the verification and GENE, will provide a follow-up letter to the NRC by September 29, 2004. Affected and 60-Day Interim Notification Plants: 60-Day Interim Report: Clinton, Oyster Creek, Brunswick 1, Brunswick 2, Nine Mile Point 2, Fitzpatrick (60-Day interim report for current operation, Reportable Condition for SLMCPR licensing submittal), Pilgrim, Vermont Yankee, Dresden 2, Dresden 3, LaSalle 1, LaSalle 2, Limerick 1, Limerick 2, Peach Bottom 2, Peach Bottom 3, Quad Cities 1, Quad Cities 2, Perry 1, Duane Arnold, Monticello, Hope Creek, Hatch 1, Hatch 2, Browns Ferry 2 Reportable Condition: Nine Mile Point 1, Fermi 2, Fitzpatrick (60-Day interim report for current operation, Reportable Condition for SLMCPR licensing submittal), Cooper.
In part, the update reads: The evaluation for the potentially unaffected plants has now been completed. The results of the screening calculation have been confirmed: this is not a reportable condition for the plants that were previously identified as a 60-Day Interim Report notification in MFN 04-081" (See above). Notified R1DO(Rogge), R2DO(Lesser), R3DO(Kozak), R4DO(Howell) and NRREO(Hodges). |
| ENS 40962 | 16 August 2004 16:04:00 | This letter provides a 10 CFR21(a)(2) 60 - Day Interim Report notification regarding a potential issue with the Level 3 trip from the narrow range water level instruments that initiate reactor scram. A conservative evaluation by GE Nuclear Energy (GENE) has determined that water level instruments may indicate high by as much as 8 inches, should the reactor water level drop below the dryer seal skirt. At issue is whether with the actual water level as much as 8 inches lower than indicated, the top of active fuel (TAF) will be uncovered for the limiting loss of feedwater event due to 1. Actual water level being lower than indicated when the Level 3 trip occurs, or 2. Failure of the Level 3 trip to occur if water level drops below the narrow range instrument variable leg tap prior to reaching the Level 3 trip setpoint. Because TAF is a Technical Specification Safety Limit, TAF uncovery for a loss of feedwater event would be a Reportable Condition. However, it would not lead to a significant safety hazard due to multiple automatic and passive protection features of a BWR. GENE has completed analysis for BWR 2/3 plants and determined that this is not a reportable condition (i.e., the TAF safety limit is protected). GENE has not completed the analyses for BWR /4-/6 plants. For these plants, GENE has determined that actual water level being lower than indicated by up to 8 inches when the Level 3 trip occurs does not lead to TAF uncovery. However, for these plants GENE has not determined if water level could drop below the narrow range instrument variable leg tap prior to reaching the Level 3 trip setpoint. Therefore, this letter is issued as a 60 - Day Interim Notification under 10 CFR21.21(a)(2) to the BWR/4-/6 plants listed in Attachment 1. The potentially affected plants are being concurrently notified of this situation by a GENE Safety Communication letter. GENE is committed to complete the evaluation by October 11, 2004. Attachment 1 Plants List Below: (Affected Plants - Evaluation incomplete) Clinton, Brunswick Units 1 & 2, Nine Mile Point Unit 2, Fermi Unit 2, Columbia, Fitzpatrick, Grand Gulf, River Bend, La Salle Units 1&2, Limerick Units 1 & 2, Peach Bottom Units 1 & 2, Perry, Cooper, Duane Arnold, Susquehanna Units 1 & 2, Hope Creek, Hatch Units 1 & 2, Browns Ferry Units 1 (plant is in an extended shutdown),2 & 3.
GENE has now completed the evaluation and has determined that all BWR/2/3/6 and ABWR plants, and most BWR/4-5 plants are not susceptible to the condition that could result in a failure of the Level 3 trip to occur. These plants are identified as Not Reportable in Attachment 1 (of GENE Part 21 final report to the NRC dated October 11, 2004). However, for same BWR/4-5 plants, GENE does not have sufficient information to determine if the condition exists that could result in a failure of the Level 3 trip to occur. These plants are notified, as a Transfer of Information under 10CFR21.21(b), as identified in Attachment 1 (of GENE Part 21 final report to the NRC dated October 11, 2004). The plants where insufficient information for a final determination have been identified by GENE as Columbia, Susquehanna 1 and 2, Browns Ferry 2 and 3. Notified R1DO (Barkley), R2DO(Ogle), R3DO (Clayton) and R4DO (Bywater) and NRR (Hodge). |