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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 49935 | 20 March 2014 13:11:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Fire-Related Unanalyzed Condition That Could Impact Credited Safe Shutdown Analysis | This is a non-emergency notification. In preparation for converting from 10 CFR 50, Appendix R, to NFPA (National Fire Protection Association) Standard 805, an update to the Brunswick Steam Electric Plant (BSEP) Safe Shutdown Analysis has been performed which identified circuit configurations where fire damage, under certain postulated fire scenarios, could impact the ability to safely shut down following a fire, in accordance with 10 CFR 50, Appendix R. Affected fire areas are CB-23E, RB1-N, RB1-S, RB2-N, RB2-S, TB1, DG-07, and DG-16E. A fire in one of these areas could potentially affect the post fire capability of the following safe shutdown systems: 1) Containment Overpressure Protection, 2) Emergency Bus Load Shed, 3) Control Room HVAC, or 4) Emergency Diesel Generator Building HVAC. This is reportable as an unanalyzed condition that significantly degrades plant safety in accordance with 10 CFR 50.72(b)(3)(ii)(B). Fire protection compensatory measures (i.e. roving fire watches) currently exist in the affected fire areas. This event did not result in any adverse impact to the health and safety of the public. The safety significance is minimal. Fire watches were already ongoing in these areas prior to the time of discovery. The conditions identified here are based on hypothetical fire scenarios that have not actually occurred. This condition has been entered into the Corrective Action Program (i.e. CR 676576). Previous similar events were reported in NRC Event Reports 47341 and 49222, and in Brunswick LERs 1-2011-002 and 1-2013-002. The NRC Resident Inspector has been notified. | Emergency Diesel Generator HVAC | |
| ENS 49222 | 29 July 2013 14:42:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Fire Related Unanalyzed Condition That Could Impact Credited Safe Shutdown Analysis | This is a non-emergency notification. In preparation for converting from 10 CFR 50. Appendix R, to NFPA (National Fire Protection Association) 805, a review of the Brunswick Steam Electric Plant (BSEP) Safe Shutdown Analysis identified conditions that may not ensure required equipment remains available under certain postulated fire scenarios. The analysis determined that the effects of a postulated fire in specific fire areas could prevent critical systems or components from performing their intended functions, potentially resulting in the inability to achieve and maintain safe shutdown. Alternate safe shutdown procedures currently credited with bringing about operator actions to mitigate a postulated fire have been found not to contain needed actions. Affected fire areas are RB1-N, RB1-S, RB2-N, RB2-S, DG-07, and DG-08E. A fire in one of these areas could potentially adversely affect Emergency Diesel Generators (EDGs) 2, 3, and 4 along with EDG Building ventilation system components that could fail vital auxiliaries (HVAC) to the affected switchgear rooms. This is reportable as an unanalyzed condition that significantly degrades plant safety in accordance with 10 CFR 50.72(b)(3)(ii)(B). Most required fire watches were already in place as a result of a previous similar event reported in Brunswick ENS Report 47341 and Brunswick LER 1-2011-002. Two additional actions have been taken to establish fire watches in areas containing electrical busses E6 and E7. This condition has been entered into the Corrective Action Program (i.e., CR 619341). The licensee has notified the NRC Resident Inspector. | Emergency Diesel Generator HVAC | |
| ENS 48253 | 30 August 2012 01:34:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Local Control Capability of Emergency Diesel Generator #2 Not Available Under Certain Conditions | On August 28, 2012, during planned maintenance on Emergency Diesel Generator No.2, a post-maintenance continuity testing associated with the Alternate Safety Shutdown (ASSD) switch on Emergency Diesel Generator No.2 (EDG 2) revealed unexpected results when the switch was taken to the LOCAL position. Troubleshooting activities determined the switch to be operating properly, meaning the contacts are actually open. However, a current path preventing isolation of the control room circuit remained. It was determined that a wire, not identified in EDG wiring diagrams, created a short between two ASSD switch contacts. At 2134 hours Eastern Daylight Time (EDT) on August 29, 2012, it was concluded that the condition may impact the ability of EDG 2 to perform its intended ASSD function. In the event of a fire, an induced fault could potentially affect the ability to locally control EDG 2. Local control of EDG 2 is credited in the safe shutdown analysis. Therefore, this condition is being reported in accordance with 10 CFR 50.72(b)(3)(ii)(B) as an unanalyzed condition. This condition did not result in any adverse impact to the health and safety of the public. This condition has no affect on the Technical Specification operability of EDG 2 and it was fully capable of performing its intended design basis accident response functions given the as-found condition of the circuitry. The initial safety significance of this condition is minimal. This condition has no affect on the Technical Specification operability of EDG 2 and it was fully capable of performing its intended design basis accident response functions given the as-found condition of the circuitry. Repairs to the ASSD switch wiring have been completed. Post maintenance testing is occurring at this time. Planning is in progress to inspect the remaining Emergency Diesel Generators for similar conditions. The licensee has notified the NRC Resident Inspector. | Emergency Diesel Generator | 05000325/LER-2012-005 |
| ENS 47452 | 17 November 2011 18:45:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition on Startup Following Maintenance Outage | On November 17, 2011, at 1345 Eastern Standard Time (EST), it was determined that an unanalyzed condition that significantly degraded plant safety existed on Unit 2 following a mid-cycle maintenance outage which required reactor vessel disassembly. Unit 2 had reached Mode 2 (i.e., Startup) during the power ascension, when elevated drywell leakage was identified and the plant was shutdown (see EN#47444). Unit 2 was subsequently operating in Mode 4 (i.e., Cold Shutdown) to allow for drywell entry. During leak investigation activities, it was determined that the reactor pressure vessel (RPV) head was not fully tensioned. This condition is being reported in accordance with 10 CFR 50.72(b)(3)(ii)(B) as a condition of the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety. The safety significance of this event was minimal. Unit 2 was in power ascension operating at a maximum of approximately 7% of rated thermal power. Control Room Operators took appropriate action to shutdown the Unit when elevated drywell leakage was identified. The NRC Senior Resident Inspector has been notified. | Reactor Pressure Vessel | |
| ENS 47341 | 13 October 2011 14:00:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Fire Related Unanalyzed Condition That Could Impact Equipment Credited in Safe Shutdown Analysis | In preparation for converting from 10 CFR 50, Appendix R, to NFPA (National Fire Protection Association) 805, a review of the Brunswick Steam Electric Plant (BSEP) Safe Shutdown Analysis identified conditions that may not ensure a protected train of equipment remains available under certain postulated fire scenarios. The analysis determined that a postulated fire in specific fire areas could cause spurious actuation of critical components, potentially resulting in loss of equipment required for Safe Shutdown. A fire in one of the specified fire areas could potentially adversely affect the following; Suppression Pool level instrument 2-CAC-LT-2602, Residual Heat Removal net positive suction head (i.e., Drywell Containment Overpressure), Reactor Core Isolation Cooling (RCIC), Emergency bus E-1, and Emergency bus E-3. The initial safety significance of this event is minimal. Fire watches have been established for the affected portions of fire areas RB2-01, RB1-01, TB1, CB-02, and CB-23E. Additionally, fire detection and suppression equipment in the affected areas is fully functional. Fire watches have been established for the affected portions of fire areas RB2-01, RB1-01, TB1, CB-02, and CB-23E. This condition has been entered Into the Corrective Action Program (i.e., CR 493784). This is reportable as an unanalyzed condition that significantly degrades plant safety in accordance with 10 CFR 50.72(b)(3)(ii)(B). Compensatory measures have been established to provide an hourly fire watch for the affected fire areas. The licensee has notified the NRC Resident Inspector. | Reactor Core Isolation Cooling Residual Heat Removal | 05000325/LER-2011-002 |
| ENS 44425 | 19 August 2008 15:10:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Alternate Safe Shutdown Local Start Capability of All Four Emergency Diesel Generators May Not Be Available Under Certain Conditions | On August 18, 2008, during biennial (i.e., every two years) non-Technical Specification related testing, Emergency Diesel Generator (EDG) No. 4 failed to start from the EDG local control panel. This occurred with the EDG aligned for Alternate Safe Shutdown (ASSD) response versus its normal alignment. The purpose of the testing is to confirm that the EDG control logic is isolated from the control room and only operable via local controls. Based on troubleshooting activities, it was discovered that this failure to start was due to improper wiring in the circuitry associated with the EDG lockout relay. The improper wiring was introduced by a modification performed on all four EDGs in 2007. At 1110 hours Eastern Daylight Time (EDT) on August 19, 2008, it was concluded that this condition impacted the ability of EDG Nos. 2, 3, and 4 to perform their intended ASSD function. The 2007, modification affected the termination point for the power to the EDG lockout relay reset coil. Per the modification, this termination point was incorrectly connected to the downstream side of the alternate shutdown isolation switch which results in a loss of power to the lockout reset when the switch is placed in local control. Therefore, in the unlikely event of a fire, an induced failure could potentially initiate a lockout signal which could not be reset from the EDG local control panel. Although the modification was installed on all four EDGs, only the local control of EDGs 2, 3, and 4 is credited in the safe shutdown analysis. This condition does not affect the Technical Specification operability of the EDGs and they remain fully capable of performing their intended design basis accident response functions. The NRC Resident Inspector has been notified. Compensatory fire watches have been established and an engineering change package to correct the condition is being developed. | Emergency Diesel Generator | 05000325/LER-2008-006 |
| ENS 41160 | 29 October 2004 20:00:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition Discovered in 480 Volt Switchgear Room | The following information was obtained from the licensee via facsimile: On October 29, 2004, at approximately 1600 hours, control room operators were informed by engineering personnel that raceways were located closer than 20 feet from the redundant division without being protected by a fire barrier wrap in accordance with the requirements of 10CFR50 Appendix R, Section III.G.2. In 480 volt switchgear room E7, Brunswick is committed to maintaining redundant safe shutdown circuits by a minimum of 20 feet with no-intervening combustibles, unless the circuits are protected by a one-hour fire barrier wrap. Two conduits containing Division II circuits were identified closer than 20 feet to their redundant counterparts with no wrap installed. The (NRC) Resident Inspector has been notified. INITIAL SAFETY SIGNIFICANCE EVALUATION The initial safety significance of this condition is considered to be minimal. When the condition was recognized, impairments were established in accordance with the fire protection program and compensatory measures were implemented. In addition, fire detection systems in the affected area have been verified to be operable. The affected area is maintained as a combustible free separation zone. CORRECTIVE ACTIONS Impairments have been initiated and appropriate compensatory measures established. The root cause and additional corrective actions are being documented in accordance with the corrective action program. | 05000324/LER-2004-003 |