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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 47755 | 19 March 2012 21:12:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Potential for Aerial Lift to Impact Service Water Piping During Seismic Event | At 1612 on 3/19/12 it was identified that an aerial lift was located in the Auxiliary Building stored in a seismic storage area near Train A and Train B safety related service water piping to Control Room Air Conditioning (CRAC) Alternate Cooling System. This resulted in both trains of Service Water being INOPERABLE per TS 3.7.8 and both trains of CRAC Alternate Cooling system per TS 3.7.11. At this time, there is no conclusive information that would support the OPERABILITY of the Service Water System during a seismic event therefore this event is being conservatively reported under 50.72(b)(3)(ii)(B), 'The nuclear plant being in an unanalyzed condition that significantly degrades plant safety.' and 50.72(b)(3)(v)(A) and (D) 'Any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structure systems that are needed for: (A) shutdown the reactor and maintain it in a safe shutdown condition, (D) mitigate the consequences of an accident.' The aerial lift was removed and the plant is no longer in the condition noted above. The NRC Resident Inspector has been informed.
On March 19, 2012, EN #47755 provided notification that both service water trains were potentially inoperable based on the potential for an improperly stored aerial lift to tip onto the adjacent service water piping during a seismic event. The notification was conservatively made due to the lack of conclusive information regarding any potential interaction between the aerial lift and the service water piping during a seismic event. Subsequent investigation and analysis determined that physical characteristics of the aerial lift are such that it would not have adversely interacted with the service water piping. Therefore, both service water trains remained operable and this condition did not meet the reportability criteria identified in 10CFR50.72. As a result, the notification made on 3/19/2012 (EN #47755) is hereby retracted. The NRC Resident Inspector has been notified. Notified R3DO (Stone). | Service water | |
| ENS 44938 | 26 March 2009 17:26:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Non-Functional Steam Exclusion Barrier | This event was additionally reported under 10 CFR 50.72(b)(3)(v)(D), 'Mitigate the Consequences of an Accident.' On 03/26/2009 at 1226 CDT, an engineering analysis determined that a steam exclusion door would not provide adequate steam exclusion protection. A metal plate, that had been installed over the door's glass window, would fail under the steam overpressure postulated during a high energy line break. The metal plate was replaced with one of a stronger design. Permanent repairs were completed on the door at 1549 on 03/26/2009. While the original plate was installed, the barrier was non-functional. In accordance with TRM 3.0.9, Section A.1 all equipment supported by that steam exclusion barrier was immediately declared inoperable. This zone includes both trains of ECCS and support equipment (i.e., SI, RHR, ICS, CCW, etc.). TS 3.0.c was entered and exited during the time the door issue was identified and repaired. All 3 trains of Auxiliary Feedwater were declared inoperable due to AFW low suction pressure trip channels were declared inoperable per TS 3.4.b.5. Power was reduced to less than 1673 MWt per TS 3.4.b.3. All mode changes were suspended per TS 3.4.b.2. Therefore, this is reportable under 10 CFR 50.72 (b)(3)(v), 'Any event or condition that at the time of discovery could have prevented the fulfillment of a safety function,' and under 10 CFR 50.72(b)(3)(ii)(B) 'any event or condition that results in the nuclear plant being in an unanalyzed condition that significantly degrades plant safety.' Containment Fan Coil Unit A was not operational during this event. The licensee notified the NRC Resident Inspector. * * * RETRACTION FROM JACK GADZALA TO PETE SNYDER ON 9/9/09 AT 1718 * * * On March 26, 2009 EN # 44938 provided notification that both trains of ESF equipment (e.g. SI, RHR, ICS, CCW, etc.) were inoperable based on an engineering analysis, which determined that a steam exclusion door in the auxiliary building would not provide adequate steam exclusion protection. Subsequent physical pressure testing of equivalent doors and engineering evaluation of the test results determined that the door remained capable of fulfilling its steam exclusion function. A series of pressure tests demonstrated that the door (including a metal plate that had been installed over the door's glass window) would withstand the steam overpressure postulated during a high energy line break. Therefore, the door remained functional and the supported ESF equipment in the auxiliary building remain operable. Consequently, this condition did not meet the reportability criteria in 10 CFR 50.72. As a result, the notification made on 3/26/09 is hereby retracted. This condition was also reported in Licensee Event Report (LER) 2009-005-00 on May 21, 2009." The licensee is planning on retracting this LER. | Auxiliary Feedwater | |
| ENS 44929 | 23 March 2009 12:31:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Non-Functional Steam Exclusion Barrier | On 3/23/09 at 0731 an I&C Supervisor transiting though a steam exclusion double door found the door sweep misaligned causing one side to catch and stick open. This prevented the door from closing automatically until the sweep was readjusted back to its original position. The door was then closed and monitored while permanent repairs took place. While the door was open and could not close automatically, the barrier was Non-Functional for Steam Exclusion. In accordance with TRM 3.0.9 Section A.1 all equipment supported by that steam exclusion barrier was immediately declared inoperable. This zone includes both trains of ECCS and support equipment (i.e., SI, RFIR, ICS, CCW, etc ). TS 3.0.c was entered and exited during the time the door could have stuck open (6 minutes) with both trains of ECCS inoperable. Permanent repairs were completed on the door at 0809 on 3/23/09. Therefore, this is reportable under 10 CFR 50.72 (b)(3)(v), 'Any event or condition that at the time of discovery could have prevented the fulfillment of a safety function, and under 10 CFR 50.72(b)(3)(ii)(B) 'any event or condition that results in the nuclear plant being in an unanalyzed condition that significantly degrades plant safety.' The licensee is also reporting this event under 10 CFR 50.72 (b)(3)(v)(D), Accident Mitigation. The licensee notified the NRC Resident Inspector.
05/18/2009 - Retraction of EN # 44929. Non�Functional Steam Exclusion Barrier. EN # 44929 provided notification that both trains of ESF equipment (e.g. SI, RHR, CCW, etc ) were inoperable due to degradation of a steam exclusion boundary door in the auxiliary building on March 23, 2009. Subsequent engineering evaluation determined that the degraded door remained capable of fulfilling its steam exclusion function during the brief (eight minute) period when it was degraded. The door had become degraded while being used for routine transit when the lower door sweep became misaligned, thereby preventing the door from completely closing. However, an engineering evaluation determined that the resulting gap was sufficiently small such that the total allowed leak path criteria were not exceeded. Additionally, the door swing was in the direction of postulated steam flow, such that the door would have been held in the closed direction by any steam overpressure postulated under accident conditions. Therefore, the door remained functional and the supported ESF equipment in the auxiliary building remained operable. Consequently, this condition did not meet the reportability criteria in 10 CFR 50.72. As a result, the notification made on 03/23/2009 is hereby retracted. The licensee notified the NRC Resident Inspector. Notified the R3DO(R. Skokowski). | ||
| ENS 44832 | 5 February 2009 21:39:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Steam Exclusion Door Gap | At 1539 on 02/05/2009 it was identified that a Steam Exclusion door seal was not flush with the door. The door seal gap was noted during a Fire Zone Inspection. It was determined that the door exceeded the allowed limit. This could have allowed steam from the Turbine Drive Auxiliary Feed Pump room into the emergency safeguards bus area. This could have resulted in both trains of ESF Equipment failing to perform their required functions. Upon discovery Specification 3.0.c was entered at 1539. Work was completed on the door to return it to functional status at 1601, and Technical Specification 3.0.c was exited at that time. The NRC Resident inspector has been notified.
Retraction of EN 44832, both trains of Engineered Safeguards Features (ESF) equipment inoperable due to a degraded steam exclusion boundary door. EN 44832 provided notification that both trains of ESF equipment were inoperable due to degradation of a steam exclusion boundary door in the turbine driven auxiliary feedwater pump room on February 5, 2009. Subsequent engineering evaluation determined that the degraded door remained capable of fulfilling its steam exclusion function during the period when it was degraded. A degraded door seal had resulted in a slight gap to exist between the door and the sill. This gap could have allowed steam from the turbine driven auxiliary feedwater pump room into the emergency safeguards bus area. However, the force that would be exerted on the door by steam overpressure postulated under accident conditions, would compress the door seal sufficiently to reduce the gap such that the total allowed leak path criteria would not be exceeded. Therefore, the door remained functional and the supported ESF equipment in the emergency safeguards bus area remained operable. Consequently, this condition did not meet the reportabllity criteria in 10 CFR 50.72. As a result, the notification made on 02/05/2009 (EN 44832) is hereby retracted. The licensee notified the NRC Resident Inspector. Notified R3DO (Peterson). | Auxiliary Feedwater | |
| ENS 44812 | 28 January 2009 18:24:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Steam Exclusion Door Inappropriately Held Open by a Door Chock | At 0834 on 01/28/2009 it was identified that a Steam Exclusion door was held open by a door chock installed on the door for less than 15 minutes. The door was open to ventilate the room during venting of carbon dioxide piping for routine maintenance. A Maintenance Mechanic was stationed at the door as per procedure. This door would have allowed steam into the emergency safeguards bus area from the Carbon Dioxide Tank room. This could have resulted in both Trains of ESF Equipment failing to perform their required functions. Upon discovery the door chock was disengaged to allow the door to self-close if required. The NRC Resident Inspector has been notified. | 05000305/LER-2009-002 | |
| ENS 44746 | 30 December 2008 17:13:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Non-Functional Steam Exclusion / Control Room Exclusion Zone (Crez) Barrier | On 12/30/08 at 1113 an Operator transiting though a steam exclusion / CREZ door found the door sweep separated from the door. This prevented the door from performing its Steam Exclusion and CREZ boundary functions as the opening, due to the missing sweep, exceeded the allowable opening size for the respective functions. Repairs were completed on the door at 1157 on 12/30/08. While the door sweep was not intact, the barrier was non-functional. In accordance with TRM 3.0.9 Section A.1 all equipment supported by that steam exclusion barrier was immediately declared inoperable. This zone includes both trains of Control Room Post Accident Recirculation and the Control Room itself. TS 3.0.c was entered when the sweep was identified degraded and exited following repairs. Therefore, this is reportable under 10CFR50.72(b)(3)(v), 'Any event or condition that at the time of discovery could have prevented the fulfillment of a safety function', and under 10CFR50.72(b)(3)(ii)(B), 'any event or condition that results in the nuclear plant being in an unanalyzed condition that significantly degrades plant safety'. The licensee notified the NRC Resident Inspector. | 05000305/LER-2008-003 | |
| ENS 44616 | 30 October 2008 14:30:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Non-Functional Steam Exclusion Barrier | On 10/30/08 at 0930 an RP Technician transiting though a steam exclusion door found a kickplate degraded. The kickplate is held on by two screws and one screw was missing. When the door was opened, the kickplate rotated and became lodged in the staircase grating. This prevented the door from closing until the technician physically lifted the plate out of the way to close the door. The door was open for less than a minute. This kickplate is not part of the door seal itself so when the door is closed it is functional. The kick plate was taped up as a temporary fix and access was restricted through the door until permanent repairs were complete. Permanent repairs were completed on the door at 1116 on 10/30/08. While the door was open and could not close automatically, the barrier was non-functional. In accordance with TRM 3.0.9 Section A.1 all equipment supported by that steam exclusion barrier was immediately declared inoperable. This zone includes both trains of ECCS and support equipment (i.e., SI, RHR, ICS, CCW, etc ). TS 3.0.c was entered and exited during the time the door was open with both trains of ECCS inoperable. Therefore, this is reportable under 10 CFR 50.72 (b)(3)(v), 'Any event or condition that at the time of discovery could have prevented the fulfillment of a safety function', and under 10 CFR 50.72(b)(3)(ii)(B) 'any event or condition that results in the nuclear plant being in an unanalyzed condition that significantly degrades plant safety'. The licensee notified the NRC Resident Inspector. | 05000305/LER-2008-002 | |
| ENS 43689 | 3 October 2007 20:30:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor | Potential Unavailability of Two Way Radio System to Support Safe Shutdown | In response to Self-Assessment Report SA014668 and preparation for the 2008 Triennial Fire Protection Inspection, a Fire Protection Improvement Plan implementation is in progress. Preliminary analysis of the major equipment, power supplies and cables associated with the two-way radio communication system (DCR-3341) indicates the potential for the plant two-way radio system to be adversely impacted and potentially unavailable to support post-fire safe shutdown operator actions and/or fire brigade fire fighting activities for a fire location in: - Fire Zone AX-35 (Control Room and AC Equipment Room) - Fire Zone TU-22 (Turbine Room) at the mezzanine elevation just outside Battery Room 1B - Fire Zone TU-98 (Battery Room 1B) This review has identified discrepancies regarding the credited means of communication required for use by Operators in response to an Appendix R fire. The safe shutdown procedures E-O-06 and E-O-07, and the Manual Action Feasibility Study (Fire Protection Engineering Evaluation FPEE-003) only credit the plant two-way radio system. However, upper tier program documents (e.g., Fire Protection Program Plan, Appendix R Design Description) do not consistently contain the same requirements. For example, the Fire Plan, Rev 7, Section 12.9 requires both the 5-channel Gai-Tronics system between key shutdown locations AND the multi-channel portable radio communications equipped with repeaters and provided for use by the plant fire brigade shall be operable at all times. The Appendix R Design Description. Rev. 5 is silent on safe shutdown communications. It is not clear at this time whether the 5-channel Gai-Tronics should also be credited in the safe shutdown procedures. If so, then the cables supporting operation of the Gai-Tronics would need to be identified and located by fire zone to determine their availability in lieu of two-way radio communications for a fire in any of the three fire zones identified above. Until this is verified, the Appendix R timeline for achieving sate shutdown may not be able to be met. Therefore, this is reportable under 10 CFR 50.72 (b)(3)(V)(A), 'Any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to: (A) shutdown the reactor and maintain it in a safe shutdown condition, (B) remove residual heat, (C) control the release of radioactive material, OR (D) mitigate the consequences of an accident. The licensee will notify the NRC Resident Inspector.
EN# 43689 provided notification that the Appendix R timeline for achieving safe shutdown may not be achievable due to potential loss of the credited two-way radio communication system due to fire and consequent need for face-to-face communications between the Operators. This position was adopted pending verification of the availability of the Gai-Tronics paging system. A subsequent engineering analysis of the major equipment, power supplies and cables associated with the Fire Protection/Appendix R two-way radio communication system, the Gai-Tronics plant paging system and the dedicated Emergency Gai-Tronics System, determined that adequate communications would have been available during a fire in the subject fire zones. Operators are familiar with and skilled in the use of the Gai-Tronics system as part of their job function. Interviews with on-shift Operators confirmed that operations would use the two-way radios; and if they failed, would then use the nearest Gai-Tronics handset station and then Emergency Gai-Tronics at specific locations for an Appendix R Dedicated Shutdown scenario. Where Operator manual actions are in close proximity to the Dedicated Shutdown panel, face-to-face communications would be achievable and timely. Consequently, the assumption of only face-to-face communications described in the Event Report would not have been necessary for all safe shutdown actions. Use of redundant communications systems (Gai-Tronics and Emergency Gai-Tronics) would have been available for fires in the subject fire zones such that the Appendix R safe shutdown time requirements would not have been significantly impacted. The loss of the two-way radio system for the identified fire zones would not have prevented the fulfillment of the safety function of systems that are needed to shut down the reactor and maintain it in a safe shutdown condition. As such, this condition is not reportable under 10CFR50.72(b)(3)(v)(A) as previously stated. Consequently, the notification made on 10/03/2007 (EN 43689) is hereby retracted. The licensee notified the NRC Resident Inspector. Notified R3DO (Lipa). | ||
| ENS 42589 | 18 May 2006 18:23:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor | Postulated Flooding That Could Impact Both Trains of the Residual Heat Removal (Rhr) System | At 1323 the Shift Manger was notified of a flooding concern that could impact both trains of the Residual Heat Removal (RHR) System. At this time both trains of RHR were declared inoperable and actions were taken to isolate the lines that could flood the RHR Pump Pits, The lines in question were from the Spent Fuel Pool Cleanup system specifically associated with the Spent Fuel Pool Dernineralizers and the Pre-filters and Post-filters. These lines were Isolated at 1345. Spent Fuel Pool cooling remained in service following isolation of the Spent Fuel Pool Cleanup system. Both RHR trains were returned to service at 1345. Initial piping analysis determined that these lines would have remained intact during and following a seismic event. The final evaluation s expected to be completed within two weeks. The NRC Resident Inspector was notified of this event by the licensee. | Residual Heat Removal | |
| ENS 41496 | 15 March 2005 22:30:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Kewaunee Plant Design for Flooding Events May Not Mitigate the Consequences of Piping System Failures | The following was provided by the licensee: While reviewing Nuclear Regulatory Commission's (NRC) memorandum regarding Task Interface Agreement (TIA), TIA 2001-02,'Design Basis Assumptions For Non-Seismic Piping Failures at Prairie Island Plant,' Kewaunee staff determined that the Kewaunee plant design for flooding events may not mitigate the consequences of piping system failures. As a minimum, and as a consequence of assuming failure of non-seismically qualified piping systems as prescribed in the TIA, water has been assumed to collect in the turbine building from a circulating water system piping failure that would result in substantial damage to Engineered Safeguards (ESF) and Safe Shutdown (SS) plant equipment, most notably electrical equipment. As a consequence of high water level in the turbine building, water could flow into the ESF equipment rooms that contain the Auxiliary Feedwater pumps, Emergency Diesel Generators and both the 480 volt and 4160 volt electrical switchgear. Water is assumed to flow into the equipment rooms by way of leakage past non-water-tight doors and the plant's unchecked floor-drain system. The expected water levels In the safeguards and electrical equipment rooms are assumed to increase to the point of causing multiple trains of both ESF and SS equipment to be unavailable to safely shutdown the plant. Kewaunee's primary mitigation strategy to combat flooding events is to recognize the event and initiate manual actions to open doors/ barriers. Opening the barriers to flooding directs the water out of the turbine building through the safeguards equipment rooms and returns it to the lake. Normally the manual actions would be expected to be performed before water level accumulates to a point of causing equipment damage. However, under the seismic failure assumptions, water levels are assumed to accumulate faster than the plant's ability to identify and react in order to assure protection of equipment required to initiate and complete a safe plant shutdown. Coincidental to the condition being reported, the plant had recently implemented additional precautionary measures to combat internal flooding events that lesson the significance of the condition being reported. Temporary pumping equipment, temporary sandbag barriers and additional personnel have been staged to minimize the consequences of previously questioned flooding events. Furthermore, a number of plant equipment design changes are being processed to further improve Kewaunee's defenses against internal flooding events. However, given the event being reported, the full scope of any additional actions is still to be determined. The NRC Resident Inspector was notified. | Emergency Diesel Generator Auxiliary Feedwater Circulating Water System | |
| ENS 41423 | 20 February 2005 01:10:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Technical Specifications Required Shutdown Due to Inoperable Auxiliary Feedwater Pumps | During continuing evaluation of the operability of the Auxiliary Feedwater (AFW) Pump discharge pressure switches, engineering determined that a high energy line break had the potential to affect the AFW Pump Suction line from the Condensate Storage Tank (CST) due to the inability of the discharge pressure switches to protect the AFW pumps from a loss of suction from the CST. At 1910 CST on 02/19/2005 it was determined that all three AFW Pumps were inoperable as a result of the condition discovered by engineering. Due to the high energy line break, there is the potential for damage to the CST supply line to the AFW Pumps (due to pipe whip resulting from a feedwater line circumferential break). Damage to the CST supply line may result in air entrainment in the AFW Pump supply and potential AFW pump damage following an automatic AFW Pump Start. Technical Specification 3.4.b.7 allows AFW Pumps to be placed in "pull-out" at less than 15% power because analysis shows that there is at least 10 minutes available for an operator to manually initiate AFW flow if needed. At 2003 CST a power reduction to <15% power was initiated to restore the operability of an AFW Pump. When power is less than 15%, the Turbine Driven AFW Pump will be placed in "pull-out" and Service Water will be aligned to the suction of the Turbine Driven AFW Pump to restore Operability of the Turbine Driven AFW Pump. When operability is restored to one AFW pump, the plant will enter a 4 hour LCO as a result of two AFW Pumps remaining out of service. Since operability of the two motor driven pumps will not be restored within the 4 hour LCO time, plant cooldown to less than 350 deg F will continue in accordance with Technical Specification 3.4.b.6. The licensee notified the NRC Resident Inspector. | Feedwater Service water Auxiliary Feedwater | |
| ENS 41406 | 12 February 2005 04:26:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Postulated Loss of Afw Pumps Following Tornado Damage to Condenstate Storage Tank | Plant engineering staff discovered that the auxiliary feedwater (AFW) system discharge pressure trip protective equipment may not operate to protect the AFW pumps from condensate storage tanks (CST) failing from a tornado event. The AFW system protection design uses discharge pressure switches to protect the AFW pumps from a loss of suction pressure. The AFW system is required equipment to support a plant shutdown which is also assumed to occur as a result of a tornado. Current analysis predicts substantial damage to the CSTs caused by a tornado. The CSTs are aligned as the AFW pumps' initial/preferred water source. The tank damage is predicted to result in a rapid, near complete, loss of water inventory due to a tornado. The CST volume loss quickly causes air to enter the pump suction piping to all three AFW pumps. Under the postulated conditions, the current discharge pressure switch design does not act fast enough to trip the pumps before significant pump damage can occur, it appears that the original design analyses failed to consider the consequences of a loss of suction source causing air to enter the pump suction. The design basis for AFW pump discharge pressure switches, in part, is assumed to protect the AFW pumps from damage due to a tornado. Given the event postulated as described, the pressure switches may not meet this design basis assumption. Therefore, all three AFW pump discharge pressure switches were declared inoperable at 2226 on 2/11/2005. To ensure the AFW Pumps are protected in the event of a tornado, compensatory actions were incorporated in Operating Procedure E-0-05 Response to Natural Events. At the earliest indication of a tornado threat (i.e. Tornado Watch, Warning or Strike) operators are directed to align Service Water to the AFW Pump Suction to ensure adequate NPSH In the event of a loss of the CST supply. These actions do not replace the automatic function but ensure the pumps are protected against loss of NPSH. Based on the proceduralized compensatory actions put in place, the AFW Pump Low discharge Pressure Trips were declared Operable but Non-Conforming at 0056 on 2/12/2005. The licensee has notified the NRC Resident Inspector. | Service water Auxiliary Feedwater |