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This telephone report is being made in accordance with 10 CFR 50.73(a)(2)(iv)(A) as an event or condition caused by an invalid signal that resulted in an automatic actuation on November 19, 2006 at 2028 CST of the following Primary Containment Isolation System (PCIS) groups:

• Group 2, Primary Containment isolations (partial) including Shutdown Cooling (SDC) valve isolation with no loss of SDC function.
• Group 6, Secondary Containment isolation, and initiation of Standby Gas Treatment System and Control Room Emergency Filtration System (complete).

The systems above functioned as designed. Reactor Protection System (RPS) bus 'B' had lost power from its Motor Generator (MG) and was found de-energized which caused the actuations. Power was restored to the RPS bus 'B' from auxiliary power, and plant systems were recovered. In addition to the actuations reported above, the following actuations also occurred and functioned as designed.

• RPS half-scram on B channel (no rods moved);
• Half-group isolations of the following PCIS Groups;

         *  Group 1,  Main Steam Isolation Valves (valves were already shut), 
         *  Group 3,  Reactor Water Clean-up (one valve shut),
         *  Group 7,  Recirculation Loop Sample Lines (valves were already shut),

The cause of the invalid signal was 'B' RPS MG set high voltage condition produced by a faulty voltage adjust potentiometer in the voltage regulator. This caused a protective relay to trip the RPS MG exciter field which caused the subsequent low voltage and loss of the 'B' RPS bus. The licensee notified the NRC Resident Inspector.

This notification is being made pursuant to NRC regulation 10 CFR 50.72(b)(3)(v)(D), 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 mitigate the consequences of an accident. At 1852 (CDT) on June 9, 2009, River Bend Station personnel discovered that a normally closed auxiliary building door was open. This door serves as part of the secondary containment boundary. At discovery, immediate action was taken to close the door. This action restored the secondary containment to the design configuration. Investigation determined that the door was last accessed at 1242 on June 9 and was most probably left open at that time. Further action is being taken to investigate the cause of the event. Secondary containment leak tightness is required to ensure that the release of radioactive materials from the primary containment is restricted to those leakage paths and associated leakage rates assumed in the accident analysis, and that fission products entrapped within the secondary containment structures will be treated by the Standby Gas Treatment System prior to discharge to the environment. With the subject door being open, the function of secondary containment would be impacted. A second door is located in the same exterior passage way as the secondary containment door found open. This door was closed during the period of time the secondary containment door was open. This second door serves a security function. However, it potentially could serve to perform the secondary containment function. An evaluation is being performed to determine the actual impact of the condition on the secondary containment function. However, based on the identified condition, this report is being made as a condition that could have prevented fulfillment of a safety function. The licensee has notified the NRC Resident Inspector.

• * * RETRACTION FROM D. WILLIAMSON TO D.PARK AT 1325 ON 6/29/09 * * *

(This event was reported by River Bend Station on 6/10/09 at 0102 (EDT). This update is being provided for the purpose of retracting that notification.) Subsequent investigation determined that the secondary containment door was left open for unknown reasons some time after 0800 on the morning of June 9. At the 1242 observation by the persons exiting the building, the door was open and it was left in that condition. However, a separate exterior door in that same passageway serves the security function, and it has been confirmed that, other than for routine access, the security door remained closed and locked during the time that the interior pressure boundary door was open. An engineering analysis has determined that the as-found condition did not defeat the function of secondary containment. While the security door is not air-tight, the maximum potential leakage past it under postulated accident conditions has been evaluated. An existing engineering calculation provides a means to determine the maximum size of a breach in the auxiliary building boundary such that the draw-down requirement prescribed by Technical Specifications is maintained. That calculation uses the additional flow area of an identified breach, in addition to the most recent test results of the standby gas treatment system (that system establishes and maintains a negative pressure in the building as part of the its design). Measurements taken on the door found that the potential flow area around it totaled 35 square inches. The current test results indicate that the standby gas treatment system can support the safety function of the auxiliary building with an analytical breach size of 230 square inches. As there is significant margin between the measured gap around the security door and the analytical value, this event was well bounded by the assumptions of the design basis of the building. As such, this event did not constitute a loss of the safety function of secondary containment. The licensee notified the NRC Resident Inspector. Notified the R4DO (D.Powers).

This event is reportable under 10 CFR 50.73(a)(2)(iv)(A) as an automatic actuation of general containment isolation signals affecting containment isolation valves in more than one system. This 60-day telephone notification is being made in accordance with 10 CFR 50.73(a)(1) for invalid actuations reported under 10 CFR 50.73(a)(2)(iv)(A). This actuation was invalid since it was caused by maintenance activities and not by any actual plant condition warranting containment isolation. On April 21, 2011, at 1310 hours, upon de-energization of the Reactor Protection System (RPS) Bus A for maintenance, an unexpected actuation of the Groups 3 and 4 outboard containment isolation valves occurred. Outboard isolations occurred in the reactor building drain and ventilation systems, and reactor closed cooling systems. Control room emergency ventilation and standby gas treatment systems also started. All systems functioned as designed, excluding those components that were already removed from service. Following the event, the RPS Bus A was re-energized and the plant was restored to normal operating condition for the current configuration per plant procedures.

The subsequent investigation found that personnel involved with work planning and control did not recognize that de-energizing RPS Bus A would create a condition which completed the Nuclear Steam Supply Shutoff System outboard isolation logic. An associated relay for the second channel had been previously de-energized to support ongoing maintenance. The cause of allowing this conflicting work to occur was determined to be inadequate procedural guidance. Planned corrective actions will enhance procedures to identify logic impacts, conflicting equipment and recommended protection schemes. There were no actual safety consequences associated with this event since all affected equipment responded as designed. The NRC Resident Inspectors have been notified.

This 60-day telephone notification is being made per the reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1) to describe an invalid actuation signal affecting more than one system. On September 19, 2011, at 1000 hours Central Daylight Time (CDT), during the performance of a maintenance activity on the Reactor/Refueling Zone Ventilation Radiation Monitor BFN-3-RM-090-0141/143, Browns Ferry Unit 3 received a Primary Containment Isolation System (PCIS) Group 6 isolation. This resulted in isolation of the reactor and refuel zone ventilation systems, initiation of the Standby Gas Treatment System (Trains A, B and C), and the initiation of the Control Room Emergency Ventilation System (Trains A and B). All plant systems responded as designed.

There are two divisions of Reactor/Refueling Zone Ventilation Radiation Monitors: BFN-3-RM-090-0140/142 and BFN-3-RM-090-0141/143. A downscale or inoperable signal in both divisions will initiate the PCIS Group 6 isolation. Prior to the performance of the maintenance activity, BFN-3-RM-090-0140/142 was functional. However, during a field walkdown after the PCIS Group 6 isolation, relay BFN-3-RLY-064-16AK62A for BFN-3-RM-090-0140/142 was discovered to be chattering. A chattering relay could cause momentary loss of continuity between the contacts which would effectively generate a spurious isolation signal from that division. Thus, when 3-RM-090-0141/143 was made inoperable by the surveillance, the PCIS Group 6 logic was made-up. The PCIS Group 6 isolation was reset at 1008 CDT. This event was entered in the Corrective Action Program as Problem Evaluation Report (PER) 434799. There were no safety consequences or impact on the health and safety of the public as a result of these events. The NRC Senior Resident Inspector was notified.

At 22:04 CDT on 9/02/2012, the Emergency Reserve Auxiliary Transformer (ERAT) transferred unexpectedly to the Reserve Auxiliary Transformer (RAT). During this transfer, the Fuel Pool Cooling and Cleanup (FC) system pump 'A' tripped and the Fuel Building Ventilation (VF) system isolated. Upper containment pool level dropped below the minimum required level per Technical Specifications (TS) 3.6.2.4 and Secondary Containment differential pressure increased above 0.25 inches vacuum per TS 3.6.4.1. Upper Containment Pool level was restored above the minimum level at 01:27 CDT on 9/3/2012 within the 4 hour completion time. The Upper Containment Pool is a part of the suppression pool makeup system used to ensure the Primary Containment function. Secondary Containment differential pressure was restored at 22:19 on 9/2/2012 when the Standby Gas Treatment System was started. Maintaining secondary containment differential pressure helps to control the release of radioactive material. This event is being reported as a condition that could have prevented the fulfillment of a safety function per 10 CFR 50.72(b)(3)(v)(B) and 10 CFR 50.72(b)(3)(v)(C). The station is currently in a 72-hour action to restore the ERAT to an operable status per TS LCO 3.8.1 Required Action A.2. Plant conditions are stable and actions are underway to repair the ERAT. The NRC Resident (Inspector) has been notified.

• * * RETRACTION ON 10/26/12 AT 1322 EDT FROM KEN LEFFEL TO DONG PARK * * *

Upper Containment Pool level dropped below the normal pool level of 827 feet-3 inches when the Fuel Pool Cooling and Cleanup system pump 'A' tripped, and was initially reported as dropping below the minimum level (825 feet-6 inches) required by Technical Specification (TS) 3.6.2.4. However, subsequent reports from the field confirmed that the lowest level reached was 827 feet 0 inches, which is greater than the minimum required TS level. Therefore, no loss of safety function occurred for the Upper Containment Pool level as a result of this event, and the event is not reportable under 50.72 (b)(3)(v)(B). The NRC Resident (Inspector) has been notified." Notified R3DO (Pelke).

LaSalle Unit 1 and LaSalle Unit 2 have both experienced an automatic reactor scram, in conjunction with a loss of offsite power. This was caused by an apparent lightning strike in the main 345kV/138kV switchyard during a thunderstorm. 138kV line 0112 has been inspected in the field, and heavy damage has been noted on the insulators on two of the three phases on a line lightning arrestor line side. The plant systems have all responded as expected. All five diesel generators started, and have loaded on to their respective buses as designed. All rods went full in on both units during the respective scrams. HPCS (High Pressure Core Spray) system was started on each unit and automatically aligned for injection for initial level control. The MSIVs (Main Steam Isolation Valves) are shut on both units with decay heat being removed via the safety relief valves. Suppression pool cooling is in progress. The licensee will notify the NRC Resident Inspector and has notified the State. Notified DHS, FEMA, USDA, HHS, DOE, NICC, EPA, and Nuclear SSA via email.

• * * UPDATE FROM DON PUCKETT TO VINCE KLCO AT 2113 EDT ON 4/17/2013 * * *

In addition to information (previously provided), LaSalle Unit 2 received a high drywell pressure signal (1.77 psig) due to loss of containment cooling from the loss of power. At the time of this high drywell pressure signal, high pressure core spray pump and 2B residual heat removal (RHR) pump was already in operation, the low pressure core spray system and 2A residual heat removal system was secured and (placed) in pull to lock. When the signal was satisfied the ECCS (Emergency Core Cooling Systems) signal was processed but only the 2C RHR pump would have started. In this case, the 2C RHR pump tripped when the signal was received. There is no evidence of reactor coolant leakage. There was no additional ECCS systems discharging into the RCS (Reactor Coolant System). As (initially stated), level was controlled using High Pressure Core Spray and level control is now being maintained using the Reactor Core Isolation Cooling (RCIC) systems. The 2C RHR pump trip is under investigation. Due to the initial loss of offsite power for both Unit 1 and Unit 2 reported at 1511 (CDT), multiple containment isolation valves isolated and closed as expected. Once initial containment isolations were verified, two Unit 2 primary containment vent and purge valves were opened to vent the Unit 2 containment. Once Unit Two containment pressure reached 1.77 (psig), these two vent valves isolated as expected. Due to the loss of offsite power, the Station Vent Stack Wide Range Gas Monitor (WRGM) and the Standby Gas Treatment Wide Range Gas Monitor (VGWRGM) also lost power. Manual sampling has been implemented and power is restored to the VGWRGM, however the VGWRGM has not been declared operable yet. Normal radiation levels have been reported from the manual sampling. (This is being reported in accordance with 10CFR50.72(b)(3)(xiii).) The licensee notified the NRC Resident Inspector and the State of Illinois. Notified the R3 IRC, NRR EO(Skeen), IRD MOC (Grant).

• * * UPDATE AT 0057 EDT ON 04/18/13 FROM MIKE LAWRENCE TO S. SANDIN * * *

After the Unit 2 primary containment vent and purge system isolated on the Unit 2 containment High Pressure signal, Venting of the Unit 1 primary containment was commenced. At 2005 CDT, Unit 1 primary containment pressure reached the Group 2 primary containment isolation system setpoint (1.77 PSIG) causing the primary containment vent and purge valves being used to vent the Unit 1 containment to isolate. Unit 1 primary containment venting was being performed through the Standby Gas Treatment system which is a filtered system. In addition to the primary containment isolation signal on high drywell pressure, an ECCS initiation on high drywell pressure also occurred. The ECCS signal resulted in an auto start of the 1C RHR system. The 1B RHR system was already running in suppression pool cooling mode. 1A RHR and LPCS had been secured to prevent overloading the common diesel generator for division 1. The common diesel generator supplies both Unit 1 and Unit 2 division 1 ESF busses. The licensee informed the NRC Resident Inspector. Notified NRR EO (Skeen), IRD MOC (Grant) and R3IRC (Louden).

• * * UPDATE AT 0947 EDT ON 04/18/13 FROM JUSTIN FREEMAN TO PETE SNYDER * * *

LaSalle has terminated the unusual event which was initiated at 1511 on 4/17/13 and reported under EN 48939. This unusual event has been terminated based on meeting the following established criteria. This report is being made in accordance with 10CFR50.72.(c)(1)(iii). 1) Off-site power has been restored to all ESF busses 2) Fuel Pool Cooling has been restored on both units 3) Primary Containment Chillers have been restored on both units 4) Drywell pressure is less than ECCS initiation setpoint 5) ECCS signals cleared to allow diesels to be placed in stand by Recovery of remaining plant systems will be managed through the Outage Control Center (OCC)." The licensee informed the NRC Resident Inspector. Notified R3DO (Orth), NRR EO (Chernoff), IRD (Grant), DHS, FEMA, USDA, HHS, DOE, NICC, EPA, and Nuclear SSA via email.

• * * UPDATE AT 1711 EDT ON 4/21/2013 FROM GREG LECHTENBERG TO MARK ABRAMOVITZ * * *

In addition to the 10 CFR 50.72 Sections initially identified, the Loss of Offsite Power was also reportable under 10 CFR 50.72(b)(3)(v)(D) as an event or condition that could have prevented the fulfillment of the safety function of systems needed mitigate the consequences of an accident. This event is considered a safety system functional failure for both Units 1 and 2. The licensee will notify the NRC Resident Inspector. Notified the R3DO (Orth).