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Event Notification Report for December 17, 1999

                    U.S. Nuclear Regulatory Commission
                              Operations Center

                              Event Reports For
                           12/16/1999 - 12/17/1999

                              ** EVENT NUMBERS **

36515  36516  36517  36518  36519  36520  36521  

+------------------------------------------------------------------------------+
|Power Reactor                                    |Event Number:   36515       |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: TURKEY POINT             REGION:  2  |NOTIFICATION DATE: 12/16/1999|
|    UNIT:  [3] [] []                 STATE:  FL |NOTIFICATION TIME: 01:01[EST]|
|   RXTYPE: [3] W-3-LP,[4] W-3-LP                |EVENT DATE:        12/15/1999|
+------------------------------------------------+EVENT TIME:        23:25[EST]|
| NRC NOTIFIED BY:  WILSON                       |LAST UPDATE DATE:  12/16/1999|
|  HQ OPS OFFICER:  LEIGH TROCINE                +-----------------------------+
+------------------------------------------------+PERSON          ORGANIZATION |
|EMERGENCY CLASS:          N/A                   |ANN BOLAND           R2      |
|10 CFR SECTION:                                 |                             |
|APRE 50.72(b)(2)(vi)     OFFSITE NOTIFICATION   |                             |
|                                                |                             |
|                                                |                             |
|                                                |                             |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR|   INIT RX MODE  |CURR PWR|  CURR RX MODE   |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|3     N          Y       100      Power Operation  |100      Power Operation  |
|                                                   |                          |
|                                                   |                          |
+------------------------------------------------------------------------------+
                                   EVENT TEXT                                   
+------------------------------------------------------------------------------+
| OFFSITE NOTIFICATION REGARDING A BUNKER-C OIL SPILL AT THE FOSSIL UNITS      |
|                                                                              |
| The following text is a portion of a facsimile received from the licensee:   |
|                                                                              |
| "Units 1 and 2 (fossil) are notifying [the Department of Environmental       |
| Resource Management (DERM)] of [a] Bunker-C oil spill on [a] permeable       |
| surface.  [It is] estimated that 100 - 150 gallons of bunker C [are]         |
| contained by an earth/rock berm around the fuel tank."                       |
|                                                                              |
| The licensee notified the NRC resident inspector.                            |
+------------------------------------------------------------------------------+

+------------------------------------------------------------------------------+
|Power Reactor                                    |Event Number:   36516       |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: BEAVER VALLEY            REGION:  1  |NOTIFICATION DATE: 12/16/1999|
|    UNIT:  [] [2] []                 STATE:  PA |NOTIFICATION TIME: 14:19[EST]|
|   RXTYPE: [1] W-3-LP,[2] W-3-LP                |EVENT DATE:        12/16/1999|
+------------------------------------------------+EVENT TIME:        13:45[EST]|
| NRC NOTIFIED BY:  GEORGE E. STOROLIS           |LAST UPDATE DATE:  12/16/1999|
|  HQ OPS OFFICER:  STEVE SANDIN                 +-----------------------------+
+------------------------------------------------+PERSON          ORGANIZATION |
|EMERGENCY CLASS:          N/A                   |BILL RULAND          R1      |
|10 CFR SECTION:                                 |                             |
|AOUT 50.72(b)(1)(ii)(B)  OUTSIDE DESIGN BASIS   |                             |
|NLCO                     TECH SPEC LCO A/S      |                             |
|                                                |                             |
|                                                |                             |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR|   INIT RX MODE  |CURR PWR|  CURR RX MODE   |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|                                                   |                          |
|2     N          Y       100      Power Operation  |100      Power Operation  |
|                                                   |                          |
+------------------------------------------------------------------------------+
                                   EVENT TEXT                                   
+------------------------------------------------------------------------------+
| UNIT 2 SERVICE WATER PUMP 'C' MAY NOT HAVE BEEN CAPABLE OF PERFORMING DESIGN |
| FUNCTION DURING A 7-DAY OUTAGE OF SERVICE WATER PUMP 'B.'                    |
|                                                                              |
| "Routine operations plant surveillance revealed, at approximately 0056 hours |
| on 11/21/99, that the discharge metal expansion joint of Beaver Valley Power |
| Station (BVPS) Unit 2 Service Water System (SWS) pump 2SWS*P21C was deformed |
| outward.  2SWS*P21C was supplying the SWS 'B' Train, since approximately     |
| 0425 hours on 11/15/99, while the train normal supply pump, 2SWS*P21B, was   |
| out of service for replacement of the pump vacuum break check valve.  Due to |
| the unknown condition of the expansion joint, 2SWS*P21C was declared         |
| inoperable, and [the] required action of Technical Specifications (TS)       |
| 3.7.4.1 was entered.  At approximately 0154 hours (same day), Standby        |
| Service Water System (SWE) Pump 2SWE*P21B was placed into service to supply  |
| [the] 'B' SWS Train, and 2SWS*P21C was removed from service.  2SWS*P21B was  |
| returned to service to supply [the] 'B' SWS Train, and the required action   |
| of TS was exited at approximately 0035 hours on 11/22/99.                    |
|                                                                              |
| "Investigation has determined that the subject expansion joint most likely   |
| deformed, at approximately 1149 hours on 11/09/99 due to water hammer of the |
| 'B' SWS Train piping during safeguards protection system slave relay         |
| testing.  Subsequent engineering assessment of the as-found condition of the |
| expansion joint is indeterminate whether the capability of the SWS system to |
| perform its intended function under all design basis events was              |
| significantly affected.  Therefore, during the time frame that 2SWS*P21C was |
| relied upon to maintain operability of the SWS (11/15/99 until 2SWS*P21B was |
| placed into service to supply 'B' SWS Train), it cannot be assured that Unit |
| 2 would have remained within the plant design basis with an additional       |
| postulated single failure of redundant SWS Train 'A' that would have         |
| prevented the train from mitigating a design basis accident.  Throughout the |
| time 2SWS*P21C was relied upon for SWS Train 'B,' the Unit SWE remained      |
| fully operable.  This notification is applicable to the 1-hour non-emergency |
| event reporting criteria of 10 CFR 50.72(b)(ii)(B), for having been 'in a    |
| condition that is outside the design basis of the plant'."                   |
|                                                                              |
| The water hammer which occurred on 11/09/99 was attributed to a failed       |
| vacuum break check valve which had corroded closed.  The SWS Pump 'C'        |
| bellows has been replaced, and required maintenance will be completed by     |
| 12/24/99.  The pump will be declared operable following post-maintenance     |
| testing.                                                                     |
|                                                                              |
| The licensee informed the NRC Resident Inspector.                            |
+------------------------------------------------------------------------------+

+------------------------------------------------------------------------------+
|Power Reactor                                    |Event Number:   36517       |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: BRUNSWICK                REGION:  2  |NOTIFICATION DATE: 12/16/1999|
|    UNIT:  [1] [2] []                STATE:  NC |NOTIFICATION TIME: 16:40[EST]|
|   RXTYPE: [1] GE-4,[2] GE-4                    |EVENT DATE:        12/16/1999|
+------------------------------------------------+EVENT TIME:        12:48[EST]|
| NRC NOTIFIED BY:  KEN CHISM                    |LAST UPDATE DATE:  12/16/1999|
|  HQ OPS OFFICER:  STEVE SANDIN                 +-----------------------------+
+------------------------------------------------+PERSON          ORGANIZATION |
|EMERGENCY CLASS:          N/A                   |ANN BOLAND           R2      |
|10 CFR SECTION:                                 |                             |
|AIND 50.72(b)(2)(iii)(D) ACCIDENT MITIGATION    |                             |
|NLCO                     TECH SPEC LCO A/S      |                             |
|                                                |                             |
|                                                |                             |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR|   INIT RX MODE  |CURR PWR|  CURR RX MODE   |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1     N          Y       100      Power Operation  |100      Power Operation  |
|2     N          Y       100      Power Operation  |100      Power Operation  |
|                                                   |                          |
+------------------------------------------------------------------------------+
                                   EVENT TEXT                                   
+------------------------------------------------------------------------------+
| REACTOR CORE ISOLATION COOLING SYSTEM DECLARED INOPERABLE ON BOTH UNITS DUE  |
| TO UNDERSIZED THERMAL OVERLOADS INSTALLED ON THREE (3) VALVES WHICH MAY HAVE |
| PREVENTED OPERATION UNDER WORST CASE CONDITIONS.                             |
|                                                                              |
| "On December 16, 1999, at 1248, the Reactor Core Isolation Cooling System    |
| was declared inoperable because the thermal overloads on three system valves |
| were determined to be sized such that the affected valves might not operate  |
| during worst case conditions.  The affected valves are:  1 (2)-E51-V8        |
| (Turbine Trip and Throttle Valve), 1(2)-E51-F019 (Minimum Flow Bypass to     |
| Torus Valve), and 1(2)-E51-F046 (Cooling Water Supply Valve).  Analysis also |
| determined that 1(2)-E41-F059 (High Pressure Core Injection Cooling System   |
| Water Supply Valve) also contains inappropriately sized thermal overloads;   |
| however, the [High Pressure Coolant Injection] System has not been declared  |
| inoperable because this valve have been repositioned to its accident         |
| position (open), and administrative measures have been taken to maintain the |
| valve in the open position.                                                  |
|                                                                              |
| "The Reactor Core Isolation Cooling system is a single-train system used to  |
| prevent overheating of the reactor fuel in the event of a reactor isolation  |
| accompanied by a loss of feedwater.  The high pressure High Pressure Coolant |
| Injection system (approximately ten times the flow rate as the Reactor Core  |
| Isolation Cooling System) remains operable.  Plant technical specifications  |
| allow continued operation for 14 days with the Reactor Core Isolation        |
| Cooling system inoperable.  The Reactor Core Isolation Cooling system is not |
| considered an engineered safety feature at the Brunswick Plant although it   |
| is included in plant technical specifications.  For these reasons, the       |
| safety significance of this event is considered to be low.  Engineering      |
| calculations are currently in progress to confirm the operability of the     |
| [High Pressure Coolant Injection] System.                                    |
|                                                                              |
| "Engineering and maintenance personnel are working to determine a corrective |
| action plan at this time."                                                   |
|                                                                              |
| The licensee informed the NRC Resident Inspector.                            |
+------------------------------------------------------------------------------+

+------------------------------------------------------------------------------+
|Power Reactor                                    |Event Number:   36518       |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: PERRY                    REGION:  3  |NOTIFICATION DATE: 12/16/1999|
|    UNIT:  [1] [] []                 STATE:  OH |NOTIFICATION TIME: 19:03[EST]|
|   RXTYPE: [1] GE-6                             |EVENT DATE:        12/16/1999|
+------------------------------------------------+EVENT TIME:             [EST]|
| NRC NOTIFIED BY:  ALAN RABENOLD                |LAST UPDATE DATE:  12/16/1999|
|  HQ OPS OFFICER:  STEVE SANDIN                 +-----------------------------+
+------------------------------------------------+PERSON          ORGANIZATION |
|EMERGENCY CLASS:          N/A                   |RONALD GARDNER       R3      |
|10 CFR SECTION:                                 |                             |
|NLTR                     LICENSEE 24 HR REPORT  |                             |
|                                                |                             |
|                                                |                             |
|                                                |                             |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR|   INIT RX MODE  |CURR PWR|  CURR RX MODE   |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1     N          Y       100      Power Operation  |98       Power Operation  |
|                                                   |                          |
|                                                   |                          |
+------------------------------------------------------------------------------+
                                   EVENT TEXT                                   
+------------------------------------------------------------------------------+
| 24-HOUR REPORT INVOLVING POSSIBLE VIOLATION OF CONDITION OF LICENSEE         |
|                                                                              |
| "Preliminary investigations indicate that during the preceding operating     |
| cycle, the Perry Plant potentially exceeded its rated power level.   A GE    |
| software change disabled feedwater temperature compensation (FWC).  Perry    |
| plant uses the 3D MONICORE output in subsequent calculations for power       |
| indication.  Perry plant reduced power to 98% as a precaution, while Reactor |
| Engineering confirms the power calculation error.  A corrective action       |
| response team has been assembled, and the scope of the software change is    |
| being reviewed for any other possible effects on the plant.                  |
|                                                                              |
| "In February of 1999, during plant coastdown to RFO7, and during subsequent  |
| removal of the '6B' feedwater heater, a condition was created in which the   |
| plant was operated at 100% power while at a reduced feedwater temperature.   |
| Preliminary investigations indicate that the plant could have violated its   |
| Operating License during this 2-day period of time.                          |
|                                                                              |
| "This condition is being reported as a possible violation of the Operating   |
| License, and accordingly, it is required to be reported (24 hour, and        |
| Licensee Event Report) under the Perry Operating License section 2.F."       |
|                                                                              |
| The licensee plans to inform the NRC Resident Inspector.                     |
|                                                                              |
| * * * UPDATE AT 2040 EST ON 12/16/99 BY S. SANDIN * * *                      |
|                                                                              |
| "The 3D MONICORE did not have a problem.  The problem was the 'databank.'    |
| This databank has cycle-specific data to support 3D MONICORE calculations.   |
|                                                                              |
| "The revision of 3D MONICORE was baseline 94 installed in 1996.  There were  |
| NO changes to the MONICORE programs.                                         |
|                                                                              |
| "The databank was the BOC7 databank installed in October 1997."              |
|                                                                              |
| A value in the databank was set to zero following instructions from the      |
| vendor.  The licensee did not recognize, and the vendor was unaware, that    |
| this data was utilized by another program.  This resulted in incorrect       |
| temperature compensation during coastdown between 02/07 through 02/09/99     |
| when the "6B" feedwater heater was removed from service.                     |
+------------------------------------------------------------------------------+

+------------------------------------------------------------------------------+
|Power Reactor                                    |Event Number:   36519       |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: SAINT LUCIE              REGION:  2  |NOTIFICATION DATE: 12/16/1999|
|    UNIT:  [1] [] []                 STATE:  FL |NOTIFICATION TIME: 20:20[EST]|
|   RXTYPE: [1] CE,[2] CE                        |EVENT DATE:        12/16/1999|
+------------------------------------------------+EVENT TIME:        20:03[EST]|
| NRC NOTIFIED BY:  DAVE FIELDS                  |LAST UPDATE DATE:  12/16/1999|
|  HQ OPS OFFICER:  STEVE SANDIN                 +-----------------------------+
+------------------------------------------------+PERSON          ORGANIZATION |
|EMERGENCY CLASS:          N/A                   |ANN BOLAND           R2      |
|10 CFR SECTION:                                 |                             |
|AOUT 50.72(b)(1)(ii)(B)  OUTSIDE DESIGN BASIS   |                             |
|                                                |                             |
|                                                |                             |
|                                                |                             |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR|   INIT RX MODE  |CURR PWR|  CURR RX MODE   |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1     N          Y       100      Power Operation  |100      Power Operation  |
|                                                   |                          |
|                                                   |                          |
+------------------------------------------------------------------------------+
                                   EVENT TEXT                                   
+------------------------------------------------------------------------------+
| UNIT 1 CABLE SEPARATION INSIDE CONTAINMENT IS NOT AS DESCRIBED IN THE NRC    |
| SAFETY EVALUATION REPORT (SER) SUBMITTED FOR AN EXEMPTION GRANTED IN 1985.   |
|                                                                              |
| "Description of Condition:                                                   |
|                                                                              |
| "FPL has identified inconsistencies between FPL's original exemption request |
| for cable separation inside containment (K1) and the NRC SER for Unit 1      |
| Appendix R.  The 1985 NRC SER granted approval stating cables were routed on |
| separate elevations, approximately 25 feet apart.  However, the NRC 1987 SER |
| granted approval stating cables were separated vertically by at least 25     |
| feet.  FPL's original submittal never stated that the cables were vertically |
| separated by 25 feet but that the design was that redundant cables were      |
| routed at separate containment elevations.  Although the 1987 NRC SER        |
| changes appear to be editorially based, it resulted in FPL not being in      |
| strict compliance with the latest NRC SER.                                   |
|                                                                              |
| "FPL has concluded that its original bases for the Unit 1 inside containment |
| K1 Appendix R exemption remain valid The probability, magnitude, and         |
| consequences of a fire in these areas inside containment is extremely low    |
| for the following reasons:                                                   |
|                                                                              |
| 1.  The containment has a large volume with a high ceiling, which would      |
| dissipate the hot gases from a fire to the upper area of containment away    |
| from the affected area.                                                      |
|                                                                              |
| 2.  The cables in the effected cable tray are self-extinguishing which will  |
| minimize fire propagation along the cables.                                  |
|                                                                              |
| 3.  The bottom cable trays are designed with solid bottoms, which            |
| effectively act as a radiant energy shield.                                  |
|                                                                              |
| 4.  The main source of combustion inside containment is the oil in the RCP   |
| motors.  However, the RCP motors are not in the vicinity of the affected     |
| cable trays.  The affected cable trays are separated from the RCP motors by  |
| the biological shield.                                                       |
|                                                                              |
| 5.  The combustible loading in this portion of containment is low and in the |
| area where the lack of  separation occurs consists of mostly cable           |
| insulation, which has a high ignition temperature.                           |
|                                                                              |
| 6.  There are no fire hazards or ignition sources in the area.               |
|                                                                              |
| 7.  All non-IEEE 383 cables are covered with a fire retardant coating.       |
|                                                                              |
| 8.  This area has fire detectors, which would provide prompt notification of |
| a fire to the control room.                                                  |
|                                                                              |
| 9.  The containment is inspected prior to operation for items that could     |
| impact sump operability; therefore, the potential for transient combustibles |
| is precluded.  In addition, the containment is a radiation control area with |
| very limited access during power operation.  The possibility of introducing  |
| new transient combustibles is very small.                                    |
|                                                                              |
| "FPL concludes that non-compliance with the NRC approved Unit 1 Appendix R   |
| exemption K1 does not constitute an operability concern.  The original bases |
| for FPL's exemption request provides reasonable assurance that a fire inside |
| containment would not prevent the loss of any safe shutdown function.        |
|                                                                              |
| "Immediate Corrective Actions:                                               |
|                                                                              |
| 1.  A containment loose debris inspection is performed following each        |
| containment entry during Modes 1-4.                                          |
|                                                                              |
| 2.  If containment entry is being provided for maintenance purposes, an      |
| inventory of all material, tools, and other items intended to go into        |
| containment is performed.                                                    |
|                                                                              |
| 3.  All cable trays on the 18 foot and 23 foot elevations were verified to   |
| be free of debris and other fire hazards prior to heatup from the previous   |
| refueling outage.                                                            |
|                                                                              |
| "The NRC resident inspector has been informed of this notification by the    |
| licensee."                                                                   |
+------------------------------------------------------------------------------+

+------------------------------------------------------------------------------+
|Power Reactor                                    |Event Number:   36520       |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: CRYSTAL RIVER            REGION:  2  |NOTIFICATION DATE: 12/16/1999|
|    UNIT:  [3] [] []                 STATE:  FL |NOTIFICATION TIME: 23:05[EST]|
|   RXTYPE: [3] B&W-L-LP                         |EVENT DATE:        12/16/1999|
+------------------------------------------------+EVENT TIME:        20:47[EST]|
| NRC NOTIFIED BY:  RICKY RAWLS                  |LAST UPDATE DATE:  12/16/1999|
|  HQ OPS OFFICER:  STEVE SANDIN                 +-----------------------------+
+------------------------------------------------+PERSON          ORGANIZATION |
|EMERGENCY CLASS:          N/A                   |ANN BOLAND           R2      |
|10 CFR SECTION:                                 |                             |
|APRE 50.72(b)(2)(vi)     OFFSITE NOTIFICATION   |                             |
|                                                |                             |
|                                                |                             |
|                                                |                             |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR|   INIT RX MODE  |CURR PWR|  CURR RX MODE   |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|3     N          Y       100      Power Operation  |100      Power Operation  |
|                                                   |                          |
|                                                   |                          |
+------------------------------------------------------------------------------+
                                   EVENT TEXT                                   
+------------------------------------------------------------------------------+
| STATE OF FLORIDA NOTIFICATION INVOLVING RESCUED LOGGER HEAD SEA TURTLE       |
|                                                                              |
| "[An] endangered Logger Head Sea Turtle was rescued at the Crystal River     |
| (CR-3) intake area.  The turtle was injured but is expected to live.  The    |
| Turtle was transported to Florida Power Corp. Mariculture Center.  The       |
| Florida Department of Environmental Protection was notified at 2047 EST on   |
| 12/26/99."                                                                   |
|                                                                              |
| The licensee also informed the NRC Resident Inspector.                       |
+------------------------------------------------------------------------------+

+------------------------------------------------------------------------------+
|Power Reactor                                    |Event Number:   36521       |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: COOK                     REGION:  3  |NOTIFICATION DATE: 12/17/1999|
|    UNIT:  [1] [] []                 STATE:  MI |NOTIFICATION TIME: 00:30[EST]|
|   RXTYPE: [1] W-4-LP,[2] W-4-LP                |EVENT DATE:        12/16/1999|
+------------------------------------------------+EVENT TIME:        20:52[EST]|
| NRC NOTIFIED BY:  SCOTT RICHARDSON             |LAST UPDATE DATE:  12/17/1999|
|  HQ OPS OFFICER:  LEIGH TROCINE                +-----------------------------+
+------------------------------------------------+PERSON          ORGANIZATION |
|EMERGENCY CLASS:          N/A                   |RONALD GARDNER       R3      |
|10 CFR SECTION:                                 |ROBERT GALLO         NRR     |
|AESF 50.72(b)(2)(ii)     ESF ACTUATION          |FRANK CONGEL         IRO     |
|                                                |                             |
|                                                |                             |
|                                                |                             |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR|   INIT RX MODE  |CURR PWR|  CURR RX MODE   |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1     N          N       0        Refueling        |0        Refueling        |
|                                                   |                          |
|                                                   |                          |
+------------------------------------------------------------------------------+
                                   EVENT TEXT                                   
+------------------------------------------------------------------------------+
| AUTOMATIC STARTING AND LOADING OF AN EMERGENCY DIESEL GENERATOR FOLLOWING A  |
| LOSS OF OFFSITE POWER TO THE TRAIN 'A' RESERVE AUXILIARY TRANSFORMER         |
| RESULTING IN A LOSS OF SPENT FUEL POOL COOLING FOR APPROXIMATELY 38 MINUTES  |
|                                                                              |
| The following text is a portion of a facsimile received from the licensee:   |
|                                                                              |
| "[At 2052 on 12/16/99, the] Unit 1 Train 'A' [emergency diesel generator     |
| (CD EDG)] automatically started and loaded following [a] loss of offsite     |
| power to the 4-kV safeguards buses on Train 'A.'  The apparent cause was the |
| trip of the Unit 2 Train 'A' reserve feed transformer (201CD) sudden         |
| overpressurization  relay.  Trip of the 201CD sudden overpressurization      |
| relay resulted in opening of breaker 12CD which feeds the Unit 1 reserve     |
| feed transformer (101CD) as well as 201CD."                                  |
|                                                                              |
| "[Both] units were defueled at the time of the event with the Unit 1 [spent  |
| fuel pit (SFP)] pump in service.  The loss of offsite power in Unit 1 caused |
| the in-service SFP pump to be de-energized.  The Unit 2 SFP pump was         |
| [manually] started at 2130 hours.  SFP temperature remained constant at 85F |
| throughout this event."                                                      |
|                                                                              |
| "The 201CD sudden overpressure relay was defeated, and offsite power was     |
| restored to the 101CD transformer.  The CD EDG will be unloaded once offsite |
| power is made available."                                                    |
|                                                                              |
| "Actuation of the 201CD reserve transformer sudden overpressure relay was a  |
| result of filling the transformer with dry nitrogen following [planned]      |
| corrective maintenance activities."                                          |
|                                                                              |
| The licensee stated that both units were defueled with fuel in the common    |
| SFP.  At the time of this event, offsite power to Unit 2 was being supplied  |
| by the normal auxiliary transformer (backfeed).  All systems functioned as   |
| required and there was nothing unusual or not understood.  At the time of    |
| this event notification, the licensee was in the process of removing the CD  |
| EDG from service and restoring the Train 'A' reserve feed transformer as the |
| offsite power supply for Unit 1.                                             |
|                                                                              |
| The licensee plans to notify the NRC resident inspector.                     |
+------------------------------------------------------------------------------+