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 85�F |
| 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. |
+------------------------------------------------------------------------------+
Page Last Reviewed/Updated Wednesday, March 24, 2021
Page Last Reviewed/Updated Wednesday, March 24, 2021