Event Notification Report for March 15, 1999
U.S. Nuclear Regulatory Commission
Operations Center
Event Reports For
03/12/1999 - 03/15/1999
** EVENT NUMBERS **
35387 35461 35462 35463 35464 35465 35466 35467 35468 35469 35470 35471
!!!!!!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!!!!
+------------------------------------------------------------------------------+
|Fuel Cycle Facility |Event Number: 35387 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: PADUCAH GASEOUS DIFFUSION PLANT |NOTIFICATION DATE: 02/20/1999|
| RXTYPE: URANIUM ENRICHMENT FACILITY |NOTIFICATION TIME: 08:44[EST]|
| COMMENTS: 2 DEMOCRACY CENTER |EVENT DATE: 02/19/1999|
| 6903 ROCKLEDGE DRIVE |EVENT TIME: 10:05[CST]|
| BETHESDA, MD 20817 (301)564-3200 |LAST UPDATE DATE: 03/12/1999|
| CITY: PADUCAH REGION: 3 +-----------------------------+
| COUNTY: McCRACKEN STATE: KY |PERSON ORGANIZATION |
|LICENSE#: GDP-1 AGREEMENT: Y |MIKE JORDAN R3 |
| DOCKET: 0707001 |CHARLEY HAUGHNEY NMSS |
+------------------------------------------------+ |
| NRC NOTIFIED BY: K. A. BEASLEY | |
| HQ OPS OFFICER: JOHN MacKINNON | |
+------------------------------------------------+ |
|EMERGENCY CLASS: N/A | |
|10 CFR SECTION: | |
|NCFR NON CFR REPORT REQMNT | |
| | |
| | |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| WATER INVENTORY CONTROL SYSTEM ACTIVATION |
| |
| A water inventory control system (WICS) activation occurred on C-360 |
| position 4 autoclave on February 19, 1999, at 1005 CST. A high level drain |
| primary alarm was received during a cylinder sampling heat cycle. The |
| safety system did operate as required (shuts off steam to the autoclave |
| which reduces the amount of condensation in the autoclave). The purpose of |
| the WICS is to limit the amount of condensate in the autoclave. The cause |
| of this actuation is being investigated. The certificate holder thinks that |
| this event might have been caused by an invalid signal, and if it is |
| determined that this event was caused by an invalid signal, this event |
| notification will be retracted at a later time. |
| |
| The safety system actuation is reportable per Safety Analysis Report, |
| Section 6.9, Table 1, Criteria J.2, Safety System Actuation due to a Valid |
| Signal, as a 24-hour event notification. |
| |
| The NRC resident inspector was notified of this event. |
| |
| ***RETRACTION on 03/12/99 at 1324 EST from W. F. Cage taken by |
| MacKinnon**** |
| |
| Subsequent investigation and troubleshooting of the autoclave systems has |
| concluded that the WICS actuated due to an invalid signal. This condition |
| is supported by the following: |
| |
| 1. The actuation occurred at a point in the heating cycle after the maximum |
| steam demand and resulting highest condensate load has passed. Past history |
| has shown that valid actuations occur during maximum steam and condensate |
| load, not afterward at lower loads. |
| |
| 2. The actuation was initiated by the primary condensate probe only. The |
| secondary condensate probe did not alarm until after the steam supply had |
| been isolated by the WICS actuation, which caused a drop in autoclave |
| pressure supplying the motive force, driving the condensate out into the |
| drain. This, and testing subsequent to the event, proved that both probes |
| were operable and would have alarmed had water in the drain actually risen |
| to the probe level. |
| |
| 3. The drain line was inspected, and no obstruction was noted that could |
| have caused a blockage or disruption of condensate flow. |
| |
| 4. Inspection of the autoclave electrical systems indicated that some of |
| the condensate probe wires were not properly, or firmly, grounded. These |
| loose connections are considered to be the most likely cause of the WICS |
| actuation. This would be an actuation from an invalid signal, i.e., not |
| what the safety system is designed to protect against. |
| |
| The NRC resident inspector was not notified of this event by the certificate |
| holder. The NRC Region 3 (Ron Gardner) and NMSS EO (Fred Combs) were |
| notified by the NRC operations officer. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35461 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: SOUTH TEXAS REGION: 4 |NOTIFICATION DATE: 03/12/1999|
| UNIT: [1] [2] [] STATE: TX |NOTIFICATION TIME: 00:07[EST]|
| RXTYPE: [1] W-4-LP,[2] W-4-LP |EVENT DATE: 03/11/1999|
+------------------------------------------------+EVENT TIME: 11:30[CST]|
| NRC NOTIFIED BY: KLIMPLE |LAST UPDATE DATE: 03/12/1999|
| HQ OPS OFFICER: CHAUNCEY GOULD +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |GARY SANBORN R4 |
|10 CFR SECTION: | |
|NINF INFORMATION ONLY | |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|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
+------------------------------------------------------------------------------+
| THE PLANT DOES NOT MEET THE DESIGN BASIS REQUIREMENT FOR RHR HEAT EXCHANGER |
| CONTROL VALVES TO BE SECURED IN A SAFE POSITION. |
| |
| DURING A REVIEW OF THE OPERATING PROCEDURES FOR THE RESIDUAL HEAT REMOVAL |
| (RHR) HEAT EXCHANGER SYSTEM, NO STEPS OR LINEUPS WERE FOUND TO IMPLEMENT THE |
| DESIGN BASIS REQUIREMENT FOR THE RHR HEAT EXCHANGER FLOW AND BYPASS CONTROL |
| VALVES TO BE SECURED IN THEIR SAFE POSITION. |
| |
| THE RHR HEAT EXCHANGER FLOW AND BYPASS CONTROL VALVES WERE FOUND TO BE |
| CONFIGURED WITH THE SOLENOID VENT VALVES IN AN ENERGIZED CONDITION, |
| THEREFORE, POTENTIALLY ALLOWING THE NON-SAFETY-RELATED POSITIONER TO CONTROL |
| THE VALVE. |
| FAILURE OF ONE OF THE NON-SAFETY POSITIONERS (DUE TO ADVERSE CONDITIONS) |
| COULD HAVE DRIVEN THE VALVES TO THEIR NON-SAFETY POSITION AT THE ONSET OF |
| A POSTULATED ACCIDENT, THEREBY PREVENTING THE IMMEDIATE AVAILABILITY OF THE |
| LOW HEAD SAFETY INJECTION FLOW. |
| |
| THIS CONDITION WAS RESOLVED AS SOON AS IT WAS IDENTIFIED BY DE-ENERGIZING |
| AND REMOVING THE FUSES FOR THE VALVES. |
| |
| THE NRC RESIDENT INSPECTOR WAS NOTIFIED BY THE LICENSEE. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35462 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: HARRIS REGION: 2 |NOTIFICATION DATE: 03/12/1999|
| UNIT: [1] [] [] STATE: NC |NOTIFICATION TIME: 09:48[EST]|
| RXTYPE: [1] W-3-LP |EVENT DATE: 03/12/1999|
+------------------------------------------------+EVENT TIME: 06:39[EST]|
| NRC NOTIFIED BY: KEITH HOLBROOK |LAST UPDATE DATE: 03/12/1999|
| HQ OPS OFFICER: JOHN MacKINNON +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |EDWARD MCALPINE R2 |
|10 CFR SECTION: | |
|ARPS 50.72(b)(2)(ii) RPS ACTUATION | |
|AESF 50.72(b)(2)(ii) ESF ACTUATION | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 A/R Y 100 Power Operation |0 Hot Standby |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| AUTOMATIC REACTOR TRIP DUE TO TURBINE TRIP AND MAIN FEEDWATER ISOLATION |
| CAUSED BY HIGH STEAM GENERATOR WATER LEVEL. |
| |
| This is a report of the following actuations: ESF-P-14, Turbine Trip and |
| Main Feedwater Isolation; auxiliary feedwater actuation; and a reactor |
| trip. |
| |
| A loss of control of the "C" feed regulating valve caused steam generator |
| water level to increase. Operators attempted to take manual control but had |
| very little control from the main control room. The steam generator water |
| level exceeded the 82.4% high steam generator water level trip setpoint |
| causing a P-14 (steam generator high level override) actuation. This |
| actuation resulted in a turbine trip/reactor trip and a loss of both running |
| main feedwater pumps. It also caused all main feedwater isolation valves to |
| close. All control rods fully inserted. No primary/secondary plant code |
| safety valves or power-operated relief valves opened. Both motor-driven and |
| the turbine-driven auxiliary feedwater pumps automatically started on low |
| steam generator water level following the reactor trip. The steam dump |
| bypass control system is operating properly and is maintaining a T(ave) of |
| 557�F. At the present time, steam generator water levels are being |
| maintained by one operating motor-driven auxiliary feedwater pump. |
| Troubleshooting is in progress to determine the failure of "C" feedwater |
| regulating valve (air-operated valve). |
| |
| The offsite electrical grid is stable, and all emergency core cooling |
| systems and the emergency diesel generators are fully operable if needed. |
| |
| The NRC resident inspector was notified of this event by the licensee. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35463 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: MCGUIRE REGION: 2 |NOTIFICATION DATE: 03/12/1999|
| UNIT: [1] [2] [] STATE: NC |NOTIFICATION TIME: 12:24[EST]|
| RXTYPE: [1] W-4-LP,[2] W-4-LP |EVENT DATE: 03/11/1999|
+------------------------------------------------+EVENT TIME: 15:00[EST]|
| NRC NOTIFIED BY: MIKE WILDER |LAST UPDATE DATE: 03/12/1999|
| HQ OPS OFFICER: DICK JOLLIFFE +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |EDWARD MCALPINE R2 |
|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 |100 Power Operation |
|2 N Y 82 Power Operation |82 Power Operation |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| - Discrepancies between Plant Fire Protection Program and 10CFR50, Appendix |
| R, Safety Evaluation Report - |
| |
| A self initiated technical audit of the McGuire Fire Protection Program |
| identified apparent deviations from the approved Fire Protection Program. A |
| potential discrepancy exists between the McGuire Fire Protection Program and |
| the NRC description provided in Safety Evaluation Report (SER), Supplement |
| 6, regarding Appendix R, Section III.G.3. Additional deviations from |
| certain licensee commitments regarding testing were also identified in the |
| audit. |
| |
| The licensee's engineering staff has evaluated deviations identified by the |
| audit team and has determined that the fire protection related systems are |
| fully operable. In addition, these deviations have no impact on achieving |
| and maintaining safe shutdown following a design bases fire event. |
| |
| McGuire Facility Operating License (FOL) NPF-9 (Unit 1) and NPF-17 (Unit 2) |
| require 24-hour notification to the NRC for deviations from the approved |
| Fire Protection Program. The above deviations are being reported under that |
| license condition criterion. A follow-up report describing the cause of the |
| deviations and corrective actions will be submitted to the NRC within 14 |
| days. |
| |
| The licensee notified the NRC resident inspector |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35464 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: POINT BEACH REGION: 3 |NOTIFICATION DATE: 03/12/1999|
| UNIT: [1] [2] [] STATE: WI |NOTIFICATION TIME: 15:08[EST]|
| RXTYPE: [1] W-2-LP,[2] W-2-LP |EVENT DATE: 03/12/1999|
+------------------------------------------------+EVENT TIME: 13:14[CST]|
| NRC NOTIFIED BY: PHIL SHORT |LAST UPDATE DATE: 03/12/1999|
| HQ OPS OFFICER: DICK JOLLIFFE +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |RONALD GARDNER R3 |
|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 |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 N Y 100 Power Operation |100 Power Operation |
|2 N Y 73 Power Operation |73 Power Operation |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| - AUXILIARY FEEDWATER SYSTEM FOR BOTH UNITS INOPERABLE FOR 4 MINUTES - |
| |
| UNIT 1 IS OPERATING AT 100% POWER, AND UNIT 2 IS OPERATING AT 73% POWER. |
| |
| ON 03/12/99, DURING AN ENGINEERING EVALUATION, LICENSEE ENGINEERING |
| DEPARTMENT PERSONNEL DETERMINED THAT THE UNIT 2 TURBINE-DRIVEN (TD) |
| AUXILIARY FEEDWATER (AFW) |
| PUMP FLOW RATE HAD BEEN SET INCORRECTLY DURING A PREVIOUSLY PERFORMED TEST. |
| LICENSEE PERSONNEL DETERMINED THAT THIS CONDITION AFFECTED THE ENTIRE AFW |
| SYSTEM. (PORTIONS OF THE AFW SYSTEM ARE SHARED BY BOTH UNITS.) |
| |
| AT 1314 CST ON 03/12/99, THE LICENSEE ADMINISTRATIVELY DECLARED THE UNIT 1 |
| TD AFW PUMP, THE UNIT 2 TD AFW PUMP, AND THE TWO SHARED MOTOR-DRIVEN (MD) |
| AFW PUMPS INOPERABLE (AND THUS, THE ENTIRE AFW SYSTEM) AND ENTERED TECHNICAL |
| SPECIFICATION LIMITING CONDITION FOR OPERATION ACTION STATEMENT 15.3.0.B |
| (a.k.a. 3.0.3). ALL FOUR AFW PUMPS REMAINED FUNCTIONAL. THIS CONDITION IS |
| CONSIDERED TO BE OUTSIDE THE DESIGN BASES OF THE PLANT. |
| |
| AT 1318 CST ON 03/12/99, THE LICENSEE TRIPPED THE UNIT 2 TD AFW PUMP AND |
| DECLARED IT INOPERABLE, ENTERED TECH SPEC LCO A/S 15.3.4.C.1 (72 HOUR |
| SHUTDOWN LCO), DECLARED THE OTHER THREE AFW PUMPS OPERABLE, AND EXITED TECH |
| SPEC LCO A/S 15.3.0.B. THUS, ONLY THE UNIT 2 PORTION OF THE AFW SYSTEM IS |
| INOPERABLE. |
| |
| THE LICENSEE IS DETERMINING CORRECTIVE ACTIONS. |
| |
| THE LICENSEE INFORMED THE NRC RESIDENT INSPECTOR. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35465 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: PEACH BOTTOM REGION: 1 |NOTIFICATION DATE: 03/12/1999|
| UNIT: [] [3] [] STATE: PA |NOTIFICATION TIME: 16:23[EST]|
| RXTYPE: [2] GE-4,[3] GE-4 |EVENT DATE: 03/12/1999|
+------------------------------------------------+EVENT TIME: 12:34[EST]|
| NRC NOTIFIED BY: PHIL BREIDENBAUGH |LAST UPDATE DATE: 03/12/1999|
| HQ OPS OFFICER: DICK JOLLIFFE +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |TOM MOSLAK R1 |
|10 CFR SECTION: | |
|AESF 50.72(b)(2)(ii) ESF ACTUATION | |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|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
+------------------------------------------------------------------------------+
| - RCIC SYSTEM VALVES CLOSED TO ISOLATE CONTAINMENT DURING MAINTENANCE |
| ACTIVITIES - |
| |
| AT 1234 ON 03/12/99, WITH UNIT 3 AT 100% POWER, A PRIMARY CONTAINMENT |
| ISOLATION OCCURRED FOR THE REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM |
| DURING RESTORATION FROM MAINTENANCE ACTIVITIES. THE ISOLATION OCCURRED WHEN |
| THE OUTBOARD STEAM ISOLATION VALVE WAS JOGGED OPEN TO REPRESSURIZE THE STEAM |
| LINE TO THE RCIC TURBINE. WHEN THE VALVE WAS JOGGED OPEN, A MOMENTARY HIGH |
| STEAM FLOW SIGNAL OCCURRED AND BOTH INBOARD AND OUTBOARD ISOLATION VALVES |
| CLOSED TO THEIR ISOLATION POSITION. |
| |
| AT 1620 ON 03/12/99, THE LICENSEE RESET ALL SYSTEMS TO NORMAL. |
| |
| THE LICENSEE INFORMED THE NRC RESIDENT INSPECTOR. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|General Information or Other |Event Number: 35466 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| REP ORG: COOPER ENERGY SERVICES |NOTIFICATION DATE: 03/12/1999|
|LICENSEE: COOPER ENERGY SERVICES |NOTIFICATION TIME: 16:27[EST]|
| CITY: GROVE CITY REGION: 1 |EVENT DATE: 03/12/1999|
| COUNTY: STATE: PA |EVENT TIME: 12:00[EST]|
|LICENSE#: AGREEMENT: N |LAST UPDATE DATE: 03/12/1999|
| DOCKET: |+----------------------------+
| |PERSON ORGANIZATION |
| |TOM MOSLAK R1 |
| |RONALD GARDNER R3 |
+------------------------------------------------+GARY SANBORN R4 |
| NRC NOTIFIED BY: JOHN M. HORNE |VERN HODGE NRR |
| HQ OPS OFFICER: DICK JOLLIFFE | |
+------------------------------------------------+ |
|EMERGENCY CLASS: N/A | |
|10 CFR SECTION: | |
|CDEG 21.21(c)(3)(i) DEFECTS/NONCOMPLIANCE | |
| | |
| | |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| - Amendment to 10 CFR Part 21 - Cooper-Bessemer KSV Emergency Diesel |
| Generator Power Piston Failure - |
| (Refer to event #32416 for additional information.) |
| |
| Following the failure of a KSV power piston due to a hydraulic lock at |
| Commonwealth Edison Zion Station in January, 1997, Cooper Cameron |
| Corporation issued a Part 21 notification letter dated May 29, 1997, |
| reference QCG-10389. The piston which failed had a minimum crown thickness |
| of 0.040 inches, but had operated successfully for several years, and failed |
| only because of the unusual hydraulic lock event. That letter recommended |
| that KSV pistons be inspected for crown thickness by ultrasonic or other |
| methods when the pistons were exposed in the course of normal maintenance |
| activities. A conservative minimum thickness limit of 0.100 inches was |
| established. |
| |
| In the 2 years since the Zion failure, a total of 198 or more pistons have |
| been measured at seven different sites. Of these, one was found initially |
| in our plant with a thickness of 0.070 inches and was destroyed. All other |
| pistons checked have been above the 0.100-inch limit, and the actual |
| measured thickness was recorded for 157 of these. All but seven of those |
| documented have had a minimum thickness of 0.150 inch or greater. The |
| distribution of minimum thickness for the pistons has been documented to be |
| from 0.100 to >0.400 inches. These results, in combination with the |
| successful operating history of the KSV engine, provide reasonable assurance |
| that all potentially defective pistons have been removed from service. |
| |
| Based on this information, Cooper-Bessemer believes it is not necessary to |
| continue to measure the thickness of piston crowns during other maintenance |
| activities. |
| |
| This 10 CFR Part 21 Amendment applies to the following plants: |
| Region 1 - Nine Mile Point and Susquehanna |
| Region 3 - Byron and Zion |
| Region 4 - Cooper, Palo Verde, Waterford, South Texas and Grand Gulf. |
| |
| (Call the NRC Operations Center for a contact telephone number.) |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35467 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: DIABLO CANYON REGION: 4 |NOTIFICATION DATE: 03/12/1999|
| UNIT: [1] [2] [] STATE: CA |NOTIFICATION TIME: 16:37[EST]|
| RXTYPE: [1] W-4-LP,[2] W-4-LP |EVENT DATE: 03/12/1999|
+------------------------------------------------+EVENT TIME: 13:00[PST]|
| NRC NOTIFIED BY: ART WELLS |LAST UPDATE DATE: 03/12/1999|
| HQ OPS OFFICER: DICK JOLLIFFE +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |GARY SANBORN R4 |
|10 CFR SECTION: | |
|DDDD 73.71 UNSPECIFIED PARAGRAPH | |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 N N 0 Hot Shutdown |0 Hot Shutdown |
|2 N Y 100 Power Operation |100 Power Operation |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| - PLANT SECURITY REPORT - |
| |
| A SAFEGUARDS SYSTEM DEGRADATION RELATED TO PERIMETER MONITORING OCCURRED. |
| COMPENSATORY MEASURES IMMEDIATELY WERE TAKEN UPON DISCOVERY. REFER TO THE |
| HOO LOG FOR ADDITIONAL DETAILS. |
| |
| THE LICENSEE INFORMED THE NRC RESIDENT INSPECTOR. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35468 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: MILLSTONE REGION: 1 |NOTIFICATION DATE: 03/12/1999|
| UNIT: [] [] [3] STATE: CT |NOTIFICATION TIME: 17:39[EST]|
| RXTYPE: [1] GE-3,[2] CE,[3] W-4-LP |EVENT DATE: 03/12/1999|
+------------------------------------------------+EVENT TIME: 17:31[EST]|
| NRC NOTIFIED BY: JOE RUTTAR |LAST UPDATE DATE: 03/12/1999|
| HQ OPS OFFICER: DICK JOLLIFFE +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |TOM MOSLAK R1 |
|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 |
+-----+----------+-------+--------+-----------------+--------+-----------------+
| | |
| | |
|3 N Y 100 Power Operation |100 Power Operation |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| -Control room personnel access door latch mechanism may not withstand HELB |
| pressure in the turbine building- |
| |
| A licensee preliminary review of an error in an analysis assumption for the |
| pressure applied to a control room personnel access door due to a high |
| energy line break (HELB) in the turbine building indicates that the door |
| latch mechanism may not withstand the applied pressure. |
| |
| The control room door has been latched with an alternative latching |
| mechanism and will remain closed against this pressure so there are no |
| current operability concerns. |
| |
| This situation results in a historical condition where Unit 3 may have |
| operated outside its design basis. Engineering evaluation is continuing. A |
| conservative decision has been made to report this condition as a condition |
| outside the design basis of Unit 3 pursuant to 10CFR50.72(b)(1)(ii)(B) |
| pending the results of this evaluation. |
| |
| The licensee informed the NRC resident inspector. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35469 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: SOUTH TEXAS REGION: 4 |NOTIFICATION DATE: 03/12/1999|
| UNIT: [] [2] [] STATE: TX |NOTIFICATION TIME: 19:02[EST]|
| RXTYPE: [1] W-4-LP,[2] W-4-LP |EVENT DATE: 03/12/1999|
+------------------------------------------------+EVENT TIME: 14:12[CST]|
| NRC NOTIFIED BY: TIM FRAWLEY |LAST UPDATE DATE: 03/13/1999|
| HQ OPS OFFICER: DICK JOLLIFFE +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |GARY SANBORN R4 |
|10 CFR SECTION: |WILLIAM BECKNER, EO NRR |
|AESF 50.72(b)(2)(ii) ESF ACTUATION |ROSEMARY HOGAN IRO |
|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
+------------------------------------------------------------------------------+
| - PARTIAL LOSS OF OFFSITE POWER DUE TO A FAULT IN A SWITCHYARD CIRCUIT |
| BREAKER - |
| |
| AT 1412 CST ON 03/12/99, DURING PLANT SWITCHYARD ACTIVITIES, UNIT 2 |
| EXPERIENCED A LOSS OF POWER TO THE #2 STANDBY TRANSFORMER DUE TO A FAULT IN |
| A SWITCHYARD CIRCUIT BREAKER. THIS CONDITION CAUSED TRAIN 'B' AND TRAIN 'C' |
| 4160-VOLT ESF BUSES TO ACTUATE ON A LOSS OF OFFSITE POWER. THE FOLLOWING |
| ESF SYSTEMS ACTUATED: TRAIN 'B' AND TRAIN 'C' ESF EDGs, ESSENTIAL COOLING |
| WATER SYSTEM, ESSENTIAL CHILLED WATER SYSTEM, COMPONENT COOLING WATER |
| SYSTEM, CONTROL ROOM HVAC SYSTEM, AND AUXILIARY FEEDWATER SYSTEM (ON RECIRC |
| MODE). |
| |
| DURING THIS EVENT, THE SPENT FUEL POOL TEMPERATURE INCREASED APPROXIMATELY |
| 1.5�F, AND THE TRAIN 'B' ESF EDG (#22) OUTPUT BREAKER FAILED TO CLOSE |
| AUTOMATICALLY. THE LICENSEE DECLARED THE TRAIN 'B' ESF EDG (#22) |
| INOPERABLE, EVEN THOUGH IT IS FUNCTIONAL (72-HOUR LCO). |
| |
| THERE WAS NO FIRE, AND NO PERSONNEL WERE INJURED. THE LICENSEE RESTORED ALL |
| SYSTEMS TO NORMAL. UNIT 2 IS STABLE AND OPERATING AT 100% POWER. |
| |
| THE LICENSEE IS INVESTIGATING THE CAUSE OF THE FAULT IN THE SWITCHYARD |
| CIRCUIT BREAKER AND THE TRAIN 'B' ESF EDG (#22) OUTPUT BREAKER FAILING TO |
| CLOSE AUTOMATICALLY. |
| |
| THE LICENSEE PLANS TO SUBMIT A LICENSEE EVENT REPORT ON THIS EVENT TO THE |
| NRC. |
| |
| THIS EVENT HAD NO IMPACT ON UNIT 1. |
| |
| THE NRC RESIDENT INSPECTOR WAS ON SITE DURING THIS EVENT AND REPORTED IT TO |
| REGION 4. |
| REGION 4 PERSONNEL WERE AWARE OF THIS EVENT PRIOR TO THE LICENSEE REPORTING |
| IT TO THE NRC OPERATIONS OFFICER. |
| |
| |
| * * * UPDATE AT 1207 ON 03/13/99 FROM TIM FRAWLEY TO JOLLIFFE * * * |
| |
| AT 1412 CST ON 03/12/99, DURING THE LOSS OF POWER TO #2 STANDBY TRANSFORMER, |
| THE LICENSEE ENTERED TECH SPEC 3.0.3 FOR UNIT 2 DUE TO THE LOSS OF POWER TO |
| TWO 13.8-KV STANDBY BUSES AND THE INOPERABLE CONDITION OF THE TRAIN 'B' ESF |
| EDG (#22); THIS CONDITION IS NOT COVERED UNDER THE ACTIONS OF TECH SPEC |
| 3.8.1.1 (AC ELECTRICAL POWER SOURCES). |
| |
| AT 1530 CST ON 03/12/99, THE LICENSEE RESTORED POWER TO THE TWO 13.8-KV |
| STANDBY BUSES. |
| |
| AT 1553 CST ON 03/12/99, THE LICENSEE EXITED TECH SPEC 3.0.3 FOLLOWING |
| SATISFACTORY PERFORMANCE OF SURVEILLANCE REQUIREMENT 4.8.1.1.1.a. THIS |
| CONDITION IS REPORTABLE TO THE NRC UNDER SECTION 2.G OF FACILITY OPERATING |
| LICENSE NPF-80. |
| |
| UNIT 2 IS STABLE AND OPERATING AT 100% POWER. |
| |
| THE LICENSEE INFORMED THE NRC RESIDENT INSPECTOR. THE NRC OPERATIONS |
| OFFICER NOTIFIED THE R4DO (GARY SANBORN), NRR EO (BILL BECKNER), AND IRO MGR |
| (ROSEMARY HOGAN). |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35470 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: SEABROOK REGION: 1 |NOTIFICATION DATE: 03/13/1999|
| UNIT: [1] [] [] STATE: NH |NOTIFICATION TIME: 16:31[EST]|
| RXTYPE: [1] W-4-LP |EVENT DATE: 03/13/1999|
+------------------------------------------------+EVENT TIME: 10:32[EST]|
| NRC NOTIFIED BY: PATRICK CYR |LAST UPDATE DATE: 03/13/1999|
| HQ OPS OFFICER: DICK JOLLIFFE +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |TOM MOSLAK R1 |
|10 CFR SECTION: | |
|HFIT 26.73 FITNESS FOR DUTY | |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|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
+------------------------------------------------------------------------------+
| A CONTRACTOR SUPERVISOR WAS DETERMINED TO BE UNDER THE INFLUENCE OF ALCOHOL |
| DURING A RANDOM TEST. THE CONTRACTOR'S ACCESS AUTHORIZATION TO THE PLANT |
| HAS BEEN TERMINATED. REFER TO THE HOO LOG FOR ADDITIONAL DETAILS. |
| |
| THE LICENSEE PLANS TO INFORM THE NRC RESIDENT INSPECTOR. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35471 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: CLINTON REGION: 3 |NOTIFICATION DATE: 03/15/1999|
| UNIT: [1] [] [] STATE: IL |NOTIFICATION TIME: 03:57[EST]|
| RXTYPE: [1] GE-6 |EVENT DATE: 03/15/1999|
+------------------------------------------------+EVENT TIME: 01:05[CST]|
| NRC NOTIFIED BY: TIM HOLLAND |LAST UPDATE DATE: 03/15/1999|
| HQ OPS OFFICER: LEIGH TROCINE +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |RONALD GARDNER R3 |
|10 CFR SECTION: | |
|AESF 50.72(b)(2)(ii) ESF ACTUATION | |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 N N 0 Cold Shutdown |0 Cold Shutdown |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| UNANTICIPATED REACTOR CORE ISOLATION COOLING (RCIC) SUCTION VALVE ISOLATION |
| DURING PERFORMANCE OF A REVISED TEST PROCEDURE |
| |
| The following text is a portion of a facsimile received from the licensee: |
| |
| "During the performance of CPS No. 9054.04 (RCIC Automatic Suction Shift |
| Test), the RCIC suction valve from the suppression pool (1E51-F031) |
| automatically isolated on a valid low steam line pressure signal. This |
| isolation was not intended to occur during the performance of this |
| procedure." |
| |
| "The isolation occurred following the removal of a simulated RCIC steam line |
| pressure signal of > 60 psig. When the current steam line pressure of 0 |
| psig was picked up by the logic, the 1E51-F031 isolated as designed." Upon |
| receipt of the valid low steam line pressure signal, all systems functioned |
| as required and there was nothing unusual or not understood. |
| |
| The licensee stated that the test procedure was newly revised and that some |
| of the steps had been swapped. The steps should have requested opening of |
| the other suction source, closure of this suction source, and then removal |
| the simulator. |
| |
| The licensee notified the NRC resident inspector. |
+------------------------------------------------------------------------------+
Page Last Reviewed/Updated Wednesday, March 24, 2021
Page Last Reviewed/Updated Wednesday, March 24, 2021