Event Notification Report for January 25, 1999
U.S. Nuclear Regulatory Commission
Operations Center
Event Reports For
01/22/1999 - 01/25/1999
** EVENT NUMBERS **
35197 35297 35298 35299 35300 35301 35302 35303 35304 35305 35306 35307
35308 35309 35310 35311
!!!!!!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!!!!
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35197 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: LIMERICK REGION: 1 |NOTIFICATION DATE: 12/27/1998|
| UNIT: [] [2] [] STATE: PA |NOTIFICATION TIME: 17:43[EST]|
| RXTYPE: [1] GE-4,[2] GE-4 |EVENT DATE: 12/27/1998|
+------------------------------------------------+EVENT TIME: 14:30[EST]|
| NRC NOTIFIED BY: JOHN HUNTER |LAST UPDATE DATE: 01/22/1999|
| HQ OPS OFFICER: LEIGH TROCINE +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |JOHN WHITE R1 |
|10 CFR SECTION: | |
|AINA 50.72(b)(2)(iii)(A) POT UNABLE TO SAFE SD | |
|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 |
+-----+----------+-------+--------+-----------------+--------+-----------------+
| | |
|2 N Y 100 Power Operation |100 Power Operation |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| HIGH PRESSURE COOLANT INJECTION (HPCI) DECLARED INOPERABLE DURING |
| SURVEILLANCE TESTING DUE TO LOW LUBE OIL PRESSURES. |
| |
| The following text is a portion of a facsimile received from the licensee: |
| |
| "Unit 2 HPCI was secured and declared inoperable after being placed in |
| service for a regularly scheduled surveillance test. During performance of |
| the pump, valve, and flow test, a lube oil pressure alarm annunciated due to |
| low lube oil pressures in various points throughout the system. All other |
| parameters were normal, and the system was operating properly. HPCI was |
| then secured, and a normal shutdown of the system was achieved. No other |
| abnormalities were identified throughout the entire evolution. |
| Investigation continues as to the cause of the low lube oil pressure |
| alarm." |
| |
| The unit was placed in a 14-day technical specification limiting condition |
| for operation as a result of this issue. |
| |
| The licensee plans to notify the NRC resident inspector. |
| |
| *** RETRACTION OF EVENT ON 01/22/99 AT 1418 EST FROM TONKINSON TAKEN BY |
| MacKINNON *** |
| |
| During surveillance testing of the Unit 2 HPCI system on 12/27/98, a low oil |
| pressure alarm was received in the control room. Lube oil pressure was |
| observed to be below the recommended value on a local pressure gauge at one |
| location. The low oil pressure was on a section of piping that supplies |
| lube oil to the governor end bearing only. All other system parameters were |
| normal, and surveillance testing acceptance criteria were satisfied. The |
| system was declared inoperable at that time. |
| |
| The ball valve supplying lube oil to the affected portion of piping was |
| removed, cleaned, and re-installed. Other lube oil system inspections |
| occurred. The lube oil was analyzed, and there was no evidence of bearing |
| degradation. |
| |
| Subsequent engineering analysis, with support from the turbine and bearing |
| manufacturers, concluded that the as-found oil pressure was sufficient to |
| supply lube oil to the governor end bearing indefinitely. No bearing damage |
| would be expected with operation at the observed oil pressure. The as-found |
| condition would not adversely affect the capability of the HPCI system to |
| fulfill safety-related functions. The HPCI system, therefore, was not |
| inoperable due to as-found lube oil pressure condition. |
| |
| The NRC resident inspector was notified of this retraction by the licensee. |
| The R2DO (Costello) was notified by the NRC operations officer. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35297 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: DIABLO CANYON REGION: 4 |NOTIFICATION DATE: 01/22/1999|
| UNIT: [1] [2] [] STATE: CA |NOTIFICATION TIME: 00:59[EST]|
| RXTYPE: [1] W-4-LP,[2] W-4-LP |EVENT DATE: 01/21/1999|
+------------------------------------------------+EVENT TIME: 18:25[PST]|
| NRC NOTIFIED BY: JOSEPHINE BROWN |LAST UPDATE DATE: 01/22/1999|
| HQ OPS OFFICER: BOB STRANSKY +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |DALE POWERS 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
+------------------------------------------------------------------------------+
| This notification is a courtesy call regarding a safeguards system |
| degradation related to a computer. Compensatory measures were taken |
| immediately upon discovery. (Refer to the HOO log for additional details.) |
| |
| The NRC resident inspector will be informed of this notification by the |
| licensee. |
| |
| *** UPDATE ON 01/22/99 AT 1430 EST FROM ART WELLS TAKEN BY MacKINNON *** |
| |
| After further review, the licensee made this a loggable report. The |
| secondary alarm system is back in service. |
| |
| The NRC resident Inspector will be notified of this event update by the |
| licensee. The R4DO (Powers) was notified by the NRC operations officer. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35298 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: INDIAN POINT REGION: 1 |NOTIFICATION DATE: 01/22/1999|
| UNIT: [] [3] [] STATE: NY |NOTIFICATION TIME: 09:38[EST]|
| RXTYPE: [2] W-4-LP,[3] W-4-LP |EVENT DATE: 01/22/1999|
+------------------------------------------------+EVENT TIME: 09:15[EST]|
| NRC NOTIFIED BY: C. KOCSIS |LAST UPDATE DATE: 01/22/1999|
| HQ OPS OFFICER: BOB STRANSKY +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |FRANK COSTELLO 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
+------------------------------------------------------------------------------+
| SINGLE FAILURE COULD PREVENT COMPLETE CONTAINMENT ISOLATION DURING VENTING |
| (Refer to event #35301 for a similar event on Unit 2.) |
| |
| The following text is a portion of a facsimile received from the licensee: |
| |
| "At approximately 0915 hours on January 22, 1999, it was determined that |
| during containment pressure relief, the containment isolation function could |
| not be completely achieved if there [was] a containment isolation signal, |
| coupled with the single failure of containment isolation valve VS-PCV-1190 |
| to close. This would occur since initiation of pressure relief results in |
| three-way valve PS-SOV-1280 (a one-inch valve with one-inch ports on the |
| weld channel supply line between VS-PCV-1190 and VS-PCV-1191) changing |
| position to isolate [the] weld channel and vent the line between the |
| containment isolation valves (VS-PCV-1190 and VS-PCV-1191) to atmosphere. |
| If a postulated event were to occur that resulted in a containment isolation |
| signal, VS-PCV-1190 must close before an interlock with PS-SOV-1280 would |
| allow that valve to change position and supply weld channel gas between the |
| containment isolation valves. Thus, during pressure relief, a single |
| failure of VS-PCV-1190 to close on a containment isolation signal could |
| result in a one-inch vent path. Immediate corrective action was taken to |
| administratively restrict containment pressure relief until corrective |
| action to assure containment integrity during containment pressure relief |
| can be established such as isolation using an installed weld channel manual |
| isolation valve. This is a condition that resulted, during past containment |
| pressure relief operations, in Indian Point 3 being outside the plant design |
| basis for containment isolation." |
| |
| The NRC resident inspector has been informed of this notification by the |
| licensee. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Other Nuclear Material |Event Number: 35299 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| REP ORG: U.S. ARMY |NOTIFICATION DATE: 01/22/1999|
|LICENSEE: U.S. ARMY |NOTIFICATION TIME: 09:48[EST]|
| CITY: FT. SHAFTER REGION: 4 |EVENT DATE: 01/21/1999|
| COUNTY: STATE: HI |EVENT TIME: [HST]|
|LICENSE#: 12-00722-06 AGREEMENT: N |LAST UPDATE DATE: 01/22/1999|
| DOCKET: |+----------------------------+
| |PERSON ORGANIZATION |
| |ROGER LANKSBURY R3 |
| |DONALD COOL NMSS |
+------------------------------------------------+DALE POWERS R4 |
| NRC NOTIFIED BY: J. HAVENNER | |
| HQ OPS OFFICER: BOB STRANSKY | |
+------------------------------------------------+ |
|EMERGENCY CLASS: N/A | |
|10 CFR SECTION: | |
|IBBF 30.50(b)(2)(ii) EQUIP DISABLED/FAILS | |
| | |
| | |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| EXCESSIVE SURFACE CONTAMINATION OF SEALED SOURCE |
| |
| The U.S. Army Radioisotope Committee, located in Rock Island, IL, reported |
| the following incident that occurred at Fort Shafter, HI. Wipe tests of a |
| chemical agent monitor containing a 10-mCi Ni-63 source indicate that the |
| sealed source is leaking. Count rates of up to 17,586 dpm/100 cm� were |
| recorded. The device has been removed from service. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35300 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: THREE MILE ISLAND REGION: 1 |NOTIFICATION DATE: 01/22/1999|
| UNIT: [1] [] [] STATE: PA |NOTIFICATION TIME: 10:06[EST]|
| RXTYPE: [1] B&W-L-LP,[2] B&W-L-LP |EVENT DATE: 01/22/1999|
+------------------------------------------------+EVENT TIME: 09:30[EST]|
| NRC NOTIFIED BY: JOHN SCHORK |LAST UPDATE DATE: 01/22/1999|
| HQ OPS OFFICER: BOB STRANSKY +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |FRANK COSTELLO 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 |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 N Y 100 Power Operation |100 Power Operation |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| BORIC ACID SYSTEM PIPING MAY NOT BE MAINTAINED AT PROPER TEMPERATURE. |
| |
| The following text is a portion of a facsimile received from the licensee: |
| |
| "CA-P-1A/B discharge piping heat trace is not maintaining adequate |
| temperatures. FSAR sec 9.2.1.2 states, 'Further, all piping, pumps, and |
| valves associated with the boric acid mix tank and the reclaimed boric acid |
| storage tanks to transport boric acid solution from them to the makeup and |
| purification system are provided with redundant electrical heat tracing to |
| ensure that the boric acid solution will be maintained 10 [�]F or more above |
| its crystallization temperature. The electrical heat tracing is controlled |
| by the temperature of the external surface of the piping systems.' |
| Temperature readings on the surface of the pipe ranged between 97.4 [�]F and |
| 171 [�]F. The heat trace setpoints are 160 [�]F. Based on recent chemistry |
| samples as high as 17,400 ppm and the crystallization curve, Figure 1A in OP |
| 1104-47B, the required temperature to prevent crystallization would be 117 |
| [�]F. Adding 10 [�]F would require a minimum 127 [�]F for the boric acid |
| solution." |
| |
| "The heat trace requirement is to ensure the boron does not crystallize and |
| prevent flow to the makeup tank. Quarterly [in-service] testing is performed |
| for CA-P-1A/B (most recently in November 1998). Although the required |
| solution temperature may not be being maintained, testing has shown that |
| these lines are not blocked and are functioning." |
| |
| "Because the temperature of the boric acid solution within the subject |
| piping cannot be confirmed via direct measurement or analysis at this time |
| to be at or above 127 [�]F, this condition has been identified as being |
| potentially outside the design basis of the plant and was reported to the |
| NRC within 1 hour in accordance with the requirements of 10 CFR |
| 50.72(a)(2)(ii)." |
| |
| "The chemical addition system pumps are currently out of service due to |
| maintenance being performed on the system, unrelated to the heat tracing. |
| The system in-service test is planned to be performed when the system is |
| returned to service. The chemical addition system in-service test performed |
| last November found the system performed as required at that time." |
| |
| "An analysis is planned to be performed to determine if the temperature of |
| the boric acid solution within the chemical addition system piping is at or |
| above the specified temperature." |
| |
| "Plans are underway to install, if necessary, additional insulation and, if |
| necessary heat tracing, to ensure the fluid temperature is maintained at the |
| correct temperature." |
| |
| "The potential outside design basis condition has been documented in CAP |
| T1999-0052 via the GPU Nuclear, Appendix B, corrective action program." |
| |
| The NRC resident inspector was notified of this event by the licensee. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35301 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: INDIAN POINT REGION: 1 |NOTIFICATION DATE: 01/22/1999|
| UNIT: [2] [] [] STATE: NY |NOTIFICATION TIME: 12:43[EST]|
| RXTYPE: [2] W-4-LP,[3] W-4-LP |EVENT DATE: 01/22/1999|
+------------------------------------------------+EVENT TIME: 12:10[EST]|
| NRC NOTIFIED BY: DENNIS CORNAX |LAST UPDATE DATE: 01/22/1999|
| HQ OPS OFFICER: STEVE SANDIN +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |FRANK COSTELLO 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 |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|2 N Y 99 Power Operation |99 Power Operation |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| SINGLE FAILURE COULD PREVENT COMPLETE CONTAINMENT ISOLATION DURING VENTING. |
| |
| The following text is a portion of a facsimile received from the licensee: |
| |
| "At approximately 12:10 hours on January 22, 1999, it was determined that |
| during containment pressure relief, the containment isolation function could |
| not be completely achieved if there [was] a containment isolation signal, |
| coupled with the single failure of containment isolation valve PCV-1190 to |
| close. The potential for this to occur exists since initiation of pressure |
| relief results in three-way solenoid valve SOV-1280 (a one-inch valve with |
| ports on [the] weld channel supply line between PCV-1190 and PCV-1191) |
| changing position to isolate [the] weld channel and vent the line between |
| the containment isolation valves (PCV-1190 and PCV-1191) to atmosphere. If |
| a postulated accident event were to occur that resulted in a containment |
| isolation signal, PCV-1190 must close before an interlock with SOV-1280 |
| would allow that valve to change position and supply weld channel gas |
| between the containment isolation valves. Thus, during the pressure relief, |
| a single failure of PCV-1190 to close on a containment isolation signal |
| could result in a one-inch vent path, which would be a monitored release |
| path and filtered by the [primary auxiliary building] exhaust system." |
| |
| "Immediate corrective action taken will administratively require the closure |
| of a manual valve ( PCV-1110-8) any time that PCV-1190 is in the open |
| position thereby precluding a pathway for [the] containment atmosphere to |
| communicate with the environment." |
| |
| This condition was discovered in response to an event reported by Indian |
| Point 3. (Refer to event #35298 for additional information.) The licensee |
| is continuing its evaluation to identify any other primary containment |
| isolation valves which may be affected. |
| |
| The licensee informed the NRC resident inspector. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Fuel Cycle Facility |Event Number: 35302 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: PORTSMOUTH GASEOUS DIFFUSION PLANT |NOTIFICATION DATE: 01/22/1999|
| RXTYPE: URANIUM ENRICHMENT FACILITY |NOTIFICATION TIME: 13:28[EST]|
| COMMENTS: 2 DEMOCRACY CENTER |EVENT DATE: 01/22/1999|
| 6903 ROCKLEDGE DRIVE |EVENT TIME: 09:30[EST]|
| BETHESDA, MD 20817 (301)564-3200 |LAST UPDATE DATE: 01/22/1999|
| CITY: PIKETON REGION: 3 +-----------------------------+
| COUNTY: PIKE STATE: OH |PERSON ORGANIZATION |
|LICENSE#: GDP-2 AGREEMENT: N |ROGER LANKSBURY R3 |
| DOCKET: 0707002 |DON COOL, NMSS EO |
+------------------------------------------------+FRANK CONGEL IRO |
| NRC NOTIFIED BY: KEITH WILLIAMSON | |
| HQ OPS OFFICER: JOHN MacKINNON | |
+------------------------------------------------+ |
|EMERGENCY CLASS: | |
|10 CFR SECTION: | |
|NBNL RESPONSE-BULLETIN | |
| | |
| | |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| FAILURE TO IMPLEMENT AN NCSA IN BUILDING X-326 BECAUSE THERE WAS NO |
| PROCEDURE FOR OPERATION OF CALIBRATION BUGGIES |
| |
| This event was reported per NRC Bulletin 91-01 as a 4-hour notification. |
| |
| The following text is a portion of a facsimile received from Portsmouth: |
| |
| "On 01/22/1999 at 0930 hours during the on-going plant-wide search for |
| abandoned equipment, three (3) calibration buggies (all of different design) |
| were discovered in the X-326 [process building] building which may meet the |
| fissile material limits. All three (3) buggies meet the requirements of |
| NCSA-Plant069; but the NCSA was not implemented in the X-326 building |
| because there was not a procedure for operation of the calibration buggies. |
| The identified equipment has had a boundary set up around them, and the |
| anomalous NCS condition report is complete." |
| |
| "Spacing and geometry of the components on the buggies were controlled such |
| that the requirements of Plant069 were met." |
| |
| "The calibration buggies found do not violate the requirements of |
| NCSA-Plant069. The problem was the failure to flow down the requirements of |
| the NCSA into a procedure for the operation of the buggies. The problem was |
| in the implementation of the NCSA. The NCSA states that it is for use in |
| building X-326, but [it] was never implemented in that building." |
| |
| The NRC resident inspector was notified of this event by the certificate |
| holder. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35303 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: LIMERICK REGION: 1 |NOTIFICATION DATE: 01/22/1999|
| UNIT: [] [2] [] STATE: PA |NOTIFICATION TIME: 14:53[EST]|
| RXTYPE: [1] GE-4,[2] GE-4 |EVENT DATE: 01/22/1999|
+------------------------------------------------+EVENT TIME: 14:00[EST]|
| NRC NOTIFIED BY: Glenn H. Stewart |LAST UPDATE DATE: 01/22/1999|
| HQ OPS OFFICER: JOHN MacKINNON +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |FRANK COSTELLO R1 |
|10 CFR SECTION: | |
|AOUT 50.72(b)(1)(ii)(B) OUTSIDE DESIGN BASIS | |
|NLTR LICENSEE 24 HR REPORT | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|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
+------------------------------------------------------------------------------+
| DISCOVERY THAT A FIRE INDUCED FAULT COULD IMPACT EQUIPMENT REQUIRED FOR SAFE |
| SHUTDOWN |
| |
| The following text is a portion of a facsimile received from the licensee: |
| |
| "On 01/21/99, at 1545 hours, an Engineering review determined that in the |
| event of a fire in Fire Area 64, 'Reactor Enclosure Cooling Water Equipment |
| Area,' a fire induced fault in the 480-VAC power cable to the 2B Reactor |
| Enclosure Cooling Water (RECW) pump motor could open the load center (LC) |
| breaker to its associated motor control center (MCC) which would impact |
| equipment required for safe shutdown in the event of a fire in that area. |
| At that time, it was believed that this situation could only occur if the 2B |
| RECW pump was operating at the time of the fire. The 2B RECW pump currently |
| is not operating. This condition is due to less than adequate circuit |
| breaker coordination which is limited to a small region on the time-current |
| coordination curve for the LC breaker and the MCC breaker. On 01/22/99, at |
| 1400 hours, further investigation revealed that fire-induced damage to an |
| auto-start pressure switch in the control circuit for the 2B RECW pump |
| located in the affected fire area could create a hot short that would cause |
| the pump to auto-start resulting in the identified impact on safe shutdown |
| equipment. This represents a condition that is outside the design basis of |
| the plant and is reportable as a 1-hour notification in accordance with |
| 10CFR50.72(b)(1)(ii)(B). This condition is also considered a noncompliance |
| with the Fire Protection Program as described in the Limerick Generating |
| Station (LGS) Updated Final Safety Analysis Report (UFSAR), Section |
| 9A.6.1.1, and is reportable as a violation of LGS, Unit 2, Operating License |
| Condition 2.C.(3), 'Fire Protection.' Accordingly, this notification is |
| also being made within 24 hours as required by LGS, Unit 2, Operating |
| License Condition 2.E. This condition has existed since June 22, 1989, the |
| date of issuance of the Low Power Operating License for LGS, Unit 2. This |
| condition does not effect the operability of the 2B RECW pump or the |
| affected MCC based on the application of single failure criterion which |
| limits an electrical failure to a single division of safety-related power. |
| A fire watch has been established in the affected fire area as an |
| appropriate compensatory measure." |
| |
| The licensee notified the NRC resident inspector. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|General Information or Other |Event Number: 35304 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| REP ORG: INTEGRATED RESOURCES, INC. |NOTIFICATION DATE: 01/22/1999|
|LICENSEE: BARKER MICORFARADAS, INC |NOTIFICATION TIME: 15:20[EST]|
| CITY: Nebraska City REGION: 4 |EVENT DATE: 01/22/1999|
| COUNTY: STATE: NE |EVENT TIME: 14:20[CST]|
|LICENSE#: AGREEMENT: Y |LAST UPDATE DATE: 01/22/1999|
| DOCKET: |+----------------------------+
| |PERSON ORGANIZATION |
| |DALE POWERS R4 |
| |FRANK COSTELLO R1 |
+------------------------------------------------+ROGER LANKSBURY R3 |
| NRC NOTIFIED BY: JOHN BROSEMER |VERN HODGE NRR |
| HQ OPS OFFICER: JOHN MacKINNON | |
+------------------------------------------------+ |
|EMERGENCY CLASS: N/A | |
|10 CFR SECTION: | |
|CCCC 21.21 UNSPECIFIED PARAGRAPH | |
| | |
| | |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| PRELIMINARY NOTIFICATION BY INTEGRATED RESOURCES OF 10-CFR-PART-21 |
| NOTIFICATION |
| |
| INTEGRATED RESOURCES, INC., CALLED TO MAKE A PRELIMINARY NOTIFICATION OF |
| INTENT TO ISSUE A 10-CFR-PART-21 NOTIFICATION ON THEMSELVES. DRESDEN |
| STATION SENT SQUARE ROOT CONVERTERS (DRESDEN STATION HAS ABOUT 100 OF THESE |
| SQUARE ROOT CONVERTERS) TO INTEGRATED RESOURCES, INC., FOR FAILURE ANALYSIS |
| BECAUSE AFTER ABOUT 5 YEARS OF USE, THESE SQUARE ROOT CONVERTERS START TO |
| FAIL. ALL THE SQUARE ROOT CONVERTERS ARE NON-SAFETY RELATED. NINE (9) OF |
| THE SQUARE ROOT CONVERTERS WERE TESTED. FAILURE ANALYSIS DETERMINED THAT |
| ALL FIVE (5) OF THE ALUMINUM ELECTROLYTIC CAPACITORY SPARGUE ELECTRIC CO. |
| (MODEL #TE1302 WITH MANUFACTURE DATE CODE OF 9322H) FAILED. ALL THE OTHER |
| SQUARE ROOT CONVERTERS' FAILURE POINT IS WHERE THE SQUARE ROOT CONVERTER |
| COULD NOT BE CALIBRATED PROPERLY. THE SQUARE ROOT CONVERTERS ARE BEING SENT |
| BACK TO THEIR MANUFACTURER (BARKER MICROFARADS, INC., LOCATED IN HILLSVILLE, |
| VA) TO DETERMINE THE FAILURE MECHANISM OF THE SQUARE ROOT CONVERTERS. |
| INTEGRATE RESOURCES, INC., STATED THAT THEY EXPECTED THE RESULTS OF THE |
| FAILURE MECHANISM OF THE SQUARE ROOT CONVERTS TO BE SENT TO THEM NEXT WEEK. |
| |
| |
| INTEGRATED RESOURCES, INC., SAID THAT NINE MILE POINT UNIT 1 HAS ONE |
| SAFETY-RELATED SQUARE ROOT CONVERTER AND ONE SAFETY-RELATED FUNCTION |
| GENERATOR AND THAT FITZPATRICK HAS TWO SAFETY-RELATED BASIC CONTROLLERS. |
| THESE TWO NUCLEAR POWER PLANTS WILL BE NOTIFIED OF THE POTENTIAL FAILURE OF |
| THESE DEVICES. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Other Nuclear Material |Event Number: 35305 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| REP ORG: TEXAS DEPARTMENT OF HEALTH |NOTIFICATION DATE: 01/22/1999|
|LICENSEE: TECHNICAL WELDING |NOTIFICATION TIME: 16:00[EST]|
| CITY: PASADENA REGION: 4 |EVENT DATE: 01/21/1999|
| COUNTY: STATE: TX |EVENT TIME: [CST]|
|LICENSE#: L02187 AGREEMENT: Y |LAST UPDATE DATE: 01/22/1999|
| DOCKET: |+----------------------------+
| |PERSON ORGANIZATION |
| |DALE POWERS R4 |
| | |
+------------------------------------------------+ |
| NRC NOTIFIED BY: HELEN WATKINS | |
| HQ OPS OFFICER: JOHN MacKINNON | |
+------------------------------------------------+ |
|EMERGENCY CLASS: N/A | |
|10 CFR SECTION: | |
|NAGR AGREEMENT STATE | |
| | |
| | |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| EXPOSURE GREATER THAN TEDE OF 5 REM (Refer to event #35306 for a similar |
| event.) |
| |
| THE FOLLOWING INFORMATION WAS RECEIVED VIA FACSIMILE TO THE NRC OPERATION |
| CENTER FROM THE TEXAS DEPARTMENT OF HEALTH BUREAU OF RADIATION CONTROL |
| (TDH-BRC) AS AN AGREEMENT STATE REPORT: |
| |
| "INCIDENT #7411 - TECHNICAL WELDING, PASADENA, TX, L02187, NOTIFIED TDH-BRC |
| OF A 5.5-R 1998 ANNUAL EXPOSURE TO A RADIOGRAPHER. 4.560 R WAS [RECEIVED] |
| DURING THE 12/98 MONITORING PERIOD. THE LICENSEE RECEIVED A VERBAL REPORT |
| FROM THE BADGE PROCESSOR ON 01/21/99. THE TDH-BRC IS INVESTIGATING." |
| |
| THERE WAS NO ADDITIONAL INFORMATION AVAILABLE ON THE FACSIMILE. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Other Nuclear Material |Event Number: 35306 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| REP ORG: TEXAS DEPARTMENT OF HEALTH |NOTIFICATION DATE: 01/22/1999|
|LICENSEE: TECHNICAL WELDING |NOTIFICATION TIME: 16:00[EST]|
| CITY: PASADENA REGION: 4 |EVENT DATE: 01/20/1999|
| COUNTY: STATE: TX |EVENT TIME: [CST]|
|LICENSE#: L02187 AGREEMENT: Y |LAST UPDATE DATE: 01/22/1999|
| DOCKET: |+----------------------------+
| |PERSON ORGANIZATION |
| |DALE POWERS R4 |
| | |
+------------------------------------------------+ |
| NRC NOTIFIED BY: HELEN WATKINS | |
| HQ OPS OFFICER: JOHN MacKINNON | |
+------------------------------------------------+ |
|EMERGENCY CLASS: N/A | |
|10 CFR SECTION: | |
|NAGR AGREEMENT STATE | |
| | |
| | |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| EXPOSURE GREATER THAN TEDE OF 5 REM (Refer to event #35305 for a similar |
| event.) |
| |
| THE FOLLOWING INFORMATION WAS RECEIVED VIA FACSIMILE TO THE NRC OPERATION |
| CENTER FROM THE TEXAS DEPARTMENT OF HEALTH BUREAU OF RADIATION CONTROL |
| (TDH-BRC) AS AN AGREEMENT STATE REPORT: |
| |
| "INCIDENT #7410 - TECHNICAL WELDING, PASADENA, TX, L02187, NOTIFIED THE |
| TDH-BRC OF A 5.2-R 1998 ANNUAL EXPOSURE TO A RADIOGRAPHER. THE DOSE DURING |
| THE [3] MONTHS OF THE YEAR RESULTED FROM CALCULATED ASSESSMENTS. THE BADGES |
| WERE LOST. THE LICENSEE CONFIRMED THE OVEREXPOSURE ON 01/20/99. THE |
| TDH-BRC IS INVESTIGATING." |
| |
| THERE WAS NO ADDITIONAL INFORMATION AVAILABLE ON THE FACSIMILE. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35307 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: RIVER BEND REGION: 4 |NOTIFICATION DATE: 01/22/1999|
| UNIT: [1] [] [] STATE: LA |NOTIFICATION TIME: 18:11[EST]|
| RXTYPE: [1] GE-6 |EVENT DATE: 01/22/1999|
+------------------------------------------------+EVENT TIME: 16:17[CST]|
| NRC NOTIFIED BY: RUSS GODWIN |LAST UPDATE DATE: 01/22/1999|
| HQ OPS OFFICER: JOHN MacKINNON +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |DALE POWERS R4 |
|10 CFR SECTION: | |
|AARC 50.72(b)(1)(v) OTHER ASMT/COMM INOP | |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 N Y 95 Power Operation |95 Power Operation |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| EMERGENCY PAGING SYSTEM AND INTERNAL TELEPHONE SYSTEM ARE INOPERABLE. |
| |
| For unknown reason at this time, the emergency paging system and the plant's |
| internal telephone system became inoperable at 1617 EST. The emergency |
| paging system is used to call plant personnel to the plant in case there is |
| an emergency. The automatic telephone system is operable, and it can |
| automatically call plant personnel if they are needed. The licensee can |
| call offsite, and the ENS (emergency notification system) telephone system |
| is fully operable. The licensee said that they own the emergency paging |
| system, and they are presently troubleshooting the system. |
| |
| The NRC resident inspector will be notified of this event notification. |
| |
| *** UPDATE ON 01/22/99 AT 2001 EST FROM RICK TAKEN BY MacKINNON *** |
| |
| The emergency paging system and the plant's internal telephone system were |
| returned to service at 1840 CST. |
| |
| The NRC resident inspector was notified of this update by the licensee. The |
| R4DO (Powers) was notified by the NRC operations officer. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35308 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: POINT BEACH REGION: 3 |NOTIFICATION DATE: 01/22/1999|
| UNIT: [1] [] [] STATE: WI |NOTIFICATION TIME: 19:10[EST]|
| RXTYPE: [1] W-2-LP,[2] W-2-LP |EVENT DATE: 01/22/1999|
+------------------------------------------------+EVENT TIME: 17:59[CST]|
| NRC NOTIFIED BY: RICK ROBBINS |LAST UPDATE DATE: 01/22/1999|
| HQ OPS OFFICER: JOHN MacKINNON +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |ROGER LANKSBURY R3 |
|10 CFR SECTION: |WILLIAM BATEMAN NRR |
|ASHU 50.72(b)(1)(i)(A) PLANT S/D REQD BY TS |FRANK CONGEL IRO |
|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 |95 Power Operation |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| TECHNICAL SPECIFICATION SHUTDOWN DUE TO WESTINGHOUSE BREAKER CONCERNS |
| |
| Due to 4160-Volt Westinghouse 50-DH-350 breaker concerns on the Unit 1 |
| reactor coolant pumps, Technical Specification 15.3.5.-2.14.b and Technical |
| Specification 15.3.5-2.16, items a and b, were entered, and this requires |
| Unit 1 to be in a Hot Shutdown condition within 8 hours from 1600 CST. The |
| licensee said that this concern was part of a Ginna event sent out for |
| Westinghouse review. The licensee believes there is a laminated plate in |
| the shoot of the breaker that has a varnish which is holding the laminated |
| plates on. At this point, there is a concern about the operability of the |
| breakers. The laminated plates have been found to degrade over time, and |
| these plates can slide down and prevent the breakers from opening/closing. |
| Reactor coolant pumps, main feedwater pumps, and the '1A05' emergency bus |
| have the Westinghouse 50-DH-350 breakers. |
| |
| The only emergency operating piece of equipment out of service is the 'P38A' |
| (motor-driven auxiliary feedwater pump), which is out if service for |
| repairs. It should be back in service within the next 2 hours. The |
| electrical grid is stable, and since Unit 2 is in a refueling outage, it is |
| not affected by the breaker problem. |
| |
| The NRC resident inspector was informed of this event by the licensee. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35309 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: NINE MILE POINT REGION: 1 |NOTIFICATION DATE: 01/22/1999|
| UNIT: [] [2] [] STATE: NY |NOTIFICATION TIME: 19:40[EST]|
| RXTYPE: [1] GE-2,[2] GE-5 |EVENT DATE: 01/20/1999|
+------------------------------------------------+EVENT TIME: 17:30[EST]|
| NRC NOTIFIED BY: ROY GREEN |LAST UPDATE DATE: 01/22/1999|
| HQ OPS OFFICER: JOHN MacKINNON +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |FRANK COSTELLO R1 |
|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 |
+-----+----------+-------+--------+-----------------+--------+-----------------+
| | |
|2 N Y 100 Power Operation |100 Power Operation |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| LATE NOTIFICATION ON REACTOR CORE ISOLATION COOLING (RCIC) BEING DECLARED |
| INOPERABLE ON 01/20/99 |
| |
| On 01/20/99 at 1730 EST, RCIC became inoperable as a result of two |
| inoperable primary containment isolation valves being closed to satisfy |
| Technical Specification 3.6.3. RCIC was declared inoperable per Technical |
| Specification 3.7.4 (14 days to restore to operable status). Initial |
| assessment conclude that an immediate report was not required. However, |
| upon additional review, it has been determined that a 4-hour report is |
| required in accordance with 10 CFR 50. 72 (b)(2)(iii)(D). RCIC was restored |
| to operable status on 01/22/99 at 1142 EST. The technical specifications |
| mentioned above were exited at the time. |
| |
| During the time period RCIC was out of service, all emergency core cooling |
| systems were fully operable. |
| |
| The NRC resident inspector was notified of this event by the licensee. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35310 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: BRUNSWICK REGION: 2 |NOTIFICATION DATE: 01/23/1999|
| UNIT: [1] [] [] STATE: NC |NOTIFICATION TIME: 09:50[EST]|
| RXTYPE: [1] GE-4,[2] GE-4 |EVENT DATE: 01/23/1999|
+------------------------------------------------+EVENT TIME: 06:38[EST]|
| NRC NOTIFIED BY: DAVE JENKINS |LAST UPDATE DATE: 01/23/1999|
| HQ OPS OFFICER: LEIGH TROCINE +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |CHRIS CHRISTENSEN R2 |
|10 CFR SECTION: |BRIAN BONSER R2 |
|AESF 50.72(b)(2)(ii) ESF ACTUATION | |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 M/R Y 25 Power Operation |0 Hot Shutdown |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| PRIMARY CONTAINMENT ISOLATIONS FOLLOWING A MANUAL REACTOR SCRAM FROM 25% |
| POWER DUE TO LOWERING TEMPERATURE IN THE BOTTOM HEAD REGION OF THE REACTOR |
| PRESSURE VESSEL DURING SINGLE REACTOR RECIRCULATION LOOP OPERATION |
| |
| The following text is a portion of a facsimile received from the licensee: |
| |
| "EVENT: On January 23, 1999, at 06:38, primary containment groups 2, 6, and |
| 8 isolations were received following a manual reactor scram. The reactor |
| scram was inserted on Unit 1 due to lowering temperature in the bottom head |
| region of the reactor pressure vessel during single reactor recirculation |
| loop operation. The cooldown was augmented by recirculation pump runback to |
| 28% demand (expected action at low power due to total feedwater flow) which |
| reduced circulation through the vessel. A technical specification shutdown |
| was not required due to bottom head temperature at the time of the reactor |
| scram. Following the manual reactor scram, reactor water level lowered to |
| 160 inches. This is below the Reactor Water Level Low Level One setpoint of |
| 166 inches. This is a normal level transient following a reactor scram and |
| was anticipated by the operating crew. Although these isolations were |
| anticipated by the operating crew, they were not explicitly discussed prior |
| to the reactor scram; therefore, this report is being made in accordance |
| with 10 CFR 50.72(b)(2)(ii). All required isolations occurred as a result |
| of the Reactor Water Level Low Level One initiation signal. Reactor water |
| level immediately swelled above the Low Level One setpoint. Group 2 |
| isolation valves include drywell equipment and floor drains, traversing |
| incore probe, residual heat removal (RHR) discharge isolation to radwaste, |
| and RHR process sampling valves. Group 6 isolation valves include |
| containment atmosphere control system and post-accident monitoring valves. |
| Group 8 isolation valves include RHR system shutdown cooling isolation |
| valves; these valves were closed prior to the isolation signal." |
| |
| "INITIAL SAFETY SIGNIFICANCE EVALUATION: Minimal. All systems responded as |
| designed from the Reactor Water Level Low Level One initiation signal." |
| |
| "CORRECTIVE ACTION(S): Isolations occurred as designed; no corrective |
| actions [are] required." |
| |
| The licensee stated that Technical Specification 3.4.9, Reactor Coolant |
| System Pressure and Temperature Limits - Normal Operation With the Core |
| Critical, specifies minimum temperatures while critical. If parameters go |
| outside these references limits, this technical specification requires the |
| parameters to be restored within 30 minutes. With temperature lowering in |
| the bottom head region, the licensee chose to manually scram the reactor to |
| restore the parameter before expiration of the 30-minute limiting condition |
| for operation. |
| |
| All rods fully inserted following the manual reactor scram. There were no |
| emergency core cooling actuations or safety injections, and none were |
| expected. None of the relief valves lifted. |
| |
| The unit is currently stable in Mode 3 (Hot Shutdown). Normal feedwater is |
| being used to supply water to the reactor vessel. The main steam isolation |
| valves are open, the turbine stop and control valves are closed, and the |
| condenser is available as a heat sink. All containment parameters appear to |
| be normal. Offsite power is available, and the emergency diesel generators |
| are operable if needed. |
| |
| NOTE: Prior to this event, the unit was operating at reduced power to |
| facilitate the performance of a recirculation |
| pump motor-generator set brush replacement. |
| |
| The licensee notified the NRC resident inspector. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35311 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: BEAVER VALLEY REGION: 1 |NOTIFICATION DATE: 01/23/1999|
| UNIT: [1] [] [] STATE: PA |NOTIFICATION TIME: 12:43[EST]|
| RXTYPE: [1] W-3-LP,[2] W-3-LP |EVENT DATE: 01/23/1999|
+------------------------------------------------+EVENT TIME: 10:24[EST]|
| NRC NOTIFIED BY: TOM COTTER |LAST UPDATE DATE: 01/23/1999|
| HQ OPS OFFICER: JOHN MacKINNON +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |FRANK COSTELLO R1 |
|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 M/R Y 73 Power Operation |0 Hot Standby |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| DURING POWER REDUCTION, THE REACTOR WAS MANUALLY TRIPPED DUE TO LOW |
| CONDENSER VACUUM. |
| |
| Prior to this event, one of four condenser waterbox sections was removed |
| from service for tube cleaning, and reactor power was increased from 90% to |
| approximately 92% power. At 0916 EST, a low condenser vacuum alarm was |
| received at which time the licensee commenced a power reduction. During |
| power reduction, condenser parameters continued to degrade due to |
| circulating water system air intrusion from a leaking isolation valve. The |
| reactor was manually tripped from approximately 73% power due to the |
| degradation of the condenser system. All rods fully inserted into the |
| reactor core, and all systems operated as expected. Both the motor-driven |
| and turbine-driven auxiliary feedwater pumps automatically started on |
| low-low steam generator water level. None of the power-operated relief |
| valves on the primary or secondary side of the plant opened. Decay heat |
| from the primary system of the plant is being dumped to the main condenser. |
| (The main condenser can handle the decay heat loads.) Both the motor-driven |
| and turbine-driven feedwater pumps were secured after main feedwater was |
| restored to service. The electrical grid is stable, and all the emergency |
| core cooling systems are fully operable is needed. |
| |
| The NRC resident inspector was notified of this event by the licensee. |
+------------------------------------------------------------------------------+
Page Last Reviewed/Updated Thursday, March 25, 2021