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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 THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED  !!!!!!!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.       |
+------------------------------------------------------------------------------+