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Event Notification Report for February 1, 2001

                    U.S. Nuclear Regulatory Commission
                              Operations Center

                              Event Reports For
                           01/31/2001 - 02/01/2001

                              ** EVENT NUMBERS **

37703  37704  37705  37706  37707  

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|General Information or Other                     |Event Number:   37703       |
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| REP ORG:  FIRSTENERGY NUCLEAR OPERATING CO.    |NOTIFICATION DATE: 01/31/2001|
|LICENSEE:  INGERSOLL-DRESSER PUMP COMPANY       |NOTIFICATION TIME: 13:09[EST]|
|    CITY:  SHIPPINGPORT             REGION:  1  |EVENT DATE:        11/27/2000|
|  COUNTY:                            STATE:  PA |EVENT TIME:        12:00[EST]|
|LICENSE#:                        AGREEMENT:  N  |LAST UPDATE DATE:  01/31/2001|
|  DOCKET:                                       |+----------------------------+
|                                                |PERSON          ORGANIZATION |
|                                                |DANIEL HOLODY        R1      |
|                                                |VERN HODGE           NRR     |
+------------------------------------------------+                             |
| NRC NOTIFIED BY:  L.W. MYERS                   |                             |
|  HQ OPS OFFICER:  DOUG WEAVER                  |                             |
+------------------------------------------------+                             |
|EMERGENCY CLASS:          N/A                   |                             |
|10 CFR SECTION:                                 |                             |
|CCCC 21.21               UNSPECIFIED PARAGRAPH  |                             |
|                                                |                             |
|                                                |                             |
|                                                |                             |
|                                                |                             |
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                                   EVENT TEXT                                   
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| PART 21 REPORT ON CAP SCREW FAILURE USED IN AN AUXILIARY FEEDWATER PUMP AT   |
| THE BEAVER VALLEY POWER STATION UNIT ONE                                     |
|                                                                              |
| One of the four cap screws on the collar of the hydraulic balancing drum of  |
| the steam-driven Auxiliary Feedwater pump (AFP) 1FW-P-2 failed.  The head of |
| the screw broke off and became lodged in the area between the stuffing box   |
| extension and the balancing drum collar, preventing 1FW-P-2 from starting on |
| 11/27/00.  The root cause of the cap screw failure was material defect.      |
| Final metallurgical analysis revealed that the failure was due to            |
| intergranular failure.  The defects noted in the fastener surface were       |
| attributed to the original manufacture of the cap screw. The probable cause  |
| of the failure was the propagation of manufacturing cracks under static      |
| preload, which caused tensile stress of approximately 88% of the yield       |
| stress of the cap screw.  Hydrogen absorption and diffusion into regions of  |
| high stress caused propagation of the cracks.  The failure was a time        |
| delayed process.                                                             |
|                                                                              |
| The material defect led to the failure of one AFP cap screw which prevented  |
| the AFP from starting. Failure of one or more Auxiliary Feedwater Pumps to   |
| start when required, would result in a major degradation of essential safety |
| related equipment, and the required Auxiliary Feedwater System may not have  |
| been able to perform its safety related function, which would constitute a   |
| substantial safety hazard.                                                   |
|                                                                              |
| Though not attributed as part of root cause for the one cap screw failure,   |
| two related noteworthy non-compliant issues were identified with the four    |
| cap screws found on the 1FW-P-2 AFP collar. An emission spectrograph test    |
| run on a cap screw showed a chromium content of 0.148% (indicating the screw |
| was carbon steel). The vendor Material Release for 1FW-P-2 (MR 912004) shows |
| that the cap screws are 410 stainless steel that should have contained 12%   |
| chromium. FENOC is not able to conclude whether operating with carbon steel  |
| cap screws (in place of the required stainless steel) could have caused the  |
| AFP to fail.                                                                 |
|                                                                              |
| The cap screws also had hardness values of 41-44 HRC (Hardness Rockwell C).  |
| The purchase specification requires 410 stainless steel with a hardness less |
| than 22 HRC. Although carbon steel bolts are less susceptible to stress      |
| corrosion cracking than stainless steel bolts, FENOC is not able to conclude |
| whether operating with carbon steel cap screws with a hardness of 41-44 HRC  |
| (in excess of the required hardness limit of 22 HRC) could have caused the   |
| AFP to fail.                                                                 |
|                                                                              |
|                                                                              |
| FENOC ALSO SUBMITTED THE FOLLOWING INFORMATION RELATED TO THE REPLACEMENT    |
| SCREWS THAT WERE ORDERED FROM FLOWSERVE CORPORATION AND MANUFACTURED BY U.S. |
| BOLT:                                                                        |
|                                                                              |
| The specified maximum hardness value was exceeded for 16 of 20 cap screws    |
| supplied for use on a balancing drum located on the Auxiliary Feedwater Pump |
| (AFP) shaft.  Exceeding the hardness limit makes these cap screws            |
| susceptible to stress corrosion cracking. Therefore, the defect, if gone     |
| undetected and installed, could have caused these cap screws to fail during  |
| their operating life.  A failed cap screw could jam and prevent a standby    |
| AFP from starting.  Failure of one or more AFPs to start when required,      |
| would result in a major degradation of essential safety related equipment,   |
| and the required Auxiliary Feedwater System may not have been able to        |
| perform its safety related function, which would constitute a significant    |
| safety hazard. As such, the defect is reportable pursuant to 10CFR Part 21   |
| requirements.                                                                |
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|Power Reactor                                    |Event Number:   37704       |
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| FACILITY: SAINT LUCIE              REGION:  2  |NOTIFICATION DATE: 01/31/2001|
|    UNIT:  [] [2] []                 STATE:  FL |NOTIFICATION TIME: 15:08[EST]|
|   RXTYPE: [1] CE,[2] CE                        |EVENT DATE:        01/09/2001|
+------------------------------------------------+EVENT TIME:        09:30[EST]|
| NRC NOTIFIED BY:  ALBRITTON                    |LAST UPDATE DATE:  01/31/2001|
|  HQ OPS OFFICER:  DOUG WEAVER                  +-----------------------------+
+------------------------------------------------+PERSON          ORGANIZATION |
|EMERGENCY CLASS:          N/A                   |LEONARD WERT         R2      |
|10 CFR SECTION:                                 |                             |
|*IND 50.72(b)(3)(v)(D)   ACCIDENT MITIGATION    |                             |
|                                                |                             |
|                                                |                             |
|                                                |                             |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR|   INIT RX MODE  |CURR PWR|  CURR RX MODE   |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|                                                   |                          |
|2     N          Y       100      Power Operation  |100      Power Operation  |
|                                                   |                          |
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                                   EVENT TEXT                                   
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| POTENTIAL FOR THE CONTROL ROOM OPERATORS TO EXCEED GENERAL DESIGN CRITERIA   |
| 19 LIMITS DURING AN ACCIDENT                                                 |
|                                                                              |
| Following evaluation and analysis, FPL identified a condition that could     |
| have potentially led to the control room operators receiving a dose in       |
| excess of GDC 19 limits during accident conditions.  St. Lucie recently      |
| identified weaknesses in the Unit 2 procedures that could result in the      |
| operators not taking action to manually align the control room outside air   |
| intakes to pressurize the control room when the control room ventilation     |
| system is operating in the emergency recirculation mode.  The procedures     |
| direct which outside air intake should be used depending on existing         |
| radiological conditions. However, the procedures do not direct the operators |
| to establish outside makeup air (i.e.. throttle open the outside air intake  |
| valves) if minimum control room pressurization is met.   In addition to the  |
| procedural weaknesses identified, the control room differential pressure     |
| indicators exhibit a small positive pressure bias that could have misled the |
| operators into believing that the control room was maintained at a positive  |
| differential pressure without the need to open the outside air intake        |
| valves.   During the course of the hypothesized accident, operation of the   |
| control room ventilation system in this condition would have been ultimately |
| discovered and corrected by the technical advisors in the Technical Support  |
| Center or Emergency Operations Facility.                                     |
|                                                                              |
| St. Lucie issued Night Orders to the operating crews to clarify operational  |
| requirements for the control room ventilation systems until the requisite    |
| procedure changes are implemented.                                           |
|                                                                              |
| The licensee notified the NRC resident inspector.                            |
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|Power Reactor                                    |Event Number:   37705       |
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| FACILITY: RIVER BEND               REGION:  4  |NOTIFICATION DATE: 01/31/2001|
|    UNIT:  [1] [] []                 STATE:  LA |NOTIFICATION TIME: 17:24[EST]|
|   RXTYPE: [1] GE-6                             |EVENT DATE:        01/31/2001|
+------------------------------------------------+EVENT TIME:        14:00[CST]|
| NRC NOTIFIED BY:  RUSS WALTON                  |LAST UPDATE DATE:  01/31/2001|
|  HQ OPS OFFICER:  DOUG WEAVER                  +-----------------------------+
+------------------------------------------------+PERSON          ORGANIZATION |
|EMERGENCY CLASS:          N/A                   |BLAIR SPITZBERG      R4      |
|10 CFR SECTION:                                 |                             |
|*PRE 50.72(b)(2)(xi)     OFFSITE NOTIFICATION   |                             |
|                                                |                             |
|                                                |                             |
|                                                |                             |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR|   INIT RX MODE  |CURR PWR|  CURR RX MODE   |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1     N          Y       100      Power Operation  |100      Power Operation  |
|                                                   |                          |
|                                                   |                          |
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                                   EVENT TEXT                                   
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| OFFSITE NOTIFICATION -  SEWAGE SPILL                                         |
|                                                                              |
| At approximately 1400 CST on 01/31/01, an untreated sewage spill,            |
| overflowing from an underground manhole cover, of approximately 200 gallons  |
| occurred outside of the Generation Support Building.  The sewage spilled     |
| into the storm drains which drain into East Creek.  The spill has been       |
| isolated and contained.  The spill occurred outside the Protected Area but   |
| remained inside the Owner Controlled Area during the event with no spillage  |
| reaching the Mississippi River.  The Louisiana Department of Environmental   |
| Quality was notified by River Bend personnel at                              |
| 1451.                                                                        |
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|Power Reactor                                    |Event Number:   37706       |
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| FACILITY: COOK                     REGION:  3  |NOTIFICATION DATE: 01/31/2001|
|    UNIT:  [] [2] []                 STATE:  MI |NOTIFICATION TIME: 18:48[EST]|
|   RXTYPE: [1] W-4-LP,[2] W-4-LP                |EVENT DATE:        01/31/2001|
+------------------------------------------------+EVENT TIME:        11:45[EST]|
| NRC NOTIFIED BY:  LEE JOHNSON                  |LAST UPDATE DATE:  01/31/2001|
|  HQ OPS OFFICER:  DOUG WEAVER                  +-----------------------------+
+------------------------------------------------+PERSON          ORGANIZATION |
|EMERGENCY CLASS:          N/A                   |DAVID HILLS          R3      |
|10 CFR SECTION:                                 |                             |
|*INC 50.72(b)(3)(v)(C)   POT UNCNTRL RAD REL    |                             |
|                                                |                             |
|                                                |                             |
|                                                |                             |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR|   INIT RX MODE  |CURR PWR|  CURR RX MODE   |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|                                                   |                          |
|2     N          Y       100      Power Operation  |100      Power Operation  |
|                                                   |                          |
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                                   EVENT TEXT                                   
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| POTENTIAL INABILITY TO CONTROL A RADIOACTIVE RELEASE                         |
|                                                                              |
| At 1145 on 1/31/01, the door to the auxiliary engineered safeguards suction  |
| side vestibule in unit 2 was found stuck open.  This bypasses the normal     |
| flowpath for both trains of ESF fans and renders them inoperable.  The plant |
| made an unrecognized entry into Technical Specification 3.0.3 until the door |
| could be closed, which occurred in approximately one minute.  This eight     |
| hour report is being made in accordance with 10 CFR 50.72 based on the       |
| temporary inability to control a possible radioactive release.               |
|                                                                              |
| The licensee notified the NRC resident inspector.                            |
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|Power Reactor                                    |Event Number:   37707       |
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| FACILITY: LASALLE                  REGION:  3  |NOTIFICATION DATE: 02/01/2001|
|    UNIT:  [1] [] []                 STATE:  IL |NOTIFICATION TIME: 01:08[EST]|
|   RXTYPE: [1] GE-5,[2] GE-5                    |EVENT DATE:        01/31/2001|
+------------------------------------------------+EVENT TIME:        21:47[CST]|
| NRC NOTIFIED BY:  SHANE MARIK                  |LAST UPDATE DATE:  02/01/2001|
|  HQ OPS OFFICER:  STEVE SANDIN                 +-----------------------------+
+------------------------------------------------+PERSON          ORGANIZATION |
|EMERGENCY CLASS:          N/A                   |DAVID HILLS          R3      |
|10 CFR SECTION:                                 |                             |
|*RPS 50.72(b)(2)(iv)(B)  RPS ACTUATION - CRITICA|                             |
|                                                |                             |
|                                                |                             |
|                                                |                             |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR|   INIT RX MODE  |CURR PWR|  CURR RX MODE   |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1     A/R        Y       100      Power Operation  |0        Hot Shutdown     |
|                                                   |                          |
|                                                   |                          |
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                                   EVENT TEXT                                   
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| UNIT 1 EXPERIENCED AN AUTOMATIC REACTOR SCRAM FOLLOWING FAILURE OF A MAIN    |
| POWER TRANSFORMER                                                            |
|                                                                              |
| "At 21:47 CST, U-1 automatically scrammed from a main turbine 'NON-EHC' trip |
| caused from a failure of a main power transformer. The main power            |
| transformers received an auto deluge signal and an acrid smell is reported   |
| in the area. The main generator tripped from the loss of the main power      |
| transformer causing the main turbine to trip, which caused an automatic      |
| reactor scram. The fast closure of the main turbine valves caused a reactor  |
| pressure spike which tripped both reactor recirculation pumps and caused two |
| safety relief valves to actuate.                                             |
|                                                                              |
| "All automatic actions initiated as designed, but the following anomalies    |
| were noted;                                                                  |
|                                                                              |
| - 1A circulating water pump tripped                                          |
| - Division 1 alternate rod insertion failed to reset on scram recovery       |
| - 1B recirculation pump received a low oil level alarm on restart attempt    |
| - U2 received an electrical perturbation from the U1 scram which resulted in |
| a loss of the 2A heater drain pump and two heaters.  Cram rods were inserted |
| in accordance with Operating procedures. U2 was stabilized at 930 MWE."      |
|                                                                              |
| All rods fully inserted.  The two safety relief valves reseated after        |
| actuation.  Decay heat is currently being removed via the bypass valves to   |
| the main condenser.  RCIC is inoperable but available, if needed. There are  |
| no challenges to offsite power and the system auxiliary transformer is fully |
| available.   The licensee is presently resetting the deluge system in order  |
| to assess if there is mechanical damage on the 1 west main power transformer |
| and will determine whether a U-1 cooldown is required to evaluate the 1B     |
| recirculation pump problem.  The NRC resident inspector was informed and is  |
| currently onsite.                                                            |
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