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Event Notification Report for February 2, 2001

                    U.S. Nuclear Regulatory Commission
                              Operations Center

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

                              ** EVENT NUMBERS **

37703  37707  37708  37709  

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|Power Reactor                                    |Event Number:   37703       |
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| FACILITY: BEAVER VALLEY            REGION:  1  |NOTIFICATION DATE: 01/31/2001|
|    UNIT:  [1] [] []                 STATE:  PA |NOTIFICATION TIME: 13:09[EST]|
|   RXTYPE: [1] W-3-LP,[2] W-3-LP                |EVENT DATE:        11/27/2000|
+------------------------------------------------+EVENT TIME:        12:00[EST]|
| NRC NOTIFIED BY:  L. W. MYERS                  |LAST UPDATE DATE:  02/01/2001|
|  HQ OPS OFFICER:  DOUG WEAVER                  +-----------------------------+
+------------------------------------------------+PERSON          ORGANIZATION |
|EMERGENCY CLASS:                                |DANIEL HOLODY        R1      |
|10 CFR SECTION:                                 |VERN HODGE           NRR     |
|CCCC 21.21               UNSPECIFIED PARAGRAPH  |                             |
|                                                |                             |
|                                                |                             |
|                                                |                             |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|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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| 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.                                                                 |
|                                                                              |
|                                                                              |
| THE LICENSEE 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 10 CFR Part 21  |
| requirements.                                                                |
|                                                                              |
| HOO NOTE:  This report was modified to identify the licensee rather than the |
| reporting organization (FirstEnergy Nuclear Operating Company) which is the  |
| owner/operator of Beaver Valley.                                             |
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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     |
|                                                   |                          |
|                                                   |                          |
+------------------------------------------------------------------------------+
                                   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.                                                            |
|                                                                              |
| * * * UPDATE AT 2007 EST ON 2/1/01 BY SHANE MARIK TO FANGIE JONES * * *      |
|                                                                              |
| "This is a follow up notification to event #37707 to enhance and clarify     |
| plant response following the post scram investigation. It was determined     |
| that a bushing/insulator failure on the 'C' phase of the 1 West Main Power   |
| Transformer failed causing the lockout of the main generator. The failed     |
| bushing/insulator is not located directly on the 1 West Main Power           |
| transformer but is located on the first main tower between the transformer   |
| and the switchyard.                                                          |
|                                                                              |
| "During the turbine trip and reactor scram the reactor vessel level          |
| instrumentation spiked causing a 'ringing phenomenon' initiated from the     |
| increase in pressure. This phenomenon was identified from the transient      |
| analyses data and seen during previous pressure transients. The ringing in   |
| the level instrumentation caused varying level indication (<1/2 second       |
| cycles) which is indication only, not a real change in reactor level. This   |
| ringing phenomenon caused to the actuations and 1/2 isolations identified    |
| during the scram.                                                            |
|                                                                              |
| "The following is offered to clarify the anomalies identified during the     |
| event.                                                                       |
| 1.  The main steam isolation valves received a 1/2 group one isolation due   |
| to the ringing phenomenon.                                                   |
| 2.  Reactor recirculation pumps tripped off due to the ringing phenomenon.   |
| 3.  Four safety relief valves opened. Previously reported as two. All four   |
| safety relief valves re-closed properly.                                     |
| 4.  Reactor building ventilation tripped due to the inboard isolation        |
| dampers closing on low voltage transient.                                    |
| 5.  The main generator voltage regulator failed to auto transfer to manual.  |
| 6.  A division one ground was received and was subsequently isolated to      |
| three alarm points associated with the main power transformers which         |
| received a deluge on the bushing fault."                                     |
|                                                                              |
| The licensee notified the NRC Resident Inspector.  The R3DO (David Hills)    |
| has been notified.                                                           |
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|Power Reactor                                    |Event Number:   37708       |
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| FACILITY: SUSQUEHANNA              REGION:  1  |NOTIFICATION DATE: 02/01/2001|
|    UNIT:  [] [2] []                 STATE:  PA |NOTIFICATION TIME: 11:12[EST]|
|   RXTYPE: [1] GE-4,[2] GE-4                    |EVENT DATE:        02/01/2001|
+------------------------------------------------+EVENT TIME:        07:30[EST]|
| NRC NOTIFIED BY:  DAVID T. WALSH               |LAST UPDATE DATE:  02/01/2001|
|  HQ OPS OFFICER:  LEIGH TROCINE                +-----------------------------+
+------------------------------------------------+PERSON          ORGANIZATION |
|EMERGENCY CLASS:          N/A                   |DANIEL HOLODY        R1      |
|10 CFR SECTION:                                 |                             |
|*DEG 50.72(b)(3)(ii)(A)  DEGRAD COND DURING OP  |                             |
|                                                |                             |
|                                                |                             |
|                                                |                             |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|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 THE MAIN STEAM ISOLATION VALVE (MSIV) MAX PATH LIMIT COULD    |
| HAVE BEEN EXCEEDED DUE TO INABILITY TO DEPRESSURIZE THE STEAM LINE           |
|                                                                              |
| The following text is a portion of a facsimile received from the licensee:   |
|                                                                              |
| "[The] Unit 2 [Reactor Core Isolation Cooling] System was removed from       |
| service to perform scheduled maintenance.  Control Room operators closed the |
| inboard containment isolation valve, HV249F007, to depressurize the steam    |
| line.  The steam line did not depressurize as expected.  Operators then      |
| closed the outboard containment isolation valve, HV249F008, and were         |
| successful in isolating the pathway and depressurizing the steam piping.  At |
| 07:30, the inboard isolation valves were declared inoperable.  Due to the    |
| inability to depressurize the steam line and following a technical review of |
| the system response and supporting data, it appears that the MSIV max path   |
| limit could have been exceeded, and therefore, this event is reportable      |
| under 10CFR50.72(b)(3)(ii), Degraded or Unanalyzed Condition."               |
|                                                                              |
| The licensee notified the NRC resident inspector.                            |
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|General Information or Other                     |Event Number:   37709       |
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| REP ORG:  ROSEMOUNT NUCLEAR INSTRUMENTS, INC.  |NOTIFICATION DATE: 02/01/2001|
|LICENSEE:  ROSEMOUNT NUCLEAR INSTRUMENTS, INC.  |NOTIFICATION TIME: 13:11[EST]|
|    CITY:  EDEN PRAIRIE             REGION:  3  |EVENT DATE:        02/01/2001|
|  COUNTY:                            STATE:  MN |EVENT TIME:             [CST]|
|LICENSE#:                        AGREEMENT:  N  |LAST UPDATE DATE:  02/01/2001|
|  DOCKET:                                       |+----------------------------+
|                                                |PERSON          ORGANIZATION |
|                                                |DAVID HILLS          R3      |
|                                                |VERN HODGE           NRR     |
+------------------------------------------------+                             |
| NRC NOTIFIED BY:  JEFFREY W. SCHMITT           |                             |
|  HQ OPS OFFICER:  FANGIE JONES                 |                             |
+------------------------------------------------+                             |
|EMERGENCY CLASS:          N/A                   |                             |
|10 CFR SECTION:                                 |                             |
|CCCC 21.21               UNSPECIFIED PARAGRAPH  |                             |
|                                                |                             |
|                                                |                             |
|                                                |                             |
|                                                |                             |
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                                   EVENT TEXT                                   
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| 10 CFR PART 21 REPORT FOR MODEL 353C AND 353C1 CONDUIT SEALS                 |
|                                                                              |
| "This notification relates to Model 353C and 353C1 conduit seals which       |
| exhibit an electrical short condition. Any model 353C and 353C1 conduit seal |
| which does not indicate a short condition is deemed acceptable and this      |
| notification is not applicable to such units.                                |
|                                                                              |
| "This notification is not applicable to units which are currently            |
| operational, or have successfully completed a functional test by the         |
| customer that would verify the insulation resistance of the lead wires.      |
|                                                                              |
| "RNII [Rosemount Nuclear Instruments, Inc.] does not have sufficient         |
| information to determine the safety impact related to plant applications.    |
| Licensees must determine the impact on plant operations and plant safety and |
| take action as deemed necessary."                                            |
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