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
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 37703 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| 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 |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| 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. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 37707 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| 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
+------------------------------------------------------------------------------+
| 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. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 37708 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| 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. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|General Information or Other |Event Number: 37709 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| 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 | |
| | |
| | |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| 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." |
+------------------------------------------------------------------------------+
Page Last Reviewed/Updated Thursday, March 25, 2021
Page Last Reviewed/Updated Thursday, March 25, 2021