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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| | |
| | |
| | |
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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 | |
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+-----+----------+-------+--------+-----------------+--------+-----------------+
|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 | |
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+-----+----------+-------+--------+-----------------+--------+-----------------+
|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 | |
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+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
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|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| |
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+-----+----------+-------+--------+-----------------+--------+-----------------+
|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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