| U.S. Nuclear Regulatory Commission
Operations Center Event Reports For
05/06/2004 - 05/07/2004 ** EVENT NUMBERS ** |
 | !!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!! | | Power Reactor | Event Number: 40636 | Facility: BRUNSWICK
Region: 2 State: NC
Unit: [1] [ ] [ ]
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: WILLIAM WOODBURY
HQ OPS Officer: MIKE RIPLEY | Notification Date: 04/03/2004
Notification Time: 05:49 [ET]
Event Date: 04/03/2004
Event Time: 03:00 [EST]
Last Update Date: 05/06/2004 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
| Person (Organization):
JOEL MUNDAY (R2)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | Y | 22 | Power Operation | 22 | Power Operation | Event Text HIGH PRESSURE COOLANT INJECTION SYSTEM INOPERABLE
"During plant startup from the Brunswick Unit 1 Refueling Outage, the High Pressure Coolant Injection (HPCI) System was declared inoperable due to pump data not meeting the acceptable value of OPT-09.2, HPCI System Operability Test. With pump flow established at the reference value of 4550 gallons per minute, the actual turbine speed was 2530 rpm, outside the acceptable range of 2485-2515 rpm.
"Initial Safety Significance Evaluation: Minimal since the Automatic Depressurization System (ADS), the Low Pressure Coolant Injection System (LPCI), and the Core Spray System are Operable. In addition the Reactor Core Isolation Cooling System (RCIC) is Operable.
"Corrective Actions: Engineering to evaluate the test results and determine corrective actions."
The licensee will notify the NRC Resident Inspector.
* * * RETRACTION TAKEN ON 5/6/04 @ 1534 EDT BY CROUCH FROM ELBERFELD * * *
The following information was obtained from the licensee via facsimile:
"The purpose of this call is to retract the notification (Event Number 40636) made by the Brunswick Steam Electric Plant, Unit No, 1, Docket No. 50-325 / License No. DPR-71, on April 3, 2004, at 0549 hours.
"On April 2, 2004, at 2324 hours, the High Pressure Coolant Injection (HPCI) system was declared inoperable for scheduled testing performed in accordance with 0PT-09.2, 'HPCI System Operability Test.' 0PT-09.2 includes, in part, Technical Specification (TS) Surveillance Requirement (SR) 3.5.1.7, to verify that, with reactor pressure < 1045 and > 945 psig, the HPCI pump unit could develop a flow rate > 4250 gpm against a system head corresponding to reactor pressure. The testing also includes in part, In-service Test (IST) program requirements for HPCI pump performance (to test for pump degradation). During this testing, water is circulated from the Condensate Storage Tank (CST) through the HPCI pump and back to the CST through a 'Bypass to CST' valve. Motive force for the pump is provided by reactor steam being routed through the HPCI turbine into the suppression pool. Due to the heat energy being deposited in the suppression pool, time for HPCI turbine/pump testing is limited.
"When tested, the HPCI pump passed the TS SR 3.5.1.7 pump pressure and flow rate test requirements. However, when HPCI turbine speed was set to the IST-required value of 2,500 rpm, as indicated by the control room indicator 1-E41-C002-4 (the preferred indication), the IST required system flow of 4,550 gpm could not be attained. Turbine speed was then adjusted, using a portable speed indicator, to attain 4,550 gpm system flow. The portable speed indicator read 2530 rpm, high, outside the acceptable range of 2,485 to 2,515 rpm. The unsatisfactory test indicated potential pump degradation. The IST program requires the equipment to be declared inoperable until unsatisfactory test results are evaluated. On April 3, 2004, at 0436 hours, the HPCI system was declared inoperable due to the unsatisfactory test. At 0549 hours, the NRC was notified (Event Number 40636) in accordance with 10 CFR 50.72(b)(3)(v)(D) as a condition that at the time of discovery could have prevented the fulfillment of the safety function of a system needed to mitigate the consequences of an accident.
"During the testing with the portable speed indicator, it was noted that the control room indicator 1-E41-C002-4 was reading high (i.e., 1-E41-C002-4 was reading 2,600 rpm compared to 2,500 rpm on the portable speed indicator). Turbine speed indicator 1-E41-C002-4 was adjusted back to within required accuracy in accordance with Work Order 306977.
"After the initial test, a thorough review of pump data, instrument calibration data, and IST program guidance was conducted. It was determined that there was no indication of pump degradation and no indication of a flow indicator problem. Having corrected a speed indicator problem, sections of 0PT-09.2 were performed on April 5, 2004, to test for pump degradation using the preferred speed indicator 1-E41-C002-4. Test results were satisfactory and demonstrated that the pump performance met the IST program requirements. The HPCI system was declared operable on April 5, 2004, at approximately 0521 hours. Investigation of this condition is documented in the corrective action program in Action Request (AR) 123551.
"NUREG-1022, Rev. 2, Section 3.2.7 (page 56) lists types of events or conditions that are generally not reportable under 10 CFR 50.72(b)(3)(v) and 10 CFR 50.73 (a)(2)(v) criteria. The list of not-reportable conditions includes:
"- Removal of a system or part of a system from service as part of a planned evolution for maintenance or surveillance testing when done in accordance with an approved procedure and the plant's TS (unless a condition is discovered that could have prevented the system from performing its function)
"On April 2, 2004, the HPCI system was removed from service for testing in accordance with an approved procedure 0PT-09.2. The system was declared inoperable due to the test results discussed above, and since the HPCI system is a single train safety system, an ENS notification was made. However, no condition was discovered that could have prevented the HPCI system from performing its functions.
"Carolina Power & Light Company, doing business as Progress Energy Carolinas, Inc., has determined that this event does not meet the 10 CFR 50.72 or 10 CFR 50.73 reporting criteria and the notification for Event Number 40636 is retracted. The resident inspector has been notified." |
| Power Reactor | Event Number: 40728 | Facility: POINT BEACH
Region: 3 State: WI
Unit: [1] [ ] [ ]
RX Type: [1] W-2-LP,[2] W-2-LP
NRC Notified By: JIM FOUSE
HQ OPS Officer: MIKE RIPLEY | Notification Date: 05/06/2004
Notification Time: 13:30 [ET]
Event Date: 05/06/2004
Event Time: 10:46 [CDT]
Last Update Date: 05/06/2004 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(A) - DEGRADED CONDITION
| Person (Organization):
CHRISTINE LIPA (R3)
TERRY REIS (NRR)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | N | 0 | Refueling | 0 | Refueling | Event Text REACTOR VESSEL HEAD NOZZLE WELD AREA FLAW
"During performance of NDE examinations of the Point Beach Nuclear Plant (PBNP) Unit 1 reactor pressure vessel (RPV) head required by the First Revised NRC Order (EA-03-009), flaw indications were identified on Nozzle 26. The ultrasonic examination (UT) signal for nozzle 26 identified flaw indications in the J-groove weld area that extend into the CRDM tube base material.
"A dye-penetrant (PT) examination of the nozzle 26 CRDM J-groove material was also performed. The PT exams showed minor surface indications that required further evaluation. Following minor excavation of the weld surface, additional examinations of the J-groove surface were performed. The results of these exams indicated the existence of flaws in the weld that do not meet accepted flaw evaluation guidance.
"Based upon preliminary analysis, it is expected that these indications would not be found acceptable under ASME standards. Therefore, this condition represents degradation of a principal safety barrier reportable under 10 CFR 50.72(b)(3)(ii)(A).
"PBNP Unit 1 Nozzle 26 is planned to be repaired prior to placing the RPV Head back into service.
"The condition was determined to be reportable at 1046 CDT on May 6, 2004. The licensee has notified the Resident Inspectors." |
| Other Nuclear Material | Event Number: 40729 | Rep Org: VA NATIONAL HEALTH PHYSICS PROGRAM
Licensee: CLEMENT J. ZABLOCKI VA MEDICAL CENTER
Region: 3
City: MILWAUKEE State: WI
County:
License #: 03-23853-01VA
Agreement: Y
Docket:
NRC Notified By: GARY WILLIAMS
HQ OPS Officer: HOWIE CROUCH | Notification Date: 05/06/2004
Notification Time: 14:47 [ET]
Event Date: 05/06/2004
Event Time: [CDT]
Last Update Date: 05/06/2004 | Emergency Class: NON EMERGENCY
10 CFR Section:
20.2201(a)(1)(i) - LOST/STOLEN LNM>1000X
| Person (Organization):
CHRISTINE LIPA (R3)
LAWRENCE KOKAJKO (NMSS)
| Event Text LOSS OF FOUR TRITIUM EXIT SIGNS
The following information was obtained from the Master Materials License Holder via facsimile:
"The loss occurred at a medical permittee authorized under the master materials license issued to the Department of Veterans Affairs, NRC License 03-23853 01 VA. The permittee is the Clement T. Zablocki VA Medical Center, Milwaukee, Wisconsin.
"The loss occurred circa 2001 and was discovered on May 6, 2004. The basis for the report is under 10 CFR 31.5(c)(10) in that radioactive materials obtained under a general license were lost and the reporting requirements in 10 CFR 20.2201 must be followed.
"Specifically, the permittee stated four tritium exit signs were apparently disposed or discarded during a facility construction project circa 2001. The signs were last seen in 2000.
"The signs were either of two models. 'The first model is a 11.5 Curie sign from Shield Source Incorporated. The second model is a 20 Curie sign from SRB Technology. The signs were manufactured circa 1994.
"The Department of Veterans Affairs will evaluate the circumstances related to the loss of radioactive materials and submit a written report to NRC, Region III, within 30 days." |
| Power Reactor | Event Number: 40730 | Facility: HARRIS
Region: 2 State: NC
Unit: [1] [ ] [ ]
RX Type: [1] W-3-LP
NRC Notified By: JOHN YADUSKY
HQ OPS Officer: HOWIE CROUCH | Notification Date: 05/06/2004
Notification Time: 16:25 [ET]
Event Date: 05/06/2004
Event Time: 12:52 [EDT]
Last Update Date: 05/06/2004 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
| Person (Organization):
KERRY LANDIS (R2)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | A/R | Y | 100 | Power Operation | 0 | Hot Standby | Event Text AUTOMATIC REACTOR TRIP AND ACTUATION OF AUXILIARY FEEDWATER SYSTEM
The following information was received from the licensee via facsimile:
"On May 6, 2004, with the reactor at 100 percent power in MODE 1, an unplanned actuation of the reactor protection system occurred. At 1252 [EDT] the reactor was automatically tripped from a power range negative flux rate trip signal. The auxiliary feedwater system actuated as expected to stabilize steam generator levels. All systems functioned as required and no other safety systems were actuated. All control rods inserted on the reactor trip. The operations staff responded to the event in accordance with applicable plant procedures. The plant stabilized at normal operating no-load reactor coolant system temperature and pressure following the reactor trip. Steam generator water levels are being maintained using normal main feedwater. All emergency core cooling system equipment is available. The plant electrical system is available and in a normal configuration. The cause of the plant trip is under investigation.
"This condition is being reported as an unplanned reactor protection system actuation and specified system actuation in accordance with 10 CFR 50.72(b)(2)(iv)(B) and10 CFR 50.72(b)(3)(iv)(A) ."
During the transient, a steam generator power-operated relief valve lifted momentarily and then re-seated. No reportable radiological release occurred during the event.
The licensee notified the NRC Resident Inspector. |
| Other Nuclear Material | Event Number: 40731 | Rep Org: MALLINCKRODT
Licensee: MALLINCKRODT
Region: 3
City: ST. LOUIS State: MO
County:
License #: 24-04-206-01
Agreement: N
Docket:
NRC Notified By: RULAND SAWYER
HQ OPS Officer: MIKE RIPLEY | Notification Date: 05/06/2004
Notification Time: 16:45 [ET]
Event Date: 04/21/2004
Event Time: [CDT]
Last Update Date: 05/06/2004 | Emergency Class: NON EMERGENCY
10 CFR Section:
20.2201(a)(1)(ii) - LOST/STOLEN LNM>10X
| Person (Organization):
CHRISTINE LIPA (R3)
LAWRENCE KOKAJKO (NMSS)
| Event Text MISSING DEPLETED URANIUM SHIPPING SHIELD
The licensee reported that an empty 45.7 lb. depleted uranium shipping shield could not be located. The shield was last reported as being shipped back to St. Louis from Mallinckrodt's Beltsville, MD facility on 5/10/03. After an investigation as to the possible location of the shield conducted in early 2004, it was declared lost on 4/21/04. The licensee has contacted NRC Region 3. | |