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Event Notification Report for November 1, 2010

U.S. Nuclear Regulatory Commission
Operations Center

Event Reports For
10/29/2010 - 11/01/2010

** EVENT NUMBERS **


46327 46375 46377 46378 46379

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!!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!!
Power Reactor Event Number: 46327
Facility: MONTICELLO
Region: 3 State: MN
Unit: [1] [ ] [ ]
RX Type: [1] GE-3
NRC Notified By: MARK PIKUS
HQ OPS Officer: CHARLES TEAL
Notification Date: 10/11/2010
Notification Time: 22:05 [ET]
Event Date: 10/11/2010
Event Time: 13:05 [CDT]
Last Update Date: 10/29/2010
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
Person (Organization):
DAVE PASSEHL (R3DO)

Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
1 N Y 95 Power Operation 95 Power Operation

Event Text

TECHNICAL SPECIFICATION DOES NOT ACCOUNT FOR POWER UPRATE

"On October 11, 2010 at 1305 CDT it was identified that the analysis of record for the Technical Specification 3.3.5.1, Table 3.3.5.1-1 function 1e and 2e, Reactor Steam Dome Pressure Permissive-Bypass timer (Pump Permissive) did not reflect current plant conditions. Specifically, the analysis was not updated to account for any increase in plant licensed power and a change to the RWCU (Reactor Water Cleanup System) isolation for enhanced ability to isolate RWCU on a line break on critical crack. The allowable value for these function is greater than or equal to 18 minutes and less than or equal to 22 minutes.

"All equipment associated with emergency core cooling function are unaffected. Discussion with General Electric indicates that a margin exists to accommodate the higher power level. Additionally, the changes to the RWCU isolation logic added leak detection instruments that will isolate RWCU earlier for the majority of pipe leaks. This discovery is being reported as an unanalyzed condition solely due to the lack of a formal analysis of current plant conditions."

The NRC Resident Inspector has been notified.


* * * RETRACTION FROM MARTIN RAJKOWSKI TO JOHN KNOKE AT 1451 ON 10/29/10 * * *

"Under NUREG-0737, Item II.K.3.18 is a regulatory requirement to implement a modification to extend the ADS [Automatic Depressurization System] to a unique event sequence that involves multiple failures including HPCI [High Pressure Coolant Injection] plus no operator action after 10 minutes. According to Item II.K.3.18, the bypass timer logic complements, but does not replace, the existing ADS actuation logic. This requirement is not associated with any design basis accident mitigation sequence at MNGP [Monticello Nuclear Generating Plant].

"The plant has performed an evaluation that addresses the changes in plant thermal power and RWCU (Reactor Water Cleanup System) enhanced isolation capabilities to the analysis of record. This evaluation concluded that Peak Clad Temperature (PCT) will remain under the 2200 ?F acceptance limit with the current Technical Specification allowable value, current plant configuration and current licensed thermal power.

GEH [General Electric] did an independent evaluation that concluded that based on use of limiting scenarios analyzed for MNGP, this condition is only a lack of formal analysis of current plant conditions and that no Substantial Safety Hazard exists. It is judged that the maximum Reactor Steam Dome Pressure Permissive-Bypass timer (Pump Permissive) setting of 22 minutes will not result in a predicted PCT higher than 2200?F with consideration of the RWCU pipe break isolation instrumentation modification.

"Since the timers with their current setpoint will protect the fuel cladding, this event does not significantly degrade safety. Therefore, the event notification made on October 11, 2010 is being retracted."

The licensee has notified the NRC Resident Inspector. Notified R3DO (Skokowski)

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Power Reactor Event Number: 46375
Facility: OCONEE
Region: 2 State: SC
Unit: [1] [2] [3]
RX Type: [1] B&W-L-LP,[2] B&W-L-LP,[3] B&W-L-LP
NRC Notified By: SANDRA SEVERANCE
HQ OPS Officer: BILL HUFFMAN
Notification Date: 10/29/2010
Notification Time: 09:39 [ET]
Event Date: 10/28/2010
Event Time: 12:40 [EDT]
Last Update Date: 10/29/2010
Emergency Class: NON EMERGENCY
10 CFR Section:
21.21 - UNSPECIFIED PARAGRAPH
Person (Organization):
REBECCA NEASE (R2DO)
PT 21 GROUP E-MAIL ()

Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
1 N Y 100 Power Operation 100 Power Operation
2 N Y 100 Power Operation 100 Power Operation
3 N N 0 Refueling 0 Refueling

Event Text

DEFECT DISCOVERED IN TUBE STEEL THAT COULD BE USED IN VARIOUS SAFETY RELATED STRUCTUAL APPLICATIONS

"On October 28, 2010, Duke Energy completed a reportability determination which concluded that a defect associated with four inch by four inch tube steel with one-half inch wall thickness is reportable under Part 21. The tube steel was procured safety-related from Mackson, Inc. on May 26, 2010. Receipt of the material occurred on June 14, 2010. During construction of the Protected Service Water (PSW) ductbank elevated cable raceway, Craft reported a longitudinal crack in the tube steel, approximately four feet in length, adjacent to a raceway fabrication weld. The crack was located in the manufacturer's longitudinal seam weld in the tube steel. Follow-up investigation and laboratory evaluation revealed that the structural steel tubing in question contains surface breaking flaws located along the centerline of the seam weld which are attributable to lack of fusion that occurred during tubing manufacture. Additional testing of samples from the same heat of material indicated that the seam weld flaw depth varied with some localized areas reaching depths of at least 40 percent through the wall thickness prior to raceway fabrication welding. According to documents received from the supplier, during dedication, the supplier performed chemical, physical and 100 percent visual exam in accordance with their accepted dedication procedures for ASTM A500 for Grade B material. However, the supplied product did not conform to the requirements of ASTM A500 in that the longitudinal butt joint was not welded across its thickness (Reference ASTM A500, Section 6.2). Duke Energy will provide follow-up written notification within 30 days pursuant to Part 21.21(d)(3)(ii).

"Initial Safety Significance: None. The defective tube steel utilized in the PSW structure was not placed into service. Tube steel sections of the same heat of material not used in pre-fabrication efforts were scrapped. Those installed were cut out or evaluated for acceptability by Engineering. The failure of this weld significantly impairs the structural properties of the hollow structural section. The generic implications associated with the potential to use these structural members in various nuclear safety-related applications at this site and at other stations results in a substantial safety hazard were it to remain uncorrected.

"Corrective Action(s):
1. Notified supplier, Mackson, Inc.
2. Re-worked all uses of the defective structural tube steel.
3. Developed additional, required testing for safety -related tube steel."

The supplier (Mackson, Inc) indicated to Oconee that no other nuclear power plants have received this type of tube steel from Mackson. Oconee also has concluded that the condition is confined to only one heat of the tubing used onsite. All the tube steel of this heat has been either disposed of, removed, or verified acceptable.

The licensee has notified the NRC Resident Inspector.

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Power Reactor Event Number: 46377
Facility: NORTH ANNA
Region: 2 State: VA
Unit: [1] [2] [ ]
RX Type: [1] W-3-LP,[2] W-3-LP
NRC Notified By: PAGE KEMP
HQ OPS Officer: DONG HWA PARK
Notification Date: 10/29/2010
Notification Time: 14:11 [ET]
Event Date: 10/29/2010
Event Time: 14:00 [EDT]
Last Update Date: 10/29/2010
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(xi) - OFFSITE NOTIFICATION
Person (Organization):
REBECCA NEASE (R2DO)

Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
1 N Y 7 Power Operation 7 Power Operation
2 N Y 100 Power Operation 100 Power Operation

Event Text

OFFSITE NOTIFICATION DUE TO INCREASE IN TRITIUM LEVELS IN A GROUND WATER SAMPLE

"This report is being made pursuant to 10 CFR 50.72(b)(2)(xi) for notification to other government agencies.

"Dominion North Anna Power Station intends to voluntarily notify state and local agencies regarding an increase in tritium levels in one (1) onsite ground water monitoring sample point. This increase in tritium levels has not exceeded any NRC regulatory dose limits nor has it exceeded the voluntary reporting limits (i.e., 20,000 picoCuries per liter) specified in NEI 07-07 Industry Ground Water Protection - Final Guidance Document. Two (2) adjacent onsite ground water monitoring sample points have not shown a similar increase. None of the eight (8) ground water monitoring sample points surrounding the station have shown any detectable levels of tritium. All indications show that the tritium in the one (1) onsite ground water monitoring sample point has not migrated to the lake or any drinking water sources.

"The station continues to monitor, sample and investigate the source of the tritium anomaly. This condition does not present a health hazard to station employees or the general public."

Normal tritium levels at the particular sample point are 3-4000 picoCuries per liter. One sample read 16,500 picoCuries per liter. Samples afterwards have return to normal readings.

The NRC Resident Inspector has been notified.

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Other Nuclear Material Event Number: 46378
Rep Org: GE HEALTH CARE
Licensee: GE HEALTH CARE
Region: 3
City: LIVONIA State: MI
County:
License #: 21-24828-01MD
Agreement: N
Docket:
NRC Notified By: EMILE POISSON
HQ OPS Officer: DONG HWA PARK
Notification Date: 10/29/2010
Notification Time: 15:34 [ET]
Event Date: 10/26/2010
Event Time: 13:00 [EDT]
Last Update Date: 10/29/2010
Emergency Class: NON EMERGENCY
10 CFR Section:
20.1906(d)(2) - EXTERNAL RAD LEVELS > LIMITS
Person (Organization):
RICHARD SKOKOWSKI (R3DO)
CHRISTIAN EINBERG (FSME)

Event Text

PACKAGE SURFACE CONTAMINATION IN EXCESS OF REPORTING LIMITS

The licensee reported receipt of a package of radioactive material with removable surface contamination on the outside of the package greater than the reporting limits of 220 dpm per cm squared. The package contained returned material, Tc-99m (Technetium), that was being shipped back to the company from a nearby customer in Rodchester Hills, MI. A wipe test performed on the external surface of the package indicated a removable contamination level of 583 dpm per cm squared.

A survey inspection of the receiving area did not find any contamination and other survey results were inconclusive. The surface contamination appears to be Tc-99m. No personnel contamination resulted from the incident and the package is being stored in a secured area.

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Power Reactor Event Number: 46379
Facility: DUANE ARNOLD
Region: 3 State: IA
Unit: [1] [ ] [ ]
RX Type: [1] GE-4
NRC Notified By: DICK FOWLER
HQ OPS Officer: DONG HWA PARK
Notification Date: 10/29/2010
Notification Time: 20:59 [ET]
Event Date: 10/29/2010
Event Time: 16:21 [CDT]
Last Update Date: 10/29/2010
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(A) - DEGRADED CONDITION
Person (Organization):
RICHARD SKOKOWSKI (R3DO)

Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
1 N N 0 Refueling 0 Refueling

Event Text

DEGRADED CONDITION - 71% THROUGH WALL INDICATION ON REACTOR RECIRC NOZZLE

"Information from a Phased Array UT [Ultrasonic Testing] examination of RRA-F002A indicates that a linear indication 6.5" long and a maximum depth of 71 % through wall extent (ID [Inner Diameter] connected) exists. The Phased Array UT examination of RRA-F002A was performed as part of scheduled outage activities.

"TRM [Technical Requirements Manual] LCO [Limiting Condition for Operation] 3.7.3 for Structural Integrity, Condition B has been entered for the Recirc nozzle/piping.

"This event is being reported as a degraded condition pursuant to 10CFR50.72(b)(3)(ii)(A) based on the fact that the indications result in a material defect in the primary coolant system which cannot be found acceptable under ASME Section XI, IWB-3600, 'Analytical Evaluation of Flaws' or ASME Section XI, Table IWB-3410-1, 'Acceptance Standards.'

"The licensee informed the NRC Resident Inspector."

Page Last Reviewed/Updated Thursday, March 29, 2012
Thursday, March 29, 2012