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Event Notification Report for February 9, 2011

U.S. Nuclear Regulatory Commission
Operations Center

Event Reports For
02/08/2011 - 02/09/2011

** EVENT NUMBERS **


46403 46545 46547 46599 46603 46604

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General Information Event Number: 46403
Rep Org: VELAN INC
Licensee: FLOWSERVE
Region:
City: QUEBEC, CANADA State:
County:
License #:
Agreement: N
Docket:
NRC Notified By: VICTOR APOSTOLESCU
HQ OPS Officer: MARK ABRAMOVITZ
Notification Date: 11/08/2010
Notification Time: 16:02 [ET]
Event Date: 09/15/2010
Event Time: [EST]
Last Update Date: 02/08/2011
Emergency Class: NON EMERGENCY
10 CFR Section:
21.21 - UNSPECIFIED PARAGRAPH
Person (Organization):
KATHLEEN O'DONOHUE (R2DO)
PART 21 GP VIA EMAIL ()
RONALD BELLAMY (R1DO)
ANN MARIE STONE (R3DO)
VINCENT GADDY (R4DO)

Event Text

PART 21 REPORT - LIMITORQUE LIMIT SWITCH DEFECT

The following report was received via fax:

"During the performance testing of our valves equipped with Limitorque SMB-00 we found that the limit switch contacts proved to be defective. These tests took place in July 2010. Flowserve was advised and sent in replacement parts that were installed by their representatives. The valve-actuator assemblies were cycled and proper operation was assessed. These valves have been shipped to Dominion Virginia. At the time we considered the issue isolated and did not pursue an in-depth corrective action response from the Supplier.

"Later in September when testing three valves equipped with Limitorque SMB-00 and two valves equipped with Limitorque SMB-2-06, we found again that the limit switch contacts were defective, exhibiting similar problems as found earlier in July. These valves are still at our factory, awaiting the response and corrective action from Flowserve.

"The limit switch boxes (4 gear train limit switches, 16 sets of contacts) appear to be identical on the two types of actuators mentioned above.

"Upon closer examination, we determined that construction and installation elements appear poorly controlled, resulting in unexpected failure to operate due to the contact blade (called finger base by the Manufacturer) not returning to a position where it can make contact again. This was documented internally on a Velan internal deviation report on September 3, 2010.

"We advised Flowserve of our findings on September 15 and issued a formal Corrective Action Request (CAR 25500-73903) on September 16, 2010, with a deadline for responding that expired on October 26, 2010. After a number of follow-ups, we managed to make contact with responsible personnel at Flowserve on October 29. An evaluation report (electrical continuity test performed on sample switch assemblies cycled 2000 times) was submitted to our attention by Flowserve. However, we determined that the test did not answer all our concerns and requested Flowserve to provide additional information. Currently the supplier is engaged in retrieving the defective parts from our facilities and performing additional examinations and tests. The Manufacturer expects to have all necessary tests, examinations and evaluations completed on or before November 19, 2010.

"Based on functional testing performed at Velan we determined that we have no record of similar defects on valve-actuator assemblies produced prior to these events, we therefore believe that the root cause is relatively recent but there is no way to know until Flowserve analyzes and evaluates the deficiency.

"This type of defect has the potential to affect other valve manufacturers who may have installed Limitorque actuators equipped with this type of limit switch but we cannot say if such deviation could create a substantial safety hazard."


* * * UPDATE FROM VICTOR APOSTOLESCU TO DONALD NORWOOD VIA FACSIMILE AT 0804 EST ON 1/24/2011 * * *

On January 14, 2011, Velan received the final report from Flowserve concerning limit switches identified in this notification. Velan has accepted the conclusions in the report.

The following is a synopsis of those conclusions: It was determined that producing a bend in the contact finger cannot occur during normal cyclic operation of the rotor. It is highly likely that the cause of the bent finger assemblies was due to the use of a flat blade screwdriver. A flat blade screwdriver can also exert enough force to damage the cotter pin hole in the spring stud. Based on testing and evaluations of all returned Velan switches and switches from Flowserve stock, a design deficiency has not been identified. Properly set switches will perform their intended functions. A maintenance update will be issued by Flowserve to guide the industry on any recommendations during their regularly scheduled maintenance outages.

Notified R1DO (Newport), R2DO (Sykes), R3DO (Bloomer), and R4DO (O'Keefe). Notified Part 21 Group via E-mail. Notified NRR and NRO via facsimile.

* * * UPDATE AT 1528 ON 2/8/2011 FROM VICTOR APOSTOLESCU TO MARK ABRAMOVITZ * * *

Velan Inc. has issued the final report on this problem with no changes from the January 14, 2011 update.

Notified R1DO (Bellamy), R2DO (McCoy), R3DO (Duncan), and R4DO (Clark). Notified Part 21 Group via E-mail.

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Part 21 Event Number: 46545
Rep Org: ABB INC.
Licensee: ABB INC.
Region: 1
City: CORAL SPRINGS State: FL
County:
License #:
Agreement: Y
Docket:
NRC Notified By: CHAD BUCHWALTER
HQ OPS Officer: VINCE KLCO
Notification Date: 01/14/2011
Notification Time: 18:52 [ET]
Event Date: 01/10/2011
Event Time: [EST]
Last Update Date: 02/08/2011
Emergency Class: NON EMERGENCY
10 CFR Section:
21.21 - UNSPECIFIED PARAGRAPH
Person (Organization):
LAWRENCE DOERFLEIN (R1DO)
BRIAN BONSER (R2DO)
MICHAEL KUNOWSKI (R3DO)
BOB HAGAR (R4DO)
PART 21 GRP by email ()

Event Text

PART 21 NOTIFICATION OF POTENTIAL DEFECT FOR OVERCURRENT RELAYS

The following information was received by facsimile:

"This letter is submitted in accordance with 10 C.F.R 21.21(d)(3)(ii) with respect to a failure to comply with the specifications associated with the COM 5, COM 9, and COM 11 Overcurrent Relays. The style numbers for the COM 5 relay are 1326D81A07A. 1326D81A05A, 1326D81A01, 1326D81A02, 1329D12A02, 1329D12A01, 1329D12A06, 1326D81A08, 1329D12A07, 1329D12A05, 1329D12A06, 1329D12A08, 1329D12A09, 1326D81A08A, and 1326D81A09A. The style numbers for the COM 9 relay are 1326D81A10A and 1326D81A10. The style numbers for the COM 11 relay are 1329D12A03, 1326D81A03, and 1326D81A03A.

"The notifying individual is Mr. Pat Wilkinson, General Manager, ABB Inc. (Distribution Automation), 4300 Coral Ridge Rd, Coral Springs FL, 33065.

"Notification regarding the subject relays is as follows: The failure to comply centers around the seismic specification of the COM 5, COM 9, and COM 11 relays. The Zero Period Acceleration (ZPA) rating for the COM 5, COM 9 and COM 11 relays were incorrectly being certified to meet a ZPA rating of 5.6g. The relays only meet a ZPA rating of 3.6g.

"On December 14th, 2010, ABB's Engineering Group, while performing a document review of the ABB's CTR-COM-5 Qualification Conclusion Report, discovered the incorrectly reported ZPA rating. The deviation was identified as a potential defect on January 10, 2011.

"The COM family relays were originally seismically qualified by Westinghouse on July 11, 1977 with a ZPA rating of 5.7g. A second seismic qualification test was performed by an outside vendor on September 12, 2001 with a reported ZPA rating of 3.6g. ABB then transferred the vendor information to its own conclusion report: CTR-COM-5 signed December 14, 2001. The CTR-COM-5 conclusion report incorrectly transcribed the ZPA rating of 3.6g from the outside vendor. The CTR-COM-5 conclusion report is the basis for ABB's Relay Selection disk, used by ABB Marketing, and ABB's quality Certificate of Conformance.

"The root cause of this issue was determined to be inadequate review and transfer of the outside vendor's seismic test data.

"ABB is taking, or has taken, the following corrective actions:

- Correct CTR-COM-5 conclusion report to 3.6g ZPA (Engineering completed request on January 11, 2011 )
- Contact all customers with potentially affected open Purchase Orders to ensure acceptance of the 3.6g ZPA rating. Orders on hold until acceptance. (Sales to complete by January 21, 2011)
- Update Certificate of Conformance template for the affected style numbers to reflect 3.6g ZPA. (Quality Assurance completed request on January 13, 2011)
- Perform a review of all qualification reports to ensure all ZPA ratings for all product families are correctly reported (Engineering to complete by February 14, 2011)
- Identification of potentially affected customers (Marketing to complete by February 15, 2011)
- Notification of potentially affected customers (Marketing to complete by February 28, 2011).
- Correct ZPA rating on the Relay Selection Disk to 3.6g ZPA (Marketing to complete by October 30, 2011)

"The customers and the quantity data are still being collected at this time. Depending upon a Licensee's specified ZPA requirements, the lower ZPA rating of relays could possibly create a substantial safety hazard. If a higher ZPA rating is required by the Licensee, please contact ABB Coral Springs Customer Support at 1-800-222-1946 or (954) 825-0606 on available solutions.

"If you have any questions regarding this notice, please contact the Quality Manager, Mr. Chad Buchwalter, directly at (954) 825-0604."

* * * UPDATE FROM CHAD BUCHWALTER TO HOWIE CROUCH VIA FAX ON 2/8/11 @ 0857 EST * * *

ABB Inc. updated their initial report to state that they had previously made 10 CFR 21.21 notifications on ZPA deviations for the COM relays (see EN #30753). Additionally, they provided a list of customers who may have received the subject relays. No nuclear power plants were listed on the customer roster.

Notified R1DO (Bellamy), R2DO (Hopper), R3DO (Duncan), R4DO (Clark) and Part 21 Group via email.

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Non-Agreement State Event Number: 46547
Rep Org: ELI LILLY AND COMPANY
Licensee: ELI LILLY AND COMPANY
Region: 3
City: INDIANAPOLIS State: IN
County:
License #:
Agreement: N
Docket:
NRC Notified By: STAN HANPTON
HQ OPS Officer: MARK ABRAMOVITZ
Notification Date: 01/17/2011
Notification Time: 10:48 [ET]
Event Date: 01/14/2011
Event Time: [EST]
Last Update Date: 02/08/2011
Emergency Class: NON EMERGENCY
10 CFR Section:
20.2201(a)(1)(i) - LOST/STOLEN LNM>1000X
Person (Organization):
MICHAEL KUNOWSKI (R3DO)
LAURA PEARSON (ILTA)
LARRY CAMPER (FSME)

This material event contains a "Less than Cat 3" level of radioactive material.

Event Text

LOST TRITIUM EXIT SIGNS

A major renovation was performed in two floors of Building 22 on the Eli Lilly Campus. This renovation was completed in December, 2010. An inventory of tritium exit signs was performed and eight exit signs on the two floors were found missing and presumed disposed of during the renovation. The exit signs each contained approximately 10 Curies of tritium. The licensee will perform wipe tests but expects negative results since the concrete surfaces were all covered with carpet, tile, or wall coverings.

Four Evenlite Model 201 serial numbers 57281, 59311, 57381, 57421
Four SRBT Model BX serial numbers C035796, C035795, C035790, C035789

* * * UPDATE AT 1555 ON 2/8/2011 FROM STAN HANPTON TO MARK ABRAMOVITZ * * *

The inventory of tritium exit signs was extended and eight additional tritium exit signs were missing. These additional exit signs were assigned to building 31 on the Eli Lilly Campus. This building had also been recently renovated. Each sign contained approximately 10 Ci tritium. Wipe tests in building 22 and 31 were performed with no activity detected.

Notified the R3DO (Duncan), ILTAB (Allston), and FSME (McConnell).

THIS MATERIAL EVENT CONTAINS A "LESS THAN CAT 3" LEVEL OF RADIOACTIVE MATERIAL

Sources that are "Less than IAEA Category 3 sources," are either sources that are very unlikely to cause permanent injury to individuals or contain a very small amount of radioactive material that would not cause any permanent injury. Some of these sources, such as moisture density gauges or thickness gauges that are Category 4, the amount of unshielded radioactive material, if not safely managed or securely protected, could possibly - although it is unlikely - temporarily injure someone who handled it or were otherwise in contact with it, or who were close to it for a period of many weeks. For additional information go to http://www-pub.iaea.org/MTCD/publications/PDF/Pub1227_web.pdf

This source is not amongst those sources or devices identified by the IAEA Code of Conduct for the Safety & Security of Radioactive Sources to be of concern from a radiological standpoint. Therefore is it being categorized as a less than Category 3 source

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Power Reactor Event Number: 46599
Facility: SAINT LUCIE
Region: 2 State: FL
Unit: [1] [2] [ ]
RX Type: [1] CE,[2] CE
NRC Notified By: LOURDES PORRO
HQ OPS Officer: DONALD NORWOOD
Notification Date: 02/07/2011
Notification Time: 22:35 [ET]
Event Date: 02/07/2011
Event Time: 22:25 [EST]
Last Update Date: 02/08/2011
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE
Person (Organization):
GEORGE HOPPER (R2DO)

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 N 0 Refueling 0 Refueling

Event Text

CONTROL ROOM EMERGENCY VENTILATION SYSTEM OUT OF SERVICE DUE TO PLANNED MAINTENANCE

"On February 07, 2011 at 2225 EST, the Control Room Emergency Ventilation system on St. Lucie Unit 1 was declared out of service due to pre-planned maintenance to upgrade the Emergency Response Data Acquisition and Display System. This maintenance renders the control room envelope out of service for a portion of the maintenance. The Technical Support Center (TSC) ventilation system is part of the Unit 1 Control Room Emergency Ventilation system, therefore, the TSC ventilation system has been rendered non-functional during the course of work activities. The TSC ventilation is expected to be returned to service in 4 hours.

"If an emergency is declared requiring TSC activation during this period, the TSC will be staffed and activated using existing emergency planning procedures. Should the TSC become uninhabitable, the TSC staff will relocate to an alternate TSC location in accordance with applicable site procedures.

"This notification is being made in accordance with 10CFR50.72(b)(3)(xiii) due to the potential loss of an Emergency Response Facility (ERF). An update will be provided once the TSC ventilation system has been restored to normal operation. The NRC Resident Inspector has been notified."

* * * UPDATE FROM LOURDES PORRO TO DONG PARK AT 0040 EST ON 2/7/11* * *

"[As of 0013 EST, the] Control Room and TSC ventilation system has been restored to normal operation."

The licensee will notify the NRC Resident Inspector.

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Power Reactor Event Number: 46603
Facility: CLINTON
Region: 3 State: IL
Unit: [1] [ ] [ ]
RX Type: [1] GE-6
NRC Notified By: DAN HUNT
HQ OPS Officer: JOHN SHOEMAKER
Notification Date: 02/08/2011
Notification Time: 15:06 [ET]
Event Date: 02/08/2011
Event Time: 08:00 [CST]
Last Update Date: 02/08/2011
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
Person (Organization):
ERIC DUNCAN (R3DO)

Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
1 N Y 97 Power Operation 97 Power Operation

Event Text

UNANALYZED EVENT COULD INITIATE HIGH PRESSURE CORE SPRAY AND OVERFILL THE REACTOR PRESSURE VESSEL

"During the performance of a Fire Protection self assessment, it was discovered that a calculation for the safe shutdown analysis has an assumed action for an operator to locally depress the internal breaker trip plunger to trip the High Pressure Core Spray (HPCS) pump in response to a fire in the main control room. However, due to personnel safety concerns related to potential arc flashing events associated with this action, the remote shutdown procedure was revised to locally close the HPCS injection valve (1E22FOO4) in lieu of depressing the internal breaker trip plunger.

"During engineering's review of this procedure and supporting calculation, it was determined that the HPCS system could be initiated due to concurrent fire induced hot short cable damage to the two automatic initiation logic instrument cables routed in the same raceway in the area. In this event, even if the HPCS breaker could be tripped or the HPCS injection valve could be closed locally, HPCS would continue to fill the reactor pressure vessel (RPV) and flood the main steam lines. Once pressure reaches the setpoint for the Main Steam Safety Relief Valves (MSSRVs), they would lift and discharge mixed-phase water through the discharge line to the suppression pool. This conservatively postulated scenario would place the MSSRVs and their associated tailpipes in an unanalyzed condition for the stresses expected during the two-phase flow event.

"While it is not expected that a failure of the MSSRV discharge line will occur, a confirmatory analysis will be performed. Compensatory measures for Multiple Spurious Operations have been determined to be adequate until the analysis is complete."

The licensee added additional fire zone surveillance to operator plant walk downs and will investigate to determine further corrective actions.

The license has notified the NRC Senior Resident Inspector.

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Power Reactor Event Number: 46604
Facility: COLUMBIA GENERATING STATION
Region: 4 State: WA
Unit: [2] [ ] [ ]
RX Type: [2] GE-5
NRC Notified By: MATT HUMMER
HQ OPS Officer: MARK ABRAMOVITZ
Notification Date: 02/08/2011
Notification Time: 20:55 [ET]
Event Date: 12/20/2010
Event Time: 04:38 [PST]
Last Update Date: 02/08/2011
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
Person (Organization):
JEFF CLARK (R4DO)

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

Event Text

FUSE FAILURE CAUSES LOW PRESSURE CORE SPRAY TO BE INOPERABLE

"On December 20, 2010, the low pressure core spray (LPCS) system was declared inoperable due to loss of power to the LPCS minimum flow valve. The minimum flow valve supports operability by providing a flow path to prevent pump damage during situations where the LPCS pump has been started in response to a transient, but reactor vessel pressure is not low enough to allow LPCS injection. The power loss was caused by the clearing of all 3 line power fuses for the motor starter for the minimum flow valve. An apparent cause evaluation concluded that the most likely cause of the fuses clearing was a random fuse failure of one of the fuses at less than design amperage attributable to a defect in the fuse solder joint.

"The Technical Specification (TS) Required Action for LCO 3.5.1 Condition A, one low pressure ECCS injection/spray subsystems inoperable, was complied with by restoring the LPCS system to operable within the allowed completion time.

"The safety functions for LPCS are to provide inventory makeup and spray cooling during large breaks in the reactor coolant system that uncover the core. All remaining ECCS subsystems were operable and at no time did this event result in the loss of a safety function. The low pressure injection function was not challenged due to all three loops of the Residual Heat Removal (RHR) system Low Pressure Coolant Injection (LPCI) mode being operable while the core spray function was satisfied by the operable High Pressure Core Spray (HPCS) system.

"This event is being reported under 10 CFR 50.72(b)(3)(v)(D) as an event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident tor a single train system.

"Historically, LPCS inoperabilities at Columbia (including initial review of this event) were not considered to be a single train system for reportability purposes. The basis for the historical consideration was assessment of LPCS inoperabilities consistent with the plant safety analysis and the associated system and safety function groupings which do not single out LPCS as a single train system. There are two pertinent groupings in the safety analyses which are aligned with the credited safety functions of LPCS. The two groupings are the low pressure injection system function (combined with LPCI), and a core spray system function (combined with HPCS). Industry precedent has been consistent with the historical position. However, recent NRC interpretations have considered safety function at the lowest system level which result in LPCS being considered as a single train performing a safety function in scope of the reportability rules in 10 CFR 50.72 and 50.73. A Licensee Event Report will be submitted for this event.

"As a result of the recent interpretation with regard to LPCS, a review of prior LPCS inoperabilities within the past three years is being performed to determine if the reporting criteria were met during prior events. If necessary, additional 10 CFR 50.72 and 10 CPR 50.73 notifications/reports will be made on prior LPCS inoperabilities .

The licensee will notify the NRC Resident Inspector.

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Thursday, March 29, 2012