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Event Notification Report for June 18, 2013

U.S. Nuclear Regulatory Commission
Operations Center

Event Reports For
06/17/2013 - 06/18/2013

** EVENT NUMBERS **


48996 48998 49098 49102 49126 49127 49128

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Part 21 Event Number: 48996
Rep Org: CURTISS WRIGHT FLOW CONTROL CO.
Licensee: CURTISS WRIGHT FLOW CONTROL CO.
Region: 1
City: EAST FARMINGTON State: NY
County:
License #:
Agreement: Y
Docket:
NRC Notified By: JOHN DEBONIS
HQ OPS Officer: CHARLES TEAL
Notification Date: 05/03/2013
Notification Time: 09:25 [ET]
Event Date: 05/03/2013
Event Time: [EDT]
Last Update Date: 06/17/2013
Emergency Class: NON EMERGENCY
10 CFR Section:
21.21(a)(2) - INTERIM EVAL OF DEVIATION
Person (Organization):
GORDON HUNEGS (R1DO)
PART 21 GROUP (EMAI)
ERIC DUNCAN (R3DO)

Event Text

INTERIM PART 21 REPORT OF POTENTIAL DEFECT IN A RELIEF VALVE BELLOWS

The following was excerpted from a fax:

(ii) Identification of the basic component supplied for such facility or such activity within the United States which may fail to comply or contains a potential defect.

Target Rock P/N: 303480-1, Bellows, Manufactured by Target Rock.

(iii) Identification of the firm supplying the basic component which fails to comply or contains a defect.

Target Rock, Business Unit of Curtiss-Wright Flow Control Corporation
1966E Broadhollow Road
East Farmingdale, NY 11735

(iv) Nature of the defect or failure to comply and the safety hazard which is created or could be created by such defect or failure to comply.

During as-found steam testing on March 5, 2013 of a Pilgrim Main Steam Safety Relief Valve (MS-SRV) (TR Model 09J-001, valve assembly S/N 5, pilot assembly S/N 23, bellows PIN 303480-1 S/N 607) a loud pop was heard and as-found testing was secured. Subsequently, the pilot assembly was removed from the valve assembly and subjected to a leak test and would not hold pressure. The pilot assembly was disassembled and a visual inspection of the P/N 303480-1 bellows convolutions revealed a through wall failure in one of the convolutions. It is noted the steam testing was performed at an offsite test facility and the valve did not fail installed in the plant.

The bellows acts as a pressure sensor responsible for initiating the opening of the MS-SRV at set pressure. Failure of the bellows does not directly impact the integrity of the Reactor Coolant System (RCS) pressure boundary, which is maintained by the bonnet assembly that surrounds it, but does impair the ability of the MS-SRV to provide over-pressure protection of the RCS. This technology has an extensive history of reliability in nuclear power systems and has been used in Commercial Nuclear Power Plants (NPPs) since the 1970s. This is the first reported incident regarding a thru wall bellows failure.

Target Rock initiated a comprehensive root cause evaluation pursuing several areas of investigation. In parallel, Entergy is conducting an independent investigation and we are cooperating with them. A complete review of our paperwork confirms all manufacturing procedures and processes were performed in accordance with all specified requirements. This includes:

- Raw material analysis
- Dimensional inspections
- Cleaning
- Heat Treatment
- Manufacturing processes
- Testing
- Review of design stresses

Preliminary metallurgical analysis of the failed bellows indicates cracks forming in an inter-granular manner as would be expected from Inter Granular Stress Corrosion Cracks (IGSCC) originating at pit like location on the interior pressurized surface. The source of this cracking is the focus of on going investigations. Target Rock has also visually inspected two other bellows of the same part number, one manufactured from the same material lot and another manufactured from an earlier material lot. Both of these bellows were installed in valves steam tested at Target Rock. One of these valves bellows was also full flow tested at Wyle Labs. Neither of these additional bellows contained pit-like locations and may indicate this potential failure mechanism is an isolated incident. However, to date, neither Target Rock nor Pilgrim can draw final conclusions with the information collected and analyzed.

The mode of failure has not been determined; however, in order to address the potential for a common mode failure, Target Rock is continuing metallurgical testing of the failed bellows and the two other bellows with the same part number. Based on these results, it is likely we will need to evaluate bellows that have been installed in other NPP as they become available.

(v) The date on which the information of such defect or failure to comply was obtained.

The as-found steam test and identification of the potential defect occurred on March 5, 2013.

(vi) In the case of a basic component which contains a defect or fails to comply, the number and location of these components in use at, supplied for, being supplied for, or may be supplied for, manufactured, or being manufactured for one or more facilities or activities subject to the regulations in this part.

The following plants are running with bellows P/N 303480-1 installed: Limerick 1 & 2, Pilgrim, and J.A. Fitzpatrick.

(vii) The corrective action which has been, is being, or will be taken; the name of the individual or organization responsible for the action; and the length of time that has been or will be taken to complete the action.

The root cause of the potential defect is not yet known as of the date of this report. Therefore, no specific corrective actions have been initiated. Target Rock Corrective Action Request CAR 13-013 will document the corrective actions when they are determined. This determination will be based on further mechanical and material evaluations. TR anticipates completing these evaluations within 45 days; however, in the event the evaluations are not completed, TR will forward another interim report within 45 days.

(viii) Any advice related to the defect or failure to comply about the facility, activity, or basic component that has been, is being, or will be given to purchasers or licensees.

Target Rock will recommend that the end user perform a detailed visual inspection of the interior convolutions of installed bellows P/N 303480-1 at the next opportunity to determine if any areas of pitting or cracking exist on the interior walls of the bellows. This is a difficult inspection to perform due to the following: internal geometry of the convolutions, a trained inspector is required and specific inspection technology is needed to yield reliable results.

* * * UPDATE FROM JOHN DEBONIS TO HOWIE CROUCH VIA EMAIL AT 1109 EDT ON 6/17/13 * * *

The following are excerpts from an email sent by Target Rock, a business unit of Curtiss-Wright Flow Control Corporation:

Our [Target Rock] investigation indicates the bellows failed due to in-situ hydrogen embrittlement and this hydrogen embrittlement may have been promoted by inadequate cleaning of the bellows. The inadequate cleaning may have induced formation of surface pits during heat treatment providing for localized concentration of hydrogen.

Based on these results, we [Target Rock] are notifying end users with the P/N 303480-1 bellows in service (listed below) to perform field inspections at the next available opportunity. Note, the ASME Code requires these valves to be as-found tested at a maximum 5-year interval. A procedure to inspect the bellows will be forwarded to the applicable plants in parallel with this notification.

In addition to this inspection Target Rock recommends, as a preventive measure, the P/N 303480-1 bellows be replaced with a P/N 300083-1 or -3 bellows, as applicable, to negate the effects of hydrogen embrittlement. Finite element analysis of the P/N 300083-1 or -3 bellows shows significantly lower stresses at plant operating conditions. The lower stress levels provide an incremental increase in safety margin so that hydrogen embrittlement need not be considered a significant degradation mechanism.

Target Rock is implementing corrective actions to improve in-process cleaning and inspection, with emphasis on cleaning prior to heat treatment, to address this root cause.

The corrective actions will be completed within 60 days of this letter.

Should you have any questions regarding this matter, please contact Steven Pauly, Vice President Energy Products at (631) 293-3800, ext. 4640.

Limerick 1 & 2 has 28, Pilgrim has 4, and FitzPatrick has 3 of these items.

Notified NRR Part 21 Group (email), R1DO (Rogge) and R3DO (Daley) via email.

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Part 21 Event Number: 48998
Rep Org: CURTISS WRIGHT FLOW CONTROL CO.
Licensee: WOLLASTON ALLOYS, INC.
Region: 1
City: CHESWICK State: PA
County:
License #:
Agreement: Y
Docket:
NRC Notified By: JAMES DRAKE
HQ OPS Officer: PETE SNYDER
Notification Date: 05/03/2013
Notification Time: 10:50 [ET]
Event Date: 05/03/2013
Event Time: [EDT]
Last Update Date: 06/17/2013
Emergency Class:
10 CFR Section:
21.21(d)(3)(i) - DEFECTS AND NONCOMPLIANCE
Person (Organization):
GORDON HUNEGS (R1DO)
KATHLEEN O'DONOHUE (R2DO)
ERIC DUNCAN (R3DO)
MARK HAIRE (R4DO)
PART 21 REACTORS (EMAI)

Event Text

PART 21 REPORT - INSUFFICIENT PROCESS CONTROL ON PUMP IMPELLER

The following is a summary of information received via fax:

"In January 2013, Curtiss-Wright Electro Mechanical Corporation completed final testing on AP1000 Reactor Coolant Pump (RCP) Serial Number 9, part number 6D70795G05, Revision 8, which contained a sand cast impeller (S/N 3021) cast by Wollaston Alloys of Braintree, MA. When it was disassembled for inspection it was discovered that a piece of an impeller blade approximately 3 inches by 2 1/2 inches had separated from the main impeller casting. The separated piece was the leading edge of one blade, and it was subsequently recovered intact from the pump test loop.

"This incident was investigated as a significant condition adverse to quality with the potential to create a substantial safety hazard; but, was deemed not a reportable incident since all cast impellers were either:
1) in CW-EMD control, or
2) exported to customers in the People's Republic of China.

"Our customers (Westinghouse Electric Company and the Chinese customers and regulatory authorities) were kept informed as the investigation progressed and root cause was identified.

"The physical cause of the failure is most likely due to a flaw present in both the cast material and weld overlay applied to the impeller blade. The original flaw was most likely a consequence of tensile overload failure due to cooling stresses introduced by the welding process. Subsequent weld repairs were insufficient in remediating the original flaw, which went undetected by NDT methods. Ultimately, AP1000 RCP Serial Number 3021 failed by high cycle fatigue followed by ductile failure.

"As a result of the above investigation, CW-EMD is concerned that the identified lack of process control at Wollaston Alloys, Inc., could result in other significant conditions adverse to quality with the potential to create a substantial safety hazard.

"Because of the nature of the issue, CW-EMD is unable to complete a full extent of condition investigation, and is reporting this issue to the Commission to ensure full awareness within the industry.

"Name and address of the individual or individuals informing the Commission:

James A. Drake, General Manager
Curtiss-Wright Electro-Mechanical Corporation
1000 Wright Way
Cheswick, Pa 15024"

* * * UPDATE FROM STEVE GRIEF TO JOHN SHOEMAKER ON 5/17/13 AT 1549 EDT * * *

Subject: Report of Potential Substantial Safety Hazard in accordance with Title 10 Code of Federal Regulations, Part 21.

Wollaston Alloys is submitting this interim report as a result of product concerns discovered by Curtiss Wright EMD during the investigation of an impeller blade failure occurring during testing as noted in Curtiss Wright EMD's notification to the NRC dated May 3, 2013. Wollaston is requesting an additional 30 days to identify any current or previous orders where 10 CFR 21 is invoked and to determine if there is evidence that a condition exists that could create a substantial safety hazard.

Notified R1DO (Schroeder), R2DO (Bartley), R3DO (Riemer), R4DO (Walker), and Part 21 Reactors via email.

* * * UPDATE FROM STEPHEN GRIEF TO PETE SNYDER ON 6/17/13 AT 1752 EDT * * *

Subject: Report of Potential Substantial Safety Hazard in accordance with Title 10 Code of Federal Regulations, Part 21.

"Wollaston Alloys, Inc. is submitting this report as an update to the interim report submitted on May 17, 2013 resulting from product concerns identified by Curtiss Wright EMD during the investigation of an impeller blade failure as noted in Curtiss Wright EMD's notification to the Nuclear Regulatory Commission (NRC), dated May 3, 2013. This investigation did not include the Curtiss Wright impeller failure since it is not considered a reportable incident, but addresses the identified concerns with regards to other products supplied to domestic users. The previously submitted interim report requested an additional 30 days to identify basic components supplied under the requirements of 10 CFR Part 21 and to determine if there is evidence that a defect or failure to comply exists.

"Name and address of the individual or individuals informing the Commission:

Stephen M. Grief, Quality Manager
Wollaston Alloys, Inc.
205 Wood Road
Braintree, MA 02184"

The complete report has been summarized as follows:

Wollaston Alloys, Inc. determined the cases in which Wollaston Alloys, Inc. supplied components to which the requirements of 10 CFR Part 21 applied. After a review of all records pertaining to those purchase orders, Wollaston Alloys, Inc. found no evidence of a defect or failure to comply.

Notified R1DO (Rogge), R2DO (Ehrhardt), R3DO (Daley), R4DO (Walker), and Part 21 Reactors via email.

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Agreement State Event Number: 49098
Rep Org: OK DEQ RAD MANAGEMENT
Licensee: BUILDING AND EARTH SCIENCES
Region: 4
City: PONCA CITY State: OK
County: USA
License #: OK-31032-01
Agreement: Y
Docket:
NRC Notified By: KEVIN SAMPSON
HQ OPS Officer: DANIEL MILLS
Notification Date: 06/07/2013
Notification Time: 09:48 [ET]
Event Date: 06/07/2013
Event Time: [CDT]
Last Update Date: 06/09/2013
Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
Person (Organization):
BLAIR SPITZBERG (R4DO)
FSME EVENTS RESOURCE (EMAI)
ILTAB (EMAI)

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

Event Text

AGREEMENT STATE REPORT - MISSING TROXLER GAUGE

The following was received from the State of Oklahoma via email:

"Building and Earth Sciences has reported that a Troxler Model 3430 (S/N 36097) was stolen from a location in Ponca City, OK. The case was in the back of a truck and some time during the night it was opened and the gauge removed. The RSO is on the way to the site now, [the Oklahoma Department of Environmental Quality] will provide more information as it becomes available."

Troxler Model 3420 Density gauges typically contain Cs-137 8 mCi and Am-241/Be 40 mCi sources.

* * * UPDATE FROM MIKE BRODERICK TO JOHN SHOEMAKER AT 1100 EDT ON 6/9/2013 * * *

The State of Oklahoma reports that the gauge has been recovered and is now in the possession of the licensee. A private citizen found the missing gauge on the side of a road near the location where the gauge was stolen in Ponca City, OK. The source rod was still locked and it is believed that no exposures have occurred. The licensee will return the gauge to a secure storage location.

Notified R4DO (Spitzberg) and FSME Events Resource and ILTAB via email.

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

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Agreement State Event Number: 49102
Rep Org: MINNESOTA DEPARTMENT OF HEALTH
Licensee: UNIVERSITY OF MINNESOTA
Region: 3
City: MINNEAPOLIS State: MN
County:
License #: 1049-211-27
Agreement: Y
Docket:
NRC Notified By: SHERRIE FLAHERTY
HQ OPS Officer: MARK ABRAMOVITZ
Notification Date: 06/10/2013
Notification Time: 16:53 [ET]
Event Date: 08/22/2012
Event Time: [CDT]
Last Update Date: 06/10/2013
Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
Person (Organization):
ERIC DUNCAN (R3DO)
FSME EVENT RESOURCES ()

Event Text

AGREEMENT STATE REPORT - POTENTIAL MEDICAL OVERDOSE

The following report was received via e-mail:

"On May 26, 2013 during a transfer of electronic treatment planning records to a new system, the University of Minnesota (license number 1049-211-27) discovered a medical event that occurred on August 20-22, 2012 at the university of Minnesota Medical Center in Minneapolis with the Nucletron HDR. The licensee reported that dosimetry staff were testing the transfer of information from previously treated patients into a brachytherapy check program, and it was discovered that in this particular case the source position data was entered into the HDR planning system incorrectly. The licensee is calculating the exact doses delivered and it appears as though the dose to unintended regions by greater than 50% for several areas. The Minnesota Department of Health was notified of the potential event on May 27, 2013. A final report will be submitted when the report is received from the licensee."

A Medical Event may indicate potential problems in a medical facility's use of radioactive materials. It does not necessarily result in harm to the patient.

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Power Reactor Event Number: 49126
Facility: WOLF CREEK
Region: 4 State: KS
Unit: [1] [ ] [ ]
RX Type: [1] W-4-LP
NRC Notified By: JIM KURAS
HQ OPS Officer: DONALD NORWOOD
Notification Date: 06/17/2013
Notification Time: 14:37 [ET]
Event Date: 06/17/2013
Event Time: 11:11 [CDT]
Last Update Date: 06/17/2013
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(i) - PLANT S/D REQD BY TS
Person (Organization):
WAYNE WALKER (R4DO)

Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
1 N Y 100 Power Operation 55 Power Operation

Event Text

TECHNICAL SPECIFICATION REQUIRED SHUTDOWN DUE TO NON-FUNCTIONAL CLASS 1E ELECTRICAL A/C UNIT

"Class 1E A/C Unit SGK05A cools safety related electrical train 'A' and was declared non-functional at 1111 hours. As a result, the following supported safety related electrical equipment was declared inoperable: 4.16 KV Bus NB01, 480 volt Buses NG01 and NG03, 120 volt Instrument AC Inverters and Buses NN11, NN13, NN01 and NN03, 125 VDC Chargers and Buses NK11, NK13, NK01 and NK03. T/S 3.0.3 was entered from T/S 3.8.7 due to two out of four 120 volt AC Inverters (NN11 and NN13) being inoperable. All electrical systems listed above remain available but are declared inoperable due to inadequate room cooling capability. Plant shutdown to mode 5 commenced at 1125 hours. No major equipment is out-of-service. All systems have functioned normally. Plant is currently at 55% power ramping down. Plant must be in mode 3 by 1811 CDT. No compensatory measures have been established.

"The NRC Resident Inspector has been notified. "

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Power Reactor Event Number: 49127
Facility: NINE MILE POINT
Region: 1 State: NY
Unit: [ ] [2] [ ]
RX Type: [1] GE-2,[2] GE-5
NRC Notified By: JASON SAWYER
HQ OPS Officer: DONALD NORWOOD
Notification Date: 06/17/2013
Notification Time: 18:16 [ET]
Event Date: 06/17/2013
Event Time: 12:27 [EDT]
Last Update Date: 06/17/2013
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE
Person (Organization):
JOHN ROGGE (R1DO)

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

LOSS OF RADWASTE / REACTOR BUILDING VENT GASEOUS EFFLUENT RADIATION MONITORING CAPABILITY

"At 1227 EDT on June 17, 2013, Nine Mile Point Unit 2 determined that the Radwaste / Reactor Building vent gaseous effluent radiation monitor had failed its source check and was, therefore, non-functional.

"This radiation monitor is necessary for accident assessment and is credited for Emergency Action Level (EAL) classification. The inability to classify an EAL due to the out-of-service Radwaste / Reactor Building vent gaseous effluent monitor is considered a loss of emergency assessment capability and is reportable per 10CFR50.72(b)(3)(xiii).

"Troubleshooting is currently in progress. The expected out-of-service time is unknown at this time. Compensatory actions are in place to take and analyze periodic grab samples in accordance with the Offsite Dose Calculation Manual.

"The NRC Resident Inspector has been notified."

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Power Reactor Event Number: 49128
Facility: PRAIRIE ISLAND
Region: 3 State: MN
Unit: [1] [2] [ ]
RX Type: [1] W-2-LP,[2] W-2-LP
NRC Notified By: WAYNE EPPEN
HQ OPS Officer: DONALD NORWOOD
Notification Date: 06/17/2013
Notification Time: 18:34 [ET]
Event Date: 06/17/2013
Event Time: 16:15 [CDT]
Last Update Date: 06/17/2013
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
Person (Organization):
ROBERT DALEY (R3DO)

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 82 Power Operation 82 Power Operation

Event Text

POTENTIAL UNANALYZED CONDITION - CARBON DIOXIDE FIRE SUPPRESSION SYSTEM DOES NOT MEET DESIGN REQUIREMENTS

"At approximately 1615 CDT on June 17, 2013, station personnel identified a potential unanalyzed condition based on the following:

"The 1998 design calculation for the carbon dioxide fire suppression system protecting the Relay Room is in part based on an unverified leakage rate for the enclosure. Recent testing and subsequent calculations have found the room leakage rate is higher than the leakage rate specified in the 1998 design calculation. As such, more carbon dioxide would be required to meet design requirements. The minimum required level for the system storage tank would not currently provide two shots of carbon dioxide (design requirement) to suppress a fire.

"The minimum required quantity in the tank would be sufficient to provide one shot of carbon dioxide to suppress a fire; however, since this doesn't meet the design requirement, a continuous fire watch was initiated."

The licensee notified the NRC Resident Inspector.

Page Last Reviewed/Updated Tuesday, June 18, 2013
Tuesday, June 18, 2013