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Event Notification Report for June 12, 2017

U.S. Nuclear Regulatory Commission
Operations Center

Event Reports For
06/09/2017 - 06/12/2017

** EVENT NUMBERS **


51998 52721 52741 52756 52780 52783 52784 52786 52796 52797 52798 52800
52801

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Fuel Cycle Facility Event Number: 51998
Facility: B&W NUCLEAR OPERATING GROUP, INC.
RX Type: URANIUM FUEL FABRICATION
Comments: HEU FABRICATION & SCRAP
Region: 2
City: LYNCHBURG State: VA
County: CAMPBELL
License #: SNM-42
Agreement: N
Docket: 070-27
NRC Notified By: DAVID SPANGLER
HQ OPS Officer: RICHARD SMITH
Notification Date: 06/10/2016
Notification Time: 12:02 [ET]
Event Date: 06/09/2016
Event Time: 13:30 [EDT]
Last Update Date: 06/09/2017
Emergency Class: NON EMERGENCY
10 CFR Section:
PART 70 APP A (b)(1) - UNANALYZED CONDITION
Person (Organization):
ANTHONY MASTERS (R2DO)
NMSS_EVENTS_NOTIFIC (EMAI)
FUELS GROUP (EMAI)

Event Text

UNANALYZED CONDITION DUE TO A CHEMICAL PROCESSING MALFUNCTION

"I. EVENT DESCRIPTION: High enriched scrap fuel material is processed in BWXT NOG-Lynchburg's Uranium Recovery facility to reclaim as much of the uranium as possible. The material is dissolved in acid and transferred to a series of horizontal columns where the acid is neutralized. The solution may be transferred to a set of accountability weigh columns for measurement prior to entering the uranium extraction process. The solution is subsequently transferred to a series of horizontal feed columns. Process water is used to periodically flush the horizontal columns during cleanup for materials accountability.

"On June 9, 2016, a BWXT Nuclear Criticality Safety (NCS) engineer was notified that a bluish tint had been observed in the favorable geometry process water connection to the horizontal columns. By procedure, a blue dye is added to the acid to aid in its identification in the event of a spill. Further evaluation determined that the favorable geometry process water line was directly connected to the horizontal column system and the presence of the blue dye indicated a potential backflow of uranium bearing solution into the water line. The favorable geometry water line is under constant water pressure. The valves controlling the water flow are normally closed. There is also a check valve in the line to prevent backflow. The line is supplied from a favorable geometry header on the mezzanine above. The header supplies water to other processes in Uranium Recovery, including an unfavorable geometry hot water heater.

"The Integrated Safety Analysis (ISA) was reviewed and an accident sequence for this potential backflow could not be identified. On June 9, 2016 at 1330 [EDT] it was the determined the accident sequence was unanalyzed and not properly documented in the ISA. Although IROFS [Items Relied on for Safety] listed for other accident sequences were applicable to the backflow scenario, the performance requirements of 10 CFR 70.61 were not maintained. As documented in the ISA, criticality was not 'highly unlikely.'

"II. EVALUATION OF THE EVENT: Backflow into the favorable geometry water line can only be achieved by a forced flow to overcome the water pressure in the line. The only source for this pressurized flow is during the transfer from the accountability weigh columns to the horizontal column system. The solution is transferred using an air diaphragm pump. A trained and qualified operator opens the valve to initiate the transfer of solution from the accountability weigh columns to the horizontal columns (an uncredited control). The transfer of solution to the horizontal columns is monitored by a trained and qualified operator (an uncredited control). The water heater is substantially upstream of the supply line to the horizontal columns. The direction of flow of the process water in the supply header is away from unfavorable geometry hot water heater. The process water header is a favorable geometry (a credited IROFS). An operator checks the process water pressure on a daily basis (a credited IROFS). If the above existing IROFS and uncredited controls were considered in an ISA accident sequence, the likelihood of a criticality could be demonstrated to be highly unlikely. However, these uncredited controls are not designated as IROFS. Although the as-found condition presented no safety concern, the scenarios as documented in the ISA did not demonstrate that the performance requirements of 10 CFR 70.61 were maintained. There was no immediate risk of a criticality or threat to the safety of workers or the public as a result of this event.

"Ill. NOTIFICATION REQUIREMENTS: BWXT is making this 24 hour report in accordance with 10 CFR 70, Appendix A, (b)(1), 'Any event or condition that results in the facility being in a state that was not analyzed, was improperly analyzed, or is different from that analyzed in the Integrated Safety Analysis, and which results in failure to meet the performance requirements of 70.61.'

"IV. STATUS OF CORRECTIVE ACTIONS: A section of the piping from the process water supply header was removed to physically isolate the process water supply from the horizontal columns. Criticality is no longer credible. The hot water heater is the only unfavorable geometry connected to the process water system in the Uranium Recovery facility. The hot water heater was assayed with a gamma survey instrument in several locations along the bottom and up the sides. No counts above background were detected. In addition, multiple liquid samples were taken from the bottom of the water heater and a cartridge filter housing prior to the hot water heater. All samples were well counted and determined to be below the Minimum Detectable Activity (MDA). An investigation of the root causes of this event is ongoing. Corrective actions will be determined as a result of the investigation."

The licensee has notified the NRC Resident Inspector and Region II personnel.

* * * UPDATE AT 0944 EDT ON 06/09/17 FROM ROBERT JOHNSON TO S. SANDIN * * *

"I. EVENT DESCRIPTION: On June 10th, 2016 BWXT NOG-Lynchburg notified the NRC of an improperly analyzed condition involving the potential backflow of uranium bearing solution from a fissile solution processing system into the Uranium Recovery process water system. This notification was recorded as Event Notification Number 51998. One of the corrective actions in response to Apparent Violation 70-27/2016-004-01 was to conduct an Extent of Cause review of 'tasks involving ancillary systems (e.g., process water, nitric acid, HF acid, compressed air, steam, etc.) in our Uranium Recovery and Specialty Fuel Facility to verify these systems have documented accident scenarios as needed to meet the requirements of 10 CFR 70.'

"During the review of a waste processing system in the Specialty Fuel Facility (SFF), an additional potential backflow scenario was identified for a waste processing system. The system consists of a series of four favorable geometry columns. Waste solution is transferred to the columns using a less than or equal to 2.5 liter pump. The transfer of solution to the columns is monitored by a trained and qualified operator. Waste solution transferred to the columns is limited to a concentration of 5 grams 235U/liter as a Routine Operating Limit. A process water line is directly connected to the column system. The water is used to further dilute the concentration of the waste solution to a level that is acceptable for discharge into the hot waste drain (less than 0.04 grams 235U/liter). The column system is equipped with vent lines that overflow to the floor. The pump is capable of over-pressurizing the columns and possibly forcing waste solution into the favorable geometry process water system if the overflow vent lines were to fail. An unfavorable geometry hot water heater is located a significant distance downstream of the connection to the process water system. The Integrated Safety Analysis (ISA) was reviewed and an accident sequence for potential backflow of solution from the waste columns into the process water system could not be identified.

"II. EVALUATION OF THE EVENT: Administrative IROFS were identified in other accident sequences of the ISA that are applicable to the waste column backflow scenario. These IROFS include the operator control of solution concentration initially transferred to the waste processing system, and the daily check of the process water pressure which limits backflow into the system. Although these IROFS were available and reliable, additional IROFS are needed to meet the performance requirements of 10 CFR 70.61. Neither the operator's monitoring of the solution transfer or the overflow through the column vent lines are credited as IROFS in the ISA. They are considered uncredited safety controls.

"If either of the uncredited safety controls and the currently existing IROFS could be considered in an accident sequence, the likelihood of a criticality could be demonstrated to be highly unlikely. However, only IROFS documented in the ISA can be credited as preventing a criticality accident. Although the as-found condition presented no safety concern, the scenarios as documented in the ISA did not demonstrate the performance requirements of 10 CFR 70.61 were maintained. On June 8, 2017 at 10:00 am it was the determined the potential backflow of solution from the waste processing system was improperly analyzed and not documented in the ISA. As documented in the ISA, criticality was not highly unlikely. The SFF waste processing system was not in operation at the time of discovery. There was no immediate risk of a criticality or threat to the safety of workers or the public as a result of this event.

"Ill. NOTIFICATION REQUIREMENT: BWXT is making this 24 hour report to update Event Notification Number 51998 in accordance with 10 CFR 70, Appendix A, (b)(1) - Any event or condition that results in the facility being in a state that was not analyzed, was improperly analyzed, or is different from that analyzed in the Integrated Safety Analysis, and which results in failure to meet the performance requirements of 70.61.

"IV. STATUS OF CORRECTIVE ACTIONS: The process water line connected to the SFF waste processing system was locked out and posted as 'Out of Service.' Additional corrective actions are to be determined.

"The Extent of Cause review for AV 70-27/2016-004-01 is complete and the results are being finalized."

The licensee informed the NRC Resident Inspector.

Notified R2DO (Suggs), NMSS Events Notification, and Fuels Group by email.

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Agreement State Event Number: 52721
Rep Org: ARKANSAS DEPARTMENT OF HEALTH
Licensee: BAXTER HEALTHCARE CORPORATION
Region: 4
City: MOUNTAIN HOME State: AR
County:
License #: GL-0026
Agreement: Y
Docket:
NRC Notified By: ANGIE D. HALL
HQ OPS Officer: DONG HWA PARK
Notification Date: 05/02/2017
Notification Time: 14:16 [ET]
Event Date: 04/29/2017
Event Time: 18:30 [CDT]
Last Update Date: 06/09/2017
Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
Person (Organization):
MARK HAIRE (R4DO)
NMSS_EVENTS_NOTIFIC (EMAI)

Event Text

AGREEMENT STATE REPORT - FIXED GAUGE WITH STUCK SHUTTER

The following report was received from the State of Arkansas via e-mail:

"The Arkansas Department of Health was notified via telephone and e-mail on Monday, May 1, 2017, at approximately 1337 [CDT] of the licensee's general licensed device stuck shutter and the failure of the shutter mechanism. The equipment failure appears to have occurred from strong thunderstorms. The lockout procedure was verbally verified with the licensee.

"The licensee's fixed gauge is a ThermoFisher Scientific, Model Number E-SFL10-XX, Serial Number SP9810, which contains approximately 1,178.42 mCi Kr-85 (approximately 43.60 GBq Kr-85).

"The original activity: 1,250 mCi Kr-85 (approximately 46.25 GBq Kr-85) as of June 1, 2016. The source's Model Number is TFC-185 and the Serial Number is QC00256.

"Licensee operations with the gauge are twenty-four hours a day. The gauge has a fail-safe closing shutter mechanism and the gauge immediately went into the safe mode electronically during the thunderstorm. The closed shutter position was verified by the visual colored indicators in place.

"There have been no known radiation exposures to personnel and/or members of the public. There have been no known radiological health and safety concerns.

"The shutter will be repaired by the manufacturer today, May 2, 2017, whom will perform a root cause analysis. The State of Arkansas is awaiting information from today's evaluation, repairs, and surveys.

"The State is awaiting a 30 day written report from the licensee."

State Event Number: ARK-2017-002

* * * UPDATE ON 6/9/17 AT 1507 EDT FROM ANGIE HALL TO BETHANY CECERE * * *

The following follow-up report was received from the State of Arkansas via e-mail:

"The [Arkansas] Department [of Health] has received the required thirty (30) day report from the licensee, along with the related exposure survey and service reports from the manufacturer.

"The licensee states, 'We have several of these gauges that have been operating normally for several years. Due to the nature of this repair and the fact that this is the only time we have ever had to do this repair, it is my belief that a component in the device failed to operate as intended. The technician scraped paint off the surface of the shutter/flag mechanism that would help the mechanism stick to the magnet. The technician also tightened and applied thread lock to the securing hardware for the magnet. Both items should have been done at the factory.'

"The manufacturer's Field Engineer (also known as Technician) stated, 'I believe the issue started with the holding magnet being loose. This may have caused the solenoid to work harder and fail, and that could have caused the board to fail.'

"The manufacturer's Radiation Safety Officer stated, 'Thermo EGS Gauging LLC. is aware of the shutter mechanism failing in the closed or failing to the closed position on the TFC-185 and the TFC-190 sensors. The R&D manufacturing team has been working on a engineering project to correct this issue. We are currently in the testing phase and will manufacture all future sensors with the updated materials and parts. There is also an upgrade retrofit kit in the works for existing sensors that are demonstrating/developing this shutter issue. Currently we are sending out Field Service engineers to the customer's site to offer temporary repairs on an as needed basis.'

"The Department is waiting on a response from the manufacturer concerning the above statement.

"There have been no known radiological health issues, radiological exposures, safety hazards, or concerns to personnel or members of the public during this event.

"Event Cause: Equipment failure/design, manufacturing, or installation error.

"Corrective Action by licensee: New equipment obtained and repairs made with engineering change to system.

"Reporting Requirement Information:

"Arkansas State Board of Health 'Rules and Regulations for Control of Sources of Ionizing Radiation' RH-1502.f.2 and RH-402.c.

"U.S. Nuclear Regulatory Commission (NRC) Regulations Title 10 CFR Section 30.50 (b) (2) - Equipment is disabled or fails to function as designed, and Title 10 CFR Section 21.21 - A failure to comply or a defect affecting a basic component that is supplied for a facility or an activity that is subject to licensing requirements.

"The licensee and manufacturer reports and surveys have been sent to Randy Erickson, Regional State Agreement Officer, U.S. NRC, Region IV Office, and the NRC Headquarters Operations Center.

"Note: Leak tests are not required for this type of gauge."

NMED Item Number 170231.

Notified R4DO (Rollins) and NMSS Events Notification by email.

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Part 21 Event Number: 52741
Rep Org: EMERSON PROCESS MANAGEMENT
Licensee: FISHER CONTROLS INTERNATIONAL
Region: 3
City: MARSHALLTOWN State: IA
County:
License #:
Agreement: Y
Docket:
NRC Notified By: KIM SAGAR
HQ OPS Officer: BETHANY CECERE
Notification Date: 05/09/2017
Notification Time: 14:36 [ET]
Event Date: 03/10/2017
Event Time: [CDT]
Last Update Date: 06/09/2017
Emergency Class: NON EMERGENCY
10 CFR Section:
21.21(a)(2) - INTERIM EVAL OF DEVIATION
Person (Organization):
DON JACKSON (R1DO)
RANDY MUSSER (R2DO)
DAVID HILLS (R3DO)
JOHN KRAMER (R4DO)
PART 21/50.55 REACTO (EMAI)

Event Text

PART 21 REPORT - POTENTIAL MINIMUM WALL THICKNESS ISSUE ON CERTAIN POSI-SEAL VALVES

The following report was received via email:

"Pursuant to 10 CFR 21.21 (a)(2), Fisher Controls International LLC ('Fisher') is providing required written interim notification of a potential failure to comply concerning minimum wall requirements in the packing box region of certain Posi-Seal valves [A11].

"On March 10, 2017, Fisher discovered a minimum wall issue while processing an order. While the order at issue was corrected prior to shipment, Fisher is currently engaged in an extent of condition investigation to determine whether the noncompliance extends beyond the specific order. Fisher anticipates completing this investigation and disseminating additional information on or before June 9, 2017. Appropriate corrective action will be initiated to prevent reoccurrence.

"Should there be any further questions concerning this matter, please contact Benjamin Ahrens, Manager, Quality by email at Benjamin.Ahrens@Emerson.com or via phone at 641-754-2249."

* * * UPDATE ON 6/9/17 AT 1708 EDT FROM DEBBIE SCHAEFER TO BETHANY CECERE * * *

The following report was received via email:

"Fisher Information Notice: FIN 2017-03
9 June 2017

"Subject:
Potential Minimum Wall Thickness Issue on Certain Posi-Seal A11 Packing Boxes

"Purpose:
The purpose of this Fisher Information Notice (FIN) is to alert affected customers that, as of 10 March 2017, Fisher Controls International LLC (Fisher) became aware of a situation which may affect the minimum wall thickness of certain Posi-Seal designed valve bodies. Fisher is informing affected customers of this circumstance in accordance with Section 21.21 (b) of 10 CFR 21.

"Discussion:
Fisher engineering identified a potential minimum wall thickness issue on a 14 [inch] Posi-Seal Type A11 body drawing in the process of fulfilling an order. The valve body drawing was such that it was possible to machine the packing box region of the valve body neck within drawing-specified tolerances but fail to meet minimum wall thickness as required by ASME B16.34. As a result of this discovery, the affected order was halted and an investigation was initiated.

"Extent of Condition:
POSI-SEAL manufactured and supplied Type A11 valve bodies from their location in N. Stonington, CT to the nuclear industry prior to being acquired by Fisher in 1985. An extent of condition investigation was conducted to identify all potentially affected body drawing numbers and instances of orders shipped by Posi-Seal and subsequently by Fisher that could have had minimum wall issues. The scope of this investigation included all sizes and appropriate pressure classes of potentially affected valves shipped from either Posi-Seal or Fisher. The investigation concluded that only the 14 [inch] size for certain body drawing numbers are affected. Other sizes or types of valves are not affected by this FIN.

"Equipment Affected by this Fisher Information Notice:
Items subject to this FIN are confined to the 14 [inch] Type A11 valve bodies having original Posi-Seal drawing/serial numbers referred to in the attached Appendix A. All affected equipment was shipped from N. Stonington. No affected equipment was shipped from the Fisher Marshalltown, IA location.

"Action Required:
Fisher has revised the affected drawings to ensure the minimum wall requirements of ASME B16.34 are met. Fisher will also work directly with the affected customers to evaluate and resolve the potential minimum wall thickness issues.

"The in-process order was stopped and will be verified to meet minimum wall thickness requirements prior to shipment.

"10 CFR 21 Implications:
Fisher requests that the recipient of this FIN review it and take appropriate action in accordance with 10 CFR 21. If there are any technical questions or concerns, please contact:

"Ben Ahrens
Quality Manager
Emerson Automation Solutions
Fisher Controls International LLC
301 South First Avenue
Marshalltown, IA 50158
Phone: (641) 754-2249
Benjamin.ahrens@emerson.com

"List of Affected Equipment

"Customer/Site Niagara Mohawk Power Corp., Nine Mile Point, Unit 2, Shipped in 1983
Purchase Order NMP2-P304D/12177
Customer Tag Numbers 2CPS*AOV104, 2CPS*AOV106, 2CPS*AOV108, 2CPS*AOV110
Posi Serial Number 19157-3
Qty 4

"Customer/Site Niagara Mohawk Power Corp., Nine Mile Point, Unit 2, Shipped in 1983
Purchase Order NMP2-P304D/12177
Customer Tag Numbers 2GTS*V51, 2GTS*V52,
Posi Serial Number 19157-35
Qty 2

"Customer/Site Union Electric Company, Callaway Plant Unit 1, Shipped in 1985
Purchase Order 14894-M-236-2
Customer Tag Numbers EF-V343, -V344, -V345, -346, GN-V057, -V058, -V059, -V060
Posi Serial Number 35287-01
Qty 8"

Notified R1DO (Welling), R4DO (Rollins), and Part 21/50.55 Reactors Group by email.

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Part 21 Event Number: 52756
Rep Org: CURTISS WRIGHT FLOW CONTROL CO.
Licensee: CURTISS-WRIGHT
Region: 1
City: HUNTSVILLE State: AL
County:
License #: N/A
Agreement: Y
Docket:
NRC Notified By: TONY GILL
HQ OPS Officer: JEFF HERRERA
Notification Date: 05/15/2017
Notification Time: 18:55 [ET]
Event Date: 03/16/2017
Event Time: [CDT]
Last Update Date: 06/09/2017
Emergency Class: NON EMERGENCY
10 CFR Section:
21.21(a)(2) - INTERIM EVAL OF DEVIATION
Person (Organization):
MIKE ERNSTES (R2DO)
PART 21/50.55 REACTO (EMAI)

Event Text

PART 21 - POTENTIAL DEFECT IN GRAYBOOT SOCKET CONTACTS

The following information was received via email:

"This letter is issued to provide an interim notification of a potential defect in certain lots of Grayboot socket contacts supplied with EQ qualified Grayboot Connector Kits. On March 16, 2017, Curtiss-Wright, Nuclear Division, Huntsville Operations was contacted by Georgia Power Vogtle Nuclear Power Plant concerning a potential defect where the socket contact tines were in a relaxed state.

"Although we have completed some testing and verification activities, additional testing is in progress now and will provide necessary information to complete our evaluation. Current testing will be completed and final conclusions made by May 31, 2017.

"At this time, based on test results, evaluations and operating experience, Curtiss Wright is confident that any potentially affected Grayboot Assemblies will continue to perform their intended safety functions. As such, if the final recommendation is to replace the potentially defective socket contact, this can be accomplished during subsequent routine maintenance activities.

"This notification is being made to comply with 60 day interim reporting requirements as defined in 10 CFR 21.21(a)(2).

"For additional information, please contact Samuel Bledsoe, EGS Products Engineering Manager (1-256-690-7852) or Tony Gill, EGS and Trentec Quality Assurance Manager (1-256-426-4558)."

* * * UPDATE PROVIDED BY TONY GILL TO JEFF ROTTON AT 1813 EDT ON 05/31/2017 * * *

The following information was provided via email:

"This letter is issued to provide final findings associated with a potential defect concerning GRAYBOOT socket contacts. This issue was initially identified in an interim report dated May 15, 2017. As documented previously, Curtiss-Wright, Nuclear Division, Huntsville Operations was contacted by Georgia Power Vogtle Nuclear Power Plant on March 16, 2017 concerning a potential defect wherein GRAYBOOT socket contact tines were in a relaxed state. This notification of a potential defect concerns model GB-1 GRAYBOOT kits supplied with two-tined, silver-plated, 12-14 AWG socket contacts.

"Based upon this scope, potentially affected kits/parts are: 1. GB-1(12-14) GRAYBOOT kits, 2. GB-1 (12-14/ 16-18) GRAYBOOT kits, and 3. GB-1-6 GRAYBOOT 12-14 AWG socket contacts.

"This issue does not affect the following: 1. Any GRAYBOOT 'A' kits/parts, 2. Any model GB-2 or GB-3 GRAYBOOT kits/parts, or 3. Any model GB-1 GRAYBOOT kits/parts with 16-18 AWG socket contacts.

"Our evaluation is documented in Report No. EGS-TR-880708-15 and is available for review at our facility in Huntsville, AL. The results identify the most likely root cause is improper heat treating of the socket contacts during manufacturing. Additional testing and analysis was performed to confirm that any affected GRAYBOOT assemblies can still preform their safety-related function and do not present a substantial safety hazard .

"The findings outlined in Report No. EGS-TR-880708-15 provide a high level of confidence that affected GRAYBOOT assemblies do not present a substantial safety hazard. This position is further validated by the lack of negative operating experience over the last 20 plus years from properly installed GRAYBOOT assemblies. However, this condition causes the contact to be more susceptible to damage from handling during connection and disconnection, and therefore the following actions are recommended:

"1. Any affected sockets in inventory should be replaced. Affected sockets in service should be replaced during routine maintenance activities.

OR

"2. In lieu of replacement, it is acceptable to perform the following [steps 1-3] to confirm a separation force greater than 0.19 lbs. This is consistent with existing Curtiss-Wright dedication acceptance criteria. It is recommended that any contacts not meeting this criteria be replaced. 1. Crimp a spare pin contact to an appropriate piece of wire. 2. Connect a force gage or 0.19 lbs. of static weight to the opposite end of the wire. 3. Insert the pin into the socket and confirm that the pin does not separate from the socket under a minimum load of 0.19 lbs.

"To confirm this deviation is not present in existing inventory or in future purchased lots, the following corrective actions have been or will be implemented by Curtiss-Wright: 1. Micro hardness testing was performed on all socket contact lots in inventory to verify their acceptability. Results confirmed that all lots were acceptable. 2. Acceptance criteria for dedication of socket contacts will be revised to include verification of acceptable contact hardness. This corrective action will be completed by June 9, 2017. No dedication of socket contacts will be performed until this corrective action is complete.

"A list of affected utilities and associated purchase orders is being developed and will be complete and submitted by June 9, 2017.

"For additional information, please contact Samuel Bledsoe, EGS Products Engineering Manager (1-256-690-7852) or Tony Gill, EGS and Trentec Quality Assurance Manager (1-256-426-4558)."

Notified R1DO (Bower), R2DO (Shaeffer), R3DO (Daley), R4DO (O'Keefe) and Part 21 Operating Reactors Group via email.

* * * UPDATE AT 1859 EDT ON 06/09/17 FROM TONY GILL TO JEFF HERRERA * * *

The following update was received via email:

"On May 31, 2017, Curtiss-Wright, Nuclear Division, Huntsville Operations issued a letter documenting final findings regarding a potential defect concerning model GB-1 GRAYBOOT kits supplied with two-tined, silver-plated, 12-14 AWG contacts. Please find that letter attached.

"Pursuant to the attached letter, please find attached a list of affected purchase orders.

"For additional information, please contact Samuel Bledsoe, EGS Products Engineering Manager (1-256-690-7852) or Tony Gill, EGS and Trentec Quality Assurance Manager (1-256-426-4558).

List of Sites Affected:
Arkansas Nuclear 1
Bruce Nuclear Power Development
Brunswick
Callaway
Calvert Cliffs
Clinton
Columbia
Cooper
CTEAM/CRIT
Davis-Besse
Diablo Canyon
Duane Arnold
Farley
FMM
Fort Calhoun
Gentilly
Ginna
Haddam Neck
Harris
Indian Point
Kewaunee
La Salle
Limerick
Millstone
Nine Mile Point
North Anna
Oconee
Oyster Creek
Peach Bottom
Pilgrim
Point Beach
Prairie Island
Quad Cities
River Bend
Saint Lucie
San Onofre
Sizewell B
South Texas
Summer
Turkey Point
Vermont Yankee
Vogtle
Waterford
Wolsong
Zion

Notified R1DO(Welling), R2DO(Suggs), R3DO(Orlikowski), R4DO(Rollins), Part-21 Reactors (via email).

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Non-Agreement State Event Number: 52780
Rep Org: KARMANOS CANCER INSTITUTE
Licensee: KARMANOS CANCER INSTITUTE
Region: 3
City: DETROIT State: MI
County:
License #: 21-04127-06
Agreement: N
Docket:
NRC Notified By: JOE RAKOWSKI
HQ OPS Officer: DONG HWA PARK
Notification Date: 06/01/2017
Notification Time: 12:25 [ET]
Event Date: 05/31/2017
Event Time: 12:45 [EDT]
Last Update Date: 06/01/2017
Emergency Class: NON EMERGENCY
10 CFR Section:
35.3045(a)(1) - DOSE <> PRESCRIBED DOSAGE
Person (Organization):
ROBERT DALEY (R3DO)
NMSS_EVENTS_NOTIFICA (EMAI)

Event Text

MEDICAL EVENT - PATIENT RECEIVED LESS DOSE THAN PRESCRIBED

"On May 31, 2017, the licensee's gamma knife model 'C' failed during treatment delivery. The couch retracted from the treatment position at 1245 [EDT] due to a clutch malfunction. Three out of five shots were delivered to the single planned lesion. The prescribed volume received a dose of 15 Gy versus the prescribed 20 Gy. Due to uncertainty regarding repair, the fixation frame was removed from the patient's head. Repair was completed at 1845 [EDT]."

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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Agreement State Event Number: 52783
Rep Org: TEXAS DEPT OF STATE HEALTH SERVICES
Licensee: TEAM INDUSTRIAL SERVICE INC
Region: 4
City: PASADENA State: TX
County:
License #: 00087
Agreement: Y
Docket:
NRC Notified By: ART TUCKER
HQ OPS Officer: DONG HWA PARK
Notification Date: 06/02/2017
Notification Time: 16:19 [ET]
Event Date: 06/01/2017
Event Time: [CDT]
Last Update Date: 06/02/2017
Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
Person (Organization):
NEIL OKEEFE (R4DO)
NMSS_EVENTS_NOTIFIC (EMAI)

Event Text

AGREEMENT STATE REPORT - RADIOGRAPHY SOURCE FAILED TO RETRACT

The following information was received from the State of Texas via email:

"On June 2, 2017, the Agency [Texas Department of State Health Services] was notified by the licensee that one of its radiography crews was unable to retract a 92.5 curie iridium - 192 source into a QSA 880D exposure device. The radiographers had performed several exposures. To perform the radiography, one end of a flexible guide tube was connected to the exposure device and the other end was connected to a stiff guide tube. After completing an exposure, the source hung up in the stiff guide tube and would not retract into the exposure device. The radiographers contacted their management who responded to the site. The stiff guide tube was examined and was found to have several indentations in it that appeared to be preventing the source from passing through the tube. The recovery team straightened out the guide tubes and was able to retract the source into the fully shielded position. The licensee stated the guide tube was not damaged on this date. The licensee has removed the stiff guide tube from service. Additional information will be provided as it is received in accordance with SA 300."

Texas Incident No.: I-9488

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Agreement State Event Number: 52784
Rep Org: ILLINOIS EMERGENCY MGMT. AGENCY
Licensee: PETNET SOLUTIONS, INC.
Region: 3
City: DES PLAINES State: IL
County:
License #: IL-01906-01
Agreement: Y
Docket:
NRC Notified By: GIBB VINSON
HQ OPS Officer: DONG HWA PARK
Notification Date: 06/02/2017
Notification Time: 16:49 [ET]
Event Date: 04/03/2017
Event Time: [CDT]
Last Update Date: 06/02/2017
Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
Person (Organization):
ROBERT DALEY (R3DO)
DANIEL COLLINS (NMSS)
NMSS_EVENTS_NOTIFIC (EMAI)

Event Text

AGREEMENT STATE REPORT - ELEVATED RADIATION LEVELS AT A RADIOPHARMACY

The following information was obtained from the State of Illinois via email:

"On May 10, 2017, PETNET Solutions, Inc., a radiopharmacy with a cyclotron, reported high area dosimetry readings inside their facility with the highest reading being 12,276 mrem for the month of April. Readings were also elevated for the first 10 days of May. There were no neutron readings present which indicates that the readings were likely not from the cyclotron itself but from the radiopharmaceutical production process. PETNET was advised to cease cyclotron operations until their team of engineers and health physicists could determine a cause. Agency Inspectors arrived the next day, May 11, 2017, along with engineers and health physicists from PETNET to conduct an investigation at the property. Interviews with PETNET staff and an inspection of the cyclotron/hot cell did not identify any immediate problems. It was determined during the investigation that non-radiation workers opposite the adjoining wall potentially received a radiation dose during this event. The non-radiation worker involved was informed of the event, and PETNET requested permission to place TLDs and do surveys along their common wall over the next week to assess the situation. This was agreed to. In the interim the licensee estimated that the dose to the workstation in this adjoining office was 154.7 mrem. This exceeds the dose limit in regulation 32 IAC 340.310(a)(3) (10 CFR 20.1301(a)(1) equivalent). However, this dose would not represent any health effects to the individual and would be equivalent to approximately two pelvic X-rays at a doctor's office. Currently, PETNET engineers believe the root cause was related to problems in the targetry and delivery systems resulting in multiple instances of F-18 escaping from a v-vial vent. This material then accumulated in the ventilation ducts leading to the filter banks near the area dosimeters on the wall. PETNET has committed to a corrective action plan including additional maintenance on the unit, an alarming radiation detector in the cyclotron vault connected to the operator's panel and additional area dosimeters on the opposite side of the adjoining wall. IEMA [Illinois Emergency Management Agency] has committed to routine site visits until further notice."

Item number: IL177005

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Agreement State Event Number: 52786
Rep Org: WA OFFICE OF RADIATION PROTECTION
Licensee: GEODESIGN, INC.
Region: 4
City: WILSONVILLE State: OR
County:
License #: WN-I0542-1
Agreement: Y
Docket:
NRC Notified By: JAMES KILLINGBECK
HQ OPS Officer: DONG HWA PARK
Notification Date: 06/02/2017
Notification Time: 19:59 [ET]
Event Date: 05/31/2017
Event Time: [PDT]
Last Update Date: 06/02/2017
Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
Person (Organization):
NEIL OKEEFE (R4DO)
DANIEL COLLINS (NMSS)
NMSS_EVENTS_NOTIFIC (EMAI)

Event Text

AGREEMENT STATE REPORT - PORTABLE GAUGE OPERATOR HANDLED UNSHIELDED SOURCE

The following report was received from the State of Washington via email:

"An operator of a portable moisture/density gauge touched and directly handled the unshielded source.

"Here is how the event occurred: A worker knelt down next to a portable moisture/density gauge and pulled it up, tilted it back, and stuck his face toward the source rod. He then extended the source rod out of the gauge housing and into the air. He then took his left hand and rubbed the source rod clean with his bare hand.

"This event was observed by a manager from Oregon Radiation Protection Services, who then reported the event to the Washington State Department of Health for investigation.

"Incident Number: WA-17-014"

The portable moisture/density gauge is a Troxler 3430 containing 0.333 GBq (9 mCi) of Cs-137 and 1.628 GBq (44 mCi) of Am-241/Be.

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Power Reactor Event Number: 52796
Facility: TURKEY POINT
Region: 2 State: FL
Unit: [3] [ ] [ ]
RX Type: [3] W-3-LP,[4] W-3-LP
NRC Notified By: MITCH GUTH
HQ OPS Officer: STEVE SANDIN
Notification Date: 06/09/2017
Notification Time: 10:17 [ET]
Event Date: 04/10/2017
Event Time: 13:47 [EDT]
Last Update Date: 06/09/2017
Emergency Class: NON EMERGENCY
10 CFR Section:
50.73(a)(1) - INVALID SPECIF SYSTEM ACTUATION
Person (Organization):
LADONNA SUGGS (R2DO)

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

Event Text

INVALID SYSTEM ACTUATION DURING TESTING

"This 60-day telephone notification is being made in accordance with the reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1) to describe an invalid actuation of several safety systems.

"On April 10, 2017 at 1347 hours [EDT] with Unit 3 in Mode 6 during performance of the Train B Engineered Safeguards Integrated Test, safety system actuations occurred prior to the expected point in the test procedure when a loss of continuity resulted while the seismic clips were being removed from a fuse. The actuations were supposed to occur at a subsequent step when the fuse was to be pulled to actuate the Hi Containment Pressure signal. As a result, the following equipment actuated: 3B, 4A and 4B High Head SI pumps; 3B Containment Spray pump; Containment Isolation and Containment Ventilation Isolations; 3A and 3B Emergency Diesel Generators; Emergency Containment Coolers. Because an actual high containment pressure signal did not exist at the time of the actuation, the actuation is considered invalid. All equipment responded as expected.

"The NRC Resident Inspector has been notified."

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Power Reactor Event Number: 52797
Facility: SUSQUEHANNA
Region: 1 State: PA
Unit: [1] [2] [ ]
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: RONALD FRY
HQ OPS Officer: JEFF HERRERA
Notification Date: 06/09/2017
Notification Time: 12:51 [ET]
Event Date: 06/09/2017
Event Time: 05:09 [EDT]
Last Update Date: 06/09/2017
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(C) - POT UNCNTRL RAD REL
Person (Organization):
BLAKE WELLING (R1DO)

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

Event Text

EXHAUST FAN BREAKER FAILURE CAUSING LOW SECONDARY CONTAINMENT DIFFERENTIAL PRESSURE

"On June 9, 2017 at 0509 [EDT], Secondary Containment Zone 3 (Unit 1 and 2 Reactor Building) differential pressure lowered to 0 [inches] WG [water gauge] during a routine restoration due to equipment failure. One of the two Unit 1 Zone 3 exhaust fan breakers experienced a failure that during procedural restoration caused Secondary Containment Zone 3 to experience a positive differential pressure. Required differential pressure per SR 3.6.4.1.1 could not be maintained.

"Zone 3 differential pressure was recovered to [greater than] 0.25 [inches] WG following restart of Unit 2 Zone 3 Secondary Containment fans. All other Zones of Secondary Containment were unaffected by this event.

"This event is being reported under 10 CFR 50.72(b)(3)(v)(C) and per the guidance of NUREG 1022 Rev 3 section 3.2.7 as a loss of a Safety Function. There is no redundant Susquehanna Secondary Containment System.

"The NRC Resident Inspector has been notified."

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Part 21 Event Number: 52798
Rep Org: ENGINE SYSTEMS, INC
Licensee: ENGINE SYSTEMS, INC
Region: 1
City: ROCKY MOUNT State: NC
County:
License #:
Agreement: Y
Docket:
NRC Notified By: TOM HORNER
HQ OPS Officer: BETHANY CECERE
Notification Date: 06/09/2017
Notification Time: 17:00 [ET]
Event Date: 06/08/2017
Event Time: [EDT]
Last Update Date: 06/09/2017
Emergency Class: NON EMERGENCY
10 CFR Section:
21.21(d)(3)(i) - DEFECTS AND NONCOMPLIANCE
Person (Organization):
LADONNA SUGGS (R2DO)
JESSE ROLLINS (R4DO)
PART 21/50.55 REACTO (EMAI)

Event Text

PART 21 REPORT - DEFECTIVE MOTOR OPERATED POTENTIOMETER

The following information is excerpted from an Engine Systems, Inc. (ESI) report received via fax:

"Component: Motor operated potentiometer without pre-position feature
P/Ns: 72-07900-100-ESI
2-04E-184-103-ESI
2-00G-009-002-ESI

"System: Emergency Diesel Generator

"Summary:
Engine Systems Inc. (ESI) began a 10CFR21 evaluation on April 12, 2017 following correspondence with Duke Energy - Catawba Nuclear Plant of a potential issue with a motor operated potentiometer (MOP) supplied by ESI. During bench testing, Duke was unable to adjust the cams of the MOP to operate the pre-position feature (also referred to as return-to-center). Further investigation revealed that the MOP supplied to Duke did not contain the desired cam arrangement.

"The evaluation was concluded on June 8, 2017 and it was determined that this issue is a reportable defect as defined by 10CFR21. For installations that rely on the pre-position function to drive generator output voltage to a predetermined setpoint in an emergency event, an incorrect cam arrangement on the motor operated potentiometer could negatively affect the emergency diesel generator [EDG] set's output voltage. This may prevent the EDG from carrying its safety-related loads in an emergency event.

"Discussion:
A motor operated potentiometer is commonly used in the voltage regulator system of emergency diesel generator sets to allow for local or remote adjustment of generator output voltage. The assembly supplied by ESI was a replacement for obsolete Portec/NEI part number 72-07900-100. ESI initiated a Supplier Deviation Notification [SDN] (Part Number Change) to replace the obsolete part with one that would serve as a functional replacement using the same mounting footprint. To accomplish this, a Basler brand MOP was selected. The "-ESI" suffix was added to designate a modification to the Basler MOP by replacing the standard 2W, open-wound potentiometer with a 3W, precision-wound potentiometer identical to the one used in the original Portec units. The SDN indicated that the replacement MOP would have the pre-position feature. Note that the SDN would later be revised to include replacements for Cooper-Bessemer part numbers 2-04E-184-103 and 2-00G-009-002 which represented the same Portec part number 72-07900-100.

"The Basler MOP selected by ESI for the assembly was model MOC2103. A review of Basler model MOC2103 indicates that it is not a return-to-center style device and instead contains a standard cam arrangement. Photos on the following page depict the differences between the two cam arrangement styles. ESI inadvertently selected the incorrect model for this application. Reviewing the qualification testing performed by ESI revealed that ESl's testing duplicated the original Portec test procedure (whereby one cam is set to close above setpoint and one cam is set to close below setpoint). However, the testing was shown to be inadequate to fully ensure the return-to-center functionality. Though the cams were properly set above and below setpoint, a further verification that the unit would return to center, even if sufficiently away from setpoint, was not performed. For installations that rely on the preposition function, it is possible that the MOPs supplied by ESI would not return the potentiometer to setpoint. Since the MOP is used to set output voltage of the emergency diesel generator, the consequences of this issue are that the EDG may not maintain voltage within its required range to support its safety-related loads.

"Root cause evaluation:
ESI selected the incorrect MOP to be used for this assembly. The original MOP used was a Portec/NEI P/N 72-07900-100 which became obsolete in the 1990s. ESI offered a replacement assembly (using the "-ESI" suffix) that included a Basler brand MOP, modified by replacing the factory supplied potentiometer to replicate the original Portec design. The original Portec MOP had the return-to-center feature and Basler offers some models with return-to-center and some without. ESI inadvertently selected the model without return-to-center.

"As a secondary cause, ESl's functional testing did not adequately verify the MOPs ability to return to center. The test procedure has steps to set the cams and limit switches as would be performed for the return-to-center style; however, the testing did not ensure that the cams would return the potentiometer back to the center, regardless of starting position.

[photos submitted with report with following captions:]

"Photo 1: Desired Return-to-Center Cam Arrangement Photo 2: As-Supplied Standard Cam Arrangement
Photos 3&4: Return-to-Center Cam Profile & Arrangement
Photo 5&6: Standard Cam Profile & Arrangement

"Evaluation of previous shipments:
ESI has supplied the replacement MOP from 2009 to 2016 on five (5) separate safety-related orders. A review of each order Indicates that the incorrect MOP was selected for the assembly.

"Affected Customers:
ESI has supplied motor operated potentiometers with the incorrect cam arrangement to the following customers.

"Part Number 72-07900-100-ESI; Customer Duke Energy - Catawba; Customer PO 196046; Qty 6; ESI Serial Numbers 3013905-1.1-1 thru 6; MOP Serial Numbers H01931980, H01995582, H01995584, H01995585, H01995586, H01995587; C-of-C Date 5/30/2015

"Part Number 72-07900-100-ESI; Customer Dremel/CFE - Laguna Verde; Customer PO 14-011/700427979; Qty 1; ESI Serial Number 3011566-3.1-1; MOP Serial Number H01798043; C-of-C Date 2/13/2015

"Part Number 2-04E-184-103-ESI; Customer APS - Palo Verde; Customer PO 500530279; Qty 3; ESI Serial Numbers 3005695-1.1-1 thru 3; MOP Serial Numbers: H00958610, H00961219, 333861; C-of-C Date 7/31/2009

"Part Number 2-04E-184-103-ESI; Customer APS - Palo Verde; Customer PO 500555659; Qty 3; ESI Serial Numbers 3008341-1.1-1 thru 3; MOP Serial Numbers H01151813, H01161548, H01161550; C-of-C Date 212/2012

"Part Number 2-00G-009-002-ESI; Customer South Texas Project; Customer PO 200410; Qty 2; ESI Serial Numbers 3015607-1.1-1 thru 2; MOP Serial Numbers H02027435, H02027436; C-of-C Date 8/19/2016

"Total Quantity 15

"Corrective Action:
Customers that have installed the motor operated potentiometer must evaluate the impact on their specific system. It may be acceptable to continue to use the MOP if other provisions are in place to ensure the EDG voltage is within an acceptable range for safety-related operation. Credit may also be taken for site acceptance testing following installation of the MOP. If the MOP is deemed acceptable for the application, no further action is necessary.

"For those customers that have determined the MOP is unacceptable for their application and/or would simply rather ensure they have the correct MOP, it may be returned to ESI for warranty replacement. Once a warranty order and Return Goods Authorization (RGA) is issued, supply of safety-related replacement MOP(s) is expected to be complete within 90 days.

"To prevent recurrence of this issue, ESI is implementing a revision to the procurement process for this assembly to update the part number of the MOP utilized. In addition, the dedication inspection and testing activities are being revised to ensure testing properly validates the return-to-center feature. This action is currently underway and will be completed prior to any future shipments."

If you have any questions, you may call:

Tom Horner
Quality Assurance Manager
Tel: (252) 977-2720

ESI Report ID: 10CFR21-0117, Rev. 0, dated 06/09/17

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Power Reactor Event Number: 52800
Facility: CLINTON
Region: 3 State: IL
Unit: [1] [ ] [ ]
RX Type: [1] GE-6
NRC Notified By: SCOTT LABUNSKI
HQ OPS Officer: JEFF HERRERA
Notification Date: 06/11/2017
Notification Time: 01:00 [ET]
Event Date: 06/10/2017
Event Time: 22:56 [CDT]
Last Update Date: 06/11/2017
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
Person (Organization):
ROBERT ORLIKOWSKI (R3DO)

Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
1 M/R Y 99 Power Operation 0 Hot Shutdown

Event Text

MANUAL REACTOR SCRAM DUE TO LOSS OF FEEDWATER HEATING

At 2256 CDT on 6/10/17, Clinton operators manually scrammed the reactor from 99 percent power due to a loss of feedwater heating. The scram was uncomplicated and the plant is stable and in mode 3. All rods inserted and decay heat is being removed by the condenser. All offsite power is available. The cause of the loss of feedwater heating is under investigation.

The NRC Resident Inspector and the State of Illinois have been notified.

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Power Reactor Event Number: 52801
Facility: INDIAN POINT
Region: 1 State: NY
Unit: [ ] [3] [ ]
RX Type: [2] W-4-LP,[3] W-4-LP
NRC Notified By: CLIFTON GATES
HQ OPS Officer: JEFF ROTTON
Notification Date: 06/11/2017
Notification Time: 14:35 [ET]
Event Date: 06/11/2017
Event Time: 09:11 [EDT]
Last Update Date: 06/11/2017
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(A) - POT UNABLE TO SAFE SD
Person (Organization):
BLAKE WELLING (R1DO)

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

Event Text

EXCESSIVE BONNET LEAKAGE FROM CHEMICAL VOLUME CONTROL SYSTEM VALVE

"At Indian Point Energy Center [IPEC], Unit 3 normal letdown was isolated due to a significant body/bonnet leak on the inlet valve to Reactor Coolant Filter. Shift team took action to isolate letdown and stop the leak. The Abnormal Operating Procedure (AOP) was entered, normal letdown was isolated per procedure, and excess letdown was placed in service to balance inventory at 61 percent Pressurizer Level. This was above the Technical Specification [3.4.9] level of 54.3 percent which was exceeded at 0911 [EDT], putting the unit in a 6-hour shutdown action statement. The valve body/bonnet was torqued, successfully eliminating the leakage. Pressurizer Level was restored to the normal control band and the AOP was exited at 1136. The SM [Shift Manager] estimated the leakage at 18 gallons per minute when leak was active. No EAL [Emergency Action Level] thresholds were exceeded. This occurrence is considered a safety system functional failure per 10 CFR 50.72(b)(3)(v)(A), requiring an 8-hour NRC report. Both IPEC units are stable at full power."

The licensee has notified the NRC Resident Inspector and the New York State Public Service Commission.

Page Last Reviewed/Updated Monday, June 12, 2017
Monday, June 12, 2017