Source: https://www.nrc.gov/reading-rm/doc-collections/event-status/event/2017/20170206en.html
Timestamp: 2019-04-24 08:41:56
Document Index: 200514653

Matched Legal Cases: ['art 21', 'ART 21', 'art 21', 'art 21', 'art 21', 'art 21']

NRC: Event Notification Report for February 6, 2017
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Event Notification Report for February 6, 2017
02/03/2017 - 02/06/2017
50900 52499 52510 52512 52514 52515 52531 52534
Part 21 Event Number: 50900
City: EAST FARMINGDALE State: NY
INTERIM PART 21 REPORT - POTENTIAL TEST INDUCED DEFECT IN A 0867F MAIN STEAM SAFETY RELIEF VALVES
The following report was received from Curtiss - Wright via email:
"This letter provides interim notification of a potential test induced defect in a 0867F Series Main Steam Safety Relief Valves (MS-SRVs) manufactured and supplied by Target Rock (TR). The information required for this notification is provided below:
"(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 0867F Series of Main Steam-Safety Relief Valves Manufactured by Target Rock. This is a 3-stage piloted valve consisting of a main valve (the 'Main') with an actuator mounted to it (the 'Topworks'). The 0867F is the latest generation of the 67F line of MS-SRVs, including the original 3-Stage and 2-Stage designs, and this product line has over 40 years of plant operational experience. Only the 0867F is under investigation. This is due to the differences between the 0867F design and the other designs.
As we understand it, the Pilgrim Station recently manually opened the Target Rock Main Steam Safety Relief Valves (MS-SRVs) as part of cooling down the reactor following a loss of offsite power. One of the four installed MS-SRVs may not have fully opened. As-found steam testing of the affected MS-SRV did not duplicate this failure; the valve opened on demand. However, the valve did not re-close as expected. Internal inspections found damaged parts in the main stage subassembly that could potentially affect the ability of the MS-SRV to operate as designed.
We are investigating potential root causes for this damage. However, we are still unable to determine if a specific defect exists. GE SIL-196, Supplement 17 determined Main Spring relaxation was caused by 'extreme dynamics encountered during limited flow testing . Valve dynamics under full flow conditions (i.e. discharge not gagged) are much less severe than those under limited flow conditions.' These extreme dynamics, under limited flow test conditions, are the focus of our investigation. Specific areas of investigation include;
a) Testing of materials to verify they are consistent with our material specifications,
b) evaluation of differences between the 0867F and earlier designs, and
c) evaluation of the differences between different limited flow test loop configurations and test procedures
The Pilgrim event occurred on January 27, 2015. As-found testing occurred on February 2, 2015.
While we have yet to determine if a specific defect exists, the following plants were supplied 0867F MS-SRVs:
- Pilgrim (Model 09J-001) Quantity Shipped = 8
- Fitzpatrick (Model 09H-001) Quantity Shipped = 4, Quantity on order= 8
- Hatch 1 and 2 (Model 09G-001) Quantity Shipped= 24, Quantity on order= 12
The following plants will be supplied 0867F MS-SRVs:
- Hope Creek (Models 14J-001, 14J-002) Quantity on order = 7
The root cause of the potential test induced defect has not yet been confirmed as of the date of this report. Therefore, no specific corrective actions have been initiated. Target Rock Problem Report 080 will document the corrective actions when they are determined and complete the 10 CFR Part 21 evaluation of the potential test induced defect. 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.
We are working with all three (4) sites to identify appropriate precautions.
"Should you have any questions regarding this matter, please contact Michael Cinque, Director of Program Management at (631 ) 293-3800."
* * * UPDATE FROM JOHN DeBONIS (VIA EMAIL) TO HOWIE CROUCH AT 1355 EDT ON 5/1/15 * * *
Curtiss-Wright provided an update to state that their root cause analysis is still in progress and they anticipate completion within 60 days.
Notified NRR Part 21 Group (via email), R1DO (Gray), and R2DO (Ehrhardt).
* * * UPDATE FROM JOHN DeBONIS (VIA EMAIL) TO STEVEN VITTO AT 1256 EDT ON 6/30/15 * * *
Curtiss-Wright provided an update to state their root cause analysis findings and corrective actions. Corrective actions are estimated to be completed within 12 months.
"The following plants were supplied 0867F MS-SRVs:
Pilgrim (Model 09J-001) Quantity Shipped = 8
FitzPatrick (Model 09H-001) Quantity Shipped = 4, Quantity on order= 8
Hatch 1 and 2 (Model 09G-001) Quantity Shipped = 24, Quantity on order= 12
"The following plants will be supplied 0867F MS-SRVs:
Hope Creek (Models 14J-001, 14J-002) Quantity on order = 7
"Valves Currently Installed
"Target Rock recommends valves currently installed be inspected to ensure the main piston shoulder has contact with the main disc stem shoulder. These inspections should be scheduled based on plant-specific indications of the potential for fretting. These inspections can be performed by removing the base assembly from the main body and physically measuring for shoulder-to-shoulder contact.
Notified NRR Part 21 Group (via email), R1DO (Dimitriadis), and R2DO (Suggs).
* * * UPDATE AT 0803 EST ON 02/03/17 FROM JOHN DEBONIS TO JEFF HERRERA * * *
"Target Rock, a business unit of Curtiss-Wright Flow Control Corporation (TR), previously submitted NID# 15428 (Ref. 1) regarding 0867F Series of Main Steam Safety Relief Valves (MSSRV). The purpose of this letter is to provide a final update on the status of the actions identified in NID# 15428 (Ref. 1 ).
"TR has developed a design change that we have verified, though analysis and qualification testing, ensures testing on the available limited flow test facilities will not impart damage to our product. Qualification included both limited and full flow testing. Target Rock confirms this design change has no effect on either National Board certification or the performance required by the applicable Specification to support the plant safety analysis.TR will offer this design change as our recommended long term solution to all utilities who currently have installed or plan to install the 0867F Series Main Steam Safety Relief Valve model in their respective plants. Should you have any questions regarding this matter, please contact me [Alex DiMeo] at (631) 293-3800."
Notified R1DO (Gray), R2DO (Walker) and Part 21 Group.
Non-Agreement State Event Number: 52499
Rep Org: GEO ENGINEERING & TESTING INC
Licensee: GEO ENGINEERING & TESTING INC
City: TAMUNING State: GU
License #: 56-18173-02
NRC Notified By: UKRIT SIRIPRUSANAN
HQ OPS Officer: MARK ABRAMOVITZ Notification Date: 01/20/2017
Event Date: 01/01/1988
Event Time: [GST]
LOST/MISPLACED MOISTURE DENSITY GAUGE
A Troxler moisture density gauge was last leak checked in 1988. The gauge handle was damaged by construction equipment and put aside sometime between 1988 and 1990. The current location of the gauge is unknown.
Troxler Model 3411B
Serial Number 8117
Sources: Cs-137 10 mCi and Am-241 60 mCi
* * * RETRACTED ON 0036 EST ON 2/3/17 FROM QUOC VO TO JEFF HERRERA * * *
"At the time of discovery HPCS was out of service in accordance with plant technical specifications, therefore the failure of the supporting equipment, DMA-FN-32, is not reportable under 10 CFR 50.72(b)(3)(v)(D).
Fuel Cycle Facility Event Number: 52512
HQ OPS Officer: DONG HWA PARK Notification Date: 01/26/2017
A contract employee had a prohibited item in the Protected Area. The employee's access to the site has been restricted.
Agreement State Event Number: 52514
Licensee: CHILDREN'S HOSPITAL OF CHICAGO - MEDICAL CENTER
AGREEMENT STATE REPORT - UNPLANNED CONTAMINATION AT HOSPITAL
"On January 25, the licensee's Radiation Safety Officer (RSO) contacted the Agency [Illinois Emergency Management Agency] to report an issue with an administration of a capsule containing I-131 which had occurred the previous afternoon. A nominal dose of 30 milliCi in capsule form was given to a child within the nuclear medicine department of the licensee's facility. Although the patient was being treated on an outpatient basis, the licensee was keeping the patient for a short time to ensure there would be no complications before being sent home. During this period, staff checked in on the patient several times and during one of the visits, discovered that rather than swallowing the capsule as instructed, the patient had spit the capsule out into their hand and was hiding the capsule. This resulted in extensive contamination of the patient's hand, clothing and the chair they were sitting in as well as the immediate surrounding area. During the process to evaluate and decontaminate the patient, additional contamination was discovered in adjacent camera rooms and corridors where the staff had traversed. Staff moved to close the department and restrict passage into/out of the nearby areas to prevent additional spread of contamination. Initial estimates suggest that the patient ingested little if any of the activity and that excessive levels were throughout the area of the nuclear medicine department. Based on this finding, barriers were erected and the department was closed for over 48 hours while assessment and decontamination efforts were ongoing.
"Agency inspectors were at the site on January 26 to perform assessments of exposure, contamination levels, potential uptake by staff and corrective action being taken by the licensee. This matter will remain open while those assessments are on going. Initial bioassay results suggest only negligible uptakes have occurred with staff. Potential exposures/uptakes continue to be evaluated by the licensee throughout the decontamination process. The licensee is exploring the potential for having additional outside resources complete the necessary decontamination and remediation steps so that the department can reopen and provide at least limited services."
Illinois Item Number: IL17002
Agreement State Event Number: 52515
Licensee: EASTMAN KODAK COMPANY
License #: C1347
AGREEMENT STATE REPORT - FAILED SEALED SOURCE LEAK TEST
"On November 1, 2016, Eastman Kodak Company informed the Department [New York State Department of Health] that an NRD Model A-2003 static eliminator sealed source leak test result indicated a measurable contamination of 0.00661 microCuries. The device was taken out of service and another leak test was performed on the unit. The second leak test results showed only 0.0003 microCuries."
New York Event Report ID No: NYDOH - NY-16-09
Power Reactor Event Number: 52531
Event Time: 14:58 [EST]
UNANALYZED CONDITION - APPENDIX R FIRE ANALYSIS
"In preparation for transitioning the Plant Hatch Fire Protection Licensing Basis from 10CFR50.48(b) (Appendix R) to 10CFR50.48(c) (NFPA 805), a weak-link and operator manual action (OMA) analysis for Information Notice (IN) 92-18 type hot shorts on motor-operated valves (MOVs) was performed to support the Plant Hatch Appendix R Safe Shutdown Analysis. As a result of the analysis, it was identified that cable impacts can bypass valve limit and torque switches, resulting in physical damage to valves required for Safe Shutdown. This would prevent the valves from being operated locally or being operated from the remote shutdown panel. These cable failures could also cause the valve motors to fail. This updated analysis has identified circuit configuration deficiencies in Fire Areas 0024 A & C (Main Control Room & Cable Spread Room), 1203F (U1 Reactor Building SE), 1205F (U1 Reactor Building NE), and 2203F (U2 Reactor Building NE). Therefore, due to the identified deficient conditions, it was determined that in the event of a postulated fire in the affected fire areas, the paths of safe shutdown on the affected unit(s) could be compromised and impact the ability to achieve safe shutdown conditions.
"Compensatory measures were already in place in accordance with the plant's Fire Hazard Analysis (FHA) as a result of previous conditions in these same fire areas. The presence of the compensatory measures in addition to automatic fire detection in these fire areas ensure that the safe shutdown paths are preserved until the degraded conditions can be repaired.
"CRs 10326399, 10326401, 10326402, 10326404 and 10326405"
The unanalyzed condition is applicable to 10CFR50.48(b) Appendix R and not to 10CFR50.48(c) (NFPA 805).
Power Reactor Event Number: 52534
HQ OPS Officer: STEVE SANDIN Notification Date: 02/03/2017
UNIT 1 MANUAL REACTOR TRIP DUE TO LOOP 1 MSIV STARTING TO FAIL CLOSED
"At 1545 EST on 2/3/17, Vogtle Unit 1 was manually tripped from 100% power when loop 1 Main Steam Isolation Valve (MSIV) started to fail closed. Non-Safety Related 4160V bus 1NA01 failed to transfer to alternate incoming power supply automatically and was successfully manually energized.
"All control rods fully inserted and AFW [Auxiliary Feedwater] and FWI [Feedwater Isolation] actuated as expected.
"Unit 1 is in Mode 3 and stable with decay heat being removed by AFW."