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Event Notification Report for October 01, 2019

U.S. Nuclear Regulatory Commission
Operations Center

Event Reports For
9/30/2019 - 10/1/2019

** EVENT NUMBERS **


54130 54290 54299 54300 54301 54302

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!!!!! THIS EVENT HAS BEEN RETRACTED.THIS EVENT HAS BEEN RETRACTED !!!!!
Power Reactor Event Number: 54130
Facility: FITZPATRICK
Region: 1     State: NY
Unit: [1] [] []
RX Type: [1] GE-4
NRC Notified By: JOHN WALKOWIAK
HQ OPS Officer: THOMAS KENDZIA
Notification Date: 06/24/2019
Notification Time: 21:18 [ET]
Event Date: 06/24/2019
Event Time: 18:15 [EDT]
Last Update Date: 09/30/2019
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
Person (Organization):
DONNA JANDA (R1DO)
Unit SCRAM Code RX Crit Initial PWR Initial RX Mode Current PWR Current RX Mode
1 N Y 100 Power Operation 100 Power Operation

Event Text

POTENTIAL UNANALYZED CONDITION DUE TO UNPROTECTED CONTROL CIRCUITS RUNNING THROUGH MUTILPLE FIRE AREAS

"During a review of industry Operating Experience it was identified that there were unprotected DC control circuits for non safety-related DC motors which are routed from the Battery Charger Rooms to other separate fire areas. Circuit Breakers used to protect the motor power conductors appear to be inadequate to protect the control conductors. The concern is that under fire safe shutdown conditions, it is postulated that a fire in one area can cause short circuits potentially resulting in secondary fires or cable fires in other fire areas where the cables are routed. The secondary fires or cable failures are outside the assumptions of the 10 CFR 50 Appendix R Safe Shutdown Analysis.

"This condition is reportable as an 8-hour ENS report in accordance with 10 CFR 50.72(b)(3)(ii)(B) as an unanalyzed condition. Requirements of the Technical Requirements Manual (TRM) for the affected fire areas will be implemented."

The licensee has notified the NRC Resident Inspector.


* * * RETRACTION FROM ROBERT GRAHAM TO HOWIE CROUCH AT 2045 EDT ON 9/30/19 * * *

"In accordance with NUREG-1022, Sections 2.8 and 5.1.2, James A. FitzPatrick Nuclear Power Plant is retracting (formally withdrawing) Licensee Event Report (LER) Number 2019-002.

"LER 2019-002 was transmitted to the NRC via letter JAFP-19-0080 dated August 23, 2019. The LER reported, under 10 CFR 50.73(a)(2)(ii)(B), the nuclear power plant being in an unanalyzed condition that significantly degraded plant safety.

"Subsequent to submittal of LER 2019-002, FitzPatrick Engineering completed analyses using more accurate input conditions. This analysis has determined no credible hot short scenario will result in damage to adjacent cables in other fire zones, showing that the postulated condition would not degrade plant safety. Therefore, James A. FitzPatrick Nuclear Power Plant is retracting LER 2019-002 [and this event notification]."

The licensee will notify the NRC Resident Inspector and the New York State Public Service Commission.


Notified R1DO (DeFrancisco).

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Agreement State Event Number: 54290
Rep Org: CALIFORNIA RADIATION CONTROL PRGM
Licensee: CONSTRUCTION TESTING SERVICES, INC.
Region: 4
City: ROCKLIN   State: CA
County:
License #: 8043-34
Agreement: Y
Docket:
NRC Notified By: K. ARUNIKA HEWADIKARAM
HQ OPS Officer: OSSY FONT
Notification Date: 09/23/2019
Notification Time: 15:31 [ET]
Event Date: 09/20/2019
Event Time: 00:00 [PDT]
Last Update Date: 09/23/2019
Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
Person (Organization):
VINCENT GADDY (R4DO)
NMSS_EVENTS_NOTIFICATION (EMAIL)
- CNSNS (MEXICO) (EMAIL)
ILTAB (EMAIL)
This material event contains a "Less than Cat 3" level of radioactive material.

Event Text

AGREEMENT STATE REPORT - STOLEN MOISTURE DENSITY GAUGE

The following was received from the California Department of Public Health - Radiologic Health Branch (RHB) via email:

"On 09/21/19, (Saturday), the RSO [radiation safety officer] contacted RHB and left a voicemail message to report a stolen moisture density gauge. The gauge involved is a Troxler Model 3440, S/N 28050 containing 9 mCi of Cs-137 and 44 mCi of Am-241. The gauge was stolen on 09/20/2019, between the hours of 2030 PDT and 0130 PDT, while it was charging at the user's residence. Union City Police was notified of the incident (Case # 190921007, Officer # 4740, Sarah Lings). Licensee will be posting a reward for the safe return of the gauge.

"Licensee failed to contact CA Office of Emergency Services (OES) regarding immediate notification. RHB will be following up on the incident."

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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Power Reactor Event Number: 54299
Facility: FERMI
Region: 3     State: MI
Unit: [2] [] []
RX Type: [2] GE-4
NRC Notified By: PAUL ANGOVE
HQ OPS Officer: DONALD NORWOOD
Notification Date: 09/30/2019
Notification Time: 01:22 [ET]
Event Date: 09/29/2019
Event Time: 22:28 [EDT]
Last Update Date: 09/30/2019
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(C) - POT UNCNTRL RAD REL
Person (Organization):
ROBERT DALEY (R3DO)
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

SECONDARY CONTAINMENT PRESSURE EXCEEDED TECHNICAL SPECIFICATION REQUIREMENT

"On September 29, 2019 at 2228 EDT, during a planned swap of Reactor Building HVAC trains, the exhaust fan discharge damper for the train being removed from service failed to close when the train was shutdown, which resulted in the Technical Specification (TS) for secondary containment pressure not being met for approximately 2 minutes and 15 seconds. The maximum secondary containment pressure observed during that time was approximately 0.1 inches of water gauge (positive).

"Secondary containment pressure was returned to within the TS operability limit of greater than or equal to 0.125 inches of vacuum water gauge (TS SR 3.6.4.1.1) by restarting the train of RBHVAC. Secondary containment pressure is currently stable. Secondary containment was declared Operable at 2235 EDT. There were no radiological releases associated with this event.

"Declaring secondary containment inoperable is reportable under 10 CFR 50.72(b)(3)(v)(C) as an event or condition that could have prevented the fulfillment of a safety function needed to control the release of radioactive material.

"The Licensee has notified the NRC Resident Inspector."

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Power Reactor Event Number: 54300
Facility: BROWNS FERRY
Region: 2     State: AL
Unit: [1] [] []
RX Type: [1] GE-4,[2] GE-4,[3] GE-4
NRC Notified By: LaGRANT MAYE
HQ OPS Officer: DONALD NORWOOD
Notification Date: 09/30/2019
Notification Time: 10:28 [ET]
Event Date: 07/31/2019
Event Time: 16:50 [CDT]
Last Update Date: 09/30/2019
Emergency Class: NON EMERGENCY
10 CFR Section:
50.73(a)(1) - INVALID SPECIF SYSTEM ACTUATION
Person (Organization):
BINOY DESAI (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

Event Text

60-DAY OPTIONAL TELEPHONIC NOTIFICATION DUE TO AN INVALID ACTUATION OF A CONTAINMENT ISOLATION SIGNAL

"This 60-day telephone notification is being made per 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 a general containment isolation signal affecting more than one system.

"On July 31, 2019, at approximately 1650 hours Central Daylight Time (CDT), Browns Ferry Nuclear Plant (BFN), Unit 1 experienced a Primary Containment Isolation System (PCIS) Group 6 isolation during performance of surveillance procedure 1-SR-3.3.6.2.3(A), Reactor/Refueling Zone Ventilation Radiation Monitor 1-RM-90-140/142 Calibration and Functional Test. The Group 6 isolation caused the initiation of Standby Gas Treatment (SBGT) Trains A, B, and C, and Control Room Emergency Ventilation (CREV) subsystem B. Unit 1 H2O2 Analyzer and Drywell Radiation Monitor CAM, 1-RM-90-256, were declared Inoperable and Technical Specifications (TS) Limiting Condition for Operation (LCO) 3.4.5 Condition B was entered. All affected safety systems responded as expected.

"Plant conditions which initiate PCIS Group 6 actuations are Reactor Vessel Low Water Level (Level 3), High Drywell Pressure, or Reactor Building Ventilation Exhaust High Radiation (Reactor Zone or Refuel Zone). At the time of the event, these conditions did not exist; therefore, the actuation of the PCIS was invalid.

"This condition was the result of two cleared fuses in the alarm logic. The apparent cause is a ground fault on the A6 Open Drain Input/Output Module.

"There were no safety consequences or impact to the health and safety of the public as a result of this event.

"This event was entered into the Corrective Acton Program as Condition Report 1537358.

"The NRC Resident Inspector has been notified of this event."

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Part 21 Event Number: 54301
Rep Org: FLOWSERVE US INC.
Licensee: FLOWSERVE US INC
Region: 1
City: RALEIGH   State: NC
County:
License #:
Agreement: Y
Docket:
NRC Notified By: MEGAN STRONG
HQ OPS Officer: KERBY SCALES
Notification Date: 09/30/2019
Notification Time: 15:55 [ET]
Event Date: 09/30/2019
Event Time: 00:00 [EDT]
Last Update Date: 09/30/2019
Emergency Class: NON EMERGENCY
10 CFR Section:
21.21(d)(3)(i) - DEFECTS AND NONCOMPLIANCE
Person (Organization):
ANNE DeFRANCISCO (R1DO)
BINOY DESAI (R2DO)
HIRONORI PETERSON (R3DO)
DAVID PROULX (R4DO)
- PART 21/50.55 REACTORS (EMAIL)

Event Text

PART 21 - VALVE MANUFACTURE ACCEPTANCE CRITERIA

The following information was received from Flowserve US Inc. via facsimile:

"Description: Contrary to the requirements of ASME Section Ill - NC-4000, Flowserve Raleigh identified that they were utilizing as standard practice, the base material acceptance criteria in lieu of welding acceptance criteria for valves with temporary attachments (i.e. - Lug removal areas). This utilization of criteria has been ongoing as far back as Flowserve's Review could determine. No specific orders or customers are identified as this is systemic to the overall process of valve manufacturing at Flowserve Raleigh.

"Evaluation: A review was completed of the ASME Code requirements by Flowserve Raleigh's Engineering and Metallurgical Process Control Departments with the following results:

"The examination of Temporary Attachment Removal Areas, to the NB/NC-2540 Examination and Repair of Forgings and Bars, and NB/NC-2570 Examination and Repair of Statically and Centrifugally Cast Products is contrary to NB/NC-5340 and 5350 acceptance criteria. However, it can be determined that Temporary Attachment Removal Areas examined to NB/NC-2500 acceptance criteria is consistent with the acceptable surface condition resulting from a welded repair performed on the same material product form. No greater risk to pressure integrity is created by the examination of Temporary Attachment Removal Areas to NB/NC-2500 acceptance criteria. The examination of Temporary Attachment Removal Areas examined to NB/NC-2500 acceptance criteria does not result in a Risk to safety relating to pressure integrity.

"Paragraph NB/NC-4435 of Article NB/NC-4000 FABRICATION AND INSTALLATION contains mandatory requirements for the examination of Components. Contrary to the requirement of NB/NC-4435 (b) (3) to examine the Nonstructural Temporary Attachment Removal Area in accordance with the Acceptance Criteria of NB/NC-5340 or NB/NC-5350 Flowserve performed these examinations in accordance with Article NB/NC-2000 in accordance with the Acceptance Standards of NB/NC-2500 for the applicable Material Product Form. NB/NC-2540 Examination and Repair of Forgings and Bars, NB/NC-2570 Examination and Repair of Statically and Centrifugally Cast Products.

"ND-4435 contains no mandatory requirements for the examination of Nonstructural Temporary Attachment Removal Area.

"Extent of Condition: This utilization of criteria has been ongoing as far back as Flowserve's Review could determine. No specific orders or customers are identified as this is systemic to the overall process of valve manufacturing at Flowserve Raleigh.

"Corrective Actions: Flowserve Raleigh Corrective Action, (CAR-393758) has been issued, and is currently in process of determining root cause and preventive action measures.

"Summation: After review by Flowserve Raleigh's Engineering and Metallurgical Process Control Departments. It is the position of Flowserve Raleigh, that in accordance with the provisions of 10 CFR Part 21, this condition, while reportable to the NRC (Nuclear Regulatory Commission), is not a significant/substantial safety hazard."

Megan Strong
Quality Manager
office: 919-831-3220
mstrong@flowserve.com

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Power Reactor Event Number: 54302
Facility: BROWNS FERRY
Region: 2     State: AL
Unit: [] [2] []
RX Type: [1] GE-4,[2] GE-4,[3] GE-4
NRC Notified By: MATTHEW SLOUKA
HQ OPS Officer: DONALD NORWOOD
Notification Date: 10/01/2019
Notification Time: 07:05 [ET]
Event Date: 10/01/2019
Event Time: 03:07 [CDT]
Last Update Date: 10/01/2019
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
Person (Organization):
BINOY DESAI (R2DO)
Unit SCRAM Code RX Crit Initial PWR Initial RX Mode Current PWR Current RX Mode
2 A/R Y 0 Startup 0 Hot Shutdown

Event Text

AUTOMATIC REACTOR SCRAM DURING REACTOR STARTUP

"On 10/1/2019 at 0307 CDT, Unit 2 was conducting a normal reactor startup and received a valid Reactor Protection System (RPS) scram. The reactor was critical in MODE 2 at the Point of Adding Heat. Operators began withdrawing Source Range Monitor (SRM) Instrumentation per procedure. When the operator depressed the SRM Drive Out pushbutton to withdraw the last two SRMs (C and D), an unexpected full Reactor Scram was received. Annunciator indication in the Main Control Room indicated a Neutron Monitoring Scram. The Intermediate Range Monitors (IRM) D, E, F, H and G all indicated Upscale High High. There were no Emergency Core Cooling System (ECCS) or Containment Isolation System actuations. All other systems functioned as designed.

"The cause of the Reactor Scram is still under investigation.

"This event requires a 4-hour report per 10 CFR 50.72(b)(2)(iv)(B), 'Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.'

"This event also requires an 8-hour report per 10 CFR 50.72(b)(3)(iv)(A), 'Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B), (1) Reactor protection system (RPS) including: reactor scram or reactor trip, except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.'

"The NRC Resident Inspector has been notified."


Page Last Reviewed/Updated Tuesday, October 01, 2019
Tuesday, October 01, 2019