Event Notification Report for May 07, 2018
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
05/06/2018 - 05/07/2018
EVENT NUMBERS
533925348953386533875338853389
!!!!! THIS EVENT HAS BEEN RETRACTED !!!!!
!!!!! THIS EVENT HAS BEEN RETRACTED !!!!!
Power Reactor
Event Number: 53392
Facility: FARLEY
Region: 2 State: AL
Unit: [1] [] []
RX Type: [1] W-3-LP,[2] W-3-LP
NRC Notified By: BLAKE MITCHELL
HQ OPS Officer: STEVEN VITTO
Region: 2 State: AL
Unit: [1] [] []
RX Type: [1] W-3-LP,[2] W-3-LP
NRC Notified By: BLAKE MITCHELL
HQ OPS Officer: STEVEN VITTO
Notification Date: 05/08/2018
Notification Time: 01:39 [ET]
Event Date: 05/07/2018
Event Time: 23:00 [CDT]
Last Update Date: 06/27/2018
Notification Time: 01:39 [ET]
Event Date: 05/07/2018
Event Time: 23:00 [CDT]
Last Update Date: 06/27/2018
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
Person (Organization):
BINOY DESAI (R2DO)
BINOY DESAI (R2DO)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | N | 0 | Hot Standby | 0 | Hot Standby |
UNANALYZED CONDITION
"On May 7, 2018 at 1041 CDT, Unit 1 performed an RCS (reactor coolant system) leakrate procedure that calculated an unidentified RCS leakrate of 0.202 gpm. The leak source investigation concluded at 2150 that the packing for the charging flow control valve (FCV) was the source of the RCS leakage when it was bypassed, which isolated the leakage. A second RCS leakrate calculation was performed after the charging flow control valve was isolated which calculated an acceptable leakrate of 0.00 gpm.
"The packing leakage from the charging flow control valve represented leakage external to containment which would result in a greater that 5 Rem dose projection to control room personnel during accident conditions which does not satisfy the GDC19 criteria described in Technical Specification Bases 3.7.10. Therefore the control room emergency filtration system would not be able to fulfill its design function resulting in an unanalyzed condition.
"This condition is being reported pursuant to 10CFR50.72(b)(3)(ii) for a 'condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety'.
"The packing leak from the charging flow control valve will remain isolated until repaired under work order SNC944374."
The NRC Resident Inspector has been notified.
* * * RETRACTION FROM BLAKE MITCHELL TO ANDREW WAUGH AT 2151 EDT ON 6/27/2018 * * *
"Retraction - There was sufficient margin in the analysis of record to account for the increased leakage in the charging flow control valve that was not known at the time of reporting. Doses at the site boundary and in the Main Control Room would have remained less than the legal limits had a Loss of Coolant Accident occurred based on plant conditions at the time."
The licensee notified the NRC Resident Inspector.
Notified the R2DO (Michel).
"On May 7, 2018 at 1041 CDT, Unit 1 performed an RCS (reactor coolant system) leakrate procedure that calculated an unidentified RCS leakrate of 0.202 gpm. The leak source investigation concluded at 2150 that the packing for the charging flow control valve (FCV) was the source of the RCS leakage when it was bypassed, which isolated the leakage. A second RCS leakrate calculation was performed after the charging flow control valve was isolated which calculated an acceptable leakrate of 0.00 gpm.
"The packing leakage from the charging flow control valve represented leakage external to containment which would result in a greater that 5 Rem dose projection to control room personnel during accident conditions which does not satisfy the GDC19 criteria described in Technical Specification Bases 3.7.10. Therefore the control room emergency filtration system would not be able to fulfill its design function resulting in an unanalyzed condition.
"This condition is being reported pursuant to 10CFR50.72(b)(3)(ii) for a 'condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety'.
"The packing leak from the charging flow control valve will remain isolated until repaired under work order SNC944374."
The NRC Resident Inspector has been notified.
* * * RETRACTION FROM BLAKE MITCHELL TO ANDREW WAUGH AT 2151 EDT ON 6/27/2018 * * *
"Retraction - There was sufficient margin in the analysis of record to account for the increased leakage in the charging flow control valve that was not known at the time of reporting. Doses at the site boundary and in the Main Control Room would have remained less than the legal limits had a Loss of Coolant Accident occurred based on plant conditions at the time."
The licensee notified the NRC Resident Inspector.
Notified the R2DO (Michel).
Part 21
Event Number: 53489
Rep Org: FISHER CONTROLS INTERNATIONAL
Licensee: FISHER CONTROLS INTERNATIONAL
Region: 3
City: MARSHALLTOWN State: IA
County:
License #:
Agreement: Y
Docket:
NRC Notified By: KIM SAGAR
HQ OPS Officer: HOWIE CROUCH
Licensee: FISHER CONTROLS INTERNATIONAL
Region: 3
City: MARSHALLTOWN State: IA
County:
License #:
Agreement: Y
Docket:
NRC Notified By: KIM SAGAR
HQ OPS Officer: HOWIE CROUCH
Notification Date: 07/05/2018
Notification Time: 20:41 [ET]
Event Date: 05/07/2018
Event Time: 00:00 [CDT]
Last Update Date: 07/05/2018
Notification Time: 20:41 [ET]
Event Date: 05/07/2018
Event Time: 00:00 [CDT]
Last Update Date: 07/05/2018
Emergency Class: NON EMERGENCY
10 CFR Section:
21.21(d)(3)(i) - DEFECTS AND NONCOMPLIANCE
10 CFR Section:
21.21(d)(3)(i) - DEFECTS AND NONCOMPLIANCE
Person (Organization):
LAURA KOZAK (R3DO)
- PART 21/50.55 REACTORS (EMAIL)
LAURA KOZAK (R3DO)
- PART 21/50.55 REACTORS (EMAIL)
PART 21 NOTIFICATION - HEX NUTS AND CAP SCREWS ON CERTAIN VEE-BALL VALVES NOT PROPERLY DEDICATED
Fisher Controls issued FIN 2018-02 for hex nuts and caps screws for securing the valve body-to-actuator on eight valves (six purchased by D.C. Cook Nuclear and two purchased by Korea Hydro and Nuclear Power) that were not commercial grade dedicated. Fisher became aware of this issue on May 7, 2018.
For technical questions, please contact:
Jacob Clos
Quality Manager
Emerson Automation Solutions
Fisher Controls International LLC
301 South First Avenue
Marshalltown, IA 50158
Phone: (641) 754-2108
Jacob.Clos@Emerson.com
Fisher Controls issued FIN 2018-02 for hex nuts and caps screws for securing the valve body-to-actuator on eight valves (six purchased by D.C. Cook Nuclear and two purchased by Korea Hydro and Nuclear Power) that were not commercial grade dedicated. Fisher became aware of this issue on May 7, 2018.
For technical questions, please contact:
Jacob Clos
Quality Manager
Emerson Automation Solutions
Fisher Controls International LLC
301 South First Avenue
Marshalltown, IA 50158
Phone: (641) 754-2108
Jacob.Clos@Emerson.com
Power Reactor
Event Number: 53386
Facility: SALEM
Region: 1 State: NJ
Unit: [] [2] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: MATT MOG
HQ OPS Officer: STEVEN VITTO
Region: 1 State: NJ
Unit: [] [2] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: MATT MOG
HQ OPS Officer: STEVEN VITTO
Notification Date: 05/07/2018
Notification Time: 05:23 [ET]
Event Date: 05/07/2018
Event Time: 03:25 [EDT]
Last Update Date: 05/07/2018
Notification Time: 05:23 [ET]
Event Date: 05/07/2018
Event Time: 03:25 [EDT]
Last Update Date: 05/07/2018
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
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL 50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
Person (Organization):
MATT YOUNG (R1DO)
MATT YOUNG (R1DO)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 2 | M/R | Y | 100 | Power Operation | 0 | Hot Standby |
MANUAL REACTOR TRIP
"This 4 and 8 hour notification is being made to report that Salem Unit 2 initiated a manual reactor trip and subsequent automatic Auxiliary Feedwater system actuation. The trip was initiated due to a 21 Reactor Coolant Pump reaching its procedural limit for motor winding temperature of 302F.
"Salem Unit 2 is currently stable in Mode 3. Reactor Coolant system pressure is 2235 PSIG and Reactor Coolant System temperature is 547 F with decay heat removal via the Main Steam Dump and Auxiliary Feedwater Systems. Unit 2 has no active shutdown technical specification action statements in effect. All control rods inserted on the reactor trip. All ECCS [emergency core cooling systems] and ESF [emergency safety function] systems functioned as expected.
"No safety related equipment or major secondary equipment was tagged for maintenance prior to this event. No personnel were injured during this event."
The NRC Resident Inspector was notified. The Lower Alloways Creek Township will be notified.
"This 4 and 8 hour notification is being made to report that Salem Unit 2 initiated a manual reactor trip and subsequent automatic Auxiliary Feedwater system actuation. The trip was initiated due to a 21 Reactor Coolant Pump reaching its procedural limit for motor winding temperature of 302F.
"Salem Unit 2 is currently stable in Mode 3. Reactor Coolant system pressure is 2235 PSIG and Reactor Coolant System temperature is 547 F with decay heat removal via the Main Steam Dump and Auxiliary Feedwater Systems. Unit 2 has no active shutdown technical specification action statements in effect. All control rods inserted on the reactor trip. All ECCS [emergency core cooling systems] and ESF [emergency safety function] systems functioned as expected.
"No safety related equipment or major secondary equipment was tagged for maintenance prior to this event. No personnel were injured during this event."
The NRC Resident Inspector was notified. The Lower Alloways Creek Township will be notified.
Power Reactor
Event Number: 53387
Facility: COOK
Region: 3 State: MI
Unit: [] [2] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: JIM HABLIK
HQ OPS Officer: JEFF HERRERA
Region: 3 State: MI
Unit: [] [2] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: JIM HABLIK
HQ OPS Officer: JEFF HERRERA
Notification Date: 05/07/2018
Notification Time: 06:42 [ET]
Event Date: 05/07/2018
Event Time: 03:36 [EDT]
Last Update Date: 05/07/2018
Notification Time: 06:42 [ET]
Event Date: 05/07/2018
Event Time: 03:36 [EDT]
Last Update Date: 05/07/2018
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
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL 50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
Person (Organization):
BILLY DICKSON (R3DO)
BILLY DICKSON (R3DO)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 2 | M/R | Y | 30 | Power Operation | 0 | Hot Standby |
MANUAL RX TRIP DUE TO HIGH-HIGH LEVEL IN MOISTURE SEPARATOR DRAIN TANK
"On May 7, 2018 at 0336 [EDT], DC Cook Unit 2 Reactor was manually tripped due to a high-high level experienced in the East Moisture Separator Drain Tank (MSDT) of the Moisture Separator Reheater (MSR).
"This notification is being made in accordance with 10 CFR 50.72(b)(2)(iv)(B), Reactor Protection System (RPS) actuation as a four (4) hour report, and under 10 CFR 50.72(b)(3)(iv)(A), specified system actuation of the Auxiliary Feedwater System, as an eight (8) hour report. The NRC Resident Inspector has been notified.
"Unit 2 is being supplied by offsite power. All control rods fully inserted. All Aux Feedwater Pumps started properly. Decay heat is being removed via the Steam Generator Power Operated Relief Valves following Main Steam Stop Valve closure at 0431 due to a slow RCS [Reactor Coolant System] cooldown. Preliminary evaluation indicates all plant systems functioned normally following the Reactor Trip. DC Cook Unit 2 remains stable in Mode 3 while conducting the Post Trip Review. No radioactive release is in progress as a result of this event."
"On May 7, 2018 at 0336 [EDT], DC Cook Unit 2 Reactor was manually tripped due to a high-high level experienced in the East Moisture Separator Drain Tank (MSDT) of the Moisture Separator Reheater (MSR).
"This notification is being made in accordance with 10 CFR 50.72(b)(2)(iv)(B), Reactor Protection System (RPS) actuation as a four (4) hour report, and under 10 CFR 50.72(b)(3)(iv)(A), specified system actuation of the Auxiliary Feedwater System, as an eight (8) hour report. The NRC Resident Inspector has been notified.
"Unit 2 is being supplied by offsite power. All control rods fully inserted. All Aux Feedwater Pumps started properly. Decay heat is being removed via the Steam Generator Power Operated Relief Valves following Main Steam Stop Valve closure at 0431 due to a slow RCS [Reactor Coolant System] cooldown. Preliminary evaluation indicates all plant systems functioned normally following the Reactor Trip. DC Cook Unit 2 remains stable in Mode 3 while conducting the Post Trip Review. No radioactive release is in progress as a result of this event."
Power Reactor
Event Number: 53388
Facility: CALLAWAY
Region: 4 State: MO
Unit: [1] [] []
RX Type: [1] W-4-LP
NRC Notified By: JEREMY MORTON
HQ OPS Officer: DAN LIVERMORE
Region: 4 State: MO
Unit: [1] [] []
RX Type: [1] W-4-LP
NRC Notified By: JEREMY MORTON
HQ OPS Officer: DAN LIVERMORE
Notification Date: 05/07/2018
Notification Time: 16:31 [ET]
Event Date: 05/07/2018
Event Time: 13:35 [CDT]
Last Update Date: 05/07/2018
Notification Time: 16:31 [ET]
Event Date: 05/07/2018
Event Time: 13:35 [CDT]
Last Update Date: 05/07/2018
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(A) - POT UNABLE TO SAFE SD 50.72(b)(3)(v)(B) - POT RHR INOP 50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
10 CFR Section:
50.72(b)(3)(v)(A) - POT UNABLE TO SAFE SD 50.72(b)(3)(v)(B) - POT RHR INOP 50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
Person (Organization):
MARK HAIRE (R4DO)
MARK HAIRE (R4DO)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | Y | 100 | Power Operation | 100 | Power Operation |
DISCOVERY OF A CONDITION THAT COULD HAVE PREVENTED FULFILLMENT OF A SAFETY FUNCTION
"On May 7, 2018, during an engineering review of mission time requirements for Technical Specification related equipment, a deficiency was discovered regarding the Emergency Operating Procedure (EOP) guidance for natural circulation cooldown with a stagnant loop. This condition could be the result of a postulated Main Steam Line Break with a loss of offsite power.
"During a natural circulation cooldown with a faulted steam generator, flow in the stagnant reactor coolant system (RCS) loop associated with the isolated faulted steam generator (SG) could stagnate and result in elevated temperatures in that loop. This becomes an issue when RCS depressurization to residual heat removal system (RHR) entry conditions is attempted. The liquid in the stagnant loop will flash to steam and prevent RCS depressurization. In this condition, the time required to complete the cooldown would be sufficiently long that the nitrogen accumulators associated with Callaway's atmospheric steam dumps and turbine driven auxiliary feedwater pump flow control valves would be exhausted. The atmospheric steam dumps and turbine driven auxiliary feedwater pump would not be capable of performing their specified safety functions of cooling the plant to entry conditions for RHR operation. This issue has been analyzed by Westinghouse in WCAP-16632-P. This WCAP determined that to prevent loop stagnation, the RCS cooldown rate in these conditions should be limited to a rate dependent on the temperature differential present in the active loops.
"The WCAP analysis was used to support a revision to the generic Emergency Response Guideline (ERG) for ES-0.2 "Natural Circulation Cooldown." Figure 1 in ES-0.2 provides a curve of the maximum allowable cooldown rate as a function of active loop temperature differential which is directly proportional to the level of core decay heat. At the time of discovery of this condition, Callaway's EOP structure did not ensure that the ES-0.2 guidance would be implemented for a natural circulation cooldown with a stagnant loop.
"Callaway has issued interim guidance to the on-shift personnel regarding this concern and is in the process of revising the applicable EOPs.
"This condition is reportable per 10 CFR 50.72(b)(3)(v) for any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to (A) Shutdown the reactor and maintain it in a safe shutdown condition, (B) Remove residual heat, or (D) mitigate the consequences of an accident."
The licensee notified the NRC Resident Inspector of this condition.
"On May 7, 2018, during an engineering review of mission time requirements for Technical Specification related equipment, a deficiency was discovered regarding the Emergency Operating Procedure (EOP) guidance for natural circulation cooldown with a stagnant loop. This condition could be the result of a postulated Main Steam Line Break with a loss of offsite power.
"During a natural circulation cooldown with a faulted steam generator, flow in the stagnant reactor coolant system (RCS) loop associated with the isolated faulted steam generator (SG) could stagnate and result in elevated temperatures in that loop. This becomes an issue when RCS depressurization to residual heat removal system (RHR) entry conditions is attempted. The liquid in the stagnant loop will flash to steam and prevent RCS depressurization. In this condition, the time required to complete the cooldown would be sufficiently long that the nitrogen accumulators associated with Callaway's atmospheric steam dumps and turbine driven auxiliary feedwater pump flow control valves would be exhausted. The atmospheric steam dumps and turbine driven auxiliary feedwater pump would not be capable of performing their specified safety functions of cooling the plant to entry conditions for RHR operation. This issue has been analyzed by Westinghouse in WCAP-16632-P. This WCAP determined that to prevent loop stagnation, the RCS cooldown rate in these conditions should be limited to a rate dependent on the temperature differential present in the active loops.
"The WCAP analysis was used to support a revision to the generic Emergency Response Guideline (ERG) for ES-0.2 "Natural Circulation Cooldown." Figure 1 in ES-0.2 provides a curve of the maximum allowable cooldown rate as a function of active loop temperature differential which is directly proportional to the level of core decay heat. At the time of discovery of this condition, Callaway's EOP structure did not ensure that the ES-0.2 guidance would be implemented for a natural circulation cooldown with a stagnant loop.
"Callaway has issued interim guidance to the on-shift personnel regarding this concern and is in the process of revising the applicable EOPs.
"This condition is reportable per 10 CFR 50.72(b)(3)(v) for any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to (A) Shutdown the reactor and maintain it in a safe shutdown condition, (B) Remove residual heat, or (D) mitigate the consequences of an accident."
The licensee notified the NRC Resident Inspector of this condition.
Power Reactor
Event Number: 53389
Facility: WATERFORD
Region: 4 State: LA
Unit: [3] [] []
RX Type: [3] CE
NRC Notified By: ARVEL HALL
HQ OPS Officer: RICHARD SMITH
Region: 4 State: LA
Unit: [3] [] []
RX Type: [3] CE
NRC Notified By: ARVEL HALL
HQ OPS Officer: RICHARD SMITH
Notification Date: 05/07/2018
Notification Time: 17:40 [ET]
Event Date: 05/07/2018
Event Time: 10:22 [CDT]
Last Update Date: 05/07/2018
Notification Time: 17:40 [ET]
Event Date: 05/07/2018
Event Time: 10:22 [CDT]
Last Update Date: 05/07/2018
Emergency Class: NON EMERGENCY
10 CFR Section:
26.719 - FITNESS FOR DUTY
10 CFR Section:
26.719 - FITNESS FOR DUTY
Person (Organization):
MARK HAIRE (R4DO)
FFD GROUP (EMAIL)
MARK HAIRE (R4DO)
FFD GROUP (EMAIL)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 3 | N | Y | 100 | Power Operation | 100 | Power Operation |
FITNESS FOR DUTY - NON-LICENSED SUPERVISOR CONFIRMED POSITIVE FOR ALCOHOL
"A non-licensed supervisor had a confirmed positive result for alcohol during a random fitness for duty test. The employee's access to the plant has been terminated.
"The NRC Resident Inspector has been notified."
"A non-licensed supervisor had a confirmed positive result for alcohol during a random fitness for duty test. The employee's access to the plant has been terminated.
"The NRC Resident Inspector has been notified."