05000293/LER-2018-006, Automatic Reactor Scram Due to Feedwater Regulating Valve Malfunction
| ML18338A107 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 11/29/2018 |
| From: | Miner P Entergy Nuclear Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 2.18.'068 LER 2018-006-00 | |
| Download: ML18338A107 (6) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(iv)(B), System Actuation |
| 2932018006R00 - NRC Website | |
text
~Entergx Letter Number 2.18.'068 November 29, 2018 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Entergy Nuclear Operations, Inc.
Pilgrim Nuclear Power Station 600 Rocky Hill Road Plymouth, MA 02360 Peter J. Miner Manager, Regulatory Assurance SUBJECT: Licensee Event Report 2018-006-00, Automatic Reactor Scram Due to Feedwater Regulating Valve Malfunction Pilgrim Nuclear Power Station Docket No. 50-293 Renewed License No. DPR-35
Dear Sir or Madam:
The enclosed Licensee Event Report 2018-006-00, Automatic Reactor Scram Due to Feedwater Regulating Valve Malfunction, is submitted in accordance with Title 1 O Code of Federal Regulations 50. 73.
If you have any questions or require additional information, please contact me at (508) 830-7127.
There are no regulatory commitments contained in this letter.
Sincerely,
~;tf}y~
PJM/r; # : Licensee Event Report 2018-006-00, Automatic Reactor Scram Due to Feedwater Regulating Valve Malfunction
Entergy Nuclear Operations, Inc.
Pilgrim Nuclear Power Station cc:
Mr. David C. Lew Regional Administrator, Region I U. S. Nuclear Regulatory Commission 2100 Renaissance Boulevard, Suite 100 King of Prussia, PA 19406-2713 Mr. John Lamb, Senior Project Manager Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 9 012 Washington, DC 20555-0001 NRC Senior Resident Inspector Pilgrim Nuclear Power Station Letter No. 2.18.068 Page 2 of 2 Letter Number 2.18.068 Licensee Event Report 2018-006-00, Automatic Reactor Scram Due to Feedwater Regulating Valve Malfunction (3 Pages)
SNRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB: NO. 3150-0104 EXPIRES: 03/31/2020 (04-2017) htto://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr1022/r3D the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 1. FACILITY NAME
~- DOCKET NUMBER
.PAGE Pilgrim Nuclear Power Station 05000-293 1 OF3
'1. TITLE Automatic Reactor Scram Due to Feedwater Regulating Valve Malfunction
- 5. EVENT DATE
- 6. LEA NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED SEQUENTIAL FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR REV MONTH DAY YEAR N/A NUMBER NO.
N/A 10 05 2018 2018
- - 006
- - 00 11 29 2018 FACILITY NAME DOCKET NUMBER N/A N/A
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)
N D 20.2201 (b)
D 20.2203(a)(3)(i)
D 50.73(a)(2)(ii)(A)
D 50.73(a)(2)(viii)(A)
D 20.2201 (d)
D 20.2203(a)(3)(ii)
D 50.73(a)(2)(ii)(B)
D 50.73(a)(2)(viii)(B)
D 20.2203(a)(1 l D 20.2203(a)(4)
D so.73(a)(2)(iii)
D 50.73(a)(2)(ix)(A)
D 20.2203(a)(2)(i)
D 50.36(c)(1)(i)(A)
[8150.73(a)(2)(iv)(A)
D 50.73(a)(2)(x)
- 10. POWER LEVEL D 20.2203(a)(2)(ii)
D 50.36(c)(1 )(ii)(A)
D 50.73(a)(2)(v)(A)
D 73.71 (a)(4l D 20.2203(a)(2)(iii)
D so.36(c)(2)
D 50.73(a)(2)(v)(B)
D 73.71 (a)(s) 100 D 20.2203(a)(2)(iv)
D so.46(a)(3)(ii)
D 50.73(a)(2)(v)(C)
D 73.77(a)(1)
D 20.2203(a)(2)(v)
D 50.73(a)(2)(i)(A)
D 50.73(a)(2)(v)(D)
D 73.77(a)(2)(i)
D 20.2203(a)(2)(vi)
D 50.73(a)(2)(i)(B)
D 50.73(a)(2)(vii)
D 73.77(a)(2)(ii)
D 50.73(a)(2)(i)(C) 0 OTHER Specify in Abstract below or in NRG Form 366A
- 12. LICENSEE CONTACT FOR THIS LEA LICENSEE CONTACT rrLEPHONE NUMBER (Include Area Code)
!Mr. Peter J. Miner - Regulatory Assurance Manager
~08-830-7127 CAUSE SYSTEM COMPONENT MANU-REPORTABLE
CAUSE
SYSTEM COMPONENT MANU-REPORTABLE FACTURER TOEPIX FACTURER TO EPIX X
SJ FCV C635 y
- 14. SUPPLEMENTAL REPORT EXPECTED
- 15. EXPECTED MONTH DAY YEAR D YES (If yes, complete 15. EXPECTED SUBMISSION DATE)
~ NO SUBMISSION DATE
~BSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)
On October 5, 2018 at 1209 hours0.014 days <br />0.336 hours <br />0.002 weeks <br />4.600245e-4 months <br />, with the reactor operating at 100% power, an automatic reactor scram occurred due to a reactor water level perturbation and receipt of a low reactor water level Reactor Protection System signal caused by a malfunction of the 'A' Feedwater Regulating Valve (FRV).
The Failure Modes Analysis attributed the probable cause to be an intermittent signal in the stepper motor and encoder control loop, which is the system that manipulates the pneumatic spool to operate the FRV's Double-Acting Piston Actuator. The as-found condition of the stepper motor and encoder electrical connectors showed that several pin sockets were not properly seated within their connector housings. Proper seating of the FRV stepper motor and encoder electrical connections is critical in ensuring error free operation of the FRVs.
There were no radiological releases due to this event. All control rods inserted fully. All other plant systems responded as designed. This event had no impact on the health and/or safety of the public.
This report is submitted in accordance with 10 CFR 50.73(a)(2)(iv)(A).
BACKGROUND 05000- 293 YEAR 2018 SEQUENTIAL NUMBER
- - 006 REV NO.
- - 00 The purpose of the Feedwater Level Control system (FWLC) is to automatically control feedwater flow to the reactor vessel, maintaining reactor vessel water level within a specified range during all operating modes.
FWLC instrumentation measures reactor vessel water level, total feedwater flowrate and total steam flowrate.
During automatic operation, these three measurements are used to control feedwater flow.
The ability to maintain vessel level within a specified range during load changes is accomplished by the three-element control signal. The total steam flow signal and the total feedwater flow signal are fed into a proportional amplifier. The output from this amplifier is the mismatch between the input signals (steam flow-feedwater flow error signal). If steam flow is greater than feedwater flow, the amplifier output is increased from its normal value, causing the system to increase feed flow to balance with steam flow. This amplifier output is fed to a second proportional amplifier that also receives a reactor vessel water level signal. Adding the reactor vessel water level signal to the steam flow-feedwater flow error signal results in a three-element control signal, which is fed to the level controller.
EVENT DESCRIPTION
On October 5, 2018 at 1209 hours0.014 days <br />0.336 hours <br />0.002 weeks <br />4.600245e-4 months <br /> with the reactor operating at 100% power, Pilgrim Nuclear Power Station automatically scrammed due to a reactor water level perturbation and receipt of a low reactor water level Reactor Protection System (RPS) signal. Investigation revealed a sudden increase in loop 'B' feedwater flow with a subsequent rapid decrease in loop 'A' feedwater flow. The flow perturbation resulted from closure of the
'A' Feedwater Regulating Valve (FRV), FV-642A. The 'B' FRV, FV-642B, responded properly by initially reducing flow then increasing flow to compensate. Feedwater flow through loop 'B' was not able to maintain reactor water level. An automatic reactor scram occurred due to receipt of a low reactor water level RPS signal. All control rods inserted fully.
CAUSE OF THE EVENT
rrhe Failure Modes Analysis attributed the probable cause to be an intermittent signal in the stepper motor and encoder control loop, which is the system that manipulates the pneumatic spool to operate the FRV's Double-
~cting Piston Actuator. The as-found condition of the stepper motor and encoder electrical connectors showed
~hat several pin sockets were not properly seated within their connector housings. Proper seating of the FRV stepper motor and encoder electrical connections is critical in ensuring error free operation of the FRVs.
CORRECTIVE ACTIONS
The entire pneumatic module which includes the stepper motor, encoder, and air spool, the associated connectors, and the stepper motor driver board were replaced on FV-642A. Post Work Testing, calibration, valve stroke and configuration review were completed satisfactorly demonstrating proper operation of the replacement components. Since these failure modes could be applicable to the FRV in the other train, the station proactively replaced the same components on FV-642B.
Page 2 of 3 (04-2017)
U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB: NO. 3150-0104 EXPIRES: 3/31/2020 LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET (See NUREG-1022, R.3 for instruction and guidance for completing this fonn http://www.nrc.gov/reading-nn/doc-collections/nureqs/staff/sr1022/r3D
, the NRG may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 3. LEA NUMBER YEAR Pilgrim Nuclear Power Station 05000- 293 2018
SAFETY CONSEQUENCES
SEQUENTIAL NUMBER
- - 006 REV NO.
- - 00 The actual consequence was an automatic reactor scram due to low reactor water level, which resulted from loss of ability to control FRV, FV-642A. All control rods inserted fully. All other plant systems responded as designed.
!There were no other actual consequences to safety of the general public, nuclear safety, industrial safety, and radiological safety for this event.
REPORT ABILITY
!This report is submitted in accordance with 1 O CFR 50.73(a)(2)(iv)(A), Any event or condition that resulted in manual or automatic actuation of any of the systems listed in paragraph (a)(2)(iv)(B) of this section including 50.73(a)(2)(iv)(B)(1 ), Reactor Protection System.
PREVIOUS EVENTS LER 2016-007-00, Manual Reactor Trip Due To Feedwater Regulating Valve Malfunction REFERENCES CR-PNP-2018-07927 Page 3 of 3