05000382/LER-2019-005-01, Automatic Reactor Scram Due to Steam Generator 1 High Level Resulting from a Main Turbine Trip and Subsequent Reactor Power Cutback Due to Failed Diode Modules with the Main Exciter
| ML20099C680 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 04/08/2020 |
| From: | Wood P Entergy Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| W3F1-2020-0012 LER 2019-005-01 | |
| Download: ML20099C680 (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)(iii) 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) |
| 3822019005R01 - NRC Website | |
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Entergy Operations, Inc.
17265 River Road Killona, LA 70057-3093 Tel (504) 464-3786 Paul Wood Manager, Regulatory Assurance 10 CFR 50.73 W3F1-2020-0012 April 8, 2020 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 In accordance with 10 CFR 50.73, Entergy is hereby submitting supplemental LER 2019-005-01 for an event that occurred on May 16, 2019.
This letter contains no new regulatory commitments.
If you have any questions or require additional information, please contact the Regulatory Assurance Manager, Paul Wood, at (504) 464-3786.
Respectfully, Paul Wood PW/jb
Subject:
Licensee Event Report (LER) 2019-005-01 Automatic Reactor SCRAM due to Steam Generator #1 High Level Resulting from a Main Turbine Trip and Subsequent Reactor Power Cutback due to Failed Diode Modules within the Main Exciter Waterford Steam Electric Station, Unit 3 (Waterford 3)
NRC Docket No. 50-382 Renewed Facility Operating License No. NPF-38
Enclosure:
Waterford 3 Licensee Event Report 2019-005-01 cc:
NRC Region IV Regional Administrator NRC Senior Resident Inspector - Waterford Steam Electric Station, Unit 3 NRR Project Manager
ENCLOSURE W3F1-2020-0012 Entergy Operations, Inc.
Waterford 3 Licensee Event Report 2019-005-01 (4 pages)
NRC FORM 366 (04-2018)
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION (04-2020)
LICENSEE EVENT REPORT (LER)
(See Page 2 for required number of digits/characters for each block)
(See NUREG-1022, R.3 for instruction and guidance for completing this form http://www.nrc/gov/reading-rm/doc-collections/nuregs/staff/sr1022/r3/)
APPROVED BY OMB: NO. 3150-0104 EXPIRES: 04/30/2020
- 1. FACILITY NAME Waterford Steam Electric Station, Unit 3
- 2. DOCKET NUMBER 05000382
- 3. PAGE 1 OF 4
- 4. TITLE Automatic Reactor Scram due to Steam Generator #1 High Level Resulting from a Main Turbine Trip and Subsequent Reactor Power Cutback due to Failed Diode Modules within the Main Exciter
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED MONTH DAY YEAR YEAR SEQUENTIAL NUMBER REV NO.
MONTH DAY YEAR FACILITY NAME DOCKET NUMBER 05 16 2019 2019 -
005
- - 01 04 08 2020 FACILITY NAME DOCKET NUMBER
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) 1 20.2201(b) 20.2203(a)(3)(i) 50.73(a)(2)(ii)(A) 50.73(a)(2)(viii)(A) 20.2201(d) 20.2203(a)(3)(ii) 50.73(a)(2)(ii)(B) 50.73(a)(2)(viii)(B) 20.2203(a)(1) 20.2203(a)(4) 50.73(a)(2)(iii) 50.73(a)(2)(ix)(A) 20.2203(a)(2)(i) 50.36(c)(1)(i)(A) 50.73(a)(2)(iv)(A) 50.73(a)(2)(x)
- 10. POWER LEVEL 20.2203(a)(2)(ii) 50.36(c)(1)(ii)(A) 50.73(a)(2)(v)(A) 73.71(a)(4) 100 20.2203(a)(2)(iii) 50.36(c)(2) 50.73(a)(2)(v)(B) 73.71(a)(5) 20.2203(a)(2)(iv) 50.46(a)(3)(ii) 50.73(a)(2)(v)(C) 73.77(a)(1) 20.2203(a)(2)(v) 50.73(a)(2)(i)(A) 50.73(a)(2)(v)(D) 73.77(a)(2)(i) 20.2203(a)(2)(vi) 50.73(a)(2)(i)(B) 50.73(a)(2)(vii) 73.77(a)(2)(ii) 50.73(a)(2)(i)(C)
OTHER Specify in Abstract below or in Steam Generator Level Deviation:
The vendor review completed to determine the cause of the Steam Generator Level deviation did not determine the definitive cause for the deviation between the two level channels. There does not appear to be an obvious difference between the Original Steam Generators (OSG) and the Replacement Steam Generators (RSG) which would cause the two level channels measuring the level within the same SG to differ during a transient. The cause of the level channel deviations is most likely attributed to multiple factors which affect the sensed level via the pressure transmitter(s).
CORRECTIVE ACTIONS
A. Loss of Excitation:
Completed corrective actions include:
a.
The Main Exciter rectifier wheel was disassembled, cleaned, repaired including replacement of failed diode modules, and reassembled.
b.
Failed diodes were sent offsite for failure analysis.
c.
Corrected cooling water leak and door seal degradation inside Main Exciter housing.
d.
Revised Main Exciter Preventive Maintenance strategy to clearly outline the scope for Major and Minor PMs of the Main Exciter.
e.
Implemented actions targeting the understanding of the proper N/A of steps in accordance with fleet administrative procedure guidance including the justification expected/required per the procedure.
B. Steam Generator Level Deviation:
Completed corrective actions include:
a.
An Emergent Issue Team established during the forced outage confirmed that the Feedwater Control System performed as expected based on receiving a level deviation in both Steam Generator 1 and 2. The level deviation was not an expected plant response following a Reactor Power Cutback. Based on discussions with the replacement Steam Generator vendor, this may be an effect related to the Replacement Steam Generators b.
Completed a vendor study to determine the cause of the Steam Generator Level Deviations following the Reactor Power Cutback Planned corrective actions include:
a.
Increase SG NR level transmitter deviation check within the feedwater control system to provide sufficient operating margin to deviation setpoint b.
Calibrate the two level control transmitters such that each output is equal during steady state operation c.
Change the transmitter response time to provide additional smoothing of the level inputs to the feedwater control system and reduce the likelihood of the channel deviation check activation SAFETY EVALUATION The actual consequences as stated in the problem statement were low. There were no other actual consequences to safety of the general public, nuclear safety, industrial safety and radiological safety for this event. The potential consequence to safety of the general public, nuclear safety, industrial safety and radiological safety of this event, if the Automatic Reactor Trip on High Steam Generator Water Level was removed, is low. High Steam Generator Water Level in Steam Generator #2 would have also generated an automatic Reactor Trip. Failing that, the Operators would have initiated a manual Reactor Trip or taken manual control to restore Steam Generator Water Level.
Steam Generator overfill is modeled in the PRA only as it relates to success of the Emergency Feedwater A/B pump, not the Main Feedwater Pumps. The overfill modeled in these PRA sequences occurs post-trip and is based on Emergency Feedwater being established for plant conditions, therefore, this PRA assessment is qualitative in nature.
PREVIOUS OCCURRENCES
CR-WF3-2017-5842: (Reported under LER 2017-002-00 and Supplement 2017-002-01.)
A reactor power cutback occurred on July 17, 2017 following a main turbine trip initiated from Isophase Bus Duct failure.
The same level deviations occurred in Steam Generators 1 and 2. However, this event did not result in Feed Water Control shifting to manual due to the subsequent reactor trip that was experienced (failure of Fast Dead Bus Transfer).