05000446/LER-1917-001, Regarding Auxiliary Feedwater System Actuation During Unit 2 Turbine Trip
| ML17290A358 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 10/05/2017 |
| From: | Dreyfuss J Vistra Energy, Vistra Operations Company |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| CP-201700743, TXX-17079 LER 17-001-00 | |
| Download: ML17290A358 (5) | |
| 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) |
| 4461917001R00 - NRC Website | |
text
CP-201700743 TXX-17079 U.S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, DC 20555-0001 10/05/2017 V!S?~'\\
ENERGY
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SUBJECT:
COMANCHE PEAK NUCLEAR POWER PLANT DOCKET NO. 50-446 John R. Dreyfuss Plant Manager Luminant P.O. Box 1002 6322 North FM 56 Glen Rose, TX 76043 0 254.897.5200 m 802.380.0894 Ref 10 CFR 50.73 AUXILIARY FEEDWATER SYSTEM ACTUATION DURING UNIT 2 TURBINE TRIP LICENSEE EVENT REPORT 446/17-001-00
Dear Sir or Madam:
Pursuant to 10CFR50.73, Vistra Operations Company LLC (Vistra OpCo), hereby submits enclosed Licensee Event Report 446/17-001-00, "Auxiliary Feedwater System Actuation During Unit 2 Turbine Trip" for Comanche Peak Nuclear Power Plant (CPNPP) Unit 2.
This communication contains no new licensing basis commitments regarding CPNPP Units 1 and 2.
If you have any questions regarding this submittal, please contact Gary L. Merka at 254-897-6613.
Sincerely, 1601 BRYAN STREET DALLAS, TEXAS 75201 0 214-812-4600 VISTRAENERGY.COM
TXX-17079 Page 2 of 2 Enclosure c-Kriss Kennedy, Region IV Margaret W. O'Banion, NRR Resident Inspectors, Comanche Peak
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 03/31/2020 (04-2017) htti:i://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr1022/r3/)
the NRG may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 3. PAGE Comanche Peak Nuclear Power Plant 05000 446 1 OF 3
- 4. TITLE Auxiliary Feedwater System Actuation During Unit 2 Turbine Trip
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED I
SEQUENTIAL I REV FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR MONTH DAY YEAR NUMBER NO.
05000 FACILITY NAME DOCKET NUMBER 08 11 2017 2017 -
001 m
00 10 05 2017 05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: {Check all that apply)
D 20.2201(b)
D 20.2203(a)(3)(i)
D 50.73(a)(2)(ii)(A)
D 50.73(a)(2)(viii)(A) 1 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)
D 20.2203(a)(4)
D 50.73(a)(2)(iii)
D 50.73(a)(2)(ix)(A)
D 20.2203(a)(2)(i)
D 50.36(c)(1)(i)(A)
[{] 50.73(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)(4)
D 20.2203(a)(2)(iii)
D 50.36(c)(2)
D 50.73(a)(2)(v)(B)
D 13.11 (a)(5)
D 20.2203(a)(2)(iv)
D 5o.46(a)(3)(ii)
D 50.73(a)(2)(v)(C)
D 13.77(a)(1) 10 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)
D OTHER Specify in Abstract below or in SUMMARY OF THE EVENT, INCLUDING DATES AND APPROXIMATE TIMES On August 11, 2017, CPNPP Unit 2 was in MODE 1 operating at approximately 10% power while increasing power following a refueling outage. After syncing the Main Generator to the grid at 1120, Operators (utility, licensed) in the Unit 2 Control Room noted increasing water level in Steam Generator (SG) 2-02. The SG 2-02 flow control bypass valve [EllS:
(SJ)(FCV)] was demanded closed, but the valve would not close and remained in mid-position. Operators then attempted to close the valve via the hand switch on the Main Control Board, but SG 2-02 water level continued rising, and at 1124 Unit 2 received a P-14 signal resulting in an automatic Turbine trip, a trip of the 2B Main Feedwater Pump, a Feedwater Isolation signal, and an automatic Auxiliary Feedwater (AFW) pump start as designed. All systems responded normally during and following the Turbine trip.
E. THE METHOD OF DISCOVERY OF EACH COMPONENT OR SYSTEM FAILURE, OR PROCEDURAL PERSONNEL ERROR Operators (utility, licensed) in the Unit 2 Control Room noted increasing level in Steam Generator (SG) 2-02 followed by an automatic Turbine trip on a P-14 signal.
II. COMPONENT OR SYSTEM FAILURES A. CAUSE OF EACH COMPONENT OR SYSTEM FAILURE The cause of the SG 2-02 flow control bypass valve malfunction is still being determined and a supplemental report providing this information will be submitted by January 25, 2018.
B. FALURE MODE, MECHANISM, AND EFFECTS OF EACH FAILED COMPONENT The cause of the SG 2-02 flow control bypass valve malfunction is still being determined and a supplemental report providing this information will be submitted by January 25, 2018.
C. SYSTEMS OR SECONDARY FUNCTIONS THAT WERE AFFECTED BY FAILURE OF COMPONENTS WITH MULTIPLE FUNCTIONS Below 25 percent load, the SG 2-02 flow control bypass valves automatically maintain the steam generator water level by using control signals from the Steam Generator water levels. The valves are air operated and are designed to fail closed upon loss of air.
D. FAILED COMPONENT INFORMATION
The SG 2-02 flow control bypass valves are 4 inch, carbon steel, globe valves. The valves are a Model ED manufactured by Fisher Controls.
Ill. ANALYSIS OF THE EVENT A. SAFETY SYSTEM RESPONSES THAT OCCURRED The Turbine Trip and Auxiliary Feedwater Systems actuated as required.
B. DURATION OF SAFETY SYSTEM TRAIN INOPERABILITY Not applicable - No safety system train inoperability resulted from this event.
C. SAFETY CONSEQUENCES AND IMPLICATIONS OF THE EVENT This event is bounded by the CPNPP Final Safety Analysis Report (FSAR) accident analysis which assumes conservative initial conditions which bound the plant operating range and other assumptions which could reduce the capability of safety systems to mitigate the consequences of the transient. Feedwater system malfunctions that result in an increase in feedwater flow are analyzed in section 15.1.2 of the FSAR. The system is analyzed to demonstrate plant behavior in the event that an excessive feedwater addition occurs due to a control system malfunction or operator error. The FSAR analysis shows that the departure from nucleate boiling ratio encountered for an excessive feedwater addition at power is above the limit value and the feedwater malfunction event at no-load is bounded by the feedwater malfunction event at full power. The event of August 11, 2017, occurred at 10% reactor power, and all safety systems and components functioned as designed. Based on the above, it is concluded that the health and safety of the public were unaffected by this condition and this event has been evaluated to not meet the definition of a safety system functional failure per 10CFR50.73(a)(2)(v).
IV. CAUSE OF THE EVENT
The AFW actuation was caused by a P-14 signal that was received due to high level in SG 2-02 related to the mechanical malfunction of a Steam Generator 2-02 flow control bypass valve. The cause of the SG 2-02 flow control bypass valve malfunction is still being determined and a supplemental report providing this information will be submitted by January 25, 2018.
V. CORRECTIVE ACTIONS
The SG 2-02 feedwater flow control bypass valve was repaired. An extent of condition review was performed and verified that all of the other flow control bypass valves were closed. Additional corrective actions related to this event are still being determined and a supplemental report providing this information will be submitted by January 25, 2018.
VI. PREVIOUS SIMILAR EVENTS
A similar reportable event occurred at CPNPP on October 3, 2015 related to a feedwater flow control valve malfunction (Unit 2 LER 446/15-002). The cause of that event was a degraded positioner 0-ring, and the details/causes of the August 11, 2017 event are believed to be sufficiently different from the October 3, 2015 event such that the previous corrective actions could not have prevented this event. Page 3
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