05000446/LER-1917-003, Regarding Manual Reactor Trip Due to Trip of Both Main Feedwater Pumps

Regarding Manual Reactor Trip Due to Trip of Both Main Feedwater Pumps

ML18033A329
Person / Time
Site:	Comanche Peak
Issue date:	01/22/2018
From:	Dreyfuss J Vistra Energy
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
CP-201800009, TXX-18001 LER 17-003-00
Download: ML18033A329 (6)

LER-1917-003, Regarding Manual Reactor Trip Due to Trip of Both Main Feedwater Pumps

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)
4461917003R00 - NRC Website

text

CP-201800009 TXX-18001 U.S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, DC 20555-0001 V!S'!~'\\

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SUBJECT:

COMANCHE PEAK NUCLEAR POWER PLANT-UNIT 2 DOCKET NO. 50-446 REACTOR TRIP DUE TO TRIP OF BOTH MFPS

Dear Sir or Madam:

John R. Dreyfuss Plant Manager Luminant P.O. Box 1002 6322 North FM 56 Glen Rose, TX 76043 o 254.897.5200 Ref 10 CFR 50.73 Enclosed is Licensee Event Report (LER) 2-17-003-00, "Unit 2 Reactor Trip Due to Trip of Both MFPs" for Comanche Peak Nuclear Power Plant (CPNPP), Unit 2.

This letter contains no new regulatory commitments.

If you have any questions regarding this submittal, please contact Garry Struble at (254) 897-6628 or garry.struble@luminant.com.

Enclosure Sincerely, COMANCHE PEAK NUCLEAR POWER PLANT-UNIT 2, REACTOR TRIP DUE TO TRIP OF BOTH MFPS LICENSEE EVENT REPORT {LER) 2-17-003-00 1601 BRYAN STREET DALLAS, TEXAS 75201 o 214-812-4600 VISTRAENERGY.COM

TXX-18001 Page 2 of 2 c-Kriss Kennedy, Region IV Balwant Singal, NRR Resident Inspectors, Comanche Peak

NRC FORM366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB: NO. 3150-0104 EXPIRES: 03/31/2020 (04-2017) htt11://www.nrc.go)l/reading-rm/doc-collections/nuregs/staff/sr1022/r3l) 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, Unit 2 05000 446 1 OF 4

4. TITLE Manual Reactor Trip due to trip of both Main Feedwater Pumps

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 NUMBER NO.

MONTH DAY YEAR 05000 FACILITY NAME DOCKET NUMBER 11 25 2017 2017 -

003 -

00 01 22 2018 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 so.13(a)(2><iii)

D 50.73(a)(2)(ix)(A)

D 20.2203(a)(2>

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><n>

D 50.36(c)(1)(ii)(A)

D 50.73(a)(2)(v)(A)

D 13.71(a)(4>

D 20.2203(a)(2)(rn>

D so.3a(c)(2>

D 50.73(a)(2)(v)(B)

D 73.71(a)(s>

D 20.2203(a)(2)(iv)

D so.4a(a><3><n>

D 50.73(a)(2)(v)(C)

D 13.77(a><1>

100 D

20.2203(a)(2)(v)

D 50.73(a)(2)(i)(A)

D 50.73(a)(2)(v)(D)

D 13.77(a><2>

D 20.2203(a)(2)(vi)

D 50.73(a)(2)(i)(B)

D 50.73(a)(2)(vii)

D 13.77(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 At time 0042 on November 25, 2017 Comanche Peak, Unit 2 entered Technical Specification 3.3.2, Engineered Safety Feature Actuation System (ESFAS) Instrumentation to perform Train A surveillance testing. While performing the test a push-to-test switch failed. Analysis determined the alignment of Train A Solid State Protection System and the failure of the test switch did not cause or contribute to the trip of both main feedwater pumps. The failed switch was replaced on November 26, 2017 and the Solid State Protection System returned to its normal alignment.

At time 2023 on November 25, 2017 Comanche Peak, Unit 2 received alarms 2-ALB-78 FWPT A TRIP and 2-ALB-8A FWPT 8 TRIP indicating a trip of both main feedwater pumps [EIIS:(SJ)(P)]. After confirming a decreasing water level in all four steam generators, the control room initiated a manual reactor trip 14 seconds after the feed pumps tripped. All safety systems responded as designed.

The cause is not determined because there is no evidence that directly confirms why the main feedwater pumps tripped.

The initiating cause was not a permanent condition and the trip signal was gone once investigation into the cause of the trip began. Main feedwater pump, main feedwater pump control system (Mark V), and Solid State Protection System diagnosis and analysis were performed with no direct cause for trip of both main feedwater pumps identified. The reactor was started up on November 29, 2017.

E. THE METHOD OF DISCOVERY OF EACH COMPONENT OR SYSTEM FAILURE, OR PROCEDURAL PERSONNEL ERROR Initial indication of both main feedwater pumps tripping was provided to the Control room operator by annunciated alarms.

Operators (Utility, Licensed) confirmed main feedwater pump trips by observing decreasing level in all four steam generators. The reactor was manually tripped approximately 14 seconds after both main feedwater pumps tripped (times as indicated by the plant computer).

II. COMPONENT OR SYSTEM FAILURES A. CAUSE OF EACH COMPONENT OR SYSTEM FAILURE No definitive component or system failure has been identified. A likely cause of the trip of both main feedwater pumps is a possible spurious actuation of abandoned relays in the Solid State Protection System. A modification performed in 1992 removed the input signals for two water hammer interlocks (steam generator low pressure and low level). The modification did not remove the relays or remove power to relays that previously would trip both main feedwater pumps. The fuses providing power to the abandoned relays were removed on both units. Subsequent corrective actions will remove relays and contacts from the system.

8. FAILURE MODE, MECHANISM, AND EFFECTS OF EACH FAILED COMPONENT Although both main feedwater pumps tripped, it cannot be positively confirmed what device actuated to cause the trips.

Since both main feedwater pumps tripped at the same time, the circuits that could have caused the trip are limited to the Solid State Protection System auxiliary relays or some malfunction of the Mark V Digital feedwater control system that caused the processors to initiate a trip of both main feedwater pumps. The most likely (or least unlikely) is a spurious actuation of an abandoned relay because an actuation of either train relay can trip both main feedwater pumps without a trip of the Main Turbine and this is the only logical component that requires a single circuit failure to cause the eveht as it occurred on November 25, 2017.

C. SYSTEMS OR SECONDARY FUNCTIONS THAT WERE AFFECTED BY FAILURE OF COMPONENTS WITH MULTIPLE FUNCTIONS This event did not involve systems or secondary functions which were affected by the possible spurious relay actuation.

D. FAILED COMPONENT INFORMATION

No specific component was determined to be failed.

Ill. ANALYSIS OF THE EVENT A. SAFETY SYSTEM RESPONSES THAT OCCURRED The Reactor Protection System responded as designed to the manual trip input by the plant operators. All plant safety systems responded as designed. Automatic start of the AFW system was the expected response and the system responded as designed.

B. DURATION OF SAFETY SYSTEM TRAIN INOPERABILITY The event reported herein did not involve the inoperability of any safety system component or system.

Unit 2, Train A Solid State Protection System was in a Slave Relay Testing alignment for the surveillance test with the failed test switch described above. This alignment did not affect the trip of both main feedwater pump event or the possible spurious relay actuation.

Unit 2, Train A Solid State Protection System was inoperable for 27 hours3.125e-4 days <br />0.0075 hours <br />4.464286e-5 weeks <br />1.02735e-5 months <br /> and 3 minutes.

C. SAFETY CONSEQUENCES AND IMPLICATIONS OF THE EVENT A loss of normal feedwater is an ANS Condition II event (Faults of Moderate Frequency). When both main feedwater pumps tripped the reactor was manually tripped and the auxiliary feedwater system automatically started to provide feedwater to the steam generators. The reactor trip on low-low level in any steam generator provides the necessary protection against a loss of normal feedwater.

No automatic safety functions were exercised other than the expected automatic start of the Auxiliary Feedwater System and all plant safety systems responded as designed during the resultant transient. This event had no impact on nuclear safety, reactor safety, radiological safety, environmental safety or the safety of the public. This event has been evaluated to not meet the definition of a safety system functional failure per 10 CFR 50.73(a)(2)(v).

IV. CAUSE OF THE EVENT

A definitive cause for the trip of both main feedwater pumps was not identified. Any failure of the main feedwater pumps or the digital feedwater control system (Mark V) that could trip both main feedwater pumps were eliminated. Several potential main feedwater pump trip signals from the Solid State Protection System were eliminated. However, a possible cause is the spurious actuation of an abandoned relay from a previous plant modification. Inputs to the abandoned relays were removed but the relays were abandoned in place with power available to the relay circuit. The abandoned relays and associated contacts have not been cycled by operation or testing for approximately 25 years.

V. CORRECTIVE ACTIONS

The main feedwater pumps, the digital feedwater control system (Mark V), and the Solid State Protection System were diagnosed and analyzed to determine a cause. The main feedwafor pumps and the digital feedwater control system were tested and no failures were identified. The Solid State Protection System analysis identified a possible cause in a spurious actuation of abandoned relays. The power fuses for these relays were removed on both Unit 1 and Unit 2. All proposed activities initiated as a result of this event are being tracked and managed in the Comanche Peak Corrective Action Program.

VI. PREVIOUS SIMILAR EVENTS

There have been no similar reportable events at Comanche Peak in the past three years. Page 4

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