05000457/LER-2009-002, Regarding Loss of Offsite Power Coincident with a Reactor Trip Due to Loss of 2C Reactor Coolant Pump

Regarding Loss of Offsite Power Coincident with a Reactor Trip Due to Loss of 2C Reactor Coolant Pump

ML092710198
Person / Time
Site:	Braidwood
Issue date:	09/28/2009
From:	Shahkarami A Exelon Nuclear
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
BW090091 LER 09-002-00
Download: ML092710198 (5)

LER-2009-002, Regarding Loss of Offsite Power Coincident with a Reactor Trip Due to Loss of 2C Reactor Coolant Pump

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(i) 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)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(iv)(B), System Actuation
4572009002R00 - NRC Website

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10 CFR 50.73 September 28, 2009 BW090091 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Braidwood Station, Unit 2 Facility Operating License No. NPF-77 NRC Docket No. STN 50-457

Subject:

Licensee Event Report 2009-002-00 -- Unit 2 Loss of Offsite Power Coincident with a Reactor Trip Due to Loss of 2C Reactor Coolant Pump The enclosed Licensee Event Report (LER) is being submitted in accordance with 10 CFR 50.73, "Licensee event report system," paragraph (a)(2)(iv)(A), as an event that resulted in a valid actuation of the reactor protection system, emergency AC electrical power system and auxiliary feedwater system. On July 30,2009, Braidwood Station Unit 2 experienced a loss of offsite power due to a system auxiliary transformer tripping on a sudden pressure relay actuation. This was followed by an automatic reactor trip from full power due to a reactor coolant pump tripping on overcurrent.

10 CFR 50.73(a) requires an LER to be submitted within 60 days following discovery of the event.

Therefore, this report is being submitted by September 28,2009.

There are no regulatory commitments contained in this letter. Should you have any questions concerning this submittal, please contact Mr. David Gullott, Regulatory Assurance Manager, at (815) 417-2800.

Amir Shahkarami Site Vice Presiden Braidwood Station

Enclosure:

LER 2009-002-00

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 0813112010 (9-2007)

, the NRC may digits/characters for each block) not conduct or sponsor, and a person is not required to respond to, the informAtion I'nllAr.linn

13. PAGE Braidwood Station, Unit 2 05000457 1

of 4

4. TITLE Unit 2 Loss of Offsite Power Coincident with a Reactor Trip Due to Loss of 2C Reactor Coolant Pump 5, EVENT DATE

6. LER NUMBER

7. REPORT DATE

8. OTHER FACILITIES INVOLVED YEAR ISEQUENTIALIREV FACILITY NAME DOCKET NUMBER MONTH DAY YEAR NUMBER NO.

MONTH DAY YEAR N/A N/A FACILITY NAME DOCKET NUMBER 07 30 2009 2009 - 002 -

00 09 28 2009 N/A N/A

9. OPERATING MODE

11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) 1 o 20.2201(b) o 20.2203(a)(3)(i) o 50.73(a)(2)(i)(C) 50,73(a)(2)(vii) o 20.2201(d) 20.2203(a)(3)(ii) 50.73(a)(2)(ii)(A) o 50.73(a)(2)(viii)(A) o 20.2203(a)(1) 20.2203(a)(4) o 50.73(a)(2)(ii)(B) 50.73(a)(2)(viii)(B) n 20.2203(a)(2)(i) 50.36(c)(1 )(i)(A) 50.73(a)(2)(iii) 50.73(a)(2)(ix)(A)

10. POWER LEVEL o 20.2203(a)(2)(ii) o 50.36(c)(1 )(ii)(A) 50.73(a)(2)(iv)(A) 50.73(a)(2)(x) o 20.2203(a)(2)(iii) o 50.36(c)(2) o 50.73(a)(2)(v)(A) 73.71(a)(4) 100 o 20.2203(a)(2)(iv) o 50.46(a)(3)(ii) 50.73(a)(2)(v)(B) o 73.71(a)(5) o 20.2203(a)(2)(v) o 50.73(a)(2)(i)(A) 50.73(a)(2)(v)(C) o OTHER o 20.2203(a)(2)(vi) 50.73(a)(2)(i)(B) o 50.73(a)(2)(v)(D)

Specify in Abstract below or in NRC Form 36M

12. LICENSEE CONTACT FOR THIS LER FACILITY NAME ITELEPHONE NUMBER Code)

David Gullott, Regulatory Assurance Manager (815) 417-2800CAUSE SYSTEM COMPONENT MANU-REPORTABLE

CAUSE

SYSTEM COMPONENT MANU-REPORTABLE FACTURER TO EPIX FACTURER TO EPIX D

EA RLY W120 Y

N/A N/A N/A N/A N/A

14. SUPPLEMENTAL REPORT EXPECTED

15. EXPECTED MONTH DAY YEAR SUBMISSION YES (Ifyes, complete 15. EXPECTED SUBMISSION DATE)

NO DATE N/A N/A N/A ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)

On July 30,2009 at 20:59 hours, Unit 2 experienced an automatic reactor trip from full power. The 2C reactor coolant pump (RCP) tripped on overcurrent follOWing the initial loss of system auxiliary transformer (SAT) 242-1 on a sudden pressure relay actuation and subsequent bus 258 transfer. This resulted in an automatic reactor trip from full power due to less than four RCPs in operation above 30% power.

The root causes of the Unit 2 reactor trip were determined to be: 1) the 2C RCP breaker tripped inappropriately due to a combination of an overcurrent trip setpoint time delay shorter than minimum acceptable, and an overcurrent drop out reset value too low; and 2) the procedure for the calibration of overcurrent protective relays does not contain guidance to take an as-found reading / calibration and functional check of the overcurrent drop out reset value.

The corrective actions include: 1) perform a functional check, and calibration of the overcurrent protective relays for the 2C RCP, and on pumps with the same overcurrent protective relay; and 2) revise the procedure for the calibration of overcurrent protective relays to include procedure steps for functional check and calibration, as required, for the high drop out instantaneous values associated with overcurrent time delay relays.

Irhere were no actual safety consequences impacting plant or public safety as a result of this event.

This event is being reported pursuant to 10 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 10 CFR 50.73(a)(2)(iv)(B) including any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical, actuation of the emergency AC electrical power system, and actuation of the PWR auxiliary feedwater system.

NRC FORM 366 (9-2007)

PRINTED ON RECYCLED PAPER

LICENSEE EVENT REPORT (LER) u.s. NUCLEAR REGULATORY COMMISSION CONTINUATION SHEET

2. DOCKET

6. LER NUMBER

3. PAGE I

SEQUENTIAL I REV 05000457 YEAR NUMBER NO, 2

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2009 002 00

/\\.

Plant Operating Conditions Before The Event

Event Date:

Unit: 2 July 30, 2009 MODE: 1 Event Time: 20:59 Reactor Power: 100 percent Unit 2 Reactor Coolant System (RC) [AB]:

Normal operating temperature and pressure B.

Description of Event

There were no structures, systems or components inoperable at the beginning of the event that contributed to the severity of the event.

NOTE: The plant is designed so that during normal operations, two of the four reactor coolant pumps (RCPs) are normally powered from the onsite power via unit auxiliary transformers (UATs), and two RCPs powered from offsite power via system auxiliary transformers (SATs) [FK]. The plant is designed so that a fault on either of the power sources will result in an automatic bus transfer (ABT) to the unaffected power source.

On July 30,2009 at 20:59 hours, Unit 2 received a sudden pressure relay (SPR) actuation on SAT 242-1. In response, the breakers for SATs 242-1 and 242-2 automatically opened (deenergizing the SATs), which deenergized busses 258 and 259 [EA]. Per design, both busses automatically transferred to UATs 241-1 and 241-2.

The 2C RCP, which is powered from bus 258, inappropriately tripped on overcurrent following the bus 258 transfer to UAT 241-2. This resulted in an automatic reactor trip from full power due to less than four RCPs in operation above 30% power. FolloWing the trip of the main generator output breakers, the UATs subsequently deenergized, and the remaining RCPs tripped. FolloWing the Unit 2 reactor trip, the 2A and 2B auxiliary feedwater (AF) [BA] pumps and the 2A and 2B emergency diesel generators [EK] auto started, per design.

Operator response to the trip was proper and all safety related systems, structures, and components operated normally during this event.

This event resulted in an Unusual Event emergency classification due to loss of offsite power greater than 15 minutes. The declaration of an Emergency Classification is reportable per 10CFR50.72(a)(1 )(i) and 10CFR50.72(a)(3) within one hour of the declaration. At 21 :12 hours on July 30,2009, state and local government agencies were notified via the Nuclear Accident Reporting System (NARS). At 21 :55 hours on July 30, 2009, the NRC Operations Center was notified via the Emergency Notification System (ENS).

Since SAT 242-1 had an SPR actuation and a possible internal fault, the plant recovery was accomplished by removing the SAT 4 kV [EB] and 6.9 kV moveable disconnect links on the transformer low side of SAT 242-1 and opening the 345 kV high side motor operated disconnect to SAT 242-1. The offsite power was restored to Unit 2 by reenergizing SAT 242-2 that was disconnected from the potentially faulted SAT 242-1. At 00:36 hours on August 2,2009, folloWing placing of the safety 4kV busses on SAT 242-2, the Unusual Event was terminated. At 15:18 hours on August 2,2009, the NRC Operations Center was notified via ENS.

This event is reportable under 10 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 10 CFR 50.73(a)(2)(iv)(B) including any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical, actuation of the emergency AC electrical power system, and actuation of the PWR auxiliary feedwater system.

NRC FORM 36M (9"2007)

PRINTED ON RECYCLED PAPER

~.

Cause of Event

LICENSEE EVENT REPORT (LER) u.s. NUCLEAR REGULATORY COMMISSION CONTINUATION SHEET

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6. LER NUMBER

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2009 002 00 It was determined that the 2C RCP trip was not due to the SPR actuation on SAT 242-1, but occurred following the ABT of the 6.9 kV feed to bus 258 from SAT 242-2 to UAT 241-2, since both 2B and 2D condensate I condensate booster pumps (CD/CB) [SD] continued to run for approximately 40 seconds after the 2C RCP trip and subsequent reactor trip. The 2B and 2D CD/CB pumps continued to run because the UATs were still energized. Following the reactor trip, a turbine trip occurs. The main generator does not trip for approximately 60 seconds after the turbine trip. The main generator is motorized from the switchyard and continues to feed power to the UATs. When the main generator output breakers in the switchyard trip, the UATs are subsequently deenergized.

Testing of the 2C RCP overcurrent protective relays found that the overcurrent dropout and the time delay associated with this relay were out of tolerance. Specifically, the C phase relay was found to trip earlier than allowed by the relay setting order (83 milliseconds versus an expected 90 to 110 milliseconds) and the current drop out settings for these relays were lower than required (83% to 75% versus the expected 90% of the trip setpoint). The combination of a shorter than intended time delay and an extended period that the overcurrent element was picked up resulted in a condition in which the breaker would trip when a trip is not warranted. The net result of these two out-of-tolerance conditions caused the 2C RCP to inappropriately trip on overcurrent.

Additionally, a review of the procedure for calibration and functional checks of the RCP overcurrent relays determined that no guidance is provided in the procedure to ensure that the overcurrent protective relays are periodically checked or the settings verified.

Therefore, the root causes of the Unit 2 reactor trip were determined to be:

1) 2C RCP breaker tripped inappropriately due to a combination of an overcurrent trip setpoint time delay shorter than minimum acceptable per relay setting order, and an overcurrent drop out reset value too low to allow the relay to reset within the time delay window of the relay; and 2)

The procedure for the calibration of overcurrent protective relays does not contain guidance to take an as-found readinglcalibration and functional check of the overcurrent drop out reset value.

D.

Safety Consequences

There were no safety consequences impacting plant or public safety as a result of this event.

The reactor trip is designed to prevent Departure from Nucleate Boiling caused by the reduction in total core flow.

This event is explicitly described under Updated Final Safety Analysis Report, section 15.3.2, "Complete Loss of Reactor Coolant Flow".

This event resulted in a loss of all offsite power to Unit 2, and a loss of forced cooling to the Unit 2 reactor and use of the Unit 2 Emergency Diesel Generators to power safety related loads. The reactor trip system responded automatically due to the loss of the 2C RCP. During the reactor shutdown, all required safety systems responded appropriately. There was no loss of any function that would have prevented fulfillment of actions necessary to 1)

Shutdown the reactor and maintain it in a safe shutdown condition, 2) Remove residual heat, 3) Control the release of radioactive material, or 4) Mitigate the consequences of an accident.

This event did not result in any radiological release to the public.

This event did not result in a safety system functional failure.

NRC FORM 36M (9*2007)

PRINTED ON RECYCLED PAPER

E.

Corrective Actions

LICENSEE EVENT REPORT (LER) U,S. NUCLEAR REGULATORY COMMISSION CONTINUATION SHEET

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2009 002 00 The corrective actions include:

Perform a functional check, and calibration of all Unit 2 RCP overcurrent protective relays (completed).

Perform a functional check, and calibration of the overcurrent protective relays on pumps with the same overcurrent protective relay as the 2C RCP.

Revise the procedure for the calibration of overcurrent protective relays to include procedure steps for functional check and calibration, as required, for the high dropout instantaneous values associated with overcurrent time delay relays.

F.

Previous Occurrences

There have been no similar Licensee Event Report events at Braidwood Station in the last three years.

G.

Component Failure Data

Manufacturer WestinghouseNomenclature Reactor Coolant Pump Protective Relay Mfg. Part Number N/A PRINTED ON RECYCLED PAPER