05000296/LER-2013-003, Regarding Automatic Reactor Shutdown Due to an Actuation of the Reactor Protection System from a Turbine Trip

Regarding Automatic Reactor Shutdown Due to an Actuation of the Reactor Protection System from a Turbine Trip

ML13120A307
Person / Time
Site:	Browns Ferry
Issue date:	04/26/2013
From:	Polson K Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LER 13-003-00
Download: ML13120A307 (10)

LER-2013-003, Regarding Automatic Reactor Shutdown Due to an Actuation of the Reactor Protection System from a Turbine Trip

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 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)(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
2962013003R00 - NRC Website

text

Tennessee Valley Authority, Post Office Box 2000, Decatur, Alabama 35609-2000 April 26, 2013 10 CFR 50.73 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Browns Ferry Nuclear Plant, Unit 3 Facility Operating License No. DPR-68 NRC Docket No. 50-296

Subject:

Licensee Event Report 50-296/2013-003-00 The enclosed Licensee Event Report provides details of an automatic reactor shutdown due to an actuation of the reactor protection system from a turbine trip. The Tennessee Valley Authority is submitting this report in accordance with Title 10 of the Code of Federal Regulations (10 CFR) 50.73(a)(2)(iv)(A).

There are no new regulatory commitments contained in this letter. Should you have any questions concerning this submittal, please contact J. E. Emens, Jr., Nuclear Site Licensing Manager, at (256) 729-2636.

Respectfully, K. J. Poison Vice President

Enclosure:

Licensee Event Report 50-296/2013-003 Automatic Reactor Shutdown Due to an Actuation of the Reactor Protection System From a Turbine Trip cc (w/ Enclosure):

NRC Regional Administrator - Region II NRC Senior Resident Inspector - Browns Ferry Nuclear Plant

ENCLOSURE Browns Ferry Nuclear Plant Unit 3 Licensee Event Report 50-296/2013-003-00 Automatic Reactor Shutdown Due to an Actuation of the Reactor Protection System From a Turbine Trip See Enclosed

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 EXPIRES 10/31/2013 (10-2010)

, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

3. PAGE Browns Ferry Nuclear Plant, Unit 3 05000296 1 of 8

4. TITLE: Automatic Reactor Shutdown Due to an Actuation of the Reactor Protection System From a Turbine Trip

5. EVENT DATE

6. LER NUMBER

7. REPORT DATE

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

N/A 05000 FACILITY NAME DOCKET NUMBER 02 25 2013 2013 -

003 -

00 04 26 2013 N/A 05000

9. OPERATING MODE

11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)

[1 20.2201(b)

El 20.2203(a)(3)(i)

[I 50.73(a)(2)(i)(C)

[I 50.73(a)(2)(vii) 0l 20.2201(d) 0l 20.2203(a)(3)(ii)

[E 50.73(a)(2)(ii)(A)

El 50.73(a)(2)(viii)(A)

El 20.2203(a)(1)

El 20.2203(a)(4)

El 50.73(a)(2)(ii)(B)

El 50.73(a)(2)(viii)(B) 0_ 20.2203(a)(2)(i) 0l 50.36(c)(1)(i)(A)

El 50.73(a)(2)(iii) 0 50.73(a)(2)(ix)(A)

10. POWER LEVEL El 20.2203(a)(2)(ii) 0l 50.36(c)(1)(ii)(A)

[

50.73(a)(2)(iv)(A)

El 50.73(a)(2)(x)

[I 20.2203(a)(2)(iii)

-] 50.36(c)(2)

El 50.73(a)(2)(v)(A)

El 73.71(a)(4)

El 20.2203(a)(2)(iv) 0l 50.46(a)(3)(ii)

El 50.73(a)(2)(v)(B)

El 73.71(a)(5) 092 El 20.2203(a)(2)(v) 0l 50.73(a)(2)(i)(A)

El 50.73(a)(2)(v)(C)

[E OTHER El 20.2203(a)(2)(vi)

El 50.73(a)(2)(i)(B)

[E 50.73(a)(2)(v)(D)

Specify in Abstract below or in indication of isolation. On February 25, 2013, at approximately 1514 hours0.0175 days <br />0.421 hours <br />0.0025 weeks <br />5.76077e-4 months <br /> CST, the isolation valves were deactivated satisfying TS LCO 3.6.1.3 Required Actions.

On February 25, 2013, at approximately 1415 hours0.0164 days <br />0.393 hours <br />0.00234 weeks <br />5.384075e-4 months <br /> CST, suppression pool water level exceeded -1.0 inches. Suppression pool water level could not be maintained within normal bands due to running HPCI system for pressure control and suppression pool water level was not able to be pumped down due to the reactor feedwater piping separation.

The BFN, Unit 3, TS LCO 3.6.2.2 requires that suppression pool water level shall be greater than or equal to -6.25 inches with and -7.25 inches without differential pressure control and less than or equal to -1.0 inches in reactor Modes 1, 2, and 3. On February 25, 2013, at approximately 1415 hours0.0164 days <br />0.393 hours <br />0.00234 weeks <br />5.384075e-4 months <br /> CST, TS 3.6.2.2 Condition A was entered due to suppression pool water level not within limits. The TS 3.6.2.2 Required Action A. 1 requires suppression pool water level to be restored within limits within two hours. On February 25, 2013, at approximately 1615 hours0.0187 days <br />0.449 hours <br />0.00267 weeks <br />6.145075e-4 months <br /> CST, TS 3.6.2.2 Condition B was entered due to required action and completion time not being met for TS 3.6.2.2 Condition A. The TS 3.6.2.2 Required Action B.1 requires the unit to be in Mode 3 (Hot Shutdown) within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in Mode 4 (Cold Shutdown) within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The BFN, Unit 3, entered Mode 3 on February 25, 2013, at approximately 1313 hours0.0152 days <br />0.365 hours <br />0.00217 weeks <br />4.995965e-4 months <br /> CST and Mode 4 on February 25, 2013, at approximately 2141 hours0.0248 days <br />0.595 hours <br />0.00354 weeks <br />8.146505e-4 months <br /> CST.

B. Status of structures, components, or systems that were inoperable at the start of the event and that contributed to the event

The reactor feedwater long cycle valve return line connection to the miscellaneous drain header caused a rapid loss of condenser vacuum which caused the turbine trip.

C. Dates and approximate times of occurrences

February 25, 2013, at 1313 hours0.0152 days <br />0.365 hours <br />0.00217 weeks <br />4.995965e-4 months <br /> CST February 25, 2013, at 1324 hours0.0153 days <br />0.368 hours <br />0.00219 weeks <br />5.03782e-4 months <br /> CST February 25, 2013, at 1326 hours0.0153 days <br />0.368 hours <br />0.00219 weeks <br />5.04543e-4 months <br /> CST February 25, 2013, at 1415 hours0.0164 days <br />0.393 hours <br />0.00234 weeks <br />5.384075e-4 months <br /> CST Reactor automatically scrammed due to actuation of the RPS from a turbine trip.

Also, TS 3.6.1.3 Condition A was entered due to one or more penetration flow paths with one PCIV inoperable. The BFN, Unit 3, entered Mode 3.

The RCIC system was initiated to control reactor water level.

The HPCI system was initiated to control reactor pressure.

The TS 3.6.2.2 Condition A was entered due to suppression pool water level not within limits.

February 25, 2013, at 1422 hours0.0165 days <br />0.395 hours <br />0.00235 weeks <br />5.41071e-4 months <br /> CST February 25, 2013, at 1514 hours0.0175 days <br />0.421 hours <br />0.0025 weeks <br />5.76077e-4 months <br /> CST February 25, 2013, at 1615 hours0.0187 days <br />0.449 hours <br />0.00267 weeks <br />6.145075e-4 months <br /> CST Indication was subsequently restored following restoration of containment isolation signals, and the Drywell CAM Inboard Return Isolation Valve was manually isolated.

Isolation valves were deactivated to satisfy TS LCO 3.6.1.3 Required Actions.

The TS 3.6.2.2 Condition B was entered due to required action and completion time not being met for TS 3.6.2.2 Condition A.

February 25, 2013, at 1649 hours0.0191 days <br />0.458 hours <br />0.00273 weeks <br />6.274445e-4 months <br /> CST The BFN reported event to the NRC.

February 25, 2013, at 2141 hours0.0248 days <br />0.595 hours <br />0.00354 weeks <br />8.146505e-4 months <br /> CST The BFN, Unit 3, entered Mode 4.

D. Manufacturer and model number (or other identification) of each component that failed during the event:

A section of reactor feedwater piping, BFN-3-MISC-003, separated resulting in a loss of condenser vacuum.

E. Other systems or secondary functions affected

There were no other systems or secondary functions affected.

F. Method of discovery of each component or system failure or procedural error

Operations personnel observed a scram turbine generator load reject annunciator in the control room resulting in a reactor scram. The reactor feedwater piping separation was discovered by Operations personnel via a remote camera.

G. The failure mode, mechanism, and effect of each failed component, if known:

The failure mode of the feedwater long cycle return line connection to the miscellaneous drain header was due to excessive vibration. The excessive vibration was a result of pipe movement due to seat leakage of one or more of the feedwater long cycle valves. The valve leakage caused flashing in the drain header which caused subsequent failure of the miscellaneous drain header pipe wall. This failure caused a loss of vacuum on the turbine which caused the turbine to trip. When the turbine tripped on low condenser vacuum, an automatic reactor scram occurred.

H. Operator actions

Operations personnel responded in accordance with Emergency Operating Instructions on Low Reactor Water Level. Also, Operations personnel responded in accordance with the Abnormal Operating Instructions for the automatic scram.

I. Automatically and manually initiated safety system responses

The BFN, Unit 3, reactor automatically scrammed due to the actuation of the RPS from a turbine trip. The MSIVs were manually closed. There was one SRV that was

manually operated to maintain reactor pressure due to the unavailability of the Main Turbine Bypass Valves upon loss of condenser vacuum. Reactor water level was controlled with the RCIC system and reactor pressure was controlled with the HPCI system.

III.

Cause of the event

A. The cause of each component or system failure or personnel error, if known:

Direct Cause The direct cause of the event was the failure of the feedwater long cycle return line connection to the miscellaneous drain header due to cyclic fatigue. The cyclic fatigue was a result of pipe movement due to seat leakage of one or more of the feedwater long cycle valves. The pipe movement caused the material stress on the pipe connection to be much greater than the endurance level of 10,000 pounds per square inch (psi) for the carbon steel material.

Root Cause Valve design used for flow control valves 3-FCV-003-0071, -0072, and -0073 is incorrect for the specified application. This led to valve leakage which caused flashing in the drain header with subsequent failure of the miscellaneous drain header pipe wall due to excessive vibration. The wall failure caused a loss of condenser vacuum with a subsequent reactor scram. Currently, the valves used in this application are solid disc gate valves. These valves are not ideally suited for severe (high differential pressure) applications.

B. The cause(s) and circumstances for each human performance related root

cause

Root Cause Station personnel do not consistently consider risk when making decisions. Work orders to replace the BFN, Unit 3, feedwater long cycle valves have been cancelled with little or no justification. This exposed BFN, Unit 3, to continued risk of failed piping downstream of the feedwater long cycle valves providing the potential for degraded, or loss of, condenser vacuum. Work orders written in 2007 to replace valves with a more suitable valve were never worked and were subsequently cancelled in 2010.

IV. Analysis of the event

The Tennessee Valley Authority (TVA) is submitting this report in accordance with Title 10 of the Code of Federal Regulations (10 CFR) 50.73(a)(2)(iv)(A) as 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: RPS which includes a reactor scram or a reactor trip, general containment isolation signals affecting containment isolation valves in more than one system or MSIVs, HPCI, and RCIC.

The BFN, Unit 3, scrammed due to the 8 inch reactor feedwater long cycle return pipe breaking at the connection with the 24 inch miscellaneous drain header which caused a rapid loss of

condenser vacuum. Once BFN, Unit 3, was shut down, a walkdown was performed to assess the 8 inch reactor feedwater long cycle return pipe for any additional damage and ensure the remainder of the piping system was per design requirements. During the walkdown, it was verified that the support spacing met code requirements for the 8 inch piping. Although no significant damage was noted during the walkdown, it was determined that the piping had been experiencing a significant amount of movement during the last operating cycle based on rub marks noted on the piping at individual support locations. It was determined that the excessive pipe movement was due to leakage past the normally closed feedwater long cycle flow control valve 3-FCV-003-0073. Also, normally closed feedwater long cycle flow control valves 3-FCV-003-0071 and 3-FCV-003-0072 were potentially leaking past their seats. When these normally closed valves leak past their seats, high pressure and temperature water was exposed to a low pressure (condenser vacuum) normally stagnant system which in turn caused the excessive pipe movement. The analysis used for this condition determined that material fatigue was the most probable cause of the failure of the connection with the 24 inch miscellaneous drain header due to the material stress being much greater than the endurance limit of 10,000 psi for the carbon steel material.

The feedwater long cycle valves are solid disc gate valves. These valves have continually been identified as leaking past their seats since installation and have been repaired numerous times.

The repair to these valves was not a permanent fix because the valves continued to leak. The leakage from these valves has caused piping erosion that has lead to through wall leaks.

Additionally, the leakage has caused failure of the feedwater long cycle piping at welding joints and at the wall of the miscellaneous drain header.

V.

Assessment of Safety Consequences

The RPS provides timely protection against the onset and consequences of conditions that threaten the integrity of the fuel barrier and the nuclear system process barrier, i.e.

fuel cladding and reactor coolant system pressure boundary respectively. The RPS is designed such that no single failure can prevent a reactor scram, and the RPS is designed to automatically shutdown the reactor based on parameters which deviate from normal.

A. Availability of systems or components that could have performed the same function as the components and systems that failed during the event:

The MSIVs were manually closed. There was one SRV that was manually operated to maintain reactor pressure due to the unavailability of the Main Turbine Bypass Valves due to the loss of condenser vacuum. All systems responded as expected to the turbine trip. No ECCS or RCIC system reactor water level initiation set points were reached. Reactor water level was controlled by the RCIC system and reactor pressure was controlled with the HPCI system.

B. For events that occurred when the reactor was shut down, availability of systems or components needed to shutdown the reactor and maintain safe shutdown conditions, remove residual heat, control the release of radioactive material, or mitigate the consequences of an accident:

All safety systems remained available during this event and operated as designed.

C. For failure that rendered a train of a safety system inoperable, an estimate of the elapsed time from discovery of the failure until the train was returned to service:

There were no safety systems rendered inoperable as a result of this event.

Therefore, TVA concluded that there was minimal safety significance for this event.

VI.

Corrective Actions

Corrective Actions are being managed by TVA's corrective action program under Problem Evaluation Reports (PERs) 516455 and 687732.

Immediate Corrective Actions

A Temporary Alteration Control Form was initiated to cut and cap the feedwater long cycle lines downstream of each feedwater long cycle valve and upstream of the miscellaneous drain header.

Corrective Actions to Prevent Recurrence

1. Issue and implement design change to replace BFN, Units 1, 2, and 3, feedwater long cycle valves.

2. Implemented a Strategic Performance Management process to reinforce and institutionalize conservative decision making principles at BFN.

3. Establish initial and continuing training requirements, and develop and deliver training to provide expected behaviors for leaders and craft that support their roles and responsibilities.

VII.

Additional Information

A. Previous similar events at the same plant:

A search of BFN Licensee Event Reports (LERs) for Units 1, 2, and 3 for the last several years identified LER 50-296/2013-002-00, Manual Actuation of Reactor Core Isolation Cooling System During Reactor Shutdown. This LER identified a similar condition concerning the reactor feedwater piping separation on the feedwater long cycle lines. The analysis for that LER was ongoing at the time of this event; therefore, the corrective actions for that LER would not have prevented this event from occurring.

A search was performed on the BFN corrective action program. Similar PERs related to the condition which caused the event reported in this LER are PERs 41131 and 52947.

PER 41131 states, "Chemistry identified a threat to the integrity of the welded pipe connection of the 8 inch long cycle return line at the miscellaneous drain header and potential condenser vacuum issues with movement observed from the line giving a strong indication of long cycle valve leakage with subsequent flashing occurring in the long cycle return line." The PER 41131 was closed to PER 52947 which identified that excessive movement of the long cycle piping caused the weld to the miscellaneous drain header to fatigue. The piping movement was caused by flashing in the line due to leakage through one or more of the feedwater long cycle valves. However, the corrective actions did not address the risk of failing to repair and maintain the valves that were the cause of the pipe movement which resulted ultimately in the failure of the 8 inch long cycle feedwater return line at the tie into the 24 inch miscellaneous drain header.

As a result of not addressing the actual failure mode and recognizing the risk associated with the long cycle return valves, BFN missed an opportunity to utilize available Operating Experience to prevent this event.

B. Additional Information

There is no additional information.

C. Safety System Functional Failure Consideration:

In accordance with Nuclear Energy Institute (NEI) 99-02, this condition is not considered a safety system functional failure.

D. Scram with Complications Consideration:

In accordance with NEI 99-02, this event is considered an unplanned scram with complications.

VIII. COMMITMENTS

There are no commitments.