05000259/LER-2011-005, Regarding Reactor Water Level Scram Due to Distracted Operations Crew

Regarding Reactor Water Level Scram Due to Distracted Operations Crew

ML11180A007
Person / Time
Site:	Browns Ferry
Issue date:	06/27/2011
From:	Polson K Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LER 11-005-00
Download: ML11180A007 (9)

LER-2011-005, Regarding Reactor Water Level Scram Due to Distracted Operations Crew

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(iv)(B), 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
2592011005R00 - NRC Website

text

Tennessee Valley Authority, Post Office Box 2000, Decatur, Alabama 35609-2000 June 27, 2011 10 CFR 50.73 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Browns Ferry Nuclear Plant, Unit 1 Facility Operating License No. DPR-33 NRC Docket No. 50-259

Subject:

Licensee Event Report 50-259/2011-005-00 On April 27, 2011, severe weather in the Tennessee Valley service area caused grid instability and a significant loss of offsite power that resulted in a scram of all three units at Browns Ferry Nuclear Plant.

At 2120 hours0.0245 days <br />0.589 hours <br />0.00351 weeks <br />8.0666e-4 months <br />, Unit 1 experienced a low reactor water level scram due to the reactor water level lowering to +2 inches. The Tennessee Valley Authority is submitting this report in accordance with 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), except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.

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 - Reactor Water Level Scram Due to Distracted Operations Crew

U.S. Nuclear Regulatory Commission Page 2 June 27, 2011 cc (w/ Enclosure):

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

ENCLOSURE Browns Ferry Nuclear Plant, Unit 1 Licensee Event Report - Reactor Water Level Scram Due to Distracted Operations Crew See Attached

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 1 05000259 1 OF 6

4. TITLE Reactor Water Level Scram Due to Distracted Operations Crew

5. EVENT DATE

6. LER NUMBER

7. REPORT DATE

8. OTHER FACILITIES INVOLVED T

R FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR SEQUENTIAL REV MONTH DAY YEAR N/A 05000 MONTH I A

YEARYEA NUMBER NO I

FACILITY NAME DOCKET NUMBER 04 27 2011 2011 005 00 06 27 2011 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)

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

[I 50.73(a)(2)(vii) 3 [1 20.2201(d)

E] 20.2203(a)(3)(ii)

El 50.73(a)(2)(ii)(A)

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

[I 20.2203(a)(1)

El 20.2203(a)(4)

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

[I 50.73(a)(2)(viii)(B)

El 20.2203(a)(2)(i)

El 50.36(c)(1)(i)(A)

[I 50.73(a)(2)(iii)

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

10. POWER LEVEL

[] 20.2203(a)(2)(ii)

El 50.36(c)(1)(ii)(A) 0 50.73(a)(2)(iv)(A)

[I 50.73(a)(2)(x)

El 20.2203(a)(2)(iii)

[I 50.36(c)(2)

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

El 73.71(a)(4) 000 El 20.2203(a)(2)(iv)

El 50.46(a)(3)(ii)

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

El 73.71(a)(5)

[I 20.2203(a)(2)(v)

El 50.73(a)(2)(i)(A)

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

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

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

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

Specify in Abstract below or in at 1622 hours0.0188 days <br />0.451 hours <br />0.00268 weeks <br />6.17171e-4 months <br /> at 1636 hours0.0189 days <br />0.454 hours <br />0.00271 weeks <br />6.22498e-4 months <br /> at 1701 hours0.0197 days <br />0.473 hours <br />0.00281 weeks <br />6.472305e-4 months <br /> at 2120 hours0.0245 days <br />0.589 hours <br />0.00351 weeks <br />8.0666e-4 months <br /> at 2238 hours0.0259 days <br />0.622 hours <br />0.0037 weeks <br />8.51559e-4 months <br /> April 28, 2011, at 1337 hours0.0155 days <br />0.371 hours <br />0.00221 weeks <br />5.087285e-4 months <br /> The first 161-kV line was lost. The other 161-kV line did not trip and provided the only sustained source of offsite power to the station during the event and recovery.

Units 1, 2, and 3 automatically scrammed due to power load unbalance and prompt turbine trips. Units 1, 2, and 3 entered Mode 3 (Hot Shutdown).

BFN declared a Notification of Unusual Event (NOUE) in accordance with Emergency Implementing Procedure (EPIP) 1, "Emergency Classification Procedure," Emergency Action Level 5.1.U - Loss of normal and alternate supply voltage to all unit-specific 4-kV [EA] shutdown boards for greater than 15 minutes and at least two Emergency Diesel Generators [EK]

(EDGs) supplying power to unit-specific 4-kV shutdown boards.

Unit 1 received a low reactor water level scram due to reactor water level lowering to +2 inches. At the time of this event, RCIC and CRD were injecting to the vessel and the reactor level band specified was +2 to +51 inches. A valid containment isolation signal was received and Groups 2, 3, 6, and 8 isolated as expected.

Operations personnel notified NRC of the low level reactor water level scram in accordance with 10 CFR 50.72(b)(3)(iv)(A).

Unit 1 entered Mode 4.

D. Other Systems or Secondary Functions Affected

None

E. Method of Discovery

The event was immediately discovered by operations personnel.

F. Operator Actions

Operations personnel immediately increased the RCIC flow controller output that increased the flow of water into the reactor vessel.

G. Safety System Response The Unit 1 RCIC turbine governor control system operated as designed. The governor control system responded properly during manual starts on April 27, 2011, at 1639 hours0.019 days <br />0.455 hours <br />0.00271 weeks <br />6.236395e-4 months <br /> and on April 28, 2011, at 0104 hours0.0012 days <br />0.0289 hours <br />1.719577e-4 weeks <br />3.9572e-5 months <br />. The ramp generator signal converter (RGSC) responded properly and trended with the flow controller output during RCIC System operation. The Electric Governor-Magnetic Pickup (EG-M) Control Box responded properly and trended with the flow controller output during RCIC System operation. The flow controller output and EG-M Control Box output both reached their peak values on April 27, 2011, at 2124 hours0.0246 days <br />0.59 hours <br />0.00351 weeks <br />8.08182e-4 months <br /> following the reactor scram. Based on the above discussion, the Unit 1 RCIC turbine governor system operated as designed.

III. CAUSE OF THE EVENT

A. Immediate Cause Reactor water level lowered below +2 inches.

B. Root Cause The cause is Operations leadership not consistently driving and reinforcing standards.

C. Contributing Factors One of the significant contributing factors is:

Operations managers and supervisors do not consistently identify and correct gaps in performance.

IV. ANALYSIS OF THE EVENT

The operations crew became distracted due to other events and lost track of water level.

V. ASSESSMENT OF THE SAFETY CONSEQUENCES The function of the RCIC System is to provide makeup coolant to the reactor to respond to transient events. The RCIC System is not an Engineered Safety Features system and no credit is taken in the safety analyses for the RCIC System operation. Based upon its overall contribution to the reduction of overall plant risk, the system and its instrumentation meet Criterion 4 of the NRC Policy Statement. The RCIC System provides makeup water to the reactor. The system initiates if vessel level drops to the low level initiation point (Level 3;

+ 2 inches). The system provides makeup water to the reactor until the reactor vessel water level reaches the high water level (Level 8) trip. In this event, Operations personnel promptly restored water level. Based on the information provided in the LER, the Unit 1 RCIC turbine governor system operated as designed and there were no safety consequences associated with this event.

VI. CORRECTIVE ACTIONS

A. Immediate Corrective Action The Operations crew involved in this event received coaching.

B. Corrective Actions to Prevent Recurrence Significant corrective actions taken to prevent recurrence include:

1. Conducted training on transient mitigation strategies.

2. Establish weekly shift manager meetings to focus on performance improvement.

3. Establish the expectation that a Human Performance review board is conducted for each department clock reset and that at a minimum a Senior Reactor Operator (SRO) and Shift Manager (SM) will be in attendance for each board.

4. Operations Manager or Operations Superintendent to conduct board walk downs with each SRO to ensure expectations are being met.

5. Schedule and conduct individual meetings for SMs and Unit Supervisors with Operations management.

6. Establish monthly Unit Supervisor leadership meetings to focus on performance improvement.

VII ADDITIONAL INFORMATION A. Failed Components None B. Previous LERs or Similar Events None

C. Additional Information

The corrective action documents for this report are specified in PER 335574 and PER 363784.

D. Safety System Functional Failure Considerations The event is not a safety system functional failure according to NEI 99-02.

E. Scram With Complications Consideration This event was not a complicated scram according to NEI 99-02.

VIII. COMMITMENTS

None