05000296/LER-2011-001, For Browns Ferry Nuclear Plant, Unit 3, Regarding Loss to Shutdown Cooling (RHR)

For Browns Ferry Nuclear Plant, Unit 3, Regarding Loss to Shutdown Cooling (RHR)

ML11199A050
Person / Time
Site:	Browns Ferry
Issue date:	07/11/2011
From:	Polson K Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LER 11-001-00
Download: ML11199A050 (11)

LER-2011-001, For Browns Ferry Nuclear Plant, Unit 3, Regarding Loss to Shutdown Cooling (RHR)

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(v), Loss of Safety Function 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)(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)(i)(B), Prohibited by Technical Specifications
2962011001R00 - NRC Website

text

Tennessee Valley Authority, Post Office Box 2000, Decatur, Alabama 35609-2000 July 11, 2011 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 50-29612011-001-00 On May 12, 2011, at 1825 hours0.0211 days <br />0.507 hours <br />0.00302 weeks <br />6.944125e-4 months <br /> Central Daylight Time (CDT), Browns Ferry Nuclear Plant (BFN)-Unit 3 was in Mode 4, as a result of the severe weather event that occurred on April 27, 2011. Relay maintenance activities in the Primary Containment Isolation System (PCIS) inadvertently de-energized a Group 2 PCIS relay, which initiated the Group 2 PCIS isolation. The Group 2 isolation caused a trip of the 3B Residual Heat Removal pump, which subsequently resulted in a loss of shutdown cooling (SDC). The relay wires were re-connected and SDC was restored on BFN-Unit 3 at 1905 hours0.022 days <br />0.529 hours <br />0.00315 weeks <br />7.248525e-4 months <br /> CDT.

The Tennessee Valley Authority is submitting a 60 day written report which is required in accordance with 10 CFR 50.73(a)(2)(v), any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to: (B)

Remove Residual Heat.

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.

Respecfffull K. J. Poison Vice President

Enclosure:

Licensee Event Report - Loss of Shutdown Cooling (RHR)

U.S. Nuclear Regulatory Commission Page 2 July 11, 2011 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 - Loss of Shutdown Cooling (RHR)

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 3 05000296 1 OF 8

4. TITLE Loss of Shutdown Cooling (RHR)

5. EVENT DATE

6. LER NUMBER

7. REPORT DATE

8. OTHER FACILITIES INVOLVED FACILITY NAME DOCKET NUMBER SEQUENTIAL REV MONTH DAY YEAR 05000 MONTH DAY YEAR YEAR NUMBER NO.

FACILITY NAME DOCKET NUMBER 05 12 2011 2011 001 00 07 11 2011 05000

.9. OPERATING MODE

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

[1 20.2201(b)

[I 20.2203(a)(3)(i)

El 50.73(a)(2)(i)(C)

[E 50.73(a)(2)(vii) 4 [1 20.2201(d)

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

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

[3 50.73(a)(2)(viii)(A)

E] 20.2203(a)(1)

[] 20.2203(a)(4)

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

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

__ 20.2203(a)(2)(i)

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

El 50.73(a)(2)(iii)

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

10. POWER LEVEL Q 20.2203(a)(2)(ii)

El 50.36(c)(1)(ii)(A)

[I 50.73(a)(2)(iv)(A)

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

El 20.2203(a)(2)(iii)

El 50.36(c)(2)

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

[1 73.71(a)(4) 000

-- 20.2203(a)(2)(iv)

[I 50.46(a)(3)(ii) 0 50.73(a)(2)(v)(B)

[1 73.71(a)(5)

El 20.2203(a)(2)(v)

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

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

[I OTHER o" 20.2203(a)(2)(vi)

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

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

Specify in Abstract below or in At 1840 hours0.0213 days <br />0.511 hours <br />0.00304 weeks <br />7.0012e-4 months <br /> CDT, electricians were directed to re-connect the wire. Operations personnel reset the Group 2 Isolation at 1845 hours0.0214 days <br />0.513 hours <br />0.00305 weeks <br />7.020225e-4 months <br /> CDT. On May 12, 2011, at approximately 1905 hours0.022 days <br />0.529 hours <br />0.00315 weeks <br />7.248525e-4 months <br /> CDT, the 3B RHR pump was returned to service and SDC was restored for BFN, Unit 3.

Moderator temperature prior to the event was 112.5 degrees Fahrenheit (F) and the highest moderator temperature recorded during the loss of SDC was 122 degrees F.

After the event, management discussions concluded that work to replace PCIS relays was to be placed on hold. Other work for BFN, Unit 3, was to be authorized by the OCC Unit Operations Manager on a case-by-case basis and this was communicated to the Shift Manager. Operations personnel were directed to review all work packages prior to restarting BFN, Unit 3.

B.

Inoperable Structures, Components, or Systems that Contributed to the Event There were no inoperable structures, components, or systems that contributed to this event.

C.

Dates and Approximate Times of Major Occurrences

On April 27, 2011, at 1636 hours0.0189 days <br />0.454 hours <br />0.00271 weeks <br />6.22498e-4 months <br /> CDT On May 12, 2011, at 1825 hours0.0211 days <br />0.507 hours <br />0.00302 weeks <br />6.944125e-4 months <br /> CDT at 1830 hours0.0212 days <br />0.508 hours <br />0.00303 weeks <br />6.96315e-4 months <br /> CDT at 1840 hours0.0213 days <br />0.511 hours <br />0.00304 weeks <br />7.0012e-4 months <br /> CDT at 1845 hours0.0214 days <br />0.513 hours <br />0.00305 weeks <br />7.020225e-4 months <br /> CDT All BFN units entered Mode 4 as a result of a severe weather event causing a loss of off-site power to all units.

Operations received an annunciation notification that the 3B RHR pump tripped and an "A" Channel Group 2 Isolation is initiated.

A report was received by Operations that there were no breaker [BKR] issues and also no apparent problems for the 3B RHR pump.

Electricians report to Operations that when they lifted a neutral wire on relay BFN-3-RLY-064-16AK56, a Group 2 PCIS Isolation occurred. Electricians were instructed to re-connect the wire.

As instructed by Operations, the electricians re-connected wire for relay BFN-3-RLY-064-16AK56, and Operations reset the Group 2 isolation.

at 1904 hours0.022 days <br />0.529 hours <br />0.00315 weeks <br />7.24472e-4 months <br /> CDT at 1905 hours0.022 days <br />0.529 hours <br />0.00315 weeks <br />7.248525e-4 months <br /> CDT at 1930 hours0.0223 days <br />0.536 hours <br />0.00319 weeks <br />7.34365e-4 months <br /> CDT at 2220 hours0.0257 days <br />0.617 hours <br />0.00367 weeks <br />8.4471e-4 months <br /> CDT On May 13, 2011, at 0053 hours6.134259e-4 days <br />0.0147 hours <br />8.763227e-5 weeks <br />2.01665e-5 months <br /> CDT Operations placed capacitor banks in manual for 3B RHR pump start Operations returned Loop II to RHR SDC and confirmed that the 3B RHR pump was running.

Operations review determined that when RHR Pump 3B tripped, the motor tripout alarm was not received.

Operations logs report time to boil of 4.13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> and 3B RHR pump in SDC.

Notification was made in accordance with 10 CFR 50.72(b)(3)(v).

D.

Other Systems or Secondary Functions Affected

There were no other systems or secondary functions affected.

E.

Method of Discovery

Several annunciation windows alarmed and alerted Operations to this event. These were:

Window 3-XA-55-3E, window 13, [3B RHR SYS II PUMP B TRIPPED].

Window 3-XA-55-3D, window 11, [RHR SD CLG FLOW LOW].

Window 3-XA-55-23C, window 26, [4160 V SD BD 3EC MOTOR OL OR TRIP].

These annunciation windows alerted Operations that the 3B RHR pump had tripped and that a Group 2 isolation had occurred.

A single annunciation window, Window 3-XA-55-8C, window 33 [, MOTOR TRIPOUT],

did not illuminate.

F.

Operator Actions

After determining the cause of the event, Operations had the neutral wire re-connected and reset the Group 2 isolation. The 3B RHR pump was restarted. Operations checked to ensure that the 38 RHR pump was running and that SDC had been restored.

G.

Safety System Responses Relay maintenance activities on a Group 1 PCIS relay interrupted the neutral in a Group 2 PCIS relay, initiating a Group 2 PCIS isolation which resulted in a loss of SDC. The relays caused a valve to close, resulting in the 3B RHR pump trip. Relays were reset and SDC was restored. SDC was lost for approximately 40 minutes. Moderator temperature prior to the event was 112.5 degrees F and the highest moderator temperature recorded during the loss of SDC was 122 degrees F. All plant equipment and systems functioned as designed, except for a single annunciation window, Window 3-XA-55-8C, window 33 [, MOTOR TRIPOUT], which did not illuminate.

Ill.

CAUSE OF THE EVENT

A.

Immediate Cause Upon discussion with Operations, the electricians were directed to lift the leads while energized. The logic prints reviewed did not show the neutral for the relays. The direct cause of this event was the lifting of the wiring that was landed on the common side of a relay coil.

B.

Root Cause The work package details did not include specific work precautions or instructions to require that jumpers be installed to prevent the loss of SDC. It was not obvious that the work package had been planned for refueling outage conditions and did not take into consideration that the desired work activities could not be performed while the RHR System was in service.

C.

Contributina Factors

1. There was inadequate training on electrical fundamentals in the area of plant wiring and plant configuration.

2. There was a lack of consistency with the decisions and actions regarding the use of the Risk Management Process for outage scope additions. The outage scope process does not evaluate risk for the addition of pre-planned packages that were intended to be worked in a different mode of plant operation.

3. Conservative decision making was not used to evaluate the work task.

Maintenance and Operations Work Control discussed continuing this work with voltage present on the relay. Engineering was not consulted.

IV.

ANALYSIS OF THE EVENT

A. Evaluation of Plant Systems/Components Originally, 17 PCIS relays were included in the forced outage work scope. After review by the Operations Manager, three PCIS relays were identified as directly affecting SDC and were removed from the scope of work. With the identification of these three relays, personnel were convinced that all relays had been identified that were high risk.

This outage work package originally was screened as low risk when it was reviewed the first time due to the evolution taking place during a normal refueling outage (RFO) with plant conditions in the configuration assumed in the earlier review. The procedure was checked "NO" for causing an Engineered Safety Feature actuation due to the package being planned for an RFO with both loops of RHR out of service and Alternate Decay Heat Removal in service.

Personnel statements indicate that maintenance personnel discussed performance of the PM task with Operations before beginning work and obtained approval.

However, personnel statements do not indicate that the pre-job briefing for the relay replacement discussed whether or not the work order was appropriately planned for the current plant conditions.

Upon discussion with Operations, i.e., the Work Control center, the Electricians were directed to lift the leads while energized. The logic prints did not show the neutral for the relays. If a better questioning attitude had been displayed by those involved, there is a higher probability that the wiring diagram would have been reviewed which would have identified the neutral wiring configuration on the relays.

The Tennessee Valley Authority is reporting this event in accordance with 10 CFR 50.73 (a)(2)(v), as any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to (B) Remove Residual Heat.

V.

ASSESSMENT OF SAFETY CONSEQUENCES

For a loss of SDC, no unique safety actions are required. In these cases, SDC is re-established using other normal SDC equipment. In cases where the RHR SDC suction line becomes inoperable, a unique requirement for cooling arises. In cases in which the reactor vessel head is off, either half of the RHR-Low Pressure Coolant Injection (LPCI) mode can be used to maintain water level. In cases in which the reactor vessel head is on and the system can be pressurized, the LPCI system, main

steam relief valves (manually operated), and RHR-torus cooling mode can be used to maintain water level and remove decay heat.

Moderator temperature prior to the event was 112.5 degrees F and the highest moderator temperature recorded during the loss of SDC was 122 degrees F. Time to boil at the time of event was 4.13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />. The small reported rise in moderator temperature would have resulted in a small decrease in the time to boil. Because of the small temperature rise and the prompt restoration of RHR SDC to maintain water level and remove decay heat, there were no safety consequences associated with this event.

VI.

CORRECTIVE ACTION

A.

Immediate Corrective Actions

Electricians re-landed neutral wire for relay BFN-3-RLY-64-16AK56.

Operations reset the Group 2 Isolation at 1845 hours0.0214 days <br />0.513 hours <br />0.00305 weeks <br />7.020225e-4 months <br /> CDT.

A work stoppage was issued by the Outage Director to cease all plant work activities until an initial investigation and review of the incident were performed.

Plant management and supervisory personnel were briefed and instructed to discuss the event with their employees to disseminate this information to the site organizations.

B.

Corrective Actions to Prevent Recurrence Determine the scope of relay maintenance procedures which require revisions to verify that approved instructions are included in the implementing work plan to identify and place jumpers needed to maintain system logic during any work involving wire lifts which are part of a series (daisy-chain) circuit configuration.

Verify that approved instructions are included in the applicable implementing work plan to identify and place jumpers needed to maintain system logic during any work involving wire lifts which are part of a daisy-chain circuit configuration.

Re-evaluate and revise existing PCIS work packages to verify that approved instructions are included in the implementing work plan to identify and place jumpers needed to maintain system logic during any work involving wire lifts which are part of a daisy-chain circuit configuration.

Plant management decided that if a refueling outage work order was being added to a forced outage per Procedure NPG-SPP-7.2.8, Outage Scope Control, the work order should be returned to a planning status for a review of the work order to verify that the work order can be performed in the current plant conditions or to identify the plant conditions needed to perform the work order.

.JRC VII.

ADDITIONAL INFORMATION

A.

Failed Components Window 3-XA-55-8C, window 33 [MOTOR TRIPOUT], did not illuminate when RHR Pump 3B tripped. This is documented in Problem Evaluation Report (PER) 368629.

B.

Previous Similar Events

A review of operations logs and the Corrective Action Program database concluded there were no previous similar events with a similar cause.

C.

Additional Information

The corrective action document associated with this event is PER 368764.

D.

Safety System Functional Failure Consideration This event is classified as a safety-system functional failure according to NEI 99-02.

E.

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

VII.

COMMITMENTS

There are no commitments.