05000296/LER-2013-002, Regarding Manual Actuation of Reactor Core Isolation Cooling System During Reactor Shutdown

Regarding Manual Actuation of Reactor Core Isolation Cooling System During Reactor Shutdown

ML13109A025
Person / Time
Site:	Browns Ferry
Issue date:	04/12/2013
From:	Polson K Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LER 2013-002-00
Download: ML13109A025 (7)

LER-2013-002, Regarding Manual Actuation of Reactor Core Isolation Cooling System During Reactor Shutdown

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
2962013002R00 - NRC Website

text

Tennessee Valley Authority, Post Office Box 2000, Decatur, Alabama 35609-2000 April 12, 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 50-296/2013-002-00 The enclosed Licensee Event Report provides details of the manual actuation of the Reactor Core Isolation Cooling system during a planned reactor shutdown. 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, Vice President

Enclosure:

Licensee Event Report 50-296/2013-002 Manual Actuation of Reactor Core Isolation Cooling System During Reactor Shutdown 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 50-296/2013-002-00 Manual Actuation of Reactor Core Isolation Cooling System During Reactor Shutdown 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 5

4. TITLE: Manual Actuation of Reactor Core Isolation Cooling System During Reactor Shutdown

5. EVENT DATE

6. LER NUMBER

7. REPORT DATE

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

MONTH DAY YEAR N/A 0500 FACILITY NAME DOCKET NUMBER 02 11 2013 2013 002 -

00 04 12 2013 N/A 05000

9. OPERATING MODE

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

[O 20.2201(b)

El 20.2203(a)(3)(i)

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

El 50.73(a)(2)(vii)

El 20.2201(d)

El 20.2203(a)(3)(ii)

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

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

[I 20.2203(a)(1)

[I 20.2203(a)(4)

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

El 50.73(a)(2)(viii)(B)

[__ 20.2203(a)(2)(i)

El 50.36(c)(1)(i)(A) 0l 50.73(a)(2)(iii)

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

10. POWER LEVEL 0l 20.2203(a)(2)(ii) 0l 50.36(c)(1)(ii)(A) 0 50.73(a)(2)(iv)(A)

El 50.73(a)(2)(x)

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

[I 50.36(c)(2)

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

El 73.71(a)(4)

El 20.2203(a)(2)(iv)

El 50.46(a)(3)(ii)

[I 50.73(a)(2)(v)(B) 0l 73.71(a)(5) 000 El 20.2203(a)(2)(v)

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

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

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

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

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

Specify in Abstract below or in NRC Forrn 366A

12. LICENSEE CONTACT FOR THIS LER FACILITY NAME TELEPHONE NUMBER (Include Area Code)

Mark Acker, Licensing Engineer 256-729-7533

14. SUPPLEMENTAL REPORT EXPECTED 0

YES (If yes, complete 15. EXPECTED SUBMISSION DATE)

ABSTRACT (Limit to 1400 spaces. i.e., approximately 15 single-spaced typewritten lines)

On February 11, 2013, at 0613 hours0.00709 days <br />0.17 hours <br />0.00101 weeks <br />2.332465e-4 months <br /> Central Standard Time (CST), the Reactor Core Isolation Cooling (RCIC) system was manually started during a planned Browns Ferry Nuclear Plant, Unit 3, reactor shutdown. A Reactor Feedwater recirculation piping separation resulted in the loss of condenser vacuum and subsequent unavailability of the Main Turbine Bypass Valves.

The RCIC system was manually started to control reactor water level in anticipation of loss of Reactor Feedwater Pumps tripping on low vacuum. Safety Relief Valves were manually operated to maintain reactor pressure. No Emergency Core Cooling System or RCIC system reactor water level initiation set points were reached. The RCIC system was removed from service on February 11, 2013, at 1449 CST.

The causal analysis for this event is ongoing. Upon completion of the causal analysis, the Tennessee Valley Authority will submit a supplement to this Licensee Event Report.

NRC FORM 366 (10-2010)

1.

Plant Operating Conditions Before the Event

At the time of the event, Browns Ferry Nuclear Plant (BFN), Unit 3, was in Mode 3 at zero percent rated thermal power during a planned reactor shutdown.

I1.

Description of Events A. Event:

On February 11, 2013, at 0613 hours0.00709 days <br />0.17 hours <br />0.00101 weeks <br />2.332465e-4 months <br /> Central Standard Time (CST), the Reactor Core Isolation Cooling (RCIC) system [BN] was manually started during a planned BFN, Unit 3, reactor shutdown. A Reactor Feedwater [SJ] recirculation piping separation resulted in the loss of condenser vacuum and subsequent unavailability of the Main Turbine Bypass Valves [V] [JI]. The RCIC system was manually started to control reactor water level in anticipation of loss of Reactor Feedwater Pumps (RFPs) [P] tripping on low vacuum. Safety Relief Valves (SRVs) [SB] were manually operated to maintain reactor pressure. No Emergency Core Cooling System (ECCS) [BJ][BO][BM] or RCIC system reactor water level initiation set points were reached. The RCIC system was removed from service on February 11, 2013, at 1449 CST.

The causal analysis for this event is ongoing. Upon completion of the causal analysis, the Tennessee Valley Authority (TVA) will submit a supplement to this Licensee Event Report (LER).

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

There were no inoperable structures, components, or systems that contributed to this event.

C. Dates and approximate times of occurrences

February 11, 2013 at 0424 CST Operations initiated a planned reactor manual scram.

February 11, 2013 at 0613 CST The RCIC system was manually started to control reactor water level.

February 11, 2013 at 1449 CST The RCIC system was removed from service.

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

A section of Reactor Feedwater recirculation piping, BFN-3-MISC-003, separated resulting is 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 received indication of decreasing condenser vacuum in the reactor feedwater recirculation system [AD].

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

A section of Reactor Feedwater recirculation piping separated resulting in a loss of condenser vacuum.

The causal analysis for this event is ongoing. Upon completion of the causal analysis, the TVA will submit a supplement to this LER.

H. Operator actions

Operations manually operated the RCIC system and SRVs to control reactor water level and pressure during the planned BFN, Unit 3, reactor shutdown.

I. Automatically and manually initiated safety system responses

Operations manually operated the RCIC system and SRVs to control reactor water level and pressure during the planned BFN, Unit 3, reactor shutdown.

III.

Cause of the event

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

The causal analysis for this event is ongoing. Upon completion of the causal analysis, the TVA will submit a supplement to this LER.

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

cause

The causal analysis for this event is ongoing. Upon completion of the causal analysis, the TVA will submit a supplement to this LER.

IV.

Analysis of the event

The 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),

which includes the RCIC system.

The causal analysis for this event is ongoing. Upon completion of the causal analysis, the TVA will submit a supplement to this LER.

In late December 2012, BFN, Unit 3, experienced an unusual rise in water in-leakage to the Radwaste building [NE], concurrent with a higher than normal river water level. The amount of in-leakage challenged the capacity of the Radwaste system [WD]. The source of in-leakage was determined to be the BFN, Unit 3, Condenser Circulating Water (CCW) [SG] conduit and the decision was made to conduct a planned outage in February 2013, to allow repairs to the CCW conduit. A manual scram was initiated on February 11, 2013 at 0424 CST to start the planned outage.

On February 11, 2013, at 0613 hours0.00709 days <br />0.17 hours <br />0.00101 weeks <br />2.332465e-4 months <br /> CST, the RCIC system was manually started to control reactor water level because a Reactor Feedwater recirculation piping separation resulted in the loss of condenser vacuum and subsequent unavailability of the Main Turbine Bypass Valves. The RCIC system was manually started to control reactor water level in anticipation of loss of RFPs tripping on low vacuum. The SRVs were manually operated to maintain reactor pressure. No ECCS or RCIC system reactor water level initiation set points were reached. The RCIC system was removed from service on February 11, 2013, at 1449 CST.

V.

Assessment of Safety Consequences

The RCIC system is designed to operate either automatically or manually following reactor pressure vessel (RPV) isolation accompanied by a loss of coolant flow from the feedwater system to provide adequate core cooling and control of the RPV water level.

The RCIC system was manually initiated during this event to control reactor water level.

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

The RCIC system was manually initiated to control reactor water level and the SRVs were manually operated to control reactor pressure. The RCIC system maintained reactor water level in the prescribed band during this event. In addition to the RCIC system, the ECCS and Automatic Depressurization System were operable and available to provide core cooling if needed.

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:

The causal analysis for this event is ongoing. Upon completion of the causal analysis, the TVA will submit a supplement to this LER.

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:

The causal analysis for this event is ongoing. Upon completion of the causal analysis, the TVA will submit a supplement to this LER.

VI.

Corrective Actions

Corrective Actions are being managed by TVA's corrective action program under Problem Evaluation Report (PER) 710206.

The causal analysis for this event is ongoing. Upon completion of the causal analysis, the TVA will submit a supplement to this LER.

VII.

Additional Information

A. Previous similar events at the same plant:

A search of BFN LERs for Units 1, 2, and 3 for the last several years did not identify any similar events.

A search was performed on the BFN corrective action program. There were no similar PERs identified.

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:

This event did not result in an unplanned scram with complications.

VIII. COMMITMENTS

There are no commitments.