05000260/LER-2009-005, Reactor Motor Operated Valve Board 2D & Residual Heat Removal Subsystem Inoperable Longer than Allowed by the Plants Technical Specifications

Reactor Motor Operated Valve Board 2D & Residual Heat Removal Subsystem Inoperable Longer than Allowed by the Plants Technical Specifications

ML092640116
Person / Time
Site:	Browns Ferry
Issue date:	09/16/2009
From:	West R Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LER 09-005-00
Download: ML092640116 (7)

LER-2009-005, Reactor Motor Operated Valve Board 2D & Residual Heat Removal Subsystem Inoperable Longer than Allowed by the Plants Technical Specifications

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 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)(v), Loss of Safety Function
2602009005R00 - NRC Website

text

Tennessee Valley Authority, Post Office Box 2000, Decatur, Alabama 35609-2000 September 16, 2009 10 CFR 50.73 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Browns Ferry Nuclear Plant Unit 2 Facility Operating License No. DPR-52 NRC Docket No. 50-260

Subject:

Licensee Event Report (LER) 50-26012009-005 The enclosed report provides details of the 480 Volt Reactor Motor Operated Valve Board 2D and Residual Heat Removal Subsystem being inoperable longer than allowed by the Plant's Technical Specifications. Therefore, TVA is reporting this in accordance with 10 CFR 50.73(a)(2)(i)(B) as an operation or condition prohibited by the Plant's Technical Specifications.

There are no new regulatory commitments contained in this letter. Should you have any questions concerning this submittal, please contact F. R. Godwin, Site Licensing and Industry Affairs Manager, at (256) 729-2636.

spectfully, R. G. West Vice President cc: See page 2

U.S. Nuclear Regulatory Commission Page 2 September 16, 2009 Enclosure cc (Enclosure):

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

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 EXPIRES 08/31/2010 (9-2007)

, the NRC may not conduct or sponsor, and a person is not required to respond to, the information (See reverse for required number of collection.

3. PAGE Browns Ferry Nuclear Plant Unit 2 05000260 1

1 of 5

4. TITLE: Reactor Motor Operated Valve Board And Residual Heat Removal Subsystem Inoperable Longer Than Allowed B, The Plant's Technical Specifications

6. EVENT DATE

6. LER NUMBER

7. REPORT DATE

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

MONTH DAY YEAR None NA FACILITY NAME DOCKET NUMBER 07 18 2009 2009 005 00 08 16 2009 None NA

9. OPERATING MODE

11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) 0 20.2201(b)

[0 20.2203(a)(3)(i) 0 50.73(a)(2)(i)(C)

[I 50.73(a)(2)(vii) 1 20.2201(d) 0 20.2203(a)(3)(ii) 0 50.73(a)(2)(ii)(A)

E0 50.73(a)(2)(viii)(A) o 20.2203(a)(1) 0 20.2203(a)(4) 0 50.73(a)(2)(ii)(B) 0 50.73(a)(2)(viii)(B) o 20.2203(a)(2)(1) 0 50.36(c)(1)(i)(A)

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

10. POWER LEVEL 0

20.2203(a)(2)(ii) 0l 50.36(c)(1)(ii)(A)

[0 50.73(a)(2)(iv)(A) 0 50.73(a)(2)(x)

EO 20.2203(a)(2)(iii) 0 50.36(c)(2) 0 50.73(a)(2)(v)(A) 0 73.71 (a)(4) ol 20.2203(a)(2)(iv) 0 50.46(a)(3)(ii)

El 50.73(a)(2)(v)(B) 0 73.71 (a)(5) 100 El 20.2203(a)(2)(v)

E0 50.73(a)(2)(i)(A) 0l 50.73(a)(2)(v)(C) 0 OTHER El 20.2203(a)(2)(vi) 10 50.73(a)(2)(i)(B)

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

Specify in Abstract below or in (If more space is required, use additional copies of (If more space is required, use additional copies of NRC Form 366A) (17)

V. ASSESSMENT OF SAFETY CONSEQUENCES

The safety consequences of this event were not significant. BFN TS LCO 3.8.7 requires that the RMOV board remain operable with the reactor in Modes 1, 2 and 3. TS LCO 3.0.4 prohibits ascending reactor mode changes unless the TS Systems required for the mode are operable. The objective of this TS provision is to ensure that TS Systems required for plant startup and operation are in service to support safe plant operation. The Auto/Manual Transfer switch must be maintained in the Auto position in order to maintain the RMOV board fully operable. With the Auto/Manual Transfer Switch in the manual position loss of the feed from 480 V Shutdown Board 2A results in a failure to automatically transfer to the alternate feed, 480 V Shutdown Board 2B.

The 480 V RMOV Board 2D has only three (3) loads, all Division 1. These are the normally open Reactor Recirculation System Pump 2B [AD] discharge valve, the normally closed Loop 1, RHR inboard low pressure core injection (LPCI) valve, and the normally open Loop 1 RHR pumps A and C minimum flow bypass valve. The failure of 480 V RMOV Board 2D to transfer on loss of the upstream feed, 480 V Shutdown Board 2A which is provided power from 4.16 kV Shutdown Board B, during a postulated Design Basis Accident - Loss of Coolant Accident (DBA-LOCA) would result in a failure of the valves to actuate including, failure of the LPCI injection to open. The postulated board failure results in a Loss of Loop 1 RHR injection path which is bounded the BFN single failure analysis described in Updated Safety Analysis Report Section 6.5 and Table 6.5-3.

The analysis concludes that peak clad temperature during a DBA-LOCA is maintained below 2200 degrees F. Therefore, BFN concludes that the health and safety of the public was not affected by this event.

VI. CORRECTIVE ACTIONS

A.

Immediate Corrective Actions

The Units 2 and 3 Reactor Building Auxiliary Unit Operator Rounds Instruction has been revised to verify the Auto/Manual Transfer switches on both the 480 V RMOV Board 2D and 2E on both Units 2 and 3 are in the Auto position. These two 480 V RMOV boards do not have a protective cover on the Auto/Manual Transfer switch. The verification is performed once per shift.

B.

Corrective Actions to Prevent Recurrence - The corrective actions are being managed by BFN's corrective action program.

For Units 1 and 2, the BFN operating instructions governing plant startup are being revised to require that the Emergency Control Switch Verification be performed prior to entry into Mode 3 or 2.

For Unit 3 which is restarting from a short forced outage the BFN operating instruction governing plant startup has been revised to require that the Emergency Control Switch Verification be performed prior to entry into Mode 3 or 2.

VII. ADDITIONAL INFORMATION

A.

Failed Components None.

B.

PREVIOUS LERS ON SIMILAR EVENTS None.U.S. NUCLEAR REGULATORY COMMISSION (9-2007)

LICENSEE EVENT REPORT (LER)

FACILITY NAME (1)

DOCKET (2)

LER NUMBER (6)

PAGE (3)

YEAR SEQUENTIAL REVISION NUMBER NUMBER Browns Ferry Nuclear Plant Unit 2 05000260 2009

-- 005

-- 00 5 of 5 NA*RA*I ivE (r more space is required, use additional copies of NRC Form 366A) (17)

C.

Additional Information

Corrective action document for this report is Problem Evaluation Report 176648.

D.

Safety System Functional Failure Consideration:

This event is not a safety system functional failure according to NEI 99-02.

E.

Scram With Comolications Consideration:

This event was not a complicated scram according to NEI 99-02.

VIII. COMMITMENTS

None.