05000259/LER-2007-003, Regarding Main Steam Line High Flow Instrument Inoperable in Excess of Technical Specification Allowable Outage Time

Regarding Main Steam Line High Flow Instrument Inoperable in Excess of Technical Specification Allowable Outage Time

ML072180157
Person / Time
Site:	Browns Ferry
Issue date:	08/03/2007
From:	O'Grady B Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LER 07-003-00
Download: ML072180157 (8)

LER-2007-003, Regarding Main Steam Line High Flow Instrument Inoperable in Excess of Technical Specification Allowable Outage Time

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)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(iv)(A), System Actuation 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)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(x)
2592007003R00 - NRC Website

text

August 3, 2007 U.S. Nuclear Regulatory Commission 10 CFR 50.73 ATTN: Document Control Desk Mail Stop OWFN, P1-35 Washington, D. C. 20555-0001

Dear Sir:

TENNESSEE VALLEY AUTHORITY - BROWNS FERRY NUCLEAR PLANT (BFN)

- UNIT 1 - DOCKET 50-259 - FACILITY OPERATING LICENSE DPR LICENSEE EVENT REPORT (LER) 50-259/2007-003-00 The enclosed report provides details of exceeding the Technical Specification allowable outage time due to an inoperable Main Steam Line Flow-High pressure instrument.

As such, in accordance with 10 CFR 50.73(a)(2)(i)(B), TVA is reporting this as any operation or condition prohibited by the units TS. There are no commitments contained in this letter.

Sincerely, Original signed by R. G. Jones for:

Brian OGrady cc: See page 2

U.S. Nuclear Regulatory Commission Page 2 August 3, 2007 Enclosure cc (Enclosure):

Ms. Eva A. Brown, Project Manager U.S. Nuclear Regulatory Commission (MS 08G9)

One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852-2739 Mr. James T. Moorman, III, Branch Chief U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, Georgia 30303-8931 NRC Resident Inspector Browns Ferry Nuclear Plant 10833 Shaw Road Athens, Alabama 35611-6970

U.S. Nuclear Regulatory Commission Page 3 August 3, 2007 DTL:DAH:BAB Enclosure cc (Enclosure):

A. S. Bhatnagar, LP 6A-C D. C. Matherly, BFT 2A-BFN R. H. Bryan, Jr., LP 4J-C W. R. Campbell, Jr. LP 6A-C J. C. Fornicola, LP 6A-C R. G. Jones, POB 2C-BFN G. V. Little, NAB 1D-BFN R. F. Marks, Jr., PAB 1C-BFN B. A. Wetzel, BR 4X-C S. A. Vance, WT 6A-K E. J. Vigluicci, WT 6A-K NSRB Support, LP 5M-C INPO:LEREvents@inpo.org EDMS WT CA - K S:lic/submit/lers/259 2007 003.doc

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION (6-2004)

LICENSEE EVENT REPORT (LER)

(See reverse for required number of digits/characters for each block)

APPROVED BY OMB NO. 3150-0104 EXPIRES 06/30/2007

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

1. FACILITY NAME Browns Ferry Unit 1

2. DOCKET NUMBER 05000259

3. PAGE 1 OF 5

4. TITLE Main Steam Line High Flow Instrument Inoperable in Excess of Technical Specification Allowable Outage Time

5. EVENT DATE

6. LER NUMBER

7. REPORT DATE

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

MONTH DAY YEAR FACILITY NAME None DOCKET NUMBER N/A 06 04 2007 2007 - 003 - 00 08 03 2007 FACILITY NAME None DOCKET NUMBER N/A

9. OPERATING MODE

11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) 1

10. POWER LEVEL 055 20.2201(b) 20.2201(d) 20.2203(a)(1) 20.2203(a)(2)(i) 20.2203(a)(2)(ii) 20.2203(a)(2)(iii) 20.2203(a)(2)(iv) 20.2203(a)(2)(v) 20.2203(a)(2)(vi) 20.2203(a)(3)(i) 20.2203(a)(3)(ii) 20.2203(a)(4) 50.36(c)(1)(i)(A) 50.36(c)(1)(ii)(A) 50.36(c)(2) 50.46(a)(3)(ii) 50.73(a)(2)(i)(A) 50.73(a)(2)(i)(B) 50.73(a)(2)(i)(C) 50.73(a)(2)(ii)(A) 50.73(a)(2)(ii)(B) 50.73(a)(2)(iii) 50.73(a)(2)(iv)(A) 50.73(a)(2)(v)(A) 50.73(a)(2)(v)(B) 50.73(a)(2)(v)(C) 50.73(a)(2)(v)(D) 50.73(a)(2)(vii) 50.73(a)(2)(viii)(A) 50.73(a)(2)(viii)(B) 50.73(a)(2)(ix)(A) 50.73(a)(2)(x) 73.71(a)(4) 73.71(a)(5)

OTHER Specify in Abstract below or in Other actions were taken to verify that all Analog Trip Unit (ATU) channels were indicating as expected for current plant conditions. The Integrated Computer System was also reviewed for abnormalities. Instrument Inspection Checklists have been re-performed (excepting instruments that were inaccessible) for systems 01 (Main Steam) [SB], 63 (Standby Liquid Control) [BR], 64 (Primary Containment) [JM], 71 (Reactor Core Isolation Cooling) [BN], 73 (High Pressure Coolant Injection)

[BJ], 74 (Residual Heat Removal System) [BO], and 75 (Core Spray) [BM]. No additional mispositioned valves were identified.

VII.

ADDITIONAL INFORMATION

A. Failed or Degraded Components None.

B. Previous LERs on Similar Events None.

C. Additional Information

Browns Ferry Corrective Action document PER 125786.

D. Safety System Functional Failure Consideration:

During this event, the remaining MSL flow instruments on MSL 1D remained in operation and would have provided the isolation function in the event of a MSL break. Therefore, this event is not considered a safety system function failure in accordance with NEI 99-02.

E. Loss of Normal Heat Removal Consideration:

The condition being reported did not involve a reactor scram.

VIII. COMMITMENTS

None.