05000260/LER-2007-002, Re Main Steam Relief Valve as Found Setpoint Exceeded Technical Specifications Lift Pressure

Re Main Steam Relief Valve as Found Setpoint Exceeded Technical Specifications Lift Pressure

ML071980146
Person / Time
Site:	Browns Ferry
Issue date:	07/16/2007
From:	Rosalyn Jones Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LER 07-002-00
Download: ML071980146 (9)

LER-2007-002, Re Main Steam Relief Valve as Found Setpoint Exceeded Technical Specifications Lift Pressure

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
2602007002R00 - NRC Website

text

July 16, 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 2 - DOCKET 50-260 - FACILITY OPERATING LICENSE DPR LICENSEE EVENT REPORT (LER) 50-260/2007-02-00 The enclosed report provides details of a failure to meet the requirements of the Technical Specifications (TS) Limiting Condition for Operation (LCO) 3.4.3 concerning main steam relief valve (MSRV) operability.

In accordance with 10 CFR 50.73(a)(2)(i)(B), TVA is reporting this as any operation or condition prohibited by the plant's 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 July 16, 2007 Enclosure cc (Enclosure):

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 Senior Resident Inspector Browns Ferry Nuclear Plant 10833 Shaw Road Athens, AL 35611-6970 Ms. Eva A. Brown, Project Manager U.S. Nuclear Regulatory Commission (MS 08G9)

One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852-2739

U.S. Nuclear Regulatory Commission Page 3 July 16, 2007 DTL:SWA:BAB Enclosure cc (Enclosure):

A. S. Bhatnagar, LP 6A-C R. H. Bryan, Jr., LP 4J-C W. R. Campbell, Jr., LP 6A-C D. C. Matherly, BFT 2A-B.

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 E. J. Vigluicci, ET 11A-K NSRB Support, LP 5M-C INPO:LEREvents@inpo.org EDMS WT CA - K S:lic/submit/lers/260-2007-02.doc

NRC FORM 366 (7-2001)

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 2

2. DOCKET NUMBER 05000260

3. PAGE 1 of 6

4. TITLE Main Steam Relief Valve As Found Setpoint Exceeded Technical Specifications Lift Pressure

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 05 16 2007 2007-002-00 07 16 2007 FACILITY NAME none DOCKET NUMBER N/A

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

9. OPERATING MODE 1

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

OTHER

10. POWER LEVEL 100 20.2203(a)(2)(vi)

X 50.73(a)(2)(i)(B) 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 (If more space is required, use additional copies of (If more space is required, use additional copies of NRC Form 366A) (17) VII.

ADDITIONAL INFORMATION

A.

Failed or Degraded Components Target Rock MSRV model No. 7567F B.

Previous LERs on Similar Events Numerous previous reports on similar events have been made from BFN and other nuclear plants. The physical phenomenon affecting MSRV lift setpoints which results in this reportable condition is well-understood, and it has been the subject of much industry study. Different mitigative approaches have been tested, but none have successfully eliminated the issue. The installation of the instrumentation logic/circuitry which will automatically open the MSRVs as appropriate during pressurization transients largely negates the conditions safety impact.

Though this phenomenon has only a relatively small impact on the MSRV function and because of the compensatory hardware mitigation which has been installed at BFN, BFN is continuing to work with other industry stakeholders toward the total elimination of this issue.

C.

Additional Information

Browns Ferry corrective action document PER 12944.

D.

Safety System Functional Failure Consideration:

The condition being reported involves only setpoint drift of varying numbers of MSRVs on Unit 2. The safety/relief function provided by these valves was not compromised at any time. A safety system functional failure did not result from this condition.

E.

Loss of Normal Heat Removal Consideration:

The condition being reported did not involve a reactor scram; therefore, did not involve a loss of normal heat removal consideration.

VIII. COMMITMENTS

None.