05000296/LER-2010-001, For Browns Ferry, Unit 3 Regarding Safety Relief Valves As-Found Setpoints Exceeded Technical Specification Lift Pressure Values

For Browns Ferry, Unit 3 Regarding Safety Relief Valves As-Found Setpoints Exceeded Technical Specification Lift Pressure Values

ML101740433
Person / Time
Site:	Browns Ferry
Issue date:	06/21/2010
From:	Polson K Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LER 10-001-00
Download: ML101740433 (7)

LER-2010-001, For Browns Ferry, Unit 3 Regarding Safety Relief Valves As-Found Setpoints Exceeded Technical Specification Lift Pressure Values

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) 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
2962010001R00 - NRC Website

text

Tennessee Valley Authority, Post Office Box 2000, Decatur, Alabama 35609-2000 June 21, 2010 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 50-296/2010-001-00 The enclosed Licensee Event Report provides details of a failure to meet the requirements of Browns Ferry Nuclear Plant, Unit 3, Technical Specification 3.4.3 concerning safety relief valve operability. The Tennessee Valley Authority is submitting this report in accordance with 10 CFR 50.73(a)(2)(i)(B), as any 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 Dan Williamson, Acting Site Licensing and Industry Affairs Manager, at (256) 729-2636.

Respectfully, Vice President cc: See page 2

U.S. Nuclear Regulatory Commission Page 2 June 21, 2010 Enclosure cc (w/ Enclosure):

NRC 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, digits/characters for each block) and a person is not required to respond to, the information collection.

AGE Browns Ferry Nuclear Plant Unit 3 05000296 1 of 5

4. TITLE: Safety Relief Valves As-Found Setpoints Exceeded Technical Specification Lift Pressure Values

5. EVENT DATE

6. LER NUMBER

7. REPORT DATE

8. OTHER FACILITIES INVOLVED MOT 7 AY EAR YEAR SEQUENTIAL REV YEAR FACILITY NAME DOCKET NUMBER HTNUMBER NO.

MONTH DAY YEAR N/A N/A FACILITY NAME DOCKET NUMBER 04 20 2010 2010 001 00

06.

21 2010 N/A N/A

9. OPERATING MODE

11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) o 20.2201(b) 0 20.2203(a)(3)(i) 0 50.73(a)(2)(i)(C)

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

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

10. POWER LEVEL 0

20.2203(a)(2)(ii)

[0 50.36(c)(1)(ii)(A) 0 50.73(a)(2)(iv)(A) 0 50.73(a)(2)(x) o 20.2203(a)(2)(iii) 0 50.36(c)(2) 0 50.73(a)(2)(v)(A) 0 73.71 (a)(4) o 20.2203(a)(2)(iv)

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

[0 73.71 (a)(5) 100 0 20.2203(a)(2)(v) 0 50.73(a)(2)(i)(A) 0 50.73(a)(2)(v)(C) 0 OTHER

[] 20.2203(a)(2)(vi) 0 50.73(a)(2)(i)(B) 0 50.73(a)(2)(v)(D)

Specify in Abstract belo or in NRC For-366A

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

Mike Oliver, Licensing Engineer 256-729-7874CAUSE SYSTEM COMPONENT MANU-REPORTABLE

CAUSE

SY MANU-REPORTABLE FACTURER TO EPIX FACTURER TO EPIX B

SB RV T020 Y

14. SUPPLEMENTAL REPORT EXPECTED

16. EXPECTED MONTH DAY YEAR SUBMISSION 0I YES (If yes, complete 15. EXPECTED SUBMISSION DATE) 0 NO DATE N/A N/A N/A ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced type"wrtten lines)

On April 20, 2010, the Tennessee Valley Authority (TVA) determined that 8 of the 13 Safety Relief Valves (SRVs), removed from Unit 3 following Cycle 14 operation, mechanically actuated at pressures greater than 3 percent above their Technical Specification (TS) setpoints, and thus, were inoperable.

Unit 3 TS Limiting Condition for Operation (LCO) 3.4.3 requires twelve (12) SRVs to be operable in reactor Modes 1, 2, and 3. With one or more required SRVs inoperable, TS 3.4.3 Action A requires the unit to be placed in Mode 3 (hot shutdown) within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in Mode 4 (cold shutdown) within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Since 8 of the 13 SRVs were found during test to have actuated above their TS setpoints, it is concluded that Unit 3 operated with inoperable SRVs longer than allowed by TS 3.4.3 Action A.

The root cause of the SRV setpoint drift is corrosion-induced bonding between the pilot disc and seating surface.

TVA is submitting this report in accordance with 10 CFR 50.73(a)(2)(i)(B), as any operation or condition prohibited by the plant's Technical Specifications.

NRC FORM 366 (9-2007)

(If more space is required, use additional copies of (If more space is required, use additional copies of V (it 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 Comoonents None B.

PREVIOUS LERS ON SIMILAR EVENTS There have been previous similar events on Unit 1 (LER 259/2008-003-00) and Unit 2 (LER 260/2009-003-00).

C.

Additional Information

Corrective action document for this report is PER 226627.

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 did not include a reactor scram.

VIII. COMMITMENTS

None