05000260/LER-2009-002, Regarding Leak in an ASME Class I Code Reactor Pressure Boundary Pipe

Regarding Leak in an ASME Class I Code Reactor Pressure Boundary Pipe

ML092150033
Person / Time
Site:	Browns Ferry
Issue date:	07/30/2009
From:	West R Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LER 09-002-00
Download: ML092150033 (6)

LER-2009-002, Regarding Leak in an ASME Class I Code Reactor Pressure Boundary Pipe

Event date:
Report date:
Reporting criterion:	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 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
2602009002R00 - NRC Website

text

Tennessee Valley Authority, Post-Office Box 2000, Decatur, AJabama 35609-2000 July 30, 2009 10 CFR 50.73 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop: OWFN, P1-35 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) 2009-002 Leak In An ASME Code Class I Reactor Pressure Boundary Pipe The enclosed Licensee Event Report is being reported in accordance with 10 CFR 50.73, "Licensee event report system," paragraph (a)(2)(ii)(A), as an event or condition that resulted in the nuclear power plant, including principal safety barriers, being seriously degraded.

There are no 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.

pespec

tfully, R. G. West Site Vice President, BFN cc: See page 2

U.S. Nuclear Regulatory Commission Page 2 July 30, 2009 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. Eugene F. Guthrie, 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

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 EXPIRES 08131/2010 (9-2007)

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

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

3. PAGE Browns Ferry Unit 2 05000260 1 of 4

4. TITLE: Leak In An ASME Class I Code Reactor Pressure Boundary Pipe

5. EVENT DATE

6. LER NUMBER

7. REPORT DATE

8. OTHER FACILITIES INVOLVED MONTH DAY YEAR YEAR I SEQUENTIAL I REV MONTH DAY YEAR FACILITY NAME DOCKET NUMBER XINUMBER NOD.

None N/A 05 31 2009 2009-002-00 07 30 2009 FACILITY NAME DOCKET NUMBER None N/A

9. OPERATING MODE

11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR 1:(Check all that apply) 4 0 20.2201(b)

El 20.2203(a)(3)(i)

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

El 50.73(a)(2)(vii)

El 20.2201(d)

El 20.2203(a)(3)(ii) 0 50.73(a)(2)(ii)(A)

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

El 20.2203(a)(1)

El 20.2203(a)(4)

C1 50.73(a)(2)(ii)(B)

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

El 20.2203(a)(2)(i)

El 50.36(c)(1)(i)(A)

El 50.73(a)(2)(iii)

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

10. POWER LEVEL

[

20.2203(a)(2)(ii)

[] 50.36(c)(1)(ii)(A)

El 50.73(a)(2)(iv)(A)

El 50.73(a)(2)(x) 0 20.2203(a)(2)(iii)

El 50.36(c)(2)

El 50.73(a)(2)(v)(A)

El 73.71(a)(4)

El 20.2203(a)(2)(iv)

El 50.46(a)(3)(ii)

[E 50.73(a)(2)(v)(B) 0l 73.71 (a)(5)

El 20.2203(a)(2)(v)

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

El 50.73(a)(2)(v)(C) 0l OTHER El 20.2203(a)(2)(vi)

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

E5 50.73(a)(2)(v)(D) 366ASpecify in Abstract below or in NRC Form

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

Steve Austin, Licensing Engineer, BFN Licensingl 256-729-2070CAUSE ISYSTEM ICOMPONENT MANU-REPORTABLE I-

CAUSE

SYSTEM COMPONENT MANU-REPORTABLE FACTURER TO EPIX FACTURER TO EPIX F]

14. SUPPLEMENTAL REPORT EXPECTED ri

15. EXPECTED MONTH DAY YEAR YES (if yes, complete 15. EXPECTED SUBMISSION DATE)

NO SUBMISSION I

DATE N/A N/A IN/A ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced type written lines)

On May 31, 2009, at approximately 1210 hours0.014 days <br />0.336 hours <br />0.002 weeks <br />4.60405e-4 months <br /> Central Daylight Time (CDT), during the performance of the Unit 2, ASME Section Xl System Leakage Test of the Reactor Pressure Vessel and Associated Piping, 2-SI-3.3.1.A, TVA identified a leak that could not be isolated. Following the confirmation that the through-wall leak was associated with an ASME Class I pressure boundary the test was terminated.

Operations maintained the reactor Mode 4 in accordance with the Technical Requirements Manual Action Statement 3.4.3.A.1. A weld performed in February of 1992, on the valve body and the tapered plug contained a defect. The root pass was made using the gas-tungsten process. This process was the standard in 1992. TVA postulates that a small amount of moisture present during the weld operation caused a defect in the weld. The defect slowly propagated and failed.

NRC FORM 366 (6-2004)

(If more space is required, use addifional copies of (If more space is required, use additional copies of (If more space is required, use additional copies of NRC Form 366A) (17) 50 degrees F above the minimum temperature required by nondestructive testing considerations, until each indication has been investigated and evaluated. Until completion of repairs, BFN maintained the reactor in accordance with these requirements. Therefore, TVA concludes that this event did not affect the health and safety of the public.

VI. CORRECTIVE ACTIONS

A.

Immediate Corrective Actions

Once it was determined that the leak was part of the ASME Class I pressure boundary, Operations placed the reactor in Mode 4.

B.

Corrective Actions to Prevent Recurrence (1)

TVA repaired the weld. The completion of the weld was followed by successful completion of 2-SI-3.3.1.A.

VII. ADDITIONAL INFORMATION

A.

Failed Comoonents None.

B.

Previous LERs on Similar Events None.

C.

Additional Information

Corrective action document for this report is Problem Evaluation Report 172551.

D.

Safety System Functional Failure Consideration:

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

E.

Scram With Complications Consideration:

This event did not result in a complicated scram according to NEI 99-02.

VIII. COMMITMENTS

None.

TVA does not consider this corrective action a regulatory requirement. TVA will track the completion of this action in the Corrective Action Program.