05000259/LER-2008-002, ASME Code Class 1 Pressure Boundary Leak on an Instrument Line Connected to the Reactor Vessel
| ML090270197 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 01/22/2009 |
| From: | West R Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LER 08-002-00 | |
| Download: ML090270197 (6) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 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 |
| 2592008002R00 - NRC Website | |
text
Tennessee Valley Authority, Post Office Box 2000, Decatur, Alabama 35609-2000 January 22, 2009 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop: OWFN, P1-35 Washington, D. C. 20555-0001 10 CFR 50.73
Dear Sir:
TENNESSEE VALLEY AUTHORITY - BROWNS FERRY NUCLEAR PLANT (BFN) -
UNIT 1 - DOCKET 50-259 - FACILITY OPERATING LICENSE DPR LICENSEE EVENT REPORT (LER) 50-259/2008-002 The enclosed report provides details of an ASME Code Class 1 Boundary Leak on an Instrument Line Connected to the Reactor Vessel. TVA is submitting this report in accordance with 10 CFR 50.73(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 commitments contained in this letter.
GWest Site Vice President, BFN cc: See page 2
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U.S. Nuclear Regulatory Commission Page 2 January 22, 2009 Enclosure cc (Enclosure):
Ms. Eva Brown, Project Manager U.S. Nuclear Regulatory Commission (MS 08G9)
One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852-2739 Rebecca L. Nease, 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 08/31/2010 (9-2007)
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information digits/characters for each block) collection.
- 3. PAGE Browns Ferry Unit 1 05000259 1 of 4
- 4. TITLE: ASME Code Class 1 Pressure Boundary Leak On An Instrument Line Connected to the Reactor Vessel
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED SEOUENTIAL REV FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR NUMBER NO.
None N/A FACILITY NAME DOCKET NUMBER 11 23 2008 2008 002 00 01 22 2009 None N/A
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)
El 20.2201(b)
[] 20.2203(a)(3)(i)
El 50.73(a)(2)(i)(C)
El 50.73(a)(2)(vii)
El 20.2201(d)
El 20.2203(a)(3)(ii)
Z 50.73(a)(2)(ii)(A)
El 50.73(a)(2)(viii)(A)
E] 20.2203(a)(1)
El 20.2203(a)(4)
El 50.73(a)(2)(ii)(B)
[I 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)
[I 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL El 20.2203(a)(2)(ii)
El 50.36(c)(1)(ii)(A)
[I 50.73(a)(2)(iv)(A)
El 50.73(a)(2)(x)
El 20.2203(a)(2)(iii)
[I 50.36(c)(2)
El 50.73(a)(2)(v)(A)
[E 73.71(a)(4)
El 20.2203(a)(2)(iv)
[I 50.46(a)(3)(ii)
[I 50.73(a)(2)(v)(B)
El 73.71 (a)(5) 000 El 20.2203(a)(2)(v)
El 50.73(a)(2)(i)(A)
El 50.73(a)(2)(v)(C)
El OTHER El 20.2203(a)(2)(vi)
El 50.73(a)(2)(i)(B)
El 50.73(a)(2)(v)(D)
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- 12. LICENSEE CONTACT FOR THIS LER NAME TELEPHONE NUMBER (Include Area Code)
Steve Austin, Licensing Engineer 256-729-2070CMOET MANU-REPORTABLE MANU-REPORTABLE
CAUSE
SYSTEM COMPONENT A
CAUSE
SYSTEM COMPONENT FACTURER TO EPIX FATRRI OEI
- 14. SUPPLEMENTAL REPORT EXPECTED
- 15. EXPECTED MONTH DAY YEAR SUBMISSION El YES (If yes, complete 15. EXPECTED SUBMISSION DATE)
Z NO DATE N/A N/A N/A ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)
On November 23, 2008, at approximately 1200 hours0.0139 days <br />0.333 hours <br />0.00198 weeks <br />4.566e-4 months <br /> Central Standard Time (CST), during the performance of the Unit 1 vessel hydrostatic test, ASME Section XI System Leakage Test of the Reactor Pressure Vessel and Associated Piping, 1-SI-3.3.1.A, a reactor pressure boundary leak was discovered on an unisolatable instrument line connected to the reactor vessel. This instrument line is an ASME Code Class 1 equivalent component, 2-inch pipe, near pressure vessel nozzle N11B. BFN entered Technical Requirements Manual (TRM) Section 3.4.3 - Structural Integrity, which requires the integrity of ASME Code Class 1, 2, and 3 equivalent components shall be maintained through the life of the plant. Unit 1 was in mode 4 at the time of discovery; it remained in mode 4 until the repairs were completed. The root cause of the event was residual stress introduced to the safe end inside diameter during initial fabrication. The N11B safe end was examined ultrasonically (UT). The through wall leak was repaired by weld overlay.
NRC FORM 366 (9-2007)
(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)
The visual inspection during the performance of 1-SI-3.3.1.A identified the through wall leak at the instrument line connection to the reactor vessel. BFN entered Technical Requirements Manual (TRM) 3.4.3, Structural Integrity, Condition A, which requires immediate restoration of the structural integrity of the affected component or maintain the reactor in mode 4 of the reactor coolant system less than 50 degrees F above the minimum temperature requires for nondestructive testing considerations, until each indication has been investigated and evaluated. Until repairs were completed, BFN maintained the reactor in accordance with these requirements. Therefore, the event did not adversely affect the safety of the public and plant personnel.
VI. CORRECTIVE ACTIONS
A.
Immediate Corrective Actions
Once TVA determined there was an un-isolatable leak in the ASME Class 1 reactor pressure boundary, the reactor was depressurized to the pre-test pressure and mode 4 was maintained.
B.
Corrective Actions to Prevent Recurrence (1)
To determine the extent of the through wall crack, the N1 1B safe end was examined ultrasonically (UT). The through wall leak was repaired by weld overlay.
BFN ultrasonically examined the remaining Unit 1 small-bore (less than 4 inches in diameter) instrument nozzle safe ends and the core delta pressure - Standby Liquid Control [BR] line nozzle safe end. The examinations did not identify any other recordable indications.
VII. ADDITIONAL INFORMATION
A.
Failed Components None.
B.
Previous LERs on Similar Events None.
C.
Additional Information
Corrective action document for this report is PER 157918.
D.
Safety System Functional Failure Consideration:
This event is not a safety system functional failure according to NEI 99-02.
E.
Scram With Complications Consideration:
This event was not a complicated scram according to NEI 99-02.
VIII. COMMITMENTS
None.
TVA does not consider the corrective action a regulatory requirement. The completion of the action will be tracked in TV(s Corrective Action Program.