05000339/LER-2002-001, For North Anna Unit 1, Reactor Vessel Head Leakage Due to Hot Short Cracking and Primary Water Stress Corrosion Cracking
| ML023180480 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 11/06/2002 |
| From: | Heacock D Virginia Electric & Power Co (VEPCO) |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 02-620 LER 02-001-00 | |
| Download: ML023180480 (12) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2) 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 |
| 3392002001R00 - NRC Website | |
text
1 OCFR50.73 Virginia Electric and Power Company North Anna Power Station P. 0. Box 402 Mineral, Virginia 23117 November 6, 2002 U. S. Nuclear Regulatory Commission Serial No.:
02-620 Attention: Document Control Desk NAPS:
JHL Washington, D. C. 20555-0001 Docket No.: 50-339 License No.: NPF-7
Dear Sirs:
Pursuant to 10CFR50.73, Virginia Electric and Power Company hereby submits the following Licensee Event Report applicable to North Anna Power Station Unit 2.
Report No. 50-339/2002-001-00 This report has been reviewed by the Station Nuclear Safety and Operating Committee and will be forwarded to the Management Safety Review Committee for its review.
Very truly yours, D. A. Heacock, Site Vice President North Anna Power Station Enclosure Commitments contained in this letter: None cc: United States Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23 T85 Atlanta, Georgia 30303-8931 Mr. M. J. Morgan NRC Senior Resident Inspector f"'
North Anna Power Station
NRC FORM 366 U S NUCLEAR REGULATORY APPROVED BY OMB NO. 3150-0104 EXPIRES 7-31-2004 7201 CoMMISSION CtC!
the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection FACILITY NAME (l)
DOCKET NUMBER (2)
PAGE (3)
NORTH ANNA POWER STATION, UNIT 2 05000 - 339 1 1OF11 TITLE (4)
Reactor Vessel Head Leakage due to Hot Short Cracking and Primary Water Stress Corrosion Cracking EVENT DATE (5)
LER NUMBER (6)
REPORT DAT (
I OTHER FACILITIES INVOLVED (8)
MONTH DAY YEAR YEAR SEQUENTIAL REVISION MONTH DAY YEAR FACILITY NAME DOCUMENTNUMBER
_NUMBER NUMBER l l
l ll05000-09 14 2002 2002.
-- 001 --
00 11 06 2002 FACILITY NAME DOCUMENT NUMBER 05000-OPERATING 1
THIS REPORT IS SUBMITTED PURSUANTTO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) (11)
MODE (9) l6l20 2201 (b) l l20 2203(a)(3)(ii) 50.73(a)(2)(n)(B) l50 73(a)(2)(ix)(A)
POWER l
20.2201(d) 20 2203(a)(4) l 50.73(a)(2)(m) 50 73(a)(2)(x)
LEVEL[(I) 020.2203(a)(1) l 50 36(c)(1)(i)(A) l 50.73(a)(2)(v)(A) l 73 71(a)(4) 20.2203(a)(2)(i) l 50.36(c)(1)(il)(A) 50.73(a)(2)(v)(A) 7371 (a)(5)
I l 20.2203(a)(2)(ii) 50.36(c)(2) l 50.73(a)(2)(v)(B)
OTHER 20.2203(a)(2)(^)
50 46(a)(3)(ii) l 50.73(a)(2)(v)(C)
Specify In Abstract below or l 20.2203(a)(2)(iv) 50.73(a)(2)(l)(A) 50.73(a)(2)(v)(D)
In (If more space is required, use additonal copies of (If more space is required, use additional copies of NRC Forn 366A) (17)
Eddy Current J-Groove Weld J-Groove Weld Coverage Eddy Current Penetration Tube OD Penetration #
1 2
3 4
5 6
7 8
9 10 11 12 13 15 17 19 21 22 23 24 25 26 27 28 29 30 31 32 33 34 35 36 37 38 39 40 41 42 43 44 Reportable (RI)
Reportable (RI)
Reportable (RI)
Reportable (RI)
Reportable (RI)
Reportable (RI)
Recordable (NRI)
Reportable (RI)
Reportable (RI)
Recordable (NRI)
Recordable (NRI)
Recordable (NRI)
Reportable (RI)
Reportable (RI)
No Detectable Indications (NDI)
Reportable (RI)
Reportable (RI)
Reportable (RI)
Recordable (NRI)
No Detectable Indications (NDI)
Reportable (RI)
Recordable (NRI)
Reportable (RI)
Recordable (NRI)
Reportable (RI)
Recordable (NRI)
Reportable (RI)
Recordable (NRI)
Recordable (NRI)
Recordable (NRI)
Reportable (RI)
Reportable (RI)
Recordable (NRI)
Reportable (RI)
Reportable (RI)
Reportable (RI)
Reportable (RI)
Reportable (RI)
Recordable (NRI)
Rerortable (RI) 3600 3600 3600 3600 3600 3600 3600 3600 3600 3600 3600 3600 3600 3600 3600 3600 3600 3600 3600 3600 3600 3600 3600 3600 3600 3600 3600 3600 3600 3600 3600 3600 3600 3600 3600 3600 3600 3600 3600 3600 No Detectable Indications (NDI)
No Detectable Indications (NDI)
No Detectable Indications (NDI)
No Detectable Indications (NDI)
No Detectable Indications (NDI)
No Detectable Indications (NDI)
No Detectable Indications (NDI)
Recordable (NRI)
No Detectable Indications (NDI)
Not Inspected Recordable (NRI)
No Detectable Indications (NDI)
No Detectable Indications (NDI)
No Detectable Indications (NDI)
No Detectable Indications (NDI)
No Detectable Indications (NDI)
No Detectable Indications (NDI)
No Detectable Indications (NDI)
No Detectable Indications (NDI)
Not Inspected No Detectable Indications (NDI)
No Detectable Indications (NDI)
No Detectable Indications (NDI)
No Detectable Indications (NDI)
No Detectable Indications (NDI)
No Detectable Indications (NDI)
No Detectable Indications (NDI)
No Detectable Indications (NDI)
Not Inspected No Detectable Indications (NDI)
Not Inspected No Detectable Indications (NDI)
No Detectable Indications (NDI)
No Detectable Indications (NDI)
No Detectable Indications (NDI)
No Detectable Indications (NDI)
Not Inspected Not Inspected No Detectable Indications (NDI)
No Detectable Indications (NDI)
(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 (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) determine the additional service life allowable before repair. The results indicate that 1.5 years of service could be achieved prior to the indication growing to 75% through-wall, 1.8 years before becoming completely through-wall, and 26.0 years remain before the flaw would become unstable.
A boat sample from the previous weld overlay repair of penetration 51 was removed and sent to Westinghouse for laboratory analysis. Analysis of the boat sample indicated that the weld overlay repair did not extend out far enough to cover previous NDE indications.
Additional information related to the structural integrity of the reactor pressure vessel head penetration nozzles, including the extent of the leakage and indications, and the inspections and repairs undertaken to satisfy regulatory requirements was provided in the written response to NRC Bulletin 2002-02.
A root cause evaluation of the event is being performed.
6.0 ACTIONS TO PREVENT RECURRENCE A management decision was made to replace the North Anna Unit 2 reactor head constructed of materials that are known to be more resistant to cracking rather than perform multiple penetration repairs.
7.0 SIMILAR EVENTS
LER 50-339/2001-003-00 documents that a qualified, visual barehead inspection of the North Anna 2 reactor pressure vessel head penetrations identified three penetrations as being rejectable due to the existence of boric acid on the reactor pressure vessel barehead surface.
8.0 MANUFACTURER/MODEL NUMBER Rotterdam Dockyard Company/Serial Number 30662
9.0 ADDITIONAL INFORMATION
None