05000286/LER-2013-004, Regarding Technical Specification (TS) Prohibited Condition Due to Leak in the Reactor Coolant Pressure Boundary Caused by Defects in a Weld on Seal Table In-Core Detector Drive E-11
| ML13141A275 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 05/13/2013 |
| From: | Ventosa J Entergy Nuclear Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-13-074 LER 13-004-00 | |
| Download: ML13141A275 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(1), Submit an LER, Invalid Actuation 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 2862013004R00 - NRC Website | |
text
.- En tergy Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, N.Y. 10511-0249 Tel (914) 254-6700 John A. Ventosa Site Vice President NL-1 3-074 May 13, 2013 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Stop O-PI -17 Washington, D.C. 20555-0001
SUBJECT:
Licensee Event Report # 2013-004-00, "Technical Specification (TS)
Prohibited Condition Due to Leak in the Reactor Coolant Pressure Boundary Caused by Defects in a Weld on Seal Table I n-Core Detector Drive E-1 1" Indian Point Unit No. 3 Docket No. 50-286 DPR-64
Dear Sir or Madam:
Pursuant to 10 CFR 50.73(a)(1), Entergy Nuclear Operations Inc. (ENO) hereby provides Licensee Event Report (LER) 2013-004-00. The attached LER identifies an event where there was a Technical Specification (TS) prohibited condition due to a leak in the Reactor Coolant Pressure Boundary caused by defects in a weld on seal table In-Core detector drive E-1 1 which is not permitted by the TS, therefore is reportable under 10 CIFR 50.73(a)(2)(i)(B). This condition was recorded in the Entergy Corrective Action Program as Condition Report CR-1P3-2013-01556.
There are no new commitments identified in this letter. Should you have any questions regarding this submittal, please contact Mr. Robert Walpole, Manager, Licensing at (914) 254-6710.
Sincerely, JAV/cbr cc:
Mr. William Dean, Regional Administrator, NRC Region I NRC Resident Inspector's Office, Indian Point 2 Ms. Bridget Frymire, New York State Public Service Commission LEREvents@inpo.org
Abstract
On March 14, 2013, during a scheduled refueling outage (RO) boric acid program walk down inspection, boron residue was identified on the fillet weld which attaches the E-11 in-core guide tube to the seal table.
Since there was no visible leak path indication that the leak had initiated at the high pressure mechanical seal connection located just above the fillet weld, the boron residue was cleaned and a surface examination [i.e., Liquid Penetrant Test (PT)] was performed on the fillet weld.
Although this surface examination did not identify any rejectable indications at the leak location, it did identify rounded indications in the weld which could have been the cause of the leakage.
No linear indications (i.e., cracks) were identified on the weld or on the guide tube base material above the toe of the weld.
Because PT results showed rounded indications in the weld and boron residue was present, the condition was judged to represent a potential through wall defect therefore a Reactor Coolant leak path.
The seal table In-core detector guide tube is part of the reactor coolant pressure boundary (RCPB).
Technical Specification (TS) 3.4.13 does not allow any RCPB leakage.
The apparent cause of the defect was OD initiated stress corrosion cracking of the stainless steel guide tube base material under the fillet weld.
It is postulated that the rounded indications on the weld metal allowed contaminants from previous mechanical joint leaks to contact the guide tube base material under the weld resulting in SSC of the tube base material.
Corrective actions included VT-2 visual examination of the remaining seal table penetrations to verify that no additional through wall leaks were present and the leaking guide tube was removed from service by cutting the tube below the leaking area and installing a welded plug to form a new RCPB.
The event had no significant effect on public health and safety.
(If more space is required, use additional copies of (If more space is required, use additional copies of NRC Form 366A) (17)
Safety Significance
This event had no significant effect on the health and safety of the public. There were no actual safety consequences from the event because there were no significant failures in the RCPB.
Periodic inspections identify leaks when they are small so that repairs can be performed to prevent RCPB degradation.
TS 3.4.13 has Surveillance Requirement 3.4.13.1 to verify RCS leakage is within limits by performance of RCS inventory balance every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
This surveillance ensures the integrity of the RCPB is maintained and provides a trend of leakage early before significant degradation.
An early warning of RCPB leakage or unidentified leakage is provided by the systems that monitor containment atmosphere radioactivity and operation of the containment sump.
Failure of a RCPB would be a Loss of Coolant Accident (LOCA).
A LOCA is analyzed in UFSAR Section 14.3.
A minor pipe break (small break) is defined as a rupture of the RCPB with a total cross-sectional area less than 1.0 square foot in which the normally operating charging system flow is not sufficient to sustain pressurizer level and pressure.
The results of analysis in UFSAR Section 14.3.3.4 concluded the limiting break was a 3 inch cold leg break.
The results of the analysis demonstrated that for a small break LOCA, the Emergency Core Cooling System will meet the acceptance criteria contained in 10CFR50.46.
The LOCA analysis of Section 14.3 are bounding for the components reported in this LER.