05000374/LER-2015-003, Regarding Reactor Recirculation Loop Discharge Isolation Valve Vent Line Leak Due to Weld Defect
| ML15279A190 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  (NPF-018) |
| Issue date: | 10/06/2015 |
| From: | Vinyard H Exelon Generation Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| RA15-065 LER 15-003-00 | |
| Download: ML15279A190 (4) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 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)(B) 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)(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 |
| 3742015003R00 - NRC Website | |
text
LaSalle County Station 2601 North 21 Road Marseilles, Illinois 61341 Exeton Generation 10 CFR 50.73 RA1 5-065 October 6, 2015 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 LaSalle County Station, Unit 2 Facility Operating License No. NPF-18 NRC Docket No. 50-374
Subject:
Licensee Event Report 2015-003-00, Reactor Recirculation Loop Discharge Isolation Valve Vent Line Leak Due to Weld Defect In accordance with 10 CFR 50.73(a)f2)(ii)(A), Exelon Generation Company (EGC), LLC, is submitting Licensee Event Report Number 2015-003-00 for LaSalle County Station Unit 2.
There are no regulatory commitments in this letter. Should you have any questions concerning this report, please contact Mr. Guy V. Ford, Regulatory Assurance Manager, at (815) 415-2800.
Respectfully, Harold T. Vinyard Plant Manager LaSalle County Station
Enclosure:
Licensee Event Report cc:
Regional Administrator NRC Region Ill NRC Senior Resident Inspector LaSalle County Station
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB: NO. 3150-0104 EXPIRES: 01/3112017 (02-2014)
Eshmated burden per response to comply vIth this mandatory collection request: 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />.
I.
Reported lessons learned are incorporated into the licensing process and ted back to industry.
Send comments regarding burden estimate to the FOIA, Privacy and Intormalion Collections LICENSEE EVENT REPORT (LER)
Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by internet e-mail to Infocollects.Resource@nrc.gov, and to the Desk Otficer, Office of Information and (See Page 2 for required number of Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC
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20503. If a means used to impose an information coltection does not display a currently valid 0MB
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control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 3. PAGE LaSalle County Station, Unit 2 05000374 1
OF 3
- 4. TITLE Reactor Recirculation Loop Discharge Isolation Valve Vent Line Leak Due to Weld Defect
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR SEQUENTIAL E/
MONTH DAY YEAR N/A N/A FACILITY NAME DOCKET NUMBER 08 07 2015 2015 003
- - 00 10 06 2015 N/A N/A
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Checkall that apply)
LI 20.2201(b)
LI 20.2203(a)(3)ç)
LI 50.73(a)f2)(i)(C)
LI 50.73(a)(2)(vii)
LI 20.2201(d)
LI 20.2203(a)(3)Qi) 50.73(a)(2)(ii)(A)
LI 50.73(a)(2)(viii)(A)
LI 20.2203(a)(1)
LI 20.2203(a)(4)
LI 50.73(a)(2)(B)(B)
LI 50.73(a)(2)(viii)(B)
LI 20.2203(a)f2)(i)
LI 50.36(c)(1)(i)(A)
LI 50.73(a)(2)(Ni)
LI 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL LI 20.2203(a)f2)fii)
LI 50.36(c)(1)(ii)(A)
LI 50.73fa)(2)(iv)(A)
LI 50.73(a)(2)(x)
LI 20.2203(a)(2)(iU)
LI 50.36(c)(2)
LI 50.73(a)(2)(v)(A)
LI 73.71 (a)(4) 00 LI 20.2203(a)(2)tv)
LI 50.46(a)(3)çi)
LI 50.73(a)(2)(v)(B)
LI 73.71(a)(5) 0 LI 20.2203(a)(2)(v)
LI 50.73(a)(2)(i)(A)
LI 50.73(a)(2)(v)(C)
LI 50.73(a)(2)(i)(B)
LI 50.73(a)f2)(v)(D)
SpeciynrAbstractbeloworin
- 12. LICENSEE CONTACT FOR THIS LER LICENSEE CONTACT ITELEPHONE NUMBER (Include Area Code)
John KowaISki, Engineering Director 1815-415-3800MANU-I REPORTABLE MANU-I REPORTABLE RACTURER I
TOEPIX FACTURER TO EPIX [AUSE SYSTEM COMPONENT
CAUSE
SYSTEM COMPONENT
- 14. SUPPLEMENTAL REPORT EXPECTED
- 15. EXPECTED MONTH I
DAY I
YEAR SUBMISSION I
LI YES (If yes, complete 15. EXPECTED SUBMISSION DATE)
NO DATE I
I ABSTRACT (Limit to 7400 spaces, i.e., approximately 75 single-spaced typewritten lines)
On August 7, 2015, Unit 2 was in Mode 3 for a planned maintenance outage. At 1300 hours0.015 days <br />0.361 hours <br />0.00215 weeks <br />4.9465e-4 months <br />, during the initial drywell entry, a steam leak was observed on the Reactor Recirculation (RR) system line 2RR94AB-3/4, which is upstream of valve 2B33-FO8OB (RR Pump Discharge Valve 2B33-F067B Inspection Port - Reactor Side Upstream Stop Valve). At 1345 hours0.0156 days <br />0.374 hours <br />0.00222 weeks <br />5.117725e-4 months <br />, the leak was determined to be pressure boundary leakage. Technical Specification 3.4.5, RCS Operational Leakage, Required Actions C.1 and C.2 were entered, which require the unit to be in Mode 3 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 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, respectively. Unit 2 entered Mode 4 at 2209 hours0.0256 days <br />0.614 hours <br />0.00365 weeks <br />8.405245e-4 months <br /> on August 7, 2015.
This condition was reported (EN 51300) on August 7, 2015 to the NRC in accordance with 10 CFR 50.72(b)(3)(ii)(A) for the pressure boundary leakage as a principal safety barrier being in a seriously degraded condition.
The cause of the steam leak was determined to be poor weld quality and vibration induced fatigue. The weld was repaired during the maintenance outage.
NRC FORM 366 (02-2014)
LaSalle County Station Unit 2 is a General Electric Company Boiling Water Reactor with 3546 Megawatts Rated Core Thermal Power.
A.
CONDITION PRIOR TO EVENT
Unit(s): 2 Event Date: August 7, 2015 Event Time: 1345 CST Reactor Mode(s): 3 Mode(s) Name: Hot Shutdown Power Level: 0 percent B.
DESCRIPTION OF EVENT
On August 7, 2015, Unit 2 was in Mode 3 for a planned maintenance outage. At 1300 hours0.015 days <br />0.361 hours <br />0.00215 weeks <br />4.9465e-4 months <br />, during the initial drywell entry, a steam leak was observed on the Reactor Recirculation (RR)[ADJ system line 2RR94AB-3/4, which is upstream of valve 2B33-FO8OB (RR Pump Discharge Valve 2B33-F067B Inspection Port
- - Reactor Side Upstream Stop Valve). At 1345 hours0.0156 days <br />0.374 hours <br />0.00222 weeks <br />5.117725e-4 months <br />, the leak was determined to be pressure boundary leakage.
Technical Specification 3.4.5, RCS Operational Leakage, Required Actions C.1 and C.2 were entered, which require the unit to be in Mode 3 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 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, respectively. Unit 2 entered Mode 4 at 2209 hours0.0256 days <br />0.614 hours <br />0.00365 weeks <br />8.405245e-4 months <br /> on August 7, 2015.
This condition was reported (EN 51300) on August 7, 2015, to the NRC in accordance with 10 CFR 50.72(b)(3)(ii)(A) for the pressure boundary leakage as a principal safety barrier being in a seriously degraded condition.
C.
CAUSE OF EVENT
The cause for the steam leak on line 2RR94AB-3/4 was determined to be poor weld quality and vibration induced fatigue due to Reactor Recirculation system operation.
D.
SAFETY ANALYSIS
The safety significance of the event was minimal. Makeup capability was adequate to compensate for the leak.
All Emergency Core Cooling Systems (ECCS) were operable and capable of fulfilling their intended safety functions during the period of excessive leakage. The event did not constitute a safety system functional failure.
E.
CORRECTIVE ACTIONS
The leak was repaired by replacing line 2RR94AB-3/4, which included removal of the inspection/vent valves and installation of a pipe cap.
F.
PREVIOUS OCCURRENCES
LER 373-2013-005-00 On April 27, 2013, LaSalle Unit 1 was in Mode 2 (Startup) following a forced outage. At 1800 hours0.0208 days <br />0.5 hours <br />0.00298 weeks <br />6.849e-4 months <br /> CDT, during a walk down of the drywell, a steam leak was observed coming from the Reactor Core Isolation Cooling Steam Supply Inboard Isolation Bypass/Warm up Valve (1 E51 -F076), a normally-closed, one inch, motor operated valve. The leak was determined to be on the valve bonnet extension-to-bonnet upper seal weld. At 2124 hours0.0246 days <br />0.59 hours <br />0.00351 weeks <br />8.08182e-4 months <br /> CDT the leak was classified as reactor coolant pressure boundary leakage, and Technical Specification (TS) 3.4.5 Condition C was entered. TS 3.4.5 Required Actions C.1 and C.2 require that the unit be in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and Mode 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
The apparent cause was a weld defect or discontinuity from the original weld construction (i.e., manufacturing, installation/construction errors, etc.) of the upper seal weld that propagated through wall as a result of system loading and conditions (i.e., high pressure steam) during normal plant operations. Corrective actions included repair of the defective seal weld area.
LER 373-2011-002-00 On February 9, 2011, LaSalle Unit 1 was in Mode 2 (Startup) following a forced outage. A steam leak was observed coming from the Reactor Core Isolation Cooling Steam Supply Inboard Isolation Bypass/Warm up Valve (1 E51 -F076), a normally-closed, one inch, motor operated valve. The leak was determined to be on the valve bonnet extension-to-bonnet upper seal weld. At 1804 hours0.0209 days <br />0.501 hours <br />0.00298 weeks <br />6.86422e-4 months <br />, the leak was classified as pressure boundary leakage, and Technical Specification (TS) 3.4.5 Condition C was entered. TS 3.4.5 Required Action C.1 and C.2 require that the unit be in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and Mode 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
The equipment apparent cause evaluation determined that the cause was a weld defect or discontinuity from the original weld construction (i.e., manufacturing, installation/construction errors, etc.) of the upper seal weld that propagated through wall as a result of system loading and conditions (i.e., high pressure steam) during normal plant operations. Corrective action included repair of the defective upper seal weld area.
LER 374-2005-002-00 On March 12, 2005, during a scheduled refueling outage on Unit 2, a pinhole leak in a Class 1 weld on the outboard Main Steam Isolation Valve drain line (2B21 -F028D) was discovered during a hydrostatic test of the reactor coolant pressure boundary. The apparent cause of the leak was a weld inclusion or defect from a Class 1 weld made in 1995.
The weld was repaired, non-destructive surface examination performed, and the hydrostatic test was re-performed successfully within acceptance criteria.
G.
COMPONENT FAILURE DATA
No component failures occurred during this event.