05000373/LER-2013-005, Regarding Technical Specification Required Shutdown Due to Pressure Boundary Leakage

Regarding Technical Specification Required Shutdown Due to Pressure Boundary Leakage

ML13177A278
Person / Time
Site:	LaSalle
Issue date:	06/26/2013
From:	Vinyard H Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RA13-036 LER 13-005-00
Download: ML13177A278 (4)

LER-2013-005, Regarding Technical Specification Required Shutdown Due to Pressure Boundary Leakage

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
3732013005R00 - NRC Website

text

10 CFR 50.73 RA13-036 June 26, 2013 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 LaSalle County Station, Unit I Facility Operating License No. NPF-11 NRC Docket No. 50-373

Subject:

Licensee Event Report 2013-005-00 Plant Shutdown Required by Technical Specifications Due to Pressure Boundary Leakage In accordance with 10 CFR 50.73(aX2XiXA) and 10 CFR 50.73(ax2XiiXA), Exelon Generation Company (EGC), LLC, is submitting Licensee Event Report Number 2013 -

005-00 for LaSalle Unit 1.

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.

Harold T. Vinyard Plant Manager LaSalle County Station

Enclosure:

Licensee Event Report cc:

Regional Administrator - NRC Region III NRC Senior Resident Inspector - LaSalle County Station

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/31/2013 (10-2010)

, the NRC may digits/characters for each block) not conduct or sponsor, and a person is not required to respond to, the i

ll ti i format on co ec on.

n

3. PAGE LaSalle County Station, Unit 1 05000373 1

OF 3

4. TITLE Technical Specification Required Shutdown Due to Pressure Boundary Leakage

5. EVENT DATE

6. LER NUMBER

7. REPORT DATE

8. OTHER FACILITIES INVOLVED MONTH DAY YEAR YEAR SEQUENTIAL REV MONTH DAY YEAR FACILITY NAME N/A DOCKET NUMBER N/A NUMBER NO NO.

FACILITY NAME DOCKET NUMBER 04 27 2013 2013 005 00 06 26 2013 N/A 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)

El 20.2203(a)(3)(i)

El 50.73(a)(2)(i)(C)

El 50.73(a)(2)(vii) 2 q 20.2201(d) q 20.2203(a)(3)(ii)

Z 50.73(a)(2)(ii)(A) q 50.73(a)(2)(viii)(A) q 20.2203(a)(1) q 20.2203(a)(4) q 50.73(a)(2)(ii)(B) q 50.73(a)(2)(viii)(B) q 20.2203(a)(2)(i) q 50.36(c)(1)(i)(A) q 50.73(a)(2)(iii) q 50.73(a)(2)(ix)(A)

10. POWER LEVEL q 20.2203(a)(2)(ii) q 50.36(c)(1)(ii)(A) q 50.73(a)(2)(iv)(A) q 50.73(a)(2)(x) q 20.2203(a)(2)(iii) q 50.36(c)(2) q 50.73(a)(2)(v)(A) q 73.71(a)(4) 006 q 20.2203(a)(2)(iv) q 50.46(a)(3)(ii) q 50.73(a)(2)(v)(B) q 73.71(a)(5) q 20.2203(a)(2)(v) 50.73(a)(2)(i)(A) q 50.73(a)(2)(v)(C) q OTHER q 20.2203(a)(2)(vi) q 50.73(a)(2)(i)(B) q 50.73(a)(2)(v)(D)

Specify in Abstract below or in E.

CORRECTIVE ACTIONS

The entirety of the original weld was ground out, including the previous weld repair, and was replaced with new weld covering the entire radius of the affected joint.

Actions to identify and order the correct replacement valve are in-progress. There is currently no spare, and the original valve manufacturer no longer exists.

F.

PREVIOUS OCCURRENCES

LER 05000373/2011-002-00 On February 9, 2011, LaSalle Unit 1 was in Mode 2 (Startup) following a forced outage. During a walk down of the drywell, a steam leak was observed coming from the Reactor Core Isolation Cooling (RI)[BN] 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 Specifications (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 />.

Corrective actions were to grind-out the defect and perform a code weld repair of the affected area. The corrective action from the most current event was to grind out the entirety of the original weld, including the previous weld repair, and to replace it with new weld covering the entire radius of the affected joint.

G.

COMPONENT FAILURE DATA

Manufacturer: Anderson Greenwood Company (A415)

Model Number: N03-6498-510