05000374/LER-2015-002, Regarding Two Main Steam Safety Relief Valves Failed Inservice Inspection Pressure Test

Regarding Two Main Steam Safety Relief Valves Failed Inservice Inspection Pressure Test

ML15105A268
Person / Time
Site:	LaSalle
Issue date:	04/15/2015
From:	Vinyard H Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RA15-010 LER 15-002-00
Download: ML15105A268 (4)

LER-2015-002, Regarding Two Main Steam Safety Relief Valves Failed Inservice Inspection Pressure Test

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 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
3742015002R00 - NRC Website

text

10 CFR 50.73 RA 15-010 April 15, 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 2015-002-00, Two Main Steam Safety Relief Valves Failed Inservice Inspection Pressure Test In accordance with 10 CFR 50.73(a)(2)(i)(B), Exelon Generation Company (EGC), LLC, is submitting Licensee Event Report Number 2015-002-00 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.

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: 01/31/2017 (02-2014)

Estimated burden per response to comply with this mandatory collection request: 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />.

Reported lessons learned are incorporated into the licensing process and fed back to industry.

Send comments regarding burden estimate to the FOIA, Privacy and Information Collections LICENSEE VENT REPORT (LER)

Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by il t l f ll i

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on an (See Page 2 for required number of Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC digits/characters for each block) 20503. If a means used to impose an information collection does not display a currently valid OMB 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 Two Main Steam Safety Relief Valves Failed Inservice Lift Inspection Pressure Test

5. EVENT DATE

6. LER NUMBER

7. REPORT DATE

8. OTHER FACILITIES INVOLVED MONTH DAY YEAR YEAR SEQUENTIAL NUMBER REV NO.

MONTH DAY YEAR FACILITY NAME N/A DOCKET NUMBER N/A 02 14 2015 2015 -

002 00 04 15 2015 FACILITY NAME N/A DOCKET NUMBER N/A

9. OPERATING MODE

11. THIS REPORT IS SUBMITTED PURSUANT TO T HE REQUIREMENTS OF 10 CFR §: (Check ail that apply) q 20.2201(b) q 20.2203(a)(3)(i) q 50.73(a)(2)(i)(C) q 50.73(a)(2)(vii) q 20.2201(d) q 20.2203(a)(3)(ii) q 50.73(a)(2)(ii)(A) q 50.73(a)(2)(viii)(A) 5 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) q 20.2203(a)(2)(iv) q 50.46(a)(3)(ii) q 50.73(a)(2)(v)(B) q 73.71 (a)(5) 000 q 20.2203(a)(2)(v) q 50.73(a)(2)(i)(A) q 50.73(a)(2)(v)(C) q OTHER q 20.2203(a)(2)(vi)

[I 50 73(a)(2)(i)(B) 73(a)(2)(v)(D)

El 50 Specify r

ct below or in orm m 366

12. LICENSEE CONTACT FOR T HIS LER LICENSEE CONTACT TELEPHONE NUMBER (Include Area Code)

Andrew Schierer, Programs Engineering Manager 815-415-3846CAUSE SYSTEM COMPONENT MANU FACTURER REPORTABLE TO EPIX

CAUSE

SYSTEM COMPONENT MANU-FACTURER REPORTABLE TO EPIX

14. SUPPLEMENTAL REPORT EXPECTED

15. EXPECTED MONTH DAY YEAR YES (If yes, complete 15. EXPECTED SUBMISSION DATE)

F] NO SUBMISSION 07 17 2015 ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)

During the February 2015 Unit 2 refueling outage L2R1 5, two main steam safety relief valves (SRV) did not pass Technical Specification Surveillance Requirement 3.4.4.1 and Inservice Testing Program lift pressure requirements.

Both SRVs lifted below their expected lift pressures. SRV 2B21-F013S was required to lift within plus or minus three percent of 1150 psi (i.e., 1150 psi plus or minus 34.5 psi) and actually lifted at 1099 psi. SRV 2B21-F013M was required to lift within plus or minus three percent of 1195 psi (i.e., 1195 psi plus or minus 35.8 psi) and actually lifted at 1145 psi.

Both SRVs were replaced during the outage. The cause of the event is under review by the vendor testing laboratory, and will be reported in a supplement to this LER upon completion.

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: February 14, 2015 Event Time: 1530 CST Reactor Mode(s): 5 Mode(s) Name: Refueling Power Level: 0%

B.

DESCRIPTION OF EVENT

During the February 2015 Unit 2 refueling outage L2R15, two main steam safety relief valves (SRV)[AD] did not pass Technical Specification (TS) Surveillance Requirement 3.4.4.1 and Inservice Testing Program lift pressure requirements. Both SRVs lifted below their expected lift pressures. SRV 2B21-F013S was required to lift within plus or minus three percent of 1150 psi (i.e., 1150 psi plus or minus 34.5 psi), but actually lifted at 1099 psi. SRV 2821-F013M was required to lift within plus or minus three percent of 1195 psi (i.e., 1195 psi plus or minus 35.8 psi), but actually lifted at 1145 psi.

This condition was discovered while Unit 2 was outside the mode of applicability for TS 3.4.4, Safety/Relief Valves (Modes 1, 2 and 3); however, multiple test failures are reportable under 10 CFR 50.73(a)(2)(i)(B) as an operation or condition prohibited by the plant's Technical Specifications.

C.

CAUSE OF EVENT

The cause of the event is under review by the testing facility vendor and will be reported in a supplement to this LER upon completion.

D.

SAFETY ANALYSIS

The safety significance of this condition was minimal. The out-of-tolerance lift pressures were discovered while the plant was in Mode 5 during a refueling outage and the SRVs were not required to be operable. Both SRVs lifted prior to their expected lift pressures, which is conservative in regards to maintaining reactor pressure vessel overpressure limits.

The cause of the event is under review by the vendor testing facility. Upon completion of the review, the cause will be evaluated for any potential safety concerns and will be reported in a supplement to this LER.

E.

CORRECTIVE ACTIONS

Both SRVs were replaced during the outage.

The cause of the event was placed under review by the vendor testing laboratory.

F.

PREVIOUS OCCURRENCES

A review of past events identified no reportable occurrences of out-of-tolerance safety relief valve pressures in the previous ten years.

G. COMPONENT FAILURE DATA

Crosby Safety Relief Valves for Main Steam Service, Style HB-65-BP, Size 6R10. ASME Section III, Class I