05000387/LER-2024-001, Main Steam Isolation Valve Leakage

Main Steam Isolation Valve Leakage

ML24144A278
Person / Time
Site:	Susquehanna
Issue date:	05/23/2024
From:	Casulli E Susquehanna, Talen Energy
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
PLA-8123 LER 2024-001-00
Download: ML24144A278 (1)

LER-2024-001, Main Steam Isolation Valve Leakage

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(iii) 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
3872024001R00 - NRC Website

text

Edward Casulli Susquehanna Nuclear, LLC Site Vi ce Pr esid ent 769 Sa l em Bou levard Berwick, PA I 8603 TALEN~

Te l. 570.542.3795 Fax 570.542. 1504 Edward.Casul l i@T alenEnergy.co m ENERGY

May 23, 2024

Attn: Document Control Desk 10 CFR 50.73 U. S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUSQUEHANNA STEAM ELECTRIC STATION LICENSEE EVENT REPORT 50-387/2024-001-00 UNIT 1 LICENSE NO. NPF-14 PLA-8123 Docket No. 50-387

Attached is Licensee Event Report (LER) 50-387/2024-001-00. The LER reports an event involving inoperability of a Main Steam Line Isolation Valve that was determined to be reportable in accordance with 10 CFR 50.7 3(a)(2)(i)(B) as a condition prohibited by Technical Specifications.

There were no actual consequences to the health and safety of the public as a result of this event.

This letter contains no new or revised regulatory commitments.

E. Casulli

Attachment: L E R 50-387 / 2024-001-00

Copy: NRC Region I Ms. J. England, NRC Senior Resident Inspector Ms. A. Klett, NRC Project Manager Mr. M. Shields, PA DEP/BRP NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB : NO. 3150-0104 EXPIRES : 04/30/2027 (04 2024) Estimated burden per response to co mply 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 ticensing process and fed back to industry. Send comments regarding burden

(¥~1 LICENSEE EVENT REPORT (LER) estimate to the FOIA, Library, and Information Collections Branch.Resource @nrc.gov. and the 0MB reviewer (T-6 A \\O M), U. S. Nuc lea-Regula10f'/

.,. ___.., (See Page 2 for required number of digits/characters for each block) Comm ission, Washington, DC 20555-0001. or by email to lnfocollectsat 0MB Office of lnfonmation and Regula\\QI'/ Affairs, (3150-0104), Atln: Desk Officer for \\he Nuclear Regula10f'/

(S ee NUR EG -1022, R.3 for instruction and guida nce for comple ting this form Commission, 725 17th Street NW, Washington, DC 20503. The NRC may not conduct or sponsor, and a person is h!!Q://www.nre.gov/read ing -[!JJ/doc-col lectio ns /o u_rgg 'a/staff l sr1 022/r 3/ ). not requir ed to respond to, a collection of information unless the document requesting or requiring the collection displays a currently va lid 0MB control number.

1. Facility Name 00 050 2. Docket Number 3. Page Susquehanna Steam Electric Station Unit 1 052 00387 1 OF 2

4. Title Main Steam Isolation Valve Leakage

5. Event Date 6. LER Number 7. Report Date 8. Other Facilities Invol ved

Month Day Year Yea r Sequential Revision Month Day Year Facility Name Dock et Number Number No. 050

03 28 2024 2024 - 001 - 00 05 23 2024 Facility Name 052 Docket Numb e r

9. Operating Mode 110. Power Level 5 0

11. This Report is Submitted Pursuant to the Requirements of 10 CFR § : (Check all that apply) 10 CFR Part 20 20.2203(a)(2)(vi) 10 CFR Part 50 50. 73(a)(2)(ii)(A) 50. 73(a)(2)(viii)(A) 73.1200(a) 20.2201(b) 20.2203(a)(3)(i) 50.36(c)(1 )(i)(A) 50. 73(a)(2)(ii)(B) 50. 73(a)(2)(viii)(B) 73.1200(b) 20.2201(d) 20.2203(a)(3)(ii) 50.36(c)(1 )(ii)(A) 50.73(a)(2)(iii) 50. 73(a)(2)(ix)(A) 73.1200(c) 20.2203(a)(1) 20.2203(a)(4) 50.36(c)(2) 50. 73(a)(2)(iv)(A) 50. 73(a)(2)(x) 73.1200(d) 20.2203(a)(2)(i) 10 CFR Part 21 50.46(a)(3)(ii) 50. 73(a)(2)(v)(A) 10 CFR Part 73 73. 1200(e) 20.2203(a)(2)(ii) 21.2(c) 50.69(g) 50.73(a)(2)(v)(B) 73.77(a)(1) 73.1200(f) 20.2203(a)(2)(iii) 50.73(a)(2)(i)(A) 50.73(a)(2)(v)(C) 73. 77(a)(2)(i) 73.1200(g) 20.2203(a)(2)(iv) [Z] 50. 73(a)(2)(i)(B) 50.73(a)(2)(v)(D) 73. 77(a)(2)(ii) 73.1200(h) 20.2203(a)(2)(v) 50. 73(a)(2)(i)(C) 50. 73(a)(2)(vii)

OTHER (S pecify here, in abstra ct, or NR G 366A).

12. Licensee Contact for this LER

Licensee Contact Phone Number (Include area code)

Brad Yarzebinski, Nuclear Regulatory Affairs Engineer 570-542-2839

13. Complete One Line for each Component Failure Described in this Report

Cause System Component Manufacturer Reportable to IRIS Cause System Component Manufacturer Reportable to IRIS

14. Supplemental Report Expected Month Day Year

15. Expected Submission Date No Ye s (If ye s, complete 15. Expected Submission Date) 07 26 2024 [Z]

16. Abstract (Limit to 1326 spaces, i.e., ap proximately 13 single-spaced typewritten li nes)

On March 28, 2024, during Local Leak Rate Testing (LLRT) conducted during the Unit 1 refueling outage, as-found leakage through the inboard Main Steam Isolation Valve (HV141 F0228) was 57,166 standard cubic centimeters per minute (seem) which exceeded the Technical Specification (TS) Surveillance Requirement 3.6. 1.3.12 limit of 100 standard cubic feet per hour (scfh)

(47,194 seem) for individual valve leakage. HV141 F022B was repaired resulting in a total penetration LLRT value of 1,568 seem.

There is preliminary indication that the condition ex isted longer than allowed by TS 3.6. 1.3. The condition is being reported in accordance with 10 CFR 50. 73(a)(2)(i) (B) as a condition prohibited by TS.

A cause evaluation is in progress. A supplement will be issued to provide information regarding the cause of the condition and additional corrective actions.

There were no actual consequences to the health and safety of the public as a result of this event.

CONDITIONS PRIOR TO EVENT

Unit 1 - Mode 5, zero (0) percent Rated Thermal Power (RTP)

Unit 2 - Mode 1, approximately 100 percent RTP

EVENT DESCRIPTION

On March 28, 2024, during Local Leak Rate Testing (LLRT) conducted during the Unit 1 refueling outage,

as-found leakage through the inboard Main Steam Isolation Valve (MSIV) (HV141F022B) [EIIS System/ Component Code: SB/ISV] was 57,166 standard cubic centimeters per minute (seem) which exceeded the Technical Specification Surveillance Requirement (TS SR) 3.6.1.3.12 limit of 100 standard cubic feet per hour (scfh) (47, 194 seem) fo r individual valve leakage. HV141 F022B was repaired resulting in an acceptable as-left penetration LLRT value of 1,568 seem.

Based on preliminary information, there is evidence that the condition existed during the last operating cycle for longer than allowed by TS 3.6.1.3. The condition is being reported in accordance with 10 CFR 50.73(a)(2)(i)(B) as a condition prohibited by Technical Specifications.

CAUSE OF EVENT

The cause of the event is under investigation and will be provided in a supplement to this LER.

ANALYSIS/SAFETY SIGNIFICANCE

The redundant MSIV (HV141F028B) in the "B" steam line had an as-found individual leakage value of 4,029 seem, which is below the TS SR 3.6.1.3.12 limit of 100 scfh (47, 194 seem). The redundant valve, bounded by engineering evaluation,

provides assurance that the dose consequences remain within the regulatory limit of 5 rem Total Effective Dose Equivalent (TEDE) for the control room and 25 rem TEDE for the low population zone and exclusion area boundary. As such, there was no loss of safety function for the redundant valve or the main steam penetrat ion. Accordingly, this event will not be counted as a safety system functional failure in the Reactor Oversight Process Performance Indicators. There were no actual consequences to the health and safety of the public as a result of this event.

CORRECTIVE ACTIONS

HV141 F022B was repaired resulting in a leak rate within the TS SR 3.6. 1.3.12 limit. Any additional corrective actions w ill be provided in a supplement to this LER.

PREVIOUS OCCURRENCES

Previous occurrences, if any, will be provided in a supplement to this LER.