05000397/LER-2025-001, Pressure Boundary Leakage Resulting in a Technical Specification Required Plant Shutdown, Degraded Condition, and Condition Prohibited by Technical Specifications

Pressure Boundary Leakage Resulting in a Technical Specification Required Plant Shutdown, Degraded Condition, and Condition Prohibited by Technical Specifications

ML25161A146
Person / Time
Site:	Columbia
Issue date:	06/10/2025
From:	David Brown Energy Northwest
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
GO2-25-085 LER 2025-001-00
Download: ML25161A146 (1)

LER-2025-001, Pressure Boundary Leakage Resulting in a Technical Specification Required Plant Shutdown, Degraded Condition, and Condition Prohibited by Technical Specifications

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
3972025001R00 - NRC Website

text

GO2-25-085 10 CFR 50.73 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001

Subject:

COLUMBIA GENERATING STATION, DOCKET NO. 50-397 LICENSEE EVENT REPORT NO. 2025-001-00

Dear Sir or Madam:

Transmitted herewith is Licensee Event Report number 2025-001-00 for Columbia Generating Station. This report is submitted pursuant to 10 CFR 50.73(a)(2)(i)(A), 10 CFR 50.73(a)(2)(ii)(A), and 10 CFR 50.73(a)(2)(i)(B).

There are no commitments being made to the Nuclear Regulatory Commission by this letter. If you have any questions or require additional information, please contact Mr. Z. K.

Dunham, Regulatory Affairs Manager, at (509) 377-4735.

Executed on this ______ day of ____________, 2025.

Respectfully, David P. Brown Site Vice President Attachment: Licensee Event Report 2025-001-00 cc:

NRC Region IV Regional Admin NRC Region IV Project Manager NRC Senior Resident Inspector/988C C.D. Sonoda - BPA David P. Brown Columbia Generating Station P.O. Box 968, MD PE23 Richland, WA 99352-0968 Ph. 509-377-8385 dpbrown@energy-northwest.com

 
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June 10, 2025 ENERGY NORTHWEST

Abstract

Columbia Generating Station

397 3

Pressure Boundary Leakage Resulting in a Technical Specification Required Plant Shutdown, Degraded Condition, and Condition Prohibited by Technical Specifications 04 12 2025 2025 001 00 Mode 3 0 percent

Valerie Lagen (509) 372-5507

At 02:23 on April 12, with the unit in Mode 3 at 0 percent power, pressure boundary leakage was identified on the bonnet vent line elbow connection of the Reactor Recirculation (RRC) Pump 1B discharge isolation valve, RRC-V-67B. Leakage from the area was initially identified during a drywell entry at 15 percent reactor power and later confirmed as pressure boundary leakage during a subsequent entry in Mode 3.

This event is being reported pursuant to 10 CFR 50.73(a)(2)(i)(A) as a Plant Shutdown Required by Technical Specifications, 10 CFR 50.73(a)(2)(ii)(A) as a Degraded or Unanalyzed Condition that Resulted in the condition of the nuclear power plant, including its principal safety barriers being seriously degraded, and 10 CFR 50.73(a)(2)(i)(B) as an Operation or Condition Prohibited by Technical Specifications.

There were no structures, systems, or components that were inoperable at the beginning of the event that contributed to the event.

 
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06 10 2025

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I Description of events:

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55&SXPS%GLVFKDUJHLVRODWLRQYDOYH55&9%>$',69@ZDVLGHQWLILHGDVSUHVVXUHERXQGDU\\OHDNDJH/HDNDJH

IURPWKHDUHDZDVLQLWLDOO\\LGHQWLILHGGXULQJGU\\ZHOOHQWU\\DWSHUFHQWUHDFWRUSRZHURQ$SULODQGODWHU

FRQILUPHGDVSUHVVXUHERXQGDU\\OHDNDJHGXULQJDVXEVHTXHQWHQWU\\LQ0RGHTechnical Specification (TS) /LPLWLQJ

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$FWLRQV C DQG&UHTXLUHWhe 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 Mode 4 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The unit was in Mode 3 at the timeofWKHHYHQW. Mode 4 was entered at 09:32 on April 12, 2025.

Plant Conditions

At the time of the event, the plant was LQ0RGHat percent poZHU.

Energy Industry Identification System (EIIS) codes are identified as [XX].

Background:

On 7/1/2023 (approximately 2 weeks after startup from the previous refueling outage) an indication was received that a Reactor Coolant System (RCS) [AB] leak existed whHQERWKContainment Monitoring System Radiation Ratemeters [IK

075] were reading higher than normal and tracking similarly. During this time Equipment Drain Radioactive (EDR) [WK]

leakage and Floor Drain Radioactive (FDR) [WK] leakage, containment temperature, and containment pressure were steady and did not indicate a leak.Chemistry performed an isotopic analysis of a grab sample of the containment atmosphere which did not indicate leakage.,Q-XO\\2023, a Technical Issues Resolution Process was initiated to analyze trends indicating potential leaks in containment.,Q$XJXVW 2023, an 2perational 'ecision-0aking,ssue 2'0,was created to establish RCS leakage monitoring criteria and identify trigger points with actions. The leakage remained constant throughout the operating cycle and did not challenge the TS RCS Operational limits of 5 gallons per minute of unidentified leakage, or an increase of 2 gallons per minute over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of unidentified leakage.

Root Cause:

Vent piping attached to the bonnet of RRC-V-B was not designed to prevent high cycle fatigue. This resulted in cyclic fatigue of a weld on a 3/4-inch vent line from the RRC-V-67B socket welded connection. This is a legacy issue associated with the original designs lack of dynamic analysis, or with the 1996 and 1997 GE analysis and testing of small-bore piping in support of Adjustable Speed Drive modifications. $GMXVWDEOHVSHHGGULYHFRQWUROVWKHVSHHGRIWKH55&>$'@

Safety Significance

The 5HDFWRU&RRODQW6\\VWHP5&6>$%@OHDN was LQLWLDOO\\estimated to be between 0.05 and 0.10 gallons per minute.

While the leak remained below the TS RCS2SHUDWLRQDOOLPLWVIRUXQLGHQWLILHGOHDNVIURPWKHWLPHRILQLWLDOLQGLFDWLRQXQWLO

SRVLWLYHLGHQWLILFDWLRQ, LIWKHOHDNKDGLQFUHDVHGDQGH[FHHGHG76XQLGHQWLILHGOHDNUDWHOLPLWVDQRUGHUO\\VKXWGRZQZRXOG

KDYHEHHQLQLWLDWHGDQGWKHUHDFWRUZRXOGKDYHEHHQSODFHGLQFROGVKXWGRZQFRQGLWLRQLQDFFRUGDQFHZLWKWKH76

Immediate Corrective Actions

The RRC-V-67B weld leak was reSDLUHG

Planned Corrective Actions

 Perform verification that cracks have not been initiated at the surfaces and roots of welds identifiedDVEHLQJVXEMHFWWR

F\\FOLFIDWLJXH

 Ensure a drywell walkdown is performed to assess evidence of leakage from the welds identified as the suspect

population.

 Implement RCS Leakage ODMI for next cycle.

 Redesign the RRC-V-67B vent line to sufficiently address the cyclic fatigue issue.

 Implement the new design of RRC-V-67B vent line.

 Perform H[WHQWRIFRQGLWLRQE\\XVHRInon-destructive examinations on LGHQWLILHG small-bore pipe weldsRQVPDOOERUHSLSH

ZHOGVWRILQGHYLGHQFHRIWKURXJKZDOOOHDNDJH

 Perform pipe stress/vibrational analysis on lines identified as being susceptible to cyclic fatigue and perform an

engineering change to remove VXVFHSWLELOLW\\WRF\\FOLFIDWLJXH.

Previous Similar Events

No similar events identified in last 10 years. Page 3

of 3

 
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