05000457/LER-2018-001, Reactor Coolant System Pressure Boundary Leak on a Steam Generator Bowl Drain Line Due to High Cycle Fatigue Cracking Initiated from Small Welding Defects

Reactor Coolant System Pressure Boundary Leak on a Steam Generator Bowl Drain Line Due to High Cycle Fatigue Cracking Initiated from Small Welding Defects

ML18340A014
Person / Time
Site:	Braidwood
Issue date:	12/06/2018
From:	Marchionda-Palmer M Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
BW180111 LER 2018-001-00
Download: ML18340A014 (4)

LER-2018-001, Reactor Coolant System Pressure Boundary Leak on a Steam Generator Bowl Drain Line Due to High Cycle Fatigue Cracking Initiated from Small Welding Defects

Event date:
Report date:
Reporting criterion:	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)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)
4572018001R00 - NRC Website

text

December 6, 2018 BW180111 U.S. Nuclear Regulatory Commission A TIN: Document Control Desk Washington, DC 20555-0001 Braidwood Station, Unit 2 Renewed Facility Operating License No. NPF-77 NRC Docket No. STN 50-457 10 CFR 50.73

Subject:

Licensee Event Report 2018-001 Reactor Coolant System Pressure Boundary Leak on a Steam Generator Bowl Drain Line due to High Cycle Fatigue Cracking Initiated from Small Welding Defects The enclosed Licensee Event Report (LER) is being submitted in accordance with 1 O CFR 50.73, "Licensee Event Report System."

There are no regulatory commitments contained in this letter. Should you have any questions concerning this submittal, please contact Mr. Francis Jordan, Regulatory Assurance Manager, at (815) 417-2800.

Respectfully, Marri Marchionda-Palmer Site Vice President Braidwood Station

Enclosure:

LER 2018-001-00 cc: NRR Project Manager - Braidwood Station Illinois Emergency Management Agency - Division of Nuclear Safety US NRC Regional Administrator, Region Ill US NRC Senior Resident Inspector (Braidwood Station)

Illinois Emergency Management Agency - Braidwood Representative

NRC FORM 366 (04-2018)

U.S. NUCLEAR REGULATORY COMMISSION !!-PPROVED BY OMB: NO. 3150*0104 EXPIRES: 03/31/2020 LICENSEE EVENT REPORT (LER)

(See Page 2 for required number of digits/characters for each block)

(See NUREG-1022, R.3 for instruction and guidance

, !he NRG may not conduct or sponsor, and a llerson is not renuired to resoond to, the infonmation collection.

H

~~~

3. Page Braidwood Station, Unit 2 05000457 1

OF 3

4. Title Reactor Coolant System Pressure Boundary Leak on a Steam Generator Bowl Drain Line due to High Cycle Fatigue Cracking Initiated from Small Welding Defects

5. Event Date Month Day Year 10 08 2018

9. Operating Mode 3

10. Power Level 000 Licensee Contact Francis Jordan

Cause

System N/A N/A

6. LEA Number

7. Report Date

8. Other Facilities Involved I

Sequential I Rev Facility Name Docket Number Year Number No.

Month Day Year N/A N/A Facility Name Docket Number 2018 - 001 00 12 06 2018 N/A N/A

11. This Report is Submitted Pursuant to the Requirements of 10 CFR §: (Check all that apply)

D 20.2201(b)

D 20.2201(d)

D 20.2203(a)(1 i D 20.2203(a)(2)(i)

D 20.2203(a)(2)(ii)

D 20.2203(a)(2)(iii)

D 20.2203(a)(2)(iv)

D 20.2203(a)(2)(v)

D 20.2203(a)(2)(vi)

D 20.2203(a)(3)(i)

~ 50.73(a)(2)(ii)(A)

D 50.73(a)(2)(viii)(A)

D 20.2203(a)(3)(ii)

D 50.73(a)(2)(ii)(B)

D 50.73(a)(2)(viii)(B)

D 20.2203(a)(4)

D 50.73(a)(2)(iii)

D 50.73(a)(2)(ix)(A)

D 50.36( c)( 1 )(i)(A)

D 50.73(a)(2)(iv)(A)

D 50.73(a)(2)(x)

D 50.36( c)( 1 )(ii)(A)

D 50.73(a)(2)(v)(A)

D 73.71(a)(4)

D 50.36(c)(2)

D 50.73(a)(2)(v)(B)

D 73.71(a)(5)

D 50.46(a)(3)(ii)

D 50.73(a)(2)(v)(C)

D 13.77(a)(1)

D 50.73(a)(2)(i)(A)

D 50.73(a)(2)(v)(D)

D 13.77(a)(2)(i)

~ 50.73(a)(2)(i)(B)

D 50.73(a)(2)(vii)

D 13.77(a)(2)(ii)

D 50.73(a)(2)(i)(C)

D other (Specify in Abstract below or in

D. Safety Consequences

There were no safety consequences impacting plant or public safety as a result of this event.

SEQUENTIAL NUMBER 001 REV NO.

00 The Updated Final Safety Analysis Report (UFSAR) bounding accident for failure described in this evaluation would be a small break loss of coolant accident (SBLOCA), which is a break of the reactor coolant pressure boundary with a total cross-sectional area less than 1.0 square foot. The SG bowl drain line leak was downstream of a drain coupling with an orifice not exceeding 0.374 inches and is bounded by the evaluations discussed in the UFSAR; therefore, there would have been no safety consequences during a design basis event. There was no loss of safety function for this event.

Completed Corrective Actions

Replacement of the failed 2D SG bowl drain line as well as the drain lines for the remaining three SGs, using more robust socket welds and smaller weld filler rods.

F. Previous Occurrences

None

G. Component Failure Data

Manufacturer Nomenclature Mfg. Part Number N/A N/A N/A N/A Page_3_ of _3_