05000278/LER-1917-001, Regarding Reactor Pressure Boundary Leakage Due to Weld Failure in One-Inch Diameter Instrument Line

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Regarding Reactor Pressure Boundary Leakage Due to Weld Failure in One-Inch Diameter Instrument Line
ML17355A003
Person / Time
Site: Peach Bottom Constellation icon.png
Issue date: 12/20/2017
From: Pat Navin
Exelon Generation Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
CCN: 17-106 LER 17-001-00
Download: ML17355A003 (4)


LER-1917-001, Regarding Reactor Pressure Boundary Leakage Due to Weld Failure in One-Inch Diameter Instrument Line
Event date:
Report date:
Reporting criterion: 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)(ii)(A), Seriously Degraded
2781917001R00 - NRC Website

text

Exelon Generation December 20, 2017 U. S. Nuclear Regulatory Commission A TIN: Document Control Desk Washington, DC 20555-0001

Subject:

Peach Bottom Atomic Power Station (PBAPS) Unit 3 Renewed Facility Operating License No. DPR-56 NRC Docket No. 50-278 Licensee Event Report (LER) 3-17-001 10CFR 50.73 Enclosed is a Licensee Event Report concerning a condition prohibited by Technical Specifications due to a reactor pressure boundary leakage from a weld. In accordance with NEI 99-04, the regulatory commitment contained in this correspondence is to restore compliance with the regulations. The specific methods that have been planned to restore and maintain compliance are discussed in the LER. If you have any questions or require additional information, please do not hesitate to contact Jim Kovalchick at 717-456-3351.

Since~

Patrick D. Navin Site Vice President Peach Bottom Atomic Power Station PDN/dnd/IR 4065691 Enclosure cc:

US NRC, Administrator, Region I US NRC, Senior Resident Inspector R. R. Janati, Commonwealth of Pennsylvania S. Gray, State of Maryland B. Watkins, PSE&G, Financial Controls and Co-owner Affairs CCN: 17-106

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 03/31/2020 (04*2017)

, the http://www.nrc.qov/readinq-rm/doc-collections/nureqs/staff/sr1022/r3/)

NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

3. PAGE Peach Bottom Atomic Power Station Unit 3 05000278 1 OF 3
4. TITLE Reactor Pressure Boundary Leakage Due to Weld Failure in One-Inch Diameter Instrument Line
5. EVENT DATE
6. LEA NUMBER
7. REPORT DATE
8. OTHER FACILITIES INVOLVED I

SEQUENTIAL I REV FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR NUMBER NO.

MONTH DAY YEAR 05000 FACILITY NAME DOCKET NUMBER 10 23 2017 2017 - 001

- 00 12 21 2017 05000
9. OPERATING MODE
11. THIS REPORT IS SUBMIITED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) 0 20.2201 (b)

D 20.2203(a)(3)(il 181 50. 73(a)(2)(ii)(A) 0 50.73(a)(2)(viii)(A) 0 20.2201 (d)

D 20.2203(a)(3)(iil 0 50.73(a)(2)(ii)(B) 0 50.73(a)(2)(viii)(B) 3 0 20.2203(a)(1)

D 20.2203(a)(4l 0 50.73(a)(2)(iii) 0 50.73(a)(2)(ix)(A)

D 20.2203(aH2Hil D 50.36(c)(1)(i)(A)

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

D 50.73(a)(2)(x)

10. POWER LEVEL D 20.2203(a)(2)(iil 0 50.36(c)(1 )(ii)(A)

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

D 13.11 (a)(4l D 20.2203(a)(2)(iiil D 50.36(c)(2)

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

D 73.71 (a)(5)

D 20.2203(a)(2)(iv)

D so.46(a)(3)(iil D 50.73(a)(2)(v)(C)

D 13.77(a)(1) 000 D 20.2203(a)(2)(v) 0 50.73(a)(2)(i)(A)

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

D 73.77(a)(2)(il 0 20.2203(a)(2)(vi)

[8J 50.73(a)(2)(i)(B) 0 50.73(a)(2)(vii)

D 73.n(a)(2)(iil D so.73(a)(2)(i)(C)

D OTHER Specify in Abstract below or m SEQUENTIAL NUMBER 001 REV NO.

00

2. 10 CFR 50.73(a)(2)(ii)(A) - Degradation of the RCS - Because of the RCS pressure boundary leakage, one of the principal safety barriers of the plant was degraded.

Cause of the Event

The section of the one-inch pipe containing the cracked socket weld was sent to an offsite laboratory for analysis. The pipe was sectioned and lack of fusion defects were identified at the root of the weld. The crack initiated at the location of these defects. The defects would have occurred when the weld was performed during a recirculation system pipe replacement in the late 1980's. It is unknown when the crack began, but normal vibration for the piping likely caused it to propagate to the surface of the weld, resulting in the identified leak.

Corrective Actions

The section of pipe and the associated fitting were replaced. Instrument lines connected to the suction and discharge lines for both of the recirculation pumps with similar configuration and subject to vibration were also replaced during the refueling outage. The new welds were performed with a 2:1 profile, which reduces their susceptibility to vibration-induced failures.

Previous Similar Occurrences A similar event occurred in September of 2005. A crack in a socket weld caused by a lack of fusion defect resulted in a 1 gpm leak in a one-inch equalizing line for a check valve on the 'A' Residual Heat Removal (AHR) injection line. The event is documented in LEA 2005-003, dated 10/28/2005. Additional information on this previous occurrence is contained in the corrective action program. Page _3_ of _3_