05000454/LER-2017-001, Regarding Volumetric and Surface Examinations of Reactor Pressure Vessel Head Penetration Nozzles Identify Indications Attributed to Primary Water Stress Corrosion Cracking and Minor Subsurface Void Enlargement from
| ML17115A068 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 04/25/2017 |
| From: | Kanavos M Exelon Generation Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| BYRON 2017-0033 LER 17-001-00 | |
| Download: ML17115A068 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(iii) 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)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability |
| 4542017001R00 - NRC Website | |
text
Byron Generating Station Exe(on Generat Rd www.exebncorp.com April 25, 2017 LTR:
BYRON 2017-0033 File:
1.10.0101 (1D.101) 2.07.0100 (5A.108)
United States Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Byron Station, Unit 1 Renewed Facility Operating License No. NPF-37 NRC Docket No. STN 50-454
Subject:
Licensee Event Report (LER) 454-2017-001-00, Byron Station Unit 1
Volumetric and Surface Examinations of Reactor Pressure Vessel Head Penetration Nozzles Identify Indications Attributed to Primary Water Stress Corrosion Cracking and Minor Subsurface Void Enlargement from Operating Stresses Enclosed is Byron Station Licensee Event Report (LER) No. 454-2017-001-00 regarding Byron Station Unit 1 indications on reactor head penetration nozzles.
This condition is reportable in accordance with 10 CFR 50.73(a)(2)(ii)(A) for any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded.
There are no regulatory commitments in this report.
Should you have any questions concerning this submittal, please contact Mr. Douglas Spitzer, Regulatory Assurance Manager, at (815) 406-2800.
Respectfully, Mark E. Kanavos Site Vice President Byron Generating Station MEKIGC/sg
Enclosure:
LER 454-2017-001-00 cc:
Regional Administrator NRC Region Ill NRC Senior Resident Inspector Byron Generating Station
NRC FORM 366 U.s. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB: NO. 31 50-0104 EXPIRES: 03/31/2020 (04-2017)
, the htIJy//www.nrc.qov/readinq-rrn/doc-collections/nureqs/staff/sr1022/r3/)
NRC may not conduct or sponsor. and a person is not required to respond to. the information collection.
I 3. PAGE Byron Station Unit 1 05000454 1
OF 4
- 4. TITLE Byron Station Unit 1 Volumetric and Surface Examinations of Reactor Pressure Vessel Head Penetration Nozzles Identify Indications Attributed to Primary Water Stress Corrosion Cracking and Minor Subsurface Void Enlargement from Operating Stresses
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITiES INVOLVED FACiLITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR SEQUENTIAL RNEOV MONTH DAY YEAR N/A N/A FACILITY NAME DOCKET NUMBER 03 03 17 2017 001 00 04 25 17 N/A N/A
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITEED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Checkallthat apply)
LI 202201(b)
LI 20.2203(a)(3)O) 50.73(a)(2)(i)(A)
LI 50.73(a)(2)(viii)(A)
LI 202201(d)
LI 20.2203(a)(3)çi)
LI 50.73(a)(2)Oi)(B)
[=1 50.73(a)(2)(viii)(B)
Defueled LI 20.2203(a)(1)
LI 20.2203(a)(4)
LI 50.73(a)(2)(iii)
LI 50.73(a)(2)(Ix)(A)
LI 20.2203(a)(2)O)
LI 50.36(c)(1)(i)(A)
LI 50.73(a)(2)(iv)(A)
LI 50.73(a)(2)(x)
- 10. POWER LEVEL LI 20.2203(a)(2)(ii)
LI 50.36(c)(1)(i)(A)
LI 50.73(a)(2)(v)(A)
LI 73.71(a)(4)
LI 20.2203(a)(2)Oii)
LI 50.36(c)(2)
LI 50.73(a)(2)(v)(B)
LI 73.71(a)(5)
LI 202203(a)(2)Ov)
LI 50.46(a)(3)(ii)
LI 50.73(a)(2)(v)(C)
LI 73.77(a)(1)
Opercent LI 20.2203(a)(2)(v)
LI 50.73(a)(2)(i)(A)
LI 50.73(a)(2)(v)(D)
LI 73.77(a)(2)O)
LI 20.2203(a)(2)(vi)
LI 50.73(a)(2)(i)(B)
LI 50.73(a)(2)(vii)
LI 73.77(a)(2)(ii)
LI 5073(a)(2)(i)(C)
LI OTHER Specify in Abstracr below or in VHP 77 nozzle Mad one unacceptable indication in the tube base material identified by volumetric (UT) examination.
The unacceptable indication was 0.315 inches in length and 0.156 inches deep (25 percent through wall) from the nozzle OD surface. This indication was located 1.823 to 2.138 from the end of the nozzle.
This condition is reportable to the NRC in accordance with 10 CFR 50.73 (a)(2)(ii)(A), as a condition that resulted in a principle safety barrier being seriously degraded.
This LER is being submitted in follow-up to three emergency notification system ENS calls made in accordance with 10 CFR 50.72(b)(3)(ii)(A):
ENS 52591 dated March 3, 2017 at 2052 EST ENS 52592 dated March 5, 2017 at 1605 EST ENS 52592, supplemented March 6, 2017 at 1750 EST.
C. Cause of Event
The cause of the P-31 recordable indication is attributed to existing welding discontinuities/minor subsurface voids opening to the surface or enlarging due to thermal and/or pressure stresses during plant operation.
The cause of the P-74, P-76 and P-77 recordable indications is attributed to Primary Water Stress Corrosion Cracking (PWSCC).
D. Safety Consequences
This event is not considered an event or condition that could have prevented fulfillment of a safety function. The indications were identified in a timely manner and repaired prior to through-wall leakage occurring. The indications were identified as part of a required periodic inspection. Potentially, if the indications remained undetected, any one could have propagated over time through the alloy 600 weld material to form a leak path through the reactor coolant boundary.
E. Corrective Actions
Immediate Actions Completed P-31
- - This indication was reduced to an acceptable dimension by manual buffing in accordance with ASME Section XI.
P-74, P-76 and P-77
- - These indications were repaired by grinding with no welding required in Accordance with ASME Section Xl.
Corrective Actions
- - Longer Term Byron Station has implemented the Ultra High Pressure Cavitation Peening (UHPCP) process at Byron Station Unit 1 as a mitigating strategy to address the potential for further PWSCC degradation. During B1R21, 100 percent of the VHPs were peened in the inside diameter and the outside diameter.
F. Previous Occurrences
Byron Station, Unit 2. Licensee Event Report (LEA) 455-2007-001-00, Reactor Pressure Vessel Head Control Rod Drive Mechanism Penetration Nozzle Weld Indication Due to an Initial Construction Weld Defect Allowing the Initiation of Primary Water Stress Corrosion Cracking.
(June 8, 2007).
of 4
Byron Station, Unit 1. Licensee Event Report 2011-002-00, Unit 1 Reactor Pressure Vessel Head Penetration Nozzle Weld Flaws Attributed to Primary Water Stress Corrosion Cracking, (May 18, 2011).
Byron Station, Unit 2. Licensee Event Report 2014-004-00, Byron Station Unit 2 Reactor Pressure Vessel Head Control Rod Drive Mechanism Penetration Nozzle Weld Indication attributed to Primary Water Stress Corrosion
- Cracking, (December 5, 2014).
A review of these LERs concluded that these events are similar; however, the causes and corrective actions taken would not have been expected to prevent this event from occurring.
G. Component Failure Data
Manufacturer Nomenclature Model MfQ. Part Number Westinghouse Reactor Vessel Integrated Head 171 8E72 N/A Package TerminationPage j of..4.