05000454/LER-2017-001

LER-2017-001, 1 OF 4

Byron Station - Unit 1
Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded
Initial Reporting
ENS 52591	10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded
4542017001R00 - NRC Website

LER 17-001-00 for Byron, Unit 1, 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 E Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
BYRON 2017-0033 LER 17-001-00
Download: ML17115A068 (5)

comments regarding burden estimate to the Information Services Branch (1--2 F43). U.S. Nuclear Regulatory Commission, Washington, DC 20555.0001 or by e-mail to Infccollects.Resource@nrc.gov and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202. (3150-0104), Office of Management and Budget Washington, DC 20503. If a means used to impose an informaton collection does not dsplay a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to the information collection.

3. LER NUMBER

2017 - 00 001

A. Plant Operating Conditions Before the Event:

Event Dates: March 3, 2017, March 5, 2017, and March 6, 2017 Unit: 1 Mode: 6 (defueled) Reactor Power: 0 percent Unit 1 Reactor Coolant System (RCS) [AB]:

Ambient operating temperature and pressure No structures, systems or components were inoperable at the start of this event that contributed to the event.

B. Description of Event:

During the Byron Station, Unit 1, spring 2017, refueling outage (B1R21), NDE personnel performed volumetric and surface examinations of the Reactor Vessel Head Penetration (VHP) nozzles in accordance ASME Code Case N729- 1, as modified by 10 CFR 50.55a(g)(6)(ii)(D). Examination results identified recordable indications for VHP nozzles 31, 74, 76, and 77 that did not meet the applicable acceptance criteria. Each of the affected VHPs is a Core Exit Thermocouple nozzle. None of the unacceptable indications were located in the Reactor Coolant System pressure boundary region.

March 3, 2017:

VHP 31 nozzle had one unacceptable indication identified by surface (PT) examination on an existing Embedded Flaw repaired weld. The unacceptable indication was 7/32 inch compared to the acceptance criteria of 3/16 inch. The indication was located approximately 3 inches from the interface of the nozzle and J-Groove weld.

March 5, 2017:

VHP 76 nozzle had two unacceptable indications in the tube base material identified by volumetric (UT) examination.

The first unacceptable indication was 0.231 inches long and 0.135 inches deep (22 percent through wall) from the nozzle OD surface. The indication was 1.945 inches to 2.18 inches from the end of the nozzle. The second unacceptable indication was 0.316 inches long and 0.153 inches deep (25 percent through wall) from the nozzle OD surface. The indication was located 1.949 inches to 2.265 inches from the end of the nozzle.

March 6, 2017:

VHP 76 nozzle had two additional unacceptable indications in the tube base material identified by volumetric (UT) examination. The third unacceptable indication was 0.354 inches long and 0.179 inches deep (29 percent through wall) from the nozzle OD surface. The indication was located 1.544 inches to 1.898 inches from the end of the nozzle.

The fourth unacceptable indication was 0.355 inches long and 0.130 inches deep (21 percent through wall) from the nozzle OD surface. The indication was located 1.544 inches to 1.898 inches from the end of the nozzle.

VHP 74 nozzle had two unacceptable indications in the tube base material identified by volumetric (UT) examination.

The first unacceptable indication was 0.394 inches long and 0. 133 inches deep (21 percent through wall) from the nozzle OD surface (through wall dimension is 21 percent through wall). The indication was located 1.822 inches to 2.216 inches from the end of the nozzle. The second unacceptable indication was 0.394 inches long and 0.128 inches deep (20 percent through wall) from the nozzle OD surface. The indication was located 1.507 inches to 1.901 inches from the end of the nozzle.

comments regarding burden estimate to the Information Services Branch (T-2 F43). U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to NEOB-10202. (3150-0104). Office of Management and Budget, Washington. DC 20503. If a means used to impose an information collection does not display a currently valid OMB control number. the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

3. LER NUMBER

2017 - 00 001 VHP 77 nozzle had 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 Xl.

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 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 (LER) 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).

comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S. Nuclear Regulatory Commission. Washington. DC 20555-0001, or by e-mail to Infocollects.Resource@nrc gov, and to the Desk Officer, Office of Information and Regulatory Affairs, used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection

3. LER NUMBER

2017 - 00 001 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 Mfg. Part Number Westinghouse Reactor Vessel Integrated Head 1718E72 N/A Package Termination