05000455/LER-2022-001-01, Volumetric Examinations of Reactor Pressure Vessel Head Core Exit Thermocouple Penetration P-75 Identified an Indication Attributed to Primary Water Stress Corrosion Cracking
| ML23243A934 | |
| Person / Time | |
|---|---|
| Site: | Byron  (NPF-066) |
| Issue date: | 08/31/2023 |
| From: | Welt H Constellation Energy Generation |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| 2023-0042 LER 2022-001-01 | |
| Download: ML23243A934 (1) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded |
| 4552022001R01 - NRC Website | |
text
Constellation.,,
August 31, 2023 L TR:
BYRON 2023-0042 File:
1D.101 5A.108 United States Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Byron Station, Unit 2 Renewed Facility Operating License No. NPF-66 NRC Docket No. STN 50-455 Constellation Energy Generation, LLC (CEG; Byron Station 4450 N. German Church Road Byron, IL 61010-9794 www.constellationenergy.com 10CFR50.73
Subject:
Licensee Event Report (LER) Supplement No. 455-2022-001-01 "Byron Station Unit 2 Volumetric Examinations of Reactor Pressure Vessel Head Core Exit Thermocouple Penetration P-75 Identified an Indication Attributed to Primary Water Stress Corrosion Cracking" Enclosed is Byron Station Licensee Event Report (LER) Supplement No. 455-2022-001-01 regarding volumetric examinations of the Reactor Vessel Head Core Exit Thermocouple Penetration P-75 identified a recordable indication that did not meet the applicable acceptance criteria on Byron Unit 2.
This condition is being submitted in accordance with 10 CFR 50.73, "Licensee Event Report System."
There are no regulatory commitments in this report.
Should you have any questions concerning this submittal, please contact Ms. Zoe Cox, Regulatory Assurance Manager, at (815) 406-2800.
Respectfully, 1/lJ)tr Harris Welt Site Vice President Byron Generating Station HW/ZC/hh
Enclosure:
LER 455-2022-001-01 cc:
Regional Administrator-NRC Region Ill NRC Senior Resident Inspector-Byron Generating Station
Abstract
During the Byron Station, Unit 2, spring 2022, refueling outage, volumetric examinations of the Reactor Vessel Head Core Exit Thermocouple Penetration P-75 identified a recordable indication that did not meet the applicable acceptance criteria.
The cause of the P-75 unacceptable indication is attributed to Primary Water Stress Corrosion Cracking.
A flaw growth analysis determined the indication was acceptable for continued operation for two refueling cycles under ASME Code Case N-729-6 ASME Section XI requirements. The indication will be repaired within the next two refuel cycles.
This event 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.
A. Plant Operating Conditions Before the Event
Event Date:
April 23, 2022 Unit: 2 MODE: 6 (Refueling)
Unit 2 Reactor Coolant System (RCS) [AB]:
YEAR 05000455 2022 Reactor Power: 000 percent SEQUENTIAL NUMBER 001 Ambient Temperature and Depressurized No structures, systems or components were inoperable at the start of this event that contributed to the event
B. Description of Event
REV NO.
01 At 08:53 on April 23, 2022, during the Byron Station, Unit 2, spring 2022 (B2R23) refueling outage, volumetric examinations of the Reactor Vessel Head Core Exit Thermocouple (CETC) Penetration P-75 identified a recordable indication that did not meet the applicable acceptance criteria. No additional indications were found, and the extent of condition was limited to Penetration 75.
During the prior Byron Station B2R19 and B2R20 refueling outages in spring 2016 and fall 2017, the Reactor Pressure Vessel Head Penetration Nozzles were mitigated from Primary Water Stress Corrosion Cracking (PWSCC) using the Ultra-High Pressure Cavitation Peening (UH PCP) process in accordance with the requirements of MRP-335, Revision 3-A. In response to a relief request submitted on August 30, 2018 (ADAMS Accession No. ML18248A060), requesting approval for alternative follow-up inspections of peening-applied reactor vessel head penetration nozzles in accordance with ASME Code Case N-729-4 for the fourth inservice inspection (ISi) interval of Byron Station, Unit 2, the Nuclear Regulatory Commission (NRC) authorized the proposed alternative on September 19, 2017 (ADAMS Accession No. ML19035A294). 10 CFR 50.55a, "Codes and standards," paragraph (g)(6)(ii)(D) has since been updated such that holders of operating licenses or combined licenses for pressurized-water reactor as of or after June 3, 2020, shall implement the requirements of ASME BPV Code Case N-729-6 instead of ASME BPV Code Case N-729-4, subject to the listed conditions, by no later than one year after June 3, 2020.
The indication at P-75 was found outside the required peening coverage area. Peening had not been performed on this area due to geometry limitations where the funnel fillet welds, and the adjacent areas are shadowed by top of the funnel. This indication is located at 184 degrees with a length of 0.197" with a depth of 0.141" from the outer diameter (OD) surface of the CETC penetration. The indication extends from 1.498" to 1.695" from the end of the nozzle. The indication is axially oriented and is at the location of one of the funnel fillet welds.
The indication is in the CETC penetration nozzle itself and not the weld associated with the anti-rotation weld for the funnel. Although this portion of the nozzle is not part of the pressure retaining boundary, it does fall within the jurisdiction of ASME Section XI and is Class 1 safety related. Byron Station completed an analysis to determine growth of the flaw in accordance with ASME Code Case N-729-6 and ASME Section XI, 2007 Edition through the 2008 Addenda. The analysis documents that the Unit 2 reactor vessel head remains operable for two fuel cycles (until the spring 2025 (B2R25) refueling outage).
This LER is being submitted in follow-up to ENS 55857 made on April 23, 2022, in accordance with 10 CFR 50. 72(b)(3)(ii)(A).
C. Cause of Event
YEAR 2022 SEQUENTIAL NUMBER 001 The cause of the P-75 unacceptable indication is attributed to PWSCC, based on the volumetric examination characterization of the indication.
D. Safety Consequences
REV NO.
01 This condition had no actual safety consequences impacting plant or public safety. This event is not considered an event or condition that could have prevented fulfillment of a safety function.
The indication is in the CETC penetration nozzle itself and not the weld associated with the anti-rotation weld for the funnel. Although this portion of the nozzle is not part of the pressure retaining boundary, it does fall within the jurisdiction of ASME Section XI and is Class 1 safety related. Byron Station has completed an analysis to determine growth of the flaw in accordance with ASME Code Case N-729-6 and ASME Section XI, 2007 Edition through the 2008 Addenda. The analysis documents that the Unit 2 reactor vessel head remains operable for two fuel cycles (until B2R25). The remaining CETC penetrations have been analyzed and reviewed during the spring 2022 refueling outage (B2R22), and no other indications were found.
Based on the B2R23 documented characteristics and dimensions of the observed volumetric indications, there was no Safety Significant Functional Failure (i.e., loss of safety function) resulting from this indication. The primary coolant pressure boundary was maintained and capable of performing its design function.
E. Corrective Actions
A flaw growth analysis determined the indication was acceptable for continued operation for two refueling cycles under ASME Code Case N-729-6 ASME Section XI requirements. Since MRP-335 Rev 3-A and ASME Code Case N-729-6 are silent on flaws found outside the required peening area, it needs to be assumed that the entire nozzle is flawed. Therefore, the volumetric re-inspection frequency can be summarized as either:
With Repair - Every other refueling outage after B2R23 until the flaw is repaired with a post volumetric exam the outage immediately following the repair.
Without Repair (Flaw Evaluation) - Every other refueling outage after B2R23 until the flaw is repaired.
NRC condition 5.2(d) on the SER of MRP-335 Rev 3-A points to the requirements of 10 CFR 50.55a if a flaw is found in a peened nozzle. In Byron relief request 14R-17 Rev. 0 (Reference 2.5.3.11), unmitigated nozzles were classified as N-729-6 items 84.20 prior to implementing the inspection relaxation of MRP 335-Rev 3-A. Since the inspection relaxation of MRP-335 Rev 3-A cannot be applied due to the flaw found, Penetration 75 stays classified as N-729-6 item 84.20 where the requirements of 10 CFR 50.55a (g)(6)(ii)(D) shall apply.
A more refined flaw growth analysis has been performed and a re-examination is required of the flaw in accordance with ASME Boiler and Pressure Vessel Code,Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components", Code Case N-729-6, Relief Request I4R-14, and MRP-335 Rev. 3-A.
F. Previous Occurrences
SEQUENTIAL NUMBER 001 REV NO.
01 No previous, similar Licensee Event Reports were identified at the Byron Station in the past three years.
G. Component Failure Data
Manufacturer Westinghouse Nomenclature Reactor Vessel Integrated Head Package Termination Model 1718E72 Mfg. Part Number N/A Page_4_ of _4_