05000368/LER-2024-001, Surface Flaw in Primary Coolant System
| ML24338A212 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 12/03/2024 |
| From: | Keele R Entergy Operations |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| 2CAN122401 LER 2024-001-00 | |
| Download: ML24338A212 (1) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded |
| 3682024001R00 - NRC Website | |
text
entergy 2CAN122401 December 3, 2024 ATTN: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555-0001 Riley D. Keele, Jr.
Manager, Regulatory Assurance Arkansas Nuclear One Tel 479-858-7826 10 CFR 50.73
Subject:
Licensee Event Report 50-368/2024-001-00, Material Defect in Primary Coolant System that Cannot be Found Acceptable Under American Society of Mechanical EngineersSection XI Arkansas Nuclear One - Unit 2 NRC Docket No. 50-368 Renewed Facility Operating License No. NPF-6 Entergy Operations, Inc. (Entergy) submits the enclosed Licensee Event Report (LER) 50-368/2024-001-00 for Arkansas Nuclear One, Unit 2. This event is reportable in accordance with 1 O CFR 50. 73(a)(2)(ii)(A) as any event or condition that resulted in the condition of the nuclear power plant, including principal safety barriers, being seriously degraded. The LER describes a Reactor Pressure Vessel Closure Head Control Element Drive Mechanism Nozzle In-Service Inspection indication which resulted in a surface material defect in the primary coolant system which did not meet acceptable limits.
This letter contains no new commitments and no revisions to existing commitments.
Should you have any questions concerning this issue, please contact Riley D. Keele Jr.,
Manager, Regulatory Assurance, at 479-858-7826.
Sincerely, RDK/ble
Enclosure:
Licensee Event Report 50-368/2024-001-00 Entergy Operations, Inc. 1448 SR 333, Russellville, AR 72802
2CAN122401 Page 2 of 2 cc:
NRC Region IV Regional Administrator NRC Senior Resident Inspector - Arkansas Nuclear One NRC Project Manager - Arkansas Nuclear One Institute of Nuclear Power Operations (LEREvents@inpo.org)
Enclosure 2CAN122401 Licensee Event Report 50-368/2024-001-00
Abstract
On October 8, 2024, Arkansas Nuclear One, Unit 2 was shut down in Mode 6 for a scheduled refueling outage. The reactor vessel closure head (RVCH) was being examined in accordance with the lnservice Inspection Program. Ultrasonic examination (UT) identified a surface flaw indication that was not acceptable under American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel code requirements. The RVCH Control Element Drive Mechanism (CEDM) Nozzle 71 was found to contain a surface flaw. On October 8, 2024, it was determined that the indication did not meet the acceptable limits as defined in ASME Code Case N-726-6.
On October 21, 2024, a half nozzle repair was performed to correct the condition on CEDM Nozzle 71. The plant remained in Mode 6 during the repairs. This event is reportable in accordance with 10 CFR 50.73(a)(2)(ii)(A) as any event or condition that could result in the condition of the nuclear power plant, including the principle safety barriers, being seriously degraded.
There were no consequences to the general safety of the public, nuclear safety, industrial safety or radiological safety.
PLANT STATUS
- 2. DOCKET NUMBER
- 3. LER NUMBER I
00368 NUMBER NO.
I YEAR SEQUENTIAL REV
~-I 001 1-0 Arkansas Nuclear One, Unit 2 (ANO-2) was shut down for a refueling outage and in Mode 6. There were no other structures, systems, or components that were inoperable at the time that contributed to the event.
EVENT DESCRIPTION
On October 8, 2024, ANO-2 reactor vessel closure head (RVCH) [RPV] was being examined in accordance with the lnservice Inspection Program. Ultrasonic (UT) examination identified a surface flaw indication that was not acceptable under American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel code requirements, the RVCH Control Element Drive Mechanism (CEDM) Nozzle 71 was found to contain a surface flaw, and it was determined that the indication did not meet the acceptable limits as defined in ASME Code Case N-729-6.
Nozzle 71 had an indication identified that breached the threshold of measured growth which would classify it as potential Primary Water Stress Corrosion Cracking (PWSCC). Supplemental examinations were performed on the indication which identified PWSCC as the flaw mechanism.
The UT and bare metal visual (BMV) examinations of Nozzle 71 did not identify any evidence of a leak path through the RVCH above the CEDM J-groove weld nor were there boron deposits on the RVCH outer diameter surface. This demonstrates the indication had not progressed to the J-groove weld triple point resulting in Reactor Coolant System (RCS) [AB] leakage onto the RVCH.
This event was reported on October 8, 2024 at 1431 CDT under 10 CFR 50.72(b)(3)(ii)(A) for degradation of a principal safety barrier (EN 57369).
This event is being reported under 10 CFR 50.73(a)(2)(ii)(A) which requires submittal of a Licensee Event Report within 60 days after the discovery for any event or condition that resulted in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded.
SAFETY ASSESSMENT
There were no consequences to the safety of the general public, nuclear safety, industrial safety, or radiological safety.
The safety significance of the flaw's presence during operation was low. The flaw did not go through the wall of the nozzle. Therefore, there was no leakage of RCS water onto the RVCH. The size of the flaw was small, therefore substantial margin existed until reaching a critical (e.g., unstable) crack size. The ANO-2 RVCH Inspection Program is in accordance with ASME Code Case N-729-6 requirements, as modified by the additional limitations set forth in 10 CFR 50.55a(g)(6)(ii)(D). This provides assurance against any credible PWSCC degradation event that would challenge nuclear safety.
A Materials Reliability Program RVCH Penetration Safety Assessment for U.S. Pressurized Water Reactor Plants (MRP-110) was performed by the industry and was able to demonstrate that there is significant margin against nozzle ejection due to circumferential cracking because of the time required for a circumferential crack to grow to the critical size, which is typically at least 330 degrees. In addition, the safety assessment also demonstrated that periodic BMV examination of the top surface of the RVCH performed at appropriate intervals (each refueling outage for ANO-2) provides assurance against significant wastage of the low-alloy steel head material, even given the assumption of a leaking nozzle.
EVENT CAUSE(S)
- 2. DOCKET NUMBER
- 3. LER NUMBER I
00368 NUMBER NO.
I YEAR SEQUENTIAL REV
~-I 001 1-0 The ANO-2 RVCH is the original RVCH with penetrations containing lnconel Alloy 600 material which is susceptible to PWSCC. PWSCC resulted in an indicated surface flaw in RVCH Nozzle 71, during the required inspection in 2R30.
Until the ANO-2 RVCH is replaced with one that does not contain susceptible material, there is a continued risk of discovering unacceptable indications in future inspections.
CORRECTIVE ACTIONS
A half-nozzle repair was performed to correct the identified condition on CEDM Nozzle 71. The half-nozzle repair involved machining away the lower section of the nozzle which contained the flaw. The remaining portion of the nozzle was welded to the RVCH to form a new pressure boundary. The new weld also attached a replacement lower nozzle that provides a means for re-attaching the guide cone. Examinations required by ASME Code Case N-729-6 performed on Nozzle 71 was documented in ANO2-RR-24-001.
The ANO-2 RVCH is scheduled to be replaced during the refueling outage scheduled in 2026.
PREVIOUS SIMILAR EVENTS
LER-2021-002-0, RVCH Nozzle 46 in 2021. Page 3
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