05000323/LER-2022-001, Reactor Coolant System Pressure Boundary Degradation
| ML22355A081 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 12/21/2022 |
| From: | Petersen D Pacific Gas & Electric Co |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| DCL-22-093 LER 2022-001-00 | |
| Download: ML22355A081 (1) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded |
| 3232022001R00 - NRC Website | |
text
Pacific Gas and Electric Company..
PG&E Letter DCL-22-093 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Docket No. 50-323, OL-DPR-82 Diablo Canyon Power Plant, Unit 2 Dennis B. Petersen Station Director Diablo Canyon Power Plant Mail code 104/5/502 P.O. Box 56 Avila Beach, CA 93424 805.545.4022 Dennis.Petersen@pge.com 10 CFR 50.73 Unit 2 Licensee Event Report 2022-001-00, Unit 2 Reactor Coolant System Pressure Boundary Degradation
Dear Commissioners and Staff,
In accordance with the requirements of 10 CFR 50.73(a)(2)(ii)(A), Pacific Gas and Electric Company (PG&E) hereby submits the enclosed Diablo Canyon Power Plant (DCPP) Unit 2 Licensee Event Report regarding a reactor coolant system boundary degradation related to a through-wall leak in a socket weld.
PG&E makes no new or revised regulatory commitments (as defined by NEI 99-04) in this report. All corrective actions identified in this letter will be implemented in accordance with the DCPP Corrective Action Program.
This event did not adversely affect the health and safety of the public.
If you have any questions or require additional information, please contact Mr. James Morris, Regulatory Services Manager, at (805) 545-4609.
Sincerely, Dennis B. Petersen jmsp/51168277-12 Enclosure cc/enc:
Mahdi 0. Hayes, NRC Senior Resident Inspector Samson S. Lee, NRR Senior Project Manager Scott A. Morris, NRC Region IV Administrator INPO Diablo Distribution A
member of t he STA~
Alliance Cal l away *
- Diablo Canyon
- Palo Verde
- Wol f Creek Date
Abstract
Diablo Canyon Power Plant, Unit 2 00323 3
Unit 2 Reactor Coolant System Pressure Boundary Degradation 10 23 2022 2022 001 00 12 21 2022 6
000
David Madsen, Licensing Engineer 805-545-6192
At 0830 PDT on 10/23/2022, during routine outage inspections on Unit 2, it was determined that the reactor coolant pressure boundary did not meet American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC)Section XI acceptance criteria due to finding a through-wall indication at a 2-inch stainless steel socket weld (Weld No. WIB-975D) on the cold leg Loop 1 vacuum refill connecting piping (Line No. 1140), and was therefore reportable.
This event is being reported per 10 CFR 50.73(a)(2)(ii)(A) as a degraded condition.
The presumed cause of the degradation was vibration-induced fatigue propagation of a flaw initiated at a weld defect. Corrective actions to address the condition consisted of performing a weld repair in accordance with ASME BPVC Section XI Case N-666-1 during the refueling outage.
There was no impact to the health and safety of the public or plant personnel.
X AB PSF Yes
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C. Status of Inoperable Structures, Systems, or Components (SSCs) that Contributed to the Event There were no SSCs that were inoperable at the start of the event that contributed to the Event.
D. Method of Discovery
This condition was discovered during a routine piping and valve walkdown as part of the stations boric acid corrosion control program.
E. Operator Actions
None were required. The condition was discovered during routine walkdowns during plant refueling. There was no active leak from the weld observed at the time of discovery.
F. Safety System Responses None required.
IV. Cause of the Problem The direct cause of the indication was vibration-induced fatigue propagation initiated at a weld flaw.
V. Assessment of Safety Consequences
The condition did not adversely affect the health and safety of the public or on-site personnel. The boric acid residue found was indicative of a through-wall leak and active leakage was not observed at the time the condition was identified in Mode 6. Calculated total RCS leakage rates for the previous Unit 2 operating cycle were negligible, indicating that any through-wall leakage was minimal. Additionally, the weld indication was localized in nature with no other accompanying indications.
VI. Corrective Actions
The weld was repaired in accordance with ASME BPVC Section XI Code Case N-666-1. The weld repair was completed on 10/31/2022 and passed required inspections. This was the only connecting piping of this configuration on the remaining RCS loops. Finalization of the cause evaluation and implementation of corrective actions will be managed in accordance with the DCPP Corrective Action Program.
VII. Additional Information
There have been no similar events at DCPP in the previous three years.
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