05000482/LER-2021-004, Low Steam Generator Level Due to Main Feedwater Valve Failure Caused Automatic Reactor Trip
| ML21291A262 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 10/18/2021 |
| From: | Bayer R Wolf Creek |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| WO 21-0037 LER 2021-004-00 | |
| Download: ML21291A262 (4) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 4822021004R00 - NRC Website | |
text
P.O. Box 411 l Burlington, KS 66839 l 620-364-8831 Robert J. Bayer Plant Manager October 18, 2021 WO 21-0037 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001
Subject:
Docket No. 50-482: Licensee Event Report 2021-004-00, Low Steam Generator Level due to Main Feedwater Valve Failure Caused Automatic Reactor Trip Commissioners and Staff:
The enclosed Licensee Event Report (LER) 2021-004-00 is being submitted pursuant to 10 CFR 50.73(a)(2)(iv)(A) regarding an Engineered Safety Features Actuation and automatic reactor trip at Wolf Creek Generating Station.
This letter contains no commitments. If you have any questions concerning this matter, please contact me at (620) 364-4015, or Ron Benham at (620) 364-4204.
Sincerely, Robert J. Bayer RJB/rlt
Enclosure:
LER 2021-004-00 cc:
S. S. Lee (NRC), w/e S. A. Morris (NRC), w/e N. OKeefe (NRC), w/e Senior Resident Inspector (NRC), w/e
Abstract
Wolf Creek Generating Station 482 3
Low Steam Generator Level due to Main Feedwater Valve Failure Caused Automatic Reactor Trip 08 18 2021 2021 004 00 10 18 2021 1
100%
Ron Benham, Director Nuclear and Regulatory Affairs (620) 364-4204 B
SJ FCV C635 Y
At 1036 Central Daylight Time (CDT) on 8/18/2021, Wolf Creek Generating Station (WCGS) experienced a reactor trip due to low level in the 'B' steam generator. WCGS was operating in MODE 1 at 100% power when the trip occurred. ENS notification#55416 was made at 1251 CDT in accordance with 10 CFR 50.72(b)(2)(iv)(B) due to reactor scram, and 10 CFR 50.72(b)(3)(iv)
(A) for an auxiliary feedwater system actuation. All control rods dropped, all equipment functioned as designed, and offsite power remained available.
The direct cause was a fracture of the valve stem for the 'B' steam generator main feedwater regulating valve, causing the valve to fail closed and resulting in a loss of feed flow control to the B steam generator. The hardware failure anaylsis performed on the valve stem determined that the fracture was due to a fatigue crack which had propagated through the stem of the valve. Tool marks within the thread root, as well as the thread root being cut deeper and narrower were identified as stress risers which allowed the crack to propagate into the material of the stem.
Dye penetrant testing of the replacement stem, and on the existing valve stems of the three other main feedwater regulating valves, showed no relevant indications. Page of 05000-
- 3. LER NUMBER YEAR SEQUENTIAL NUMBER REV NO.
CAUSE
The direct cause was a fatigue crack propagated through the stem of AEFCV0520, causing the stem to fracture, resulting in a loss of feed flow control to the B Steam Generator.
The root cause was tool marks within the thread root caused local stress risers, which allowed multiple cracks to initiate.
In addition, the thread root was also cut deeper and narrower than allowed by specification. This created an additional stress riser which allowed the cracks to propagate into the body of the valve stem.
CORRECTIVE ACTIONS
Actions taken:
AEFCV0520 was disassembled and then reassembled with a replacement valve stem. The replacement valve stem was satisfactorily inspected (no visible indications) using dye penetrant testing prior to installation. Dye penetrant testing of the stems on the other three main feedwater control valves was also performed to look for surface indications of cracking. No relevant indications were identified on any of the other three stems.
Actions planned:
A new Preventive Maintenance (PM) activity will be created to replace the stems on the four MFRVs on a 7.5-year interval. The frequency of 7.5 years for this time-directed stem replacement PM was set based on the life span of the valve stem that failed, and therefore considers a valve stem with sharper than allowed threads and tooling marks. In addition, a methodology and scope for verifying root depth of threads on replacement stems for critical valves will be implemented.
SAFETY SIGNIFICANCE
There were no safety consequences impacting plant or public safety from this event. All control rods dropped, offsite power remained available, and all safety equipment operated appropriately and as designed. There was no loss of any function that would have prevented fulfillment of actions necessary to: shutdown the reactor and maintain it in a safe shutdown condition, remove residual heat, control the release of radioactive material, or mitigate the consequences of an accident.
OPERATING EXPERIENCE/PREVIOUS EVENTS None 3
3 Wolf Creek Generating Station 482 2021 004 00