05000391/LER-2020-004, Steam Generators Degraded Due to Axial Outside Diameter Stress Corrosion Cracking
| ML21007A022 | |
| Person / Time | |
|---|---|
| Site: | Watts Bar  |
| Issue date: | 01/07/2021 |
| From: | Anthony Williams Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| WBL-20-068 LER 2020-004-00 | |
| Download: ML21007A022 (9) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded |
| 3912020004R00 - NRC Website | |
text
[IE Tennessee Valley Authority, Post Office Box 2000, Spring City, Tennessee 37381 January 7, 2021 WBL-20-068 ATTN : Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Watts Bar Nuclear Plant, Unit 2 Facility Operating License No. NPF-96 NRC Docket No. 50-391 10 CFR 50.73
Subject:
Licensee Event Report 391/2020-004-00, Steam Generators Degraded Due to Axial Outside Diameter Stress Corrosion Cracking This submittal provides Licensee Event Report (LER) 391/2020-004-00. This LER provides details concerning a degraded condition associated with the steam generator tubes at Watts Bar Nuclear Plant (WBN) Unit 2. This condition is being reported as a condition of one of the plant's principal safety barriers being seriously degraded in accordance with 10 CFR 50.73(a)(2)(ii)(A).
There are no regulatory commitments contained in this letter. Please direct any questions concerning this matter to Tony Brown, WBN Licensing Manager, at (423) 365-7720.
Anthony L. Williams IV Site Vice President Watts Bar Nuclear Plant
U.S. Nuclear Regulatory Commission WBL-20-068 Page 2 January 7, 2021
Enclosure:
LER 391/2020-004-00 "Steam Generators Degraded Due to Axial Outside Diameter Stress Corrosion Cracking" cc (Enclosure):
NRG Regional Administrator - Region II NRG Senior Resident Inspector - Watts Bar Nuclear Plant NRG Project Manager - Watts Bar Nuclear Plant
U.S. Nuclear Regulatory Commission WBL-20-068 Page 3 January 7, 2021 MAB: DCB bee (Enclosure):
C. Y. Barajas J. Barstow D. Bost M. A. Brown J. M. Casner C. C. Chandler V. L. Dennis R. E. Detwiler S. M. Douglas K. D. Hulvey B. A. Jenkins J. T. Johnson J. T. Polickoski T. Rausch C. L. Rice M. P. Sokolowich J. R. Staggs A. L. Williams IV ECM
ENCLOSURE Tennessee Valley Authority Watts Bar Nuclear Plant Unit 2 LER 391/2020-004-00 "Steam Generators Degraded Due to Axial Outside Diameter Stress Corrosion Cracking"
Abstract
At 1311 EST on November 11, 2020, it was determined, after evaluation of the Watts Bar Nuclear Plant (WBN)
Unit 2 Steam Generator (SG) tube eddy current test data collected during the ongoing refueling outage, that the WBN Unit 2 Reactor Coolant System pressure boundary did not meet the performance criteria for SG tube structural integrity. Specifically, SG number 3 failed the condition monitoring assessment for conditional burst probability.
The cause of the degradation in the SGs, and particularly SG number 3, is axial outside diameter stress corrosion cracking (ODSCC) of the Alloy 600 mill annealed (MA) SG tubing coincident with the carbon steel tube support plate intersections. Corrective actions taken include plugging, and stabilizing if required, all SG tubes as required by the SG program. Corrective actions to prevent recurrence include a planned mid-cycle SG inspection and steam generator replacement.
This event is being reported to the Nuclear Regulatory Commission (NRG) under 10 CFR 50.73(a)(2)(ii)(A) as a condition that resulted in the plant's principal safety barriers being seriously degraded.
I.
Plant Operating Conditions Before the Event
Watts Bar Nuclear Plant (WBN) Unit 2 was in Mode 5 at O percent rated thermal power (RTP).
II.
Description of Event
A. Event Summary At 1311 EST on November 11, 2020, it was determined, after evaluation of the Watts Bar Nuclear Plant (WBN) Unit 2 Steam Generator (SG)[EIIS:HX] tube eddy current test data collected during the U2R3 refueling outage, that the WBN Unit 2 Reactor Coolant System (RCS)[EIIS:AB] pressure boundary did not meet the performance criteria for SG tube [EIIS:TBG] structural integrity. Specifically, SG number 3 failed the condition monitoring assessment for conditional burst probability in accordance with Generic Letter (GL) 95-05, "Voltage Based Repair Criteria." The conditional probability of burst limit in GL 95-05 is 1 E-02; however, this limit was exceeded for SG 3. WBN has completed tube plugging and additional corrective actions are in progress.
This event is being reported to the Nuclear Regulatory Commission (NRC) under 10 CFR 50. 73(a)(2)(ii)(A) as a condition that resulted in the plant's principal safety barriers being seriously degraded.
B. Status of structures, components, or systems that were inoperable at the start of the event and that contributed to the event No inoperable structures, systems, or components contributed to this condition.
C. Dates and approximate times of occurrences
Date 11/11/20 11/11/20 Time (EST) 1311 1611 Event Received preliminary condition monitoring report that indicated SG#3 failed accident induced conditional burst probability..
Event Notification 54994 made to NRC.
D. Manufacturer and model number of each component that failed during the event
The degraded steam generators are Model D3 SGs manufactured by Westinghouse.
E. Other systems or secondary functions affected
No other systems or secondary functions were affected. Page 2 of 5 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB: NO. 3150-0104 EXPIRES: 08/31/2023 (08-2020)
LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET
- 3. LERNUMBER YEAR Watts Bar Nuclear Plant, Unit 2 05000-391 2020 SEQUENTIAL NUMBER
- - 004
F. Method of discovery of each component or system failure or procedural error
The degraded condition of the steam generators was discovered during periodic eddy current testing (ECT) of the steam generator tubes.
G. Failure mode, mechanism, and effect of each failed component REV NO.
- - 00 The failure mechanism of the SG tubes that resulted in failing to meet condition monitoring is axial outside diameter stress corrosion cracking (ODSCC) of the SG tubing at the carbon steel tube support plate (TSP) intersections.
H. Operator actions
No operator actions were required.
I.
Automatically and manually initiated safety system responses
None required.
Ill.
Cause of the Event
A. Cause of each component or system failure or personnel error The degradation mechanism leading to the equipment failure event is axial ODSCC of the Alloy 600 mill annealed (MA) SG tubing coincident with carbon steel TSP intersections. The potential for this degradation to occur in Alloy 600MA tubing has been widely documented through industry operating experience (OE). Although the degradation mechanism was an expected occurrence at Watts Bar Unit 2, the growth rate identifed was greater than projected. Operating temperatures of the tube material are elevated at WBN Unit 2 and coupled with localized crevice chemistry at the TSP intersections creates an undesirable condition that leads to initiation and growth of ODSCC.
No human performance root causes are associated with this event.
IV.
Analysis of the Event
WBN Unit 2 is a four loop Pressurized Water Reactor (PWR) provided by Westinghouse. The SGs installed at WBN Unit 2 are Westinghouse Model D3 SGs with Alloy 600 MA tubing. This particular SG Model and tubing has a history of tube degradation that requires them to be replaced relatively early in plant life. Replacement SGs have already been procured for WBN Unit 2 and they were scheduled for replacement during the fifth refueling outage for WBN Unit 2. Page 3 of 5 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB: NO. 3150-0104 EXPIRES: 08/31 /2023 (08-2020)
- 3. LERNUMBER YEAR Watts Bar Nuclear Plant, Unit 2 05000-391 2020 SEQUENTIAL NUMBER
- - 004 REV 1110.
- - 00 During the WBN Unit 2 third refueling outage, a 100 percent eddy current inspection of all four SGs was performed. This inspection identified higher than projected degradation from axial ODSCC of the SG tubing coincident with carbon steel TSP intersections. The results show that the End-of-Cycle 3 condition monitoring SG primary-to-secondary leak rates and conditional probabilities of tube burst for SG1, SG2 and SG4 are well within their respective allowable accident analysis limits of 3 gallons per minute (gpm)/SG and 1 E-02. The SG3 condition monitoring primary-to-secondary leak rate is also within the allowable limit, however, the conditional probability of tube burst for SG3 was calculated to be 3.0050E-02 which exceeds the limit. The conditional probability of burst refers to the probability that the SG tube burst pressure associated with one or more flaw indications in the faulted SG will be less than the maximum pressure differential associated with a postulated Main Steam Line Break assumed to have occurred over the prior operating interval.
V.
Assessment of Safety Consequences
An assessment of safety significance using the guidance of GL 95-05 was performed for the as found condition of the SGs. This assessment considered the probability of a MSLB, the probability of a SG TSP being displaced, and the probability of a tube rupture from the TSP displacement.
This resulted in a change in the large early release frequency (LERF) of much less than 1 E-7 per year, which is very small.
A. Availability of systems or components that could have performed the same function as the components and systems that failed during the event Not applicable.
B. For events that occurred when the reactor was shut down, availability of systems or components needed to shutdown the reactor and maintain safe shutdown conditions, remove residual heat, control the release of radioactive material, or mitigate the consequences of an accident Not applicable.
C. For failure that rendered a train of a safety system inoperable, an estimate of the elapsed time from the discovery of the failure until the train was returned to service Not applicable.
VI.
Corrective Actions
This event was entered into the Tennessee Valley Authority's (TVA) Corrective Action Program and is being tracked under Condition Report (CR) 1651444. Page 4 of 5 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0 MB: NO. 3150-0104 EXPIRES: 08/31/2023 (08-2020)
LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET
- 3. LERNUMBER Watts Bar Nuclear Plant, Unit 2 05000-391
A. Immediate Corrective Actions
YEAR 2020 SEQUENTIAL NUMBER
- - 004 REV NO.
- - 00 Upon identifying the condition, a _condition report was generated and a report was made to the NRC. SG tubes were plugged, and stabilized if required, prior to returning the plant to MODE 4 in accordance with program requirements.
B. Corrective Actions to Prevent Recurrence or to reduce probability of similar events occurring in the future As a result of the identified condition, a mid-cycle inspection of SG tubes will be performed at WBN Unit 2. Actions are in progress to move up the planned replacement of the WBN Unit 2 steam generators.
VII.
Previous Similar Events at the Same Site
No previous similar events have been reported at WBN.
VIII.
Additional Information
There is no additional information.
IX.
Commitments
There are no new commitments. Page 5 of 5