05000316/LER-2022-001, Automatic Reactor Trip Due to Steam Generator High-High Level
| ML23004A205 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 01/04/2023 |
| From: | Ferneau K Indiana Michigan Power Co |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| AEP-NRC-2023-01 LER 2022-001-00 | |
| Download: ML23004A205 (1) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 3162022001R00 - NRC Website | |
text
An MP Company BOUNDLESS ENERGY-January 4, 2023 Docket No.: 50-316 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk 11555 Rockville Pike Rockville, MD 20852 Donald C. Cook Nuclear Plant Unit 2 LICENSEE EVENT REPORT 316/2022-001-00 Indiana Michigan Power Cook Nuclear Plant One Cook Place Bridgman, Ml 49106 indianamichiganpower.com AEP-NRC-2023-01 10 CFR 50.73 Automatic Reactor Trip Due to High-High Steam Generator Level In accordance with 10 CFR 50. 73, Licensee Event Report (LER) System, Indiana Michigan Power Company, the licensee for Donald C. Cook Nuclear Plant Unit 2, is submitting as an enclosure to this letter the following report:
LER 316/2022-001-00: Automatic Reactor Trip Due to High-High Steam Generator Level There are no commitments contained in this submittal.
Should you have any questions, please contact Mr. Michael K.
Scarpello, Regulatory Affairs Director, at (269) 466-2649.
Sincerely, K~£9~
Site Vice President SJM/sjh
Enclosure:
Licensee Event Report 316/2022-001-00: Automatic Reactor Trip Due to High-High Steam Generator Level
U.S. Nuclear Regulatory Commission Page 2 c:
EGLE - RMD/RPS J.B. Giessner-NRC Region Ill M. G. Menze - AEP Ft. Wayne NRC Resident Inspector N. Quilico - MPSC R. M. Sistevaris - AEP Ft. Wayne S. P. Wall-NRC, Washington D.C.
A. J. Williamson - AEP Ft. Wayne AEP-NRC-2023-01
Enclosure to AEP-NRC-2023-01 Licensee Event Report 316/2022-001-00: Automatic Reactor Trip Due to High-High Steam Generator Level
Abstract
On November 10, 2022, at 0744 hours0.00861 days <br />0.207 hours <br />0.00123 weeks <br />2.83092e-4 months <br />, the Donald C. Cook Nuclear Plant Unit 2 Reactor automatically tripped from 44 percent power, due to reaching the automatic trip setpoint for a high-high level in the #3 Steam Generator (SG).
Following the trip, Unit 2 was supplied by offsite power. All control rods fully inserted. The Auxiliary Feedwater Pumps started and operated properly, and decay heat removal was through the Steam Dump System. All required equipment operated as expected, and the trip was not complicated.
The high-high level in the #3 SG was caused by unrecoverable feedwater oscillations that occurred during power ascension.
A preliminary root cause of the oscillations was the organization's failure to recognize the effects of digital positioner tuning on feedwater system stability. A root cause evaluation is ongoing at the time of the submittal of this Licensee Event Report
{LER). A supplement will be submitted if the causes or corrective actions are significantly different than described in this LER.
The event is reportable in accordance with 10 CFR 50.73(a)(2)(iv)(A), System Actuation, due to the valid actuation of the Reactor Protection System and the Auxiliary Feedwater System, as a result of the automatic reactor trip.
~ -
l
EVENT DESCRIPTION
SEQUENTIAL NUMBER 001 REV NO.
00 On November 10, 2022, at 0744 hours0.00861 days <br />0.207 hours <br />0.00123 weeks <br />2.83092e-4 months <br />, the Donald C. Cook Nuclear Plant (CNP) Unit 2 Reactor automatically tripped from 44 percent power, due to reaching a high-high level automatic trip setpoint in 2-OME-3-3, "Number 3 Steam Generator" (#3 SG) [SB][SG].
Following the reactor trip, Unit 2 was supplied by offsite power. All control rods fully inserted. The Auxiliary Feedwater Pumps [BA][P] started as required and operated properly. Decay heat removal was through the Condenser Steam Dump System [Jl][COND]. All required equipment operated as expected, and the trip was not complicated.
Prior to the event, CNP had completed a scheduled refueling outage by synchronizing the generator [EL][GEN] to the electrical grid on November 9, 2022, at 0011 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br />.
On November 10, 2022, at 0515 hours0.00596 days <br />0.143 hours <br />8.515212e-4 weeks <br />1.959575e-4 months <br />, operators commenced power ascension at 1.6 Megawatts per hour, starting from 26% power. At 0522 hours0.00604 days <br />0.145 hours <br />8.630952e-4 weeks <br />1.98621e-4 months <br />, feedwater system oscillations were observed on SG water level indications [JB][L 1] and operators performed actions to place 2-PP-1E, East Main Feed Pump [SJ][P] in manual Differential Pressure (DP) control and raise pump speed.
In response to abnormal SG level indications becoming synchronized, operators began to stabilize power at 44%.
The East Main Feed Pump was placed in speed control and pump speed was raised from 4300 RPM to 4500 RPM. Operators attempted to control SG level by adjusting feedwater regulating valve controllers [JB][FCO]. SG level rapidly increased and actions performed were unable to prevent level from rising to the reactor trip setpoint of 67% in the #3 SG, and automatically tripping the reactor.
Investigation following the trip determined that the feedwater oscillations that occurred during power ascension was due in part to the method used by operators to operate the Main Feed Pump with respect to the digital controls and feedwater system configuration. Configuration of the system is based on 100% power output and two main feed pumps in operation. Validation was obtained by providing pertinent input data to the site simulator team to evaluate the scenario. The simulator results demonstrated that operating one Main Feed Pump in manual DP mode at lower than program DP resulted in unrecoverable feedwater oscillations and SG levels.
Event Notification 56216 was submitted in accordance with 10 CFR 50.72(b)(2)(iv)(B), Reactor Protection System (RPS) [JG] actuation as a four (4) hour non-emergency report, and under 10 CFR 50.72(b)(3)(iv)(A), specified system actuation of the Auxiliary Feedwater System [BA], as an eight (8) hour non-emergency report.
The event is also reportable in accordance wlth 10 CFR 50.73(a)(2)(iv)(A), System Actuation, due to the valid actuation of the RPS and the Auxiliary Feedwater System, as a result of the automatic reactor trip.
COMPONENT There were no component failures that contributed to the event.
CAUSE OF THE EVENT
YEAR 2022
- SEQUENTIAL NUMBER 001 REV NO.
00 The high-high level in the #3 SG was caused by unrecoverable feedwater oscillations that occurred during power ascension. A preliminary root cause of the oscillations was the organization's failure to recognize the effects of digital positioner tuning on Unit 2 feedwater regulating valves which resulted in system instabilities. A root cause evaluation is ongoing at the time of the submittal of this Licensee Event Report (LER). A supplement will be submitted if the causes or corrective actions are significantly different than described in this LER.
ASSESSMENT OF SAFETY CONSEQUENCES
NUCLEAR SAFETY All equipment operated as designed, and there was no actual or potential nuclear safety hazards that resulted from the automatic reactor trip due to reaching the automatic trip setpoint for a high-high level in the #3 SG.
INDUSTRIAL SAFETY There was no actual personnel safety hazard that resulted from the automatic reactor trip due to reaching the automatic trip setpoint for a high-high level in the #3 SG.
RADIOLOGICAL SAFETY There was no actual or potential radiological safety hazard or radiological release that resulted from the automatic reactor trip due to reaching the automatic trip setpoint for a high-high level in the #3 SG.
PROBABILISTIC RISK ASSESSMENT (PRA)
The safety significance of the Unit 2 trip during power ascension can be estimated by calculating the Conditional Core Damage Probability (CCDP) and Conditional Large Early Release Probability (CLERP) of the transient initiating event. A PRA risk assessment was performed and determined the estimated CCDP and the estimated CLERP of this event. A comparison of the CCDP and CLERP estimates to regulatory guidance resulted in the conclusion that this event was of very low safety significance.
CORRECTIVE ACTIONS
COMPLETED ACTIONS YEAR 2022
- SEQUENTIAL NUMBER 001 REV NO.
00 Engineering direction was provided to operate the Unit 2 Main Feed Pumps in speed control and raise the pump speed to achieve a higher DP during power ascension. The current operating procedure used allows for both DP and speed control to operate the main feed pump. Therefore, a procedure change was not required to support the recommendation.
PLANNED ACTIONS Adjust digital tuning of the feedwater regulating valve positioners to improve feedwater system stability.
PREVIOUS SIMILAR EVENTS
A review of LERs for the past five years involving reactor trips was performed. None of the trips occurred from similar causes. Page 4 of 4