05000410/LER-2022-001, Regarding Automatic Reactor Scram Due to Low Reactor Water Level During Maintenance
| ML22159A196 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 06/03/2022 |
| From: | Sterio A Constellation Energy Corp |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NMP2L2812 LER 2022-001-00 | |
| Download: ML22159A196 (7) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 4102022001R00 - NRC Website | |
text
,
NMP2L2812 June 03, 2022 Constellation>>
U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Nine Mile Point Nuclear Station, Unit 2 Renewed Facility Operating License No. NPF-69 Docket No. 50-41 O 10 CFR 50.73
Subject:
NMP2 Licensee Event Report 2022-001, Revision 0, Automatic Reactor Scram due to Low Reactor Water Level During Maintenance In accordance with the reporting requirements contained in 10 CFR 50.73(a)(2)(v)(B), please find enclosed NMP2 Licensee Event Report (LER) 2022-001, Revision 0, Automatic Reactor Scram due to Low Reactor Water Level During Maintenance There are no regulatory commitments contained in this letter.
Should you have any questions regarding the information in this submittal, please contact Brandon Shultz, Site Regulatory Assurance Manager, at (315) 349-7012.
Respectfully,
~J-,k__
Alexander Sterio Plant Manager, Nine Mile Point Nuclear Station AS/JA
Enclosure:
NMP2 Licensee Event Report 2022-001, Revision 0, Automatic Reactor Scram due to Low Reactor Water Level During Maintenance cc:
NRC Regional Administrator, Region I NRC Resident Inspector NRC Project Manager
Enclosure NMP2 Licensee Event 2022-001, Revision 0 Automatic Reactor Scram due to Low Reactor Water Level During Maintenance Nine Mile Point Nuclear Station, Unit 2 Renewed Facility Operating License No. NPF-69
Abstract
On Thursday, April 5, 2022, at approximately 0222 hours0.00257 days <br />0.0617 hours <br />3.670635e-4 weeks <br />8.4471e-5 months <br /> with power level at approximately 100 percent, Nine Mile Point Unit 2 (NMP2) experienced a feedwater level control transient during maintenance on the Feedwater Flow Control Valve (LV1 OB),
which resulted in a low reactor water level scram. This event is reportable per 1 0CFR 50. 73(a)(2)(iv)(A) as any event or condition that resulted in a manual or automatic actuation of any of the systems listed in 1 0CFR50. 73(a)(2)(iv)(B). Following the reactor scram, all plant systems responded per design.
The root cause of this event was incorrect procedure guidance for nulling the demand and position signal in N2-SOP-06, "Feedwater Failures, Attachment 1, Restoration of 2FWS-LV10 Control". Corrective actions included revising N2-SOP-06, to correctly null the demand signal to the potentiometer intended for control.
I.
DESCRIPTION OF EVENT
A. PRE-EVENT PLANT CONDITIONS:
Prior to the event, NMP2 was in Mode 1, operating at approximately 100 percent rated thermal power. Feedwater level control was controlling in automatic with Feedwater Level Control Valve LV1 0A controlling in AUTO on the backup position sensor due to the primary position sensor being non-functional. Feedwater Level Control Valve LV1 OB was controlling in MANUAL with both primary and backup position sensors non-functional.
Feedwater flow control is provided by a flow control valve in the discharge line of each reactor feed pump. The feedwater flow control valves may be controlled either in automatic or manual. Each feedwater flow control valve is provided with a primary and a secondary sensor that provides position feedback to the feedwater level control system.
B. EVENT:
On Thursday, April 5, 2022, at appr9ximately 0222 hours0.00257 days <br />0.0617 hours <br />3.670635e-4 weeks <br />8.4471e-5 months <br />, NMP2 experienced a reactor scram due to reactor water low level while performing maintenance on Feedwater Level Control Valve LV10B.
The potentiometer of the primary position sensor was being replaced on Feedwater Level Control Valve LV1 OB.
Feedwater Level Control Valve LV1 OB was placed in a lock-up condition to support replacement of the potentiometer primary position sensor to allow Operations to regain automatic valve functionality. After the potentiometer was insta'lled, Operations attempted to restore Feedwater Level Control Valve LV1 OB to the primary position using N2-SOP-06, Feedwater Failures, Attachment 1. Upon restoring Feedwater Level Control Valve LV1 OB primary positioner, the position feedback signal was mismatched (80% open and demanded position was 43%), causing Feedwater Level Control Valve LV10B to rapidly close which subsequently caused reactor water level to lower. Feedwater Level Control Valve LV10A attempted to open to maintain reactor water level but was unsuccessful, resulting in a low reactor water level scram. High Pressure Core Spray and Reactor Core Isolation Cooling systems actuated as designed to maintain level, and plant personnel successfully recovered feedwater level control.
The condition was reported to the NRC on April 5, 2022 at approximately 0608 hours0.00704 days <br />0.169 hours <br />0.00101 weeks <br />2.31344e-4 months <br /> pursuant to the requirements.of 1 0CFR50. 72(b )(2)(iv)(A), 50. 72(b )(2)(iv)(B), and 50. 72(b)(3)(iv)(A) (Event Notification#55821)
Nine Mile Point Unit 1 (NMP1) was unaffected.
The event has been entered into the plant's corrective action program as IR 04490223.
C. INOPERABLE STRUCTURES, COMPONENTS, OR SYSTEMS THAT CONTRIBUTED TO THE EVENT:
No other systems, structures, or components contributed to this event.
D. DATES AND APPROXIMATE TIMES OF MAJOR OCCURRENCES AND OPERATOR ACTIONS:
The dates, times, major occurrences, and operator actions for this event are as follows.
April 2, 2022:
2249 - The primary position feedback for Feedwater Level Control Valve LV1 OB failed April 5, 2022:
0000 - Shortly after midnight, Instrument Maintenance Department (IMD) began the prejob brief and maintenance activity for replacement of the Feedwater Level Control Valve LV1 OB primary potentiometer 0137 - Operations unlocked Feedwater Level Control Valve LV10B by taking the S10B switch to NORMAL to allow IMO technicians to get initial position readings. Once readings were captured and documented, Operations placed the valve back into the lock-up condition by moving the S10B switch to reset per N2-SOP-06. IMD then matched the replacement potentiometer resistance to the originally installed potentiometer and installed the new primary potentiometer.
0222 - After completion of the primary potentiometer replacement and adjustment, Operations repositioned the control switch to NORMAL, resetting the lockup condition of the valve and returning Feedwater Level Control Valve LV1 OB to service. Upon returning to service, the feedwater flow control system erroneously read the Feedwater Level Control Valve LV1 OB position at approximately 800/o versus the required demand signal of 43% causing it to drive the valve closed. Feedwater Level Control Valve LV1 DA responded however, reactor water level lowered and scrammed on Low Water Reactor Water Level.
0224 - Turbine tripped
E. METHOD OF DISCOVERY
This event was discovered by Reactor Operators when feedwater flow control valves began moving and Reactor scram annunciation was received.
F. SAFETY SYSTEM RESPONSES:
All safety systems responded per design.
11..
CAUSE OF THE EVENT
The root cause of this event was determined to be incorrect procedure guidance for nulling the demand and position signal in procedure N2-SOP-06, Feedwater Failures, Attachment 1, Restoration of 2FWS-LV10 Control. When the NORMAL/RESET switch (S10B) is placed to reset, the signal sentto the MIA (manual-automatic) station is from the non-selected position feedback potentiometer and not the potentiometer selected with the corresponding Feedback Potentiometer selector switch (S2B). This procedure error was introduced when the modification was implemented and not recognized.
Ill.
ANALYSIS OF THE EVENT
The automatic reactor scram is reportable per 10CFR 50.72(b)(2)(iv)(B) and 10CFR 50.73(a)(2)(iv)(A). It is defined under paragraph 1 0CFR 50. 73(a)(2)(iv)(A) as any event or condition that resulted in a manual or automatic actuation of any of the systems listed in 1 0CFR50:73(a)(2)(iv)(B).
Loss of feedwater transient events are considered in the design basis of the plant with multiple automatic and manual recovery paths. The High Pressure Core Spray and Reactor Core Isolation Cooling systems actuated as designed to maintain level, and plant personnel successfully recovered feedwater control shortly after the trip. It is judged that the safety significance of this event is low, and the event did not pose a threat to the health and safety of the public or plant personnel. All other plant systems performed per design without complications.
IV.
CORRECTIVE ACTIONS
A. ACTION TAKEN TO RETURN AFFECTED SYSTEMS TO PRE-EVENT NORMAL STATUS:
A riew model of potentiometers were installed on Feedwater Level Control Valve LV1 0A and LV1 OB.
B. ACTION TAKEN OR PLANNED TO PREVENT OCCURRENCE:
Revised N2-SOP.:Q6, Attachment 1 to correctly null the demand signal to the potentiometer intended for control.
V.
ADDITIONAL INFORMATION
A. FAILED COMPONENTS:
No component failures were associated with this event. An IRIS entry was completed for the event. (#525435)
B. PREVIOUS LERs ON SIMILAR EVENTS:
None.
C. THE ENERGY INDUSTRY IDENTIFICATION SYSTEMS (EIIS) COMPONENT FUNCTION IDENTIFIER AND SYSTEM NAME OF EACH COMPONENT OR SYSTEM REFERRED TO IN THIS LER:
COMPONENT Feedwater Level Control System Feedwater System Reactor Protection System High Pressure Coolant Injection System Reactor Core Isolation Cooling
/
D. SPECIAL COMMENTS:
None IEEE 803 FUNCTION IDENTIFIER NIA NIA NIA NIA NIA IEEE 805 SYSTEM IDENTIFICATION JB SJ JC BJ BN Page 5
of 5