05000410/LER-2022-001, Regarding Automatic Reactor Scram Due to Low Reactor Water Level During Maintenance

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)

LER-2022-001, Regarding Automatic Reactor Scram Due to Low Reactor Water Level During Maintenance

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(iv)(A), System Actuation
4102022001R00 - NRC Website

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Constellation >>

10 CFR 50.73 NMP2L2812 June 03, 2022

U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Nine M ile Point Nuclear Station, Unit 2 Renewed Facility Operating License No. NPF-69 Docket No. 50 -41 O

Sub ject: 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, Rev ision 0, Automat ic Reactor Scram due to Low Reactor Water Level During Maintenance

There are no regulatory comm itments conta ined in this letter.

Should you have any questions regarding the informa tion in this submittal, please contact Brandon Shultz, Site Regu latory Assurance Manager, at (315) 349 -7012.

Respectfully,

~J-,k__

Alexander Sterio Plant Manager, Nine Mile Po int Nuclea r Station

AS /J A

Enclosure : NMP2 Licensee Event Report 2022 -001, Rev ision 0, Automat ic Reactor Scram due to Low Reactor Water Level During Maintenance

cc : NRC Regional Administrator, Reg ion 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 NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB: NO. 3150-0104 EXPIRES : 08 /31/2023 (08-2020).

Nine Mile Po int Unit 2 05000 410 1 OF 5

4. Title

Automatic Reactor Scram due to Low Reactor Water Level During Maintenance

5. Event Date 6. LER Number 7. Report Date 8. Other Facilities Involved

Month Day Year Number No. Mon t t------+ - - -+----+---+--- - --+-- -+------, f----+-- ---iNIA 05000 NIA 04 05 2022 2022 - 001 00 06 03 2022 Facility Name Docke t Number NIA 05000 NIA

9. Operating Mode 110. Power Level 100

11. This Report is Submitted Pursuant to the Requirements of 1 0 CFR §: (Check all that apply) 10 CFR Part 20 D 20.2203(a)(2)(vi) 50.36(c)(2) 50. 73(a)(2)(iv)(A) 50.73(a)(2)(x)

20.2201(b) D 20.2203(a)(3)(i) 50.46(a)(3)(ii) 50. 73(a)(2)(v)(A) 10 CFR Part 73 20.2201(d) D 20.2203(a)(3)(ii) 50.69(9) 50. 73(a)(2)(v)(B) 73.71{a)(4) 20.2203(a)(1) D 20.2203(a)(4) 73.71(a)(5) 50.73(a)(2)(i)(A) 50.73(a)(2)(v)(C) 20.2203(a)(2)(i) 10 CFR Part 21 50.73(a)(2)(i)(B) 50. 73(a)(2)(v)(D) 73. 77(a)(1 )(i) 20.2203(a)(2)(ii) D 21. 2(c) 50.73(a)(2)(i)(C) 50. 73(a)(2)(vii) 73. 77(a)(2)(i) 20.2203(a)(2)(iii) 10 CFR Part 50 50. 73(a)(2)(ii)(A) 50. 73{a)(2)(viii)(A) 73. 77(a)(2)(ii) 20.2203(a)(2)(iv) 50.36(c)(1 )(i)(A) 50. 73(a)(2)(ii)( B) 50. 73(a)(2)(viii)(B) 20.2203(a)(2)(v) 50.36(c)(1 )(ii)(A) 50.73(a)(2)(iii) 50. 73(a)(2)(ix)(A)

OTHER (Spec ify here, in abstract, or NRC 366A ).

12. Licensee Contact for this LER

licensee Contact Phone Number (Include area code)

Brandon Shultz, Site Regulatory Assurance Manager (315 ) 349-7012

13. Complete One Line for each Component Failure Described in this Report

Cause System Component Manufacturer Reportable to IRIS Cause System Component Manufacturer Reportable to IRIS D NIA NIA NIA y NIA NIA NIA NIA NIA

14. Supplemental Report Expected Month Day Y ear 15. Expected Submission Date 0 ID No Yes (If yes, complete 15. Expected Subm ission Date) NA NA NA

16. Abstract (limit to 1560 spa c es, i.e., appr oxi mately 15 sin gle-spaced ty pewr itten line s)

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 condit ion 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, Attachment 1 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:

IEEE 803 IEEE 805 FUNCTION SYSTEM COMPONENT IDENTIFIER IDENTIFICATION

Feedwater Level Control System NIA JB Feedwater System NIA SJ Reactor Protection System NIA JC High Pressure Coolant Injection System NIA BJ Reactor Core Isolation Cooling NIA BN

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D. SPECIAL COMMENTS:

None