05000327/LER-2019-003, Automatic Reactor Trip Due to Negative Rate Trip as a Result of a Dropped Control Rod

Automatic Reactor Trip Due to Negative Rate Trip as a Result of a Dropped Control Rod

ML19297G404
Person / Time
Site:	Sequoyah
Issue date:	10/23/2019
From:	Rasmussen M Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LER 2019-003-00
Download: ML19297G404 (8)

LER-2019-003, Automatic Reactor Trip Due to Negative Rate Trip as a Result of a Dropped Control Rod

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(v), Loss of Safety Function
3272019003R00 - NRC Website

text

Tennessee Valley Authority, Sequoyah Nuclear Plant, P.O. Box 2000, Soddy Daisy, Tennessee 37384 October 23, 2019 10 CFR 50.73 ATTN:

Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Sequoyah Nuclear Plant, Unit 1 Renewed Facility Operating License No. DPR-77 NRC Docket No. 50-327

Subject:

Licensee Event Report 50-327/2019-003-00, Automatic Reactor Trip due to Negative Rate Trip as a Result of a Dropped Control Rod The enclosed licensee event report provides details concerning an automatic reactor trip following a dropped control rod and an inoperable steam generator pressure transmitter affecting the Engineered Safety Features Actuation System. These events are being reported, in accordance with 10 CFR 50.73(a)(2)(iv)(A), as an event that resulted in an automatic actuation of the reactor protection system and the auxiliary feedwater system; 10 CFR 50.73(a)(2)(i)(B), as an event that resulted in a condition prohibited by Technical Specifications; and in accordance with 10 CFR 50.73(a)(2)(v)(D), as an event that resulted in a condition which could have prevented the fulfillment of a safety function necessary to mitigate the consequences of an accident. A supplement to this LER is planned for April 30, 2020.

There are no regulatory commitments contained in this letter. Should you have any questions concerning this submittal, please contact Mr. Andrew McNeil, acting Site Licensing Manager, at (423) 843-8098.

Respectfully, Matthew Rasmussen Site Vice President Sequoyah Nuclear Plant Enclosure: Licensee Event Report 50-327/2019-003-00 cc:

NRC Regional Administrator - Region II NRC Senior Resident Inspector-Sequoyah Nuclear Plant printed on recycled paper

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 03/31/2020 (04-2018)

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Sequoyah Nuclear Plant Unit 1 05000327

3. Page 1 OF7

4. Title Automatic Reactor 7 rip due to Negative Rate Trip as a Result of a Dropped Control Rod

5. Event Date

6. LER Number

7. Report Date

8. Other Facilities Involved Month Day Year Year Sequential Number Rev No.

Month Day Year Facility Name NA Docket Number 05000 08 27 2019 2019

- 003

- 00 10 23 2019 Facility Name Docket Number NA 05000 a OperatingMode 1

) l~l No 04 30 2020 Abstract (Limit to 1400 spaces, i.e., approximately14 single-spaced typewrittenlines)

On August 27, 2019, at 0109 eastern daylight time, Sequoyah Nuclear Plant, Unit 1, automatically tripped as a result of Control Bank D Control Rod H-8 dropping into the core. All safety related equipment operated as designed, all control rods fully inserted as required, and auxiliary feedwater automatically initiated as expected. Additionally, during a post trip review, itwas identified that the Steam Generator #3 pressure transmitter, 1-PT-1-23, demonstrated sluggish behavior. This pressure transmitter had previously been identified as inoperable during a Unit 1 reactor tripoccurring April 14, 2019. Maintenance was performed following the April event and the transmitter was declared operable. After further review, it was determined that the transmitter was inoperable since November of 2015.

For Control Rod H-8, troubleshooting was performed and did not identify any immediate concerns with the Control Rod Drive Mechanism (CRDM) System. With no cause identified, a monitoring plan was developed to enable identification of a cause in the event the condition was repeated prior to the Unit 1 refueling outage. For 1-PT-1-23, work orders were completed to backfill and blow out the sensing lines and the transmitter was declared operable on August 28, 2019. The causes of these events are still under investigation. When the final investigations are completed, a supplement to this LER will be provided.

NRC FORM 366 (04-2018)

I.

Plant Operating Conditions Before the Event

At the time of the event, Sequoyah Nuclear Plant (SQN) Unit 1 was in Mode 1 at 100 percent rated thermal power.

II.

Description of Event

A.

Event Summary

YEAR 2019 SEQUENTIAL NUMBER

• 003 REV NO.

- 00 Prior to beginning the coast-down for the Unit 1 refueling outage, scheduled to begin October 12, 2019, Control Bank D was stepped out of the core from 220 steps to 228 steps.

The rods were in this position for approximately 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. At 0109 eastern daylight time (EDT) on August 27, 2019, SQN Unit 1 automatically tripped due to a negative rate trip as a result of Control Bank D, Control Rod, H-8 [EMS: AA] dropping into the core. Investigation revealed Control Rod H-8 dropped into the core approximately 1.3 seconds before all other control rods.

Post-trip troubleshooting was performed and did not identify any immediate concerns with the Control Rod Drive Mechanism (CRDM) System. With no cause identified, a monitoring plan was developed to enable site personnel to identify a cause in the event the condition was repeated prior to the Unit 1 refueling outage. Additionally, rod control current-order testing was performed to ensure the rod control system functioned as designed. Results from the current-order testing identified acceptable results from 0 to approximately 20 steps and 220 to 231 steps. A temporary modification was implemented to energize both the stationary gripper and the movable gripper to ensure Control Rod H-8 does not drop. The modification would not prevent Control Rod H-8 from dropping into the core upon opening of the reactor trip breakers.

Following the reactor trip, all safety related equipment operated as designed. No complications were experienced.

The Unit 1 reactor trip is reportable in accordance with 10 CFR 50.73(a)(2)(iv)(A), as an event that resulted in an automatic actuation of the Reactor Protection System [EMS: JC] and the Auxiliary Feedwater (AFW) [EMS: BA] System.

Additionally, during a post trip review, itwas identified that Steam Generator (SG) [ENS: SB] #3 pressure transmitter [EMS: PT], 1-PT-1-23, demonstrated sluggish behavior. The sluggish response could have challenged the Engineered Safety Features Actuation System (ESFAS)

[EMS: JE] input function associated with SG pressure. 1-PT-1-23 is one of three pressure transmitters used to monitor SG #3 pressure. These three transmitters are used, with their instrument loops, to provide two-out-of-three logic (per SG) required by technical specifications for ESFAS. The affected portion of ESFAS included actuation signals for Safety Injection (SI)

[EMS: BQ] and Steam Line Isolation (SLI).(04-2018)

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7

3. LER NUMBER Sequoyah Nuclear Plant Unit 1 05000-327 YEAR SEQUENTIAL NUMBER REV NO.

- 00 B.

2019 003 Maintenance was performed to backfill the sense line and the transmitter tubing was removed and checked to verify that there were no blockages. 1-PT-1-23 was declared operable on August 28 at 0142 EDT.

1-PT-1-23 had previously been identified as inoperable during a Unit 1 reactor trip occurring April 15, 2019. Maintenance was performed following the April event and 1-PT-1-23 was declared operable. A past operability evaluation determined that the maintenance performed in April did not correct the condition and 1-PT-1-23 was actually inoperable since November of 2015.

With 1-PT-1-23 inoperable, Technical Specification (TS) Limiting Condition for Operation (LCO) 3.3.2.D for ESFAS Instrumentation required the channel be placed in trip within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or be in Mode 3 within 78 hours9.027778e-4 days <br />0.0217 hours <br />1.289683e-4 weeks <br />2.9679e-5 months <br />. Failing to complete the Required Actions led to a condition prohibited by TS, and is reportable under 10 CFR 50.73(a)(2)(i)(B). Additionally, a past operability evaluation determined other steam line pressure channels in addition to 1-PT-1-23 were removed from service for testing since 2015. With 1-PT-1-23 and an additional instrument for that channel inoperable, ESFAS instrumentation would not have provided sufficient logic for actuation. This condition could have prevented the fulfillment of a safety function necessary to mitigate the consequences of an accident, which is reportable under 10 CFR 50.73(a)(2)(v)(D).

Status of structures, components, or systems that were inoperable at the start of the event and contributed to the event:

No inoperable structures, components, or systems contributed to this event.

C.

Dates and approximate times of occurrences

Date/Time (EDT)

Description

November 2015 Trace analysis of plant computer data indicated that 1-PT-1-23 pressure response was lagging. A condition report (CR) was initiated, but 1-PT-1-23 was not repaired. 1-PT-1-23 was not recognized as inoperable.

April 15,2019, 1229 Following a Unit 1 trip associated with the loss of an operating Main Feedwater Pump, a post trip review identified that 1-PT-1-23 pressure response was lagging. TS LCO 3.3.2.D was entered.

April 17,2019, 0332 Maintenance was completed with the sensing line blown down.

1-PT-1-23 was declared operable and TS LCO 3.3.2.D was exited.

However, 1-PT-1-23 was still inoperable.Page 3 of 7

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W APPROVED BY OMB: NO. 3150-0104 EXPIRES: 03/31/2020 Estimated burden perresponse to comply with this mandatory collection request: 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />. Reported lessons learned are incorporated intothe licensing processand fed back to industry. Send comments regarding burden estimate to the Information Services Branch (T-2 F43), U. S. Nuclear Regulatory Commission, Washington, DC 20556-0001,or bye-mail to lnfocollects.Resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC 20503. If a means used to impose an information collection does not display a currently valid OMB control number, the NRC may not conductor sponsor, anda personis notrequired to respondto,the informationcoilection.

Sequoyah Nuclear Plant Unit 1 05000-327 YEAR 2019

3. LBRNUMBER SEQUENTIAL NUMBER 003 Date/Time (EDT)

Description

August 26, 2019 2332 Control Bank D stepped out to 228 steps.

August 27, 2019, 0109 The Unit 1 reactor tripped on power range high neutron flux rate due to Control Rod H-8 dropping into the core. Unit 1 entered Mode 3.

Main control room (MCR) operators took appropriate actions in accordance with applicable procedures.

August 27, 2019, 0424 A post trip review identified that 1-PT-1-23 pressure response was lagging. TS LCO 3.3.2.D was entered.

August 28, 2019, 0142 Maintenance was completed and 1-PT-1-23 was declared operable.

TS LCO 3.3.2.D was exited.

November 2015-August28, 2019 Various maintenance and testing activities were performed (cumulative time 29 hours3.356481e-4 days <br />0.00806 hours <br />4.794974e-5 weeks <br />1.10345e-5 months <br /> and 43 minutes) with both 1-PT-1-23 and additional channels removed from service.

D.

Manufacturer and model number of each component that failed during the event

The causes of these events are still under investigation. When the final investigations are completed, a supplement to this LER will be provided.

E.

Other systems or secondary functions affected

1-PT-1-23 provides input to the ESFAS.

F.

Method of discovery of each component or system failure or procedural error

MCR alarms and annunciators provided indication to the operators during the reactor trip.

1-PT-1-23 was discovered to be inoperable during a post trip review.

G.

Failure mode, mechanism, and effect of each failed component:

The causes of these events are still under investigation. When the final investigations are completed, a supplement to this LER will be provided.

REV NO.

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3. LB* NUMBER Sequoyah Nuclear Plant Unit 1 05000-327 YEAR 2019 SEQUENTIAL NUMBER

• 003 REV NO.

- 00 H.

Operator actions

MCR operators responded to the reactor trip, as required. They promptly identified the condition and took appropriate actions.

I.

Automatically and manually initiated safety system responses

The reactor protection system, including main feedwater isolation and AFW start, responded to the trip, as designed. All rods fully inserted as required.

III.

Cause of the Event

A.

Cause of each component or system failure or personnel error

The causes of these events are still under investigation. When the final investigations are completed, a supplement to this LER will be provided.

B.

Cause(s) and circumstances for each human performance related root cause

The causes of these events are still under investigation. When the final investigations are completed, a supplement to this LER will be provided.

IV.

Analysis of the Event

The plant safety system responses during and after the reactor trip were bounded by the responses described in the Updated Final Safety Analysis Report (UFSAR). The UFSAR Chapter 15 event that most closely matches the reactor trip is the Rod Cluster Control Assembly Misalignment. The degraded response of 1-PT-1-23 did not cause abnormal degradation of, or stress upon, the principle safety barriers (fuel cladding, reactor coolant system, or containment). Therefore, this event did not adversely affect the health and safety of plant personnel or the general public.

V.

Assessment of Safety Consequences

There were no actual safety consequences as a result of the reactor trip.

A probabilistic risk assessment (PRA) evaluation was completed for the inoperable 1-PT-1-23.

The evaluation considered 1-PT-1-23 inoperable in conjunction with not maintaining a two-out-of-three logic for ESFAS instrumentation (SI signal on low SG pressure, SLI signal on low SG pressure, SLI signal on a High Negative Rate) for a duration covering 29 hours3.356481e-4 days <br />0.00806 hours <br />4.794974e-5 weeks <br />1.10345e-5 months <br /> and 43 minutes (28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br /> and 10 minutes previously reported in LER 1-2019-002 plus 93 minutes since previous LER), and the safety significance was determined to be low.Page 5 of 7(04-2018)

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V Sequoyah Nuclear Plant Unit 1 05000-327 YEAR 2019 aLBtNUIVBER SEQUENTIAL NUMBER 003 A.

Availability of systems or components that could have performed the same function as the components and systems that failed during the event:

None.

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:

The event did not occur when the reactor was shut down.

C.

For failure that rendered a train of a safety system inoperable, an estimate of the elapsed time from discovery of the failure until the train was returned to service:

1-PT-1-23 was declared inoperable and returned to service in approximately 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br />.

VI.

Corrective Actions

The reactor trip event was entered into the Tennessee Valley Authority Corrective Action Program (CAP) under CR 1544222. The inoperability of 1-PT-1-23 was entered into the CAP under CR 1544227.

A.

Immediate Corrective Actions

For Control Rod H-8, troubleshooting was performed and did not identify any immediate concerns with the Control Rod Drive Mechanism (CRDM) System. With no cause identified, a monitoring plan was developed to enable site personnel to identify a cause in the event the condition was repeated prior to the Unit 1 refueling outage. Additionally, rod control current-order testing was performed to ensure the rod control system functioned as designed. Results from the current-order testing identified acceptable results from 0 to approximately 20 steps and 220 to 231 steps. However, unexpected anomalies were identified, during the test, which will be further investigated. Westinghouse was consulted on the issue and concluded that the current-order issue is a potential failure mechanism in conjunction with another potential failure mechanism. Actions were put into place to not exceed 220 steps to ensure reliability prior to beginning the refueling outage to preclude another event. A temporary modification was implemented to energize both the stationary gripper and movable gripper to ensure Control Rod H-8 does not drop. The modification would not prevent Control Rod H-8from dropping into the core upon opening of the reactor trip breakers.

For 1-PT-1-23, work orders were completed to backfill and blow out the sensing lines.

REV NO.

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Sequoyah Nuclear Plant Unit 1 05000-327 YEAR 2019 a LER NUMBER SEQUENTIAL NUMBER 003 B.

Corrective Actions to Prevent Recurrence or to reduce probability of similar events occurring in the future:

The causes of these events are still under investigation, completed, a supplement to this LER will be provided.

VII.

Previous Similar Events at the Same Site

When the final investigations are A review of previous reportable events related to an automatic reactor trip due to a negative rate trip identified an automatic reactor trip as a result of Control Rod, H-8, falling into the Unit 1 core reported in LER 1-2015-001. The root cause of the event was failure of a maintenance procedure to provide inspection guidance and acceptance criteria on CRDM vertical panel connections. The corrective action to prevent recurrence was revising the maintenance procedure and periodic preventive maintenance of CRDM connections.

Additionally, a review of previous reportable events for the past three years identified that LER 1-2019-002 involved a condition prohibited by technical specifications and an event that could have prevented the fulfillment of a safety function associated with inoperable pressure transmitter, 1-PT-1-23. The maintenance performed on the transmitter, to restore operability, was unsuccessful and not identified until another reactor trip approximately four months later.

The corrective action was to create preventative maintenance instructions to clear sensing lines for Main Steam transmitters.

VIII.

Additional Information

There is no additional information.

IX.

Commitments

There are no commitments.

REV NO.

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