05000327/LER-2019-002, Regarding Steam Generator Pressure Transmitter Degraded Sensing Line Causes Condition Prohibited by Technical Specifications

Regarding Steam Generator Pressure Transmitter Degraded Sensing Line Causes Condition Prohibited by Technical Specifications

ML19207A069
Person / Time
Site:	Sequoyah
Issue date:	07/26/2019
From:	Rasmussen M Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LER 2019-002-00
Download: ML19207A069 (7)

LER-2019-002, Regarding Steam Generator Pressure Transmitter Degraded Sensing Line Causes Condition Prohibited by Technical Specifications

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(v), Loss of Safety Function
3272019002R00 - NRC Website

text

Tennessee Valley Authority, Sequoyah Nuclear Plant, P.O. Box 2000, Soddy Daisy, Tennessee 37384 July 26, 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 50-327/2019-002-00, Steam Generator Pressure Transmitter Degraded Sensing Line Causes Condition Prohibited by Technical Specifications The enclosed licensee event report provides details concerning an inoperable steam generator pressure transmitter affecting the Engineered Safety Features Actuation System. This event is being reported, in accordance with 10 CFR 50.73(a)(2)(i)(B), as an event that resulted in a condition prohibited by Technical Specifications. Additionally, this event is being reported, 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.

There are no regulatory commitments contained in this letter. Should you have any questions concerning this submittal, please contact Mr. Jonathan Johnson, Site Licensing Manager, at (423)843-8129.

Respectfully, r^ Matthew Rasmussen Site Vice President Sequoyah Nuclear Plant Enclosure: Licensee Event Report 50-327/2019-002-00 cc:

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

NRC FORM 366 (04-2018)

NRC FORM 366 (04-2018)

U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 03/31/2020

, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

LICENSEE EVENT REPORT (LER)

(See Page 2 for required number of digits/characters for each block)

(See NUREG-1022, R.3 for instruction and guidance for completing this form http://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr1022/r3/)

1. Facility Name Sequoyah Nuclear Plant Unit 1

2. Docket Number 05000327

3. Page 1 OF 6

4. Title Steam Generator Pressure Transmitter Degraded Sensing Line Causes Condition Prohibited by Technical Specifications

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 05 29 2019 2019

- 002

- 00 07 26 2019 Facility Name Docket Number NA 05000

9. Operating Mode 1

)

No N/A N/A N/A Abstract (Limit to 1400 spaces, i.e., approximately 14 single-spaced typewritten lines)

On April 15, 2019, during a post trip review, it was identified that Steam Generator #3 pressure transmitter, 1-PT-1-23, had demonstrated sluggish behavior during the transient. The pressure input is utilized by the Engineered Safety Features Actuation System for several postulated events. Operations declared the instrument inoperable, and entered the appropriate Technical Specifications (TS) Limiting Condition for Operation (LCO) until maintenance activities restored operability. Investigation revealed that the instrument response exceeded the TS channel check acceptance criteria. Historical search and past operability evaluation completed May 29, 2019, determined that previous identification was not effectively resolved, and that there were multiple periods where other transmitters were removed from service at the same time as the affected transmitter. This led to a condition prohibited by TS, and a condition that could have prevented the fulfillment of a safety function. The cause of the component failure was debris found in the pressure transmitter sensing line due to a lack of regularly scheduled preventative maintenance. The corrective action is to ensure preventative maintenance instructions are created to clear sensing lines for Main Steam transmitters.

NRC FORM 366 (04-2018)

Page 2 of 6 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

Following a unit trip on April 14, 2019, it was noted, during a post trip review, that Steam Generator (SG) [EIIS: SB] #3 Steam Pressure Transmitter, 1-PT-1-23 [EIIS: PT], exhibited a sluggish response. This sluggish response could have challenged the Engineered Safety Features Actuation System (ESFAS) [EIIS: JE] input function associated with SG pressure.

It was determined on April 15, 2019, at 1229 eastern daylight time (EDT), that the pressure transmitter sensing line was partially blocked. Technical Specifications (TS) Limiting Condition for Operation (LCO) 3.3.2.D for ESFAS Instrumentation was entered at that time.

Maintenance was performed to blow down the sensing line, and 1-PT-1-23 was restored to operable status on April 17, 2019 at 0332 EDT.

A past operability evaluation (POE) completed on May 29, 2019, determined that 1-PT-1-23 was outside of its acceptance criteria during the transient conditions. The POE also revealed that this condition had been identified during a previous transient in 2015, but was not adequately dispositioned, which made 1-PT-1-23 inoperable since 2015 (not recognized at the time). With 1-PT-1-23 inoperable, TS 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 perform a plant shutdown within 78 hours9.027778e-4 days <br />0.0217 hours <br />1.289683e-4 weeks <br />2.9679e-5 months <br /> and exit the mode of applicability. 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, the POE determined other 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).

B.

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

There were no inoperable structures, systems, or components that contributed to this event.

NRC FORM 366 (04-2018)

Page 3 of 6 (04-2018)

U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 03/31/2020

, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

LICENSEE EVENT REPORT (LER)

CONTINUATION SHEET

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

2019

- 002

- 00

C. Dates and approximate times of occurrences

Date/Time (EDT)

Event 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.

November 2015 -

April 2019 Various maintenance and testing activities were performed (cumulative time 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br /> 10 minutes) with additional channels removed from service.

April 14, 2019, 0320 Unit 1 tripped following a loss of an operating Main Feedwater Pump April 15, 2019, 1229 Post trip review identified that 1-PT-1-23 pressure response was lagging. A CR was initiated, TS LCO 3.3.2.D was entered.

April 17, 2019 0332 Maintenance was completed with sensing line blown down. 1-PT-1-23 was restored to operable status and TS LCO 3.3.2.D was exited

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

The subject transmitter is a safety-related, 10-50 milliAmp Foxboro Model E11GM, with a process range of 0-1200 pounds per square inch gage. The component identification at SQN is SQN-1-PT-001-0023-F.

E.

Other systems or secondary functions affected

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

F.

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

The described condition was identified during a review of key parameters as part of a post trip review.

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

The sensing line was discovered partially blocked due to long-term build-up of solids. The identified blockage did not affect indicated pressure during steady state operations, but did provide sluggish indication and response when a significant load rejection occurred.

NRC FORM 366 (04-2018)

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LICENSEE EVENT REPORT (LER)

CONTINUATION SHEET

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

2019

- 002

- 00 1-PT-1-23 is one of three pressure transmitters used to monitor SG #3 steam pressure.

These three transmitters are used, with their instrument loops, to provide two-out-of-three logic (per SG) required by TS for ESFAS. The affected portion of ESFAS included actuation signals for Safety Injection [EIIS: BQ] and Steam Line Isolation (SLI). With the instrument line partially clogged, the pressure transmitter was inoperable and not able to respond in the time as required by Technical Specifications and the Final Safety Analysis Report (FSAR).

H. Operator actions

Operators completed review of key plant parameters, and submitted a CR for the identified pressure response anomaly.

I.

Automatically and manually initiated safety system responses

There were no automatic or manual safety system responses associated with this event.

III.

Cause of the Event

A.

Cause of each component or system failure or personnel error

The cause of the component failure was debris found in the pressure transmitter sensing line due to a lack of regularly scheduled preventative maintenance.

B.

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

There was no identified human performance related root cause.

IV.

Analysis of the Event

The degraded response of 1-PT-1-23 did not cause abnormal degradation of, or stress upon, the principle safety barriers (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

The discovery of the condition and subsequent repair encompassed a period of approximately 39 hours4.513889e-4 days <br />0.0108 hours <br />6.448413e-5 weeks <br />1.48395e-5 months <br />. Investigation of instrument behavior revealed a questionable behavioral response extending back several years (back to 2015). Transmitter operation at steady state power met channel checks as required by TS Surveillance Requirement 3.3.2.1, but only exhibited degradation during plant transients. A period of three years was selected to

NRC FORM 366 (04-2018)

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U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 03/31/2020

, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

LICENSEE EVENT REPORT (LER)

CONTINUATION SHEET

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

2019

- 002

- 00 assess the probabilistic risk assessment (PRA) evaluation based on NUREG-1022 to provide a bounding analysis period. Considering 1-PT-1-23 inoperable for three years 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 Isolation signal on a High Negative Rate) for a duration of 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br /> and 10 minutes over the specified three year span, the safety significance was determined to be low.

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

When the degraded condition was recognized on April 15, 2019, the appropriate LCO was entered, and the affected transmitter was returned to service within the required Completion Time. For past operability evaluation, 1-PT-1-23 was inoperable, and that condition placed the ESFAS function from two-out-of-three logic to two-out-of-two logic configuration. The ESFAS channel associated with the 1-PT-1-23 should have been tripped, which would have satisfied redundancy requirements, or the unit placed in a lower mode. With the exception of the times noted in the POE totaling 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br /> 10 minutes, the remaining SG pressure transmitters remained functional. The SI ESFAS signal is also initiated by high containment pressure or low pressurizer pressure, and the SLI signal is also initiated by high-high containment pressure. These utilize instrumentation different than 1-PT-1-23.

B.

For events that occurred when the reactor was shut down, availability of systems or components needed to shut down 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 39 hours4.513889e-4 days <br />0.0108 hours <br />6.448413e-5 weeks <br />1.48395e-5 months <br />.

VI.

Corrective Actions

Corrective Actions are being managed by Tennessee Valley Authoritys corrective action program under CRs 1507948 and 1516260.

NRC FORM 366 (04-2018)

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LICENSEE EVENT REPORT (LER)

CONTINUATION SHEET

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

2019

- 002

- 00 A.

Immediate Corrective Actions

The pressure transmitter was declared inoperable. Maintenance was initiated to blow down the transmitter sensing line, which was completed and subsequently returned to service.

B.

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

The corrective action is to ensure preventative maintenance instructions are created to clear sensing lines for Main Steam transmitters.

VII. Previous Similar Events at the Same Site

There were no previous similar events at SQN occurring within the last three years.

VIII. Additional Information

There is no additional information.

IX.

Commitments

There are no commitments.