05000498/LER-2015-001, Regarding Manual Reactor Trip Due to Lowering Steam Generator Levels and Valid Auxiliary Feedwater System Actuation Following a Manual Main Turbine Trip

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Regarding Manual Reactor Trip Due to Lowering Steam Generator Levels and Valid Auxiliary Feedwater System Actuation Following a Manual Main Turbine Trip
ML16067A086
Person / Time
Site: South Texas 
Issue date: 02/18/2016
From: Gerry Powell
South Texas
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
NOC-AE-16003324 LER 2015-001-00
Download: ML16067A086 (8)
v • d • e


LER-2015-001, Regarding Manual Reactor Trip Due to Lowering Steam Generator Levels and Valid Auxiliary Feedwater System Actuation Following a Manual Main Turbine Trip
Event date:
Report date:
Reporting criterion: 10 CFR 50.73(a)(2)(iv)(A), System Actuation
10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded
10 CFR 50.73(a)(2)(viii)(A)
10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition
10 CFR 50.73(a)(2)
10 CFR 50.73(a)(2)(x)
10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor
10 CFR 50.73(a)(2)(v), Loss of Safety Function
10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown
10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
10 CFR 50.73(a)(2)(vii), Common Cause Inoperability
10 CFR 50.73(a)(2)(i)
4982015001R00 - NRC Website
v • d • e

text

Nuclear Operating Company South Texas Proiect Electric Generating Station P.O. Box 28,9 Wadsworth, Texas 77483

The event did not have an adverse effect on the health and safety of the public.

There are no commitments in this letter.

If there are any questions, please contact Wendy Brost at (361) 972-8516 or me at (361) 972-7566.

G. T. Powell Site Vice President web

Attachment:

Unit 1 LER 2015-001-00 STI: 34259540

NOC-AE-1 6003324 Page 2 of 2 cc:

(paper copy)

(electronic copy)

.Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 1600 East Lamar Boulevard Arlington, TX 76011-4511 Lisa M. Regner Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North (08 H04) 11555 Rockville Pike Rockville, MD 20852 NRC Resident Inspector U. S. Nuclear Regulatory Commission P. O. Box 289, Mail Code: MNl16 Wadsworth, TX 77483 Morgan, Lewis & Bockius LLP Steve Frantz, Esquire U.S. Nuclear Regulatory Commission Lisa M. Regner NRG South Texas LP John Ragan Chris O'Hara Jim von Suskil CPS Energqy Kevin Polio Cris Eugster L. D. Blaylock Cramn Caton & James, P.C.

Peter Nemeth City of Austin Elaina Ball John Wester Texas Dept. of State Health Services Richard A. Ratliff Robert Free

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB: NO. 3150-0104 EXPIRES: 1013112018 (11-2015)

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

1. FACILITY NAME 2

OKTNME

.PG South Texas Unit 1 000481O

4. TITLE Manual Reactor Trip due to Lowering Steam Generator Levels and Valid Auxiliary Feedwater System Actuation Following a Manual Main Turbine Trip
5. EVENT DATE
6. LER NUMBER 7..REPORT DATE
8. OTHER FACILITIES INVOLVED MONH AY YEA YAR SEQUENTIAL REV MONTH DAY YEAR FcUYNM o~TNME MNH DY EA YER NUMBER NO.

J N/A 05000 FAcILITY NAME DOcKET NUMBER 12 21 2015 2015 001 00 02 18 2016 N/A 05000

9. OPERATING MODE.
11. THIS REPORT IS SUBMITT'ED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)

D] 20.2201 (b)

[] 20.2203(a)(3)(i)

[]50.73(a)(2)(ii)(A)

[] 50.73(a)(2)(viii)(A) 1 D 20.2201 (d)

ElI 20.2203(a)(3)(ii)

[] 50.73(a)(2) (ii)(B)

[] 50.73(a)(2)(Viii)(B) 20.2203(a)(2)(i)

[] 50.36(c)(1 )(i)(A)

[]50.73(a)(2)(iv)(A)

D] 50.73(a)(2)(x)

10. POWER LEVEL D] 20.2203(a)(2)(ii)

[] 50.36(0)(1)(ii)(A)

[]50.73(a)(2)(v)(A)

[]73.71 (a)(4)

[] 20,2203(a)(2)(iii)

D] 50,36(c)(2)

D] 50.73(a)(2)(V)(B)

[]73.71 (a)(5)

D] 20.2203(a)(2)(iv)

[] 50.46(a)(3)(ii)

[]50.73(a)(2)(v)(C)

[] 73.77(a)(1) 48%

jJ 20.2203(a)(2)(v)

[] 50.73(a)(2)(i)(A)

[] 50.73(a)(2)(v)(D)

[] 73.77(a)(2)(i)

LI 20.2203(a)(2)(vi)

[] 50.73(a)(2)(i)(B)

LI 50.73(a)(2)(vii)

[] 73.77(a)(2)(ii)

LI 50.73(a)(2)(i)(C)

[OTHER Specify in Abatract below or in summary of the event On December 21, 2015, STP Unit 1 power ascension following a refueling outage was in progress and the reactor was at approximately 48 percent rated thermal power. At approximately 1450 hours0.0168 days <br />0.403 hours <br />0.0024 weeks <br />5.51725e-4 months <br />, Operators observed Reactor Coolant System (RCS) temperature fluctuations due to turbine load swings caused by an oscillating Main Turbine Governor Valve (GV), GV2.

At 1453, Main Turbine demand rose approximately 5 percent and GV2 continued to cycle.

At 1455 hours0.0168 days <br />0.404 hours <br />0.00241 weeks <br />5.536275e-4 months <br />, the Group 1 Steam Dumps opened for approximately 23 seconds. At 1456, Operators commenced load reduction on the Main Turbine to attempt to lower turbine demand. Operators observed power lowering but there was no effect on the GV2 oscillations.

At 1508 and 1510, the Group 2 Steam Dumps modulated open and closed while the Group 1 Steam Dumps remained closed due to a failure of the valve positioners.

At 1519, Operators manually tripped the Main Turbine. With reactor power less than 50 percent, as expected, the reactor did not automatically trip when the turbine tripped. When the Main Turbine Trip signal was received, the steam dump valve positioners were bypassed as designed and the Group 1, 2, and 3 Steam Dumps momentarily opened. Following the turbine trip, the steam dumps returned to a modulation mode of operation.

At 1524, a Main Feedwater Isolation occurred and the loss of feedwater resulted in lowering steam generator (SG) levels. This was due to the failure of the Group 1 Steam Dumps to modulate in response to the Main Turbine load changes, which resulted in a significant difference between steam flow and feedwater flow. Operators attempted to manually reduce feedwater flow but were not able to prevent the Main Feedwater Isolation.

B. Cause of component failure

The cause of the GV2 failure was an intermittent groutnd on the signal wire to the LVDT for GV2 that was the result of a small score in the insulation of the LVDT signal wiring. This wiring was vendor supplied and there is no documented history of this wiring ever being replaced or reworked. The insulation was most likely damaged during the initial installation.

The aggressive fluctuations in steam flow due to the GV2 oscillations caused the spring clips in the Group 1 Steam Dumps to become dislodged, causing the valves to be unresponsive to modulation demands. The steam dumps regularly cope with changes in steam flow during normal operation; however, the aggressive steam flow fluctuations that the Group 1 Steam Dutmps experienced in this event challenged the design of the Steam Dump system, resulting in the malfunction of the spring clips on the positioners.

C. Systems or secondary futnctions that were affected by failutre of components with multiple functions The failed components described in the narrative, Steam Dump Group 1 and GV2, do not have multiple functions that affect other systems. The failures of these components contributed to the eventual Main Turbine trip and reactor trip.

D. Failed component information (Energy Industry Identification System (EIIS) designators provided in

{brackets})

High Pressure Turbine Govemnor Valve Position Transmitter {ZT}

Manuifacturer: Westinghouse Electric Corporation Model: 677J444G21 Steam Dumps Valve Positioner {V}

Manufacturer: Bailey Controls Model: AV112000 NRC FORM 360A (02-2014)

III. Analysis of the event

A. Safety system responses that occurred The Reactor Protection System and AFW systems both responded to this event.

B. Duration of safety system inoperability

There were no SSCs that were inoperable at the start of the event that contributed to the event.

C. Safety consequences and implications

No Technical Specification LCOs were entered due to this event. Operators manually tripped the reactor following the Main Feedwater isolation.

For the Probabilistic Risk Assessment (PRA) analysis, the initiating event is classified as a Total Loss of Main Feedwater (TLMFW) - the isolation of main feedwater led to decreasing levels in the SG which would have inevitably resulted in an automatic reactor trip. The TLMIFW event is a modeled initiating event, and no risk significant equipment was confirmed out of service.

The STP PRA was used to estimate the relevant metrics for a reactor trip, Conditional Core Damage Probability (CCDP) and Conditional Large Early Release Probability (CLERP), given that the TLMIFW initiating event actually occurred. The CCDP and CLERP were determined to be 5.99E-07 and 3.36E-08, respectively, indicating very low risk significance.

The resulting risk of this event is well within the NRC acceptance criteria of less than 1E-06 events per year for the CCDP and less than 1E-07 events per year for the CLERP, as outlined in Regulatory Guide 1.174.

The event was of very low risk significance and no radioactive release occurred; therefore, there was no adverse effect on the health and safety of the public.

IV. Cause of the event

Prior to and following the manual trip of the Main Turbine, the Group 1 Steam Dumps did not respond as expected for the load shed, resulting in a Main Feedwater Isolation due to rising SG level. Operators then initiated a manual reactor trip due to lowering SG levels and the AFW system actuated automatically on a SG low level signal. There were no human performance errors that contributed to the event.

V. Corrective actions

As a corrective action, STP replaced the LVDT and the associated cables for GV2. Inspections were also performed on all Unit 1 governor and throttle valves following the reactor trip to ensure that the condition was limited to GV2. Inspections will be performed on the cables and wiring associated with the LVDTs and servo valves for the governor and throttle valves in Unit 2 during the next Unit 2 refueling outage.

Visual Inspections were performed on all Unit 1 Steam Dump Groups following the reactor trip. Repairs to the Group 1 Steam Dumps Were completed on December 23, 2015 and the spring clips were verified to be within tolerance.

VI. Previous similar events

An Operating Experience review was conducted as part of the Cause Evaluation performed for this event.

Several failures of the High Pressure Governor valves due to loose or faulty connections, however, none of these failures resulted from insulation damage or shield grounding.

Several failures related to the steam dump valves were reviewed and none of these failures resulted from the spring clips being dislodged following a secondary transient. A similar event (Condition Report 08-43 13) consisting of valve oscillations of IliP Turbine GV1 led to perturbations in the secondary that cycled Electro-Hydraulic lines greater than six inches. There was no report of steam dump failures at that time.

One previous STP Unit 1 LER (2000-007-00) has been submitted related to governor valves and the steam dumps with a subsequent manual reactor trip. The cause of this event was a failed logic card and missing screw on the steam dump actuator hand wheel.

NRC FORM 3S66A (02-2014)

Nuclear Operating Company South Texas Proiect Electric Generating Station P.O. Box 28,9 Wadsworth, Texas 77483

The event did not have an adverse effect on the health and safety of the public.

There are no commitments in this letter.

If there are any questions, please contact Wendy Brost at (361) 972-8516 or me at (361) 972-7566.

G. T. Powell Site Vice President web

Attachment:

Unit 1 LER 2015-001-00 STI: 34259540

NOC-AE-1 6003324 Page 2 of 2 cc:

(paper copy)

(electronic copy)

.Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 1600 East Lamar Boulevard Arlington, TX 76011-4511 Lisa M. Regner Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North (08 H04) 11555 Rockville Pike Rockville, MD 20852 NRC Resident Inspector U. S. Nuclear Regulatory Commission P. O. Box 289, Mail Code: MNl16 Wadsworth, TX 77483 Morgan, Lewis & Bockius LLP Steve Frantz, Esquire U.S. Nuclear Regulatory Commission Lisa M. Regner NRG South Texas LP John Ragan Chris O'Hara Jim von Suskil CPS Energqy Kevin Polio Cris Eugster L. D. Blaylock Cramn Caton & James, P.C.

Peter Nemeth City of Austin Elaina Ball John Wester Texas Dept. of State Health Services Richard A. Ratliff Robert Free

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB: NO. 3150-0104 EXPIRES: 1013112018 (11-2015)

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

1. FACILITY NAME 2

OKTNME

.PG South Texas Unit 1 000481O

4. TITLE Manual Reactor Trip due to Lowering Steam Generator Levels and Valid Auxiliary Feedwater System Actuation Following a Manual Main Turbine Trip
5. EVENT DATE
6. LER NUMBER 7..REPORT DATE
8. OTHER FACILITIES INVOLVED MONH AY YEA YAR SEQUENTIAL REV MONTH DAY YEAR FcUYNM o~TNME MNH DY EA YER NUMBER NO.

J N/A 05000 FAcILITY NAME DOcKET NUMBER 12 21 2015 2015 001 00 02 18 2016 N/A 05000

9. OPERATING MODE.
11. THIS REPORT IS SUBMITT'ED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)

D] 20.2201 (b)

[] 20.2203(a)(3)(i)

[]50.73(a)(2)(ii)(A)

[] 50.73(a)(2)(viii)(A) 1 D 20.2201 (d)

ElI 20.2203(a)(3)(ii)

[] 50.73(a)(2) (ii)(B)

[] 50.73(a)(2)(Viii)(B) 20.2203(a)(2)(i)

[] 50.36(c)(1 )(i)(A)

[]50.73(a)(2)(iv)(A)

D] 50.73(a)(2)(x)

10. POWER LEVEL D] 20.2203(a)(2)(ii)

[] 50.36(0)(1)(ii)(A)

[]50.73(a)(2)(v)(A)

[]73.71 (a)(4)

[] 20,2203(a)(2)(iii)

D] 50,36(c)(2)

D] 50.73(a)(2)(V)(B)

[]73.71 (a)(5)

D] 20.2203(a)(2)(iv)

[] 50.46(a)(3)(ii)

[]50.73(a)(2)(v)(C)

[] 73.77(a)(1) 48%

jJ 20.2203(a)(2)(v)

[] 50.73(a)(2)(i)(A)

[] 50.73(a)(2)(v)(D)

[] 73.77(a)(2)(i)

LI 20.2203(a)(2)(vi)

[] 50.73(a)(2)(i)(B)

LI 50.73(a)(2)(vii)

[] 73.77(a)(2)(ii)

LI 50.73(a)(2)(i)(C)

[OTHER Specify in Abatract below or in I.Description of reportable event

A. Reportable event classification

This event is reportable under §50.73(a)(2)(iv) (A) as an event or condition that resulted in a manual actuation of the Reactor Protection System and also as an event or condition that resulted in an automatic actuation of the Auxiliary Feedwater (AFW) system.

B. Plant operating conditions prior to event

On December 21, 2015, Unit 1 was operating in Mode 1 at 48 percent power. Unit 1 was returning to power operation following refueling outage 1RE19.

C. Status of structures, systems, and components (SSCs) that were inoperable at the start of the event and that contributed to the event There were no SSCs that were inoperable at the start of the event that contributed to the event.

D. Narrative summary of the event

On December 21, 2015, STP Unit 1 power ascension following a refueling outage was in progress and the reactor was at approximately 48 percent rated thermal power. At approximately 1450 hours0.0168 days <br />0.403 hours <br />0.0024 weeks <br />5.51725e-4 months <br />, Operators observed Reactor Coolant System (RCS) temperature fluctuations due to turbine load swings caused by an oscillating Main Turbine Governor Valve (GV), GV2.

At 1453, Main Turbine demand rose approximately 5 percent and GV2 continued to cycle.

At 1455 hours0.0168 days <br />0.404 hours <br />0.00241 weeks <br />5.536275e-4 months <br />, the Group 1 Steam Dumps opened for approximately 23 seconds. At 1456, Operators commenced load reduction on the Main Turbine to attempt to lower turbine demand. Operators observed power lowering but there was no effect on the GV2 oscillations.

At 1508 and 1510, the Group 2 Steam Dumps modulated open and closed while the Group 1 Steam Dumps remained closed due to a failure of the valve positioners.

At 1519, Operators manually tripped the Main Turbine. With reactor power less than 50 percent, as expected, the reactor did not automatically trip when the turbine tripped. When the Main Turbine Trip signal was received, the steam dump valve positioners were bypassed as designed and the Group 1, 2, and 3 Steam Dumps momentarily opened. Following the turbine trip, the steam dumps returned to a modulation mode of operation.

At 1524, a Main Feedwater Isolation occurred and the loss of feedwater resulted in lowering steam generator (SG) levels. This was due to the failure of the Group 1 Steam Dumps to modulate in response to the Main Turbine load changes, which resulted in a significant difference between steam flow and feedwater flow. Operators attempted to manually reduce feedwater flow but were not able to prevent the Main Feedwater Isolation.

B. Cause of component failure

The cause of the GV2 failure was an intermittent groutnd on the signal wire to the LVDT for GV2 that was the result of a small score in the insulation of the LVDT signal wiring. This wiring was vendor supplied and there is no documented history of this wiring ever being replaced or reworked. The insulation was most likely damaged during the initial installation.

The aggressive fluctuations in steam flow due to the GV2 oscillations caused the spring clips in the Group 1 Steam Dumps to become dislodged, causing the valves to be unresponsive to modulation demands. The steam dumps regularly cope with changes in steam flow during normal operation; however, the aggressive steam flow fluctuations that the Group 1 Steam Dutmps experienced in this event challenged the design of the Steam Dump system, resulting in the malfunction of the spring clips on the positioners.

C. Systems or secondary futnctions that were affected by failutre of components with multiple functions The failed components described in the narrative, Steam Dump Group 1 and GV2, do not have multiple functions that affect other systems. The failures of these components contributed to the eventual Main Turbine trip and reactor trip.

D. Failed component information (Energy Industry Identification System (EIIS) designators provided in

{brackets})

High Pressure Turbine Govemnor Valve Position Transmitter {ZT}

Manuifacturer: Westinghouse Electric Corporation Model: 677J444G21 Steam Dumps Valve Positioner {V}

Manufacturer: Bailey Controls Model: AV112000 NRC FORM 360A (02-2014)

III. Analysis of the event

A. Safety system responses that occurred The Reactor Protection System and AFW systems both responded to this event.

B. Duration of safety system inoperability

There were no SSCs that were inoperable at the start of the event that contributed to the event.

C. Safety consequences and implications

No Technical Specification LCOs were entered due to this event. Operators manually tripped the reactor following the Main Feedwater isolation.

For the Probabilistic Risk Assessment (PRA) analysis, the initiating event is classified as a Total Loss of Main Feedwater (TLMFW) - the isolation of main feedwater led to decreasing levels in the SG which would have inevitably resulted in an automatic reactor trip. The TLMIFW event is a modeled initiating event, and no risk significant equipment was confirmed out of service.

The STP PRA was used to estimate the relevant metrics for a reactor trip, Conditional Core Damage Probability (CCDP) and Conditional Large Early Release Probability (CLERP), given that the TLMIFW initiating event actually occurred. The CCDP and CLERP were determined to be 5.99E-07 and 3.36E-08, respectively, indicating very low risk significance.

The resulting risk of this event is well within the NRC acceptance criteria of less than 1E-06 events per year for the CCDP and less than 1E-07 events per year for the CLERP, as outlined in Regulatory Guide 1.174.

The event was of very low risk significance and no radioactive release occurred; therefore, there was no adverse effect on the health and safety of the public.

IV. Cause of the event

Prior to and following the manual trip of the Main Turbine, the Group 1 Steam Dumps did not respond as expected for the load shed, resulting in a Main Feedwater Isolation due to rising SG level. Operators then initiated a manual reactor trip due to lowering SG levels and the AFW system actuated automatically on a SG low level signal. There were no human performance errors that contributed to the event.

V. Corrective actions

As a corrective action, STP replaced the LVDT and the associated cables for GV2. Inspections were also performed on all Unit 1 governor and throttle valves following the reactor trip to ensure that the condition was limited to GV2. Inspections will be performed on the cables and wiring associated with the LVDTs and servo valves for the governor and throttle valves in Unit 2 during the next Unit 2 refueling outage.

Visual Inspections were performed on all Unit 1 Steam Dump Groups following the reactor trip. Repairs to the Group 1 Steam Dumps Were completed on December 23, 2015 and the spring clips were verified to be within tolerance.

VI. Previous similar events

An Operating Experience review was conducted as part of the Cause Evaluation performed for this event.

Several failures of the High Pressure Governor valves due to loose or faulty connections, however, none of these failures resulted from insulation damage or shield grounding.

Several failures related to the steam dump valves were reviewed and none of these failures resulted from the spring clips being dislodged following a secondary transient. A similar event (Condition Report 08-43 13) consisting of valve oscillations of IliP Turbine GV1 led to perturbations in the secondary that cycled Electro-Hydraulic lines greater than six inches. There was no report of steam dump failures at that time.

One previous STP Unit 1 LER (2000-007-00) has been submitted related to governor valves and the steam dumps with a subsequent manual reactor trip. The cause of this event was a failed logic card and missing screw on the steam dump actuator hand wheel.

NRC FORM 3S66A (02-2014)

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