05000327/LER-2011-003, For Sequoyah, Unit 1, Reactor Trip as a Result of Turbine Control Card Failure

For Sequoyah, Unit 1, Reactor Trip as a Result of Turbine Control Card Failure

ML11241A105
Person / Time
Site:	Sequoyah
Issue date:	08/24/2011
From:	John Carlin Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LER 11-003-00
Download: ML11241A105 (7)

LER-2011-003, For Sequoyah, Unit 1, Reactor Trip as a Result of Turbine Control Card Failure

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 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)(viii)(B) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(ix)(A) 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)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
3272011003R00 - NRC Website

text

Tennessee Valley Authority, Post Office Box 2000, Soddy Daisy, Tennessee 37384-2000 August 24, 2011 10 CFR 50.73 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Sequoyah Nuclear Plant, Unit 1 Facility Operating License No. DPR-77 NRC Docket Nos. 50-327

Subject:

License Event Report 327/2011-003, "Unit 1 Reactor Trip As a Result of Turbine Control Card Failure" The enclosed licensee event report provides details concerning an automatic reactor trip and automatic engineered safety feature actuation of auxiliary feedwater following the failure of a turbine control analog electro-hydraulic signal conditioning card. The Tennessee Valley Authority (TVA) is submitting this report in accordance with 10 CFR 50.73 (a)(2)(iv)(A)\\ a condition that resulted in automatic actuation of the reactor protection system.

The cause of the event is still under investigation; therefore, a supplement of this event report is forthcoming.

There are no regulatory commitments contained in this letter. Should you have any questions concerning this submittal, please contact G. M. Cook, Sequoyah Site Licensing Manager, at (423) 843-7170.

Respecully, I

Vi Preside'nt Sequoyah Nuclear Plant

Enclosure:

Licensee Event Report - Unit 1 Reactor Trip As a Result of Turbine Control Card Failure cc:

NRC Regional Administrator-Region II NRC Senior Resident Inspector - Sequoyah Nuclear Plant

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/31/20 (10-2010)

, the NRC may sfor each block) not conduct or sponsor, and a person is not required to respond to, the digits/characters finformation collection.

3. PAGE Sequoyah Nuclear Plant Unit 1 05000327 1 OF 6

4. TITLE:

Unit 1 Reactor Trip As a Result of Turbine Control Card Failure

5. EVENT DATE

6. LER NUMBER

7. REPORT DATE

8. OTHER FACILITIES INVOLVED SSEQUENTIAL REV MFACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR NUMBER NO.

MONTH DAY YEAR FACILITY NAME DOCKET NUMBER 06 26 2011 2011 -

003 00 08 24 2011

9. OPERATING MODE

11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) 1 E] 20.2201(b) 13 20.2203(a)(3)(i)

[]

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

[I 50.73(a)(2)(vii) 13 20.2201(d) 13 20.2203(a)(3)(ii)

F1 50.73(a)(2)(ii)(A) 13 50.73(a)(2)(viii)(A) 13 20.2203(a)(1) 13 20.2203(a)(4) 13 50.73(a)(2)(ii)(B) 13 50.73(a)(2)(viii)(B)

[]

20.2203(a)(2)(i) 1] 50.36(c)(1)(i)(A) 13 50.73(a)(2)(iii) 13 50.73(a)(2)(ix)(A)

10. POWER LEVEL 13 20.2203(a)(2)(ii) 1] 50.36(c)(1)(ii)(A)

ED 50.73(a)(2)(iv)(A)

E-50.73(a)(2)(x) 100 El 20.2203(a)(2)(iii)

[] 50.36(c)(2) 13 50.73(a)(2)(v)(A) 13 73.71(a)(4) 13 20.2203(a)(2)(iv) 13 50.46(a)(3)(ii) 13 50.73(a)(2)(v)(B) 13 73.71(a)(5) 13 20.2203(a)(2)(v)

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

[I 50.73(a)(2)(v)(C) 13 OTHER 13 20.2203(a)(2)(vi) 13 50.73(a)(2)(i)(B) 13 50.73(a)(2)(v)(D)

Specify in Abstract below or in June 26, 2011 at 1625:50 DST Three steam dump valves did not close as expected following the trip. Operators turn steam dump control off in accordance with Emergency Subprocedure ES-0.1 "Reactor Trip Response," a subprocedure of procedure E-0. Decay heat removal via the steam generator atmospheric relief valves was used in accordance with, Emergency Subprocedure ES-01.

D.

Other Systems or Secondary Functions Affected

No other systems or secondary functions were affected by this event.

E.

Method of Discovery

Control room alarms alert operators to the start of the event.

F.

Operator Actions

At the time of the turbine load rejection, control bank D rods automatically inserted 10 steps over 7 seconds from an initial rod height of 220 steps withdrawn. The reactor operator misdiagnosed the reason for the insertion and incorrectly placed the control rods in manual; the reactor tripped approximately 12 seconds later with all rods fully inserting. Had the rods remained.in automatic, rod height of control bank D would have been approximately 198 steps withdrawn at the time of the reactor trip. The difference in reactivity worth from the actual versus projected rod height of control bank D is not significant and is bounded by the accident analysis that assumes the control rod of highest worth remains fully withdrawn. The operator's actions of placing the rods in manual-was not in compliance with plant procedures and did not meet expectations set by Operations Management or Training. Once the reactor tripped, Operations entered emergency operations procedure E-0 "Reactor Trip or Safety Injection," as required by procedure. The steam dump system initially functioned as expected (all valves opened); afterwards, Operations manually turned off the steam dump system because three of the valves did not close when expected.

G.

Safety System Responses:

The plant responded as expected for the conditions of the reactor trip.

II1.

CAUSE OF THE EVENT

A.

Immediate Cause:

The immediate cause of the reactor trip was a turbine trip above 50 percent rated thermal power due to a failed turbine control AEH card.

B.

Root Cause:

The root cause of this event is being evaluated and will be provided in the supplement to this LER.

IV.

ANALYSIS OF THE EVENT

SQN Unit 1 was operating in Mode 1 at approximately 100 percent rated thermal power.

At 1615 DST, the reactor automatically tripped following a turbine trip from greater than 50 percent rated thermal power (P-9 interlock). Prior to the event, reactor coolant system (RCS) pressure was approximately 2234 pounds per square inch gauge (psig).

Following the turbine control card failure, the turbine valves closed and the turbine load reduction caused a rise in RCS temperature. This caused RCS volume to increase with a corresponding increase in RCS pressure. RCS pressure increased above 2335 psig, the setpoint of the pressurizer power operated relief valves (PORVs), before the reactor trip. Both PORVs opened briefly (approximately 3 seconds); approximately 7 seconds later, one PORV re-opened. Both PORVs reclosed at the proper pressure. Because of the turbine load reduction, RCS average temperature increased to approximately 586 degrees Fahrenheit, which is 3 degrees Fahrenheit above the Technical Specification (TS) 3.2.5 limit of 583 degrees Fahrenheit. The TS 3.2.5, Departure from Nucleate Boiling (DNB) Parameters, action statement requires that the parameter be restored to a value within its limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 5 percent of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The TS 3.2.5 limit for average temperature was exceeded for a period of approximately 7 seconds which is well within the TS 3.2.5 action statement requirement. Following the reactor trip, average temperature rapidly decreased as a result of loss of nuclear heat generation to approximately 545 degrees Fahrenheit, then increased to its no-load value of 547 degrees Fahrenheit. Following the trip, all safety related equipment operated as designed. The auxiliary feedwater system automatically actuated as expected on loss of the main feedwater pumps. The main feedwater pumps were available for recovery using approved plant procedures following the scram. Initially, the steam dump system functioned as expected (all valves opened). Subsequently, the steam dump system was manually turned off because three of the valves did not close when expected. As a consequence, decay heat removal was provided via the atmospheric relief valves.

11

V.

ASSESSMENT OF SAFETY CONSEQUENCES

As discussed in the above "Analysis of The Event," following the trip, all safety related equipment operated as designed, the auxiliary feedwater system actuated as expected and decay heat removal was provided using the atmospheric relief valves. In addition, the DNB parameter for average temperature momentarily increased to above the TS 3.2.5 limit to 586 degrees Fahrenheit. However, the parameter was restored to within the TS 3.2.5 limit well within the allowance of the associated action statement and all other DNB parameters remained within limits during this event. As a result, this event did not adversely affect the health and safety of plant personnel or the general public.

VI.

CORRECTIVE ACTIONS

A.

Immediate Corrective Actions

The immediate corrective action was to replace the failed AEH signal conditioning card.

B. Corrective Actions to Prevent Recurrence:

The corrective actions are being managed through the SQN Corrective Action Program.

Corrective actions to prevent recurrence are being developed to address the root cause and will be provided in the supplement to this LER.

VII.

ADDITIONAL INFORMATION

A.

Failed Components:

The failed component was a Westinghouse AEH signal conditioning card.

B.

Previous LERs on Similar Events:

A review of previous reportable events for the past three years did not identify any

previous similar events

C.

Additional Information

None.

D. Safety System Functional Failure:

This event did not result in a safety system functional failure in accordance with 10 CFR 50.73(a)(2)(v).

E.

Unplanned Scram with Complications:

This event did not result in an unplanned scram with complications. The main feedwater pumps were available for recovery using approved plant procedures following the scram. Although the steam dump system valves opened as expected, the steam dump system was later turned off because three of the valves did not close when expected. The atmospheric relief valves were available and were subsequently used for decay heat removal.

VIII.

COMMITMENTS

None.