05000261/LER-2006-001, Re Manual Reactor Trip Due to Failure of a Turbine Governor Valve Electro-Hydraulic Control Card

Re Manual Reactor Trip Due to Failure of a Turbine Governor Valve Electro-Hydraulic Control Card

ML063630040
Person / Time
Site:	Robinson
Issue date:	12/21/2006
From:	Kapopoulos E Progress Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RNP-RA/06-0120 LER-06-001-00
Download: ML063630040 (5)

LER-2006-001, Re Manual Reactor Trip Due to Failure of a Turbine Governor Valve Electro-Hydraulic Control Card

Event date:
Report date:
Reporting criterion:	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)(iv)(A), System Actuation 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 10 CFR 50.73(a)(2)(iv)(B), System Actuation
2612006001R00 - NRC Website

text

10 CFR 50.73 Progress Energy Serial: RNP-RA/06-0120 DEC 2 1 2006 United States Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261/LICENSE NO. DPR-23 LICENSEE EVENT REPORT NO. 2006-001-00 MANUAL REACTOR TRIP DUE TO FAILURE OF A TURBINE GOVERNOR VALVE ELECTRO-HYDRAULIC CONTROL CARD Ladies and Gentlemen:

The attached Licensee Event Report is submitted in accordance with the requirements of 10 CFR 50.73. Should you have any questions regarding this matter, please contact Mr. C. T. Baucom, Supervisor - Licensing/Regulatory Programs, at (843) 857-1253.

Sincerely, Ernest J. Kapopoulos, Jr.

Plant General Manager H. B. Robinson Steam Electric Plant, Unit No. 2 CTB/grs Attachment c:

Dr. W. D. Travers, NRC, Region 11 Mr. C. P. Patel, NRC, NRR NRC Resident Inspector, HBRSEP Progress Energy Carolinas, Inc...,,

Robinson Nuclear Plant 3581 West Entrance Road Hartsville, SC 29550

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NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 EXPIRES 6-30-2007 (6-2004)

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

3. PAGE H. B. Robinson Steam Electric Plant, Unit No. 2 05000261 1 OF 4

4. TITLE Manual Reactor Trip Due to Failure of a Turbine Governor Valve Electro-Hydraulic Control Card

5. EVENT DATE

6. LER NUMBER

7. REPORT DATE

8. OTHER FACILITIES INVOLVED FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR SEQUENTIAL REV NUMBER NO MONTH DAY YEAR 05000 FACILITY NAME DOCKET NUMBER 10 25 2006 2006 - 001 - 00 12 21 2006 05000

9. OPERATING MODE

11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)

El 20.2201(b)

El 20.2203(a)(3)(i)

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

El 50.73(a)(2)(vii)

[E] 20.2201(d)

El 20.2203(a)(3)(ii)

El 50.73(a)(2)(ii)(A)

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

El 20.2203(a)(1)

El 20.2203(a)(4)

El 50.73(a)(2)(ii)(B)

E] 50.73(a)(2)(viii)(B)

[I 20.2203(a)(2)(i)

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

[I 50.73(a)(2)(iii)

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

10. POWER LEVEL El 20.2203(a)(2)(ii)

El 50.36(c)(1)(ii)(A) 0 50.73(a)(2)(iv)(A)

E] 50.73(a)(2)(x)

E] 20.2203(a)(2)(iii)

El 50.36(c)(2)

El 50.73(a)(2)(v)(A) 0l 73.71(a)(4) 100%

E] 20.2203(a)(2)(iv)

El 50.46(a)(3)(ii)

El 50.73(a)(2)(v)(B)

[] 73.71 (a)(5)

E] 20.2203(a)(2)(v)

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

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

E] OTHER Specify in Abstract below or in El 20.2203(a)(2)(vi)

El 50.73(a)(2)(i)(B)

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

I.

DESCRIPTION OF EVENT

At 0247 hours0.00286 days <br />0.0686 hours <br />4.083995e-4 weeks <br />9.39835e-5 months <br /> on October 25, 2006, with H. B. Robinson Steam Electric Plant (HBRSEP),

Unit No.

2, in Mode 1 at 100% power, control room operators responded to alarms received for steam flow greater than feed flow on all three steam generators [EIIS System:Component AB:SG].

Turbine first stage pressure [EIIS System:Component TA:PI] indicated 0 psig.

Three of the four turbine governor valves [EIIS System:Component TA:FCV] indicated closed, the fourth indicated an intermediate position.

The control rods [EIIS System AA] began automatically inserting, as expected, due to the difference between reactor coolant average temperature and reference temperature, which is based on turbine first stage pressure.

The reactor operator placed the rod control system in manual based on a misdiagnosis that a turbine first stage pressure channel had failed, since indicated pressure was 0 psig.

The net megawatt recorder indicated that electrical generation had rapidly reduced from 742 to 0 megawatts.

The control room operators diagnosed the event as a 100% load rejection and initiated a manual reactor trip at 0248 hours0.00287 days <br />0.0689 hours <br />4.100529e-4 weeks <br />9.4364e-5 months <br />, 68.8 seconds following the start of the event.

A pressurizer power operated relief valve (PORV)

[EIIS System:Component AB:RV] lifted and reseated during the event.

As a result of lifting, the pressurizer PORV has a current leakage rate of approximately 0.03 gpm.

This leakage is being monitored in accordance with site procedures.

The auxiliary feedwater system (AFW)

[EIIS System BA] also initiated during the event, as expected, due to the low steam generator level that resulted from rapid closure of the turbine governor valves.

II.

CAUSE OF EVENT

Investigation of this event was conducted using the Corrective Action Program and documented in Significant Nuclear Condition Report (NCR) 210311.

This investigation found that the root cause of this event was a component failure in one of the five logic cards in the turbine governor valve electro-hydraulic control (EHC) system [EI!S System JJ].

The following five EHC system logic cards were replaced:

1.

Digital-to-Analog (D/A) Converter

2.

Up-Down Counter

3.

High-Thresold Logic (HTL) Gate 6

4.

Input Expander

5.

HTL Gate 1 Technical representatives from Westinghouse and Data Techniques (a Westinghouse vendor) tested the removed logic cards.

Three of the five cards were found to have anomalies.

A weak transistor on the up-down counter card may have caused the counter to count incorrectly and could have caused the condition experienced.

The vendor recommended that this card be replaced with a new style card.

The D/A converter card was found to be slightly out of calibration.

The vendor determined that the small offset would not have an effect on the system.

The HTL Gate 6 card had an intermittent failure of a single-gate circuit which is not believed to have contributed to this event.

III.

ANALYSIS OF EVENT

The condition described in this Licensee Event Report is reportable in accordance with 10 CFR 50.73(a)(2)(iv)(A),

any event or condition that resulted in manual or automatic actuation of any of the systems listed in 10 CFR 50.73(a)(2)(iv)(B).

The health and safety of the public and plant personnel were not impacted by this event.

The required safety functions were maintained and the operating parameters of the plant remained within required safety limits.

The steam generator PORVs and a pressurizer PORV lifted as designed to relieve system pressure.

Based on a review of plant data, if the control room operators had taken no action, the automatic trip logic would have been satisfied and would have tripped the unit within 1 to 2 seconds of the manual trip.

Given this short duration, no additional safety significance would have resulted if the manual trip had not been initiated.

The plant responded as designed and the operating crew manually tripped the reactor in response to the conditions they observed.

The AFW system initiated during the event, as expected, due to the low steam generator level that resulted from rapid closure of the turbine governor valves.

During the event the reactor operator placed the rod control system in manual based on a misdiagnosis that a turbine first stage pressure channel had failed.

Turbine first stage pressure channel failure indications were present during the initial stages of the event and some of the indications for a first stage pressure channel failure and a 100% load rejection are common.

Once the load rejection was properly diagnosed, the crew took the appropriate and conservative actions to place the plant in a safe condition.

The Updated Final Safety Analysis Report, Section 15.2.2, Loss of External Load, includes the assumption that the rod control system is in manual at the start of the event.

Additionally, the investigation of this event included simulator modeling to understand the impact on plant parameters due to having the rod control system in manual.

This modeling found the impact to be minimal, and further found that the pressurizer PORV would have lifted if the rod control system had been left in automatic.

Therefore, placing the rod control system in manual did not significantly affect the event.

IV.

CORRECTIVE ACTIONS

Immediate Corrective Actions:

The five logic cards in the EHC system were replaced.

Operators have reviewed information and received simulator training on the proper response to turbine governor valve failures with regard to placing the rod control system in manual.

Planned Corrective Actions

In accordance with vendor recommendations, the EHC system up-down counter card will be replaced with a new style card (1B51049-101) during the next refueling outage, which is currently scheduled to begin in April 2007.

V.

ADDITIONAL INFORMATION

Failed Component Information:

The failed EHC system card was a Westinghouse up-down counter card, Part Number 2822A82.

Previous Similar Events

Licensee Event Report 1998-003-00 On April 25, 1998, at approximately 1334 hours0.0154 days <br />0.371 hours <br />0.00221 weeks <br />5.07587e-4 months <br />, steam generator (SG) level deviation alarms were received on all three SGs.

Control room operators observed that turbine first stage pressure was rapidly decreasing and reactor control rods were automatically inserting.

The control board operator placed the rod control system in manual based on a misdiagnosis that a turbine first stage pressure transmitter had failed low.

Reactor coolant system (RCS) pressure increased to approximately 2335 psig resulting in opening of a pressurizer PORV.

The initiating event was the inadvertent closing of the turbine governor valves, which reduced steam flow and resulted in an increase in steam generator pressure.

The increased steam generator pressure resulted in shrinkage of the steam generator levels below the reactor trip setpoint.

Troubleshooting of the EHC system, which controls the turbine governor valves, was performed by maintenance personnel assisted by a vendor technical representative.

Various failure modes of the EHC system were investigated.

Of the EHC system failures investigated, only changes to the impulse channel (which monitors turbine first stage pressure) corresponding to a pressure spike of approximately 35 psig, were found to duplicate the turbine governor valve response recorded during the event.

No equipment failures were identified that would have resulted in the initiation of the event and the root cause of the event could not be definitively determined.'A