ML18011B075

LER 95-011-01:on 951105-06,reactor Trip/Si Signal Generated & Unplanned ESF Actuation Occurred During Troubleshooting. Reactor Trip Caused by Equipment Failure.Component Replaced. W/951205 Ltr

ML18011B075
Person / Time
Site:	Harris
Issue date:	12/03/1995
From:	Donahue J, Verrilli M CAROLINA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
HPN-95-113, LER-95-011-01, LER-95-11-1, NUDOCS 9512110464
Download: ML18011B075 (10)

Text

PH.lCDR.I ACCEI.E RATED Rl l)S IY P ROC ESSli G 1

REGULATORY INFORMATION DISTRXBUTION SYSTEM (RIDS)

ACCESSION NBR:9512110464 DOC.DATE: 95/12/03 NOTARIZED: NO DOCKET FACIL:50-400 Shearon Harris Nuclear Power Plant, Unit 1, Carolina 05000400 AUTH. NAME AUTHOR AFFILIATION VERRILLI,M. Carolina Power & Light Co.

Carolina Power & Light Co. 'ONAHUE,J.W.

P RECIP.NAME RECIPIENT AFFXLIATION

SUBJECT:

LER 95-011-01:on 951105-06,reactor trip/SI signal generated

& unplanned ESF actuation occurred during troubleshooting. I Reactor trip caused by equipment failure. Component replaced.

W/951205 ltr. 0 DISTRIBUTION CODE: IE22T COPIES RECEIVED LTR ENCL SIZE TITLE: 50.73/50.9 Licensee Event Report (LER), Inciden&Rpt, etc.

NOTES:Application for permit renewal filed. 05000400 RECIPIENT COPIES RECIPIENT COPIES XD CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD2-1 PD 1 1 LE,N 1 1 INTERNAL: ACRS 1 1 AEOD/S PD/~B 2 2 AEOD/S PD/RRAB 1 1 ~FIZZ 'CENTER+ 1 1 NRR/DE/ECGB 1 1 +NRR/-'DE/EEI'B 1 1 NRR/DE/EMEB 1 1 NRR/DRCH/HHFB 1 1 NRR/DRCH/HICB 1 1 NRR/DRCH/HOLB 1 1 NRR/DRCH/HQMB 1 1 NRR/DRPM/PECB 1 1 NRR/DSSA/SPLB 1 1 NRR/DSSA/SRXB 1 1 RES/DSIR/EIB 1 1 RGN2 FILE Ol 1 1 D

EXTERNAL: L ST LOBBY WARD 1 1 LITCO BRYCE,J H 2 2 0 NOAC MURPHY,G.A 1 1 NOAC POORE,W. 1 1 NRC PDR 1 1 NUDOCS FULL TXT 1 1 C

U j M

N YOTE TO ALL"RIDS" RECIPIENTS:

PLEASE ICELP L'S TO REDUCE 4VXSTE! COSTA(:T'I'I IE DOCI:iIE4T CONTROL DESk, ROO! I P1.37 (EAT, 504.~OS' TO I.LIXII%ATE >'OI:R %ANIL FROil DISTRIDL'I IOY. LIS'I'S I:OR DOCUi!I:.X'I'S 'YOI ')Oi "I'L'I:.D!

FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 26 ENCL 26

Carolina Power & Light Company Harris Nuclear Plant PO Box 165 New Hill NC 27562 DEC 0 S 1995 U.S. Nuclear Regulatory Commission Serial: HNP-95-113 ATTN: NRC Document Control Desk 10CFR50.73

,Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT UNIT 1 DOCKET NO. 50-400 LICENSE NO. NPF-63 LICENSEE EVENT REPORT 95-011-00 Gentlemen:

In accordance with Title 10 to the Code of Federal Regulations, the enclosed Licensee Event Report is submitted. This report concerns an unplanned Reactor Trip/Safety Injection event and an ESF Actuation during subsequent testing.

Sincerely, J. W. D nahue General Manager Harris Plant MV Enclosure c: Mr. S. D. Ebneter (NRC - RII)

Mr. N. B. Le (NRC - PM/NRR)

Mr. D. J. Roberts (NRC - HNP) rtt5121 10464 rst51203 PDR ADQCK 05000400 8 , , PDR State Road 1134 New Hill NC

U. S. Nuclear Regulatory Commission Document Control Desk / LER 95-011 Page 2 CC: Ms. D. B. Alexander Ms. P. B. Brannan Mr. R. K. Buckles (LIS)

Mr. W. R. Campbell (BNP)

Mr. J. M. Collins Mr. John Paul Cowan Ms. S. D. Floyd Ms. S. F. Flynn Mr. H. W. Habermeyer, Jr.

Ms. T. A. Head (GLS File)

Mr. G. D. Hicks (BNP)

Mr. M. D. Hill Mr. W. J. Hindman Mr.'R. M. Krich (RNP)

Mr. P. M. Odom (RNP)

Mr. C. W. Martin (BNP)

Mr. R. D. Martin Admiral K. R. McKee Mr. J. P. McKone Mr. J. W. Moyer (RNP)

Mr. W. R. Robinson Mr. G. A. Rolfson Mr. C. T. Sawyer Mr. R. S. Stancil Mr. J. P. Thompson (BNP)

Mr. T. D. Walt HNP Real Time Training INPO NLS File: HI/A-2D (L. M. Randall)

Nuclear Records: HO-950241

NRC FORM 366 U.S. N CLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 (495) EXPIRES 04/30/96 ES)lMATEO BURDEN PER RESPONSE TO COMPLY WITH THI MANDATORY

)NFOPMAT)ON COuECT)ON REOUEST: 500 HRS. REPORTED LESSONS LEARNED ARE LICENSEE EVENT REPORT (LER) INCORPORATED INTO THE UCENSING PROCESS ANO FEO BACK TO WOUSTRZ.

FORWARD COMMENTS REGARDOIG BURDEN ESTIMATE TD THE )NFORMATION ANO RECORDS MANAGEMENT BRANCH IN F33L US. NUCLEAR REGUIATORY COMMLSSIO)L (See reverse for required number of WASHINGTON, OC 20555000), ANO TO THE PAPERWORK REDUCTION PRMECT Q)50.

digits/characters for each block) 0)04L OFFICE OF MANAGEMENT ANO BUDGET, WASHINGTON, DC 20503.

FACILITY NAME II) DOCKET NUMBER l3) PAGE I3)

Shearon Harris Nuclear Plant - Unit ¹ 1 50-400 1OF3 TITLE I4)

Reactor Trip/Safety Injection during Solid State Protection System testing due to the failure of a relay contact, and unplanned ESF actuation during troubleshooting following the Reactor Trip/Sl.

EVENT DATE (5) LER NUMBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED (6)

SEOUENTIAL REVISION FACIUTY NAME DOCKETNUMBER MONTH BAY YEAR MONTH OAY YEAR NUMBER NUMBER FACIUTY NAME DOCKETNUMBER 05 95 95 011 00 12 03 95 OPERATING THIS REPORT IS SUBMITTED PUR SUANT TO THE REQUIREMENTS OF 10 CFR B: (Chock one o r more) (11)

MODE (9) 20.2201(b) 20.2203(a) (2)(v) 50.73(a) (2) (I) 50.73(a) (2) (viii)

POWER 100 20.2203(a) (1) 20.2203(a)(3)(l] 50.73(a)(2) (ii) 50.73(a)(2)(x)

LEVEL (10) 20.2203(a) (2) (I) 20.2203(a)(3) (o) 50.73(o) (2) (iii) 73.71 20.2203(a) (2) (ii) 20.2203(a) (4) 50.73(a)(2)(iv) OTHER 20.2203(a)(2) (iii) 50.36(c)(1) 50.73(a)(2)(v) Specify in Abstract below or in NRC Form 366A 20.2203(o)(2)(iv) 50.36(c)(2) 50.73(a)(2)(vii)

LICENSEE CONTACT FOR THIS LER (12)

NAME TELEPHONE NUMBER (IiClude Area Code)

Michael Verrilli, Sr. Analyst - Licensing (919) 362-2303 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

REPORTABLE REPORTABLE CAUSE SYSTEM COMPONENT MANUFACTURER CAUSE SYSTEM COMPONENT MANUFACTURER TO NPROS TO NPROS RLY CNTR P297 SUPPLEMENTAL REPORT EXPECTED (14) EXPECTED MONTH OAY YEAR YES SUBMISSION (If yos, comploto EXPECTED SUBMISSION DATE). DATE (15)

ABSTRACT (Limit to 1400 spaces, i.o., approximately 15 singlo-spaced typowrinon lines) (16)

On November 5, 1995, A-Train Engineered Safety Feature Actuation System (ESFAS) slave relay surveillance testing was in progress. While performing the test portion that verifies the operability of the Main Steam Line Isolation Signal circuitry, a "Low Steam Line Pressure Reactor Trip/SI" signal was generated. This occurred when a contact failed to maintain continuity on the K809 relay and eliminated the "block" function to the "A" Main Steam Isolation Valve (MSIV), which resulted in the "A" MSIV closing. As the valve closed and turbine throttle valve position remained constant, increased steam load was carried by "B" and "C" Steam Generators (SGs). 'Ibis resulted in a pressure decrease in the "B" and "C" main steam lines due to increased steam flow. The rate compensation feature associated with the low steam line pressure SI initiated a low steam line pressure Reactor Trip/SI Signal for "C" steam line. Automatic systems responded as required and main control room operators took appropriate actions to stabilize the plant in Mode-3 (Hot Standby). An Unusual Event was declared at 0805 due to the ECCS actuation. The UE was then exited at 0912 based on the termination of Sl.

On November 6, 1995, continued ESFAS slave relay testing was being performed following the Reactor Trip/SI event from the previous day. During this testing the Auxiliary Feedwater (AFW) Flow Control Valves fully opened from their original throttled position. While this valve actuation was in accordance with system design, it was not recognized in the procedure. Thus, operators performing the test did not expect the valve operation. The test being performed is normally performed in Mode-1 with the AFW FCV's fully open. With the plant in Mode-3, the valves are throttled to control Steam Generator (SG) level. The operators recognized that the valves were opening and reestablished Steam Generator level control prior to exceeding the normal operating band. During the investigation of this actuation, a similar occurrence was identified where the AFW FCV's partially opened during testing on October 30, 1994. At that time the actuation was not recognized as reportable, but should have been.

NRC fORM SSSA US. NUCLEAR REGULATOllT COMMISSION t4.99I LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITT NAME it] DOCKET LER NUMGER IGI PAGE IS)

SEOUENTIAL REVISION NUMBER NUMBER Shearon Harris Nuclear Plant ~

Unit //1 50400 2 OF'3 95 - 011 00 TEXT Pl moro spice r's rorprief, oso oriÃ6roor copies ol A'RC Form SSQI IITt EVENT DESCRIPTION:

On November 5, 1995, with the plant operating in Mode-1 at 100% power, A-Train Engineered Safety Feature Actuation System (ESFAS, EIIS Code: JE) slave relay surveillance testing (OST-1044) was in progress. While performing section 7.3 of the test procedure, the portion of the test that verifies the operability of the Main Steam Line Isolation Signal circuitry, a "Low Steam Line Pressure Reactor Trip/Safety Injection (SI)" signal was generated. This occurred at 0737 hours0.00853 days <br />0.205 hours <br />0.00122 weeks <br />2.804285e-4 months <br />, when a contact failed to maintain continuity on the K809 relay and eliminated the "block" function to the "A" Main Steam Isolation Valve (MSIV, EIIS Code: SB), which resulted in the "A" MSIV closing. As the valve closed and turbine throttle valve position remained constant, steam load was carried by "B" and "C" Steam Generators (S/Gs). This resulted in a pressure decrease in the "B" and "C" main steam lines due to increased steam flow. The rate compensation feature associated with the low steam line pressure SI initiated a low steam line pressure Reactor Trip/Sl Signal for "C" steam line. Automatic safety equipment functioned as required except for the "closed" indication on one valve, SP-941, (Hydrogen Monitor isolation valve, EIIS Code: TK). It was later determined that the valve closed as required but experienced a position indication proximity switch problem.

Due to the SI, the Reactor Coolant System (RCS, EIIS'Code: AB) was being filled to solid plant conditions by the injection flow. The main control room operators proceeded through the Emergency Operating Procedure (EOP) network as required to secure SI. Although progress through the EOP procedure flow paths was timely, it did not prevent the plant from going solid. The solid plant conditions resulted in an increase in RCS pressure, which lifted a pressurizer PORV. A liquid/steam mixture was released by the pressurizer PORV to the Pressurizer Relief Tank (PRT). During this time the PORV actually cycled approximately 58 times. This resulted in an overpressure condition in the PRT and one of the two rupture disks ruptured as required to limit pressure in the tank. Approximately 1200 gallons of water from the PRT spilled over into the containment sump through the rupture disk. The control room staff took actions necessary to stabilize the plant and establish operation in Mode-3 (Hot Standby). An Unusual Event (UE) was declared at 0805 due to the ECCS actuation. The UE was then exited at 0912 based on termination of SI and stabilization of the plant in Mode-3. Operations personnel performing the test at the time of the Reactor Trip/SI signal took prompt and appropriate actions to preserve existing plant and system conditions, which were critical in identifying the deficient K809 relay contacts.

On November 6, 1995, continued ESFAS Slave Relay testing was being performed following the Reactor Trip/SI event from the previous day to verify the operability of other SSPS "block function" relays. At 1509 hours0.0175 days <br />0.419 hours <br />0.0025 weeks <br />5.741745e-4 months <br />, while performing section 7.3 of the A-Train, K635 Relay SI Block Circuit Test (OST-1044), the Auxiliary Feedwater (AFW, EIIS Code: BA) Flow Control Valves (FCV's) fully opened from their original throttled position. While this valve actuation was in accordance with system design, it was not recognized in the procedure. Thus, operators performing the test did not expect the valve motion. Based on the guidance of NUREG-1022, the opening of the AFW FCV's was considered an unplanned ESF actuation. OST-1044 is a quarterly interval surveillance test that is normally performed in Mode-1 with the AFW FCV's fully open. Due to the fact that these valves have been open during previous testing, no valve motion occurred. With the plant in Mode-3 following the Reactor Trip/SI, the valves were throttled to control Steam Generator (SG) level. The operators recognized that the valves were opening and reestablished Steam Generator level control prior to exceeding the normal operating band. During the investigation of this actuation, a similar occurrence was identified where the AFW FCV's partially opened during testing on October 30, 1994. At that time the actuation was not recognized as reportable, but should have been. Corrective actions for this occurrence involved a revision to the K640 Relay "Go Circuit" portion of the test procedure, but the "block circuit" portion of the test was not recognized, nor revised as needed.

CAUSE:

The cause of the Reactor Trip/SI was component/equipment failure. One set of contacts on the K809 "blocking" relay failed to maintain continuity as required when the block signal was generated. Additional analysis will be performed to determine the cause and failure mode of the relay contacts. The Unplanned ESF Actuation was caused by an inadequate procedure. Section 7.3 of OST-1044 did not provide proper guidance to ensure that operators performing the test were aware that the AFW Flow Control Valves would receive an open signal while testing the K635 block The October 30, 1994 AFW FCV actuation was also caused by a deficient procedure and was not reported due 'ircuit.

to a misunderstanding of reporting requirements on the part of Operations personnel that were involved with testing at the time.

NRC FOAM 3BBA US. NUCLEAR REGULATORY COMMISSION I4-BS)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME 0] DOCKET LER NUMBER IB) PAGE I3)

SEQUENTIAL REVISION NUMBER NUMBER Shearon Harris Nuclear Plant - Unit //1 05000400 3 OF 3 95 - 011 - 00 TEXT i)fmort sptotrs rtrfoitd. ost tdffrd)rrtfsopss of fff)C &m 3SQI (17)

SAFETY SIGNIFICANCE:

There were no significant safety consequences as a result of either event. The reactor trip and safety injection actuations were initiated by plant conditions resulting from an equipment failure. Safety systems responded as required to ensure plant safety and operators appropriately responded to verify system response and stabilize the plant in Mode-3. The Unplanned ESF Actuation was also a result of plant systems responding as designed. Though the opening of the AFW Flow Control Valves was unexpected by operators performing the testing, S/G level was never outside of the normal operating band.

PREVIOUS SIMILARLKRs:

No similar LERs have been reported.

CORRECTIVE ACTIONS COMPLETED:

1. The K809 Block Relay was replaced and A-Train ESFAS Slave Relay Testing was completed.

2. An engineering analysis, including a walkdown inspection was completed for the secondary plant to evaluate possible water hammer effects.

3. An inspection was performed in the containment building for the areas in the vicinity of the PRT that contain environmentally qualified equipment.

4. An inspection was performed to assess the condition of the PRT Rupture disks. This included a "foreign material exclusion" inspection for rupture disk fragments.

5. An evaluation was performed to assess the effects of the water (approximately 1,200 gallons) that spilled into the containment sump from the PRT rupture disk.

6. An inspection and evaluation was performed to assess the condition of the Pressurizer PORV that liAed and cycled during the event.

For each of the above inspections and/or evaluations (C/A's 2-6) no discrepancies were identified that would adversely effect or preclude plant startup.

CORRECTIVE ACTIONS PLANNED:

1. Additional investigation into the cause and failure mode will be performed for the K809 relay contacts that failed during testing.

2. Surveillance procedures OST-1044 and OST-1045 will be revised to provide operators guidance pertaining to the operation of AFW Flow Control Valves during K635 Relay testing.

3 Reporting requirements related to unplanned ESF actuations will be reemphasized with Operations personnel.

~

4. Additional review will be completed to ensure that similar AFW System testing deficiencies do not exist in other procedures.

EIIS CODES:

Engineered Safety Feature Actuation System, EIIS System Code: JE Component Code: RLY CNTR Main Steam System, EIIS System Code: SB Component Code:ISV Hydrogen Monitor Isolation Valve, EIIS System Code: TK Component Code: ISV Reactor Coolant System, EIIS System Code: AB Auxiliary Feedwater System, EIIS System Code: BA Component Code: FCV

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