ML18012A174

LER 95-011-01:on 951105,reactor Trip/Safety Injection Occurred During Solid State Protection Sys Testing Due to Failure of Relay Contact & Unplanned ESF Actuation Occurred During troubleshooting.W/960322 Ltr

ML18012A174
Person / Time
Site:	Harris
Issue date:	03/22/1996
From:	Donahue J, Verrilli M CAROLINA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
HNP-96-044, HNP-96-44, LER-95-011, LER-95-11, NUDOCS 9603250308
Download: ML18012A174 (5)

Text

CATEGORY REGULATORY INFORMATION DISTRIBUTION SYSTEM 1y (RIDS)

ACCESSION NBR:9603250308 DOC.DATE: 96/03/22 NOTARIZED: NO DOCKET FACIL:50-400 Shearon Harris Nuclear Power Plant, Unit 1, Carolina 05000400 AUTH. NAME AUTHOR AFFILIATXON VERRILLI,M. Carolina Power & Light Co.

DONAHUE,J.W. Carolina Power 6 Light Co.

RECIP.NAME RECIPIENT AFFILIATION

SUBJECT:

LER 95-011-01:on 951105,reactor trip/safety injection during SSPS testing due to failure of a relay contact a unplanned ESF actuation. Caused by component/equipment failure.

Block relay replaced & testing completed.W/960322 ltr.

DISTRIBUTION CODE: IE22T COPIES RECEIVEDrLTR I ENCL TITLE: 50.73/50.9 Licensee Event Report (LER), Incident Rpt, etc.

J SIZE:

NOTES:Application for permit renewal filed. 05000400 Q RECIPXENT COPIES RECIPIENT COPXES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD2-1 PD 1 1 LE,N 1 1 INTERNAL: ACRS 1 1 AEOD/SPD RAB 2 2 AEOD/SPD/RRAB 1 1 N 1 1 NRR/DE/ECGB 1 1 R DZ EEGB 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 D RES/DSIR/EIB 1 1 RGN2 FILE 01 1 1 0

EXTERNAL: L ST LOBBY WARD 1 1 LITCO BRYCEPJ H 2 2 NOAC MURPHYPG A 1 1 NOAC POORE,W. 1 1 C NRC PDR 1 1 NUDOCS FULL TXT 1 1 N

NoTE To nr.r Rins" r<EczpiENT :

PLEASE HELP US TO REDUCE WASTEI CONTACT THE DOCUMENT CONTROL 5D-5(EXT DESKS'OOM OWFN 415 2083) TO ELIMINATE YOUR NAME FROM DISTRiBUTION LISTS FOR DOCUMENTS YOU DON'T NEED I PULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 26 ENCL 26

Carolina Power & Light Company Harris Nuclear Plant PO 8ox 165 New Hill NC 27562 HAR 2 U.S. Nuclear Regulatory Commission 2 lgy6 Serial: HNP-96-044 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 95-011-01 Gentlemen:

H In accordance with Title 10 to the Code of Federal Regulations, the enclosed supplement to Licensee Event Report 95-011 is submitted. This supplement provides additional information pertaining to the failure mode of the relay contacts that failed to remain closed and caused the unplanned Reactor Trip/Safety Injection event on November 5, 1995.

Sincerely, J. W. Donahue 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) 9603250308 960322 g(PP PDR ADOCK 05000400 PDR State Road H3A New Hill NC

N C FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150.0104 EXPIRES 04/30/96 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS MANOATORT INfORMATION COLLECTION REDDEST: 50D HRS. REPORTED LESSONS LEARNED ARE LICENSEE EVENT REPORT (LER) INCORPORATED UITO THE UCENSING PROCESS ANO f<D BACK TO INDUSTRY.

fORWARO COMMENTS REGARDING BURDEN ESTIMATE TO THE INfORMATIOH ANO RECORDS MANAGEMENT BRANCH IT% f33L US. NUCLEAR REGULATORT COMMISSIOIL (See reverse for required number of WASHB(GTOIL OC 205550001, ANO TO THE PAPERWORK REDUCTION PROJECT l3150.

digits/characters for each block) OIO(L Off(CE Of MANAGEMENT ANO BUDGET, WASHUIGTOH. OC 20503.

FACIL(TY NAME (1) DOCKET NUMBER (21 PAGE (3I Shearon Harris Nuclear Plant - Unit I)( 1 50-400 1 OF 3 TITLE (41 Reactor Trip/Safety Injection during Solid State Protection System testing due to the failure of a relay contact, end unplanned ESF actuation during troubleshooting following the Reactor Trip/Sl.

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

SEQUENTIAL FACILITYNAME DOCKET NUMBER REVISION MONTH OAY MONTH OAY YEAR NUMBER NUMBER FACIUTY NAME DOCKET NUMBER 05 95 95 011 01 03 29 96 OPERATING THIS REPORT IS SUBMITTED PUR SUANT TO THE REQUIREMENTS OF 10 CFR Bi (Check one o r more) (11)

MODE (9) 20.2201(b) 20.2203(B) (2) {v) 50.73(B)(2)(I) 50. 73(B) (2) (viii)

POWER 100 20.2203(B) (1) 20.2203(B) {3){I) 50.73(B)(2)(ii) 50.73(B) (2) {x)

LEVEL (10) 20.2203(B)(2)(l) 20.2203(B)(3) (ii) 50.73(B){2)(iii) 73.71 20.2203(a) {2)(ii) 20.2203(B)(4) 50.73(B) (2)(iv) OTHER 20.2203(B) (2)(iii) 50.36(c)(1) 50.73(B)(2)(v) Specify in Abstract below or in NRC Form 3BBA 20.2203(B) (2)(iv) 50.36(c)(2) 50.73(B)(2)(vii)

LICENSEE CONTACT FOR THIS LER (12I TELEPHONE NUMBER (Inclod<<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 STSTEM COMPONENT MANUFACTURER TO NPROS TO NPRDS RLY JE P297 CNTR SUPPLEMENTAL REPORT EXPECTED <14) EXPECTED MONTH OAY YEAR YES SUBMISSION

{Ifyes, comple(B EXPECTED SUBMISSION DATE). X NO DATE (15)

ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten 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). This resulted in a pressure decrease in the "B" and "C" main steam lines due to increased steam fiow. The rate compensation feature associated with the low steam line pressure Sl 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, l995, 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-l 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.

This LER revision provides additional information from laboratory failure analysis of the K809 relay contacts which failed to remain closed and caused the unplanned Reactor Trip/Safety Injection event on November 5, l995.

N I

NRC PORM 366A US. NUCLEAR REGUIATORY COMMISSIOII F4.9Q LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME iii DOCKET LER NUMRER I6) PAGE n)

SEOUENTIAL REVLTION NUMBER NUMRER Shearon Harris Nuclear Plant ~ Unit //1 50400 2 OF 3 95 - 011 - 01 TEXT iN mao oooo e myoid oso odif(illooo4o ol NRC &an 36W FIT)

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 (Sl)" 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 Sl initiated a low steam line pressure Reactor Trip/SI 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 plan 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 (PR'Q. 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 Sl 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 circuit. The October 30, 1994 AFW FCV actuation was also caused by a deficient procedure and was not reported due to a misunderstanding of reporting requirements on the part of Operations personnel that were involved with testing at the time.

Additional investigation was performed by Potter & Brumfield Products Division to determine the failure mechanism for the K809 relay contact failure. This investigation revealed that the low current load from the SSPS test circuit did not have enough arc energy to burn through the surface tarnish, which is naturally present on silver contacts of this age. Corrective actions have been taken to procedurally address this conditon.

NRC FORM 366A US. NUCLEAR REGUIATORT COMMISSION

+BBI LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACIIITT NAME III OOCXET EER NUMBER (6) PAGE ui SEOUENFIAE REVISION NUMBER NUMBER Shearon Harris Nuclear Plant ~

Unit ¹1 50400 3 OF 3 95 - 011 - 01 TEXT pf evrv spvrv ir r Ovvaf. vsv Addled vvpiev vl ANC Fan 3aQI iiii SAFETY SICNIFICANCE:

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 SIMILAR LERs:

No similar LERs have been reported.

CORRECTIVE ACTIONS COMPLETED:

l. 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.

Note: 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.

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

8. Reporting requirements related to unplanned ESF actuations were reemphasized with Operations personnel.

9. Additional review was completed to ensure that similar AFW System testing deficiencies do not exist in other procedures.

10. Additional investigation into the cause and failure mechanism for the K809 relay contacts was performed. The results are provided in this supplement.

Testing procedures OST-1044 and OST-1045 were revised to require cycling the test switches several times between the "NORMAL"and "PUSH TO TEST" positions prior to depressing the test switch. This will help ensure that test switch contacts are wiped sufficiently to remove any tarnish prior to opening of the slave relay contact.

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