ML18012A261

LER 96-002-07:on 960508,failed to Properly Perform Tech Spec Surveillance Testing.Caused by Personnel Error.Personnel Counseled,Procedure Revised & Continued in-program Tech Spec Testing Program review.W/960607 Ltr

ML18012A261
Person / Time
Site:	Harris
Issue date:	06/07/1996
From:	Donahue J, Verrilli M CAROLINA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
HNP-96-095, HNP-96-95, LER-96-002-04, LER-96-2-4, NUDOCS 9606110405
Download: ML18012A261 (15)

Text

CATEGORY 3y REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9606110405 DOC.DATE: 96/06/07 NOTARIZED: NO DOCKET FACIL:50-400 Shearon Harris Nuclear Power Plant, Unit 1, Carolina 05000400 AUTH. NAME AUTHOR AFFILIATION VERRILIiM. Carolina Power s Light Co.

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

RECIP.NAME RECIPIENT AFFILIATION 4

SUBJECT:

LER 96-002-07:on 960508,failed to properly perform Tech Spec surveillance testing. Caused by'ersonnel error. Personnel counseled, procedure revised & continued in-program Tech Spec testing program review.W/960607 ltr.

DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR ENCL SIZE:

TITLE: 50.73/50.9 Licensee Event Report (LER), Incident Rpt, etc.

NOTES:Application for permit renewal filed. 05000400 G

0 RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD2-1 PD 1 1 LE,N 1 1 INTERNAL: ACRS 1 1 AE 2 2 AEOD/SPD/RRAB 1 1 ILE CENTER 1 1 NRR/DE/ECGB 1 '1 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 D 1 1 NRR/DSSA/SRXB 1 1

'ES/DSIR/EIB 1 1 RGN2 FILE 01 1 1 0 EXTERNAL: L ST LOBBY WARD 1 1 LITCO BRYCE,J H 2 2 NOAC MURPHY,G.A 1 1 NOAC POORE,W. 1 1 NRC PDR 1 1 NUDOCS FULL TXT 1 1 N

NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM OWFN 5D-5(EXT. 4l5-2083) TO ELIMINAT" YOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!

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 JUN 7 1996 U.S. Nuclear Regulatory Commission Serial: HNP-96-095 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 96-002-07 Gentlemen:

In accordance with Title 10 to the Code of Federal Regulations, the enclosed revision to Licensee Event Report 96-002 is submitted. This revision reports additional Technical Specification Testing deficiencies identified during the on-going Technical Specification testing program review.

Sincerely, J. W. Donahue General Manager Harris Plant MV Enclosure c: Mr. J. B. Brady (NRC - HNP)

Mr. S. D. Ebneter (NRC - RII)

Mr. N. B. Le (NRC - PM/NRR) 9606ii0405 960607 05000400 PDR ADOCK P Dl t 12000/t State Road ii34 New Hill NC

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150.0104 ILB5] EXPIRES 04/30/96 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS MANDATORY INfORMATION COLLECTION REDDEST: 500 HRS. REPORTED lESSONS LEARNED ARE LICENSEE EVENT REPORT (LER) (NCORPORATED INTO THE UCEHS(NG PROCESS AHD FED BACK TO (HOUSTRY.

fORWARO COMMENTS REGARDING BURDEN ESTIMATE TO THE U(FORMATION AND RECORDS MANAGEMENT BRANCH IT 6 f331 US. NUClEAR REGULATORY COMMISSIOH.

(See reverse for required number of WASHINGTON. OC 20555000l, ANO TO THE PAPERWORK REDUCTION PR(uECT 0(50.

digits/characters for each block) 0(0(1 OFFICE OF MANAGEMENT AND BUDGET, WASHWGTON. OC 20503.

FAOILrrY NAME (tl DOCKET NUMBER (2l PAGE (3)

Harris Nuclear Plant Unit-1 50-400 1 OF 10 TITLE (4)

Failure to properly perform Technical Specification surveillance testing.

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

FACIUTY NAME DOCKETNUMBEA SEOUENTIAL REVISION MONTH OAY YEAR YEAR MONTH OAY YEAR NUMBER NUMBER 05000 FACILITYNAME DOCKET NUMBER 8 96 96 002 07 6 7 96 05000 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) X 50.73(a)(2)(i) 50.73(a) (2) (viii)

POWER 20.2203(a) (1) 20.2203(a) (3) (i] 50.73 (a) (2) (ii) 50.73(a)(2)(x) 1004 LEVEL (10) 20.2203(a) (2) (i] 20.2203(a)(3)(ii) 50.73(a)(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) SpecI(y In Abstract below or ln NRC Form 366A 20.2203(a)(2)(iv) 50.36(c) (2) 50.73(a)(2)(vii)

LICENSEE CONTACT FOR THIS LER (12)

NAME TELEPHONE NUMBER (Include )Vea Codel Michael Verrilli Sr. Analyst - Licensing (9191 362-2303 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DES CRIBED IN THIS REPORT (13)

REPORTABLE REPORTABIE CAUSE SYSTEM COMPONENT MANUFACTURER CAUSE SYSTEM COMPONENT MANUFACTURER TO NPROS TO NPROS SUPPLEMENTAL REPORT EXPECTED (14) MON'TH OAY YEAR EXPECTED YES SUBMISSION X (If yos, complete EXPECTED SUBMISSION DATE). NO DATE (15) 6 15 96 ABSTRACT (Limit to 1400 spaces, i.o., approximately 15 single.spacod typowritton lines) (16)

On January 17, 1996, during a Technical Specification testing program review, a failure to perfo~ required surveillance testing during a planned maintenance outage in October, 1994 was identified. Specifically, on October 30, 1994, the plant was shut down and taken to Mode-5 (Cold Shutdown). This outage exceeded 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and per Technical Specification requirements Engineered Safety Feature slave relay testing was required. To satisfy this requirement OST-1083 and OST-1084 were satisfactorily completed, however the slave relay circuits for the CSIP Alternate Mini-Flow Isolation Valves, 1CS-746 and 1CS-752 were not tested due to an error that occurred during a procedure revision in June 1993.

In September 1992, a plant modification was completed on the Charging/Safety Injection Pump (CSIP) Alternate Mini-Flow System that necessitated revisions to Operations Surveillance Test procedures OST-1083 and OST-1084. These revisions were completed in June 1993 and removed the slave relay testing for CSIP Alternate Mini-Flow Isolation Valves (1CS-746 and 1CS-752) from OST-1083 and OST-1084 and transferred the testing requirement to procedure OST-1809. The cause of the Technical Specification violation was personnel error during the June 1993 procedure revision process for OST-1083 and OST-1084. OST-1809 was successfully performed during Refueling Outage 6 on September 8, 1995. This test verified the operability of these circuits, thus no immediate corrective action was required upon identification of the deficiency. Additional corrective actions included personnel counseling, appropriate procedure revisions and the continuation of an in-progress Technical Specification testing program review.

One additional Technical Specification testing deficiency (item 24) was identified during the on-going comprehensive Technical Specification testing program review and is being reported by this supplemental report.

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION F(-96(

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (II OOCKET LER NUMBER (6) PAGE(3)

SFOUENTIAL REVISION NUMBER NUMBER Shearon Harris Nuclear Plant ~

Unit //I 50400 2 OF 10 96 - 002 - 07 TEXT pf mort sposofs rorfvpod, vso oddi(bool sopAts of HRC form 366(f (IT(

EVENT DESCRIPTION:

On January 17, 1996, a failure to perform Technical Specification surveillance testing during an October 1994 planned maintenance outage was identified. The identification of this condition was a result of an on-going comprehensive Technical Specification testing program review that began in September 1995 following submittal of LER 95-07.

Specifically, on October 30, 1994, the plant was shut down and taken to Mode-5 (Cold Shutdown) for a planned maintenance outage. This outage exceeded 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and therefore, per Technical Specification 4.3.2.1 requirements, Engineered Safety Feature (ESF) slave relay testing was required for the 13 ESF relays delineated in Table 4.3-2 that had not been tested in the last 92 days due to being at full power operations. To satisfy this requirement OST-1083 and OST-1084 were completed on November 3, 1994. However,'he slave relay circuits for the CSIP Alternate Mini-Flow Isolation Valves (ICS-746 and ICS-752, EIIS Code BQ-ISV) were not tested due to an error that occurred during procedure revisions performed in June 1993 on Engineered Safety Feature (ESF) 18-Month Slave Relay Operations Surveillance Test procedures OST-1083 and OST-1084. This error involved inappropriately removing the slave relay testing for the 1CS-746 and ICS-752 circuits from OST-1083 and OST-1084 and transferring the testing requirement to procedure OST-1809 (Refueling Water Storage Tank switchover to the Containment sumps), which is also an 18-month ESF response time test. OST-1809 was not performed following the October maintenance outage, thus resulting in the testing omission and Technical Specification violation.

During the investigation of this event, personnel performing the Technical Specification testing program review verified that OST-1809 had been successfully performed during Refueling Outage 6 on September 8, 1995, which verified the operability of the affected circuits.

The June 1993 revisions to OST-1083 and OST-1084 were performed to incorporate a plant modification (PCR-6547) on the Charging/Safety Injection Pump (CSIP) Alternate Mini-Flow lines. This modification removed the previously installed relief valves and provided an "open" signal to 1CS-746 and 1CS-752 upon receipt of a Safety Injection signal.

This condition was determined to be a violation of the Technical Specification surveillance test periodicity requirement and is being reported per 10CFR50.73(a)(2)(i)(b).

The following additional Technical Specification testing deficiencies have been identified by the on-going comprehensive Technical Specification testing program review:

Slave relays (K635 & K640) for the Auxiliary Feedwater (AFW) Flow Control Valves (EIIS BA-FCV) were not tested within their required quarterly surveillance interval following Refueling Outage (RFO) 5 in 1994 through RFO 6 in October 1995. This was a result of inadequate technical reviews associated with the plant modification (PCR-6502) that installed the auto-open signal to these valves. PCR-6502 specified the slave relay surveillance testing interval as once per 18 months per Technical Specification 4.7.1.2, but failed to identify the quarterly requirement contained in Technical Specification 4.3.2.1.. Both of these relays were subsequently tested following RFO 6, which verified their operability. This condition was identified on February 1, 1996 with the plant operating in Mode-1 at 100% power.

Testing for manual Safety Injection (SI, EIIS-BQ) and Containment Spray (CS, EIIS-BE) actuation has not fully tested all switch contacts within the required 18 month surveillance test interval per Technical Specification 4.3.2.1.

The Operations Surveillance Test Procedures (OST-1825 & OST-1826) that verify the operability of, the actuation circuits, only test one of the two manual actuation switches for each signal once per 18 months, thus resulting in the Technical Specification violation. The alternate test switch has been satisfactorily tested approximately once per 36 months due to test performance staggering. Based on this previous testing, the SI and CS switches are currently operable. However, one set of CS switches will become inoperable on March 3, 1996 and one SI switch will become inoperable on March 19, 1996. This condition has existed since initial development of OST-1825 & OST-1826 and was identified on February 12, 1996 with the plant operating in Mode-I at 100% power.

RM A (4-95)

'NRC FORM 366A US. NUCLEAR REGULATORT COMMISSION I4.96I LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITT KAME ni OOCXET LER NUMBER I6) PAGE a SEOUENTIAL REVISION NUMBER NUMBER Shearon Harris Nuclear Plant ~

Unit ¹I 50400

- 3 OF 10 96 002 - 07 TEXT P/ rooro opooo ro revvvaf. vro orrrrrrloool oopr'oo ol h'RC foror 36N 0 TI Quarterly surveillance testing was not performed for the "B" Charging/Safety Injection Pump Mini-Flow Isolation Valve (1CS-196) Slave Relay (K601) (EIIS Code BQ-ISV,RLY) during Refueling Outage 6 in 1995. This slave relay should have been tested during the performance of OST-1086 in August of 1995, but the "B" CSIP was inoperable and 1CS-196 was under clearance at the time of the test. This was noted in the procedure, but the test was considered satisfactory by the operations control room staff due to a deficiency in the acceptance criteria section of OST-1086. The acceptance criteria, which was changed during a recent procedure revision, did not specify the need to test the 1CS-196 slave relay, so no Equipment Inoperable Record was generated for tracking purposes, which would have required subsequent testing. The operability of the slave relay was successfully verified by performing OST-1086 in December 1995. This condition was identified on February 17, 1996 with the plant operating in Mode-1 at 100% power.

Eleven Maintenance Surveillance Test (MST) Procedures were identified that did not verify automatic isolation of the effluent pathway on a loss of power for the associated radiation monitor (EIIS Code IL-MON) as required by Technical Specification 4.3.3.10. This affected six radiation monitors and their related pathways. This testing omission was created when the MSTs were inappropriately revised in 1993 in an effort to eliminate procedure steps that were thought to be redundant and unnecessary. Inattention to detail and an incomplete understanding of the Technical Specification testing requirements for the radiation monitor circuitry caused the deficiency. The testing requirements for these radiation monitors were removed from Technical Specifications in May 1995, but due to the time frame of the deficiency, the condition constitutes a violation and is therefore included in this LER supplement.

This condition was identified on February 14, 1996 with the plant operating in Mode-1 at 100% power.

Proper overlap testing has not been performed for an actuation of the Fuel Handling Building Emergency Ventilation System (EIIS-VG) originating from a high radiation alarm signal, as per Technical Specification 4.9.12.

OST-1048 tests this feature, but does not include a particular section of cable (¹12913M-SA) between the North Spent Fuel Pool Radiation Monitor (RM-1FR-3567A-SA, EIIS Code IL-MON) and the South Spent Fuel Pool Radiation Monitor (RM-IFR-3564A-SA). This condition has existed since initial development of OST-1048 and was identified on February 19, 1996 with the plant operating in Mode-1 at 100% power. Technical Speciiflication compliance is satisfied by the current operability of other FHB radiation monitors.

Logic testing for the Control Room Emergency Filtration Fans (R-2 "A" and "B", EIIS Code VI-FAN) has inadequately verified all automatic fan start signals associated with a Control Room Isolation Actuation. OST-1825 and 1826 have properly verified that the fans automatically start upon receipt of a safety injection actuation signal, however, the operability of a parallel circuit path that provides an automatic start signal on high radiation has not been verified during past testing. This condition has existed since initial development of the applicable surveillance procedures and constitutes a violation of Technical Specification 4.3.2.1. surveillance requirements. This condition was identified on February 22, 1996 Lvith the plant operating in Mode-1 at 100% power.

Logic testing for the Reactor Auxiliary Building Electrical Equipment Protection Room Inlet Isolation Valves (1CZ-7 & 1CZ-8, EIIS Code VF-V) has not properly verified the operability of each actuation circuit path. These valves receive a thermal overload bypass signal from two parallel sources; a Control Room Ventilation Isolation Signal and a signal from the Emergency Safeguards Sequencer. Surveillance testing has properly verified the operability these circuits from the Emergency Safeguards Sequencer. However, a portion of the thermal overload bypass circuit for Control Room Isolation has not been verified when the signal is generated from high radiation.

This constitutes a violation of Technical Specification 4.3.2.1 surveillance requirements. This condition has existed since initial surveillance procedure development and was identified on February 26, 1996 with the plant operating in Mode-1 at 100% power.

Trip Actuating Device Operational Testing has not been adequately performed for the Main Feedwater Pump trip signal following a safety injection actuation. OST-1825 actuates the safety injection switch and then verifies that the Main Feedwater Pumps trip, but due to the process involved during this testing, which includes lifting several leads and installing jumpers in Auxiliary Relay Panel (ARP-10, EIIS Code SJ-PL), a section of internal wiring in ARP-10 has not been verified. This testing deficiency has existed since initial surveillance procedure development and constitutes a violation of Technical Specification 4.3.2.1 surveillance requirements. This condition was identified on March 4, 1996 with the plant operating in Mode-1 at 100% power.

U.S. NUCLEAR REGULATORY COMMISSION NRC FORM 366A l(-96)

LICENSEE EVENT REPORT (LEB)

TEXT CONTINUATION FACILITY NAME (I) DOCKET LER NUMBER (6) PAGE(3)

SEOUENTIAL REYISION YEAR NUMBER NUMBER Shearon Harris Nuclear Plant ~

Unit )I'1 50100 4 OF 10 96 - 002 - 07 TEXT Pl moro sposo r's rorrvvod, vso odÃrioool oopr'os of NRC Form 36@) (I))

Ten Maintenance Surveillance Test Procedures were identified that did not perform a channel out of service alarm test or control room annunciation verification on the effluent channels of four separate radiation monitors (EIIS Code IL) as required by Technical Specification 4.3.3.11, Table 4.3-9. This condition has existed since initial development of the applicable surveillance test" procedures and was identified on March 11, 1996 with the plant operating in Mode-I at 100% power.

10. Surveillance testing has not been performed to properly verify the operability of fourteen blocking relays in the Emergency Safeguards Sequencer (ESS) Panels. (EIIS Code EK-PL,RLY) These relays function to ensure that ESF components start from the sequencer's load block program by blocking normal process demand signals and by preventing non-essential safety loads from energizing during load blocks 1 through 8. This testing deficiency has existed since initial surveillance procedure development and constitutes a violation of Technical Specification surveillance requirement 4.8.1.1.2. This condition was identified on March 21, 1996 with the plant operating in Mode-I at 100% power, and at 1500 hours0.0174 days <br />0.417 hours <br />0.00248 weeks <br />5.7075e-4 months <br /> both Emergency Sequencers were declared inoperable. Due to this, Technical Specification 4.0.3 was entered, which allowed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to demonstrate operability of the sequencers.

Special surveillance test procedures (OST-901ST and OST-9019T) were developed to perform this testing. At 1500 hours0.0174 days <br />0.417 hours <br />0.00248 weeks <br />5.7075e-4 months <br /> on the next day, March 22, 1996, testing to demonstrate sequencer operability was not complete, so the plant entered Technical Specification 3.0.3, which required shutdown to Mode-3 by 2200 hours0.0255 days <br />0.611 hours <br />0.00364 weeks <br />8.371e-4 months <br />. At 1804, the plant commenced a load decrease to comply with this requirement. At 2139, the plant was taken off line and at 2152, Mode-3 was achieved. Testing was completed to verify proper operation of the A-train sequencer at 2336 on March 22, 1996 and for the B-train sequencer at 1430 on March 23, 1996.

Surveillance testing has not been performed to verify proper operation of the Containment Fan Cooler Post-Accident Dampers (CV-D1, CV-D3, CV-DS, CV-D7 / EIIS Code VA-DMP). These dampers receive an open signal from the Emergency Safeguards Sequencer and directly from the associated fan cooler starting circuitry. Previous surveillance testing did not verify operability of the signal circuitry originating from the sequencer, failed to consider the existence of a parallel path within the start signal circuitry from the four fan cooler units and did not properly verify that the dampers were actually open. This condition was identified on March 26, 1996, at which time the plant was shutdown in Mode-5 (Cold Shutdown). This constitutes a violation of Technical Specification surveillance requirement 4.6.2.3. To verify operability of these dampers, testing was developed and performed on March 27, 1996. This testing identified that two of the post-accident dampers did not fully open as required and returned closed. Following lubrication of one damper and a minor modification to increase the output of the other damper actuator, satisfactory results were obtained and the dampers and their associated fan cooler units were returned to service on March 28, 1996.

12. Surveillance testing has not been performed to verify proper operation of one relay contact that inhibits the Essential Services Chilled Water Chillers (WC2A-SA &B-SB, EIIS Code: KM -CHU) from starting until load block g8 on the Emergency Safeguards Sequencers. Test procedures have not documented verification of this process. This condition was identified o'n April 16, 1996 with the plant operating in Mode-1 at 100% power. This testing deficiency has existed since initial surveillance procedure development and constitutes a violation of Technical Specification surveillance requirement 4.8.1.1.2.

13. Surveillance testing has not been performed to verify proper operation of a relay contact that bypasses the anti-recycle feature for starting the Essential Services Chilled Water Chillers (WC2A-SA &B-SB, EIIS Code KM-CHU).

The anti-recycle feature prevents more than one chiller start within a 30 minute period for equipment protection purposes. This anti-recycle feature is bypassed upon receipt of an automatic start signal from the Emergency Safeguards Sequencer. Verification of this bypass function has not been included in past surveillance testing. This condition was identified on April 16, 1996 with the plant operating in Mode-1 at 100% power. This testing deficiency has existed since initial surveillance procedure development and constitutes a violation of Technical Specification surveillance requirement 4.8.1.1.2.

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION

((-SSI LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILIT'Y NAME (il OOCKET LER NUMBER (SI PAGE(3)

SEQUENTIAL REYISION YEAR NUMBER NUMBER Shearon Harris NUclear Plant ~ Unit ¹1 50400 5 OF 10 96 - 002 - 07 TEXT pf sppss spsssir sssrspsd. pss sdd6ml ropes pl A'RC Form JSQI (I TI

14. Surveillance testing has not been performed to verify proper operation of the Containment Spray automatic sump swapover logic. The Containment Spray Pump Refueling Water Storage Tank Suction Valves (1CT-26SA &ICT-71SB) receive an automatic shut signal when their respective Containment Sump Suction Valves reach the full open position as indicated via contacts on their full open limit switches. This limit switch function has been verified by using a switched jumper that simulates limit switch operation. Tlierefore, continuity through'the limit switch has not been verified. This condition was identified on April 17, 1996 with the plant operating in Mode-1 at 100% power.

This testing deficiency has existed since initial surveillance procedure development and constitutes a violation of Technical Specification surveillance requirement 4.3.2.1.

15. Surveillance testing has not been performed to verify proper operation of the Containment Spray Pump sump suction valves (ICT-105SA &ICT-102SB, EIIS Code BE- V) following actuation of relay K741. These valves receive an automatic open signal on Refueling Water Storage Tank (RWST) Lo-Lo level via slave relays K739 and K741. A parallel path exists from each of these relays and past surveillance testing has only verified proper operation of the suction valves from the K739 relay. This condition was identified on April 17, 1996 with the plant operating in Mode-1 at 100% power. This testing deficiency has existed since initial surveillance procedure development and constitutes a violation of Technical Specification surveillance requirement 4.3.2.1.

16. Surveillance testing has not been performed to verify proper operation of the Computer Room Dampers adjacent to the Main Control Room, following a Control Room Isolation Signal. These dampers (CK-D7-1&2, CK-D4-1&2, CK-D8-1&2 and CK-B11-1&2, EIIS Code VI-DMP) receive a signal from relay K603 to place the Computer Room in Recirculation, but have not been included in previous surveillance testing. This condition was identified on April 17, 1996 with the plant operating in Mode-I at 100% power. This testing deficiency has existed since initial surveillance procedure development and constitutes a violation of Technical Specification surveillance requirement 4.3.2.1.

17. Surveillance testing has not been performed to verify proper operation of the Control Room Isolation Signal for various Computer and Communication Room HVAC components in addition to the dampers listed in item 416 above.

These components receive actuation signals following a Control Room Isolation Signalbut have not been included in previous surveillance testing. This condition was identified on April 17, 1996 with the plant operating in Mode-1 at 100% power. This testing deficiency has existed since initial surveillance procedure development and constitutes,a violation of Technical Specification surveillance requirement 4.3.2.1.

18. Surveillance testing has not been performed to verify proper operation of the Emergency Safeguards Sequencer load block 4 starting circuit path for the IA-SA and IB-SB Containment Spray Pumps (EIIS Code: BE-P). In addition, a parallel Containment Spray Pump starting circuit path from ESS load block 2 has not been independently verified. A combination of testing has verified the pumps automatic start circuitry properly functions, thus no immediate operability concern exists. However, previous testing has not clearly documented which of the two circuit paths provided the automatic pump start signal. This condition was identified on April 23, 1996 with the plant operating in Mode-I at 100% power. This testing deficiency has existed since initial surveillance procedure development and constitutes a violation of Technical Specification surveillance requirement 4.8.1.1.2'.

19. Surveillance testing has not been performed to verify proper operation of Breakers 1A3A-SA & 1B3A-SB (EIIS Code: EB-BKR). These breakers are required to trip open during the ESS load shedding process. Previous testing has not included verification and documentation that the breakers have tripped open. Based on a review of historic computer printout logs from the most recent refueling outage (RFO ¹6), the breakers have operated as required, thus no immediate operability concern exists. This condition was identified on April 24, 1996 with the plant operating in Mode-1 at 100% power. This testing deficiency has existed since initial surveillance procedure development and constitutes a violation of Tcchnical Specification surveillance requirement 4.8.1.1.2.

NRC FORM 3BBA US. NUCLEAR REGULATORY COMMISSION

)4-95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (I) OOCKET LER NUMBER (6) PAGE Ii)

SEOUENTIAL RDtISION YEAR NUMBER NUMBER Shearon Harris Nuclear Plant Unit A'1 50 400 6 OF 10 96 - 002 - 07 TEXT ff! mhvo spooo fs fohfviod, vso odChleool oopfos of frfrC fofm 3hriShu Ill)

20. Surveillance testing has not been performed to verify proper operation the Gross Failed Fuel Detector Isolation Valve (1CC-304) following a Safety Injection signal. This valve is required to shut following a Safety Injection signal to isolate the Gross Failed Fuel Detector, but verification and documentation of this actuation has not been included in previous testing. This condition was identified on April 24, 1996 with the plant operating in Mode-1 at 100% power.

This testing deficiency has existed since initial surveillance procedure development and constitutes a violation of Technical Specification surveillance requirement 4.3.2.1.

21. Surveillance testing has not been performed to verify proper operation of the Computer Room and Communication Room Dampers adjacent to the Main Control Room, following a Control Room Isolation Signal. These dampers (CK-Dl l-1&2 and CK-D12-1&2) fail open following a Control Room Isolation Signal and during previous surveillance testing have only been verified by observation of "not-shut" indication. This condition was identified on February 26, 1996 with the plant operating in Mode-1 at 100% power. This testing deficiency has existed since initial surveillance procedure development and constitutes a violation of Technical Specification surveillance requirement 4.3.2.1. Due to an administrative oversight during the on-going Technical Specification testing program review, this item was not included in Revision P3 to this LER as it should have been to meet the 30-day reporting requirement.

The reportability determination process has been enhanced to prevent recurrence.

(Note: Failure to initially-identify this reportable condition and include it in Revision P3 caused a break in the numerical sequence of identified deficiencies. Items 18 - 20 will be included in a future revision to this LER.)

22. Technical Specification surveillance 4.8.1.1.2.f.8 requires a verification that electrical loads automatically connected to the safety bus during Emergency Safeguards Sequencer loading, do not exceed the Emergency Diesel Generators (EDG) continuous rating of 6500 KW. During the Technical Specification Testing Program review, several components were identified that should have been included in this calculation, but had not been during previous testing. When combined, these loads amount to an additional 50 KW of load. Based on the calculated post-accident EDG load of approximately 4000 KW, the 6500 KW limit was not exceeded, thus no operability concern exists.

This surveillance requirement was relocated from Technical Specifications to the EDG Reliability Program by issuance of amendment 60 to the Harris Operating License. Though no longer a Technical Specification requirement, the failure to include these loads durirtg past testing represents a historical Technical Specification violation. This testing deficiency has existed since initial surveillance procedure development and was identified on April 25, 1996 with the plant operating in Mode-1 at 100% power.

23. Surveillance testing has not been performed to verify proper operation of several dampers in the Engineered Safety Features (ESF) Ventilation System (EIIS Code: VF-DMP). These dampers receive indirect actuation signals from their associated fans (AH-12 1A-SA, AH-13 1A-SA, AH-16 1A-SA & AH-16 IB-SB) during Emergency Safeguards Sequencer loading. Damper actuation has not been verified or documented during previous testing. This condition was identified on April 25, 1996 with the plant operating in Mode-1 at 100% power. This testing deficiency has existed since initial surveillance procedure development and constitutes a violation of Technical Specification surveillance requirement 4.8.1.1.2.f.

24. Surveillance testing has not been performed to verify proper operation of the inhibit interlock circuits (UR-3 & UR-4) associated with the Primary Shield Cooling Fans (S-2 lA-SA and S-2 1B-SB, EIIS Code: VA-Fan) and Reactor Support Cooling Fans (S-4 1A-SA and S-4 1B-SB, EIIS Code: VA-Fan). These inhibit circuits block fan operation signals during execution of the Emergency Safeguards Sequencer automatic loading process. This condition was identified on May 8, 1996 with the plant operating in Mode-1 at 100% power. This testing deficiency has existed since initial surveillance procedure development and constitutes a violation of Tcchnical Specification surveillance requirement 4.8.1.1.2.f.

NRC fORM 366A U.S. NUClEAR REGULATORY COMMISSION

)6 96)

LICENSEE EVENT REPORT tLER)

TEXT CONTINUATION FACILITY NAME II) OOCKET lER NUMBER IB) PAGE I3)

SEOUEN'TIAL REYISION YEAR NUMBER NUMBER Shaaron Harris Nuclaar Plant ~

Unit )I1 50400 7 OF 10 96 - 002 - 07 TEXT p/moro sprso N roqod, oso oddinorrslsopks ol hi)C Form SBQ) IIT)

CAUSE:

The cause of the original Technical Specification violation was personnel error during the June 1993 procedure revision process for OST-1083 and OST-1084. The testing requirements for the slave relay circuits for the CSIP Alternate Mini-Flow Isolation Valves (ICS-746 and ICS-752) were inappropriately transferred to OST-1809, which was not identified or scheduled as a Mode-5 "event related" surveillance test.

Cause For Additional Items Identified:

Item 1:

The failure to adequately test the K635 and K640 slave relays for the AFW Flow Control Valves was caused by inadequate technical reviews associated with plant modification PCR-6502. This resulted in deficient surveillance test procedures developed to verify the operability of the automatic open signal for the flow control valves on a quarterly basis.

Items 2 3 4 5:

Each of these items were caused by inadequate surveillance test procedures that resulted from incorrectly interpreting how to implement Technical Specification testing requirements. The test procedures for the Safety Injection and Containment Spray manual actuation switches, as well as the FHB Emergency Ventilation system, were based upon this incorrect interpretation and have been deficient since initial development. The radiation monitor MST revisions completed in 1993, were intentionally performed to eliminate what was considered to be redundant and unnecessary testing steps. This decision was also based on the incorrect testing requirement interpretation, as was the revision to OST-1086 that resulted in the acceptance criteria section not listing 1CS-196, and subsequently resulting in the failure to test the valve.

Items 6 throu h 24:

Each of the additional items contained in the revisions to this LER were identified as a result of the on-going Technical Specification testing program review and were caused by inadequate surveillance test procedures. In the case of item )I)11, the two post-accident dampers failed to fully open during testing due to improper actuator sizing and inadequate lubrication and preventive maintenance methods.

SAFETY SIGNIFICANCE:

There were no adverse safety consequences as a result of this event. The CSIP Alternate Mini-Flow Isolation Valve circuits were tested satisfactorily on September 8, 1995 to verify operability. This testing provides confidence that had an accident occurred requiring CSIP mini-flow protection due to the re-pressurization of the Reactor Coolant System during a safety injection, the isolation valves would have opened to prevent pump damage.

There were no adverse safety consequences as a result of the additional items contained in this LER revision. In each case where applicable, subsequent testing was performed that verified the operability of the effected component or circuit. In the case of item Pl 1, where two of the Containment Fan Cooler Post-Accident Dampers failed to completely open during testing, consequences were minimal. These dampers are required to be open in a post-accident condition within containment to allow a high velocity fan discharge flow to selected areas of containment to accelerate temperature mixing and heat removal. Assuming the failure of these two dampers to open during an accident scenario, combined with the postulated worst case single failure of one safety related electrical supply bus, engineering review has determined that adequate air flow would still exist to ensure containment cooling. This conclusion is based on the availability of one train of Containment Spray and the fact that one fan would remain operable in each Containment Fan Cooler unit. The discharge air flow from each fan would not exit through the post-accident dampers, but would still prov'ide air mixing in containment via the seismic class 1 concrete air shafts.

PREVIOUS SIMII.AR EVENTS:

Previous events have been submitted as LERs related to surveillance testing deficiencies caused by procedural inadequacies.

LER 95-07, which was submitted on September 28, 1995, contained a corrective action to perform a comprehensive Technical Specification testing program review and it was during this review process that the CSIP Alternate Mini-Flow Isolation Valve condition was identified. This review is being performed by a multi-discipline team and is still in progress.

The additional item reported in this supplement was identified as a result of the on-going Technical Specification testing program review.

R RM A

e NRC FORM 366A US. NUCLEAR REGULATORY COMMISSION L4.SSI LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (I) OOCKET LER NUMBER Ni PAGE ISI SEOUENTIAI. RotISION NUMBER NUMBER Shaaron Harris Nuclaar Plant ~

Unit A'1 50400 8 OF 10 96 - 002 - 07 TEXT pl moro spssv r's rvrrvisd, vsv vrrdrsdrrrol sop<is of pNC Form JSS<U Ilil CORRECTIVE ACTIONS COMPLETED:

1. Slave relay testing was satisfactorily completed on September 8, 1995 that verified the operability of the CSIP Alternate Mini-Flow Isolation Valve circuits.

2. Personnel involved in the June 1993 procedure revision process for OST-1083,and OST-1084 were counseled.

3. Surveillance procedures OST-1083 and OST-1084 were revised on February 16, 1996 to include the requirements for CSIP Alternate Mini-Flow Isolation Valve slave relay testing.

CORRECTIVE ACTIONS PLANNED:

1. The comprehensive Technical Specification testing program review that identified this condition is currently in progress and will be completed as described in LER 95-07.

CORRECTIVE ACTIONS FOR ADDITIONALITEMS IDFNTIFIED:

Upon completion of the Technical Specification Testing Program review, appropriate 18-month (refueling) interval surveillance procedures will be revised to incorporate each of the identified deficiencies. This may include development of new procedures and/or separation of testing requirements into several existing procedures. These actions will be performed as addressed in the pending Harris Plant response to NRC Generic Letter 96-01.

~I<em OST-1044 was revised in December 1995 and OST-1045 was revised in February 1996. These revisions incorporated K635 and K640 slave relay testing on a quarterly basis.

Item 2:

To address the SI switch that would have become inoperable on March 16, 1996, a Request for Exigent License Amendment was submitted to the NRC on February 16, 1996. This requested a one-time extension of the testing interval for testing the SI switch, due to the hazards involved with testing while on-linc. Operations Surveillance Test procedure (OST-9016T) was revised and performed to test the CS switches. Additionally, a new OST will be developed to properly test each Safety Injection and Containment Spray manual actuation switch once every 18 months.

Item 3:

OST-1086 will be revised to enhance the acceptance criteria to ensure that testing of the "B" CSIP Normal Mini-Flow Isolation (1CS-196) is included.

Item 4:

The radiation monitor operability and testing requirements were moved from Technical Specification 4.3.3.10 to the Off-Site Dose Calculation Manual (ODCM) in May 1995. Upon identification of this condition, the effected radiation monitors were declared inoperable. Appropriate MST procedure revisions were completed and the tests performed, to fully verify the automatic pathway isolation function of the radiation monitors. To ensure compliance, additional procedure changes and/or ODCM revisions will be completed to clarify the testing requirements and enhance the performance of future testing.

Item 5:

OST-1048 will be revised to test the FHB Train A Emergency Ventilation actuation from Radiation Monitor RM-1FR-3567A-SA, which will properly include the previously omitted cable.

A

NRC FORM 366A U.S. NUCLEAR RLGULATORT COMMISSION l4 BSI LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACIUTT NAME (I) OOCKET LER NUMBER IBI PAGE ITI SLOUENTIAL RETISIOK TEAR NUMBER NUMBER Shaaron Harris Nuclaar Plant ~

Unit //1 50 400 9 OF 10 96 - 002 - 07 TEXT pf mort evsofs soqvvod. vsv vdd Iaosl sopis of fffIC&m 36iW Il TI Item 6:

MST-l0359 and MST-10361 were revised on February 21, 1996 and successfully performed. This verified the operability of the parallel R-2 fan start circuit on high radiation.

Item 7:

OST-9017T was developed and successfully performed on February 27, 1996. This verified the operability of the thermal overload bypass circuit used during the Control Room Isolation Signal when generated from high radiation. To ensure compliance, future testing of this circuit will be incorporated with a revision to the appropriate maintenance surveillance test procedures or a newly developed operations surveillance test procedure.

Item 8:

OST-1825 and OST-1826 will be revised to properly test appropriate internal wiring in ARP-IO.

Item 9:

Immediate corrective actions included declaring the affected radiation monitors inoperable and placing the deficient MST procedures on administrative hold until they could be revised. These procedures were subsequently revised and performed as needed to properly demonstrate the operability of the radiation monitors. This was completed on March 14, 1996.

Item 10:

Immediate corrective actions for this item included declaring both Emergency Safeguards Sequencers inoperable and complying with the testing and plant shutdown requirements of Tcchnical Specification 4.0.3 and 3.0.3. Testing was completed to verify operability of the A-train sequencer at 2336 on March 22, 1996 (OST-9018T) and for the B-train sequencer at 1430 on March 23, 1996 (OST-9019T). To ensure compliance, the appropriate surveillance test procedures will be revised to include future testing of the blocking relays.

Item 11:

Following corrective maintenance, which included a modification to increase the actuator spring rate for damper ICV-DI, testing was satisfactorily performed on March 28, 1996 and the post accident dampers and their associated fan cooler units were returned to service. Preventive maintenance for these dampers was enhanced by generating a checklist (CL-ME0023) that includes requirements for periodic lubrication and inspection. This was completed on 4/19/96. To ensure compliance, the appropriate surveillance test procedures will be revised to include future testing of the post-accident dampers.

Items 12 & 13:

Immediate corrective actions for these items included declaring both Essential Services Chilled Water Chiller units inoperable. Testing (OST-9020T and EPT033) was then satisfactorily completed on April 16, 1996 to verify operability of both chiller units. To ensure compliance, the appropriate surveillance test procedures will be revised to include future testing of the chiller unit inhibit and anti-recycle bypass functions.

Item 14:

No immediate operability concern existed as a result of this condition due to the performance of a special test (OST-1809T) performed on June 6, 1995. To ensure compliance, the appropriate surveillance test procedure(s) will be revised to include future testing of the limit switch function/continuity.

M A ~

I

NRC FORM 3BBA US. NUCLEAR REGULATORY COMMISSION t4ast LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME lu OOCXET LER NUMBER (B1 PAGE 13)

SEQUENTIAl REVISION YEAR NUMBER NUMBER Shearon Harris Nuclear Plant ~

Unit //'1 50.400 10 OF 10 96 - 002 - 07 TEXT p/ mort sposo N rorvi od, vso oddi drool sopr'os ol NRC Form 3BatU lr 7)

Item 15:

No immediate operability concern existed as a result of this condition due to the performance of quarterly surveillance testing (OST-1083 &1084) performed on March 26, 1996, which verified actuation of valves 1CT-102 and 1CT-105. To ensure compliance, the appropriate surveillance test procedure(s) will be revised to include future testing of the actuation signal from both slave relays.

Item 16 & 17:

Testing was performed on April 17, 1996 to verify the proper operation of the Computer and Communication Room HVAC components following a Control Room Isolation Signal (MST-10362). To ensure compliance, the appropriate surveillance test procedure(s) will be revis'ed to include future testing of these dampers.

Item 18:

No immediate operability concern existed as a result of this condition. To ensure compliance, the appropriate surveillance test procedure(s) will be revised to include future testing of the Containment Spray Pump automatic start circuitry.

Item 19:

Based on a review of OST-1823 and OST-1824 ERFIS computer historic log printouts, the 1A3A-SA and IB3A-SB breakers operated as required during testing performed in Refueling Outage 6. To ensure compliance, the appropriate surveillance test procedure(s) will be revised to include future testing of these breakers.

Item 20; Operability of valve 1CC-304 was verified on March 26, 1996 during the performance of OST-1083. To ensure compliance, the appropriate surveillance test procedure(s) will be revised to include future testing of ICC-304 following a Safety Injection signal.

Item 21:

Testing was performed on February 26, 1996 to verify proper operation of dampers CK-Dl 1-1&2 and CK-D12-1&2 (OST-9017T). To ensure compliance, the appropriate surveillance test procedure(s) will be revised to include future testing of these dampers.

Item 22:

An Engineering Evaluation (ESR 00226) was initiated on April 25, 1996 and concluded that the additional 50KW loading would not exceed the EDG's continuous rating of 6500 KW. To ensure compliance, the appropriate test procedure(s) and/or engineering calculations will be revised.

Item 23:

Testing was satisfactorily performed on April 25, 1996 to verify the operability of each damper that received an indirect signal from it's associated ESF Ventilation System fan. To ensure compliance, the appropriate surveillance test procedure(s) will be revised to include future testing of these dampers.

Item 24:

Testing was satisfactorily performed on May 8, 1996 to verify the operability of the UR-3 and UR-4 inhibit circuits. To ensure compliance, the appropriate surveillance test procedure(s) will be revised to include future testing of these circuits.

EIIS CODES:

High Head Safety Injection: BQ Auxiliary Feedwater Flow Control Valves: BA-FCV Containment Spray: BE Containment Ventilation: VA Fuel Handling Building Ventilation: VG Control Room Emergency Ventilation: VI Reactor Auxiliary Building Ventilation: VF Main Feedwater: SJ Radiation Monitoring: IL Emergency Sequencers: EK