ML18012A795

LER 96-023-01:on 961114,design Deficiency in EDG Protection Circuitry Was Identified.Caused by Inadequate Original Plant Design.Surveillance Test Procedures OST-1013 & OST-1073 revised.W/970529 Ltr

ML18012A795
Person / Time
Site:	Harris
Issue date:	05/29/1997
From:	Donahue J, Johnny Eads CAROLINA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
HNP-97-110, LER-96-023, LER-96-23, NUDOCS 9706050101
Download: ML18012A795 (10)

Text

C A.'l.'EGORY J.REGULATOINPORMATION DISTRIBUTION TEM (RIDE)ACCESSION NBR:9706050101 DOC.DATE: 97/05/29 NOTARIZED:

NO FACIL;.50-400 Shearon.Harris Nuclear Power Plant, Unit 1, Carolina AUTH.NAME AUTHOR AFFILIATION EADSiJ.Carolina Power&Light Co.DONAHUE,J.W.

Carolina Power&Light Co.RECIP.NAME RECIPIENT AFFILIATION DOCKET¹05000400

SUBJECT:

LER 96-023-01:on 961114,design deficiency in EDG protection circuitry was identified.

Caused by inadequate original plant design.Surveillance test procedures OST-1013&OST-1073 revised.W/970529 ltr.DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR ENCL'IZE: TITIE: 50.73/50.9 Licensee Event Report (LER), Incident Rpt, etc.A r T NOTES:Application for permit renewal filed.05000400 RECIPIENT ID CODE/NAME PD2-1 PD INTERNAL: ACRS AEOD/SPD/RRAB NRR/DE/ECGB'RR/DE/EMEB NRR/DRCH/HICB NRR/DRCH/HQMB NRR/DSSA/SPLB RES/DET/EIB EXTERNAL: L ST LOBBY WARD NOAC POORE,W.NRC PDR COPIES LTTR ENCL 1 1 1 1 1 1 1 1 1 1--1 1 1 1 1 1 1-1 1'1 1 1 1 RECIPIENT ID CODE/NAME ROONEYiV B FILE CENT R B NRR/DRCH/HHFB NRR/DRCH/HOLB NRR/DRPM/PECB NRR/DSSA/SRXB RGN2 FILE'1 LITCO BRYCE,J H NOAC QUEENERIDS NUDOCS FULL TXT COPIES LTTR ENCL 1 1 2 2 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 R D 0 N NOTE TO ALL"RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE.TO HAVE YOUR NAME OR ORGANIZATION REMOVED FROM DISTRIBUTION L1STS OR REDUCE THE NUMBER OF COPIES RECEIVED BY YOU OR YOUR ORGANIZATION, CONTACT THE DOCUMENT CONTROL DESK (DCD)ON EXTENSION 415-2083 FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 25 ENCL 25 f

Carolina Power&Light Company Harris Nuclear Plant PO Box 165 New Hill NC 27562 MAY 29 1997 U.S.Nuclear Regulatory Commission ATTN: NRC Document Control Desk Washington, DC 20555 Serial: HNP-97-110 10CFR50.73 SHEARON HARRIS NUCLEAR POWER PLANT UNIT 1 DOCKET NO.50-400 LICENSE NO.NPF-63 LICENSEE EVENT REPORT 96-023-01 Sir or Madam: In accordance with Title 10 to the Code of Federal Regulations, the enclosed revision to Licensee Event Report 96-023 is submitted.

This revision provides additional information related to the design deficiency identified in the Emergency Diesel Generator protection circuitry.

Sincerely, J.W.Donahue Director of Site Operations Harris Plant JHE Enclosure c: Mr.J.B.Brady (HNP Senior NRC Resident)Mr.L.A.Reyes (NRC Regional Administrator, Region II)Mr.N.B.Le (NRC-NRR Project Manager)970b050iOf 970529 PDR ADQCK 05000400 8 PDR II!Illl/IIIIIIIIIIIIIIIIIIIIIIII////III/

State Road 1134 New Hill NC l I j+

U.S.'Nuclear Regulatory Commission Document Control Desk/HNP-97-110 Page 2 of 2 CC: Mr.T.C.Bell Mr.H.K.Chernoff (RNP)Mr.B.H.Clark Mr.G.W.Davis Ms.S.F.Flynn Ms.J.P.Gawron (BNP)Mr.H.W.Habermeyer Mr.M.D.Hill Mr.W.J.Hindman Ms.C.W.Hobbs (HEEC)Mr.R.M.Krich Mr.M.B.Keef (HEEC)Ms.W.C.Langston Mr.C.W.Martin (BNP)Mr.R.D.Martin Mr.J.W.McKay Mr.P.M.Odom (RNP)Mr.W.R.Robinson Mr.G.A.Rolfson Mr.R.F.Saunders Mr.C.N.Sweely Mr.M.A.Turkal (BNP)Mr.T.D.Walt Mr.R.L.Warden (RNP)HNP Real Time Training INPO Harris Licensing File Nuclear Records

~X l NRC FORM 366 IL95)U.S.LEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)(See reverse for required number of digits/characters for each block)ROVED BY OMB NO.3150-0104 EXPIRES 04/30/98 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THS MANDATORY INFORMATION COLLECTION REQUEST: 500 HRS.REPORTED LESSONS LEARNED ARE INCORPORATED UITO THE UCENSWG PROCESS AND FEO BACK TO INDUSTRY.FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION ANO RECORDS MANAGEMENT BRANCH IT4(F33L US.NUCLEAR REGULATORY COMM)SSION.

WASHiNGTON, OC 205550001.

ANO TO THE PAPERWORK REDUCTION PROJECT (3150.0104L OFFICE OF MANAGEMENT ANO BUDGET, WASHINGTON, OC 205(KL FACILITY NAME (1I Harris Nuclear Plant Unit-1 DOCKET NUMBER (21 50-400 PAGE (3)1 OF3 TrrLE (41 Design deficiency in Emergency Diesel Generator protection circuitry.

MONTH DAY YEAR EVENT DATE{5)LER NUMBER (6)YEAR SEOUENTIAL NUMBER REVISION NUMBER MONTH OAY YEAR REPORT DATE (7)FACIUTY NAME OTHER FACILITIES INVOLVED (8)DOCKET NUMBER 05000 14 96 96-023-01 29 97 FACIUTY NAME DOCKET NUMBER 05000 OPERATING MODE (9)POWER LEVEL (10)100%20.2201(b) 20.2203(a)

(1)20.2203(a)(2)(i)20.2203(a)

(2)(ii)20.2203{a)

(2)(iii)20.2203(a)(2)(iv) 20.2203{a)(2)(v) 20.2203(a)(3)(i)20.2203(a)

(3)(ii)20.2203(a)

(4)50.36(c)(1) 50.36(c)(2) 50.73(a){2)(i)X 50.73(a)(2)(ii) 50.73(a)(2)(iii) 50.73(a)(2)(iv)50.73(a){2)(v)50:73(a)(2)

{vii)50.73(a)(2)

{viii)50.73(a)(2)(x) 73.71 OTHER Specify In Abstract below or in NRC Form 366A THIS REPORT IS SUBMITTED PUR SUANT TO THE REQUIREMENTS OF 10 CFR 5: (Check one or more)(11)LICENSEE CONTACT FOR THIS LER Johnny Eads Project Engineer-Licensing (12)TELEPHONE NUMBER (Inerude Ares Code)(919)362-2646 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DES CRIBED IN THIS REPORT (13)CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE TO NPROS CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE TO NPROS SUPPLEMENTAL REPORT EXPECTED (14)YES{If yes, complete EXPECTED SUBMISSION DATE).X NO EXPECTED SUBMISSION DATE (15)MONTH OAY YEAR ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single.spaced typewritten lines){16)On November 14, 1996, with the plant operating in Mode-1 at 100%power, a design deficiency was identified in the protection circuitry for the safety-related Emergency Diesel Generators (EDG).Section 8.3.1.1.2.14.g of the Harris Plant Final Safety Analysis Report (FSAR)states that"Protection is provided for the diesel generator and the safety related electrical system during periodic testing of the diesel generator coincident with a loss of off-site power by the voltage restrained overwurrent relay (51V)at the diesel generator feeder.This relay senses over-current due to overloading of the diesel generator in conjunction with reduction of voltage.The relay is arranged to trip the feeder breaker to the diesel generator." During an engineering review resulting from NRC Generic Letter 9641, the ability of the 51V relay to provide the described protection during a loss of off-site power (LOOP)event was questioned.

Subsequent investigation concluded on December 4, 1996, that the relay would not provide this protection.

As a result of this condition, if a LOOP had occurred while the EDG was synchronized to the off-site electrical grid during periodic testing, the undervoltage relays for the safety-related 6.9 kV bus may not have actuated and the associated emergency sequencer would not have recognized the LOOP condition and sequenced the safety bus loads as required.This condition was reported to the NRC on December 4, 1996 per 10 CFR 50.72 via the emergency notification system as operation outside the design basis of the plant.Immediate corrective actions included revising the applicable EDG test procedures to verify that stable grid voltage exists prior to paralleling and declaring the EDG inoperable during this testing.Additional corrective action included the design and installation of a modification to the EDG protection circuitry to return the system to its original functional design basis.The initial 10 CFR 50.73 LER was submitted on December 16, 1996.This revision provides additional information related to the deficient 51V relay configuration.

NRC FORM 36BA I4BSI LICENSEE EVENT REPORT (LER)TEXT CONTINUATION US, NUCLEAR REGULATORY COMMISSION FACILITY NAME III Shearon Harris Nuclear Plant-Unit 01 DOCKET 50400 LER NUMBER IB)'YEAR SEOUENTIAL REVISION NUMBER NUMBER 96-023-01 PAGE ISI 2 OF 3 TEXT rrl mare spese m reerdrsC ese edSrkwl eoples ol ffRC Furm 3SSlU (I T]EVENT DESCRIPTION:

On November 14, 1996, with the plant operating in Mode-1 at 100%power, a design deficiency was identified in the protection circuitry for the safety-related Emergency Diesel Generators (EDG).Section 8.3.1.1.2.14.g of the Harris Plant Final Safety Analysis Report (FSAR)states that"Protection is provided for the diesel generator and the safety related electrical system during periodic testing of the diesel generator coincident with a loss of off-site power by the voltage restrained over-current relay (51V)at the diesel generator feeder.This relay senses over-current due to overloading of the diesel generator in conjunction with reduction of voltage.The relay is arranged to trip the feeder breaker to the diesel generator." During an engineering review resulting from NRC Generic Letter 96-01, the ability of the 51V relay to provide the described protection during a loss of off-site power (LOOP)event was questioned.

Subsequent investigation concluded on December 4, 1996, that the relay would not provide this protection.

To simplify the discussion, the concern will be described for Division A of the onsite power system but is also applicable to Division B.To perform required Technical Specification surveillance testing of the EDGs, it is necessary to connect an EDG in parallel with the off-site power system.This is accomplished by connecting a running-EDG to its associated safety bus (closing EDG output breaker 106)while the safety bus remains connected to its associated non-safety bus through the two tie breakers (104 and 105).The non-safety bus is connected to the off-site power system via either the Start-up Transformer or the Unit Auxiliary Transformer.

This electrical distribution system configuration allows the EDG to assume the additional load required for testing.In the original design, if an EDG was in the test mode and a LOOP event occurred, the'logic would have held the tie breakers between the non-safety bus (1D)and safety bus (1A-SA)closed with the objective of producing an oveiload condition on the safety bus while dragging the voltage down to allow operation of the bus undervoltage relay or the voltage controlled overcurrent relay (51V).The 51V is operational in the test mode only.The original EDG logic used the safety-related 6.9 kV bus undervoltage relays to detect a LOOP.The premise of the logic was that if the off-site power source was lost then the load on the diesel from connected loads would exceed of, the EDG capacity and cause an undervoltage condition.

To ensure that the connected load was large enough, the original logic inhibited, if the EDG was in the test configuration, the tripping of the 105 breaker by a LOOP detection relay (CR1/1748).

However, if the available loading on the safety bus and non-safety bus did not exceed the capacity of the EDG, the UV on the safety bus (or the 51V)would not actuate and the sequencer would not receive a signal to perform its LOOP program.In summary, the evaluation of possible loading conditions revealed that the EDG would not respond as described in the FSAR when distribution system load was insufficient to actuate the 51V or UV relays.This condition was reported to the NRC on December 4, 1996 per 10 CFR 50.72 via the emergency notification system as operation outside the design basis of the plant.This condition was then reported in LER 96-023, dated December 16, 1996.Corrective actions identified at that time included declaring the EDG inoperable during periodic testing and entering the appropriate Technical Specification Action Statements.

In addition, the EDG protection circuitry was modified to provide protection from a LOOP during EDG load testing.The EDG circuitry modification returns the onsite power system to its original functional design basis and minimizes the need for operator action if a LOOP occurs during EDG testing.This change also eliminates the need to declare the EDG inoperable during periodic testing.However, the modification required deviations from existing licensing commitments.

Specifically, the deviations included: 2)The use of non-class 1E equipment to provide inputs to a Class 1E device in support of a safety function.The modification design takes credit for the response of the LOOP relay and its supporting power supply and inputs;and The use of operator action as a contingency response in case the non-class 1E equipment fails to provide the automatic action.

NRC FORM'36BA i4-BS)LICENSEE EVENT REPORT (LER)TEXT CONTINUATION US.NUClEAR REGUEATORY COMMISSION FACIlITY NAME II)Shee)on Harris Nuclear Plant-Unit N1 DOCKET 50400 lER NUMBER (6)YEAR SEOUENTIAI.

REYIION NUMBER NUMBER 66-023-01 PAGE l3)'OF 3 TEXt fff mort spooo rs rorfoiorl oso oW(oroof oopfos of ffi)C Form 36&9 I)1)EVENT DESCRIPTION: (cont.)Because of these deviations, the EDG protection circuitry modification was determined to be an Unreviewed Safety Question.The proposed change was submitted to the NRC in-accordance with 10 CFR 50.59(c)and 10 CFR 50.90 on April 18, 1997.On May 8, 1997, the NRC issued HNP License Amendment No.72 approving the changes to the EDG protection circuitry.

CAUSE: This condition was caused by inadequate original plant design.SAFETY SIGNIFICANCE:

There were no actual safety consequences associated with this event.However, the HNP Probabilistic Safety Assessment (PSA)was used to estimate the increase in core damage risk if the EDGs are assumed to be unavailable when operated in parallel with off-site power.The model includes events representing the EDG being unavailable.

The probability of the EDGs being unavailable was increased to account for an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per year per EDG.The increase in annual core damage frequency was found to be less than one percent, which is not considered to be risk significant in accordance'with the EPRI PSA Applications Guide.The HNP PSA model includes initiating events for a number of transients, such as reactor trip, turbine trip, loss of off-site power, and inadvertent safety injection.

The model also includes various loss of coolant accidents, steam generator tube rupture, and ATWS scenarios.

The calculated increase in annual core damage frequency of less than one percent represents the increased risk from the greater number of hours of the EDGs being unavailable.

Operator action to return the EDG being tested to service is not credited.The change in core damage frequency includes the relative importance of the EDGs in mitigating the effects of the various scenarios included in the PSA.\This condition is being reported in accordance with 10 CFR 50.73.a.2.ii(B) as a condition that was outside the design basis of the plant.PREVIOUS SIMILAR EVENTS: There have been no previous deficiencies reported related to the design of EDG protection circuitry.

CORRECTIVE ACTIONS COMPLETED:

1.Surveillance test procedure OST-1013 (lA-SA EDG Operability Test-Monthly Interval)was revised on December 4, 1996.2.Surveillance test procedure OST-1073 (1B-SB EDG Operability Test-Monthly Interval)was revised on February 17, 1997.3.A modification to the EDG protection circuitry to return the system to its original functional design basis was installed and turnover completed on May 26, 1997 during Refueling Outage 7.Implementation of this modification included appropriate revisions to testing procedures and the Final Safety Analysis Report.CORRECTIVE ACTIONS PLANNED: None.

14