ML18012A824

LER 97-015-00:on 970602,inadequate Auxiliary Feedwater Sys Flow Control Valve Surveillance Testing Deficiency Was Identified.Caused by Failure to Recognize Impact on TS 4.7.1.2.1.Readjusted AFW FCV Actuator spring.W/970702 Ltr

ML18012A824
Person / Time
Site:	Harris
Issue date:	07/02/1997
From:	Donahue J, Verrilli M CAROLINA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
HNP-97-137, LER-97-015, LER-97-15, NUDOCS 9707090073
Download: ML18012A824 (6)

Text

CATEGORY REGULATORY INFORMATION DISTRIBUTION SYSTEM ly (RIDS)

ACCESSION NBRo9707090073 DOC DATEs 97/07/02

~ NOTARIZEDe NO DOCKET FACIL:56- 30" Shearon Harris Nuclear Power Plant, Unit 1, Carolina 05000400 AUTH. NAME AUTHOR AFFILIATION VERRILLI,M. Carolina Power & Light Co.

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

RECIP.NAME RECIPIENT AFFILIATION

SUBJECT:

LER 97-015-00:on 970602,inadequate auxiliary feedwater sys flow control valve surveillance testing deficiency C identified. Caused by failure to recognize impact on TS 4.7.1.2.1.Readjusted AFW FCV actuator spring.

DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR ) ENCL ( SIZE:

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

E NOTES:Application for permit renewal filed, 05000400 RECIPIENT COPIES RECIPIENT COPIES 0 ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD2-1 PD 1 1 ROONEY,V 1 1 R INTERNAL: ACRS 1, 1 AEODPZ'D/RAB 2 2 AEOD/SPD/RRAB 1 1 FILE CENTER 1 1 NRR/DE/ECGB 1 1 NRR 15K/E 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/DET/EIB 1 1 RGN2 FILE 01 1 1 EXTERNAL: L ST LOBBY WARD 1 1 LITCO BRYCE,J H 1 1 0 NOAC POORE,W. 1 1 NOAC QUEENERiDS 1 1 NRC PDR 1 1 NUDOCS FULL TXT 1 1 NOTE TO ALL "RIDS" RECZPZENTS:

PLEASE HELP US TO REDUCE NASTEl CONTACT THE DOCUMENT CONTROL DES@,

ROOM OWPN 5D-5(EXT. 415-2083) TO ELIMINATE YOUR NAME PROM DISTRIBUTION LISTS POR DOCUMENTS YOU DON'T NEEDl FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 25 ENCL 25

Carolina Power 8 Light Company Harris Nuclear Plant PO Box 165 New Hill NC 27562 JuL a 1997 U.S. Nuclear Regulatory Commission Serial: HNP-97-137 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 97-015-00 Sir or Madam:

In accordance with Title 10 to the Code of Federal Regulations, the enclosed Licensee Event Report is submitted. This report describes inadequate testing for the Auxiliary Feedwater System Flow Control Valves.

Sincerely, J. W. D nahue Director of Site Operations Harris Plant MV Enclosure W&~> I c: Mr. J. B. Brady (HNP Senior NRC Resident)

Mr. L. A. Reyes (NRC Regional Administrator, Region II)

Mr. V. L. Rooney (NRC - NRR Project Manager) 9707090073 970702 lllllllllllllllllilllllllllllllllllllll PDR ADQCK 05000400 PDR State Road 1134 New Hill NC

9 NRC R)RM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 H.%l EXPIRES 04/30/98 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THI MANDATORY INFORMATION COllECTION REQUEST: 500 HRS. REPORTED lESSONS LEARNED ARE INCORPORATED (NTO THE UCENSING PROCESS ANO FED BACK TO INDUSTRY.

LICENSEE EVENT REPORT (LER) FORWARD COMMENTS REGARDU(G BURDEN ESTIMATE TO THE INFORMATION ANO RECORDS MANAGEMENT BRANCH IT@ F33L US. NUCLEAR REGULATORY COMM(SSlON, (See reverse for required number of WASHINGTON, OC 205550001. ANO TO THE PAPERWORK REDUCTION PROIECT (3(50 0(04L OFFICE OF MANAGEMENT AND BUDGET. WASHINGTOIL OC 209K digits/characters for each block)

DOCKET NUMBER (3( PAGE (3I FACIUTY NAME (1)

Harris Nuclear Plant Unit-1 50-400 1 OF 3 TTTLE (4(

Inadequate Auxiliary Feedwater System Flow Control Valve surveillance testing resulting in Technical Specification violation.

EVENT DATE (5) LER NUMBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED (8)

FAC(UTY NAME DOCKETNUMBER SEOUENTIAL REVSION MONTH OAY YEAR MONTH OAY YEAR NUMBER NUMBER FACIUTY NAME DOCKET NUMBER 97 97 015 00 2 97" 05000 OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR 5: (Check one or more) (11)

MODE (9) 20.2201(b) 20.2203(a)(2)(v) 50.73(a)(2)(i) 50.73(a) (2) (viii) 20.2203(a)(1) 20.2203(a)(3)(i) 50.73(a)(2) (ii) 50.73(a) (2) (x)

POWER 0%

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 50.36(c) (1) 50.73(a) (2) (v) Specity in Abstract below

20. 2203(a) (2) (iii) or in NRC Form 3BBA 20.2203(a)(2) (iv) 50.36(c)(2) 50.73(a) (2) (vii)

LICENSEE CONTACT FOR THIS LER (12)

NAME TELEpHoNE NUMBER (rircrude Area cede)

Michael Verrilli Sr. Analyst - Licensing (919) 362-2303 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT l13)

REPORTABLE REPORTABLE CAUSE SYSTEM COMPONENT MANUFACTURER CAUSE SYSTEM COMPONENT MANUFACTURER TO NPROS TO NPROS

<,r, 4(,,A.B YEAR SUPPLEMENTAL REPORT EXPECTED (14) EXPECTED MONTH OAY YES SUBMISSION (It yes, complete EXPECTED SUBMISSION DATE).

X NO DATE (15)

ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewrinen lines) l16)

On June 2, 1997, with the plant in mode 4 for refueling outage 7, a deficiency was identified related to previous testing of the Auxiliary Feedwater (AFW) System flow control valves (FCV). Specifically, Technical Specification (TS) 4.7.1.2.1 requires the motor driven AFW pump FCVs to open upon receipt of an auto open signal. This requirement was added by amendment 42 to the facility operating license following a plant modification (PCR-6502), completed in April 1994 during refueling outage 5. PCR-6502 was installed to provide an automatic open feature for the FCVs, which would allow AFW flow to be throttled during plant start-up activities. Operations surveillance test procedures have tested these valves on a quarterly basis, but did not verify their ability to open during high differential pressure conditions that exist during plant start-up while in mode 3, with low Steam Generator pressures present. In May 1994, Operations personnel identified that the FCVs experienced sticking problems when S/G pressures were very low while in mode 3. To address this condition, maintenance was performed on the valve actuators and procedures were revised to alert operators of this condition and provide guidance on how to open the FCVs if they stuck closed. During additional. investigation in March, 1996, testing confirmed that the FCVs would not open from the fully shut position if steam generator pressures were lessQan 320 psig.

However, during these previous instances, plant personnel did not consider the condition to be mode limiting or reportable as a technical specification surveillance requirement violation.

This condition was caused by the failure to recognize the impact on TS 4.7.1.2.1 that the AFW FCV actuator deficiencies created when low steam generator pressures were present while in mode 3. This resulted in past surveillance testing that did not completely satisfy the requirement for the AFW FCVs to respond as required per Technical Specification 4.7.1.2.1.

Corrective actions included readjusting the AFW FCV actuator spring pre-loading and testing the FCVs to verify their operability. The site surveillance test scheduling process will also be revised to ensure that future testing of the FCVs is erformed with the hi h differential ressure conditions resent,

HRC FORM 36BA US. NUCLEAR REGULATORY COMMISSION IL-95l LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACIUTY HAME 11) DOCKET LER NUMBER IBI PAGE LTI SEOUENOAL REVlQON YEAR NUMBER NUMBER Shaaron Harris Nuciaar Plant ~

Unit ftl 50400 2 OF 3 97 - 015 - 00 TEKT rSF sooso sposors soqopod, ooo odif6v>>l sop>>s or NRC Fano NSV IITi EVENT DESCRIPTION:

On June 2, 1997, with the plant in mode 4 for refueling outage 7, a deficiency was identified related to previous testing of the Auxiliary Feedwater (AFW) System flow control valves (FCVs, 1AF-49, 1AF-50, & 1AF-51, EIIS Code: BA-FCV). Specifically, Technical Specification 4.7.1.2.1 requires the motor driven AFW pump FCVs to open upon receipt of an auto open signal. This requirement was added by amendment 42 to the facility operating license following a plant modification (PCR-6502) installed in April 1994 during refueling outage 5. PCR-6502 was installed to provide an automatic open feature for the FCVs, which would allow AFW flow to be throttled during plant start-up activities.

Operations Surveillance Test procedures (OST-1044 and OST-1045) have tested these valves on a quarterly basis, but did not verify their ability to open during high differential pressure conditions that exist during plant start-up while in mode 3 with low Steam Generator pressures present.

In May 1994, Operations personnel identified that the FCVs experienced sticking problems when S/G pressures were very low while in mode 3. To address this condition, maintenance was performed on the valve actuators and procedures were revised to alert operators of this condition and provide guidance on how to open the FCVs if they were stuck closed. During additional investigation in March, 1996, testing confirmed that the FCVs would not open from the fully shut position if steam generator pressures were less than 320 psig. However, during these previous instances, plant personnel did not consider the condition to be mode limiting or reportable as a technical specification surveillance requirement violation.

The FCV issue was raised again during plant start-up following refueling outage 7, when testing (EPT-711) revealed that the valves would not open at low steam generator pressures. Investigation determined that the FCV actuator spring pre-loading was not set to allow maximum valve performance. This was resolved by readjusting the actuator spring pre-loading which allowed the valve to open as designed. Testing was performed on June 3, 1997 to confirm the valves operability, prior to mode 3 entry, which proved that the FCV material equipment problems had been resolved and that the FCVs would open against the high differential pressures associated with low steam generator pressure.

Based on the above, during the period from refueling outage 5 to May, 1997, Technical Specification 4.7.1.2.1 for the testing of the auto open signal was not properly performed. This constitutes a violation of Technical Specifications and is being reported per 10CFR50.73.a.2.i.

The turbine driven AFW pump FCVs were not effected by this condition. PCR-6502 did not install an automatic open feature for these valves. Also, high differential pressures would not exist across these FCVs since the turbine driven AFW pump discharge pressure is controlled by existing steam generator pressure.

CAUSE:

This condition was caused by the failure to recognize the impact on TS 4.7.1.2.1 that the AFW FCV actuator deficiencies created when low steam generator pressures were present while in mode 3. This resulted in past surveillance testing that did not completely satisfy the requirement for the AFW FCVs to respond as required per Technical Specification 4.7.1.2.1.

SAFETY SIGNIFICANCE:

There were no adverse safety consequences as a result of this event. The AFW FCV valves have operated as required in modes 1 and 2. The deficient condition only existed during plant start-up while in mode 3 with low steam generator pressures. Procedures were in place to provide guidance for opening a stuck AFW FCV if needed for controlling Steam Generator water level while shutdown or during plant start-up.

NRC FORM 366A LLS. NUCLEAR REGUIATORT COMMISSION

)4 96)

LlCENSEE EVENT REPORT (LER) 7DT COibiTINUATlON FACILITT NAME )1) OOCXET LER NUMBER I6) PAGE 0)

SEOUENTML REVISION TEAR NUMBER NUMBER Shearon Harris Nuclear Plant ~ Unit ¹1 50400 3 OF 3 97 - 015 00 TEXT Pl mvrv srrssvds rsqvdsd, vsv srrd)rdmsl ssrrrss v! JVRC form 36QI Ill)

PREVIOUS SIMILAREVENTS:

LER )l'95-15 was submitted on January 11, 1996 and reported the failure to perform response time testing on the AFW FCVs following the installation of the auto-open modification. Though related to testing of the AFW FCVs, the corrective actions contained in LER 95-15 focused primarily on the response time issue and would not have prevented the event discussed in this LER.

CORRECTIVE ACTIONS COMPLETED:

1. The AFW FCV actuator spring pre-loading was adjusted June 3; 1997 to ensure that the valves would open during periods of low steam generator pressure. Testing was also performed on June 3, 1997, prior to mode 3 entry, to verify the operability of the FCVs.

2. This event was reviewed in detail by plant management. This included a discussion on how this condition should have been previously identified as a TS compliance issue and emphasized the importance of management aggressively addressing and resolving plant problems.

CORRECTIVE ACTIONS PLANNED:

1. The site surveillance test scheduling process will be revised to ensure that the AFW FCVs are tested once per 1S months with the high differential pressure present. This will be completed by August 15, 1997.

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