ML17325A736

LER 88-004-00:on 880420,discovered Six of Twenty Main Steam Safety Valves Lift Setpoints Out of Spec During Surveillance Testing.Caused by Testing Method & Setpoint Drift.Safety Valves Setpoints reset.W/880512 Ltr

ML17325A736
Person / Time
Site:	Cook
Issue date:	05/12/1988
From:	Droste J, Will Smith AMERICAN ELECTRIC POWER CO., INC., INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
LER-88-004, LER-88-4, NUDOCS 8805260368
Download: ML17325A736 (6)

Text

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REGULAT 'NFORMATION DISTRIBUTION YSTEM <RIDS)

ACCESSION NBR 88052*03*8 DOC. DATE: 86/05/12 NOTARIZED: NQ DQCl'8 =

FAC IL: 50-316 Donald C. Cook Nuclear Power Plant. Unit 2> indiana 'c 0500031a AUTH. NANE *UTHOR AFFILIATION DROSTE> J. B. *merican Electric Poeer Co. > Inc.

SMITH> W. G. American Electric Power Co. > Inc.

SNITH> W. G. Indiana Nichigan Poeer Co. < formerly Indiana 'c Nichi gan Ele REC IP. NAME RECIPIENT AFFILIATION

SUBJECT:

LER SS-004-00: on 880420> discovered six of twenty main steam safety valves surveillance lift setpoints out of specification during testing. Caused bg testing method ~i setpoint drift. Safety valves setpoints reset. W/880512 ltr.

DISTRIBUTION CODE: IE22D COPIES RECEIVED: LTR J ENCL ( SIZE:

TITLE: 50. 73 Licensee Event Report <LER) Incident Rpt> etc.

NOTES:

REC IP IENT CQP IES RECIP IENr COP I ES ID CODE/MANE LTTR ENCL ID CODE/MANE LTTR ENCL PD3-.1 LA 1 1 PD3-1 PD 1 1 STANG> J 1

.INTERNAL: ACRS MICHELSQN 1 1 ACRS MOELLER 2 2 AEQD/DOA 1 1 AEOD/DSP/NAS 1 1 AEOD/DSP/ROAB 2 .

2 AEOD/DSP/TPAB 1 1

• RN/DCTS/DAB 1 1 DEDRO 1 1 NRR/DEST/ADS 7E 1 0 NRR/DEST/CEB SH 1 1 NRR/DEBT/ESB SD 1 1 NRR/DEST/ICSB 7 1 1 NRR/DEST/NEB 9H 1 1 NRR/DEST/NTB 9H 1 1 NRR/DEST/PSB SD 1 NRR/DEST/RSB SE 1 1 NRR/DEST/SGB 8D 1 1 NRR/DLPG/HFB 10 1 1 NRR/DLPG/GAB 10 1 1 NRR/DOE*/EAB 11 1 1 NRR/DREP/RAB 10 1 1 NRR/DREP/RPB 10 2 2 NRR/DRIS/SIB 9A 1 1 N - ~ /ILRB12 1 1 NUDOCS-ABSTRACT 1 1 EG- Q2 1 1 RES TELFORD> J 1 1 RES/DE/EIB 1 1 RES/DRPS DEPY 1 1 RGN3 FILE 01 1 EXTERNAL: EG5G WlLLIANS>S 4 4 FORD BLDG HOY, A 1 H ST LOBBY WARD 1 1 LPDR 1 1 NRC PDR 1 1 NSIC HARRIS> J 1 1 NSIC NAYS> G 1 1 TOTAL NUMBER OF COPIES REQUIRED: LTTR 46 ENCL 45

NRC Form 368 U.S. NUCLEAR REGULATORY COMMISSION (943) APPROVED OMB NO. 31504)IOS EXPIRES: 8/31/58 LICENSEE EVENT REPORT (LER)

FACILITY NAME (II DOCKET NUMBER l2) PAGE 3)

D. C. Cook TITLE Iel Nuclear Plant Unit 2 050003 I $ 1OFP4 Main Steam 'Safety Valves Out of Specification Due to A arent Set oint Drift EVENT DATE ISI LER NUMBER (81 REPORT DATE (7I OTHER FACILI'TIES INVOLVED (8)

MONTH DAY YEAR YEAR SEOVSNTrAL cos ss8vtsloN MONTH OAY YEAR FACILITYNAMES OOCKFT NUMBER(S)

NUMSEsr  %$ NVMSER 0 5 0 0 0 042 088 8 8 0 0 0 512 8 8 0 5 0 0 0 THIS REPORT IS SUBMITTED PURSUANT T 0 THE REQUIREMENTS OF 10 CFR (): (Chech one or more of the follovflnpl (11)

OP E RATING MODE (9) 20A02(8) 20.406 I el 50.73(sl(2) Ov) 73.71(8)

POWER 20A05( ~ l(ll(II 50.38(cl (I ) 50.73(s)(2) (v) 73.71(c)

LEVEL p 8 p 20A05(s) (1)(4) 50.38 (c) (2) 50.73(e) (2)(v4) OTHER ISpeclfy In Ahstrect jp

j%Nj$$  : . P 40~~', 20A05( ~ )(1((4)l 20A06(e) (I ) (Ivl 50.73(s) 12)(ll 50.73( ~ ) (2) (4) 50.73(e)l2)(vBI)(A) 50.73(sl(2) (vlBII 8)

Oeiow end ln Text, HRC Form 366AI

%sestet':.i.s.? s?,ice 20A05(e) (I ) (vt 50.73(sl(2)(41) 60.73( ~ ) l2) (x)

LICENSEE CONTACT FOR THIS LER (12I NAME TELEPHONE NUMBER AREA CODE J. B. Droste erintendent Maintenance Su COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13) 9 MANUFAC. EPORTABLE 9 5U'O SYSTE' MANUFAC.

CAUSE SYSTEM COMPONENT NpROS . CAUSE COMPONENT TURER TURER o  ?

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B RV D2 43 , AMN I%Ãj5%:

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SUPPLEMENTAL REPORT EXPECTED (Ie) MONTH OAY

?

YEAR EXPECTED SUBMISSION DATE (16)

YES (if yes, complete EXPECTED SUBMISSION DATE) NO ABSl'RACT ILImlt to le00 speces, I.e., epproxlmetely fifteen sinpleepece typevrvitten linnl (15)

Between April 19 at 1610 hours0.0186 days <br />0.447 hours <br />0.00266 weeks <br />6.12605e-4 months <br /> and April 20 at 1811 hours0.021 days <br />0.503 hours <br />0.00299 weeks <br />6.890855e-4 months <br />, 1988, with the Unit 2 reactor in Mode 1 (power operation) at 77 percent thermal power, six of the twenty Main Steam Safety Valves (MSSV) lift setpoints were found to be out of specification during surveillance testing. The MSSV lift setpoints ranged from 16 psi below to 6 psi above the Technical Specification required range. In each case the MSSVs'ift setpoints were corrected and the safety valves left operable prior to completion of the surveillance test procedure (STP).

The apparent MSSV setpoint drift could have been attributable to two factors, 1) testing method, and; 2) setpoint drift due to valve design/

application. The investigation concluded that the old testing method had a high probability of contributing to the apparent MSSV setpoint drift.

The immediate corrective action, as required by the Surveillance Test Procedure, was to reset the safety valves setpoints to within their specified ranges utilizing an improved testing method. To prevent recurrance, future MSSV setpoints will be tested with the improved testing method. This will more accurately relfect the MSSV setpoints.

8805260368 880512 0500031((I PDR ADOCK S

NRC Form 388 (943)

NRC form 388A U.S. NUCLEAR REGULATORY COMMISSION (94)3)

LICENSEE E . T REPORT (LER) TEXT CONTINUA APPROVE'O OMB NO, 3150%104 EXPIRES: 8/31/88 FACILITY NAME (1) COCKET NUMBER (2) LER NUMBER (5) PAGE I3)

YEAR +so SEGUENTIAL SsPj REVISION NUMBER iNS NUMBER D. C. Cook Nuclear Plant Unit 2 0 5 0 0 0 3 ] 0 0 4 0 0 02" 0 4 TEXT //F moro 4/>>co /4 SII/rorL I>>o ////OI>>/ HRC Form 38((43/ (17)

Conditions Prior to Occurance Unit, Two 77 Percent Reactor Thermal Power Descri tion of Event Between April 19 at 1610 hours0.0186 days <br />0.447 hours <br />0.00266 weeks <br />6.12605e-4 months <br /> and April 20 at 1811 hours0.021 days <br />0.503 hours <br />0.00299 weeks <br />6.890855e-4 months <br />, 1988, six of the twenty Unit 2 Main Steam Safety Valves (MSSV) (EIIS/SB-RV) were found out of specification during surveillance testing. The MSSV's lift setpoints lift setpoints ranged from 16 psi below to 6 psi above the Technical Specification required setpoints. In each case the MSSV was declared inoperable prior to performance of the surveillance test procedure. Upon discovery of the low setpoint values the MSSV setpoint was immediately adjusted to bring the value within the Technical Specification required range. The action statement for Technical Specification 3.7.1.1 was complied with during the performance of the testing.

The main steam header of each of the four steam generators (EIIS/SB-SG) in Unit 2 are equipped with 5 safety valves for a total of 20 valves per unit.

The required relief pressure setpoint ranges and their as-found conditions for MSSV's are listed below for the valves found out of specification:

Technical Valve Specification As-Found Valve Identification Steam Required Range Setpoint Date Serial No. No. Generator ( si ) ~(si I 4/19/88 BN¹ 6341 SV-3-2 1074-1096 1066 4/19/88 BN¹6324 SV-1A-3 1054-1076 1082 4/20/88 BN¹6335 SV-2B-3 1064-1086 1062 4/20/88 BN¹6340 SV-3-3 1074-1096 1058 4/20/88 BN¹6326 SV-1A-4 1054-1076 1045 4/20/88 BN¹6339 SV-2B-4 1064-1086 1060 With the exception of the MSSV's there were no inoperative structures, components, or systems that contributed to this event.

NRC FORM SMA *U 8 QPQ l 988 0 824 538/455 (94)3)

8 NRC Form 366A U.S, NUCLEAR REGULATORY COMMISSION

($ 83)

LICENSEE E T REPORT (LER) TEXT CONTINUA . APPROVEO OMB NO, 3160-0104 EXPIRES: 8/31/88 FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)

SEOI/ENTIAL REVISION NUMBER i or NUMBER D. C. Cook Nuclear Plant Unit 2 6 0 0 0 3 8 8 0 3 OF 0 4 TEXT //P'/Rord BPBco /4 Bdrr/RRL Boo a//I/Boon/I HRC Forrrr 388l'4/ (13)

Cause of Event The apparent MSSV setpoint drift could have been attributed to two factors,

1) testing method, and; 2) setpoint drift due to valve design/application.

The investigation concluded that the old testing method was inherently less accurate and had a high probability of contributing to the apparent MSSV setpoint drift.

Anal sis of Event The safety valve setpoints discovered in this event. were found to be out of compliance with the Technical Specification (T/S) 3.7.1.1 requirements and therefore reportable per 10CFR50-73 (a)(2)(i)(B).

The following PSAR Chapter 14 accident analyses consider secondary-side pressure relief:

l. Loss of External Electrical Load (Appendix 14C.3.6)

2. Loss of Normal Feedwater (Appendix 14C.3.7)

3. Loss of All A.C. Power to the, Station Auxiliaries

4. Steam Generator Tube Rupture (14.2.4)

5. Loss of Reactor Coolant from Small Ruptured Pipes or from Cracks in Large Pipes which Actuates the ECCS (Appendix 14E.l)

The low as-found setpoints would not have affected the capacity to relieve the Steam Generator pressures during any analyzed event. Since the low "as-found" lift setpoints would have resulted in the early opening of the MSSV's, and the deviation from the T/S setpoints was small, this situation would not have adversely affected plant safety. The. lowest "as-found" setpoint found during this event was 1045 psi. This value is higher than lift the lowest as-found lift setpoint found during a similar event on November 2, 1985 in Units 1 and 2 and is bounded by the safety analysis submitted for that event. The only foreseeable ramification (of the low setpoint deviation) would be a potential actuation of the safety valve prior to the acutation of the steam generator power relief valve which is usually set to actuate at 1025 psi. This foreseeable ramification should not have occurred in,the present case, since the lowest psi. Two MSSV's in the train where one valve failed high were within the lift setpoint found was 1045 acceptable range. Since their operability conforms with the design criteria for over pressure protection and since the setpoint deviation was small (6 psi) the as-found setpoint would not have affected the capacity to relieve the steam generator pressure during any analyzed event.

NRC FORM SBBA *U S GPO 19884 824 5381466 (84)3)

NRC Form 368A U.S NUCLEAR REGULATORY COMMISSION (94)3)

LICENSEE E I T REPORT ILER) TEXT CONTINUA APPROVED OM8 NO, 3150-0104 EXPIRES: 8/31/88 FACILITY NAME ul DOCKET NUMSER (2) LER NUMBER (6) PAGE (3)

YEAR en.'5 SEGVENTIAL pic REVISION RH NUMBER :CP'r. NUMBER D. C. Cook Nuclear Plant Unit 2 0 s 0 o 0 31 688 0 0 040F04 TEXT //mere e>>ce /4 BEI/)er/, Iree ~ /I/Oee/HRC Form 86549/ (IT)

Corrective Action The immediate corrective action, as required by the Surveillance Test Procedure, was to reset the safety valves setpoints to within their specified ranges. The lift setpoints were adjusted and tested utilizing the improved testing method. To prevent recurrance, future MSSV setpoints will be tested with the improved testing method. This will provide a more accurate determination of the MSSV setpoints.

Failed Component Identification Main Steam Safety Valve Plant Designation: SV-lA-3 (('; SV-2B-3 6 4; SV-3-2 6 3 Manufacturer: Dresser Consolidated Valves Model: 3707RA-RT22 EIIS Code: SB-RV Previous Similar Events LER 50-315/86-020-00 LER 50-315/87-011-00 NRC FORM 346A *U.S.GPO:1986 0.524 538/455 (94)3)

Indiana Michigan Power Company Cook Nuclear Plant PO- Box -'58 Srtdgrnan. Ml 49106 616 465 5901 INOIANA NICHIGAM May 12, 1988 United States Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555 Operating License DPR-58 Docket No. 50-316 Document Control Manager:

In accordance with the criteria established by 10 CFR 50.73 entitled Licensee Event Re ortin S stem, the following report is being submitted:

88-004-00 Sincerely, W. G. Smith, Jr.

Plant Manager WGS:clw Attachment cc: D. H. Williams, Jr.

A. B. Davis, Region III M. P. Alexich R. F. Kroeger H. B. Brugger R. W. Jurgensen NRC Resident Inspector J. F. Stang, NRC R. C. Callen G. Charnoff, Esq.

Dottie Sherman, ANI Library D. Hahn INPO PNSRC A. A. Blind P. A. Barrett/P. Lauzau