ML17329A044

LER 90-016-01:on 901221,determined That Two of Three Safety Valves Had Lift Settings Outside TS 4.4.3 Acceptance Criteria.Possibly Caused by Steam Cutting in Disc Insert & Nozzle.Valves Satisfactorily tested.W/910607 Ltr

ML17329A044
Person / Time
Site:	Cook
Issue date:	06/07/1991
From:	Blind A, Weber G INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-90-016, LER-90-16, NUDOCS 9106120235
Download: ML17329A044 (5)

Text

ACCELERATED DISTRIBUTION DEMONST$&TION SYSTEM

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REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9106120235 DOC.DATE: 91/06/07 NOTARIZED: NO DOCKET g FACIL:50-315 Donald C. Cook Nuclear Power Plant, Unit 1, Indiana & 05000315 AUTH. NAME AUTHOR AFFILIATION WEBER,G.A. ~ Indiana Michigan Power Co. (formerly Indiana & Michigan Ele BLIND,A.A. , Indiana Michigan Power Co. (formerly Indiana -& Michigan Ele RECIPIENT AFFILIATION R

'ECIP.NAME 90-016-01:on 901221,determined that two of three safety

SUBJECT:

LER valves had lift settings outside TS 4.4.3 acceptance criteria. Possibly caused by steam cutting in dxsc insert &

nozzle. Valves satisfactorily tested.W/910607 ltr.

DISTRIBUTION CODE'E22T COPIES RECEIVED LTR ENCL 2 SIZE LIncident TITLE: 50.73/50.9 Licensee Event Report (LER), Rpt, etc.

NOTES RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD3-1 LA 1 1 PD3-1 PD 1 1 COLBURN,T. '1 1 INTERNAL: ACNW AEOD/DS P/TPAB NRR/DET/ECMB 9H 1,

2 1

2 1

1 AEOD/DOA AEOD/ROAB/DS P NRR/DET/EMEB 7E 1

2 1

1

.2 1

NRR/DLPQ/LHFB10 1 1 NRR/DLPQ/LPEB10'RR/DREP/PRPB11 1 1 NRR/DOEA/OEAB 1 1 2 2 NRR/DST/SELB 8D 1 1 NRR/DST/SICB8H3 1 1 NRR/DST/RPLB8D1 1 1 NRR/DST/SRXB 8E 1 1 REGALE~ 0~ 1 1 RES/DSIR/EIB 1 1 RGN3~QQLE 0 1+ 1 1 EXTERNAL EG &G BRYCE E J ~ H 3 3 L ST LOBBY WARD 1 1 NRC PDR 1 1 NSIC MURPHY,G.A 1 1 NSIC POORE,W. 1 1 NUDOCS FULL TXT 1 1 R D

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D NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT. 20079) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!

FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 31 ENCL 31

Indiana Michigan~

Power Company

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Cook Nuclear Plant One CooK Place Bridgrnan, Ml 49106 616 465 5901 INDIANA NICHIGAN POWER June 7, 1991 United States Nuclear Regulatory Commission Document Control Desk Rockville, Maryland 20852 Operating Licenses DPR-58 Docket No. 50-315 Document Control Manager:

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

90-016-01 Sincerely, A.A. Blind Plant Manager AAB:sb Attachment c D.H. Williams, Jr.

A.B. Davis, Region III E. E. Fitzpatrick P.A. Barrett B.F. Henderson R.F. Kroeger B. Walters . Ft. Wayne NRC Resident Inspector T. Colburn NRC J.G. Keppler M.R. Padgett G. Charnoff,, Esq.

Dottie Sherman, ANI -Library D. Hahn INPO S.J. Brewer/B.P. Lauzau B.A. Svensson .

9i06i202 910607 PDR ADOCK 0 0003l.=

8 FDR

NRC F<<m 355 U.S. NUCLEAR REOULATORY COMMISSION (9 $ 3 I APPROVED OMB NO. 31500101

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LlCENSEE EVENT REPORT (LER) EXPIRES; 5/31/SS FACILITY NAME I'I) DOCKET NUMBER (2) PA E D. C. Cook Nuclear Plant Unit 1 o 5 o o o31 5>oF03 Failure of two pressurizer safety valves to meet Technical Specification required surveillance test criteria EVENT DATE l5) LER NUMBER (5) REPORT DATf {7) OTHER FACILITIES INVOLVED (5)

MONTH DAY YfAR YfAR SFOUENTIAL MONTH OAY YEAR FACILITYNAMES DOCK'ET NUMBER(5)

NUMBER NS NUMBER 0 5 0 0 0 1 2 21 909 0 016 01 06 07 9 1 0 5 0 0 0 THIS REPORT IS SUBMITTED PURSUANT T 0 THE REQUIREMENTS OF 10 CFR (i: (Chtc>> one ot mote Of the foliem'nPI (11 OPE RAT( NO MODE (5) 5 20A02(bl 20A05(cl 50.73( ~ ) (2)(ivl 73.71(4)

POWER 20A05(cl(1){il 50.35(c) ill 50.7 3(t ) (2)(r) 73.71(cl LEVEL p p 0 20A05(t ) {I)(4) 50.35(c) {2) 50.73(el(2) (v4) OTHER (Sptcify in Abc(rect

"~y'  % 'xcc Oeiorr end In Text, NRC Form 10AOS(e) (1 )(iii) X 50.73(el {2)(i) 50.73( ~ I (2) (riii)(A) 366A) 20A05 {c ) (I ) (Iv) 50.73( ~ l(2)(4) 50.73( ~ ) (2lbi4) (B) .

%4xci .hi 20AOS( ~ ) (I I {v) (i( i) '0.73(c)(2) 50.73(cl(2)(x)

LICENSEE CONTACT FOR THIS LER {12)

NAME TELEPHONE NUMBER G. A. Weber - Plant Engineering Superintendent AREA CODE 16 465 -59 01 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

MANUFAC EPORTABLE MANUFAC. EPORTABL CAUSE SYSTEM COMPONENT TURER TO NPROS COMPONE NT TURER TO NPRDS

+LE.'YSTEM A 3 R V C 710 (nt i(LIST. Nk+4%

SUPPLEMENTAL REPORT 'EXPECTED ()Cl MONTH DAY YEAR EXP E CTED SUBMISSION DATE (15)

Yf5 fifyet, compittt EXPECTED SVShtiSSIOhi OATEI NO ABSTRACT (Limit tO te00 tPtcet. I ~, ePProeimtttiy fiftttn tlnPlt tPtte tyPtnntttn finn( (15)

This revision is being submitted to provide additional information regarding the Cause Description and Corrective Actions Statement..

On December valves sent to 21, 1990 a

it was determined that two of three pressurizer safety test laboratory off site for testing required by Technical Specification 4.4.3 had been found with li.ft settings outside of the acceptance criteria. Acceptable settings are between 2461 psig and 2509 psig.

Valve 1-SV-45B lifted at 2451 psig and valve 1-SV-45C at 2548 psig.

An evaluation of the test data revealed that, the valve with a 10 psi below acceptance criteria would have had no impact on the operability lift setting of of affected components, as lift it would have lifted prematurely. The valve with a setting exceeding the acceptable range did so by 39 psi. Calculations performed indicate that if the reactor coolant system pressure had reached 2548 psig, no damage would have been incurred by pipe, fittings, valves and/or the pressurizer. The UFSAR was reviewed for impact of the as-found setpoints.

The high setpoi.nt would have been offset by the additional capacity of power operated relief valves and full safety valve capacity i.s not required. Also reviewed were criteria associated with Departure from Nucleate Boiling. No adverse. consequences applied to safety analyses.

NRC Fntm 355 (9 83 I

U.S. NUCLEAR REOULATORY COMMISSION NRC Form 3SSA (9431 LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APPRovEooMSN0.3(so-oloo EXPIRES: 8/3'I/88 ~

FACILITY NAME (I) OOCKET NUMBER (31 LER NUM8ER ISI PACE (3)

YEAR SEQUENTIAL w+P,'EVISION NUMSER NUMSER D. C. Cook Nuclear Plant Unit 1 0 5 o 0 o 3 159 0 0 1 6 0 1 0 2 oF 03 TEXT /// moro Eooco /F ror/u/rorL u>> I/tuuoo/H/IC Form 38843/ (IT)

This revision is being submitted to provide additional information regarding the Cause Description and corrective actions taken.

Conditions Prior to Occurrence Unit One << Mode 5 (Cold Shutdown).

Unit.Two - Mode 1 (Power Operation).

esc it o of Event On December valves 21, 1990 it was determined that two of the three pressurizer lift settings outside Technical Specification safety (EZZS/AB-RV) had 3.4.3 acceptance criteria. The safety valves are tested at a test laboratory using steam at nominal temperature and pressure, as required by Technical Specification Surveillance 4.4.3. The valves are required to lift between 2461 and 2509 psig. Replacement valves tested by the same laboratory have been installed in place of the three valves removed for testing. Valve 1-SV 45B lifted at 2451 psig and 1-SV-45C lifted at 2548 psig. The third valve, SV-45A, was acceptable. Technical Specification 4.4.3 requires that each pressurizer code safety valve be demonstrated operable per Section XI of the ASME Boiler and Pressure Vessel Code, 1974 Edition.

Cause of Event The vendor could not conclusively determine the cause for 1-SV-'45B or 1-SV-45C to have unacceptable lift setpoints.

However, the vendor reported that inspection of 1-SV-45C revealed that the disc insert and nozzle showed signs of steam cutting. The steam cuts were evidenced by the removal of 0.002 inches of material du ing lapping. Steam cuts cause leakage, which will tend to increase the lift setpoint pressure.

The cause for 1-SV-45B to have a low lift setpoint could not be determined.

These failures are considered isolated events and not indicative'of generic failures of the Pressurizer Safety Valves. The previous reportable event (LER 50-316/89-04) was attributed to setpoint drift and is also considered to be an isolated event.

Anal sis of Event Test data was assessed following receipt. The valve which was found with a lift setting of 10 psi below acceptance criteria would have had no impact on the operability of affected components. It would have lifted prematurely which and the plant maintained. in a safe condition. The valve with a lift setting exceeded acceptance criteria did so by 39 psi. Calculations subsequently performed indicate that, had the reactor coolant system reached the setpoint lift pressure of 1-SV-45C, no damage would have been incurred by the pipe, fittings, valves, and/or the pressurizer.

The UFSAR Chapter 14 Safety analyses applicable to Unit 1 were reviewed .for the impact of "as found" safety valve opening setpoints of 2451 psig for 1-SV-45B and 2548 psig for 1-SV-45C.

~ C.S. CPOr ISSS ESP-SS> Poll I NRC FORM SESA IO O'I,

NRC Form 388A Ug. NUCLEAR REOULATORY COMMISSION 1843)

LICENSEE EVENT REPORT (I.ER) TEXT CONTINUATION APPROVEO OMB NO. 3I50&IIM EXPIRES: 8/31/BB FACILITYNAME )1) DOCKET NUMBER l3) LER NUMBER )8) PACE LE)

YEAR PS& SEOUSNTIAL NUM 84

~< REVISION vega NUM>>%4 D. C. Cook Nuclear Plant Unit 1 o 5 o o o3 15 90 0 1 6 0 1 p 30F p 3 TEXT ///more <<r>>c>> /1 />>1/r>>IL Iree e//oer'on/ //RC %%dmr 3554'c/ ) IT) sis of Eve t Continued For over-pressurization cases, the additional capacity of the PORVs would offset the higher final setpoint, and the full safety valve capacity is not required for any case. Therefore, the increased setpoint of the higher value is of no, concern.

For Departure from Nucleate Boiling (DNB) cases, the lowered setpoint would have been reached for only one case, the loss of external load/turbine trip case, minimum feedback, with pressure control. The accident analysis minimum DNBR for thi.s case is approximately 1.79, and the acceptable value is 1.45, so the slight pressure reduction i.s easily offset by the existing margin.

Zn conclusion, the specified "as found" pressurizer safety valve setpoint would have no consequence on the applicable safety analyses. The failure of the valves to meet the surveillance test criteria are being reported under 10CFR50.73 (a)(2)(i),(B), "Any operati.on or condition prohibited by the plant's Techni.cal Specifications."

Corrective Action r Valves 1-SV-45B and 1-SV-45C required lapping of the disc and nozzle seats and were satisfactorily teated for steam set pressure and seat leakage.

Failed Com onent Identification

'Pressurizer Safety Valve Plant Designation: 1-SV-45B and 1-SV-45C Manufacturers Crosby Valve Company Model: HB-86-BP EZIS Code: AB-RV revious Similar Events LER 50-316/89-04 L 8 O'POI 988 S)/r SS9 >>O'I r N4C CORM SSSA