ML17332A758

LER 95-003-00:on 950203,two Pressurizer Safety Valves Failed to Meet TS Required Surveillance Test Criteria.Partially Disassembled Valves 2-SV-45A & 2-SV-45C.W/950508 Ltr

ML17332A758
Person / Time
Site:	Cook
Issue date:	05/08/1995
From:	Blind A, Weber G INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-95-003, LER-95-3, NUDOCS 9505150175
Download: ML17332A758 (6)

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CEI.ERATED l<IDS PROCI'.SSIiG)

REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9505150175 DOC.DATE: 95/05/08 NOTARIZED: NO DOCKET FACIL:50-316 Donald C. Cook Nuclear Power Plant, Unit 2, Indiana M 05000316 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 P RECIP.NAME RECIPIENT AFFILIATION

SUBJECT:

LER 95-003-00:on 950203,two pressurizer safety valves failed to meet TS required surveillance test criteria. Partially disassembled valves 2-SV-45A & 2-SV-45C.W/950508 ltr.

DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR jENCL 0 TITLE: 50.73/50.9 Licensee Event Report (LER), IncidentQ Rpt, etc.

SIZE:

NOTES:

RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD3-1 PD 1 1 HICKMAN,J 1 1 INTERNAL: AEODJSPD/RAB 2 2 AEOD/SPD/RRAB 1 1 ILE'ENTERS 1 1 ,NRR/DE/ECGB 1 1

/DE/EELB 1 1 NRR/DE/EMEB 1 1 NRR/D IS P/PI PB 1 1 NRR/DOPS/OECB 1 1 NRR/DRCH/HHFB 1 1 NRR/DRCH/HICB 1 1 NRR/DRCH/HOLB 1 1 NRR/DSSA/SPLB 1 1 NRR/DSSA/SPSB/B 1 1 NRR/DSSA/SRXB 1 1 RES/DSIR/EIB 1 1 RGN3 FILE 01 1 1 EXTERNAL: L ST LOBBY WARD '

1 1 LITCO BRYCE,J H 2 2 D NOAC MURPHY,G.A 1 1 NOAC POORE,W. 1 NRC PDR 1 1 NUDOCS FULL TXT 1 1 U

VOTE TO ALL"RIDS" RECIPIENTS:

I'LEASE I) E LP I:S TO REDUCE iVKSTE! CONTACT'I'IIE DOC!: ifEXT CONTROL DESk. ROOM I'1.37 (EXT. 304.2083 ) 'I 0 ELlhilb:KTE 5'OUR i'AifLFf(Oif DIS'I'R I!IUTION L IS'IS I'OR DOCI.'ifI'.X'I'S 5'Ol.'ON "I' I I I)!

FULL TEXT CONVERS1ON REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 26 ENCL 26

indiana Michigan Power Company Cook Nuclear Plant One Cook Place Bridgrnan, Ml 49106 616 465 5901 INQMNA NlCHIGAN PQWER May 8, 1995 United States Nuclear Regulatory Commission Document Control Desk Rockville, Maryland 20852 Operating Licenses DPR-74 Docket No. 50-316 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:

95-003-00 Sincerely, A. A. Blind Plant Manager

/mr Attachment c J. B. Martin, Region E. E. Fitzpatrick III P. A. Barrett R. F. Kroeger M.'A. Bailey Ft. Wayne NRC Resident Inspector J. B. Hickman NRC J. R. Padgett G. Charnoff, Esq.

D. Hahn INPO S. J. Brewer 9505150175 950508 PDR S

ADOCK 05000316 PDR y@$

NRC FORM 366 UCLEAR REGULATOAY COMMISSION PROVED BY OMB NO. 3150%104 15 92l EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECnON REOUEST. 50.0 HRS, FORWARD LICENSEE EVENT REPORT (LER) COMMENTS REGARDING BURDEN ESTIMATE TO THE iNFORMATCN AND RECORDS MANAGEMENT BRANCH (MNBB 77141. U.S. NUCLEAR REGULATORY COMMISSION. WASHINGTON, DC 205554001. ANO TO THE PAPERWORK REDUCTION PROJECT 131500104), OFFICE OF (See reverse for required number of digits/characters for each block) MANAGEMENTANO BUDGET. WASHINGTON, DC 20503.

FACILITY NAME (I) DOCKET NUMBER (2) PAGE (3)

D. C. COOK NUCLEAR PLANT UNIT 2 05000 316 1OF 4

(') FAILURE OF TWO PRESSURIZER SAFET TECHNICAL SPECIFICATION RE UIRED SURVEILLANCE 'TEST CRITERIA EVENT DATE 5 LER NUMBEA 6 REPORT NUMBER 7 OTHER FACILITIES INVOLVED 6 SEOUENTIAL REVISION FACIUlYNAME DOCKET NUMBER MONTH DAY YEAR YEAR MONTH DAY NUMBER NUMBER 05000 FACIUTY NAME DOCKET NUMBER 02 03 95 95 003 00 05 08 95 05000 OPERATING THIS REPORT IS SUBMIT)ED PURSUA NT TO THE REQUIREMENTS OF 10 CFR E: (Check one or m ore) 11 MODE (9) 20.402(b) 20.405(c) 50.73(a) (2) (iv) 73.71(b)

POWER 20.405(a) (1) (i) 50.36(c)(1) 50.73(a)(2)(v) 73.71(c)

LEVEL (10) 100 20.405(a) (1) (ii) 50.36(c) (2) 50.73(a) (2) (vii) OTHER 20.405(a) (1) (iii) X 50.73(a) (2) (i) 50.73(a) (2) (viii)(A) (speciry in Absrracr below ana in ress. NRC 20.405(a) (1)(iv) 50.73(a) (2) (ii) 50.73(a) (2) (viii)(B) Form 366A) 20.405(a) (1) (v) 50.73(a) (2) (iii) 50.73(a) (2) (x)

LICENSEE CONTACT FOR THIS LER 12 NAME TELEPHQNE NUMBER FncIvae Area coae)

G. A. WEBER PLANT ENGINEERING SUPERINTENDENT 616-465-5901 COMPLETE ONE LINE FOA EACH COMPONENT FAILUAE DESCRIBED IN THIS REPORT 13

~I~ix x:s'i 9'jj" P CAUSE SYSTEM COMPONENT REPORTABLE REPORTABLE MANUFACTURER CAUSE SYSTEM MANUFACTURER TO NPRDS TO NPRDS X AB RV C710 SUPPLEMENTAL REPORT EXPECTED 14 EXPECTED MONTH DAY YEAR YES SUBMISSION (II yes. compIere EKPECTED SUBMISSION DATE) No X DATE (15)

ABSTRACT (umit to 1400 spaces. i.e.,'approximately 15 single-spaced typewritten lines) (16)

On October 5, 1994 with Unit 2 in Mode 6 (no 'fuel), the Pressurizer Safety Valves in service prior to the Unit 2 Refueling Outage were removed for shipment to an offsite lab for set point testing. On February 3, 1995, with Unit 2 in Mode 1 at 100 percent reactor thermal power, Wyle Laboratories determined that two of three Unit 2 Pressurizer Safety Valves were found with lift settings outside the Technical Specification acceptance criteria. Acceptable settings are between 2461 psig and 2509 psig. Valve 2-SV-45A was found to have a lift setpoint of 2524 psig, and valve 2-SV-45C had a lift setpoint of 2538 psig. There was no safety significance because the safety valves would still have limited the peak transient pressure to 2615 psig in the event of an overpressure transient . This is below the Technical Specification safety limit of 2735 psig. No specific cause for the drift was determined. Both 2-SV-45A and 2-SV-45C were partially disassembled (retaining spring compression) and inspected. No problems were noted. The nozzle and disc seating surfaces were lapped and

'olished. The valves were reassembled and tested satisfactorily.

NRC FORM 366 15.92)

IQL NRC FORMSSBA U.S. NUCLEAR REGULATORY COMMISSION IBB9) APPROVED DMS NO. 31500104 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVENT REPORT ILERI INFORMATION COLLECTION REOUEST: 500 HRS, FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINuATION AND RFPORTS MANAGEMENT BRANCH IP-530). U.S. NUCLEAR REGULATORY COMMISSION. WASHINGTON, OC 20555. AND TO THE PAPERWORK REDUCTION PROJECT )3)500)0i). OFFICE OF MANAGEMENTANO BUDGET, WASHINGTON, DC 20503.

FACILITY NAME 11) DOCKET NUMBER )21 LER NUMBER (51 PAGE LS)

SEQUENTIAL FF"'EV>SION YEAR NVMFE4 i NVMIIFA D. C. COOK NUCLEAR PLANT UNIT 2 p 5 p p(p 3 1 6 9 5 0 0 3 0 0 02oF 0 4 TEXT N mpm uwce N mqvppd. Pw arFlponsl HRC %%dmr 355431 <) TI Conditions Prior to Occurrence Unit 2 - Mode 1, 100 percent Reactor Thermal Power (reflects condition at the time that Wyle Labs determined the event)

Descri tion of Event On October 5, 1994 with Unit 2 in Mode 6 (no fuel), the Pressurizer Safety Valves in service prior to the Unit 2 Refueling Outage were removed for shipment to an offsite lab for set point testing. On February 3, 1995, with Unit 2 in Mode 1 at 100 percent reactor thermal power, it was determined that two of the three Pressurizer Safety Valves (EIIS/AB-RV), Crosby Valve Model Numbers HB-BP-86 and HB-86-BP, had lift settings outside Technical Specification 3.4.3 acceptance criteria. Technical Specification Surveillance 4.4.3 requires that each Pressurizer Code Safety Valve be demonstrated operable per Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10CFR50, Section 50.55a. The safety valves are tested at a test laboratory using steam at nominal temperature and pressure. The valves are required to lift at 2485 psig +/- 1 percent, (i.e. between 2461 and 2509 psig). Valve 2-SV-45A lifted at 2524 psig and 2-SV-45C lifted at 2538 psig. The third valve, 2-SV-45B, had an acceptable lift setpoint of 2500 psig.

The test report was completed by Wyle Labs on March 8, 1995, and forwarded to AEPSC in Columbus, Ohio. The test report was sent to the D.C. Cook Plant in early April 1995. A condition report was written by the plant on April 7, 1995, and this event was determined reportable on April 13, 1995.'ause of Event The phenomena of pressurizer safety valve setpoint drift outside of design criteria experienced at the D.C. Cook Plant is similar to that experienced throughout the nuclear industry for pressurizer as well as main steam safety valves. Crosby Valve and Wyle Labs were contacted to discuss the test results, but no specific cause for the drift was determined.

NRC Fonn 35BA 1509)

NRC FORM 366A US, NUCLEAR REGULATORY COMMISSION IBBBI APPROVED 0MB NO.31500104 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVENT REPORT ILER) INFORMATION COLLECTION REQUESTI 50.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPOR'TS MANAGEMENT BRANCH IP.530I. U.S. NUCLEAR REGULATORY COMMISSION. WASHINGTON. DC 20555. AND TO THE PAPERWORK REDUCTION PROJECT 13150010OI. OFFICE OF MANAGEMENTAND BUDGET. WASHINGTON. DC 20503.

FACILITY NAME 111 DOCKET NUMBER 121 LER NUMBER 161 PAGE 13I YEAR Cq SEGVSNTIAL REVISION NUMBER NVMOSR D. C. COOK NUCLEAR PLANT UNIT 2 o 5 o o o 3 1 6 9 5 0 0 30 003 OF 0 4 TEXT llfmoro Jooco Io IoooRRS VM ooorooool HRC Form 36SA Bl I I TI Anal sis of Event This event has been determined reportable under the provisions of 10CFR50.73 (a) (2) (i) (B) as an operation prohibited by Technical Specification 3.4.3, which requires all of the Pressurizer Safety Valves to be operable with a lift setting'of 2485 psig +/- 1 percent.

The as-found lift setpoints of safety valves 2-SV-45A and 2-SV-45C did not have any actual impact on the Reactor Coolant System (RCS) since the pressure would not have exceeded the maximum transient limit of 2735 psig per ASME BEcPV Section'III, which is 110 percent of the design pressure (2485 psig). There was no impact on the health or safety of the public.

Safety Valve 2-SV-45C (worst case) had a lift setpoint of 2538 psig. The valve would have reached it's rated capacity at an RCS pressure of 2615 psig (2538 psig plus 3 percent accumulation). Valve 2-SV-45A would have attained its full rated lift at 2600 psig (2524 psig plus 3 percent).

Also, the reactor vessel and pressurizer were designed to ASME BkPV Section III which permits a maximum transient pressure of 2735 psig, 110 percent of design pressure (2485 psig). The RCS piping, valves and fittings are designed to ANSI B31.1, 1967 Edition, which permits a maximum transient pressure of 2985 psig, 120 percent of design pressure (2485 psig). In addition, the entire RCS was hydro tested to 3107 psig, 125 percent of design pressure (2485 psig), to demonstrate system integrity prior to initial operation.

In conclusion, this event did not have any safety significance and did not represent a hazard to the public health and safety. The safety limit of 2735 psig would not have been exceeded since the maximum RCS pressure would not have exceeded 2615 psig (2-SV-45C setpoint of 2538 plus 3 percent).

Corrective Action Valves 2-SV-45A and 2-SV-45C were partially disassembled (retaining spring compression) and the nozzle and disc seating surfaces were lapped and polished. The valves were then reassembled, retested and reset to comply with the acceptance criteria set point, and to verify that the seat leakage requirements were met.

NR C Fono 366A 16491

NRC FORM SSSA U.S. NUCLEAR REGULA'TORY COMMISSION ISBS) APPROVEO 0MB NO.3)500)04 ESTIMATED BURDEN PEA RESPONSE TO COMPLY WTH THIS LICENSEE EVENT REPORT ILERI INFORMATION COLLEC'TION REQUEST: 500 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND AEPORTS MANAGEMFNT BRANCH IP.530), U.S. NUCLEAR REGULATOAY COMMISSION. WASHINGTON. OC 20555, AND TO THE PAPERWORK REDUCTION PROJECT I31500104). OFFICE OF MANAGEMENTANO BUOGFT,WASHINGTON. OC 20503.

FACILITY NAME 111 DOCKET NUMBER )21 LER NUMBER IS) PAGE 13)

YEA A ~R)oo' SEOUSNTIAL OEVISION NMMOFA NUMBER D. C. COOK NUCLEAR PLANT UNIT 2 o s o o o3 16 5 0 0 3 000 4 OF 0 4 TEXT IR more eoece N reerered. o)e eeeeana'IYRC Form 3054'el 117)

Corrective Action continued Since a specific root cause could not be determined, no preventive action is planned at this time. AEPSC is continuing to participate in an ongoing dialogue with other utilities in an effort to identify ways to improve valve performance. Finally, as part of Unit 1 Cycle 16 and the Unit 2 Cycle 12, AEPSC is pursuing a change in pressurizer safety valve set point tolerances from +/- 1 percent to +/- 3 percent. For Unit 1 the Westinghouse analysis has been received and a request will be submitted to the NRC in the near future. It is anticipated that the analysis for Unit 2 will be completed by the end of 1995.

Failed Com onent Identification Pressurizer Safety Valve Plant Designation: 2-SV-45A Manufacturer: Crosby Valve Company Model: HB-BP-86 EIIS Code: AB-RV Pressurizer Safety Valve Plant Designation: 2-SV-45C Manufacturer: Crosby Valve Company Model: HB-86-BP EIIS: AB-RV Previous Similar Events

'ER:

50-315/90-16, 92-09, 94-04 LER: 50-316/89-04, 92-06 NAC Form 35BA ISe9)