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#REDIRECT [[05000315/LER-1994-004, :on 940406,three Pressurizer Safety Valves Sent to Offsite Test Lab for Testing for Failure to Meet TS Acceptance Criteria.Cause Not Determined.Nozzle & Disc Seating Surfaces Lapped & Polished]]
| number = ML17331B385
| issue date = 05/10/1994
| title = LER 94-004-00:on 940406,three Pressurizer Safety Valves Sent to Offsite Test Lab for Testing for Failure to Meet TS Acceptance Criteria.Cause Not Determined.Nozzle & Disc Seating Surfaces Lapped & polished.W/940510 Ltr
| author name = Blind A, Weber G
| author affiliation = INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
| addressee name =
| addressee affiliation = NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
| docket = 05000315
| license number =
| contact person =
| document report number = LER-94-004, LER-94-4, NUDOCS 9405190369
| document type = LICENSEE EVENT REPORT (SEE ALSO AO,RO), TEXT-SAFETY REPORT
| page count = 6
}}
 
=Text=
{{#Wiki_filter:ACCELERATED ITRIBUTION                          DEMONS'ATION                SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
ACCESSION NBR:9405190369            DOC.DATE: 94/05/10 NOTARIZED: NO                DOCKET g FACIL:50-315 Donald C. Cook Nuclear Power Plant, Unit 1, Indiana M                  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 RECIP.NAME          RECIPIENT AFFILIATION
 
==SUBJECT:==
LER  94-004-00:on 940406,three pressurizer safety valves, sent to to off site test lab for testing failed to meet TS acceptance criteria. Cause not determined. Nozzle & disc seating surfaces lapped          & polished.W/940510      ltr.
DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR ( ENCL TITLE: 50.73/50.9 Licensee Event Report (LER), Incident Rpt, etc.
0    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: AEOD/DOA                      1    1      AEOD/DSP/TPAB            1    1 AEOD/ROAB/DS P              2    2      NRR/DE/EELB              1    1 NRR/DE/EMEB                  1    1      NRR/DORS/OEAB            1    1 NRR/DRCH/HHFB                1    1                  I NRR/DRCH/H CB            1    1 NRR/DRCH/HOLB                1    1      NRR/DRI L/RPEB          1    1 NRR/DRSS/PRPB                2    2      NRR DSSA/SPLB            1    1 NRR/DSSA/SRXB                1    1    GREG            02      1    1 RES/DSIR/EIB                1    1              FILE    01      1    1 EXTERNAL: EG&G BRYCE,J.H                2    2      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 NOTE TO ALL "RIDS" RECIPIENTS:
PLEASE HELP US TO REDUCE iVASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT. 20079) TO ELIMINATEYOUR NAME FROiVi DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!
FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED'TTR                27  ENCL    27
 
Indiana Michig Power Company Cook Nuclear Plant One Cook Place Bridgrnan, Ml 49106 616 465 5901 INDlANA NICHlGi4N POWE'R May 10,      1994 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:
94-004-00 A. A.      Blind Plant Manager
/sb c'.
Attachment B. Martin, Region E. E. Fitzpatrick P. A. Barrett III 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 q n.~,<
i~ U l,s  I v
'7405190369 940510 PDR      ADOCK 05000315 S                        PDR
 
NRC FORM 366                                            U.S. NUCLEAR REGULATORY COMMISSION                      APPROVED BY OMB NO. 3150.0104 (5 92)                                                                                                                        EXP IR ES  5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMADON COLLECTION REQUEST: 50.0 HRS.            FORWARD LICENSEE EVENT REPORT (LER)                                                    COMMENTS REGARDUIG BURDEN ESDMATE TO THE INFQRMATICN AND RECORDS MANAGEMENTBRANCH (MNBB Tr ta), U.S. NUCLEAR REGULATORY COMMISSION> WASHINGTON, DC 20555 0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150010a), OFFICE OF (See reverse for required number of digits/characters for each block)                  MANAGEMENTAND BUDGET, WASHINGTON, DC 20503, FACILITYNAME (I)                                                                                        DOCKET NUMBER (2)                                PAGE  P)
D. C. COOK NUCLEAR PLANT  UNIT 1                                                                                  05000 315                          10F 3 FAILURE OF THREE PRESSURIZER SAFETY VALVES TO'MEET TECHNICAL SPECIFICATION RVEILLANCE TEST CRITERIA EVENT DATE 5                            LER NUMBER 6                      REPORT NUMBER 7                    OTHER FACILITIES INVOLVED 6 REVISION FACIUTY NAME                          DOCKET NUMBER SEQUENTIAL MONTH          DAY        YEAR    YEAR                                          MONTH      DAY NUMBER            NUMBER                                                                        05000 FACILITYNAME                          DOCKET NUMBcR 04          06          94      94                004              0          05        10    94                                            05000 OPERATING                        THIS REPORT IS SUBMITTED PURSUA NT TO THE REQUIREMENTS OF 10 CFR E: Check one or m ore 11 MODE (9)                          20A02(b)                                20.405(c)                          50.73(a) (2)(iv)                  73.71(b)
POWER                          20.405(a) (1) (i)                        50.36(c) (1)                        M73(a)(2)(v)                      73.71(c)
LEVEL (10)                0      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)        (Speoty in Atrctract tretow and in Text, NRC 20.405(a) (1) (iv)                      50.73(a) (2) (ii)                  50.73(a)(2)(viii)(B)        Form 366A) 2o.405(a)(1)(v)                          50.73(a)(2) (iii)                  50.73(a)(2)(x)
LICENSEE CONTACT FOR THIS LER 12 NAME                                                                                                              TELEPHONE NUMBER Fnctvde Area code)
G. A.        WEBER  PLANT ENGINEERING SUPERINTENDENT                                                        (616) 465-5902 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT 13 REPORTABLE                                                                            REPORTABLE CAUSE        SYSTEM        COMPONENT      MANUFACTURER                                          CAUSE  SYSTEM    COMPONENT        MANUFACTURER TO NPRDS            , D.                                                                TO NPRDS
                                                                              ;j.x RV              C710
                                                                              ', ~
SUPPLEMENTAL REPORT EXPECTED 14                                                          EXPFCTFD      MONTH      DAY    YEAR YES                                                                                                                    SUBMISSION NO 0( yee, cornptete ExPEGTED sUBMIssIQN DATE)                                                                              DATE (15)
ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single. spaced typewritten lines) (16)
On all Apri.l 6, 1994 with Uni.t 1 in Mode 5 (Cold Shutdown) three of the pressurizer safety valves, which were sent to an off site it  was determined that test laboratory for testing, were found with Technical Specification acceptance criteria. Acceptable settings are between, lift  settings outside of the 2461 psig and 2509 psig. Valve 1-SV-45A was found to have a 2536 psig, valve 1-SV-45B had a                            lift    setpoint of 2535 psig and 1-SV-45C had a lift      setpoint of lift      setpoint of 2538 psig.
There was no safety-signi.ficance since the worst case (1-SV-45C-of 2538 psig) would result in a maximum transient pressure of 2615 psig (2538 lift    setpoint psig plus 3 percent accumulation to attain its full rated lift). This is below the Technical Specification safety limit of 2735 psig.
All three valves were partially disassembled (retaining spring compression) and inspected.              No problems were noted.                      The nozzle and disc seating surfaces were lapped and polished.                        The valves were reassembled and tested satisfactori.ly. The safety valve conditions experienced- at the D.C. Cook Plant are similar to current i.ndustry trends/concerns.                                          Since a specific Root Cause could not be determined, no preventive action i.s planned at this time.
However, we will be evaluating the test methods and industry activities pertaining to the pressurizer safety valves.
NRC FORM 366 (5-92)
 
REQUIRED NUMBER OF DIGITS/CHARACTERS FOR EACH BLOCK BLOCK          NUMBER OF TITLE NUMBER    DIGITS/CHARACTERS UP TO 46                  FACILITY NAME 8 TOTAL DOCKET NUMBER 3 IN ADDITION TO 05000 VARIES                  PAGE NUMBER UP TO 76                  TITLE 6 TOTAL EVENT DATE 2 PER BLOCK 7 TOTAL 2 FOR YEAR LER NUMBER 3 FOR SEQUENTIAL NUMBER 2 FOR REVISION NUMBER 6 TOTAL REPORT DATE 2 PER BLOCK UP TO 18 *. FACILITY NAME 8 TOTAL DOCKET NUMBER OTHER FACILITIES INVOLVED 3 IN ADDITION TO 05000 OPERATING MODE 10                                        POWER LEVEL 1
REQUIREMENTS OF 10 CFR CHECK BOX THAT APPLIES UP TO 50 FOR NAME 12                                        LICENSEE CONTACT 14 FOR TELEPHONE CAUSE VARIES 2 FOR SYSTEM 13        4 FOR COMPONENT                  EACH COMPONENT FAILURE 4 FOR MANUFACTURER NPRDS VARIES 1
14                                        SUPPLEMENTAL REPORT EXPECTED CHECK BOX THAT APPLIES 6 TOTAL 15                                        EXPECTED SUBMISSION DATE 2 PER BLOCK
 
NRC FORM 366A                                                                U.S. NUCLEAR REGULATORY COMMISSION APPROVED OMB NO. 31500104 r (669)
EXPIRES: 4/30I92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVENT REPORT (LER)                                                INFORMATION COLLECTION REOUESTI 50.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION                                                  AND REPORTS MANAGEMENT BRANCH (F630), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, OC 20555, AND TO 1HE PAPERWORK REDUCTION PRO)ECT (31500104), OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503.
FACILITY NAME (1)                                                                  DOCKE'T NUMBER (2)                  LER NUMBER (6)                  PAGE (3)
YEAR      SEOUENTIAL      REVISION Hr5 NUMBER        NUMSFR D. C.      COOK NUCLEAR PLANT  UNIT                                    1        o  5    o  o    o  315 94              0 0    4 0        0  0  2 OF    0 3 TEXT (ifmore 4Pece Je eetrr9ed, Iree eddttr'one) NRC Forrrr 3IJSA'et (12)
Conditions Prior to Occurrence:
Unit One  Mode 5 (Cold Shutdown                                        following refueling).
Descri tion of Event:
On    April 6, 1994,                      it      was determined that all three safety pressurizer safety valves, Crosby Valve Model HB-86-PB, (EZIS/AB-RV) had                                                lift    settings outside Technical Specification 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.
2485 psig plus or minus 1 percent, (i.e. between 2461 and 2509 psig). Valve The valves are required to              lift    at 1-SV-45A was found to have a lift        setpoint of 2535 psig and 1-SV-45C had a lift    setpoint of 2536 psig, valve 1-SV-45B had a lift  setpoint of 2538 psig.
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 safety valve conditions experienced at the D.C. Cook Plant are similar to current industry trends/concerns.                                        The phenomenon of safety valve setpoint drift outside of design tolerances is common in the Nuclear industry.
However, as yet, no cause for the drift has been determined.
1-SV-45A, 1-SV-45B and 1-SV-45C were partially disassembled (retaining spring compression) and inspected.                                    No problems were noted.              The nozzle and disc seating surfaces were lapped and polished. The valves were reassembled and tested satisfactorily. The test program (test facility and procedure) for the Pressurizer Safety Valves has not changed from previous testing.
Anal sis of Event:
This event has been determined to be reportable under the provisions of 10CFR5073(a)(2)(i)(B) as an operation prohibited by Plant 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 1-SV-45A, 1-SV-45B and 1-SV-45C did not have any actual impact on the Reactor Coolant System (RCS) since the safety valves were not challenged during the last fuel cycle. There was no potential impact since the RCS would not have exceeded the maximum transient limit of 2735 psig, which is 110 percent of design pressure (2485 psig).
There was no impact on the health or safety of the public.
Safety Valve 1-SV-45C (worst case) had a pressure would have to reach a pressure of 2615 psig (2538 psig plus 3 percent lift    setpoint of 2538 psig. The RCS accumulation) for this valve to attain its full rated lift. Valve 1-SV-45A would have attained its rated 1-SV-45B would have attained its rated lift    at 2612 psig (2536 psig plus 3 percent) and lift    at 2612 psig (2535 plus 3 percent).
NRC Form 366A (669)
 
NRC FORM 366A                                                      U.S. NUCLEAR REGULATORY COMMISSIO (64)9)                                                                                                                  APPROVED OMB NO. 31504))04 EXPIRES) 4/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVENT REPORT (LER)                                            INFORMATION COLLECTION REGUESTI 500 HRS. FORWARD COMMENTS REGARDING BUADEN ESTIMATE TO THE RECORDS TEXT CONTINUATION                                                AND REPORTS MANAGEMENT BRANCH IP430), U.S. NUCLEAR AEGULATOAYCOMMISSION, WASHINGTON, OC 20555, AND TO THE PAPERWORK REDUCTION PAOJECT (31500104), OFFICE OF MANAGF MENT AND BUDGET, WASHINGTON, OC 20503.
FACILITY NAME (1)                                                        DOCKET NUMBER (21                      LFR NUMBER (6)                    PAGE (3)
YEAR        SSOVSNTIAL      REVISION
                                                                                                              )'.yi(  NVMSSR        NVM 54 D. C. COOK NUCLEAR PLANT  UNIT                                1      0  5  0  0    0  3  I    5 9 4          0    0 4          00 03        oF 0 TEXT /// mttt tptct B nqvnd. IItttddldoIM//Y/ICFdmt 3/)BA'4/ (I2)
Anal sis of Event continued!
The reactor vessel and pressurizer were designed to ASME BGPV Section which permits a maximum transient. pressure of 2735 psig, 110 percent of design III 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 (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 exceed 2615 psig (1-SV-45C setpoint of 2538 psig plus 3 percent).
In addition to this Safety Analysis, an additional Safety Evaluation is being performed to review the combined effect of recent Main Steam Safety Valve setpoints (reported in LER 50-315/94-001 and 94-003) in conjunction with the lift as-found          lift      setpoints of the Pressurizer Safety Valves. This evaluation is scheduled to be completed by June 30, 1994. We do not anticipate any adverse conditions to be identified during this evaluation, however, an updated LER will be submitted if deemed necessary.
Corrective Action:
The nozzle and disc seating surfaces were lapped and polished for all three safety valves. Retests were satisfactorily completed for steam set pressure and seat leakage.
The      spring pressure on 1-SV-45A was adjusted by 1/3                                flat to      bring the set pressure back into tolerance.
Since a specific Root Cause could not be determined, no preventive action is planned at this time. However, we will continue to follow industry activities pertaining to safety valve setpoint                              drift.
Failed Com onent Identification:
Pressurizer Safety Valve Plant Designation:                                    1-SV-45A, 1-SV-45B and 1-SV-45C Manufacturer: Crosby Valve Company Model: HB-86-BP EIIS Code: AB-RV Previous Similar Events:
LERS        50-315/90-16, 92-09 LER:        50-316/89-04, 92-06 NRC Form 366A (669)}}

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