ML17265A614

LER 99-003-00:on 990301,two Main Steam non-return Check Valves Were Declared Inoperable Due to Exceedance of Acceptance Criteria.Caused by Changes in Methodology & Matls.Packing Gland Torque Will Be Adjusted.With 990331 Ltr

ML17265A614
Person / Time
Site:	Ginna
Issue date:	03/31/1999
From:	MECREDY R C, ST MARTIN J T ROCHESTER GAS & ELECTRIC CORP.
To:	VISSING G S NRC (Affiliation Not Assigned), NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-99-003, LER-99-3, NUDOCS 9904080027
Download: ML17265A614 (14)

Text

'ATEGORY REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)ACCESSION NBR:9904080027 DOC.DATE: 99/03/31 NOTARIZED:

NO(SCIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G ,'AUTH.NAME.AUTHOR AFFILIATION.ST.MARTIN,J.T.

Rochester Gas&, Electric Corp.MECREDY,R.C.

Rochester Gas&Electric Corp.RECIP.NAME RECIPIENT AFFILIATION DOCKET 05000244 VISSING,G.S.

C A Q 05000244 0 NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72).

SUBJECT:

LER 99-003-00:on 990301,two main steam non-return check valves were declared inoperable due to exceedance of acceptance criteria.Caused by changes in m'ethodology

&matls.Packing gland torque will be adjusted.With 990331 ltr.DISTRIBUTION CODE: IE22T COPIES RECEIVED: LTR I ENCL (SIZE: TITLE: 50.73/50.9 Licensee Event Report (LER), Incident Rpt, etc.RECIPIENT ID CODE/NAME PD1-1 PD INTERNAL: AEOD RAB~ENZ NRR/DRCH/HQMB NRR/DSSA/SPLB RGN1 FILE 01 EXTERNAL: L ST LOBBY WARD NOAC POORE,W.NRC PDR COPIES LTTR ENCL 1 1 2 2 1 1 1 1 1 1 1 1 1 1 1 1 1 1 RECIPIENT ID CODE/NAME VISSING,G.

AEOD/SPD/RRAB NRR/DRCH/HOHB NRR/DRPM/PECB RES/DET/EIB LMITCO MARSHALL NOAC QUEENER,DS NUDOCS FULL TXT COPIES LTTR ENCL ,1 1 1 1 1 1 1'1 1 1 1 1 1 1 1 D N NOTE TO ALL"RZDS" RECIPZENTS:

PLEASE HELP US TO REDUCE HASTE.TO HAVE YOUR NAME OR ORGANIZATION REMOVED FROM DISTRIBUTION LISTS OR REDUCE THE NUMBER OF COPIES RECEIVED BY YOU OR YOUR ORGANIZATION, CONTACT THE DOCUMENT CONTROL DESK (DCD)ON EXTENSION 415-2083 FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 18 ENCL 18 I I I AND gM~giESTH CAS li'gD".'E~IC RFC&Crv~89&5 4~=.S"E'C."$

5'.7 V'Sa-'-.sg', i+CE.';6 Ac-27K March 31, 1999 U.S.Nuclear Regulatory Commission Document Control Desk Attn: Guy S.Vissing Project Directo'rate I-1 Washington, D.C.20555

Subject:

LER 1999-003, Two Valves Declared Inoperable Results in Condition Prohibited by Technical Specifications R.E.Ginna Nuclear Power Plant Docket No.50-244

Dear Mr.Vissing:

The attached Licensee Event Report LER 1999-003 is submitted in accordance with 10 CFR 50.73, Licensee Event Report System, item (a)(2)(i)(B),"Any operation or condition prohibited by the plant's Technical Specifications".

Ver truly yours, Robert C.Mecred xc: Mr.Guy S.Vissing (Mail Stop SC2)Project Directorate I-1 Division of Reactor Projects-I/II Office of Nuclear Reactor Regulation U.S.Nuclear Regulatory Commission Washington, D.C.20555 Regional Administrator, Region I U.S.Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 U.S.NRC Ginna Senior Resident Inspector 9904080027 99033i PDR ADOCK 05000244 8 PDR 0 I NRC FORM 366 IB ISSSI U.S.NUCLEAR REGULATORY COMMISSION FACILITY NAME I11 R.E.Ginna Nuclear Power Plant LICENSEE EVENT REPORT (LER)(See reverse for required number of digits/characters for each block)PAGE (3)1 OF 6 DocKET NUMBER I2I 05000244 fN5lt'a Jeep r rVp Be'I'o'N'gAL(s%88P' information coaction request: 50 hrs.Reported lessons learned are incorporated into the licensing process and fed back to industry.Forward comments regarding burden estimate to the Records Management Branch (TA F33).U.S.Nudear Regulatory Commission, Washington, DC 205554001.

and to the Paperwork Reduction project (31504I04), Office of Management and Budget, Washington, DC 20503.If an.information coBection does not display a cunently valid OMB control number, the NRC may not conduct or sponsor.and a TITLE (el Two Valves Declared Inoperable Results in Condition Prohibited by Technical Specifications EVENT DATE (5)LER NUMBER (6)REPORT DATE (7)OTHER FACILITIES INVOLVED (8)MONTH OAY YEAR SEOUENTIAL REYISION NUMBER NUMBER MONTH OAY FACIUTY NAME DOCKET NUMBER 05000 03 01 OPERATING MODE (9)POWER LEVEL (10)1999 0 1999-003-00 03 31 1999 FACIUTY NAME 20.2201(b) 20.2203(a)(1) 20.2203(a)(2)(i)20.2203(a)(2)(ii) 20.2203(a)(2)(iii)20.2203(a)

(2)(iv)20.2203(a)

(2)(v)20.2203(a)(3)(i) 20.2203(a)

(3)(ii)20.2203(a)(4) 50.36(c)(1) 50.36(c)(2)

X 50.73(a)(2)(i)(B) 50 73(a)(2)(u) 50.73(a)(2)(iii)50.73(a)(2)(iv) 50.73(a)(2)(v)50.73(a)(2)(vii)

THIS REPORT IS SUBMITTED P URSUANT TO THE REQUIREMENTS OF 10 CFR E: (Check DOCKET NUMBER 05000 one or morel (11)50.73(a)(2)(viii)50.73(a)(2)(x)73.71 OTHER Specify ln Abstract below or in NRC Form 366A NAME John T.St.Martin-Technical Assistant LICENSEE CONTACT FOR THIS LER (12)TELEPHONE NUMBER (rneivde Aree Cadet (716)771-3641 CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABIE TO EPIX CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABlE TO EPIX SUPPLEMEIITAL REPORT EXPECTED (14)YES (If yes, complete EXPECTED SUBMISSION DATE).X No EXPECTED SUBMISSION DATE (15)MONTH OAT ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)(16)On March 1, 1999, at approximately 1707 EST, it was determined that the required torque to initiate valve disc closure for the two main steam non-return check valves was greater than the acceptance criteria specified in plant test procedures.

Immediate corrective action was to declare both valves inoperable and enter Technical Specification Limiting Condition'or Operation 3.0.3.Following an evaluation of the test data by Nuclear Engineering Services, it was determined that the valves were oper'able.

The plant exited Limiting Condition for Operation 3.0.3.The underlying cause of the event was changes in the methodology and materials for packing these valves, which resulted in a greater than anticipated shaft breakaway torque.Corrective action to prevent recurrence is outlined in Section V.B.

1 NRC FORM 366A IB 1999)LICENSEE EVENT REPORT (LER)TEXT CONTINUATION U.S.NUCLEAR REGULATORY COMMISSION

'ACILITY NAME I1)R.E.Ginna Nuclear Power Plant DOCKET l2)05000244 LER NUMBER I6I SMUENTIAL RBBBIBN NUMBER NUMBER 1999-003-00 PAGE (3I 2 OF 6 TEXT (If more space is required, use additional copies of NRC Form 366A/I17)PRE-EVENT PLANT CONDITIONS:

Since 1992, Performance Monitoring technicians have performed surveillance test procedure PT-2.10.15,"Main Steam Non-Return Check Valve Closure Verification", using the test methodology established by Nuclear Engineering Services (NES).Performance of test procedure PT-2.10.15 satisfies Ginna Station Improved Technical Specifications (ITS)Surveillance Requirement (SR)3.7.2.2 and satisfies the requirements of Section XI of the ASME Code for these valves.The required torque to initiate valve disc closure (breakaway torque)for the main steam non-return check valves (CV-3518 and CV-3519)has consistently been measured significantly lower than the acceptance criteria specified within the test procedure (600 ft-lbs).On March 1, 1999, the plant was in Mode 3, cooling down to Mode 4 for a scheduled refueling outage.Both main steam isolation valves (MSIVs)were closed.At approximately 1707 EST, Performance Monitoring technicians were performing procedure PT-2.10.15.

The technicians were utilizing a calibrated torque wrench with a range of 0 to 600 ft-lbs, as they had in previous years.The technicians could not initiate valve disc closure (achieve breakaway torque), even at the full range of the torque wrench.They consulted'with supervision, and initiated a plant ACTION'Report to document the inability to achieve check valve disc movement up to 600 ft-Ibs of torque.DESCRIPTIOA OF EVENT: A.DATES AND APPROXIMATE TIMES OF MAJOR OCCURRENCES:

March 1, 1999, 1707 EST: Event Date and Time and Discovery Date and Time.March 1, 1999,.1 734 EST: Both main steam non-return check valves are declared inoperable.

0 March 1, 1999, 1930 EST: Engineering Technical Evaluation determines that both main steam non-return check valves are operable.March 1, 1999, 2018 EST: The Plant enters Mode 4, where ITS LCO 3.7.2 is not applicable.

ITS LCO 3.0.3 for the main steam non-return check valves is exited.B.EVENT: On March 1, 1999, the plant was in Mode 3, cooling down to Mode 4 for a scheduled refueling outage.Both main steam isolation valves (MSIVs)were closed, as specified by the Initial Conditions for test procedure PT-2.10.15.

The Performance Monitoring technicians notified the Shift Supervisor of the failure of the main steam non-return check valves to meet the closure torque acceptance criteria of test procedure PT-2.10.15.

NRC FORM 366A (6 IBBS)U.S.NUCLEAR REGULATORY COMMISSION

'ACILITY NAME (1)R.E.Ginna Nuclear Power Plant DOCKET (2)05000244 LER NUMBER (6)TEAR SEOUENTIAL RENSION NUMBER NUMBER 1999-OO3 LICENSEE EVENT REPORT (LER)TEXT CONTINUATION PAGE (3)3 OF 6 TEXT llf more spaceis required, use additional copies of NRC Form 366Al (17)The Shift Supervisor reviewed ITS Limiting Condition for Operation (LCO)3.7.2, notified the NES staff of the'event, and requested an engineering technical evaluation.

At approximately 1734 EST the Shift Supervisor declared both valves CV-3518 and CV-3519 inoperable based on exceeding the acceptance criteria of test procedure PT-2.10.15.

As specified in ITS LCO Required Action 3.7.2.E.1, with"one or more valves inoperable in flowpath from each steam generator (SG)", immediate entry into ITS LCO 3.0.3 is required.The Shift Supervisor directed entry into ITS LCO 3.0.3 at this time.Performance Monitoring technicians obtained a torque wrench of larger range and again attempted to achieve breakaway torque.At approximately 700 ft-Ibs torque, the valve d(sc for CV-3518 started to close, and at approximately 900 ft-Ibs torque the valve disc for CV-3519 started to close.These as-found breakaway torque values were provided to NES staff.NES staff performed an engineering technical evaluation of this event.At approximately 1930 EST, NES staffhadreviewed an engineering analysis (Design'Analysis DA-ME-92-147) that had been performed previously for these valves, and determined that the as-found breakaway torque was within the bounds of the analysis.This information was provided to the Shift Supervisor.

While the valves were now capable of being declared operable, the plant continued the planned cooldown and entered Mode 4 at approximately 2018 EST on March 1, 1999.In Mode 4, ITS LCO 3.7.2 is not applicable, and ITS LCO 3.0.3 was formally exited at this time.The entry into ITS LCO 3.0.3 as a result of declaring both CV-3518 and CV-3519 inoperable is considered to be a condition prohibited by Technical Specifications.

Entry into ITS LCO 3.0.3 for any reason or justification is considered reportable per the NRC guidance in NUREG-1022 Revision 1.INOPERABLE STRUCTURES, COMPONENTS, OR SYSTEMS THAT CONTRIBUTED TO THE EVENT: None D.,OTHER SYSTEMS OR SECONDARY FUNCTIONS AFFECTED: None E.METHOD OF DISCOVERY:

This event was discovered by Performance Monitoring technicians who were performing a routine surveillance test during the plant cooldown.

.~

NRC FORM 366A (B IBBB)LICENSEE EVENT REPORT (LERj TEXT CONTINUATION U.S.NUCLEAR REGULATORY COMMISSION FACILITY NAME (1)DOCKET (2)LER NUMBER (6)PAGE (3)R.E.Ginna Nuclear Power Plant 05000244 VEAR SEOUENTIAE REVISION NUMBER NUMBER 1999-003-00 4 OF 6 TEXT ilf more space is required, use additional copies of NRC Form 366Al (17)OPERATOR ACTION: The Shift Supervisor reviewed ITS LCO 3.7.2 and declared both valves CV-3518 and CV-3519 inoperable based on exceeding the acceptance criteria of test procedure PT-2.10.15.

The Shift Supervisor directed entry into ITS LCO 3.0.3 at this time.The Shift Supervisor notified NES staff of the event, and requested an engineering technical evaluation.

The operators continued the process of performing a plant cooldown per operating procedure 0-2.2,"Plant Shutdown from Hot Shutdown to Cold Conditions".

After the plant was in Mode 4, ITS LCO 3.7.2 was not applicable and LCO 3.0.3 was exited for the main steam non-return check valves.G.SAFETY SYSTEM RESPONSES:

None III.CAUSE OF EVENT: A.IMMEDIATE CAUSE: The immediate cause of the condition prohibited by Technical Specifications was entering ITS LCO Required Action 3.7.2.E.1 for two valves inoperable, which required immediate entry into ITS LCO.3.0.3.B.INTERMEDIATE CAUSE: The intermediate cause of entry into ITS LCO 3.7.2.E.1 was the decision to declare both main steam non-return check valves inoperable for exceeding the acceptance criteria of Steps 6.1.3 and 6.2.3 of test procedure PT-2.10.15.

ROOT CAUSE: The underlying cause for exceeding the acceptance criteria was changes in the methodology and materials for packing these valves, instituted during the previous outage.These changes in methodology and vendor-recommended replacement shaft bushing materials were made in order to provide improved shaft sealability and vibration mitigation, and resulted in a greater than anticipated shaft breakaway torque.Over time, during the previous plant operating cycle, heat and moisture were absorbed by the packing, which caused the shaft friction to increase to the as-found values of 700 and 900 ft-lbs, which were higher than anticipated, based on testing results from previous years.

NRC FORM 366A (B.IBBS)LICENSEE EVENT REPORT (LER)TEXT CONTINUATION U.S.NUCLEAR REGULATORY COMMISSION FACILITY NAME I1)R.E.Ginna Nuclear Power Plant DOCKET I2)05000244 LER NUMBER I6)YEAR SEQUENTIAL REVISION NUMBER NUMBER I 999-003-00 PAGE I3)5 OF 6 TEXT (If more spaceis required, use additional copies of ftlRC Form 366A/I17)IV.ANALYSIS OF EVENT: This event is reportable in accordance with 10 CFR 50.73, Licensee Event Report System, item (a)(2)(i)(8),"Any operation or condition prohibited by the plant's Technical Specifications".

Declaring both main steam non-return check valves inoperable resulted in entry into ITS LCO 3.0.3.Since the plant entered ITS LCO 3.0.3, this condition is reportable.

An assessment was performed considering both the safety consequences and implications of this event with the following results and conclusions:

There were no operational or safety consequences attributed to not meeting the acceptance criteria specified in procedure PT-2.10.15 because: o The acceptance criteria in test procedure PT-2.10.15 was conservatively chosen in 1992 to be well below the value calculated in Design Analysis DA-ME-92-147.

This conservative value had been utilized as the acceptance criteria in test procedure PT-2.10.15, prior to defining the operability requirements in ITS SR 3.7.2.2.The engineering technical evaluation performed on March 1, 1999, determined that the as-found breakaway torque values for the non-return check valves were within this previous analysis.o The two MSIVs isolate steam flow from the secondary side of the steam generators (SGs)following a Design Basis Accident (DBA).Both MSIVs were closed, as specified in the Initial Conditions of test procedure PT-2.10.15, prior to initiation of the surveillance test on March 1, 1999.The MSIVs are designed to work with the main steam non-return check valves, located immediately downstream of each MSIV, to preclude the blowdown of more than one SG following a steam line break (SLB).Based on the above, it can be concluded that the public's health and safety was assured at all times.V.CORRECTIVE ACTION: A.ACTION TAKEN TO RETURN AFFECTED SYSTEMS TO PRE-EVENT NORMAL STATUS: Immediate corrective action was to declare both valves inoperable and enter ITS LCO 3.0.3.Following an evaluation of the test data by NES, it was determined that the valves were operable.The plant exited ITS LCO 3.0.3.The plant is still in the 1999 refueling outage.

NRC FORM 366A (61998)LICENSEE EVENT REPORT (LER)TEXT CONTINUATION U.S.NUCLEAR REGULATORY COMMISSION FACILITY NAME (1)R.E.Ginna Nuclear Power Plant DOCKET (2)05000244 LER NUMBER (6)9(RU(N(IAL RENSIOII IIUMBIR 1999-003-00 PAGE (3)6 OF 6 TEXT llf more space is required, use additional copies of NRC Form 366A/(17)B.ACTION TAKEN OR PLANNED TO PREVENT RECURRENCE o Packing gland torque for these check valves will be adjusted to a value specified by the IST Engineer.An as-found baseline breakaway torque value will be obtained for each valve during the 1999 outage.0 The design analysis will be revised to provide acceptance criteria, both for the ASME Code degradation value and for determination of valve operability.

I A"reference value" will be established in accordance with ASME/ANSI OM-1987 Part 10 for breakaway torque for these valves.This value will be included in a future revision to test procedure PT-2.10.1 5.VI.ADDITIONAL INFORMATION:

A.FAILED"COMPONENTS:

None B.PREVIOUS LERs ON SIMILAR EVENTS: A similar LER event historical search was conducted with the following results: No documentation of similar LER events with the same root cause at Ginna Nuclear Power Plant could be identified.

C.SPECIAL COMMENTS: None fi