ML17265A660

LER 99-006-00:on 990421,start of turbine-driven Auxiliary Feedwater Pump Was Noted.Caused by MOV Being Left in Open Position.Closed Manual Isolation Valve to Secure Steam to Pump.With 990521 Ltr

ML17265A660
Person / Time
Site:	Ginna
Issue date:	05/21/1999
From:	Mecredy R, St Martin J ROCHESTER GAS & ELECTRIC CORP.
To:	Vissing G NRC (Affiliation Not Assigned), NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-99-006, LER-99-6, NUDOCS 9905280059
Download: ML17265A660 (14)

Text

0 CATEGORY e REGULATORY XNPORMATXON DZSTR1BUTZON SYSTEM (RIDS)

ACCESSION NBR:9905280059 DOC.DATE: 99/05/21 NOTARIZED- NO DOCKET FACZL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G 05000244 AUTH.NAME ~ AUTHOR AFFILIATION ST MARTIN,J.T. Rochester Gas Ec Electric Corp.

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

RECZP.NAME RECZPXENT AFFILXATXON VISSZNG, G. S.

SUBJECT:

LER 99-006-00:on 990421,start of turbine-driven auxiliary feedwater pump was noted. Caused by'MOV being left in open position. Closed manual isolation valve to secure steam to pump. With 990521 ltr.

DISTRIBUTION CODE: ZE22T COPIES RECEIVED:LTR 1 ENCL I SIZE: E TITLE: 50.73/50.9 Licensee Event Report (LER), Incident Rpt, etc.

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

RECIPIENT COPIES RECIPIENT COPIES ZD CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL LPD1-1 PD, 1 1 'ZSSZNG,G. 1 1 INTERNAL: AEOD/SPD/RRAB 1 1 PILE CENTER+0 1 1 NRR/DZPM/XOLB 1 1 R'/DZPM/IOME 1 1 NRR/DRIP/REXB 1 1 NRR/DSSA/SPLB 1 1 RES/DET/EZB 1 1 RGN1 PILE 01 1 1 EXTERNAL: L ST LOBBY WARD 1 1 LMITCO MARSHALL 1 1 .D NOAC POORE,W. 1 1 NOAC QUEENER, DS 1 1 NRC PDR 1 1 NUDOCS FULL TXT 1 1 N

NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE HASTE. TO HAVE YOUR NAME OR ORGA'NIZATION OR REDUCE THE NUMBER OF COPIES RECEIVED BY YOU REMOVED FROM DISTRI BUTION LISTS OR YOUR ORGANIZATION, CONTACT THE DOCUMENTT CONTROL DESK (DCD) ON EXTENSION 415-2083 CON FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 16 ENCL 16

4ND ROCIIESTER GAS AND ELECTRIC CORPORATION ~ 89 EASTAVENLIE, ROCHESTER, N. Y Iddd9 LI00I AREA CODE7ld Sdd-2r 00 ROBERT C. MECREDY V<e President t4vcteor Ooerotions May 21, 1999 U.S. Nuclear Regulatory Commission Document Control Desk Attn: Guy S. Vissing Project Directorate I-1 Washington, D.C. 20555

Subject:

LER 1999-006, Valve in Unexpected Position Results in Start of Turbine-Driven Auxiliary Feedwater Pump R.E. Ginna Nuclear Power Plant Docket No. 50-244

Dear Mr. Vissing:

The attached Licensee Event Report LER 1999-006 is submitted in accordance with 10 CFR 50,73, Licensee Event Report System, item (a) (2) (iv), which requires a report of, "Any event or condition that resulted in a manual or automatic actuation of any engineered safety feature (ESF), including the reactor protection system (RPS) ".

Very truly yours, a..d obert C. Mecre xc: Mr. Guy S. Vissing (Mail Stop SC2)

Project Directorate I-1 Division of Reactor Projects -

Office of Nuclear Reactor Regulation I/II U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Regional Administrator, Region I U. S. Nuclear Regulatory Commission 898 475 Allendale Road King of Prussia, PA 19406 U.S. NRC Ginna Senior Resident Inspector 9'tt05280059 990521 PDR ADGCK 05000244 S PDR

C NRC ORM 366 U.S. NUCLEAR REGULATORY COMMISSION IB 1BBB)

ECLAT'aukefI r ra"9ke~AW'p~A'L'PJNP coBection request: 50 hrs. Reported lessons learned 'nformation are inceporated into the licensing process and fed back to LICENSEE EVENT REPORT (LER) hdusay. Forward comments regarding burden estimate to the Records Management Branch (T4 F33), U.S, Nudear Regulatory Cornmisskxt, Wash!ngton, DC 205550001, and to (See reverse for required number of the Pa pervrork ReducBon Pro)oct {31504104), 05ce of digits/characters for each block) Management and Budget, Washington, DC 20503. If an ktformaUon coBecUon does not display a currently val OMB control number. the NRC may not conduct or sponsor, and a FACILITY NAME I1l DOCKET NUMBER lzl PAGE I3I R. E. Ginna Nuclear Power Plant 05000244 1 OF 6 TITLE te)

Valve in Unexpected Position Results in Start of Turbine-Driven Auxiliary Feedwater Pump EVENT DATE (5) LER NUMBER {6) REPORT DATE {7) OTHER FACILITIES INVOLVED {8)

FACIUTY NAME OOCKETNVMBER SLOUTNTIAL RENSION MONTH OAY YEAR MONTH OAY YEAR NUMBER NUMBER 05000 04 21 1999 1999 - 006 - 00 05 21 1999 FACIUTY NAME DOCKET NVMBER 05000 OPERATING THIS REPORT IS SUBMITTED PU RSUANT TO THE REQUIREMENTS OF 10 CFR 5: (Check one or more) (11)

MODE {9) 20.2201 (b) 20.2203(a) (2) (v) 50.73(a) (2) (i) (6) 60.73(a)(2) (viii)

POWER 20.2203(9) (1) 20.2203(a)(3)(I) 50.73(a)(2)(ii) 50.73(a)(2)(x)

LEVEL (10) 002 20.2203(a) {2)(i) 20.2203{a)(3)(ii) 50.73(a) 12)(iii) 73.71 20.2203(a)(2) (ii) 20.2203{a) (4) X 50.73(a)(2)(iv) OTHER

20. 2203(a) (2) (iii) 50.36(c) (1) 50.73(a)(2)(v) Specify in Abstract below 20.2203{a) {2)(iv) 50.36(c)(2) 50.73(a)(2) {vii) or In NRC Form 366A LICENSEE CONTACT FOR THIS LER (12)

NAME TELEpHoNE NvMBER orroarde Area code)

John T. St. Martin - Technical Assistant (716) 771-3641 REPORTABLE REPORTABLE CAUSE SYSTEM COMPONENT MANUFACTURER 'fo EPIX CAUSE SYSTEM COMPONENT MANUFACTURER TO EPIX SUPPLEMENTAL REPORT EXPECTED (l4] EXPECTEO MONTH OAY YES NO SUBMISSION (If yes, complete EXPECTED SUBMISSION DATE). X OATE {f5}

ABSTRACT {Limitto 1400 spaces, i.e., approximately 15 single*spaced typewritten lines) (16)

On April 21, 1999, at approximately 1947 EDST, the plant was in Mode 2 at approximately 2% reactor power with the reactor coolant system being maintained at a temperature of 549 degrees F and a pressurizer pressure of 2235 PSIG.

Testing of a motor-operated valve which isolates steam to the turbine-driven auxiliary feedwater pump was being completed, and the valve was,left in the open position.

The manual isolation valve was then opened, admitting steam to the turbine-driven auxiliary feedwater pump. This started the turbine-driven auxiliary feedwater pump.

Immediate action was to isolate auxiliary feedwater flow from the turbine-driven auxiliary feedwater pump and close the manual isolation valve to secure steam to the pump.

The underlying cause of this event was the motor-operated valve was left in the open position following diagnostic testing. This event occurred at a time when the valve would normally have been expected to be closed.

Corrective action to prevent recurrence is outlined in Section V.B.

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (8 1998)

L(CENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1) DOCKET (2) LER NUMBER (6) PAGE (3)

WA 8(BIIBI)gt IIEYISIOII IIUMBN IIBMBBI R.E. Ginna Nuclear Power Plant 05000244 .-

1999 006 00 2 OF 6 TEXT llfmore spaceis required, use edditionel copies of NRC Form 366AI (17)

PRE-EVENT PLANT CONDITIONS II On April 21, 1999, at approximately 1947 EDST, the plant was in Mode 2, holding at approximately 2%

reactor power preparing for startup from the 1999 refueling outage. The reactor coolant system (RCS) was being maintained at a temperature of approximately 549 degrees F and a pressurizer (PRZR) pressure of .

approximately 2235 PSIG. The steam generator (SG) atmospheric relief valves (ARVs) were being operated to maintain constant RCS temperature. When the ARVs operate, significant steam flow noise is present in the area around the ARVs.

Testing of motor-operated valve (MOV) 3505A (MOV steam admission valve to the turbine-driven auxiliary feedwater (TDAFW) pump from the "A" SG) was being completed per Maintenance Procedure M-64.1.2 (MOVATS Testing of Motor-Operated Valves) and Test Procedure PT-50.7 (Differential Pressure Testing of TDAFW Pump Steam Supply Valves MOV-3504A and/or MOV-3505A). Procedure M-64.1.2 is a generic procedure, which allows for the installation and removal of test equipment, and the acquisition and analysis of test data. M-64.1.2 is performed by a team from the Reliability Test group, composed of a team leader and an electrician. Procedure PT-50.7 is a test procedure to obtain specific differential pressure test data of the listed MOVs, and is performed by Performance Monitoring and Reliability Test personnel in conjunction with Control Room operators. MOV-3505A is in the vicinity of the ARVs. Performance Monitoring personnel had completed procedure PT-50.7 and had verified that MOV-3505A was in the closed position, as specified in PT-50.7.

II. 'ESCRIPTION OF EVENT:

A. DATES AND APPROXIMATE TIMES OF MAJOR OCCURRENCES:

o April 21, 1999, 1947 EDST: Event date and time.

o April 21, 1999, 1947 EDST: Discovery date and time.

o April 21, 1999, 1948 EDST: TDAFW pump is secured.

EVENT:

.Prior to the start of M-64.1.2, during a discussion in the Control Room, it had been determined that confirmation stroking of MOV-3505A, required by a step in M-64.1.2, would be deferred and documented as complete when a similar step is performed in 'periodic test procedure PT-16Q-T, "Auxiliary Feedwater Turbine Pump - Quarterly" Personnel from the Reliability Test group,

~

Performance Monitoring, and Control Room operators were involved in this discussion.

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (B.I998)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1) DOCKET (2) LER NUMBER (6) PAGE (3) 8EOUEKEIA( BEMBIOK KUMBN KUMBN R.E. Ginna Nuclear Power Plant O5OOO244 -

1999 006 00 3 OF 6 TEXT ilfmore spaceis required, use edditionel copies of NRC Form 366AI (17)

Reliability Test personnel submitted an isolated work area request (" hold" request) to hold closed V-3505 (the manual isolation valve to MOV-3505A) during MOYATS testing. MOY diagnostic tests were then performed per procedure M-64.1.2 on MOV-3505A. As required, the DC power was removed from MOV-3505A during these tests. Removal of DC power removes position indication for MOY-3605A on the Main Control Board (MCB). MOY-3505A was stroked several times while performing the MOV diagnostic testing, and was left in the open position by Reliability Test personnel, as permitted by M-64.1.2. When the testing portion of M-64.1.2 was completed, the Reliability Test personnel began to remove the MOV diagnostic test equipment in preparation for restoring MOV-3505A to service.

During this restoration process, the Reliability Test personnel informed the Control Room operators to begin restoring MOV-3505A to service. Communications between Reliability Test personnel and Control Room operators did not adequately include a discussion of the current valve position for MOV-3505A, nor that DC power was still in the process of being restored for the MOV. Since DC power had not been restored for MOV-3505A, there were no MCB annunciator alarms or other MCB indications to alert the Control Room operators that MOV-3605A was open.

After requesting release of the "hold" from the Reliability Test team leader, the Control Room operators proceeded to remove the holds associated with MOV-3505A and to realign valves. A Control Room operator directed an Auxiliary Operator (AO) to open V-3505,'the manual isolation valve for MOV-3605A. The AO started to open this valve, which is in the vicinity of the ARVs. As this valve was being opened, steam was beginning to be supplied to the TDAFW pump, and the pump started. This steam flow would otherwise have been a direct indication of the improper condition, but due to the significant noise in the area, the AO did not detect steam flow as he opened V-3605. There were no alarm indications available in the Control Room to alert the Control Room operators to the start of the TDAFW pump.

The AFW System systems engineer was in the vicinity of the TDAFW pump during these activities (which is on next floor directly below the ARVs, MOV-3505A, and V-3606) anticipating performance of procedure PT-16Q-T, which would stroke MOY-3605A and operate the TDAFW pump. He observed the TDAFW pump starting and notified the Control Room of this condition.

Around the same time, the Reliability Test electrician contacted the Control Room operators by telephone, requesting that DC power be restored to MOV-3505A. This action would auto-close MOV-3605A, which was believed (by the Control Room operators) to have already been done prior to the release of the hold for V-3606. The Control Room operators, having been notified by the AFW systems engineer, observed feedwater flow from the TDAFW pump on the MCB, and then realized there had been an unexpected start of the TDAFW. The Control Room operators promptly secured feedwater flow from the TDAFW pump by manually closing air-operated discharge flow control valves at the Main Control Board. The Control Room operator then notified the AO to locally re-close V-3605 to secure steam to the TDAFW pump.

A review of data indicated that during the time the TDAFW pump was operating, there was a slight cooldown of the RCS of approximately two (2) degrees F and a TDAFW pump run of about one (1) minute.

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (B.IBBB)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME l1) DOCKET I2) LER NUMBER l6) PAGE l3)

SB)UBITIAL REVISION IIUMBFR IIUMBBI R.E. Ginna Nuclear Power Plant 05000244 1999 - - 00 006 4 OF 6 TEXT /If more space is required, use addi Iional copies of NRC Form 366A/ l17)

C. 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 identified by the AFW systems engineer, who observed the TDAFW pump starting and notified the Control Room of this condition.

F. OPERATOR ACTION:

After confirming the start of the TDAFW pump, the Control Room operators promptly secured feedwater flow from the TDAFW pump and directed the AO to secure steam to the TDAFW pump.

The Control Room operators subsequently notified higher supervision and notified the NRC per 10CFR50.72 (b} (2} (ii}, non-emergency four hour notification, at approximately 2230 EDST on April 21, 1999.

G. SAFETY SYSTEM RESPONSES:

None III. CAUSE OF EVENT A. IMMEDIATECAUSE:

The immediate cause of the start of the TDAFW pump was opening V-3505 with MOV-3505A in an unexpected position (open}.

8. INTERMEDIATE CAUSE:

The intermediate cause for MOV-3505A being in an unexpected position was the deferral of steps in M-64.1.2 requiring confirmation stroking and not verifying the position of MOV-3505A prior to removing the hold tag for V-3505.

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

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITV NAME (1) DOCKET (2) LER NUMSER (6) PAGE (3)

S(BII(tnIAL II(VISIBM It(1MB@ ttIIMBER R.E. Ginna Nuclear Power Plant 05000244 tees - ooe - oo 5 OF 6 TEXT llfmore space is required, use additional copies of NRC Form 366AJ (17)

C. ROOT CAUSE:

The underlying cause for MOV-3505A being in an unexpected position was less than adequate guidance in procedure M-64.1.2 to ensure the proper sequence of completion activities. Steps were allowed to be deferred, which no longer assured that restoration of DC power, confirmation stroking, and final valve position are completed prior to removal of the hold on the isolated work area.

These errors were cognitive errors. Performance Monitoring personnel left MOV-3605A closed at the completion of PT-50.7, but the subsequent re-positioning of MOV-3505A by Reliability Test personnel per M-64.1.2 was not a'dequately communicated to the other groups. Control Room operators removed the hold on V-3505, but the Reliability Test team leader and Control Room operators made a cognitive error in assuming that MOV-3605A was closed at this time. These errors were associated with activities that were covered by approved procedures. However, there was not sufficient detail included within these procedures with respect to requirements for valve, cycling. Significant steam flow noise in the area of the ARVs where the AO opened the manual isolation valve contributed to the error, in that the AO could not detect steam flow to the TDAFW pump as he opened this valve. Personnel involved in this event included licensed operators, a non-licensed operator, and other licensee personnel from the Performance Monitoring and Reliability Test groups ~

IV. ANALYSIS OF EVENT:

This event is reportable in accordance with 10 CFR 50.73, Licensee Event Report System, item (a) (2) (iv),

which requires a report of, "Any event or condition that resulted in a manual or automatic actuation of any engineered safety feature (ESF), including the reactor protection system (RPS)". The start of an AFW pump is an actuation of an ESF.

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 or implications attributed to the unexpected start of the TDAFW pump because:

The start of the TDAFW pump occurred with acceptable levels in both SGs.

The additional feedwater flow from the TDAFW pump was promptly secured and V-3606 was closed to secure steam to the TDAFW pump to minimize the cooldown of the reactor coolant system (RCS).

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION IBIBBB)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME 11) DOCKET 12) LER NUMBER IB) PAGE {3)

SEBVENIML RDBBIBN NUMBER NUMBER R.E. Ginna Nuclear Power Plant 05000244 1999 QQ6 QQ 6 OF 6 TEXT llfmore space is required, use additional copies ol iVRC Form 366Al I17) o The limiting case for this event is a sudden increase in feedwater flow at hot zero power. The case analyzed in the Ginna Station Updated Final Safety Analysis Report (UFSAR), Section 15.1.2 involves a step increase in feedwater flow to both steam generators to 110% of the nominal full power flow rate initiated at hot zero power with the reactor control in manual rod control. The plant condition at the time of this event was Mode 2. The specific criterion used to evaluate a feedwater malfunction event initiated at hot zero power is that the maximum reactivity insertion rate that results from the cooldown should be less than the reactivity insertion rate that has been analyzed in the rod withdrawal from a subcritical condition analysis. This event resulted in a cooldown of approximately 2 degrees F, which resulted in a negligible reactivity change.

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:

Feedwater flow from the TDAFW pump was secured and steam to the TOAFW pump was secured to minimize the cooldown of the RCS.

ACTION TAKEN OR PLANNED TO PREVENT RECURRENCE:

MOV diagnostic testing procedures, including M-64.1.2, will be revised to identify a final required valve position following diagnostic testing.

The lessons learned from this event will be incorporated into the planning and coordination of MOV. diagnostic testing.

VI. ADDITIONALINFORMATION:

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 could be identified. However, LERs96-004, 96-008, and 96-010 and 96-011 are similar events (start of an auxiliary feedwater pump) with different root causes.

C. SPECIAL COMMENTS:

None