ML17263A667

LER 94-007-00:on 940427,feedwater Transient Occurred Due to Loss of Ability to Control Feedwater Regulating Valve, Causing Lo Lo SG Level Reactor Trip.Caused by Improperly Secured Stroke Adjust Set screw.W/940527 Ltr

ML17263A667
Person / Time
Site:	Ginna
Issue date:	05/27/1994
From:	Mecredy R, St Martin J ROCHESTER GAS & ELECTRIC CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-94-007, LER-94-7, NUDOCS 9406030044
Download: ML17263A667 (11)

Text

ACCELERATED DISTRIBUTION DEMONSTRATION SYSTEM REGULATORY INFORMATION DISTRIBUTION- SYSTEM (RIDS)

ACCESSION NBR:9406030044 DOC.DATE: 94/05/27 NOTARIZED: NO DOCKET g FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G 05000244 AUTH. NAME AUTHOR AFFILIATION ST.MARTIN,J.T. Rochester Gas & Electric Corp.

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

RECIP.NAME RECIPIENT AFFILIATION .R I

SUBJECT:

LER 94-007-00:on 940427,feedwater transient occurred due to loss of ability to control feedwater regulatinq valve, causing lo lo SG level reactor trip. Caused by improperly D secured stroke adjust set screw.W/940527 ltr.

DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR TITLE: 50.73/50.9 Licensee Event Report (LER), j ENCL Incident NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72).

/ SIZE:

Rpt, etc.

05000244 A 8

RECIPIENT COPIES RECIPIENT D COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL D PD1-3 PD 1 1 JOHNSON,A 1 1 INTERNAL: AEOD/DOA 1 1 AEOD/DSP/TPAB 1 1 AEOD/ROAB/DSP 2 2 NRR/DE/EELB 1 1 NRR/DE/EMEB 1 1 NRR/DORS/OEAB 1 1 NRR/DRCH/HHFB 1 1 NRR/DRCH/HICB 1 1 NRR/DRCH/HOLB 1 1 NRR/DRIL/RPEB 1 1 NRR/DRSS/PRPB 2 2 NRR/~SSA/S PLB 1 1 NRR/DSSA/SRXB 1 1 REGFFL~ 02 1 1 RES/DSIR/EIB 1 1 G 1 FILE 01 1 . 1 I

EXTERNAL: EG&G BRYCE,J.H NRC PDR 2

1 2

1 L ST LOBBY WARD NSIC MURPHY,G.A 1,1 1

'

NSIC POOREiW ~ 1 1 NUDOCS FULL TXT 1 1 D

A D

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 27 ENCL 27

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~ ~ ~ v r A sr 4'r roe rr Scare ROCHESTER GAS AND ELECTRIC CORPORATION 89 EAST AVENUE, ROCHESTER N. Y. 14649-0001 ROBERT C. MECREOY TELEPHONE Vice Preridr nr AicEA coDE 716 546'2700 Cinna Crurlesr Producuon May 27, 1994 U.S. Nuclear Regulatory Commission Attn: Allen R. Johnson PWR Project Directorate I-3 Document Control Desk Washington, DC 20555

Subject:

LER 94-007, Feedwater Transient, Due to Loss of Ability to Control Feedwater Regulating Valve, Causes a Lo Lo Steam Generator Level Reactor Tl lp R.E. Ginna Nuclear Power Plant Docket No. 50-244 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 attached Licensee Event Report LER 94-007 is hereby submitted.

This event has in no way affected the public's health and safety.

Very truly yours, g.P~g'~/r X.C~

Robert C. Mecredy xc: U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 Ginna USNRC Senior Resident Inspector 9406030044 'ss40527 PDR ADOCK 05000244 C~ PDR

NRC FORH 366 U.S. NUCLEAR REGULATORY C(NHI SSIOH PROVED BY MS NO. 3150-0104 (5-92) EXPIRES 5/31/95 ESTIHATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.

LICENSEE EVENT REPORT (LER) FORWARD COMMENTS REGARDING BURDEN ESTIMA'IE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (HNBB 7714), U.S. NUCLEAR REGULATORY COMMISSION, (See reverse for required nunber of digits/characters for each block) WASHINGTON, DC 20555-000'I AND TO THE PAPERWORK REDUCTION PROJECT (3140 0104), OFFICE OF MANAGEHENT AND BUDGET WASHINGTON DC 20503.

FAGILITY NAME (1) R. E ~ Ginna Nuclear Power Plant DOCKET HNNIER (2) PAGE (3) 05000244 10F9 TITLE (4) Feedwater Transient, Due to Loss of Ability to Control Feedwater Regulating Valve, Causes a Lo Lo Steam Generator Level Reactor Trip EVENT DATE 5 LER NINBER 6 REPORT DATE 7 OTHER FACILITIES INVOLVED 8 SEQUENTIAL REVISION FACILITY NAME DOCKET NUMBER HONTH DAY YEAR YEAR MONTH DAY YEAR NUMBER NUMBER 04 27 94 94 --007-- 00 05 27 FACILITY HAHE DOCKET NUMBER OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR  : (Check one or more 11 MODE (9) N 20.402(b) 20.405(c) X 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)

~

045 50.73(a)(2)(vii)

LEVEL (10) 20.405(a)(1)(ii) 50.36(c)(2) OTHER 20.405(a)(1)(iii) 50.73(a)(2)(i) 50.73<a)(2)(vIII)(A) (Specify in 20.405(a)(1)(iv) 50.73(a)(2)(ii) 50.73<a)(2)(viii)(B) Abstractand in Text, belo~

20.405(a)(1)(v) 50.73(a)(2)(iii) 50.73(a)(2)(x) NRC Form 366A LICENSEE CONTACT FOR THIS LER 12 NAME John T. St. Martin - Director, Operating Experience TELEPHONE NUMBER (Include Area Code)

(315) 524.4446 C(NPLETE ONE LINE FOR EACH COHPOHENT FAILURE DESCRIBED IN THIS REPORT 13 REPORTABLE REPORTABLE CAUSE SYSTEH COMPONENT MANUFACTURER CAUSE SYSTEM COMPONENT MANUFACTURER TO NPRDS TO NPRDS B JB LCV B042 Pj's)kj~ ';.

SUPPLEMENTAL REPORT EXPECTED 14 EXPECTED MONTH DAY YEAR YES SUBHI SSI ON (If yes, coaplete EXPECTED SUBMISSION DATE). X NO DATE (15)

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

On April 27, 1994, at approximately 1407 EDST, with the reactor at approximately 454 reactor power, the ability to control the "A" main feedwater regulating valve was lost. At 1410 EDST, the reactor tripped on Lo Lo level ((/= 17%) in the "A" Steam Generator. The Control Room operators performed the actions of procedures E-0 and ES-0.1.

The underlying cause was determined to be an improperly secured stroke adjust set screw for the valve positioner signal diaphragm assembly for the "A" main feedwater regulating valve. This event is NUREG-1022 Cause Code (B), "Design, Manufacturing, Construction/Installation."

Immediate corrective action was to install a new valve positioner of a previous design. Corrective action to preclude repetition is outlined in Section V (B).

NRC FORM 366 (5-92)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSIOH PROVED BY QHI NO. 3150-0104 (5.92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY MITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.

FORNARD COMMENTS REGARDIHG BURDEN ESTIMATE TO LICENSEE EVENT REPORT (LER) THE INFORMATION AHD RECORDS MANAGEHENT BRANCH TEXT CONTINUATION (MHBB 7714), U.S. NUCLEAR REGULATORY COMMISSION, NASHINGTON, DC 20555-0001 AND TO THE PAPERHORK REDUCTION PROJECT (3140.0104), OFFICE OF MANAGEMENT AHD BUDGET llASHINGTON DC 20503.

FACILITY NAME 1 DOCKET NMBER 2 LER NUMBER 6 PAGE 3 SEQUENTIAL REVISION YEAR R.E. Ginna Nuclear Power Plant 05000244 94 -- 007-- 00 2 OF 9 TEXT (If more space is required, use additional copies of HRC Form 366A) (17)

PRE-EVENT PLANT CONDITIONS The plant achieved full power operation on April 23, 1994, after completion of the 1994 annual refueling and maintenance outage. On April 26, 1994, the plant was manually shut down to repair a small steam leak on an instrument fitting on the high pressure (HP) turbine. Due to stability problems with control, of feedwater flow to the "A" Steam Generator (S/G) at steady state conditions, the valve positioner for the "A" main feedwater regulating valve (MFRV) was replaced during this brief shutdown.

On April 27, 1994, a load increase was in progress, controlled by Plant Operating Procedure 0-1.2, "Plant -Startup from Hot Shutdown to Full Load". The plant was at approximately 45% reactor power.

Preparations were being made to start a second main feedwater pump, per System Operating Procedure T-4.F, "Restoring 1B Feedwater Pump to Service After Maintenance or Power Reduction".

II. DESCRIPTION OF EVENT A. DATES AND APPROXIMATE TIMES OF MAJOR OCCURRENCES:

o April 27, 1994, 1410 EDST: Event date and time.

o April 27, 1994, 1410 EDST: Discovery date and time.

o April 27, 1994, 1410 EDST: Control Room operators verify both reactor trip breakers open, and all control and shutdown rods inserted.

o April 27, 1994, 1412 EDST: Control Room operators manually stop the operating main feedwater pump to limit a reactor coolant system cooldown.

o April 27, 1994, 1416 EDST: Control Room operators manually close both main steam isolation valves to limit a reactor coolant system cooldown.

o April 27, 1994, 1440 EDST: Plant stabilized at hot shutdown condition.

NRC FORM 366A (5-92)

NRC FORM 366A U.S. NUCLEAR REGULATORY CQOIISSION APPROVED BY Q(B HO. 3150-0104 (5-92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS IHFORMATION COLLECTIOH REQUEST: 50.0 HRS.

FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO LICENSEE EVENT REPORT (LER) THE INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S. NUCLEAR REGULATORY COMMISSION, TEXT CONTINUATION WASHINGTON, DC 20555-0001 AND TO THE PAPERWORK REDUCTION PROJECT (31/0-0104), OFFICE OF MANAGEMENT AND BUDGET WASHINGTON DC 20503.

FACILITY NAME 1 DOCKET HQIBER 2 LER NUMBER 6 PAGE 3 SEOUENTIAL REVISION YEAR R.E. Ginna Nuclear Power Plant 05000244 M

94 -- 007-- 00 3 OF 9 TEXT (If more space is required, use additional copies of HRC Form 366A) (17)

B. EVENT:

On April 27, 1994, a load increase was in progress, following the brief plant shutdown on April 26, 1994. Plant Operating Procedure 0-1.2 was being followed to control the load increase.

Only one main feedwater pump was in operation, and preparations were being made to start the second main feedwater pump and to continue with the load increase.

On April 27, 1994, at approximately 1407 EDST, the Control Room operators noticed a slight decrease in feedwater flow to the "A" S/G. They attempted to manually increase feedwater flow, but the "A" MFRV did not respond to the demand signal to open the valve from the Main Control Board. At approximately 1408 EDST, Main Control Board annunciator G-22, "ADFCS System Trouble" alarmed, due to a deviation between the demand signal and actual position of the "A" MFRV. Level continued to decrease in the "A" S/G. The Shift Supervisor ordered a rapid load reduction, in an attempt to decrease the need for feedwater flow. Within two minutes, power had been decreased by approximately 15%, and actual feedwater flow being delivered to the "A" S/G exceeded steam flow. However, level in the "A" S/G decrea'sed to ( 17oI resulting in a reactor trip on S/G Lo Lo level, at 1410 EDST.

The Control Room operators performed the immediate actions of Emergency Operating Procedure E-O, "Reactor Trip or Safety Injection", and transitioned to Emergency Operating Procedure ES-0.1', "Reactor Trip Response", when all it was verified that both reactor trip breakers were open, injection control not and shutdown rods actuated or were inserted, and safety was required. During performance of ES-0.1, the Control Room operators noted that an anticipated reactor coolant system (RCS) cooldown was occurring, and manually stopped the operating main feedwater pump. In addition, both main steam isolation valves (MSIVs) were manually closed by the Control Room operators.

These actions mitigated the RCS cooldown.

During this event, pressurizer (PRZR) level decreased below the setpoint for letdown isolation, closing the letdown isolation valves and deenergizing the PRZR heaters. After PRZR level was restored above the setpoint, the Control Room Foreman directed that letdown and PRZR heaters be restored to service. The plant was subsequently stabilized in hot shutdown (at approximately 1440 EDST) using Plant Operating Procedures 0-3, "Hot Shutdown with Xenon Present", and 0-2, "Plant Shutdown".

HRC FORM 366A (5-92)

HRC FORM 366A .S. NUCLEAR REGUULTORY COMMISSION PROVED BY OIB NO. 3150-0104 (5-92) EXP I RES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTIOH REQUEST: 50.0 HRS.

FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO LICENSEE EVENT REPORT (LER) THE INFORMATIOH AHD RECORDS MANAGEMENT BRANCH TEXT CONTINUATION (MNBB 7714), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001 AND TO THE PAPERWORK REDUCTION PROJECT (3140-0104), OFFICE OF MANAGEMENT AND BUDGET WASHINGTON DC 20503.

FACILITY NAME 1 DOCKET NMBER 2 i LER NMBER 6 PAGE 3 YEAR SEQUEH'TIAL REVISION R.E. Ginna Nuclear Power Plant 05000244 94 -- 007 00 4 OF 9 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

C~ INOPERABLE STRUCTURES I COMPONENTS I OR SYSTEMS THAT CONTRIBUTED TO THE EVENT:

None D. OTHER'YSTEMS OR SECONDARY FUNCTIONS AFFECTED:

None E. METHOD OF DISCOVERY:

This event was apparent due to Main Control Board indications of the loss of ability to control feedwater flow to the "A" S/G.

The reactor trip was immediately apparent due to alarms and indications in the Control Room.

F. OPERATOR ACTION:

After the reactor trip, the Control Room operators performed the actions of Emergency Operating Procedures E-O, "Reactor Trip or Safety Injection" and ES-0.1, "Reactor Trip Response". The operating main feedwater pump was manually stopped, and the MSIVs were manually closed to limit further RCS cooldown. The plant was stabilized at hot shutdown. Subsequently, the Control Room operators notified higher supervision and the Nuclear Regulatory Commission per 10CFR50.72, Non-Emergency, 4 Hour Notification.

G. SAFETY SYSTEM RESPONSES:

None III. CAUSE OF EVENT A. IMMEDIATE CAUSE:

The reactor trip was due to "A" S/G Lo Lo level ((/= 17%),

caused by decreased feedwater flow to the "A" S/G.

B. INTERMEDIATE CAUSE:

The decreased feedwater flow to the "A" S/G was due to loss of ability to control the "A" MFRV, caused by the valve positioner for the "A" MFRV not responding to the demand open signal.

HRC FORM 366A (5-92)

HRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION PROVED BY OMB NO. 3150-0104 (5-92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 COMMENTS REGARDING BURDEN ESTIMATE TO HRS.'ORWARD LICENSEE. EVENT REPORT (LER) THE INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S. NUCLEAR REGULATORY COMMISSIOH, TEXT CONTINUATION WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150 0104), OFFICE OF MANAGEMENT AND BUDGET WASHINGTON DC 20503.

FACILITY NAME 1 DOCKET NWBER 2 LER NIÃBER 6) PACE 3 SEQUENTIAL REVISION YEAR

'R.E. Ginna Nuclear Power Plant 05000244 94 -- 007-- ,00 5 OF 9 TEXT (lf more space is required, use additional copies of HRC Form 366A) (17)

C. ROOT CAUSE:

The underlying cause of the valve positioner for the "A" MFRV not properly responding to a change in input, demand signal from the controller on the Main Control Board was an improperly secured stroke adjust set screw. The set screw was found in a backed out condition in the valve position signal diaphragm assembly for the "A" MFRV. It is postulated that this stroke adjust allen head set screw did not have adequate thread sealant to prevent it from backing out of the signal diaphragm assembly cover when subjected- to vibration. With the set screw backed out, the signal diaphragm was restricted from responding to an input demand signal to open the MFRV. Note that the set screw is a factory-set adjustment. This particular set screw backed out when subjected to less than twelve hours of operation, and this failure mode has not occurred in other valve positioners that have 'accumulated thousands of hours of operation. This event is NUREG-1022 Cause Code (B), "Design, Manufacturing, Construction/Installation."

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 manual or automatic actuation of any engineered safety feature (ESF) including the reactor protection system (RPS)". The "A" S/G Lo Lo level reactor trip was an automatic actuation of the RPS.

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

o There were no safety consequences or implications attributed to the reactor trip because:

• The two reactor trip breakers opened as required.

• All control and shutdown rods inserted as designed.

• The plant was stabilized at hot shutdown.

HRC FORM 366A (5-92)

NRC FORH 366A U.S NUCLEAR REGULATORY COMMISSION APPROVED BY (HGI NO. 3150-0104 (5-92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLEC'IIOH REQUEST: 50.0 HRS ~

FORNARD COMMENTS REGARDING BURDEN ESTIMATE TO LICENSEE EVENT REPORT (LER) tHE INFORMATION AND RECORDS MANAGEMENT BRANCH TEXT CONTINUATION (MNBB 7714), U.S. NUCLEAR REGULATORY COMMISSION, MASHIHGTON, DC 20555-0001, AND TO THE PAPERNORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET MASHINGTON DC 20503.

FACILITY NAME 1 DOCKET HINSER 2 LER NIMBER 6 PAGE 3 I YEAR SEQUENTIAL REVISION R.E. Ginna Nuclear Power Plant 05000244 94 -- 007 00 6 OF 9 TEXT (If more space is required, use additional copies of NRC Form 366A) (17) o The Ginna Updated Final Safety Analysis Report (UFSAR) transient, as described in Chapter 15.2.6, "Loss of Normal Feedwater", describes a condition where the reactor trips on Lo Lo S/G level. This UFSAR transient was reviewed and compared to the plant response for this event. The UFSAR transient is a complete loss of Main Feedwater (MFW) at full power, with only one auxiliary feedwater (AFW) pump available one (1) minute after the loss of MFW, and secondary steam relief (i.e., decay heat removal) through the safety valves only. The protection against a loss of MFW includes the reactor trip on Lo Lo S/G level and the start of the AFW pumps. These protection features operated as designed.

Based on the above evaluation, the plant transient of April 27, 1994, is bounded by the UFSAR Safety Analysis assumptions.

o Technical Specifications (TS) were reviewed with respect to the post trip review data. 'he following are the results of that review:

'oFollowing the reactor trip, PRZR water level decreased approximately 12.5  % due to a moderate RCS cooldown.

This cooldown occurred during the post trip recovery period. This cooldown was bounded by the plant accident analysis, and did not exceed the TS limit of 100 degrees F per hour. Additional mitigation was provided by closing the MSIVs and stopping -the main feedwater pump. TS 3.1.1.5 states, in part, that when the RCS temperature is at or above 350 degrees F, at least 100 KW of PRZR heaters will be operable. TS 3.1.1.5 also states, in part, that inoperable due to heaters, restore the PRZR'to operable if the PRZR,is status within six (6) hours. PRZR water level was restored above the setpoint for letdown isolation within two (2) minutes, restoring the PRZR heaters to operable status, well before the six '(6) hour action statement.

NRC FORM 366A (5-92)

NRC FORH 366A U.S. NUCLEAR REGULATORY COMHISSION APPROVED BY QGI NO. 3150-0104 (5.92) EXPIRES 5/31/95 ESTIHATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REOUEST: 50.0 HRS.

FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO LICENSEE EVENT REPORT (LER) THE INFORMATION AND RECORDS MANAGEHENT BRANCH TEXT CONTINUATION (MNBB 7714), U.ST NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001 AND TO THE PAPERWORK REDUCTION PROJECT (3140 0104), OFFICE OF HANAGEMENT AND BUDGET WASHINGTON DC 20503.

FACILITY NAHE 1 DOCKET NINBER 2 LER NUMBER 6 PAGE 3 SEOUENTIAL REVISION YEAR R.E. Ginna Nuclear Power Plant 05000244 94 -- 007-- 00 7 OF 9 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

• Both S/G levels decreased following the reactor trip. "A" S/G level decreased to < 04, and "B" S/G level decreased to

< 14'. This is an expected transient. TS 4.3.5.5 states that in order to demonstrate that a reactor coolant loop is operable, the S/G water level shall be >/= 169. Thus, both coolant loops were. inoperable, even though both loops were still in operation and performing their intended function of decay heat removal. Both S/Gs were available as a heat sink, and sufficient AFW flow was maintained for adequate steam release from both S/Gs. TS 3.1.1.1(c) states, in part, that except for special tests, when the RCS temperature is at or above 350 degrees F with the reactor power less than or equal to 130 MWT (8.5%), at least one reactor coolant loop and its associated S/G and reactor coolant pump shall be in operation. Both reactor coolant loops were in operation, but the S/Gs were inoperable due to level indication. Both loops were restored to operable status when S/G levels were restored to >/= 164 ("B" S/G level in less than four (4) minutes, and "A" S/G level in approximately ten (10) minutes).

Based on the above and a review of post trip data and past plant transients, it can be concluded that the plant operated as designed, that there were no unreviewed safety questions, and that the public's health and safety was assured at all times.

NRC FORM 366A (5-92)

NRC FORH 366A U.S. NUCLEAR REGULATORY C(SOIISSIOH APPROVED BY MB NO. 3150-0104 (5-92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.

FORWARD COMMENTS REGARDIHG BURDEN'STIMATE TO LICENSEE EVENT REPORT (LER) THE INFORMATION AND RECORDS MANAGEMENT BRANCH TEXT CONTINUATION (MHBB 7714), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001 AND TO THE PAPERWORK REDUCTION PROJECT (3140-0104), OFFICE OF MANAGEMENT AND BUDGET 'WASHINGTON DC 20503.

FACILITY NAME 1 DOCKET NQIBER 2 LER HINBER 6 PAGE 3 YEAR SEQUENTIAL REVISION R.E. Ginna Nuclear Power Plant 05000244 M

94 -- 007-- 00 8 OF 9 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

V. CORRECTIVE ACTION A. ACTION TAKEN TO RETURN AFFECTED SYSTEMS TO PRE-EVENT NORMAL STATUS:

o The valve positioner for the "A" MFRV (Bailey Model AV112100I or "Type AV1") had been replaced on April 26, 1994, with another Type AV1 positioner. This specific positioner failed in less than twelve hours of use and was replaced after this event (on April 28, 1994) with a Bailey Model 5321030A10 (or "5321030") valve positioner.

The Bailey Model 5321030 is the original model of valve positioner for the MFRV application and had operated successfully since plant startup in 1969. This positioner model was changed out in 1991 as part of EWR 4773, which installed the Advanced Digital Feedwater Control System (ADFCS). The Type AV1 positioners have had a history of reliability and stability concerns in this MFRV application.

o The valve positioner for the "B" MFRV was also replaced on April 28, 1994, for the reasons discussed above.

o Valve. ramp and step change diagnostic testing was performed for both MFRVs to verify proper valve positioning and response.

B. ACTION TAKEN OR PLANNED TO PREVENT RECURRENCE:

o The positioners for both MFRVs were replaced with Model 5321030 positioners as discussed above.

HRC FORM 366A (5-92)

NRC FORM 366A .S. NUCLEAR REGULATORY CQOIISSIOH PROVED BY OMB NO. 3150-0104 (5-92) EXPIRES 5/31/95 I

ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTIOH REDUEST: 50.0 HRS.

FORWARD COMMENTS REGARDIHG BURDEH ESTIMATE TO LICENSEE EVENT REPORT (LER) THE INFORMATION AND RECORDS MANAGEMENT BRANCH TEXT CONTINUATION (MHBB 7714), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001 AHD TO THE PAPERWORK REDUCTION PROJECT (3140.0104), OFFICE OF MANAGEMENT AND BUDGET WASHINGTON DC 20503.

FACILITY NAME 1 DOCKET NUMBER 2 LER NUMBER 6 PAGE 3 YEAR SEOUENTIAL REVISION R.E. Ginna Nuclear Power Plant 05000244 M

94 -- 007-- 00 9 OF 9 TEXT (If more space is required, use additional copies of HRC Form 366A) (17)

VI. ADDITIONAL INFORMATION A. FAILED COMPONENTS:

The failed component was the stroke adjust allen head set screw for the valve positioner signal diaphragm assembly for the "A" MFRV. The positioner is a Model AV112100 positioner, manufactured by Bailey Controls Inc.

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 iden-tified; however, LER 93-006 was a similar event, in that there was loss of ability to control the "A" MFRV; and LERs88-005, 90 007 I and 90-0 1 0 were simi lar events with different root causes.

C. SPECIAL COMMENTS:

None HRC FORM 366A (5-92)