ML17264A605

LER 96-012-00:on 960820,feedwater Transient Occurred,Due to Closure of Feedwater Regulating Valve,Causing Lo Lo Steam Generator Level Reactor Trip.Sgs Were Restored & Missing Screw in 1/P-476 Was replaced.W/960919 Ltr

ML17264A605
Person / Time
Site:	Ginna
Issue date:	09/19/1996
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-96-012, LER-96-12, NUDOCS 9609270247
Download: ML17264A605 (17)

Text

CATEGORY REGULA'Y INFORMATION DISTRIBUTION SYSTEM (RIDS)

'l ACCESSION NBR:9609270247 DOC.DATE: 96/09/19 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 a Electric Corp.

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

RECIP.NAME RECIPIENT AFFII IATION VISSING.G.S.

C

SUBJECT:

LER 96-012-00:on 960820,feedwater transient occurred,due to closure of feedwater regulating valve, causing lo lo steam generator level reactor trip. SGs were restored a missing screw in 1/p-476 was replaced. W/960919 ltr.

DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR J ENCL TITLE: 50.73/50.9 Licensee Event Report (LER), Incident Rpt, etc. J SIZE: E Q

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

RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD1-1 PD 1 1 VISSING,G. 1 1 INTERNAL: AEOD SPD/RAB 1 1 AEOD/SPD/RRAB 1 1 ILE C NRR/DE/EELB E~ 1 1

1 NRR/DE/ECGB 1 1 1 NRR/DE/EMEB 1 1 NRR/DRCH/HHFB 1 1 NRR/DRCH/HICB 1 1 NRR/DRCH/HOLB 1 1 NRR/DRCH/HQMB 1 1 NRR/DRPM/PECB 1 1 NRR/DSSA/SPLB 1 1 D NRR/DSSA/SRXB 1 1 RES/DSIR/EIB 1 1 RGN1 FILE 01 1 1 0 EXTERNAL: L ST LOBBY WARD 1 1 LITCO BRYCE,J H 1 1 NOAC MURPHY,G.A 1 1 NOAC POOREgW. 1 1 NRC PDR 1 1 NUDOCS FULL TXT 1 1 U NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM OWFN SD-5(EXT ~ 415-2083) TO ELIMINATE YOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!

FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR '23 ENCL 23

AND ROCHESTER GAS AND E1ECTRIC CORPORAT1ON ~ 89 EASTAVENUF, ROCHESTER, N. Y 1d6d9.0D01 AREA CODE716 546-27M ROBERT C. MECREDY V ee president seuc~eor Opesotions September 19 1996 U.S. Nuclear Regulatory Commission Document Control Desk Attn: Guy S. Vissing Project Directorate I-1 Washington, D.C. 20555

Subject:

LER 96-012, Feedwater Transient, Due to Closure of Feedwater Regulating Valve, Causes a Lo Lo Steam Generator Level Reactor Trip R.E. Ginna Nuclear Power Plant Docket No. 50-244

Dear Mr. Vissing:

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 96-012 is hereby submitted.

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

Very truly yours, Robert C. Mecred xc: Mr. Guy'S. Vissing (Mail Stop 14C7)

PWR Project Directorate I-1 Washington, D.C. 20555 U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 Ginna Senior Resident Inspector 9b09270247 9b09i9 PDR ADQCK 05000244 S PDR

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NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSIO APPROVED BY OMB NO. 3150<104 (4-95) EXPIRES 04/30/9B ESTIMATED BURDEN PER RESPONSE To COMPLY WITH THIS MANDATORY INFORMATION COLLECTION REQUEST: 50.0 HRS.

I L CENSEE EVENT REPORT (LER) REPORTED LESSONS LEARNED ARE INCORPORATED INTO THE UCENSING PROCESS AND FED BACK TO INDUSTRY. FORWARD COMMENTS REGARDING BURDEN ESTIMATE To THE (See reverse for required number of INFORMATION AND RECORDS MANAGEMENT BRANCH IT.6 F33),

digits/characters for each block) U.S. NUCI.EAR REGULATORY COMMISSION, WASHINGTON, Dc 20555.0001, AND TO THE PAPERWORK REDUCTION PROJECT FACIUTY NAME I1) OOCKET NUMBER IR) PAGE )3)

R.E. Ginna Nuclear Power Plant 05000244 1 OF8 TITLE I4)

Feedwater Transient, Due to Closure of Feedwater Regulating Valve, Causes a Lo Lo Steam Generator Level Reactor Trip EVENT DATE (5) LER NUMBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED (6)

FACILITY NAME DOCKET NUMBER SEQUENTIAL REVISION MONTH DAY YEAR NUMBER NUMBER MONTH DAY YEAR FACKJTY NAME OOCKET NUMBER 08 20 96 96 012 00 09 96 OPERATING THIS REPORT IS SUBMITTED PUR SUANT TO THE REQUIREMENTS OF 10 CFR 5) (Check ono or mote) (11)

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

POWER 20.2203(a)(1) 20.2203(a) (3) (i) 50.73(a) (2) (ii) 50.73(a) (2) (x)

LEVEL (10) 20.2203(a)(2)(i) 20.2203(a) l3) BI) 50.73(a) (2) (iii) 73.71

20. 2203(a) (2) (ii) 20.2203(a) (4) X 50.73(a)(2)(iv) OTHER 20.2203la)(2) (iii) 50.36(c) ( I) 50.73(a)(2) (v) Specify in Abstract bolo W or in NRC Form 366A 20.2203(a) (2) (iv) 50.36(c) (2) 50.73(a)(2)(vii)

LICENSEE CONTACT FOR THIS LER l12)

NAME TELEPHONE NUMBER (Ioolode Aree Code)

John T. St. Martin - Technical Assistant (716) 771-3641 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DES CRIBED IN THIS REPORT (13)

SYSTEM COMPONENT MANUFACTURER CAUSE SYSTEM COMPONENT REPORTABLE CAUSE TO NPRDS MANUFACTURER TO NPRDS SJ TD R369 SUPPLEMENTAL REPORT EXPECTED (14) EXPECTED MONTH DAY YEAR YES 6 UBMIssl0 N (If yes, complete EXPECTED SuBMISSION DATE). X NO DATE (15)

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

On August 20, 1996, at approximately 1442 EDST, with the plant in Mode 1 at approximately 100% steady state reactor power, the "B" main feedwater regulating valve went to the fully closed position. At 1443 EDST, the reactor tripped on Lo Lo level in the "B" Steam Generator. The Control Room operators performed the actions of procedures E-0 and ES-0.1. Following the reactor trip, all systems operated as designed, and the reactor was stabilized in Mode 3.

The underlying cause of the closure of the "B" main feedwater regulating valve was determined to be a loss of electrical continuity, caused by a missing screw in the current-to-pressure transducer for the "B" main feedwater regulating valve.

Corrective action was to replace the missing screw.

This event is NUREG-1022 Cause Code (A).

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

NRC FORM 366 (4.95)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (4-95)

LICENSEE EVENT REPORT (LER)

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

YEAR SEQUENTIAL REVISION NUMBER NUMBER R.E. Ginna Nuclear Power Plant 05000244 2 OF 8 96 012 00 TEXT ilfmore space is required, use addidonal copies of ftVRC Form 386Ai (17)

PRE-EVENT PLANT CONDITIONS:

On August 20, 1996, the plant was in Mode 1 at approximately 100% steady state reactor power. At approximately 1442 EDST, the Control Room operators received several Main Control Board Annunciator alarms. These alarms indicated that there was a problem in the Advanced Digital Feedwater Control System (ADFCS), and that a main feedwater regulating valve (MFRV) was now in manual control. The Control Room operators observed that the "B" MFRV had closed and feedwater flow to the "B" SG was not adequate for 100% steady state power operation. The Control Room operators responded to these alarms and attempted to restore adequate flow to the "B" Steam Generator (SG) by opening the MFRV.

Attempts were unsuccessful, and water level in the "B" SG was rapidly decreasing due to the loss of feedwater flow to that SG.

DESCRIPTION OF EVENT'.

DATES AND APPROXIMATE TIMES OF MAJOR OCCURRENCES:

August 20, 1996, 1442 EDST: Valve positioner failure.

August 20, 1996, 1443 EDST: Event date and time.

August 20, 1996, 1443 EDST: Discovery date and time.

August 20, 1996, 1444 EDST: Control Room operators verify both reactor trip breakers open and verify all control and shutdown rods inserted.

August 20, 1996, 1450 EDST: Control Room operators manually close both main steam isolation valves to limit a reactor coolant system cooldown.

August 20, 1996, 1453 EDST: Control Room operators manually stop both main feedwater pumps to limit a reactor coolant system cooldown.

August 20, 1996, 1545 EDST: Plant is stabilized in Mode 3.

EVENT:

On August 20, 1996, at approximately 1443 EDST, the plant was in Mode 1 at approximately 100% steady state reactor power. Feedwater flow to the "B" SG was inadequate, and water level in the "B" SG was rapidly decreasing. When the "B" SG level was at 20% (and still decreasing),

the Control Room Foreman ordered a manual reactor trip. Before the Control Room operators performed a manual reactor trip, the reactor automatically tripped on Lo Lo level in the "B" SG

(( 17%)

NRC FORM 366A (4-95)

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1 NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (4-95)

LXCENSEE EXTENT REPORT (LER)

TEXT CONTINUATION FACILITYNAME (1) DOCKET LER NUMBER (6) PAGE (3)

YEAR SEQUENTIAL REVISION NUMBER NUMBER 4 OF 8 R.E. Ginna Nuclear Power Plant 05000244 96 012 00 TEXT iifmore spece is required, use eddiuonal copies of PVRC F'arm 366AJ (17)

E. METHOD OF DISCOVERY:

This event was immediately apparent due to Main Control Board indication of inadequate feedwater flow to the "B" SG. The reactor trip was immediately apparent due to plant response and alarms and indications in the Control Room.

F. OPERATOR ACTION:

After the reactor trip, the Control Room operators performed the appropriate actions of Emergency Operating Procedures E-0 and ES-0.1. Feedwater flow to the "A" SG was stopped to mitigate the

'ncrease in "A" SG level. The MSIVs were manually closed and both MFW pumps stopped to limit further RCS coo!down. Appropriate actions were taken to restore level in the "B" SG and to minimize level increase in the "A" SG.

The setting for lifting of the SG atmospheric relief valves (ARV) was lowered from 1050 PSIG to minimize a subsequent RCS heatup (and prevent PRZR overpressure). The plant was stabilized in Mode 3.

Subsequently, the Control Room operators notified higher supervision and the NRC per 10 CFR 50.72 (b) (2) (ii), non-emergency four hour notification, at approximately 1755 EDST on August 20, 1996.

G. SAFETY SYSTEM RESPONSES:

AII safeguards equipment functioned properly. Both motor-driven AFW pumps started when "B" SG level decreased below 17% after the reactor trip. The turbine-driven AFW pump started as per design, due to a starting signaI from AMSAC. Main feedwater isolation occurred on high level in

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the "A" SG (i.e., 85% narrow range level).

III. CAUSE OF EVENT:

A. IMMEDIATECAUSE:

The immediate cause of the reactor trip was due to "B" SG Lo Lo level (( 17%), caused by inadequate feedwater flow to the "B" SG.

B. INTERMEDIATE CAUSE:

The intermediate cause of the inadequate feedwater flow to the "B" SG was the closure of the "B" MFRV, caused by the current-to-pressure (I/P) transducer not responding to the input demand signal. This resulted in loss of input demand signal to the "B" MFRV valve positioner.

NRC FORM 366A (4-95)

II NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (4-95)

LICENSEE EVENT REPORT (LER)

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

YEAR SEQUENTIAL REVISION NUMBER NUMBER R.E. Ginna Nuclear Power Plant 05000244 5 OF 8 96 012 00 TEXT llfmore speceis required, use edditionel copies of NRC Form 386'A/ (17)

ROOT CAUSE:

The underlying cause of the loss of input demand signal to the "B" MFRV valve positioner was a loss of electrical continuity from the terminal block to the circuit board on the terminal block inside the current-to-pressure transducer (I/P-476) that supplies air pressure to the "B" MFRV.

This loss of continuity was the result of a missing screw which caused an unreliable input signal connection, resulting in loss of the signal to the transducer, and caused the output air signal to decrease to minimum. On minimum air pressure, the MFRV goes fully closed.

The basic design of the Rosemount Model 3311 I/P transducer (I/P-476) is significantly different when compared to other instrumentation. The mounting of the circuit board to the terminal block is unique, and special instructions or guidance were absent in the manufacturer's technical manual. Four screws are installed in the terminal block in these Rosemount transducers. Two are used for field wire connections, and two are used to hold down the terminal block connection board.

This event is NUREG-1022 Cause Code (A), "Personnel Error". A Human Performance Enhancement System (HPES) evaluation was initiated for this event. The HPES evaluation concluded that, in the event a screw was discovered missing on the terminal block for these transducers, it had been a previously accepted practice for Instrument and Control (I(AC) technicians not to replace the screw, and to reconnect any wiring onto a different screw, as long as it was the same electrical point, same terminal block, and same terminal number. This practice does not affect electrical continuity for transducers of a different design, since no screws on the terminal block hold down the terminal block connection board. However, on Rosemount transducers, ail four screws are required for their specific function.

This error was a cognitive error, in that the IRC technicians did not understand the detailed function of each screw, and did not recognize that their practice could cause unreliable connections in the transducer. This error was not contrary to any approved procedures and is not covered in detail in any procedure. There are no unusual characteristics of the locations for any of these transducers.

The failure of the "B" MFRV I/P transducer meets the NUMARC 93-01, "Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants", definition of a "Maintenance Preventable Functional Failure".

NRC FORM 366A (4-95)

II NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (4.95)

LICENSEE %WENT REPORT (LER)

TEXT CONTINUATION FACILITYNAME (1) DOCKET LER NUMBER (6) PAGE (3)

YEAR SEQUENTIAL REVISION NUMBER NUMBER R.E. Ginna Nuclear Power Plant 05000244 6 OF 8 96 012 00 TEXT iifmore speceis required, use edditionel copies of NRC Form 366A j (17)

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 "B" SG Lo Lo level reactor trip was an automatic actuation of the RPS, and MFW isolation and AFW pump starts are actuations of an ESF component.

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 reactor trip because:

o The two reactor trip breakers opened as required.

o AII control and shutdown rods inserted as designed.

o The plant was stabilized in Mode 3.

The Ginna Station Improved Technical Specifications (ITS) Limiting Conditions for Operation (LCOs) and Surveillance Requirements (SRs) were reviewed with respect to the post trip review data. The following are the results of that review:

PRZR pressure decreased below 2205 PSIG during the transient after the reactor trip.

During this time a thermal power step >10% occurred due to the reactor trip, which is within the limits of ITS LCO 3.4.1. Therefore, compliance with ITS was maintained. The RCS temperature DNB limit (577.5 degrees F) was not approached. Additional mitigation was provided by closing the MS)Vs and stopping the MFW pumps. Minimum PRZR pressure was approximately 2092 PSIG.

After the reactor trip, the RCS cooled down to approximately 539 degrees F and was subsequently stabilized at 547 degrees F. The cooldown was within the limits of ITS LCO 3.4.3. In addition, the required shutdown margin was maintained at all times during the RCS cooldown.

Both SG levels decreased following the reactor trip. "B" SG level decreased below 16%

indicated narrow range level SR 3.4.5.2 states that in order to demonstrate that a reactor

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coolant loop is operable, the SG water level shall be >/= 16%. Thus, the "B" coolant loop was inoperable, even though it was still in operation and performing its intended function of decay heat removal.

Both SGs were available as a heat sink, and sufficient AFW flow was maintained for adequate steam release from both SGs. The "8" coolant loop was restored to operable status when SG level was restored to >/= 16%, in approximately thirty-five (35) minutes.

This is within the limits of ITS LCO 3.4.5 ACTION A.

NRC FORM 366A (4-95)

I NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (4.95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITYNAME (1) DOCKET LER NUMBER (6) PAGE (3)

YEAR SEQUENTIAL REVISION NUMBER NUMBER R.E. Ginna Nuclear Power Plant 05000244 7 OF 8 96 012 00 TEXT iifmore space is required, use eddi tionel copies of NRC Form 366A/ (17) o The Ginna Station 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 SG 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 AFW pumps 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 SG level and the start of AFW pumps. These protection features operated as designed. Based on the above evaluation, the plant transient of August 20, 1996, is bounded by the UFSAR Safety Analysis assumptions.

o The UFSAR transient, as described in Chapter 15.1.2, "Increase in Feedwater Flow at Full Power", describes a condition where the automatic operation of the main feedwater isolation provided protection from potential SG overfill and damage to the turbine and steam piping due to water carryover. Prudent operator action provided the necessary action to reduce SG level. The high level in the "A" SG that resulted during the transient is bounded by the UFSAR Safety Analysis assumptions.

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.

V. CORRECTIVE ACTION:

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

The SGs were restored to operable status when SG level in the "B" SG increased above 16% level, by addition of AFW. Subsequently, levels were restored to their normal operating levels.

The missing screw in I/P-476 was replaced.

Both MFRVs were operated fully open and fully closed from the Main Control Board hand controller to verify proper valve positioning and response.

B. ACTION TAKEN OR PLANNED TO PREVENT RECURRENCE:

o There are six (6) Rosemount Model 3311 I/P transducers in use at Ginna Station. All 6 were inspected. In addition to the missing screw for I/P-476, a broken field wire connection screw was found in I/P-466 (for the "A" MFRV), and the field wire was landed on one of the terminal board screws. A terminal board screw was missing in the transducer for the "B" SG atmospheric relief valve, and was later found in a nearby conduit. The configurations of all Rosemount transducers were restored to approved configurations. ILC technicians have been made aware of the unusual arrangement of the terminal block screws in Rosemount transducers.

NRC FORM 366A (4-95)

I NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION I4-95)

LICENSEE EVZRFZ REPORT (LER)

TEXT CONTINUATION FACILITY NAME I1) DOCKET LER NUMBER I6) PAGE I3)

YEAR SEQUENTIAL REVISION NUMBER NUMBER R.E. Ginna Nuclear Power Plant 05000244 8 OF 8 96 012 00 TEXT ilfmore speceis required, use eddirional copies of NRC Form 366A/ I17)

Calibration procedures for all 6 Rosemount transducers have been changed to ensure that all four screws are in place and wires are landed on the correct terminal points.

Nuclear Training Work Requests (NTWR) have been written to incorporate the lessons learned into the l&C training program.

Vl. ADDITIONALINFORMATION:

A. FAILED COMPONENTS:

The failed component (I/P-476) was a Rosemount Model 3311 I/P transducer.

PREVIOUS LERs ON SIMILAR EVENTS:

A similar LER event historicai 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.

However, LERs93-006 (due to connecting screw for linkage feedback arm) and 94-007 (due to set screw backing out of valve position signal diaphragm assembly) were similar events, in that there was a loss of ability to control a MFRV which resulted in a reactor trip. LERs85-006, 88-003,88-005, 90-007,90-010, 92-002, and 92-003 were similar events (reactor trip from Lo SG level) with different root causes.

C. SPECIAL COMMENTS:

None NRC FORM 366A I4-95)