ML17265A613

LER 99-002-00:on 990227,discovered That Surveillance Had Not Been Performed at Frequency,Per Ts.Caused by Personnel Error.Procedure O-6.13 Will Be Evaluated for Enhancement Documentation of Completion of ITS Srs.With 990329 Ltr

ML17265A613
Person / Time
Site:	Ginna
Issue date:	03/29/1999
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-99-002, LER-99-2, NUDOCS 9904080022
Download: ML17265A613 (10)

Text

~ CATEGORY la REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9904080022 DOC.DATE: 99/03/29 NOTARIZED: NO DOCKET 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 F Electric Corp.

RECIP.NAME RECIPIENT AFFILIATION VISSING,G.S.

SUBJECT:

LER 99-002-00:bn 990227,discovered that surveillance had not been perfohned at frequency,per TS.Caused by personnel error. Procedure 0-6.13 will be evaluated for enhancement documentation of completion of ITS SRs.With 990329 ltr.

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

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 S 2 2 AEOD/SPD/RRAB 1 1 CENT 1 1 NRR/DRCH/HOHB 1 1

~DRG - QMB 1 1 NRR/DRPM/PECB 1 1 NRR/DSSA/SPLB 1 1 RES/DET/EIB 1 1 RGN1 FILE 01 1 1 D

EXTERNAL: L ST LOBBY WARD 1 1 LMITCO MARSHALL 1 1 NOAC POORE,W. 1 1 NOAC QUEENER,DS 1 1 0 NRC PDR 1 1 NUDOCS FULL TXT 1 1 N

T NOTE TO ALL "RZDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE. TO HAVE YOUR NAME OR ORGANIZATION REMOVED FROM DISTRZBUTION LISTS OR REDUCE THE NUMBER OF COPIES RECEIVED BY YOU OR YOUR ORGANIZATION, CONTACT THE DOCUMENT CONTROL DESK (DCDj ON EXTENSION 415-2083 FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 18 ENCL 18

AND g~gPjP~Pw ~ 'qP ~,P 'V'(', e'>'4 i <v'A A>> i".E s March 29, 1999 U. S. Nuclear Regulatory Commission Document Control Desk Attn: Guy S. Vissing Project Directorate I-1 Washington, D.C. 20555

Subject:

LER 1999-002, Surveillance Not Performed, Due to Personnel Error, Resulted in Violation of Technical Specifications 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) (i) (B), which requires a report of, "Any operation or condition prohibited by the plant's Technical Specifications",

the attached Licensee Event Report LER 1999-002 is hereby submitted.

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

Very ruly yours, Robert C. Me redy 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

'7904080022 9'70329 PDR ADOCK 05000244 8 PDR

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION g(PF'IPJ(DQY OMB NOB 3$ 50$ IPWILS 0+63@2001 (9 1998).

infotmaUon cot(ection request: 50 hrs. Reported lessons learned aro incorporated into the licensing process and fod back to LICENSEE EVENT REPORT (LER) industry. Fonvard comments regarding burden esUmato to tho Records Management Branch (TW F33). U.S. Nuc(oar Regulatory Commission, Washington, DC 20555000(, and to (See reverse for required number of tho Papetwotk Reduction Project (315(40104), Ofrco of digits/characters for each block) Management and Budget, Washington, DC 20503. If an information coBection does nct display a currently valid OMB control number. the NRC may nct conduct or sponsor. and a FACILITY NAME (1) 00CKET NUMBER (2I PAGE (3)

R.E. Ginna Nuclear Power Plant 05000244 1 OF 5 TITLE te)

Surveillance Not Performed, Due to Personnel Error, Resulted in Violation of Technical Specifications EVENT DATE (5) LER NUMBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED (8)

FACILITY N AME DOCKET NUMBER SEQUENTIAL REVISION MONTH OAY YEAR YEAR MONTH OAY YEAR NUMBER NUMB(R 05000 02 27 1999 1999 - 002 - 00 03 29 1999 FACILITYNAME DOCKET NUMBER 05000 OPERATING THIS REPORT IS SUBMITTED P URSUANTTO THE REQUIREMENTS OF 10 CFR EI (Check ono or more) (11)

MODE (9) 20.2201(b) 20.2203(a)(2)(v) 50.73(a) (2)(l) 50.73(a)(2) (vill)

POWER 20.2203(a)(1) 20.2203(a)(3)(l) 50.73(a)(2)(II 50.73(a) (2) (xl LEVEL (10) 70 20.2203(a)(2) (I) 20.2203(a) (3) (n) 50.73(a)(2)(III 73.71 20.2203(a) (2) (II) 20.2203(a)(4) 50.73(a) (2)((v OTHER 20.2203(a) (2) (Ill) 50.36(c)(1) 50.73(a) (2)(v Specify in Abstract below or 20.2203(a)(2)(lv) 50.36(c) (2) 50.73(a) (2) (vl (n NRC Form 366A LICENSEE CONTACT FOR THIS LER (12)

NAME TELEPHONE NUMBER Uscexfe Ates Code)

John T. St. Martin - Technical Assistant (716) 771-3641 REPORTABLE REPORTABLE CAUSE SYSTEM COMPONENT'ANUFACTURER CAUSE SYSTEM COMPONENT MANUFACTURER TO EPIX TO EPIX SUPPLEMENTAL REPORT EXPECTED R4) MONTH OAY EXPECTED YES SUBMISSION (If yes, complete EXPECTED SUBMISSION DATE). X NO DATE ((5)

ABSTRACT (Llmlt to 1400 Spaces, I.e., approximately 15 single-spaced typewritten Hnos) (16)

On February 27, 1999, at approximately 1600 EST, the plant was in coastdown prior to the 1999 refueling outage, in Mode 1 at approximately 70% steady state reactor power. It was discovered that a surveillance had not been performed at the frequency required by the plant's Technical Specifications. This constituted a missed surveillance, which is a condition prohibited by the plant's technical specifications.

Control Room operators verified that the plant was currently in compliance with these surveillance requirements. No immediate corrective action was needed.

The underlying cause of not performing this surveillance was a personnel error.

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

0 NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (6 1996)

LICENSEE EVENT REPORT (LER)

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

SEQUENTIAL REVISI yEAR ON NUMBER NUMBE R.E. Ginna Nuclear Power Plant 05000244 2 OF 5 1999 002 00 TEXT ill more spaceis reouired, use additional copies ol NRC Form 366AI (171 PRE-EVENT PLANT CONDITIONS:

On February 27, 1999, at approximately 1600 EST, the plant was in coastdown prior to the 1999 refueling outage, in Mode 1 at approximately 70% steady state reactor power. One of the responsibilities of the Control Room operators is to perform the computer checks in accordance with plant procedure S-26.1 (Computer Program Check). Control Room operators had just completed these computer checks, to meet the Surveillance Requirements (SR) of the Ginna Station Improved Technical Specifications (ITS). Step 5.4 of plant procedure 0-6.13 (Daily Surveillance Log) requires that the computer program checks be performed every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

II. DESCRIPTION OF EVENT:

A. DATES AND APPROXIMATE TIMES,OF MAJOR OCCURRENCES o February 26, 1999, 2318 EST: SR 3.2.3.1 is performed.

o February 27, 1999, 0728 EST: It was intended that SR 3.2.3.1 be performed.

o February 27, 1999, 1418 EST: Event date and time.

o February 27, 1999, 1539 EST: SR 3.2.3.1 is performed.

o February 27, 1999, 1600 EST: Discovery date and time.

o February 27, 1999, 2326 EST: SR 3.2.3.1 is performed.

EVENT:

On February 27, 1999, at approximately 1600 EST, the plant was in coastdown prior to the 1999 refueling outage, in Mode 1 at approximately 70% steady state reactor power. Control Room operators had just completed computer checks, to meet the SR of the ITS. Specifically, SR 3.2.3.1, which verifies operability of the Axial Flux Difference (AFD) monitor, had been performed at approximately 1539 EST. Performance of this SR is required at a Frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, per ITS SR 3.2.3.1. Ginna Station administrative requirements, as listed in Operations Procedure 0-6.13, "Daily Surveillance Log", are to perform this SR once during every eight hour time block. One of the operators mentally questioned whether he had actually performed SR 3.2.3.1 when he had previously performed these computer checks at approximately 0728 EST. Note that proper performance of SR 3.2.3.1 results in the receipt of several alarms which are received on the Plant Process Computer System (PPCS) and documented on the PPCS alarm printout.

The Control Room operator conducted additional self-checking. He realized his error during the second check of the day when, using the procedure, he could not recall performing step 5.4 of procedure S-26.1. He immediately informed the Control Room Foreman about the potential missed surveillance. After a check of the PPCS alarm printout, both the Foreman and the Control Room Operator realized that the surveillance had, in fact, been missed. They discovered that there had been no PPCS alarms generated around the timeframe of 0738 EST, and concluded that SR 3.2.3.1 had not been performed at 0728 EST. Review of the PPCS alarm printout confirmed that SR 3.2.3.1 had been performed at approximately 2318 EST on February 26, 1999.

~-

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (6.1998)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME I1) DOCKET (2) LER NUMBER (6) PAGE I3)

SEQUENTIAL REVISI NUMBER NUMBE R.E. Ginna Nuclear Power Plant 05000244 - - 3 OF 5 1999 002 00 TEXT fffmore spece is rerfuired, use eddi tionel copies of NRC Form 366Af )171 In accordance with ITS SR 3.2.3.1, this SR was due within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Applying the applicability requirements'of ITS SR 3.0.2, the specific Frequency is met if the surveillance is performed within 1.25 times the interval specified (1.25 times 12 results in 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />). Thus, the latest SR 3.2.3.1 could be performed would have been 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> after 2318 hours0.0268 days <br />0.644 hours <br />0.00383 weeks <br />8.81999e-4 months <br />, or approximately 1418 EST on February 27, 1999. Time 1418 on February 27 is the Event date and time.

Although SR 3.2.3.1 had been missed at 0728 EST, the AFD monitor continued to be fully functional. Review of PPCS alarms confirmed that SR 3.2.3.1 had been properly performed at all other times in February 1999, at the frequency required by the plant's Technical Specifications and procedure 0-6.13.

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 indicated during mental questioning of previous. actions, and was formally discovered by review of PPCS alarm printouts.

F. OPERATOR ACTION:

The Control Room operators identified that SR 3.2.3.1 had inadvertently not been performed at approximately 0728 EST on February 27. At the time of discovery (1600 EST), Control Room operators verified that the plant was currently in compliance with these surveillance requirements and that the AFD monitor was, and had been, operable. No immediate corrective action was needed.

G. SAFETY SYSTEM RESPONSES:

None III. CAUSE OF EVENT:

A. IMMEDIATE CAUSE:

The immediate cause of the condition prohibited by Technical Specifications was a missed surveillance in that SR 3.2.3.1 was not performed within 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> after 2318 EST on February 26.

r t

O NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (6.1999)

LICENSEE EVENT REPORT (LERJ TEXT CONTINUATION FACILITY NAME (1) OOCKET (2) LER NUMBER (6) PAGE (3)

REVISI YEAR SEQUENTIAL ON NUMBER NUMBE R.E. Ginna Nuclear Power Plant 05000244 - - 4 5 1999 002 00 OF TEXT filmore speco is recurred, uso eddidonel copies of NRC Form 366AJ 1171 B. INTERMEDIATE CAUSE:

The intermediate cause of the missed surveillance was inadvertently not performing SR 3.2.3.1 at approximately 0728 EST on February 27.

ROOT CAUSE:

The underlying cause of the missed surveillance was a personnel error by an RG(tIE licensed operator, who performed several separate tasks associated with some routine ITS SRs that are performed using the PPCS, and inadvertently did not perform ITS SR 3.2:3.1 at that time. This error was a cognitive error in that the licensed operator did not recognize or detect that he had failed to perform SR 3.2.3.1 at 0728 EST. Inadvertently not performing SR 3.2.3.1 was contrary to approved procedures, in that the procedure required the SR be performed every eight hours. There were no unusual characteristics in the Control Room that directly contributed to the error.

IV. ANALYSIS OF EVENT:

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

(B), which requires a report of, "Any operation or condition prohibited by the plant's Technical Specifications". The missed surveillance is a condition prohibited by the plant's techncial specifications.

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 missed surveillance because:

The AFD is a measure of axial power distribution skewing to the top or bottom half of the core. The AFD is. defined as the difference in normalized flux signals between the top and bottom halves of a two section excore neutron detector in each detector well. For convenience, this flux difference is converted to provide flux difference units. The allowed range of the AFD is used in the nuclear design process to help ensure that core peaking factors and axial power distributions meet safety analysis requirements.

SR 3.2.3.1 is the verification that the AFD monitor is operable. This is normally accomplished by iritroducing a signal into the PPCS to verify control room annunciation of AFD not within the target band. The Frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is sufficient to ensure operability of the AFD monitor, since under normal plant operation the AFD is not expected to significantly change.

SR 3.2.3.1 had been properly performed at all times prior to this event, and was properly performed at approximately 1539 EST and 2326 EST on February 27, confirming that the AFD monitor was operable.

The frequency of monitoring the AFD by the PPCS is nominally once per minute. This monitoring continued throughout this event.

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION 16 19)8)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION

~ FACILITY NAME I1) DOCKET I2) LER NUMBER I6) PAGE I3)

REVISI YEAR SEOUENTIAL ON NUMBER NUMBE R.E. Ginna Nuclear Power Plant 05000244 - 5 OF 5 1999 002 00 TEXT illmore speceis required, use eddaionel copies ol ltiRC Form J66AI 117) o With thermal power (90% of reactor thermal power (RTP), the AFD may be outside the target'band provided that the deviation time is restricted. It is intended that the plant is operated with the AFD within the target band about the target flux difference.

o Inoperability of the alarm does not necessarily prevent the actual AFD values from being available (e.g., from the computer logs or hand logs). AFD values for the preceding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> can be obtained from the hourly PPCS printouts or hand logs. If the AFD monitor had been inoperable during operation at (90% RTP, the AFD measurement is monitored at a Surveillance Frequency of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to ensure that the AFD is within its limits.

Based on the above, it can be concluded 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:

Since the AFD monitor was operable at the time of discovery, no immediate corrective actions were needed.

B. ACTION TAKEN OR PLANNED TO PREVENT RECURRENCE:

o Procedure 0-6.13 will be evaluated for the enhancement of documentation of the completion of ITS SRs.

o Lessons learned from this event were discussed with the licensed operator who inadvertently did not perform SR 3.2.3.1. Emphasis was placed on attention to detail.

o Operations supervision will review this LER and corrective actions with all operating shifts.

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 at Ginna Station could be identified.

C. SPECIAL COMMENTS:

None