ML17264B046

LER 97-003-01:on 970730,bistable Instrument Trip Setpoint Could Have Exceeded Allowable Value.Caused by Insufficient Existing Margin Between Trip Setpoint & Allowable Value. Held Switches in Tripped configuration.W/970929 Ltr

ML17264B046
Person / Time
Site:	Ginna
Issue date:	09/29/1997
From:	Martin J, Mecredy R ROCHESTER GAS & ELECTRIC CORP.
To:	Vissing G NRC (Affiliation Not Assigned), NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-97-003, LER-97-3, NUDOCS 9710030201
Download: ML17264B046 (12)

Text

CATEGORY 1~

REGULAT Y INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9710030201 DOC.DATE: 97/09/29 NOTARIZED: NO DOCKET FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G 05000244 AUTH. NAME AUTHOR AFFILIATION MARTIA I J. T.'. Rochester Gas a Electric Corp.

MECREDY,R.C. Rochester Gas 6 Electric Corp.

RECIP.NAME RECIPIENT AFFILIATION VISSING,G.S.

SUBJECT:

LER 97-003-01:on 970730,identified that high steam flow bistable instrument trip setpoint plus instrument unceitainty counld exceeded. Caused by condition prohibited by Tech Specs. Bistable proving switches placed.W/970929 ltr.

DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR ENCL SIZE-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). 0500n24~ G RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD1-1 PD 1 1 VISSING,G. 1 1 INTERNAL: AEQB/BTBPtAB.. 2 2 AEOD/SPD/RRAB 1 1 FILE CEN 1 1 NRR/DE/ECGB 1 1 NRR= E EELB 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 NRR/DSSA/SRXB 1 1 RES/DET/EIB 1 1 D RGN1 FILE 01 1 1 EXTERNAL: L ST LOBBY WARD 1 1 LITCO BRYCE,J H 1 1 NOAC POOREEW. 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 WASTE. 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 24 ENCL 24

AND ROCHESTER GAS AND EIECTRIC CORPORATION ~ 89 EAST AVENUE, ROCHESTER, N Y,'4649 KOI D AREA COGENT I6 546 W'N Sos-Rr c .v:cR-DY

't"ce c ence >

cc ec' 0< I c" s September 29, 1997 U.S. Nuclear Regulatory Commission Document Control Desk Attn: Guy S. Vissing Project Directorate I-1 Washington, D.C. 20555

Subject:

LER 97-003, Revision 1, Bistable Instrument Trip Setpoint (plus Instrument Uncertainty) Could Exceed Allowable Value, Causes a Condition Prohibited by Plant 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 97-003 (Revision 1) is hereby submitted. A further assessment is provided in this supplement, as committed to in the original LER.

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

Very truly yours, Robert C. Mecredy xc: Mr. Guy S. Vissing (Mail Stop 14B2)

PWR Project Directorate I-1 Washington, D.C. 20555 U.S. Nuclear Regulatory Commission

~IN D3 C3 Region I 475 Allendale Road King of Prussia, PA 19406 Ginna Senior Resident Inspector .. RI!liI!I!Eii!iLLi!II!I~

97f003020i 970929 05000244 PDR ADQCK S PDR

NRC FORM 366 U.S. N LEAR REGULATORY COMMISSIO APPROVED BY OMB NO. 3150-0104 (4.95) EXPIRES 04/30/98 ESTIMATED BURDEN PER RESPONSE To COMPLY WITH THIS MANDATORY INFORMATION COLLECTION REQUEST: 50.0 HRS.

LICENSEE EVENT REPORT (LER) REPORTED LESSONS LEARNED ARE INCORPORATED INTO THE LICENSING 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 F331, digits/characters for each block) U.s. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555.000), AND To THE PAPERWORK REDUCTION PROJECT FACILITYNAME (1) DOCKETNUMBBl(2) FAD B (3)

R.E. Ginna Nuclear Power Plant 05000244 1OF6 TrrLE (4)

Bistable Instrument Trip Setpoint (plus Instrument Uncertainty) Could Exceed Allowable Value, Causes a Condition Prohibited by Plant Technical Specifications EVENT DATE (5) LER NUMBER (61 REPORT DATE (7) OTHER FACILITIES INVOLVED (81 FACILITY NAME DOCKET NUMBER SEQUENTIAL REVISION MONTH OAY NUMBER NUMBER MONTH OAY YEAR FACILITY NAME DOCKET NUMBER 07 30, 97 97 003 01 09 29 97 OPERATING THIS REPORT IS SUBMITTED PUR SUANTTO THE REQUIREMENTS OF 10 CFR E: (Check one or more) (11)

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

POWER 20.2203(a) (11 20.2203(a)(3)(i) 50.73(a) (2) (ii) 50.73(a)(2)(x)

LEVEL (10) 100 20.2203(a)(2)(i) 20.2203(a)(3)(ii) 50.73(a) (2) (iii) 73.71 20.2203(a)(2) (ii) 20.2203(a)(4) 50.73(a) (2) (iv) OTHER 20.2203(a) (2) (iii) 50.36(c) (1) 50.73(a)(2)(v) Specify in Abstract below or in NRC Form 366A 20.2203(a)(2) (iv) 50.36(c)(2) 50.73(a)(2)(vii)

LICENSEE CONTACT FOR THIS LER (12)

TELEPHONE NUMBER (Iicivde Arae Code)

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

REPORTABLE REPORTABLE CAUSE SYSTEM COMPONENT MANUFACTURER To NPRDS CAUSE SYSTEM COMPONENT MANUFACTURER To NPRDS SUPPLEMENTAL REPORT EXPECTED (14) EXPECTED MONTH DAY YEAR YES SUBMISSION (If yes, complete EXPECTED SUBMISSION DATE). X NO DATE (15)

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

On July 30, 1997, at approximately 1430 EDST, the plant was in Mode 1 at approximately 100% steady state reactor power. It was identified that the High Steam Flow Bistable Instrument Trip Setpoint plus the Instrument Uncertainty could exceed the "Allowable Value" listed in the Ginna Station Improved Technical Specifications.

Immediate corrective action was to declare all four channels of High Steam Flow inoperable and enter Limiting Condition for Operation 3.0.3. When action was taken to place the affected bistable proving switches in the tripped position, the plant exited Limiting Condition for Operation 3.0.3.

The cause of this condition was that the existing margin between the Trip Setpoint and the Allowable Value is insufficient to account for process and instrument uncertainties.

Corrective action to prevent recurrence is outlined in Section V.B. Additional corrective actions are identified in this supplement.

NRC FORM 366 (4.951

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)

SEQUENTIAL REVISION NUMBER NUMBER 2 OF 6 R.E. Ginna Nuclear Power Plant 05000244 97 003 01 TEXT Ilfmore space is required, use additional copies of fiRC Form 366AI (17)

PRE-EVENT PLANT CONDITIONS:

Personnel from Nuclear Engineering Services (NES) had been evaluating the impact of instrument uncertainties, including validation of actual plant-specific data. On July 30, 1997, at approximately 1430 EDST, the plant was in Mode 1 at approximately 100% steady state reactor power. In activities unrelated to plant conditions, NES personnel identified that the High Steam Flow Bistable Instrument Trip Setpoint, when added to the Instrument Uncertainty, could exceed the "Allowable Value" listed in the Ginna Station Improved Technical Specifications (ITS).

II. DESCRIPTION OF EVENT:

A. DATES AND APPROXIMATE TIMES OF MAJOR OCCURRENCES:

July 30, 1997, 1430 EDST: Event date and time.

July 30, 1997, 1430 EDST: Discovery date and time.

July 30, 1997, 1527 EDST: High Steam Flow Bistable proving switches are placed in the tripped position, thus restoring the channels to operable status.

B. EVENT:

On July 30, 1997, at approximately 1430 EDST, the plant was in Mode 1 at approximately 100%

steady state reactor power. In activities unrelated to plant conditions, NES personnel reported to Operations supervision that the current High Steam Flow Bistable Instrument Trip Setpoint, when added to the Instrument Uncertainty, could exceed the Allowable Value for this Engineered Safety Feature Actuation System (ESFAS) Instrumentation, as listed in ITS Table 3.3.2-1, Function 4.d.

Since the specified Trip Setpoint pIus the uncertainty could be nonconservative with respect to the Allowable Value, all four channels of High Steam Flow were declared inoperable at approximately 1438 EDST. There is no ITS Limiting Condition for Operation (LCO) to address the condition of four channels inoperable, so ITS LCO 3.0.3 was entered.

The Plant Operations Review Committee (PORC) reviewed this condition. The High Steam Flow Trip Setpoint is normally met at approximately 10% reactor power (increasing), since steam flow per steam generator (SG) at full power is approximately 3.3E6 Ibm/hr. Thus the bistables are automatically tripped when above 10% power. PORC determined that it was acceptable to place the bistable proving switches for the affected bistables in the tripped position.

NRC FORM 366A (4-95)

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

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1) DOCKET LER NUMBER I6) PAGE I3)

YEAR SEQUENTIAL REVISION NUMBER NUMBER R.E. Ginna Nuclear Power Plant 3 OF 6 05000244 97 003 01 TEXT /lfmore space is required, use additional copies of NRC Form 366AI I17)

The Control Room operators placed the affected bistable proving switches in the tripped position.

This action ensured that any variation in the Trip Setpoint would not affect the completion of affected functions for High Steam Flow. Since the four affected channels of High Steam Flow were now performing their intended function by being maintained in the tripped configuration and all ITS required surveillances were met, these four channels were declared operable at approximately 1527 EDST on July 30, 1997.

C. INOPERABLE STRUCTURES, COMPONENTS, OR SYSTEMS THAT CONTRIBUTED TO THE EVENT:

None D. OTHER SYSTEMS OR SECONDAR< FUNCTIONS AFFECTED:

None E. METHOD OF DISCOVERY:

This condition was self-identified by an NES engineer while performing validation of calculated instrument uncertainties.

F. OPERATOR ACTION:

When notified of the results of the impact of the instrument uncertainty on operability of the High Steam Flow Bistables, Operations supervision notified the Shift Supervisor of this condition.

Operations supervision also notified the NRC Resident Inspector.

Following notification, the Control Room operators declared all affected functions provided by the four channels of High Steam Flow inoperable and entered ITS LCO 3.0.3. After the affected bistable proving switches were placed in the tripped position and holds placed on these switches, they returned the four channels to operable status and exited ITS LCO 3.0.3.

G. SAFETY SYSTEM RESPONSES:

None

~

NRC FORM 366A I4-95)

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

LZCENSEE EVENT REPORT (LER)

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

YEAR SEQUENTIAL REVISION NUMBER NUMBER 4 OF 6 R.E. Ginna Nuclear Power Plant 05000244 97 003 01 TEXT llfmore spaceis required, use additional copies of NRC Form 366AJ (17)

III. CAUSE OF EVENT:

A. IMMEDIATECAUSE:

The immediate cause of the condition prohibited by Technical Specifications was entry into ITS LCO 3.0.3 due to four channels of High Steam Flow being declared inoperable.

B. INTERMEDIATE CAUSE:

The. intermediate cause of all four channels of High Steam Flow being declared inoperable was a potential for instrument uncertainty to allow the Trip Setpoint to exceed the Allowable Value as listed in ITS Table 3.3.2-1 for ESFAS Instrumentation.

C. ROOT CAUSE:

The underlying cause of the potential to exceed the Allowable Value was that the existing margin between the High Steam Flow Trip Setpoint ar)d the ITS Allowable Value is insufficient to account for process and instrument uncertainties 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." Having all four channels of High Steam Flow inoperable is a condition prohibited by the Ginna Station Improved Technical Specifications.

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

o The current Trip Setpoint and Allowable Value parameters for High Steam Flow had not changed since September 1969. However, the power rating for Ginna Station was uprated from 1300 MWt to the current 1520 MWt in the early 1970s. There was no change to the High Steam Flow setpoints, even though other main steam isolation functions (e.g., the HIGH-HIGH Steam Flow Trip Setpoint and Allowable Value) were changed at that time. The affected High Steam Flow setpoints should also have been changed at that time.

NRC FORM 366A (4-95)

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

LICENSEE EVENT REPORT (LER)

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

YEAR SEQUENTIAL REVISION NUMBER NUMBER 5 OF 6 R.E. Ginna Nuclear Power Plant 05000244 97 003 01 TEXT (Ifmore spaceis required, use eddi tionel copies of NRC Form 366Ai (17)

There are three coincident parameters which must be reached in order for this function to cause a main steam line isolation: high steam flow, safety injection, and low average temperature in the reactor coolant system. A License Amendment Request (LAR) has been submitted to change the ITS Allowable Value, while maintaining the same Trip Setpoint. It has been demonstrated that the change to the high steam flow parameter does not delay the time at which this isolation signal would be reached for any analyzed accident, since the steam flow value is reached much earlier in the accident scenario than the other parameters. Consequently, assuming worst case drift considerations, there was no impact on any analyzed accident.

A review of the operating history of Ginna Station did not identify any occurrence of a steam line break requiring main steam isolation.

Increases in heat removal by the secondary system are discussed in the Ginna Station Updated Final Safety Analysis Report (UFSAR), Chapter 15.1. Accidents analyzed included inadvertent opening of a Steam Generator relief/safety valve and steam system piping failures. Protection is provided by:

a. safety injection system actuation

b. reactor trip C. isolation of main feedwater lines

d. trip of MSIVs Below 15% reactor thermal power, a reactor trip would occur with smaller load changes.

However, steam line isolation is not credited for terminating uncontrolled power excursions for small steam breaks. Also, the fact that a 10% load increase could be accommodated implies that steam line isolation would not occur or be required between 15% and 100% RTP.

The Ginna Station setpoint analyses have been primarily based on using manufacturer recommended instrument drift and uncertainty values. As a result of converting to the ITS, Ginna Station changed the definition of "fuel cycle" from 18 months to 24 months, which required a verification of instrumentation historical performance data. This review indicated that the instrumentation subjected to ITS surveillances bounded (and in most cases greatly exceeded) the manufacturer recommended values.

Based on the above, it can be concluded that the public's health and safety was assured at all times.

NRC FORM 366A I4 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 6 OF 6 R.E. Ginna Nuclear Power Plant 05000244 97 003 01 TEXT llfmore spaceis required, use additional copies of NRC Form 866'A) (17)

V. CORRECTIVE ACTION:

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

The bistable proving switches were placed in the tripped configuration, and the affected functions for High Steam Flow were restored to operable status at approximately 1527 EDST on July 30, 1997.

ACTION TAKEN OR PLANNED TO PREVENT RECURRENCE:

The affected bistable proving switches are held in the tripped configuration.

The Trip,Setpoint, Instrument Uncertainty, and Allowable Value for High Steam Flow have been evaluated. The ITS Allowable Value, as stated in ITS LCO Table 3.3.2-1, Function 4.d, will be changed to reflect an appropriate and achievable setpoint.

The affected bistabie proving switches will be placed in their normal configuration after NRC approval of the License Amendment Request for the change in the Allowable Value.

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 Nuclear Power Station could be identified.

C. SPECIAL COMMENTS:

None NRC FORM 366A (4-95)