ML18101A185

LER 94-002-04:on 940119,RCS Flow Calculations Indicated That Either RCS Flow Was Low or That Unit May Have Operated Above 4311 Megawatts.Caused by Personnel Error.Steam & Feedwater Flow Circuitry Was updated.W/940812 Ltr

ML18101A185
Person / Time
Site:	Salem
Issue date:	08/12/1994
From:	Hagan J, Wiltsee F Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-94-002, LER-94-2, NUDOCS 9408230078
Download: ML18101A185 (8)

Text

Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 Salem Generating Station

. ' August 12, 1_994 U. s. Nuclear Regulatory Commission Document Control Desk

-Washington, DC 20555

Dear Sir:

SALEM GENERATING STATION LICENSE NO. DPR-75 DOCKET NO. *50-311 UNIT NO. 2 SUPPLEMENTAL LICENSEE EVENT REPORT 94-002-04 This supplemental Licensee Event Report is being submitted pursuant to Code of Federal Regulations 10CFR 50.73. It provides additional corrective action as well as the results of further investigation and testing.

• Sincerely yours, MJPJ:pc Distribution

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/(f-?J' i vY'"" . I 95*2189 REV 7*92

NRC FORM 366 .S. NUCLEAR REGULATORY COMMISSION *APPROVED BY OMB NO. 3150-0104 (5-92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS

.INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD LICENSEE EVENT REPORT (LER) COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF (See reverse for required number of digits/characters for each block) MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (1) DOCKET NUMBER (2) II PAGE (3)

~ .. 1°"' l"'an<>i-<1tini;i: Station - Unit 2 05000 311 111 OF 06 TITLE( l 4 Reactor Power Higher Than Indicated And Subsequent Failure To Enter Technical

~n<>,.if'i,.aHnn ~ 1 Due To Inonerable Nuclear Instrumentation.

EVENT DATE (5\ LER NUMBER 16 REPORT NUMBER (7) OTHER FACILITIES INVOLVED (8)

FACILITY NAME DOCKET NUMBER SEQUENTIAL REVISION.

MONTH DAY YEAR YEAR MONTH DAY YEAR NUMBER NUMBER 05000 FACILITY NAME DOCKET NUMBER fll 94 . 002 04 . OB 12 94 05000 OPERATING THIS REPORT IS SUBMITIED PURSUANT TO THE REQUIREMENTS OF 10 CFR !i: (Check one or more 1111 MODE (9) 1 20.402(b) 20.405(c) 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).

LEVEL (10) 100 20.405(a)(1)(ii) 50.36(c)(2) 50.73(a)(2)(vii) OTHER 20.405{a)(1)(iii) X 50.73(a)(2)(i) 50.73(a)(2)(viii)(A) (Specify in Abstract 11--+----"....;..:_..;..;......:..__ _ _ _ _t-=-+--~,...,...;..,...,,....-----+--+-....,..--.,....,..,...,...,....,~~-___, below and in Te><t, NRC 20.405(a)(1)(iv) 50.73(a)(2)(ii) 50.73(a)(2)(viii)(B) Form 366A) 20.405(a)(1)(v) 50.73(a)(2)(iii) 50.73(a)(2)(x)

LICENSEE CONTACT FOR THIS LER. (12)

- NAME TELEPHONE NUMBER (Include Area Code) l<'_ H- Wilt.,,ee - LER Coordinator (609) 339-5163 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (131 REPORTABLE REPORTABLE CAUSE SYSTEM COMPONENT MANUFACTURER GAUSE SYSTEM COMPONENT MANUFACTURER TONPRDS TO NPRDS SUPPLEMENTAL REPORT EXPECTED (14) EXPECTED MONTH DAY YEAR YES SUBMISSION X

I (If yes, complete EXPECTED SUBMISSION DATE)

NO DATE (15) 02 28 95 ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16)

On 1/19/94, review of Unit 2 Fuel Cycle 8 c.alorimetric and Reactor Coolant System (RCS) flow calculations indicated the Unit may have operated >:3411 megawatts (thermal) due to reactor thermal power >

indicated. Power was reduced by 3% to compensate for an estimated 2.5% error in indicated power. Technical Specification 3.0.3 was not entered on 1/19/94 when Nuclear Instrumentation (NI) power range was inoperable. The NI was readjusted on 1/21/94. INITIAL Data showed a potential indication error ranging from 2*. 5% to 4. 6%. SUBSEQUENT engineering evaluation has determined that the plant operated 1.4% >

rated thermal power during cycle 7 and 2.58% > rated thermal power during Cycle 8. Based upon completed evaluations/analyses, the safety of Unit 2 was not compromised. Existing overtemperature*delta temperature *(OTDT) and overpower delta temperature (OPDT) setpoints provided adequate margin with manual rod control and all rods fully withdrawn. New setpoints have been established and the OTDT, OPDT, steam flow, and feedwater (FW) flow circuitry have been revised for full power operation and automatic rod control. The failure to readjust the NI on 1/19/94 will be covered in Licensed Operator Requalification Training. A supplemental report detailing the cause of the feedwater flow indication error will be submitted after the completion of 2R8, presently scheduled to begin in October, 1994.

NRC FORM 366 (5-92)

REQUIRED NUMBER OF DIGITS/CHARACTERS FOR EACH BLOCK BLOCK NUMBER OF TITLE NUMBER DIGITS/CHARACTERS 1 UP TO 46 FACILITY NAME 8 TOTAL 2 DOCKET NUMBER 3 IN ADDITION TO 05000 3 VARIES PAGE NUMBER 4 UP TO 76 TITLE 6 TOTAL 5 EVENT OATE 2 PER BLOCK 7 TOTAL 2 FOR YEAR 6 LER NUMBER 3 FOR SEQUENTIAL NUMBER -

2 FOR REVISION NUMBER 6 TOTAL 7 REPORT DATE 2 PER BLOCK UP TO 18 - FACILITY NAME 8 OTHER FACILITIES INVOLVED 8 TOTAL -- DOCKET NUMBER 3 IN ADDITION TO 05000 9 1 OPERATING MODE 10 3 POWER LEVEL 1

11 REQUIREMENTS OF 10 CFR CHECK BOX THAT APPLIES UP TO 50 FOR NAME 12 LICENSEE CONTACT 14 FOR TELEPHONE CAUSE VARIES 2 FOR SYSTEM 13 4 FOR COMPONENT EACH COMPONENT FAILURE 4 FOR MANUFACTURER NPRDS VARIES 1

14 SUPPLEMENTAL REPORT EXPECTED CHECK BOX THAT APPLIES 6 TOTAL 15 EXPECTED SUBMISSION DATE 2 PER BLOCK

Salem Generating station DOCKET NUMBER LICENSEE EVENT REPORT (LER) TEXT CONTINUATION LER NUMBER PAGE Unit 2 5000311 94-002-04 2 of 6 PLANT AND SYSTEM IDENTIFICATION:

Westinghouse - Pressurized Water Reactor Energy Industry Identification System (EIIS) codes are identified in the text as {xx}

IDENTIFICATION OF OCCURRENCE:

Reactor Power Higher Than I:r:idicated And subsequent Failure To Enter Technical Specification 3~0.3 Due To Inoperable Nuclear Instrumentation

• Event Date: 1/19/94 Prior submittal Date: 6/29/94 Supplement Report Date: 8/12/94 This report was initiated by Incident Report Nos.94-027 and 94-0-77.

CONDITIONS PRIOR TO OCCURRENCE:

Mode 1 Reactor Power 100% - Unit Load 1180 MWe DESCRIPTION OF OCCURRENCE:

On January 19, 1994, review of Unit 2 Fuel Cycle 8 calorimetric and Reactor Coolant System (RCS) {AB} flow calculations indicated that either RCS flow was low or that the Unit may have operated above the 3411 megawatts (thermal), specified in Operating License Condition 2.C.(1). Power was reduced by 3% to conservatively compensate for an estimated 2.5% error in indicated power.

Data from a single feedwater {SJ} flow tracer test on February 3, 1994, showed a potential indication error as high as 4.6%. To avoid exceeding 100% reactor power, administrative controls were implemented to limit Reactor thermal power to 95% by calorimetric. In addition, nuclear instrumentation (NI) {JC} was adjusted due to the identified error. Existing overtemperature delta temperature (OTDT) and overpower delta temperature (OPDT) setpoints provided adequate margin, as long as rod control was maintained in manual with all rods not fully withdrawn. The Unit was maintained in manual rod control when all rods were not fully withdrawn until new setpoints for OTDT and OPDT could be established.

New OTDT and OPDT setpoints have been established and on March 13, 1994, the OTDT and OPDT circuitry was updated to reflect revised full power operating conditions and rod control was th*en returned to.automatic. In addition, the steam and feedwater flow circuitry have been updated to reflect the revised full power

LICENSEE EVENT REPORT (LEE) TEXT CONTINUATION Salem Generating station DOCKET NUMBER LER NUMBER PAGE Unit 2 5000311 94-002-04. 3 of 6 DESCRIPTION OF OCCURRENCE: (cont'd) operating conditions. On March 22, 1994, the feedwater flow nozzle constants in the calorimetric calculation procedure and in the on

• line calorimetric computer were increased, which effectively derates the Unit by 5% rated thermal power.

The Nuclear Regulatory Commission* (NRC) was notified of the potential overpower event pursuant to 10CFR50.72(b) (1) (ii) (B).

On March 3, 1994, subsequent review determined that the NI should have been readjusted on January 19, 1994,.following identification of the potential overpower condition. As such, the NI power range was inoperable until the NI was readjusted on January 21, 1994, and a failure to enter Technical Specification 3.0.3 occurred.

Subsequent engineering evaluation has determined that the plant operated 1.4% greater than rated thermal power during Cycle 7, from May 1992 through March 1993, and 2.58% greater than rated thermal power during Cycle 8, from July 1993 through January 1994.

ANALYSIS OF OCCURRENCE:

Nuclear instrumentation trip setpoints ensure that safety limits for the reactor core and reactor coolant system are not exceeded during normal operation and design basis anticipated operational

• occurrences.

Review of Fuel Cycle 8 calorimetric and Reactor Coolant System flow calculations, show the Unit's Operating License Condition maximum Reactor power level of 3411 megaw~tts (thermal) may have been exceeded. Initial assessment determined this event resulted from .a potential error of 2.5% in actual Reactor thermal power higher than shown by NI. Data from a single feedwater flow tracer test showed a potential indication error as high as 4.6%.

To avoid exceeding 100% reactor power, administrative controls were implemented to limit Reactor thermal power to 95% by calorimetric.

In addition, the NI was adjusted for the. indicated error. Evaluation*

of the OTDT.and OPDT setpoints showed adequate margin for the existing installed values, provided that no uncontrolled rod withdraw events occurred. Correspondingly, the Unit was maintained in manual

• rod control when all rods were not fully withdrawn to prevent uncontrolled rod withdraw events.

Subsequent engineering evaluation shows the. plant operated 1.4%

greater than rated thermal power during Cycle 7 and 2.58% greater than rated thermal power during Cycle 8.

New OTDT and OPDT setpoints have been established and the appropriate circuitry has been updated to reflect revised full power operating

Salem Generating Station DOCKET NUMBER LICENSEE EVENT REPORT (LER) TEXT CONTINUATION LER NUMBER PAGE Unit 2 5000311 94-002-04 4 of 6 ANALYSIS OF OCCURRENCE: (cont'd) conditions, and rod control has been returned to automatic~ In addition, steam and feedwater flow circuitry have been updated to reflect the revised full power operating conditions. Feedwater flow nozzle constants in both the calorimetric calculation procedure and the on line calorimetric computer were increased, which ef:fectively derates the Unit by 5% rated thermal power.

Subsequent analysis determined the NI should .have been adjusted following the conservative 3% reduction.in reactor power to eliminate the possibility of operating the Unit above its licensed rated thermal power. Therefore, the NI power range was inoperable until the NI was readjusted and a failure to enter TS 3.0.3 occurred.

APPARENT CAUSE OF OCCURRENCE:

The cause of the feedwater flow indication error is presently under investigation.

The failure to readjust the NI on January 19, 1994 occurred due to personnel error by Operations personnel and was a direct consequence of the immediate concern and focus to operate the Unit within its licensed rated thermal power.

PRIOR SIMILAR OCCURRENCES:

A review of documentation did not show any prior similar occurrence of this event.

SAFETY SIGNIFICANCE:

This event is reportable pursuant to 10CFR50.73(a) (2) (i) (B) due the inoperability of the nuclear instrumentation as a result of the event and the subsequent failure to enter TS 3.0.3.

Initial safety assessment by Westinghouse, of the potential effect of operating Salem Unit 2 at 104.5% power, showed no adverse consequence for Loss of Cooling Accidents (LOCAs). This determination was made because depending on the analysis involved, either power level is not an*initial condition in the analyses or there is sufficient margin in the analyses to mitigate the effects of the event. Similarly, no adverse consequences are shown for the LOCA Containment analysis. A Salem specific analysis, based on full power operation at 3600 MWT (WCAP 13131), has not been reviewed by the NRC and as such, is not part of the Salem licensing basis. However, the evaluation model used for the long-term LOCA mass and energy release calculations was documented in WCAP 10325 for generic application. This model has been reviewed and approved by the NRC and has been used in the analysis of other plants.

Salem Generating station

• DOCKET NUMBER LICENSEE EVENT REPORT (LER) TEXT CONTINUATION LER NUMBER. PAGE Unit 2 5000311 .94-002-04 5 of 6 SAFETY SIGNIFICANCE: (cont'd)

Subsequent Westinghouse analysis has been performed which examined

. potential effects of having operated Unit 2 at power levels up to 104.5% rated power. This analysis, documented in NFSI-94-201 addressed each licensing basis LOCA a.nd rion-LOCA event and the impact of the overpower operation upon each event. For all LOCA and some non~LOCA *events, engineering evaluation confirmed that no significant safety concern existed. This is because either the licensing

" analysis was unaffected by the overpower operation or that more than sufficient margin already existed to offset adverse consequences*

associated with overpower operation. For the remaining non-LOCA

• ' events, there was insufficient margin or sensitivities to assess the impact of overpower operation or to reach a conclusion without additional detailed analyses. Therefore, further analyses were performed to address these.events. Based upon the completed evaluations and results from the analyses, the safety of Unit 2 was not compromised .

.': CORRECTIVE ACTION:

Administrative controls were implemented to limit Reactor thermal power to 95% of rated thermal power by calorimetric and rtucle.ar instrumentation was adjusted due to the identified error. The Unit was maintained in manual rod control when all rods were not fully withdrawn. This was done to prevent uncontrolled rod withdraw events until new setpoints for OTDT and OPDT were established, to reflect

~evised.full power operating conditions.

The OTDT and OPDT circuitry was updated to reflect the revised full power operating conditions and rod control was returned to automatic. The steam and feedwater flow circuitry were also updated to reflect the revised full power operating conditions. The feedwater flow nozzle constants in the calorimetric calculation procedure and the *On line calorimetric computer were increased by 5%

to effectively derate the Unit by 5% rated thermal power, which removed the need for administrative controls on reactor power.

Ultrasonic flow measurement devices have been installed on all four FW headers and a test was conducted to determine the actual FW flow.

The results of the ultrasonic feedwater flow test have been incorporated into an engineering evaluation, which determined that the plant operated 1.4% greater thari rated thermal power during Cycle 7, from May 1992 through March 1993, and 2.58% greater than rated thermal power during Cycle 8, from July 1993 through January 1994.

The accuracy of the installed flow nozzles will be periodically

. assessed using the installed ultrasonic flow meters in conjunction with reviewing changes to plant parameters. Removal, replacement,

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station DOCKET NUMBER LER NUMBER PAGE Unit 2 5000311 94-002-04 6 of 6 CORRECTIVE ACTION: (cont'd) and inspection of the nozzles will be performed during refueling outage 2R8.

The failure to readjust the NI on January 19, 1994, following the reactor power reduction, will be covered in Licensed Operator Requalification Training for 1994 - 1995.

A supplemental report detailing the cause of the f eedwater flow indication error will be submitted after the completion of 2R8, presently scheduled to begin in October, 1994 .

. MJPJ:pc SORC Mtg.94-063