ML18100A947

LER 94-002-01:on 940119,Cycle 8 Calorimeter & RCS Flow Calculation Determined Unit Operated Above 3411 Megawatts. Caused by Feedwater Flow Indication.Controls Implemented to Limit Reactor Power to 95%.W/940309 Ltr

ML18100A947
Person / Time
Site:	Salem
Issue date:	03/09/1994
From:	Hagan J, Pastva M Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-94-002, LER-94-2, NUDOCS 9403180103
Download: ML18100A947 (6)

Text

, ' I' e OPS~G .I Public Service Electric and Gas Company P.O. Box .236 Hancocks Bridge, New Jersey 08038 Salem Generating Station March 9, 1994 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 94-002-01 This supplemental Licensee Event Report is peing submitted pursuant

• to Code of Federal Regulations 10CFR 50.73. It clarifies the previously submitted event analysis.

Sincerely yours, MJPJ:pc Distribution 9403180103 940309 PDR ADOCK 050003il S PDR The paver is in your hands.

95-2189 REV 7-92

NRC FORM 366 . U.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 REOUES:r: 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) MANAGEMEN"T: AND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (1) o.ocKET NUM,BER (2) PAGE (3)

Salem Generating Station - Unit 2 05000 311 1 OF 04 TITLE (4)

Reactor Power Higher Than Indicated.

EVENT DATE Isl LER NUMBER (6 REPORT NUMBER (7) OTHER FACILITIES INVOLVED (8 FACILITY NAME DOCKET NUMBER SEQUENTIAL REVISION MONTH DAY YEAR YEAR MONTH DAY YEAR NUMBER NUMBER 'I . 05000 FACILITY NAM~ DOCKET NUMBER 01 19 94 94 -- 002

-- 01 03 09 94 ' 05000 OPERATING. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check one or more (11 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 ii1;!lltf11ii 20.405(a)(1)(iii) x 50.73(a)(2)(i) 50.73(a) (2) (viii) (A) (Specify in Abstract below and in Teict, NRG 20.405(a)(1 )(iv)

• 50.73(a)(2)(ii) 50.73(a)(2)(viii)(B) i 1 20.405(a)(1)(v) 50.73(a)(2)(iii) 50.73(a)(2)(x)

Form 366A)

LICENSEE CONTACT FOR THIS LER 12)

NAME TELEPHONE NUMBER (Include Area Code)

M. J; Pastva, Jr. - LER Coordinator (609) 339-5165 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

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

ABSTRACT ~imit to 1400 spaces, i.e., approximately 15 sin~e-paced typewritten lines) (16)

On /19/94, review of Unit uel Cycle 8 calorimetric and Reactor Coolant System flow calculations, determined the Unit. may have operated above the 3411 megawatts (thermal), specified in Operating License Conditiori 2.C. (1). This results from Reactor thermal power being higher than indicated* by. nuclear instrumentation. Preliminary data shows a potential indication error ranging from 2.5% to as high as 4.5%, resulting from f eedwater flow being higher than indicated.

To avoid exceeding 100% reactor power, administra~ive controls have been implemented to limit power to 95% by calorimetric. Nuclear instrumentation has been adjusted due to the identified error.

Existing overtemperature delta temperature (OTDT). and overpower delta temperature (OPDT) setpoints provide adequate margin, as long as rod control is in manual when all rods are not fully withdrawn .. The Unit will be maintained in manual rod control when all' rods are not fully withdrawn until new setpoints for OTDT and OPDT are established, to reflect revised full power operating conditions. ~ The cause of the f eedwater flow indication error is under investigation. It is anticipated that on or before 3/31/94, a supplement to this report will be provided to detail results of further inv~stigation and testing and safety significance assessment of this event.

NRG 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 VARIE.S PAGE NUMBER I

4 UP TO 76 TITLE 6 TOTAL 5 EVENT DATE 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 i 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 DAT!=

2 PER BLOCK*

1*

LICENSEE EVENT REPORT (LER) TEXT CONTINU~TION Salem Generating Station DOCKET NUMBER LER NUMBER PAGE Unit 2

• 5000311 94-002-01 2 of 4 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 Indicated Event Date: 1/19/94 Supplement Report Date: 3/9/94 This report was initiated by Incident Report No.94-027.

CONDITIONS PRIOR TO OCCURRENCE:

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

Ori January 19, 1994, review of Unit 2 Fuel Cycle 8, calorimetric and Reactor Coolant System (RCS) 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. Preliminary data from a single feedwater flow tracer test on February 3, 1994 shows a potential indication error as high as 4.5%. To avoid exceeding 100%

reactor power,*administrative controls have been implemented to limit Reactor thermal power to 95% by calorimetric. In ;addition, nuclear instrumentation has been adjusted due to the identified error. -

Existing overtemperature delta temperature (OTDT) and overpower delta temperature (OPDT) setpoints provide adequate margin, as long as rod control is in manual when all rods are not fully withdrawn. The Unit will be maintained in manual rod control when all rods are not fully withdrawn until new setpoints for OTDT and OPDT are established, to reflect revised full power operating conditions.

• The NRC was notified of this event per 10CFR50.72,(b) (1) (ii) (B).

ANALYSIS OF OCCURRENCE:

Nuclear instrumentation trip setpoints ensure that s~fety limits for the reactor core and reactor coolant system are npt exceeded during normal operation and design basis anticipated operational occurrences.

Review of Fuel Cycle 8 calorimet~ic and Reactor Coolant System flow

- - - - - ---~------~

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station DOCKET NUMBER LER NUMBER PAGE Unit 2 5000311 94-002-01 3 of 4 ANALYSIS OF OCCURRENCE: (cont'd) calculations, shows the Unit's Operating License Condition maximum Reactor power level of 3411 megawatts (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 nuclear instrumentation. Preliminary data from a single f eedwater flow tracer test shows a potential indication error as high as 4.5%.

To avoid exceeding 100% reactor power, administrative controls have been implemented to limit Reactor thermal power to 95% by calorimetric. In addition, nuclear instrumentation has been adjusted for the indicated error. Evaluation of the OTDT and OPDT setpoints shows adequate margin for the existing installed values, provided there are no uncontrolled rod withdraw events. As such, the Unit will be maintained in manual rod control when all rods are not fully withdrawn. This will prevent uncontrolled rod withdraw events until new setpoints for OTDT and OPDT are established, to reflect revised full power operating conditions. -

APPARENT CAUSE OF OCCURRENCE:

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

PRIOR SIMILAR OCCURRENCES:

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

SAFETY SIGNIFICANCE:

This is reported pursuant to the requirements of 10CFR50.73(a) (2) (i) (B) due to error introduced to the nuclear instrumentation as a result of the event.

Initial safety assessment by westinghouse, of the_ potential effect of operating Salem Unit 2 *at 104. 5% power, shows 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

t LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station DOCKET NUMBER LER NUMBER PAGE Unit *2 5000311 . 94-002-01 4 of 4 I

SAFETY SIGNIFICANCE: (cont'd) used for the long-term LOC~ 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.

With regard to non-LOCA events, power level is bot:h an initial condition and a basis for the setpoints of both the Reactor Protec~ion System and Engineered Safety Feature Actuation System. Following final assessment of the feedwater flow verification test results, the.

potential impact on safety significance for RPS settings and initial condition$ for non-LOCA. events will be assessed.

CORRECTIVE ACTION:

Administrative controls have been implemented to limit Reactor thermal power to 95% by calorimetric and.nuclear instrumentation has been adjusted due to the identified error.

The Unit will be maintained in manual rod control.when all rods are not fully withdrawn. This will prevent uncontrolled rod withdraw events until new setpoints for OTDT and OPDT are established, to reflect revised full power operating conditions.

_It is anticipated that on or before March 31, 199~, a supplement to this report will be provided to detail the results of further investigation and testing and safety significance assessment of this event.

nag~r -

at ions MJPJ:pc SORC Mtg.94-021