ML18100B170

LER 94-002-03:on 940119,RCS Flow Calculations Indicated Unit May Have Operated at 3411 Megawatts Due to Personnel Error. Ultrasonic Flow Measurement Devices Have Been Installed on All Four FW headers.W/940620 Ltr

ML18100B170
Person / Time
Site:	Salem
Issue date:	06/20/1994
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 9407050084
Download: ML18100B170 (8)

Text

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• Public. Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 Salem Generating Station June 29, 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 . I UNIT NO. 2 SUPPLEMENTAL LICENSEE EVENT REPORT 94-002-03 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, n

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\_ G *J 'v ..i** J 9407050084 940620 PDR ADOCK 05000311 S PDR The power is in your hands.

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 REGULATOR.Y COMMISSION, 'NASHINGTt;>N, 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) PAGE (3)

Salem Generating Station - Unit 2 05000 311 1 OF 06 TITLE (4) Reactor Power Higher Than Indicated And Subsequent Fa.ilm::e To Enter TE;!chnical Specification 3.0.3 Due To Inoperable Nuclear Instrumentation*.

EVENT DATE (5) LEA NUMBER (6 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 01 19 94 94 002 03. 06 29 94 050.00 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 2b.405(a) (1) (iii) X. so. 13{a)(2) (il *50. 73 (a) (2) (viii) (A) (Specify in Abstract below and in Tex1, NRG 20.405(a) (1) (iv) 50.73(a) (2) (ii) 50. 73 (a) (2) (viii) (B) Form 366A)

I . 20.405(a)(1 )(v) 50.73(a)(2)(iii)

LICENSEE CONTACT FOR THIS LEA.. 12) 50.73(a)(2)(x)

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)

.. MONTH DAY YEAR EXPECTED X

I YES (If yes, complete .EXPECTED SUBMISSION DATE)

NO SUBMISSIQN DATE (15) 08 15 94 ABSTRACT ~imit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16) *. . .

On /19/94, review of Unit 2 Fuel Cycle 8 calorimetric and Reactor Coolant System (RCS) flow*ca:lculations 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 specificationi3~0.3*was hot entered on 1/19/94 when Nuclear Instrumentation (NI) power range was inoperable. The NI was readjusted on 1/21/94. Data showed a potential indication error ranging from 2.5% to as high as 4.6%.

Based upon completed evaluations and results from 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 Requalif ication Training for 1994 - 1995. Testing results will be incorporated into an engineering evaluation which will document the actual FW flow and reactor power. It is anticipated that a supplement to this report will be submitted by 8/15/94 to detail results of further event investigation/testing.

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 VARIES ... PAGE NUMBER 4 UP.TO 76 __ TITLE.

6 TOTAL 5 EVENT DATE 2 PER BLOCK .

7 TOTAL 2 FOR YEAR 6

• l,Efl NUMBER 3 FOR SEQUENTIAL NUMBER 2 FOR REVISION NUMBER

• • '* 6 TOTAL 7 REPORT DATE 2 PER BLOCK UP TO 18 -- FACILITY NAME 8  : OTH.ER 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 APP.LIES UP TO 50 FOR NAME 12 LICENSEE CONTACT 14 FORTELEPHONE 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 L

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station DOCKET.NUMBER LER NUMBER PAGE Unit 2 5000311 94-002-03 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 Indicated And Subsequent Failure To Enter Technical Specification 3*.0.3 Due To Inoperable Nuclear Instrumentation Event Date: 1/19/94 Prior Submittal Date: 3/30/94 Supplement Report Date: 6/29/94 This report was initiated by Incident Report Nos.94-027 and 94-077.

CONDITIONS PRIOR TO OCCURRENCE:

Mode 1 Reactor Power 100% - Unit Load 1180 MWe DESCRIPTlON 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

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station DOCKET NUMBER LER NUMBER PAGE Unit 2 5000311 94-002-03 3 of 6 DESCRIPTION OF OCCURRENCE: (cont'd) revised full power operating conditions and rod control was then returned to automatic. In addition, the steam flow and feedwater flow circuitry have been updated to reflect the revised full power operating conditions. On March 22, 1994, the feedwater flow nozzle flow 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 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.

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 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 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.

New OTDT and OPDT setpoints have been established and the appropriate circuitry has been updated to reflect revised full power operating conditions, and rod control has been returned to automatic. In addition, steam flow and feedwater flow circuitry have been updated to reflect the revised full power operating conditions. Feedwater flow

LICENSEE. EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station DOCKET NUMBER LER NUMBER PAGE Unit 2 5000311 94-002-03 4 of 6 ANALYSIS OF OCCURRENCE: (cont'd) nozzle flow constants in both the calorimetric calculation procedure and the on line calorimetric computer were increased, which effectively 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 applicatibn. This model has been reviewed and approved by the NRC and has been used in the analysis of other plants.

• LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating station DOCKET NUMBER LER NUMBER PAGE Unit 2 5000311 94-002-03 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 and non-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 nuclear 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 revised 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 nozzle flow 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.

Testing has been conducted at Alden Laboratories to determine the flow profiles expected in the Salem FW piping configuration. A preliminary report has been prepared to document the results of the FW flow test using the ultrasonic flow devices. This report is being finalized and will be incorporated into an engineering evaluation which will document the actual FW flow, as well as reactor thermal power.

The failure to readjust the NI on January 19, 1994, following the

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station DOCKET NUMBER LER NUMBER PAGE Unit 2 5000311 94-002-03 6 of 6 CORRECTIVE ACTION: (cont'd) reactor power reduction, will be covered in Licensed Operator Requalification Training for 1994 - 1995.

It is anticipated that a supplement to this report will be submitted by August 15, 1994 to detail results of further event investigation/testing.

General ager -

Salem Operations MJPJ:pc SORC Mtg.94-050