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{{#Wiki_filter:e Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge; New Jersey 08038 Salem Generating Station U. s. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555  
{{#Wiki_filter:e OPS~G Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge; New Jersey 08038 Salem Generating Station February 17, 1994 U. s. Nuclear Regulatory Commission Document Control Desk Washington, DC                 20555


==Dear Sir:==
==Dear Sir:==
SALEM GENERATING STATION LICENSE NO. DPR-75 DOCKET NO. 50-311 UNIT NO. 2 LICENSEE EVENT REPORT 94-002-00 February 17, 1994 This Licensee Event Report is being submitted pursuant to the requirements of Code of Federal Regulation 10CFR50.73(a)
(2) (i) (B). Issuance of this report is required within thirty (30) days of event discovery.
MJPJ:pc Distribution 9402240010 940217 PDR ADOCK 05000311 S . PDR 1 he power is in your hands. Sincerely yours, gan General Manager -Salem Operations 95-2189 REV 7-92 


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 LICENSEE EVENT REPORT (LER) INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD 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) PAGE (3) Salem Generating Station -Unit 2 05000 311 1 OF 04 TITLE (4) Reactor Power Higher Than Indicated
SALEM GENERATING STATION LICENSE NO. DPR-75 DOCKET NO. 50-311 UNIT NO. 2 LICENSEE EVENT REPORT 94-002-00 This Licensee Event Report is being submitted pursuant to the requirements of Code of Federal Regulation 10CFR50.73(a) (2) (i) (B).
.. EVENT DATE (5) LEA NUMBER (6 REPORT NUMBER (7\ OTHER FACILITIES INVOLVED (8) SEQUENTIAL REVISION FACILITY NAME DOCKET NUMBER MONTH DAY YEAR' YEAR NUMBER MONTH DAY *YEAR 05000 NUMBER FACILITY NAME DOCKET NUMBER 01 19 94 94 --002 --00 02 17 05000 94 OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR &sect;: CChecl< one or more (11) MODE (9) 1 20.402(b) 20.405(c)
Issuance of this report is required within thirty (30) days of event discovery.
: 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)
Sincerely yours, gan General Manager -
OTHER I-20.405(a)(1)(iii) x 50.73(a)(2)(i) 50.73(a) (2) (viii) (A) (Specify in Abstract 20.405(a)(1)(iv) 50.73(a)(2)(ii) 50.73(a) (2) (viii) (B) below and in Text, NRC Form 366A) 20.405(a)(1)(v) 50.73(a) (2) (iii) 50. 73 (a)(2)(x)
Salem Operations MJPJ:pc Distribution 9402240010 940217 PDR ADOCK 05000311 S           .             PDR 1he power is in your hands.
LICENSEE CONTACT FOR THIS LEA 12) NAME TELEPHONE NUMBER (lnclu.de Area Code) M. J. Pastva, Jr. -*LER Coordinator (609) 339-5165 COMPLETE ONE LINE FOR' EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13) CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE CAUSE SYSTIEM COMPONENT MANUFACTURER REPORTABLE TO NPRDS TO NPRDS ;:;: SUPPLEMENTAL REPORT EXPECTED (14) EXPECTED MONTH DAY YEAR I YES SUBMISSION X (If yes, complete EXPECTED SUBMISSION DATE) NO DATE (15) 03 31 94 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 calorimetric and Reactor Coolant System flow calculations, determined the Unit may have .. operated above the.3411 megawatts (thermal)
95-2189 REV 7-92
I sp'ecified in Operating License Condition 2.C.(l). .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.
N~..C    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.
To avoid exceeding 100% reactor power, administrative controls have been implemented to limit power to 95% by calorimetric.
FACILITY NAME (1)                                                                                        DOCKET NUMBER (2)                                    PAGE (3)
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.
Salem Generating Station - Unit 2                                                                                        05000 311                              1 OF   04 TITLE (4)
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.
Reactor Power Higher Than Indicated EVENT DATE (5)                           LEA NUMBER (6                  REPORT NUMBER (7\                      OTHER FACILITIES INVOLVED (8)
It is anticipated that on or before 3/31/94, a supplement to this report will be provided to detail results of further investigation and testing and safety significance assessment of this event. NRC FORM 366 (5-92)
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    --      00        02        17        94                                                05000 OPERATING                      THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR &sect;: CChecl< one or more (11)
BLOCK NUMBER f 2 3 4 5 6 7 8 9 10 11 12 13 14 15 REQUIRED NUMBER OF DIGITS/CHARACTERS FOR EACH BLOCK NUMBER OF DIGITS/CHARACTERS TITLE UP TO 46 FACILITY NAME 8 TOTAL 3 IN ADDITION TO 05000 DOCKET NUMBER VARIES PAGE NUMBER UP TO 76 TITLE 6 TOTAL 2 PER BLOCK EVENT DATE 7 TOTAL 2 FOR YEAR 3 FOR SEQUENTIAL NUMBER LER NUMBER 2 FOR REVISION NUMBER 6 TOTAL 2 PER BLOCK REPORT DATE UP TO 18 FACILITY NAME 8 TOTAL-DOCKET NUMBER OTHER FACILITIES INVOLVED 3 IN ADDITION TO 05000 .. 1 OPERATING MODE 3 POWER LEVEL 1 REQUIREMENTS OF 10 CFR CHECK BOX THAT APPLIES UP TO 50 FOR NAME 14 FOR TELEPHONE LICENSEE CONTACT* CAUSE VARIES 2 FOR SYSTEM 4 FOR COMPONENT EACH COMPONENT FAILURE 4 FOR MANUFACTURER NPRDS VARIES 1 CHECK BOX THAT APPLIES SUPPLEMENTAL REPORT EXPECTED 6 TOTAL 2 PER BLOCK EXPECTED SUBMISSION DATE LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station Unit 2 DOCKET NUMBER 5000311 PLANT AND SYSTEM IDENTIFICATION:
MODE (9)                 1      20.402(b)                             20.405(c)                               50. 73 (a)(2)(iv)                  73.71 (b)
Westinghouse
POWER                          20.405(a)(1)(i)                       50.36(c)(1)                             50.73(a)(2)(v)                     73.71 (c)
-Pressurized Water Reactor LER NUMBER 94-002-00 PAGE 2 of 4 Energy Industry Identification system (EIIS) codes are identified in the text as {xx} IDENTIFICATION OF OCCURRENCE:
LEVEL (10)              100      20.405(a)(1)(ii)                       50.36(c)(2)                             50.73(a)(2)(vii)                   OTHER x                                                                        (Specify in Abstract I-20.405(a)(1)(iii)                     50.73(a)(2)(i)                         50.73(a) (2) (viii) (A) below and in Text, 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)
Reactor Power Higher Than Indicated Event Date: 1/19/94 Report Date: 2/17/94 This report was initiated by Incident Report Noo 94-027. CONDITIONS PRIOR TO OCCURRENCE:
LICENSEE CONTACT FOR THIS LEA 12)
Mode 1 Reactor Power 100% -Unit Load 1180 MWe DESCRIPTION OF OCCURRENCE:
NAME                                                                                                                TELEPHONE NUMBER (lnclu.de Area Code)
on 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.
M. J. Pastva, Jr. -*LER Coordinator                                                                                  (609) 339-5165 COMPLETE ONE LINE FOR' EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)
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.
REPORTABLE                                                                                    REPORTABLE CAUSE        SYSTEM        COMPONENT      MANUFACTURER                                        CAUSE    SYSTIEM    COMPONENT          MANUFACTURER TO NPRDS                                                                                      TO NPRDS
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)
SUPPLEMENTAL REPORT EXPECTED (14)                                                         EXPECTED          MONTH      DAY    YEAR X
(1) (ii) (B). ANALYSIS OF OCCURRENCE:
I YES (If yes, complete EXPECTED SUBMISSION DATE)
Nuclear instrumentation trip setpoints ensure that safety limits for the reactor core and reactor coolant system are not exceeded during normal operation and basis operational occurrences.
NO SUBMISSION DATE (15)          03        31      94 ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16)
Review of Fuel Cycle 8 calorimetric and Reactor Coolant System flow LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station Unit 2 DOCKET NUMBER 5000311 ANALYSIS OF OCCURRENCE: (cont'd) LER NUMBER 94-002-00 PAGE 3 of 4 calculations, shows the Unit's Operating License Condition maximum Reactor power level of 3411 megawatts may have been exceeded.
On 1/19/94, review of Unit 2 Fuel cycle 8 calorimetric and Reactor                                                                            ..
Initial assessment determined this event resulted from a potential error of 2.5% in actual Reactor thermal power higher than shown by nuclear instrumentation.
Coolant System flow calculations, determined the Unit may have operated above the.3411 megawatts (thermal) I sp'ecified in Operating License Condition 2.C.(l). .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.
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.
To avoid exceeding 100% reactor power, administrative controls have been implemented to limit power to 95% by calorimetric. Nuclear instrumentation has been adjusted due to the identified error.
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.
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 investigation and testing and safety significance assessment of this event.
This will prevent uncontrolled rod withdraw events until new setpoints for OTDT and OPDT are established, to reflect revised full power operating conditions.
NRC FORM 366 (5-92)
 
REQUIRED NUMBER OF DIGITS/CHARACTERS FOR EACH BLOCK BLOCK          NUMBER OF TITLE NUMBER    DIGITS/CHARACTERS f              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                                      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
 
LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station      DOCKET NUMBER    LER NUMBER        PAGE Unit 2                          5000311        94-002-00        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 Report Date:   2/17/94 This report was initiated by Incident Report Noo 94-027.
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) 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 safety limits for the reactor core and reactor coolant system are not exceeded during normal operation and de~ign basis anticipat~d operational occurrences.
Review of Fuel Cycle 8 calorimetric and Reactor Coolant System flow
 
LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station        DOCKET NUMBER    LER NUMBER        PAGE Unit 2                            5000311        94-002-00      3 of 4 ANALYSIS OF OCCURRENCE: (cont'd) calculations, shows the Unit's Operating License Condition maximum Reactor power level of 3411 megawatts (therm~l) 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:
APPARENT CAUSE OF OCCURRENCE:
The cause of the feedwater flow indication error is presently under investigation.
The cause of the feedwater flow indication error is presently under investigation.
PRIOR SIMILAR OCCURRENCES:
PRIOR SIMILAR OCCURRENCES:
A review of documentation did not show any prior similar occurrence of this event. SAFETY SIGNIFICANCE:
A review of documentation did not show any prior similar occurrence of this event.
This is reported pursuant to the requirements of 10CFR50.73(a)
SAFETY SIGNIFICANCE:
(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 events such Loss of Cooling Accidents (LOCAs) and the LOCA Containment Integrity This determination was made because depending on the analysis involved, either power level is not an initial condition in the analyses, there is sufficient margin in the analyses to mitigate the effects of the event, or because credit can be taken for items outside of the licensing basis. With regard to non-LOCA events,. power level is both an initial condition and a basis for the setpoints of both the Reactor Protection 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 conditions for
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 events such Loss of Cooling Accidents (LOCAs) and the LOCA Containment Integrity analys~s. This determination was made because depending on the analysis involved, either power level is not an initial condition in the analyses, there is sufficient margin in the analyses to mitigate the effects of the event, or because credit can be taken for items outside of the licensing basis. With regard to non-LOCA events,. power level is both an initial condition and a basis for the setpoints of both the Reactor Protection 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 conditions for
LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station Unit 2 SAFETY SIGNIFICANCE: (cont'd) DOCKET NUMBER 5000311 non-LOCA events will be assessed.
 
CORRECTIVE ACTION: LER NUMBER 94-002-00 PAGE 4 of 4 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.
LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station     DOCKET NUMBER      LER NUMBER      PAGE Unit 2                         5000311          94-002-00      4 of 4 SAFETY SIGNIFICANCE: (cont'd) non-LOCA events will be assessed.
It is anticipated that on or before March 31, 1994, a supplement to this report will be provided to detail the results of further investigation and testing and safety significance assessment of this event. MJPJ:pc SORC Mtg. 94-017}}
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, 1994, a supplement to this report will be provided to detail the results of further investigation and testing and safety significance assessment of this event.
MJPJ:pc SORC Mtg. 94-017}}

Revision as of 10:31, 21 October 2019

LER 94-002-00:on 940119,determined That Unit May Have Operated Above 3,411 MW Specified in OL Condition 2.C.(1). Administrative Controls Implemented to Power to 95% by Calorimetric
ML18100A889
Person / Time
Site: Salem PSEG icon.png
Issue date: 02/17/1994
From: Hagan J, Pastva M
Public Service Enterprise Group
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-94-002-01, LER-94-2-1, NUDOCS 9402240010
Download: ML18100A889 (6)
v • d • e


Text

e OPS~G Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge; New Jersey 08038 Salem Generating Station February 17, 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 LICENSEE EVENT REPORT 94-002-00 This Licensee Event Report is being submitted pursuant to the requirements of Code of Federal Regulation 10CFR50.73(a) (2) (i) (B).

Issuance of this report is required within thirty (30) days of event discovery.

Sincerely yours, gan General Manager -

Salem Operations MJPJ:pc Distribution 9402240010 940217 PDR ADOCK 05000311 S . PDR 1he power is in your hands.

95-2189 REV 7-92

N~..C 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) PAGE (3)

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

..

Reactor Power Higher Than Indicated 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 -- 00 02 17 94 05000 OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: CChecl< 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 x (Specify in Abstract I-20.405(a)(1)(iii) 50.73(a)(2)(i) 50.73(a) (2) (viii) (A) below and in Text, 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 LEA 12)

NAME TELEPHONE NUMBER (lnclu.de 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 SYSTIEM COMPONENT MANUFACTURER TO NPRDS TO NPRDS

SUPPLEMENTAL REPORT EXPECTED (14) EXPECTED MONTH DAY YEAR X

I YES (If yes, complete EXPECTED SUBMISSION DATE)

NO SUBMISSION DATE (15) 03 31 94 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 calorimetric and Reactor ..

Coolant System flow calculations, determined the Unit may have operated above the.3411 megawatts (thermal) I sp'ecified in Operating License Condition 2.C.(l). .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, administrative 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 investigation and testing and safety significance assessment of this event.

NRC FORM 366 (5-92)

REQUIRED NUMBER OF DIGITS/CHARACTERS FOR EACH BLOCK BLOCK NUMBER OF TITLE NUMBER DIGITS/CHARACTERS f 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 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

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station DOCKET NUMBER LER NUMBER PAGE Unit 2 5000311 94-002-00 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 Report Date: 2/17/94 This report was initiated by Incident Report Noo 94-027.

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) 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 safety limits for the reactor core and reactor coolant system are not exceeded during normal operation and de~ign basis anticipat~d operational occurrences.

Review of Fuel Cycle 8 calorimetric and Reactor Coolant System flow

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station DOCKET NUMBER LER NUMBER PAGE Unit 2 5000311 94-002-00 3 of 4 ANALYSIS OF OCCURRENCE: (cont'd) calculations, shows the Unit's Operating License Condition maximum Reactor power level of 3411 megawatts (therm~l) 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 feedwater 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 events such Loss of Cooling Accidents (LOCAs) and the LOCA Containment Integrity analys~s. This determination was made because depending on the analysis involved, either power level is not an initial condition in the analyses, there is sufficient margin in the analyses to mitigate the effects of the event, or because credit can be taken for items outside of the licensing basis. With regard to non-LOCA events,. power level is both an initial condition and a basis for the setpoints of both the Reactor Protection 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 conditions for

--- - -

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station DOCKET NUMBER LER NUMBER PAGE Unit 2 5000311 94-002-00 4 of 4 SAFETY SIGNIFICANCE: (cont'd) 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, 1994, a supplement to this report will be provided to detail the results of further investigation and testing and safety significance assessment of this event.

MJPJ:pc SORC Mtg.94-017

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