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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 |
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| ==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)
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| (2) (i) (B). Issuance of this report is required within thirty (30) days of event discovery.
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| 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
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| 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 §: 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.
| | Reactor Power Higher Than Indicated EVENT DATE (5) LEA NUMBER (6 REPORT NUMBER (7\ OTHER FACILITIES INVOLVED (8) |
| The cause of the f eedwater flow indication error is under investigation.
| | 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) |
| 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)
| | MODE (9) 1 20.402(b) 20.405(c) 50. 73 (a)(2)(iv) 73.71 (b) |
| 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:
| | POWER 20.405(a)(1)(i) 50.36(c)(1) 50.73(a)(2)(v) 73.71 (c) |
| Westinghouse
| | 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) |
| -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: | | LICENSEE CONTACT FOR THIS LEA 12) |
| 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:
| | NAME TELEPHONE NUMBER (lnclu.de Area Code) |
| Mode 1 Reactor Power 100% -Unit Load 1180 MWe DESCRIPTION OF OCCURRENCE:
| | M. J. Pastva, Jr. -*LER Coordinator (609) 339-5165 COMPLETE ONE LINE FOR' EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13) |
| 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.
| | REPORTABLE REPORTABLE CAUSE SYSTEM COMPONENT MANUFACTURER CAUSE SYSTIEM COMPONENT MANUFACTURER TO NPRDS TO NPRDS SUPPLEMENTAL REPORT EXPECTED (14) EXPECTED MONTH DAY YEAR X |
| 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.
| | I YES (If yes, complete EXPECTED SUBMISSION DATE) |
| 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
| | NO SUBMISSION DATE (15) 03 31 94 ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16) |
| *. The NRC was notified of this-event per 10CFR50.72(b)
| | On 1/19/94, review of Unit 2 Fuel cycle 8 calorimetric and Reactor .. |
| (1) (ii) (B). ANALYSIS OF OCCURRENCE:
| | 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. |
| 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.
| | 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. |
| 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.
| | 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. |
| Initial assessment determined this event resulted from a potential error of 2.5% in actual Reactor thermal power higher than shown by nuclear instrumentation.
| | NRC FORM 366 (5-92) |
| 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. | | 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 |
| This will prevent uncontrolled rod withdraw events until new setpoints for OTDT and OPDT are established, to reflect revised full power operating conditions. | | 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. | | 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. |
| This will prevent uncontrolled rod withdraw events until new setpoints for OTDT and OPDT are established, to reflect revised full power operating conditions. | | CORRECTIVE ACTION: |
| 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}} | | 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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LER 98-012-01:on 980725,noted That Afs Was Operated with Less than Required Number of Operable AFW Pumps.Caused by Improper Procedure Implementation.Runout Protection Pressure Device for 22 AFW Pumps Was Returned to Svc.With 980928 Ltr
ML18106A8821998-09-21021 September 1998
LER 98-013-00:on 980820,noted Surveillance of Containment Penetration Overcurrent Protection Devices Missed.Caused by Human Error.Satisfactorily Tested Apprpriate Breakers & Disciplined Involved Personnel.With 980921 Ltr
ML18106A8791998-09-16016 September 1998
LER 96-006-01:on 960717,determined That non-radioactive Liquid Basin Radwaste Monitor Inoperable During Low Head Conditions.Caused by Inadequate Design Change Package.Design Change 1EC3663-01 Has Been Installed.With 980916 Ltr
ML18106A8801998-09-0808 September 1998
LER 98-013-00:on 980806,operation with TS Required Equipment OOS Was Noted.Caused by Human Error.Reviewed Processes & Practices Re Safety Sys Status Control,Procedure Rev & Extra Training.With 980908 Ltr
ML18106A8531998-08-27027 August 1998
LER 98-007-00:on 980730,reactor Coolant Instrument Line through-wall Leak Was Noted.Cause of Event Has Not Yet Been Determined.Assembled Root Cause Team & Replaced Affected tubing.W/980827 Ltr
ML18106A8521998-08-27027 August 1998
LER 98-011-00:on 980803,ESFA During a 4KV Automatic Transfer Test Was Noted.Caused by Premature Release of Control Console Pushbutton Due to Inadequate Procedural Step.Revised procedure.W/980827 Ltr
ML18106A8421998-08-24024 August 1998
LER 98-012-00:on 980725,discovered That Plant Had Operated in Modes 1 & 2 w/twenty-two AFW Pumps Inoperable.Caused by Failure to Restore Pump Runout Protection Pressure Device to Svc.Returned Subject Device to svc.W/980824 Ltr
ML18106A8431998-08-24024 August 1998
LER 98-009-00:on 980810,failure to Post Continuous Firewatch as Required by Fire Protection Plan Noted.Caused by Failure to Recognize Concurrent Conditions.Continuous Firewatch Was Posted Immediately & Repaired Smoke detectors.W/980824 Ltr
ML18106A8141998-08-13013 August 1998
LER 98-010-00:on 980714,determined That Leakage from Boron Injection Tank Exceeded Max Allowable ECCS Leakage from Sources Outside Containment.Caused by Leaking 2SJ404 Manual Sample valve.2SJ404 Valve repaired.W/980813 Ltr
ML18106A8201998-08-13013 August 1998
LER 98-012-00:on 980715,potential to Exceed Rating of Piping Due to Isolation of Overpressure Protection Line Was Noted. Caused by Inadequate Procedural Guidance.Appropriate Operations Dept Procedures Have Been revised.W/980813 Ltr
ML18106A6931998-06-29029 June 1998
LER 98-003-00:on 980122,inappropriate Plugging of Tubes R9C60 & R10C60 in Salem Unit 2 Sg,Was Performed.Caused by Failure of Qualification,Verification & Validation Process. Tubes Reviewed to Verify No Others Inappropriately Plugged
ML18106A6471998-06-0404 June 1998
LER 98-011-00:on 980505,improper Isolation of Single Cell Battery Charger from 125 Vdc Battery Was Noted.Caused by Inadequate 10CFR50.59 Applicability Review.Placed Procedure SC.MD-CM.ZZ-0024(Q) on Administrative hold.W/980604 Ltr
ML18106A6421998-06-0101 June 1998
LER 98-010-00:on 931019,reactor Pressure Vessel Insp Plugs Were Out of Configuration,Was Noted.Caused by Personnel Error.Proper Configuration Was Restored Shortly After Discovery Prior to Entering Mode 2.W/980601 Ltr
ML18106A6431998-05-29029 May 1998
LER 98-006-01:on 980227,determined Incorrect Scaling Error of First Stage Pressure Transmitter Existed.Caused by Human Error.Revised Setpoint Calculation SC-MS002-01 & Revised Associated Instrument Calibr Database info.W/980529 Ltr
ML18106A6141998-05-18018 May 1998
LER 98-008-00:on 970814,failure to Test 21 & 22 AF 40 Valves in Closed Direction as Required by TS 4.0.5 Was Noted.Caused by Inadequate Design Mod Process.Motor Driven 21/22 AF 40 Valves Were Tested IAW Revised procedure.W/980518 Ltr
ML18106A5901998-05-0101 May 1998
LER 98-009-00:on 980405,epoxy Missing from Terminals of H Analyzer Was Noted.Caused by Inadequate Development of Procedure.H Analyzers Were Repaired & Review of Other Safety Related Equipment in Containment Was performed.W/980501 Ltr
ML18106A5611998-04-20020 April 1998
LER 98-008-00:on 980323,inadequate Testing of Salem Unit 1 Containment Air Locks Resulted in Entering TS 3.0.3.Caused by less-than-adequate Work Practices During Replacement of Equalizing Valve.Salem Unit 2 Airlocks Were Inspected
ML18106A6061998-04-0101 April 1998
Corrected LER 98-004-00:on 980302,failure to Comply W/Tss 4.11.1.1.2 & 3.3.3.8 Was Noted.Caused by Organizational Deficiency.Steps Have Been Taken to Correctly Document Safety Factors.Corrects Prior Similar Occurrences
ML18106A4451998-04-0101 April 1998
LER 98-004-00:on 980302,failure to Perform TS 4.11.1.1.2 & 3.3.3.8 Was Noted.Caused by Organizational Deficiency.Steps Were Taken to Correctly Document Safety factors.W/980401 Ltr
ML18106A4351998-03-30030 March 1998
LER 98-006-00:on 980227,incorrect Scaling of First Stage Turbine Impulse Pressure Transmitters Noted.Cause Indeterminate.Implemented Procedure Changes & re-scaled Affected Turbine Impulse Pressure transmitters.W/980330 Ltr
ML18106A3961998-03-20020 March 1998
LER 98-005-00:on 980219,inoperability of Twelve EDG Fuel Oil Transfer Pump (FOTP) Noted.Caused by Installation of Incorrect Control Switch.Installed Correct off-auto-manual Switch & Verified Operability of Twelve FOTP.W/980320 Ltr
ML18106A5781998-03-20020 March 1998
Corrected LER 98-005-00:on 980219,inoperability of 12 Fuel Oil Transfer Pump (Fotp),Noted.Caused by Installation of Incorrect Control Switch.Field Insp Performed to Verify Configuration of Switches for 11,21 & 22 FOTPs
ML18106A4021998-03-20020 March 1998
LER 98-007-00:on 980218,failure to Establish Containment Integrity (Closure) Prior to Fuel Movement Was Noted.Caused by Failure to Identify & Include Condensate Pot Vent in Appropriate Valve Lineup.Valves identified.W/980320 Ltr
ML18106A4031998-03-20020 March 1998
LER 98-006-00:on 980221,ESF Actuation of 11 & 12 Auxiliary Feedwater Pumps Occurred.Caused by Human Error.Operators Promptly Established Feedwater to All SG & Restored Proper Water levels.W/980320 Ltr
1999-08-26
[Table view]
Category:RO)
MONTHYEARML18107A5031999-08-26026 August 1999
LER 99-006-00:on 990729,determined That SG Blowdown RMs Setpoint Was non-conservative.Caused by Inadequate ACs for Incorporating Original Plant Licensing Data Into Plant Procedures.Blowdown Will Be Restricted.With 990826 Ltr
ML18107A4691999-07-28028 July 1999
LER 99-008-00:on 990714,determined That Limit Switch Cables Were Subject to Multiple Hot Shorts in Same Fire Area.Caused by Inadequate Original Post Fire Safe Shutdown Analysis.All Limit Switch Cables for MOVs Were Reviewed.With 990728 Ltr
ML18107A4441999-07-0606 July 1999
LER 99-007-00:on 990605,surveillance for Quadrant Power Tilt Ratio (QPTR) Was Missed.Caused by Human Error.Qptr Calculation Was Performed & Personnel Involved Have Been Held Accountable IAW Pse&G Policies.With 990706 Ltr
ML18107A4211999-07-0202 July 1999
LER 99-005-00:on 990605,11 Containment Declared Inoperable. Caused by Valves 11SW72 & 11SW223 Both Leaking.Procedure S1.OP-ST.SW-0010(Q) Was Enhanced to Provide Specific Instructions to Ensure Proper Sequencing.With 990702 Ltr
ML18107A4321999-07-0101 July 1999
LER 99-006-01:on 990501,determined That There Was No Flow in One of Four Injection Legs.Caused by Sticking of Valve in Safety Injection Discharge Line to 21 Cold Leg.Valve Was Cut Out of Sys & Replaced.With 990701 Ltr
ML18107A4331999-07-0101 July 1999
LER 99-002-01:on 990405,determined That 2SA118 Failed as Found Leakrate Test.Caused by Foreign Matl Found in 2SA118 valve.2SA118 Valve Was Cycled Several Times & Seat Area Was Air Blown in Order to Displace Foreign Matl.With 990701 Ltr
ML18107A3951999-06-17017 June 1999
LER 99-004-00:on 990520,reactor Tripped from 100% Power,Due to Negative Flux Trip Signal from Nuclear Instrumentation. Cause Has Not Been Determined.Discoloration Was Identified on One of Penetrations.With 990617 Ltr
ML18107A3661999-06-0909 June 1999
LER 99-003-00:on 990513,unplanned Entry Into TS 3.0.3 Was Made.Caused by Human error.Re-positioned Creacs Supply Fan Selector Switches & Revised Procedures S1 & S2.OP-ST.SSP-0001(Q).With 990609 Ltr
ML18107A3551999-06-0202 June 1999
LER 99-005-00:on 990504,failure to Meet TS Action Statement Requirements for High Oxygen Concentration in Waste Gas Holdup Sys Occurred.Caused by Inability of Operators. Existing Procedures Will Be Evaluated.With 990602 Ltr
ML18107A3541999-06-0101 June 1999
LER 99-006-00:on 990501,HHSI Flow Balance Discrepancy Was Noted During Surveillance.Caused by Sticking of Check Valve in SI Discharge Line to 21 Cold Leg.Valve 21SJ17,was Cut Out of Sys & Replaced.With 990601 Ltr
ML18107A2931999-05-12012 May 1999
LER 99-002-00:on 990413,determined That Number 12 Auxiliary Bldg Exhaust Fan Was Rotating Backwards.Caused by mis-wiring of Motor Due to Human Error by Maint technician.Mis-wiring Was Corrected & Fan Was Returned to Svc.With 990512 Ltr
ML18107A2781999-05-10010 May 1999
LER 99-004-00:on 990411,automatic Actuation of ESF Occurred During Reactor Vessel Head Removal in Support of Refueling Operations.Caused by High Radiation Condition.Containment Atmosphere Was Monitored.With 990505 Ltr
ML18107A2791999-05-0404 May 1999
LER 99-003-00:on 990406,all Salem Unit 2 Chillers Rendered Inoperable.Caused by Human Error.Lessons Learned from Event Were Communicated to All Operators by Including Them in Night Orders.With 990504 Ltr
ML18107A2741999-05-0303 May 1999
LER 99-002-00:on 990405,determined That Containment Isolation Valve Failed as Found Leakrate Test.Caused by Foreign Matl Blocking Valves from Closing.Check Valve Mechanically Agitated.With 990504 Ltr
ML18107A2351999-04-23023 April 1999
LER 99-001-00:on 990330,MSSV Failed Lift Set Test.Caused by Setpoint Variance Which Is Result of Aging.Valves Were Adjusted & Retested to Ensure TS Tolerance.With 990423 Ltr
ML18106B1471999-03-29029 March 1999
LER 99-001-00:on 990228,reactor Scram Was Noted as Result of Turbine Trip.Caused by Operator Error.Lesson Plans Revised to Explicitly Demonstrate Manner in Which Valve Functions. with 990329 Ltr
ML18106B0701999-02-16016 February 1999
LER 98-015-00:on 981208,inadvertent Discharge Through RHR Relief Valve During Startup Was Noted.Caused by Operator Performing Too Many Tasks Simultaneously.Appropriate Actions Have Been Taken IAW Policies & Procedures.With 990216 Ltr
ML18106B0491999-01-28028 January 1999
LER 98-007-01:on 980730,reactor Coolant Instrument Line through-wall Leak Was Noted.Caused by Transgranular Stress Corrosion Cracking.Replaced Affected Tubing.With 990128 Ltr
ML18106B0401999-01-18018 January 1999
LER 98-016-00:on 981219,ECCS Leakage Was Outside of Design Value.Caused by Leakage Past Seat of 21RH34 Manual Drain. Valve 21RH34 Was Reseated.With 990118 Ltr
ML18106B0081998-12-24024 December 1998
LER 97-001-01:on 970215,failure to Perform TS Surveillance of Component Cooling Water Sys Check Valves Occurred.Caused by Inadequate Communication Between EOP Group & IST Reviewers.Procedure Revised.With 981224 Ltr
ML18106B0021998-12-17017 December 1998
LER 98-015-01:on 980924,improper Installation of Test Equipment to RPS Occurred.Caused by Inadequate 10CFR50.59 Applicability Reviews During Past Revs.Revised Procedures. with 981217 Ltr
ML18106A9551998-11-0303 November 1998
LER 96-013-01:on 960711,concluded That Current Gain & Bias Settings Had Rendered Overtemperature Delta Temp Protection Channels Inoperable.Caused by Scaling Error.Licensee Will Revise Scaling Calculations.With 981105 Ltr
ML18106A9451998-10-30030 October 1998
LER 97-004-01:on 970408,failure to Comply with TS Action Statement,Dg Start & Inadequate Surveillance Testing,Was Noted.Caused by Inadequate Tracking of Inoperable Equipment. Discussed Event & Lessons Learned.With 981022 Ltr
ML18106A9491998-10-22022 October 1998
LER 98-015-00:on 980924,identified Improper Installation of Test Equipment to Rps.Cause Indeterminate.Procedures for Installation of Test Equipment for Collection of State Point Data Were Placed on Administrative Hold.With 981022 Ltr
ML18106A9301998-10-21021 October 1998
LER 98-014-00:on 980725,noted Improper Calibr of Liquid Radwaste Effluent Line Radiation Monitor.Caused by Inattention to Detail by Maint Personnel.Channel Calibr Was Successfully Performed on 1R18 on 980821.With 981019 Ltr
ML18106A9071998-10-0101 October 1998
LER 98-014-00:on 980918,discovered That Fire Barrier Matl for HVAC Ducts Does Not Meet Required Level of Fire Resistance.Cause Indeterminate.Established Appropriate Compensatory Actions for Fire Barriers.With 981001 Ltr
ML18106A8951998-09-28028 September 1998
LER 98-012-01:on 980725,noted That Afs Was Operated with Less than Required Number of Operable AFW Pumps.Caused by Improper Procedure Implementation.Runout Protection Pressure Device for 22 AFW Pumps Was Returned to Svc.With 980928 Ltr
ML18106A8821998-09-21021 September 1998
LER 98-013-00:on 980820,noted Surveillance of Containment Penetration Overcurrent Protection Devices Missed.Caused by Human Error.Satisfactorily Tested Apprpriate Breakers & Disciplined Involved Personnel.With 980921 Ltr
ML18106A8791998-09-16016 September 1998
LER 96-006-01:on 960717,determined That non-radioactive Liquid Basin Radwaste Monitor Inoperable During Low Head Conditions.Caused by Inadequate Design Change Package.Design Change 1EC3663-01 Has Been Installed.With 980916 Ltr
ML18106A8801998-09-0808 September 1998
LER 98-013-00:on 980806,operation with TS Required Equipment OOS Was Noted.Caused by Human Error.Reviewed Processes & Practices Re Safety Sys Status Control,Procedure Rev & Extra Training.With 980908 Ltr
ML18106A8531998-08-27027 August 1998
LER 98-007-00:on 980730,reactor Coolant Instrument Line through-wall Leak Was Noted.Cause of Event Has Not Yet Been Determined.Assembled Root Cause Team & Replaced Affected tubing.W/980827 Ltr
ML18106A8521998-08-27027 August 1998
LER 98-011-00:on 980803,ESFA During a 4KV Automatic Transfer Test Was Noted.Caused by Premature Release of Control Console Pushbutton Due to Inadequate Procedural Step.Revised procedure.W/980827 Ltr
ML18106A8421998-08-24024 August 1998
LER 98-012-00:on 980725,discovered That Plant Had Operated in Modes 1 & 2 w/twenty-two AFW Pumps Inoperable.Caused by Failure to Restore Pump Runout Protection Pressure Device to Svc.Returned Subject Device to svc.W/980824 Ltr
ML18106A8431998-08-24024 August 1998
LER 98-009-00:on 980810,failure to Post Continuous Firewatch as Required by Fire Protection Plan Noted.Caused by Failure to Recognize Concurrent Conditions.Continuous Firewatch Was Posted Immediately & Repaired Smoke detectors.W/980824 Ltr
ML18106A8141998-08-13013 August 1998
LER 98-010-00:on 980714,determined That Leakage from Boron Injection Tank Exceeded Max Allowable ECCS Leakage from Sources Outside Containment.Caused by Leaking 2SJ404 Manual Sample valve.2SJ404 Valve repaired.W/980813 Ltr
ML18106A8201998-08-13013 August 1998
LER 98-012-00:on 980715,potential to Exceed Rating of Piping Due to Isolation of Overpressure Protection Line Was Noted. Caused by Inadequate Procedural Guidance.Appropriate Operations Dept Procedures Have Been revised.W/980813 Ltr
ML18106A6931998-06-29029 June 1998
LER 98-003-00:on 980122,inappropriate Plugging of Tubes R9C60 & R10C60 in Salem Unit 2 Sg,Was Performed.Caused by Failure of Qualification,Verification & Validation Process. Tubes Reviewed to Verify No Others Inappropriately Plugged
ML18106A6471998-06-0404 June 1998
LER 98-011-00:on 980505,improper Isolation of Single Cell Battery Charger from 125 Vdc Battery Was Noted.Caused by Inadequate 10CFR50.59 Applicability Review.Placed Procedure SC.MD-CM.ZZ-0024(Q) on Administrative hold.W/980604 Ltr
ML18106A6421998-06-0101 June 1998
LER 98-010-00:on 931019,reactor Pressure Vessel Insp Plugs Were Out of Configuration,Was Noted.Caused by Personnel Error.Proper Configuration Was Restored Shortly After Discovery Prior to Entering Mode 2.W/980601 Ltr
ML18106A6431998-05-29029 May 1998
LER 98-006-01:on 980227,determined Incorrect Scaling Error of First Stage Pressure Transmitter Existed.Caused by Human Error.Revised Setpoint Calculation SC-MS002-01 & Revised Associated Instrument Calibr Database info.W/980529 Ltr
ML18106A6141998-05-18018 May 1998
LER 98-008-00:on 970814,failure to Test 21 & 22 AF 40 Valves in Closed Direction as Required by TS 4.0.5 Was Noted.Caused by Inadequate Design Mod Process.Motor Driven 21/22 AF 40 Valves Were Tested IAW Revised procedure.W/980518 Ltr
ML18106A5901998-05-0101 May 1998
LER 98-009-00:on 980405,epoxy Missing from Terminals of H Analyzer Was Noted.Caused by Inadequate Development of Procedure.H Analyzers Were Repaired & Review of Other Safety Related Equipment in Containment Was performed.W/980501 Ltr
ML18106A5611998-04-20020 April 1998
LER 98-008-00:on 980323,inadequate Testing of Salem Unit 1 Containment Air Locks Resulted in Entering TS 3.0.3.Caused by less-than-adequate Work Practices During Replacement of Equalizing Valve.Salem Unit 2 Airlocks Were Inspected
ML18106A6061998-04-0101 April 1998
Corrected LER 98-004-00:on 980302,failure to Comply W/Tss 4.11.1.1.2 & 3.3.3.8 Was Noted.Caused by Organizational Deficiency.Steps Have Been Taken to Correctly Document Safety Factors.Corrects Prior Similar Occurrences
ML18106A4451998-04-0101 April 1998
LER 98-004-00:on 980302,failure to Perform TS 4.11.1.1.2 & 3.3.3.8 Was Noted.Caused by Organizational Deficiency.Steps Were Taken to Correctly Document Safety factors.W/980401 Ltr
ML18106A4351998-03-30030 March 1998
LER 98-006-00:on 980227,incorrect Scaling of First Stage Turbine Impulse Pressure Transmitters Noted.Cause Indeterminate.Implemented Procedure Changes & re-scaled Affected Turbine Impulse Pressure transmitters.W/980330 Ltr
ML18106A3961998-03-20020 March 1998
LER 98-005-00:on 980219,inoperability of Twelve EDG Fuel Oil Transfer Pump (FOTP) Noted.Caused by Installation of Incorrect Control Switch.Installed Correct off-auto-manual Switch & Verified Operability of Twelve FOTP.W/980320 Ltr
ML18106A5781998-03-20020 March 1998
Corrected LER 98-005-00:on 980219,inoperability of 12 Fuel Oil Transfer Pump (Fotp),Noted.Caused by Installation of Incorrect Control Switch.Field Insp Performed to Verify Configuration of Switches for 11,21 & 22 FOTPs
ML18106A4021998-03-20020 March 1998
LER 98-007-00:on 980218,failure to Establish Containment Integrity (Closure) Prior to Fuel Movement Was Noted.Caused by Failure to Identify & Include Condensate Pot Vent in Appropriate Valve Lineup.Valves identified.W/980320 Ltr
ML18106A4031998-03-20020 March 1998
LER 98-006-00:on 980221,ESF Actuation of 11 & 12 Auxiliary Feedwater Pumps Occurred.Caused by Human Error.Operators Promptly Established Feedwater to All SG & Restored Proper Water levels.W/980320 Ltr
1999-08-26
[Table view]
Category:TEXT-SAFETY REPORT
MONTHYEARML20217A9931999-09-30030 September 1999
NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data
ML18107A5581999-09-30030 September 1999
Monthly Operating Rept for Sept 1999 for Salem,Unit 2.With 991014 Ltr
ML18107A5571999-09-30030 September 1999
Monthly Operating Rept for Sept 1999 for Salem,Unit 1.With 991014 Ltr
ML18107A5301999-08-31031 August 1999
Monthly Operating Rept for Aug 1999 for Salem,Unit 2.With 990913 Ltr
ML18107A5311999-08-31031 August 1999
Monthly Operating Rept for Aug 1999 for Salem,Unit 1.With 990913
ML18107A5031999-08-26026 August 1999
LER 99-006-00:on 990729,determined That SG Blowdown RMs Setpoint Was non-conservative.Caused by Inadequate ACs for Incorporating Original Plant Licensing Data Into Plant Procedures.Blowdown Will Be Restricted.With 990826 Ltr
ML18107A5201999-08-12012 August 1999
Rev 0 to Sgs Unit 2 ISI RFO Exam Results (S2RFO#9) Second Interval,Second Period, First Outage (96RF).
ML18107A4811999-07-31031 July 1999
Monthly Operating Rept for July 1999 for Salem,Unit 1.With 990813 Ltr
ML18107A4821999-07-31031 July 1999
Monthly Operating Rept for July 1999 for Salem,Unit 2.With 990813 Ltr
ML18107A4691999-07-28028 July 1999
LER 99-008-00:on 990714,determined That Limit Switch Cables Were Subject to Multiple Hot Shorts in Same Fire Area.Caused by Inadequate Original Post Fire Safe Shutdown Analysis.All Limit Switch Cables for MOVs Were Reviewed.With 990728 Ltr
ML18107A4441999-07-0606 July 1999
LER 99-007-00:on 990605,surveillance for Quadrant Power Tilt Ratio (QPTR) Was Missed.Caused by Human Error.Qptr Calculation Was Performed & Personnel Involved Have Been Held Accountable IAW Pse&G Policies.With 990706 Ltr
ML18107A4211999-07-0202 July 1999
LER 99-005-00:on 990605,11 Containment Declared Inoperable. Caused by Valves 11SW72 & 11SW223 Both Leaking.Procedure S1.OP-ST.SW-0010(Q) Was Enhanced to Provide Specific Instructions to Ensure Proper Sequencing.With 990702 Ltr
ML18107A4331999-07-0101 July 1999
LER 99-002-01:on 990405,determined That 2SA118 Failed as Found Leakrate Test.Caused by Foreign Matl Found in 2SA118 valve.2SA118 Valve Was Cycled Several Times & Seat Area Was Air Blown in Order to Displace Foreign Matl.With 990701 Ltr
ML18107A4321999-07-0101 July 1999
LER 99-006-01:on 990501,determined That There Was No Flow in One of Four Injection Legs.Caused by Sticking of Valve in Safety Injection Discharge Line to 21 Cold Leg.Valve Was Cut Out of Sys & Replaced.With 990701 Ltr
ML18107A5211999-07-0101 July 1999
Rev 0 to Sgs Unit 2 ISI RFO Exam Results (S2RFO#10) Second Interval,Second Period,Second Outage (99RF).
ML18107A4351999-06-30030 June 1999
Monthly Operating Rept for June 1999 for Salem,Unit 1.With 990713 Ltr
ML18107A4341999-06-30030 June 1999
Monthly Operating Rept for June 1999 for Salem,Unit 2.With 990713 Ltr
ML20196H8621999-06-30030 June 1999
NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data, June 1999 Rept
ML18107A3951999-06-17017 June 1999
LER 99-004-00:on 990520,reactor Tripped from 100% Power,Due to Negative Flux Trip Signal from Nuclear Instrumentation. Cause Has Not Been Determined.Discoloration Was Identified on One of Penetrations.With 990617 Ltr
ML18107A3661999-06-0909 June 1999
LER 99-003-00:on 990513,unplanned Entry Into TS 3.0.3 Was Made.Caused by Human error.Re-positioned Creacs Supply Fan Selector Switches & Revised Procedures S1 & S2.OP-ST.SSP-0001(Q).With 990609 Ltr
ML18107A3551999-06-0202 June 1999
LER 99-005-00:on 990504,failure to Meet TS Action Statement Requirements for High Oxygen Concentration in Waste Gas Holdup Sys Occurred.Caused by Inability of Operators. Existing Procedures Will Be Evaluated.With 990602 Ltr
ML18107A3441999-06-0101 June 1999
Interim Part 21 Rept Re Premature Over Voltage Protection Actuation in Circuit Specific Application in Dc Power Supply.Testing & Evaluation Activities Will Be Completed on 990716
ML18107A3541999-06-0101 June 1999
LER 99-006-00:on 990501,HHSI Flow Balance Discrepancy Was Noted During Surveillance.Caused by Sticking of Check Valve in SI Discharge Line to 21 Cold Leg.Valve 21SJ17,was Cut Out of Sys & Replaced.With 990601 Ltr
ML18107A3681999-05-31031 May 1999
Monthly Operating Rept for May 1999 for Salem Generating Station,Unit 1.With 990611 Ltr
ML18107A3721999-05-31031 May 1999
Monthly Operating Rept for May 1999 for Salem Generating Station,Unit 2.With 990611 Ltr
ML18107A2931999-05-12012 May 1999
LER 99-002-00:on 990413,determined That Number 12 Auxiliary Bldg Exhaust Fan Was Rotating Backwards.Caused by mis-wiring of Motor Due to Human Error by Maint technician.Mis-wiring Was Corrected & Fan Was Returned to Svc.With 990512 Ltr
ML18107A2781999-05-10010 May 1999
LER 99-004-00:on 990411,automatic Actuation of ESF Occurred During Reactor Vessel Head Removal in Support of Refueling Operations.Caused by High Radiation Condition.Containment Atmosphere Was Monitored.With 990505 Ltr
ML18107A2791999-05-0404 May 1999
LER 99-003-00:on 990406,all Salem Unit 2 Chillers Rendered Inoperable.Caused by Human Error.Lessons Learned from Event Were Communicated to All Operators by Including Them in Night Orders.With 990504 Ltr
ML18107A2741999-05-0303 May 1999
LER 99-002-00:on 990405,determined That Containment Isolation Valve Failed as Found Leakrate Test.Caused by Foreign Matl Blocking Valves from Closing.Check Valve Mechanically Agitated.With 990504 Ltr
ML18107A3711999-04-30030 April 1999
Corrected Monthly Operating Rept for Apr 1999 for Salem Generating Station,Unit 1
ML18107A3151999-04-30030 April 1999
Submittal-Only Screening Review of Salem Generating Station Individual Plant Exam for External Events (Seismic Portion), Rev 1
ML18107A2991999-04-30030 April 1999
Monthly Operating Rept for Apr 1999 for Salem Unit 1.With 990514 Ltr
ML18107A2971999-04-30030 April 1999
Monthly Operating Rept for Apr 1999 for Salem Unit 2.With 990514 Ltr
ML18107A2351999-04-23023 April 1999
LER 99-001-00:on 990330,MSSV Failed Lift Set Test.Caused by Setpoint Variance Which Is Result of Aging.Valves Were Adjusted & Retested to Ensure TS Tolerance.With 990423 Ltr
ML18107A2881999-04-0707 April 1999
Rev 0 to NFS-0174, COLR for Salem Unit 2 Cycle 11.
ML18107A1821999-03-31031 March 1999
Monthly Operating Rept for Mar 1999 for Salem,Unit 1.With 990414 Ltr
ML18107A1831999-03-31031 March 1999
Monthly Operating Rept for Mar 1999 for Salem,Unit 2.With 990414 Ltr
ML18106B1471999-03-29029 March 1999
LER 99-001-00:on 990228,reactor Scram Was Noted as Result of Turbine Trip.Caused by Operator Error.Lesson Plans Revised to Explicitly Demonstrate Manner in Which Valve Functions. with 990329 Ltr
ML18106B1021999-02-28028 February 1999
Monthly Operating Rept for Feb 1999 for Salem Unit 2.With 990315 Ltr
ML18106B1011999-02-28028 February 1999
Monthly Operating Rept for Feb 1999 for Salem Unit 1.With 990315 Ltr
ML18106B0931999-02-25025 February 1999
Part 21 Rept Re Possible Defect in Swagelok Pipe Fitting Tee,Part Number SS-6-T.Caused by Crack Due to Improper Location of Heated Bar.Only One Part Out of 7396 Pieces in Forging Lot Was Found to Be Cracked.Affected Util,Notified
ML18106B0701999-02-16016 February 1999
LER 98-015-00:on 981208,inadvertent Discharge Through RHR Relief Valve During Startup Was Noted.Caused by Operator Performing Too Many Tasks Simultaneously.Appropriate Actions Have Been Taken IAW Policies & Procedures.With 990216 Ltr
ML18106B0551999-02-0101 February 1999
Part 21 Rept Re Possible Matl Defect in Swagelok Pipe Fitting Tee,Part Number SS-6-T.Defect Is Crack in Center of Forging.Analysis of Part Is Continuing & Further Details Will Be Provided IAW Ncr Timetables.Drawing of Part,Encl
ML18106B0561999-01-31031 January 1999
Monthly Operating Rept for Jan 1999 for Salem Generating Station,Unit 2.With 990212 Ltr
ML18106B0571999-01-31031 January 1999
Monthly Operating Rept for Jan 1999 for Salem Generating Station,Unit 1.With 990212 Ltr
ML20205P1671999-01-31031 January 1999
a POST-PLUME Phase, Federal Participation Exercise
ML18106B0441999-01-29029 January 1999
Part 21 Rept Re Possible Defect in Swagelok Pipe Fitting Tee Part Number SS-6-T.Caused by Crack in Center of Forging. Continuing Analysis of Part & Will Provide Details in Acoordance with NRC Timetables
ML18106B0491999-01-28028 January 1999
LER 98-007-01:on 980730,reactor Coolant Instrument Line through-wall Leak Was Noted.Caused by Transgranular Stress Corrosion Cracking.Replaced Affected Tubing.With 990128 Ltr
ML18106B0401999-01-18018 January 1999
LER 98-016-00:on 981219,ECCS Leakage Was Outside of Design Value.Caused by Leakage Past Seat of 21RH34 Manual Drain. Valve 21RH34 Was Reseated.With 990118 Ltr
ML18106B0251998-12-31031 December 1998
Monthly Operating Rept for Dec 1998 for Salem Unit 2.With 990115 Ltr
1999-09-30
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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
