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| {{#Wiki_filter: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:Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 Salem Generating Station August 3, 1992 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-70 DOCKET NO. 50-272 UNIT NO. 1 LICENSEE EVENT REPORT 92-015-00 August 3, 1992 This Licensee Event requirements of the 50. 73 (a) (2) (i) (B). thirty (30) days of Report is being submitted pursuant to the Code of Federal Regulations lOCFR This report is required to be issued within event discovery. | | |
| MJP:pc Distribution 120055 The Enerav Peoole 9208120067 920803 PDR ADOCK 05000272 5 PDR Sincerely yours,* c. A Vondra General Manager -Salem Operations 95-2189 (10M) 12-89 NRC FORM 366 (6-89) U.S. NUCLEAR REGULATORY COMMISSION APPROVED OMB NO. 3150-0104 LICENSEE EVENT REPORT (LERI FACILITY NAME (1) $alem Generatin9 Station -Unit 1 TITLE (4) EXPIRES: 4/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENT BRANCH (P-530). U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT 13150-0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503. I DOCKET NUMBER (2) I PAGE 131 0 15 I 0 I 0 I 0 I 2 17 I 2 1 I OF 0 14 Potential For Pressurizer PORV Loss of
| | SALEM GENERATING STATION LICENSE NO. DPR-70 DOCKET NO. 50-272 UNIT NO. 1 LICENSEE EVENT REPORT 92-015-00 This Licensee Event Report is being submitted pursuant to the requirements of the Code of Federal Regulations 10CFR |
| & Position Indication During An Accident EVENT DATE (5) LER NUMBER (6) REPORT DATE (71 OTHER FACILITIES INVOLVED (8) MONTH DAY YEAR YEAR :tJ tt MONTH DAY YEAR FACILITY NAMES DOCKET NUMBERISI Salem Unit 2 o I 5 I O I o I O 13 I 111 a I 1 a 12 9 2 912 -a 11 1 s OPERATING MODE (9) THIS REPORT IS SUBMITTED PURSUANT TO THE OF 10 CFR §: (Chock ono or moro of tho following) | | : 50. 73 (a) (2) (i) (B). This report is required to be issued within thirty (30) days of event discovery. |
| (11) 5 20.402(b) 20.405(c) 50.73(1)(2)(iv) 73.71(b) ,....__ -,....__ 20.406(1)(1 | | Sincerely yours,* |
| )(i) 50.38lcllll 50.73(1)(2)(v) 73.71 (c) ,....__ -,....__ 20.405(1)(1 | | : c. A Vondra General Manager - |
| )(Ii) 50.381cl(2) 50.73(1)(2) lvii) OTHER !Spacify in Abstroct ,....__ -,....__ bslow tJnd in Text, NRC Form 20.405(1)(1 | | Salem Operations MJP:pc Distribution 120055 The Enerav Peoole 9208120067 920803 PDR ADOCK 05000272 95-2189 (10M) 12-89 5 PDR |
| )(iii) x 50.73(1)(2)(i) 50.73loll2Jlviii)(A) 366A) ,...._ -20.405(0)(1 | | |
| )(Iv) 50.73(*112)(ii) 50.73(oll2Hviii)(B) | | NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION (6-89) APPROVED OMB NO. 3150-0104 EXPIRES: 4/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD LICENSEE EVENT REPORT (LERI COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENT BRANCH (P-530). U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT 13150-0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503. |
| ,....__ -20.406l*HllM 50,73(o)(2)1iiil 50.73(o)(2)(x)
| | FACILITY NAME (1) DOCKET NUMBER (2) I PAGE 131 |
| POWER I LEVEL -<101 01 0 I 0 a1111= LICENSEE CONTACT FOR THIS LER (12) NAME TELEPHONE NUMBER AREA CODE M. J. Pollack -LER Coordinator 6 I O I 9 3 I 3 I 9 1-I 2 IO I 21 2 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT 113) CAUSE SYSTEM COMPONENT I I I I I I I I MANUFAC* TUR ER I I I I I I SUPPLEMENTAL REPORT EXPECTED (14) n YES (If y6S, complot* EXPECTED SUBMISSION DATE) NO ABSTRACT (Limit to 1400 spactJs, i.e .* approximat11/y fifttJtJn single-space typewritten lines) 116) I I I I I I I I MANUFAC* TUR ER I I I I I I EXPECTED SUBMISSION
| | $alem Generatin9 Station - Unit 1 TITLE (4) |
| * DATE 1151 MONTH DAY YEAR I I I On July 2, 1992, it was discovered that during certain postulated design basis accidents (inside containment steamline break, inside containment feedwater line break or loss-of-coolant accident), loss of required valve control and position indication for the PRl and PR2 Pressurizer Power Operated Relief Valves (PORVs) may have resulted.
| | I0 15 I 0 I 0 I 0 I 2 17 I 2 1 OF 0 14 I Potential For Pressurizer PORV Loss of Cont~ol & Position Indication During An Accident EVENT DATE (5) LER NUMBER (6) REPORT DATE (71 OTHER FACILITIES INVOLVED (8) |
| This was due to the PR1/PR2 environmentally qualified (EQ) valve limit switches and control solenoids sharing a 125 VDC circuit breaker with non EQ pressure switches. | | MONTH DAY YEAR YEAR :tJ SE~~~~~~AL tt ~~~~~~ MONTH DAY YEAR FACILITY NAMES DOCKET NUMBERISI Salem Unit 2 o I 5 I O I o I O 13 I 111 a I 1 a 12 9 2 912 - a 11 1 s OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE R~QUIREMENTS OF 10 CFR §: (Chock ono or moro of tho following) (11) |
| Moisture intrusion in the non EQ components could have tripped the common circuit breaker resulting in loss of valve control and position indication. | | MODE (9) 5 20.402(b) |
| The non EQ components are pressure switches which monitor the normal air supply header pressure going to the PORV. In the event there is loss of control air, the pressure switches cause air accumulators (aux air supply) to provide backup air for the PORVs. Tech. Spec. 3.3.3.7 (both Salem Units) is applicable in Modes 1, 2, and 3. The PORVs are listed in this table. Contrary to this Tech. Spec., the PORVs position indication, via limit switch, may have become inoperable during a design basis accident. | | ,....__ 20.405(c) 50.73(1)(2)(iv) |
| The root cause of this event is "Management/QA Deficiency". | | ,....__ 73.71(b) |
| Due to inadequate design, inadequate design review, and inadequate administrative controls, these design changes were implemented without providing separation of the lE EQ components and non EQ components. | | POWER LEVEL I 01 0 I 0 20.406(1)(1 )(i) 20.405(1)(1 )(Ii) |
| Separation of the associated PORV circuits has been completed. | | ,....__ 50.38lcllll 50.381cl(2) 50.73(1)(2)(v) 50.73(1)(2) lvii) |
| A review will be performed to ensure there are no other cases similar to this event. NRC Form 366 (6-89) | | ,....__ 73.71 (c) |
| LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station Unit 1 D.QCKET NUMBER 5000272 PLANT AND SYSTEM IDENTIFICATION: | | OTHER !Spacify in Abstroct |
| Westinghouse | | <101 ,....__ |
| -Pressurized Water Reactor LER NUMBER 92-015-00 PAGE 2 of 4 Energy Industry Identification System (EIIS) codes are identified in the text as {xx} IDENTIFICATION OF OCCURRENCE: | | x - ,....__ bslow tJnd in Text, NRC Form a1111= |
| Potential for Pressurizer Power Operated Relief Valves loss of control and position indication during postulated accidents Discovery Date: 7/2/92 Report Date: 8/03/92 This report was initiated by Incident Report Nos. 92-420 and 92-421. CONDITIONS PRIOR TO OCCURRENCE: | | 20.405(1)(1 )(iii) 50.73(1)(2)(i) 50.73loll2Jlviii)(A) 366A) 20.405(0)(1 )(Iv) 50.73(*112)(ii) 50.73(oll2Hviii)(B) 20.406l*HllM 50,73(o)(2)1iiil |
| Unit 1: Mode 5 {Cold Shutdown) | | - 50.73(o)(2)(x) |
| Unit 2: Mode 5 DESCRIPTION OF OCCURRENCE: | | LICENSEE CONTACT FOR THIS LER (12) |
| On July 2, 1992, it was discovered that during certain postulated design basis accidents (inside containment steamline break, inside containment feedwater line break or loss-of-coolant accident), loss of required valve control and position indication for the PR1 and PR2 Pressurizer Power Operated Relief Valves (PORVs) {AB} may have resulted. | | NAME TELEPHONE NUMBER AREA CODE M. J. Pollack - LER Coordinator 6 I O I 9 3 I 3 I 9 1- I 2 IO I 21 2 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT 113) |
| This was due to the limit switches and control solenoids sharing a 125 VDC circuit breaker with non environmentally qualified (EQ) components. | | MANUFAC* MANUFAC* |
| With this configuration, moisture intrusion in the non-EQ components could have tripped the common circuit breaker, resulting in loss of valve control and position indication. | | CAUSE SYSTEM COMPONENT TUR ER TUR ER I I I I I I I I I I I I I I I I I I I I I I I I I I I I SUPPLEMENTAL REPORT EXPECTED (14) MONTH DAY YEAR EXPECTED n ~ |
| This situation was discovered during preparation of an Engineering Evaluation related to EQ design classifications. | | SUBMISSION |
| The non EQ components are pressure switches which monitor the normal air supply pressure in the headers going to the PORV. Technical Specification 3.3.3.7 (for both Salem Units) states, "The accident monitoring instrumentation channels shown in Table 3.3-11 shall be operable." It is applicable in Modes 1, 2, and 3. The PORVs are listed in this table. Contrary to this Technical Specification, the prior configuration of the PORV's position indication, via limit could have become inoperable during a design basis accident. | | * DATE 1151 YES (If y6S, complot* EXPECTED SUBMISSION DATE) NO I I I ABSTRACT (Limit to 1400 spactJs, i.e .* approximat11/y fifttJtJn single-space typewritten lines) 116) |
| | On July 2, 1992, it was discovered that during certain postulated design basis accidents (inside containment steamline break, inside containment feedwater line break or loss-of-coolant accident), loss of required valve control and position indication for the PRl and PR2 Pressurizer Power Operated Relief Valves (PORVs) may have resulted. This was due to the PR1/PR2 environmentally qualified (EQ) valve limit switches and control solenoids sharing a 125 VDC circuit breaker with non EQ pressure switches. Moisture intrusion in the non EQ components could have tripped the common circuit breaker resulting in loss of valve control and position indication. The non EQ components are pressure switches which monitor the normal air supply header pressure going to the PORV. In the event there is loss of control air, the pressure switches cause air accumulators (aux air supply) to provide backup air for the PORVs. Tech. |
| | Spec. 3.3.3.7 (both Salem Units) is applicable in Modes 1, 2, and 3. The PORVs are listed in this table. Contrary to this Tech. Spec., the PORVs position indication, via limit switch, may have become inoperable during a design basis accident. The root cause of this event is "Management/QA Deficiency". Due to inadequate design, inadequate design review, and inadequate administrative controls, these design changes were implemented without providing separation of the lE EQ components and non EQ components. Separation of the associated PORV circuits has been completed. A review will be performed to ensure there are no other cases similar to this event. |
| | NRC Form 366 (6-89) |
| | |
| | LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station D.QCKET NUMBER LER NUMBER PAGE Unit 1 5000272 92-015-00 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: |
| | Potential for Pressurizer Power Operated Relief Valves loss of control and position indication during postulated accidents Discovery Date: 7/2/92 Report Date: 8/03/92 This report was initiated by Incident Report Nos. 92-420 and 92-421. |
| | CONDITIONS PRIOR TO OCCURRENCE: |
| | Unit 1: Mode 5 {Cold Shutdown) |
| | Unit 2: Mode 5 DESCRIPTION OF OCCURRENCE: |
| | On July 2, 1992, it was discovered that during certain postulated design basis accidents (inside containment steamline break, inside containment feedwater line break or loss-of-coolant accident), loss of required valve control and position indication for the PR1 and PR2 Pressurizer Power Operated Relief Valves (PORVs) {AB} may have resulted. This was due to the limit switches and control solenoids sharing a 125 VDC circuit breaker with non environmentally qualified (EQ) components. With this configuration, moisture intrusion in the non-EQ components could have tripped the common circuit breaker, resulting in loss of valve control and position indication. This situation was discovered during preparation of an Engineering Evaluation related to EQ design classifications. |
| | The non EQ components are pressure switches which monitor the normal air supply pressure in the headers going to the PORV. |
| | Technical Specification 3.3.3.7 (for both Salem Units) states, "The accident monitoring instrumentation channels shown in Table 3.3-11 shall be operable." It is applicable in Modes 1, 2, and 3. The PORVs are listed in this table. Contrary to this Technical Specification, the prior configuration of the PORV's position indication, via limit s~itch, could have become inoperable during a design basis accident. |
| APPARENT CAUSE OF OCCURRENCE: | | APPARENT CAUSE OF OCCURRENCE: |
| The root cause of this event is "Management/QA Deficiency", per NUREG 1022, "Licensee Event Report system" (i.e., potential for the loss of LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating station Unit 1 DOCKET NUMBER 5000272 APPARENT CAUSE OF OCCURRENCE: (cont'd) LER NUMBER 92-015-00 PAGE 3 of 4 PORV position indication during a design basis event as required by Regulatory Guide 1.97, "Instrumentation For Light-Water Cooled Nuclear Power Plants To Assess Plant And Environs Conditions During And Following An Accident"). | | The root cause of this event is "Management/QA Deficiency", per NUREG 1022, "Licensee Event Report system" (i.e., potential for the loss of |
| | |
| | LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating station DOCKET NUMBER LER NUMBER PAGE Unit 1 5000272 92-015-00 3 of 4 APPARENT CAUSE OF OCCURRENCE: (cont'd) |
| | PORV position indication during a design basis event as required by Regulatory Guide 1.97, "Instrumentation For Light-Water Cooled Nuclear Power Plants To Assess Plant And Environs Conditions During And Following An Accident"). |
| Design modifications involving PORV enhancement were completed in 1981 for Unit 1 and 1984 for Unit 2. Due to inadequate design, inadequate design review, and inadequate administrative controls, these design changes were implemented without providing separation of the lE EQ components (limit switch upgrades) and non EQ components (pressure switches) on the same circuit without considering the possible effects during a design basis event. The pressure switches are not required to be in the EQ program since the valves can only be considered as auxiliary devices not required to mitigate accident conditions. | | Design modifications involving PORV enhancement were completed in 1981 for Unit 1 and 1984 for Unit 2. Due to inadequate design, inadequate design review, and inadequate administrative controls, these design changes were implemented without providing separation of the lE EQ components (limit switch upgrades) and non EQ components (pressure switches) on the same circuit without considering the possible effects during a design basis event. The pressure switches are not required to be in the EQ program since the valves can only be considered as auxiliary devices not required to mitigate accident conditions. |
| ANALYSIS OF OCCURRENCE: | | ANALYSIS OF OCCURRENCE: |
| In Modes 1, 2, and 3, the PORVs relieve Reactor Coolant System (RCS) {AB} pressure during all design transients up to and including the design step load decrease with steam dump operation. | | In Modes 1, 2, and 3, the PORVs relieve Reactor Coolant System (RCS) |
| The PORVs minimize the possibility of an undesirable opening of the spring loaded pressurizer code safety valves. However, they are not credited in the plant design basis analysis. | | {AB} pressure during all design transients up to and including the design step load decrease with steam dump operation. The PORVs minimize the possibility of an undesirable opening of the spring loaded pressurizer code safety valves. However, they are not credited in the plant design basis analysis. |
| As stated previously, Technical Specification 3.3.3.7 requires the PORVs position indication to be operable in the event of a design basis accident. | | As stated previously, Technical Specification 3.3.3.7 requires the PORVs position indication to be operable in the event of a design basis accident. This is in addition to any of three other position indicators (i.e., tailpipe temperatures, Pressurizer Relief Tank Temperature or Pressurizer Relief Tank Pressure). |
| This is in addition to any of three other position indicators (i.e., tailpipe temperatures, Pressurizer Relief Tank Temperature or Pressurizer Relief Tank Pressure). | | In the 125 VDC Cabinet, one circuit breaker provided power to each PORV PR1/PR2 for control, position indication and auxiliary air supply control. The pressure switches in the auxiliary air control circuit are exempt from the EQ program. Therefore, a postulated design basis accident could have caused a short circuit. This would have tripped the breaker resulting in loss of PR1/PR2 control due to the loss of power to solenoid valves, position indication, and loss of the auxiliary air supply solenoid valves. |
| In the 125 VDC Cabinet, one circuit breaker provided power to each PORV PR1/PR2 for control, position indication and auxiliary air supply control. The pressure switches in the auxiliary air control circuit are exempt from the EQ program. Therefore, a postulated design basis accident could have caused a short circuit. This would have tripped the breaker resulting in loss of PR1/PR2 control due to the loss of power to solenoid valves, position indication, and loss of the auxiliary air supply solenoid valves. The PORVs are not credited in the accident analysis; therefore, the health and safety of the public would not have been affected as a result of this event. However, since the accident monitoring channels for PORV position indication may not have remained operable during a postulated design basis accident, per Technical Specification 3.3.3.7, this event is reportable in accordance with Code of Federal Regulations 10CFR50.73(a) | | The PORVs are not credited in the accident analysis; therefore, the health and safety of the public would not have been affected as a result of this event. However, since the accident monitoring channels for PORV position indication may not have remained operable during a postulated design basis accident, per Technical Specification 3.3.3.7, this event is reportable in accordance with Code of Federal Regulations 10CFR50.73(a) (2) (i) (B). |
| (2) (i) (B). This concern was identified by Quality Assurance (QA) during a review of lE equipment located in harsh environments cross checked against equipment on the Environmental Equipment Master List. Those cases LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station Unit 1 DOCKET NUMBER 5000272 ANALYSIS OF OCCURRENCE: (cont'd) LER NUMBER 92-015-00 PAGE 4 of 4 identified by QA have been resolved. | | This concern was identified by Quality Assurance (QA) during a review of lE equipment located in harsh environments cross checked against equipment on the Environmental Equipment Master List. Those cases |
| Engineering evaluations were prepared discussing these cases. Since 1984, the design change process controls have been significantly enhanced such that recurrence of this event is unlikely. | | |
| The process now includes specialty reviews requiring Program Analysis Group (EQ engineers) interface. | | LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station DOCKET NUMBER LER NUMBER PAGE Unit 1 5000272 92-015-00 4 of 4 ANALYSIS OF OCCURRENCE: (cont'd) identified by QA have been resolved. Engineering evaluations were prepared discussing these cases. |
| CORRECTIVE ACTION: The PORV circuits identified as deficient have been modified to separate the PRl and PR2 non EQ components from the PRl and PR2 EQ controls and indication circuits (limit switches PR1-LS1/LS2 and PR2-LS1/LS2 and associated relays). This event has been discussed with applicable Nuclear Electrical Engineering Department personnel. | | Since 1984, the design change process controls have been significantly enhanced such that recurrence of this event is unlikely. The process now includes specialty reviews requiring Program Analysis Group (EQ engineers) interface. |
| A review will be performed to ensure there are no other EQ components sharing power suppiies with non EQ components which could compromise the function of the EQ component. | | CORRECTIVE ACTION: |
| Additional corrective action will be taken, based upon results of the General anager Salem Operations MJP:pc SORC Mtg. 92-088}} | | The PORV circuits identified as deficient have been modified to separate the PRl and PR2 non EQ components from the PRl and PR2 EQ controls and indication circuits (limit switches PR1-LS1/LS2 and PR2-LS1/LS2 and associated relays). |
| | This event has been discussed with applicable Nuclear Electrical Engineering Department personnel. |
| | A review will be performed to ensure there are no other EQ components sharing power suppiies with non EQ components which could compromise the function of the EQ component. Additional corrective action will be taken, based upon results of the review,/~~~propriate. |
| | ~~~ |
| | General anager Salem Operations MJP:pc SORC Mtg. 92-088}} |
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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: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
Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 Salem Generating Station August 3, 1992 U. S. Nuclear Regulatory Commission Document control Desk Washington, DC 20555
Dear Sir:
SALEM GENERATING STATION LICENSE NO. DPR-70 DOCKET NO. 50-272 UNIT NO. 1 LICENSEE EVENT REPORT 92-015-00 This Licensee Event Report is being submitted pursuant to the requirements of the Code of Federal Regulations 10CFR
- 50. 73 (a) (2) (i) (B). This report is required to be issued within thirty (30) days of event discovery.
Sincerely yours,*
- c. A Vondra General Manager -
Salem Operations MJP:pc Distribution 120055 The Enerav Peoole 9208120067 920803 PDR ADOCK 05000272 95-2189 (10M) 12-89 5 PDR
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION (6-89) APPROVED OMB NO. 3150-0104 EXPIRES: 4/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD LICENSEE EVENT REPORT (LERI COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENT BRANCH (P-530). U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT 13150-0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.
FACILITY NAME (1) DOCKET NUMBER (2) I PAGE 131
$alem Generatin9 Station - Unit 1 TITLE (4)
I0 15 I 0 I 0 I 0 I 2 17 I 2 1 OF 0 14 I Potential For Pressurizer PORV Loss of Cont~ol & Position Indication During An Accident EVENT DATE (5) LER NUMBER (6) REPORT DATE (71 OTHER FACILITIES INVOLVED (8)
MONTH DAY YEAR YEAR :tJ SE~~~~~~AL tt ~~~~~~ MONTH DAY YEAR FACILITY NAMES DOCKET NUMBERISI Salem Unit 2 o I 5 I O I o I O 13 I 111 a I 1 a 12 9 2 912 - a 11 1 s OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE R~QUIREMENTS OF 10 CFR §: (Chock ono or moro of tho following) (11)
MODE (9) 5 20.402(b)
,....__ 20.405(c) 50.73(1)(2)(iv)
,....__ 73.71(b)
POWER LEVEL I 01 0 I 0 20.406(1)(1 )(i) 20.405(1)(1 )(Ii)
,....__ 50.38lcllll 50.381cl(2) 50.73(1)(2)(v) 50.73(1)(2) lvii)
,....__ 73.71 (c)
OTHER !Spacify in Abstroct
<101 ,....__
x - ,....__ bslow tJnd in Text, NRC Form a1111=
20.405(1)(1 )(iii) 50.73(1)(2)(i) 50.73loll2Jlviii)(A) 366A) 20.405(0)(1 )(Iv) 50.73(*112)(ii) 50.73(oll2Hviii)(B) 20.406l*HllM 50,73(o)(2)1iiil
- 50.73(o)(2)(x)
LICENSEE CONTACT FOR THIS LER (12)
NAME TELEPHONE NUMBER AREA CODE M. J. Pollack - LER Coordinator 6 I O I 9 3 I 3 I 9 1- I 2 IO I 21 2 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT 113)
MANUFAC* MANUFAC*
CAUSE SYSTEM COMPONENT TUR ER TUR ER I I I I I I I I I I I I I I I I I I I I I I I I I I I I SUPPLEMENTAL REPORT EXPECTED (14) MONTH DAY YEAR EXPECTED n ~
SUBMISSION
- DATE 1151 YES (If y6S, complot* EXPECTED SUBMISSION DATE) NO I I I ABSTRACT (Limit to 1400 spactJs, i.e .* approximat11/y fifttJtJn single-space typewritten lines) 116)
On July 2, 1992, it was discovered that during certain postulated design basis accidents (inside containment steamline break, inside containment feedwater line break or loss-of-coolant accident), loss of required valve control and position indication for the PRl and PR2 Pressurizer Power Operated Relief Valves (PORVs) may have resulted. This was due to the PR1/PR2 environmentally qualified (EQ) valve limit switches and control solenoids sharing a 125 VDC circuit breaker with non EQ pressure switches. Moisture intrusion in the non EQ components could have tripped the common circuit breaker resulting in loss of valve control and position indication. The non EQ components are pressure switches which monitor the normal air supply header pressure going to the PORV. In the event there is loss of control air, the pressure switches cause air accumulators (aux air supply) to provide backup air for the PORVs. Tech.
Spec. 3.3.3.7 (both Salem Units) is applicable in Modes 1, 2, and 3. The PORVs are listed in this table. Contrary to this Tech. Spec., the PORVs position indication, via limit switch, may have become inoperable during a design basis accident. The root cause of this event is "Management/QA Deficiency". Due to inadequate design, inadequate design review, and inadequate administrative controls, these design changes were implemented without providing separation of the lE EQ components and non EQ components. Separation of the associated PORV circuits has been completed. A review will be performed to ensure there are no other cases similar to this event.
NRC Form 366 (6-89)
LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station D.QCKET NUMBER LER NUMBER PAGE Unit 1 5000272 92-015-00 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:
Potential for Pressurizer Power Operated Relief Valves loss of control and position indication during postulated accidents Discovery Date: 7/2/92 Report Date: 8/03/92 This report was initiated by Incident Report Nos.92-420 and 92-421.
CONDITIONS PRIOR TO OCCURRENCE:
Unit 1: Mode 5 {Cold Shutdown)
Unit 2: Mode 5 DESCRIPTION OF OCCURRENCE:
On July 2, 1992, it was discovered that during certain postulated design basis accidents (inside containment steamline break, inside containment feedwater line break or loss-of-coolant accident), loss of required valve control and position indication for the PR1 and PR2 Pressurizer Power Operated Relief Valves (PORVs) {AB} may have resulted. This was due to the limit switches and control solenoids sharing a 125 VDC circuit breaker with non environmentally qualified (EQ) components. With this configuration, moisture intrusion in the non-EQ components could have tripped the common circuit breaker, resulting in loss of valve control and position indication. This situation was discovered during preparation of an Engineering Evaluation related to EQ design classifications.
The non EQ components are pressure switches which monitor the normal air supply pressure in the headers going to the PORV.
Technical Specification 3.3.3.7 (for both Salem Units) states, "The accident monitoring instrumentation channels shown in Table 3.3-11 shall be operable." It is applicable in Modes 1, 2, and 3. The PORVs are listed in this table. Contrary to this Technical Specification, the prior configuration of the PORV's position indication, via limit s~itch, could have become inoperable during a design basis accident.
APPARENT CAUSE OF OCCURRENCE:
The root cause of this event is "Management/QA Deficiency", per NUREG 1022, "Licensee Event Report system" (i.e., potential for the loss of
LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating station DOCKET NUMBER LER NUMBER PAGE Unit 1 5000272 92-015-00 3 of 4 APPARENT CAUSE OF OCCURRENCE: (cont'd)
PORV position indication during a design basis event as required by Regulatory Guide 1.97, "Instrumentation For Light-Water Cooled Nuclear Power Plants To Assess Plant And Environs Conditions During And Following An Accident").
Design modifications involving PORV enhancement were completed in 1981 for Unit 1 and 1984 for Unit 2. Due to inadequate design, inadequate design review, and inadequate administrative controls, these design changes were implemented without providing separation of the lE EQ components (limit switch upgrades) and non EQ components (pressure switches) on the same circuit without considering the possible effects during a design basis event. The pressure switches are not required to be in the EQ program since the valves can only be considered as auxiliary devices not required to mitigate accident conditions.
ANALYSIS OF OCCURRENCE:
In Modes 1, 2, and 3, the PORVs relieve Reactor Coolant System (RCS)
{AB} pressure during all design transients up to and including the design step load decrease with steam dump operation. The PORVs minimize the possibility of an undesirable opening of the spring loaded pressurizer code safety valves. However, they are not credited in the plant design basis analysis.
As stated previously, Technical Specification 3.3.3.7 requires the PORVs position indication to be operable in the event of a design basis accident. This is in addition to any of three other position indicators (i.e., tailpipe temperatures, Pressurizer Relief Tank Temperature or Pressurizer Relief Tank Pressure).
In the 125 VDC Cabinet, one circuit breaker provided power to each PORV PR1/PR2 for control, position indication and auxiliary air supply control. The pressure switches in the auxiliary air control circuit are exempt from the EQ program. Therefore, a postulated design basis accident could have caused a short circuit. This would have tripped the breaker resulting in loss of PR1/PR2 control due to the loss of power to solenoid valves, position indication, and loss of the auxiliary air supply solenoid valves.
The PORVs are not credited in the accident analysis; therefore, the health and safety of the public would not have been affected as a result of this event. However, since the accident monitoring channels for PORV position indication may not have remained operable during a postulated design basis accident, per Technical Specification 3.3.3.7, this event is reportable in accordance with Code of Federal Regulations 10CFR50.73(a) (2) (i) (B).
This concern was identified by Quality Assurance (QA) during a review of lE equipment located in harsh environments cross checked against equipment on the Environmental Equipment Master List. Those cases
LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station DOCKET NUMBER LER NUMBER PAGE Unit 1 5000272 92-015-00 4 of 4 ANALYSIS OF OCCURRENCE: (cont'd) identified by QA have been resolved. Engineering evaluations were prepared discussing these cases.
Since 1984, the design change process controls have been significantly enhanced such that recurrence of this event is unlikely. The process now includes specialty reviews requiring Program Analysis Group (EQ engineers) interface.
CORRECTIVE ACTION:
The PORV circuits identified as deficient have been modified to separate the PRl and PR2 non EQ components from the PRl and PR2 EQ controls and indication circuits (limit switches PR1-LS1/LS2 and PR2-LS1/LS2 and associated relays).
This event has been discussed with applicable Nuclear Electrical Engineering Department personnel.
A review will be performed to ensure there are no other EQ components sharing power suppiies with non EQ components which could compromise the function of the EQ component. Additional corrective action will be taken, based upon results of the review,/~~~propriate.
~~~
General anager Salem Operations MJP:pc SORC Mtg.92-088