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| {{#Wiki_filter:* | | {{#Wiki_filter:PS~G Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038-0236 Nuclear Business Unit DEC 18 1995 LR-N95231 U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen: |
| * Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038-0236 Nuclear Business Unit DEC 18 1995 LR-N95231 U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:
| | LICENSEE EVENT REPORT 272/95-012-000 SALEM GENERATING STATION - UNIT 1 FACILITY OPERATING LICENSE NO. DPR-70 DOCKET NO. 50-272 This Licensee Event Report entitled "Adequacy of Turbine Driven Auxiliary Feed Water Pump Enclosures" is being submitted pursuant to the requirements of the Code of Federal Regulations 10CFR50.73(a) (1). |
| LICENSEE EVENT REPORT 272/95-012-000 SALEM GENERATING STATION -UNIT 1 FACILITY OPERATING LICENSE NO. DPR-70 DOCKET NO. 50-272 This Licensee Event Report entitled "Adequacy of Turbine Driven Auxiliary Feed Water Pump Enclosures" is being submitted pursuant to the requirements of the Code of Federal Regulations 10CFR50.73(a) | | Sincerely, . |
| (1). SORC Mtg. 95-149 .Attachment JHA/tcp C Distribution LER File n , ... 9512270322 951218 *PDR ADOCK 05000272 S PDR Sincerely, . -l-2 <? tl2ui..__ | | -l-2 <? tl2ui..__ |
| Clad. Warren General Manager -Salem Operations 95*2168 REV. 6/94
| | Clad. Warren General Manager - |
| *
| | Salem Operations SORC Mtg. 95-149 |
| * Attachment A PSE&G Commitments for LER 272/95-012-000 The following items represent PSE&G commitments made to the Nuclear Regulatory Commission related to LER 272/95-012-000.
| | .Attachment JHA/tcp C Distribution LER File Or*~.-. n , ... |
| The commitments are' as follows: Design change alternatives are under consideration to eliminate the potential for overpressurization of the TDAFP enclosure. | | 9512270322 951218 |
| The design change will be described in the LER supplement and will be implemented prior to restart of the units. Other HELB calculations involving similar plant configurations will be reviewed to verify the correctness of the input assumptions. | | *PDR ADOCK 05000272 S PDR 95*2168 REV. 6/94 |
| A schedule for completion of this review will be provided in the supplement to this LER'. Any additional corrective actions identified as a result of the dontinuing investigation into this occurence will be provided in the supplement. | | |
| NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 14-95) EXPIRES 04/30/98 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITTi THIS MANDATORY INFORMATION COLLECTION REQUEST: 50.0 HRS. LICENSEE EVENT REPORT (LER) REPORTED LESSONS LEARNED ARE INCORPORATED INTO THE LICENSING PROCESS AND FED BACK TO INDUSTflY. | | Attachment A PSE&G Commitments for LER 272/95-012-000 The following items represent PSE&G commitments made to the Nuclear Regulatory Commission related to LER 272/95-012-000. The commitments are' as follows: |
| FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE (See reverse for required number of INFORMATION AND RECORDS MANAGEMENT BRANCH IT-6 F33), digits/characters for each block) U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT FACILITY NAME 111 DOCKET NUMB9' 121 PAGE 131 Salem Generating Station -Unit 1 05000272 1 OF 3 TITLE 141 Adequacy of Turbine Driven Auxiliary Feed Water Pump Enclosures EVENT DATE 151 LER NUMBER 161 REPORT DA TE 171 OTHER FACILITIES INVOLVED 181 YEAR I SEQUENTIAL I REVISION FACILITY NAME DOCKET NUMBER MONTH DAY YEAR NUMBER NUMBER MONTH DAY YEAR Salem Generating Station -Unit 2 05000311 FACILITY NAME DOCKET NUMBER 12 11 76 95 --012 --000 12 18 95 -I OPERATING I | | Design change alternatives are under consideration to eliminate the potential for overpressurization of the TDAFP enclosure. The design change will be described in the LER supplement and will be implemented prior to restart of the units. |
| * I THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: !Check one or more) 1111 MODE 191 20.2201 (b) 20.2203(a)(2Jlv) 50.731a)(2)(i) 50.73(aJl2)(viiiJ I POWER I I 20.2203(aJl1 I 20.2203(a)(3Jlil x 50.73(a)(2)(ii) 50.73(a)(2)(x)
| | Other HELB calculations involving similar plant configurations will be reviewed to verify the correctness of the input assumptions. A schedule for completion of this review will be provided in the supplement to this LER'. |
| LEVEL 1101 0 20.2203(a)l2)(i) 20.2203(a)(3Jliil | | Any additional corrective actions identified as a result of the dontinuing investigation into this occurence will be provided in the supplement. |
| : 50. 7 3 (a)(2)(iii) 73.71 -20 .2203 (a)(2)(ii) 20.2203(a)(4) 50.73(a)(2lli"'.I OTHER 20.2203(a)(2)(iii) 50.36(c)(l I 50.73(a)(2)(v)
| | |
| Specify in Abstract below or in NRC Form 366A 20.2203 (a)(2)(iv) 50.36(c)(2) | | NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 14-95) EXPIRES 04/30/98 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITTi THIS MANDATORY INFORMATION COLLECTION REQUEST: 50.0 HRS. |
| : 50. 73 (a)(2)(vii)
| | REPORTED LESSONS LEARNED ARE INCORPORATED INTO THE LICENSEE EVENT REPORT (LER) LICENSING PROCESS AND FED BACK TO INDUSTflY. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE (See reverse for required number of INFORMATION AND RECORDS MANAGEMENT BRANCH IT-6 F33), |
| LICENSEE CONTACT FOR THIS LER (12) NAME TELEPHONE NUMBER (Include Are. Code) Greg Cranston, Manager -Nuclear Engineering, 609 339 1955 Mee anical --*** 1m-* l'!'l'F. nw1 J,T'NF. 'Ii' "'l'D | | U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC digits/characters for each block) 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT FACILITY NAME 111 DOCKET NUMB9' 121 PAGE 131 Salem Generating Station - Unit 1 05000272 1 OF 3 TITLE 141 Adequacy of Turbine Driven Auxiliary Feed Water Pump Enclosures 171 I |
| -* *m -* 'N .. *1* li'l TT.TTRF. | | EVENT DATE 151 LER NUMBER 161 REPORT DATE OTHER FACILITIES INVOLVED 181 MONTH DAY YEAR YEAR I SEQUENTIAL NUMBER REVISION NUMBER MONTH DAY YEAR FACILITY NAME Salem Generating Station - Unit 2 FACILITY NAME DOCKET NUMBER 05000311 12 11 76 95 -- 012 -- 000 12 18 95 DOCKET NUMBER I |
| Tl l'lfln T'tJ 'l'HTA 'DF.Pni>rri , , ":l \ CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE I CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE TONPRDS TO NPRDS SUPPLEMENTAL REPORT EXPECTED (141 EXPECTED MONTH DAY YEAR IYES 'NO SUBMISSION 02 15 96 X (If yes, complete EXPECTED SUBMISSION DATE). DATE 1151 ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (161 In early November 1995 it was discovered that assumptions in the High Energy Line Break (HELB) analysis for the turbine driven auxiliary feed water pump (TDAFP) enclosure did not match as-built conditions and could allow pressure in the enclosure. | | I I* |
| to exceed the enclosure design pressure during a HELB. A detailed analysis is being performed to determine the extent of the potential overpressurization and the affect on the structural integrity of the TDAFP enclosure. | | OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: !Check one or more) 1111 MODE 191 20.2201 (b) 20.2203(a)(2Jlv) 50.731a)(2)(i) 50.73(aJl2)(viiiJ 20.2203(aJl1 I 20.2203(a)(3Jlil x 50.73(a)(2)(ii) 50.73(a)(2)(x) |
| The cause of this occurrence is attributed to inadequate verification of assumptions in the calculations performed to evaluate previously identified as-built design deficiencies and inaccurate design drawings. | | I POWER LEVEL 1101 I0 I 20.2203(a)l2)(i) 20.2203(a)(3Jliil 50. 7 3 (a)(2)(iii) 73.71 20 .2203 (a)(2)(ii) 20.2203(a)(4) 50.73(a)(2lli"'.I OTHER 20.2203(a)(2)(iii) 50.36(c)(l I 50.73(a)(2)(v) Specify in Abstract below or in NRC Form 366A 20.2203 (a)(2)(iv) 50.36(c)(2) 50. 73 (a)(2)(vii) |
| Corrective actions include a review of other HELB calculation assumptions and evaluation of design changes to eliminate the potential TDAFP enclosure overpressure. | | LICENSEE CONTACT FOR THIS LER (12) |
| This condition is being reported in accordance with 10CFR50.73(a) | | NAME TELEPHONE NUMBER (Include Are. Code) |
| (2) (ii) . A four hour 10CFR50.72 notification was made on November 16, 1995. NRC FORM 366 14-95) | | Greg Cranston, Manager - Nuclear Engineering, 609 - 339 - 1955 Mee anical |
| *
| | *** 1m-* l'!'l'F. nw1 J,T'NF. 'Ii' "'l'D F.lH~H -* *m - * 'N .. ~n *1* li'l TT.TTRF. r'l~R~'D Tl l'lfln T'tJ 'l'HTA 'DF.Pni>rri , , ":l \ |
| * NRC FORM 366A (4-95) U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER) TEXT CONTINUATION FACILITY NAME 111 DOCKET LER NUMBER 161 Salem Generating Station -Unit 1 05000272 YEAR I SEQUENTIAL I REVISION NUMBER NUMBER 95 --012 --000 TEXT (If mor8 spac8 is rtJquirtJd, ustJ additional copies of NRC Form 366AJ (171 PLANT AND SYSTEM IDENTIFICATION Westinghouse
| | REPORTABLE REPORTABLE CAUSE SYSTEM COMPONENT MANUFACTURER CAUSE SYSTEM COMPONENT MANUFACTURER TONPRDS TO NPRDS X IYES (If yes, SUPPLEMENTAL REPORT EXPECTED (141 complete EXPECTED SUBMISSION DATE). |
| -Pressurized Water Reactor Auxiliary Feedwater System (AF) {BA}* Auxiliary Building Ventilation System (ABS) {VF} PAGE 131 2 OF 3 | | I 'NO EXPECTED SUBMISSION DATE 1151 MONTH 02 DAY 15 YEAR 96 ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (161 In early November 1995 it was discovered that assumptions in the High Energy Line Break (HELB) analysis for the turbine driven auxiliary feed water pump (TDAFP) enclosure did not match as-built conditions and could allow pressure in the enclosure. to exceed the enclosure design pressure during a HELB. A detailed analysis is being performed to determine the extent of the potential overpressurization and the affect on the structural integrity of the TDAFP enclosure. The cause of this occurrence is attributed to inadequate verification of assumptions in the calculations performed to evaluate previously identified as-built design deficiencies and inaccurate design drawings. Corrective actions include a review of other HELB calculation assumptions and evaluation of design changes to eliminate the potential TDAFP enclosure overpressure. |
| | This condition is being reported in accordance with 10CFR50.73(a) (2) (ii) . |
| | A four hour 10CFR50.72 notification was made on November 16, 1995. |
| | NRC FORM 366 14-95) |
| | |
| | NRC FORM 366A (4-95) |
| | * LICENSEE EVENT REPORT (LER) |
| | * U.S. NUCLEAR REGULATORY COMMISSION TEXT CONTINUATION FACILITY NAME 111 DOCKET LER NUMBER 161 PAGE 131 YEAR I SEQUENTIAL NUMBER I REVISION NUMBER Salem Generating Station - Unit 1 05000272 2 OF 3 95 -- 012 -- 000 TEXT (If mor8 spac8 is rtJquirtJd, ustJ additional copies of NRC Form 366AJ (171 PLANT AND SYSTEM IDENTIFICATION Westinghouse - Pressurized Water Reactor Auxiliary Feedwater System (AF) {BA}* |
| | Auxiliary Building Ventilation System (ABS) {VF} |
| * Energy Industry Identification system (EIIS) codes and component function identifier codes appear in the text as {SS/CCC}. | | * Energy Industry Identification system (EIIS) codes and component function identifier codes appear in the text as {SS/CCC}. |
| IDENTIFICATION OF OCCURRENCE Event Date: Unit 1: December 11, 1976 (Initial Plant Criticality) | | IDENTIFICATION OF OCCURRENCE Event Date: Unit 1: December 11, 1976 (Initial Plant Criticality) |
| Unit 2: August 2, 1980 (Initial Plant Criticality) | | Unit 2: August 2, 1980 (Initial Plant Criticality) |
| Date Determined to be Reportable: | | Date Determined to be Reportable: November 16, 1995 CONDITIONS PRIOR TO OCCURRENCE Unit 1: Defueled, o % Reactor Power Unit 2: Mode 5, o % Reactor Power DESCRIPTION OF OCCURRENCE - |
| November 16, 1995 CONDITIONS PRIOR TO OCCURRENCE Unit 1: Unit 2: Defueled, Mode 5, o % Reactor Power o % Reactor Power DESCRIPTION OF OCCURRENCE | | In early November 1995 it was discovered that assumptions in the HE~B analysis for the turbine driven auxiliary feed water pump (TDAFP) {BA/P} |
| -In early November 1995 it was discovered that assumptions in the analysis for the turbine driven auxiliary feed water pump (TDAFP) {BA/P} enclosure did not match as-built conditions and could potentially al1ow pressure in the enclosure to exceed the enclosure design pressure during a HELB. The ABS 6 damper {VF/DMP} is located -in the wall separating the TDAFP enclosure and the adjacent pipe chase. The ABS 6 damper automatically opens to protect the TDAFP enclosure from overpressurization during a HELB in the enclosure.
| | enclosure did not match as-built conditions and could potentially al1ow pressure in the enclosure to exceed the enclosure design pressure during a HELB. |
| The HELB analysis performed for a steam line break in the TDAFP enclosure implicitly assumed that the ABS 6 damper opened instantaneously when the pressure in the enclosure reached the setpoint. | | The ABS 6 damper {VF/DMP} is located -in the wall separating the TDAFP enclosure and the adjacent pipe chase. The ABS 6 damper automatically opens to protect the TDAFP enclosure from overpressurization during a HELB in the enclosure. The HELB analysis performed for a steam line break in the TDAFP enclosure implicitly assumed that the ABS 6 damper opened instantaneously when the pressure in the enclosure reached the setpoint. |
| Significant delays actually exist between the pressure reaching the setpoint .and the initiation of the opening stroke of ABS 6 damper, and the damper reaching the full open position. | | Significant delays actually exist between the pressure reaching the setpoint .and the initiation of the opening stroke of ABS 6 damper, and the damper reaching the full open position. The analysis also implicitly assumed that the damper full open position provided 100 percent of the damper opening for pressure relief. In the full open position the damper is actually open only 45 degrees, limiting the area available to less than 100 percent. It has been determined that with these delays and the less than fully open damper, the peak pressure in the enclosure following a steam line break may exceed the design pressure of the TDAFP enclosure. |
| The analysis also implicitly assumed that the damper full open position provided 100 percent of the damper opening for pressure relief. In the full open position the damper is actually open only 45 degrees, limiting the area available to less than 100 percent. It has been determined that with these delays and the less than fully open damper, the peak pressure in the enclosure following a steam line break may exceed the design pressure of the TDAFP enclosure. | | A detailed analysis is being performed to determine the extent of the potential overpressurization and the affect on the structural integrity of the TDAFP enclosure throughout the postulated HELB event. |
| A detailed analysis is being performed to determine the extent of the potential overpressurization and the affect on the structural integrity of the TDAFP enclosure throughout the postulated HELB event. NRC FORM 366A 14-95) | | NRC FORM 366A 14-95) |
| *
| | |
| * NRC FORM 366A (4-95) U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER) TEXT CONTINUATION FACILITY NAME 111 DOCKET LER NUMBER 161 Salem Generating Station -Unit 1 05000272 YEAR I SEQUENTIAL I REVISION NUMBER NUMBER 95 --012 --000 TEXT (If more space is required, use additional copies of NRC Form 366A) 117) SAFETY SIGNIFICANCE PAGE 131 3 OF 3 While the detailed analysis to determine the affect on the integrity of the TDAFP enclosure is not yet complete, an assumed failure of the TDAFP enclosure resulting from overpressurization due to a steam line break would allow steam to escape to surrounding areas. The escaping steam has the potential to render equipment in the area, including the motor driven auxiliary feed water pumps, inoperable.
| | NRC FORM 366A (4-95) |
| | * LICENSEE EVENT REPORT (LER) |
| | TEXT CONTINUATION |
| | * U.S. NUCLEAR REGULATORY COMMISSION FACILITY NAME 111 DOCKET LER NUMBER 161 PAGE 131 YEAR I SEQUENTIAL NUMBER I REVISION NUMBER Salem Generating Station - Unit 1 05000272 3 OF 3 95 -- 012 -- 000 TEXT (If more space is required, use additional copies of NRC Form 366A) 117) |
| | SAFETY SIGNIFICANCE While the detailed analysis to determine the affect on the integrity of the TDAFP enclosure is not yet complete, an assumed failure of the TDAFP enclosure resulting from overpressurization due to a steam line break would allow steam to escape to surrounding areas. The escaping steam has the potential to render equipment in the area, including the motor driven auxiliary feed water pumps, inoperable. |
| The supplement to this LER will provide a more detailed safety significance based on the results of the on-going structural analysis. | | The supplement to this LER will provide a more detailed safety significance based on the results of the on-going structural analysis. |
| APPARENT CAUSE OF OCCURRENCE The cause of this occurrence is attributed to inadequate verification of the unstated assumptions in the calculations performed to evaluate previously identified as-built design deficiencies. | | APPARENT CAUSE OF OCCURRENCE The cause of this occurrence is attributed to inadequate verification of the unstated assumptions in the calculations performed to evaluate previously identified as-built design deficiencies. During that evaluation, the primary focus was on the deficiencies being evaluated and the input assumptions_ were not questioned. A contributing factor is that one of the design drawings erroneously shows the ABS6 damper as normally open. |
| During that evaluation, the primary focus was on the deficiencies being evaluated and the input assumptions_ | | PRIOR SIMILAR OCCURRENCES Two prior similar occurrences involving unverified calculation assumptions have been identified within- the past five years. The first, as reported in LER 272/91-036-01, concerned incorrect input assumptions in the steam line break analyses. The second, as reported in LER 272/95-027-00, concerned inaccurate assumptions in dose calculations. Corrective actions taken in response to these events could not have prevented the event reported here. |
| were not questioned. | | CORRECTIVE ACTIONS Design change alternatives are under consideration to eliminate the potential for overpressurization of the TDAFP enclosure. The design change will be described in the LER supplement and will be implemented prior to restart of the units. - |
| A contributing factor is that one of the design drawings erroneously shows the ABS6 damper as normally open. PRIOR SIMILAR OCCURRENCES Two prior similar occurrences involving unverified calculation assumptions have been identified within-the past five years. The first, as reported in LER 272/91-036-01, concerned incorrect input assumptions in the steam line break analyses. | | Other HELB calculations involving similar plant configurations will be reviewed to verify the correctness of the input assumptions. A schedule for completion of this review will be provided in the supplement to this LER. |
| The second, as reported in LER 272/95-027-00, concerned inaccurate assumptions in dose calculations. | | Any additional corrective actions identified as a result of the continuing investigation into this occurrence will be provided in the supplement. |
| Corrective actions taken in response to these events could not have prevented the event reported here. CORRECTIVE ACTIONS Design change alternatives are under consideration to eliminate the potential for overpressurization of the TDAFP enclosure. | |
| The design change will be described in the LER supplement and will be implemented prior to restart of the units. -Other HELB calculations involving similar plant configurations will be reviewed to verify the correctness of the input assumptions. | |
| A schedule for completion of this review will be provided in the supplement to this LER. Any additional corrective actions identified as a result of the continuing investigation into this occurrence will be provided in the supplement. | |
| NRC FORM 366A (4-951}} | | NRC FORM 366A (4-951}} |
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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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PS~G Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038-0236 Nuclear Business Unit DEC 18 1995 LR-N95231 U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:
LICENSEE EVENT REPORT 272/95-012-000 SALEM GENERATING STATION - UNIT 1 FACILITY OPERATING LICENSE NO. DPR-70 DOCKET NO. 50-272 This Licensee Event Report entitled "Adequacy of Turbine Driven Auxiliary Feed Water Pump Enclosures" is being submitted pursuant to the requirements of the Code of Federal Regulations 10CFR50.73(a) (1).
Sincerely, .
-l-2 <? tl2ui..__
Clad. Warren General Manager -
Salem Operations SORC Mtg.95-149
.Attachment JHA/tcp C Distribution LER File Or*~.-. n , ...
9512270322 951218
- PDR ADOCK 05000272 S PDR 95*2168 REV. 6/94
Attachment A PSE&G Commitments for LER 272/95-012-000 The following items represent PSE&G commitments made to the Nuclear Regulatory Commission related to LER 272/95-012-000. The commitments are' as follows:
Design change alternatives are under consideration to eliminate the potential for overpressurization of the TDAFP enclosure. The design change will be described in the LER supplement and will be implemented prior to restart of the units.
Other HELB calculations involving similar plant configurations will be reviewed to verify the correctness of the input assumptions. A schedule for completion of this review will be provided in the supplement to this LER'.
Any additional corrective actions identified as a result of the dontinuing investigation into this occurence will be provided in the supplement.
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 14-95) EXPIRES 04/30/98 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITTi THIS MANDATORY INFORMATION COLLECTION REQUEST: 50.0 HRS.
REPORTED LESSONS LEARNED ARE INCORPORATED INTO THE LICENSEE EVENT REPORT (LER) LICENSING PROCESS AND FED BACK TO INDUSTflY. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE (See reverse for required number of INFORMATION AND RECORDS MANAGEMENT BRANCH IT-6 F33),
U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC digits/characters for each block) 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT FACILITY NAME 111 DOCKET NUMB9' 121 PAGE 131 Salem Generating Station - Unit 1 05000272 1 OF 3 TITLE 141 Adequacy of Turbine Driven Auxiliary Feed Water Pump Enclosures 171 I
EVENT DATE 151 LER NUMBER 161 REPORT DATE OTHER FACILITIES INVOLVED 181 MONTH DAY YEAR YEAR I SEQUENTIAL NUMBER REVISION NUMBER MONTH DAY YEAR FACILITY NAME Salem Generating Station - Unit 2 FACILITY NAME DOCKET NUMBER 05000311 12 11 76 95 -- 012 -- 000 12 18 95 DOCKET NUMBER I
I I*
OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: !Check one or more) 1111 MODE 191 20.2201 (b) 20.2203(a)(2Jlv) 50.731a)(2)(i) 50.73(aJl2)(viiiJ 20.2203(aJl1 I 20.2203(a)(3Jlil x 50.73(a)(2)(ii) 50.73(a)(2)(x)
I POWER LEVEL 1101 I0 I 20.2203(a)l2)(i) 20.2203(a)(3Jliil 50. 7 3 (a)(2)(iii) 73.71 20 .2203 (a)(2)(ii) 20.2203(a)(4) 50.73(a)(2lli"'.I OTHER 20.2203(a)(2)(iii) 50.36(c)(l I 50.73(a)(2)(v) Specify in Abstract below or in NRC Form 366A 20.2203 (a)(2)(iv) 50.36(c)(2) 50. 73 (a)(2)(vii)
LICENSEE CONTACT FOR THIS LER (12)
NAME TELEPHONE NUMBER (Include Are. Code)
Greg Cranston, Manager - Nuclear Engineering, 609 - 339 - 1955 Mee anical
- 1m-* l'!'l'F. nw1 J,T'NF. 'Ii' "'l'D F.lH~H -* *m - * 'N .. ~n *1* li'l TT.TTRF. r'l~R~'D Tl l'lfln T'tJ 'l'HTA 'DF.Pni>rri , , ":l \
REPORTABLE REPORTABLE CAUSE SYSTEM COMPONENT MANUFACTURER CAUSE SYSTEM COMPONENT MANUFACTURER TONPRDS TO NPRDS X IYES (If yes, SUPPLEMENTAL REPORT EXPECTED (141 complete EXPECTED SUBMISSION DATE).
I 'NO EXPECTED SUBMISSION DATE 1151 MONTH 02 DAY 15 YEAR 96 ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (161 In early November 1995 it was discovered that assumptions in the High Energy Line Break (HELB) analysis for the turbine driven auxiliary feed water pump (TDAFP) enclosure did not match as-built conditions and could allow pressure in the enclosure. to exceed the enclosure design pressure during a HELB. A detailed analysis is being performed to determine the extent of the potential overpressurization and the affect on the structural integrity of the TDAFP enclosure. The cause of this occurrence is attributed to inadequate verification of assumptions in the calculations performed to evaluate previously identified as-built design deficiencies and inaccurate design drawings. Corrective actions include a review of other HELB calculation assumptions and evaluation of design changes to eliminate the potential TDAFP enclosure overpressure.
This condition is being reported in accordance with 10CFR50.73(a) (2) (ii) .
A four hour 10CFR50.72 notification was made on November 16, 1995.
NRC FORM 366 14-95)
NRC FORM 366A (4-95)
- LICENSEE EVENT REPORT (LER)
- U.S. NUCLEAR REGULATORY COMMISSION TEXT CONTINUATION FACILITY NAME 111 DOCKET LER NUMBER 161 PAGE 131 YEAR I SEQUENTIAL NUMBER I REVISION NUMBER Salem Generating Station - Unit 1 05000272 2 OF 3 95 -- 012 -- 000 TEXT (If mor8 spac8 is rtJquirtJd, ustJ additional copies of NRC Form 366AJ (171 PLANT AND SYSTEM IDENTIFICATION Westinghouse - Pressurized Water Reactor Auxiliary Feedwater System (AF) {BA}*
Auxiliary Building Ventilation System (ABS) {VF}
- Energy Industry Identification system (EIIS) codes and component function identifier codes appear in the text as {SS/CCC}.
IDENTIFICATION OF OCCURRENCE Event Date: Unit 1: December 11, 1976 (Initial Plant Criticality)
Unit 2: August 2, 1980 (Initial Plant Criticality)
Date Determined to be Reportable: November 16, 1995 CONDITIONS PRIOR TO OCCURRENCE Unit 1: Defueled, o % Reactor Power Unit 2: Mode 5, o % Reactor Power DESCRIPTION OF OCCURRENCE -
In early November 1995 it was discovered that assumptions in the HE~B analysis for the turbine driven auxiliary feed water pump (TDAFP) {BA/P}
enclosure did not match as-built conditions and could potentially al1ow pressure in the enclosure to exceed the enclosure design pressure during a HELB.
The ABS 6 damper {VF/DMP} is located -in the wall separating the TDAFP enclosure and the adjacent pipe chase. The ABS 6 damper automatically opens to protect the TDAFP enclosure from overpressurization during a HELB in the enclosure. The HELB analysis performed for a steam line break in the TDAFP enclosure implicitly assumed that the ABS 6 damper opened instantaneously when the pressure in the enclosure reached the setpoint.
Significant delays actually exist between the pressure reaching the setpoint .and the initiation of the opening stroke of ABS 6 damper, and the damper reaching the full open position. The analysis also implicitly assumed that the damper full open position provided 100 percent of the damper opening for pressure relief. In the full open position the damper is actually open only 45 degrees, limiting the area available to less than 100 percent. It has been determined that with these delays and the less than fully open damper, the peak pressure in the enclosure following a steam line break may exceed the design pressure of the TDAFP enclosure.
A detailed analysis is being performed to determine the extent of the potential overpressurization and the affect on the structural integrity of the TDAFP enclosure throughout the postulated HELB event.
NRC FORM 366A 14-95)
NRC FORM 366A (4-95)
- LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION
- U.S. NUCLEAR REGULATORY COMMISSION FACILITY NAME 111 DOCKET LER NUMBER 161 PAGE 131 YEAR I SEQUENTIAL NUMBER I REVISION NUMBER Salem Generating Station - Unit 1 05000272 3 OF 3 95 -- 012 -- 000 TEXT (If more space is required, use additional copies of NRC Form 366A) 117)
SAFETY SIGNIFICANCE While the detailed analysis to determine the affect on the integrity of the TDAFP enclosure is not yet complete, an assumed failure of the TDAFP enclosure resulting from overpressurization due to a steam line break would allow steam to escape to surrounding areas. The escaping steam has the potential to render equipment in the area, including the motor driven auxiliary feed water pumps, inoperable.
The supplement to this LER will provide a more detailed safety significance based on the results of the on-going structural analysis.
APPARENT CAUSE OF OCCURRENCE The cause of this occurrence is attributed to inadequate verification of the unstated assumptions in the calculations performed to evaluate previously identified as-built design deficiencies. During that evaluation, the primary focus was on the deficiencies being evaluated and the input assumptions_ were not questioned. A contributing factor is that one of the design drawings erroneously shows the ABS6 damper as normally open.
PRIOR SIMILAR OCCURRENCES Two prior similar occurrences involving unverified calculation assumptions have been identified within- the past five years. The first, as reported in LER 272/91-036-01, concerned incorrect input assumptions in the steam line break analyses. The second, as reported in LER 272/95-027-00, concerned inaccurate assumptions in dose calculations. Corrective actions taken in response to these events could not have prevented the event reported here.
CORRECTIVE ACTIONS Design change alternatives are under consideration to eliminate the potential for overpressurization of the TDAFP enclosure. The design change will be described in the LER supplement and will be implemented prior to restart of the units. -
Other HELB calculations involving similar plant configurations will be reviewed to verify the correctness of the input assumptions. A schedule for completion of this review will be provided in the supplement to this LER.
Any additional corrective actions identified as a result of the continuing investigation into this occurrence will be provided in the supplement.
NRC FORM 366A (4-951
