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| | issue date = 10/18/1991 | | | issue date = 10/18/1991 |
| | title = LER 91-031-00:on 910923,main Steam Line Isolation Actuation Signal Occurred Upon Receipt of High Steam Line Flow Signal, Coincident W/Low Steam Line Pressure Signal.Caused by Design Inadequacy.Flow Measuring Sys modified.W/911018 Ltr | | | title = LER 91-031-00:on 910923,main Steam Line Isolation Actuation Signal Occurred Upon Receipt of High Steam Line Flow Signal, Coincident W/Low Steam Line Pressure Signal.Caused by Design Inadequacy.Flow Measuring Sys modified.W/911018 Ltr |
| | author name = POLLACK M J, VONDRA C A | | | author name = Pollack M, Vondra C |
| | author affiliation = PUBLIC SERVICE ELECTRIC & GAS CO. OF NEW JERSEY | | | author affiliation = PUBLIC SERVICE ELECTRIC & GAS CO. OF NEW JERSEY |
| | addressee name = | | | addressee name = |
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| {{#Wiki_filter:e Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 Salem Generating Station U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 | | {{#Wiki_filter:e PS~G Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 Salem Generating Station October 18, 1991 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 91-031-00 October 18, 1991 This Licensee Event Report is being submitted pursuant to the requirements of the Code of Federal Regulations lOCFR 50.73(a} (2) (iv). This report is required to be issued within thirty (30) days of event discovery. | | |
| MJP:pc Distribution T h c-::: ;:: .. .., , | | SALEM GENERATING STATION LICENSE NO. DPR-70 DOCKET NO. 50-272 UNIT NO. 1 LICENSEE EVENT REPORT 91-031-00 This Licensee Event Report is being submitted pursuant to the requirements of the Code of Federal Regulations 10CFR 50.73(a} (2) (iv). This report is required to be issued within thirty (30) days of event discovery. |
| * D c *'i -, t 9 i *1 | | Sincerely yours, C. A Vondra General Manager - |
| * PDR ADOCK 05000272 S PDR Sincerely yours, C. A Vondra General Manager -Salem Operations 95-2189 (10M) 12-E | | Salem Operations MJP:pc Distribution T h c- ::: ~- ;:: .. .., , |
| *; .. NRC FORM 366 16-89) U.S. NUCLEAR REGULATORY COMMISSION e APPROVED OMB NO. 3150-0104 LICENSEE EVENT REPORT (LERI EXPIRES: 4/30/92 ESTIMATED BURDEN PER RESPONSE TD COMPLY WTH THIS JNFORMATION COLLECTION REQUEST: 50.0 HAS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS AND F.IEPOATS MANAGEMENT BRANCH IP-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 121 I PAGE 3 Salem Generating Station -Unit 1 o I 5 I o I o I o I 2 I 7 I 2 1 OF O I 4 TITLE (4) ESF Actuation:
| | * D c *'i -, t ~ |
| Main Steam Line Isolation Signal Due To Design Concern EVENT DATE (51 LER NUMBER 161 REPORT DATE (7) OTHER FACILITIES INVOLVED (Bl MONTH DAY YEAR YEAR >t | | 9 i *1 c128~:~2t.'1 * .~11 :;~18 PDR ADOCK 05000272 95-2189 (10M) 12-E S PDR |
| :}( | | |
| MONTH DAY YEAR FACILITY NAMES DOCKET NUMBEAISI 019 2 I 3 9 1 911 -0 I 311 -oJ 0 mi 0 1 I 8 911 THIS REPORT IS SUBMITTED PURSUANT TO THE OF 10 CFR §: (Chock one or more of rhe following) | | NRC FORM 366 16-89) |
| (11) OPERATING. | | LICENSEE EVENT REPORT (LERI U.S. NUCLEAR REGULATORY COMMISSION e APPROVED OMB NO. 3150-0104 EXPIRES: 4/30/92 ESTIMATED BURDEN PER RESPONSE TD COMPLY WTH THIS JNFORMATION COLLECTION REQUEST: 50.0 HAS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS AND F.IEPOATS MANAGEMENT BRANCH IP-530), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO |
| 11* MODE (9) "" 20.402(b) 20.405lcl x 73.71(b) 50.73(1112)(iv)
| | *; THE PAPERWORK REDUCTION PROJECT 13150-0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503 . |
| POWER I 20.4-06ioi(1)(il | | FACILITY NAME (1) DOCKET NUMBER 121 I PAGE 3 Salem Generating Station - Unit 1 o I 5 I o I o I o I 2 I7 I 2 1 OF O I 4 TITLE (4) |
| -,_ 60.361cl i 11 50.73ioll2)(vi 73.71 (cl o 1 o 1 o-2a.40s1.1111w1
| | ESF Actuation: Main Steam Line Isolation Signal Due To Design Concern EVENT DATE (51 LER NUMBER 161 REPORT DATE (7) OTHER FACILITIES INVOLVED (Bl MONTH DAY YEAR YEAR >t SE~~~~~~AL :}( ~~';.~~~ MONTH DAY YEAR FACILITY NAMES DOCKET NUMBEAISI 019 2 I3 9 1 911 - 0 I 311 - oJ 0 mi 0 1 I8 911 THIS REPORT IS SUBMITTED PURSUANT TO THE R~QUIREMENTS OF 10 CFR §: (Chock one or more of rhe following) (11) |
| -50.361cll2l
| | OPERATING.11 |
| ,._..... 50.73le)12llvii) 50.7311112llviiillAI 50.7311112llviiillBI 50.731111211xl OTHER fSpdcify in Absrrllct bslow and in Tsxr. NRC Form 366A) --
| | * MODE (9) "" 20.402(b) 20.405lcl ,_x 50.73(1112)(iv) 73.71(b) |
| = 20.40511111 lliiil 20.40511111111*) | | POWER I |
| 20.40511111 | | ~*~~*-~-,--~~~~---! |
| )(Y) --50.73(111211il | | 20.4-06ioi(1)(il 60.361cl i 11 |
| ..._ 60.73(1112)(ii)
| | ,._..... 50.73ioll2)(vi 73.71 (cl L~;-;,rL o 1 o 1 o- 2a.40s1.1111w1 50.361cll2l 50.73le)12llvii) OTHER fSpdcify in Absrrllct |
| ,....._ 50.73(o)(2)(iii)
| | - - bslow and in Tsxr. NRC Form 1 |
| LICENSEE CONTACT FOR THIS LER 112) NAME TELEPHONE NUMBER AREA CODE M. J. Pollack -LER Coordinator COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT 1131 CAUSE SYSTEM COMPONENT I I I I I I I I TURER I I I I I I .*.* *.*.:*.* *.*. .-.--.-. .-.* .,. __ ._ . -.-. -.*. -:-: _.,. *.-. .-.--.-. .-.--.*.*.-. _.,. ,.,._.,.,.,. | | 1tlll~11lllllll1llll= |
| .-.-.-.-.
| | 20.40511111 lliiil 50.73(111211il 50.7311112llviiillAI 366A) 20.40511111111*) 60.73(1112)(ii) 50.7311112llviiillBI 20.40511111 )(Y) |
| _._.,_._._, _._._._,_._.
| | - 50.73(o)(2)(iii) 50.731111211xl LICENSEE CONTACT FOR THIS LER 112) |
| **:-*:-: ;.:.;.: *:-:-:*:-:*:
| | NAME TELEPHONE NUMBER AREA CODE M. J. Pollack - LER Coordinator COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT 1131 CAUSE SYSTEM COMPONENT MANUFAC-TURER COMPONENT MANUFAC-TURER A ~6o~;~giE *::.*.:1::::1:**:*1,:::.1....1.:::1*:*,:,:::::*:::*1: |
| _.,._._. -.-..-.:.-.*
| | 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 (141 MONTH DAY YEAR EXPECTED n YES (If yes. comp/ere EXPECTED SUBMISSION DATE) M ABSTRACT (Limit to 1400 spaces. I.e .. spproximatoly f;frBBn singfe.space typewrirren lines) 116) |
| SUPPLEMENTAL REPORT EXPECTED (141 n YES (If yes. comp/ere EXPECTED SUBMISSION DATE) M NO ABSTRACT (Limit to 1400 spaces. I.e .. spproximatoly f;frBBn singfe.space typewrirren lines) 116) COMPONENT
| | NO SUBMISSION DATE 1151 I I I On 9/23/91 at 1414 hours, a Main Steam Isolation (MSI) actuation occurred. At the time, the Unit was in Mode 4 and heating up in preparation for startup. The MSI signal occurred upon receipt of a High Steamline Flow signal coincident with a Low Steamline Pressure signal. In Mode 4, the bistables for low steamline pressure are tripped providing half the logic signal required for MSI. At the time of the event, the No. 14 Stea~ Generator (S/G) Steamline flow channel No. I bistable was in the tripped position to support channel functional testing. When No. 11 S/G steamline flow channels I and II bistables tripped the MSI signal logic was satisfied. MSI is an Engineered Safety Feature. Similar MSI events have occurred (reference LERs 272/90-019-00 and 272/90-027-00). The root cause of this event is attributed to design inadequacy associated with the main steamline flow transmitter sensing lines during plant heatup. Assessment of this event, by Maintenance personnel, was that the false high steam flow signals were not caused by failed components. The false signals cleared after a few hours. At the time of the event, Maintenance was performing channel monthly functional testing as per Tech. Spec. |
| :-: I I I I I I I I TURER I I I I I I EXPECTED SUBMISSION DATE 1151 A
| | Surveillances 4.3.1.1.1 and 4.3.2.1.1. The channel I and channel II flow transmitters functional testing, for all 4 steamlines, was successfully completed on 9/23/91 and 9/28/91, respectively. |
| *::.*.:1::::1:**:*1,:::.1
| | Engineering is continuing to evaluate the modifications made to the main steam flow measuring system completed during the Unit 1 ninth refueling outage. |
| .... 1.:::1*:*,:,:::::*:::*1: | | NRC Form 366 (6-89) |
| MONTH DAY YEAR I I I On 9/23/91 at 1414 hours, a Main Steam Isolation (MSI) actuation occurred.
| | |
| At the time, the Unit was in Mode 4 and heating up in preparation for startup. The MSI signal occurred upon receipt of a High Steamline Flow signal coincident with a Low Steamline Pressure signal. In Mode 4, the bistables for low steamline pressure are tripped providing half the logic signal required for MSI. At the time of the event, the No. 14 Generator (S/G) Steamline flow channel No. I bistable was in the tripped position to support channel functional testing. When No. 11 S/G steamline flow channels I and II bistables tripped the MSI signal logic was satisfied.
| | LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station DOCKET NUMBER LER NUMBER PAGE Unit 1 5000272 |
| MSI is an Engineered Safety Feature. Similar MSI events have occurred (reference LERs 272/90-019-00 and 272/90-027-00).
| | ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ |
| The root cause of this event is attributed to design inadequacy associated with the main steamline flow transmitter sensing lines during plant heatup. Assessment of this event, by Maintenance personnel, was that the false high steam flow signals were not caused by failed components. | | 91-031-00 2 of 4 PLANT AND SYSTEM IDENTIFICATION: |
| The false signals cleared after a few hours. At the time of the event, Maintenance was performing channel monthly functional testing as per Tech. Spec. Surveillances 4.3.1.1.1 and 4.3.2.1.1.
| | Westinghouse - Pressurized Water Reactor Energy Industry Identification System (EIIS) codes are identified in the text as rxxl IDENTIFICATION OF OCCURRENCE: |
| The channel I and channel II flow transmitters functional testing, for all 4 steamlines, was successfully completed on 9/23/91 and 9/28/91, respectively.
| | Engineered Safety Feature Actuation; Main Steamline Isolation Due to Equipment/Design Concerns Event Date: 9/23/91 Report Date: 10/18/91 This report was initiated by Incident Report No. 91-664. |
| Engineering is continuing to evaluate the modifications made to the main steam flow measuring system completed during the Unit 1 ninth refueling outage. NRC Form 366 (6-89)
| | CONDITIONS PRIOR TO OCCURRENCE: |
| LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station Unit 1 DOCKET NUMBER LER NUMBER PAGE 5000272 91-031-00 2 of 4 PLANT AND SYSTEM IDENTIFICATION:
| | Mode 4 (Hot Shutdown); Reactor in preparation for startup Tavg at approximately 250°F and increasing DESCRIPTION OF OCCURRENCE: |
| Westinghouse
| | On September 23, 1991, at 1414 hours, with the Unit in Mode 4 {Hot Shutdown) and heating up in preparation for startup, a Main Steamline Isolation (MSI) {JEI actuation occurred. The MSI signal occurred upon receipt of a "High Steamline Flow" signal coincident with an existing "Low Steamline Pressure" signal. |
| -Pressurized Water Reactor Energy Industry Identification System (EIIS) codes are identified in the text as rxxl IDENTIFICATION OF OCCURRENCE:
| | In Mode 4, the low steamline pressure bistables are tripped due to plant conditions, providing half of the logic signal required for MSI. The high steamline flow logic requires indication of high flow in one (1) out of two (2) channels per Steam Generator (S/G) in two (2) of the fciur (4) S/Gs. At the time of the event, the No. 14 Steam Generator (S/G) Steamline flow channel No. I bistable was in the tripped position to support channel functional testing. When No. 11 SIG steamline flow channels I and II bistables tripped the Main Stearnline Isolation signal logic was satisfied. |
| Engineered Safety Feature Actuation; Main Steamline Isolation Due to Equipment/Design Concerns Event Date: 9/23/91 Report Date: 10/18/91 This report was initiated by Incident Report No. 91-664. CONDITIONS PRIOR TO OCCURRENCE:
| | Several minutes after the MSI signal, No. 12 S/G Channel II high steamline flow bistables tripped. |
| Mode 4 (Hot Shutdown);
| | MSI is an Engineered Safety Feature (ESF). Therefore, on September 23, 1991 at 1520 hours, this event was reported to the Nuclear Regulatory Commission in accordance with Code of Federal Regulations 10CFR 50. 72 (b) (2) (ii). |
| Reactor in preparation for startup Tavg at approximately 250°F and increasing DESCRIPTION OF OCCURRENCE:
| | APPARENT CAUSE OF OCCURRENCE: |
| On September 23, 1991, at 1414 hours, with the Unit in Mode 4 {Hot Shutdown) and heating up in preparation for startup, a Main Steamline Isolation (MSI) {JEI actuation occurred.
| | Similar MSI actuations occurred on June 3, 1990 and August 12, 1990 (reference LERs 272/90-019-00 and 272/90-027-00, respectively). The root cause of this event (and the prior events) is attributed to |
| The MSI signal occurred upon receipt of a "High Steamline Flow" signal coincident with an existing "Low Steamline Pressure" signal. In Mode 4, the low steamline pressure bistables are tripped due to plant conditions, providing half of the logic signal required for MSI. The high steamline flow logic requires indication of high flow in one (1) out of two (2) channels per Steam Generator (S/G) in two (2) of the fciur (4) S/Gs. At the time of the event, the No. 14 Steam Generator (S/G) Steamline flow channel No. I bistable was in the tripped position to support channel functional testing. When No. 11 SIG steamline flow channels I and II bistables tripped the Main Stearnline Isolation signal logic was satisfied.
| | |
| Several minutes after the MSI signal, No. 12 S/G Channel II high steamline flow bistables tripped. MSI is an Engineered Safety Feature (ESF). Therefore, on September 23, 1991 at 1520 hours, this event was reported to the Nuclear Regulatory Commission in accordance with Code of Federal Regulations lOCFR 50. 72 (b) (2) (ii). APPARENT CAUSE OF OCCURRENCE:
| | LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station DOCKET NUMBER LER NUMBER PAGE |
| Similar MSI actuations occurred on June 3, 1990 and August 12, 1990 (reference LERs 272/90-019-00 and 272/90-027-00, respectively).
| | =u=n=i~t=-=1=--~~~~~~~~~~~~~~5~0~0~0~2~7~2'==----~~~~~9~1~--=-0~3~1--_=o~o'---~~~3=--=of 4 APPARENT CAUSE OF OCCURRENCE: (cont'd) design inadequacy associated with the main steamline flow transmitter sensing lines during plant heatup. Assessment of this event, by Maintenance personnel, was that the false high steam flow signals were not caused by failed components. The false signals cleared, on their own, after a few hours. |
| The root cause of this event (and the prior events) is attributed to LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station DOCKET NUMBER LER NUMBER PAGE 4 APPARENT CAUSE OF OCCURRENCE: (cont'd) design inadequacy associated with the main steamline flow transmitter sensing lines during plant heatup. Assessment of this event, by Maintenance personnel, was that the false high steam flow signals were not caused by failed components.
| | At the time of the event, Maintenance was performing channel monthly functional testing as per Technical Specification Surveillances 4.3.1.1.1 and 4.3.2.1.1. The channel I and channel II flow transmitters functional testing, for all four (4) steamlines, was successfully completed on September 23, 1991 and September 28, 1991, respectively. |
| The false signals cleared, on their own, after a few hours. At the time of the event, Maintenance was performing channel monthly functional testing as per Technical Specification Surveillances 4.3.1.1.1 and 4.3.2.1.1.
| | The Salem design arrangement for main steamline flow differential pressure measurement includes two (2) taps (to provide redundancy) on the high-and low pressure side of the main steamline venturi. |
| The channel I and channel II flow transmitters functional testing, for all four (4) steamlines, was successfully completed on September 23, 1991 and September 28, 1991, respectively.
| | Attached to the taps are 1" manual globe valves. Steam is directed through 1" pipe to condensate pots located near the high pressure tap. The condensate is then directed to a Rosemount model 1153HD5 differential pressure transmitter via a 3/8" line. |
| The Salem design arrangement for main steamline flow differential pressure measurement includes two (2) taps (to provide redundancy) on the high-and low pressure side of the main steamline venturi. Attached to the taps are 1" manual globe valves. Steam is directed through 1" pipe to condensate pots located near the high pressure tap. The condensate is then directed to a Rosemount model 1153HD5 differential pressure transmitter via a 3/8" line. Steamline flow measuring design concerns were previously identified via LER 272/88-017-01.
| | Steamline flow measuring design concerns were previously identified via LER 272/88-017-01. That LER addresses steamline flow measurement drift with the Unit at power. Engineering believes that the drift concern and this recent event appear to be related to design concerns associated with the transmitter sensing lines. Possible causes of these events are discussed in a detailed study completed in support of LER 272/88-017-01. |
| That LER addresses steamline flow measurement drift with the Unit at power. Engineering believes that the drift concern and this recent event appear to be related to design concerns associated with the transmitter sensing lines. Possible causes of these events are discussed in a detailed study completed in support of LER 272/88-017-01. | |
| ANALYSIS OF OCCURRENCE: | | ANALYSIS OF OCCURRENCE: |
| MS! protection is applicable in Mode 1 (Power Operation) , Mode 2 (Startup), and Mode 3 (Hot Standby). | | MS! protection is applicable in Mode 1 (Power Operation) , Mode 2 (Startup), and Mode 3 (Hot Standby). It is provided to mitigate the consequences of various design base accidents including main steamline rupture and steam generator primary to secondary tube rupture. |
| It is provided to mitigate the consequences of various design base accidents including main steamline rupture and steam generator primary to secondary tube rupture. In Mode 4, the reactor is subcritical with Tavg between 200°F and 350°F. Decay heat is removed either by the Residual Heat Removal system or via steaming from the steam generators. | | In Mode 4, the reactor is subcritical with Tavg between 200°F and 350°F. Decay heat is removed either by the Residual Heat Removal system or via steaming from the steam generators. Makeup water to the S/Gs can be supplied by either a Condensate Pump or by an Auxiliary Feedwater Pump. In Mode 4, the Auxiliary Feedwater System |
| Makeup water to the S/Gs can be supplied by either a Condensate Pump or by an Auxiliary Feedwater Pump. In Mode 4, the Auxiliary Feedwater System {BAJ is not required to be operable. | | {BAJ is not required to be operable. |
| At the time of the actuation, decay heat removal was being accomplished using the Residual Heat Removal (RHR) System {BP} via No. 11 RHR Pump. When the MS! signal was actuated, the four MS7 Main Steamline Drain Valves closed as designed. | | At the time of the actuation, decay heat removal was being accomplished using the Residual Heat Removal (RHR) System {BP} via No. 11 RHR Pump. When the MS! signal was actuated, the four MS7 Main Steamline Drain Valves closed as designed. The other valves which close on a MS! signal were already closed; i.e., MS167s (main steamline isolation valves) and MS18s (main steamline warm-up valves). Since the actuation was not the result of an actual plant need for Main Steam Isolation, this event did not affect the health |
| The other valves which close on a MS! signal were already closed; i.e., MS167s (main steamline isolation valves) and MS18s (main steamline warm-up valves). Since the actuation was not the result of an actual plant need for Main Steam Isolation, this event did not affect the health LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station Unit 1 ANALYSIS OF OCCURRENCE: | | |
| DOCKET NUMBER 5000272 (cont'd) LER NUMBER 91-031-00 PAGE 4 of 4 or safety of the public. However, since Main Stearn Isolation is an ESF system, this event is reportable to the Nuclear Regulatory Commission in accordance with Code of Federal Regulations 10CFR50.73(a) | | LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station DOCKET NUMBER LER NUMBER PAGE Unit 1 5000272 91-031-00 4 of 4 ANALYSIS OF OCCURRENCE: (cont'd) or safety of the public. However, since Main Stearn Isolation is an ESF system, this event is reportable to the Nuclear Regulatory Commission in accordance with Code of Federal Regulations 10CFR50.73(a) (2) (iv). |
| (2) (iv). CORRECTIVE ACTION: As stated in the Apparent Cause of Occurrence section, the channel I and channel II flow transmitters functional testing, for all four (4) stearnlines, was successfully completed on September 23, 1991 and September 28, 1991, respectively. | | CORRECTIVE ACTION: |
| System Engineering is continuing to evaluate {during the current fuel cycle) modifications made to the main steam flow measuring system completed during the Unit 1 ninth refueling outage (April 1991). Upon completion of this evaluation, recommendations for additional changes to the system will be considered for implementation. | | As stated in the Apparent Cause of Occurrence section, the channel I and channel II flow transmitters functional testing, for all four (4) stearnlines, was successfully completed on September 23, 1991 and September 28, 1991, respectively. |
| The modifications which had been completed include: 1. the addition of a summator in the steam flow protection loop for all channels and 2. drilling holes at the low point of the instrument line to process line nozzles for all eight lower flow rate measuring instrument line taps. MJP :.pc SORC Mtg. 91-107 General Manager -Salem Operations}} | | System Engineering is continuing to evaluate {during the current fuel cycle) modifications made to the main steam flow measuring system completed during the Unit 1 ninth refueling outage (April 1991). |
| | Upon completion of this evaluation, recommendations for additional changes to the system will be considered for implementation. The modifications which had been completed include: 1. the addition of a summator in the steam flow protection loop for all channels and 2. |
| | drilling holes at the low point of the instrument line to process line nozzles for all eight lower flow rate measuring instrument line taps. |
| | General Manager - |
| | Salem Operations MJP :.pc SORC Mtg. 91-107}} |
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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
[Table view] |
Text
e PS~G Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 Salem Generating Station October 18, 1991 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 91-031-00 This Licensee Event Report is being submitted pursuant to the requirements of the Code of Federal Regulations 10CFR 50.73(a} (2) (iv). 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 T h c- ::: ~- ;:: .. .., ,
9 i *1 c128~:~2t.'1 * .~11 :;~18 PDR ADOCK 05000272 95-2189 (10M) 12-E S PDR
NRC FORM 366 16-89)
LICENSEE EVENT REPORT (LERI U.S. NUCLEAR REGULATORY COMMISSION e APPROVED OMB NO. 3150-0104 EXPIRES: 4/30/92 ESTIMATED BURDEN PER RESPONSE TD COMPLY WTH THIS JNFORMATION COLLECTION REQUEST: 50.0 HAS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS AND F.IEPOATS MANAGEMENT BRANCH IP-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 121 I PAGE 3 Salem Generating Station - Unit 1 o I 5 I o I o I o I 2 I7 I 2 1 OF O I 4 TITLE (4)
ESF Actuation: Main Steam Line Isolation Signal Due To Design Concern EVENT DATE (51 LER NUMBER 161 REPORT DATE (7) OTHER FACILITIES INVOLVED (Bl MONTH DAY YEAR YEAR >t SE~~~~~~AL :}( ~~';.~~~ MONTH DAY YEAR FACILITY NAMES DOCKET NUMBEAISI 019 2 I3 9 1 911 - 0 I 311 - oJ 0 mi 0 1 I8 911 THIS REPORT IS SUBMITTED PURSUANT TO THE R~QUIREMENTS OF 10 CFR §: (Chock one or more of rhe following) (11)
OPERATING.11
- MODE (9) "" 20.402(b) 20.405lcl ,_x 50.73(1112)(iv) 73.71(b)
POWER I
~*~~*-~-,--~~~~---!
20.4-06ioi(1)(il 60.361cl i 11
,._..... 50.73ioll2)(vi 73.71 (cl L~;-;,rL o 1 o 1 o- 2a.40s1.1111w1 50.361cll2l 50.73le)12llvii) OTHER fSpdcify in Absrrllct
- - bslow and in Tsxr. NRC Form 1
1tlll~11lllllll1llll=
20.40511111 lliiil 50.73(111211il 50.7311112llviiillAI 366A) 20.40511111111*) 60.73(1112)(ii) 50.7311112llviiillBI 20.40511111 )(Y)
- 50.73(o)(2)(iii) 50.731111211xl LICENSEE CONTACT FOR THIS LER 112)
NAME TELEPHONE NUMBER AREA CODE M. J. Pollack - LER Coordinator COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT 1131 CAUSE SYSTEM COMPONENT MANUFAC-TURER COMPONENT MANUFAC-TURER A ~6o~;~giE *::.*.:1::::1:**:*1,:::.1....1.:::1*:*,:,:::::*:::*1:
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 (141 MONTH DAY YEAR EXPECTED n YES (If yes. comp/ere EXPECTED SUBMISSION DATE) M ABSTRACT (Limit to 1400 spaces. I.e .. spproximatoly f;frBBn singfe.space typewrirren lines) 116)
NO SUBMISSION DATE 1151 I I I On 9/23/91 at 1414 hours0.0164 days <br />0.393 hours <br />0.00234 weeks <br />5.38027e-4 months <br />, a Main Steam Isolation (MSI) actuation occurred. At the time, the Unit was in Mode 4 and heating up in preparation for startup. The MSI signal occurred upon receipt of a High Steamline Flow signal coincident with a Low Steamline Pressure signal. In Mode 4, the bistables for low steamline pressure are tripped providing half the logic signal required for MSI. At the time of the event, the No. 14 Stea~ Generator (S/G) Steamline flow channel No. I bistable was in the tripped position to support channel functional testing. When No. 11 S/G steamline flow channels I and II bistables tripped the MSI signal logic was satisfied. MSI is an Engineered Safety Feature. Similar MSI events have occurred (reference LERs 272/90-019-00 and 272/90-027-00). The root cause of this event is attributed to design inadequacy associated with the main steamline flow transmitter sensing lines during plant heatup. Assessment of this event, by Maintenance personnel, was that the false high steam flow signals were not caused by failed components. The false signals cleared after a few hours. At the time of the event, Maintenance was performing channel monthly functional testing as per Tech. Spec.
Surveillances 4.3.1.1.1 and 4.3.2.1.1. The channel I and channel II flow transmitters functional testing, for all 4 steamlines, was successfully completed on 9/23/91 and 9/28/91, respectively.
Engineering is continuing to evaluate the modifications made to the main steam flow measuring system completed during the Unit 1 ninth refueling outage.
NRC Form 366 (6-89)
LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station DOCKET NUMBER LER NUMBER PAGE Unit 1 5000272
~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
91-031-00 2 of 4 PLANT AND SYSTEM IDENTIFICATION:
Westinghouse - Pressurized Water Reactor Energy Industry Identification System (EIIS) codes are identified in the text as rxxl IDENTIFICATION OF OCCURRENCE:
Engineered Safety Feature Actuation; Main Steamline Isolation Due to Equipment/Design Concerns Event Date: 9/23/91 Report Date: 10/18/91 This report was initiated by Incident Report No.91-664.
CONDITIONS PRIOR TO OCCURRENCE:
Mode 4 (Hot Shutdown); Reactor in preparation for startup Tavg at approximately 250°F and increasing DESCRIPTION OF OCCURRENCE:
On September 23, 1991, at 1414 hours0.0164 days <br />0.393 hours <br />0.00234 weeks <br />5.38027e-4 months <br />, with the Unit in Mode 4 {Hot Shutdown) and heating up in preparation for startup, a Main Steamline Isolation (MSI) {JEI actuation occurred. The MSI signal occurred upon receipt of a "High Steamline Flow" signal coincident with an existing "Low Steamline Pressure" signal.
In Mode 4, the low steamline pressure bistables are tripped due to plant conditions, providing half of the logic signal required for MSI. The high steamline flow logic requires indication of high flow in one (1) out of two (2) channels per Steam Generator (S/G) in two (2) of the fciur (4) S/Gs. At the time of the event, the No. 14 Steam Generator (S/G) Steamline flow channel No. I bistable was in the tripped position to support channel functional testing. When No. 11 SIG steamline flow channels I and II bistables tripped the Main Stearnline Isolation signal logic was satisfied.
Several minutes after the MSI signal, No. 12 S/G Channel II high steamline flow bistables tripped.
MSI is an Engineered Safety Feature (ESF). Therefore, on September 23, 1991 at 1520 hours0.0176 days <br />0.422 hours <br />0.00251 weeks <br />5.7836e-4 months <br />, this event was reported to the Nuclear Regulatory Commission in accordance with Code of Federal Regulations 10CFR 50. 72 (b) (2) (ii).
APPARENT CAUSE OF OCCURRENCE:
Similar MSI actuations occurred on June 3, 1990 and August 12, 1990 (reference LERs 272/90-019-00 and 272/90-027-00, respectively). The root cause of this event (and the prior events) is attributed to
LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station DOCKET NUMBER LER NUMBER PAGE
=u=n=i~t=-=1=--~~~~~~~~~~~~~~5~0~0~0~2~7~2'==----~~~~~9~1~--=-0~3~1--_=o~o'---~~~3=--=of 4 APPARENT CAUSE OF OCCURRENCE: (cont'd) design inadequacy associated with the main steamline flow transmitter sensing lines during plant heatup. Assessment of this event, by Maintenance personnel, was that the false high steam flow signals were not caused by failed components. The false signals cleared, on their own, after a few hours.
At the time of the event, Maintenance was performing channel monthly functional testing as per Technical Specification Surveillances 4.3.1.1.1 and 4.3.2.1.1. The channel I and channel II flow transmitters functional testing, for all four (4) steamlines, was successfully completed on September 23, 1991 and September 28, 1991, respectively.
The Salem design arrangement for main steamline flow differential pressure measurement includes two (2) taps (to provide redundancy) on the high-and low pressure side of the main steamline venturi.
Attached to the taps are 1" manual globe valves. Steam is directed through 1" pipe to condensate pots located near the high pressure tap. The condensate is then directed to a Rosemount model 1153HD5 differential pressure transmitter via a 3/8" line.
Steamline flow measuring design concerns were previously identified via LER 272/88-017-01. That LER addresses steamline flow measurement drift with the Unit at power. Engineering believes that the drift concern and this recent event appear to be related to design concerns associated with the transmitter sensing lines. Possible causes of these events are discussed in a detailed study completed in support of LER 272/88-017-01.
ANALYSIS OF OCCURRENCE:
MS! protection is applicable in Mode 1 (Power Operation) , Mode 2 (Startup), and Mode 3 (Hot Standby). It is provided to mitigate the consequences of various design base accidents including main steamline rupture and steam generator primary to secondary tube rupture.
In Mode 4, the reactor is subcritical with Tavg between 200°F and 350°F. Decay heat is removed either by the Residual Heat Removal system or via steaming from the steam generators. Makeup water to the S/Gs can be supplied by either a Condensate Pump or by an Auxiliary Feedwater Pump. In Mode 4, the Auxiliary Feedwater System
{BAJ is not required to be operable.
At the time of the actuation, decay heat removal was being accomplished using the Residual Heat Removal (RHR) System {BP} via No. 11 RHR Pump. When the MS! signal was actuated, the four MS7 Main Steamline Drain Valves closed as designed. The other valves which close on a MS! signal were already closed; i.e., MS167s (main steamline isolation valves) and MS18s (main steamline warm-up valves). Since the actuation was not the result of an actual plant need for Main Steam Isolation, this event did not affect the health
LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station DOCKET NUMBER LER NUMBER PAGE Unit 1 5000272 91-031-00 4 of 4 ANALYSIS OF OCCURRENCE: (cont'd) or safety of the public. However, since Main Stearn Isolation is an ESF system, this event is reportable to the Nuclear Regulatory Commission in accordance with Code of Federal Regulations 10CFR50.73(a) (2) (iv).
CORRECTIVE ACTION:
As stated in the Apparent Cause of Occurrence section, the channel I and channel II flow transmitters functional testing, for all four (4) stearnlines, was successfully completed on September 23, 1991 and September 28, 1991, respectively.
System Engineering is continuing to evaluate {during the current fuel cycle) modifications made to the main steam flow measuring system completed during the Unit 1 ninth refueling outage (April 1991).
Upon completion of this evaluation, recommendations for additional changes to the system will be considered for implementation. The modifications which had been completed include: 1. the addition of a summator in the steam flow protection loop for all channels and 2.
drilling holes at the low point of the instrument line to process line nozzles for all eight lower flow rate measuring instrument line taps.
General Manager -
Salem Operations MJP :.pc SORC Mtg.91-107
