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| | issue date = 08/13/1998 | | | issue date = 08/13/1998 |
| | title = 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 | | | title = 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 |
| | author name = BAKKEN A C, DUCA P J | | | author name = Bakken A, Duca P |
| | 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:.. * | | {{#Wiki_filter:.. |
| * Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038-0236 Nucle'ar Business Unit U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen: | | PS~G * |
| LER 311/98-010-00 AUG 13 1998 LR-N980397 SALEM GENERATING STATION -UNIT 2 FACILITY OPERATING LICENSE NO. DPR-75 DOCKET NO. 50-311 This Licensee Event Report entitled "ECCS Leakage Outside Design Basis Value" is being submitted pursuant to the requirements of the Code of Federal Regulations | | * Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038-0236 Nucle'ar Business Unit AUG 13 1998 LR-N980397 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen: |
| ****1 OCFR50.73 (a)(2)(ii)****. | | LER 311/98-010-00 SALEM GENERATING STATION - UNIT 2 FACILITY OPERATING LICENSE NO. DPR-75 DOCKET NO. 50-311 This Licensee Event Report entitled "ECCS Leakage Outside Design Basis Value" is being submitted pursuant to the requirements of the Code of Federal Regulations |
| Attachment PJD/ c Distribution LER File 3.7 9808210157 980813 PDR ADOCK 05000311 | | ****1 OCFR50.73 (a)(2)(ii)****. |
| * S PDR The power is in your hands. Sincerely, A. C. Bakken Ill General Manager Salem Operations 95-2168 REV. 6/94 NRC FORM 366 U.S. NUCLEAR REGULATOR (6-1998) LICENSEE EVENT REPORT (LER) (See reverse for required number of digits/characters for each block) FACILITY NAME (1) SALEM UNIT 2 TITLE(4) MISSION ECCS LEAKAGE OUTSIDE DESIGN BASIS VALUE. APPROVED BY 0. 3150-0104 EXPIRES 06/30/2001 Estimated burden p ponse to comply with this mandatory information collection request: 50 hrs. Reported lessons learned are incorporated into the licensing process and fed back to industry. | | Sincerely, A. C. Bakken Ill General Manager Salem Operations Attachment PJD/ |
| Forward comments regarding burden estimate to the Records Management Branch (T-6 F33), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and to the Paperwork Reduction Project (3150-0104), Office of Management and Budget, Washington, DC 20503. If an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
| | c Distribution LER File 3.7 9808210157 980813 PDR ADOCK 05000311 |
| DOCKET NUMBER (2) 05000311 !PAGET OF 4 EVENT DATE (5) LER NUMBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED (8) MONTH DAY YEAR YEAR SEQUENTIAL REVISION MONTH DAY YEAR FACILITY NAME DOCKET NUMBER NUMBER NUMBER 07 14 98 98 -010 00 08 13 98 FACILITY NAME DOCKET NUMBER Salem Unit 1 05000272 OPERATING 1 100% x NAME TELEPHONE NUMBER (Include Area Code) Philip J. Duca Jr. , Salem Licensing Engineer (609) 339-2381 COMPLETE AILURE DESC BED IN THIS REPORT (13) CAUSE SYSTEM COMPONE MANUFACTURER REPORTABLE CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE NT TO EPIX TO EPIX EXPECTED MONTH DAY YEAR YES (If yes, complete EXPECTED SUBMISSION DATE). ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16) On 7 /14/98 Operations personnel determined that leakage f.rom the Boron Injection Tank exceeded the maximum allowable ECCS leakage from sources outside Containment.
| | *S PDR The power is in your hands. |
| The leakage path was determined to be past the valve seat of a manual sample valve (2SJ404) from the Boron Injection Tank. The valve was subsequently isolated.
| | 95-2168 REV. 6/94 |
| With the valve isolated, ECCS leakage returned to within UFSAR limits. The cause of the event was the leaking 2SJ404 manual sample valve. During the period of leakage, there were no actual safety consequences to the health and safety of the public or to the plant staff. Estimated potential post accident offsite doses remain well within 10CFR100 limits. Estimated post accident onsite beta skin dose is within the General Design Criterion 19 (GDC 19) limit, while estimated whole body and thyroid doses are slightly above the GDC 19 limits based on conservative estimates.
| |
| Estimated core damage frequency during the period of the leakage, at approximately 3.7xE-6, decreases potential significance.
| |
| This event is reportable pursuant to 10CFR50.73(a)2(ii) "any event or. condi tion ...............
| |
| that resulted in the nuclear power plant being:...... (B) in a condition that was outside the design basis of the plant".
| |
| 1---*---. NRC FORM 366A (6-1998) U.S. NUCLEAR REG ATORY COMMISSION FACILITY NAME (1) SALEM UNIT 2 LICENSEE EVENT REPORT (LER) TEXT CONTINUATION DOCKET (2) NUMBER(2) 05000311 TEXT (If more space is required, use additional copies ofNRC Form 366A) (17) PLANT AND SYSTEM* IDENTIFICATION Westinghouse
| |
| -Pressurized Water Reactor LER NUMBER (6) PAGE (3) NUMBER NUMBER YEAR I SEQUENTIAL I REVISION 2 0 F 98 0 1 0 00 High-Pressure Safety Injection System/Sample Valve {BQ/SMV}*
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| 4
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| * Energy Industry System {EIIS} codes and component function identifier codes appear as (SS/CCC) CONDITIONS PRIOR TO OCCURRENCE At the time of identification, Salem Unit 2 was operating at 100% Power. DESCRIPTION OF OCCURRENCE On 7/14/98 Operations personnel determined that leakage from the Boron Injection Tank exceeded the maximum allowable ECCS Leakage as defined by UFSAR Section 6.3.2.11.
| |
| The leakage was determined to be approximately 0.25 gpm (60,000cc/hour) past the valve seat of a manual sample valve (2SJ404) from the Boron Injection Tank. The leakage was flowing into the RHR sump in the Auxiliary Building, outside of Containment.
| |
| UFSAR Section 6.3.2.11 addresses leakage during the recirculation phase of an accident, and states that "The total leakage resulting from all sources is about 3800 cc/hour as described in UFSAR Section 15.4.1. Recirculation loop leakage sources are summarized in Table 6.3-12. Leakage is monitored by procedure to ensure this leak rate is not exceeded." Leakage in excess of 3800 cc/hour placed the plant in a condition outside that assumed for design basis accident conditions.
| |
| For about two weeks prior to the event, efforts were being focused on reducing an elevated reactor coolant system (RCS) unidentified leakage. Initially this leakage was thought to be within containment.
| |
| The continuing investigations included determining the source of increasing in-leakage to the RHR sump located in the Auxiliary Building.
| |
| These investigations used station procedures.
| |
| One of these procedures provides a program for monitoring leakage from systems outside Containment that could contain highly radioactive fluids following an accident.
| |
| The objective of the program is to detect and correct any degradation of the pressure boundaries for these systems and thereby reduce post-accident dose rates and airborne activity.in the Auxiliary Building in accordance with Technical Specification 6.8.4.a. NRC FORM 366A (6-1998)
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| NRC FORM 366A (6-1998) U.S. NUCLEAR REGULATORY COMMISSION FACILITY NAME (1) SALEM UNIT 2 LICENSEE EVENT REPORT (LER) TEXT CONTINUATION DOCKET (2) NUMBER(2) 05000311 TEXT (If more space is required, use additional copies ofNRCForm 366A) (17) LER NUMBER (6) PAGE (3) YEAR I SEQUENTIAL I REVISION 3 0 F NUMBER NUMBER 98 0 1 0 00 4 Following identification that 2SJ404 was the source of the leakage, 2SJ8 was closed, isolating 2SJ404. Valve 2SJ8 is the root valve in series with 2SJ404. Subsequent to closing 2SJ8, leakage into the RHR sump stopped. A follow-up RCS leak rate confirmed that the unidentified leakage was significantly reduced. CAUSE OF OCCURRENCE The cause of the event was the leaking 2SJ404 manual sample valve. PRIOR SIMILAR OCCURRENCES 1996, 1997 ano 1998 LERs were reviewed for similar occurrences. | | NRC FORM 366 U.S. NUCLEAR REGULATOR MISSION APPROVED BY 0. 3150-0104 EXPIRES 06/30/2001 (6-1998) Estimated burden p ponse to comply with this mandatory information collection request: 50 hrs. Reported lessons learned are incorporated into LICENSEE EVENT REPORT (LER) the licensing process and fed back to industry. Forward comments regarding burden estimate to the Records Management Branch (T-6 F33), U.S. |
| No similar events were identified. | | Nuclear Regulatory Commission, Washington, DC 20555-0001, and to the (See reverse for required number of Paperwork Reduction Project (3150-0104), Office of Management and digits/characters for each block) Budget, Washington, DC 20503. If an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection. |
| SAFETY CONSEQUENCES AND IMPLICATIONS The activity level of the reactor coolant was slightly elevated due to a suspected single fuel pin leak. However, there was no appreciable change to radiological conditions in the area of the 2SJ404 due to the leakage through the valve. Routine effluent monitoring and dose assessment for the period showed the dose to be well within technical specification limits. Therefore, during the period of leakage, there were no safety consequences to the health and safety of the public or to the plant staff. In an effort to assess the potential safety consequences, dose estimates were conducted. | | FACILITY NAME (1) |
| The estimates were performed using the actual measured leakage through 2SJ404 (0.25 gpm). These estimates showed offsite whole body accident dose to be comparable with the plant design basis for both the Exclusion Area Boundary (EAB) and the Low Population Zone (LPZ) and well within lOCFR Part 100 limits. Offsite accident thyroid doses for the EAB and LPZ while greater than the design basis, were well within the Part 100 limits. Potential onsite beta skin dose was estimated to remain within the General Design Criterion (GDC) 19 limit. However, whole body and thyroid doses, assuming a source term (the Salem design basis accident source) per the guidance of Regulatory Guide 1.4, are estimated to be slightly above the GDC 19 limits. The potential consequences of the event depend on the core damage frequency, which for Salem is 4.45xE-5/year. | | SALEM UNIT 2 DOCKET NUMBER (2) 05000311 !PAGET OF 4 TITLE(4) |
| While it is impossible to exactly NRC FORM 366A (6-1998) NRC FORM 366A (6-1998) U.S. NUCLEAR REGULATORY COMMISSION FACILITY NAME (1) SALEM UNIT 2 LICENSEE EVENT REPORT (LER) TEXT CONTINUATION DOCKET (2) NUMBER(2) 05000311 TEXT (If more space is required, use additional copies of NRC Form 366A) (17) LER NUMBER (6) PAGE (3) YEAR I 4 OF 4 98 0 1 0 00 identify a time when the 2SJ404 leakage became a significant component of the RCS unidentified leakage, formal troubleshooting to identify the leakage began two weeks prior to identification of *the leak. If this period is increased to a month, the core damage frequency for the period is approximately 3.7xE-6. CORRECTIVE ACTIONS 1. The leakage flow path was isolated restoring ECCS leakage from systems outside Containment to within UFSAR limits. 2. A work order has been initiated to inspect and repair the 2SJ404 valve. 3. Operator training will be provided on this event during an upcoming requalification cycle. This training will include discussion of the design basis leakage criteria for primary coolant sources outside containment. | | ECCS LEAKAGE OUTSIDE DESIGN BASIS VALUE. |
| 1 4. The program to reduce leakage from primary coolant sources outside containment (required by Technical Specification 6.8.4.a) will be reviewed and appropriate revisions will be NRC FORM 366A (6-1998)}}
| | EVENT DATE (5) LER NUMBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED (8) |
| | MONTH DAY YEAR YEAR SEQUENTIAL REVISION MONTH DAY YEAR FACILITY NAME DOCKET NUMBER NUMBER NUMBER 07 14 98 98 -010 00 08 13 98 FACILITY NAME DOCKET NUMBER Salem Unit 1 05000272 OPERATING 1 100% x NAME TELEPHONE NUMBER (Include Area Code) |
| | Philip J. Duca Jr. , Salem Licensing Engineer (609) 339-2381 COMPLETE AILURE DESC BED IN THIS REPORT (13) |
| | CAUSE SYSTEM COMPONE MANUFACTURER REPORTABLE CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE NT TO EPIX TO EPIX EXPECTED MONTH DAY YEAR YES (If yes, complete EXPECTED SUBMISSION DATE). |
| | ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16) |
| | On 7 /14/98 Operations personnel determined that leakage f.rom the Boron Injection Tank exceeded the maximum allowable ECCS leakage from sources outside Containment. |
| | The leakage path was determined to be past the valve seat of a manual sample valve (2SJ404) from the Boron Injection Tank. The valve was subsequently isolated. With the valve isolated, ECCS leakage returned to within UFSAR limits. |
| | The cause of the event was the leaking 2SJ404 manual sample valve. |
| | During the period of leakage, there were no actual safety consequences to the health and safety of the public or to the plant staff. Estimated potential post accident offsite doses remain well within 10CFR100 limits. Estimated post accident onsite beta skin dose is within the General Design Criterion 19 (GDC 19) limit, while estimated whole body and thyroid doses are slightly above the GDC 19 limits based on conservative estimates. Estimated core damage frequency during the period of the leakage, at approximately 3.7xE-6, decreases potential significance. |
| | This event is reportable pursuant to 10CFR50.73(a)2(ii) "any event or. |
| | condi tion ............... that resulted in the nuclear power plant being:...... (B) in a condition that was outside the design basis of the plant". |
| | |
| | . NRC FORM 366A U.S. NUCLEAR REG ATORY COMMISSION (6-1998) |
| | LICENSEE EVENT REPORT (LER) |
| | TEXT CONTINUATION FACILITY NAME (1) DOCKET (2) LER NUMBER (6) PAGE (3) |
| | NUMBER(2) |
| | SALEM UNIT 2 05000311 YEAR I SEQUENTIAL NUMBER IREVISION NUMBER 2 0F 4 98 0 1 0 00 TEXT (If more space is required, use additional copies ofNRC Form 366A) (17) |
| | PLANT AND SYSTEM* IDENTIFICATION Westinghouse - Pressurized Water Reactor High-Pressure Safety Injection System/Sample Valve {BQ/SMV}* |
| | * Energy Industry ~dentification System {EIIS} codes and component function identifier codes appear as (SS/CCC) |
| | CONDITIONS PRIOR TO OCCURRENCE At the time of identification, Salem Unit 2 was operating at 100% Power. |
| | DESCRIPTION OF OCCURRENCE On 7/14/98 Operations personnel determined that leakage from the Boron Injection Tank exceeded the maximum allowable ECCS Leakage as defined by UFSAR Section 6.3.2.11. The leakage was determined to be approximately 0.25 gpm (60,000cc/hour) past the valve seat of a manual sample valve (2SJ404) from the Boron Injection Tank. The leakage was flowing into the RHR sump in the Auxiliary Building, outside of Containment. UFSAR Section 6.3.2.11 addresses leakage during the recirculation phase of an accident, and states that "The total leakage resulting from all sources is about 3800 cc/hour as described in UFSAR Section 15.4.1. Recirculation loop leakage sources are summarized in Table 6.3-12. Leakage is monitored by procedure to ensure this leak rate is not exceeded." Leakage in excess of 3800 cc/hour placed the plant in a condition outside that assumed for design basis accident conditions. |
| | For about two weeks prior to the event, efforts were being focused on reducing an elevated reactor coolant system (RCS) unidentified leakage. |
| | Initially this leakage was thought to be within containment. The continuing investigations included determining the source of increasing in-leakage to the RHR sump located in the Auxiliary Building. |
| | These investigations used station procedures. One of these procedures provides a program for monitoring leakage from systems outside Containment that could contain highly radioactive fluids following an accident. The objective of the program is to detect and correct any degradation of the pressure boundaries for these systems and thereby reduce post-accident dose rates and airborne activity.in the Auxiliary Building in accordance with Technical Specification 6.8.4.a. |
| | NRC FORM 366A (6-1998) |
| | |
| | I~ |
| | **~========~*==========~*~====~ |
| | NRC FORM 366A (6-1998) |
| | U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER) |
| | TEXT CONTINUATION FACILITY NAME (1) DOCKET (2) LER NUMBER (6) PAGE (3) |
| | NUMBER(2) |
| | SALEM UNIT 2 05000311 YEAR ISEQUENTIAL NUMBER IREVISION NUMBER 3 0F 4 98 0 1 0 00 TEXT (If more space is required, use additional copies ofNRCForm 366A) (17) |
| | Following identification that 2SJ404 was the source of the leakage, 2SJ8 was closed, isolating 2SJ404. Valve 2SJ8 is the root valve in series with 2SJ404. Subsequent to closing 2SJ8, leakage into the RHR sump stopped. A follow-up RCS leak rate confirmed that the unidentified leakage was significantly reduced. |
| | CAUSE OF OCCURRENCE The cause of the event was the leaking 2SJ404 manual sample valve. |
| | PRIOR SIMILAR OCCURRENCES 1996, 1997 ano 1998 LERs were reviewed for similar occurrences. No similar events were identified. |
| | SAFETY CONSEQUENCES AND IMPLICATIONS The activity level of the reactor coolant was slightly elevated due to a suspected single fuel pin leak. However, there was no appreciable change to radiological conditions in the area of the 2SJ404 due to the leakage through the valve. Routine effluent monitoring and dose assessment for the period showed the dose to be well within technical specification limits. |
| | Therefore, during the period of leakage, there were no safety consequences to the health and safety of the public or to the plant staff. |
| | In an effort to assess the potential safety consequences, dose estimates were conducted. The estimates were performed using the actual measured leakage through 2SJ404 (0.25 gpm). These estimates showed offsite whole body accident dose to be comparable with the plant design basis for both the Exclusion Area Boundary (EAB) and the Low Population Zone (LPZ) and well within 10CFR Part 100 limits. Offsite accident thyroid doses for the EAB and LPZ while greater than the design basis, were well within the Part 100 limits. |
| | Potential onsite beta skin dose was estimated to remain within the General Design Criterion (GDC) 19 limit. However, whole body and thyroid doses, assuming a source term (the Salem design basis accident source) per the guidance of Regulatory Guide 1.4, are estimated to be slightly above the GDC 19 limits. |
| | The potential consequences of the event depend on the core damage frequency, which for Salem is 4.45xE-5/year. While it is impossible to exactly |
| | |
| | NRC FORM 366A (6-1998) |
| | NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (6-1998) |
| | LICENSEE EVENT REPORT (LER) |
| | TEXT CONTINUATION FACILITY NAME (1) DOCKET (2) LER NUMBER (6) PAGE (3) |
| | NUMBER(2) |
| | SALEM UNIT 2 05000311 YEAR I SE~~ 1=~~ 4 OF 4 98 0 1 0 00 TEXT (If more space is required, use additional copies ofNRC Form 366A) (17) identify a time when the 2SJ404 leakage became a significant component of the RCS unidentified leakage, formal troubleshooting to identify the leakage began two weeks prior to identification of *the leak. If this period is increased to a month, the core damage frequency for the period is approximately 3.7xE-6. |
| | CORRECTIVE ACTIONS |
| | : 1. The leakage flow path was isolated restoring ECCS leakage from systems outside Containment to within UFSAR limits. |
| | : 2. A work order has been initiated to inspect and repair the 2SJ404 valve. |
| | : 3. Operator training will be provided on this event during an upcoming requalification cycle. This training will include discussion of the design basis leakage criteria for primary coolant sources outside containment. 1 |
| | : 4. The program to reduce leakage from primary coolant sources outside containment (required by Technical Specification 6.8.4.a) will be reviewed and appropriate revisions will be made~ |
| | NRC FORM 366A (6-1998)}} |
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Existing Procedures Will Be Evaluated.With 990602 Ltr ML18107A3541999-06-0101 June 1999 LER 99-006-00:on 990501,HHSI Flow Balance Discrepancy Was Noted During Surveillance.Caused by Sticking of Check Valve in SI Discharge Line to 21 Cold Leg.Valve 21SJ17,was Cut Out of Sys & Replaced.With 990601 Ltr ML18107A2931999-05-12012 May 1999 LER 99-002-00:on 990413,determined That Number 12 Auxiliary Bldg Exhaust Fan Was Rotating Backwards.Caused by mis-wiring of Motor Due to Human Error by Maint technician.Mis-wiring Was Corrected & Fan Was Returned to Svc.With 990512 Ltr ML18107A2781999-05-10010 May 1999 LER 99-004-00:on 990411,automatic Actuation of ESF Occurred During Reactor Vessel Head Removal in Support of Refueling Operations.Caused by High Radiation Condition.Containment Atmosphere Was Monitored.With 990505 Ltr ML18107A2791999-05-0404 May 1999 LER 99-003-00:on 990406,all Salem Unit 2 Chillers Rendered Inoperable.Caused by Human Error.Lessons Learned from Event Were Communicated to All Operators by Including Them in Night Orders.With 990504 Ltr ML18107A2741999-05-0303 May 1999 LER 99-002-00:on 990405,determined That Containment Isolation Valve Failed as Found Leakrate Test.Caused by Foreign Matl Blocking Valves from Closing.Check Valve Mechanically Agitated.With 990504 Ltr ML18107A2351999-04-23023 April 1999 LER 99-001-00:on 990330,MSSV Failed Lift Set Test.Caused by Setpoint Variance Which Is Result of Aging.Valves Were Adjusted & Retested to Ensure TS Tolerance.With 990423 Ltr ML18106B1471999-03-29029 March 1999 LER 99-001-00:on 990228,reactor Scram Was Noted as Result of Turbine Trip.Caused by Operator Error.Lesson Plans Revised to Explicitly Demonstrate Manner in Which Valve Functions. with 990329 Ltr ML18106B0701999-02-16016 February 1999 LER 98-015-00:on 981208,inadvertent Discharge Through RHR Relief Valve During Startup Was Noted.Caused by Operator Performing Too Many Tasks Simultaneously.Appropriate Actions Have Been Taken IAW Policies & Procedures.With 990216 Ltr ML18106B0491999-01-28028 January 1999 LER 98-007-01:on 980730,reactor Coolant Instrument Line through-wall Leak Was Noted.Caused by Transgranular Stress Corrosion Cracking.Replaced Affected Tubing.With 990128 Ltr ML18106B0401999-01-18018 January 1999 LER 98-016-00:on 981219,ECCS Leakage Was Outside of Design Value.Caused by Leakage Past Seat of 21RH34 Manual Drain. Valve 21RH34 Was Reseated.With 990118 Ltr ML18106B0081998-12-24024 December 1998 LER 97-001-01:on 970215,failure to Perform TS Surveillance of Component Cooling Water Sys Check Valves Occurred.Caused by Inadequate Communication Between EOP Group & IST Reviewers.Procedure Revised.With 981224 Ltr ML18106B0021998-12-17017 December 1998 LER 98-015-01:on 980924,improper Installation of Test Equipment to RPS Occurred.Caused by Inadequate 10CFR50.59 Applicability Reviews During Past Revs.Revised Procedures. with 981217 Ltr ML18106A9551998-11-0303 November 1998 LER 96-013-01:on 960711,concluded That Current Gain & Bias Settings Had Rendered Overtemperature Delta Temp Protection Channels Inoperable.Caused by Scaling Error.Licensee Will Revise Scaling Calculations.With 981105 Ltr ML18106A9451998-10-30030 October 1998 LER 97-004-01:on 970408,failure to Comply with TS Action Statement,Dg Start & Inadequate Surveillance Testing,Was Noted.Caused by Inadequate Tracking of Inoperable Equipment. Discussed Event & Lessons Learned.With 981022 Ltr ML18106A9491998-10-22022 October 1998 LER 98-015-00:on 980924,identified Improper Installation of Test Equipment to Rps.Cause Indeterminate.Procedures for Installation of Test Equipment for Collection of State Point Data Were Placed on Administrative Hold.With 981022 Ltr ML18106A9301998-10-21021 October 1998 LER 98-014-00:on 980725,noted Improper Calibr of Liquid Radwaste Effluent Line Radiation Monitor.Caused by Inattention to Detail by Maint Personnel.Channel Calibr Was Successfully Performed on 1R18 on 980821.With 981019 Ltr ML18106A9071998-10-0101 October 1998 LER 98-014-00:on 980918,discovered That Fire Barrier Matl for HVAC Ducts Does Not Meet Required Level of Fire Resistance.Cause Indeterminate.Established Appropriate Compensatory Actions for Fire Barriers.With 981001 Ltr ML18106A8951998-09-28028 September 1998 LER 98-012-01:on 980725,noted That Afs Was Operated with Less than Required Number of Operable AFW Pumps.Caused by Improper Procedure Implementation.Runout Protection Pressure Device for 22 AFW Pumps Was Returned to Svc.With 980928 Ltr ML18106A8821998-09-21021 September 1998 LER 98-013-00:on 980820,noted Surveillance of Containment Penetration Overcurrent Protection Devices Missed.Caused by Human Error.Satisfactorily Tested Apprpriate Breakers & Disciplined Involved Personnel.With 980921 Ltr ML18106A8791998-09-16016 September 1998 LER 96-006-01:on 960717,determined That non-radioactive Liquid Basin Radwaste Monitor Inoperable During Low Head Conditions.Caused by Inadequate Design Change Package.Design Change 1EC3663-01 Has Been Installed.With 980916 Ltr ML18106A8801998-09-0808 September 1998 LER 98-013-00:on 980806,operation with TS Required Equipment OOS Was Noted.Caused by Human Error.Reviewed Processes & Practices Re Safety Sys Status Control,Procedure Rev & Extra Training.With 980908 Ltr ML18106A8531998-08-27027 August 1998 LER 98-007-00:on 980730,reactor Coolant Instrument Line through-wall Leak Was Noted.Cause of Event Has Not Yet Been Determined.Assembled Root Cause Team & Replaced Affected tubing.W/980827 Ltr ML18106A8521998-08-27027 August 1998 LER 98-011-00:on 980803,ESFA During a 4KV Automatic Transfer Test Was Noted.Caused by Premature Release of Control Console Pushbutton Due to Inadequate Procedural Step.Revised procedure.W/980827 Ltr ML18106A8421998-08-24024 August 1998 LER 98-012-00:on 980725,discovered That Plant Had Operated in Modes 1 & 2 w/twenty-two AFW Pumps Inoperable.Caused by Failure to Restore Pump Runout Protection Pressure Device to Svc.Returned Subject Device to svc.W/980824 Ltr ML18106A8431998-08-24024 August 1998 LER 98-009-00:on 980810,failure to Post Continuous Firewatch as Required by Fire Protection Plan Noted.Caused by Failure to Recognize Concurrent Conditions.Continuous Firewatch Was Posted Immediately & Repaired Smoke detectors.W/980824 Ltr ML18106A8141998-08-13013 August 1998 LER 98-010-00:on 980714,determined That Leakage from Boron Injection Tank Exceeded Max Allowable ECCS Leakage from Sources Outside Containment.Caused by Leaking 2SJ404 Manual Sample valve.2SJ404 Valve repaired.W/980813 Ltr ML18106A8201998-08-13013 August 1998 LER 98-012-00:on 980715,potential to Exceed Rating of Piping Due to Isolation of Overpressure Protection Line Was Noted. Caused by Inadequate Procedural Guidance.Appropriate Operations Dept Procedures Have Been revised.W/980813 Ltr ML18106A6931998-06-29029 June 1998 LER 98-003-00:on 980122,inappropriate Plugging of Tubes R9C60 & R10C60 in Salem Unit 2 Sg,Was Performed.Caused by Failure of Qualification,Verification & Validation Process. Tubes Reviewed to Verify No Others Inappropriately Plugged ML18106A6471998-06-0404 June 1998 LER 98-011-00:on 980505,improper Isolation of Single Cell Battery Charger from 125 Vdc Battery Was Noted.Caused by Inadequate 10CFR50.59 Applicability Review.Placed Procedure SC.MD-CM.ZZ-0024(Q) on Administrative hold.W/980604 Ltr ML18106A6421998-06-0101 June 1998 LER 98-010-00:on 931019,reactor Pressure Vessel Insp Plugs Were Out of Configuration,Was Noted.Caused by Personnel Error.Proper Configuration Was Restored Shortly After Discovery Prior to Entering Mode 2.W/980601 Ltr ML18106A6431998-05-29029 May 1998 LER 98-006-01:on 980227,determined Incorrect Scaling Error of First Stage Pressure Transmitter Existed.Caused by Human Error.Revised Setpoint Calculation SC-MS002-01 & Revised Associated Instrument Calibr Database info.W/980529 Ltr ML18106A6141998-05-18018 May 1998 LER 98-008-00:on 970814,failure to Test 21 & 22 AF 40 Valves in Closed Direction as Required by TS 4.0.5 Was Noted.Caused by Inadequate Design Mod Process.Motor Driven 21/22 AF 40 Valves Were Tested IAW Revised procedure.W/980518 Ltr ML18106A5901998-05-0101 May 1998 LER 98-009-00:on 980405,epoxy Missing from Terminals of H Analyzer Was Noted.Caused by Inadequate Development of Procedure.H Analyzers Were Repaired & Review of Other Safety Related Equipment in Containment Was performed.W/980501 Ltr ML18106A5611998-04-20020 April 1998 LER 98-008-00:on 980323,inadequate Testing of Salem Unit 1 Containment Air Locks Resulted in Entering TS 3.0.3.Caused by less-than-adequate Work Practices During Replacement of Equalizing Valve.Salem Unit 2 Airlocks Were Inspected ML18106A6061998-04-0101 April 1998 Corrected LER 98-004-00:on 980302,failure to Comply W/Tss 4.11.1.1.2 & 3.3.3.8 Was Noted.Caused by Organizational Deficiency.Steps Have Been Taken to Correctly Document Safety Factors.Corrects Prior Similar Occurrences ML18106A4451998-04-0101 April 1998 LER 98-004-00:on 980302,failure to Perform TS 4.11.1.1.2 & 3.3.3.8 Was Noted.Caused by Organizational Deficiency.Steps Were Taken to Correctly Document Safety factors.W/980401 Ltr ML18106A4351998-03-30030 March 1998 LER 98-006-00:on 980227,incorrect Scaling of First Stage Turbine Impulse Pressure Transmitters Noted.Cause Indeterminate.Implemented Procedure Changes & re-scaled Affected Turbine Impulse Pressure transmitters.W/980330 Ltr ML18106A3961998-03-20020 March 1998 LER 98-005-00:on 980219,inoperability of Twelve EDG Fuel Oil Transfer Pump (FOTP) Noted.Caused by Installation of Incorrect Control Switch.Installed Correct off-auto-manual Switch & Verified Operability of Twelve FOTP.W/980320 Ltr ML18106A5781998-03-20020 March 1998 Corrected LER 98-005-00:on 980219,inoperability of 12 Fuel Oil Transfer Pump (Fotp),Noted.Caused by Installation of Incorrect Control Switch.Field Insp Performed to Verify Configuration of Switches for 11,21 & 22 FOTPs ML18106A4021998-03-20020 March 1998 LER 98-007-00:on 980218,failure to Establish Containment Integrity (Closure) Prior to Fuel Movement Was Noted.Caused by Failure to Identify & Include Condensate Pot Vent in Appropriate Valve Lineup.Valves identified.W/980320 Ltr ML18106A4031998-03-20020 March 1998 LER 98-006-00:on 980221,ESF Actuation of 11 & 12 Auxiliary Feedwater Pumps Occurred.Caused by Human Error.Operators Promptly Established Feedwater to All SG & Restored Proper Water levels.W/980320 Ltr 1999-08-26
[Table view] Category:RO)
MONTHYEARML18107A5031999-08-26026 August 1999 LER 99-006-00:on 990729,determined That SG Blowdown RMs Setpoint Was non-conservative.Caused by Inadequate ACs for Incorporating Original Plant Licensing Data Into Plant Procedures.Blowdown Will Be Restricted.With 990826 Ltr ML18107A4691999-07-28028 July 1999 LER 99-008-00:on 990714,determined That Limit Switch Cables Were Subject to Multiple Hot Shorts in Same Fire Area.Caused by Inadequate Original Post Fire Safe Shutdown Analysis.All Limit Switch Cables for MOVs Were Reviewed.With 990728 Ltr ML18107A4441999-07-0606 July 1999 LER 99-007-00:on 990605,surveillance for Quadrant Power Tilt Ratio (QPTR) Was Missed.Caused by Human Error.Qptr Calculation Was Performed & Personnel Involved Have Been Held Accountable IAW Pse&G Policies.With 990706 Ltr ML18107A4211999-07-0202 July 1999 LER 99-005-00:on 990605,11 Containment Declared Inoperable. Caused by Valves 11SW72 & 11SW223 Both Leaking.Procedure S1.OP-ST.SW-0010(Q) Was Enhanced to Provide Specific Instructions to Ensure Proper Sequencing.With 990702 Ltr ML18107A4321999-07-0101 July 1999 LER 99-006-01:on 990501,determined That There Was No Flow in One of Four Injection Legs.Caused by Sticking of Valve in Safety Injection Discharge Line to 21 Cold Leg.Valve Was Cut Out of Sys & Replaced.With 990701 Ltr ML18107A4331999-07-0101 July 1999 LER 99-002-01:on 990405,determined That 2SA118 Failed as Found Leakrate Test.Caused by Foreign Matl Found in 2SA118 valve.2SA118 Valve Was Cycled Several Times & Seat Area Was Air Blown in Order to Displace Foreign Matl.With 990701 Ltr ML18107A3951999-06-17017 June 1999 LER 99-004-00:on 990520,reactor Tripped from 100% Power,Due to Negative Flux Trip Signal from Nuclear Instrumentation. Cause Has Not Been Determined.Discoloration Was Identified on One of Penetrations.With 990617 Ltr ML18107A3661999-06-0909 June 1999 LER 99-003-00:on 990513,unplanned Entry Into TS 3.0.3 Was Made.Caused by Human error.Re-positioned Creacs Supply Fan Selector Switches & Revised Procedures S1 & S2.OP-ST.SSP-0001(Q).With 990609 Ltr ML18107A3551999-06-0202 June 1999 LER 99-005-00:on 990504,failure to Meet TS Action Statement Requirements for High Oxygen Concentration in Waste Gas Holdup Sys Occurred.Caused by Inability of Operators. Existing Procedures Will Be Evaluated.With 990602 Ltr ML18107A3541999-06-0101 June 1999 LER 99-006-00:on 990501,HHSI Flow Balance Discrepancy Was Noted During Surveillance.Caused by Sticking of Check Valve in SI Discharge Line to 21 Cold Leg.Valve 21SJ17,was Cut Out of Sys & Replaced.With 990601 Ltr ML18107A2931999-05-12012 May 1999 LER 99-002-00:on 990413,determined That Number 12 Auxiliary Bldg Exhaust Fan Was Rotating Backwards.Caused by mis-wiring of Motor Due to Human Error by Maint technician.Mis-wiring Was Corrected & Fan Was Returned to Svc.With 990512 Ltr ML18107A2781999-05-10010 May 1999 LER 99-004-00:on 990411,automatic Actuation of ESF Occurred During Reactor Vessel Head Removal in Support of Refueling Operations.Caused by High Radiation Condition.Containment Atmosphere Was Monitored.With 990505 Ltr ML18107A2791999-05-0404 May 1999 LER 99-003-00:on 990406,all Salem Unit 2 Chillers Rendered Inoperable.Caused by Human Error.Lessons Learned from Event Were Communicated to All Operators by Including Them in Night Orders.With 990504 Ltr ML18107A2741999-05-0303 May 1999 LER 99-002-00:on 990405,determined That Containment Isolation Valve Failed as Found Leakrate Test.Caused by Foreign Matl Blocking Valves from Closing.Check Valve Mechanically Agitated.With 990504 Ltr ML18107A2351999-04-23023 April 1999 LER 99-001-00:on 990330,MSSV Failed Lift Set Test.Caused by Setpoint Variance Which Is Result of Aging.Valves Were Adjusted & Retested to Ensure TS Tolerance.With 990423 Ltr ML18106B1471999-03-29029 March 1999 LER 99-001-00:on 990228,reactor Scram Was Noted as Result of Turbine Trip.Caused by Operator Error.Lesson Plans Revised to Explicitly Demonstrate Manner in Which Valve Functions. with 990329 Ltr ML18106B0701999-02-16016 February 1999 LER 98-015-00:on 981208,inadvertent Discharge Through RHR Relief Valve During Startup Was Noted.Caused by Operator Performing Too Many Tasks Simultaneously.Appropriate Actions Have Been Taken IAW Policies & Procedures.With 990216 Ltr ML18106B0491999-01-28028 January 1999 LER 98-007-01:on 980730,reactor Coolant Instrument Line through-wall Leak Was Noted.Caused by Transgranular Stress Corrosion Cracking.Replaced Affected Tubing.With 990128 Ltr ML18106B0401999-01-18018 January 1999 LER 98-016-00:on 981219,ECCS Leakage Was Outside of Design Value.Caused by Leakage Past Seat of 21RH34 Manual Drain. Valve 21RH34 Was Reseated.With 990118 Ltr ML18106B0081998-12-24024 December 1998 LER 97-001-01:on 970215,failure to Perform TS Surveillance of Component Cooling Water Sys Check Valves Occurred.Caused by Inadequate Communication Between EOP Group & IST Reviewers.Procedure Revised.With 981224 Ltr ML18106B0021998-12-17017 December 1998 LER 98-015-01:on 980924,improper Installation of Test Equipment to RPS Occurred.Caused by Inadequate 10CFR50.59 Applicability Reviews During Past Revs.Revised Procedures. with 981217 Ltr ML18106A9551998-11-0303 November 1998 LER 96-013-01:on 960711,concluded That Current Gain & Bias Settings Had Rendered Overtemperature Delta Temp Protection Channels Inoperable.Caused by Scaling Error.Licensee Will Revise Scaling Calculations.With 981105 Ltr ML18106A9451998-10-30030 October 1998 LER 97-004-01:on 970408,failure to Comply with TS Action Statement,Dg Start & Inadequate Surveillance Testing,Was Noted.Caused by Inadequate Tracking of Inoperable Equipment. Discussed Event & Lessons Learned.With 981022 Ltr ML18106A9491998-10-22022 October 1998 LER 98-015-00:on 980924,identified Improper Installation of Test Equipment to Rps.Cause Indeterminate.Procedures for Installation of Test Equipment for Collection of State Point Data Were Placed on Administrative Hold.With 981022 Ltr ML18106A9301998-10-21021 October 1998 LER 98-014-00:on 980725,noted Improper Calibr of Liquid Radwaste Effluent Line Radiation Monitor.Caused by Inattention to Detail by Maint Personnel.Channel Calibr Was Successfully Performed on 1R18 on 980821.With 981019 Ltr ML18106A9071998-10-0101 October 1998 LER 98-014-00:on 980918,discovered That Fire Barrier Matl for HVAC Ducts Does Not Meet Required Level of Fire Resistance.Cause Indeterminate.Established Appropriate Compensatory Actions for Fire Barriers.With 981001 Ltr ML18106A8951998-09-28028 September 1998 LER 98-012-01:on 980725,noted That Afs Was Operated with Less than Required Number of Operable AFW Pumps.Caused by Improper Procedure Implementation.Runout Protection Pressure Device for 22 AFW Pumps Was Returned to Svc.With 980928 Ltr ML18106A8821998-09-21021 September 1998 LER 98-013-00:on 980820,noted Surveillance of Containment Penetration Overcurrent Protection Devices Missed.Caused by Human Error.Satisfactorily Tested Apprpriate Breakers & Disciplined Involved Personnel.With 980921 Ltr ML18106A8791998-09-16016 September 1998 LER 96-006-01:on 960717,determined That non-radioactive Liquid Basin Radwaste Monitor Inoperable During Low Head Conditions.Caused by Inadequate Design Change Package.Design Change 1EC3663-01 Has Been Installed.With 980916 Ltr ML18106A8801998-09-0808 September 1998 LER 98-013-00:on 980806,operation with TS Required Equipment OOS Was Noted.Caused by Human Error.Reviewed Processes & Practices Re Safety Sys Status Control,Procedure Rev & Extra Training.With 980908 Ltr ML18106A8531998-08-27027 August 1998 LER 98-007-00:on 980730,reactor Coolant Instrument Line through-wall Leak Was Noted.Cause of Event Has Not Yet Been Determined.Assembled Root Cause Team & Replaced Affected tubing.W/980827 Ltr ML18106A8521998-08-27027 August 1998 LER 98-011-00:on 980803,ESFA During a 4KV Automatic Transfer Test Was Noted.Caused by Premature Release of Control Console Pushbutton Due to Inadequate Procedural Step.Revised procedure.W/980827 Ltr ML18106A8421998-08-24024 August 1998 LER 98-012-00:on 980725,discovered That Plant Had Operated in Modes 1 & 2 w/twenty-two AFW Pumps Inoperable.Caused by Failure to Restore Pump Runout Protection Pressure Device to Svc.Returned Subject Device to svc.W/980824 Ltr ML18106A8431998-08-24024 August 1998 LER 98-009-00:on 980810,failure to Post Continuous Firewatch as Required by Fire Protection Plan Noted.Caused by Failure to Recognize Concurrent Conditions.Continuous Firewatch Was Posted Immediately & Repaired Smoke detectors.W/980824 Ltr ML18106A8141998-08-13013 August 1998 LER 98-010-00:on 980714,determined That Leakage from Boron Injection Tank Exceeded Max Allowable ECCS Leakage from Sources Outside Containment.Caused by Leaking 2SJ404 Manual Sample valve.2SJ404 Valve repaired.W/980813 Ltr ML18106A8201998-08-13013 August 1998 LER 98-012-00:on 980715,potential to Exceed Rating of Piping Due to Isolation of Overpressure Protection Line Was Noted. Caused by Inadequate Procedural Guidance.Appropriate Operations Dept Procedures Have Been revised.W/980813 Ltr ML18106A6931998-06-29029 June 1998 LER 98-003-00:on 980122,inappropriate Plugging of Tubes R9C60 & R10C60 in Salem Unit 2 Sg,Was Performed.Caused by Failure of Qualification,Verification & Validation Process. Tubes Reviewed to Verify No Others Inappropriately Plugged ML18106A6471998-06-0404 June 1998 LER 98-011-00:on 980505,improper Isolation of Single Cell Battery Charger from 125 Vdc Battery Was Noted.Caused by Inadequate 10CFR50.59 Applicability Review.Placed Procedure SC.MD-CM.ZZ-0024(Q) on Administrative hold.W/980604 Ltr ML18106A6421998-06-0101 June 1998 LER 98-010-00:on 931019,reactor Pressure Vessel Insp Plugs Were Out of Configuration,Was Noted.Caused by Personnel Error.Proper Configuration Was Restored Shortly After Discovery Prior to Entering Mode 2.W/980601 Ltr ML18106A6431998-05-29029 May 1998 LER 98-006-01:on 980227,determined Incorrect Scaling Error of First Stage Pressure Transmitter Existed.Caused by Human Error.Revised Setpoint Calculation SC-MS002-01 & Revised Associated Instrument Calibr Database info.W/980529 Ltr ML18106A6141998-05-18018 May 1998 LER 98-008-00:on 970814,failure to Test 21 & 22 AF 40 Valves in Closed Direction as Required by TS 4.0.5 Was Noted.Caused by Inadequate Design Mod Process.Motor Driven 21/22 AF 40 Valves Were Tested IAW Revised procedure.W/980518 Ltr ML18106A5901998-05-0101 May 1998 LER 98-009-00:on 980405,epoxy Missing from Terminals of H Analyzer Was Noted.Caused by Inadequate Development of Procedure.H Analyzers Were Repaired & Review of Other Safety Related Equipment in Containment Was performed.W/980501 Ltr ML18106A5611998-04-20020 April 1998 LER 98-008-00:on 980323,inadequate Testing of Salem Unit 1 Containment Air Locks Resulted in Entering TS 3.0.3.Caused by less-than-adequate Work Practices During Replacement of Equalizing Valve.Salem Unit 2 Airlocks Were Inspected ML18106A6061998-04-0101 April 1998 Corrected LER 98-004-00:on 980302,failure to Comply W/Tss 4.11.1.1.2 & 3.3.3.8 Was Noted.Caused by Organizational Deficiency.Steps Have Been Taken to Correctly Document Safety Factors.Corrects Prior Similar Occurrences ML18106A4451998-04-0101 April 1998 LER 98-004-00:on 980302,failure to Perform TS 4.11.1.1.2 & 3.3.3.8 Was Noted.Caused by Organizational Deficiency.Steps Were Taken to Correctly Document Safety factors.W/980401 Ltr ML18106A4351998-03-30030 March 1998 LER 98-006-00:on 980227,incorrect Scaling of First Stage Turbine Impulse Pressure Transmitters Noted.Cause Indeterminate.Implemented Procedure Changes & re-scaled Affected Turbine Impulse Pressure transmitters.W/980330 Ltr ML18106A3961998-03-20020 March 1998 LER 98-005-00:on 980219,inoperability of Twelve EDG Fuel Oil Transfer Pump (FOTP) Noted.Caused by Installation of Incorrect Control Switch.Installed Correct off-auto-manual Switch & Verified Operability of Twelve FOTP.W/980320 Ltr ML18106A5781998-03-20020 March 1998 Corrected LER 98-005-00:on 980219,inoperability of 12 Fuel Oil Transfer Pump (Fotp),Noted.Caused by Installation of Incorrect Control Switch.Field Insp Performed to Verify Configuration of Switches for 11,21 & 22 FOTPs ML18106A4021998-03-20020 March 1998 LER 98-007-00:on 980218,failure to Establish Containment Integrity (Closure) Prior to Fuel Movement Was Noted.Caused by Failure to Identify & Include Condensate Pot Vent in Appropriate Valve Lineup.Valves identified.W/980320 Ltr ML18106A4031998-03-20020 March 1998 LER 98-006-00:on 980221,ESF Actuation of 11 & 12 Auxiliary Feedwater Pumps Occurred.Caused by Human Error.Operators Promptly Established Feedwater to All SG & Restored Proper Water levels.W/980320 Ltr 1999-08-26
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217A9931999-09-30030 September 1999 NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data ML18107A5581999-09-30030 September 1999 Monthly Operating Rept for Sept 1999 for Salem,Unit 2.With 991014 Ltr ML18107A5571999-09-30030 September 1999 Monthly Operating Rept for Sept 1999 for Salem,Unit 1.With 991014 Ltr ML18107A5301999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Salem,Unit 2.With 990913 Ltr ML18107A5311999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Salem,Unit 1.With 990913 ML18107A5031999-08-26026 August 1999 LER 99-006-00:on 990729,determined That SG Blowdown RMs Setpoint Was non-conservative.Caused by Inadequate ACs for Incorporating Original Plant Licensing Data Into Plant Procedures.Blowdown Will Be Restricted.With 990826 Ltr ML18107A5201999-08-12012 August 1999 Rev 0 to Sgs Unit 2 ISI RFO Exam Results (S2RFO#9) Second Interval,Second Period, First Outage (96RF). ML18107A4811999-07-31031 July 1999 Monthly Operating Rept for July 1999 for Salem,Unit 1.With 990813 Ltr ML18107A4821999-07-31031 July 1999 Monthly Operating Rept for July 1999 for Salem,Unit 2.With 990813 Ltr ML18107A4691999-07-28028 July 1999 LER 99-008-00:on 990714,determined That Limit Switch Cables Were Subject to Multiple Hot Shorts in Same Fire Area.Caused by Inadequate Original Post Fire Safe Shutdown Analysis.All Limit Switch Cables for MOVs Were Reviewed.With 990728 Ltr ML18107A4441999-07-0606 July 1999 LER 99-007-00:on 990605,surveillance for Quadrant Power Tilt Ratio (QPTR) Was Missed.Caused by Human Error.Qptr Calculation Was Performed & Personnel Involved Have Been Held Accountable IAW Pse&G Policies.With 990706 Ltr ML18107A4211999-07-0202 July 1999 LER 99-005-00:on 990605,11 Containment Declared Inoperable. Caused by Valves 11SW72 & 11SW223 Both Leaking.Procedure S1.OP-ST.SW-0010(Q) Was Enhanced to Provide Specific Instructions to Ensure Proper Sequencing.With 990702 Ltr ML18107A4331999-07-0101 July 1999 LER 99-002-01:on 990405,determined That 2SA118 Failed as Found Leakrate Test.Caused by Foreign Matl Found in 2SA118 valve.2SA118 Valve Was Cycled Several Times & Seat Area Was Air Blown in Order to Displace Foreign Matl.With 990701 Ltr ML18107A4321999-07-0101 July 1999 LER 99-006-01:on 990501,determined That There Was No Flow in One of Four Injection Legs.Caused by Sticking of Valve in Safety Injection Discharge Line to 21 Cold Leg.Valve Was Cut Out of Sys & Replaced.With 990701 Ltr ML18107A5211999-07-0101 July 1999 Rev 0 to Sgs Unit 2 ISI RFO Exam Results (S2RFO#10) Second Interval,Second Period,Second Outage (99RF). ML18107A4351999-06-30030 June 1999 Monthly Operating Rept for June 1999 for Salem,Unit 1.With 990713 Ltr ML18107A4341999-06-30030 June 1999 Monthly Operating Rept for June 1999 for Salem,Unit 2.With 990713 Ltr ML20196H8621999-06-30030 June 1999 NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data, June 1999 Rept ML18107A3951999-06-17017 June 1999 LER 99-004-00:on 990520,reactor Tripped from 100% Power,Due to Negative Flux Trip Signal from Nuclear Instrumentation. Cause Has Not Been Determined.Discoloration Was Identified on One of Penetrations.With 990617 Ltr ML18107A3661999-06-0909 June 1999 LER 99-003-00:on 990513,unplanned Entry Into TS 3.0.3 Was Made.Caused by Human error.Re-positioned Creacs Supply Fan Selector Switches & Revised Procedures S1 & S2.OP-ST.SSP-0001(Q).With 990609 Ltr ML18107A3551999-06-0202 June 1999 LER 99-005-00:on 990504,failure to Meet TS Action Statement Requirements for High Oxygen Concentration in Waste Gas Holdup Sys Occurred.Caused by Inability of Operators. Existing Procedures Will Be Evaluated.With 990602 Ltr ML18107A3441999-06-0101 June 1999 Interim Part 21 Rept Re Premature Over Voltage Protection Actuation in Circuit Specific Application in Dc Power Supply.Testing & Evaluation Activities Will Be Completed on 990716 ML18107A3541999-06-0101 June 1999 LER 99-006-00:on 990501,HHSI Flow Balance Discrepancy Was Noted During Surveillance.Caused by Sticking of Check Valve in SI Discharge Line to 21 Cold Leg.Valve 21SJ17,was Cut Out of Sys & Replaced.With 990601 Ltr ML18107A3681999-05-31031 May 1999 Monthly Operating Rept for May 1999 for Salem Generating Station,Unit 1.With 990611 Ltr ML18107A3721999-05-31031 May 1999 Monthly Operating Rept for May 1999 for Salem Generating Station,Unit 2.With 990611 Ltr ML18107A2931999-05-12012 May 1999 LER 99-002-00:on 990413,determined That Number 12 Auxiliary Bldg Exhaust Fan Was Rotating Backwards.Caused by mis-wiring of Motor Due to Human Error by Maint technician.Mis-wiring Was Corrected & Fan Was Returned to Svc.With 990512 Ltr ML18107A2781999-05-10010 May 1999 LER 99-004-00:on 990411,automatic Actuation of ESF Occurred During Reactor Vessel Head Removal in Support of Refueling Operations.Caused by High Radiation Condition.Containment Atmosphere Was Monitored.With 990505 Ltr ML18107A2791999-05-0404 May 1999 LER 99-003-00:on 990406,all Salem Unit 2 Chillers Rendered Inoperable.Caused by Human Error.Lessons Learned from Event Were Communicated to All Operators by Including Them in Night Orders.With 990504 Ltr ML18107A2741999-05-0303 May 1999 LER 99-002-00:on 990405,determined That Containment Isolation Valve Failed as Found Leakrate Test.Caused by Foreign Matl Blocking Valves from Closing.Check Valve Mechanically Agitated.With 990504 Ltr ML18107A3711999-04-30030 April 1999 Corrected Monthly Operating Rept for Apr 1999 for Salem Generating Station,Unit 1 ML18107A3151999-04-30030 April 1999 Submittal-Only Screening Review of Salem Generating Station Individual Plant Exam for External Events (Seismic Portion), Rev 1 ML18107A2991999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for Salem Unit 1.With 990514 Ltr ML18107A2971999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for Salem Unit 2.With 990514 Ltr ML18107A2351999-04-23023 April 1999 LER 99-001-00:on 990330,MSSV Failed Lift Set Test.Caused by Setpoint Variance Which Is Result of Aging.Valves Were Adjusted & Retested to Ensure TS Tolerance.With 990423 Ltr ML18107A2881999-04-0707 April 1999 Rev 0 to NFS-0174, COLR for Salem Unit 2 Cycle 11. ML18107A1821999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for Salem,Unit 1.With 990414 Ltr ML18107A1831999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for Salem,Unit 2.With 990414 Ltr ML18106B1471999-03-29029 March 1999 LER 99-001-00:on 990228,reactor Scram Was Noted as Result of Turbine Trip.Caused by Operator Error.Lesson Plans Revised to Explicitly Demonstrate Manner in Which Valve Functions. with 990329 Ltr ML18106B1021999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for Salem Unit 2.With 990315 Ltr ML18106B1011999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for Salem Unit 1.With 990315 Ltr ML18106B0931999-02-25025 February 1999 Part 21 Rept Re Possible Defect in Swagelok Pipe Fitting Tee,Part Number SS-6-T.Caused by Crack Due to Improper Location of Heated Bar.Only One Part Out of 7396 Pieces in Forging Lot Was Found to Be Cracked.Affected Util,Notified ML18106B0701999-02-16016 February 1999 LER 98-015-00:on 981208,inadvertent Discharge Through RHR Relief Valve During Startup Was Noted.Caused by Operator Performing Too Many Tasks Simultaneously.Appropriate Actions Have Been Taken IAW Policies & Procedures.With 990216 Ltr ML18106B0551999-02-0101 February 1999 Part 21 Rept Re Possible Matl Defect in Swagelok Pipe Fitting Tee,Part Number SS-6-T.Defect Is Crack in Center of Forging.Analysis of Part Is Continuing & Further Details Will Be Provided IAW Ncr Timetables.Drawing of Part,Encl ML18106B0561999-01-31031 January 1999 Monthly Operating Rept for Jan 1999 for Salem Generating Station,Unit 2.With 990212 Ltr ML18106B0571999-01-31031 January 1999 Monthly Operating Rept for Jan 1999 for Salem Generating Station,Unit 1.With 990212 Ltr ML20205P1671999-01-31031 January 1999 a POST-PLUME Phase, Federal Participation Exercise ML18106B0441999-01-29029 January 1999 Part 21 Rept Re Possible Defect in Swagelok Pipe Fitting Tee Part Number SS-6-T.Caused by Crack in Center of Forging. Continuing Analysis of Part & Will Provide Details in Acoordance with NRC Timetables ML18106B0491999-01-28028 January 1999 LER 98-007-01:on 980730,reactor Coolant Instrument Line through-wall Leak Was Noted.Caused by Transgranular Stress Corrosion Cracking.Replaced Affected Tubing.With 990128 Ltr ML18106B0401999-01-18018 January 1999 LER 98-016-00:on 981219,ECCS Leakage Was Outside of Design Value.Caused by Leakage Past Seat of 21RH34 Manual Drain. Valve 21RH34 Was Reseated.With 990118 Ltr ML18106B0251998-12-31031 December 1998 Monthly Operating Rept for Dec 1998 for Salem Unit 2.With 990115 Ltr 1999-09-30
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PS~G *
- Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038-0236 Nucle'ar Business Unit AUG 13 1998 LR-N980397 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:
LER 311/98-010-00 SALEM GENERATING STATION - UNIT 2 FACILITY OPERATING LICENSE NO. DPR-75 DOCKET NO. 50-311 This Licensee Event Report entitled "ECCS Leakage Outside Design Basis Value" is being submitted pursuant to the requirements of the Code of Federal Regulations
- 1 OCFR50.73 (a)(2)(ii)****.
Sincerely, A. C. Bakken Ill General Manager Salem Operations Attachment PJD/
c Distribution LER File 3.7 9808210157 980813 PDR ADOCK 05000311
- S PDR The power is in your hands.
95-2168 REV. 6/94
NRC FORM 366 U.S. NUCLEAR REGULATOR MISSION APPROVED BY 0. 3150-0104 EXPIRES 06/30/2001 (6-1998) Estimated burden p ponse to comply with this mandatory information collection request: 50 hrs. Reported lessons learned are incorporated into LICENSEE EVENT REPORT (LER) the licensing process and fed back to industry. Forward comments regarding burden estimate to the Records Management Branch (T-6 F33), U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001, and to the (See reverse for required number of Paperwork Reduction Project (3150-0104), Office of Management and digits/characters for each block) Budget, Washington, DC 20503. If an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
FACILITY NAME (1)
SALEM UNIT 2 DOCKET NUMBER (2) 05000311 !PAGET OF 4 TITLE(4)
ECCS LEAKAGE OUTSIDE DESIGN BASIS VALUE.
EVENT DATE (5) LER NUMBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED (8)
MONTH DAY YEAR YEAR SEQUENTIAL REVISION MONTH DAY YEAR FACILITY NAME DOCKET NUMBER NUMBER NUMBER 07 14 98 98 -010 00 08 13 98 FACILITY NAME DOCKET NUMBER Salem Unit 1 05000272 OPERATING 1 100% x NAME TELEPHONE NUMBER (Include Area Code)
Philip J. Duca Jr. , Salem Licensing Engineer (609) 339-2381 COMPLETE AILURE DESC BED IN THIS REPORT (13)
CAUSE SYSTEM COMPONE MANUFACTURER REPORTABLE CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE NT TO EPIX TO EPIX EXPECTED MONTH DAY YEAR YES (If yes, complete EXPECTED SUBMISSION DATE).
ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16)
On 7 /14/98 Operations personnel determined that leakage f.rom the Boron Injection Tank exceeded the maximum allowable ECCS leakage from sources outside Containment.
The leakage path was determined to be past the valve seat of a manual sample valve (2SJ404) from the Boron Injection Tank. The valve was subsequently isolated. With the valve isolated, ECCS leakage returned to within UFSAR limits.
The cause of the event was the leaking 2SJ404 manual sample valve.
During the period of leakage, there were no actual safety consequences to the health and safety of the public or to the plant staff. Estimated potential post accident offsite doses remain well within 10CFR100 limits. Estimated post accident onsite beta skin dose is within the General Design Criterion 19 (GDC 19) limit, while estimated whole body and thyroid doses are slightly above the GDC 19 limits based on conservative estimates. Estimated core damage frequency during the period of the leakage, at approximately 3.7xE-6, decreases potential significance.
This event is reportable pursuant to 10CFR50.73(a)2(ii) "any event or.
condi tion ............... that resulted in the nuclear power plant being:...... (B) in a condition that was outside the design basis of the plant".
. NRC FORM 366A U.S. NUCLEAR REG ATORY COMMISSION (6-1998)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1) DOCKET (2) LER NUMBER (6) PAGE (3)
NUMBER(2)
SALEM UNIT 2 05000311 YEAR I SEQUENTIAL NUMBER IREVISION NUMBER 2 0F 4 98 0 1 0 00 TEXT (If more space is required, use additional copies ofNRC Form 366A) (17)
PLANT AND SYSTEM* IDENTIFICATION Westinghouse - Pressurized Water Reactor High-Pressure Safety Injection System/Sample Valve {BQ/SMV}*
- Energy Industry ~dentification System {EIIS} codes and component function identifier codes appear as (SS/CCC)
CONDITIONS PRIOR TO OCCURRENCE At the time of identification, Salem Unit 2 was operating at 100% Power.
DESCRIPTION OF OCCURRENCE On 7/14/98 Operations personnel determined that leakage from the Boron Injection Tank exceeded the maximum allowable ECCS Leakage as defined by UFSAR Section 6.3.2.11. The leakage was determined to be approximately 0.25 gpm (60,000cc/hour) past the valve seat of a manual sample valve (2SJ404) from the Boron Injection Tank. The leakage was flowing into the RHR sump in the Auxiliary Building, outside of Containment. UFSAR Section 6.3.2.11 addresses leakage during the recirculation phase of an accident, and states that "The total leakage resulting from all sources is about 3800 cc/hour as described in UFSAR Section 15.4.1. Recirculation loop leakage sources are summarized in Table 6.3-12. Leakage is monitored by procedure to ensure this leak rate is not exceeded." Leakage in excess of 3800 cc/hour placed the plant in a condition outside that assumed for design basis accident conditions.
For about two weeks prior to the event, efforts were being focused on reducing an elevated reactor coolant system (RCS) unidentified leakage.
Initially this leakage was thought to be within containment. The continuing investigations included determining the source of increasing in-leakage to the RHR sump located in the Auxiliary Building.
These investigations used station procedures. One of these procedures provides a program for monitoring leakage from systems outside Containment that could contain highly radioactive fluids following an accident. The objective of the program is to detect and correct any degradation of the pressure boundaries for these systems and thereby reduce post-accident dose rates and airborne activity.in the Auxiliary Building in accordance with Technical Specification 6.8.4.a.
NRC FORM 366A (6-1998)
I~
- ~========~*==========~*~====~
NRC FORM 366A (6-1998)
U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1) DOCKET (2) LER NUMBER (6) PAGE (3)
NUMBER(2)
SALEM UNIT 2 05000311 YEAR ISEQUENTIAL NUMBER IREVISION NUMBER 3 0F 4 98 0 1 0 00 TEXT (If more space is required, use additional copies ofNRCForm 366A) (17)
Following identification that 2SJ404 was the source of the leakage, 2SJ8 was closed, isolating 2SJ404. Valve 2SJ8 is the root valve in series with 2SJ404. Subsequent to closing 2SJ8, leakage into the RHR sump stopped. A follow-up RCS leak rate confirmed that the unidentified leakage was significantly reduced.
CAUSE OF OCCURRENCE The cause of the event was the leaking 2SJ404 manual sample valve.
PRIOR SIMILAR OCCURRENCES 1996, 1997 ano 1998 LERs were reviewed for similar occurrences. No similar events were identified.
SAFETY CONSEQUENCES AND IMPLICATIONS The activity level of the reactor coolant was slightly elevated due to a suspected single fuel pin leak. However, there was no appreciable change to radiological conditions in the area of the 2SJ404 due to the leakage through the valve. Routine effluent monitoring and dose assessment for the period showed the dose to be well within technical specification limits.
Therefore, during the period of leakage, there were no safety consequences to the health and safety of the public or to the plant staff.
In an effort to assess the potential safety consequences, dose estimates were conducted. The estimates were performed using the actual measured leakage through 2SJ404 (0.25 gpm). These estimates showed offsite whole body accident dose to be comparable with the plant design basis for both the Exclusion Area Boundary (EAB) and the Low Population Zone (LPZ) and well within 10CFR Part 100 limits. Offsite accident thyroid doses for the EAB and LPZ while greater than the design basis, were well within the Part 100 limits.
Potential onsite beta skin dose was estimated to remain within the General Design Criterion (GDC) 19 limit. However, whole body and thyroid doses, assuming a source term (the Salem design basis accident source) per the guidance of Regulatory Guide 1.4, are estimated to be slightly above the GDC 19 limits.
The potential consequences of the event depend on the core damage frequency, which for Salem is 4.45xE-5/year. While it is impossible to exactly
NRC FORM 366A (6-1998)
NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (6-1998)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1) DOCKET (2) LER NUMBER (6) PAGE (3)
NUMBER(2)
SALEM UNIT 2 05000311 YEAR I SE~~ 1=~~ 4 OF 4 98 0 1 0 00 TEXT (If more space is required, use additional copies ofNRC Form 366A) (17) identify a time when the 2SJ404 leakage became a significant component of the RCS unidentified leakage, formal troubleshooting to identify the leakage began two weeks prior to identification of *the leak. If this period is increased to a month, the core damage frequency for the period is approximately 3.7xE-6.
CORRECTIVE ACTIONS
- 1. The leakage flow path was isolated restoring ECCS leakage from systems outside Containment to within UFSAR limits.
- 2. A work order has been initiated to inspect and repair the 2SJ404 valve.
- 3. Operator training will be provided on this event during an upcoming requalification cycle. This training will include discussion of the design basis leakage criteria for primary coolant sources outside containment. 1
- 4. The program to reduce leakage from primary coolant sources outside containment (required by Technical Specification 6.8.4.a) will be reviewed and appropriate revisions will be made~
NRC FORM 366A (6-1998)