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Category:LICENSEE EVENT REPORT (SEE ALSO AO RO)
MONTHYEAR05000312/LER-1986-016, Errata to LER 86-016-02:on 861003,decay Heat Sys Pump B Tripped Due to Abnormal Indication on Panel H2SFB Causing Bistable B Trip & HV-20002 to Close.Caused by Arc in Sump Level Indicator.Indicators Will Be Repainted1987-11-18018 November 1987 Errata to LER 86-016-02:on 861003,decay Heat Sys Pump B Tripped Due to Abnormal Indication on Panel H2SFB Causing Bistable B Trip & HV-20002 to Close.Caused by Arc in Sump Level Indicator.Indicators Will Be Repainted 05000312/LER-1987-036, Errata to LER 87-036-01:on 870608,discovered That Pipe Supplying Water for Bearing Cooling for Reactor Bldg Spray Pump Blocked.Preventive Maint Procedure for Pump Bearing Cooling Water Supply & Return Lines Instituted1987-11-18018 November 1987 Errata to LER 87-036-01:on 870608,discovered That Pipe Supplying Water for Bearing Cooling for Reactor Bldg Spray Pump Blocked.Preventive Maint Procedure for Pump Bearing Cooling Water Supply & Return Lines Instituted 05000312/LER-1987-035, Corrected Page 1 to LER 87-035-00:on 870625,during Startup Test Procedure 199 Continuous Fire Watch Abandoned from Monitoring Alarm Panel H4FCP5 for 3 H.Caused by Uncontrolled Discharges of Carbon Dioxide Exposing Personnel to1987-07-27027 July 1987 Corrected Page 1 to LER 87-035-00:on 870625,during Startup Test Procedure 199 Continuous Fire Watch Abandoned from Monitoring Alarm Panel H4FCP5 for 3 H.Caused by Uncontrolled Discharges of Carbon Dioxide Exposing Personnel to Risk 05000312/LER-1985-016, Corrected Pages of Rev 1 to LER 85-016-011986-12-31031 December 1986 Corrected Pages of Rev 1 to LER 85-016-01 ML20147B3341978-07-10010 July 1978 /03L-1 on 780710:Maintenance on Safety Sys Hydraulic Snubber Previously Found to Have Low Fluid Accumulator Level Revealed 2 Deteriorated o-rings Made W Polyurethene Seals Instead of ethylene-propylene Seals 1987-07-27
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20211H7921999-08-13013 August 1999 Safety Evaluation Supporting Amend 126 to License DPR-54 ML20195D1901999-05-0606 May 1999 Annual Rept ML20195H8571998-12-31031 December 1998 1998 Annual Rept for Smud. with ML20155D4801998-10-27027 October 1998 Amend 3 to Rancho Seco DSAR, Representing Updated Licensing Basis for Operation of Permanently Shutdown & Defueled Rancho Seco Nuclear Facility During Permanently Defueled Mode ML20248C4301998-05-0606 May 1998 Annual Rept, Covering Period 970507-980506 ML20249A7831997-12-31031 December 1997 1997 Smud Annual Rept ML20198G8271997-08-22022 August 1997 Safety Evaluation Supporting Amend 125 to License DPR-54 ML20217D3271997-07-30030 July 1997 Update of 1995 Decommissioning Evaluation for Rancho Seco Nuclear Generating Station ML20140A6371997-05-0606 May 1997 Annual Rept, Covering Period 960507-970506 ML20140G4481997-05-0101 May 1997 Part 21 Rept Re Potential Defect in Component of Dsrv & Dsr Enterprise Standby Diesel Generator Sys.Recommends That Springs Be Inspected on Periodic Basis,Such as During Refueling Outages ML20137W8151997-03-20020 March 1997 Amend 1 to Post Shutdown Decommissioning Activities Rept ML20141J2711996-12-31031 December 1996 Smud 1996 Annual Rept ML20138L1231996-11-13013 November 1996 Smud Rancho Seco Incremental Decommissioning Action Plan, Rev 0,961113 ML20129E7151996-10-14014 October 1996 Defueled SAR for Rancho Seco ML20059H6821994-01-17017 January 1994 Revised Rancho Seco Quality Manual ML20058K3841993-12-0909 December 1993 Part 21 Rept Re Potential Defect in Component of Dsrv & Dsr Enterprise Standby DG Sys,Regarding Potential Problem W/ Subcover Assembled Atop Power Head ML20056E5171993-08-31031 August 1993 Technical Review Rept, Tardy Licensee Actions ML20059K1981993-05-0606 May 1993 Annual Rept, Covering Period from 920501- 930506,consisting of Shutdown Statistics,Narrative Summary of Shutdown Experience & Tabulations of Facility Changes, Tests & Experiments,Per 10CFR50.59(b) ML20128C9641993-02-0202 February 1993 Informs Commission of Status of Open Issues & Progress of Specified Facilities Toward Decommissioning ML20127H2301993-01-15015 January 1993 Part 21 Rept Re Potential Defeat in Component of Dsrv & Dsr Enterprise Standby DG Sys.Starting Air Distributor Housing Assemblies Installed as Replacement Parts at Listed Sites ML20126B0421992-12-17017 December 1992 Final Part 21 Rept Re Potential Problem W/Steel Cylinder Heads.Initially Reported on 921125.Caused by Inadequate Cast Wall Thickness at 3/4-inch-10 Bolt Hole.Stud at Location Indicated on Encl Sketch Should Be Removed ML20125C7161992-12-0707 December 1992 Part 21 Rept Re Possibility for Malfunction of Declutching Mechanisms in SMB/SB-000 & SMB/SB/SBD-00 Actuators. Malfunction Only Occurs During Seismic Event.Balanced Levers May Be Purchased from Vendor.List of Affected Utils Encl ML20127P5861992-11-23023 November 1992 Followup to 921005 Part 21 Rept Re Potential Defect in SB/SBD-1 Housing Cover Screws.Procedure Re Replacement of SBD-1 Spring Cover Bolts Encl.All Fasteners Should Be Loosened & Removed.List of Affected Utils Encl ML20126E6771992-08-0303 August 1992 Rev 7 to Rancho Seco Quality Manual NL-90-451, Monthly Operating Rept for Oct 1990 for Rancho Seco Nuclear Generating Station1990-10-31031 October 1990 Monthly Operating Rept for Oct 1990 for Rancho Seco Nuclear Generating Station ML17348B5061990-10-0909 October 1990 Part 21 Rept Re Zener Diode VR2 on Power Supply Board 9 1682 00 106 Possibly Being Installed Backwards ML20059G9621990-09-10010 September 1990 Safety Evaluation Supporting Amend 115 to License DPR-54 NL-90-443, Monthly Operating Rept for Aug 1990 for Rancho Seco1990-08-31031 August 1990 Monthly Operating Rept for Aug 1990 for Rancho Seco ML20217A5711990-08-28028 August 1990 Final Engineering Rept,Assessment of Spent Fuel Pool Liner Leakage NL-90-439, Monthly Operating Rept for Jul 1990 for Rancho Seco Nuclear Generating Station1990-07-31031 July 1990 Monthly Operating Rept for Jul 1990 for Rancho Seco Nuclear Generating Station ML20055F8591990-07-16016 July 1990 Special Rept 90-11:on 900613,06,25,18,21 & 28,fire Barriers Breached More than 7 Days & Not Made Operable in 14 Days. Corrective Actions:Operability of Fire Detectors Verified on One Side of Breached Barriers NL-90-423, Monthly Operating Rept for June 1990 for Rancho Seco Nuclear Generating Station1990-06-30030 June 1990 Monthly Operating Rept for June 1990 for Rancho Seco Nuclear Generating Station ML20055C6291990-05-21021 May 1990 Special Rept 90-08:on 900419,fire Pump Batteries Inoperable When Surveillance Procedure SP.206 Not Performed by Due Date.Caused by Test Frequency Incorrectly Changed from Weekly to Monthly.Surveillance Schedule Revised ML20055C6301990-05-21021 May 1990 Special Rept 90-09:on 900424,25,30,31 & 0502,fire Barriers Inoperable for More than 7 Days,Per Tech Spec 3.14.6.2 Requirement.Hourly Fire Watches Established & Penetrations & Doors Returned to Operable Status ML20058B6521990-05-0404 May 1990 Rev 0 to ERPT-M0216, Property Loss Study for Rancho Seco Nuclear Generating Station in Long Term Defueled Mode ML20248E0121989-09-13013 September 1989 Supplemental Part 21 Rept Re Potential Problem W/Six Specific Engine Control Devices in Air Start,Lube Oil, Jacket Water & Crankcase Sys.Initially Reported on 890429. California Controls Co Will Redesign Valve Seating NL-89-634, Monthly Operating Rept for Aug 1989 for Rancho Seco Nuclear Generating Station1989-08-31031 August 1989 Monthly Operating Rept for Aug 1989 for Rancho Seco Nuclear Generating Station NL-89-598, Monthly Operating Rept for Jul 1989 for Rancho Seco Nuclear Generating Station1989-07-31031 July 1989 Monthly Operating Rept for Jul 1989 for Rancho Seco Nuclear Generating Station ML20247B3611989-07-17017 July 1989 Safety Evaluation Supporting Amend 112 to License DPR-54 NL-89-556, Monthly Operating Rept for June 1989 for Rancho Seco Nuclear Station1989-06-30030 June 1989 Monthly Operating Rept for June 1989 for Rancho Seco Nuclear Station ML20245E1161989-06-20020 June 1989 Safety Evaluation Supporting Amend 111 to License DPR-54 ML20245B6651989-06-15015 June 1989 Part 21 Rept 150 Re Potential Defect in Component of Dsr Standby Diesel Generator.Cause of Failure Determined to Be Combination of Insufficient Lubrication to Bushings.Listed Course of Action Recommended at Next Scheduled Engine Maint ML20245A1991989-06-0909 June 1989 Safety Evaluation Supporting Amend 110 to License DPR-54 ML20248B9361989-06-0505 June 1989 Safety Evaluation Supporting Amend 107 to License DPR-54 ML20248B6321989-06-0505 June 1989 Safety Evaluation Supporting Amend 108 to License DPR-54 ML20248B9621989-06-0505 June 1989 Safety Evaluation Supporting Amend 109 to License DPR-54 NL-89-517, Monthly Operating Rept for May 1989 for Rancho Seco Nuclear Generating Station1989-05-31031 May 1989 Monthly Operating Rept for May 1989 for Rancho Seco Nuclear Generating Station ML20247P1761989-05-30030 May 1989 Safety Evaluation Accepting Generic Ltr 83-28,Item 4.5.2 Re on-line Testing of Reactor Trip Sys ML20247N7491989-05-30030 May 1989 Special Rept 89-13:on 890328-29,specific Activity of Primary Sys Exceeded Limits in Administrative Procedure & Chemistry Control Commitment.Possibly Caused by Power Reduction.Isotopic Analysis Continued ML20247K7261989-05-23023 May 1989 Safety Evaluation Supporting Amend 105 to License DPR-54 1999-08-13
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d.) 1 feltC Perm 3BSA U.S. NUCLEAA RETULATORY COMMIS$ ION LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Apenovso oms No aisomo.
EXPIRES: 8/31/88 PACILITV NAast H) DOCKET NUMBER (2)
LER NUMBER (6) PAGE (3)
RANCHO SEC0 NUCLEAR GENERATING STATION vraa "M.7,7 ' O *32 -
UNIT NO. 1 serr_
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. ancre wnn On August 8 and August 14, 1985, while in cold shutdown, the Decay Heat Removal System (DHS) suction block valve (HV-20002) automatically closed on a high Reactor Coolant System (RCS) pressure signal, thus resulting in a temporary loss of the DHS system capability. In both cases, DHS flow was re-established in eleven minutes or less, and no noticeable increases in the incore temperatures were detected.
HV-20002 is a motor operated valve located in the decay heat removal drop line which is closed during plant operation and must be opened to initiate decay heat removal operation. HV-20002 is interlocked with Core Flooding Tank isolation valve HV-26514 and with RCS pressure transmitter PT-21099 to help provide decay heat removal pump suction header over-pressure protection in conjunction with the DHS relief valves. The DHS design pressure is 300 psig. RCS transmitter PT-21099, located off Reactor Coolant System (RCS) "B" hot leg, provides a RCS pressure signal. A cable exists between the RCS transmitter PT-21099 and SFAS cabinet H4SAB2. DHS utilizes components within cabinet H4SAB2 such as the power supply unit, buffers, and the test modules.
These components do not affect / actuate the SFAS system. Automatic closure of HV-20002 occurs when the RCS pressure exceeds 255 psig. The RCS pressure recorded by operations personnel at the time of the events was approximately 230 psig. PT-21099 was replaced and calibrated during the Cycle 7 Refueling Outage and a successful maintenance test was performed following the events to ensure the proper operability of the decay heat valve interlock and associated instrumentation. An on-going Root Cause Investigation notes that several transmitters have served in this capacity. Upgrades have been done with respect to the Environmental Qualification Program. The more recently used Rosemont transmitters appear to be more sensitive to this voltage spiking than previous transmitters used.
There were three similar trips of the decay heat removal system, each following spurious closures of HV-20002. Alternate decay heat removal capability was available via the RCP-D/0TSG, which was operating concurrently with the DHS, during the December 29th incident. The second incident, at 1618 hours0.0187 days <br />0.449 hours <br />0.00268 weeks <br />6.15649e-4 months <br /> on December 30, 1985, followed the shutdown of RCP 1 The incident was mitigated by resetting the valve closure circuit, reopening the drop line valves, opening the MOV breaker, and restarting the DHS "A" at 1624 hours0.0188 days <br />0.451 hours <br />0.00269 weeks <br />6.17932e-4 months <br /> the same day. The third incident on December 31 was mitigated by restarting RCP-D/0TSG cooling, and re-establishing DHS "A" operation; both within 30 minutes of the trip. Due to the presence of alternate decay heat removal capability (RCP/0TSG), these three trips were determined not reportable in accordance with 10 CFR Part 50.73.
l 8701130281 861231
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m-LICENSEE EVENT REPORT (LER) TEXT CONTINUATION i.PeRovEo oMe mo. uso-oio4 EXPtRES: 8/31/08 FAC1UTY NAast (1) DOCKET NUMBER (2) LER NUM8ER (6) FAGE (3)
RANCHO SEC0 NUCLEAR GENERATING STATION yp. .
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0l1 0l 3 OF Ql4 wira-. - . = -. ee a-wnn The spurious decay heat isolation signal was traced to a shielded instrument cable improperly routed (cable number 1R1S04B6A) through Channel B power trays and conduit to a penetration, as documented in NCR S-5263, Revision 3. This instrument cable was originally routed at plant startup. The voltage spike on the PT-21099 instrument circuit appeared, in a test, when the power circuit to Motor Operated Valve HV-20002 was energized. The spikes caused the decay heat "B" interlock to actuate. The existing instrument cable was abandoned in place. A new instrument cable was routed in instrumentation cable raceways since this event. Testing of the new instrument loop will be completed by December 1,1986, once the Safety Features Panel involved is available.
Further investigation of the power cable tray involved revealed that 47 additional power cables were also routed with the HV-20002 power cable.
Any one of these power cables could, and most likely did, cause a voltage spike similar to that found during testing of HV-20002. The problem was corrected by moving the PT-21099 circuit out of the power circuits raceway. The District also checked the routing of Pressure Transmitter PT-21092 circuitry which provides the high pressure isolation signal for HV-20001, and found it in a similar configuration to PT-21099.
This routing will be corrected prior to plant restart.
The NEPM construction specifications and corresponding inspection documents were issued subsequent to the occurrence of these problems so that adherence to these procedures should prevent recurrence.
- The CRTS (Cable Routing Tracking System) is a computer program and set of administrative controls that has tracked electrical cable configuration since its installation in July 1980. The cable in question was installed prior to CRTS' use. The CRTS program has the ability to discern intermixing among Quality Class 1 Channels A, B, C, and D cables and prevent it, as well to preclude the installation of instrument cables within power cable raceways for Quality Class 2 and 3 cables.
The CRTS staff is composed of a full time qualified engineer who acts as an administrator; an engineering aid; and a data entry operator. Four additional engineers were added to the CRTS staff in order to execute a comprehensive program at Rancho Seco to rid the system data base of its discrepancies and deviations. The comprehensive program is being undertaken with continuing overview by the Rancho Seco Quality Department.
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- U S GPO 1986 0624 638,465
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, LICENSEE EVENT REPORT (LER) TEXT CONTINUATION muovEO ous No. 3 so-oto.
EXPIRES: 8/31/8B FACILify NAasE (1) DOCKET NUMBER (2)
LER NUM8ER 16) PAGE(3)
RANCHO SEC0 NUCLEAR GENERATING STATION vpa 55=m -
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. .a .,.w.v. .ms wnn The comprehensive CRTS correction program includes a rigorous statistical survey of plant electrical equipment, raceways, terminations, and cables in order to ensure that the CRTS computer model and its reports accurately reflect the plant's electrical configuration. The correction program schedule has eight parts. The first six items are to be completed prior to plant Restart. The remaining items are to be complete before the conclusion of the Cycle 8 Refueling Outage.
- 1. Improve Efficiency of CRTS Program Operations.
- 2. Resolve Identified CRTS Problems.
- 3. Perform Engineering Analyses of Problems.
- 4. Perform Walkdowns of the CRTS Problems that Require Field Verification.
- 5. Complete the CRTS Data Base Verification after Data Base Update.
- 6. Document and Revise Where Needed the Existing Administrative Control Mechanism by Which Engineers, Designers, and Constructors Use CRTS.
- 7. Establish Annual Software Verification Drogram for CRTS with Rancho Seco Quality Department.
- 8. Establish Improved CRTS Electrical Configuration (NEP 4109) Configuration Control Procedures.
During the above program those cable configurations found to be in an unacceptable arrangement with respect to Nuclear Safety will be reconfigured as necessary to achieve conformance prior to plant Restart.
These actions will enhance the Cable Raceway Tracking System and rid it of its discrepancies and deviations, making it the accurate and precise engineering and construction tool it was designed to be. CRTS, thereby, will continue to prevent the installation of instrument cables into control cable or power cable raceways. Furthermore, should any other original plant CRTS occurrences of similar nature exist, these will have been detected and corrected.
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