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Category:REPORTABLE OCCURRENCE REPORT (SEE ALSO AO LER)
MONTHYEARML20217G7821997-07-22022 July 1997 Special Rept:On 970227,ECCS Sys Was Actuated & Injected Water in Rcs.Declared HPCI Subsystem Operable Following Restoration of Subsystem to Normal Standby Lineup ML20097D4661992-05-27027 May 1992 Special Rept,Reflecting Quad-Cities Nuclear Power Station, Unit 2 Summary Status of Fuel Performance at End of Cycle 11. No Sipping Performed on Reload Fuel Assemblies at End of Cycle 11 ML20066J6331991-01-31031 January 1991 Forwards Summary Status of Fuel Performance for End of Cycle 11 ML20207P1881987-01-0606 January 1987 Special Rept:Summarizes Status of Fuel Performance as of End of Cycle 8.Of 2031 Assemblies used,289 Determined to Have Failed & Discharged as Leaker Assemblies.Aug 1986 Sipping Tests Showed No Indication of Assembly Failure ML20082P6541983-12-0505 December 1983 Telecopy RO 50-83/20-01T:on 831202,review of Ultrasonic Test Data Obtained During Current Refuel Outage Revealed Circumferential Linear Indication in Heat Affected Zone of Weld 02B-S9.Further Info Expected within 14 Days ML20078J8531983-10-12012 October 1983 Telecopy Ro:On 831011,during Performance of Ultrasonic Insp of Cleanup Sys Piping,Crack Indication Discovered in Heat Affected Zone of Weld 12S-S27 of Line 2-1202-6A.Further Info to Be Provided within 14 Days ML20084S1401977-05-13013 May 1977 Ro:On 770512,unit Experienced Sudden Turbine Control Valve Opening Resulting in Feedwater Flow Increase & Reactor Scram from APRM flux.Short-term Reactivity Increase Corresponding to Reactor Period of Less than 5s Experienced ML20084S1531977-05-0909 May 1977 Ro:On 770507,station Experienced Momentary Short Period of Less than 5s During Shutdown Margin Demonstration.Similar Experience Occurred on 770504.Addl Info Will Be Submitted in RO 50-254/77-23 ML20084S1901977-02-17017 February 1977 Telecopy Ro:On 770217,info Received by Station Indicated That MAPLHGR Limit Curves Given by Tech Specs Shall Be Additionally Lowered.Reduction Resulted from Review of ECCS Analysis.Maplhgr Curves Lowered Effective 770217 ML20084S2041977-01-18018 January 1977 Telecopy Ro:On 770118,info Received by Station Indicated That MAPLHGR Limit Curves Shall Be Lowered.Reduction Resulted from Review of ECCS Analysis.Curves Lowered ML20084S3501977-01-0404 January 1977 RO-50-254/76-37:on 761205,limit for Having No More than 2 Ci Activity in Radwaste Tank Farm in 24h Exceeded. Caused by Deterioration of Effectiveness of Radwaste Sys to Perform as Designed.Radwaste Sys Being Updated ML20084S3551976-12-30030 December 1976 RO 50-254/76-36:on 761202,cooling Water Suction Header Common to Both RHR Containment Cooling Loop 1A & Diesel Generator 1 Cooling Water Pump Airlocked.Caused by Procedure Inadequacy.Maint Procedures Will Be Revised ML20084S2861976-12-28028 December 1976 RO 50-254/76-38:on 761215,GE Notified Util That NRC Discovered Errors in Code Inputs to ECCS App K Analysis, Caused by GE Incorrectly Applying Data for ECCS App K Analysis.Maplhgr Curves Reduced by 4% ML20084S3391976-12-16016 December 1976 Ro:On 761215,preliminary Info Received by Station Indicated That MAPLHGR Curves Should Be Lowered by 4%.Change Resulted from Review of ECCS Analysis.Curves Lowered ML20084S3611976-12-16016 December 1976 RO 50-254/76-35:on 761203,discovered Discrepancy Between Tech Spec & Nedo 20360 Rod Worth Minimizer Operable Rated Power.Orders Written in Daily Order Book Requiring Rod Worth Minimizer Operability Below 20% ML20084S3701976-12-0303 December 1976 Ro:On 761203,info Received Indicated That Rod Worth Minimizer (RWM) Operability Requirements Should Be Changed from 10% to Below 20% Rated Power.Caused by Review of Rod Drop Accident Analysis ML20084S3791976-11-22022 November 1976 Updated RO 50-254/76-33:on 761029,repairs to RCIC Pump Consisted of Rebuilding Pump Casing & Installing New Rotary Element.Pump Reassembled & Repairs Completed by 761111.Welds on Piping Acceptable ML20084S3771976-11-12012 November 1976 RO 50-254.76-34-on 761104,electromatic Relief Valves 1-203-3C & 1-203-3E Failed to Open When Actuated from Control Room.Caused by Excessive Steam Leakage Into Area Below Valve Disc.Investigation in Progress W/Manufacturer ML20084S3941976-11-10010 November 1976 RO 50-254/76-33:on 761029,while Performing RCIC Sys Pump Operability Surveillance,Discovered That Pump Could Not Achieve Flow & Pressure Required.Caused by Two of Five Stages Being Severely Damaged.Pump Being Rebuilt ML20084S5421976-10-30030 October 1976 Updated RO 50-254/76-26:on 760803,rod Worth Minimizer Not Operable for Withdrawal of First Twelve Control Rods to Fully Withdrawn Position While in Startup Mode.Caused by Burned Out Wire Runs on Relay Board.Software Modified ML20084S4031976-10-13013 October 1976 RO 50-254/76-32:on 761001,main Chimney Monitoring Sys & Reactor Bldg Vent Sample Sys Not Functioning Properly.On 760919,flexible Sample Hose Wrapped W/Tape to Stop Possible Leak.Caused by Failure of Flexible Suction Hose ML20084S5641976-10-0404 October 1976 Supplemental RO 50-254/76-25:electrical Nitrogen Vaporizers Installed on Nitrogen Makeup Sys.Installation Should Prevent Future Recurrences ML20084S4761976-10-0404 October 1976 RO 50-254/76-31:on 760920,station Informed by GE of Error in Reload 2 Licensing Submittal in Determining Max Change in Critical Power Ratio Due to Abnormal Operating Transient. Caused by Incorrect Analysis in Preparing Submittal ML20084S4831976-09-21021 September 1976 RO Re Notification by GE of Oversight in Reload 2 Licensing Submittal Leading to Possible Nonconservative Operation During Cycle 3.Work Request to Lower Rod Block Monitor Line to 10% at Full Flow Initiated ML20084S5031976-09-0909 September 1976 RO 50-254/76-29:on 760809,surveillance of Primary Containment Oxygen Concentration Revealed Increase in Concentration from 4.2 to 4.8%.Caused by Instrument Drift. Oxygen Analyzer Recalibr ML20084S5151976-09-0707 September 1976 RO 50-254/76-28:on 760809,position Indication Lost on Reactor Water Cleanup Sys Isolation Valve Mo 1-1201-2.Caused by Relay 595-125 Shorting Out & Burning Up Control Transformer.Relay & Transformer Replaced ML20084S5781976-08-26026 August 1976 RO 50-254/76-24:on 760727,reactor Bldg to Suppression Chamber Vacuum Breaker Pressure Switch PS-1-1622B Tripped at 0.536 Psid.Caused by Instrument Setpoint Drift.Switch Recalibr ML20084S5241976-08-19019 August 1976 RO 50-254/76-27:on 760806,ECCS Analysis Performed by GE for BWR-3 Type Plant Resulted in Calculated Peak Clad Temp Greater than 2,200 F for Reduced Core Flows.Caused by Severity of Conservatisms Associated W/Using App K ML20084S5511976-08-16016 August 1976 RO 50-254/76-26:on 760803,rod Worth Minimizer Multiple Output Distributor Error Detected.Caused by Burned Out Wire Runs on Relay Board.Board Repaired & Hardware & Software Mods & Procedural Changes Being Considered ML20084P7411976-06-23023 June 1976 Telecopy Ro:Initial Swipe Survey of Nuclear Fuel Svcs NFS-4 Cask Indicated Three of 38 Tests Exceeded Limits ML20084P7551976-06-17017 June 1976 Telecopy Ro:On 760617,HPCI Sys Motor Speed Changer Failed to Come Off Low Speed Stop During Monthly Surveillance.Caused by Motor Speed Changer Linkage Being Bound.Linkage Freed ML20084S4131976-03-30030 March 1976 Telecopy Ro:On 760330,suppression Chamber Water Level Instrumentation Found Miscalibrated.Suppression Chamber Water Level Immediately Returned to Normal.Addl Info Will Be Submitted in RO 50-265/76-04 within 14 Days ML20084Q1751976-03-26026 March 1976 Telecopy Ro:On 760326,chemical Waste Sample Tank Discharged to River at Rate in Excess of Limits ML20084S4151976-01-0909 January 1976 Telecopy Ro:On 760108,while in Cold Shutdown,Crack Indications Found on a & B Loops in valve-to-pipe Junctions on Sides of Bypass Valve That Cannot Be Isolated & in Heated Zones ML20084S5091976-01-0505 January 1976 RO 50-254/76-1:on 760105,w/unit in Cold Shutdown for Refueling,Pinhole Leak Found in Fillet Weld of 3/4 Inch Drain Line.Caused by Degradation of Reactor Coolant Primary Boundary.Work Request Being Issued ML20084S5311975-12-31031 December 1975 RO 50-265/75-47:on 751231,w/unit Operating at 805 Mwe, Reactor Core Isolation Cooling (RCIC) Trip Throttle Valve Could Not Be Reset Following Successful Monthly Surveillance Functional Testing.Hpci Tested & Found Operable ML20084U2081975-02-25025 February 1975 Ro:On 750212,following Verification of Reactor core,mixed- Oxide Fuel Assembly Identification Number Stamped in Wrong Orientation Discovered.Caused by Mfg Error.Fuel Insp Procedure Changed to Verify Number Orientation as Correct ML20084U2121975-02-20020 February 1975 Ro:On 750212,during Core Spray Operational Hydrostatic Test, Water Observed Overflowing Reactor Bldg Floor Drain Sump 1B.Caused by Premature Actuation of Core Spray Discharge Header Relief Valves.Test Procedure Amended ML20085C9241974-11-15015 November 1974 Ro:On 741011,control Rod Drive N-11 Jammed Fully Inserted Past Position 00 & Would Not Withdraw After Increasing Drive Pressure.Caused by Broken Seals on Stop Piston.Drive N-11 Replaced ML20085C9181974-11-15015 November 1974 Ro:On 741017,during Removal of Faulty Intermediate Range Monitor 18 Detector,Detector Became Stuck in Shuttle Tube. Caused by Coupling Between Shuttle & Drive Tubes Overtightened.New Shuttle Tube Ordered ML20084K5181974-06-26026 June 1974 Telecopy RO Re Setpoint Drift of Standby Liquid Control a Relief Valve.Valve Reset & Tested ML20084K6451974-06-22022 June 1974 AO 50-265/74-12:on 740612,high Differential Pressure Noted Across Combined High Efficiency & Carbon Filters.Caused by Exhausted High Efficiency Prefilter.Original Carbon Filters, New Rough Prefilters & New Efficiency Prefilters Installed ML20084K5461974-06-19019 June 1974 Telecopy Ro:On 740619,water Leak Discovered at Pressure Test Connection & Feedwater Line.Caused by Cracked Weld at Weldolet Line.Repairs in Process ML20084K5541974-06-19019 June 1974 Telecopy Ro:On 740619,water Leak Discovered at Pressure Test Connection & Feedwater Line.Caused by Cracked Weld at Weldolet Line.Repairs in Process ML20084K5891974-06-14014 June 1974 Telecopy Ro:On 740613,level Switch LIS-2-263-72D Failed. Switch Lightly Pressed & Functioned Normally.New Switch Ordered ML20084K5941974-06-14014 June 1974 Telecopy Ro:On 740613,level Switch LIS-2-263-72D Failed. Switch Functioned Normally When Lightly Pressed.New Switch to Be Installed ML20084K6291974-06-13013 June 1974 Telecopy Ro:Level Switch Failed to Actuate Control Room Annunciator.Cause Under Investigation ML20084L0441974-06-10010 June 1974 Telecopy Ro:On 740609,high Water Conductivity Discovered in Reactor.Shutdown & Water Cleanup Initiated ML20084K8401974-06-10010 June 1974 Telecopy Ro:On 740610,feedwater Valve Failed.Caused by Wall Thickness Being Less than Design Min.Cleanup, Decontamination & Water Processing Initiated ML20084K8361974-06-10010 June 1974 Telecopy Ro:On 740610,feedwater Low Flow Valve Failed.Caused by Not Maintaining Min Wall Thickness.Radiation Released Not Above Normal Level.Decontamination,Cleanup & Water Processing Initiated 1997-07-22
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217A9931999-09-30030 September 1999 NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data SVP-99-204, Monthly Operating Repts for Sept 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With1999-09-30030 September 1999 Monthly Operating Repts for Sept 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With ML20217A1691999-09-22022 September 1999 Part 21 Rept Re Engine Sys,Inc Controllers,Manufactured Between Dec 1997 & May 1999,that May Have Questionable Soldering Workmanship.Caused by Inadequate Personnel Training.Sent Rept to All Nuclear Customers ML20212J0501999-09-21021 September 1999 Safety Evaluation Re Licensee Implementation Program to Resolve USI A-46 at Plant,Per GL 87-02,Suppl 1 SVP-99-179, Monthly Operating Repts for Aug 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With1999-08-31031 August 1999 Monthly Operating Repts for Aug 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With ML20210L8661999-08-0202 August 1999 Safety Evaluation Accepting License 60-day Response to GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Movs SVP-99-155, Monthly Operating Repts for July 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With1999-07-31031 July 1999 Monthly Operating Repts for July 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With SVP-99-148, Monthly Operating Repts for June 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With1999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With ML20196H8621999-06-30030 June 1999 NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data, June 1999 Rept ML20195K1481999-06-16016 June 1999 Safety Evaluation Authorizing Relief Request RV-23A for Duration of Current 10 Yr IST Interval on Basis That Compliance with Code Requirements Would Result in Hardship Without Compensating Increase in Level of Quality & Safety SVP-99-123, Monthly Operating Repts for May 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With1999-05-31031 May 1999 Monthly Operating Repts for May 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With ML20195B2591999-05-19019 May 1999 Rev 66a to CE-1-A,consisting of Proposed Changes to QAP for Dnps,Qcs,Znps,Lcs,Byron & Braidwood Stations SVP-99-104, Monthly Operating Repts for Apr 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With1999-04-30030 April 1999 Monthly Operating Repts for Apr 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With SVP-99-102, Summary Rept of Changes,Tests & Experiments Completed, Covering Period 990201-0430. with1999-04-30030 April 1999 Summary Rept of Changes,Tests & Experiments Completed, Covering Period 990201-0430. with ML20205Q5291999-04-16016 April 1999 SER Concluding That Quad Cities Nuclear Power Station,Unit 1,can Be Safely Operated for Next Fuel Cycle with Weld O2BS-F4 in Current Condition Because Structural Integrity of Weld Will Be Maintained ML20205J6011999-04-0707 April 1999 Safety Evaluation Accepting Proposed Merger of Calenergy Co, Inc & Midamerican Holdings Co for Quad Cities Nuclear Power Station,Units 1 & 2 SVP-99-071, Monthly Operating Repts for Mar 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With1999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With ML20205C5671999-03-19019 March 1999 Simulator Four-Yr Certification Rept ML20207D2341999-03-0101 March 1999 Post Outage (90 Day) Summary Rept, for ISI Exams & Repair/Replacement Activities Conducted 981207-1205 ML20204B1571999-02-28028 February 1999 Monthly Operating Repts for Feb 1999 for Quad Cities,Units 1 & 2.With SVP-99-021, Quarterly Summary SER of Changes,Tests & Experiments Completed, Covering Period of 981101-990131,IAW 10CFR50.59 & 10CFR50.71(e).With1999-01-31031 January 1999 Quarterly Summary SER of Changes,Tests & Experiments Completed, Covering Period of 981101-990131,IAW 10CFR50.59 & 10CFR50.71(e).With ML20205D1311998-12-31031 December 1998 1998 Decommissioning Funding Status Rept for Yr Ending 981231 for Quad Cities Nuclear Power Station,Units 1 & 2 ML20205M7061998-12-31031 December 1998 Unicom Corp 1998 Summary Annual Rept. with SVP-99-007, Monthly Operating Repts for Dec 1998 for Quad Cities Nuclear Power Station,Units 1 & 2,IAW GL 97-02 & TS 6.9.With1998-12-31031 December 1998 Monthly Operating Repts for Dec 1998 for Quad Cities Nuclear Power Station,Units 1 & 2,IAW GL 97-02 & TS 6.9.With ML20196C8391998-11-30030 November 1998 Rev 0 to GE-NE-B13-01980-030-2, Assessment of Crack Growth Rates Applicable to Induction Heating Stress Improvement (IHSI) Recirculation Piping in Quad Cities Unit 1 SVP-98-364, Monthly Operating Repts for Nov 1998 for Quad Cities Nuclear Power Station,Units 1 & 2.With1998-11-30030 November 1998 Monthly Operating Repts for Nov 1998 for Quad Cities Nuclear Power Station,Units 1 & 2.With ML20196G1241998-11-30030 November 1998 COLR for Quad Cities Unit 1 Cycle 16 ML20196D9651998-11-30030 November 1998 Safety Evaluation Supporting Relief Requests CR-21 & CR-24, Respectively.Relief Request CR-23,proposed Alternative May Be Authorized,Per 10CFR50.55a & Relief Request CR-22 Was Withdrawn by Licensee ML20196C8731998-11-30030 November 1998 Rev 0 to GE-NE-B13-01980-30-1, Fracture Mechanics Evaluation on Observed Indications at Two Welds in Recirculation Piping of Quad Cities,Unit 1 Station ML20196A9761998-11-20020 November 1998 Safety Evaluation Re Licensee 180-day Response to GL 95-07, Thermal Binding of Safety-Related Power-operated Gate Valves ML20196A4191998-11-19019 November 1998 Safety Evaluation Accepting QA TR CE-1-A,Rev 66 Re Changes in Independent & Onsite Review Organization by Creating NSRB SVP-98-346, Monthly Operating Repts for Oct 1998 for Quad Cities Nuclear Power Station,Units 1 & 2.With1998-10-31031 October 1998 Monthly Operating Repts for Oct 1998 for Quad Cities Nuclear Power Station,Units 1 & 2.With SVP-98-358, Summary Rept of Changes,Tests & Experiments Completed, Including SEs Covering Period on 980716-1031.With1998-10-31031 October 1998 Summary Rept of Changes,Tests & Experiments Completed, Including SEs Covering Period on 980716-1031.With SVP-98-326, Monthly Operating Repts for Sept 1998 for Quad Cities Nuclear Power Station,Units 1 & 2.With1998-09-30030 September 1998 Monthly Operating Repts for Sept 1998 for Quad Cities Nuclear Power Station,Units 1 & 2.With ML20153D0191998-09-18018 September 1998 Part 21 Rept Re Defect in Gap Conductance Analyses for co- Resident BWR Fuel.Initially Reported on 980917.Corrective Analyses Performed Demonstrating That Current Operating Limits Bounding from BOC to Cycle Exposure of 8 Gwd/Mtu ML20153C6771998-09-17017 September 1998 Part 21 Rept Re Defect Relative to MCPR Operating Limits as Impacted by Gap Conductance of co-resident BWR Fuel at Facilities.Operating Limit for LaSalle Unit 2 & Quad Cities Unit 2 Will Be Revised as Listed ML20151T2711998-09-0404 September 1998 Safety Evaluation Accepting Licensee Response to NRC Bulletin 95-002 ML20151Y7261998-08-31031 August 1998 Monthly Operating Repts for Aug 1998 for Quad Cities Nuclear Power Station ML20237E2331998-08-21021 August 1998 Revised Pages of Section 20 of Rev 66 to CE-1-A, QA Topical Rept ML20151Y7301998-07-31031 July 1998 Revised MOR for Jul 1998 for Quad Cities Nuclear Power Station,Units 1 & 2 ML20237A6251998-07-31031 July 1998 Monthly Operating Repts for July 1998 for Quad Cities Nuclear Power Station,Unit 1 & 2 SVP-98-328, Summary Rept of Changes,Tests & Experiments Completed, Including SEs Covering Period of 971001-980715,per 10CFR50.59 & 10CFR50.71(e).With1998-07-15015 July 1998 Summary Rept of Changes,Tests & Experiments Completed, Including SEs Covering Period of 971001-980715,per 10CFR50.59 & 10CFR50.71(e).With SVP-98-249, Monthly Operating Repts for June 1998 for Quad Cities Nuclear Power Station,Units 1 & 21998-06-30030 June 1998 Monthly Operating Repts for June 1998 for Quad Cities Nuclear Power Station,Units 1 & 2 SVP-98-215, Monthly Operating Repts for May 1998 for Quad Cities Nuclear Power Station Units 1 & 21998-05-31031 May 1998 Monthly Operating Repts for May 1998 for Quad Cities Nuclear Power Station Units 1 & 2 ML20247N6281998-05-19019 May 1998 Rev 2 to COLR for Quad Cities Unit 2 Cycle 15 ML20216C0561998-04-30030 April 1998 Safe Shutdown Rept for Quad Cities Station,Units 1 & 2, Vols 1 & 2.W/22 Oversize Figures SVP-98-176, Monthly Operating Repts for Apr 1998 for Quad Cities Nuclear Power Station,Units 1 & 21998-04-30030 April 1998 Monthly Operating Repts for Apr 1998 for Quad Cities Nuclear Power Station,Units 1 & 2 ML20217D0281998-04-22022 April 1998 Part 21 Rept Re Additive Constants Used in MCPR Determination for Siemens ATRIUM-9B Fuel by Core Monitoring Sys Were Found to Be non-conservative.SPC Personnel Notified All Customers w/ATRIUM-9B Lead Test Assemblies ML20217G3951998-04-0808 April 1998 TS 3/4.8.F Snubber Functional Testing Scope Quad Cities Unit 2 TS (Safety-Related) Snubber Population 129 Snubbers SVP-98-128, Monthly Operating Repts for Mar 1998 for Quad Cities Nuclear Station Units 1 & 21998-03-31031 March 1998 Monthly Operating Repts for Mar 1998 for Quad Cities Nuclear Station Units 1 & 2 1999-09-30
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Text
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- Comm: :lth Edison ,
. e Quad-Citi ic! ear Power Station
' E4 Post Offic 216 \
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- Cordova, Illinois 61242 Telephone 309/654-2241 q' /A FAP-72-146 fl((llr}'9, July 24, 1972 7 8g
%A;5?M ,W Mr. J. F. O' Leary 4 O)d#
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Director, Directorate of Licensing /. 4y U. S. Atomic Energy Commission 0 0 0N l Washington, D. C. 20545 50-265 j
Dear Mr. O' Leary:
Ref: Quad-Cities Nuclear Power Station Unit 2-DPR-30 Appendix A Sections 6.6.A.3 and 6.6.B.3 The purpose of thi,s letter is to inform you of the details regarding the fire in two electrical cable trays in the Reactor Building at Quad-Cities Station Unit 2. This incident occurred on' July 16, 1972, at 1:30 a.m.
.This letter was prepared by Messrs. B. II . Temple,;.N. J.
Kalvianakis, and W. C. Lui, members of a special committee appointed by Messrs. II . K. Hoyt (Superintendent of Generating Station - Nuclear) and F. A. Palmer (Superintendent, Quad-Cities Nuclear Power Station) to investigate the incident.
Description of Incident During startup testing on Unit 2 at Quad-Cities Nuclear Power Station with the reactor operating at 8 p ermal power, t
Reactor Water Recirculation Motor-Generator Set 2B tripped and the indicating lights for the following equipment were initially observed to be out:
Reactor Water Recirculation Suction Valve MO-2-202-4B Reactor Water Recirculation Discharge Valve MO-2-202-5B Reactor Water Recirculation Equilizer Bypass Valve MO 202-9B.
Reactor Water Recirculation Discharge Bypass Valve MO 202-7B Drywell Cooling Blower 2C, 2D, and 2E Standby Liquid Control System Reactor thermal power dropped to 60% after the recirculation pump tripped. An operator was dispatched to investigate and attempt to reset the Reactor Water Recirculation System in preparation for restart of 2B Recirculation Motor-Generator Set 8304060 DR ADO 2] % PDR g$$
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l Mr. J . F. O' Leary July 24,1972 while a second operator was dispatched to investigate the Standby Liquid Constrol System and replace the blown fuse.
It was discovered that none of theThe valves that had lost their voltage on their supply indicating lights could be reset.
bus, Bus 28/29-5, was checked and found to be normal. One of the Station Operating Engineers was notified while a search for abnormal conditions in the station was in progress. At 3:30 a.m.
a small fire was discovered in the two electrical cable trays in the TIP (Traversing In-Core Probe) room in Unit 2. The fire appeared to have extinguished itself. No smoke was detected in the room when it was entered. Some remaining sparks were extinguished with a portable CO2 fire extinguisher.
Immediate Actions ,
When the source of'the trouble was discovered an orderly shut-down of the reactor was initiated. The All equipment affected by the fire was taken out of service. turbine-generator was off the system at 5:37 a.m. and all rods were inserted by 9: 16 a.m.
Investigation An investigation was performed to determine the cause of the fire. Engineers from Commonwealth Edison's Station Electrical Engineering and Production Departments were on site to investigate and concurred with the corrective action recommended by General Electric. Investigation revealed the following:
- 1. Location The fire was located in the TIP room in the southwest quad-rant of the 595' elevation of the Reactor Building in cable pan sections 603T (top) and 603B (bottom). The trays are 30" wide and 6" deep. The fire had confined itself to approx-imately 5 feet at the end of cable tray sections 603T and 603B. The major burning was in the bottom cable tray, 603B.
The cable trays in this area are ladder type trays and run against the drywell at elevations 614'6" and 616' above the drywell penetrations which are at elevation 611' . The only I cables affected were those going through penetration 100F.
All of the electrical leads coming out of penetration 100F were led up into the bottom ladder tray and spliced onto the incoming cable after their jackets had been stripped back.
Nineteen of the total of 24 damaged cables had been routed in the bottom cable tray. The remaining 5 cables had been routed in the top cable tray and looped down through the ladder section for splicing to the penetration leads in the lower ladder section directly above penetration 100F.
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l Mr. J. F. O' Leary July 24, 1972 j
- 2. Location of Hottest Area of Fire The apparent hot spot of the fire was located approximately 3 feet from the end of the bottom pan and towards the front edge of the pan. This was also the location of the majority of the conductor splices.
- 3. Area of spread The fire had fairly well consumed all of the insulation in the front half of the lower pan at the apparent fire center.
It had spread along the loosely stacked cables to the point where they all came together in a tight bundle (approximately 2 feet from.the fire center) to continue on along the pan.
The fire extinguished itself as soon as it reached this tightly packed bundle and at the top of the vertical run of i the penetration leads where they dropped out of the ladder tray. . One 2-conductor number 14 cable (25799) in the top
. pan directly above the fire hot spot had its insulation burned i off. Its conductors had burned throu6h in two places as a result of electrical arcing on two of the rungs of the ladder section. Four reactor protection system cables were also routed through penetration 100F. As specified by design they
, were in separate conduit and were completely unaffected by the fire.
- 4. Cause of Fire Cable 22770, the 3 conductor 1/0 power feed for Drywell Cooling Blower 2D, was buried near the bottom of the apparent hottest spot of the fire. After removal of the upper cables, it was observed that phase T3 (C) conductor was burned through at the end of the splice which had the penetration leads crimped into it. The cable was burned completely through with face pitting and fusion. This type of burn is typical of a high resis tence j oint failure. The 2 pieces of the conductor were completely separated and there was no apparent arcing to the other phases of the cables, to any of the surrounding wires, or the cable pan.
Repair Procedure
- 1. The conductor bundles from the penetration were all opened and the individual conductors were cut back to good conductor and good insulation. This left approximately a three foot lead a
from the penetration.
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. O O Mr. J . F . O ' Le ary July 24, 1972 No indications of over heating , ere discovered. Since the original splices were approximately 30,000 circular mills short on copper at the penetration end, all drywell cooling blower feeds will be respliced in the following sequence:
A. All feeds through penetration 100F in Unit 2 drywell.
B. All feeds through penetration 104B in Unit 2 drywell.
C. Unit #1 cables outside of the drywell as each drywell cooler can be taken out of service.
After completion of the cable repairs, tests were conducted to verify the correct operation of the affected system components.
All required tests were completed prior to startup. The unit was returned to ser.vice at 3:45 p.m. on July 22, 1972.
Conclusion y_Invea.tigation indicated that the fire originated from a bad
--splice in the C phase of the power feed for the 2D Drywell
-Gooler. Although there was little visible evidence of arcing, it is reasonable to assume that the initial arc from this cable contributed to the spread of the fire. It was apparent that the fire did not spread out of the main zone by radiation or conduction. All power cable protective devices functioned pro-perly and were instrumental in containing the spread of the fire.
Had the fire occurred in an area where both division I and II engineered safeguard system cables were present, the separation system used throughout the plant would have prevented spread of the fire from one division to the other. A complete list of cables damaged by the fire is attached.
Very truly yours, COMMONWEALTH EDISON COMPANY Quad-Cities N clear Power Station
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. O Mr . J . F . O ' Le ary July 2 4, 19 72 i
- 2. A new 12" x 12" wire tray was run from the original pan just inside the TIP room wall directly to the penetration.
- A 45 elbow sloping down to the penetration was installed on the wire duct for the purpose of making the cable splices.
> 3 The cables in the bottom tray were rerouted into the new 1
wire tray.
! 4. The cables in the upper pan were left there until they were rolled out for splicing.
5 The rerouted cables were then cut back to good conductor and good insulation and spliced to the penetration leads as per original procedures with the exception of the drywell cooling blower power feeds.
- 6. For the drywell ' cooling blower power feeds , the splices were accomplished as follows:
A. A standard 1/0 YS25 Burndy sleeve was crimped to the 1/0 cable with a Burndy MY29-3 indenting tool.
B. The seven #10 101 strand conductors from the pene-i tration were inserted into the other end of the 1/0 connector sleeve.
C. Three solid #10 copper filler pins were then inserted into the connector sleeve with the #10 stranded con-ductors. This gave a total copper cross section of i 102,800 circular mills verses 105,600 circular mills for 1/0 cable. The connector was then indented using a MY29-3 indenting tool set on the 1/0 index. Several I of the splices were made on a trial basis. They were 2
then sectioned at the dents in the connector. The cross-section of the 7 #10 101 strand conductors plus d
the 3 #10 filler pins were a homogeneous cross-section of copper. The #10 solid conductors were not visually identifiable in the cross-section. The entire splice was of very good quality. !
7 The connectors were then insulated with GE6380 tape and the l splice completed with a Scotchcast 82-A2 splicing kit.
Corrective Action All remaining drywell cooling blower feeds for both Unit 1 ana 2 were physically inspected for evidence of overheating at the splice, i
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REACTOR WATER RECIRCULATION l CABLE # EQUIPMENT NAME 22857 Suction valve MO-2-202-4B Motor 22864 Discharge valve MO-2-202-5B Motor 22871 Discharge Bypass valve MO-2-202-7B Motor 22883 Equilizer bypass valve MO-2-202-98 Motor 22859 Suction valve M0-2-202-4B Limit switches 22866 Discharge valve MO-2-202-5B Limit switches 22873 Discharge bypass valve MO-2-202-73 Limit switches 22885 Equilizer bypass valve MO-2-202-9B Limit switches 20592 Current Transformer at 2B pump motor 25779 Discharge valve MO-2-202-5B limit switches DRYWELL COOLING ,
22430 Blower 2C 22770 Blower 2D 22435 Blower 2E TIP (TRAVERSING IN-CORE PROBE) 25183 Channel #1 Indexing Mechanism 25185 Channel #2 Indexing Mechanism 25769 Channel #3 Indexing Mechanism 25771 Channel #4 Indexing Mechanism 25773 Channel #5 Indexing Mechanism STANDBY LIQUID CONTROL 26340 Shut-off valve 1101-1 position indication RESIDUAL HE AT REMOVAL 20687 Injection manual valve 1001-33B position indication.
22573 Shutdown cooling isolation Valve 1001-50 limit switch.
PLANT EVACUATION 24240 Reactor Building Siren - S-35 RE ACTOR VALVES AND EQUIPMENT 26336 Reactor head cooling drain valve A0-2-220-47 26326 Reactor head seal instrument Shut-off valve A0-2-220-52 l
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