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Category:LICENSEE EVENT REPORT (SEE ALSO AO RO)
MONTHYEARML20211G6101997-09-29029 September 1997 LER 97-S01-00:on 970829,discovered That Shotgun Shell Was Missing from Ammunition Carrier.Caused by Improper Issuance & Control of Weapons & Ammunition.Missing Shell Immediately Replaced 05000298/LER-1993-0301993-08-0909 August 1993 LER 93-030-00:on 930708,discovered Inadequate Inservice Testing of Svc Water Sys Check Valves Due to Flow Instrument Calibr Error & Personnel Error in Specifying Sys Flow Requirement.Revised Calibr Factor developed.W/930809 Ltr 05000298/LER-1993-0291993-08-0606 August 1993 LER 93-029-00:on 930708,partially Withdrawn Control Rod 26-31 Scrammed.Caused by RPS Scram Signal from Two of Four Reactor Pressure Scram Switches.Scram Reset & Pressure Testing completed.W/930806 Ltr 05000298/LER-1993-0081993-04-22022 April 1993 LER 93-008-00:on 930328,4,160-volt Breakers 1BG,1GB & 1GE Tripped Due to Actuation of 1GS Breaker Lockout Relay.Caused by Oversight in Design Change Installation Instructions. Work Stopped & Design Change Implemented 05000298/LER-1993-0071993-04-22022 April 1993 LER 93-007-00:on 930326,determined That ECCS Pump Compartment Cooler Power Supply Design Defect Could Have Prevented Adequate Containment Heat Removal.Caused by Design Errors.Mods Planned to Provide Convection Cooling 05000298/LER-1993-0011993-03-18018 March 1993 LER 93-001-00:on 930225,several Design Discrepancies in SW & Reactor Equipment Cooling Sys Identified.Caused by Piping Configuration Error During Plant Const.Mechanical & Electrical Mods Will Be made.W/930318 Ltr 05000298/LER-1982-014, Amended LER 82-014/03X-1:on 820607,rejectable Indications Found in Heat Affected Zones of Welds BJ-13,BJ-15,BJ-20 & BJ-23.Caused by Improper Fabrication Techniques.Affected Piping & Fittings Replaced1983-02-0303 February 1983 Amended LER 82-014/03X-1:on 820607,rejectable Indications Found in Heat Affected Zones of Welds BJ-13,BJ-15,BJ-20 & BJ-23.Caused by Improper Fabrication Techniques.Affected Piping & Fittings Replaced 05000298/LER-1982-005, Amended LER 82-005/03X-1:on 820221,surveillance Test Indicated MSIV 86A Had Closing Time Faster than Allowed by Tech Specs.Caused by Loss of Nitrogen Preload & Fluid from Hydraulic Sys Accumulator.Accumulator Recharged & Tested1982-10-20020 October 1982 Amended LER 82-005/03X-1:on 820221,surveillance Test Indicated MSIV 86A Had Closing Time Faster than Allowed by Tech Specs.Caused by Loss of Nitrogen Preload & Fluid from Hydraulic Sys Accumulator.Accumulator Recharged & Tested ML20204D4181978-12-0808 December 1978 /01T-0 on 781124:while Checking RR MG Set 1B Speed control,1 Amp fuse(F2) in Drive Amplifier Circuit Was Pulled Out & Checked.Caused Rapid Increase in Pwr Level,Tripping RR MG Set 1B ML20148A7231978-12-0707 December 1978 /03L-0 on 781111:during S.P.6.4.1.2. (Withdrawn Control Rod Operability), Control Rod 30-31 Was Found Uncoupled from Its Control Rod Drive. Cause Cannot Be Determined Until Control Rod Drive Is Removed & Examined ML20150D3371978-11-13013 November 1978 /03L-0 on 781016:cable HP58,Div 1,was Found Routed Thru Div 11 Riser Under Panel 9-3.Caused by Personnel Error During Constr.Redundant Sys Were Operable ML20150D3191978-10-13013 October 1978 /03L-0 on 780918:during Normal Oper,After Cooling Down the Torus,Rhr Heat Exchanger 1B Bypass RHR-MO-66B Failed to Oper,Due to Loose Set Screw on Yoke of Limiterque SMB-3 Operator ML20150D3561978-09-28028 September 1978 /03L-0 on 780912:Cooper-Bessemer Diesel Engine Type KSV-16-T Main Bearing Overheated.Main Bearing 10 Was Damaged Due to Minimal Oil Flow During Engine Coastdown. Caused by Foreign Particles in Oil 1997-09-29
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20212K9781999-09-30030 September 1999 Safety Evaluation Accepting USI A-46 Implementation Program ML20217A9931999-09-30030 September 1999 NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data ML20217G7461999-09-30030 September 1999 Monthly Operating Rept for Sept 1999 for Cooper Nuclear Station 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 ML20212C5001999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Cooper Nuclear Station ML20211D6491999-08-25025 August 1999 Part 21 Rept Re Nonconformance within LCR-25 safety-related Lead Acid Battery Cells Manufactured by C&D.Analysis of Cells Completed.Analysis of Positive Grid Matl Shows Nonconforming Levels of Calcium within Positive Grid Alloy ML20210R0381999-07-31031 July 1999 Monthly Operating Rept for July 1999 for Cooper Nuclear Station ML20210J2921999-07-29029 July 1999 Special Rept:On 990406,OG TS & Associated Charcoal Absorbers Were Removed from Svc.Caused by Scheduled Maint on Hpci. Evaluation of Offsite Effluent Release Dose Effects Was Performed to Ensure Plant Remained in Compliance ML20209H8281999-07-15015 July 1999 Safety Evaluation Accepting GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves, for Cooper Nuclear Station ML20211A9981999-07-12012 July 1999 Draft,Probabilistic Safety Assessment, Risk Info Matrix, Risk Ranking of Systems by Importance Measure ML20196H8621999-06-30030 June 1999 NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data, June 1999 Rept ML20209E1061999-06-30030 June 1999 Monthly Operating Rept for June 1999 for Cns.With ML20196B3851999-06-17017 June 1999 Summary Rept of Facility Changes,Test & Experiments,Per 10CFR50.59 for Period 970901-990331.Summary of Commitment Changes Made During Same Time Period Also Encl ML20195K2851999-05-31031 May 1999 Monthly Operating Rept for May 1999 for Cooper Nuclear Station.With ML20206P0481999-05-12012 May 1999 Safety Evaluation Concluding That NPP Established Acceptable Program to Verify Periodically design-basis Capability of safety-related MOVs at CNS & Adequately Addressed Actions Requested in GL 96-05 ML20206J0811999-05-0404 May 1999 Rev 14 to CNS QA Program for Operation ML20206P9751999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for Cooper Nuclear Station ML20205Q0891999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for Cooper Nuclear Station.With ML20204G8951999-03-15015 March 1999 CNS Inservice Insp Summary Rept Fall 1998 Refueling Outage (RFO-18) ML20207M9231999-03-12012 March 1999 Amended Part 21 Rept Re Cooper-Bessemer Ksv EDG Power Piston Failure.Total of 198 or More Pistons Have Been Measured at Seven Different Sites.All Potentially Defective Pistons Have Been Removed from Svc Based on Encl Results ML20204B3701999-03-11011 March 1999 SER Accepting Third 10-year Interval Inservice Insp Plan Requests for Relief for RI-17,Rev 1 and RI-25,Rev 0.Request for Relief RI-13,Rev 2 Involving Snubber Testing & Is Being Evaluated in Separate Report ML20204C9751999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for Cooper Nuclear Station ML20199E6751999-01-14014 January 1999 Monthly Operating Rept for Dec 1998 for Cooper Nuclear Station ML20195B9191998-12-31031 December 1998 1998 NPPD Annual Rept. with ML20196J9641998-12-0707 December 1998 Safety Evaluation Accepting Licensee Third 10-yr Interval Inservice Insp Plan Request for Relief RI-27,rev 1 ML20198D2471998-11-30030 November 1998 Monthly Operating Rept for Nov 1998 for Cooper Nuclear Station.With ML20196A2861998-11-23023 November 1998 SER Re Core Spray Piping Weld for Cooper Nuclear Station. Staff Concluded That Operation During Cycle 19 Acceptable with Indication re-examined During RFO 18 ML20196A5241998-11-23023 November 1998 Safety Evaluation Accepting Proposed Alternative to Use UT Techniques Qualified to Objectives of App Viil as Implemented by PDI Program in Performing RPV Shell Weld & Shell to Flange Weld Examinations ML20196A5061998-11-23023 November 1998 Safety Evaluation Re Flaw Indication Found in Main Steam Nozzle to Shell Weld NVE-BD-N3A at Cns.Plant Can Be Safely Operated for at Least One Fuel Cycle with Indication in as-is Condition ML20196C4241998-11-20020 November 1998 Rev 1 to Cooper Nuclear Station COLR Cycle 19 ML20195H1761998-11-17017 November 1998 SER Authorizing Proposed Alternative in Relief Requests RV-06,RV-07,RV-09,RV-11,RV-12 & RV-15 Pursuant to 10CFR50.55a(a)(3)(ii).RV-08 Granted Pursuant to 10CFR50.55a(f)(6)(i) & RV-13 Acceptable Under OM-10 ML20195F8601998-10-31031 October 1998 Monthly Operating Rept for Oct 1998 for Cooper Nuclear Station.With ML20155D9961998-10-31031 October 1998 Rev 0 to GE-NE-B13-01980-24, Fracture Mechanics Evaluation on Observed Indication at N3A Steam Outlet Nozzle to Shell Weld at Cooper Nuclear Station ML20154Q5661998-10-0505 October 1998 Rev 0 to CNS COLR Cycle 19 ML20154L5381998-09-30030 September 1998 Monthly Operating Rept for Sept 1998 for Cooper Nuclear Station.With ML20151Z6141998-09-16016 September 1998 SER Accepting Util Responses to NRC Bulletin 95-002 for Cooper Nuclear Station ML20154F7931998-08-31031 August 1998 Rev 0 to J11-03354-10, Supplemental Reload Licensing Rept for CNS Reload 18,Cycle 19 ML20153B1101998-08-31031 August 1998 Monthly Operating Rept for Aug 1998 for Cooper Nuclear Station ML20237E7771998-08-20020 August 1998 Revised COLR Cycle 18 for Cooper Nuclear Station ML20151Q1211998-08-14014 August 1998 Rev 0 to Control of Hazard Barriers ML20237C0591998-07-31031 July 1998 Monthly Operating Rept for Jul 1998 for Cooper Nuclear Station ML20236R9131998-07-20020 July 1998 SER Accepting Rev 13 to Quality Assurance Program for Operation Policy Document for Plant ML20236P2971998-07-0707 July 1998 Rev 2 to NPPD CNS Strategy for Achieving Engineering Excellence ML20236R0931998-06-30030 June 1998 Monthly Operating Rept for June 1998 for Cooper Nuclear Station ML20249A7701998-05-31031 May 1998 Monthly Operating Rept for May 1998 for Cooper Nuclear Station ML20247G6131998-05-13013 May 1998 Part 21 Rept Re Defect Contained in Automatic Switch Co, Solenoid Valves,Purchased Under Purchase Order (Po) 970161. Caused by Presence of Brass Strands.Replaced Defective Valves ML20247G0951998-04-30030 April 1998 Monthly Operating Rept for Apr 1998 for Cooper Nuclear Station ML20237B6861998-04-24024 April 1998 Vols I & II to CNS 1998 Biennial Emergency Exercise Scenario, Scheduled for 980609 ML20217A1531998-04-16016 April 1998 Closure to Interim Part 21 Rept Submitted to NRC on 970929. New Date Established for Completion of Level I & 2 Setpoint Project Committed to in .Final Approval of Setpoint Calculations Will Be Completed by 980531 ML20216G5331998-03-31031 March 1998 Monthly Operating Rept for Mar 1998 for Cooper Nuclear Station 1999-09-30
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5' COOPER NUCLI AR ST AllON 9? PO Ef 0 x 98, BROWNVf LLE., NE BR ASKA 683?J x -
Nebraska Public Power District u cc - ON w m ers.3.i>
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CNSS933202 August 6, 1993 U.S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555
Dear Sir:
Cooper Nuclear Station Licensee Event Report 93-029, Revision 0, is forwarded as an attachment to this letter.
Sincerely, r
i-R. L. Cardner Plant Manager RLG/j u Attachment cc: J. L. Milboan G. R. Horn J. M. Meacham R. E. Wilbur V. L. Wolstenholm D. A. Whitman INPO Records Center NRC Resident Inspector R.J. Singer CNS Training CNS Quality Assurance 130036 9308130125 930806 PDR ADOCK 05000298 fYS 1 -
S PDR i
s NRCParm 3tt U S. NUCLEAR KEGULATO'tY COMMtS$ ION APPROVED OMB WO.215 4 0104
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LICENSEE EVENT REPORT (LER)
Fl4tLITY hAME H) DUCEET NUMBER (2) P'6E 43' Cooper Nuclear Station o 151010 l 0 l 2 l 9 l 8 1 lOFl0 l3 TITLE tai Spurious RPS Actuation During Performance Of The ASME Class 1-N System 1.eakage Test EVENT DATE (5) LER NUMSER 46) REPORT DATE (71 OTHE R F ACILITIEE INVOLVED (S)
MONTH DAY YEAR YEAR ley h*' :
[$ MONTM DAY YEAR F ACILITV h AMES DOCKET NUMBEHt$)
015!o]ojol l l 1
l 0l 7 0l 8 9 3 9l 3 0l2l9 0l0 0l 8 0l6 9l 3 oistogogog OPE uT M THIS REPORT f5 SUBMITTED PUR5 DANT TO THE REQUIREMENTS OF 10 CFR {: (Chece one or mom af the fehowg1011 "N 8 N 20.e2t : 20. mm y so.73c.H2H 73 71te)
POWER 20 4066aH1HG 50.36sclui SO.73teH2Hv) 73.71(c)
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oTHER<s e ,.A6,r-r be ow eno en Test. NRC f ann 20 40BleH1 Hei) 60.73Lal[2Hil 50.73talt2Hve4HA) .166A1 20 405deH1}ne) 50.73saH2)hel 50.73 tax 2HvHiHB) 20 406aaH1 Hv) 90.731sH2He,il 50.73 eH2Hal LtCENSEE CONTACT FDR THit LER H2)
GAME TELEPwONE NUMBER ARE A CODE.
John R. Myers 4;Ol2 8;2 3 5g3 1 1 8 1 1l1 COMPLETE ONf LINE FOR EACH COMPONENT F AILURE DESCRfBED IN THIS REPORT (13P CAUSE SYSTEM CDMPONfNT
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$USMi&5 ION s VES (19 ven. commere EXPECTED SUBnUS90N DA TE) NO l l l u.sTR ACT omir , re m r e.. .-me., , u .,- r,-r- ae nm On July 8, 1993, at 11:42 pm, a Reattor Protection System (RPS) scram signal was generated from two of the four Reactor Pressure-Scram switches, resulting in the scramming of partially withdrawn control Rod 26-31. The Reactor Coolant System was pressurized to perform the ASME Class 1-N System Leakage Test in accordance with ASME Code requirements prior to startup from the 1993 Refueling Outage. In concert with this test, other activities were being performed, including replacement of instrument line excess flow check valves and scram time testing. The common sensing line for the two Reactor Pressure-Scram switches was apparently vibrated during activities associated with replacing an excess flow check valve, resulting in a minor pressure perturbation at the switches. With the Reactor at approximately 1030 psig, only a small margin existed between the Reactor Coolant System pressure and the switches' setpoint.
The scram was reset, and pressure testing completed without further difficulties. Procedures related to ASME Class IN system leakage testing will be revised to require better coordination of activities in the vicinity of the High Pressure-Scram switches and sensing lines which could result in an RPS actuation.
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NRC Faste 306 W. 3)
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s N';C form 3e64 -
U.S. NUCLEAR REGULATORY COMMISSION UCENSEE EVENT REPORT ILER) TEXT CONTINUATION AreRovEo oms No. 3 so-oio.
EXPIRES: 8/3U88 f ACILITV NAME 4:ll DOCF.ET NUMBER (2) LER NUMBER (6) PAGE (3) .
naR "th"W' ' n's"llC Cooper Nuclear Station o l5 l0 jo lo l2 l 9l 8 9l3 -
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0l0 0l2 OF 0 l3 TEXT <a-, a-. . w. - .e-w une m m4v nn A. Event Description On July 8,1993, at 11:42 pm, a Reactor Protection System (RPS) scram signal was generated from NBI-PS-55C and D, two of the Reactor Pressure-Scram switches. As a result, Control Rod 26-31, which was partially withdrawn for performing scram timing, automatically scrammed.
The Reactor Coolant System was pressurized to approximately 1030 psig to:
perform the ASME Ciass 1-N System Leakage Test, in accordance with ASME Code requirements, prior to startup from the 1993 Refueling' Outage. In.
concert with the leakage testing, other activities, including replacement of instrument line excess flow check valves and scram time testing, were being performed. In response to the scram, correct plant response was verified and an investigation of the cause was begun,.since the immediate indications did not identify a specific cause for the actuation. Further investigation found that at the time of the actuation replacement of an excess flow check valve was in progress on an instrument line emanating i' rom the penetration which contains the common sensing line for these Reactor Pressure-Scram switches.
B. Plant Status The plant was in cold shutdown, in the 1993 Refueling Outage. The Reactor was pressurized to npproximately 1030 psig, undergoing the ASME.
~
Class 1-N System Leakage. test, with a Reactor coolant temperature of approximately 190 degrees Fahrenheit.
C. Basis for Report Actuation of the Reactor Protection System resulting in the scram of a single, partially withdrawn, control rod, report.able in accordance with the criteria prescribed by 10CFR50. 73(a)(2)(iv) .
D. Cause The common sensing line for the two Reactor Pressure-Scram switches was apparently vibrated while completing the activities associated with replacing the excess flow check valve, resulting in a minor pressure perturbation at the switches. With the Reactor at approximately 1030 psig, only a small margin existed between the-Reactor Coolant System pressure and the switches' setpoint. Under these conditions, a minor pressure perturbation would be sufficient to actuate the switches.
~
The procedures for performing the pressure testing, excess flow check valve testing and maintenance, and control rod scram timing did not contain guidance to ensure that the activities being conducted would not result in a scram should minor pressure perturbations occur.
E. Safety Sirnificance Reactor Coolant System pressure was being maintained at approximately 1030 psig, below the nominal 1045 psig setpoint of the Reactor Pressure-Scram switches. Under these conditions, the spurious pressure perturbation did .not represent a challenge to the integrity of the Reactor Coolant System.
NRC FORM 3e6A . *U.S. GPoe 199 S 470- 589 - 0 C010 nc3)
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2MC hem 386A .
"'*- s LICENSEE EVENT REPORT (LER) TEXT CONTINUATION U1 NUCLEAR REGULATNtY COMMISSION APPRovfo cus No. s.'so4ios '
I EXPIRES: 9/31/SB
.{
f ACILITY NAME (18 ' DOCKET NUMBER (21 !
LER NUMBER f6) PAGE t3) v5^a "Stt%^' OTJ'#2 Cooper Nuclear Station O0 TEKT 4W more spece a ruoaned, ese addeonn! NRC hum 386ksi(118 o l5 l0 l0 l0 l 2 l9 y 8 9l3 _
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F. Safety Implications 6
t The high pressure scram is intended'to prevent exceeding the Technical Specification safety limit of 1337 psig for Reactor Coolant System overpressurization during operational transient conditions. Under transient conditions, a scram at 1045 psig will ensure that the safety limit is not challenged. The. pressure testing which was'being performed is conducted only while the Reactor is shutdown, and thus the pressure '
rise associated with an operational transient condition is not-applicable.
I G. Corrective Action i
The scram was reset, and a reduced pressure maintained while investigating the scram. Subsequently, pressure testing was completed i without further difficulties. Procedures related to ASME Class IN system leakage testing will be revised to require better coordination of activities in the vicinity of the High Pressure-Scram switches and
- sensing lines which could result in an RPS actuation.
H. Similar Events .
LER 87-01 discusses Reactor scrams which occurred during performance of ,
similar testing in 1987. Procedures were revised, however, the >
revisions did not place appropriate restrictions upon the operational '
conditions required to prevent RPS actuations.
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I WRC FORM 38ea * .5. CT N 14811- 5.10 '*H 9 dWP 49 53; t.
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05000298/LER-1993-001 | LER 93-001-00:on 930225,several Design Discrepancies in SW & Reactor Equipment Cooling Sys Identified.Caused by Piping Configuration Error During Plant Const.Mechanical & Electrical Mods Will Be made.W/930318 Ltr | | 05000298/LER-1993-007 | LER 93-007-00:on 930326,determined That ECCS Pump Compartment Cooler Power Supply Design Defect Could Have Prevented Adequate Containment Heat Removal.Caused by Design Errors.Mods Planned to Provide Convection Cooling | | 05000298/LER-1993-008 | LER 93-008-00:on 930328,4,160-volt Breakers 1BG,1GB & 1GE Tripped Due to Actuation of 1GS Breaker Lockout Relay.Caused by Oversight in Design Change Installation Instructions. Work Stopped & Design Change Implemented | | 05000298/LER-1993-029 | LER 93-029-00:on 930708,partially Withdrawn Control Rod 26-31 Scrammed.Caused by RPS Scram Signal from Two of Four Reactor Pressure Scram Switches.Scram Reset & Pressure Testing completed.W/930806 Ltr | | 05000298/LER-1993-030 | LER 93-030-00:on 930708,discovered Inadequate Inservice Testing of Svc Water Sys Check Valves Due to Flow Instrument Calibr Error & Personnel Error in Specifying Sys Flow Requirement.Revised Calibr Factor developed.W/930809 Ltr | |
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