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Category:LICENSEE EVENT REPORT (SEE ALSO AO RO)
MONTHYEAR0CAN129805, LER 98-S02-00:on 981124,security Officer Found Not to Have Had Control of Weapon for Period of Approx 3 Minutes Due to Inadequate self-checking to Ensure That Weapon Remained Secure.All Security Officers Briefed.With1998-12-11011 December 1998 LER 98-S02-00:on 981124,security Officer Found Not to Have Had Control of Weapon for Period of Approx 3 Minutes Due to Inadequate self-checking to Ensure That Weapon Remained Secure.All Security Officers Briefed.With 1CAN099401, LER 94-V01-00:on 940804,functional Test Procedure for Hydrogen Analyzer Was Not Adequate to Alert Plant Personnel.Caused by Personnel Oversight.Functional Test Procedure Has Been Revised1994-09-0707 September 1994 LER 94-V01-00:on 940804,functional Test Procedure for Hydrogen Analyzer Was Not Adequate to Alert Plant Personnel.Caused by Personnel Oversight.Functional Test Procedure Has Been Revised 05000368/LER-1990-0151990-07-30030 July 1990 LER 90-015-00:on 900628,channel Functional Tests for Logarithmic Power Level Nuclear Instrumentation Found to Be Inadequate.Caused by Personnel Error Re Plant Procedure Development.Plant Startup Procedure revised.W/900730 Ltr 05000368/LER-1990-0141990-07-27027 July 1990 LER 90-014-00:on 900626,reactor Trip Occurred from Approx 30% Rated Thermal Power on Low Departure from Nucleate Boiling Ratio.Caused by Erroneous Indication Received by Control Element Assembly Calculator 1.W/900727 Ltr 05000313/LER-1990-0041990-07-0202 July 1990 LER 90-004-00:on 900531,degraded Fire Barrier Penetration Seal Discovered in 2-inch Metal Sleeve Through Floor Slab & Conduit Contained within Sleeve.Caused by Personnel Oversight.Fire Barrier Sealed & Log updated.W/900702 Ltr 05000368/LER-1990-0131990-06-29029 June 1990 LER 90-013-00:on 900530,degraded Fire Barrier Penetration Discovered During Routine Tour of Auxiliary Bldg.Caused by Improper Sealing of Penetration While in Breached Condition. Fire Barrier Sealed & Personnel counseled.W/900629 Ltr 05000368/LER-1990-0111990-05-25025 May 1990 LER 90-011-00:on 900425,discovered That post-accident Hydrogen Analyzers Did Not Appear Independent Due to Sharing Common Nitrogen Bottle Supplies.Caused by Design Oversight. Plant Mod Re Nitrogen Supplies initiated.W/900525 Ltr 05000368/LER-1989-0221990-05-25025 May 1990 LER 89-022-02:on 891114,4,160-volt Ac ESF Electric Bus Unexpectedly Deenergized During post-maint Testing on Auxiliary Relay.Caused by Inadequate Test Controls.Job Order Changed to Include Test switch.W/900525 Ltr 05000368/LER-1950-368, :on 860722,while Conducting Routine Plant Tour, Fire Watch Personnel Found Asleep at Fire Watch Stations. Fire Watch Personnel Replaced.Similar Event Reported in LER 50-368/83-0431986-08-29029 August 1986
- on 860722,while Conducting Routine Plant Tour, Fire Watch Personnel Found Asleep at Fire Watch Stations. Fire Watch Personnel Replaced.Similar Event Reported in LER 50-368/83-043
05000368/LER-1986-006, Corrected LER 86-006-00:on 860421,reactor Protective Sys Actuated on Two of Four Core Protection Circulator Channels. Caused by Electrical Noise from Starting Reactor Coolant Pump.Channel C Circuit Card Replaced1986-05-30030 May 1986 Corrected LER 86-006-00:on 860421,reactor Protective Sys Actuated on Two of Four Core Protection Circulator Channels. Caused by Electrical Noise from Starting Reactor Coolant Pump.Channel C Circuit Card Replaced 05000368/LER-1979-026, Oversize Updated LER 79-026/03X-1:on 790323,fire Dampers Not Installed Per Design Specs.Caused by Installation Qa.Fire Dampers Cited Installed as Required.Aperture Card Available in PDR1984-06-0707 June 1984 Oversize Updated LER 79-026/03X-1:on 790323,fire Dampers Not Installed Per Design Specs.Caused by Installation Qa.Fire Dampers Cited Installed as Required.Aperture Card Available in PDR 05000313/LER-1983-023, Updated LER 83-023/03X-6:on 830916-1120,fire Protection Deficiencies Discovered.Caused by Inadequate Original Installation Specs & Acceptance Criteria.Walkdown Insp in Progress.Insp Program Being Developed1983-12-20020 December 1983 Updated LER 83-023/03X-6:on 830916-1120,fire Protection Deficiencies Discovered.Caused by Inadequate Original Installation Specs & Acceptance Criteria.Walkdown Insp in Progress.Insp Program Being Developed 05000368/LER-1983-045, Revised LER 83-045/03X-4:on 830916-1123,fire Protection Deficiencies Identified.Cause & Corrective Actions Listed1983-12-16016 December 1983 Revised LER 83-045/03X-4:on 830916-1123,fire Protection Deficiencies Identified.Cause & Corrective Actions Listed 05000368/LER-1983-020, Telecopy LER 83-020/01T-0:on 830519,fire Watch Posted in Wrong Location for Period Exceeding Tech Spec Limit.Further Details Will Be Provided1983-05-19019 May 1983 Telecopy LER 83-020/01T-0:on 830519,fire Watch Posted in Wrong Location for Period Exceeding Tech Spec Limit.Further Details Will Be Provided 05000313/LER-1983-007, Revised LER 83-007/03X-1:on 830308 & 21,reactor Bldg Pressure Switches PS-2403 & PS-2401 Found to Be Out of Tolerance,Opening at 18.9 Psia.Pressure Switches Reset at 18.5 Psia Per Calibr Procedure1983-04-19019 April 1983 Revised LER 83-007/03X-1:on 830308 & 21,reactor Bldg Pressure Switches PS-2403 & PS-2401 Found to Be Out of Tolerance,Opening at 18.9 Psia.Pressure Switches Reset at 18.5 Psia Per Calibr Procedure ML20064H9241978-12-19019 December 1978 /03L-0 on 781129:Chilled Water Return Containment Isolation valve,2CV3850-2,was Taken Out of Svc to Repair Control Room Communication Breaker Contacts.Breaker Was Changed to Type AB Contacts ML20147H4441978-12-19019 December 1978 /03L-0:on 781121,borated Water Storage Tank Boron Concentration Fell Below Tech Specs Limit.Caused by Open Condensate Valve CS-38.Valve Closed Increasing Boron Concentration ML20064H9071978-12-19019 December 1978 /03L-0 on 781128:both Doors of the Personnel Hatch Doors Were Simultaneously Opened,Caused by a Sheared Mechanical Door Interrlock Gear Key.Mechanical Interlock Repairs Made & Containment Air Lock Declared Operable ML20064J1741978-12-0707 December 1978 /03L-0 on 781109:during Mode 3 Oper,Emergency Diesel Generator B Tripped from 100% Load.Caused by Damage to Rod & Main Bearings,Crankshaft & Pistons,Perhaps Due to Loose Baseplate Mounting Screw ML20064E9581978-11-20020 November 1978 /01T-0 on 781107:received Notice That Speed Switch Made by Dynalog Corp Had Design Problem That Could Cause Failure.Speed Switch Replaced ML20064E9561978-11-20020 November 1978 /03L-0 on 781107:during Post-Core Hot Functional Testing B Condensate Storage Tank Dropped Below Tech Spec Minimum.Caused by Inability of Makeup Plants to Produce High Demand for Condensate ML20064E9611978-11-20020 November 1978 /01T-0 on 781107:during Mode 3 Oper,Noted That First Cycle Plant Protec Sys Setpoint Changes Req by Rosemont Pressure Transmitter Substitution Were Less than Conservative.Caused by Failure to Change Setpoint ML20064E9721978-11-20020 November 1978 /01T-0 on 781108:technician Installed Replacement Speed Switch in B Emergency Diesel Generator W/O Having Job Order Approved by Shift Supervisor.Due to Forgetfulness of Instru Technician ML20064F1021978-11-16016 November 1978 /01T-0 on 781102:during Post-Core Hot Functional Test,Prior to Initial Criticality Mode 3 Oper,Lift Setpoints of Main Steam Line Code Safety Valves Were Nonconservative. Caused by Incorrect Identification of Simmer Point ML20064E5471978-11-0303 November 1978 /03L-0 on 781006:during Post-Core Hot Functional Test,Prior to Initial Criticality,Mode 3 Oper,Emergency Feedwater Valve 2CV-1025-1,hydraulic Pump Motor Failed. Motor Was Replaced & Tested Successfully ML20064E5291978-11-0101 November 1978 /03L-0 on 781005:During Post-Core Hot Functional Testing,Pressurizer Code Safety Valve Was Gagged to Eliminate Weepage.Weepage Occurred Following Corrective Maintenance.Valve Was Allowed to Cool & Proper Oper Proved ML20204B4171978-10-0606 October 1978 /01T-0:on 780925,during Surveillance Test,Control Rod Driveline Breaker a Failed to Trip on First Attempt. Caused by Trip Mechanism Mechanical Arm Being Out of Adjustment.Arm Adjustment Corrected 1998-12-11
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217L8931999-10-31031 October 1999 Rev 1 to BAW-10235, Mgt Program for Volumetric Outer Diameter Intergranular Attack in Tubesheets of Once-Through Sgs ML20212L1141999-10-0101 October 1999 Safety Evaluation Granting Request for Exemption from Technical Requirements of 10CFR50,App R,Section III.G.2.c 0CAN109902, Monthly Operating Repts for Sept 1999 for Arkansas Nuclear One,Units 1 & 2.With1999-09-30030 September 1999 Monthly Operating Repts for Sept 1999 for Arkansas Nuclear One,Units 1 & 2.With ML20216J6271999-09-27027 September 1999 Rev 0 to CALC-98-R-1020-04, ANO-1 Cycle 16 Colr ML20212F5261999-09-22022 September 1999 SER Approving Request Reliefs 1-98-001 & 1-98-200,parts 1,2 & 3 for Second 10-year ISI Interval at Arkansas Nuclear One, Unit 1 0CAN099907, Monthly Operating Repts for Aug 1999 for Ano,Units 1 & 2. with1999-08-31031 August 1999 Monthly Operating Repts for Aug 1999 for Ano,Units 1 & 2. with ML20211F4281999-08-25025 August 1999 Safety Evaluation Concluding That Licensee Provided Acceptable Alternative to Requirements of ASME Code Section XI & That Authorization of Proposed Alternative Would Provide Acceptable Level of Quality & Safety 0CAN089904, Monthly Operating Repts for July 1999 for Ano,Units 1 & 2. with1999-07-31031 July 1999 Monthly Operating Repts for July 1999 for Ano,Units 1 & 2. with ML20210K8831999-07-29029 July 1999 Non-proprietary Addendum B to BAW-2346P,Rev 0 Re ANO-1 Specific MSLB Leak Rates 0CAN079903, Monthly Operating Repts for June 1999 for Ano,Units 1 & 2. with1999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Ano,Units 1 & 2. with ML20207E7231999-06-0202 June 1999 Safety Evaluation Authorizing Proposed Alternative Exam Methods Proposed in Alternative Exam 99-0-002 to Perform General Visual Exam of Accessible Areas & Detailed Visual Exam of Areas Determined to Be Suspect ML20196A0191999-05-31031 May 1999 Monthly Operating Repts for May 1999 for Arkansas Nuclear One,Units 1 & 2.With ML20196A6251999-05-31031 May 1999 Non-proprietary Rev 0 to TR BAW-10235, Mgt Program for Volumetric Outer Diameter Intergranular Attack in Tubesheets of Once-Through Sgs ML20195D1991999-05-28028 May 1999 Probabilistic Operational Assessment of ANO-2 SG Tubing for Cycle 14 ML20206M7711999-05-11011 May 1999 SER Accepting Relief Request from ASME Code Section XI Requirements for Plant,Units 1 & 2 0CAN059903, Monthly Operating Repts for Apr 1999 for Ano,Units 1 & 2. with1999-04-30030 April 1999 Monthly Operating Repts for Apr 1999 for Ano,Units 1 & 2. with ML20206F0691999-04-29029 April 1999 Safety Evaluation Accepting Licensee Re ISI Plan for Third 10-year Interval & Associated Requests for Alternatives for Plant,Unit 1 ML20205M6941999-04-12012 April 1999 Safety Evaluation Granting Relief for Second 10-yr Inservice Inspection Interval for Plant,Unit 1 ML20205D6061999-03-31031 March 1999 Safety Evaluation Supporting Licensee Proposed Approach Acceptable to Perform Future Structural Integrity & Operability Assessments of Carbon Steel ML20205R6351999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for Ano,Units 1 & 2. with ML20205D4711999-03-26026 March 1999 SER Accepting Util Proposed Alternative to Employ Alternative Welding Matls of Code Cases 2142-1 & 2143-1 for Reactor Coolant System to Facilitate Replacement of Steam Generators at Arkansas Nuclear One,Unit 2 ML20204B1861999-03-15015 March 1999 Safety Evaluation Authorizing Licensee Request for Alternative to Augmented Exam of Certain Reactor Vessel Shell Welds,Per Provisions of 10CFR50.55a(g)(6)(ii)(A)(5) 0CAN039904, Monthly Operating Repts for Feb 1999 for Ano,Units 1 & 2. with1999-02-28028 February 1999 Monthly Operating Repts for Feb 1999 for Ano,Units 1 & 2. with ML20212G6381999-02-25025 February 1999 Ano,Unit 2 10CFR50.59 Rept for 980411-990225 ML20203E4891999-02-11011 February 1999 Rev 1 to 97-R-2018-03, ANO-2,COLR for Cycle 14 ML20199F0351998-12-31031 December 1998 Monthly Operating Repts for Dec 1998 for Ano,Units 1 & 2 ML20198M7841998-12-29029 December 1998 SER Accepting Util Proposal to Use ASME Code Case N-578 as Alternative to ASME Code Section Xi,Table IWX-2500 for Arkansas Nuclear One,Unit 2 0CAN129805, LER 98-S02-00:on 981124,security Officer Found Not to Have Had Control of Weapon for Period of Approx 3 Minutes Due to Inadequate self-checking to Ensure That Weapon Remained Secure.All Security Officers Briefed.With1998-12-11011 December 1998 LER 98-S02-00:on 981124,security Officer Found Not to Have Had Control of Weapon for Period of Approx 3 Minutes Due to Inadequate self-checking to Ensure That Weapon Remained Secure.All Security Officers Briefed.With ML20196F4911998-12-0101 December 1998 SER Accepting Request for Relief ISI2-09 for Waterford Steam Electric Station,Unit 3 & Arkansas Nuclear One,Unit 2 ML20198D2441998-11-30030 November 1998 Monthly Operating Repts for Nov 1998 for Ano,Units 1 & 2. with ML20199F7401998-11-16016 November 1998 Rev 9 to ANO-1 Simulator Operability Test,Year 9 (First Cycle) ML20195B4801998-11-0707 November 1998 Rev 20 to ANO QA Manual Operations ML20195C4841998-11-0606 November 1998 SER Accepting QA Program Change to Consolidate Four Existing QA Programs for Arkansas Nuclear One,Grand Gulf Nuclear Station,River Bend Station & Waterford 3 Steam Electric Station Into Single QA Program 0CAN119808, Monthly Operating Repts for Oct 1998 for Ano,Units 1 & 2. with1998-10-31031 October 1998 Monthly Operating Repts for Oct 1998 for Ano,Units 1 & 2. with ML20197H0741998-10-29029 October 1998 Rev 1 to Third Interval ISI Program for ANO-1 ML20155C1351998-10-26026 October 1998 Rev B to Entergy QA Program Manual ML17335A7641998-10-22022 October 1998 LER 98-004-00:on 980923,inadvertent Actuation of Efs Occurred During Surveillance Testing.Caused by Personnel Error.Personnel Involved with Event Were Counseled & Procedure Changes Were Implemented.With 981022 Ltr ML20154J2471998-10-0909 October 1998 SER Accepting Inservice Testing Program,Third ten-year Interval for License DPR-51,Arkansas Nuclear One,Unit 1 0CAN109806, Monthly Operating Repts for Sept 1998 for ANO Units 1 & 2. with1998-09-30030 September 1998 Monthly Operating Repts for Sept 1998 for ANO Units 1 & 2. with ML20154E2171998-09-28028 September 1998 Follow-up Part 21 Rept Re Defect with 1200AC & 1200BC Recorders Built Under Westronics 10CFR50 App B Program. Westronics Has Notified Bvps,Ano & RBS & Is Currently Making Arrangements to Implement Design Mods 0CAN099803, Monthly Operating Repts for Aug 1998 for ANO Units 1 & 2. with1998-08-31031 August 1998 Monthly Operating Repts for Aug 1998 for ANO Units 1 & 2. with ML20237B7671998-08-19019 August 1998 ANO REX-98 Exercise for 980819 ML18066A2771998-08-13013 August 1998 Part 21 Rept Re Deficiency in CE Current Screening Methodology for Determining Limiting Fuel Assembly for Detailed PWR thermal-hydraulic Sa.Evaluations Were Performed for Affected Plants to Determine Effect of Deficiency ML20236X2351998-08-0505 August 1998 Part 21 Rept Re Defect Associated W/Westronics 1200AC & 1200BC Recorders Built Under Westronics 10CFR50,App B Program.Beaver Valley,Arkansas Nuclear One & River Bend Station Notified.Design Mod Is Being Developed 0CAN089804, Monthly Operating Repts for July 1998 for Ano,Units 1 & 21998-07-31031 July 1998 Monthly Operating Repts for July 1998 for Ano,Units 1 & 2 ML20196C7831998-07-30030 July 1998 Summary Rept of Results for ASME Class 1 & 2 Pressure Retaining Components & Support for ANO-1 ML20155H7161998-07-15015 July 1998 Rev 1 to 96-R-2030-02, Revised Reactor Vessel Fluence Determination ML20236R0531998-06-30030 June 1998 Monthly Operating Repts for June 1998 for Ano,Units 1 & 2 ML20249B7791998-06-22022 June 1998 Part 21 Rept Re Findings,Resolutions & Conclusions Re Failure of Safety Related Siemens 4KV,350 MVA,1200 a Circuit Breakers to Latch Closed ML20249B5091998-06-15015 June 1998 SG ISI Results for Fourteenth Refueling Outage 1999-09-30
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I . Arkansas Power & Light Company o%N a 3B 17G RusseHville. AR 72801 l Tel 501964 3100 J I
l May 25, 1990 2CAN059914 U. S. Nuclear Regulatory Commission i Document Control Desk .
Mail-Station P1-137 i Washington, D. C. 20555
SUBJECT:
Arkansas Nuclear One - Unit 2 Docket No. 50-368 License No NPF-6 Licensee Event Report No. 50-368/89-022-02 i
Gentlemen:
In accordance with-10CFR50.73(a)(2)(iv), attached is the subject supplemental
. report concerning inadequate post maintenance test controls which resulted in deenergizing a 4160 VAC Engineered Safety Features electric bus unexpectedly-while performing-post maintenance testing on an auxiliary relay. This report .
is being supplemented to provide a revised schedule for the implementation of the post maintenance testing program. I Very truly yours, i
/
E. C. Ewing i General Manager ?
Technical Support and Assessment ECC/0M/sgw Attachment cc: Regional Administrctor !
Region IV U. S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 1000 '
Arlington, TX 76011 ,
INP0 Records Center Suite 1500 1100 Circle 75 Parkway Atlar,ta, GA 30339-3064 gg y ,
9005300231 900523 //
PDR S
ADOCK 05000368 An Entergy Company - y!
PDC-
'tg Foro 1062.01A NRC Fors 366 U.S. Nuclear Regulatory Commission (9 B3) Approved OMB No. 3150 0104 Expires: 8/3.'/85
- LICEN$[E EVENT REPORT (L E R)
TAUDh hAME (1) Arkansas Nuclear One, Unit Two 100CAETNUMBER(2)(PAGE(U 10151010101 31 61 Bill 0Fl013 TITLE (4) Inadequate Post Maintenance Test Controls Resulted in Deenergizing a 4160 VAC Eagineered Safety Features Electric Bus Unexpectedly While Performing Post Maintenance Testing on an Auxiliary Relay EVENT DATE (5) LER NLMELR (6) REPORT DATI;(7)
OTHER FACILITIES INVOLVED (6) levision 15equentia l' I Month Day Year Year Number t
i fNmber Month Day Year Facility Names Docket Number (si 11 1 0 5 0 0 0
- 1. 11 4 81 9 81 9' -
Of 21 21 - 1 01 2 01 5 21.5 91 01 0 5 0 0 0 OPERA'ING ITH. 5 REPORT 15 5UBMITTED PUR5UANT
O THE REQUllsEMEN15 OF 10 CFR 5:
M30f (9) 5 (Check one or more of the followina) (11)
POWER) 20.402(b) l_ l 20.405(c)
LEVEll
_I 20.405(a)(1)(1)
I l,XI 50.73 a)(2)(v) a)(2)(iv) l_l 73.71(b)
(10) 101010 t 50.36(c)(1) _t 50.73 l l 20.405(a)(1)(11) l_
l_I 50.36(c)(2) l _ l 50.73 a)(2)(vit) l_ I 73.71(c) i
_ l 20.405(a)(1)(iii) l l_ l 50.73(a)(2)(viii)(A)l _ l Othen (Specify in Abstract below and i
_ l 20.405(a)(1)(iv) l_ 1 50.73(a)(2)(t) ii) l_ l 50.73(a)(2)(viii)(B)l in Text, NRC Form l l 20.405(a)(1)(v) I_l150.73(a)(2)(iii) 50.7.l(a)(2)( l I 50.73(a)(2)(x) l 366A)
LICEN5EE CONTACT FOR THIS LER (12)
Ge l Telephong Number Dana Millar, Nuclear Safety and Licensing Specialist tArea l ICode i 1510111916l41-13111010 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DE5CRIBED IN THIS REPORT (13) l l l 4Reportablel
. Causel$ystem i I I IReportablej LNanonent IManufacturert in NPRDS I ICause Systeml Component IManufacturer to NPRDS I I I I I I I I i 1 1 1 I I i i i 11 1 I I I I I i I I i i i l l I I I i i l I I i l l I I SIiPPLEMENT REPORT EXPECTl;D (14)
I I I I i l I i l l EXPECTED Mont h Day Year
.I'l Yes (If yes complete Expected Submission Date) til No i SUBMISSION l AB5 TRACT (Limit to 1400 spaces, i.e., approximately fifteen single-space typewritten I DALE (15) i I 1 I i 1 lines) (16)
On November 14, 1989, maintenance personnel initiated a post maintenance test, using instructions in a maintenance job order, to simulate an undervoltage on a 480 VAC Engineered Safety Features (ESF) Motor Control Center (2B5) by placing a jumper across the 2B5 undervoltage relay contacts. Immediately following this step, the normal of fsite power feeder breaker to the associated 4160 VAC ESF bus (2A3) unexpectedly opened resulting in the loss of power, to 2A3. The electrical bus deenergized as designed, The test steps provided in the job order did not identify that 2A3 would deenergize as part of the test. When 2A3 was deenergized, a low Pressure Safety Injection (LPSI) pump, which was supplying flow for decay heat removal and a Service Water pump deenergized. A standby LPSI pump powered from the redundant 4160 VAC ESF electrical bus was started in approximately one minute and flow reestablished. Since the plant had been shutdown for several days prior to this event 'he reactor decay heat levels were low and the momentary interruption of flow did not result in any significant Reactor Coolant System temperature or pressure increases. The test was teenluated and satisfactorily completed. The root cause of this event was determined to be inadequate post maintenance test controls.' An evaluation of the controls tLat are in place has been performed and the appropriate station procedures will be revised to reflect the results of the evaluation. This event is reportable pursuant to 10CFR50.73(a)(2)(iv).
a l
4
- g. %e Form 2062 # B NRC F:ro 366A U.S. Nuclear Regulatory Commission (9-83) Approved DMB No. 31'20-0104 LICENSEE EVENT REPORT (LER) TEXT CONTINVATION FACILin hAME (1) (DOCAET NUMBER (2) l LER NUMBE R (6) l FAGE (3) l l l j5equentiali l Revision!
Arkansas Nuclear One, Unit Two l l Year Numbea Number l 101510!0l01 31 61 81 81 9 -- 01 21 2 -- 01 2101210F10I3 TEM (If more space is required, use additional NRC form 366A's) (17)
A. Plant Status l
.' At the tinie of occurrence of this event Arkansas Nuclear One, Unit two (AND 2) was im Mode $
(Cold Shutdown). teactor Coolant System (RCS) (AB) temperature was approximately 178 degrees Fahrenheit and RCS pressure was about 250 psia. The seventh refueling outt.ge (2R7) for AND-?
commenced September 25, 1989 and ended November 22, 1989. .
t B. Event Description ,
During refueling outage 2R7, detailed as-built wiring verification inspections of several Control Rone cabinets were performed by engineering personnel. On November 10, 1989, it was discovered that an amiliary relay (27-1X/2B5), actuated by undervoltage relays which molitor voltage on a 480 VAC Engineered Safety Features (ESf) Motor Control Center (MCC) (285), was not correctly wired.
MCC 2B5 receives power from a 4160 VAC ESF electrical bus (2A3), which is normally energized by the stations offsite power system, in the unlikely event of a loss of offsite electrical poner or degraded offsite power voltage conditions, the offsite power scurce is automatically disconnected and an Emergency Diesel Generator (EDG) (EK) (2K4A) is started to supply 2A3 and 2B5 with power.
These functions (opening of offsite power feeder breaker and EDG start) are initiated by the 285 bus undervoltage relays which upon detection of a low voltage conditioit actuate to energize auxiliary relay 27 1X/2B5. When the auxiliary relay is erargized it functions to close contacts f in each of two redundant automatic starting circuits for the associated EDG and to provide an open signal to the normal offsite electrical feeder breaker to 2A3 (2A309). During the wiring intpection, it was found that one of the contacts on 27-1X/2B5 which provides one of the tedundant EDG automatic starting circuits was not wired into the circuit as indicated on the applicable design drawings. ,
Upon a request by engineering, a job order was issued and the wiring discrepancy corrected. To ensure the wiring was correctly performed, engineering personnel were contacted to provide recommendations for testing the circuit.
The testing method, decided upon between engineering, maintenance and operations was included in the maintenance $b order, which was used to correct the wiring error, to provide the necessary steps to perform the test. The proposed test included placing the start handswitch for 2K4A in the pull-to-lock position to prevent actual ttarting of the EDG and verifying both of the redundant EDG emergency start relays would energize when an undervoltage condition was simulated on 2B5.
At approximately 0245 hours0.00284 days <br />0.0681 hours <br />4.050926e-4 weeks <br />9.32225e-5 months <br /> on November 14, 1989, maintenance personnel initiated the test by placing a jumper across the 2B5 bus undervoltage relay contacts to simulate undervoltage on the bus, immediately following this step, the normal offsite power feeder breaker (2A309) to the >
4160 VAC ESF bus unexpectedly opened resulting in loss of power to bus 2A3. Sinct the handswitch for 2K4A was in a pull i o-lock position, the EDG did not autcoatically start to provide electrical power for 2A3 and its associated loads. When power was lost to 2A3, a Low Pressure Safety injection-(LPSI) pump (BP P), which was supplying the shutdown cocling system (SDC) for decay heat removal,
, anti a operating Service Water (SW) pump (Bl*P) deenergized as designed.
[ Operations personnel restored SDC flow in approximately one minute by starting a standby LPS.1 pump powered from the redundant 4160 VAC ESF electrical bus. Bus 2A3 was reenergized in approximately three minutes from offsite power by reclosing breaker 2A309 and ths SW pump restarted.
An evaluation of the event was conducted and it was determined that personnel developing and reviewing the test instructions failed to recognize that auxiliary relay 27-1X/285 was also used to automatically open the offsite power feeder breaker (2A309) to bus 2A3 and therefore, performance
' of the test as written would result in d*energi2ing the electrical bus.
C. Safety Significance Electrical bus 2A3 deenergized as designed when the undervoltage condition was simulated on 2B5.
Since the plant had been shutdown for several days prior to this event, core decay heat levels were low. The unexpected loss of decay heat removal flow did not result in any significant increase in RCS temperature or pressure.
4 %
Ftre 3062.01B NRC F: m 360A U.S. Nuclear R:gulatory Commission j9-83)' Approved OMB No. 3150-0104
, Expires: 8/31/85 LICENSEE EVENT REPORT (LER) TEXT CONTINUATION
~
FACILITY KAME (1) IDDCAE1 NUMBER (2) ! LfR NUMEER (6) l FAGE (3) l l l 5eouentiell (Revision l Arkantes Nuclear One Unit Two I i Year Number Number i 101510101013161 Bill 9 --
01 ?l 2 --
06 Pl01310F1013 TEXT (If more space is required, use additional hRC Form 366A's) (17)
The loss of the Service Weter pump resulted in the associated EDG being inoperable until the pump '
was returned to service. However, the EDG was already prevented from automatically starting by placing the EDG start handswitet in pull to-lock as directed by the test in the job order.
As a result of this event, there were no significant safety concerns.
D. Root Cause The root cause of this e'. e r wcs determined to be inadequate post maintenance test controls. The steps necestery to perfors, the post maintenance testing were determined by engineering personnel and it was decided to implement the test by including the test instructions in the maintenance job order used to correct the wiring problem. A fomal technical review (e.g. , independent review) of the testing steps was not performed prior to performing the activity.
E. Basis for Reportability This event is being reported pursuant to 10CFp.50.73(a)(2)(iv), as an ursplanned actuation of an ESF system.
A 10CFR50.72(b)(2)(11) notification to the NkC was made on November 17, 1989 at 0050 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br />.
F. Corrective Actions A test switch located in the electrical trip circuit for 2A309 should have been opened to preclude the inadvertent tripping of the breaker and loss of power to 2A3. The job order instructions were changed to include this test switch and the test was performed satisfactorily with no further unexpected events.
An interim memorandum has been prepared to require an impact statement on significant post main-tenance tests implemented by job orders. The memorandum was issued on December 19, 1989. A procedure revision and training on the revision was completed by January 15, 1990.
An evaluation of the controls of the post maintenance testing that are currently in place has been performed and has resulted in the identification of more extensive procedure revisions than originally anticipated. A two phase plan to implement enhancements to plant procedures has been developed. The first stage will identify and revise the applicable station administrative pro-cedures. The second phase will be to develop a post maintenance testieg guideline. The purpose of the guideline is to aid a planner in developing work packages and to provide consistency in post maintenance testing.
As actions have been taken to implement the two phase plan, it has been identified that additional procedures and training will be necessary to ensure a satisfactory program is established. The pro-cedures, guidelines and training for the post maintenance testing program will be completed and the program implemented by July 31, 1990. After implementation of the program, each corrective mainten-ance job order which is issued for 'Q', 'S' or 'F' components will be evaluated for post maintenance testing guidelines using the approved plant procedure " Control of post-Maintenance Testing".
Additionally an evaluation of the current preventative maintenance procedures will be performed to assess the need for testing enhancements. This evaluation will be completed by October 1, 1990, G. Additional Information A similar event due to inadequate work controls was reported in LER 50-313/88 023 00.
The 10CFR50.72 notification was not made in a timely manner following occurrence of the event due to a lack of understanding that the unplanned opening of the offsite power feeder breaker to the 4160 VAC ESF electrical bus should be considered an ESF actuation. Prior to this event, Arkansas Fower and Light Company did not consider that the actuation of this auxiliary relay should be considered en ESF actuation and was, therefore, not reported under 10CFR50.72 or 10CFR50.73 reporting criteria.
To enhance the current process used in evaluating events or plant conditions for r9 portability, additional guidance has been provided by Plant Licensing to Operations personnel regarding 10CFR50.72 notifications. Additionally, training will be provided for Operations personnel on 10CFR$0.72 reporting criteria. This training will be ccepleted by June 22, 1990.
Energy Industry Identification System (Ells) codes are included in the text as [XX).
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