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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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16E01.N43103 July 2, 1990 1 CAM 079001 U. S. Nuclear Regulatory Commission Document Control Desk Mail Station P1-137 Washington, D. C. 20555
SUBJECT:
Arkansas Nuclear One - Unit 1 Docket No. 50-313 License No. DPR-51 Licensee Event Report 50-313/90-004-00 1
Gentlemen:
In accordance with 10CFR50.73(a)(2)(i)(B), attached is the subject report concerning a degraded fire barrier penetration as the result of personnel oversight and procedure inadequacy.
Very truly yours, ff)
~~
E. C. Ewing
~~b General Manager, Technical Support and Assessment ECE/DBS/sgw Attachment cc: Regional Administrator Region IV U. S. Nuclear Regult.itory Commission 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas 76011 '
INPO Records Center
. Suite 1500 1100 Circle 75 Parkway Atlanta, GA 30339-3064 9no7opoo45 900702 ADOCK C3000313
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. Form 1062.01A NRC Fom 366 U.S. Nuclear Regulatory Commission (6 89) Approved OMB No. 3150 0104 Expires: 4/30/92 L1C(NS(( [V[NT REPORT (L t R)
FA;1LITt MME (1) Arkansas huclear One, Unit One IDOCMT NJMBER (2) IPAGE (3) 10151010101 31 Il 3!110F1013 TITLE (4) Degracea Fire Barrier Fenetration as the Result of Personnel Oversight and Procedural Inadequacy EVENT DATE (5) ,ER NJISER (6? REPORT DATE (7) OTHER F ACILITIES JNVOLVED (6) l I Jequentiali, I Revision l l Month Day Year Year '
Number Number IMonthi Day Year Facility Names Docket Number (sh l l 0 5 0 0 0 01 5 31 1 91 0 91 0 1--1 0 1 01 a '~
. ; 0 ! 0_ l 01 71 Of 2 91 01 0 5 0 0 0 OPERAING . lTn 5 REPORT 15 50BMITTED PUR5UANl TO THE REQVIREMENT5 0F 10 CFR 5 MDDE (9) l NI (Check one or more of the followinL) (11)
POWER l l_I 20.402(b) 6,,,6 20.405(c) l_ l 50.7)(a)(2)(iv) l_l 73.71(b)
LEVELI 1 20.405(a)(1)(1) l.,,,,1 50. 36(c )(1) 50.73(a)(2)(v) l_I 73.71(c)
(10) 101810 ,,,1 1 20.405(a)(1)(ii) l
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50.73(a)(2)(vii) l
~~i 20.405(a)(1)(iii) 7l 50.36(c)(2)
~ 50.73(a)(2)(1)
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l 50.73(a)(2)(viii)(A)l_l Other (Specify inAbstract below and
~ l 20.405(a)(1)(tv) - $0.73(a)(2)(ii) 11 50.73(a)(2)(viii)(B)) -In Text, NRC Form l 20.405(a)(1)(v) 50.73Ja)(2)(iii) l~l 50.73(a)(2)(x) 1 366A)
',1CEN5EE C0hTACT FOR THIS LER (12) home Daryll Saulsberry l Telephone Number l Area l Nuclear Safety and Licensing Specialist (Code l 1510111916l41 13111010 COMPLETE ONE LIhE FOR EACH COMP 0hEhT FAILURE DE5CRIBED IN THI5 REPORT (13) l l l l Reportable l l l iReportablel Causelsystem Component Manufacturer to NPR05 Cause System Comporent Manufacturert to NPRD5 I I I I I i 1 1 1 1 1 I I l- t 1 l l I I '
I i 1 1 I I I I 1 I I I I i _1 SUPPLEMENT REPORT EkPECTED (14) l EkPECiED Month Day Year i SUBM15510N l l'1 Yes (if yes, complete fxpected submission Date) (Il No l DATE (15) l l l 1 . I AB5 TRACT (Limit to 14D0 spaces, i.e., approximately fifteen single space typewritten lines) (16)
On May 31, 1990 at 1330, while conducting a fire barrier penetration seal inspection as part of a comprehensive inspection program initiated as part of a Generic Letter 86*10 evaluation, a degraded fire barrier was discovered by personnel within the fire protection group at Arkansas Nuclear One. The deficient seal consisted of a 2 inch metal sleeve through a floor slab and a 1\ it.ch conduit contained within the sleeve. A review of past documentation revealed this condition has existed prior to a general fire barrier inspection walk down conducted in 1983. Since this condition was not identified during this walk down or subsequent Technical Specification surveillances, the root cause of this condition has been determined to be personnel error and oversight regarding incorrect procedure identification ci penetration number 97 0038. Upon discovery of this condition, the corresponding fire detection system was verified operable, a fire watch was posted in accordance with Technical Specification requirements, the fire barrier was sealed, and the applicable fire print and penetration log updated.
In addition, the fire barrier inspection procedure will be revised and a training program will be implemented for fire barrier inspectors. The degraded fire barrier penetration seal is not a significant safety concern considering the fire preventative measures currently available but is reportable pursuant to 10CFR50.73(a)(2)(1)(B), as a condition prohibited by Technical Specifications.
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Form 1062.01B NRC Form 366A U.$. Nuclear Regulatory Commission (609) Approved OMB No. 3150 0104 Expires: 4/30/9' LICENSEE EYENT REPORT (LER) TEXT CONT]NUAT10N FACILITY MME (1) (DOCAEl NJMBER (2) ( ,[R NJ45[R ((O l FAGE (3) l l l , seguential I Revision) l l_ Year Number Number l Arkansas Nuclear One. Unit One 10151010101 31 11 31 91 0 -
0 I 01 4 **
OIO 10ltl0F1013 ItKI (If more space is required, use socitional NRC Form 366A's) (17)
A. Plant Status At the time this condition was discovered. Unit 1 (AND 1) was in power operations at 80 percent.
Reactor Coolant System (RCS) (AB) temperature was 579 segrees Fahrenheit and reactor coolant ;
system pressure was approximately 2155 psig. 1
- 8. Event Description On May 31,1990 at 1330, while conducting a fire barrier penetration seal inspection as part of a j comprehensive inspection progree initiated to ensure installed seals are in accordance with tested 1 configurations or have adequate basis for installation (i.e., Generic Letter 6610 evaluation), j a degraded fire barrier was discovered by personnel within the fire protection group at Arkansas Nuclear One (ANO). The degraded fire barrier consisted of a 2 inch metal sleeve extending approximately 3 inches above the floor slab to approximately 22 inches below the floor slab between the cable spreading room and the solid waste filler storage room. A Ik inch conduit passing through the 2 inch metal sleeve was surrounded t>y an open annulus which dio not contain a fire retardant seal.
The sleeve appears to have been used as an equipment drain line, at one time, with the portion of the drain line extending through the floor slab modified into a sleeve and subsequently utilited for the routing of conduit through the fire barrier. The sleeve and conduit pass through penetration number 97 0038 in room 97. The sleeve passing through the floor slab was surrounded by an adequate seal and was properly identified in the fire barrier inspection procedure. The conduit within the sleeve was not surrounded by a seal nor was it identified in the inspection procedure. Documentation pertaining to installation of the conduit indicates that the conduit was routed prior to a major fire barrier walk down effort, conducted in 1983, which served to fiele Orify the adem .y of penetration seals located in either an NRC required fire barrier or insurant .. ...ed fire barrier. The results of the fire barrier walk down effort were used to supply baseline data for future inspections of fire barrier penetration seals.
C. Root Cause Fire barrier penetration seals inspected during the walk down effort of 1983, including penetration number 97-0038, were either found containing a satisfactory fire barrier seal or were modified to confore with approved fire barrier sealant standards. Historical documentation of penetration t
number 97 0038 indicates that no deficiencies were found with this fire barrief seal during the l 1983 walk down. Since existing documentation indicates that the routing of conduit was performed I prior to the 1983 walk down, the concition should have been identified during these inspections since inspection guidance was available to the inspector. However, the sleeve configuration was such that it could have misled the inspector to overlook the gap and accept the fire barrier penetration as satisfactory. Additionally, several Technical Specification surveillance procedures for fire barrier penetrations have been performed since 1983 and also have friled to identify the deficient fire barrier penetration seal. Therefore, the root cause of this condition has been determined to be personnel error and oversight related to the failure to identify an inadequate fire barrier seal during the 1963 walk down effort. A contributing factor associated with this condition may be attributed to the fact that the sleeve through which penetration number 97 0038 passes was not correctly identified in the procedure used to perform Technical Specification inspections. Technical Specification surveillances were conducted by maintenance personnel; whereas, the current fire barrier penetration seal inspection program is being conducted by the ANO fire protection group.
D. Corrective Actions l
Upon discovery of this condition, the fire detection system for the cable spreading room was verified operable and a fire watch was posted in accordance with Technical Specification requirements, The fire barrier was sealed through a job request initiated to ensure the annulus l between the sleeve and conduit was adequately enclosed. In response to identifying the fire barrier penetration for future inspections, the new fire barrier penetration designation for i
sleeve (97 0127) has been listed on fire print 97-1 and entered in Penetration Log FB 00-L1, This l Should be effective in providing a cue to inform the fire barrier inspector that this penetration
. exists and requires inspection during future inspections.
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Form 1062.01B NRC Fors $66A U.$. Nuclear Regulatory Commission (6 89) Approved OMB ho. 3150-0104 LIC[NS(( IV[NT REPORT (LER) TEXT CONTINUAi!ON 1 iACILITY h8ME (1) lDOCMT NMER (2) l ,[R N M ER (6) l PAGE (3) l l pequentiall Revisiont l LYeer Numbee Number l Arkansas Nuclear One. Unit One 10151010101 31 11 31 91 0 ~ 0 1 01 4 --
0i0 101310Fl013 MT(Ifmorespaceisrequired,usesoditionalNRCForm366A's)(17) '
Additionally, fire barrier inspection procedure (1405.016) will be revised to correctly identify the new designated fire barrier penetration. This action will be completed prior to December 1, 1990. i These actions are in addition to the current fire barrier seal inspection program which is part cf the AND Business Plan (Action D.5.c) scheduled for completion prior to December 31, 1991. The objectives of the assessment program are to verify the physical configuration of Technical $pecification penetration seals, perfore evaluations of seal designs when deviations are identified, develop a data base and procedures for seal configuration sianagement, and the correction of identified deficiencies.
To provide additional guidance to the fire barrier penetration seal inspector on the correct ,
method of inspecting fire barrier penetrations, a training program will be developed addressing .
the identification of deficient conditions. The training program also will present a discussion (
of penetration sealant material and p9ssible conditions rendering particular sealant materials >
deficient. This program is scheduled for development prior to December 1, 1990. '
E. Safety $ignificance This condition has potential cafety significance considering that the deficient fire barrier seal provides protection for the cable spreading room. A fire spreading to the cable spreailing room could result in degraded plant control due to possible conductor damage associated with Control Room instrumentation. The degree of damage to Control Room instrumentation is dependent on the nature and extent of the fire within the cable spreading area. Plant control in the event of a fire in the cable spreading room is addressed through abnormal operating procedure 1203.02.
In actuality, the fire preventative measures currently evallable make the spread of fire in these areas only remotely possible. These measures include a fixed fire detection system in the cable spreading room which provides alarm annunciation in the ANO 1 Control Room, fire suppression equipment in the form of fire extinguishers, fire water hose reels, and an automatically actuated system. Fire Brigade personnel, specifically trained in fire fighting, are available at all times in the unlikely event a fire were to occur. Although the seal was degraded, the availability of .
detection instrumentation, suppression equipment, and Fire Brigade personnel provide adequate protection against fire propagation. Therefore, there is not a safety concern related to the j degraded seal.
- At the ANO site there are approximately 10,000 total penetrations through p1' ant fire barriers. l Approximately 1500 penetrations have already been reverified with only the condition addressed in this report being identified as a deficient fire barrier penetration seal. Therefore, considering the small population of deficient penetrations which have been identified, the safety concerns as they relate to potentially existent conditions are relatively small for the remaining number of seals which have not been inspected at this time.
F. Basis for Reportability This condition is reportable pursuant to 10CFR50.73(a)(2)(1)(B), as a condition prohibited by Technical Specifications.
G. Additional Information f A condition involving an inadequate fire barrier seal in conjunction with personnel related l oversight was previously reported in 50-368/88-018 and 50 368/90-01F 00.
Revisions to this licensee event report may be submitted in the future if additional inadequate fire barrier seals are identified as part of the current fire barrier inspection program.
Energy Industry Identification System (E!!$) codes are identified in the text as (XX).
l