05000289/LER-1997-001, Forwards Corrected Copy of LER 97-001-00 Re Condition Found During Review of GL 96-06

From kanterella

(Redirected from 05000289/LER-1997-001)

Jump to navigation Jump to search

Forwards Corrected Copy of LER 97-001-00 Re Condition Found During Review of GL 96-06
Person / Time
ML20136B753
Site:	Crane
Issue date:	02/17/1997
From:	James Knubel GENERAL PUBLIC UTILITIES CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20136B750	List: 05000289/LER-1997-001, Forwards Corrected Copy of LER 97-001-00 Re Condition Found During Review of GL 96-06 05000289/LER-1997-001-02, :on 970117,potential Overpressurization of Piping Between Closed Reactor Building Containment Isolation Valves Occurred Due to Inadequate Design Code Guidance. Results of Review Will Be Documented.Pages 2 & 3 Corrected
References
6710-97-2058, GL-96-06, GL-96-6, NUDOCS 9703110108
Download: ML20136B753 (1)
v • d • e

LER-1997-001, Forwards Corrected Copy of LER 97-001-00 Re Condition Found During Review of GL 96-06

Event date:
Report date:
2891997001R00 - NRC Website
v • d • e

text

_.

1 l

GPU Nuclear. ine.

(

Route M1 South NUCLEAR

'cn 0"'ce Bon *ao

%ddletown, PA 17057-C'80-1 Tel 717 94 7621 CORRECTED COPY February 17, 1997 6710-97-2058 l

U. S. Nuclear Regulatory Conunission Attn: Document Control Desk Washington, DC 20555 l

Dear Sir:

Subject:

Three Mile Island Nuclear' Station, Unit 1 (TMI-1)

Operating License No. DPR-50 Docket No. 50-289 LER 97-001-00 i

The purpose of this letter is to transmit TMI-l Licensee Event Report (LER) No. 97-001-00 regarding a condition found during a review of Generic Letter 96-06, that a design-basis accident could potentially pressurize certam piping segments between containment isolation valves beyon.d stresses allowed by the code of record for TMI-1. The affected piping was found to be operable in that the resulting stresses were within the ASME Code Section III, Appendix F criteria. The abstract provides a brief description of the event.

The event did not adversely affect the health and safety of the public.

Sincerely,

~

J.Knubel Vice President and Director, TMI j

MRK Attachment cc: Administrator, NRC Region I TMI Senior NRC Resident Inspector TMI Senior NRC Project Manager 9703110108 970304 I

PDR ADOCK 05000289 PDR S

Licensee Event Report

70 | 71 | 72 | 73 | 74 | 75 | 76 | 77 | 78 | 79 | 80 | 81 | 82 | 83 | 84 | 85 | 86 | 87 | 88 | 89 |

90 | 91 | 92 | 93 | 94 | 95 | 96 | 97 | 98 | 99 | 00 | 01 | 02 | 03 | 04 | 05 | 06 | 07 | 08 | 09 |

10 | 11 | 12 | 13 | 14 | 15 | 16 | 17 | 18 | 19 | 20 | 21 | 22 | 23 | 24 | 25 |

		Reporting criterion
05000289/LER-1997-001, Forwards Corrected Copy of LER 97-001-00 Re Condition Found During Review of GL 96-06	Forwards Corrected Copy of LER 97-001-00 Re Condition Found During Review of GL 96-06
05000289/LER-1997-001-02, :on 970117,potential Overpressurization of Piping Between Closed Reactor Building Containment Isolation Valves Occurred Due to Inadequate Design Code Guidance. Results of Review Will Be Documented.Pages 2 & 3 Corrected	on 970117,potential Overpressurization of Piping Between Closed Reactor Building Containment Isolation Valves Occurred Due to Inadequate Design Code Guidance. Results of Review Will Be Documented.Pages 2 & 3 Corrected	10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(b)(2)(ii)
05000289/LER-1997-001-01, :on 970117,potential Overpressurization of Piping Between Closed Reactor Building Isolation Valves Occurred Due to Inadequate Design Code Guidance.Results of This Review Will Be Documented	on 970117,potential Overpressurization of Piping Between Closed Reactor Building Isolation Valves Occurred Due to Inadequate Design Code Guidance.Results of This Review Will Be Documented	10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(b)(2)(ii)
05000289/LER-1997-002, :on 970129,potential Inability of Startup Feedwater Block Valves to Fully Close Following Main Steam Line Break Occurred Due to MOV Program Weakness.Increased Torque Switch Settings on Valve Operators	on 970129,potential Inability of Startup Feedwater Block Valves to Fully Close Following Main Steam Line Break Occurred Due to MOV Program Weakness.Increased Torque Switch Settings on Valve Operators	10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(b)(2)(ii)
05000289/LER-1997-003-01, Forwards LER 97-003-01 Re Potential Overpressurization of Makeup Pump Suction Piping,Due to Inadequate Test & Operating Procedures	Forwards LER 97-003-01 Re Potential Overpressurization of Makeup Pump Suction Piping,Due to Inadequate Test & Operating Procedures	10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition
05000289/LER-1997-004, :on 970228,declared Auxiliary & Fuel Handling Bldg Ventilation Sys Inoperable.Cause Related to Wording in TS Bases & Fsar.All Sys Maint,Testing & Design Mods Performed	on 970228,declared Auxiliary & Fuel Handling Bldg Ventilation Sys Inoperable.Cause Related to Wording in TS Bases & Fsar.All Sys Maint,Testing & Design Mods Performed	10 CFR 50.73(b)(2)(ii)
05000289/LER-1997-005-01, :on 970410,declared motor-operated Reactor Bldg Purge Valves Inoperable Due to Inadequate Closing Torque. Caused by Deficient Design Analysis.Will Perform Evaluation of Valve Specific Seating Torque Test Data	on 970410,declared motor-operated Reactor Bldg Purge Valves Inoperable Due to Inadequate Closing Torque. Caused by Deficient Design Analysis.Will Perform Evaluation of Valve Specific Seating Torque Test Data	10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(b)(2)(ii)
05000289/LER-1997-005, Has Been Cancelled	Has Been Cancelled
05000289/LER-1997-006-01, :on 970424,reactor Bldg Fan Motors Unqualified Cable Termination Occurred Due to Incorrect Engineering Guidance.Caused by Personnel Error.Qualifiable Repair of AH-E-1A Fan Motor Termination Was Installed on 970323	on 970424,reactor Bldg Fan Motors Unqualified Cable Termination Occurred Due to Incorrect Engineering Guidance.Caused by Personnel Error.Qualifiable Repair of AH-E-1A Fan Motor Termination Was Installed on 970323	10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2) 10 CFR 50.73(b)(2)(ii)
05000289/LER-1997-006, Forwards LER 97-006-00,resulting from Violation of TS 3.3 & Declaring Reactor Bldg fans,AH-E-1B & C,Inoperable on 970424	Forwards LER 97-006-00,resulting from Violation of TS 3.3 & Declaring Reactor Bldg fans,AH-E-1B & C,Inoperable on 970424
05000289/LER-1997-006-02, Re Unqualified Cable Terminations.Ler 97-006-00 Retracted	Re Unqualified Cable Terminations.Ler 97-006-00 Retracted
05000289/LER-1997-007, :on 970621,generator Output Breaker Failed Resulting in Loss of Offsite Power & Rt.Caused by Failure of Generator Breaker GB-102.Both Failed Breakers Were Replaced	on 970621,generator Output Breaker Failed Resulting in Loss of Offsite Power & Rt.Caused by Failure of Generator Breaker GB-102.Both Failed Breakers Were Replaced	10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition 10 CFR 50.73(a)(2) 10 CFR 50.73(b)(2)(ii)
05000289/LER-1997-008, :on 970621,power to Control Rod Drive Mechanisms Was Lost & Reactor Tripped.Caused by Long Trip Insertion Times at Tmi.Rods Were Trip Tested Per Surveillance Procedure 1303.11.1	on 970621,power to Control Rod Drive Mechanisms Was Lost & Reactor Tripped.Caused by Long Trip Insertion Times at Tmi.Rods Were Trip Tested Per Surveillance Procedure 1303.11.1	10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2) 10 CFR 50.73(b)(2)(ii)
05000289/LER-1997-009-01, :on 970725,engineering Analysis of Loss of a Train DC Power W/Loss of Offsite Power & LOCA Was Determined.Caused by Plant Designers Error.Trained Operators on Methodology to Regain Control of Es Equipment	on 970725,engineering Analysis of Loss of a Train DC Power W/Loss of Offsite Power & LOCA Was Determined.Caused by Plant Designers Error.Trained Operators on Methodology to Regain Control of Es Equipment	10 CFR 50.73(a)(2) 10 CFR 50.73(b)(2)(ii)
05000289/LER-1997-009, Forwards LER 97-009-00,re Condition Which Was Evaluated & Determined to Be Reportable on 970725.Condition Was Identified as Result of Review of Crystal River 3 LER for Applicability at TMI-1	Forwards LER 97-009-00,re Condition Which Was Evaluated & Determined to Be Reportable on 970725.Condition Was Identified as Result of Review of Crystal River 3 LER for Applicability at TMI-1
05000289/LER-1997-010, Forwards LER 97-010-00,concerning Inoperable PORV Resulting from Violation of TS 4.2.2,in that,post-maintenance Test Was Not Performed on PORV Following Replacement During 11R Refueling	Forwards LER 97-010-00,concerning Inoperable PORV Resulting from Violation of TS 4.2.2,in that,post-maintenance Test Was Not Performed on PORV Following Replacement During 11R Refueling
05000289/LER-1997-010-01, :on 971012,PORV Inoperability Was Determined. Caused by mis-wiring & Failure to Perform post-maintenance Test Following Replacement During 11R Refueling Outage.Porv Replaced During 12R Refueling Outage Was Wired Correctly	on 971012,PORV Inoperability Was Determined. Caused by mis-wiring & Failure to Perform post-maintenance Test Following Replacement During 11R Refueling Outage.Porv Replaced During 12R Refueling Outage Was Wired Correctly	10 CFR 50.73(a)(2)(i)

Retrieved from "https://kanterella.com/w/index.php?title=05000289/LER-1997-001,_Forwards_Corrected_Copy_of_LER_97-001-00_Re_Condition_Found_During_Review_of_GL_96-06&oldid=4219370"