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Category:CORRESPONDENCE-LETTERS
MONTHYEARML20217M0721999-10-19019 October 1999 Forwards NRC Rept Number 17, Requal Tracking Rept from Operator Licensing Tracking Sys.Rept Was Used by NRC to Schedule Requalification Exam for Operators & Record Pass Dates ML20217D8361999-10-11011 October 1999 Provides NRC with Summary of Activities at TMI-2 During 3rd Quarter of 1999 ML20217F8271999-10-0707 October 1999 Forwards Pmpr 99-13, CNWRA Program Manager Periodic Rept on Activities of CNWRA, for Fiscal Reporting Period 990828- 0924.Diskette Containing Pmpr in Wordperfect 8 Is Encl. All Variances Are Expressed with Regard to Current Plans ML20212L1831999-10-0101 October 1999 Responds to Recent Ltr to Wd Travers Expressing Concerns Re Millstone NPPs & Continued Lack of Emergency Mgt Plan for Eastern Long Island.Nrc Continues to Monitor Performance of Millstone to Ensure Adequate Protection to Public Health ML20212L2081999-10-0101 October 1999 Responds to Recent Ltrs to President Wj Clinton,Chairman Jackson & Commissioners & Wd Travers,Expressing Concerns Re Millstone NPPs & Continued Lack of Mgt Plan for Eastern Long Island.Nrc Continues to Monitor Plant Performance ML20212K1241999-10-0101 October 1999 Responds to Recent Ltrs to Chairman Jackson,Commissioners & Wd Travers,Expressing Concern Re Millstone Npps.Nrc Continues to Monitor Performace of Millstone to Ensure That Public Health & Safety,Adequately Protected ML20212L1971999-10-0101 October 1999 Responds to Recent Ltr to Chairman Jackson & Commissioners Expressing Concerns Re Millstone NPPs & Continued Lack of Emergency Mgt Plan for Eastern Long Island.Nrc Continues to Monitor Plant Performance to Ensure Public Health & Safety ML20212J9991999-10-0101 October 1999 Responds to Recent Ltr to President Clinton,H Clinton, Chairman Jackson &/Or Wd Travers Expressing Concern Re Millstone Npps.Nrc Continues to Monitor Performance of Plant to Ensure That Public Health & Safety Adequately Protected ML20212L0061999-10-0101 October 1999 Discusses GL 97-06 Issued by NRC on 971231 & Gpu Response for Three Mile Island .Staff Reviewed Response & Found No New Concerns with Condition of SG Internals or with Insp Practices Used to Detect Degradation of SG Internals ML20212L2171999-10-0101 October 1999 Responds to Recent Ltr to President Wj Clinton,Chairman Jackson & Commissioners,Wd Travers & Ferc,Expressing Concerns Re Millstone NPPs & Continued Lack of Emergency Mgt Plan for Eastern Long Island ML20212K8771999-09-30030 September 1999 Informs of Completion of mid-cycle PPR of Three Mile Island on 990913.No Areas Identified in Which Licensee Performance Warranted Addl Insp Beyond Core Insp Program.Provides Historical Listing of Plant Issues & Insp Schedule ML20212K8551999-09-30030 September 1999 Informs That During 990921 Telcon Between P Bissett & F Kacinko,Arrangements Were Made for Administration of Licensing Exams at Facility During Wk of 000214.Outlines Should Be Provided to NRC by 991122 ML20216J6581999-09-28028 September 1999 Provides Info as Requested of Licensees by NRC in Administrative Ltr 99-03, Preparation & Scheduling of Operator Licensing Exams ML20212J0011999-09-27027 September 1999 Forwards Insp Rept 50-289/99-07 on 990828.No Violations Noted ML20212E1971999-09-16016 September 1999 Forwards Rev 11 of Gpu Nuclear Operational QAP, Reflecting Organizational Change in Which Functions & Responsibilities of Nuclear Safety & Technical Support Div Were Assigned to Other Divisions ML20212A2101999-09-13013 September 1999 Forwards Rev 3 of Gpu Nuclear Post-Defueling Monitored Storage QAP for Three Mile Island Unit 2, Including Changes Made During 1998.Description of Changes Provided on Page 2 ML20216G4151999-09-0909 September 1999 Forwards Pmpr 99-12, CNWRA Program Manager Periodic Rept on Activities of CNWRA, for Fiscal Reporting Period 990731- 0827.All Variances Expressed with Regard to Current Operations Plans ML20211M5861999-09-0202 September 1999 Forwards non-proprietary & Proprietary Response to NRC 990708 RAI Re TS Change Request 272,reactor Coolant Sys Coolant Activity.Proprietary Encl Withheld ML20211M6591999-09-0101 September 1999 Forwards Errata Page to 990729 Suppl to TS Change Request 274,to Reflect Proposed Changes Requested by . Page Transmitted by Submitted in Error ML20211L2401999-09-0101 September 1999 Submits Response to NRC AL 99-02, Operator Reactor Licensing Action Estimates ML20211H3731999-08-27027 August 1999 Responds to NRC 990810 RAI Re TMI LAR 285 & TMI-2 LAR 77 Re Changes Reflecting Storage of TMI-1 Radioactive Matls in TMI-2 Facility.Revised License Page mark-up,incorporating Response,Encl ML20211H4001999-08-27027 August 1999 Responds to NRC 990810 RAI Re TMI-1 LAR 285 & TMI-2 LAR 77 Re Changes to Clarify Authority to Possess Radioactive Matls Without Unit Distinction.Revised License Page mark-up, Incorporating Response Encl ML20211K2391999-08-23023 August 1999 Forwards fitness-for-duty Performance Data Repts for TMI, Oyster Creek & Corporate Headquarters Located in Parsippany, Nj ML20211H5041999-08-20020 August 1999 Forwards Proprietary & non-proprietary Rept MPR-1820,rev 1, TMI Nuclear Generating Station OTSG Kinetic Expansion Insp Criteria Analysis. Affidavit Encl.Proprietary Rept Wihheld 05000289/LER-1999-007, Forwards LER 99-007-01 Re Increasing Failure Rate of ESAS Relays.Rept Supplements Preliminary Info Re Determination of Root Cause & Long Term Corrective Actions.Changes Made for Supplement Are Indicated in Bold Typeface1999-08-20020 August 1999 Forwards LER 99-007-01 Re Increasing Failure Rate of ESAS Relays.Rept Supplements Preliminary Info Re Determination of Root Cause & Long Term Corrective Actions.Changes Made for Supplement Are Indicated in Bold Typeface ML20211A4261999-08-19019 August 1999 Forwards Insp Rept 50-289/99-04 on 990606-0717.Two Severity Level 4 Violations Occurred & Being Treated as Noncited Violations ML20211H3571999-08-19019 August 1999 Forwards Itemized Response to NRC 990712 RAI Re TS Change Request 248 Re Remote Shutdown Sys,Submitted on 981019 ML20211A3931999-08-12012 August 1999 Requests NRC Concurrence with Ongoing Analytical Approach as Described in Attachment,Which Is Being Utilized by Gpu Nuclear to Support Detailed License Amend Request to Revise Design Basis for TMI-1 Pressurizer Supports ML20210R4691999-08-11011 August 1999 Forwards Update 3 to Post-Defueling Monitored Storage SAR, for TMI-2.Update 3 Revises SAR to Reflect Current Plant Configuration & Includes Minor Editorial Changes & Corrections.Revised Pages on List of Effective Pages ML20210N7601999-08-10010 August 1999 Informs That NRC Staff Reviewed Applications Dtd 990629, Which Requested Review & Approval to Allow Authority to Possess Radioactive Matl Without Unit Distinction Between Units 1 & 2.Forwards RAI Re License Amend Request 285 ML20210N7191999-08-0606 August 1999 Forwards Notice of Partial Denial of Amend to FOL & Opportunity for Hearing Re Proposed Change to TS 3.1.12.3 to Add LCO That Would Allow Continued HPI Operation ML20210L3831999-07-30030 July 1999 Responds to NRC 990617 RAI Re OTSG Kinetic Expansion Region Insp Acceptance Criteria That Was Used for Dispositioning Indications During Cycle 12 Refueling (12R) Outage ML20210K7371999-07-30030 July 1999 Forwards Rev 2 to 86-5002073-02, Summary Rept for Bwog 20% Tp LOCA, Which Corrects Evaluation Model for Mk-B9 non- Mixing Vane Grid Previously Reported in Util to Nrc,Per 10CFR50.46 ML20210L1151999-07-28028 July 1999 Confirms Two Senior Management Changes Made within Amergen Energy Co,Per Proposed License Transfer & Conforming Administrative License Amends for TMI-1 05000289/LER-1999-009, Forwards LER 99-009-00 Re 990626 Event Involving Partial Loss of Offsite Power & Subsequent Automatic Start of EDG 1A.Commitments Made by Util Are Contained in long-term Corrective Actions Section1999-07-22022 July 1999 Forwards LER 99-009-00 Re 990626 Event Involving Partial Loss of Offsite Power & Subsequent Automatic Start of EDG 1A.Commitments Made by Util Are Contained in long-term Corrective Actions Section ML20216D4001999-07-22022 July 1999 Provides Summary of Activities at TMI-2 During 2nd Quarter of 1999 ML20210B8231999-07-21021 July 1999 Forwards Exemption from Certain Requirements of 10CFR50.54(w) for Three Mile Island Nuclear Station,Unit 2 in Response to Licensee Application Dtd 990309,requesting Reduction in Amount of Insurance for Unit to Amount Listed ML20210G9471999-07-15015 July 1999 Forwards Pmpr 99-10, CNWRA Program Manager Periodic Rept on Activities of CNWRA, for Fiscal Reporting period,990605- 0702.Diskette Containing Pmpr in Wordperfect 8 Format Is Also Encl ML20209H9401999-07-15015 July 1999 Forwards Copy of Environ Assessment & Findings of No Significant Impact Re Application for Exemption Dtd 990309. Proposed Exemption Would Reduce Amount of Insurance for Onsite Property Damage Coverage as Listed ML20209G2451999-07-15015 July 1999 Advises That Suppl Info in Support of Proposed License Transfer & Conforming Adminstrative License Amends,Submitted in & Affidavit,Marked Proprietary,Will Be Withheld from Public Disclosure,Per 10CFR2.790(b)(5) ML20212K1701999-07-13013 July 1999 Submits Concerns Re Millstone & Continued Lack of Emergency Mgt Plan for Eastern Long Island.Nrc Should Provide Adequate Emergency Planning in Case of Radiological Accident ML20216D9861999-07-12012 July 1999 Forwards RAI Re 981019 Application Request for Review & Approval of Operability & SRs for Remote Shutdown Sys. Response Requested within 30 Days of Receipt of Ltr ML20209G5861999-07-0909 July 1999 Forwards Insp Rept 50-289/99-05 on 990510-28.No Violations Noted ML20209F2571999-07-0909 July 1999 Forwards Staff Evaluation Rept of Individual Plant Exam of External Events Submittal on Three Mile Nuclear Station, Unit 1 ML20209D8451999-07-0808 July 1999 Forwards Insp Rept 50-289/99-06 on 990608-11.No Violations Noted.Overall Performance of ERO Very Good & Demonstrated, with Reasonable Assurance,That Onsite Emergency Plans Adequate & That Util Capable of Implementing Plan ML20209D6291999-07-0808 July 1999 Forwards Notice of Withdrawal & Corrected TS Pages 3-21 & 4-9 for Amend 211 & 4-5a,4-38 & 6-3 for Amend 212,which Was Issued in Error.Amends Failed to Reflect Previously Changes Granted by Amends 203 & 204 ML20209D5141999-07-0808 July 1999 Forwards RAI Re 981019 Application & Suppl ,which Requested Review & Approval of Revised Rc Allowable Dose Equivalent I-131 Activity Limit with Max Dose Equivalent Limit of 1.0 Uci/Gram.Response Requested within 30 Days 05000289/LER-1999-008, Forwards LER 99-008-00 Re Discovery of Degraded But Operable Condition of RB Emergency Cooling Sys.Condition Did Not Adversely Affect Health & Safety of Public1999-07-0202 July 1999 Forwards LER 99-008-00 Re Discovery of Degraded But Operable Condition of RB Emergency Cooling Sys.Condition Did Not Adversely Affect Health & Safety of Public ML20196J3981999-07-0101 July 1999 Responds to GL 98-01,Suppl 1, Y2K Readiness of Computer Sys at Npps. Y2K Readiness Disclosure for TMI-1 Encl ML20209C1131999-07-0101 July 1999 Forwards Signed Agreement as Proposed in NRC Requesting Gpu Nuclear Consent in Incorporate TMI-1 Thermo Lag Fire Barrier Final Corrective Action Completion Schedule Commitment of 000630 Into Co Modifying License 1999-09-09
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARML20217D8361999-10-11011 October 1999 Provides NRC with Summary of Activities at TMI-2 During 3rd Quarter of 1999 ML20217F8271999-10-0707 October 1999 Forwards Pmpr 99-13, CNWRA Program Manager Periodic Rept on Activities of CNWRA, for Fiscal Reporting Period 990828- 0924.Diskette Containing Pmpr in Wordperfect 8 Is Encl. All Variances Are Expressed with Regard to Current Plans ML20216J6581999-09-28028 September 1999 Provides Info as Requested of Licensees by NRC in Administrative Ltr 99-03, Preparation & Scheduling of Operator Licensing Exams ML20212E1971999-09-16016 September 1999 Forwards Rev 11 of Gpu Nuclear Operational QAP, Reflecting Organizational Change in Which Functions & Responsibilities of Nuclear Safety & Technical Support Div Were Assigned to Other Divisions ML20212A2101999-09-13013 September 1999 Forwards Rev 3 of Gpu Nuclear Post-Defueling Monitored Storage QAP for Three Mile Island Unit 2, Including Changes Made During 1998.Description of Changes Provided on Page 2 ML20216G4151999-09-0909 September 1999 Forwards Pmpr 99-12, CNWRA Program Manager Periodic Rept on Activities of CNWRA, for Fiscal Reporting Period 990731- 0827.All Variances Expressed with Regard to Current Operations Plans ML20211M5861999-09-0202 September 1999 Forwards non-proprietary & Proprietary Response to NRC 990708 RAI Re TS Change Request 272,reactor Coolant Sys Coolant Activity.Proprietary Encl Withheld ML20211M6591999-09-0101 September 1999 Forwards Errata Page to 990729 Suppl to TS Change Request 274,to Reflect Proposed Changes Requested by . Page Transmitted by Submitted in Error ML20211L2401999-09-0101 September 1999 Submits Response to NRC AL 99-02, Operator Reactor Licensing Action Estimates ML20211H3731999-08-27027 August 1999 Responds to NRC 990810 RAI Re TMI LAR 285 & TMI-2 LAR 77 Re Changes Reflecting Storage of TMI-1 Radioactive Matls in TMI-2 Facility.Revised License Page mark-up,incorporating Response,Encl ML20211H4001999-08-27027 August 1999 Responds to NRC 990810 RAI Re TMI-1 LAR 285 & TMI-2 LAR 77 Re Changes to Clarify Authority to Possess Radioactive Matls Without Unit Distinction.Revised License Page mark-up, Incorporating Response Encl ML20211K2391999-08-23023 August 1999 Forwards fitness-for-duty Performance Data Repts for TMI, Oyster Creek & Corporate Headquarters Located in Parsippany, Nj 05000289/LER-1999-007, Forwards LER 99-007-01 Re Increasing Failure Rate of ESAS Relays.Rept Supplements Preliminary Info Re Determination of Root Cause & Long Term Corrective Actions.Changes Made for Supplement Are Indicated in Bold Typeface1999-08-20020 August 1999 Forwards LER 99-007-01 Re Increasing Failure Rate of ESAS Relays.Rept Supplements Preliminary Info Re Determination of Root Cause & Long Term Corrective Actions.Changes Made for Supplement Are Indicated in Bold Typeface ML20211H5041999-08-20020 August 1999 Forwards Proprietary & non-proprietary Rept MPR-1820,rev 1, TMI Nuclear Generating Station OTSG Kinetic Expansion Insp Criteria Analysis. Affidavit Encl.Proprietary Rept Wihheld ML20211H3571999-08-19019 August 1999 Forwards Itemized Response to NRC 990712 RAI Re TS Change Request 248 Re Remote Shutdown Sys,Submitted on 981019 ML20211A3931999-08-12012 August 1999 Requests NRC Concurrence with Ongoing Analytical Approach as Described in Attachment,Which Is Being Utilized by Gpu Nuclear to Support Detailed License Amend Request to Revise Design Basis for TMI-1 Pressurizer Supports ML20210R4691999-08-11011 August 1999 Forwards Update 3 to Post-Defueling Monitored Storage SAR, for TMI-2.Update 3 Revises SAR to Reflect Current Plant Configuration & Includes Minor Editorial Changes & Corrections.Revised Pages on List of Effective Pages ML20210L3831999-07-30030 July 1999 Responds to NRC 990617 RAI Re OTSG Kinetic Expansion Region Insp Acceptance Criteria That Was Used for Dispositioning Indications During Cycle 12 Refueling (12R) Outage ML20210K7371999-07-30030 July 1999 Forwards Rev 2 to 86-5002073-02, Summary Rept for Bwog 20% Tp LOCA, Which Corrects Evaluation Model for Mk-B9 non- Mixing Vane Grid Previously Reported in Util to Nrc,Per 10CFR50.46 ML20210L1151999-07-28028 July 1999 Confirms Two Senior Management Changes Made within Amergen Energy Co,Per Proposed License Transfer & Conforming Administrative License Amends for TMI-1 05000289/LER-1999-009, Forwards LER 99-009-00 Re 990626 Event Involving Partial Loss of Offsite Power & Subsequent Automatic Start of EDG 1A.Commitments Made by Util Are Contained in long-term Corrective Actions Section1999-07-22022 July 1999 Forwards LER 99-009-00 Re 990626 Event Involving Partial Loss of Offsite Power & Subsequent Automatic Start of EDG 1A.Commitments Made by Util Are Contained in long-term Corrective Actions Section ML20216D4001999-07-22022 July 1999 Provides Summary of Activities at TMI-2 During 2nd Quarter of 1999 ML20210G9471999-07-15015 July 1999 Forwards Pmpr 99-10, CNWRA Program Manager Periodic Rept on Activities of CNWRA, for Fiscal Reporting period,990605- 0702.Diskette Containing Pmpr in Wordperfect 8 Format Is Also Encl ML20212K1701999-07-13013 July 1999 Submits Concerns Re Millstone & Continued Lack of Emergency Mgt Plan for Eastern Long Island.Nrc Should Provide Adequate Emergency Planning in Case of Radiological Accident 05000289/LER-1999-008, Forwards LER 99-008-00 Re Discovery of Degraded But Operable Condition of RB Emergency Cooling Sys.Condition Did Not Adversely Affect Health & Safety of Public1999-07-0202 July 1999 Forwards LER 99-008-00 Re Discovery of Degraded But Operable Condition of RB Emergency Cooling Sys.Condition Did Not Adversely Affect Health & Safety of Public ML20196J3981999-07-0101 July 1999 Responds to GL 98-01,Suppl 1, Y2K Readiness of Computer Sys at Npps. Y2K Readiness Disclosure for TMI-1 Encl ML20209C1131999-07-0101 July 1999 Forwards Signed Agreement as Proposed in NRC Requesting Gpu Nuclear Consent in Incorporate TMI-1 Thermo Lag Fire Barrier Final Corrective Action Completion Schedule Commitment of 000630 Into Co Modifying License ML20196J7651999-06-29029 June 1999 Provides Updated Info Re Loss of Feedwater & Loss of Electric Power Accident Analyses to Support TS Change Request 279 Re Core Protection Safety Limit,As Discussed at 990616 Meeting ML20196J7701999-06-29029 June 1999 Forwards LAR 285 for License DPR-50,clarifying Authority to Possess Radioactive Matls Without Unit Distinction,So That After Transfer of TMI-1 License to Amergen,Radioactive Matls May Continue to Be Moved Between TMI-1 & TMI-2 Units ML20209C0391999-06-29029 June 1999 Forwards LAR 77 to License DPR-73,clarifying Authority to Possess Radioactive Matls Without Unit Distinction,So That After Transfer of TMI-2 License to Amergen,Radioactive Matl May Continue to Be Moved Between TMI-1 & TMI-2 Units ML20196G2061999-06-23023 June 1999 Requests That NRC Update Current Service Lists to Reflect Listed Personnel Changes That Occurred at TMI 05000289/LER-1999-006, Forwards LER 99-006-00,providing Complete Description,Extent of Condition & Actions Taken in Association with Determination of Inability of Pressurizer Support Bolts to Meet FSAR Requirements1999-06-23023 June 1999 Forwards LER 99-006-00,providing Complete Description,Extent of Condition & Actions Taken in Association with Determination of Inability of Pressurizer Support Bolts to Meet FSAR Requirements ML20196D2171999-06-17017 June 1999 Forwards Pmpr 99-9, CNWRA Program Manager Periodic Rept on Activities of CNWRA, for Fiscal Reporting period,990508- 0604.New Summary Personnel Table Was Added to Rept Period.Matl Scientist Joined Staff Period ML20196A0431999-06-15015 June 1999 Providess Notification That Design Verification Activities Related to Calculations Supporting Analytical Values Identified in Gpu Nuclear Ltr to NRC Has Been Completed 05000289/LER-1999-004, Forwards LER 99-004-00,re Discovery of Emergency FW Pump Bearing Failure During Performance of Oil Change on 990510. Event Was Determined Reportable IAW 10CFR50.73,since Pump Was Determined to Be Inoperable Longer than TS AOT1999-06-0909 June 1999 Forwards LER 99-004-00,re Discovery of Emergency FW Pump Bearing Failure During Performance of Oil Change on 990510. Event Was Determined Reportable IAW 10CFR50.73,since Pump Was Determined to Be Inoperable Longer than TS AOT ML20212K2541999-06-0808 June 1999 Submits Concerns Re Millstone NPP & Continued Lack of Emergency Mgt Plan for Eastern Long Island.Requests That NRC Provide Adequate Emergency Planning in Case of Radiological Accident ML20212K2671999-06-0808 June 1999 Submits Concerns Re Millstone NPP & Continued Lack of Emergency Mgt Plan for Eastern Long Island.Requests That NRC Provide Adequate Emergency Planning in Case of Radiological Accident ML20195E2751999-06-0404 June 1999 Informs That PCTs & LOCA Lhr Limits Submitted in Util Ltr for LOCA Reanalysis Performed in Support of TMI-1 20% Tube Plugging Amend Request Have Been Revised.Revised PCT & LOCA Lhr Limit Values Are Provided on Encl Table 1 ML20195E3281999-06-0404 June 1999 Forwards Application for Amend to License DPR-50,modifying Conditions Which Allow Reduction in Number of Means for Maintaining Decay Heat Removal Capability During Shutdown Conditions ML20195C5721999-06-0202 June 1999 Forwards Description of Gpu Nuclear Plans for Corrective Actions for 1 H Fire Barriers in Fire Zones AB-FZ-3,AB-FZ-5, AB-FZ-7,FH-FZ-2 & Previous Commitments for Fire Zones CB-FA-1 & FH-FZ-6 ML20207E2561999-05-25025 May 1999 Submits 30-day Written Rept on Significant PCT Change in ECCS Analyses at TMI-1 ML20195B2461999-05-21021 May 1999 Forwards Itemized Response to NRC 990506 RAI for TS Change Request 279 Re Core Protection Safety Limit ML20206R6461999-05-13013 May 1999 Forwards Rev 39 of Modified Amended Physical Security Plan for TMI 05000289/LER-1999-003, Forwards LER 99-003-00, Discovery of Condition Outside UFSAR Design Basis for CR Habitability, Which Was Determined Reportable on 990310.Rept Is Being Submitted Four Weeks Later than Required,Per Discussion with NRC1999-05-0707 May 1999 Forwards LER 99-003-00, Discovery of Condition Outside UFSAR Design Basis for CR Habitability, Which Was Determined Reportable on 990310.Rept Is Being Submitted Four Weeks Later than Required,Per Discussion with NRC ML20206K6301999-05-0707 May 1999 Provides Addl Info Re TMI-1 LOFW Accident re-analysis Assumptions for 20% Average SG Tube Plugging as Discussed on 990421 ML20206H0781999-04-30030 April 1999 Forwards Rev 0 to 1092, TMI Emergency Plan. Summary of Changes Encl ML20206J4811999-04-30030 April 1999 Provides Summary of Activities at TMI-2 During First Quarter of 1999.TMI-2 RB Was Not Inspected During Quarter.Routine Radiological Surveys of Auxiliary & Fuel Handling Bldgs Did Not Identify Any Significant Adverse Trends ML20206E4121999-04-27027 April 1999 Requests That TS Change Request 257 Be Withdrawn ML20206C5211999-04-23023 April 1999 Requests Mod to Encl Indemnity Agreement Number B-64,on Behalf of Gpu & Affiliates,Meed,Jcpl,Penelec & Amergen Energy Co,Llc.Ltr Supersedes & Withdraws 990405 Request Submitted to NRC ML20206C8261999-04-22022 April 1999 Submits Financial Info IAW Requirements of 10CFR50.71(b) & 10CFR140.21 1999-09-09
[Table view] Category:UTILITY TO NRC
MONTHYEARML20059H5351990-09-10010 September 1990 Forwards Encls 1-3 of Generic Ltr 90-07 Re Operator Licensing Exam Schedule ML20059G0641990-08-31031 August 1990 Advises That Util Agreed to Revised Frequency of Once Every 12 Months for Corrective Actions Audits Per Tech Spec Change Request 65 Based on 900718 & 19 Discussions ML20059F1691990-08-30030 August 1990 Requests Exemption from Requirements of 10CFR50,App J, Section III.D.1(a) for Facility Re Schedule Requirements for Connecting Type a Testing w/10-yr Inservice Insp Interval, Per 10CFR50.12(a)(2) ML20064A4661990-08-30030 August 1990 Responds to 900803 SALP Rept 50-289/89-99.TMI Does Not Expect to Be Lead Plant for Installation of Advanced Control Sys.Maint Backlog Goals Established.Info on Emergency Preparedness & Engineering/Technical Support Encl ML20059C8791990-08-29029 August 1990 Forwards TMI-1 Semiannual Effluent & Release Rept for Jan - June 1990, Including Executive Summary of Effluent Release Rept,Disposal & Effluent Release Data & Assessment of Radiation Doses.No Changes to ODCM for Reporting Period ML20059D5491990-08-29029 August 1990 Responds to NRC Re Notice of Violation & Proposed Imposition of Civil Penalty Re Personnel Inattentiveness & Failure of Site Managers to Correct Condition.Shift & Immediate Supervisor Discharged ML20059C7851990-08-27027 August 1990 Forwards Rev 5 to Sys Description 3184-007, Solid Waste Staging Facility, Updating Minor Changes to Pages 6,8,9 & 13 ML20059C1091990-08-24024 August 1990 Forwards Rev 6 to Physical Security Contingency Plan.Rev Withheld ML20059B8251990-08-24024 August 1990 Forwards Payment of Civil Penalty in Amount of $50,000,per NRC ML20056B4651990-08-20020 August 1990 Corrects Statement Made in 900716 Response to NRC Bulletin 90-001, Loss of Fill-Oil in Rosemount Transmitters. Identified That Only Half of Operating Crews Provided W/ Briefing on Bulletin ML20058Q1851990-08-17017 August 1990 Requests That Distribution List for TMI-2 Correspondence Be Updated to Be Consistent W/Recently Implemented Organizational Changes.Ee Kintner,Mb Roche & Wj Marshall Should Be Deleted ML20058Q1721990-08-13013 August 1990 Forwards TMI-2 Effluent & Offsite Dose Rept,First Quarter 1990, Update ML20058Q1821990-08-13013 August 1990 Advises That Util Will No Longer Provide Annual Update to Dewatering Sys for Defueling Canisters Sys Description,Per NRC .W/Completion of Defueling & Shipment of All Defueling Canisters Offsite,Sys Has Been Deactivated ML20058M7201990-08-0303 August 1990 Forwards Rev 2 to TER 3232-019, Div Technical Evaluation Rept for Processed Water Disposal Sys. Mods Include Elimination of Pelletizer & Relocation of Druming Station to Discharge of Blender/Dryer ML20055J4581990-07-27027 July 1990 Responds to Violations Noted in Insp Rept 50-289/90-10. Corrective Actions:Missing Support Brace on Cable Tray Support Found & Corrected ML20055J4561990-07-27027 July 1990 Advises That Info Contained in Generic Ltr 90-06,not Applicable to Current Nonoperating & Defueled Condition of Facility.Generic Ltr Will Be Reevaluated,If Decision Made to Restart Facility ML20055H6901990-07-20020 July 1990 Forwards Rev 25 to TMI-2 Organization Plan for NRC Review & Approval.Rev Proposes Consolidation of Plant Operations & Maint Sections Into Plant Operation & Maint Section ML20055G4431990-07-19019 July 1990 Forwards Rev 12 to 990-1745, TMI-1 Fire Hazards Analysis Rept & Update 9 to FSAR for TMI-1 ML20055G8781990-07-19019 July 1990 Discusses Compliance W/Reg Guide 1.97 Re Containment High Range Radiation Monitors,Per 900507-11 Insp.Physical Separation of Power Cables & Required Isolation Will Be Provided to Satisfy Reg Guide Category 1 Requirements ML20055F9601990-07-11011 July 1990 Forwards, 1990 TMI Nuclear Station Annual Emergency Exercise Scenario to Be Conducted on 900912.W/o Encl ML20044A9531990-07-0909 July 1990 Forwards Util Response to Weaknesses Identified in Maint Team Insp Rept 50-289/89-82.Corrective Actions:Engineering Personnel Reminded to Assure Documented Approval Obtained Prior to Proceeding W/Work ML20055E0481990-07-0505 July 1990 Documents Action Taken by Util to Improve Heat Sink Protection Sys & Current Status of Sys.Main Feedwater Logic Circuits Modified Prior to Startup from 8R Outage to Eliminate Potential for Inadvertent Isolation ML20055E0011990-07-0202 July 1990 Forwards Revs 1 & 2 to Topical Rept 067, TMI-1 Cycle 8 Core Operating Limits Rept, Per Tech Spec 6.9.5.4 ML20055C9971990-06-28028 June 1990 Forwards Rev 27 to Physical Security Plan.Rev Withheld ML20055D2071990-06-28028 June 1990 Forwards Certification of TMI-1 Simulation Facility,Per 10CFR55.45.b.5.Resumes of Personnel Involved Encl. Resumes Withheld (Ref 10CFR2.790(a)(6)) ML20055D0861990-06-25025 June 1990 Documents Deviation from Requirements of Reg Guide 1.97,per Insp on 900507-11.Based on Most Limiting Analysis,Existing Range of 0-1,200 Psi Sufficient.Deviation Consistent W/B&W Owners Group Task Force Evaluation of Reg Guide ML20043H4851990-06-18018 June 1990 Forwards Application for Amend to License DPR-50,consisting of Tech Spec Change Request 179 ML20043H4031990-06-18018 June 1990 Responds to Generic Ltr 90-04, Request for Info on Status of Licensee Implementation of Generic Safety Issue Resolved W/Imposition of Requirements or Corrective Actions. ML20043F9921990-06-11011 June 1990 Forwards Listing of Exam Ref Matl Sent on 900601 in Response to 900505 Ltr ML20043F0661990-06-0404 June 1990 Forwards Inservice Insp Data Rept for Period 880816-900304. Owner Rept for Repairs or Replacements Performed on ASME Section XI Class 1 & 2 Components,Also Encl ML20055C9041990-05-23023 May 1990 Advises That App a to Rept Is Set of Recommendations from Safety Advisory Board on Possible Research Opportunities ML20043B2391990-05-18018 May 1990 Revises Commitments in Encl Met Ed 800430 Ltr Re QA of Diesel Generator Fuel Oil.Requirement for QC Review for Acceptability Prior to Filling Diesel Generator Fuel Oil Storage Tanks Deleted from Procedure ML20043A5441990-05-16016 May 1990 Discusses Status of Safety & Performance Improvement Program Portion of B&W Owners Group EOP Review Project ML20043A5311990-05-15015 May 1990 Responds to Violations Noted in Insp Rept 50-289/89-82. Corrective Actions:Periodic Insp Program Established Utilizing Checklist for Stored Equipment & Existing Tool Rooms Will Be Purged of Controlled or Unneeded Matls ML20043A2321990-05-11011 May 1990 Forwards TMI-1 Reactor Bldg 15-Yr Tendon Surveillance (Insp Period 5) Technical Rept 069.Evaluations Conclude That Test & Insp Results Demonstrate TMI-1 Reactor Bldg post- Tensioning Sys in Good Condition ML20042G2741990-05-0404 May 1990 Forwards Semiannual Update of Projects Listed in Categories A,B & C of long-range Planning Program Integrated Schedule ML20012F2621990-04-0202 April 1990 Responds to Violation Noted in Insp Rept 50-289/89-26. Corrective Actions:Util Policy of Shift Supervisor Involvement in Bypassing & Resetting Safety Sys Expanded to Include Shutdown Conditions & Technicians Briefed ML20012F2611990-04-0202 April 1990 Provides Supplemental Response to Station Blackout Rule. Target Reliability of 0.975 Chosen for Emergency Diesel Generators.Diesel Generator Reliability Program May Change Based on Final Resolution of Generic Issue B-56 ML20012F2731990-03-30030 March 1990 Confirms 900328 Conversations & Provides Technical Basis for Planned Actions to Correct Present Power Limitation Due to High Steam Generator Secondary Side Differential Pressure. Main Turbine Will Be Tripped from 80% Power ML20042D8281990-03-23023 March 1990 Fulfills Requirements of Tech Spec Section 4.19.5.a Re once-through Steam Generator Tubes post-inservice Insp Rept for Unscheduled Outage 8U-1 ML20012D7001990-03-22022 March 1990 Forwards Util Response to Generic Ltr 90-01 Re NRC Regulatory Impact Survey.Site Mgt & Staff Hour Categories Added to Response ML20012D7121990-03-21021 March 1990 Forwards Rev 0 to TMI-1 Cycle 8 Core Operating Limits Rept. ML20012C4771990-03-12012 March 1990 Responds to Generic Ltr 89-19, Request for Action Re Resolution of USI A-47, 'Safety Implication of Control Sys in LWR Nuclear Power Plants,' Per 10CFR50.54(f). Current Design Adequate W/O Addl Tech Specs ML20012B8241990-03-12012 March 1990 Forwards Application for Tech Spec Change Request 199 to License DPR-50,revising Tech Specs Re Steam Generator Tube Insp Requirements ML20055C3931990-02-23023 February 1990 Documents Interpretation of Tech Spec 4.19.5.a Re once- Through Steam Generator Tube post-inservice Insp Rept for Refueling Interval 8R.Total of Eight Tubes Removed from Svc by Plugging ML20011F5251990-02-23023 February 1990 Documents Interpretation of Tech Spec 5.3.1.1 Re Design Features of Fuel Assemblies in Light of Issuance of Generic Ltr 90-02.Tech Spec Change Request Re Utilization of Dummy Fuel Rods or Open Water Channels Will Be Filed by 900420 ML20011F6651990-02-22022 February 1990 Forwards Updated Status Summary of Consideration of TMI-1 PRA Recommendations as of 891231.Changes to Torque Switch Settings for DH-V-4A & B Will Be Implemented in Refueling Outage 8 Re Closing Against High Differential Pressure ML20006C2901990-01-26026 January 1990 Provides Addl Info Supporting Deferral of Seismic Qualification Util Group Walkdowns to 10R Outage.Performance of Walkdowns Provide Proper Scheduling & Priority for Resolution of USI A-46 for TMI-1 ML20011E1221990-01-26026 January 1990 Forwards Response to Generic Ltr 89-13, Svc Water Sys Problems Affecting Safety-Related Equipment. Audit Rept Determined That Operation of Decay Heat Closed Cooling Water Sys Consistent W/Design Basis Documents ML19354E8601990-01-25025 January 1990 Requests Approval for Use of B&W Steam Generator Plugs Mfg W/Alternate Matl (nickel-base Alloy/Alloy 600).Alloy 600 Has Superior Corrosion Resistance to Primary Water Stress Corrosion Cracking 1990-09-10
[Table view] |
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o GPU Nuclear Corporation Nuclear o
=ers:r8o Middletown, Pennsylvania 17057-0191 717 944 7621 TELEX 84 2386 Writer's Direct Dial Nurnber:
March 19, 1985 5211-85-2035 Office of Nuclear Reactor Regulation Attn: J. F. Stolz, Chief Operating Reactor Branch No. 4 Division of Licensing U.S. Nuclear Regulatory Comission Washington, DC 20555
Dear Mr. Stolz:
Three Mile Island Nuclear Station Unit 1 (TMI-1)
Operating License No. DPR-50 Docket No. 50-289 Pump and Yalve Inservice Testing (IST)
Your letter of October 23, 1984 provides a supplemental evaluation (SSE) on Pump and Valve Inservice Testing (IST) and your letter of December 11, 1984 concludes that the updated IST Program which GPUN submitted on July 10, 1984 was acceptable with the exception of the relief requests which were denied.
It is the intent of this letter to initiate appeal for further review of certain requests for exemption from ASME Code Section XI test requirements due to impracticality in accordance with 10CFR50.55a(g). GPUN has estimated a cost of $1,500,000 for plant modifications that would be needed in order to perform these additional tests. We request that further review take into consideration these costs and other cost aspects, in addition to the safety benefit to be derived from testing, as can be applied in a " cost / benefit" analysis.
Also, additional relief is requested as described in Attachment 1 (item A.2.b) regarding the pump differential pressure calculation for the Boric Acid Mix Pumps (CA-PIA /B). A revision to the applicable portion of our IST program submittal (Reference 4) is included in Attachment 2.
Most of the IST open items have been resolved. Attachment 1 addresses each of the items included in your SSE and provides additional GPUN commitments for testing which will resolve some of these items. Under each item in Attachment 1 for which " cost / benefit" analysis is needed, we present the technical justification for exemption due to the impracticality of testing along with the estimated cost for plant modifications that would be required to perform the tests which are in question.
8503250012 850319 PDR ADOCK 05000289 giA7 p PDR
'f GPU Nuclear Corporation is a subsidiary of the General Public Utilities Corporation
m 5211-85-2035 March 19, 1985 The largest area of concern deals with additional requirements for isolation valve testing which would require $1,000,000 in plant modifications. In '
regard to the inclusion of additional valves in T.S. Table 3.1.6.1, GPUN does not feel that it is the intent of the regulation that individual elements of the IST Program be included in Technical Specifications. T.S. Section 4.2.2 currently ~ specifies that Inservice Testing of ASME Code Class 1, 2, and 3 pumps and valves be performed in accordance with applicable code and addenda as required by 10CFR50.55a(g), except where relief has been granted.
TMI-1 Technical Specification Section 4.2.2 provides standard language referencing 50.55a(g) and was included in TMI-1 Technical Specifications as requested by NRC in a letter dated November 17, 1976. .GPUN asserts that duplication within the Technical Specifications serves no useful purpose and detracts from the safety impact as described in NUREG-1024, Section 2.2 (Reference 4).
Modification costs are just one aspect of the costs involved. Other aspects should also be considered (i.e., ALARA, lost operating time due to testin exposure' to errors due to additional testing and replacement power costs)g, although it would be difficult to estimate these costs with any degree of accuracy.
Replacement power costs are usually estimated at approximately $500,000 per s day.. Testing, where it is necessary during the transition to cold shutdown, would result in $50,000 per day in the cost of fuel oil. Although ASME Code Section XI (IW-3426) allows leakage limits to be established by the plant l owner, unnecessary valve maintenance would result in driving the personnel exposures and the costs much higher if the acceptance criteria were established arbitrarily below that allowable by plant design in order to be conservative.
Additionally .the performance of the tests which are in question places the plant in a non-routine condition. This results in a certain amount of added risk, and introduces some possibility of maintenance or operator error, i.e.
through valve misalignments, etc. The cost impact as a result of additional risk in performing these tests would be small and difficult to estimate.
However, the benefit gained from testing will probably also be small insofar as the public health and safety is concerned.
In summary, we believe that a " systematic and documented" (cost / benefit) analysis of the relevant and material factors would show that the subject isolation valve testing would not provide a substantial increase in the overall protection of the public health and safety, and any requirement to add additional valves to T.S. Table 3.1.6.1 is not needed since inservice testing of ASME code valves is already required by T.S. Section 4.2.2. The Proposed Rule, " Revision of Backfitting Process for Power Reactors" states that all of
5211-85-2035 March 19, 1985 those factors for such a systematic and documented analysis described therein are currently in use by the Commission and are addressed in the Regulatory Analysis Guidelines of the U.S. Nuclear Regulatory Commission (NUREG-BR-0058),
SECY-83-321 (NRC Manual), and in the CRGR Charter. Therefore, in the absence of a final rule on "Backfitting," we request in this appeal that those factors of such an analysis be applied.
Sincerely,
. D. Ikill Director, TMI-1 HDH/MRK/spb Attachment cc: Joel Page J. Thoma R. Conte
References:
(1) Letter from J. Stolz to H. Hukill dated October 23, 1984.
(2) Letter from J. Stolz to H. Hukill dated April 20, 1981.
(3) Letter from J. Stolz to H. Hukill dated December 11, 1984.
(4) Letter from H. Hukill to J. Stolz dated July 10, 1984.
(5) NUREG-1024 Technical Specificatic..s - Enhancing the Safety Impact, November, 1983.
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Attachment 1 PUMP AND VALVE INSERVICE TESTING (IST) - OPEN ITEMS Item designations correspond to those of the Supplement Safety Evaluation Report (Reference 1).
< Item A.1 a) SW-P2A/B (Screen House Ventilation Equipment Pum AH-P3A/B (Control Building Chilled Water Pumps) ps) i Flow measuring instrumentation for SW-P2A/B and AH-P3A/B will be added prior to Cycle 7 startup. GPUN will continue to record and evaluate inlet pressure, differential pressure, and bearing vibration (pump in-board vibration for AH-P3A/B and motor bearing vibration for SW-P2A/B since the SW-P2A/B impellers are submerged below river level). This item is resolved.
Item A.2 a) EF-P1, EF-P2A/B (Emergency Feedwater Pumps)
BS-PIA /B (Reactor Building Spray Pumps)
DH-PIA /B (Decay Heat Removal Pumps)
DC-PIA /B (Decay Heat Closed Cooling Water Pumps) t These pumps will be tested quarterly. This item is resolved. -
b) CA-PIA /B (Boric Acid Mix Pumps)
The supplemental evaluation (SSE) (Reference 1) denies GPUN's previous j request to delete these pumps and associated flowpath valves CA-V177 and i WDL-V361) from the IST Program. The SSE does not, however, reflect the i- relief which is now being requested subsequent to further discussion with NRC staff on this item.
In a conference call with the NRC on May 21, 1984, GPUN understood that a revised test program would be acceptable to resolve this item by addition of WDL-P13A/B (and the associated flowpath valves) to the IST Program.
The Reclaimed Boric Acid Recycle Pumps (WDL-P13A/B) are similar in 1 function to CA-PIA /B in that they can supply concentrated boric acid to
- the RCS. Although ASME Section XI, IWP-1100 states that emergency powered pumps should be included in the IST Program, and WDL-P13A/B are not emergency powered, they are included in the IST Program at NRC request.
The addition of WDL-P13A/B into the program provides further assurance of the capability to supply concentrated boric acid to the Reactor Coolant System in order to meet the intent of the ASME Code Section XI.
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Additionally, since CA-PIA /B are positive displacement pumps, we feel that it would not be meaningful to calculate A P for CA-PIA /B. Therefore, we ,
wish to amend our IST program submittal (Reference 4) to request further relief in accordance with 10 CFR 50.55a(g)(5) as shown in Attachment 2 which provides a revision to pages 1, 7 and 8 of Table A-2 of that program submittal. This change including the justification for relief is indicated by change bars.
Without recirculation capability, the only method of testing these pumps would be to inject into the Reactor Coolant System (RCS). The resulting reactivity changes would affect plant operations adversely and would result in significant volumes of additional radioactive waste. The cost of plant modifications for CA-PIA /B to install the additional recirculation line (including the associated heat trace, flow meter, pressure gauges, etc.) is estimated at $500,000.
It is impractical to test the subject pumps during plant operation. The revised test program, which includes WDL-P13A/B meets the intent of the code. Therefore, the requested relief is justified.
Item B.1 a) CF-V4A/B (Core Flood Discharge Check Valves)
The SSE (Reference 1) states that TMI-1 procedure SP 1300-3T should be 4
revised to specify a 150 psid minimum test pressure for CF-V4A/B. The test method employed by SP 1300-3T, however, pressurizes the upstream side of CF-V4A/B to 460-599 psig while the downstream side of CF-V4A/B is open to atmosphere. This ensures a test pressure greater than 150 psid. -
Therefore, there is no need to change the procedure.
The SSE (Reference 1) states that SP 1300-3T should be revised to indicate that testing must be accomplished during heatup from cold shutdown as required by the Event V Order dated April 20, 1981. The Order for l Modification of License concerning Primary Coolant System Pressure Isolation Valves does not specify that periodic leakage testing necessarily be accomplished during heatup. In accordance with the Order and T.S. 4.2.7, tests conducted during cooldown when valve dissassembly is not involved are considered as valid a test as a test perforined during heatup. Therefore, there is no significant safety benefit to require changing the procedure or further restricting the test conditions.
A failure of CF-V4A/B, which isolate the Core Flood Tanks from the RCS, does not lead to a LOCA outside centainment. Therefore, CF-V4A/B are not of the WASH 1400 Event V configuration and are not included in the Event V Order dated April 20, 1981. Attached to and referenced in the Event V Order is the Technical Evaluation Report (TEP.) Primary Coolant System Isolation Valves dated October 24, 1980, by the NRC's contractor, Franklin Research Center. Page 4 of the TER states that CF-V4A/B is not a valve configuration of concern and that they need not be included in the testing. Therefore, addition of these valves to Technical Specification (T.S.) Table 3.1.6.1. is not justified.
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b) RC-V4, RC-V23 (MOV and Check Valve in the Decay Heat Auxiliary Spray Line)
These valves are only used to spray the pressurizer when RCS pressure is less than 400 psig. During normal plant operation these valves are shut and do not automatically open; therefore, they are entirely passive (RC-V4 is procedurally required to be shut when the RCS is greater than 400 psig). Their leak tightness will be verified by current RCS leakage calculations. The cost for plant modifications to allow individual pressure isolation valve testing of RC-V4 and RC-V23 is estimated to be
$100,000. GPUN concludes that these costs are not warranted based on the increment of safety to be gained by individual valve testing, and the relief which is being requested is therefore justified.
c) HU-V107A/B/C/D MU-V86A/B, MU-V95 (Makeup System Check Valves)
These valves are in the HPI line from the MU Pumps to the RCS. Please note that there are many (at least four) high pressure valves in series in these lines. The MOVs (MU-V16A-D) open when RCS pressure drops below 1600 psig. Failure to perform a pressure isolation function seems unlikely with the number of redundant valves involved. Figure 1 and the sixth paragraph in Section 3.2 of the Franklin Research Center TER (Reference 2 Attachment 2) do not support an Event V configuration for these valves.
The HPI check valves were disassembled in 1980 to correct manufacturing l quality assurance problems. TMI-1 has not operated since that time.
l Therefore, there is reasonable assurance that these valves can perform a pressure isolation function.
The cost for modification to allow pressure isolation testing is estimated to cost $250,000 for MU-V107A/B/C/D and $350,000 for MU-V86A/B, MU-V94.
GPUN concludes that $600,000 modification costs are not warranted based on the increment of safety to be gained by individual valve testing, and the relief which is being requested is therefore justified.
i d) DH-V1, DH-V2 (Decay Heat Removal System Pump Suction Valves from Loop B Hot Leg)
These valves are two motor operated valves in series; and therefore, do not exhibit Event V valve configuration per Figure 1 of the Franklin Research Center TER. Therefore, there is no demonstrated need to add these valves to T.S. Table 3.1.6.1. In addition, when the RCS is greater than or equal to 400 psig, these valves are closed by procedure and are individually interlocked closed. Therefore, these valves are passive.
The cost of modification to allow individual valve pressure isolation testing is estimated to cost $75,000.
DH-V1 and DH-V2 were disassembled during 1976 and 1983 (see GPUN letter dated June 13, 1984 for the details on maintenance performed). Disc and seating surfaces were found satisfactory in both instances. GPUN concludes that there is reasonable assurance that DH-V1 and DH-V2 can Al-3
perform a pressure barrier function and that modification costs are not warranted based on the increment of safety to be gained by individual leak testing of DH-V1 and DH-V2 and the relief which is being requested is therefore justified.
Item B.2 - Full Stroke Testino of Check Valves a) CF-V4A/B (Core Flood Tanks Discharge Check Valves)
CF-V4A/B and CF-VSA/B are the same size, model number, and manufacturer.
If disassembly of CF-V4(A or B), or CF-V5(A or B) shows a degraded condition which would make its full stroke capability questionable, the remaining three valves will be disassembled at the same outage. This is incompliancewiththeSSE(Reference 1). Therefore, this item is resolved.
b) C0-V16A/B, EF-VilA/B, EF-V13 EF-V12A/B (Emergency Feed Pump Suction and Discharge Check Valves)
Relief was granted. Therefore, this item is resolved.
c) DH-V14A/B, DH-V16A/B (Decay Heat Removal Pumps Suction and Discharge Check Valves)
The SSE (Reference 1) states that licensee proposes to part stroke these valves quarterly and perform a full stroke test each refueling utilizing flowrate equal to or greater than the maximum assumed in the Safety Analysis Report. The staff concurs with the testing proposed and grants relief from full stroke testing at quarterly or cold shutdown intervals.
GPUN wishes to point out that the TMI-1 IST Program (Reference 3) includes tests for full flow through DH-V16A/B and 73% opening of DH-V14A/B. Our request for relief therein states that a full stroke test of DH-V14A/B is not practical, since it would necessitate the spray down of the entire Reactor Building to achieve full flow and further states that 73% opening of DH-V14A/B provides reasonable assurance that the valve will open fully. GPUN interprets the SSE to indicate that 73% opening of DH-V14A/B is accepted as being equivalent to a full stroke test. Therefore, this item is resolved.
d) MS-V9A/B (Main Steam Supply Check Valves to Steam Driven Emergency FeedwaterPump)
These valves supply steam from the OTSGs to the steam driven Emergency FeedwaterPump(EF-P1). During a cold shutdown, it is impractical to stroke test these valves (full stroke or partial stroke) since the steam which would be needed to operate these valves is not available during cold shutdown conditions.
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o Full stroke testing of MS-V9A/B is also impractical due to other limitations during plant conditions when steam is available. EF-P1 must be tested using the recirculation line to the Condensate Storage Tank
- bypassing the OTSG. This is to prevent degradation of the OTSGs by
- excessive thermal stress cycling of the emergency feedwater nozzles. The i number of thermal cycles on the emergency feedwater nozzles is limited to
- (40) cycles over the life of the plant. Due to the small size of the i recirculation line. EF-P1 cannot be tested at full capacity; and MS-V9A/B will not open fully. Under these restrictions it is only possible to obtain approximately 48% flow which corresponds to about 80% opening of MS-V9A/B.
Plant modifications which would be required to perfonn full stroke tests of MS-V9A/B either by piping in auxiliary steam or by replacing the
- recirculation piping with larger piping capable of recirculating the full 3 EFW pump capacity would introduce exorbitant cost. GPUN has not fully examined the cost and safety impact of modifications which would be required to test MS-V9A/B, however, we do not feel that such modifications would be beneficial.
MS-V9B was disassembled for IST examination purposes in late 1984 and found to be in excellent condition. Since no indication of potential degradation was found, this provides additional assurance of the continued capability of MS-V9A/B to open fully when needed.
It is impractical to test MS-V9A/B when steam is not available and it is also impractical to perform a full stroke test on MS-V9A/B. GPUN concludes that quarterly testing of MS-V9 at 48% flow (80% open) when steam is available meets the intent of the ASME Code Section XI and the -
relief which is being requested is therefore justified.
4 e) BS-V21A/B (Reactor Building Spray Pump Suction Check Valves from Sodium Thiosulfate Tanks)
The line from the Sodium Thiosulfate Tank has been cut and blind-flanged.
BS-V21A/B are no longer included in the IST Program. Therefore, this item is resolved.
f) BS-V52A/B (Sodium Hydroxide Tank to Decay Heat Pumps Suction Header Check Valves)
Relief was granted. Therefore, this item is resolved.
g) Fluid 31ock System Check Valves (Li Gate Containment Isolation Valves) quid Supply Check Valves to Selected Until the fluid block system is physically removed. GPUN will either maintain the manual valve in each relevant line administratively closed or
, cut and cap the lines. In either case, the option to use this system will .
not be maintained until the system is removed. Therefore, this item is resolved.
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h) MU-V73A/B/C, MU-V107A/B/C/D (Makeup Pumps Discharge Check Valves to the Cold Legs)
Relief was granted. Therefore, this item is resolved.
- 1) MU-V14A/B (Makeup Pumps Suction Stop Check Valves)
Relief was granted. Therefore, this item is resolved.
j) MU-V94, MU-V95, MU-V86A/B (HPI Cross Connect Piping and HPI Discharge
, CheckValves)
Relief was granted. Therefore, this item is resolved.
Item B.3 a) EF-V3 (Emergency River Water Suction Source Check Valve)
Full stroke testing of EF-V3 each refueling is not practical for the reason that such a test would introduce contaminants into the OTSGs. GPUN is pursuing the feasibility of removing the valve internals for EF-V3 per 10CFR50.59 evaluation. Part stroke testing of EF-V3 will be performed in
, accordance with the IST Program until the valve internals are removed.
This valve was disassembled on 12/5/84 (for IST purposes) and found to be in excellent condition. Therefore, there is reasonable assurance that EF-V3 would open if required.
Item B.4 ,
a) FW-V12A/B (Main Feedwater Check Valves)
GPUN will develop a method to verify the full closure capability of FW-V12A/B before startup from the cycle six refueling outage. Until that time, NRC has agreed that testing of FW-V12A/B will not be required based on disassembly and repair of the valves in 1980. Therefore, this item is resolved for Cycle 5 operation.
Item B.5 a) BS-V30A/B(ReactorBuildingSprayDischargeCheckValves)
If a disassembly / inspection reveals that the full stroke capability of the disassembled valve may be in question, GPUN will disassemble and inspect the other valve at the same outage. This commitment is in compliance with the SSE. Therefore, this item is resolved.
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._~ _ , _ _ . _ _.__._ ______ _ ._._..___ _ _.__ _ _ _ _ _ _ _ _ . . _ _ _ _ _ . _ _ _ _ _ _ _
TABLE A-2 THREE MILE ISLAND - UNIT NO. 1 '
PERIODIC INSERVICE INSPECTION PROGRAM - (PUMPS)
EXCEPTIONS TO ASME XI REQUIREMENTS PAGE 1 of 8 Rev. 1 (3/14/85)
ASME III ASME XI CODE CLASS ASME XI EXCEPTION TESTING PERFORMED IN LIEU PUMP NAME PUMP NO. CODE CLASS EQUIVALENT REQUESTED
- JUSTIFICATION OF CODE REQUIREMENT CONTROL BUILDING AH-P3A 3 Non-Nuclear Q See Note 3 Flow metering will be CHILLED WATER AH-P3B installed prior to startup for Cycle 7.
Tb See Note 13 None Lubr. Level See Note 13 None BUILDING SPRAY BS-PIA 2 N-2 None N/A N/A BS-PIB BORIC ACID CA-PIA 3 Non-Nuclear Quarterly Testing See Note 12 Refueling interval testing CA-PIB Q See Note 12 Q will be caluclated ,
Pi See Note 12 None 2 AP See Note 12 None ]'
I 7' "' Tb See Note 5 None Five Minute runtime See Note 12 Run until system is stable DECAY HEAT DC-PIA 3 Non-Nuclear None N/A N/A CLOSED COOLING DC-PIB WATER DECAY HEAT DH-PIA 2 N-2 None N/A N/A REMOVAL DH-PIB ,,
U 2-DECAY HEAT DR-PIA 3 Non-Nuclear V See Note 1 Motor vibration will be 9 RIVER WATER DR-PIB measured. @
Tb See Note 2 None ti Lubr. Level See Note 2 None n2
- SEE ASME SECTION XI FOR DEFINITION OF TEST QUANTITIES 0013A l
TABLE A-2 -
THREE MILE ISLAND - UNIT NO. 1 PERIODIC INSERVICE INSPECTION PROGRAM - (PUMPS)
EXCEPTIONS TO ASME XI REQUIREMENTS Page 7 of 8 Rev. 1 (3/14/85) i JUSTIFICATION NOTES (Cont.)
be drained and then flushed with Nuclear Service Closed Cooling Water. The drain and flush water is drained to the Reactor Building Sump and this produces large quantities of water that must be processed through the Liquid Waste Disposal System. However, flow rate will be measured during refueling outages, when river water is pumped through the cooling coils in accordance with Technical Specification requirements.
Note 12 For CA-PIA /B and WDL-P13A/B, GPUN requests relief from the measurement of Q, A P, Pj, and the 5 minute run time. In addition, relief is requested to perform testing only during refueling outages.
Without recirculation capability, the only method of testing these pumps is to inject into the Reactor 20 Coolant Makeup System. The resulting reactivity changes would affect plant operations adversely and would 7'
result in significant volumes of radioactive waste. For these reasons it is impractical to test the subject pumps during operation. The appropriate test interval is each refueling.
- In a conference call with NRC on May 21, 1984, the following test program was agreed upon in order to meet the intent of the ASME Code Section XI test requirements:
- 1. Tests of the subject pumps will be conducted each refueling interval.
- 2. Pump differential pressure ( AP) will be calculated, not measured, since pump inlet pressure (Pj) is not available (there are no existing pressure gauge taps).
- 3. Flow rate (Q) will be calculated using tank level change over time, since installed
. flow measuring instuments do not exist.
GPUN proposes to calculate pump differential pressure (AP) for WDL-P13A/B only. Since CA-PIA /B are positive displacement pumps, relief is requested from the ASME Section XI, IWP-3110 requirement to calculate AP for CA-PIA /B. The calculation of AP for CA-PIA /B would not be meaningful since the flow rate is fixed solely by the displacement of the cylinder and the speed of the pump both of which are held constant, while AP is only a function of system resistance (backpressure). Therefore, such a calculation would be an unnecessary exercise for the operator.
0014A
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TABLE A-2 '
THREE MILE ISLAND - UNIT NO. 1 i PERIODIC INSERVICE INSPECTION PROGRAM - (PUMPS)
] EXCEPTIONS TO ASME XI REQUIREMENTS Page 8 of 8 Rev. 1 (3/14/85) ,
JUSTIFICATION NOTES (Cont.)
Tests each refueling to verify the required flowrate for CA-PIA /B while pumping to the Makeup Tank will demonstrate positively that CA-P1A/B can perform its safety function by providing the required boric acid capacity at nomal Makeup Tank backpressures. For this test, the Makeup Tank will be pressurized to its
{ normal operating range. Pump vibration will be measured while pumping to the Makeup Tank and Makeup Tank level change over time will be used to determine the flow rate.
) ASME Section XI IWP-3500 states that pumps under test should be run for at least 5 minutes under conditions j as stable as the system permits prior to taking data. To minimize radioactive waste, the subject pumps will i be run until the system is stablized and then data will be recorded. GPUN believes this meets the intent of
- IWP-3500.
ASME Section XI, IWP-1100 states that emergency powered pumps should be included in the IST Program.
g WDL-P13A/B are not emergency powered, but are included in the IST Program as requested by NRC. The addition "e of WDL-P13A/B in the program provides further assurance of the capability to supply concentrated boric acid j to the Reactor Coolant Makeup System.
j Note 13
- The pump and motor form an integral unit. The pump bearings are located in the motor. There is no i 4
lubrication level on the pump that can be checked. Also yearly bearing temperatures will not be measured
) since the bearings are deep inside the motor end caps.
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l 0014A