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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
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{ GPU Nuclear, loc.
(- Route 441 South NUCLEAR Post Office Box 480 Middletown, PA 17057-0480 Tel717 944-7621 May 21,1999 1920-99-20269 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Ladies and Gentlemen:
Subje.t: Thw Mile Island Nuclear Station, Unit 1 (TMI-1)
Operating License No. DPR-50 Docket No. 50-289 Additional Information - Technical Specification Change Request No. 279 -
Core Protection Safety Limit This letter provides itemized responses (Attaciunent 1) to the NRC Request for Additional Information dated May 6,1999. Find design verification of the identified values provided in response to Question No.
1 is estimated to be completed by June 11,1999. GPU Nuclear will notify NRC when the design verification is complete.
If m.y additional information is needed, please contact Mr. David J. Distel, Nuclear Licensing and Regulatory Affairs at (973) 316-7955.
Sincerely, 7
G 4<
James W. Lange ach Vice President and Director, TMl
/DJD Attachments: (1) RAI Responses (2) Single Failure Capability Overview of the TMI-l Emergency Feedwater System \
(3) Emergency Feedwater System Schematic - .
A cc: Administrator, Region I TMI-l Senior Pwject Manager TMI-l Senior Resident inspector
.; l File No. 98195 ,,..
o{
o 9905290199 990521 PDR ADOCK 05000289 P PM ,
L- _
\ .
METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER AND LIGHT COMPANY PENNSYLVANIA ELECTRIC COMPANY d/b/a i GPU ENERGY GPU NUCLEAR, Inc.
1 i Three Mile Island Nuclear Station, Unit 1 Operating License No. DPR-50 Docket No. 50-289 Technical Specification Change Request No. 279 Response to Request for Additional Information (RAI)
COMMONWEALTH OF PENNSYLVANIA )
) SS:
COUNTY OF DAUPHIN )
This GPUN Inc. response to the NRC Staff s RAI on Technical Specification Change Request 279 is submitted in support ofLicensee's request to change Appendix A to Operating License No. DPR-50 for Three Mile Island Nuclear Station, Unit 1. All statements contained in this submittal have been reviewed, and all such statements made and matters set forth therein are true and correct to the best of my knowledge.
1 I
BY #
Ece President and Direhor, ThU Sworn and subscribed before me this g/ yo M ,1999.
(
Jnac b ;
/.~ Noh Public g Notarial Seal i Suzanne C. Miklosik, Notary Public
- @22 Mile's"fA R"%
Member, Pennsylvana Associategn of Ncp r<{
L
1920-99-20269 Attachment 1 Page1of5 ATTACHMENT 1
- 1. NRC Ouestion Define specifically with respect to TSCR 279 what the total EFW (emergency feedwater) flow requirement is, assuming 20 percent average tubes plugged in the once-through-steam generators, including the flow breakdown, i.e., how much flow is necessary for decay heat removal, pump recirculation, bearing cooling, instrument inaccuracy, etc. and describe how the flow balance will be assured and maintained over time.
Response
The bounding design basis events for emergency feedwater (EFW) flow requirements that were reanalyzed for Technical Specification Change Request (TSCR) No. 279 are the Loss of Feedwater (LOFW) and Loss of All AC Power.
The specific event, acceptance criteria, and minimum required EFW flow are tabulated below.
Event Acceptance Criteria Minimum Required Notes EFW Flow
- Loss ofFeedwater e Thermal Power <112% 550 gpm total delivered . Single failure leaves (RCPs ON) flow @ 1065 psia 2 of the 3 EFW pumps e RCS < 2750 psig OTSG pressure available to provide flow.
- Pressurizer does not (275 gpm/OTSG) e No anticipatory reactor go solid trips credited.
- Analysisis sensitive to total net flow but not flow distribution between the OTSGs Loss of All AC e Thermal Power <112% 350 gpm total delivered . Only the TDP is available
?ower fl w @ 1065 psia on loss of all AC power e RCS < 2750 psig OTSG pressure e Resultant Dose
<10CFR100
- The accident acceptance criteria were met using these EFW flow rates in the analysis. Lower flows, although not currently analyzed, may also meet the acceptance criteria.
o 1
!~ 4 1920-99-20269
[' Attachment 1
. L Page 2 of 5
)
L The LOFW event assumes that the EFW system delivers 550 gpm total flow to the once-through-steam generators (OTSG) at event pressure. . Assuming a single failure in the 1
- EFW system (See Attachment 2 for single failure discussion), the system must be able to deliver required flow to the OTSGs with 2 of the 3 pumps and only one (1) control valve
.(EF-V-30) open to each generator. All three'of the possible two-pump combinations must .i be capable of delivering the required flow (i.e.,2 motor driven pumps (MDP), the turbine ;
driven pump (TDP) and MDP "A", 'or, the TDP and MDP "B"). I iThe Loss of All AC Power event assumes that the EFW system delivers 350 gpm total flow
- to the OTSGs at the event pressure. With both onisite and off-site AC power unavailable, only the TDP is ~available. Therefore, the TDP must be capable of delivering the required l '
flow (350 gpm total delivered flow). This event must cope with decay heat only, as tiie
- reactor coolant pumps (RCPs) are tripped with the loss of motive power.
Significant differences in EFW flow requirements are due to:
1 Loss of Feedwater Loss of All AC Power e Reactor Coolant Pump Heat e No Reactor Coolant Pump Heat e Reactor trip on resulting high RCS pressure e Immediate R.cactor trip on loss of AC power i The capability of the EFW system to deliver the above design basis flow requirements will !
l continue to be verified as required by the TMI-1 Technical Specifications (once per l refueling interval or whenever the reactor is in cold shutdown for more than 30 days). The !
surveillance test acceptance criteria for the EFW pumps and the method for generating those criteria are as follows:
- EFW Pumo Test Acceptance Criteria The surveillance' test acceptance criteria for each individual EFW pump are developed 4
- from the most limiting flow requirement. The LOFW event (RCPs ON) requires l 550 gpm total EFW flow delivered to the OTSGs at 1055 psia from any two of the three !
l EFW pumps.' The approximate system head requirement with 550 gpm total flow
- delivered to the OTSGs at event pressure is 2662 feet . System head includes the effects of elevation differences, friction and velocity head due to flow and generator pressure. l The system head was established using a hydraulic model benchmarked to actual system l L data to validate' accuracy.
Since two pumps are available to deliver flow, individual pump performance is acceptable ifit develops this approximate value of head while delivering approximately half of the total flow requirement (e.gc, ~ 275 gpm), to the generator. The uncenainties due to instrumentation used b measure head and flow will be included in the values to establish the acceptance criteria used in the surveillance test. )
l e
' ~ Calculation is complete and being design verified.
a.__
1920-99-20269 Attachment 1 -
Page 3 of 5
. EFWTDP Test Acceptance Ciiteria for Loss of All AC Power Additional acceptance criteria were developed for the turbine driven EFW pump based on the flow requirement for the Loss of All AC Power event. The event requires 350 gpm total EFW flow delivered to the OTSGs at 1065 psia. Only the TDP is available and it alone must deliver the flow requirement. The approximate system head requirement at 350 gpm total delivered flow to the OTSGs at event pressure is 2540 feet 4 System head includes the effects ofelevation differences, friction and velocity head due to flow and generator pressure.
The TDP performance was evaluated with respect to the TDP acceptance criteria for both LOFW and Loss of All AC Power events. The evaluation
- determined that the LOFW event is more limiting and, therefore, establishes the operability acceptance criteria for the turbine driven EFW pump.
TMI-l is currently revising the EFW refueling interval surveillance test procedure for the upcoming 13R outage (Fall 1999) to incorporate the new pump acceptance criteria. The surveillance test will operate each pump individually, injecting water from the condensate storage tanks into a single OTSG. The plant will be at cold shutdown with the OTSG depressurized during the test. Flow to the steam generators will be throttled until total flow delivered to the generator is at least the required rate (~ 275 gpm) including instrument uncertainty.
- Pump discharge pressure measured at pressure instrumentation locations downstream of the pump recirculation and bearing / seal cooling water branch lines, combined with pump suction pressure, will produce a conservative total dynamic head (TDH). EFW pump TDH acceptance criteria will assure the pump adequately meets the design basis head requirements including instrument uncertainty.-
- Analysis of past EFW system test results show that all pumps wil! meet the above i acceptance criteria. Test data also shows that the performance of all three pumps is essentially unchanged since initial installation in 1974.
The test acceptance criteria described above are based on total EFW flow delivered to the OTSGs. Changes in the pump flow breakdown (recirculation and bearing / seal cooling) will not affect operability provided the pump produces adequate head while deliverina required flow to the OTSGs. The surveillance testing is representative of normal system conditions, in that bearing / seal cooling and pump recirculation will operate normally during the test (i.e., are not isolated or throttled to improve test results). Therefore, using pump head and delivered flow alone as test acceptance criteria is appropriate to verify operability even if the pump flow distribution changes over the life of the pir.nt. I
- Calculation is complete and being design verified.
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1920-99-20269 Attachment 1 Page 4 of 5 As good engineering practice, additional information (e.g., total pump flow) will be recorded during refueling interval testing to assess the overall condition of the pumps. This l data allows trending of changes in recirculation or bearing / seal cooling flows. However, flow balancing of the EFW system is not performed, and branch line flows do not afTect the
' EFW pump acceptance criteria. The OTSG delivered flow is the key parameter (vice total pump flow) to which pump operability is appropriately verified to deliver required flow and total dynamic head.
Quarterly surveillance testing, as required by TMI-l Technical Specifications, operates each pump on recirculation and is used to monitor the EFW pumps for changes in recirculation and bearing / seal cooling flows between refueling outages.
- 2. NRC OuestioD
. Assuming maximum allowed pump degradation (for all three EFW pumps) as permitted by the ASME Code, compare the established EFW flow requirement above with the EFW flow that will !
be delivered assuming the worst-case pump combination (assuming single failure). j i
Response
As recommended by NUREG 1482, TMI-l will take action before the pumps degrade beyond the most restrictive ofeither ,
1 (1) the operability test acceptance criteria described in response to Question 1 above, or ;
(2) 10% degradation allowed by ASME Section XI and OM-6.
EFW pump degradation has been evaluated and determined to be limited by the test acceptance '
criteria based on the design basis event required flow and TDH. Allowing 10% degradation per i ASME Section XI and OM-6 would not produce acceptable pump performance for operability. -l Therefore, TMI-l will maintain the appropriate operability acceptance criteria that are more conservative than code limits for the EFW pumps.
- 3. NRC Ouestian l In responding to Question 1 of the NRC's letter dated March 10,1999, regarding TMI-l TSCR 279, you discussed the loss of feedwater accident and stated in a letter dated March 26,1999, that the EFW system was conservatively assumed to deliver flow (550 gpm) to the steam generators starting 43 seconds after the initiation signal with any (emphasis added) combination i ofEFW pumps (1 TDP or 2 MDPs or 1 TDP and 1 MDP). However, the information you
. provided during the April 23,1999, meeting with the staffindicated that the maximum predicted flow capability of the TDP is 536 gpm. Please either confirm the capability of the TDP to deliver the required 550 gpm flow, or amend your response to Question 1 to clarify the correct combination of pumps necessary to achieve the required 550 gpm flow.
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'1920 9t)-20269 Attachment 1 Page 5 of 5 This information is necessary for the staff to accurately understand and evaluate the TMI-l EFW pump capability as it applies to the flow requirement of 550 gpm total flow to the once-through-steam generators as stated in your March 26,1999, letter.
Response
l The GPU Nuclear letter to the NRC dated March 26,1999 (1920-99-20088), discussed EFW flow rates, initiation and delivery time and pump combinations in conjunction with the LOFW event. The LOFW analysis uses 550 gpm EFW flow to the steam generators beginning 43 seconds after the low OTSG level initiation signal. In this analysis, all EFW pumps are l bounded by the 43 second delay assumption. As such, the analysis does not rely upon any specific pump individually, or in combination, to deliver 550 gpm steam generator flow. . EFW flows actually beginning prior to the assumed 43 second delay or in excess of 550 gpm would benefit the event and are therefore not credited.
The GPU Nuclear March 26,1999, response cited the TDP alone as a possible EFW pump combination, assuming it could deliver 550 gpm to the OTSGs during the LOFW event.
However, the EFW system design basis does not require the TDP alone to provide the required system flow for the .LOFW event, as discussed in the response to Question 1 above.
Information presented during the April 23,1999, NRC meeting correctly describes the EFW system design basis as withstanding a single active failure. As such, system function relies upon the TDP to operate as a single pump only in the Loss of All AC Power event. No reliance has been placcd on the TDP alone to satisfy required EFW system flow for any other design basis event. This discussion clarifies the information previously provided in the GPU Nuclear letter dated March 26,1999.
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Attachment 2 Page1of2
, ATTACHMENT 2 Single Failure Capability Overview of the TMI-l Emergency Feedwater System Instrumentation and Control The Emergency Fcedwater System (EFW) is automatically initiated and controlled by the single failure proof Heat Sink Protection System (HSPS). Each of two actuation trains (A and B) are separately powered from diverse supplies initiated with 2 out of 4 channel logic. Four (4) instrument channels are powered from diverse IE supplies.
The worst case active failure (entire HSPS Train) affects only one pump and one flow control valve in each flow path. The system meets the following criteria:
- 1. No single failure will prevent the system design function. Sufficient redundancy to neither cause unwanted actuation nor prevent necessary system operation.
- 2. Mechan :nally and electrically separated so that a fault in one channel will not affect another.
HSPS Train A HSPS Train B Initiates MDP EF-P-2A Initiates MDP EF-P-2B Opens MS-V-13A & B starting the TDP EF-P-1 Opens MS-V-13A & B starting the TDP EF-P-1 Transfers flow control valves EF-V-30A & C from Transfers flow control valves EF-V-30B & D from 0" Startup Range (SU) to 25" SU/50% Operating 0" Startup Range (SU) to 25" SU/50% Operating Range OTSG level setpoint. Range OTSG level setpoint.
Mechanical Desian The EFW System was modified to make it single failure proof and safety grade including:
- 1. Parallel EFW flow control valves (EF-V-30C & D) to the system
- 2. Removed low OTSG pressure isolation signal to the flow contro? valves
- 3. Cavitating venturi in the injection line to each OTSG
- 4. Locked open pump recirculation lines Mechanical components in the system required to change state are:
- 1. Three pumps and associated drivers (EF-P-1, EF-P-2A/B),
- 2. Parallel flow control valves (EF-V-30A/B/C/D), and,
- 3. Steam admission and pressure regulating valves (MS-V-13 A/B, MS-V-6).
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- Attachment 2 Page 2 of2 The failure of any one mechanical component would only result in either:
- 1. . Loss of one EFW pump .
- 2. Loss of one flow control valve associated with one OTSG
- 3. Failure of MS-V-6, causing a loss of the turbine driven pump
- 4. Failure of one MS-V-13 valve, causing a delay in the start of the TDP
- Electrical Desian -
Power supplies for the significant components of the EFW system are shown on Attachment 3.
The motor driven pumps are powered from separate' emergency diesel backed IE switchgear.
The IC ES power supply provides power to both MS-V-2A and MS-V-2B. However, the valves are normally open and are not required to change state to perform their safety function.
Therefore, no single electrical failure will prevent performance of the EFW system safety function.
' Environmental Considerations The EFW system function considers the consequences of design basis events. These consequences include pipe whip, jet impingement and harsh environment (pressure, temperature, radiation and humidity).
The EFW system is not subjected to pipe whip orjet impingement effects from high energy line breaks or LOCAs. The system function continues in the post-LOCA radiation environment.
. Instruments located inside the Reactor Building have been environmentally qualified for LOCA and HELB temperature and pressure conditions. HSPS equipment located in the' control building is not subject to the harsh temperature or pressures resulting from LOCA or HELB events.-
EFW pumps and valves are located in the Intermediate Building. This equipment is not subjected to harsh temperature or pressures caused by LOCAs or HELBs in the Reactor Building. EFW equipment in the Intermediate Building has been environmentally qualified for the worst-case HELB environment (steam line break). The Intermediate Building has available volume to provide sufficient time for the operator to detect and isolate a feed line break in the building b.Gre the flood level can affect EFW system operation.
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