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Category:CORRESPONDENCE-LETTERS
MONTHYEARDD-99-12, Informs That Time Provided by NRC Regulation within Which Commission May Act to Review DD-99-12 Has Expired.With Certificate of Svc.Served on 9910281999-10-28028 October 1999 Informs That Time Provided by NRC Regulation within Which Commission May Act to Review DD-99-12 Has Expired.With Certificate of Svc.Served on 991028 ML20217N8261999-10-25025 October 1999 Discusses Errata Re 991021 Filing of Northeast Nuclear Energy Co Answer to Request for Hearing & Petition to Intervene B17901, Submits Statement of Personal Qualification (NRC Form 398) Along with Supporting Certification of Medical Exam by Facility Licensee (NRC from 396) in Support of License Renewal for PM Miner.Encls Withheld,Per 10CFR2.790(a)(6)1999-10-18018 October 1999 Submits Statement of Personal Qualification (NRC Form 398) Along with Supporting Certification of Medical Exam by Facility Licensee (NRC from 396) in Support of License Renewal for PM Miner.Encls Withheld,Per 10CFR2.790(a)(6) B17886, Requests Permission to Utilize Code Case N-623, Deferral of Insps of Shell-to-Flange & Head-to-Flange of Reactor Vessel,Section Xi,Div 1, for Millstone Unit 21999-10-18018 October 1999 Requests Permission to Utilize Code Case N-623, Deferral of Insps of Shell-to-Flange & Head-to-Flange of Reactor Vessel,Section Xi,Div 1, for Millstone Unit 2 05000336/LER-1999-012, Forwards LER 99-012-00,re Unrecoverable CEA Misalignment Entry Into TS 3.0.3 on 990917.Commitments Made by Util Are Encl1999-10-15015 October 1999 Forwards LER 99-012-00,re Unrecoverable CEA Misalignment Entry Into TS 3.0.3 on 990917.Commitments Made by Util Are Encl B17900, Forwards Revised Assumptions Used in Fuel Handling Accident Analysis & Summary of Results.List of Regulatory Commitments Are Listed in Attachment 11999-10-14014 October 1999 Forwards Revised Assumptions Used in Fuel Handling Accident Analysis & Summary of Results.List of Regulatory Commitments Are Listed in Attachment 1 ML20217G9491999-10-14014 October 1999 Forwards Exemption from Requirements of 1-CFR50,App E, Section IV.F.2.c,re Conduct of full-participation Exercise in Sept 1999,at Plant,Units 1,2 & 3 ML20217D9671999-10-12012 October 1999 Forwards Copy of Transcript of Public Meeting Held by NRC Staff & NNECO on 990825 at Waterford,Connecticut on Decommissioning Program for Millstone,Unit 1.Without Encl ML20217D3011999-10-0707 October 1999 Forwards Request for Addl Info Re Util 990118 Request for Amend to License NPF-49 to Allow full-core Offloads to Spent Fuel Pool During Core Offloads to Spent Fuel Pool During Core Offload Events ML20217F0031999-10-0606 October 1999 Forwards Original Petition to Intervene Being Filed on Behalf of Clients,Connecticut Coalition Against Millstone & Long Island Coalition Against Millstone,Iaw Provisions of 10CFR2.714 ML20217P1201999-10-0606 October 1999 Informs NRC of Proposed Acquisition of Parent Holding Company of Central Maine & Requests NRC Concurrence,Based on Threshold Review,That Proposed Acquisition Does Not,In Fact, Constitute Transfer Subject to 10CFR50.80 B17892, Requests Withdrawal of License Amend Application Re 24-month SG Tube Insp Surveillance Extensions,Submitted in Util 950726 & s to NRC1999-10-0505 October 1999 Requests Withdrawal of License Amend Application Re 24-month SG Tube Insp Surveillance Extensions,Submitted in Util 950726 & s to NRC ML20217B4711999-10-0404 October 1999 Informs That Staff Did Not Identify Any Safety Concerns Re Licensee Proposals to Modify Commitments Made for Action Items 4.2.1,4.2.2,4.5.1 & 4.5.2 of GL 83-28 by Providing Addl Justifications or Safety Bases for Changes 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 ML20212J3051999-10-0101 October 1999 Discusses GL 97-06 Re Degradation of SG Internals,Dtd 971230.GL Requested Each PWR Licensee to Submit Info That Will Enable NRC Staff to Verify Whether PWR SG Internals Comply & Conform to Current Licensing Basis for Facilities 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 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 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 ML20212J2451999-10-0101 October 1999 Informs That Util 980807 & 990629 Responses to GL 98-01, Y2K Readiness of Computer Sys at NPPs Acceptable.Nrc Considers Subj GL to Be Closed for Units 2 &3 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 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 ML20217A9271999-09-30030 September 1999 Discusses Investigation Conducted at Millstone Nuclear Power Station by NRC OI Region I on 980319 to Determine If Contract Training Instructor Was Terminated for Raising Concerns About Quality of Training Matls ML20217B3221999-09-30030 September 1999 Refers to Investigation Rept 1-1997-035 Conducted at Millstone Nuclear Power Station by NRC Ofc of Investigation Field Ofc,Region I on 970915 to Determine Whether Former Health Physics Technician Discriminated Against ML20212J6621999-09-30030 September 1999 Informs of Completion of mid-cycle PPR of Units 2 & 3 on 990916.Identified Several Recent Instances in Which Condition Repts Were Not Initiated,Resulting in Untimely or Inadequate C/As.Historical Listing of Plant Issues Encl B17887, Requests Exemption from Requirements of 10CFR140.11(a)(4) Which Requires Licensees to Maintain Secondary Financial Protection Beyond Primary Layer1999-09-28028 September 1999 Requests Exemption from Requirements of 10CFR140.11(a)(4) Which Requires Licensees to Maintain Secondary Financial Protection Beyond Primary Layer ML20216J1341999-09-28028 September 1999 Ltr Contract:Task Order 49, Millstone Units 2 & 3 Employee Concerns Program Insp, Under Contract NRC-03-98-021 B17883, Forwards Mnps Unit 3 ISI Summary Rept,Cycle 6, IAW ASME Section XI,IWA-62301999-09-27027 September 1999 Forwards Mnps Unit 3 ISI Summary Rept,Cycle 6, IAW ASME Section XI,IWA-6230 B17890, Provides Response to GL 99-02, Laboratory Testing of Nuclear-Grade Activated Charcoal1999-09-27027 September 1999 Provides Response to GL 99-02, Laboratory Testing of Nuclear-Grade Activated Charcoal B17888, Informs That There No Longer Exists Need to Maintain Millstone Unit 2 SRO License for CA Hines,License SOP-10741-01,effective 9909151999-09-24024 September 1999 Informs That There No Longer Exists Need to Maintain Millstone Unit 2 SRO License for CA Hines,License SOP-10741-01,effective 990915 B17884, Forwards NRC Form 536, Operator Licensing Exam Data, for Mnps,Units 2 & 3,per Administrative Ltr 99-031999-09-24024 September 1999 Forwards NRC Form 536, Operator Licensing Exam Data, for Mnps,Units 2 & 3,per Administrative Ltr 99-03 ML20212F4961999-09-20020 September 1999 Forwards Insp Repts 50-245/99-08,50-336/99-08 & 50-423/99-08 on 990615-0809.Four Violations of NRC Requirements Occurred & Being Treated as Ncvs,Consistent with App C of Enforcement Policy 05000336/LER-1999-001, Forwards LER 99-001-00 Re Thermal Reactor Power Limit That Was Exceeded.Commitments Made by Util Encl1999-09-20020 September 1999 Forwards LER 99-001-00 Re Thermal Reactor Power Limit That Was Exceeded.Commitments Made by Util Encl B17867, Requests Relief from Requirements of 10CFR50.55a(g),IAW ASME Section XI for Millstone,Unit 3.Util Requests Relief from Performing Visual Exam of Reactor Pressure Supports to Extent Required by Code for Class 1 Supports1999-09-17017 September 1999 Requests Relief from Requirements of 10CFR50.55a(g),IAW ASME Section XI for Millstone,Unit 3.Util Requests Relief from Performing Visual Exam of Reactor Pressure Supports to Extent Required by Code for Class 1 Supports B17865, Provides NRC Staff with Updated Proposed Rev of FSAR Section 14.6.3, Radiological Consequences of SG Tube Failure. Updated Proposed Rev Will Replace Info Provided in Attachment 3 of Initial Submittal1999-09-16016 September 1999 Provides NRC Staff with Updated Proposed Rev of FSAR Section 14.6.3, Radiological Consequences of SG Tube Failure. Updated Proposed Rev Will Replace Info Provided in Attachment 3 of Initial Submittal B17876, Informs That Util Will Adopt Last Approved Northeast Util QA Program (Nuqap) Tr,Rev 21,dtd 990630,as Unit 1 Nuqap,Per Decision to Permanently Cease Operations at Subject Plant1999-09-16016 September 1999 Informs That Util Will Adopt Last Approved Northeast Util QA Program (Nuqap) Tr,Rev 21,dtd 990630,as Unit 1 Nuqap,Per Decision to Permanently Cease Operations at Subject Plant B17882, Forwards NRC Forms 398 & 369 in Support of License Renewal for Bb Parrish,License SOP-10399-2.Encl Withheld Per 10CFR2.790(a)(6)1999-09-15015 September 1999 Forwards NRC Forms 398 & 369 in Support of License Renewal for Bb Parrish,License SOP-10399-2.Encl Withheld Per 10CFR2.790(a)(6) B17881, Forwards NRC Form 398 & NRC Form 396 for Le Olsen,License SOP-10398-2.Encl Withheld Per 10CFR2.790(a)(6)1999-09-15015 September 1999 Forwards NRC Form 398 & NRC Form 396 for Le Olsen,License SOP-10398-2.Encl Withheld Per 10CFR2.790(a)(6) B17859, Forwards up-to-date Distribution Lists for NRC Correspondence to NNECO & NUSCO.Side-bars Indicate Changes from Previous Lists Provided to NRC on 9801281999-09-15015 September 1999 Forwards up-to-date Distribution Lists for NRC Correspondence to NNECO & NUSCO.Side-bars Indicate Changes from Previous Lists Provided to NRC on 980128 B17880, Forwards NRC Form 398 & NRC Form 396 for Rf Martin,License SOP-10397-0.Encls Withheld Per 10CFR2.790(a)(6)1999-09-15015 September 1999 Forwards NRC Form 398 & NRC Form 396 for Rf Martin,License SOP-10397-0.Encls Withheld Per 10CFR2.790(a)(6) B17872, Informs of Election to Consolidate Previous Commitments Re Work Observation Program with Two New Programmatic Commitments Listed in Attachment 11999-09-14014 September 1999 Informs of Election to Consolidate Previous Commitments Re Work Observation Program with Two New Programmatic Commitments Listed in Attachment 1 ML20212A9011999-09-10010 September 1999 Forwards Environ Assessment & Finding of No Significant Impact Re Application for Exemption,Dtd 990803.Proposed Exemption Would Provide Relief from Requirement of 10CFR50 ML20212A3171999-09-10010 September 1999 Discusses Investigation Rept 1-1998-045 Conducted on 981112 to Determine If Former Senior Health Physics Technician Being Denied Employment at Millstone in Retaliation for Having Raised Safety Concerns in Past.Synopsis Encl B17838, Forwards Revs 34 & 35 to Physical Security Plan.Explanation of Changes Provided as Attachment 1.Without Encls1999-09-10010 September 1999 Forwards Revs 34 & 35 to Physical Security Plan.Explanation of Changes Provided as Attachment 1.Without Encls ML20212A7501999-09-10010 September 1999 Forwards Staff Requirements Memo Response,Dtd 990525,which Provides Actions NRC Plans for Continued Oversight of safety-conscious Work Environ & Employee Concerns Program ML20211J9291999-09-0303 September 1999 Forwards mark-ups & Retypes of Proposed Conforming License Changes Required in Connection with Transfers Being Sought in 990615 Application of Montaup Electric Co & New England Power Co for Transfer of Licenses & Ownership Interests ML20211K5171999-09-0202 September 1999 Expresses Appreciation for Support Provided for NRC Public Meeting on 990825 05000336/LER-1999-010, Forwards LER 99-010-00,documenting 990804 Event of Failure to Perform ASME Section XI IST on Pressurizer Relief Line Flow Control Sample Valve Following Maint Activities.List of Util Commitments Contained in Attachment 11999-09-0202 September 1999 Forwards LER 99-010-00,documenting 990804 Event of Failure to Perform ASME Section XI IST on Pressurizer Relief Line Flow Control Sample Valve Following Maint Activities.List of Util Commitments Contained in Attachment 1 ML20216H0591999-09-0202 September 1999 Responds to Re Issues Submitted by Cullen on Behalf of Several Petitioners Concerning Offsite Emergency Prepardeness for Millstone Nuclear Power Station ML20211N9241999-09-0101 September 1999 Forwards Document Classification Form for Insertion Into Emergency Planning Services Department Procedures ML20211H0741999-08-30030 August 1999 Discusses GL 92-01,Rev 1, Rv Structural Integrity, Issued by NRC on 950519 & NNECO Responses for Millstone Unit 2 & 980715.Informs That Staff Revised Info in Rvid & Released Info as Rvid Version 2 Based on Response Review 1999-09-30
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARML20217N8261999-10-25025 October 1999 Discusses Errata Re 991021 Filing of Northeast Nuclear Energy Co Answer to Request for Hearing & Petition to Intervene B17886, Requests Permission to Utilize Code Case N-623, Deferral of Insps of Shell-to-Flange & Head-to-Flange of Reactor Vessel,Section Xi,Div 1, for Millstone Unit 21999-10-18018 October 1999 Requests Permission to Utilize Code Case N-623, Deferral of Insps of Shell-to-Flange & Head-to-Flange of Reactor Vessel,Section Xi,Div 1, for Millstone Unit 2 B17901, Submits Statement of Personal Qualification (NRC Form 398) Along with Supporting Certification of Medical Exam by Facility Licensee (NRC from 396) in Support of License Renewal for PM Miner.Encls Withheld,Per 10CFR2.790(a)(6)1999-10-18018 October 1999 Submits Statement of Personal Qualification (NRC Form 398) Along with Supporting Certification of Medical Exam by Facility Licensee (NRC from 396) in Support of License Renewal for PM Miner.Encls Withheld,Per 10CFR2.790(a)(6) 05000336/LER-1999-012, Forwards LER 99-012-00,re Unrecoverable CEA Misalignment Entry Into TS 3.0.3 on 990917.Commitments Made by Util Are Encl1999-10-15015 October 1999 Forwards LER 99-012-00,re Unrecoverable CEA Misalignment Entry Into TS 3.0.3 on 990917.Commitments Made by Util Are Encl B17900, Forwards Revised Assumptions Used in Fuel Handling Accident Analysis & Summary of Results.List of Regulatory Commitments Are Listed in Attachment 11999-10-14014 October 1999 Forwards Revised Assumptions Used in Fuel Handling Accident Analysis & Summary of Results.List of Regulatory Commitments Are Listed in Attachment 1 ML20217P1201999-10-0606 October 1999 Informs NRC of Proposed Acquisition of Parent Holding Company of Central Maine & Requests NRC Concurrence,Based on Threshold Review,That Proposed Acquisition Does Not,In Fact, Constitute Transfer Subject to 10CFR50.80 ML20217F0031999-10-0606 October 1999 Forwards Original Petition to Intervene Being Filed on Behalf of Clients,Connecticut Coalition Against Millstone & Long Island Coalition Against Millstone,Iaw Provisions of 10CFR2.714 B17892, Requests Withdrawal of License Amend Application Re 24-month SG Tube Insp Surveillance Extensions,Submitted in Util 950726 & s to NRC1999-10-0505 October 1999 Requests Withdrawal of License Amend Application Re 24-month SG Tube Insp Surveillance Extensions,Submitted in Util 950726 & s to NRC B17887, Requests Exemption from Requirements of 10CFR140.11(a)(4) Which Requires Licensees to Maintain Secondary Financial Protection Beyond Primary Layer1999-09-28028 September 1999 Requests Exemption from Requirements of 10CFR140.11(a)(4) Which Requires Licensees to Maintain Secondary Financial Protection Beyond Primary Layer B17883, Forwards Mnps Unit 3 ISI Summary Rept,Cycle 6, IAW ASME Section XI,IWA-62301999-09-27027 September 1999 Forwards Mnps Unit 3 ISI Summary Rept,Cycle 6, IAW ASME Section XI,IWA-6230 B17890, Provides Response to GL 99-02, Laboratory Testing of Nuclear-Grade Activated Charcoal1999-09-27027 September 1999 Provides Response to GL 99-02, Laboratory Testing of Nuclear-Grade Activated Charcoal B17884, Forwards NRC Form 536, Operator Licensing Exam Data, for Mnps,Units 2 & 3,per Administrative Ltr 99-031999-09-24024 September 1999 Forwards NRC Form 536, Operator Licensing Exam Data, for Mnps,Units 2 & 3,per Administrative Ltr 99-03 B17888, Informs That There No Longer Exists Need to Maintain Millstone Unit 2 SRO License for CA Hines,License SOP-10741-01,effective 9909151999-09-24024 September 1999 Informs That There No Longer Exists Need to Maintain Millstone Unit 2 SRO License for CA Hines,License SOP-10741-01,effective 990915 05000336/LER-1999-001, Forwards LER 99-001-00 Re Thermal Reactor Power Limit That Was Exceeded.Commitments Made by Util Encl1999-09-20020 September 1999 Forwards LER 99-001-00 Re Thermal Reactor Power Limit That Was Exceeded.Commitments Made by Util Encl B17867, Requests Relief from Requirements of 10CFR50.55a(g),IAW ASME Section XI for Millstone,Unit 3.Util Requests Relief from Performing Visual Exam of Reactor Pressure Supports to Extent Required by Code for Class 1 Supports1999-09-17017 September 1999 Requests Relief from Requirements of 10CFR50.55a(g),IAW ASME Section XI for Millstone,Unit 3.Util Requests Relief from Performing Visual Exam of Reactor Pressure Supports to Extent Required by Code for Class 1 Supports B17876, Informs That Util Will Adopt Last Approved Northeast Util QA Program (Nuqap) Tr,Rev 21,dtd 990630,as Unit 1 Nuqap,Per Decision to Permanently Cease Operations at Subject Plant1999-09-16016 September 1999 Informs That Util Will Adopt Last Approved Northeast Util QA Program (Nuqap) Tr,Rev 21,dtd 990630,as Unit 1 Nuqap,Per Decision to Permanently Cease Operations at Subject Plant B17865, Provides NRC Staff with Updated Proposed Rev of FSAR Section 14.6.3, Radiological Consequences of SG Tube Failure. Updated Proposed Rev Will Replace Info Provided in Attachment 3 of Initial Submittal1999-09-16016 September 1999 Provides NRC Staff with Updated Proposed Rev of FSAR Section 14.6.3, Radiological Consequences of SG Tube Failure. Updated Proposed Rev Will Replace Info Provided in Attachment 3 of Initial Submittal B17881, Forwards NRC Form 398 & NRC Form 396 for Le Olsen,License SOP-10398-2.Encl Withheld Per 10CFR2.790(a)(6)1999-09-15015 September 1999 Forwards NRC Form 398 & NRC Form 396 for Le Olsen,License SOP-10398-2.Encl Withheld Per 10CFR2.790(a)(6) B17880, Forwards NRC Form 398 & NRC Form 396 for Rf Martin,License SOP-10397-0.Encls Withheld Per 10CFR2.790(a)(6)1999-09-15015 September 1999 Forwards NRC Form 398 & NRC Form 396 for Rf Martin,License SOP-10397-0.Encls Withheld Per 10CFR2.790(a)(6) B17859, Forwards up-to-date Distribution Lists for NRC Correspondence to NNECO & NUSCO.Side-bars Indicate Changes from Previous Lists Provided to NRC on 9801281999-09-15015 September 1999 Forwards up-to-date Distribution Lists for NRC Correspondence to NNECO & NUSCO.Side-bars Indicate Changes from Previous Lists Provided to NRC on 980128 B17882, Forwards NRC Forms 398 & 369 in Support of License Renewal for Bb Parrish,License SOP-10399-2.Encl Withheld Per 10CFR2.790(a)(6)1999-09-15015 September 1999 Forwards NRC Forms 398 & 369 in Support of License Renewal for Bb Parrish,License SOP-10399-2.Encl Withheld Per 10CFR2.790(a)(6) B17872, Informs of Election to Consolidate Previous Commitments Re Work Observation Program with Two New Programmatic Commitments Listed in Attachment 11999-09-14014 September 1999 Informs of Election to Consolidate Previous Commitments Re Work Observation Program with Two New Programmatic Commitments Listed in Attachment 1 B17838, Forwards Revs 34 & 35 to Physical Security Plan.Explanation of Changes Provided as Attachment 1.Without Encls1999-09-10010 September 1999 Forwards Revs 34 & 35 to Physical Security Plan.Explanation of Changes Provided as Attachment 1.Without Encls ML20211J9291999-09-0303 September 1999 Forwards mark-ups & Retypes of Proposed Conforming License Changes Required in Connection with Transfers Being Sought in 990615 Application of Montaup Electric Co & New England Power Co for Transfer of Licenses & Ownership Interests 05000336/LER-1999-010, Forwards LER 99-010-00,documenting 990804 Event of Failure to Perform ASME Section XI IST on Pressurizer Relief Line Flow Control Sample Valve Following Maint Activities.List of Util Commitments Contained in Attachment 11999-09-0202 September 1999 Forwards LER 99-010-00,documenting 990804 Event of Failure to Perform ASME Section XI IST on Pressurizer Relief Line Flow Control Sample Valve Following Maint Activities.List of Util Commitments Contained in Attachment 1 ML20216H0591999-09-0202 September 1999 Responds to Re Issues Submitted by Cullen on Behalf of Several Petitioners Concerning Offsite Emergency Prepardeness for Millstone Nuclear Power Station ML20211N9241999-09-0101 September 1999 Forwards Document Classification Form for Insertion Into Emergency Planning Services Department Procedures B17851, Forwards Semiannual fitness-for-duty Program Performance Data for 990101-990630 for Millstone Nuclear Power Station, Units 1,2 & 3,IAW 10CFR26.71(d)1999-08-27027 August 1999 Forwards Semiannual fitness-for-duty Program Performance Data for 990101-990630 for Millstone Nuclear Power Station, Units 1,2 & 3,IAW 10CFR26.71(d) B17855, Forwards NRC Forms 398 & 396 in Support of License Renewal for SRO TE Grilley,SOP-4053-04.Encl Withheld,Per 10CFR2.7901999-08-17017 August 1999 Forwards NRC Forms 398 & 396 in Support of License Renewal for SRO TE Grilley,SOP-4053-04.Encl Withheld,Per 10CFR2.790 B17849, Forwards Second Quarter Backlog Performance Rept for 1999, Which Represents Fourth Rept on Mnps Performance Since Restart of Unit 3 & First Status Update for Unit 2.No Regulatory Commitments Are Contained in Ltr1999-08-16016 August 1999 Forwards Second Quarter Backlog Performance Rept for 1999, Which Represents Fourth Rept on Mnps Performance Since Restart of Unit 3 & First Status Update for Unit 2.No Regulatory Commitments Are Contained in Ltr B17854, Forwards Monthly Operating Rept for July 1999 for Millstone Nuclear Power Station,Unit 2,per TS 6.9.1.7.Revised Repts for May & June Also Encl Which Reflect Correct Faulty Printometer Readings1999-08-14014 August 1999 Forwards Monthly Operating Rept for July 1999 for Millstone Nuclear Power Station,Unit 2,per TS 6.9.1.7.Revised Repts for May & June Also Encl Which Reflect Correct Faulty Printometer Readings B17850, Forwards First Lhc Quarterly Assessment Rept for Assessment Performed 990621 to 990701.NNECO Taking Appropriate Actions to Address Observations in Rept1999-08-11011 August 1999 Forwards First Lhc Quarterly Assessment Rept for Assessment Performed 990621 to 990701.NNECO Taking Appropriate Actions to Address Observations in Rept B17837, Forwards COLR for Cycle 7, for Millstone Unit 3,IAW TS 6.9.1.6.Explanation of Changes to COLR Also Encl1999-08-0707 August 1999 Forwards COLR for Cycle 7, for Millstone Unit 3,IAW TS 6.9.1.6.Explanation of Changes to COLR Also Encl B17657, Requests Schedular Exemption from Emergency Plan Exercise Requirements of 10CFR50,App E,Part Iv,Section F,Paragraph 2.c.Requests That Nrc/Fema Evaluated Exercise Be Conducted in Mar 2000 Rather than Sept 19991999-08-0303 August 1999 Requests Schedular Exemption from Emergency Plan Exercise Requirements of 10CFR50,App E,Part Iv,Section F,Paragraph 2.c.Requests That Nrc/Fema Evaluated Exercise Be Conducted in Mar 2000 Rather than Sept 1999 B17845, Forwards Revised Commitment for Surveillance Scheduling & Tracking.Options for Surveillance Scheduling & Tracking Methodologies to Be Incorporated in Standardized Station Surveillance Program Are Currently Being Reconsidered1999-08-0202 August 1999 Forwards Revised Commitment for Surveillance Scheduling & Tracking.Options for Surveillance Scheduling & Tracking Methodologies to Be Incorporated in Standardized Station Surveillance Program Are Currently Being Reconsidered B17831, Informs NRC Staff That Change 3 to Rev 25 of Mnps Emergency Plan Was Implemented on 990715.Change Removes Facility Organizational Charts from Emergency Plan & Identifies Relocation to Nuqap1999-07-26026 July 1999 Informs NRC Staff That Change 3 to Rev 25 of Mnps Emergency Plan Was Implemented on 990715.Change Removes Facility Organizational Charts from Emergency Plan & Identifies Relocation to Nuqap B17834, Forwards Proprietary Revised NRC Form 398,which Certifies That SL Doboe Has Completed Eligibility Requirements for Sro,Per 10CFR55.31.Proprietary Info Withheld1999-07-20020 July 1999 Forwards Proprietary Revised NRC Form 398,which Certifies That SL Doboe Has Completed Eligibility Requirements for Sro,Per 10CFR55.31.Proprietary Info Withheld B17836, Forwards Revised NRC Form 396 & Supporting Physician Rept for Licensed Operator Restricted from Licensed Duties, Effective 990628,due to Medical Condition.Without Encl1999-07-20020 July 1999 Forwards Revised NRC Form 396 & Supporting Physician Rept for Licensed Operator Restricted from Licensed Duties, Effective 990628,due to Medical Condition.Without Encl B17811, Submits Response to NRC AL 99-02,requesting That Licensees Provide Numerical Estimates of Licensing Actions to Be Expected to Be Submitted in Fy 2000 & 20011999-07-20020 July 1999 Submits Response to NRC AL 99-02,requesting That Licensees Provide Numerical Estimates of Licensing Actions to Be Expected to Be Submitted in Fy 2000 & 2001 ML20210S9911999-07-18018 July 1999 Requests NRC Intervene for All Shareholders of New England Electric System & to Help with Merger with National Grid Group & That NRC Petition Security & Exchange Commission to Investigate Matter Relative to No Shareholder Options B17835, Forwards Rev 33 to Millstone Station Physical Security Plan, Per 10CFR50.54(p)(2).Licensee Determined That Changes Do Not Decrease Effectiveness of Plan.Rev Withheld from Public Disclosure,Per 10CFR2.790(d)(1)1999-07-16016 July 1999 Forwards Rev 33 to Millstone Station Physical Security Plan, Per 10CFR50.54(p)(2).Licensee Determined That Changes Do Not Decrease Effectiveness of Plan.Rev Withheld from Public Disclosure,Per 10CFR2.790(d)(1) B17818, Provides NRC Staff with Change to TS Bases Sections 3/4.5.2 & 3/4.5.3, ECCS Subsystems for Info Only.Change Was Reviewed & Approved by Unit 3 Plant Operations Review Committee IAW Provisions of 10CFR50.591999-07-16016 July 1999 Provides NRC Staff with Change to TS Bases Sections 3/4.5.2 & 3/4.5.3, ECCS Subsystems for Info Only.Change Was Reviewed & Approved by Unit 3 Plant Operations Review Committee IAW Provisions of 10CFR50.59 B17824, Forwards Monthly Oeprating Rept for June 1999 & Revised Monthly Operating Rept for May 1999 for Millstone Unit 21999-07-13013 July 1999 Forwards Monthly Oeprating Rept for June 1999 & Revised Monthly Operating Rept for May 1999 for Millstone Unit 2 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 B17816, Provides Certification That M Lettrich,Has Completed Eligibility Requirements,Per 10CFR55.31 for Operator License.Util Requests That Licensing Action Be Taken for Named Individual1999-07-0101 July 1999 Provides Certification That M Lettrich,Has Completed Eligibility Requirements,Per 10CFR55.31 for Operator License.Util Requests That Licensing Action Be Taken for Named Individual B17801, Forwards 10CFR50.59 Annual Rept for Period Jan-Dec 1998. Various Changes That Were Initiated in Previous Yrs & Completed in 1998,also Incorporated Into Annual Rept1999-06-30030 June 1999 Forwards 10CFR50.59 Annual Rept for Period Jan-Dec 1998. Various Changes That Were Initiated in Previous Yrs & Completed in 1998,also Incorporated Into Annual Rept B17819, Forwards Rev 17 to FSAR & Addendum 6 to Annual Rept.Nneco Recently Completed Review of Unit 2 Design & Licensing Bases Which Resulted in Changes to FSAR Provided in Encl 1.Encl 2 Includes Info Covering Changes Not Included in1999-06-30030 June 1999 Forwards Rev 17 to FSAR & Addendum 6 to Annual Rept.Nneco Recently Completed Review of Unit 2 Design & Licensing Bases Which Resulted in Changes to FSAR Provided in Encl 1.Encl 2 Includes Info Covering Changes Not Included in B17780, Forwards Rev 21,Change 0 to Northeast Utilities QAP (Nuqap) TR, IAW 10CFR50.54(a)(3).Changes to TR Are Shown as Text in Bold Italics1999-06-30030 June 1999 Forwards Rev 21,Change 0 to Northeast Utilities QAP (Nuqap) TR, IAW 10CFR50.54(a)(3).Changes to TR Are Shown as Text in Bold Italics B17723, Responds to NRC Request for Info Re GL 98-01, Y2K Readiness of Computer Sys at Npps. Y2K Readiness Disclosure for Units 2 & 3 Encl.Without Encl1999-06-29029 June 1999 Responds to NRC Request for Info Re GL 98-01, Y2K Readiness of Computer Sys at Npps. Y2K Readiness Disclosure for Units 2 & 3 Encl.Without Encl B17767, Forwards Rev 12 to FSAR & Addendum 3 to Annual Rept, for Millstone Unit 3,per 10CFR50.71(e) & 10CFR50.4(b)(6). No New Regulatory Commitments Contained in Ltr1999-06-29029 June 1999 Forwards Rev 12 to FSAR & Addendum 3 to Annual Rept, for Millstone Unit 3,per 10CFR50.71(e) & 10CFR50.4(b)(6). No New Regulatory Commitments Contained in Ltr 1999-09-03
[Table view] Category:UTILITY TO NRC
MONTHYEARML20064A8071990-09-11011 September 1990 Forwards Core Operating Limits Rept for Four & Three Loop Operation,Per Tech Spec 6.9.1.6.d A08900, Responds to Expressing NRC Views on Access to Util Internal or third-party Assessment Repts.Believes Internal Analysis to Support Amend of One of Util NRC OLs to Authorize Higher Power Level Is within NRC Purview & Access1990-09-11011 September 1990 Responds to Expressing NRC Views on Access to Util Internal or third-party Assessment Repts.Believes Internal Analysis to Support Amend of One of Util NRC OLs to Authorize Higher Power Level Is within NRC Purview & Access B13628, Forwards Correction to Error Discovered in Util 900727 Response to Notice of Violation from Insp Rept 50-336/90-09. Statement Corrected to Read That Contract Personnel That Have long-term Assignments,Certified as Testing Personnel1990-09-0707 September 1990 Forwards Correction to Error Discovered in Util 900727 Response to Notice of Violation from Insp Rept 50-336/90-09. Statement Corrected to Read That Contract Personnel That Have long-term Assignments,Certified as Testing Personnel B13624, Forwards Addl Info Re Relief Request from ASME Code Section XI Requirements for Temporary Repair to Piping Adjacent to Valve 2-SW-97A,per 900817 Telcon1990-09-0505 September 1990 Forwards Addl Info Re Relief Request from ASME Code Section XI Requirements for Temporary Repair to Piping Adjacent to Valve 2-SW-97A,per 900817 Telcon A08977, Advises of Commitment to Install Hardened Wetwell Vent at Facility,In Response to NRC .Util Will Be Proceeding W/Initial Design & Engineering of Hardened Vent, to Support Installation During 1993 Refueling Outage1990-09-0404 September 1990 Advises of Commitment to Install Hardened Wetwell Vent at Facility,In Response to NRC .Util Will Be Proceeding W/Initial Design & Engineering of Hardened Vent, to Support Installation During 1993 Refueling Outage B13596, Forwards Millstone Unit 3 Individual Plant Exam for Severe Accident Vulnerabilities, Per Generic Ltr 88-20.Rept Identified That No Major Severe Accident Vulnerabilities Requiring Corrective Action Needed1990-08-31031 August 1990 Forwards Millstone Unit 3 Individual Plant Exam for Severe Accident Vulnerabilities, Per Generic Ltr 88-20.Rept Identified That No Major Severe Accident Vulnerabilities Requiring Corrective Action Needed B13626, Informs That Info Provided in Re safety-related Equipment Classification Programs Also Applicable for Unit 3 of Facility1990-08-31031 August 1990 Informs That Info Provided in Re safety-related Equipment Classification Programs Also Applicable for Unit 3 of Facility B13618, Forwards Addl Info on Util 900815 Relief Request from ASME Code Section XI Requirements,Per 10CFR50.55a(g)(6)(i),for Repairs to Pipe 3SWP-006-050-031990-08-24024 August 1990 Forwards Addl Info on Util 900815 Relief Request from ASME Code Section XI Requirements,Per 10CFR50.55a(g)(6)(i),for Repairs to Pipe 3SWP-006-050-03 ML20059C2061990-08-23023 August 1990 Forwards Vols 1 & 2 to Semiannual Radioactive Effluents Release Rept Jan-June 1990, Per 10CFR50.36a.Rept Includes Summary of Quantities of Solid Radwaste & Liquid & Gaseous Effluents A08918, Responds to Notice of Violation & Proposed Imposition of Civil Penalty Re Insp Rept 50-245/90-08.Mitigation of Civil Penalty Requested.Corrective Action:List of Procedural Changes Provided1990-08-22022 August 1990 Responds to Notice of Violation & Proposed Imposition of Civil Penalty Re Insp Rept 50-245/90-08.Mitigation of Civil Penalty Requested.Corrective Action:List of Procedural Changes Provided B13610, Requests Relief from ASME Boiler & Pressure Vessel Code Section XI Requirements Re Mods to Pipe 3SWP-006-050-03. Results of Insps & Required Repairs Will Determine Schedule for Future Insps1990-08-15015 August 1990 Requests Relief from ASME Boiler & Pressure Vessel Code Section XI Requirements Re Mods to Pipe 3SWP-006-050-03. Results of Insps & Required Repairs Will Determine Schedule for Future Insps B13595, Notifies NRC That Utils Volunteer to Participate in Emergency Response Data Sys Project for All Four Nuclear Units,Per Generic Ltr 89-15 & NUREG-13941990-08-14014 August 1990 Notifies NRC That Utils Volunteer to Participate in Emergency Response Data Sys Project for All Four Nuclear Units,Per Generic Ltr 89-15 & NUREG-1394 B13607, Requests Relief from ASME Code Section XI to Reflect Mod to Piping Adjacent to Valve 2-SW-97A,in Response to Generic Ltr 90-051990-08-10010 August 1990 Requests Relief from ASME Code Section XI to Reflect Mod to Piping Adjacent to Valve 2-SW-97A,in Response to Generic Ltr 90-05 A08845, Responds to NRC Backfit Analyses of Hardened Wetwell Vent, Contained in .Benefit of Installing Hardened Wetwell Vent to Satisfy Basic Design Objective of Preventing core-melt Event Not Sufficient for Immediate Decision1990-08-0808 August 1990 Responds to NRC Backfit Analyses of Hardened Wetwell Vent, Contained in .Benefit of Installing Hardened Wetwell Vent to Satisfy Basic Design Objective of Preventing core-melt Event Not Sufficient for Immediate Decision ML20058N2181990-08-0707 August 1990 Notification of Change in Senior Operator Status.Util Determined That Need to Maintain Senior OL of LS Allen No Longer Exists.Determination Effective 900719 ML20058M8321990-08-0707 August 1990 Discusses Spent Fuel Racks Poison Surveillance Coupon Boraflex Degradation.Visual Exam of Remaining Surveillance Coupons Revealed Similar Situation Existed in All Coupon Samples B13590, Responds to NRC 900718 Request for Addl Info Re Util 900418 Request for Schedular Exemption from 10CFR50,App E,Section IV.F.3 to Allow Dec 1990 Full Participation Exercise to Be Exchanged W/Oct 1991 Partial Participation Exercise1990-08-0101 August 1990 Responds to NRC 900718 Request for Addl Info Re Util 900418 Request for Schedular Exemption from 10CFR50,App E,Section IV.F.3 to Allow Dec 1990 Full Participation Exercise to Be Exchanged W/Oct 1991 Partial Participation Exercise A08881, Responds to NRC Re Violations Noted in Insp Rept 50-423/90-08.Corrective Action:Operators Directly Involved W/Event Removed from Licensed Duties & Counseled by Operations Manager on Causes of Event1990-07-31031 July 1990 Responds to NRC Re Violations Noted in Insp Rept 50-423/90-08.Corrective Action:Operators Directly Involved W/Event Removed from Licensed Duties & Counseled by Operations Manager on Causes of Event B13594, Forwards Rev 13 to Physical Security Plan.Rev Withheld (Ref 10CFR73.21(b) & 2.790(a))1990-07-30030 July 1990 Forwards Rev 13 to Physical Security Plan.Rev Withheld (Ref 10CFR73.21(b) & 2.790(a)) ML20055J4621990-07-27027 July 1990 Advises That Need to Maintain OL or Senior OL for Listed Individuals No Longer Exists,Effective 900701 B13585, Provides Supplemental Response to Generic Ltr 89-19 Re Request for Action Concerning Resolution of Unresolved Safety Issue A-47.Plant Procedures Modified to Provide Operability Verification of Steam Generator Protection Sys1990-07-26026 July 1990 Provides Supplemental Response to Generic Ltr 89-19 Re Request for Action Concerning Resolution of Unresolved Safety Issue A-47.Plant Procedures Modified to Provide Operability Verification of Steam Generator Protection Sys A08565, Responds to NRC 900302 Request for Addl Info Re LPCI Swing Bus Transfer Design & Single Failure Vulnerabilities1990-07-26026 July 1990 Responds to NRC 900302 Request for Addl Info Re LPCI Swing Bus Transfer Design & Single Failure Vulnerabilities B13592, Requests Extension to Respond to NRC Backfit Analyses of Hardened Wetwell Vent1990-07-24024 July 1990 Requests Extension to Respond to NRC Backfit Analyses of Hardened Wetwell Vent ML20063P9791990-07-23023 July 1990 Notification of Change in SL Jackson Status Effective 900701,due to Permanent Reassignment within Util B13563, Discusses Util Comprehensive Review of Options Re Degraded Grid Undervoltage Protection.Confirms Previous Conclusion That Splitting Loss of Normal Power Logic Would Have Overall Adverse Impact on Plant Safety1990-07-20020 July 1990 Discusses Util Comprehensive Review of Options Re Degraded Grid Undervoltage Protection.Confirms Previous Conclusion That Splitting Loss of Normal Power Logic Would Have Overall Adverse Impact on Plant Safety B13566, Advises That Util Has Reasonable Assurance That Safety Relief Valves Operable & Will Perform as Expected Until Next Outage1990-07-20020 July 1990 Advises That Util Has Reasonable Assurance That Safety Relief Valves Operable & Will Perform as Expected Until Next Outage B13588, Corrects 900703 Submittal of Results of Second in-cycle Insp of Steam Generators.All Tubes W/Cracks Stacked & Plugged1990-07-18018 July 1990 Corrects 900703 Submittal of Results of Second in-cycle Insp of Steam Generators.All Tubes W/Cracks Stacked & Plugged A08822, Submits Addl Info Requested Re Util Response to Generic Ltr 88-11, Radiation Embrittlement of Reactor Vessel Matls. Charpy Impact Use Values for Welds Provided in Evaluation of Irradiated Capsule W-971990-07-18018 July 1990 Submits Addl Info Requested Re Util Response to Generic Ltr 88-11, Radiation Embrittlement of Reactor Vessel Matls. Charpy Impact Use Values for Welds Provided in Evaluation of Irradiated Capsule W-97 ML20055G5331990-07-18018 July 1990 Forwards Decommissioning Financial Assurance Certification Rept B13587, Requests Temporary Waiver of Compliance from Tech Spec 3.5.F.2 Re Emergency Diesel Generator (EDG) Limiting Condition for Operation.Waiver Would Extend Available Time to Repair Damaged Electrical Components1990-07-16016 July 1990 Requests Temporary Waiver of Compliance from Tech Spec 3.5.F.2 Re Emergency Diesel Generator (EDG) Limiting Condition for Operation.Waiver Would Extend Available Time to Repair Damaged Electrical Components ML20055D3461990-07-0303 July 1990 Responds to NRC Bulletin 90-001 Re Loss of Fill Oil Transmitters Mfg by Rosemount.Operability Determinations Performed & Documented for All Rosemount 1153 & 1154 Transmitters at Facility B13545, Forwards Rev 3 to Updated FSAR for Millstone Unit 31990-06-29029 June 1990 Forwards Rev 3 to Updated FSAR for Millstone Unit 3 ML20055D7191990-06-29029 June 1990 Amends 900604 Rev 13 to QA Program ML20055D3481990-06-29029 June 1990 Forwards Addl Info Re Facility Crdr & Isap,Including Justification for Human Engineering Discrepancies Dispositioned for No Corrective Action B13531, Forwards Rev 8 to Updated FSAR for Millstone Unit 21990-06-29029 June 1990 Forwards Rev 8 to Updated FSAR for Millstone Unit 2 B13550, Responds to Generic Ltr 90-04, Request for Info on Status of Generic Safety Issues Resolved W/Imposition of Requirements or Corrective Actions. Implementation & Completion Tables for staff-imposed Requirements Encl1990-06-27027 June 1990 Responds to Generic Ltr 90-04, Request for Info on Status of Generic Safety Issues Resolved W/Imposition of Requirements or Corrective Actions. Implementation & Completion Tables for staff-imposed Requirements Encl B13499, Forwards Corrected Tech Specs Page 3/4 9-1 for Incorporation Into Proposed Amend 36 to License DPR-21 Re Auxiliary Electrical Sys1990-06-26026 June 1990 Forwards Corrected Tech Specs Page 3/4 9-1 for Incorporation Into Proposed Amend 36 to License DPR-21 Re Auxiliary Electrical Sys ML20043F8721990-06-11011 June 1990 Corrects Name of Vendor Supplying Replacement Plug Valves, Per Util 900511 Ltr.Replacement Bolts,Not Valves,Purchased from Cardinal Industrial Products Corp ML20043H0161990-06-0808 June 1990 Requests Exemption from App J to 10CFR50 for 12 Valves in Reactor Bldg Closed Cooling Water Sys.Valves Not within Definition of Containment Isolation Valves in App J & Not Required to Be Tested ML20043E8831990-06-0505 June 1990 Requests NRC Authorization to Use Plugs Fabricated of nickel-chromium-iron Uns N-06690 Matl Alloy 690 to Plug Tubes in Steam Generators of Plant ML20043D0451990-05-30030 May 1990 Discusses Proposed Rev to Tech Specs Re Facility ESF Actuation Sys Instrumentation Trip Setpoint,Per 900330 Ltr ML20042H0311990-05-0909 May 1990 Discusses Steam Generator Safety Assessment.Concludes That Continued Operation Through Remainder of Current Cycle 10 Fully Justified ML20042F0941990-04-30030 April 1990 Provides Addl Info Re Environ Impact of 900226 Application for Amend to License NPF-49,revising Tech Specs to Allow Containment Pressure to Increase to 14 Psia During Modes 1-4,per NRC Request ML20042F0661990-04-30030 April 1990 Responds to NRC 900404 Ltr Re Violations Noted in Safety Insp Rept 50-336/90-01 on 900120-0305.Corrective Action:Ler 90-004 Submitted on 900430 to Document Condition Prohibited by Plant Tech Specs ML20042E8331990-04-27027 April 1990 Forwards Annual Environ Protection Plan Operating Rept for 1989, & Monitoring Marine Environ of Long Island Sound at Millstone Nuclear Power Station Annual Rept 1989. ML20012E2681990-03-23023 March 1990 Responds to NRC 900226 Ltr Re Violations Noted in Insp Rept 50-423/89-23.Corrective Actions:Requirement to Review All Changes on Safety Sys for Potential Operating Procedure Changes Stressed to Operations & Engineering Personnel ML20012C3141990-03-13013 March 1990 Forwards Info Re Insp of Facility Emergency Operating Procedures,Per 900119 Ltr ML20012B4111990-03-0202 March 1990 Provides Addl Info Requested to Clarify Changes Proposed to Tech Spec Action Statements for Inoperable Accumulator B13453, Advises That Mods to Steam Generator Blowdown Sample Sys Completed,Per Violations Noted in Insp Rept 50-423/89-14 & Salp.Sys Will Be Evaluated for Next 2 Wks to Ensure Blowdown Radiation Monitor Operating Reliably1990-02-26026 February 1990 Advises That Mods to Steam Generator Blowdown Sample Sys Completed,Per Violations Noted in Insp Rept 50-423/89-14 & Salp.Sys Will Be Evaluated for Next 2 Wks to Ensure Blowdown Radiation Monitor Operating Reliably ML20011F7541990-02-26026 February 1990 Notifies That Jh Parillo Reassigned & No Longer in Need of License SOP-10263-2 as of 900219 1990-09-07
[Table view] |
Text
<.3 IMHtTHIIAST trrILITIIIS 3 H ARTFoRo. CONNECTICUT,06101 (203) 666-6911 L L J October 24, 1979 Docket No. 50-336 Director of Nuclear Reactor Regulation Attn: Mr. R. Reid, Chief Operating Reactors Branch #4 U. S. Nuclear Regulatory Commission Washington, D. C. 20555
References:
(1) E. L. Conner telecopy dated October 16, 1979.
(2) W. G. Counsil letter to R. Reid dated September 28, 1979.
Gentlemen:
Millstone Nuclear Power Station, Unit No. 2 Feedwater System Piping Please find enclosed as Appendix #1, Northeast Nuclear Energy Company's (NNECO) response to questions raised by NRC relative to fatigue crack growth and repair options applicable to the feedwater system piping cracks. This information is essentially comprised of that which was presented to the NRC Staff orally in Bethesda, on October 19, 1979.
During that meeting, a question was raised by the Staf f relative to the probable cause for crack initiation and the rationala for its present depth of approximately 100 mils. i.s a member of the Westinghouse Owner's Group on Feedwater Pipe Cracking, we are currentl3 involved in an extensive program aimed at assessing this cracking phenomena. Feedwater piping strain, temperature, and acceleration have been measured at several plant sites including Millstone Unit No. 2. This data has nct revealed ang significant dynamic strain which could be construed to have caused the crack initiation. Significant thermal stratification in the first horizontr.1 run of the feedwater piping has been measured at all plants during plant startup, coincident with low feedwater flow rates. Stresses calculated from this steady-state stratified thermal profile do not reveal levels sufficient to cause crack initiation or appreciable crack growth. It is our belief, however, that this existence of a zone of temperature instability is the cause of crack initiation and crack growth.
NNECO has had considerable experience with two-temperature fluid regimes in feedwater systems as a result of inspections performed on the Millstone Unit No.1 RPV feedwater nozzles beginning in 1974. This inspection and subsequent thermocouple data revealed a thermal instability phenomena of sufficient magnitude to initiate and grow tracks to a certain depth. At that time, ti - General Electric Company initiated a ecmprehensive program to rasolve these crackit.g occurrences.
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This program culminated in the publication of two (2) CE reports, NEDE-21480 and NEDE-21821, which document in detail, experimentation and analysis to support continued safe operation of these BWR plants.
PWR feedwater piping temperature measurements have not confirmed the presence of high cycle ( sul hz) water temperature fluctuations during these periods of thermal stratification. However, tests performed LA GE at their two-temperature test facility and the Moss Landing Facility demonstrate a trend tovatd thermal fluctuations in the vicinity of the two-temperature flui ! .erface. Assuming that fluid temperature fluctuations on the order of 300*F are present during low flow conditions (Millstone Unit No. 2 measurements reveal a potential of 350*F), it can be seen from Table 3-1 of NEDE-21821 that cracking can initiate af ter approximately 400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br /> of hot standby operation. It is estimated that the thermal skin stresses for carbon steel are approximauely 2/3 of tuose for stainless steel. The depth to which these skin effects penetrate is repre-sented by Figures 3-2 through 3-5. It is noteworthy that very little aetal response to wasar temperature fluctuation is felt beyond the 0.1 inch depth, pal ticularly for the higher frequency casec.
There is no evidence of a driving force which could cause appreciable crack growth beyond the 100 mil depth to occur in the Millstone Unit No. 2 feedwater piping system. This is supported by analytical and experimental ob;ervations of the Millstone Unit No. 2 system. Conditions which have caused Aarger crack growth at other plants could be:
(1) The aetallurgical condition of the material.
(2) The residual stress distribution.
(3) Counterbore geometry.
(4) External loadf ug conditions.
Regardless of the cause, it is important to note that none of the above conditions change with time. That is, the state of stress intensity factor at the feedwater pipe crack tips can be described by information obtained from the installed instrumentation and the ultrasonic examinations throughout future plant operation.
The reexamination of these crack indications is scheduled to be performed during the upcoming October, 1979 outage. This and the other planned interim examina-tions will confirm that the crack growth is within the predicted crack growth rates.
As stated in Reference (2), if any increase in crack depth is evidenced, vithin the tolerance of the ultrasonic examinations, the cracks will be repaired. The results of this inspection will b plotted and compared directly agair.sc those taken in August, 1979. It is judged that any appreciable change in crack front geometry or depth wLil be readily apparent.
With regard to operational philosophy of the auxiliary feedwater system during plant startup, NNECO will continue to operate in the continuous feed mode whenever possible. That is, once auxiliary feedwater flow is established, abrupt changes in flow rate will be avoided in order to minimize u e potential for crack propaga-tion.
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A permanent repair is planned for the 1980 refueling outage. At that time, sufficient data will have been reduced and causative mechanisms identified such that positive solutions can be employed to minimize the potential for further cracking. This approach results in the significant advantages of minimizing personr.el exposure and reducing plant outage time. Performing a 19 pair at this time is further complicated by the recent sick-out organize 3 by the local craft unions. Resolution of this situation may not recur for several weeks. To effect a repair at this time may require solicitation of non-union personnel and qualification and indoctrination to the various repair procedures. It is anticipated that this could add substan-tially to the duration of the outage, if the repair were to be undertaken at this time. Def erring the replacement, therefore, results in substantial man-rem and economic benefits.
Should NNECO's anticipation of no detectable crack growth be realized, the 1980 replacement program is concluded to be technically defensible and appropriate.
An interim ultrasonic examination, prior co March 1, 1980 will be conducted to verify the absence of crack growth. The leak detection equipment will remain functional and be monitored without change from the current program until the issue is permanently resolved.
We trust you find the above information responsive to the Reference (1) requests and questions raised by the NRC Staff during our October 19, 1979 meeting.
Very truly yours, NORTHEAST NUCLE /I FNERGY COMPANY
- l- 2 .
W. G. Counsil Vice President Attachments
c; e DOCKET NO. 50-336 APPENDIX #1 MILLSTONE NUCLEAR POWER ST..IION, UNIT NO. 2 RESPONSES TO ADDITIONAL INFORMATION REQUIRED TO ASCERTAIN ACTIONS NECESSARY REGARDING THE FEEDWATER SYSTEM PIPING 12 8 i .'.T OCTOBER, 1979
PURPOSE In response to the NRC request of October 16, 1979, the additional information required to ascertain actions necessary regarding the feedwater system, which was presented to the NRC on October 19, 1979, in Bethesda, Maryland, is hereby submitted.
Question 1 In your letter of September 28, 1979, you stated that thermal variations (stratification) was observed during low flow conditions. Address the potential for cracx propagation during these low flow conditions (low cycle f atigue).
Provide a quantitative analysis regarding crack growth rates during the thermal transient cycle.
Response
The inspections conducted in response to NRC I6E Bulletin No. 79-13 revealed linear circumferential indications adjacent to the steam generator feedwater nozzle safe-end to pipe and pipe to elbow welds in both feedwater piping loops. The largest observed linear indications near each weld were subsequently mapped by ultrasonic inspection in terms of through-wall depth and circumferential length. In addition, the arec of interest was instrumented in order to establish the mechanical and thermal feedwater piping loading conditions durin o startup and full-power operation. The results of the instrumentation data specifically applicable to Millstone Unit No. 2 are presented in Attachment #1.
Ynowing the crack sizes and orientation of the feedwater piping flaws, Westinghouse assessed the potential for crack growth in a conservative manner considering not only the original design basis transients but also the additional thermal loading derived from the instrumentation results at the subject unit. The assessment of crack growth for the linear indications in the feedwater piping was performed by Westinghouse and it is enclosed as Attachment #2.
The results of the conservative assessment indicate that the largest 11aw in the feedwater piping system will not grow significantly. Furthermore, the final flaw size for the worst location is a factor of five (5) smaller than the -
established critical flaw size.
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.- e Question 2 Assuming the analysis requested above predicts crack growth at a suf ficiently low rate to ensure adequate safety margins can be maintained until a permanent repair can be made at the June,1980 refueling outage, provide the details of an augmented inspection program which verifies that crack growth has not occurred at a rate faster than predicted by the analysis.
Response
The results of the crack growth assessment confirm that crack growth will occur at a sufficiently low rate. Therefore, adequate safety margins will be maintained until the June / July,1980 refueling outage.
The augmented inspection program for the observed feedwater pi 'ag flaws between Octobe) 31, 1979 and the next refueling outage was evaluated consistent with the present understanding of the feedwater piping cracking phenomena. Based on this and the results of the Millstone Unit No. 2 feedwater piping system instrumentation data, it is important to note that thermal loading conditions exist durf ag plant startup and plant shutdown operations which have the potential to induce further crack growth. Therefore, it is imperative that the inspection frequency be established consistent with the objective to minimize the potential for crack growth in the interim period.
Based on the above, it is proposed that the feedwater piping flaws be inspected as a minimum prior to March 1,1980 to verify that the actual crack growth is within the predicted crack growth. In addition, it is proposed to conduct additional inspections of the feedwater piping flaws at any plant cold shutdown of more than two (2) days duration between October 31, 1979 and the next refueling outage. However, the time interval between successive inspections shall be more than three (3) calendar months. All results of the additioral inspections conducted in the interim period will be submitted to the NRC within seven (7) days followf.ng the completion of the examination.
Question 3 In the proposed repair / replacement program you submitted, you stated that the removal of the shield wall section can be made within design bases limitations.
- a. Provide the technical information supporting your conclusion.
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Response
The original shield wall model was modified to include a six (6) foot diameter by twelve (12) i.ch deep cutout. Imposing the design basis loading conditions indicates that the shield wall cutout will not degrade the load bearing capa-bility of the shield wall. The design basis loading conditions were derived from the subject unit Final Safety Analysis Report.
Question 3.b Provide assurance that the method of removing part of the concrete wall by drilling and chipping will not damage the concrete left in place and the existing reinforcing bars. Describe the quality assurance procedures which will be used during thc concrete removal operation.
Response
The shield wall removal by drilling and chip,.nc sill be performed in accordance with the work procedures submitted to the NRC on September 28, 1979. Industry experience indicates that the proposed removal process will not damage the concrete left in place and the existing reinforcing bars.
As part of the QA Category I work procedures referenced above, Quality Control will monitor and perform a final inspection of the shield wall cutout to ensure that the shield wall removal was accomplished within the specified dimensional requirements and that the minimum specified covers exist for the specified rebar.
Question 3.c Describe the procedure which will be used if the reinforcing bars must be removed.
Response
A grinding disk will be used to remove reinforcing bars.
Question 3.d If replacement on the removed shield wall segment is required, ad/.res; the following:
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.: e (1) Define the concrete mix which will be used to fill the recess in the shield wall.
(2) Describe the procedure for reinforcirg bar replacement.
(3) Define the method to be used to ensure compatibility of the new and old concrete, especially the measures planned to limit shrinkage of the new concrete. Discuss the degree of working together that can be expected from the new concrete and existing wall.
Response
Based on the fact that the steam generator feedwater nozzle / safe-end to pipe weld ' will be subject to future periodic inspection, it is concluded that the shield wall cutout will remain as is to provide the required accessibility.
Question 4 Provide the details for material removal as discussed in repair Option B.
Also, provide the detailed procedures for the weld repair on the ID of the pipe should the wall thickness be reduced from Code limits. Address the mock-up used to qualify the welding procedures, for training and to qualify the welders / welding machine operators.
Response
The Option B repair method consists of pipe removal outside the shield wall and repair of the steam generator feedwater nozzle safe-end to piping indications from the ID.
CE Chattanooga has developed a grinding fixture which will be positioned inside the pipe. Prior to positioning of the grinding fixture, a liquid penetrant examination will be performed to map the inside pipe / safe-end area containing the linear indications. A six-inch grinding wheel attached to the positioning fixture will be used to remove any linear indications. The width of the grinding wheel is approximately one inch. After the grinding operation, a final liquid penetrant examination will be performed to ensure complete removal of the linear ind ications.
The weld repair will be performed with the Jiametrics internal welding fixture in accordance with applicable code requirements. The weld repair criteria will 1281. ::'
y be based on fatigue considerations rather than minimum wall thickness limitations specified by the ASME Section III code. Therefore, it is our intent to weld repair all ground-out areas from the ID in order to eliminate all stress intensi fication locations. The final weld repair contour will be polished and blended uniformly to the adjacent base material.
All welders will be qualified to the approved weld procedure. The weld procedure will be qualified in accordance with the applicable code requirements.
Question 5 Describe the simulation for welder training and qualification to account for the limited access between the shield wall and steam generator in Option A.
Provide any details regarding the consideration of automated welding to make the nozzle to pipe repair.
Response
The Option A repair requires the specified shield wall removal in order to gain access to the steam generator nozzle safe-end to pipe welds. After shield wall removal, the feedwater pipe will be cut between the steam generator nozzle to safe-end weld and the safe-end to pipe weld as specified in the work procedures submitted to the NRC on September 28, 1979.
The weld procedure for the Option A repair method will be qualified in accordance with the applicable code requirements. All welders will be qualified to position 6G (QW-405 ASME Section IX) which essentially qualifies them to perform welding in any position. In addition, all weldcrs will be trained in the steam generator nozzle / shield wall mock-up to simulate the welding conditions in the limited access area.
Based on our investigation of four manufacturers supplying automated welding tools, it is concluded that no automated welding tools exist at the present time which will physically fit into the limited access area.
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.: e Question 6 State if a UT baseline examination will be performed for the nozzle to piping weld if Option A is used.
Response
A baseline volumet' examination will be performed for the repair nozzle to piping welds (Option A) . However, based on the requirements for future inspec-tions of the subject velds by radiography and the fact that radiography has been used to detect the feedwater piping indications, it is concluded that the baseline volumetric examination will be accomplished radiog aphically.
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- DOCKET No. 50-336 ATTACHMENT #1 MILLSTONE NUCLEAR POWER STATION, UNIT No. 2 FEEDWATER SYSTEM PIPING 1281 .':1 OCTOBER, 1979
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SUMMARY
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PRCFILE 2 EVENT NUMBER OF OCCURRENCES 0F EVENTS(I) 0F PROFILE PROFILE -
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DOCKET NO. 50-336 ATTACHMENT #2 MILLS *IONE NUCLEAR POWER STATION, UNIT NO. 2 FEEDWATER SYSTEM PIPING 1281 .X' OCTOBER, 1979
s:
ASSESSMENT OF GROWTH OF FEEDWATER LINE FLAWS MILLSTONE II 0 d W.H. Bamford The purpose of this work is to estimate the future growth of a flaw located in the counterbore region near the feedwater nozzle safe-end-to-pip e vield. The flaw of interest has been confirced by UT to be apprcxirately 0.10 inches deep, and oriented circurferentially.
As a result of the location of this flaw, instrumentation was in-stalled to monitor the temperature fluctuations in one loop. Results shotted inat in a certain flow rate range the water stratifies, produ-cing significant stresses which are potentiall importcnt for crack growth.
The types of stratification produced were typical of those observed in other plants, but not as severe. The observed stratifications were classified under five different types, as shown in Figure 1. The tem-perature difference from top to bottom of the pipe for profile 1 was measured at about 350*F, whereas for other plants it has been found
'~~
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to be as high as 450*F.
~'
A three dimensional finite element stress analysis has been completed for each of the five temperature profiles in Figure 1, and transient studies have shown that the five profiles represent limiting conditions compared viith the stress results obtained for any transient step in be-tvieen the profiles.
To acco ,plish a fctigue crack growth analysis, the system design tran- -
sients for corral, upset and test conditions were corbired t:ith the cycies ci stress from stratification, which occurs durir.g hot standby c craticn. As sho.:n in Figure 2, there are approximately nine cycles of varicus degrees wi'ich for the purpose of this coalysis, we will a r s.- ., occur er:" tire hot st:r.dby cccurs.
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A, tabulation of the cycle types used in the crack growth analysis, along with applicable stresses, is provided in Table 1. Tables 2* through 5 show the stresses at various locations around the pipe as a result of the stratification.
- The actual stresses from the three dimensional analysis were used for the fatigue crack growth analysis, except in two cases, where compres-sive stresses far exceeded the yield stress in corression. The locat-ion is at the top of the pipe, and the condition occurs only when the pipe is nearly filled with cold water (profile 1) at low flow. For this case, tensile residual stress values were assumed to exist, equal to the yield strength. This is seen at locations 1 and 2 in Tables 2 and 4. This assumption is considered to be extremely conservative.
Crack growth was calculated at each of thirteen' locations around the pipe for periods of 1, 2, 3 and 4 years, assuming an initial flaw of 0.100 inches deep, extending entirely around the inside of the pipe.
' A fatigue crac.k growth law which accounts for mean stress or R ratio i
(omin./ e max.) as - ell as the presence of the water environment was used. The law is shown in Figure 3.
Results of the crack growth analysis are shown in Table 6 for each of the locations considered. These results show that the observe flaws will not grow significantly during the next years service. The final flaw size for the worst location is a factor of 5 smaller than the critical flew size for the pipe, us shor in figure 4.
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> i i i i 2 3 4 5 678910 20 50 40 50 60 70 EC 90 00 STRESS !!JTf t.'SITY Ft.CTOR R AfJGE 2, oK IKSI T I FATIGUE CRACK GROWTP DATA FOR SA.503, CliSS 2 ANb CLASS 3 AND SA 533, F4 #.E 2 GRADE B, CLASS 1 STE'ELS
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m: s TABLE -1 TRE';51Ef;TS USED IN FATIGt'E CRAtr. GE C'.5 H A*: ALYS15 MILLSTONE II ,
I Cycles Inside Surface (. tress Outside Surf..ce St-Min. Max.
l Description (40 years) Max. {
i Hot Standby 1 50 ) ~7 Hot Standby 2 1500 )
These stresses are dependent on circumferent' Hot Standby 3 500 ) position. See Tables 2 through 5.
Hot Standoy 4 2500 )
l 15000 10.47 7.71 8.49 7 Unit Load-Unload "
i Step Increase /0ecrease 2000 9.56 7.87 7.89 7 8.42 3.79 7.
Partial Loss of Flow 40 23.7 40
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23.02 8.42 3.18 7.
Loss of Lo~ad 22.69 8.21 2.88 7.
Reactor Trip 400 11.23 0.0 10.04 0.
' Secondary Leak Test 200 l
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STRESS RESULTS - AXIAL DIRECTION ,
g CONDITION 5 HOT STANDBY f1 h
Inside Surface Outside Surface Min. Max. Min.
Location Max.
(Ksi) (Ksi) t 40.0 0.0 -40.0 0.0 1
40.0 0.0 -40.0 0.0
! 2 9.46 -23.0 8.56 8.29 3
35.12, 4.63 11.23 7.02
'4 68.97 - 1.43 13:20 4.66 5
66.05 - 7.28 12.61 1.71 6
46.17 - 8.06 10.83 -0.33
[ 7 7.27 7.34 3.69 8 24.37
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Location Max.
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Inside Surface Max. Min.
Max. Min.
Location (Ksi)
(Ksi')
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Max. Min. _
' Location (Ksi)
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9.60 3.24
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