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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 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 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 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 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 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 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 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 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 ML20055G5331990-07-18018 July 1990 Forwards Decommissioning Financial Assurance Certification Rept 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 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 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 ML20055D3481990-06-29029 June 1990 Forwards Addl Info Re Facility Crdr & Isap,Including Justification for Human Engineering Discrepancies Dispositioned for No Corrective Action ML20055D7191990-06-29029 June 1990 Amends 900604 Rev 13 to QA Program 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 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 ML20006G1581990-02-21021 February 1990 Forwards Response to & Comments on Initial SALP Rept 50-423/88-99 for Period 880601 - 891015.Procedures Revised to Permit Operators to Adjust Area Monitors to Reduce Nuisance Alarms 1990-09-07
[Table view] |
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i February 7, 1990 Docket No. 50-336 B13443 Re: ECT Inspection Mr. William T. Russell Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406
Dear Mr. Russell:
Millstone Nuclear Power Station, Unit No. 2 Steam Generator Insoection On June 5, 1989,II) NRC Region I requested that Northeast Nuclear Energy
. Company (NNECO) perform an in-cycle steam gentrator eddy-current testing (ECT) examination for M On July 12,1989,gstone Unit No.
. NNECO 2 during responded bythe letter current to the Cycle Region10 operation.
I request and committed to perform an ECT inspection of the Millstone Unit No. 2 steam generators with an estimated shutdown time frame between September and October of 1989. . A more definitive gegedule and detailed woik scope was provided by letter dated August 30,-1989.
In continuing dialogue with the NRC Staff, NNECO had been requested to present the latest steam generator ECT results and an overall safety assessment of the steam generators to the Staff in a meeting to be held on February 8,1990, in the NRC Headquarters in White Flint, Maryland. In preparation for that meeting, NNEC0 intended to submit relevant information to allow the NRC sufficient time for review of this matter. On February 1, 1990, NNECO requested from the NRC Staff, a one-week extension for submittal of the steam generator ECT results and an overall safety assessment of the Millstone Unit No.: 2 steam generators. This extension was needed to finalize and document I
(1) W. T. Russell letter to E. J. Mroczka, " Millstone Unit No. 2 Mid-Cycle Steam Generator Inspection," dated June 5, 1989.
(2) E. J. Mroczka letter to W. T. Russell, " Millstone Nuclear Power Station, Unit No. 2, Steam Generator Inspection," dated July 12, 1909.
(3) E. J. Mroczka letter to W. T. Russell,
- Millstone Nuclear Power Station, Unit No. 2, Steam Generator Inspection," dated August 30, 1989.
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Mr. William T. Russell
> B13443/Page 2 1 February 7, 1990 NNECO's internal safety assessment. The NRC Staff granted this extension and subsequently- rescheduled the February 8,1990, meeting to February 22, 1990.
Attachment I to this letter provides NNECO's safety assessment for Millstone Unit - No. 2's steam generators with an overall conclusion, based upon our analysis of current Millstone Unit No. 2 testing results and other industry !
data, that continued operation through the current Cycle 10 is fully justi- l fied. '
Should you need additional information or clarification prior to the '
February 22, 1990, meeting, please contact us.
Very truly yours, 5
NORTHEAST NUCLEAR ENERGY COMPANY 4
E. 'Kffro'czka f Senitrr Vice President Attachment l
cc: G. S. Vissing, NRC Project Manager, Millstone Unit No. 2 i W. J. Raymond, Senior Resident Inspector, Millstone Unit Nos. 1, 2, and 3 U.S. Nuclear Regulatory Commission, Document Control Desk
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3 I Docket No. 50-336 B13443 .
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Millstone Nuclear Power Station, Unit No. 2 Steam. Generator Safety Assessment February 1990
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p Mr.' William T. Russell 513443/ Attachment I/Page 1 February 7,1990 SAFETY ASSESSMENT OF THE MILLSTONE 2 STEAM GENERATORS i
DirRODUCTION E
The purpose of this assessment is to evaluate the structura'l integrity of the
' Millstone Unit No. 2 (MP2) steam generators (SGs) with regard to safety and reliability. The SG tube bundle forms a portion of the primary pressure boundary. Corrosion induced degradation of the tubes and tube supports has occurred as a result of past operation of the unit. This degradation is identified by nondestructive inspection of the SG tubes. A in-cycle inspec-tion of the SG tubes has been recently completed. The intent of this addi-tional inspection was to identify any stress corrosion cracks which may have developed and to ascertain whether the corrosion process was under control.
The following discussion addresses the effect of known and potential degra-dation on the integrity of the pressure boundary and its implications with regard to safety and reliability for the remainder of the current operating .,
cycle. I DISCUSSION Corrosion in'duced degradation of the tubes and supports has been identified in I the MP2 SGs by various nondestructive and destructive inspections including ,
eddy current testing (ECT), ultrasonic testing, radiography, fiberoptics and !
removed tube examination. Corrosion of the tubes has occurred as a result of I
concentrations of corrosive chemical specias on the secondary side of the o tubes, primarily in sludge areas within 13 2nches from the top of the tube-
! sheet. Two corrosive mechanisms, pitting and stress corrosion, are principal-ly responsible for the existing tube degradation. Both the hot leg and cold leg sides of the SG have been affected by each mechanism to various degrees.
Tube support degradation has occurred principally on the hot leg side of all support levels. l o Pitting Pitting has occurred in the MP2 SG tubes as a result of acidic chemistry conditions and sludge build up on the tubesheet early in the life of the SGs. The pitting has occurred in all four SG plenums. Tubes with pits ,
identified by ECT as having depths of greater than or equal to 40 percent throughvall are either repaired or removed from service. Improvements in l secondary side water chemistry and chemical cleaning in 1985 which i removed a large portion of the sludge pile effectively stopped the L progression of pitting. Prior to 1985, a large inventory of pits less !
than 40 percent throughwall was created. These pits have not shown any j '
significant growth over the last several cycles, indicsting that the
- - conditions and environment which originally promoted the pitting no l longer exists.
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1 Mr. William T. Russell
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813443/ Attachment I/Page 2 ,
February 7,1990 i' o Tube Support Degradation Tube support degradation of the MP2 SGs has occurred as a result of acidic chemistry conditions early in the operating life of the SGs. The acid conditions caused the carbon steel material to corrode and form magnetite. The build up. of the magnetite corrosion product in turn caused denting of the tubes and shifting of the supports. Elimination of the corrosive acidic environment has stopped the further degradation of ,
the supports. The progression of support degradation, as measured by tube denting, has been approximately zero since Cycle 7. A fiber optic inspection of the first eggerate tube support was performed during the EOC9 Refueling Outage (RFO). This inspection revealed an eggerate con-dition much better than laboratory testing of the model boiler eggerates and. better than assumed in previous evaluations. Although tube bowing '
was noted, no loss of support was identified by the fiberoptic inspec-tion. No examination of the tube supports was performed during this outage.
o Stress Corrosion Cracking Stress corrosion cracking has developed in the MP2 SG tubes at the top of the tubesheet on both the hot let and cold leg sides. The cracking is OD initiated and circumferential1y oriented. The cracking was first dis-covered in a leaking tube (SG 1H L25/R19) in January 1987. At the time the safety significance of a circumferential1y oriented crack vas assessed and operation of the unit was concluded to be safe with a reduc-tion of the primary-to-secondary leak rate to 0.15 gpm.
Subsequent ECT inspections during the EOC8 RF0 identified 26 additional cracks, all in the SG 1 hot leg. In addition, during the outage, Tube L25/R19 was removed for examination. Based on the results of the exami-nation of L25/R19 and the ECT inspection results, the safety significance of the cracking was reassessed and found to be acceptable with a further reduction of the primary-to-secondary leak rate to 0.10 gpm.
Destructive examination of the leaking Tube L25/R19, removed after almost a year in operation while the tube was plugged, confirmed the presence of an OD initiated stress corrosion circumferential crack. The tube con-tained a continuous circumferential crack which was throughwall over 1908, and greater than 50 percent throughwall over 260*. The remaining 100' of the tube circumference contained several individual microcracks less than 50 percent throughvall. Based on examination of the fracture surface and review of MP2 SG chemistry, the best estimate of the causative chemical species was determined to be caustic.
Evaluation of the observed crack size indicated that the tube should not have been able to withstand operating loads and therefore the crack must have grown during the time the tube vas plugged. Careful evaluation of the fracture surface showed that the continuous segment ref the crack h , + - < - = a--r-- g - - ' -
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Mr.'Villiam T. Russell B13443/ Attachment I/Pege 3 February 7, 1990 +
actually initiated as approximately 20 separate microcracks which even- ,
tually linked up by stress corrosion. It was postulated that the crack developed by the following scenario:
- 1. Caustic concentrating in the sludge area initiated circumferential1y oriented microcracks on the OD surface of the tube..
- 2. Existing microcracks propagated as additional microcracks initiated.
- 3. As the microcracks propagated, the ID of the tube was breached when a crack was propagated between 14' and 21' around the circumference.
This aspect ratio is typical for stress corrosion of SG tubes.
- 4. The throughwall portion of the crack enlarged and the tube began to leak. The remaining ligaments between the microcrack provided sub-stantial structural strength. The initial 0.10 gpm leak rate indi-cated a throughwall opening of approximately 35'.
- 5. The tube was plugged. ,
- 6. The remaining ligaments between microcracks stress corroded during the operating period while f.he tube was plugged, forming the observed continuous crack.
The postulated cracking scenario was reviewed with Vestinghouse Electric Corporation. Vestinghouse agreed that the postulated scenario was reasonable and closely matched results observed for tests of circumfer-ential cracks in North Anna SG tubes. Using the postulated cracking scenario and the observed leak rate, calculations were performed which demonstrated that the crack in Tube L25/R19 shoved acceptable " leak before break" behavior and at the initial leak rate of 0.10 gpm main -
tained the structural margins required by Regulatory Guide 1.121. -
Corrective actions to control the stress corrosion cracking af the tubes were instituted in Cycle 9. These actions consisted of on-line botic ,
acid addition, begun approximately two months into Cycle 9, and further l improvement in bulk vater chemistry impurity levels. The addition of P boric acid was expected to slow the propagation of incipient caustic
! stress corrosion cracks and halt the formation of nev caustic stress corrosion cracks. The unit successfully operated the entire Cycle 9 p without the development of any primary-to-secondary leaks.
l l An inspection of the MP2 SGs was performed during the EOC9 RF0 using the best equipment and techniques available in the industry at the time.
l' Improvements in the equipment and techniques had been made since the l- previous refueling outage inspection. Following the completion of all l
EOC9 nondestructive testing, E9 tube ends were identified as being l cracked at the top of the tubesheet. Cracks were identified in all four L
plenums with 85 percent of the cracking taking place in SG 1, with a l
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Mr. Villiam T. Russell 513443/ Attachment I/Page 4 February 7, 1990 60 percent /40 percent split between the hot leg and cold leg, respec-tively. The location of cracking within the- SG tube bundle corresponds with both a corrosive environment and high tube stress. The sludge pile provides the corrosive environment, while high tube stress develops as a c- result of tubesheet denting and support corrosion. The support corrosion causes shifting of the tube bundle and boving of the tube. Tube boving has been demonstrated by the inability to successfully pass long rigid objects (such as sleeves and stabilizers) through the top of.tubesheet region of some tubes. The fact that. support corrosion as measured by tube denting has been inactive over the last three operating cycles pro-vides reason to suspect that the tube stresses at the top of the tube-sheet have been present since 1985 or earlier and are not increasing.
The identification of 309 cracks during the EOC9 RF0 created the impres-sion that the number of cracks was rapidly increasing. However, increased rotating probe testing (468 tube ends EOC8 versus 12,556 tube ends EOC9), reduced probe speed and a tighter crack determination criteria contributed to the large number of cracks detected during the !
'EOC9 RFO. In addition, the criteria for determining which tubes to test !
vith -the rotating probe technique changed from bobbin Distorted Roll !
Indications (DRI) EOC8 to 100 percent of the crack boundary region EOC9.
An added effect was that the EOC9 rotating probe test was run at a slover i speed than during EOC8 (200 rpm and 0.1 inch per second versus 300 rpm I and 0.2 inch per second). The slover speed reduced the level of noise !
generated by the rotation of the probe past the tube transition and allowed for cracks of smaller signal amplitude to be more easily detected. A comparison of the limited E0C9 ultrasonic test results with !
the EOC9 rotating probe results demonstrated that the rotating probe data F analysis criteria used at EOC8 did not identify certain cracks, therefore, the analysis criteria was tightened.
l l The number and apparent growth of cracks in the cold leg did not appear l l to be consistent with the expected Arrhenius behavior with temperature !
i- typical of caustic induced stress corrosion. Normally, the hot leg would be expected to show a larger growth rate. One possible explanation was !
that the cold leg cracks which developed to a state identifiable by bobbin coil ECT during E0C9 RF0 actually initiated during Cycle 8 vhile I caustic chemistry conditions existed in the SGs, and propagated slover .
l due to the lover temperature. Variation in sludge consistency between l l the hot leg and the cold leg, variations in diffusion parameters with l .
temperature, solubility variations with temperature, variations in boric ,
acid effectiveness with temperature and other unknown factors also may !
have contributed to the observed cold leg cracking. This indicated that additional cold leg crack propagation could be expected to occur during Cycle 10.
Based on review and interpretation of the E0C9 inspection results, the corrective actions to mitigate stress corrosion of the tubes were con-sidered to be effective. The historical ECT review indicated a dovnvard trend in the rate of cracking. The cracks which were identified as new
- i. at the EOC9 vere consistent with the propagation of incipient caustic
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Mr. Villiam T. Russell
' B13443/ Attachment I/Page 5 February 7, 1990 cracks in a boric acid environment. The safety significance of the '
cracking was reassessed following the EOC9 RF0 inspection and found to be '
acceptable.- Projections of the number of new cracks expected to form in Cycle 10 were made assuming effective corrective actions. . On a best-estimate- basis, 60 new cracks were expected by the in-cycle inspection and 75 new cracks by EOC10. Upper bound estimates were 200 new cracks by '
the in-cycle inspection and 250 new cracks by E0010. If an active cor-rosion process were assumed, over 1,000 new cracks would have been expected by EOC10.
During October 1989..MP2 was shut dovn to perform an in-cycle inspection 1 of the SGs. The purpose.of this inspection was to identify stress cor-l rosion cracks which had developed to a detectable size since the EOC9 RF0 '
inspection. and to ensure that the corrosion process responsible for the cracking was under control. The inspection was performed with the best equipment and techniques available at the time and reflected improvements which had been made since the EOC9 RF0 inspection. The inspection pro- '
gram consisted of an examination for cracks by rotating pancake coil probe testing of all tubes within the region where cracks vould be expected to occur, plus bobbin coil probe testing for pits. In addition, ultrasonic tests were performed on tubes which RPC indicated that cracks were possibly present to obtain additional information on crack length and depth.
A total of 104 tube _ ends were identified as being cracked at the top of the tubesheet. Cracks were identified in all four plenums, with 90 percent of the cracking taking place in SG1, with an approximate 60 percent /40 percent split between the hot leg and cold leg, respec-v tively. Figure 1 illustrates the location and number of cracks. in a representative SG plenum. The location and distribution of the cracking is almost identical to the EOC9 RF0 inspection results. All tubes with identified cracks were staked and plugged. Determination of whether a tube was cracked was made conservatively and it is possible that some of the repaired tubes may not actually contain cracks. An additional 11 tubes were preventatively staked and plugged, since a positive determi-nation could not be made that a crack was not present in these tubes.
+
A comparison of the February 1989 (E0C9) and October 1989 rotating probe data was performed for all but six tubes identified as cracked during the October examination. No rotating pancake coil probe data was collected in- February on six cracked tubes. The extrapolated results from this comparison identified 35 of the 104 cracks (35 percent) as not being detectable by ECT in February. A total of 69 cracks (65 percent of the total) were present in February. Identification of the cracks using February 1989 data was possible by lovering the threshold for crack determination to the level equivalent to that used for the October 1989 examination. The reasons for identifying cracks in October which vere present (but not identified) in February included:
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'Mr."Villiam T. Russell '
B13443/ Attachment I/Page 6 February 7,1990 ,
i 1. Improved training of the analysts prior to the October examination,
- 2. Independent review of data tapes in October versus an independent
review of RPC graphics only, in February, 3.. Lovering the initial reporting threshold to include all "possible" ,
cracks versus "best estimate" cracks. ,
- 4. Ultrasonic testing of all tubes with "possible" cracks as identified by rotating pancake coil probe.
The small number and dovnvard trend of new cracks present during the October 1989 inspection is consistent with the best estimate prediction.
This prediction was based on the observed effect of boric acid treatment on caustic stress corrosion in laboratory tests. The addition of boric acid to the MP2 SGs was expected to halt the initiation of new cracks, and slow the propagation of existing incipient cracks. The 35 new cracks are likely cracks which vere at an incipient stage prior to the boric acid addition and have now propagated to a detectable size.
Based on results from the previous _three inspections, the cracks in the MP2 SG tubes can be characterized as follows: the cracks are located at the top of the tubesheet, within approximately 0.2 inches from where the expansion transition meets the nominal tube diameter, and are circumfer-entially oriented. The macrocrack, as defined by rotating probe ECT, is ;
j made up of several discontinuous microcracks which are separated by liga-ments of sound material. Circumferential extents of the macrocracks ranged from 14' to 329' by RPC and up to 360' by ultrasonic testing (UT).
Crack depths up to 100 percent throughvall vere estimated by UT. The ,
discontinuous nature of the array of microcracks was confirmed by UT and removed tube examinations. In general, the microcracks were located at ,
different axial planes, typically within a 0.1 inch band. A tube with a discontinuous array of microcracks is substantially stronger than a tube with a continuous crack of equivalent circumferential extent and depth.
, The actual circumferential extent and depth of the macrocrack is rela-L tively independent of the operating time. The circumferential extent is
! controlled by the number of microcracks which have initiated around the I
circumference and the uniformity of the stresses in the tube at the top of the tubesheet. The depth of the individual microcracks is also con-trolled by the stress field. Once a microcrack has passed the incipient stage, the crack vill propagate to the depth supported by the stress field fairly rapidly compared to the length of an operating cycle.
Projection of the number of cracked tubes expected by the EOC10 vere made by extrapolating the current cracking trends. On this b, sis, 45 tubes '
l' vill be identified as cracked during the E0C10 RF0 inspection. The cracks are expected to be the result of either the propagation of incip-ient cracks, which developed during the time period caustic conditions
- _ _ _ ~ . _ _ . _ _ _ _ , - __.__
i Mr. Villiam T. Russell 313443/ Attachment I/Page 7 '
February 7, 1990 existed in the SGs, or the identification of existing cracks which vere previously below NDE detection thresholds.
A distribution of the circumferential extent of the macrocracks, as indi-cated by RPC, has been present for each inspection where cracks vere observed. The distribution of circumferential extents has remained the same, regardless of the operating. period between inspections. This is illustrated in Figure 2. This indicates that once macrocracks have formed, the overall population vill not show any significant growth in circumferential extent. The distribution of the cracks observed during the October 1989 inspection can be modeled by a gamma distribution with a median of 90', a mean of 102', and a standard deviation of 67', Figure 3.
It. is expected that the new cracks projected to be identified during the EOC10 refueling outage vill continue to follow a gamma distribution of the circumferential extents. Cracks with large circumferential extents are potentially more challenging from a structural standpoint. Using the gamma distribution model, statistical inferences can be made with regard to the number of observed macrocracks which vill have large circumfer-ential extents. Given a projection of 45 cracks,.the number of macro-cracks which vill exceed 240' at the 95 percent confidence level is four and the expected number is two, Figure 4. To date 36 cracks have been found which exceed 240' in circumferential extent, including the tube which leaked at 0.1 gpm in January 1987 (Tube L25/R19 254' circumfer-ential extent).
Three tubes were removed from the MP2 SG during the October 1989 shutdown for burst testing and destructive examinations. The tubes contained the '
following circumferential cracks:
TUBE RPC EXTENT UT EXTENT UT MAX DEPTH SG2 Cold L118/R14 291' 340' Clustar 80%
SG1 Cold L52/R22 191' 155' 80%
110' 60%
SG1 Cold L145/R23 88' 80' 40%
65' 40%
Tube SG 2C L118/R14 was considered a vorst case crack based on the combi-nation of circumferential extent and depth as indicated by nondestructive inspections. This tube was pressurized with water at increasing pressure until bursting at a pressure of 11,200 psi. The failure was axial, coin- .
ciding with a groove created by the removal process. Tube SG IC L52/R22 was also tested, developing a leak at 8500 psi. This tube could not be burst because of test equipment limitations. A virgin tube vould be expected to fail axially at a burst pressure between _1,000 psi and 12,000 psi.
,1
Mr."Villiam T. Russell B13443/ Attachment I/Page 8 February 7, 1990 The burst. pressures of the removed tubes compare quite favorably with the normal operating differential pressure between the primary and secondary side -of the tube of 1375 psi and the maximum accident pressure, with dynamic effects considered, of 3150 psi. Applying .the Regulatory Guide 1.121 safety factors of 3 on normal operating loads and 1.4 on accident loads produce pressures of 4125 psi and 4450 psi respectively. Despite the large extent of cracking and potential weakening of the tube as a result .of removal from the SG, both of the tested tubes fully met the Regulatory Guide requirements.
For circumferential cracks of the type and location found in the MP2 SG, a substantial portion of the cross-sectional area must be corroded before the tube would no longer fail axially at virgin tube pressure in a burst test. The presence of the tubesheet provides restraint to bulging of the tube in the region of the crack and effectively increases the corroded area which can be tolerated without loss of margin to burst. In addi-tion, the stiffness of the tube, the tube supports, and the presence of ,
adjacent tubes restrain bending of asymmetric circumferential cracks, '
such that the corroded area averaged over the circumference of the tube can 'be used to provide a reasonable approximation of the burst strength of the tube. It is estimated that a tube vill continue to fail axially with no loss of margin to burst with circumferential cracks up to approx-imately 58 percent throughvall averaged over 360'. The Regulatory Guide 1.121 safety factors to burst continue to be met by a uniform 77 percent throughvall 360' circumferential crack. These estimates do not consider the additional strength provided by ligaments between the microcracks and the presence of microcracks at separate elevations.
Regulatory Guide 1.121 requires an operational allowance for tube degra-dation that may occur before the next scheduled tube inspection. SG tube rupture scenarios are restricted to single tube rupture events. There-fore, the operational allowance must incorporate both leak-before-break and frequency challenge criteria. The allovable operating period before the next inspection outage vill be the length of time required to accumu-late not more than one crack which could potentially exceed Regulatory Guide 1.121 margins.
Following the in-cycle inspection, UT depth information was analyzed to determine the depth profile of the cracks. This information was then used to calculate an equivalent throughwall degradation when averaged over the 360' circumference of the tube. Equivalent depths vere also calculated for the tube which leaked in January 1987 and the tubes removed for destructive examination. Similar to the distribution of circumferential extents, the distribution of equivalent depths was modeled by a gamma distribution with a mean of 22.3 percent throughvall and a standard deviation of 15.4 percent throughvall, Figure 5. A similar . distribution of equivalent depths is expected for any cracks which are identified in the future. Using this distribution, statistical inferences can be made with regard to the number of tubes with large equivalent depth cracks. Given a projection of 45 cracks, not more than
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.Mr. Villiam T. Russell B13443/ Attachment I/Page 9 February 7, 1990 two cracks are expected to have equivalent depths greater than 58 percent throughwall, Figure 6. In addition, not more than one crack vill have an equivalent depth greater than 77 percent at the 95 percent confidence level. Even this postulated low probability crack would not erode Regulatory Guide 1.121 margins due to mitigating factors known to be present- such as the discontinuous nature of actual cracks and the strengthening provided by the tubesheet.
The tube with the array of cracks with largest equivalent depth was tube 1H L25/R19, which leaked at 0.1 gpm in January 1987. This tube had an equivalent depth of 73 percent throughvall when averaged over 360' as indicated by the destructive exam. Tube 2C L118/R14, which burst axially in testing at 11,200 psi, had the largest equivalent crack of the tubes i removed in October 1989, estimated at 55 percent throughwall by UT and j 49 percent throughwall by the destructive examination. All of the l remaining tubes for which UT data was available had equivalent depths less than 55 percent throughvall, with the exception of tube IC L45/R13 and tube IC L28/R36 which had estimated equivalent depths of 70 percent throughwall and 60 percent throughvall, respectively. In all cases, the Regulatory Guide 1.121 margins to burst were satisfied, and in all but three cases, the tube vould fail axially in a burst test without loss of ;
margins when compared to a virgin tube.
The results of all nondestructive and destructive testing performed to date, including the burst tests, fully support the assumptions regarding the corrosion mechanism and crack morphology used to determine '.he allow-able leak rate of 0.1 gpm. The vast majority of cracks found to date l have been relatively small in circumferential extent and shallow in equivalent depth, such that margins to burst have not been affected. The o few cracks with large extents and equivalent depth either exhibited or i
vould be expected to exhibit a leak before break behavior. The number of cracks identified by the inspections has decreased by a factor of 3, i demonstrating that the corrosion process is understood and under control.
The redection in the number of cracks further reduess the risk of a tube rupture and enhances safety. Improved crack detection methods and con- i servative plugging criteria assure that existing cracks, which could ;
potentially compromise Regulatory Guide 1.121 requirements, have been detected and plugged. The risk of a tube rupture is considered to be extremely lov. s In summary, based on our conservative assessment of the in-cycle steam generator tube inspection results (UT and ECT) and the available burst test data, it is concluded that the observed stress corrosion cracking j intent of the applicable structural integrity and meets the primary-to-secondary leakage criteria of Regulatory Guide 1.121.
CONCLUSIONS Continued operation of the MP2 SGs for Cycle 10 vith regard to concerns associated with degraded tube and support conditions has been evaluated and
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Mr.'Villiam T. Russell B13443/ Attachment I/Page 10 February 7, 1990 ,
found. to be safe in accordance with applicable regulatory guidelines. ECT performed during the October 1989 outage has identified tube defects due to
. corrosion which could potentially compromise the margins of safety required by Regulatory Guide 1.121. .These identified defects have been removed from service. Pitting and cracking of MP2 SG tubes have been analyzed and found to exhibit acceptable leak before break behavior. If additional pitting or cracking. were to develop'during the remainder of the operating cycle, struc-tural_ integrity of the tube primary pressure boundary. with the appropriate '
margins of safety required by Regulatory Guide 1.121, is maintained provided ,
the 0.1 gpm primary-to-secondary leakage rate of the MP2 technical specifica-tions is. not exceeded. Leakage in excess of the technical specification limits requires shutdown of the unit, ensuring safety.
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'CIRCUMFERENTIALLCRACK EXTENTS .
- ;
MILLSTONE 2 STEAM GENERATOR TUBES i NUMBER OF CRACKS
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MACROCRACK CIRCUMFERENTIAL EXTENT FIGURE 3 9
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_ ___ _______________________2__mm______..__ _ _ _,_ _ , ______g
PROJECTED CIRCUMFERENTIAL CRACK EXTENTS MILLSTONE 2 STEAM GENERATOR TUBES FOR A POPULATION OF 45 CRACK 8 EOC10 NUMBER OF CRACKS 10 94 95% CONFIDENCE 8_
9M CONMENCE O EXPECTED NUMBER 7 -
1 6
t 5 -
4
1 3 -
l 2 -
1
- ' ' '
O 180 210 240 270 300
! CIRCUMFERENTIAL EXTENT (DEGREES) !
! FIGURE 4 '
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CIRCUMFERENTIAL-CRACK DEPTHS MILLSTONE 2 STEAM GENERATOR-TUBES
. UT EXAM NUMBER OF CRACK 8 20 GAMMA DISTRIBUTION PjPJ L25/R19 (LEAK) 15 -
W- OCT-1989-INSPECTION MEAN = 22.3%
- 10 -- - - - - - -
sTAnoARo ouviations is:4w 1
5 - -
0 I -
O 5 10 15 20 25 30 35 40 45 50 55 60 65 70 75 80 85 90 95100 AIVERAGE CRACK DEPTH (% TW)
FIGURE 5
_1 PROJECTED CIRCUMFERENTIAL CRACK DEPTHS-MILLSTONE 2 STEAM GENERATOR TUBES FOR A POPULATION OF 46 CRACK 8 EOC10..
NUMBER OF CRACKS 8 -
% 95% CONFIDENCE 7
, !' 90% CONFIDENCE i 6- O EXPECTED NUMBER l
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l 45 50 -- 55 60 65 70 77 MERAGE DEPTH (% THROUGHVELL)
FIGURE 6
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