Similar Documents at Surry |
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Category:CORRESPONDENCE-LETTERS
MONTHYEARML18152B3571999-10-22022 October 1999 Requests Relief from Temporary Repair of Through Wall Leak Discovered on 30 Inch Component Cooling Heat Exchanger Discharge Pipe Associated with Service Water Sys Common to Surry Units 1 & 2 ML18152B3581999-10-14014 October 1999 Submits Response to Violations Noted in Insp Repts 50-280/98-201 & 50-281/98-201.Corrective Actions:Visual Insps Were Completed on Accessible Coatings Inside Containment for Both Units 1 & 2 ML18152B3541999-10-12012 October 1999 Requests Use of Code Case N-532 & N-619,per Provisions of 10CFR50.55a(a)(3).Detailed Info Supporting Request,Encl. Attachment 2 Includes Technical White Paper That Provides Further Technical Info ML18152B3401999-09-27027 September 1999 Requests That Ma Walker Be Removed from List of Individuals Scheduled to Take Exam IAW Guidance Provided in NUREG-1021, Operator Licensing Exam Stds for Power Reactors ML18152B3391999-09-27027 September 1999 Forwards Revised Relief Request IWE-3,which Now Includes Addl Visual Exam Requirement After post-repair or Mod Pressure Testing Is Completed,Per Telcon with NRC ML18152B3381999-09-21021 September 1999 Forwards in Triplicate,Applications for Renewal of License for Bf Jurewicz & JW Heide.Without Encls ML18152B3331999-09-17017 September 1999 Forwards Revised 180-day Response to NRC GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves. ML18152B6671999-09-17017 September 1999 Forwards Two NRC Forms 536,containing Info on Proposed Site Specific Operator Licensing Exam Schedules & Estimated Number of Applicants Planning to Take Exams And/Or Gfes,In Response to NRC Administrative Ltr 99-03 ML18152B3341999-09-16016 September 1999 Requests Relief from Specific Requirements of Subsection IWE of 1992 Edition with 1992 Addenda of ASME Section XI Re Containment Liner Examination Requirements,For North Anna & Surry Power Stations ML18152B4521999-09-14014 September 1999 Forwards Comments on Review of Preliminary Accident Sequence Precursor Analysis of Operational Event That Occurred at Plant,On 980508,as Reported in LER 98-009 ML20216F1381999-09-0808 September 1999 Forwards Retake Exam Repts 50-280/99-302 & 50-281/99-302 on 990824.One SRO Applicant Who Received re-take Operating Test Passed re-take Exam ML18152B4501999-09-0808 September 1999 Submits in Triplicate,Application for Renewal of License for Rd Scherer,Iaw 10CFR55.57.Requests That Certification of Medical Exam by Facility Licensee,Nrc Form 396,be Withheld from Public Disclosure,Per 10CFR2.790(a)(6) ML18152B4471999-09-0101 September 1999 Requests That NRC Remove Listed Labels from Distribution ML18152B3711999-08-27027 August 1999 Forwards LER 99-005-00,per Plant TS Table 3.7.6.Rept Has Been Reviewed by Station Nuclear Safety & Operating Committee.Commitment Made by Util,Listed ML18152B3851999-08-23023 August 1999 Forwards Revised TS Basis Pages for TS 3.1.B,deleting Reactor Vessel Toughness Data Duplicated in Ufsar.Ref to Applicable UFSAR Section Included in TS Basis ML18152B3661999-08-20020 August 1999 Provides Medical Status Rept for E Washington,As Required by License Conditions.Summary of E Washington Current Physical Exam & Pertinent Lab Data Attached.Encl Withheld,Per 10CFR2.790(a)(6) ML18152B3651999-08-20020 August 1999 Requests Removal of License Condition from Sh Wightman Operator License SOP-21538.Updated NRC Form 396 Is Encl.Form NRC 396 Withheld,Per 10CFR2.790(a)(6) ML18152B3681999-08-20020 August 1999 Submits 30-day Rept Re Two Instances in Which Conditions of Approval in Coc Were Not Observed in Making Shipment.Two Type B Shipments Using Model CNS 8-120B Package Were Made After Expiration of QA Program Approval Between 990531-0628 ML18152B3801999-08-18018 August 1999 Forwards Technical Rept NE-1206,Rev 0, Surry Unit 2,Cycle 16 Startup Physics Tests Rept, Summarizing Results of Physics Testing Program Performed After Initial Criticality on 990525 ML18152B3781999-08-13013 August 1999 Forwards ISI Summary Rept for Surry Power Station,Unit 2 for 1999 Refueling Outage.Rept Provides Summary of Examination Performed During Outage for Third ISI Interval.No New Commitments Were Made ML18152B3751999-08-13013 August 1999 Forwards LER 99-004-00,IAW 10CFR50.73.Commitment Made by Util,Listed ML20210Q7661999-08-0606 August 1999 Requests Exception to 10CFR50.4 Requirement to Provide Total of Twelve Paper Copies When Submitting Surry & North Anna UFSAR Updates.Seek Approval to Submit Only Signed Original & One CD-ROM Version,Per Conversation with J Skoczlas ML18152B4081999-08-0606 August 1999 Forwards Response to NRC 990520 & 0525 RAIs Re North Anna & Surry Responses to GL 95-07, Pressure Locking & Thermal Binding of SR Power-Operated Gate Valves. ML18152B4091999-08-0505 August 1999 Forwards Vepc semi-annual fitness-for-duty Program Performance Data Rept for 990101-990630,IAW 10CFR26.71(d) ML18152B4001999-07-29029 July 1999 Requests Relief from Certain Impractical Requirements of ASME Section XI Code Associated with Partial Exams Conducted During 1998 Surry Unit 1 Refueling Outage.Relief Request SR-020 Encl ML18152B3981999-07-28028 July 1999 Forwards 60-day Response to GL 99-02, Lab Testing of Nuclear-Grade Activated Charcoal. Commitments Contained in Ltr,Listed ML18152B3971999-07-26026 July 1999 Provides Estimates of Licensing Actions Expected to Be Submitted in Fys 2000 & 2001,in Response to NRC AL 99-02 ML18152B3961999-07-23023 July 1999 Forwards Preliminary,Uncertified License Application & Medical Certification for License to Operate Surry Power Station Units 1 & 2 for Ds Cobb.Encl Withheld,Per 10CFR2.790 (a)(6) ML18151A6281999-07-23023 July 1999 Forwards Revised Epips,Including Rev 19 to EPIP-4.02,rev 14 to EPIP-4.16,rev 8 to EPIP-4.21 & Rev 7 to EPIP-4.30.EP & EPIPs Continue to Meet Stds of 10CFR50.47(b) ML18152B3991999-07-23023 July 1999 Requests That License for Jz Laplante Be Canceled as License Is No Longer Required ML18152B3931999-07-16016 July 1999 Forwards Updated NRC Form 396 & Ltr,Which Documents Medical Status of Mb Gross,License SOP-20476-2.Encl Withheld from Public Disclosure,Per 10CFR2.790(a)(6) ML18152B4401999-07-0101 July 1999 Informs NRC That on 990511,Dominion Resources,Inc,Executed Amended & Restated Agreement & Plan of Merger with Consolidated Natural Gas Co ML18152B4371999-06-24024 June 1999 Forwards Response to NRC Request for Clarification of Relief Requests Submitted on 990212,requesting Relief from Performing Hydrostatic Testing for Certain Small Diameter Class 1,RCS Pressure Boundary Connections ML18152B4361999-06-22022 June 1999 Forwards Response to RAI Re Surry & North Anna Power Stations,Units 1 & 2,per GL 96-06 ML18152B4301999-06-0303 June 1999 Informs of Util Intention to Revise Schedule for Submittal of License Renewal Applications for Surry & North Anna Power Stations to March 2002 ML18152B4261999-05-28028 May 1999 Provides Formal Notification of Effect of Recent Organizational Restructuring on OLs of North Anna & Surry Power Stations,Per NRC 990513 Telcon Request ML18152B4221999-05-27027 May 1999 Forwards Info Concerning Changes to ECCS Evaluation Models & Application in Existing Licensing Analyses for Surry & North Anna Power Stations,Units 1 & 2 ML18152B4231999-05-26026 May 1999 Informs That Vepc Will Revise 180 Day Response to NRC GL 96-05,within 120 Days of Date of Ltr to Incorporate Commitment to Participate in Joint Owners Group Program as Applicable ML18152B4211999-05-25025 May 1999 Forwards Rev 1 to Relief Request P-11 to Clarify Original Intent of Request by Specifically Requesting Relief from Requirements of Section 6.1 of OM-6 ML18152A4741999-05-19019 May 1999 Forwards Completed Registration Form for Renewal of ASTs at Surry Nuclear Power Station,Iaw Section 9VAC 25-91-100.F ML18152B4171999-05-17017 May 1999 Provides Notification of Number of Steam Generator Tubes That Were Plugged During Spring 1999 Refueling Outage Planned ISI ML20217D6621999-05-14014 May 1999 Forwards NRC Operator Licensing Exams 50-280/99-301 & 50-281/99-301 (Including Completed & Graded Exams) for Tests Administered on 990329-0401 & 990412-15.Nine Candidates Passed (& One Failed) Exam ML18152A3701999-05-13013 May 1999 Submits Proposal to Use Provisions of ASME Section XI Code Case N-597 for Analytical Evaluation of Class 1,2 & 3 Carbon & Low Alloy Steel Piping Components Subjected to Wall Thinning as Result of Flow Accelerated or Other Corrosion ML18152B4121999-05-0303 May 1999 Forwards Application for Renewal of License for SV Ross. Encl Withheld Per 10CFR2.790(a)(6) ML18152B4101999-04-29029 April 1999 Forwards Scope & Objectives for 990803 Surry Power Station Emergency Exercise.Without Encls ML18152B6561999-04-23023 April 1999 Forwards Annual Radioactive Effluent Release Rept for Surry Power Station,Jan-Dec 1998, Which Includes Summary of Quantities of Radioactive Liquid & Gaseous Effluents & Solid Waste Released During CY98 ML18152B6491999-04-13013 April 1999 Forwards MOR for Mar 1999 for Surry Power Station,Units 1 & 2.MOR for Feb 1999 Incorrectly Stated Gross Electrical Energy Generated (Mwh) for Unit 2.Rept Should Have Stated Monthly Figure as 568965.0 ML18151A5851999-03-31031 March 1999 Forwards Rept on Status of Decommissioning Funding for Each of Four Nuclear Power Reactors,Per 10CFR50.75(f)(1) ML18152A2801999-03-30030 March 1999 Forwards Summary of Structural Integrity Evaluation of Thermally Induced Over Pressurization of Containment Penetration Piping During DBA for SPS & Naps,Units 1 & 2,per GL 96-06.Draft Proposed UFSAR Revised Pages,Encl ML18153A2721999-03-29029 March 1999 Forwards LER 99-002-00 Per 10CFR50.73.Listed Commitments Contained in Ltr 1999-09-08
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARML18152B3571999-10-22022 October 1999 Requests Relief from Temporary Repair of Through Wall Leak Discovered on 30 Inch Component Cooling Heat Exchanger Discharge Pipe Associated with Service Water Sys Common to Surry Units 1 & 2 ML18152B3581999-10-14014 October 1999 Submits Response to Violations Noted in Insp Repts 50-280/98-201 & 50-281/98-201.Corrective Actions:Visual Insps Were Completed on Accessible Coatings Inside Containment for Both Units 1 & 2 ML18152B3541999-10-12012 October 1999 Requests Use of Code Case N-532 & N-619,per Provisions of 10CFR50.55a(a)(3).Detailed Info Supporting Request,Encl. Attachment 2 Includes Technical White Paper That Provides Further Technical Info ML18152B3391999-09-27027 September 1999 Forwards Revised Relief Request IWE-3,which Now Includes Addl Visual Exam Requirement After post-repair or Mod Pressure Testing Is Completed,Per Telcon with NRC ML18152B3401999-09-27027 September 1999 Requests That Ma Walker Be Removed from List of Individuals Scheduled to Take Exam IAW Guidance Provided in NUREG-1021, Operator Licensing Exam Stds for Power Reactors ML18152B3381999-09-21021 September 1999 Forwards in Triplicate,Applications for Renewal of License for Bf Jurewicz & JW Heide.Without Encls ML18152B3331999-09-17017 September 1999 Forwards Revised 180-day Response to NRC GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves. ML18152B6671999-09-17017 September 1999 Forwards Two NRC Forms 536,containing Info on Proposed Site Specific Operator Licensing Exam Schedules & Estimated Number of Applicants Planning to Take Exams And/Or Gfes,In Response to NRC Administrative Ltr 99-03 ML18152B3341999-09-16016 September 1999 Requests Relief from Specific Requirements of Subsection IWE of 1992 Edition with 1992 Addenda of ASME Section XI Re Containment Liner Examination Requirements,For North Anna & Surry Power Stations ML18152B4521999-09-14014 September 1999 Forwards Comments on Review of Preliminary Accident Sequence Precursor Analysis of Operational Event That Occurred at Plant,On 980508,as Reported in LER 98-009 ML18152B4501999-09-0808 September 1999 Submits in Triplicate,Application for Renewal of License for Rd Scherer,Iaw 10CFR55.57.Requests That Certification of Medical Exam by Facility Licensee,Nrc Form 396,be Withheld from Public Disclosure,Per 10CFR2.790(a)(6) ML18152B4471999-09-0101 September 1999 Requests That NRC Remove Listed Labels from Distribution ML18152B3711999-08-27027 August 1999 Forwards LER 99-005-00,per Plant TS Table 3.7.6.Rept Has Been Reviewed by Station Nuclear Safety & Operating Committee.Commitment Made by Util,Listed ML18152B3851999-08-23023 August 1999 Forwards Revised TS Basis Pages for TS 3.1.B,deleting Reactor Vessel Toughness Data Duplicated in Ufsar.Ref to Applicable UFSAR Section Included in TS Basis ML18152B3651999-08-20020 August 1999 Requests Removal of License Condition from Sh Wightman Operator License SOP-21538.Updated NRC Form 396 Is Encl.Form NRC 396 Withheld,Per 10CFR2.790(a)(6) ML18152B3681999-08-20020 August 1999 Submits 30-day Rept Re Two Instances in Which Conditions of Approval in Coc Were Not Observed in Making Shipment.Two Type B Shipments Using Model CNS 8-120B Package Were Made After Expiration of QA Program Approval Between 990531-0628 ML18152B3661999-08-20020 August 1999 Provides Medical Status Rept for E Washington,As Required by License Conditions.Summary of E Washington Current Physical Exam & Pertinent Lab Data Attached.Encl Withheld,Per 10CFR2.790(a)(6) ML18152B3801999-08-18018 August 1999 Forwards Technical Rept NE-1206,Rev 0, Surry Unit 2,Cycle 16 Startup Physics Tests Rept, Summarizing Results of Physics Testing Program Performed After Initial Criticality on 990525 ML18152B3781999-08-13013 August 1999 Forwards ISI Summary Rept for Surry Power Station,Unit 2 for 1999 Refueling Outage.Rept Provides Summary of Examination Performed During Outage for Third ISI Interval.No New Commitments Were Made ML18152B3751999-08-13013 August 1999 Forwards LER 99-004-00,IAW 10CFR50.73.Commitment Made by Util,Listed ML18152B4081999-08-0606 August 1999 Forwards Response to NRC 990520 & 0525 RAIs Re North Anna & Surry Responses to GL 95-07, Pressure Locking & Thermal Binding of SR Power-Operated Gate Valves. ML20210Q7661999-08-0606 August 1999 Requests Exception to 10CFR50.4 Requirement to Provide Total of Twelve Paper Copies When Submitting Surry & North Anna UFSAR Updates.Seek Approval to Submit Only Signed Original & One CD-ROM Version,Per Conversation with J Skoczlas ML18152B4091999-08-0505 August 1999 Forwards Vepc semi-annual fitness-for-duty Program Performance Data Rept for 990101-990630,IAW 10CFR26.71(d) ML18152B4001999-07-29029 July 1999 Requests Relief from Certain Impractical Requirements of ASME Section XI Code Associated with Partial Exams Conducted During 1998 Surry Unit 1 Refueling Outage.Relief Request SR-020 Encl ML18152B3981999-07-28028 July 1999 Forwards 60-day Response to GL 99-02, Lab Testing of Nuclear-Grade Activated Charcoal. Commitments Contained in Ltr,Listed ML18152B3971999-07-26026 July 1999 Provides Estimates of Licensing Actions Expected to Be Submitted in Fys 2000 & 2001,in Response to NRC AL 99-02 ML18151A6281999-07-23023 July 1999 Forwards Revised Epips,Including Rev 19 to EPIP-4.02,rev 14 to EPIP-4.16,rev 8 to EPIP-4.21 & Rev 7 to EPIP-4.30.EP & EPIPs Continue to Meet Stds of 10CFR50.47(b) ML18152B3991999-07-23023 July 1999 Requests That License for Jz Laplante Be Canceled as License Is No Longer Required ML18152B3961999-07-23023 July 1999 Forwards Preliminary,Uncertified License Application & Medical Certification for License to Operate Surry Power Station Units 1 & 2 for Ds Cobb.Encl Withheld,Per 10CFR2.790 (a)(6) ML18152B3931999-07-16016 July 1999 Forwards Updated NRC Form 396 & Ltr,Which Documents Medical Status of Mb Gross,License SOP-20476-2.Encl Withheld from Public Disclosure,Per 10CFR2.790(a)(6) ML18152B4401999-07-0101 July 1999 Informs NRC That on 990511,Dominion Resources,Inc,Executed Amended & Restated Agreement & Plan of Merger with Consolidated Natural Gas Co ML18152B4371999-06-24024 June 1999 Forwards Response to NRC Request for Clarification of Relief Requests Submitted on 990212,requesting Relief from Performing Hydrostatic Testing for Certain Small Diameter Class 1,RCS Pressure Boundary Connections ML18152B4361999-06-22022 June 1999 Forwards Response to RAI Re Surry & North Anna Power Stations,Units 1 & 2,per GL 96-06 ML18152B4301999-06-0303 June 1999 Informs of Util Intention to Revise Schedule for Submittal of License Renewal Applications for Surry & North Anna Power Stations to March 2002 ML18152B4261999-05-28028 May 1999 Provides Formal Notification of Effect of Recent Organizational Restructuring on OLs of North Anna & Surry Power Stations,Per NRC 990513 Telcon Request ML18152B4221999-05-27027 May 1999 Forwards Info Concerning Changes to ECCS Evaluation Models & Application in Existing Licensing Analyses for Surry & North Anna Power Stations,Units 1 & 2 ML18152B4231999-05-26026 May 1999 Informs That Vepc Will Revise 180 Day Response to NRC GL 96-05,within 120 Days of Date of Ltr to Incorporate Commitment to Participate in Joint Owners Group Program as Applicable ML18152B4211999-05-25025 May 1999 Forwards Rev 1 to Relief Request P-11 to Clarify Original Intent of Request by Specifically Requesting Relief from Requirements of Section 6.1 of OM-6 ML18152B4171999-05-17017 May 1999 Provides Notification of Number of Steam Generator Tubes That Were Plugged During Spring 1999 Refueling Outage Planned ISI ML18152A3701999-05-13013 May 1999 Submits Proposal to Use Provisions of ASME Section XI Code Case N-597 for Analytical Evaluation of Class 1,2 & 3 Carbon & Low Alloy Steel Piping Components Subjected to Wall Thinning as Result of Flow Accelerated or Other Corrosion ML18152B4121999-05-0303 May 1999 Forwards Application for Renewal of License for SV Ross. Encl Withheld Per 10CFR2.790(a)(6) ML18152B4101999-04-29029 April 1999 Forwards Scope & Objectives for 990803 Surry Power Station Emergency Exercise.Without Encls ML18152B6561999-04-23023 April 1999 Forwards Annual Radioactive Effluent Release Rept for Surry Power Station,Jan-Dec 1998, Which Includes Summary of Quantities of Radioactive Liquid & Gaseous Effluents & Solid Waste Released During CY98 ML18152B6491999-04-13013 April 1999 Forwards MOR for Mar 1999 for Surry Power Station,Units 1 & 2.MOR for Feb 1999 Incorrectly Stated Gross Electrical Energy Generated (Mwh) for Unit 2.Rept Should Have Stated Monthly Figure as 568965.0 ML18151A5851999-03-31031 March 1999 Forwards Rept on Status of Decommissioning Funding for Each of Four Nuclear Power Reactors,Per 10CFR50.75(f)(1) ML18152A2801999-03-30030 March 1999 Forwards Summary of Structural Integrity Evaluation of Thermally Induced Over Pressurization of Containment Penetration Piping During DBA for SPS & Naps,Units 1 & 2,per GL 96-06.Draft Proposed UFSAR Revised Pages,Encl ML18153A2721999-03-29029 March 1999 Forwards LER 99-002-00 Per 10CFR50.73.Listed Commitments Contained in Ltr ML18153A3421999-03-26026 March 1999 Provides Updated Medical Status Rept for Wb Gross in Accordance with License SOP-20476-02,Docket 55-5228,as Amended by 980320 License Amend.Informs That Gross Exhibits No Performance Problems & Will Continue on Current Medicine ML20204H0331999-03-17017 March 1999 Forwards Rev 5 to PSP for Surry & North Anna Power Stations & Associated Isfsis.Description & Justification for Changes Included with Plan Rev.Rev 5 to PSP Withheld Per 10CFR73.21 ML18153A3411999-03-15015 March 1999 Forwards Signed Applications & Medical Certificates for Initial License at Surry Power Station Units 1 & 2 for Listed Individuals.Without Encls 1999-09-08
[Table view] Category:UTILITY TO NRC
MONTHYEARML18153C3661990-09-20020 September 1990 Forwards Topical Rept VEP-NE-3-A, Qualification of WRB-1 CHF Correlation in VEPCO Cobra Code. ML18153C3701990-09-18018 September 1990 Forwards Addl Info Re Facility Containment Isolation Valve Type C Test,Per 900914 10CFR50,App J Exemption Request ML18152A2341990-09-14014 September 1990 Requests Exemption from 10CFR50,App J Section III.D.3 Re Local Leak Rate Testing During Every Reactor Shutdown. Basis & Justification for Exemption Encl ML20059J3341990-09-13013 September 1990 Forwards Rev 15 to Nuclear Security Personnel Training & Qualification Plan.Rev Withheld ML18151A2901990-08-31031 August 1990 Forwards Rev 10 to Updated FSAR for Surry Power Station Units 1 & 2,representing Second Updated FSAR Submitted This Yr ML18153C3451990-08-29029 August 1990 Forwards Proprietary Semiannual Fitness for Duty Program Performance Data Rept for 900103-0630.Rept Includes Summaries of Mgt Sanctions Imposed,Actions Taken to Correct Program Weaknesses & Events Reported to Nrc.Encl Withheld ML18153C3381990-08-22022 August 1990 Responds to NRC 900723 Ltr Re Violations Noted in Insp Rept 50-280/90-21 & 50-281/90-21.Corrective actions:as-found-as- Left Conditions of Auxiliary Feedwater Evaluated & Found Operable ML18153C3391990-08-22022 August 1990 Requests Approval for Use of Plugs Fabricated of nickel- chromium-iron Uns N-06690 Matl (Alloy 690) to Plug Tubes in Steam Generators for Mechanical & Welded Applications ML18153C3161990-08-0101 August 1990 Provides Supplemental Response to NRC 900629 Ltr Re Electrical Crossties,Load Shedding on Nonblackout Unit & Emergency Diesel Generator Reliability.Emergency Diesel Generator Reliability Program in Place,Per Reg Guide 1.155 ML18153C3171990-08-0101 August 1990 Resubmits Synopsis of Changes to Updated Operational QA Program Topical Rept Vep 1-5A ML18153C3091990-07-30030 July 1990 Provides Outline of Plan to Meet 10CFR50 App G Requirements, for Low Upper Shelf Energy Matls,Per NRC 900521 Request ML18153C3061990-07-30030 July 1990 Forwards Revised Tech Spec Pages,Addressing Constitution of Quorum & Timeliness of Mgt Safety Review Committe Meeting Minutes,Per NRC Request ML18153C3051990-07-26026 July 1990 Advises That Util Submitted Decommissioning Funding Plan & Financial Assurance Info W/Isfsi License Application ML18153C3041990-07-26026 July 1990 Responds to NRC 900626 Ltr Re Violations Noted in Insp Rept 50-281/90-20.Corrective Actions:Leaking Drain Plug & Upper Drain Plug on Motor Replaced W/Oil Drain Assemblies Composed of Piping & Valves ML18153C3031990-07-26026 July 1990 Advises of Withdrawal of Request for NRC Review & Approval of Engineering Evaluation 8.Revised Evaluation Will Be Maintained Onsite for NRC Audit During Future Insps,Per Generic Ltr 86-10 ML18153C3101990-07-26026 July 1990 Forwards Decommissioning Financial Assurance Certification Rept..., Nuclear Decommissioning Trust Agreement & Nonqualified Nuclear Decommissioning Trust Amended & Restated Trust Agreement, Per 10CFR50.75 ML18153C2901990-07-18018 July 1990 Responds to NRC Bulletin 90-001, Loss of Fill-Oil in Transmitters Mfg by Rosemount. Listed Transmitters Compiled.Transmitters Found Installed within Reactor Protection or ESFAS Have Been Replaced ML18153C2861990-07-12012 July 1990 Requests Cancellation of Operator Licenses for Listed Individuals.Licenses No Longer Required ML18153C2871990-07-11011 July 1990 Responds to Violations Noted in Insp Repts 50-280/90-18 & 50-281/90-18.Corrective Actions:Permanent Drain Line Installed & Matrix Which Describes Proper Ventilation Alignment for Plant Conditions Provided for Personnel Use ML18153C2851990-07-0606 July 1990 Forwards Response to Generic Ltr 90-04 Re Status of Generic Safety Issues ML18153C2831990-07-0303 July 1990 Advises That MW Hotchkiss No Longer Needs Operator License SOP-20548-1.Cancellation of License Requested ML18153C2821990-07-0303 July 1990 Requests Exemption from 10CFR50,App J,Paragraph III.A.6(b), Which Requires That When Two Consecutive Periodic Type a Tests Fail to Meet Applicable Acceptance Criteria Type a Test Shall Be Performed at Plant ML18153C3591990-06-28028 June 1990 Responds to SALP Repts 50-280/90-16 & 50-281/90-16 for Period 890701-900331.Corrective Actions Focus on Issues of Maint Backlog,Maint planning,post-maint Testing,Staffing & Procurement ML18153C2611990-06-21021 June 1990 Responds to NRC 900522 Ltr Re Violations Noted in Insp Repts 50-280/90-07 & 50-281/90-07.Corrective actions:as-built Configurations of 120-volt Ac & Dc Vital & Semivital Panel Breaker Installations Verified to Be Acceptable ML18153C2581990-06-18018 June 1990 Forwards Reissued Semiannual Radioactive Effluent Release Rept,Jul-Dec 1989. Rept Contains Info Re SR-89,Sr-90 & Fe-55 Analytical Results for Liquid Composite Samples ML18153C2591990-06-18018 June 1990 Forwards Response to NRC 900524 Request for Addl Info Re NRC Bulletin 88-004, Potential Safety-Related Pump Loss. Engineering Will Initiate Study to Evaluate Enhancements of Cooldown & Possible Heatup Operation ML18153C2541990-06-15015 June 1990 Forwards Corrected Tech Specs Page 3.1-3,per Identification of Typo in 900522 Application for Amends to Licenses DPR-32 & DPR-37 ML18153C2521990-06-14014 June 1990 Responds to NRC 900515 Ltr Re Violations Noted in Insp Repts 50-280/90-09 & 50-281/90-09.Corrective Actions:Abnormal Procedures & Fire Contingency Action Procedures Being Upgraded Via Technical Procedure Upgrade Program ML18153C2501990-06-0808 June 1990 Confirms That Primary Policy Re Onsite Property Damage Insurance,Provided by Nuclear Mutual Limited ML18153C2361990-05-29029 May 1990 Responds to NRC 900427 Ltr Re Violations Noted in Insp Repts 50-280/90-14 & 50-281/90-14.Corrective Actions:Personnel Involved Counseled as to Importance of Properly Recording & Reporting Surveillance Data ML18153C1971990-04-24024 April 1990 Responds to Unresolved Items Noted in Insp Repts 50-338/89-12,50-339/89-12,50-280/88-19 & 50-281/88-19 Re Secondary Sys Containment Leakage & Concludes Leakage Need Not Be Quantified & Not Included in as-found Leakage ML18153C1961990-04-20020 April 1990 Forwards Facility Previous Tests & Projected Leakage Totals for Type C Testing for Valves & Penetrations,Per 900108 Ltr ML18153C1901990-04-18018 April 1990 Requests That Operator License OP-20447-1 for Ja Yourish Be Cancelled ML18153C1861990-04-0505 April 1990 Requests Exemption from 10CFR50,App J,Paragraph III.A.6(b). Util Implemented Corrective Action Program Which Meets Intent of Regulation in Establishing Containment Integrity ML18153C1671990-03-30030 March 1990 Submits Supplemental Response to 10CFR50.63, Loss of All AC Power. Understands That Load Mgt Schemes for Both Blackout & Nonblackout Units Allowed by Station Blackout Rule ML18151A2551990-03-30030 March 1990 Forwards Rev to, Corporate Emergency Response Plan & Rev to, Corporate Plan Implementing Procedures. ML18151A4941990-03-29029 March 1990 Forwards Listed Info Re Licensee Guarantees of Payment of Deferred Premiums,Per 10CFR140.21(e) ML18153C1631990-03-27027 March 1990 Responds to Violations Noted in Insp Repts 50-280/86-05 & 50-281/86-05.Corrective Action:Surveillance Tests Being Performed in Accordance W/Administrative Requirements of Station Procedures & Plans Implementing New Review Process ML18153C1551990-03-20020 March 1990 Clarifies 900108 Request for Exemption from 10CFR50,App J Re Type C Testing Requirements ML18153C1511990-03-19019 March 1990 Responds to Generic Ltr 89-19, Request for Action Re Resolution of USI A-47. Operability & Surveillance Requirements for Steam Generator Overfill Protection Sys Will Be Incorporated in Tech Spec Change ML18153C1661990-03-16016 March 1990 Discusses Waiver of Compliance Re Containment Vacuum Sys Operability,Per 900315 & 16 Telcons ML18153C1471990-03-14014 March 1990 Discusses Functional Test for High Setpoint for PORVs ML18153C1571990-03-12012 March 1990 Forwards List of Emergency Operating Procedures in Preparation for 900402-12 Insp.Vol I Is Emergency Operating Procedure Set & Consists of 47 Notebooks.Vol II Contains Fire Contingency Action (App R) Procedures ML18153C1371990-03-0808 March 1990 Forwards Suppl to 1986 Inservice Insp Summary Rept,Adding Two Missing NIS-2 Forms Containing Info Re Replacement of Bolting Matl on 1-RC-SV-1551C (Flange a) & 1-RC-HCV-1556A ML18153C1281990-03-0101 March 1990 Submits 1989 Annual Steam Generator Inservice Insp Rept Results.No Steam Generator Tubes Plugged in 1989 ML18153C1261990-03-0101 March 1990 Responds to Generic Ltr 90-01 Re NRC Regulatory Impact Survey.Survey Covers Type of Insp,Audit or Evaluation by NRC Resident,Nrc Regional Ofc,Nrc Teams & INPO ML18153C1231990-02-22022 February 1990 Responds to NRC 900123 Ltr Re Violations Noted in Insp Repts 50-280/89-32 & 50-281/89-32.Corrective Actions:Use of Lab Hood Attached to F-2 Fan Suction Prohibited & Contaminated & Radioactive Items Removed from Hood ML18152A4881990-02-0606 February 1990 Responds to NRC 891222 Ltr Re Violations Noted in Insp Repts 50-280/89-34 & 50-281/89-34 on 891029-1125.Corrective Actions:Steps in Operating Procedure 2-OP-1.3 Associated W/ Valve Test Being Evaluated for Inclusion in OP-7.1.1 ML18153C0991990-02-0101 February 1990 Withdraws 891018 Application for Amends to Licenses DPR-32 & DPR-37,increasing Pressurizer Safety Valve Setpoint Tolerance to +/- 3% of Nomical Lift Setpoint.Emergency Tech Spec Change Granted on 891116 Provided Modified Tolerances ML18153C0951990-01-29029 January 1990 Forwards Response to Generic Ltr 89-13, Svc Water Sys Problems Affecting Safety-Related Equipment. Belief in Appropriateness to Address Generic Ltr 89-13 Concerns within Context of Established Programmatic Improvements Noted 1990-09-20
[Table view] |
Text
s VIHOrNIA Er.ncraic ann PownN CObf PANY Ricnw oxo.Vs uorn:A 2020 October 4, 1979 Pir. Ja-es P. O'Reilly, Director Office of Inspection & Enforcement D Cy Serial No. 817 PSE&C/Ct1Rjr:mac: wang U.S. Nuclesr Regulatory Comission Wb Region II 101 Marietta Street, Suite ~i100 o }g gt Docket No. 50-280 bJ1J.!Saat Atisnta, Georgia 30303 -l.icense No. DPR-32
Dear Mr. O'Reilly:
RESPONSE TO I.E. LETTER DATED SEPTEM3ER 7, 1970 SURRY POWER STAT [dN UNIT 1
. ,e purpose of this letter is to provide responses-to your letter of September 7, 1079 where you identified four items that required resolution prior to Unit I being returned to service. Our responses are as follows:
- 1. . Seven piping runs were selected for analysis in response to item one II) of your letter of Septem5er 7, 1979. These piping runs meet the requirements of it s ene in the letter in that they are samples of safety-related oiping systems six inches in diameter or less. In addition, these particular piping runs have the following characteristics:
- a. These piping runs are located inside the Reactor Containment Structure and are deamed inaccessible during plant operation.
- h. This piping is required for Safe Shutdown as defined and identified by Vepco station personnel.
It is our understanding that the selection of these seven piping runs has been reviewed by NRC personnel during discussions with Vepco at the Surry Power Station on September 14,: 1979. These seven piping runs have been analyzed as five' sample problems: '
System ES No. Problem No.
~
Safety Injection ' '
f1105'A?'
3000 1105 A3 Safety In'jection 1106 Al'
~
. 3003 1106 A4 Resistance Temperature Detection 1103 AS 3001 Resistance Tempe*ature Detection 1103 A4
- 3002 Resistance Temperature Detection 1103 A5 3004 D
~
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, Mr.' James P. 0'Re111y, Director 2
~. It should be noted that an MKS is a present day isometric piping
. configuration drawing, whereas an MSX is the original piping configuration
~
drawing.
The con'pletion of the analysis .of these seven piping runs coupled with the analysis; performed pertaining to the March 13, 1979 Show Cause Order will 4
complete the reanalysis of the Safe Shutdown piping runs 2 1/2 inches diameter and_ greater inside the Reactor Containment. For the purpose =of ,
seismic review,.-Safe, Shutdown requirerpents are. considered to be all.the
, Engineered Safeguards' as defined in the FSAR plus Auxiliary Feedwater and Residual Heat Removal wit at lesst;one heat exchanger.
~ '
We are advised by the St6nd WWebster Enginee$1'ng Corp' oration that during the . design of >the Su.rry Power Station, the Boston Stone 8 Webster office >
prepared original MSK drawirigssto show support locationsf and functions for' 21/2 to 6 inch' pipe.. The MSK &awings were' based upon approved pipe drawings (FP drawings) with additional infohnition pertaining to support locations and loads. These MSK drawings were transmitted to the
- construction site for use by the hanger designers', and to the fabricator of shoo supplied hangers, if, applicable, to develop construction drawings. Only.a limited number of these construction drawings have taen recovered to date.
The pipe stress problems for the present sample program tre based on. site
. generated ismittries of the oiping . systems (MXS @awings) as they presently exist in the plant .(as opposed to original MSK drawings). These MKS drawings show field verified pipe dimensions, locations and function of the hangers, and include _small diameter (21" vhes and less) branch piping to either the first anchor or second resu t. In addition, field as-built data is developed for all supports on 2 W to 6 inch piping.
The field MKS's are compared to presently avai:cble design documents in order to identify differences. Where significant differences are found, 3
the configuration defined on the field MKS's are reanalyzed using simplified methods (and in some cases NUPIPE computer methods) using approved criteria.
The analysis that is presently underway on the seven safe shutdown piping i . runs in the-sample is in accordance with the criteria and methods that have been reviewed with NRR during~their evaluation of.the March 13, 1979
'Show Cause reanal'l sis. These methods are consistent with the Surry FSAR c
and the recent " Safety Evaluation by the Office .of Nuclear Reactor
-Regulation",7and'have been accepted by the NRC. ,
- Thefivesampleplroblemswerereviewed,as.outiined-inthefollowing.five
- steps: , ,
~1) :A comparison of.the'present day MKS's was made to.the'available;
' ~ design documents.(MSK's).. This comparison; indicated that the
- following~. supports are not installed in accordance with the' ,
> presently available design documents:
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'3, Problem ;
.Line Present Day.
Configuration N o .' -System . Designation .
.; . .
3000 SI 3"-51-147-1503 Original AH moved around ' '
elbow and changed to strut.
3003 . SI -3"-51-145-1503 1 Original RH1 moved around >
elbow and. changed'to , .
strut.- .
,:(1). VC/L'C missing -
3001 RTD- '2"-RC-146-1502 .
.2"-RC-149-1502" '(1) LC" missing /(1) VC .
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. _. . ; added .
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.(?)VC/LC(m'issing
-3"-RC-147-1502
-(1) VC/LCfadded 300? ,RTD 3"-RC-}16-150?1 -(1 , Anchor < missing
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VC/LC missing -
~
"2"-RC1117-1502~ ' ((11)'VC niissing
'1) VC/LC missing
~
3004 RTO ~'2"-RC-132-1502~ (2) LC missing 3"-RC-131-1502** (1) Anchor missing (1).VC/LC missing AC = ' Axial lCohstraint. .
VC='-Vertical'Constra' int
LC =' Lateral Constraint- "
'RH = Rod ilanger ,
s .
- 2) The, presently existing HKS's~for the five problems'were reviewed
- using simplified analysis methods'.- The results of this'. review '
showed'that additional supports were.. required las follows:- ,
Problem No. ,
' System' '
~ Required.- ,
3000 >
'Sk: 4 1 Anchoe, ~'lVC/Lb, ,
- 1. Guide-
' ?3003 SI 3 Guides;.;llLC, 1 Anchor: 1'
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- 23001, 3002,A- RT" , 1 LC, 1sVC, 1;VC/LC for ,
- 3004 ,
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'3)' !Since;the;simplifieit method is; overly conservative, these five ,
- problems were then modeled and.run 'on the NUPIPE4 ciomputer _ ^ '
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. program using presentlyiexisting physical data'and approved: ..
O ~ cc ' ~ criter,ia .as previously described;" -
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LThe three RTD, bfpass lines are the most . difficult small pipe n systems in.the plant to layout and support becauseJof the.
thermal expansion of the pipe involved and because of the thermal motions imposed at tefeinations (and joints) of the:
pipe. In addition to controls on arrangements to meet stress.
the. NSSevendor,. technical requirements control relative ~
- criterian'thekbypas'sillegt, length'of 1ccat.ionion the loop' piping, and 3 absolute lengths of some segments of' the bypass legs. .
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The seven MKS?sidetail;579 linedfeet ofipipe of-which 273 feet >
are 2 1/2 to 6 inches and the 7emaining'305 lineal feat.are' ,
branchipiping4, inches.and smaller. The analysis using NUPIPE indicated that'all 273flineal 'fest of: piping L1/2 to 6 inches I w'a's Veithin allowable str'ess limits.' 2, l 4j Less than 3 feet (or less than one percent) of the branch pipe 2 .
i- inches and smaller exceeded the allowable deadweight stress -
- limits when multiplied by a stress intensification factor. .> >
Multiplication of non-cyclic stressL(such as_ dead-weight Stress) by an; intensification factor is a conservatism not explicitly '
i required by the B31.1' code,1955. .Also, allowable stresses were computed using minimum material properties, whereas actualf ,
! properties shown by test results typically eiceed minimums such s
. that an allowable stress calculated from the real' material '
strength would be significantly higher. '
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None of the pipe > exceeded the' allowable thermal stress > and most ,
was determined to be less than half of allowable thermalt '-
! stress. The pipe was therefore well supported:to preclude ..
~'effect of therma,1 stress which is continuously;present in the !
" RTD t ~ pass loops._ - .
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s Stress from the postulated earthquake condition exceeded
~
allowable; in-only 3. feet of branch pipe 2 inches and sm' aller '
when calculated'with a low damping factor. These stres'ses were.
.such thatislightly higher damping values would re' ally act to? 7 '
reduce vibratory response. Using current ASME code criteria,the stresses would not exceed real~ ultimate strength,:so no' failures ','
. would occuriand public safety"would not be affected, although-
- safety margins are r, educed. -
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The computers pipe stress analysis lof thel preserdly existing . ,- i
- configurations.using;NUPIPE indicated that. fewer,added supports:
- i. / are required toimeet approved criteria than was indicated'by the t
- conservative simplified methods. The 'added supports required by mL
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"the NUPIPE! analysis,Jusing approved Show Cause criteria,-are;asi GJ q ,
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- Additional Supports ,.
. Problem No. Svsten ' Required v' 3000 ' SI- Hone 3003 SI 1 suoport:to be removed
-3001 RTD 3 supports .
3002 RTD 3 supports 3004 RTD 3 supports.
- 4) The supports to be added will be based upon final approved
, NUPIPE stress runs in accordance with approved criteria.
5 )' ' Existing supports'en 2 1/2 to 6 inch piping are being reviewed and will be rean'alyzed using loads from the NUPIPE analysis, or in some cases to generic loads using approved criteria. The analysis using NUPIPE computer methods and approved Show Cause i
. criteria indicates th.at all 2 1/2 to.6 inch piping in the five-sample problems meet allowable stress limits. The branch piping 2 inch and smaller; which was necessarily incluaed in the computer model for.the larger piping, also met allowable stress limits with the exception'of. about one percentief. the small .
.' ' piping. We do 'not believe that safety is impaired and believe that the piping is fully capable of performing its function in.
the event of a design earthquake. However, any modifications ~
identified in the five sample. problems as a result of the use.of '
Show Cause criteria will be completed prior to startup.
~
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- 2. Anchor Bolt Test Program ST-75 was conducted to' inspect sanples- of 'all size anchor bolts'(Phillips Red Head Self-Drilling' Anchors) on baseplates- '
of. pipe supports for safety-related piping. . The objective of the prograi was to inspect approximately 200 bolts by measuring the anchor size, depth
- ' of set plug, shell projection, and thread engagement, and by . torquing the- "
bolts;to.a.value that would demonstrate a' factor safety of two. The , ,
. initial attempt was to obtain the sample by testing .one bolt per -!
accessible support inside.the containment, excluding those supports
/
previously tested under the Anchor Bolt Sampling Program, ST-39, and ;
excluding;those~ supports which had anchor bolt modifications made as a. ,
result _of the NRC Order to Show;Cause. Due to the: limited size of the , '
> sample obtained,J the criteria were expanded .to include.on,el bolt per accessible. baseplate inside the containment, allowing more than one. ,
'haseplate per support to be i_nspected and allowing inspection of different' anchor. bolts on those: supports inspected undar ST-39.JA sufficient number "of anchor boltsloutside.the 7containment was inc16dedito bring the totai
' sample size.to 'approximately 200 ~ anchor bolts. Of th'e final sample size. .
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of 209 b'olts, 35 were outside the containment. This methodology for: '
- selecting bolts-was; discussed'with your inspector, Mr. R.(M. Compton, and it. was . agreed that the main objective.was not' toiselect bolts' previously'-
inspected nor any,new bolts.'iThe sampling program met this objective.
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6 1 3 a s The results;of'the Anchor Bolt Test Program show that 9S.2% {199 anchor
- holts).of the' sample . size of 209; anchor bolts were acceptable and have demonstrated a factor of safety ' of two? This sample size: represents 195 '
- 1baseplates on 164 supports. Ten anchor bolts were found not to be
- - Lacceptable' for the following reasons. Two anchor bolts did not'have sufficientithrea'd engagement, one anchor (bolt hadino thread engagement and one anchor bolt had only 1/8" of. thread engagement.' .Six anchor bolts had the bolt and'shell turning together. Five;of these were on one-b'aseplate r 4 end rotated 'when the back-off torque was ' applied in attempting. to remove .
the bolt and the: sixth anchor bolt rotated:during;appifcation of the proof
< load torque. Two ' anchor bolts were found on one. baseplate with the shells: .
rusted to the extent that a back-off torque off'250.ft.-lbs. was' recessary to remove the bolt and no measurements could be obtained for the shell or i depth of set plug. Results of this program will be maintained at Surry Power Station as Special Test-75 should you require any further l
~
4 4 inspection. The detailsief the results1have,been= r,eytewed< by your Messrs.
E. J. Vallish and R. H. Compton; ' All' deficiencies? identified in this program will be corrected prior to re,tur,ni,ng the Unit (to servic'. e s .
. .. ..w.,. .r.
- 3. We are incorporating flexibility considerations..into'the-analysis of the concrete' anchors forlthe thirteen pipe supports with baseplates-in the five sample piping runs.) Current; support, loads are preliminary.since pipe stress analysis,is sincompletc.;at this time.l 4eyerthelessewe-are p ,
.oroceedingiith, flexibility review per the Stone & Webster! simplified 50-STEP procedure or more sophisticated computer analysis, and will use '
the final oipe stress loading to confirm the preliminary baseplate results. Flexibility analysis'is complete on five of these supports and, results are# acceptable using . preliminary loads. All reanalysis -is being.
F performed using field verified piping and support information. If
' hardware modifications are required as analysis continues, appropriate action will be taken in accordance with the plant Technical Specifications.
4 An additional number of supports'are beingfevaluated fordflexibility.
considerations as a result-of a previous; bolt sampling program.
~
Approximately 94 supports are involved with this program with evaluations "
complete on 39.. A total of 35 supports have been found to1be acceptable _-
with flexibility criteria included ~as described above. .Any baseplates or, ,
anchor-bolts found to be overstressed dueitoLfinal (NUPIPE). loading ~n .
~
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- conditions with computer analy"s for the baseplates will .be subject to a appropriate. action in accordanes with the; plant Technical. Specifications. -
, 4. SpecialTest15: NhichcoveredtheAnchorBoltTest'Programincluded.a '
baseplate visual? inspection. The visual inspectio_n : involved a walkdown of 1,, all seismic Category;I piping systems inside the containment covered by:
[ -
. the NRC Order, to Show Cause and I.E. Bulletin 79.-14 to look for any gross-
~
N defects: such:asimissing or loose bolts and< nuts, and stewed bolts which, ,
would adversely; affect thesability of th'e sup' port to perform i_s.. intended ,
- ~ ;functioni A total- of.88 piping isometric 1 drawings 4MKS's) were walked' *
,dewn in the? inspection. Fifty-onel drawing's~'w ere for the;NRC Order to Show?
Cause. and 37: drawings were for the piping covered by:I.E. Bulletin 79-14. _
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. All noted discrepancies were tabulated and the line rewalked to verify.and; evaluate the discrepancies. These, items;which;were found to bc
deficiencies incisde five U-bolts ihichJwerelloose. three baseplates which were not flush with the concrete surface, four baseplates with oversize-
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bolt holes which required piste
' Linstalledat.'an'angleand_ requip [Eashers,one.springhangerwhichwas' ed. realignment;, one bolt missing in.a
- U-bolt, one baseplate which required regrouting, ch6 skewed b31t, one. "
' anchor bolt missing,1 two rod hangers not connectedsto:the piping, .one ,
i
'oneb'rokenanchoribolti,'a'ndSnewasherinstalled washer behind a which baseplate.' sas'Five need*14, lpipe suppbrtifiscrepancies were found on:the MK drawings prepared for.use in complying with I.E. Bulletin 70-14 These discrepancies include.one U-bolt and.one support which.were shown on the drawings but not installed, and two supports and one U-bolt which werej , , ,
- installed but not shown'on the drawings.; These discrepancies will be
' reflected in the finalized MKS drawi,ngs which will:be used for analysist under I.E. Bulletin 79-14. ,
i A repair procedure was wr.itten as part of ST-75 and included repairs for those deficiencias noted in the visual . inspection as well .as those I
deficiencies which resulted from or were noted as a result of the Anchor' Bolt Test Program.. All of'these deficiencies wil.1 be corrected prior to i- start-up.
! The deficiencies found were small in numSer. Therefore, we believe the failure rate of the visual inspection ,to be acceptable. The results of the: visual inspection program have been reviewed by your Mr. E. J. Vallish.
, 'Approximately I?.5.MKS's make up the I.E. Bulletin 79-14 scope cf work.
Seven of these have been analyzed as the sample problems. We believe the remaining 118 MKS's can be analyzed in about five months and we will make j every'e.ffort to expedite this schedule. If during this review any j modifications are identified, they will be reported-to the station and l evaluated under tne requirements of the plant Technical Specifications.: ,
j
< Very trulv yours, 4
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W. C. Spencer- .
. Vice President ~ Power Station
' Engineering and Construction Services ,
cc: Mr. Victor'Stell'o,. Director ~ .
y Office _oftInspection~and EnforcementL 1
' Mr. . Harold R. Dent' o n,t Director . .
Office of Nuclear" Reactor Regulation <
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