Forwards Addl Info Re NUREG-0737,Item II.D.1 on Performance Testing of Relief & Safety Valves,Per 850920 Commitment. Response to Request Numbers 2,4,5,6b & 7 Also Encl.Remaining Info Will Be Submitted by 860228ML18144A006 |
Person / Time |
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Site: |
Surry |
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Issue date: |
10/31/1985 |
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From: |
Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
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To: |
Harold Denton, Varga S Office of Nuclear Reactor Regulation |
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References |
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RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.1, TASK-TM 85-615A, TAC-44622, TAC-44623, TAC-44633, NUDOCS 8511050123 |
Download: ML18144A006 (32) |
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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
\,,.
1..
e VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIROINIA 23261 W. L. STEWART VICE PRESIDENT October 31, 1985 NUCLEAR OPERATIONS Mr. Harold R. Denton, Director Serial No. 85-615A Office of Nuclear Reactor Regulation E&C/TLG:acm Attn: Mr. Steven A. Varga, Chief Docket Nos. 50-280 Operating Reactors Branch No. 1 50-281 Division of Licensing License Nos. DPR-32 U. S. Nuclear Regulatory Commission DRP-37 Washington, D. C. 20555 Gentlemen:
VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNIT NOS. 1 AND 2 ADDITIONAL INFORMATION RELATED TO NUREG-0737, ITEM II.D.I PERFORMANCE TESTING OF RELIEF AND SAFETY VALVES In our letter dated September 20, 1985, Serial No.85-615, we stated we would forward to you the requested information related to the above subject by October 31, 1985. We noted that the information requested required input from our architect-engineer (Stone & Webster), our vendor (Westinghouse) and ourselves. We were to review and approve the information and forward it to you.
In reviewing the information supplied by the architect-engineer and the vendor, it has become apparent that additional information and review are necessary to supply a complete response to your request. A partial submittal (safety valve information only) is enclosed which includes the information for request numbers 2, 4, 5, 6b and 7. The remaining information will be submitted to you no later than February 28, 1986. The need for additional time was discussed with the NRC Surry Project Manager on October 10, 1985.
Very truly yours, K2) t\ a,J..;J ~
~ W. L. Stewart Enclosure
'/----~-
~g1 !050t23es10~-1-----,,
p R ADOCK 05000280 1
PDR
...* .. e e Vrnou;,IA ELECTRIC AND PoWER COMPANY TO Mr. Harold R. Denton cc: Dr. J. Nelson Grace Regional Administrator NRC Region II Mr. D. J. Burke NRC Resident Inspector Surry Power Station
e ENCLOSURE VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNIT NOS. 1 AND 2 ADDITIONAL INFORMATION RELATED TO NUREG-0737, ITEM II.D.I PERFORMANCE TESTING OF RELIEF AND SAFETY VALVES e
Question
- 2. On page 4 of Ref~rence 3, it was stated that safety valve blowdown in excess of 5% had been analyzed in conjunction with the Westinghouse Owners Group program on safety valves and ~he results from these analyses showed no adverse effect on plant safety. Reference 3 also indicated that a discussion of the analyses was presented in the Westinghouse Report WCAP-10105 (Reference 4). However, we were unable to find the aforesaid information in the Westinghouse report. Please provide documentati.on which demonstrates that the increased blowdown will not affect the safe operation of the reactor cooling system.
Response
The impact on plant safety of pressurizer safety valve blowdown in excess of 5% was evaluated for Surry Units 1 and 2. The re$ults of this evaluation showed no adverse effects on plant safety.
Safety valve blowdowns in excess of that assumed in the Surry FSAR will have the following effects on the events in which safety valve actuation occurs:
- 1. Increased pressurizer water level during and following the valve blowdown,
- 2. Lower pressurizer pressure during and followi~g valve blowdown,
- 3. Increased inventory loss through the valve.
e e The impact of the increased relief valve blowdowns with respect to the above effects was evaluated for the Surry FSAR events in which safety valve actuation occurs, (i.e., Loss of External Electrical Load and Locked Rotor).
For the Loss of External Electrical Load event, results from sensitivity analyses performed for a 4-loop plant were used for the evaluation. These analyses investigated the effects of different blowdown rates on the event.
Similar results are expected for a 3-loop *plant. The results of these analyses showed only marginal increases in pressurizer water volume and the maximum pressurizer water levels were well below the level at which liquid relief would occur. The Surry FSAR analysis results show th/it a small -
increase in pressurizer water volume, due to increased safety valve blowdown, will not result in liquid relief. The sensitivity analyses also showed that peak RCS pressures were unaffected by the increased blowdowns.
The increased blowdowns did result in lower pressurizer pressure and increased RCS inventory loss, however, these had no adverse impact on the event and adequate decay heat removal was maintained. These conclusions can also be applied to the Loss of Normal Feedwater/Station Blackout results.
For the Locked ~-otor event, increased safety valve blowdowns have little impact on the event. As analyzed and presented in the Surry FSAR, the opening and closing of the safety valve occurs over a short time period (3 seconds). As a resu1t, there is little change in either pressurizer level or RCS inventory. Increased safety valve blowdowns would have no impact on peak pressure, peak clad temperature, or DNBR as these occur prior to the closing of the safety valve.
. Question:
e
- 4. The magnitude of the maximum bending moment induced at the inlet flanges of the safety valves and PORVs during valve discharge were not provided in the* Surry transmittal. Give the predicted values of the maximum moments for these valves so that the valve operability can be evaluated.
Response
Tables IA, lB, 2A, and 2B provide the calculated moments at the safety relief valve and power operated relief valve inlets due to the safety valve
- discharge loading condition.
e e SURRY UNIT 1 TABLE lA Safety Valve Inlet Loads Due to SRV Discharge Type of_ Valve Number Loading Condition SV/1551C sv /255 lB SV /1551A Loads (Ft-Lbs) Loads (Ft-Lbs) Loads (Ft-Lbs)
Mt Mb1 Mb2 Mt Mb1 Mb2 Mt Mb1 Mb2 Thermal*.': 1371 1043 -3951 780 804 -2892 -2245 -6360 -1293 Dead Load -93 186 -530 -161 295 -907 44 -230 -147 OBE (I) 69 131 15 87 139 54 155 41 36 DBE (I) 120 236 25 142 296 88 246 65 67 OBE (A) 150 14 35 103 13 35 78 45 30 DBE (A) 341 38 95 246 35 93 197 126 78 SRV Discharge 7094 1080 3251 9177 3252 2428 58 5166 2305
(
Reference:
7)
Where OBE (I) = Operational Basis Earthquake Inertia DBE (I) = Design Basis Earthquake Inertia OBE (A) = Operational Basis Earthquake Anchor Movement DBE (A) = Design Basis Earthquake Anchor ~ovement Mt = Torsional Moment Mb = Bending !foment
- = Maximum Thermal Load Component with Sign from all Thermal Conditions Analyzed Note: Dynamic loads such as OBE(I), DBE(!), OBE(A), DBE(A) and SRV Dischaige are considered as+/-.
- .. .. e SURRY UNIT 1 TABLE lB PORV Inlet Loads Due to SRV Discharge
_Type of Valve Number Loading Condition PCV/1455 PCV/1456 Loads (Ft-Lbs) Loads (Ft-Lbs)
Mt Mb1 Mb2 Mt Mb1 Mb2 Thermal~~ -720 2022 90 -1318 1312 -679 Dead Load 31 -10 -211 42 -7 -224 OBE (I) 85 112 114 103 96 99 DBE (I) 137 224 221 194 184 196 OBE (A) 1 25 2 1 18 1 DBE (A) 2 60 5 3 43 3 SRV Discharge 6 28 0 6 47 4
(
Reference:
7)
Where OBE (I) = Operational Basis Earthquake Inertia DBE (I) = Design Basis Earthquake Inertia OBE (A) = Opera_tional Basis Earthquake Anchor Movement DBE (A) = Design Basis Earthquake Anchor Movement Mt = Torsional Moment Mb -* Bending Moment
- = Maximum Thermal Load Component with Sign from all Thermal Conditions Analyzed Note: Dynamic loads such as OBE(I), DBE(I), OBE(A),_DBE(A) and SRV Discha~are considered as+/-.
e SURRY UNIT 2 TABLE 2A Safety Valve Inlet Loads Due to SRV Discharge Type of- Valve Number
- i,oading Condition SV/2551A SV/2551B SV/2551C Loads (Ft-Lbs) Loads (Ft-Lbs) Loads (Ft-Lbs)
Mt Mb1 Mb2 Mt Mb1 Mb2 Mt Mb1 ~2 Thermalir -1904 3902 -3114 -2002 2322 4177 2799 -6849 245i Dead Load 9 -106 48 1 -343 -724 -53 -292 -108 OBE (I) 102 80 57 114 150 88 78 82 101 DBE (I). 180 155 119 230 312 183 154 165 201 OBE (A) 88 48 41 209 80 87 68 44 8 DBE (A) 186 129 103 469 221 218 84 59 10 SRV Discharge 4268 4415 6451 9195 7969 8657 2189 9229 5889
(
Reference:
8)
Where OBE (I) = Operational Basis Earthquake Inertia DBE (I) = Design Basis Earthquake Inertia OBE (A) = Operational Basis Earthquake Anchor Movement DBE (A) = Design Basis Earthquake Anchor Movement Mt = Torsional Moment Mb
-:r
= Bending Moment
= Maximum Thermal Load Component with Sign from all Thermal Conditions Analyzed Note: Dynamic loads such as OBE(I), DBE(I), OBE(A), DBE(A) and SRV Discharge are considered as+/-.
e SURRY UNIT 2 TABLE 2B PORV Inlet Loads Due to SRV Discharge Type of Valve Number
- Loading Condition PCV/2455 PCV/2456 Loads (Ft-Lbs) Loads (Ft-Lbs)
Mt Mb1 Mbz Ht Mb1 Mb2 Thermal* -711 1120 407 -972 -264 -1096 Dead Load -77 -1 227 -61 .o 104 OBE (I) 70 46 279 54 38 206 DBE (I)* 143 93 589 111 76 436 OBE (A) 0 1 0 0. 0 0 DBE (A) 0 1 0 1 2 0 SRV Discharge 8 17 5 2 15 7
(
Reference:
8)
Where*
OBE (I) = Operational Basis Earthquake Inertia DBE (I) = Design Basis Earthquake Inertia OBE (A) = Operational Basis Earthquake Anchor Movement DBE (A) = Design Basis Earthquake Anchor Movement Ht = Torsional Moment Mb = Bending Moment
-!: = Maximum Thermal Load Component with Sign from all Thermal Conditions Analyzed Note: Dynamic load_s such as OBE(I), DBE(!), OBE(A), DBE(A) and SRV Dischar_ge are considered as+/-.
e e Question
- 5. NUREG-0737, Item II.D.1 requires the qualification of the control circuits of the plant specific PORVs for design-basis transients and accidents.
However, such qualification does not have to be submitted for this review, if it has already been included in the submittal to fulfill the requirements of 10CFRS0.49. Verify whether the in-plant PORV control circuits have been included in the 10CFRS0.49 review or provide the necessary evaluation to demonstrate that the requirement of NUREG-0737, Item II.D.l concerning control circuitry has been met, if the PORV circuitry was not reviewed under 10CFRS0.49.
Response
. The in-plant PORV control circuits have been reviewed and included in the 10CFRS0.49 and the Regulatory Guide 1.97 reviews. Included in these.
reviews and qualified are the MOVs, SOVs, limit switches, associated cables and electrical interfaces. A Regulatory Guide 1.97 implementation schedule has been submitted which includes work which will provide redundant limit switches and safety valve position indication consistant with the Regulatory Guide.
e e Question:
- 6. The thermal-hydraulic analysis of the safety/PORV Inlet and discharge
_piping was discussed on pages 15 and 16 of Reference 3. In the discussions only information concerning safety valve discharge conditions was provided. The PORV discharge conditions were not addressed.
- a. Provide a discussion of the thermal-hydraulic analysis of the PORV discharge conditions and give the input parameters used in the analysis.
- b. Since the ASME Code requires the derating of the safety valves to 90 percent of actual flow capacity, the safety .valve analysis should be based on flow equal to or higher than 111 percent of the valve flow rating unless a different flow rate can be justified.
Discuss the pertinent input data used in the therma_l-hydraulic
! . analysis to demonstrate that the above derating requirement was I'
- ~ met.
I
Response
6b. The safety valve input data used in the thermal-hydraulic analysis did not explicitly account for the ASME code derating factor, however, the slug flow analysis performed by using WATSLUG code is considered to be conservative by more than the derating factor. The slug after passing through the valve had a density set equal to 6 lb/ft 3 due to flashing based upon the energy in the 400°F slug upstream of the valve. The downstream slugs are treated as square-edged slugs which is conservative, and the slugs from each safety valve are combined into a larger square-.edged slug which is also conservative. The overall conservatism of our analysis (fluid transient and stress. analyses) is demonstrated by Figure 3A. 3A. 6 of Attachment A to be significantly larger than 11 percent. Attachment A contains a general description and verification of the WATSLUG program against RELAP 5/MOD 1 and EPRI test results. The allowable stress used is 1. 8 sh. This is conservative for SRV
- discharge loads case which includes DBE load-ing.
Question:
- 7. The discussion of the piping stress analysis on pages 17 t.hrough 20 of Reference 3 did not include the stress evaluations for the safety/PORV piping and supports and other pertinent data. The following informa-tion is required in order to complete the evaluation of the piping stress analysis.
- a. Give numerical values of the cutoff frequency and total analysis time for safety valve and PORV discharge used in the NUPIPE computer analyses.
- b. Were the same load combination equations and stress. limits given on page 19 of Reference 3 used to evaluate both the piping upstream and downstream of the safety valves and PORVs, and explain why the inlet piping (Class 1) and the discharge piping stresses (Class 2/3) are evaluated by the same standard.
- c. Provide a comparison between the worst stresses and allowable stresses for the inlet and discharge piping to show that the piping system is adequate for all.loading combinations.
- d. Provide the same comparison for the piping support stresses as in item c above.
- e. Provide a copy of the thermal-hydraulic and stress analysis report for both the inlet and discharge piping.
Response
7a. The cut-off frequency and mode, and the total integration time for the NUPIPE *computer analysis for the safety valve discharge loading condition for Surry Power Station Uni ts 1 and 2 are as follows:
Unit 1 Unit 2 Cut-off frequency (HZ) 502 543 Cut..:off mode 196 181 Total integration time 0.9 0.9 (seconds)
Reference 1 2 The analysis for the safety valve discharge loading included both t.:.e inlet and discharge piping of the safety as well as relief valve piping from the pressurizer nozzle to the pressurizer relief tank nozzle.
7b. The applicable code per Updated Final Safety Analysis Report (LTSAR)
Table 4.1-9 for the piping analysis, USAS B31.l.l 1955, "Code for Pressure Piping," does not contain any piping classification such as Class 1 and Class 2/3 piping. Accordingly, the piping, both upstre3m and doi.rnstre_am of the safety valves and POR\'s, is evaluated to a sing:e set o_f load combination equations and allowable stress limits. The
. loading combinations and allowable stress limits utilized in the analysis are con.sistent with UFSAR and Westinghouse reco1111Dendations.
(References 3 and 4) e e 7c. Tables 3 and 4 provide a comparison of maximum stress level for Surry Units 1 and 2 at any location in the piping system (between the pressurizer nozzle and the pressurizer relief tank nozzle), to the code allowable limits. The comparison indicates that the piping system meets the allowable stress limits for all loading combinations.
7d. Tables 5 and 6 provide a list of attributes for two representative restraints associated with the PSRV system. The calculated stresses/loads for the various restraint components of the system are compared to their respective allowables. These allowables used are those specified in the AISC Manual for Steel Construction (7th Ed, 1970) and manufacturer allowable limits.
The comparison ind1cates that the stresses/loads for these two restraints, which are typical of others within the PSRV system, are within allowable.
stresses/loads permitted.
7e. The thermal-hydraulic and stress analysis for the pressurizer safety and relief valve piping system has been performed in accordance with the ASA B31.1 Code for Pressure Piping 1955 Edition, as noted on page 19 of our October 31, 1984 submittal. Surry Power Station is an ASA B31.1 plant and unlike an* ANSI l?Jl.7 plant, Code for Nuclear Power Piping,does not require formal stress reports.* As noted, the analysis has been performed, however we do not have the reports as have been requested.
e Table 3 HAXIffiJM STRESS LEVEL
SUMMARY
SURRY UNIT 1 Maximum Point Calculated Allowable Equation Criteria No. Stress (psi) Stress (psi) 8 s Sh 465 8053 15990 p + SDl ~ 1. 0 9N sp + SDL + SOBET ~ 1.25 sh 465 10308 19188 9F s p + SDL + (SfiBET + s6cc); ~ 1
- 8 sh 200 28604 28782 10 STH <SA= ( 1. 25 SC + 0 . 25 Sh) 300 36769 27435 11 s p + SDL + STH ~ Sh + SA 3.00 39595 43425 Note: It is acceptable to exceed EQ(lO) for thermal expansion stress as long as the requirements of EQ (11) a re met.
e Table 4 MAXIMUM STRESS LEVEL
SUMMARY
SURRY UNIT 2 Maximum Point Calculated Allowable Equation Criteria No. Stress (psi) Stress (psi) 8 sp + 5DL ~ 5H 248 8187 15990 9N s 5 2 248 12988 19188 p + DL + SOBET ~ 1. SH 9F Sp+ SDL + (S2DBET +szocc )\ < 1.8 SH 38 22797 28782 10 STH ~ 1. 25 Sc+ 0.25 SH= SA 38 37666 27435 11 sp + 5DL + STH ~SA+ SH 38 41617 43425 Note: It is acceptable to exceed EQ(lO) for thermal expansion stress as long as the requirements of EQ(ll) are met.
~ Notes for Tables 3 and 1
111
~. S p
= Longitudinal Pressure Stress (psi)
SDL = Deadload Stress (psi)
Operational Basis Earthquake Stress (includes anchor mov~ents)
SDBET = Design Basis Earthquake Stress (includes anchor movements)
SOCC = Safety Valve Discharge Time History Loads SC= Allowable Stress at Ambient Temperature= 18750 psi SH= Allowable Stress at Maximum Operating Temperature= 15990 psi at 470°F (See Note 3 below)
STii = Thermal Stress (psi)
- 2. The pipe stress reanalysis and pipe support modifications associ-ated with this design change are based on allowable stresses as provided in ANSI B31. l Code-1973 Edition. This later Code used a different method liQ__ establish allowable stresses and incorporated test data performed subsequent to the ASA B31. l-1955 (including Code Case N7)
Code.
- 3. The allowable stress used is lower of SA376 TP316 material (piping upstream of safety and relief valves) at 650°F and SA312 TP304 (piping downstream of safety and relief valves) at 470°F.
e TABLE 5 SURRY UNIT I RESTRAINT H-906 Calculate Allowable Item Stress/Load Stress/Load Reference Snubber 4186 lbs 10350 lbs Corner & Ladda Catalog Tube Steel Axial Stress 0 psi 21600 psi AISC Manual of Steel Bending Stress 17798 psi 21600 psi Construction (7th ED)
Shear Stress 1207 psi 14400 psi ~ ~ 0.6 Fy ls - 0.4 FY Integral Welded Attachment (Lug)
Bending Stress Negligible 11874 psi AISC Manual of Steel Shear Stress 2093 psi 7760 psi Constr,uction (7th ED)
~ = 0.6 Fy S = 0.4 FY Critical Weld 0:27 in. Fillet 0.31 in. Fillet AISC Manual of Steel (Required) (Provided) Construction (7th ED)
Drill co-Maxi-Bolt Anchor Bolts Tension 5983 lbs* 10142 lbs S'WEC:-STS-ACll-1 Shear 1057 lbs* 8000 lbs Baseplate Stres~ 24949 psi 27000 psi AISC Manual of Steel Construction (7th ED)
FB = . 75 FY Reference 5
- Note: Bolt interaction= 0.72 < 1.0 e
-TABLE 6 SURRY UNIT II RESTRAINT H-900 Actual Allowable
- It~ Stress/Load Stress/Load Reference Snubber 18000 lbs 21000 lbs Corner & Ladder Catalog Tube Steel Axial/Bending AISC Manual of Steel Interaction 0.234 l Construction (7th ED)
Shear Stress 427 psi ,14400 psi ~ : 0.6 FY
~ - 0.4 FY Integral Welded Attachment (Trunnion)
Bending Stress 5294 psi 11874 psi AISC Manual of Steel 3226 psi 7760 psi* Construction (7th-Ed)
F = 0. 6 FY
~ = 0.4 FY Critical \oi'eld
- 0. 16 in. Fillet 0.28 in. Fillet AISC Manual of Steel (Required) (Provided) Construction (7th ED)
Drillco-Maxi-Bolt Anchor Bolts Tension 8029 lbs: 16593 lbs S\oi'EC-STS-ACll-1
- Shear 4401 lbsr 11259 Baseplate Stress 11889 psi 27000 psi AISC Manual of Steel Construction (7th ED)
FB = 0.6 FY Reference 6
- Note: Bolt interaction= 0.87 < 1.0
~eferences:
- 1. SWEC EMD calculation No. 14937.03-NP(B)-003-XF, Rev. O, dated Novl!mber 23, 1984, "Time-history analysis for the pressuri*zer safety and relief valve piping."
- 2. SWEC EMD calculation No. 14937.03-NP(B)-001-XF, Rev. O, dated November 27, 1984, "Time-history analysis for the pressurizer safety and relief valve piping."
- 3. UFSAR for Surry Power Station - Units 1 and 2, Section lSA.
- 4. Review of pressurizer safety valve performance as observed in the EPRI Safety and Relief *valve Test Program, dated June 1982,
Westinghouse Corporation.
- 5. SWEC EMD Calculation - 12846.22-NP(B)-Z-630-036 Virginia Power-Surry Unit No. 1, "Pipe Support Calculations for: Pressurizer Safety and Relief Tank Piping MKS-124Al and A2 Problem 630," dated February 28, 1985.
- 6. SWEC _EMD -calculation - 14937.03-NP(B)-001-ZB-Oll, Surry Unit No. 2, "Pipe support calculation for: Pressurizer Safety and Relief Tank Piping, DCP-84-72," dated March 2, 1985.
e e
- 7. S\¥EC EMD Calculation: 12846.22.NP(B)-030-X12 Rev. 2, dated February 26, 1985, "Pipe stress analysis of the pressurizer safety and relief system with time history - pipe stress Problem 630," computer run number R2586008, Job No.: 3132, dated December 6, 1984.
- 8. SWEC EMD calculation: 14937.03-NP(B)-002-XE, Rev. O, dated December 6, 1984, "Stress Analysis for Pressurizer Safety and ,Relief System Piping ~ Problem 2000" computer run number R2586012, Job No.:
2842 dated November 13, 1984.
~AFETY EVALUATION QUESTIONS t,1:REG Oi 3 7
~
SURRY UNITS 1 & 2
- l. G4neral Description The-pu'rl)oa~ of W,1.TSLOG (Ref. l) is to detem~e forcing functior.s on piping 1ystems during vater slug discharge events for subsequent i~put to piping dyu&m.ic &A&l.ysis.
- The analysi* ~ baaed upon rigid' body motion of the generally subcooled water slug and id~ gas representations of the steam or air using rigid column theory to facilitate tracking the several water-ste.am or vater-air interfaces. The driving force i.s the ste.am pressure between the valve and the slug, less friction and other losses, and back pressure. Density changes due to-poasible local f~shing of the-water slug are c~nsidered.
Raving recourse co the control volume theory, the subsequa~t 1egment force calculation i1 carried cue. - *
'!'he* input consists of complet~ piping system geomet-:-y, pipe dimensions, valve flow characteristics, valve opening time, detail upstre.am steam conditions, and initial downstream steam o~ air conditions, while the output contains forcing functions for each piping segment baaed upon flow velocities, pressures, and de:;siciea during the water slug diach.arge event. Forces ara written on cape for direct input to NUPil't-S'J (MI-llO).
(Ref. 2).
- 2. Program Verification The WATSLUG model of the test problem is di.&grammed in Figure 3A.3.A-l .nd the mr?I?E-SW model is diagrm:mied in Figure 3A.3 .A-2. WATS!.OG is veri!ied by com;,ar:!.ng tbe solution of this test problem co the results for the same problem obtained by an inde?endenc analytical approach (RE!.Al'5/~0D l, ~ef. 3) as shown in Figures 3A.3.A-3 and 3A.3.A-4 and by c01:parison of ?redicteci ve:-sus measured support reactions. NUPn'E-SW (ME-llO) ge~eraced support reactions due :o the ~AIS1.0G forcing functions vere compared'with ~e:-:::e~:a:
me.asurements from & test run cf this problem (:::PRI Test 908, Ref. 3) as shown in Figures 3A.3.A-5 and 3A.3.A-6.
The WA~SlUG generated forcing functions and tbe resultant NUl'!PE-~J support re.actions co!Zll)are favorably vith the RD.A.PS/MOD l pt'edicted forcing tunctio~s and the !PRI me&sured suppot't reactions, respectively.
- 3. References----
- l. \JA.!SLOG" CME-212) computer code by J. S. Hsieh and 0. A. Van ~yne, Ver. 0, Rev. 3, December 1982 &nd the related documentation calculation 576.4iO.l-NP(B)-038-FD, Rev. 2, '"water Slug Disc~rge i~
P1pi~g System (~AISLOG) - PTeproduction Version 3", dated ~.arch J, l9S:.
- 2. NU?IP!-SW, ME-llO, V03,I.l4 (created S2.095), "Computer code :or Scre~s Analysis of Nuclear Piping",
- 3. "A;,plication of RI:..APS/~OD 1 for calculation of Safety and Re lie: Va.:se
- >ischarge Piping Hydrodynamic Loads", **nce:-im Report, ~rch 1962, ':,~*
!~ce:-mou~ta1n !echnologies, lnc., Idaho Fal:s; :daho, Prcjec: !".a.~ager R. K. Bouse.
- I r.nnJT DATA :OR ~ATSt:G
- TOT.AI. n?SIDE :!UC'!'ION
?!P! ~o. u::NGTH (!t) D L ~ (:t) FAC':'OR l l6.l25 0.408 O.Ol.5 2 l.2.56.3 0.5054 0. Ol.S 3 6.3 *.562 0.948 0. 0 l.'.3 ORinCI OPINING OISCRARG? FT...OW
~
A~~~ (~_.2)
... ~
..:....;. (S ec ) COE:T!C!~ RAT! (lbm/sec) 0.0253 O.Ol.S 0.805 120.83 tTPSn!AM STL\M CONDITIONS PRl:SstnU:
P!U:SSUR.E (?SL\) R.!SE ?_o\T! (!SI)
- lee 269C. ~o.
DOWNS~ GAS CONDITIONS PRESSURE (?SL\) D~S!T! C;"1~)
l;. 0.099i5
~CT::S:
e*
_COTOFF ctTOF: I!mGRATION
-~E '!'llOtm.C'! T~
SJ 433 Hz 0.0009 S&c. O.S Sec. l0%
P!r! TO'IAL OUTSIDE
- SECTION u:NGTH (Ft) DUHITDt. (IN) ~nnc:n:ss (nl) ilt!Gcr! (~~ 1Ft) l 4 .* 73 8.62.5 0.906 . 74. 71 2 12 * .31 6.625 0.864 53 .* 16-3 12.43 6.625 0.28 18.97 4 69.0 12.15 0.688 sa. 60 5 l.l 12. 75 l * .s -
6 l.O 8.625 0.322 2S.53 1 O.S3 6.625 0.432 :S.57 6
EaoT * !COLD
- YOONG' s ~0Dut.US OF P,!P_!
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SIGMINT NO, WAT'S'9UG MODEL ,1111 NUMIU I
- il'l!!SUIIIIZER T'O 1.4 I
Z
3
- 1.G T'Q t.1 FIGURE_ 3A.3.A
- I WAiSL.UG MOOEl. CF EPR!
SAMPt..E PROBL.EM *
- - - - ~ - *1
.. I E] D.o*
IOG UO MIi
@ 120 JO 121. LEGEND, O* NODE NUii Bf 11 JII
- SUPPORT I 110 1111 140 A* MAIi POIHJ e
40 60 145 1:50 EJ 2i.1* E]* IPIU TlH CRH SI HG~ENT ~O.
10 411
© Ill IH Q
- NUPIP(-IW MODEL PIP( SECTION NO.
<D DO 6 180 I
N
@ 240 O"I I
2.D'
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@ 0 1D 110 1811 180 IH uo 2,:s SUPPORT 2 0 1 e
411.4 SUPPORT 4 SUPPORT~ WE 34/35 rn
\Al, 3Z/l3 FIGURf 5A. 3.A- 2 NUPIPE-SW MODEL EPRI SAMPLE. PROBLEM
D0,000 - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ,
0.0 r e
.A w
u
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0 IL N
J-N I
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Ill II)
' I
-ID0,000 .
200,000* L-----..I------L-----L------r:&--------1~----------........
0.0 0.1 0.f 0.J
'0.4 TIME tSEC)
LEOEND FIGURE 3A. 5.A- 5 WATSLU8 COMPARISON OF SEGMENT 2
llfLAP II/MOD I FORCINO FUNCTION
1)0,000 r
-.J IJJ 80,000 u
a:
~
If) l-I z N Ill S0,000 00 :I:
I I!)
Ill en 0.0 --
-20,000
-10,000 L------IL-------'1....----_...----=-------------------....._____,
,0.0 0.1 O.l 0.1 0.4 TIME (SEC)
LEGEND FIGURE 5A. 3.A - 4 WATSLU8 COMPARISON OF SEGMENT 3
RfLAPG/MOO I FORCING FUNCTION
e e
~-
10,:co---------------------
--* 4C,OOO
-z~
t,~
0
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= '"
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lo&,
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lo&,
en
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0.1 ----~----....J o.z 0.4 TIME (SEC) 1.!GENO ----
NUP111( *SW
EJl"I i!ST 111£SUL1'S
) FIGURE ~A.3.A *'
COMPARISON OF SEGMENT~, I SUPPORT ~EAC710N I
~
.-,t , . ,
, , . ... 1 e '
160,000---------------------,
80,000
- 119 0
.Q z
Q I-0 C:
Lu -80,000 a:
I-a:
0 Q.
Q.
Cl)
N
-160,000 I-z Lu
~
C, Lu (J)
-240,000 EPRI PEAK
-320,000 400,000"'-----i......-------------...__-_ __
0.1 0.2 o. 3 o. 4 TIME (SEC)
LEGEND NU PIPE- SW FIGURE 3A.3A-~
EPRI TEST RESULTS COMPARISON OF SEGMENT 2 SUPPORT REACT ION