Letter Sequence Other |
---|
|
Results
Other: ML20137H372, ML20197B076, ML20204G924, ML20205T170, ML20206B329, ML20206B459, ML20206F887, ML20207K386, ML20207K441, ML20207K446, ML20207K506, ML20207K512, ML20207P779, ML20207P991, ML20207P993, ML20209E329, ML20209F187, ML20209G043, ML20210A740, ML20210A748, ML20210A757, ML20210T436, ML20210T655, ML20210T686, ML20211D992, ML20211E058, ML20211E084, ML20211E110, ML20211G583, ML20211N368, ML20212M102, ML20214Q988, ML20214Q998, ML20214S836, ML20214S958, ML20215G796, ML20215H964, ML20215H973, ML20215J855, ML20215J871, ML20215L534, ML20216D570, ML20216E230, ML20234C109, ML20235E520, ML20235F508, ML20245C018
|
MONTHYEARML20137H3721985-11-26026 November 1985 Memorandum & Order Granting Extension of 851130 Deadline for Environ Qualification of Electrical Equipment to 860531 & Approving Proposal to Allow Operation w/35% Reactor Power Limit During Interim.Served on 851127 Project stage: Other ML20211E0841986-02-20020 February 1986 Issue a to Fort St Vrain:Delayed Firewater Cooldown;Effect of Liner Cooling on Orifice Valve Temps Project stage: Other ML20209F1871986-03-18018 March 1986 Fort St Vrain Steam Generator Temps During Interruption of Forced Cooling from 105% Power Project stage: Other ML20214S9581986-09-25025 September 1986 Requests That Util Take Listed Actions Re Three Repts Submitted Concerning Firewater Cooldown Analysis,Per NRC Review & 860917 Telcon.Response to Item 1 Requested within 15 Days of Ltr Date & Item 2 at Time Analysis Submitted Project stage: Other ML20211E0581986-09-30030 September 1986 Effect of Delayed Firewater Cooldown W/Loss of Liner Cooling on Pcrv Temps Project stage: Other ML20215G7961986-10-10010 October 1986 Advises That Info Requested in NRC Re Best Estimate Schedule for Resolution of Problems Encountered W/Firewater Cooldown Analysis Will Be Submitted as Part of Util Graduated rise-to-power Program Project stage: Other ML20197B0761986-10-22022 October 1986 Informs That Util Will Update & Submit Rept on Chernobyl Accident by 861126.Update Will Ctr on Graphite Related Concerns,Including Analysis of Worst Case Explosive Gas Mixtures & Comparison of Reactor Kinetics Behavior Project stage: Other ML20211G5831986-10-22022 October 1986 Anticipates Completion of Steam Generator Analysis & App R Modeling Reanalysis Work by Feb 1987,per 860918 Telcon W/Nrc Re Steam Generator Cool Down Studies for App R Project stage: Other ML20215L5341986-10-23023 October 1986 Staff Requirements Memo Re Commission Briefing in Washington,Dc on Status of Facility Project stage: Other ML20207K5121986-11-13013 November 1986 Fort St Vrain Calculations for Circulator Temp-Related Operating Limits Project stage: Other ML20207K5011986-12-0404 December 1986 Effect of Firewater Cooldown Using Economizer-Evaporator- Superheater (EES) Bundle on Steam Generator Structural Integrity. Draft Rept of Steam Generator Ability to Withstand post-App R Firewater Cooldown Transient Encl Project stage: Draft Other ML20207K4461986-12-12012 December 1986 Issue a to Effect of Firewater Cooldown Using Reheater on Steam Generator Structural Integrity Project stage: Other ML20211N3681986-12-12012 December 1986 Forwards Restart Interaction Schedule,Per 861205 Request Project stage: Other ML20207K5061986-12-22022 December 1986 Issue a to Effect of Intentional Depressurization on Cooldown from 39% Power Using One Reheater Module (1-1/2 H Delay) Project stage: Other ML20207K4411986-12-23023 December 1986 Issue a to Economizer-Evaporator-Superheater (EES) Cooldown from 39% & 78% Power Using Condensate or Firewater (1.5 H Delay) Project stage: Other ML20207K3861986-12-30030 December 1986 Forwards Analyses Supporting Power Operation Up to 39% Power Based on Safe Shutdown Cooling Following 90 Min Interruption of Forced Circulation.Conclusions of Repts Listed.Corrective Actions for LERs 86-020 & 86-026 Also Listed Project stage: Other ML20207P7791987-01-0707 January 1987 Forwards Current Integrated Schedule for Restart & Power Ascension Activities.Schedule Incorporates Consolidated Schedular Info on Both Interaction Activities.Updates Will Be Provided Twice Per Month.W/One Oversize Graph Project stage: Other ML20207P9931987-01-13013 January 1987 SAR for Tech Spec Limiting Condition for Operation 4.3.1 Change Permitting Safe Shutdown Cooling W/Evaporator- Economizer-Superheater Project stage: Other ML20207P9911987-01-15015 January 1987 Proposed Tech Specs,Requiring Both evaporator-economizer Superheater Sections & Both Reheater Sections to Be Available During Operation at Power as Min Number of HXs Project stage: Other ML20207P9891987-01-15015 January 1987 Application for Amend to License DPR-34,requiring Both evaporator-economizer Superheater Sections & Both Reheater Sections to Be Available During Operation at Power as Min Number of Operable HXs Project stage: Request ML20207P9871987-01-15015 January 1987 Forwards Application for Amend to License DPR-34,changing Tech Specs to Require Both evaporator-economizer Superheater Sections & Both Reheater Sections to Be Available During Operation at Power.Fee Paid Project stage: Request ML20211E1101987-01-26026 January 1987 Rev a to Engineering Evaluation of Procedure to Recover from Actuation of Steam Line Rupture Detection/Isolation Sys for Power Levels Through P2 Project stage: Other IR 05000267/19870021987-01-30030 January 1987 Partially Withheld Insp Rept 50-267/87-02 on 870106-09 (Ref 10CFR73.21).No Violations or Deviations Noted.Major Areas Inspected:Matl Control & Accounting Project stage: Request ML20210A7571987-01-30030 January 1987 Fort St Vrain 1987 Power Ascension Plan Project stage: Other ML20210A7481987-01-30030 January 1987 Requests Concurrence to Start Up & Operate Facility Through Graduated Rise to Power Up to 100% of Rated Power,Subj to Listed Constraints. Fort St Vrain 1987 Power Ascension Plan Encl Project stage: Other ML20210A7401987-02-0202 February 1987 Forwards Updated Nrc/Public Svc Co of Colorado Restart Interaction Schedule, Reflecting Current Target Dates & Recently Completed Items Project stage: Other ML20209G0431987-02-0202 February 1987 Forwards Current Integrated Schedule for Plant Restart & Power Ascension Activities.W/One Oversize Encl Project stage: Other ML20210N8831987-02-0303 February 1987 Forwards Request for Addl Info on 861230 & 870115 Submittals Re Analysis of Firewater Cooldown from 82% of Full Power Project stage: RAI ML20210P0191987-02-0505 February 1987 Summary of 870113 Meeting W/Util Re Completion of Equipment Qualification Program & Program & Approvals Required for Plant Restart Project stage: Meeting ML20210T6861987-02-0505 February 1987 Rev a to Engineering Evaluation of Reanalysis of FSAR Accidents/Transients Relying on EES Cooling. W/Four Oversize Drawings Project stage: Other ML20211D9921987-02-0505 February 1987 Issue a to Economizer-Evaporator-Superheater Cooldowns for Equipment Qualification & App R Events W/Vent Lines (1.5 H Delay) Project stage: Other ML20210T6551987-02-0606 February 1987 Provides Results of Confirmatory Analyses for FSAR Accidents Which Utilize Either EES or Reheater Section of Steam Generator for DHR Project stage: Other ML20210T4361987-02-11011 February 1987 Requests Publication of Fr Notice of Consideration of Issuance of Amend to License DPR-34 & Proposed NSHC Determination & Opportunity for Hearing on 870115 Request Re Operation of evaporator-economizer-superheater Sections Project stage: Other ML20211E9791987-02-12012 February 1987 Forwards Proposed Agenda & Slides for 870226 Meeting W/ Commission & Staff to Secure Commission Approval for Full Power Operation of Facility Project stage: Meeting ML20211D8901987-02-17017 February 1987 Forwards Response to NRC 870203 Request for Addl Info Re Firewater Cooldown from 82% of Full Power,Per Util 861230 & s Project stage: Request ML20207Q7941987-03-0303 March 1987 Forwards Second Request for Addl Info Re Util Analysis of Firewater Cooldown from 82% of Full Power Operation,Based on Review of 861230,870115 & 0217 Submittals Project stage: Approval ML20212M1021987-03-0505 March 1987 Forwards Assessment of Addl Concerns Per 870226 Request Re Restart of Facility.Events & Circumstances Involving Contracts Had No Adverse Impacts Re Plant Safety or Safe Plant Operations Project stage: Other ML20204G9241987-03-20020 March 1987 Forwards Restart & Power Ascension Schedule,Incorporating Consolidated Schedular Info on NRC-util Interaction Activities.Brief Narrative Description of Scope of Each Line Item Activity Also Encl.W/One Oversize Encl Project stage: Other ML20205B3441987-03-20020 March 1987 Forwards Response to NRC 870303 Second Request for Addl Info Re Firewater Cooldown from 82% of Full Power (Safe Shutdown Cooling) Project stage: Request ML20205M8901987-03-30030 March 1987 Forwards Third Request for Addl Info Re Util 861230,870115 & 0217 Submittals Concerning Analysis of Firewater Cooldown from 82% of Full Power.Major Concerns Re Effects of Transient Loading Due to Seismic Motion or Flow Project stage: RAI ML20205T1701987-04-0101 April 1987 Forwards Oversize Current Integrated Schedule for Facility Restart & Power Ascension Activities Required for Equipment Qualification Completion Certification,Startup/Plant Criticality & Power Ascension to 82%.Related Info Encl Project stage: Other ML20206B6031987-04-0101 April 1987 Forwards Comments Re Implication of Chernobyl Reactor Accident.Design Differences Between Fort St Vrain & Chernobyl Preclude Accident Similar to Chernobyl from Occurring at Fort St Vrain Project stage: Approval ML20206B4591987-04-0303 April 1987 Forwards Summary of Equipment Qualification (EQ) Insp Conducted by NRR & IE on 870126-30.EQ Program Approved. Detailed Results of Insp Will Be Provided Project stage: Other ML20206B3291987-04-0707 April 1987 Submits Daily Highlight.Public Svc Co of Colorado Authorized to Restart & Operate Facility HTGR at Level of Up to 35% Full Power.Facility Out of Operation Since 860531,when Shut Down for Equipment Qualification Mods Project stage: Other ML20206F8871987-04-10010 April 1987 Submits Requested Addl Info for Analysis of Firewater Cooldown for 82% Power Operation,Per Project stage: Other ML20209E3291987-04-27027 April 1987 Provides Written Authorization to Operate Reactor at Up to 35% Full Power,Per Section IV of 870406 Confirmatory Order Modifying License DPR-34 Project stage: Other ML20215H9641987-04-30030 April 1987 Forwards Updated Ga Technologies Procedure 909410, Buckle Users Manual, Per 870330 Request.Manual Updated to Include Revs to Computer Code Required by High Temps & Short Times Assumed for Steam Generator Tube Stress Analysis Project stage: Other ML20215H9731987-04-30030 April 1987 Revised Buckle Users Manual:Creep Collapse of Thin-Walled Circular Cylindrical Shells Subj to Radial Pressure & Thermal Gradients Project stage: Other ML20215J8711987-05-0404 May 1987 Rev a to Evaluation of Test Data for Confirmation of Fire Water Flow Rate to Circulator Water Turbine During EES Cooldown for Safe Shutdown Cooling Project stage: Other ML20215J8551987-05-0404 May 1987 Forwards Rev a to EE-EQ-0057, Evaluation of Test Data for Confirmation of Fire Water Flow Rate to Circulator Water Turbine During EES Cooldown for Safe Shutdown Cooling Project stage: Other 1987-01-07
[Table View] |
|
---|
Category:CORRESPONDENCE-LETTERS
MONTHYEARML20205G8771999-03-26026 March 1999 Forwards Copy of Cover Page from Fort St-Vrain Welding Manual,Which Had Been Listed as Encl on Page 4 of 990325 Reply to EA 98-081.Cover Page Had Been Inadvertently Left Out with Original Reply ML20197H8811998-12-0101 December 1998 Forwards Proposed Change to Fsv ISFSI Physical Protection Plan in Which Commitment Is Made to Provide Feature to Security Posture for Facility ML20236R9191998-07-20020 July 1998 Ltr Contract:Mod 4 to Task Order 27, Task Area No 4 of Basic Contract - Fort St Vrain Insp Under Contract NRC-02-95-003 ML20199H8141997-11-21021 November 1997 Responds to Requesting Clarification as to Whether Increase in Tritium & Iron-55 Contamination Limits That Were Approved for Plant Apply to All Licensees ML20198K1931997-10-10010 October 1997 Provides Supplemental Info in Support of Util Proposed Rev to Physical Security Plan for Plant Plant Isfsi.Plan Withheld,Per 10CFR2.790(d) & 10CFR73.21 ML20198H5601997-09-16016 September 1997 Final Response to FOIA Request for Documents.Documents Listed in App a Being Released in Entirety ML20141F3521997-05-14014 May 1997 Forwards Proposed Issue 4 of Physical Security Plan for Fort St Vrain ISFSI for Review & Approval.Encl Withheld,Per 10CFR2.790(d) ML20141C8611997-05-0909 May 1997 Informs of Approval of Fsv Final Survey Rept & Effluent Pathway Survey Plan & Supporting Analysis ML20141K9881997-05-0505 May 1997 Forwards Amend 89 to License DPR-34 & Supporting Safety Evaluation.Amend Designates All Elements of Approved Decommissioning Plan as License Termination Plan ML20138G2701997-04-28028 April 1997 Provides Response to NRC Comments Re Proposed Sampling & Survey Plan for Fsv Effluent Pathway.Response Documents Fsv Liquid Effluent Discharge Pathway Areas Are Acceptable for Release for Unrestricted Use IAW Draft NUREG/CR-5849 ML20148D4651997-04-24024 April 1997 Forwards Revised Interim Ltr Rept Which Describes Procedures & Results of Confirmatory Survey of Group E Effluent Discharge Pathway Areas at Fsv Station NUREG/CR-5849, Requests That Licensee Provide Evidence That Average Contamination Levels in Group E Effluent Discharge Pathway Areas Meet Averaging Criteria in Draft NUREG/CR-58491997-04-23023 April 1997 Requests That Licensee Provide Evidence That Average Contamination Levels in Group E Effluent Discharge Pathway Areas Meet Averaging Criteria in Draft NUREG/CR-5849 ML20138B0511997-04-22022 April 1997 Forwards Copy of Proposed Amend to Fsv NPDES Permit, Wastewater Discharge Permit CO-0001121 Requested to Support Repowering Activities,Iaw Section 3.2.d of Fsv Non-Radiological Ts,App B to License DPR-34 ML20140E1061997-04-10010 April 1997 Forwards Confirmatory Survey of Group Effluent Discharge Pathway Areas for Fsv Nuclear Station,Platteville,Co ML20137S4481997-04-0808 April 1997 Informs That Decommissioning Activities at Fsv Are Complete & NRC Issued Exemption from Requirements of 10CFR50.54(w) in .Property Damage Insurance Policy Is Maintaned to Protect Fsv balance-of-plant Assets ML20137S0821997-04-0707 April 1997 Forwards Insp Rept 50-267/97-01 on 970310-11.No Violations Noted ML20137S1691997-04-0707 April 1997 Fifth Partial Response to FOIA Request for Documents. Forwards Documents Listed in App K ML20137S5421997-04-0707 April 1997 Forwards Final Quarterly 10CFR50.59 Rept of Changes,Tests & Experiments Affecting Decommissioning of Fort St Vrain Nuclear Station ML20148D5951997-04-0404 April 1997 Forwards Confirmatory Survey for Fsv Nuclear Station, Psc,Platteville,Co, Final Rept ML20137R6921997-04-0404 April 1997 Informs of Approval for Request for Addl 45 Days to Remedy Deficiencies Identified in NRC Re Financial Assurance Mechanism for Fsv Decommissioning Costs ML20137J8051997-03-31031 March 1997 Third Partial Response to FOIA Request for Documents.Records in App F Encl & Will Be Available in Pdr.App G & H Records Withheld in Part (Ref FOIA Exemptions 5 & 7) ML20148D5811997-03-26026 March 1997 Forwards Confirmatory Survey Plan for Group E Effluent Discharge Pathway Areas at Fsv Nuclear Station, Covered in Final Survey Rept,Vol 6 ML20137G7361997-03-25025 March 1997 Requests Addl Time for Util to Respond to NRC Comments in Re Financial Assurance Mechanism for Fort St Vrain Decommissioning Costs ML20137G9521997-03-24024 March 1997 Forwards Quarterly 10CFR50.59 Rept for Period 961201-970228 Re Changes,Tests & Experiments for Fort St Vrain Decommissioning ML20137H1131997-03-24024 March 1997 Second Partial Response to FOIA Request for Documents. Forwards Documents Listed in App D.Documents Also Available in Pdr.Documents Listed in App E Withheld in Part (Ref FOIA Exemption 6) ML20137C0181997-03-18018 March 1997 Documents That No Personnel Has Received Radiation Exposure at Fsv in 1997 or at Any Time Subsequent to ML20137C0061997-03-18018 March 1997 Documents That There Have Been No Activities Involving Release of Radioactive Matls from Fsv Nuclear Station That Potentially Could Have Affected Environ,Subsequent to Previous Radiological Envion Operating Rept ML20136G1201997-03-11011 March 1997 Forwards Annual Radioactive Effluent Release Rept for Jan- Dec 1996 & Jan-Mar 1997. All Effluent Releases Completed as of 960703.Repts on Activities After 960703 Reflect Disposal of Solid Waste ML20136B1331997-02-28028 February 1997 First Partial Response to FOIA Request for Documents. Documents Listed in App a Already Available in Pdr.Forwards App B Documents.App C Documents Being Withheld in Entirety (Ref FOIA Exemption 5) ML20135D7891997-02-27027 February 1997 Forwards Responses to Comments Re Fort St Vrain Final Survey Rept ML20135D9531997-02-27027 February 1997 Forwards Copy of Amend to Util Npdes,Wastewater Discharge Permit CO-0001121,which Clarifies That Monitoring of Farm Pond Outlet Required When Industrial Wastewater Being Discharged Through Upstream Goosequill Ditch ML20135A8711997-02-14014 February 1997 Requests That Encl Deficiencies Identified in Financial Assurance Mechanism for Fort St Vrain Decommissioning Cost Be Addressed within 45 Days ML20134D1551997-01-31031 January 1997 Forwards Util Responses to NRC Comments Provided in NRC Ltr Re Sampling & Survey Plan Used for Final Radiological Survey of Liquid Effluent Pathway at Ft St Vrain ML20134C8481997-01-30030 January 1997 Forwards Draft Confirmatory Survey Rept for Fsv Nuclear Station,Psc,Platteville,Co Providing Info on Essap Activities on 960930-1003 ML20133L4961997-01-0707 January 1997 Forwards Comments That Need to Be Resolved Before Final Approval of Util Submittal Entitled, Proposed Sampling & Survey Plan for Effluent Pathway,Ft St Vrain Final Survey Program ML20133E0481997-01-0202 January 1997 Forwards Comments to Fsv Nuclear Station, Decommissioning Project Final Survey Rept (Volumes 4-11), for Consideration ML20132G0421996-12-23023 December 1996 Forwards Insp Rept 50-267/96-05 on 961203-05.No Violations Noted ML20132F2841996-12-19019 December 1996 Forwards Quarterly 10CFR50.59 Rept of Changes,Tests & Experiments Affecting Decommissioning of Plant,Covering Period of 960901-1130 ML20133A8591996-12-16016 December 1996 Forwards Original & Copy Transcripts of Public Hearing,Held on 961203 in Platteville,Co Re Decommissioning & License Termination of Util Ft Saint Vrain Nuclear Generating Station ML20133N0011996-12-0404 December 1996 Recommends That NRC Require License to Modify Submission of Unexecuted Draft Trust Agreement Remaining Decommissioning Costs for Ft St Vrain Nuclear Generating Station in Listed Ways ML20135B3861996-11-25025 November 1996 Informs That NRC Reviewed Util 961114 Submittal (P-96096) Entitled, Fort St Vrain Final Emergency Response Plan, & Meets Requirements of 10CFR50.54(q) ML20135A5861996-11-25025 November 1996 Submits Suppl Info Re Annual Environ Rept for 1995 Operation of Fsv ISFSI ML20135A6361996-11-20020 November 1996 Submits Copy of Describing Discharge Practices for Groundwater Seeping Into Fsv'S Reactor Building Sump ML20134L4721996-11-14014 November 1996 Notifies NRC That Util Adopted Fsv ISFSI Emergency Response Plan to Direct Emergency Response for Radiological Accidents Occuring at Site,Until 10CFR50 License Is Terminated ML20134F4351996-10-30030 October 1996 Forwards Sections 1,2,6 & 8 from Survey Packages F0015, F0039 & F0126 & Sections 1,2 & 6 from Survey Package F0115 to Support on-site NRC Insp ML20134G5991996-10-30030 October 1996 Forwards Volumes 1-12 to Final Survey Rept for Groups A,B,C Rev 1,D Rev 1,E,F Rev 1 & G-J for NRC Approval in Support of Forthcoming Request for Termination of Fsv 10CFR50 License ML20133D7691996-10-22022 October 1996 Forwards Preliminary Rept Re Orise Support of NRC License Insp at Fsv on 960930-1003 ML20136B1411996-10-15015 October 1996 FOIA Request for Documents Re NOV Addressed to Scientific Ecology Group Re NRC Insp Rept 50-267/94-03 & OI Investigation Repts 4-94-010 & 4-95-015 ML20128M6181996-10-0404 October 1996 Forwards Ltr from PSC to Co Dept of Public Health & Environ Describing Monitoring Practices at Plant ML20128G8041996-10-0101 October 1996 Forwards Fsv Decommissioning Fire Protection Plan Update 1999-03-26
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARML20205G8771999-03-26026 March 1999 Forwards Copy of Cover Page from Fort St-Vrain Welding Manual,Which Had Been Listed as Encl on Page 4 of 990325 Reply to EA 98-081.Cover Page Had Been Inadvertently Left Out with Original Reply ML20197H8811998-12-0101 December 1998 Forwards Proposed Change to Fsv ISFSI Physical Protection Plan in Which Commitment Is Made to Provide Feature to Security Posture for Facility ML20198K1931997-10-10010 October 1997 Provides Supplemental Info in Support of Util Proposed Rev to Physical Security Plan for Plant Plant Isfsi.Plan Withheld,Per 10CFR2.790(d) & 10CFR73.21 ML20141F3521997-05-14014 May 1997 Forwards Proposed Issue 4 of Physical Security Plan for Fort St Vrain ISFSI for Review & Approval.Encl Withheld,Per 10CFR2.790(d) ML20138G2701997-04-28028 April 1997 Provides Response to NRC Comments Re Proposed Sampling & Survey Plan for Fsv Effluent Pathway.Response Documents Fsv Liquid Effluent Discharge Pathway Areas Are Acceptable for Release for Unrestricted Use IAW Draft NUREG/CR-5849 ML20148D4651997-04-24024 April 1997 Forwards Revised Interim Ltr Rept Which Describes Procedures & Results of Confirmatory Survey of Group E Effluent Discharge Pathway Areas at Fsv Station ML20138B0511997-04-22022 April 1997 Forwards Copy of Proposed Amend to Fsv NPDES Permit, Wastewater Discharge Permit CO-0001121 Requested to Support Repowering Activities,Iaw Section 3.2.d of Fsv Non-Radiological Ts,App B to License DPR-34 ML20140E1061997-04-10010 April 1997 Forwards Confirmatory Survey of Group Effluent Discharge Pathway Areas for Fsv Nuclear Station,Platteville,Co ML20137S4481997-04-0808 April 1997 Informs That Decommissioning Activities at Fsv Are Complete & NRC Issued Exemption from Requirements of 10CFR50.54(w) in .Property Damage Insurance Policy Is Maintaned to Protect Fsv balance-of-plant Assets ML20137S5421997-04-0707 April 1997 Forwards Final Quarterly 10CFR50.59 Rept of Changes,Tests & Experiments Affecting Decommissioning of Fort St Vrain Nuclear Station ML20148D5951997-04-0404 April 1997 Forwards Confirmatory Survey for Fsv Nuclear Station, Psc,Platteville,Co, Final Rept ML20148D5811997-03-26026 March 1997 Forwards Confirmatory Survey Plan for Group E Effluent Discharge Pathway Areas at Fsv Nuclear Station, Covered in Final Survey Rept,Vol 6 ML20137G7361997-03-25025 March 1997 Requests Addl Time for Util to Respond to NRC Comments in Re Financial Assurance Mechanism for Fort St Vrain Decommissioning Costs ML20137G9521997-03-24024 March 1997 Forwards Quarterly 10CFR50.59 Rept for Period 961201-970228 Re Changes,Tests & Experiments for Fort St Vrain Decommissioning ML20137C0061997-03-18018 March 1997 Documents That There Have Been No Activities Involving Release of Radioactive Matls from Fsv Nuclear Station That Potentially Could Have Affected Environ,Subsequent to Previous Radiological Envion Operating Rept ML20137C0181997-03-18018 March 1997 Documents That No Personnel Has Received Radiation Exposure at Fsv in 1997 or at Any Time Subsequent to ML20136G1201997-03-11011 March 1997 Forwards Annual Radioactive Effluent Release Rept for Jan- Dec 1996 & Jan-Mar 1997. All Effluent Releases Completed as of 960703.Repts on Activities After 960703 Reflect Disposal of Solid Waste ML20135D7891997-02-27027 February 1997 Forwards Responses to Comments Re Fort St Vrain Final Survey Rept ML20135D9531997-02-27027 February 1997 Forwards Copy of Amend to Util Npdes,Wastewater Discharge Permit CO-0001121,which Clarifies That Monitoring of Farm Pond Outlet Required When Industrial Wastewater Being Discharged Through Upstream Goosequill Ditch ML20134D1551997-01-31031 January 1997 Forwards Util Responses to NRC Comments Provided in NRC Ltr Re Sampling & Survey Plan Used for Final Radiological Survey of Liquid Effluent Pathway at Ft St Vrain ML20134C8481997-01-30030 January 1997 Forwards Draft Confirmatory Survey Rept for Fsv Nuclear Station,Psc,Platteville,Co Providing Info on Essap Activities on 960930-1003 ML20133E0481997-01-0202 January 1997 Forwards Comments to Fsv Nuclear Station, Decommissioning Project Final Survey Rept (Volumes 4-11), for Consideration ML20132F2841996-12-19019 December 1996 Forwards Quarterly 10CFR50.59 Rept of Changes,Tests & Experiments Affecting Decommissioning of Plant,Covering Period of 960901-1130 ML20133A8591996-12-16016 December 1996 Forwards Original & Copy Transcripts of Public Hearing,Held on 961203 in Platteville,Co Re Decommissioning & License Termination of Util Ft Saint Vrain Nuclear Generating Station ML20133N0011996-12-0404 December 1996 Recommends That NRC Require License to Modify Submission of Unexecuted Draft Trust Agreement Remaining Decommissioning Costs for Ft St Vrain Nuclear Generating Station in Listed Ways ML20135A5861996-11-25025 November 1996 Submits Suppl Info Re Annual Environ Rept for 1995 Operation of Fsv ISFSI ML20135A6361996-11-20020 November 1996 Submits Copy of Describing Discharge Practices for Groundwater Seeping Into Fsv'S Reactor Building Sump ML20134L4721996-11-14014 November 1996 Notifies NRC That Util Adopted Fsv ISFSI Emergency Response Plan to Direct Emergency Response for Radiological Accidents Occuring at Site,Until 10CFR50 License Is Terminated ML20134F4351996-10-30030 October 1996 Forwards Sections 1,2,6 & 8 from Survey Packages F0015, F0039 & F0126 & Sections 1,2 & 6 from Survey Package F0115 to Support on-site NRC Insp ML20134G5991996-10-30030 October 1996 Forwards Volumes 1-12 to Final Survey Rept for Groups A,B,C Rev 1,D Rev 1,E,F Rev 1 & G-J for NRC Approval in Support of Forthcoming Request for Termination of Fsv 10CFR50 License ML20133D7691996-10-22022 October 1996 Forwards Preliminary Rept Re Orise Support of NRC License Insp at Fsv on 960930-1003 ML20136B1411996-10-15015 October 1996 FOIA Request for Documents Re NOV Addressed to Scientific Ecology Group Re NRC Insp Rept 50-267/94-03 & OI Investigation Repts 4-94-010 & 4-95-015 ML20128M6181996-10-0404 October 1996 Forwards Ltr from PSC to Co Dept of Public Health & Environ Describing Monitoring Practices at Plant ML20128G8041996-10-0101 October 1996 Forwards Fsv Decommissioning Fire Protection Plan Update ML20128G0481996-09-30030 September 1996 Submits Rev to Psco Definitions of Contents of Documentation Packages Re Fsv Final Survey Project ML20129C0421996-09-20020 September 1996 Forwards Quarterly Submittal of 10CFR50.59 Rept of Changes, Tests & Experiments for Facility Decommissioning,Covering Period of 960601-0831 ML20133D7601996-09-16016 September 1996 Forwards Confirmatory Survey Plan for Fsv Nuclear Station Decommissioning Project,First Final Survey Rept Submittal- Vols 1-5.NRC Comments Incorporated.Spending Plan Attached ML20117P0711996-09-13013 September 1996 Describes Util Plans to Remove Bldg 28 from Plant Facility ML20129A4431996-09-11011 September 1996 Describes Util Plans for Demonstrating That Liquid Effluent Pathway & Surrounding Open Land Areas Satisfy 10 Mrem/Yr Criteria Provided in Plant Final Survey Plan ML20117K5291996-09-0404 September 1996 Provides Notification That Util Will Be Revising Financial Assurance Mechanism That Will Be Used to Cover Remaining Costs of Decommissioning Plant ML20117C7281996-08-22022 August 1996 Discusses Impact of Final Decommissioning Rule & Requests NRC Concurrence That Requirements to Submit & Obtain Approval of License Termination Plan Have Been Satisfied ML20116P3431996-08-16016 August 1996 Describes Actions to Remove Structures & Equipment Items from Fort St Vrain Facility for NRC Info.Requests That NRC Advise Util of Wishes to Perform Confirmatory Survey of Any Parts of New Fuel Storage Building Before 960903 ML20133D7551996-08-14014 August 1996 Provides Environ Survey & Site Assessment Program'S (Essap) Comments Re Review of Fsv Nuclear Station Decommissioning Project Final Survey Rept ML20116M0771996-08-14014 August 1996 Provides Suppl Response to Re Insp Rept 50-267/96-01 in Jan 1996 Re NRC Concerns About Fsv Final Survey Program.Specifically,Bias in Instrumentation Response Overestimating Amount of Contamination Present ML20116M1841996-08-13013 August 1996 Forwards Util Responses to NRC Comments in Re Use of in-situ Gamma Spectroscopy to Measure Exposure Rates During Plant Final Survey.Approval to Use in-situ Gamma Spectroscopic instrument,Microspec-2,requested ML20116K0061996-08-0909 August 1996 Submits Fort St Vrain Nuclear Station Decommissioning Project Final Survey Rept ML20116M1241996-08-0808 August 1996 Responds to NRC Bulletin 96-004, Chemical,Galvanic,Or Other Reactions in Spent Fuel Storage & Transportation. Informs That Modular Vault Dry Storage Sys Is Not Susceptible to Problems Addressed in Bulletin ML20116F3611996-08-0202 August 1996 Submits Revised Documentation for Fort St Vrain Final Survey Program ML20116F8141996-08-0202 August 1996 Informs of Util Intent to Modify Fort St Vrain Control Room,Which Will Make Certain Final Survey Locations Unavailable for Further Review.Final Survey Efforts Are Complete ML20116A4511996-07-19019 July 1996 Requests NRC Approval of Proposed Method to Fsv Final Survey Plan to Determine Exposure Rates in Prestressed Concrete Reactor Vessel 1999-03-26
[Table view] Category:UTILITY TO NRC
MONTHYEARML20059H9471990-09-14014 September 1990 Forwards Rev to Proposed Physical Security Plan,Per .Encl Withheld (Ref 10CFR73.21) ML20064A6841990-09-14014 September 1990 Notifies That Ms Frazier License OP-5608-1 Terminated Effective 900906,per 10CFR50.74 ML20059K2091990-09-14014 September 1990 Requests That Dj Trumblee Reactor Operator License Be Terminated,Effective 900912 ML20059L5831990-09-14014 September 1990 Forwards Application for Amend to License DPR-34,changing Tech Spec Design Features Section to Remove CRD & Orifice Assemblies from Core Regions Defueled in Support of Plant Closure Activities ML20059D3231990-08-15015 August 1990 Forwards fitness-for-duty Performance Data Rept for Plant ML20058N5661990-08-0808 August 1990 Requests Exemption from 10CFR50.54(w)(1) Requirements for Min Property Insurance Coverage ML20058L8031990-08-0303 August 1990 Forwards GA-C18103, Fort St Vrain Cycle 3 Core Performance, Per Request for Addl Info Re Proposed Amend to Tech Spec Which Would Allow Util to Complete Defueling of Facility ML20058M1931990-07-31031 July 1990 Advises of Termination of Kj Einig Reactor Operator License, Effective 900727 ML20056A1791990-07-30030 July 1990 Responds to NRC Re Violations Noted in Insp Rept 50-267/90-07.Corrective Actions:Electrical & Mechanical Craft Personnel Assigned to Planning & Scheduling Organization ML20055H8091990-07-23023 July 1990 Forwards Addl Info in Support of 900622 SAR for Isfsi. Proprietary SAR Withheld (Ref 10CFR2.790) ML20056A1711990-07-20020 July 1990 Discusses Decommissioning Funding for Plant.Util Revised Decommissioning Financial Plan,Based on Safstor,Currently in Effect & Funds Will Be Accumulated Per Plan Provisions ML20055H6571990-07-20020 July 1990 Responds to NRC Bulletin 90-001, Loss of Fill Oil in Transmitters Mfg by Rosemount. No Rosemount Model 1153 or 1154 Transmitters Presently in Use at Plant ML20058P6891990-07-20020 July 1990 Forwards Security Plan for Proposed ISFSI for NRC Review,Per 900622 & 0412 Ltrs.Plan Withheld (Ref 10CFR73.21) ML20055G8601990-07-20020 July 1990 Responds to NRC Re Violations Noted in Insp Rept 50-267/90-07.Corrective Actions:Site Mgt Recognized Applicability of 10CFR50.72(b)(2)(vi) Re Previous Day Notification & Immediately Notified NRC ML20055H2811990-07-20020 July 1990 Forwards Rev 4 to Updated Fire Protection Plan ML20055G7061990-07-20020 July 1990 Forwards Application for Amend to License DPR-34,revising Tech Spec 6.1 to Remove Equipment to Support Plant Closure ML20059D1361990-07-13013 July 1990 Applies for Exemption from Payment of Annual Fees for FY90, Per 10CFR171 ML20055E0401990-07-0303 July 1990 Forwards Issue 3 to FPOR-18, Fire Protection Operability Requirements, as Result of Rev 3 to Facility Fire Protection Plan ML20055E0331990-07-0202 July 1990 Informs of Plans to Remove B Helium Circulator from Plant Prestressed Concrete Reactor Vessel in Mannner Similar to Previous Circulator Maint Outages.Requests Timely Ack of NRC Concurrence W/Undertaking ML20055E9171990-06-29029 June 1990 Resubmits Semiannual Radioactive Effluent Release Rept Jul-Dec 1989, as Result of Omitted Tables ML20044A8281990-06-29029 June 1990 Notifies of Termination of Wj Ashmore,License SOP-43244 & Requests That Reactor Operator License Be Terminated Effective 900620,per 10CFR50.74 ML20055D3931990-06-29029 June 1990 Forwards Response to Generic Ltr 90-04,requesting Info on Status of Generic Safety Issues.W/O Encl ML20055D1591990-06-22022 June 1990 Forwards Proposed Change to Updated Fsar,App B Re Use of Controls Other than Change Notice Sys to Implement Mods Such as Cutting & Capping of Piping,In Order to Obtain Positive Equipment Isolation ML20043J0581990-06-15015 June 1990 Forwards Proposed Issue 5 to Fort St Vrain Nuclear Generating Station Defueling Emergency Response Plan, Deleting Offsite Emergency Response Capabilities,Per Util 890816 Defueling SAR & NUREG-0654 ML20043D3291990-05-25025 May 1990 Forwards Response to NRC 900330 Ltr Re Violations Noted in Insp Rept 50-267/90-04.Encl Withheld (Ref 10CFR73.21) ML20055D2051990-05-25025 May 1990 Responds to Violations Noted in Insp Rept 50-267/90-06. Corrective Actions:Extensive Training on Radiological Emergency Response Plan Implementing Procedures Will Be Provided to Operating Crews Beginning in May 1990 ML20043C8981990-05-25025 May 1990 Responds to NRC 900427 Ltr Re Violations Noted in Insp Rept 50-267/90-05.Corrective Actions:Mgt Will Continue to Stress Requirement for Procedural Compliance on Ongoing Basis ML20043B3691990-05-22022 May 1990 Describes Util Current Plans for Removal of Certain Plant Equipment Items.Components to Be Removed Have No Required Nor Useful Function During Planned or Postulated Remaining Defueling or Shutdown Conditions ML20043A9491990-05-11011 May 1990 Forwards Issue 59 to AOP-K-1, Environ Disturbances - Earthquake. ML20042F9771990-05-0404 May 1990 Forwards Assessment of Possible Sources of Water in Plant Core,Means of Detecting Water in Pcrv & Effects of Water Ingress on Reactivity,For Info ML20042F9011990-05-0101 May 1990 Forwards Conceptual Plan & Cost Estimates for Early Dismantlement of Fort St Vrain Pcrv, Per NRC 900315 Request ML20042E8911990-04-27027 April 1990 Requests 30-day Extension to 900529 to Submit Response to Violations Noted in Insp Rept 50-267/90-04 ML20042F3111990-04-26026 April 1990 Forwards Application to Amend License DPR-34,providing Tech Spec Changes Needed to Complete Defueling ML20012D4711990-03-19019 March 1990 Requests Approval of Encl Exemption Request Allowing Util to Accumulate Decommissioning Funds Beyond Termination of Operations ML20012B7541990-03-0707 March 1990 Informs of Electronic Transfer of Fees on 900307 for Invoice IO547 ML20011F4691990-02-22022 February 1990 Responds to NRC 900116 Ltr Re Violations Noted in Insp Rept 50-267/89-23.Violation Re Fire Doors 13 & 9 Between 480-volt Ac Switchgear Room & Bldg 10 Propped Open on 891129 Disputed.Improved Communications Underway ML20006E7821990-02-15015 February 1990 Forwards Responses to 891004 Questions Re Decommissioning Financial Plan ML20011E8941990-02-0707 February 1990 Responds to 890719 NRC Bulletin 89-002 Re safety-related Anchor/Darling Model S350W Swing Check Valves.Licensee Reviewed Plant Data Base & Design Documents & Concludes That No Anchor/Darling Model S350W Valves Installed at Plant ML20011E7861990-02-0707 February 1990 Forwards Overview of Evaluation on Compliance W/Environ Qualification Regulation (10CFR50.49) During Defueling of Facility.During Defueling,Pcrv Will Be Maintained at Nearly Atmospheric Pressures & Rapid Depressurization Not Possible ML20011E5031990-02-0606 February 1990 Notifies of Termination of Ma Rhoton Employment W/Util, Effective 900206 ML20006D6621990-02-0606 February 1990 Responds to Generic Ltr 89-13, Svc Water Sys Problems Affecting Safety-Related Equipment. Currently Established Preventive Maint & Tech Spec Surveillance Activities on HXs Remaining in Svc Will Continue ML20006B2091990-01-26026 January 1990 Forwards Fire Operability Requirements 3 & 14,Issues 3 & 4, Respectively,To Fire Protection Program Plan,Section FP.6.1. Changes Cover Halon Sys & Basis & Records Ctr Halon Sys ML20006B7791990-01-25025 January 1990 Forwards Application for Amend to License DPR-34,changing Section 7.1 of Tech Specs.Specifically,Senior Vice President,Nuclear Operations Revised to Vice President, Nuclear Operations & Nuclear Training Manager Revised ML20006A5001990-01-19019 January 1990 Notifies of Util Intent to Bore 10 Concrete Core Drills Into Prestressed Concrete Reactor Vessel Barrel Section at Various Elevations Found Necessary for Decommissioning Planning.Drillings Wil Begin on 900205.Assessment Encl ML20006A5041990-01-19019 January 1990 Responds to NRC 891218 Ltr Re Weaknesses Noted in Insp Rept 50-267/89-22.Corrective Actions:During Power Operations Three Licensed Reactor Operators Will Be Present in Control & Short in Emergency Public Address Sys Repaired ML20005G5001990-01-12012 January 1990 Forwards Approved Proposed Change to NPDES Permit to Use Halocide Product for Microbiological Control in Smaller Cooling Tower ML19354D8581990-01-0505 January 1990 Forwards Safety Assessment of Proposed Changes to Fire Protection Program Plan,Per P Erickson Request.Changes Cover Acceptance Criteria for Shutdown Cooling Following Fire & New Fire Protection Cooldown Trains 1 & 2 ML20005F5531990-01-0303 January 1990 Certifies That Util Has Implemented fitness-for-duty Program Meeting 10CFR26 Requirements ML20005D8921989-12-21021 December 1989 Forwards Proposed Change to Plant NPDES Permit ML19354D7441989-12-0808 December 1989 Requests Relief from Requirements of Generic Ltr 89-10 Re motor-operated Valves Since Plant Permanently Shut Down Since 890818 & Defueling Initiated on 891127 1990-09-14
[Table view] |
Text
.
h Public Service ~ mm 2420 W. 26th Avenue, Suite 1000, Denver, Colorado 80211 April 10, 1987 Fort St. Vrain Unit No. 1 P-87122 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Attention: Mr. H. N. Berkow, Director Standardization and Special Projects Directorate Docket No. 50-267
SUBJECT:
Additional Information for Analysis of Firewater Cooldowi for 82% Power Operation
REFERENCES:
(SeeAttachment1)
Dear Mr. Berkow:
Reference I requested that PSC provide information concerning the !
firewater cooldown from 82% of full power, in addition to that which was presented by PSC in References 2 and 3. PSC provided responses to NRC's first request for additional information in Reference 4. .
PSC provided responses to NRC's second request (Reference 5) for l additional information in Reference 6.
Attachment 2 provides PSC's responses to the NRC's third request (Reference 7) for additional information concerning PSC's original submittals in References 2 and 3. This additional information is provided to enable the NRC to complete its review of PSC's analysis of firewater cooldown from 82% of full power.
0 ott 8704140391 870410 0 gDR ADOCK 05000267 PDR L
h
P-87122
. Page 2
. April 10, 1987 If you have any questions about the responses in Attachment 2, please contact Mr. M. H. Holmes at (303) 480-6960.
Very truly yours, t
hH.L.Brey, Manager Nuclear Licensing and Fuels Division HLB /AW:jmt Attachments cc: Regional Administrator, Region IV Attention: Mr. J. E. Gagliardo, Chief Reactor Projects Branch Mr. R. E. Farrell Senior Resident Inspector Fort St. Vrain
Attachment 1 to P-87122 Page 1 REFERENCES 4
- 1) NRC letter, Heitner to Williams, dated February 3, 1987 (G-87031)
- 2) PSC letter, Warembourg to Berkow, dated December 30, 1986 (P-86683)
- 3) PSC letter, Williams to Berkow, dated January 15, 1987 (P-87002) f 4) PSC letter, Brey to Berkow, dated February 17, 1987 (P-87055)
- 5) NRC letter, Heitner to Williams, dated March 3, 1987 (G-87060)
- 6) PSC letter, Brey to Berkow, dated March 20, 1987 (P-87110) 1
- 7) NRC letter, Heitner to Williams, dated March 30, 1987 (G-87100) <
A I
1 i
)
i l
i 1
I
c.
Attachment 2 to P-87122 '
Page 1 ,
REQUEST FOR ADDITIONAL INFORMATION i
NRC REQUEST 1 The stress evaluation regions in the steam generator during the single cycle cooldown from power using firewater in the EES tube bundles did not consider the effects due to flow induced vibration or dynamic loading due to an SSE.
- a. Provide justification for not including the effects due to flow-induced vibration,
- b. Provide the basis, method and results of the stress evaluations l '
which include the effects due to SSE loading .for the Reheater tubes and their supports, the EES tubes and the EES tube support structure, the superheater helical tube bundle and its support structure, and the superheater downcomer.
PSC RESPONSE 1 la. Flow Induced Vibration Effects:
The effects of flow induced vibration were not included in the
, evaluation of the steam generator during the single cycle cooldown from power using firewater in the EES tube bundles. In the current safe shutdown cooling and Appendix R cooling analyses, the maximum helium flows are less than 4% of rated values. In testing which duplicated full operational flow velocities described in FSAR Appendices A.13.6 and A.13.7, flow induced vibration stresses were generally less than predicted and small fractions of ASME Code allowables. Therefore loadings induced by helium flows of only 4% of rated flow are considered negligible.
Ib, SSE Loading Effects:
The primary stresses due to an SSE were considered in conjunction with normal and upset operational loadings in the steam generator design report, GADR-9. SSE loadings were not combined with other i
rare events nor was there a requirement for this combination. Fort St. Vrain steam generators were designed to the ASME Boiler and Pressure Vessel Code, 1965 edition as amended by all addenda up to i and including the Winter 1966 addenda. The above ASME code edition did not contain provisions for emergency and faulted conditions. The Summer 1968 addenda to Section III of the ASME Code first addressed emergency and faulted conditions. The first regulatory guidance for
- combining SSE loadings with emergency and faulted events was Regulatory Guide 1.48, " Design Limits and Loading Combinations for
- Seismic Category 1 Fluid System Components," May 1973. This regulatory guide was not backfitted to older plants including Fort
- St. Vrain. Regulatory Guide 1.48 was subsequently withdrawn in 1985, and was replaced by the July 1981 revision of Standard Review Plan i
v.
. Attachment 2 to P-87122 Page 2 3.9.3 "ASME Code Class 1, 2, and 3 Components, Component Supports, and Core Support Structures."
It is therefore concluded that combining the SSE with faulted events is a current regulatory requirement but not applicable to plants constructed prior to Regulatory Guide 1.48 in 1973. This is further substantiated by the NRC Systematic Evaluation Program (SEP) conducted in the late 1970's and early 1980's which evaluated the oldest eleven nuclear power plants to selected current licensing requirements. Load combinations were an SEP topic but were deleted because load combinations were also an unresolved safety issue (USI).
Load combinations remain an unresolved safety issue today, and are USI Item B-6. The following is a quote from Page 2.B.6-2 of NUREG-0933, A Prioritization of Generic Safety Issues, " Reactors constructed prior to 1972 did not have design requirements which included SSE, LOCA, and pipe cooling considerations."
Based on the above information, PSC concludes that the NRC position, expressed in Reference 7, that SSE loadings must be combined with the loadings of other rare events initiating a firewater cooldown (i.e.,
tornadoes, HELB, Appendix R fires) represents a change in a previously accepted staff position. It is not apparent to PSC that a substantial additional safety benefit would result from this new position. However, without accepting that the requested load combination is applicable to Fort St. Vrain, we are responding to your specific request for additional information to facilitate a timely review of PSC's safe shutdown cooling analyses. The responses in fact illustrate that SSE loadings comprise only a small portion of the total loading during normal operation as well as during firewater cooldowns.
EES Tube Bundle:
The effects of the SSE event on the Superheater II tubes are already included in the GA Document 909204, Issue N/C, evaluation, but were not specifically identified as such. The stresses due to dead weight on Page 16 on GA Document 909204 include the seismic contribution.
These stress values were obtained from Section 6.3.4, Page 20 of GADR-9.
See the Attached Table Ib, for a detailed stress evaluation including stress allowables for the Superheater II. From Table Ib it can be seen that seismic induced stress constitutes only a small fraction (6%) of the total stress which in turn is within code allowable limits.
Superheater Downcomer:
Since design temperature for the Superheater Downcomer is 1221 degrees F (GADR-9 Summary Table) and the tube temperature during the firewater cooldown event is 300 degrees F (Page A-5 of GA Document 909204), the restrained thermal expansion stresses will be reduced accordingly. The pressure stresses are lower by a factor of almost three during the firewater cooldown event (Table 5.1 of GA Document 909204). The SSE stress of 614 psi (GADR-9 Section 6.4.5.2, Page 5) 1
_ ___ * ' ~
a-Attachment 2 to P-87122 Page 3 ,
will remain the same. The allowable stresses during the firewater cooldown event are greater than the 100 percent power condition by about a factor of two, so the firewater cooldown event has much higher safety margins than the 100 percent power conditions.
Reheater Tubes:
The critical failure mechanism of a reheater tube has been identified as creep collapse under external pressure. The seismic loading on the FSV reheater tubes produces very low stress levels (550 psi, Section 6.5.4, Page 193 of GADR-9). A low level vibratory bending load superimposed onto an external pressure creep collapse load of a tube has no significant effect. The SSE bending stresses in the tube are in the axial direction, while the collapse stresses are in the hoop direction. Only if the tube bending load was large would the secondary effects from the additional tube ovality caused by bending of an already curved tube require evaluation.
Reheater Tubes and Supports:
Because the reheater tubes and their supports are basically at the same temperature during the EES firewater cooldown event, the interaction loads between the tubes and support structure due to restrained thermal expansion are negligible. The thermal stresses due to tube wall radial gradient and restrained thermal expansion are negligible for the EES firewater cooldown event. The 100 percent load condition evaluated in GADR-9 considered the SSE event in conjunction with the dead weight and helium drag loadings, (3220 psi, Section 6.5.4, Page 96 of GADR-9). This combination of dead weight, SSE and pressure met the ASME Code primary load requirements at 100 percent power. Because the loads and stresses from pressure and thermal effects during the EES firewater cooldown event are lower than the 100 percent loads and stresses, and the allowable loads and stresses for this one time short term faulted event are higher, the SSE when combined with the firewater EES cooldown event will not fail the reheater tubes.
l l
l l
Attachment 2 to P-87122 '
'Ibble Ib.
SiWar II 'Ibbes imriiwy Stress GAIR-9 Section Stress Dir=+im Owiitim calm 1ation 6.3.4 page No. Hoop Iong.
(See Note 1) psi. psi.
Flow Ind. Vib. Insignificant -
0 0 Deadweight Insignificant 19 0 0 Seismic 0 hoop 2156 long. 20 0 2156 Pressure hoop 3291(-613)/1822 28 1107 -
Pressure long. 389(-613)/1822 28 -
131 Bearhug 5702(175)/136 39 0 Wall thermal 7337 (See Note 2) 42916(135)/210 32 27589 27589 (See Note 3) gradient Allowable Stress carhinations Stress, psi Primary Mernbrane stress (See Note 4) 1107 131 17920 (Pressure only)
Primary Maubrane plus Primary Bending stress 1107 2287 26880 (Pressure and SSE)
Total stress, inc. Secondary Bending and Thermal 28696 37213 78000 (Pressure, SSE, Bearhug and Wall thermal gradient)
Notes:
- 1. This calculation converted full power operation stresses (first number) to firewater cooldown stresses using nultipliers/ dividers obtained frcrn Table 5.1 of GA Document 909204. For convenience a copy of Table 5.1 of GA Doc. 909204 is attached. The full power operation stresses were obtained fran GADR-9 Seton 6.3.4, page nunbers as imilcated.
- 2. Bearhug stress is a secondary stress caused by restrained thermal expansion of the helical tube bundle by the tube support plates.
- 3. Tube wall thermal gradient stress is a secondary stress caused by i tenperature difference through the tube wall frcan the heat flux.
4.a Allowable Stress for faulted conditions:
Primary Mesubrane Allowable Stress is S .
Primary Mernbrane plus Primary Bending Y Allowable Stress is 1.5S Primary Menbrane plus Primary Bending plus all Secondary (Tota 1Y stress)
Allowable Stress is 3Sa*
Material properties of SB-163 Grade 2, (Sanicro 31) were taken fran:
ASME B&W Section III Code Case 1342-1. S = 17920 psi 8 385 0F.
ASME B&W Section III Sunmer 1968 Edition. SY = 26000 psi @ 385 F.
8 (See Note'4b) 1 4.b This is the first edition of the ASME B&W Code where limits are provided for Faulted conditions. l
Table 5.1 SUPERHEATER II LOADINGS I.OADING PLATE TEMP. "F TUBE TEMP. "F TEMP. DIFF. PLT.-TBE. TUBE EXTERNAL INTERNAL TUBE CONDITION INLET OUTLET INLET OUTLET INLET DUTLET WALL DT F PRESSURE PRESSURE P.DIFF.
FULL 1099. 1132. 963. 1097. 136. 35. 210. 690. 2512. 1822.
LOAD
. QUARTER 932. 1032. 875. 1032. 57. O. 87.5 589. 2419. 1830. EDN LOAD TRANSIENT 1110. 1135. 995. 1090. 45.
9A @20s.
- 15. ? 469. 2713. 2244. $E!
g g r+'
TRANSIENT 1110. 1135 997. 1100. 113. 25. ? 469, 2713. 2244.
9A @50s. h TRANSIENT 1085. 117.V . 965. 1105. 120. 15. ? 469. 2713. 2244.
9A @l00s.
TRANSIENT 1015. 1045. 915. 1015. 100. 30. ? 469. 2713. 2244.
9A @300s.
EES F/W 455, 301. 286. 294. 169. 7. 128. 700. 87. -613.
COOLDOWN 75% POWER EES F/W 468. 302. 293. 296. 175. 6. 135. 700. 87. -613.
C00LDOWN 105% POWER
$# -e Information on full and quarter load and transient 9A obtained from Ref. 7, Section 6.3, p 16. 7N gg (Excerpted from GA Document 909204,ft/A) $$
e-Attachmant 2 to P-87122 Page 4 ,
NRC REQUEST 2 In Attachment 7 of P-86683, it is stated on Page 3 that during the 90 minute period the tube metal temperature is less than 1300 degrees F, while calculation No. 68-02 of Attachment 9 indicates that the maximum of all tube temperatures in this transient is 780 degrees F.
Provide a reconciliation and basis of these two values, indicating which is the correct value.
PSC RESPONSE 2 The maximum tube temperature during the first 90 minutes varies from 1000 degrees F at the start to 780 degrees F at time 90 minutes into the interruption of forced circulation event following reactor operation at 105% power. See Figures 7-1 and 7-2 in PSC Response 7.
The structural evaluation of the steam generator during a delayed firewater cooldown was being performed concurrently with the system analysis. Temperatures and pressures for the structural evaluation were conservatively chosen to assure that they would envelope those from the final system analysis. A more appropriate temperature value for the first 90 minute period would be 1000 degrees F. However, the 1300 degrees F used for the first 90 minute period in the reheater structural evaluation is conservative.
l l
1 I
1 l
o.
l Attachment 2 to P-87122 l i
Page 5 ,
NRC REQUEST 3 Provide the detailed methodology used for performing the creep collapse analysis of the reheater tube, and the User's Manual for the computer program " BUCKLE". Provide assurance that the time to reach yield stress in the maximum stressed point is shorter than the time at which the ovality becomes unbounded, i.e., show that collapse can not occur at a maximum stress which is lower than the yield stress.
PSC RESPONSE 3 The GA Buckle computer code solves the equations used in the Pan method (Ref. 4 of GA document 909204, Issue N/C), which is a published and industry accepted method. The buckling time is defined as the time at which an element of the tube reached a stress level equal to the material yield stress at the temperature of the tube.
The code tracks the distorted shape, and when the combination of shape and loading produces a stress equal to the yield stress, the tube is deemed to have failed. This is conservative as further deformation is necessary to cause the plastic hinge to form through the tube wall, which may allow the tube to collapse. Work hardening which wculd increase the time to failure is ignored in this analysis.
Because of the high temperatures assumed for the analysis and the short times involved, revisions to the computer code were required.
The GA Users Manual is being updated to incorporate these revisions, and will be released and submitted to the NRC by April 30, 1987.
For a long tube to collapse under the influence of external pressure loading without the yield stress of the material being exceeded, the tube must be thin walled relative to its mean diameter. This hypothesis can be shown to be valid by using the critical pressure from Equation 11-12 of Reference 1 to determine the hoop stress due to this critical external pressure for various mean diameter to wall thickness ratios. The tube mean diameter to wall thcikness ratio has to be generally much greater than twenty for external pressure buckling of a tube to occur before the material yield stress is reached. Above a ratio of twenty, collapse can occur without any elements of the tube reaching the yield stress of the material. The Fort St. Vrain reheater tubes are thick walled, having a diameter to wall thickness ratio of eight. Therefore the reheater tubes could not collapse with an element stress lower than the material yield ;
stress. l i
Reference for PSC Response 3: l
- 1) Timoshenko, S. P., and Gere, J. M. Theory of Elastic Stability, Second Edition, McGraw-Hill Book Co., Inc.
)
i I 1
i
o-Attachment 2 to P-87122 Page 6 NRC REQUEST 4 Provide a basis for the statement on Page 16 of attachment 7 that "the structural integrity of the steam generator is not likely to be compromised due to excessive tube / plate interaction loads during the EES firewater cooldown event," in particular during simultaneous SSE loading.
PSC RESPONSE 4 The structural integrity of the steam generator is not likely to be compromised due to excessive tube / plate interaction loads during the EES firewater cooldown event because the summation of the tube stresses during the event are less than during the 100 percent power condition. The restrained thermal expansion (Bearhug) loading is classified as a secondary loading, the other secondary category load is the tube wall thermal gradient. Restrained thermal expansion (Bearhug) stresses increase to 7337 psi during the firewater cooldown event from 5702 psi during the 100 percent power condition. The tube wall thermal gradient stresses decrease to 27589 psi during the firewater cooldown event from 42916 psi during the 100 percent power condition. The summation of these secondary category loadings for the firewater cooldown event is therefore less than during the 100 percent power condition. The total longitudinal stress, including primary and secondary stresses which include a 2156 psi SSE load contribution, would be 37213 psi as opposed to a faulted allowable stress limit of 78000 psi. See Table Ib in PSC Response 1 for details of the stress contributions from SSE, deadweight, pressure, flow induced vibration, restrained thermal expansion (Bearhug), and tube wall thermal gradient.
r-j
. Attachment 2 i to P-87122 Page 7 l i
NRC REQUEST 5 It is stated that Sanicro 31 reheater tube material is a European Alloy 800 - type material, which met the required material specifications of Alloy 800 Grade 2 at the time when the steam generators were built. Provide the chemical composition of this material and its strength levels (i.e., yield and ultimate strengths at appropriate temperatures). In addition address the metallurgical treatment of Sanicro 31 in comparison with the treatment of the ASME Code Alloy 800 material. Further, explain why the Sanicro 31 will not degrade to below the minimum levels allowed for the Code material because of the temperature cycles which the steam generator tubes experience.
PSC RESPONSE 5 Clarifying the NRC Request 5: When the Fort St. Vrain steam generators were designed, Ni-Fe-Cr alloys, Grade 1 and Grade 2 (the modern equivalents which are Alloy 800 and Alloy 800H, respectively),
were included in Code Cases 1325-2 and 1342-1. The Fort St. Vrain steam generator tubing was manufactured by Sandvik, to the ASME B&PV Code specification Ni-Fe-Cr alloy, Grade 2 as given in Code Case 1342-1. This material was given the designation Sanicro 31 by Sandvik.
The chemical composition of Sanicro 31 used in the construction of the Fort St. Vrain steam generators is the sama as published in ASME B&PV Code Case 1342 ' (see Figure 5-1). The material properties (i.e., yield and ultima ' strengths at appropriate temperatures) of Sanicro 31 (Ni-Fe-Cr alloy, Grade 2) used in the construction of the Fort St. Vrain steam generators satisfy the requirements of ASME B&PV Code Case 1342-1 (see Figure 5-1). This Sanicro 31 material was annealed to approximately 2100 degrees F, as given in Code Case 1342-1, to meet all the requirements of Ni-Fe-Cr alloy, Grade 2.
The Fort St. Vrain steam generator tubes were designed using the ASME Code Case 1342-1 design stress intensities including consideration for temperature cycling over the design life of the steam generators.
However, in late 1975 it was recognized that potential tube failures may occur in the Fort St. Vrain steam generators during service due to the presence of cold formed (non-heat treated) tight radius bends in the Sanicro 31 tubes. These concerns and programs are discussed in FSAR Appendix A.13.27. The characteristics of concern in the cold formed small radius bends, were recrystallization, low ductility rupture and creep-fatigue damage. Of the three identified concerns, low ductility rupture was shown in Appendix A.13.27 of the FSAR to no longer be a problem. The recrystallization and creep-fatigue damage in the cold formed small radius bends were also subsequently found not to be problems as reported in the Electric Power Research Institute (EPRI) Report EPRI-HTGR 86-03, " Properties of Recrystallized Ni-Fe-Cr alloy,H and Associated HTGR Steam Generator Design Implications," dated July 1986. The tests reported in EPRI-HTGR 86-03 were performed on a FSV archive heat of Sanicro 31 that met all the specification requirements of Ni-Fe-Cr alloy,H.
Therefore, it was concluded that there are no identified
O
Attachment 2-to P-87122 i Page 8 t -
, i metallurgical problems that could degrade the service life because of the temperature cycles which the steam generator tubes experience.
k f
k i
a i
i i
?
4 1
I i
4 i
i 4
)
1 i - < ,. .,,----- -- - .. ..-.---.---- , - - - - . , , - , , . - , . _ . . - , . - - , , , . . - - . , . - - . . . . . - . . ~ . - , - - - - , , - - - , , - , - , . - - , , . -- - - - , , , , . -
ATTuccewr '2-Fi c,u ne 5-1 T N '"
CASE (PAcas I oF ?-)
jgy INTERPRETATIONS OF ASME BOILER AND PRESSURE VESSEL CODE Approved by Council Alarch 14,1966 Case 1342-1 (Special Ruling) tilechanical Propertics:
Requirements for Nickel-Iron-Chromium Grade 1* Grade 2" Alloy (All Product Forms) ,
Section 111 Tensile Strength, psi, minimum , a,000 _0,000 Inquiry: May . nickel-iron-chromium Yield Strength, _
allovs he used in the construction of 0.2 per cent Class A Nuclear Vessels in accord. Offset, psi, and with Section 111, and what re. minimum 30,000 25,000 quirements apply to these materials. Elongation in Reply: It ;is the opinion of the com- cc ;in$u$ 30 30 mittee that nickcl ,ron-chromium i alloys may be used in the construction of
- Anncaled ut opproximately 1800 F Class A Nuclear Vessels and the fol- " Annealed at approximately 2100 F lowing specified special requirements apply in addition to the requirements 2) Fabrication Requirements, for various product forms specified in Welded fabrication shall conform to Section Ill. the applicable requirements of Section
- 1) Materials conforming to Specifica- 111.
Lions 5B-163 and/or ASTM IM07-63T, 11408-63T, U409-63T and conforming to
" '. pr edure and performanec t,i c 7ollowing requirements as t qualifications shall he conducree ns _
chemical compos,i tion and mechanical prescribed in Section IX, Part B, except that the tensile strength of the reduced P' P " S section specimen shall be not less Chemical Composition: th an th e minimum tensile strensth Per Cent specified for the base material.
Nickel 30.0 -35.0 Chromium 19.0 -23.0 b)E.id.ing c shall be done by any Ir u il lance welding process or combination of Carbon processes capable of meeting the 0.10 max. requirements.
ManFanese 1.50 max.
Sulphur 0.015 max. c) Welds that are exposed to the Silicon 1.00 max. corrosive action of the contents of the Copper 0.75 max. vessel should have a resistance to cor-Aluminum 0.15- 0.60 rosion that is not substantially -less Titanium 0.15- 0.60 than that of the base metal. The use of Meeting of January 14, 1966 Reprinted from MECHANICAL ENGINEERING May,1966, a pub!! cation of the American Society of Mechanical Engineers.
Pages 319 thru 320 replace pages 205 and 206.
319
Arrneumeur '2.
CASE (continuedl/ % NI" A Gorz.e 6-1 1342-1 (pac # rz. or. t)
INTERPRETATIONS OF ASME BOILER AND PRESSURE VESSEL CODE filler metal that will deposit weld Table 1.1 metal with practically the same com- -
position as the material joined is Sir in Limiting Factors recommended. When the manufacturer is 1 of the opinion that a physically-better Strain, Per Cent Factors joint can be made by departure from these limits, filler metal of a different 0.10 0.90 composition may be used provided the 0.09 0 .11 9 strength of the weld metal at the 0.08 0.!!!!
' operatinF temperature is not appreciahly 0.07 0.!!6 less than that of the high-alloy material 0.06 0.!!3
~ ~
to be welded, and the user is satisfied 0.05 . 0.!!0 g that its resistan ce to corrosion is 0.04 0.77 satisfactory for the intended service.
0.03 0.73 i
d) A separate welding procedure 0.02 0.60
- and performance qualification shall be 0.01 0.63
- made for this innterial.
i e) llent treatment after forming or j fabrication is neither required nor
- prohibiicd.
Table 2 i 3) llesi::n stress intensiiv values are Yield Strength values I
listed in Table 1.
- 4) Yield strength values are listed in 3 Table 2. Grade 1 Grade 2 Table 1 Min. Spec. T.S. 30000 250_00_, ._
Design Stress Intensity, psi' Tem. F Temperoiure not exceeding F Grade 1 Grade 2 100 30000 2i>000 j gg 3 600 21500 100 20,000 16,700 200 20,000 16,700 300 2r>000 19200 300 20,000 16,700 400 25000 17600 r,00 24100 16500 400 20,000 II',000 l 500 20,000 14,900 (>00 23000 1(>200 600 20,000 14,600 799 3.g ,gg g 7,ggg 650 20,000 14,400 800 23000 Irir,00 700 20,000 14,300 750 20,000 14,200 000 20,000 14,000
- The design st re ri a intensity v.il u r is
! controlled by the IcNect value of $ of the !
j minimum specified yield strength ut ruum I
t e mp era ture, or 0.9 of the 0.27, cffset 1 yield strength ut temperuture.
l 320
~ '
ry
. Attachment 2 to P-87122 Page 9 NRC REQUEST 6 It is implied that Sanicro 31 is equivalent to an Ni-Fe-Cr alloy, H material, e.g., SB-163 or SB-407. If this is the case, justify the value of 23,300 psi at 380 degrees F for Sm. Current ASME Code values indicate Sm values of 16,700 psi for 58-163 and 20,000 psi for SB-407 at 380 degrees F.
PSC RESPONSE 6 The values of Sm on Page 16 of the GA Technologies, Inc. Report 909204, issue N/C were not used in the analysis, and were incorrectly added to the report. The lines that reference Sm should be removed.
The correct Sm value for Sanicro 31 (SB-163 Grade 2) at 380 degrees F is 16,000 psi. This Sm value was obtained by interpolation from the ASME B&PV Code Case 1342-1 (see Figure 5-1 in PSC Response 5). The Sm value was not required nor used in the current analysis, s
]
J
..e, - . -- m-m --~e--a - - - - < - - > . - - - - - - - - - - , - - - - - , - -----.n- ,. - - - - - - -
a-Attachment 2
. to P-87122 Page 10 NRC REQUEST 7 Figure 4.4 is incomplete as submitted. The temperature profile for the time in hours from 0 to 2 is missing. Please supply this information.
PSC RESPONSE 7 The temperatures in the steam generator during the period of no forced circulation (1-1/2 hours) are equal to or less than the normal operating temperatures and thus were not included in Figure 4.4 of GA Technologies, Inc. Report 909204, Issue N/C. As stated in PSC Response 2, temperature / pressure profiles were conservatively established to assure they enveloped the final system analysis. The final system analysis has circulators started at 1-1/2 hours into the transient and not at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
A study was performed by GA Technologies in early 1986 to define bulk temperatures in the steam generator during an interruption of forced
, cooling (10FC) following reactor operation at 105% power for periods as long as five hours. The study considered the steam generator to be either dry (empty) or full with retention of the water and steam inventory. The analysis was performed using the RATSAM computer code (Reference 1), and the analysis was documented in Reference 2.
Figures 7-1 and 7-2 (Figures 2 and 3 from Reference 2) show bul k steam generator temperatures varying with time. At time two hours the bulk steam generator temperatures range from about 620 degrees F to 750 degrees F which are well within the normal operating range.
Retention of the water and steam inventory, Figure 7-2, has only a small effect on bulk steam generator temperatures.
References for PSC Response 7:
- 1) R. K. Deremer and T. Shih, "RATSAM: A Computer Program to I Analyze the Transient Behavior of the HTGR Primary Coolant System During Accidents," GA Technologies, Inc. Report GA-A13705, May )
1977.
- 2) G. J. Cadwallader and A. Bagierek, "FSV Steam Generator Temperatures During 10FC From 105% Power," GA Technologies, Inc.
GA Document 908784, Issue N/C, March 19, 1986.
- - - , ' - - =r -
FIGU R E 2 FSV STEAM GENERATOR TEMPERATURES - CASE 1 (WET) .
1000-P 7 CURVE IDENT.
n _
l k _
n
_ v e 900 -
REHEATER tti _
Q A v _ ~
i _
SUPERHEATER II
~
E m 3 800 "
3Q - 3 SUPERHEATER I E
An C* "
- EVAPORATOR (D L-A 'A 1 O M "
_' g Ld F
700 < -
A A
7.d -
C
.__ y :
i NOO o y
~
ECONOMIZER f
d 600 1 1
=
1 a
NM T i
y
_1 I STM GEN OUTLET b_ _
lti -ID y -
o_ ,
I 500 i i i i iiii i i i i i i i i i i i i g@
0 1 2 3 4 5
$ sh-i s
e TIME CHOURS)
IDFC 9 105% POWER "s
s s
FIGURE 3
FSV STEAM GENERATOR TEMPERATURES - CASE 2 CDRY) -
1000-O I CURVE IDENT.
n -
it- -
- O to 900 REHEATER sv
~
A
< i SUPERHEATER II K
11 g 800 c p <
\ - <
SUPERHEATER I c oc w -
O o_ _ -
r 4 --- --
n a e=r? S Ov EVAPORATOR u T p
a ne w 700 <
~
r r* -
y z
x U v 0F V ECON 0MIZER x
600 [^" "^7 g~
.t.
y w STM GEN OUTLET it. -
M - o?
Ps 500 i i i i iiii i i i i i i i i i i i 1 3G
-3 ape j
e 0 1 2 3 4 5 j TIME CHOURS)
P IDFC 9105% POWER i
o-
. Attachment 2 to P-87122 Page 11 NRC REQUEST 8 The possibility of accumulated creep-fatigue damage contributing to tube failure is not explicitly addressed in this submittal. Please provide an evaluation of this potential failure mechanism in the context of the creep-collapse failure mode that was assumed.
PSC RESPONSE 8 Prior creep-fatigue damage in a steam generator tube can affect the
- tube's creep-collapse failure mode if the prior accumulated creep-fatigue damage is high. The Fort St. Vrain Steam Generators were designed and licensed to the ASME B&PV Code Section III (through the Winter 1966 Addenda) and eleven Code Cases including Code Case 1342-
, 1. These ASME code rules did not require creep-fatigue damage evaluation. However, the fatigue analysis performed in the steam generator design report, GADR-9, showed low fatigue damage even in those regions of the steam generator where there were significant cyclic loads. Because the heat flux loads in the reheater are low, there is no reason to expect that there will be significant creep-4 fatigue damage in the reheater tubes particularly adjacent to the small radius bends where the maximum ovality is expected to occur.
(It is at these maximum ovality points that creep-collapse would firstoccur.)
Creep-rupture damage (a component of creep-fatigue damage) induced during the transient event analyzed is also small because of the low stress levels involved. The following table gives the creep-rupture damage increments calculated for both the previously defined 105%
power case and the recommended 87.5% power case.
Power Time to Actual Creep-Rupture Level Temp. Press. Stress Rupture Event Time Damage Fraction
% deg. F psi ksi hr hr 105 1660 350 3.5 879 2 0.0022 87.5 1500 350 3.5 34690 2 0.0001 Creep-fatigue damage (obtained by summing creep-rupture damage and fatigue damage) has to reach a value of 1.0 before failure is predicted. Since the creep-rupture damage fraction sustained during the transient event is small and prior accumulated creep-fatigue damage has not been shown to be large, this creep-fatigue phenomenon does not have a significant influence on the predicted times for i
creep-collapse of the tubes.
o.
-Attachment 2 to P-87122 Page 12 NRC REQUEST 9 With regard to the Fort St. Vrain (FSV) Technical Specifications, 5.3.11 - Steam Generator Bimetallic Welds Surveillance and SR 5.3.12
- Steam Generator Tube Leaks Surveillance, describe the results of the surveillance requirements including the number of times and the dates that they have been implemented. Provide any other information that is available concerning degradation of the steam generator tubes.
PSC RESPONSE 9 FSV surveillance test SR 5.3.11, which requires the volumetric examination of steam generator bimetallic welds for indications of subsurface defects at five calendar year intervals, was approved as part of Amendment No. 39 to the FSV Operating License.
Implementation of this testing at FSV is per Inservice Inspection (ISI) Criterion C, which states that this surveillance requirement shall be implemented before the beginning of fuel cycle 5. Since the FSV plant is still in the early life of fuel cycle 4, no examinations per SR 5.3.11 have been performed to date.
FSV surveillance test SR 5.3.12, which requires an extensive evaluation of each steam generator tube leak, was approved as part of Amendment No. 45 to the FSV Operating License. There are no new tube i
examination results available at this time because there have been no new tube leaks since the last (second) leak occurred in the Superheater II tube bundle of module B-2-3. The results of that tube leak metallographic evaluation were reported (GA Report GA-C17042) to the NRC in PSC letter, Warembourg to Collins, dated September 16, 1983 (P-83311). This report found no significant or unusual degradation of the Sanicro 31 subheader tubing that was examined by GA Technologies, Inc.
l Similar results were reported for the first tube leak that occurred
, in the Superheater II tube bundle of module B-1-1 in December of l 1977. This leak and results of GA Technologies, Inc. metallographic l examination were reported in GA Report GA-D15395 (December 1977).
1 4
l l