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Category:CORRESPONDENCE-LETTERS
MONTHYEARML20217P3791999-10-21021 October 1999 Forwards NRC Form 396 & NRC Form 398 for Renewal of Licenses SOP-20607-1 & SOP-20610-1.Without Encls ML20217N2521999-10-20020 October 1999 Provides Supplemental Info Re 990405 Containment Insp Program Requests for Relief RR-L-1 & RR-L-2,in Response to 991013 Telcon with NRC ML20217K7541999-10-15015 October 1999 Forwards Rev 1 to Unit 1,Cycle 9 & Unit 2 Cycle 7 Colrs,Iaw Requirements of TS 5.6.5.Figure 5, Axial Flux Difference Limits as Function of Percent of Rated Thermal Power for RAOC, Was Revised for Both Units ML20217G6751999-10-13013 October 1999 Requests Withholding of Proprietary Info Contained in Application for Amend to OLs to Implement Relaxations Allowed by WCAP-14333-P-A,rev 1 ML20217G1071999-10-0707 October 1999 Informs That on 990930,NRC Staff Completed mid-cycle PPR of Vogtle & Did Not Identify Any Areas in Which Performance Warranted More than Core Insp Program.Nrc Plans to Conduct Core Insps at Facility Over Next Six Months ML20216J9041999-10-0101 October 1999 Forwards Response to RAI Re GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves ML20216J9161999-10-0101 October 1999 Forwards Response to NRC 990723 RAI Re GL 95-07, Pressure Locking & Thermal Binding of SR Power-Operated Gate Valves ML20217B0141999-10-0101 October 1999 Forwards Insp Repts 50-424/99-06 & 50-425/99-06 on 990725- 0904 at Vogtle Units 1 & 2 Reactor Facilities.Determined That One Violation Occurred & Being Treated as non-cited Violation ML20212E8751999-09-20020 September 1999 Forwards Response to NRC GL 99-02, Lab Testing of Nuclear Grade Activated Charcoal. Description of Methods Used to Comply with Std Along with Most Recent Test Results Encl ML20212E7481999-09-20020 September 1999 Requests Approval Per 10CFR50.55a to Use Alternative Method for Determining Qualified Life of Certain BOP Diaphragm Valves than That Specified in Code Case N-31.Proposed Alternative,Encl ML20212C2191999-09-16016 September 1999 Forwards NRC Form 536, Operator Licensing Exam Data, Which Is Current Need for NRC Operator Licensing Exams for Years 2000 Through 2003 of Plant Vogtle,Per Administrative Ltr 99-03 ML20211Q4801999-09-0101 September 1999 Informs That on 990812-13,Region II Hosted Training Managers Conference on Recent Changes to Operator Licensing Program. List of Attendees,Copy of Slide Presentations & List of Questions Received from Participants Encl ML20211J5291999-08-30030 August 1999 Forwards Snoc Copyright Notice Dtd 990825,re Production of Engineering Drawings Ref in VEGP UFSAR ML20211J5251999-08-30030 August 1999 Forwards Response to NRC 990727 RAI Re GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design Basis Accident Conditions ML20211J7381999-08-27027 August 1999 Informs That Licensee Vessel Data Is Different than NRC Database Based on Listed Info,Per 990722 Request to Review Rvid ML20211E9251999-08-23023 August 1999 Forwards fitness-for-duty Performance Data for Jan-June 1999,as Required by 10CFR26.71(d).Data Reflected in Rept Covers Employees at Vogtle Electric Generating Plant ML20210V0881999-08-16016 August 1999 Forwards Insp Repts 50-424/99-05 & 50-425/99-05 on 990620- 0724.No Violations Noted.Vogtle Facility Generally Characterized by safety-conscious Operations,Sound Engineering & Maintenance Practices ML20210Q4611999-08-0505 August 1999 Informs That NRC Plans to Administer Generic Fundamentals Exam Section of Written Operator Licensing Exam on 991006 for Vogtle.Requests Info Re Individuals Who Will Take Exam. Sample Registration Ltr Encl ML20210L2181999-08-0202 August 1999 Forwards NRC Form 396 & Form 398 for Renewal of Listed Licenses,Iaw 10CFR55.57.Without Encl ML20210N1191999-08-0202 August 1999 Discusses 990727 Telcon Between Rs Baldwin & R Brown Re Administration of Licensing Exam at Facility During Wk of 991213 ML20210G3351999-07-27027 July 1999 Forwards Second Request for Addl Info Re GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design Basis Accident Conditions ML20210E0121999-07-23023 July 1999 Forwards Second Request for Addl Info Re GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves ML20210D9341999-07-22022 July 1999 Discusses Closure of TACs MA0581 & MA0582,response to Requests for Info in GL 92-01,rev 1,suppl 1, Reactor Vessel Structural Integrity ML20210C8011999-07-21021 July 1999 Provides Response to NRC AL 99-02,which Requests That Addressees Submit Info Pertaining to Estimates of Number of Licensing Actions That Will Be Submitted for NRC Review for Upcoming Fy 2000 & 2001 ML20210E0431999-07-15015 July 1999 Forwards Insp Repts 50-424/99-04 & 50-425/99-04 on 990502- 0619.Two Violations of NRC Requirements Identified & Being Treated as non-cited Violations Consistent with App C of Enforcement Policy ML20209H3881999-07-14014 July 1999 Forwards Revs 1 & 2 to ISI Program Second 10-Year Interval Vogtle Electric Generating Plant Unit 1 & 2 ML20209C4041999-07-0101 July 1999 Forwards Rev 29 to VEGP Units 1 & 2 Emergency Plan.Rev 29 Incorporates Design Change Associated with Consolidation of Er Facilities Computer & Protues Computer.Justifications for Changes & Insertion Instructions Are Encl ML20196H8081999-06-28028 June 1999 Discusses 990528 Meeting Re Results of Periodic PPR for Period of Feb 1997 to Jan 1999.List of Attendees Encl ML20212J2521999-06-21021 June 1999 Responds to NRC RAI Re Yr 2000 Readiness at Nuclear Power Plants.Gl 98-01 Requested Response on Status of Facility Y2K Readiness by 990701 ML20196F9171999-06-21021 June 1999 Forwards Owner Rept for ISI for Vogtle Electric Generating Plant,Unit 1 Eighth Maint/Refueling Outage. Separate Submittal Will Not Be Made to NRC on SG Tubes Inspected During Subj Outage ML20195F8031999-06-11011 June 1999 Forwards Changes to VEGP Unit 1 Emergency Response Data Sys (ERDS) Data Point Library.Changes Were Completed on 990308 While Unit 1 Was SD for Refueling Outage ML20207E7421999-06-0303 June 1999 Refers to from NRC Which Issued Personnel Assignment Ltr to Inform of Lm Padovan Assignment as Project Manager for Farley Npp.Reissues Ltr with Effective Date Corrected to 990525 ML20207F6201999-06-0202 June 1999 Sixth Partial Response to FOIA Request for Documents.Records in App J Encl & Will Be Available in Pdr.App K Records Withheld in Part (Ref FOIA Exemptions 7) & App L Records Completely Withheld (Ref FOIA Exemption 7) ML20207D9861999-05-28028 May 1999 Informs That,Effective 990325,LM Padovan Was Assigned as Project Manager for Plant,Units 1 & 2 ML20207D2701999-05-19019 May 1999 Forwards Insp Repts 50-424/99-03 & 50-425/99-03 on 990321- 0501.One Violation of NRC Requirements Identified & Being Treated as non-cited Violation Consistent with App C of Enforcement Policy ML20206M5141999-05-11011 May 1999 Informs That NRC Ofc of Nuclear Regulation Reorganized Effective 990328.As Part of Reorganization,Div of Licensing Project Mgt Created.Rl Emch Section Chief for Vogtle. Reorganization Chart Encl ML20206U4061999-05-11011 May 1999 Confirms Telcon with J Bailey Re Mgt Meeting Scheduled for 990528 to Discuss Results of Periodic Plant Performance Review for Plan Nuclear Facility Fo Period of Feb 1997 - Jan 1999 05000424/LER-1998-006, Forwards LER 98-006-03 Re Motor Control Ctr Breaker Buckets Not Being Seismically Qualified.Rev Is Submitted to Document Results of Seismic Testing That Demonstrated That No Condition Outside Design Basis of TS Requirements Exi1999-05-10010 May 1999 Forwards LER 98-006-03 Re Motor Control Ctr Breaker Buckets Not Being Seismically Qualified.Rev Is Submitted to Document Results of Seismic Testing That Demonstrated That No Condition Outside Design Basis of TS Requirements Existed ML20206D6411999-04-29029 April 1999 Forwards Vogtle Electric Generating Plant Radiological Environ Operating Rept for 1998 & Vogtle Electric Generating Plant Units 1 & 2 1998 Annual Rept Annual Radioactive Effluent Release Rept ML20206D5881999-04-29029 April 1999 Forwards Rept Which Summarizes Effects of Changes & Errors in ECCS Evaluation Models on PCT for 1998,per Requirements of 10CFR50.46(a)(3)(ii).Rept Results Will Be Incorporated Into Next FSAR Update ML20206D6951999-04-28028 April 1999 Provides Update of Plans for VEGP MOV Periodic Verification Program Re GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Movs ML20206C2241999-04-21021 April 1999 Forwards Revised Monthly Operating Repts for Mar 1999 for Vogtle Electric Generating Plant,Units 1 & 2.Page E2-2 Was Iandvertently Omitted from Previously Submitted Rept on 990413 ML20206A6371999-04-21021 April 1999 Forwards SE Authorizing Licensee Re Rev 9 to First 10-yr ISI Interval Program Plan & Associated Requests for Relief (RR) 65 from ASME Boiler & Pressure Vessel Code ML20205Q3351999-04-15015 April 1999 Forwards Insp Repts 50-424/99-02 & 50-425/99-02 on 990214-0320.Three Violations Identified & Being Treated as Non-Cited Violations ML20205T2351999-04-0909 April 1999 Informs That on 990317,B Brown & Ho Christensen Confirmed Initial Operator Licensing Exam Scheduled for Y2K.Initial Exam Date Scheduled for Wk of 991213 for Approx 10 Candidates ML20205K7501999-04-0505 April 1999 Informs That Effective 990329,NRC Project Mgt Responsibility for Plant Has Been Transferred from Dh Jaffe to R Assa ML20209A3741999-04-0505 April 1999 Submits Several Requests for Relief for Plant from Code Requirements Pursuant to 10CFR50.55a(a)(3) & (g)(5)(iii).NRC Is Respectfully Requested to Approve Requests Prior to Jan 1,2000 ML20205H3481999-03-31031 March 1999 Forwards Georgia Power Co,Oglethorpe Power Corp,Municipal Electric Authority of Ga & City of Dalton,Ga Status of Decommissioning Funding for Each Reactor or Part of Reactor Owned for OLs NPF-68 & NPF-81 ML20205F9091999-03-29029 March 1999 Submits Rept of Number of SG Tubes Plugged During Plant Eighth Maintenance/Refueling Outage (1R8).Inservice Insps Were Completed on SGs 1 & 4 on 990315.No Tubes Were Plugged ML20205G0761999-03-26026 March 1999 Provides Results of Individual Monitoring for 1998.Encl Media Contains All Info Required by Form NRC 5.Without Encl 1999-09-20
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARML20217P3791999-10-21021 October 1999 Forwards NRC Form 396 & NRC Form 398 for Renewal of Licenses SOP-20607-1 & SOP-20610-1.Without Encls ML20217N2521999-10-20020 October 1999 Provides Supplemental Info Re 990405 Containment Insp Program Requests for Relief RR-L-1 & RR-L-2,in Response to 991013 Telcon with NRC ML20217K7541999-10-15015 October 1999 Forwards Rev 1 to Unit 1,Cycle 9 & Unit 2 Cycle 7 Colrs,Iaw Requirements of TS 5.6.5.Figure 5, Axial Flux Difference Limits as Function of Percent of Rated Thermal Power for RAOC, Was Revised for Both Units ML20217G6751999-10-13013 October 1999 Requests Withholding of Proprietary Info Contained in Application for Amend to OLs to Implement Relaxations Allowed by WCAP-14333-P-A,rev 1 ML20216J9161999-10-0101 October 1999 Forwards Response to NRC 990723 RAI Re GL 95-07, Pressure Locking & Thermal Binding of SR Power-Operated Gate Valves ML20216J9041999-10-0101 October 1999 Forwards Response to RAI Re GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves ML20212E7481999-09-20020 September 1999 Requests Approval Per 10CFR50.55a to Use Alternative Method for Determining Qualified Life of Certain BOP Diaphragm Valves than That Specified in Code Case N-31.Proposed Alternative,Encl ML20212E8751999-09-20020 September 1999 Forwards Response to NRC GL 99-02, Lab Testing of Nuclear Grade Activated Charcoal. Description of Methods Used to Comply with Std Along with Most Recent Test Results Encl ML20212C2191999-09-16016 September 1999 Forwards NRC Form 536, Operator Licensing Exam Data, Which Is Current Need for NRC Operator Licensing Exams for Years 2000 Through 2003 of Plant Vogtle,Per Administrative Ltr 99-03 ML20211J5291999-08-30030 August 1999 Forwards Snoc Copyright Notice Dtd 990825,re Production of Engineering Drawings Ref in VEGP UFSAR ML20211J5251999-08-30030 August 1999 Forwards Response to NRC 990727 RAI Re GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design Basis Accident Conditions ML20211J7381999-08-27027 August 1999 Informs That Licensee Vessel Data Is Different than NRC Database Based on Listed Info,Per 990722 Request to Review Rvid ML20211E9251999-08-23023 August 1999 Forwards fitness-for-duty Performance Data for Jan-June 1999,as Required by 10CFR26.71(d).Data Reflected in Rept Covers Employees at Vogtle Electric Generating Plant ML20210L2181999-08-0202 August 1999 Forwards NRC Form 396 & Form 398 for Renewal of Listed Licenses,Iaw 10CFR55.57.Without Encl ML20210C8011999-07-21021 July 1999 Provides Response to NRC AL 99-02,which Requests That Addressees Submit Info Pertaining to Estimates of Number of Licensing Actions That Will Be Submitted for NRC Review for Upcoming Fy 2000 & 2001 ML20209H3881999-07-14014 July 1999 Forwards Revs 1 & 2 to ISI Program Second 10-Year Interval Vogtle Electric Generating Plant Unit 1 & 2 ML20209C4041999-07-0101 July 1999 Forwards Rev 29 to VEGP Units 1 & 2 Emergency Plan.Rev 29 Incorporates Design Change Associated with Consolidation of Er Facilities Computer & Protues Computer.Justifications for Changes & Insertion Instructions Are Encl ML20196F9171999-06-21021 June 1999 Forwards Owner Rept for ISI for Vogtle Electric Generating Plant,Unit 1 Eighth Maint/Refueling Outage. Separate Submittal Will Not Be Made to NRC on SG Tubes Inspected During Subj Outage ML20212J2521999-06-21021 June 1999 Responds to NRC RAI Re Yr 2000 Readiness at Nuclear Power Plants.Gl 98-01 Requested Response on Status of Facility Y2K Readiness by 990701 ML20195F8031999-06-11011 June 1999 Forwards Changes to VEGP Unit 1 Emergency Response Data Sys (ERDS) Data Point Library.Changes Were Completed on 990308 While Unit 1 Was SD for Refueling Outage 05000424/LER-1998-006, Forwards LER 98-006-03 Re Motor Control Ctr Breaker Buckets Not Being Seismically Qualified.Rev Is Submitted to Document Results of Seismic Testing That Demonstrated That No Condition Outside Design Basis of TS Requirements Exi1999-05-10010 May 1999 Forwards LER 98-006-03 Re Motor Control Ctr Breaker Buckets Not Being Seismically Qualified.Rev Is Submitted to Document Results of Seismic Testing That Demonstrated That No Condition Outside Design Basis of TS Requirements Existed ML20206D5881999-04-29029 April 1999 Forwards Rept Which Summarizes Effects of Changes & Errors in ECCS Evaluation Models on PCT for 1998,per Requirements of 10CFR50.46(a)(3)(ii).Rept Results Will Be Incorporated Into Next FSAR Update ML20206D6411999-04-29029 April 1999 Forwards Vogtle Electric Generating Plant Radiological Environ Operating Rept for 1998 & Vogtle Electric Generating Plant Units 1 & 2 1998 Annual Rept Annual Radioactive Effluent Release Rept ML20206D6951999-04-28028 April 1999 Provides Update of Plans for VEGP MOV Periodic Verification Program Re GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Movs ML20206C2241999-04-21021 April 1999 Forwards Revised Monthly Operating Repts for Mar 1999 for Vogtle Electric Generating Plant,Units 1 & 2.Page E2-2 Was Iandvertently Omitted from Previously Submitted Rept on 990413 ML20209A3741999-04-0505 April 1999 Submits Several Requests for Relief for Plant from Code Requirements Pursuant to 10CFR50.55a(a)(3) & (g)(5)(iii).NRC Is Respectfully Requested to Approve Requests Prior to Jan 1,2000 ML20205H3481999-03-31031 March 1999 Forwards Georgia Power Co,Oglethorpe Power Corp,Municipal Electric Authority of Ga & City of Dalton,Ga Status of Decommissioning Funding for Each Reactor or Part of Reactor Owned for OLs NPF-68 & NPF-81 ML20205F9091999-03-29029 March 1999 Submits Rept of Number of SG Tubes Plugged During Plant Eighth Maintenance/Refueling Outage (1R8).Inservice Insps Were Completed on SGs 1 & 4 on 990315.No Tubes Were Plugged ML20205G0761999-03-26026 March 1999 Provides Results of Individual Monitoring for 1998.Encl Media Contains All Info Required by Form NRC 5.Without Encl ML20205H4051999-03-25025 March 1999 Forwards Info on Status of Decommissioning Funding for Each Reactor or Part of Reactor Owned for OLs NPF-68 & NPF-81,as Requested IAW 10CFR50.75(f)(1) ML20205H3891999-03-25025 March 1999 Forwards Info on Status of Decommissioning Funding for Each Reactor or Part of Reactor Owned for OLs NPF-68 & NPF-81,as Requested IAW 10CFR50.75(f)(1).Page 2 in Third Amend Power Sales Contract of Incoming Submittal Not Included ML20205A9441999-03-25025 March 1999 Forwards VEGP Unit 1 Cycle 9 Colr,Per TS 5.6.5.d ML20205H3811999-03-24024 March 1999 Forwards Info on Status of Decommissioning Funding for Each Reactor or Part of Reactor Owned for OLs NPF-68 & NPF-81,as Requested IAW 10CFR50.75(f)(1) ML20205H3621999-03-22022 March 1999 Forwards Info on Status of Decommissioning Funding for Each Reactor or Part of Reactor Owned for OLs NPF-68 & NPF-81, as Requested IAW 10CFR50.75(f)(1) ML20204G4361999-03-18018 March 1999 Forwards Summary Rept of Present Level & Source of on-site Property Damage Insurance Coverage for Vegp,Iaw Requirements of 10CFR50.54(w)(3) ML20204C0591999-03-17017 March 1999 Forwards Rev 0 to WCAP-15160, Evaluation of Pressurized Thermal Shock for Vegp,Unit 2 & Rev 0 to WCAP-15159, Analysis of Capsule X from Vegp,Unit 2 Reactor Vessel Radiation Surveillance Program ML20207K9551999-03-11011 March 1999 Forwards Response to Rai,Pertaining to Positive Alcohol Test of Licensed Operator.Encl Info Provided for NRC Use in Evaluation of Fitness for Duty Occurrence.Encl Withheld,Per 10CFR2.790(a)(6) ML20207L9721999-03-10010 March 1999 Forwards Rev 15 to EPIP 91104-C of Manual Set 6 of Vogtle Epips.Without Encl ML20207B0191999-02-25025 February 1999 Forwards Fitness for Duty Performance Data for six-month Reporting Period 980701-1231,IAW 10CFR26.71(d) 05000424/LER-1998-009, Forwards LER 98-009-00 Re Event in Which Improper Testing Method Resulted in Inadequate Surveillances on 9812291999-01-27027 January 1999 Forwards LER 98-009-00 Re Event in Which Improper Testing Method Resulted in Inadequate Surveillances on 981229 ML20199F7701999-01-13013 January 1999 Submits Revised Response to RAI Re Licensee 980713 Proposed Amend to Ts,Eliminating Periodic Response Time Testing Requirements on Selected Sensors & Protection Channels. Corrected Copy of Table,Encl ML20199F7981999-01-13013 January 1999 Forwards Corrected Pages to VEGP-2 ISI Summary Rept for Spring 1998 Maint/Refueling Outage. Change Bar in Margin of Affected Pages Denotes Changes to Rept ML20199G1381999-01-13013 January 1999 Forwards Copy of Permit Renewal Application Package for NPDES Permit Number GA0026786,per Section 3.2 of VP Environ Protection Plan 05000424/LER-1998-007, Forwards LER 98-007-00,re Inadequate Surveillances Due to Improperly Performed Response Time Testing,On 981215,IAW 10CFR50.731999-01-13013 January 1999 Forwards LER 98-007-00,re Inadequate Surveillances Due to Improperly Performed Response Time Testing,On 981215,IAW 10CFR50.73 ML20198F6131998-12-18018 December 1998 Forwards Revised Certification of Medical Exam Form for License SOP-21147.Licensee Being Treated for Hypertension. Util Requests That Individual License Be Amended to Reflect Change in Status ML20198L6631998-12-18018 December 1998 Forwards Amend 37 to Physical Security & Contingency Plan. Encl 1 Provides Description & Justification for Changes & Encl 2 Contains Actual Amend 37 Pages.Amend Withheld,Per 10CFR73.21 ML20198D9291998-12-16016 December 1998 Forwards Requested Info Re Request to Revise TSs Elimination of Periodic Pressure Sensor Response Time Tests & Elimination of Periodic Protection Channel Response Time Tests ML20198D9991998-12-16016 December 1998 Forwards Responses to 980916 RAI Re Response to GL 97-01, Degradation of Control Rod Drive Mechanism Nozzle & Other Vessel Closure Head Penetrations ML20198D8171998-12-14014 December 1998 Forwards NRC Form 396 & Form 398 for Renewal of License OP-20993.Without Encls ML20206N3051998-12-0808 December 1998 Submits RAI Re Replacement of Nuclear Instrument Sys Source & Intermediate Range Channels & post-accident Neutron Flux Monitoring Sys 1999-09-20
[Table view] Category:UTILITY TO NRC
MONTHYEARELV-02056, Forwards Operator Exam Schedule for Facility,Per Generic Ltr 90-07 Request,Including Number of Candidates to Be Examined During NRC Site Visits,Requalification Schedules & Number of Candidates to Participate in Generic Fundamentals Exam1990-09-0606 September 1990 Forwards Operator Exam Schedule for Facility,Per Generic Ltr 90-07 Request,Including Number of Candidates to Be Examined During NRC Site Visits,Requalification Schedules & Number of Candidates to Participate in Generic Fundamentals Exam ELV-01599, Discusses Mods to HED-1114 Re Plant Dcrdr,Per . Amber Monitor Light Covers Installed for Spare Pumps to Make Status of Pumps Readily Apparent to Operator1990-09-0404 September 1990 Discusses Mods to HED-1114 Re Plant Dcrdr,Per . Amber Monitor Light Covers Installed for Spare Pumps to Make Status of Pumps Readily Apparent to Operator ELV-02059, Clarifies 900409 Response to 900323 Confirmation of Action Ltr.Util Made 31 Successful Start Attempts for Diesel Generator (DG) 1A & 29 Successful Start Attempts for DG 1B1990-08-30030 August 1990 Clarifies 900409 Response to 900323 Confirmation of Action Ltr.Util Made 31 Successful Start Attempts for Diesel Generator (DG) 1A & 29 Successful Start Attempts for DG 1B ELV-01956, Forwards Listed Documents in Response to Request for Addl Info Re Settlement Monitoring Program,Per 900614 Request1990-08-30030 August 1990 Forwards Listed Documents in Response to Request for Addl Info Re Settlement Monitoring Program,Per 900614 Request ELV-02050, Responds to Violations Noted in Insp Repts 50-424/90-08 & 50-425/90-08.Corrective Actions:Administrative Procedures Controlling Verification & Validation of Emergency Operating Procedures Will Be Evaluated & Revised as Required1990-08-30030 August 1990 Responds to Violations Noted in Insp Repts 50-424/90-08 & 50-425/90-08.Corrective Actions:Administrative Procedures Controlling Verification & Validation of Emergency Operating Procedures Will Be Evaluated & Revised as Required ELV-02028, Forwards Fitness for Duty Performance Data for First Six Month Period,Per 10CFR26.71(d)1990-08-22022 August 1990 Forwards Fitness for Duty Performance Data for First Six Month Period,Per 10CFR26.71(d) ELV-02022, Forwards Revised LER Re Apparent Personnel Error Leading to Unsecured Safeguards Info.Ler Withheld1990-08-22022 August 1990 Forwards Revised LER Re Apparent Personnel Error Leading to Unsecured Safeguards Info.Ler Withheld ELV-02027, Forwards Rev 0 to Core Operating Limits Rept, for Cycle 3, Per Amends 32 & 12 to Licenses NPF-68 & NPF-79,respectively1990-08-20020 August 1990 Forwards Rev 0 to Core Operating Limits Rept, for Cycle 3, Per Amends 32 & 12 to Licenses NPF-68 & NPF-79,respectively ELV-01973, Submits Rept Re Results of Leakage Exams Conducted During Spring 1990 Refueling Outage,Per TMI Item III.D.1.1.None of Identified Leakage Considered Excessive.Work Orders Issued in Effort to Reduce Leakage to Level as Low Practical1990-08-14014 August 1990 Submits Rept Re Results of Leakage Exams Conducted During Spring 1990 Refueling Outage,Per TMI Item III.D.1.1.None of Identified Leakage Considered Excessive.Work Orders Issued in Effort to Reduce Leakage to Level as Low Practical ELV-01918, Responds to NRC 900612 Request for Comments & Suggestions on Draft risk-based Insp Guide.Util Conducting Individual Plant Exam & Will Withhold Comment on risk-based Insp Guide Until Completion1990-08-0303 August 1990 Responds to NRC 900612 Request for Comments & Suggestions on Draft risk-based Insp Guide.Util Conducting Individual Plant Exam & Will Withhold Comment on risk-based Insp Guide Until Completion ELV-01943, Responds to Violation & Proposed Imposition of Civil Penalty in Insp Repts 50-424/90-11 & 50-425/90-11.Corrective Action: Complete Audit of Contents of Safeguards Info Container Performed & Unassigned Safeguards Info Dispositioned1990-07-27027 July 1990 Responds to Violation & Proposed Imposition of Civil Penalty in Insp Repts 50-424/90-11 & 50-425/90-11.Corrective Action: Complete Audit of Contents of Safeguards Info Container Performed & Unassigned Safeguards Info Dispositioned ELV-01949, Forwards Info Re Status of Pen Branch Fault Investigation. Investigations Conducted So Far Still Indicate That Pen Branch Fault Not Capable1990-07-26026 July 1990 Forwards Info Re Status of Pen Branch Fault Investigation. Investigations Conducted So Far Still Indicate That Pen Branch Fault Not Capable ELV-01500, Forwards Nuclear Decommissioning Funding Plan for Plant.Info Provides Assurance That NRC Prescribed Min Funding Will Be Available to Decommission Facilities1990-07-25025 July 1990 Forwards Nuclear Decommissioning Funding Plan for Plant.Info Provides Assurance That NRC Prescribed Min Funding Will Be Available to Decommission Facilities ML20055H6441990-07-23023 July 1990 Submits Summary of Snubber Types & Sample Plans for Functional Testing to Be Performed During Sept 1990 Outage ML20044B0311990-07-13013 July 1990 Forwards Vogtle Electric Generating Plant Unit 1 Reactor Containment Bldg 1990 Integrated Leakage Rate Test Final Rept. ML20044B1541990-07-12012 July 1990 Responds to NRC 900612 Ltr Re Violations Noted in Insp Repts 50-424/90-08 & 50-425/90-08.Corrective Actions:Eop Step Deviation Documents to Be Upgraded,Adding More Justification & Temporary Change Issued to Correct EOP Deficiencies ELV-01867, Responds to Violations Noted in Insp Repts 50-424/90-10 & 50-425/90-10.Corrective Action:Level Indication Error Corrected After Discrepancy Discovered1990-07-12012 July 1990 Responds to Violations Noted in Insp Repts 50-424/90-10 & 50-425/90-10.Corrective Action:Level Indication Error Corrected After Discrepancy Discovered ML20055F1651990-07-0909 July 1990 Forwards Comments Re NUREG-1410 ELV-01858, Advises That Full Compliance W/Violation Will Not Be Achieved Until Nov 1990,when Evaluation of VP-2693 Complete1990-07-0606 July 1990 Advises That Full Compliance W/Violation Will Not Be Achieved Until Nov 1990,when Evaluation of VP-2693 Complete ML20044A8851990-07-0606 July 1990 Forwards Response to NRC Question on Steam Generator Level Instrumentation Setpoints,Per Revised Instrument Line Tap Locations.Tap Location Will Be Changed from Above Transition Cone to Below Transition Cone ELV-01834, Forwards Response & Comments to Regulatory Effectiveness Review Rept.Encl Withheld (Ref 10CFR73.21)1990-06-28028 June 1990 Forwards Response & Comments to Regulatory Effectiveness Review Rept.Encl Withheld (Ref 10CFR73.21) ML20044A2791990-06-25025 June 1990 Responds to Generic Ltr 90-04, Request for Info on Status of Licensee Implementation of Generic Safety Issues Resolved W/Imposition of Requirements or Corrective Actions. Table Indicating Status of Each Generic Safety Issue Encl ML20043J0171990-06-22022 June 1990 Discusses Corrective Actions for Plant Site Area Emergency, Per 900514 Ltr.Jacket Water High Temp Switches Calibr for Diesel Generators,Using Revised Calibr Procedure ML20043H3061990-06-15015 June 1990 Forwards Rev 3 to ISI-P-014, Inservice Insp Program, for Review & Approval,Per Tech Spec 4.0.5 Re Surveillance Requirements.Rev Includes Withdrawal of Relief Requests RR-45,47,48 & 54 ML20043G2071990-06-12012 June 1990 Forwards Amend 18 to Physical Security & Contingency Plan. Amend Withheld (Ref 10CFR73.21) ML20043G1021990-06-0606 June 1990 Requests Temporary Waiver of Compliance from Requirements of Action Statement 27 of Tech Spec 3.3.2 for Period of 6 H When Two Operating Control Room Emergency Filtration Sys Trains Shut Down for Required Testing ML20043E6901990-06-0505 June 1990 Forwards Rev 12 to Emergency Plan & Detailed Description & Justification of Changes.W/O Rev ML20043G7651990-06-0505 June 1990 Forwards Rev 13 to Emergency Plan & Description & Justification of Changes ML20043B5991990-05-25025 May 1990 Forwards Scope & Objectives Re 1990 Annual Emergency Preparedness Exercise to Be Conducted on 900801 ML20043B5981990-05-24024 May 1990 Responds to Violations Noted in Insp Rept 50-424/90-05 on 900217-0330.Corrective Actions:Locked Valve Procedure Revised to Eliminate Utilization of Hold Tag on Valves Required by Tech Specs to Be Secured in Position ML20043B6291990-05-22022 May 1990 Forwards Rev 5 to ISI-P-008, Inservice Testing Program, Per Tech Specs 4.0.5 Re Surveillance Requirements & Generic Ltr 89-04 ML20043B6351990-05-22022 May 1990 Forwards Rev 2 to ISI-P-016, Inservice Testing Program, Per Generic Ltr 89-04, Guidance on Developing Acceptable Inservice Testing Programs. ML20042H0601990-05-14014 May 1990 Forwards Summary of Corrective Actions for 900320 Site Area Emergency Due to Loss of Offsite Power Concurrent W/Loss of Onsite Emergency Diesel Generator Capability.Truck Driver Disciplined for Lack of Attention ML20042G7301990-05-11011 May 1990 Forwards Revised Pages for May 1989,Jan & Mar 1990 Monthly Operating Repts for Vogtle Electric Generating Plant,Units 1 & 2.Revs Necessary Due to Errors Discovered in Ref Repts ML20042E2911990-04-18018 April 1990 Forwards Amend 17 to Security Plan.Amend Withheld (Ref 10CFR2.790) ML20042E7481990-04-0909 April 1990 Requests Approval to Return Facility to Mode 2 & Subsequent Power Operation,Per 900320 Event Re Loss of Offsite Power Concurrent W/Loss of Onsite Emergency Diesel Generator Capability ML20012E9001990-03-28028 March 1990 Provides Supplemental Response to Station Blackout Rule,Per NUMARC 900104 Request.Mods & Associated Procedure Changes Identified in Sections B & C W/Exception of Mods to Seals Will Be Completed 1 Yr from Acceptance of Analysis ML20012E8581990-03-28028 March 1990 Suppls Response to NRC Bulletin 88-010,Suppl 1 Re Traceability Reviews on Molded Case Circuit Breakers Installed in safety-related Applications.All Breakers Procured & Installed in Class 1E Equipment Reviewed ML20012E9761990-03-27027 March 1990 Requests Withdrawal of Inservice Insp Relief Requests RR-45, RR-47,RR-48 & Conditional Withdrawal of RR-54 Based on Reasons Discussed in Encl,Per 900206 Conference Call ML20012D8561990-03-22022 March 1990 Submits Special Rept 1-90-02 Re Number of Steam Generator Tubes Plugged During 1R2.One of Four Tubes Exceeded Plugging Limit & Required Plugging.Remaining Three Tubes Plugged as Precautionary Measure.No Defective Tubes Detected ML20012D6641990-03-22022 March 1990 Provides Followup Written Request for Waiver of Compliance to Make Tech Spec 3.04 Inapplicable to Tech Spec 3.8.1.2 to Permit Entry Into Mode 5 W/Operability of Diesel Generator a & Associated Load Sequencer Unverified ML20012D3681990-03-19019 March 1990 Forwards Proprietary & Nonproprietary Suppl 2 to WCAP-12218 & WCAP-12219, Supplementary Assessment of Leak-Before-Break for Pressurizer Surge Lines of Vogtle Units 1 & 2, Per 900226 Request.Proprietary Rept Withheld (Ref 10CFR2.790) ML20012D3401990-03-19019 March 1990 Submits Response to 891121 Request for Addl Info Re Settlement Monitoring Program.Current Surveying Procedures Used by Plant to Monitor Settlement of Major Structures Outlined in Procedure 84301-C.W/41 Oversize Drawings ML20012D6631990-03-15015 March 1990 Responds to Generic Ltr 89-19 Re Resolution of USI A-47 on Safety Implications of Control Sys in Lwrs.Overfill Protection Sys Sufficiently Separate from Control Portion of Main Feedwater Control Sys & Not Powered from Same Source ML20012C4681990-03-0606 March 1990 Provides Summary Rept of Property Damage Insurance Levels, Per 10CFR50.54(w)(1) ML20012B2891990-03-0606 March 1990 Forwards Plant Pipe Break Isometrics,Vols 1 & 2 & Advises That Encl Figures Have Been Revised to Be Consistent W/Pipe Analysis in Effect at Time That Unit 2 Received Ol,Including Revs Through 890930.W/309 Oversize Figures ML20012B2421990-03-0606 March 1990 Forwards Cycle 3 Radial Peaking Factor Limit Rept & Elevation Dependent Peaking Factor Vs Core Height Graph ML20011F5291990-02-26026 February 1990 Withdraws 881107 Proposed Amend to Tech Spec 3.8.1.1, Revising Action Requirements for Inoperable Diesel Generator to Clarify Acceptability of Air Roll Tests on Remaining Operable Diesel Generator ML20011F5261990-02-26026 February 1990 Forwards 1989 Annual Rept - Part 1. Part 2 Will Be Submitted by 900501 ML20011E8911990-02-12012 February 1990 Advises That Hh Butterworth No Longer Employed by Util 1990-09-06
[Table view] |
Text
a: -
LGehrgia Power dompany .
~; .
.i
'M r 333 Piedmyit Avenue' i ~ "
Atlanta, Georgia 30308 ' f e Tefcohone 404.526-(726.
I
L Post Office dox 4545 .
- Atlanta, Georg'a 30302.
}
5(
Georgia Power - )
a ! C;c A G3ter the southem ekttic sntem l
- %ce President and General Manager '
. Vogtte Project
, November 11, 1983 Mr.-Harold ~R. Eenton,' Director ..
File: X6BB06 Office of Nuclear-Reactor-Regulation _
Log: GN-281 U..S.: Nuclear Regulatory Commission
' Washington, D.C. 20555
~
j F
NRC DOCKET NUMBERS.50-424 AND 50-425 CONSTRUCTION PERMIT NUMBERS CPPR-103 AND CPPR-109 ~
V0GTLE ELECTRIC GENERATING PLANT - UNITS 1 AND 2 I
' ARBITRARY INTERMEDIATE PIPE BREAKS 4
Dear Mr. Denton:
+ ;
Georgia Power' Company (GPC) has followed closely the recent activities of- i U the Nuclear Regulatory Commissicn (NRC) staff .and the nuclear industry
-related to the treatment of design basis pipe breaks in high energy piping a
systems. 1In particular, it is noted that the NRC staff has expressed an interest .in the industry's propesal to modify the current pipe break criteria to: eliminate from design consideration those intermediate breaks'
- generally referred to as arbitrary intermediate breaks, i.e. those break )
locations.which, based on stress analysis, are below the stress limits and/or I
[ .the cumulative usage factors specified in the current NRC criteria, but i
I' -are selected 'to provide a minimum of two breaks between terminal ends. NRC 1 staff and-industry' discussions with the' Advisory Committee on Reactor Safe-
~
!. 1 guards (ACRS) on March.29 and June 2,~1983 have indicated general agreement i with' these; objectives and recognition that elimination of the arbitrary
. intermediate breaks offers' considerable benefits due to the deletion of the l associated pipe whip restraints and other provisions currently incorporated p
.in plant designs to mitigate the effects of such breaks. i 1
The' break selection criteria currently employed by GPC for Plant Vogtle is l- .taken from NRC Branch Technical Positions ASB 3-1 and MEB 3-1 and described in section 3.6 of the Plant Vogtle Final Safety Analysis Report (FSAR). These ,
documents require that pipe' breaks be considered at terminal ends and at I intermediate locations where' stresses or cumulative usage factors exceed
[; specified limits. If two intermediate locations cannot be determined based i
~
on the above, i.e. stresses and cumulative usage factors are below specified l r limits,.then the two highest' stress locations are selected.
t GPC concurs with the nuclear industry in the belief that current knowledge F -and experience supports the conclusion that designing for the arbitrary l ' intermediate breaks is not justified and that this requirement should be deleted. :This conclusion is supported by extensive operating experience goof nin over.80 operating U.S. plants and a number of similar plants overseas in which no piping failures have been known-to occur that would suggest
/l
~
8311180049 831111 PDR ADOCK 05000424
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r, '
- 1 m <
t Mr.: Harold R. Denton, Director File: X6BB06
< November 11,L1983. .,
Log: GN-281 Page'2--
~
- , r Lthe4 need to-design protective features to mitigate'the' dynamic effects of' Jarbitrary intermediate breaks.-LArbitrary intermediate breaks are often.
- postulated at. locations ~where stresses are well below the ASME Code allow- '
ables and within a few percent _of the. stress leiels at other points in the same system. This results in complicated protective features being
.provided"for-specific break. locations in.thelpiping system that provide M y;
little to enhance overall: plant safety.- .
L '
'In practice,' consideration'of these two arbitrary intermediate breaks is.
particularlyl difficult because the location of the high stress points may
~
move several.tbnes as the seismic design and analysis of structures and l' piping develops. The; industry ~ recognizes ~that the revised NEB 3-1, which was included-in the July 1981 revision to the Standard Review Plan (NUREG- '~
'0800), provides criteria for'not having to relocate intermediate break
~
' points when highest stress locations shift.as a result of piping reanalysis.
[ As a practical matter, however,-these criteria provide'little relief, since 4
the burden is:on the designer to prove,that not postulating breaks at.
r -relocated highest stress points doesinot degrade safety. This may require extensive additional analysis of break / target interactions for the relocated breakLpoints and could result in design, fabrication and installation of
. additional pipe whip restraints at the relocated break points and elimination
.of previously installed restraints at abandoned break points. Early determi-nation of exact break locations is quite important because of all of the
- secondary effects of the pipe break to be considered.
J The benefits to be realized from the elimination of the arbitrary intermediate Lbreak locations center primarily around the elimination of the associated l pipe whip restraints and other structural provisions to mitigate the con-- ,
' sequences of these breaks. While a substantial reduction in capital costs t .for these' restraints and~ structures can be realized immediately, there are
' also significant operational benefits to be realized over the 40 year life
~
I of the plant.
Access during plant operation for such activities as maintenance and inservice inspection is improved due to the elimination of' congestion
- created Lby these restraints arul the supporting structural steel, and in some cases'due to the need to remove some. restraints to gain access to welds. ~In addition to the decrease in maintenance effort, a significant reduction in man-rem exposure can be realized through fewer ~ manhours spent
'in radiation areas. Also, theineed to verify adequate cold and hot clear-l ances between pipes and restraint during initial heatup, which requires '
additional hold points during this already critical'startup phase, can he dispensed with..
t-e
- ,, w , . -w . . .. , . ,_ _ .-- ___...--_.-___._...-.._.._.._,_,m... .._.,-,-._.-.._,.w._,,,_.
-. . - -. = . . .- --
, a
. MrdHarold-R.L Dentyn, Direttcr. File: ' X6BB06 cNovember.11, 1983 ,
/ Log: GN-281
. Page53 Recovery from unusual plcnt conditions would also be improved by elimination of this congestion. In'thelevent -of a radioactive release or spill inside the plant, decontamination-operations weald be much more effective if the complexLshapes, represented by_the stre2tural frameworks supporting the restraints, were eliminated. This-results_in decreasing' man-rem exposures-associated with decontamination and restoration activities. Similarly,
~
access'for control of fires within these areas of the_ plant would beJimproved, especially under. low visibilityfconditions. Substantia 1'overall benefits
. . in these' areas would be realized by reducing the number of whip restraints required.
By design, whip restraints fit closely around the high energy piping with gaps' typically being on the order of half an inch. These restraints and their supporting steel increase the' heat loss to the surrounding environment significantly. Also, because thermal' movement of the piping system during
-startup and shutdown could deform the piping insulation against the fixed '
whip' restraint,1the insulation must be cut back in these areas,~ creating convection gaps adjacent'to the restraint, which also increases heat loss to.the environment._ This is a major contributor to the tendency of many containments to operate at temperatures-near technical-specification limits.
- The elimination of wnip restraints associated with arbitrary intermediate' '
breaks would assist in controlling-the normal environmental temperatures
- and improve system operational efficiency.
'For the above reason, GPC requests NRC approval for the application of the following alternative pipe break criteria (excluding the RCS primary-loop) on. Plant Vogtle: '
ASME Section III Piping'Inside Containment e Piping systems shall be designed to accommodate pipe breaks at terminal ends'and locations where the stress or usage factor criterion of MEB 3-1 are exceeded. No arbitrary intermediate breaks will be postulated when the stress and/or usageifactor criterion arecnot exceeded.
e For_ breaks that must be taken, the design will accommodate pipe whip, i mjet' impingement and compartment pressurization resulting from mechanistic treatment of the break. Current acceptable methods for limiting break opening, moderate and low energy exclusions, limited duration operation, etc.-may still be-applied.
~ ASME Section.III and Seismically Designed Non-ASME Section III Piping Outside Containment e Piping systems shall be designed to accommodate pipe breaks at terminal ends and locations where the stress criterion of MEB 3-1 are exceeded.
No arbitrary intermediate breaks will be postulated when the stress criterion are not exceeded.
,,r-o '-++'ey-.w,-., , , , ry--,-uptw-, ev--.. ,-mn----w-ve,,,- .,m,--w-,-. -t,.,r--Ww-e-.,,.,,--=., v.er----=,e-.wre---+-e.,re,-,m, ,-ea -y .--.r-y,--.,w.,- -.
g . . , , -.
_ .~ . <
lKr. Harold'k.'Denton, Director ' File: fX6BB05:
>J . November 11,'1983 Log: GN-281.
-Page 41
~
~
For breaks that must Lbc taken, the design will accommodate pipe whip
~
e
- and. jet impingement' effects resulting from mechanistic treatment of Lthe break. Current acceptable methods -for limiting break opening,.
moderate and low er.ergy exclusions, limited duration operation, etc.
. may'still,be' applied, e-J For' environmental qualification of-equipm nt and structural design-
.of compartments or enclosures traversed by high energy piping systems,..
breaks'wi11' continue to be postulated in accordance with the present project criteria, i.e. in each compartment or enclosure traversed by the high energy piping system, non-mechanistic breaks are postulated at the location.that results in the most severe environmental consequences.
Therefore, elimination of'the arbitrary intermediate breaks will not
' impact the environmental qualification program or planc structural design.
Application of the alternative' pipe. break criteria described above will not alter Plant Vogtle's commitment to qualify in the design of structures,
~
systems.and. components important to safety. The Plant Vogtle quality
- , - assurance program'will continue to ensure that' structures, systems and components important to safety are (esigned, fabricated, erected and tested
-'to the quality standards commensurate with the safety function to be per-l formed. "
The number of pipe breaks (terminal end and intermed#. ate) currently postulated in the Plant Vogtle pipe break analysis and associated pipe whip restraints is summarized in Attachment A. This attachment also identifies the number of pipe: breaks and pipe whip restrants to be elimianted from the design
.through application of the proposed alternate crteria.
Attachment B provides specific identification of the arbitrary intermediate breaks and pipe whip restraints to'be-eliminated from=the design, the system in which the breaks and associated restraints are located and the FSAR figure which shows their physical location within a given system.
l i GPC has evaluated the potential cost savings and operational benefits that result from the elimination of arbitrary intermediate breaks. These-benefits
. include $3-4 million savings in analysis, design, fabrication and installation of ' associated pipe whip restraints and jet impingement barriers and 400-500
. man-rem in dose reductions for both Vogtle units over the 40 year plant life.
. This dose reduction results in an additional operational cost savings of $2-3 l million. A detailed breakdown of the benefits realized by the climination
[ of the crbitrary intermediate breaks is provided in Attachment C. The actual benefits that GPC will realize are expected to be higher than these due to the hidden factors and intangibles that are difficult to identify at this
' time. It .is clear, however, that elimination of the arbitrary intermediate breaks is both safety effective and cost effective.
e p
+ v+w -cw. c-w,--wy wr , +m- -- r----.r- --re, --e-w - - - r -r+-'- m , ~. 4-v =- * , --
.r. -* - - -w-*n ----v..ar---er--~- r v-'a---+ - * - w-- +~--
ti Mr.1 Harold.R..Denton, Director . File: X63B06 November 11,11983 Log: GN-281 Page 5 s
The percentage of the total potential benefits that can be realized by GPC for Plant Vogtle becomes_a matter of_ timing due to the advanced stage
~
of-design and construction. To make it possible for GPC to realize the maximum benefits afforded by this proposed change in the pipe break criteria, immediate attention by the NRC is requested.
GPC believen' that elimination of the arbitrary intermediate breaks is entirely independer.t of the NRC's current consideration of changes in the criteria for breaks in the primary coolant loops _and, selected application of leak before break me:hodology at high stress points, and that these efforts can proceed independently.
We would appreciate a favorable response to the proposed change in the pipe break criteria by January 6, 1984.
You s truly
/ / _
/ 4-D. O. Fos er.
DOF/sw xc: R. A. Thomas D. E. Dutton O. Batum J. A. Bailey' M. Malcom G. Bockhold, Jr.
L.'T. Gucwa M. A.~ Miller j G. F. Trowbridge, Esquire l
t.
l l
L l
. i Attachment A d
SUMMARY
OF BREAKS AND RESTRAINTS (PER UNIT) i PMW VMEE Total Summary To be Deleted No. of No. of No. of No. of Breaks Restraints Breaks Restraints Inside containment e ASME Class 1 Branch Lines
, Terminal Ends 79 35
) Intermediate - High Stress
- 81 19 i - Arbitrary 22 6 22 6 4 e ASME Class 2 i Terminal Ends 88 44
- Intermediate - High Stress
- - Arbitrary 70 83 70 83
. e Total 348 192 92 89 I Outside Containment i
e ASME Class 2 & 3 and -Seismically Designed
~
ANSI B31.1 Piping Terminal Ends 134 18 l Intermediate - High Stress
- 4 2 i - Arbitrary 90 21 90 21 i
e Total 228 41 90 21 Total - Inside & Outside Containment 576 233 182 110
- Stress and/or usage factor criterion of MEB 3-1 are exceeded.
i
F' .
- Attechment B Pags 1 of 6 POSTULATED ARBITRARY INTERMEDIATE PIPE BREAKS - PLANT V0GTLE
[This-table ~provides specific identification of.the arbitrary intermediate breaks (event' number) and pipe whip' restraints (PWRs) to be eliminated from the Plant Vogtle design, the system in which the breaks and associated PWRs are located.and the FSAR figure which shows their physical location within a
.given system. An asterisk in the last column denotes that the pipe break event number and PWR (where applicable) were identified after= submittal of the' FSAR and.as a result'are not shown in FSAR Figure 3.6.1-1. The event number and PWR number will be included in a future FSAR amendment.
FSAR Pipe Break Figure 3.6.1-1 System Event Number PWR Number Sheet Number INSIDE CONTAINMENT - ASME SECTION III, CLASS 1 Safety Injection (Cold Leg) - P-0301-C 27 Loop 1 -
Safety Injection (Cold Leg) - P-0304-C 32.
Loop 2 -
Safety Injection (Cold Leg) - P-0309-C PBR-91 38 Loop 3 Safety Injection (Cold Leg) - P-2067-C 35 Loop.41
. Chemical & Volume Control P-0121-C 78 (Charging) P-0122-C. PBR-153 78
?
Safety Injection (Hot Leg) - P-0422-B PBR-87 20 Loop 2 P-0423-B PBR-86 20 Reactor Coolant - Pressurizer P-0105-C 8 Surge l Seal Whter Injection - Loop 1 P-2501-C 195 Seal Water Injection - Loop 2 P-2505-C 196 P-2506-C 196
- Seal Water Injection - Loop 3 P-2511-C 197 Seal Water Injection - Loop 4 P-2513-C 198 P-2515-C 198 Reactor Coolant Drain - P-0342-C 65 Loop ~1 P-0343-C PBR-128 65 l
Reactor Coolant Drain - P-0346-C 66 Loop 2 P-0347-C PBR-131 66
}
Attachment B Pag 2 2 of 6 FSAR Pipe Break Figure 3.6.1-1 System Event Ncmber PWR Number Sheet Number
. Reactor Coolant Drain - P-0350-C 67 Loop'3_ P-0351-C -
67 Reactor Joolant Drain - P-0357-C 68 Loop-4-(Excess Letdown)
INSIDE CONTAINMENT - ASME SECTION III, CLASS 2 Chemical & Volume Control P-2122-C 80
-(Charging) P-2123-C 80 Chemical & Volsme Control P-2117-C PBR-162 87 (Letdown) P-2119-C 85
+
Chemical & Volume Control P-0124-C 78 (Charging) P-0125-C PBR-156 78 ,
Auxiliary Feedwater - Loop 1 P-2265-C ?BR-277,279 156 P-2266-C PBR-278,280,281,282 156 P-2267-C 156 Auxiliary Feedwater - Loop 2 P-2271-C PBR-290,291 157 P-2272-C PBR-292,293,315 157 P-2273-C PBR-294,295,296 157 P-2274-C 157 Auxiliary Feedwdter.- Loop 4 P-2261-C PBR-269,272 159 P-2262-C PBR-271,273,274,275 159 Steam Generator Blowdown - 'P-1649-C 132 Loop-1 - P-1645-C 133 P-1641-C 130 P-1642-C PBR-188 130 Steam Generator Blowdown - P-1672-C 135 Loop 2 P-1673-C 135 P-1666-C 139 P-1667-C PBR-287 139 l Steam Generator Blowdown - P-1652-C 141 Loop 3 P-1653-C 141
< P-1659-C PBR-228,229 146 P-1660-C PBR-230 146 Steam Generator Blowdown - P-1675-C 147 Loop 4 P-1634-C 147 P-1637-C PBR-182 149 3
P-1636-C 149
- , - _ _ . . , - - - . . _ _. - _ _ _ - _ _ _ _ _ . . . ,- . _ _ - . - . _ - ~ . _ - , _ _ , . . . . - . - _ -
( . . .. - .
3; Attechment B Pega 3 cf 6
{
FSAR Pipe Break Figure 3.6.1-1 System Event Number PWR Number Sheet Number Steam Generator Wet Layup - P-1029-C 152.
Loop 1 P-1030-C -
152
. Steam Generator Wet Layup - P-1037-C 153 Loop 2 P-1038-C- 152 Steam Generator Wet Layup - P-1033-C 154 Loop 3 P-1034-C 154 Steam Generator Wet Layup - P-1025-C 155 Loop 4 P-1026-C 155 Main Steam - Loop 1 P-1004-C PBR-17,18,19 2 P-1005-C PBR-20,22 2 Main Steam - Loop 2 P-1010-C PBR-63,64,65 3 P-1011-C PBR-66,68 3 Main Steam - Loop 3 P-1007-C PBR-70,71,72 3 P-1008-C PBR-73,75 3 Main Steam - Loop 4 P-1001-C PBR-10,11,12 2 P-1002-C. PBR-13,15 2 Main Feedwater - Loop 1 P-1108-C PBR-38,40,41,42,120 5 P-1109-C PBR-43,45 5 Main Feedwater - Loop 2 P-1111-C PBR-55,57,58,59,121 6 P-1112-C PBR-60,54 6 Main Feedwater - Loop 3 P-1101-C PBR-46,49,48,50,122 6 P-1102-C PBR-51,53 6 Main Feedwater - Loop 4 P-1105-C PBR-8,5,6,4,123 5 P-1106-C PBR-3,1 5 Chemical & Volume Control P-2518-C PBR-316 195 (Seal Water Injection - Loop 1)
Chemical & Volume Control P-2520-C PBR-321 196
( (Seal Water Injection - Loop 2) P-2521-C PBR-320 196 i
l Chemical & Volume. Control P-2523-C 197 (Seal Water Injection - Loop 3) P-2524-C PBR-322,317 197 l
j- Chemical & Volume Control P-2526-C PER-323 198 (Seal Water Injection - Loop 4) P-2527-C PBR-318 198 l
l Safety Injection (Accumulator P-2027-C 61 Drain) - Loop'1 P-2028-C 61 l
I
e Attachment B Page 4 of 6 FSAR Pipe Break Figure 3.6.1-1 System Event Number PWR Number Sheet Number Safety Injection (Accumulator P-2031-C 62 Drain) Loop 2 P-2032-C _
62
. Safety. Injection (Accumulator P-2035-C 63 Drain) - Loop 3 P-2038-C 63 Safety Injection (Accumulator P-2041-C 64 Drain) - Loop 4 P-2044-C 64 OUTSIDE CONTAINMENT Auxiliary Feedwater P-22Po-C. 174*
P-2290-C 174*
P-2293-C 174*
Chemical & Volume Control P-2136-C 88 (Charging) P-2137-C 88 P-2142-C 89 P-2143-C 89 Chemical & Volume Control P-2177-C 91 (Charging) P-2176-C 91 P-2159-C 93 P-2158-C 93 P-2164-C 98 P-2163-C 98 P-2162-C 98 P-2151-C 100 P-2152-C 100 P-2153-C 97 P-2192-C 103 P-2193-C
^
104 P-2126-C 117 P-2125-C 117
! P-2131-C 118*
l P-2132-C 118*
Chemical & Volume Control P-2186-C 106 (Letdown) P-2187-C 106 P-2198-C 108 P-2197-C 108 P-2113-C 109 P-2112-C' 109 Main Steam - Loops 1 & 4 P-1055-C PBR-201, PBR-218 2 P-1054-C PBR-217 2 9
Main Feedwater - Loops 1 & 4 P-1149-C PBR-241, PBR-242 5 Auxiliary Feedwater - Train A P-2204-C 190 P-2201-C 190 u -- - . - _ . _ . _ . . , _ ~ . _ . . . _ _ . _ _ . . _ . _ _ _ _ . _ _ _ _ . . _
Attachment B Pega 5 of 6 FSAR j Pipe Break Figure 3.6.1-1 l System Event Number PWR Number Sheet Number q Auxiliary Feedwater - Train B P-2207-C 189 P-2209-C _ 189 Auxiliary Feedwater - Train C 'P-2216-C 180 P-2215-C -180 P-2240-C 180*
P-2239-C 180 Steam Generator Blowdown - P-1601-C PBR-301 119 Loop 1 P-1602-C 119 Steam Generator Blowdown - P-1605-C PBR-302 120 Loop 2 P-1606-C- 120
- Steam Generator Blowdown - P-1609-C PBR-303 121 Loop 3 P-1610-C 121 Steam Generator Blowdown - P-1620-C
~
PBR-304 122 Loop 4 P-1621-C 122 Steam Generator Blowdown - P-1627-C 125 Common P-1628-C 125 Main Steam (Atmos. Dump) - P-1087-C 70 Loop 1 P-1088-C 70 Main Steam (Atmos. Dump) - P-1090-C >
73 Loop 4 - P-1091-C 73 Main Steam - Loop 1 P-1084-C
- Main Feedwater - Loops 2 & 3 P-1119-C PBR-249 6 ,
P-1118-C PBR-247 6 P-1117-C 6 Main Feedwater - Loops 1 & 4 P-1140-C PBR-234, PBR-263 7 P-1139-C 7 P-1137-C 7 Main Steam - Loops 2 & 3 P-1064-C PBR-206 3 P-1068-C PBR-208, PBR-209, 3 PBR-222 P-1074-C 4 P-1075-C 4 Main Steam - Loops 1 & 4 P-1061-C PBR-213 4 P-1062-C 4 Auxiliary Feedwater - Train A P-2276-C 186 P-2277-C 186
Atte.chment B Pign 6 cf 6 FSAR Pipe Break Figure 3.6.1-1 System Event Number PWR Number Sheet Number
- Auxiliary Feedwater - Train C P-2241-C 179*
P-2278-C -
179 Auxiliary Feedwater P-2236-C 170
^
.P-2231-C 169 Main-Steam - Loop 2 P-1017-C P-1018-C
--Main Steam (Atmos. Dump) - P-1093-C 71 Loop 2 P-1094-C PBR-260 71 Main Steam (Atmos. Dump) - P-1096-C PBR-257 72 Loop 3 P-1097-C ,
72 Main Steam - Loop 2 P-1125-C PBR-262
' Auxiliary Steam P-1129-C
P-1153-C
- I e- .- . . . . - , - - -
-r .-,. . - e----r,,e-,,r-+-,---- c,,-----+r-e--,-.-- -,----e -v --- + = , - , .- t -r-,- +=t --y-*-------wwv.--
ATTACHMilff C
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e SUt91ARY OF BENEFITS FOR TIJE ELIMINATION OF ARBITRARY INTERMELIATE PIPE BREAKS -
PLANT VOCTII UNITS 1 & 2 Changes Resulting from Break Elimination Cost Savings (1983 Rates) Operational Benefits Elimination of 110 Pipe Whip Restraints e Design, Fabrication and Installation Costs
- e Potential improvement in quality of inservice (PWRs) Per Unit inspection (ISI) e Dome Reduction Costs e Cose reduction free improved personnel access during maintenance, ISI and recovery from unusual plant conditions, e.g., radioactive spills, fires, etc.
e Improved capability to recover from unusual plant conditions, e.g., decontamination following radioactive spills, access for fire fightir.g. etc.
e Redu, 1 system heat loss resulting from improved insulation design e Dose reduction and improved construction schedule by eliminating the need to set and maintain PWR clearance gaps Elimination of Jet Barriers and/or o Barrier Design, Fabrication and Installation e Dose reduction from improved personnel cccess Equipment Relocation Costs
- during maintenance and recovery free unusual plant con 4itions, e.g., radioactive spills, e Dose Reduction Costs fires, etc.
- Relocation Costs e Improved capability to recover from unusual plant conditions, e.g., decontamination following radioactive spills, access for fire fighting, etc.
Elimination of Analyses Associated e Jet Impingement Load and Pipe Whip Analyses e Improved system layout and design for future with the Dynamic Effects and Loadina Costs
- plant modifications Conditions TOTAL SAVINGS (UNITS 1 AND 2) $5-7 Million 400-500 man-rea in dose redudtion over the 40 year plant life.
- 0ne Time Cost Applicable to Both Units.
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