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Category:CORRESPONDENCE-LETTERS
MONTHYEARML17241A5001999-10-21021 October 1999 Forwards Rev 3 to Emergency Response Data Sys (ERDS) Data Point Library for St Lucie Unit 1.Rev Provides Replacement Pages & Follows Format Recommended by NUREG 1394, ERDS Implementation, Rev 1,App C ML17309A9981999-10-19019 October 1999 Forwards Revised Epips,Including Rev 3 to EPIP-10 & Rev 25 to HP-202.EPIP-10 Added Onsite Monitoring Points,Made Administrative Changes & Incorporated New Attachments & HP-202 Added Red Team Survey Points ML20217F6171999-10-0808 October 1999 Forwards Insp Repts 50-335/99-11 & 50-389/99-11 on 990827 & 990907-09.No Violations Identified.Matl Encl Contained Safeguards Info as Defined by 10CFR73.21 & Disclosed to Unauthorized Individuals Prohibited by Section 147 of AEA ML17241A4811999-10-0101 October 1999 Reports Number of Tubes Plugged During Unit 1 Refueling Outage SL1-16,per TS 4.4.5.5.a ML20212M1601999-09-28028 September 1999 Refers to 990908 Engineering Meeting Conducted at NRC Region II to Discuss Engineering Issues at Lucie & Turkey Point Facilities.List of Attendees & Copy of Presentation Handout Encl ML17241A4701999-09-25025 September 1999 Forwards Info Requested by NRC Staff During 990916 Telcon to Complete Staff Review of Request for risk-informed Extension of Action Completion/Aot Specified for Inoperable Train of LPSI Sys at Plant ML17241A4721999-09-24024 September 1999 Forwards Rev 1 to Plant Change/Mod (PCM) 99016 to St Lucie Unit 1,Cycle 16 COLR, IAW TS 6.9.1.11.d.Refueling Overhaul Activities Are Currently in Progress & Reactor Operations for Cycle 16 Are Scheduled to Commence in Oct 1999 ML17241A4681999-09-22022 September 1999 Requests Restriction Be Added to Senior Operator License SOP-21093 for TE Bolander.Nrc Forms 369,encl.Encl Withheld Per 10CFR2.790(a)(6) ML17241A4671999-09-20020 September 1999 Forwards Completed NRC Form 536, Operator Licensing Exam Data, for St Lucie Units 1 & 2,as Requested in Administrative Ltr 99-03, Preparation & Scheduling of Operator Licensing Exams. ML17241A4581999-09-13013 September 1999 Forwards Info Requested by NRC Staff During 990630 & 0816 Telcons,To Complete Review of Proposed License Amend for Fuel Reload Process Improvement Program ML17241A4531999-08-31031 August 1999 Informs That No Candidates from St Lucie Plant Will Be Participating in PWR Gfes Being Administered on 991006 ML17241A4521999-08-31031 August 1999 Withdraws Relief Request 16 & Suppl Relief Request 15 with Info Requested During 990526 Telephone Conference Re ISI Insp Plan,Third 10-yr Interval ML17241A4501999-08-26026 August 1999 Informs That FPL Has Reviewed Reactor Vessel Integrity Database,Called RVID2,re Closure of GL 92-01,rev 1,suppl 1. Requested Corrections & Marked Up Pages from Rvid 2 Database Summary Repts That Correspond to Comments,Attached ML17241A4371999-08-13013 August 1999 Forwards fitness-for-duty Program Performance Data for six- Month Period Ending 990630,per 10CFR26.71(d) ML17241A4461999-08-11011 August 1999 Requests That W Rept Entitled, Evaluation of Turbine Missile Ejection Probability Resulting from Extending Test Interval of Interceptor & Reheat Stop Valves at St Lucie Units 1 & 2, Be Withheld from Public Disclosure L-99-171, Forwards Rev 56 to Physical Security Plan.Summary of Changes & Marked Up Copy of Revised Pages Also Encl.Encls Withheld from Public Disclosure Per 10CFR2.790(a)(3)1999-07-29029 July 1999 Forwards Rev 56 to Physical Security Plan.Summary of Changes & Marked Up Copy of Revised Pages Also Encl.Encls Withheld from Public Disclosure Per 10CFR2.790(a)(3) ML17309A9911999-07-26026 July 1999 Forwards Revised EPIPs & Revised Procedures That Implement Emergency Plan as Listed.Procedures Provides Instruction for Operational Support Ctr (OSC) Chemistry Supervisor to Establish Remote Labs at Locations Specified ML17241A4471999-07-22022 July 1999 Requests That Rev 1 to WCAP-14732 & Rev 1,Add 1 to WCAP-14732 Be Withheld from Public Disclosure ML17241A4221999-07-22022 July 1999 Forwards List of Proposed Licensing Actions for St Lucie Units 1 & 2,planned During Fys 2000 & 2001,in Response to Administrative Ltr 99-02, Operating Reactor Licensing Action Estimates. ML17241A4151999-07-22022 July 1999 Forwards Revised Relief Request 25 for Second 10-yr ISI Interval for Unit 2 ML17241A4101999-07-16016 July 1999 Forwards FP&L Supplemental Response to NRC Request for Info Re Y2K Readiness at Nuclear Power Plants ML17309A9881999-07-0707 July 1999 Forwards Rev 5 to EPIP-03, Emergency Response Organization Notification/Staff Augmentation. Rev 5 to EPIP-03 Was Revised to Transfer EP Responsibilities from Training Manager to Protection Svcs Manager ML20209F1541999-07-0606 July 1999 Informs That NRC in Process of Conducting Operational Safeguards Response Evaluations at Nuclear Power Reactors. Plant Chosen for Such Review Scheduled for Wk of 990823-26 ML17241A4011999-06-30030 June 1999 Forwards Info Copy of Florida Wastewater Permit (FL0002208) (Formerly NPDES Permit) Mod,Which Was Issued by Florida Dept of Environ Protection on 990604 ML17241A3971999-06-30030 June 1999 Forwards Suppl Response to GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Movs, as Requested in 990317 Ltr ML17355A3661999-06-30030 June 1999 Forwards Florida Power & Light Topical QA Rept, Dtd June 1999.Encl I Includes Summary of Changes Made to Topical QA Rept Since 1998 ML17241A3951999-06-29029 June 1999 Provides Response to NRC Request for Info Re Y2K Readiness at Nuclear Power Plants.Gl 98-01, Y2K Readiness of Computer Sys at Nuclear Power Plants, Requested Response on Status of Facility Y2K Readiness by 990701.Disclosure Encl ML17241A3731999-06-17017 June 1999 Supplements Relief Requests 4,11 & 13 for Third ten-year ISI Interval with Info Requested During 990526 Telcon.Expedited Review Is Requested by 990730 to Avoid Negatively Impacting Upcoming St Lucie Unit 1 Refueling Outage (SL1-16) ML17241A3641999-06-14014 June 1999 Submits Supplement to Relief Request 24 with Info Requested by Nrc.In Addition Relief Request 24 Is Identical to St Lucie Unit 1 Relief Request 4 for Third ISI Interval Being Supplemented by FPL Ltr L-99-139 ML20195F3871999-06-11011 June 1999 Final Response to FOIA Request for Documents.App a Records Being Withheld in Entirety (Ref FOIA Exemption 5) IA-99-247, Final Response to FOIA Request for Documents.App a Records Being Withheld in Entirety (Ref FOIA Exemption 5)1999-06-11011 June 1999 Final Response to FOIA Request for Documents.App a Records Being Withheld in Entirety (Ref FOIA Exemption 5) L-99-129, Forwards Rev 55 to Physical Security Plan,Summary of Changes & Marked Up Copy of Revised Pages.With Directions for Incorporating Rev Into Plan & Copies of Replacement Pages.Rev Withheld,Per 10CFR2.790(a)(3)1999-06-0909 June 1999 Forwards Rev 55 to Physical Security Plan,Summary of Changes & Marked Up Copy of Revised Pages.With Directions for Incorporating Rev Into Plan & Copies of Replacement Pages.Rev Withheld,Per 10CFR2.790(a)(3) ML17241A3561999-06-0707 June 1999 Forwards Rept Containing Brief Description & Summary of SEs for Changes,Tests & Experiments Which Were Approved for Unit 3 During Period of 970526-981209 ML17241A3601999-06-0707 June 1999 Forwards Correction to Annual Radiological Environ Operating Rept for CY98.Util Has Identified Transcription Error on Last Page of Attachment C of Rept,Results from Interlaboratory Comparison Program 1998 ML20195F3941999-05-27027 May 1999 FOIA Request That Memo from J Calvo to Fl Lebdon Re TIA - St Lucie,Unit 1 Environ Qualification of Woodward Governor Controls Be Placed in PDR ML17241A3461999-05-24024 May 1999 Forwards Revised Relief Request 22 to Clarify Several Areas of Relief.Nrc Action Is Requested to Be Complete by Aug 1999 to Support Planning for Spring 2000 Unit 2 Refueling Outage ML17241A3391999-05-20020 May 1999 Forwards Notification of Change to Small Break LOCA ECCS Evaluation Model Used for St Lucie Unit 1.Anomaly Was Discovered & Corrected That Resulted in Reducing Calculated PCT for Limiting SBLOCA by More than 50 F ML17241A3371999-05-20020 May 1999 Forwards Util Suppl to GL 95-07 Response Re pressure-locking & Thermal Binding of safety-related power-operated Gate Valves,In Response to NRC Second RAI Dtd 990225 ML20207C7531999-05-17017 May 1999 Discusses Issue Identified by FPL in Feb 1998 Involving Potential for Fire to Cause Breach of Rc Sys High/Low Pressure Interface Boundary & NRC Decision for Exercise of Enforcement Discretion ML17241A3301999-05-17017 May 1999 Forwards LER 99-004-00 Re as Found Cycle 10 Psv Setpoints Outside TS Limits,Which Occurred on 990415.Root Cause Determination Not Yet Complete.Suppl to Include Root Cause & Corrective Actions Will Be Submitted ML17309A9821999-05-10010 May 1999 Forwards Rev 36 to St Lucie Emergency Plan, Per 10CFR50.54(q).Executive Summary & Summary of Changes Incorporated by Rev,Encl IR 05000335/19980141999-04-29029 April 1999 Provides Confirmation of NRC Staff Conclusions Re Cited & non-cited Violations in Insp Rept 50-335/98-14 & 50-389/98-14.Utils Position Re Consideration of Multiple Spurious Actuations in Event of Fire,Reiterated ML17241A3221999-04-29029 April 1999 Provides Confirmation of NRC Staff Conclusions Re Cited & non-cited Violations in Insp Rept 50-335/98-14 & 50-389/98-14.Utils Position Re Consideration of Multiple Spurious Actuations in Event of Fire,Reiterated ML17229B1071999-04-28028 April 1999 Forwards 1998 Annual Environ Operating Rept for St Lucie Unit 2. Rept Includes Discussions of 5-inch Barrier Net Maint & Taprogge Condenser Tube Cleaning Sys Ball Loss,As Agreed at First Biennial Sea Turtle Meeting Held on 980120 ML17229B1051999-04-22022 April 1999 Requests That Listed Individuals Be Placed on Official Serve List for Nuclear Matl Safety & Safeguards Info Notices ML17229B1061999-04-21021 April 1999 Notifies NRC of Change in Medical Status of Licensed Operator Pf Farnsworth (Docket 55-21285,license SOP-21094). NRC Form 3996, Medical Exam Certification, Encl.Encl Withheld Per 10CFR2.790(a)(6) ML17309A9851999-04-15015 April 1999 Requests That NRC Review Denial of Appeal from Assessment of Fees Assessed in 981101 Invoice RS0062-99 & Assessment of Fees in Invoice RS0182-99 Which Was Also Denied in 990305 Ltr.Both Invoices Are for Fees Re Inspector GG Warnick ML20205M0431999-04-13013 April 1999 Eighth Partial Response to FOIA Request for Records.App Q & R Records Encl & Being Made Available in PDR ML17229B0951999-04-0808 April 1999 Requests Approval of Encl Revised Relief Request 6,in Response to 990322 Telcon with NRC & 10CFR55.55a(a)(3). Request States That Visual VT-3 Exams Will Be Conducted IAW IWA-2213 & Repairs Will Be IAW Util ASME Section IX Program ML17229B0821999-04-0707 April 1999 Requests Approval of Interim Relief Request 26 Re Repair Requirements for Class 2 ECCS Piping,Per 10CFR50.55a(a)(3) & 50.55a(g)(iii).Alternative Actions Apply Guidance of GLs 91-18 & 90-05 & ASME Code Case N-513.Evaluation,encl 1999-09-28
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARML17241A5001999-10-21021 October 1999 Forwards Rev 3 to Emergency Response Data Sys (ERDS) Data Point Library for St Lucie Unit 1.Rev Provides Replacement Pages & Follows Format Recommended by NUREG 1394, ERDS Implementation, Rev 1,App C ML17309A9981999-10-19019 October 1999 Forwards Revised Epips,Including Rev 3 to EPIP-10 & Rev 25 to HP-202.EPIP-10 Added Onsite Monitoring Points,Made Administrative Changes & Incorporated New Attachments & HP-202 Added Red Team Survey Points ML17241A4811999-10-0101 October 1999 Reports Number of Tubes Plugged During Unit 1 Refueling Outage SL1-16,per TS 4.4.5.5.a ML17241A4701999-09-25025 September 1999 Forwards Info Requested by NRC Staff During 990916 Telcon to Complete Staff Review of Request for risk-informed Extension of Action Completion/Aot Specified for Inoperable Train of LPSI Sys at Plant ML17241A4721999-09-24024 September 1999 Forwards Rev 1 to Plant Change/Mod (PCM) 99016 to St Lucie Unit 1,Cycle 16 COLR, IAW TS 6.9.1.11.d.Refueling Overhaul Activities Are Currently in Progress & Reactor Operations for Cycle 16 Are Scheduled to Commence in Oct 1999 ML17241A4681999-09-22022 September 1999 Requests Restriction Be Added to Senior Operator License SOP-21093 for TE Bolander.Nrc Forms 369,encl.Encl Withheld Per 10CFR2.790(a)(6) ML17241A4671999-09-20020 September 1999 Forwards Completed NRC Form 536, Operator Licensing Exam Data, for St Lucie Units 1 & 2,as Requested in Administrative Ltr 99-03, Preparation & Scheduling of Operator Licensing Exams. ML17241A4581999-09-13013 September 1999 Forwards Info Requested by NRC Staff During 990630 & 0816 Telcons,To Complete Review of Proposed License Amend for Fuel Reload Process Improvement Program ML17241A4531999-08-31031 August 1999 Informs That No Candidates from St Lucie Plant Will Be Participating in PWR Gfes Being Administered on 991006 ML17241A4521999-08-31031 August 1999 Withdraws Relief Request 16 & Suppl Relief Request 15 with Info Requested During 990526 Telephone Conference Re ISI Insp Plan,Third 10-yr Interval ML17241A4501999-08-26026 August 1999 Informs That FPL Has Reviewed Reactor Vessel Integrity Database,Called RVID2,re Closure of GL 92-01,rev 1,suppl 1. Requested Corrections & Marked Up Pages from Rvid 2 Database Summary Repts That Correspond to Comments,Attached ML17241A4371999-08-13013 August 1999 Forwards fitness-for-duty Program Performance Data for six- Month Period Ending 990630,per 10CFR26.71(d) ML17241A4461999-08-11011 August 1999 Requests That W Rept Entitled, Evaluation of Turbine Missile Ejection Probability Resulting from Extending Test Interval of Interceptor & Reheat Stop Valves at St Lucie Units 1 & 2, Be Withheld from Public Disclosure L-99-171, Forwards Rev 56 to Physical Security Plan.Summary of Changes & Marked Up Copy of Revised Pages Also Encl.Encls Withheld from Public Disclosure Per 10CFR2.790(a)(3)1999-07-29029 July 1999 Forwards Rev 56 to Physical Security Plan.Summary of Changes & Marked Up Copy of Revised Pages Also Encl.Encls Withheld from Public Disclosure Per 10CFR2.790(a)(3) ML17309A9911999-07-26026 July 1999 Forwards Revised EPIPs & Revised Procedures That Implement Emergency Plan as Listed.Procedures Provides Instruction for Operational Support Ctr (OSC) Chemistry Supervisor to Establish Remote Labs at Locations Specified ML17241A4221999-07-22022 July 1999 Forwards List of Proposed Licensing Actions for St Lucie Units 1 & 2,planned During Fys 2000 & 2001,in Response to Administrative Ltr 99-02, Operating Reactor Licensing Action Estimates. ML17241A4471999-07-22022 July 1999 Requests That Rev 1 to WCAP-14732 & Rev 1,Add 1 to WCAP-14732 Be Withheld from Public Disclosure ML17241A4151999-07-22022 July 1999 Forwards Revised Relief Request 25 for Second 10-yr ISI Interval for Unit 2 ML17241A4101999-07-16016 July 1999 Forwards FP&L Supplemental Response to NRC Request for Info Re Y2K Readiness at Nuclear Power Plants ML17309A9881999-07-0707 July 1999 Forwards Rev 5 to EPIP-03, Emergency Response Organization Notification/Staff Augmentation. Rev 5 to EPIP-03 Was Revised to Transfer EP Responsibilities from Training Manager to Protection Svcs Manager ML17241A4011999-06-30030 June 1999 Forwards Info Copy of Florida Wastewater Permit (FL0002208) (Formerly NPDES Permit) Mod,Which Was Issued by Florida Dept of Environ Protection on 990604 ML17241A3971999-06-30030 June 1999 Forwards Suppl Response to GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Movs, as Requested in 990317 Ltr ML17355A3661999-06-30030 June 1999 Forwards Florida Power & Light Topical QA Rept, Dtd June 1999.Encl I Includes Summary of Changes Made to Topical QA Rept Since 1998 ML17241A3951999-06-29029 June 1999 Provides Response to NRC Request for Info Re Y2K Readiness at Nuclear Power Plants.Gl 98-01, Y2K Readiness of Computer Sys at Nuclear Power Plants, Requested Response on Status of Facility Y2K Readiness by 990701.Disclosure Encl ML17241A3731999-06-17017 June 1999 Supplements Relief Requests 4,11 & 13 for Third ten-year ISI Interval with Info Requested During 990526 Telcon.Expedited Review Is Requested by 990730 to Avoid Negatively Impacting Upcoming St Lucie Unit 1 Refueling Outage (SL1-16) ML17241A3641999-06-14014 June 1999 Submits Supplement to Relief Request 24 with Info Requested by Nrc.In Addition Relief Request 24 Is Identical to St Lucie Unit 1 Relief Request 4 for Third ISI Interval Being Supplemented by FPL Ltr L-99-139 L-99-129, Forwards Rev 55 to Physical Security Plan,Summary of Changes & Marked Up Copy of Revised Pages.With Directions for Incorporating Rev Into Plan & Copies of Replacement Pages.Rev Withheld,Per 10CFR2.790(a)(3)1999-06-0909 June 1999 Forwards Rev 55 to Physical Security Plan,Summary of Changes & Marked Up Copy of Revised Pages.With Directions for Incorporating Rev Into Plan & Copies of Replacement Pages.Rev Withheld,Per 10CFR2.790(a)(3) ML17241A3601999-06-0707 June 1999 Forwards Correction to Annual Radiological Environ Operating Rept for CY98.Util Has Identified Transcription Error on Last Page of Attachment C of Rept,Results from Interlaboratory Comparison Program 1998 ML17241A3561999-06-0707 June 1999 Forwards Rept Containing Brief Description & Summary of SEs for Changes,Tests & Experiments Which Were Approved for Unit 3 During Period of 970526-981209 ML20195F3941999-05-27027 May 1999 FOIA Request That Memo from J Calvo to Fl Lebdon Re TIA - St Lucie,Unit 1 Environ Qualification of Woodward Governor Controls Be Placed in PDR ML17241A3461999-05-24024 May 1999 Forwards Revised Relief Request 22 to Clarify Several Areas of Relief.Nrc Action Is Requested to Be Complete by Aug 1999 to Support Planning for Spring 2000 Unit 2 Refueling Outage ML17241A3371999-05-20020 May 1999 Forwards Util Suppl to GL 95-07 Response Re pressure-locking & Thermal Binding of safety-related power-operated Gate Valves,In Response to NRC Second RAI Dtd 990225 ML17241A3391999-05-20020 May 1999 Forwards Notification of Change to Small Break LOCA ECCS Evaluation Model Used for St Lucie Unit 1.Anomaly Was Discovered & Corrected That Resulted in Reducing Calculated PCT for Limiting SBLOCA by More than 50 F ML17241A3301999-05-17017 May 1999 Forwards LER 99-004-00 Re as Found Cycle 10 Psv Setpoints Outside TS Limits,Which Occurred on 990415.Root Cause Determination Not Yet Complete.Suppl to Include Root Cause & Corrective Actions Will Be Submitted ML17309A9821999-05-10010 May 1999 Forwards Rev 36 to St Lucie Emergency Plan, Per 10CFR50.54(q).Executive Summary & Summary of Changes Incorporated by Rev,Encl ML17241A3221999-04-29029 April 1999 Provides Confirmation of NRC Staff Conclusions Re Cited & non-cited Violations in Insp Rept 50-335/98-14 & 50-389/98-14.Utils Position Re Consideration of Multiple Spurious Actuations in Event of Fire,Reiterated IR 05000335/19980141999-04-29029 April 1999 Provides Confirmation of NRC Staff Conclusions Re Cited & non-cited Violations in Insp Rept 50-335/98-14 & 50-389/98-14.Utils Position Re Consideration of Multiple Spurious Actuations in Event of Fire,Reiterated ML17229B1071999-04-28028 April 1999 Forwards 1998 Annual Environ Operating Rept for St Lucie Unit 2. Rept Includes Discussions of 5-inch Barrier Net Maint & Taprogge Condenser Tube Cleaning Sys Ball Loss,As Agreed at First Biennial Sea Turtle Meeting Held on 980120 ML17229B1051999-04-22022 April 1999 Requests That Listed Individuals Be Placed on Official Serve List for Nuclear Matl Safety & Safeguards Info Notices ML17229B1061999-04-21021 April 1999 Notifies NRC of Change in Medical Status of Licensed Operator Pf Farnsworth (Docket 55-21285,license SOP-21094). NRC Form 3996, Medical Exam Certification, Encl.Encl Withheld Per 10CFR2.790(a)(6) ML17309A9851999-04-15015 April 1999 Requests That NRC Review Denial of Appeal from Assessment of Fees Assessed in 981101 Invoice RS0062-99 & Assessment of Fees in Invoice RS0182-99 Which Was Also Denied in 990305 Ltr.Both Invoices Are for Fees Re Inspector GG Warnick ML17229B0951999-04-0808 April 1999 Requests Approval of Encl Revised Relief Request 6,in Response to 990322 Telcon with NRC & 10CFR55.55a(a)(3). Request States That Visual VT-3 Exams Will Be Conducted IAW IWA-2213 & Repairs Will Be IAW Util ASME Section IX Program ML17229B0821999-04-0707 April 1999 Requests Approval of Interim Relief Request 26 Re Repair Requirements for Class 2 ECCS Piping,Per 10CFR50.55a(a)(3) & 50.55a(g)(iii).Alternative Actions Apply Guidance of GLs 91-18 & 90-05 & ASME Code Case N-513.Evaluation,encl ML17229B0851999-04-0505 April 1999 Requests Approval of Encl Relief Request 25 Which Proposes to Use Alternative Requirements of ASME Code Case N-613 in Lieu of Requirements of ASME Section XI Figures IWB-2500-7(a) & IWB-2500-7(b).Action Requested by Aug 1999 ML17309A9791999-03-31031 March 1999 Forwards Revised EPIPs Including Rev 2 to EPIP-00,rev 2 to EPIP-09,rev 2 to EPIP-10 & Rev 10 to HP-207.Summary of Revs Listed ML17309A9761999-03-23023 March 1999 Forwards Revised Epips,Including Rev 4 to EPIP-03, Er Organization Notification/Staff Augmentation, Rev 3 to EPIP-05, Activation & Operation of OSC & Rev 14 to HP-200, HP Emergency Organization. Changes to Epips,Discussed ML17229B0691999-03-19019 March 1999 Transmits TS Pages Requested by NRC for Use in Issuance of Proposed License Amend Re SFP Storage Capacity,Per Soluble Boron Credit ML17229B0721999-03-16016 March 1999 Requests Approval of Enclosed Relief Requests 23 & 24 Re ISI Plan for Second ten-year Interval.Nrc Action Is Requested to Be Complete by Aug 1999 to Support Planning for Spring 2000 Unit 2 Refueling Outage ML17355A2631999-03-12012 March 1999 Forwards FPL Decommissioning Fund Status Repts for St Lucie, Units 1 & 2 & Turkey Point,Units 3 & 4.Rept for St Lucie, Unit 2 Provides Status of Decommissioning Funds for All Three Owners of That Unit ML17229B0481999-03-10010 March 1999 Informs That Util Delivered Matls Requested in Encl 1 of NRC Ltr by Hand on 990308,as Requested by NRC Ltr Dtd 990218 1999-09-25
[Table view] Category:UTILITY TO NRC
MONTHYEARML17223A9401990-09-13013 September 1990 Forwards Evaluation of Potential Safety Impact of Failed Control Element Assemblies on Limiting Transients for Facility ML17223A9341990-09-10010 September 1990 Forwards Addl Info Re Generic Implications & Resolution of Control Element Assembly (CEA) Failure at Facility,Per NRC Request.Description of Testing Program for Old Style CEAs in Unit 1 Core Encl L-90-315, Advises That Util Has Completed Evaluation of NUREG-0737, Item II.D.1,SER Item 81990-08-30030 August 1990 Advises That Util Has Completed Evaluation of NUREG-0737, Item II.D.1,SER Item 8 ML17223A9201990-08-28028 August 1990 Forwards Forms NIS-1 & NIS-2, Owners Rept for Inservice Insps as Required by Provisions of ASME Code Rules, Per 900725 Ltr ML17223A8911990-08-20020 August 1990 Forwards Corrected Monthly Operating Repts for Jul 1990 for St Lucie Units 1 & 2 & Summary of Operating Experience ML17348A5041990-08-17017 August 1990 Forwards fitness-for-duty Program Performance Data for Jan-June 1990 L-90-301, Discusses Generic Implications & Resolution of Control Element Assemblies Failure at Plant1990-08-16016 August 1990 Discusses Generic Implications & Resolution of Control Element Assemblies Failure at Plant ML17223A8751990-08-0909 August 1990 Responds to Violations Noted in Insp Rept 50-335/90-14. Corrective Actions:Rcs Flow Determination by Calorimetric Procedure Repeated W/Supervisor of Individual Observing & Individual Counseled by Supervisor IR 05000335/19900141990-08-0909 August 1990 Responds to Violations Noted in Insp Rept 50-335/90-14. Corrective Actions:Rcs Flow Determination by Calorimetric Procedure Repeated W/Supervisor of Individual Observing & Individual Counseled by Supervisor ML17348A4701990-07-27027 July 1990 Forwards Rept Detailing Investigative Analysis of Unsatisfactory Blind Specimen Results,Identification of Causes & Corrective Actions Taken by Lab to Prevent Recurrence,Per Unsatisfactory Performance Testing ML17223A8621990-07-25025 July 1990 Advises That NIS-1 & NIS-2 Forms,As Part of Inservice Insp Rept,Will Be Submitted by 900831 ML17348A4281990-07-25025 July 1990 Forwards Decommissioning Financial Assurance Repts for Plants,Per 10CFR50.33(k) & 50.75(b) ML17223A8631990-07-25025 July 1990 Submits Addl Info Re Implementation of Programmed Enhancements Per Generic Ltr 88-17, Loss of Dhr. All Mods for Unit 1 Completed & Operational.Mods for Unit 2 Schedule for Upcoming Refueling Outage L-90-271, Responds to NRC Ltr Re Violations Noted in Insp Repts 50-335/90-09 & 50-389/90-09.Corrective Actions:Procedural Expectation Re Hanging & Removal of Deficiency Tags Will Be Reemphasized to Personnel Generating Work Orders1990-07-20020 July 1990 Responds to NRC Ltr Re Violations Noted in Insp Repts 50-335/90-09 & 50-389/90-09.Corrective Actions:Procedural Expectation Re Hanging & Removal of Deficiency Tags Will Be Reemphasized to Personnel Generating Work Orders ML17223A8581990-07-19019 July 1990 Forwards Implementation Status of 10CFR50.62 Mod at Facility Re Requirements for Reduction of Risk from ATWS Events for Light Water Cooled Nuclear Power Plants ML17223A8491990-07-18018 July 1990 Responds to NRC Bulletin 90-001, Loss of Fill Oil in Transmitters Mfg by Rosemount. No Rosemount Transmitters Models 1153 Series B,1153 Series D & 1154 Mfg Prior to 890711 Supplied by Different Vendor ML17223A8521990-07-17017 July 1990 Forwards Addl Info Requested Re Generic Implications & Resolution of Control Element Assembly Failure at Plant.Encl Confirms Util Intent to Follow C-E Regulatory Response Group Action Program IR 05000335/19900131990-07-0909 July 1990 Responds to Violations Noted in Insp Repts 50-335/90-13 & 50-389/90-13.Corrective Actions:Maint Personnel Counseled & Aware of Importance of Verifying Design Configuration Requirements ML17223A8421990-07-0909 July 1990 Responds to Violations Noted in Insp Repts 50-335/90-13 & 50-389/90-13.Corrective Actions:Maint Personnel Counseled & Aware of Importance of Verifying Design Configuration Requirements ML17348A3881990-07-0505 July 1990 Requests Audit of NRC Records to Independently Verify Reasonableness of Charges Assessed Against Util,Per 10CFR170 Svcs ML17223A8391990-07-0303 July 1990 Forwards Results of Beach Survey Procedure & Reduction of Field Survey Data,Per Tech Spec 4.7.6.1.1.Unit 1 Updated Fsar,Section 2.4.2.2,concluded That Dune Condition Acceptable Per Tech Spec 5.1.3 ML17223A8381990-07-0202 July 1990 Requests Termination of Operator License for s Lavelle.Util Also Requests That Ltr Be Withheld (Ref 10CFR2.790) L-90-239, Forwards Rev 6 to Guard Training & Qualification Plan.Rev Withheld (Ref 10CFR73.21)1990-07-0202 July 1990 Forwards Rev 6 to Guard Training & Qualification Plan.Rev Withheld (Ref 10CFR73.21) ML17223A8371990-06-27027 June 1990 Provides Details of Implementation Plan Re Recommendations & Schedular Requirements in Generic Ltr 89-10,per 891228 Ltr.Design Basis Review of safety-related motor-operated Valves & Determination of Switch Settings in Progress ML17308A4981990-06-27027 June 1990 Responds to Generic Ltr 90-04 Re Request for Info on Status of Licensee Implementation of Generic Safety Issues Resolved W/Imposition of Requirements or Corrective Actions ML17223A8341990-06-19019 June 1990 Forwards Corrected Proposed Tech Spec Figure 3.4-2 Per 900207 Application for Amend to License NPF-16,incorporating Revised Pressure/Temp Limits & Results of Revised Low Temp Overpressure Protection Analysis Into Tech Specs ML17223A8241990-06-18018 June 1990 Forwards Revised Combined Semiannual Radioactive Effluent Release Rept for Jan-June 1988. ML17223A8271990-06-18018 June 1990 Forwards Ma Smith 900601 Ltr to WR Cunningham of EPA Requesting Mod to Plant NPDES Permit to Permit Cleaning of Facility & to Establish Discharge Limits for Chemical Cleaning Wastes ML17348A2981990-06-12012 June 1990 Forwards Rev 16 to Topical QA Rept. ML17223A6761990-05-31031 May 1990 Advises That Air Operated safety-related Components Will Perform All Design Basis Events,Per 881227 Ltr.All Actions Required by Generic Ltr 88-14 Complete for Plant ML17348A2651990-05-29029 May 1990 Submits Rept Detailing Investigative Analysis of Unsatisfactory Blind Specimen Results,Identification of Causes & Corrective Actions Taken by Lab to Prevent Recurrence,Per 10CFR26,App A.2.8(e)(4) ML17223A6741990-05-22022 May 1990 Forwards Info Re Status of 10CFR50.62 Mods to Meet ATWS Requirements as of 900515.Plant Change/Mod Package Necessary for Installing ATWS Will Be Issued by 900630.Hardware Procurement for Diverse Scram Sys Approx 90% Complete ML17223A6361990-05-0808 May 1990 Forwards Final Response to NRC Bulletin 88-010, Nonconforming Molded-Case Circuit Breakers. One Untraceable Circuit Breaker Installed in Unit 2 Qualified SPDS & Replaced W/Traceable Breaker ML17223A6281990-04-21021 April 1990 Forwards St Lucie Unit 2 Annual Environ Operating Rept, Vol 1 1989. ML17223A6081990-04-13013 April 1990 Responds to Violations Noted in Insp Repts 50-335/90-02 & 50-389/90-02.Corrective Actions:Nuclear Plant Supervisor Required to Remain in Control Room During Significant Changes in Power Operation & Preventive Maint Upgraded ML17223A6071990-04-0505 April 1990 Responds to NRC Bulletin 89-001, Failure of Westinghouse Steam Generator Tube Mechanical Plugs. Removal & Replacement of Cold Leg Side Plugs of Heat Number 3513 for Unit 1 Completed During Refueling Outage ML17308A4911990-04-0202 April 1990 Forwards Description & Summary of Safety Evaluations of Plant Changes/Mods Reportable Per 10CFR50.59.Repair &/Or Replacement of Protective Coatings on Surfaces Inside Bldg Pose No Unreviewed Safety Question ML17223A5931990-03-30030 March 1990 Forwards Status of 10CFR50.62, Requirements for Reduction of Risk from ATWS Mods at Plant as of 900315.Diverse Scram Sys Module Prototype Fabrication in Progress ML17223A5921990-03-27027 March 1990 Forwards Addl Info on Proposed License Amend Re Increased Max Allowable Resistance Temp Detector Delay Time,Per 891219 Telcon & Advises That Util Request to Increase Plant Resistance Temp Detector Response Time Remain Unchanged ML17223A5831990-03-19019 March 1990 Forwards Response to Generic Ltr 89-19, Request for Action Re Resolution of USI A-47, 'Safety Implications of Control Sys in LWR Nuclear Power Plants,' Per 10CFR50.54(f) ML17347B6191990-03-13013 March 1990 Provides Listing of Property Insurance Programs ML17223A5531990-03-0909 March 1990 Submits Results of Investigation of Error Detected in Dose Assessment During 900124 NRC Evaluated Exercise at Plant. Operator Error Caused Keyboard Hangup Requiring Computer Restart ML17223A5451990-03-0808 March 1990 Forwards Revised Tech Specs Re Steam Generator Tube Repairs, Per 890602 Telcon & Subsequent Discussions W/Nrc ML17308A4871990-03-0707 March 1990 Forwards Response to Eight Audit Questions & Licensing Bases Criteria to Resolve Station Blackout Issue.Util Currently Has Procedures to Mitigate Effects of Hurricanes & Tornados Which Meet or Exceed NUMARC 87-00 Guidelines ML17347B5881990-03-0101 March 1990 Responds to Generic Ltr 90-01, Request for Voluntary Participation in NRC Regulatory Impact Survey. Info Covers Time Spent by Key Power Plant Managers in Responding to Operational Insps & Audits ML17347B6031990-02-27027 February 1990 Requests Approval to Use Code Case N-468 at Plants ML17223A5321990-02-26026 February 1990 Forwards CEN-396 (L)-NP, Verification of Acceptability of 1-Pin Burnup Limit of 60 Mwd/Kg for St Lucie Unit 2. ML20012A0011990-02-26026 February 1990 Notifies That Followup Actions Completed on Schedule & Incorporated Into Rev 25 to Plant Physical Security Plan,Per NRC 890605 Request ML17223A5411990-02-26026 February 1990 Provides Addl Info Re Proposed License Amends Re Moderator Temp Coefficient Surveillance Requirements,Per 891026 & 900109 Telcons IR 05000335/19890241990-02-22022 February 1990 Responds to Violations Noted in Insp Repts 50-335/89-24 & 50-389/89-24.Corrective Actions:Applicable Procedures Changed to Clarify Which Spaces & Blocks Required to Be Completed on Plant Work Order & QC Supervisor Counseled 1990-09-13
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REQfj.AT INFORMATION DISTR I IO STEM (R IDS)ACCESSIQN NBR: 8708310039 DOC.DATE.: 87/08/25 NOTARIZED:
NO FACIL: 59-389 St.Lucie Planti Unit 2i Florida Power 5 Light Co.AUTH.NAME AUTHOR AFFILIATION WQODYi C.O.Florida Power Zr Light Co.REC I P.NAME.RECIPIENT AFFILIATION Document'ontrol Branch (Document Control Desk)
SUBJECT:
Forwards revised large breal LOCA EGCS performance results for limiting break size for facility which gustifies incresed steam generator tube plugging limit of up to 1430 Tubes.DISTRIBUTION CODE: A001D CQP IES RECEIVED: LTR ENCL'I ZE: TITLE: QR Submittal:
General Distribution NOTES: DOCKET 0 05000389 RECIPIENT ID CODE/NAME PD2-2 LA TOURIQNYi E COPIES RECIPIENT LTTR ENCL ID CODE/NAME 0 PD2-2 PD 1 COP IES LTTR ENCL 5 INTERNAL: ARM/DAF/LFMB
~NRR/DEBT/CEB NRR/DEBT/RSB NRR/PMAS/ILRB 01 EXTERNAL: EQSQ BRUSKE, 8 NRC PDR 0 1 1 1 1 1 1 1 1 1 1 h!RR/DEST/ADS NRR/DEBT/MTB h!RR/DOEA/TSB QQC/HDS2 RES/DE/EIB LPDR NSIC 1 1 1 1 1 1 1 0 1 1 1 1 1 1 TOTAL NUMBER OF COPIES REQUIRED: LT1R 21 ENCL 18
-4 I'I 4"1 Jttftf'4<<'<<.4<<4 I l]EK')J'r4))" I>IJ'I Ji 1 tl 4 4l'LTCJ/'<<,fl" 4 t'4 I 4)~/~IJ"J P"7, 4, g, 4 L fJ 41~1 jf 1), I I I I)~l 4 tt),'I)+I<<'4 4 I Ii'4/l y g't"j"<<44 K'1 JJ rt.*wg", 4<<)t,)t)t/4 IJ/.L I" 4<<)/I I I<<4 eely I~1 t 4'I KJ, l<<g" 41" 4 III),$4 I'/4 44~
P.X 14000, JUNO BEACH, FL 33408 0420>y9l/6, AUGUSTJ 2 5 1987 L-87-327 U.S.Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C.20555 Gentlemen:
Re: St.Lucie Unit No.2" Docket No.50-389 Lar e Break'LOCA Anal isis Florida Power&-Light Company (FPL)has reanalyzed the St.Lucie Unit 2 Large Break LOCA Analysis.The new LOCA Analysis supersedes the analysis of record submitted by FPL letter L-86-37, dated January 3, l986, which supported a steam generator tube plugging limit of l250 average length tubes per steam generator.
The attached revised Large Break LOCA Analysis justifies an increased steam generator tube plugging limit of up to l430 tubes.This re-analysis was performed using the NRC-approved June l985 version of the Combustion Engineering (CE)Large Break LOCA evaluation model.Other plant parameter changes were incorporated in this analysis in an effort to bound future cycles and possible plant changes.The results of the analysis demonstrate a peak clad temperature (PCT)of 2I07 F, a peak local clad oxidation percentage of 7.62%and a peak core wide oxidation percentage of less than 0.70%.These results demonstrate compliance with I OCFR50.46 acceptance criteria of 2200oF, l7%and I%, respectively.
The re-analysis predicts I F higher PCT than the PCT predicted in the current Reference Analysis.Although this change in PCT does not require submittal of the revised analysis to the NRC, it is being submitted to justify an increased steam generator tube plugging limit and to call to the staff's attention the use of the June l 985 version of the CE Large Break LOCA evaluation model for St.Lucie Unit 2.Very truly yours, C.O.Woo Group Vi resident Nuclear Energy COW/E JW/gc At tachment cc: Dr.J.Nelson Grace, Regional Administrator, Region II, USNRC Senior Resident Inspector, USNRC, St.Lucie Plant 8708$t 0039 BQP00+89 70826 pDR ADO~pDR P E JW I/02 I/I an FPL Group company I, I d-~~~44 4~4 v~Ilf~I'I t hl 4~4 NM A~4 II!<<A~"~d t'~l 4 44 VIVII~" I 4 I'" I It'ld 4 At!'LL'V V 4 I 4 I d lthI, N I,!~I!~1 4~'4 i'I I 4 4~'~l 4'=4 1 4 I~*<<4 I'1 4 It I d~',I f.~4 IL V~h)44<<~~I 4 LA!'v 4\I tl I!I 4 N, IU d~4 4 ('d I'4'.4 LL'I (I'4 I!4 NU~/N 4 1~I ht.dv,-I AC 4-VU t!~I 4 4'.MV!4 I>>,<<A'"';4Ih Al~~I, I!I.4 t,",..4 1!f, I 4 IIL,'l t.4,4:.l"..VV C 4 U I~let dt II 4=<<N II 4 d 4 4 I!.V!3 I~41 V 1 I 14'~NLAUlAV!4 4 I'~t'Md!c 4".!L<1~I!'4!!Vt', d 1 hl~~~I Large Break LOCA ECCS Performance Results for the Limiting Break Size for St.Lucie 2 Large 8reak LOCA ECCS Performance Introduction and Summar An ECCS performance analysis was performed for St.Lucie Unit 2 to demonstrate compliance with lOCFR50.46 which presents the NRC Acceptance Criteria for Emergency Core Cooling Systems for light water-cooled reactors (Reference 1).The analysis evaluates various plant changes utilizing the recently approved C-E June 1985 version (Reference 2)of the large break loss-of-coolant (LOCA)evaluation model.This model differs from the evaluation model applied in licensing St.Lucie 2 CycIe 3 (Reference 3).The revised large break LOCA evaluation model approved by NRC includes, changes to: (1)the cladding deformation/rupture models based on NUREG-0630, (2)the steam cooling models applied at and above the rupture location for reflood rates below one inch per second.(3)COMPERC-II allowing safety injection pump delivery before the safety injection tanks have emptied, (4)CEFLASH-4A numerical methods, (5)the stagnation properties used in the Moody break flow model, (6)the nodalization scheme used in CEFLASH-4A, and (7)the axial power shape used in the analyses.Of these changes, items (1)and (2)introduce beneficial effects on the calculated results at and above the clad rupture location.Items (3)through (6)have a negligible or small beneficial impact.The axial power shape has an adverse impact on results;however, the Cycle 3 analysis already incorporated the new shape.The current analysis complies with the conditions for NRC approval of the revised model.These conditions require that application of the model include: a break spectrum study utilizing the adverse axial power shape to deter~inc the limiting break size;a determination of whether no single failure is worse than assuming the worst single failure;and assurance that the revised steam cooling heat transfer is not allowed to exceed heat transfer predictions based on the FLECHT correlation.
A break spectrum analysis was performed to determine the limiting large break.In addition, the analysis assumed 1430 plugged tubes per steam generator, fuel parameters which bound current and expected conditions, augmentation penalty of unity, an initial safety injection tank (SIT)gas pressure of 200 psig, reduced RCS vessel and core bypass flow, and an end-of-cycle temperature coast down.The analysis justifies an allowable peak linear heat generation rate (PLHGR)of 13.0 kw/ft.'This PLHGR is equal to the existing limit for St.Lucie Unit 2.The method of analysis and detailed results which support this value are presented herein.Method of Anal sis The method of analysis is based upon C-E's June 1985 large break LOCA ECCS evaluation model which is described in References 4 through 10 and was approved by the NRC in Reference 2.The Reference Cycle, St.Lucie 2 Cycle 3 analysis (Reference 3)utilized the previously approved large break LOCA evaluation model.Except for the model and various plant parameters differ-ences described above, the method of analysis is identical to the Reference Cycle large break LOCA ECCS performance analysis.Blowdown hydraulics, refill/reflood hydraulics and hot rod temperature calculations were performed with fuel parameters which bound the current fuel cycle and expected conditions for future cycles at a reactor power level of 2754 Mwt.The blowdown hydraulics calculations'ere performed with the CEFLASH-4A code (Reference 7)while the refill/reflood hydraulics calculations were performed with the COMPERC-II code (Reference 8).The hot rod clad temperature and clad oxidation calculations were performed with the STRIKIN-II
~~
and PARCH codes (Reference 11 and 12, respectively).
Fuel performance calcu-lations were performed using the FATES-3A version of the C-E's NRC approved fuel performance code (Reference 13 and 14)with the grain size restriction as required by the NRC (Reference 15).Most of the ECCS analysis input parameters are the same as those of the Reference Cycle (Reference 3).1n particular the limiting axial shape used is the same as that used in the Reference Cycle and is consistent with the selection procedure documented in Reference 9 and approved by the NRC in Reference 2.A summary of the significant input parameters and initial conditions for the present and the reference analysis are shown in Table l.The major differences and their impact on the Peak Clad Temperature (PCT)are discussed below.This analysis accounts for steam generator U-tube plugging of up to 1430 average length tubes per generator compared to 1250 for the Reference Cycle (Reference 3).In addition, this analysis used an initial safety injection tank pressure of 200 psig and an augmentation penalty of unity compared to values of 570 psig and 1.01, respectively, for the Reference Cycle.Based on Reference 16 a favorable increase in the initial containment wall temperature of 90 F (compared to a value of 60 F for the Reference Cycle)was used.A break spectrum analysis was performed incorporating the above.To bound future fuel cycles, the limiting break determined from the break spectrum analysis was reanalyzed with a limiting set of radiation enclosure data.An assessment was made of the impact of reducing the core bypass flow such that vessel flow can be reduced (from 363,000 gpm to 359,700 gpm)while maintaining the same core flow.An evaluation of a temper ature.coastdown to 520 F at the end of the cycle was also performed.
Steam generator tube plugging increases the resistance to flow passing through the prima'ry side of the steam generator, thereby inhibiting steam venting from the core outlet plenum to the break.This reduces the refill/reflood rates and increases the peak cladding temperature.
This analysis assumes 1430 plugged tubes per generator; however, it also conservatively bounds plugging fewer than 1430 tubes in either or both steam generators, since this would reduce the flow resistance and reduce the peak clad temperature.
The reduction in the augmentation penalty results in an increase hot assembly average channel PLHGR.The hot assembly average channel influences the radiation heat transfer between the hot rod of the hot and the average rod of the hot assembly.Higher power of the average the hot assembly results in reduced heat transfer from the hot rod to surrounding rods resulting in a higher PCT.of the PLHGR assembly rod of its Reducing the SIT initial gas pressure resuIts in a slight increase in the refill time.Increased refill time means a longer period of adiabatic heat up.This consequently results in a higher PCT.Increasing the initial containment walI temperature results in an increase in reflood flow into the core.This helps to lower the peak clad temperature.
Reducing the vessel flowrate by less than 1%with a co~responding decrease in core bypass flow has a minimal impact on the PCT.Studies performed for other C-E plants have shown that~educing the vessel and core flowrates by 16%increases PCT by less than 10 F.This slight sensitivity would be further reduced if the core flow remains the same.
Temperature coastdown-.at EOC does not adversely affect PCT.Explicit physics calculations for EOC coastdown conditions confirmed that the Reference Cycle core parameters (e.ges axial and radial power distributfons, and PLHGR)conservatively bound EOC coastdown conditions.
The only adverse impact of EOC coastdown is the effect which the reduced coolant temperature has on the blowdown hydraulics.
However, previous studies have shown this to be a small effect, and one which is offset by the significantly lower fuel stored energy at EOC burnup relative to the limiting burnup used in the current analysiz.Results Table 2 provides the results of the break spectrum.Double-Ended Slot at Pump Discharge (DES/PD)b~eaks were judged to be non-limiting based on fuel average temperatures at TAD (Time of Annulus Downflow)which defines the end of blowdown portion of the transient.
As expected, the break spectrum analysis determined the 0.6 DEG/PD break to be the limiting break.The previously approved evaluation model demonstrated a weak sensitivity to PCT due to the va~ious break sizes.This is also true for the June 1985 evaluation model as shown in Table 2.However, the new leak flow model incorporated in the June 1985 evaluation model introduces a shift fn the limiting break size due to the small change in the leak flow characteristics.
This is consistent with other C-E plants which utilized the June 1985 evaluation model.Table 3 presents the results of the limiting break reanalyzed with a conservative set of radiation enclosure data.Table 4 presents a list of the significant parameters displayed graphically for the break.The results of the evaluation confirm that 13.0 kw/ft is an acceptable value for the PLHGR in the present analysis.The peak clad temperature and maximum local and core wide clad oxidation values as shown in Table 3, are well below the 10CFR50.46 acceptance limits of 2200 F, 17%and 1%respectively.
The 0.6 OEG/PD produced the highest clad temperature of 2107 F and a peak local oxidation of 7.62%compared to the acceptance criteria of 2200 F and 17%0 respectively.
The 0.6 OEG/PD also resulted in the highest core wide oxidation of less than 0.7%which is well below the 1%acceptance criteria.A review of the the effects of initial operating conditions on these results was performed.
It was determined that over the ranges of operating conditions allowed by the Technical Specification, a PLHGR of 13.0 kw/ft is an acceptable limit.Conclusions The results of the ECCS performance evaluation for the present analysis for St.Lucie Unit 2 demonstrated a peak clad temperature of 2107 F, a peak local clad oxidation percentage of 7.62%, and a peak core wide oxidation percentage of less than 0.7%compared to the acceptance criteria of 2200 F, 17%and 1%~espectively.
Therefore, operation of St.Lucie Unit 2 at a core power level of 2754 Neth (102%of 2700 Nwth)and a PLHGR of 13.0 kw/ft is in conformance with 10CFR50.46.
Table 1 St.Lucie-Unit 2 Si nificant Parameters and Initial Conditions For Break S ectrum Stud (1)Parameters Reference~Ccl e Present~Anal aia Core Power at 102%of Nominal (MMt)Core Average Linear Heat Rate at 102%of Nominal (kw/ft)~Peak Linear Heat Generation Rate (PLHGR)Hot Assembly, Hot Channel (kw/ft)PLHGR Hot Assembly, Average Channel (kw/ft)Core Inlet Temperature (OF)Core Outlet Temperature
('F)Vessel Flow (10 ibm/hr)Core Flow((10 ibm/hr)Gap conductance at PLHGR'Btu/hr-ft'-'F)
(2)Fuel Centerline Temperature at PLHGR (F)Fuel Average Temperature at PLHGR (4F)Hot Rod Gas Pressure (psia)Hot Rod Burnup (NMO/MTU)Number of Tubes Plugged per Steam Generator Augmentation Factor Safety Injection Tank (SIT)gas pressure (psig)Initial Containment Temperature (F)2754 4.90 13.0 11.45 552 603.8 136.1 131.1 1416 3228 2078 1118 1038 1250 1.01 570 60 2754 4.90 13.0 11.57 552 603.8 136.1 131.1 1460 3296 2102 1118 1038 1430 1.00 200 90 Hot rod radiation enclosure, and core and vessel flowrates weie not changed for the break spectrum study.Their impact is subsequently evaluated based on the limiting break size determined by this study.STRIKIN-II values at hot rod burnup which yields highest peak clad temperature.
Table 2 St.Lucie-Unit 2 Break Spectrum-Results Break Size TAO()Time, Seconds Fuel Average Temperature at TAO F Peak Clad Temperature OF 0.8 OEG/PO 0.8 OEG/PO 0.4 OEG/PO 0.8 OES/PO 0.6 OES/PO 0.4 OES/PO 20.0 22.6 27.6 18.0 20.2 25.4 1077 1123 1074 991 992 984 2061 2065 2034 (b)(b)(b)(a)Time of annulus downflow-end of blowdown.(b)Slot breaks were judged to be non-limiting based on their significantly lower fuel ave~age temperature at TAO.and because the r eflood heat transfer applicable to the slot breaks is no worse than'he conservative heat transfer applied to the guillotine breaks.Oouble-Ended Guillotine at Pump Oischarge.
Oouble-Ended Slot at Pump Oischarge.
Table 3 St.Lucie-Unit 2 tnitial Conditions and Results for Limitin Break Size 0.6 OEG/PO Reference~Cnl e Pt esent~Anal sis Initial Conditions Peak Linear Heat Generation Rate (kw/ft)Radiation Enclosure x-factor Peak Linear Heat Generation Rate (PLHGR)Hot Assembly, Average Channel (kw/ft)13.0 2.19 11.45 13.0 2.00 11.80 Results Peak Clad Temper ature ('F)Time of Peak Clad Temperature (Seconds)Time of Clad Rupture (Seconds)Peak Local Clad Oxidation (%)Total Core-Wide Clad Oxidation (%)2106 259 55.85 16.12~0.70 2107 266 44.74 7.62+0.70'Lower x-factor indicates flatter power distribution in the vicinity of the hot rod.
Table a St.Lucie Unit 2 Variables Plotted as a Function ot Time for the Limitin Lar e Break l/ariaaie Fiaure Number Cor e Power Pressure in Center Hot Assembly Node Leak Flow Hot Assembly Flow (below hot spot)Hot Assembly Flow{above hot spot)Hot Assembly guality Containment Pressure Mass Added to Core During Reflood Peak Claa Temperature Hot Spot Gap Conductance Peak Local Clad Oxidation Temperature of Fuel Centerline, Fuel Average, Clad and Coolant at Hottest Node Hot Spot Heat Transfer Coefficient Hot Rod internal Gas Pressure 1 2 3 4 5 6 7 8 9 10 11 12 13 14
Reference:
l.Acceptance Crfterfa for Emergency Core Cooling Systems for Light Mater Cooled Nuclear Power Reactors, Federal Register, Vol.39, No.3, January 4, 1974.2.Letter, D.N.Crutchfield (NRC)to A.E.Scherer (C-E),"Safety Eval-uation of Combustion Engineerfng ECCS Large Break Evaluation Model and Acceptance for Referencfng of Related Licensing Topical Reports", July 31, 1986.3.Letter C.O.Woody (FPL)to F.J.Miraglia (NRC),"St.Lucie Unit No.2 Docket No.50-389 CE Large Break LOCA Analysis", January 3, 1986, L-86-37.4.Letter, A.E.Scherer (C-E)to J.R.Miller (NRC),'LD-81-095, Enclosure 1-P,"C-E ECCS Eva1uatfon Nodel Flow Slockage Analysis".(Proprietary), December 15, 1981.5.Letter, A.E.Scherer (C-E)to C.0.Thomas (NRC), LO-86-027,"Responses
-to questions on C-E's Revfsed Evaluation Model for Large Break LOCA Analysis", (Proprietary), June 17, 1986.6.Letter, A.E.Scherer (C-E)to C.0.Thomas (NRC), L0-85-032," Revision to C-E Model for Large Break LOCA Analysis", July 3, 1985.7.CENPO-133, Supplement 5-P,"CEFLASH-4A.
A FORTRAN77 Ofgftal Computer Program for Reactor Slowdown Analysis", June 1985.8.CENPD 134, Supplement 2-P,"CONPERC-ll, A Program for Emergency Reffll-Reflood of the Core", June 1985.9.CENPO-132-P.
Supplement 3-P,"Calculative Nethods for the C-E Large Break LOCA Evaluation Model for the Analysis of C-E and M Designed NSSS", June 1985.
1O.Letter, A.E.Scherer (C-E)to C.O.Thomas (NRC), L0-85-050, Enclosure,"Supplemental Material for Inclusion in CENPO-132.
Supplement 3-P", (Proprietary), November 5, 1985.11.CENP0-135, Supplement 2-P,"STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program (Modifications)", February 1975.CENP0-135-P, Supplement 4-P,"STRIKIN-II.
A Cylindrical Geometry Fuel Rod Heat Transfer Program", August 1976.CENP0-135-P, Supplement 5-P,"STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program", April 1977.12.CENP0-138-P, and Supplement 1-P."PARCH, A FORTRAN IV Oigital Program to Evaluate Pool Boiling, Axial Rod and Coolant Heatup", February 1975.CENPO-138 Supplement 2-P,"PARCH-A FORTRAN-IV Qigital Program to Evaluate Pool Boiling, Axial Rod and Coolant Heatup", January 1977.13.CENP0-139-P-A,"C-E Fuel Evaluation Model Topical Report", July 1974.14.CEN-161(B)-P,"Improvements to Fuel Evaluation Model Topical Report", July 1981.15.Letter from R.A.Clark (NRC)to A.E.Lundvall, Jr.(BGSE), dated March 31, 1983.16.Letter, J.L.Perryman (FPSL)to E.L.Trapp (C-E), FRN-86-404,"St.Lucie 2 Large Break LOCA Reevaluation", November 10, 1986.3183-4 3-4-87 FIgura 1 ST.LUCIE UNIT 2 O.B x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG CORE POSER 1 2001 1 0000 80QQ 6000 I~O 4000 20GO 000q a a a a a a a a a a C4 a a a a P)a a a a a a a iQ T lNK IN SEC Figllf8 2 ST.LUCIE UNIT 2 0.8 x OOUBLE ENDED GUILLOTINE BREAK IN PUMP OISCHARGE LEG PRESSURE IN CENTER HOT ASSEMBLY NOOE 2400 0 2000 0 16GO 0 g 1200.G 8GG G 4GG.G CD CD CD CD CD CD CQ CD CD CD CD OD CD CD CU CD CD CD CD T I l1E l N SEC Figure 3~ST.LUCIE UNIT 2 O.B x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG LEAK FLON iZGOQO.-PUMP SIDE---REACTOR VESSEL SIDE 100000 8GQQO.CQ GOGGO.4GOGG 2GOGO C)C)CD O CD CQ C)CD C)CD C)CQ C)O O CV CD C)CD T I f1E I'!4 SEC FIgUre 4 ST.LUCIE UNIT 2 0.8 x OOUBLE ENOED GUILLOTINE BREAK IN PUMP OISCHARGE LEG HOT ASSEMBLY FLOW, BELOW HOT SPOT 30 000 20 000 iO.GGG CO Kl GGG-IG GGG CD-20 GGG-30 GGO CD CD CD C)CD (0 C)CD CD CU C)CD C)Ca CD CD CD CU CD CI CD CD
,~~F1gUfB 5 ST.LUCIE UNIT 2 0.8 x DOUBLE ENDED GUILLOTINE 8REAK IN PUMP DISCHARGE LEG HOT ASSEMBLY FLOW, ABOVE HOT SPOT 30 000 20 000 10 GGG CCl GGG-LG GGG C)-20.0GG-3G GGG CD CD CD CD CD CD (D CD CD CD Q3 CD CD CD C4 CD CD CD CD T iiIK t N SEC Figure 8 ST.LUClf UNIT 2 O.B x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG HOT ASSEMBLY OUAUTY NODE 13, BELOW HOTTEST REGION NQGE 14, AT HOTTEST REGION NOOE 1$, ABOVE HOTTEST REGION j 0000 GGGG GGGG II il jl ll Ii///////'//j I I t I/" i 4GGG//2GOG GGGG CD CD CD CD C)>@~do 0 hill pv,~c no.