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Category:CORRESPONDENCE-LETTERS
MONTHYEARML20217F9941999-10-15015 October 1999 Discusses FPC 970819 Request for Temporary Relief from ASME Code Section XI Requirements to Repair ASME Class 3 Nuclear Service & Decay Heat Sea Water System Piping.Forwards SE Containing Results of Staff Review ML20217J5171999-10-13013 October 1999 Informs That on 990930,NRC Staff Completed mid-cycle PPR of Plant,Unit 3 & Did Not Identify Any New Areas That Warranted More than Core Insp Program.Previously Planned Regional Initiative Insp of safety-related Mod Will Be Performed 3F1099-14, Requests Copy of NRC Radtrad Code & Copy of User Instructions.Conditions for Receiving Code Listed1999-10-13013 October 1999 Requests Copy of NRC Radtrad Code & Copy of User Instructions.Conditions for Receiving Code Listed 3F1099-11, Provides Info on Requested Minor Permit Mod of Encl NPDES Permit.No New Regulatory Commitments Are Made1999-10-0404 October 1999 Provides Info on Requested Minor Permit Mod of Encl NPDES Permit.No New Regulatory Commitments Are Made ML20212L0771999-10-0404 October 1999 Forwards SER Accepting Licensee Relief Requests 98-012 Through 98-018 Involving Containment Insps at Crystal River Unit 3 Pursuant to 10CFR50.55a(a)(3)(i) & 10CFR50.55a(a)(3)(ii) ML20217D6551999-10-0101 October 1999 Requests That Natl Communication Sys Arrange for Licensee Participation in Government Emergency Telecommunications Service,Per NRC Info Notice 99-025 ML20212J8481999-10-0101 October 1999 Forwards Safety Evaluation Re Second 10 Yr Interval ISI Program Requests for Relief 98-009-II.Reliefs Granted for 98-009-II,Parts B & C & 98-010-II & 98-011-II 3F0999-03, Notifies of Approved Change to NPDES Permit Applicable to Crystal River Unit 3 IAW Section 3.2.3 of Epp.Proposed Change Was Approved on 990914 by State of Fl & Provided in Attachment1999-09-27027 September 1999 Notifies of Approved Change to NPDES Permit Applicable to Crystal River Unit 3 IAW Section 3.2.3 of Epp.Proposed Change Was Approved on 990914 by State of Fl & Provided in Attachment 3F0999-18, Notifies NRC That Due Date for Commitment Common to Ltrs 980115 & 980209 Will Be Extended.Revised Completion Date for Cable Ampacity Project Is 0003311999-09-27027 September 1999 Notifies NRC That Due Date for Commitment Common to Ltrs 980115 & 980209 Will Be Extended.Revised Completion Date for Cable Ampacity Project Is 000331 ML20212F7251999-09-23023 September 1999 Discusses Staff Review of Util 980330 Response,As Suppl on 990514,to GL 97-06, Degradation of SG Internals. Staff Concludes That Licensee Responses to GL Provide Reasonable Assurance That Condition of SG Internals Acceptable ML20212F7331999-09-23023 September 1999 Discusses Util Licensing Action for GL 98-01, Year 2000 Readiness of Computer Systems at Nuclear Power Plants. NRC Ack Efforts Util Completed to Date in Preparing Crystal River,Unit 3 for Y2K Transition 3F0999-20, Forwards Summary Re Justification to Defer USI A-46 Commitment,Per Work Needed to Resolve GL 87-03, Verification of Seismic Adequacy of Mechanical & Electrical Equipment in Operating Reactors,Usi A-461999-09-21021 September 1999 Forwards Summary Re Justification to Defer USI A-46 Commitment,Per Work Needed to Resolve GL 87-03, Verification of Seismic Adequacy of Mechanical & Electrical Equipment in Operating Reactors,Usi A-46 ML20212E6741999-09-21021 September 1999 Forwards Safety Evaluation Accepting Proposed EAL Changes Submitted by ,As Supplemented by 981120,990713 & 0831 Ltrs,Incorporating Guidance in NUMARC/NESP-007,Rev 2, Methodology for Development of Eals 3F0999-01, Forwards FPC Crystal River Unit 3 Plant Reference Simulator Four-Year Simulator Certification Rept Sept 1995-Sept 1999, Per 10CFR55.45(b)(5)(ii) & 10CFR55.45(b)(5)(iv)1999-09-17017 September 1999 Forwards FPC Crystal River Unit 3 Plant Reference Simulator Four-Year Simulator Certification Rept Sept 1995-Sept 1999, Per 10CFR55.45(b)(5)(ii) & 10CFR55.45(b)(5)(iv) 3F0999-19, Provides Clarification of Minor Inconsistency Identified During Review of NRC SE for Plant Third 10-year Interval Inservice Insp Program Plan & Associated Requests for Relief1999-09-15015 September 1999 Provides Clarification of Minor Inconsistency Identified During Review of NRC SE for Plant Third 10-year Interval Inservice Insp Program Plan & Associated Requests for Relief ML20212F3141999-09-13013 September 1999 Forwards Insp Rept 50-302/99-05 on 990704-0814.Violations Noted,But Being Treated as non-cited Violations ML20211L9081999-09-0303 September 1999 Informs of Completion of Licensing Action for GL 92-08, Thermo-Lag 330-1 Fire Barriers, Dtd 921217,for Crystal River Unit 3 ML20211Q7581999-09-0101 September 1999 Forwards Summary of 990812-13 Training Managers Conference in Atlanta,Georgia Re Recent Changes to Operator Licensing Program.List Conference Attendees,Copy of Presentation Slides & List of Participant Questions Encl 3F0899-23, Provides Addl Info in Response to Several NRC Staff Questions Needed to Complete Review of Request to Adopt NEI 97-03,Draft Final Rev 3, Methodology for Development of Eals1999-08-31031 August 1999 Provides Addl Info in Response to Several NRC Staff Questions Needed to Complete Review of Request to Adopt NEI 97-03,Draft Final Rev 3, Methodology for Development of Eals ML20211G7111999-08-30030 August 1999 Modifies Approval of 980521 Request for Exception to 10CFR50.4(b)(6) & Grants Util Approval to Submit Copies of Future Updates to FSAR as Listed ML20211G7031999-08-30030 August 1999 Informs of Approval of Util 980521 Request for Exception to 10CFR50.4(b)(6),allowing Util to Submit Updates to Plant Ufsar.Ltr Modifies That Approval & Grants Util Approval 3F0899-07, Provides Formal Notification to NRC of FPC Plans Relative to Renewal of Crystal River Unit 3,FOL DPR-72.FPC Plans to Submit Application for License Renewal by End of 20021999-08-27027 August 1999 Provides Formal Notification to NRC of FPC Plans Relative to Renewal of Crystal River Unit 3,FOL DPR-72.FPC Plans to Submit Application for License Renewal by End of 2002 ML20212C1351999-08-27027 August 1999 Requests Withholding of Proprietary Version of Enhanced Spent Fuel Storage Project Engineering Input 3F0899-20, Forwards six-month fitness-for-duty Program Performance Data for Period 990101-990630,IAW 10CFR26.711999-08-26026 August 1999 Forwards six-month fitness-for-duty Program Performance Data for Period 990101-990630,IAW 10CFR26.71 3F0899-05, Forwards Response to NRC 990716 RAI Re Proposed Alternate Repair Criteria for Axial Tube End crack-like Indications in Crystal River Unit 31999-08-20020 August 1999 Forwards Response to NRC 990716 RAI Re Proposed Alternate Repair Criteria for Axial Tube End crack-like Indications in Crystal River Unit 3 3F0899-17, Submits Relief Request 99-0001-RR,seeking NRC Approval for Evaluation Performed by Util on through-wall Flaw in Nuclear Svc & Decay Heat Sea Water (RW) Sys,Per Guidance of GL 90-051999-08-19019 August 1999 Submits Relief Request 99-0001-RR,seeking NRC Approval for Evaluation Performed by Util on through-wall Flaw in Nuclear Svc & Decay Heat Sea Water (RW) Sys,Per Guidance of GL 90-05 3F0899-16, Informs That Licensee Is Requesting State of Fl Dept of Environ Protection to Make Changes in Plant NPDES Permit to Modify Conditions on Use of Biocide in Instrument Air Compressor Sys.No New Commitments Are Made in Submittal1999-08-19019 August 1999 Informs That Licensee Is Requesting State of Fl Dept of Environ Protection to Make Changes in Plant NPDES Permit to Modify Conditions on Use of Biocide in Instrument Air Compressor Sys.No New Commitments Are Made in Submittal 3F0899-02, Forwards Rev 2 to Cycle 11 COLR IAW Plant TS Section 5.6.2.18.Rev 1 of Cycle 11 COLR Was Not Submitted Due to Administrative Error.Changes Made in Rev 1 Listed & Incorporated in Encl Rev 21999-08-16016 August 1999 Forwards Rev 2 to Cycle 11 COLR IAW Plant TS Section 5.6.2.18.Rev 1 of Cycle 11 COLR Was Not Submitted Due to Administrative Error.Changes Made in Rev 1 Listed & Incorporated in Encl Rev 2 3F0899-06, Forwards Monthly Operating Rept for July 1999 for Crystal River,Unit 3,per ITS 5.7.1.2.Revised Repts for Apr,May & June 1999,also Encl.Data on Line Item 6 Updated to Agree with More Accurate Computer Point That Measures Value1999-08-13013 August 1999 Forwards Monthly Operating Rept for July 1999 for Crystal River,Unit 3,per ITS 5.7.1.2.Revised Repts for Apr,May & June 1999,also Encl.Data on Line Item 6 Updated to Agree with More Accurate Computer Point That Measures Value 05000302/LER-1997-038, Forwards LER 97-038-01,IAW 10CFR50.73(c).Submittal Also Provides Notification That Commitment Common to LER 97-038-00 & Reply to NOV 50-302/97-16 Has Been Revised & Revised Commitment Has Been Implemented1999-08-13013 August 1999 Forwards LER 97-038-01,IAW 10CFR50.73(c).Submittal Also Provides Notification That Commitment Common to LER 97-038-00 & Reply to NOV 50-302/97-16 Has Been Revised & Revised Commitment Has Been Implemented ML20210Q4511999-08-0505 August 1999 Informs That NRC Plans to Administer Generic Fundamentals Exam Section of Written Operator Licensing Exam on 991006 ML20210P0741999-08-0505 August 1999 Forwards SE Accepting Licensee 980416 & 1130 Ltrs Re Third 10-year Interval ISI Program Plan & Associated Requests for Relief for Plant,Unit 3 3F0799-30, Forwards List of Licensing Actions Currently Estimated for Fys 2000 & 2001,in Response to Administrative Ltr 99-02,dtd 9906031999-07-29029 July 1999 Forwards List of Licensing Actions Currently Estimated for Fys 2000 & 2001,in Response to Administrative Ltr 99-02,dtd 990603 ML20210G8551999-07-27027 July 1999 Forwards Insp Rept 50-302/99-04 on 990523-0703.One Violation Identified & Being Treated as Noncited Violation 3F0799-09, Provides Response to NRC 990625 Telcon RAI Re Util Use of Relief Request 98-009-II for Plant ASME Section XI, Inservice Insp Second Interval.Ltr Established No New Regulatory Commitments1999-07-19019 July 1999 Provides Response to NRC 990625 Telcon RAI Re Util Use of Relief Request 98-009-II for Plant ASME Section XI, Inservice Insp Second Interval.Ltr Established No New Regulatory Commitments ML20209H5211999-07-16016 July 1999 Forwards Request for Addl Info Re Licensee Proposed Alternate Repair Criteria for Axial Tube End crack-like Indications in CR-3 once-through Steam Generators in Order to Complete Review ML20209G3231999-07-15015 July 1999 Forwards Biological Opinion Issued by Natl Marine Fisheries (NMFS) of Dept of Commerce.Nmfs Concluded That Operation of Cw Intake Sys of Crystal River Not Likely to Jeopardize Existence of Species Listed in Biological Opinion ML20209G3481999-07-15015 July 1999 Transmits Natl Marine Fisheries Svc (NMFS) Biological Opinion Based on Review of Continued Use of Cw Intake Sys at Crystal River Energy Complex.Concludes That Continued Use of Cw Intake Sys Not Likely to Adversely Affect Gulf Sturgeon 3F0799-21, Forwards Copy of Revised NPDES Permit IAW Section 3.2.3 of Unit 3 Environ Protection Plan,Per 990430 Request to Allow Use of Biocide in Station Air Compressor Cooling Sys. Wastewater Permit FL0000159 Issued 990630 Also Encl1999-07-14014 July 1999 Forwards Copy of Revised NPDES Permit IAW Section 3.2.3 of Unit 3 Environ Protection Plan,Per 990430 Request to Allow Use of Biocide in Station Air Compressor Cooling Sys. Wastewater Permit FL0000159 Issued 990630 Also Encl 3F0799-05, Requests Exemption from 10CFR70.51, Matl Balance,Inventory & Records Requirements, as It Relates to 10CFR70.51(d) Re Physical Inventory of SNM for Crystal River Unit 3.Detailed Justification for Request,Encl1999-07-14014 July 1999 Requests Exemption from 10CFR70.51, Matl Balance,Inventory & Records Requirements, as It Relates to 10CFR70.51(d) Re Physical Inventory of SNM for Crystal River Unit 3.Detailed Justification for Request,Encl 3F0799-25, Forwards License Renewal Applications for Four Individuals, IAW 10CFR55.57.Without Encl1999-07-14014 July 1999 Forwards License Renewal Applications for Four Individuals, IAW 10CFR55.57.Without Encl 3F0799-26, Provides Notice of Change in Status for Senior Operator,Iaw 10CFR50.74(a).RD Demontfort,License Number SOP 20528-2,has Been Reassigned & No Longer Requires License Effective 9907301999-07-14014 July 1999 Provides Notice of Change in Status for Senior Operator,Iaw 10CFR50.74(a).RD Demontfort,License Number SOP 20528-2,has Been Reassigned & No Longer Requires License Effective 990730 3F0799-22, Provides Update & Rev to Submittal Made by Util Ltr with Regard to EAL Classification Methodology for Unit 3.Reponses to NRC Staff Questions Provided as Attachment D to Ltr & Reflects Discussions Held1999-07-13013 July 1999 Provides Update & Rev to Submittal Made by Util Ltr with Regard to EAL Classification Methodology for Unit 3.Reponses to NRC Staff Questions Provided as Attachment D to Ltr & Reflects Discussions Held 3F0799-03, Forwards Rev 5-0 to Safeguards Contingency Plan,Replacing Current Rev to Safeguards Contingency Plan,Rev 4,in Entirety.Rev Withheld,Per 10CFR73.211999-07-0808 July 1999 Forwards Rev 5-0 to Safeguards Contingency Plan,Replacing Current Rev to Safeguards Contingency Plan,Rev 4,in Entirety.Rev Withheld,Per 10CFR73.21 3F0799-02, Submits Rev 7-3 to Physical Security Plan,Replacing Current Rev to CR-3 Physical Security Plan,Rev 7-2,in Entirety.Rev Withheld,Per 10CFR73.211999-07-0808 July 1999 Submits Rev 7-3 to Physical Security Plan,Replacing Current Rev to CR-3 Physical Security Plan,Rev 7-2,in Entirety.Rev Withheld,Per 10CFR73.21 ML20196L1261999-07-0707 July 1999 Discusses Closeout of TAC MA0538 Re License Response to RAI Re GL 92-01,Rev 1,Suppl 1, Rv Structural Integrity, Issued on 950519 to Plant,Unit 3 3F0799-10, Submits Copy of Historical NPDES Permit Rev That Was Made in 1997 Re Use of Biocide at Crystal River Unit 31999-07-0707 July 1999 Submits Copy of Historical NPDES Permit Rev That Was Made in 1997 Re Use of Biocide at Crystal River Unit 3 ML20196J4991999-07-0101 July 1999 Advises That Info Contained in ,Which Included TR BAW-2346P,will Be Withheld from Public Disclosure,Per 10CFR2.790(b)(5) ML20209C0811999-06-25025 June 1999 Forwards Overdue Controlled Document Transmittals for Listed Documents 3F0699-06, Submits Final Response to GL 98-01,Suppl 1 Re Year 2000 Readiness of Nuclear Power Plants.Year 2000 Readiness Disclosure for Crystal River,Unit 3,encl1999-06-23023 June 1999 Submits Final Response to GL 98-01,Suppl 1 Re Year 2000 Readiness of Nuclear Power Plants.Year 2000 Readiness Disclosure for Crystal River,Unit 3,encl 1999-09-03
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEAR3F1099-14, Requests Copy of NRC Radtrad Code & Copy of User Instructions.Conditions for Receiving Code Listed1999-10-13013 October 1999 Requests Copy of NRC Radtrad Code & Copy of User Instructions.Conditions for Receiving Code Listed 3F1099-11, Provides Info on Requested Minor Permit Mod of Encl NPDES Permit.No New Regulatory Commitments Are Made1999-10-0404 October 1999 Provides Info on Requested Minor Permit Mod of Encl NPDES Permit.No New Regulatory Commitments Are Made 3F0999-03, Notifies of Approved Change to NPDES Permit Applicable to Crystal River Unit 3 IAW Section 3.2.3 of Epp.Proposed Change Was Approved on 990914 by State of Fl & Provided in Attachment1999-09-27027 September 1999 Notifies of Approved Change to NPDES Permit Applicable to Crystal River Unit 3 IAW Section 3.2.3 of Epp.Proposed Change Was Approved on 990914 by State of Fl & Provided in Attachment 3F0999-18, Notifies NRC That Due Date for Commitment Common to Ltrs 980115 & 980209 Will Be Extended.Revised Completion Date for Cable Ampacity Project Is 0003311999-09-27027 September 1999 Notifies NRC That Due Date for Commitment Common to Ltrs 980115 & 980209 Will Be Extended.Revised Completion Date for Cable Ampacity Project Is 000331 3F0999-20, Forwards Summary Re Justification to Defer USI A-46 Commitment,Per Work Needed to Resolve GL 87-03, Verification of Seismic Adequacy of Mechanical & Electrical Equipment in Operating Reactors,Usi A-461999-09-21021 September 1999 Forwards Summary Re Justification to Defer USI A-46 Commitment,Per Work Needed to Resolve GL 87-03, Verification of Seismic Adequacy of Mechanical & Electrical Equipment in Operating Reactors,Usi A-46 3F0999-01, Forwards FPC Crystal River Unit 3 Plant Reference Simulator Four-Year Simulator Certification Rept Sept 1995-Sept 1999, Per 10CFR55.45(b)(5)(ii) & 10CFR55.45(b)(5)(iv)1999-09-17017 September 1999 Forwards FPC Crystal River Unit 3 Plant Reference Simulator Four-Year Simulator Certification Rept Sept 1995-Sept 1999, Per 10CFR55.45(b)(5)(ii) & 10CFR55.45(b)(5)(iv) 3F0999-19, Provides Clarification of Minor Inconsistency Identified During Review of NRC SE for Plant Third 10-year Interval Inservice Insp Program Plan & Associated Requests for Relief1999-09-15015 September 1999 Provides Clarification of Minor Inconsistency Identified During Review of NRC SE for Plant Third 10-year Interval Inservice Insp Program Plan & Associated Requests for Relief 3F0899-23, Provides Addl Info in Response to Several NRC Staff Questions Needed to Complete Review of Request to Adopt NEI 97-03,Draft Final Rev 3, Methodology for Development of Eals1999-08-31031 August 1999 Provides Addl Info in Response to Several NRC Staff Questions Needed to Complete Review of Request to Adopt NEI 97-03,Draft Final Rev 3, Methodology for Development of Eals ML20212C1351999-08-27027 August 1999 Requests Withholding of Proprietary Version of Enhanced Spent Fuel Storage Project Engineering Input 3F0899-07, Provides Formal Notification to NRC of FPC Plans Relative to Renewal of Crystal River Unit 3,FOL DPR-72.FPC Plans to Submit Application for License Renewal by End of 20021999-08-27027 August 1999 Provides Formal Notification to NRC of FPC Plans Relative to Renewal of Crystal River Unit 3,FOL DPR-72.FPC Plans to Submit Application for License Renewal by End of 2002 3F0899-20, Forwards six-month fitness-for-duty Program Performance Data for Period 990101-990630,IAW 10CFR26.711999-08-26026 August 1999 Forwards six-month fitness-for-duty Program Performance Data for Period 990101-990630,IAW 10CFR26.71 3F0899-05, Forwards Response to NRC 990716 RAI Re Proposed Alternate Repair Criteria for Axial Tube End crack-like Indications in Crystal River Unit 31999-08-20020 August 1999 Forwards Response to NRC 990716 RAI Re Proposed Alternate Repair Criteria for Axial Tube End crack-like Indications in Crystal River Unit 3 3F0899-16, Informs That Licensee Is Requesting State of Fl Dept of Environ Protection to Make Changes in Plant NPDES Permit to Modify Conditions on Use of Biocide in Instrument Air Compressor Sys.No New Commitments Are Made in Submittal1999-08-19019 August 1999 Informs That Licensee Is Requesting State of Fl Dept of Environ Protection to Make Changes in Plant NPDES Permit to Modify Conditions on Use of Biocide in Instrument Air Compressor Sys.No New Commitments Are Made in Submittal 3F0899-17, Submits Relief Request 99-0001-RR,seeking NRC Approval for Evaluation Performed by Util on through-wall Flaw in Nuclear Svc & Decay Heat Sea Water (RW) Sys,Per Guidance of GL 90-051999-08-19019 August 1999 Submits Relief Request 99-0001-RR,seeking NRC Approval for Evaluation Performed by Util on through-wall Flaw in Nuclear Svc & Decay Heat Sea Water (RW) Sys,Per Guidance of GL 90-05 3F0899-02, Forwards Rev 2 to Cycle 11 COLR IAW Plant TS Section 5.6.2.18.Rev 1 of Cycle 11 COLR Was Not Submitted Due to Administrative Error.Changes Made in Rev 1 Listed & Incorporated in Encl Rev 21999-08-16016 August 1999 Forwards Rev 2 to Cycle 11 COLR IAW Plant TS Section 5.6.2.18.Rev 1 of Cycle 11 COLR Was Not Submitted Due to Administrative Error.Changes Made in Rev 1 Listed & Incorporated in Encl Rev 2 05000302/LER-1997-038, Forwards LER 97-038-01,IAW 10CFR50.73(c).Submittal Also Provides Notification That Commitment Common to LER 97-038-00 & Reply to NOV 50-302/97-16 Has Been Revised & Revised Commitment Has Been Implemented1999-08-13013 August 1999 Forwards LER 97-038-01,IAW 10CFR50.73(c).Submittal Also Provides Notification That Commitment Common to LER 97-038-00 & Reply to NOV 50-302/97-16 Has Been Revised & Revised Commitment Has Been Implemented 3F0899-06, Forwards Monthly Operating Rept for July 1999 for Crystal River,Unit 3,per ITS 5.7.1.2.Revised Repts for Apr,May & June 1999,also Encl.Data on Line Item 6 Updated to Agree with More Accurate Computer Point That Measures Value1999-08-13013 August 1999 Forwards Monthly Operating Rept for July 1999 for Crystal River,Unit 3,per ITS 5.7.1.2.Revised Repts for Apr,May & June 1999,also Encl.Data on Line Item 6 Updated to Agree with More Accurate Computer Point That Measures Value 3F0799-30, Forwards List of Licensing Actions Currently Estimated for Fys 2000 & 2001,in Response to Administrative Ltr 99-02,dtd 9906031999-07-29029 July 1999 Forwards List of Licensing Actions Currently Estimated for Fys 2000 & 2001,in Response to Administrative Ltr 99-02,dtd 990603 3F0799-09, Provides Response to NRC 990625 Telcon RAI Re Util Use of Relief Request 98-009-II for Plant ASME Section XI, Inservice Insp Second Interval.Ltr Established No New Regulatory Commitments1999-07-19019 July 1999 Provides Response to NRC 990625 Telcon RAI Re Util Use of Relief Request 98-009-II for Plant ASME Section XI, Inservice Insp Second Interval.Ltr Established No New Regulatory Commitments ML20209G3481999-07-15015 July 1999 Transmits Natl Marine Fisheries Svc (NMFS) Biological Opinion Based on Review of Continued Use of Cw Intake Sys at Crystal River Energy Complex.Concludes That Continued Use of Cw Intake Sys Not Likely to Adversely Affect Gulf Sturgeon 3F0799-25, Forwards License Renewal Applications for Four Individuals, IAW 10CFR55.57.Without Encl1999-07-14014 July 1999 Forwards License Renewal Applications for Four Individuals, IAW 10CFR55.57.Without Encl 3F0799-21, Forwards Copy of Revised NPDES Permit IAW Section 3.2.3 of Unit 3 Environ Protection Plan,Per 990430 Request to Allow Use of Biocide in Station Air Compressor Cooling Sys. Wastewater Permit FL0000159 Issued 990630 Also Encl1999-07-14014 July 1999 Forwards Copy of Revised NPDES Permit IAW Section 3.2.3 of Unit 3 Environ Protection Plan,Per 990430 Request to Allow Use of Biocide in Station Air Compressor Cooling Sys. Wastewater Permit FL0000159 Issued 990630 Also Encl 3F0799-05, Requests Exemption from 10CFR70.51, Matl Balance,Inventory & Records Requirements, as It Relates to 10CFR70.51(d) Re Physical Inventory of SNM for Crystal River Unit 3.Detailed Justification for Request,Encl1999-07-14014 July 1999 Requests Exemption from 10CFR70.51, Matl Balance,Inventory & Records Requirements, as It Relates to 10CFR70.51(d) Re Physical Inventory of SNM for Crystal River Unit 3.Detailed Justification for Request,Encl 3F0799-26, Provides Notice of Change in Status for Senior Operator,Iaw 10CFR50.74(a).RD Demontfort,License Number SOP 20528-2,has Been Reassigned & No Longer Requires License Effective 9907301999-07-14014 July 1999 Provides Notice of Change in Status for Senior Operator,Iaw 10CFR50.74(a).RD Demontfort,License Number SOP 20528-2,has Been Reassigned & No Longer Requires License Effective 990730 3F0799-22, Provides Update & Rev to Submittal Made by Util Ltr with Regard to EAL Classification Methodology for Unit 3.Reponses to NRC Staff Questions Provided as Attachment D to Ltr & Reflects Discussions Held1999-07-13013 July 1999 Provides Update & Rev to Submittal Made by Util Ltr with Regard to EAL Classification Methodology for Unit 3.Reponses to NRC Staff Questions Provided as Attachment D to Ltr & Reflects Discussions Held 3F0799-02, Submits Rev 7-3 to Physical Security Plan,Replacing Current Rev to CR-3 Physical Security Plan,Rev 7-2,in Entirety.Rev Withheld,Per 10CFR73.211999-07-0808 July 1999 Submits Rev 7-3 to Physical Security Plan,Replacing Current Rev to CR-3 Physical Security Plan,Rev 7-2,in Entirety.Rev Withheld,Per 10CFR73.21 3F0799-03, Forwards Rev 5-0 to Safeguards Contingency Plan,Replacing Current Rev to Safeguards Contingency Plan,Rev 4,in Entirety.Rev Withheld,Per 10CFR73.211999-07-0808 July 1999 Forwards Rev 5-0 to Safeguards Contingency Plan,Replacing Current Rev to Safeguards Contingency Plan,Rev 4,in Entirety.Rev Withheld,Per 10CFR73.21 3F0799-10, Submits Copy of Historical NPDES Permit Rev That Was Made in 1997 Re Use of Biocide at Crystal River Unit 31999-07-0707 July 1999 Submits Copy of Historical NPDES Permit Rev That Was Made in 1997 Re Use of Biocide at Crystal River Unit 3 ML20209C0811999-06-25025 June 1999 Forwards Overdue Controlled Document Transmittals for Listed Documents 3F0699-12, Provides Suppl Info for LAR 240,rev 0 & Pump Curve for EFP-3 to Facilitate Review,As Requested1999-06-23023 June 1999 Provides Suppl Info for LAR 240,rev 0 & Pump Curve for EFP-3 to Facilitate Review,As Requested 3F0699-06, Submits Final Response to GL 98-01,Suppl 1 Re Year 2000 Readiness of Nuclear Power Plants.Year 2000 Readiness Disclosure for Crystal River,Unit 3,encl1999-06-23023 June 1999 Submits Final Response to GL 98-01,Suppl 1 Re Year 2000 Readiness of Nuclear Power Plants.Year 2000 Readiness Disclosure for Crystal River,Unit 3,encl 3F0699-08, Provides Updated Info to Licensee Response to GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design-Basis Accident Conditions. Ltr Establishes No New Regulatory Commitments1999-06-21021 June 1999 Provides Updated Info to Licensee Response to GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design-Basis Accident Conditions. Ltr Establishes No New Regulatory Commitments 3F0699-09, Forwards FPC 1998 Annual Financial Repts for Two Participating co-owners of Crystal River Unit 3.Financial Statements & Independent Auditors Repts for City of Alachua,Fl,Encl1999-06-0404 June 1999 Forwards FPC 1998 Annual Financial Repts for Two Participating co-owners of Crystal River Unit 3.Financial Statements & Independent Auditors Repts for City of Alachua,Fl,Encl 3F0599-21, Submits Addendum to B&W Owners Group Topical Rept BAW-2346P, Rev 0.Addendum Includes Leak Rate Values Based on CR-3 Plant Specific Main Steam Line Break Tube Loads1999-05-28028 May 1999 Submits Addendum to B&W Owners Group Topical Rept BAW-2346P, Rev 0.Addendum Includes Leak Rate Values Based on CR-3 Plant Specific Main Steam Line Break Tube Loads 3F0599-10, Submits Changes Made to Crystal River,Unit 3 Its,As Required by ITS 5.6.2.17.Encl Provides Revs to Plant ITS Bases That Will Update NRC Copies of Its.Instructions for Updating ITS, Encl1999-05-26026 May 1999 Submits Changes Made to Crystal River,Unit 3 Its,As Required by ITS 5.6.2.17.Encl Provides Revs to Plant ITS Bases That Will Update NRC Copies of Its.Instructions for Updating ITS, Encl ML20207E4341999-05-25025 May 1999 Submits 30-day Written Rept on Significant PCT Changes in ECCS Analysis for ANO-1.CRAFT2 Limiting PCT for ANO-1 Was Bounded by 1859 F PCT Calculated at 2568 Mwt for Crystal River 3 Cold Leg Pump Discharge Break Size of 0.125 Ft 3F0599-22, Forwards non-proprietary Version of B&Wog Topical Rept BAW-2346NP, Alternate Repair Criteria for Tube End Cracking in Tube-to-Tubesheet Roll Joint of Once-Through Sgs1999-05-21021 May 1999 Forwards non-proprietary Version of B&Wog Topical Rept BAW-2346NP, Alternate Repair Criteria for Tube End Cracking in Tube-to-Tubesheet Roll Joint of Once-Through Sgs 3F0599-18, Forwards 1998 Annual Radiological Environ Operating Rept for Crystal River,Unit 3. Rept Is Submitted in Accordance with CR-3 ITS 5.7.1.1(b) & Section 6.6 of ODCM1999-05-14014 May 1999 Forwards 1998 Annual Radiological Environ Operating Rept for Crystal River,Unit 3. Rept Is Submitted in Accordance with CR-3 ITS 5.7.1.1(b) & Section 6.6 of ODCM 3F0599-17, Submits Update Response to GL 97-06, Degradation of SG Internals. Ltr Establishes No New Regulatory Commitments1999-05-14014 May 1999 Submits Update Response to GL 97-06, Degradation of SG Internals. Ltr Establishes No New Regulatory Commitments 3F0599-07, Submits Guarantee of Payment of Deferred Premiums for CR-3 in Accordance with 10CFR140.21.Internal Cash Flow Projection Was Prepared in Accordance with Suggested Format Outlined in Reg Guide 9.4 Dtd Sept 19781999-05-14014 May 1999 Submits Guarantee of Payment of Deferred Premiums for CR-3 in Accordance with 10CFR140.21.Internal Cash Flow Projection Was Prepared in Accordance with Suggested Format Outlined in Reg Guide 9.4 Dtd Sept 1978 3F0599-03, Provides Update Curves for Facility Pressure/Temp Limits Rept,Rev 2 & Updated Rev Bar ITS Pages Associated with LAR, in Response to NRC RAI Re Subject LAR1999-05-12012 May 1999 Provides Update Curves for Facility Pressure/Temp Limits Rept,Rev 2 & Updated Rev Bar ITS Pages Associated with LAR, in Response to NRC RAI Re Subject LAR 3F0599-05, Responds to 990402 RAI Re Third 10-year Interval ISI Program Plan Requests for Relief.Util Revised Relief Requests 98-010-II,98-003-PT,98-005-PT & 98-001-SS Based on Responses to Rai.Revised Relief Requests Encl1999-05-12012 May 1999 Responds to 990402 RAI Re Third 10-year Interval ISI Program Plan Requests for Relief.Util Revised Relief Requests 98-010-II,98-003-PT,98-005-PT & 98-001-SS Based on Responses to Rai.Revised Relief Requests Encl 3F0599-08, Forwards Licensee Clarification of Info Provided in Amend 171 Re post-LOCA Boron Dilution Precipitation Prevention.Ltr Establishes No New Regulatory Commitments1999-05-0303 May 1999 Forwards Licensee Clarification of Info Provided in Amend 171 Re post-LOCA Boron Dilution Precipitation Prevention.Ltr Establishes No New Regulatory Commitments 3F0599-09, Forwards Crystal River Unit 3 Radioactive Effluent Release Rept - 1998 & Revised Crystal River Unit 3 Radioactive Effluent Release Rept - 1997. Licensee Informs That ODCM & PCP Were Not Revised During 19981999-05-0101 May 1999 Forwards Crystal River Unit 3 Radioactive Effluent Release Rept - 1998 & Revised Crystal River Unit 3 Radioactive Effluent Release Rept - 1997. Licensee Informs That ODCM & PCP Were Not Revised During 1998 3F0499-24, Forwards Summary of Proposed Changes to Crystal River,Unit 3 NPDES Permit,That Are Being Submitted to Florida Dept of Environ Protection.Proposed Change Will Allow Use of Scale Inhibitor,Biocides & Foam Control Agent1999-04-30030 April 1999 Forwards Summary of Proposed Changes to Crystal River,Unit 3 NPDES Permit,That Are Being Submitted to Florida Dept of Environ Protection.Proposed Change Will Allow Use of Scale Inhibitor,Biocides & Foam Control Agent 3F0499-09, Forwards FPC Annual Financial Rept & Annual Financial Repts for Eight of Ten Participating co-owners of Crystal River Unit 3 Nuclear Station.Outstanding Annual Financial Rept Will Be Submitted by 9907301999-04-30030 April 1999 Forwards FPC Annual Financial Rept & Annual Financial Repts for Eight of Ten Participating co-owners of Crystal River Unit 3 Nuclear Station.Outstanding Annual Financial Rept Will Be Submitted by 990730 3F0499-23, Submits Repts Required by App B,Environ Protection Plan,Of Crystal River,Unit 3 Operating License.Fl Dept of Environ Protection Has Provided Clarification Re Ph Monitoring Requirements1999-04-23023 April 1999 Submits Repts Required by App B,Environ Protection Plan,Of Crystal River,Unit 3 Operating License.Fl Dept of Environ Protection Has Provided Clarification Re Ph Monitoring Requirements 3F0499-18, Informs of Recent Senior Management Change at Fpc,Which Will Not Affect Std Recipients of Incoming NRC Correspondence. Updated Util Mailing List,Encl1999-04-20020 April 1999 Informs of Recent Senior Management Change at Fpc,Which Will Not Affect Std Recipients of Incoming NRC Correspondence. Updated Util Mailing List,Encl 3F0499-05, Forwards Rev 19 to Radiological Emergency Response Plan. Changes to Plan Marked with Vertical Bars in Left Margin1999-04-16016 April 1999 Forwards Rev 19 to Radiological Emergency Response Plan. Changes to Plan Marked with Vertical Bars in Left Margin 3F0499-08, Forwards FPC Annual ITS Dose Rept for Period Jan-Dec 1998. Rept Provides person-rem Radiation Exposures,According to Work & Job Function,At CR-3 for Period Jan-Dec 19981999-04-16016 April 1999 Forwards FPC Annual ITS Dose Rept for Period Jan-Dec 1998. Rept Provides person-rem Radiation Exposures,According to Work & Job Function,At CR-3 for Period Jan-Dec 1998 1999-09-27
[Table view] Category:UTILITY TO NRC
MONTHYEAR3F0990-11, Forwards Final Status Update Re Design & Operations Verification for Instrument Air Sys Per Generic Ltr 88-141990-09-20020 September 1990 Forwards Final Status Update Re Design & Operations Verification for Instrument Air Sys Per Generic Ltr 88-14 3F0990-05, Forwards 1990 Inservice Insp Summary Rept. Rept Contains Owners Data Rept,Data Summary Sections for Class 1,2 & 3 Components,Rept for Repair & Replacements & Listing of Exams1990-09-14014 September 1990 Forwards 1990 Inservice Insp Summary Rept. Rept Contains Owners Data Rept,Data Summary Sections for Class 1,2 & 3 Components,Rept for Repair & Replacements & Listing of Exams 3F0990-08, Forwards Response to Violations Noted in Insp Rept 50-302/90-23.Corrective Actions:Training Session Conducted to Stress Importance of Attachments Being Part of Work Package When Required by Procedure1990-09-13013 September 1990 Forwards Response to Violations Noted in Insp Rept 50-302/90-23.Corrective Actions:Training Session Conducted to Stress Importance of Attachments Being Part of Work Package When Required by Procedure 3F0890-23, Discusses Decommissioning Financial Assurance Rept Dtd 900726.Amount Util Collecting Exceeds Amount Necessary Based on NRC Formula & Does Not Include Estimated Cost of Removal & Disposal of Nonradioactive Structures & Matls1990-08-30030 August 1990 Discusses Decommissioning Financial Assurance Rept Dtd 900726.Amount Util Collecting Exceeds Amount Necessary Based on NRC Formula & Does Not Include Estimated Cost of Removal & Disposal of Nonradioactive Structures & Matls ML20028G8251990-08-29029 August 1990 Advises That Supplemental Response to Insp Rept 50-302/89-18 Will Be Submitted by 901030 ML20059G1721990-08-24024 August 1990 Submits Info Re Change in Operator License Status.Re Rawls & Wa Stephenson Senior Reactor Licenses Should Be Terminated, Effective 900817 Due to Reassignment from Position Requiring Licenses 3F0890-15, Forwards Fitness for Duty Program Performance Data for Period of Jan-Jun 19901990-08-23023 August 1990 Forwards Fitness for Duty Program Performance Data for Period of Jan-Jun 1990 3F0890-21, Forwards Semiannual Radioactive Release Rept Jan-June 1990 for Crystal River Unit 3 & Rev 15 to Crystal River Unit 3 Odcm1990-08-23023 August 1990 Forwards Semiannual Radioactive Release Rept Jan-June 1990 for Crystal River Unit 3 & Rev 15 to Crystal River Unit 3 Odcm ML20059A0031990-08-16016 August 1990 Responds to NRC 900613 Request for Addl Info Re Util 871222 Response to Violations Noted in Insp Rept 50-302/87-30 3F0890-03, Forwards Rev 11 to Inservice Insp - Pump & Valve Program, Crystal River Unit 31990-08-16016 August 1990 Forwards Rev 11 to Inservice Insp - Pump & Valve Program, Crystal River Unit 3 3F0890-10, Advises That Util Completed Mods to Comply w/10CFR50.62 Requirements Re Reduction of Risk from ATWS Events1990-08-16016 August 1990 Advises That Util Completed Mods to Comply w/10CFR50.62 Requirements Re Reduction of Risk from ATWS Events 3F0890-05, Forwards Addl Info Re Response to Mode 3 Loca,Per 900711 Request,Providing Background of Factors Considered During Evaluation of Tech Spec Change Request 1741990-08-10010 August 1990 Forwards Addl Info Re Response to Mode 3 Loca,Per 900711 Request,Providing Background of Factors Considered During Evaluation of Tech Spec Change Request 174 ML20058L2001990-08-0202 August 1990 Provides Current Status of Reg Guide 1.97 Activities. Pressurizer Heater Status & Main Steam Safety/Relief Valve Position Indications Completed 3F0790-10, Forwards Justification for Continued Operation Re Emergency Diesel Generator Block Loading Voltage Dips.Util Will Install Higher Accuracy Relays to Improve on Accuracy & Repeatability of Load Intervals1990-07-18018 July 1990 Forwards Justification for Continued Operation Re Emergency Diesel Generator Block Loading Voltage Dips.Util Will Install Higher Accuracy Relays to Improve on Accuracy & Repeatability of Load Intervals 3F0790-06, Forwards Rev 1 to Cycle 8 Core Operating Limits Rept, Correcting Typo in Note 1 of Figures 1-8 & Note 2 of Figures 1,2 & 4-81990-07-12012 July 1990 Forwards Rev 1 to Cycle 8 Core Operating Limits Rept, Correcting Typo in Note 1 of Figures 1-8 & Note 2 of Figures 1,2 & 4-8 ML20055D5561990-06-29029 June 1990 Forwards 890505 & s to Be Placed in Plant File 3F0690-15, Informs That Tech Spec Actions Identified in Util Addressed as Part of Tech Spec Improvement Program,Per 900518 Request for Programmed Enhancements for Generic Ltr 88-171990-06-29029 June 1990 Informs That Tech Spec Actions Identified in Util Addressed as Part of Tech Spec Improvement Program,Per 900518 Request for Programmed Enhancements for Generic Ltr 88-17 3F0690-22, Forwards Annual Financial Repts for 1989 for Orlando Utils Commission & Cities of Bushnell,Leesburg,Ocala & Tallahassee,Per 10CFR50.71(b)1990-06-27027 June 1990 Forwards Annual Financial Repts for 1989 for Orlando Utils Commission & Cities of Bushnell,Leesburg,Ocala & Tallahassee,Per 10CFR50.71(b) 3F0690-19, Responds to Deviations Noted in Insp Rept 50-302/90-15. Corrective Actions:All Reg Guide 1.97 Category 1 Instruments on Main Control Board Marked & Engineered Safeguards Matrix Indicating Lights Arranged in Unique Array1990-06-25025 June 1990 Responds to Deviations Noted in Insp Rept 50-302/90-15. Corrective Actions:All Reg Guide 1.97 Category 1 Instruments on Main Control Board Marked & Engineered Safeguards Matrix Indicating Lights Arranged in Unique Array ML20043H5781990-06-21021 June 1990 Responds to Generic Ltr 89-06, Spds. All Open Issues Identified in NRC SER & Plant SPDS Satisfies NUREG-0737, Item I.D.2 & Suppl 1 ML20044A6121990-06-21021 June 1990 Forwards Payment of Civil Penalty,Per NRC 900524 Order Based on Findings in Insp Rept 50-302/89-09 ML20043J0221990-06-21021 June 1990 Submits Info in Support of Tech Spec Change Request 175,Rev 1,Suppl 1 Re Spent Fuel Pool Storage Capacity at Plant ML20043H4151990-06-21021 June 1990 Forwards Response 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. 3F0690-18, Responds to Violations Noted in Insp Rept 50-302/90-09. Corrective Action:Temporary Lettering Placed on Index Plates for Consistency W/Previous Markings Immediately Following Discovery of Error1990-06-20020 June 1990 Responds to Violations Noted in Insp Rept 50-302/90-09. Corrective Action:Temporary Lettering Placed on Index Plates for Consistency W/Previous Markings Immediately Following Discovery of Error ML20043H2901990-06-15015 June 1990 Forwards Results of Refuel 7 once-through Steam Generator (OTSG) Eddy Current Insp,Per Tech Spec Section 4.4.5.5.Eight Defective Tubes & Two Administratively Plugged Tubes in OTSG a Resulted from Review of Insp Data ML20044A0641990-06-13013 June 1990 Forwards Rev 5 to Physical Security Plan,Replacing Currently Approved Rev 4 W/Amends.Rev Withheld (Ref 10CFR73.21) ML20043G2271990-06-12012 June 1990 Suppls 900604 Ltr Describing How Control Room Habitability Dose Would Be Adversely Affected by Elimination of Reactor Bldg Flood Vol Unless Overly Conservative Failure Postulations Also Changed.Quarterly Updates to Be Provided ML20043G3711990-06-12012 June 1990 Forwards 1990 Internal Cash Flow Projection for Plant Which Updates Utils Utilization of Alternative (E) ML20043G2361990-06-12012 June 1990 Forwards Info Re Fire Protection Sys Reliability for Providing Water to Intermediate Bldg Following High Energy Line Break.B&W Issued Contract to Verify Mass/Energy Release & Motor Starters & Terminal Blocks Insulated ML20043F1601990-06-0404 June 1990 Provides Supplemental Info on Reactor Bldg Flooding,Per Util 900517 Ltr Describing Resolution Plan Being Pursued.Util Has Limited Vol of Water Contributed by Borated Water Storage Tank & Sodium Hydroxide Tank to Flood Level ML20043C8721990-05-31031 May 1990 Forwards Rev 0 to Crystal River Unit 3 Cycle 8 Core Operating Limits Rept, Per Tech Spec 6.9.1.7 ML20043B8431990-05-24024 May 1990 Describes Alternative Testing of Reactor Bldg Spray Suction Valves BSV-1 & BSV-8,per Generic Ltr 89-04.At Least One Valve Will Be Disassembled & Inspected During Each Refueling Outage Using Alternate Insp Method ML20043B2771990-05-18018 May 1990 Advises of Mod to Original Commitment & Plans to Store Seal Ring in Original Storage Location During Plant Operations, Per 790606 Ltr.Mod Minimizes Personnel Exposure & Enhances Seal Plate Leak Tightness Due to Less Handling ML20043B0321990-05-17017 May 1990 Provides Details of Resolution Util Will Pursue Re Reactor Bldg Flooding Detailed in Encl LER 90-005.Mod Will Be Installed to Add Alarm in Main Control Room to Indicate When Flood Level Reaches Point & Operator Action Begins ML20043A5441990-05-16016 May 1990 Discusses Status of Safety & Performance Improvement Program Portion of B&W Owners Group EOP Review Project ML20042G7681990-05-10010 May 1990 Submits Unsatisfactory Performance Testing Incident Repts, Per 10CFR26, Fitness for Duty. Results of Three Positive Blind Test Samples Certified to Contain No Drugs ML20043A4931990-05-10010 May 1990 Forwards Executed Amend 8 to Indemnity Agreement B-54 ML20042G2101990-05-0707 May 1990 Provides Notice of MW Kirk Permanent Reassignment from Position & Requests Termination of Senior Reactor Operator License 20481-1,effective 900505 ML20042E8611990-04-26026 April 1990 Forwards Annual Financial Repts for Six Participating Owners of Plant ML20012F2941990-03-30030 March 1990 Forwards Nonproprietary Rev to 51-1176431-02, Crystal River 3 Reactor Vessel...Temp Overpressure Protection, in Support of Tech Spec Change Request 174,Suppl 1 Re Response to Generic Ltr 88-11 ML20042D8321990-03-30030 March 1990 Provides Supplemental Response to Station Blackout Rule Implementation & Affirms That Diesel Generator Target Reliability Will Be Maintained ML20012E2041990-03-23023 March 1990 Requests Temporary Waiver of Compliance Granted on Tech Spec 3.9.8.2 Re DHR Power Source Requirements ML20012D9471990-03-21021 March 1990 Requests Approval of Capsule Withdrawal Schedule in Table 3-19 of BAW-1543,Rev 3, Master Integrated Reactor Vessel Surveillance Program to Allow Plant Refuel 7 Outage Plans to Continue on Schedule.Changes Needing NRC Approval Listed ML20012D7371990-03-19019 March 1990 Responds to Generic Ltr 89-19 Re Resolution of USI A-47. Util Will Implement Appropriate Sys to Protect Against Overfill Concerns ML20012C4501990-03-13013 March 1990 Forwards Listing of Insurance Policies in Place for Plant as of 900225 ML20012B4821990-03-0707 March 1990 Forwards Inservice Insp Pump & Valve Program Relief Request V-371 Proposing Alternate Acceptance Criteria Requirements for IWP-4150 for Fluctuations in Hydraulic Instrument Readings ML20012B3651990-03-0101 March 1990 Lists Five Addl Drugs Included in 10CFR26 Re fitness-for- Duty Testing Program.Specimens Identified as Positive on Initial Screening Will Be Confirmed Using Gas Chromatography or Mass Spectrometry at Listed cut-off Levels ML20012A4041990-02-27027 February 1990 Forwards 1989 Annual Rept of Personnel Exposure in Accordance w/10CFR20.407 & Tech Spec 6.9.1.5.(a) & Annual Rept of Facility Changes,Tests & Experiments in Accordance w/10CFR50.59 ML20006G1731990-02-23023 February 1990 Advises That Util Voluntarily Agrees to Participate in Emergency Response Data Sys Proposed in Generic Ltr 89-15 ML20011F2411990-02-21021 February 1990 Provides Followup on & Documents Discussion in 900116 Meeting Re Emergency Diesel Generator Loading 1990-09-20
[Table view] |
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e,e Florida Power C O R POR ATION May 24, 1989 3F0589-20 U. S. Nuclear Regulatory Commission Attention: Document Control Desk
' Washington, D.C. 20555
Subject:
Crystal River Unit 3 Docket No. 50-302 Operating License No. DPR-72 High Energy Line Break
Dear Sir:
Florida Power Corporation (FPC) has had discussions with the staff
'regarding the bases provided in our December 16, 1989 exemption requestLfor schedular relief from the requirements of GDC-4. In particular, the staff requested FPC to elaborate on certain technical details of our undue risk considerations. FPC has reviewed the basis provided and has included additional information where appropriate.
Attachment i reflects the original exemption request supplemented with this information.
FPC has started design vork on 10 of the break locations. The work includes both jet shield and pipe whip restraints. These 10 locations
.were selected as they were the same under both the old criteria [GAI Report 1811) and new criteria (submitted on March 31, 1989]. In I addition, these break locations included those having the most potential for damage. These locations include 4 main steam line breaks, 2 main feedwater line breaks, 2 steam generator blowdown line breaks, and 2 auxiliary steam line breaks (branch lines off the main steam line to the emergency feed. water pump turbine). The main steam and main feedwater breaks have the largest potential for energy release in the event of a' break. These breaks already have some existing shielding and pipe whip restraints. Attachment 3 shows a complete tabulation of all breaks postulated using the revised criteria.
8906050106 890524 gDR ADOCK0500ggg2 goof
' s POST OFFICE BOX 219
- CRYSTAL RIVER, FLORIDA 326294219 + (904) 795-6486 A Florida Progress Company
i t
May 24, 1989 3F0589-20 Page 2 The schedule provided in the exemption request was based on the information available at the' time of the submittal. The resolution of concerns with the criteria and the experience gained through the work ;
on the 10 break locations will allow development of a firm schedule. I We anticipate completion in 1990 depending on the need for and scheduling of outages during the next cycle of operation.
Sincerely, Rolf C. Widell, Director Nuclear Operations Site Support RCW/JWT/sdr Attachment xc: Regional Administrator, Region II Senior Resident Inspector
n ..
May 20, 1989
~3F0589-20 ATTACHMENT 1 BASES FOR SPECIFIC EXEMPTION NRC. regulations in 10CFR50.12(a) provide.the requirements for seeking specific exemptions. Under Section 50.12, an exemption may be. granted if the' exemption will not present an undue risk to the public health and. safety and~special circumstances are present. FPC believes the standards for an exemption are satisfied by considering the following:
- 1. No Undue Risk L The proposed schedular exemption will present no undue risk to public' health and safety because:
- a. The principle safety systems have existed since original plant construction and the implementation.of GAI #1811.
Therefore, HELB interactions are'quite likely to have been properly _'dispositioned. We are unable to locate sufficient documentation to rely upon.this as permanent resolution and are, therefore, reestablishing our demonstration of compliance with a more appropriate HELB design basis. Even thoughLFPC failed to consider explicitly HELB effects,.many other criteria (electrical separation, Appendix R, etc.)
provide some degree of protection by reducing the likelihood that any one break can cause loss of safety function by impacting multiple trains.
FPC has not identified any information which would reduce our confidence that the original HELB work was adequately accomplished. Nevertheless, FPC will improve the level of demonstrated compliance as we implement the revised criteria,
- b. FPC has examined the pipe stress analysis in the piping systems covered by the HELB criteria to get an understanding of the actual stress levels in the piping compared to the specified allowables at the postulated terminal end break locations. The results of that examination are shown in Attachment 2 (revised). The data shows that postulated break locations are not highly stressed. The reduced stress levels ,
in the. piping systems give FPC further assurance that a terminal end break is a very low probability event.
The staff requested FPC to demonstrate that calculated stress levels using the original design codes (generally ANSI
'B31.1). envelope those calculated using more current codes (ASME III, NC) which form the basis for current staff positions on AIB's, etc. Attachment 4 is a portion of that review. This review continues to demonstrate very low stress levels and hence low break probability.
1
e ,May-20,.1989 "U 3F0589-20 l
'm \
- c. CR-3 has functional redundancy in the High Pressure Injection
'(HPI) System and the Emergency ' Feedwater (EF) System. That' is, the plant can be' safely shutdown;with continued decay 3 heat removal- in' Hot Shutdown conditions, with secondary heat removal (EF) or primary feed-and-bleed-(HPI). Without-t completing a full' scale systems interaction analysis, it can be concluded.that one of these two systems will be available to'aitigate a break in'aither of the two buildings'in which HELBs are postulated. No HPI System components are located in the Intermediate Building (IB). Consequently, no HELB event in the IB can affect the HPI System performance.
- There are EFW System components located within the Auxiliary Building (AB).- FPC has evaluated.the location of this equipment.and. potential HELBs to determine if any adverse interactions are. evident. Based on our evaluation, FPC'can conclude that the EFW System performance will not be affected by a HELB in the AB.
L ,
The staff requested FPC to expand this consideration in at least=two areas: (1) retention of RCS inventory maintenance capability and-(2) possible remote or local operability should automatic function be lost. Attachment 6 is a further discussion of Makeup System reliability.
The survivability of make-up, letdown., MFW isolation, EFW, RB cooling, and RC drain tank isolation capability were more exhaustively evaluated in the Appendix R reviews. These reviews postulated complete loss of all make-up equipment in each of'8 fire zones in the Auxiliary Building and 3 fire zones in the Turbine Building concurrent with removing from service each of~the three make-up pumps. The review demonstrated safe shutdown capability in each case. There are only 12 breaks in the Auxiliary Building. None of these breaks are likely to cause damage in excess of the Appendix R postulation. One break is even less likely to damage all equipment in a zone than would a fire. The Appendix R analysis is, therefore, generally considered bounding and sufficient to demonstrate retained safe shutdown capability even with unprotected interactions. The means utilized in achieving this level of function survivability was the verification of. existing features or modifications to provide a sufficient degree of compartmentalization, such as spatial separation or barriers. FPC chose not to rely on post-fire maintenance, operator action, or exemption. The largest modification effort was directed at separating A and B trains of shutdown systems'by floor, i.e. most A train equipment L powered and controlled via circuits on the 119' elevation and B train circuits cn1 the 958 elevation. Such degree of compartmentalization will assure survivability of safe shutdown systems to function.
2
__--_-__-______-_-__-_m
h May 20,.1989 3F0589-20
.WithLregard to remote manual, local, or manual capability.
Esurviving a HELB where automatic function is lost, FPC provides.the following additional support. -The majority of added " targets":were associated with the addition'of an automatic capability to existing manual capability. . Since.
.there1were-approximately=10,000 potentia) targets and several hundred: modification packages.which may have added such-equipment,.it has not been possible to draw any formal conclusions regarding numbers or percentages added, contrasted with pre-existing equipment.. Even though actual
- g. quantitative values are not-readily retrievable, the largest modifications-(EFIC,' Remote' Shutdown, Post Accident:
Monitoring, etc.).do not generally impact local controls.
- Those should remain protected. Further, Appendix R reviews described above'do' demonstrate a large degree of
. survivability with normal shutdown, even though automatic HELB mitigating systems were not included in this review.
As noted above,.there are relatively few break locations in the Auxiliary-Building ~. This is due, in part,Jto the fact that.one lif the-high energy lines is~the auxiliary steam line that provides steam heating to'the reactor coolant waste evaporators. That line has remained under.the original.
design. criteria (Report #1811]'for high energylline break.
As.such, safety-related equipment in the vicinity of that line must1be protected from. postulated leak cracks at.the
-most adverse location. As a part of reviewing the design basis for CR-3 for HELB,.FPC determined that the. automatic isolation circuitry for the auxiliary steam line that would protect-the environment of the Auxiliary Building would-only work for breaks,-not cracks. To justify continued operation, FPC decided to keep this'line' closed at all times until a permanent fix could be installed. Therefore, this part of the HELB concern at CR-3 has been eliminated until the
~
permanent fix is installed. The permanent fix will consider
.all the HELB effects, i.e., jet impingement and environmental r . qualification.
- d. CR-3 is located in an area which ja considered to be r ' seismically inactive. The CR-3 p?. ping is designed for approximately 4 times the expected ground motion (0.03g based on current seismic hazards methodology). This fact combined with the historical-perspective of earthquake activity discussed in FSAR Section 2.5.4, Seismology, leads FPC to conclude that the seismic load contribution to the potential for HELBs is significantly overstated.
e .- Increased attention given to the integrity of feedwater piping systems brought about by the Surry event and the actions required by Bulletin 87-01, Thinning of Pipe Walls in Nuclear Power Plants, reduces the probability of a HELB.
FPC's program which is described in our letter dated 3
1
6 May 20, 1989 3F0589-20 September-3, 1987 began with the last refueling outage and is ongoing. As stated in our bulletin response, FPC's program provides additional assurance that adequate structural integrity remains in our high energy feedwater piping systems and CR-3 can continue to operate safely.
- f. Licensed operators are trained to mitigate accidents initiated by HELBs. Scenarios are routinely run during simulator training providing opportunities for operators to diagnose conditions caused by HELB and mitigate resulting
. transients. Emergency / Abnormal operating procedures address
-compensatory measures which can be taken in the event that certain equipment fails to actuate or does not perform its intended function. These procedures and training provide added assurance that the plant can be safely shutdown in the event of HELB. Attachment 5 provides a list of instrumentation that could be used to indicate a HELB in a particular system.
For these reasons, FPC believes that adequate protection of the public health and safety will be maintained during the period relief will be in effect.
i l
l 4
t May 20,.1989 3F0589-20 MTKHENT 2 POSTUIATED TERMINAL END FIPE BREAK STRESS IEVEIS AND BREAK THRESH 0ID IJMITS MATERIAL AND STRESS TABLE SLMEARY SYSTEM (S) MATERIAL PIPING ALIDWABLES BREAK THRESEDID Sh SA .8 SA .8 (Sh+SA)
E ,0ISG,FW,EEW,AS A106 GR B 15000 22500 18000 30000 E-1 & W -2 A376TP304 18062 27952 22362 36811
~ MU-6 THRU -10 -A312/27.9 .17500 27875 22300 36300 E-4,-5,-12, & -13 A312/26.9 14000 26875 21500- 32700 MU-14 THRU 17 A312/27.0 18062 27013 21610 36060 E-11 A312/28.2 18062 28200 22560 37010
- RIGCROUS ANALYSIS ** % OF DATA IN BREAK PIPE PRIMARY SEC'RY- 70IAL THRESHOLD DECENDING NLMBER HAT'L
- SIRESS STRESS STRESS BREAK CRDER STRESS *************
MAIN STEAM SYSTEM (TABLE C-2-1)
MS-1 A106 GR B 5729 4915 10644 35.5 41.2 E -12 MS-5 A106 GR B 5445 752 6197 20.7 35.5 MS-1 E-9 A106 GR B 5934 4233 10167 33.9 33.9 MS-9 E-12 A106 GR B 6797 5550 12347 41.2 27.7 MS-16 MS-15 A106 GR B 8257 0 8257 27.5 27.5 MS-15 E -16 A106 GR B 8296 0 8296 -27.7 20.7 E-5 MS-17 A106 GR B 2733 0 2733 9.1 15.2 MS-26 E-18 A106 GR B 2760 0 2760 9.2 15.0 E -37 MS-19 A106 GR B 2711 0 2711 9.0 14.9 MS-36 MS-20 A106 GR B 2867 0 2867 9.6 14.1 MS-23 E-23 A106 GR B 4234 0 4234 14.1 13.7 MS-29 MS-24 A106 GR B 3917 0 3917 13.1 13.7 E-25 E-25 A106 GR B 4117 0 4117 13.7 13.1 MS-24 E -26 A106 GR B 4570 0 4570 15.2 12.8 E -32 MS-29 A106 N B 4105 0 4105 13.7 11.9 MS-35 MS-30 A106 GR B 3351- 0 3351 11.2 11.4 E -38 E -31 A106 GR B 3246 0 3246 10.8 11.2 MS-30 E -32 A106 GR B 3852 0 3852 12.8 10.8 E -31 MS-35 A106 GR B 3559 0 3559 11.9 9.6 E -20 E -36 A106 GR B 4480 0 4480 14.9 9.2 MS-18 MS-37 A106 GR B 4509 0 4509 15.0 9.1 MS-17 MS-38 A106 GR B 3414 0 3414 11.4 9.0 E -19 I
OISG BLOWDOWN SYSTEM (TABIE C-2-1)
MS-41 A106 GR B 10358 14664 25022 83.4 99.1 E -42 E-42 A106 GR B 15051 14664 29715 99.1 83.4 MS-41
m 9
- 1 May 20, 1989
'3F0589-20 -l
]
- RIGCEDUS ANALYSIS ** '% OF DATA IN IEEAK PIPE PRIMARY SEC'RY TUIAL THRESHOID IECENDING NLMBER . MAT'L
- STRESS STRESS STRESS IEEAK OEEER STRESS *************
FEE [ MATER SYSTEM (TABLE C-2-2)
FW-1 ' A106 GR B 5903 1635 7538 25.1 84.7 FW-28B L, 'FW-6 A106 GR B 9421 867 10288 34.3- 60.7 FW-28A FW-10. A106 GR B 11846 5352 17198- 57.3 57.3 FW-10 FW-14 A106 GR B 5235 1700 6935 23.1 41.7 FW-31 FW-20 A106 GR B'. 10506 601 11107 37.0. 37.0 FW-20' FW-24 'A106 GR B'- 6942 -1664 8606' 28.7 -34.3 FW-6 FW-28A-A106 GR B- ~5771 10931- 16702' 60.7 33.9 FW-32 FW-28B A106 Gt B 6928. 15245 -22173 84.7 28.7 FW-24 FW-28C A106 GR B. 7757 157 7914 26.4 26.4 FW-28C-FW-29 A106 GR B 7459 .103 7562 25.2 25.2 FW-29 1FW-30 A106 GR B- 4899 125. 5024 16.7 25.1 FW-1 FW-31 A106 GR B 4758 7509 -12267 41.7 24.9 FW FW-31 A106'GR B 6753 0 6753 22.5 23.1 FW-14 FW-32 A106 GR B 7480 0 .7480 24.9 22.5 FW-31 FN-32 A106 GR B 6867 3296 10163 33.9 16.7 FW-30 MAKEUP AND PURIFICATION SYSTEM (TABLE C-2-3)
' W-1 A376TP304 5505' 754= 6259 17.0 86.0 MJ-16 E-2 -A376TP304 5373 1097 6470 17.6 62.7 MJ-5
' W A312/26.9 8151 1666 9817 30.0 57.7 MJ-10
. E-5 A312/26.9 14185 6330 20515 62.7 35.5 MJ-8 MJ-6 .A312/27.9 4349 1475- 5824 16.0 31.2 E-13.
MJ-7; A312/27.9 5682 1281 6963 19.2 30.0 MJ-4 MJ-8 A312/27.9 6418 6481 -12899 35.5. 29.3 MJ-12 MJ-9 A312/27.9 6860 151. 7011 19.3 28.5' KI-17 MJ-10 A312/27.9 3507 12862 16369 57.7 24.3 KI-15 MJ-11 A312/28.2 5482 492 5974 16.1 23.4 MJ-14 MJ-11 7A312/28.2- 3949 169 4118 11.1 19.3 MJ-9
. MJ-11 ' A312/28.2 4517- 749 5266 14.2 19.2 MJ-7 MJ-12 A312/26.9 7602 1981 9583 29.3 .17.6- MJ-2 M1-13 A312/26.9 8237 1981 10218 31.2' 17.0' MJ-1
- MJ-14.- A312/27.0 4458 615 5073 14.1 16.1 MJ-11 MJ-14 A312/27.0 8447 0 8447 23.4 16.0 MJ-6 Mi-15 A312/27.0 4058 4698 8756 24.3 14.2 MJ-11 MJ-16 A312/27.0 4918 18587 23505 86.0 14.1 MJ-14 MJ-17 A312/27.0 5928 4353 10281 28.5 11.1 MJ-11 EMERGENCY FEEDWATER SYSTEM (TABIE C-2-5)
EF-11 A106 GR B 4859 2 4861 16.2 18.8 EF-13 EF-13 A106 GR B 5650 0 5650 18.8 16.2 EF-11
LMay 20,-1989-
. 3FT)589-20
- RIGCROUS ANMXSIS ** % OF DMA B BREAK PIPE PRIMME . SEC'RY TUIAL THRESICID IECENDING NLMBER MAT'L
- SIRESS STRESS STRESS BREAK OREER STRESS *************
AUXILIARY STEAM SYSTEM (TABIE C-2-4)
AS-1 . A106 GR B 3355 5798 9153 32.2 42.4 AS-5 AS-1 A106 GR B 6430 4065 10495 35.0. 35.0 -AS-1 AS-1 A106 GR B 4841 9 4850 16.2 32.2~ AS-1~
AS-1 A106 GR B 6724' 418 7142 23.8 28.0 AS-12 AS-5 A106 GR B 5589 58 5647 18.8 26.7 AS-4 AS-5 A106 GR B 10175 2559 12734 42.4 26.0 AS-4 AS-4 -A106 GR B 6435 1589 8024 26.7 24.0 AS-13 AS-4 A106 GR B 7591 210 7801 26.0 23.8 AS-1 AS-10 A106 GR B 6561 473 7034 23.4 23.4 AS-10 AS-12 A106 GR B 5454 2944 8398 28.0 18.8 AS AS-13 -A106 GR B- 7188' O 7188 24.0 16.2 AS-1
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May 20, 1989
- 3 F0589-2 0 ATTACHMENT 3 HELB BREAKS A review of postulated break locations was performed to determine the extent of HELB protective devices installed during original plant construction and subsequent modifications and upgrades. The table below summarizes the review findings.
- Breaks # Breaks Breaks With Line # # Breaks Partial Completely No Potential Intent System Size Breaks Restrained Shielded Shielded Tarcet_,
MS 24" 4a 4 4 0 0 MS 1-1/2" 9 0 0 0 5b OTSG BD 3" 2 2 0 2 0 AS 6" 3c 3 2 0 0 AS 4" 1 0 0 0 0 EF 6" 2 0 0 0 0 FW 18" 2a 2 1 0 0 FW 6" 7 0 1 0 0 MU 2-1/2" 6 0 0 0 0 MU 3" 4 0 0 0 0 MU 2 0 0 0 0 0 Total 42 11 8 2 5 Footnotes:
- a. Design of required jet shields to completely mitigate breaks is in process.
-b. 5 breaks in 1-1/2" MS branch lines will blowdown and whip into the IB floor approximately 2 ft. away.
- c. Design of jet shields to completely mitigate 2 of the 3 breaks is in process.
6
May 20, 1989 3F0589-20 In conclusion, there are 42 postulated HELB locations outside
-containment, 7 breaks are either shielded or have'no significant jet impingement or pipe whip consequences, 11 breaks are restrained for pipe whip (including all postulated breaks in the 24' MS, 18" FW, 6"
'OTSG blowdown lines),-and 8 postulated breaks are partially shielded by existing jet shield devices.
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I May 20, 1989 i 3F0589-20 f 1
l l ATTACHMENT 4 PIPING ANALYSIS METHODS I
{
i The design basis of piping at CR-3 which is not reactor coolant piping is the USAS Code 331.1 (1967). As part of the review for the I development of the updated HELB criteria, a comparison of the basis behind B31.1 and the more recent ASME Code Section III (1974) was performed. The basis of each Code and a comparison of ASME (1974)
I versus B31.1 (1967) methods for 10 High Energy Piping Analysis problems are discussed in the following paragraphs.
BASIS OF B31.1 (1967)
The fatigue piping component tests of A.R.C. Markl in the 1950's were used as a basis to develop stress intensification factors (SIF's) for B31.1 components. The fatigue tests used a girth buttweld as the baseline component, with other piping components (i.e., elbows, tees, etc.) fatigue lives scaled to the buttweld baseline. The SIF represented " peak stresses" or " secondary stresses", which are considered to be self-limiting stresses from a code viewpoint.
Based on these tests, bending moment loadings for longitudinal thermal expansion stresses were intensified (SIF applied); however, torsional moments were not. Longitudinal pressure stresses were calculated based on classical pipe stress methods. There was no distinction between primary and secondary stresses in this Code (1967).
BASIS OF ASME III (1974)
In 1974, the ASME Code,Section III was released, which departed somewhat from the previous stress calculation methods of the 1971 ASME Code (which parallelled B31.1). The major changes included:
- 1) Update of the SIF tables to reflect better fatigue data for some piping components such as welds, reducers, and branches.
- 2) Adopted intensification of the resultant moment (square root-sum-of-the- squares) as the basis for piping stress.
- 3) Code now discriminates between primary (0.75 i) and secondary (i) stresses in applying the SIF.
- 4) Other minor changes such as use of mean-wall fiber stress, and defining upset loading allowables (primary) to 1.2 Sh.
l Some of these changes would result in higher calculated stresses (items 1 and 2); and some would result in lower calculated stresses (3 i and 4). A direct comparison was required to assess the overall effect on break postulation, since the break threshold stress was also increased from .8 (Sh+Sa) for B31.1 to .8 (1.2 Sh+Sa) for ASME (1974).
8
_ _ _ _ _ = _ _ _ . _ _ _ _
May 20, 1989
.3F0589-20 Longitudinal pressure stress calculational methods were unchanged from the B31.1 code.
COMPARISON OF RESULTS In order to assess the real significance of the ASME Section III Code (1974) changes as compared to B31.1. (1967) , ten (10) high energy line analysis problems were reviewed. This review included (4) main steam, (2) feedwater, and (4) decay heat system problems, with piping sizes ranging from'24 inch schedule 60 to 6 inch schedule 80, including 48 individual piping components. The 48 components selected represented identified crack locations (i.e., highest stress), and included an evaluation of'ASME (1974) stresses (based on intensification of resultant moment) which were then compared to the B31.1 reported stresses from the existing piping analysis packages.
The results show:
. No new break locations when using ASME Section'III (1974) methods.
. Some piping has been qualified to ASME (1974) and then reconciled
'to B31.1. Breaks / cracks were postulated on reported stress levels for-these analysis problems.
. Of 48 components, only 3 had total ASME primary plus secondary stresses larger than the existing B31.1 reported stresses.
Minimum ASME based break margins of at least 20% (i.e., maximum primary plus' secondary stress / break threshold ratios no larger than 0.80)were identified, most below 50%. The ASME calculated stresses did not cause break threshold margins to increase by more than 10%.
. Even though B31.1 methods intensified only bending moments, the inclusion of the 0.751 (primary stress) requirement of ASME (1974) resulted in generally lower ASME (1974) stresses than B31.1 values. The increase break threshold also contributed to overall higher break threshold margins for ASME based break analysis methods.
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.May 20, 1989 3F0589-20 1
l ATTACHMENT.5 INSTRUMENTATION TO DETECT A HELB SYSTEM: MAKE-UP BREAK # ALERTING CONDITIONS MU-1,2 1) Decreasing Letdown Flow MCB Indication (MU FI)
- 2) Dec. Letdown Temp MCB Indication (MU TI)-
- 3) Dec. Prefilter delta p MCB Indication (MU-81-DPI)
- 4) Dec. Letdown Radiation MCB Indication / Alm.
(RM-L1)
- 5) Dec. Pressurizer Lvl. MCB Ind./ Alm. (RC LI1, 2) (RC-1-LS2 )
- 6) Dec. Filter delta p MCB Indication (MU-18-DPI)
- 7) Dec. Makeup Tank Lvl. MCB Ind./ Alm. (MU LIR1) (MU-14-LS1)
- 8) Dec. Makeup Tank Press. MCB Ind./ Alm. (MU PIR1) (MU-17-PS)
MU-5, 6, 7, 8, 1) Dec. Makeup Flow MCB Indication (MU-24-FI).
9, 10, 11, 15, 2) Dec. Total Seal Inj. Flow MCB Ind./ Alm. (MU FI) (MU-27-FS) 16, 17
- 3) Dec. RCP Seal Inj. Flow MCB Ind./ Alm, (MU FI1, 2, 3, 4 3) (MU FS1, 2, 3, 4)
- 4) Dec. Makeup Tank Lvl. MCB Ind./ Alm. (MU LIR1) (MU-14-LS1)
- 5) Dec. Makeup Tank Press. MCB Ind./ Alm. (MU PIR1) (MU-17-PS)
- 6) Dec. Pressurizer Lvl. MCB Ind./ Alm. (RC LI1, 2) (RC-1-LS2) i f
10
May 20, 1989 3F0589-20 SYSTEM: MAIN STEAM POSTULATED BREAK ALERTING CONDITION' MS-1 1. Depressurization of OTSG A - as sensed by MS-106-PT, MS-107-PT, MS-108-PT, and MS-109-PT
- 2. Initiation and alarm of MSLI
- 3. Automatic closure and alarm of MFWLI
- 4. Start and alarm of both EFW pumps
- 5. EFW Vector Logic isolation of bad'OTSG MS-5 Same as MS-1 MS-9 1. Depressurization of OTSG B - as sensed by MS-110-PT, MS-111-PT, MS-112-PT, and MS--
113-PT
- 2. Initiation and alarm of MSLI
- 3. Automatic closure and alarm of MSWLI
- 4. ' Start and alarm of both EFW pumps
- 5. EFW Vector Logic isolation of bad OTSG MS-12 Same as MS-9 MS-41 1. Depressurigation of OTSG A
- 2. Initiation of both EFW pumps on low level in OTSG A MS-42 1. Depressurization of OTSG B
- 2. Initiation of both EFW pumps on low level in OTSG B MS-45 Same as MS-9 MS-46 Same as MS-1 MS-47 Same as MS-1 MS-48 Same as MS-9 ,
i MS-49 Same as MS-9 MS-50 Same as MS-1 MS-51 Same as MS-1 MS-52 & -53 Same as MS-1 1
11
________-__-_______--______D
i May 20, 1989 3F0589 SYSTEM: AUXILIARY STEAM POSTULATED BREAK ALERTING CONDITION AS-1 1. Depressurization of OTSG B - as sensed by MS-110-PT through MS-113-PT
- 2. Initiation and alarm of MSLI
- 3. Automatic closure and alarm of MFWI
- 4. . Start and alarm of both EFW pump
- 5. Vector Logic isolation of bad OTSG AS-5 1. Depressurization of OTSG A - as sensed by MS-106-PT through MS-109-PT
- 2. Initiation and, alarm of MSLI
- 3. Automatic closure and alarm of MFWI
- 4. . Start and alarm of both EFW pumps
- 5. EFW Vector Logic isolation of bad OTSG AS-4 1. Depressurization of OTSG A & OTSG B - as sensed by MS-106-PT through MS-113-PT
- 2. Initiation and alarm of MSLI
- 3. Automatic closure and alarm of MFWI
- 4. Start and alarm of motor driven EF pump
- 5. Alarm that turbine driven EF pump failed to start AS-12 Same as AS-4 12
7,a=.
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May 20, 1989. !
'3F0589 l 1
SYSTEM: MAIN FEEDWATER/ EMERGENCY FEEDWATER - S.G. "B" BREAK ALERTING CONDITIONS FW-14 1. OTSG pressure decrease (MS-110, 111, 112, 113-PT)
- 2. OTSG level decrease (SP-29, 30, 31, 32-LT)
- 3. Initiate EFW & alarm of both EFW pumps started
-4. Increase flow (SP-8B-FI1, FI2)
- 5. High RCS p essure trip
- 6. Turbine trip EF-13 1. OTSG pressure decrease-(same as FW-14)
- 2. Main FW & MS isolation
- 3. High EFW flow to good OTSG
- 4. Alry.a of both EFW pumps started
- 5. EFW Vector Logic isolated bad generator FW-20 1. OTSG pressure increase (same as FW-14) 2.- OTSG level decrease (same as FW-14)_
- 3. Initiate EFW & alarm of both EFW pumps started
- 4. Increase flow (SP-7B-FI1) (SP-12B-DPT)
- 5. Decrease flow (SP-8B-FI1, FI2)
- 6. High RCS pressure trip
~7.- Turbine trip FW-24, FW-32 1. OTSG pressure increase (same as FW-14)
- 2. OTSG 1evel decrease (same as FW-14)
- 3. Initiate EFW & alarm of both EFW pumps started
- 4. Decrease flow (same as FW-20, conditions 4 &
5)
- 5. High RCS pressure trip
- 6. Turbine trip l-13 l
i May.20, 1989
- 3F0589-20 SYSTEM: MAIN FEEDWATER/ EMERGENCY FEEDWATER - S.G. "A" l
BREAK ALERTING CONDITIONS FW-1 1. OTSG pressure decrease (MS-106, 107, 108, 109- .
PT)
- 2. OTSG 1evel decrease (SP-25, 26, 27, 28-LT)
- 3. Initiate EFW & alarm of both EFW pumps started !
- 4. Increase flow (SP-8A-FI1, FI2)
- 5. High RCS pressure trip
- 6. Turbine trip EF-11 1. OTSG pressure decrease (same as FW-1) i
- 2. Main FW & MS isolation
- 3. High EFW flow to good OTSG
- 4. Alarm of both EFW pumps started
- 5. EFW Vector Logic isolated bad generator FW-6, FW-28 1. OTSG pressure increase (same as FW-1)
- 2. OTSG level decrease (same as FW-1) !
- 3. Initiate EFW & alarm of both EFW pumps started
- 4. Increase flow (SP-7A-FI1) (SP-12A-DPT) ;
- 5. Decrease flow (SP-8A-FI1, FI2) l
- 6. High RCS pressure trip '
- 7. Turbine trip i FW-10, FW-31 1. OTSG pressure increase (same as FW-1)
- 2. OTSG 1evel decrease (same as FW-1)
- 3. Initiate EFW & alarm of both EFW pumps started
- 4. Decrease flow (same as FW-6, conditions 4 & 5)
- 5. High RCS pressure trip
- 6. Turbine trip l Refer to Attachment 6 for a discussion of the reliability of the I Makeup System when the breaks are considered.
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14 L. - - - - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
l m I V, ,
May'20, 1989 3 F0589-2 0..
b- ATTACHMENT 6
~HIGH' ENERGY LINE BREAKS j MAKE-UP SYSTEM RELIABILITY l
The.High Energy Line Break (HELB) analysis performed Crystal 1 River
-Unit 3 identified twelve ~-(12) credible break locations, as described-in Impell Report No.- 03-0920-1186,. dated March 1989. None of the
_ identified breaks can prevent the MU System from performing its inventory control function.
The redundancy built into the Make-Up System is sufficient to overcome any challenge to system function presented by a High Energy Line Break.
Two'(2) independent supply lines ensure borated water is supplied to the Make-Up - Pump -(.MUP) suction header.
A' common suction header supplies three (3), 100% capacity MUP's.
A disabled portion of this header.can be isolated'aither remotely or manually, ensuring two (2) of three (3) MUP's are supplied with
. water.
Theithree (3), 100% capacity MUP's discharge to a common header, allowing-any' single pump to supply multiple injection' lines. This common discharge header contains both manual and remotely operated valves which can isolate any disabled section, ensuring three (3) of five (5) potential flow paths are available to supply the Reactor Coolant' System (RCS) with~ borated water.
There are five (5) potential flow paths for borated water into the RCS - four'(4) High Pressure Injection (HPI) lines, and one (1)
Reactor Coolant Pump seal injection line. E Jh of these five (5) lines is independent and can be placed in service or isolated from the control room.
Keeping the redundance built into the Make-Up System in mind, each
. potential break location has been reviewed to discount any credible threat to system function.
Break MU-1:
Potential break location is downstream of the letdown coolers and can be isolated remotely from the control room. .This break is in the-vicinity of the four (4) HPI injection lines and the four (4) remotely operated valves used to place them in service. The possibility of this break disabling all four lines is considered extremely remote.
- The remote operated valves needed to place an injection line'in service range between 4 and 19 feet from the break.
15
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t lMay 20, 1989 3F0589-20
.The four (4) remotely operated valves and the potential break are on slightly different elevations ( 2 ft.).
The potential break would have to drastically change its angle of incidence by 120 0 to impinge upon and disable each remotely operated valve.
The four (4) injection line penetrations span 18 ft., precluding the' possibility of pipe whip impacting and disabling all four (4) lines.
Break MU-2:
This potential break location is slightly downstream of break MU-1, just prior to the black orifice on the 119' elevation. This location can potentially impact MU System components on both the 95' and 119' elevation. There are no MU System components on the 119' elevation associated with RCS inventory control. This break cannot totally disable those MU System components associated with RCS inventory control for the same reasons outlined under break MU-1.
Break MU-5:
This potential break is in the normal make-up line, just prior to penetration #435. The short run of pipe available for whip (approximately 4_ft.) is not sufficient to impact other injection lines (two range from 10 to 16 feet away). Remotely operated valves necessary to place injection lines in service range from 12 to 24 feet away, and are on slightly different elevations (12 ft.). The probability _of this impinging jet changing its angle of incidence by 1200 and disabling remotely operated valves at varying distances is extremely remote.
Breaks MU-6, 7. 8. 9. 10:
This potential breaks MU-7, 8, 9, and 10 are located in the South Auxiliary Building on the 95' elevation, with MU-6 immediately above on the 119' elevation. Break MU-6 can be dismissed, since there isn't any equipment associated with the make-up function on that level.
Breaks MU-7, 8, 9, and 10 located upstream and downstream of the reactor coolant pump seal injection filters cannot credibly disable the four (4) HPI injection lines for the following reasons:
- These four (4) breaks are located in a separate room. Any potential jet would have to travel over 20 feet, with an accuracy of i2 0 just to exit the room through a 3 ft. wide door.
- After leaving the room, the impingir.g jet would have to travel an additional 24 ft. just to contact the nearest injection line.
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May 20, 1989 3F0589-20 The impinging jet can contact only three (3) of four (4) remotely operated valves needed to place an injection line in service, the fourth shield by an existing concrete wall.
The impinging jet would have to travel in excess of 60 feet, with an angle of incidence accurate to less than 10, to contact the i furthest remotely operated valve.
Break MU-11:
This potential break occurs on the 95' elevation of the Auxiliary Building, in the make-up valve room. The break occurs at an anchor, which will lessen the impact of pipe whip, if any. The resulting jet, constrained from changing direction, cannot disable the four (4) supply lines which run through the room. Consider that:
Two of the injection lines are 12 and 18 feet from the break, j respectively, and in different directions. A constrained jet would have to change its angle of incidence by 1800 just to contact both lines.
The valve room contains remotely operated valves used to isolate cections of the common discharge header. Even if the impinging jet disabled both remotely operated valves, two (2) manually
- g. operated valves are available which serve the same function.
These valves may be operated from outside the make-up valve room via remote handwheels, thus, the operator is not in danger.
Breaks MU-15, 16. 17:
These potential breaks are located at the discharge of each make-up pump. While these pumps are located in a common room on the 95' elevation'of the Auxiliary Building, each is separated by a concrete partition. A break at any pump discharge will not disable the remaining two (2) pumps. In addition, all remotely operated valves required to isolate the suction and discharge of the disabled pump are located outside of the make-up pump room.
There is adequate redundance built into the Make-Up System to ensure continued RCS inventory control following any postulated Make-Up System HELB. A single HELB in the MU System is not capable of disabling multiple injection lines, control valves and pumps - the targets are simply too numerous and diversely located.
l 17
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