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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] |
Text
K e# ^
325$0W THIS DOCUMENT CONTAINS Y$$ POOR QUALITY PAGE3 Florida Power -
C09 PO R A T IO N September 4, 1980 File: 3-0-3-a-3 e
Mr. Robert W. Reid Branch Chief Operating Reactors Branch #4 Division of Operating Reactors U. S. Nuclear Regulatory Commission Washington, D. C. 20555
Dear Mr. Reid:
The concern of new instrumentation to detect inadequate core cooling (ICC) has again come into the NRC limelight. This letter serves to reiterate the Florida Power Corporation (FPC) position that no addition-al instrumentation is required. Enclosure 1 is provided as our evalua-tion of instrumentation to detect ICC and supports our position.
Operator guidelines using currently installed instrumentation to detect conditions of ICC were submitted to you in response to NRC staff re-quests (i.e., IE Bulletin 79-05C and Item 2.1.3.b of NUREG-0578).
Specifically, our letter of November 14, 1979, submitted operator guidelines addressing loss of RCS inventory with and without the reactor coolant pumps operating plus guidance on loss of natural cirulation due to loss of the heat sink. These guidelines have been incorporated into plant operating procedures since January 1, 1980. Based upon incorporation of the guidelines plus the incorporation of redundant subcooling meters, we contend Crystal River Unit 3 can be operated safely with the presently installed instrumentation.
A further requirement of NUREG-0578 was to provide a description of any additional instrumentation or controls preposed to supplement the exist-ing instrumentation and controls. By letter dated February 15, 1980, we summarized our evaluation of instrumentation giving an unambiguous, easy-to-interpret indication of ICC. Our conclusion reached was no additional instrumentation could better detect ICC conditions should they be allowed to occur.
FPC has yet to receive comments on the ICC quidelines or the evaluation of additional instrumentation. Therefore, no further actions will be taken by FPC until Staff review and concurrence with the previous sub-mittals is obtained. It is also our position we have met the require-ments related to ICC as set forth in NUREG-0578 and IE Bulletin 79-05C.
8009090 4 ? [ So -JO A General Office 3201 inirty-fourin street soutn . P O Box 14042, st Petersburg. Fionaa 33733 e 813-866 5151
- 7 m-4
,7 R.'W Rei d '.-
Page Two ' September _4, 1980-
' Again, Enclosure provides the basis for our position on additional
.. instrumentation - to detect -ICC. We urge the Staff to review - this and oreviously. submitted documents ~ prior to . issuance of any further requests.--
Very truly yours, FLORIDA POWER. CORPORATION y%
Patsy . Baynard Manager ..
Nuclear Support Servf ces Enclosure Baynard(WO6)D1-1
.7
.e r
e i
? ,f e
- 9
= STATE OF FLORIDA COUNTY OF-PINELLAS P. - Y. Baynard ' states that she' is the Manager, Nuclear Support Services Department of Florida ~ Power Corporation; that she is authorized on the part of said company to sign and file with the Nuclear Regulatory Com-mission the. information attached hereto; and that' all such statements made and matters set forth therein are true and correct to the best of her knowledge, information and belief.-
.WNm2
-d(/ P. Y., Saynard Subscribed and. sworn to before me, a Notary Public in and for the State and County above named, this 4th day of September,1980.
haLAAO2 h Yb= Ag
(/ Notary Public /
- Notary P'ublic, State of. Florida at Large,
- My Commission Expires: June 8, 1984 PYB/MAHNotary(DN-98)
{ - ~
- x. h.
G ENCLOSURE 1 EVALUATION OF INSTRUMENTATION TO DETECT INADEQUATE CORE COOLING Baynard(WO6)Dl-1 I l
1 l
L
- TABLE OF CG1 TENTS 3
PAGE
1.0 BACKGROUND
1 2.0 DEFINITION OF INADEQUATE CORE COOLING 2 3.0 OPERATOR GUIDELINES FOR INADEQUATE CORE COOLING 3
=>
4.0 DISCUSSION OF METH005 TO DETECT INADEQUATE CORE COOLING 5 4.1 Core Outlet D1ermocouples 5
'~
. 4.2 Axial Incore Thermocouples 6 4.3 Ultrasonic Techniques 6
, 4.4 Neutron and Gama Beams 7 l
4.5 Differential Pressure Transmitters 8
5.0 CONCLUSION
S 10 APPENDIX A - NUREG-0578 POSITION ON INSTRUMENTATION A-1 FOR DETECTION OF INADEQUATE CORE COOLING A-2
' AND CLARIFICATION FROM H. R. DENTON'S
, LETTER OF OCTOBER 30, 1979 TABLE 4.0 PROPOSED INADEQUATE CORE COOLING INDICATIONS COMPARED TO ESTABLISHED CRITERA FIGURE 3.1 CORE EXIT THERMOCOUPLE TEMPERATURE FOR INADEQUATE CORE COOLING FIGURE 4.1 LAYOUT OF CORE THERMOCOUPLE FIGURE 4.1 LOCATION OF THERMOCOUPLE I
i
- i -
De W -
. ,- .m . r, - -- ,- _. r , ~ -
r
1.0 BACKGROUND
The major concerns raised in the aftermath of the TMI-2 accident were 7 identified in the "TMI 2 LESSONS LEARNED TASK FORCE STATUS REPORT, NUREG-0578". Section 2.1.3.b of that report addressed additional
- p instric.entation which could assist in the detection of inadequate core cooling. The NRC position on additional instruentation was that
.s."
" licensees shall provide a description of any additional
.O instr uentation or controls (primary or backup) prdposed for the plant .... giving an unambiguous, easy-to-interpret
.; indication of inadequate core cooling."....
-t.
7, Subsequently, the NRC's position was clarified and amplified in Enclosure 1.
I to H. R. Denton's letter of October 30, 1979 to all operating nuclear power plants entitled " Discussion of Lessons Learned Short Term ,
..., Requirements". This letter addressed the following requirements for any l.. - additional instr uentation proposed. (The complete clarification is reproduced in Appendix A.) .
- a. Design of new instrumentation should provide an unambiguous indication of inadequate core cooling. l
- b. The indication should have the following properties:
], (1) It must indicate the existence of inadequate core cooling '
1
,. caused by various phenomena.
- 1. (2) It must not erroneously indicate core cooling because of i-the presence of an unrelated phenomena. l k*
- c. The indication must give advanced warning of the approach of inadequate core cooling.
- d. The indication must cover the full range from normal operation l
, 1 g to complete core uncovering. l 9
H. R. Denton's letter of October 30, 1979, clarificd the requirements that any investigation of additional instrunentation include an evaluation of
.. reactor water level indication.
In responsa +c NUREG-0578 B&W has developed operator guidelines for action to recover from a condition of inadequate core cooling using existing instrunentation (References 1-5). The evaluation provided in the following sections reviews the adequacy of existing and prop'osed
( instrunentation to indicate inadequate core cooling (ICC). To perform this review, it is important to understand when ICC actually occurs, what ,
1 operator actions occur prior to ICC, and the guidelines followed once ICC ;
has occured. The next two sections describe ICC and the actions taken before and af ter ICC is indicated. These sections are then followed by a comparison of existing and proposed equipment for indicating ICC which conclude with a section describing why the existing installed 3
instrunentation provides the best indication.
2.0 DEFINITION OF INADEQUATE CORE COOLING 1
8 In a depressurization event, the reactor coolant system (RCS) must first reach saturation conditions before there is any danger of inadequate core cooling. Subsequently if the RCS inventory is reduced and uncovery of the core begins, temperatures in the uncovered region will increase causing superheating. It is important to note in this discussion that inadequate Eb' 4 fsc^- l core cooling deet. not begin until3reactor vessel (RV) water inventory i l
f alls below the top of the core thus resulting in an increasing fuel clad temperature.
i a== .
- n. 3.0 OPERATING PHILOSOPHY AND GUIDELfNES FOR INADEQUATE CORE COOLING i
The gotis of the operator prior t' ICC are different than those once ICC I has occured. Prior to an indication that ICC has occured, the operator is taking actions which will stabilize pressure and refill the RCS. The goal is to re-establish the subcooling margin at the high pressure condition or
.. cooldown and depressurize to low pressure injection plant conditions.
Indication that ICC has occured changes the operator's guidance because the goal of refilling at the high pressure cannot be attained. The operator at this point is instructed to partially depressurize using the f PORY to increase RCS inventory addition rate. Note: If this fails the b
operator is instructed to further depressurize and' establish low pressure l injection (LPI). These last two steps are based on conscious decisions
, that recovery at the higher pressure is not possible and that ,
. - depressurization will cause more immediate core voiding, but in the longer r term will result in improved core cooling by increased RCS inventory.
I Based on this logic it is important that the indication not be ambiguous and not occur prematurely. It is important to provide as much time as ,
l f' possible for recovery at the higher pressure which leads to the preferred mode of operation.
Synptoms of an overcooling transient are similar to the small break loss of coolant transient up to the point of inadequate core cooling. At this 1
point, if the operator has taken actions for inadequate core cooling when in fact overcooling exists, an unnecessary serious transient would result.
Thus, the operator must not proceed with the inadequate core cooling actions until ir.sdequate core cooling is confirmed.
l
- . The following sections describe the actual operator actions taken prior to ICC and those once ICC is indicated.
=
3.1 Operator Actions During Approach to ICC Operator actions during the approach to an inadequate core cooling condition are stamarized as follows:
,.- 1. Initiate HPI
- 2. Maintain OTSG level .
- 3. Trip RC pumps if ESFAS initiated ay low RC pressure
- 4. Monitor incore thermocoupie temperatures to determine if inadequate core cooling exists.
These actions are verified when saturation conditions exist. No further
,, actions are taken until thermocouple temperatures reach a predetermined
- l. temperature from Small Break Operating Guidelines (see Figure 3.1-1, Curve 1). This indicates that superheating is ccurring, that fuel clad ;
temperature has increased 'above saturaction and that inadequate core a cooling exists. l
~
3.2 Operator Actions Once ICC is Indicated Once inadequate core cooling is indicated the operator is instructed to 1 take the following actions:
, 1. Start one RCP per loop
- 2. Depressurize operative OTSG(s) to 400 psig as rapidly as possible
!- 3. Open the PORV to maintain RCS pressure within 50 psi of OTSG pressure
- 4. Continue cooldown by maintaining 100*F/hr decrease in secondary saturation temperature to achieve 150 psig RCS pressure FIGURE 3.1-1 CORE EXIT THER4* 0 COUPLE TE'.iPERATURE FOR INADEQUATE CC11E COOLING l
i
. 1200 - -
. 1100 -
CURVE #2 B T TCLAD LESS THAN 1500*F 5 1000 -
?
E
,, E 900 -
E O
CURVE.#1 v
. \
u '
T CLA0 LESS THAN 1400 F l
\
i 700 -
1 i
T
- 600 -
500 -
d 400 ' ' ' - -
200 600 1000 1400 1800 2200 Pressure, psia
-- , ,n ,- -,-,,~v - -
w-, -e- - * , , ,ww- e w, ----,w,,e,, uer ,w-r-w----- -,v ,- v-~,-- --,--, -
L These actions are taken to reduce RC pressure thus increasing HPI flow and RCS inventory addition rate. If thermocouple temperature continues to e
I rise above a higher predetermined temperature which indicates a
_ significant increase in fuel clad temperature (see Figure 3.1-1, Curve 2)
I the operator should:
.. 1. Start all RCPs I
' 2. Depressurize OTSG(s) to atmospheric pressure
- 3. Open the'PORV to depressurize the RCS and allow LPI to restore core cooling i
4.0 DISCUSSION OF METHODS TO DETECT INADEQUATE CORE COOLING The following methods of indicating core cooling were examined in this F
', , evaluation: 1 i
~
i
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- 1. Existing core thermocouples l
- 2. Additional axial core thermocouples
[-
- 3. UltrasonicRVleveiindication
- 4. Neutron or gamma beam RV level indication
- 5. Differential pressure (dp) transmitters for RV level inoication The capabilities and evaluations associated with each type of indication
, are discussed below. Table 4.0-1 provides a summary of the methods and l their capabilities.
)
l- 4.1 Core Outlet Thennocouples The existing core thermocouple instruments indicate inadequate core cooling when interpreted using the operator guidelines of Referen -
2 and 3. The location of these thermocouples provides indication d :" ply 1
increased temperatures at the top of the core when the top of the core ,
I
. reaches conditions of inadequate cooling. The locations of the i thermocouples in the core and fuel assembly are shown on figures 4.1-1 and 4.1-2.
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- g ' Figure 4.1-1 Layout of Core Thennocouoles a e, INLET y INLET B
- p. . INTERilEDI ATE // /
RANGE DETECTOR SOURCERANGEy 3
, DETECTOR . ,. .
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DETECTOR ff Y A INLET INLET @:CCNTROLR00 SOURCE: K e ; INSTRUMENT TUBE (52 THERMOCOUPLES)
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Figure, 4.1-2 Location of Themoccuole 4
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m 4.2 Axial Incore Themocouples Additional thermocouples installed axially in the incore instrument guide
-- tube will provide an indication of the extent of inadequate core cooling; but, an indication that the middle of the core is inadequately cooled will not elicit any further operator action over and above the actions taken when the top of the core indicates inadequate core cooling. There would be no change in operator guidance even if this thermocouple information I
were available.
,' 4.3 Ultrasonic Techniques Several methods of ultrasonic techniques were considered. These included using existing internal structures as wave guides, installing an
,, externally excited ultrasonic vibrating rod and installing a head tr.ounted 1
, transducer. In simple applications, all of these methods have been
. proven. However, in the reactor vessel the core provides a heat source l
which changes the density of the fluid. The fluid changes state from a single phase liquid to a two phase fluid, and finally to a single phase vapor. Ultrasonic level measurement techniques are frequently used where there is a sharp density change at the fluid interface. The ?evel created in a reactor vessel as a result of a LOCA will be a frothy, two-phase mixture height rather than a fixed phase interface. The variable density change will not provide an easy-to-interpret indication, and could provide an ambiguous output signal.
l i
The ambiguous signal could lead the operator to believe that the core was inadequately cooled when in fact sufficient heat transfer was causing the frothy condition and adequate cooling was in progress. As a consequence of the incorrect belief, the operator would take the incorrect actions of depressurizing the RCS.
i
~. .
4.4 Neutron and Gamma Beams Neutron and gama beams have been used successfully to determine the level of fluid in a vessel. The application of. this method to a RV level would be the use of the core as a source and use the existing out of core detectors to monitor the water level through changes in count rate.
1 Normally, the detector count rate decreases at rates characteristic of the
- ' various mechanisms of neutron production that exist following a reactor i trip.
.. One concept of water level measurement uses the installed source range
,, detectors which respond to a decrease in water density. As water level decreases, the detector output increases. However, if the water level decreases to below the top of the core, the detector output decreases.
The intensity of the neutron beam and thus detector output would be very dependent on previous power history, thus requiring calibration prior to eac,h use of the instrunent. This is not reasonable during accident conditions. For this reason, further investigation of this method was terminated. A more detailed discussion of the application of this nuclear radiation method is included in Reference 6.
(
Another concept of RV water level measurement system has been tested at three reactor sites. The system employs BF3 neutron detectors above and below the reactor vessel. Data was collected and extrapolated to determine neutron count rate between one and six days af ter shutdown as a function of water level above the core. The data showed a relatively slow increase in count rate as the water level decreased from a full condition, n , with a marked increase in count rate when the water level reached five feet above the top of the core. At this level water was still above the hot leg nozzles. This indication system is capable'of providing a discrete data point indicating that reactor vessel level is five feet
r- '
l
., above the core. Evaluation of tha remaining data requires interpretation I by the operator to determine the correct reactor vessel water level. The capability of this instrunent must be evaluated imediately af ter a
- f shutdown to show its effectiveness in a high background level which would be the case following a LOCA.
4.5 Differential Pressure Transmitters
., The use of differential pressure transmitters to measure reactor vessel L- level was considered. Three level measurement ranges, one across the r reactor vessel, a second across the hot leg, and one combining these ranges, were evaluated. '
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s.
The first, a reactor vessel differential pressure (dp) measurement, would require new penetrations in an incore nozzle at the bottom of the reactor
,... vessel and at the top in a control rod drive mechanism (CRDM) closure. An L instrunent could be installed to provide a differential pressure between the bottom of the core and the top of the reactor vessel, but the differential pressure (dp) would be affected by not only the water level
[
4 head, but also by shock loss, friction loss, and flow acceleration loss.
During forced flow conditions, the shock loss, friction loss, and flow j acceleration loss terms dominate the signal.
8 Additionally, the magnitude of these terms varies depending on the density, and thus riowrate, of the pumped fluid. Due to the changing magnitude of these terms, it is not possible to compensate the dp signal i to achieve a water level from head only. During stagnant boiloff, the j 1 i decay heat in the core will cause the level of coolant in the core region to swell to a level greater than that in the downcomer region of the reactor vessel. A dp level measurement would measure the collapsed level !
l L in the-downcomer region. A swelled level of 12 feet might be indicated by a collapsed level of between 7.4 and 8.625 feet, depending on system i I
l
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pressure.
Th3 unpredictable peak power distribution and decay heat level
,P , preclude compensating the dp signal for this error. Although the parameter of interest in this case is the mixture height, the dp cell
-l would measure a collapsed level which means that under some conditions s .-
this signal would be ambiguous, and could lead to premature
] depressurization of the plant by the operator's misinterpretation of the indication.
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l'
,, The second method, a hot leg differential pressure measurement would l, require new penetrations at the bottom of the hot leg and the vent line at
.. the top of the hot leg. This instrunent would provide a dp tignal and not
(. an actual water level. In this instance, measuring any water level woulo 7- be a valid indication that the core was covered. During flow conditions, the output signal wuld be affected by the same effects as the reactor
[- vessel dp signal discussed above. However, the hot leg dp signal could be l- temperature compensated. The fact that the hot leg contains coolant would
.} ~ indicate that the core was covered and thus no new operator actions for inadequate core cooling would be required. However, if the operator takes
) actions for inadequate core cooling based on only a level in the hot leg l then he would be taking incorrect actions for some casualties which could 1
also be indicated by a level in the hot leg; i.e., overcooling, partial l L.
steam voiding in the hot leg caused by transients.
The third method, a differential pressure measurement from the bottom of the reactor vessel to the top of the hot leg, would require new penetrations in an incore nozzle at the bottom of the reactor vessel and
, at the vent line at the top of the hot leg. This range is a comt ination of the two previou,s instrument ranges. It provides an advantage over the hot leg level measurement in that it can measure the entire RV level span, but it would still exhibit the same ambiguity as the reactor vessel dp P. .
L described earlier. In addition, due to the greatly expanded range, the inaccuracy of the instrunent'would be greater, perhaps as large as + 4.0 feet. This measurement would be inaccurate in the hot leg range and would
]. be ambiguous in the reac. tor vessel range as discussed above.
All three methods of dp level measurement require additional :tructural penetrations or modifications. Addf tionally, the operator would not be
~
directed to take action until he ccnfirmed the existence of inadequate s.
core cooling with the core exit trermocouple, thus these additions would not change any operator guidance.
L
5.0 CONCLUSION
S As has been discussed, no proposed method of indication of inadequate core L cooling would meet all the established criteria. The introduction of l- ambiguous information provided by some proposed systems of inadequate core cooling indication would cause operator confusion. This confusion could lead to incorrect and unsafe actions in some situations; i.e., premature depressurization during LOCAs, or incorrect actions during overcooling events.
Reliance on existing core exit thermocouples and previously published operator guidelines for interpreting the available information is the best and most direct method of determining that the inadequate core !
cooling condition has ocurred. The existing instrunentation in the B&W designed nuclear steam supply system is able to detect inadequate core cooling. The incore thermocouples provide an unambiguous indication of the existence of inadequate core cooling, and will not erroneously
_ indicate inadequate core cooling. The thermocouples provide the most discriminating capability of defining the existence of inadequate core cooling.
l' e
- l The basis for this conclusion is further suppcrted by the following:
The recently installed T meter provides a long term sat )
-- 1 indication of the approach to inadequate core cooling since saturation conditions must be achieved prior to the onset of l
[i inadequate core cooling. Saturation conditions would be reached a 1 significant time before inadequate core cooling, th T the operator !
I L. would be alerted to'the condition. e )
The existing core thermocouples will indicate the imediate approach, the existence of and termination of the inadequate core cooling condition.
The instrisnents will ensure direct, appropriate interpretation of
{ plant conditions by the operator when used in conjunction with previously published operator guidelines.
Each proposed reactor vessel level measurement system concept fails to pr' ovide any additional aid to the operator for detection of inadequate core cooling. Core cooling is directly indicated by temperature measurement, not level measurement. Secondly, each of 4
the level measurement concepts fails to meet all of the established criteria as outlined in Table 4.0-1.
The potentially ambiguous information previoed by the proposed RV level indication instrtsnent systems could lead to unsafe ano '
incorrect actions if the operator acted on the level indication.
No new or additional detectors are required to cover the full range of plant conditions. i Adequate core cooling is determined by core heat removal capabilities. It is directly indicated by the reactor coolant system temperature / pressure relationship. The approach to inadequate core cooling is indicated in sufficient time by the T meter to allow the operator to take mitigating action. If 1
l 1
.7 '
. l' '
his actions are unsuccessful and inadequate core heat removal
- [f' conditions exists, sufficient indication for the operator is
-- 'available'by means of the core thermocouples. As superheated conditions are reached the thermocouple temperature will increase.
If additional operator actions of partial depressurization of the RCS are successful and he can regain control of the core heat removal, the thermocouple indication will provide the necessary
{'
feedback to tell him that his actions were effective.
T a/
- l. 'paF 9
. c. It is B&W's g technical judgement that the existing plant sensors provide a
' !-- reliable and accurate method of detecting the approach to and existence of inadequate core cooling for all modes of plant operation.
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TABLC 4.0-1 1[ VEL MEAE":tMtNT MCIH00 IAllCH MfEI EXI5TllIG Citi!!stI A
~ ~" --~'
LEVEL MEASINtfM(NT METH005 '~~
. CRITDelA ' TxTst ing-' ~~AW4it iosF iieutron or Nanked in Order of BW Subcooling incore incor.* Ultra- Gamed ibt leg itV Awigned Priority _,__ Monitor f/C T/C Sonics Sem* $PIIS level AP
- 1. Must be direct indica- X X tion of ICC
- 2. Unambiguous - not X X erroneously indicate
!CC
- 3. Cover full range from X X X normal operation to core uncovery
- 4. Provide advanced warn. X X X X ing of ICC
- 5. Isnanbiguous _ indicate ICC during pumped high X X void fraction and stagnant bulloff
- 6. Nu major structural X X X X th4nges to plant I. Ilnenbiguous - meects safety grdde criteria **
I alkvelop usik is still required to prove capability of this method inanediately af ter shutthsen.
- %t ate-est the-as t hardware to meet saf ety grade criteria is siot available to comply with the schedule installation d4tc.
~, .
l APPENDIX A p*
,,, NUREG-0578 POSITION ON INSTRUMENTATION FOR DETECTION
- 0F INADE00 ATE CCRE COOLING AND CLARIFICATION FROM H. R. OENTON'S LETTER OF OCTOBER 30, 1979 POSITION Licensees shall provide a description of any additional instrunentation or controls (primary or backup) proposed for the plant to supplement those devices cited in the preceding section giving an unambiguous, easy-to-interpret
, indication of inadequate core cooling. A description of the functional design L. requirements for the system shall also be included. A description of the r- procedures to be used with the proposed equipment, the analysis used in developing these procedures, and 'a schedule for installing the equipment shall be provided.
CLARIFICATION
- 1. Design of new instrunentation should provide an unambiguous indication of
, inadequate core cooling. This may require new measurements to or a
. synthesis of existing measurements which meet safety-grade criteria.
- 2. The evaluation is to include reactor water level indication.
- 3. A comitment to provide the necessary analysis and to study advantages of various instrunents to monitor water level and core coolng is required in the response to the September 13, 1979 letter.
, . 4. The indication of inadequate core cooling must be unambiguous, in that, it
.should have the following properties:
,, a) it must indicate the existence of inadequate core cooling caused by various phenomena (i.e. , high void fraction pumped flow as well as stagnant boil.off).
b) It must not erroneously indicate inadequate core cooling because of.the presence of an unrelated phenomenon.
A-1
. _ _ . . _ . _ _ _ _ .__a
, - AkPENDIX A (Cont'd)
- 5. The . indication must give advanced warning of the approach of inadequate
{ core cooling.
, . , 6. The indication must cover the full range from normal operation to I complete core uncovering. For example, if water level is chosen as the 2-- unambiguous indication, then the range of the instrument (or
}
instruments) must cover the full range from normal water : level to the bottom of the core.
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e P*
4 1
A-2
, REFERENCES i
- 1. Small Break Operating Guidelines, B&W Document 69-1106001-00, November 1979
- 2. Small Break Operating Guidelines, B&W Document 69-1106003-00, November 1979
- 3. Small Break Operating Guidelines, B&W Document 69-1106002-00, November 1979 3
- 4. Inadequate Core Cooling Decay Heat Removal System Mode of Operation, B&W Docunent 69-1106921-00, December 1979
- 5. Inadequate Core Cooling - DNB at Power, Site Instruction 3/4/9/187, 5/355, 7/364, S/172, 11/191, 14/402 dated December 21, 1979 .
"~
- 6. Analysis Sunnary in Support of Inadequate Core Cooling Guidelines, B&W
, Document 86-1105508-01, December 5, 1979 r-u k.
09
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