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Category:CORRESPONDENCE-LETTERS
MONTHYEARBSEP-99-0170, Forwards Proprietery Notification of Change in Operator Status.Individual Name,Docket Number,License Number & Effective Date of Expiration,Encl.Proprietary Info Withheld, Per 10CFR2.790(a)(6)1999-10-19019 October 1999 Forwards Proprietery Notification of Change in Operator Status.Individual Name,Docket Number,License Number & Effective Date of Expiration,Encl.Proprietary Info Withheld, Per 10CFR2.790(a)(6) BSEP-99-0161, Submits Response to NRC RAI Re Relief Request VRR-03.VRR-03 Requested Relief from full-stroke Open Exercise Requirements for Supply Check Valves to air-operated Valves 1(2)-RNA-V313,1(2)-RNA-V314,1(2)-RNA-V350 & 1(2)-RNA-V3511999-10-13013 October 1999 Submits Response to NRC RAI Re Relief Request VRR-03.VRR-03 Requested Relief from full-stroke Open Exercise Requirements for Supply Check Valves to air-operated Valves 1(2)-RNA-V313,1(2)-RNA-V314,1(2)-RNA-V350 & 1(2)-RNA-V351 ML20217G1191999-10-0808 October 1999 Informs That on 990930,NRC Staff Completed mid-cycle PPR of Brunswick NPP & Did Not Identify Any Areas Where Performance Warranted More than Core Insp Program.Nrc Plan to Conduct Only Core Insps at Facility Over Next Five Months ML20217C6931999-10-0606 October 1999 Forwards Insp Repts 50-324/99-06 & 50-325/99-06 on 990801- 0911 at Brunswick Reactor Facility.No Violations Were Identified BSEP-99-0114, Forwards Supplemental Biological Assessment Submitted by CP&L Providing Updated Data for 1998 & 1999. Pictures of Intake Canal at Diversion Structure During High Tide Conditions Also Encl1999-10-0404 October 1999 Forwards Supplemental Biological Assessment Submitted by CP&L Providing Updated Data for 1998 & 1999. Pictures of Intake Canal at Diversion Structure During High Tide Conditions Also Encl BSEP-99-0157, Submits Annual Rept Summarizing Effect of Changes & Errors in Accepted loss-of-coolant Accident ECCS Evaluation Models Applicable to Bsep,Units 1 & 2 IAW 10CFR50.46(a)(3)(ii)1999-10-0404 October 1999 Submits Annual Rept Summarizing Effect of Changes & Errors in Accepted loss-of-coolant Accident ECCS Evaluation Models Applicable to Bsep,Units 1 & 2 IAW 10CFR50.46(a)(3)(ii) BSEP-99-0142, Forwards Proprietary Updated List of Home Addresses for Individuals Licensed to Operate CP&L Bsep,Units 1 & 2. Proprietary Info Withheld,Per 10CFR2.790(a)(6)1999-09-23023 September 1999 Forwards Proprietary Updated List of Home Addresses for Individuals Licensed to Operate CP&L Bsep,Units 1 & 2. Proprietary Info Withheld,Per 10CFR2.790(a)(6) BSEP-99-0158, Provides Proprietary Notification of Change in Operator Status.Individual Name,Docket Number,License & Effective Date of Expiration Provided in Encl to Ltr.Proprietary Info Withheld,Per 10CFR2.790(a)(6)1999-09-21021 September 1999 Provides Proprietary Notification of Change in Operator Status.Individual Name,Docket Number,License & Effective Date of Expiration Provided in Encl to Ltr.Proprietary Info Withheld,Per 10CFR2.790(a)(6) BSEP-99-0153, Forwards Monthly Operating Repts for Brunswick Steam Electric Plant,Units 1 & 2.CP&L Is Submitting Encl Rept of Operating Statistics & SD Experience for Aug 19991999-09-14014 September 1999 Forwards Monthly Operating Repts for Brunswick Steam Electric Plant,Units 1 & 2.CP&L Is Submitting Encl Rept of Operating Statistics & SD Experience for Aug 1999 BSEP-99-0147, Forwards Response to NRC Telcon RAI Re Relief Requests RR-5, 12,13,14,22,23,24 & 25,per Inservice Insp Program for Third 10-yr Interval1999-09-14014 September 1999 Forwards Response to NRC Telcon RAI Re Relief Requests RR-5, 12,13,14,22,23,24 & 25,per Inservice Insp Program for Third 10-yr Interval BSEP-99-0151, Notifies NRC That Unit 1 Digtial FW Control Sys Upgrade Was Completed on 990831 During Plant Sd.Action Completes Y2K Remediation Activities for BSEP1999-09-0808 September 1999 Notifies NRC That Unit 1 Digtial FW Control Sys Upgrade Was Completed on 990831 During Plant Sd.Action Completes Y2K Remediation Activities for BSEP BSEP-99-0149, Forwards Two Proprietary License Renewal Applications Consisting of NRC Form 398 & NRC Form 396,for Operators Licensed at Plant.Proprietary Encls Withheld1999-09-0202 September 1999 Forwards Two Proprietary License Renewal Applications Consisting of NRC Form 398 & NRC Form 396,for Operators Licensed at Plant.Proprietary Encls Withheld BSEP-99-0150, Forwards Proprietary Info Re Positive Drug Test for Operator Licensed on CP&L Bsep,Units 1 & 2,in Response to NRC .Proprietary Info Withheld,Per 10CFR2.790(a)(6)1999-09-0202 September 1999 Forwards Proprietary Info Re Positive Drug Test for Operator Licensed on CP&L Bsep,Units 1 & 2,in Response to NRC .Proprietary Info Withheld,Per 10CFR2.790(a)(6) BSEP-99-0112, Forwards Summary of Exam Results for Feedwater Sparger & Nozzle Exams Performed During RFO13.Evaluation of Exam Results Documented in ESR 98-00333, Unit 2 Feedwater Sparger Evaluation Based on B214R1 IVVI Exam Results, Encl1999-09-0101 September 1999 Forwards Summary of Exam Results for Feedwater Sparger & Nozzle Exams Performed During RFO13.Evaluation of Exam Results Documented in ESR 98-00333, Unit 2 Feedwater Sparger Evaluation Based on B214R1 IVVI Exam Results, Encl ML20211K9001999-08-27027 August 1999 Forwards Insp Repts 50-324/99-05 & 50-325/99-05 on 990620-0731.Two Violations Occurred & Being Treated as NCVs BSEP-99-0146, Forwards Revised EPIPs for Bsep,Units 1 & 2.List of Revised Procedures & Summary of Changes Are Encl1999-08-26026 August 1999 Forwards Revised EPIPs for Bsep,Units 1 & 2.List of Revised Procedures & Summary of Changes Are Encl BSEP-99-0148, Forwards Rev 0 to NF-908.03, Brunswick Unit 2,Cycle 14 Neutronics Startup Rept, IAW Section 13.4.2.1 of UFSAR for Brunswick Steam Electric Plant,Units 1 & 21999-08-25025 August 1999 Forwards Rev 0 to NF-908.03, Brunswick Unit 2,Cycle 14 Neutronics Startup Rept, IAW Section 13.4.2.1 of UFSAR for Brunswick Steam Electric Plant,Units 1 & 2 BSEP-99-0141, Forwards Proprietary Info Re Expiration of Operator Licenses SOP-20812-1,OP-20977-1 & OP-21217,per 10CFR50.74(b) Re Notification of Changes in Operator Status.Proprietary Info Withheld,Per 10CFR2.790(a)(6)1999-08-25025 August 1999 Forwards Proprietary Info Re Expiration of Operator Licenses SOP-20812-1,OP-20977-1 & OP-21217,per 10CFR50.74(b) Re Notification of Changes in Operator Status.Proprietary Info Withheld,Per 10CFR2.790(a)(6) BSEP-99-0130, Requests Relief from ASME Boiler & Pressure Vessel Code, Section Xi,Iaw 10CFR50.55a(g)(5)(iii) & NRC GL 90-05, Guidance for Performing Temporary Non-Code Repair of ASME Code Class 1,2 & 3 Piping. Regulatory Commitments,Encl1999-08-25025 August 1999 Requests Relief from ASME Boiler & Pressure Vessel Code, Section Xi,Iaw 10CFR50.55a(g)(5)(iii) & NRC GL 90-05, Guidance for Performing Temporary Non-Code Repair of ASME Code Class 1,2 & 3 Piping. Regulatory Commitments,Encl BSEP-99-0132, Requests Delay in Providing Update on Util Intended Actions Re GL 96-06,until 180 Days After NRC Approval of Generic Technical Basis.List of Regulatory Commitments,Encl1999-08-20020 August 1999 Requests Delay in Providing Update on Util Intended Actions Re GL 96-06,until 180 Days After NRC Approval of Generic Technical Basis.List of Regulatory Commitments,Encl BSEP-99-0134, Forwards Monthly Operating Repts for July 1999 for BSEP, Units 1 & 2.Revised Rept for June 1999 for Unit 2,reflecting Info on 990628 Forced Shutdown of Plant,Encl1999-08-13013 August 1999 Forwards Monthly Operating Repts for July 1999 for BSEP, Units 1 & 2.Revised Rept for June 1999 for Unit 2,reflecting Info on 990628 Forced Shutdown of Plant,Encl ML20210S9061999-08-11011 August 1999 Informs That GE Document Entitled, Addl Info Regarding 1.09 Cycle Specific SLMCPR for Brunswick Unit 1 Cycle 12, Will Be Withheld from Public Disclosure Pursuant to 10CFR2.790(b)(5) & Section 103(b) of Atomic Energy Act BSEP-99-0128, Forwards Revised Relief Request RR-17,applicable to Remainder of Third 10-year ISI Program for Plant1999-08-11011 August 1999 Forwards Revised Relief Request RR-17,applicable to Remainder of Third 10-year ISI Program for Plant ML20210S9121999-08-11011 August 1999 Informs That Document NEDC-31624P,Suppl 2,Rev 6,entitled, Loss-Of-Coolant Accident Analysis Rept for Brunswick Steam Electric Plant Unit 2,Reload 13,Cycle 14 Will Be Withheld from Public Disclosure Pursuant to 10CFR2.790(b)(5) ML20210P8221999-08-11011 August 1999 Advises That Info Contained in Which Submitted Document,Prepared by Siemens Power Corp,EMF-2168(P) Rev 0, Marked Proprietary,Will Be Withheld from Public Disclosure, Per 10CFR2.790(b)(5) & Section 103(b) of AEA of 1954 ML20210P9381999-08-10010 August 1999 Forwards SE Accepting Licensee & Suppls & 0517,which Submitted Assessment of Impact on Operation of Unit 1 with Crack Indications of 2.11, 6.36 & 1.74 Inches in Three Separate Jet Pump Risers ML20210P8911999-08-10010 August 1999 Forwards SE Authorizing Relief Requests CIP-01,02,06,07,08, 09,10,11 (with Certain Exceptions) & 12-18,for Second 10-year ISI Interval.Requests CIP-04 & 05 Would Result in Hardship,Relief CIP-03 Not Required & CIP-11 Denied in Part BSEP-99-0110, Forwards Rev 0 to ESR 99-00279, B214R1 RPV Hydrotest Bolted Connection Corrective Action Evaluation, IAW Subparagraph IWB-3144(b) of 1989 Edition of ASME B&PV Code, Section XI1999-08-0505 August 1999 Forwards Rev 0 to ESR 99-00279, B214R1 RPV Hydrotest Bolted Connection Corrective Action Evaluation, IAW Subparagraph IWB-3144(b) of 1989 Edition of ASME B&PV Code, Section XI BSEP-99-0123, Forwards Rev 54 to Radiological Emergency Response Plan (RERP) & Revised Epips,Including Re 44 to OPEP-02.1,rev 4 to OPEP-02.1.1 & Rev 7 to OPEP-02.6.26.Summary of Changes Provided in Encl 21999-08-0505 August 1999 Forwards Rev 54 to Radiological Emergency Response Plan (RERP) & Revised Epips,Including Re 44 to OPEP-02.1,rev 4 to OPEP-02.1.1 & Rev 7 to OPEP-02.6.26.Summary of Changes Provided in Encl 2 ML20210Q4581999-08-0505 August 1999 Informs That NRC Plans to Administer Generic Fundamentals Exam Section to Written Operator Licensing Exam on 991006. Authorized Representative of Facility Must Submit List of Individuals to Take exam,30 Days Before Exam Date ML20210P2031999-08-0505 August 1999 Discusses Staff Response to GL 92-01,Rev 1,Suppl 1, Rv Structural Integrity for Plant,Units 1 & 2 ML20210N2041999-08-0505 August 1999 Forwards SE Accepting Response to GL 87-02, Verification of Seismic Adequacy of Mechanical & Electrical Equipment in Operating Reactors,Unresolved Safety Issues (USI) A-46, & Suppl 1 BSEP-99-0127, Forwards Rev 3 to Physical Security & Safeguards Contingency Plan, for Brunswick Steam Electric Plant,Units 1 & 2.Without Encl1999-08-0505 August 1999 Forwards Rev 3 to Physical Security & Safeguards Contingency Plan, for Brunswick Steam Electric Plant,Units 1 & 2.Without Encl BSEP-99-0124, Forwards Proprietary Medical Status Rept & NRC Form 396 for Individual Holding License SOP-20811-1. Proprietary Encls Withheld1999-07-30030 July 1999 Forwards Proprietary Medical Status Rept & NRC Form 396 for Individual Holding License SOP-20811-1. Proprietary Encls Withheld ML20210G2941999-07-28028 July 1999 Discusses Public Meeting Conducted on 990720 to Present Results of Periodic Plant Performance Review for Brunswick Facility for Period of May 1997 to January 1999.List of Attendees Encl BSEP-99-0122, Informs That,Effective 990702,NRC Operator License for Individual Licensed on BSEP Units 1 & 2,expired Because Individual Employment at BSEP Was Terminated.Proprietary Encl Containing Name,Docket & License Number Withheld1999-07-20020 July 1999 Informs That,Effective 990702,NRC Operator License for Individual Licensed on BSEP Units 1 & 2,expired Because Individual Employment at BSEP Was Terminated.Proprietary Encl Containing Name,Docket & License Number Withheld BSEP-99-0121, Submits Response to Administrative Ltr 99-02, Operating Reactor Licensing Action Estimates. Estimates That 14 Submittals Will Be Made During Fy 2000 & Nine Will Be Made During Fy 2001 for Plant1999-07-19019 July 1999 Submits Response to Administrative Ltr 99-02, Operating Reactor Licensing Action Estimates. Estimates That 14 Submittals Will Be Made During Fy 2000 & Nine Will Be Made During Fy 2001 for Plant BSEP-99-0111, Forwards New Affidavit Supporting Withholding of Proprietay Info Provided by GE Entitled, Addl Info Re 1.09 Cycle Specific SLMCPR for Brunswick Unit 1 Cycle 12 & Included in Util 980223 LAR1999-07-19019 July 1999 Forwards New Affidavit Supporting Withholding of Proprietay Info Provided by GE Entitled, Addl Info Re 1.09 Cycle Specific SLMCPR for Brunswick Unit 1 Cycle 12 & Included in Util 980223 LAR ML20210E1181999-07-19019 July 1999 Forwards Insp Repts 50-324/99-04 & 50-325/99-04 on 990509- 0619.Three Violations of NRC Requirements Occurred & Being Treated as non-cited Violations Consistent with App C of Enforcement Policy ML20209G1081999-07-13013 July 1999 Forwards Radiological Emergency Response Plan & Revised Epips,Including Rev 4 to OPEP-02.1.1 & Rev 7 to OPEP-02.6.26 BSEP-99-0106, Informs That Units 1 & 2 Odcm,Spec 7.3.10, Gaseous Radwaste Treatment Sys Requires Gaseous Radwaste Treatment Sys to Be in Operation Whenever Main Condenser Air Ejector Sys Is in Operation.Special Rept Encl1999-06-28028 June 1999 Informs That Units 1 & 2 Odcm,Spec 7.3.10, Gaseous Radwaste Treatment Sys Requires Gaseous Radwaste Treatment Sys to Be in Operation Whenever Main Condenser Air Ejector Sys Is in Operation.Special Rept Encl BSEP-99-0103, Submits follow-up Info Re Interlaboratory Comparison Program for 1998 Radiological Environ Operating Rept,Suppl 11999-06-24024 June 1999 Submits follow-up Info Re Interlaboratory Comparison Program for 1998 Radiological Environ Operating Rept,Suppl 1 BSEP-99-0100, Forwards Rev 0 to Calculation 2B11-0001, Core Shroud B214R1 Structural Evaluation, Which Provides Evaluation of Unit 2 Core Shroud Insp Results1999-06-23023 June 1999 Forwards Rev 0 to Calculation 2B11-0001, Core Shroud B214R1 Structural Evaluation, Which Provides Evaluation of Unit 2 Core Shroud Insp Results BSEP-99-0105, Forwards Rev 6 to OPEP-02.6, Severe Weather. Summary of Changes Encl1999-06-23023 June 1999 Forwards Rev 6 to OPEP-02.6, Severe Weather. Summary of Changes Encl ML20196H8191999-06-21021 June 1999 Informs That on 980915,Commission Suspended SALP Program for an Interim Period Until NRC Completes Review of Process for Assessing Performance at NPPs ML20212J0541999-06-17017 June 1999 Responds to Requesting That NRC Staff ...Allow BWR Plants Identified to Defer Weld Overlay Exams Until March 2001 or Until Completion of NRC Staff Review & Approval of Proposed Generic Rept,Whichever Comes First BSEP-99-0091, Notifies That Operator Licensed on Units 1 & 2,has Been Reassigned to non-licensed Activities & Operator License Considered Expired.Proprietary Info Encl.Proprietary Info Withheld,Per 10CFR2.790(a)(6)1999-06-14014 June 1999 Notifies That Operator Licensed on Units 1 & 2,has Been Reassigned to non-licensed Activities & Operator License Considered Expired.Proprietary Info Encl.Proprietary Info Withheld,Per 10CFR2.790(a)(6) ML20196A8111999-06-14014 June 1999 Ack Receipt of 990401 Response to NRC Ltr Issued on 990302 Re OI Investigation to Determine Whether Former BSEP Manager Threatened to Fire Employees Who Brought Safety Concerns to Nrc.Concluded That No Violation Occurred ML20195J1971999-06-14014 June 1999 Advises That GE-NE-523-B13-01920-56,Rev 1,submitted with 990202 Affidavit,Marked Proprietary,Will Be Withheld from Public Disclosure,Per 10CFR2.790(b)(5) & Section 103(b) of AEA of 1954,as Amended BSEP-99-0090, Provides Notification That Data Point D23C0315 Has Been Restored to Original Configuration During RFO on 9905141999-06-0909 June 1999 Provides Notification That Data Point D23C0315 Has Been Restored to Original Configuration During RFO on 990514 1999-09-08
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARBSEP-99-0170, Forwards Proprietery Notification of Change in Operator Status.Individual Name,Docket Number,License Number & Effective Date of Expiration,Encl.Proprietary Info Withheld, Per 10CFR2.790(a)(6)1999-10-19019 October 1999 Forwards Proprietery Notification of Change in Operator Status.Individual Name,Docket Number,License Number & Effective Date of Expiration,Encl.Proprietary Info Withheld, Per 10CFR2.790(a)(6) BSEP-99-0161, Submits Response to NRC RAI Re Relief Request VRR-03.VRR-03 Requested Relief from full-stroke Open Exercise Requirements for Supply Check Valves to air-operated Valves 1(2)-RNA-V313,1(2)-RNA-V314,1(2)-RNA-V350 & 1(2)-RNA-V3511999-10-13013 October 1999 Submits Response to NRC RAI Re Relief Request VRR-03.VRR-03 Requested Relief from full-stroke Open Exercise Requirements for Supply Check Valves to air-operated Valves 1(2)-RNA-V313,1(2)-RNA-V314,1(2)-RNA-V350 & 1(2)-RNA-V351 BSEP-99-0114, Forwards Supplemental Biological Assessment Submitted by CP&L Providing Updated Data for 1998 & 1999. Pictures of Intake Canal at Diversion Structure During High Tide Conditions Also Encl1999-10-0404 October 1999 Forwards Supplemental Biological Assessment Submitted by CP&L Providing Updated Data for 1998 & 1999. Pictures of Intake Canal at Diversion Structure During High Tide Conditions Also Encl BSEP-99-0157, Submits Annual Rept Summarizing Effect of Changes & Errors in Accepted loss-of-coolant Accident ECCS Evaluation Models Applicable to Bsep,Units 1 & 2 IAW 10CFR50.46(a)(3)(ii)1999-10-0404 October 1999 Submits Annual Rept Summarizing Effect of Changes & Errors in Accepted loss-of-coolant Accident ECCS Evaluation Models Applicable to Bsep,Units 1 & 2 IAW 10CFR50.46(a)(3)(ii) BSEP-99-0142, Forwards Proprietary Updated List of Home Addresses for Individuals Licensed to Operate CP&L Bsep,Units 1 & 2. Proprietary Info Withheld,Per 10CFR2.790(a)(6)1999-09-23023 September 1999 Forwards Proprietary Updated List of Home Addresses for Individuals Licensed to Operate CP&L Bsep,Units 1 & 2. Proprietary Info Withheld,Per 10CFR2.790(a)(6) BSEP-99-0158, Provides Proprietary Notification of Change in Operator Status.Individual Name,Docket Number,License & Effective Date of Expiration Provided in Encl to Ltr.Proprietary Info Withheld,Per 10CFR2.790(a)(6)1999-09-21021 September 1999 Provides Proprietary Notification of Change in Operator Status.Individual Name,Docket Number,License & Effective Date of Expiration Provided in Encl to Ltr.Proprietary Info Withheld,Per 10CFR2.790(a)(6) BSEP-99-0153, Forwards Monthly Operating Repts for Brunswick Steam Electric Plant,Units 1 & 2.CP&L Is Submitting Encl Rept of Operating Statistics & SD Experience for Aug 19991999-09-14014 September 1999 Forwards Monthly Operating Repts for Brunswick Steam Electric Plant,Units 1 & 2.CP&L Is Submitting Encl Rept of Operating Statistics & SD Experience for Aug 1999 BSEP-99-0147, Forwards Response to NRC Telcon RAI Re Relief Requests RR-5, 12,13,14,22,23,24 & 25,per Inservice Insp Program for Third 10-yr Interval1999-09-14014 September 1999 Forwards Response to NRC Telcon RAI Re Relief Requests RR-5, 12,13,14,22,23,24 & 25,per Inservice Insp Program for Third 10-yr Interval BSEP-99-0151, Notifies NRC That Unit 1 Digtial FW Control Sys Upgrade Was Completed on 990831 During Plant Sd.Action Completes Y2K Remediation Activities for BSEP1999-09-0808 September 1999 Notifies NRC That Unit 1 Digtial FW Control Sys Upgrade Was Completed on 990831 During Plant Sd.Action Completes Y2K Remediation Activities for BSEP BSEP-99-0150, Forwards Proprietary Info Re Positive Drug Test for Operator Licensed on CP&L Bsep,Units 1 & 2,in Response to NRC .Proprietary Info Withheld,Per 10CFR2.790(a)(6)1999-09-0202 September 1999 Forwards Proprietary Info Re Positive Drug Test for Operator Licensed on CP&L Bsep,Units 1 & 2,in Response to NRC .Proprietary Info Withheld,Per 10CFR2.790(a)(6) BSEP-99-0149, Forwards Two Proprietary License Renewal Applications Consisting of NRC Form 398 & NRC Form 396,for Operators Licensed at Plant.Proprietary Encls Withheld1999-09-0202 September 1999 Forwards Two Proprietary License Renewal Applications Consisting of NRC Form 398 & NRC Form 396,for Operators Licensed at Plant.Proprietary Encls Withheld BSEP-99-0112, Forwards Summary of Exam Results for Feedwater Sparger & Nozzle Exams Performed During RFO13.Evaluation of Exam Results Documented in ESR 98-00333, Unit 2 Feedwater Sparger Evaluation Based on B214R1 IVVI Exam Results, Encl1999-09-0101 September 1999 Forwards Summary of Exam Results for Feedwater Sparger & Nozzle Exams Performed During RFO13.Evaluation of Exam Results Documented in ESR 98-00333, Unit 2 Feedwater Sparger Evaluation Based on B214R1 IVVI Exam Results, Encl BSEP-99-0146, Forwards Revised EPIPs for Bsep,Units 1 & 2.List of Revised Procedures & Summary of Changes Are Encl1999-08-26026 August 1999 Forwards Revised EPIPs for Bsep,Units 1 & 2.List of Revised Procedures & Summary of Changes Are Encl BSEP-99-0141, Forwards Proprietary Info Re Expiration of Operator Licenses SOP-20812-1,OP-20977-1 & OP-21217,per 10CFR50.74(b) Re Notification of Changes in Operator Status.Proprietary Info Withheld,Per 10CFR2.790(a)(6)1999-08-25025 August 1999 Forwards Proprietary Info Re Expiration of Operator Licenses SOP-20812-1,OP-20977-1 & OP-21217,per 10CFR50.74(b) Re Notification of Changes in Operator Status.Proprietary Info Withheld,Per 10CFR2.790(a)(6) BSEP-99-0148, Forwards Rev 0 to NF-908.03, Brunswick Unit 2,Cycle 14 Neutronics Startup Rept, IAW Section 13.4.2.1 of UFSAR for Brunswick Steam Electric Plant,Units 1 & 21999-08-25025 August 1999 Forwards Rev 0 to NF-908.03, Brunswick Unit 2,Cycle 14 Neutronics Startup Rept, IAW Section 13.4.2.1 of UFSAR for Brunswick Steam Electric Plant,Units 1 & 2 BSEP-99-0130, Requests Relief from ASME Boiler & Pressure Vessel Code, Section Xi,Iaw 10CFR50.55a(g)(5)(iii) & NRC GL 90-05, Guidance for Performing Temporary Non-Code Repair of ASME Code Class 1,2 & 3 Piping. Regulatory Commitments,Encl1999-08-25025 August 1999 Requests Relief from ASME Boiler & Pressure Vessel Code, Section Xi,Iaw 10CFR50.55a(g)(5)(iii) & NRC GL 90-05, Guidance for Performing Temporary Non-Code Repair of ASME Code Class 1,2 & 3 Piping. Regulatory Commitments,Encl BSEP-99-0132, Requests Delay in Providing Update on Util Intended Actions Re GL 96-06,until 180 Days After NRC Approval of Generic Technical Basis.List of Regulatory Commitments,Encl1999-08-20020 August 1999 Requests Delay in Providing Update on Util Intended Actions Re GL 96-06,until 180 Days After NRC Approval of Generic Technical Basis.List of Regulatory Commitments,Encl BSEP-99-0134, Forwards Monthly Operating Repts for July 1999 for BSEP, Units 1 & 2.Revised Rept for June 1999 for Unit 2,reflecting Info on 990628 Forced Shutdown of Plant,Encl1999-08-13013 August 1999 Forwards Monthly Operating Repts for July 1999 for BSEP, Units 1 & 2.Revised Rept for June 1999 for Unit 2,reflecting Info on 990628 Forced Shutdown of Plant,Encl BSEP-99-0128, Forwards Revised Relief Request RR-17,applicable to Remainder of Third 10-year ISI Program for Plant1999-08-11011 August 1999 Forwards Revised Relief Request RR-17,applicable to Remainder of Third 10-year ISI Program for Plant BSEP-99-0127, Forwards Rev 3 to Physical Security & Safeguards Contingency Plan, for Brunswick Steam Electric Plant,Units 1 & 2.Without Encl1999-08-0505 August 1999 Forwards Rev 3 to Physical Security & Safeguards Contingency Plan, for Brunswick Steam Electric Plant,Units 1 & 2.Without Encl BSEP-99-0123, Forwards Rev 54 to Radiological Emergency Response Plan (RERP) & Revised Epips,Including Re 44 to OPEP-02.1,rev 4 to OPEP-02.1.1 & Rev 7 to OPEP-02.6.26.Summary of Changes Provided in Encl 21999-08-0505 August 1999 Forwards Rev 54 to Radiological Emergency Response Plan (RERP) & Revised Epips,Including Re 44 to OPEP-02.1,rev 4 to OPEP-02.1.1 & Rev 7 to OPEP-02.6.26.Summary of Changes Provided in Encl 2 BSEP-99-0110, Forwards Rev 0 to ESR 99-00279, B214R1 RPV Hydrotest Bolted Connection Corrective Action Evaluation, IAW Subparagraph IWB-3144(b) of 1989 Edition of ASME B&PV Code, Section XI1999-08-0505 August 1999 Forwards Rev 0 to ESR 99-00279, B214R1 RPV Hydrotest Bolted Connection Corrective Action Evaluation, IAW Subparagraph IWB-3144(b) of 1989 Edition of ASME B&PV Code, Section XI BSEP-99-0124, Forwards Proprietary Medical Status Rept & NRC Form 396 for Individual Holding License SOP-20811-1. Proprietary Encls Withheld1999-07-30030 July 1999 Forwards Proprietary Medical Status Rept & NRC Form 396 for Individual Holding License SOP-20811-1. Proprietary Encls Withheld BSEP-99-0122, Informs That,Effective 990702,NRC Operator License for Individual Licensed on BSEP Units 1 & 2,expired Because Individual Employment at BSEP Was Terminated.Proprietary Encl Containing Name,Docket & License Number Withheld1999-07-20020 July 1999 Informs That,Effective 990702,NRC Operator License for Individual Licensed on BSEP Units 1 & 2,expired Because Individual Employment at BSEP Was Terminated.Proprietary Encl Containing Name,Docket & License Number Withheld BSEP-99-0111, Forwards New Affidavit Supporting Withholding of Proprietay Info Provided by GE Entitled, Addl Info Re 1.09 Cycle Specific SLMCPR for Brunswick Unit 1 Cycle 12 & Included in Util 980223 LAR1999-07-19019 July 1999 Forwards New Affidavit Supporting Withholding of Proprietay Info Provided by GE Entitled, Addl Info Re 1.09 Cycle Specific SLMCPR for Brunswick Unit 1 Cycle 12 & Included in Util 980223 LAR BSEP-99-0121, Submits Response to Administrative Ltr 99-02, Operating Reactor Licensing Action Estimates. Estimates That 14 Submittals Will Be Made During Fy 2000 & Nine Will Be Made During Fy 2001 for Plant1999-07-19019 July 1999 Submits Response to Administrative Ltr 99-02, Operating Reactor Licensing Action Estimates. Estimates That 14 Submittals Will Be Made During Fy 2000 & Nine Will Be Made During Fy 2001 for Plant ML20209G1081999-07-13013 July 1999 Forwards Radiological Emergency Response Plan & Revised Epips,Including Rev 4 to OPEP-02.1.1 & Rev 7 to OPEP-02.6.26 BSEP-99-0106, Informs That Units 1 & 2 Odcm,Spec 7.3.10, Gaseous Radwaste Treatment Sys Requires Gaseous Radwaste Treatment Sys to Be in Operation Whenever Main Condenser Air Ejector Sys Is in Operation.Special Rept Encl1999-06-28028 June 1999 Informs That Units 1 & 2 Odcm,Spec 7.3.10, Gaseous Radwaste Treatment Sys Requires Gaseous Radwaste Treatment Sys to Be in Operation Whenever Main Condenser Air Ejector Sys Is in Operation.Special Rept Encl BSEP-99-0103, Submits follow-up Info Re Interlaboratory Comparison Program for 1998 Radiological Environ Operating Rept,Suppl 11999-06-24024 June 1999 Submits follow-up Info Re Interlaboratory Comparison Program for 1998 Radiological Environ Operating Rept,Suppl 1 BSEP-99-0100, Forwards Rev 0 to Calculation 2B11-0001, Core Shroud B214R1 Structural Evaluation, Which Provides Evaluation of Unit 2 Core Shroud Insp Results1999-06-23023 June 1999 Forwards Rev 0 to Calculation 2B11-0001, Core Shroud B214R1 Structural Evaluation, Which Provides Evaluation of Unit 2 Core Shroud Insp Results BSEP-99-0105, Forwards Rev 6 to OPEP-02.6, Severe Weather. Summary of Changes Encl1999-06-23023 June 1999 Forwards Rev 6 to OPEP-02.6, Severe Weather. Summary of Changes Encl BSEP-99-0091, Notifies That Operator Licensed on Units 1 & 2,has Been Reassigned to non-licensed Activities & Operator License Considered Expired.Proprietary Info Encl.Proprietary Info Withheld,Per 10CFR2.790(a)(6)1999-06-14014 June 1999 Notifies That Operator Licensed on Units 1 & 2,has Been Reassigned to non-licensed Activities & Operator License Considered Expired.Proprietary Info Encl.Proprietary Info Withheld,Per 10CFR2.790(a)(6) BSEP-99-0090, Provides Notification That Data Point D23C0315 Has Been Restored to Original Configuration During RFO on 9905141999-06-0909 June 1999 Provides Notification That Data Point D23C0315 Has Been Restored to Original Configuration During RFO on 990514 BSEP-99-0092, Informs That,Effective 990513,NRC Operator License for Individual Licensed on BSEP Units 1 & 2,expired.Proprietary Encl Containing Name of Individual,Docket Number,License Mumber & Date of Expiration Encl.Proprietary Encl Withheld1999-06-0808 June 1999 Informs That,Effective 990513,NRC Operator License for Individual Licensed on BSEP Units 1 & 2,expired.Proprietary Encl Containing Name of Individual,Docket Number,License Mumber & Date of Expiration Encl.Proprietary Encl Withheld BSEP-99-0085, Forwards Proprietary Rev 67 to EPIP OPEP-App a, Emergency Response Resources & non-proprietary Rev 53 to Radiological Emergency Response Plan, Rev 15 to EPIP OPEP- 03.8.2 & Rev 17 to EPIP OPEP-04.2.Proprietary Info Withheld1999-05-27027 May 1999 Forwards Proprietary Rev 67 to EPIP OPEP-App a, Emergency Response Resources & non-proprietary Rev 53 to Radiological Emergency Response Plan, Rev 15 to EPIP OPEP- 03.8.2 & Rev 17 to EPIP OPEP-04.2.Proprietary Info Withheld BSEP-99-0078, Provides Addl Info Re Jet Pump Riser Weld Insp Results,Per 990406 Telcon with NRC Re Effect of worst-case Crack on LPCI Sys Flow & PCT1999-05-17017 May 1999 Provides Addl Info Re Jet Pump Riser Weld Insp Results,Per 990406 Telcon with NRC Re Effect of worst-case Crack on LPCI Sys Flow & PCT BSEP-99-0080, Forwards Revs Made to TS Bases Between 980605 & 990507 Without Prior NRC Approval1999-05-17017 May 1999 Forwards Revs Made to TS Bases Between 980605 & 990507 Without Prior NRC Approval BSEP-99-0083, Forwards Two Proprietary License Renewal Applications Consisting of NRC Forms 398 & 396,for Operators Licensed on CP&L Bsep,Units 1 & 2,IAW 10CFR55.57.Proprietary Info Withheld from Public Disclosure IAW 10CFR2.790(a)(6)1999-05-14014 May 1999 Forwards Two Proprietary License Renewal Applications Consisting of NRC Forms 398 & 396,for Operators Licensed on CP&L Bsep,Units 1 & 2,IAW 10CFR55.57.Proprietary Info Withheld from Public Disclosure IAW 10CFR2.790(a)(6) BSEP-99-0056, Forwards 1998 Radiological Environ Operating Rept for BSEP, Units 1 & 2. Encl 1 Provides List of Regulatory Commitments.Vols II & Iii,Sample Analysis Data for Jan-June 1998 & Jul-Dec 1998,respectively1999-05-13013 May 1999 Forwards 1998 Radiological Environ Operating Rept for BSEP, Units 1 & 2. Encl 1 Provides List of Regulatory Commitments.Vols II & Iii,Sample Analysis Data for Jan-June 1998 & Jul-Dec 1998,respectively BSEP-99-0073, Withdraws Request for Approval of Proposed QAP Change Which Was Submitted by1999-05-12012 May 1999 Withdraws Request for Approval of Proposed QAP Change Which Was Submitted by BSEP-99-0062, Forwards Latest Revs of COLR & Suppl Reload Licensing Rept, for Bsep,Unit 2.Proprietary Rev 6,Suppl 2 to NEDC- 31624P & EMF-2168P,included.Proprietary Encls Withheld, Per 10CFR2.790 & 10CFR9.171999-05-11011 May 1999 Forwards Latest Revs of COLR & Suppl Reload Licensing Rept, for Bsep,Unit 2.Proprietary Rev 6,Suppl 2 to NEDC- 31624P & EMF-2168P,included.Proprietary Encls Withheld, Per 10CFR2.790 & 10CFR9.17 BSEP-99-0079, Forwards Proprietary One License Renewal Application Consisting of NRC Form 398, Personal Qualification Statement - Licensee & NRC Form 396, Certification Of... Proprietary Info Withheld,Per 10CFR2.790(a)(6)1999-05-0707 May 1999 Forwards Proprietary One License Renewal Application Consisting of NRC Form 398, Personal Qualification Statement - Licensee & NRC Form 396, Certification Of... Proprietary Info Withheld,Per 10CFR2.790(a)(6) BSEP-99-0066, Submits Revised Relief Requests Which Provides Alternative to ASME B&PV Code Requirements for Insp of Containments. Revised Copies of Relief Requests CIP-03,CIP-04,CIP-05, CIP-10,CIP-16 & CIP-17 Provided1999-05-0404 May 1999 Submits Revised Relief Requests Which Provides Alternative to ASME B&PV Code Requirements for Insp of Containments. Revised Copies of Relief Requests CIP-03,CIP-04,CIP-05, CIP-10,CIP-16 & CIP-17 Provided ML20195D6301999-04-30030 April 1999 Forwards Natl Marine Fisheries Svc Biological Opinion Based on Review of Continued Use of Cooling Water Intake Sys at Bsep.Rept Reviews Effects of Activity on Loggerhead,Kemps Ridley,Green,Hawksbill & Leatherback Sea Turtles BSEP-99-0071, Forwards Notice of Expiration of Operator License,Per 10CFR50.75(b).Details Withheld Per 10CFR2.790(a)(6)1999-04-27027 April 1999 Forwards Notice of Expiration of Operator License,Per 10CFR50.75(b).Details Withheld Per 10CFR2.790(a)(6) BSEP-99-0069, Discusses Rev of Relief Request RR-17, Leakage at Bolted Connections, for ISI Program for Third ten-yr Interval.Util & NRC Were Unable to Reach Agreement Re Use of ASME Code Case N-566-1 as Permanent Basis for Relief Request RR-171999-04-26026 April 1999 Discusses Rev of Relief Request RR-17, Leakage at Bolted Connections, for ISI Program for Third ten-yr Interval.Util & NRC Were Unable to Reach Agreement Re Use of ASME Code Case N-566-1 as Permanent Basis for Relief Request RR-17 BSEP-99-0063, Notifies NRC of Plans to Install Mod for Bsep,Units 1 & 2 Erds.Physical Computer Sys Changeout Activities Will Temporarily Affect Availability of ERDS Computer Points Via NRC Data Link1999-04-21021 April 1999 Notifies NRC of Plans to Install Mod for Bsep,Units 1 & 2 Erds.Physical Computer Sys Changeout Activities Will Temporarily Affect Availability of ERDS Computer Points Via NRC Data Link BSEP-99-0050, Informs NRC That Certain Category C,D & E Piping Welds in Units 1 & 2 Have Not Been Inspected in Accordance with Frequencies Provided by NRC GL 88-01, NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping,1999-04-0606 April 1999 Informs NRC That Certain Category C,D & E Piping Welds in Units 1 & 2 Have Not Been Inspected in Accordance with Frequencies Provided by NRC GL 88-01, NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping, BSEP-99-0051, Forwards Response to Ofc of Investigations Rept 2-1998-014 to Support Final Resolution of Allegations.Response in Encl 1 Withheld.Public Version of Response in Encl 21999-04-0101 April 1999 Forwards Response to Ofc of Investigations Rept 2-1998-014 to Support Final Resolution of Allegations.Response in Encl 1 Withheld.Public Version of Response in Encl 2 BSEP-99-0015, Forwards Response to NRC 981221 RAI Re Cp&Ls Response to GL 95-07, Pressure Locking & Thermal Binding of SR Power-Operated Gate Valves1999-03-31031 March 1999 Forwards Response to NRC 981221 RAI Re Cp&Ls Response to GL 95-07, Pressure Locking & Thermal Binding of SR Power-Operated Gate Valves 1999-09-08
[Table view] Category:UTILITY TO NRC
MONTHYEARML20064A8161990-09-14014 September 1990 Forwards Info Re Performance of NDE Exam of Feedwater Nozzles & Safe Ends During 1989/1990 Maint/Refueling Outage, Per NUREG-0619,Section 4.4.3.1(2) ML20059L0911990-09-12012 September 1990 Confirms That Fee Electronically Transferred to Dept of Treasury for Payment of NRC Review Fees ML20059G2881990-09-0606 September 1990 Forwards Proprietary NEDC-31624P,Rev 2,Class III, Brunswick Steam Electric Plant Units 1 & 2 SAFER/GESTR-LOCA LOCA Loss of Coolant Accident Analysis, for NRC Approval.Rept Withheld ML17348B4941990-08-30030 August 1990 Forwards Semiannual 10CFR26 fitness-for-duty Program Data for 900103-0630.Mgt Decision Made to Utilize Alcohol Breath Instruments as Screening Devices for Unscheduled Work Call Outs in Determining fitness-for-duty BSEP-90-0579, Informs That Util Plans to Replace Pilot Valve Assemblies on Listed Safety Relief Valves Prior to Startup1990-08-24024 August 1990 Informs That Util Plans to Replace Pilot Valve Assemblies on Listed Safety Relief Valves Prior to Startup BSEP-90-0564, Forwards Response to Violations Noted in Insp Repts 50-324/90-19 & 50-325/90-19.Corrective Actions:Procedures to Shutdown Reactor Feed Pump Using Manual Control Approved on 900627 & Real Time Training Provided to Operators1990-08-16016 August 1990 Forwards Response to Violations Noted in Insp Repts 50-324/90-19 & 50-325/90-19.Corrective Actions:Procedures to Shutdown Reactor Feed Pump Using Manual Control Approved on 900627 & Real Time Training Provided to Operators BSEP-90-0565, Responds to Violations Noted in Insp Repts 50-324/90-06 & 50-325/90-06.Corrective Actions:Personnel Counseled on Closure of Locked High Radiation Area Doors & Preventive Maint Program Established for Reactor & Turbine Bldg Gates1990-08-16016 August 1990 Responds to Violations Noted in Insp Repts 50-324/90-06 & 50-325/90-06.Corrective Actions:Personnel Counseled on Closure of Locked High Radiation Area Doors & Preventive Maint Program Established for Reactor & Turbine Bldg Gates ML20058N1671990-08-0909 August 1990 Forwards Updated Tech Spec Pages Re 900314 Application for Amends to Licenses DPR-71 & DPR-62.Amends Permit Removal of Rod Sequence Control Sys & Reduces Rod Worth Minimizer Cutoff Setpoint to 10%-rated Thermal Power BSEP-90-0535, Forwards Monthly Status Rept on Plant Integrated Action Plan.New Target Completion Date for D27 Will Be 9008151990-08-0707 August 1990 Forwards Monthly Status Rept on Plant Integrated Action Plan.New Target Completion Date for D27 Will Be 900815 ML20056A6931990-08-0707 August 1990 Advises of Plans to Complete Remaining Portions of Licensed Operator Requalification Exams During Oct-Dec 1990 BSEP-90-0545, Forwards Monthly Operating Rept for Jul 1990 for Brunswick Steam Electric Plant Units 1 & 2 & Corrected Operating Data Rept for June 1990,Unit 21990-08-0303 August 1990 Forwards Monthly Operating Rept for Jul 1990 for Brunswick Steam Electric Plant Units 1 & 2 & Corrected Operating Data Rept for June 1990,Unit 2 ML20063Q0501990-08-0303 August 1990 Forwards Corrections to Plant Updated Fsar,Amend 8,per ML20056A0291990-07-30030 July 1990 Forwards Response to NRC 900206 Request for Addl Info Re Util 870817 License Amend Request,Extending Expiration Dates for OLs ML20056A0611990-07-27027 July 1990 Advises That Util Plans to Perform Next Integrated Leak Rate Test During Next Reload 8 Outage,Currently Scheduled for Sept 1991 ML20056A1801990-07-25025 July 1990 Forwards Info Re NPDES Permit Noncompliance Reported to Permit Agency for Month of Mar 1990,per 830426 Agreement BSEP-90-0524, Informs of Change to Info in NRC Insp Repts 50-324/90-01 & 50-325/90-01 Paragraph 3.b of Details1990-07-24024 July 1990 Informs of Change to Info in NRC Insp Repts 50-324/90-01 & 50-325/90-01 Paragraph 3.b of Details ML20055G6771990-07-20020 July 1990 Responds to NRC Re Violations Noted in Insp Repts 50-324/90-14 & 50-325/90-14.Corrective Actions:Event Included in Third Quarter Real Time Training for Personnel, Emphasizing Importance of Detection of Plant Deficiencies ML20055F9431990-07-12012 July 1990 Advises That Stated Amount Electronically Transferred to Us Dept of Treasury,Nrc on 900711 for Payment of Operator License Exam Fees for Listed Insp Invoices BSEP-90-0459, Suppls Response to Violations Noted in Insp Repts 50-324/90-06 & 50-325/90-06 Re Locked High Radiation Area Boundary.Three Options to Eliminate Design Problems W/Locked High Radiation Boundaries Will Be Submitted by 9008171990-06-29029 June 1990 Suppls Response to Violations Noted in Insp Repts 50-324/90-06 & 50-325/90-06 Re Locked High Radiation Area Boundary.Three Options to Eliminate Design Problems W/Locked High Radiation Boundaries Will Be Submitted by 900817 ML20058K3551990-06-28028 June 1990 Responds to NRC Confirmation of Actions Ltr Re Licensed Operator Training Program Review.Curriculum Development Unit of Nuclear Training Section Will Develop Self Assessment Criteria to Increase Effectiveness of Training Program ML20055D2751990-06-28028 June 1990 Informs That Senior Operator License SOP-20687 for Rj Zuffa No Longer Needed & Should Be Terminated Upon Receipt of Ltr ML20055D7881990-06-27027 June 1990 Forwards Updated Tech Spec Pages to 890329 Application for Amends to Licenses DPR-71 & DPR-62 to Permit Expanded Operating Domain for Facility ML20055C8201990-06-18018 June 1990 Forwards Rev 0 to Rept of 1989/90 Inservice Insp for Brunswick Steam Electric Plant Unit 2, Per ASME Boiler & Pressure Vessel Code,Section Xi,Article IWA 6220(b) ML20055H5501990-06-14014 June 1990 Discusses Confirmation of Concurrence Re Qualification of Fifth Operating Shift.Util Determined That Crew C Will Be Prepared for NRC Testing by 900723 & 24 to Provide Fifth Crew ML20043G4711990-06-13013 June 1990 Forwards Tech Specs to Support 900228 License Amend Request Re Fuel Cycle 8 Reload ML20043G4671990-06-13013 June 1990 Forwards Proposed Tech Specs Re Turbine Control Valves & Fire Detectors,To Support 900315 License Amend Request ML20043G4701990-06-13013 June 1990 Withdraws 900419 Request for License Amend Re one-time Extension of Diesel Generator Surveillance Interval.Util Investigating Methods of Dealing W/Situation & Will Keep NRR Project Manager Informed of Plans ML20043F1921990-06-0606 June 1990 Advises of Intentions to Delay Submittal of Permanent Change to Tech Spec 3/4.8.1,pending Resolution of Lessons Learned from Incident at Plant ML20043F0431990-06-0606 June 1990 Advises That Correct Date for Remedial Exercise Is 900622, Per Insp Repts 50-324/90-13 & 50-325/90-13.Procedures Will Be Revised by 900706 to Ensure Adequate Accountability for All Personnel within Protected Area ML20043C9021990-06-0101 June 1990 Responds to Violations Noted in Insp Repts 50-324/90-14 & 50-325/90-14.Corrective Actions:Box Cover Closed Following Notification of Discrepancy & Human Performance Evaluation Initiated to Ensure Proper Corrective Actions Initiated ML20043D1851990-05-30030 May 1990 Forwards Response to NRC 900206 Request for Addl Info Re Util 870817 Proposed Change to Expiration Dates of Ols. Responses to Remaining Questions Will Be Submitted by 900629 ML20043B7821990-05-25025 May 1990 Responds to Violations Noted in Insp Repts 50-324/88-36 & 50-325/88-36.Allowable Working Loads for 1/2-inch Hilti Anchor Bolts Currently Used at Facilities Adequate Per NRC Info Notice 86-094.No Further Action Planned ML20043A3941990-05-16016 May 1990 Forwards Tech Spec Re 900419 Amend Request for One Time Extension of Diesel Generator Surveillance Requirements ML20043G4991990-05-14014 May 1990 Forwards Corrected Graph for 900531 Integrated Action Plan Status Rept Illustrating Level 1 Items Projected as of 900331 ML20042G5411990-05-0909 May 1990 Forwards Response to Request for Addl Info Re 891010 License Amend Request to Raise Allowable Containment Leak Rate from 0.5 to 1.0 Vol Percent Per Day.Core Thermal Power Level to Be Used in Calculation Is 2,550 Mwt ML20042F6111990-05-0202 May 1990 Responds to Generic Ltr 89-19, Request for Action Re Resolution of USI A-47, 'Safety Implication of Control Sys in LWR Nuclear Power Plants.' Plant Procedures Provide Adequate Guidance for Operation of Overfill Protection Sys ML20042E8531990-04-27027 April 1990 Discusses Relocation of Primary/Secondary Containment Isolation Valve Listings to Plant Procedures.Relocation Will Help Avoid Future Expenditure of NRC Resources for Review & Processing of License Amend Requests ML20042E7281990-04-27027 April 1990 Provides Tech Spec Interpretation Evaluation of Offsite/ Onsite Electrical Distribution Sys,Per NRC Request During 900123 Telcon.Submittal of Amend Request Will Be Made by 900531 ML20012F4871990-04-0606 April 1990 Forwards Response to Regulatory Effectiveness Review.Encl Withheld (Ref 10CFR73.21) ML20012F4701990-04-0404 April 1990 Forwards Addl Info to Support Util Finding of NSHC for 900228 Application for Amend to License DPR-71,revising MCPR Safety Limit Specified in Tech Spec 2.1.2 from 1.04 to 1.07 for Unit 1,Cycle 8 Operation ML20042E1421990-04-0303 April 1990 Forwards Monthly Status Rept on Plant Integrated Action Plan ML20012E8681990-03-30030 March 1990 Submits Supplemental Response to Station Blackout Rule Per NUMARC 900104 Request.One Change to Coping Assessment Calculation Was Necessary & Deviations from NUMARC 87-00 Methodology Identified ML20012E4621990-03-27027 March 1990 Discusses Licensee Response to NRC 891010 Request for Recirculation Pipe Welds.Ultrasonic Testing Records Package on Weld Samples Assembled & Given to BNL During 891206 Facility Visit.Requested Welds Shipped to BNL on 900301 ML20012D9761990-03-22022 March 1990 Responds to Insp Repts 50-324/88-36 & 50-325/88-36 Re Util Response to IE Bulletin 79-02.Verification That Pipe Support Base Plate Flexibility Accounted for in Calculation of Anchor Bolt Loads Addressed ML20012C4601990-03-15015 March 1990 Forwards Control Loop Voltage Drop Calculation Inadvertently Omitted from 900313 Ltr Providing Addl Info Re Ari/Rpt ML20012E1701990-03-13013 March 1990 Forwards Addl Info Re Ari/Rpt Sys Power Supply,Per NRC 900110 Request ML20012A4271990-02-28028 February 1990 Forwards Rev 0 to Brunswick Unit 2,Cycle 9,Core Operating Limits Rept. ML20012A2491990-02-28028 February 1990 Forwards Supplemental Response to Violations Noted in Insp Repts 50-324/89-26 & 50-325/89-26.Corrective Actions:Ncr S-89-089 Initiated on F075 Event & Involved Operations Personnel Counseled on Events ML20011F3821990-02-26026 February 1990 Confirms Amount Electronically Transferred to Us Dept of Treasury,Nrc on 900223 for Payment of NRC Review Fees of 10CFR50 Applications & 10CFR55 Svcs Per 10CFR170,for Period of 890101-0617 for Listed Invoices ML20011F3371990-02-23023 February 1990 Forwards Corrected Pages to Amend 171 to License DPR-62 Issued on 900206.Amend Changes Tech Specs to Add Footnote to Tables 2.2.1-1 & 3.2.2-2 for Adjustment of Main Steam Line Radiation Monitors When Water Chemistry Sys in Svc 1990-09-06
[Table view] |
Text
'. &
Carolina Power & Light Companj
. January 17, 1979 FILE: NG-3514(B) SERIAL: GD-79-166 Office of Nuclear Reactor Regulation ATTENTION: Mr. T. A. Ippolito, Chief Operating Reactors Branch No. 3 United States Nuclear Regulatory Commission Washington, D. C. 20555 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NO. 1 DOCKET No. 50-325 LICENSE NO. DPR-71 FUEL CYCLE No. 2 - RELOAD LICENSING
Dear Mr. Ippolito:
In accordance with the Code of Federal Regulations, Title 10, Part 50.90 and PIrt 2.101, Carolina Power & Light Company hereby requests revisions to the Technical Specifications for its Brunswick Steam Electric Plant, Unit No. 1. These revisions are necessary to complete the first refueling of Unit No. 1 and begin Cycle 2 operation, and we are submitting them to the Staff as a supplement to our December 29, 1978, Reload Licensing Submittal. Changes to the Technical Specifications are indicated by a vertical line in the right-hand margins of the affected pages which are attached.
A revised safety limit MCPR of 1.07 has been used by General Electric (GE) in their NED0-24166 document and is, therefore, incorporated into our revised Technical Specification pages. The new and the revised graphs of MAPLHGR vs. Planar Average Exposure were prepared using tabulated data from GE's NEDO-24165 document.
The Unit No. 1 vfueling outage began on January 13, 1979, and ascent to power is presently schat ad to begin on March 25, 1979. Due to the over-lapping of outages for Unite 1 and 2, we wish to stress the importance that the Staff review be completed and the appropriate license amendments be issued in a time period that allows Unit No. 1 to return to power in accord with this schedule.
In accordance with 10CFR170.12(c), we have determined that this request constitutes a Class III amendment because it involves a single technical issue. Our check for $4,000 is enclosed as payment for this amendment fee.
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IllSTRUMENTATION 3/4.3.4 CONTROL ROD WITHDRAWAL BLOCK INSTRUMEilTATION LIMITING CONDITION FOR OPERATIO*1 3.3.1 The control rod withdrawal block instrumentation shown in Table 3.3.4-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.4-2.
APPLICABILITY: As shown in Table 3.3.4-1.
ACTION:
- a. With a control rod withdrawal block instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.4-2, declare the channel inoperable until he channel is restored to OPERABLE status with its Trip Setpoirc adjusted consistent with the Trip Setpoint value.
- b. With the requirements for the minimum number of OPERABLE channels not satisfied for one trip system, POWER OPERATION may continue provided that either:
- 1. The inoperable channel (s) is restored to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or
- 2. The redundant trip system is demonstrated OPERABLE within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> until the inoperable channel is restored to OPERABLE status, and the inoperable channel is restored to OPERABLE status within 7 days, or
- 3. For the Rod Block Monitor only, THERMAL POWER is limited such that MCPR will remain above 1.07 assuming a single l error that results in complete withdrawal of any single control rod that is capable of withdrawal.
- 4. Otherwise, place at least one trip system in the tripped condition within the next hour.
- c. With the requirements for the minimum number of OPERABLE channels not satisfied for both trip systems, place at least one trip system in the tripped condition within one hour,
- d. The provisions of Specification 3.0.3 are not applicable in OPERATIONAL CONDITION 5.
SURVEILLANCE REQUIREMENTS 4.3.4 Each of the above required control rod withdrawal block instrumen-tation channels shall be deronstrated OPERABLE by the performance of a CHANNEL CHECK, CHANNEL CALIBRATION and a CHANNEL FUNCTIONAL TEST during the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.4-1.
BRUNSWICK-UNIT 1 3/4 3-39
2.2 LIMITING SAFETY SYSTEM SETTINGS BASES .
2.2.1 REACTOR PROTECTION SYSTEM IflSTRUMENTATION SETPOINTS The Reactor Protection System Instrumentation Setpoints speci-fied in Table 2.2.1-1 are the values at which the Reactor Trips are set for each parameter. The Trip Setpoints have been selected to ensure that the reactor core and reactor coolr.nt system are prevented from exceeding their safety limits,
- l. Intermediate Range Monitor, heutron Flux - High The IRM system consists of 8 chambers, 4 in each of the reactor trip systems. The IRM is a 5 decade 10 range instrument. The trip set-ooint of 120 divisions is active in each of the 10 ranges. Thus as the IRM is ranged uo to accommodate the increase in power level, the trip setpoint is also ranged up. Range 10 allows the IRM instruments to remain on scale at higher power levels to provide for additional overlap.
and also permits calibration at these higher powers.
The most significant source of reactivity change during the power increase are due to control rod withdrawal. In order to ensure that the IRM provides the required protection, a range of rod withdrawal accidents have been analyzed, Section 7.5 of the FSAR. The most severe case involves an initial condition in which the reactor is just subcritical and the IRM's are not yet on scale. Additional conservatism was taken in this analysis by assuming the IRM channel closest to the rod being wi thdrawn i r. bypassed. The results of this analysis show that the reactor is shutdown and peak power is limited to 1% of RATED THERMAL POWER, thus maintaining MCPR above 1.07. Based on this. analysis, the l IRM provides protection against local control rod errors and continuous withdrawal of control rods in sequence and provides backup protection for the APRM.
- 2. Average Power Range Monitor For operation at low pressure and low flow during STARTUP, the APRM scram setting of 15" of RATED THERMAL POWER provides adequate thermal margin between the setpoint and the Safety Limits. This margin accom-modates the anticipated maneuvers associated with power plant startup.
Effects of increasing pressure at zero or low void content are minor, cold water from sources available during startup, is not much colder than that already in the system, temperature coefficients are small and control rod patterns are constrained by the RSCS and RWM. Of all BRUNSWICK-UNIT 1 B 2-9
POWER DISTRIBUTION LIMITS 3/4.2.2 APRM SETPOINTS LIMITING CONDITION FOR OPERATION 3.2.2 The flow biased APRM scram trip setpoint (S) and rod block trip set-point (Sgg) shall be established according to the following relationships:
S < (0.66W + 54%) T S
RB 1 (0.66W + 42%) T wnere: S and 5 are in percent of RATED THERMAL POWER, W = Loch recirculation flow in percent of rated flow, T = Lowest value of the ratio of design TPF divided by the MTPF obtained for any class of fuel in the core (T 1 1.0),and Design TPF for 8 x 8 fuel = 2.45 Design TPF for 8 x 8R fuel = 2.48 APPLICABILITY: CONDITION 1, when THERMAL POWER > 25% of RATED THERMAL POWER.
ACTION:
With 5 or S RA exceeding the allowable value, initiate corrective action within 15 minutes and continue corrective action so that S and S oB are within the required limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.2.2 The MTPF for each class of fuel shall be determined, the value of T calculated, required:
and the flow biased APRM trip setpoint adjusted, as
- a. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,
- b. Whenever THERMAL POWER has been increased by at least 15% of RATED THERMAL POWER and steady state operating conditions have been established, and
- c. Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with a LIMITING CONTROL ROD PATTERN for MTPF.
BRUNSWICK-UNIT 1 3/4 2-4
Bases Table E 3.2.1 't SIGt41 FICA!iT IrlPUT PARAMETERS TO THE l
LOSS-OF-C00LAt4T ACCIDEriT AtiALYSIS FOR BRdilSWICK-ViilT 1 Plant Parameters ;
Core lhermal Power ........... . . . . . . . . 25 31 IMt , which cc,rresponds to 105% of rated steam flot. *
{
Vessel Steam Output .......... .
6
. 10.96 x 10 Lbm/h which corresponds to 105% of rated steam flow Vessel Steam Dome Pressure.... .... 1055 psia Recirculation Line Break Area ,
for Large Breaks - Discharge
- Suction 2.4 ft' (DBA) 1.9 ft (807. DBA) 4.2 ft2 Number of Drilled Bundles 560 l
" "" #8 PEAK TECHflICAL If1ITI AL SPECIFICATI0fi DESIGri MIfilMUM LIfiEAR HEAR AXIAL CRITICAL FUEL BU! IDLE GEfiERATIOl1 RATE PE AKIfiG POWER
' FUEL TiPES GEOMETRY (kw/ f t ) FACTOR RAT 10 "
ALL 8x8 13.4 1.4 1.2 A more detailed list of input to each model and its source is presented in Section II of Reference 1.
- This power level meets the Appendix K requirement of 102%.
, ** T o account for the 27 uncertainty in bundle power required b, Aprendix K, 9'T calculatien is per f orrmd wi t h an MO7 e f 1.15 (i.e., 1. 2 d ivided by 1.t+0.! )
- et a bundle with an initial !!CPE of 1.20.
3RUtiSWICK - UtilT 1 B 3/4 2-2
REACTIVITY CONTROL SYSTEMS BASES CONTROL RODS (Continued) potential effects of the rod ejection accident are limited. The ACTION statements permit variations from the basic requirements but at the same time impose more restrictive criteria for continued operation. A limita-tion on inoperable rods is set such that the resultant effect on total rod worth and scram shape will be kept to a minimum. The requirements for the various scram time measurements ensure that any indication of systematic problems with rod drives will be investigated on a timely basis.
Damage within the control rod drive mechanism could be a generic problem, therefore with a control rod immovable because of excessive friction or mechanical interference, operation of the reactor is limited to a time period which is reasonable to determine the cause of the inoperability and at the same time prevent operation with a large number of inoperable control rods.
Control rods that are inoperable for other reasons are permitted to be taken out of service provided that those in the ncn-fully-inserted position are consistent with the SHUTDOWN MARGIN requirements.
The number of control rods permitted to be inoperable could be more than the eight allowed by the specification, but the occurrence of eight inoperable rods could be indicative of a generic problem and the reactor must be shutdown for investigation and resolution of the problem.
The control rod system is analyzed to bring the reacor subcritical at a rate fast enough to prevent the MPCR from becoming less than 1.07 l during the limiting power transient analyzed in Section 14.3 of the FSAR.
This analysis shows that the negative reactivity rates resulting from the scram with the average response of all the drives as given in the specifications, provide the required protection and MPCR remains greater than 1.07. The occurrence of scram times longer than those specified l
should be viewed as an indication of a systemic problem with the rod drives and therefore the surveillance interval is reduced in order to prevent operation of the reactor for long periods o# time with a potentially serious problem.
Control rods with inoperable accumulators are declared inoperable and Speci fication 3.1.3.1 then applies. This prevents a pattern of inoperable accumulators that would result in less , reactivity insertion BRUNSWICK - UNIT 1 B 3/4 1-2
POWER DISTRIBUTION LIMITS BASES
- 1. General Electric Company A...lytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K, NED0-20566, January, 1976
- 2. General Electric Refill Reflood Calculation (Supplement to SAFE Code Description) transmitted to USAEC by letter, G. L. Gyorey to V. Stello, Jr., dated December 20, 1974.
- 3. Letter from J. A. Jones, Carolina Power and Light Company to B. C. Rusche, NRC transmitting Amendment 31 to the Brunswick Unit 1 Docket No. 50-325, dated November 26, 1975.
- 4. General Electric BWR Generic Reload Application for 8 x 8 Fuel, NEDD-20360, Revision 1, November 1974
- 5. R. B. Linford, Analytical Methods of Plant Transient Evaluations for the GE BWR, February 1973 (NEDO-10302).
- 6. Letter from J. A. Jones, Carolina Power and Light Company, to B. C. Rusche, NRC dated May 7,1976.
BRUNSWICK - UNIT 1 B 3/4 2-6
' s SPECIAL TEST EXCEPTIONS 3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS LIMITING CCNDITION FOR OPERATION 3.10.3 The requircments of Specifications 3.9.1 and 3.9.3 and Table 1.2 may be suspended to permit the reactor node switch to be locked in the Startup position and to allow up t o three control rod.s to be withdrawn for l shutdown raargin demonstrations provided at l e a s t. the following require-nents are satisfie' J.
- a. The source range nonitors are OPERABLE with the RPS circuitry shorting links removed per Specification 3.9.2,
- b. The rod worth ninimizer is OPERABLE per Specification 3.1.4.1 and is programmed for the shutdown margin denonstration, and
- c. The " notch-override" control shall not be used during movement of the control rods.
APPLICABILITY: CONDITION 5, during shutdown margin demonstrations.
ACTION With the requirements of the above specification not satisfied, irrae-diately restore the reactor mode switch to the Refuel position.
SURVEILLANCE REQUIREMENTS 4.10.3 Within 30 ninutes prior to the performance of a shutdown margin denonstration verify that;
- a. The source range monitors are OPERABLE per Specification
- 3. 9.2, a nd
- b. The rod worth ninimizer is OPERABLE with the required pronram, per Speci fication 3.1.4.1, BRUNSWICK - UNIT 1 3/4 10-3
T
_ABLE 3.3.4-2 en
- o g CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION SETPOINTS w
5 h TRIP FUNCTION AND INSTRUMENT NUMBER TRIP SETPOINT ALLOWABLE VALUE E
5 1. APRM (C51-APRM-CH.A,B,C,D,E,F)
- a. Upscale (Flow Biased) < (0.66 W + 425) T* < (0.66 W + 425) T*
- b. Inoperative fiA fli PF fiA KTPF
- c. Downscale s 3/125 of full ;cale > 3/125 of full scale
- d. Upscale (Fixed) [12%ofRATEDTHERMALPOWER [125ofRATEDTHERMALPOWER
- 2. _ ROD BLOCK MONITOR (C51-RBM-CH.A,B)
- a. Upscale < (0.66W + 40;) T* < (0.66 W + 40% ) T* l
- b. Inoperative NA MTPF NA MIPF
}{ c. Downscale > 3/125 of full scale > 3/125 of full scale SOURCE RA1GE MONITORS (C51-SRM-K600A,B,C,D)
- a. Detector not full in NA 5
"A 5
- b. Upscale < 1 x 10 cps < 1 x 10 cps
- c. Inoperative NA NA
- d. Downscale > 3 cps > 3 cps
- 4. INTERMEDIATE RANGE MONITORS (C51-IRM-K601A,B,C,D,E,F,G,H)
- a. Detector not full in MA NA
- b. Upsmale < 103/125 of full scale < 108/125 of full scale
- c. Inoperative NA NA
- d. Downscale > 3/125 of full scale s 3/125 of full scale
- T = 2.45 for 8 x 8 fuel T = 2.48 for (8 x 8)R fuel
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FUEL TYPE 8DRB28311 (8x8R)
MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLEGR)
VERSUS PLANAR AVERACE EXPOSURE FIGURE 3.2.1-4
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FUEL TYPE 8DRB265L (8x8R)
MAXIMUM AVERACE PLANAR LINEAR llEAT GENERATION RATE (MAPLllGR) .
VERSUS PLALiR AVERAGE EXPOSURE FIGURE 3.2.1-3
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FUEL TYPE 2 (8x8)
EiXIMUM AVERAGE PLA';AR LINEAR l! EAT GENERATION RATE (MAPLilGR)
VERSUS PLANAR AVERAGE EXPOSURE FIGGE 3. 2.1- 2
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FUEL TYPE 1 (8x8)
MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLilGR)
VERSUS PLA.!AR AVERAGE EXPOCURE FIGURE 3. 2.1- 1
I PC'.ER DIS TRIBUT IC'; L 111 TS 3/4.2.3 M I '; I:*0. ' CRITIC /1 PU.Ei, R AT I O L IMIT II.S CONDIM r0k Gi'E!KT IM
- 3. 2. 3 The M'NIMUM CR!TICAL POWER 'lATIO (FCPR), as a function of core flow, shali ,e equai to or greater than :tCpg <. tne t' 7 s r.c. . in figure 3.2.3-1 where
?!CPR - 1.22 tor lioC2 t o (U C - 2000 'r.-lD/t)*
!!CPR = 1.21 tor (EOC2 - 2000 'rJD/t) to (EOC2 - 1 '
,?fdD/t)
':C P;' - 1.2M !or (FuC2 - 1000 .lD / t ) to I:OC2 APPLIU4!LITY:
CCNDI T I. .'; 1, nhe . T Hi f&'_ 'tT.E R 255 RATED THERMAL
' u.. u A C.T.1.0.'. :
With MCPR less thy 'rn: o ppl i e. :.L l e l i ni t de te r: i r.ed t rc:, F i gu re 3.2.3-1, initiate : o r r e c. t i v e :c t ion ' i th i' 15 ii ma t c" ,:nd car;i ra cor rec ti', -
ac* ion so that .Ric om.:I tu or qcea ter than Lt.e irolicable limit w . I h i n 4 'ua" or re d o..e T H F ' . '., c i ' .' ,. R t o l e v, t h a r. J b uf RAIED THEk"AL i>/.nR within the n4 .t
. . i.uars SURV E il i S.C! Ri C,ili I'" ';TS
. 1' .1 MCPR ',ha l l tm de te n . i ned t o L;e eauat to or greater th.n the applicable l mi t ce ten ined f ra . Figure ' .9. a-l :
- a. At leas t or.ce per .'a huurs ,
- b. Wher ner THERVAL IUWER has been increased by at least 155 01 U ATl 0 THi!d'3! F f '.-?! D and eteady state operating conditions havt been estaulished, ar.J c Init.ially ar.J at least cnce per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating witn a LIMITING CONTROL ROD PATTERN for "CPR.
- BOC2 Bea,inninR of Cycle 2 EOC2 Cad o: Cyc 1, . 3 BRUNS'.'IC K-UNIT 1 3/4 2-0
IiWER DISTRIBUTIC', LIMITS r , ,.
i- .J 3/4.2.2 APr " 5F T PU:',TS The tuel e l eold in e integrity seifety l im i t s of Specification 2.1 were based on
.i Tu l'Ai. Pl.AK I M; F/u;1 t lK u t 2.43 lor 8:;8 luel and 2.43 for 8xSR inel. the scram setting and rod block f unctions of the ;H " instrur ents must be adjusted to ensure that t he 'iCPR does r.J t become les s
- na r 1.0 in the degraded situation. Tne scra.- s et t i r.g s and red blac k s e t t i ns; 3 are adjusted in a c c o r d a r,c t. with, tne forc:ula in this specificaticn when the combination o f THER"T,L PC.;E P u.d p ea t. flux indicnes a TOTM PE/J.It;C FACTOR greater inan 2.03. The ne'hcd used to determine the design TPF shall be con-sinent .;ith the >thod used to determine the 'iTPF-3/4.'>3 I . U""'
- C R I T I CM v'A % i10 Tne required operating limit MCPP's at steady state operating conditions as specified in Specification 3.2.3 are derived from the established fuel cladding intejrity Satety L l ahnurual oparational transier.ts'j yit "CP;, of 1.07, FCr any and Cperating abnormal an analysis of tran-sient analysis evaluaticn eith the initial condition of the reactor Leing at the ',teadj state operating linit, it is required that the resul ting MCPR Joes not decrease Leicw the Safety Lmit "CPR at any tine during the transient assum n; instrument trip setting as given in S pec i fica ticn 2.2.1, To assure that the fuel cladding 'ntegrity Safety Limit is r.ot exceeded during any anticipated abncrmal operacioral transient, the most limiting transients have been analy:ed to datermine which result in the largest reduction in CRITICAL PCUER 1ATIO (CPR), The type of transients evaluated
,eere losa of flo.., increase ,o pressure and power, positive reactivity insertion, and coolant tem erature decrease.
The limiting transient which determines the required steady stat. MCPR l i:li t is the turbine trip with failure of the turbine by pass. This transient yield the largest . MCPR. When added to the Safety Linit MCFR of 1.07 the required minimum operating limit MCPR of Specification l 3.2.3 is obtained- Prior to the analysis of abnormal operational tran-sients an initial fuel bundle MCPR was determined. This parameter is based on the bundle fic s calculatcd by a GE rulti-channel steady state flow distribution nodel as described in Section 4.4 of NEDO-20360 \"> and on core paraneters shown in Reference 3, response to Itens 2 and 9.
BRU"SWICK - UNIT 1 B 3/a 2-3
R E ACT I V I T Y C0 tit RO.L S Y S T.ly._S ROD BLOCK MONITOR LIMITING CONDITION _EUR OPERATION ._ ____ _ _ _ , _ _ _ _ . ____
3.l.4.3 Both Rod Black Monitor (RBM) channels shall be OPERABLE.
T@ PL IC ABI L I T Y_ :
CONDITICU 1, when THEPMAL POWER is greater th_n the preset po.,er level of the RWM and RSES.
AC T 10ti:
- d. With one RCM Channel inoperdble, PO!.TR OPERATIOil may continue provided that either
- 1. The inoperanle RBM channel is restored to OPERABLE status within ?4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or
- 2. The redundant RBM is demonstrated CPERABl.E within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and at lea r+ once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> until the inoperable RBM is restared to GPiRACLE status, and the inoperable RBM is restored to WilRABLE status within 7 days, or
- 3. THERMAL P0.!ER is limited such that MCPR will remain above 1.07 assuning a single error that results in complete l withdrawol of any single control rod that is capable of withdrawal, Otherwise, trip at least one rod block monitor channel,
- b. With both RBM channels inoperable, trip at least one rod block monitor channel within one hour.
SURVEILLANCE REQUIRLMENTS ___ _
4.1.4.3 Each of the above required RBM channels shall be demanstrated OPERABLE by perforunce of a CHANNLL FUNCTIONAL TEST and CHANUf L CALI-BRATION at the frequencies and during the OPERATIONAL CONDITIONS specified in Table 4.3.4-1.
BRUNSWICK-UNIT l 3/4 1-17
2 .1 S A F E T Y L I M I.T.S.
BASES
.----.-.- ..----- - .-._.-- - - = .. -
Tm fuel clcddirg, reattor pressure vessel and priracy system p u i. 3 a. Un prir ciN1 barriers tc tne release of radioactive '1at-erials to the envir m . Eifety lir .ts are established to protect the integrity of m se 'urriers during r,ceral plant operations and antici-pated transients iho fuel ciaddina :ntegrity linit is set such T.nat no f uel d. 7;e is calculated to occur if the linit is not. violated. Be-cause fuel dram is not directly otrervable, a step-back approach is used to establis' a Ea fety Linit such that the i l N I KJ" C R I T I C A L P G'.-lE R RATIO (CPR) is no less than 1.07. !'CPR I.07 represents a conserva- l tive margin relative to th. conditicns required to raintain fuel cladding integrity. The .~ fl cladding is cne of the physical barriers which separate the radiacctive oatr: rials f ecr tne environs. The integrity of this Lladding L.:rr ier in related t o i ts relative freedon f rom perfor-ations ce crack.ing. Altho..qh some corrosion or use related cracking na) cccur during the life of tr.e claddin1, fission product migratica frcm this source is increnentally cumulative and continuously neasurable.
Fuel cladding perf nca t ion s , hov,evc r , can resul t from thernal stresses which occur frac reactor operation sienificantly above design conditions and the Linitin' % fety , s t e. Settings. 'clhi l e fi ss ion product ni gra-tion teori cldJina ;mrforation is just as leasurenle as that from use rel a ted crack i r.g , t he then tally caused cladding perforations signal a tnreshold, beyoaa which still 7reater thermal stresses may cause gross rather tnan incrwentel claddina Jetericration. Therefore, the fuel cladJing Safety L ti t is defined with a narr;in to the conditions which would produce onset of transition hoiling, PCPR of 1.0. These con-ditions represent a significant departure from the conditicn intended by design for planned operation.
2.1.1 THER'GL PBER (Low Pressure or Low Flow)
The use of the GEXL correlation is not valid for all critical power calcuidtions at pressures oeloa 300 psia or core flows less than 101 rs rated flow. Therefore the fuel cladding integrity linit is established by other neans. This is done by establishing a limiting condition on core THER"AL PC'.ER with the following basis. Since tha pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low po..or and fio.ss will always be greater than 4.5 psi. Analyses s ho.s that with a flow of 23 x 10' lbs/hr bundle flow, bundle pressure drop is nearly independent of Lundle pae: e,d has a value of 3.5 psi, Thus, t."e Sundle flow with a , 5 psi crivinn head will be greater than 23 x 103 lbs/hr Full scale ATLM test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assenbly critical power at this flow i s approxinately 3. 35 M'-lt . With the design peaking factors, this corresponds to a IHERMAL PCWER of more than 509 of RATED THERMAL POWER. Thus, a THER"AL PCWLR linit of 251 of RATED THERMAL PC',!ER for reactor pressure below S00 psia is conservative.
BRUNSWICh-UNIT 1 B J-l
- u. -
1 2.0 SAFETY LIMITS T.';D LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMI_TS THERMAL. PCWER (Loa Proscure or Lc. Flc..';
2 .1.1 THERMAL PC'lER shall not exceed 25L of RATED THE RML POWER with the renctor .essel stean dom pressure less than 800 ps;a or core flow less than 10 . o f ca tec f I .re.
APPLICABIllTY: CONDIT IC';5 1 and 2.
ACT 10'.
With THERMAL PO'-lER exceeding 25 of RATED THERMAL PC'ER and thc reactor sessel s team dc:Je pressure less than i"20 psia cr core low less than 101 of ratej tiow, Le it a t leas t HOT SHaTDC'..N wi thin 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> .
THERMALLOWLR(HiohPressureandHiohFlc.d 2.1.2 The MINIMUM CRITICAL PCWER RATIO (MCPR) shall not be less tnan 1.07 with the reactor vessel stean. dote pre .sure greater than 800 psia l and core flu grea Ne than 10~ of rated flce.
APPLICABILITY:
CONDITIONS 1 and 2.
ACTION:
With MCPR less than 1 07 and the reactcr vessel steam dome pressuro l ureatcr than 800 psia and core flcw greater than 10' of ratcd flos. be in at least HOT SHUIDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
REACTOR COOLANT B STEM PRESSURE The reactor coolant system pressure, as measured in the reactor h 2.1.3 vessel stean do:ne, shall not exceed 1325 piig.
APPLICABILIT(: CONDITIONS 1, 2, 3 and 4 ACTICN-With the reactor coulant system pressure, as neasured in the reactor vessel s tean: done, above 1325 psig, t,e in at least HOT SHUTCOWN with reactor coolant system pressure < 1325 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
BRUNSWICK-UNIT 1 2-1
};r. T. A. Ippolito January 17, 1979 If your staff has any questions concerning the attached information, we will be glad to discuss them e!.ther by telephone or at a meeting with repre-sentatives of your staff.
Yours very truly,
/?
"Q g-' .A) Q- , (/ Qg4 E. E. Utley Senior Vice President Pouer Supply JAM /mf Enclosures Sworn to and subscribed before me this 17th day of January, 1979 Notary Public My Commission Expires: October 4, 1981
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