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Category:CORRESPONDENCE-LETTERS
MONTHYEARML20217M3371999-10-20020 October 1999 Forwards Notice of Docketing of License SNM-2506 Amend Application.Notice Has Been Forwarded to Ofc of Fr for Publication ML20217M1111999-10-19019 October 1999 Forwards Insp Repts 50-282/99-14 & 50-306/99-14 on 990920- 22.One Violation Noted & Being Treated as Ncv.Insp Focused on Testing & Maint of Heat Exchangers in High Risk Sys ML20217F4331999-10-15015 October 1999 Forwards Rev 39 to Security Plan.Changes Do Not Decrease Effectiveness of Security Plan.Rev Withheld,Per 10CFR73.21 ML20217C2351999-10-0606 October 1999 Forwards Insp Repts 50-282/99-12 & 50-306/99-12 on 990823-0917.No Violations Noted.Insp Consisted of Selected Exam of Procedures & Representative Records,Observation of Activities & Interviews with Personnel ML20212J8811999-09-28028 September 1999 Forwards Preliminary Accident Sequence Precurson Analysis of Operational Event That Occurred at Plant,Unit 1 on 990105, for Review & Comment.Comment Requested within 30 Days of Receipt of Ltr IR 05000272/19990071999-09-28028 September 1999 Forwards Insp Repts 50-272/99-07 & 50-306/99-07 on 990721- 0831.One Potentially Safety Significant Issue Identified Dealing with Control Room Special Ventilation System.Four Addl Issues of Low Safety Significance Identified ML20212G7171999-09-24024 September 1999 Submits Semiannual Status Update on Project Plans for USAR Review Project & Conversion to Its.Conversion Package Submittal Continues to Be Targeted for Aug of 2000 ML20212G9801999-09-23023 September 1999 Refers to Resolution of Unresolved Items Identified Re Security Alarm Station Operations at Both Monitcello & Prairie Island ML20212F5121999-09-20020 September 1999 Forwards Response to NRC , Preparation & Scheduling of Operator Licensing Examinations ML20212D8401999-09-16016 September 1999 Discusses 990902 Telcon Between D Wesphal & R Bailey Re Administeration of Retake Exam at Prairie Island During Wk of 991206.NRC May Make Exam Validation Visit to Facility During Wk of 991116 ML20217H2331999-09-10010 September 1999 Forwards Security Insp Repts 50-282/99-10 & 50-306/99-10 on 990809-12.Two Findings,Each of Low Risk Significance Identified & Being Treated as non-cited Violations Consistent with App C of Enforcement Policy ML20217H5661999-09-0909 September 1999 Discusses 990907 Pilot Plan Mgt Meeting Re Results to-date of Pilot Implementation of NRC Revised Reactor Oversight Process at Prairie Island & Quad Cities.Agenda & Handouts Provided by Utils Encl ML20212A9241999-09-0909 September 1999 Discusses Plans Made During 990902 Telephone Conversation to Inspect Licensed Operator Requalification Program at Prairie Island During Weeks of 991101 & 991108.Requests That Written Exams & Operating Tests Be Submitted by 991022 ML20212B0511999-09-0909 September 1999 Forwards Insp Repts 50-282/99-11 & 50-306/99-11 on 990816-20.One Issue of Low Safety Significance Was Identified & Being Treated as Ncb ML20211Q7641999-09-0808 September 1999 Informs That NRC Tentatively Scheduled Initial Licensing Exam for Plant Operator License Applicants During Wk of 000515,in Response to D Westphal ML20211N8631999-09-0707 September 1999 Withdraws 970814 Request for Exemption from 10CFR50,App R, Section III.G.2, Fire Protection of Safe Shutdown Capabilities ML20211K5911999-09-0101 September 1999 Informs That Util Reviewed Rvid Data Base,As Requested in NRC .Summary of Proposed Changes & Observed Differences Are Included in Encl Tables ML20211L0211999-09-0101 September 1999 Provides Notification That License Amends 141 & 132 & Associated License Conditions 6 & 7 Have Been Fully Implemented ML20211K5931999-08-31031 August 1999 Forwards License Amend Request for License SNM-2506, Proposing Change to License Conditions 6,7 & 8 & TSs App a of License by Permitting Inclusion of Bpras & Thimble Plug Devices in Sf Assemblies Stored in TN-40 Casks ML20211Q6041999-08-31031 August 1999 Forwards Rev 19 to USAR for Pingp,Per 10CFR50.71(e).Rev Brings USAR up-to-date as of 990228,though Some Info Is More Recent.Attachment 1 Contains Descriptions & Summaries of SE for Changes,Tests & Experiments,Per 10CFR50.59 ML20211K2591999-08-27027 August 1999 Forwards NSP Co Fitness for Duty Program Performance Data for Six Month Period Ending 990630 ML20211D3541999-08-24024 August 1999 Discusses GL 95-07 Re Pressure Locking & Thermal Binding of safety-related Power Operated Gate Valves.Forwards SE Re Response to GL 95-07 ML20211C7601999-08-19019 August 1999 Confirms NRC Intent to Meet with NSP & Ceco on 990807 in Lisle,Il to Discuss with Region III Pilot Plants,Any Observations,Feedback,Lessons Learned & Recommendations Relative to Implementation of Pilot Program ML20211B8311999-08-19019 August 1999 Forwards Request for Relief 8 Re Limited Exams Associated with Unit 1 Third ten-year Interval Inservice Insp Program. Licensee Requests Relief Due to Impractibility of Obtaining 100% Exam Coverage for Affected Items ML20211B5711999-08-19019 August 1999 Forwards Second 90-day Rept for Implementation of Voltage Based Repair Criteria at Prairie Island Unit 1.Rept Fulfills Requirements of Section 6.b of Attachment 1 to GL 95-05 ML20211C2311999-08-19019 August 1999 Forwards Unit 1 ISI Summary Rept,Interval 3,Period 2 Refueling Outage Dates 990425-0526,Cycle 19 971212-990526. Rept Identifies Components Examined,Exam Methods Used,Exam Number & Summarizes Results ML20211B0561999-08-18018 August 1999 Provides Addl Info on Proposed Rev to Main Steam Line Break Methodology ,in Response to NRC Staff Request Made in 990416 Telcon.Nuclear Svcs Corp Rept PIO-01-06, Analysis Rept Structural Analyses of Main Steam Check... Encl ML20211B2621999-08-17017 August 1999 Forwards Insp Repts 50-282/99-09 & 50-306/99-09 on 990719-22.No Violations Noted.Insp Included Review & Evaluation of Current Emergency Preparedness Performance Indicators ML20211C7371999-08-17017 August 1999 Discusses Closure of Staff Review Re Generic Implication of Part Length Control Rod Drive Mechanism Housing Leak on 980123.Enclosed NRC 980811 & 1223 Ltrs Responded to WOG Positions Re Corrective Actions ML20210T5661999-08-12012 August 1999 Forwards RAI Re & Suppl ,which Requested Exemptions from TSs of Section III.G.2 of 10CFR50 App R,To Extent That Specifies Separation of Certain Redundant Safe Shutdown Circuits with fire-related Barriers ML20210R7021999-08-12012 August 1999 Forwards Insp Repts 50-282/99-06 & 50-306/99-06 on 990601- 0720.One NCV Occurred,Consistent with App C of Enforcement Policy ML20210P5191999-08-11011 August 1999 Discusses GL 92-01,Rev 1,Supp 1, Rv Integrity, Issued by NRC on 950519 & NSP Responses for PINGP & 951117. Staff Reviewed Info in Rvid & Released Info as Rvid Version 2.Requests Submittal of Comments Re Revised Rvid by 990901 ML20210G5061999-07-30030 July 1999 Responds to Administrative Ltr 99-02, Operating Reactor Licensing Action Estimates 05000282/LER-1999-007, Forwards LER 99-007-00,re Loss of CR Special Ventilation Function.One New Commitment Was Made in Rept as Indicated in Corrective Action Section Statement in Bold Italics1999-07-23023 July 1999 Forwards LER 99-007-00,re Loss of CR Special Ventilation Function.One New Commitment Was Made in Rept as Indicated in Corrective Action Section Statement in Bold Italics ML20210J4991999-07-22022 July 1999 Forwards Rev 18 to USAR for Pingp,Bringing USAR up-to-date as of 990228,though Some Info More Recent.Safety Evaluation Summaries Also Encl ML20209J0941999-07-15015 July 1999 Forwards SER Finding Rev 7 to Topical Rept NSPNAD-8102, Reload Safety Evaluation Methods for Application to PI Units, Acceptable for Ref in Plant Licensing Actions ML20209H8051999-07-14014 July 1999 Forwards Summary of non-modification Safety Evaluation Number 515 Re Storage of Fuel Inserts,Per Insp Rept 72-0010/99-201 ML20209D4181999-07-0707 July 1999 Informs That Util Has Changed Listed TS Bases Pages Attached for NRC Use.Util Made No New Commitments in Ltr ML20209H8361999-07-0202 July 1999 Forwards Operator Licensing Exam Repts 50-282/99-301(OL) & 50-306/99-301(OL) for Tests Administered During Week of 990517-21.Two Applicants Passed All Sections of Exam & Issued Reactor Operator Licenses to Operate Pings ML20196J9681999-07-0101 July 1999 Informs That in Sept 1998,Region III Received Rev 20 to Portions of Util Emergency Plan Under 10CFR50.54(q).Based on Determination That Changes Do Not Decrease Effectiveness of Licensee Emergency Plan,No NRC Approval Required ML20209C3951999-07-0101 July 1999 Forwards Supplemental Response to GL 95-07, Pressure Locking & Thermal Binding of SR Power-Operated Gate Valves ML20209B7541999-07-0101 July 1999 Final Response to GL 98-01,Suppl 1 Re Y2K Readiness of Computer Sys.Sys Remediated as Required for Plant Operation. Contingency Plans Developed to Mitigate Impact of Y2K-induced Events at Key Rollover Dates ML20196J8941999-06-30030 June 1999 Transmits Util Comments on Draft Regulatory Guide DG-1074, Steam Generator Tube Integrity. Licensee Recommends That NRC Focus on Several Important Listed Areas Considered Principal Concerns & Contentions ML20209F0391999-06-30030 June 1999 Forwards Insp Repts 50-282/99-04 & 50-306/99-04 on 990407-0531.Violation Noted.Notice of Violation or Civil Penalty Will Not Be Issued,Based on NRC Listed Decision to Exercise Discretion ML20209C3011999-06-29029 June 1999 Forwards Annual Rept of Corrections to NSP ECCS Evaluation Models,Iaw 10CFR50.46.Since All Analyses Remain in Compliance,No Reanalysis Is Required or Planned ML20209B5751999-06-24024 June 1999 Submits Revised Relief Request for Limited Examinations Associated with Third 10-yr ISI Examination Plan.Attached Is Unit 1 Relief Request 7,rev 1 Which Addresses Limited Examinations ML20196F3871999-06-23023 June 1999 Forwards Revised Pages 71,72 & 298 of Rev 7 to NSPNAD-8102, Prairie Island Nuclear Power Plant Reload Safety Evaluation Methods for Application to PI Units, Per Discussions with Nrc.Approved Version of Rept Will Be Issued 05000282/LER-1999-006, Forwards LER 99-006-00 Re Discovery That Manual SI Actuation Switch Had Not Been Tested on Staggered Basis During Integrated SI Test.Two New Commitments Are Indicated in Corrective Action Section Statement in Bold Italics1999-06-18018 June 1999 Forwards LER 99-006-00 Re Discovery That Manual SI Actuation Switch Had Not Been Tested on Staggered Basis During Integrated SI Test.Two New Commitments Are Indicated in Corrective Action Section Statement in Bold Italics ML20196D5501999-06-18018 June 1999 Forwards Individual Exam Results for Licensee Applicants Who Took May 1999 Initial License Exam.In Accordance with 10CFR2.790,info Considered, Proprietary. Without Encls ML20196A6741999-06-17017 June 1999 Refers to 990517-20 Meeting with Util in Welch,Minnesota Re Licensee Initiatives in Risk Area & to Establish Dialog Between SRAs & Licensee PRA Staff 1999-09-09
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARML20217F4331999-10-15015 October 1999 Forwards Rev 39 to Security Plan.Changes Do Not Decrease Effectiveness of Security Plan.Rev Withheld,Per 10CFR73.21 ML20212G7171999-09-24024 September 1999 Submits Semiannual Status Update on Project Plans for USAR Review Project & Conversion to Its.Conversion Package Submittal Continues to Be Targeted for Aug of 2000 ML20212F5121999-09-20020 September 1999 Forwards Response to NRC , Preparation & Scheduling of Operator Licensing Examinations ML20211N8631999-09-0707 September 1999 Withdraws 970814 Request for Exemption from 10CFR50,App R, Section III.G.2, Fire Protection of Safe Shutdown Capabilities ML20211K5911999-09-0101 September 1999 Informs That Util Reviewed Rvid Data Base,As Requested in NRC .Summary of Proposed Changes & Observed Differences Are Included in Encl Tables ML20211L0211999-09-0101 September 1999 Provides Notification That License Amends 141 & 132 & Associated License Conditions 6 & 7 Have Been Fully Implemented ML20211Q6041999-08-31031 August 1999 Forwards Rev 19 to USAR for Pingp,Per 10CFR50.71(e).Rev Brings USAR up-to-date as of 990228,though Some Info Is More Recent.Attachment 1 Contains Descriptions & Summaries of SE for Changes,Tests & Experiments,Per 10CFR50.59 ML20211K5931999-08-31031 August 1999 Forwards License Amend Request for License SNM-2506, Proposing Change to License Conditions 6,7 & 8 & TSs App a of License by Permitting Inclusion of Bpras & Thimble Plug Devices in Sf Assemblies Stored in TN-40 Casks ML20211K2591999-08-27027 August 1999 Forwards NSP Co Fitness for Duty Program Performance Data for Six Month Period Ending 990630 ML20211C2311999-08-19019 August 1999 Forwards Unit 1 ISI Summary Rept,Interval 3,Period 2 Refueling Outage Dates 990425-0526,Cycle 19 971212-990526. Rept Identifies Components Examined,Exam Methods Used,Exam Number & Summarizes Results ML20211B8311999-08-19019 August 1999 Forwards Request for Relief 8 Re Limited Exams Associated with Unit 1 Third ten-year Interval Inservice Insp Program. Licensee Requests Relief Due to Impractibility of Obtaining 100% Exam Coverage for Affected Items ML20211B5711999-08-19019 August 1999 Forwards Second 90-day Rept for Implementation of Voltage Based Repair Criteria at Prairie Island Unit 1.Rept Fulfills Requirements of Section 6.b of Attachment 1 to GL 95-05 ML20211B0561999-08-18018 August 1999 Provides Addl Info on Proposed Rev to Main Steam Line Break Methodology ,in Response to NRC Staff Request Made in 990416 Telcon.Nuclear Svcs Corp Rept PIO-01-06, Analysis Rept Structural Analyses of Main Steam Check... Encl ML20210G5061999-07-30030 July 1999 Responds to Administrative Ltr 99-02, Operating Reactor Licensing Action Estimates 05000282/LER-1999-007, Forwards LER 99-007-00,re Loss of CR Special Ventilation Function.One New Commitment Was Made in Rept as Indicated in Corrective Action Section Statement in Bold Italics1999-07-23023 July 1999 Forwards LER 99-007-00,re Loss of CR Special Ventilation Function.One New Commitment Was Made in Rept as Indicated in Corrective Action Section Statement in Bold Italics ML20210J4991999-07-22022 July 1999 Forwards Rev 18 to USAR for Pingp,Bringing USAR up-to-date as of 990228,though Some Info More Recent.Safety Evaluation Summaries Also Encl ML20209H8051999-07-14014 July 1999 Forwards Summary of non-modification Safety Evaluation Number 515 Re Storage of Fuel Inserts,Per Insp Rept 72-0010/99-201 ML20209D4181999-07-0707 July 1999 Informs That Util Has Changed Listed TS Bases Pages Attached for NRC Use.Util Made No New Commitments in Ltr ML20209C3951999-07-0101 July 1999 Forwards Supplemental Response to GL 95-07, Pressure Locking & Thermal Binding of SR Power-Operated Gate Valves ML20209B7541999-07-0101 July 1999 Final Response to GL 98-01,Suppl 1 Re Y2K Readiness of Computer Sys.Sys Remediated as Required for Plant Operation. Contingency Plans Developed to Mitigate Impact of Y2K-induced Events at Key Rollover Dates ML20196J8941999-06-30030 June 1999 Transmits Util Comments on Draft Regulatory Guide DG-1074, Steam Generator Tube Integrity. Licensee Recommends That NRC Focus on Several Important Listed Areas Considered Principal Concerns & Contentions ML20209C3011999-06-29029 June 1999 Forwards Annual Rept of Corrections to NSP ECCS Evaluation Models,Iaw 10CFR50.46.Since All Analyses Remain in Compliance,No Reanalysis Is Required or Planned ML20209B5751999-06-24024 June 1999 Submits Revised Relief Request for Limited Examinations Associated with Third 10-yr ISI Examination Plan.Attached Is Unit 1 Relief Request 7,rev 1 Which Addresses Limited Examinations ML20196F3871999-06-23023 June 1999 Forwards Revised Pages 71,72 & 298 of Rev 7 to NSPNAD-8102, Prairie Island Nuclear Power Plant Reload Safety Evaluation Methods for Application to PI Units, Per Discussions with Nrc.Approved Version of Rept Will Be Issued 05000282/LER-1999-006, Forwards LER 99-006-00 Re Discovery That Manual SI Actuation Switch Had Not Been Tested on Staggered Basis During Integrated SI Test.Two New Commitments Are Indicated in Corrective Action Section Statement in Bold Italics1999-06-18018 June 1999 Forwards LER 99-006-00 Re Discovery That Manual SI Actuation Switch Had Not Been Tested on Staggered Basis During Integrated SI Test.Two New Commitments Are Indicated in Corrective Action Section Statement in Bold Italics ML20195G4281999-06-0909 June 1999 Notifies That Amsac/Dss Mods Completed & TS 138/129 Has Been Fully Implemented 05000282/LER-1999-005, Forwards LER 99-005-00 Re Containment Inservice Purge Sys Not Isolated During Heavy Load Movement Over Fuel.Event Has Indicated That Level of Performance Expected by Mgt Has Not Yet Been Achieved1999-06-0707 June 1999 Forwards LER 99-005-00 Re Containment Inservice Purge Sys Not Isolated During Heavy Load Movement Over Fuel.Event Has Indicated That Level of Performance Expected by Mgt Has Not Yet Been Achieved ML20207F4301999-06-0101 June 1999 Forwards 1999 Unit 1 SG Insp Results,Per TS 4.12.E.1. Following Insp 84 Tubes Were Plugged for First Time ML20196L2461999-05-21021 May 1999 Forwards Rev 0 to COLR for Pingp,Unit 1 Cycle 20, IAW TS Section 6.7.A.6 ML20195C6861999-05-21021 May 1999 Forwards Rev 17 to USAR for Prairie Island Nuclear Generating Plant.Attachment 1 Contains Descriptions & Summaries of SEs for Changes,Tests & Experiments Made Under Provisions of 10CFR50.59 During Period Since Last Update ML20206U6781999-05-17017 May 1999 Forwards Revised Emergency Response Plan Implementing Procedures,Including Rev 15 to F3-3,rev 15 to F3-16,rev 14 to F3-22 & Table of Contents ML20206U7131999-05-17017 May 1999 Forwards Revised EOF Emergency Plan Implementing Procedures, Including Table of Contents & Rev 2 to F8-10, Record Keeping in Eof. with Updating Instructions ML20206T2461999-05-17017 May 1999 Forwards Off-Site Radiation Dose Assessment for Jan-Dec 1998, Rev 0 to Annual Radiactive Effluent Rept for 980105- 990103 & Effluent & Waste Disposal Annual Rept Solid Waste & Irradiated Fuel Shipments,Jan-Dec 1998 ML20206R0401999-05-13013 May 1999 Forwards Application for Amends to Licenses DPR-42 & DPR-60, Removing Plant Organization Requirement,Imposed in Amend 141/132 That Plant Manager,Who Has Responsibility for Overall Safe Operation of Plant,Report to Corporate Officer ML20206Q0871999-05-13013 May 1999 Forwards Result of Evaluation Re Ultrasonic Exams of SG Number 22 Performed in Accordance with ASME Boiler & Pressure Vessel Code Section Xi.Procedure Used for Evaluation Contained in WCAP-14166,submitted for Review ML20206F9381999-05-0303 May 1999 Forwards Response to NRC 990304 RAI Re GL 96-05 Program at Pingp.Licensee Commitments Are Identified in Encl as Statements in Italics ML20206J3851999-05-0303 May 1999 Forwards 1998 Annual Radiological Environmental Monitoring Rept 05000282/LER-1999-004, Forwards LER 99-004-00 Re Discovery of Inadequate Sp That Demonstrates Operability of SFP Special Ventilation Sys.Two New NRC Commitments Are Contained in Corrective Action Section of Rept in Bold Italics1999-05-0303 May 1999 Forwards LER 99-004-00 Re Discovery of Inadequate Sp That Demonstrates Operability of SFP Special Ventilation Sys.Two New NRC Commitments Are Contained in Corrective Action Section of Rept in Bold Italics ML20206E1761999-04-28028 April 1999 Forwards Revised TS Pages for Amends 144 & 135 to Licenses DPR-42 & DPR-60,respectively,to Update Controlled Manual or File ML20205S3221999-04-20020 April 1999 Forwards Application for Amends to Licenses DPR-42 & DPR-60, Changing Implementation Date for Relocation from TS to UFSAR of Requirements in TS 3.1.E & Flooding Shutdown Requirements of TS 5.1 ML20205P9891999-04-12012 April 1999 Requests Approval for Proposed Alternatives to Liquid Penetrant Requirements of N-518.4 of 1968 ASME Boiler & Pressure Vessel Code.Results of Analysis & Summary of Tests Performed & Tests Results Are Encl ML20205Q0191999-04-12012 April 1999 Forwards Application for Amend to License DPR-42 & DPR-60, Relocating Shutdown Margin Requirements from TS to COLR 05000282/LER-1998-010, Forwards LER 98-010-01 Re Discovery That 32 App R Related MOVs Are Susceptible to Physical Damage by Fire Induced Hot Shorts.Rept Provides Addl Details on Current Plans for Completing C/As Committed to in Original LER1999-04-0808 April 1999 Forwards LER 98-010-01 Re Discovery That 32 App R Related MOVs Are Susceptible to Physical Damage by Fire Induced Hot Shorts.Rept Provides Addl Details on Current Plans for Completing C/As Committed to in Original LER ML20205P9221999-04-0101 April 1999 Submits Relief Request 8,rev 0 Which Addresses Limited Exams Associated with Unit 2 Third ten-year Interval Inservice Insp Program.Util Requests Relief Per 10CFR50.55a(q)(5)(iii) Due to Impracticality of Obtaining 100% Exam Coverage ML20205E8371999-03-31031 March 1999 Submits Four Copies of Rev 38 to Prairie Island Security Plan,Per 10CFR50.54(p).Changes Do Not Decrease Effectiveness of Security Plan.Encl Withheld,Per 10CFR73.21 ML20196K7831999-03-31031 March 1999 Forwards Decommissioning Funding Status Rept for Monticello & Prairie Island Nuclear Generating Plants,Per Requirements of 10CFR50.75(f)(1) ML20205Q5051999-03-30030 March 1999 Forwards Inservice Insp Summary Rept Interval 3,Period 1 & 2 Refueling Outage Dates 981109-1229 Cycle 19,970327- 981229. Rept Identifies Components Examined,Exam Methods Used,Exam Number & Summarized Results ML20205H5731999-03-29029 March 1999 Submits Required 1998 Actual & 1999 Projected Cash Flow Statements for Monticello Nuclear Generating Plant & PINGP, Units 1 & 2.Encl Contains Proprietary Info.Proprietary Info Withheld,Per 10CFR2.790(b)(1) ML20205C6561999-03-26026 March 1999 Submits Semiannual Update on Project Plans for USAR Review Project & Conversion to ITS ML20204H3371999-03-19019 March 1999 Forwards Application for Amend to Licenses DPR-42 & DPR-60, Removing Dates of Two NRC SERs & Correcting Date of One SER Listed in Section 2.C.4, Fire Protection 1999-09-07
[Table view] Category:VENDOR/MANUFACTURER TO NRC
MONTHYEARML20151K6201988-03-25025 March 1988 Requests That Proprietary WCAP-11787, Prairie Island Units 1 & 2 Evaluation for Tube Vibration Induced Fatigue, Be Withheld from Public Disclosure (Ref 10CFR2.790(b)(1)) ML20235Y4451987-09-14014 September 1987 Authorizes Withholding of Proprietary WCAP-11525, Probabilistic Evaluation of Reduction in Turbine Valve Test Frequency, Per Encl Affidavit AW-80-027 (Ref 10CFR2.790) ML20215G4951987-04-0707 April 1987 Requests Withholding of Vol 2 to WCAP-10924, Westinghouse Large-Break LOCA Best Estimate Methodology:Application to Two-Loop PWRs Equipped W/Upper Plenum..., Per 10CFR9.5(4). Rept Will Be Separately Submitted Under 10CFR2.790 NRC-87-3216, Forwards Proprietary Vol 2 to WCAP-10924, Westinghouse Large-Break LOCA Best-Estimate Methodology:Application to Two-Loop PWRs Equipped W/Upper Plenum Injection, to Support Early 1988 Facility Startup.Rept Withheld (Ref 10CFR9.5(4))1987-04-0707 April 1987 Forwards Proprietary Vol 2 to WCAP-10924, Westinghouse Large-Break LOCA Best-Estimate Methodology:Application to Two-Loop PWRs Equipped W/Upper Plenum Injection, to Support Early 1988 Facility Startup.Rept Withheld (Ref 10CFR9.5(4)) ML20215F5271986-09-10010 September 1986 Requests Withholding of Proprietary Slide Presentation Entitled, Nrc/Northern States Power/Westinghouse Joint Technical Exchange:Shape Memory Alloy Sleeving (Ref 10CFR2.790).Affidavit Encl ML20214P8301986-08-0505 August 1986 Requests Rev 1 to WCAP 10929, Technical Bases for Eliminating Large Primary Loop Pipe Rupture as Structural Design Basis... & Rev 1 to WCAP 10931, Toughness Criteria..., Be Withheld (Ref 10CFR2.790) ML20154L4771986-02-25025 February 1986 Requests That Proprietary Prairie Island Unit 1 Reactor Vessel Upper Internals Replacement Safety Evaluation Be Withheld (Ref 10CFR2.790).Affidavit Encl ML20136H9031985-11-14014 November 1985 Requests Withholding Proprietary Info Re Best Estimate LOCA Methodology as Applied to PWRs W/Upper Plenum Injection from Public Disclosure Per 10CFR2.790.Supporting Affidavit Encl ML20198F3561985-09-27027 September 1985 Requests That Proprietary WCAP-10929, Technical Bases for Eliminating Large Primary Loop Pipe Rupture... & WCAP-10931, Toughness Criteria... Be Withheld from Public Disclosure Per 10CFR2.790 ML20126A2651985-03-0707 March 1985 Requests That Proprietary WCAP-10756, Prairie Island Units 1 & 2 Steam Generator Sleeving Rept (Mechanical Sleeves), Be Withheld from Public Disclosure (Ref 10CFR2.790). Affidavit Encl ML20106J5371984-09-18018 September 1984 Requests Proprietary WCAP-10639, Technical Bases for Eliminating Large Primary Loop Pipe Rupture as Structural Design Basis for Prairie Island Unit 1, Be Withheld Per 10CFR2.790.Affidavit Encl ML20072L4311983-03-17017 March 1983 Authorizes Utilization of Encl Affidavit AW-77-47 in Support of Northern States Power Request to Withhold Proprietary Info Per 10CFR2.790 ML20070C9801982-12-0909 December 1982 Forwards Summary of NRC Review of Exxon Nuclear Co Thermal Hydraulic Methodology for Mixed Cores in Pwrs.Encl Toprod Sensitivity & Thermal Margin Studies Submitted on 820618 & 29,respectively,remain Valid.Comments Requested ML1116511951980-04-11011 April 1980 Forwards Proprietary & Nonproprietary Versions of Structural Analysis of RCS for Postulated LOCA, Preliminary Issue.Applications for Withholding Encl. Proprietary Version Withheld (Ref 10CFR2.790) ML19262B5721979-12-19019 December 1979 Forwards Westinghouse Evaluation of 791002 Tube Rupture Event.Concurs W/Util Recommendations in LER 79-027/01T-0 Re Operation of Turbine Driven Auxiliary Feed Pump & Steam Supply Valve from Faulted Steam Generator ML19263C4091979-02-12012 February 1979 Forwards Rept XN-NF-78-44,Generic Analysis of Control Rod Ejection Transient for Pwrs. 1988-03-25
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, a E(DN NUCLEAR COMPANY,Inc.
11$1#ws/lquidrAusf
- r. a au f.at acum< wadaptemmt Phone:(509)375-8100 Teler: 15 2878 December 9, 1982 GCC:048:82 Mr. J. J. Holonich Core Performance Branch Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555
SUBJECT:
ENC Thermal-Hydraulic Analyses for PWR Mixed Cores
Dear Mr. Holonich:
With your review of ENC's thermal-hydraulic methodology for mixed cores, including consideration of fuel with different rod sizes nearing completion, we thought it might be helpful to summarize the review as well as related analyses and developments from our perspective. This summary is provided in Attachment A.
As a result of your review, we found that automating the interface between core and subchannel analyses provides significant implementation advantages over our previous way of doing such analyses and this approach will be used in future work by ENC. As a result of our recent work with the automated interface in XCOBRA-IIIC, we believe our prior analyses to be accurate though they were cumbersome to perform. Thus, we believe our bounding analysis for ENC TOPROD fuel in Prairie Island Units 1 and 2 that was provided in submittals of June 18, 1982 to Gene Hsii and June 29, 1982 to Joe Holonich continue to be valid and thus supports future cycles of ENC TOPROD fuel in Prairie Island Units 1 and 2. For reference, I have provided these submittals again here as Attachments B and C.
~
We will be sending a copy of this letter and attachments to Northern States Power with a recommendation that it be used by them in supporting the validity of the current analysis. We would appreciate comments you may have regarding this matter.
Sincerely, G. C. Cooke, Manager Plant Transient Analysis GCC:GF0:gf l Attachments As noted l
8212140411 821209 goaaoocxosooom . . . _ . . . . .
f
ATTACHMENT A_
SUWiARY OF THE NRC'S REVIEW 0F ENC'S THERMAL HYDRAULIC METHODOLOGY FOR MIXED CORES IN PWRs The present detailed NRC review of EhC's thermal margin methodology for PWRs started with the review for ENC TOPROD reload fuel in Prairie Island Unit 1. The ENC calculations were performed with ENC's XCOBRA-IIIC code using a manual interface between core wide flow distribution analyses and limiting assembly MDNBR subchannel analyses. In the case of Prairie Island Unit 1, the original overpower MDNBRs being calculated with the W-3 correlation were on the order of 1.8 to 2.0. Reactor conditions were such that strong interassembly crossflow did not occur, and the use of an axial average flow factor from the core-wide flow distribution analysis was found to be adequate to establish limiting assembly flow in the MDNBR subchannel cnalysis.
Upon review of the TOPROD work, the NRC was concerned as to the general validity of the approach and was further concerned that crossflow at the interassembly boundary may impact MDNBR especially when the neighboring assembly has a different rod size from the limiting assembly. In response, ENC performed additional sensitivity studies. These calculations involved single and double octant subchannel models and were performed at much higher power conditions in order to produce MDNBRs near the 95:95 limit for the W-3 correlation. The purpose of going to a higher power beyond that applicable to Prairie Island was to examine the general validity issue.
In addition to going to higher power in the sensitivity studies, different axial power profiles were considered. At the higher power level
interassembly crossficw was much larger than in the original Prairie Island core thermal hydraulic calculations. For these new conditions ENC found it necessary to tighten the convergence parameter in the XCOBRA-IIIC calcula-tions and also found it necessary to use more care in selecting the limiting assembly inlet mass velocity in the MDNBR subchannel calculations. With respect to the latter item, selecting the inlet mass velocity, it was necessary to go beyond simply an axial average flow factor and to choose an inlet n. ass velocity which balanced conservatisms in enthalpy rise against non-conservatisms in assembly average mass velocity at the MDNBR elevation.
In view of the close agreement between single and double octant MDNBR results, it was concluded that the XCOBRA-IIIC-based core thermal hydraulics methodology with a manual interface between core and subchannel calculations (and without explicit account of interassembly crossflow) does provide an accurate calculation (approximately 1.0%) of MDNBR. Further, it was concluded that the single octant approach was generally applicable since in the case of non-peripheral MDNBR locations, it agreed closely with the double octant model and since it will be bounding for peripheral MDNBR locations. A less conservative but more accurate MDNBR calculation for peripheral MDNBR locations would require a double octant calculation.
Aside from the complication of choosing between single or double octant MDNBR subchannel models, the ENC methodology did have one major drawback.
Specifically, this was the fact that the analyst must have specialized training and experience to correctly establish the inlet mass velocity for the MDNBR subchannel analysis. The balancing of conservatisms in enthalpy rise against non conservatisms in mass velocity was straightforward in principle, but did require additional XCOBRA-IIIC sensitivity analyses to establish the variation of MDNBR with enthalpy rise and with assembly average mass velocity.
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Furthermore, in the case of double octant calculations where the limiting assembly mass velocity varies with axial location, a trial and error process was required to establish the inlet mass velocity for the final MDNBR calculation.
In view of the potential for errors in application and the general cumbersomeness of the method, a new approach was taken by ENC. Specifically, an automated crossflow boundary condition feature was added to XCOBRA-IIIC.
This new approach did not change MDNBR results, but it did make it much easier and straightfoward to calculate MDNBR for any given core situation. General agreement has been found in audit calculations by the NRC between XCOBRA-IIIC analyses with automated crossflow boundary conditions and its own methods for core thermal hydraulics analysis.
To conclude, it is nrted that in its applications and in additional sensitivity studies for 0.C. Cook Unit 2 fuel geometries, ENC has found that the limiting assembly location in the core is not an important factor with respect to MDNBR. Secondly, it has been found that radial averaging of crossflow boundary conditions around the boundaries of the limiting assembly and applying this averaged crossflow boundary condition to the MDNBR l subchannel channel yields an accurate calculation of MDNBR. These findings l
l have consistently been found to be the case since the review of ENC's PWR l
l thermal hydraulic methodology was begun for ENC TOPROD fuel at Prairie Island Unit 1. Finally, it is noted that none of the developments during this review of ENC's methodology has resulted in any significant change to the very considerable MDNBR margin to limits in ENC's analyses for Prairie Island Unit l 1.
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4 To : Gene Hsii, June 18, 1982. (informal)
ATTACHMENT B PRAIRIE ISLAND TOPROD SENSITIVITY STUDY Thermal margin results for the DNB limiting two pump coastdown event at Prairie Island Units 1 and 2 are reported in Appendix A.13, Table A.4, of the '
Exxon Nuclear Company TOPR00 thermal margin analysis document, XN-NF #-
56(P), Supplement 1. The results reported for a full TOPROD core are an MDNBR of 1.85 before accounting for rod bow and an MDNBR of 1.60 af ter a conservative account of rod bow as discussed in Appendix A.13 of XN-NF-80-56(P) Supplement
- 1. These results for a full core of TOPROD fuel have been confirmed as bounding of mixed cores of ENC TOPROD and ENC standard fuel in Prairie Island Units 1 and 2.
When calculating thermal margins, ENC methodology involves a core flow
' distribution analysis and an MDNBR subchannel analysis both of which are performed with ENC's XCOBRA-IIIC computer code. Crossflow boundary condi-tions are indirectly accounted for in the MDNBR subchannel analysis by selecting an input mass velocity that best matches the limiting assembly flow results from the core flow analysis. This is done by balancing conservatisms in enthalpy rise against nonconservatisms in mass velocity at the calculated MDNBR elevation in the MDNBR subchannel analysis.
In view of a concern that ENC's indirect approach to handling crossflow may not be accurate, a sensitivity study has been performed to bound the maximum potential error in the TOPROD analysis. Specifically, the impact to MDNBR of using the minimum mass velocity from the core flow analysis versus an average mass velocity which balances conservatisms in enthalpy rise against non conservatisms in flow, was evaluated. This evaluation was made for the limiting conditions at the time of MDNBR in the Prairie Island TOPROD two-pump coastdown transient event.
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The minimum flow over the active region of the fuel was found to be 3%
less than the average flow that balances enthalpy rise error against flow error at the NDNBR elevation (0.95 flow factor versus 0.98 flow factor). The associated reduction in MDNBR was evaluated to be 1.5%. Conservatisms(item d in Appendix A.10 of XN-NF-80-56(P), Supplement 1) in the initial MDNBR of 2.24 osed in the Prairie Island TOPROD transient analysis bound this 1.5%
reduction in MDNBR.
i e
l-4 0
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To : J:e Holonich,. June 29, 1982 (informal)
ATTACHMENT C PRAIRIE ISLANL - THERMAL MARGIN As supplemental information to ENC's most recent evaluation of Prairie Island TOPROD therma'l margin (.0une 18, 1982, transmittal to Gene Hsii),
please note the following:
- 1) The calculated results for MDN3R at the time of limiting conditions during the two-pump coastdown transient are as follows:
MDNBR = 1.929 in transient XCOBRA-IllC core flow and MONBR subchannel calculations MDNBR = 1.892 in a quasi-steady core flow and MDNBR subchannel calculations.
These MDNBR results both reflect use of an axial minimum flow factor 1 of 0.95 in the single octant M3NBR subchannel calculations.
- 2) Examination of the core flow results shows that a 0.98 flow factor would balance conservatisms in enthalpy rise against non conservatisms in mass velocity at the MDNBR elevation. Use of the 0.98 flow factor in the quasi-steady MDNBR evaluation gave an MDNBR result of 1.919 (versus 1.892 for the 0.95 flow factor). Thus use of'a 0.95 flow factor yield a 1.5% reduction in MDNBR.
- 3) The MDNBR results cited above all pertain to a full core of TOPROD fuel at' Prairie Island Units I or 2. This case has been reconfirmed as bounding of mixed cores of TOPROD and ENC standard fuel. The radial
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. power distribution used in the core flow analysis closely corresponds to and yields equivalent results to the Cycle 9 (2/3 TOPR00, 1/3 standard) distribution given in Figure A.1 of XN-NF-81-56(P), Supplement 1.
Use, reproduction. transtmttel or dec60sure of the above enformata is subsect to the restraction on the fwSt or title page of this document.
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