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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
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Northem States Power Company 414 N:co!Iet Mah Minneapolis, Minnesota 55401 1t '
Telephone (612) 330 5500 March 3. 1992 U S Nutlear Regulatory Commission Attn: Document Control Desk Vashington, DC 20555 PRAIRIE IS1ANL NUCLEAR CENERATING PLANT Docket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 Small Break LOCA Analysia. ,
We have conpleted a new Small Break Loss of Coolant Accident (LOCA) analysis.
This analysis is being subinitted per your request in Inspection Report 90015 and 90016 (dated October ?6 1990). The results of the analysis are document in Attachment 1. The andv=is is documented in the form of a draf t Updated Safety Analysis Report section (this report section is a drsft but the analysis results are final), This attachment will be incorporated into the next revision to the Updated Safety Analysis Report (June, 1992).
This accident was reanalyzed to correct an input error in auxiliary feedwater flow and several analysis concerns. The new peak cladding temperature is 107' F.
Please contact us it you have questio..s concerning these reports.
kllb%&
Themas M Parker Manager Nucleat Support Services c: Regional Administrator Region III, NRC Senior Resident Inspector, NRC NRR Froject Manager, NRC J E S11 berg -
Attachments:
- 1. Draft Updated Safety Analysis Report Section 14.7 0 G O D ? .>
92O3090191 920303 PDR ADOCK 0D000282 q(
30 t p PDR )f
l Prairie Island Units 1 and 2 Updated Safety Analysis Report 1
1 14.7 Lms of ReacInr Coolant.ftom SmdLEnplumU'ipes or Fr'Enfracks in Lnrag_Einn which Amrate the Emenency Core Coollr&Sntttu 14.7-1 Acceptance Cntena A minor pipe break (small break), as considered in this section, is defined as m rupture of the reattor coolant pressure boundag with a total cross-sectional area less than 1.0 ft in which the nonnally operating charging system flow is not sufficient to sustain pressurizet level rJ pressure.
This is considered a Condition III event, an infrequent fault. ,
The Acceptance Criteria for the loss-of-coolant accident is described in 10 CFR 50.46 as follows:
(a) The calculated maximum fuel element cladding temperature shall not exceed 2200 F.
(b) The calculated total oxidation'of the cladding shall nowhere exceed 0.17 times the total chdding thickness before oxidation.
(c) The calculated t Ja! an.ount of hydrogen generated from the chemical reaction of the
'hdding with water or steam shall not exceed 0.01 times the hypothetical amount that sould be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.
(d) Calculated changes in core geometry shall be such that the core remains amenable to cooling.
(e) After any calculated sucessful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long hved radioactivity remaining in the core.
These criteda were established to provide ignificant margin in ECCS performance following a LOCA.
14.7.2 pfelqdption of Sr"gil Break LOCA Transient
~
Ruptures of small cross-section win cause loss of the coolant at a rate which can be accomodated by the charging pumps. These pumps would maintain an operational water level in the pressurizer permitting the operator to execute an orderly shutdown. The coolant which would be released to the ,
. containment contains the fission products existing at equillboui.
The maximum break size for which the rormal makeup system can maintain the pressurizer level is obtained by comparing the calculated flow from the Reactor Coolant System through the postulated break against the charging pump makeup flow at normal Reactor Coolant System pressure; i.e.,
2250 psia. A makeup flow mte fmm one charging pump is typically adequate to sustain pressurizer level at 2~250 psia for a break through a 0.375 inch diameter hole. This break results in the loss of approximately 17.25 lbm/sec.
__ , . . . . __.m..-._.... _ _ . . _ . _ _ . - - _ _ . _ . _ . . _ . . . _ _ _ _ .
Prairie Island Units 1 and 2 I Updated Safety Analysis Report i
Should a larger break occur, depressurization of the Reactor Coolant System causes Guid into the loops from the pressurizer resulting ir. a pressure and level decrease in the pressurizer. Reactor trip occurs when the low pressurizer pressure trip setpoint is reached. During the early part of the small break transient, the effect of the b~ak flow is not strong enough to overcome the Gow maintained by the reactor coolant pumps through the core as they are coasting down following reactor trip.
Therefore, upward now through the core is maintained. The Safety injection System is actuated when the appropriate setpoint is reached. The consequences of the accident are limited in two ways:
- 1. Control rod insertion and void foanation in'the core cause a rapid reduction of the nuclear power to a residual level corresponding to the delayed tission and fission product decay.
- 2. Iniection of borated water ensures sufficient flooding of the core to prevent excessive clad tempemtures.
Before the break occurt e plant is in an equilibrium condition; i.e., the heat generated in the core is being removed via the acondary system. During blowdown, heat from fission product decay, hat internals, and the vessel continues to be transferred to the Reactor Coolant System. The heat transfer between the Reactor Coolant System and the secondary system may be in either direction depending on the relative temperate.es. In the case of continued heat addition to the secondary, system pressure increases and steam dump may occur. Makeup to the secondary side is automatically provided by the auxiliary feedwater pumps. The safety injection signal stops normal feedwater now by closing the main feedwater isolation valves and initiates auxiliary feedwater flow by starting auxiliary feedwater pumps. The secondary flow aids in the reduction of Reactor Coolant System pressures.
%%n the RCS deptessurizes to the accumulator cover gas pressure, the cold leg acc mulators begin to inject water into the reactor coolant loops. Due to the loss of offsite power usumption, the ractor coolant pumps are asssumed to be tripped at the time of reactor trip during the accident and the effects of pump coastdown am included in the blowdown analyses.
14.7.3 Small Break LOCA Evajuation Model The NOTRUMP computer code is used in the analysis of loss-of-coolant accidents due to s;nall breaks in the reactor coolant system. The NOTRUMP computer code is a state-of-the-art one-dimensional general network code ;onsisting of a number of advanced features. Among these features are the calculation of thermal non-equilibrium in all Guid volumes, flow regime-dependent
~
- drift Gux calculations with counter-current flooding limitations, mixture level tracking logic in multiple-stacked fluid nodes, and regime-dependent heat transfer correlations. . The NOTRUMP small break, LOCA emergency core cooling system (ECCS) evaluation model was developed to n determine the RCS response to design basis scall break LOCAs and to address the NRC concerns expressed in NUREG-0511, " Generic Evaluation of Feedwater Transients and Small Break Loss-of Coolant Accidents in Westinghouse-Designed Operating Plants."
4 t
.-.-mv ., , - = . , < . . ~ . - , ,. . -.,-s.r .- , - . ,---4 _ _ . _.-_ , __ _ m. _ m ~_ m -- -+
_ - _ . - . - - - - - . - _ . . ,. -- - ~. . _-
Prairio Island Units 1 and 2 Updated Safety Analysis Report i
In NOTRUMP, the RCS is nodalized into volumes interconnected by flowpaths. The broken loop is modeled explicitly with the intact loops lumped in;o a second loop. The transieni behavior of the ;
system is determined from the governing consenation equations of mass, energy, and momentum.
applied throughout the system. A detailed description of NOTRUMP is given in WCAP-10054-P-A and WCAP 10079-P-A.
The use of NOTRUMP in the analysis involves, among other things, the representation of the reactor core as heated control volumes with an associated bubble rise model to permit a transient mixture height calculation. The multinode capability of the prognm enables an explicitly and detailed spatial representation of various system components. In particular, it enables 1 proper calculation 'of the behavior of the loop seal during a loss-of-coolant transient.
Citdding themial analyses are performed with the LOCTA-IV WCAP-8301-P code which uses the RCS pressure, fuel rod power history, stea:n flow past the uncovered part of the core, and mixture height history from the NOTRUMP hydraulic calculations, as input.
A schematic representation of the computer code interfres is given in Figure 14.71.
This model was developed to resolve TMI Action Item II.K.3.30. NRC acceptance of this model '
for Prairie Island was documented in an NRC staff letter dated June 6,1985.
14.7 Small Break Inout Parameters and Initial Conditions Table 14.7-1 lists important input parameters and initial conditions used in the small break analysis.
The axial power distribution and core decay power assumed for the small break analysis are shown in Figures 14.7-2 and 14.7-3.
Safety injection flow rate to the Reactor Coolant System is a function of the system pressure is used as part of the input. ' The Safety Injection System (SI) was assumed to be delivering to the RCS l 25 seconds after the generation of a safety injection signal.
For this analysis, the SI delivery considers injection fic~ which is depicted in Figure 14.7-4 as a function of RCS pressure. This figure represents injection flow from one degmded SI pump spilling
~
to either RCS pressure (if break size is smaller than SI injection line diameter), or to O psig containn"nt pressure (if break size is greater than or eaual to tise SI injection line diameter). The 25 second delay includes time requimd for diesel startup and loading of the safety injection pumps
-onto the emergency buses. Fbw from the RHR pumps does not affect the analysis since their shutoff head is lower that RB pressure during the time portion of the (nuisient co 1sidered here.
Also, minimum safeguards E aergency Core Cooling System capability and opembility has been asssumed in this analysis.
The hydmulic analyses are performed with the NCTRUMP code using 102% of the licensed core
. power. The core thermal transient analyses are performed with the LOCTA-IV code using 102% of the licensed core power.
. . ..- _ . , . - . ~.. . . _.---- . . . - . . . - . . .. - . - - ~ - .. -- .
- I Prairie Island Units 1 and 2 Updated Safety Analysis Report 14.7 5 Small Break Resulu As noted previously, the calculated peak cladding temperature resulting from a small break LOCA is less than that calculated for a large break LOCA. A range of small break analyses are presented which establishes the limiting break size. The results of these analyses are summarized in Tables 14.7-2 and 14.7-3. Figures 14.7-5 throuc;h 14.7-11 present the principal parameters of interest for the small break ECCS analyses. For the cases analyzed, the following transient parame:ers are included except hot spot clad temperature when the core is not uncovered. .
- a. RCS Pressure
- b. Core mixture height ,
- e. Hot spot clad tempemture i For the limitng break analyzed (6 inch), the following additional transient pararneters are present (Figures 14.7-12 through 14.7-14):
- a. Core steam flow rate
- b. Core heat transfer coefficient
- c. Hot spot fluid temperature-
' The maximum peak clad temperature for the breaks analyzed is 1077'F. These results are well below all Acceptance Criteria limits of 10 CFR 50.46, and in no case is limiting when compared to the results presented for larne breaks.
i J
1 w w-- --o y s v -w,a- w+,-,m,. -.4---i.-- #- . ~ , . _ - - - ~ w .- . - . - . .,+v -_r2 . . - -.- - - -,e a - -
+e w-
._ 4-_ __.,___._ _ _ . _ . . _ _ _ . _ . _ . . . . ___ . _ - -
Prairie Island Units 'i and 2 l Updated Safety Analysis Report ]
n ,
l TABLE 14.7-1 t
INPUT PARAMETERS USED IN Tim SMALL BREAK LOCA ANALYSIS Parametel - IDPJll Core Power 107, of 1650 MWt Peak Linear Power (kW/ft) 15.096 kW/ft (includes 102% factor)
- Total Peaking Factor 2.50 Power Shape ~ See Figure 14.7-2 Fuel Assembly Array 14X14 OFA Accumula' tor Conditions:
Cover Gas Pressure 710 psig Water Volume 1250 ft3 Total Volume 2000 ft 3 Pumped Safety Injection Flow See Figure 14.7-4 Steam Generator Initial Pressure 664 psia Stern Generator Tube Plugging Level 10 %
- Reactor Trip Signal 1700 psia .
- Safety Injaction Signal 1700 psia Rod Drop Time - 2.4 seccada
~
^ 2.0 seconds Reactor Trip Signal Delay Time l
l-I.
r 1
- .. . . _ . . __,_ .m..._.-._. . .-
Prairie Island Units 1 and ?
Updated Safety Analysis Report TABLE 14.7-2 a
SMALL BREAK LOCA TIME SEQUE'4CE OF EVENTS Break Size EVENT 4.0Inen. 6.0 Inch 8.0 Inch Break Initiation, sec. 0.0 0.0 0.0 Reactor Trip Signal, sec. 8.0 5.6 5.2 Safety Injection Signal, sec. 8.0 5.6 5.2 I Top of Core Uncovered, sec. ~'60 - 140 - 78 Accumulator Injection Begins, sec. ~ 350 ~ 130 ~ 80 Peak Clad Temperature Occurs, sec.
~178 ~192 ~ 116 Top of Core Recovered, sec. ~180 -195 - 122 TABLE 14.7-3 SMALL BREAK LOCA ANALYSIS RESULTS Bmak Size RESULT ,
4.0 Inch l 6.0 Inch - l 8.0 Inch Peak Clad Temperature, *F 834 1077 1053 Peak Clad Temperature Location, ft. 10.5 10.75 10.5
~
Local ZrH O 2 Reaction (max), % 0.0333 0.0339 0.0337 Local Zr/H 2O Reaction Location, ft. 10.5 10.75 10.5 Hot Rod Burst Time, sxonds NA NA NA Hot Rod Burst Location, ft. NA NA NA
]
N
PRAIKIE ISLA O 1
CCPE PRESSUFF,CCRE N FLOW,M1XTURE LEVEL 0 AND FUEL RCD POWER L T HISTORY O '
R C U i M O< TIME < CORE COVERED A P
If
=
f Figure 14.7-1 Code interface Descriptiori for Small Break Model 4
PRAIRIE ISLAND SBLOCA . ANALYSIS Small Break LOCA Power Shape 3
l 2,5 <. . ~ -
~ ,
p- % ~ ,
2 -
p V f 3- ,
o 1.5 a - -
- n A i.1 1
f f - -* \
3 1
0.5 I- --
0 -
0 2 4 6 8 10 12
, Height (ft)
SELOCA Shape FQ Umit
__ c SELOCA POWE)v SHAPE ,
PRAIRIE ISLAND UNITS 1 AND 2 FIGURE 14.7 2
_=
i' e
i
t
', PRAIRIE ISLerD e
l 1
l ioC _
-~
7 . ..
~
2 " *OTAl. Anb!CU Ai. HEAT (WIT'H '% SHUTOO*N) l C:
j.'so-1 -
i[ 7 5'
- e : ~
N N
N l
l w
N q IC*2 r 5 ~; 2 3,
2 -
50-3 i _
ii i i s t !I
- ! I f Ii t!!I I I I I I t til i f ! . 1,.).,J t ti t i i i f f r i_
in f . 3 3 ICO2 5 106 2 -S 102 2 g to5 2 5 iC' TIME #TER SHUTCC*N (SECONCSI-1 f
I .
Figure 14.7-3 Care Power After 11eactor Trip (Applie to all Small areaks) 4
- ,a -c, , .. , ,, -
l l
i i
]
PRAIRIE ISLAND SBLOCA ANALYSIS HHSI Flov> - 1 Degraded Pump r 2,500 f%
2,000 'R '
NA.
R I A
- ? E
'E 1,500 T i-
- c. b-e 8
m b-y e 1,000 e i :- a.
i @ '
e k,
A s
, 500 - - - -
- r. N
- l. K >
j=
0
'54' -
O 10 20 30 40 50 injected Flow (Ibm /sec) l Spill to RCS Spill to O psig l
l l
l l=
L HHSI FLOW RATE PRAIRIE ISLAND IJNITS 1 AND 2 FIGURE 14.74 t
l
, , , , , ,,n , . , , , -
g
. __ _ _ _ __ _ . _ . .. . . _ . _ .- ~
2400.
l l l 2220. -
I i
2000. - - - -
l 2
1800.
m k ^
i u 1600. ' - - - - - -- --
l m
m W
tz 1400, a.. l l
u u
N g-- ,
g 1220, 1003.
\s N 020. - ' - - -
\
\
600.6 - --
sx N s
'd . 52. 100. 15E 200. 250. 300. 550. 400. 452. 523.
TIME (SEC) i
.:.0* SMAU. DR SAK LOC A PRAIRIEISLAND UNITS 1 AtlD 2 . - _ . .
FIGURE 14.7 5 f
~
l
-, -- ,<~ --r-- nw-,
l 4
5 3, -
l l
- i
'N l l
- 26. - ', - d- ! ,
)
\ s l, 1 kN I I i l
1 cu.'
3,,
Myg 1 -
l
\A a l' \ A r
h l 24 -
\ ,l \ . f~i .
1
/,a j ; i
~
! / ;
= 2c,.. --
i / ,
\
I '
L.,
I
.d(i '
/
s
, i l' j t aa ; -
ej ;
O I {
x j =
4 i '
E ; 9' W l t
1-3 L; t i
-d
- 16. --
] p iI l -~
- 4 , -;
l
~
i l ,
- 12. \ t' 1
I
?
I e
^ I) 50. 100. 152. 200. 250. 530. 550, 4 'J 0 . 4h0. s21 .
J T i t1E ISEC) t-4 4
y-4.0* SMALL BREAK LOCA 4 CORE MIXTURE LEVEL PRAIRIE ISLAND UNITS 1 AND 2 FIGURE 14.7 6 I
l m . -- ,- ---e -. , - -
4 2400.
2200, ---
, i i i I
2000. 1 i 2 1900.I I 1 G !
S 1600. I l l 0
a l
i g 1400. -
l N
c.
m 1200.
M k - %
u N' 8 1000, e O
=
800.
N \
600.-
3 400.
\ ~~~
w 200,JI 25. 50, 75. 100. 125. 150. 175. 200. 225, 250.
TIME ISEC1 t
J 4
6.0" SMALL BREAK LOCA PRAIRIE ISLAND UNITS 1 AND 2 I FIGl.'RE 14.7-7 l i.
1
'9 3 0. -
l .i
's f i i
i
\ ' l { l a9. 1
\ '
2 .
h.' ,1 l
~ l
'L /\ l O, ,
- 24
\ l ..
' l \-l 1 /
d \, j 'l
-~
Q 22. \
p.-
t I b
ci \
- ~ , ! t I!
x 40. ,
8 ,
E $ ,
t~ l i l c :e. ,
.\ l l
'6.
- 4,
/\l
'2 ._
' IJ . 25. SH. 75. 120. [25. .:J. 1/5. 242, 225. d ;0..
T I .ME ISEC1 N
6.0* SMALL BREAK LOCA cons m aamsL i PRAIRIE ISLAND UNITS 1 AND 2 t
FIGURE 14.7-8 l
l l
I l
4
'1:0.
1 4 f I ie 3 r,
I i
I 900.
\h E l l 300. .
w- l
c u u 700. ' -
B w I
~
i 600.
/
V 530. ,
\ NX l 400. -
120. 140. 160. 180. 200. 220. 240. 260.
TIME (S) 8.0* SMALL BREAK LOCA PRAIRlE ISLAND UNITS 1 AND 2 FIGURE 14.7 9 1
1 1
2400. ,
2200. i
.. l l 2000.
l l
1900. _L -
E l 1
G S 1600. ..-
d a
$ 1400.
d a
oc 1200.
N s
g 1000.
N -
W N l E
800. \'s ,
600. -
\ r-400. ---
s ,
i agg 2 .
D. 20. 40. 60. - - _ . . -80. 100. 120. 140, 160.
TIME (SEC) 8.0* CMALL BREAK LOCA PRAIRIE ISLAND UNITS 1 AND 2 FIGURE 14.710 l
1.
l i
l l
L
32.j i j --- --
I f
\ t 27,gl. + ' i j l l
- N ,r'N ,
i \. / \ l ! !
d b. .
'\ l l 1 i i/
]
i L 22,E '
1.
l yA
'~
l / I
\ i
- u. . _ _
\ ' _/
i \ l [
\ ,
w" l ;
";, 17 3 LJ I
)
f !
, % \ t )'
- r. 15.I--- \. i
/
I
\ l '
U 12.3 o . \
l i
\
i \
- 10. \
I i
\'
7.5 -- ----
- 5. - -+ - "
.! ',l I,-
I l p'r- a >
l' . 20. 40. 6d. 09 100. 1 2 k. . - 40. 1C2.
~
f(ME ($CC) s -
0.0* SMALL CREAK LOCA i PRAIRIE ISLAND UNITS 1 AND 2 FIGURE 14.7-11
^
300.. i 1 i
. l. '
fj j I i 3 C O . 7-y ,
i i !
l 7CD.L ,
l 3 1 I I i E 500.
ca ! l a i I
2 l 3 :<j0. {- i 1
s !
l 5 i i l 1 8 .ac.
- a -
I i
Y d t I 1 l 5 <co, .
)
l l'
' hi c
y
' I 200. '
Mpm . ; \'
100, r;-
l p!\{ 1
^>
-\
\,f
- y Nw I -
/
O ,' \+b IdC. .25. 150. 175. 200. ca.
- c. 2 5, . SJ. 15. a_e.
r!.ME ISEC1 6.O* SMALL BREAK LOCA L CORE ON VAPOR FLOW PRAIRIE ISLAND UNITS 1 AND 2
' FIGURE 14.7-12 l
I
( .
104, i
r- ~
i -t i i -
/ l t i i l > > s s
/ 1 i i i _
I l l l
~
l l Y
T i N t 10 - .
(_ , i M i i i /i 3 co '
i ! ! _/i \ /
~
l ! I I/~ \ /
s t )' I I Il \/
8 _
il _
L' A ,
' 3 E 102 '
l 3 '
. I ', l E .
I I l i M \ I
% N Id I I i
10 l' 200. 220. 240. 250.
120. 140. 160. 180.
TIME (S) 6.0* SMALL BREAK LOCA HOT SPOT RCD SURFACE HEAT TRANSFER COEFFICtENT PRAIRIE ISLAND UNITS 1 AND 2 ,
FIGURE 14.7-13 i
m . _ _ _ _
9 4
-'~
i 1
50:
I 75: -
I 700 g .
i
/ l '
u u 650 .
r 2
2 C
c.
soo
, =
a w
j' 500
.n N
i V
'O 220 2o 260 120 140 iso 180 200 TIME (S) 6.0* SMALL BR*J.AK LOCA HOT SPOT FLut0 PRAIRIE ISLAND UNITS 1 AND 2 TEMPERATURE FIGURE 14,7-14
.. - J