Letter Sequence Acceptance Review |
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Results
Other: A02000, Forwards Addl Info Requested in NRC Re Transient & Accident Analyses in Basic Safety Rept, A02204, Provides Addl Info Requested by NRC to Facilitate Review of Cycle 5 Operation, A03722, Forwards Addl Info Re Effects of Loss of Offsite Power on Steam Generator Tube Rupture & Main Steam Line Break Analyses, B10325, Forwards Reload Safety Analysis Submitted in Support of Cycle 5 Reload.Rept Includes Results of Boron Dilution,Loss of Reactor Coolant Flow,Seized Rotor Reanalysis & Control Element Assembly Ejection & Withdrawal at Critical, B10960, Discusses Steam Generator Tube Rupture Reassessment for Cycle 6 Operation.Radiological Doses Will Be Increased as Result of Lower Reseat Pressure.Releases Expected to Be Significantly Below 10CFR100 Criteria, B11003, Informs That Util QA-verified 831212 Reanalysis of Steam Generator Tube Rupture Event.Reported Results Remain Valid. Verification Completed Prior to Startup for Cycle 6, B11098, Provides Summary of Radiological Results Re Steam Generator Tube Rupture Reanalysis.Table Containing Doses for All Analyzed Cases Encl, ML20033A519, ML20039C038, ML20087K388, ML20116E663
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MONTHYEARA02000, Forwards Addl Info Requested in NRC Re Transient & Accident Analyses in Basic Safety Rept1981-10-27027 October 1981 Forwards Addl Info Requested in NRC Re Transient & Accident Analyses in Basic Safety Rept Project stage: Other B10325, Forwards Reload Safety Analysis Submitted in Support of Cycle 5 Reload.Rept Includes Results of Boron Dilution,Loss of Reactor Coolant Flow,Seized Rotor Reanalysis & Control Element Assembly Ejection & Withdrawal at Critical1981-11-17017 November 1981 Forwards Reload Safety Analysis Submitted in Support of Cycle 5 Reload.Rept Includes Results of Boron Dilution,Loss of Reactor Coolant Flow,Seized Rotor Reanalysis & Control Element Assembly Ejection & Withdrawal at Critical Project stage: Other ML20033A5191981-11-17017 November 1981 Cycle 5 Refueling Reload Safety Analysis Project stage: Other B10353, Application to Amend License DPR-65 Incorporating Revisions to Tech Specs to Support Cycle 5.Completion of Revisions & Reanalysis Required for Item II.K.3.30 of NUREG-0737 Scheduled for 8207011981-12-17017 December 1981 Application to Amend License DPR-65 Incorporating Revisions to Tech Specs to Support Cycle 5.Completion of Revisions & Reanalysis Required for Item II.K.3.30 of NUREG-0737 Scheduled for 820701 Project stage: Request ML20039C0381981-12-17017 December 1981 Revised Tech Specs Re Cycle 5 Reload Analyses Project stage: Other ML20040E2041982-01-12012 January 1982 Forwards Safety Evaluation of Westinghouse Basic Safety Rept.Transient Analysis for Startup of Inactive Reactor Coolant Pump,Excess Load & Loss of Normal Feedwater Acceptable Project stage: Approval A02204, Provides Addl Info Requested by NRC to Facilitate Review of Cycle 5 Operation1982-02-0404 February 1982 Provides Addl Info Requested by NRC to Facilitate Review of Cycle 5 Operation Project stage: Other B10960, Discusses Steam Generator Tube Rupture Reassessment for Cycle 6 Operation.Radiological Doses Will Be Increased as Result of Lower Reseat Pressure.Releases Expected to Be Significantly Below 10CFR100 Criteria1983-12-12012 December 1983 Discusses Steam Generator Tube Rupture Reassessment for Cycle 6 Operation.Radiological Doses Will Be Increased as Result of Lower Reseat Pressure.Releases Expected to Be Significantly Below 10CFR100 Criteria Project stage: Other B11003, Informs That Util QA-verified 831212 Reanalysis of Steam Generator Tube Rupture Event.Reported Results Remain Valid. Verification Completed Prior to Startup for Cycle 61984-01-17017 January 1984 Informs That Util QA-verified 831212 Reanalysis of Steam Generator Tube Rupture Event.Reported Results Remain Valid. Verification Completed Prior to Startup for Cycle 6 Project stage: Other A03635, Forwards Info & Comments in Response to 831114 Request for Various Confirmatory Evaluations & Justification for Analytical Models Utilized in Cycle 6 Safety Analyses1984-02-0101 February 1984 Forwards Info & Comments in Response to 831114 Request for Various Confirmatory Evaluations & Justification for Analytical Models Utilized in Cycle 6 Safety Analyses Project stage: Request ML20087K3881984-03-13013 March 1984 Submits Details of Followup Actions on Amend 90 to License DPR-65.Adequate Justification Provided to Demonstrate Applicability of Current Large Break LOCA Evaluation for Operation W/O Thermal Shield Project stage: Other B11098, Provides Summary of Radiological Results Re Steam Generator Tube Rupture Reanalysis.Table Containing Doses for All Analyzed Cases Encl1984-04-13013 April 1984 Provides Summary of Radiological Results Re Steam Generator Tube Rupture Reanalysis.Table Containing Doses for All Analyzed Cases Encl Project stage: Other ML20098B7771984-09-14014 September 1984 Forwards Info to Resolve Concerns Identified During Feb 1984 Meeting Re Large Break Loss of Coolant,Steam Line Break & Steam Generator Tube Rupture Accident Evaluations Submitted to Support Cycle 6 Operation,Per Amend 90 to License DPR-65 Project stage: Meeting ML20112B5431985-01-0909 January 1985 Responds to NRC 831230 Request for Addl Info.Assumption of Complete Mixing Not Concern from DNB Standpoint for Facility Since No Boiling Occurred in Hot Channel for Steam Line Breaks W/Different Mixing Factors Project stage: Request ML20116E6631985-04-22022 April 1985 Discusses Followup Actions to SER for Amend 90 to License DPR-65.New Steam Line Break Generator Tube Rupture Analysis Will Be Submitted for Review 4 Months After Startup from Current Refueling Outage Project stage: Other A03722, Forwards Addl Info Re Effects of Loss of Offsite Power on Steam Generator Tube Rupture & Main Steam Line Break Analyses1985-11-0808 November 1985 Forwards Addl Info Re Effects of Loss of Offsite Power on Steam Generator Tube Rupture & Main Steam Line Break Analyses Project stage: Other ML20212M5601987-01-21021 January 1987 Advises That 840914,850102 & 1108 Supplemental Cycle 6 Analyses Sufficient for Satisfying NRC Request for Confirmatory Calculations.Acceptance of Retran Model Should Not Be Construed as Acceptance for Future Licensing Project stage: Acceptance Review 1984-02-01
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January 21, 1987 Distribution Docket No. 50-336
' Docket File ACRS (10)
NRC/ Local POR PKreutzer PBD-8 Reading DJaffe FMiraglia OGC-Beth Mr. Edward J. Mroczka, Senior Vice President EJordan NThompson Nuclear Engineering and Operations BGrimes JPartlow Northeast Nuclear Energy Company Gray File 3.5a P. O. Box 270 Hartford, Connecticut 06141-0270
Dear Mr. Mroczka:
By [[letter::B10987, Forwards ASME Section Iii,Class 2 & 3 & ANSI B31.1.0 Piping Analysis, Ultrasonic Indications in Steam Generator Top Head Dome Weld. Evaluation Verified Structural Integrity of Plant Steam Generators,Per NRC|letter dated December 30, 1983]], the NRC staff forwarded Amendment No. 90 to Facility Operating License No. DPR-65.
The amendment authorized Cycle 6 operation for Millstone Unit No. 2.
The NRC Safety Evaluation accompanying the amendment addressed various aspects of the evaluations submitted to support Cycle 6 operation.
The staff documented several concerns relating to the analysis of steam line breaks and steam generator tube ruptures and requested that these concerns be addressed by confirmatory calculations.
The staff's concerns included the possible effect of a loss of of fsite power coincident with these analyzed events and the mixing factors considered in the steam line break analysis.
The staff has reviewed the supplemental Cycle 6 analyses provided in letters dated September 14, 1984, January 2, and November 8, 1985, and has concluded that these analyses are sufficient for satisfying our request for confirmatory calculations. We note, however, that the steam generator tube rupture was performed using a RETRAN model for Millstone Unit No. 2.
While we find that this model was adequate for responding to our request, our acceptance of this calculation should not be construed as an acceptance of the Millstone Unit No. 2 RETRAN model for future licensing analyses.
Should NNECo decide to utilize the RETRAN code for future licensing analyses, you should submit the Millstone Unit No. 2 RETRAN model for staff review and approval at that time.
Based upon the above, we consider the issue of supplemental Cycle 6 analyses to be closed.
I l
Sincerely,
/s/
D. H. Jaffe, Project Manager PWR Project Directorate #8 Division of PWR Licensing-B cc:
See next page 0701300161 070121 PDH ADOCK 05000336
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Mr. Edward J. Mroczka Millstone Nuclear Power Station Northeast Nuclear Energy Company Unit No. 2 cc:
Gerald Garfield, Esq.
Mr. Stephen E. Scace Day, Berry & Howard Superintendent Counselors at Law Millstone Nuclear Power Station City Place P. O. Box 128 Hartford, Connecticut 06103-3499 Waterford, Connecticut 06385 Regional Administrator, Region I Mr. Wayne D. Romberg U.S. Nuclear Regulatory Commission Vice President, Nuclear Operations Office of Executive Director for Northeast Nuclear Energy Company Operations P. O. Box 270 631 Park Avenue Hartford, Connecticut 06141-0270 King of Prussia, Pennsylvania 19406 Mr. Charles Brinkman, Manager Washington Nuclear Operations C-E Power Systems Combustion Ergineering, Inc.
7910 Woodmont Avenue Bethesda, Maryland 20814 Mr. Lawrence Bettencourt, First Selectman Town of Waterford Hall of Records - 200 Boston Post Road Waterford, Connecticut 06385 Northeast Utilities Service Company ATTN:
Mr. Richard M. Kacich, Manager Generation Facilities Licensing Post Office Box 270 Hartford, Connecticut 06141-0270 Kevin McCarthy, Director Radiation Control Unit Department of Environmental Protection State Office Building Hartford, Connecticut 06106 Mr. Theodore Rebelowski U.S. NRC P. O. Box 615 Waterford, Connecticut 06385-0615 Office of Policy & Management ATTN:
Under Secretary Energy Division 80 Washington Street Hartford, Connecticut 06106
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