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Category:CORRESPONDENCE-LETTERS
MONTHYEARML20217K4391999-10-18018 October 1999 Provides Response to RAI to Support USI A-46 Program Submittal for North Anna Power Station,Units 1 & 2.Rev 10 to BNL Rept 52361,encl ML20217H3301999-10-14014 October 1999 Forwards Rev 0 to COLR for North Anna 2 Cycle 14 Pattern Su. No New Commitments Intended by Ltr ML18152B3541999-10-12012 October 1999 Requests Use of Code Case N-532 & N-619,per Provisions of 10CFR50.55a(a)(3).Detailed Info Supporting Request,Encl. Attachment 2 Includes Technical White Paper That Provides Further Technical Info ML20216K1681999-10-0101 October 1999 Forwards Vols I-VIII of Rev 35 to UFSAR for Naps.Rev Also Includes Update to Chapter 17 of Ufsar,Which Contains Operational QA Program.Changes to Program Description Do Not Reduce Commitments Contained Therein ML20212J9101999-10-0101 October 1999 Forwards SE Accepting Licensee 990916 & 27 Relief Requests IWE-3 for Plant.Se Addresses Only IWE-3 Due to Util Urgent Need for Relief.Requests IWE-7 & IWE-8 Will Be Addressed at Later Date ML18152B3391999-09-27027 September 1999 Forwards Revised Relief Request IWE-3,which Now Includes Addl Visual Exam Requirement After post-repair or Mod Pressure Testing Is Completed,Per Telcon with NRC ML20212G5091999-09-22022 September 1999 Forwards in Triplicate,Applications for Renewal of Licenses for Listed Individuals.Encls Withheld,Per 10CFR2.790(a)(6) ML18152B6671999-09-17017 September 1999 Forwards Two NRC Forms 536,containing Info on Proposed Site Specific Operator Licensing Exam Schedules & Estimated Number of Applicants Planning to Take Exams And/Or Gfes,In Response to NRC Administrative Ltr 99-03 ML18152B3331999-09-17017 September 1999 Forwards Revised 180-day Response to NRC GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves. ML18152B3341999-09-16016 September 1999 Requests Relief from Specific Requirements of Subsection IWE of 1992 Edition with 1992 Addenda of ASME Section XI Re Containment Liner Examination Requirements,For North Anna & Surry Power Stations ML20211N2531999-09-0808 September 1999 Responds to Request to Exceed 60,000 Mwd/Mtu Lead Rod Burnup in Small Number of Fuel Rods in North Anna Unit 2.Informs That NRC Offers No Objection to Requested Use of Rods in Reconstituted Fuel Assembly.Se Supporting Request Encl ML18152B4471999-09-0101 September 1999 Requests That NRC Remove Listed Labels from Distribution ML20211L9151999-09-0101 September 1999 Forwards Response to NRC Request for Comments Re Closure of Review of Response to GL 92-01,Rev 1,Suppl 1, Reactor Vessel Structural Integrity ML20211J2211999-08-31031 August 1999 Approves Request to Remove Augmented ISI (Aii) Program for RCS Bypass Lines from North Anna Licensing Basis.Se Re Request to Apply LBB to Eliminate Augmented Insp Program on RCS Bypass Lines Encl ML20211H4131999-08-27027 August 1999 Informs That Util Revised Encl Bases for TS 2.2.1, Reactor Trip Sys Instrumentation Setpoints, Discussing Steam Flow/ Feed Flow Mismatch Portion of Steam Flow/Feed Flow Mismatch & Low SG Water Level Reactor Trip Setpoint ML20138B3241999-08-23023 August 1999 Forwards Draft Response to Question 1 Re NAPS USI A-46 ML20211D9041999-08-20020 August 1999 Forwards Revised Pages to Third Ten Year ISI Program & Relief Requests, Replacing Pages in 990408 Submittal ML20211B3871999-08-17017 August 1999 Requests Permission to Routinely Discharge from SW Reservior to Waste Heat Treatment Facility Under Existing Vpdes Permit Through Outfalls 108 & 103.Discharges Are Scheduled to Commence on 990907,due to High Priority Placed on Project ML20210T0671999-08-13013 August 1999 Informs of Completion of Review of Proposed Revs of Schedule for Withdrawal of Rv Surveillance Capsules Submitted by VEPCO on 981217.Approves Proposed Revs.Forwards Safety Evaluation ML20210Q9841999-08-12012 August 1999 Forwards Rev 1 to Vepc COLR for North Anna Unit 2,Cycle 13 Pattern Ud, Per TS 6.9.1.7.d.COLR Was Revised to Include Temp Coastdown Operation at End of Cycle 13 ML18152B4081999-08-0606 August 1999 Forwards Response to NRC 990520 & 0525 RAIs Re North Anna & Surry Responses to GL 95-07, Pressure Locking & Thermal Binding of SR Power-Operated Gate Valves. ML20210Q7661999-08-0606 August 1999 Requests Exception to 10CFR50.4 Requirement to Provide Total of Twelve Paper Copies When Submitting Surry & North Anna UFSAR Updates.Seek Approval to Submit Only Signed Original & One CD-ROM Version,Per Conversation with J Skoczlas ML18152B4091999-08-0505 August 1999 Forwards Vepc semi-annual fitness-for-duty Program Performance Data Rept for 990101-990630,IAW 10CFR26.71(d) ML20210N2921999-08-0505 August 1999 Discusses Which Submitted Proposed TSs Bases Change for Containment Leakage. Licensee Changes to Bases May Be Subj to Future Insps or Audits ML20210J8861999-08-0202 August 1999 Provides Clarification to Commitment Made in Identifying Extent by Which Existing Plant Design Complied with RG 1.97,specifically Re Variable, Radiation Exposure Rate ML20210F6121999-07-28028 July 1999 Forwards Supplemental Info on Proposed Irradiation of Fuel Rods Beyond Current Lead Rod Burnup Limit,Documenting Info Provided During 990624 Meeting & Suppl Original Submittal ML18152B3971999-07-26026 July 1999 Provides Estimates of Licensing Actions Expected to Be Submitted in Fys 2000 & 2001,in Response to NRC AL 99-02 ML20209E7621999-07-0909 July 1999 Provides Addl Info to Justify Use of Less than One Gpm Detectable Leakage Rate to Establish Required Margin for Crack Stability in LBB Analysis,Per 980623 Application on Reactor Coolant Loop Bypass Lines 05000338/LER-1999-005, Forwards LER 99-005-00,IAW 10CFR50.73.Commitment Made by Util Encl1999-07-0808 July 1999 Forwards LER 99-005-00,IAW 10CFR50.73.Commitment Made by Util Encl ML20209E3711999-07-0202 July 1999 Forwards Insp Repts 50-338/99-03 & 50-339/99-03 on 990425-0605.Violations Being Treated as Noncited Violations ML18152B4401999-07-0101 July 1999 Informs NRC That on 990511,Dominion Resources,Inc,Executed Amended & Restated Agreement & Plan of Merger with Consolidated Natural Gas Co ML18152B4371999-06-24024 June 1999 Forwards Response to NRC Request for Clarification of Relief Requests Submitted on 990212,requesting Relief from Performing Hydrostatic Testing for Certain Small Diameter Class 1,RCS Pressure Boundary Connections ML20196G2581999-06-23023 June 1999 Discusses Closure of GL 92-01,rev 1,suppl 1,reactor Vessel Structural Integrity ML20212J2951999-06-22022 June 1999 Forwards Corrected Markup & Typed Version of Affected Pages. Requests That Attached Pages for Those Previously Provided in 990506 Submittal Be Replaced & Incorporated Into NRC Review of Proposed TS ML18152B4361999-06-22022 June 1999 Forwards Response to RAI Re Surry & North Anna Power Stations,Units 1 & 2,per GL 96-06 ML20196F1151999-06-22022 June 1999 Forwards Relief Requests NDE-047 & NDE-048 for North Anna Power Station,Unit 1 Re ASME Section XI ISI Program ML20196G2211999-06-21021 June 1999 Forwards Licensee Sampling & Testing Obligations Re Vpdes Permit VA0052451 Reissuance Application.Details of Requests for Sampling & Testing Waivers,Included ML20195J7011999-06-15015 June 1999 Forwards Revised EPIP 2.01 Which Corrects Typo That Was Found in Step 10 of Procedure.Rev Does Not Implement Actions That Decrease Effectiveness of EP ML20195J1391999-06-11011 June 1999 Submits Addl Info as Addendum to Original Application Which Proposed Use of Three Chemicals in Bearing Cooling Tower at North Anna Power Station,Per Reissuance of Vpdes Permit ML18152B4301999-06-0303 June 1999 Informs of Util Intention to Revise Schedule for Submittal of License Renewal Applications for Surry & North Anna Power Stations to March 2002 ML20195C6601999-06-0101 June 1999 Forwards Response to NRC 990216 RAI Re Summary Rept of USI A-46 Program ML20207C9851999-05-28028 May 1999 Requests Regrading of Rt Robinson 990408 Written Exam,Based on Listed Reasons.Answer C for Question 18 Is Requested to Be Reconsidered as Correct or Question Be Deleted ML18152B4261999-05-28028 May 1999 Provides Formal Notification of Effect of Recent Organizational Restructuring on OLs of North Anna & Surry Power Stations,Per NRC 990513 Telcon Request ML18152B4221999-05-27027 May 1999 Forwards Info Concerning Changes to ECCS Evaluation Models & Application in Existing Licensing Analyses for Surry & North Anna Power Stations,Units 1 & 2 ML18152B4231999-05-26026 May 1999 Informs That Vepc Will Revise 180 Day Response to NRC GL 96-05,within 120 Days of Date of Ltr to Incorporate Commitment to Participate in Joint Owners Group Program as Applicable ML20206U7441999-05-20020 May 1999 Informs That NRC Unable to Conclude That NAPS Has Met Intent of Supplement 4 to GL 88-20.RAI Re Fire Area of IPEEE Encl. Response Requested within 90 Days of Submittal Date ML20207A8541999-05-20020 May 1999 Forwards RAI Re Licensee Listed Responses to GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves. Response Requested within 90 Days of Submittal Date ML20195B5381999-05-14014 May 1999 Forwards Rev 8,Change 2 to North Anna Units 1 & 2 IST Programs for Pumps & Valves. Summaries of Program Changes Provided for Each Unit IST Program.Relief Requests Have Been Removed from IST Programs ML18152A3701999-05-13013 May 1999 Submits Proposal to Use Provisions of ASME Section XI Code Case N-597 for Analytical Evaluation of Class 1,2 & 3 Carbon & Low Alloy Steel Piping Components Subjected to Wall Thinning as Result of Flow Accelerated or Other Corrosion ML20206L4661999-05-10010 May 1999 Forwards SE Accepting Request to Delay Submitting Plant, Unit 1 Class Piping ISI Program for Third Insp Interval Until 010430,to Permit Development of Risk Informed ISI Program for Class 1 Piping 1999-09-08
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARML20217K4391999-10-18018 October 1999 Provides Response to RAI to Support USI A-46 Program Submittal for North Anna Power Station,Units 1 & 2.Rev 10 to BNL Rept 52361,encl ML20217H3301999-10-14014 October 1999 Forwards Rev 0 to COLR for North Anna 2 Cycle 14 Pattern Su. No New Commitments Intended by Ltr ML18152B3541999-10-12012 October 1999 Requests Use of Code Case N-532 & N-619,per Provisions of 10CFR50.55a(a)(3).Detailed Info Supporting Request,Encl. Attachment 2 Includes Technical White Paper That Provides Further Technical Info ML20216K1681999-10-0101 October 1999 Forwards Vols I-VIII of Rev 35 to UFSAR for Naps.Rev Also Includes Update to Chapter 17 of Ufsar,Which Contains Operational QA Program.Changes to Program Description Do Not Reduce Commitments Contained Therein ML18152B3391999-09-27027 September 1999 Forwards Revised Relief Request IWE-3,which Now Includes Addl Visual Exam Requirement After post-repair or Mod Pressure Testing Is Completed,Per Telcon with NRC ML20212G5091999-09-22022 September 1999 Forwards in Triplicate,Applications for Renewal of Licenses for Listed Individuals.Encls Withheld,Per 10CFR2.790(a)(6) ML18152B3331999-09-17017 September 1999 Forwards Revised 180-day Response to NRC GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves. ML18152B6671999-09-17017 September 1999 Forwards Two NRC Forms 536,containing Info on Proposed Site Specific Operator Licensing Exam Schedules & Estimated Number of Applicants Planning to Take Exams And/Or Gfes,In Response to NRC Administrative Ltr 99-03 ML18152B3341999-09-16016 September 1999 Requests Relief from Specific Requirements of Subsection IWE of 1992 Edition with 1992 Addenda of ASME Section XI Re Containment Liner Examination Requirements,For North Anna & Surry Power Stations ML18152B4471999-09-0101 September 1999 Requests That NRC Remove Listed Labels from Distribution ML20211L9151999-09-0101 September 1999 Forwards Response to NRC Request for Comments Re Closure of Review of Response to GL 92-01,Rev 1,Suppl 1, Reactor Vessel Structural Integrity ML20211H4131999-08-27027 August 1999 Informs That Util Revised Encl Bases for TS 2.2.1, Reactor Trip Sys Instrumentation Setpoints, Discussing Steam Flow/ Feed Flow Mismatch Portion of Steam Flow/Feed Flow Mismatch & Low SG Water Level Reactor Trip Setpoint ML20138B3241999-08-23023 August 1999 Forwards Draft Response to Question 1 Re NAPS USI A-46 ML20211D9041999-08-20020 August 1999 Forwards Revised Pages to Third Ten Year ISI Program & Relief Requests, Replacing Pages in 990408 Submittal ML20210Q9841999-08-12012 August 1999 Forwards Rev 1 to Vepc COLR for North Anna Unit 2,Cycle 13 Pattern Ud, Per TS 6.9.1.7.d.COLR Was Revised to Include Temp Coastdown Operation at End of Cycle 13 ML20210Q7661999-08-0606 August 1999 Requests Exception to 10CFR50.4 Requirement to Provide Total of Twelve Paper Copies When Submitting Surry & North Anna UFSAR Updates.Seek Approval to Submit Only Signed Original & One CD-ROM Version,Per Conversation with J Skoczlas ML18152B4081999-08-0606 August 1999 Forwards Response to NRC 990520 & 0525 RAIs Re North Anna & Surry Responses to GL 95-07, Pressure Locking & Thermal Binding of SR Power-Operated Gate Valves. ML18152B4091999-08-0505 August 1999 Forwards Vepc semi-annual fitness-for-duty Program Performance Data Rept for 990101-990630,IAW 10CFR26.71(d) ML20210J8861999-08-0202 August 1999 Provides Clarification to Commitment Made in Identifying Extent by Which Existing Plant Design Complied with RG 1.97,specifically Re Variable, Radiation Exposure Rate ML20210F6121999-07-28028 July 1999 Forwards Supplemental Info on Proposed Irradiation of Fuel Rods Beyond Current Lead Rod Burnup Limit,Documenting Info Provided During 990624 Meeting & Suppl Original Submittal ML18152B3971999-07-26026 July 1999 Provides Estimates of Licensing Actions Expected to Be Submitted in Fys 2000 & 2001,in Response to NRC AL 99-02 ML20209E7621999-07-0909 July 1999 Provides Addl Info to Justify Use of Less than One Gpm Detectable Leakage Rate to Establish Required Margin for Crack Stability in LBB Analysis,Per 980623 Application on Reactor Coolant Loop Bypass Lines 05000338/LER-1999-005, Forwards LER 99-005-00,IAW 10CFR50.73.Commitment Made by Util Encl1999-07-0808 July 1999 Forwards LER 99-005-00,IAW 10CFR50.73.Commitment Made by Util Encl ML18152B4401999-07-0101 July 1999 Informs NRC That on 990511,Dominion Resources,Inc,Executed Amended & Restated Agreement & Plan of Merger with Consolidated Natural Gas Co ML18152B4371999-06-24024 June 1999 Forwards Response to NRC Request for Clarification of Relief Requests Submitted on 990212,requesting Relief from Performing Hydrostatic Testing for Certain Small Diameter Class 1,RCS Pressure Boundary Connections ML18152B4361999-06-22022 June 1999 Forwards Response to RAI Re Surry & North Anna Power Stations,Units 1 & 2,per GL 96-06 ML20196F1151999-06-22022 June 1999 Forwards Relief Requests NDE-047 & NDE-048 for North Anna Power Station,Unit 1 Re ASME Section XI ISI Program ML20212J2951999-06-22022 June 1999 Forwards Corrected Markup & Typed Version of Affected Pages. Requests That Attached Pages for Those Previously Provided in 990506 Submittal Be Replaced & Incorporated Into NRC Review of Proposed TS ML20195J7011999-06-15015 June 1999 Forwards Revised EPIP 2.01 Which Corrects Typo That Was Found in Step 10 of Procedure.Rev Does Not Implement Actions That Decrease Effectiveness of EP ML18152B4301999-06-0303 June 1999 Informs of Util Intention to Revise Schedule for Submittal of License Renewal Applications for Surry & North Anna Power Stations to March 2002 ML20195C6601999-06-0101 June 1999 Forwards Response to NRC 990216 RAI Re Summary Rept of USI A-46 Program ML18152B4261999-05-28028 May 1999 Provides Formal Notification of Effect of Recent Organizational Restructuring on OLs of North Anna & Surry Power Stations,Per NRC 990513 Telcon Request ML20207C9851999-05-28028 May 1999 Requests Regrading of Rt Robinson 990408 Written Exam,Based on Listed Reasons.Answer C for Question 18 Is Requested to Be Reconsidered as Correct or Question Be Deleted ML18152B4221999-05-27027 May 1999 Forwards Info Concerning Changes to ECCS Evaluation Models & Application in Existing Licensing Analyses for Surry & North Anna Power Stations,Units 1 & 2 ML18152B4231999-05-26026 May 1999 Informs That Vepc Will Revise 180 Day Response to NRC GL 96-05,within 120 Days of Date of Ltr to Incorporate Commitment to Participate in Joint Owners Group Program as Applicable ML20195B5381999-05-14014 May 1999 Forwards Rev 8,Change 2 to North Anna Units 1 & 2 IST Programs for Pumps & Valves. Summaries of Program Changes Provided for Each Unit IST Program.Relief Requests Have Been Removed from IST Programs ML18152A3701999-05-13013 May 1999 Submits Proposal to Use Provisions of ASME Section XI Code Case N-597 for Analytical Evaluation of Class 1,2 & 3 Carbon & Low Alloy Steel Piping Components Subjected to Wall Thinning as Result of Flow Accelerated or Other Corrosion ML20206H0221999-05-0303 May 1999 Informs That Licensee Changes Bases for TS 3/4.6.1.2, Containment Leakage. Changes Allow Use of Other NRC Staff Approved/Endorsed Integrated Leak Test Methodologies to Perform Containment Leakage Rate Testing.Ts Bases Page,Encl ML20206G9481999-05-0303 May 1999 Informs NRC That Insp of 58 Accessible safety-related Pipe Supports Completed in Response to NOV from Insp Rept 50-338/98-05 & 50-339/98-05.Commitments Made Include Plans to Perform Assessment of Welding & Welding Insp ML20205T1181999-04-16016 April 1999 Requests NRC Approval Prior to Proposed Irradiation of Fuel Rods Beyond Current Lead Rod Burnup Limit.Nrc Concurrence with Irradiation Program Requested by End of June 1999 ML20205P1891999-04-0808 April 1999 Forwards ISI Program for Third ten-yr ISI Interval for North Anna Unit 1 for Class 1,2 & 3 Components & Component Support.Third ten-yr Insp Interval for North Anna Unit 1 Begins on 990501.Page 2-26 of Encl Not Included ML20205K3631999-04-0505 April 1999 Requests That Relief Request IWE-3 Be Removed from 980804 Relief Requests Submitted to Nrc.Subject Relief Request Was Inadvertently Retained in Attachment 1 for Unit 1 ML20205K2191999-04-0101 April 1999 Forwards Response to NRC 990106 RAI Re Util Summary Rept on USI A-46 Program,Submitted 970527.Calculations & Corrected Table 11.1-1,encl ML18151A5851999-03-31031 March 1999 Forwards Rept on Status of Decommissioning Funding for Each of Four Nuclear Power Reactors,Per 10CFR50.75(f)(1) ML18152A2801999-03-30030 March 1999 Forwards Summary of Structural Integrity Evaluation of Thermally Induced Over Pressurization of Containment Penetration Piping During DBA for SPS & Naps,Units 1 & 2,per GL 96-06.Draft Proposed UFSAR Revised Pages,Encl ML20204H0331999-03-17017 March 1999 Forwards Rev 5 to PSP for Surry & North Anna Power Stations & Associated Isfsis.Description & Justification for Changes Included with Plan Rev.Rev 5 to PSP Withheld Per 10CFR73.21 ML20205E2701999-02-25025 February 1999 Forwards Rept on Status of Decommissioning Funding for North Anna Power Station,Units 1 & 2.Trust Agreement Between Old Dominion & Bankers Trust Co,Effective 990301,attached ML20207A8741999-02-25025 February 1999 Draft Response to NRC Telcon Re Licensee Request for Approval of LBB Evaluation in Support of Elimination of Augmented Insp Program on RCS Loop Bypass Lines.Response Justifies Use of Less than One Gpm Detectable Leakage Rate ML18152B5401999-02-11011 February 1999 Requests Relief from Specific Requirements of Subsection Iwl of 1992 Edition with 1992 Addenda of ASME Section Xi,Per 10CFR50.55a(a)(3) ML20203C8181999-02-0505 February 1999 Forwards Response to NRC 981217 Telcon RAI Re risk-basis of Nitrogen Accumulator Action Statement to Complete NRC Review of 951025 Proposed TS Changes 1999-09-27
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VINGINIA l$LECTHIC AND Pownn COMPANY l l RICHMOND,VIHOINIA 202 61 l
i May 6, 1996 l
l U. S. Nuclear Regulatory Commission Serial No.96-208 Attention: Document Control Desk Washington, D.C. 20555 Docket Nos. 50-338 l
50-339 License Nos. NPF-4 l l NPF-7 l l ;
Gentlemen:
1 1
YRGINIA ELECTRIC AND POWER COMPANY l NORTH ANNA POWER STATION UNITS NO.1 AND 2 NOTIFICATION OF INTENTION TO USE LEAD FUEL ASSEMBLIES WITH ADVANCED CLADDING MATERIALS This letter provides notification of Virginia Electric and Power Company plans to load four l lead test fuel assemblies supplied by Framatome Cogema Fuels (FCF), formerly the B&W j
Fuel Company, into North Anna Unit 1. It is our intention to begin irradiation of these assemblies in North Anna Unit 1 Cycle 13, which is currently scheduled to begin l operation in June,1997. This program will be limited to the irradiation of the four lead test l assemblies, with the objective of demonstrating the performance of the FCF fuel
- assembly design features in the North Anna units under operating conditions typical of our normal fuel management.
The North Anna lead test assemblies will be very similar to Mark-BW17 assemblies l previously irradiated in other Westinghouse-designed reactors. However, the North Anna assemblies will incorporate several new features, including use of an advanced zirconium-based alloy, designated as MS, for the fuel assembly structural tubing. The fuel i rod cladding in these assemblies will be fabricated from two advanced zirconium based alloys, M4 and MS, which have previously been approved for use as cladding materials in
- demonstration assemblies in the McGuire Unit 1 and Three Mile Island Unit 1 reactors.
l Additional information on the design features of the lead test assemblies is provided in Attachment 1, along with a summary of the evaluations that will be performed to support their use at North Anna.
i The Design Section of the North Anna Technical Specifications currently defines the fuel
. rod cladding material as either Zircaloy-4 or ZlRLO. The references listed in Administrative Section 6.9 for the Core Operating Limits Report (COLR) address evaluations with Zircaloy-4 or ZlRLO rod cladding materials. Therefore, use of these lead test assemblies will require an amendment to the operating license, in the form of a 9605130450 960506 PDR ADOCK 05000338 P PDR
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l license condition and changes to the appropriate Technical Specifications. Further, use l l of a material other than Zircaloy-4 or ZlRLO for the cladding material will require an exemption to the requirements of 10 CFR 50.46, which requires the use of an approved ECCS evaluation model for reactors with Zircaloy clad fuel.
- In addition, the NRC's July 29,1986 safety evaluation report which approved our normal reload nuclear design methodology (VEP-FRD-42 Rev.1-A) specified that, in its present form, our methodology could be applied only to Westinghouse-supplied fuel in Westinghouse-supplied reactors. Therefore, we will also be seeking NRC concurrence that our standard reload design methodology may be applied to the North Anna cores
, which contain the four FCF lead test assemblies. The remainder of the fuel in the cores will continue to be supplied by Westinghouse.
We plan to submit our proposed license amendments for the lead assemblies by August 1,1996, for your review and approval. To provide sufficient time for including the lead assembly features into the North Anna Unit 1 Cycle 13 reload core design, we will be requesting that NRC approve our submittal by February 1,1997.
While it is our intent to insert these assemblies into Unit 1 for three consecutive cycles of irradiation, we will also request that the proposed license conditions be applicable to both North Anna Units 1 and 2 to allow additional flexibility in the irradiation schedule for the lead test assemblies. To support this request, the evaluations in our submittal will be performed to support operation of the FCF assemblies in either North Anna Unit 1 and 2.
Any effect of using the lead test assemblies will be incorporated into each appropriate cycle specific reload analysis.
l If you have any questions about the proposed lead test assembly program, please contact us.
Very truly yours, James P. O'Hanlon Senior Vice President - Nuclear Attachment
cc: U. S. Nuclear Regulatory Commission Region ll 101 Marietta Street, N. W.
Suite 2900 Atlanta, Georgia 30323 R. D. McWhorter NRC Senior Resident Inspector North Anna Power Station 1
1
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i
s Attachment i Planned Evaluations to Support Use of FCF Lead Test Assemblies in North Anna General Description The North Anna lead test assemblies will be very similar to Mark-BW17 assemblies i previously irradiated in the McGuire Units 1 and 2, Catawba Units 1 and 2, and Trojan reactors. However, the North Anna assemblies will be an advanced Mark-BW17 design, incorporating several new features, including: mid-span mixing grids, an advanced (fine ,
mesh) debris filter bottom nozzle, a quick disconnect top nozzle, a floating top end grid I (only the middle grids on the Mark-BW17 design ' float'), and use of an advanced zirconium-based alloy, designated as MS, for the fuel assembly structural tubing. The fuel rod cladding in these assemblies will be fabricated from two advanced zirconium based alloys, M4 and MS. The majority of the rods will use M5 for the fuel rod cladding, but two of the assemblies will also contain a limited number of fuel rods with cladding fabricated from the M4 alloy. These two alloys have previously been used as cladding materials for limited numbers of fuel rods in demonstration assemblies in the McGuire Unit 1 and Three Mile Island Unit 1 reactors. The North Anna lead test assemblies will differ frorn these demonstration assemblies in using advanced alloys as the cladding material for all fuel rods in the assemblies, as well as using alloy M5 for the guide thimbles.
Evaluations to be Performed Evaluation of the lead test assemblies will be performed jointly by FCF and Virginia Electric and Power Company. These evaluations will include both testing and analyses, and will address all aspects of safety, including mechanical, thermal hydraulic, neutronic, transient, and accident analyses.
. Material testing for the new cladding alloys includes strength measurements, corrosion behavior, and swelling and rupture characteristics. Because the M5 alloy will comprise the vast majority of the cladding and because M4 closely parallels the Zircaloy-4 alloy specification, the brittle fracture testing (cold water plunge tests) and the high temperature oxidation rate testing were conducted only for the M5 alloy. The results of these tests indicate that the 17 percent local oxidation limit remains applicable and that the Baker-Just metal-water reaction rate correlation remains conservative. Swelling and rupture data, for use in benchmarking the rupture characteristics of the LOCA evaluation model, have been obtained for both alloys.
The testing covers the expected range of parameter variation, providing sufficient basis for an adjustment to the assumed materials characteristics if required.
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Mechanical and hydraulic tests will also be conducted on components and a prototype fuel assembly. Component testing will address holddown spring and nozzle compression, hydraulic pressure drop, grid dynamic crush testing, and general functional tests. The prototype mechanical tests, comprising measurement of assembly static and dynamic responses to various steady state and impact loading situations, have been completed. Additional tests of assembly pressure drop, control rod drop time, life and wear, and flow induced vibration susceptibility will also be performed. ,
Using approved methods and established design limits, FCF will verify that the static and dynamic structural characteristics of the lead test assemblies are compatible with the resident fuel in the core. The lead test assemblies will be designed to maintain I their mechanical integrity through the planned operating life of the fuel. The j evaluations will address normal operation, faulted conditions (seismic and LOCA), and shipping and handling loads. The lead test assemblies will also be designed for compatibility with all core intemais, instrumentation, and control rod assemblies.
I I . FCF evaluations of the fuel rod mechanical and thermal perfomlance for the lead test assemblies will be completed using their approved codes and methods (BAW-10162P-A and BAW-10084P-A). The fuel rod strength analysis will be performed in accordance with BAW-2133P, " Mark-BW Advanced Cladding Fuel Rod Evaluation,"
which was submitted to the NRC, and referenced in subsequent analyses, to support Technical Specifications changes for the use of the advanced cladding materials at other reactors. The mechanical analyses will address shipping and handling, stress, creep collapse, strain and fatigue. The thermal performance analyses will demonstrate that criteria for fuel rod temperature and intemal pressure are met. The transient portions of these evaluations will initially be performed using generic FCF input for bounding plant transient conditions. During the design phase for each actual operating cycle, plant and cycle specific information will be provided to FCF by the l Virginia Electric and Power Company to allow verification that the generic data remain i applicable, or - if necessary - to allow reevaluation for the actual operating conditions. l l
These cycle specific data will be generated with Virginia Electric and Power Company !
codes using methcds which are consistent with FCF's methodology approved by the l NRC. !
. The thermal hydraulic design evaluation of the lead test assemblies will be performed by FCF using their approved methods and information on the characteristics of the core intemals and the resident fuel supplied by Virginia Electric and Power Company. {
Mixed and full core evaluations will be performed to account for the lead test assembly l
- impacts on minimum DNBR, pressure drop, fuel assembly lift and lateral flow i
! I velocities. It is expected that the DNBR performance for the lead test assemblies will be bounded by'the resident fuel. The thermal hydraulic (DNB) evaluation of the l l
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l applicable North Anna reload cores would then be performed by Virginia Electric and Power Company assuming a full core of Westinghouse fuel.
l . Virginia Electric and Power Company will perform the neutronic evaluation of the
! cores containing the FCF lead test assemblies in accordance with our normal reload nuclear design methodology (VEP FRD-42 Rev.1-A). The nuclear design will ensure that the lead test assemblies are not placed in the highest rod power density locauons.
. Virginia Electric and Power Company will assess the impact of the lead test i assemblies on the non-LOCA core accident analyses. Any lead test assembly design features which differ from the resident fuel and which might affect input to the safety l l analyses will be defined, and the impact of these design changes on the analyses of l l record will be assessed. Because the lead test assemblies will not be placed in the highest rod power density locations, it is expected that they will be bounded by the ,
safety analyses performed for the resident fuel.
. The LOCA performance of the lead test assemblies will be calculated with the l l Framatome Technologies RSG LOCA Evaluation Model. This evaluation model is l described in BAW-10168 and has been approved by the NRC. Inputs which are l descriptive of the Mark-BW17 lead test assemblies will be compiled and incorporated into a large break LOCA model representing the North Anna units. The treatment of i swelling and rupture in the advanced Mark-BW17 LOCA calculations will be based on ongoing evaluations of the rupture testing data. The rupture characteristics of the M4 l alloy are essentially the same as those of Zircaloy-4. Thus the modeling of NUREG-l 0630 is directly applicable to that alloy. The rupture characteristics of the M5 alloy are similar to M4 and Zircaloy-4, but a judgement as to the applicability of NUREG-0630 has not yet been finalized. If the rupture characteristics of the M5 alloy are determined to be within the range of applicability of the NUREG-0630 data l
correlations, the material properties as determined within NUREG-0630 for Zircaloy-4 will be applied. However, if the ongoing evaluations determine that a meaningful difference exists between the rupture properties determined in NUREG-0630 and those which would be appropriate for the M5 alloy, the material properties for the NUREG-0630 model will be rederived or adjusted based on the rupture test results.
The LOCA calculations would then be conducted with the revised material properties appropriate for the advanced alloy.
1 The calculational base to be developed will include a bumup sensitivity study, a 3-break mini-spectrum, and a 2-elevation K, validation study. These studies will be conducted for the M5 cladding because its properties differ from Zircaloy-4 more than the M4 material l properties. Upon completion of the studies, the predicted peak cladding temperatures will be compared to those for the resident Westinghouse fuel, and a LOCA differential result, DPCT, will be established. If the DPCT appears to be strongly dependent on material Page 3 of 4
4 properties which differ for the M4 and M5 alloys, selected calculations will be redone using the M4 properties to establish a DPCT specific to the M4 alloy. The DPCT(s) will be applied to the licensing calculational results for the resident fuel design to provide the licensing basis for the lead test assemblies. It is expected that the DPCT(s) will be substantially negative, so that the LOCA analyses of record (based on a full core of Westinghouse fuel) will remain bounding for the cores which incorporate the FCF lead test assemblies. At this time, there are no plans to impose peaking requirements on the lead test assemblies for the LOCA analyses which differ from those applied to the remainder of the core.
l l
Where any differences exist between the North Anna Unit 1 and Unit 2 designs, a bounding approach will be taken for the aforementioned evaluations to support operation of the FCF assemblies in either North Anna core. This will ensure that the lead test assemblies are technically capable of operating in either unit. This degree of flexibility l may be desirable if reinsertion is delayed to allow time (off critical path) to perform more l extensive characterization of the assemblies after one or two cycles of operation: upon completion of the examinations, rather than waiting a year or more for the next refueling outage to reinsert the LTAs, the assemblies could instead be incorporated into the next s cycle of the other unit. Similarly, if NRC review and approval of the program can not support the proposed use in North Anna 1 Cycle 13, it would be possible to irradiate the
- lead test assemblies in the North Anna Unit 2 reactor rather than wait a full cycle to conduct the program in North Anna Unit 1. Any effect of using the lead test assemblies
- will be incorporated into the cycle specific reload analyses for each applicable cycle.
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