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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 ML20216E6661999-09-10010 September 1999 Forwards Analysis of Capsule W Virginia Power North Anna Unit 1 Nuclear Power Plant Reactor Vessel Matl Surveillance Program, for Capsule Withdrawn on 980922 ML20211N2531999-09-0808 September 1999 Responds to Request to Exceed 60,000 Mwd/Mtu Lead Rod Burnup in Small Number of Fuel Rods in North Anna Unit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 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 ML18152B4471999-09-0101 September 1999 Requests That NRC Remove Listed Labels from Distribution ML18151A6301999-09-0101 September 1999 Submits Summary of 990812-13 Training Managers Conference on Recent Changes to Operator Licensing Program.Meeting Covered Changes to Regulations,Examination Stds,New Insp Programs & Other Training Issues.List of Attendees Encl 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 ML18152B3701999-08-30030 August 1999 Informs That Virginia Electric & Power Co Hereby Granted Approval to Submit Original & One Paper Copy & 6 CD-ROM Copies of Updates to FSAR as Listed,In Response to ,Requesting Exception to 10CFR50.4(b)(6) 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 ML18152B3811999-08-23023 August 1999 Forwards Safety Evaluation,Granting Relief Request IWL-3 Re ISI Programs,Per Util 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 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 ML18152B4091999-08-0505 August 1999 Forwards Vepc semi-annual fitness-for-duty Program Performance Data Rept for 990101-990630,IAW 10CFR26.71(d) ML18152B3731999-08-0505 August 1999 Informs That NRC Plans to Administer Gfes of Written Operator Licensing Exam on 991006 at North Anna & Surry. Requests Info Re Individuals Taking Exam,Personnel Who Will Have Access to Exam.Sample Registration Ltr Encl 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 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 ML18152B3951999-06-30030 June 1999 Informs That Effective 990706,GE Edison Has Been Assigned as Project Manager,Project Directorate II-1,for North Anna Power Station.Edison Will Continue in Assignment as Project Manager for Surry Power Station 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 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 ML18152B4361999-06-22022 June 1999 Forwards Response to RAI Re Surry & North Anna Power Stations,Units 1 & 2,per GL 96-06 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 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 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 ML20216E6661999-09-10010 September 1999 Forwards Analysis of Capsule W Virginia Power North Anna Unit 1 Nuclear Power Plant Reactor Vessel Matl Surveillance Program, for Capsule Withdrawn on 980922 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 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 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 ML18152B4361999-06-22022 June 1999 Forwards Response to RAI Re Surry & North Anna Power Stations,Units 1 & 2,per GL 96-06 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 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 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
[Table view] |
Text
VIRGINIA EimcTRIC ann POWER COMPANY Riemioso, You;iNir 2.us!
August 27, 1999 U.S. Nuclear Regulatory Commission Serial No.99-447 Attention: Document Control Desk NL&OS/GSS/ETS R0 Washington, D.C. 20555 Docket Nos.
50-338/339 License Nos. NPF-4/7 Gentlemen:
VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNITS 1 AND 2 TECHNICAL SPECIFICATION BASES CHANGE 1
STEAM FLOW / FEED FLOW MISMATCH SETPOINT i
Virginia Electric and Power Company has revised the Bases for Technical Specifications 2.2.1," Reactor Trip System instrumentation Setpoints." Changes update the Bases section to discuss the Steam Flow / Feed Flow Mismatch portion of the i
" Steam Flow / Feed Flow Mismatch and Low Steam Generator Water Level" reactor trip setpoint. The Bases discussion previously included a steam flow / feed flow mismatch value in pounds-mass per hour. The value has been revised to include steam flow as a percentage of the rated thermal power which is consistent with the limiting safety system setting required by Technical Specification 2.2.1.
We are providing these Technical Specification Bases changes for your information.
The Technical Specifications Bases changes have been reviewed and approved by the Station Nuclear Safety and Operating Committee and the Management Safety Review Committee. It has been determined that these changes do not involve an unreviewed safety question as defined in 10 CFR 50.59.
A discussion and the Technical Specifications Bases changes are provided in Attachments 1 and 2 respectively.
There are no commitments made in this letter. If you have any further questions, please contact us.
Very truly yours,
\\
020021 l
Leslie N. Hartz Vice President - Nuclear Engineering and Services Attachments O
3 9909020004 990827 PDR ADOCK 05000338 P
PDR
o 1
cc:, U.S. Nuclear Regulatory Commission L
Region ll l
Atlanta Federal Center 61 Forsyth Street, SW Suite 23T85 Atlanta, Georgia 30303 Mr. M. J. Morgan NRC Senior Resident inspector North Anna Power Station Commissioner Department of Radiological Health Room 104A 1500 East Main Street Richmond, Virginia 23219 -
Mr. J. E. Reasor Old Dominion Electric Cooperative Innsbrook Corporate Center 4201 Dominion Blvd.
Glen Allen, Virginia 23060 l
l l
l l
4
4 Discussion of Bases Change I
North Anna Power Station Units 1 and 2 Virginia Electric and Power Company
r.
^
l 1
I l
DISCUSSION OF CHANGE l
Introduction The Steam Flow - Feed Flow Mismatch Reactor Trip Setpoint specified in the North Anna Unit 1 and Unit 2 Technical Specifications Bases for Section 2.2-1 is 6
incorrect. Specifically, the flow value of 1.616
- 10 lbs/ hour represents 40% of l
the pre-core uprate design flow for RATED THERMAL POWER conditions (i.e.,
4.04
For consistency between units and to avoid future bases changes, the Technical Specification Bases Section for the Unit 1 and Unit 2 Steam - Feedwater Flow Mismatch Reactor Trip are being changed so that the Trip Setpoint is expressed in terms of a percentage of nominal steam flow (i.e., Flownom) at RATED THERMAL POWER instead of a specific flow value given in terms of Ibs/ hour.
Background
i in 1998, two precision feedwater flow tests (one per unit) were performed at North Anna in order to quantify the accuracy of the secondary calorimetric using both feedwater flow and steam flow. The results of the flow tests determined that measured feedwater flow was very accurate when compared to the reference flow instrumentation. In addition, it was also determined that the measured steam flow was reading approximately 2.0% higher than feedwater flow and thus, the units were not operating at their RATED THERMAL POWER. Subsequent to the flow tests, the North Anna secondary calorimetric was revised so that it would be based on feedwater flow. As part of the design change package (DCP) used to implement this change to the secondary calorimetric program, the impact of increasi g power by approximately 2.0% on the Westinghouse 7300 Protection and Control System was assessed.
This assessment determined that the increased power level would not adversely affect the 7300 Protection and Control System and that no immediate instrument scaling changes were necessary.
It was demonstrated that the Reactor Protection System would satisfy the applicable Safety Analysis assumptions and Technical Specifications requirements.
However, during the review it was determined that it would be advantageous to normalize the steam flow portion of the 7300 Protection System to " Reference Feedwater Flow" as determined by the FLOWCALC portion of the secondary calorimetric computer program. This normalization process enhances the accuracy of the Reactor Trips, ESFAS Actuations, and indications developed from this parameter.
In addition to normalizing steam flow to feedwater flow, it was also decided that i
the Steam Flow - Feed Flow Mismatch Reactor Trip Setpoint would be changed Page 1 of 5
y from 34% of Flownom to 40% of Flow consistent with the limit stated in nom Technical Specifications, Table 2.2-1. In the mid 1980's, the Steam Flow - Feed Flow Mismatch Reactor Trip Setpoint was changed from the nominal value of 40% of Flownom to 34% of Flownom to account for feedwater flow fouling. Since the performance of the recent flow tests, feedwater flow has been proven to be accurate and thus it has been decided to change the Steam Flow - Feed Flow l
Mismatch Reactor Trip Setpoint back to the nominal value of 40% of Flownom.
During the preparation of the DCP, to implement the scaling changes discussed above, it was discovered that the setpoint value in the Technical Specification Bases Section for the Steam Flow - Feed Flow Mismatch Reactor Trip was i
incorrect.
Discussion Since the mid-1980s, the nominal 100% power flowrate for North Anna Units 1 and 2 has changed two times. The nominal flowrate for pre-core uprate was 4.04
- 10e !bs/ hour. After the core uprate in 1986, the nominal flowrate for 100%
8 power conditions increased'to 4.26
- 10 lbs/ hour, in 1993 after the Steam Generator Replacement Project (SGRP), the nominal flowrate changed again 8
and decreased to 4.247
- 10 lbs/ hour. During this time, the Steam Flow - Feed j
Flow Mismatch Reactor Trip Setpoint value given in the Bases Section remained 8
8 at 31.616 + 10 lbs/ hour for Unit 1 and > 1.616
Further, as stated above, the actual setpoint installed in the plant for Unit 2 is 8
34% of Flow which equates to 1.444
- 10 lbs/ hour (Post -SGRP). For Unit 1, nom the Steam Flow - Feed Flow Mismatch Reactor Trip Setpoint is set at 34% of the 8
Pre-SGRP Flow value which equates to 1.448
- 10 lbs/ hour. As can be seen, nom keeping the Bases up to date ucing an actual flow value given in terms of Ibs/ hour would have required multiple changes to the Bases Section over time.
The Trip Setpoint value given in Technical Specifications, Table 2.2-1, item 14,
" Steam Flow - Feed Flow Mismatch Reactor Trip" is 5 40% of full steam flow at RATED THERMAL POWER.
This setpoint permits changes in the nominal flowrate and thus changes to the actual trip setpoint. Further, the setpoint in the table permits (he actual trip setpoint to be set more conservatively if required.
The Steam Flow - Feed Flow Mismatch Reactor Trip coincident with Steam Generator low Level is considered to be a backup reactor trip and is not credited in the Chapter 15 Safety Analysis.
However, according to the North Anna UFSAR, Section 7.2.2.3.5 this reactor trip function is credited as a backup to the Steam Generator Low-Low Level Reactor Trip in the IEEE 279-1971 scenario for the interaction between protection and control for power levels 3 50%.
Therefore, the implied Design Basis Limit for this Reactor Trip function is 50%
power or 50% of nominal flow at RATED THERMAL POWER. Specifying the trip setpoint for this function as 5 40% of nominal steam flow at RATED THERMAL Page 2 of 5
U, l
P,0WER provides a TOTAL' ALLOWANCE of 10% 'of. nominal steam flow 'at RATED THERMAL-POWER. The Channel Statistical Allowance (CSA) for the Steam Flow - Feed Flow Mismatch Reactor Trip is + 7.31% of Flownom.
- Subtracting the CSA value of 7.31% of Flownom from the TOTAL ALLOWANCE VALUE of 10% of Flownom yields a margin to the Design Basis Limit of 2.69% of Flownom for power levels t 50E Based on this analysis, the Steam Flow - Feed Flow Mismatch Reactor Trip Setpoint given in Table 2.2-1, item 14, is acceptable and provides positive margin ~to the Design Basis Limit. Changing the Technical Specifications Bases Section for the Steam Flow - Feed Flow Mismatch Reactor Trip to match the nomenclature given in Table 2.2-1 will more accurately reflect j
actual plant conditions and will have no impact on the Safety Analysis, Technical Specifications or on plant operations.
Specific Changes i
i Unit 1 Technical Specifications The third sentence in the Unit 1 Technical Specifications Bases Section 2.2.1 for
- the Steam /Feedwater Flow Mismatch and Low Steam Generator Water Level
. Reactor Trip currently states:
The Steam /Feedwater Flow Mismatch portion of this trip is activated when the steam fiow exceeds the feedwater flow by 31.616
- 10e Ibs/ hour of full steam flow at RATED THERMAL POWER."
The third sentence in the Unit 1 Technical Specifications Bases Section 2.2.1 for L
the Steam / Fee.dwater Flow Mismatch and Low Steam Generator Water Level is modified as follows:
"The Steam /Feedwater Flow Mismatch portion of this trip is activated before steam flow exceeds feedwater flow by 40% of nominal steam flow at RATED THERMAL POWER."
Unit 2 Technical Specifications The' third sentence in the Unit 2 Technical Specifications Bases Section 2.2.1 for the Steam /Feedwater Flow Mismatch and Low Steam Generator Water Level Reactor Trip currently states:-
i "The Steam /Feedwater Flow Mismatch portion of this trip is activated 8
l when the steam flow exceeds the feedwater flow by > 1.616
- 10 lbs/ hour of full steam flow at RATED THERMAL POWER."
Page 3 of 5
The third sentence in the Unit 2 Technical Specifications Bases Section 2.2.1 for the Steam /Feedwater Flow Mismatch and Low Steam Generator Water Level is modified as follows:
"The Steam /Feedwater Flow Mismatch portion of this trip is activated before steam flow exceeds feedwater flow by 40% of nominal steam flow at RATED THERMAL POWER."
Safety Significance This correction to the Unit 1 and Unit 2 Technical Specifications Bases for the Steam /Feedwater Flow Mismatch coincident with Low Steam Generator Water Level Reactor Trip does not create an unreviewed safety question as described below:
The Bases change does not increase probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report.
Changing the setpoint value given in the Bases for the Steam Flow - Feed Flow Mismatch podion of the reactor trip from an incorrect flow value given in terms of lbs/ hour to a flow value given in percent of nominal flow that is consistent with the trip setpoint specified in Table 2.2-1 has no impact on the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report. Note that the Steam Flow - Feed Flow Mismatch coincident with Steam Generator Low Level Reactor Trip is not credited in the Chapter 15 Safety Analysis.
This Bases change will now express the Trip Setpoint for the Steam Flow -
Feed Flow Mismatch portion of this reactor trip in a similar manner as it is presented in Technical Specification Table 2.2-1. This change will have no impact on the probability of occurrence or the consequence of any accident previously evaluated.
The Bases change does not create the possibility of an accident or malfunction of a different type than any evaluated previously in the safety analysis report.
This Bases change corrects a setpoint value that does not reflect current plant conditions. The revised basis will express the Trip Setpoint for the Steam Flow - Feed Flow Mismatch portion of this reactor trip in a similar manner as it is presented in Technical Specification Table 2.2-1.
This change will make the bases consistent with the Technical Specification Page 4 of 5
i limit. Therefore it is concluded that no new or different kind of an accident or malfunction from any previously evaluated has been created.
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This Bases change does not result in a reduction of margin of safety as defined in the basis for any Technical Specifications.
The margin of safety is not reduced as a result of this Bases change since the change has no effect on any safety analysis assumptions.
As previously stated, the Steam Flow - Feed Flow Mismatch coincident with Steam Generator Low Level Reactor Trip is not credited in the Chapter 15 Safety Analysis. This trip function is intended to serve as a backup for the Steam Generator (SG) Low-Low Level Reactor Trip during an IEEE 279-1971 Protection and Control System interaction when the SG Low-Low Level Reactor Trip is not functional. Further, it has been demonstrated j
that this trip function still retains a positive margin of 2.69% of Flow with nom respect to the implied Design Basis Limit of < 50% of Flow Therefore, nom.
the proposed Bases changes do not reduce the margin of safety as it i
applies to this reactor trip function.
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