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| | document type = Legal-Affidavit, Letter, Report, Technical | | | document type = Legal-Affidavit, Letter, Report, Technical |
| | page count = 338 | | | page count = 338 |
| | project = TAC:MF7338, TAC:MF7337 | | | project = TAC:MF7337, TAC:MF7338 |
| | stage = Other | | | stage = Supplement |
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| F(x,y,z)=F(x,y,z)M(x,y,z)/(UMTMTTILT)F(x,y,z)=F(x,y,z)M(x,y,z)/(UMTMTTILT)F(x,y)=F(x,y)M(x,y)/(UMRTILT) | | F(x,y,z)=F(x,y,z)M(x,y,z)/(UMTMTTILT)F(x,y,z)=F(x,y,z)M(x,y,z)/(UMTMTTILT)F(x,y)=F(x,y)M(x,y)/(UMRTILT) |
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| (,,) F(x,y,z) F(x,y,z) NP(x,y,z)(,,)(,)F(x,y,z)F(x,y) Section 6.1.1 Change: -20-15-10-50510152025-40-35-30-25-20-15-10-50510152025303540LOCA FQ Margin (Limit -FQ)/LIMIT*100AFD (%)Figure 1Example LOCA FQ Margin Versus AFDTypicalSteady State AFDOperating RangeNegative HFPAFD LimitPostive HFPAFD Limit FSection 6.3.1 Change: FSection 6.4.1 Change: | | (,,) F(x,y,z) F(x,y,z) NP(x,y,z)(,,)(,)F(x,y,z)F(x,y) Section 6.1.1 Change: 15-10-50510152025-40-35-30-25-20-15-10-50510152025303540LOCA FQ Margin (Limit -FQ)/LIMIT*100AFD (%)Figure 1Example LOCA FQ Margin Versus AFDTypicalSteady State AFDOperating RangeNegative HFPAFD LimitPostive HFPAFD Limit FSection 6.3.1 Change: FSection 6.4.1 Change: |
| Section 6.4.2 Change: F(x,y,z)F(x,y)(,)F(x,y)F(x,y) RNP(x,y) | | Section 6.4.2 Change: F(x,y,z)F(x,y)(,)F(x,y)F(x,y) RNP(x,y) |
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| * Fz) that yields a MDNBR equal to the DNBR limit defines a single point along a MATP curve. An example MATP curve is shown in Figure 14 of Reference 1. The MATP limits are used to ensure the DNB design basis for Conditions I and II transients is satisfied by comparing calculated total peaking factors from a core nodal simulator (e.g. SIMULATE-3) against the MATP limits for a series of power distributions. Core power distributions are generated (by the core nodal simulator) as a function of rod position, burnup, xenon concentration, inlet temperature, and core power level. The method to generate these power distributions is described in Reference 4. DNBR margin is computed for each fuel assembly to verify positive margin exists for the allowable core power versus axial flux difference (AFD) limit used to establish the F(I) portion of the OTT trip function. The same methodology is used at a different state point condition to develop the core operational AFD limits. The operational AFD limits protect the core against the limiting Condition II DNB transient, typically the Loss of Flow transient. In summary, the Duke Energy MATP methodology is a generic methodology that relies on the NRC-approved VIPRE-01 model, fuel design, CHF correlation, and DNBR limit. It is applicable to any reactor or fuel type provided the NRC approvals for the unit specific methodology components are obtained. | | * Fz) that yields a MDNBR equal to the DNBR limit defines a single point along a MATP curve. An example MATP curve is shown in Figure 14 of Reference 1. The MATP limits are used to ensure the DNB design basis for Conditions I and II transients is satisfied by comparing calculated total peaking factors from a core nodal simulator (e.g. SIMULATE-3) against the MATP limits for a series of power distributions. Core power distributions are generated (by the core nodal simulator) as a function of rod position, burnup, xenon concentration, inlet temperature, and core power level. The method to generate these power distributions is described in Reference 4. DNBR margin is computed for each fuel assembly to verify positive margin exists for the allowable core power versus axial flux difference (AFD) limit used to establish the F(I) portion of the OTT trip function. The same methodology is used at a different state point condition to develop the core operational AFD limits. The operational AFD limits protect the core against the limiting Condition II DNB transient, typically the Loss of Flow transient. In summary, the Duke Energy MATP methodology is a generic methodology that relies on the NRC-approved VIPRE-01 model, fuel design, CHF correlation, and DNBR limit. It is applicable to any reactor or fuel type provided the NRC approvals for the unit specific methodology components are obtained. |
| Since NRC approval has been granted for the VIPRE-01 model, CHF correlation, and DNBR limits in Page 3 References 2 and 3, application of the Duke Energy Core DNB limit and MATP methodology to the Harris and Robinson plants is appropriate. 4.0 Conclusion The methodology described in DPC-NE-2004-P for development of limits to verify the required DNB design basis is satisfied is independent of the Westinghouse unit analyzed. The methodology does require an NRC approved: VIPRE-01 model for the unit being evaluated Critical Heat Flux (CHF) correlation for the fuel type Statistical DNBR limit for the fuel type Satisfying these three criteria allow use of the Duke Energy thermal-hydraulic analysis methodology, including the Core Safety Limit generation and MATP limit approach described in Reference 1 for plants using the Westinghouse reactor control and protection system. | | Since NRC approval has been granted for the VIPRE-01 model, CHF correlation, and DNBR limits in Page 3 References 2 and 3, application of the Duke Energy Core DNB limit and MATP methodology to the Harris and Robinson plants is appropriate. 4.0 Conclusion The methodology described in DPC-NE-2004-P for development of limits to verify the required DNB design basis is satisfied is independent of the Westinghouse unit analyzed. The methodology does require an NRC approved: VIPRE-01 model for the unit being evaluated Critical Heat Flux (CHF) correlation for the fuel type Statistical DNBR limit for the fuel type Satisfying these three criteria allow use of the Duke Energy thermal-hydraulic analysis methodology, including the Core Safety Limit generation and MATP limit approach described in Reference 1 for plants using the Westinghouse reactor control and protection system. |
| 5.0 References 1.DPC-NE-2004-PA, Rev. 2a, "Core Thermal-Hydraulic Methodology Using VIPRE-01" 2.DPC-NE-2005-P-A, Rev. 5, "Duke Energy Thermal-Hydraulic Statistical Core Design Methodology" 3.EMF-92-153(P)(A), Rev. 1, "HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel" 4.DPC-NE-2011-PA, Rev. 1a, "Nuclear Design Methodology Report For Core Operating Limits of Westinghouse Reactors" | | 5.0 References 1.DPC-NE-2004-PA, Rev. 2a, "Core Thermal-Hydraulic Methodology Using VIPRE-01" 2.DPC-NE-2005-P-A, Rev. 5, "Duke Energy Thermal-Hydraulic Statistical Core Design Methodology" 3.EMF-92-153(P)(A), Rev. 1, "HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel" 4.DPC-NE-2011-PA, Rev. 1a, "Nuclear Design Methodology Report For Core Operating Limits of Westinghouse Reactors"}} |
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MONTHYEARRA-15-0031, Application to Revise Technical Specifications for Methodology Report DPC-NE-1008-P, Revision 0, Nuclear Design Methodology Using CASMO-5/Simulate-3 for Westinghouse Reactors.2015-08-19019 August 2015
Application to Revise Technical Specifications for Methodology Report DPC-NE-1008-P, Revision 0, Nuclear Design Methodology Using CASMO-5/Simulate-3 for Westinghouse Reactors.
Project stage: Request
RA-16-0003, Application to Revise Technical Specifications to Adopt Methodology Reports DPC-NF-2010, Revision 3, Nuclear Physics Methodology for Reload Design, and DPC-NE-2011-P, Revision 2, Nuclear Design Methodology Report...2016-02-0303 February 2016
Application to Revise Technical Specifications to Adopt Methodology Reports DPC-NF-2010, Revision 3, Nuclear Physics Methodology for Reload Design, and DPC-NE-2011-P, Revision 2, Nuclear Design Methodology Report...
Project stage: Request
RA-16-0020, Withdrawal of License Amendment Request Regarding Methodology Reports DPC-NF-2010 and DPC-NE-2011-P2016-04-0707 April 2016
Withdrawal of License Amendment Request Regarding Methodology Reports DPC-NF-2010 and DPC-NE-2011-P
Project stage: Withdrawal
RA-16-0023, Supplemental Information for License Amendment Request Regarding Methodology Report DPC-NE-1008-P2016-05-0404 May 2016
Supplemental Information for License Amendment Request Regarding Methodology Report DPC-NE-1008-P
Project stage: Supplement
2016-02-03
[Table View] |
Shearon Harris, Unit 1 and H.B. Robinson, Unit 2 - Supplemental Information for License Amendment Request Regarding Methodology Report DPC-NE-1008-P| ML16125A420 |
| Person / Time |
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| Site: |
Harris, Robinson  |
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| Issue date: |
05/04/2016 |
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| From: |
Elnitsky J
Duke Energy Progress |
|---|
| To: |
Document Control Desk, Office of Nuclear Reactor Regulation |
|---|
| Shared Package |
|---|
| ML16125A444 |
List: |
|---|
| References |
|---|
| RA-16-0023, TAC MF7337, TAC MF7338 DPC-NE-1008, Rev. 0, DPC-NE-2011, Rev. 2, DPC-NF-2010, Rev. 3 |
| Download: ML16125A420 (338) |
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Category:Legal-Affidavit
MONTHYEARRA-23-0313, Baffle-Former Bolt (Bfb) Subsequent Inspection Interval Evaluation2023-12-14014 December 2023
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Response to Request for Additional Information Regarding Technical Specification Changes to Support Self-Performance of Core Reload Design and Safety Analyses
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Robinson Fuel Meth Pre-sub Meeting - Presentation Affidavit for Closed Pre-Submittal Meeting with Duke Energy Progress to Discuss Fuel Reload Design Methodology Reports and Proposed LAR Re H.B. Robinson and Shearon Harris Plants
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License Amendment Request for Main Steam Safety Valve Lift Setting Tolerance Change
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Application to Revise Technical Specifications to Adopt Methodology Report DPC-NE-3008-P, Revision 0, Thermal-Hydraulic Models for Transient Analysis.
RA-15-0041, Attachment 1, Affidavit and Attachment 3, Pre-Submittal Meeting Presentation Materials on DPC-NF-2010-A and DPC-NE-2011-P Methodologies (Redacted)2015-09-18018 September 2015
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Application to Revise Technical Specifications for Methodology Report DPC-NE-1008-P, Revision 0, Nuclear Design Methodology Using CASMO-5/Simulate-3 for Westinghouse Reactors.
RA-15-0004, Application to Revise Technical Specifications for Methodology Report DPC-NE-2005-P, Revision 5, Thermalhydraulic Statistical Core Design Methodology2015-03-0505 March 2015
Application to Revise Technical Specifications for Methodology Report DPC-NE-2005-P, Revision 5, Thermalhydraulic Statistical Core Design Methodology
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License Amendment Request for Revision to Technical Specification Core Operating Limits Report for Realistic Large Break LOCA Analysis
RA-11-008, Progress Energy - Evidence of Guarantee of Payment of Deferred Premiums2011-04-14014 April 2011
Progress Energy - Evidence of Guarantee of Payment of Deferred Premiums
RNP-RA/09-0054, Attachments VI & VII to Serial: RNP-RA/09-0054 - Request for Technical Specifications Change to Section 3.3.1 Reactor Protection System Instrumentation2009-06-19019 June 2009
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RNP-RA/07-0126, Request for Technical Specifications Change to Section 3.6.8 Isolation Valve Seal Water System2007-11-29029 November 2007
Request for Technical Specifications Change to Section 3.6.8 Isolation Valve Seal Water System
ML0721203582007-07-20020 July 2007
Supplemental Declarations of Robert L. Smith, Judy Hogan, Worth Glover, Jr., Tony A. Hackney, Elbert Green, Derry J. Smith, Jr. and Beverly Ann D'Aquanni
RNP-RA/05-0062, Response to Request for Additional Information Regarding Technical Specifications Change Request to Section 3.8.4 DC Sources - Operating2005-07-13013 July 2005
Response to Request for Additional Information Regarding Technical Specifications Change Request to Section 3.8.4 DC Sources - Operating
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Response to NRC Request for Additional Information Regarding Loss of Coolant Accident Alternative Source Term Dose Analysis
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Joint Appendix Volume III: Pages 1299 - 1556, Petition to Review a Final Decision of the Nrc. Affidavit of Eric A. Mccartney, Michael J. Devoe, Benjamin Wesley Morgan, and Deposition of Gordon R. Thompson Regarding Contention EC-6 Volmen 3
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Affidavit of Gareth W. Parry, Stephen F. Lavie, Robert L. Palla and Christopher Gratton in Support of NRC Staff Brief and Summary of Relevant Facts, Data and Arguments Upon Which the Staff Proposes to Rely at Oral Argument on Environmental
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Affidavit of Kenneth C. Heck in Support of NRC Staff'S Written Summary
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Transmitting Affidavit of J. A. Jones, Marked for Identification as Applicant Exhibit Pp, & Revisions to Applicant'S Proposed Findings of Fact & Conclusion of Law in Form of Supplemental Initial Decision (Construction Permits)
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NRC Staff Motion to Accept Transcript Corrections
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Applicant'S Response to Conservation Council of North Carolina and Wake Environment, Inc.'S Motion to Remand to Licensing Board for Further Hearings
2023-08-30
[Table view]
Category:Letter
MONTHYEARIR 05000261/20230042024-01-31031 January 2024
Integrated Inspection Report 05000261/2023004
IR 05000400/20230042024-01-30030 January 2024
Integrated Inspection Report 05000400/2023004
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Unit, No. 2 - Exemption from Select Requirements of 10 CFR Part 73 (EPID L-2023-LLE-0047 (Security Notifications, Reports, and Recordkeeping and Suspicious Activity Reporting))
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Revision to Reactor Vessel Material Surveillance Capsule Withdrawal Schedule
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Request for Withholding Information from Public Disclosure, H. B. Robinson Steam Electric Plant, Unit No. 2
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Notification of Inspection and Request for Information for NRC Problem Identification and Resolution Inspection
IR 05000261/20234202023-11-30030 November 2023
Security Baseline Inspection Report 05000261/2023420 (Cover Letter with Report)
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Fire Protection Team Inspection Report 05000261/2023010
ML23317A3462023-11-14014 November 2023
Duke Fleet - Correction Letter to License Amendment Nos. 312 & 340 Issuance of Amendments Regarding the Adoption of Technical Specifications Task Force Traveler TSTF-554, Revision 1
IR 05000400/20230032023-11-0909 November 2023
Integrated Inspection Report 05000400/2023003
IR 05000261/20230032023-11-0707 November 2023
Integrated Inspection Report 05000261 2023003 and 07200060 2023001
ML23226A0862023-10-12012 October 2023
Issuance of Amendment No. 277 Regarding Revision of TSs to Add High-High Steam Generator Level Function to Table 3.3.2-1 and Remove Obsolete Content from TSs 2.1.1.1 and 5.6.5.b
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Communication from C-10 Research & Education Foundation Regarding NextEra Common Emergency Fleet Plan License Amendment Request and Related Documents Subsequently Published
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Issuance of Amendment No. 199 Regarding Administrative Changes to the Renewed Facility Operating License and Technical Specifications
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Issuance of Alternative to Steam Generator Welds
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Issuance of Amendments Regarding the Adoption of Technical Specifications Task Force Traveler TSTF-554, Revision 1
IR 05000400/20230052023-08-23023 August 2023
Updated Inspection Plan for Shearon Harris Nuclear Power Plant, Unit 1 (Report 05000400/2023005)
ML23235A0552023-08-23023 August 2023
Notification of an Fire Protection Team Inspection (FPTI) (NRC Inspection Report 05000261/2023010) and Request for Information (RFI)
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RQ Inspection Notification Letter
IR 05000261/20230052023-08-21021 August 2023
Updated Inspection Plan for H.B. Robinson Steam Electric Plant (Report 05000261/2023005)
IR 05000261/20230022023-08-0707 August 2023
Integrated Inspection Report 05000261/2023002
IR 05000400/20234022023-07-26026 July 2023
Security Baseline Inspection Report 05000400/2023402
IR 05000400/20230022023-07-24024 July 2023
Integrated Inspection Report 05000400/2023002
IR 05000400/20234402023-07-17017 July 2023
Special Inspection Report 05000400/2023440 and Preliminary Greater than Green Finding and Apparent Violation Cover Letter
IR 05000261/20234022023-05-30030 May 2023
Cyber Security Inspection Report 05000261/2023402
ML23145A1602023-05-25025 May 2023
Submittal of Updated Final Safety Analysis Report (Revision No. 30), Independent Spent Fuel Storage Installation Safety Analysis Report (Revision No. 28), Technical Specifications Bases Revisions, Quality
IR 05000400/20243012023-05-15015 May 2023
Notification of Licensed Operator Initial Examination 05000400/2024301
IR 05000400/20230012023-05-10010 May 2023
Integrated Inspection Report 05000400 2023001
IR 05000400/20234042023-05-0404 May 2023
Cyber Security Inspection Report 05000400/2023404
IR 05000261/20230012023-05-0202 May 2023
Integrated Inspection Report 05000261/2023001
ML23118A0762023-05-0101 May 2023
Approval for Use of Specific Provision of a Later Edition of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI
ML23118A1392023-04-28028 April 2023
Submittal of Updated Final Safety Analysis Report (Amendment 65), Technical Specification Bases Revision, Report of Changes Pursuant to 10 CFR 50.59 and Summary of Commitment Changes
IR 05000400/20234032023-04-0505 April 2023
Security Baseline Inspection Report 05000400/2023403
IR 05000400/20230102023-03-15015 March 2023
Comprehensive Engineering Team Inspection (CETI) Inspection Report 05000400/2023010
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William States Lee III 1 and 2 - Issuance of Amendments Regarding the Relocation of the Emergency Operations Facility
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Annual Assessment Letter for Shearon Harris Nuclear Power Plant - NRC Inspection Report 05000400/2022006
IR 05000261/20220062023-03-0101 March 2023
Annual Assessment Letter for H.B. Robinson Steam Electric Plant, Unit 2 (Report 05000261/2022006)
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Information Request for the Cyber-Security Baseline Inspection, Notification to Perform Inspection 05000261/2023402
ML23041A2272023-02-13013 February 2023
2022 Q4 Robinson_Workflow Final
ML23033A5272023-02-0808 February 2023
Correction of Typographical Errors Incurred During Issuance of License Amendment No. 196
IR 05000400/20220042023-02-0707 February 2023
Integrated Inspection Report 05000400/2022004
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Notification of Target Set Inspection and Request for Information (NRC Inspection Report 05000400/2023403)
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Issuance of Amendment No. 274 Regarding Revision of Technical Specification 3.8.1 Surveillance Requirement 3.8.1.16
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Issuance of Amendment No. 273 Regarding Adoption of TSTF-577, Revised Frequencies for Steam Generator Tube Inspections, Revision 1
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2023 Requalification Program Inspection - H.B. Robinson Nuclear Plant
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301, NRC Operator License Examination Report & Cover Letter Merged
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McGuire Nuclear Station and Shearon Harris Nuclear Power Plant Authorization of RA-19-0352 Regarding Use of Alternative for RPV Head Closure Stud Examinations
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Issuance of Amendments Regarding the Adoption of Technical Specifications Task Force Traveler TSTF-541, Rev. 2
IR 05000400/20220032022-11-10010 November 2022
Integrated Inspection Report 05000400/2022003
IR 05000261/20220032022-11-0303 November 2022
Integrated Inspection Report 05000261 2022003
2024-01-08
[Table view]
Category:Report
MONTHYEARRA-23-0313, Baffle-Former Bolt (Bfb) Subsequent Inspection Interval Evaluation2023-12-14014 December 2023
Baffle-Former Bolt (Bfb) Subsequent Inspection Interval Evaluation
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Revision to Reactor Vessel Surveillance Capsule Withdrawal Schedule
RA-23-0120, Supplemental Information Regarding Addition of Feedwater Isolation on Steam Generator Level High-High to Technical Specification 3.3.22023-05-31031 May 2023
Supplemental Information Regarding Addition of Feedwater Isolation on Steam Generator Level High-High to Technical Specification 3.3.2
RA-23-0080, Refueling Outage 32 Steam Generator Tube Inspection Report Supplement Pursuant to Technical Specifications Task Force (TSTF) Traveler TSTF-577, Revision 1, Revised Frequencies for Steam Generator Tube2023-04-0505 April 2023
Refueling Outage 32 Steam Generator Tube Inspection Report Supplement Pursuant to Technical Specifications Task Force (TSTF) Traveler TSTF-577, Revision 1, Revised Frequencies for Steam Generator Tube
RA-22-0302, Technical Specifications (TS) Section 5.6.6, Post Accident Monitoring Instrumentation Report for Inoperable Containment Sump Water Level (Wide Range)2022-11-0101 November 2022
Technical Specifications (TS) Section 5.6.6, Post Accident Monitoring Instrumentation Report for Inoperable Containment Sump Water Level (Wide Range)
RA-22-0239, Technical Specifications Section 5.6.6. Post Accident Monitoring Instrumentation Report for Inoperable Safety Valve Position (Primary)2022-08-0909 August 2022
Technical Specifications Section 5.6.6. Post Accident Monitoring Instrumentation Report for Inoperable Safety Valve Position (Primary)
RA-22-0017, Supplement to Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections2022-01-0606 January 2022
Supplement to Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections
RA-21-0312, Technical Specifications Section 5.6.6 Post Accident Monitoring Instrumentation Report for Inoperable Containment Sump Water Level (Wide Range)2021-11-22022 November 2021
Technical Specifications Section 5.6.6 Post Accident Monitoring Instrumentation Report for Inoperable Containment Sump Water Level (Wide Range)
IR 05000261/20210052021-08-25025 August 2021
Updated Inspection Plan for H. B. Robinson Steam Electric Plant, Unit 2 (Report 05000261/2021005)
RA-21-0097, Notification of Permit Revision Request Regarding Copper and Zinc Limits2021-03-15015 March 2021
Notification of Permit Revision Request Regarding Copper and Zinc Limits
RA-20-0381, CFR 50.54( Q) Screening Evaluation Form for Revisions2020-12-0808 December 2020
CFR 50.54( Q) Screening Evaluation Form for Revisions
RA-20-0335, Response to Request for Additional Information Regarding License Amendment Request to Reduce the Minimum Required Reactor Coolant System Flow Rate and Update the List of Analytical Methods Used in .2020-11-24024 November 2020
Response to Request for Additional Information Regarding License Amendment Request to Reduce the Minimum Required Reactor Coolant System Flow Rate and Update the List of Analytical Methods Used in .
RA-20-0032, License Amendment Request to Reduce the Minimum Required Reactor Coolant System Flow Rate and Update the List of Analytical Methods Used in the Determination of Core Operating Limits2020-03-0606 March 2020
License Amendment Request to Reduce the Minimum Required Reactor Coolant System Flow Rate and Update the List of Analytical Methods Used in the Determination of Core Operating Limits
RA-19-0403, Submittal of the Summary Technical Report for the Reactor Pressure Vessel Surveillance Program Capsule Z2019-10-23023 October 2019
Submittal of the Summary Technical Report for the Reactor Pressure Vessel Surveillance Program Capsule Z
RA-19-0223, Annual Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models Pursuant to 10CFR 50.462019-05-30030 May 2019
Annual Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models Pursuant to 10CFR 50.46
RA-18-0179, 90-Day Special Report2018-10-0404 October 2018
90-Day Special Report
ML18230A7562018-08-18018 August 2018
Fault Investigation, Responses to Mr. W. R. Butler'S Letter of May 16, 1975
ML18230A7592018-08-18018 August 2018
Fault Investigation, Progress Report, Volume 1 of 2
HNP-18-023, Report of Changes Pursuant to 10 CFR 50.59 and Summary of Commitment Changes2018-05-0202 May 2018
Report of Changes Pursuant to 10 CFR 50.59 and Summary of Commitment Changes
RNP-RA/18-0024, Report of Changes Pursuant to 10 CFR 50.59(d)(2)2018-04-0202 April 2018
Report of Changes Pursuant to 10 CFR 50.59(d)(2)
RA-17-0040, Providing Methodology Reports DPC-NE-1008-P, Revision 0, DPC-NF-2010, Revision 3, and DPC-NE-2011-P, Revision 2. (Non-Proprietary Versions)2017-08-15015 August 2017
Providing Methodology Reports DPC-NE-1008-P, Revision 0, DPC-NF-2010, Revision 3, and DPC-NE-2011-P, Revision 2. (Non-Proprietary Versions)
HNP-17-040, Snubber Program Plan2017-06-0101 June 2017
Snubber Program Plan
HNP-17-006, Submittal of Cycle 20 Activity Report2017-01-30030 January 2017
Submittal of Cycle 20 Activity Report
HNP-16-083, Mitigating Strategies Assessment Report for Flooding Hazard Information2016-12-21021 December 2016
Mitigating Strategies Assessment Report for Flooding Hazard Information
ML16281A5102016-12-15015 December 2016
Staff Assessment of the Reactor Vessel Internals Aging Management Program Plans
RNP-RA/16-0087, Technical Specifications Section 5.6.6 Post Accident Monitoring Instrumentation Report for Inoperable High Range Containment Area Radiation Monitors2016-10-31031 October 2016
Technical Specifications Section 5.6.6 Post Accident Monitoring Instrumentation Report for Inoperable High Range Containment Area Radiation Monitors
RNP-RA/16-0078, Technical Specifications Section 5.6.6, Post Accident Monitoring Instrumentation Report for Inoperable High Range Containment Area Radiation Monitors2016-10-0505 October 2016
Technical Specifications Section 5.6.6, Post Accident Monitoring Instrumentation Report for Inoperable High Range Containment Area Radiation Monitors
ML16280A2002016-10-0505 October 2016
Response to Request for Additional Information Related to the Pressurized Water Reactor Internals Program Plan for Aging Management of Reactor Internals
RA-16-0024, Supplemental Information for License Amendment Request Regarding Methodology Report DPC-NE-3008-P2016-10-0303 October 2016
Supplemental Information for License Amendment Request Regarding Methodology Report DPC-NE-3008-P
RNP-RA/16-0073, Response to Request for Additional Information Regarding License Amendment Request to Revise Reactor Coolant System Pressure and Temperature Limits2016-09-14014 September 2016
Response to Request for Additional Information Regarding License Amendment Request to Revise Reactor Coolant System Pressure and Temperature Limits
RNP-RA/16-0038, Transition Report, Revision 1, Transition to National Fire Protection Association Standard 805, with Attachments a, I, J, L, M, and V. Pages 22-1742016-05-25025 May 2016
Transition Report, Revision 1, Transition to National Fire Protection Association Standard 805, with Attachments a, I, J, L, M, and V. Pages 22-174
RA-16-0023, Supplemental Information for License Amendment Request Regarding Methodology Report DPC-NE-1008-P2016-05-0404 May 2016
Supplemental Information for License Amendment Request Regarding Methodology Report DPC-NE-1008-P
ML15253A4102016-03-11011 March 2016
Enclosure 2 Screening Analysis Report
HNP-16-017, Transmittal of Summary of a 10 CFR 50.54(q) Evaluation2016-02-29029 February 2016
Transmittal of Summary of a 10 CFR 50.54(q) Evaluation
RA-15-0042, Application to Revise Technical Specifications to Adopt Methodology Report DPC-NE-3008-P, Revision 0, Thermal-Hydraulic Models for Transient Analysis.2015-11-19019 November 2015
Application to Revise Technical Specifications to Adopt Methodology Report DPC-NE-3008-P, Revision 0, Thermal-Hydraulic Models for Transient Analysis.
RA-15-0047, Annual Report of Changes Pursuant to 10 CFR 50.462015-11-17017 November 2015
Annual Report of Changes Pursuant to 10 CFR 50.46
ML15301A5572015-11-0202 November 2015
Supplement to Staff Assessment of Response to 10 CFR 50.54(f) Information Request Flood Causing Mechanisms Reevaluations
ML15280A1992015-10-19019 October 2015
Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50 Section 50.54(f) Seismic Hazard Reevaluations for Recommendation 2.1 of the Near-Term Task Force Review
RA-15-0031, Application to Revise Technical Specifications for Methodology Report DPC-NE-1008-P, Revision 0, Nuclear Design Methodology Using CASMO-5/Simulate-3 for Westinghouse Reactors.2015-08-19019 August 2015
Application to Revise Technical Specifications for Methodology Report DPC-NE-1008-P, Revision 0, Nuclear Design Methodology Using CASMO-5/Simulate-3 for Westinghouse Reactors.
RNP-RA/15-0053, Compliance Letter and Final Integrated Plan in Response to the March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order No. EA-12-04)2015-08-19019 August 2015
Compliance Letter and Final Integrated Plan in Response to the March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order No. EA-12-04)
RA-15-0022, Annual Report of Changes Pursuant to 10 CFR 50.462015-05-14014 May 2015
Annual Report of Changes Pursuant to 10 CFR 50.46
ML15126A0892015-04-30030 April 2015
EAL Bases
HNP-15-027, Flood Hazard Reevaluation Report, Revision 12015-04-0101 April 2015
Flood Hazard Reevaluation Report, Revision 1
RNP-RA/15-0018, Technical Specifications (TS) Section 5.6.6, Post Accident Monitoring (PAM) Instrumentation Report, 14-Day Report for the Inoperability of the Power Range Neutron Flux Channel2015-02-26026 February 2015
Technical Specifications (TS) Section 5.6.6, Post Accident Monitoring (PAM) Instrumentation Report, 14-Day Report for the Inoperability of the Power Range Neutron Flux Channel
RNP-RA/14-0037, Response to NRC Request for Additional Information Regarding License Amendment Request to Modify Technical Specification 3.4.12, Low Temperature Overpressure Protection System2014-04-0808 April 2014
Response to NRC Request for Additional Information Regarding License Amendment Request to Modify Technical Specification 3.4.12, Low Temperature Overpressure Protection System
HNP-14-035, Seismic Hazard and Screening Report (CEUS Sites), Response to NRC 10 CFR 50.54(f) Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.32014-03-27027 March 2014
Seismic Hazard and Screening Report (CEUS Sites), Response to NRC 10 CFR 50.54(f) Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3
RNP-RA/14-0012, Flood Hazard Reevaluation Report, Response to NRC 10 CFR 50.54(f) Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3 and 9.3....2014-03-12012 March 2014
Flood Hazard Reevaluation Report, Response to NRC 10 CFR 50.54(f) Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3 and 9.3....
RNP-RA/14-0011, Revision to Response to Recommendation 2.3 Seismic Walkdown of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident2014-02-27027 February 2014
Revision to Response to Recommendation 2.3 Seismic Walkdown of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident
ML13365A2912014-02-19019 February 2014
Interim Staff Evaluation Regarding Overall Integrated Plan in Response to Order EA-12-049 (Mitigation Strategies)
ML13364A2142014-02-12012 February 2014
Interim Staff Evaluation Relating to Overall Integrated Plan in Response to Order EA-12-049 (Mitigation Strategies)
2023-07-12
[Table view]
Category:Technical
MONTHYEARRA-23-0313, Baffle-Former Bolt (Bfb) Subsequent Inspection Interval Evaluation2023-12-14014 December 2023
Baffle-Former Bolt (Bfb) Subsequent Inspection Interval Evaluation
RA-23-0141, Revision to Reactor Vessel Surveillance Capsule Withdrawal Schedule2023-07-12012 July 2023
Revision to Reactor Vessel Surveillance Capsule Withdrawal Schedule
RA-23-0120, Supplemental Information Regarding Addition of Feedwater Isolation on Steam Generator Level High-High to Technical Specification 3.3.22023-05-31031 May 2023
Supplemental Information Regarding Addition of Feedwater Isolation on Steam Generator Level High-High to Technical Specification 3.3.2
RA-23-0080, Refueling Outage 32 Steam Generator Tube Inspection Report Supplement Pursuant to Technical Specifications Task Force (TSTF) Traveler TSTF-577, Revision 1, Revised Frequencies for Steam Generator Tube2023-04-0505 April 2023
Refueling Outage 32 Steam Generator Tube Inspection Report Supplement Pursuant to Technical Specifications Task Force (TSTF) Traveler TSTF-577, Revision 1, Revised Frequencies for Steam Generator Tube
RA-22-0302, Technical Specifications (TS) Section 5.6.6, Post Accident Monitoring Instrumentation Report for Inoperable Containment Sump Water Level (Wide Range)2022-11-0101 November 2022
Technical Specifications (TS) Section 5.6.6, Post Accident Monitoring Instrumentation Report for Inoperable Containment Sump Water Level (Wide Range)
RA-22-0239, Technical Specifications Section 5.6.6. Post Accident Monitoring Instrumentation Report for Inoperable Safety Valve Position (Primary)2022-08-0909 August 2022
Technical Specifications Section 5.6.6. Post Accident Monitoring Instrumentation Report for Inoperable Safety Valve Position (Primary)
RA-22-0017, Supplement to Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections2022-01-0606 January 2022
Supplement to Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections
RA-21-0097, Notification of Permit Revision Request Regarding Copper and Zinc Limits2021-03-15015 March 2021
Notification of Permit Revision Request Regarding Copper and Zinc Limits
RA-20-0381, CFR 50.54( Q) Screening Evaluation Form for Revisions2020-12-0808 December 2020
CFR 50.54( Q) Screening Evaluation Form for Revisions
RA-20-0335, Response to Request for Additional Information Regarding License Amendment Request to Reduce the Minimum Required Reactor Coolant System Flow Rate and Update the List of Analytical Methods Used in .2020-11-24024 November 2020
Response to Request for Additional Information Regarding License Amendment Request to Reduce the Minimum Required Reactor Coolant System Flow Rate and Update the List of Analytical Methods Used in .
RA-20-0032, License Amendment Request to Reduce the Minimum Required Reactor Coolant System Flow Rate and Update the List of Analytical Methods Used in the Determination of Core Operating Limits2020-03-0606 March 2020
License Amendment Request to Reduce the Minimum Required Reactor Coolant System Flow Rate and Update the List of Analytical Methods Used in the Determination of Core Operating Limits
RA-19-0403, Submittal of the Summary Technical Report for the Reactor Pressure Vessel Surveillance Program Capsule Z2019-10-23023 October 2019
Submittal of the Summary Technical Report for the Reactor Pressure Vessel Surveillance Program Capsule Z
RA-18-0179, 90-Day Special Report2018-10-0404 October 2018
90-Day Special Report
RA-17-0040, Providing Methodology Reports DPC-NE-1008-P, Revision 0, DPC-NF-2010, Revision 3, and DPC-NE-2011-P, Revision 2. (Non-Proprietary Versions)2017-08-15015 August 2017
Providing Methodology Reports DPC-NE-1008-P, Revision 0, DPC-NF-2010, Revision 3, and DPC-NE-2011-P, Revision 2. (Non-Proprietary Versions)
HNP-17-040, Snubber Program Plan2017-06-0101 June 2017
Snubber Program Plan
ML16280A2002016-10-0505 October 2016
Response to Request for Additional Information Related to the Pressurized Water Reactor Internals Program Plan for Aging Management of Reactor Internals
RNP-RA/16-0073, Response to Request for Additional Information Regarding License Amendment Request to Revise Reactor Coolant System Pressure and Temperature Limits2016-09-14014 September 2016
Response to Request for Additional Information Regarding License Amendment Request to Revise Reactor Coolant System Pressure and Temperature Limits
RA-16-0023, Supplemental Information for License Amendment Request Regarding Methodology Report DPC-NE-1008-P2016-05-0404 May 2016
Supplemental Information for License Amendment Request Regarding Methodology Report DPC-NE-1008-P
RA-15-0042, Application to Revise Technical Specifications to Adopt Methodology Report DPC-NE-3008-P, Revision 0, Thermal-Hydraulic Models for Transient Analysis.2015-11-19019 November 2015
Application to Revise Technical Specifications to Adopt Methodology Report DPC-NE-3008-P, Revision 0, Thermal-Hydraulic Models for Transient Analysis.
RA-15-0031, Application to Revise Technical Specifications for Methodology Report DPC-NE-1008-P, Revision 0, Nuclear Design Methodology Using CASMO-5/Simulate-3 for Westinghouse Reactors.2015-08-19019 August 2015
Application to Revise Technical Specifications for Methodology Report DPC-NE-1008-P, Revision 0, Nuclear Design Methodology Using CASMO-5/Simulate-3 for Westinghouse Reactors.
RNP-RA/14-0011, Revision to Response to Recommendation 2.3 Seismic Walkdown of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident2014-02-27027 February 2014
Revision to Response to Recommendation 2.3 Seismic Walkdown of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident
ML13365A2912014-02-19019 February 2014
Interim Staff Evaluation Regarding Overall Integrated Plan in Response to Order EA-12-049 (Mitigation Strategies)
ML13364A2142014-02-12012 February 2014
Interim Staff Evaluation Relating to Overall Integrated Plan in Response to Order EA-12-049 (Mitigation Strategies)
ML14030A1022014-01-29029 January 2014
Mega-Tech Services, LLC, Technical Evaluation Report Regarding the Overall Integrated Plan for Shearon Harris Nuclear Power Plant, Unit 1, TAC No.: MF0874
ML14027A0632014-01-24024 January 2014
Mega-Tech Services, LLC, Technical Evaluation Report Regarding the Overall Integrated Plan for H. B. Robinson Steam Electric Plant, Unit 2, TAC No.: MF0720
ML13267A2122013-09-30030 September 2013
Enclosure 1, Transition Report - NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001 Edition
ML13270A1762013-09-24024 September 2013
Redacted - Office of Nuclear Reactor Regulation (NRR) Reactor Systems Branch (Srxb) Support of Region II Inspection of H. B. Robinson Treatment of Voids in Systems That Are Important to Safety
ML12331A1752012-11-26026 November 2012
Draft Bypass Fiber Quantity Test Plan
ML12278A3992012-08-31031 August 2012
WCAP-17077-NP, Rev 1, PWR Vessel Internals Program Plan for Aging Management of Reactor Internals at Robinson Nuclear Plant
ML1210707162012-05-17017 May 2012
Letter Report on the Evaluation of Cables from the HEAF Fire Event at the H.B. Robinson Steam Electric Plant
ML12067A1802012-02-29029 February 2012
ANP-3011Q1(NP), Revision 000, Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis. Enclosure 4 Response to Request for Additional Information (Non-Proprietary)
ML12068A1332012-02-23023 February 2012
Calculation RNP-M/MECH-1815, Revision 1, Evaluation of Emergency Diesel Generator Starting Capability at 150 PSIG
ML1008904242010-03-0202 March 2010
NAI-1478-001, Revision 1, HNP CSAT Volume, Flow and Naoh Concentration Range Revisions.
ML1008905932010-01-31031 January 2010
Application for Revision to Technical Specification Core Operating Limits Report References, ANP-2853(NP), Rev. 0, Realistic Large Break LOCA Summary Report.
ML1002508582010-01-18018 January 2010
Holtec Report, No. HI-2043321, Rev. 6, Critically Safety Analyses of BWR Fuel Without Credit for Boraflex in the Racks at the Harris Nuclear Power Station.
ML1003402842010-01-12012 January 2010
NAI-1478-001, Revision 0, HNP CSAT Volume, Flow and Naoh Concentration Range Revisions
RNP-RA/09-0081, WCAP-17077-NP, Revision 0, PWR Vessel Internals Program Plan for Aging Management of Reactor Internals at Robinson Nuclear Plant.2009-07-31031 July 2009
WCAP-17077-NP, Revision 0, PWR Vessel Internals Program Plan for Aging Management of Reactor Internals at Robinson Nuclear Plant.
RNP-RA/09-0054, Attachments VI & VII to Serial: RNP-RA/09-0054 - Request for Technical Specifications Change to Section 3.3.1 Reactor Protection System Instrumentation2009-06-19019 June 2009
Attachments VI & VII to Serial: RNP-RA/09-0054 - Request for Technical Specifications Change to Section 3.3.1 Reactor Protection System Instrumentation
ML0921500542009-05-31031 May 2009
ANP-2693(NP), Revision 0, Loss of Forced Reactor Coolant Flow Analysis for Harris Nuclear Plant, Unit 1, Enclosure 4 to Serial: HNP-09-068
ML0821900132008-08-0707 August 2008
Monthly Operating Reports Second Quarter 2008
HNP-06-128, Engineering Report on: Evaluation of Charging-Safety Injection Pump Motor Bearing Temperatures at Progress Energy Harris Nuclear Plant2006-10-12012 October 2006
Engineering Report on: Evaluation of Charging-Safety Injection Pump Motor Bearing Temperatures at Progress Energy Harris Nuclear Plant
ML0632802412006-10-0505 October 2006
Shearon Harris, Letter Report NAI-1282-001, Gothic Charging Pump Room Heatup Analysis for Shearon Harris.
RNP-RA/06-0081, Steam Generator Alternate Repair Criteria for Tube Portion within the Tubesheet, WCAP-16627-NP2006-08-31031 August 2006
Steam Generator Alternate Repair Criteria for Tube Portion within the Tubesheet, WCAP-16627-NP
RNP-RA/06-0027, ANP-2512(NP), Loss of Forced Reactor Coolant Flow Analysis.2006-03-31031 March 2006
ANP-2512(NP), Loss of Forced Reactor Coolant Flow Analysis.
ML0525105042005-08-31031 August 2005
Framatome Anp, Inc 77-5069740-NP-00, Shearon Harris Criticality Evaluation.
ML0507004082004-02-20020 February 2004
EMF-3030(NP), Revision 0, Robinson Nuclear Plant, Realistic Large Break LOCA Analysis, February 2004, Non-Proprietary Version
RNP-RA/03-0075, Technical Basis for RPV Head CRDM Nozzle Inspection Interval H.B. Robinson Steam Electric Plant, Unit No. 2, References 9-1 Through E-62003-07-31031 July 2003
Technical Basis for RPV Head CRDM Nozzle Inspection Interval H.B. Robinson Steam Electric Plant, Unit No. 2, References 9-1 Through E-6
RNP-RA/03-0031, Response to Request for Additional Information Re Application for Renewal of Operating License, Attachment III, Pages 356 - 5042003-04-28028 April 2003
Response to Request for Additional Information Re Application for Renewal of Operating License, Attachment III, Pages 356 - 504
ML0311207052003-04-18018 April 2003
Review of 90-day Steam Generator Tube Inservice Inspection Report for a Refueling Outage in 2001
ML0305202692003-02-15015 February 2003
Follow-up Report, Reference Event #39516
2023-07-12
[Table view] |
Text
JOHN ELNITSKY Senior Vice President Nuclear Engineering 526 South Church Street, EC-07H Charlotte, NC 28202 704-382-4371 John.Elnitsky@duke-energy.com
SUBJECT:
SUPPLEMENTAL INFORMATION FOR LICENSE AMENDMENT REQUEST REGARDING METHODOLOGY REPORT DPC-NE-1008-P
REFERENCES:
Application to Revise Technical Specifications for Methodology Report DPC-NE-1008-P Revision 0, "Nuclear Design Methodology Using CASMO-5/SIMULATE-3 for Westinghouse Reactors"Application to Revise Technical Specifications to Adopt Methodology Reports DPC-NF-2010 Revision 3 "Nuclear Physics Methodology for Reload Design" and DPC-NE-2011-P Revision 2, "Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors"Withdrawal of License Amendment Request Regarding Methodology Reports DPC-NF-2010 and DPC-NE-2011-PShearon Harris Nuclear Power Plant, Unit No. 1 and H. B. Robinson Steam Electric Plant, Unit 2 - Withdrawal of Requested Licensing Action Regarding Duke Energy Progress, Inc., Application to Revise Technical Specifications to Adopt Methodology Reports Submitted to NRC for Acceptance Review (CAC Nos. MF7337 AND MF7338)
Attachment 1 Affidavit of John Elnitsky
Nuclear Design Methodology Using CASMO-5/SIMULATE-3 for Westinghouse ReactorsNuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors
Nuclear Design Methodology Using CASMO-5/SIMULATE-3 for Westinghouse ReactorsNuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors
EVALUATION OF THE PROPOSED CHANGE
McGuire Nuclear Station, Units 1 and 2 Issuance of Amendments Regarding Revision 1 to DPC-NE-1005-P, Nuclear Design Methodology Using CASMO-4/SIMULATE-3 MOX (TAC Nos. MD7409 and MD7410)Catawba Nuclear Station, Units 1 and 2 Issuance of Amendments Regarding Revision 1 to DPC-NE-1005-P, Nuclear Design Methodology Using CASMO-4/SIMULATE-3 MOX (TAC Nos. MD7407 and MD7408)Oconee Nuclear Station, Units 1, 2, and 3, Issuance of Amendments Regarding the Use of CASMO-4/SIMULATE-3 Methodology for Reactor Cores Containing Gadolinia Bearing Fuel (TAC Nos. ME4646, ME4647, and ME4648)Topical Report on Physics Methodology for Reloads: McGuire and Catawba Nuclear Station, McGuire Nuclear Station, Units 1 and 2 Re: Issuance of Amendments (TAC Nos. MB3222 and MB3223)Catawba Nuclear Station, Units 1 and 2 Re: Issuance of Amendments (TAC Nos. MB3343 and MB3344)
Catawba Nuclear Station, Units 1 and 2 and McGuire Nuclear Station, Units 1 and 2, Re: Topical Report DPC-NF-2010, Nuclear Physics Methodology for Reload Design, Revision 2Acceptance for Referencing of Topical Report DPC-NE-2011-P, "Duke Power Company Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors
Attachment 3 Proposed Technical Specification Changes (Mark-up)
Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.2 Annual Radiological Environmental Operating Report (continued) In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible. 5.6.3 Radioactive Effluent Release Report The Radioactive Effluent Release Report covering the operation of the unit shall be submitted in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.B.1. 5.6.4 DELETED 5.6.5 CORE OPERATING LIMITS REPORT (COLR) a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following: HBRSEP Unit No. 2 1. Shutdown Margin (SOM) for Specification 3.1.1; 2. Moderator Temperature Coefficient limits for Specification 3.1.3; 3. Shutdown Bank Insertion Limits for Specification 3.1.5; 4. Control Bank Insertion Limits for Specification 3.1.6; 5. Heat Flux Hot Channel Factor (Fa(Z)) limit for Specification 3.2.1; 6. Nuclear Enthalpy Rise Hot Channel Factor limit for Specification 3.2.2; (continued) 5.0-24 Amendment No. 212 No changes to this page. Included for information only. Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued) 7. Axial Flux Difference (AFD) limits for Specification 3.2.3; and 8. Boron Concentration limit for Specification 3.9.1. b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC. The approved version shall be identified in the COLR. These methods are those specifically described in the following documents: HBRSEP Unit No. 2 1. Deleted 2. XN-NF-84-73(P), "Exxon Nuclear Methodology for Pressurized Water Reactors: Analysis of Chapter 15 Events," approved version as specified in the COLR. 3. XN-NF-82-21 (A), "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations," approved version as specified in the COLR. 4. Deleted 5. XN-75-32(A), "Computational Procedure for Evaluating Rod Bow," approved version as specified in the COLR. 6. Deleted. 7. Deleted 8. XN-NF-78-44(A), "Generic Control Rod Ejection Analysis," approved version as specified in the COLR 9. XN-NF-621(A), "XNB Critical Heat Flux Correlation," approved version as specified in the COLR. 10. Deleted 11. XN-NF-82-06(A), "Qualification of Exxon Nuclear Fuel for Extended Burnup," approved version as specified in the COLR. 12. Deleted 13. Deleted. (continued) 5.0-25 Amendment No. 227 No changes to this page. Included for information only. Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued) HBRSEP Unit No. 2 14. Deleted 15. Deleted 16. ANF-88-054(P), "PDC-3: Advanced Nuclear Fuels Corporation Power Distribution Control for Pressurized Water Reactors and Application of PDC-3 to H.B. Robinson Unit 2," approved version as specified in the COLR. 17. ANF-88-133 (P)(A), "Qualification of Advanced Nuclear Fuels' PWR Design Methodology for Rod Burnups of 62 Gwd/MTU," approved version as specified in the COLR. 18. ANF-89-151(A), "ANF-RELAP Methodology for Pressurized Water Reactors: Analysis of Non-LOCA Chapter 15 Events," approved version as specified in the COLR. 19. EMF-92-081 (A), "Statistical Setpoint/Transient Methodology for Westinghouse Type Reactors," approved version as specified in the COLR. 20. EMF-92-153(P)(A), "HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel," approved version as specified in the COLR. 21. XN-NF-85-92(P)(A), "Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results," approved version as specified in the COLR. 22. EMF-96-029(P)(A), "Reactor Analysis System for PWRs," approved version as specified in the COLR. 23. EMF-92-116, "Generic Mechanical Design Criteria for PWR Fuel Designs," approved version as specified in the COLR. 24. EMF-2103(P)(A), "Realistic Large Break LOCA Methodology for Pressurized Water Reactors," approved version as specified in the COLR. (continued) 5.0-26 Amendment No. 227 Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT <COLR) (continued) Insert 1 (see next page) 25. EMF-2310(P)(A), "SRP Chapter 15 Non-LOCA Methodology for Pressurized Water approved version as specified in the COLR. 26. BAW-10240(P)(A), "Incorporation of M5 Properties in Framatome ANP Approved Methods," approved version as specified in the COLR. 27. EMF-2328(P)(A), "PWR Small Break LOCA Evaluation Model, S-RELAP5 Based,* approved version as specified in the COLR. DPC-NE-2005-P-A, "Thermal-Hydraulic Statistical Core Design Methodology," approved version as specified in the COLR. c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SOM, transient analysis limits, and accident analysis limits) of the safety analysis are met. d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC. 5.6.6 Post Accident Monitoring <PAM) Instrumentation Report When a report is required by Condition B or G of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status, (continued) HBRSEP Unit No. 2 5.0-27 Amendment No. 2 4 4
No changes to this page. Included for information only. ADMINISTRATIVE CONTROLS 6.9.l.6 CORE OPERATING LIMITS 6.9.1.6.l Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT CCOLR). plant prncedure PLP-106. prior to each reload cycle. or prior to any remaining portion of a reload cycle. for the fo 11 owing: a. b. c. d. e. SHUTDOWN MARGIN limits for Specification 3/4.1.1.2. Moderator Temperature Coefficient Positive and Negative Limits and 300 ppm surveillance limit for Specification 3/4.1.1.3. Shutdown Bank Insertion Limits for Specification 3/4.1.3.5. Control Bank Insertion Limits for Specification 3/4.1.3.6. Axial Flux Difference Limits for Specification 3/4.2.1. f. Heat Flux Hot Channel Factor. . K(Z). and VCZ) for Specification 3/4.2.2. g. Enthalpy Rise Hot Channel Factor. . and Power Factor Multiplier. PF6H for Specification 3/4.2.3. h. Boron Concentration for Specification 3/4.9.1. 6.9.1.6.2 The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC at the time the reload analyses are Eerformed. and the approved revision number shall be identified in the CO R. a. XN-75-27(P)(A). "Exxon Nuclear Neutronics Design Methods for Pressurized Water Reactors." approved version as specified in the COLR. (Methodology for Sgecification 3.1.1.2 -SHUTDOWN MARGIN -MODES 3. 4 and 5. 3.1.1.3 -Moderator Temperature Coefficient. 3.1.3.5 -Shutdown Bank Insertion Limits. 3.1.3.6 -Control Bank Insertion Limits. 3.2.1 -Axial Flux Difference. 3.2.2 -Heat Flux Hot Channel Factor. 3.2.3 -Nuclear Enthalpy Rise Hot Channel Factor. and 3.9.1 -Boron Concentration). b. ANF-89-151(P)(A). "ANF-RELAP Methodology for Pressurized Water Reactors: Analysis of Non-LOCA Chapter 15 Events." approved version as specified in the COLR. (Methodology for Specification 3.1.1.3 -Moderator Temperature Coefficient. 3.1.3.5 -Shutdown Bank Insertion limits. 3.1.3.6 -Control Bank Insertion Limits. 3.2.1 -Axial Plux Difference. 3.2.2 -Heat Flux Hot Channel Factor. and 3.2.3 -Nuclear Enthalpy Rise Hot Channel Factor). c. XN-NF-82-21(P)(A). "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations." approved version as specified in the COLR. (Methodology for Specification 3.2.3 -Nuclear Enthalpy Rise Hot Channel Factor). SHEARON HARRIS -UNIT 1 6-24 Amendment No. 94 No changes to this page. Included for information only. ADMINISTRATIVE CONTROLS 6.9.1.6 CORE OPERATING LIMITS REPORT !Continued) d. e. f. g. XN-75-32(P)(A), "Computational Procedure for Evaluating Fuel Rod Bowing,* approved version as specified in the COLR. (Methodology for Specification 3.2.2 -Heat Flux Hot Channel Factor, and 3.2.3 -Nuclear Enthalpy Rise Hot Channel. Factor). EMF-84-093(P)(A), "Steam Line Break Methodology for PWRs," approved version as specified in the COLR. (Methodology for Specification 3.1.1.3 -Moderator Temperature Coefficient, 3.1.3.5 -Shutdown Bank Insertion Limits, 3.1.3.6 -Control Bank Insertion Limits, and 3.2.3 -Nuclear Enthalpy Rise Hot Channel Factor). ANP-3011 (P), "Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis,* 1 Revision 1, as approved by NRC Safety Evaluation dated May 30, 2012. (Methodology for Specification 3.2.1
- Axial Flux Difference, 3.2.2 -Heat Flux Hot Channel Factor, and 3.2.3 -Nuclear Enthalpy Rise Hot Channel Factor).
- XN-NF-78-44(NP)(A), "A Generic Analysis of the Control Rod Ejection Transient for Pressurized Water Reactors," approved version as specified in the COLR. (Methodology for Specification 3.1.3.5 -Shutdown Bank Insertion Limits, 3.1.3.6 -C.ontrol Bank Insertion Limits, and 3.2.2 -Heat Flux Hot Channel Factor). SHEARON HARRIS -UNIT 1
- 13e B. 3, 4 3. 4 15 Insert 2 ADMINISTRATIVE CONTROLS 6.9.1.6 CORE OPERATING LIMITS REPORT (Continued) (Methodology for Specification 3.1.1.3 -Moderator Temperature Coefficient, 3.1.3.5 -Shutdown Bank Insertion Limits, 3.1.3.6 -Control Bank Insertion Limits, 3.2.1 -Axial Flux Difference, 3.2.2 -Heat Flux Hot Channel Factor, and 3.2.3 -Nuclear Enthalpy Rise Hot Channel Factor). o. Mechanical Design Methodologies XN-NF-81-58(P}(A}, "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model," approved version as specified in the COLR. ANF-81-58(P)(A}, "RODEX2 Fuel Rod Thermal Mechanical Response Evaluation Model," approved version as specified in the COLR. XN-NF-82-06(P}(A}, "Qualification of Exxon Nuclear Fuel for Extended Burn up," approved version as specified in the COLR. ANF-88-133(P}(A}, "Qualification of Advanced Nuclear Fuels' PWR Design Methodology for Rod Bumups of62 GWd/MTU," approved version as specified in the COLR. XN-NF-85-92(P}(A), "Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results," approved version as specified in the COLR. (see next page) EMF-92-116(P)(A), "Generic Mechanical Design Criteria for PWR Fuel Designs," approved version as specified in the COLR. (Methodologies for Specification 3.2.1 -Axial Flux Difference, 3.2.2 -Heat Flux Hot Channel Factor, and 3.2.3 -Nuclear Enthalpy Rise Hot Channel Factor}. DPC-NE-2005-P-A, "Thermal-Hydraulic Statistical Core Design Methodology." approved version as specified in the COLR. (Methodology for Specification 3.2.3-Nuclear Enthalpy Rise Hot Channel Factor} 6.9.1.6.3 The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, nuclear timits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met. . 6.9.1.6.4 The CORE OPERA TING LIMITS REPORT, including any mid-cycle revisions or supplements, shall be provided, upon issuance for each reload cycle, to the NRC Document Control Desk, with copies to the Regional Administrator and Resident Inspector. 6.9.1.7 STEAM GENERATOR TUBE INSPECTION REPORT A report shaU be submitted within 180 days after the initial entry into HOT SHUTDOWN following completion of an inspection performed in accordance with Specification 6.8.4.1. The report shall include: a. The scope of inspections performed on each SG, b. Degradation mechanisms found, c. Nondestructive examination techniques utilized for each degradation mechanism, SHEARON HARRIS -UNIT 1 6-24c Amendment No.1 4 8
Attachment 5 DPC-NE-1008, "Nuclear Design Methodology Using CASMO-5/SIMULATE-3 for Westinghouse Reactors" (Redacted)
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Figure 3-12 cont'd Harris Unit 1Cycle18 Assembly Average Power Distribution Comparisons 100.0% FP, 192 EFPD, Control D at 218 SWD 3-32 a-c
Figure 3-12 cont'd Harris Unit 1 Cycle 18 Assembly Average Power Distribution Comparisons 100.0% FP, 485 EFPD, Control D at 218 SWD 3-35 a-c
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Attachment 6 DPC-NF-2010, Revision 3, "Nuclear Physics Methodology for Reload Design" and Technical Justification of Changes
K1K2
T1K1T2K2
=( )10= (/°)
,,=( )10= (/°)
=( )10= (/%)
=+
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Available Rod Worth Required Rod Worth
=1+Bias+UC+Ux1+Ux2+
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UL(C) =UL(C) ()=++()
UL(C=()=+()
Incore instrument at corelocation L-05 is used for RVLIS
Technical Justification of Changes for Revision 3
Identification of the report location of each change refers back to the Revision 2a page numbering.
Attachment 8 DPC-NE-2011, Revision 2, "Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors" and Technical Justification of Changes (Redacted)
=1+Bias+(UC+Ux1+Ux2+...)
DNBM=MinMARP(x,y)RPP(x,y)MARP(x,y)100
CFMM=MinMAXLHRLHR(x,y,z)MAXLHR
LOCA LIMIT KW/FT CORE HEIGHT FT
% MARGIN % OFFSET LEGEND Bank D (swd) LOCA MARGIN VS. AXIAL OFFSET
% MARGIN % OFFSET LOFA MARGIN VS. AXIAL OFFSET
F(x,y,z) K(Z)F(x,y,z) .K(Z)F F(x,y,z)UMTMTTILT[F(x,y,z)M(x,y,z)F(x,y,z)
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F(x,y,z)F(x,y,z)F (x,y,z)F(x,y,z)M(x,y,z)F (x,y,z)MBLP=(,,)F (x,y,z)100%F(x,y,z)F(x,y,z)F (x,y,z) F(x,y)F(x,y,z)F(x,y)F(x,y)F(x,y)F(x,y)F(x,y)F(x,y)F (x,y)
F(x,y)F (x,y)F(x,y)M(x,y)F(x,y,z),F(x,y,z)F (x,y,z)F(x,y)F (x,y)
K(Z) Core Height, ft
MAX. ALLOWABLE TOTAL PEAK AXIAL PEAK LOCATION, X/L
% OF RATED THERMAL POWER FLUX DIFFERECE ( I) %
ROD BANK POSITION (STEPS WITHDRAWN) PERCENT OF RATED THERMAL POWER
Technical Justification of Changes for Revision 2 (Redacted)
refers back to the Revision 1a page numbering.
Section 1.3, Applicability of the Method:
Section 7,
References:
Appendix A:
Section 4.1 Change:
Section 4.5 Change:
Section 4.6 Change:
Appendix B Addition:
FF,F, (,)
Section 6.1 Change: Section 6.3 Change: Section 6.3.1 Change:
Section 6.1, equation on page 6-1 and removal of TILT definition: F(x,y,z)UMTMTTILT[F(x,y,z)M(x,y,z) Section 6.1, text and equation on page 6-3: FF(x,y,z)F(x,y,z)=F(x,y,z)M(x,y,z)/(UMTMTTILT)Section 6.2, equation on page 6-4: F(x,y,z)UMTMTTILT[F(x,y,z)M(x,y,z)]Section 6.2, text and equation on page 6-5: FF(x,y,z)
F(x,y,z)=F(x,y,z)M(x,y,z)/(UMTMTTILT)Section 6.3, equation on page 6-6 and removal of TILT definition: F(x,y)UMRTILT[F(x,y)M(x,y)]Section 6.3, text and equation on page 6-7: FM(x,y)F(x,y)F(x,y)=F(x,y)M(x,y)/(UMRTILT)
F(x,y,z)=F(x,y,z)M(x,y,z)/(UMTMTTILT)F(x,y,z)=F(x,y,z)M(x,y,z)/(UMTMTTILT)F(x,y)=F(x,y)M(x,y)/(UMRTILT)
(,,) F(x,y,z) F(x,y,z) NP(x,y,z)(,,)(,)F(x,y,z)F(x,y) Section 6.1.1 Change: 15-10-50510152025-40-35-30-25-20-15-10-50510152025303540LOCA FQ Margin (Limit -FQ)/LIMIT*100AFD (%)Figure 1Example LOCA FQ Margin Versus AFDTypicalSteady State AFDOperating RangeNegative HFPAFD LimitPostive HFPAFD Limit FSection 6.3.1 Change: FSection 6.4.1 Change:
Section 6.4.2 Change: F(x,y,z)F(x,y)(,)F(x,y)F(x,y) RNP(x,y)
F(x,y,z)F(x,y,z)F(x,y,z)F(x,y,z)F(x,y,z)
F(x,y,z)F(x,y,z)F(x,y,z)F(x,y,z)
F(x,y,z)F(x,y,z)F(x,y,z)F(x,y,z)
F(x,y,z),F(x,y,z)F (x,y,z)F(x,y)F (x,y)
Change B-1: Appendix B Changes, third paragraph, page B-2:
Attachment 9 Application of the DPC-NE-2004-PA Operating Limits and Maximum Allowable Total Peak Methodology to Harris and Robinson Nuclear Plants
Page 1 Application of the DPC-NE-2004-PA Operating Limits and Maximum Allowable Total Peak Methodology to Harris and Robinson Nuclear Plants 1.0 Background DPC-NE-2004-PA, "Core Thermal-Hydraulic Methodology Using VIPRE-01", Reference 1, describes Duke Energy's NRC-approved steady state thermal hydraulics analysis methodology for McGuire and Catawba Nuclear Stations using the VIPRE-01 computer code. Key elements of the report include a description of the code inputs and correlations used to develop McGuire/Catawba VIPRE-01 models. The methodology used to determine allowable thermal-hydraulic operating limits in terms of core power level, reactor coolant system temperature and pressure, and three dimensional core power distributions is also presented. This methodology is used to determine allowable operating limits that provide DNB protection for nuclear plants that utilize a Westinghouse NSSS control and protection system. This method is described in Section 5 of Reference 1. Section 5 is applicable to the Harris and Robinson nuclear units as well as McGuire and Catawba.
2.0 Applicability The thermal-hydraulic methodology in Reference 1 is intended to interface with a Westinghouse plant reactor protection system (RPS) to ensure the required DNB design basis is satisfied. McGuire, Catawba, Harris and Robinson nuclear units all employ a Westinghouse RPS system to prevent fuel damage from occurring during Condition I and II events. DNB protection is accomplished through the use of the over-temperature delta-temperature (OTT) and over-power delta-temperature (OPT) trip functions, in combination with the following trips which limit the applicable range over which the OTT and OPT trip functions must provide DNB protection. High pressurizer pressure trip Low pressurizer pressure trip Low reactor coolant flow trip Core DNB limits are determined for a range of operating conditions to ensure that the DNB design basis is satisfied. The protected space in terms of power level and pressure is defined by the OTT/OPT trip, high pressurizer pressure trip, and low pressurizer pressure trip. An upper range on temperature is set by a combination of the Hot Leg boiling limit and the steam generator safety valve actuation. The low reactor coolant system (RCS) flow trip provides DNB protection by limiting the reactor coolant flow that must be considered. The methodology described in Reference 1 requires the following three elements to be in place. A NRC-approved VIPRE-01 model for the unit being evaluated A NRC-approved critical heat flux correlation for the fuel type A NRC-approved statistical DNBR limit for the fuel type VIPRE-01 models have been developed to represent the Harris and Robinson core geometry and AREVA's HTP fuel product in current operation. The VIPRE-01 model developed, along with the code inputs and correlations used to the represent AREVA's Advanced W 17x17 HTP and 15x15 HTP fuel products are described in Appendices H and I of DPC-NE-2005-P-A, "Duke Energy Thermal-Hydraulic Statistical Core Design Methodology", Reference 2.
Page 2 The critical heat flux (CHF) correlation applicable to AREVA's Advanced W 15x15 and 17x17 HTP fuel designs is the HTP CHF correlation developed in Reference 3. The applicability of this correlation with the VIPRE-01 computer code was also demonstrated in Appendices H and I of Reference 2. In addition, Reference 2 describes the statistical core design (SCD) DNBR limit for each reactor considering plant specific uncertainties. All three methodology components (VIPRE-01 model, CHF correlation and statistical DNBR limit) in Reference 2 are used to calculate core DNB limits. These limits are determined as a function of power level, reactor coolant system temperature, and reactor coolant system pressure using a reference radial peak power, FH, and the pin-by-pin power distribution specific to the applicable fuel design evaluated in Reference 2 Appendices H and I. Each point along the line corresponds to a constant DNBR value based on the SCD DNBR limit. DNB limit lines are provided in the core safety limits figure presented in Technical Specifications Figure 2.1-1 or provided in the cycle specific Core Operating Limits Report (COLR).
In summary, the Duke Energy thermal-hydraulic methodology used to provide DNB protection for Westinghouse plant instrumentation systems is not dependent upon the plant where it is applied. As a result, the method is as applicable to the Harris and Robinson nuclear units as well as McGuire and Catawba. It does however, rely on the use of NRC-approved VIPRE-01 model, CHF correlation, and DNBR limit as specified in Reference 2 for each plant.
3.0 Maximum Allowable Peaking Limits The Duke Energy maximum allowable total peaking (MATP) methodology is described in Section 5 of Reference 1. This method is independent of the reactor design or fuel type and consists of DNB calculations performed with an NRC-approved model, fuel design, critical heat flux correlation, and DNBR limit. The methodology is used to develop local fuel peaking limits to prevent DNB for a fixed reactor core operating condition defined by a thermal power level, pressure, RCS flow and reactor coolant temperature. The MATP limits developed are lines of constant MDNBR for a range of axial peak magnitudes with the location of the peak varied from the bottom to the top of the core. For a given axial peak magnitude (Fz) and axial location, the total peak (FH
- Fz) that yields a MDNBR equal to the DNBR limit defines a single point along a MATP curve. An example MATP curve is shown in Figure 14 of Reference 1. The MATP limits are used to ensure the DNB design basis for Conditions I and II transients is satisfied by comparing calculated total peaking factors from a core nodal simulator (e.g. SIMULATE-3) against the MATP limits for a series of power distributions. Core power distributions are generated (by the core nodal simulator) as a function of rod position, burnup, xenon concentration, inlet temperature, and core power level. The method to generate these power distributions is described in Reference 4. DNBR margin is computed for each fuel assembly to verify positive margin exists for the allowable core power versus axial flux difference (AFD) limit used to establish the F(I) portion of the OTT trip function. The same methodology is used at a different state point condition to develop the core operational AFD limits. The operational AFD limits protect the core against the limiting Condition II DNB transient, typically the Loss of Flow transient. In summary, the Duke Energy MATP methodology is a generic methodology that relies on the NRC-approved VIPRE-01 model, fuel design, CHF correlation, and DNBR limit. It is applicable to any reactor or fuel type provided the NRC approvals for the unit specific methodology components are obtained.
Since NRC approval has been granted for the VIPRE-01 model, CHF correlation, and DNBR limits in Page 3 References 2 and 3, application of the Duke Energy Core DNB limit and MATP methodology to the Harris and Robinson plants is appropriate. 4.0 Conclusion The methodology described in DPC-NE-2004-P for development of limits to verify the required DNB design basis is satisfied is independent of the Westinghouse unit analyzed. The methodology does require an NRC approved: VIPRE-01 model for the unit being evaluated Critical Heat Flux (CHF) correlation for the fuel type Statistical DNBR limit for the fuel type Satisfying these three criteria allow use of the Duke Energy thermal-hydraulic analysis methodology, including the Core Safety Limit generation and MATP limit approach described in Reference 1 for plants using the Westinghouse reactor control and protection system.
5.0 References 1.DPC-NE-2004-PA, Rev. 2a, "Core Thermal-Hydraulic Methodology Using VIPRE-01" 2.DPC-NE-2005-P-A, Rev. 5, "Duke Energy Thermal-Hydraulic Statistical Core Design Methodology" 3.EMF-92-153(P)(A), Rev. 1, "HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel" 4.DPC-NE-2011-PA, Rev. 1a, "Nuclear Design Methodology Report For Core Operating Limits of Westinghouse Reactors"