ML20147A039: Difference between revisions

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{{#Wiki_filter:Page 1 Information Removed from Amendment 63 of the Updated Final Safety Analysis Report Shearon Harris Nuclear Power Plant, Unit 1 (HNP)
{{#Wiki_filter:Page 1 Information Removed from Amendment 63 of the Updated Final Safety Analysis Report Shearon Harris Nuclear Power Plant, Unit 1 (HNP)
Impacted Related Discussion Sections/Tables/Figures Sections                             The original reactor vessel head (RVH) was replaced 3.5.1.1.2.1.1; 3.9.4.1; 3.9.4.3;     during refueling outage 22 (RFO-22). The 4.5.1.3; 4.5.2.1; 5.2.3.1; 5.2.3.2.2; replacement RVH includes integrated control rod 5.2.4.2.1; 5.3.1.2; 5.3.1.3.2;       drive mechanism (CRDM) latch housings that are 5.3.3.1; 5.3.3.3; 5.3.3.5; 5.3.3.7;   welded to the head penetration nozzles with a full 15.4.8.1.1; 18.2.2.3                 penetration butt weld that eliminates the lower canopy welds of the original design. The CRDMs were Tables                               replaced during RFO-22 as well. The replacement 3.2.1-1; 3.5.1-4; 3.9.2-1; 5.2.1-1;   CRDM integrated rod travel housing is a one-piece 5.2.3-1; 5.2.3-2; 5.3.1-1; 5.3.1-2;   design that eliminates the top cap. Information 5.3.3-1                              related to the original RVH that is no longer applicable to the HNP design was replaced by Figures                              information pertinent to the new RVH.
Impacted Sections/Tables/Figures Related Discussion Sections 3.5.1.1.2.1.1; 3.9.4.1; 3.9.4.3; 4.5.1.3; 4.5.2.1; 5.2.3.1; 5.2.3.2.2; 5.2.4.2.1; 5.3.1.2; 5.3.1.3.2; 5.3.3.1; 5.3.3.3; 5.3.3.5; 5.3.3.7; 15.4.8.1.1; 18.2.2.3 Tables 3.2.1-1; 3.5.1-4; 3.9.2-1; 5.2.1-1; 5.2.3-1; 5.2.3-2; 5.3.1-1; 5.3.1-2; 5.3.3-1 Figures 3.9.4-1; 3.9.4-2; 5.3.3-1 The original reactor vessel head (RVH) was replaced during refueling outage 22 (RFO-22). The replacement RVH includes integrated control rod drive mechanism (CRDM) latch housings that are welded to the head penetration nozzles with a full penetration butt weld that eliminates the lower canopy welds of the original design. The CRDMs were replaced during RFO-22 as well. The replacement CRDM integrated rod travel housing is a one-piece design that eliminates the top cap. Information related to the original RVH that is no longer applicable to the HNP design was replaced by information pertinent to the new RVH.
3.9.4-1; 3.9.4-2; 5.3.3-1 Section                              The Boraflex Monitoring Program was removed since 18.1.1.12                            Boraflex is no longer credited in the HNP criticality analysis for spent fuel storage racks. This change is associated with Renewed Operating License Amendment No. 167 (ADAMS Accession No. ML18204A286).
Section 18.1.1.12 The Boraflex Monitoring Program was removed since Boraflex is no longer credited in the HNP criticality analysis for spent fuel storage racks. This change is associated with Renewed Operating License Amendment No. 167 (ADAMS Accession No. ML18204A286).
Section                               Information related to operating organization 13.1.2                                personnel functions, responsibilities, and authorities was updated to reflect current organizational structure.
Section 13.1.2 Information related to operating organization personnel functions, responsibilities, and authorities was updated to reflect current organizational structure.
Sections                             High-energy line breaks (HELBs) inside the Reactor 3.6.2.1.1.3; 3.6A.2.1.2.2; 3.6A.2.3; Auxiliary Building (RAB) have been analyzed and 3.6A.6.4                             evaluated using a new GOTHIC-based computer model and other supporting calculations. Content has Tables                               been revised to clarify that Generic Letter 87-11 no 3.6A-4; 3.6A-13                       longer requires a minimum of two arbitrary intermediate breaks to be postulated between Figures                               terminal piping ends. Tables and figures showing pipe 3.6A-12-CALC; 3.6A-12-PLOT-C;         whip restraints, break identification numbers, and 3.6A-12-PLOT-D; 3.6A-12-PLOT-F;       break locations were revised to delete terminal-end 3.6A-12-PLOT-G; 3.6.A-25-CALC;       breaks exempted by the analysis and to correct 3.6A-25-PLOT-A; 3.6A-25-PLOT-B;       administrative errors.
Sections 3.6.2.1.1.3; 3.6A.2.1.2.2; 3.6A.2.3; 3.6A.6.4 Tables 3.6A-4; 3.6A-13 Figures 3.6A-12-CALC; 3.6A-12-PLOT-C; 3.6A-12-PLOT-D; 3.6A-12-PLOT-F; 3.6A-12-PLOT-G; 3.6.A-25-CALC; 3.6A-25-PLOT-A; 3.6A-25-PLOT-B; 3.6A-25-PLOT-C High-energy line breaks (HELBs) inside the Reactor Auxiliary Building (RAB) have been analyzed and evaluated using a new GOTHIC-based computer model and other supporting calculations. Content has been revised to clarify that Generic Letter 87-11 no longer requires a minimum of two arbitrary intermediate breaks to be postulated between terminal piping ends. Tables and figures showing pipe whip restraints, break identification numbers, and break locations were revised to delete terminal-end breaks exempted by the analysis and to correct administrative errors.
3.6A-25-PLOT-C Section                               The seismic monitoring magnetic tape recorder and 3.7.4                                analog playback system were replaced with a digital recorder and digital display system. Information was Table                                updated to reflect this change.
Section 3.7.4 Table 3.7.4-1 The seismic monitoring magnetic tape recorder and analog playback system were replaced with a digital recorder and digital display system. Information was updated to reflect this change.  
3.7.4-1


Page 2 Impacted Related Discussion Sections/Tables/Figures Sections                             Vendor-generated Chapter 15 dose consequence 15.0A.1.1; 15.1.5.3; 15.1.5.3.3;     analyses were replaced by applying the in-house dose 15.1.5.3.5; 15.1.5.3.6; 15.2.6.3.4;   computational tool, LOCADOSE, to generate dose 15.2.6.3.5; 15.3.3.3; 15.3.3.3.2;     results using Alternative Source Term (AST) 15.3.3.3.3; 15.3.3.4; 15.4.3.2.3;     methodology consistent with the current HNP 15.4.7.3; 15.4.8.3.3; 15.6.2.3.1.2;   licensing basis. This includes revising the Main Steam 15.6.2.3.1.4; 15.6.2.3.1.5;           Line Break (MSLB) analysis to eliminate the fuel 15.6.3.4; 15.6.3.4.3; 15.6.3.4.6;     failure source term dose analysis case to reflect that 15.6.3.4.7; 15.6.5.4.2; 15.6.5.5;     per Regulatory Guide 1.183, Appendix E, Assumptions 15.6.5.5.1; 15.6.5.5.2; 15.6.5.5.3;   for Evaluating the Radiological Consequences of a 15.6.5.5.4; 15.6.5.5.5; 15.7.1.3.2;   PWR Main Steam Line Break Accident, and the NRC-15.7.1.3.3; 15.7.4.2.5; 15.7.4.2.6   approved AST methodology for HNP, when no fuel failure is predicted to occur due to the MSLB accident Tables                               then two iodine spiking cases should be evaluated in 15.1.5-5; 15.1.5-6; 15.2.6-6;         lieu of assuming a fuel failure case.
Page 2 Impacted Sections/Tables/Figures Related Discussion Sections 15.0A.1.1; 15.1.5.3; 15.1.5.3.3; 15.1.5.3.5; 15.1.5.3.6; 15.2.6.3.4; 15.2.6.3.5; 15.3.3.3; 15.3.3.3.2; 15.3.3.3.3; 15.3.3.4; 15.4.3.2.3; 15.4.7.3; 15.4.8.3.3; 15.6.2.3.1.2; 15.6.2.3.1.4; 15.6.2.3.1.5; 15.6.3.4; 15.6.3.4.3; 15.6.3.4.6; 15.6.3.4.7; 15.6.5.4.2; 15.6.5.5; 15.6.5.5.1; 15.6.5.5.2; 15.6.5.5.3; 15.6.5.5.4; 15.6.5.5.5; 15.7.1.3.2; 15.7.1.3.3; 15.7.4.2.5; 15.7.4.2.6 Tables 15.1.5-5; 15.1.5-6; 15.2.6-6; 15.3.3-6; 15.3.3-7; 15.4.3-5; 15.4.3-6; 15.4.8-5; 15.4.8-6; 15.6.2-2; 15.6.3-6; 15.6.3-13; 15.6.5-15; 15.6.5-16; 15.6.5-17; 15.6.5-18; 15.7.1-2; 15.7.4-2; 15.7.4-4 Vendor-generated Chapter 15 dose consequence analyses were replaced by applying the in-house dose computational tool, LOCADOSE, to generate dose results using Alternative Source Term (AST) methodology consistent with the current HNP licensing basis. This includes revising the Main Steam Line Break (MSLB) analysis to eliminate the fuel failure source term dose analysis case to reflect that per Regulatory Guide 1.183, Appendix E, Assumptions for Evaluating the Radiological Consequences of a PWR Main Steam Line Break Accident, and the NRC-approved AST methodology for HNP, when no fuel failure is predicted to occur due to the MSLB accident then two iodine spiking cases should be evaluated in lieu of assuming a fuel failure case.
15.3.3-6; 15.3.3-7; 15.4.3-5; 15.4.3-6; 15.4.8-5; 15.4.8-6; 15.6.2-2; 15.6.3-6; 15.6.3-13; 15.6.5-15; 15.6.5-16; 15.6.5-17; 15.6.5-18; 15.7.1-2; 15.7.4-2; 15.7.4-4 Sections                              Beginning with the cycle 23 core design, non-LOCA 4.2.2.3; 4.3.1.1; 4.3.1.3; 4.3.1.4.1; nuclear design and safety analyses were performed 4.3.1.4.2; 4.3.1.5; 4.3.1.6;         using Duke Energy methods approved for use by 4.3.2.2.1; 4.3.2.2.2; 4.3.2.2.3;     license amendments issued by the NRC. The changes 4.3.2.2.4; 4.3.2.2.5; 4.3.2.2.6;     made reflect the transition to the Duke nuclear 4.3.2.2.7; 4.3.2.2.8; 4.3.2.2.9;     design and safety analysis methods used to confirm 4.3.2.3; 4.3.2.3.1; 4.3.2.3.2;       the acceptability of postulated Chapter 15 accidents, 4.3.2.3.5; 4.3.2.4.11; 4.3.2.4.13;   and to describe the nuclear software used to 4.3.2.4.15; 4.3.2.5; 4.3.3; 4.3.3.1; calculate nuclear physics data core power 4.3.3.2; 7.7.1.3.1                   distributions. The change to Section 7.7.1.3.1 reflects the removal of the reference to the PDC-3 core Tables                               monitoring software.
Sections 4.2.2.3; 4.3.1.1; 4.3.1.3; 4.3.1.4.1; 4.3.1.4.2; 4.3.1.5; 4.3.1.6; 4.3.2.2.1; 4.3.2.2.2; 4.3.2.2.3; 4.3.2.2.4; 4.3.2.2.5; 4.3.2.2.6; 4.3.2.2.7; 4.3.2.2.8; 4.3.2.2.9; 4.3.2.3; 4.3.2.3.1; 4.3.2.3.2; 4.3.2.3.5; 4.3.2.4.11; 4.3.2.4.13; 4.3.2.4.15; 4.3.2.5; 4.3.3; 4.3.3.1; 4.3.3.2; 7.7.1.3.1 Tables 4.1.1-2; 4.3.2-1; 4.3.2-2; 4.3.2-3; 4.3.2-21; 4.3.2-38; 4.3.2-39 Beginning with the cycle 23 core design, non-LOCA nuclear design and safety analyses were performed using Duke Energy methods approved for use by license amendments issued by the NRC. The changes made reflect the transition to the Duke nuclear design and safety analysis methods used to confirm the acceptability of postulated Chapter 15 accidents, and to describe the nuclear software used to calculate nuclear physics data core power distributions. The change to Section 7.7.1.3.1 reflects the removal of the reference to the PDC-3 core monitoring software.
4.1.1-2; 4.3.2-1; 4.3.2-2; 4.3.2-3; 4.3.2-21; 4.3.2-38; 4.3.2-39 Section                               Information related to redundant safety grade flow 9.2.2.5                               instrumentation for detecting loss of component cooling water to the oil coolers of the reactor coolant pump motors was removed. The removal of these sentences removed redundant, irrelevant, and incorrect statements. The abandoned flow switches do not perform a design function.
Section 9.2.2.5 Information related to redundant safety grade flow instrumentation for detecting loss of component cooling water to the oil coolers of the reactor coolant pump motors was removed. The removal of these sentences removed redundant, irrelevant, and incorrect statements. The abandoned flow switches do not perform a design function.  


Page 3 Impacted Related Discussion Sections/Tables/Figures Section                             Information related to containment jib crane was 18.2.6.2                           removed due to the cranes permanent removal.
Page 3 Impacted Sections/Tables/Figures Related Discussion Section 18.2.6.2 Tables 6.1.1-2; 6.2.5-4 Information related to containment jib crane was removed due to the cranes permanent removal.
Tables 6.1.1-2; 6.2.5-4 Sections                            A new GOTHIC containment model was created for 6.2.1.1.3.2, Appendix 6.2A;         the HNP Chapter 6 containment analysis. The new 9.2.1.3.1; 9.2.5.3.1               GOTHIC model replaced the existing CONTEMPT analysis of record, and the FSAR was updated Tables                             accordingly. Additionally, information related to 6.2.1-50; 9.2.1-5; 9.2.1-10;       material thermal conductivity, volumetric heat 9.2.1-11                           capacity, average node spacing, and heat rejection was removed on the basis that it was excessively detailed information.
Sections 6.2.1.1.3.2, Appendix 6.2A; 9.2.1.3.1; 9.2.5.3.1 Tables 6.2.1-50; 9.2.1-5; 9.2.1-10; 9.2.1-11 A new GOTHIC containment model was created for the HNP Chapter 6 containment analysis. The new GOTHIC model replaced the existing CONTEMPT analysis of record, and the FSAR was updated accordingly. Additionally, information related to material thermal conductivity, volumetric heat capacity, average node spacing, and heat rejection was removed on the basis that it was excessively detailed information.
Sections                           Information related to the power range high negative 3.9.1.1.2; 7.2.1.1.2               neutron flux rate trip was removed as aligned with the update to the HNP Technical Specifications per Tables                              License Amendment No. 175 (ADAMS Accession No.
Sections 3.9.1.1.2; 7.2.1.1.2 Tables 7.2.1-1; 7.2.1-3 Information related to the power range high negative neutron flux rate trip was removed as aligned with the update to the HNP Technical Specifications per License Amendment No. 175 (ADAMS Accession No. ML19225C069). Additionally, parameter values associated with Core Thermal Overpower Trips were removed and point to the Core Operating Limits Report in accordance with changes to the HNP Technical Specifications per License Amendment No.
7.2.1-1; 7.2.1-3                    ML19225C069). Additionally, parameter values associated with Core Thermal Overpower Trips were removed and point to the Core Operating Limits Report in accordance with changes to the HNP Technical Specifications per License Amendment No.
161 (ADAMS Accession No. ML17250A202).
161 (ADAMS Accession No. ML17250A202).
Sections                           Content has been updated to reflect the approval of 15.0; 15.0.3.1; 15.0.3.2; 15.0.3.3; Duke Energy methodologies per License Amendment 15.0.3.4; 15.0.4; 15.0.5; 15.0.6;   Nos. 157, 164, and 171 (ADAMS Accession Nos.
Sections 15.0; 15.0.3.1; 15.0.3.2; 15.0.3.3; 15.0.3.4; 15.0.4; 15.0.5; 15.0.6; 15.0.7; 15.0.8; 15.0.10.3; 15.0.11.2; 15.0.11.2.1; 15.0.11.2.4; 15.0.11.3; 15.0.13; 15.1.1; 15.1.3; 15.1.5; 15.2.3; 15.2.8; 15.3.1; 15.3.2; 15.3.3; 15.4.1; 15.4.2; 15.4.3.1; 15.4.3.2; 15.4.3.3; 15.4.4; 15.4.6; 15.4.7; 15.4.8; 15.5.1; 15.6.1; 15.6.3 Tables 15.0.3.6; 15.0.13-1; 15.0.3-5; 15.0.6-2; 15.0.8-1; 15.1.3-1; 15.1.3-2; 15.1.3-3; 15.1.3-4; 15.1.5-3; 15.2.3-1; 15.2.3-2; 15.2.3-3; 15.2.3-4; 15.2.3-5; 15.2.3-6; 15.2.8-1; 15.2.8-2; 15.2.8-3; 15.2.8-4; 15.2.8-5; 15.2.8-7; 15.2.8-8; 15.3.2-1; 15.3.2-2; 15.3.2-3; 15.3.2-4; Content has been updated to reflect the approval of Duke Energy methodologies per License Amendment Nos. 157, 164, and 171 (ADAMS Accession Nos.
15.0.7; 15.0.8; 15.0.10.3;         ML17102A923, ML18060A401, and ML19290F980, 15.0.11.2; 15.0.11.2.1;             respectively). Content related to previous 15.0.11.2.4; 15.0.11.3; 15.0.13;   Framatome, Inc. methods and computer codes that 15.1.1; 15.1.3; 15.1.5; 15.2.3;     are no longer applicable following the 15.2.8; 15.3.1; 15.3.2; 15.3.3;     implementation of Duke Energy methods has been 15.4.1; 15.4.2; 15.4.3.1; 15.4.3.2; removed.
ML17102A923, ML18060A401, and ML19290F980, respectively). Content related to previous Framatome, Inc. methods and computer codes that are no longer applicable following the implementation of Duke Energy methods has been removed.
15.4.3.3; 15.4.4; 15.4.6; 15.4.7; 15.4.8; 15.5.1; 15.6.1; 15.6.3 Tables 15.0.3.6; 15.0.13-1; 15.0.3-5; 15.0.6-2; 15.0.8-1; 15.1.3-1; 15.1.3-2; 15.1.3-3; 15.1.3-4; 15.1.5-3; 15.2.3-1; 15.2.3-2; 15.2.3-3; 15.2.3-4; 15.2.3-5; 15.2.3-6; 15.2.8-1; 15.2.8-2; 15.2.8-3; 15.2.8-4; 15.2.8-5; 15.2.8-7; 15.2.8-8; 15.3.2-1; 15.3.2-2; 15.3.2-3; 15.3.2-4;


Page 4 Impacted Related Discussion Sections/Tables/Figures 15.3.3-1; 15.3.3-2; 15.3.3-3; 15.3.3-4; 15.3.3-5; 15.4.1-1; 15.4.1-2; 15.4.1-3; 15.4.1-4; 15.4.2-1; 15.4.2-2; 15.4.2-2a; 15.4.2-3; 15.4.2-4; 15.4.2-5; 15.4.3-1; 15.4.3-2; 15.4.3-3; 15.4.3-4a; 15.4.3-6; 15.4.3-7; 15.4.3-8; 15.4.3-9; 15.4.6-2; 15.4.7-1; 15.4.8-1; 15.4.8-2; 15.4.8-3; 15.4.8-4b; 15.5.1-1; 15.5.1-2; 15.5.1-3; 15.5.1-4; 15.6.1-1; 15.6.1-2; 15.6.1-3; 15.6.1-4 Figures 15.0.5-4; 15.0.5-5; 15.0.5-6; 15.1.3-1; 15.1.3-2; 15.1.3-3; 15.1.3-4; 15.1.5-1; 15.1.5-2; 15.1.5-3; 15.1.5-4; 15.1.5-5; 15.1.5-6; 15.2.3-1; 15.2.3-2; 15.2.3-3; 15.2.3-4; 15.2.3-5; 15.2.3-6; 15.2.3-7; 15.2.3-9; 15.2.3-10; 15.2.3-11; 15.2.3-12; 15.2.8-1 through 15.2.8-32; 15.3.2-1; 15.3.2-2; 15.3.2-3; 15.3.2-4; 15.3.2-5; 15.3.2-6; 15.3.2-7; 15.3.3-1 through 15.3.3-13; 15.4.1-1; 15.4.1-2; 15.4.1-3; 15.4.1-4; 15.4.2-1; 15.4.2-2; 15.4.2-3; 15.4.2-5; 15.4.2-6; 15.4.2-7; 15.4.2-8; 15.4.2-9; 15.4.2-10; 15.4.3-1; 15.4.3-2; 15.4.3-3; 15.4.3-4; 15.4.3-6; 15.4.3-7; 15.4.3-8; 15.4.3-10; 15.4.3-11; 15.4.3-12; 15.4.3-13; 15.4.3-14; 15.4.3-15; 15.4.8-9 through 15.4.8-16; 15.5.1-1 through 15.5.1-14; 15.6.1-1 through 15.6.1-7}}
Page 4 Impacted Sections/Tables/Figures Related Discussion 15.3.3-1; 15.3.3-2; 15.3.3-3; 15.3.3-4; 15.3.3-5; 15.4.1-1; 15.4.1-2; 15.4.1-3; 15.4.1-4; 15.4.2-1; 15.4.2-2; 15.4.2-2a; 15.4.2-3; 15.4.2-4; 15.4.2-5; 15.4.3-1; 15.4.3-2; 15.4.3-3; 15.4.3-4a; 15.4.3-6; 15.4.3-7; 15.4.3-8; 15.4.3-9; 15.4.6-2; 15.4.7-1; 15.4.8-1; 15.4.8-2; 15.4.8-3; 15.4.8-4b; 15.5.1-1; 15.5.1-2; 15.5.1-3; 15.5.1-4; 15.6.1-1; 15.6.1-2; 15.6.1-3; 15.6.1-4 Figures 15.0.5-4; 15.0.5-5; 15.0.5-6; 15.1.3-1; 15.1.3-2; 15.1.3-3; 15.1.3-4; 15.1.5-1; 15.1.5-2; 15.1.5-3; 15.1.5-4; 15.1.5-5; 15.1.5-6; 15.2.3-1; 15.2.3-2; 15.2.3-3; 15.2.3-4; 15.2.3-5; 15.2.3-6; 15.2.3-7; 15.2.3-9; 15.2.3-10; 15.2.3-11; 15.2.3-12; 15.2.8-1 through 15.2.8-32; 15.3.2-1; 15.3.2-2; 15.3.2-3; 15.3.2-4; 15.3.2-5; 15.3.2-6; 15.3.2-7; 15.3.3-1 through 15.3.3-13; 15.4.1-1; 15.4.1-2; 15.4.1-3; 15.4.1-4; 15.4.2-1; 15.4.2-2; 15.4.2-3; 15.4.2-5; 15.4.2-6; 15.4.2-7; 15.4.2-8; 15.4.2-9; 15.4.2-10; 15.4.3-1; 15.4.3-2; 15.4.3-3; 15.4.3-4; 15.4.3-6; 15.4.3-7; 15.4.3-8; 15.4.3-10; 15.4.3-11; 15.4.3-12; 15.4.3-13; 15.4.3-14; 15.4.3-15; 15.4.8-9 through 15.4.8-16; 15.5.1-1 through 15.5.1-14; 15.6.1-1 through 15.6.1-7}}

Latest revision as of 13:54, 11 December 2024

Amendment 63 to Updated Final Safety Analysis Report, Information Removed from Amendment
ML20147A039
Person / Time
Site: Harris Duke Energy icon.png
Issue date: 05/15/2020
From:
Duke Energy Progress
To:
Office of Nuclear Reactor Regulation
Shared Package
ML20147A016 List:
References
RA-20-0134
Download: ML20147A039 (4)
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Page 1 Information Removed from Amendment 63 of the Updated Final Safety Analysis Report Shearon Harris Nuclear Power Plant, Unit 1 (HNP)

Impacted Sections/Tables/Figures Related Discussion Sections 3.5.1.1.2.1.1; 3.9.4.1; 3.9.4.3; 4.5.1.3; 4.5.2.1; 5.2.3.1; 5.2.3.2.2; 5.2.4.2.1; 5.3.1.2; 5.3.1.3.2; 5.3.3.1; 5.3.3.3; 5.3.3.5; 5.3.3.7; 15.4.8.1.1; 18.2.2.3 Tables 3.2.1-1; 3.5.1-4; 3.9.2-1; 5.2.1-1; 5.2.3-1; 5.2.3-2; 5.3.1-1; 5.3.1-2; 5.3.3-1 Figures 3.9.4-1; 3.9.4-2; 5.3.3-1 The original reactor vessel head (RVH) was replaced during refueling outage 22 (RFO-22). The replacement RVH includes integrated control rod drive mechanism (CRDM) latch housings that are welded to the head penetration nozzles with a full penetration butt weld that eliminates the lower canopy welds of the original design. The CRDMs were replaced during RFO-22 as well. The replacement CRDM integrated rod travel housing is a one-piece design that eliminates the top cap. Information related to the original RVH that is no longer applicable to the HNP design was replaced by information pertinent to the new RVH.

Section 18.1.1.12 The Boraflex Monitoring Program was removed since Boraflex is no longer credited in the HNP criticality analysis for spent fuel storage racks. This change is associated with Renewed Operating License Amendment No. 167 (ADAMS Accession No. ML18204A286).

Section 13.1.2 Information related to operating organization personnel functions, responsibilities, and authorities was updated to reflect current organizational structure.

Sections 3.6.2.1.1.3; 3.6A.2.1.2.2; 3.6A.2.3; 3.6A.6.4 Tables 3.6A-4; 3.6A-13 Figures 3.6A-12-CALC; 3.6A-12-PLOT-C; 3.6A-12-PLOT-D; 3.6A-12-PLOT-F; 3.6A-12-PLOT-G; 3.6.A-25-CALC; 3.6A-25-PLOT-A; 3.6A-25-PLOT-B; 3.6A-25-PLOT-C High-energy line breaks (HELBs) inside the Reactor Auxiliary Building (RAB) have been analyzed and evaluated using a new GOTHIC-based computer model and other supporting calculations. Content has been revised to clarify that Generic Letter 87-11 no longer requires a minimum of two arbitrary intermediate breaks to be postulated between terminal piping ends. Tables and figures showing pipe whip restraints, break identification numbers, and break locations were revised to delete terminal-end breaks exempted by the analysis and to correct administrative errors.

Section 3.7.4 Table 3.7.4-1 The seismic monitoring magnetic tape recorder and analog playback system were replaced with a digital recorder and digital display system. Information was updated to reflect this change.

Page 2 Impacted Sections/Tables/Figures Related Discussion Sections 15.0A.1.1; 15.1.5.3; 15.1.5.3.3; 15.1.5.3.5; 15.1.5.3.6; 15.2.6.3.4; 15.2.6.3.5; 15.3.3.3; 15.3.3.3.2; 15.3.3.3.3; 15.3.3.4; 15.4.3.2.3; 15.4.7.3; 15.4.8.3.3; 15.6.2.3.1.2; 15.6.2.3.1.4; 15.6.2.3.1.5; 15.6.3.4; 15.6.3.4.3; 15.6.3.4.6; 15.6.3.4.7; 15.6.5.4.2; 15.6.5.5; 15.6.5.5.1; 15.6.5.5.2; 15.6.5.5.3; 15.6.5.5.4; 15.6.5.5.5; 15.7.1.3.2; 15.7.1.3.3; 15.7.4.2.5; 15.7.4.2.6 Tables 15.1.5-5; 15.1.5-6; 15.2.6-6; 15.3.3-6; 15.3.3-7; 15.4.3-5; 15.4.3-6; 15.4.8-5; 15.4.8-6; 15.6.2-2; 15.6.3-6; 15.6.3-13; 15.6.5-15; 15.6.5-16; 15.6.5-17; 15.6.5-18; 15.7.1-2; 15.7.4-2; 15.7.4-4 Vendor-generated Chapter 15 dose consequence analyses were replaced by applying the in-house dose computational tool, LOCADOSE, to generate dose results using Alternative Source Term (AST) methodology consistent with the current HNP licensing basis. This includes revising the Main Steam Line Break (MSLB) analysis to eliminate the fuel failure source term dose analysis case to reflect that per Regulatory Guide 1.183, Appendix E, Assumptions for Evaluating the Radiological Consequences of a PWR Main Steam Line Break Accident, and the NRC-approved AST methodology for HNP, when no fuel failure is predicted to occur due to the MSLB accident then two iodine spiking cases should be evaluated in lieu of assuming a fuel failure case.

Sections 4.2.2.3; 4.3.1.1; 4.3.1.3; 4.3.1.4.1; 4.3.1.4.2; 4.3.1.5; 4.3.1.6; 4.3.2.2.1; 4.3.2.2.2; 4.3.2.2.3; 4.3.2.2.4; 4.3.2.2.5; 4.3.2.2.6; 4.3.2.2.7; 4.3.2.2.8; 4.3.2.2.9; 4.3.2.3; 4.3.2.3.1; 4.3.2.3.2; 4.3.2.3.5; 4.3.2.4.11; 4.3.2.4.13; 4.3.2.4.15; 4.3.2.5; 4.3.3; 4.3.3.1; 4.3.3.2; 7.7.1.3.1 Tables 4.1.1-2; 4.3.2-1; 4.3.2-2; 4.3.2-3; 4.3.2-21; 4.3.2-38; 4.3.2-39 Beginning with the cycle 23 core design, non-LOCA nuclear design and safety analyses were performed using Duke Energy methods approved for use by license amendments issued by the NRC. The changes made reflect the transition to the Duke nuclear design and safety analysis methods used to confirm the acceptability of postulated Chapter 15 accidents, and to describe the nuclear software used to calculate nuclear physics data core power distributions. The change to Section 7.7.1.3.1 reflects the removal of the reference to the PDC-3 core monitoring software.

Section 9.2.2.5 Information related to redundant safety grade flow instrumentation for detecting loss of component cooling water to the oil coolers of the reactor coolant pump motors was removed. The removal of these sentences removed redundant, irrelevant, and incorrect statements. The abandoned flow switches do not perform a design function.

Page 3 Impacted Sections/Tables/Figures Related Discussion Section 18.2.6.2 Tables 6.1.1-2; 6.2.5-4 Information related to containment jib crane was removed due to the cranes permanent removal.

Sections 6.2.1.1.3.2, Appendix 6.2A; 9.2.1.3.1; 9.2.5.3.1 Tables 6.2.1-50; 9.2.1-5; 9.2.1-10; 9.2.1-11 A new GOTHIC containment model was created for the HNP Chapter 6 containment analysis. The new GOTHIC model replaced the existing CONTEMPT analysis of record, and the FSAR was updated accordingly. Additionally, information related to material thermal conductivity, volumetric heat capacity, average node spacing, and heat rejection was removed on the basis that it was excessively detailed information.

Sections 3.9.1.1.2; 7.2.1.1.2 Tables 7.2.1-1; 7.2.1-3 Information related to the power range high negative neutron flux rate trip was removed as aligned with the update to the HNP Technical Specifications per License Amendment No. 175 (ADAMS Accession No. ML19225C069). Additionally, parameter values associated with Core Thermal Overpower Trips were removed and point to the Core Operating Limits Report in accordance with changes to the HNP Technical Specifications per License Amendment No.

161 (ADAMS Accession No. ML17250A202).

Sections 15.0; 15.0.3.1; 15.0.3.2; 15.0.3.3; 15.0.3.4; 15.0.4; 15.0.5; 15.0.6; 15.0.7; 15.0.8; 15.0.10.3; 15.0.11.2; 15.0.11.2.1; 15.0.11.2.4; 15.0.11.3; 15.0.13; 15.1.1; 15.1.3; 15.1.5; 15.2.3; 15.2.8; 15.3.1; 15.3.2; 15.3.3; 15.4.1; 15.4.2; 15.4.3.1; 15.4.3.2; 15.4.3.3; 15.4.4; 15.4.6; 15.4.7; 15.4.8; 15.5.1; 15.6.1; 15.6.3 Tables 15.0.3.6; 15.0.13-1; 15.0.3-5; 15.0.6-2; 15.0.8-1; 15.1.3-1; 15.1.3-2; 15.1.3-3; 15.1.3-4; 15.1.5-3; 15.2.3-1; 15.2.3-2; 15.2.3-3; 15.2.3-4; 15.2.3-5; 15.2.3-6; 15.2.8-1; 15.2.8-2; 15.2.8-3; 15.2.8-4; 15.2.8-5; 15.2.8-7; 15.2.8-8; 15.3.2-1; 15.3.2-2; 15.3.2-3; 15.3.2-4; Content has been updated to reflect the approval of Duke Energy methodologies per License Amendment Nos. 157, 164, and 171 (ADAMS Accession Nos.

ML17102A923, ML18060A401, and ML19290F980, respectively). Content related to previous Framatome, Inc. methods and computer codes that are no longer applicable following the implementation of Duke Energy methods has been removed.

Page 4 Impacted Sections/Tables/Figures Related Discussion 15.3.3-1; 15.3.3-2; 15.3.3-3; 15.3.3-4; 15.3.3-5; 15.4.1-1; 15.4.1-2; 15.4.1-3; 15.4.1-4; 15.4.2-1; 15.4.2-2; 15.4.2-2a; 15.4.2-3; 15.4.2-4; 15.4.2-5; 15.4.3-1; 15.4.3-2; 15.4.3-3; 15.4.3-4a; 15.4.3-6; 15.4.3-7; 15.4.3-8; 15.4.3-9; 15.4.6-2; 15.4.7-1; 15.4.8-1; 15.4.8-2; 15.4.8-3; 15.4.8-4b; 15.5.1-1; 15.5.1-2; 15.5.1-3; 15.5.1-4; 15.6.1-1; 15.6.1-2; 15.6.1-3; 15.6.1-4 Figures 15.0.5-4; 15.0.5-5; 15.0.5-6; 15.1.3-1; 15.1.3-2; 15.1.3-3; 15.1.3-4; 15.1.5-1; 15.1.5-2; 15.1.5-3; 15.1.5-4; 15.1.5-5; 15.1.5-6; 15.2.3-1; 15.2.3-2; 15.2.3-3; 15.2.3-4; 15.2.3-5; 15.2.3-6; 15.2.3-7; 15.2.3-9; 15.2.3-10; 15.2.3-11; 15.2.3-12; 15.2.8-1 through 15.2.8-32; 15.3.2-1; 15.3.2-2; 15.3.2-3; 15.3.2-4; 15.3.2-5; 15.3.2-6; 15.3.2-7; 15.3.3-1 through 15.3.3-13; 15.4.1-1; 15.4.1-2; 15.4.1-3; 15.4.1-4; 15.4.2-1; 15.4.2-2; 15.4.2-3; 15.4.2-5; 15.4.2-6; 15.4.2-7; 15.4.2-8; 15.4.2-9; 15.4.2-10; 15.4.3-1; 15.4.3-2; 15.4.3-3; 15.4.3-4; 15.4.3-6; 15.4.3-7; 15.4.3-8; 15.4.3-10; 15.4.3-11; 15.4.3-12; 15.4.3-13; 15.4.3-14; 15.4.3-15; 15.4.8-9 through 15.4.8-16; 15.5.1-1 through 15.5.1-14; 15.6.1-1 through 15.6.1-7

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