ML15196A570

Submittal of Updated Final Safety Analysis Report, Revision 24

ML15196A570
Person / Time
Site:	Oconee
Issue date:	06/30/2015
From:	Batson S L Duke Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML15196A571	List: ML15196A558 ML15196A560 ML15196A570 ML15197A315 ONS-2015-072, Oconee, Units 1, 2, and 3, Submittal of Updated Final Safety Analysis Report, Revision 24 ONS-2015-072, Oconee, Units 1, 2, and 3, Updated Final Safety Analysis Report, Revision 24, Chapter 2 - Site Characteristics ONS-2015-072, Oconee, Units 1, 2, and 3, Updated Final Safety Analysis Report, Revision 24, Chapter 2 - Site Characteristics, Tables - Appendix 2A. Tables ONS-2015-072, Oconee, Units 1, 2, and 3, Updated Final Safety Analysis Report, Revision 24, Chapter 8 - Electric Power ONS-2015-072, Updated Final Safety Analysis Report, Revision 24, Chapter 9 - Auxiliary Systems
References
ONS-2015-072
Download: ML15196A570 (72)

Text

~~9ENERGY° Vice President Oconee Nuclear Station Duke Energy~ONOIVP I 7600 Rochester Hwy Seneca, SC 29672 ONS-201 5-072 o; 864.873.3274 f:864.873.4208 Scott.Batson@duke -energy. corn June 30, 2015 Attn: Document Control Desk 10 CER 50.71(e)U.S. Nuciear Regulatory Commission 10 CFR 50.59(d)11555 Rockville Pike 10 CFR 54.37(b)Rockville, MD 20852-2746

Subject:

Duke Energy Carolinas, (LLC) (Duke Energy)Oconee Nuclear Station, Units 1, 2, and 3 Docket Nos. 50-269, 50-270, 50-287 Updated Final Safety Analysis Report, Revision 24 Pursuant to 10 CFR 50.71(e), and in accordance with 10 CFR 50.4, Duke Energy hereby submits the Oconee Nuclear Station Updated Final Safety Analysis Report (UFSAR), Revision 24. The effective date of the UFSAR revision is December 31, 2014, as indicated at* the bottom of each page. Changes made in Rev. 24 are indicated by side bars.The Oconee UFSAR, Revision 24, is enclosed on one compact disk (CD). The contents are in Adobe Acrobat Portable Document Format (pdf). As required by NRC guidance for electronic submissions, Attachment 1 provides a listing of the document components that comprise the enclosed CD. Attachment 2 provides the List of Effective Pages for Tables and Figures.Attachment 3 provides insertion instructions for those receiving hardcopy distribution.

Attachment 4 provides a listing of items removed in the 2014 UFSAR update.Attachment 5 provides the report of changes, tests, and experiments performed pursuant to 10 CFR 50.59.In addition, 10 CFR 54.37(b) requires that after the renewed license is issued, the UFSAR update must include any systems, Structures and components (SSCs) newly identified that would have been subject to an aging management review or evaluation of time-limited aging analysis in accordance with 10 CFR 54.21. The UFSAR update must describe how the effects of aging are managed such that the intended function(s) in 10 CFR 54.4(b) will be effectively maintained during the period of extended operation.

A review was completed to determine whether any newly-identified SSCs existed in support of submitting UFSAR Revision 24. As a result of this review, there were no newly-identified SSCs for which aging management reviews or time-limited aging analyses would apply.This submittal document contains no new or revised regulatory commitments.

If you have any questions regarding this submittal, please contact Susan Perry at (864) 873-4370.www.duke-energy comn ONS-201 5-072 Document Control Desk June 30, 2015 D Page 2 I declare under penalty of perjury that the foregoing is true and correct. Executed on June 30, 2015.Sincerely, Scott L. Batson Vice President Oconee Nuclear Station Attachments:

1. Document Components on CD 2. List of Effective Pages (LOEP) for Tables and Figures 3. Update Insertion Instructions (for hardcopy distribution only)4. List of Removed Items D 5. 10 CFR 50.59 Report

Enclosure:

CD: Oconee Nuclear Station Updated Final Safety Analysis Report, 2014 Update -Rev 24 cc: Mr. Victor McCree, NRC Region II Administrator (CD)U.S. Nuclear Regulatory Commission Marquis One Tower 245 Peachtree Center Ave., NE, Suite 1200 Atlanta, GA 30303-1257 Mr. James R. Hall, Project Manager (CD)U.S. Nuclear Regulatory Commission 11555 Rockville Pike Mail Stop O-8G9A Rockville, MD 20852 Mr. Eddy Crowe (hardcopy)

Senior Resident Inspector Oconee Nuclear Station (r'.ONS-201 5-072 Document Control Desk I June 30, 2015 Attachment 1 Document Components on CD (2 pages)

D ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 1, Page 1 Filename 000 ONS UFSAR Rev 24 Title Page 001 ONS UFSAR Rev 24 Oh 1 Text 002 ONS UFSAR Rev 24 Oh 1 Tables 003 ONS UFSAR Rev 24 Oh 1 Figures 004 ONS UFSAR Rev 24 Oh 2 Text 005 ONS UFSAR Rev 24 Oh 2 Tables 006 ONS UFSAR Rev 24 Oh 2 Figures 007 ONS UFSAR Rev 24 Oh 3 Text 008 ONS UFSAR Rev 24 Oh 3 Tables 009 ONS UFSAR Rev 24 Oh 3 Figures 010 ONS UFSAR Rev 24 Oh 4 Text 011 ONS UFSAR Rev 24 Oh 4 Tables 012 ONS UFSAR Rev 24 Oh 4 Figures 013 ONS UFSAR Rev 24 Oh 5 Text 014 ONS UFSAR Rev 24 Oh 5 Tables 015 ONS UFSAR Rev 24 Oh 5 Figures 016 ONS UFSAR Rev 24 Oh 6 Text 017 ONS UFSAR Rev 24 Oh 6 Tables 018 ONS UFSAR Rev 24 Oh 6 Figures 019 ONS UFSAR Rev 24 Oh 7 Text 020 ONS UFSAR Rev 24 Oh 7 Tables 021 ONS UFSAR Rev 24 Oh 7 Figures 022 ONS UFSAR Rev 24 Oh 8 Text 023 ONS UFSAR Rev 24 Oh 8 Tables 024 ONS UFSAR Rev 24 Oh 8 Figures 025 ONS UFSAR Rev 24 Oh 9 Text 026 ONS UFSAR Rev 24 Oh 9 Tables 027 ONS UFSAR Rev 24 Oh 9 Figures File Size 82 KB 309 KB 159 KB 4,668 KB 911 KB 4,788 KB 7,501 KB 2,259 KB 1,727 KB 10,260 KB 1,006 KB 1,202 KB 1,323 KB 1,000 KB 1,686 KB 5,074 KB 950 KB 1,964 KB 13,914 KB 1,433 KB 366 KB 8,618 KB 548 KB 441 KB 32,654 KB 1,482 KB 1,029 KB 24,022 KB ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 1, Page 2 Filename File Size 028 ONS UFSAR Rev 24 Oh 10 Text 611 KB 029 ONS UFSAR Rev 24 Oh 10 Tables 141 KB 030 ONS UFSAR Rev 24 Oh 10 Figures 734 KB 031 ONS UFSAR Rev 24 Oh 11 Text 838 KB 032 ONS UFSAR Rev 24 Oh 11 Tables 479 KB 033 ONS UFSAR IRev 24 Oh 11 Figures 285 KB 034 ONS UFSAR Rev 24 Oh 12 Text 406 KB 035 ONS UFSAR Rev 24 Oh 12 Tables 210 KB 036 ONS UFSAR Rev 24 Oh 13 Text 676 KB 037 ONS UFSAR Rev 24 Oh 13 Tables 74 KB 038 ONS UFSAR Rev 24 Oh 13 Figures 422 KB 039 ONS UFSAR Rev 24 Oh 14 Text 449 KB 040 ONS UFSAR Rev 24 Oh 14 Tables 130 KB 041 ONS UFSAR Rev 24 Oh 15 Text 1,904 KB 042 ONS UFSAR Rev 24 Oh 15 Tables 2,210 KB 043 ONS UFSAR Rev 24 Oh 15 Figures 5,528 KB 044 ONS UFSAR Rev 24 Oh 16 Text 98 KB 045 ONS UFSAR Rev 24 Oh 17 Text 101 KB 046 ONS UFSAR Rev 24 Oh 18 Text 785 KB 047 ONS UFSAR Rev 24 Oh 18 Tables 104 KB pm d ONS-201 5-072 Document Control Desk SJune 30, 2015 Attachment 2 List of Effective Pages (LOEP) for Tables and Figures (40 pages) 9.b ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 2, Page 1 OCONEE UFSAR -2014 UPDATE List of Effective Pages (LOEP) for Tables The purpose of this list is to assure that the pages in the Tables section of your manual match the most recent issue, as well as to show a full accounting of all tables, including those that have been deleted. The earliest effective date, 12/31/00, was used when all tables were re-issued.Effective Date Table No.Table Title Chapter 1 12/31/00 12/31/00 12/31/00 Chapter 2 1-1 1-2 1-3 Key Dates in Oconee History Engineered Safeguards Equipment Deleted Per 1997 Update 12/31/00 2-1 1970 Population Distribution 0-10 Miles 12/31/00 2-2 2010 Projected Population Distribution 0-10 Miles 12/31/00 2-3 1970 Population

~Distribution 0-50 Miles 12/31/00 2-4 2010 Projected Population Distribution 0-50 Miles 12/31/00 2-5 1970 Cumulative Population Density 0-50 Miles 12/31/00 2-6 2010 Projected Cumulative Population Density 0-50 Miles 12/31/08 2-7 Frequency of Tropical Cyclones in Georgia, South Carolina and North Carolina Plus Coastal Waters 12/31/08 2-8 Mean Monthly Thunderstorm Days and Thunderstorms for Nuclear Plant Site 12/31/08 2-9 Duration and Frequency (in Hours) of Calm and Near-Calm Winds Average of Three Locations (1/59 -12/63)12/31/08 2-10 Annual Surface Wind Rose For Greenville, South Carolina (1/59 -12/63)12/31/08 2-1 1 Percent Frequency of Wind Speeds at Various Hours Through the Day -Greenville, S.C. (1/59 -12/63)12/31/08 2-12 Duration and Frequency of Calm and Near-Calm Winds Average of Three Locations (1/59 -12/63)12/31/08 2-13 Percentage Distribution of Athens, Georgia Annual Winds at 0630 Eastern Standard Time (800-1300 Feet Above Ground)12/31/08 2-14 Percentage Distribution of Athens, Georgia Annual Winds at 0630 Eastern Standard Time (2300-2800 Feet Above Ground)12/31/08 2-15 Average Wind Direction Change with Height, Athens, Georgia, by Lapse Rates in the Lowest 50 Meters-Two Years of Record Effective Date 12/31/14 ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 2, Page 2 Effective Table Table Title Date No.12/31/08 2-16 67.50 Sector Wind Direction Persistence Duration (in Hours) Greenville, S.C.WBAS 12/31/08 2-17 112.50 Sector Wind Direction Persistence Duration (in Hours) Greenville, S.C.WBAS 12/31/08 2-18 Surface Temperature (0 F) Clemson, S.C. (68 Years of Record)12/31/08 2-19 Surface Precipitation (Inches) Clemson, S.C. (71 Years of Record)12/31/08 2-20 Precipitation

-Wind Statistics

-Greenville, S.C. 1959-1 963 (By Precipitation Intensities) 12/31/08 2-21 Pasquill Stability Categories for Greenville, South Carolina 12/31/08 2-22 Pasquill Stability Category and Supplemental Data for Greenville, S.C.12/31/08 2-23 Average Temperature Difference

(°F) at Minimum Temperature Time 12/31/08 2-24 Joint Frequency Distribution of Wind Speed and Wind Direction for each Stability Class, for Greenville-Spartanburg, South Carolina for 1975 12/31/08 2-25 Joint Frequency Distribution of Wind Speed and Wind Direction for each Stability Class, for Greenville-Spartanburg, South Carolina for 1968-1972 12/31/08 2-26 Joint Frequencies of Wind Direction and Speed by Stability Class (March 1970 -March 1972)12/31/08 2-27 Joint Frequency Tables of Wind Direction and Speed by Atmospheric Stability-Low and High Level (January 1975 -December 1975)12/31/08 2-28 Composite Poorest Diffusion Conditions Observed for Each Hour of Day (Based on 30 Months of Data)12/31/08 2-29 Dispersion Factors Used for Accident and Routine Operational Analyses X/Q 12/31/08 2-30 Determining Appropriate Dispersion Factors 12/31/08 2-31 Oconee Nuclear Station XIQ at Critical Receptors to 5 Miles (Depleted by Dry Deposition) 12/31/08 2-32 Oconee Nuclear Station D/Q at Critical Receptors to 5 Miles 12/31/08 2-33 Oconee Nuclear Station X/Q at Critical Receptors to 5 Miles (Non-Depleted) 12/31/08 2-34 Relative Concentration, X/Q, Frequency Distribution Without Wind Speed Correction 12/31/08 2-35 Gas-Tracer Experimental Results From January 15 -March 11, 1970 12/31/08 2-36 Relative Concentration, X/Q, Frequency Distribution With Wind Speed Correction 12/31/08 2-37 Comparative Wind Speed Data Effective Date 12/31/14 ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 2, Page 3 Effective Table Table Title Date No.12/31/08 2-38 Supplemental Data Oconee Meteorological Survey (Tower Data) For Period of June 1, 1968 Thru May 31, 1969 12/31/08 2-39 Supplemental Data -Joint Frequency Distribution 12/31/08 2-40 Deleted Per 2008 Update 12/31/08 2-41 Deleted Per 2008 Update 12/31/08 2-42 Deleted Per 2008 Update 12/31/08 2-43 Deleted Per 2008 Update 12/31/08 2-44 Supplemental Data -SF 6 Detector Readings -Test Date: January 28, 1970 12/31/08 2-45 Deleted Per 2008 Update 12/31/08 2-46 Deleted Per 2008 Update 12/31/08 2-47 Deleted Per 2008 Update 12/31/08 2-48 Deleted Per 2008 Update 12/31/08 2-49 Deleted Per 2008 Update 12/31/08 2-50 Deleted Per 2008 Update 12/31/08 2-51 Deleted Per 2008 Update 12/31/08 2-52 Deleted Per 2008 Update 12/31/08 2-53 Deleted Per 2008 Update 12/31/08 2-54 Deleted Per 2008 Update 12/31/08 2-55 Deleted Per 2008 Update 12/31/08 2-56 Deleted Per 2008 Update 12/31/08 2-57 Deleted Per 2008 Up)date 12/31/08 2-58 Deleted Per 2008 Up)date 12/31/08 2-59 Deleted Per 2008 UpJdate 12/31/08 2-60 Deleted Per 2008 Up3date 12/31/08 2-61 Deleted Per 2008 12/31/08 2-62 Deleted Per 2008 Up3date 12/31/08 2-63 Deleted Per 2008 Up3date 12/31/08 2-64 Deleted Per 2008 Up3date 12/31/08 2-65 Deleted Per 2008 12/31/08 2-66 Deleted Per 2008 Up3date Effective Date 12/31/14 ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 2, Page 4 Effective Table Table Title Date No.12/31/08 2-67 Deleted Per 2008 12/31/08 2-68 Deleted Per 2008 12/31/08 2-69 Deleted Per 2008 12/31/08 2-70 Deleted Per 2008 12/31/08 2-71 Deleted Per 2008 12/31/08 2-72 Deleted Per 2008 Up3date 12/31/08 2-73 Deleted Per 2008 Up3date 12/31/08 2-74 Deleted Per 2008 Up3date 12/31/08 2-75 Deleted Per 2008 12/31/08 2-76 Deleted Per 2008 Up3date 12/31/08 2-77 Deleted Per 2008 Update 12/31/08 2-78 Deleted Per 2008 Up3date 12/31/08 2-79 Deleted Per 2008 Upodate 12/31/08 2-80 Deleted Per 2008 Up3date 12/31/08 2-81 Deleted Per 2008 Up3date 12/31/08 2-82 Deleted Per 2008 Up3date 12/31/08 2-83 Deleted Per 2008 12/31/08 2-84 Deleted Per 2008 Up3date 12/31/08 2-85 Deleted Per 2008 Up3date 12/31/08 2-86 Deleted Per 2008 Up3date 12/31/08 2-87 Deleted Per 2008 Up3date 12/31/08 2-88 Deleted Per 2008 Update 12/31/08 2-89 Deleted Per 2008 Update 12/31/08 2-90 Deleted Per 2008 Update 12/31/08 2-91 Deleted Per 2008 Update 12/31/08 2-92 Deleted Per 2008 Update 12/31/00 2-93 Soil Permeability Test Results 12/31/00 2-94 Significant Earthquakes in the Southeast United States (intensity V or Greater)12/31/00 2-95 Velocity Measurements 12/31/00 2-96 Core Measurements Effective Date 12/31/14 ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 2, Page 5 Effective Table Table Title Date No.Chapter 3 12/31/04 3-1 System Piping Classification 12/31/14 3-2 System Component Classification 12/31/04 3-3 Summary of Missile Equations 12/31/00 3-4 List of Symbols 12/31/14 3-5 Properties of Missiles -Reactor Vessel & Control Rod Drive 12/31/03 3-6 Properties of Missiles -Steam Generator 12/31/00 3-7 Properties of Missiles -Pressurizer 12/31/00 3-8 Properties of Missiles -Quench Tank and Instruments 12/31/00 3-9 Properties of Missiles -System Piping 12/31/00 3-10 Missile Characteristics 12/31/00 3-1 1 Depth of Penetration of Concrete 12/31/00 3-12 Containment Coatings 12/31/00 3-13 Service Load Combinations for Reactor Building 12/31/00 3-14 Accident, Wind, and Seismic Load Combinations and Factors for Class 1 Concrete Structures 12/31/00 3-15 Inward Displacement of Liner Plate 12/31/03 3-16 Stress Analysis Results 12/31/00 3-17 Stress Analysis Results 12/31/03 3-18 Stress Analysis Results 12/31/03 3-19 Stress Analysis Results 12/31/03 3-20 Stress Analysis Results 12/31/03 3-21 Stress Analysis Results 12/31/00 3-22 Bent Wire Test Results 12/31/09 3-23 Auxiliary Building Loads and Conditions 12/31/00 3-24 Mark-BZ Fuel Assembly Seismic and LOCA Results at 600 0 F 12/31/00 3-25 Deleted Per 1996 Update 12/31/00 3-26 Stress Limits for Seismic, Pipe Rupture and Combined Loads 12/31/00 3-27 Deleted Per 1999 Update 12/31/00 3-28 Deleted Per 2004 Update Effective Date 12/31/14 p June 30,201 Effective Table Table Title Date No.12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31(00 12/31/00 12/31/00 3-29 3-30 3-31 3-32 3-33 3-34 3-35 3-36 3-37 3-38 3-39 3-40 3-41 3 -42 3-43 3-44 3-45 3-46 3-47 3-48 3-49 3-50 3-51 3-52 3-53.3-54 3-55 3-56 3-57 3-58 Deleted Per 2004 Update Deleted Per 2004 Update Deleted Per 2004 Update Deleted Per 2004 Update Deleted Per 2004 Update Deleted Per 2004 Update Deleted Per 2004 Update Deleted Per 2004 Update Deleted Per 2004 Update Deleted Per 2004 Update Deleted Per 2004. Update Deleted Per 2004 Update Deleted Per 2004 Update Deleted Per 2004 Update Deleted Per 2004 Update Deleted Per 2004 Update Deleted Per 2004 Update Deleted Per 2004 Update Deleted Per 2004 Update Deleted Per 2004 Update Deleted Per 2004 Update Deleted Per 2004 Update Deleted Per 2004 Update Deleted Per 2004 Update Deleted Per 2004 Update Deleted Per 2004 Update Deleted Per 2004 Update Deleted Per 2004 Update Deleted Per 2004 Update Deleted Per 2004 Update Effective Date 12/31/14 June 30,201 Effective Table Table Title Date No.12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/13 Chapter 4 12/31/09 12/31/09 12/31/12 12/31/12 12/31/00 12/31/12 12/31/09 12/31/00 12/31/12 12/31 /12 12/31/12 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 3-59 3-60 3-61 3-62 3-63 3-64 3-65 3-66 3-67 3-68 4-1 4-2 4-3 4-4 4-5 4-6 4-7 4-8 4-9 4-10 4-11 4-12 4-13 4-14 4-15 4-16 4-17 4-18 Deleted Per 2004 Update Deleted Per 2004 Update Deleted Per 2004 Update Deleted Per 2004 Update Deleted Per 2004 Update Deleted Per 2004 Update Deleted Per 2004 Update Deleted Per 2004 Update Deleted Per 2004 Update Electrical Equipment Seismic Qualification Core Design, Thermal, and Hydraulic Data Fuel Assembly Components Nuclear Design Data Typical Fuel Cycle Excess Reactivity, HFP Samarium Effective Multiplication Factor Shutdown Margin Calculation for Typical Oconee Fuel Cycle Moderator Temperature Coefficient (For the First Cycle)BOL Distributed-Temperature Moderator Coefficients, 100% Power, 1200 ppm Boron (O1C01)BOL Distributed-Temperature Moderator Coefficients, vs. Power, No Xenon BOL Distributed-Temperature Moderator Coefficient, 100% Full Power Power Coefficients of Reactivity pH Characteristics Design Methods Deleted Per 1999 Update Deleted Per 1997 Update Internals Vent Valve Materials Vent Valve Shaft & Bushing Clearances Control Rod Assembly Data Effective Date 12/31/14 ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 2, Page 8 Effective Table Table Title Date No.12/31/00 4-19 Axial Power Shaping Rod Assembly Data 12/31/12 4-20 Burnable Poison Rod Assembly Data 12/31/00 4-21 Control Rod Drive Mechanism Design Data 12/31/13 4-22 Fuel Assembly/APSR Compatibility 12/31/09 4-23 Fuel Assembly Design Descriptions 12/31/13 4-24 Design Information for Current Demonstration Programs vs. Typical FAs Chapter 5 12/31/00 5-1 Reactor Coolant System Pressure Settings 12/31/04 5-2 Transient Cycles for ROS Components Except Pressurizer Surge Line 12/31/10 5-3 Stress Limits for Seismic, Pipe Rupture, and Combined Loads 12/31/04 5-4 Reactor Coolant System Component Codes 12/31/14 5-5 Materials of Construction 12/31/00 5-6 Summary of Primary Plus Secondary Stress Intensity for Components of the Reactor Vessel 12/31/00 5-7 Summary of Cumulative Fatigue Usage Factors for Components of the Reactor Vessel 12/31/04 5-8 Stresses Due to a Maximum Design Steam Generator Tube Sheet Pressure Differential of 2,500 psi at 650°F 12/31/04 5-9 Ratio of Allowable Stresses to Computed Stresses for a Steam Generator Tube Sheet Pressure Differential of 2,500 psi 12/31/04 5-10 Fabrication Inspections 12/31/03 5-1 1 Reactor Vessel Design Data 12/31/03 5-12 Reactor Vessel -Physical Properties (Oconee 1 )12/31/03 5-13 Reactor Vessel -Chemical Properties (Oconee 1)12/31/00 5-14 Reactor Vessel -Mechanical Properties (Oconee 2 & 3)12/31/04 5-15 Reactor Coolant Flow Distribution with Less than Four Pumps Operating 12/31/11 5-16 Reactor Coolant Pump -Design Data (Oconee 1)12/31/01 5-17 Reactor Coolant Pump -Design Data (Oconee 2, 3) (Data.Per Pump)12/31/00 5-18 Reactor Coolant Pump Casings -Code Allowables (Applies to Oconee 2 & 3)12/31/00 5-19 Deleted Per 2000 Update 12/31/04 5-20 Steam Generator Design Data (Data Per Steam Generator)

Effective Date 12/31/14 ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 2, Page 9 Effective Table Table Title Date No.12/31/04 5-21 Reactor Coolant Piping Design Data 12/31/06 5-22 Pressurizer Design Data 12/31/00 5-23 Operating Design Transient Cycles for Pressurizer Surge Line 12/31/00 5-24 Evaluation of Reactor Vessel Pressurized Thermal Shock Toughness Properties at 48 EFPY -Oconee Unit 1 12/31/00 5-25 Evaluation of Reactor Vessel Pressurized Thermal Shock Toughness Properties at 48 EFPY -Oconee Unit 2 12/31/00 5-26 Evaluation of Reactor Vessel Pressurized Thermal Shock Toughness Properties at 48 EFPY -Oconee Unit 3 12/31/00 5-27 Evaluation of Reactor Vessel Extended Life (48EFPY) Charpy V-Notch Upper-Shelf Energy -Oconee Unit 1 12/31/00 5-28 Evaluation of Reactor Vessel Extended Life (48EFPY) Charpy V-Notch Upper-Shelf Energy -Oconee Unit 2 12/31/00 5-29 Evaluation of Reactor Vessel Extended Life (48EFPY) Charpy V-Notch Upper-Shelf Energy -Oconee Unit 3 Chapter 6 12/31/00 6-1 Deleted Per 1995 Update 12/31/00 6-2 Deleted Per 2000 Update 12/31/00 6-3 Quality Control Standards for Engineered Safeguards Systems 12/31/04 6-4 Engineered Safeguards Piping Design Conditions 12/31/00 6-5 Single Failure Analysis Reactor Building Spray System 12/31/00 6-6 Single Failure Analysis For Reactor Building Cooling System 12/31/14 6-7 Reactor Building Penetration Valve Information 12/31/05 6-8 High Pressure Injection System Component Data 12/31/09 6-9 Low Pressure Injection System Component Data 12/31/00 6-10 Core Flooding System Components Data 12/31/05 6-11 Single Failure Analysis -Emergency Core Cooling System 12/31/00 6-12 Oconee Nuclear Station Analysis of Valve Motors Which May Become Submerged Following A LOCA 12/31/00 6-13 Equipment Operational During An Accident and Located Outside Containment 12/31/00 6-14 Equipment Operational During an Accident and Located Within the Containment Effective Date 12/31/14 ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 2, Page 10 Effective Table Table Title Date No.12/31/05 6-15 Emergency Core Cooling Systems Performance Testing 12/31/00 6-16 Deleted Per 1999 Update 12/31/00 6-17 Deleted Per 1999 Update 12/31/00 6-18 Inventory of Iodine Isotopes in Reactor Building (at t = o)12/31/05 6-19 Single Failure Analysis for Reactor Building Penetration Room Ventilation System 12/31/00 6-20 Parameters for Boron Precipitation Analysis 12/31/03 6-21 Summary of Calculated Containment Pressures and Temperatures for LOCA Cases 12/31/13 6-22 Containment Response Analyses Initial Conditions 12/31/13 6-23 Containment Structural Heat Sink Data 12/31/13 6-24 Accident Chronology for Limiting Break for Equipment Qualification 12/31/13 6-25 Minimum Acceptable Combinations of Containment Heat Removal Equipment Performance 12/31/13 6-26 Engineered Safety Feature Assumptions in Containment Response Analyses 12/31/08 6-27 Summary of Calculated Containment Pressures and Temperatures for Secondary System Pipe Rupture Cases 12/31/00 6-28 Steam Generator Compartment Pressure Response Flowpath Discharge Coefficients 12/31/03 6-29 Peak Pressure Mass and Energy Release Data 12/31/03 6-30 RELAP5 Long-Term Mass and Energy Release Data 12/31/13 6-31 EFLOW/FATHOMS Long-Term Mass and Energy Releases 12/31/03 6-32 Steam Line Break Mass and Energy Releases for Double-Ended Guillotine Break 12/31/05 6-33 NPSH Available and Required for LPI and BS Pumps (Limiting Flow Case)12/31/08 6-34 Deleted Per 2008 Update 12/31/03 6-35 ROTSG Peak Pressure Mass and Energy Release Data Chapter 7 12/31/12 7-1 Reactor Trip Summary 12/31/12 7-2 Engineered Safeguards Actuation Conditions 12/31/12 7-3 Engineered Safeguards Actuated Devices Effective Date 12131/14 June 30,201 Effective Table Table Title Date No.12/31/13 12/31/06 12/31/05 Chapter 8 12/31/02 12/31/00 12/31/14 12/31/14 12/31/00 12/31/00 12/31/00 Chapter 9 7-4 7-5 7-6 Characteristics of Out-of-Core Neutron Detector Assemblies NNI Inputs to Engineered Safeguards ICS Transient Limits 8-1 Loads to be Supplied from the Emergency Power Source 8-2 Single Failure Analysis for 125 Volt DC Switching Station Power Systems 8-3 Single Failure Analysis for the Keowee Hydro Station 8-4 Single Failure Analysis for the Emergency Electrical Power Systems 8-5 Single Failure Analysis for 125 Volt DC Instrumentation and Control Power System 8-6 Single Failure Analysis for the 120 Volt AC Vital Power System 8-7 125 Volt DC Panelboard Fault Analysis 12/31/00 9-1 Spent Fuel Cooling System Data, Units 1, 2 12/31/00 9-2 Spent Fuel Cooling System Data, Oconee 3 12/31/00 9-3 Component Cooling System Performance Data (For Normal Operation on a Per Oconee Basis)12/31/00 9-4 Cooling Water Systems Component Data (Component Data on a Per Unit Basis)12/31/06 9-5 Chemical Addition and Sampling System Component Data 12/31/11 9-6 High Pressure injection System Performance Data 12/31/05 9-7 High Pressure injection System Component Data 12/31/00 9-8 Low Pressure Injection System Performance Data 12/31/09 9-9 Low Pressure Injection System Component Data 12/31/00 9-10 Coolant Storage System Component Data (Component Quantities for Three Units)12/31/00 9-11 Ventilation System Major Component Data 12/31/02 9-12 Deleted Per 2002 Update 12/31/00 9-13 Component Cooling System Component Data (Component Data on a Per Unit Basis)12/31/10 9-14 SSF System Main Components Effective Date 12/31/14 June 30,201 Effective Table Table Title Date No.12/31/10 12/31/11 12/31/00 12/31/00 12/31/00 9-15 9-16 9-17 9-18 9-19 SSF Primary Valves SSF Instrumentation Design Basis Tornado Missiles And Their Impact Velocities Design Basis Tornado Missiles Minimum Barrier Thicknesses Codes and Specifications For Design of Category I Structures Chapter 10 12/31/00 12/31/09 Chapter 11 10-1 10-2 Condensate/Feedwater Reserves (each unit)Parameter Indication Location for EFW System 12/31/00 11-1 Potential Radioactive Waste Quantities from Three Units 12/31/00 11-2 Estimated Maximum Rate of Accumulation Radioactive Wastes Per Operation 12/31/00 11-3 Yearly Average Activity Concentrations in the Station Effluent for Three Units, Each Operating with One Percent Defective Fuel 12/31/04 11-4 Escape Rate Coefficients for Fission Product Release 12/31/00 11-5 Reactor Coolant Activity 12/31/12 11-6 Waste Disposal System Component Data (Component Quantities for Three Units)12/31/04 11-7 Process Radiation Monitors Chapter 12 12/31/00 12/31/00 12/31/04 Chapter 13 12/31/00 12/31/00 Chapter 14 12/31/00 12/31/00 12-1 12-2 12-3 13-1 13-2 14-1 14-2 Parameters Used for Shielding Analyses Principal Shielding Area Radiation Monitors Deleted Per 1991 Update Deleted Per 1999 Update Tests Prior to Initial Fuel Loading Postcriticality Tests Effective Date 12/31/14 June 30,201 Effective Table Table Title Date No.Chapter 15 12/31/10 12/31/13 12/31/08 12/31/04 12/31/03 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/12 12/31/12 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/03 15-1 15-2 15-3 15-4 15-5 15-6 15-7 15-8 15-9 15-10 15-11 15-12 15-13 15-14 15-15 15-16 15-17 15-18 15-19 15-20 15-21 15-22 15-23 15-24 15-25 15-26 15-27 15-28 Reg. Guide 1.183 Fuel Handling Accident Source Term Rod Ejection Accident Analysis Results Deleted Per 2008 Update Deleted Per 2004 Update Steam Line Break Accident -With Offsite Power Case Sequence of Events Summary of LOCA Break Spectrum Break Size and Type Deleted Per 1997 Up]date Deleted Per 1995 Up]date Deleted Per 1995 Up]date Deleted Per 1995 Up]date Deleted Per 1995 Up]date Deleted Per 1995 Up]date Deleted Per 1995 Up]date Deleted Per 2004 Up]date Total Core Activity for Maximum Hypothetical Accident Summary of Transient and Accident Doses Including the Effects of High Burnup Reload Cores with Replacement Steam Generators Deleted Per 2000 Up]date Deleted Per 2000 Up]date Deleted Per 1995 Up]date Deleted Per 1995 Up]date Deleted Per 1995 Up]date Deleted Per 1995 Up]date Deleted Per 1995 Up3date Deleted Per 1997 Up3date Deleted Per 2001 Update Deleted Per 1995 Up3date Deleted Per 2003 Update HPI Flow Assumed in Core Flood Line Small Break LOCA Analyses Effective Date 12/31/14 0NS-201 5-072 Document Control Desk June 30, 2015 Attachment 2, Page 14 Effective Table Table Title Date No.12/31/03 15-29 HPI Flow Assumed in RCP Discharge Small Break LOCA Analyses 12/31/03 15-30 HPI Flow Assumed in HPI Line Small Break LOCA Analyses 12/31/08 15-31 Deleted Per 2008 Update 12/31/14 15-32 Summary of Transient and Accident Cases Analyzed 12/31/13 15-33 Methodology Topical Reports and Computer Codes Used in Analyses 12/31/14 15-34 Summary of Input Parameters for Accident Analyses Using Computer Codes 12/31/12 15-35 Trip Setpoints and Time Delays Assumed in Accident Analyses 12/31/13 15-36 Startup Accident Sequence of Events 12/31/08 15-37 Rod Withdrawal at Power Accident-Peak RCS Pressure Analysis Sequence of Events 12/31/03 15-38 Rod Withdrawal at Power Accident -Core Cooling Capability Analysis Sequence of Events 12/31/08 15-39 Cold Water Accident Sequence of Events 12/31/08 15-40 Loss of Flow Accidents Four RCP Coastdown from Four RCP Initial Conditions Sequence of Events 12/31/03 15-41 Loss of Flow Accidents Two RCP Coastdown from Four RCP Initial Conditions Sequence of Events 12/31/03 15-42 Loss of Flow Accidents One RCP Coastdown from Three RCP Initial Conditions Sequence of Events 12/31/03 15-43 Loss of Flow Accidents Locked Rotor from Four RCP Initial Conditions Sequence of Events 12/31/08 15-44 Loss of Flow Accidents Locked Rotor from Three RCP Initial Conditions Sequence of Events 12/31/10 15-45 Control Rod Misalignment Accidents

-Dropped Rod Accident Sequence of Events 12/31/13 15-46 Turbine Trip Accident Sequence of Events 12/31/11 15-47 Steam Generator Tube Rupture Accident Sequence of Events 12/31/03 15-48 Steam Line Break Accident -Without Offsite Power Case Sequence of Events 12/31/08 15-49 Small Steam Line Break Accident Sequence of Events 12/31/09 15-50 Failed Fuel Source Term for the Rod Ejection Accident (Curies)Effective Date 12/31/14 ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 2, Page 15 Effective Table Table Title Date No.12/31/09 15-51 Reactor Coolant System Fission Product Source Activities

-500 EPFD Equilibrium Cycle 12/31/03 15-52 Deleted Per 2003 Update 12/31/03 15-53 Deleted Per 2003 Update 12/31/03 15-54 Deleted Per 2003 Update 12/31/03 15-55 Deleted Per 2003 Update 12/31/14 15-56 Deleted Per 2014 Update 12/31/14 15-57 Deleted Per 2014 Update 12/31/03 15-58 Parameters Used To Determine Hydrogen Generation 12/31/01 15-59 Deleted Per 2001 Update 12/31/14 15-60 Deleted Per 2014 Update 12/31/09 15-61 Control Room Atmospheric Dispersion Factors (x/Qs)12/31/14 15-62 Results of LBLOCA AnalYSeS for Mark-B-HTP Full Core Sequence of Events 12/31/14 15-63 Results of LBLOCA Analyses for Full Core Mark-B-HTP; Gadolinia Fuel Pins 12/31/11 15-64 Results of 102% FP SBLOCA Analyses for Full Core Mark-B-HTP 12/31/12 15-65 Dose Equivalent Iodine (DEl) Calculation 12/31/12 15-66 Dose Equivalent Xenon (DEX) Calculation 12/31/14 15-67 Peak Cladding Temperature Results of LBLOCA Reanalyses for Mark-B-HTP Full Core to Address Error Corrections, UO 2 Fuel Pins 12/31/14 15-68 Peak Cladding Temperature Results of LBLOCA Reanalyses for Mark-B-HTP Full Core to Address Error Corrections, Gadolinia Fuel Pins Chapter 18 12/31/14 18-1 Summary Listing of the Programs, Activities and TLAA Effective Date 12/31/14 ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 2, Page 16 OCONEE UFSAR -2014 UPDATE List of Effective Pages (LOEP) for Figures The purpose of this list is to assure that the pages in the Figures section of your manual match the most recent issue as well as to show a full accounting of all figures, including those that have been deleted. The earliest effective date, 12/31/00, was used when all figures were re-issued.Effective Figure Figure Title Date No.Chapter 1 12/31/00 1-1 Duke Power Service Area 12/31/00 1-2 General Arrangement, Floor Plan Elevation 758+0 12/31/00 1-3 General Arrangement, Floor Plan Elevation 771 +0 and Elevation 775+0 12/31/00 1-4 General Arrangement, Floor Plan Elevation 783+9 12/31/04 1-5 General Arrangement, Floor Plan Elevation 796+6 12/31/00 1-6 General Arrangement, Floor Plan Elevation 809+3 12/31/07 1-7 General Arrangement, Floor Plan Elevation 822+0 12/31/00 1-8 General Arrangement, Floor Plan Elevation 838+0 and Elevation 844+0 12/31/00 1-9 General Arrangement, Sections Chapter 2 12/31/00 2-1 General Location 12/31/00 2-2 Topography within 5 Miles 12/31/00 2-3 General Area Map 12/31/00 2-4 Site Plan 12/31/11 2-5 Radioactive Effluent Site Boundaries 12/31/00 2-6 Population Centers within 100 Miles 12/31/08 2-7 Forecast of High-Pollution-Potential Days in the U.S.12/31/08 2-8 Annual Surface Wind Rose for Greenville, South Carolina, WBAS (1959-1 963)12/31/08 2-9 Upper Air Wind Rose-Athens, Georgia. 800-1300 ft above ground. (Dec 1954- Nov 1961)12/31/08 2-10 Upper Air Wind Rose-Athens, Georgia. 2300-2800 ft above ground.(Dec 1959 -Nov 1961)12/31/08 2-11 Cumulative Probability of Wind Direction Persistence Duration at Greenville, SC Effective Date 12/31/14 ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 2, Page 17 Effective Figure Figure Title Date No.1 2/31/08 2-12 Precipitation Surface Wind Rose for Greenville, South Carolina, WBAS (1959- 1963)12/31/08 2-13 Surface Wind Direction Frequency Distribution-During Low-Level Temperature Inversion Conditions 12/31/08 2-14 Maximum Topographic Elevation versus Distance (NNE and N sectors)12/31/08 2-15 Maximum Topographic Elevation versus Distance (NE sector)12/31/08 2-16 Maximum Topographic Elevation versus Distance (ENE sector)12/31/08 2-17 Maximum Topographic Elevation versus Distance (ESE and E sectors)12/31/08 2-18 Maximum Topographic Elevation versus Distance (SSE and SE sectors)12/31/08 2-19 Maximum Topographic Elevation versus Distance (SSW and S sectors)12/31/08 2-20 Maximum Topographic Elevation versus Distance (WSW and SW sectors)12/31/08 2-21 Maximum Topographic Elevation versus Distance (WNW and W sectors)12/31/08 2-22 Maximum Topographic Elevation versus Distance (NW sector).12/31/08 2-23 Maximum Topographic Elevation versus Distance (NWW sector)12/31/08 2-24 Relative Elevations of Meteorological 12/31/08 2-25 Annual Surface Wind Rose (October 19, 1966 -October 31, 1967)12/31/08 2-26 Precipitation Surface Wind Rose (October 19,. 1966 -October 31, 1967)12/31/08 2-27 Surface Wind Frequency Distribution during Low-Level Temperature Inversion Conditions (October 19, 1966 -October 31, 1967)12/31/08 2-28 Wind Rose for Tower Winds (June 19, 1967 -May 31, 1968)12/31/08 2-29 Frequency Distribution for Tower Winds During Low-Level Temperature Inversion Conditions (June 19, 1967 -May 31, 1968)12/31/08 2-30 Precipitation Wind Rose for Tower Winds (June 19, 1967 -May 31, 1968)12/31/08 2-31 General Building Arrangements 12/31/08 2-32 Plot Plan and Site Boundary During Early Meteorological Studies 12/31/08 2-33 SF 6 Gas Tracer Test Background Sample Points 12/31/08 2-34 SF 6 Gas Tracer Test Release Point 12/31/08 2-35 Deleted Per 2008 Update 12/31/08 2-36 Deleted Per 2008 Update Effective Date 12/31/14 Atachete 2,Nag.1 Effective Figure Figure Title Date No.12/31/08 2-37 SF 6 Gas Tracer Test Release and Sample Stations 12/31/08 2-38 Approximate Terrain at Nuclear Site 12/31/00 2-39 Location of Municipal Water Supply Intakes 12/31/00 2-40 Areal Groundwater Survey 12/31/00 2-41 Groundwater Survey at Station Site 12/31/00 2-42 Well Permeameter Test Apparatus 12/31/00 2-43 Formulae for Determining Permeability 12/31/00 2-44 Regional Geologic Map 12/31/00 2-45 Topographic Map of Area 12/31/00 2-46 Location and Topographic Map 12/31/00 2-47 Strike and Dip of Joint Pattern 12/31/00 2-48 Earthquake Epicenters 12/31/00 2-49 Regional Techtonics 12/31/00 2-50 Ground Motion Spectra 12/31/00 2-51 Recommended Response Spectra 12/31/00 2-52 Ground Motion Spectra 12/31/00 2-53 Recommended Response Spectra 12/31/00 2-54 Ground Motion Spectra 12/31/00 2-55 Recommended Response Spectra 12/31/00 2-56 Subsurface Profile 12/31/00 2-57 Subsurface Profile 12/31/00 2-58 Subsurface Profile 12/31/00 2-59 Subsurface Profile 12/31/00 2-60 Subsurface Profile 12/31/00 2-61 Subsurface Profile 12/31/00 2-62 Subsurface Profile 12/31/00 2-63 Subsurface Profile 12/31/00 2-64 Subsurface Profile 12/31/00 2-65 Boring Plan 12/31/00 2-66 Core Boring Record, Boring Log NA-i Effective Date 12/31/14 ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 2, Page 19 Effective Date 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00*12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 rigure No.l-gure ti tie 2-67 2-68 2-69 2-70 2-71 2-72 2-73 2-74 2-75 2-76 2-77 2-78 2-79 2-80 2-81 2-82 2-83 2-84 2-85 2-86 2-87 2-88 2-89 2-90 2-91 2-92 2-93 2-94 2-95 2-96 Core Boring Record, Boring Log NA-i Core Boring Record, Boring Log NA-2 Core Boring Record, Boring Log NA-2 Core Boring Record, Boring Log NA-3 Core Boring Record, Boring Log NA-3 Core Boring Record, Boring Log NA-4 Core Boring Record, Boring Log NA-4 Core Boring Record, Boring Log NA-5 Core Boring Record, Boring Log NA-5 Core Boring Record, Boring Log NA-6 Core Boring Record, Boring Log NA-6 Core Boring Record, Boring Log NA-7 Core Boring Record, Boring Log NA-7 Core Boring Record, Boring Log NA-8 Core Boring Record, Boring Log NA-8 Core Boring Record, Boring Log NA-9 Core Boring Record, Boring Log NA-9 Core Boring Record, Boring Log NA-9 Core Boring Record, Boring Log NA-10 Core Boring Record, Boring Log NA-10 Core Boring Record, Boring Log NA-10 Core Boring Record, Boring Log NA-i11 Core Boring Record, Boring Log NA-i11 Core Boring Record, Boring Log NA-12 Core Boring Record, Boring Log NA-12 Core Boring Record, Boring Log NA-13 Core Boring Record, Boring Log NA-13 Core Boring Record, Boring Log NA-14 Core Boring Record, Boring Log NA-14 Core Boring Record, Boring Log NA-15 Effective Date 12/31/14 Atachete 2,Nag.2 Effective Figure Figure Title Date No.12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 Chapter 3 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 2-97 2-98 2-99 2-100 2-101 2-102 2-103 2-104 2-105 2-106 2-107 2-108 2-109 2-110 2-111 2-112 2-113 2-114 2-115 2-116 2-117 2-118 Core Boring Record, Boring Log NA-15 Core Boring Record, Boring Log NA-16 Core Boring Record, Boring Log NA-16 Core Boring Record, Boring Log NA-16 Core Boring Record, Boring Log NA-17 Core Boring Record, Boring Log NA-17 Core Boring Record, Boring Log NA-17 Core Boring Record, Boring Log NA-18 Core Boring Record, Boring Log NA-18 Core Boring Record, Boring Log NA-18 Core Boring Record, Boring Log NA-18 Core Boring Record, Boring Log NA-19 Core Boring Record, Boring Log NA-19 Core Boring Record, Boring Log NA-19 Core Boring Record, Boring Log NA-19 Core Boring Record, Boring Log NA-20 Core Boring Record, Boring Log NA-20 Core Boring Record, Boring Log NA-20 Core Boring Record, Boring Log NA-21 Core Boring Record, Boring Log NA-21 Seismic Field Work Location Map Diagrammatic Cross Section through Seismic Lines 3-1 3-2 3-3 3-4 3-5 3-6 3-7 Frequency and Mode Shapes -Auxiliary Building -North South Direction Frequency and Mode Shapes -Auxiliary Building -East West Direction Auxiliary Building Mass Model Auxiliary Building -East West Direction

-Seismic Model Results Auxiliary Building -North South Direction

-Seismic Model Results Example Spectrum Curves Reactor Building -Seismic Model Results Effective Date 12131114 ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 2, Page 21 Effective Figure Figure Title Date No.12/31/00 3-8 Reactor Building -Seismic Model Results 12/31/06 3-9 Main Steam System West Generator Problem Number 1-01-08 12/31/06 3-10 Core Flooding Tank lA Problem Number 1-53-9 12/31/06 3-11 Low Pressure Injection System West Generator Problem Number 1-53-9 12/31/06 3-12 RCP Piping to HPI Letdown Coolers Problem Number 1-55-03 12/31/06 3-13 RCP Piping to HPI Letdown Coolers Problem Number 1-55-03 12/31/06 3-14 RCP Piping to HPI Letdown Coolers Problem Number 1-55-03 12/31/06 3-15 RCP Piping to HPI Letdown Coolers Problem Number 1-55-03 12/31/00 3-16 Seismic Analysis of Component Coolers 12/31/00 3-17 Seismic Analysis of Component Coolers 12/31/00 3-18 Seismic Analysis of Component Coolers 12/31/00 3-19 Reactor Building Typical Details 12/31/11 3-20 Typical Electrical and Piping Penetrations 12/31/00 3-21 Details of Equipment Hatch and Personnel Hatch 12/31/00 3-22 Reactor Building Finite Element Mesh 12/31/00 3-23 Reactor Building Finite Element Mesh 12/31/00 3-24 Reactor Building Thermal Gradient 12/31/00 3-25 Reactor Building Isostress Plot Wall and Dome 12/31/00 3-26 Reactor Building Isostress Plot Wall and Base 12/31/00 3-27 Reactor Building Finite Element Mesh Wall Buttresses 12/31/00 3-28 Reactor Building Isostress Plot for Buttresses 12/31/00 3-29 Temperature Gradient at Buttress 12/31/00 3-30 Buttress Reinforcing Details 12/31/00 3-31 Reactor Building Equipment Hatch Mesh 12/31/00 3-32 Reactor Building Penetration Loads 12/31/00 3-33 Reactor Building Model for Liner Plate Analysis for Radial Displacement 12/31/00 3-34 Reactor Building Model for Liner Analysis for Anchor Displacement 12/31/00 3-35 Reactor Building -Results from Tests on Liner Plate Anchors 12/31/00 3-36 Location of Plugged Sheaths Effective Date 12/31/14 ONS-201 5-072 Document Control Desk.June 30, 2015 Attachment 2, Page 22 Effective Figure Figure Title Date No.12/31/00 3-37 Reactor Building Instrumentation for Unit 1 12/31/00 3-38 Turbine Building Cross-Section at Line 21 12/31/00 3-39 Deleted Per 1996 Update 12/31/00 3-40 Deleted Per 1996 Update 12/31/00 3-41 Deleted Per 1996 Update 12/31/00 3-42 Deleted Per 1996 Update 12/31/00 3-43 Deleted Per 1996 Update 12/31/00 3-44 Deleted Per 1996 Update 12/31/00 3-45 Deleted Per 1996 Update 12/31/00 3-46 Deleted Per 1996 Up3date 12/31/00 3-47 Deleted Per 1996 Update 12/31/00 3-48 Deleted Per 1996 Update 12/31/00 3-49 Deleted Per 2004 Up0date 12/31/00 3-50 Deleted Per 2004 Up3date 12/31/00 3-51 Deleted Per 2004 Up3date 12/31/03 3-52 Seismic, Thermal, and Dead Load Analytical Model for the Pressurizer Surge Line Piping (Units 2 and 3)12/31/03 3-53 Deleted Per 2003 12/31/03 3-54 Deleted Per 2003 Up3date 12/31/03 3-55 Deleted Per 2004 Up3date 12/31/03 3-56 Deleted Per 2004 Up3date 12/31/00 3-57 Directions and Velocities of the Coolant Flow in the Reactor 12/31/00 3-58 Location of Instrumentation Surveillance Specimen Holder Tubes and the Plenum Cylinder Tubes 12/31/00 3-59 Location of the Instrumentation in the Specimen Holder Tube 12/31/00 3-60 Location of the Accelerometer in Plenum Cylinder Tube Chapter 4 12/31/00 4-1 Burnable Poison Rod Assembly 12/31/00 4-2 Deleted Per 1999 Update 12/31/00 4-3 Deleted Per 1999 Update Effective Date 12/31114 ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 2, Page 23 Effective Figure Figure Title Date No.12/31/00 4-4 Typical Pressurized Fuel Rod 12/31/12 4-5 Typical Boron Concentration Versus Core Life 12/31/12 4-6 Typical BPRA Concentration and Distribution 12/31/00 4-7 Typical Control Rod Locations and Groupings 12/31/00 4-8 Typical Uniform Void Coefficient 12/31/00 4-9 Deleted Per 1995 Update 12/31/00 4-10 Typical Rod Worth Versus Distance Withdrawn 12/31/00 4-11 Percent Neutron Power Versus Time Following Trip 12/31/00 4-12 Power Spike Factor Due to Fuel Densification 12/31/00 4-13 Power Peaking Caused by Dropped Rod (Oconee Unit 1, Cycle 1)12/31/00 " 4-14 Azimuthal Stability Index Versus Moderator Coefficient From Three Dimensional Case (Oconee Unit 1, Cycle 1)12/31/00 4-15 Azimuthal Stability Index with Compounded Error Versus Moderator Coefficient Calculated From Three Dimensional Case (Oconee Unit 1, Cycle 1 )12/31/00 4-16 Azimuthal Stability Index Versus Moderator Coefficient From Three Dimension Case (Oconee Unit 2, Cycle 1 )12/31/00 4-17 Azimuthal Stability Index with Compounded Error Versus Moderator Coefficient From Three Dimensional Case (Oconee Unit 2, Cycle 1)12/31/00 4-18 Deleted Per 1997 Update 12/31/00 4-19 Deleted Per 1995 Update 12/31/00 4-20 Deleted Per 1995 Update 12/31/00 4-21 Flow Regime Map for the Hot Unit Cell 12/31/00 4-22 Flow Regime Map for the Hot Control Rod Cell 12/31/00 4-23 Flow Regime Map for the Hot Wall Cell 12/31/00 4-24 Flow Regime Map for the Hot Corner Cell 12/31/00 4-25 Deleted Per 1996 Update 12/31/00 4-26 Reactor Vessel and Internals General Arrangement 12/31/00 4-27 Reactor Vessel and Internals Cross Section 12/31/00 4-28 Core Flooding Arrangement 12/31/00 4-29 Internals Vent Valve Clearance Gaps 12/31/00 4-30 Internals Vent Valve Effective Date 12/31/14 Atachete 2,Nag2 Effective Figure Figure Title Date No.12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/04 Chapter 5 4-31 4-32 4-33 4-34 4-35 4-36 4-37 4-38 Control Rod Assembly Axial Power Shaping Rod Assembly Deleted Per 1999 Update Control Rod Drive -General Arrangement Deleted Per 1999 Update Deleted Per 1999 Update Typical Fuel Assembly Westinghouse 177 Fuel Assembly 12/31/00 5-1 Reactor Coolant System (Unit 1)12/31/00 5-2 Reactor Coolant System (Units 2 & 3)12/31/00 5-3 Reactor Coolant System, Arrangement Plan (Unit 1)12/31/00 5-4 Reactor Coolant System, Arrangement Elevation (Unit 1)12/31/04 5-5 Reactor Coolant System, Arrangement Plan (Unit 2)12/31/04 5-6 Reactor Coolant System, Arrangement Elevation (Unit 2)12/31/04 5-7 Reactor Coolant System, Arrangement Plan (Unit 3)12/31/04 5-8 Reactor Coolant System, Arrangement Elevation (Unit 3)12/31/04 5-9 Reactor and Steam Temperatures versus Reactor Power (Replacement Steam Generators) 12/31/00 5-10 Points of Stress Analysis for Reactor Vessel 12/31/04 5-11 Location of Replacement Steam Generator Weld 12/31/00 5-12 Deleted Per 1991 Update 12/31/00 5-13 Deleted Per 1991 Update 12/31/03 5-14 Reactor Vessel Outline (Unit 1 ) (Shown with original reactor vessel head)12/31/03 5-15 Reactor Vessel Outline (Unit 2) (Shown with original reactor vessel head)12/31/03 5-16 Reactor Vessel Outline (Unit 3) (Shown with original reactor vessel head)12/31/00 5-17 Reactor Coolant Controlled Leakage Pump (Unit 1)12/31/00 5-18 Reactor Coolant Pump Estimated Performance Characteristic (Unit 1 )Effective Date 12/31/14 ONS-20 15-072 Document Control Desk June 30, 2015 Attachment 2, Page 25 Effective Date 12/31/00 12/31/00 12/31/04 12/31/04 12/31/00 12/31/00 12/31/04 12/31/04 12/31/00 12/31/00 12/31/08 12/31/03 12/31/03 12/31/03 12/31/03 Chapter 6 12/31/05 12/31/02 12/31/06 12/31/06 12/31/00 12/31/00 12/31/00 12/31/00 12/31/09 12/31/00 12/31/00 12/31/00 12/31/00 Figure No.Figure Title 5-19 5-20 5-21 5-22 5-23 5-24 5-25 5-26 5-27 5-28 5-29 5-30 5-31 5-32 5-33 6-1 6-2 6-3 6-4 6-5 6-6 6-7 6-8 6-9 6-10 6-11 6-12 6-13 Reactor Coolant Pump (Units 2, 3)Reactor Coolant Pump Estimated Performance Characteristic (Units 2, 3)Flow Diagram of Bingham Reactor Coolant Pump -Piping Diagram Flow Diagram of Bingham Reactor Coolant Pump -Piping Diagram Code Allowables and Reinforcing Limits Nozzles and Bowls Code Allowables, Cover Steam Generator Outline Deleted Per 2004 Update Turbine Generator Speed Response Following Load Rejection Pressurizer Outline Reactor Coolant System Arrangement Elevation (Typical)Reactor Coolant System Arrangement

-Plan (Typical)Jet Impingement Load on the Replacement Steam Generator Deleted Per 2003 Update Replacement Reactor Vessel Closure Head Outline Flow Diagram of Emergency Core Cooling System Flow Diagram of Reactor Building Spray System Reactor Building Cooling Schematic Reactor Building Purge and Penetration Ventilation System Reactor Building Spray Pump Characteristics Reactor Building Cooler Heat Removal Capacity Reactor Building Cooler Heat Removal Capability as a Function of Air-Steam Mixture Flow Reactor Building Post-Accident Steam-Air Mixture Composition Reactor Building Isolation Valve Arrangements Deleted Per 1993 Update Deleted Per 1993 Update Deleted Per 1993 Update Deleted Per 1999 Update Effective Date 12/31/14 I ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 2, Page 26 Effective Figure Figure Title Date No..12/31/00 6-14 Deleted Per 1999 Update 12/31/00 6-15 Deleted Per 1991 Update 12/31/00 6-16 High Pressure Injection Pump Characteristics 12/31/00 6-17 Low Pressure Injection Pump Characteristics 12/31/00 6-18 Low Pressure Injection Cooler Capacity 12/31(07 6-19 Control Rooms 1-2 And 3 Locations 12/31/00 6-20 General Arrangement Control Room 1-2 12/31/00 6-21 General Arrangement Control Room 3 12/31/00 6-22 Penetration Room Ventilation Fan And System Characteristics 12/31/00 6-23 Penetrations In Penetration Room 809'3" Floor And Wall Areas 12/31/00 6-24 Penetrations In Penetration Room 838'0" Floor 12/31/00 6-25 Penetration Rooms Details, Mechanical Openings 12/31/00 6-26 Penetration Rooms Details, Electrical Openings 12/31/00 6-27 Penetration Rooms Details Construction Details 12/31/03 6-28 ONS ROTSG Peak Pressure Analysis 12/31/03 6-29 ONS ROTSG Peak Pressure Analysis 12/31/03 6-30 ONS ROTSG Peak Pressure Analysis 12/31/03 6-31 ONS ROTSG Peak Pressure Analysis 12/31/03 6-32 ONS ROTSG Peak Pressure Analysis 12/31/03 6-33 ONS ROTSG Peak Pressure Analysis 12/31/03 6-34 ONS ROTSG Peak Pressure Analysis 12/31/03 6-35 ONS ROTSG Peak Pressure Analysis 12/31/13 6-36 Oconee Large Break LOCA Long-term Containment Response 12/31/13 6-37 Oconee Large Break LOCA Long-term Containment Response 12/31/03 6-38 Deleted Per 2003 Update 12/31/03 6-39 Deleted Per 2003 Update 12/31/03 6-40 Deleted Per 2003 Update 12/31/03 6-41 Deleted Per 2003 Update 12/31/08 6-42 Oconee Steam Line Break: Containment Pressure 12/31/08 6-43 Oconee Steam Line Break: Containment Temperature Effective Date 12/31/14 Atachete 2,Nag.2 Effective Figure Figure Title Date No.12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/03 12/31/00 12/31/00 12/31/00 12/31/00 Chapter 7 12/31/13 12/31/00 12/31/03 12/31/12 12/31/13 12/31/13 12/31/00 12/31/13 12/31/11 12/31/13 12/31/00 12/31/09 12/31/09 12/31/00 12/31/00 12/31/00 12/31/00 6-44 LOCA-Mass Release for the Subcompartment Pressure Response Analysis 6-45 LOCA-Energy Release Rate for the Subcompartment Pressure Response Analysis 6-46 LOCA-Reactor Compartment Pressure Response 6-47 LOCA-Steam Generator Compartment Vent Discharge Coefficient 6-48 LOCA-Steam Generator Compartment Pressure Response 6-49 Deleted Per 2003 Update 6-50 LOCA-Mass Released to the Reactor Building 6-51 LOCA-Energy Released to the Reactor Building 6-52 LOCA-Reactor Building Pressure 6-53 Deleted Per 1997 Update 7-1 7-2 7-3 7-4 7-5 7-6 7-7 7-8 7-9 7-10 7-11 7-12 7-13 7-14 7-15 7-16 7-17 Reactor Protective System Typical Pressure Temperature Boundaries Typical Power Imbalance Boundaries Rod Control Drive Controls Engineered Safeguards Protection System Nuclear Instrumentation System Nuclear Instrumentation Flux Range Nuclear Instrumentation Detector Locations Nuclear Instrumentation Detector Locations

-(Unit 1)Nuclear Instrumentation Detector Locations

-(Unit 2 & 3)Automatic Control Rod Groups -Typical Worth Value Versus Distance Withdrawn Control Rod Drive Logic Diagram*Control Rod Electrical Block Diagram Integrated Control System Core Thermal Power Demand -Integrated Control System Integrated Master -Integrated Control System Feedwater Control -Integrated Control System Effective Date 12/31/14 Atachete 2,Nag.2 Effective Figure Figure Title Date No.12/31/04 12/31/00 12/31/00 12/31/00 12/31100 12/31/00 12/31/00 12/31/00 12/31/00 Chapter 8 12/31/13 12/31/00 12/31/14 12/31/14 12/31/13 12/31/13 12/31/06 12/31/00 12/31/00 Chapter 9 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 12/31/00 7-18 7-19 7-20 7-21 7-22 7-23 7-24 7-25 7-26 8-1 8-2 8-3 8-4 8-5 8-6 8-7 8-8 8-9 9-1 9-2 9-3 9-4 9-5 9-6 9-7 9-8 9-9 Reactor and Steam Temperatures Versus Reactor Power (Replacement Steam Generator)

Reactor Control -Integrated Control System Incore Detector Locations Incore Monitoring Channel Deleted Per 1997 Update Deleted Per 1997 Update Deleted Per 1997 Update Deleted Per 1997 Update Control Room Layout Single Line Diagram Site Transmission Map Typical 6900 Volt and 4160 Volt Unit Auxiliary

-Single Line Diagram Typical 600 Volt and 208 Volt ESG Auxiliaries

-Single Line Diagram Typical DC and AC Vital Power System -Single Line Diagram Keowee DC Power System -Single Line Diagram 230 KV SWYD One Line 125V DC Deleted Per 1997 Update 125/250 VDC Station Aux. Circuits Fuel Storage Rack (Module)Fuel Storage Rack (Assembly)

Spent Fuel Pool Outline Oconee 1, 2 Spent Fuel Pool Outline Oconee 3 Spent Fuel Cooling System Deleted Per 1990 Update Fuel Handling System (Units 1&2 -Page 1 and Unit 3 -Page 2)Component Cooling System Condenser Circulating Water System Effective Date 12/31/14 ONS-201 5-072 Document Controi Desk June 30, 2015 Attachment 2, Page 29 Effective Figure Figure Title Date No.1 2131106 9-10 High Pressure Service Water System 12/31/07 9-11 Low Pressure Service Water System 12/31/14 9-12 Low Pressure Service Water System 12/31/00 9-13 Recirculated Cooling Water System 12/31/00 9-14 Deleted Per 1997 Update 12/31/03 9-15 Chemical Addition and Sampling System 12/31/07 9-16 Chemical Addition and Sampling System 12/31/14 9-17 High Pressure Injection System 12/31/00 9-18 High Pressure Injection System 12/31/04 9-19 Low Pressure Injection System 12/3 1/04 9-20 Coolant Storage System 12/31/01 9-21 Coolant Treatment System 12/31/00 9-22 Post-Accident Liquid Sample System 12/31/00 9-23 Post-Accident Containment Air Sample System 12/31/12 9-24 Control Room Area Ventilation and Air Conditioning System 12/31/00 9-25 Spent Fuel Pool Ventilation System Unit 1 and 2 12/31/00 9-26 Spent Fuel Pool Ventilation System Unit 3 12/31/11 9-27 Auxiliary Building Ventilation System Unit I and 2 12/31/01 9-28 Auxiliary Building Ventilation System Unit 3 12/31/00 9-29 ,Deleted Per 1998 Update 12/31/13 9-30 SSF General Arrangements Longitudinal Section 12/31/13 9-31 SSF General Arrangements Plan Elevation 777' and 754'12/31/13 9-32 SSF General Arrangements Plan Elevation 797+0 12/31/13 9-33 SSF General Arrangements Plan Elevation 817+0 12/31/13 9-34 SSF General Arrangements Transverse Section 12/31/00 9-35 SSF RC Makeup System 12/31/13 9-36 SSF Auxiliary Service Water System 12/31/13 9-37 SSF HVAC Service Water System & SSF Diesel Cooling Water System 12/31/13 9-38 SSF Diesel Air Starting System 12/31/00 9-39 SSF Sump System Effective Date 12/31/14 Atachete 2,Nag.3 Effective Figure Figure Title Date No.12131113 12/31/00 12/31/00 12/31/00 12/31/14 12/31/14 12/31/14 9-40 9-41 9-42 9-43 9-44 9-45 9-46 SSF 41 60V/600V/208V Electrical Distribution SSF 125 VDC Auxiliary Power Systems Essential Siphon Vacuum System Siphon Seal Water System Protected Service Water PSW AC Electrical Distribution PSW DC Electrical Distribution Chapter 10 12/31/07 12/31/00 12/31/00 12/31/02 12/31/00 12/31/07 12/31/00 12/31/14 12/31/04 Chapter 11 12/31/00 12/31/00 12/31/00 12/3 1/1 1 12/31/00 12/31/00 Chapter 13 12/31/14 12/31/00 12/31/14 12/31/14 10-1 10-2 10-3 10-4 10-5 10-6 10-7 10-8 10-9 11-1 11-2 11-3 11-4 11-5 11-6 13-1 13-2 13-3 13-4 Main Steam and Auxiliary Steam System High Pressure Turbine Exhaust and Steam Seal System High Pressure Turbine Exhaust and Steam Seal System Moisture Separator and Reheater Heater and Drain System Vacuum System Condensate System Main Feedwater System Emergency Feedwater System OTSG Recirculation System 3" Liquid Waste Discharge Liquid Waste Disposal System Gaseous Waste Disposal System Waste Water Collection Basin Deleted Per 1999 Update Deleted Per 1997 Update Duke Energy Corporation Structure Deleted Per 1999 Update Nuclear Generation Department Nuclear Generation

-Oconee Nuclear Site Effective Date 12/31/14 Atachete 2,Nag.3 Effective Figure Figure Title Date No.12/31/00 12/31/00 12/31/12 12/31/14 13-5 13-6 13-7 13-8"At the Controls" Definition

-Unit 1 & 2"At the Controls" Definition

-Unit 3 Deleted Per 2012 Update Nuclear & PMC Organizational Structure Chapter 15 12/31/13 12/31/13 12/31/13 12/31/13 12/31 /13 12/31/13 12/31/00 12/31/00 12/31/00 12/31/00 12/31/03 12/31/03 12/31/03 12/31/03 12/31/03 12/31/03 12/31/03 12/31/08 12/31/08 15-1 Startup Accident 15-2 Startup Accident 15-3 Startup Accident 15-4 Startup Accident 15-5 Startup Accident 15-6 Startup Accident 15-7 Deleted Per 1998 Update 15-8 Deleted Per 1998 Update 15-9 Deleted Per 1998 Update 15-10 Deleted Per 1998 Update 15-11 Rod Withdrawal at Power Accident -Peak RCS Pressure Analysis Power 15-12 Rod Withdrawal at Power Accident -Peak ROS Pressure Analysis ROS Temperatures 15-13 Rod Withdrawal at Power Accident -Peak RCS Pressure Analysis Pressurizer Level 15-14 Rod Withdrawal at Power Accident -Peak RCS Pressure Analysis RCS Pressure 15-15 Rod Withdrawal at Power Accident -Core Cooling Capability Analysis Power 15-16 Rod Withdrawal at Power Accident -Core Cooling Capability Analysis RCS Temperatures 15-17 Rod Withdrawal at Power Accident -Core Cooling Capability Analysis Pressurizer Level 15-18 Cold Water Accident -RCS Flow 15-19 Loss of Coolant Flow Accidents

-Four RCP Coastdown From Four RCP Initial Conditions Analysis -RCS Flow Effective Date 12/31/14 ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 2, Page 32 Effective Figure Figure Title Date No.12/31/08 15-20 Loss of Coolant Flow Accidents

-Four RCP Coastdown From Four ROP Initial Conditions Analysis -Power 12/31/08 15-21 Loss of Coolant Flow Accidents

-Four RCP Coastdown From Four RCP. Initial Conditions Analysis -RCS Temperature 12/31/08 15-22 Loss of Coolant Flow Accidents

-Four RCP Coastdown From Four RCP Initial Conditions Analysis -Pressurizer Level 12/31/08 15-23 Loss of Coolant Flow Accidents

-Four RCP Coastdown From Four RCP Initial Conditions Analysis -RCS Pressure 12/31/13 15-24 Loss of Coolant Flow Accidents

-Four RCP Coastdown From Four RCP Initial Conditions Analysis -DNBR 12/31/03 15-25 Loss of Coolant Flow Accidents

-Two RCP Coastdown From Four RCP Initial Conditions Analysis -RCS Flow 12/31/10 15-26 Control Rod Misalignment Accidents

-Dropped Rod Analysis -Neutron Power 12/31/10 15-27 Control Rod Misalignment Accidents

-Dropped Rod Analysis -RCS Temperatures 12/31/10 15-28 Control Rod Misalignment Accidents

-Dropped Rod Analysis -Pressurizer Level 12/31/13 15-29 Rod Ejection Accident -BOC Four RCPs -Power 12/31/13 15-30 Rod Ejection Accident -BOC Three RCPs -Power 12/31/13 15-31 Rod Ejection Accident -BOC HZP -Power 12/31/13 15-32 Rod Ejection Accident -EOC Four RCPs -Power 12/31/13 15-33 Rod Ejection Accident -EOC Three RCPs -Power 12/31/13 15-34 Rod Ejection Accident -EOC HZP -Power 12/31/13 15-35 Deleted Per 2013 Update 12/31/13 15-36 Rod Ejection Accident -BOC Four RCPs -RCS Pressure 12/31/00 15-37 Deleted Per 1999 Update 12/31/00 15-38 Deleted Per 1999 Update 12/31/00 15-39 Deleted Per 1999 Update 12/31/03 15-40 Steam Line Break Accident -With Offsite Power -Steam Line Pressure 12/31/03 15-41 Steam Line Break Accident -With Offsite Power -Break Flowrate 12/31/03 15-42 Steam Line Break Accident -With Offsite Power -RCS Temperature 12/31/03 15-43 Steam Line Break Accident -With Offsite Power -Reactivity Effective Date 12/31/14 ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 2, Page 33 Effective Figure Figure Title Date No.12/31/00 15-44 LOCA -Large Break Analysis Code Interfaces 12/31/00 15-45 Deleted Per 2000 Update 12/31/00 15-46 Deleted Per 1990 Update 12/31/00 15-47 Deleted Per 1997 Update 12/31/00 15-48 Deleted Per 1997 Update 12/31/00 15-49 Deleted Per 2000 Update 12/31/08 15-50 LOCA -Peak Cladding Temperature vs. Break Size for LBLOCA Spectrum 12/31/00 15-51 Deleted Per 1997 Update 12/31/00 15-52 Deleted Per 1995 Update 12/31/00 15-53 Deleted Per 1995 Update 12/31/00 15-54 Deleted Per 1995 Update 12/31/00 15-55 Deleted Per 1995 Update 12/31/00 15-56 Deleted Per 1995 Update 12/31/00 15-57 Deleted Per 1995 Update 12/31/00 15-58 Deleted Per 1995 Update 12/31/00 15-59 Deleted Per 1995 Update 12/31/00 15-60 Deleted Per 1995 Update 12/31/00 15-61 Deleted Per 1995 Update 12/31/00 15-62 Deleted Per 1995 Update 12/31/00 15-63 Deleted Per 1995 Update 12/31/00 15-64 Deleted Per 1995 Update 12/31/00 15-65 Deleted Per 1995 Update 12/31/00 15-66 Deleted Per 1995 Update 12/31/00 15-67 Deleted Per 1995 Update 12/31/00 15-68 Deleted Per 1995 Update 12/31/00 15-69 Deleted Per 1995 Update 12/31/00 15-70 Deleted Per 1995 Update 12/31/00 15-71 Deleted Per 1995 Update 12/31(00 15-72 Deleted Per 1995 Update Effective Date 12/31/14 ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 2, Page 34 Effective Figure Figure Title Date No.12/31/00 15-73 Deleted Per 1995 Update 12/31/00 15-74 Deleted Per 1995 Update 12/31/00 15-75 Deleted Per 1995 Update 12/31/00 15-76 Deleted Per 1995 Update 12/31/00 15-77 Deleted Per 1995 Update 12/31/00 15-78 Deleted Per 1995 Update 12/31/00 15-79 Deleted Per 1995 Up3date 12/31/00 15-80 MHA- Integrated Direct Dose 12/31/00 15-81 Deleted Per 1995 Update 12/31/00 15-82 Deleted Per 2000 Up3date 12/31/00 15-83 Deleted Per 1995 Update 12/31/00 15-84 Deleted Per 2000 Up3date 12/31/00 15-85 Deleted Per 2000 Up3date 12/31/00 15-86 Deleted Per 1997 Update 12/31/00 15-87 Deleted Per 2000 Up)date 12/31/00 15-88 Deleted Per 1995 Update 12/31/03 15-89 Post-Accident Hydrogen Control- Reactor Building Arrangement 12/31/00 15-90 Deleted Per 1995 Update 12/31/00 15-91 Deleted Per 1995 Update 12/31/00 15-92 Deleted Per 1995 Up)date 12/31/00 15-93 Deleted Per 1995 Up)date 12/31/00 15-94 Deleted Per 1995 Update 12/31/00 15-95 Deleted Per 1995 Update 12/31/00 ,15-96 Deleted Per 1995 Up)date 12/31/00 15-97 Deleted Per 1995 Up)date 12/31/00 15-98 Deleted Per 1995 Up)date 12/31/00 15-99 Deleted Per 1995 Up3date 12/31/00 15-1 00 Deleted Per 1995 Update 12/31/00 15-101 Deleted Per 1995 Update 12/31/00 15-102 Deleted Per 1995 Update Effective Date 12/31/14 ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 2, Page 35 Effective Figure Figure Title Date No.12/31/00 15-103 Deleted Per 1995 Update 12/31/00 15-104 Deleted Per 1995 Update 12/31/00 15-1 05 Deleted Per 1995 Update 12/31/00 15-1 06 Deleted Per 1995 Update 12/31/00 15-1 07 Deleted Per 1995 Update 12/31/00 15-1 08 Deleted Per 1995 Update 12/31/00 15-1 09 Deleted Per 1995 Update 12/31/00 15-110 "Deleted Per 2001 Update 12/31/00 15-111 Deleted Per 2003 Update 12/31/14 15-112 Deleted Per 2014 Update 12/31/03 15-113 Rod Withdrawal at Power Accident -Core Cooling Capability Analysis RCS Pressure 12/31/13 15-114 Rod Withdrawal at Power Accident -Core Cooling Capability Analysis DNBR 12/31/08 15-115 Cold Water Accident -Core Average Temperature 12/31/08 15-116 Cold Water Accident -Power 12/31/08 15-117 Cold Water Accident -Cold Leg Temperature 12/31/08 15-118 Cold Water Accident -RCS Pressure 12/31/03 15-119 Loss of Coolant Flow Accidents

-Two RCP Coastdown from Four RCP Initial Conditions Analysis -Power 12/31/03 15-120 Loss of Coolant Flow Accidents

-Two RCP Coastdown from Four RCP Initial Conditions Analysis -RCS Temperature 12/31/03 15-121 Loss of Coolant Flow Accidents

-Two RCP Coastdown from Four RCP Initial Conditions Analysis -Pressurizer Level 12/31/03 15-1 22 Loss of Coolant Flow Accidents

-Two RCP Coastdown from Four RCP Initial Conditions Analysis -RCS Pressure 12/31/13 15-1 23 Loss of Coolant Flow Accidents

-Two RCP Coastdown from Four RCP Initial Conditions Analysis -DNBR 12/31/03 15-1 24 Loss of Coolant Flow Accidents

-One RCP Coastdown from Three RCP Initial Conditions Analysis -RCS Flow 12/31/03 15-125 Loss of Coolant Flow Accidents

-One RCP Coastdown from Three RCP Initial Conditions Analysis -Power 12/31/03 15-126 Loss of Coolant Flow Accidents

-One RCP Coastdown from Three RCP Initial Conditions Analysis -RCS Temperature Effective Date 12/31/14 ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 2, Page 36 Effective Figure Figure Title Date No.12/31/03 15-127 Loss of Coolant Flow Accidents

-One RCP Coastdown from Three RCP Initial Conditions Analysis -Pressurizer Level 12/31/03 15-128 Loss of Coolant Flow Accidents

-One RCP Coastdown from Three RCP Initial Conditions Analysis -ROS Pressure 12/31/13 15-129 Loss of Coolant Flow Accidents

-One RCP Coastdown from Three RCP Initial Conditions Analysis -DNBR 12/31/03 15-1 30 Loss of Coolant Flow Accidents

-Locked Rotor From Four RCP Initial Conditions Analysis -RCS Flow 12/31/03 15-131 Loss of Coolant Flow Accidents

-Locked Rotor From Four RCP Initial Conditions Analysis -Power 12/31/03 15-1 32 Loss of Coolant Flow Accidents

-Locked Rotor From Four RCP Initial Conditions Analysis -RCS Temperature 12/31/03 15-1 33 Loss of Coolant Flow Accidents

-Locked Rotor From Four RCP Initial Conditions Analysis -Pressurizer Level 12/31/03 15-1 34 Loss of Coolant Flow Accidents

-Locked Rotor From Four RCP Initial Conditions Analysis -RCS Pressure 12/31/08 15-135 Loss of Coolant Flow Accidents

-Locked Rotor From Four RCP Initial Conditions Analysis -DNBR 12/31/08 15-1 36 Loss of Coolant Flow Accidents

-Locked Rotor From Three RCP Initial Conditions Analysis -ROS Flow 12/31/08 15-1 37 Loss of Coolant Flow Accidents

-Locked Rotor From Three RCP Initial Conditions Analysis -Power 12/31/08 15-1 38 Loss of Coolant Flow Accidents

-Locked Rotor From Three RCP Initial Conditions Analysis -ROS Temperatures 12/31/08 15-1 39 Loss of Coolant Flow Accidents

-Locked Rotor From Three RCP Initial Conditions Analysis -Pressurizer Level 12/31/08 15-140 Loss of Coolant Flow Accidents

-Locked Rotor From Three RCP Initial Conditions Analysis -RCS Pressure 12/31/11 15-141 Loss of Coolant Flow Accidents

-Locked Rotor From Three RCP Initial Conditions Analysis -DNBR 12/31/00 15-142 Intentionally Blank 12/31/10 15-143 Control Rod Misalignment Accidents

-Dropped Rod -RCS Pressure 12/31/13 15-144 Control Rod Misalignment Accidents

-Dropped Rod -DNBR 12/31/13 15-145 Turbine Trip Accident -Steam Generator Pressure 12/31/13 15-146 Turbine Trip Accident -RCS Temperatures Effective Date 12/31/14 ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 2, Page 37 Effective Figure Figure Title Date No..1 2/31/13 15-147 Turbine Trip Accident-Pressurizer Level 12/31/13 15-148 Turbine Trip Accident -RCS Pressure 12/31/13 15-149 Turbine Trip Accident -Power 12/31/11 15-150 Steam Generator Tube Rupture -Power 12/31/11 15-151 Steam Generator Tube Rupture -Break Flow 12/31/11 15-152 Steam Generator Tube Rupture -ROS Pressure 12/31/11 15-153 Steam Generator Tube Rupture -Pressurizer Level ,12/31/11 15-154 Steam Generator Tube Rupture -Steam Generator Pressure 12/31/11 15-1 55 Steam Generator Tube Rupture -Steam Generator Level 12/31/11 15-156 Steam Generator Tube Rupture -RCS Temperatures 12/31/03 15-1 57 Steam Line Break Accident -With Offsite Power -Power 12/31/03 15-158 Steam Line Break Accident -With Offsite Power -ROS Pressure 12/31/03 15-1 59 Steam Line Break Accident -With Offsite Power -Core Inlet Flow 12/31/03 15-1 60 Deleted Per 2003 Update 12/31/03 15-161 Steam Line Break Accident -Without Offsite Power -Steam Line Pressure 12/31/03 15-1 62 Steam Line Break Accident -Without Offsite Power -RCS Temperatures

.12/31/03 15-163 Steam Line Break Accident -Without Offsite Power -RCS Flow 12/31/03 15-164 Steam Line Break Accident -Without Offsite Power -Reactivity 12/31/03 15-165 Steam Line Break Accident -Without Offsite Power -Power 12/31/03 15-1 66 Steam Line Break Accident -Without Offsite Power -RCS Pressure 12/31/13 15-1 67 Steam Line Break Accident -Without Offsite Power -DNBR 12/31/03 15-1 68 Small Steam Line Break -Steam Mass Flows 12/31/03 15-169 Small Steam Line Break -Steam Line Pressures 12/31/11 15-1 70 Small Steam Line Break -Main Feedwater Mass Flows 12/31/03 15-171 Small Steam Line Break -RCS Temperatures 12/31/03 15-1 72 Small Steam Line Break -Core Average Power 12/31/03 15-1 73 Small Steam Line Break -RCS Hot Leg Pressure 12/31/14 15-1 74 Deleted Per 2014 Update 12/31/03 15-1 75 Oconee -No CHRS Flow Effective Date 12/31/14 ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 2, Page 38 Effective Figure Figure Title Date No.12/31/01 15-176 Deleted Per 2001 Update 12/31/03 15-1 77 Lower Bound Containment Pressure Used in Large Break LOCA 12/31/14 15-178 Deleted Per 2014 Update 12/31/14 15-179 Deleted Per 2014 Update 12/31/14 15-1 80 Deleted Per 2014 Update 12/31/14 15-1 81 Deleted Per 2014 Update 12/31/14 15-182 Deleted Per 2014 Update 12/31/14 15-183 Deleted Per 2014 Update 12/31/14 15-184 Deleted Per 2014 Update 12/31/14 15-185 Deleted Per 2014 Update 12/31/14 15-186 Deleted Per 2014 Update 12/31/14 15-187 Deleted Per 2014 Update 12/31/14 15-188 Deleted Per 2014 Update 12/31/14 15-189 Deleted Per 2014 Update 12/31/14 15-1 90 Deleted Per 2014 Update 12/31/14 15-191 Deleted Per 2014 Update 12/31/14 15-1 92 Deleted Per 2014 Update 12/31/14 15-193 Deleted Per 2014 Update 12/31/14 15-1 94 Deleted Per 2014 Update 12/31/14 15-195 Deleted Per 2014 Update 12/31/14 15-1 96 Deleted Per 2014 Update 12/31/14 15-1 97 Deleted Per 2014 Update 12/31/14 15-198 Deleted Per 2014 Update 12/31/14 15-1 99 Deleted Per 2014 Update 12/31/14 15-200 Deleted Per 2014 Update 12/31/14 15-201 Deleted Per 2014 Update 12/31/14 15-202 Deleted Per 2014 Update 12/31/14 15-203 Deleted Per 2014 Update 12/31/14 15-204 Deleted Per 2014 Update 12/31/14 15-205 Deleted Per 2014 Update Effective Date 12/31/14 ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 2, Page 39 Effective Figure Figure Title Date No.12/31/14 15-206 Deleted Per 2014 Up3date 12/31/14 15-207 Deleted Per 2014 Up3date 12/31/14 15-208 Deleted Per 2014 Up3date 12/31/14 15-209 Deleted Per 2014 Up3date 12/31/14 15-210 Deleted Per 2014 Up3date 12/31/14 15-211 Deleted Per 2014 Up3date 12/31/14 15-212 Deleted Per 2014 Upadate 12/31/08 15-213 77% of 2568 MWt, Mark-B-HTP Mixed-Core SBLOCA Spectrum Analysis 12/31/08 15-214 0.075 ft 2 CLPD, 77% of 2568 MWt, Mark-B-HTP Mixed-Core SBLOCA Case-Pressure 12/31/08 15-215 0.075 ft2 CLPD, 77% of 2568 MWt, Mark-B-HTP Mixed-Core SBLOCA Case- Break and ECCS Mass Flow Rates 12/31/08 15-216 0.075 ft 2 CLPD, 77% of 2568 MWt, Mark-B-HTP Mixed-Core SBLOCA Case- Hot Channel Levels 12/31/08 15-217 0.075 ft2 CLPD, 77% of 2568 MWt, Mark-B-HTP Mixed-Core SBLOCA Case- Peak Cladding Temperature 12/31/08 15-218 0.075 ft 2 CLPD, 77% of 2568 MWt, Mark-B-HTP Mixed-Core SBLOCA Case- HC Vapor Temperature at Core Exit 12/31/11 15-219 Mark-B-HTP Full-Core BOL LBLOCA -Reactor Vessel Upper Plenum Pressure 12/31/11 15-220 Mark-B-HTP Full-Core BOL LBLOCA -Break Mass Flow Rates 12/31/11 15-221 Mark-B-HTP Full-Core BOL LBLOCA -Hot Channel Mass Flow Rates 12/31/11 15-222 Mark-B-HTP Full-Core BOL LBLOCA -Core Flooding Rates 12/31/11 15-223 Mark-B-HTP Full-Core BOL LBLOCA- Hot Pin Fuel & Clad Temperatures at Ruptured Location 12/31/1 1 15-224 Mark-B-HTP Full-Core BOL LBLOCA -Hot Pin Fuel & Clad Temperatures at Unruptured Location 12/31/11 15-225 Mark-B-HTP Full-Core BOL LBLOCA -Quench Front Advancement 12/31/11 15-226 Mark-B-HTP Full-Core BOL LBLOCA -Hot Pin Heat Transfer Coefficients 12/31/11 15-227 102% of 2568 MWt, Full Core Mark-B-HTP SBLOCA Break Spectrum Analysis Effective Date 12/31/14 ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 2, Page 40 Effective Figure Figure Title Date No.12/31/11 15-228 0.15ft 2 CLPD, 102% of 2568 MWt, Full Core Mark-B-HTP SBLOCA -Pressure 12/31/11 15-229 0.15ft 2 CLPD, 102% of 2568 MWt, Full Core Mark-B-HTP SBLOCA -Break and ECCS Mass Flow Rates 12/31/11 15-230 0.15ft2 CLPD, 102% of 2568 MWt, Full Core Mark-B-HTP SBLOCA -RV Collapsed Liquid Level & Hot Channel Mixture Level 12/31/11 15-231 0.15ft 2 CLPD, 102% of 2568 MWt, Full Core Mark-B-HTP SBLOCA -Hot Pin Peak Clad Temperature 12/31/11 15-232 0.15ft 2 CLPD, 102% of 2568 MWt, Full Core Mark-B-HTP SBLOCA-Hot Channel Vapor Temperature at Core Exit Effective Date 12/31/14 ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 3 Update Insertion Instructions (for hardcopy distribution only)(3 pages)

ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 3, Page 1 Update Insertion Instructions (for hardcopy distribution only)1. Replace List of Effective Pages (LOEP) for Tables and Figures with the 2014 LOEP Update.2. Replace entire text portions for each chapter with the updated text portion (including the Table of Contents, List of Figures, and List of Tables).3. Update Tables and Figures according to the instructions below.NOTE: Tables and Figures from prior year were re-issued in order to remove revision bars.Chapter 3 Chapter 4 Chapter 5 Chapter 6 Remove Table 3-2 (5 pages)Table 3-5 Table 3-68 (8 pages)Table 4-22 Table 4-24 Table 5-5 (2 pages)Table 6-7 (7 pages)Table 6-22 Table 6-23 Table 6-24 Table 6-25 Table 6-26 (2 pages)Table 6-31(2 pages)Figure 6-36 Figure 6-37 Table 7-4 Figure 7-1 (16 pages)Figure 7-5 (8 pages)Figure 7-8 Figure 7-10 Table 8-3 (1 page)Table 8-4 Insert Table 3-2 (6 pages)Table 3-5 Table 3-68 (8 pages)Table 4-22 Table 4-24 Table 5-5 (2 pages)Table 6-7 (7 pages)Table 6-22 Table 6-23 Table 6-24 Table 6-25 Table 6-26 (2 pages)Table 6-31 (2 pages)Figure 6-36 Figure 6-37 Table 7-4 Figure 7-1 (16 pages)Figure 7-5 (8 pages)Figure 7-8 Figure 7-10 Table 8-3 (2 pages)Table 8-4 Chapter 7 Chapter 8 Figure 8-1 (l1ix 17)Figure 8-3 (11lx17) (pg 1 of 2)Figure 8-4 (11 x 17) (3 pages)Figure 8-5 (11 x 17)Figure 8-1 (11 x 17)Figure 8-3 (11 x 17) (pg 1 of 2)Figure 8-4 (11 x 17) (3 pages)Figure 8-5 (11 x 17)

ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 3, Page 2 Remove Insert Chapter 9 Chapter 10 Chapter 13 Chapter 15 Figure 9-12 Figure 9-17 Figure 9-40 new new new Figure 10-8 Figure 13-1 Figure 13-3 Figure 13-4 Figure 13-8 Table 15-2 Table 15-32 (2 pages)Table 15-33 Table 15-34 (8 pages)Table 15-36 Table 15-46 Table 15-56 Table 15-57 Table 15-60 Table 15-62 (2 pages)Table 15-63 (2 pages)new new Figure 9-12 Figure 9-17 Figure 9-40 Figure 9-44 Figure 9-45 Figure 9-46 Figure 10-8 Figure 13-1 Figure 13-3 Figure .13-,4 Figure 13-8 Table 15-2 Table 15-32 (2 pages)Table 15-33 Table 15-34 (8 pages)Table 15-36 Table 15-46 Table 15-56 (Deleted)Table 15-57 (Deleted)Table 15-60 (Deleted)Table 15-62 (2 pages)Table 15-63 (2 pages)Table 15-67 Table 15-68 (2 pages)Figure 15-1 Figure 15-2 Figure 15-3 Figure 15-4 Figure 15-5 Figure 15-6 Figure 15-24 Figure 15-29 Figure 15-30 Figure 15-31 Figure 15-32 Figure 15-33 Figure 15-34 Figure 15-35 Figure 15-36 Figure 15-1 Figure 15-2 Figure 15-3 Figure 15-4 Figure 15-5 Figure 15-6 Figure 15-24 Figure 15-29 Figure 15-30 Figure 15-31 Figure 15-32 Figure 15-33 Figure 15-34 Figure 15-35 Figure 15-36 ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 3, Page 3 Chapter 15 (continued)

Remove Figure 15-112 Figure 15-114 Figure 15-1 23 Figure 15-1 29 Figure 15-144 Figure 15-145 Figure 15-1 46 Figure 15-147 Figure 15-148 Figure 15-1 49 Figure 15-1 67 Figure 15-1 74 Figure 15-178 thru 15-196 (19 pages)Figure 15-1 97 thru 15-212 (16 pages)Insert Figure 15-112 (Deleted)Figure 15-114 Figure 15-1 23 Figure 15-1 29 Figure 15-144 Figure 15-145 Figure 15-146 Figure 15-1 47 Figure 15-148 Figure 15-149 Figure 15-1 67 Figure 15-1 74 (Deleted)Figure 15-1 78 thru 15-1 96 (Deleted)

(1 page)Figure 15-197 thru 15-212 (Deleted)

(1 page)Chapter 18 Chpe 8 Table 18-1 (4 pages) Table 18-1 (4 pages)

ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 4 List of Removed Items (2 pages)

ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 4, Page 1 List of Removed Items 1. Section 7.4.3.1.2, System Design -Turbine Driven EFW Pump (TDEFWP)The fifth paragraph was a single sentence which stated, "Once automatically started, the TDEFWP will continue to operate until manually secured by the operator or shutdown by the MSLB circuitry." Since this sentence would be corrected similarly to the sentence revised in the fourth paragraph (described below), the fifth paragraph was removed.Section 7.4.3.1.2, fourth paragraph under subheading "Turbine Driven EFW Pump (TDEFWP)", was revised as follows since AFIS had been installed on all three units: "Once automatically started, the TDEFWP will continue to operate until manually secured by the operator or disabled by an AFIS (Unit 1) or SeLB (Uit 2I 3 " n, 3) signal." This change is based on the NRC-approved AFIS installation for all three units.2. Section 9.2.3, Auxiliary Service Water System, was deleted and now refers users to UFSAR Section 9.7, Protected Service Water System. The Auxiliary Service Water System was replaced by the Protected Service Water System as detailed in the Safety Evaluation Report dated August 13, 2014.3. Following the transition of all Oconee units to full-core Mark-B-HTP loading patterns, the existing UFSAR Chapter 15 description and results of the LBLOCA analyses and full-power SBLOCA analyses for Mk-B1 1 fuel and Mark-B-HTP fuel in a mixed-core configuration were removed as obsolete information.

a. Section 15.14.4.1.2, Limiting Linear Heat Rate Analysis (LOCA Limit) for Mk-B1 1 (M5) Fuel, was deleted since it describes LBLOCA analyses applicable to Mark-B1 1 fuel. Since all Oconee units have completed the transition to full-core Mark-B-HTP loading patterns, the existing UFSAR Chapter 15 description and results of the LBLOCA analyses for Mark-Bl1 fuel are no longer applicable.

Corresponding to the deletion of Section 15.14.4.1.2, the following UFSAR Figures and Table(s) were also deleted: Figures15-112, 15-1 74, 15-1 78,15-179, 15-180,15-181, 15-182,15-183, 15-184, and 15-185; Table 15-56.b. Section 15.14.4.1.3, Mixed Core Mark-B-HTP LOCA LHR Limits, was deleted since it describes LBLOCA analyses applicable to Mark-B1 1 and Mark-B-HTP fuel in a mixed-core configuration.

Since all Oconee units have completed the transition to full-core Mark-B-HTP loading patterns, the existing UFSAR Chapter 15 description and results of the LBLOCA analyses for mixed-core configurations of Mark-B1 1 and Mark-B-HTP fuel are no longer applicable.

Corresponding to the deletion of Section 15.14.4.1.3, the following UFSAR Figures and Table(s) were also deleted: Figures15-198, 15-199,15-200, 15-201,15-202, 15-203,15-204, 15-205, and 15-206; Table 15-60.c. Section 15.14.4.1.4, Mixed Core Mark-B1l LOCA LHR Limits, was deleted since it describes LBLOCA analyses applicable to Mark-B1 1 and Mark-B-HTP fuel in a mixed-core configuration.

Since all Oconee units have completed the transition to full-core Mark-B-HTP loading patterns, the existing UFSAR Chapter 15 description and results of the LBLOCA analyses for mixed-core configurations of Mark-B1 1 and Mark-B-HTP fuel are no longer applicable.

ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 4, Page 2 d. Section 15.14.4.2.1, Mark-B-11 SBLOCA and Break Spectrum Results, was deleted since it describes SBLOCA analyses applicable to Mark-B11I fuel. Since all Oconee units have completed the transition to full-core Mark-B-HTP loading patterns, the existing UFSAR Chapter 15 description and results of the SBLOCA analyses for full cores of Mark-B1 1 fuel are no longer applicable.

Corresponding to the deletion of Section 15.14.4.2.1, the following UFSAR Figures and Table should also be deleted: Figures15-186, 15-1 87,15-188, 15-189,15-190, 15-191,15-192, 15-193,15-194, 15-195,15-196, and 15-197;Table 15-57.e. Section 15.14.4.2.2, Mixed Core Mark-B-HTP SBLOCA and Break Spectrum Results, was deleted since it describes SBLOCA analyses applicable to Mark-B1 1 and Mark-B- HTP fuel in a mixed-core configuration.

Since all Oconee units have completed the transition to full core Mark-B-HTP loading patterns, a separate UFSAR section for description and results of the SBLOCA analyses for mixed-core configurations of Mark-Bi11 and Mark-B-HTP fuel is no longer applicable.

Corresponding to the deletion of Section 15.14.4.2.2, the following UFSAR Figures should also be deleted: Figures15-207, 15-208,15-209, 15-210,15-211, and 15-212.4. Sections 18.3.19.1, Master Integrated Reactor Vessel Surveillance Program, 18.3.19.2, Cavity Dosimetry Program, 18.3.19.3, Fluence and Uncertainty Calculations, and 18.3.19.4, Pressure Temperature Limit Curves, were combined into a single section, 18.3.19, Reactor Vessel Integrity Program. The change updates the nomenclature and better defines the different components of the Program. The time limited aging analysis (TLAA) and surveillance and monitoring activities are defined with clear separation.

Finally, the need to periodically exchange cavity dosimetry, update fluence transport calculations and predict end of life fluence is emphasized.

Section 18.3.19.5, Monitoring Effective Full Power Years, was deleted. The implication that effective full power years (EFPY) are used to determine when updates are needed to TLAAs instead of fluence values is being corrected by deleting this section. A review of the Safety Evaluation Report (NU3REG-1 723) determined that EFPY was not part of the NRC's understanding of the need for TLAA fluence updates, so the correction is supported by the wording in the SER. The changes do not affect License Renewal commitments or change any of the regulatory requirements.

ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 5 10 CER 50.59 Report (18 pages)

ONS-201 5-072 Document Controi Desk June 30, 2015 Attachment 5, Page 1 Title: EC 100744 (Non-LAR Scope Only) -Making the Reverse Osmosis System Operable (Reverse Osmosis to Clean Silica from the SFPs and BWSTs)(10 CFR 50.59 Revision 0) (AR 00437862 / 01337862)(1)

Summary: Engineering Change (EC) 100744 addresses making a Reverse Osmosis (RO) System operational that is to be used for treating the contents of the Units 1 & 2 Spent Fuel Pool (SFP), the Unit 3 SEP, the Unit I Borated Water Storage Tank (BWST), the Unit 2 BWST, and the Unit 3 BWST by removing silica. This EC is only addressing making the flowpath functional for processing the SFPs and the BWSTs. The RO Unit, piping, and electrical supplies to the RO System components were or will be installed by other ECs. The scope of two other ECs that install air operated valves in the BWST flowpath are also credited for RO System operation on the BWST. The RO Unit has been installed by another EC, but it was installed as an inert object, with no justification for its being operational.

EC 100744 will allow the RO skid to be powered, and allow the opening of isolation valves associated with the flowpath for cleaning silica from the SFPs and BWSTs. These collective ECs have not made the RO Unit operational because it was determined that NRC approval was required for the use of the RO System. NRC approval has now been received.The RO System will utilize a Boric Acid Recovery System (BARS) that is supplied by Diversified Technology Services (DTS). The BARS unit is designed to remove silica, while recovering boric acid to the maximum extent possible.

This unit is used to remove silica by the reverse osmosis process, and so is typically called the Reverse Osmosis Unit (RO Unit). The reverse osmosis process uses high pressure applied to a solution on one side of a semi-permeable membrane.Some minerals, salts, and colloidal solids are unable to pass through the membrane and are rejected into a waste stream, while the remainder of the solution passes through the membrane and is collected for return to the system. The RO Unit that is to be used for the cleanup of the BWST and SFP has been designed to recover a high percentage of boron.The elevated concentrations of silica in the SFPs and BWSTs are due to the decomposition of Boraflex, which was used in the spent fuel racks in the SFPs. This decomposition has caused the Reactor Coolant System silica concentration to be above that recommended by the fuel vendor. Reduction of the silica content would also allow the use of "zinc addition" in the Reactor Coolant System (RCS) for dose rate reduction.

Zinc addition would be accomplished by another EC.In the early stages of the Engineering Change process for EC 100744, Duke Energy determined that operating the RO System during Unit operation could not be implemented under the 10 CFR 50.59 process and would require Nuclear Regulatory Commission (NRC) approval before implementing the activity.

As such, Duke Energy submitted a License Amendment Request (LAR) to the NRC to review and approve the design features and controls that would be used to ensure that Operation of a Reverse Osmosis (RO) system during Unit operation does not significantly impact the BWST or SFP function or other plant equipment.

The LAR also included new technical specification changes associated with the RO System. The LAR provided technical justification for periodic limited operation of the RO System during Unit operation.

Duke Energy evaluated the effect of potential failures and identified precautionary measures that must be taken before and during RO System operation, and required operator actions to protect affected structures, systems, and components.

The NRC has reviewed and approved the LAR.(1) The AR (Activity Record) number is used to track the 10 CFR 50.59 report in Duke Energy's database.

ONS-201 5-072 Document Controi Desk June 30, 2015 Attachment 5, Page 2 There were aspects of the EC that were not included in the LAR or were changed from the information included in the LAR submittals.

The aspect not included was a new automatic shutdown circuit that was added to address ambient temperature effects in Room 349. The new circuit was added by a separate EC, but the effects of the RO Unit increasing the room temperature and the use of the protection circuit as part of the operation of the RO Unit are addressed in the 10 CFR 50.59 evaluation for EC 100744. The aspect that was changed from the information contained in the LAR related to the tables submitted that contain time, boron, and water levels.A summary of the justification for the responses to the evaluation questions is now provided.The RO System itself is not required to operate to prevent any design basis events or accidents.

The ambient temperature shutdown circuit could inadvertently shutdown the RO Unit, but since the RO System is not used to prevent eventsfaccidents, it does not cause accidents previously evaluated in the UFSAR. So the ambient temperature shutdown circuit does not cause any accidents or events if it causes inadvertent loss of power to the RO Unit. The RO Unit will add heat to Room 349 while the unit is running and a smaller amount after it is shutdown and cools to ambient. The ambient temperature shutdown circuit is being added by another EC that will shutdown the RO Unit before Room 349 reaches temperatures that have been previously found to be acceptable for protection of equipment in the room that has a 10 CFR 50.59 design function.

Thus, the activity does not result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the UFSAR.The new tables that provide requirements for operation of the RO System are supported by calculations that meet the criteria described in the LAR to the NRC. These criteria include that the maximum time period that the RO System can operate without makeup is limited by water inventory rather than boron concentration.

In addition, the changes in water level can be easily recognized by plant personnel and low level is alarmed prior to decreasing below the TS low water level limit. Also, that in all cases the boron concentration stays above the TS minimum boron concentration prior to reaching the low water level alarm level or at the end of the 7 days, which will be specified as the maximum RO operating period. The LAR noted that the specific values used in the table that was submitted may be changed if the Oconee calculation was revised to support the change. These criteria were met by the new tables.The ambient temperature shutdown circuit is designed to ensure that the heat from the RO Unit does not cause the room temperature at any time to go above the previously established room temperature limits. The design of the ambient temperature shutdown circuit is QA-1 and is seismically qualified, which includes all subcomponents that make up the ambient temperature circuits.

Non-QA power to the monitoring circuit is used. This is acceptable since the same power source is used for both the RO Unit and the ambient temperature shutdown circuit. The power supply to the ambient temperature circuit comes off the supply to the RO Unit and the circuitry is otherwise redundant so that a single failure could not cause the circuit to be inoperable, since the circuit is redundant/single failure proof. If power is lost to the ambient temperature shutdown circuit, there is no credible way for power to the RO Unit to continue uninterrupted.

An additional feature of the circuit is that power is required to keep the main contacts of the contactor supplying power to the RO Unit closed. If power is lost, stored mechanical energy (spring) is used to open the circuit contacts/breaker.

This offers additional assurance that the RO Unit cannot continue to receive power for this power interruption event.This last feature is not specifically required because it must trip on high temperature regardless of this feature. Thus, no credible single failure will prevent the circuit from shutting off power to ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 5, Page 3 the RO Unit when the setpoint is reached. The circuitry and equipment of the ambient temperature shutdown equipment is designed to withstand the environmental conditions that would be present up to the time it is called upon to act. The components used for the ambient temperature circuit are not digital and are not tied to components that are digital related. Thus, the activity will not result in a more than minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the UFSAR.The RO System itself is not required to operate to mitigate any design basis events or accidents.

The ambient temperature shutdown circuit could inadvertently shutdown the RO Unit, but the RO System is not used to mitigate events/accidents and will not increase any consequences of any accidents evaluated in the UFSAR. The new tables are supported by calculations that meet the criteria described in the LAR to the NRC. The ambient temperature shutdown circuit is designed to ensure that the heat from the RO Unit does not cause the room temperature at any time to go above the previously established room temperature limits. The changes to the tables and the addition of the ambient temperature shutdown circuit do not prevent any systems or components from performing their design functions as described in the UFSAR or the NRC's safety evaluation for the operation of the RO System during any accidents previously evaluated in the UFSAR. Thus, the activity does not result in more than a minimal increase in the consequences of an accident previously evaluated in the UFSAR.The RO System itself is not required to operate to mitigate any design basis events or accidents, so any malfunctions of the RO System operation itself does not affect dose. The ambient temperature shutdown circuit could inadvertently shutdown the RO Unit, but since the RO System is not used to mitigate events/accidents, it does not cause accidents previously evaluated in the UFSAR. The new tables are supported by calculations that meet the criteria described in the LAR to the NRC. The ambient temperature shutdown circuit is designed to ensure that the heat from the RO Unit does not cause the room temperature at any time to go above the previously established room temperature limits. The changes to the tables and the addition of the ambient temperature shutdown circuit do not prevent any systems or components from performing their design functions as described in the UFSAR or the NRC's safety evaluation for the operation of the RO System. The ambient temperature shutdown circuit is designed such that a single failure will not prevent it from shutting off the RO Unit if the room temperature reaches its setpoint.

Thus, the activity does not result in more than a minimal increase in the consequences of a malfunction of an SSC important to safety previously evaluated in the UFSAR.The new tables are supported by calculations that meet the criteria described in the LAR to the NRC. The ambient temperature shutdown circuit is designed to ensure that the heat from the RO Unit does not cause the room temperature at any time to go above the previously established room temperature limits. The changes to the tables and the addition of the ambient temperature shutdown circuit do not prevent any systems or components from performing their design functions as described in the UFSAR or the NRC's safety evaluation for the operation of the RO System. There are no new accidents that the non-LAR portion of the EC can create.Thus, the activity does not create a possibility for an accident of a different type than previously evaluated in the UFSAR.The new tables are supported by calculations that meet the criteria described in the LAR to the NRC. The ambient temperature shutdown circuit is designed to ensure that the heat from the RO Unit does not cause the room temperature at any time to go above the previously established room temperature limits. The changes to the tables and the addition of the ambient ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 5, Page 4 temperature shutdown circuit do not prevent any systems or components from performing their design functions as described in the UFSAR or the NRC's safety evaluation for the operation of the RO System. The components used for the ambient temperature circuit are not digital and are not tied to components that are digital related. Thus, this activity does not create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in the UFSAR.The fission product barriers are the fuel pellet, cladding, reactor coolant pressure boundary, and containment.

This activity does not modify a fission product barrier, nor does this activity affect a controlling numerical value for a parameter established during the licensing review as presented in the UFSAR for a parameter used to determine the integrity of a fission product barrier. Thus, the modification does not result in a design basis limit for a fission product barrier as described in the UFSAR being exceeded or altered.Specific calculation methods of evaluation regarding the calculation used to determine the values used in the table that contains the boron and level information are not described in the LAR, NRC's safety evaluation, or the UFSAR. This activity does not revise a computer code or calculation that is described in the UFSAR. It also does not change a design or analysis code described in the UFSAR. Thus, the activity does not involve a change in a method of evaluation described in the UFSAR. Thus, the modification does not result in a departure from a method of evaluation in the UFSAR.The non-LAR scope did not require any changes, additions, or deletions to Technical Specifications (TS).The responses to the 10 CFR 50.59 questions were all no. There were also no changes, additions, or deletions needed for the TSs. Thus, the activity can be implemented without prior NRC approval.Title: Revise UFSAR Section 8.3.1.2 to Address 126 Day Keowee Reservoir Water Storage Supply for a Keowee Hydro Unit -Reference PIP 0-08-3902 (10 CFR 50.59 Revision 0) (AR 00438727 I 01338727)Summary: UFSAR Section 8.3.2.1 contains a paragraph as follows: The independent Keowee units, along with the alternate circuits, provide the required redundancy to assure reliable emergency power. Storage capacity of the Keowee reservoir and naturally occurring minimum streamflow are such that the generating units can provide continuous emergency power following an accident.

The Keowee reservoir, between its normal elevation and maximum planned drawdown, has sufficient stor'age which, when combined with minimum recorded streamflow on the Keowee River will permit a hydro unit to carry continuously one nuclear unit's emergency auxiliary loads for 126 days.UFSAR Section 8.3.1.2 is to be revised to remove the statement that provides information that the Keowee reservoir would permit a hydro unit to carry continuously one nuclear unit's emergency auxiliary loads for 126 days. The preceding statement to the one that is being removed is being reworded to remove the word "continuous" to avoid confusion in that continuous can imply going on forever and to remove the word "minimum" since the statement ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 5, Page 5 is now more general. The removed sentence in the paragraph indicated that the Keowee Hydro Units (KHU) were not expected to operate forever.The new paragraph is revised as follows: The independent Keowee units, along with the alternate circuits, provide the required redundancy to assure reliable emergency power. The Keowee reservoir and its water supplies (e.g., incoming streamflow) provide the water for the Keowee units so they can provide emergency power following an accident.The proposed UFSAR change is to remove a statement about the specified number of days that the Keowee reservoir is available for one KHU to carry continuously one nuclear unit's emergency auxiliary loads. Based on research, it was concluded that this information was provided by Duke originally to the Atomic Energy Commission (AEC) as part of a response to a Preliminary Safety Analysis Report (PSAR) question regarding the reliability of the emergency power supplies for Oconee. The Oconee design basis accident is for three units to have a Loss of Offsite Power, not just one. The second sentence in the paragraph is being reworded to remove the word "continuous" to avoid confusion in that continuous can imply going on forever and to remove the word "minimum" since the statement is now more general. The third sentence in the paragraph indicated that the KHUs were not expected to operate forever. The second sentence is also being reworded to clarify the concept that the Keowee reservoir and its water supplies are what provide the water for the KHUs. The third sentence is being deleted since the information is not part of the Keowee design function as discussed below.The KHUs have a design function to provide emergency power to the three Oconee units. The following information addresses whether the 126 days of reservoir supply is part of the design function for Keowee supplying this emergency power function.There was a question by the Atomic Energy Commission during the PSAR time period in which the AEC desired information regarding the reliability of the emergency power generation sources. Duke responded to this request with a number of aspects that was to show how reliable these sources were, including the KHUs. It was in this discussion that Duke addressed the Keowee reservoir could provide a storage capacity for a hydro unit carrying one nuclear unit's emergency auxiliary Loads for 126 days. The early version of the FSAR also contained similar information to support statements made about the reliability of the KHUs as emergency power sources. The current version of the UFSAR uses this same type information in justification for the reliability of the hydro units as an emergency power source. The AEC's safety evaluations for the original FSARs did not address the lake reservoir from the perspective of supplying the KHUs for emergency power. The AEC review of the emergency power systems was oriented toward the design of the Keowee units themselves and the power distribution system, as well as the use of the Lee Steam Station gas turbines as backup power and the other two units.The last sentence in the affected UFSAR paragraph is being removed since it is considered to be non-licensing basis information that was initially provided as general information to the AEC with respect to the response to a PSAR question.

The statement in the UFSAR relating to the 126 day capacity of the reservoir does not correspond to the 3 unit site's design basis accident, which is a 2 unit LOOP and a 1 unit LOCA/LOOP.

The UFSAR statement is the reservoir's water supply for only one nuclear unit's emergency auxiliary loads.

ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 5, Page 6 The UFSAR statement being changed also does not relate to a required volume of water in Lake Keowee. The statement refers to a volume of water based on normal elevation and maximum planned drawdown, but the normal lake level elevation is not a requirement.

The maximum planned drawdown can be 775 feet based on a Technical Specification

3.8.1 level

limit. No normal elevation foot level was specified in the statement.

No requirement could be found in early licensing correspondence to maintain any level above 775 feet. At the time that the statement was initially in the UFSAR, there was no lake level requirement except for the TS LCO level limit. There is now a lake level limit in the Selected Licensee Commitments for lake water supply, but this commitment was not part of Oconee's licensing basis at the time that the 126 day value was initially included in the UFSAR.Based on the subjectivity of the 126 day capacity statement in the UFSAR and the inconsistency with the Oconee design basis accident of 3 units having a LOOP, as well as finding no other pertinent information that utilizes the 126 day time period, the conclusion is made that this 126 day time period was not used in any accident analyses.

The original Unit 1 and Units 2 & 3 safety evaluations also do not address the Keowee reservoir's capacity from the perspective of the water used for the KHU's emergency power supplies.

The only requirement for lake level at the time of the initial UFSAR information about 126 days was the TS requirement for a lake level of 775 feet elevation, which was not to ensure a volume of lake water for the KHUs. This 126 day time period appears to have been just general information that was included in the original FSAR information.

Based on information in NEI 96-07 Revision 1 regarding what a design function consists of, the 126 days is not credited in safety analyses to meet NRC requirement and is not necessary to comply with regulations, license conditions, orders or technical specifications.

In addition, the reservoir water supply with respect to supplying the KHUs is not a function that, if not performed, would initiate a transient or accident that the plant is required to withstand since the KHUs are not initiators of transients or accidents, but are used for mitigation of accidents.

Based on the NEI 96-07 guidance and the discussion above, the KHUs' design function does not include the 126 day period included in the UFSAR.Thus, the activity does not result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the UFSAR.The KHUs would not be used in an emergency power function unless a loss of offsite power (LOOP) occurred, the changes to the UFSAR paragraph would not cause any accident previously evaluated in the UFSAR. Thus, the change does not result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the UFSAR.The KHU(s) themselves are not changed and will still be able to perform their design basis function of providing adequate emergency power to the Oconee units for a Loss of Coolant Accident (LOCA)/LOOP and other loss of offsite power events. The only change is to revise/remove the statements about a volume of water that may or may not be available in the Keowee reservoir to supply the KHUs. The required minimum lake level is controlled by Technical Specifications and Selected Licensee Commitments.

The minimum water level is related to the volume available.

If the water supply is not available, either with the existing or revised UFSAR information, the KHUs will not perform their function of providing emergency power supply to Oconee. But this effect will be the same if the water is not available.

Therefore there is not a more than minimal increase in the consequences of an accident previously evaluated in the UFSAR.

ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 5, Page 7 The UFSAR change does not affect the single failure design of the Oconee emergency power system. The change is removing non-design function information about a volume of water that may or may not be available to the KHUs. The required minimum lake level is controlled by TSs and SLCs. The minimum water level is related to the volume available.

The effect to Oconee would be the same if the KHUs were not available due to loss of the lake source. If no emergency power from a KHU is available, the consequences will be the same for any condition when emergency power is needed. Thus, the activity does not result in more than a minimal increase in the consequences of a malfunction of an SS0 important to safety previously evaluated in the UFSAR.The UFSAR change does not cause any accidents of a different type. The lake provides water for cooling functions at Oconee, but this water is returned to the lake. The function of the lake and the KHUs are not changed. The revision/removal of information regarding a potentially available (since no required volumes are specified) volume of reservoir water does not cause the KHUs or the lake to operate in a manner different than currently evaluated in the UFSAR.The station blackout (380) event already assumes the loss of the KHUs as an emergency power source. Thus, the activity does not create a possibility for an accident of a different type than previously evaluated in the UFSAR.Lake Keowee provides the source of water for the KHUs. If the reservoir water supply is lost, the effect of running out of water for the KHU emergency power supply will be the same. The KHUs emergency power supply will be lost. The change does not create any new potential for the loss of multiple emergency power sources. Thus, there are no new malfunctions that can be created with a different result. Thus, the activity does not create a possibility for a malfunction of an SS0 important to safety with a different result than any previously evaluated in the UFSAR.The fission product barriers are the fuel pellet, cladding, reactor coolant pressure boundary, and containment.

This activity does not modify a fission product barrier, nor does this activity affect a controlling numerical value for a parameter established during the licensing review as presented in the UFSAR for a parameter used to determine the integrity of a fission product barrier. Thus, the activity does not result in a design basis limit for a fission product barrier as described in the UFSAR being exceeded or altered.This activity does not revise a computer code or calculation that is described in the UFSAR. It also does not change a design or analysis code described in the UFSAR. No method of evaluation regarding the 126 days of availability was found to be described in the UFSAR. Thus, the activity does not involve a change in a method of evaluation described in the UFSAR. Thus, the UFSAR change does not result in a departure from a method of evaluation in the UFSAR.Title: Oconee Unit 3 Cycle 28 Reload 10 CFR 50.59 Evaluation (AR 00443458 1 01343458)Summary: This activity installs the core designed for Oconee Nuclear Station Unit 3 Cycle 28 (03028). The 03028 Reload Design Safety Analysis Review (REDSAR), performed in accordance with Engineering Directives Manual EDM-501, "Engineering Change Program for Nuclear Fuel", and the 03028 Reload Safety Evaluation confirms the UFSAR accident analyses remain bounding with respect to predicted O3C28 safety analysis physics parameters (SAPP), and fuel thermal ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 5, Page 8 and mechanical performance limits. The SAPP method is described in topical report DPC-NE-3005-PA, "U FSAR Chapter 15 Transient Analysis Methodology".

The O3C28 core reload is similar to past cycle core designs, with a design generated using NRC approved methods. Applicable Technical Specifications have been reviewed and no changes are required for the operation of O3C28. This 10 CFR 50.59 evaluation concluded that no prior NRC approval is necessary for O3C28 operation.

Title: EC 91875, Rev. 75 (OD500927)

-Keowee AC Power Supply Tie-ins to PSW, EC 91873, Rev. 16 (OD500922)

-PSW Power Feed Installation, 10 CFR 50.59 Evaluation Rev. 1 (AR 00444590 / 01344590)Sum mary: This 10CFR50.59 Evaluation addresses engineering change packages EC 91875, Rev. 75 and EC 91873, Rev. 16, which have been developed to provide an alternate QA-1 power supply to the PSW system. EC 91875 installs, at Keowee, the following:

• PSW switchgear cabinets (KPF-1 and KPF-2).* Incoming PSW switchgear breakers KPF-9 and KPF-10.* Outgoing PSW switchgear breakers KPF-11I and KPF-1 2.* New tap connections to the existing KHU segregated buses.* Power cable between the new tap connections on the KHU segregated buses and breakers KPF-9 and KPF-10.* Cable tray and cable tray supports.* Control switches and instrumentation in the Keowee control room.* New protective relaying and modifications to existing protective relaying.* Control power from the existing Keowee 1 25VDC Power System.EC 91873 installs the following:

• New and existing (spare) underground power cable from outgoing PSW switchgear breakers KPF-11 and KPF-12 to PSW switchgear B6T and B7T in the PSW Building.The cable will be routed in cable trenches and duct banks.*Three new PSW MCCs, 1XPSW, 2XPSW, and 3XPSW, in the Unit 1/2/3 Auxiliary Building.

The MCCs will be powered from a load center in the PSW Building via underground power cables. (NOTE: EC 91873, Rev. 16 only includes a partial turnover of new PSW MCCs 1XPSW, 2XPSW, and 3XPSW, which permits the MCCs to be energized and to power prescribed loads.*Underground power cable from the PSW Building load center (PX13) to the Auxiliary Building MCCs 1XPSW, 2XPSW, and 3XPSW, via underground duct bank.*New instrument cable from new MCCs 1XPSW, 2XPSW, and 3XPSW to existing terminal cabinets.

ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 5, Page 9 AR 444477444477is a single 100FR50.59 Screening that was prepared for the above ECs. The screening concluded that certain aspects of these ECs resulted in adverse effects to design functions of the KHUs as described in the UFSAR, and in controlling the electrical distribution system in a manner that is outside the reference bounds of its design, as described in the UFSAR. Therefore, the scope of this 100FR5O.59 Evaluation will address these specific aspects of EC 91875 and EC 91873.10CFR50.59 Screening AR 444477444477determined that EC 91875 and EC 91873 resulted in the following adverse effects to design functions as described in the UFSAR (Screening Question 1): a) The existing differential protective relay for each KHU (30 87GB-I and 30 87GB-2) is being replaced with three (3) single phase differential relays for each KHU (87GB1-Z0, -Y0, -X0 and 87GB2-Z0, -Y0, -X0) for differential protection of each Keowee electric generator.

The existing differential protective relays provide an alarming function, while the replacement differential protective relays will provide an emergency lockout (trip) of the respective KHU in the event of a phase-to-phase fault in either the emergency electric power system or in the new PSW electric distribution system. Installation of the replacement relays is considered to have an adverse effect on the design function of the KHUs because a failure of any one of the new 87GB relays may initiate a trip of the respective KHU and prevent the KHU from performing its credited design function.b) A current transformer, a 5OB overcurrent relay and a 62-timer relay are being added upstream of PSW electrical distribution system feeder breakers KPF-9 and KPF-10 to provide breaker fault detection and protection for these circuits.

If breakers KPF-9 or KPF-10 are tripped and subsequently fail to open, the breaker failure scheme becomes active. If the 508 overcurrent relay detects continued fault current for the time setting of the 62-timer relay, the 62-timer relay will initiate an emergency lockout (trip) of the respective KHU. Installation of the 62-timer relay is considered to have an adverse effect on the design function of the KHUs because a failure of the relay may initiate a trip of the respective KHU and prevent the KHU from performing its credited design function.c) The existing KHU generator buses, which are segregated and metal-enclosed, are being modified to provide a means of routing new power cable from the KHU buses to new PSW feeder breakers KPF-9 and KPF-10. A new bolted connection will be made to each of the generator buses. Flexible electrical connections will be installed on each of the segregated phase buses and connected to newly installed transition junction boxes. Cable terminations will be made inside the transition junction boxes and the cables will be routed to breakers KPF-9 and KPF-10. These new components (bolted connections, flexible electrical connectors, transition junction boxes, cabling and termination points) create new potential fault locations for the KHU generator buses. A fault on any of the new components may result in a lockout (trip) of the associated KHU and prevent the KHU from performing its credited design function.10CFR50.59 Screening AR 444477444477also determined that EC 91875 and EC 91873 involves an SS0 that will be controlled in a manner that is outside the reference bounds of its design as described in the UFSAR (Screening Question 4):

ONS-201 5-072.Document Control Desk O June 30, 2015 Attachment 5, Page 10 The existing Keowee frequency and voltage out-of-tolerance (COT) protection logic ensures that KHU generator frequency and voltage are within analyzed limits before loads are placed onto the KHU buses. This protection is accomplished either by blocking the closure or by tripping existing emergency power system breakers.

Since this protection is performed automatically, no operator action is required to verify adequate frequency and voltage during either the initial automatic addition of emergency loads onto the KHU buses or during the subsequent manual addition of non-emergency loads (supplied from the main feeder buses)onto the KHU buses. However, new PSW breakers KPF-9 and KPF-1 0 have not been provided with automatic frequency and voltage OCT protection, and therefore will require that Operations personnel verify that the voltage and frequency of a running KHU is acceptable before placing PSW system loads onto the KHU buses. The PSW system. loads will be the only loads at Oconee that can be added onto the KHU buses without automatic frequency and voltage COT protection.

Normal operation and failures of the new 87GB relays and the new 62-timer relays, and faults'within the new components installed to route power from the existing KHUs to the new PSW feeder breakers.

KPF-9 and KPF-1 0, have been evaluated and shown to result in an increase in the probability of a Station Blackout accident that is less than 10%. Therefore, installation of the new 87GB relays, the new 62-timer relays, and the new components installed to route power from the existing KHUs to the new PSW feeder breakers KPF-9 and KPF-10 does not result in more than a minimal increase in the frequency of occurrence of an accident.S The manual addition of PSW system electrical loads to a KHU generator will not result in an overload condition where the combination of the current design basis electrical loads and the PSW system electrical loads exceed the rated capacity of the Keowee generators.

It will also not result in a KHU generator loading transient that causes a frequency out-of-tolerance (COT)trip or a voltage COT trip of emergency power path breakers.

Manual verification of proper KHU frequency and voltage prior adding PSW system loads will also not increase the probability of a KHU failure. Therefore, the manual addition of PSW system electrical loads onto a KHU generator will not initiate a Station Blackout and does not result in any increase in the frequency of occurrence of an accident.The new protective relaying components, the components that make up the new power path between the KHU generators and the new PSW switchgear, and the potential Keowee generator trips that they may introduce, have been evaluated and determined to be within the existing single failure design criteria of the KHU generator and the emergency electric power system. Also, these components will be furnished as QA-1 (as required), will be seismically qualified, and will be seismically installed.

Hence, EC 91875 and EC 91873 will not result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety, and will have no increase in the consequences of a either an accident or a malfunction of an SSC important to safety.The new protective relaying components and the components that make up the new power path between the KHU generators and the new PSW switchgear will not introduce any equipment failures or malfunctions that are new or that have a different result than what is considered credible for the existing protective relaying and power path components in these systems.Therefore, the new protective relaying and power path components will not create a new type of p accident or a malfunction with a different result.The manual verification of proper KHU frequency and voltage prior to the addition of PSW system electrical loads onto the Keowee generator buses, and the addition of the PSW system ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 5, Page 11 electrical loads themselves have been evaluated and shown not to prevent the KHU generators from performing their credited design functions.

Hence, the manual addition of PSW system electrical loads onto the Keowee generator buses will not increase either the probability or consequences of accidents or malfunctions, and will not create a new type of accident or a malfunction with a different result.Accident analyses assume that electrical power is provided to the essential equipment that is required to protect fuel cladding, the reactor coolant pressure boundary, and the containment.

For accidents which include the consideration of the loss of offsite power (LOCA and main steam line break), the KHUs are credited for providing the required electrical power to these essential systems within 23 seconds. The new 87GB relays, the new 62-timer relays, the new components installed to route power from the existing KHU generator buses to new breakers KPF 9 and KPF-10, and the manual loading of the PSW system electrical loads onto the KHU generators will have no impact on the ability of at least one of the two redundant KHUs to automatically start and align to the emergency buses within 23 seconds. Therefore, the new relays, new power path components, and new manual loading of the KHU generators do not result in a design basis limit for a fission product barrier as described in the UFSAR being exceeded or altered.A revision to Technical Specifications will not be required.Title: EC 91834, REV. 29, (00100950)

-U1 HPI ALTERNATE POWER FEED FROM PSW, 10 CFR 50.59 EVALUATION, REV. 0 (AR 00445172 / 01345172)Summary: This 10 CFR 50.59 Evaluation addresses engineering change package EC 91834, Rev. 29, which has been developed to provide an alternate power source from the PSW electrical distribution system to selected components in the Unit 1 High Pressure injection (HPI) system and the Unit 1 Reactor Coolant System (RCS) to support core reactivity control, reactor coolant inventory control, and reactor coolant pressure control functions following events (fire and HELB) that damage the essential 416OVAC power system on the affected unit. The components that will be furnished with an alternate power supply from the PSW electrical distribution system are: 1A and 1 B HPI Pumps -Two motor-operated power transfer switches (1 HPISXTRN001 and 1 HPISXTRN002) will be installed to allow remote switching of the power source for the 1A and 1 B HPI pumps. Power will be available from either the pumps normal power supply or from the PSW electrical distribution system. The motor operated transfer switches will be operated via new control switches that were installed in the main control room (MCR) by EC 91830. Since only one switchgear in the PSW electrical distribution system has been provided to support Unit 1 HPI pump operation, a separate manual alignment switch (1HPISXALGN001) will be installed for the purpose of selecting which HPI pump (1A or 1B) will be powered and operated from the designated PSW switchgear.

1 A HPI BWST Suction Valve (1 HP-24) -A new power transfer switch (1 HPISXTRN003) will be installed to allow remote switching of the source of motive and control power for valve 1 HP-24. Power will be available from either the valves normal power supply or ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 5, Page 12 from the PSW electrical distribution system. The power transfer switch will be operated by a new control switch that was installed in the MCR by EC 91830.1 A HPI to RCS Loop A Injection Valve (1HP-26) -A new power transfer switch (1 HPISXTRN004) will be installed to allow remote switching of the source of motive and control power for valve 1 HP-26. Power will be available from either the valves normal power supply or from the PSW electrical distribution system. The power transfer switch will be operated by a new control switch that was installed in the MCR by EC 91830.*R.CS Loop Higqh Point Vent Valves (1RC-155, -156. -157 and -158) and Reactor Vessel Head Vent Valves (1RC-159 and -160) -New control switches installed within the control circuits of valves 1RC-155, -156, -157, -158, -159 and 160 by EC 91830 will be wired into the PSW electrical distribution system. These control switches will permit the switching of the source of motive and control power for valves 1RC-155, -156, -157,-158, -159 and 160. Power will be available from either the valves normal power supply or from the PSW electrical distribution system.In addition, EC 91834 will replace existing manually operated RCP Seal Flow Isolation Valve (1 HP-i139) and RCP Seal Flow Control Bypass Valve (1HP-1 40) with electric motor operated (EMO) valves and will power these valves from the PSW electrical distribution system. These new EMO valves will provide the ability to remotely control flow to the RCP seals after a postulated failure of the existing RCP Seal Flow Control Valve (1 HP-31). Finally, EC 91834 will add a PSW piping tie-in to the existing low pressure service water (LPSW) system line to provide the capability to provide cooling water to the HPI pump motor coolers from the PSW system.AIR 444959 is the 10 CFR 50.59 Screening that was prepared for EC 91834, Rev. 29. The screening concluded that only one aspect of EC 91834 resulted in an adverse effect to the design function of an SSC, as described in the UFSAR. Therefore, the scope of this 10 CFR 50.59 Evaluation will address the components installed by EC 91834 that created this adverse effect.10 CFR 50.59 Screening AR 444959444959concluded that the installation of new power transfer switch 1HPISXTRN003 vA'ill have an adverse effect on the design function of valve 1HP-24. A motor contactor within power transfer switch 1HP1SXTRNOO3 must be energized and its contacts must be in the "closed" position for the existing normal power supply to be available to valve 1 HP-24. Failure of the contactor will result in the normal power supply to valve 1 HP-24 being unavailable.

Without its normal power supply, valve I1HP-24 will not be able to actuate in response to an engineered safeguards actuation signal.A failure of new power transfer switch 1 HPISXTRN003 will render the normal power supply to valve 1 HP-24 unavailable and thereby prevent valve 1 HP-24 from operating when required.However, neither the operation of valve 1HP-24 nor its failure to operate will initiate any of the accidents or transients that have been evaluated in the UFSAR. Therefore, the installation of power transfer switch 1 HPISXTRN003 will not result in an increase in the frequency of an accident previously evaluated in the UFSAR.Failures of new power transfer switch 1 HPISXTRN003 have been evaluated and determined to be within the existing single failure design criteria of the HPI system. Also, the new power transfer switch will be furnished as QA-1, will be seismically qualified, and will be seismically installed.

Hence, the new power transfer switch will not result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety, and will have no ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 5, Page 13 increase in the consequences of either an accident or a malfunction of an SSC important to safety.New power transfer switch 1HPISXTRN003 will not introduce any equipment failures or malfunctions that are new or that have a different result than what is considered credible for the existing HPI system. Therefore, the new power transfer switch will not create a new type of accident or a malfunction with a different result.The MSLB, REA, small-break LOCA and SGTR accident analyses credit the availability and performance of the HPI system to protect fuel cladding, the reactor coolant pressure boundary, and the containment.

Valve 1 HP-24 is specifically credited for being one of two redundant flow paths between the BWST and the HPI pumps and for being sufficiently open within 14 seconds of receiving an ES signal to allow the required flow from the HPI system. New power transfer switch 1 HPISXTRN003 will have no impact on the ability of at least one of the two redundant flow paths between the BWST and the HPI pumps (via valve 1HP-24 or valve 1HP-25) to automatically open and be sufficiently open within 14 seconds of receiving an ES signal.Therefore, the installation of new power transfer switch 1 HPISXTRN003 does not result in a design basis limit for a fission product barrier as described in the UFSAR being exceeded or altered.A revision to the Technical Specifications will not be required.Title: EC 91857, REV. 19 (00200953)

UNIT 2 HPI ALTERNATE POWER (OUTAGE), EC 91858, REV. 7 (0D200954)

UNIT 2 HPI ALTERNATE POWER (PRE-OUTAGE), 10 CFR 50.59 EVALUATION, REV. 0 (AR 00445198 / 01345198)Summary: This 10 CFR 50.59 Evaluation addresses engineering change packages EC 91857, Rev. 19, and EC 91858, Rev. 7, which have .been developed to provide an alternate power source from the PSW electrical distribution system to selected components in the Unit 2 High Pressure Injection (HPI) system and the Unit 2 Reactor Coolant System (RCS) to support core reactivity control, reactor coolant inventory control, and reactor coolant pressure control functions following events (fire and HELB) that damage the essential 4160VAC power system on the affected unit. The components that will be furnished with an alternate power supply from the PSW electrical distribution system are: 2A and 2B HPI Pumps -Two motor-operated power transfer switches (2HPISXTRN001 and 2HPISXTRN002) will be installed to allow remote switching of the power source for the 2A and 2B HPI pumps. Power will be available from either the pumps normal power supply or from the PSW electrical distribution system. The motor operated transfer switches will be operated via new control switches that were installed in the main control room (MCR) by EC 91852. Since only one switchgear in the PSW electrical distribution system has been provided to support Unit 2 HPI pump operation, a separate manual alignment switch (2HPISXALGN001 ) will be installed for the purpose of selecting which HPI pump (2A or 2B) will be powered and operated from the designated PSW switchgear.

• 2A HPI BWST Suction Valve (2HP-24) -A new power transfer switch (2HPISXTRN003) will be installed to allow remote switching of the source of motive and control power for valve 2HP-24. Power will be available from either the valves normal ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 5, Page 14 power supply or from the PSW electrical distribution system. The power transfer switch will be operated by a new control switch that was installed in the MCR by EC 91852.* 2A HPI to RCS Loop A Iniection Valve (2HP-26) -A new power transfer switch (2HPISXTRN004) will be installed to allow remote switching of the source of motive and control power for valve 2HP-26. Power will be available from either the valves normal power supply or from the PSW electrical distribution system. The power transfer switch will be operated by a new control switch that was installed in the MCR by EC 91852.*RCS Loop Higqh Point Vent Va yves (2RC-155, -156, -157 and -158) and Reactor Vessel Head Vent Valves (2RC-159 and -160) -New control switches installed within the control circuits of valves 2RC-155, -156, -157, -158, -159 and 160 by EC 91852 will be wired into the PSW electrical distribution system. These control switches will permit the switching of the source of motive and control power for valves 2RC-155, -156, -157,-158, -159 and 160. Power will be available from either the valves normal power supply or from the PSW electrical distribution system.In addition, EC 91857 and EC 91858 will replace existing manually operated RCP Seal Flow Isolation Valve (2HP-1 39) and RCP Seal Flow Control Bypass Valve (2HP-1 40) with electric motor operated (EMO) valves and will power these valves from the PSW electrical distribution system. These new EMO valves will provide the ability to remotely control flow to the RCP seals after a postulated failure of the existing RCP Seal Flow Control Valve (2HP-31).

Finally, EC 91857 and EC 91858 will add a PSW piping tie-in to the existing Low Pressure Service Water (LPSW) system line to provide the capability to provide cooling water to the HPI pump motor coolers from the PSW system.AR 445081445081is the 10 CFR 50.59 screen that was prepared for EC 91857, Rev. 19, and EC 91858, Rev. 7. The screen concluded that two aspects of EC 91857 and EC 91858 resulted in adverse effects to the design function of an SS0, as described in the UFSAR. Therefore, the scope of this 10 CFR 50.59 Evaluation will address the components installed by EC 91857 and EC 91858 the created these adverse effects.10 CFR 50.59 Screen AR 445081445081concluded that the installation of new power transfer switch 2HPISXTRN003 will have an adverse effect on the design function of valve 2HP-24. A motor contactor within power transfer switch 2HPISXTRN003 must be energized and its contacts must be in the "closed" position for the existing normal power supply to be available to valve 2HP-24.Therefore, for the existing normal power supply to be available to valve 2HP-24, a motor contactor within power transfer switch 2HPISXTRN003 must be energized and must change from the "open" to the "closed" position.

Failure of the contactor will result in the normal power supply to valve 2HP-24 being unavailable.

Without its normal power supply, valve 2HP-24 will not be able to actuate in response to an engineered safeguards actuation signal.10 CFR 50.59 Screen AR 445081445081concluded that the installation of new power transfer switch 2HPISXTRN004 will have an adverse effect on the design function of valve 2HP-26. A motor contactor within power transfer switch 2HPISXTRN004 must be energized and its contacts must be in the "closed" position for the existing normal power supply to be available to valve 2HP-26.Therefore, for the existing normal power supply to be available to valve 2HP-26, a motor contactor within power transfer switch 2HPISXTRN004 must be energized and must change from the "open" to the "closed" position.

Failure of the contactor will result in the normal power supply to valve 2HP-26 being unavailable.

Without its normal power supply, valve 2HP-26 will not be able to actuate in response to an engineered safeguards actuation signal.

3 a ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 5, Page 15 A failure of new power transfer switch 2HPISXTRN004 will render the normal power supply to valve 2HP-26 unavailable and thereby prevent valve 2HP-26 from operating when required.However, neither the operation of valve 2HP-26 nor its failure to operate will initiate any of the accidents or transients that have been evaluated in the UFSAR. Therefore, the installation of power transfer switch 2HPISXTRN004 will not result in an increase in the frequency of an accident previously evaluated in the UFSAR.Failures of new power transfer switch 2HPISXTRN004 have been evaluated and determined to be within the existing single failure design criteria of the HPI system. Also, the new power transfer switch will be furnished as QA-1, will be seismically qualified, and will be seismically installed.

Hence, the new power transfer switch will not result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety, and will have no increase in the consequences of either an accident or a malfunction of an SSC important to safety.New power transfer switch 2HPISXTRN004 will not introduce any equipment failures or malfunctions that are new or that have a different result than what is considered credible for the existing HP! system. Therefore, new power transfer switch 2HPISXTRN004 will not create a new type of accident or a malfunction with a different result.The MSLB, REA, small-break LOCA and SGTR accident analyses credit the availability and performance HPI system to protect fuel cladding, the reactor coolant pressure boundary, and the containment.

Valve 2HP-24 is specifically credited for being one of two redundant flow paths between the BWST and the HPI pumps and for being sufficiently open within 14 seconds of receiving an ES signal to allow the required flow from the HPI system. New power transfer switch 2HPISXTRN003 will have no impact on the ability of at least one of the two redundant flow paths between the BWST and the HPI pumps (via valve 2HP-24 or valve 2HP-25) to automatically open and be sufficiently open within 14 seconds of receiving an ES signal.Therefore, the installation of new power transfer switch 2HPISXTRN003 does not result in a design basis limit for a fission product barrier as described in the UFSAR being exceeded or altered.The MSLB, REA, small-break LOCA and SGTR accident analyses credit the availability and performance HPI system to protect fuel cladding, the reactor coolant pressure boundary, and the containment.

Valve 2HP-26 is specifically credited for aligning the HPI discharge header to the reactor vessel and for being sufficiently open within 14 seconds of receiving an ES signal to allow the required flow from the HPI system. New power transfer switch 2HPISXTRN004 will have no impact on the ability of the HPI discharge path to automatically open and be sufficiently open within 14 seconds of receiving an ES signal. Therefore, the installation of new power transfer switch 2HPISXTRN004 does not result in a design basis limit for a fission product barrier as described in the UFSAR being exceeded or altered.A revision to the Technical Specifications will not be required.

SIs' , ONS-201 5-072 Document Control Desk~June 30, 2015 Attachment 5, Page 16 Title: EC 109523, Revision 004- UI&2 BWST RECIRC ISOLATION VALVES, 10 CFR 50.59 Revision 0 (AR 00445282 / 01345282)Summary: The Nuclear Regulatory Commission (NRC) issued amendments for the Reverse Osmosis (RO)system by correspondence dated April 30, 2014. The connection from the BWSTs to the RO System includes redundant automatically actuated, safety-related, seismically qualified isolation valves between the RO System supply piping and the Borated Water Storage Tanks (BWSTs).These automatically actuated isolation valves also isolate the BWST recirculation pump from the BWST. The isolation valves actuate to close on declining BWST level before BWST TS level is reached, thereby isolating RO and Spent Fuel Pool Cooling (SFPC) purification systems from BWST prior to entering Reactor Building Emergency Sump (RBES) recirculation phase following drawdown of the BWST. Isolation of RO and SEPC purification systems prevents unanalyzed consequences from leakage from BWST into those systems' piping when in the RBES recirculation phase. The redundant control circuitry of the isolation valves contain digital pressure transmitters which actuate on low BWST level to close the valves. Actuation of any one of 4 digital pressure transmitters will close both redundant valves and secure the BWST recirculation pump. Loss of power to the circuit, or loss of air to the valves will result in the valves closing.The digital pressure transmitter used in this application has self-diagnostics and is configured to D fail LO which closes the redundant valves. The digital pressure transmitter is demonstrated to be a very reliable device and has received a generic qualification from EPRI for use in mild environment nuclear applications.

The evaluation demonstrates that the proposed control circuitry for the redundant isolation valves will preserve the current licensing basis. The activity will not create more than a minimal increase in the frequency or consequences of accidents or malfunctions of SSCs important to safety. The proposed activity will not create the potential for a new type of unanalyzed event, has no impact on the fission product barriers and does not affect evaluation methodology.

The answers to all eight evaluation questions is negative, no Tech Spec modification, deletion, or addition is required and no UFSAR revision is required.

Therefore under 10 CFR 50.59 it is permissible to implement this modification without prior approval from the NRC.Title: EC 91877, REV. 30, PSW SYSTEM HEADER; EC 91878, REV. 1, REPLACE ASW SYSTEM WITH PSW SYSTEM; EC 106526, REV. 0, INSTALL MOTOR FEEDER CABLES FOR PSW PUMPS; EC 111881, REV. 12, PSW PUMP ROOM DUCT EXTENSION; 10 CFR 50.59 EVALUATION REV. 0 (AR 00445554 / 01345554)Summary: EC 91877, EC 91878, EC 106526 and EC 111881 will replace the existing ASW system with the PSW system. The new PSW system will provide an increase in mechanical capacity, as compared to the ASW system, such that it will be capable of injecting raw water into the secondary side of the steam generators without first depressurizing the steam generators.

The new PSW system will also be furnished with the necessary power-operated valves, b instrumentation, and controls to permit both initiation and operation of the system from the I main control room.

ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 5, Page 17 Replacement of the ASW system with the PSW system will require NRC approval based on the following:

• Duration of Decay Heat Removal -After a loss of Lake Keowee event, the stored water inventory in the CCW intake and discharge piping is credited for providing 37 days of decay heat removal capability via the existing ASW system. However, the PSW system will only be credited for providing decay heat removal for a minimum of 30 days, regardless of whether it is placed into operation to mitigate a loss of Lake Keowee event or a tornado event. In addition, the stored water inventory in the CCW intake and discharge piping has only been demonstrated adequate for 30 days of PSW system operation.

As such, the replacement of the ASW system with the PSW system results in an increase in the probability of failure of the system that is credited for providing 37 days of decay heat removal after a loss of Lake Koewee event.Therefore, the new PSW system results in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the UFSAR.*Use of Alternate Criteria for Postulating High Energy Line Breaks (HELBs) -The current ONS criteria for postulating HELBs would require that HELBs be considered in the new PSW system based on its normal operating pressure.

However, alternate criteria, which is based on the actual time that the system is in operation, has been applied to the PSW system. Using this alternate criteria, HELBs in the PSW system are not required to be postulated.

The application of this alternate criteria constitutes a departure from a method of evaluation described in the UFSAR used in establishing the design basis.Replacement of the ASW system with the PSW system will not require a revision to the Technical Specifications.

Title: EC 0000110806

-005 -UNIT 1 & 2 600/208VAC CABLE TRAY SYSTEM FOR PSW SCENARIO -10 CFR 50.59 Evaluation of Selected Unit 1 & 2 Auxiliary Building Penetrations) (AR 00445630 / 01345630)Summary: The installation of four (4) penetrations in the walls, floors, and ceilings of the Unit 1 Auxiliary Building were determined to affect from radiation the qualification of equipment located proximate to the penetration installation areas. Using the guidance of NEI 96-07, the appropriate review of such adverse effects is documented by performing the 10 CER 50.59 Evaluation summarized here. Engineering Change (EC) 110806 provided the evaluations to update the Oconee Equipment Qualification Criteria Manual (EQCM) to demonstrate that with these four (4)penetrations installed, plant equipment proximate to the penetration locations continued to be capable of reliably performing their important to safety functions, and the penetration installations do not create initiators of accidents previously evaluated in the UFSAR.Consequences for accidents previously described in the UFSAR are not increased, nor will these penetration installations increase accident dose consequences through a malfunction of an SSC important to safety previously evaluated in the UFSAR. These activities do not create the possibility for an accident of a different type than any previously evaluated in the UFSAR, nor do they create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in the UFSAR. No design basis for a fission product barrier

-~ p ONS-201 5-072 Document Control Desk June 30, 2015 Attachment 5, Page 18 as described in the UFSAR is being exceeded or altered, and these activities do not result in a departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses.Title: Revise SLC 16.9.16, Reactor Building Polar Crane and Auxiliary (Control Rod Drive) Hoist -(RCS System Open) (AR 00445799 / 01345799)Sum mary: This proposed activity is a change to Selected Licensee Commitment 16.9.16, Reactor Building Polar Crane and Auxiliary (Control Rod Drive) Hoist -(RCS System Open). The change is a revision to the Bases of the SLO. This change is needed to allow more flexibility in the use of the polar crane hoist and auxiliary hoist over the fuel transfer canal.The Bases are to be revised as follows: Current Version: Use of either reactor building polar crane hoist and auxiliary (CRD) hoist over the fuel transfer canal when the reactor vessel head is removed is restricted to those operations necessary for the fuel handling and core internals operations.

Proposed Revised Bases: Use of either reactor building polar crane hoist and auxiliary (CRD) hoist over the fuel transfer canal when the reactor vessel head is removed is restricted to those operations necessary for the fuel handling and core internals operations or other specific limited activities that have been analyzed appropriately to demonstrate that the activity is incapable of resultinq in a radioloqical release in the event of a load drop onto the reactor core.*By not allowing activities which could cause a radiological release, this SLC change is bounded by the Fuel Handling accident already evaluated in the UFSAR.