ML21210A135

2 to Updated Final Safety Analysis Report, Chapter 15, Structures and Construction (EPID L-2020-LRO-0076) - Redacted

ML21210A135
Person / Time
Site:	Surry
Issue date:	09/30/2020
From:	Virginia Electric & Power Co (VEPCO)
To:	Office of Nuclear Reactor Regulation
Thomas V
Shared Package
ML21208A006	List: ML21208A367 ML21208A409 ML21208A418 ML21208A424 ML21208A429 ML21210A125 ML21210A126 ML21210A127 ML21210A128 ML21210A129 ML21210A133 ML21210A135 ML21210A136 ML21210A137 ML21210A138 ML21210A139 ML21210A140
References
20-325, EPID L-2020-LRO-0076
Download: ML21210A135 (194)

Text

Intentionally Blank Revision 52Updated Online 09/30/20 SPS UFSAR 15-i Chapter 15: Structures and Construction Table of Contents Section Title Page 15.1 STRUCTURES AND MACHINERY ARRANGEMENT . . . . . . . . . . . . . . . . . . 15.1-1 15.1 Reference Drawings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.1-2 15.2 STRUCTURAL DESIGN CRITERIA . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.2-1 15.2.1 General . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.2-1 15.2.2 Normal Wind Loading . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.2-2 15.2.3 Tornado Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.2-2 15.2.4 Seismic Design . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.2-4 15.2.5 Hydrostatic Loadings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.2-6 15.2 References. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.2-8 15.3 MATERIAL . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.3-1 15.3.1 Concrete . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.3-1 15.3.1.1 Cement. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.3-1 15.3.1.2 Admixtures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.3-1 15.3.1.3 Water . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.3-1 15.3.1.4 Aggregates . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.3-2 15.3.1.5 Proportioning . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.3-2 15.3.2 Reinforcing Steel. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.3-2 15.4 CONSTRUCTION PROCEDURES AND PRACTICES . . . . . . . . . . . . . . . . . . . 15.4-1 15.4.1 Codes of Practice. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.4-1 15.4.2 Concrete . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.4-2 15.4.3 Reinforcing Steel. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.4-4 15.4.4 Construction Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.4-5 15.4.5 Construction Practice . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.4-5 15.4.6 Quality Assurance Program (Construction Phase) . . . . . . . . . . . . . . . . . . . . . . . 15.4-5 15.5 SPECIFIC CONTAINMENT STRUCTURAL DESIGNS . . . . . . . . . . . . . . . . . . 15.5-1 15.5.1 Containment Structure. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.5-1 15.5.1.1 General. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.5-1 15.5.1.2 Design Criteria. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.5-2 15.5.1.3 Buoyant Loads . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.5-4 15.5.1.4 Dynamic Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.5-4 15.5.1.5 Static Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.5-8

Revision 52Updated Online 09/30/20 SPS UFSAR 15-ii Chapter 15: Structures and Construction Table of Contents (continued)

Section Title Page 15.5.1.6 Reinforcing Steel Arrangement . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.5-12 15.5.1.7 Penetration Design. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.5-14 15.5.1.8 Steel Liner and Penetrations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.5-15 15.5.1.9 Materials . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.5-29 15.5.1.10 Construction Procedures and Practices . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.5-31 15.5.1.11 Missiles and Piping Rupture . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.5-34 15.5.1.12 Ground Water Protection and Corrosion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.5-36 15.5.1.13 Testing and Inservice Surveillance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.5-36 15.5.2 Reactor Pressure Vessel Head Replacement Project (Applicable to Unit 1 and Unit 2) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.5-41 15.5.2.1 Codes and Specifications. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.5-41 15.5.2.2 Liner Restoration . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.5-42 15.5.2.3 Reinforcing Steel Restoration . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.5-42 15.5.2.4 Concrete Restoration . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.5-43 15.5.2.5 Post Modification Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.5-44 15.5 References. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.5-45 15.5 Reference Drawings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.5-46 15.6 OTHER CLASS I STRUCTURES. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.6-1 15.6.1 Other Structures. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.6-1 15.6.2 Reactor Coolant System Supports. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.6-2 15.6.2.1 Design Basis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.6-4 15.6.2.2 Description. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.6-6 15.6.3 Containment Internal Structure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.6-10 15.6 References. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.6-13 15.7 MASONRY WALLS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.7-1 15.7 References. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.7-2 Appendix 15A Seismic Design for the Nuclear Steam Supply System and Miscellaneous Components . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15A-i

Revision 52Updated Online 09/30/20 SPS UFSAR 15-iii Chapter 15: Structures and Construction Table of Contents (continued)

Section Title Page 15A.1 GENERAL SEISMIC DESIGN CRITERIA FOR THE NUCLEAR STEAM SUPPLY SYSTEM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15A-1 15A.2 SEISMIC DESIGN CRITERIA FOR PIPING, VESSELS, SUPPORTS AND REACTOR VESSEL INTERNALS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15A-1 15A.2.1 Loading Condition Definitions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15A-2 15A.2.1.1 Normal Conditions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15A-2 15A.2.1.2 Upset Conditions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15A-2 15A.2.1.3 Emergency Conditions. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15A-2 15A.2.1.4 Faulted Conditions. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15A-2 15A.2.2 Piping, Vessels, and Supports . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15A-3 15A.2.3 Reactor Vessel Internals . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15A-4 15A.2.3.1 Design Criteria for Normal Operation. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15A-4 15A.2.3.2 Design Criteria for Abnormal Operation. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15A-4 15A.3 GENERAL ANALYTICAL PROCEDURE FOR SEISMIC DESIGN . . . . . . . . . 15A-4 15A.3.1 Mechanical Equipment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15A-5 15A.3.2 Earthquake Experience-Based Method Developed for Unresolved Safety Issue (USI) A-46 for Seismic Verification of Equipment. . . . . . . . . . . . . . . . . . 15A-7 15A.3.3 Reactor Coolant Loops and Supports . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15A-8 15A.3.4 Anchor Bolts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15A-11 15A.3.5 Piping Systems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15A-13 15A.3.5.1 Reanalysis Methods and Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15A-14 15A.3.5.2 Verification of Analysis Methods . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15A-16 15A.3.5.3 Soil Structure Interaction. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15A-18 15A.4 MOVEMENT OF REACTOR COOLANT SYSTEM COMPONENTS. . . . . . . . 15A-22 15A.5 TESTS TO DEMONSTRATE THE CONSERVATISM OF THE LIMIT CURVES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15A-23 15A.5.1 Westinghouse Topical Reports . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15A-23 15A.5.2 Framatome Computer Programs (Unit 1 only) . . . . . . . . . . . . . . . . . . . . . . . . . . 15A-24 15A.5.2.1 BWSPAN. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15A-25 15A.5.2.2 BIJLAARD . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15A-25 15A.5.2.3 FERMETURE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15A-25 15A.5.2.4 SYSTUS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15A-25 15A.5.2.5 RCCM-ASME . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15A-25

Revision 52Updated Online 09/30/20 SPS UFSAR 15-iv Chapter 15: Structures and Construction Table of Contents (continued)

Section Title Page 15A.6 REACTOR COOLANT LOOP (RCL) PIPING REANALYSIS SUBSEQUENT TO LEAK BEFORE BREAK AND SNUBBER ELIMINATION . . . . . . . . . . . . . . . 15A-26 15A References. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15A-26

Revision 52Updated Online 09/30/20 SPS UFSAR 15-v Chapter 15: Structures and Construction List of Tables Table Title Page Table 15.2-1 Structures, Systems, and Components Designed for Seismic and Tornado Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.2-9 Table 15.2-2 Damping Factors for Class I Structures . . . . . . . . . . . . . . . . . . . . . . . . . 15.2-28 Table 15.5-1 Containment Structural Loading Criteria. . . . . . . . . . . . . . . . . . . . . . . . 15.5-47 Table 15.5-2 Capacity Reduction Factor for Concrete . . . . . . . . . . . . . . . . . . . . . . . . 15.5-48 Table 15.5-3 Missile Dimensions and Weights Required to Penetrate Plate of Varying Thicknesses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.5-48 Table 15.5-4 Comparison of Stresses Under Test Pressure With Stresses Under Incident Conditions and Earthquake Plus Incident Conditions . . . . . . . . . . . . . . 15.5-49 Table 15.6-1 Steam Generator Support Materials . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.6-14 Table 15.6-2 Reactor Coolant Pump Support Materials . . . . . . . . . . . . . . . . . . . . . . . 15.6-14 Table 15.6-3 Summary of Stress for Failure of Reactor Coolant Pump Support During Normal Operation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.6-15 Table 15A-1 Loading Conditions and Stress Limits . . . . . . . . . . . . . . . . . . . . . . . . . . 15A-32 Table 15A-2 Minimum Margins of Safety . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15A-34 Table 15A-3 Tests and Test Results on SA 106B Carbon Steel Pipe Specimens (Internal Pressurization = 3000 psia) . . . . . . . . . . . . . . . . . . . . . . . . . . . 15A-35 Table 15A-4 Tests and Test Results on 304 Stainless Steel Specimens (Internal Pressurization = 3000 psia) . . . . . . . . . . . . . . . . . . . . . . . . . . . 15A-36 Table 15A-5 Level of Stress as a Percentage of Code Allowable Stress . . . . . . . . . . 15A-37 Table 15A-6 Factors of Safety for Component Supports Under Design-Basis Seismic and Normal Operating Loads Surry Units 1 and 2. . . . . . . . . . 15A-38 Table 15A-7 Factors of Safety for Component Supports Under Combined Accident Loads Surry Units 1 and 2 . . . . . . . . . . . . . . . . . . . . . . . . . . . 15A-39 Table 15A-8 Calculated Stress as Percentage of Code Allowable Stress (REFERENCE 69) 15A-40 Table 15A-9 Factors of Safety for Steam Generator and Reactor Coolant Pump Supports (REFERENCE 70) . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15A-41

Revision 52Updated Online 09/30/20 SPS UFSAR 15-vi Chapter 15: Structures and Construction List of Figures Figure Title Page Figure 15.1-1 Site Plan . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.1-3 Figure 15.1-2 Plot Plan . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.1-4 Figure 15.5-1 Reactor Containment Waterproofing . . . . . . . . . . . . . . . . . . . . . . . . . . 15.5-50 Figure 15.5-2 Containment Loading Plot . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.5-51 Figure 15.5-3 Reinforcing Details Equipment Access Hatch Opening . . . . . . . . . . . 15.5-54 Figure 15.5-4 Reinforcing Details Sections Through Ring Beam Equipment Access Hatch . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.5-55 Figure 15.5-5 Reinforcing Details Personnel Hatch Opening . . . . . . . . . . . . . . . . . . 15.5-56 Figure 15.5-6 Reinforcing Details Sections Through Ring Beam Personnel Hatch . 15.5-57 Figure 15.5-7 Wall and Mat Joint . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.5-58 Figure 15.5-8 Section-Typical Bridging Bar . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.5-59 Figure 15.5-9 Typical Electrical Penetration Sleeve With Flanges . . . . . . . . . . . . . . 15.5-61 Figure 15.5-10 Typical Piping Penetrations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.5-62 Figure 15.5-11 Personnel Hatch Assembly . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.5-64 Figure 15.5-12 Typical Liner Details . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.5-65 Figure 15.5-13 Containment Loading Plot . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.5-66 Figure 15.5-14 Machine Shop Replacement Fac.; South Elevation . . . . . . . . . . . . . . . 15.5-67 Figure 15.5-15 Machine Shop Replacement Fac.; Site Plan . . . . . . . . . . . . . . . . . . . . 15.5-68 Figure 15.5-16 Machine Shop Replacement Fac.; East/West elevs. . . . . . . . . . . . . . . 15.5-69 Figure 15.6-1 Reactor Neutron Shield Tank Assembly . . . . . . . . . . . . . . . . . . . . . . . 15.6-16 Figure 15.6-2 Steam Generator Support Assembly . . . . . . . . . . . . . . . . . . . . . . . . . . 15.6-17 Figure 15.6-3 Reactor Coolant Pump Supports General Arrangement . . . . . . . . . . . 15.6-18 Figure 15.6-4 Pressurizer Support . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.6-19 Figure 15A-1 Envelope for 0.5% Damping Ground Response Spectra-Average Gmax 15A-42 Figure 15A-2 Envelope for 0.5% Damping Ground Response Spectra-Average Gmax + 50% . . . . . . . . . . . . . . . . 15A-43 Figure 15A-3 Envelope for 0.5% Damping Ground Response Spectra-Average Gmax -50% . . . . . . . . . . . . . . . . . 15A-44 Figure 15A-4 Design Limits Compared to Experimental Points, SA 106B Carbon Steel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15A-45 Figure 15A-5 Design Limits Compared to Experimental Points, 304 Stainless Steel 15A-46

Revision 52Updated Online 09/30/20 SPS UFSAR 15-vii Chapter 15: Structures and Construction List of Figures (continued)

Figure Title Page Figure 15A-6 Typical Stress-Strain Curve, 304 Stainless Steel . . . . . . . . . . . . . . . . . 15A-47 Figure 15A-7 Typical Stress-Strain Curve, Inconel 600 . . . . . . . . . . . . . . . . . . . . . . 15A-48 Figure 15A-8 Typical Stress-Strain Curve, SA 302 Grade B. . . . . . . . . . . . . . . . . . . 15A-49

Revision 52Updated Online 09/30/20 SPS UFSAR 15-viii Intentionally Blank

Revision 52Updated Online 09/30/20 SPS UFSAR 15.1-1 CHAPTER 15 STRUCTURES AND CONSTRUCTION 15.1 STRUCTURES AND MACHINERY ARRANGEMENT The site arrangement, plot plan, and the general arrangement of equipment within the principal Class I structures are shown on the Figures and Reference Drawings listed in the following tabulation:

Item Reference Drawing Site Plan Figure 15.1-1 Plot Plan Figure 15.1-2 and Reference Drawing 1 Containment Structure and Containment Auxiliary Structures Reference Drawings 2 through 8 Auxiliary Building Reference Drawings 9, 10, 11, & 12 Fuel Building Reference Drawings 13 & 14 Control Area Reference Drawing 15

Revision 52Updated Online 09/30/20 SPS UFSAR 15.1-2 15.1 REFERENCE DRAWINGS The list of Station Drawings below is provided for information only. The referenced drawings are not part of the UFSAR. This is not intended to be a complete listing of all Station Drawings referenced from this section of the UFSAR. The contents of Station Drawings are controlled by station procedure.

Drawing Number Description

1. 11448-FY-1D Plot Plan

2. 11448-FM-1A Machine Location: Reactor Containment, Elevation 47'- 4"

3. 11448-FM-1B Machine Location: Reactor Containment, Elevation 18'- 4"

4. 11448-FM-1C Machine Location: Reactor Containment, Elevation 3'- 6"

5. 11448-FM-1D Machine Location: Reactor Containment, Elevation 27'- 7"

6. 11448-FM-1E Machine Location: Reactor Containment; Sections A-A, E-E,

& Z-Z

7. 11448-FM-1F Machine Location: Reactor Containment; Sections B-B, X-X,

& Y-Y

8. 11448-FM-1G Machine Location: Reactor Containment, Sections C-C & D-D

9. 11448-FM-5A Arrangement: Auxiliary Building

10. 11448-FM-5B Arrangement: Auxiliary Building, Unit 1

11. 11448-FM-5C Arrangement: Auxiliary Building

12. 11448-FM-5D Arrangement: Auxiliary Building

13. 11448-FM-9A Arrangement: Fuel Building, Sheet 1

14. 11448-FM-9B Arrangement: Fuel Building, Sheet 2, Unit 1

15. 11448-FA-1E Control and Relay Room Service Building

Revision 52Updated Online 09/30/20 SPS UFSAR 15.1-3 Figure 15.1-1 SITE PLAN

REDACTED Revision 52Updated Online 09/30/20 SPS UFSAR 15.2-1 15.2 STRUCTURAL DESIGN CRITERIA 15.2.1 General The structures, systems, and components of the Surry Power Station, Units No. 1 and No. 2, are classified into groupings requiring seismic, tornado or conventional design. The effects of the Power Uprating to a core power of 2546 MWt on pipe stress and supports were reviewed for the systems listed below. The review determined that the existing piping and support configuration is adequate to withstand the increase in pressure and temperature associated with the Power Uprating.

Systems: Main Steam Condensate Extraction System H. P. Heater Drain L. P. Heater Drains Reactor Coolant Class I design encompasses those structures, systems or components of reactor facilities that are essential to the prevention of accidents that could affect the public health and safety, or to the mitigation of their consequences.

Structures, systems, and components are designed, fabricated, and constructed to performance standards that will enable the facility to withstand, without loss of capability to protect the public, the additional forces that might be imposed by:

1. The operating-basis earthquake and the design-basis earthquake.

2. Tornados and other local site effects including flooding conditions, winds, and ice. Radiation levels that constitute a hazard to the public are defined in 10 CFR 50.67.

A Class I structure is designed for resistance to seismic loadings in accordance with Section 15.2.4 and for tornados, where applicable, in accordance with Section 15.2.3. There are some structures, systems, or components whose loss or failure by earthquake will not affect the public health or safety and will permit safe station shutdown, although their loss could interrupt power generation. These structures, systems, or components are not designed for specific seismic or tornado loadings.

Structures not designed for seismic or tornado loadings are designed according to Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings (AISC-1963), and Building Code Requirements for Reinforced Concrete (ACI 318-63, Part IVA -

Working Stress Design).

Revision 52Updated Online 09/30/20 SPS UFSAR 15.2-2 These structures are designed for dead, live, and normal wind loads using allowable stress levels given in the above codes.

Some structures, systems and components of the station are necessary for a safe and orderly shutdown during a tornado. These structures are designed for tornado loadings, and systems and components are protected by tornado-resistant structures.

A list of the structures, systems, and components designed to satisfy seismic and/or tornado criteria is given in Table 15.2-1.

15.2.2 Normal Wind Loading All structures were designed to withstand the following wind loads applied to the projected area of all surfaces:

Elevation 26 ft. 6 in. to Elevation 56 ft. 6 in., 30 lb/ft2 Elevation 57 ft. 6 in. to Elevation 75 ft. 6 in., 35 lb/ft2 Elevation 75 ft. 6 in. to Elevation 130 ft. 0 in., 45 lb/ft2 Elevation 131 ft. 0 in. and above, 55 lb/ft2 Roofs were designed for uplift using 1.25 times the wind load taken at the corresponding elevation of the roof.

Members subject to stresses produced by this wind load combined with live and dead loads were proportioned for stresses 33-1/3% greater than conventional working stresses, provided that the section thus required is not less than that required for the combination of dead and live loads computed without the one-third increase.

15.2.3 Tornado Criteria Section 2.2 outlines the probability of a tornado occurring at the site. Although no structural damage is known to have resulted to a reinforced concrete building in a tornado (Reference 1), the structures and systems so indicated in Table 15.2-1 are designed to ensure safe shutdown of the reactor when subjected to tornado loadings. The Seismic Class I and Tornado Criterion T structures discussed in Sections Section 15.2.3 and Section 15.2.4, respectively, are primarily of reinforced-concrete construction. The principal components that transmit horizontal and vertical loads to the foundation are the reinforced-concrete roof and floor slabs, and both interior and exterior reinforced-concrete walls. Since these components act as diaphragms, tending to minimize stress concentrations that might otherwise occur (in a column, for example), and their thicknesses are usually controlled by requirements for biological shielding or tornado and interior missile protection, stresses and strains are generally not significant. For these reasons, calculated stresses and strains for selected principal structural components have been omitted from this report. In addition, test data and analytical studies, in accordance with Appendix C of

Revision 52Updated Online 09/30/20 SPS UFSAR 15.2-3 Reference 13, have confirmed that 2-foot thick, reinforced-concrete test specimens, with similar spans and steel reinforcement as those found in SPS Tornado Criterion T structures (Table Table 15.2-1), will not experience a ductility ratio, , in excess of applicable industry code allowable limits (i.e., 10), when subjected to tornado load effects, as described in SPS UFSAR, Section 15.2.3.

The tornado model used for design has the following characteristics:

Rotational velocity 300 mph Translation velocity 60 mph Pressure drop 3 psi in 3 sec Overall diameter 1200 ft Radius of maximum winds 200 ft Applicable structures are designed to resist a maximum wind velocity associated with a tornado of 360 mph, which is obtained by adding the rotational and translational velocities.

Structures and systems are checked for tornado pressure loading, vacuum loading, and the combination of these two.

The tornado wind velocity is converted to an equivalent pressure, which is applied to the structures uniformly using the formula:

P = 0.00256 V2 where:

P = equivalent pressure, lb/ft2 V = wind velocity, mph This pressure is multiplied by applicable shape factors and drag coefficients as given in ASCE Paper No. 3269 by Thomas W. Singell (Reference 2), and applied to the silhouette of the structure.

A reduction of the full negative pressure differential is made when venting of the structures is provided. The amount of the reduction is a function of the venting area provided.

Tornado wind loads are combined with other loads as described in Section 15.5.1.2.

Tornado and earthquake loads are not considered to act simultaneously. A uniform wind velocity and a nonuniform atmospheric pressure gradient is incorporated in the design of the containment structure.

Structural design criteria for tornado loading for the containment structure are given in Table 15.5-1 and Section 15.5.1.5.

Revision 52Updated Online 09/30/20 SPS UFSAR 15.2-4 It is assumed that a tornado could generate either of the following potential missiles:

1. Missile equivalent to a wooden utility pole 40 feet long, 12-inch diameter, weighing 50 lb/ft3 and traveling in a vertical or horizontal direction at 150 mph.

2. Missile equivalent to a 1-ton automobile traveling at 150 mph.

The design assumes maximum wind forces and partial vacuum to occur simultaneously with the impact of either of the missiles singly. Allowable stresses do not exceed 90% of the certified minimum yield strength of the steel, the capacity reduction factor given in Section 15.5.1.2 times the certified minimum yield strength of the reinforcing steel, and 75% of the ultimate strength of the concrete. The allowable stress limits of 0.9 Fy (steel superstructures) and 0.9 fy and 0.75 fc (reinforced concrete structures) apply to stresses from the overall structural response due to tornado load effects. These stresses are located away from the tornado missile impact zone and outside any yield-line patterns that may develop during the tornado missile impact.

It is noted that the physical configuration of certain plant components does not provide complete physical protection against tornado-generated missiles. The vulnerable surface area for each component was assessed probabilistically using the Tornado Missile Risk Evaluator Methodology (Reference 12) and it was determined that the risk to the plant is acceptably low, such that the additional missile protection need not be provided. Refer to Table 15.2-1 for identification of these components.

15.2.4 Seismic Design Class I structures, systems, and components designed to resist seismic forces are listed in Table 15.2-1. The design is based on two separate seismic criteria: the operating-basis earthquake (OBE) and the design-basis earthquake (DBE), as described in Section 2.5.

The seismic analysis of Class I structures, such as the containment structure, auxiliary building, fuel building, service building (including the control room), and safeguard areas, was based on the modal analysis response spectra technique. Major equipment-supporting structures, such as steam generator supports, reactor coolant pump supports, and pressurizer supports, were treated in an identical manner. Acceleration response spectra for the OBE and DBE are given on Figures 2.5-5 and 2.5-6.

Seismic loading includes the horizontal or vertical responses acceleration or combinations of both where the effects, as measured by the separate acceleration components, of horizontal and vertical accelerations are combined to produce maximum stress intensities, taking into account any potential adverse effect due to phase of the separate accelerations.

Damping factors for the structures, systems, and components are given in Table 15.2-2.

The design of the containment structures is based on ultimate strength design and loading factors as described in Section 15.5.1.2. Maximum allowable stress levels for both the

Revision 52Updated Online 09/30/20 SPS UFSAR 15.2-5 operating-basis earthquake and the design-basis earthquake are based on proportions of the minimum yield strength.

For other Class I structures, the operating-basis earthquake loading is combined with dead, live, and other static loads. Normal wind or tornado loadings are not assumed to occur simultaneously with the earthquake loading. Members are proportioned for stresses 33-1/3%

greater than conventional working stresses, provided that the section thus required is not less than that required for the combination of dead and live loads computed without the one-third increase.

Allowable soil-bearing values are increased one-third.

For Class I structures other than the containment structure, the design basis earthquake is combined with static loads using loading combinations given in Table 15.5-1. For these structures under the design-basis earthquake loading, the allowable stresses do not exceed 90% of the certified minimum yield strength for structural steel, the capacity reduction factor, given in Section 15.5.1.2, times the certified minimum yield strength for reinforcing steel, and the capacity reduction factor times the specified strength for concrete. Allowable soil bearing values are increased by one-half.

To allow for unimpeded relative motions between structures, a rattlespace is provided between the:

1. Containment structures and the auxiliary building.

2. Containment structures and the fuel building.

3. Containment structures and the containment auxiliary structures around the periphery of each containment.

4. Fuel building and auxiliary building.

5. Auxiliary building and control area.

In general, the periphery of the rattlespace between buildings is arranged to prevent material entering the space, with the inner areas left as a void.

Maximum relative motions between adjoining structures are included in the stress analyses of all piping that extends from one building to another.

Type A sand, as described in Section 2.4.3.3, was removed from under the fuel building, auxiliary building, and control area and replaced by a dense graded granular fill material as described in Section 2.4.5.1.

The analytical procedure used for the nuclear steam supply system is described by Section 15A.3 of Appendix 15A.

The reactor protection system, engineered safety feature (ESF) circuits, and the emergency power system are designed so that they will not lose the capability to shut the plant down and

Revision 52Updated Online 09/30/20 SPS UFSAR 15.2-6 maintain it in a safe shutdown condition under operating-basis earthquake or design-basis earthquake conditions. For the design-basis earthquake, permanent deformation of the equipment is allowable, provided that the capability to perform its function is maintained.

Typical protection system equipment is subjected to type tests under simulated seismic accelerations to demonstrate its ability to perform its functions. Type testing was performed using conservatively large accelerations and applicable frequencies. Analyses were done for structures that were not done for the reactor protection system equipment. However, the peak accelerations and frequencies were checked against those derived by structural analyses of operational and design-basis earthquake loadings.

A Westinghouse topical report, WCAP-7397-L, provides the seismic evaluation of safety-related equipment. The type tests covered by this report are applicable to the Surry Power Station, with the exception of the process control equipment, which is covered in a supplement to WCAP-7397-L.

The emergency switchgear has been tested under seismic conditions, and manufacturers test data are available. The emergency generator and control panels are identical with those used in locomotives, and have been tested under severe conditions, but no seismic tests have been made.

The control board was designed to withstand earthquake conditions, and an analysis was performed to verify the adequacy of the seismic design, but tests were not performed.

The emergency batteries are supported on rigid reinforced concrete pedestals firmly anchored to the floor. Steel retaining members integral with the pedestals prevent the batteries from being dislodged under seismic excitation.

15.2.5 Hydrostatic Loadings Finish ground grade at the station is at Elevation 26 ft. 6 in. Natural ground water level is at approximately Elevation 4 ft. 0 in.

The exterior wall of the containment structure extends approximately 66 feet below the finished ground level. Water-resistant membrane protection for this structure is defined in Sections 15.5.1.9 and 15.5.1.10. External pumps for reducing the hydrostatic head on the containment structure are described in Section 15.5.1.3. This latter section also discusses the effect of the buoyant pressure on the containment structure.

Exterior surfaces of walls of other Class I structures with floor levels below Elevation 26 ft.

6 in. are covered with a mopped-on bitumastic coating to establish a water-resistant membrane.

The roofs of safety-related structures and select nonsafety-related structures with the potential to impact structures, systems, or components important to safety are evaluated to withstand hydrostatic surcharge loading associated with a Beyond Design Basis (BDB) Local

Revision 52Updated Online 09/30/20 SPS UFSAR 15.2-7 Intense Precipitation (LIP) rainfall event. Roof parapets of the Turbine Building, Service Building, and Condensate Polishing Building feature cutouts to passively limit the hydrostatic surcharge loading induced by the postulated BDB event.

Revision 52Updated Online 09/30/20 SPS UFSAR 15.2-8

15.2 REFERENCES

1. V. C. Gilberton and E. E. Mageanu, Tornadoes, AIA Technical Reference Guide, TRG 13-2, U. S. Weather Bureau.

2. T. W. Singell, Wind Forces on Structures: Forces on Enclosed Structures, Journal of the Structural Division of the ASCE, July 1958.

3. Deleted.

4. Letter, NRC to Vepco, Serial #85-885, dated December 4, 1985

5. ASME Boiler and Pressure Vessel Code,Section III, Division I, Code Case N-411, Alternative Damping Values for Seismic Analysis of Piping Sections, American Society of Mechanical Engineers, 345 E. 47th Street, New York, NY 10017, dated September 17, 1984.

6. NRC Bulletin No. 88-11: Pressurizer Surge Line Thermal Stratification, USNRC, December 20, 1988.

7. Virginia Power Letters Serial Nos. 89-006A dated May 3, 1989 and 89-006B dated November 13, 1989 to United States Nuclear Regulatory Commission.

8. Revised report on the Reanalysis of Safety-Related Piping Systems - Surry Power Station, Unit 1, August 1979, by Stone & Webster Engineering Corporation.

9. Report on the I.E.Bulletin 79-14, Analysis for As-Built Safety-Related Piping Systems -

Surry Power Station - Unit 2, July 1981, by Ebasco Services, Inc.

10. Report on the Reanalysis of Safety-Related Piping Systems - Surry Power Station - Unit 2, Rev. 1, April 1980, by Ebasco Services, Inc.

11. Mitsubishi Heavy Industries, LTD., Design Report DG KCS-03-0008, Dominion Generation, Surry Power Station Unit 2, Control Rod Drive Mechanism Design Report, Rev. 3.

12. NEI 17-02, Rev. 1B, Tornado Missile Risk Evaluator (TMRE), September 2018, as implemented and approved at Shearon Harris Nuclear Power Plant (ML18347A385).

13. SWECO 7703, Missile-Barrier Interaction - A Topical Report, Stone & Webster Engineering Corporation, Boston, MA, September 1977

Revision 52Updated Online 09/30/20 Table 15.2-1 STRUCTURES, SYSTEMS, AND COMPONENTS DESIGNED FOR SEISMIC AND TORNADO CRITERIA (Refer to the equipment classification list (Q-list) for a more comprehensive list of components. See Note 1.)

Earthquake Tornado Item Criterion Criterion Sponsora Note Structures Reactor containment SW Reinforced-concrete substructure I P Reinforced-concrete superstructure I T Reinforced concrete interior shields and walls I NA Steel plate liner I P P for containment integrity Piping, duct, and electrical penetrations and P for shield wall and critical system shield wall I P penetrations only Personnel access hatch I P P for containment integrity Equipment access hatch I P P for containment integrity Equipment hatch platform I T SW T for tornado winds only Cable vault and cable tunnel I T SW SPS UFSAR Pipe tunnel to containment from auxiliary building I T SW Main steam valve house I T SW Auxiliary steam generator feed pump cubicle I T SW Cubicle for main steam and feedwater isolation valves I T SW Containment spray pump house, below grade I T SW EL. +27.5 and below Recirculation spray and low-head safety injection pump cubicle and pipe tunnel I P SW Safeguards ventilation room I NA SW 15.2-9

a. HISTORICAL information, see Note 2.

Table 15.2-1 (CONTINUED)

Revision 52Updated Online 09/30/20 STRUCTURES, SYSTEMS, AND COMPONENTS DESIGNED FOR SEISMIC AND TORNADO CRITERIA (Refer to the equipment classification list (Q-list) for a more comprehensive list of components. See Note 1.)

Earthquake Tornado Item Criterion Criterion Sponsora Note Structures (continued)

Auxiliary building SW Reinforced-concrete structure I T Steel superstructure I NA Vacuum equipment area I NA Fuel building SW Reinforced-concrete structure I T T for horizontal missile only Steel superstructure I T T for tornado winds only Spent-fuel storage rack I P P for horizontal missile only Fuel-handling trolley support structure I T Over spent-fuel only Service building SW Control room I T SW SPS UFSAR Emergency switchgear and relay room I T SW Battery rooms I T SW Mechanical equipment room-3 I T SW (Air-conditioning equipment room) For control room and relay room only Reactor trip breaker cubicle I T SW Emergency diesel-generator rooms SW Reinforced-concrete floor I T Walls, excluding louvers I T Roof slab I T

a. HISTORICAL information, see Note 2. 15.2-10

Table 15.2-1 (CONTINUED)

Revision 52Updated Online 09/30/20 STRUCTURES, SYSTEMS, AND COMPONENTS DESIGNED FOR SEISMIC AND TORNADO CRITERIA (Refer to the equipment classification list (Q-list) for a more comprehensive list of components. See Note 1.)

Earthquake Tornado Item Criterion Criterion Sponsora Note Structures (continued)

Turbine building NA NA SW By design, building collapse will not damage any Class I structures and components during earthquake, or tornado-resistant structures and components during tornado Mechanical Equipment Room-4 NA NA SW See note for Turbine Building Mechanical Equipment Room-5 I T Low-level intake structure I T SW T for emergency service water cubicle pump only (Circulating water pump intake structure)

High-level intake structures I T SW T, no missile protection required Seal pits I T SW T, no missile protection required SPS UFSAR High-level intake canal I NA SW Fire-pump house I T SW Engine-driven pump only Fuel-oil transfer pump vault I T SW Boron recovery tank dikes I T SW Systems Reactor coolant system Steam generators I P W Steam generator supports I P SW Reactor coolant pumps b I P W

a. HISTORICAL information, see Note 2. 15.2-11

Table 15.2-1 (CONTINUED)

Revision 52Updated Online 09/30/20 STRUCTURES, SYSTEMS, AND COMPONENTS DESIGNED FOR SEISMIC AND TORNADO CRITERIA (Refer to the equipment classification list (Q-list) for a more comprehensive list of components. See Note 1.)

Earthquake Tornado Item Criterion Criterion Sponsora Note

b. All references to pumps include drivers.

Systems (continued)

Reactor coolant system (continued)

Reactor coolant pump supports I P SW Pressurizer and pressurizer heaters I P W Pressurizer support I P SW Pressurizer relief tank I P W Reactor vessel Reactor core support structure I P W Reactor control rod guide structure I P W Fuel assemblies I P W Control rod and drive shaft assemblies I P W SPS UFSAR Incore instrumentation thimbles I P W Reactor vessel supports and neutron shield tank I P SW Control rod drive mechanisms I P W Reactor coolant piping, valves, and supports b I P W Reactor coolant bypass piping, valves, and I P W supports Pressurizer surge line I P W, SW c Pressurizer spray lines, valves, and supports I P SW 15.2-12

a. HISTORICAL information, see Note 2.

b. All references to piping and valves include root valves connecting to non-Class I systems, and valve operators.

Table 15.2-1 (CONTINUED)

Revision 52Updated Online 09/30/20 STRUCTURES, SYSTEMS, AND COMPONENTS DESIGNED FOR SEISMIC AND TORNADO CRITERIA (Refer to the equipment classification list (Q-list) for a more comprehensive list of components. See Note 1.)

Earthquake Tornado Item Criterion Criterion Sponsora Note

c. Pressurizer surge line was reanalyzed per NRC Bulletin 88-11, dated December 20, 1988.

Systems (continued)

Reactor coolant system (continued)

Pressurizer safety and relief valves I P W Safety injection system Accumulators and supports I NA W Low-head safety injection pumps and piping I P W P for containment integrity Boric acid injection tanks and piping I P W Piping, valves, and supports I NA SW Except drain/sample lines Containment spray system Refueling water storage tank I NA SW Containment spray pumps I NA SW SPS UFSAR Piping, valves, and supports I NA SW Except recirculation lines Refueling water chemical addition tank I NA SW Recirculation spray systems Recirculation spray pumps and piping I P SW P for containment integrity Recirculation spray heat exchangers I NA SW Reactor containment sump and screens I NA SW Piping, valves, and supports I NA SW 15.2-13

Table 15.2-1 (CONTINUED)

Revision 52Updated Online 09/30/20 STRUCTURES, SYSTEMS, AND COMPONENTS DESIGNED FOR SEISMIC AND TORNADO CRITERIA (Refer to the equipment classification list (Q-list) for a more comprehensive list of components. See Note 1.)

Earthquake Tornado Item Criterion Criterion Sponsora Note Systems (continued)

Containment vacuum system Process vent I NA SW Vacuum pump piping, valves, and supports I NA SW Chemical and volume control system Boric acid storage tanks I NA W Boric acid transfer pumps I P W Boric acid blender I P W Charging/safety injection pumps I P W Regenerative heat exchanger I P W Nonregenerative heat exchanger I P W Mixed-bed demineralizers I P W SPS UFSAR Reactor coolant filter I P W Volume control tank I P W Seal-water heat exchanger I P W Seal-water filter I P W Excess letdown heat exchanger I P W Piping, valves, and supports Boric acid piping I P SW Feed and bleed piping I P* SW 15.2-14

a. HISTORICAL information, see Note 2.

Table 15.2-1 (CONTINUED)

Revision 52Updated Online 09/30/20 STRUCTURES, SYSTEMS, AND COMPONENTS DESIGNED FOR SEISMIC AND TORNADO CRITERIA (Refer to the equipment classification list (Q-list) for a more comprehensive list of components. See Note 1.)

Earthquake Tornado Item Criterion Criterion Sponsora Note Systems (continued)

Chemical and volume control system (continued)

Piping, valves, and supports (continued)

Hydrogen, nitrogen, and vent piping for I P SW volume control tank Residual heat removal system Residual heat removal pumps I P W Residual heat exchangers I P W Piping, valves, and supports I P SW Boron recovery system Gas stripper I P SW Gas stripper overhead condenser I P SW SPS UFSAR Primary drain tank I P SW Primary drain tank vent chiller condenser I P SW Overhead gas compressors I P SW Gas stripper surge tank I P SW Component cooling system Component cooling pumps I P SW Component cooling water heat exchangers I P SW Component cooling surge tank I P SW 15.2-15

a. HISTORICAL information, see Note 2.

Table 15.2-1 (CONTINUED)

Revision 52Updated Online 09/30/20 STRUCTURES, SYSTEMS, AND COMPONENTS DESIGNED FOR SEISMIC AND TORNADO CRITERIA (Refer to the equipment classification list (Q-list) for a more comprehensive list of components. See Note 1.)

Earthquake Tornado Item Criterion Criterion Sponsora Note Systems (continued)

Component cooling system (continued)

Piping, valves, and supports For residual heat exchangers I P* SW For fuel pool coolers I P* SW P for horizontal missile only RCP thermal barrier supply piping from the I P SW upstream check valves to the RCPs Charging pump component cooling water system Charging pump component cooling water I P SW pumps Charging pump mechanical seal coolers I P SW SPS UFSAR Charging pump seal cooling surge tank I P SW Charging pump intermediate seal coolers I P SW Piping, valves, and supports I P SW Fuel pool cooling system Fuel pool pumps I P SW P for horizontal missile only Fuel pool coolers I P* SW P for horizontal missile only Piping, valves, and supports connecting above I P SW P for horizontal missile only equipment to spent-fuel pool

a. HISTORICAL information, see Note 2.

15.2-16

Table 15.2-1 (CONTINUED)

Revision 52Updated Online 09/30/20 STRUCTURES, SYSTEMS, AND COMPONENTS DESIGNED FOR SEISMIC AND TORNADO CRITERIA (Refer to the equipment classification list (Q-list) for a more comprehensive list of components. See Note 1.)

Earthquake Tornado Item Criterion Criterion Sponsora Note Systems (continued)

Compressed air system Critical system interface components - local instrument air accumulators/bottled sources and associated piping, valves, and supports Interfacing systems:

Reactor coolant system - Pressurizer PORVs I P Component cooling system - Containment I P isolation trip valves from the RHR HXs Main steam/feedwater system - Steam I P supply admission valves to the turbine driven auxiliary feedwater pump Ventilation vent system - Dampers - Fuel I N/A SPS UFSAR building filtered exhaust, containment air compressor cubicle exhaust, safeguards area normal exhaust, charging pump normal and filtered ventilation, and containment exhaust isolation Service water system Engine-driven emergency service water pumps I P* SW Charging pump service water pumps I P SW Charging pump lubricating oil coolers I P SW 15.2-17

a. HISTORICAL information, see Note 2.

Table 15.2-1 (CONTINUED)

Revision 52Updated Online 09/30/20 STRUCTURES, SYSTEMS, AND COMPONENTS DESIGNED FOR SEISMIC AND TORNADO CRITERIA (Refer to the equipment classification list (Q-list) for a more comprehensive list of components. See Note 1.)

Earthquake Tornado Item Criterion Criterion Sponsora Note Systems (continued)

Service water system (continued)

Service water piping, valves, and pipe supports for:

Recirculation spray heat exchangers I NA SW Component cooling heat exchangers I P* SW Engine-driven emergency service water I P* SW pump Emergency diesel-generator cooling I P SW Control room air-conditioning equipment I P* SW condensers Charging pump lube-oil coolers I P* SW SPS UFSAR Diesel-oil tank for emergency service water I P* SW pump Fire protection system Engine-driven fire pump I P SW Diesel-oil tank (300 gallons) I P SW Yard hydrant piping system I NA SW

a. HISTORICAL information, see Note 2.

15.2-18

Table 15.2-1 (CONTINUED)

Revision 52Updated Online 09/30/20 STRUCTURES, SYSTEMS, AND COMPONENTS DESIGNED FOR SEISMIC AND TORNADO CRITERIA (Refer to the equipment classification list (Q-list) for a more comprehensive list of components. See Note 1.)

Earthquake Tornado Item Criterion Criterion Sponsora Note Systems (continued)

Fuel handling system Manipulator crane in containment I P W Crane will be parked and secured so no damage to reactor control rod drive mechanisms can occur during earthquake Movable platform with hoist in fuel building I NA SW Platform will be parked and secured so no damage to fuel can occur during earthquake or tornado Fuel-handling trolley in fuel building I NA SW Trolley will be parked and secured during tornado warning periods so no damage to spent fuel can occur Fuel transfer tube with blind flange I P SW P for containment isolation Fuel elevator in fuel building I NA SW SPS UFSAR Ventilation system Ventilation equipment for safeguards ventilation I NA SW room Ductwork from safeguards ventilation room to I NA SW ventilation vent no. 2 Ductwork for containment purge system I P SW P for containment isolation penetrating containment between and including isolation butterfly valves

a. HISTORICAL information, see Note 2.

15.2-19

Table 15.2-1 (CONTINUED)

Revision 52Updated Online 09/30/20 STRUCTURES, SYSTEMS, AND COMPONENTS DESIGNED FOR SEISMIC AND TORNADO CRITERIA (Refer to the equipment classification list (Q-list) for a more comprehensive list of components. See Note 1.)

Earthquake Tornado Item Criterion Criterion Sponsora Note Systems (continued)

Ventilation system (continued)

Air-conditioning equipment for main control I P SW room and relay room Emergency main control and relay room I P SW ventilation Ventilation vent no. 2 I NA SW Control rod drive ventilation fans NA P SW Main steam system Steam piping from main steam lines to auxiliary I P SW steam generator feed pump turbine Main steam piping from steam generators to and I P SW including main steam trip valve SPS UFSAR Main steam piping, valves, and supports from NA NA SW Check was made that design-basis trip valves to and including turbine stop valves earthquake would not cause failure to function Turbine steam bypass piping, valves, and NA NA SW supports to condenser

a. HISTORICAL information, see Note 2.

15.2-20

Table 15.2-1 (CONTINUED)

Revision 52Updated Online 09/30/20 STRUCTURES, SYSTEMS, AND COMPONENTS DESIGNED FOR SEISMIC AND TORNADO CRITERIA (Refer to the equipment classification list (Q-list) for a more comprehensive list of components. See Note 1.)

Earthquake Tornado Item Criterion Criterion Sponsora Note Systems (continued)

Circulating water system Condenser NA NA SW Check was made that condenser water boxes will not fail during earthquake Circulating water piping, valves, and supports I P* SW P, underground from high-level intake canal to circulating water discharge tunnel to and including condenser intake butterfly valve; condenser discharge butterfly valve Circulating water discharge tunnel I P SW Circulating water pump vacuum breaker NA NA SW No credible failure mode for passive vacuum breaker Condensate and feedwater system SPS UFSAR Emergency condensate storage tank I P SW Emergency condensate makeup tank I P SW Auxiliary steam generator feed pumps I P* SW

a. HISTORICAL information, see Note 2.

15.2-21

Table 15.2-1 (CONTINUED)

Revision 52Updated Online 09/30/20 STRUCTURES, SYSTEMS, AND COMPONENTS DESIGNED FOR SEISMIC AND TORNADO CRITERIA (Refer to the equipment classification list (Q-list) for a more comprehensive list of components. See Note 1.)

Earthquake Tornado Item Criterion Criterion Sponsora Note Systems (continued)

Condensate and feedwater system (continued)

Piping, valves, and supports Emergency condensate storage tank to I P SW auxiliary steam generator feed pump From auxiliary steam generator feed pumps I P SW to steam generator feed lines Steam generator feed lines inside containment I P SW to and including first isolation check valve outside containment Primary vent and drain system Primary drain cooler I P W Piping, valves, and supports I P SW SPS UFSAR Primary drain transfer tank I P SW Secondary vent and drain system Steam generator blowdown piping, valves, and I P SW supports inside containment to and including first isolation trip valves

a. HISTORICAL information, see Note 2.

15.2-22

Table 15.2-1 (CONTINUED)

Revision 52Updated Online 09/30/20 STRUCTURES, SYSTEMS, AND COMPONENTS DESIGNED FOR SEISMIC AND TORNADO CRITERIA (Refer to the equipment classification list (Q-list) for a more comprehensive list of components. See Note 1.)

Earthquake Tornado Item Criterion Criterion Sponsora Note Systems (continued)

Gaseous waste disposal system Waste gas decay tanks I P SW Waste gas recombiner system I NA SW Waste gas compressors I NA SW Waste gas filter I NA SW Process vent blowers I NA SW Waste gas piping, valves, and supports from I NA SW stripper to process vent Process radiation monitoring system Process vent particulate monitor I NA SW Process vent gas monitor I NA SW SPS UFSAR Spray recirculation heat exchanger service I NA SW water monitors Area radiation monitoring system Main control room monitor I P SW

a. HISTORICAL information, see Note 2.

15.2-23

Table 15.2-1 (CONTINUED)

Revision 52Updated Online 09/30/20 STRUCTURES, SYSTEMS, AND COMPONENTS DESIGNED FOR SEISMIC AND TORNADO CRITERIA (Refer to the equipment classification list (Q-list) for a more comprehensive list of components. See Note 1.)

Earthquake Tornado Item Criterion Criterion Sponsora Note Systems (continued)

Instrumentation and control All instrumentation and control to operate and monitor operation of critical system component shown above during MCA or controlled shutdown Systems include:

Reactor protection (in part) I P W Safeguards initiation I N/A W/SW Containment isolation I P W/SW Reactor control (in part) I P W Includes trip breakers Steam generator water level control system I P W SPS UFSAR Reactor makeup control I P W Nuclear instrumentation (in part) I P W Non-nuclear process instrumentation I P W/SW Circulating water intake canal low level I P N/A isolation system Electrical systems Emergency diesel-generators I P SW Fuel-oil day tanks I P SW Fuel-oil transfer pumps I P SW 15.2-24

a. HISTORICAL information, see Note 2.

Table 15.2-1 (CONTINUED)

Revision 52Updated Online 09/30/20 STRUCTURES, SYSTEMS, AND COMPONENTS DESIGNED FOR SEISMIC AND TORNADO CRITERIA (Refer to the equipment classification list (Q-list) for a more comprehensive list of components. See Note 1.)

Earthquake Tornado Item Criterion Criterion Sponsora Note Systems (continued)

Electrical systems (continued)

Underground fuel-oil storage tanks I P SW Fuel-oil piping, valves, and supports, I P SW P for piping to protected generators emergency diesel-generators Station service batteries and chargers I P SW Vital bus and inverters I P SW Emergency station service transformers I P SW Emergency station service switchgear I P SW Control panel boards I P SW Pressurizer heater control group only I P SW Cable to critical components, instruments, and I P SW Cable passing through unprotected areas SPS UFSAR controls as shown above will be in rigid conduit Miscellaneous Reactor containment crane I P SW Piping I P SW

a. HISTORICAL information, see Note 2.

15.2-25

Table 15.2-1 (CONTINUED)

Revision 52Updated Online 09/30/20 STRUCTURES, SYSTEMS, AND COMPONENTS DESIGNED FOR SEISMIC AND TORNADO CRITERIA (Refer to the equipment classification list (Q-list) for a more comprehensive list of components. See Note 1.)

Earthquake Tornado Item Criterion Criterion Sponsora Note Legend W - Westinghouse Electric Corporation.

SW - Stone & Webster Engineering Corporation.

I - Refers to Seismic Class I criteria. All Class I components and structures are designed to resist the operating-basis earthquake within allowable working stresses. A check has been made to determine that failure to function will not occur with a design-basis earthquake.

T - Refers to structures that will not fail during the design tornado.

P - Refers to systems and components that will not fail during the design tornado since they are protected by tornado resistant structures or by being buried underground.

P* - Refers to systems and components that are not provided with complete physical protection from tornado-generated missiles, but have been evaluated using the Tornado Missile Risk Evaluator Methodology (Reference 12) and it has determined that the risk to the plant associated with the partially exposed SSC is sufficiently low such that complete physical protection from tornado-generated missiles need not be provided.

NA - Not applicable.

SPS UFSAR NOTES:

15.2-26

Table 15.2-1 (CONTINUED)

Revision 52Updated Online 09/30/20 STRUCTURES, SYSTEMS, AND COMPONENTS DESIGNED FOR SEISMIC AND TORNADO CRITERIA (Refer to the equipment classification list (Q-list) for a more comprehensive list of components. See Note 1.)

Earthquake Tornado Item Criterion Criterion Sponsora Note

1. CAUTION, this table shall only be used for the classification of structures. Refer to the PAMS database for the classification of systems and components. A list of structures, systems, and components, like those in Table 15.2-1, was provided as part of the licensing application to permit a determination to be made as to the general suitability of the classification given and the design approach applied. Since the time of original plant licensing, an equipment classification listing (Q- List), was developed and subsequently replaced with a database (PAMS) to provide a more comprehensive and up-to-date list of individual components and their classifications than does this table, which only provides a general list of systems and components. According to the SPS current licensing basis, structures required to withstand the effects of a design basis tornado (Tornado Criterion "T") are also required to be designed to Seismic Category I requirements (Seismic Criterion "I"). Hence, all structures classified as "T" must also be classified as "I", but not necessarily vice versa. The Q-List and PAMS database only provide an input field for the more encompassing Seismic Category I classification for structures and do not provide a separate input field to identify those Seismic Criterion "I" structures that must also meet the Tornado Criterion "T" classification. Hence, SPS UFSAR, Table 15.2-1, was updated to be consistent with the SPS current licensing basis to reflect both the Seismic Criterion "I" and Tornado Criterion "T" classifications for structures at SPS in response to US NRC RIS 2015-06. For the classifications of systems and components at SPS, designed to be functional under Seismic Class I, Seismic Criterion "I", refer to the PAMS database.

SPS UFSAR

2. The information in the sponsor column designates the division of responsibility between Westinghouse and Stone & Webster for the original design of listed structures, systems, and components. These designations are considered HISTORICAL and are not intended or expected to be updated for the life of the plant.

15.2-27

Revision 52Updated Online 09/30/20 SPS UFSAR 15.2-28 Table 15.2-2 DAMPING FACTORS FOR CLASS I STRUCTURES Component Percent of Critical Damping

1. Reactor vessel internals and control rod assembly drives

a. Welded assemblies 1.0

b. Bolted assemblies 2.0

c. Control rod assembly drives 5.0a

2. Reinforced concrete reactor support structure, including the reactor 5.0 vessel

3. Vital piping systems

a. Carbon steel 0.5 OBE, 1.0 DBEb

b. Stainless steel 0.5 OBE, 1.0 DBEb

4. Containment structure and foundation 5.0

5. Steel framed structures, including supporting structures and foundations

a. Bolted 2.5

b. Welded 1.0

6. Concrete structures aboveground

a. Shear-wall type 5.0

b. Rigid-frame type 5.0

7. Mechanical equipment, including pumps, fans, and similar items 2.0

a. For Surry Unit 2 control rod assembly drives, the Percent of Critical Damping used is 5% as justi-fied in Mitsubishi Heavy Industries Design Report, Reference 11.

b. In accordance with reference 4, the damping values of ASME Code Case N-411 (reference 5) have been approved for use at Surry as an alternate, for both the operating-basis earthquake and the design-basis earthquake.

The values specifically are: five percent below frequency of 10Hz; linear reduction from five percent to two percent between 10Hz and 20Hz and two percent above 20Hz. These damping values are used in the following situations and with the following additional considerations:

a. For seismic analyses in cases where new piping is added, existing systems are modified, existing systems are re-evaluated for support optimization.

b. For seismic analyses using response spectrum methods and not for seismic analyses using time-history analyses methods.

c. When these damping values are used, the +/-15% peak broadening criteria of Regulatory Guide 1.122, Development of Floor Design Response Spectra for Seismic Design of Floor Supported Equipment or Components, will be used.

d. When these alternate damping values are used, they are used in a given analyses in their entirety.

e. When these damping values are used together with changes in the support arrangement that increases the flexibility of piping systems, the predicted maximum displacements are reviewed to ensure that such displacements do not cause adverse interaction with adjacent structures, components or equipments.

Revision 52Updated Online 09/30/20 SPS UFSAR 15.3-1 15.3 MATERIAL 15.3.1 Concrete See Section 15.5.2.4 for the description of the concrete used for the Reactor Pressure Vessel Head Replacement Project.

15.3.1.1 Cement All cement used was an approved American brand conforming to the specification for Portland cement, ASTM Designation C150, Type II, low alkali. It is suitable for Class I structures because of its lower heat of hydration and improved resistance to sulphate attack. A low-content alkali was specified to minimize the possibility of reaction with aggregates. Certified copies of mill tests, showing that the cement meets or exceeds the ASTM requirements for Portland cement, were furnished by the manufacturer. An independent testing laboratory performed tests on the cement for compliance with the specifications.

15.3.1.2 Admixtures An air-entraining agent was used in the concrete in an amount sufficient to entrain from 3 to 5% air by volume of the concrete. This agent conformed to the requirements of Standard Specification for Air-Entraining Admixtures for Concrete, ASTM C260, when tested in accordance with Standard Method of Testing Air-Entraining Admixtures for Concrete, ASTM C233.

The air-entraining agent was added separately to the batch in solution in a portion of the mixing water. The solution was batched by means of a mechanical dispenser capable of accurate measurement, and in a manner that ensured uniform distribution of the agent throughout the batch during the specified mixing period.

Water-reducing agents were used when their use was approved in writing. Water-reducing agents were Master Builders NB-100, type R or N, manufactured by Master Builders of Cleveland, Ohio. Type N is normal NB-100 and is used when a normal rate of hardening is required. Type R contains a retarder and is used in warm weather to reduce the rate of hardening and to avoid cold joints.

Calcium chloride was not used under any circumstances.

15.3.1.3 Water Mixing water was obtained from a deep well and was kept clean and free from injurious amounts of oils, acids, alkalies, salts, organic materials, or other substances deleterious to concrete or steel. The quality of the water was the equivalent of that suitable for drinking. The water was continuously checked and tested for compliance with the above requirements by an independent testing laboratory.

Revision 52Updated Online 09/30/20 SPS UFSAR 15.3-2 15.3.1.4 Aggregates Fine and coarse aggregates conformed to the requirements of the Standard Specifications for Concrete Aggregates ASTM C33. Aggregates were evaluated for potential chemical alkali reactivity. Aggregates were free from any materials that could have been deleteriously reactive in any amount sufficient to have caused excessive expansion of mortar or concrete. All aggregates were tested for compliance with the above requirements by an independent testing laboratory.

15.3.1.5 Proportioning Proportioning of structural concrete conformed to ACI 301, Chapter 3. Working-stress-type concrete and ultimate-strength-type concrete conformed to the requirements of ACI 301, Paragraph 302. Ultimate-strength-type concrete was used in the construction of the foundation mat, exterior wall, and dome of the reactor containment. In general, working-stress-type concrete was used for other areas. Concrete mixes had a 28-day specified strength of 3000 psi, except as otherwise noted on the engineering drawings.

Proportions of ingredients were determined and tests conducted by an independent laboratory in accordance with the method detailed in ACI 301, Paragraph 308, for combinations of materials established by trial mixes.

The maximum slump of mass concrete, as defined in ACI 301, Chapter 14, in general did not exceed 3 inches. Slump of other concrete conformed to ACI 301, Paragraph 305. The samples for the slump tests were taken at the end of the last conveyor, chute, or pipeline before the concrete was placed in the forms.

The close and complex spacing of reinforcing steel in the heavily reinforced sections surrounding the equipment and personnel hatches results in the use of concrete with a maximum slump of 5 inches. The results of strength tests indicate that the 5-inch slump concrete will have a minimum compressive strength of approximately 4000 psi at 28 days. This is considerably higher than the nominal stipulated value of 3000 psi at 28 days used for design purposes, and demonstrates that the structural strength of the containment would not be jeopardized by the use of concrete with a slump of 5 inches.

15.3.2 Reinforcing Steel Except for the No. 14 and No. 18 reinforcing bars for the foundation mat, exterior wall, and dome of the containment structure, all reinforcing conforms to Grade 40 (or higher strength steel) of the Standard Specification for Deformed Billet-Steel Bars for Concrete Reinforcement ASTM A615.

For No. 14 and No. 18 reinforcing bars and splices for the foundation mat, exterior wall, and dome of the containment structure, see Section 15.5.1.9. See Section 15.5.2.3 for the description of the reinforcing steel used for the Reactor Pressure Vessel Head Replacement Project.

Mill Test Reports showing chemical and physical properties were obtained and evaluated for each heat of steel used in making all reinforcing steel furnished.

Revision 52Updated Online 09/30/20 SPS UFSAR 15.4-1 15.4 CONSTRUCTION PROCEDURES AND PRACTICES See Section 15.5.2 for the description of the restoration of the construction opening used for the Reactor Vessel Head Replacement Project.

15.4.1 Codes of Practice Materials and workmanship conformed to the following codes and specifications:

ACI 301-66 Structural Concrete for Buildings and all specifications of the American Society for Testing and Materials referred to in Section 105 and declared to be a part of ACI 301-66 as is fully set forth therein.

ACI 304 Recommended Practice for Measuring, Mixing, and Placing Concrete.

ACI 305 Recommended Practice for Hot Weather Concreting.

ACI 306 Recommended Practice for Cold Weather Concreting.

ACI 318-63 Building Code Requirements for Reinforced Concrete.

ACI 347 Recommended Practice for Concrete Formwork.

See Section 15.5.2.1 for the description of the codes and specifications used for the restoration of the construction opening used for the Reactor Pressure Vessel Head Project.

Section III of the ASME Boiler and Pressure Vessel Code for Nuclear Vessels was used as a guide in the selection of materials, design stresses, and fabrication of the steel containment liner.

ACI 301-66, Specifications for Structural Concrete for Buildings, together with ACI 347-63, Recommended Practice for Concrete Formwork, and ACI 318-63, Building Code Requirements for Reinforced Concrete, formed the basis for the concrete specifications.

ACI 301-66 was supplemented as necessary with mandatory requirements relating to types and strengths of concrete, including minimum concrete densities, proportioning of ingredients, reinforcing steel requirements, joint treatments, and testing agency requirements.

Admixtures, types of cement, bonding of joints, embedded items, concrete curing, additional test specimens, additional testing services, cement and reinforcing steel mill test report requirements, and additional concrete test requirements were specified in detail.

Concrete protection for reinforcement, preparation, and cleaning of construction joints, concrete mixing, delivering, placing, and curing, with the following exceptions, equaled or exceeded the requirements of ACI 301:

Section 1404 (a) - Maximum slump was generally restricted to 3 inches to permit placing concrete in the heavily reinforced containment structures. The slump was increased to 5 inches in the areas of the containment wall adjacent to the equipment and personnel hatches where the large

Revision 52Updated Online 09/30/20 SPS UFSAR 15.4-2 steel inserts and additional reinforcing steel required a more plastic mix for adequate concrete placement. All concrete mixes were designed and tested before use. All concrete mixes used in the work were fully documented.

Section 1404 (b) - Maximum placing temperature of the concrete when deposited conformed to the requirements of ACI 305-59, Recommended Practice for Hot Weather Concreting.

Section 1404 (c) - Minimum placing temperature of the concrete when deposited conformed to the requirements of ACI 306-66, Recommended Practice for Cold Weather Concreting.

15.4.2 Concrete Concrete ingredients were batched in a batch plant and transferred to transit mix trucks for mixing, agitating, and delivering to the point of placement. Water was added to the mix with the other ingredients before the truck left the batch plant area. Batching and mixing otherwise conformed to ACI 301, Chapter 7.

Placing of concrete was by bottom-dump buckets, concrete pumps, or by conveyor belt.

Bottom-dump buckets did not exceed 4 yd3 in size. The discharge of concrete was controlled so that concrete could be effectively compacted around embedded items and near the forms.

For placing of concrete for the wall and dome of the containment structure, see Section 15.5.1.10. See Section 15.5.2.4 for the description of the concrete used for the Reactor Pressure Vessel Head Replacement Project.

Vertical drops greater than 6 feet for any concrete were not permitted, except where suitable equipment was provided to prevent segregation. All concrete placing equipment and methods were subjected to the approval of the structural engineer.

The surfaces of all construction joints were thoroughly treated to remove all laitance and to expose clean, sound aggregate. Surfaces of fresh concrete were roughened by cutting with an air-water jet after the initial concrete set had occurred, but before the concrete had reached its final set. After cutting, the surface was washed and rinsed. Where the use of an air-water jet was not advisable in any specific instance, then that surface was roughened by hacking with hand tools or other satisfactory means to produce the requisite clean surface.

Before placing subsequent concrete lifts, the surfaces of all construction joints were thoroughly cleaned and wetted, and all excess water that was not absorbed by the concrete was removed. Horizontal construction joints were then covered by a 0.50-inch-thick layer of sand/cement grout of the same sand/cement ratio as the concrete, and new concrete was then placed immediately against the fresh grout.

Revision 52Updated Online 09/30/20 SPS UFSAR 15.4-3 Curing and protection of freshly deposited concrete conformed to ACI 301, Chapter 12, using curing compounds conforming to ASTM C309.

For curing of the top surface of the containment foundation mat, see Section 15.5.1.10.

Concrete strength tests were performed in accordance with ACI 301, Chapter 16, Section 1602 (a), Paragraph 4, supplemented as follows.

No fewer than two sets of compression test specimens for each mix design of concrete placed were taken during the first two days of placing concrete, or at least one set of test specimens for each 250 yd3 placed. Thereafter, one set of test specimens was taken for each 250 yd3, or fraction thereof, for each mix design of concrete placed in any one day. In addition, one set of specimens was taken whenever, for any reason, the materials, methods of concreting, or proportioning were changed.

The test specimens for compressive strength were cylinders 6 inches in diameter and 12 inches long. Each set consisted of five specimens, at least one of which was tested at 7 days and three at 28 days age. The remaining cylinder was retained at the laboratory for further tests at 60 days age if the result of the previous tests made such a test desirable.

Concrete strength tests were evaluated by the engineers in accordance with ACI 214-65, Recommended Practice for Evaluation of Compression Test Results of Field Concrete, and ACI 301-66, Chapter 17.

Strengths of working-stress-type concrete were considered satisfactory if the average of any five consecutive strength tests of the laboratory-cured specimens at 28-days age was equal to or greater than the specified compressive strength, f'c, of the concrete.

Strengths of ultimate-strength-type concrete were considered satisfactory if the average of any three consecutive strength tests of the laboratory cured specimens at 28-days age was equal to or greater than the specified compressive strength, f'c, of the concrete.

If any tests for individual cylinders or group of cylinders failed to reach the specified compressive strength, f'c, of the concrete, the responsible engineers were immediately notified to determine if further action would be required.

The field tests for slump of Portland cement concrete were in accordance with ASTM C143. Any batch not meeting specified requirements was rejected.

Slump tests were made frequently during concrete placement and each time concrete test specimens were made.

Statistical quality control of the concrete was maintained by a computer program. This program analyzed compression test results reported by the testing laboratory in accordance with methods recommended by ACI 214, Recommended Practice for Evaluation of Compression Test Results of Concrete.

Revision 52Updated Online 09/30/20 SPS UFSAR 15.4-4 15.4.3 Reinforcing Steel Placing of reinforcing steel conformed to the requirements of Chapter 5 of ACI 301, Structural Concrete for Buildings, and Chapter 8 of ACI 318, Building Code Requirements for Reinforced Concrete. See Section 15.5.2.3 for the description of the placement of the reinforcing steel used for the Reactor Pressure Vessel Head Replacement Project.

All Cadweld splices were made in accordance with the instructions issued by the manufacturer, Erico Products, Inc., Cleveland, Ohio.

In order to qualify operators for making Cadweld process joints, each operator was required to demonstrate to the Senior Quality Control Engineer his ability to make an acceptable fixed joint using the Cadweld process. Cadwelders were requalified after every 200 Cadwelds. Testing was by tensile testing a Cadweld made under simulated field conditions.

The ends of the reinforcing steel bars to be joined by the Cadweld process were square cut by the fabricator. Ends of the bars were then thoroughly cleaned of all rust, scale, grease, oil, water, or other foreign matter before the joints were made.

Welding was performed using the Metallic Arc Welding Process with coated electrodes, or the Metallic Inert Gas Shielding Welding Process (MIG) using bare wire. The filler metal for the Metallic Arc Welding Process conformed to ASTM A316, Coated Arc Welding Electrodes, Classification E-10016-D2 or E-10018-D2.

The filler metal for the MIG welding process was a spooled bare wire 0.30 inch or 0.35 inch in diameter, Linde or Arcos Type 515. The shielding gas used for the MIG welding process was Linde C-25, a mixture of 75% argon and 25% carbon dioxide.

The ends of the bars to be joined by butt welding were prepared by sawing or flame cutting, and dressing by grinding, where necessary, to form a single vee butt joint.

Mill test reports of the heats of steel used for making the rebars were obtained by the Senior Quality Control Engineer to confirm the grade of steel welded. Where preheating was required, temperatures were checked with Tempilstiks.

In order to qualify welders for work on the reinforcing steel bars, each welder made a reinforcing bar test weld in the horizontal fixed position, welding vertically up. Each test weld was sectioned through the center of the weld by power sawing and machining. The cross-sectioned surface was etched with a 10% solution of nitric acid and water. The etched surface was examined by the field welding supervisor, who determined the qualification of the welding operator.

Tack welding of rebar was not permitted.

Special criteria for placing reinforcement for the containment structure are provided in Section 15.5.1.6.

Revision 52Updated Online 09/30/20 SPS UFSAR 15.4-5 15.4.4 Construction Procedures The portion of the site to be covered by structures was cleared, and general excavation performed to the underside of the foundations for the various buildings. In general, this excavation was from elevation +34 to +10, with some building foundations slightly higher or lower. The major Class I structures (except the Fuel Building and main steam valve enclosure structures) are supported on mat foundations; the Fuel Building and the main steam valve enclosure structures are supported on pile foundation. For additional construction procedures for other Class I structures, see Section 15.6.1.

15.4.5 Construction Practice Vepco maintained quality control personnel on the site at all times to serve as qualified inspectors in all phases of work, so as to ensure and document that all construction operations met the rigid requirements of the specifications as outlined in the quality assurance report. The qualification of welding procedures and welders was performed in accordance with Part A of Section IX of the ASME Boiler and Pressure Vessel Code or, for structural steel, in accordance with American Welding Society requirements.

Concrete was sampled and tested during construction, in accordance with ACI 318, to ensure compliance with the specifications. A competent independent testing laboratory was retained to design the concrete mixes, take samples, perform all tests of aggregates and concrete cylinders, and report to Vepco for approval.

Special practices to be followed for the containment liner are contained in Section 15.5.1.8.

15.4.6 Quality Assurance Program (Construction Phase)

The descriptions of the quality assurance program during the construction phase have been deleted. These activities have been completed and the descriptions are no longer needed for the operational phase. The NRC-approved Operational Quality Assurance Program is described in Chapter 17.

Revision 52Updated Online 09/30/20 SPS UFSAR 15.4-6 Intentionally Blank

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-1 15.5 SPECIFIC CONTAINMENT STRUCTURAL DESIGNS 15.5.1 Containment Structure 15.5.1.1 General For arrangement of the containment structure, see Reference Drawings 1 through 7.

Each of the reactor containment structures is similar in design and construction to that of the Connecticut Yankee Atomic Power Plant at Haddam, Connecticut. Each is a steel-lined, heavily reinforced concrete structure with vertical cylindrical wall and hemispherical dome supported on a flat base mat. Below grade, the containment structures are constructed inside a cofferdam of steel sheet piling. The structures are soil-supported. The base of the foundation mats is located approximately 66 feet below finished ground grade.

Each containment structure has an inside diameter of 126 ft. 0 in. The spring line of the dome is 122 ft. 1 in. above the top of the foundation mat. The inside radius of the dome is 63 ft.

0 in. The interior vertical height is 185 ft. 1 in., and the base mat is 10 ft. 0 in. thick. The steel liner for the wall is 3/8-inch thick, except over the base mat, where 0.25-inch and 0.75-inch plate is used. The steel liner for the dome is 0.50-inch thick. A waterproof membrane, as shown in Figure 15.5-1, is placed below the containment structural mat and carried up the containment wall to ground level. Attached to and entirely enveloping the part of the structure below grade, the membrane protects the structure from the effects of ground water and the steel liner from external hydrostatic pressure. Ground water immediately adjacent to the containment structure is kept below the top surface of the foundation mat by pumping as required.

Access to the containment structure is provided by a 7 ft. 0 in. i.d. personnel hatch penetration, and a 14 ft. 6 in. i.d. equipment hatch penetration. Other smaller containment structure penetrations include hot and cold pipes, main steam and feedwater pipes, fuel transfer tube, and electrical conductors.

The reinforced concrete structure has been designed to withstand all loadings and stresses anticipated during the operation and life of the unit. The steel lining is attached to and supported by the concrete. The liner functions primarily as a gastight membrane, and transmits incident loads to the concrete. The containment structure does not require the participation of the liner as a structural component. No credit has been taken for the presence of the steel liner in designing the containment structure to resist seismic force or other design loads.

The steel wall and dome liners are protected from potential interior missiles by interior concrete shield walls. CRDM missile protection is provided by a concrete shield on Unit 1 and a steel shield on Unit 2. The base mat liner is protected by a 1.50 to 2-foot thick concrete cover, except where a 0.75-inch-thick liner plate was used beneath the reactor vessel incore instrumentation, and at a drainage trench where floor grating provides additional protection.

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-2 As an added precaution against water seepage that might penetrate the waterproofing membrane in small quantities, pipe sumps are provided in each of the instrument observation pits located outside the cylindrical wall of the containment but within the waterproofing membrane.

The sumps penetrate the base mat and terminate in the porous concrete immediately below the mat.

Pumps are provided to remove ground water outside the waterproofing membrane, as described in Section 15.5.1.3.

15.5.1.2 Design Criteria The design of the containment structures is based on:

1. Biological shielding requirements.

2. The temperature and pressure generated by the design-basis accident (DBA), Section 14.5.2.

3. The operating and design-basis earthquakes discussed in Section 2.5

4. Severe weather phenomena.

5. The maximum calculated power level of 2597 MWt.

The design-basis accident was selected as the design basis for the containment structure because all other bases would result in lower temperatures and pressures. The containment structure is also designed for the normal subatmospheric operating conditions. Further, the containment structure is designated for a leakage rate not to exceed 0.1% of the contained volume per day at 45 psig.

The minimum operating pressure for the containment is 10.1 psia with about 1.0 psia additional partial water vapor pressure. The resulting total containment pressure is approximately 11.1 +/- 0.5 psia. The temperature of the containment air fluctuates between a maximum temperature of 125°F and a minimum of 75°F during normal operation, and 60°F during shutdown, depending upon the ambient temperature of available service water. The normal operating pressure allows accessibility for inspection and minor maintenance during operation without requiring containment pressurization or the use of supplementary breathing equipment for personnel.

The containment structure is designed by ultimate strength methods conforming to ACI 318-63, Part IV-B. Design load criteria based on ACI requirements and others given below conform to current containment design.

The ultimate load capacity of the containment structure as modified by the safety provisions of ACI 318-63, Section 1504, is not less than that required to meet the containment structural loading criteria.

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-3 Loads imposed on the containment shell design include:

1. Dead load.

2. DBA pressure.

3. Temperature rise in liner associated with DBA.

4. Normal operating temperature gradients.

5. Earthquake.

6. Wind loads, including tornado winds.

Loads imposed on the containment mat design include:

1. Mat and interior structures during construction.

2. Dead load for complete structure and contents.

3. Dead load and DBA pressure and liner loading.

4. Dead load, DBA pressure, liner loading, and earthquake.

5. Dead load and earthquake.

The ultimate load capacity of the containment structure, as modified by the safety provisions of ACI 318-63, Section 1504, is not less than that required to satisfy the following structural loading criteria, tabulated in Table 15.5-1.

The seismic design coefficients and critical damping factors used in the design of the reactor containment structure are given in Section 15.5.1.4. The average acceleration spectra curves are included in Section 2.5. The earthquake loads include the horizontal or vertical acceleration, or a combination of both where the effects, as measured by the stresses resulting from the separate acceleration components, of horizontal and vertical ground accelerations are combined algebraically.

The load capacity of the tension members is based on the guaranteed minimum yield strength of the reinforcing steel. Load capacities of flexural and compression members are determined in accordance with the Building Code Requirements for Reinforced Concrete, ACI 318. The load capacity so determined is decreased by a reduction factor multiplier , to compensate for small adverse variations in material and workmanship. The reduction factors are listed in Table 15.5-2.

The load capacity reduction factor for stresses in concrete produced by tornado-carried missiles, in combination with other tornado-produced stresses as given in Loading Criteria 5, is 0.75.

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-4 The dominant design load is the 45-psig containment design pressure, which creates major tensile membrane stresses in the reinforcing steel, coincident with moments at the junction of the containment wall and mat.

The design tornado wind loading and pressure drop criteria are stated in Sections 2.2 and 15.2.3.

Since the DBA pressure load is greater than the negative pressure load of tornadoes, the containment structure is able to maintain its integrity and permit an orderly shutdown on the reactor unit should a tornado strike the structure.

15.5.1.3 Buoyant Loads Yard elevation is at +26 ft. 6 in.; the base of the containment mat is at Elevation -39 ft. 7 in.

Six seepage drains are provided to drain the area beneath the containment structure. Four drains extend down to Elevation -65 ft. 0 in., and two drains extend down to Elevation -105 ft. 0 in.

These drains terminate in a 12-inch thick, crushed-rock layer placed immediately below the mat and through which water can travel to the edge of the cofferdam. Seepage from these drains, and other seepage into the cofferdam, collects inside the cofferdam around the base of the mat. Two pumps located in a cubicle adjacent to the instrument well remove all subsurface seepage water.

To prevent loss of pumping capability, the system design permits access to critical areas, such as the interior drainage header, the pump cubicle, and backwash facilities. This will permit maintenance and continued operation of the drainage system, thereby preventing water levels from reaching the top of the containment base mat and exerting hydrostatic pressure on the top of the mat liner.

The pumps are controlled to maintain the water level in this space between a high of Elevation -32.75 ft. and a low of Elevation -33.4 ft., a range of 0.7 feet which is equivalent to a fluctuation in buoyant pressure under the structure of +/-22 lb/ft2 from the mean value. A local high level indicator comes in if the water exceeds Elevation -32.6 ft. The dead load of the structure and its contents is 7200 lb/ft2. This fluctuation in buoyant pressure amounts to 0.31% of the dead load weight.

In the unlikely event of multiple pump failure for a sufficient period of time for the ground water to rise to finished ground grade at Elevation +26 ft. 6 in. the buoyant pressure would increase to a maximum value of 4150 lb/ft2, which amounts to less than 60% of the dead load structure. Therefore, flotation of the containment is not credible.

15.5.1.4 Dynamic Analysis Analyses were conducted to determine response stresses in the containment structure due to the application of seismic loading. Earthquake ground motion values were applied simultaneously in the horizontal and vertical directions. Vertical ground motions were assigned a magnitude equal to two-thirds of the horizontal motions. The magnitudes of the operating-basis earthquake and the design-basis earthquake are derived and assigned as described in Section 2.5. Design loading

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-5 conditions combined with seismic loading and allowable stress levels are stated in Section 15.5.1.2.

The earthquake loading was analyzed using a Stone & Webster program, Container Vessel Seismic Analysis, based upon the dynamic analysis of a containment structure by Messrs. Hansen, Holley, and Biggs of MIT.

The general analytical model of the containment structure responding to horizontal earthquake forces is a coupled two-mass system in which the wall and dome comprise one mass and the base slab and internals comprise the second mass. This model responds to three degrees of freedom: flexure in the wall and dome, translation, and rocking of the structure as a unit. The model includes the first three modes of vibration.

The stiffness of the wall and dome was obtained through formulas recommended by Professor R. V. Whitman of MIT, based on work by G. N. Bycroft.

The output of the computer program was spot-checked by manual analysis, which confirmed the program basis.

Another independent manual analysis that considered the internals as a third coupled mass resulted in loading values that were not greater than those obtained from the analysis of the two-mass system.

A preliminary analysis of response to vertical earthquake forces using a single-mass system showed that these forces are not controlling factors in the design.

When computing the response of the reinforced concrete containment structure to earthquake forces, the value of 5% of critical damping was used with the design earthquake acceleration of 0.07g. This is an overall value that includes the damping in both the reinforced concrete structure and the soil. The magnitudes of earthquake forces applied to the structure were obtained from the response spectrum for 0.07g at zero period and 5% critical damping at the calculated frequency of the structure, and then distributed over the structure in accordance with the relative motions of the structure as determined by dynamic analysis.

The force derived by use of this damping factor was used for the entire reinforced concrete containment. The value of 5% of critical damping, together with the damping factors for other systems, structures, and equipment, is listed in Table 15.2-2.

The value of 10% of critical damping was used with the design-basis earthquake of 0.15g on the basis of increased cracking in the concrete and increased movement in the concrete and soil.

To verify the damping used for design, an analysis of the soil structure interaction damping was made in accordance with the procedures suggested in Analysis of Foundation Vibrations, by

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-6 Robert V. Whitman, Proceedings of a Symposium organized by the British National Section of the International Association for Earthquake Emergency.

Damping factors for soil were calculated for the rigid body translation and rocking. Flexure damping was assessed as suggested by Newmark.

For each of the four modes of vibration, energy losses in structural flexure, sliding, and rocking were calculated and proportioned to determine the total system energy loss, thereby defining the damping to be used in spectrum response.

This analysis demonstrated that the damping factors used for design and the resulting seismic response characteristics are conservative.

Earthquake load criteria are included in the loading criteria described in Section 15.5.1.2.

Operating and design-basis earthquake factors are each combined with other loads, including the design-basis accident pressure. Resulting shears are computed by the computer program.

Lateral earth pressure under seismic loadings on the containment mat was determined by computing the lateral resistance developed in the soil as the structure responds in flexure, translation, and rocking. In this analysis, the translational restraining force has two components, a shear across the base of the structure and lateral soil pressures on the side wall of the containment structure developed by its displacement relative to its static position.

The spring constant, that is, force per unit of lateral displacement by shear, for a circular rigid base on an elastic half space is given by Bycroft (Reference 1) as:

32 ( 1 - u ) Gr k x = -----------------------------------o-7 - 8u where:

G = shear modulus ro = radius of base u = Poissons Ratio Note: Values in consistent units For usual values of u this reduces approximately to:

kx = 5Gro

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-7 The horizontal pressure on the side wall of the containment structure can be evaluated from the theories of horizontal subgrade reaction. From Terzaghi (Reference 2) the relation between horizontal deflection and pressure at any point is given by:

P-k h = ----

yh where:

P = horizontal pressure at soil structure interface y = horizontal deflection of soil at interface kh = coefficient of horizontal subgrade reaction further:

nh z k h = -------

B where:

nh = coefficient dependent upon physical properties of the soil z = depth below free surface of soil B = width of loaded area, which may be taken as diameter of containment structure For purposes of this analysis, a value of n = 40 tons/ft3 was selected from tables presented by Terzaghi. This value is appropriate to dense sand above the ground water table. It is a conservative value, since the higher the coefficient, the stiffer the soil, and the greater the loads imposed upon the side walls of the structure.

The rotational, translational, and flexural deflections of the structure were determined from response analysis and added so as to obtain maximum deflections. The lateral soil pressures on the side wall of the structure were then computed for these total deflections using the theory of horizontal subgrade reaction.

In determining these pressures, the side wall of the structure was assumed to be rigid radially, since radial deflection of the side wall would reduce relative soil-structure deflections, and thus the soil forces acting upon the structure.

The analysis was performed for both the operating-basis earthquake of 0.07g and the design-basis earthquake of 0.15g. The analysis for a 0.15-g earthquake indicates a lateral force of 300 lb/ft2 at Elevation -8 ft. 6 in. which defines approximately the magnitude of this component.

It should be noted that these forces, if included in the seismic loadings on the structure, would reduce the base shear and vertical bending stresses in the shell. Accordingly, they are not

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-8 included when computing such stresses in the shell and thereby contribute to the conservation of the design.

Rocking motion of the containment structure was considered in the determination of the natural frequency, the distribution of inertia forces, and in the amplitudes of motions.

The containment wales supporting the cofferdam structure do not affect consideration of horizontal pressure under seismic loading on the containment wall.

Four circular concrete wales originally supported the sheet steel cofferdam in which the containment structure is founded. The top wale, Wale A, has been partially removed at several points to permit completion of adjacent structures; in this condition it does not impose any restraint on the containment structure. The bottom wale, Wale D, is in the lower plane of the containment mat and below the plane of the wall, and offers no restraint. Wale C extends from a height of 4 ft. to 8 ft. above the base of the wall. Wale B extends from a height of 17 ft. 6 in. to 21 ft. 6 in. above the base of the wall. These two wales are approximately 3 ft. 9 in. from the containment wall, and the space between the wales and wall is backfilled with pervious fill. Under seismic loading, the distribution of the lateral earth pressure through the cofferdam wales would not have any different effect than if these pressures were applied directly to the structure.

15.5.1.5 Static Analysis The containment structure was analyzed and designed for all loading conditions combined with load factors as outlined in Section 15.5.1.2. The forces, shears, and moments in the structural shell were obtained from a computer program based on Numerical Analysis of Unsymmetrical Bending of Shells of Revolution, by B. Budiansky and P. P. Radkowski, published in the American Institute of Aeronautics and Astronautics Journal, dated August 1963.

Forces, moments, and shears in the base slab were obtained from a Stone & Webster computer program, Flat Circular Mat Foundations for Nuclear Secondary Containment Structures. The program analyzes a flat circular plate supported on an elastic foundation and computes the discontinuity stresses at the junction of the mat and cylinder, and the soil bearings pressure.

Discontinuity stresses, shears, and moments at the junction of the cylinder and mat were determined using an analogy to the Hardy Cross method for distributing fixed-end moments in continuous frames. The theoretical fixed-end moments obtained from the shell and mat computer analysis were balanced in proportion to the relative stiffness of the mat and cylinder.

An independent, manual computation, based on Theory of Plates and Shells, by S. Timoshenko, at a few selected points produces forces, shears, and moments substantially the same as those produced by the computer programs for the shell and the mat.

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-9 The containment shell program used to derive stresses in the shell assumes an isotropic material. The program does not include considerations of temperature gradients due to the thermal loadings across the containment wall.

To compute maximum stresses due to the thermal load, six general strain equations were derived, one equation for each of the four principal areas of reinforcing steel and one for each major axis of the steel liner. These equations relate strain to position, temperature, and incident stress for each item considered. To solve these general strain equations, six additional equations were used: four equations for strain compatibility, which equate radial and longitudinal strains, and two equations for load equilibrium. The solution of these equations for incident conditions gives the stress in each of the principal areas of reinforcing steel and the stress on the steel liner.

These equations permit the thermal stresses to be considered separately without modification of the major shell program.

The thermal operating load in the containment concrete wall, combined with incident condition loadings, produces a stress difference of approximately 6000 psi between the reinforcing steel adjacent to the inside face of the wall and the reinforcing steel adjacent to the outside face of the wall. This difference exists in both the longitudinal steel and the hoop reinforcing steel.

To permit the addition of these stresses to those obtained from the containment shell program, without exceeding the maximum, the containment shell program stresses are limited to 3000 psi below the maximum allowable design stress. This approach is considered extremely conservative since it limits the design stress in the interior layers of reinforcing steel to approximately 6000 psi, less than the maximum allowable design stress permitted on the exterior layers of reinforcing steel.

Structural failure cannot occur, however, until the interior reinforcing steel exceeds yield.

Up to that point plastic yielding of the outside reinforcing would be controlled by the elastic behavior of the interior steel.

In the solution of the general strain equations, the effect of the concrete has been ignored, since it is assumed to be cracked and incapable of carrying any of the tensile loads considered.

The dead load of the concrete is also ignored, as this was found to have little effect on the hoop stresses. This assumption also provides a more conservative result.

The loads exerted on the concrete shell by the thermal effects of the exposed steel liner were obtained from the calculations discussed above. The equivalent pressure, p, equals the hoop stress, f, in the steel liner multiplied by the liner thickness, t, and divided by the radius of the liner,

r. The computed equivalent pressure associated with 1.5 times incident pressure equals 5.45 psi.

Stiffness factors were used to distribute computed fixed-end moments derived from an analysis of the containment cylindrical wall, considered as a shell with a fixed-end moment, and

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-10 from an analysis of the containment mat, considered as a flat circular plate with uniform fixed-edge moment.

Stiffness factors for the cylinders were computed from formulas given in Raymond J.

Roarks book, Formulas for Stress and Strain, for long, thin-walled cylinders. Stiffener factors for the mat were computed from formulas for circular flat plates with uniform edge moment, from the same source.

Variation of the modulus of elasticity of the concrete to differentiate between uncracked and cracked concrete was not considered in determining the stiffness factors chosen.

Use of such a variable would modify the distribution of the moments and shear forces to some degree, but it is not believed that this would significantly affect the accuracy of the results.

The safety factor inherent in the present design would accommodate such small variations.

The actual distribution of the moments and forces at the junction of the wall and mat are a function of the relative stiffness of each member. This is determined by the design approach used.

Provided the total forces are distributed between the two areas under consideration, differences of distribution due to theoretical variations of the theoretical value of Youngs Modulus for concrete are not considered likely to improve the results beyond the accuracy obtained with the assumptions already used.

The methods for computing soil pressures under the mat were based upon an analogy to E. P. Popovs Method of Successive Approximations for Beams on an Elastic Foundation, published in the Proceedings of the ASCE, Separate No. 18, dated May 1950. The program computes the deflection at the center of the mat relative to a point on the mat at the intersection of the center line of the containment shell walls. The elastic curve of the mat deflection is assumed to be parabolic between these two points. Multiplying the deflection by the subgrade spring constant, the program then provides a parabolic soil pressure curve, which is combined with the rectangular soil pressure curves to provide final soil pressures under the mat. The subgrade spring constant is derived from Professor R. V. Whitmans formula:

4G k = -------------------------

( 1 - U )R where:

k = Spring constant G = Shear modulus of subgrade material R = Radius of mat U = Poissons ratio of subgrade material

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-11 While the subgrade reaction varies with depth, a single typical value for the reaction was used which is representative of the zone at the level being considered. The shear modulus was computed using a formula developed by Hardin and Black (Reference 3) from observed soil samples, and substantiated by dynamic triaxial tests of the soil. The stiffness of the soil was also based on work by G. N. Bycroft which is referred to in the Section 15.5.1.4.

A variable soil pressure conforming to the deformation of the mat was used in determining the stresses in the structure.

Maximum wind velocity associated with a tornado is given as 360 mph. This velocity was converted to an equivalent pressure using the formula P =.00256V2, where P = equivalent pressure, lb/ft2 and V = wind velocity, mph. Wind pressure was distributed over the containment dome in accordance with the methods given in Wind Stresses in Domes, by P. Gondikas and M. G. Salvadori, published in ASCE proceedings No. 2616, dated October 1960.

Wind pressure was distributed over the containment cylindrical shell in accordance with the methods given in Wind Forces on Structures, by T. W. Singell, published in ASCE proceedings No. 1710, dated July 1958.

Tornado wind loads were combined with other loads as described in Section 15.5.1.2.

An analysis of the containment structure indicated that resulting membrane stresses due to tornado wind loading in the dome reinforcing are less than 5000 psi, and that discontinuity stresses at the junction of the dome and cylinder are somewhat less.

The wind loading on the cylindrical shell creates bending, direct and shear stresses. The bending and direct stresses in the horizontal reinforcing equal 16,000 psi.

An investigation of overturning due to wind shows that the resultant (DL + wind) falls within the Kern point radius of the cylinder, indicating that the vertical reinforcing will not be subject to tensile forces from this load.

Containment torsional loadings from wind were considered negligible, in view of the ideal shape of the containment when considered as a torsion resistant shell supplemented by the diagonal reinforcing throughout the walls provided to resist earthquake loads.

The Stone & Webster computer programs for the reactor containment base slab, cylindrical wall, and dome use a constant Youngs Modulus of Elasticity and Poissons Ratio. No attempt was made to assign varying numerical values to these factors to differentiate between the relative amount of cracking in different parts of the structure.

The output of the mat program furnishes the following information:

1. Radial and tangential bending moments and vertical shear at five-foot intervals along horizontal radii from the center of the mat, spaced at 30-degree intervals.

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-12

2. Discontinuity stresses at the junction of the mat and cylinder.

3. Soil pressure.

The output of the shell program furnished forces, shears, and moments at 1-foot intervals in the height of the cylindrical wall, and at one-degree intervals in the height of the dome. Similar information is furnished at each of 16 equidistant points on the circumference of the vessel at each level considered.

Scaled load plots obtained from the computer programs for moment, shear, deflection, longitudinal force, and hoop tension are shown in Figure 15.5-2 for each of three design load conditions. The fourth design load condition did not govern design and is not represented.

The following assumptions were made:

a. The dead and live structural loads are included in all three of the design load cases.

b. Pressure load, factored and unfactored, is the dominant load condition.

c. Wind loading replaces earthquake loads where wind loads exceed earthquake loads.

d. Tornado loads are included under the general category of wind loads discussed above.

e. Buoyant water loads as discussed in Section 15.5.1.3 are substantially less than dead loads.

f. Earthquake loads, both for the operating-basis earthquake and the design-basis earthquake, are included in the analysis.

g. Thermal load from the liner is converted into an equivalent pressure and added to the incident pressure load when computing moments, shear, and tension associated with the design-basis accident.

h. Thermal load from the concrete is discussed in Section 15.5.1.5. Stresses resulting from this load are combined with incident pressure load stresses.

15.5.1.6 Reinforcing Steel Arrangement The foundation mat of the containment structure is reinforced with both top and bottom layers of reinforcing. Bottom mat reinforcing is placed in a rectangular grid pattern with layers at 90 degrees to each other. Reinforcing for the top of the mat consists of concentric circular bars combined with radial bars. The reinforcement pattern for the top of the mat is arranged to permit maintaining a uniform spacing of the vertical wall rebars that extend into the mat. Splices in adjacent parallel rebar in the mat are in general not less than 4 feet apart.

Hoop tension in the cylinder wall is resisted by horizontal bars located near both the outer and inner surfaces of the wall. All horizontal circumferential bars, including those in the dome, have their joints staggered at a minimum of 3 feet apart.

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-13 Longitudinal tension in the cylinder wall is resisted by two rows of vertical bars, one near the interior face and the other near the exterior face of the wall. Vertical bars are placed in groups of 20 bars of equal length. These are arranged so that no adjacent group in the same or opposite face of the wall has splices closer than 6 feet vertically.

See Section 15.5.2.3 for the description of the splicing scheme used for the Reactor Pressure Vessel Head Replacement Project.

The dome reinforcing consists of layers of rebar placed radially extending from the vertical reinforcing of the cylindrical wall and horizontal layers of circumferential hoop bars. Layers are located near both the inner and outer faces of the concrete. The radial pattern of the reinforcing steel terminating in the containment dome results in a high degree of redundancy of reinforcing steel in the dome. Bars are terminated beyond a point where there is more than twice the amount of steel required for design purposes. The rate of convergence of these bars, and low-stress requirements dictated by the arrangement, produces a low bond stress. In a limited number of cases where bars are terminated close to the center of the dome, anchorage stresses are more critical, and bars are hooked to provide the required anchorage. Near the crown, the rebars are welded to a concentric ring cast in the concrete.

Radial shear loads generated by internal pressure resulting from the design-basis accident are resisted by rebars inclined at 45 degrees with the horizontal and extending between the surfaces of both the vertical reinforcing closest to the interior and exterior faces of cylinder wall.

This radial shear will vary from a maximum at the base of the wall where the foundation mat restrains the independent movement of the wall to zero at some level above the mat. Anchorage bond stresses in these shear bars is kept below allowable stress levels to minimize potential cracking of the concrete. In addition, sufficient longitudinal and circumferential reinforcing is carried to the base of the wall to carry all potential loads without assistance from the radial shear reinforcing.

The tangential shears resulting from the earthquake loading are resisted by rebars inclined at approximately 45 degrees in each direction, in the plane of the wall parallel to the main reinforcing steel.

Minimum concrete cover for all principal reinforcing steel of the containment structure exceeds the requirements of ACI 318, Paragraph 808(d), which states, Concrete protection for reinforcement shall in all cases be at least equal to the diameter of the bars. The largest and principal reinforcing bar is No. 18, which requires a minimum cover of only 2-3/8 inches by the code.

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-14 15.5.1.7 Penetration Design Penetration through the containment structure is divided into one of the following three categories:

1. Pipe penetrations nine inches in diameter or less.

No special structural reinforcing is provided for penetrations nine inches in diameter or less.

Penetrations in this category are located to avoid interference with the reinforcing steel.

2. Pipe penetrations greater than nine inches and up to 3 ft. 6 in. in diameter.

For penetrations greater than nine inches, and up to and including 3 ft. 6 in. diameter, supplementary reinforcement is provided in amount and distribution such that area requirements for reinforcement are adequately satisfied.

At all these size penetrations, reinforcing steel interrupted by the openings is terminated at each side of the opening. Supplementary reinforcing was placed parallel to the interrupted bars to provide bar continuity. Horizontal, diagonal, and vertical bars were used to effectively frame the opening. The total area of reinforcement provided in any plane is not less than twice the area of steel interrupted or cut by the opening, with half of this placed on each side of the opening.

Additional reinforcing around these openings is not less than 20 feet in length, and of sufficient length to develop the full ultimate strength of the bar in ultimate bond stress to conform to the requirements of ACI 318, Section 1801(C 2). Horizontal bars are considered as top bars for this purpose.

3. Openings larger than 3 ft. 6 in. in diameter.

The two openings in this category are the 7 ft. 0 in.-diameter personnel access hatch and the 14 ft. 6 in.-diameter equipment access hatch. Details of the additional reinforcement provided around the equipment access hatch and personnel access hatch are shown in Figures 15.5-3 through 15.5-6, inclusive.

These penetrations are analyzed by means of a computer program (Reference 4). This program analyzes a ring beam based on the method of virtual work. The program assumes the ring beam to be isolated from the containment shell and loaded in two planes. The analysis includes the effect of the stiffened ring and the moments introduced by transferring external loads from the shell at the perimeter of the ring to the center line of the beam.

The ring beams are designed to resist biaxial bending moments, axial tension, torsion, and biaxial shear resulting from loading criteria listed in Section 15.5.1.2. The biaxial bending moments and axial tension are assumed to be resisted by the reinforcing bars only, the concrete being neglected. The torsional and biaxial shear stresses are assumed to be resisted entirely by binders placed radially around the penetrations. Torsion is computed by the formulas for torsion in a rectangular beam. The principal circumferential and meridional reinforcing is extended to the

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-15 inner face of the ring beam and bent at right angles, hereby providing additional shear resistance, the availability of which is considered in the design.

The normal pattern of membrane stress in the cylinder wall is interrupted in the area adjacent to the stiffened openings. This redistribution of stress was investigated by means of a computer program, based upon a paper by B. Budiansky and P. Radkowski (Reference 5). For this investigation, a flat circular plate with a radius equal to three times the distance from the center of the opening to the outside face of the stiffening ring beam was used to establish the stress pattern.

The movement of both the stiffened ring and the adjacent shell was compared to determine if significant discontinuity stresses were present. Extra reinforcement was added to regions of marked deviation from the normal pattern to keep the discontinuity effects to the level at which they can be considered negligible. The gross concrete area of the ring section was used to determine the section stiffness and rigidity.

15.5.1.8 Steel Liner and Penetrations The containment structure has an inside diameter of 126 ft. 0 in., and an interior vertical height of 185 ft. 1 in., measured from the top of the foundation mat to the center of the dome. The cylindrical steel wall liner is 3/8 inch thick, the hemispherical dome liner plate is 0.50 inch thick, and the flat base liner is 0.25 inch and 0.75 inch thick.

See Section 15.5.2.2 for the description of the restoration of the steel liner for the Reactor Pressure Vessel Head Replacement Project.

The top of the containment dome at Surry has a 39-inch penetration that was used during construction. This penetration is sealed by a welded plug on the liner side and a bolted plate on the outer end, and is filled with sandbags.

The steel lining is attached to and supported by the concrete; the liner functions primarily as a gastight membrane. The steel wall and dome liner are protected from potential interior missiles by interior concrete shield walls. CRDM missile protection is provided by a concrete shield on Unit 1 and a steel shield on Unit 2. The base liner is protected by a 1.50- to 2-foot-thick concrete mat, except in two areas where 0.75-inch-thick liner plate is used beneath the reactor vessel incore instrumentation, and at a drainage trench where floor grating provides additional protection.

The steel liner is designed to withstand the effects of all temperature, earthquake, and pressure loads, including the effect of the subatmospheric operating pressure.

The liner stress limits and their associated strains are limited to the stress criteria given in Paragraph N-1314 of Section III of the ASME Boiler and Pressure Vessel Code for nuclear vessels, and to basic primary stress levels taken from Table N-421 of that Code. The liner material SA-442-GR60 has a specified NDT that is at least 80°F below the minimum liner operating temperature, and considerably more than 120°F below the design-basis accident temperature.

Under either of these conditions, the liner material is able to accommodate at least 60% plastic

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-16 strain without cracking. A strain of this magnitude is at least 80 times greater than the maximum strain that will be imposed on the liner.

Reference to a generalized fracture analysis diagram shows that the Crack Arrest Temperature (CAT) curve crosses the NDT +80°F line at approximately 60% of the span between the Fracture Temperature Elastic (FTE) and the Fracture Temperature Plastic (FTP) ordinates, indicating that the steel can be strained to 60% of the required strain to fracture without cracking, even in the presence of large flaws.

To demonstrate that this plate material can accommodate plastic strains of this magnitude when biaxially stressed, tests were conducted on samples of 3/8-inch-thick plate of identical specification to the steel to be used in this containment liner, at a temperature of 90°F above the NDT of the steel. In these tests, the plate samples were each laid across a 23-inch-diameter ring, and a 3-5/8-inch-diameter mandrel was forced into the plate at the center line of the ring. In all cases, the mandrel deformed the plate by an amount in excess of 4 inches before shearing through.

The design-basis earthquake can be expected to produce tremors to the extent of not more than 8 to 10 cycles, and the operating-basis earthquake not more than 4 to 5 cycles.

Operating pressure variations from 9.5 psia to 14.7 psia can be expected to occur not more than 150 times during the lifetime of the unit, since personnel access is permitted under subatmospheric conditions. Temperature variations from 70°F to 105°F resulting from seasonal swings and shutdowns of the unit can be expected to occur not more than 600 times during the lifetime of the unit.

The containment liner is designed for 1500 cycles of operating pressure variations, 6000 cycles of temperature variation, and 20 cycles of design-basis earthquake, all simultaneously applied.

The containment liner is also designed for one cycle of design-basis accident pressure, one cycle of design-basis accident temperature, and ten cycles of design-basis earthquake, all considered simultaneously applied.

The containment liner is designed, within allowable working stresses, to withstand a vacuum increase of not less than 1.5 psi. The shell and dome plate liner is capable of withstanding an internal pressure of 3 psia, and the bottom mat liner is capable of withstanding an internal pressure of 8 psia, with reference to standard atmospheric conditions outside the containment.

The change in barometric pressure due to tornadoes is not expected to exceed 3 psig, and the change due to maximum hurricane will be approximately 1.1 psig. These pressure changes will result in a decrease in the atmospheric pressure, which will decrease the differential between atmospheric pressure and the containment structure ambient pressure, thereby decreasing the potential for stresses in the containment.

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-17 The accumulated effects of the above are evaluated in accordance with Paragraph N-415.1 of the ASME Boiler and Pressure Vessel Code,Section III, Nuclear Vessels.

The steel containment liner is securely anchored to the concrete wall and dome with Nelson stud-type concrete anchors. Failure could occur by stud failure in shear or tension, by studs pulling out from the concrete, or by studs tearing off from the liner plate. Tests conducted by Northeastern University, Boston, Massachusetts, using 1/2-inch-diameter studs and 3/8-inch-thick plate, show that shear failure occurs in the stud adjacent to the weld connecting the stud to the plate; in no instance was the plate damaged. Tests conducted for the stud manufacturer under the direction of Dr. I. M. Viest indicate that, with the manufacturers recommended depth of embedment of the stud in concrete, the ultimate strength of the stud material can be developed in direct tension.

The principal design load imposed on the studs is due to the subatmospheric pressure operating condition, with the anchor lattice spacing based on considerations of plate buckling. A safety factor greater than 10 is provided against stud failure in tension.

Shear due to design-basis accident conditions and earthquake will result in stresses less than the allowable working stresses.

In addition to the concrete stud anchors, the wall and base mat sections are anchored and joined at the intersection of the vertical wall and the base mat with a continuous steel skirt embedded and anchored in the concrete.

All anchors are designed so that failure occurs in the anchor, thereby assuring that the leaktightness of the containment liner will be maintained during and after anchor failure.

Probable mode of failure will be one of random stud failure due to poor workmanship during stud attachment. This type of failure will result in separation of the stud from the liner without impairment of the liner ductility or integrity.

Loss of random anchor points will not trigger a chain reaction, since the design load on each stud is low compared with the stud load capability. Design spacing of these studs is such that a group of at least 10 adjacent studs would have to fail to cause a liner plate to reach its yield stress under design operating conditions. Even with this unlikely condition, the loads on the studs adjacent to this area would remain within their safe load capability.

As shown in Figure 15.5-7, the liner was welded to a skirt ring which in turn is embedded and anchored into the concrete mat. The skirt-to-liner juncture and the skirt-to-mat anchorage were proportioned to develop the full strength of the liner. Under DBA conditions, the liner at the base juncture will be under a state of biaxial compressive strain, due primarily to thermal effects.

All thermally hot pipes penetrating the reinforced concrete containment wall pass through individual sleeves that are approximately 1 foot in diameter larger than the pipe, and project inward a distance of approximately 2 feet from the liner. A typical application is shown in

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-18 Figure 15.5-10. The pipe is welded to a thick cap that is an integral part of the end of the penetration sleeve.

Each penetration sleeve with a thermally hot pipe penetration is equipped with two water-cooled heat exchangers to limit the temperature of the liner and the concrete in contact with the sleeve. One heat exchanger is located inside the sleeve encompassing its length (inner unit);

the other is located outside the penetration sleeve in proximity to the liner. Either of the heat exchangers will provide adequate cooling for the penetration if the other is out of service. The associated component cooling water system has two independent lines. One line circulates water through the outer unit; the other circulates water through the inner unit. The inner unit limits the radial heat flow resulting from convection and thermal radiation from the thermally hot pipe penetration, to keep the temperature of the concrete in contact with the sleeve within allowable limits. In addition, the inner unit controls the longitudinal heat flow resulting from conduction from the same heat source, thus limiting the temperature of the liner and temperature gradient along the sleeve to keep the resulting thermal stresses in the liner and sleeve within the limits set forth in Section III of the ASME Pressure Vessel Code. The outer unit also limits the longitudinal heat flow, providing independent thermal protection of the penetration sleeve and liner.

The circumferential groove in the attachment plate, between the sleeve and penetration with its outside threaded connection, serves as a test chamber for the testing of the welds joining the attachment plate and penetration.

All penetrations are anchored in the reinforced concrete containment wall. The anchor strength is equal to the full yield strength of the pipe with regard to torsion, bending, and shear, and to the maximum possible pipe jet reaction. All stresses induced in the liner by these combinations of loadings are only those reflected by the resulting distortions in the reinforced concrete containment wall, and are minor in intensity. So, loads will not be imposed on the liner, thereby preserving its integrity.

All highly stressed insert plates at penetrations and equipment supports that are welded into the liner to transfer loads into the concrete have been ultrasonically tested to check for possible laminations. Tests were conducted on all plates where analysis showed a higher than average stress field, although all such plates are stressed well below the allowable limits for the materials.

These tests show that no faults exist in the insert plates.

The pipes anchored to the containment penetrations between containment isolation valves constitute an extension of the containment, and are designed in accordance with the USA Standard Code for Pressure Piping - Power Piping, USAS B31.1.0-1967, with respect to materials and allowable stress. Analyses of stresses due to thermal expansion and shock loadings from earthquake, pipe jet reaction, and other causes were made using established digital computer calculation techniques.

In order to determine the loading combinations that act on a penetration, the pipe line passing through the penetration sleeve was assumed to have failed transversely at several

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-19 locations along its run. The location at which the reaction of the ensuing jet of fluid flowing from the broken end first causes the pipe to completely yield, in either bending or torsion, was taken as the design case from which all resultant combinations of penetration loading were determined for that particular pipe line. The maximum stress allowed on any individual element of the penetration is 90% of the minimum yield point.

The intent of this criterion is to keep the material assembly components within the elastic range of the material. Under operating conditions of pressure, temperature, and external loads, the stresses in the assembly will be within the limits established in Section III of the ASME Pressure Vessel Code.

As a part of the issues identified in NRC GL 96-06, isolated containment penetration piping with confined fluid was reviewed for susceptibility to thermal over-pressurization following a DBA. The linear elastic analysis criteria stipulated in the 1989 version of the ASME Boiler and Pressure Vessel Code Section III, Appendix F was used for structural integrity evaluation. The internal pressure in piping penetrations during a design basis accident (LOCA or MSLB) was calculated by taking into account the difference in the expansion of the fluid and the pipe, the temperature increase immediately following the DBA and credit for a limited amount of circumferential strain in the pipe. The analysis established that thermally induced over-pressurization of isolated water-filled piping sections in the containment boundary could not jeopardize the ability of the accident mitigating systems to perform their safety functions and could not lead to a breach of containment integrity (Reference 10).

All liner seams were strength-welded. Small steel channels welded continuously along the edges of their flanges to the liner plate cover the plate weld seams, in a manner similar to those installed at the Connecticut Yankee Station. These channels are zoned into test areas by dams welded to the ends of the sections of the channels. Fittings are provided in the channels for periodic testing of the weld seams for leaktightness under pressure. Typical liner details are shown in Figure 15.5-12. Testing of the liner is described in Section 5.5.

To transfer the stress adequately around penetration openings, or to transfer the pipeyield load adequately to the concrete within the limits of this material, whichever is larger, the liner is reinforced in accordance with the rules set forth in the ASME Boiler and Pressure Vessel Code, 1968,Section III, Nuclear Vessels.

All major equipment and pipe loads are carried on the interior concrete structure or by the neutron shield tank. A 1.50- to 2-foot-thick concrete slab placed over the bottom mat steel liner provides anchorage and support for other equipment located in the base of the containment structure. The neutron shield tank skirt is attached to the containment mat by 1.50-inch-diameter anchor bolts. The skirt support was welded to the liner, and the entire weld, including the anchor bolts, covered by test channels. The internal concrete structure is attached to the containment mat by lengths of 3-inch by 6-inch steel bars which, placed horizontally, intersect the steel plate liner as shown in Figure 15.5-8. The main vertical reinforcing steel bars were welded to the top and

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-20 bottom faces of these bars, thus providing bar continuity without creating multiple penetrations through the liner.

The 1.50-foot-thick concrete slab is anchored through the steel liner plate in a similar manner using 7-inch by 0.50-inch bars, as shown in Figure 15.5-8. These bars, termed bridging bars, form an integral part of the steel liner, and conform to the material and workmanship specifications of the steel liner. All welded joints are covered by test channels and tested as all other liner plate joints.

Access to the containment structure is provided by a 7 ft. 0 in. i.d. personnel hatch and a 14 ft. 6 in. i.d. equipment hatch. Other smaller containment structure penetrations include hot and cold pipes, main steam and feedwater pipes, fuel transfer tube, and electrical conductors.

Electrical conductors penetrating the containment structure range in size from No. 16 AWG thermocouple leads to 1-inch-diameter solid copper rods for 4160V power circuits. Each penetration group passes through 8-inch-diameter steel sleeves. The sleeves were welded into the containment liner with a test channel around the weld for periodic leak testing, as shown in Figure 15.5-9 (Amphenol electrical penetration depicted).

The basic Amphenol electrical penetration consists of an eight-inch steel tube with bolted-on flanges, through which pass the sealed conductors. The hermetically sealed connectors, as shown in Figure 15.5-9, were bench-tested for leaktightness.

Each flange is held tightly in place with eight bolts that draw the flange against a high temperature sealing ring and a backing plate welded to the sleeve. Each flange is tapped for leak testing. A make-up method is used to determine the penetration leakage by applying a test pressure equal to greater than containment design pressure (45 psig) between the o-ring seals. An electrical connector may be replaced, if necessary, without welding or cutting the containment liner or sleeve.

The design and qualifications of the Amphenol electrical mating connectors are based upon the requirements of military specification number MIL-C-5015. Connector design is such that silastic components are provided in the connector to feed through the interface. This type of interface has been proven adequate to meet the environmental requirements of MIL-C-5015.

Additional capability to withstand elevated temperatures is provided in the material used for the sealing members.

The original tests conducted at the Amphenols shop consisted of the following:

Connectors installed in the flanges normally operate at ambient conditions of 105°F and 9.75 psia, and were tested for leak rate and tagged for integrity before shipment to the job. A test facility was set up by the manufacturer suitable for 50 psig, with provisions for thermocycling from 32° to 300°F. A thermocycle run of at least three cycles was made on one of each type flange. A time interval of 30 minutes was

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-21 allowed between the thermocycles. The leak rate test after thermocycling was made at 50 psig and 300°F. Each completed flange had a leak rate of less than 1 x 10-6 cc/sec per assembled flange. All flanges were leak tested at 50 psig and 300°F. Helium gas was used in the test facility.

For the Amphenol triaxial cable penetrations a more detailed procedure for the thermocycle test was followed in shop test:

The type sample consisted of a containment side flange disk with hermetic assemblies welded in place. A thermocouple was installed to monitor disk temperature. The disk was stabilized at 32°F and then placed in an oven heated previously to 280°F. On entrance of the disk, the oven temperature was reduced, straight line, to 150°F over a 60-minute period. The disk was removed and cooled to 100°F, while the oven was reheated to 280°F. The disk was then returned to the oven and the oven temperature reduced to 150°F as before. The highest metal temperature reached during this cycle was recorded and a 50-psig helium leak test was conducted at this metal temperature for all discs of this type.

Each Amphenol penetration assembly, without external cable mating connectors, was tested in the factory to demonstrate insulation resistance of at least 1000 megohms at 1000V dc. In addition, each penetration has passed an overpotential test. After initial installation, each penetration with external cables connected was tested at 1000V dc for 5 minutes.

Containment electrical penetrations now in use at the Surry Power Station were manufactured by either Amphenol Space and Missile Systems, Conax Corporation, or Westinghouse Electric Corporation. Nitrogen pressure is not required for penetration functional capability; however, each penetration is capable of being pressurized with nitrogen for leak detection purposes. RTV-8112, THIOKOL, or POLYSYLFONE are used to provide a tight seal around conductors.

Amphenol penetration electrical connectors were tested by D. G. OBrien, Inc. in 1972. The purpose of this test was to demonstrate operability during simulated LOCA conditions. The connectors passed the test with no less than 34 megohms internal resistance while retaining complete electrical continuity. The test had no observable physical effect on the connector assembly or cable. No connectors are associated with the CONAX or Westinghouse type penetrations; however, both manufacturers have provided data regarding the performance of the materials used in their penetrations. This information includes thermal performance, radiation resistance, and chemical resistance tests. All data indicate excellent performance characteristics for a LOCA environment.

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-22 All containment structure piping penetrations consist of a basic containment insert, plus additional items, as required for the individual services. Two basic types of penetrations are used for piping systems:

1. Unsleeved - These penetrations consist of piping installed through the containment wall without a sleeve around the outside of the piping. Unsleeved penetrations are used for cold piping systems (temperature of the fluid in the piping is less than 150°F) when only one pipe passes through the penetration.

2. Sleeved - These penetrations have a sleeve around the outside of the piping. Sleeved penetrations are used for all multiple piping systems passing through one penetration and for all thermally hot (over 150°F) piping systems, both single and multiple. Typical piping penetrations are shown in Figure 15.5-10.

The main steam and feedwater penetrations are provided with adequate space between the piping and the sleeve for the necessary pipe insulation, and for a pipe coil outside the insulation through which component cooling water is circulated. This cooling coil reduces the temperature of the sleeve and prevents any excessive heating of the concrete in contact with the sleeve. All welded seams subjected to containment pressure are leaktested by introducing air through each test boss. In addition, the sleeve end is drilled and tapped, as shown in the details, so that any leakage between pipe wall and sleeve end can be detected during periodic containment leakage testing.

The liquid and gas pipe penetration assemblies, in nearly all instances, consist of more than one pipe inside the penetration sleeve. The diameter of the sleeve depends on the number and size of the pipes installed in a given penetration. Each of these penetrations was tested with air using the same procedure as that used for the steam and feedwater penetrations.

A 20-inch o.d. fuel transfer tube penetration is provided for fuel transfer between the refueling canal in the containment structure and the spent-fuel pool in the fuel building. The penetration consists of a 20-inch stainless steel pipe installed inside a 26-inch pipe, as shown in detail in Figure 15.5-10, Sheet 1. The inner pipe acts as the transfer tube, and connects the containment refueling canal with the spent-fuel pool. The outer pipe is welded to the containment liner, and provision is made, by use of a special seal ring, for air leak testing of all welds essential to the integrity of the penetration. Bellows expansion joints are provided on the outer pipe to compensate for any differential movement between the two pipes.

The equipment hatch is a 14-ft. 6-in. single closure penetration. The equipment hatch cover is mounted inside the containment structure and is double gasketed with a leakage test tap between the o-rings. The equipment hatch cover is provided with a hoist with two point suspension and a sliding rail for storage. A positive locking device is furnished to prevent circular swing. The equipment hatch was designed, fabricated, and stamped in accordance with the ASME Boiler and Pressure Vessel Code,Section III, Class B. A removable concrete tornado missile shield protects the equipment hatch and acts as equivalent shielding.

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-23 The equipment hatch opening is analyzed in two basic steps, using the actual space curve shape. In the first step, it is assumed that the pattern of stress concentration at the junction of the cylinder and the ring beam is the same as if the cylinder were of infinite radius and the ring and the cylinder were in the same plane; that is, the cylinder wall is assumed to be a flat plate. The forces are imposed on this flat plate at such a distance from the opening that they are not influenced by the opening. These forces are membrane meridional (vertical) loads, circumferential (hoop) loads, and tangential shear loads. An outward longitudinal force is delivered to the inside face of the ring beam by a bearing plate which is welded to the liner plate.

This force is caused by incident pressure acting on the hatch.

The first step of analysis is performed by using the Stone & Webster Shell I computer program, which is based on general first-order linear plate theory modified by Sanders. This program is developed from a numerical analysis of shells of revolution as published by Bernard Budiansky and Peter Radkowski in the AIAA Journal, Vol. 1, No. 8, August 1963.

In the second step, the ring is isolated and analyzed, taking into account its actual geometry (curved in two planes). The loads imposed on the ring beam have been obtained at the junction of the ring and normal shell in the first step, plus the incident pressure loads and temperature effects on the ring surface.

The second step of the analysis is performed by means of the Stone & Webster computer program, Reinforcing Opening in a Cylindrical Structure. This program analyses structures curved in space about two axes. The loads imposed on the ring are the membrane forces at the juncture of the thickened ring and the normal shell, as obtained from the first step, and a modified incident pressure acting on the ring beam and hatch cover.

The analysis of the isolated ring is based on the theory of curved beams, as demonstrated in Seely and Smith, Advanced Mechanics of Materials. Although the theory in this textbook is confined to the one-dimensional, curved beam, the assumptions set forth are extended to the two-dimensional case. These assumptions permit a simplified calculation of the stresses and deformations.

The ring, loaded in two planes, is statically indeterminate to the sixth degree. The analysis for the ring in space consists of cutting the ring, imposing six unknown loads at the cut section, and solving for the six unknown forces by equating differential deflections and rotations on either side of the cut to zero. The six unknown forces are: direct force, a torsional moment, a transverse shear, a radial shear, and bending moments about two axes. The curvature of the ring is considered in obtaining the bending moment strains and effects on total strain energy.

An emergency airlock is provided through the equipment hatch for emergency access to the containment. The airlock is flanged to the outside of the equipment hatch cover utilizing a double o-ring seal, and has an outside diameter of 6 ft 0 in., and a length of 12 ft 8.50 in. A 30-inch-diameter door is located at each end of the air lock. The air lock doors, which swing toward the center of the containment, are interlocked so that one door cannot be operated unless

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-24 the other is closed. These are mechanical interlocks, and provisions have been made for deliberate violation of the interlock by use of a special tool. This tool shall be kept under administrative control.

Each door is equipped with a valve for equalizing the pressure across the door. At no time can the equalizing valves on both doors be open simultaneously, and in no case can an equalizing valve be open on one door while the other is operating.

The operations required at each station for engagement or release of the interlocks, for operation of the equalizing valve, and for opening or closing the door, are accomplished by rotation of a single handwheel. Provisions have been made to allow operation of the outer door from inside the containment and the inner door from outside the containment, in addition to local operation.

Both doors are designed to withstand the containment test pressure of 52 psig. Each door is also designed to withstand 8.0 psia pressure in the containment structure with full atmospheric pressure outside. The interior door is provided with an additional securing device to facilitate testing the air lock to the maximum test pressure when the containment structure is at 8.0 psia.

All shafts penetrating the door or bulkhead have double packing. A blind flanged emergency air port is provided on the air lock outside containment. A light is provided inside the air lock and is powered near the air lock for communication.

A track is provided for emergency air lock removal via a cart. The track consists of two continuously supported rails that extend through the equipment hatch barrel onto the platform.

Chicago Bridge and Iron Company designed and installed the track inside the equipment hatch barrel, and Stone & Webster designed the identical mating rails on the platform.

The tornado missile shield outside the containment equipment hatch has been modified to provide a labyrinth passage to the air lock. The missile shield slabs are fastened to the equipment hatch platform, which consequently has been modified. The equipment hatch platform has sufficient structural steel to withstand tornado wind loads on the attached missile shields.

The design, fabrication, and testing of the emergency air lock was performed by Chicago Bridge and Iron Company according to the ASME Code Section III, Subsection NE, 1971 Edition through the Winter 1972 Addenda. Welder procedures and performance qualifications were controlled under ASME Code Section IX.

The personnel hatch is a 7 ft. 0 in. i.d. double closure penetration as shown in Figure 15.5-11. Each closure head is hinged, double gasketed with a leakage test tap between the o-rings. Both doors are interlocked so that in the event one door is open, the other cannot be actuated. Both doors are furnished with a pressure equalizing connection. The equalizing valves are manually operated by persons entering or leaving the personnel hatch. The personnel hatch was designed, fabricated and stamped in accordance with the ASME Boiler and Pressure Vessel

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-25 Code,Section III, Class B. The personnel hatch is externally protected from tornado missiles by concrete shield walls and roof.

An 18-inch-diameter manway on the inner door of the personnel airlock is also provided for emergency egress from the containment. A positive locking device prevents inadvertent opening of the emergency manway. Manway position indication is provided in the control room. Alarm indication is also provided in the control room, and on the control panels on either side of the personnel airlock inner door, to indicate whenever the manway locking bar is not in the proper position to prevent inadvertent opening of the manway.

Material for the liner and penetrations is carbon steel plates conforming to ASTM A442, Grade 60, which has a specified minimum tensile strength of 60,000 psi, a minimum guaranteed yield strength of 32,000 psi, and a guaranteed minimum elongation of 25% in a standard 2-in.

specimen. The liner has sufficient ductility to tolerate local deformation without rupture. This material has a nil ductility transition temperature of -20°F, which is 80°F below the normal minimum shutdown temperature given in Section 5.4.1.

Steel items, except backing plates and anchors, gas testing channels, equipment hatch bolts, and equipment hatch nuts are made to fine grain practice and normalized. In addition, steel items other than the above have passed NDT tests performed in accordance with the following:

1. Material 5/8 inch and thicker was tested by the Drop Weight Test method in accordance with ASTM E 208.

2. Material less than 5/8 inch thick was tested by the Drop Weight Tear Test method as developed by the U.S. Naval Research Laboratory (NRL Report 6300).

3. Material 5/8 inch and thicker has an NDT no higher than -20°F.

The liner plates were ordered to conform to standard mill practice with regard to thickness tolerances. Therefore, the 3/8-inch-thick cylindrical shell liner plate ranges in thickness from 0.365 inches to 0.406 inches. The 0.50-inch-thick hemispherical dome liner plate ranges in thickness from 0.490 inches to 0.535 inches, and the 0.25-inch-thick flat base liner plate ranges in thickness from 0.240 inches to 0.285 inches.

Physical and chemical properties of materials used in the construction of the containment liner, weldability tests, and liner thickness were checked by the Stone & Webster Field Quality Control Organization on a random sampling basis.

All welding procedures and tests required in Section IX of the ASME Boiler and Pressure Vessel Code for Welding Qualifications were adhered to in the selection of weld rod material, weld rod flux, heat treatment, and qualification of the welding procedures and the performance of welding machines and welding operators engaged in the construction of the containment liner.

The welding qualification included 180-degree bend tests of weld material. These procedures ensure that the ductility of welded seams was comparable to the ductility of the containment liner plate material.

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-26 Section III of the ASME Boiler and Pressure Vessel Code for Nuclear Vessels was used as a guide in the selection of materials.

Erection of the steel liner followed completion of the concrete mat. The 3/8-inch-thick steel wall liner was erected to approximately Elevation +60 ft. The 0.25-inch-thick mat liner plate was installed on top of the concrete foundation mat during this period. On completion of the wall liner to Elevation +60 ft. and completion of the mat liner, all welds were checked for compliance with the approved weld inspection and gas test requirements. Work on the liner was then stopped until the containment interior concrete structure was completed, the polar crane was erected, and the concrete containment wall was completed to ground grade (Elevation 26 ft. 6 in.).

The 3/8-inch-thick steel wall liner was erected from Elevation +60 ft. to Elevation +92 ft.

6 in., and the containment liner completed with the construction of the 0.50-inch-thick steel dome liner. A-1 welds were inspected and gas-tested for compliance with the weld requirements.

The reinforced concrete wall, above ground grade, was completed, following as closely as practical the construction of the wall liner.

The reinforced concrete dome was constructed upon completion of the dome liner.

The steel wall liner was braced internally and locally with temporary bracing to prevent distortion during concrete placement. The exterior concrete forms were supported from the placed concrete and tied to form a tension ring.

Cantilevered steel strongbacks were used in the construction of the concrete dome to support the steel dome liner against deformation due to the weight of reinforcing steel formwork and wet concrete. Strongbacks were cantilevered from the completed concrete of the dome.

The containment liner is not a coded pressure vessel, so there was no section of the ASME Boiler and Pressure Vessel Code for Nuclear Vessels directly applicable to its design and construction. However, to ensure that good engineering practices were followed, certain portions of Section III of the Code were reviewed for suggested guidance as to design and construction practices that should be incorporated in the liner specifications. Those sections reviewed for information were:

• N-511 Certification of Materials by Vessel Manufacturer

• N-512 Material Identification

• N-513 Examination During Fabrication

• N-514 Repair of Material by Welding

• N-515 Forming Shell Sections and Heads

• N-518 Attachments

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-27

• N-519 Cutting Plates and Other Products

• N-521 Welding Processes

• N-522 Welding Qualifications and Weld Records

• N-523 Precautions for Welding

• N-524 Assembly

• N-526 Finished Longitudinal and Circumferential Joints

• N-527 Miscellaneous Welding Requirements

• N-528 Repair of Weld Defects

• N-531 Preheating

• N-541 Modification of Section IX - Welding Procedure Qualification Requirements

• N-611 Inspection, General

• N-612 Qualification of Inspectors, Engineering Specialists, and Inspection Agencies

• N-613 Access for Inspector

• N-614 Inspection of Materials

• N-615 Marking on Plates and Other Material

• N-616 Final Inspection

• N-620 Inspection of Welding

• N-622 Check of Welder and Welding Operator Performance Qualifications

• N-623 Check of Nondestructive Examination Methods

• N-625 Ultrasonic Examination of Welded Joints

• N-626 Magnetic Particle Examination

• N-627 Liquid Penetrant Examination

• N-713 Pneumatic Test

• N-714 Pressure Test Gauges

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-28 The liner attachments are Nelson concrete anchors, welded on a triangular pattern to the wall and dome liner, and cast in the containment concrete as the concrete was poured against the liner. The attachment spacing was determined by the procedure (Reference 6) set forth for buckling of a cylindrical shell under combined axial and uniform lateral pressure where each attachment constitutes a buckling wave nodal point and was so spaced that the critical buckling stress will take place in plastic range of the liner material. The liner dome was treated in a similar manner. Maximum variation from the correct stud location, where relocation was necessary to avoid an obstruction, did not exceed 1.50 inches. The bottom mat liner was covered with 1.50- to 2-foot-thick reinforced concrete slab to protect it from both pressure and temperature loadings, so that it will remain virtually unstressed.

All penetrations are anchored into the concrete containment structure wall with a loading resistance level greater than the plastic strength of the penetration pipe. Openings in the liner plate are reinforced with reinforcing plate, and/or collar, sized to develop the full relief of the liner plate. The stress around each reinforced opening was analyzed in accordance with the appropriate procedure (Reference 7).

Departure from the original specified out-of-roundness tolerance of the reactor containment liners was necessary due to erection difficulties. Attempts were made to obtain the specified tolerance by means of an adjustable ring girder and supplementary anchorage to the cofferdam.

As work progressed above the cofferdam level, it was found that it was impractical to obtain the specified liner tolerance.

A thorough review was made of the necessity for this close tolerance, and it was found unnecessarily restrictive.

The liner shell and dome are studded to the concrete and the plate is essentially plane within an equilateral triangle, 12 inches at the base and bounded by studs at the apexes of the triangle.

The response of each individual triangular element to its own particular loading system establishes the adequacy of the structure as a whole. Therefore, actual roundness of the shell has no effect on liner performance.

The following revised out-of-roundness tolerances were adopted after a thorough review of the problem.

1. The out-of-roundness tolerance shall not exceed plus or minus 3 inches from the true radius.

2. The maximum plus or minus deviation from a true circular form shall not deviate more than 0.25 inches from a straight line in any 14-inch space in any plane in any location on the liner.

The revised out-of-roundness tolerances have no adverse effect on the buckling strength of the liner, and ensure that plate buckling between studs will not occur in the elastic range.

The adjustable ring girder was found to be of limited value during the erection of the liner, due to the many liner penetrations and the stiffness of the liner shell. Therefore, the ring girder

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-29 was used for rounding the shell only in areas where its application was found advantageous by the liner fabricator.

15.5.1.9 Materials See Sections 15.5.2.2, 15.5.2.3, and 15.5.2.4 for descriptions of the construction materials used for the Reactor Pressure Vessel Head Replacement Project.

15.5.1.9.1 Concrete The description of concrete materials is given in Section 15.3.1.

See Section 15.5.2.4 for the description of the concrete used for the Reactor Pressure Vessel Head Replacement Project.

15.5.1.9.2 Porous Concrete Porous concrete is used under the base mat to provide drainage for the containment structure. The type of concrete is formed by the omission of the fine aggregate from a standard structural concrete mix. The mix was designed to have a 28-day compressive strength greater than 1000 psi.

Water porosity tests were performed earlier in an independent laboratory for porous concrete, using 6-inch by 12-inch cylinders prepared in the laboratory by compacting the material in three layers with standard tamping rods. A varying number of strokes, ranging from 10 to 40 for each layer, were used for different cylinders. After the concrete test cylinders had been properly cured, the amount of water that would flow through the 12-inch length of specimen during a three-minute period with a constant head of 4 inches of water above the top of each cylinder was determined. Results indicated water porosities of from 28 to 47 gpm/ft2, depending upon the amount of compaction and resulting density of the cylinders.

The porosity determined by the laboratory tests indicated that the four-inch porous concrete layer under the base mat provides adequate drainage, since the leakage through the membrane waterproofing of the container would be minor. This layer serves as the collection means for the seepage removal system in the mat, described in Section 15.5.1.3.

15.5.1.9.3 Reinforcing Steel Special large-size reinforcing bars, No. 14 and No. 18, used in the construction of the reactor containment structure, are steel of 50,000 psi minimum yield point, conforming to Grade 40 of the Standard Specification for Deformed Billet-Steel Bars for Concrete Reinforcement, ASTM A615 as modified to meet the following chemical and physical requirements:

• Carbon 0.35% maximum

• Manganese 1.25% maximum

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-30

• Silicon 0.15 to 0.25%

• Phosphorus 0.05% maximum

• Sulphur 0.05% maximum

• Minimum yield strength 50,000 psi.

• Elongation 16% minimum in a 2-inch test sample

• Tensile strength 70,000 to 90,000 psi For these special chemistry bars, all ingots were identified and all billets were stamped with identifying heat numbers. All bundles of bars were tagged with the heat number as they came off the rolling mill. A special stamp marking was rolled into all bars conforming to this special chemistry, to identify them as processing the chemical and mechanical qualities specified.

See Section 15.5.2.3 for the description of the reinforcing steel used for the Reactor Pressure Vessel Head Replacement Project.

The engineers quality assurance inspectors witnessed, on a random basis, the pouring of the heats and the physical and chemical tests performed by the fabricator. Bars containing inclusions, or failing to conform to the required chemical and physical requirements, were rejected.

One 12-inch-long test sample was furnished to the engineers from a finished bar from each heat of the special chemistry rebars, to permit independent verification of physical and chemical analysis tests by the engineers.

Test specimens for the special chemistry rebars conformed to Section 10.1.1 of ASTM A615 and were Standard 0.505-inch-diameter specimens with 2-inch gauge length. Rate of loadings was such that the tension-tested sample was brought to the yield point in not less than 2 minutes.

For containment structure, reinforcing steel, consisting of No. 11 bars and smaller, is of 40,000 psi minimum yield point, conforming to Grade 40 of the Standard Specification for Deformed Billet-Steel for Concrete Reinforcement, ASTM A615.

The reinforcing steel for structures other than the containment structures is described in Section 15.4.3.

15.5.1.9.4 Cadweld Splices Cadweld reinforcing steel splices, Type T full tension splices, as manufactured by Erico Products, Inc., Cleveland, Ohio, were used to splice 50,000 psi minimum yield point reinforcing bar sizes No. 14 and No. 18. These splices, including the sleeves, develop tensile strengths not less than 90% of the minimum ultimate strength of the reinforcing bar. The average value of two

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-31 or more successive splices develop at least the minimum ultimate strength of the rebar.

Information for splices other than No. 14 and No. 18 reinforcing bars is given in Section 15.4.3.

See Section 15.5.2.3 for the description of Cadwelds, including operator qualification and tensile testing, used for the Reactor Pressure Vessel Head Replacement Project.

15.5.1.9.5 Waterproofing Membrane The waterproofing membrane is a flexible polyvinyl chloride sheet having a minimum thickness of 40 mils. Associated adhesives and tapes consist of the membrane manufacturers recommended material for the application conditions.

15.5.1.10 Construction Procedures and Practices After performing the general excavation described in Section 15.4.4, two 149-ft.

5.25-in.-diameter cofferdams were constructed, one for each reactor. The cofferdams consist of interlocking steel sheet piles supported by a system of heavily reinforced concrete internal ring wales. The top of the sheet piles is at Elevation +10 ft. and tip grade is at Elevation -48 ft. The interior of the cofferdams was excavated to approximately Elevation -41 ft. Seepage drains were then driven through a 12-inch layer of crushed stone placed in the bottom of the excavation, as described in Section 15.5.1.3.

A 2-inch-thick concrete leveling slab was placed over the crushed stone and 40-mil-thick vinyl waterproof membrane placed over this concrete. A 4-inch layer of porous concrete was then placed over the membrane to protect the membrane and to serve as an internal drainage system, as described in Section 15.5.1.12.

Porous concrete was also placed around the sides of the cofferdam to fill the space between the cofferdam and the edge of the concrete mat, and to provide a form for the mat concrete. The waterproof membrane was extended vertically in this area, and protected by concrete block.

The reinforcing steel, steel bridging bars as described in Section 15.5.1.8, and other miscellaneous steel inserts required in the containment mat were placed, and the concrete poured.

The mat was constructed in six sections.

The 3/8-inch-thick steel wall liner was then erected to Elevation +60 ft on the containment wall. The steel mat liner plates were installed on top of the concrete mat. All welds were checked for compliance with the approved weld inspection and gas test requirements. The containment interior concrete structure was then built on the mat liner. On completion of the interior concrete structure, the polar crane was erected.

The exterior containment concrete wall was constructed to approximately Elevation 24 ft.

6 in. during the construction of the interior concrete. On completion of the concrete substructure, a vinyl waterproof membrane was attached to the exterior concrete surface with adhesives. The membrane completely encloses the containment structure below grade.

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-32 The space between the cofferdam and the containment structure was then backfilled with crushed stone compacted in 6-inch layers. A 2-foot-thick layer of compacted impervious fill was placed at Elevation -4.0 ft. to seal the area and to minimize the amount of ground water seeping into the area.

The liner was then completed and finished with the construction of the 0.50-inch-thick steel dome, with all welds inspected and gas-tested. The steel dome liner was supported during erection with open web steel trusses.

See Section 15.5.2.2 for the description of the restoration of the steel liner during the Reactor Pressure Vessel Head Replacement Project.

The reinforced concrete wall above ground grade was completed, following as closely as practical the construction of the wall liner.

The completed steel wall liner was braced internally and locally with temporary bracing to prevent distortion during concrete placement. The exterior concrete forms were supported from the preceding concrete.

Cantilevered steel strongbacks were used in the construction of the concrete dome to support the steel dome liner, reinforcing steel, formwork, and wet concrete against deformation.

Strongbacks were cantilevered from the completed concrete of the wall or the dome.

Careful inspection of the dome was maintained during concrete placing and until the concrete had definitely taken initial set. Concrete buckets used during the first two lifts of the dome were limited to 2 yd3 in size. Bucket sizes were increased after the second lift had set, when placing results of these lifts were satisfactory and warranted such a move.

Concrete in the wall and dome of the containment structure was poured in uniform 6-foot lifts around the entire circumference. Each lift was constructed in approximately 18-inch layers.

See Section 15.5.2.4 for the description of the concrete used for the Reactor Pressure Vessel Head Replacement Project.

Concrete forms were used on the exterior of the concrete dome to a line 50 degrees above the horizontal. The permanent steel liner served as the inner form for pouring concrete. For the area where exterior forms were used, the concrete points were in horizontal planes. Above the 50 degree line, the remainder of the dome concrete was poured as one lift.

Particular care was taken to check the special markings of the No. 14 and No. 18, 50,000-psi minimum-yield rebars for the containment structure.

See Section 15.5.2.3 for the description of welded splices and Cadwelds, including operator qualification and tensile testing, used for the Reactor Pressure Vessel Head Replacement Project.

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-33 Welded splices conform to Recommended Practices for Welding Reinforcing Steel, Metal Inserts, and Connections, in Reinforced Concrete Construction, AWS D12.1. Bars spliced by metallic arc welding develop not less than 90% of the minimum ultimate strength of the reinforcing bar, and the average of two or more successive splices develop at least the minimum ultimate strength of the bar.

Structural ductility was maintained by staggering critical splices where possible. Full scale pressure tests conducted in May 1967, on a recently completed concrete containment structure (Reference 8) in which similar Cadweld splices and welded splices were used, showed no stress concentrations or lack of structural ductility. Locations of splice groups were not discernible from inspection of the test crack patterns.

All Cadweld Process Type T joints were visually inspected. The visual inspection included inspection of the ends of the bars for dryness and cleanliness prior to fitting the sleeve over the ends. It also included inspection of the completed splice for properly filled joints to ensure that filler metal was visible at both ends of the sleeve and at the top hole in the center of the sleeve. Randomly selected splices were removed from the structure and strength-tested for compliance with the specification. Joints that did not meet all these inspection criteria were replaced.

Randomly selected Cadweld Type T splices were removed from the containment structure and tensile-tested for compliance with the specifications, in accordance with the following schedule:

1. One out of first 10 splices.

2. Three of next 100 splices.

3. One out of each subsequent unit of 100 splices.

Welding inspection of reinforcing bars was by quality control inspectors. Radiographic inspection, dye-penetrant inspection, magnetic-particle inspection, or other nondestructive inspection methods for welded joints was performed on a random basis under the direction of the Senior Quality Control Engineer.

All welds were visually inspected. Any cracks, porosity, or other defects were removed by chipping or grinding until sound metal was reached, and then repaired by welding. Peening was not permitted.

Completed welded splices were selected on a random basis and removed from the structure with suitable lengths of adjacent bars. These removed splices were tensile-tested for compliance with the specifications in accordance with the same schedule followed for the Cadweld Type T splices.

Tack welding of special chemistry rebar was not permitted.

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-34 15.5.1.11 Missiles and Piping Rupture 15.5.1.11.1 Interior Missiles Most of the high-pressure piping and equipment of the primary coolant system is located within containment cubicles protected by reinforced concrete walls and floors with a minimum thickness of 2 feet. The control rod drive mechanisms are provided with a separate missile shields (reinforced concrete for Unit 1 and steel for Unit 2). These structures will terminate the flight of any conceivable missile. Openings in the charging floor required for ventilation or access are covered by steel grating, which is designed to provide adequate missile protection. Openings in cubicle walls for overpressure blowoff protection are directed in a manner that will minimize the possibility of missiles striking the containment liner. An analysis of the missile hazard has been performed and the conclusions are as follows:

Missiles could be either concrete or steel. Because of lower density and lower strength, a concrete missile would have to be an order of magnitude heavier than a steel missile of comparable diameter and velocity for it to cause the same damage on impact with a steel shell. Also, in the context of the design-basis accident, there are more potential steel missiles and these have been studied in detail.

The most penetrating steel missile for a given mass and velocity would be rod-shaped, impacting end-on; therefore, rods of various diameters and weights have been investigated.

Missile velocities of 100 fps might be generated by rupture of a reactor coolant loop1, and this value has been used with penetration equations developed by D. A. Davenport (Reference 9) to estimate their penetrating capability.

Table 15.5-3 summarizes the results of the analysis. Inspection of these results indicates that, except for the containment liner, at 100 fps, the required weight and dimensions for penetration of the metal thicknesses of interest are not credible for missile sizes that can be postulated within the reactor containment. The metal thicknesses shown in the table bracket the thicknesses of interest for the containment liner and piping systems. The analysis for the containment liner does not consider the added resistance to penetration afforded by the interaction between the concrete containment structure and the containment liner. This added resistance will not permit penetration by missiles of credible weight and size. Major components, such as the steam generator, have greater shell thicknesses than the values in the table, and therefore will not be penetrated by the postulated missiles.

All potential missiles were evaluated, and those that constitute a hazard to either the liner or adjacent equipment, by virtue of their velocity and/or size, are restrained by local barriers or other mechanical means.

1. As discussed in Section 15.6.2, it is no longer necessary to consider the dynamic effects of a postulated rupture of the primary reactor coolant loop piping. However, pipe ruptures of other high-energy lines are still postulated.

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-35 15.5.1.11.2 Exterior Missiles The Surry Power Station site is approximately 12 miles from the nearest commercial airport at Newport News, Virginia, and 15 miles from Langley Air Force Base. The site is not on the normal approach path to either of these air fields.

An analysis of hypothetical aircraft accidents indicates that the most likely missile that might penetrate the reactor containment would be a turbojet rotor. Calculations show that the 2-ft.

6-in.-thick containment dome would withstand without penetration the direct impact of a 1500-lb rotor impacting at a velocity of 400 mph. The 4-ft. 6-in.-thick containment walls would withstand a similar missile impacting at a velocity of 980 mph. These velocities are considerably in excess of low-level aircraft approach speeds.

Tornado-generated missiles are discussed in Section 15.2.3, and include two potential missiles:

1. Missile equivalent to a wooden utility pole 40 feet long, 12 inches in diameter, weighing 50 lb/ft3, and traveling in a vertical or horizontal direction at 150 mph.

2. Missile equivalent to an automobile weighing one ton, traveling at 150 mph. Neither of these missiles would penetrate the reactor containment.

15.5.1.11.3 Pipe Rupture Incident The containment internal structure is designed to accommodate the loading due to rupture of the reactor coolant and connecting piping1, or main steam or feedwater piping. Incident rupture was considered in only one line at a time. The support system was designed to preclude damage to or rupture of any of the lines as a result of the incident. The snubber and key systems are designed to transmit rupture thrusts from a steam generator into the containment internal structures. In determining the steam generator support reactions, the system was reduced to a dynamic model consisting of a suitable number of masses and resistance elements under impulse loading. The structural support system resilience and mass was included in the model. The dynamic problem was solved by numerical methods, using a thrust-time history as loading. Resistance, dynamic amplification of the thrust, and rebound forces were calculated versus time. Design of the support system was based upon stress levels defined in Section 15.5.1.8. The reactor vessel and support system were similarly treated.

The steam lines are strapped to the crane wall at intervals selected to prevent a whipping pipe from contacting the liner. The straps are designed so that no interference with the normal thermal expansion modes of the steam lines results.

1. As discussed in Section 15.6.2, it is no longer necessary to consider the dynamic effects of a postulated rupture of the primary reactor coolant loop piping. However, pipe ruptures of other high-energy lines are still postulated.

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-36 15.5.1.12 Ground Water Protection and Corrosion The ground-water level external to the membrane protection of the exterior surfaces of the containment structure will be kept below the top surface of the foundation mat by pumps, as described in Section 15.5.1.3.

If water penetrates or otherwise circumvents the membrane, it drains to a layer of porous concrete directly below the mat and above the membrane. This 4-inch-thick layer of porous concrete serves as a horizontal drain under the entire structure. The porous layer is vented by two 4-inch-diameter pipes that extend from the underside of the mat into a subsurface cubicle adjacent to the outside of the containment structure. This cubicle is inside the waterproof membrane.

Access is provided by a concrete shaft from ground level. The 4-inch-diameter vent pipes are installed to discharge water to the floor of the cubicle at a level three feet below the mat liner; thus, flooding of the cubicle would have to occur before any hydrostatic head would be applied to the steel liner. A water level alarm is installed in the cubicle, and pumps are used as necessary to remove the water. Vertical drainage to the base of the mat is aided by three vertical inspection shafts, and various tunnels and cubicles located adjacent to the exposed exterior face of the concrete containment wall, in which the concrete is exposed.

Cathodic protection is not provided, since adequate corrosion protection of the embedded reinforcing is otherwise provided. Research by the National Bureau of Standards and other references indicates that cathodic currents damage the bond between the reinforcing steel and concrete. This bond softening is due to the gradual concentration of sodium and potassium ions.

In time, the alkali concentration becomes strong enough to attack the steel.

The surface of the steel liner in contact with concrete is not subject to corrosion because of the alkaline nature of the concrete. The interior exposed surface of the liner is protected by one coat of inorganic zinc silicate primer with one top coat of epoxy enamel. These materials were used during construction. The repair coatings currently used are selected in accordance with administrative controls. No other protective coatings or insulation are considered necessary.

15.5.1.13 Testing and Inservice Surveillance 15.5.1.13.1 General The completed containment structure was tested for structural integrity by subjecting the structure to an air pressure test equal to 115% of the design pressure. The structure was first carefully surveyed, measured, and inspected for cracks prior to the test, and all measurements recorded. All measurements were related to an independent datum. The pressure was then raised in 10-psi increments to the 115% test pressure (52 psig) and held at that pressure for one hour.

Pressure was then reduced to complete the containment liner leak rate test described in Section 5.3.

During the 48-hour period, visual examination was made of the containment exterior surface for cracks and crack patterns as well as distortion.

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-37 Visual and instrumented observations at each pressure increment were made of the containment response during the test. Crack patterns were observed and their development noted.

Temperature, barometric pressure, and weather conditions were recorded hourly during the test period.

A further detailed dimensional survey was made of the structure on completion of the tests to record recovery of the structure.

15.5.1.13.2 Test Instrumentation Instrumentation was designed to provide control and information on containment response during the air pressure test. Measurements were made of the radial deflection of the containment wall at selected locations from the top of the mat to the spring line of the dome. Vertical deflections were measured at the top of the mat and at the top of the dome. Additional measurements were made around the equipment access hatch and in other areas where stresses were critical.

Strain gauges were attached to the steel liner to record strains at the junction with the mat liner, at mid-height, and at the spring line of the dome. Additional strain gauges were attached to the liner around the equipment access and personnel hatches.

Exterior visual observations, above grade, were obtained using engineers scales attached to the structure and read by transits placed nearby. Transit measurements were calibrated with independent datum points. Readings obtained by this method were considered accurate to within 0.10 inch.

Exterior deformations below grade were measured by linear variable differential transducers (LVDTs) mounted in the two pits provided for this purpose. Linear variable differential transducers recorded displacements in mils, which is an accuracy in excess of that required. Linear variable differential transducers were also used to measure displacements of the concrete rings surrounding the equipment access and personnel access hatches.

Electrical strain gauge rosettes and conventional strain gauges, reading in microinches per inch, were used to monitor strains in the liner. Since major inaccuracies with this type of gauge have resulted from inadequate installation techniques, particular attention was given to the technique used.

Redundancy of instrumentation was obtained by multiplicity of points at which measurements were made, so that loss or damage to any one station was not critical.

The range of strains and deformations expected at the 45-psig test pressure was as follows:

1. Maximum vertical elongation of the structure, not more than 1.5 inches.

2. Increase in containment diameter, not more than 1.4 inches.

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-38

3. Maximum width of new cracks or increase in existing cracks, not more than 0.03 inch per crack.

4. After containment pressure was reduced to atmospheric, the residual width of new cracks or the increased width of existing cracks, not more than 0.01 inch.

5. There was no visual distortion of the liner plate.

The containment structure remained in the elastic range during the pressure test, and there was permanent distortion in the liner or in the concrete once the pressure was reduced to atmospheric or below. However, it was fully expected that there would be small residual cracks in the concrete as a result of shrinkage in the concrete.

Under the test program outlined herein, all instruments and measuring devices were installed just prior to the test, and normal care and protection was adequate. Items damaged for any reason were readily replaced at the initiation of the test.

15.5.1.13.3 Comparison of Test Results The selection of a test pressure, which was 115% of design pressure, was based primarily on the fact that a similar reinforced concrete containment structure for Connecticut Yankee Atomic Power Plant has been tested and accepted at 115% of its design pressure; thus, a comparative case history of structural response has been created that permits valid comparison of similar design.

The selection of 115% test pressure also conforms to the ASME Boiler and Pressure Vessel Code,Section III, Nuclear Vessels, Subsection B, Requirements for Class B Vessels, paragraph N-1312(d). This relates to metal vessels that perform the same function as a reinforced concrete containment structure.

A comparison of stresses under 115% test pressure with those in the structure under incident conditions is given in Table 15.5-4. As a sensitivity analysis, the stresses associated with 125%

pressure are also included. Incident stresses shown result from incident pressure, dead load, loads due to temperature effects on the steel liner, and temperature gradients through the concrete.

Stresses resulting from earthquake combined with incident loads are shown separately.

Scaled load plots comparing moments, shears, tension, and deflections resulting from the structural proof test pressure with moments, shears, tension, and deflections due to the unfactored design incident conditions are shown in Figure 15.5-13. A comparison of the test load with the hypothetical incident load conditions should include a review of the load plots in Figure 15.5-2 Sheet 2. This shows the increase in moments, radial shear, hoop tension, vertical tension, and radial deflection (deformation) imposed on the structure by the incident and test load. It can be seen that the test load conditions exceed incident conditions in all cases, except that of radial deflection.

The distribution of stress varies between the structural elements under apparently similar load conditions because of the contradictory action of the containment steel liner. Under test

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-39 conditions, the steel liner was in a state of biaxial tension and gave considerable assistance to the reinforcing steel, particularly to the longitudinal reinforcing. Under incident conditions, the steel liner is subjected to a point where it is restrained in compression by the reinforcing steel. This effect is greater in the dome than in the cylindrical wall, due to the increased thickness of the dome liner and the shape factor of the dome.

This also provides an additional factor of safety against ultimate failure of the structure. In the event of excessive yield in the reinforcing steel, the liner will act as a tensile membrane that would assist the reinforcing steel. This assistance would be significant, since the liner will bring a considerable reserve of energy to bear, for which design credit has not otherwise been claimed.

The test pressure of 52 psig, based on 115% design pressure, created stresses equal to or greater than the incident stresses in the following critical areas:

1. Foundation mat, where test stresses are 30 to 40% above incident conditions.

2. Large access openings, such as equipment and personnel hatches, where test stresses are comparable with incident stresses.

It is recognized that the average stress levels attained under the test conditions in the principal longitudinal and circumferential steel are below these stresses resulting from incident conditions. This is considered acceptable when the test is associated with dimensional strain measurements, when such a test provides confirmation of structural continuity and structural ductility with the concrete cracked, and when the steel is shown to carry the load in tension according to design assumptions.

An analysis of the crack pattern of the concrete under test conditions confirms stress distribution in the structure, and also reveals areas of stress concentrations. In fact, a pattern of severe local cracking would indicate structural weakness more effectively than considerations of average stress levels.

Measured response of the structure, as indicated by increase in height, diameter, and degree of recovery, together with measurements of local deformations, is extremely important in predicting structural response to incident conditions. The structural response to the test pressure is of sufficient magnitude to allow simple direct measurements of deformations without the need for high-precision methods of measurement.

In summary, it is not possible to exactly duplicate incident stress conditions with a pressure test. An increase in the test pressure above 115% would only preserve and amplify the present stress anomalies, without furnishing more meaningful data. In addition, such a test would endanger or damage the container by seriously overstressing critical areas, or it would require a container design modification directed specifically to withstand the higher pressure test without proportionate improvement in withstanding the incident condition. Modification of the containment design to obtain closer test verification of structural integrity under the test pressure would require specific redesigning for test conditions of the critical areas in the foundation mat

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-40 and at the large openings. Such redesigning would not improve the capability of the containment structure to meet the incident load conditions. A design meeting both incident and test conditions is not considered practical in this type of containment design.

The 115% pressure test provided a valid test of all criteria areas with stresses equal to or greater than incident conditions; in less critical areas, the pressure test provided sufficient information to permit a reliable evaluation of the complete structural response under incident conditions.

The average anticipated crack width at the 45-psig test pressure was 0.015 inch.

A rectangular crack pattern was anticipated, with vertical cracks spaced 12 to 15 inches on centers, and horizontal cracks spaced approximately 2 feet on centers. Horizontal crack spacing was primarily controlled by the horizontal construction joints.

The average crack width was related to the anticipated increase in containment diameter, the anticipated vertical elongation of the structure, and the crack spacing. It was assumed that the total containment extension was equal to the sum of the number of cracks multiplied by the average crack width in each direction.

Maximum summer temperature and minimum winter temperature difference is approximately 95°F. Annual average temperature variation is 40°F at the station site.

During unit operation, the annual maximum thermal cycling temperature variation is approximately 45°F.

Ambient temperature variations of this magnitude, +20°F, or even the extreme +45°F, will not reopen by any significant amount the crack pattern created in the structure by the test pressure of 45 psig.

The width of thermal cycling cracks was significantly less than the 0.010 allowed for exterior members by ACI 318.

The stresses given in Table 15.5-4 are the results obtained from computer programs referred to in the following sections:

• Section 15.5.1.5 Numerical Analysis of Unsymmetrical Bending of Shells of Revolution

• Section 15.5.1.4 Container Vessel Seismic Analysis

• Section 15.5.1.5 Flat Circular Mat Foundations for Nuclear Secondary Containment Structures

• Section 15.5.1.7 Nuclear Containment Structure Access Opening - Stone & Webster Computer Program

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-41 At large openings, the stresses due to thermal load were obtained by converting the thermal effect to a pressure equivalent, as described in Section 15.5.1.5.

Since all of the shears in the wall and dome were taken by the reinforcing, the effects of shrinkage and creep are not included.

15.5.1.13.4 Inservice Surveillance Tests Periodic structural testing of the containment structure is not planned, since it would provide no more information on the containment structure capability than that obtained from the initial test. In fact, periodic testing would cumulatively damage the concrete in the structure to the point where the test itself would be the major cause of structural deterioration.

The inservice stress and environmental conditions are not of a nature or magnitude such that any significant deterioration of the reinforcing steel or concrete could reasonably be expected, and periodic testing for structural purposes could be duplicated if at any time further tests were required. The minimum test level required to verify continued structural integrity would be no less than the 115%, or 52-psig initial test pressure.

Periodic inspection of the steel liner is accomplished by a type A leak rate test in accordance with 10 CFR 50 Appendix J. All welded joints and all penetrations of the liner are designed for periodic halogen gas testing.

In summary, no basis exists for attempting to develop structural performance information from leak rate tests conducted at moderate pressures.

15.5.2 Reactor Pressure Vessel Head Replacement Project (Applicable to Unit 1 and Unit 2)

The Reactor Pressure Vessel (RPV) Head Replacement Project created and restored a construction opening in the reactor containment structure in accordance with administrative procedures and the design control program. The opening was used to facilitate the movement of original and replacement RPV heads in and out of the reactor containment structure. The opening was restored to meet the original design bases of the containment structure.

15.5.2.1 Codes and Specifications ACI 318-63 is the design code for the restored containment structure. The restored structure meets all applicable design loads and load combinations required by ACI 318-63.

Concrete placement, curing, and repair were in accordance with ACI 301-99 with the incorporation of Hot Weather Concreting per ACI 305R-99, as appropriate, or with the incorporation of Cold Weather Concreting per ACI 306.1/ACI 306R, as appropriate. The use of ACI 301-99 is in accordance with Section 2.2 of ANSI N45.2.5-74.

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-42 Concrete mix proportioning was per ACI 211.1-91 (reapproved 1997) in accordance with Table A of ANSI N45.2.5-74.

Bechtel specifications (References 11-18) address:

• reinforcing steel procurement, testing, and placement

• Cadweld reinforcing steel splices procurement, testing, and installation

• concrete mix design, testing and placement

• structural steel and materials procurement 15.5.2.2 Liner Restoration The cut section of the containment liner plate was rewelded to the liner plate with a full penetration weld. The weld was tested to ensure no leakage. In addition, the full penetration weld was covered by a seal welded leak chase channel to facilitate testing.

Replacement material was purchased for the liner plate, Nelson studs, and leak chase channels. The Nelson studs, and leak chase channels were used for the reinstallation of the plate and the leak chase channel system. Reference 18 requires the liner plate material to be ASTM A-516-Grade 60 (or better), fine-grained and normalized.

15.5.2.3 Reinforcing Steel Restoration The reinforcing steel bars cut during the creation of the opening were re-installed using Cadweld splices or welding, as required, in accordance with References 14, 15, and 19.

Reinforcing steel bars that were damaged during the creation of the opening were repaired in accordance with References 13 and 19 or were replaced with reinforcing steel procured in accordance with Reference 12. New N18 reinforcing steel used for containment wall restoration conforms to either ASTM A615 Grade 60 and/or ASTM A706 Grade 60, and meets or exceeds the additional elongation and chemical composition requirements described in Section 15.5.1.9.3 for the containment structure existing reinforcing steel.

In lieu of the Cadweld testing protocol, described in Section 15.5.1.10, which was used during original construction, Cadweld testing was performed in accordance with Dominions Operational Quality Assurance Program Topical Report which includes:

• In-process testing of Cadweld splices in accordance with Subsubparagraph CC-4333.5.2 of ASME B&PVC Section III Division 2 (1995 Edition, 1996 Addenda).

• Cadweld Splice System Qualification in accordance with Subparagraph CC-4333.2 of ASME B&PVC Section III Division 2 (1995 Edition, 1996 Addenda).

• Cadweld Operator Qualification in accordance with Subparagraph CC-4333.4 of ASME B&PVC Section III Division 2 (1995 Edition, 1996 Addenda).

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-43

• Cadweld Testing Frequency in accordance with Subsubparagraph CC-4333.5.3 of ASME B&PVC Section III Division 2 (1995 Edition, 1996 Addenda).

To minimize the size of the construction opening, the Cadweld splice locations were not staggered as described in Section 15.5.1.10. Section 805 of ACI 318-63 does not require staggered Cadweld splices if the splice can develop in tension at least 125 percent of the specified yield strength of the reinforcing steel bar. The minimum acceptance criteria for the Cadweld splice testing in Reference 15 is that the minimum tensile strength of each sample tested shall be equal to or exceed 125 percent of the yield strength of the reinforcing steel bar. Also, the splicing scheme for the RPVH Replacement Project construction opening is similar to that used during the closure of the original construction opening.

15.5.2.4 Concrete Restoration As discussed in Dominions Operational Quality Assurance Program Topical Report commits to ANSI N45.2.5-74 (with clarifications) for satisfying the quality assurance requirements for installation, inspection, and testing of structural concrete during the operational phase of Surry Power Station. Section 2.2 of ANSI N45.2.5-74 requires that the installation, inspection, and test activities be performed in accordance with the latest codes. Tables A and B of ANSI N45.2.5-74 provide the requirements for the qualification and in-process testing of the concrete ingredients and concrete mix.

The concrete was replaced and the restored structure tested in accordance with ASME B&PVC Section XI, Articles IWL 4000 and IWL 5000, respectively. In accordance with the guidance of Table A of ANSI N45.2.5-74 concrete mix design is based on ACI 211.1-91 (reapproved 1997). The activities associated with placement of concrete were performed in accordance with References 11 and 17, which meet the requirements of ACI 301 and ANSI N45.2.5-74. In-process sampling, testing, and acceptance requirements for all repair material were in accordance with Table B of ANSI N45.2.5-74. Reference 11 provides the testing frequencies, sampling and testing standards, and acceptance criteria for concrete ingredients and concrete mix. The concrete had a minimum 5-day strength of 3000 psi.

The water used for the concrete mix was evaluated in accordance with the requirements of AASHTO T-26, as specified in Table A of ANSI N45.2.5-74. The water testing and acceptance criteria included in Reference 11 required that the water used during the restoration was free of harmful levels of contaminants.

The cement used in the new concrete was Type II Low Alkali (as defined in ASTM C 150).

For RPV Head Replacement Project, the restoration of the containment wall used size 57 (25 mm to 4.75 mm) coarse aggregate due to the limited size of the opening and the use of pour ports/bird mouths for concrete placement. Both fine and coarse aggregates were tested in accordance with the requirements of ANSI N45.2.5-74 to ensure acceptable physical

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-44 characteristics and that they were free of harmful levels of alkali reactivity and deleterious substances (acceptance criteria are defined in ASTM C 33).

Admixtures used to modify the concrete mix properties met the requirements of ASTM standards and were used in accordance with the manufacturers written procedures and applicable ACI standards (primarily ACI 211.1-91 (reapproved 1997) for mixing and ACI 301-99 for placement). Reference 16 prohibited the use of admixtures with chlorides. Uniformity of admixture lots was verified with Infrared Spectrophotometry in accordance with Table B of ANSI N45.2.5-74.

In its ready mix state, the new concrete had an air content of 4.5% (+/- 1.5%) at the point of placement. This is consistent with Table 6.3.3 of ACI 211.1-91 (reapproved 1997) for the maximum aggregate size being used in the concrete mix (1" for Size No. 57 coarse aggregate) and air-entrained concrete.

A water-reducing admixture was utilized in the concrete resulting in a maximum slump of 8 1/2 inches at point of placement based on the footnote to Table 6.3.1 of ACI 211.1-91 (reapproved 1997), which approves higher concrete slump (than the recommended 1 inch to 4 inch slump) when chemical admixtures are used provided that there are no signs of segregation or excessive bleeding.

15.5.2.5 Post Modification Testing The nondestructive examination of the containment liner was in accordance with Safety Guide 19, Nondestructive Examination of Primary Containment Liners with the following changes: after vacuum box testing of the liner seam weld and installation of the channel, the channel to liner weld was tested by a static pressure test (decay test) with an acceptance criteria of zero leakage. Soap bubble testing was used to identify leakage. Leaking areas of the joint were repaired and retested. In addition, following the containment building pressure test, the channel was re-pressurized and an as-found LLRT, meeting ANS 56.8-1994 requirements, was performed.

Prior to placing the containment structure in-service, a containment pressure test that bounds the calculated peak containment internal pressure was performed in accordance with IWL Article 5000 of the ASME B&PVC Section XI. The surface of the replacement concrete at the temporary construction opening was examined in accordance with IWL-5250 prior to pressurization, at test pressure, and following completion of pressurization.

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-45

15.5 REFERENCES

1. G. N. Bycroft, Forced Vibrations of a Rigid Circular Plate on a Semi-Infinite Elastic Space and on an Elastic Stratum, Philosophical Transactions, Royal Society, London, Series A, Vol. 248, pp. 327-368.

2. Karl Terzaghi, Evaluation of Coefficients of Subgrade Reaction, Geotechnique, Vol. 5, pp. 297-326, 1955.

3. B. O. Hardin and W. L. Black, Vibration Modulus of Normally Consolidated Clay, Symposium on Wave Propagation and Dynamic Properties of Soils, University of New Mexico, 1967.

4. Stone & Webster Engineering Corporation, Nuclear Containment Structure Access Opening.

5. B. Budiansky and P. Radkowski, Numerical Analysis of Unsymmetrical Bending of Shells of Revolution, AIAA Journal, August 1963.

6. S. Gere and S. Timoshenko, Theory of Elastic Stability, second edition.

7. J. N. Goodier and S. Timoshenko, Theory of Elasticity, second edition.

8. Stone & Webster Engineering Corporation, Report on Pressure Testing of Reactor Containment for Connecticut Yankee Atomic Power Plant, Connecticut Yankee Atomic Power Company, Haddam, Connecticut, 1967.

9. D. A. Davenport, Penetration of Reactor Containment Shells, Nuclear Safety, Vol. 2, No. 2, December 1960.

10. Letter dated March 30, 1999, Serial No.99-134, From Virginia Power to the NRC, Supplemental Response to Generic Letter 96-06.

11. Bechtel Specification 24841-120-C-101, Revision 6, Technical Specification for Material Testing Services, April 25, 2003. (Unit 1)

Bechtel Specification 24841-120-C-101, Revision 7, Technical Specification for Material Testing Services, December 3, 2003. (Unit 2)

12. Bechtel Specification 24841-120-C-303, Revision 2, Technical Specification for Purchase of Reinforcing Steel, February 28, 2003.

13. Bechtel Specification 24841-120-C-304, Revision 2, Technical Specification for Installation of Reinforcing Steel (Rebars), February 28, 2003.

14. Bechtel Specification 24841-120-C-309, Revision 1, Technical Specification for Purchase of Cadweld Rebar Splices, February 28, 2003.

15. Bechtel Specification 24841-120-C-310, Revision 2, Technical Specification for Installation of Cadweld Rebar Splices, February 28, 2003.

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-46

16. Bechtel Specification 24841-120-C-321, Revision 9, Technical Specification for Purchase of Ready Mix Concrete Qualified as Safety-Related, May 20, 2003.

17. Bechtel Specification 24841-120-C-322, Revision 3, Technical Specification for Placement of Ready Mix Concrete Qualified as Safety-Related, February 28, 2003.

18. Bechtel Specification 24841-120-C-502, Revision 3, Technical Specification for Purchase of Non-Safety Related and Safety Related Structural Steel and Materials, February 28, 2003.

19. Special Processes Manual for Surry Nuclear Power Station RPV Head Replacement Project, Revision 3, Bechtel Job 24841, May 12, 2003. (Unit 1)

20. Special Processes Manual for Surry Nuclear Power Station RPV Head Replacement Project, Revision 4, Bechtel Job 24841, September 25, 2003. (Unit 2) 15.5 REFERENCE DRAWINGS The list of Station Drawings below is provided for information only. The referenced drawings are not part of the UFSAR. This is not intended to be a complete listing of all Station Drawings referenced from this section of the UFSAR. The contents of Station Drawings are controlled by station procedure.

Drawing Number Description

1. 11448-FM-1A Machine Location: Reactor Containment, Elevation 47'- 4"

2. 11448-FM-1B Machine Location: Reactor Containment, Elevation 18'- 4"

3. 11448-FM-1C Machine Location: Reactor Containment, Elevation 3'- 6"

4. 11448-FM-1D Machine Location: Reactor Containment, Elevation 27'- 7"

5. 11448-FM-1E Machine Location: Reactor Containment; Sections A-A, E-E,

& Z-Z

6. 11448-FM-1F Machine Location: Reactor Containment; Sections B-B, X-X,

& Y-Y

7. 11448-FM-5A Arrangement: Auxiliary Building

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-47 Table 15.5-1 CONTAINMENT STRUCTURAL LOADING CRITERIA Case Loading Combination Required Load Capacity of Structure 1 Operating plus DBA = (1.0 +/- 0.05)D + 1.5P + 1.0 (T + TL) 2 Operating plus DBA plus = (1.0 +/- 0.05)D + 1.0P + 1.0 (T + TL) + 1.5E operating-basis earthquake 3 Operating plus DBA plus = (1.0 +/- 0.05)D + (1.25P) + (T' + TL') + 1.0HE design-basis earthquake 4 Operating plus 1.25 DBA and 1.25 = (1.0 +/- 0.05)D + (1.25P) + (T' + TL') + 1.25E operating basis earthquake 5 Operating plus tornado loading = (1.0 +/- 0.05)D + 1.0T' + 1.0C Legend DBA - Design-basis accident.

C - Load due to negative pressure and horizontal wind velocity resulting from tornado and missiles. For description of tornado, refer to Section 15.2.3.

D - Dead load of structure and contents including effect of earth and hydrostatic pressures, buoyancy, ice and snow loads. To provide for variations in the assumed dead load, the coefficient for the dead load components is adjusted by +/-5% as indicated in the above formulas to provide the maximum stress levels.

P - Pressure load from DBA. Pressure for containment design is 45 psig.

T - Load due to maximum temperature gradient through the concrete shell and mat based on temperature associated with 1.5 DBA pressure.

T' - Load due to maximum temperature gradient through the concrete shell and mat based on normal operating temperature.

TL - Load exerted by the exposed liner based upon temperature associated with 1.5 times DBA pressure.

TL'- Load exerted by the exposed liner based upon temperatures associated with 1.25 times DBA pressure.

T - Load due to maximum temperature gradient through the concrete shell and mat based upon temperature associates with 1.0 times DBA pressure.

TL - Load exerted by the exposed liner based upon temperature associated with 1.0 times DBA pressure.

E - Operating-basis earthquake loading. Based on a ground acceleration of 0.07g horizontally at zero period and a damping factor of 5%. For description of the operating-basis earthquake, refer to Section 2.5.

HE - Design-basis earthquake loading. Based on a ground acceleration of 0.15g horizontally at zero period and a damping factor of 10%. For description of the design-basis earthquake, refer to Section 2.5.

Note: Normal wind loadings replace earthquake loads where they exceed earthquake loadings.

Normal wind or tornado loads are not considered coincident with earthquake loads.

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-48 Table 15.5-2 CAPACITY REDUCTION FACTOR FOR CONCRETE Member Reduction Factor Tension and flexure 0.90 Diagonal tension, bond and anchorage 0.85 Table 15.5-3 MISSILE DIMENSIONS AND WEIGHTS REQUIRED TO PENETRATE PLATE OF VARYING THICKNESSES Material Missile Diameter, in.

1 2 3 4 5 Reactor Containment Liner Plate, 3/8 in.

Weight, lb 21.1 42 64 85 106 Length, in. 95 48 32 24 19 4 in. Sch. 160 pipe or 0.531 in wall thickness Weight, lb 40.2 80.3 120.5 160.6 200.8 Length, in. 181 90 60 45 36 6 in. Sch. 160 pipe or 0.718 in. wall thickness Weight, lb 68.8 137.5 206.3 275.0 343.8 Length, in. 309 155 103 78 62 8 in. Sch. 160 pipe or 0.906 in. wall thickness Weight, lb 109.0 218.0 327.0 436.0 545.0 Length, in. 514 245 164 123 99 10 in. Sch. 160 pipe or 1.125 in. wall thickness Weight, lb 176.0 325.0 528.0 704.0 880.0 Length, in. 790 395 264 198 159

Revision 52Updated Online 09/30/20 Table 15.5-4 COMPARISON OF STRESSES UNDER TEST PRESSURE WITH STRESSES UNDER INCIDENT CONDITIONS AND EARTHQUAKE PLUS INCIDENT CONDITIONS Earthquake Plus 115% Test Stress psi 125% Test Stress psi Incident Stress, psi Incident Stress, psi (52 psig) (57 psig)

Top bars at cylinder wall 25,700 26,500 33,500 38,700 Bottom bars near mat center 26,300 30,500 29,000 38,700 Cylinder WaIl (approximately mid-height)

Circumferential reinforcing Inner 22,300 22,300 21,000 22,800 Outer 28,600 28,600 20,000 21,700 Longitudinal reinforcing Inner 19,800 21,600 9200 10,000 At base of wall 12,900 19,300 16,400 19,300 Outer 27,900 29,650 9200 10,000 Diagonal reinforcing 28,300 33,600 15,500 16,900 Diagonal (radial) shear reinforcing at base of wall 15,700 16,500 20,200 23,200 SPS UFSAR Dome Radial reinforcing Inner 28,900 28,900 14,200 15,500 Outer 34,500 34,500 13,800 15,000 Circumferential reinforcing Inner 28,900 28,900 14,200 15,500 Outer 34,500 34,500 13,800 15,000 Large Openings Equipment access hatch 32,000 33,500 31,300 33,700 15.5-49 Personnel access hatch 30,200 31,700 30,700 34,300

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-50 Figure 15.5-1 REACTOR CONTAINMENT WATERPROOFING

Revision 52Updated Online 09/30/20 Figure 15.5-2 (SHEET 1 OF 3)

CONTAINMENT LOADING PLOT SPS UFSAR 15.5-51

Revision 52Updated Online 09/30/20 Figure 15.5-2 (SHEET 2 OF 3)

CONTAINMENT LOADING PLOT SPS UFSAR 15.5-52

Revision 52Updated Online 09/30/20 Figure 15.5-2 (SHEET 3 OF 3)

CONTAINMENT LOADING PLOT SPS UFSAR 15.5-53

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-54 Figure 15.5-3 REINFORCING DETAILS EQUIPMENT ACCESS HATCH OPENING

Revision 52Updated Online 09/30/20 Figure 15.5-4 REINFORCING DETAILS SECTIONS THROUGH RING BEAM EQUIPMENT ACCESS HATCH SPS UFSAR 15.5-55

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-56 Figure 15.5-5 REINFORCING DETAILS PERSONNEL HATCH OPENING (Figure 15.5-6)

Revision 52Updated Online 09/30/20 Figure 15.5-6 REINFORCING DETAILS SECTIONS THROUGH RING BEAM PERSONNEL HATCH SPS UFSAR 15.5-57

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-58 Figure 15.5-7 WALL AND MAT JOINT

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-59 Figure 15.5-8 (SHEET 1 OF 2)

SECTION-TYPICAL BRIDGING BAR

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-60 Figure 15.5-8 (SHEET 2 OF 2)

SECTION-TYPICAL BRIDGING BAR

Revision 52Updated Online 09/30/20 Figure 15.5-9 TYPICAL ELECTRICAL PENETRATION SLEEVE WITH FLANGES SPS UFSAR 15.5-61

Revision 52Updated Online 09/30/20 Figure 15.5-10 (SHEET 1 OF 2)

TYPICAL PIPING PENETRATIONS SPS UFSAR 15.5-62

Revision 52Updated Online 09/30/20 Figure 15.5-10 (SHEET 2 OF 2)

TYPICAL PIPING PENETRATIONS SPS UFSAR 15.5-63

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-64 Figure 15.5-11 PERSONNEL HATCH ASSEMBLY

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-65 Figure 15.5-12 TYPICAL LINER DETAILS

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-66 Figure 15.5-13 CONTAINMENT LOADING PLOT

Revision 52Updated Online 09/30/20 Figure 15.5-14 MACHINE SHOP REPLACEMENT FAC.; SOUTH ELEVATION SPS UFSAR 15.5-67

Revision 52Updated Online 09/30/20 Figure 15.5-15 MACHINE SHOP REPLACEMENT FAC.; SITE PLAN SPS UFSAR 15.5-68

Revision 52Updated Online 09/30/20 Figure 15.5-16 MACHINE SHOP REPLACEMENT FAC.; EAST/WEST ELEVS .

SPS UFSAR 15.5-69

Revision 52Updated Online 09/30/20 SPS UFSAR 15.5-70 Intentionally Blank

Revision 52Updated Online 09/30/20 SPS UFSAR 15.6-1 15.6 OTHER CLASS I STRUCTURES Class I structures other than the reactor containment structure are listed in Table 15.2-1. The major structures include the auxiliary building, control room area, including switchgear and relay rooms; fuel building; emergency diesel-generator rooms; containment auxiliary structures that contain main steam and feedwater isolation valves, recirculation spray and low-head safety injection pump cubicles, auxiliary steam generator feed pump cubicle, and safeguards ventilation room; and circulating water intake structures, including the high-level canal.

The fuel building, the main steam and feedwater isolation valve section of the containment auxiliary structures, and the refueling water storage tanks are supported on reinforced-concrete mats on concrete-filled steel pipe piles. All other structures are soil-supported on reinforced-concrete mats or spread footings. All Class I structures designed to meet tornado missile criteria, as listed in Table 15.2-1, are enclosed with heavily reinforced, 2-foot-thick concrete walls and roof slabs with all openings shielded against missiles.

Class I structures are designed to resist the operating-basis earthquake without exceeding allowable working stresses, where allowable stresses are one-third above the normal applicable code normal working stress. For concrete structures, a 5% critical damping function is assumed, and the accelerations selected from the acceleration response spectrum curves are considered in conjunction with the natural frequency of each structure. A check has been made to ensure that collapse-type failures will not occur under the design-basis earthquake. For this condition, a 10%

damping factor is assumed for concrete structures, and stresses are limited to not more than 120%

of the minimum yield point stress. In practice, the controlling feature of the design of these structures was to the satisfaction of the operating-basis earthquake requirements with the limited 5% damping factor.

The high-level intake canal is formed by excavating to Elevation +5 ft. from an average grade of approximately 35 feet. Earth fill dikes constructed on either side of the canal bring the finished height to Elevation +36 ft. throughout the length of the canal. The interior surfaces of the canal are lined with a 4.5-inch-thick reinforced-concrete slab. Under drains and pressure relief valves are provided to prevent uplift of the concrete liner by unbalanced hydrostatic pressure.

15.6.1 Other Structures All other structures are designed to adequately support all dead, live, and wind loads. Where necessary, subsurface walls and slabs are designed to resist the horizontal component of the soil with applicable surcharge and hydrostatic pressures.

Structural steel design conforms to the 1963 issue of the Specification for the Design, Fabrication and Erection of Structural Steel for Buildings of the American Institute of Steel Construction, except as noted herein. Plastic design methodology, in accordance with Part 2 of the 1969 issue of the Specification for the Design, Fabrication and Erection of Structural Steel for Buildings of the American Institute of Steel Construction, has been used to modify the main bents

Revision 52Updated Online 09/30/20 SPS UFSAR 15.6-2 of the Fuel Building steel superstructure. All concrete work is designed in accordance with the Building Code Requirements for Reinforced Concrete, serial designation 318-63 of the American Concrete Institute. Access and egress requirements, as well as fire ratings of walls and floor systems, satisfy the requirements of the Basic Building Code of the Building Officials Conference of America, 1966 issue.

Under the design-basis accident loading, the allowable stresses do not exceed 90% of the minimum yield strength of the structural steel. From mill test reports, the yield strength of structural steel is 42,000 psi, with an ultimate strength of 63,000 psi. Using a minimum yield of 36,000 psi for A36 steel, the design-allowable stress is 90% of 36,000 = 32,400 psi.

32, 400 Design-allowable stress for structural steel is ------------------ = 51.5% of the ultimate strength.

63, 000 Tests on special reinforcing steel with a minimum yield of 50,000 psi have resulted in yield strength of 55,500 psi, with an ultimate strength of 90,000 psi; with a design-allowable stress of 90% of minimum yield, the design-allowable stress is 0.9 x 50,000 = 45,000 psi.

45, 000 Design-allowable stress on reinforcing is ----------------- = 50% of ultimate strength.

90, 000 Concrete continues to increase in strength beyond the 28-day strength of 3000 psi. The Bureau of Reclamation Concrete Manual indicates that Type II cement concrete can be expected to increase in strength approximately 30% in 1 year from the 28-day strength.

Approximate 28-day strength for 3000-psi concrete from test reports = 3800 psi Design allowable 85% of 3000 psi = 2500 psi Ultimate strength in 1 year = 1.3 x 3800 = 4950 psi 2500 Design allowable is ----------- = 51% of ultimate strength in 1 year.

4950 The above figures show that, for structures designed for the design-basis accident loading, structural steel and reinforced concrete are designed at approximately 50% of their ultimate strength. In the design of concrete structural members under design-basis accident conditions, concrete strength is not the controlling factor.

Allowable soil bearing values for foundations are determined from the soil boring logs and the results of triaxial shear tests of the soil. Applicable factors of safety are applied to the test results.

15.6.2 Reactor Coolant System Supports The reactor coolant system includes the reactor vessel, three steam generators, three reactor coolant pumps, and a pressurizer for each unit. Structures are provided to support these heavy vessels and equipment, and to ensure system integrity during normal operation and design-basis accident conditions.

Revision 52Updated Online 09/30/20 SPS UFSAR 15.6-3 The primary equipment supports of the reactor coolant system are designed to withstand the design-basis earthquake acting simultaneously with an instantaneously applied pipe break. The original configuration of the Reactor Coolant System (RCS) equipment supports included ten large-bore (12-inch diameter) hydraulic snubbers per loop to carry the loads from postulated reactor coolant system, main steam line and feedwater pipe ruptures.

Subsequently, studies to address Unresolved Safety Issue A-2 (effects of asymmetric pressure loads resulting from PWR primary loop ruptures) concluded that the probability of rupture of the primary coolant loop is extremely small, and that the presence of a pipe crack could be detected by leakage well before the crack grew to critical size which would cause rupture.

These leak-before-break analyses, documented in References 1 and 2, were submitted to the NRC on behalf of the Westinghouse Owners Group, which included Surry Power Station.

NRC Generic Letter 84-04, Safety Evaluation of Westinghouse Topical Reports Dealing with Elimination of Postulated Pipe Breaks in PWR Primary Loops, provided the NRC staff safety evaluation concluding that, provided certain specific conditions are met, an acceptable technical basis exists so that asymmetric pressure loads resulting from pipe breaks in the reactor coolant system primary loop need not be considered as a design basis for the reviewed plants. The plant-specific conditions were a limitation on the maximum bending moments in the primary loop piping for normal operating and seismic loadings, and the existence of an adequate reactor coolant leakage detection system. The affected plants were encouraged to submit requests for partial exemptions from General Design Criterion 4 (GDC-4), to permit elimination of pipe rupture restraints required to protect against these previously postulated breaks.

Because a number of the large-bore snubbers served primarily as pipe rupture restraints, Surry proceeded with an exemption request to allow elimination of 6 of the 10 snubbers per loop, based on application of leak-before-break in accordance with Generic Letter 84-04. (The other 4 large-bore snubbers on the upper steam generator support were required for lateral loads due to a postulated longitudinal split of the main steam line.) The reactor coolant loop system was re-evaluated with all snubbers on the steam generator lower support and the reactor coolant pump supports eliminated, to assure that the conditions of pressure, deadweight, thermal, seismic, and remaining pipe rupture effects, would not result in unacceptable stress levels or factors of safety.

Largely independent analyses were performed by Westinghouse and Stone & Webster in accordance with the original division of design responsibilities. Interface force allowable limits at NSSS boundaries were assured and support design load interfaces were reviewed for acceptance.

The exemption request to allow elimination of 18 snubbers per unit was filed with the NRC on November 5, 1985 (Reference 3). The detailed technical basis (Reference 4) provided separate attachments addressing load evaluation, leakage detection, and net safety balance. The proposed design changes were discussed with the NRC and resulting NRC concerns were addressed (References 5 & 6). The GDC-4 limited scope revision (Reference 7) was subsequently published (effective May 12, 1986) permitting the use of leak-before-break technology to justify elimination of the dynamic effects of primary loop breaks from the design basis of PWRs. With

Revision 52Updated Online 09/30/20 SPS UFSAR 15.6-4 the publication of the notice of revision to GDC-4, a license amendment to the plant design basis was requested to implement the snubber reductions under the new rule (Reference 8). By letter dated June 16, 1986, the NRC approved the license amendment (#108) for both units (Reference 9). The snubber eliminations were implemented by Design Changes 85-04-1 and 86-12-2.

Subsequently, the NRC issued Generic Letter 87-11, Relaxation of Arbitrary Intermediate Breaks which provided the revised Mechanical Engineering Branch position eliminating the need to postulate Arbitrary Intermediate Breaks. As stated in the generic letter, the elimination of Arbitrary Intermediate Breaks allows elimination of the associated pipe whip restraints and jet impingement shields. Because the stresses in the main steam lines inside containment are well below the stress criteria for required mandatory intermediate breaks, the only breaks which need to be postulated are terminal end breaks which do not apply lateral loads to the steam generator.

Therefore, the governing lateral loads on the steam generator become those imposed by the main feedwater break, which are low enough that only a single snubber in each pair will be required to carry the load. Analyses were performed by Westinghouse and Stone & Webster similar to those performed for the earlier large-bore snubber reduction to ensure piping stress levels and component factors of safety were acceptable. In addition, it was necessary to verify that the basis for the previous license amendment based upon leak-before-break of the primary loop remained valid. Comparison of results with the lower large-bore snubber of each pair removed as a result of eliminating the main steam line split vs. the results in the previous leak-before-break submittals, confirmed that there were no significant reduction in margins of safety. Therefore, elimination of the lower large-bore snubber of each pair does not compromise the bases for the previous leak-before-break analysis, namely:

1. The loading on the primary loop piping is still enveloped by the generic analyses submitted on behalf of the A-2 Owners Group, and accepted by the NRC staff in Generic Letter 84-04, and specifically for Surry by NRC letter dated June 16, 1986; and

2. The reactor coolant system equipment, piping, and supports continue to have acceptable margins of safety under licensed loading conditions other than the now-eliminated ruptures of the primary loop piping and Arbitrary Intermediate Break of the main steam lines.

15.6.2.1 Design Basis All supports in the reactor coolant system are designed to withstand the design-basis earthquake acting simultaneously with an instantaneously applied pipe break. As discussed above, it is no longer necessary to consider the dynamic effects of a postulated rupture of the primary reactor coolant loop. However, single ruptures are postulated to occur in either the pressurizer surge or other reactor coolant branch lines, the main steam piping, or the main feedwater piping. In general, two types of piping failures are considered: a double-ended rupture, or a longitudinal rupture on either the horizontal or vertical axis of the pipe. The longitudinal rupture area was taken to be equal to the area of the double-ended rupture for these piping failures. Stresses in the main steam piping inside containment have been reviewed and, in

Revision 52Updated Online 09/30/20 SPS UFSAR 15.6-5 accordance with NRC Generic Letter 87-11, are sufficiently low that no intermediate break need be postulated. Only terminal end breaks need be postulated; in accordance with Generic Letter 87-11, it is not necessary to postulate longitudinal rupture at terminal end breaks.

Therefore, only vertical loads are considered to be applied to the steam generators due to a postulated main steam line break. For all postulated breaks, the pipe rupture loads are combined with design-basis seismic loads by the square-root-sum-of-squares (SRSS) method.

The peak value of the pipe thrust for any of the main steam piping breaks considered is approximately 620,000 lb; and the peak value for the pipe thrust for the pressurizer surge pipe break is approximately 195,000 lb. These thrust values are equal to P x A, the system pressure times rupture area.

For the pressurizer surge line break and other branch line breaks, the load versus time transients of these breaks are provided by a computer program that analyzes the shock wave initiated at the break as it passes through the complete piping loop. Results from this program are used as forcing functions in a structural dynamic program that results in the dynamic loadings of the supports. For the main steam line and feedwater line breaks, the dynamic forces were applied only at the steam generator nozzles, because the primary reactor coolant loop piping remains intact. The peak values of the pipe thrust for the postulated piping breaks were computed as Cr x P x A, the thrust coefficient times the system pressure times the rupture area. These values were used as the basis for developing conservative time-history forcing functions of the postulated breaks. In addition to the thrust loading, jet impingement effects were included as appropriate. For the main steam line vertical break, credit was taken in calculating the thrust loading for the flow area reduction of the flow restrictors installed at the nozzle during the steam generator replacement project; the jet impingement loading is not reduced by the flow reducers.

The time-history forcing functions are input into the structural dynamic analysis program that calculates maximum loadings on the supports. For each analyzed break, the maximum support loads are determined and then combined by SRSS summation with design-basis earthquake loads, and then added directly to the loadings due to normal operation. Combined stresses are maintained within 90% of the minimum yield point of the structural material used. For the RPV sliding foot supports, LOCA loads are developed using the AREVA methodology described in Chapter 14 Section 14.5.3.4.1.

All welding is in accordance with Section IX of the ASME Code, and all welds are examined by either radiographic, sonic, dye penetrant, or magnetic particle techniques, depending on the material and the state of stress at the weld.

The seismic restraints (snubbers) that are installed on the piping systems throughout the plant and that are required to protect the reactor coolant system or any other safety-related system are subject to operability and surveillance requirements contained within the Technical Specifications. Vepco has established a program and procedures for inspecting, testing, and maintaining snubbers in compliance with the Technical Specification requirements. A listing of all safety-related hydraulic and mechanical snubbers is maintained by Surry Power Station.

Revision 52Updated Online 09/30/20 SPS UFSAR 15.6-6 15.6.2.2 Description 15.6.2.2.1 Reactor Vessel Support The reactor vessel is supported by six sliding foot assemblies mounted on the neutron shield tank. The support feet are designed to restrain lateral and rotational movement of the reactor vessel while allowing thermal expansion. The neutron shield tank is a double-walled cylindrical structure that transfers the loadings to the heavy reinforced-concrete mat of the containment structure. The tank also serves to minimize gamma and neutron heating of the primary concrete shield, and to attenuate neutron radiation outside of the primary shield to acceptable limits (Section 11.3.2.1). The neutron shield tank assembly and material listing are shown on Figure 15.6-1.

Sliding support blocks mounted on top of the shield tank support the reactor vessel. These sliding support blocks permit radial thermal expansion of the reactor vessel, while preventing translation, rotation, or uplifting1. The support blocks are also designed to adjust to the correct height for plumbing the reactor vessel and for distributing the load properly among the six supports.

The loading conditions used in the analysis of the neutron shield tank were the simultaneous accelerations of a design-basis earthquake, the thrust forces exerted by the reactor vessel due to a double-ended reactor coolant pipe rupture,2 and the dead weight of the system and the tank itself, with stresses not to exceed allowable working stresses, and with no loss of integrity or function.

The neutron shield tank has been designed using the theory of a shell structure, dynamically loaded in both horizontal and axial planes, which results in meridional and circumferential stresses at all points along its length. The stresses in the tank were determined by using the methods and theories of Timoshenko for plates and shells, elasticity, and elastic stability. All membrane stress levels were held to the limits stated in Section VIII of the ASME Code; all membrane-plus-bending stresses were held to 90% of yield point.Section VIII of the ASME Boiler and Pressure Vessel Code was used as a guide in the fabrication and welding of the tank. A code stamp was not required, since this is not a code pressure vessel but a supporting structure for the reactor pressure vessel.

All material employed in the fabrication of the tank was new and conformed to ASTM Standards. The tank shell was constructed from ASTM A516, Grade 60, and the six sliding support blocks were made of maraged steel. The material has an NDT of -20° to -40°F. Drop weight tests were performed to determine the nil ductility transition temperature of the deposited weld metal in welding the ASTM A516, Grade 60 material with an NDT of -40°F. The maraged

1. It is no longer necessary for the RPV sliding foot supports to restrain uplift based on new LOCA loads developed as discussed in Section 14.5.3.4.1. The combined seismic and LOCA loads when added to dead weight are not sufficient to create uplift in the supports.

2. As discussed previously in this section, it is no longer necessary to consider the dynamic effects of a postulated rupture of the primary reactor coolant loop piping. However, pipe ruptures of reactor coolant branch lines must still be considered.

Revision 52Updated Online 09/30/20 SPS UFSAR 15.6-7 steel was ultrasonically tested for flaws to the quality level of MIL-1-8950. Flaws detected ultrasonically were verified by X-ray. Maraged steel pieces with verified flaws larger than 1/32 in.

were rejected. The maraged steel had a maximum hardness of 35 Rockwell C and a minimum grain size of 6, in accordance with ASTM E-112 and MIL-Std-430, Macrograph Standards for Steel Bars, Billets, and Blooms. The nonmetallic inclusion content for each billet was determined in accordance with ASTM E45. Fracture toughness tests were performed on maraged steel in accordance with ASTM Specification, Proposed Recommended Practice for Plane-Strain Fracture Toughness Testing of High-Strength Metallic Materials Using a Fatigue-Cracked Bend Specimen, Part 31, ASTM Standards.

All welds, where possible, were 100% radiographed in accordance with Paragraph UW-51 of the ASME Boiler and Pressure Vessel Code Section VIII, Division 1. Other welds that could not be radiographed were dye-penetrant checked at root pass, intermediate depths at half-inch increments, and the final pass, or magnetic-particle tested, in accordance with Appendices VI and VIII,Section VIII, ASME Boiler and Pressure Vessel Code. The surfaces of welds were ground to a surface condition suitable for the inspection procedure employed. Defects in welds were removed by chipping, grinding, or arc gouging until sound metal was reached. The resulting cavity was rewelded, employing an approved procedure.

After shop fabrication, the completed tank was subjected to a hydrostatic test of 15 psi, measured at the top of the tank. No water loss was observed for a 24-hour period. The tank was then leak tested with air at 5 psi gauge, applying soapsuds to all welds accessible from outside the tank. Leaks were repaired and the tank retested until no leakage was detected. All tests were recorded and certified. After installation of the neutron shield tank at the job site, the tank was hydrostatically retested.

15.6.2.2.2 Steam Generator Support The steam generator support consists of two (upper and lower) cast rings and associated suspension rods, lateral restraints and snubbers. The lower ring, which carries the steam generator weight, is suspended by means of three pipe columns. Hydraulic snubber cylinders and rigid lateral guides connect the upper casting to the steam generator cubicle structure to allow guided thermal expansion of the steam generator outward from the reactor during normal operation, while resisting movement during seismic and pipe break conditions. Due to the design of this support system (i.e., pin-ended connections at all member joints) lamellar tearing of the supports could not occur. The steam generator support assembly and material listing is shown on Figure 15.6-2. The supports do not have any heavy section intersecting member weldments.

The major materials used in the construction of the steam generator supports are listed in Table 15.6-1. The difference between the operating temperature and the NDT of the material ensures toughness and ductility of the steam generator supports under all operating conditions.

Welding associated with the supports was conducted in accordance with Section IX of the ASME code.

Revision 52Updated Online 09/30/20 SPS UFSAR 15.6-8 In addition to numerous inspections and tests carried out by the material suppliers and fabricators, all of the components for these supports were subject to inspection during fabrication and installation by Stone & Webster Engineering Corporation. All welds were subjected to examination by either the magnetic particle, liquid penetrant, or radiographic methods. The Vascomax 350 CVM and 300 CVM materials, and the A-352 grade LC3 casting materials were subjected to examination by the magnetic particle and ultrasonic methods. A visual inspection of a portion of these supports is required by the Technical Specifications.

The upper restraining ring is composed of two girth straps coupled together by studs to form a continuous ring. The studs are 1.25-inch 12UNF-2A Vascomax 18% Ni, maraging grade 350, with nickel cadmium coating. The nuts are 1.25-inch 12UNF-3B Vascomax 18% Ni, maraging grade 250, with Helicoil inserts. The studs and nuts are designed to minimize stress concentration during manufacture, and the studs are coated for environmental protection. The studs are pretensioned across a joint flange spacer block, which serves to reduce bending stresses in the studs.

A total of nine machined shoe openings are welded to each vessel girth strap. These shoe openings accommodate nine keys which themselves are fastened by dowels to the large upper restraint casting which is shown on Figure 15.6-2. These key and shoe openings function to allow vertical thermal expansion of the steam generator within the upper restraint casting, but will restrict lateral movement resulting from forces generated during a seismic event and/or a major pipe break applying lateral loads to the steam generator.

In a seismic event and/or a major pipe break applying lateral loads to the steam generator, the shoe openings in the vessel girth straps act against the keys, which results in a tangential load on the girth straps. The subject studs are designed to accommodate the maximum tangential load resulting from this accident condition. Existing space restrictions and restraint design required a limitation on stud size and quantity which necessitates the use of an ultra-high strength bolting material. The studs have a minimum yield strength of 326,000 psi. The nuts have a minimum yield strength of 150,000 psi.

The upper restraint casting is anchored to the containment structure (approximate 47-foot level) through rigid lateral guides oriented in the direction perpendicular to the outward thermal movement of the steam generator, and by two horizontal large-bore hydraulic snubbers, which permit the thermal movement of the steam generators outward from the reactor. The large-bore snubbers are 12-inch-diameter Pathon snubbers which have been refurbished and upgraded for increased reliability by Paul-Monroe/Remco Hydraulics under Design Change 85-05. The modifications included chroming the cylinder inner diameters; replacement of all non metallic seals with extended-service-life Tefzel seals or metallic seals; installation of poppet-type self-cleaning control valves for improved performance; conversion to the standard snubber hydraulic fluid (General Electric SF-1154); and incorporation of test-in-place features. In addition, the original common reservoir serving a number of snubbers has been replaced by individual pressurized reservoirs installed in more readily accessible locations in lower radiation

Revision 52Updated Online 09/30/20 SPS UFSAR 15.6-9 areas outside the biological shield walls. The reservoirs are plated and the tubing and fittings are stainless steel for corrosion resistance.

The lower restraining ring is also connected to the steam generator cubicle concrete structure by couplings consisting of two end plates installed perpendicular to one another. The end plates have machined dovetails and are joined by a connector plate with mating dovetails.

Although the couplings are resistant to corrosion cracking, additional protection is provided by enclosing each coupling in a rubber boot filled with silicone lubricant. The boot and lubricant are compatible with the coupling and are also radiation resistant.

15.6.2.2.3 Reactor Coolant Pump Support The reactor coolant pumps are supported by a four-legged suspended structure. (Four small-bore snubbers originally installed in the upper frame structure of the pump support were eliminated under Design Changes 85-04-1 and 86-12-2. The dynamic characteristics of the reactor coolant pump were not significantly affected by removal of these snubbers as discussed in Reference 6.) The reactor coolant pump support assembly and material listing are shown on Figure 15.6-3. The supports do not have any heavy section intersecting member weldments.

The major materials used in the construction of the reactor coolant pump supports are listed in Table 15.6-2. Welding associated with the supports was conducted in accordance with Section IX of the ASME Code.

In addition to numerous inspections and tests carried out by the material suppliers and fabricators, all of the components for these supports were subject to inspection during fabrication and installation by Stone & Webster Engineering Corporation. All welds were subjected to examination by either the magnetic particle, liquid penetrant, or radiographic methods. The Vascomax 350 CVM and 300 CVM materials, and the A-352 grade LC3 casting materials were subjected to examination by the magnetic particle and ultrasonic methods. A visual inspection of a portion of these supports is required by the Technical Specifications; however, there is no formal inspection program for all components of the supports during the life of the facility. Major inspections potentially involving disassembly can be conducted on an as-needed basis.

During normal operation the loads and stresses for piping, component connections, and other remaining component supports are not sufficient to cause the failure of the reactor coolant system piping, should there be a complete failure of the reactor coolant pump supports. The maximum stresses that can be expected in the reactor coolant piping as a result of failure of the reactor coolant pump supports during normal operation are summarized in Table 15.6-3. These loads are within the allowable nozzle loads for both the steam generator nozzle and the reactor pressure vessel nozzles. The allowable stresses for the reactor coolant pipe material (A376 Tp 316) are also summarized below. While several of these values are above yield at 650°F, they are all less than 50% of the materials ultimate strength at that temperature. The reactor coolant pipe material has an Sn (code-allowable for normal operation) of 16 ksi at 650°F, and the faulted allowable stress would be 1.85 Sn or 28.8 ksi. All of the loads summarized below are within this

Revision 52Updated Online 09/30/20 SPS UFSAR 15.6-10 faulted allowable, with the exception of the pressurizer inlet. However, thermal stresses have been conservatively included, and their deletion brings the stress levels well within the allowables.

During any postulated accident condition, i.e. seismic and/or pipe breaks, a concurrent complete failure of the reactor coolant pump support would result in unacceptable consequences throughout the reactor coolant loop piping, in terms of loads and stresses. If these supports were to fail during operation, it would be detected, as there is vibration monitoring instrumentation on both the shaft and the frame of the reactor coolant pumps. The amount of vibration is indicated in the control room, and any excessive vibration would cause an annunciator alarm to sound in the control room. To trigger the annunciator alarms, vibration greater than 3 mils on the frame and greater than 15 mils on the pump shaft must occur.

15.6.2.2.4 Pressurizer Support The pressurizer vessel is mounted in a rigid support ring girder suspended by three hanger columns from above. Antisway brackets are welded to the shell of the pressurizer to accommodate shear blocks on the ring; the ring girder is laterally supported by a reinforcement plate attached to embedments in the concrete structure. In addition, lateral support for dynamic loads is provided near the vessels center of gravity by four gapped restraints1 at lugs on the pressurizer which transmit the loads into baseplates on the concrete floor. The lateral gapped restraints and hanger shear blocks and reinforcement plate are able to take all incident loads while allowing the pressurizer vessel to expand radially and vertically. The pressurizer support assembly and material listing are shown on Figure 15.6-4.

15.6.2.2.5 Evaluation The dynamic analysis program accounts for dynamic amplification of the support forces.

These forces are then combined with the design-basis seismic loadings by SRSS summation to ensure that the supports are conservatively designed to withstand the condition of a pipe break occurring as a result of an earthquake. Rigid quality assurance criteria during fabrication ensure conformance with the conservative design.

15.6.3 Containment Internal Structure The reactor containment internal structure is a reinforced concrete structure that furnishes:

1. Supports and restraints for all internal equipment and piping including the polar crane and jib crane.

2. Missile shielding for the containment steel liner and main steam lines against internally generated missiles.

3. Biological shielding for station operators inside the containment structure under all phases of reactor operation.

1. The four lateral gapped restraints were installed by Design Changes 85-04-1 and 85-05-2 to replace the four snubbers in the original support configuration.

Revision 52Updated Online 09/30/20 SPS UFSAR 15.6-11 The structure is designed to withstand the design-basis earthquake together with the simultaneous loss-of-coolant accident (LOCA) 1 , without loss of function. Clearance is maintained between all internal structures and the steel liner of the reactor containment shell to permit differential earthquake motion. The steam generator cubicles and the pressurizer cubicle are designed to withstand an internal differential pressure load of 35 psi resulting from the postulated double-ended primary coolant pipe break. The primary shield is designed to withstand an internal pressure of 100 psig resulting from a hypothetical reactor coolant pipe break within the primary shield.

The differential pressure rise within the cubicles is controlled by open and shielded vent spaces in each cubicle, which permit rapid pressure equalization within the containment structure.

This transient pressure condition has been calculated by Stone & Websters CUPAT Program, using input from the LOCTIC Program.

Temperature differentials between cubicles are considered coincident with the pressure differentials. The short duration of the transient accident relative to the low thermal conductivity of the concrete is such that no significant temperature gradient occurs across the walls. Also, the transient accident is not considered to add to the differential cubicle wall loadings.

Structural concrete design conforms to the requirements of ACI 318, Part IV-B, Ultimate Strength design. Maximum stresses are limited to 90% of the minimum yield point in bending, or 85% of the minimum yield point in diagonal tension, bond, and anchorage.

Special large-size reinforcing steel bars No. 14 and No. 18 are controlled chemistry steel of 50,000 psi yield point, otherwise conforming to the requirements of ASTM A408. All other reinforcing steel is steel of 40,000 psi yield point conforming to ASTM A-15 and ASTM A305.

A stainless-steel-lined fuel transfer canal and reactor refueling cavity is incorporated in the concrete structure above the reactor vessel. A 0.25-inch-thick stainless-steel plate is used to prevent leakage of water from these areas into the containment structure.

Portions of the biological shield walls in the steam generator cubicles are composed of removable precast, reinforced-concrete sections. The wall sections are designed for nondestructive removal to assist future servicing of the steam generators.

Structural steel framing is used as bracing along the top, corners, and ends of the removable shield wall sections. The bracing components conform to ASTM A-36 specifications for structural steel. The precast concrete wall sections have an ultimate compressive strength of

1. As discussed in Section 15.6.2, leak-before-break analyses have demonstrated that the probability of a rupture of the primary reactor coolant loop piping is extremely low, and it is no longer necessary to consider the dynamic effects of such a break. However, the requirements for design of containment and compartments under pressures associated with a postulated primary reactor coolant loop LOCA remain unchanged.

Revision 52Updated Online 09/30/20 SPS UFSAR 15.6-12 3000 psi at 28 days, with steel reinforcement conforming to ASTM specifications for A615, grade 40.

Revision 52Updated Online 09/30/20 SPS UFSAR 15.6-13

15.6 REFERENCES

1. Westinghouse Topical Report WCAP 9558, Revision 2, Mechanistic Fracture Evaluation of Reactor Coolant Pipe Containing a Postulated Circumferential Throughwall Crack, May 1981.

2. Westinghouse Topical Report WCAP 9787, Tensile and Toughness Properties of Primary Piping Weld Metal for Use in Mechanistic Fracture Evaluation, May 1981.

3. Letter from Vepco to NRC,

Subject:

Request for Partial Exemption from General Design Criterion 4, dated November 5, 1985 (Serial No.85-136).

4. Letter from Vepco to NRC,

Subject:

Request for Partial Exemption from General Design Criterion 4 - Supplement, dated December 3, 1985 (Serial No. 85-136A).

5. Letter from Vepco to NRC,

Subject:

Partial Exemption from General Design Criterion 4 -

Request for Additional Information, dated December 27, 1985 (Serial No. 85-136B).

6. Letter from Vepco to NRC,

Subject:

Partial Exemption from General Design Criterion 4 -

Request for Additional Information, dated January 14, 1986 (Serial No. 85-136C).

7. Amendment to General Design Criterion 4 (GDC-4), 10 CFR Part 50, Appendix A, published in Federal Register 51 FR 12502, effective May 12, 1986.

8. Letter from Vepco to NRC,

Subject:

Proposed License Amendment - GDC 4, dated April 30, 1986 (Serial No.86-245).

9. Letter from NRC to Vepco transmitting Surry Unit 1 and 2 License Amendments No. 108 and related safety evaluations, dated June 16, 1986.

10. Letter from Vepco to NRC,

Subject:

Generic Letter 87-11, dated September 12, 1988 (Serial No.88-371).

11. Manual of Steel Construction, 7th Edition, American Institute of Steel Construction.

Revision 52Updated Online 09/30/20 SPS UFSAR 15.6-14 Table 15.6-1 STEAM GENERATOR SUPPORT MATERIALS Supplemental Toughness Mill Test Material Specifications Product Form Requirements Tests Reports NDE A-352grLC3 mod. casting yes yes yes yes Vascomax CVM 350 & 300 forging yes yes yes yes A-106grB 14 in. pipe none none yes yes Table 15.6-2 REACTOR COOLANT PUMP SUPPORT MATERIALS Supplemental Toughness Mill Test Material Specifications Product Form Requirements Tests Reports NDE A-106grB 14 in., 6 in. pipe none none yes yes A-105 GR 11 forging none none yes yes AlSl-4340 forging, plate none none yes no Vascomax CVM 350 & 300 forgings yes yes yes yes A-285grC plate none none yes no A-193grB7 bar none none yes no

Revision 52Updated Online 09/30/20 Table 15.6-3

SUMMARY

OF STRESS FOR FAILURE OF REACTOR COOLANT PUMP SUPPORT DURING NORMAL OPERATION Location Loop Stress (psi)

Steam Generator Outlet A 20,291 B 17,388 C 20,743 Reactor Vessel Inlet A 16,943 B 18,337 C 21,060 Crossover Leg A 21,940 B 13,292 C 20,196 Pressurizer Inlet C 32,728 Material Properties (A376 Tp 316)

°F Syield Sult 100 30 ksi 75 ksi SPS UFSAR 600 18.8 ksi 71.8 ksi 650 18.5 ksi 71.8 ksi 15.6-15

Revision 52Updated Online 09/30/20 SPS UFSAR 15.6-16 Figure 15.6-1 REACTOR NEUTRON SHIELD TANK ASSEMBLY

Revision 52Updated Online 09/30/20 SPS UFSAR 15.6-17 Figure 15.6-2 STEAM GENERATOR SUPPORT ASSEMBLY

Revision 52Updated Online 09/30/20 Figure 15.6-3 REACTOR COOLANT PUMP SUPPORTS GENERAL ARRANGEMENT SPS UFSAR 15.6-18

Revision 52Updated Online 09/30/20 SPS UFSAR 15.6-19 Figure 15.6-4 (SHEET 1 OF 2)

PRESSURIZER SUPPORT

Revision 52Updated Online 09/30/20 SPS UFSAR 15.6-20 Figure 15.6-4 (SHEET 2 OF 2)

PRESSURIZER SUPPORT

Revision 52Updated Online 09/30/20 SPS UFSAR 15.7-1 15.7 MASONRY WALLS Concrete masonry walls are used throughout the plant to provide barriers for radiation shielding, fire protection and personnel separation. Those walls utilized in the construction of Seismic Class I structures are not designed or intended to act as bearing or for transmitting building shear forces. These walls are not used as major load-bearing walls and are not included as part of the overall building shear wall system.

All of the concrete masonry walls that are in proximity to or have attachments for safety-related or equipment such that wall failure could affect a safety-related system were identified and analyzed in accordance with the requirements of IE Bulletin 80-11 (Reference 1).

The reevaluation of the masonry walls was performed based upon criteria for reevaluating concrete masonry walls submitted in Reference 2, which uses as its acceptance criteria the allowable stresses specified in ACI-531-79, Building Code Requirements for Concrete Masonry Structures. Review of test data and the literature substantiates the use of these allowable stresses.

The review also included research of acceptable damping percentages, analysis techniques, in-plane effects, arch action and local stress valves.

The reevaluation criteria considered loads from both safety- and non-safety-related attachments as well as relative interstory displacements between building elevations where applicable. All applicable loads and load combinations specified in the Surry FSAR for concrete design were included in the reevaluation. A review of the walls determined that the walls are not subjected to tornado missiles or depressurization, pipe whip or jet impingement loads. The global review of the walls included seismic inertia loads, interstory displacement loads for both in-plane and out-of-plane effects, equipment loads, and wind loads where applicable.

The local review included discontinuities such as openings and the mechanism for local load transfer into the walls. This included a review of potential local block pull out as well as possible overstress within individual blocks due to attached equipment. Multiwythe walls were also reviewed to ensure the integrity of the collar joint. Calculated shear and tension stresses across the collar joints were compared against allowable values that were conservatively chosen to account for potential small areas of voids or other discontinuities.

At the completion of the response to IE Bulletin 80-11, all identified masonry block walls were evaluated and, modified, as required, to meet the acceptance criteria. The results of this reevaluation program were transmitted to the Nuclear Regulatory Commission. The analysis results of the masonry wall reevaluation program indicated that of the walls requiring reanalysis, 79 walls were acceptable, two walls were acceptable after equipment was removed from the wall, and 31 walls were modified to meet acceptance criteria. Seven safety-related masonry walls in the fuel building were not acceptable under extreme loading conditions and were replaced with blow-off siding. An additional 217 non-safety related walls were also reviewed to ensure that they did not endanger safety-related equipment. Following the approval of responses to IE Bulletin 80-11 by the Nuclear Regulatory Commission, all subsequent modifications involving

Revision 52Updated Online 09/30/20 SPS UFSAR 15.7-2 masonry block walls are evaluated under the Nuclear Design Control Program, which continues to invoke the technical requirements of IE Bulletin 80-11 (References 3 & 4).

15.7 REFERENCES

1. IE Bulletin 80-11, Masonry Wall Design, Nuclear Regulatory Commission, Office of Inspection and Enforcement, May 8, 1980.

2. Letter from B.R. Sylvia, VEPCO, to James P. OReilly, NRC. November 3, 1980, IE Bulletin No. 80-11 Interim Report, Surry Power Station Units 1 & 2, North Anna Power Station Units 1 & 2, Serial No. 878.

3. Letter from Chandu P. Patel, NRC, Office of Nuclear Reactor Regulation, Masonry Wall Design, IE Bulletin 80-11 Surry Power Station Unit Nos. 1 and 3 (TAC Nos. 42867 and 42868), August 11, 1988, Serial No.88-551.

4. Letter from Bart C. Buckley, NRC, Office of Nuclear Reactor Regulation, Surry Units 1 and 2

- Safety Evaluation Masonry Wall Design, IE Bulletin 80-11, (TAC Nos. 42867 and 42868),

October 2, 1989.

Revision 52Updated Online 09/30/20 SPS UFSAR 15A-i Appendix 15A Seismic Design for the Nuclear Steam Supply System and Miscellaneous Components

Revision 52Updated Online 09/30/20 SPS UFSAR 15A-ii Intentionally Blank

Revision 52Updated Online 09/30/20 SPS UFSAR 15A-1 APPENDIX 15A SEISMIC DESIGN FOR THE NUCLEAR STEAM SUPPLY SYSTEM AND MISCELLANEOUS COMPONENTS 15A.1 GENERAL SEISMIC DESIGN CRITERIA FOR THE NUCLEAR STEAM SUPPLY SYSTEM This Appendix is based on Appendix B to the initial FSAR.

All Class I components of the nuclear steam supply system are designed in accordance with the following criteria:

1. Primary operating stresses, when combined with the operating-basis earthquake seismic stresses resulting from a dynamic analysis using a response spectrum normalized to a maximum horizontal ground acceleration of 0.07g and a simultaneous vertical ground acceleration of two-thirds the horizontal, are maintained within the allowable stress limits in Table 15A-1.

2. Primary operating stresses when combined with the design-basis earthquake seismic stresses resulting from a dynamic analysis using a response spectrum normalized to a maximum horizontal ground acceleration of 0.15g and a simultaneous vertical ground acceleration of two-thirds of the horizontal, are limited so that the function of the component or system shall not be impaired, preventing a safe and orderly shutdown of the unit. Further, the primary operating stresses are maintained within the allowed stress limits in Table 15A-1.

No loss of function requires that rotating equipment will not seize, pressure vessels will not rupture, and supports will not collapse or deform to such a degree as to cause failure of the supported equipment. In addition, systems required to be leaktight will remain leaktight, and engineered safeguards will perform intended functions.

15A.2 SEISMIC DESIGN CRITERIA FOR PIPING, VESSELS, SUPPORTS AND REACTOR VESSEL INTERNALS Following discussions with the Staff of the Atomic Energy Commission (AEC) Division of Reactor Licensing during the Construction Permit Application Review for Diablo Canyon Unit 1, the criteria presented in WCAP-5890, Revision 1 (Reference 1), for the generation of limit curves were modified. Details of the manner in which this modification was developed are given in Section 15A.5.1.

Revision 52Updated Online 09/30/20 SPS UFSAR 15A-2 The loading conditions employed in the design of Class I piping, vessels, and supports are enumerated and defined in Section 15A.2.1. The allowable stress limits associated with the various loading conditions are shown in Table 15A-1. Since the reactor vessel internals must also satisfy deformation limits to be able to perform their function, i.e., allow core shutdown and cooling, the vessel internals are discussed separately in Section 15A.2.3.

15A.2.1 Loading Condition Definitions The loading condition definitions given below are taken from Section III of the ASME Boiler and Pressure Vessel Code, Summer 1968 Addenda.

15A.2.1.1 Normal Conditions Any condition in the course of system start-up, operation in the design power range, and system shutdown, in the absence of upset, emergency or faulted conditions.

15A.2.1.2 Upset Conditions Upset conditions are any deviations from normal conditions anticipated to occur often enough that design should include a capability to withstand the conditions without operational impairment. The upset conditions include those transients which result from any single operator error or control malfunction, transients caused by a fault in a system component requiring its isolation from the system, transients due to loss of load or power, and any system upset not resulting in a forced outage. The estimated duration of an upset condition shall be included in the design specifications. The upset conditions include the effect of the operating-basis earthquake for which the system must remain operational or must regain its operational status.

15A.2.1.3 Emergency Conditions Any deviations from normal conditions which require shutdown for correction of the conditions or repair of damage in the system. The conditions have a low probability of occurrence but are included to provide assurance that no gross loss of structural integrity will result as a concomitant effect of any damage developed in the system.

15A.2.1.4 Faulted Conditions Those combinations of conditions associated with extremely low probability postulated events whose consequences are such that the integrity and operability of the nuclear energy system may be impaired to the extent where considerations of public health and safety are involved. Such considerations require compliance with safety criteria as may be specified by jurisdictional authorities. Among the faulted conditions may be a specified design-basis earthquake for which safe shutdown is required.

Revision 52Updated Online 09/30/20 SPS UFSAR 15A-3 15A.2.2 Piping, Vessels, and Supports The reasons for selection of the above-mentioned loading conditions and allowable stress limits are as follows:

1. When subjected to the operating-basis earthquake, the nuclear steam supply system is designed to be capable of continued safe operation. Equipment and supports needed for this purpose are required to operate within normal design limits, as shown in Table 15A-1. The load combination corresponding to the upset loading condition is the normal load, plus the operating-basis earthquake load.

2. In the case of the design-basis earthquake, it is necessary to ensure that components required to shut the unit down and maintain it in a safe shutdown condition do not lose their capability to perform their safety function. This capability is ensured by maintaining the emergency stress limits as shown in Table 15A-1.

3. For the highly unlikely but postulated case of pipe rupture, a reactor coolant branch or other potentially governing break, the effects of the pipe rupture will not cause failure propagation to the reactor coolant piping. The load combination corresponding to the faulted loading condition is the design-basis earthquake and/or design-basis accident load.

4. For the extremely remote event of simultaneous occurrence of a design-basis earthquake and postulated pipe rupture of a reactor coolant system branch line or other potentially controlling break, the Class I piping and component are checked for no loss of function, i.e., the capability to contain fluid, allow fluid flow, and perform vital engineered safeguards functions. This is ensured by limiting the various stress combinations within the faulted condition design limits shown in Table 15A-1.

The minimum margin of safety between the design limit stress and the expected collapse condition is for the case of pure tension, and is defined as:

S ultimate - S design S design Under more realistic operating conditions, piping and vessels will always experience some combination of tension and bending. For these combined load cases, the margin of safety is greater than that for pure tension, as shown by the limit curves contained in WCAP-5890, Rev. 1, and shown in Figures 15A-4 and 15A-5. Therefore, it is conservative to base the margin of safety on pure tension. Table 15A-2 illustrates the margin of safety between the stress limits for various load conditions and the expected failure or collapse condition for typical materials.

Revision 52Updated Online 09/30/20 SPS UFSAR 15A-4 Plastic or limit analyses conducted within the limits of the faulted condition were performed considering plastic material behavior, including, as required, modifications of material stiffness characteristics, formation of plastic hinges and other non-linear effects, as determined in detailed structural analysis, and provided in standard stress reports.

15A.2.3 Reactor Vessel Internals 15A.2.3.1 Design Criteria for Normal Operation The internals and core are designed for normal operating conditions and subjected to loads of mechanical, hydraulic, and thermal origin. The response of the structure under the operating-basis earthquake is included in this category as well as operational transients (upset conditions).

The stress criteria established in Section III of the ASME Boiler and Pressure Vessel Code, Article 4, have been adopted as a guide for the design of the internals and core, with the exception of those fabrication techniques and materials not covered by the Code, such as the fuel rod cladding. Seismic stresses are combined in the most conservative way and are considered primary stresses.

The members are designed under the basic principles of: (1) maintaining deflections within acceptable limits, (2) keeping the stress levels within acceptable limits, and (3) preventing fatigue failures.

15A.2.3.2 Design Criteria for Abnormal Operation The abnormal design condition assumes blowdown effects due to a pipe break1 combined (by SRSS combination) in the most unfavorable manner with the effects associated with the design-basis earthquake.

For this condition, the criteria for acceptability are that the reactor is capable of safe shutdown and that the engineered safety features are able to operate as designed. Consequently, the limitations established on the internals for these types of loads are concerned principally with the maximum allowable deflections. Additional stress criteria for critical structures under normal operation, plus the design-basis earthquake and blowdown excitation, ensure that the structural integrity of the components is maintained.

15A.3 GENERAL ANALYTICAL PROCEDURE FOR SEISMIC DESIGN The design and analysis of Class I components of the nuclear steam supply system utilizes the response spectrum approach.

1. As discussed in Section 15.6.2, it is no longer necessary to consider the dynamic effects of a postulated rupture of the primary reactor coolant loop piping. However, pipe ruptures of reactor coolant branchlines are still postulated.

Revision 52Updated Online 09/30/20 SPS UFSAR 15A-5 The dynamic analysis is performed using the normal mode methods. The inertial properties of the models are characterized by the mass and mass moment of inertia of each mass point. The stiffness properties are characterized by the moment of inertia, area, shear shape factor, torsion constant, Youngs modulus, and shear modulus.

Table 15.2-2 gives the damping ratios to be used in the dynamic analysis of components.

15A.3.1 Mechanical Equipment The Westinghouse-supplied Class I mechanical components for Surry that require a seismic analysis were determined and checked for seismic adequacy by employing the following procedure:

1. The manufacturers drawings were reviewed to classify the component.

a. If the component fell within a category that was previously analyzed using a multi-degree-of-freedom model and shown to be relatively rigid, then a static seismic analysis was performed to check equipment seismic adequacy.

b. If the component could not be categorized as similar to previously analyzed components, then a seismic modal analysis was performed, using multi-degree-of-freedom dynamic models.

2. Stresses and deflections were checked to ensure that they were within allowable limits and did not result in loss of function.

Typical Class I mechanical equipment of the engineered safety feature (ESF) systems supplied by Westinghouse was originally analyzed on a worstplant basis using a multi-degree-of-freedom modal analysis. The term worstplant basis is defined, for the particular component in question, as the most severe seismic response spectra applicable to any Westinghouse plant employing that particular piece of equipment. All contributing modes were considered. A sufficient number of masses was included in the mathematical models to ensure that coupling effects of members within the component were properly considered. The results of these analysis indicated that the models contained more masses than necessary, and that future analyses of comparable equipment could be considerably simplified by considering fewer masses, or merely performing a simple static analysis.

The method of dynamic analysis used a proprietary computer code called WESTDYN. This code uses inertia values, member sectional properties, elastic characteristics, support and restrain data characteristics, and the appropriate seismic response spectrum as input. Both horizontal and vertical components of the seismic response spectrum are applied simultaneously. The modal participation factors are combined with the mode shapes and the appropriate seismic response spectra to give the structural response for each mode. The internal forces and moments are computed for each mode from which the modal stresses are determined. The stresses are then summed using the square-root-sum-of-squares method, except for the major components in the

Revision 52Updated Online 09/30/20 SPS UFSAR 15A-6 reactor coolant loop. As discussed in Section 15A.3.3, for the reactor coolant loop analysis, the combination of modal responses for closely spaced modes is performed using the grouping methodology in Regulatory Guide 1.92.

In analyzing equipment to resist seismic loads, the vertical seismic spectrum, equal to two-thirds of the horizontal response spectrum, is used to determine the acceleration appropriate to the vertical frequency. An idealized umbrella spectrum was used in the analyses. The floor response spectra at the Surry site are encompassed by the umbrella spectrum used in the dynamic analysis of Westinghouse-supplied equipment.

Typical Class I engineered safeguards tanks supplied by Westinghouse, e.g., for boric acid, accumulator, and boron injection, were analyzed using the method above, with the combined horizontal and vertical seismic excitation occurring simultaneously in conjunction with normal loads. Hydrodynamic analyses of these tanks have been performed using the methods described in TID 7024 (Reference 2).

Selected critical Class I ESF valves supplied by Westinghouse have been analyzed using the above method, and the results indicate that their fundamental natural frequency is sufficiently separated from the building frequency. The results indicate that the total stress, considering all modes, is far below the allowable stress limit. Motors attached to motor-operated valves have been included in the mathematical models.

The deflections and stresses obtained from the seismic analysis are added to the deflections and stresses associated with the operational mode of the mechanical equipment to verify that clearances are not exceeded and the stresses are within allowable limits. This criterion ensures that this equipment will perform the intended function under seismic conditions.

All mechanical equipment analyzed had fundamental modes in the rigid range, with the exception of loop stop valves and some vertical tanks.

The fundamental frequency of vertical tanks and the loop stop valves was greater than 9 Hz, which is outside the resonance range of the structures in which they are housed. The component supports were modified to remove the fundamental frequency of the item from the resonance range of the structure, by providing stiffer supports to increase the fundamental frequency of the component. The selection of the type of restraint used was dependent upon the component analyzed and the structure surrounding the component.

Restraints, snubbers, or other devices are not used to preclude resonance of the electrical and control systems equipment for seismic loading. Protection system equipment that is typical of the design has been subject to tests under simulated seismic accelerations to demonstrate the ability to perform and complete its function. These data are contained in WCAP-7397-L (Reference 3).

Revision 52Updated Online 09/30/20 SPS UFSAR 15A-7 The seismic loads for the design and analysis of Class I mechanical components, including pumps, valves, heat exchangers, and tanks within Stone & Websters scope of responsibility, were based on the ground response spectra (GRS) shown in Figures 2.5-5 and 2.5-6 or amplified floor response spectra (ARS), depending on their location. Seismic loads were developed for the operating-basis earthquake (OBE) and the design-basis earthquake (DBE). The spectra used in the evaluation of Class I mechanical components were based on damping values consistent with those indicated in Table 15.2-2. Where applicable, seismic loads were combined with the results from other load cases such as thermal and dead load. Constraints such as snubbers or other appropriate devices are utilized wherever necessary to meet design requirements.

All Class I mechanical components are designed to withstand the operating-basis earthquake and to function through the design-basis earthquake to safe shutdown. Vendors supplying the components are required by specification to design the components to function, as outlined above, under the seismic loadings. The vendor is required to validate component integrity under the specified seismic conditions.

15A.3.2 Earthquake Experience-Based Method Developed for Unresolved Safety Issue (USI) A-46 for Seismic Verification of Equipment In response to U.S. Nuclear Regulatory Commission Generic Letter 87-02 on USI A-46, Verification of Seismic Adequacy of Mechanical and Electrical Equipment in Operating Reactors, a Generic Implementation Procedure (GIP) was developed by the Seismic Qualification Utilities Group (SQUG). The criteria and methodology in Revision 3 of the GIP (Reference 40), as modified and supplemented by the NRC Supplemental Safety Evaluation Reports (SSERs) 2 and 3 (References 44 & 45) may be used, with certain additional considerations, as an alternative to other licensing basis methods for seismic design and verification of existing, modified, new and replacement equipment classified as safety-related, NSQ or seismic category 1.

Considerations that are additional to the GIP pertain to the following issues:

• Use of GIP Method A for estimating seismic demand.

• Additional criteria applicable for the design and analysis of new flat bottom vertical tanks.

• Applicability of Part II, Section 5 of the GIP for conduit and cable tray raceways.

• Use of criteria associated with damping values, static coefficient and expansion anchor safety factors for equipment anchorage evaluations conforming to the current, conservative, licensing basis commitments.

• Documentation of the results of the Screening Verification and Walkdown in Section 4.6 of the GIP may be limited to the use of walkdown checklists. It is not necessary to complete the Screening Verification Data Sheets (SVDS).

Revision 52Updated Online 09/30/20 SPS UFSAR 15A-8

• It is not necessary to identify essential relays and perform functionality screening as defined in Section 6 of the GIP. Relays designated as Class 1E are evaluated by comparing seismic capacity to seismic demand.

• The GIP method is generally applicable only for equipment located in mild environment.

However, with case-by-case justification, it may be used for equipment in harsh environment.

Guidance for the use of the GIP for the seismic design and verification of mechanical and electrical equipment, including a discussion of the above exceptions, is provided in an engineering procedure (Reference 62).

15A.3.3 Reactor Coolant Loops and Supports The original configuration of the reactor coolant system equipment supports included ten large-bore hydraulic snubbers per loop to carry the loads from postulated pipe ruptures of the reactor coolant system, main steam line, and feedwater line. Subsequent fracture mechanics analyses, submitted to the NRC on behalf of the Westinghouse Owners Group, demonstrated that the probability of rupture of the primary coolant loop is extremely small, and that the presence of a pipe crack could be detected by leakage well before the crack grew to a critical size which would cause rupture. NRC Generic Letter 84-04 (Reference 4) provided the NRC staff safety evaluation of these leak-before-break analyses, concluding that, provided certain specific conditions are met, the dynamic effects of a postulated pipe break in the reactor coolant system primary loop need not be considered as a design basis for the reviewed plants, including Surry Units 1 and 2. Subsequently, Generic Letter 87-11 (Reference 5) eliminated the need to postulate Arbitrary Intermediate Breaks and allowed removal of the associated pipe whip restraints and jet impingement shields. Because the stresses in the main steam lines inside containment are well below the stress criteria for required mandatory intermediate breaks, the only breaks which need be postulated are terminal end breaks which do not apply lateral loads to the steam generator.

Based on the relief provided by these two relaxations of criteria, the reactor coolant system supports have been modified to eliminate eight of the ten large-bore snubbers per loop of the reactor coolant system. These efforts are discussed in Section 15.6.2; additional information is contained in References 6-15.

The reactor coolant loop system was re-evaluated with the snubbers eliminated to assure that the conditions of pressure, deadweight, thermal, seismic, and remaining pipe rupture effects, would not result in unacceptable stress levels or factors of safety. Two essentially independent analyses of a representative single primary loop were performed by Westinghouse Electric Corporation and Stone & Webster Engineering Corporation, in accordance with the original division of design responsibilities. Westinghouse performed deadweight, pressure, thermal and seismic analyses using a simplified model as the run of record to obtain piping stresses. Stone &

Webster performed analyses using a model incorporating a detailed representation of the support

Revision 52Updated Online 09/30/20 SPS UFSAR 15A-9 members, principally to obtain component support loads under normal and accident loadings.

Both analytical and models were revisions to existing models and incorporated changes due to the steam generator replacements as well as the snubber elimination efforts.

These analyses incorporated the loads from deadweight, internal pressure, thermal expansion, seismic events (OBE and DBE), and the dynamic effects of pipe ruptures of other systems (controlling breaks, for example), in the main steam line, pressurizer surge line, main feedwater line, etc.) No other hydraulic transient loading was considered as significant.

For seismic analysis, the soil structure interaction amplified response spectra for 0.5 percent critical equipment damping (OBE) and 1 percent equipment damping (DBE) were used with appropriate bump factors as discussed in Section 15A.3.5.3. These damping values are lower than those in Regulatory Guide 1.61, Damping Values for Seismic Design of Nuclear Power Plants and in ASME Code Case N-411, Alternative Damping Values for Seismic Analysis of Classes 1, 2, and 3 Piping. The vertical and horizontal earthquake responses were combined for piping analysis as described in Section 15A.3.3. For component support analysis, the responses to the three directions of earthquake loading were combined by SRSS. The combination of modal responses for closely spaced modes was performed using the grouping methodology in Regulatory Guide 1.92.

The Westinghouse analysis used the WESTDYN code and a simplified representation of the component supports as stiffness matrices. The component support stiffness matrices were supplied by Stone & Webster. The WESTDYN computer code has been utilized on numerous Westinghouse plants, and was reviewed and found acceptable by the NRC for the Surry units in 1974. A detailed description of the WESTDYN method of analysis is given below.

The code uses as input system geometry, inertia values, member sectional properties, elastic characteristics, support and restraint data characteristics, and the appropriate Surry seismic response spectrum of 0.5% critical damping. Both horizontal and vertical components of the seismic response spectrum are applied simultaneously.

With these input data, the overall stiffness matrix [K] of the three-dimensional piping system is generated (including translational and rotational stiffness). Restraints are deleted, and the stiffness matrix is inverted to give the flexibility matrix [F] of the system.

[F] = [K]-l A product matrix is formed by the multiplication of the flexibility and mass matrices. This product matrix forms the dynamic matrix [D] from which the modal matrix is computed.

[D] = [F] [M]

The modal spectral matrices are generated using a modified Jacobi method.

(2 [M] - [K]) X = O

Revision 52Updated Online 09/30/20 SPS UFSAR 15A-10 From this information, the modal participation factor is combined with the mode shapes and the appropriate seismic response spectra to give the structural response for each mode.

The Stone & Webster analysis used the STARDYNE computer code and a model incorporating a detailed representation of the support members. STARDYNE is a public domain computer program and is recognized as a Category 1 computer program suitable for nuclear safety-related work. The program used is maintained under Stone & Websters Quality Control procedures. The following modules of STARDYNE, Version 3, Level H, were used:

• STAR (Static and Modal Extraction Analysis)

• DYNRE4 (Seismic Response Spectrum Analysis)

• DYNRE6 (Time History Transient Analysis)used only for evaluating pipe rupture loadings Dynamic analyses were performed of the controlling pipe ruptures in the pressurizer surge line, main steam line, and main feedwater lines. The originally-postulated terminal and intermediate breaks were reviewed by Stone & Webster to determine those breaks which would cause the most severe loadings on the revised support configuration with snubbers removed.

Time-history forcing functions were applied to the detailed model representing these potentially limiting breaks, to obtain maximum member loads with the revised support configuration. These loads were combined by SRSS with the seismic DBE loads and then summed with deadweight and pressure loads.

The results of the two independent analyses with revised support configuration established that the frequencies of most vibrational modes are virtually unchanged by the snubber eliminations. Comparison of the interface loads calculated by the two models was performed to ensure that the results of the two models were consistent; the significant interface loads were found to be within 15%, which is considered to be good agreement. The analyses demonstrated that the piping components and supports are stressed within acceptable limits, and adequate safety margins exist in a seismic event. The maximum level of stress (percentage) compared to the Code allowable at the highest stress point in each leg of the reactor coolant loop for thermal, deadweight, and seismic conditions are given in Table 15A-5.

In addition, the maximum resultant bending moment in the primary coolant loop piping under combined deadweight, pressure, thermal, and design basis seismic loadings is 28,860 inch-kips. This value is less than the enveloped value in the Westinghouse topical report, WCAP-9558, Revision 2 (Reference 6), and also less than 42,000 inch-kips which was identified in NRC Generic Letter 84-04 as the maximum allowable moment for the Westinghouse Owners Group plants for justifying that pipe rupture need not be postulated in the primary reactor coolant loop piping.

Revision 52Updated Online 09/30/20 SPS UFSAR 15A-11 For combined normal operating and seismic loads, the location of the maximum stress is the steam generator outlet elbow; for the main steam line rupture loading, the location of maximum stress is in the steam generator inlet elbow; for the pressurizer surge line break, the maximum stress occurs in the hot leg where the surge line intersects.

Stone & Webster evaluated the calculated maximum loadings on the supports of the modified support configuration with eight of the large-bore snubbers eliminated. The factors of safety (allowable load/combined load) for the combined deadweight, pressure, thermal, and design basis seismic loads are given in Table 15A-6. Similar factors of safety under the combined deadweight, pressure, thermal, and SRSS combination of design basis seismic loads plus maximum pipe rupture loads, are given in Table 15A-7. These tables demonstrate that adequate factors of safety exist under all loading conditions.

The results of these analyses confirm that the large-bore snubber eliminations do not compromise the bases for the previous leak-before-break analysis, namely:

1. The loading on the primary loop piping is still enveloped by the generic analyses submitted on behalf of the A-2 Owners Group and accepted by the NRC staff in Generic Letter 84-04, and specifically for Surry by NRC letter dated June 16, 1986; and

2. The reactor coolant system equipment, piping, and supports continue to have acceptable margins of safety under licenced loading conditions other than the now-eliminated ruptures of the primary loop piping and Arbitrary Intermediate Break of the main steam lines.

The inertial forces and moments are computed for each mode from which the modal stresses are determined. The stresses are then summed using the square-root-sum-of-squares method.

The maximum stresses in the reactor coolant loop piping imposed by the normal loads plus loads associated with the design-basis earthquake are below the allowable stress limit. The stress levels in the reactor coolant loop piping are provided in Table 15A-7 (References 9 and 11).

15A.3.4 Anchor Bolts The majority of anchor bolts originally installed at the Surry Power Station were shell-type Phillips self-drilling anchors. A minimum safety factor of four was used. Cyclic loads and the effect of baseplate flexibility were not specifically considered; however, supports, baseplates, and anchor bolts were designed to withstand the maximum force exerted by seismic and thermal conditions.

In response to IE Bulletin 79-02 (Reference 16), all pipe support baseplates were analyzed considering baseplate flexibility, and modifications were made when baseplates and/or anchor bolts were found inadequate. Wedge-type Hilti bolts were installed in accordance with manufacturers requirements based on onsite testing conducted by Hilti, Inc.

Revision 52Updated Online 09/30/20 SPS UFSAR 15A-12 A finite element analysis was performed using the ANSYS computer program as provided by an owners group organized by Teledyne Engineering Services. A description of this program was submitted to the NRC as Technical Report TR-3501-1, Revision 1 (Reference 17). In some cases where the support plate could not be modeled in the computer program, hand calculations were performed, allowing sufficient margin for baseplate flexibility and prying action.

The factors of safety of four for wedge-type anchors and five for shelltype anchors were used to determine the anchor bolt allowable loads for the reanalysis. All baseplates were reanalyzed to ensure that these factors of safety were met. Where the factors of safety were not met by analysis, modifications were provided to ensure the appropriate factor of safety. The original design for anchor bolts at Surry was based on a factor of safety of four for all anchor bolts, based on a design concrete strength of 3000 psi. In conjunction with IE Bulletin 79-02, a concrete inspection program was performed to demonstrate a concrete strength of 4000 psi, which would provide the factor of safety of five required by the Bulletin. Thirty-two Windsor probe tests were performed at various locations throughout the plant (Surry Units 1 and 2) to provide data for the evaluation. The results of this program show a 95% confidence level of at least 4000 psi concrete. Therefore, the analysis was based on 4000 psi concrete with the factors of safety of four and five required by Bulletin 79-02.

No special design requirements for the anchor bolts to withstand cyclic loads were applied.

Testing performed for the owners group, the results of which are presented in Technical Report TR-3501-1 indicates that cyclic loading on the anchors does not result in a general reduction of the ultimate capacity of the anchor. Bolts for shell-type anchors (Phillips Red-Head self-drilling anchors) were tightened snugly, but were not preloaded. Wedge anchors (Hilti bolts) were preloaded to the design allowable load.

To ensure that the design requirements have been met for the installed anchor bolts, an inspection and testing program was conducted. Under this program, one anchor bolt per accessible base plate was inspected and tension tested to at least the anchor bolt design load.

Anchor bolt installations which were suspect based on the visual inspection were tension tested to at least the anchor bolt design allowable load (20% of the manufacturers ultimate) and evaluated for a factor of safety of five by tension testing to five times the design load, or determining the anchor capacity based on the results of the visual inspection. When the anchor was found to be inadequate as a result of the evaluation, or of slippage greater than 1/16 inch under the tension test, the baseplate was reanalyzed with that bolt missing. The remaining bolts on the baseplate were inspected and tested for adequacy under the higher redistributed load when the reanalysis was acceptable, or for the original design load when the loads could not be redistributed.

Inspection and test results showed that 97% of the baseplates were acceptable. All anchor bolts that were inadequate or damaged were replaced to ensure adequacy of the anchorage system.

In order to evaluate operability of each Seismic Category I piping system, the anchor bolt inspection and test results were recorded on a system basis. The system designations shown in Table 15.2-1 were used in conjunction with a QA Category I piping line table to determine

Revision 52Updated Online 09/30/20 SPS UFSAR 15A-13 systems for Bulletin 79-02 purposes. Of the 14 systems for which anchor bolt inspections were performed, 12 of the systems had acceptance percentages greater than 95%. The acceptance percentages for the other two systems, the reactor coolant system and the residual heat removal system, were 94.7% and 92.1%, respectively. All systems with baseplates inaccessible for bolt inspection and testing had remote visual inspections performed to ensure that all anchor bolts were present and no gross deficiencies existed. For the two systems with less than 95%

acceptance, the inaccessible baseplates were further evaluated to ensure high design factors of safety greater than those required by the Bulletin. Review of the baseplates where anchor bolts were found to be inadequate did not indicate any common characteristic (i.e., floor plates, wall plates, or ceiling plates) which would necessitate further inspection and testing of the inaccessible baseplates with a particular characteristic.

Piping systems 2 in. in diameter or less were originally designed by a chart analysis method.

To ensure adequacy of the baseplates and anchor bolts in justifying operability of the small-bore piping, a sampling program was initiated. Five 2-inch lines were selected as representative of the small bore piping. Three safety injection lines and two chemical volume and control lines, which have a total of 22 supports with 43 baseplates, were analyzed in this effort. Baseplate analysis efforts show anchor bolt factors of safety ranging from 5.2 to 638, with the majority of anchor bolts having design factors of safety above 60.

Seventy-three anchor bolts on 12 of the small-bore baseplates were inspected and tension tested. Sixty-eight of these anchor bolts (93%) were accepted. The baseplates for the five rejected anchors were reanalyzed with the discrepant bolts missing and all were found acceptable and within the allowable limits. All anchor bolts which were inadequate or damaged were replaced to ensure adequacy of the anchorage system.

The small-bore piping baseplates and anchor bolts were designed and installed by the same architect-engineer and contractor that performed the work on the large-bore piping. Therefore, based on the above results, which are consistent with the large-bore anchor bolt program, a sufficient degree of conservatism exists in the baseplates and anchor bolts of the small-bore piping to justify acceptance of this piping.

All Seismic Category I pipe supports on masonry walls were resupported without attachment to the masonry walls.

IE Bulletin 79-02 inspection details were provided to the NRC by References 18, 19, and 20.

15A.3.5 Piping Systems The Stone & Webster PSTRESS/SHOCK 2 computer code was used in the seismic analyses of certain systems at Surry Units 1 and 2. This code summed earthquake loadings algebraically, which is unacceptable for reasons set forth by the NRC in IE Bulletin 79-14 (Reference 21) and in a March 13, 1979 Order to Show Cause (Reference 22). As a result, Vepco reanalyzed

Revision 52Updated Online 09/30/20 SPS UFSAR 15A-14 safety-related systems originally analyzed by SHOCK 2, modified those systems as necessary, depending on the results of the reanalyses, and provided support for the acceptability of the analysis methods used on the remaining Seismic Class I systems.

Portions of the following systems were identified as having been analyzed with SHOCK 2:

• Pressurizer spray and relief.

• Low-head safety injection.

• High-head safety injection.

• Containment and recirculation spray.

• Residual heat removal.

• Component cooling water.

• Service water.

• Main steam.

• High-pressure steam.

• Feedwater.

• Auxiliary feedwater.

• Containment vacuum.

• Fire protection (Unit 1 only).

• Diesel muffler exhaust (Unit 1 only).

Vepco has reanalyzed all pipe stress problems originally analyzed by SHOCK 2. All supports were reanalyzed and modifications completed as necessary.

Reanalysis and safety evaluation details are given in References 23 through 26.

15A.3.5.1 Reanalysis Methods and Results As the original analysis used an algebraic intramodal summation technique, the safety-related piping system supports and attached equipment were reanalyzed with acceptable methods. The reevaluation included a dynamic computer analysis using NUPIPE programs, which incorporated a lumped mass response spectra modal analysis technique.

Revision 52Updated Online 09/30/20 SPS UFSAR 15A-15 The floor response spectra used in the reanalysis included the original amplified response spectra specified in this appendix. In some cases, piping was reanalyzed utilizing ARS that were developed using SSI techniques. The peaks in the amplified floor response spectra were broadened by +/-15% in accordance with Regulatory Guide 1.122 to account for variation in material properties and approximations in modeling.

The piping systems were modeled as three-dimensional lumped mass systems which included considerations of eccentric masses at valves and appropriate flexibility and stress intensification factors. The dynamic analysis procedures meet the criteria specified in this appendix, and are acceptable. The resultant stresses and loads from the reanalysis were used to evaluate piping, supports, nozzles, and penetrations.

Based on NRC review of the computer codes used for reanalysis, independent check analyses, and a review of modeling methods used by the Licensee, the NRC found the procedures and methods used in reanalyzing these problems acceptable.

The reanalysis included problems involving the reactor coolant system boundary and the supports associated with those problems. Since the reactor coolant system boundary is inside containment and all of the supports have been modified as necessary, there is no potential for a loss-of-coolant accident in the event of a design-basis earthquake.

At the request of the NRC, its consultant, EG&G, performed audit pipe stress calculations of five Surry 1 problems using the NUPIPE computer code. The results of the EG&G audit compared favorably with Vepcos results.

The piping support designs for affected system piping were inspected by Vepco to verify the location, orientation, support clearances, and support type. Any deviations were incorporated into piping reanalyses. These piping systems were also verified by the NRC Office of Inspection and Enforcement.

The pipe supports were reevaluated in cases where the original support design loading was exceeded as a result of piping reanalysis. In cases where the original support capacity was exceeded, the support reevaluation included the consideration of baseplate flexibility and a verification of actual field construction of the support. Where concrete expansion anchor bolts were used, their capacities, without compromising the originally committed safety margin, were also included in the reevaluation.

The pipe break criteria of this appendix were reviewed in connection with the possible effect of changes of the high-stress point resulting from the reanalyses. Results of the evaluation of the effect the reanalysis has on the pipe break criteria show that no new whip restraints are required. Therefore, the reanalysis has not changed the pipe break protection.

Revision 52Updated Online 09/30/20 SPS UFSAR 15A-16 The design and analysis of the supports and attached equipment are in accordance with the criteria specified in this appendix. The piping systems and supports were designed to the allowable limits of ANSI B31.1 for the gross properties, and to the limits of ANSI B31.7 Appendix F, for local stress considerations as per the criteria of this appendix. A reanalysis of the pressurizer surge line to account for the effect of thermal stratification and striping was performed in accordance with the requirements of ASME Boiler and Pressure Vessel Code,Section III 1986 with addenda through 1987 incorporating high cycle fatigue as required by NRC Bulletin 88-11, dated December 20, 1988.

15A.3.5.2 Verification of Analysis Methods The following computer codes and analysis methods have been identified as the current basis for the facility piping design:

1. NUPIPE/S&W

2. NUPIPE

3. Static analysis methods

4. PIPESTRESS 15A.3.5.2.1 NUPIPE/S&W Stone & Webster has submitted documentation on the NUPIPE/S&W code to the NRC.

This code calculates intramodal and intermodal responses according to the provision in Regulatory Guide 1.92 (Reference 27). A review of the code listing by the NRC staff has confirmed this statement. The option used by Vepco specifies an intramodal combination consisting of the addition of the absolute value of the responses due to the vertical earthquake component and the root-mean-square combination of the responses due to the two horizontal earthquake components. Additional documentation has also been submitted by the originators of this code (Quadrex), providing detailed information on the methods of modal combination.

Vepco solved three NRC benchmark piping problems and the solutions showed acceptable agreement with the benchmark solutions. In addition, a confirmatory problem (No. 323A) was provided to Brookhaven National Laboratory for confirmatory solution. A comparison of the solutions demonstrated good agreement (within about 10%).

15A.3.5.2.2 NUPIPE Ebasco Services, Inc., has submitted documentation to the NRC on the NUPIPE computer code, which was used in the piping reanalysis of Unit 2. This code is considered acceptable for analyses for both units.

Revision 52Updated Online 09/30/20 SPS UFSAR 15A-17 This code has previously been reviewed and has been found to satisfy the requirements of Regulatory Guide 1.92. Ebasco solved three of the NRC benchmark piping problems, and its solutions were found to agree closely with the benchmark solutions. They also provided a confirmatory problem (2508A), which was solved by the Brookhaven National Laboratory.

Comparison of the solutions showed good agreement.

15A.3.5.2.3 Static Analysis Methods Static analysis methods, which were used in cases not subjected to computerized seismic analysis, are based on simple beam formulations, wherein seismic stress levels are controlled through use of pre-established seismic spans. These simple beam formulations were utilized to calculate maximum allowable spans based upon an assumed acceleration factor of 1.5 times the peak acceleration obtained from the response spectra. In calculating the maximum span lengths, it was conservatively assumed that a longitudinal pressure stress of 4000 psi and a maximum deadweight stress of 1500 psi were present in the pipe. This combined value of 5500 psi was subtracted from the allowable stress (1.8 Sh for pressure and deadweight and seismic) to obtain a seismic allowable stress.

Calculating maximum spans by this procedure results in maximum allowable spans greater than the deadweight spans recommended in ANSI B31.1. Thus, dead weight governs and provides a greater number of supports resulting in closely spaced restraints. To minimize effects of concentrated weights, restraints were placed as required at valves and other concentrated masses.

For Surry, piping 6 inches in diameter and smaller was generally analyzed using the simplified static method, with the option of utilizing more rigorous methods available to the analyst. Piping 2 inches and below was shown on the piping drawings diagrammatically (i.e.,

without detailed dimensions). The stress engineers located supports during the installation process working at the site with erection isometric sketches.

The stress analysis was performed by assuming many simple supported straight beams, the spans of which are governed by deadload spacing requirements of ANSI B31.1. The piping fundamental frequencies associated with these maximum allowable spans (9.7 to 13.6 cycles per sec) are not in resonance with the building in which they are located (2 to 8 cycles per sec). The method of equivalent static analysis outlined in this procedure was compared with the NRCs Standard Review Plan 3.7.2 (Reference 28) and found to be acceptable.

15A.3.5.2.4 PIPESTRESS The Unit 2 RVLIS piping and Head Vent piping were reconfigured to accommodate the head assembly upgrade package. Reanalysis of the piping was performed using PIPESTRESS (Reference 68).

Revision 52Updated Online 09/30/20 SPS UFSAR 15A-18 The PIPESTRESS program is a finite element computer program which performs linear elastic analysis of piping systems using the stiffness method of finite element analysis; the displacements of the joints of a given structure are considered basic unknowns. The dynamic analysis by the modal synthesis method utilizes known maximum accelerations produced in a single degree of freedom model of a certain frequency. The principal program assumptions are as follows:

• It is a linearly elastic structure.

• Simultaneous displacement of all supports is described by a single time-dependent function.

• Lumped mass model satisfactorily replaces the continuous structure.

• Modal synthesis is applicable.

• Rotational inertia of the masses has negligible effect.

The results obtained from the pipe stress program PIPESTRESS have been compared with the following:

• ASME Benchmark problem results, Pressure Vessel and Piping 1972 computer programs verification, American Society of Mechanical Engineers.

• Longhand calculations - PIPESTRESS is compatible with NRC Regulatory Guide 1.92.

A synthesis of closely spaced modes is provided based on equation 4 of Regulatory Guide 1.92.

• Benchmark confirmatory piping analysis problems were reviewed by the NRC and Brookhaven National Laboratory.

The PIPESTRESS program is used to determine stresses and loads in the piping systems due to restrained thermal expansion, deadweight, seismic inertia and anchor movements, externally applied loads such as jet-loads, and transient forcing functions such as created by fast relief valve opening and closing, fast check valve closure after pipe breaks in main feedwater line, fast valve closure in main steam line, etc. PIPESTRESS analyzes piping systems in accordance with ANSI and ASME codes.

15A.3.5.3 Soil Structure Interaction Soil structure interaction amplified response spectra (SSI-ARS) were used in reanalyzing the piping systems for those cases where the original amplified response spectra did not give satisfactory results. Based upon review of Vepcos information submitted by References 29 and 30, the NRC informed Vepco by Reference 31 and 32 that SSI-ARS was acceptable.

Revision 52Updated Online 09/30/20 SPS UFSAR 15A-19 The amplified floor response spectra (ARS) for three levels in the containment, base mat, operating floor and spring line were computed using the multi-layered elastic half-space method and the finite element methods. The results of these analyses were compared for frequency and acceleration of the floor response spectra. The elastic half-space method gave acceleration values that were larger than the finite element method for the operating floor and the spring line. The finite element method gave accelerations slightly higher than the elastic half-space method for the containment base mat. Since no piping systems are located at, and would not use, the base mat spectra for analysis, it was concluded the elastic half-space method would be used for the reevaluation because that would be conservative. The time history used for this comparison was the original design time history used in the original design of the plant, along with the original damping values.

The same floor response spectra were generated for the Regulatory Guide 1.60 (Reference 33) requirements anchored at 0.15g, along with the Regulatory Guide 1.61 (Reference 34) damping values for comparison with the original earthquake input requirements.

The time history and the damping values are considered as a consistent set of design parameters.

The comparison of the original FSAR design requirements and the Regulatory Guide 1.60 and 1.61 set of values shows that the responses are very consistent, and that the original design requirements are adequate.

The ground-response spectra at the base of the reactor containment structure were calculated and plotted using SHAKE. The response spectra were calculated for three soil profiles, represented by the average low-strain shear modulus, Gmax, calculated from seismic cross-hole surveys, Gmax plus 50%, and Gmax minus 50%. The spectra for each soil profile are plotted on Figures 15A-1, 15A-2, and 15A-3, respectively. Also plotted on these figures is the envelope for 0.5% damping, as presented on Figure 2.5-6.

A study of the effects of the variation of the soil properties was undertaken. The response spectra for the three locations in the containment building were computed for five variations of the soil properties. Variation one considered the computed strain dependent properties using the best estimate of the in-situ properties as input to computer code SHAKE; variation two used the in-situ properties plus 50% as input to the computer code SHAKE; variation three used the in-situ properties minus 50% as input to the computer code SHAKE; variation four considered the first iteration value of the computer code SHAKE using the in-situ properties as input; and variation five used the measured values (low strain) of the soil properties. This study indicated that the response of the structure to the variations in the soil properties is essentially limited to the amplitude of the floor response spectra. It was determined that an increase of the values of the response spectra already used in piping stress calculations by a factor of 1.50 would be acceptable. This increase in the acceleration value for the floor response spectra results in a conservative reanalysis.

Revision 52Updated Online 09/30/20 SPS UFSAR 15A-20 To further verify that this increase (1.5) is conservative, the NRC staff conducted an independent study of the variation of soil properties used in the dynamic analyses. First the staff confirmed the adequacy of the average soil properties selected by Vepco and then considered parametric studies of these properties. The results of this effort indicated that a variation of +/-25%

for the input shear modulus (G max ) would accommodate uncertainties in the in-situ soil properties. The results of this variation appear to bound the possible range in soil properties based on staff experience with other site studies. Therefore, Vepcos studies for +/-50% and the increase (1.5 factor) in the response spectra are conservative.

Because the soil shear moduli used in the generation of amplified floor response spectra depend upon the level of strain induced by earthquake motion, the amplified floor response spectra are not in direct proportion to the maximum ground acceleration. Therefore, an investigation of the effects of earthquakes smaller than the design-basis earthquake was also undertaken. For the purpose of this study, amplified floor response spectra were computed for various average strain compatible shear moduli, each due to a peak horizontal ground acceleration ranging from 0.15 to 0.05g. Vepco has provided the resulting family of amplified floor response spectra at the operating floor, which show the design-basis earthquake spectrum to envelope the other spectra due to smaller earthquakes. This demonstrated that the effects of design-basis earthquake are not exceeded by those of smaller earthquakes.

The computer codes used in the reanalysis for the soil structure interaction were:

1. SHAKE

2. PLAXLY

3. REFUND

4. KINACT

5. FRIDAY The computer code SHAKE is a public domain program and was used to compute only the strain-dependent properties of the supporting soil under the structures. Because this code was only used to compute soil properties, no further verification was necessary.

The computer code PLAXLY is a proprietary code and was qualified by comparison to the existing public domain computer code FLUSH. Amplified response spectra for the containment operating floor computed by both codes were compared.

The computer code REFUND computes the frequency dependent compliance functions for a multi-layered elastic half-space. This code is a proprietary code and was qualified by comparing the results of a sample problem with the results published in the literature.

Revision 52Updated Online 09/30/20 SPS UFSAR 15A-21 The computer code KINACT is a proprietary code and is used to compute the translation and rotation time history at the base of the structure from the design time history applied at the free ground surface. This code was qualified by comparing the results of a sample problem to the results of the computer code PLAXLY.

The computer code FRIDAY uses the results of REFUND and KINACT to compute the floor response spectra for each mass point in the mathematical model of the structure. The code is a proprietary program and was qualified by comparing the results of a sample problem with the results of the public domain program STARDYNE.

Additional soil-structure interaction analysis was performed for the Service Building, Safeguards Building, emergency diesel generator rooms, and Containment Spray Pump House to develop in-structure response spectra (ISRS) at certain spectral damping values that were not originally developed for these buildings. A discussion of the methods and criteria used in developing these ISRS was provided to the NRC in Reference 42, as part of the resolution to Generic Letter 87-02 (Reference 39) and Unresolved Safety Issue A-46. These methods and criteria were found adequate and acceptable in the NRCs Safety Evaluation, as indicated in Reference 43. This analysis utilized synthetic time histories as free-field motions. Three synthetic time-histories were developed such that their spectra at 5% damping closely matched the corresponding UFSAR horizontal and vertical ground spectra for Surry, consistent with Figures 2.5-5 and 2.5-6 for OBE and DBE respectively. A very close fit was reached to ensure that no lack of energy occurs at any frequency of interest. The three time histories were statistically independent. Soil and structures were modeled in detail, as discussed below.

The low strain soil properties were obtained from Section 2.4. Industry Standard Code SHAKE was used to perform dynamic analyses of the soil profile to generate the strain compatible soil properties. To account for uncertainties in the soil properties, three low strain soil properties were considered for each seismic input, in accordance with the recommendations in the Standard Review Plan (SRP - NUREG 0800). They are: best estimate, lower bound (shear modulus equal to half the best estimate shear modulus), and upper bound (shear modulus equal to twice the best estimate shear modulus). Thus, three strain compatible soil profiles were developed consistent with soil strains induced by the Housner input.

The structures were modeled as three-dimensional stick models with lumped mass and with six degrees of freedom at each node. Eccentricities were explicitly considered at each modeled elevation to account for the effects of torsion and rocking. The damping values were in accordance with Table 15.2-2. For the cases where different portions of the structures were assigned different damping, composite modal damping values were generated based on the strain energy weighted approach.

For each structure, the proper foundation embedment was considered and frequency dependent impedance and scattering functions were calculated for each strain compatible soil case. In addition, the deconvolved time-histories at the foundation levels were verified according

Revision 52Updated Online 09/30/20 SPS UFSAR 15A-22 to the recommendation of the SRP such that their response spectra (envelop of three soil cases) are not less than sixty percent of the surface spectra. The building models were used together with the proper impedance and scattering functions and for each strain compatible soil case, the three orthogonal time-histories were applied individually. The In Structure Response Spectra (ISRS) in this effort were developed at each elevation for each structure, and peak broadened +15% and

-15% to account for uncertainties and variabilities in the structural frequencies, in accordance with Regulatory Guide 1.122.

Additional information on soil-structure interaction can be found in Chapter 2.

15A.4 MOVEMENT OF REACTOR COOLANT SYSTEM COMPONENTS The criterion for movement of the reactor pressure vessel, under the worst combination of loads, i.e., normal plus the design-basis earthquake plus reactor coolant pipe rupture loads1, is that the movement of the reactor vessel will not exceed the clearance between a reactor coolant pipe and the surrounding concrete.

The relative motions between reactor coolant system components will be controlled by the structures that are used to support the reactor pressure vessel, steam generators, pressurizer and reactor coolant pumps.

Piping runs that are external to the plant or between buildings and that would affect the health and safety of the public are dynamically stress analyzed. Necessary earthquake stops, constraints, or anchors are judiciously located to withstand motion, but allow for thermal movement.

The dynamic seismic stresses are calculated using the appropriate operational-basis earthquake and design-basis earthquake response spectrum. The design criteria for the analysis of Class I piping systems are in accordance with the code requirements of ANSI B31.1. Pressure, deadload, thermal, and seismic pipe stresses are combined in accordance with the code.

The structures to which the piping is attached are also designed to withstand these loads.

Included in the pipe stress analysis are horizontal and vertical differential motion caused by rotation, translation, and flexure of the respective structures assumed to be out of phase with each other, plus the relative motion from earthquake orbital displacement of the founding soil.

In most cases, piping runs are stress-analyzed by system, thus determining the effect of branch lines joining the main run or other piping.

1. As discussed in Section 15.6.2, it is no longer necessary to consider the dynamic effects of a postulated rupture of the primary reactor coolant loop piping. However, pipe ruptures of reactor coolant branch lines, the main steam lines are still postulated.

Revision 52Updated Online 09/30/20 SPS UFSAR 15A-23 In the analysis of piping running between structures, and to different elevations within the same structure, consideration has been given to differential motion of the piping supports and anchors.

The building displacements caused by an earthquake, which include rocking and translation of the structures, as well as the relative motion from orbital displacement of the foundation soil, are evaluated at the elevation or elevations at which the piping is supported. These displacements are then applied to the supports and anchors as external movements to the piping system, and stresses are calculated.

For the analysis of piping supported by different structures, an out-of-phase condition is always assumed, and the direction of the displacements applied to the piping supports and anchors are chosen to reflect the out-of-phase condition to yield the most conservative results.

15A.5 TESTS TO DEMONSTRATE THE CONSERVATISM OF THE LIMIT CURVES Tests performed at Westinghouse Material Testing Laboratory in Pittsburgh demonstrate the conservatism of the limit curves presented in WCAP-5890, Revision 1 (Reference 35). Carbon steel and stainless steel pipes have been tested under various combinations of axial and transverse loads to determine failure loads. Specimens about 1.5 foot long have been cut from 1.5-inch nominal diameter Schedule 160 pipes. The materials employed were SA 106B carbon steel and Type 304 stainless steel. These specimens were kept internally pressurized to 3000 psia for the entire duration of the tests. Tables 15A-3 and 15A-4 summarize the tests that have undergone evaluation and the results of this evaluation.

Standard ASTM tensile specimens have been modeled from pieces of the test pipes and stress-strain curves determined. These curves have been conservatively approximated with trapezoidal stress-strain curves as indicated in WCAP-5890, Revision 1. The limit curves for both SA 106B carbon steel and Type 304 stainless steel for the test conditions have been calculated and are reported in Figures 15A-4 and 15A-5, respectively. The experimental points, i.e., stress intensities versus axial stress as listed in Tables 15A-3 and 15A-4 are shown in Figures 15A-4 and 15A-5. Also shown in these figures are the limit curves as calculated by the use of the trapezoidal stress-strain curves up to the ultimate stress. Comparisons between the experimental points and the design limit curves show the conservatism of the latter.

15A.5.1 Westinghouse Topical Reports WCAP-5890, Rev. 1, has been replaced by WCAP-7287 (Reference 36). The revisions affected limits for the combination of normal loads plus design-basis earthquake loads plus pipe rupture loads associated with a loss-of-coolant accident. The changes reflected agreement with the staff of the Atomic Energy Commission (AEC) Division of Reactor Licensing on the stress

Revision 52Updated Online 09/30/20 SPS UFSAR 15A-24 limits for the above-mentioned load combinations. Details of the manner in which the revisions were developed are as follows:

1. Material data used to develop stress-strain curves.

Typical stress-strain curves of type 304 stainless steel (Figure 15A-6), Inconel 600 (Figure 15A-7) and SA 302B low alloy steel (Figure 15A-8) at 600°F were generated from tests using graphs of applied load versus cross-head displacement as automatically plotted by the recorder of the tensile test apparatus. The scale and sensitivity of the test apparatus recorder ensured accurate measurement of the uniform strain.

For materials other than these three, stress-strain curves were developed by conservative use of pertinent available material data (i.e., lowest values of uniform strain and initial strain hardening). Where the available data were not sufficient to develop a reliable stress-strain curve, three standard ASTM tensile tests of the material in question were performed at design temperature. These data were conservatively applied in developing a stress-strain curve as described above.

2. The ordinate (stress) of the stress-strain curves was normalized to the measured yield strength.

3. Twenty percent of uniform strain as defined on the curve developed under Item 1 was used as the allowed membrane strain.

4. The normalized stress ratio was established at 20% of uniform strain on the normalized stress-strain curves developed under item 2.

5. The value of the membrane stress limit was established.

6. The normalized stress ratio in item 4 was multiplied by the applicable code yield strength at the design temperature to get the membrane stress limit. The actual physical properties as determined from standard ASTM tensile tests on specimens from the same heats was allowed as an alternate method of determining the membrane stress limit. Sufficient documentation was provided to support the actual material properties used.

7. Limit curves for the combination of local membrane and bending stresses were developed.

The limit curves were developed by using the analytical approach presented in WCAP-5890, Revision 1, and the stress-strain curve up to the membrane stress limit as developed under item 5. These limit curves were within the limit curves discussed with the staff of the AEC Division of Reactor Licensing during meetings on November 30 and December 1, 1967, for the same materials.

15A.5.2 Framatome Computer Programs (Unit 1 only)

This section describes computer programs that were used by Framatome ANP for the dynamic and static analysis of Class 1 equipment and components during the process of qualifying the Unit 1 replacement reactor vessel closure head to ASME Section III requirements.

Revision 52Updated Online 09/30/20 SPS UFSAR 15A-25 These computer programs meet the requirements of the Dominion and Framatome ANP software validation programs. The validation program meets the requirements of 10 CFR 50 Appendix B, ASME NQA 1, and ANSI N45.2. The software validation compliance was verified during an onsite quality audit of the replacement closure head vendor. Audit results and objective evidence of the software validation are available in the Framatome ANP audit file. These programs provide results that are essentially the same or more conservative than the analyses of record.

15A.5.2.1 BWSPAN BWSPAN (Reference 63) is designed to perform analysis in accordance with the ASME Boiler and Pressure Vessel Code,Section III Nuclear Power Plant Components and the ANSI B31.1 Power Piping Code. This code has been specifically used for evaluating the configuration of the RVLIS piping routed from the closure head up to and including RX Vessel Vent Line to RVLIS Isolation Valve, 1-RC-603, Rx Vessel Vent Line to RVLIS Isolation Vent Valve, 1-RC-36 (including associated drain valve 1-RC-186) and to a location in the run of the pipe just upstream of valve 1-RC-185.

15A.5.2.2 BIJLAARD BIJLAARD (Reference 64) is designed to calculate local stresses in a cylindrical or spherical shell induced by a nozzle or support.

15A.5.2.3 FERMETURE FERMETURE (Reference 65) is designed to calculate the loadings used for the closure analysis. FERMETURE calculates the stud load components for a given set of temperature and pressure values. Additionally, FERMETURE verifies the leak tightness of the vessel closure.

15A.5.2.4 SYSTUS SYSTUS (Reference 66) is designed to analyze the thermal-mechanical behavior of beams and solid structures in two or three dimensions.

15A.5.2.5 RCCM-ASME RCCM-ASME Program (Reference 67) is a special postprocessor of SYSTUS that allows manipulation of SYSTUS results for stress analyses in accordance with the rules defined by the ASME Code Section III including stresses linearization, usage factor calculation and thermal ratchet analysis.

Revision 52Updated Online 09/30/20 SPS UFSAR 15A-26 15A.6 REACTOR COOLANT LOOP (RCL) PIPING REANALYSIS SUBSEQUENT TO LEAK BEFORE BREAK AND SNUBBER ELIMINATION Table 15A-5 identifies the maximum level of stress as a percentage of the Code allowable stress for the analysis that was performed for implementation of Leak Before Break (LBB), which allowed for the removal of large snubbers connected to the primary loop piping. Reanalysis of the Reactor Coolant Loop piping and supports was performed subsequent to the implementation of LBB to support implementation of several plant changes and refinement of analytical modeling, which include:

• Measurement Uncertainty Recapture (MUR) power uprate

• Implementation of the 15 x 15 Upgrade Fuel Design

• Nuclear Safety Advisory Letter NSAL-11-2 (Reference 71) concerning the assumed stiffness of the Reactor Vessel Lower Radial Keys.

The resultant updated stresses as a percentage of the Code allowable stress and factors of safety for supports are shown in Table 15A-8 and Table 15A-9 respectively.

In the stress reanalyses performed by Westinghouse, and documented in References 69 and 70, two additional RCL branch line pipe break cases, consisting of RHR Suction and Accumulator Injection line breaks, were added, which were not analyzed previously.

Addition of these postulated breaks increased the faulted stresses. However, the recalculated stresses still remain below the code allowable stress of 2.4 Sh. The recalculated stresses for analyzed loading conditions are shown in Table 15A-8. The change only affects the faulted condition evaluation. All the other stresses (Pressure, Deadweight, Thermal, Seismic OBE and DBE) are unchanged. The recalculated margins for pipe supports based upon Shaw calculation 13019801-P-0001 (Reference 71) are shown in Table 15A-9. The stated change affects only the faulted loads due to the two additional pipe break cases analyzed. There is no change in other loads, including seismic OBE and DBE loads.

15A REFERENCES

1. R. A. Wiesemann, R. E. Tome, and R. Salvatore, Ultimate Strength Criteria to Ensure No Loss of Function of Piping and Vessels under Earthquake Loadings, WCAP-5890, Revision 1.

2. TID-7024, Chapter 6.

3. Seismic Testing of Electrical and Control System Equipment, WCAP-7397-L.

4. NRC Generic Letter 84-04, Safety Evaluation of Westinghouse Topical Reports Dealing with Elimination of Postulated Pipe Breaks in PWR Primary Main Loops, dated February 1, 1984.

Revision 52Updated Online 09/30/20 SPS UFSAR 15A-27

5. NRC Generic Letter 87-11, Relaxation In Arbitrary Intermediate Pipe Rupture Requirements, dated June 19, 1987.

6. Westinghouse Topical Report WCAP 9558, Revision 2, Mechanistic Fracture Evaluation of Reactor Coolant Pipe Containing a Postulated Circumferential Throughwall Crack, May 1981.

7. Westinghouse Topical Report WCAP 9787, Tensile and Toughness Properties of Primary Piping Weld Metal for Use in Mechanistic Fracture Evaluation, in May 1981.

8. Letter from Vepco to NRC,

Subject:

Request for Partial Exemption from General Design Criterion 4, dated November 5, 1985 (Serial No. 85136).

9. Letter from Vepco to NRC,

Subject:

Request for Partial Exemption from General Design Criterion 4 - Supplement, dated December 3, 1985 (Serial No. 85-136A).

10. Letter from Vepco to NRC,

Subject:

Partial Exemption from General Design Criterion 4 -

Request for Additional Information, dated December 27, 1985 (Serial No. 85-136B).

11. Letter from Vepco to NRC,

Subject:

Partial Exemption from General Design Criterion 4 -

Request for Additional Information, dated January 14, 1986 (Serial No. 85-136C).

12. Amendment to General Design Criterion 4 (GDC-4), 10 CFR Part 50, Appendix A, published in Federal Register 51 FR 12502, effective May 12, 1986.

13. Letter from Vepco to NRC,

Subject:

Proposed License Amendment - GDC 4, dated April 30, 1986 (Serial No.86-245).

14. Letter from NRC to Vepco transmitting Surry Unit 1 and 2 License Amendments No. 108 and related safety evaluations, dated June 16, 1986.

15. Letter from Vepco to NRC,

Subject:

Generic Letter 87-11, dated September 12, 1988 (Serial No.88-371).

16. U.S. Nuclear Regulatory Commission, Pipe Support Base Plate Designs Using Concrete Expansion Anchor Bolts, IE Bulletin 79-02, March 8, 1979 (Revision 1), August 20, 1979 (Revision 2).

17. Teledyne Engineering Services, Summary Report. Generic Response to USNRC I&E Bulletin No. 79-02, Base Plate/Concrete Expansion Anchor Bolts, Technical Report TR-3501-1, Revision 1.

18. Letter from Vepco to NRC,

Subject:

IE Bulletin 79-02, dated December 7, 1979.

19. Letter from W. C. Spencer, Vepco to J. P. OReilly, NRC,

Subject:

Supplemental Response to IE Bulletin 79-02, Revision 2, Surry Power Station Units 1 and 2, dated January 30, 1980.

Revision 52Updated Online 09/30/20 SPS UFSAR 15A-28

20. Letter from W. C. Spencer, Vepco, to J. P. OReilly, NRC,

Subject:

Final Report on IE Bulletin 79-02, Revision 2, Surry Power Station Unit 2, dated August 4, 1980.

21. U.S. Nuclear Regulatory Commission, Seismic Analyses for As-Built Safety-Related Piping Systems, IE Bulletin 79-14, July 2, 1979.

22. Letter from H. R. Denton, NRC, to W. L. Proffit, Vepco,

Subject:

Show Cause Orders for Surry Power Station, Units 1 and 2, dated March 13, 1979.

23. Letter from W. C. Spencer, Vepco, to H. R. Denton, NRC, transmitting Report on the Reanalysis of Safety-Related Piping Systems, Surry Power Station Unit 1, dated June 5, 1979.

24. Letter from H. R. Denton, NRC, to W. L. Proffit, Vepco,

Subject:

Order Lifting the Suspension of Facility Operation Required by the Order to Show Cause Dated March 13, 1979, for the Surry Power Station, Unit 1, dated August 22, 1979.

25. Letter from W. C. Spencer, Vepco, to H. R. Denton, NRC, transmitting Report on the Reanalysis of Safety-Related Piping Systems, Surry Power Station Unit 2, dated February 22, 1980.

26. Letter from H. R. Denton, NRC, to W. L. Proffit, Vepco,

Subject:

Order Lifting the Suspension of Facility Operation Required by the Order to Show Cause Dated March 13, 1979, for the Surry Power Station, Unit 2, March 26, 1980.

27. U.S. Nuclear Regulatory Commission, Combining Modul Responses and Spatial Components in Seismic Response Analysis, Regulatory Guide 1.92, February 1976.

28. Seismic System Analysis, Standard Review Plan 3.7.2, November 24, 1975.

29. Letter from W. C. Spencer, Vepco, to V. Stello, NRC,

Subject:

Response to NRC Information Request on Soil-Structure Interaction, dated May 2, 1979.

30. Letter from W. C. Spencer, Vepco, to D. G. Eisenhut, NRC,

Subject:

Operating-Basis Earthquake Reanalysis of Piping Systems, Surry Power Station Units 1 and 2, dated September 13, 1979.

31. Letter from NRC to Vepco,

Subject:

Soil-Structure Interaction Position, dated March 25, 1979.

32. Letter from D. G. Eisenhut, NRC, to W. L. Proffit, Vepco,

Subject:

Soil-Structure Interaction Conclusions for Surry Power Station Units 1 and 2, dated November 15, 1979.

33. U.S. Nuclear Regulatory Commission, Design Response Spectra for Seismic Design of Nuclear Power Plants, Regulatory Guide 1.60, December 1973.

34. U.S. Nuclear Regulatory Commission, Damping Values for Seismic Design of Nuclear Power Plants, Regulatory Guide 1.61, October 1973.

Revision 52Updated Online 09/30/20 SPS UFSAR 15A-29

35. WCAP-5890, Revision 1.

36. WCAP-7287.

37. U.S. NRC Bulletin No. 88-11: Pressurizer Surge Line Thermal Stratification, USNRC, December 20, 1988.

38. Virginia Power Letters Serial Nos. 89-006A dated May 3, 1989 and 89-006B dated November 13, 1989 to United States Nuclear Regulatory Commission.

39. NRC Generic Letter 87-02, Verification of Seismic Adequacy of Mechanical and Electrical Equipment in Operating Reactors, Unresolved Safety Issue (USI) A-46, February 19, 1987.

40. Generic Implementation Procedure (GIP) for Seismic Verification of Nuclear Plant Equipment, Revision 3, Updated 05/16/97, prepared by Seismic Qualification Utility Group (SQUG) and sent to the NRC by letter dated May 16, 1997.

41. Letter from NRC to Vepco,

Subject:

Computer Program for Pipe Stress Analysis of ASME Class 2 and 3 Piping, dated June 16, 1993. (Serial No.93-395)

42. Letter from W. L. Stewart, Virginia Power, to U.S. Nuclear Regulatory Commission, September 18, 1992, Virginia Electric and Power Company, Surry Power Station Units 1 and 2, North Anna Power Station Units 1 and 2, Response to Supplement 1 of Generic Letter 87-02 SQUG Resolution of Unresolved Safety Issue A-46, Serial Number 92-384.

43. NRC letter to Virginia Power, Serial No.92-763, Safety Evaluation of NAPS Units 1 and 2 and SPS Units 1 and 2, 120 day response to Supplement No. 1 to Generic Letter 87-02, November 20, 1992.

44. Supplemental Safety Evaluation Report No. 2 (SSER No. 2) on Seismic Qualification Utilities Groups Generic Implementation Procedure, Revision 2, Corrected February 14, 1992, for Implementation of GL 87-02 (USI A-46), Verification of Seismic Adequacy of Equipment in Older Operating Nuclear Plants, U.S. Nuclear Regulatory Commission, May 22, 1992.

45. Supplemental Safety Evaluation Report No. 3 (SSER No. 3) on the Review of Revision 3 to the Generic Implementation Procedure for Seismic Verification of Nuclear Power Plant Equipment, Updated May 16, 1997, U.S. Nuclear Regulatory Commission, December 4, 1997.

46. Virginia Power Letter to Document Control Desk (NRC), Surry Power Station, Units 1 and 2, Summary Report for Resolution of Unresolved Safety Issue (USI) A-46, November 26, 1997.

47. NRC letter to Virginia Power, Surry Power Station, Units 1 and 2 - Request for Additional Information Regarding Seismic Qualification of Mechanical and Electrical Equipment (GL 87-02), February 23, 1998.

Revision 52Updated Online 09/30/20 SPS UFSAR 15A-30

48. Virginia Power letter to NRC, Surry Power Station, Units 1 and 2 - Response to Request dated February 23, 1998 for Additional Information on USI A-46 Summary Report (GL 87-02),

April 22, 1998.

49. NRC letter to Virginia Power, Surry Power Station, Units 1 and 2 - Request for Additional Information Regarding Seismic Qualification of Mechanical and Electrical Equipment (GL 87-02), September 11, 1998.

50. Virginia Power letter to NRC, Surry Power Station, Units 1 and 2 - GL 87 Response to Request for Additional Information dated September 11, 1998 on USI A-46 Summary Report, January 5, 1999.

51. Virginia Power Letter to Document Control Desk (NRC), Surry Power Station, Units 1 and 2, Completion of Outlier Resolution - USI A-46 Program, GL 87 Verification of Seismic Adequacy of Mechanical and Electrical Equipment, February 4, 2000.

52. NRC Letter to Virginia Power, Serial Number 00-398, transmitting Safety Evaluation by the Office of Nuclear Reactor Regulation Evaluation of Virginia Electric and Power Company Response to Supplement No. 1 to Generic Letter 87-02, Surry Units 1 and 2, July 25, 2000.

53. NRC (Brian Sheron) letter to SQUG (Neil Smith), Clarification of the Staffs Position Regarding Incorporation of the GIP Method as a Revision to the Plant Licensing Basis, June 19, 1998.

54. NRC letter to SQUG, Review of Seismic Qualification Utility Groups Report on the Use of the Generic Implementation Procedure for New and Replacement Equipment and Parts, June 23, 1999.

55. Implementation Guidelines for Seismic Qualification of New and Replacement Equipment/Parts (NARE) Using the Generic Implementation Procedure (GIP), Revision 4, Prepared for EPRI and SQUG by MPR Associates, July 2000.

56. Guideline for the Seismic Technical Evaluation of Replacement Items (STERI) for Nuclear Power Plants, EPRI Report NP-7484, February 1993.

57. Generic Seismic Technical Evaluations of Replacement Items (GSTERI) for Nuclear Power Plants, EPRI Report TR-104871, May 1995.

58. Generic Seismic Technical Evaluations of Replacement Items for Nuclear Power Plants - Item Specific Evaluations, including Supplement 1, EPRI Report SU-105849, September 1997.

59. Critical Characteristics for Acceptance of Seismically Sensitive Items (CCASSI), EPRI Report TR-112579, September 2000.

60. EPRI NP-6041-SL, Revision 1, A Methodology for Assessment of Nuclear Plant Seismic Margin, August 1991.

Revision 52Updated Online 09/30/20 SPS UFSAR 15A-31

61. BNL 52361, Seismic Design and Evaluation Guidelines for the Department of Energy High Level Waste Storage Tanks and Appurtenances, Brookhaven National Laboratory, Upton, New York, October 1995.

62. Dominion General Engineering Nuclear Standard STD-GN-0038, Seismic Qualification of Equipment.

63. BWSPAN - Pipe Stress Analysis, Version 10.2 (2A-4 Rev. T), Users Manual, Developed by Framatome ANP.

64. BIJLAARD - Local Shell Stress Calculation, Version 1.0, Developed by Framatome ANP.

65. FERMETURE - Closure Analysis, Version AXP 2.2, Developed by Framatome ANP.

66. SYSTUS - General Finite Element System, Developed by Framatome ANP Module SYSNUKE Version 233 AXP 3 for Closure Head Stress and Fatigue Analysis Module NUKE Version 1.03 SGI for Adapter Tube and Vent Pipe Stress and Fatigue Analysis.

67. RCCM-ASME - Pressure Retaining Boundary Analysis, according to ASME Code Developed by Framatome ANP, Version AXP 2.3.12 for Closure Head Stress and Fatigue Analysis Version 2.3.13 SGI and 2.3.14 SGI for Adapter Tube and Vent Pipe Stress and Fatigue Analysis.

68. PIPESTRESS Computer Software, DST Computer Services S.A.

69. Surry 1 and 2 RCL Piping Stress Analysis and Surge Stratification Analysis for RTSR, Westinghouse Calculation No. CN-PAFM-10-36, Rev. 1, January 2013.

70. Factor of Safety for Primary Components and Supports with Reduced Snubber Configuration, Calculations No. 14937.24-NMB-333-GA, Rev. 0, Addendum 00B, August 2013

71. Impact of Change in Lower Radial Key Stiffness Value, Westinghouse Nuclear Safety Advisory Letter NSAL-11-2, June 2011.

Revision 52Updated Online 09/30/20 SPS UFSAR 15A-32 Table 15A-1 LOADING CONDITIONS AND STRESS LIMITS Pressure Vessels Loading Conditions Stress Limits

1. Normal conditions a. Pm Sm

b. Pm(or PL)+PB 1.5Sm (Note 1)

c. Pm(or PL)+PB+Q 3.0Sm (Note 2)

2. Upset conditions a. Pm Sm

b. Pm (or PL)+PB 1.5Sm (Note 1)

c. Pm(or PL)+PB+Q 3.0Sm (Note 2)

3. Emergency conditions a. Pm 1.2Sm, or Pm Sy, whichever is larger

b. Pm (or PL)+PB 1.5(1.2Sm), (Note 3) or Pm (or PL)+PB 1.5 (Sy) (Note 3) whichever is larger

4. Faulted conditions Design limit curves of WCAP-5890, Rev. 1, (Note 4) as modified by Section 15A.5.1 where:

Pm = primary general membrane stress intensity PL = primary local membrane stress intensity PB = primary bending stress intensity Q = secondary stress intensity Sm = stress intensity value from ASME B&PV Code,Section III, Nuclear Vessels Sy = minimum specified material yield Pressure Piping (Note 6)

Loading Conditions Stress Limits

1. Normal conditions Pm S

2. Upset conditions Pm 1.2S

3. Emergency conditions Pm 1.8S

4. Faulted conditions Design limit curves of WCAP-5890, Revision (Note 4) 1, as modified by Section 15A.5.1 where Pm = principal stress S = Allowable stress from USAS B31.1, Code for Power Piping 7 Equipment Supports

1. Normal conditions Within working limits

2. Upset conditions Within working limits

Revision 52Updated Online 09/30/20 SPS UFSAR 15A-33 Table 15A-1 (CONTINUED)

LOADING CONDITIONS AND STRESS LIMITS Equipment Supports (continued)

3. Emergency conditions Within material yield strength after load (Note 5) redistribution

4. Faulted conditions Within material yield strength after load (Note 5) redistribution Note 1: The limits on local membrane stress intensity (PL 1.5Sm) and primary membrane plus primary bending stress intensity (PM (or PL) + PB 1.5Sm) need not be satisfied at a specific location if it can be shown by means of limit analysis or by tests that the specified loadings do not exceed 2/3 of the lower bound collapse load as per paragraph N-417.6(b) of the ASME B&PV Code,Section III, Nuclear Vessels.

Note 2: In lieu of satisfying the specific requirements for the local membrane (PL 1.5Sm) or the primary plus secondary stress intensity (PL + PB + Q 3Sm) at a specific location, the structural action may be calculated on a plastic basis and the design will be considered to be acceptable if shakedown occurs, as opposed to continuing deformation, and if the deformation which occur prior to shakedown do not exceed specified limits, as per paragraph N-417.6(a)(2) of the ASME B&PV Code,Section III, Nuclear Vessels.

Note 3: The limits on local membrane stress intensity (PL 1.5Sm) and primary membrane plus primary bending stress intensity (Pm (or PL) + PB 1.5Sm) need not be satisfied at a specific location if it can be shown by means of limit analysis or by test that the specified loadings do not exceed 120% of 2/3 of the lower bound collapse load as per paragraph N-417.10(c) of the ASME B&PV Code,Section III, Nuclear Vessels.

Note 4: As an alternate to the design limit curves that represent a pseudo plastic instability analysis, a plastic instability analysis may be performed in some specific cases considering the actual strainhardening characteristics of the material, but with the yield strength adjusted to correspond to the tabulated value at the appropriate temperature in Table N-424 or N-425, as per paragraph N-417.11c of the ASME B&PV Code,Section III, Nuclear Vessels. These specific cases will be justified on an individual basis.

Note 5: Higher stress values can be adopted if a valid limit or plastic instability analysis of the support and supported component/system is performed.

Note 6: As required by NRC Bulletin 88-11, pressurizer surge line is re-evaluated in accordance with the requirements of ASME Boiler and Pressure Vessel Code,Section III, Subsection NB, 1986 with addenda through 1987 incorporating high cycle fatigue.

Revision 52Updated Online 09/30/20 SPS UFSAR 15A-34 Table 15A-2 MINIMUM MARGINS OF SAFETY Loading Conditions Material Upset Emergency Faulted Conditions Conditions Conditions (Note 1)

SA302 Grade B 200% 150% 27%

Inconel 600 228% 172% 43%

316 SST 222% 169% 60%

A212 Grade B 346% 272% 55%

Note: Based upon the limit curves computed using Section 15A.5.1.

Revision 52Updated Online 09/30/20 Table 15A-3 TESTS AND TEST RESULTS ON SA 106B CARBON STEEL PIPE SPECIMENS (INTERNAL PRESSURIZATION = 3000 PSIA)

Pseudo-elastic Pseudo-elastic Stress Pseudo-elastic Axial Bending Stress Intensity Stress Normalized Normalized to the Normalized to the Strain Percent to the Yield Stress Yield Stress Yield Stress (gauge length)

Pure tension (no weld) 1.736 0.0 1.770 22.61% (12 in.)

1.840 0.0 1.847 22.32% (12 in.)

Pure bending a 0.10 >2.348 2.382 >35.4X (1 in.)

Tension + bending (no weld) >1.375 >1.030 2.440 >7.75% (6 in.)

>1.585 >0.565 2.180 --

>1.845 >0.266 2.145 --

Compression + bending (no weld) >1.130 1.08 2.410 --

Pure tension (circumf. weld) 1.852 0.0 1.886 20.05% (12 in.)

Pure bending (circumf. weld) 0.10 >2.580 2.614 >30.19X (1.5 in.)

Pure bending b 0.10 2.460 2.494 14.51% (2 in.)

SPS UFSAR (rejected circumf. weld)

a. The limit capability of the test apparatus has been reached before failure of these specimens was approached.

b. This test was performed on a welded pipe specimen that has been rejected by the inspector prior to the test for gross lack of penetration in weld.

This was the only test in which the weld failed. The specimen exhibited substantial ductility prior to failure.

15A-35

Revision 52Updated Online 09/30/20 Table 15A-4 TESTS AND TEST RESULTS ON 304 STAINLESS STEEL SPECIMENS (INTERNAL PRESSURIZATION = 3000 PSIA)

Pseudo-elastic Pseudo-elastic Stress Pseudo-elastic Axial Bending Stress Intensity Stress Normalized Normalized to the Normalized to the Strain Percent to the Yield Stress Yield Stress Yield Stress (gauge length)

Pure tension (no weld) 2.495 0.0 2.53 52. 12 (12 in.)

Pure bending a +0.10 >2.91 >2.945 >30.0% (1 in.)

Tension + bending (no weld) >+2.09 >0.42 >2.55 --

Pure bending (circumf. weld) +0.10 >3 27 >3.30 >25.0% (1.5 in.)

Pure tension (circumf. weld) +2.46 0.0 2.49 44.6% (12 in.)

a. The limit capability of the test apparatus has been reached before failure of these specimens was approached.

SPS UFSAR 15A-36

Revision 52Updated Online 09/30/20 SPS UFSAR 15A-37 The following information is HISTORICAL and is not intended or expected to be updated for the life of the plant.

Table 15A-5 LEVEL OF STRESS AS A PERCENTAGE OF CODE ALLOWABLE STRESS Loading Hot Leg Crossover Leg Cold Leg Code Allowable Stress Thermal 38.6% 15.6% 7.4% SA Pressure + Deadweight 68.0% 48.7% 53.3% 1. 0Sha Pressure + Deadweight + 65.5% 65.0% 60. 0% 1. 2Sha OBE Pressure + Deadweight + 49.6% 51.1% 48.9% 1. 8Sha DBE In addition, the pipe Stresses in the primary reactor coolant loop piping under combined accident loadings from pressure, deadweight, plus SRSS of design-basis earthquake and controlling pipe ruptures are less than 1.8Sh, which is conservative both respect to the allowables per Section 15A.5 and to 2. 4Sh which is permitted by the current ASME Code.

a. Sh =15 Ksi

Revision 52Updated Online 09/30/20 SPS UFSAR 15A-38 The following information is HISTORICAL and is not intended or expected to be updated for the life of the plant.

Table 15A-6 FACTORS OF SAFETY FOR COMPONENT SUPPORTS UNDER DESIGN-BASIS SEISMIC AND NORMAL OPERATING LOADS SURRY UNITS 1 AND 2 Component Factor of Safetya Original Modified Designb Designc Final Designd Steam Generator Shell >20.0 >20.0 >20.0 Steam Generator Upper Support Component 19.2 15.9 13.7 Upper Guide 7.3 4.6 3.3 Snubber 14.2 14.2 6.3 Steam Generator Lower Support Hanger Rod 1.8 1.8 1.8 Swivel End Coupling 16.3 13.4 12.2 Steam Generator Foot Vertical Force 2.7 2.8 2.8 Tangential Force 16.0 12.4 12.4 Reactor Coolant Pump Foot Vertical Force 5.5 5.2 5.3 Tangential Force 15.4 15.8 15.9 Radial Force 15.4 15.0 12.8 Reactor Coolant Pump Support Upper Vertical 5.5 5.0 5.0 Upper Horizontal 5.0 5.0 5.0 Lower Vertical 3.6 3.6 3.6 Lower Diagonal 4.6 4.6 4.6

a. Factor of Safety - (Allowable Load)/(TotaL Load of Deadweight, Pressure, Thermal and DBE)

b. Original design incorporated ten large-bore snubber per loop for primary reactor coolant system pipe rupture loads.

c. Modified design implemented elimination of six large-bore snubbers based on leak-before-break; n the four remaining large-bore snubbers were required to carry loads of main steam line rupture.

d. Final design incorporates only two large-bore snubbers on Steam Generator, following elimination of lateral loads due to postulated AIB of Main Steam Line.

Revision 52Updated Online 09/30/20 SPS UFSAR 15A-39 The following information is HISTORICAL and is not intended or expected to be updated for the life of the plant.

Table 15A-7 FACTORS OF SAFETY FOR COMPONENT SUPPORTS UNDER COMBINED ACCIDENT LOADS SURRY UNITS 1 AND 2 Component Factor of Safetya Original Modified Designc Designd Final Designe Steam Generator Shell 2.5 2.4 8.3 Steam Generator Upper Support Component 2.8 2.8 7.5 Upper Guide 1.1 1.1 2.6 Snubber 1.3 1.4 1.9 Steam Generator Lower Support Hanger Rod 1.7 1.7 1.7 Swivel End Coupling 2.3 2.3 2.3 Steam Generator Footb Vertical Force 4.0 4.2 4.3 Tangential Force 7.4 7.2 7.3 Reactor Coolant Pump Footb Vertical Force 11.9 11.3 10.9 Tangential Force >20.0 >20.0 >20.0 Radial Force >20.0 >20.0 >20.0 Reactor Coolant Pump Support Upper Vertical 4.5 4.3 4.2 Upper Horizontal 5.3 4.8 4.6 Lower Vertical 3.8 3.5 3.5 Lower Diagonal 3.0 3.0 3.0

a. Factor of Safety - (Allowable Load)/[TotaL Load of Deadweight, Pressure, Thermal and SRSS (DBE+Pipe Rupture)].

b. Allowable loads from Westinghouse specification are higher for pipe rupture case; this results in higher normal factors of safety for some components for this case compared with factors of safety for design basis seismic loads only (Table 15A-4).

c. Original design incorporated ten large-bore snubber per loop for primary reactor coolant system pipe rupture loads.

d. Modified design implemented elimination of six large-bore snubbers based on leak-before-break; n the four remaining large-bore snubbers were required to carry loads of main steam line rupture.

e. Final design incorporates only two large-bore snubbers on Steam Generator, following elimination of lateral loads due to postulated AIB of Main Steam Line.

Revision 52Updated Online 09/30/20 SPS UFSAR 15A-40 Table 15A-8 CALCULATED STRESS AS PERCENTAGE OF CODE ALLOWABLE STRESS (REFERENCE 69)

Code Crossover Allowable Loading Hot Leg Leg Cold Leg Stress Thermal 38.6% 15.6% 7.4% SA (unchanged) (unchanged) (unchanged)

Pressure + Deadweight 68.0% 48.7% 53.3% 1.1 Sha (unchanged) (unchanged) (unchanged)

Pressure + Deadweight + OBE 65.5% 65.0% 60.0% 1.2 Sha (unchanged) (unchanged) (unchanged)

Pressure + Deadweight + DBE 49.6% 51.1% 48.9% 1.8 Sha (unchanged) (unchanged) (unchanged)

Pressure + Deadweight + {(DBE)2 + 50.8% 53.9% 98.1% 2.4 Shb (LOCA/Pipe Rupture)2}1/2

a. Sh = 15 ksi

b. The faulted case maximum calculated stress which includes maximum stress for worst case LOCA/pipe break is compared against ASME code allowable stress of 2.4 Sh. This is acceptable under current industry practice and widely used by Westinghouse in the reanalysis of RCL piping for their plants. Based upon the material test data performed by Westinghouse, documented in WCAP-5890, Rev. 1 which was later replaced by WCAP-7287 and shown summarily in Tables 15A-3 plus 15A-4 and Figures 15A-4 plus 15A-5, the design limit curves are conservative compared to test results shown.

Revision 52Updated Online 09/30/20 SPS UFSAR 15A-41 Table 15A-9 FACTORS OF SAFETY FOR STEAM GENERATOR AND REACTOR COOLANT PUMP SUPPORTS (REFERENCE 70)

Factor of Component Safety Steam Generator Upper Support:

Component 7.5 Upper Guide 2.5 Snubber 1.5 Steam Generator Lower Support:

Hanger Rod 1.2 Swivel End Coupling 1.7 Reactor Coolant Pump Support:

Upper Vertical 3.1 Upper Horizontal 3.4 Lower Vertical 2.6 Lower Diagonal 2.2

Revision 52Updated Online 09/30/20 SPS UFSAR 15A-42 Figure 15A-1 ENVELOPE FOR 0.5% DAMPING GROUND RESPONSE SPECTRA-AVERAGE Gmax

Revision 52Updated Online 09/30/20 SPS UFSAR 15A-43 Figure 15A-2 ENVELOPE FOR 0.5% DAMPING GROUND RESPONSE SPECTRA-AVERAGE Gmax + 50%

Revision 52Updated Online 09/30/20 SPS UFSAR 15A-44 Figure 15A-3 ENVELOPE FOR 0.5% DAMPING GROUND RESPONSE SPECTRA-AVERAGE Gmax -50%

Revision 52Updated Online 09/30/20 Figure 15A-4 DESIGN LIMITS COMPARED TO EXPERIMENTAL POINTS, SA 106B CARBON STEEL SPS UFSAR 15A-45

Revision 52Updated Online 09/30/20 Figure 15A-5 DESIGN LIMITS COMPARED TO EXPERIMENTAL POINTS, 304 STAINLESS STEEL SPS UFSAR 15A-46

Revision 52Updated Online 09/30/20 Figure 15A-6 TYPICAL STRESS-STRAIN CURVE, 304 STAINLESS STEEL SPS UFSAR 15A-47

Revision 52Updated Online 09/30/20 SPS UFSAR 15A-48 Figure 15A-7 TYPICAL STRESS-STRAIN CURVE, INCONEL 600

Revision 52Updated Online 09/30/20 Figure 15A-8 TYPICAL STRESS-STRAIN CURVE, SA 302 GRADE B SPS UFSAR 15A-49

Revision 52Updated Online 09/30/20 SPS UFSAR 15A-50 Intentionally Blank