ML22112A038
| ML22112A038 | |
| Person / Time | |
|---|---|
| Site: | LaSalle, 05200002 |
| Issue date: | 01/31/1997 |
| From: | ABB Combustion Engineering |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20148A597 | List:
|
| References | |
| NUDOCS 9705090171 | |
| Download: ML22112A038 (1) | |
Text
{{#Wiki_filter:.. - .- - .- .. . . _ - . _ . - - . - . . -- i O: the System 80+ standardplant i i i l Design Control Document , !o ! Volume 3 t ,O A RIE Combustion Engineering, Inc. MnWW
1 i O i 4 i l l Copyright C 1997 Combustion Engineering, Inc., All Rights Reserved. Warning, Legal Notice and Disclaimer of Liability The design, engineering and other information contained in this document have been ] prepared by or for Combustion Engineering, Inc. in connection with its application to the j United States Nuclear Regulatory Commission (US NRC) for design certification of the j System 80+ nuclear plant design pursuant to Title 10, Code of Federal Regulations l Part $2. No use of any such information is authorized by Combustion Engineering, Inc. ! except for use by the US NRC and its contractors in connection with review and , approval of such application. Combustion Engineering, Inc. hereby disclaims all ! responsibihty and liability in connection with unauthorized use of such information. Neither Combustion Engineering, Inc. nor any other person or entity makes any warranty or representation to any person or entity (other than the US NRC in connection with its review of Combustion Engineering's application) conceming such information or its use, except to the extent an express warranty is made by Cembustion Engineering, Inc. to j its customer in a written contract for the sale of the goods or services described in this i document. Potential users are hereby warned that any such information may be I unsuitable for use except in connection with the performance of such a written contract j by Combustion Engineering, Inc. ' Such information or its use are subject to copyright, patent, trademark or other rights of Combustion Engineering, Inc. or of others, and no license is granted with respect to i such rights, except that the US NRC is authorized to make such copies as are necessary for the use of the US NRC and its contractors in connection with the Combustion Engineering, Inc. application for design certification. Publication, distribution or sale of this document does not constitute the performance of l engineering or other professional services and does not create or establish any duty of ! care towards any recipient (other than the US NRC in connection with its review of Combustion Engineering's application) or towards any person affected by this document. I l l ll For information address: Combustion Engineering, Inc., Nuclear Systems Licensing. 2000 Day Hill Road; Windsor, Connecticut 06095 9
System 80+ Design ControlDocument - Introduction Certified Design Material - 1.0 ' Introduction 2.0 System and Structure ITAAC 3.0 Non-System ITAAC 4.0 Interface Requirements
. 5.0 Site Parameters Approved Design Material - Design & Analysis 1.0 General Plant Description 2.0 Site Characteristics 3.0 Design of Systems, Structures & Components 4.0 Reactor 5.0 RCS and Connected Systems ,
6.0 Engineered Safety Features , 7.0 Instrumentation and Control 8.0 Electric Power 9.0 Auxiliary Systems 10.0 Steam and Power Conversion
-- 11.0 Radioactive Waste Management 12.0 Radiation Protection 13.0 Conduct of Operations l 14.0 Initial Test Program 15.0 Accident Analyses 16.0- Technical Specifications 17.0 Quality Assurance 18.0 Human Factors l 19.0 Probabilistic Risk Assessment I 20.0 Unresolved and Generic Safety Issues I Approved Design Material - Emergency Operations Guidelines i 1.0 Introduction j 2,0 Standard Post-Trip Actions i Diagnostic Actions !
3.0 4.0 Reactor Trip Recovery 5.0 Loss of Coolant Accident Recovery 6.0 Steam Generator Tube Rupture Recovery 7.0 Excess Steam Demand Event Recovery l 8.0 Loss of All Feedwater Recovery 9.0 Loss of Offsite Power Recovery 10.0 Station Blackout Recovery 11.0 Functional Recovery Guideline
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*1 1.7 . Drawings and Diagrams 1.7.1 D Elecidcal, li-L ' Mand Control Drawings ,;
. . . .. i The systems of interest are the Reactor Protective System (RPS), Alternate Protection System (APS), and ,
' the Engineered Safety Features Actuation System (ESFAS). These three systems provide reactor trips l and Engineered Safety Feature (ESP) systems actuation for limiting events as determined by the safety .
! analysis for the plant. Table 1.7-1 provides a listing of Safety Related Electrical, Instrumentation and Control Drawings.
. - The functional block diagrams listed in Table 1.7-1 include the RPS, ESFAS, and APS as shown in- - j Figures 7.2-12,= 7.3-3 and 7.7-12. The interface logic is shown in Figure 7.2-19. Other fit"wes at the end of Sections 7.2 and 7.3 provide more detailed logic on various portions of these systems. ! c . . Functional control logic diagrams for the Shutdown Cooling System and Safety Injection System are provided in Section 7.6. l ! Measurement Channel Block Diagrams (MCBDs), Plant Protection System (PPS) design drawings and component Functional Logic Diagrams are identified in the applicable sections. The MCBDs show all channels which are safety-related. The drawings apply to the RPS, APS, ESFAS, i and to the Post-Accident Monitoring requirements in the Approved Design Material. g 1.7.2 Piping and Instrumentation I% _ i I . Table 1.7-2 provides a list of valve identifiers used in the Approved Design Material valve lists. Piping and instrumentation diagram symbols which are used on the piping and instrumentation diagrams listed
- l. in Table 1.7-3 are presented in Figures 1.7-1,1.7-2, and 1.7-3.
i Table 1.7-1 Safety-Related Electrical, Instrumentation and Control Drawings 1 ! Drawing No. Title Section I Figure 7.21 PPS Basic Block Diagram 7.2 : Figure 7.2-2 PPS Functional Interface and Testing Diagram 7.2 ; Figure 7.2-3 = Typical PPS IAw Reactor Coolant Flow Trip Setpoint Operation 7.2 ~ Figure 7.2-4 Typical PPS Measurement Channel Functional Diagram (Pressurizer 7.2 Pressure Wide Range) Figure 7.2-5 Reed Switch Position Transmitter Assembly Schematic 7.2 Figure 7.2-6 Reed Switch Position Transmitter Cable Assemblies 7.2 ! Figure 7.2-7 Core Protection Calculator (CEA Calculators) 7.2
-( Figure 7.2-8 Ex-Core Neutron Flux Monitoring System ' 7.2 ..
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System 80+ Design ControlDocument Table 1.7-1 Safety-Related Electrical, Instrumentation and Control Drawings (Cont'd.) Drawing No. Title Section Figure 7.2-9 Reactor Coolant Pump Speed Sensors Typical for Each Reactor Coolant 7.2 Pump Figure 7.2-10 Trip Logic Calculator Functional Block Diagram 7.2 Figure 7.2-11 PPS Bistable Trip Logic Functional Block Diagram 7.2 Figure 7.212 PPS Reactor Trip System Simplified Functional Logic Diagram 7.2 Figure 7.2-13 Typical PPS Channel Functional Bistable Trip Channel Bypass 7.2 Figure 7.2-14 Typical PPS Channel Functional RPS Initiation Logic 7.2 Figure 7.2-15 Typical PPS Variable Setpoint Operation (Manual Reset) 7.2 Figure 7.2-16 PPS Testing Overlap 7.2 Figure 7.2-17 Interface and Test Processor Block Diagram 7.2 Figure 7.2-18 Typical PPS Channel Contact BistaMe Interface Diagram 7.2 Figure 7.2-19 Plant Protection System laterface Logic Diagram 7.2 Figure 7.2-20 MCBD Symbols, Notes and Abbreviations 7.2 Figure 7.2 21a RCS Loop 1 Temperatures (Narrow) MCBD 7.2 Figure 7.2 21b RCS loop 2 Temperatures (Narrow) MCBD 7.2 Figure 7.2-22a RCS Loop 1 Temperatures (Wide) MCBD 7.2 Figure 7.2-22b RCS Loop 2 Temperatures (Wide) MCBD 7.2 Figure 7.2-23a Reactor Coolant Pump Pressure MCBD 7.2 Figure 7.2 23b Reactor Coolant Pump Speed MCBD 7.2 Figure 7.2-24 Pressurizer Pressure MCBD 7.2 Figure 7.2-25 Nuclear Instrumentation MCBD 7.2 Figure 7.2-26 Containment Pressure MCBD 7.2 Figure 7.2-27a Steam Generator-1 Level (Wide) MCBD 7.2 Figure 7.2-27b Steam Generator-2 Level (Wide) MCBD 7.2 Figure 7.2-28a Steam Generator-1 Pressure MCBD 7.2 Figure 7.2-28b Steam Generator-2 Pressure MCBD 7.2 Asyvoved Desisps Metode! Introduction Page 1.7 2
System 80+ Design ControlDocument (D C/ Table 1.7-1 Safety-Related Electrical, Instrumentation and Control Drawings , I (Cont'd.) i Drawing No. Title Section Figure 7.2-29a Steam Generator-1 Level (Narrow) MCBD 7.2 Figure 7.2-29b Steam Generator-2 level (Narrow) MCBD 7.2 Figure 7.2-30 Steam Generator Primary D/P MCBD 7.2 Figure 7.3-la ESFAS Functional Logic (SIAS) 7.3 Figure 7.3-lb ESFAS Functional logic (CSAS, CIAS) 7.3 Figure 7.3 Ic ESFAS Functional Logic (EFAS 1, EFAS 2) 7.3 , Figure 7.3-Id ESFAS Functional Iegic (MSIS) 7.3 Figure 7.3-2 ESF-CCS legic Diagram for Typical Selective 2 out of 4 Actuation 7.3 Figure 7.3-3 Functional Diagram of Engineered Safety Features Component Contraol 7.3 System (ESF-CCS) Q Figure 7.3-4 Typical Electrical Interface for Pane!-Mounted Switches and Status 7.3
) Indicators Figure 7.3-5 Diesel Loading Sequence - Simplified Logic Diagram 7.3 Figure 7.3-6 Diesel leading Sequence - Simplified Test logic Diagram 7.3 Figure 7.3-7 ESF-CCS Test logic - Simplified legic Diagram 7.3 Figure 7.3-8a Typical FCLD for a Solenoid-Operated Valve 7.3
+ Figure 7.3-8b Typical Electrical Interface for a Solenoid-Operated Valve 7.3 Figure 7.3-9a Typical FCLD for a Modulating Valve with Solenoid-Operator 7.3 Figure 7.3-9b Typical Electrical Interface for a Modulating Valve with Solenoid Operator 7.3 Figure 7.3-10a Typical MOV Functional Interface Design 7.3 Figure 7.310b Typical Electrical Interface for a Motor-Operated Valve 7.3 Figure 7.3-11 Typical FCLD for a Full Throw Motor-Operated Valve 7.3 Figure 7.3-12 Typical FCLD for a Throttling Motor-Operated Valve 7.3 Figure 7.313a Typical FCLD for a Contactor-Operated Component 7.3 f'T Figure 7.3-13b Typical Electrical Interface for a Contactor-Operated Component 7.3 Figure 7.3-14a Typical FCLD for a Circuit Breaker-Operated Component 7.3 Aenprevent Design Meterse! Intron 6uctstwr Pope 1.7-3
System 80+ Design ControlDocument Table 1.7-1 Safety-Related Electrical, Instrumentation and Control Drawings (Cont'd.) Drawing No. Title Section Figure 7.3-14b Typical Electrical Interface for a Circuit Breaker-Operated Component 7.3 Figure 7.3-15a Typical FCLD for a Modulating Component 7.3 Figure 7.315b Typical Electrical Interface for a Modulating Component 7.3 Figure 7.3-16 Typical ESF Initiation to Actuation legic Functional Diagram 7.3 Figure 7.3-17 Simplified Schematic for Thermal Overload 7.3 Figure 7.318 In<omainment Refueling Water Storage Tank MCBD 7.3 Figure 7.3-19 Emergency Feedwater MCBD 7.3 Figure 7.3-20a Safety injection Tank 1 MCBD 7.3 7.3 i Figure 7.3-20b Safety injection Tank 2 MCBD Figure 7.3-20c Safety injection Tank 3 MCBD 7.3 Figure 7.3 20d Safety Injection Tank 4 MCBD 7.3 Figure 7.3-21 Containment Spray MCBD 7.3 Figure 7.3-22 Shutdown Cooling MCBD 7.3 Figure 7.3-23 Safety injection MCBD 7.3 1 i Figure 7.3-24 Safety Depressurization MCBD 7.3 Figure 7.3-25 Diverse Manuai Actuation of ESF Systems 7.3 j Figure 7.4-1 Interface Diagram for Division A Master Transfer Switches 7.4 Figure 7.4-2 Interface Diagram for Division N1 Master Transfer Switches 7.4 Figure 7.5-1 Diverse Display of Post Accident Monitoring Category 1 Parameters 7.5 Figure 7.5-2 IIJTC Sensor-HJTC/ Splash Shield 7.5 Figure 7.5-3 Heated Junction Thermocouple Probe Assembly 7.5 Figure 7.5-4 IlJTC Sensor and Separator Tube 7.5 Figure 7.5-5 In-Core Instrumentation IAcations 7.5 Figure 7.5-6 Electrical Diagram of IlJTC 7.5 Figure 7.5-7 IIJTC System Processing Configuration 7.5 AMrosed Design MaterW = hstroducteors Page 1.7-4
System 80+ Design ControlDocument l
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( Table 1.7-l' Safety-Related Electrical, Instrumentation and Control Drawings (Cont'd.) Drawing No. Title Section Figure 7.5 Pressurizer Level MCBD 7.5 Figure 7.6-la Functional Control Logic Shutdown Cooling System 7.6 Figure 7.6 lb Functional Control IAgic Shutdown Cooling System 7.6 Figure 7.6-Ic Functional Control Imgic Shutdown Cooling System 7.6 Figure 7.6-2 Functional Control logic Safety injection System 7.6 Figure 7.6-3 Safety-Related Interlock Test Method 7.6 Figure 7.7-12a Alternate Protection System Block Diagram 7.7 Figure 7.712b Diverse Turbine Trip and Emergency Feedwater Actuation 7.7 Figure 7.7-24 Alternate Protection System (ARTS) MCBD 7.7 Figure 7.7-25a Alternate Protection System (AFAS-1) MCBD 7.7 f% Figure 7.7-25b Alternate Protection System (AFAS-2) MCBD 7.7 Figure 7.7-27 IRWST/ Reactor Cavity Flooding System MCBD 7.7 . Figure 7.7-28 Holdup Volume / Reactor Vessel Cavity Flooding System MCBD 7.7 Figure 7.7-29 Reactor Vessel Cavity Flooding System MCBD 77 l (/ 1 494& W a**W- MM Page 1.7-5
System 80+ Design ControlDocument Table 1.7-2 Valve List Identifiers Valve Type Symbol Angle A Ball B Swing Check C Packless D Butterfly F Globe G Spring Loaded Check L Needle N Plug P Relief R Safety S Gate T Vacuum Breaker V Three Way W Excess Flow Check X Pressure Regulator Operator Type Pneumatic Diaphragm D Hand H Motor M None N Motor Modulating (Valve Position Proportion) 0 Pneumatic Piston P Solenoid S Pneumatic Vane V O Atuwesed Desiers Material kstroductas Pope 1.7-6
System 80+ Design ControlDocument O V Table 1.7-3 System 80+ Flow Diagram Matrix Piping and Figure Instrumentation Number Title Flow Diagram No. Section 1.7-1 Piping and Instrumentation Diagram Symbols E-ALWR-310-100 1.7 1.7-2 Flow Diagram Symbols and Legends F40042-01 (and 02) 1.7 5.1.2-1 Reactor Coolant System P&lD E-ALWR-310-110 5.1 5.1.2-2 Reactor Coolant Pump P&lD E-.ALWR-310-111 5.1 5.1.2-3 Pressurizer and Safety Depressurization System E-ALWR-310-Il2 5.1 P&lD 5.4.7-3 Shutdown Cooling System Flow Diagram 5.4 6.2.3-1 Annulus Ventilation System F413-06 6.2 6.2.5-1 Containment Hydmgen Recombiner System F40541 6.2 6.3.2-1 A Safety injection System P&lD E-ALWR-310-130 6.3 6.3.2-1 B Safety injection System P&lD E-ALWR-310-131 6.3 6.3.2-IC Safety injection System P&lD E-ALWR-310-132 6.3 6.3.2-I D Safety injection System Flow Diagram (Short-Term) 6.3 h b 6.3.2-l E Safety injection System Flow Diagram (Short-Term) 6.3 6.3.2-lF Safety injection System Flow Diagram (Imag-Term) 6.3 6.3.2-1G Safety injection System Flow Diagram (leng-Term) 6.3 6.8-3 In-Containment Water Storage System P&lD E-ALWR-310-133 6.8
- 8.3.1 l ' Non-Class IE Auxiliary Power System Main E71340-01 8.3 One-Line Diagram 8.3.1-2 Class IE Auxiliary Power System Main One-Line E713-00-03 8.3 Diagram 8.3.2-1 Non-Class 1E 125 VDC and 208/120VAC E713-00-02 8.3
, Instrumentation and Contiol, and 125VDC Onsite Alternate AC Source and 250VDC Power Supply Systems 8.3.2-2 Class lE 125VDC and Vital 120VAC E713-00-04 8.3 Instrumentation and Control Power Supply System 9.1 3 Spent Fuel Pool Cooling and Cleanup P&lD E-ALWR-310-140 9.1 9.2.1 1 Station Service Water System F412-01 (tluu 04) 9.2 9.2.2-1 Component Cooling Water System F411-01 (thru 18) 9.2 9.2.3-1 Demineralized Water Makeup System FSK-9-21 9.2 9.2.6-1 Condensate Storage System F414-01 9.2 , 9.2.8-1 Turbine Building Cooling Water System FSK-9-7 9.2 v' Approved Deelen Metenie! . Jhsroduction Page 1.7 7 1 l
System 80+ Design ControlDocument Table 1.7-3 System 80+ Flow Diagram Matrix (Cont'd.) Piping and j Figure Instrumentation Number - Title Mow Diagram No. Section 9.2.9-1 Chilled Water System F413-07-01 (thru 16) 9.2 9.3.1-1 Instrument Air System F422-01-01 (and 02) 9.3 Station Air System F422-03 9.3 9.3.1-2 Breathing Air System F422-02 9.3 9.3.1-3 Process Sampling System P&ID E-ALWR-310-150 9.3 9.3.2-1 Containment Building Floor Drain System F410-01 9.3 9.3.3-1 Reactor Building Subsphere Floor Drain System F410-02-01 (and 02) 9.3 9.3.3-2 Nuclear Annex Radioactive Floor Drain System F410-03-01 (and 03) 9.3 9.3.3-3 Nuclear Annex Nonradioactive Floor Drain System F410-03-02 (and 04) 9.3 9.3.3-4 CVCS Area Floor Drain System F41043-05 (and 06) 9.3 9.3.3-5 Chemical and Volume Control System Flow Diagram E-ALWR-310-120 9.3 9.3.4-1 (thru 123) 9.4-2 Control Complex Ventilation System F413-01-01 (and 02) 9.4 Fuel Building Ventilation System F413-02 9.4 9.4 3 9.4-4 Subsphere Building Cooling F413-03 9.4 9.4 5 Subsphere Building Ventilation System F41304 9.4 9.4 6 Containment Cooling and Ventilation System F413-05 9.4 9.4-7 Diesel Building Ventilation System F413-09 (and 11) 9.4 9.4-8 Nuclear Annex Ventilation System F413-10 9.4 9.4-9 Radwaste Building Ventilation System FSK-38-1 A 9.4 9.4-10 CCW Heat Exchanger Structure Ventilation System FSK-39-1 A 9.4 9.5.1-1 Fire Protection Water Distribution System F406-01 9.5 9.5.4-1 Diesel Generator Engine Fuel Oil System F417-01 (and 02) 9.5 9.5.5-1 Diesel Geocator Engine Cooling Water System F418 01 9.5 9.5.6-1 Diesel Generator Engine Starting Air System F420-01 (and 02) 9.5 9.5.7-1 Diesel Generator Engine Lube Oil System F419-01 (thru 03) 9.5 F419-04 9.5.8-1 Diesel Generator Engine Air intake and E*ust F421-01 9.5 System 9.5.9-1 Diesel Generator Building Sump Pump System F410-04-01 9.5 10.1-2 Main Steam and Feedwater System P&lD E-ALWR-310-Il5 10.1 10.3.2-1 Main Steam and Extraction Steam Syctems FSK-3-1 10.3 10.4.1-1 Main Condenser FSK-4-ID 10.4 10.4.2-1 Main Condenser Evacuation System FSK-5-1 10.4 10.4.3-1 Turbine Gland Scaling System FSK-16-1 10.4 Approved Design Matenial- kstredvetion Page r.7-8
System 80+ Design controlDocument (] (/ Table 1.7-3 System 80+ Flow Diagram Matrix (Cont'd.) Piping and Figure Instrumentation Number Title Mow Diagram No. Section 10.4.5-1 Circulating Water System FSK-2-1 10,4 10.4.6 1 Condensate Cleanup Systent FSK-4-7-1 10.4 10.4.7 1 Condensate Feedwater and lleater Drain Systems FSK-6-1 10.4 10.4.8-1 Steam Generator Blowdown System FSK-32-13 10.4 10.4.9-1 Emergency Feedwater System F401-01 (and 02) 10.4 10A-1 Emergency Feedwater System Schematic App. 10A 11.2-1 Liquid Waste Management System FSK-91 1 (A thru D) 11.2 l1.3 1 Gaseous Waste Management System FSK-31-2A i1.3 q 11.4-1 Solid Waste Management System FSK-31-3 (A and B) 11.4
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System 80+ Design ControlDocument
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NOTES: 03308.1 - FSK - 01 A - S FOR GENERAL NOTES AND TABLES REFER TO DWG F400 01 & F400 02. STONE & WEBSTER ENGINEERING CORP REFERENCE DRAWING UST DWG NO 03308.XX. DRAWING TITLE FSK 2-1 FLOW DLAGRAM - CIRCULATING WATER SYSTEM FSK-3-1 FLOW DIAGRAM - MAIN STEAM & EXTRACTION STEAM SYSTEMS FSK-4-ID FLOW DIAGRAM - MAIN CONDENSER FSK-4-3 FLOW DIAGRAM - CONDENSATE STORAGE SYSTEM FSK-4-7 FIDW DIAGRAM - CONDENSATE CLEANUP SYSTEM FSK-5-1 FLOW DIAGRAM - MAIN CONDENSER EVACUATION SYSTEM FSK-6-1 FLOW DIAGRAM - CONDENSATE, FEEDWATER, & HEATER DRAIN SYSTEMS FSK-9 7 FLOW DIAGRAM - TURBINE BUILDING COOUNG WATER SYSTEM FSK-9-10 FLOW DIAGRAM - TURBINE BUILDING SERVICE WATER SYSTEM FSK-9-21 FLOW DIAGRAM - DEMINERAUZED WATER MAKEUP SYSTEM , O) i ! U FSK-16-1 FLOW DIACRAM TURBINE GLAND SEAUNG SYSTEM FSK-31 1 A FIDW DIAGRAM - LIQUID WASTE MANAGEMENT SYSTEM FSK 31 IB FLOW DIAGRAM - UQUID WASTE MANAGEMENT SYSTEM FSK-31 lC FLOW DIAGRAM LIQUID WASTE MANAGEMENT SYSTEM FSK-31 ID FLOW DIAGRAM - LIQUID WASTE MANAGEMENT SYSTEM FSK-312A FLOW DIAGRAM GASEOUS WASTE MANAGEMENT SYSTEM FSK-31-3A FLOW DIAGRAM - SOUD WASTE MANAGEMENT SYSTEM FSK 3i-3B FLOW DIAGRAM - SOUD WASTE MANAGEMENT SYSTEM FSK-32-13 FLOW DIAGRAM - STEAM GENERATOR BIDWDOWN SYSTEM FSK-38-I A FLOW DIAGRAM - RADWASTE BUILDING VENTILATION SYSTEM FSK 39-I A FLOW DIAGRAM - CCW HLAT EXCHANGER STRUCTURE VT.NTILATION SYSTEM OT () How Diagram Index Figure 1.7-3 A$4woved Design Matenfal- krtroduction Page 1.7-17
. ~ . . - - . - - - - . .. .-. . - -- --. .
System 80+ Design ControlDocumart [ 1.8 L Regulatory Compliance, Industry Codes and Standards ! System 80+ compliance with U.S. NRC Regulatory Guides, Generic Letters, Bulletins, and elements of the Standard Review Plan is documented in this section. Regulatory Guides, the guide date or revision, l and System 80+ compliance with applicable Guides are summarized in Table 1.8-1.
~
Operational expericnce information highlighted in Regulatory Bulletins and Letters has been incorporated
- into the System 80+ designs. Generic Letters and NRC Bulletins from 1980 through December,1993 are identified in Tables 1.0-2 and 1.8-3. The applicability of each Generic letter or Bulletin to .
System 80+ is assessed, wit 0 additional information for applicable issues provided in the referenced sections of this report. i System 80+ deviations from the U.S. NRC Standard Review Plan, NUREG-0800 [ LWR Edition, June 1987), are listed in Table 1.8-4. Specific sections are also identified where further details relevant to , t each SRP deviation are discussed. Site-specific compliance with individual Standard Review Plan sections is provided in Table 1.8-5. Table 1.8-6 identifies the industrial Codes and Standards, and code editions, invoked for certification of l j the System 80+ Standard Design. Where a particular structure, system, or component requires a code l edition different from that liste.t in Table 1.8-6, an explanation of such difference is provided in the > ! appropriate text. Other Codes and Standards that are utilized but not invoked as essential for design certification are incorporated into the individual chapters of this document. Revisions to accepted l industry codes applied to System 80+ will be evaluated on a case-by-case basis. ((The applicability of , 3 code editions will be confirmed by the Combined License applicant in the site-specific Safety Analysis l . .. Report.))! 4 i ASME Section III, Division 1 Code Cases applicable to System 80+ are identified in Table 1.8-7. . Except for N-122-1, these Code Cases are consistent with those identified in Regulatory Guide 1.84, Revision 29, for design and fabrication, or Regulatory Guide 1.85, Revision 29, for mateiials and testing, i that were in effect on July ?l,1993. ((Later code cases that may be used for System 80+ piping and i
- piping support design will be provided with the plant-specific information.))!
L Cross-references to subsections of this report discussing Unresolved and Generic Safety Issues, the Three , i Mile Island Rule [10 CFR 50.34 (f)], and new NRC policy issues (SECY-93-087) are provided in Tables , ! 1.8-8,1.8-9, .$nd 1.8-10. l I f 1 O ; 3 COL information item; see DCD Introduction Section 3.2. l t Appmed Deadp nionand- beehenien (agg) page 1.g.1 , l l w sn . 4 .,r ,. . . . - ,,. , . - - -
System 80+ Design ControlDocument Table 1.8-1 NRC Regulatory Guide Applicability Analysis to System 80+ Document Title Date Section RG 1.1 - Net Positive Suction Head for Emergency Core Cooling and 11n0 6.3; 6.5 Containment Heat Removal System Pumps RG 1.2 - Thermal Shock to Reactor Pressure Vessels - GDC 35 Withdrawn RG 1.3 - N/A (BWR) RG 1.4 - Assumptions Used for Evaluating the Potential Radiological Rev. 2 6.3.3.6; 12 Consequences of a LOCA for Pressurized Water Reactors 6n4 RG 1.5 - N/A (BWR) RG 1.6 - Independence Between Redundant Standby (On-Site) Power 3ni 8.1.4.2 Sources and Between Their Distribution Systems RG 1.7 - Supplement - Control of Combustible Gas Concentration in Rev. 2 6.2.5; 12 Containment Following LOCA 11/78 RG 1.8 - Qualification & Training of Personnel for Nuclear Power Rev. 2 13 Plants 4/87 RG 1.9 - Selection, Design, Qualification, and Testing of Emergency Rev. 3 8.!.4.2 Diesel Generator Units used as Class IE Onsite Electrical Power 7/93 Systems RG 1.10 - Mechanical (Cadweld) Splices in Reinforcing Bars of Withdraum Concrete Containments RG 1.11 - Instrument Lines Penetrating Primary Reactor Contamment 3nt 7.1.2.15; 6.2.4.1.1 RG 1.12 - Instrumentation for Earthquakes Rev.1 3.7 4n4 RG 1.13 - Spent Fuel Storage Facility Design Basis Rev. 2 9.1 12/81 RG 1.14 - Reactor Coolant Pump Flywheel Integrity Rev.1 5.4.1.1 8n3 RG 1.15 - Testing of Reinforemg Bars for Concrete Structures Withdrawn RG 1.16 - Reporting of Operating Information Rev. 4 Not Applicable 8n5 RG 1.17 - Protection Against Industrial Sabotage Withdrawm RG 1.18 - Structural Acceptance Tests for Concrete Primary Reactor Withdrawn Containments RG 1.19 - Nondestructive Examination of Pnmary Containment Welds Withdrawm RG 1.20 Comprehensive Vibration Assessment Program for Reactor Rev. 2 3.9.2.4 Internals During Preoperational and Initial Startup Testing Sn6 Apprewmf Design Matenial httroduction Page 1.8-2
System 80+ Desian controlDocument O , V Table 1.8-1 NRC Regulatory Guide Applicability Analysis to System 80+ (Cont'd.) Document Title Date Section RG 1.21 - Measuring, Evaluating, and Reporting Radioactivity in, and Rev.I 11.1; 11.5.1.1; Releases from Nuclear Power Plants 6n4 11.5.2.1 (COL applicant) RG 1.22 - Periodic Testing of Protection Systems Actuation Functions 2n2 7.1.2.17; 8.1.4.2 RG 1.23 - Onsite Meteorological Programs Not Applicable RG 1.24 - Assumptions Used for Evaluating the Potential Radiological 3/72 15.7 Consequences of a Pressurized Water Reactor Radioactive Gas Storage ' Tank Failure
- RG 1.25 - Assumptions Used for Evaluating the Potential Radiological 3n2 15.7 Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors RG 1.26 - Quality Group Classifications and Standards Rev. 3 3.2.2
- 10.4 2n6 RG 1.27 - Ultimate Heat Sink Rev. 2 9.2.5 1/76 RG 1.28 - Quality Assurance Program Requirements Rev. 3 17 and Repon
, p CENPD-210-A, l
' 1 (Design and Construction) 8/85 Rev. 7 (Section 111.2 .1 )
RG 1.29 - Seismic Design Classification Rev. 3 3.2.1; 7.1.2.18; 9n8 10.4.9 RG 1.30 - Quality Assurance Requirements for the Installation. 8/72 RG 1.28 Rev. 3 lasp ction and Testing of lastrumentation and Electrical Equipment applied instead of RG 1.30 RG 1.31 - Control of Ferrite Content in Stainless Steel Weld Metal Rev. 3 5.2.3.4.2.1 408 RG 1.32 - Criteria for Safety-Related Electric Systems for Nuclear Rev. 2 8.1.4.2 Power Plants 2n7 RG 1.33 - Quality Assurance Program Requirements (Operation) Rev. 2 Not Applicable 2n8 RG 1.34 - Control of Electroslag Weld Properties 12n2 5.2.3.3.2.2 RG 1.35 - Inservice inspection of Ungrouted Tendons in Prestressed Rev. 3 Not Applicable Conctete Containment 7/90 (Concrete containment) RG 1.36 - Nonmetallic Thermal Insulation for Austenitic Stainless Steel 2n3 5.2.3.2.3; 10.3.2.3.4 RG 1.37 - Quality Assurance Requirements for Cleaning of Fluid 3n3 5.2.3.4.1.2.1; d Systems and Associated Components of Water-Cooled Nuclear Power 10.3.6.2 (Note A) . Plants Answend Denkn niemW husekcaian (2n5) Pope 1.8-3 i
System 80+ Design ControlDocument Table 1.8-1 NRC . Regulatory Guide Applicability Analysis to System 80+ (Cont'd.) Document Title Date Section RG 1.38 - Quality Assurance Requirements for Packaging, Shipping, Rev.2 17 (Note A) Receiving, Storage, and llandling of Items for Water Cooled Nuclear 5n7 Power Plants RG 1.39 - Housekeeping Requirements for Water Cooled Nuclear Rev. 2 6.4 (Note A) Power Plants 9n7 RG 1.40 - Qualification Tests of Continuous-Duty Motors Installed 3n3 3.11 Inside the Containment of Water-Cooled Nuclear Power Plants RG 1.41 - Preoperational Testing of Redundant On-site Electric Power 3/73 8.1.4.2; 14 Systems to Verify Proper lead Group Assignments RG 1.42 - Interim Licensing Policy on as low as Practicable for Withdrawn Gaseous Radioiodine Releases from Light Water-Cooled Nuclear Power Reactors LG 1.43 - Control of Stainless Steel Weld Cladding of lew-Alloy Steel Sn3 5.2.3.3.2.1 Components RG 1.44 - Control of the Use of Sensitized Stainless Steel Sn3 5.2.3.4.1.1.1 RG 1.45 - Reactor Coolant Pressure Boundary Leakage Detection Sn3 5.2.5; 7.1.2.20: Systems 11.5.1.1 RG 1.46 - Protection Against Pipe Whip Inside Containment - Withdrawn RG 1.47 - Bypassed and inoperable Status Indication for Nuclear 5/73 7.1.2.21; 8.1.4.2 Power Plant Safety Systems RG 1.48 - Design Limits and Loading Combinations for Seismic Withdrawn Category 1 Fluid System Components RG 1.49 - Power Levels of Nuclear Power Plants Rev.I 1.1.2 1203 RG 1.50 - Control of Preheat Temperature for Welding of low-Alloy Sn3 5.2.3.3.2.1 Steel RG 1.51 - Inservice Inspection of ASME Code Class 2 Nuclear Power Withdrawm Components RG 1.52 - Design, Testing, and Maintenance Criteria for Post-Accident Rev. 2 6.5 Engineered Safety Feature Atmosphere Cleanup System Air Filtration 3n8 and Adsorption Units of Light Water-Cooled Nuclear Power Plants RG 1.53 - Application of the Single-Failure Criterion to Nuclear Power 6n3 7.1.2.9; 8.1.4.2 Plant Protection Systems RG 1.54 - Quality Assurance Requirements for Protective Coatings 6n3 6.1.2.1 Applied to Water-Cooled Nuclear Power Plants RG 1.55 - Concrete Placement in Category 1 Structures Withdrawm RG 1.56 Maintenance of Water Purity in Boiling Water Reactors Rev.1 N/A (BWR) 708 Approved Design Ataterial. antroduction Page 1.8-4
System 80+ oesign controlDocument n V) Table 1.8-1 NRC Regulatory Guide Applicability Analysis to Systern 80+ (Cont'd.) Document Title Date Section RG 1.57 - Design Limits and Loading Combinations for Metal Primary 6n3 3.8 Reactor Containment System Components RG 1.58 - Qualification of Nuclear Power Plant inspection, Withdrawn Examination and Testing Personnel RG 1.59 - Design Basis Floods for Nuclear Powet Plants Rev. 2 2 807 RG 1.60 - Design Response Spectra for Seismic Design of Nuclear Rev.I 2; 3.7 Power Plants 12n3 RG 1.61 - Damping Values for Seismic Design of Nuclear Power 1093 3.7.1.3 Plants RG 1.62 Manual Initiation of Protective Actions 10n3 7.1.2.22; 8.1.4.2 RG 1.63 - Electric Penetration Assemblics in Containment Structures Rev. 3 3.8.2; 8.1.4.2 for Nuclear Power Plants 2/87 1 1 RG 1.64 - Quality Assurance Requirements for the Design of Nuclear Withdrawn l Power Plants J ( w\ RG 1.65 - Materials and Inspections for Reactor Vessel Closure Studs 10/73 5.3.1.7
%.) Withdrawn RG 1.66 - Nondestructive Examination of Tubular Products RG 1.67 - Installation of Overpressure Protection Devices Withdrawn RG 1.68 - Initial Test Programs for Water-Cooled Power Reactors Rev.2 14.2.7.1; 8n8 7.1.2.24; 8.1.4.2 RG 1.68.1 - Preoperational and initial Startup Testing of Feedwater and Rev.I N/A (BWR)
Condensate Systems for Boiling Water Power Plants in7 RG 1.68.2 - Initial Startup Test Program to Demonstrate Remote Rev.I 14.2.7.3 , Shutdown Capability for Water-Cooled Nuclear Power Plants 7/78 l RG 1.68.3 - Preoperational Testing of Instrument and Control Air 4/82 14 Systems PG 1.69 - Concrete Radiation Shields for Nuclear Power Plants 12n3 12 RG 1.70 - Standard Format and Contents of Safety Analysis Reports Rev. 3 All for Nuclear Power Plants 1108 RG 1.71 - Welder Qualification for Areas of Limited Accessibility 12n3 5.2.3.3.2.3 RG 1.72 - Spray Pond Piping Made from Fiberglass-Reinforced Rev. 2 9.2.5 Thermosetting Resin 1108 RG 1.73 - Qualification Tests of Electric Valve Operators Installed in4 3.11; 7.1.2.25 Inside the Containment of Nuclear Power Plants C/ RG 1.74 - Quality Assurance Terms and Definitions Withdrawn Apswend Dessprs Material- kutroductkus Page 1.8 5
System 80 + Design ControlDocument Table 1.8-1 NRC Regulatory Guide Applicability Analysis to System 80+ (Cont'd.) Document Title Date Section RG 1.75 - Physical Independence of Electric Systems Rev. 2 7.1.2.10; 8.1.4.2 9n8 RG 1,76 - Design Bases Tomado for Nuclear Power Plants 4n4 2 RG 1.77 - Assumptions Used for Evaluating a Control Rod Ejection Sn4 15.4.5 Accident for Pressurized Water Reactors RG 1.78 - Assump: ions for Evaluating the Habitability of a Nuclear 6n4 6.4;9.4 Power Plant Control Room During a Postulated Hazardous Chemical Release RG 1.79 - Preoperational Testing of Emergency Core Cooling Syst:ms Rev.1 3.1.33; 14.2.7.2 for Pressurized Water Systems 9/75 RG 1.80 - Preoperational Testing of Instrument Air Systems Withdrawn See RG 1.68.3 RG 1.81 - Shared Emergency and Shutdown Electric Systems for Rev.I 1.2.1.3 Multi-Unit Nuclear Power Plants ins RG 1,82 - Water Sources for lAng-Term Recirculation Cooling Rev.I 1.2; 3.8 Following a less-of-Coolant Accident 11/85 RG 1.83 - Inservice Inspection of Pressurized Water Reactor Ste un Rev.1 5.2.4.1 Generator Tubes 7n5 RG 1.84 - Design and Fabrication Code Case Acceptability ASME Rev. 29 5.2.1.2 Section III Division 1 7/93 RG 1.85 - Materials Code Case Acceptability ASME Rev. 29 5.2.1.2 Section III Division 1 7/93 RG 1.86 - Termination of Operating Licenses for Nuclear Reactors 6n4 Not Applicable RG 1.87 - Guidance for Construction of Class 1 Components in Rev.1 Not Applicable Elevated Temperature Reactors 655 RG 1.88 - Collection, Storage, and Maintenance of Nuclear Power Withdrawn Plant Quality Assurance Records RG 1.89 - Environmental Qualification of Certal Electric Equipment Rev.1 3.11; 7.1.2.5; Imponant to Safety for Nuclear Power Plants 6/84 8.1.4.2 RG 1.90 - Inservice inspection of Prestressed Concrete Containment Rev.1 Not Applicable Structures with Grouted Tendons 807 (Concrete contamment) RG 1.91 Evaluations of Explosions Postulated to Occur on Rev.1 Not Applicable Transpottation Routes Near Nuclear Power Plants 2n8 RG 1.92 - Combining Modal Responses and Spatial Components in Rev.1 3.7.2.7 Seismic Response Analysis 256 RG 1.93 - Availability of Electric Power Sources 1254 8.1.4.2 Approved Desupn Material kutroduction Page 1.8-6
i i i l l ) System 80+ Design ControlDocument o , Table 1.8-1 NRC Regulatory Guide Applicability Analysis to System 80+ (Cont'd.) ; Document Title Date Section RG 1.94 - Quality Assurance Requirements for Installation, inspection, Rev.1 3.8 (Note A) , and Testing of Structural Concrete and Structural Steel 4n6 i During the Construction Phase of Nuclear Power Plants l RG 1.95 - Protection of Nuclear Power Plant Control Room Operators Rev.1 6.4 Against en Accidental Chlorine Release - 197 RG 1.% - Design of Main Steam Isolation Valve Leakage Control Rev.I N/A (BWRs) Systems for Boiling Water Reactor Nuclear Power Plants 6/76 RG 1.97 -Instrumentation for Light-Water-Cooled Nuclear Power Rev.3 3.1; 7.1.2.26; 7.5; Plants To Assess Plant and Environs Conditions During and Following 5/83 10.4.9; 11.5.1.1 an Accident RG 1.98 - Assumptions Used for Evaluating the Potential Radiological 3/76 N/A (BWRs) Consequences of a Radioactive Offgas System Failure in a Boiling Water Reactor l RG 1.99 - Radiation Embrittlement of Reactor Vessel Materials Rev. 2 5.3.1.6.7 5/88 7 RG 1.100 - Seismic Qualification of Electric and Mechanical Rev. 2 3.10 . Equipment for Nuclear Power Plants 6/88 RG 1.101. Emergency Planning and Preparedness for Nuclear Power Rev. 3 NUREG-0654 Reactors 8/92 applies with respect to System 80+ emergency facilities described in Sections 13.3.3.1 and 13.3.3.3 RG 1.102 - Flood Protection for Nuclear Power Plants Rev.I 2 9/76 RG 1.103 - Post Tensioned Prestressing Systems for Concrete Reactor Withdrawm Vessels and Containment RG 1.104 - Overhead Crane llandling Systems for Nuclear Power Withdrawn Plants , RG 1.105 - Instrument Serpoints for Safety-Related Systems Rev. 2 7.1.2.27 2/86 RG 1.106 - Thermal Overload Protection for Electric Motors on Rev.I 7.1.2.28 Motor-Operated Valves 3/77 RG 1.107 - Qualifications for Cement Grouting for Prestressing Rev.I Not Applicable Tendons in Containment Structures 2/77 (Concrete containment) N RG 1.108 - Periodic Testing of Diesel Generator Units Used as Onsite Withdrawn See RG.1.9 Electric Power Systems at Nuclear Power Plants Anweved Design hinterial- kHroduceion Page 1.8 7
System 80 + Design ControlDocument Table 1.8-1 NRC Regulatory Guide Applicability Analysis to System 80+ (Cont'd.) Document Title Date Section RG 1.109 - Calculation of Annual Doses to Man From Routine Rev.I 11.2.6.3: Releases of Reactor Effluents for the Purpose of Evaluating Compliance 10n7 11.3.6.4; 11.3.7.2 with 10 CFR Part 50. Appendix ! RG 1.110 - Cost-Benefit Analysis for Radwaste Systems for Light- 3n6 12 Water-Cooled Nuclear Power Reactors RG 1.111 - Methods for Estimating Atmospheric Transport and Rev.1 2.3.5 Dispersion of Gaseous Effluents in Routine Releases From Light- 707 Water-Cooled Reactors RG 1.112 - Calculation of Releases of Radioactive Materials in Gaseous 5/77 Not Applicable and Liquid Effluents from Light-Water-Cooled Power Reactors RG 1.113 - Estirnating Aquatic Dispersion of Effluents from Accidental Rev.I Not Applicable and Routine Reactor Releases for the Purpose of implementing 4n7 Appendix I RG 1.114 - Guidance to Operators at the Controls and to Senior Rev. 2 Not Applicable Operators in the Control Room of a Nuclear Power Plant 5/89 RG 1.115 - Protection Against 1.ow-Trajectory Turbine Missiles Rev.1 3.5 707 RG 1.116 - Quality Assurance Requirements for Installation. Sn7 3.9 (Note A) ; Inspection and Testing of Mechanical Equipment and Systems RG 1.117 - Tornado Design Classification Rev.1 3.1.2 408 1 RG 1.118 - Periodic Testing of Electric Power and Protection Systems Rev. 2 7.1.2.7; 8 608 RG 1.119 - Survei!!ance Program for New Fuel Assembly Designs Withdrawn RG 1.120 - Fire Protection Guidelines for Nuclear Power Plants Rev.I 7.1.2.29 11n7 RG 1.121 - Bases for Plugging Degraded PWR Steam Generator Tubes 806 5.4 RG 1.122 - Development of Floor Design Response Spectra for Seismic Rev.1 3.7.2 Design of Floor-Supported Equipment or Components 2n8 RG 1.123 - Quality Assurance Requirements for Control of Withdrawn Procurement of items and Services for Nuclear Power Plants RG 1.124 - Service Limits and Loading Combinations for Class 1 Rev.1 5.4.14 Linear Type Component Supports 108 RG 1.125 - Physical Models for Design and Operation of Hydraulic Rev.1 3 Structures and Systems for Nuclear Power Plants 1008 RG 1.126 - An Acceptable Model and Related Statistical Methods for Rev.1 4.2 the Analysis of Fuel Densification 3n8 Thved Desipro Motoriel- Jhtroduction Page 1.8 8
System 80 + Design ControlDocument s r Table 1.8-1 NRC Regulatory Guide Applicability Analysis to System 80+ (Cont'd.) Document Title Date Section RG 1.127 - Inspection of Water-Control Structures Associated with Rev.1 Not Applicable Nuclear Power Plants 3n8 RG 1.128 - Installation Design and Installation of Large Lead Storage Rev.I 8.1;8.3.2 ; Batteries for Nuclear Power Plants 10n8 RG 1.129 - Maintenance, Testing, and Replacement of Large lead Rev.I 8.1.4.2 Storage Batteries for Nuclear Power Plants 2n8 RG 1.130 - Service Limits and Loading Combinations for Rev.1 5.4.14 Class 1 Plate-and-Shell-Type Component Supports 1008 RG 1.131 - Qualification Tests of Electric Cables, Field Splices, and Rev.I 8.1.4.2 Connections for Light-Water-Cooled Nuclear Power Plants 809 (Draft) RG 1.132 - Site Investigations for Foundations of Nuclear Power Plants Rev.I Not Applicable 3n9 AG 1.133 - Loose-Part Detection Program for the Primary Systems of Rev.1 7.1.2.30; Light-Water-Cooled Reactors 5/81 7.7.1.6.3 p RG 1.134 - Medical Evaluarion of Licensed Personnel for Nuclear Rev. 2 Not Applicable Q Power Plants 4/87 RG 1.135 - Normal Water level and Discharge at Nuclear Power 9n7 Not Applicable Plants RG 1.136 - Materials, Construction, and Testing of Concrete Rev. 2 Not Applicable Containments 6/81 (Concrete containment) RG 1.137 - Fuel-Oil Sptems for Standby Diesel Generators Rev.I 9.5 10n9 RG 1.138 - Laboratory Investigations of Soils for Engineering Analysis 4n8 Not Applicable and Design of Nuclear Power Plants RG 1.139 - Guidance for Residual Heat Removal Sn8 5.4.7 RG 1.140 - Design Testing, and Maintenance Criteria for Normal Rev.I 9.4; 6.2.3; 11.3 Ventilation Exhaust System Air Filtration and Adsorption Units of 1009 Ligb>Wmr-Cooled Nuclear Power Plants RG 1.141 - Containment Isolation Provisions for Fluid Systems 4n8 6.2.4 RG 1.142 - Safety-Related Concrete Structures for Nuclear Power Rev.1 3.8 Plants 10/81 RG 1.143 - Design Guidance for Radioactive Waste Management Rev.I 11 Systems, Structures, and Components Installed in Light-Water-Cooled 10n9 Nuclear Power Plants RG 1.144 - Auditing of Quality Assurance Programs for Nuclear Power Withdrawn Plants I 4proveer outen Morww kutrodvedon Pope 1.s-s
System 80 + Design ControlDocument Table 1.8-1 NRC Regulatory Guide Applicability Analysis to System 80+ (Cont'd.) Document Title Date Section RG 1.145 - Atmospheric Dispersion Models for Potential Accident Rev.I 2 Consequence Assessment at Nuclear Power Plants 2/83 RG 1.146 - Qualification of Quality Assurance Program Audit Withdrawn Personnel for Nuclear Power Plants RG 1.147 - In-service inspection Code Case Acceptability, ASME Rev.10 5.2.1.2 Section XI, Division 1 7/93 RG 1.143 - Functional Specification for Active Valve Assemblies in 3/81 3; 5; 6 Systems important to Safety in Nuclear Power Plants RG 1.149 - Nuclear Power Piant Simulation Facilities for Use in Rev.I Not Applicable Operator License Examinations 4/87 RG 1.150 - Ultrasonic Testing of Reactor Vessel Welds During Pre- Rev.1 5.1.4; 5.3.1.3 service and in-service Exammations 2/83 RG 1.151 - Instmment Sensing Lines 7/83 3.2.1; 3.2.2; 7.1.2.31 RG 1.152 - Criteria for Programmable Digital Computer System 11/85 7.1.2.32 Software in Safety-Related Systems of Nuclear Power Plants RG 1.153 - Criteria for Power, lastrumentation, and Control Portion of 12/85 5.1.4; 7.1.2.13 Safety Systems RG 1.154 - Format and Content of Plant-Specific Pressurized Thermal 1/87 Not Applicable Shock Safety Analysis Reports for PWRs (Plant-Specific) RG 1.155 - Station Blackout 8/88 8.1; 10.4.8; 10.4.9; 14.2.12 RG 1.156 - Environmental Qualification of Connection Assemblies for 11/87 7.1.2.33 Nuclear Power Plants RG 1.157 - Best-Estimate Calculations of Emergency Core Cooling 5/89 Not Used System Performance (Appendix K model used) RG 1.158 - Qualification of Safety-Related Lead Storage Batteries for 2/89 8.3.2 Nuclear Power Plants RG 1.159 - Assuring the Availability of Funds for Decommissioning 8/90 Not Applicable Nuclear Reactors (Plant Specific) RG 1.160 - Monitoring the Effectiveness of Maintenance at Nuclear 6/93 Not Applicable Power Plants (Plant Specific) RG 4.15 - Quality Assurance for Radiological Monitoring Programs Rev.I 11.1; 11.5.1.1; (Norms.1 Operations)- Effluent Streams and the Environment 2/79 11.5.1.4 (COL applicant) RG 5.1 - Serial Numbering of Fuel Assemblies for Light-Water-Cooled 12/72 4.2.4.4 Nuclear Power Reactors Amromt onian nestew keroduction rose 1.s-to
System 80+ oesign contiot Document /O V Table 1.8-1 NRC Regulatory Guide Applicability Analysis to System 80+ (Cont'd.) Document Title Date Section RG 5.12 General Use of Locks in the Protection and Control of 11/73 13.6 Facilities and Special Nuclear Materials RG 5.20 - Training, Equipping, and Qualifying of Guards and 1/74 13.6 Watclunen RG 5.44 - Perimeter Intrusion Alarm Systems Rev. 2 13.6 5/80 RG 5.57 - Shipping and Receiving Control of Strategic Special Nuclear Rev.1 4.2.3.1.5 Material 6/80 RG 5.65 - Vital Area Access Controls, Protection of Physical Security 9/86 9.4; 9.2.2; 13.6 Equipment, and Key and Imck Controls RG 5.66 - Access Authorization Program for Nuclear Power Plants 6/91 13.6 RG 8.8 - Information Relevant to Ensuring the Occupational Radiation Rev. 3 11; 12 Exposures at Nuclear Power Statior.s will be ALARA 6/78 RG 8.10 - Operating Philosophy for Maintaining Occupational Rev.1-R 12 Radiation Exposures As Low As Is Reasonably Achievable 5/77 , i Rev.2 7.1.2.34; [m 's RG 8.12 - Criticality Accident Alarm Systems 10/88 7.7.1.1.10; 11.5.1.1 RG 8.19 - Occupational Radiation Dose Assessment in Light-Water Rev.I 12 Reactor Plants-Design Stage Man-Rem Estimates 6/79 RG 8.38 - Control of Access to High and Very Iligh Radiation Areas 6/93 12 of Nuclear Plants i Note A: The QA Program Description for System 80+ (CENPD-210-A) commits to NQA-2. l
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C'N %/
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Approved Desigrs Materlat . kutroduction Pope 1.8-11
System BC + Design controlDocument Table 1.8-2 Generic Letters Applicability Analysis to System 80+ No. Title Comment Relocation of Technical Specification Tables of Instrument Response Time Limits Ch 16 93-08 93-07 Modification of the Technical Specification Administrative Control Requirements Ch 16 for Emergency and Security Plans 93-06 Research Results on Generic Safety issue 106, " Piping and the Use of Highly 20.2.30 Combustible Gases in Vital Areas' 93-05 Line-Item Technical Specifications improvements to Reduce Surveillance Ch 16 Requirements for Testing during Power Operation Rod Control System Failure and Withdrawal of Rod Control Cluster Assemblies N/Al83 93-04 Verification of Plant Records N/Al31 93-03 93-02 NRC Public Workshop on Commercial Grade Procurement and Dedication N/Ald! 93-01 Emergency Response Data System Test Program N/Al33 92-09 Limited Participation by NRC in the IAEA International Nuclear Event Scale N/Al33 92-08 Fire Endurance Test Acceptance Criteria for Fire Barrier Systems Used to N/Al31 Sup!I Separate Redundant Safe Shutdown Trains Within the Same Fire Area (Draft] 92-08 Thermo-Lag 330-1 Fire Barriers N/Alli oT07 Office Nuclear Reactor Regulation Reorganization N/At4) 92-06 Operator Licensing National Examination Schedule N/Al31 924)5 NRC Workshop on the Systematic Assessment of Licensee Performance [SALP] N/Al33 Program 92-04 Resolution of the issues Related to Reactor Vessel Water Level Instrumentation in N/Atil BWRs 92-03 Compilation of the Current Licensing Basis: Request for Voluntary Participation N/Al41 in Pilot Program N/A'1l 92-02 Resolution of Generic issue 79, "Unanalyzed Reactor Vessel [PWR] Thermal Stress during Natural Convection Cooldown" 92-01 Reactor Vessel Structural Integrity N/Aldl Rev 1 l N/A'1l 91 19 Informaticn to Addressees Regarding New Telephone Numbers for NRC Offices Located in One White Flint North l l N/A'1 i 91 18 Information to Licensees Regarding Two NRC Inspection Manual Sections on Resolution of Degraded and Nonconforming Conditions and on Operability I 91-17 Generic Safety issue 29, " Bolting Degradation or Failures in Nuclear Power 20.2.7 Plants" O Approved Desiger Material- krtraduction Page 1.812
l System 80+ oesign controlDocument
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- - Q. l - NJ l Table 1.8-2 Generic Letters Applicability Analysis to System 80+ (Cont'd.) No. Title Comment ! 91 16 Licensed Operators' and Other Nuclear Facility Personnel Fitness for Duty N/Al33 91-15 Operating Experience Feedback Report, Solenoid-Operated Valve Problems 3.9.3.2 at U.S. Reactors 91 14 Emergency Telecommunications N/Aldi 91-13 Request for Information Related to the Resolution of Generic Issue 130, N/Aldl j
" Essential Service Water System Failures at Multi-Unit Sites," Pursuant to 10CFR50.54(f) 91-12 Operator Licensing National Examination Schedule N/AI31 91 11 Resolution of Generic Issues 48, "LCOs for Class 1E Vital Instrument 20.2.11:
Buses," and 49, " Interlocks and LCOs for Class IE Tie Breakers" Pursuant 20.2.12 j to 10 CFR 50.54(f) 91 10 Explosives Searches at Protected Area Portals N/AI31 91-09 Modification of Surveillance Intetval for the Electrical Protective Ch 16 l Assemblies in Power Supplies for the Reactor Protection System Q 91-08 Removal of Component Lists from Technical Specifications Ch 16 91-07 GI-23,
- Reactor Coolant Pump Seal Failures" and Its Possible Effect on 20.2.5 i Station Blackout I
91-06 Resolution of Generic issue A-30, " Adequacy of Safety-Related DC Power 20.2.61; 1 Supplies" Pursur.nt to 10 CFR 50.54(f) 8.3.2 91-05 License Commercial - Grade Procurement and Dedication Programs N/Aldi 91 04 Changes in Technical Specification Surveillance Intervals to .ucommodate N/At51; a 24-Month Fuel Cycle 18 Month Fuel l Cycle: Ch 16 91-03 Reporting of Safeguards Events N/A'lI 91-02 Reponing Mishaps involving LLW Forms Prepared for Disposal N/At31 91 01 Removal of the Schedule for the Withdrawal of Reactor Vessel Material 10.7 Specimens from Technical Specificatior; Ch 16 90-09 Alternative Requirements for snubber Visual Inspection Intervals and N/A ts); Corrective Actions Inspection , Intervals based 1 on 18 Month l refueling cycle 9048 Simulation Facility Exemptions N/Al31 9047 Operator Licensing National Examination Schedule N/Al31 p. h , i Approvent Deeign Atatorial Arrtroeweian Pope 1.8-13 I l
System 80+ Design ControlDocument Table 1.8-2 Generic Letters Applicability Analysis to System 80+ (Cont'd.) No. Title Comment 9046 Resolution of Generic Issue 70, " Power Operated Relief Valve and Block 5.4.13; Valve Reliability", and Generic issue 94, " Additional Low-Temperature 5.2.2.10; Overpressure Protection for Light-Water Reactors", Pursuant to 10 CFR 20.2.18 50.54(O 90-05 Guidance for Performing Temporary Non-Code Repair of ASME Code N/AI31 Class 1,2, and 3 Piping 90-04 Request for Information on the Status of License Implementation of Ch 20 Generic Safety issues Resolved with imposition of Requirements or Corrective Actions 9043 Relaxation of Staff Position in Generic Letter 83-28, item 2.2 Pan 2 Ch 17 Sup!! " Vendor Interface for Safety-Related Components" 90-03 Relaxation of Staff Position in Generic Letter 83-28, Item 2.2 Part 2 Ch 17
" Vendor Interface for Safety-Related Components" 9042 Alternative Requirements for Fuel Assemblies in the Design Features N/At51 Sup!I Section of Technical Specifications 90-02 Alternative Requirements for Fuel Assemblies in the Design Features Ch 16 Section of Technical Specifications 90-01 Request for Voluntary Participation in the NRC N/Al43 Regulatory Impact Survey l
N/A'1 89-23 NRC Staff Response to Questions Pertaining to implementation of 10 CFR Part 26 89 22 Potential for increased Roof leads and Plant Area Flood Runoff Depth at 2.4 Licensed Nuclear Power Plants Due to Recent Change in Probable Maximum Precipitaion Criteria Developed by the National Weather Service 89-21 Request for Information Concerning Status of Implementation of Ch 20 Unresolved Safety Issue Requirements 89-20 Protected Area Long-Term llousekeeping N/Al31 89-19 Request for Action Related to Resolution of Unresolved Safety issue A-47 Ch 20
" Safety Implication of Control System in LWR Nuclear Power Plant" Pursuant to 10 CFR 50.54(0 89-18 Resolution of Unresolved Safety issue A-17. " Systems Interactions in Ch 20 Nuclear Power Plants" 89-17 Planned Administrative Changes to the NRC Operator Licensing Written N/Al31 Examination Process 89-16 Installation of a Hardened Wetwell Vent N/AIU 89-15 Emergency Response Data System N/AI31 89-14 Line-Item Improvements in Technical Specifications - Removal of the 3.25 Ch 16 Limit on Extending Surveillance Intervals SR 3.0.2 Approved Dessper MeterW krtroductiorr Page 1.8-14
System 80+ oes/gn contro/ Document 1 j ( 'V Table 1.8-2 Generic Letters Applicability Analysis to System 80+ (Cont'd.) No. Title Comment 1 89-13 Service Water System Problems Affecting Safety-Related Equipment N/Al31 Sup!1 89-13 Service Water System Problems Affecting Safety-Related Equipment 9.2.1 89-12 Operator Licensing Examinations N/Al31 89-11 Resolution of Generic Issue 101 " Boiling Water Reactor Water level N/Alli Redundancy" 89-10 Information on Schedule and Grouping, and Staff Responses to Additional N/AI31 Supt 6 Public Questions 89-10 Inaccuracy of Motor-Operated Valve Diagnostic Equipment N/Alli Sap!5 N/A'1l 89-10 Consideration of Valve Mispositioning in BWRs Sup!4 89-10 Consideration of the Results of NRC-Sponsored Tests of Motor-Operated N/Alli Sup!3 Valves ; 89-10 Availability of Program Descriptions N/Ald! ! /^\ Sup!2 89-10 Results of the Public Workshops 3.9.6.2 4 Supi1 89-10 Safety-Related Motor-Operated Valve Testing and Surveillance 3.9.6.2 89-09 ASME Section til Component Replacements N/Al31 89-08 Erosion / Corrosion - Induced Pipe Wall Thinning 3.6 l l i 89-07 Power Reactor Safeguards Contingency Planning for Surface Vehicle N/At31 Supt 1 Bombs 89-07 Power Reactor Safeguards Contingency Planning for Surface Vehicle App 13A Bombs 89-06 Task Action Plan item 1.D.2 - Safety Parameter Display System - 10 CFR 20.2.91 50.54(0 89-05 Pilot Testing of the Fundamentals Examination N/Al31 8944 Guidance on Developing Acceptable laservice Testing Programs 6.6 89-03 Operator Licensing National Examibation Schedule N/At31 89-02 Actions to improve the Detection of Counterfeit and Fraudulently Marketed N/Al33 Products 89-01 Implementation of Programmatic Controls for Radiological Effluent Ch 16 em Technical Specifications in the Administrative Controls Section of the Technical Specifications and the Relocation of Procedural Details of RETS (U ) to the Offsite Dose Calculation Manual or to the Process Control Program Approved Design Material- kstroducaion Page 1.8-15
System 80 + Design ControlDocument Table 1.8-2 Generic Letters Applicability Analysis to System 80+ (Cont'd.) 91 No. Title Comment 88-20 Individual Plant Examination of External Events for Severe Accident Ch 19 Sup!4 Vulnerabilities 88-20 Completion of Containment Performance Improvement Program and Ch 19 Supi3 Forwarding of Insights for Use in the Individual Examination for Severe Accident Vulnerabilities 88-20 Accident Management Strategies for Consideration in the Individual Plant Ch 19 Sup!2 Examination Process 88-20 Initiation of the Individual Plant Examination for Severe Accident N/A, Issue Sup!! Vulnerabilities - 10 CFR 50.54(f) covered by PRA (Ch 19) 88-20 Individual Plant Examtnation for Severe Accident Vulnerabilities N/A, Issue covered by PRA (Ch 19) 88-19 Use of Deadly Force by 1.icense Guards to Prevent Theft of SNM N/A13) 88-18 Plant Record Storage on Optical Discs N/AI31 88 17 I.oss of Decay Heat Removal 5.4.7;7.7.1 88-16 Removal of Cycle-Specific Parameter Limits from Technical Specifications Ch 16 88-15 Electric Power Systems- Inadequate Control over Design Process Ch 8 PRA design process 88-14 Instrument Air Supply System Problems Affecting Safety-Related 9.3.1 Equipment 88-13 Operator Licensing Examination N/AI3I 88-12 Removal of Fire Protection Requirements from Technical Specifications Ch 16 88-11 NRC Position on Radiation Embrittlement of Reactor vessel Materials and 5.3.1.6 its Impact on Plant Operations 88-10 NRC Position on Intergranular Stress Corrosion Cracking [lGSCC) in N/AI31 Supt 1 BWR Austenitic Stainless Steel Piping 88-10 Purchase of GSA Approved Security Containers N/A131 88-09 Pilot Testing of Fundamentals Exammations N/AI31 88-08 Mail Sent or Delivered to the Office of Nuclear Reactor Regulation N/Ald! 88-07 Modified Enforcement Policy Relating to 10 CFR 50.49, " Environmental N/A'lI Qualification of Electrical Equipment important to Safety for Nuclear Power Plants" t 88-06 Removal of Organization Charts from Technical Specification N/As) Administrative Control Requirements 88-05 Boric Acid Corrosion of Carbon Steel Reactor Vessel Boundary 3.11 Components in PWR Plants Approved Desiges Matedel kstrodacthm Page 1.8-16
System 80+ Design ControlDocument l O I D Table 1.8-2 Generic Letters Applicability Analysis to System 80+ (Cont'd.) l No. Title Comment 88-04 Distribution of Gems Irradiated in Research Reactors N/Alll 88-03 Resolution of Generic Safety Issue 93, " Steam Bindings of Auxiliary 20.2.25 Feedwater Pumps" 88-02 Integrated Safety Assessment Program 11 Level III PRA, Ch 19 88-01 NRC Position on Intergranular Stress Corrosion Cracking in BWR N/All! Sup!1 Austenitic Stainless Steel Piping 88-01 NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping N/All! 87-16 Transmittal of NUREG-1262, " Answers to Questions at Public Meeting N/Aldl Regarding Implementation of Title 10, Code of Federal Regulations, Part 55 on Operators' Licenses" 87-15 Policy Statement on Deferred Plants N/Al31 87-14 Operator Licensing Examinations N/Al31 87 13 Integrity of Requalification Examinations at Non-Power Reactors N/AI31 87-12 Loss of Residual Heat Removal While the Reactor Coolant System is See GL 88-17 V Partially Filled 87-11 Relaxation in Arbitrary Intermediate Pipe Rupture Requirements 3.6.2; 3.6.3 87 10 Implementation of 10 CFR 73.57, Requirements for FBI Criminal History N/A131 Checks 87-09 Sections 3.0 and 4.0 of the Standard Technical Specifications on the Ch 16 Applicability of Limiting Conditions for Operation and Surveillance Requirements 87-08 Implementation of 10 CFR 73.55 Miscellaneous Amendments and Search N/Al31 Requirements 87-07 Information Transmittal of Final Rulemaking for Rev.s to Operator N/Al31 Licensing - 10 CFR 55 and Conforming Amendments 87-06 Periodic Verification of Leak Tight Integrity of Pressure Isolation Valves N/Al31 87-05 Request for Additional Information - Assessment of License Measures to N/Alli Mitigate and/or Identify Potential Degradation of Mark I Drywells 87-04 Temporary Exemption from Provisions of the FBI Criminal History Rule N/Al31 for Temporary Workers 87 03 Verification of Seismic Adequacy of Mechanical and Electrical Equipment N/AI31 in Operating Reactors, Unresolved Safety Issue A-46 ,m 87-02 Supplemental Safety Evaluation Report No. 2 on SQUG Generic N/AI31 (] Supl1 Implementation Procedure. Rev. 2 Approved Design Material- introduction Pa.ge 1.8-17
System 80+ Design ControlDocument Table 1.8-2 Generic Letters Applicability Analysis to System 80+ (Cont'd.) No. Title Comment 87-02 Verification of Seismic Adequacy of Mechanical and Electrical Equipment N/AW in Operating Reactors, USI A-46 87-01 Public Availability of the NRC Operatc* Licensing Examination Question N/AW Bank 86-17 Availability of NUREG-Il69, " Technical Findings Related to Generic Issue N/AlI C-8; BWR Main Steam Isolation Valve leakage and Leakage Treatment Methods" 86-16 Westinghouse ECCS Evaluation Models N/AN 86-15 Information Relating to Compliance with 10 CFR 50.49, " Environmental See GL 88-07 Qualification of Electric Equipment important to Safety for Nuclear Power Plants." 86-14 Operator Licensing Examinations N/AW 86-13 Potential Inconsistency between Safety Analyses and Technical Ch 16 Specifications 86-12 Criteria for Unique Purpose Exemption from Conversion from the Use of N/AW HEU Fuel 86-11 Distribution of Products Irradiated ir. Research Reactors N/AW 86-10 Fire Endurance Test Acceptance Criteria for Fire Barrier Systems Used to 9.5.1 Sup!! Separate Redundant Safe Shutdown Trains Within the Same Fire Area 86-10 Implementation of Fire Protection Requirements 9.5.1 86-09 Technical Resolution of Generic Issue No. B-59-(N-1) Loop Operation in N/AW BWRs and PWRs 86-08 Availability of Supplement 4 to NUREG-0933, "A Prioritization of Generic Ch 20 Safety issues" 86-07 Transmittal of NUREG-1190 Regarding the San Onofre Unit 1 Loss of N/AW Power and Water Hammer Event 86-06 Implementation of TMI Action item II.K.3.5, " Automatic Trip of Reactor 20.2.121 Coolant Pumps" l 86-05 Implementation of TMI Action Item II.K.3.5, " Automatic Trip of Reactor N/Alll Coolant Pumps * [B&W) 8644 Policy Statement on Engineering Expertise on Shift N/AW 86-03 Applications for License Arr4ndments N/AW l 86-02 Technical Resolution of Generic issue B nermal Hydraulic Stability N/AW [BWR] N/A'1l 86-01 Safety Concerns Associated with Pipe Breaks in the BWR Scram System 85-22 Potential for Loss of Post-LOCA Recirculation Capability due to Insulation N/AW Debris Blockage l Approved Design Material knroduction Page 1.8-18
Sysfem 80+ Design ControlDocument V) Table 1.8-2 Generic Letters Applicability Analysis to System 80+ (Cont'd.) No. Title Comment 85-21 Not Issued. 85-20 Resolution of Generic Issue 69 [B&W) N/Atil 85-19 Reporting Requirements on Primaq Coolant lodine Spikes Ch 16 85-18 Operator Licensing Examinations N/Al31 85-17 Availability of Supplement 2 and 3 to NUREG-0933 See GL 8648 N/A'1l 85-16 High Boron Concentrations 85-15 Information Relating to the Deadlines for Compliance with 10 CFR 50.49, See GL 88-07
" Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plant" 85-14 Commercial Storage at Power Reactor Sites of Low-Level Radioactive N/Al33 Waste Not Generated by the Utility 85-13 Transmittal of NUREG-il54 Regarding the Davis-Besse Loss of Main and N/Al43 Auxiliary Feedwater Event 85-12 Implementation of TMI Action Item II.K.3.5, " Automatic Trip of Reactor N/Alli Coolant Pumps" [ Westinghouse]
n 85-11 Completion of Phase II of " Control of Heavy Imads at Nuclear Power N/AI43 k~ Plants" NUREG-0612 85-10 Technical Specifications for Generic Letter 83-28, items 4.3 and 4.4 N/Al!I [B&W) 85-09 Technical Specifications for Generic Letter 83-28, Item 4.3 (Westinghouse N/A l'1 NSSS) 85-08 Rev. of NRC Form 439, " Report of Terminating Individual's Occupational N/Al31 Addenda Exposure" 85-08 10 CFR 20.408 Termination Reports - Format N/AI31 85-07 Implementation of Integrated Schedules for Plant Modifications N/AI33 8546 Quality Assurance Guidance for ATWS Equipment that is not Safety- N/A'll Related 85-05 Inadvertent Boron Dilution Events 15.4.6 85-04 Operator Licensing Examinations N/Al31 85-03 Clarification of Equivalent Control Capacity for Standby Liquid Control N/Atti Systems [BWR) 85-02 Staff Recommended Actions Stemming from NRC Integrated Program for Ch 20; the Resolution of Unresolved Safety issues Regarding Steam Generator 5.4.2; 7.7.1.6; Tube Integrity 10.3.5
) 85-01 Fire Protection Policy Steering Committee Report 9.5.1 Approved Design Materkt - ktroduction Page 1.8-19
T System 80+ Design Contro/ Document Table 1.8-2 Generia Letters Applicability Analysis to System 80+ (Cont'd.) No. Title Comment 84-24 Certification of Compliance to 10 CFR 50.49, " Environmental N/At31 Qualification of Electric Equipment Important to Safety for Nuclear Power Plants
- 84-23 Reactor Vessel Water Level Instrumentation in BWRs N/ADI 84-22 Not issued.
84-21 Ieng Term Low Power Operation in Pressurized Water Reactors N/ADI l N/A'I 84-20 Scheduling Guidance for License Submittals of Reloads that involve Unreviewed Safety Questions 84-19 Availability of Supplement I to NUREG-0933, "A Prioritization of Generic See GL 85-17 Safety issues" 84-18 Filing of Applications for Licenses and Amendments N/Ald! 84-17 Annual Meeting to Discuss Recent Development Regarding Operator N/At31 Training, Qualifications, and Examinations 84-16 Adequacy of On-Shift Operating Experience for Near Term Operating N/ADI License Applicants 84-15 Proposed Staff Actions to improve and Maintain Diesel Generator 8.3.1.1 Reliability 84-14 Replacement and Requalification Training Program N/ADI 84-13 Technical Specification for Snubbers Ch 20; Ch 16 84-12 Compliance with 10 CFR Part 61 and Implementation of the Radiological N/At31 Effluent Technical Specifications and Attendant Process Control Program 84 11 Inspections of BWR Stainless Steel Piping N/AUI 84-10 Administration of Operating Tests Prior to initial Criticality N/ADI 84-09 Recombiner Capability Requirements of 10 CFR 50.44(cX3)(II) 6.2.5 84-08 Interim Procedures for NRC Management N/Aldl of Plant-Specific Lackfitting 84-07 Procedural Guidance for Pipe Replacement at BWRs N/AUI 84-06 Operator and Senior Operator License Examination Criteria for Passing N/AI31 Grade 84 05 Change to NUREG-1021, " Operator Licensing Exammer Standards" N/Al33 84-04 Safety Evaluation of Westinghouse Topical Repons Dealing with N/ADI Elimination of Postulated Pipe Breaks in PWR Primary Main Loops 84-03 Availability of NUREG-0933, "A Prioritization of Generic Safety issues" See GL 84-19 84-02 Notice of Meeting Regarding Facility Staffing N/ADI 84-01 NRC Use of the Terms, "Important to Safety and Safety Related* N/Aldl Approved Design Meterial Introduction Page 1.8-20
System 80 + Design ControlDocument V Table 1.8-2 Generic Letters Applicability Analysis to System 80+ (Cont'd.) No. Title Comment 83 44 Availability of NUREG-1021, " Operator Licensing Examiner Standards" N/AI31 ; 83-43 Reponing Requirements of 10 CFR Pan 50, Sections 50.72 and 50.73, and N/Aldi Standard Technical Specifications 8342 Clarification to Generic Letter 81-07 Regarding Response to NUREG-0612, 9.1.4
" Control of Heavy Imads at Nuclear Power Plants" 83-41 Fast Cold Stans of Diesel Generators N/Aldi 83-40 Operator Licensing Examination N/AI31 83-39 Voluntary Survey of Licensed Operators N/Al31 83-38 NUREG-0%5, "NRC Inventory of Dams" N/Aldi 83-37 NUREG-0737 Technical Specifications Ch 16 83-36 NUREG-0737 Technical Specifications [BWR] N/Alll 83-35 Clarification of TMI Action Plan item II.K.3.31 20.2.121 83 34 Not Issued.
NRC Positions on Cenain Requirements of Appendix R to 10 CFR 50 9.5.1 n s 83-33 (_ l 83-32 NRC Staff Recommendations Regarding Operator Action for Reactor Trip 9.9.1.1.11 and ATWS 83-31 Safety Evaluation of " Abnormal Transient Operating Guidelines * [B&W) N/All! 83-30 Deletion of Standard Technical Specification Surveillance Requirement N/At5] 4.8.1.1.2.d.6 of Diesel Generator Testing 83 29 Not Issued. 83-28 Required Actions Based on Generic implications of Salem ATWS Events N/Al33 Supt 1 83-28 Required Actions Based on Generic Implications of Salem ATWS Events N/Al31 83-27 Surveillance Intervals in Standard Technical Specifications See GL 91-04 83-26 Clarification of Surveillance Requirements for Diesel Fuel impurity Level N/Aldl Tests 83-25 Not issued. 83-24 TMI Task Action item I.G.1, "Special Low Power Testing and Training", N/AlI Recommendations for BWRs 83-23 Safety Evaluation of " Emergency Response Guidelines * [C-E] N/Al41 83-22 Safety Evaluation of " Emergency Response Guidelines * [ Westinghouse] N/Atti 83-21 Clarification of Access Control Procedures for Law Enforcement Visits N/Aldi 73 N/A'1l
- 83-20 Integrated Scheduling for Implementation of Plant Modifications Approved Design Material- entroduction Page 1.8-21
System 80+ Design controlDocument Table 1.8-2 Generic Letters Applicability Analysis to System 80+ (Cont'd.) No. Title Comment 83 19 New Procedures for Providing Public Notice Concerning Issuance of N/Aldi Amendment to Operating Licenses 83 18 NRC Staff Review of the BWR Owner's Group Control Room Survey N/Atti Program 83-17 Integrity of the Requalification Exammation for Renewal of Reactor N/At31 Operator and Senior Reactor Operator Licenses 83-16 Transmittal of NUREG-0977 Relative to the ATWS Events at Salem N/Aldl Generating Station, Unit No.1 83-15 Implementation of Regulatory Guide 1.150, " Ultrasonic Testing of Reactor 5.1.4; 5.3.1 Vessel Welds during Preservice and Inservice Examinations, Rev.1" 83-14 Definition of " Key Maintenance Personnel" N/At3) 83-13 Clarification of Surveillance Requirements for HEPA Filters and Charcoal Ch 16 Adsorber Units in Standard Technical Specifications on ESF Cleanup Systems 83-12A Issuance of NRC Form 398 - Personal Qualifications Statement - License N/Ald! 83-12 Issuance of NRC Form 398 - Personal Qualifications Statement - License N/Aldi 83 11 License Qualification for Performing Safety Analyses in Support of N/Aldl Licensing Actions 83-10f Resolution of TMI Action item II.K.3.5, " Automatic Trip of Reactor N/Alll Coolant Pumps" (B&W NSSS) 83-10e Resolution of TMI Action item II.K.3.5, " Automatic Trip of Reactor N/Al11 Coolant Pumps" (B&W NSSS) 83-10d Resolution of TMI Action Item II.K.3.5, " Automatic Trip of Reactor N/Alll Coolant Pumps" (W NSSS) 83-10c Resolution of TMl Action item II.K.3.5, " Automatic Trip of Reactor N/Alll Coolant Pumps" (W NSSS) 83-10b Resolution of TM1 Action Item II.K.3.5, " Automatic Trip of Reactor See GL 86-06 Coolant Pumps" (C-E NSSS) 83-10a Resolution of TMI Action Item II.K.3.5., " Automatic Trip of Reactor See GL 86-06 Coolant Pumps" (C-E NSSS) 83-09 Review of Combustion Engineering Owner's Group Emergency Procedures See GL 83-23 Guideline Program 83-08 Modification of Vacuum Breakers on Mark I Containments N/A!!! 83-07 The Nuclear Waste Policy Act of 1982 N/Ald! 83-06 Certificates and Revised Format for Reactor Operator and Senior Reactor N/At31 Operator Licenses 83-05 Safety Evaluation of " Emergency Procedure Guidelines, Rev. 2", NEDO- N/Alll 24934, June 1982 [BWRs] Approwd Design Material httro6wtion Page 1.8-22
i !. 1 i i System 80+ Des /an controlDocument i 1 O ,O Table 1,8-2 Generic Letters Applicability Analysis to System 80+ (Cont'd.) ! l No. Title Comment l
' N/A'1 83-04 Regional Workshops Regarding Supplement I to NUREG-0737, Requirernents for Emergency Response Capability 8343 Not Issued.
83-02 NUREG-0737 Technical Specifications (BWRs] N/Al11 I N/A)) 83-01 Operator Licensing Examination Site Visit l N/A'1 82-39 Problems with the Submittals of 10 CFR 72.21 Safeguards Information for Licensing Review 82-38 Meeting to Discuss Recent Developments for Operating Licensing N/Al31 Examinations 82-34/37 Not issued. 82-33 Supplement I to NUREG-0737 - Requirements for Emergency Response Ch 20
- Capability l
N/A'1 82-32 Subjects Requirements 82-31 Not issued. l N/A'1
/ 82-30 Filing Relating to 10 CFR 50 Production and Uplization Facilities 82-29 Not Issued.
82 28 Inadequate Core Cooling Instrumentation System 6.3.5;7.5 82 27 Transmittal of NUREG4763 and NUREG4783 [BWRs] N/Al1 82-26 NUREG-0744 Rev.1 - Pressure Vessel Material Fracture Toughness 5.3.1 l N/A'1 82-25 Integrated IAEA Exercise for Physical Inventory at LWRs 82-24 Safety / Relief Valve Quencher loads: Evaluation for BWR Mark 11 and Ill N/AlI Containments 82-23 Inconsistency between Requirements of 10 CFR 73.40(d) and Standard N/A'I l Technical Specifications for Performing Audits for Safeguards Contingency Plans 82 22 Steam Generator Tube Integrity N/Al31 82 Technical Specifications for Fire Protection Audits N/AI31 l N/A'1 82-20 Guidance for Implementing Standard Review Plan Rule l N/A'1 82-19 Submittal of Copies of Document to NRC 82 18 Reactor Operator and Senior Reactor Operator Requalification N/Al31 Examinations 82 17 Inconsistency Between Requirements of 10 CFR 50.54(t) and Standard N/Al31
.b Technical Specifications for Performing Audits of Emerpncy Preparedness
,' U Programs 82-16 NUREG-0737 Technical Specifications N/A l'1 AMwevenf Des 4pr nieterial hstroetuetigwr Page 1.8-23 v - ,,
System 80 + Design controlDocument Table 1.8-2 Generic Letters Applicability Analysis to System 80+ (Cont'd.) No. Title Comment 82-15 Not Issued. N/A'1l 82-14 Submitta! of Documents to the Nuclear Regulatory Commission 82-13 Reactor Operator and Senior Reactor Operator Examinations N/Al31 82-12 Nuclear Power Plant Staff Working Hours N/Al31 82 11 Transmittal of NUREG-0916 Relative to the Restart of R. E. Ginna N/A'll Nuclear Power Plant 82-10 Post-TMI Requirements Ch 20 N/A'1l 82-09 Environmental Qualification of Safety-Related Electrical Equipment Transmittej of NUREG-0909 Relative to the Ginna Tube Rupture N/A'1l 82-08 Transmittal of NUREG-0909 Relative to the Ginna Tube Rupture N/A'1l 82-07 82-06 Not issued. 82-05 Post-TMI Requirements See GL 82-10 82-04 Use of INPO SEE-IN Program N/At41 82-03 High Burnup MAPLHGR Limits [BWRs] N/Al11 82-02 Nuclear Power Plant Staff Working Hours N/Al31 82-01 New Applications Survey N/AI41 81-40 Qualifications of Reactor Operators N/Al33 N/A'Il 81 39 NRC Volume Reduction Policy 81-38 Storage of Low Level Radioactive Wastes at Power Reactor Site N/Al31 81-37 ODYN Code Reanalysis Requirements [BWRs] N/Alll 81-36 Revised Schedule for Completion of TMI Action Plan item II.D.1, Relief 5.4.13; and Safety Valve Testing 20.2.106 81-35 Safety Concems Associated with Pipe Breaks in the BWR Scram System N/AI1 81-34 Safety Concerns Associated with Pipe Breaks in the BWR Scram System N/Alll 81-33 Not Issued. 81-32 NUREG-0737, Item li.K.3.44 - Evaluation of Anticipated Transients N/Alll Combined with Single Failure [BWRs] 81-31 Not issued. 81-30 Safety Concerns Associated with Pipe Breaks in the BWR Scram System N/Al31 81 29 Simulator Examinations N/At3] 81-28 Steam Generator Overfill N/At31 81-27 Privacy and Proprietary Material in Emergency Plans N/Al31 l Approved Design Material Introduction Page 1.8-24 i
System 80+ Design controlDocument V Table 1.8-2 Generic Letters Applicability Analysis to System 80+ (Cont'd.) No. Title Comment l N/A'1 81-26 Safety Concerns Associated with Pipe Breaks in the BWR Scram System 81-25 Change in Implementing Schedule for Submission and Evaluation of N/AI31 Upgraded Emergency Plans 81-24 Multi-Plant Issue B-56 Control Rods Fail to Fully Insert [BWRs] N/Alll l N/A'1 81-23 A INPO Evaluation Reports l N/A'1 81-23 INPO Plant Specific Evaluation Report 81-22 Engineering Evaluation of the II.B. Robinson Reactor Coolant System N/Aldl leak on 1/27/81 81-21 Natural Circulation Cooldown 5.4.1 81-20 Safety Concerns Associated with Pipe Breaks in the BWR Scram System N/All! 81 19 Thermal Shock to Reactor Pressure Vessels 5.3.1 81-18 BWR Scram Discharge System - Clarification of Diverse Instnamentation N/Alli Requirements 81-17 Functional Criteria for Emergency Response Facilities N/At3) A 81-16 NUREG-0737 Item I.C.1 SER on Abnormal Transient Operating N/Alil Guidelines (B&W] Environmental Qualification of Class IE Electrical Equipment - l N/A'1 81-15 Clarification of Staffs Handling of Proprietary Information 81 14 Seismic Qualification for Auxiliary Feedwater System 10.4.9 81 13 SER for GEXL Correlation for 8x8R Fuel Reload Applications N/Alll 81-12 Fire Protection Rule 9.5.1.3.6 R1 81-11 BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle N/Al8l Cracking 81-10 Post-TMI Requirements for the Emergency Operations Facility N/AI31 81-09 BWR Scram Discharge System N/Alll 81-08 ODYN Code [BWRs] N/Alll 81-07 Control of Heavy Loads 9.1.4 81-06 Periodic Updating of Final Safety Analysis Reports N/Al31 81-05 Information Regarding the Program for Environmental Qualification of N/Al31 l Safety-Related Electrical Equipment j 81-04 Emergency Procedures and Training for Station Blackout Events N/Al33 (m G
! 81-03 Implementation of NUREG 0313 [BWRs] N/Alll ;
81-02 Analysis, Conclusion and Reconunendation Concerning Operator Licensing N/Al31 Approved Design Matonial krsroduction Page 1.8-25 I l
System 80+ Design ControlDocument Table 1.8-2 Generic Letters Applicability Analysis to System 80+ (Cont'd.) i No. Title Comment 81-01 Qualification of Inspection, Examination and Audit Personnel N/Al31 80-109 Guidelines for SEP Soil-Structure Interaction Reviews App 3.7B 80-106 Report on ECCS Cladding Models, NUREG-0630 Addressed by GL 80-01 80-99 Technical Specifications Revisions for Snubber Surveillance Ch 16; Ch 20 80-35 Effect of a DC Power Supply Failure on ECCS Performance 8.3.2; 15.3 80-30 Clarifications of the Term ' Operable" as it Applies to the Single-Failure Ch 16 Criterion for Safety Systems Required by Technical Specifications 80-19 Resolution of Fnhanced Fission Gas Release Concern 1.6; 4.2.1; 4.2.3 8041 NUREG-0630, " Cladding, Swelling, and Rupture - Models for LOCA 1.6; 4.2.3: Analysis" Ch 6 Exclusion criteria for Not Applicable (N/A) items: I'l The item specifically identifies another design or vendor. 121 The item is specific to components, structures or systems which are not included in the System 80+ Standard Design. 131 The item is relevant to plant operations or is specific to a panicular plant design. 141 The item includes no design requirements. 151 The item is not mandatory but is an alternative which can be implemented as desired by an applicant. O Approved Design MotorAel krtroduction Page 1.8-26
System 80+ Design controlDocument ( V Table 1.8-3 NRC Bulletins Applicability Analysis to System 80+ No. Title Comment 93-03 Resolution of issues Related to Reactor Vessel Water Level Instrumentation N/Alli in DWRs 93-02 and Debris Plugging of Emergency Core Cooling Suction Strainers N/Al31 Sup!1 93-01 Release of Patients after Brachytherapy Treatment with Remote Afterloading N/At2) Devices 92-03 Release of Patients after Brachytherapy N/Al2) 92-02 Safety Concerns Relating to "End of Life" of Aging Theratronics Teletherapy N/At21 Units 92-01 Failure of Thermo-Lag 330 Fire Barrier System to Perform its Specified Fire N/AI3I 1 Sup!1 Endurance Function 92-01 Failure of Thermo-Lag 330 Fire Barrier System to Maintain Cabling and N/AI3I Wide Cable Trays and Small Conduits Free from Fire Damage 91-01 Reporting Loss of Criticality Safety Controls N/A'I l 90 Loss of Thermal Margin Caused by Channel Box Bow N/Al33 90-01 and Loss of Fill-Oil in Transmitters Manufactured by Rosemount N/At21 Supt 1 89-03 Potential Loss of Required Shutdown Margin During Refueling Operation N/Al31 8942 Stress Corrosion Cracking of High-Hardness Type 410 Stainless Steel N/Al33 Internal Preloaded Bolting in Anchor Darling Model S350W Swing Check Valves or Valves of Similar Design 8941 and Failure of Westinghouse Steam Generator Tube Mechanical Plugs N/All! Sup!1,2 88-11 Pressurizer Surge Line Thermal Stratification 5.2, 5.4, 3.9A 88-10 and Nonconforming Molded-Case Circuit Breakers N/Al31 Supt 1 8849 Thimble Tube Thinning in Westinghouse Reactors N/Alli 88-08 and Thermal Stress in Piping Connected to Reactor Coolant Systems 3.9, 3.9A Sup! 1,2,3 88-07 and Power Oscillations in Boiling Water Reactors N/Alli Sup!1 8846 Actions to be Taken for the Transportation of Model No. Spec 2-T N/Ag2r Radiographic Exposure Device 4 , 'd 4 proved Dess/ pre Mefords/ hetroduceion re,pe 7.8-27
ytem 80+ Design ControlDocument Table 1.8-3 NRC Bulletins Applicability Analysis to System 80+ (Cont'd.) No. Title Comment 88-05 and Nonconforming Materials Supplied by Piping Suppliers, Inc. N/A (2) Sup!1,2 88-04 Potential Safety-Related Pump less App 19.8A; RAI 440.70; 5.4.7.2 88-03 Inadequate Latch Engagement in HFA Type Latching Relays Manufactured by N/AI31 GE Company 88-02 Rapidly Propagating Fatigue Cracks in Steam Generator Tubes [ Westinghouse) N/AI 'l N/A'1l 88-01 Defects in Westinghouse Circuit Breakers 87-02 and Fastener Testing to Determine Conformance with Applicable Material N/Al31 Supt 1 Specifications 87-01 Thinning of Pipe Walls in Nuclear Power Plants 3.6 Defective Teletherapy Timer that May Not Terminate Treatment Dose N/Ag2) 86-04 8643 Potential Failure of Multiple ECCS Pumps Due to Single Failure of Air- N/AMI Operated Valve in Minimum Flow Recirculation Line Static 'O" Ring Differential Pressure Switches N/Ag21 86-02 86-01 Minimum Flow Imgic Problems that Could Disable RWR Pumps [BWRs] N/Alll 85-03 and Motor-Operated Valve Common Failures During Plant Transients Due to N/At31 Supt 1 improper Switch Settings 85-02 Undervoltage Trip Attachments of Westinghouse DB-50 Type Reactor Trip N/Al!! Breakers 85-01 Steam Bincling of Auxiliary Feedwater Pumps 10.4.9 84-03 Refueling Cavity Water Seal 9.1 i 84-02 Failures of General Electric Type HFA Relays in Use in Class IE Safety N/Alli Systems 84-01 Cracks in Boiling Water Reactor Mark I Containment Vent Headers N/Att j 83-08 Electrical Ctreuit Breakers with an Undervoltage Trip Feature used in Safety- N/AI31 Related Applications N/Az)t 83-07 and Apparently Fraudulent Products Sold by Ray Miller, Inc. Sup!1,2 83-06 Nonconforming Materials Supplied by Tube-Line Corporation Facilities N/At21
~
83-05 ASMf Code Pumps and Spare Parts Manufactured by the Haywood Tyler N/At21 Pump Company 8344 Failure of Undervoltage Trip Function of Reactor Trip Breakers , N/AW 83-03 Check Valve Failures in Raw Water Cooling Systems of Diesel Generators , 9.5.5;396 83-02 Stress Corrosion Cracking in Large-Dlameter Stainless Steel Recirculation N/AIU System Piping at BWR Plants ) 83-01 Failure of Trip Breakers to Open on Automatic Trip Signal N/Al33 82-04 Deficiencies in Primary Containment Electrical Penetrak Assemblies , N/Al31 Approveef Design Meterial kutroduction Page 1.85 l
l System 80 + oesign ControlDocument
/~'s i d Table 1.8-3 NRC Bulletins Applicability Analysis to System 80+ (Cont'd.)
No. Title Comment 82-03 Stress Corrosion Cracking in Thick-Wall, Large Diameter, Stainless Steel, N/AIU Recirculation System Piping at BWR Plants 82-02 Degradation of Threaded Fasteners in the Reactor Coolant Pressure Boundary 5.2.3; 5.3.1 of PWR Plants 82-01 and Alteration of Radiographs of WeldsE Piping Subassemblies N/At31 Rev 1 81-03 Flow Blockage of Cooling Water to Safety System Components N/Al31 81-02 and Failure of Gate Type Valves to Close against Differential Pressure 3.9.6; GL Sup!1 89-10 81-01 Surveillance of Mechanical Snubbers N/AI31 80-24 Prevention of Damage Due to Water Leakage inside Containment Ch 20 (HF); N/Al31 N/A'1 l 80-20 Failures of Westinghouse Type W-2 Spring Rerum to Neutral Control Switches 80 F Failure of Mercury-Wetted Matrix Relays in Protection Systems of Plants N/At2] Designed by C-E 80-18 Maintenance of Adequate Minimum Flow Through Centrifugal Charging Ch 20 Pumps Following Secondary-Side High-Energy Line Rupture (GI-23) g , Q 80-15 Possible less of Emergency Notification System with 1.oss of Off-Site Power N/Aldl GL 91-14 80-11 Masonry Walt Design N/At21; 3.8.4.6 80-08 Examination of Containment Liner Penetration Welds N/AI33 80-06 Engineered Safety Feature Reset Controls N/Al33 80-05 Vacuum Conditions Resulting in Damage to Chemical Volume Control System 11.2 Holdup Tads 80-04 Arsysis of a PWu main Steamline Break with Continued Feedwater Addition 6.2; Ch 20 (GI 125.11.07) 80-03 Loss of Charcoal from Standard Type II,2-inch, Tray Absorber Cells 6.4;9.4.1 Exclusion criteria for Not Applicable (NIA) items: I'3 The item specifically identifies another design or vendor. I23 The item is specific to components, structures or systems which are not included in the System 80+ Standard Design. I33 The item is relevant to plant operations or is specific to a particular plant design. l*3 The item includes no design requirements.
\.J t51 The item is not mandatory but is an alternative which can be implemented as desired by an applicant.
Approved Design Materie!- hegoducaiort Page 1.8-29
i I System 80+ Design ControlDocument Table 1.8-4 Deviations from the U.S. NRC Standard Review Plan O: Comment or SRP Section/ritte Summary Description of Deviation Sectilon 2.5.2 Vibratory Ground Motion - OBE is not considered in the design basis. Table 2.0-1; Rev. 2 2.5.2.7 3.6.2 Determination of Rupture The application of leak-before-break methodology 3.6.2.1; locations and Dynamic Effects eliminates dynamic effects of postulated pipe mpture 3.6.3 Associated With the Postulated in the System 80+ Standard Design for Class 1 Rupture of Piping - Rev.1, piping with a diameter of twelve inches or greater July 1981 and for the main steam line. A leakage crack is postulated in place of a circumferential break, or longitudinal break, or through-wall crack if justified by leak-before break analyses. Exceedance of 2.4 S., using Eq. (12) or Eq. (13) of 3.6.2.1.4 Paragraph NB-3653 of the ASME Code, is included as criteria for determining intermediate break locations in Class I piping. 3.7.3 Seismic Subsystem Analysis - Alternate analysis methods are employed for piping 3.7.3.1 Rev. 2 systems. , 3.7.3 Seismic Subsystem Analysis - No explicit range of the fundamental frequencies of 3.7.3.4; Rev. 2 components and equipment with respect to the 3.7.3.8 dominant frequencies of the support structure is made. 3.10 Seismic and Dynamic Qualification tests will be performed at the time of 3.10 Qualification of Mechanical and specific equipment procrement. Methodology and Electrical Equipment criteria are summarized. 4.2 Fuel System Design - Rev. 2, With the application of the limiting factor for fuel 4.2.1.2 July 1981 assembly lateral deflection to the fuel assembly structure, no specific limit on lateral fuel rod deflection is provided. The Chapter 15 safety analysis uses the DNB convolution criterion for fuel failure; not the 95/95 Specified Acceptable Fuel Design Limit described in Section 4.4. Post-irradiation programs will be described in site-specific SARs. 4.4 Thermal and Hydraulic Design The effects of fuel densification are not included in 4.4.2.9 l
- Rev.1. July 1981 the calculation of total heat flux factor and linear heat generation rate because it is negligible.
O ApprovedDesign Atatorial-hstroduction Page 1.8-30
System 80+ Design ControlDocument A Table 1.8-4 Deviations from the U.S. NRC Standard Review Plan (Cont'd.) Comment or , SRP Sectionfritle Summary Description of Deviation Section 4.5.1 Control Rod Drive Stmetural The usage of control drive structural 4.5.1.2 Materials - Rev. 2, July 1981 material with a yield strength greater than 90 Kpsi is limited to the steel ball in the vent valve on the top of the CEDM and beanng inserts in the motor assembly. In lieu of the ASTM A262 Method E as 4.5.1.3 required in Regulatory Guide 1.44, ASTM A708 Strauss Test is employed in the System 80+ Standard Design for demonstrating freedom from sensitization in the fabricated unstabilized austenitic stainlesssteel. 4.5.2 Reactor Intemal and Core ASTM A708 Strauss Test is used for 4.5.2.3 j Support Materials - Rev. 2, sensitization test in fabricated unstabilized : July 1981 austenitic stainless steel. t , 4.6 Functional Design of Control No isolation between the CEDMs and the 4.6.2.2 Rod Drive System - Rev.1, CEAs is required because no non-
) July 1981 essential elements are involved at the d interface between these two systems.
5.2.1.1 Compliance with the Codes Specific codes and editions are identified 1.8 and Standards Rule,10 CFR i in Table 1.8-6 for design certification. 50.55a - Rev. 2, July 1981 5.2.3 Reactor Coolant Pressure The electroslag weld process is not used 5.2.3.3 Boundary Materials - Rev. 2, in the fabrication of any RCPB Jaly 1981 components. The specific recommendations of Regulatory Guide 1.71 for welder qualification for areas of limited accessibility are not completely followed, but performance qualifications for welders for those areas are conducted in accordance with the requirements of ASME Code Sections III and LX. The ASTM A708 Strauss Test or ASTM 5.2.3.4 A262 Practice E (modified Strauss) is used for sensitization test of fabricated unstabilized stainless steel. 5.3.1 Reactor Vessel Materials - - Actual reactor vessel materials will be 5.3.1.5 Rev.1, July 1981 tested at the time of material g' procurement. Test requirements are ( N described in Section 5.3.1.5. Approved Design menerw- huroducGon Page 1.8-31
l i System 80 + oesign controlDocument 1 Table 1.8-4 Deviations frorn the U.S. NRC Standard Review Plan (Cont'd.) 0l1 Comment or SRP Section/ Title Summary Description of Deviation Section 7.6.1 Residual Heat Removal (RHR) Interlocks for RHR suction isolation 7.6.1.1 System - Rev. 3, Apri! 1984 valves are not diverse. 6.2.1.1. A PWR Dry Contamments. The containment design pressure criteria 6.2.1.1.3 Including Subatmospheric for CP stage are not applicable to the Containments - Rev. 2, July System 80+ Standard Design. 1981 Analytical results of inadvertent operation 6.2.1.1 of containment heat removal systems exhibit that no special provisions against damage from external pressure conditions are required in th: System 80+ Standard Design. 6.2.1.2 Subcompartment Analysis - Due to application of leak-before-break, 6.2.1.2 Rev. 2, July 1981 the dynamic effects of pipe ruptures in ( containment subcompartments is not considered. Mass and Energy Release Metal-water reaction energy is not 6.2.1.3 I 6.2.1.3 Analysis for Postulated Loss- included in the mass / energy source terms l of-Coolant Accidents - Rev.1, since this energy has been shown to have ( July 1981 a very small effect on the containnwnt ! pressure. 6.2.2 Containment Heat Removal The in-containment refueling water 6.2.2.2; Systems - Rev 4, October storage tank eliminates the switchover to 6.5 1985 the recirculation mode of operation of the containment spray system. 6.2.4 Containment isolation System - The Chapter 15 dose analysis showed the l Rev. 2, July 1981 acceptability of 30 second closure times i for the purge valves. l 6.4 Control Room Habitability Operator wash room and kitchen are 6.4 l System - Rev. 2, July 1981 located outside the emergency zone. l 6.5.3 Fission P'oduct Control The System 80+ analysis assumes more App.ISA Systems and Structures - Rev. than 50% mixing. 2, July 1981 6.6 laservice inspection of Class 2 The ISI program is summarized in 6.6 ) and 3 Components - Rev.1, Section 6.6, but lists of components to be July 1981 inspected will be provided as part of the owner / operator's detailed mspectiori J program. l Some limits may be imposed on welding 6.6.2 area accessibility. , 1 AppromiDesign Atatorin!-Introduction Page 1.8-32
+
- Sy' tem 80+ Design ControlDocument fh (v) Table 1.8-4 Deviations from the U.S. NRC Standard Review Pian (Cont'd.)
Comment or SRP Section/ Title Summary Description of Deviation Section 11.1 Source Terms - Rev. 2, July Cost-benefit analysis for radioactive 11.1 1981 waste management systems is defered to the site-specific application. Cost-benefit analysis for radwaste augments used in the calculation of effluent releases to the environment is defered to the site-specific application. 11.2 Liquid Waste Management Cost-benefit analysis for liquid waste 11.2.6.4 Systems - Rev. 2, July 1981 management systems is deferred to the site-specific application due to the site-specific nature of population dose ; analyses. The plant transients which might occur 11.2.2 less frequently than once per fuel cycle are not taken into account for the design of waste collection tanks and waste sample tanks. 11.3 Gaseous Waste Management Cost-berefit analyses for gaseous waste 11.3.6.5 Systems - Rev. 2, July 1981 management systems is deferred to the , site-specific application. 12.2 Radiation Sources - Rev. 2, The shielding analysis will be performed ' July 1981 subsequent to component procurement and detailed piping design (layout).
)
12.3; Radiation Protection Design The shielding analysis will be performed 12.4 Features - Rev. 2, July 1981 subsequent to component procurement and detailed piping layouts. 15.1.5 Steam System Piping Failures Fuel rod failures are assumed based on 4.4.4.1 l Inside and Outside of the DNB convolution method. Containment (PWR) - Rev. 2, July 1981 Leak-Before-Break analysis and criteria 3.6.2.2.1 are applied to the Main Steam Line. l 15.4.6 Chemical and Volume Control Any single active component failure or 13.4.6.1 l i System Malfunction that single operator error has a negligible Results in a Decrease in Boron adverse impact on the accident Concentration in the Reactor consequences. Coolant (PWR) - Rev.1, July , 1981 I i 0 L)% i l ApprovedDesign A4aterial-kutroduction (2/95) Page 1.8-33 .
Sy ~ tem 80 + Design ControlDocument Table 1.8-4 Deviations from the U.S. NRC Standard Review Plan (Cont'd.) Comment or SRP Section/ Title Summary Description of Deviation Section 15.6.3 Radiological Consequences of The dose in the exclusion area boundag 15.6.3.3; Steam Generator Tube Failure for the postulated accident with an Table 15.6.3-9 (PWR) - Rev. 2, July 1981 accident initiated iodine spike and two single failures is calculated. 15.8 Anticipated Trar.sients Without ATWS events are not within the design 19.3.3.8; Scram - Rev.1, July 1981 basis and, therefore, their analysis are 19.4.13 presented as part of the PRA. 16.0 Technical Specifications - Rev. The System 80+ Technical Specif cation 16.0 1, July 1981 input will be based on NUREG-14 32. O O Approved Design Material
- knroduction page 1,s.34 i
i System 80+ Design controlDocument l Table 1.8-5 Standard Review Plan Compliance Comments Comment or SRP Section/ Title Summary Description of Deviation Section 2.1.1 Site Location and Description - The System 80+ Standard Design is 2.0; Rev. 2, July 1981 based on a set of site-related parameters Table 2.0-1; j which were selected to envelope most 2.1.1 potential nuclear power plant sites in the United States. 2.1.2 Exclusion Area Authority and No specific parameters on exclusion 2.1.2 Control Rev. 2, July 1981 area authority and control were employed in the evaluation of the System 80+ Standard Design. 2.1.3 Population Distribution - Site-specific SARs will provide 2.1.3 Rev. 2, July 1981 information relevant to requirements of the acceptance criteria. No specific ; parameters were employed in the i evaluation of the System 80+ Standard Design. 2.2.1; identification of Potential Site-specific SARs will provide data to 2.2.1; 2.2.2 Hazards in Site Vicinity - Rev. ensure that siting criteria for the System 2.2.2 [m L)
\ 2, July 1981 80+ Standard Design are met.
2.2.3 Evaluation of Potential Site-specific SARs will evaluate the 2.2.3 { Accidents - Rev. 2, July 1981 influence of site-specific, offsite potential accidents on the plant design. 2.3.1 Regional Climatology - Rev. 2, A set of regional climatology parameters 2.3.1; July 1981 was selected to envelope most potential Table 2.0-1 nuclear power plant sites in the United States. 2.3.2 Local Meteorology - Rev. 2, Local meteorological information will be 2.3.2; July 1981 presented in site-specific SARs. Table 2.0-1; 2.3.4; 2.3.5 2.3.3 Onsite Meteorological Onsite meteorological programs and 2.3.3 - Measurements Programs - Rev. measurements will be presented in site-
- 2. July 1981 specific SARs.
2.3.4 Short-term Dispersion in lieu of site meteorological data, a 2.3.4 Estimates for Accidental specified set of atmospheric conditions is Atmospheric Releases - Rev.1, employed to determine the values of July 1981 short-term diffusion estimates for the System 80+ Standard Design accident analyses. 2.3.5 long-term Diffusion Estimates In lieu of site meteorolo@a! data, 2.3.5 -
,Q - Rev. 2, July 1981 conservative atmospheric conditions are GI specified to determine the values cf long-term diffusion estimates.
i ANweved Design ninewW humedwedoor (11/96) Page 1.8-35
i System 80+ Design ControlDocument Table 1.8-5 Standard Review Plan Compliance Comments (Cont'd.) Comment or SRP Section/ Title Summary Description of Deviation Section 2.4.1 Hydrologic Description - Rev. The site-specific SAR will demonstrate 2.4; 2, July 1981 that the site parameters specified in the Table 2.0-1 System 80+ Standard Design are met. 2.4.2, Floods - Rev. 3, April 1989 The site-specific SAR will demonstrate 2.4; that the site paramete:s specified in the Table 2.01 System 80+ Standard Design are met. 2.5.1 Basic Geologic and Seisnaic Informati on will be provided in site- 2.5.1 Information - Rev. 2, July specific SARs to show that the System 1981 80+ envelope is met. 2.5.2 Vibratory Ground Motion - Information will be provided in site- 2.5.2.6; Rev.2 specific SARs to show that the System Table 2.0-1 80+ envelope is met. 2.5.2 Vibratory Ground Motion - The complete historical record of 2.5.2.1 Rev. 2 (Continued) earthquakes in the site region will be listed in the site-specific SAR 2.5.3 Surface Faulting - Rev. 2, July The site-specific SAR will present an 2.5.3 1981 evaluation to demonstrate compliance , with the SRP acceptance criteria. l l 2.5.4 Stability of Subsurface The site-specific SAR will present an 2.5.4 l Materials and Foundations - evaluation to demonstrate compliance Rev. 2 July 1981 with the SRP acceptance criteria. 2.5.5 Stability of Slopes - Rev. 2, The site-specific SAR will present an 2.5.5 July 1981 evaluation for stability of slopes to demonstrate compliance with the SRP acceptance criteria. 3.3.1 Wind leading - Rev. 2, July In lieu of site-specific value, the design 3.3.1.1; 1981 wind velocity of 110 mph, at the height Table 2.0-1 of 33 feet above nominal ground elevation with an importance factor of 1.11 is used ! as the most severe wind velocity for a 100 year recurrence interval. 3.4.1 Flood Protection - Rev. 2, July Compliance will be based on a site- 3.4.4 J 1981 specific evaluation l 3.4.2 Analysis Procedures - Rev. 2, A description of analysis procedures will 3.4.5 July 1981 be detailed in site-specific SAR. 3.5.1.5 Site Proximity Missiles Justification will be provided in the site- 3.5.1.5 (Except Aircraft) - Rev.1, specific SAR. July 1981 3.5.1.6 Aircraft Hazards - Rev. 2, Justification will : e provided in the site- 3.5.1.6 July 1981 specific SAR. Approved Design Material
- kstrocksetion (11/96) Page 1.8-36 l
System 80+ Design ContwlDocument p b Table 1.8-5 Standard Review Plan Compliance Comments (Cont'd.) i j Comment or ! SRP Sectionfritle Summary Description of Deviation Section j 6.4 Control Room Habitability Site-specific requirements are ensured 6.4 System - Rev. 2, July 1981 through interface requirements. Toxic gas releases shall be addressed in the site-specific SAR 9.2.1 Station Service Water Systern - SSWS pump structure is addressed with 9.2 Rev. 4, June 1985 site-specific interface requirements. 9.2.2 Reactor Auxiliary Cooling Component Cooling Water Hx structure 9.2.2 Water Systems - Rev. 3, June is addressed with site-specific ingace 1986 requirements. 9.2.4 Potable and Sanitary Water This system is not wholly within the 9.2.4 Systems - Rev. 2, July 1981 scope of the System 80+ design. Interface requirements are provided. 9.2.5 Ultimate Heat Sink - Rev. 2, The UHS is not within the System 80+ 9.2.5 July 1981 design. Interface requirements are provided. Site-specific SARs will demonstrate 9.2.5.2 v compliance with specific requirements of Regulatory Guide 1.27. 10.4.5 Condenser Circulating Water System is site-specific and is addressed 10.4.5 System with interface requirements. 12.1 Assuring that Occupational Operational radiation protection programs 12.1 Radiation Exposures are As will be provided in the site-specific SAR. Low As is Reasonably Achievable - Rev. 2, July 1981 17.5 Operational Radiation This information will be provided by the Protection Program - Rev. 2, owner / operator. July 1981 13.3 Emergency Planning - Rev. 2, This information will be provided in the 13.3 July 1981 site-specific SAR. 13.4 Operational Review - Rev. 2, This information will be provided by the 13.4 July 1981 owner / operator. 13.6 Physical Security - Rev. 2, This information will be provided by the 13.6 July 1981 owner / operator. i 14.2 Initial Plant Test Program - Certain required information will be 14.2 Final Safety Analysis Report - provided by the owner / operator. Rev. 2, July 1981 v , Approved Design nieseriel kooducalon Page 1.847
1 I l l System 80+ Design ControlDocument l Table 1.8-6 System 80+ Industrial Codes and Standards I Code Edition Title ANSI /American Concrete Ins *itute [ACI] 318 1989 Building Code Requirements for Reinforced Concrete,1991 Printing [(349 1985 Code Requirements for Nuclear Safety-Related Concrete Structures}}l'I ANSI /American Institute of Steel Construction [AISC] flN690 1984 Specificationfor the Design, Fabrication, and Erection of Steel Safety-Related Structures for Nuclear Facilities}}Ul 1989 Manual of Steel Construction, Allowable Stress Design, Ninth Edition ANSI /American Nuclear Society [ANS) 2.8 1992 Determining Design Basis Flooding of Power Reactor Sites 13.1 1993 Guide to Sampling Airborne Radioactive Materials in Nuclear Facilities
$1.1 1983 Nuclear Safety Criteria for the Design of Stationary PWR Plants 55.4 1993 Gaseous Radioactive Waste Processing Systems for Light Water Reactor Plants 56.2 1989 Containment Isclation Provisions for Fluid Systems after a LOCA 57.2 1976 Design Objectives for LWR Spent Fuel Storage Facilities at Nuclear Power Stations 58.1 1982 Plant Design Against Missiles 58.2 1988 Design Basis for Protection of LWRs against Effects of Pipe Rupture 58.8 1984 Time Response Design Criteria for Safety-Related Operator Action 58.9 1987 Single Failure Criteria for LWR Safety Related Fluid Systems ANSI /American Petroleum Institute [ API]
650 l 1988 Welded Steel Tanks for Oil Storage ANSI /American Society of Civil Engineers 7 l 1990 l Minimum Design 1.oads for Building and Other Structures [ ANSI M 8.1] ANSI /American Society of Mechanical Engineers [ASME] BPVC 1989 Section 11; Materials Specifications
\lBPVC 1989 Section 111; Rules for Construction of Nuclear Power Plant Compor:cnts; l Division }}}l'l BPVC 1989 Section V, Non-Destmetive Examination BPVC 1989 Section VIII; Rules for Construction of Pressure Vessels BPVC 1989 Section IX; Qualification Standard for Welding and Brazing i BPVC 1989 Section XI; Rules for Inservice Inspection of Nuclear Power Plant Components; Editions and Addenda As Applicable AG-1 1991 Code on Nuclear Air and Gas Treatment B31.1 1992 Power Piping OMS /O 1990 Standards and Guides for Operation and Maintenance of Nuclear Power Plants; l
through 1992 Addenda. i HI NRC Staff approval is required prior to implementing a change in this information; see DCD Introduction O t Section 3.5. l ApprovedDesign Atatorial-Introduction (2/95) Page 1.8-38
Sy~ tem 80 + Des /gn ControlDocument n Table 1.8-6 System 80+ Industrial Codes and Standards (Cont'd.) (v) Code Edition Title ANSI /American Society of Meclanical Engineers [ASME] (Continued) NQA-1 1989 Quality Assurance Program Requirements for Nuclear Facilities, and NQA-lb-1991 Addenda Quality Assurance Requirements for Nuclear Power Plants, and NQA-2a-1990 j NQA-2 1989 Addenda l ANSI / Institute of Electrical and Electronics Engineers 7-4.3.2 1982 Application Criteria for Programmable Digital Computer Systems in Safety Systems of Nuclear Power Generating Stations [ ANSI /IEEE/ANS] 279 1971 Criteria for Protection Systerns for Nuclear Power Generatmg Stations 308 1980 Criteria for Class IE Power Systems for NPGS 317 1983 Electrical Penetration Assemblies in Containment Structures for NPGS 323 1974 Qualifying Class IE Equipment for NPGS 344 1987 Recommended Practices for Seismic Qualification of Class IE Equipment for Nuclear Power Generating Stations 379 1977 Application of the Single Failure Criterion to NPGS Safety Systems 382 1980 Qualification of Actuators for Power Operated Valve assemblies with Safety-Related Functions of Nuclear Power Plants 384 1981 Criteria for Independence of Class IE Equipment and Circuits v 387 1984 Criteria for Diesel-Generator Units Applied as Standby Power Supplies for NPGSs 420 1982 Design and Qualification of Class IE Control Boards, Panels, and Racks used in NPGSs 422 1986 Guide for Design and Installation of Cable Systems in Power Generating Stations 497 1981 Criteria for Post-Accident Monitoring Instrumentation for NPGSs 603 1980 Criteria for Safety Systems for NPGSs 741 1990 Criteria for protection of Class IE Power Systems and Equipment in NPGSs ANSI / Instrument Society of America [ISA) S67.04 1988 Serpoints for Nuclear Safety-Related Instrumentation ANSI / National Fire Protection Association [hTPA] 78 1989 Lighting Protection Code 101 1991 Safety to Life from Fire in Buildings and Structures 803 1988 Standard for Fire Protection for Light Water Nuclear Power Plants Electric Power Research Institute /NSAC 08 1986 l The Reliability of Emergency Diesel Generators at U.S. Nuclear Power Plants \ i
&J Approved Design Material-Irnroduction Page 1.8-39
Syotem 80+ Design C7ntrolDocument Table 1.8-7 ASME Section III Code Cases Applicable to System 80+ h Case Title N-4-l l [1337-11] Special Type 403 Modified Forgings or Bars, Class 1 and CS; 7/13/87. N-604 Material for Core Suppon Structures: 7/27/88. N-71-15 Additional Materials for Subsection NF, Classes 1,2,3 and MC Component Suppons Fabricated by Welding; 12/16/89. N-122-1 Evaluation of the Design of Rectangular Cross-Section Attachments on Class-1 Piping; 7/27/92. N-192-2 Use of Braided Flexible Connectors, Class 2 and 3; 9/17/87. N-247 Cenified Design Repon Summary for Component Standard Suppon, Class 1,2,3, ami MC; 1/21/88. N-249-10 Additional Materials for Subsection NF, Classes 1,2,3 and MC Comh! Suppons Fabricated without Welding; 5/06/89. N-262 Resistance Spot Welding for Stmetural Use in Component Suppons: 7/2S/88. N-284 Metal Contamment Shell Buckling Design Methods; Section III, Division 1, Class MC; 8/25/80. N-309-1 Identification of Material for Component Suppons; 7/28/88. N-313 Alternate Rules for half-Coupling Branch Connections, Class 2; 11/28/36. N-318-4 Evaluation of the Design of Rectangular Cross Section Attachments on Class 2 or 3 Piping; 12/11/89. N-319-1 Evaluation of Stresses in Butt Welded Elbows for Class 1 Piping: 7/24/89. N-391-1 Evaluation of the Design of Hollow Circular Cross Section Welded Attachments on Class 1 Piping; 7/24/89. N-392-1 Evaluation of the Design of Hollow Circular Cross Section Welded Attachments on Class 2 and 3 Pipings; 12/11/89. N 393 Repair Welding Structural Steel Rolled Shapes and Plates for Components Suppons; 7/30/89. N-411-1 Alternative Damping Values for Response Spectra Analysis for Class 1,2, and 3 Piping; 2/20/89. N-420 Linear Energy Absorbing Suppons for Subsection NF, Class 1, 2, and 3 Construction 2/14/88 N-430 Alternative Requirements for Welding Workmanship and Visual Acceptance Criteria for Class 1,2,3, and MC Linear-type and Standard Supports; 2/28/89. N 433 Non-threaded Fasteners for Class 1,2, and 3 Components Piping Suppons; 12/16/89. N-474-1 Design Stress Intensities and Yield Strength Values for UNS NO6690 with a minimum specific Yield Strength of 35 ksi, Class 1 Components;-3/05/90. N-476 Class 1,2, 3, and MC Linear Component Supports - Design Criteria for Single Angle Members, Subsection NF; 5/06/89. N-498 Alternative Rules for 10-year Hydrostatic Pressure Testing for Class 1 and 2 Systems, Section XI, Division 1. Approved Design Materin! Introduction (11/96) Page 1.8M
1 System 80+ oesign controlDocument l (~'\ i C/ ~ Table 1.8-8 Cross-Reference for Unresolved and Generic Safety Issues l l No. Title SectionUI l 3 Setpoint Drift in Instrumentation 7.1.2.27; 20.2.1 14 PWR Pipe Cracks 3.6; 6.6; 10.1; 10.4.7; I 20.2.2 1 15 Radiation Effects on Reactor Vessel Supports 5.4.14.2; 20.2.3 22 Inadvertent Boron Dilution Events 7.7.1.1.10; 15.4.6; 20.2.4 l 23 Reactor Coolant Pump Seal Failures 5.4.1.3; 8.1.4.2; ' 9.3.4; 20.2.5 24 Automatic ECCS Switchover to Recirculation 6.8; 20.2.6 29 Bolting Degradation or Failures in Nuclear Plants 3.9.3; 4.5.2.1; 5.2; 20.2.7 36 Loss of Service Water 9.2.1; 20.2.8 ) 43 Reliability of Air Systems 9.3.1; 20.2.9 45 inoperability of Instruments Due to Extreme Cold Weather 9.4; Ch 16;
/G 20.2.10 48 LCO for Class IE Vital Instrument Buses in Operating Reactors 8.3.2; Ch 16; 20.2.11 49 Interlocks and LCOs for Redundant Class IE Tie Breaker 8.3.1.2; Ch 16; 20.2.12 51 Proposed Requirements for Improving Reliability of Open Cycle Service 9.2.1; 9.2.2; 9.2.5; Water Systems 20.2.13 57 Effects of Fire Protection System Actuation on Safety Related Equipment 9.5.1; 20.2.14 64 Identification of Protection System Instrument Sensing Lines 7.1.2.31; 20.2.15 66 Steam Generator Requirements 5.4.2; 10.3.5; 10.4.1.2; 20.2.16 67.3.3 Steam Generator Staff Actions; Improved Accident Monitoring 7.5; 20.2.17
' 70 PORV and Block Valve Reliability 6.7; 20.2.18 75 Genetic Implications of ATWS Events at Salem 7.7.1.1.11; 20.2.19 78 ' Monitoring of Fatigue Transient Limits for Reactor Coolant System 3.9.1; 20.2.20 79 Unanalyzed Reactor Vessel Thermal Stress during Natural Convection 3.9; 5.3; 20.2.21 Cooldown i 82 Beyond Design Bases Accidents in Spent Fuel Pools 9.1.2; 9.1.3; 9.1.4; 20.2.22
% .]
AwrowdDwign AdaterW-hr@odrcDon Page 1.8-41 l
System 80+ Design ControlDocument Table 1.8-8 Cross-Reference for Unresolved and Generic Safety Issues (Cont'd.) No. Title SectionUI 83 Control Room Habitability 6.4; 9.4.1; 20.2.23 87 Failure of HPCI Steam Line Without Isolation (in BWRs) 3.9.6; 20.2.24 93 Steam Binding of Auxiliary Feedwater Pumps 10.4.9.3.2: 20.2.25 94 Additional LTOP for Light Water Reactors 5.2.2.10; 5.2.3; 5.3; 20.2.26 99 Loss of RHR Capability in PWRs 5.4.7; App 19.8A; 20.2.27 103 Design for Probable Maximum Precipitation 2.0; 3.1.2; l 20.2.28 105 Interfa5c g~ Systems LOCA at LWRs App SE; 20.2.29 106 Piping and Use of Highly Combustible Gases in Vital Areas; Fire 9.5.10; 20.2.30 Protection ,_ 113 Dynamic Qualification and Testing of Large Bore Hydraulic 3.9.3.4; 20.2.31 Snubbers 118 Tendon Anchorage Failure 3.8; 20.2.32 119.1 Pipe Rupture Requirements 3.6.2.1; 3.9.2.5; 3.9.3.1: 20.2.33 119.2 Pipe Damping Values 3.7.1.3; 20.2.34 119.3 Decoupling OBE from SSE 2.5; 3.7; 20.2.35 119.5 leak Detection Requirements 3.6.3.3; 5.2.5; 7.7.1.6; 20.2.36 120 On-Line Testability of Protection Systems Ch 16; 20.2.37 121 Hydrogen Control for Large, Dry PWR Containments 3.8; 6.2.5; 19.11; 20.2.38 122.2 Initiating Feed and Bleed 7.5.1.1.5; 10.4.9; 20.2.39 124 Auxiliary Feedwater System Reliability 10.4.9; 20.2.40 125.I.3 SPDS Availability 7.5; 7.7.1; 18.7.1; 20.2.41 125.11.7 Reevaluate Provision to Automatically isolate Feedwater from Steam 10.4.9; 20.2.42 Generator During Line Break 128 Electrical Power Reliability 8.3; 20.2.43 130 Essential Service Water Pump Failure at Multiplant Sites 1.2.1.3: 9.2.1; 20.2.44 135 Integrated Steam Generator Issues 5.4.2; 6.7.2; 7.3.1; 7.5.1; 10.3.2; 10.4; 15.6.3; 20.2.45 Apprend Desigrr Material Introduction (2/95) Page 1.8-42
System 80 + Design ControlDocument Table 1.8-8 Cross-Reference for Unresolved and Generic Safety Issues (Cont'd.) No. Title Sectientu 142 leakage Through Electrical isolators in Instrumentation Circuits 20.2.46 143 Availability of Chilled Water Systems 9.2.9; 9.4; 20.2.47 153 less of Essential Service Water in LWRs 9.2.1; Ch 16; 20.2.48 155.1 More Realistic Source Term Assumptions 3.11; 6.5; App 15A; 20.2.49 A-1 Water Hammer 5.4.7; 6.3.1; 6.5.1; 7.7.1.1.4; 9.2; 10.3.2.2; 10.4.7; 10.4.9; 20.2.50 A-2 Asymmetric Blowdown leads on RCS 3.6.2; 3.6.3: 20.2.51 A-4 C E Steam Generator Tube Integrity 5.4.2; 10.3.5; 10.3.6; 10.4.1; 10.4.6; 10.4.8; 20.2.52 A-9 Anticipated Transients without Scram (ATWS) 7.1.1; 7.3; 7.7.1.1.11; 19.0; 20.2.53 ( 5.2; 5.4.14; ( "/ A-12 Fracture Toughness of Steam Generator & RCP Supports 20.2.54 A-13 Snubber Operability Assurance 3.9.3.4; 20.2.55 A-17 Systems lateraction 3.6; 19.0; 20.2.56 A 24 Qualification of Class IE Safety Related Equipment 3.9.2.2; 3.10; 3.11; 20.2.57 A-25 Non-Safety leads on Class IE Power Source; 7.1.1; 8.1.3; 8.1.4; 1 8.3.1.2.7; 20.2.58 A-26 Reactor Vessel Pressure Transient Protection 5.2.2; App SA; 7.2; 20.2.59 A-29 Plant Design for Reduction of Vulnerability to Sabotage App 13A: 20.2.60 A-30 Adequacy of Safety Related DC Power Supplies 8.3.2; 20.2.61 A-31 RHR Shutdown Requirements 5.4.7; 6.7; 8.1.2; 8.2; 10.I; 10.4.9; 20.2.62 A-35 Adequacy of Offsite Power Systems 8.1.3; 8.2; 8.3.1; ' 20.2.63 A-36 Control of Heavy leads Near Spent Fuel 9.1.4; App 19.8A; 20.2.64
. A-40 Seismic Design: Short Term Program 2.5; 3.7; 20.2.65 Q A-43 Containment Emergency Sump Performance 6.8; App 19.8A; 20.2.66 Approwd Design Meterial. Mroductiort Page 1.8D
i I System 80+ Design ControlDocument Table 1.8-8 Cross-Reference for Unresolved and Generic Safety Issues (Cont'd.) No. Title SectionUI A-44 Station Blackout 8.1.4.2;8.3.1.1.5; 19.0; 20.2.67 l A-45 Shutdown Decay Heat Removal Requirements 5.4.7; App 19.8A; 20.2.68 A-47 Safety implications of Control Systems 6.3.2; 7.3.1.1.10.3; 7.3.2; 10.4.7; 10.4.9; 20.2.69 A-49 Pressurized Thermal Shock 5.2.2.11; 5.3; 20.2.70 B-5 Ductility of Two-Way Slabs & Shells: Steel Containments 3.6; 3.8; 20.2.71 B-17 Criteria For Safety-Related Operator Actions Ch 7; 20.2.72 B-36 Develop Design, Testing, and Maintenance Criteria for Atmosphere 9.4; 20.2.73 Clean-up System Air Filtration and Adsorption Units for sngineered Safety Features Systems and for Normal Ventilation Systems B-53 Load Break Switch 8.1;8.2;8.3; 20.2.74 B 56 Diesel Generator Reliability 8.1; 8.3.1.1.4; 20.2.75 B-60 loose Parts Monitoring System 7.7.1.6.3; 20.2.76 B-61 Allowable ECCS Equipment Outage Periods Ch 16; 20.2.77 B 63 Isolation of Low Pressure System Connected to The Reactor Coolant 3.2.2; 3.9.6.2; Pressure Boundary App SE; 20.2.78 B-66 Control Room Infiltration Measurements 6.4; 9.4.1; 20.2.79 C-1 Assurance of Continuous Long-term Capability of Hermetic Seals on 3.11; Ch 6; 1cstntmentation and Electrical Equipment Ch 15; 19.11.4; l 20.2.80 C-2 Study of Containment Depressurization by inadvertent Spray 3.8; 7.3.1; 18.3; r Operation 20.2.81 1 6.3; 20.2.82 C-4 Statistical Methods for ECCS Analysis C-5 Decay Heat Update 6.2; 6.3; 15.0; 20.2.83 C-10 Effective Operation of Containment Sprays in a LOCA 6.5; 15.6.5; Ch 16; 20.2.84 i C-12 g Primary Sy' tem Vibration Assessment 3.9.2; 7.7.16; 20.2.85 l i HF 5.1 x.1 Control Stations 9.5.2; 18.7.1.6.2; 20.2.86 HF 5.2 Review cri: .ria for Human Factors Aspects of Advanced Controls 18.3; 18.4; 18.7; and Instrumentation 20.2.87 I.C.1 (1-4) Short Term Accident Analysis and Procedures Rev. 1.6; 20.2.88 1.C.9 Long-term Program Plan for Upgrading of Procedures 20.2.89 l Approved Desigrs Materiel kutroduction Page 1.8-44
i
)
i Sy-tem 80+ oesign controloccument O V Table 1.8-8 Cross-Reference for Unresolved and Generic Safety Issues (Cont'd.) No. Title SectionlU I.D.1 Control Room Design Reviews 18.0; 20.2.90 I.D.2 Control P.oom Design; Plant Safety Parameter Display Console 7.7.1.4; 7.7.1.7; 13.3.3; 18.7.1; 20.2.91 1.D.3 Control Room Design; Safety System Status Monitoring 7.1.2.21; 20.2.92 1.D.4 Control Room Design Standard 18.3; 18.4; 18.7; 20.2.93 1.D.5 (1) Control Room Design; Improved Instrumentation Research - 18.0; 20.2.94 Operator / Process Communication I.D.5 (2) Control Room Design; Improved Instrumentation Research - 7.1.2.21; 7.2.1.1.5; Plant Status and Post-Accident Monitoring 7.5.1.1.5; 18.7.1.8; 20.2.95 1.D.5 (3) Control Room Design; On-Line Reactor Surveillance System 5.2.5; 7.1.2.20; 7.7.1.6; 20.2.% I.D.5 (4) Control Room Design; Improved Instrumentation Research - 5.2.5.1.2.1; 7.5.1; Process Monitoring Instrumentation 7.5.2.5; 20.2.97 3.2; 17.1; 20.2.98 m I.F.1 Quality Assurance; Expand Quality Assurance List for Equipment
) Important to Safety I.F.2 (6,9) Quality Assurance; Develop More Detailed QA Criteria 17.1; 20.2.99 I.G.2 Scope of Preoperational and Low-Power Testing Program 20.2.100 II.B. I Safety Review Consideration; Reactor Coolant System Vents 6.7.1.2.1; 6.7.2.1.1; 7.5.1; 20.2.101 II.B.2 Safety Review Consideration; Plant Shielding to Provide Post 12.2.3; 12.3.1.2; Accident Access to Vital Areas 12.3.1.3 20.2.102 11.B.3 Safety Review Consideration; Post Accident Sampling System 9.3.2; 20.2.103 II.B.8 Rulemaking Proceedings on Degraded Core Accident; Hydrogen 19.11; 20.2.104 Rule, Severe Accident, Etc.
II.C.4 Reliability Engineering 17.3; 19.15; 20.2.105 II.D.1 Coolant System Valves; Testing Requirements 5.4.13.4.1; 6.7; 7.7.1.1.11; 20.2.1 % II.D.3 Coolant System Valves; Valve Position Indication 5.2.5.1.2.1; 5.4.13; 7.7.1.6; 20.2.107 II.E.1.1 Auxiliary Feedwater System Evaluation 10.4.9; 20.2.108 II.E.1.2 Auxiliary Feedwater System Automatic Initiation and Flow Indication 7.3; 7.7; 10.4.9; 20.2.109
/sT App SD; 8.3.1.1.2; Q! .II.E.3.1 Decay Heat Removal; Reliability of Power Supplies for Natural Circulation 20.2.110 Apiproved Design n0eterial- hveroducelan Page 1.8-45
Sy tem 80+ Design ControlDocument Table 1.8-8 Cross-Reference for Unresolved and Generic Safety Issues (Cont'd.) No. Title Sectional II.E.4.1 Containment Design; Dedicated Penetrations 6.2.4; 6.2.5; 20.2.111 II.E.4.2 Contr.inment Design; Isolation Dependability 3.1; 6.2.4; 9.4.6; 20.2.112 II.E.4.4 (1-5) Containment Design; Purging 6.2.4; 9.4.6; Ch 16; 20.2.I13 II.E.6.1 In-situ Testing of Valves-Test Adequacy Study 3.9.6; 20.2.114 II.F.1 Additional Accident Monitoring Instrumentation 7.5.1.1.5; 20.2.115 II.F.2 Identification and Recovery from Conditions Leading to Inadequate 7.5.1.1.7; 20.2.116 Core Cooling II.F.3 Instrumentation for Monitoring Accident Conditions 7.5.1.1.5; 20.2.117 II.G.1 Power Supplies for Pressurizer Relief Valves, Block Valves, and 6.7; 7.5.2.5; Level Indicators 20.2.118 II.J.3.1 Organization and Staffing to Oversee Design and Construction 20.2.119 l II.K.1 (1,2,4(a-c),7,8,11,1 7-23) Measures to Mitigate Small Break 6.3.3; 20.2.120 LOCAs and Feedwater Accidents; NRC Bulletins l II.K.1 (3,4d,5,6,9,10,13-16,24-28) Measures to Mitigate Small Break 6.3.3; 20.2.120 LOCAs and less of Feedwater Accidents: NRC Bulletins 4
!!.K.3 (2,5,6,8,25,30,31,55) Final Recommendations of B&O Task Force 6.3.3; 9.3.4.3.2; to Mitigate Accidents 19.3.3.1; 20.2.121 Ill.A. I .2 (1-3) Upgrade Licensee Emergency Suppon Facilities 13.3.3.1; 13.3.3.2; 13.3.3.3; 20.2.122-124 III.D.1.1 (2) Review Information on Provisions for Leak Detection 11.5; 12.1; 12.3; 20.2.125 Ill.D.3.3 (1-4) In-Plant Radiation Monitoring 7.5.1.1.5; 11.5 12.3.4; 20.2.126 III.D.3.4 Control Room Habitability 6.4; 20.2.127 UI USIs and GSis not applicable to the System 80+ Standard Plant Design are identified in Chapter 20, along with the corresponding reason.
O l l I Approved Dsslyn Material Jntroduction (2/95) Page 1.8 46 I l
System 80+ - Design ControlDocument r% b Table 1.8-9 Cross-Reference for the TMI Rule (10 CFR 50.34f) No. Title Sectiontu (1)(i) Plant / Site Specific PRA (II.B.8) 19.15; 20.2.104 (1)(iiXA) Simplified AFWs Reliability Analysis (II.E.1.1) 10.4.9; 19.6; 20.2.108 Design Review of AFWs (II.E.1.1) 10.4.9; (1)(ii)(B) 20.2.108 . (1)(iiXC) Evaluation of AFWs Flow Design Bases and Criteria 10.4.9 Evaluation of RCP Seal Damage Following Small Break LOCA with 5.4.1.3; (1)(iii) IDP (II.K.2.16 & II.K.3.25) 8.1.4.2; 9.3.4; 20.2.121 Analysis of Probability of Small Break LOCA Caused by PORV 19.15.2.1.2; (1)(iv) (II.K.3.2) 20.2.121 (1)(v) Evaluation of Effectiveness of High Pressure Coolant injection (BWRs N/A Only) (II.K.3.16) (t)(vi) Reduction of Challenges to Relief Valves (BWRs Only)(II.K.3.16) N/A
- (1)(vii) Feasibility and Risk Assessment of ADS Design Modifications (BWRs N/A Only) (II.K.3.I8)
/"N Effect of Core Cooling Modes Under Accident Conditions (BWRs N/A (t)(viii) (] ' Only) (II.K.3.21) (1)(ix) Study of Additional Space Cooling Needs for RCIC & HPIC (BWRs N/A Only) (II.K.3.24) (t)(x) Study ADS Capability During and Following Accident Conditions N/A (BWRs Only)(II.K.3.28) (1Xxi) Evaluate Alternate Depressurization Methods (BWRs Only) N/A (II.K.3.45) (1)(xii)(A) Compare Costs and Benefits of Alternative Hydrogen Control Systems 6.2.5 l (1)(xii)(B) Verify Compliance with (f)(2Xix) of Selected Hydrogen Control 6.2.5.1.2; l System App 19.llK (1)(xii)(C) Evaluate Design, Function & Layout of Alternative Hydrogen Control 6.2.5.1.2; Systems App 19.11K (2)(1) Simulator Capability (COL Requirement) (I.A.4.2) Ch 20 (2)(ii) Plugram to Improve Procedures (COL Requirement) (I.C.9) 20.2.89 (2Xiii) Control Room Design (I.D.I.) 18.0; 20.2.90 (2Xiv) Safety Parameter Display Console (I.D.2) 7.7.1.4; 7.7.1.7; 13.3.3; I8.7.1; 20.2.91 A (f (2)(v) Indication of Bypassed and Operable Status of Safety Systems (I.D.3) 7.1.2.21; 20.2.92 Anaroved oneten arenariet . knroduction 12/9 5) Page 1.s-47
Sy' tem 80 + Design C?ntrolDocument l Table 1.8-9 Cross-Reference for the TMI Rule (10 CFR 50.34f) (Cont'd.) h 1 No. Title Sectionlu (2)(vi) RCS High Point Venting (II.B.1) 6.7.1.2.1; 6.7.2.1.1; 7.5.1; 20.2.101 (2)(vii) Radiation and Shielding Design Review (II.B.2) 12.2.3; 12.3.1.2: 12.3.1.3; 20.2.102 (2)(viii) Post-Accident Sampling System (II.B.3) 9.3.2; 20.2.103 (2)(ix)(A) Hydrogen Control System for IdI*IClad/ Metal-Water Reaction 6.2.5; 19.11.4; (II.B.8) - Capability to Maintain < 10% App 19.1IK (2)(ix)(B) Hydrogen Control System for 100% Clad / Metal - Water Reaction 6.2.5; (ll.B.8)- Assure No local Pockets That Could Cause Loss of App 19.11K Containment or Mitigating Features (2)(ix)(C) Hydrogen Control System for 100% Clad / Metal- Water Reaction 19.11.4 (ll.B.8) - Equipment Need for Safe Shutdown Qualified for Environment (2)(ix)(D) Hydrogen Control System for 100% Clad / Metal - Water Reaction N/A (ll.B.8) - Inadvertent Actuation of Inerting System (2)(x) Qualification of RCS Relief, Safety and PORV Block Valves (II.D.1) 5.4.13.4.1; 6.7; i 7.7.1.1.11; 20.2.106 (2)(xi) Indication of Relief and Safety Valve Position in Control Room 5.2.5.1.2.1; (II.D.3) 5.4.13; 7.7.1.6; 20.2.107 (2)(xii) Actuation and Indication of AFW Flow in Control Room (II.E.1.2) 7.3; 7.7; 10.4.9; 20.2.109 (2)(xiii) Capability to Maintain Natural Circulation and Pressurizer Heaters App SD; with Onsite Power (ll.E.3.1) 8.3.1.1.2; 20.2.I10 (2)(xiv)(A) Containment isolation System (II.E.4.2) - Automatic Isolation of 6.2.4; 9.4.6; Non-Essential Systems 20.2.112 (2)(xiv)(B) Containment Isolation System (ll.E.4.2) - Two Isolation Barriers in 6.2.4 Series for Each Non-Essential Penetration 20.2.112
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(2)(xiv)(C) Containment Isolation System (ll.E.4.2)- Reset of CIAS Does Not 6.2.4 Result in Valves Opening 20.2.112 (2)(xiv)(D) Containment Isolation System (II.E.4.2) - CIAS Setpoint as IAw as 6.2.4 Compatible with Normal Operation 20.2.112 (2)(xiv)(E) Containment Isolation System (11 E.4.2) - Automatic Isolation of All 6.2.4 Paths to The Environs on High Radiation 20.2.112 (2)(xv) Containment Purging / Venting Capability with Reliable Isolation 6.2.4; 9.4.6; During Accident (II.E.4.4) TS 3.9; 20.2.113 (2)(xvi) Establish Design Criterion for Actuations of ECS and RPS (BWRs N/A Only) (ll.E.5.1) Approved Design Afsterial- krtroduction Page 1.8-48
' Sv-tem 80+ Destan contrat Document Table 1.8-9 Cross-Reference for the TMI Rule (10 CFR 50.34f) (Cont'd.)
Title SectionW . No. , Provide instrumentation to Measure, Record and Indicate in the 7.5.1.1.5; ! . (2)(xvii) f 20.2.115 Control Room (ll.F.1) Control Room Indication of inadequate Core Cooling - Saturation 7.5.1.1.7; ! l (2)(xviii) ' 20.2.116 . Meter (II.F.2) Provide Post-Accident Monitoring Ins:nunentation (ll.F.3) 7.5.1.1.5-(2Xxix) ! 20.2.117 Power Supplies for RCS Relief and Block Valves and level 6.7; 7.5.2.5; b (2Xxx) 20.2.118 Indicators (II.G.1) } N/A (2Xxxi) Auxiliary Heat Removal System Design (BWRs Only) (II.K.I.22) FMEA on Integrated Control System (B&W Only) (II.K.2.9) N/A (2Xxxii) ~ Anticipatory RPS Trip on L.oss of MFW and Turbine Trip (B&W N/A (2Xxxiii) + Only) (II.K.2.10) 4 N/A (2Xxxiv) Recording of Post-Accident Reactor Vessel Water level (BWRs ~ Only) (ll.K.3.23) Onsite Technical Support Center, and Tecludcal Operations Center, 13.3.3.1; . (2Xxxv) 13.3.3.2; ) f and Emergency Operations Facility (III.A.I.2) 13.3.3.3; 20.2.122-124 l , Leakage Control and Detection Design and Program for Systems 11.5; 12.1; (2Xxxvi) 12.3; 20.2.125 Outside Contamment (III.D.I.1) Monitoring of inplant and Airborne Radioactivity (Ill.D.3.3) 7.5.1.1.5; 11.5;
- (2)(xxvii) 12.3.4; 20.2.126 ,
Evaluate Potential Pathways That May lead to Control Room 6.4; 20.2.127 (2)(xxviii) Habitability Problems Under Accident Conditions (III.D.3.4) Administrative Procedures for Evaluating Industry Operating, Design, N/A 1 (3Xi) and Construction Experience Dur'ng Design and Construction (I.C.5)
. i J
Ensure that QA List Contains All Systems, Structures and 3.2 .7.1; (3)(ii) Components important to Safety Per Criterion II of 10 CFR 50 20.2.98 Appendix B (1.F.1) Quality Assurance Program (I.F.2) 17.1; 20.2.99 (3Xiii) Provision of Dedicated Contamment Penetration for Future 20.2.104 l 1 (3)(iv) Installation of Systems to Prevent Containment Failure (II.B.8) i Containment integrity During Hydrogen Burn (or Inerting) for 100% 3.8.2-(3)(v)(A) Clad / Metal - Water Reaction (ll.B.8) App 19.llK; l 20.2.104 ! Containment Structural leading from inadvertent Actuation of N/A (3Xv)(B) Inerting System (ll.B.8) External Hydrogen Recombiners (ll.E.4.1) 6.2.4; 6.2.5; (3Xvi) 20.2.111 Management Plan for Design and Construction Activities (COL 20.2.119 (3)(vil) Requirement) (II.J.3.1) til N/A indicates items which are Not Applicable to the System 80+ Standard Plant Design. wDeep neenerter keenocakm west reser.sts
Syntem 80+ Design controlDocument Table 1.8-10 Cross-Reference for New NRC Policy Issues (SECY-93-087) No. Title Sectiont 'l I.A Use of a Physically Based Source Term 3.11; 6.5; App ISA I.B Anticipated Transients Without Scram 7.3; 7.7.1.1.11 I.C Mid-Loop Operation App 19.8A I.D Station Blackout 8.1.4.2.; 8.3.1.1.5 1.E Fire Protection 9.5.1 I.F Intersystem Loss-of-Coolant Accident App SE I.G Ilydrogen Control 6.2.5; 19.11.3
!.H Core Debris Coolability 19.11.3 1.1 l{igh-Pressure Core Melt Ejection 19.11.3 1.J Containment Performance 19.11.3
!.K Dedicated Containment Vent Penetration 19.15.5 I.L Equipment Survivability 19.11.4.4 ,
I.M Elimination of Operating-Basis Earthquake 2.5; 3.7 I.N Inservice Testing of Pumps and Valves 3.9.6; 5.2.4; 6.6 II.A Industry Codes and Standanis 1.8 II.B Electrical Distribution 8.2; 8.3 II.C Seismic Hazard Curves and Design Parameters 19.7.5 ll.D Leak-Before-Break 3.6.2.1.3; 3.6.3 II.E. Classification of Main Steamlines in Boiling Water Reactors N/A II.F Tornado Design Basis 2.3.2.1 l II.G Containment Bypass 6.2.2; App 5F 11.11 Containment leak Rate Testing 3.8.2.7 11.1 Post-Accident Sampling System 9.3.2 II.J Level of Detail 1.1.1; 1.2.2 II.K Prototyping N/A II.L ITAAC 14.3 II.M Reliability Assurance Progrero 17.3 Approved Design Motorial . introduction (2/95) Page 1.8 50
Sy.~ tem 80 + o sign controlDocument 1 l l s,_./ Table 1.8-10 Cross-Reference for New NRC Policy Issues (SECY-93-087) (Cont'd.) No. Title Sectiont 'l II.N Site-Specific Probabilistic Risk Assessments and Analysis of External 17.3; 19.7; Events 19.15 II.O Severe Accident Mitigation Design Alternatives 19.15.5 l II.P Generic Rulemaking Related to Design Certification N/A Defense Against Common-Mode Failures in DigitalInstrumentation 7.2; 7.3; 7.7; II.Q and Control Systems App 7A II.R Steam Generator Tube Rupture . 19.15.2.1.2 II.S PRA Beyond Design Certification 17.3; 19.7; 19.15 II.T Control Room Annunciator (Alarm) Reliability 7.7 Ill.A Regulatory Treatment of Nonsafety Systems in Passive Designs N/A III.B Definition of Passive Failure - N/A III.C SBWR Stability (Passive Design) N/A III.D Safe Shutdown Requirements (Passive Design) N/A A N/A Ill.E Control Room Habitability (Passive Design) III.F Radionuclide Attenuation (Passive Design) N/A _ III.G Simplification of Offsite Emergency Planning 15.6.5 III.H Role of the Passive Plant Control Room Operator N/A i 1 l
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l l 1
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x, I'l "N/A" indicates items which are Not Applicable to the System 80+ Standard Plant Design. l l
^;; .::: Design Metww-Introduceion (2/96) Page 1.8-51
l l System 80+ Design ControlDocument n d 1.9 System 80+ Standard Design Interfaces This section provides a listing of the interfaces as used in 10 CFR 52.47(a). The System 80+ Standard Design includes an essentially complete nuclear plant, except for structures, systems and components which require site-specific design. These structures, systems and components are not included in the System 80+ design certification and shall be provided by the applicant (owner / operator) during site specific engineering. ((To ensure that the design of these items is compatible with the System 80+ Standard Design, interface requirements must be satisfied by the applicant.))' In general, interface requirements for applicant-supplied structures, systems and components that relate to a specific mechanical or electrical system are covered in the appropriate chapter; (the word "shall" is used to identify interface requirements included in descriptive text). Interface requirements which have sufficient significance to safety are specified in the Certified Design Material. Table 1.9-1 provides an index to all sections of this report containing interface requirements. Site-specific assumptions on which the System 80+ Standard Design is based are presented in Section 1.2.1, Principal Site Characteristics, and Chapter 2.0, Site Envelope Characteristics. ((The applicant (owner / operator) shall verify that the chosen site is enveloped by the characteristics given in Sections 1.2.1 and 2.0.))' These site-specific characteristics must be compatible with the System 80+ design envelopes, but they are not considered interface requirements as used in 10 CFR 52.47(a). - NJ
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1 COL information item; see DCD Introduction Section 3.2.
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4prowd Design Meterial. htmduction Page 1.91
System 80 + Design ControlDocument Table 1.9-1 Index of System, Structure or Component Interface Requirements for System 80+ System, Structure or Component Section Buildings / Structures Administration Building 1.2.1.4.1.1 Personnel Access Portal 1.2.1.4.1.2 Switchyard 8.2 Warehouse 1.2.1.4.1.3 j ^ Emergency Operations Facility 13.3.3.2 Bulk Gas Storage 9.5.10.1.2 Station Service Water Pump Structure 9.2.1.1.4 Ultimate Heat Sink, including SSWS Intake / Discharge 9.2.5.1.3 Potable and Sanitary Water System Structure 9.2.4 ) Systems Condenser Circulating Water System, including Normal lleat Sink, Pump 10.4.5.1; Structure, Intake and Discharge, and Turbine Building Service Water System 9.2.10.2.1 Offsite Power System, including Switchyard 8.1 Potable and Sanitary Water Systems, including Sewage Treatment 9.2.4.1 i Security System 13.6.1 i Service Water Pump Structure Ventilation System 9.4.8.1.2 Layout and Equiptnent for the Laboratory Facilities 13.3.3.4 Layout and Equipment for the Onsite Decontamination Facilities 13.3.3.6 I Filtered Water Source 9.2.3.2 9.2.4 Communications (Off-site) 9.5.2.2.5 Components l Component Cooling Water Heat Exchanger Materials 9.2.2 Condenser Materials Specification 10.3.6.2; 10.4.1.2 Toxic Gas Monitors 9.4.1.1; Fig. 9.4-2 l l l O Agnproved Design Material Introduction Page 1.9 2
l l l System' 80 + Desian controlDocument i Il 1.10 System 80+ COL Information V The System 80t Standard Design represents an essentially complete nuclear plant. However, certain topics are more appropriately addressed by a COUapplicant referencing this DCD. Table 1.10-1 identifies ADM sections where descriptions of COL information items are presented.
- Neither the table listings nor the descriptions within the cited ADM sections are intended to constitute requirements for the COL applicant. This information is provided only for purposes of facilitating a COL applicant's preparation of its COL application.
Table 1.10-1 COL License Information 1 COL No. FSER No. Section Subject l l1 1.8, 6.6.1 Applicable editions of code cases, industry codes and standards 1-2 1.9 Design interfaces 2-1 2.0-1 2.0 Site Parameters 2-2 2.1 1 2.1 Geography and Demography Information 2-3 2.2-1 2.2 Industrial, Transportation and Military Hazards 2-4 2.3 1 2.3, 15.0.4 Mearology informatoi
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n 2-5 2.4 1 2.4 HyoIologic Information 2-6 2.5 1 2.5 Geology, Seismology and Geotechnical Information 3-1 3.4-1 3.4 Flood Analysis 3-2 3.5 1 3.5 Missile Protection 3-3 3.6-1 3.6 Final Piping Design Information 3-4 3.7 1 3.7 Site and Plant-Specific Seismic Design Information 3-5 3.8 1 3.8, Tbl 3.8A 7 Site and Plant-Specific Structural Design Information 3-6 3.9-1 3.9 Site and Plant Specific Information for Mechanical System and Components Design 3-7 3.10-1 3.10 Seismic and Dynamic Qualification Program Details 3-8 3.11-1 3.11 Environmental Qualification Program Details 4-1 4.2.7-1 4.2.3.2.10 Online fuel failure monitoring and post-irradiation surveillance 4-2 4.4.4-1 7.2.1.1.2.5 CPC/CEAC software testing and change control 5-1 5.2.1.1 1 5.2.1.1, 5.2.1.2 ASME Code edition, Addenda and Code cases for cons:metion of the reactor coolant pressure boundary components
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System 80+ oesign control Document Table 1.10-1 COL License Information (Cont'd.) h COL No. FSER No. Section Subject 5-2 5.2.2.2-1, 5.2.2.3-1, 5.2.2.10.2.2, Verification of the material propenies and end-5.3.11,5.3.2-1 5.2.3.1, 5.3.2, of-life fluence and resulting P-T limits and 5.4.14.3 LTOP temperatures 5-3 5.2.4-1, 6.6-1 5.2.4,6.6.1, PSI and ISI program plans for NRC staff review 12.3.1.2 5-4 5.4 1 5.4.2.5 Steam Generator tube inservice inspectiou program 55 SF 1 SF (5.6.3), 6.4.1.2 Leakage monitoring progre.m 6-1 6.1.1 - 1, 6.1.1-2, 6.1.1.1, Engineered Safety Feature Systems materials 6.1.1-3, 6.1.1-4 5.2.3.3.2.1 selection and fabrication 6-2 6.3.7 1 6.3.4.1, 6.3.4.2 Periodic testing of the safety injection system 6-3 6.4-2 6.4.1.2, 6.4.2.2 Protection against the effects of toxic substance 14.2.12.1.103 releases (including TM1 Ill.D.3.4) 6-4 6.4-3 6.4.1.1 Control room habitability system 6-5 6.4.1.2 Pump seal leakage procedure 6-6 6.5 1 6.5.4.1,6.5.5 Contamment spray system operability 6-7 6.2.4 1 Table 6.2.41 Contamment isolation details 6-8 6.8.2.2.1 IRWST screen area margin analysis 71 7.1.2.7 Integrated response time for protection system 7-2 7.4.1.1.8.2 Operating procedures for SCS- l 73 7.3.1.1.10 Procedures for r.: moving ESFAS signals during 1 plant testing 7-4 7.3.2.1 Procedures for ESFAS Reset 7-5 7.3.2.3.2 ESFAS setpoint analyses 76 7.4.2.5.2 Cold diutdown procedures 77 7.5.2.5.10 Administrative controls associated with PAMI 8-1 8.3.1 1 8.1.4.5, 8.3.1.1.6, Electrical power systems sizing, testing, j Tbls 8.3.12,3,4 calibration and maintenance
' Itis 8.3.2-3,4 '
I 9-1 9.1.2.2.2, 9.1.4.2, Administrative controls and procedures 9.1.4.3, 9.1.4.4, associated with fuel storage and handling 9.1.4.6 systems 9-2 9.2.1 1 9.2.1.1.4, 9.2.1.4, Organic fouling and inorganic buildup in the 9.2.5.4, 20.2.13 SSWS (including GSI-51) 9-3 9.2.1 2 9.2.1.2.1.2 Station service water system pump structure 9-4 9.2.4-1 9.2.4.2 Potable and sanitary water systems , l 9-5 9.5.1.5-1 9 5.1.11, 19.8.1.2, Administrative controls for BTP CMED 9.51 19.15.3.2 conformance and fire brigade l 9-6 9.5.1.12 Fire llazards Analysis kunndDu4ps AsetwW-kreekceien (1/9H Pope 1.10 2
Srtem 80+ Design ControlDocument u ; O Q Table 1.10-1 COL License Information (Cont'd.) . l 7 COL No. FSER No. Section Subject 9-7 9.5.2-1, 9.5.2-2, 9.5.2.1, 9.5.2.2.5, Communications systems l 9.5.2-3 9.5.2.2.6 l Security lighting system l 9-8 9.5.3.2.2 l 9-9 9.5.4.1-1 8.3.1.1.4.11, Diesel operator training l 8.3.1.1.4.13 9-10 9.5.4.1-2, 9.5.4.2-1, 9.5.4, 9.5.5, 9.5.6, Diesel generator auxiliary support systems l 9.5.5-1, 9.5.5-2, " J.7, 9.5.8, 9.5.9 9.5.6-1, 9.5.6-2, 9.5.7-1, 9.5.8-1, 9.5.9-1 9-11 9.2.5-1 9.2.5.1.3 Protected area perimeter abutting or crossing a l body of water 9-12 9.5.1.2.1.2-1 9.5.1.2 Procedures and training for using transfer l switches 10-1 10.2 1 10.2.1 Turbine valve closing time 10-2 10.3-1 10.3.2.2 Steam hammer prevention 10-3 10.4.4-1 10.4.4.2.4.1 Pressure drops between the steam generator l Q 10 10.4.7-2, 10.4.9-2 10.4.7.2.5, nozzles and each system valve Avoidance of water hammer in the condensate, 10.4.9.1.2 feedwater, and emergency feedwater systems 10-5 10.4.9-3 10.4.9.3, Steam binding in the emergency feedwater pump i 10.4.9.5.2 11 1 11.1-1, 11.5-1 11.1, 11.5.1.1 Conformance with Appendix B to 10 CFR 20 Appendix I to 10 CFR 50, ANSI N13.1, R.G. l 1.21 and R.G. 4.15 j 11-2 11.4 1 11.4.1.1, Site-specific solid waste management system 11.4.2.3.1 operating procedures 11-3 11.5-2 11.5.1.4 Procedures in accordance with Position C of R.G. 4.15 11-4 11.2.1-1 11.2.2.3, 11.2.5, Setpoints for radiation monitors; Offsite dose l 11.2.6.1, 11.3.1.1, calculations. j 11.5.1.2.3.I f 11 5 11.5.1 1 11.5.1.1, 11.5.2.2, Operation and maintenance manual for 11.5.2.4, 11.5.2.6 monitoring and sampling liquid and gaseous process and effluent streams 12-1 12.1.1-1 12.1.1.2, 12.1.3 Operatiotial ALARA policy 12-2 12.1.2-1 Tb1 11.5-4, Shielding analysis (including TMI II.B.2); I?. l .2.1,12.2.2.1, Environmental qualification criteria and 12.2.3, 12.3.2.2 maintenance procedures 12-3 12.1.3-1 No specific New Regulatory Guides ADM Section ) l l L ;.; ' Design hienenet heroducaion (2/95) Pope 1.10-3
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System 80+ Design ControlDocument l i Table 1.10-1 COL License Information (Cont'd.) h l
' Subject COL No. FSER No. Section i 12-4 12.3.1-1 12.3.1.8 List of vital areas 12-5 12.3.4-1 12.3.4 Area radiation monitor location 12-6 12.3.4-2 12.3.4 Portable airborne iodine monitor (TMI Ill.D.3.3) 12-7 12.4.5-1 12.4.5 Dose assessment 12-8 12.5-1 12.5 Health physics program 13-1 13.1-1 13.1 Organizational structure of the site operator (including TMI 1. A.I.4, II.J.3.1) 13-2 13.3-1 13.3.1, 13.3.2 Site-specific emergency planning 13-3 13.3.3.1.7 TSC communication equipment 13 4 13.3-2 13.3.3.2, 13.3.3.4, TMI III.A.I.2, " Upgrade Licensee Emergency 13.3.3.6 Support Facilities" 13-5 13.4-1 13.4 Site operator review and audit l
13-6 13.5-1 13.5 Plant procedures (TMI 1.C.1, l.C.5, & l.C.9) 13-7 1.2.13, 13.6 Site security and sabotage protection 13-8 13.2-1 13.2, 13.3.3.1.4 Training [TMI I.A.4.2, l.A.4.l(2), II.K.l(26)] 13-9 13.5-2 13.5.2 Administrative Control Procedures TMI II.J.4.1, ll.K. l(10) 14-1 14.2.2.1 Organiration and Staffing 14-2 14.2.3-1, 14.2.3-2, 14.2.1.1, 14.2.3, Testing program procedures and schedules 14.2.3-3, 14.2.3 4, 14.2.4, 14.2.10.1, including scoping documents, startup 14.2.4-1, 14.2.10-1, 14.2.11 administrative manual, test conditions, test 14.2.11-1 methodologies, data collection and reduction, reconciliation methods, and initial fuel load and criticality procedures 14-3 14.2.6-1 14.2.6 Retention of Test Records 14-4 14.2.9-1 14.2.9 Trial use of plant operating and emergency procedures 14-5 14.2.12.2-1 14.2.7.1.3 Testing of personnel monitors and radiation survey instruments 14-6 14.2.13-1 14.2.7.5 Security system detailed test description and acceptance criteria 15-1 15.7.3.4 Liquid tank failure minimum dilution flow 15-2 15. A.3.2-1 6.1.2.2 Quantity of electrical cable insulation 17 1 17.1 1, 17.2-1 11.2.1.2, 11.3.1.2, Construction and Operation QA (including TMI 11.4.1.2, 17.1, I.F.2, ll.J.3.1) 17.2 17-2 17.3.1-1, 17.3.5-1 17.3.1, 17.3.5, D-RAP completion 17.3.7. 17.3.13 Approwd Des %rr Meterint kutroductxwr (11/96) Page 1.104
Syntem 80+ D~& contrat occumart (
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Table 1.10-1 COL License Inforrnation (Cont'd.) COL No. FSER No. Section Subject 17-3 17.3.9-1 17.3.1, 17.3.7, Operations reliability assurance process 17-4 17.3.9, 17.3.10, implementation l l 17.3.13, ' Tbl 19.15-1, 20.2.105 18-1 18.9.3.2 Validation of operating ensemble 18-2 18.6.1.3.4-1 13.2 Operator training on " Plant Safety Parameter i Display Console" 19-1 19-1 19.15.3.1 Vulnerability of the SSWS intake structure to > l tornado-generated debris 19-2 19.7.5.3 Elements of the plant affecting the performance l of systems in seisthic events 19-3 19-5, 19-6, 19-7 19.15 Details of the layout of the critical components l for fire and flood, interaction of internal flood sources, and effects of fire suppression systems on other systems 19-4 19-8 19.7.5.3 Development of detailed seismic walkdown 19-9 Talt 19.15-1 procedures to verify as-built SSC HCLPFs and seismic vulnerability of SSCs 19-5 19-10, 19 11 6.5.5, Calculation of specific flow rate and l Tb1 19.15-1 consideration of shielding requirements for local operator actions for the emergency containment spray backup system 19-6 19.1.4-1, 19 12 19.7.5.3, 19.15, Update of PRA to include final design detail and l Tb1 19.15-1 site-specific information including examination i of all external event hazards and analysis using l l site-specific spectra 19-7 19-14 19.15.6 List of risk significant SSCs for D-RAP and 19-9 Table 19.15-1 operations reliability assurance process 19-8 19-15 19.15.6, Consideration of risk important operator actions 19-16 Tbl 19.15-1 in developing procedures, training and human 19-19 reliability related programs, and systems to address in severe accident management and aligning the alternate AC source (AAC) procedures 19-9 19.3.7-1, 19.3.9-1, 19.8 1.2, Establishment of administrative controls, outage 19.3.9-2, 19.8A (2.1.3), management, procedures and training for control 19.3.9.2-1, 19-17 Tbl 19.15-1 of fire and flood barriers, containment closure 19-18 capability, switchyard availability, and alternate ! equipment during shutdown operations i 19-10 19.7.5.1 Use of seismically rugged electrical equipment l g Tbl 19.15.1 L.) l I i AnweM Denkrs Menenief krtroclucalen 11/9 71 Page 1.10-5 l l 1
I i System 80+ Deslan control Document - t I n Effective Page Listing , h Chapter 2 Pages Date i,ii 1/97 l , . i ll 11/% , iv Original-v 1/97 2.0-1 Original 2.0-2, 2.0-3 1/97 2.1-1 Original 2.2-1 2/95 4 2.3-1 through 2.3-6 Original ; 2.4-1 Original ( 2.5-1 Original 2,5-2 1/97 2.5 3 through 2.5-5 Original i 2.5-6,2.5 1/97 2.5-8 through 2.5-50 Original 2.5-51 1/97 2.5 52, 2.5-53 Original J.
-Neuwd Dee4pn nennennt she aweetenuarcs 11/9 7) . Pege I. I
l i i I system 80+ oeska controt oocument i 1 t'm) Chapter 2 Contents (./ Tigt l 2.0 Site Envelope Characteristics . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.0-1 2.1 Geography and Demography . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.1-1 2.1.1 Site Location and Description ... ....... ......... .. . . ... .. . .. ... 2.1-1 ; 2.1.2 Exclusion Area Authority and Control . . , . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.1-1 2.1.3 Population Distribution . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.1-1 2.2 Nearby Industrial, Transportation and Military Facilities . . . . . . . . . . . . . . . . 2.2-1 2.2.1 Aircraft Hazards . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.2- 1 2.2.2 Transportation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.2- 1 2.2.3 Other Industrial Hazards On and Offsite . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.2-1 2.3 Meteorology . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2. 3- 1 2.3.1 Regional Climatology . ... ..... .... ....... .. .... .. .... .. ... ... 2.3-1 2.3.2 Local Meteorology . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.3-2 2.3.3 Onsite Meteorological Measurenv.nts Programs ..... . .... .. ... .. ..... .. 2.3-2 2.3.4 Short-term (Accident) Diffusion Estimates (x/Q) . . . . .. ..... .... ...... .. 2.3-3 2.3.5 Long-term (Routine) Diffusion Estimates (x/Q) . . . . . . . . . . . . ... .. ..... .. 2.3-3 2.3.6 Onsite (Accident) Diffusion Estimates (x/Q) . . .... ... .... . . .... .... . .. 2.3-3 m 1 4 2.4 Ilydrologic Fr.p---:- sing ....................... . .. ..... .. .... 2.4-1 2.4.1 External Floods . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.4-1 2.4.2 Internal Floods . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.4- 1 2.5 Geology, Seismology, and Geotechnical Fnym;ing ..... . . . ... ..... ... 2.5-1 2.5.1 Basic Geologic And Seismic Information . .. . ..... ...... . ... . ,.... . .. 2.5-3 2.5.2 Vibratory Ground Motion . . . . . . . . . . . . . . . ... .. ... .. .. ...... . .. . 2.5-4 2.5.3 Surface Faulting . . . . . . . . . . . . . . . . . . . . . . .................... 2.5-10 2.5.4 Stability of Subsurface Materials and Foundations . . . . . . . . . . . . . . . . . . . . . . 2.5-10 2.5.5 Stability of Slopes . . . . . . . . . . . . . . . . . . . . . . . . . . .......... ..... 2 5-12 Appendix 2A Characteristics of Generic Soil Sites . . . .......................2A-1 Appendix 2B Chameteristics of Selected Control Motions . . . . . . . . . . . . . . . . . . . . . . . 2B-1 Appendix 2C Strain-Compatible Modulus and Damping Values ............ ...... 2C-1
" :Deeigns neenerief. Slee Characteristics (11/96) &
Sy3 tem 80+ Design CortrolDocument Chapter 2 Tables g Page 2.0-1 Envelope of Plant Site Design Parameters . ............... ... . .... . 2.0-1 2.3-1 Radiological Dilution Factors (x/Q) ... .......... . ..... ..... . 2.3-4 2.3-2 Onsite Accident x/Q Values at the Control Room North Air Intake . . ...... .. 2.3-5 2.3-3 Onsite Accident x/Q Values at the Control Room South Air Intake . .... . ..... . 2.3-5 2.3-4 Onsite Accident x/Q Values at the Center of the Control Room ..... .... ... 2.3-6 2.3-5 Onsite Accident x/Q Values at the Doorway to the Control Room . . . . . . . ..... 2.3-6 2.3-6 Atmospheric Dilution Factor (x/Q) Reduction Factors .. .. .. . .. .. ... . . ... . 2.3-7 2.3-7 Unit Vent Release Point Characteristics ....................... ... . 2.3-7 2.5-1 Digitized Spectral Ordinates for the Envelope of the Free-Field Surface Spectra for the Horizontal Motion,5% Damping . . . ............ . ...... . ... 2.5-13 2.5-2 Digitized Spectral Ordinates for the Envelope of the Free-Field Surface Spectra for the Vertical Motion,5% Damping . . . . . . ..... . . . . . . . . . . . . . . . . 2.5- 14 Chapter 2 Figures Page 2.5-1 Sample Soil Sites . .. .......... ......... ............. . 2.5-15 2.5-2 Range of Shear Wave Velocities for all Cases Considered . . . . . . . . . . . . . 2.5-16 2.5-3 Variation of Shear Modulus with Strain for System 80+ ....... .... . . . 2.5-17 2.5-4 Variation of Damping Ratio with Strain for System 80+ .............. . 2.5-18 2.5-5 Selected Control Motions for System 80+ Seismic Design CMSI, CMS 2, CMS 3 . . 2.5-19 2.5-6 Generation of Control Motion CMS 2 . . . . .. ........ . . . . . . . . . 2.5-20 2.5-7 Selected Smooth Spectrum and Spectrum for Synthetic Time History H1; CMS 2 Control Motion . ........ ... .............. . . . . . . . . 2.5-21 2.5-8 Selected Smooth Spectrum and Spectrum for Synthetic Time History H2; CMS 2 Control Motion . . . . . . . . . . . ........... ............. 2.5-22 2.5-9 Selected Smooth Spectrum and Spectrum for Vertical Synthetic Time History V; CMS 2 Control Motion . . . . . . . . ... ...... ... ............. . 2.5-23 2.5-10 Spectra at Ground Surface for Cases A-1, B-1, B-2 and B-4 Using Synthetic Time Hi:::ory CMS 2 H1. . ... .. . ................. ... ....... 2.5-24 2.5-11 Spectra at Ground Surface for Cases C-1, C-2, C-3, and D-1 Using Synthetic Time History CMS 2 Hl . . . . . . . . ... .... ....... ......... ... .. 2.5-25 2.5-12 Spectra at Ground Surface for Cases B-1.5, B-3.5, and C-1.5 Using Synthetic Time History CMS 2 H l . . . . . . . . . . . . . . . . . . .. ....... . .... ..... 2.5-26 2.5-13 Spectra at Foundation Level for Cases A-1, B-1, and B-4 Using Synthetic Time l History CMS 2 H1. . .......... ... ..... ................ 2.5-26 j 2.5-14 Spectra at Foundation Level for Cases C-1, C-2, C-3 and D-1 Using Synthetic Time j History CMS 2 Hl . . ............... .. .. . ....... ..... 2.5-27 l 2.5-15 Spectra at Foundation Level for Cases B-1.5, B-3.5, and C-1.5 Using Synthetic Time l History CMS 2 H l . . . . . . . . . . . . . . . . . . . . .. ... ........ ...... 2.5-28 j 2.5-16 Spectra at Ground Surface for Cases A-1, B-1, B-2, B-3, and B-4 Using Synthetic 1 Time History CMS 2 H2 . .. ....... ... ... . ........ .. 2.5-29 j l i AMvoved Design Material Site Characteristics Pageiv i
i System 80+ oestan contrar ocewnent Chapter 2 Figures (Cont'd.). ;
; i Page 4 L 2.5-17 Spectra at Ground Surface for Cases C-1, C-2, C-3, and D-1 Using Synthetic Time ; History CM S2 H2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.5-30 l
- 2.5-18 Spectra at Ground Surface for Cases B-1.5, B-3.5 and C-1.5 Using Synthetic Time i History CMS 2 H2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.5-31 2.5-19 Spectra at Foundation Level for Cases A-1, B 1, and B-4 Using Synthetic Time ;
i History CMS 2 H2 . . . . . . . . . . ...............................2.5-32 i 2.5-20 Spectra at Foundation Level for Cases C-1, C-2, C 3. and D-1 Using Synthetic Time History CMS 2 H2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.5-33 ; 2.5-21 Spectra at Foundation Level for Cases B-1.5, B-3.5 and C-1.5 Using Synthetic j Time History CMS 2 H2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.5-34 2.5 22 Spectra at Ground Surface for Cases A-1, B-2, B-3, and B-4 Using Venical j CMS 2 Synthetic Time History . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.5-35 ! 2.5-23 Spectra at Ground Surface for Cases C-1, C-2, C-3, and D-1 Using Venical i CMS 2 Synthetic Time History . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.5-36 j
- 2.5-24 Spectra at Ground Surface for Cases B-1.5, B-3.5, and C-1.5 Using Venical CMS 2 Synthetic Time History . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.5-37 l
2.5-25 Spectra at Foundation Level for Cases A-1, B-1, and B-4 Using Vertical CMS 2 Synthetic Time History . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.5-38 : } ' 2.5-26 Spectra at Foundation Level for Cases C-1, C-2, C-3 and D-1 Using Venical ' CMS 2 Synthetic Time History . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.5-39 2.5-27 Spectra at Foundation Level for Case,. B-1.5, B-3.5, and C-1.5 Using Vertical , ! CMS 2 Synthetic Time History . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.5-40 i 2.5-28 Selected Smooth Spectrum and Spectrum for Synthetic Time History H1 CMSI .... 2.5-41 i 2.5-29 Selected Smooth Spectrwn and Spectrum for Synthetic Time History H2 CMSI .... 2.5-42 !
- 2.5-30 Selected Smooth Spectrum and Spectrum for Synthetic Vertical Time ,
H istory CM S I . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.5-43 2.5-31 Selected Smooth Spectrum and Spectrum for Synthetic Time History H1 CMS 3 . . . 2.5-44 ; E 2.5-32 Selected Smooth Spectrum and Spectrum for Synthetic Time History H2 CMS 3 . . . 2.5-45 ; 2.5-33 Selected Smooth Spectrum and Spectrum for Synthetic Venical Time , H istory CM S3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.5-46 i ! 2.5-34 Spectra at Ground Surface for All Soil Cases Using Synthetic Time i H istory CMS 3 H l . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.5-47 l l 2.5-35 Spectra at Foundation level for All Soil Cases Using Synthetic Time H istory CMS 3 H 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.5-48 : 3 2.5-36 Spectra at Ground Surface for All Soil Cases Using Venical CMS 3 Synthetic Time H istory . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.5-49 l 2.5-37 Spectra at Foundation Level for All Soil Cases Using Venical CMS 3 Synthetic l Time H istory . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , 2.5-50 t 2.5-38 Horizontal and Venical Free Field Spectra at Foundation Level . . . . . . . . . . . . 2.5-51 l 2.5-39 Envelope of Free-Field Surface Spectra Horizontal Motion,5% Damping . . . . . . . 2.5-52 2.5-40 Envelope of Free-Field Surface Spectra Venical Motion,5% Damping . . . . . . . . . 2.5-53 [ 1
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l Sy~ tem 80+ Design ControlDocument I i 2.0 Site Envelope Characteristics ! The System 80+= tt) Standard Design is designed on the basis of a set of assumed site-related parameters. These parameters wete selected to envelope most potential nuclear power plant sites in the United States. A summary of the assumed parameters is provided in Table 2.0-1. (( Detailed site characteristics will be provided by the COL applicant referencing the System K)+ Standard Design for any specific application.))2 These characteristics will be reviewed and compared to the enveloping assumptions of Table 2.0-1. Should specific site parameters or characteristics be outside the envelope of assumptions established by Table 2.0-1, the applicant will demonstrate that the design satisfies the requirements imposed by the specific site parameters and conforms to all design commitments and acceptance criteria described in this report. The remainder of this chapter identifies specific assumptions related to site characteristics that are employed in the evaluation of the System 80+ Standard Design. Table 2.0-1 Envelope of Plant Site Design Parameters Ground Water Mulmum Level: 2 feet below grade Hood (or Tsunami) LevelHi Maximum Level: I foot below grade Precipitation Maximum rainfall rate: 19.4 in/hr. and 6.2 in/5 min.[21 Maximum snow design load: 50 lb/sq. ft. Design Temperatures Ambient 1% Exceedance Values Maximum: 100'F dsy bulb /77'F coincident wet bulb 80'F wet bulb (non<oinc; dent) Minimum: -10*F (
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i System 80+ is a trademark of Combustion Engineering. Inc. w 2 COL information item: see DCD Introduction Section 3.2. Anwomt Dee+ nonaww. sne csueenwkates rege 2.0-r
Sy~ tem 80+ Design ControlDocument Table 2.0-1 Envelope of Plant Site Design Parameters (Cont'd.) Design Temperatures (Cont'd.) 0% Exceedance Values (Historical Limit excluding peaks < 2 hours) Maximum: 115'F dry bulb /80'F coincident wet bulb 81*F wet bulb (non-coincident) Minimum: -40'F Station Service Water inlet: 95'18 33 Condenser Circulating Water Inlet: s100*F Extreme Wind Basic Wind Speed: 110 mph importance Factors: * .0t41 fg,33:51 Tornadof61 Maximum tomado wind speed: 330 mph Rotational Speed: 260 mph Translational velocity: 70 mph Radius: 150 ft Maximum pressure differential: 2.4 psi Rate of pressure drop: 1.7 psi /sec Missile spectra: per SRP 3.5.1.4 Spectrum 11 Soil Properties l Minimum Bearing Capacity (demand): 12 ksf (static)l81 Best Estimate of Minimum Shear Wave Velocity: 700 ft/secI71 Best Estimate of Liquefaction Potential: None (at site-specific SSE level) Seismology l SSE Response Spectra: See Figures 2.5-38,39,40 l O
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Sy' tem 80+ Design ControlDocument O U Table 2.0-1 Envelope of Plant Site Design Parameters (Cont'd.) I
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l l Aircraft Ilazards Plant to airport distance, and Smi.< D <10mi. with annual 2 operation less than 500D flights or D > 10mi. with an annual operation , less than 1000D2 fl ghts l l (D = distance in mile') l D >5mi. with an annual operation less l Plant to edge of military training routes, and l than 1000 flights (D = distance in miles) Plant to edge of Federal airway, holding pattern, or D > 2mi, airport approach natterc. (D = distance in miles) l Meteorology Short-term dilution factor x/Q l.0x10r3; EAB = 0.5 mile l 1eng-term dilution factor x/Q 2.2x10-5; LPZ = 2.0 miles l l Notes:
~g Q [1] Probable maximum flood level (PMF), as defined in ANSI /ANS-2.8, ' Determining Design Basis Flooding at Power Reactor Sites.'
[2] Maximum value for I hour I sq. mile PMP with ratio of 5 minutes to I hour PMP of 0.32, as found in National Weather Service Publication llMR No. 52. [3] Maximum normal power and normal shutdown temperature of the Station Ser' ice Water System Intake based on one percent exceedance meteorologic conditions. See Section 9.2.5.1.3 for Ultimate licat Sink temperature interface requirement for a design basis accident concurrent with a loss-of-offsite power. [4] 50-year recurrence interval; value to be utilized for design of non-safety-related structures only. [5] 100-year recurrence interval; value to be utilized for design of safety-related stnictures only. [6] 10,000,000-year tornado recurrence interval, with associated parameters based on the NRC's interim position on Regulatory Guide 1.76. Pressure effects associated with potential offsite explosions are assumed to be non-controlling for the design. [7] Site profiles are given in Section .!.5, Profiles include consideration of variability of soil propenies. The lower bound of best estimate of soil shear wave velocity defines the lower bound of dynamic Soil-Structure Interaction analysis of the superstructure. [8] Bearing capacity is def'med at the foundation level of the Nuclear island str.icture. l Ch V Approvenf Doobyn Materiel Site Characteristics W97) Page 2.0-3
- l System 80+ oesten controlDocument !
l (j 2.1 Geography and Demography - (( Site-specific information on geography and demography will be provided by the COL applicant referencing the System 80+ design.)) l 2.1.1 Site Location and Description -
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No specific assumptions were employed in the evaluation of the System 80+ Standard Design. Site-specific information will include site and location description. 2.1.1.1 Site Location Considerations of site location for the System 80+ plants are made for the purpose of minimizing the risk significance of site-dependent characteristics on plant design. Sections 2.2 through 2.5 discuss System 80+ plant siting considerations for site-dependent design parameters. Chapters 3,9, and 13 also discuss ; System 80+ design features incorporated to mitigate the consequences of site-dependent and site-independent parameters. 2.1.1.2 Site Area Map The site area map is site-specific and will be provided with other site-dependent information. , i 2.1.1.3 Boundaries for Establishing Emuent Release Limits Boundary distances have been selected for an exclusion area and a low population zone as defined in 10 CFR 100.11. These distances are employed in Sections 2.3.4 and 2.3.5 to calculate estimates for diffusion of radiological releasee. A site specific application will address the combined influences of site characteristics and local meteorology on diffusion of releases. ' 2.1.2 Exclusion Area Authority and Control No specific assumptions were employed in the evaluation of the System 80+ Standard Design. Site-specific information will include exclusion area authority and control.
- 2.1.3 Population Distribution No specific assumptions were employed in the evaluation of the System 80+ Standard Design. Site-specific information will include population distribution. ;
r3 3 COL information itern; see DCD Introduction Section 3.2. Amwwwed Dostyre Adesorief S/le Checaeristics rege 2.f.f i
i Sv-tem 80+ Desian contmlDocument 2.2 Nearby Industrial, Transportation and Military Facilities 1 I Industrial, transportation and military hazards are discussed below. (( Site-specific information on industrial, transportation, and military hazards will be provided by the COL applicant referencing the > System 80+ design.))l ! i 2.2.1 Aircraft Hazards ; i r A site is acceptable for the System 80+ without further review if the distances from the plant meet the { following requirements: j e The plant-to-airport distance D is between 5 and 10 statute miles, and the projected annual : number of operations is less than 500 D2 flights, or the plant-to-airport distance D is greater than 10 statute miles, and the projected annual number of operations is less than 1000 D' flights. l e The plant is at least 5 statute miles from the edge of military training routes, including low-level ! l training routes, except for those associated with a usage greater than 1000 flights per year, or ! where activities (such as practice bombing) may create an unusual stress situat on. e The plant is at least 2 statute miles beyond the nearest edge of a federal airway, holding pattern, . 1 or airport approach pattern. l l 2 3 If the above site proximity acceptance criteria are not met, or if sufficiently hazardous military activities i are identified, a detailed review of the aircraft hazards must be performed to qualify a specific site for the System 80+ plant. l l . 2.2.2 Transpottation Site-specific information will include hazards related to transportation. ; 4 The ultimate heat sink, which is not included in the System 80+ scope for design certification, provides l the source of cooling water for all safety-related plant systems and components during all modes of ! operation. Interface requirements (Section 9.2.5.1.3) are identified to eliminate the potential impacts on plant operations from boat or barge accident events. 2.2.3 Other Industrial Hazards On and Offsite [ Site-specific information will include onsite and offsite industrial hazards. i l l l 1 8 COL information item; see DCD Introduction Section 3.2. ! i 4 gummer Des @m asessedst - Ses owsoeurt; alce (2/95/ pope 2.2 7 ; ~
System 80 + Design ControlDocument D Q 2.3 Meteorology Meteorology parameters are as specified in Table 2.0-1. (( Site-specific meteorology information will be provided by the COL applicant referencing the System 80+ design.))' 2.3.1 Regional Climatology Site-specific information will include regional climatology. t i 2.3.1.1 Low Temperature Effects The effects on the plant from low terrperature events, such as frost, snow fall, and ice cover, are considered in the design process. Structures are designed to withstand loadings in excess of the loads generated from combinations of snow, ice, and rain. Ventilation paths are designed and reviewed to verify that they are free from snow blockage. i 2.3.2 Local Meteorology No specific assumptions were employed in the evaluation of the System 80+ Standard Design other than the assumptions in Sections 2.3.4 and 2.3.6 that establish the values of relative concentrations for the offsite accident analyses and control room operator dose analysis as presented in Chapter 15 and , Section 6.4, respectively. O b As indicated in Table 2.0-1, pressure effects and missile spectra associated with the design tornado are . considered to be controlling. Site-specific information will include an evaluation to assure that this assumption is not violated for the specific site selected or will include analysis results for any potential hazards that are more limiting than the parameters given in Table 2.01. 2.3.2.1 Tornados Tornado characteristics are as specified in Table 2.0-1. Tornado-generated missiles are considered in the System 80+ Standard Design (See Section 3.5). Structures housing safety-related equipment are designed to withstand the loadings generated by 330 mph winds. Tornados that occur at the plant site, causing extensive damage to the switchyard and a prolonged loss l of offsite power, are quantitatively evaluated. Section 3.3 provides additional information regarding design for tornado loading. 2.3.3 Onsite Meteorological Measurements Programs i No specific assumptions were employed in the evaluation of the System 80+ Standard Design. Information related to onsite meteorological measurements programs will be included in the site-specific meteorology information. a
)
3 COL information item: see DCD Introduction Section 3.2. , 4 proved Desiers Acesene!. She Character 6rtics Pope 2.31
System 80+ Design control Document 2.3.4 Short-term (Accident) Diffusion Estimates (x/Q) Atmospheric relative concentration of radiological releases is expressed as x/Q, where x is the concentration in curies per cubic meter at the receptor and Q is the rate of release in curies per second. Calculation of site-specific values of x/Q will be provided with a site-specific application that includes the meteorological measurements program. In lieu of site meteorological data, an assumed set of atmospheric conditions is employed as follows to determine the values of x/Q for the System 80+ Standard Design accident analyses. Ground-level 0-2 hour atmospheric dilution factors (x/Q) were calculated at 0.5 mile Exclusion Area Boundary (EAB) using a ground-level bivariate normal, or Gaussian diffusion model modified for sourc. configuration (i.e. building wake) and lateral plume meander under neutral and/or stable atmospheric conditions. The methodology used in the development of accident EAB x/Q's followed the guidance presented in NRC Regulatory Guide 1.145. Input parameters used included meteorological data representative of an 80-90th percentile U.S. commercial nuclear power plant site. For time periods greater than 2 Lours (i.e.,0-8 hours, 8-24 hours,1-4 days, and 4-30 days), x/Q values were determined for a 2.0 mile Low Population Zone (LPZ) using logarithmic interpolation techniques which are also described in Regulatory Guide 1.145. Table 2.3-1 presents the accident ground-level x/Q values at both the EAB and LPZ receptors. The values of x/Q in Table 2.3-1, when combined with plant design and operational limit characteristics, yield acceptable doses following postulated accidents. This analytical combination is discussed in Section 15.0.4. Details of the dose calculation methodology are given in Appendix 15A. 2.3.5 Long-term (Routine) Diffusion Estimates (x/Q) Annual average atmospheric dilution factors (x/Q) and relative deposition factors (D/Q) at the worst case locations for various respective pathways, were calculated using the methodology as presented in NRC Regulatory Guide 1.111 and stack release point characteristics provided in Table 2.3-7. The annual average x/Q value utilized for the residence which resulted in the highest stfsite dose via the plume 3 4 submersion and inhalation pathways was 7.2 x 10 5 sec/m with a proportional D/Q value of 1.3 x 10 m-2. The annual average x/Q value utilized for the worst case food pathway, including vegetable. meat, 2 and milk receptors was 1.5 x 10-5 sec/m3 with a proportional D/Q value of 2.3 x 10-8 m . A site-specific application will determine site-specific values oflong term x/Q for comparison to the values used for the System 80+ standard design routine doses. 2.3.6 Onsite (Accident) Diffusion Estimates (x/Q) Onsite accident I hour (i.e., applicable for 0-8 hours) atmospheric dilution factors (x/Q) were calculated at the north south fresh air intakes to the control room using a time based building wake model as described by Reference 1. The north and south air intakes are located on the northwest and southeast corners of the nuclear annex, respectively. In addition, x/Qs were also calculated at the center of the control room and at the doorway leading to the control room itself. The x/Q values were calculated using the same meteorological database as that referenced in Section 2.3.4 along with specific source to receptor distances and building configurations used for wake considerations. For all distances, a form of the x/Q wake model associated with normal atmospheric diffusion was utilized. For time periods greater than 8 hours (i.e., 8-24 hours,1-4 days, and 4-30 days) the 0-8 hour time based building wake x/Q values were adjusted using appropriate wind and occupancy factors from Reference 2. A second adjusunent was made to the x/Q values so that the control room doses presented in Table 6.4-1 (calculated as described in Section 6.4.3) would represent limiting values. This adjustment consists of factors necessary to raise AS4weved Design Material . Site characteristics Page 2.3 2
System 80+ Design ControlDocument p/ i the calculated doses to their respective limits. Sets of such factors were determined for each event, with the minimum factor then representing the event. Individual x/Q values were then increased by the minimum factor for any event involving that particular source and receptor. In this way limiting doses are reported involving maximum acceptable values for each x/Q. Tables 2.3-2 and 2.3-3 present resultant x/Q values for the various release points to both the north and south air intakes, respectively. Tables 2.3-4 and 2.3-5 present resultant x/Q values at the center of the control room and at the doorway to the control room, respectively. As a result of the dual air inteke/ automatic selection feature design of the System 80+ control room air intake system, the reduction factors as presented in Table 2.3-6 can be applied for the north and south air intake x/Q values. It should be noted that the reduction factors are only applicable for releases occurring from the unit vent on top of the containment shield building and the containment shield building wall itself. Calculation of site-specific x/Q values for the control room will be provided with a site-specific application for comparison to the values used for the System 80+ standard design. References For Section 2.3
- 1. Ramsdell, J. V., " Diffusion in Building Wakes for Ground-Level Releases," Atmospheric Environment, Volume 24B, Number 3 Pages 377-388.
(O_) 2. Murphy, K. G. and Campe, K. M., " Nuclear Power Plant Control Room Ventilation System Design for Meeting General Criterion 19," 1974 Proceeding of the 13th AEC Cleaning Conference, San Francisco, California, CONF-740807, U.S. Atomic Energy Commission, Washington, D.C. Table 2.3-1 Radiological Dilution Factors (x/Q) Distance Time Period Dilution Factor (sec/ cubic meter) 3 EAB (0.5 mile) 0-2 hours 1.00 x 10 LPZ (2.0 miles) 0-8 hours 1.35 x 10 4 LPZ (2.0 miles) 8-24 hours 1.00 x 10 4 LPZ (2.0 miles) 1-4 days 5.40 x 10-5 LPZ (2.0 miles) 4-30 days 2.20 x 10-5 .O Approved Desepts Material- Site Characteristics Page 2.3-3
System 80+ Design ControlDocument Table 2.3-2 Onsi' Accident x/Q Values at the Control Room North Air Intake Release Point 0-8 Hour 8-24 Hour 1-4 Day 4-30 Day Unit Vent - At Tol, of Containment Shield Building 1.55E-3ti31 1.25E-3t31 6.39E-4l31 2.62E-4!31 Center Main Steam Valve 6.87E-3 PJ Pl Pl House Number 1 Center Main Steam Valve 1.90E 3 P1 Pl P1 Ilouse Number 2 Edge Main Steam Valve 3.88E-2 121 PJ Pl House Number 1 Edge Main Steam Valve 8.43E-3 Pl 121 pl House Number 2 Contaimnent Shield 5.65E-3131 4.93E-3131 2.65E-3l33 1.28E-3t31 Building Wall Notes: 3 [1] 1.55E-3 = 1.55 x 10'3 sec/m . [2] Release period not greater than 8 hours. [3] Can be reduced by the appropriate reduction factor presented in Table 2.3-6. O Table 2.3-3 Onsite Accident x/Q Values at the Control Room South Air Intake Release Point 0-8 Hour 8-24 Hour 1-4 Day 4-30 Day Unit Vent - At Top of Containment Shield Building 1.42E-3II33 1.18E-3I33 5.45E-4131 1.66E-4I31 Center Main Steam Valve 1.52E-3 P1 Pl P1 House Number 1 Center Main Steam Valve 6.87E-3 Pl Pl Pl House Number 2 Edge Main Steam Valve 6.88E 3 [2] [2] PJ House Number ! Edge Main Steam Valve 3.88E-2 91 PJ PJ House Number 2 Containment Shield 5.65E-3131 4.74 E-3l31 2.43E-3l33 1.02E-3t31 Building Wall Notes: - l [1] 1.42E-3 = 1.42 x 10'3 sec/m3 . [2] Release period not greater than 8 hours. l [3] Can be reduced by the appropriate reduction factor presented in Table 2.3-6. l l l Approved Design Material site Charactenistics Page 2.M l l l
I I l l System 80+ Design controlDocument i
) /N V Table 2.3-4 Onsite Accident x/Q Values at the Center of the Control Room ;
Release Point 0-8 Hour F-24 Hour 1-4 Day 4-30 Day Unit Vent - At Top l of Containment Shield Building 3.20E-#1 2.84 E-3 1.42E 3 5.83E-4 l Center Main Steam Valve 6.08E-3 [2] [2] [2] l House Number 1 , Center Main Steam Valve 6.72E-3 [2] [2] [2] House Number 2 Edge Main Steam Valve 2.06E-2 [2] [2] [2] House Number 1
- Edge Main Steam Valve 2.28E-2 (2) [2] [2]
House Number 2 Containment Shield 5.88E-2 4.97E-2 2.65E-2 1.33E-2 Building Wall Notes: 3 [1] 3.20E-3 = 3.20 x 10'3 sec/m . [2] Release period not greater than 8 hours. . O Table 2.3-5 Onsite Accident x/Q Values at the Doorway to the Control Room Release Point 0-8 Hour 8-24 Hour 1-4 Day 4-30 Day Unit Vent - At Top of Containment Shield Building 2.27E-3Ill 2.02 E-3 1.06E-3 4.32E-4 [2] Center Main Steam Valve 6.08E-3 [2] [2] House Number 1 Center Main Steam Valve 6.72 E-3 [2] [2} [2] House Number 2 Edge Main Steam Valve 2.06E-2 [2] [2] [2] House Number 1 _ Edge Main Steam Valve 2.28E-2 [2] [2] [2]
!!ouse Number 2 Containment Shield 1.92E-2 1.63E-2 8.93E-3 4.39E-3 Building Wall Notes:
[1] 2.27E-3 = 2.27 x 10-3 sec/m3. [2] Release period not greater than 8 hours. b Approved Deedpar Adeferief S/fe (:horectorirtics Pepe 2.15 e
System 80+ Design ControlDocument 1 Table 2.3-6 Atmospheric Dilution Factor (x/Q) Reduction Factors Time Period Reduction Factor i 1 0-8 Hour 4 8-24 Hour 5 1-4 Day 7 4 30 Day 10 Table 2.3-7 Unit Vent Release Point Characteristics Characteristic Valueill Height Above Grade (feet) 194 Height Above Annulus Building Dc ifeet)[2] 20 Relative Temperature Difference Between Effluent and Ambient Air (*F) 0 Stack Shape Right Circular Cylinder Stack Diameter (feet) 7.66 Vent Design Flow Rate (CFM) 178,650 Effluen' '.xit Velocity (ft/sec)l31 60 Notes: [1] Approximate values to be used in calculating site-specific atmospheric dilution factors. [2] Unit vent stack located on top of Annulus Building Dome. t [3] Corresponds to design flow rate. Anrewd tusion untew. site crwoctentics rog ,2.2-s
l System 80+ Deslan controlDocument G Q 2.4 Hydrologic Engineering Hydrologic engineering parameters are as specified in Table 2.0-1. (( Site-specific hydrologic information will be provided by the COL applicant referencing the System 80+ Standard Design.))l This information will include the following considerations as appropriate for the specific site: external floods; probable maximum flood on streams and rivers; potential dam failures; probable maximum surge and seiche flooding; probable muimum tsunami loading; ice effects; cooling water canals and reservoirs; channel diversions; flood protection requirements; cooling water supply; groundwater; potential accidental release of liquid effluents in ground and surface water, and technical specifications and operation requirements. 2.4.1 External Floods 1 The site-specific flooding projections will consider severe precipitation, snow melt, flooding due to ice cover, river flooding, ocean flooding, tsunami flooding, seiche effects, wave and storm surge effects, hurricane effects, high lake levels and any other effects appropriate for the specific site. Intake structures will be designed to preclude the potential for debris blockage caused by wind, wave, and other effects. 2.4.2 Internal Floods See Section 3.4. O V l I i v l I I COL information item; see DCD Introduction Section 3.2. AMeeved Desipre neoneriel Site Cterecsonistics 2.41
a System 80 + o sign controlDocument f] V 2.5 Geology, Seismology, and Geotechnical Engineering (( Site-specific geology, seismology, and geotechnical engineering information and characteristics of the site and region surrounding the site, including site acceptance criteria, will be provided by the COL applicant referencing the System 80+ Standard Design.))3 To cover a range of possible site conditions where System 80+ may be constructed, a range of generic site conditions was selected for geologic and seismologic evaluation (Figure 2.5-1). The System 80+ is a standard plant design to be built on a suitable site. The basis for selecting any particular site is documented in the site-specific Safety Analysis Report (SAR). Site geologic features, seismological features, liquefaction potential, site instability, ground rupture and man-made conditions are included in the site-specific SAR. Site-specific investigations, including borings, are conducted in accordance with 10 CFR 50 and 100 (Reference 2) and Standard Review Plan 2.5 (Reference 3). Any deviations from Reference 3 are identified and justified in the site-specific SAR. The total depth to bedrock for each site condition and the dynamic soil properties (in terms of maximum shear wave velocities and their variation with depth, and in terms of the variations of modulus and damping with strain) were established to cover a wide range of sites and to provide reasonably conservative results. Using these site conditions and the variations of maximum shear wave velocities, 13 cases were developed; 12 soil cases and one rock outcrop case. For analysis of the superstructure, the rock outcrop case was further broken into two subcategories. One subcategory covers the case where the structure is sitting on rock, with its embedded portion not restrained. The second subcategory covers cases where the backfill material has rock type characteristics. Further discussions on the rock outcrop q case is provided in Appendix 3.7-B. The cases selected are summarized in Section 2.5.2, and more V details for each case are included in Appendix 2A. For the System 80+ seismic design, three control motions were developed which, when combined cover the majority of potential sites in the continental U.S. The twelve generic soil sites and one rock site were evaluated for each of the control motions. To cover sites with deep soil deposits, a control motion with a Regulatory Guide 1.60 spectral shape is j used as the input motion to the ground surface of each site. To cover shallow soil sites, two rock ; motions applied at a hypothetical rock outcrop are used. The selection of the two rock outcrop motions i was performed using low frequency content consistent with industry-wide accepted response spectra, and i high frequency content that exceeds the current industry practice. The enrichment of the rock outcrop , motions with high frequency content is consistent with recent studies on Eastern North America seismicity l and is a proactive measure of the System 80+ design in anticipation of future trends in the industry regarding seismic motions. The control motions described in this section are intended to provide future owners of System 80+ design with high confidence that the design is suitable for most sites in the United States. ! l To assess whether a site is suitable for construction of the System 80+ Standard Design, both the following Acceptance Criteria (Site Conditions and SSE Ground Motions) must be satisfied. i (^) V
' COL information item; see DCD Introduction Sect'.on 3.2. !
Approved Desipts Mate *ie!- Site Charactenstics Page 2.5-1
i Syotem 80+ Design controlDocument i
- 1. Site Conditions
- I
- The soil profile should have a (low-strain) shear wave velocity profile within the range shown in Figure 2.5-2. Although the soil response of a specific new soil profile could differ from the results obtained for each of the cases included in Reference 1, it would be covered by the envelope of the soil cases considered in the soil response analyses of Reference 1.
- A soil site having a total depth to bedrock greater than that shown in Figure 2.5-1 is acceptable, because it would be covered by the soil cases analyzed.
- All rock sites (with no soil deposits below the foundation level) are acceptable.
- 2. (SSE) Ground Motion l The acceptance criteria for the Ground Motion are given in Section 2.5.2.5.3.
2.5.1 Basic Geologic And Seismic Information The objective of this section is to describe geologic and seismic features that affect the site under review. Site-specific regional and site physiography, geomorphology, stratigraphy, lithology, and tectonics information will be included in the site-specific information. O O Aoproved Des @ Matenial Site Characterktics (1/97) Page 2.5 2
System 80+ oesian controt Document i 2.5.2 Vibratory Ground Motion Site-specific geological, seismological, and geotechnical data will be included in the site-specific information. 2.5.2.1 Seismicity: The complete historical record of earthquakes in the region will be included in the site-specific data. At that time, all available information pertaining to and concerning epicenter coordinates, depth of focus, origin time, highest intensity, magnitude, seismic moment, source mechanism, source dimensions, , distance from site, strong motion recordings, and earthquake-induced geologic failures will be provided for each event. 2.5.2.2 Geologic and Tectonic Characteristics of Site and Region I in the site-specific information, tectonic provinces will be established based on the development and characteristics of the current tectonic regime of the region and the pattern and level of historic seismicity. l This will lead to a determination of the earthquake-potential of all identified geologic structures within the regions. ,
- l 2.5.2.3 Correlation of Earthquake Activity with Geologic Structure or Tectonic Provinces l i
The relationship between earthquake activity history and the geologic structure or tectonic prov;nc s of a region will be included in the site-specific information. Detailed accounts comparing and contrasting . O the geologic structure (s) involved in the earthquake activity with other areas within the tectonic provinces will be supplied. 2.5.2.4 Maximum Earthquake Potential The free field control motion described in Section 2.5.2.5 will be shown to envelop the maximum , possible vibratory ground motion at the site, by the site-specific analysis. This determinatior, will be ; based on the maximum credible earthquake associated with each site-specific geologic structure, or on l the maximum historic earthquake associated with each tectonic province, and will be supplied in the site- i specific SAR. i 2.5.2.5 Seismic Wave Transmission Characteristics of the Site 2.5.2.5.1 Control Motion The Control Motion design response spectra are anchored to a 0.3g peak ground acceleration. They were '. developed with the objective of being in full compliance with the SRP requirements as well as the EPRI ' ALWR requirements report. Again, to cover a maximum range of possible sites where the System 80+ standard design may be constructed, three separate control motion spectra were developed. These are: e Control Motion Spectrum 1 (CMSI): This spectrum is included for application at the free-field ground surface. It is identical to Regulatory Guide 1.60 (R.G.1.60) spectrum and it is , i considered in order to cover sites with deep soil deposits. Furthermore, because CMSI is a standardized response spectrum shape, it is considered as an appropriate control motion for both rock and soil sites. l 4powwont Deelen Menenlof. Sise Chareceerienics Pege 2.5-3 l l l
System 80+ Design ControlDocument
- Control Motion Spectrum 2 (CMS 2): This is a rock outcrop spectrum and is developed to cover sites typical of Eastern Nonh America which could be subjected to earthquakes with high frequency content.
- Control Motion Spectrum 3 (CMS 3): This is a rock outcrop spectrum and is developed based on recommendations of the NUREG/CR-0098 (Reference 4) primarily to cover lower frequency motions which may not be covered by CMS 2. It is also greatly enhanced in the high frequency range to cover canhquakes with high frequency content. The maximum spectral acceleration range is extended to 15 Hz, as opposed to 8 Hz which is used in NUREG/CR-0098 motions.
All of the above Control Motion Spectra are shown in Figure 2.5-5. All three motions (CMS 1, CMS 2, and CMS 3) are used for application at rock sites. For soil sites, CMS 2 and CMS 3 are intended for application at the rock outcrop, and CMS 1 is intended for application at the free-field ground surface. All three motions are applied to each of the 13 sites to conservatively cover all combinations. The logic for selection process of each of these control motion spectra is described in more detail below: Selection Process for CMSI The spectrum shape corresponding to this control motion is as per the requirements of R.G.1.60. This spectrum shape is chosen in order to be in full compliance with the SRP Section 2.5 requirements as well as the EPRI ALWR requirements, and is intended to cover deep soil sites. The control motion is anchored to a peak ground acceleration of 0.3g for the two horizontal directions and the venical direction. Selection Process for CMS 2 The spectrum shape corresponding to this control motion is for application at the rock outcrop surface, is an 84 percentile curve, and is developed considering NUPEG/CR-0098 recommendations as well as ground motions deemed appropriate for the Eastern Nonh American continent. The intent of this spectral shape is to cover various soil sites overlaying a competent material as well as having rock outcrop motion characteristics typical of Eastern North America. The construction of this spectrum shape is shown in Figure 2.5-6. As can be noted from this figure, the spectral ordinates were kept equal to those obtained using NUREG/CR-0098 for frequencies lower than 3.3 Hz, with maximum ground velocity of 24 in/sec/g, which again is typical of expected eanhquakes for the Eastern United States. For higher frequencies, panicularly above 10 Hz, the selected spectral ordinates are based upon ground motion estimates appropriate for Eastern Nonh America and, as can be seen, are significantly higher than those obtained using the NUREG/CR-0098. This control motion is anchored to a peak ground acceleration of 0.3g and peak ground velocity of 7.2 in/sec for the two horizontal directions, in the venical direction, the control motion is anchored to a peak ground acceleration of 0.2g and peak ground velocity of 4.8 in/sec. The selection of 0.2g at the rock outcrop for the venical direction leads to venical spectra at the ground surface that equal or exceed the horizontal spectra at the ground surface over a significant range of frequencies for most of the soil cases. O Approved Design nieterial Site Characteristics Pope 2.5-4 6
..- - . .- . . - . - . - - . . _ _ = - - . . . - . - - . .- - . .-
t Synom 80+ Desian contralDocument blarelan Process for CMS 3 The spectrum shape corresponding to this control motion is developed for application to rock outcrop l surface, is an 84 percentile curve, and is in full compliance with the recommendations of NUREG/CR- i 0098 with maximum ground velocity of 36 in/sec/g representing typical sites in Western North America. i CMS 3 is greatly enriched in the high frequency end of the spectrum to cover earthquakes with high , j frequency content. The maximum spectral acceleration range extends from 2.2 Hz to 15 Hz. Again, this .
- control motion is anchored to a peak ground acceleration of 0.3g for the two horizontal directions and ;
0.2g for the vertical direction. ; t ! Synthetic Time Histories 4
- Synthetic time histories were generated for each of the components, Horizontal-1, Horizontal-2 and
Vertical, of each of the control motions CMS 1, CMS 2 and CMS 3, respectively. The spectral ordinates l calculated for each synthetic time history and the corresponding smooth spectra are shown in Figures 2.5- ; 7 through 2.5-9 for the CMS 2 motion Figures 2.5-28 through 2.5-30 for the CMSI motion, and Figures ! 2.5-31 through 2.5-33 for the CMS 3 motion. The spectral ordinates of each synthetic time history conservatively envelop the target smooth spectra at a sufficient number of frequency points to satisfy the ! SRP Section 2.5 criteria for development of synthetic time histories, j
.f
! The characteristics of each synthetic time history (accelerogram, velocity and displacement time histories ; and Power Spectral Density (PSD) ) are presented in Appendix 2B. The average PSD of CMSI fully ) , complies to the SRP Section 3.7.1, Appendix A guidelines for Power Spectral Densities of motions that are based on a Regulatory Guide 1.60 spectral shape. For all three motions CMS 1, CMS 2 and CMS 3, the synthetic time histories in the three directions are statistically independent with correlation coefficients ! less than 0.2. 3 2.5.2.5.2 Genede Soil Sites Generic soil sites were selected by first choosing four generic site categories. These categories were chosen to represent appropriate total thickness of soil overlying bedrock. The four categories are shown i schematically in Figure 2.5-1. Site Category A consists of approximately 51 feet of soil overlying ! bedrock; 51 feet is the approximate embedment depth selected for the System 80+. The soils in site Category B extend to a depth of 100 feet and those in Categories C and D extend to depths of 200 and 300 feet, respectively. ) One case was selected for Category A and one case for Category D; these were designated Case A-1 and Case D-1. Four cases were initially selected for site Category B; these were designated Cases B-1, B-2, B-3 and B-4. Three cases were initially selected for site Category C; these were designated Cases C-1, C-2 and C-3. Upon examination of the results of the response analyses for these cases, three additional ~ cases were added. The additional cases were designated Cases B-1.5, B-3.5 and C-1.5. These latter cases were selected to provide an estimate of the response at frequencies that were not considered to be < adequately covered by the other cases. The variations of maximum shear wave velocities with depth assigned for each case are summarized in Appendix 2A Figures 2A-2 through 2A-13. The shear wave velocity distributionwith depth was selected to provide a reasonably wide range and also to provide significant contrast in velocities at certain depths for a selected number of cases. The range of maximum shear wave velocities used for all the cases Os considered in this study is presented in Figure 2.5-2. More details about each case are given in Appendix 28. Anumn onow aanneer- sw caweeenssce neue 2.5-s
System 80+ Design ControlDocument The variation of shear modulus with shear strain was based on using the upper curve from the range published by Seed and Idriss (Reference 5) as shown in Figure 2.5-3. The variations of damping with j - shear strain was based on the lower curve from the range published by the same authors, as shown in Figure 2.5-4. 2.5.2.0.3 Site Acceptance Criteria The CMS 1, CMS 2, and CMS 3 control motions were developed for application in the seismic design of the System 80+ Standard Design. Site evaluations and demonstrations that acceptance criteria have been l met will be included in site-specific information. See also Section 3.7. According to these acceptance criteria: l 1. For a rock site, site-specific free-field ground surface response spectra at 5 % of critical damping for the horizontal and vertical directions, will be developed and compared to the envelope of the CMS 1, CMS 2, and CMS 3 control motions (all with 5% of critical damping). These envelope spectra are shown in Figure 2.5-38 for the horizontal and vertical directions. Figure 2.5-38 is the envelope for the three CMS control motions whose individual horizontal components are depicted in Figure 2.5-5. If the site-specific response spectra are enveloped by the response spectra in Figure 2.5-38, the site is acceptable for construction of the System 80+ Standard Plant Design. Alternatively, if the site-specific spectra exceed either response spectrum in Figure 2.5-38 at any frequency, a site-specific evaluation can be performed. This evaluation will consist of a site-specific structural dynamic analysis and generation of in-structure response spectra at seven critical locations to be compared to the design response spectra in Figures 3.7D-1 through 3.7D-21. If the in-structure design spectra from the site-specific evaluation are enveloped by the in-structu.c design spectra for each of the critical locations as identified below, the site is acceptable for construction of the System 80+ Standard Plant Design. l
- Foundation Basemat Elevation +50 ft. Figures 3.7D-1,3.7D-2, and 3.7D-3 l
- Interior Stmeture Elevation +91.75 ft. Figures 3.7D-4,3.7D-5, and 3.7D-6 l
- Control Room Area 1 Elevation +115.5 ft. Figures 3.7D-10,3.7D-II, and 3.7D-12 l
- Control Room Area 2 Elevation +115.5 ft. Figures 3.7D-13,3.7D-14, and 3.7D-15 l
- Top of Steel Containment Vessel Elevation +251 ft. Figures 3.7D-16,3.7D-17, and 3.7D-18 l
- Interior Structure Elevation +146 ft. Figures 3.7D-7,3.7D-8, and 3.7D-9 l
- Shield Building Elevation +263.5 ft. Figures 3.7D 19,3.7D-20, and 3.7D-21 l 2. For a deep or shallow soil site, site-specific response spectra at 5% of critical damping for the horizontal and vertical directions at the free-field ground surface will be developed. The site-specific free-field surface spectra will then be compared to the envelope of the CMSI spectra and the surface l spectra developed from CMS 2 and CMS 3 control motions (all with 5% of critical damping). These envelope ground surface spectra are shown in Figures 2.5-39 and 2.5-40 for the horizontal and the l
vertical directions, respectively. Tables 2.5-1 and 2.5-2 define the spectra of Figures 2.5-39 and 2.540 in digitized format. Approveef Design Meterial Site Chorectoristics (1/97) Page 2.5 6
System 80+ Design ControlDocument If the site-specific response spectra are enveloped by the spectra in Figures 2.5-39 and 2.5-40, the site V is acceptable for construction of the System 80+ Standard Plant Design. Alternatively, if the site-specific spectra exceed either spectrum in Figures 2.5-39 and 2.5-40 at any frequency, a site-specific evaluation can be performed. This evaluation will consist of a site-specific < soil structure interaction analysis and generation of in-structure response spectra at seven critical locations to be compared to the design response spectra in Figures 2.5-39 and 2.5-40. If the spectra from the site-specific evaluation are enveloped by the design spectra for each of the critical locations as identified in item 1 above, the site is acceptable for construction of the System 80+ Standard Plant Design. 2.5.2.5.4 Site Specific Seismic Spectra Site-specific seismic design response spectra for use in the design and qualification of site-specific structures, systems, and components not included in the design certification scope for System 80+ standard plants will be developed and provided with site-specific information. The following criteria shall be used in developing the minimum site-specific seismic design requirements.
- 1. The horizonta' and vertical free-field ground surface site-specifi: response spectra shall be developed using approved NRC procedures.
- 2. The System 80+ certified design horizontal and vertical hulatory Guide 1.60 design response spectrum shapes anchored to 0.30g peak ground acceleration shall be scaled throughout their entire
_ fregt;ncy range such that the minimum spectral amplitudes of the certified design spectra are equal to the maximum spectral amplitudes of the horizontal and vertical site-specific ground motion spectra, (' respectively, in the 5 to 10 hertz frequency range. l l
- 3. The resulting design response spectra shall be defined as the minimum seismic design requirement for ,
design and qualification of site specific structures, systems, and components for the System 80+ l standard plant. 2.5.2.6 Safe Shutdown Earthquake For the Safe Shutdown Earthquake (SSE), the following Peak Ground Accelerations (PGA) were considered: 1 CMS 1 motion. Horizontal PGA = 0.3g Vertical PGA = 0.3g CMS 2 motion: Horizontal PGA = 0.3g Vertical PGA = 0.2g 1 l l t V i l
.% J Dee> neew She Detectorireics (1/97) Page 2.5-7
Sy~ tem 30 + Design ControlDocument CMS 3 motion: Horizontal PGA = 0.3g Vertical PGA = 0.2g The associated spectral characteristics of each motion are presented in Section 2.5.2.5 and Appendix 2B. 2.5.2.7 Site Response 2.5.2.7.1 Method of An.dysis The response of each soil case was obtained using an equivalent linear response analysis for both the shear wave and compression wave. The response analysis methodology of the computer code SHAKE l was used. For the shear wave propagation, synthetic time history H1 of motion CMS 2 was applied as the input rock outcrop motion. The strain-compatible modulus and damping values were then obtained for that soil case. The computed strain iterated soil properties of the twelve soil cases using the H1 time history of the CMS 2 motion were used in all the soil and SSI analyses in order to retain these properties l as standard for the soil media. These properties were then used without further modifications for the l analysis involving synthetic time history H2 of CMS 2 and time histories H1 and H2 of CMS 3 as the input rock outcrop motions. (The strain-compatible modulus and damping values thus obtained are listed in , the tables included in Appendix 2C). For analyses involving the vertical component, the strain-compatible shear moduli were converted to constrained moduli assuming a Poisson's ratio of 0.4 for all sublayers. The strain-compatible damping values were multiplied by 1/3 to provide an estimate of the damping associated with the propagation of p-waves. l The same strain iterated soil properties were also used in the SSI analyses involving the CMS 1 motion. As discussed in Section 3.7.1, in the SSI analyses, the CMS 1 motion was applied at the free-field ground surface. 2.5.2.7.2 Results Spectra curves from the CMS 2 analyses are presented in Figures 2.5-10 through 2.5-27. Figures 2.5-10 through 2.5-12 show the spectral ordinates calculated at the ground surface for all casea considered using synthetic time history (CMS 2) H1 as input motion. The corresponding spectra calculated at the foundation level are shown in Figures 2.5-13 through 2.5-15. The spectra calculated at the free field ground surface using synthetic time history (CMS 2) H2 are presented in Figures 2.5-16 through 2.5-18 and those at the foundation level in the free field are presented in Figures 2.5-19 through 2.5-21. The corresponding spectra for the vertical component are presented in Figures 2.5-22 through 2.5-27. Spectra curves from the CMS 3 analyses are presented in Figures 2.5-34 through 2.5-37. Figure 2.5-34 shows the spectral ordinates calculated at the ground surface for all cases considered using synthetic time history (CMS 3) H1 as input motion. The corresponding H1 spectra calculated at the foundation level are shown in Figure 2.5-35. Figure 2.5-36 shows the spectral ordinates calculated at the ground surface for all cases considered using the CMS 3 vertical synthetic time history as input motion. The corresponding vertical spectra calculated at the foundation level are shown in Figure 2.5-37. The responses for the 12 soil cases were obtained using conservative approaches for selecting the free field rock outcrq motion, the range of soil profiles including depths, variation of sher.r wave velocities with depth and velocity contrasts together with the dynamic material properties. Approved Desse n Material Site Characteristics Page 2.5-8
System 80+ Design controlDocument
^Qf} The results depicted in Figures 2.5-10 through 2.5-27 are applicable to a wide range of soil deposits.
Thus, a soil profile for which the distribution of maximum shear wave velocities with depth is within the range shown in Figure 2.5-2 would have a response well covered by the results although the results for a specific new case could differ from the results obtained for each of the cases analyzed. Potential site instability or ground rupture due to steep topography, soft soils, liquefaction or fault rupture are treated as site-specific issues. The enveloping analyses performed were based on the distribution of maximum shear wave velocities with depth and thus did not require specification of a depth to water table at the site. Therefore, the water table can be at any depth as long as the variations of maximum shear wave velocities with depth are within the range discussed above and provided that any local site instability issues are resolved. 2.5.3 Surface Faulting System 80+ g.! ants will not be designed to withstand surface faulting related to earthquakes. Site-specific surface and subsubce geological and geophysical information to demonstrate that evidence of a potential for surface faulting has not been found will be included in the site-specific information. 2.5.4 Stability of Subsurface Materials and Foundations Subsurface material parameters are as specified in Table 2.0-1. Site-specific information relating to stability of subsurface materials and foundations resulting from site geotechnical and geophysical investigations will be included in the site-specific information. Information for the specific site will O O include: geologic features underlying the site; propenies of materials underlying the site and a description of the state of the art methods used to determine the static and dynamic engineering properties of foundation soils and rock in the site area; data pertaining to soil layers (including their thicknesses, densitics, moduli, and Poisson's ratios) between the basemat and the underlying rock stratum; sensitivity of surface motions due to the effects of depth on degradation in soil stiffness and damping for deep soil sites, degradation sensitivity to soil type (i.e., sand or silt), and effect of Poisson's ratio on vertical soil response; engineering classification and description of materials supporting the structural foundations; data concerning the extent of Seismic Category I excavations and backfills; groundwater conditions relative to foundation stability of safety-related structures; and liquefaction potential including testing methods used in the evaluation. 2.5.4.1 Geologic Features Site-specific information will include geologic features underlying the site. 2.5.4.2 Properties of Underlying Materials State-of-the-art methods used to determine the static and dynamic engineering properties of foundation soils and rocks in the site area will be included in the site-specific information. 2.5.4.3 Relationship of Foundation to Underlying Materials L Plot plans and profiles of site explorations will be included in the site-specific information.
- O
- o ,, ~.,. cm. - . ,.,. u ,
System 80 + oesign controlDocument 2.5.4.4 Soil and Rock Characteristics Results for geophysid investigations performed at the site, including compression and shear wave velocity data frora borings, are included in the site-specific information. 2.5.4.5 Excavation and Mckfill Site-specific information provided will include sources, quantities and engineering properties of borrow materials; compaction requirements; results of field compaction tests; and fill material properties such as moisture content, density, permeability, compressibility, and gradation. 2.5.4.6 Groundwater Conditions Site-specific information will include critical cases of groundwater conditions during plant construction and plant life. Also, soil properties assumed in design will be confirmed for all groundwater conditions. 2.5.4.7 Re=ponse of Soil and Rock to Dynamic Loading Site investigations will determine the effects of prior carthquakes on the soils and rocks in the vicinity of the site (e.g., field seismic surveys, dynamic tests on samples of foundation soils and rock). Information provided will show that design assumptions regarding variation of shear wave velocity and material damping are applicable to the site. 2.f Liquefaction Potential Proven testing methods (e.g., triaxial shear tests) will be used to demonstrate the soils under and adjacent to structural foundations are stable against classical liquefaction. 2.5.4.9 Earthquake Design Basis Refer to Sections 2.5.2.6,2.5.2.7 and Appendix 2B for a summary of the derivation of the Safe Shutdown Earthquakes. Seismic in combination with any other hazards is to be evaluated to assess site materials under dynamic conditions. 2.5.4.10 Bearing Capacity Static analyses of the underlying material supporting the loads of fills, embankments and foundations will be performed to determine stability, deformation and settlement of site materials. The method used to establish the site-specific soil bearing capacity will be described in the site-specific information. 2.5.4.11 Criteria and Design Methods The criteria, analysis techniques, and factors of safety employed in evaluating the stability of the foundations of the plant structures will be included in the site-specific information. 2.5.4.12 Techniques to Improve Subsurface Conditions if it is necessary to improve subsurface conditions, plans, summaries of specifications, and methods of quality control will be described in the site-specific information for all techniques employed. Approved Design Meterini Site Characteristics Page 2.5-10
System 80+ Design ControlDocument o
) 2.5.5 Stability of Slopes ;
No specific assumptions were employed in the evaluation of the System 80+ Standard Design. The stability of all natural and manc.ade slopes, including embankments and dams, that are vital to the safety of the nuclear plant will be included in the site-specific information. 2.5.5.1 Slope Characteristics The site-specific in'ormation will describe the characteristics of the slope by including slope profiles, a discussion of prop'enies of all natural and constructed slopes and embankments, and a description of ' groundwater N r,eepage conditions. 2.5.5.2 Design Criteria and Analyses The site-specific information will present the design criteria and the analytical methods and results which demonstrate design margin for all Seismic Category I slopes. 2.5.5.3 Investigations of Borings, Shafts, Pits, Trenches The site-specific information will present borings and soil tests performed for slope stability studies and dam and dike analyses. O 2.5.5.4 Properties of Borrow Material, Compaction and Excavation Specifications O The site-specific information will describe the excavation, backfill and borrow material for any dams, dikes and embankment slopes. It also provides construction procedures and control of such canhworks. References for Section 2.5
- 1. Idriss, I.M., "Eanhquake Ground Motions - Selection of Control Motion and Development of Generic Soil Sites".
- 2. 10 CFR Parts 50 and 100, " Reactor Site Criteria".
- 3. Standard Review Plan 2.5, " Geology and Seismology", Revision June 1987.
- 4. NUREG/CR-0098, " Development of Criteria for Seismic Review of Selected Nuclear Power Plants", N.M. Newmark, W.J. Hall, May 1978.
- 5. Seed, H.B. and Idriss, l.M. (1970), " Soil Moduli and Damping Factors for Dynamic F esponse Analyses", Report No. EERC 70-10, University of California, Earthquake Eng. Research Center, Berkeley.
f% N Anwooed Design Menerint
- Site Characteristics Pope 2.5-r r
System 80+ Design controlDocumont Table 2.5-1 Digitized Spectral Ordinates for the Envelope of the Free-Field Surface Spectra for the Horizontal Motion, 5% Damping Frequency Gh) Spectral Ordinate at 5% Damping (g) 0.100 0.01000 0.200 0.0500 0.300 0.150 0.400 0.280 0.500 0.510 0.700 0.600 0.950 0.950 1.20 0.950 1.75 1.85 1.90 1.85 2.00 1.95 2.60 1.95 3.10 1.65 4.20 1.65 l 5.00 1.45 11.0 1.45 20.0 0.910 l 30.0 0.910 40.0 0.570 ) 100.0 j 0.570 O l Apprmed Desy nesterial Site Characterhtics Page 2.512
System 80+ ,, Design controlDocument em Table 2.5-2 Digitized Spectral Ordinates for the Envelope of the Free-Field Surface ; Spectra for the Vertical Motion, 5% Damping Frequency (Hz) Spectral Ordinate at 5% Damping (g) 0.100 0.01000 i i 0.0300 l 0.200 0.600 0.150 1.00 0.500 - 1.50 1.63 1 1.80 1.63 , 2.00 2.10 2.50 2.10 l 2.75 1.50 3.00 1.50 l l 3.60 2.05 4.50 2.05 5.50 1.45 7.00 1.45 l 8.00 1.07 10.0 1.07 13.0 1.40 16.0 1.40 38.0 0.650 50.0 0.600 4 100.0 0.600 O Anwevent Design Meterda!= Site Characterianics Pope 2.513
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. Page 7 Abstract ...................................................... 2B-1 . Methodology for Development of Target Power Spectral Densities .................. 2B-1 References . , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ................ 2B-1 Figures 1 ~ 2B-1 Synthetic Acceleration, Velocity and Displacement Time Histories. CMS 1, l Component H1 ...........................................2B-2 2B-2 Synthetic Acceleration. Velocity and Displacement Time Histories, CMS 1, 3 Component H2 . . ......................................... 2B-3 ( 2B-3 Synthetic Acceleration, Velocity and Displacement Time Histories, CMS 1, Vertical Component ............. ................... ....... 2B-4 2B-4 Average Power Spectral Density CMS 1, Components H1 and H2 . . . . . . . . . . . . 2B-5 2B-5 Average Power Spectral Density, CMS 1, Vertical Component . . . . . . . . . . . . . . . 2B-6 l . 2B-6 Synthetic Acceleration Time History, CMS 2, Component H1 . . . . . . . . . . . . . . 2B-7 2B-7 Velocity - Synthetic Time History, CMS 2, Component H1 . . . . . . . . . . . . . . . . 2B-8 2B Displacement - Synthetic Time History, CMS 2, Component H1. ............ 2B-9 2B-9 Average Power Spectral Density, CMS 2, Component H1 . . . . . . . . . . . . . . . . . 2B-10 2B-10 Synthetic Acceleration Time History, CMS 2, Component H2 .............. 2B-11 2B-11 Velocity - Synthetic Time History, CMS 2, Component H2 . . . . . . . . . . . . . . . . 2B-12 28-12 _ Displacement - Synthetic Time History, CMS 2, Component H2 . . . . . . . . . . . . . 2B-13 2B-13 Average Power Spectral Density, CMS 2, Component H2 . . . . . . . . . . . . . . . . . 2B-14 _28-14 Synthetic Acceleration Time History, CMS 2, Vertical Component ..... ..... 2B-15 2B-15 Velocity - Synthetic Time History, CMS 2, Vertical Component . . . . . . . . . . . . . . 2B-16 2B Displacement - Synthetic Time History, CMS 2, Vertical Component . . . . . . . . . . 2B-17 ) l 2B-17 Average Power Spectral Density, CMS 2, Vertical Component . . . . . . . . . . . . . . 2B-18 2B 18 Synthetic Acceleration, Velocity and Displacement Time Histories, CMS 3, Component H1 ........................................... 2B-19
- 2B-19 Synthetic Accelerat.on, Velocity and Displacement Time Histories, CMS 3, Component H2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -. . . . . . . . . . . . 2B-20 2B Synthetic ~ Acceleration, Velocity and Displacement Time Histories, CMS 3, Vertical Component ........................................ 2B-21 3 2B-21 Average Power Spectral Density, CMS 3, Component H1 . . . . . . . . . . . . . . . . . 2B-22 2B-22 Average Power Spectral Density, CMS 3, Component H2 . . . . ............ 2B-23 ,
\ 2B-23 Average Power Spectral Density, CMS 3, Vertical Component . . . . . . . . . . . . . . 2B-24 !
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Sy, tem 80+ Design controlDocument n b Abstract The synthetic time histories generated to represent the horizontal components H1, H2 and the vertical component of control motions CMS 1, CMS 2 and CMS 3 are presented in this Appendix. The acceleration, velocity and displacement time histories of control motion CMS 1 are shown in Figures 2B-1 through 2B-3. The average Power Spectral Densities (PSD) of the CMSI synthetic time histories and their respective target PSDs are shown in Figures 2B-4 and 2B-5. The acceleration, velocity and displacement time histories of control motion CMS 2 are shown in Figures 2B-6 through 2B-8, 2B-10 through 2B-12, and 28-14 through 2B-16. The Power Spectral Densities (PSD) of the CMS 2 synthetic time histories and their respective target PSDs are shown in Figures 2B-9, 2B-13 and 2B-17. The acceleration, velocity and displacement time histories of control motion CMS 3 are shown in Figures 2B-18 through 2B-20. The average Power Spectral Densities (PSD) of the CMS 3 synthetic time histories and their respective target PSDs are shown in Figures 2B-21 through 2B-23. The selection process for CMS 1, CMS 2 and CMS 3 is given in Section 2.5.2.5.1. The average PSDs for the CMS 1, CMS 2, and CMS 3 control time histories were developed using the procedure described in SRP, Section 3.7.1 Appendix A. The target PSD for CMSI (horizontal) motion (Reg. Guide 1.60 horizontal) was obtained directly from the SRP, Section 3.7.1, Appendix A. The !q g methodology for the development of the target PSDs for CMSI (vertical), CMS 2 and CMS 3 is described below. Methodology for Development of Target Power Spectral Densities The development of target PSDs for CMSI (vertical), and the rock outcrop motions CMS 2 and CMS 3 was performed using principles of Random Vibration Theory (RVT). Details of this method as well as the mathematical formulation are described in References 1 and 2. The basic approach is that the target PSD is developed by an iterative process. At each step of the iteration, the PSD is refined to produce a spectrum that closely matches the target response spectrum. Adjustments to the PSD are made at the frequency ranges that do not produce a close spectral match, and the final target PSD is obtained when the desired spectrum convergence is achieved. The minimum check is set at 80% of the target PSD, consistent with SRP guidelines. The development of the target PSDs was performed using the 2 % damped spectrum as the target spectmm for each control motion. References for Appendix 2B
- 1. Boore, D.M., " Stochastic Simulation of High-Frequency Ground Motions Based on Seismological Models of the Radiated Spectra", Bulletin of the Seismological Society of America, Volume 73, Number 6, pp. I865-1894.
- 2. Boore, D. M., and Joyner, W. B., "A Note on the Use of Random Vibration Theory to Predict
(] V Peak Amplitudes of Transient Signals", Bulletin of the Seismological Society of America, Volume 74, Number 5, pp. 2035-2039. October 1984. Approwd Desiger Material- Site Characteristics Page 281
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i system 80+ oestan contratoocument - (7 Appendix 2C . V i Strain-Compatible Modulus and Damping Values . i r Contents Page Abstract . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2C-1 Tables 2C-1 Strain Compatible Modulus and Damping Values, Horizontal Motions; Peak Acceleration = 0.3 g, Cases A-1 and B-1 ........................2C-1 Strain Compatible Modulus and Damping Values, Horizontal Motions; 2C-2 Peak Acceleration = 0.3 g, Cases B-2 and B-3 . . . . . . . . . . . . . . . . . . . . . . . . 2C-2 2C-3 Strain Compatible Modulus and Damping Values, Horizontal Motions; Peak Acceleration = 0.3 g, Cases B-4 and C . . . . . . . . . . . . . . . . . . . . . . . . 2C-3
/Q 2C-4 Strain Compatible Modulus and Damping Values, Horizontal Motions;
- C/ Peak Acceleration = 0.3 g, Cases C-2 and C-3 ........................2C-4 > I 2C-5 Strain Compatible Modulus and Damping Values, Horizontal Motions; Pe* Acceleration = 0.3 g, Case D-1 ................... .......... 2C-5 ; 2C-6 Strain Compatible Modulus and Damping Values, Horizontal Motions; Peak Acceleration = 0.3 g, Cases B-1.5 and B-3.5 . . . . . . . . . . . . . . . . . . . . . . 2C-6 2C-7 Strain Compatible Modulus and Damping Values, Horizontal Motions; Peak Acceleration = 0.3 g, Case C-1.5 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2C-7 2C-8 Modulus and Damping Values, Venical Motion; Peak Acceleration = 0.2 g, Cases A- 1 and B- 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2C-8 2C-9 Modulus and Damping Values, Venical Motion; Peak Acceleration = 0.2 g, Cases B-2 and B-3 ..... ..... .... .......................... 2C-9 2C-10 Modulus and Damping Values, Vertical Motion; Peak Acceleration = 0.2 g, Cases B-4 and C-1 ....... .................................. 2C-10 2C-Il Modulus and Damping Values, Vertical Motion; Peak Acceleration = 0.2 g, . Cases C-2 and C-3 . . . . . . . . . . . . . . . . . . . . ..................... 2C-11 2C-12 Modulus and Damping Values, Vertical Motion; Peak Acceleration = 0.2 g, Case D-1 ................................................ 2C-12 2C-13 Modulus and Damping Values, Vertical Motion; Peak Acceleration = 0.2 g. Cases B- 1.S and B-3.5 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2C-13 2C-14 Modulus and Damping Values, Vertical Motion; Peak Acceleration = 0.2 g, Case C- 1.5 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2C-14 i ( .- ()- Amme oecon unww su cnueewieacs roue s , n
System 80+ Design controlDocument -3 ] Abstract The strain-compatible modulus and damping values obtained using component H1 are presented in Tables 2C-1 through 2C-7 for all the cases considered in this project. Identical values of modulus and damping values were used in conjunction with component H2. For the vertical motions, the constrained moduli were obtained based on the appropriate strain-compatible shear moduli in Tables 2C-1 through 2C-7 and a Poisson's ratio of 0.4. The damping values for the vertical motions were considered to be approximately 1/3 those listed in Tables 2C-1 through 2C-7. The actual constrained moduli and damping values used with the vertical component are listed in Tables 2C-8 through 2C-14. , Table 2C-1 Strain Compatible Modulus and Damping Values, Horizontal Motions; Peak Acceleration = 0.3 g, Cases A-1 and B-1 Case A - 1 , No. Depth Range Avg. Depth Shear Modulus Damping Ratio (ft.) (ft.) (ksf) , 1 0 to 5 2.5 12,215 0.008 /n \ 0.012 Q :' 3 5 to 10 10 to 20 7.5 15.0 13,243 13,719 0.017 4 20 to 30 25.0 14,211 0.022 j 5 30 to 40 35.0 14,740 0.025 ) 6 40 to 52 46.0 15,437 0.027 7 Below 52 Base 97,000 0.020 l Case B - 1 ~ l No. Depth Range Arg. Depth Shear Modulus Damping Ratio (ft) (ft) (ksf) 1 0 to 5 2.5 12,218 0.008
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System 80+ Design ControlDocument Table 2C-2 Strain Compatible Modulus and Damping Values, Horizontal Motions; Peak Acceleration = 0.3 g, Cases B-2 and B-3 h case B - 2 No. Depth Range Avg. Depth Shear Modulus Damping Ratio (ft) (ft) (ksf) 1 0 to 5 2.5 3,852 0.013 2 5 to 10 7.5 3,677 0.025 3 10 to 20 15.0 3,397 0.039 4 20 to 30 25.0 3,252 0.048 5 30 to 40 35.0 3,203 0.055 6 40 to 52 46.0 3,208 0.062 7 52 to 60 56.0 3,207 0.067 8 60 to 80 70.0 3,254 0.072 9 80 to 100 90.0 3,468 0.075 10 Below 100 Base 97,000 0.020 Case B - 3 No. Depth Range Avg. Depth Shear Modulus Damping Ratio (ft) (ft) (ksf) 1 0 to 5 2.5 906 0.021 2 5 to 10 7.5 775 0.040 3 10 to 20 15.0 635 0.064 4 20 to 30 25.0 523 0.088 5 30 to 40 35.0 468 0.101 6 40 to 52 46.0 468 0.107 7 52 to 60 56.0 500 0.1 % 8 60 to 80 70,0 548 0.104 9 80 to 100 90.0 633 0.098 10 Below 100 Base 97,000 0.020 O Approved Design Materint Site Characteristics Pope 2C-2
System 80+ Design Control. Document
! Table 2C-3 Strain Compatible Modulus and Damping Values, Horizontal Motions; Peak Acceleration = 0.3 g, Cases B-4 and C-1 Case B - 4 No. Depth Range Avg. Depth Shear Modulus Damping Ratio (ft) (ft) (ksf) 1 0 to 5 2.5 869 0.025 2 5 to 10 7.5 687 0.050 3 10 to 20 15.0 543 0.079 4 20 to 30 25.0 450 0.099 5 30 to 40 35.0 419 0.112 6 40 to 52 46.0 372 0.128 7 52 to 60 56.0 17,508 0.019 8 60 to 80 70.0 18,474 0.021 9 80 to 100 90.0 19,903 0.022 10 Below 100 Base 97,000 0.020 i
\ Case C - 1 No. Depth Range Avg. Depth Shear Modulus Damping Ratio (ft) (ft) (ksi) 1 0 to 5 2.5 12,161 0.008 2 5 to 10 7.5 13,204 0.013 3 10 to 20 15.0 13,520 0.018 4 20 to 30 25.0 14,024 0.023 5 30 to 40 35.0 14,558 0.026 6 40 to 52 46.0 15,150 0.029 7 52 to 60 56.0 15,548 0.031 8 60 to 80 70.0 16,249 0.034 9 80 to 100 90.0 17,738 0.034 10 100 to 120 110.0 20,709 0.032 11 120 to 140 130.0 19,927 0.036 12 140 to 160 150.0 21,602 0.036 0.036
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1 System 80 + Design ControlDocument Table 2C-4 Strain Compatible Modulus and Damping Values, Horizontal Motions; Peak Acceleration = 0.3 g, Cases C-2 and C-3 h i Case C - 2 No. Depth Range Avg. Depth Shear Modulus Damping Ratio (ft) (f0 (ks0 1 0 to 5 2.5 912 0.020 5 to 10 7.5 786 0.039 2 10 to 20 15.0 650 0.062 3 4 20 to 30 25.0 545 0.085 30 to 40 35.0 517 0.093 5 40 to 52 46.0 517 0.097 6 7 52 to 60 56.0 530 0.100 60 to 80 70.0 587 0.097 8 80 to 100 90.0 622 0.100 9 100 to 120 110.0 898 0.083 10 120 to 140 130.0 856 0.087 11 140 to 160 150.0 951 0.085 12 160 to 180 170.0 1,029 0.085 13 180 to 200 190.0 919 0.093 14 15 Below 200' Base 97,000 0.020 Case C - 3 No. Depth Range Avg. Depth Shear Modulus Damping Ratio (f0 (ft) (ks0 1 0 to 5 2.5 887 0.023 2 5 to 10 7.5 728 0.045 3 10 to 20 15.0 579 0.073 4 20 to 30 25.0 494 0.092 5 30 to 40 35.0 487 0.C97 6 40 to 52 46.0 470 0.107 7 52 to 60 56.0 496 0307 8 60 to 80 70.0 534 0.107 9 80 to 100 90.0 602 0.104 10 100 to 120 110.0 24,260 0.015 11 120 to 140 130.0 23,506 0.019 12 140 to 160 150.0 25,028 0.021 13 160 to 180 170.0 26,775 0.022 ' 14 180 to 200 190.0 26,340 0.024 15 Below 200 Base 97,000 0.020 l Approwwd Desipre Atatorial. Site Characterktics Page 2C 4 i
System 80+ o sign controlDocument O Table 2C-5 Strain Compatible Modulus and Damping Values, Horizontal
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No. Depth Range Avg. Depth Shear Modulus Damping Ratio (ft) (ft) (ksf) 1 0 to 5 2.5 905 0.021 2 5 to 10 7.5 767 0.041 . 3 10 to 20 15.0 618 0.067 4 20 to 30 25.0 504 0.091 5 30 to 40 35.0 481 0.098 6 40 to 52 46.0 509 0.098 7 52 to 60 56.0 529 0.101 j 8 60 to 80 70.0 585 0.098 9 80 to 100 90.0 602 0.104 10 100 to 120 110.0 888 0.084 120 to 140 130.0 847 0.08'; 11 12 140 to 160 150.0 916 0.088 160 to 180 170.0 976 0.089 A 13 14 180 to 200 190.0 855 0.098 15 200 to 220 210.0 31,726 0.016 16 220 to 240 230.0 30,588 0.020 17 240 to 260 250.0 30,793 0.024 18 260 to 280 270.0 31,840 0.024 19 280 to 300 290.0 31,597 0.025 20 Below 300 Base 97,000 0.020 I 1 1
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System 80+ Design controlDocument Table 2C-6 Strain Compatible Modulus and Damping Values, Horizontal Motions; Peak Acceleration = 0.3 g, Cases B-1.5 and B-3.5 Case B - 1.5 No. Depth Range Avg. Depth Shear Modulus Damping Ratio (ft) (ft) (ks0 1 0 to 5 2.5 7,658 0.010 2 5 to 10 7.5 7,595 0.019 3 10 to 20 15.0 7,600 0.027 4 20 to 30 25.0 7,454 0.036 30 to 40 35.0 7,434 0.041 5 6 40 to 52 46.0 7,522 0.045 7 52 to 60 56.0 7,745 0.047 8 60 to 80 70.0 8,287 0.047 9 80 to 100 90.0 8,901 0.049 Below 100 Base 97,000 0.020 10 _ Case B - 3.5 No. Depth Range Avg. Depth Shear Modulus Damping Ratio (ft) (ft) (ks0 1 0to5 2.5 1,246 0.026 2 5 to 10 7.5 982 0.050 3 10 to 20 15.0 7'O 0.080 4 20 to 30 25.0 a88 0.109 5 30 to 40 35.0 604 0.112 6 40 to 52 46.0 655 0.110 i 7 52 to 60 56.0 6,568 0.036 ) 8 60 to 80 70.0 6,850 0.038 ! 9 80 to 100 90.0 7,211 0.041 10 Below 100 Base 97,000 0.020 1 i i l l i l l Annoved Des &n MaterW Site Characteristics Psge 2C-6 l
i System 80+ Deslan controlDocument (M Q Table 2C-7 Strain Compatible Modulus and Damping Values, Horizontal Motions; Peak Acceleration = 0.3 g, Case C-1.5 Case C - 1.5 No. Depth Range Avg. Depth Shear Modulus Damping Ratio (ft) (ft) (ksf) 1 0 to 5 2.5 3,862 0.012 2 5 to 10 7.5 3,726 0.024 3 10 to 20 15.0 3,583 0.034 4 20 to 30 25.0 3,478 0.043 5 30 to 40 35.0 3,504 0.047 6 40 to 52 46.0 3,653 0.049 7 52 to 60 56.0 3,700 0.052 8 60 to 80 70.0 3,851 0.056 9 80 to 100 90.0 4,191 0.056 10 100 to 120 110.0 4,953 0.053 11 120 to 140 130.0 4,640 0.060 12 140 to 160 150.0 5,157 0.057 g 13 160 to 180 170.0 5,653 0.056 d 14 180 to 200 190.0 5,595 97,000 0.057 0.020 15 Below 200 Base I i i n w Aparewd Design neareriel. Site Characterkaics Page 2C.7 l
T System 80+ oesign controlDocument Table 2C-8 Modulus and Damping Values, Vertical Motion; Peak Acceleration = 0.2 g, Cases A-1 and B-1 Case A - 1 No. Depth Range Avg. Depth Constrained Modulilll Damping Ratiot21 (ft) (ft) (ksf) 0 to 5 2.5 73,290 0.0027 1 5 to 10 7.5 79,458 0.0040 2 10 to 20 15.0 82,314 0.0057 3 4 20 to 30 25.0 85,266 0.0073 30 to 40 35.0 88,440 0.0083 5 6 40 to 52 46.0 92,622 0.0090 7 Below 52 Base 302.000 0.0067 Case B - ! No. Depth Range Avg. Depth Constrained Moduli Damping Ratio (ft) (ft) (ksf) 1 0 to 5 2.5 73,308 0.0027 2 5 to 10 7.5 79,362 0.0040 3 10 to 20 15.0 81,648 0.0060 4 20 to 30 25.0 84,546 0.0077 5 30 to 40 35.0 87,780 0.0087 6 40 to 52 46.0 91,974 0.0090 7 52 to 60 56.0 %,156 0.0097 8 60 to 80 70.0 99,966 0.0103 9 80 to 100 90.0 106,950 0.0110 10 Below 100 Base 302.000 0.0067 I1 Constrained moduli are obtained using the corresponding strain-compatible shear moduli for the cases listed in Tables 2C 1 through 2C-7, and a Poisson's ratio of 0.4. [2] Damping ratios are approximately 1/3 the corresponding strain-compatible damping ratios listed in these tables. Approved Des}grr Atatorial. Site Characteristics page 2c.g
System 80+ Deslan comrolDocument Table 2C-9 Modulus and Damping Values, Vertical Motion; Peak Acceleration == 0.2 g, Cases B-2 and B-3 Case B - 2 , No. Depth Range Avg. Depth Constrained Modulill3 Damping Ratiot2) (ft) (ft) (ksf) 1 0 to 5 2.5 23,112 0.0043 2 5 to 10 7.5 22,062 0.0083 3 10 to 20 15.0 20,382 0.0130 4 20 to 30 25.0 19,512 0.0160 5 30 to 40 35.0 19,218 0.0183 6 40 to 52 46.0 19,248 0.0207 7 52 to 60 56.0 19,242 0.0223 8 60 to 80 70.0 19.524 0.0240 9 G0 to 100 90.0 20,808 0.0250 10 Below 100 Base 302,000 0.0067 Case B - 3 No. Depth Range Avg. Depth Constrained Moduli Damping Ratio (ft) (ft) (ksf) 1 0 to 5 2.5 5,436 0.0070 2 5 to 10 7.5 4,650 0.0133 3 10 to 20 15.0 3,810 0.0213 4 20 to 30 25.0 3,138 0.0293 5 30 to 40 35.0 2,808 0.0337 6 40 to 52 46.0 2,808 0.0357 7 52 to 60 56.0 3,000 0.0353 8 60 to 80 70.0 3,288 0.0347 9 80 to 100 90.0 3,798 0.0327 10 Below 100 Base 302,000 0.0067 D I'l Constrained moduli are obtained using the corresponding strain-compatible shear moduli for the cases listed in i_ Tables 2C-1 through 2C-7, and a Poisson's ratio of 0.4. l (21 Damping ratios are approximately 1/3 the corresponding strain-compatible damping ratios listed in these tables.
. Approwed Dessper 30etersel* Site Cluerectorietics Page 2C-9 l 1
System 80+ oesign controloocument Table 2C-10 Modulus and Damping Values, Vertical Motion; Peak Acceleration = 0.2 g, Cases B-4 and C-1 Case B - 4 No. Depth Range Avg. Depth Constrained Modulitt) Damping Ratiot2) (ft) (ft) (ksf) 1 0 to 5 2.5 5,214 0.0083 2 5 to 10 7.5 4,122 0.0167 3 10 to 20 15.0 3,258 0.0263 4 20 to 30 25.0 2,700 0.0330 5 30 to 40 35.0 2,514 0.0373 6 40 to 52 46.0 2,232 0.0427 7 52 to 60 56.0 105,048 0.0063 8 60 to 80 70.0 110,844 0.0070 , 9 80 to 100 90.0 119,418 0.0073 10 Below 100 Base 302,000 0.0067 Case C - 1 No. Depth Range Avg. Depth Constrained Modull Damping Ratio (ft) (ft) (ksi) 1 0 to 5 2.5 72,966 0.0027 2 5 to 10 7.5 79,224 0.0043 3 10 to 20 15.0 81,120 0.0060 4 20 to 30 25.0 84,144 0.0077 5 30 to 40 35.0 87,348 0.0087 6 40 to 52 46.0 90,900 0.0097 7 52 to 60 56.0 93,288 0.0103 8 60 to 80 70.0 97,494 0.0113 9 80 to 100 90.0 106,428 0.0113 10 100 to 120 110.0 124,254 0.0107 l 11 120 to i40 130.0 119,562 0.0120 l 12 140 to 160 150.0 129,612 0.0120 13 160 to 180 170.0 139,788 0.0120 14 180 to 200 190.0 136,872 0.0127 15 Below 200 Base 302,000 0.0067 i I Ill Constrained moduli are obtained using the corresponding strain-compatible shear moduli for the cases listed in 4 1 Tables 2C-1 through 2C-7, and a Poisson's ratio of 0.4 (2) Damping ratios are appt nimately 1/3 the corresponding strain <ompatible damping ratios listed in these tables. Approved Design Material- Site Characterisks Page 2C.10
System 80+ oesign controloocument Table 2C-11 Modulus and Damping Values, Vertical Motion; Peak Acceleration = 0.2 g, Cases C-2 and C-3 Case c - 2 No. Depth Range Avg. Depth Constrained ModuliU1 Damping Ratiot2) (ft) (ft) (ksf) 1 0 to 5 2. 5 5,472 0.0067 2 5 to 10 7.2 4,716 0.0130 3 10 to 20 15.0 3,900 0.0207 4 20 to 30 25.0 3,270 0.0283 5 30 to 40 35.0 3,102 0.0310 6 40 to 52 46.0 3,102 0.0323 7 52 to 60 56.0 3,180 0.0333 8 60 to 80 70.0 3,522 0.0323 9 80 to 100 90.0 3,732 0.0333 10 100 to 120 110.0 5,388 0.0277 11 120 to i40 130.0 5,136 0.0290 12 140 to 160 150.0 5,706 0.0283 13 160 to 180 170.0 6,174 0.0283 14 180 to 200 190.0 5,514 0.0310 15 Below 200 Base 302,000 0.0067 O Case C - 3 No. Depth Range Avg. Depth Constrained Moduli Damping Ratio (ft) (ft) (ksf) 1 0 to 5 2.5 5,322 0.0077 2 5 to 10 7.5 4,368 0.0150 3 10 to 20 15.0 3,474 0.0243 4 20 to 30 25.0 2,964 0.0307 5 30 to 40 35.0 2,922 0.0323 6 40 to 52 46.0 2,820 0.0357 ; 7 52 to 60 56.0 2.976 0.0357 l 8 60 to 80 70.0 3,204 0.0357 i 9 80 to 100 90.0 3,612 0.0347 10 100 to 120 110.0 145,560 0.0050 11 120 to 140 130.0 141,036 0.0063 l 12 140 to 160 150.0 150,168 0.0070 13 160 to 180 170.0 160,650 0.0073 14 180 to 200 190.0 158,040 0.0080 15 Below 200 Base 302,000 0.0067 U3 Constrained moduli are obtained using the corresponding strain <ompatible shear moduli for the cases listed in (A) b Tables 2C 1 through 2C 7, and a Poisson's ratio of 0.4. [2] Damping ratios are approximately 1/3 the corresponding strain-compatible damping ratios listed in these tables. 4prend Design nienwiel Sne Chwecewis6cs Pope 2C-11 l l
System 80+ Design ControlDocument Table 2C-12 Modulus and Damping Values, Vertical Motion; Peak Acceleration = 0.2 g, Case D-1 Case D - 1 Depth Range Avg. Depth Constrained Modulill! Damping Ratior2; No. (ft) (ft) (ksf) 0 to 5 2.5 5.430 0.0070 1 5 to 10 7.5 4,602 0.0137 2 10 to 20 15.0 3,708 0.0223 3 4 20 to 30 25.0 3,024 0.0303 35.0 2,886 0.0327 5 30 to 40 40 to 52 46.0 3,054 0.0327 6 52 to 60 56.0 3,174 0.0337 7 60 to 80 70.0 3,510 0.0327 8 80 to 100 90.0 3,612 0.0347 9 100 to 120 110.0 5,328 0.0280 10 120 to 140 130.0 5,082 0.0292 11 140 to 160 150.0 5,4 % 0.0293 12 13 160 o I80 170.0 5,856 0.0297 180 to 200' 190.0 5,130 0.0327 14 200 to 220 210.0 190,356 0.0053 15 16 220 to 240 230.0 183,528 0.0067 17 240 to 260 250.0 184,758 0.0080 18 260 to 280 270.0 191,040 0.0080 19 280 to 300 290.0 189,582 0.0083 20 Below 300 Base 302,000 0.0067 til Constrained moduli are obtained using the corresponding strain-compatible shear moduli for the cases listed in Tables 2C 1 through 2C-7, and a Poisson's ratio of 0.4. 121 Damping ratios are approximately 1/3 the corresponding strain-compatible damping ratios listed in these tables. Approved Design Materiel. Sete Characterisk Page 2C-12
System,8,0 + oesign controlDocument l , Table 2C-13 Modulus and Damping Values, Vertical Motion; Peak Acceleration = 0.2 g, Cases B-1.5 and B-3.5 Case B - 1.5 ; No. Depth Range Avg. Depth Constrained Moduliul Damping Ratiot21 (ft) (ft)- (ksf) 1 0 to 5 2.5 73,308 0.0033 2 5 to 10 7.5 79,362 0.0063 3 10 to 20 15.0 81,648 0.0090 4 20 to 30 25.0 84,546 0.0120 5 30 to 40 35.0 87,780 0.0137 , 6 40 to 52 46.0 91,974 0.0150 7 52 to 60 56.0 96,156 0.0157 8 60 to 80 70.0 99,966 0.0157 , 9 80 to 100 90.0 106,950 0.0163 10 Below 100 Base 302,000 0.0067 l l Case B - 3.5 l No. Depth Range Avg. Depth Constrained Moduli Damping Ratio V (ft) (ft) (ksf) 1 0 to 5 2.5 7,476 0.0087 2 5 to 10 7.5 5,892 0.0167 3 10 to 20 15.0 4,620 0.0267 ; i 4 20 to 30 25.0 3,528 0.0363 5 30 to 40 35.0 3,624 0.0373 6 40 to 52 46.0 3,930 0.0367 7 52 to 60 56.0 39,408 0.0120 8 60 to 80 70.0 41,100 0.0127 9 80 to 100 90.0 43,266 0.0137 10 Below 100 Base 302,000 0.0067 ; l l
,O Ul Constrained moduli are obtained using the corresponding strain <ompatible shear moduli for the cases listed in \,_ Tables 2C-1 through 2C-7, and a Poisson's ratio of 0.4.
Ul Damping ratios are approximately 1/3 the corresponding strain-compatible damping ratios listed in these tables. Anproved Deegro n0neerial. Site Characteristks Page 2C.13
i System 80+ Design controlDocument Table 2C-14 Modulus and Damping Values, Vertical Motion; Peak Acceleration = 0.2 g, Case C-1.5 Case C - 1.5 No. Depth Range Avg. Depth Constrained Modullu) Damping Ratiot2) (ft) (ft) (ksf) 1 0 to 5 2.5 23,172 0.0040 2 5 to 10 7.5 22,256 0.0080 3 10 to 20 15.0 21,498 0.0113 4 20 to 30 25.0 20,868 0.0143 5 30 to 40 35.0 21,024 0.0157 6 40 to 52 46.0 21,918 0.0163 7 52 to 60 56.0 22,200 0.0173 8 60 to 80 70.0 23,106 0.0187 9 80 to 100 90.0 25,146 0.0187 10 100 to 120 110.0 29,718 0.0177 11 120 to 140 130.0 27,840 0.0200 12 140 to 160 150.0 30,942 04190 13 160 to 180 170.0 33,918 0.0187 14 180 to 200 190.0 33,570 0.0190 15 Below 200 Base 302,000 0.0067 UI Constrained moduli are obtained using the corresponding strain-compatible shear moduli for the cases listed in Tables 2C-1 through 2C 7, and a Poisson's ratio of 0.4, (21 Damping ratios are approximately 1/3 the corresponding strain-compatible damping ratios listed in these tables. Alnwend Design Materniel- Site Characteristics Page 2C-14
Sy7 tem 80+ Design controlDocument
;( ,9) Effective Page' Listing Chapter 3
- Pages Date Pages Date i,ii 1/97 3.5-1 through 3.5-8 Original lii, iv 2/95 v Original 3.6-1 through 3.6-33 Original vi, vii 11/% 3.6-34, 3.6-35 11/96 viii 2/95 3.6-36 through 3.6-41 Original lx Original 3.6-42 11/%
3.6-43 Original 3.1-1 through 3.1-5 Original 3.6-44 11/% 3.1-6 11/% 3.6-45 through 3.6-53 Original 3.1-7, 3.1-8 Original 3.1-9 11/% 3.7 1 Original , 3.1-10 through 3.1-17 Original 3.7-2 11/% 3.1-18 2/95 3.7-3 through 3.7-17 Original ; 3.1-19 11/96 3.7-18 2/95 ,_ 3.1-20 through 3.1-25 Original 3.7-19 through 3.7-27 Original ( 3.1-26 2/95 3.7-28 11/% v- Original 3.1-27 Original 3.7-29 3.1-28 11/% 3.7-30 11/% ; 3.1-29, 3.1-30 Original 3.7-31 Original 3.7-32 11/96 3.2-1 2/95 3.7-33 through 3.7-37 Original 3.2-2,3.2-3 Original 3.7-38 11/96 3.2-4 11/96 3.7-39 through 3.7-69 Original 3.2-5 through 3.2-12 2/95 3.7-70 11/96 3.2-13 11/% 3.7-71 Original 3.2-14 through 3.2-17 2/95 3.2-18 11/% 3.8-1 through 3.8-7 Original 3.2-19, 3.2 20 2/95 3.8-8 2/95 3.2-21 11/% 3.8-9 through 3.8-23 Original 3.2-22 through 3.2-31 Original 3.8-24 through 3.8-36b 1/97 3.2-32 11/96 3.8-37 through 3.8-79 Original 3.2-33 through 3.2-48 Original 3.9-1 through 3.9-17 Original 3.3-1 through 3.3-3 Original 3.9-18, 3.9-19 11/% 3.9-20 through 3.9-56 Original 3.4-1 through 3.4-3 Original 3.9-57 11/96 Normd w aneurw Decem er ssc msn renei
Syntem 80+ oesign controlDocument Effective Page Listing (Cont'd.) Chapter 3 h Pages Date 3.9-58 through 3.9-62 Original 3.9-63 11/96 3.9-64 through 3.9-67 Original 3.9-68 through 3.9-71 11/96 3.9-72 Original 3.9-73 11/96 3.9-74 through 3.9-86 Original 3.9-87 2/95 3.9-88 through 3.9-206 Original 3.101 through 3.10-11 Original 3.11-1 through 3.11-15 Original O O AAnvend Des # Atatene! Design of SSC (1/p7) page g
System 80+ Design ControlDocument , t l v) Chapter 3 Contents Page ! 3.0 Design of Structures, Components, Equipment, and Systems . . . . . . . . . . . 3.1-1 3.1 Conformance with NRC General Design Criteria . . . . . . . . . . . . . . . . . . . 3.1-1 , 3.1.1. Criterion 1 - Quality Standards and Records ............. .......... 3.1-1 ; 3.1.2 Criterion 2 - Design Bases for Protection Against Natural Phenomena .. ..... 3.1-1 l-3.1.3 Criterion 3 - Fire Protection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1-2 3.1.4 Criterion 4 - Environmental and Missile Design Bases . . . . . . . . . . . . . . . . . 3.1-2 ; 3.1.5 . Criterion 5 - Sharing of Structures, Systems, and Components . . . . . . . . . . . . 3.1-3 3.1.6 Criterion 10 - Reactor Design . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1-3 3.1.7 Criterion 11 - Reactor inherent Protection '. . . . . . . . . . . . . . . . . . . . . . . . . 3.1-4 , 3.1.8 Criterion 12 - Suppression of Reactor Power Oscillations . . . . . . . . . . . . ... 3.1-4 3.1.9 Criterion 13 - Instrumentation and Control . . . . . . . . . . . . . . . . . . . . . . . . 3.1-4 3.1.10 Criterion 14 - Reactor Coolant Pressure Boundary . . . . . . . ..... . ...... 3.1-6 3.1.11 Criterion 15 - Reactor Coolant System Design . . . . . . . . ....... .... . 3.1-7 3.1.12 Criterion 16 - Con *.ainment Design . . . . . . . . . . . . . . . . . . .......... 3.1-7 3.1.13 Criterion 17 - Electrical Power Systems . . . . . . . . . . . . . . . . . . . . . . . . .. 3.1-8 3.1.14 Criterion 18 - Inspection and Testing of Electrical Power Systems . . . . . . . . . . 3.1-9 i 3.1.15 Criterion 19 - Control Room . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1-9 3.1.16 Criterion 20 - Protection System Functions . . . . . . . . . . . . . . . . ....... 3.1-10 /O 3.1.17 Criterion 21 - Protection System Reliability and Testability . . . . . ........ 3.1-11 V 3.1.18 Criterion 22 - Protection System Independence . . . . . . . . . . . . . . . . . . . . . 3.1-11 3.1.19 Criterion 23 - Protection System Failure Modes . . . . . . . . . . . . . . . . . . . . 3.1-12 3.1.20 Criterion 24 - Separation of Protection and Control Systems . . . . . . . . . . . . 3.1- 12 3.1.21 Criterion 25 - Protection System Requirements for Reactivity l Control Malfunctions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1-13 3.1.22 Criterion 26 - Reactivity Control System Redundancy and Capability . . . . . . . 3.1-13 3.1.23 Criterion 27 - Combined Reactivity Control Systems Capability . . . .... . 3.1-14 3.1.24 Criterion 28 - Reactivity Limits . . . ... .... ..... . .. ....... .. .. 3.1-14 3.1.25 Criterion 29 - Protection Against Anticipated Operational Occurrences . . . . . . 3.1-15 3.1.26 Criterion 30 - Quality of Reactor Coolant Pressure Boundary . . . . ... .... 3.1-15 ' 3.1.27 Criterion 31 - Fracture Prevention of Reactor Coolant Pressure Boundary . . . . 3.1-16 3.1.28 Criterion 32 - Inspection of Reactor Coolant Pressure Boundary . . . . . . . . . . 3.1-17 3.1.29 Criterion 33 - Reactor Coolant Makeup . . . . . . . . . . . . . . . . . . . . . .. . 3.1-17 3.1.30 Criterion 34 - Residual Heat Removal ................ ......... 3.1-18 l 3.1.31 Criterion 35 - Emergency Core Cooling . . . . . . . . . . . . . . ........... 3.1-19 3.1.32 Criterion 36 - Inspection of Emergency Core Cooling System .... . . . . . 3.1-20 3.1.33 Criterion 37.- Testing of Emergency Core Cooling System . . . ...... ... 3.1-20 3.1.34 Criterion 38 - Containment Heat Removal . . . . . . . . . . . .. . . . . . . . . . . . . 3.1-20 3.1.35 Criterion 39 - Inspection of Containment Heat Removal System . . . . . . . . .. 3.1-21 , 3.1.36 Criterion 40 - Testing of Containment Heat Removal System . . . . . . . . . . . . 3.1-21 ; 3.1.37 Criterion 41 - Containment Atmosphere Cleanup ........ .......... 3.1-22 l 3.1.38 Criterion 42 - Inspection of Containment Atmosphere Cleanup Systems . . . . . 3.1-22
'3.1.39 Criterion 43 - Testing of Containment Atmosphere Cleanup Systems . 3.1-23 n) 3.1.40' Criterion 44 - Cooling Water . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1-23 v.
Anvevent Design n0etersel . Dneign of SSC 42/9 5) Pege G
System 80+ oesign ControlDocument Chapter 3 Contents (Cont'd.) $ Page 3.1.41 Criterion 45 - Inspection of Cooling Water System .... .. .......... 3.1-24 3.1.42 Criterion 46 - Testing of Cooling Water System . . . . . .. .. 3.1-24 3.1.43 Criterion 50 - Containment Design Basis ...... ......... . ...... 3.1-24 3.1.44 Criterion 51 - Fracture Prevention of Containment Pressure Boundary . . . . . . 3.1-25 3.1.45 Criterion 52 - Capability for Containment Leakage Rate Testing . . . . . ... 3.1-25 3.1.46 Criterion 53 - Provisions for Containment Testing and Inspection . . . . . . . .. 3.1-26 3.1.47 Criterion 54 - Piping Systems Penetrating Containment . . . . ....... .... 3.1-26 3.1.48 Criterion 55 - Reactor Coolant Pressure Boundary Penetrating Containment . . 3.1-27 3.1.49 Criterion 56 - Primary Containment Isolation . . . . . .. . .... ..... 3.1-28 3.1.50 Criterion 57 - Closed System Isolation Valves . . . . . . . . . ... .... 3.1-28 3.1.51 Criterion 60 - Control of Releases of Radioactive Material to the Environment . . . . . . . . ....... ... .. ..... ... ....... 3.1-28 3.1.52 Criterion 61 - Fuel Storage and Handling and Radioactivity Control . . . ..... 3.1-29 3.1.53 Criterion 62 - Prevention of Criticality in Fuel Storage and llandling . . ..... 3.1-29 3.1.54 Criterion 63 - Monitoring Fuel and Waste Storage .. . ... .. . . .. 3.1-30 3.1.55 Criterion 64 - Monitoring Radioactivity Releases . ........ . . .. 3.1-30 3.2 Classification of Structures, Components, and Systems . .. ... . 3.2-1 3.2.1 Seismic Classification ..... . . .. . .. . .. . .... 3.2-1 3.2.2 System Quality Group Classifications (Safety Class) .. . . . ... ... 3.2-2 3.3 Wind and Tornado Loadings . . . ........ .... ... ... . . 3.3-1 3.3.1 Wind Loadings . . .. ........ ........... ...... . . 3.3-1 3.3.2 Tornado loadings . . . . ......... ... .......... ..... 3.3-1 3.3.3 Effect of Failure of Structures or Components Not Designed for Wind and Tornado loads . .... .. ....... . .. ... ... ..... . 3.3-2 3.4 Water Level (Flood) Design . . ... . . . . ... ..... .. . 3.4-1 3.4.1 Flood Elevations ...... ..... . .. . .. . . ... .. 3.4-1 3.4.2 Phenomena Considered in Design Load Calculation . . . . . . . .. ..... . 3.4-1 3.4.3 Flood Source Application . . .......... . ........ . .... 3.4-1 3.4.4 Flood Protection . ............. .. . . . ... ..... 3.4-1 3.4.5 Analytical and Test Procedures ........... . ....... ... . 3.4-3 3.5 Missile Protection . ... . . ... .......... .. . 3.5-1 3.5.1 Missile Selection and Description . . . .. . .... .. .. .. .. 3.5-1 3.5.2 Structures, Systems, and Components to be Protected from Externally Generated Missiles . . . . . . . . . .... ... . .. . . ....... 3.5-4 3.5.3 Barrier Design Procedures . ...... .... ... . .. ...... .. . 3.5-4 3.5.4 General Design Bases .... ..... . .......... ...... . .. 3.5-5 3.6 Protection Against Dynamic Effects Associated with the Postulated Rupture of Piping . . . . . . .... ...... ..... .. .. ... 3.6-1 3.6.1 Postulated Piping Failures in Fluid Systems . .. .... . ..... . . 3.6-1 Apptwed Design Materia! . Design of SSC (2/95) Pageiv
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[ System 80+ oenior correralocewrmut , 3 .
- (V Chapter 3 Contents (Cont'd.)
] !
Page j ., t t
- 3.6.2 Determination of Break IAcations and Dynamic Effects Associated with the _
Postulated Rupture of Piping . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-10 i ! 3.6.3 . I.eak-Before-Break Evaluation Procedure . . . . . . . . . . . . . . . . . . . . . . . . 3.6-26 li i
' Appendix 3.6A SupplementalInformation on Design and Analysis for Pipe Whip . 3.6A-1 l 3
5 3.7 hiemic Design . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-1 ; 3.7.1 _ Seismic input .. . ........ .. .. ... .. ...... ... . . ...... .... .
. 3.7-1 l 3.7.2 Seismic System Analysis . . ..... . .... ...... . .. .. .. ...... .... 3.7-3 l 4 3.7.3 Seismic Subsystem Analysis - . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-18 l
- 3.7.4 Seismic Instrumentation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-32 !
' 3.7.5 Seismic Category I Tanks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-35 l i
i Appendix 3.7A Coupled Reactor Coolant System Seismic Results . . . . . . . . . . . . 3.7A-1 j- Appendix 3.7B Soil Structure Interaction (SSI) Analysis Methodology anxi ; Results - Nuclear Island Strucutes . . . . . . . . . . . . . . . . . . . . . . 3.7B-1 Appendix 3.7C . Soil Structure Interaction'(SSI) Analysis Methodology and Results - Other Seismic Category I Structures . . . . . . . . . . . . . . . 3.7C-1 ! Appendix 3.7D Sample In-Structure Response Spectra . . . . . . . . . . . . . . . . . . . 3.7D- 1
! k 3.8 Design of Category I Stnictmus . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-1 3.8.1 - Concrete Containment .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . 3.8-1 .
Steel Containment . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . .. 3.8-1 3.8.2 i i 3.8.3 Concrete and Structural Steel Internal Structures . . . . . . . . . . . . . . . . . . . . 3.8-19 , I 3.8.4 Other Seismic Category I and Seismic Category II Structures . . . . . . . . . . . . 3.8-22 l 3.8.5 Foundations ..... ..... ................... .. ........... 3.8-34 Appendix 3.8A Structural Design Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8A-1 4 Appendix 3.8B - Design Details of Seismic Category I Structures . . . . . . . . . . . . 3. 8B-1 Mechanical Systems and Components . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-1 3.9
- 3.9.1 Special Topics for Mechanical Components . . . . . . . . . . . . . . . . . . . . . . . . 3.9-1
< 3.9.2 Dynamic System Analysis and Testing . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-18 3.9.3 ASME Code Class 1,2 and 3 Components. Component Supports and Class CS Core Support Structures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-26 3.9.4 Control Element Drive Mechanisms . . . . . . . . . . . . . . . . . ........... 3.9-43 3.9.5 Reactor Vessel Core Support and Internals Structures . . . . . . . . . . . . . . . . . 3.9-50 3.9.6 Testing of Pumps and Valves . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-55 Appendix 3.9A SupplementalInformation on Criteria and Analysis of System 80+
Distribution Systems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9A-1
. 3.10 Seismic and Dyna =Ic Qualification of Mechanical and Electrical Equipment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.10-1 . 3.10.1 Seismic Qualification Criteria . . . . . . . . . . . . . . . . . . . . . ........... 3.10-1 Appmod DeeQn anese69- DeaQn of SSC Pmpe v
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System 80+ Design ControlDocument ; Chapter 3 Contents (Cont'd.) h Page 3.10.2 Seismic and Dynamic Qualification of Electrical Equipment . . . . . . . . . . . . 3.10-2 3.10.3 Seismic and Dynamic Qualification of Mechanical Equipment including Motors ......... . . . ... .............. .. . . . 3.10-6 3.10.4 Mechanical and Electrical Equipment Qualification Records . . . . . ..... . 3.10-10 3.10.5 Administrative Control of Component Qualification . . . . . . . . ...... . 3.10-11 3.11 Environmental Design of Mechanical and Electrical Equipment ... . ... 3.11-1 3.11.1 Equipment Identification and Environmental Conditions . ......... . . 3.11-2 3.11.2 Qualification Tests and Analyses . . . . . . . . . . . ............ .... 3.11-2 3.11.3 Qualification Test Results . . . . . . . ........... . ... ...... 3.11-6 3.11.4 Class IE Instrumentation Loss of Ventilation Effects . . . .. ........... 3.11-6 3.11.5 Chemical Spray, Humidity, Submergence, and Power Supply Voltage and Frequency Variation ...... . . . ... . . .. . .. ... 3.11-7 3.11.6 Radiation Environmental Qualification . . .. . ..... . .. ..... 3.11-8 Appendix 3.11 A Typical Envirnomental Conditions and Test Profiles for r Structures and Components .. . . .. ... . .... 3.11 A-1 Appendix 3.11B Identification, Location and Typical Environmental Conditions of Equipment .. . . .. ......... ... . ... 3.llB-1 O Chapter 3 Tables Page 3.2-1 Classification of Structures, Systems, and Components .. .. . ...... 3.2-4 3.2-2 Safety Class 1,2 & 3 Valves . . . . . . . . ............. ....... . 3.2-25 3.2-3 Relationship of Safety Class to Code Class . . .. .. .... . .. .. . 3.2-48 3.5-1 Kinetic Energy of Potential Missiles ............. .... ..... .. . 3.5-7 3.5-2 Design Basis Tornado Missiles and their impact Velocities ....... .... 3.5-8 3.5-3 Minimum Acceptable barrier Thickness Requirements for Local Damage Prediction Against Tornado Generated Missiles . .... . .. ... .. 3.5-8 3.6-1 High- and Moderate-Energy Fluid Systems . ,, . . .. . 3.6-37 3.6-2 Systems Required for Safe Shutdown and/or to Mitigate the Consequences of a Design-Basis Accident . .. .... . ... ......... 3.6-38 3.6-3 High-Energy Lines Within Containment . . . . ... . ... . . . . 3.6-39 3.6-4 High-Energy Lines Outside Contaimnent . . . .. . ..... . 3.6-48 3.7-1 Damping Values . . . . . . . . . ....... .. .... . . 3.7-38 3.8-1 Design Loadings for Steel Containment .. . ........ .. ......... 3.8-37 3.8-2 Loading Combinations for Steel Containment . . .. ... ... . .... . 3.8-38 3.8-3A Stress Intensity Results for the Steel Containment V ssel . . .. . .. . 3.8-40 3.8-3B Service Level A Stress Analysis, Simplified Elastic-Plastic Analysis . . . . 3.8-41 3.8-3C Stability Evaluation for the Steel Containment Vessel .. .... ..... . 3.8-42 3.8-3D Ultimate Pressure Capacity . . . .. . . . . . .. ... . 3.8-42 Approved Design Material. Design of SSC (11/96) Page vi
1 i i System 80+ Design ControlDocument i q f V
/ Chapter 3 Tables (Cont'd.)
l Page 3.8-3E Stress Intensity Allowables and Results for the Steel Containment Vessel . . . . 3.843 3.8-4 Codes and Specifications for Design of Category I Structures . . . . . . . . . . . . 3.8-43 3.8-5 Load Combinations for Category I Structures . . . . . . . . . . . ....... . 3.8-43 3.9-1 Transients Used h1 Stress Analysis of Code Class 1 and CS Components . . . . . 3.9-68 3.9-2 Loading Combinations for ASME Code Class 1,2, and 3 Components and Component Supports . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-73 3.9-3 Stress Limits for ASME Code Class 1 Components, Piping, and Component Supports . . . . . . . . . . . . . . . . . . . . . . . . . . . . ......... ...... 3.9-74 3.9-4 Seismic I Active Valves . . . . . . . . . . . . . . . . . . . . . . . - . . . . . . . . . . 3.9-75 3.9-5 Stress Criteria for Safety-Related ASME Class 2 and Class 3 Vessels . . . . . . . 3.9-83 3.96 Stress Criteria for ASME Code Class 2 and Class 3 Inactive Pumps and Pump Supports ...................................... . 3.9-83 3.9-7 Design Criteria for Active Pumps and Pump Suppons . . . . . . . . . . . . . . . . 3.9-84 3.9-8 Stress Criteria for Safety-Related ASME Class 2 and Class 3 Inactive Valves . . . . . . . . . . . ................................ .. 3.9-85 3.9-9 Stress Criteria for Safety Related ASME Class 2 and Class 3 Active Valves . . . 3.9-86 3.9-10 Loading Combinations for ASME Section III Class 1 Piping . . . . . . .... . 3.9-87 3.9-11 Loading Combinations for ASME Section III Classes 2 and 3 Piping . . . . . . . 3.9-87 3.9-12 Leading Conditions and 1. cad Combination Requirements for ASME Code / Class 1, 2, and 3 Piping Supports . . . . . . . . . ..... ............ . 3.9-88 k 3.9-13 Stress Limits for CEDM Pressure Housings . . . . . . . . . . . . . . . . . . . . 3.9-89 3.9-14 Stress Limits for Core Support and Internal Structures Design and Service Loads 3.9-90 3.9-15 Inservice Testing of Safety-Related Pumps and Valves . . . . . ........... 3.9-91 3.9-16 Pressure Isolation Valves . . . . . . . . . . . . . . . . . . . . . . . . . ......,. 3.9-180 3.9-17 Reactor Vessel Internals Arrangement Comparison . . . . . . . . . . . . . . ... 3.9-181 3.9-18 Nominal Dimensional Comparison Reactor Pressure Vessel Internals . . . . . . . 3.9-183 t 3.11-1 Ventilation Areas . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.11-14 3.11-2 Summary of Assigned Radiological Equipment Qualification Level .. ..... 3.11-15 Chapter 3 Figures 3.3-1 Wind Pressure Distribution Coefficients . . . . . . . . . . .......... .... 3.3-3 3.6-1 Typical Crush Pipe Whip Restraint Configuration . . . . . . . . . .. .. . . . 3.6-53 3.7-1 Synthetic Time History H1 Spectra vs. Target Spectra for CMS 1 (2, 5 and 7 % Damping) . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-39 3.7 2 Synthetic Time History H2 Spectra vs. Target Spectra for CMS 1 (2,5 and 7% Damping) ........ ....... ........... 3.7-40 3.7-3 Synthetic Vertical Time History Spectra vs. Target Spectra for CMSI (2, 5 and 7 % Damping) . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-41 3.7-4 Synthetic Time History H1 Spectra vs. Target Spectra for CMS 2 (1 and 2 % Damping) . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . 3.7-42 .Q (.J 3.7-5 Synthetic Time History H1 Spectra vs. Target Spectra for CMS 2 (5 and 7% Damping) ....... ........... . . . . . . . . . . . . 3.7-4 3 Anwmd onw monww one or ssc num rose mv
Srtem 80+ Design ControlDocument Chapter 3 Figures (Cont'd.) h Page 3.7-6 Synthetic Time History H2 Spectra vs. Target Spectra for CMS 2 (1 and 2% Damping) .......... . .... . . .... .. 3.7-44 3.7-7 Synthetic Time History H2 Spectra vs. Target Spectra for CMS 2 (S and 7% Damping) .................. ... ..... ... 3.7-45 3.7-8 Synthetic Vertical Time History Spectra vs. Target Spectra for CMS 2 (1 and 2% Damping) . . . ..... ...... ... . . . . . 3.7-46 3.7-9 Synthetic Vertical Time History Spectra vs. Target Spectra for CMS 2 (5 and 7% Damping) . . .. . .... ... .. . 3.7-47 3.7-10 Synthetic Time History HI Spectra vs. Target Spectra for CMS 3 (1,2,5 and 7% Damping) . . . . . . . . ... ...... . 3.7-48 3.7-11 Synthetic Time History H2 Spectra vs. Target Spectra for CMS 3 (1,2,5 and 7% Damping) .. . ..... .... . .. ... 3.7-49 3.7-12 Synthetic Venical Time History Spectra vs. Target Spectra for CMS 3 (1,2,5 and 7% Damping) .. ...... .. . .. .. 3.7-50 3.7-13 Schematic Diagram of Interior Structure, Shield Building, FB,CVCS .... ... ........... .. . . .. . 3.7-51 3.7-14 Schematic Diagram of Interior Structure, Shield Building, DG-1, DG-2 . ... . .... . ...... ...... .. ........ 3.7-52 3.7-15 Schematic Diagram of Interior Structure, Shield Building, EFW1 (Horizontal), EFW2 (Horizontal) . . .. ... . . . 3.7-53 3.7-16 Schematic Diagram of Interior Structure, Shield Building, EFW1 (Vertical), EFW2 (Vertical) . . . . . ....... .... . 3.7-54 3.7-17 Schematic Diagram of Interior Structure, Shield Building, CAA, CAB . . . . . 3.7-55 3.7-18 Finite Element Model of Steel Containment Vessel .. . .. . .. . 3.7-56 3.7 19 Schematic of Combined NI Structures (Elevation Looking South) . . . . 3.7-57 3.7-20 Schematic of Combined NI Structures (Elevation Looking West) . ... . . . 3.7-58 3.7-21 Schematic Diagram of the SASSI Analysis Process Using CMS 2 and CMS 3 Motions . ...... ......, .... . ..... .. ...... .. 3.7-59 3.7-22 Schematic Diagram of the SASSI Analysis Process Using CMS 1 Motions . . . . 3.7-60 3.7-23 Reactor Coolant System Seismic Analysis Model ... . . ... ... 3.7-61 3.7-24 Pressurizer Seismic Analysis Model . . .. .. ... . ........ 3.7-62 3.7-25 Typical Surge Line Seismic Analysis Model ... . . .... 3.7-63 3.7-26 Reactor Internals Horizontal Seismic Analysis Model . . ......... .. 3.7-64 3.7-27 Reactor Internals Nonlinear Horizontal Seismic Model . . .... 3.7-65 3.7-28 Core Seismic Model; One Row of 17 Fuel Assemblies . .. . .. . . 3.7-66 3.7 29 Reactor Internals Linear Vertical Seismic Model . ........ ... ..... 3.7-67 3.7-30 Reactor Internals Nonlinear Vertical Seismic Model . . .. .. ...... 3.7-68 3.7-31 Core-Support Barrel Upper Flange Finite-Element Model ... . . . .. 3.7-69 3.7-32 Damping Value for Seismic Analysis of Piping . . .. . ,. .. 3.7-70 3.7-33 Proportional Damping . . . . . . .... . ... .. .. . .... 3.7-71 3.8-1 Containment Details ....... ... . . ... .. . . . ... 3.8-44 3.8-2 Category I Structures - Typical Feedwater Penetration .. ..... .... 3.8-47 3.8-3 Three-Dimensional ANSYS Containment Model .. . ..... .... . . 3.8-54 3.8-4 Axisymmetric ANSYS Containment Model .. . . .. .... ... 3.8-55 l 3.8-5 Nuclear Island Structures . . .. . ...... .. . . .. .. ... . 3.8-57 Approved Design Material- Design of SSC (2/95) Pope wh7
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svstem aos o.aar, cara oocam.,,r O ca ater3 vis re-(c *'a> ! i Page 3.9-1 Reactor Coolant System Supports Diagram ...................... 3.9-184' 4 3.9 2 Summary of Analytical Methodology . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-185 3.9-3 ASHSD Finite Element Model of the CSB System . . . . . . . . . . . . . . . . . . . 3.9-186 3.9-4 Control Element Shroud Tube Finite Element Model ._ . . . . . . . . . . . . . . . . 3.9-187 l 3.9-5 Lower Support Structure Instrument Nozzle Assembly Finite Element Model . . 3.9-188 ! 3.9 ICI Suppon Tube; Outer Position Finite Element Model . . . . . . . . . . . . . . . 3.9-189 3.9-7 . Skewed Beam Support Columns Finite Element Model . . . . . . . . . . . . . . . 3.9-190 ; 3.9-8 Control Element Drive Mechanism . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-191 3.9-9 Reactor Vertical Arrangement . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 . 9- 192 3.9-10 Core Support Barrel Assembly . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-193 ; 3.9-11 Reactor Vessel Core Support Barrel Snubber Assembly . . . . . . . . . . . . . . . 3.9-194 : 3.9-12 In-Core Instrument Support Structure . . . . . . . . .................. 3.9-195
.3.9-13 Core Shroud Assembly . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-1%
3.9-14 Upper Guide Structure Assembly . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-197 3.9-15 In-Core Instrumentation Guide Tube System . . . . . . . . . . . . . . . . . . . . . . 3.9-198 3.9-16 Typical inservice Testing Connections . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-199 ,i l l l O 1 Anwevenf De*4rr nieterde!
- Denker of SSC Pope hr
System 80+ Design controlDocument i ( 3.0 Design of Structures, Components, Equipment, and Systems 3.1 Conformance with NRC General Design Criteria In this section, brief discussions are presented in response to the current General Design Criteria for Nuclear Power Plants, Appendix A to 10 CFR 50. These discussions summarize the mannet in which the principal design features meet the individual criteria and include references to sections of the safety analysis report where more detailed information is given. i 3.1.1 Criterion 1 - Quality Standards and Records Structures, systems, and components important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed. Where generally recognized codes and standards are used, they shall be identified and evaluated to determine their applicability, adequacy, and sufficiency and shall be supplemented or modified as necessary to assure a quality product in keeping with the required safety function. A quality assurance program shall be established and implemented in order to provide adequate assurance that these structures, systems, and components will satisfactorily perform their safety functions. Appropriate records of the design, fabrication, erection, and testing of structures, systems, and components important to safety shall be maintained by or under the control of the nuclear power unit licensee throughout the life of the unit.
Response
The structures, systems, and components described in the Approved Design Material (ADM) are ! classified according to their importance in the prevention and mitigation of accidents using the classification system described in ANSI /ANS 51.1. Each safety-related component is given a safety class designation. The codes, standards, and quality control applicable to each component and its safety class designation are identified in Section 3.2. Where applicable, design and fabrication are in accordance with the codes required in 10 CFR 50.55a. The quality assurance program conforms with the requirements of 10 CFR 50, Appendix B, " Quality Assurance Criteria for Nuclear Power Plants," and is presented in Chapter 17. Chapter 14 describes initial tests and operations to assure performance ofinstalled equipment commensurate with the importance of the safety function. The design, fabrication, and quality programs for components not included in the ANSI classification system are governed by industry codes appropriate to the application. Details of conformance to these codes are found in the appropriate ADM sections. 3.1.2 Criterion 2 - Design Bases for Protection Against Natural Phenomena Structures, systems, and components important to safety shall be designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricares, ficods, tsunami, and seiches without loss of capability to perform their safety functions. The design bases for these structures, systems, and components shall reflect: (1) appropriate consideration of the most severe of the natural phenomena that ! have been historically reported for the site and surrounding area, with sufficient margin for the limited a:cacy, quantity, and period of time in which the historical data have been accumulated; (2) appropriate ) mmbinations of the effects of normal and accident conditions with the effects of the natural phenomena; and, (3) the importance of the safety functions to be performed. n;=.=: outen mesaw- onion or ssc rose 2.1 1
System 80 + Design controlDocument
Response
The structures, systems, and components designated Seismic Category 1 are designed to withstand, without loss of function, the effects of any one of the most severe natural phenomena, including flooding, hurricanes, tornadoes, and the Safe Shutdown Eanhquake (SSE) (refer to Chapter 2). Design criteria for wind and tornado, flood and earthquake are discussed in Sections 3.3, 3.4, and 3.7, respectively. The seismic design of safety-related structures, systems, and components is consistent with conservative structural envelopes. These " envelopes" have been selected based on the design basis earthquakes at the majority of potential plant sites in the continental U.S., using current containment structure designs. In the design stage, normal operating and acc: dent loads are appropriately combined with the seismic loads and allowable stress limits and deformatio is are defined so that safety functions are not jeopardized. 3.1.3 Criterion 3 - Fire Protection Structures, systems, and components imporant to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions. Noncombustible and heat resistant materials shall be used wherever practical throughout the unit, particularly in locations such as the containment and control room. Fire detection and fighting systems of appropriate capacity and capability shall be provided and designed to minimize the adverse effects of fires on structures, systems, and components important to safety. Firefighting systems shall be designed to assure that their rupture or inadvertent operation does not significantly impair the safety capability of these structures, systems, and components.
Response
All pressure boundary components and structures and the attendant auxiliary systems in System 80+" W design scope are designed to minimize the probability and effects of fires and explosions. High grade noncombustible and fire resistant materials are used for components located in the containment, components of engineered safety feature systems, and throughout the unit wherever practical. A detailed functional description of the Fire Protection System is provided in Section 9.5.1. 3.1.4 Criterion 4 - Environmental and Missile Design Bases Structures, systems and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss of coolant accidents. These structures, systems, and components shall be appropriately protected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids, that may result from equipment failures and from events and conditions outside the nuclear power unit. However, dynamic effects associated with postulated pipe ruptures in nuclear power units may be excluded from the design basis when analyses reviewed and approved by the Commission demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping. O i System 80+ is a trademark of Combustion Engineering, Inc. Appmed Des # MetwW Desy of S5C Page 3.12
System 80+ Design ControlDocument n ~
Response
Structures, systems, and components important to safety are designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant-accidents (see Section 3.11). Where appropriate, Standardized Functional Descriptions will include design requirements to ensure that these structures, systems, and components will be appropriately protected against dynamic effects (including the effects of missiles, pipe whipping, and discharge of fluids) that may result from equipment failures, postulated accidents, and from events and conditions outside the nuclear power unit. , r The reactor building is capable of withstanding the effects of missiles originating outside the containment such that no credible missile can result in a LOCA. The control room is designed to withstand such missiles as may be directed toward it and still maintain the capability of controlling the plant. 3.1.5 Criterion 5 - Sharing of Structures, Systems, and Components Structures, systems, and components important to safety shall not be shared arnong nuclear power units unless it can be shown that such sharing will not significantly impair their ability to perform their safety functions, including, in the event of an accident in one unit, an orderly shutdown and cooldown of the remaining units.
Response
' k/ The System 80+ design is based on non-shared systems. 3.1.6 Criterion 10 - Reactor Design I The reactor core and associated coolant, control and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of , normal operation, including the effects of anticipated operational occurrences. j i
Response
Specified Acceptable Fuel Design Limits (SAFDLs) are stated in Section 4.4.1. Operation within the operating limits (Limiting Conditions for Operation) specified by the Technical Specifications will keep the reactor fuel within the SAFDLs for normal operation and during any Anticipated Operational Occurrence. The plant is designed such that operation within Limiting Conditions for Operation with safety system settings not less conservative than the Limiting Safety System Settings prescribed in the Technical Specifications results in confidence that SAFDLs will not be exceeded as a result of any Anticipated Operational Occurrence. Operator action, aided by the control systems and monitored by plant instrumentation, maintains the plant within Limiting Conditions for Operation during normal operation. See the following sections for additional information:
- Fuel System Design, Section 4.2
- Reactor Coolant System, Chapter 5 Newmd outen moww onion or ssc ray, 3.1.s l
System 80 + Design ControlDocument
- Shutdown Cooling System, Section 5.4.7
- Reactor Protective System, Section 7.2
- Analysis of Anticipated Operational Occurrences, Chapter 15
- Technical Specifications, Chapter 16 3.1.7 Criterion 11 - Reactor 19herent Protection The reactor core and associated coolant systems shall be designed so that in the power operating range the net effect of the prompt inherent nuclear feedback characteristics tends to compensate for a rapid increase in reactivity.
Response
In the power operating range, the combined response of the fuel temperature coefficient, the moderator temperature coefficient, the moderator void coefficient, and the moderator press:are cooTicient to an increase in reactor power will be a decrease in reactivity; i.e., the inherent nuclear feedback characteristics will not be positive. The reactivity coefficients for this reactor are discussed in detail in Section 4.3. 3.1.8 Criterion 12 - Suppression of Reactor Power Oscillations The reactor core and associated coolant, control, and protection systems shall be designed to assure that power oscillations which can result in conditions exceeding specified acceptable fuel design limits are not possible or can be reliably and readily detected and suppressed.
Response
Power level oscillations do not occur. The effect of the negative power coefficient of reactivity (see GDC 11, Section 3.1.7), together with the coolant temperature program maintained by control of regulating rods and soluble boron, provides fundamental mode stability. Power level is continuously monitored by neutron flux detectors (Chapter 7). Power distribution oscillations are detected by neutron flux detectors. Axial mode oscillations are suppressed by means of part-strength or full-strength neutron absorber rods. All other modes of oscillation are expected to be convergent. Monitoring and protective requirements imposed by Criteria 10 and 20 are discussed in Sections 3.1.6, 3.1.16 and in Chapter 4. 3.1.9 Criterion 13 - Instrumentation and Control Instmmentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for Anticipated Operational Occurrence, and for accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operatmg ranges. Approved Desip MaterW. Desip of SSC Page 3.14
Sy' tem 80 + Design ControlDocument p Response: V Instrumentation is provided to monitor significant process variables which can affect the fission process, the integrity of the reactor core, the Reactor Coolant Pressure Boundary (RCPB) and their associated systems. Controls are provided for the purpose of maintaining these variables within the limits prescribed for safe operation. Instrumentation for the containment and its associated systems can be found in the appropriate chapters and in the site-specific Safety Analysis Report. The principal process variables to be monitored and controlled are:
- Neutron flux level (reactor power)
- CEA positions
- Neutron flux distribution (at various axial positions)
- Reactor coolant temperature and pressure
- Reactor coolant pump speed
- Pressurizer level
- Steam generator level and pressure In addition, Departure from Nucleate Boiling Ratio (DNBR) margin and Local Power Density (LPD) margin, in kW/ft, are also monitored.
(V,,) The Plant Protection System (PPS) consists of the Reactor Protective System (RPS) and the Engineered Safety Features Actuation Systein (ESFAS). The RPS is designed to monitor NSSS operating conditions and to initiate reliable and rapid reactor shutdown if monitored variables or combinations of monitored variables deviate from the permissible operating range to a degree that a safety limit may be reached. The ESFAS is designed to monitor plant variables and to actuate Engineered Safety Feature (ESF) systems during a design basis event. The following are provided to monitor and maintain control over the fission process during transient and steady state periods over the lifetime of the core: e Redundant channels of ex-core nuclear instrumentation, which constitute the primary means of monitoring the fission process for protection, control and low power operation.
- Redundant and diverse CEA position indicating systems for each CEA.
- Manual and automatic control of reactor power by means of CEAs.
- Manual regulation of coolant boron concentration.
- A Boronometer, which determines the boron concentration in the reactor coolant by neutron p absorption, provided as a backup to the primary method of determining soluble poison y/ concentration by routine sampling and analysis of reactor coolant.
ApprowlDesign Materia! Design of SSC Page 3.15
System 80+ Design ControlDocument
- In-core instrumentation, provided to supplement information on core power distribution and to enable calibration of ex-core flux detectors.
The non-nuclear instrumentation measures temperatures, pressures, flows and levels in the Reactor Coolant System and main steam and auxiliary systems and is used to maintain these variables within the prescribed limits. The instrumentation and control systems are described in detail in Chapter 7. The Boronometer is discussed in Sections 7.7.1.1.7 and 9.3.2 while the process radiation monitor is discussed in Section 9.3.2. When it is required that a variable be monitored during and after a Design Basis Event (DBE), in addition to normal operation, the results of analysis of the course of the event are used to ensure that the instruments provided will cover the range anticipated for the event conditions. 3.1.10 Criterion 14 - Reactor Coolant Pressure Boundary The reactor coolant pressure boundary shall be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture.
Response
The reactor coolant pressure boundary is defined in accordance with 10 CFR 50.2(v) and ANSI /ANS 51.1 (see response to GDC 55, Section 3.1.48). Reactor Coolant System components are designed to meet the requirements of the ASME Code, Section III. To establish operating pressure and temperature limitations during startup and shutdown of the Reactor Coolant System, the fracture toughness rules defined in the ASME Code, Section III, are followed. Quality control, inspection, and testing are performed as required by ASME Section III and allowable reactor pressure-temperatore operations are specified to ensure the integrity of the Reactor Coolant System. The reactor coolant pressure boundary is designed to accommodate the system pressures and temperatures attained under all expected modes of unit operatwn including all anticipated transients, and maintain the stresses within applicable limits. Piping and equipment pressure parts of the reactor coolant pressure boundary are assembled and erected by welding unless applicable codes permit flanged or screwed joints. Welding procedures are employed which produce welds of complete fusion and free of unacceptable defects. All welding procedures, welders, and welding machine operators are qualified in accordance with the requirements of Section IX of the ASME Boiler and Pressure Vessel Code for the materials to be welded. Qualification records, including the results of the procedure and perfonnance qualification tests and identification symbols assigned to each welder are maintained. The pressure boundary has provisions for in-service inspection in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, to ensure continuance of the structurel and leak-tight integrity of the boundary (see response to GDC 32, Section 3.1.28). For the reactor vessel, a material surveillance program conforming with the requirements of Appendix H to 10 CFR 50 is provided. O Appeswd Design Matenal
- Design of SSC i11/96) Page 3.1-6
_ _ _ _ _ ._ _ _ __ __ _ . _ _ _ _ _ _ . . _ _ _ _ . ~ . _ _ _ _ _ _ _ . _ _ _ _ System 80+ Deslan ControlDocument 4 3.1.11 Criterion 15 - Reactor Coolant System Design i The Reactor Coolant System (RCS) and associated auxiliary, control, and protection systems shall be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operational occurrences.
Response
- ' The design criteria and bases for the reactor coolant pressure boundary are described in the response to Criterion 14.
The operating conditions -for normal steady state and transient plant operations are established conservatively. Normal operating limits are selected so that an adequate margin exists between them and the design limits. The plant control systems are designed to ensure that plant variables are maintained ! well within the established operating limits. The plant transient response characteristics and pressure and j temperature distributions during normal operations are considered in the design as well as the accuracy
- and response of the instruments and controls. These dr. sign techniques ensure that a satisfactory margin 3
~is maintained between the plant's normal operating conditions, including design transients, and the design limits for the reactor coolant pressure coundary. ;
Plant control systems function to minimize the d2viations from normal operating limits in the event of most Anticipated Operational Occurrences. Whr.re control systems response would be inadequate or fail upon demand, the Plant Protection System functions to mitigate the consequences of such events. , The Plant Protection System functions to mitigate the consequences in the event of accidents. Analyses i show that the design limits for the reactor coolant pressure boundary are not exceeded in the event of any , i ANSI /ANS 51.1 conditions. 4 3.1.12 Criterion 16 - Cantainment Design Reactor containment and associated systems shall be provided to establish an essentially leak-tight barrier
- against the uncontrolled release of radioactivity to the environment and to assure that the containment design conditions important to safety are not exceeded for as long as postulated accident conditions l require. ,
)
- Response: 1 i l The containment system is designed to protect the public from the consequences of a LOCA, based on l the equivalent energy release of a postulated break of reactor coolant piping up to and including a double- [
!' ended break of the largest reactor coolant pipe. l t The containment vessel, shield building, and the associated Engineered Safety Feature systems are [ designed to safely withstand all internal and external environmental conditions that may reasonably be j expected to occur during the life of the plant, including both short- and long-term effects following a LOCA. ! Leak-tightness of the containment system and short- and long-term performance following a LOCA are .
. analyzed in Section 6.2.
t l w 'oneen anesaw. cons no er ssc rare 2. s.1 i
l System 80+ Design controlDocurnent \ l 3.1.13 Criterion 17 - Electrical Power Systems An onsite electric power system shall be provided to permit functioning of structures, systems and components important to safety. The safety function for each system (assuming the other system is not ! functioning) shall be to provide sufficient capacity and capability to assure that (1) specified acceptable i fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded as a l result of Anticipated Operational Occurrences and (2) the core is cooled and containment integrity and l other vital functions are maintained in the event of postulated accidents. l The onsite electric power supplies, including batteries, and the onsite electric distribution system, shall have sufficient independence, redundancy and testability to perform their safety functions assuming a single failure. Electrical power from the transmission network to the onsite electric distribution system shall be supplied by two physically independent circuits (not necessarily on separate right of ways) designed and located so as to minimize to the extent practical any likelihood of their simultaneous failure under operating and postulated accident and environmental conditions. A switchyard common to both circuits is acceptable. Each of these circuits shall be designed to be available in sufficient time following a loss of all onsite alternating current power supplies and the other offsite electric power circuit, to assure that specified acceptable fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded. One of these circuits shall be designed to be available with5 a few seconds following a loss-of-coolant-accident to assure that the core cooling, containment integrity and other vital safety functions are maintained. Provisions shall be included to minimize the probability of losing electric power from any of the remaining supplies as a result of, or coincident with, the loss of power generated by the nuclear power unit, the loss of power from the transmission network or the loss of power from the onsite electric power supplies.
Response
The System 80 + Standard Design is provided with an onsite electric power system and an offsite electric power system to permit functioning of structures, systems and components important to safety in full compliance with the requirements of this criterion as described in Chapter 8. The onsite electric power system consists of separate, redundant and independent distribution systems and dedicated power supplies with sufficient capacity, capability, and testability to perform their safety functions assuming a single failure. The offsite electric power system consists of two physically independent circuits from the station switchyard. Each circuit is inunediately available and has sufficient capacity and capability to perform its safety function. Provisions are made to minimize the probability of losing electric power from any of the remaining supplies as a result of, or coincident with, the loss of power generated by the nuclear power unit. O Approved Des &n Material Des &n of SSC Page 3.1-8
- i System 80+ Destan comrarcoeumont ,
3.1.14. Critarion 18 - Inspection and Testing of Electdct i Power Systems Electrical power systems important to safety shall be designed to permit appropriate periodic inspection -i and testing of important areas and features, such as wiring, insulation, connections, and switchboards, . to assess the continuhy of the systems and the condition of their components. The systems shall be t designed.with a capability to test periodically (1) the operability and functional performance of the 1 components of the system such as onsite power sources, relays,' switches, and buses, and (2) the l operability of the systems as a whole and, under conditions as close to design as practical, the fuh j operation sequence that brings the system into operation, including operation of applicable portions of ,
- the protection system, and the transfer of power among the nuclear power unit, the offsite power system '
and the onsite power system. ,
- Response
I I Electrical power systems important to safety are designed to permit appropriate periodic inspection and ' , testing of important areas and features, such as wiring, insulation, connections, and switchboards, to assess the continuity of the systems and to detect deterioration, if any, of their components. Capability
' is provided to periodically test the operability and functional performance of the system components. The ,
diesel generator sets will be started and loaded periodically on a routine basis, and relays, switches, and buses will be inspected and tested for operation and availability on an individual basis. ; Transfer from normal to emergency sources of power will be made to check the operability of the systems i and the full operational sequence that brings the systems into operation. i Refer to Section 8.3.1, 8.3.2 and Technical Specifications Section 3.8 for more detailed informatbn. 7 3.1.15 Criterion 19 - Control Room 4 L A control room shall be provided from which actions can be taken to operate the nuclear unit safely under
- normal conditions and to maintain it in a safe condition under accident conditions, including loss-of- ;
coolant-accidents. Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposure in excess of 5 rem
- whole body, or its equivalent to any part of the body, for the duration of the accident.
Equipment at appropriate locations outside the control room shall be provided (1) with a design capability , for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold l shutdown of the reactor through the use of suitable procedures. Response: . . All control stations, switches, controllers, and indicators necessary to operate or shut the unit down and maintain safe control of the facility are located in the control room. !- The design of the control room permits safe occupancy during abnormal conditions. The employment of non-combustible and fire retardant materials in the construction of the control room, the limitation of combustible supplies, the location of fire fighting equipment, and the continuous presence of a highly L trained operator will minumze the possibility that the control room will become uninhabitable. Radiation exposure levels following design basis accidents are maintained below allowable levels by proper design of shielding and ventilation. The Control Room Ventilation System is designed to ensure that the post l Anemed seaw neeanw.ceenn er sse nrass roue s. r.s 1>,m
,e --o -e -e --
. ,-- - --m-, , 9 ~--.-r ' -, ,-- , op - ----m
System 80+ Design ControlDocument accident radiation exposure to the control room operators due to inhalation and submersion are below allowable levels (see Section 6.4) and to recirculate cool control room air as discussed in Sections 9.4.1 and 12.2. Radiation detectors and alarms are provided. Emergency lighting is provided as discussed in Section 9.5.3. Alternate local controls and instruments are available for equipment required to bring the plant to and maintain a hot standby condition. It is also possible to attain a cold shutdown condition from locations outside of the control room through the use of suitable procedures. Refer to Section 7.4.1.1.10. A discussion of the unit's control room is provided in Section 7.7.1.3 with human factors issues discussed in Chapter 18. A discussion of the hot and cold shutdown capability is provided in Section 7.4 for the systems required for safe shutdown. Discussion regarding adequate radiation protection for the unit's control facilities is provided in Section 6.4 and in Chapter 12, 3.1.16 Criterion 20 - Protection System Functions The protection system shall be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety.
Response
A Plant Protection System (PPS), consisting of a Reactor Protective System (RPS) and an Engineered Safety Features Actuation System (ESFAS), is provided. The RPS automatically initiates a reactor trip when any of the monitored process variables reach a trip setpoint. The ESFAS automatically actuates Engineered Safety Feature (ESF) systems and their support systems when any of the monitored process variables reach a predetermined serpoint. The trip setpoints of the RPS are selected to ensure that Design Basis Events (DBEs) which are expected to occur once or more during the life of the nuclear generating station do not cause the violation of SAFDLs. The reactor trips also help the ESF systems in mitigating the consequences of DBEs which are expected to occur once during the life of several plants as well as arbitrary combinations of unplanned events and degraded systems that are never expected to occur, to within acceptable limits. Reactor trip is accomplished by de-energizing the Control Element Drive Mechanism (CEDM) coils through the interruption of the CEDM power supply either automatically or manually. The CEDM power supply is a pair of full capacity motor-generator sets. The path to the CEDMs is interrupted by opening the Reactor Trip Switchgear. With the CEDM coils de-energized, the CEAs are released to drop into the core by gravity, rapidly inserting negative reactivity to shut the reactor down. The CEDMs are described in Section 4.2, the specific reactor trips used are described in Section 7.2. The ESF systems are actuated to minimize the effects of incidents which could occur. Controls are provided for manual actuation of the ESF system. The process variables which automatically actuate the ESF system and the circuitry arrangements for the ESFAS are discussed in Section 7.3. The ESF systems are discussed in Chapter 6. The SAFDL on linear heat rate and DNBR are intended to enforce the principal thermal hydraulic design basis given in Section 4.4.1 (i.e., the avoidance of thermally induced fuel damage during normal steady state operation and during Anticipated Operational Occurrences). Approamt Design Material
- Design of SSC Page 3.1-10
i System 80+ Deshrn ControlDocument i 3.1.17 Criterion 21 - Protection System R Hahility and Testability The protection system shall be designed for high functional reliability and in-service testability . i commensurate with the safety functions to be performed. Redundancy and independence designed into ; the protection system shall be sufficient to assure that (1) no single failure results in loss of the protection , ! function and (2) removal from service of any component or channel does not result in loss of the required J minimum redundancy unless the acceptable reliability of protection system operation'can be otherwise . ; demonstrated. The protection system shall be designed to permit periodic testing of its functioning when j a
- the reactor is in operation, including a capability to test channels independently to determine failures and j losses of redundancy that may have occurred.
J Respome: I. . The PPS is designed to provide high functional reliability and in-service testability. The protection :
- system is designed to comply with the requirements ofIEEE 279-1971, " Criteria for Protection Systems ,
for Nuclear Power Generating Stations," and IEEE 603-1980, " Criteria for Safety Systems for Nuclear j l ' Power Generating Stations," and other standards as noted in Section 7.1.2. No credible single failure will result in loss of the protection function. The protection channels are independent with respect to wire 3 i routing, senson mounting, and supply of power. Each channel of the protection system, including the sensors, up to the Reactor Trip Switchgear System j
- (RTSS) and ESFAS actuation devices, is capable of being checked during reactor operation. Process (
sensors of each channel in the protection systems are checked by comparison of the redundant process l ! sensor values using the discrete indications and alarms on the control room panels as described in Section 7.7.1.3.1. - Discrepancies among redundant channel sensors beyond specified limits are alarmed as ! l . described in Section 7.7.1.4.3 and Chapter 18. : The RTSS and ESFAS are described in Chapter 7. , To minimize inadvertent actuation of an ESF system or an inadvertent reactor trip, the protection systems ; utilize a coincidence of two logics to operate. In addition, the channel being tested is bypassed so that i F the protection system converts to a two-out-of-three logic while maintaining the coincidence of two. This ' allows periodic testing and operation of the various protective functions without reducing the availability of the protection systems, a i 3.1.18 Criterion 22 - Protection System Indes'-T_= ! . l The protection system shall be designed to assure that the effects of natural phenomena, and of normal l i operating, maintenance, testing and postulated accident conditions on redundant channels do not result in loss of the protection function or shall be demonstrated to be acceptable on some other defined basis. Design techniques, such as functional diversity or diversity in component design and principles of
. operation, shall be used to the extent practical to prevent loss of the protection function. l a i
- I
- Response:
The protection systems conform to the independence requirements of IEEE 279-1971. Four independent ; measurement channels, complete with sensors, sensor power supplies, signal conditioning units, and bistable trip functions are provided for each protective parameter monitored by the protection systems j
- ^ _,_ L Dee&n aseenniet Dee@n of SSC Page 3.1 11 1
..-- ---.--+ ,,,n , . . n. , 4 ,, -. . _ . . ., - ,,.
System 80 + Design ControlDocument except for the CEA position sensors which are two-fold redundant. The measurement channels are provided with a high degree of independence by separate connection of the channel sensors to the process systems. Refer to Chapter 7 for a more detailed discussion of the protection systems. Power to the protection system channels is provided by independent vital power supply buses. The power supply systems are discussed in Chapter 8. Functional diversity is incorporated into the system design, to the extent practical, to prevent loss of the protective function. Whenever an RPS trip function is required it is frequently backed up by other trip functions. The ESFAS actuation signals are used to actuate two independent ESF trains. Where it is practical, an ESFAS can be generated by more than one parameter. The Alternate Protection System augments reactor trip and emergency feedwater actuation by using separate and diverse non-lE trip logic from that used by the Plant Protection System. The design goals are accomplished without excessive complexity by using only four channels for each parameter. This allows for testing and maintenance of a channel without reducing the system to a single channel for trip, which would make the system susceptible to spurious trip or actuation signals. The protection systems are each functionally tested to ensure satisfactory operation prior to installation in the plant. Environmental and seismic qualifications are also performed utilizing type tests, specific equipment tests, appropriate analyses, or prior operating experience. For further information, refer to Sections 3.10 and 3.11. 3.1.19 Criterion 23 - Protection System Failure Modes The protection system shall be designed to fail into a safe state or into a state demonstrated to be acceptable on some other defined basis if conditions such as disconnection of the system, loss of energy (e.g., electric power, instrument air) or postulated adverse environments (e.g., extreme heat or cold, fire, pressure, steam, water, and radiation) are experienced.
Response
The Plant Protection System trip channels are designed to fail into a safe state or into a state established as acceptable in the event of loss of power supply. A failure is assumed to occur in only one channel (i.e., a single failure). This channel can be placed into bypass which places the RPS/ESFAS local coincidence logic into a two-out-of-three configuration which retains the coincidence of tw fui irip initiation. Refer to Sections 7.2 and 7.3 for Failure Modes and Effects Analysis information. A loss of power to CEDM coils will cause the CEAs to insert into the core. Redundancy, channel independence and separation are incorporated into the protection system design to minimize the possibility of the loss of a protective function. The loss of offsite power will cause the standby diesel electric generators to start and energize the ESF trains which have actuation signals present. 3.1.20 Criterion 24 - Separation of Protection and Control Systems The protection system shall be separated from control systems to the extent that failure of any single control system componer.t or channel, or failure or removal from service of any single protection system component or cha.uiel which is common to the control and protection systems leaves intact a system satisfying all reliability, redundancy, and independence requirements of the protection system. Appromt Design Material Desbyn of SSC Page 3.1-12
.. - . . _ . _ .=- _ .
System 80+ oesign controlDocument c (s Interconnection of the protection and control systems shall be limited so as to assure that safety is not significantly impaired.
Response
Protection system components and control system components are electrically and functionally isolated from each other. See Sections 7.2,7.3 and 7,7.1.1.13 for details. The protection systems are designed so that they can sustain one channel in a tripped condition and one channel by passed indefinitely and still provide their safety function. Where control and protection systems have identical sensor input requirements, redundant Class IE sensors that are used independently by each channel of the protection system may also be used by the control system. Fo r each sensed parameter, the control system monitors all four redundant instrument channels, which are interfaced to the control system via fiber-optic interfaces to ensure electrical independence. Within the control system, signal validation logic is used to detect bypassed or failed sensors, thereby ensuring that they cause no erroneous control system actions. The control system signal validation logic is described in Section 7.7.1.1.13. The design ensures that with a sensor or channel in bypass, another sensor can fail with no resulting control system action. Therefore, with one channel in bypass, the protection system remains in an effective two-out-of-three configuration, meeting the required single failure criteria. 3.1.21 Cciterion 25 - Protection System Requirements for Reactivity Control Malfunctions O) V The protection system shall be designed to assure that specified acceptable fuel design limits are not i exceeded for any single malfunction of the reactivity control systems such as accidental withdrawal (not ejection or dropout) of control rods.
Response
Shutdown of the reactor is accomplished by the opening of the RTSS circuit breakers which interrupts power to the CEDM coils. Actuation of the circuit breakers is independent of any existing control signals. The protection systems are designed such that SAFDLs are not exceeded for any single malfunction of the reactivity control systems, including the withdrawal of a single full- or part-strength CEA. Analyses of possible reactivity control system malfunctions are discussed in Chapter 15. The various CEA related DBEs for which the protection systems are designed are discussed in Section 7.2. 3.1.22 Criterion 26 - Reactivity Control System Redundancy and Capability , l Two independent reactivity control systems of different design principles shall be provided. One of the systems shall use control rods, preferably including a positive means for inserting the rods, and shall be capable of reliably controlling reactivity changes to assure that under conditions of normal operation, including Anticipated Operational Occurrences, and with appropriate margin for malfunctions such as stuck rods, specified acceptable fuel design limits are not exceeded. The second reactivity control system x shall be capable of reliably controlling the rate of reactivity changes resulting from planned, normal
') power changes (including xenon burnout) to assure acceptable fuel design limits are not exceeded. One of the systems shall be capable of holding the reactor core subcritical under cold conditions. l l
heromtonieunwint.onie orssc np 3.1 13
System 80+ oesign controlDocument Resp 4: Two independent reactivity control systems of different design principles are provided. The first system, using Control Element Assemblies (CEAs), includes a positive means (gravity) for inserting CEAs and is capable of reliably controlling reactivity changes to assure that under conditions of normal operation, including Anticipated Operational Occurrences, SAFDLs are not exceeded. The CEAs can be mechanically driven into the core. The appropriate margin for stuck rods is provided by assuming in the analyses of anticipated operational occurrences that the highest worth CEA does not fall into the core. The second system, using neutron absorbing soluble boron, is capable of reliably compensating for the rate of reactivity changes resulting from planned normal power changes (including Xenon burnup) such that SAFDLs are not exceeded. This system is capable of holding the reactor suberitical under cold conditions. Either system is capable of making the core subcritical from a hot operating condition and holding it subcritical in the hot standby condition. Either system is able to insert negative reactivity at a rate sufficient to prevent exceeding SAFDLs as the result of a power change (i.e., the positive reactivity added by Xenon burnup). 3.1.23 Criterion 27 - Combined Reactivity Control Systems Capability The reactivity control systems wil be designed to havr. a combined capability, in conjunction with poison addition by the emergency core cooling system, of reliably controlling reactivity changes to assure that under postulated accident conditions and with apprcpriate margin for stuck rods the capability to cool the core is maintained.
Response
Dissolved boron addition capability provided by the Safety injection System (Chapter 6) in concert with the control rod (CEA) system will be such that under postulated accident conditions (Chapter 15), even with the CEA of highest wonh stuck out of the core, adequz e reactivity control is available to maintain short- and long-term capability to cool the core. 3.1.24 Criterion 28 - Reactivity Limits The reactivity control systems shall be designed with appropriate limits on the potential amount and rate of reactivity increase to assure that the effects of postulated reactivity accidents can neither (1) result in damage to the reactor coolant pressure boundary greater than limited local yielding nor (2) sufficiently disturb the core, its support structures or other reactor pressure vessel internals to impair significantly the capability to cool the core. These postulated reactivity accidents shallinclude consideration of rod ejection (unless prevented by positive means), rod dropout, steam line rupture, changes in reactor coolant temperature and pressure, and cold water addition. O l Apprend Design Meterial. Design of SSC Page 3.1 14
.. ~ __ - _ -. .. . . _.
System 80+ Design controlDocument ( Response: The bases for Control Element Assembly (CEA) design include ensuring that the reactivity worth of any one CEA is not greater than a preselected maximum value. The CEAs are divided into two sets, a shutdown set and a regulating set, further subdivided into groups as necessary. Administrative procedures and interlocks assure that only one group is withdrawn at a time, and that the regulating groups are withdrawn only after the shutdown groups are fully withdrawn. The regulating groups are programmed to move in sequence and within limits which prevent the rate of reactivity addition and the worth of individual CEAs from exceeding limiting values. The maximum rate of reactivity addition which may be produced by the Chemical and Volume Control System is too low to induce any significant pressure forces which might rupture the reactor coolant pressure boundary or disturb the reactor vessel internals. The reactor coolant pressure boundary (Chapter 5) and the reactor internals (Chapter 4) are designed to appropriate codes (refer for instance, to the response to Criterion 14) and will accommodate the static and dynamic loads associated with an inadvertent, sudden release of energy, such as that resulting from a CEA ejection or steam line break (Chapter 15), without rupture and with limited deformation which will not impair the capability of cooling the core. 3.1.25 Criterion 29 - Protection Against Anticipated Operational Occurrences The protection and reactivity control systems shall be designed to assure an extremely high probability
,Q of accomplishing their safety functions in the event of anticipated operational occurrences.
V
Response
Plant events, designated in ANSI /ANS 51.1, " Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants," have been carefully considered in the design of the protection and l reactivity control systems. Consideration of redundancy, independence and testability in the design, coupled with careful component selection, overall system testing, and adherence to detailed quality l assurance requirements, assure an extremely high probability that safety functions are accomplished in ) the eyes d Design Basis Events (DBEs). Detailed discussions of the protection systems are provided in Chapter 7. Design quality assurance is discussed in Chapter 17. The analysis of DBEs is contained in Chapter 15. 3.1.26 Criterion 30 - Quality of Reactor Coolant Pressure Boundary Components which are part of the reactor coolant pressure boundary shall be designed, fabricated, erected, and tested to the highest quality standards practical. Means shall be provided for detecting and, to the extent practicable, identifying the location of the source of reactor coolant leakage.
Response
The reactor coolant pressure boundary components are designed, fabricated, erected and tested in accordance with the ASME Code Section 111. All components are classified Safety Class 1 or 2, in accordance with the ANSI /ANS 51.1, " Nuclear Safety Criteria for the Design of Stationary PWR Plants," (-). definitions for safety classes and the reactor coolant pressure boundary. Accordingly, they receive all of the quality measures appropriate to that classification. Anwovent Design Adesorial- Design of SSC Page 3.1 15
System 80+ Design ControlDocument Means are provided for the identification of the source of reactor coolant leakage. These include the detection of leakage to systems connected to the reactor coolant pressure boundary as well as leakage from the boundary into the containment. Instrumentation is provided to indicate and record makeup flow rate and integrated makeup flow to the primary water system. This instrumentation permits detection of suddenly occurring leaks and those which are gradually increasing. 3.1.27 Criterion 31 - Fracture Prevention of Reactor Coolant Pressure Boundary The reactor coolant pressure boundary shall be designed with sufficient margin to assure that when stressed under operating, maintenance, testing, and postulated accident conditions: (1) the boundary behaves in a nonbrittle manner; and, (2) The probability of rapidly propagating fracture is minimized. The design shall reflect consideration of service temocratures and other conditions of the boundary material under operating, maintenance, testing, and pu.11ated accident conditions and the uncenainties in determining: (1) material properties; (2) the effects of irradiation on material propenies: (3) residual, steady state, and transient stresses; and, (4) size of flaws.
Response
All the reactor coolant pressure boundary components are designed and constructed in accordance with ASME Section III and comply with the test and inspection requirements of these codes. The test and inspection requirements assure that flaw sizes are limited so that the probability of failure by rapid propagation is extremely remote. Particular emphasis is placed on the quality control applied to the reactor vessel on which tests and inspections exceeding ASME code requirements are performed. The tests and inspections performed on the reactor vessel are summarized in Section 5.2.4.1. Carbon and low alloy steel materials which form pan of the pressure boundary are tested in accordance with the requirements of the fracture toughness requirements for materials, ASME Code Section III. Nonductile failure prevention will be ensured by utilizing the appropriate sections of the ASME Code. Excessive embrittlement of the reactor vessen material due to neutron radiation is prevented by providing an annulus of coolant water between the reactor core and the vessel. In addition, to minimize the effects of irradiation on material toughness propenies of core beltline materials, restrictions on upper limits for j residual elements that directly influence the RTN or shift are required by the design specification. Specifically, upper limits are placed on copper, nickel, phosphorous, sulfur, and vanadium. Further, the reactor vessel is forged such that no welds occur in the active core region. The maximum integrated fast neutron flux exposure of the reactor vessel wall opposite the midplane of the core is less than 6.2 x 10" nyt. This value assumes a sixty-year vessel design life and an eighty percent plant capacity factor. The maximum expected increase in transition temperature is about 79'F. i The actual change in material toughness propenies due to irradiation will be verified periodically during ! plant lifetime by a material surveillance program. Based on an initial RTgo7 of 10*F, operating restrictions will be applied as necessary to limit vessel stresses. The thermal stresses induced by the injection of cold water into the vessel, following a LOCA, have been examined. Analyses have shown that there is no gross yielding across the vessel wall when using the minimum specified yield strength in the ASME Boiler and Pressure Vessel Code, Section III. ; Approved Design hinterial Design of SSC page 3. r.16
System 80+ Deslan ControlDocument
, 3.1.28 Criterion 32 - Inspection or Reactor Coolant Pressure Bonadary .
Components which are part of the reactor coolant pressure boundary shall be designed to permit:
*- Periodic inspection and testing ofimportant areas and features to assess their structural and leak-tight integrity; and
- an appropriate material surveillance program for the reactor pressure vessel. ,
4 E j Response: ;
- Provisions have been made in the design for inspection, testing, and surveillance of the Reactor Coolant '
System boundary as required by ASME Boiler and Pressure Vessel Code Section XI. C-E recommends a reactor vessel surveillance program to the owner. The reactor vessel surveillance program capability
- provided to the owner conforms with ASTM-E-185-73, " Practice for Conducting Surveillance Tests for ,
Light Water Cooled Nuclear Reactor Vessels," as revised in 1982. Sample pieces taken from the same l 3 . l material used in fabrication of the reactor vessel are installed between the core and the vessel inside wall. These samples will be removed and tested by the owner at intervals during vessel life to provide an i indication of the extent of the neutron embrittlement of the vessel wall. Charpy tests will be performed on the samples to develop a Charpy transition curve. By comparison of this curve with the Charpy curve i and drop weight tests for specimens taken at the beginning of the vessel life, the change of RTmyr Will I 4 be determined and operating procedures adjusted as required. See Chapter 5 for further details. The surveillance program capability provided to the owner has provisions which comply with the NRC ' regulation, " Reactor Vessel Material Surveillance Program Requirements," 10 CFR 50, Appendix H, published in the Federal Register in July 1983. The only exception between the recommended 3 surveillance program and the requirements presented in Appendix H is the following:
- Appendix H, Section II.C.2 - Attachments to the reactor vessel.
In adhering to the requirement of placing the surveillance specimens as close as possible to the - reactor vessel wall, the capsule holders are attached to the cladding of the reactor vessel and are not major load-bearing components. By such placement, temperature, flux spectra, and fluence ; differences between the surveillance specimens and the reactor vessel are minimized, thereby s i permitting more accurate assessment of the changes in the reactor vessel properties. 3.1.29 Criterion 33 - Reactor Coolant Makeup i A system to supply reactor coolant makeup for protection against small breaks in the reactor coolant pressure boundary shall be provided. The system safety function shall be to assure that specified j acceptable fuel design limits are not exceeded as a result of reactor coolant loss due to leakage from the , reactor coolant pressure boundary and rupture of small piping or other small components which are part of the boundary. The system shall be designed to assure that for onsite electrical power system operation ; (assumitc offsite power is not available) and for offsite electrical power system operation (assuming onsite power is not available) the system safety function can be accomplished using the piping, pumps , and valves used to maintain coolant inventory during normal reactor operation. i 4O i a Anm.e o w aneww c w or ssc reo, s.1-r7 L _ _ = - _ ,_. , , _ . __
System 80+ Design ControlDocument
Response
Reactor coolant makeup during normal operation is provided by the Chemical and Volume Control System (CVCS). The design incorporates a high degree of functional reliability by provision of redundant components and an alternate path for charging. The charging pumps can be powered from either onsite or offsite power sources, including the alternate AC generator. The system is described in Section 9.3.4. The CVCS has the capability of replacing the flow loss to the containment due to leaks in small reactor coulant lines such as instrument and sample lines. These lines have 7/32 inch diameter by 1 inch long flow restricting devices to limit loss of RCS coolant due to postulated pipe breaks in CVCS piping. The CVCS is not required to perform any safety related function, such as accident mitigation, or required to perform a safe shutdown. This does not, however, compromise the " defense in depth" provided by the system as the normal means of maintaining RCS inventory and primary water chemistry. In designing the CVCS as non-safety grade, the following safety functions are performed by dedicated safety systems. Boration and makeup for design basis events will be provided by the Safety Injection System. Pressure control will be provided by the Safety Depressurization System. The Safety Injection System and the Safety Deprosurization System are described in further detail in Sections 6.3 and 6.7, respectively. The Chemical and Volume Control System is designed as a non safety related system. Because the CVCS is essential for the day to day operation of the plant, it has been provided with a high degree of reliability and redundancy and has been designed in accordance with accepted industry standards and quality assurance commensurate with its importance to plant operations. Design criteria, including ASME Code classification assignments, are in accordance with ANSI /ANS 51.1, Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants". This requires that portions of the CVCS within the RCPB, and all portions which assure containment isolation will have a rigorous safety classification in accordance with these specific functional performance requirements. l 3.1.30 Criterion 34 - ResidualIIcat Removal A system to remove residual heat shall be provided. The system safety function shall be to transfer fission product decay heat and other residual heat from the reactor core at a rate such that specified acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary are not exceeded. Suitable redundancy in components and features, and suitable interconnections, leak detection and isolation capabilities shall be provided to assure that for onsite electrical power system operation (assuming offsite power is not available) and for offsite electrical power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.
Response
Residual heat removal capability is provided by the Shutdown Cooling System for reactor coolant temperatures less than 350*F. For temperatures greater than 350'F, this function is provided by the steam generators. The Emergency Feedwater (EFW) System provides a dedicated, independent, safety-related means of supplying secondary side, quality feedwater to the steam generator (s) for removal of heat l and prevention of reactor core uncovery. The design incorporates sufficient redundancy, interconnections, leak detection, and isolation capability to ensure that the residual heat removal function I can be accomplished, assuming a single active failure. Within appropriate design limits, either system will remove fission product decay heat at a rate such that SAFDLs and the design conditions of the reactor coolant pressure boundary will not be exceeded. AMwemf Design Material. Des @ of SSC (2/95) Page 3.1 18
l Sv' tem 80+ Desier ControlDocumert i
.l
- 1 The Shutdown Cooling System and the steam generator auxiliaries are designed to operate either from offsite power or from onsite power sources.
i Further discussion is included in Section 5.4.7 for the Shutdown Cooling System and in Chapter 10 for i the Steam and Power Conversion System. 3.1.31 Criterion 35 - Emergency Core Cooling
' A system to provide abundant emergency core cooling shall be provided. The system safety function , shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) clad i metal-water reaction is limited to negligible amounts.
Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation _ , and containment capabilities shall be provided to assure that for onsite electrical power system operation (assuming offsite power b not available) and for offsite electrical power system operation (assuming p onsite power is not avalable) the system safety function can be accomplished, assuming a single failure. ;
. Response: ;
- Emergency core cooling is provided by the Safety Injection System (SIS)(described in Section 6.3). The system is designed to provide cooling water to remove heat at a rate sufficient to maintain the fuel in a l coolable geometry and to assure that zirconium-water reaction is limited to a negligible amount (less than . ;
t one percent). Detailed analysis has been performed, utilizing models complying with 10 CFR 50, I Appendix K, "ECCS Evaluation Models," to verify that the system performance is adequate to meet the , intent of the "Accapta-e Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power l Reactors" of 10 CFR 50.46(b).
'\ The system design includes provisions to assure that the required safety functions are accomplished with ;
either onsite or offsite electrical power, assuming a single failure of any component (qualified as described below). The single failure may be an active failure
- during the short-term cooling phase of l safety injection or an active or limited leakage passive failure" during the long-term cooling phase of safety injection.
l ~ i An active failure is a malfunction, excluding passive failure, of a component which relies on mechanical movement to complete its ir. tended function upon demand. Check valves which receive regular exercise -
' to ensure operability are treated as pasdve components. Examples of active failures include the failure of .l' a valve to rnove to its correct position, or the failure of a pump, fan, or diesel generator to start. Spurious action of a powend component originating within the actuation system or its supponing systems shall be
. regarded as an active failure, unless specific design features or operating restrictions preclude such spunous ,
. action.
I* A passive failure is defined as the blockage of a process flow path or a breach in the integrity of a ! component or piping (e.g., a piping failure). ; Amsassed pos(ps aseender Desden er sac tr r/ps/ pay.17.ra
System 80+ Design ControlDocument Though the SIS is designed to accommodate a limited leakage passive failure during die long-term cooling phase, it does not accommodate arbitrary large leakage passive failures, such as the complete double-ended severance of piping, which are extremely low probability events. The layout and arrangement will be such that the limited leakage passive failure does not preclude ainimum acceptable long-term cooling capability. Where building design is not relied upon to mitigate and contain leakage from the SIS passive failure, suitable automatic isolation and auxiliary equipment must be provided by the owner, as necessary. 3.1.32 Criterion 36 -Inspection of Emergency Core Cooling System The emergency core cooling system shall be designed tc, permit appropriate periodic inspection of important components, such as spray rings in the reactor pressure vessel, water injection nozzles, and piping to assure the integrity and capability of the system.
Response
The SIS is designed to facilitate access to all critical components. All pumps, heat exchangers, valves and piping external to the containment structure are readily accessible for periodic inspection to ensure system leak-tight integrity. Vr.lves, piping and tanks inside the containment may be inspected for leak-tightness during plant shutdowns for refueling and maintenance. Reactor vessel internal structures, reactor coolant piping and water injection nozzles are designed to permit visual inspection for wear due to erosion, corrosion or vibration, and nondestructive inspection techniques where these are applicable and desirable. Details of the inspection program are described in Chapters 5,6, and 16. 3.1,33 Criterion 37 - Testing of Emergency Core Cooling System The emergency core cooling system shall be designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leak-tight integrity of its components, (2) the operability and performance of the active components of the system, and (3) the operability of the system as a whole and, under conditions as close to design as practical, the performance of the full operational sequence that brings the system into operation, including operation of the applicable portions of the protection system, the transfer between normal and emergency power sources, and the operation of the associated cooling water sydem.
Response
The SIS is provided with testing capability to demonstrate system and component operability. Testing can be conducted during normal plant operation with the test facilities arranged not to interfere with the performance of the systems or with the initiation of control circuits, as described in Section 6.3 and Chapter 14. 3.1.34 Criterion 38 - Containment Heat Removal A system to remove heat from the reactor containment shall be provided. The system function shall be to reduce rapidly, consistent with the functioning of other associated systems, the containment pressure and tenp- cre following any loss-of-coolant-accident and maintain them at acceptably low levels. Approwd Design Material . Design of SSC Page 3. b20
..c
System 80+ oestan contrat Document l i A . i V Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that for onsite electrical power system operation ; j (assuming offsite power is not available) and for offsite electrical power system operation (assuming l 1 onsite power is not available) the system safety function can be accomplished, assuming a single failure. j i. Response: , The Containment Spray System consists of two completely independent subsystems. The heat removal capacity of the flow from either containment spray subsystem is adequate to keep the containment pressure and temperature below design conditions for any size break in the RCS piping up to and l i including a double-ended break of the largest reactor coolant pipe, with an unobstructed discharge from i both ends. l Borated water is sprayed downward by the system from the upper regions of the containment to cool the i atmosphere. Cooling is provided by the ultimate heat sink via the containment spray heat exchangers. j Cooling reduces the containment pressure and temperature following a major loss-of-coolant-accident. Suitable redundancy in components and features is designed into the Containment Spray System to maintain the pressure and temperature conditions below containment design even in the event of a single : failure, including the loss of onsite or offsite electrical power. l 3.1.35 Criterion 39 -Inspection of Cantainnwat Heat Removal System 2 - The containment heat removal system shall be designed to permit appropriate periodic inspection of : important components, such as the torus, sumps, spray nozzles, and piping to assure the integrity and capability of the system. . l r
Response
l i All essential equipment of the Containment Spray System is located outside the containment, except for ' spray headers, nozzles, containment sump, In-containment Refueling Water Storage Tank and associated piping. These components include two containment spray pumps, two containment spray heat exchangers j and independent containment spray headers, e 4 The detailed arrangement and layout of system piping, pumps, heat exchangers, and valves will provide the separation, availability, and accessibility required for periodic inspection. Nozzle inspection capability will be provided as well. 3.1.36 Criterion 40 - Testing of Contamment Heat Removal System The containment heat removal system shall be designed to permit appropriate periodic pressure ext functional testing to assure (1) the structural and leak-tight integrity of its components, (2) the operability and performance of the active components of the system, and (3) the operability of the system as a whole, and, under conditions as close to the design as practical, the performance of the full operational sequence _ that brings the system into operation, including operation of applicable portions of the protection system, the transfer between normal and emergency power sources, and the operation of the associated cooling water system. _ i Denisn neeennet. Doeten of SSC Pope 3.121 wg a e - , I w .-- , -, -- am--.,w-, - - ,--. t. ~-,-.-,n
I System 80+ Design ControlDocument I
Response
System piping, valves, pumps, heat exchangers, and other components of the containment heat removal system are arranged so that each component can be tested periodically for operability. Testing can be ; l conducted during normal plant operation with the test facilities arranged not to interfere with the performance of the system or with the initiation of control circuits, as described in Section 6.2. l The performance testing of containment spray pumps is conducted at some time other than refueling. The pumps are aligned to take suction from and return flow to the In-containment Refueling Water Storage Tank (IRWST). Flow and head are recorded by the installed instmmentation. Heat exchanger operation may be verified during any operating mode by circulating water through the containment spray heat exchanger and back to the IRWST. Actuator-operated valves can be cycled from the control room, and operation verified by observing control room indication. Check valves will be tested to ensure that the valves operate properly. These valves inchide the IRWST check valves and the valves on the inlets and outlets of the containment spray pumps. 3.1.37 Critedon 41 - Containment Atmosphere Cleanup Systems to control fission products, hydrogen, oxygen, and other substances which may be released into the reactor containment shall be provided as necessary to reduce, consistent with the functioning of other associated systems, the concentration and quantity of fission products released to the environment following postulated accidents, and to control the concentration of hydrogen or oxygen and other substances in the containment atmosphere following postulated accidents to assure that containment integrity is maintained. Each system shall have suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) its safety function can be accomplished, assmning a single failure.
Response
Two systems, namely the Containment Spray and Containment Hydrogen Recombiner Systems, are provided to control fission products, hydrogen, oxygen, and other substances which may be released into the reactor containment. The Containment Spray and Containment Hydrogen Recombiner Systems are designed with redundancy of vital components so that a single failure does not prevent performance of the safety function coincident with a loss of offsite power. The systems are described in detail in Sections 6.2.5 and 6.5. 3.1.38 Criterion 42 -Inmection of Containment Atmosphere Cleanup Systems The containment atmosphere cleanup systems shall be designed to permit appropriate periodic inspection i of important components, such as filter frames, ducts, and piping to assure the integrity and capability of the systems. Approved Design Material- Design of SSC Page 3.1-22
System 80+ Design controloocument ( Response: The containment atmosphere cleanup systems are designed and located so that they can be inspected periodically as required. Inspection of the Containment Spray System function relative to iodine removal is treated in the response to Criterion 39. All major components of the Containment Hydrogen Recombiner System are located outside containment and are readily accessible for periodic inspection. Purge piping and valves located inside containment may be inspected during plant shutdown. 3.1.39 Criterion 43 - Testing of Containment Atmosphere Cleanup Systems The containment atmosphere cleanup systems shall be designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leak-tight integrity of its components, (2) the operability and performance of the active components of the systems such as fans, filters, dampers, pumps, arxi valves and (3) the operability of the systems as a whole and, under conditions as close to design as practical, the performance of the full operational sequence that brings the systems into operation, including operation of applicable portions of the protection system, the transfer between normal and emergency power sources, and the operation of associated systems.
Response
Testing of the Containment Spray System shall be conducted to assure structural and leaktight integrity, , p and operability and performance in accordance with Criterion 40. In addition, performance testing will i be conducted on all components of the Containment Spray System. These tests are normally conducted while the plant is operating. System design includes provisions which allow component testing with sufficient safeguards to prevent accidental containment spray. See Sections 6.2.2 and 6.5 for details. . The Containment flydrogen Recombiner System is designed to permit periodic testing for structural and leaktight integrity of components and for operability of the system and individual components. Testing may be conducted during normal plant operation or shutdown. See Section 6.2.5 for details. 3.1.40 Criterion 44 - Cooling Water A system to transfer heat from structures, systems, and components important to safety, to an ultimate heat sink shall be provided. The system safety function shall be to transfer the combined heat load of these structures, systems, and components under normal operating and accident conditions. I Suitable redundancy in components and features, and suitable interconnections, leak detection, and isolation capabilities shall be provided to assure that for onsite electrical power system operation (assuming offsite power is not available) and for offsite electrical power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure. Responset The cooling water systems which function to remove the combined heat load from structures, systems, and components important to safety under normal operating and accident conditions are the Component Cooling Water System and the Station Service Water System. The Component Cooling Water System l \ is a closed loop system which removes heat from nuclear safety related and potentially radioactive ! systems. The Station Service Water System removes heat from the Component Cooling Water System Amend ouir noww.onie or ssc rap a.s.n
System 80+ Design ControlDocument and transfers it to the atmosphere through the Ultimate Heat Sink. The Station Service Water System is described in Section 9.2.1 and the Component Cooling Water System is described in Section 9.2.2. 3.1.41 Criterion 45 - Inspection of Cooling Water System The cooling water system shall be designed to permit appropriate periodic inspection of important components, such as heat exchangers and piping, to assure the integrity and capability of the system.
Response
The important components are located in accessible areas. These components have suitable manholes, handholes, inspection ports, or other appropriate design and layout features to allow periodic inspection. See Sections 9.2.1 and 9.2.2 for details. 3.1.42 Criterion 46 - Testing of Cooling Water System The cooling water system shall be designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leak-tight integrity of its components, (2) the operability and the performance of the active components of the system, and (3) the operability of the system as a whole, and under conditions as close to design as practical, the performance of the full operational sequence that brings the system into operation for reactor shutdown and for loss-of-coolant-accidents, including operation of applicable portions of the protection system and the transfer between normal and emergency power sources.
Response
The design provides for periodic testing of active components of the cooling water systems for operability and functional performance. Preoperational pe.4ormance tests of the components are required to be made by the manufacturer. An initial system flow test demonstrates proper functioning of the system. Thereafter, periodic tests ensure that components are functioning properly. Cooling water system valves may be connected to the preferred power source at any time during reactor operation to demonstrate operability. Many active components are operated normally, thereby demonstrating operability. Remotely operated valves are exercised and actuation circuits tested. The automatic actuation circuitry, valves, and pump breakers also may be checked when integrated system tests are performed during a planned cooldown of the Reactor Coolant System. Provisions have been made to permit periodic leakage tests to verify the continued leak-tight integrity of the systems. Reftr to Sections 9.2.1 and 9.2.2 for details. 3.1.43 Criterion 50 - Containment Design Basis The reactor containment structure, including access openings, penetrations, and the containment heat removal system shall be designed so that the containment structure and its internal compartments can accommodate, without exceeding the design leakage rate and, with sufficient margin, the calculated pressure and temperature conditions resulting from any LOCA. This margin shall reflect consideration of (1) the effects of potential energy sources which have not been included in the determination of the peak conditions, such as energy in steam generators and energy from metal-water and other chemical reactions that may result from degradation but not total failure of emergency core cooling functioning, Apoved Deshipre Material Desiger of SSC Page 3.1-24
System 80+ Design ControlDocument A Q (2) the limited experience and experimental data available for defm' ing accident phenomena and containment responses, and (3) the conservatism of the calculational model and input parameters.
Response
The containment structure, including access openings and penetrations, is designed to accommodate, without exceeding the design leak rate, the transient peak pressure and temperature associated with a LOCA up to and including a double <nded rupture of the largest reactor coolant pipe. The containment structure and Engineered Safety Feature systems have been evaluated for various combinations of energy release. The analysis accounts for system thermal and chemical energy, and for nuclear decay heat. The Safety injection System is designed such that no single active failure could result in significant metal-water reaction (see Section 6.2.1). 3.1.44 Criterion 51 - Fracture Pmention of Containment Pressure Boundary The reactor containment boundary shall be designed with sufficient margin to assure that under operating, maintenance, testing, and postulated accident conditions (1) its ferritic materials behave in a nonbrittle manner, and (2) the probability of rapidly propagating fracture is minimized. The design shall reflect consideration of service temperatures and other conditions of the containment boundary material during operation, maintenance, testing and postulated accident conditions, and the uncertainties in determining (1) material properties (2) residual, steady-state, and transient stresses, and (3) size of flaws. Q Respee-(V The material selected for the containment vessel is SA-537 Class 2. The actual mechanical and chemical properties of the material are within the limits of minimum ductility defined in the 1989 ASME Code Material Specifications Part A for SA-537/SA-537M. ) The contaimnent vessel is built to Subsection NE of Section III of the ASME Boiler and Pressure Vessel Code. l The design of the vessel reflects consideration of all ranges of temperature and loading conditions which apply to the vessel during operation, maintenance, testing and postulated accident conditions. All seam welds in the vessel are 100 percent radiographed, and the acceptance standards of the radiographs ensure that flaws in welds do not exceed the maximum allowed by the ASME Code. 1 Steady state arx! transient stresses are calculated in accordance with accepted methods (see Section 3.8). 3.1.45 Criterion 52 - Capability for Containment Leakage Rate Testmg The reactor containment and other equipment which may be subjected to containment test conditions shall ! be designed so that periodic integrated lealage rate testing can be conducted at containment design pressure. l
.";;;;.::Denker nieterne! Desiprr of SSC Page 3.125
System 80+ Design ControlDocument
Response
The containment vessel is designed so that integrated leak rate testing can be performed at design pressure after completion and installation of penetrations and equipment in accordance with the requirement of Appendix J of 10 CFR 50 (see Section 6.2.6). 3.1.46 Criterion 53 - Provisions for Containment Testing and Inspection The reactor containment shall be designed to permit (1) appropriate periodic inspection of all important areas, such as penetrations, (2) an appropriate surveillance program, and (3) periodic testing at containment design pressure of the leak-tightness of penetrations which have resilient seals and expansion bellows.
Response
The absence of insulation on the containment vessel permits periodic inspection of the exposed surfaces of the vessel. The lower portions of the containment vessel are totally encased in concrete and will not be accessible for inspection. It is contemplated that there will be no need for any special in-service surveillance program due to the rigorous design, fabrication, inspection and pressure testing the containment vessel receives prior to operation. Provisions are made to permit periodic testing at containment design pressure of penetrations which have resilient seals or expansion bellows to allow leak-tightness to be demonstrated (refer to Section 6.2.6). 3.1.47 Criterion 54 - Piping Systerns Penetrating Containment Piping systems penetrating primary reactor containment shall be provided with leak detection, isolation, and containment capabilities having redundancy, reliability, and performance capabilities which reflect the importance to safety of isolating these piping systems. Such piping systems shall be designed with a capability to test periodically the operability of the isolation valves and associated apparatus and to determine if valve leakage is within acceptable limits.
Response
l Piping systems which penetrate containment are designed to provide the required isolation and testing capabilities. These piping systems are provided with test connections to allow periodic leak detection tests to be performed, in accordance with 10 CFR 50, Appendix J. The Engineered Safety Features Actuation System circuitry provides the means for testing isolation valve operability. For a discussion of penetration design, refer to Section 6.2.4, Containment Isolation System. For additional related discussion, see the responses to General Design Criteria 55,56, and 57 (Sections 3.1.48 through 3.1.50). O' Approved Des / pre Atatorial. Designs of SSC (2/95) Page 3.1-26 ; l
-- - . . . . - - ~ - .- .- . - - . - - - -
t System 80+ ~ Desian controlDocument , 3.1.48 Critedon 55 - Reactor Coolant Pnssure Boundary Penetrating Contmenment Each line that is part 'of the' reactor coolant pressure boundary and that penetrates primary reactor l containment shall be provided with containment isolation valves as follows, unless it can be demonstrated that the containment isolation provisions for a specific class of lines, such as instrument lines, are .r acceptable on some other defined basis: l e One locked closed isolation valve inside and one locked closed isolation valve outside containment; or [ e one automatic isolation valve inside and one locked closed isolation valve outside containment; or
- e one locked closed isolation valve inside and one automatic isolation valve outside containment.
4 A simple check valve may not be used as the automatic isolation valve outside containment; or , i e one automatic isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment. s isolation valves outside containment shall be located as close to containment as practical and upon loss of actuating power, automatic isolation valves shall be designed to take the position that provides greater safety. . Other appropriate requirements to minimize the probability or consequences of an accidental rupture of I these lines or of lines connected to them shall be provided as necessary to assure adequate safety. . Determination of the appropriateness of these requirements, such as higher quality in design, fabrication, 7 a and testing, additional provisions for in-service inspection, protection against more severe natural
- phenomena, and additional isolation valves and containment, shall include consideration of the population density and physical characteristics of the site environs.
l )
Response
7 The System 80+ reactor coolant system pressure boundary is defined in accordarvvith ANSI /ANS 51.1 4- and 10 CFR 50, Section 50.2(v). All reactor coolant pressure boundary lines penetrating containment meet the isolation criteria of GDC 55 using the following basis for specific lines in addition to those noted above. ] e Safety injection lines, as shown on Figures 6.3.2-1 A, 6.3.2-1B, and 6.3.2-IC, are used to mitigate the consequences of accidents and therefore do not receive an automatic closure signal and are not locked closed, e When in the shutdown cooling mode of operation the Shutdown Cooling System is an extension of the reactor coolant pressure boundary. In this mode the system is isolated from the environment by two isolation valves in series. e The charging and seal injection lines shown on Figure 9.3.4-1 have automatic valves outside the containment which do not receive a closure signal (CIAS). This is because it is desirable to maintain charging and seal injection flow as long as the charging pumps are in operation. 4prowsef Des (pe A8sewW Des (pr of SSC pape 3.1-27 ,
- . - , . ,.-e , + , - .
System 80+ Design controlDocument i 3.1.49 Criterion 56 - Primary Containment Isolation Each line that connects directly to the containment atmosphere and penetrates primary reactor containment shall be provided with containment isolation valves as follows, unless it can be demonstrated that the containment isolation provisions for a specific class of lines, such as instnunent lines, are acceptable on some other defined basis: e One locked closed isolation valve inside and one locked closed isolation valve outside containment; or e One automatic isolation valve inside and one locked closed isolation valve outside containment; or e One locked closed isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment, or
- One automatic isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment.
Isolation valves outside ontainment shall be located as close to the containment as practical and upon loss of actuating power, ar er.atic isolation valves shall be designed to take the position that provides greater safety.
Response
System 80+ fluid systems comply with the requirements of GDC 56 with the following exceptions: Lines which connect directly to the containment atmosphere and are used for mitigating the effects of accidents are connected to a closed piping system outside containment, which is isolated from the l environment in accordance with the requirements of GDC 55. In addition, the capability far remote 1 double isolation at the containment boundary is provided in accordance with GDC 56. 3.1.50 Criterion 57 - Closed System Isolation Valves i Each line that penetrates primary reactor containment and is neither part of the icaor coolant pressure boundary nor connected directly to the containment atmosphere shall have at least one contduraent isolation valve which shall be either automatic, or locked closed, or capable of remote nanual operabon. ! This valve shall be outside containment rat iocated as close to the containment as practical. A simple check valve may not be used as the automatie isolation valve. l l
Response
1 The systems that fall into the category described in GDC 57 comply with containment isolation ! requirements as specified in the containment isolation system sections of Chapter 3. l l 3.1.51 Criterion 60 - Control of Releases of Radioactive Material to the Environment The nuclear power unit design shall include means to control suitably the release of radioactive materials in gaseous and liquid effluents and to handle radioactive solid wastes produced dming normal reactor operation, including Anticipated Operational Occurrences. Sufficient holdup capacity shall be provided novemt o*sion M*ww- Dessen or ssc 11 tam rose 2.rce
. - - . - - - _ - . - - .. - .- - ~- .. . -
System 80+ Desion CoMrol Document . i f i for mention of gaseous and liquid effluents containing radioactive materials, particularly where ! unfe nible site environmental conditions can be expected to impose unusual operational limitations upon the resease of such effluents to the environment. ;
Response
i I The sources and expected quantities of radioactive materials produced during normal reactor operation, j
~ including Anticipated Operational Occurrences, is presented in Chapter 11. The radioactive waste l . systems to suitably control the release of these materials in gaseous and liquid effluents and to handle radioactive solid wastes are described in Sections 11.2 through 11.4. .l 1 . 3.1.52 Criter11on 61 - Fuel Storage and Handling and Radioactivity Control j The fuel storage and handling, radioactive waste, and other systems which may contain radioactivity shall f
. be designed to assure adequate safety under norrnal and postulated accident conditions. These systems ! shall be designed (1) with a capability to permit appropriate periodic inspection and testing of components l important to safety, (2) with suitable shielding for radiation protection, (3) with appropriate containment, ! confinement, and filtering systems, (4) with a residual heat removal capability having reliability and l testability that reflects the importance to safety of decay heat and other residual heat removal, and (5) to j prevent significant reduction in fuel storage coolant inventory under accident conditions, j
- Response
- - !
J Fuel storage and handling and fuel pool cooling are discussed in Section 9.1. Most of the components ! and systems in this category are in frequent use and no special testing is required. Those systems and j > components important to safety that are not normally operating are tested periodically; e.g., the fuel , handling equipment (prior to each refueling).
- The spent fuel storage racks are located to provide sufficient shielding water over stored fuel assemblies i to limit radiation at the surface of the water to no more than 2.5 mrem /hr during the storage period. The l exposure time during refueling is limited so that the integrated dose to operating personnel does not i l
exceed the limits of 10 CFR 20. The accidental release of the maximum activity content of a gas decay i tank will not result in doses in excess of 500 mrem whole body. See Chapter 11 for details. f Cooling for the spent fuel pools is designed to prevent damage to fuel in the storage facilities that could result in radioactivity release to the plant operating areas or the plant environs. 3.1.53 Criterion 62 - Prevention of Criticality in Fuel Storage and Handling j Criticality in the fuel storage and handling system shall be prevented by physical systems or processes, ! . preferably by use of geometrically safe configurations.
Response
Both new and spent fuel assemblies are stored in parallel rows designed in accordance with ANSl/ANS 51.1. ; Normal procedures require that new fuel be stored in dry air or fully submerged in borated water,
' and that spent fuel be stored fully submerged in borated water.
The fuel storage and handling system is described in Section 9.1. Ammw necon anneuw.ne on er ssc - rene 2.12s
System 80+ Design ControlDocument Design of the new and spent fuel racks assures a kmof less than 0.98 for the new and 0.95 for the spent fuel assemblies. 3.1.54 Critenon 63 - Monitoring Fuel and Waste Storage Appropriate systems shall be provided in fuel storage and radioactive waste systems and associated handling areas (1) to detect conditions that may result in loss of residual heat removal capability and excessive radiation levels and (2) to initiate appropriate safety actions.
Response
Instrumentation is provided in the Pool Cooling and Purification System which will detect a loss of residual heat removal capability. Appropriate safety actions are initiated by operator responses. The instrumentation and system relationships are discussed in Section 9.1. Refer to Section 9.1 for a discussion of Fuel Storage and Handling and to Chapter 11 for a discussion of the area and ventilation system radiation monitoring. 3.1.55 Critedon 64 - Monitoring Radioactivity Releases Means shall be provided for monitoring the reactor containment atmosphere, spaces containing components for recirculation of loss-of-coolant-accident fluids, effluent discharge paths, and the plant environs for radioactivity that may be released from normal operations, including Anticipated Operational Occurrences, and from postulated accidents.
Response
Provisions are made for monitoring the contairunent atmosphere, the facility effluent discharge paths, the operating areas within the plant and the facility environs for radioactivity that could be released from normal operation, from anticipated transients, and from an accident. Some liquid and gaseous effluent will contain radioactive matter. The Waste Management System functions to remove radioactive material from these wastes by filtration and ion exchange prior to discharge. Liquid wastes are sampled, and if the contained activity meets applicable limits, they may be released with continuous radiation monitoring to the plant dilution flow canal. Gaseous wastes are processed through carbon absorbers to retain and delay radioactive fission gases prior to release. The gas is then monitored during release through the plant vent. The Condenser Air Removal System discharge is monitored for gaseous activity. The ventilation system discharges are monitored for gaseous activity. Radioactive waste management and monitoring as well ar, area monitoring are discussed in Chapter 11. O Approved Design Atatedel Design of SSC Page 3 r.30
i System 80+ onelan contrat Document ;
, \
}
3.2 Clamalfication of Structures, Connponents, and Systems j [- j j 3.2.1 hi==Re Classifiestion ' i i Structures, systems, and components which are important to safety and designed to remain funct:onal in ; the event of a Safe Shutdown Earthquake (SSE) are designated as Seismic Category I. Seismic Category I structures, systems, and components are those r.acssary to ensure: j
- l
- The integrity of the reactor coolant pressure boundary; l 1 .
- The capability to achieve safe shutdown of the reactor and keep it in a safe shutdown condition; l ;
] and j i
*~ The capability to prevent or mitigate the conseque .=. of accidents that could result in potential l f 4- offsite exposures in excess of 10 CFR 100 guidelines, j i
i p The selection of Category I structures, systems, and components is in accordance with the definition ! E above and the guidance provided by Regulatory Guide 1.29. Individual components in Category I j systems are classified by reference to the safety classes assigned in accordance with ANSI /ANS 51.1 (see !
. Section 3.2.2). ' All components in Safety Classes 1,2, and 3 are Seismic Category I. l i
Structures, systems and components which do not perform a nuclear safety related function and whose l continued function is not required are classified Non-Nuclear Safety (NNS) (see Section 3.2.2). NNS i structures, systems and components whose structural failure or interaction could degrade the functioning l of a Seismic Category I structure, system, or component to an unacceptable safety level or could result in an incapacitating injury to an occupant of the control room are designated as Seismic Category II and , are designed and constructed so that the SSE will not cause such failure in a manner that would adversely ~ l e , affect a safety system. Structures, systems, and equipment which have no enhanced seismic design requirements in addition to those imposed by building codes are designated Non-Seismic (NS). The seismic category and safety and quality classification of structures, systems, and components within the System 80+ Standard Design are listed in Table 3.2-1. The safety class is also shown on the P& ids ~ (Chapters 5, 6, and 9). Seismic Category I includes all mechanical components within the safety class . - boundaries and extends to the first seismic restraint beyond the boundary. All fuel racks are also i designated as Scismic Category I. Structures, systems, or components whose failure could reduce the 1 performance of a safety function by a Seismic Category I component to an unacceptable safety level are designed to Seismic Category II requirements for structural integrity only or are separated to the extent required to eliminate that possibility.' NNS structures, systems or components whose failure could cause flooding of safety-related structures, systems or components are designed to Seismic Category I requirements. This ensures that any structures, systems, or components that could potentially have a disabling interaction with Seismic Category I structures, systems, or components are either prevented from doing.so or are designed.to meet Seismic Category I or 11 structural integrity requirements, 4 depending on the function of the component. Structural integrity requirements may be demonstrated by dynamic or equivalent static analyses, testing, or a combination thereof. Analyses of Seismic Category II structures, systems, and components are in accordance with the seismic input and methodology criteria described in Sections 3.7.2 and 3.7.3. m e o > asses,w ca.> er sse szess esp. u -r
- . -- . . .. - - . . . - . ~ - .- - - -
System 80+ Design ControlDocurnent The listing of major elects;al components is found in Section 3.11, which also includes safety and quality classifications. Electrical structures, systems, and components not classified as Seismic Category I but whose failure could represent a hazard to the operator or could interfere with the performance of required safety functions of electrical stmetures, systems and components, are classified as Seismic Category II (Reference 1). Any electrical system or structure or component not in Seismic Category I or II is considered Non-Seismic (see Section 3.10). The use of the Seismic Category II designation for electrical components is limited to non-safety control system components which are designed and documented to maintain structural integrity during an SSE. Piping supports and component suppons are of the same seismic category as the piping and components to which they apply. Instrument sensing lines and their suppons are designed in accordance with the seismic category criteria of Regulatory Guida 1.151. For purposes of this discussion, the motors and solenoids used to provide motive power to mechanical components are treated as part of the mechanical component. 3.2.2 System Quality Group Classifications (Safety Class) In general, fluid system components imponant to safety are classified in accordance with ANSI /ANS 51.1 (Reference 2). With the exception of portions of the Chemical and Volume Control System, Safety Class 1,2,3 and NNS of ANSI /ANS 51.1 are equivalent to Quality Groups A, B, C and D of Regulatory Guide 1.26. Portions of the CVCS are designated as Safety Class NNS but are designed and constructed to Quality Group C standards. The criteria establish safety classes which are used as a guide to the selection of codes, standards, and quality assurance provisions for the design and construction of the components. The safety class designations are also used as a guide to those fluid system components to be classified as Seismic Category I and II (see Section 3.2.1). The Safety Class definitions in ANSI /ANS 51.1 are sununarized as follows:
- Safety Class 1 (SC-1) applies to pressure-retaining ponions and supports of mechanical equipment that form pan of the RCPB whose failure could cause a loss of reactor coolant in excess of the reactor coolant nonnal makeup capability and whose requirements are within the scope of the ASME Boiler and Pressure Vessel Code, Section Ill.
- Safety Class 2 (SC-2) applies to pressure-retaining portions and supports of primary containment and other mechanical equipment, requirements for which are within the scope of the ASME Boiler and Pressure Vessel Code, Section III, that are not included in SC-1 and are designed and relied upon to accomplish the nuclear safety functions defined in ANSI /ANS 51.1, Section 3.3.1.2.
- Safety Class 3 (SC-3) applies to equipment, not included in SC-1 or -2, that is designed and relied upon to accomplish the nuclear safety functions defined in ANSI /ANS 51.1, Section 3.3.1.3.
- Non-Nuclear Safety (NNS) applies to equipment that is not in Safety Class 1, 2, or 3. This equipment is not relied upon to perform a nuclear safety function.
The safety classifications of major components which are in the System 80+ design scope are listed in Table 3.2-1. Seismic category designations and quality assurance requirements are also included. Safety classes and safety class changes are shown on the system P& ids. Safety Class 1,2, and 3 valves are listed in Table 3.2-2. Approved Design MaterM Design of SSC Page 32-2
i System 80 + ' oesten contrat Document 4 Instrument sensing lines and their supports are designed in accordance with the ASME code class
. requirements of Regulatory Guide 1.151. .
All pressure containing components in Safety Classes 1,2, and 3 are designed, manufactured, and tested i j in accordance with the rules of the ASME Boiler and Pressure Vessel Code, Section III. Components i f designated NNS are designed and constmeted with appropriate consideration of the intended service using
- applicable industry codes and standards. The relationship between safety class and code class is shown f in Table 3.2-3. A higher code class may be used for a component without changing the safety class or affecting the balance of the system in which it is located.
Fracture toughness requirements are imposed on materials for pressure retaining parts of ASME Class i 2 and 3 System 80+ Standard Design components. Test methods, acceptance, and exemption criteria
- are in conformance with the ASME Code, Section III.
The safety classification system is also used to identify those components to which the requirements of 10 CFR $0, Appendix B, are applicable. Components in Safety Classes 1,2, and 3 are designed and ! manufactured under a rigorous quality assurance program reflecting the requirements of 10CFR50, Appendix B and are designated under the Quality Class Category in Table 3.2-1, Components which do j j not serve a safety- related function are not subject to the quality assurance requirements of 10CFR50, Appendix B and are designated in Table 3.2-1. The Quality Assurance Program is described in Chapter
- 17. j Piping supports and component supports are of the same safety class and have the same QA requirements !
as the piping and components to which they apply. i The use of the above outlined safety and quality classification systems meets the intent of Regulatory
- Guide 1.26 and the requirements of 10 CFR 50.55a.
References for Section 3.2 4 1. " Seismic Qualification of C-E Instrumentation Equipment," Combustion Engineering, Inc., l CENPD-182, Revision 1, May 1977. l f
- 2. " Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants,"
ANSI /ANS 51.1.1983. l i a 4 r I j I t
)
O : 1 Amsment assen Asseener.$wys of SSC - Pape 3.2-3 :
System 80+ Design ControlDocument Table 3.2-1 Classification of Structures, Systems, and Components h Safety Seismic Quality Component Identification Class Category Locationt2sj* Classt2n* l Reactor Coolant System Reactor Vessel i I RC 1 Steam Generators (primary / secondary) 1/2 [1]* I RC 1 Pressurizer 1 I RC 1 l Reactor Coolant Pumps [2,3,9]* 1 I RC 1 Piping within Reactor Coolant Pressure 1/2 [4] I RC 1 Boundary [5] Control Element Drive Mechanisms [6] [6] RC 1 Core Support Structures and 3 i RC 1 laternals Structures [7] Fuel Assemblies [8] 2 i RC 1 Control Element Assemblies [8] 3 I RC 1 Closure 11ead Lift Rig NNS 11 [1 0] RC 2 l licated Junction Thermocouple Probe 1/3 [12] I RC 1 Assembly ll1TC Pressure liout.ing 1 I RC 1 ICI Cable Tray Suppon Frame 3 I RC 1 ICI 11olding Frame NNS NS RC 3 ICI Guide Tubes 1 I RC 1 ICI Guide Tube Supports 1 I RC 1 ICI Seal 11ousing 1 I RC 1 ICI Seal Table 1 I RC 1 Piping [27] I/2 I RC 1 Valves [27] 1/2 i RC 1 In-containment Water Storage System IRWST 3 I RC 1 lloldup Volume Tank 3 I RC 1 Pressure Relief Dampers 3 I RC 1 Cavity Flooding System Piping 2 I RC 1 Valves 2 I RC 1 Safety Depressurization System l Valves 1/2/NNS I/II RC 1/2 Piping 1/2/NNS 1/11 RC 1/2 l Spargers NNS 11 RC 2 Safety Irdection System Safety injection Pumps 2 i RB 1 Safety injection Tanks 2 1 RC 1 l Piping [24,27] 1/2 1 RB/RC 1 Valves [27] 1/2 1 RB/RC 1 O Refer to Notes at end of table. Aptwowd Desbyrs Meternia! Desspre of SSC (11/96) Page 3.2-4
System 80+ oesias controlDocument A O Table 3.2-1 Classification of Structures, Systerns, and Components (Cont'd.) l Safety Seismic Quality Component Identification Class Category Locationt25) Classt2,1 l Shutdown Cooling System Shutdown Cooling Heat Exchangers 2/3 [1] I RB 1 Shutdown Cooling Pumps 2 I RB 1 Shutdown Cooling Mini-Flow Heat 2/3 [1] I RB 1 Exchanger Piping [27] 1/2/3 I RB/RC 1 Valves [27] 1/2/3 I RB/RC 1 Containment Spray System Containment Spray Pumps 2 I RB 1 Contamment Spray Heat Exchangers 2/3 [1] I RB 1 Containment Spray Mini-Flow Heat 2/3 [1] I RB 1 l Exchanger Spray Nozzles 2 I RC 1 Piping [27] 2/3 I RB/RC 1 Valves [27] 2 I RB/RC 1 Chemical and Volume Control System (CVCS) Regenerative Heat Exchanger 2 I RC 1 q) Letdown Heat Exchanger 2/NNS [1,34] I I RC NA 1 2 Seal Injection Heat Exchanger NNS [34] Purification Ion Exchangers NNS [34] I NA 2 4 Deborating Ion Exchanger NNS [34] I NA 2 Volume Control Tank NNS [34] I NA 2 Chemical Addition Package NNS NS NA 3 Boric Acid Batching Tank NNS NS NA 3 Charging Pumps NNS [34] I NA 2 Dedicated Seal Injection Pump NNS [34] I NA 2 Dedicated Seal Injection Pump NNS [34] I NA 2 Suction Stabilizer / Pulsation Dampener Boric Acid Makeup Pumps NNS [34] I NA 2 Reactor Makeup Water Pumps NNS NS NA 3 Boric Acid Concentrator NNS NS NA 2 Pre-holdup lon Exchanger NNS [34] I NA 2 Charging Pump Mini-flow Heat NNS [34] I NA 2 Exchanger ! Boric Acid Condensate Ion Exchanger NNS NS NA 2 I Reactor Drain Pumps NNS [34] I NA 2 Holdup Pumps NNS NS NA 3 j Reactor Drain Tank NNS NS RC 2 Holdup Tank NNS NS YA 2 Equipment Drain Tank NNS [34] I NA 2
)
Reactor Makeup Water Tank NNS NS YA 2 ' [L . Gas Stripper Purification Filters NNS [34] NNS [34] I I NA NA 2 2 j i L..= ouse neeeuw owen or ssc tznsi rare 12-5 l
System 80+ oesign controlDocument Table 3.2-1 Classification of Structures, Systems, and Cornponents (Cont'd.) Safety Seismic Quality Component Identification Class Category Locationt2sj Classt29) l CVCS (Cont'd.) Reactor Drain Filter NNS [34] I NA 2 Seal Injection Filters NNS [34J I NA 2 Reactor Makeup Water Filter NNS NS NA Boric Acid Filter NNS [34] I NA Letdown Strainer NNS [34] I NA - Pre-holdup Strainer NNS [34] I NA 2 Boric Acid Condensate IX Strainer NNS NS NA 3 lon Exchanger Drain IIcader Strainer NNS NS NA 3 Boric Acid Batching Strainer NNS NS NA 3 Chemical Addition Strainer NNS NS NA 3 Boric Acid Storage Tank [33] NNS [34] I YA 2 Boric Acid Batching Eductor NNS NS NA 2 letdown Orifices 2 I RC 1 Piping [27] 1/2/3/NNS [35] I/NS RC/NA/YA 1/2 Valves [27] 1/2/3/NNS [35] I/NS RC/NA/YA 1/2 Emergency Feedwater System Cavitating Venturi 2 I RC 1 Motor-Driven Emergency Feedwater 3 I RB 1 Pumps Steam-Driven Emergency Feedwater 3 I RB 1 Pumps Emergency Feedwater Pump Turbines 3 I RB 1 Emergency Feedwater Storage Tanks 3 I NA 1 Piping [27] 2/3 I NA/RB/RC 1 Valves [27] 2/3 I NA/RB/RC 1 Fuel llandling System Refueling Machine NNS II RC 2 Fuel Transfer System NNS 11 RC/NA 2
- 1. Transfer Carriage NNS Il RC/NA 2
- 2. Upending Machine NNS Il RC/NA 2
- 3. Hydraulic Power Unit NNS II RC/NA 2 Fuel Transfer Tube, Valve, Stand NNS !I RC/NA 2 CEA Change Platform NNS 11 RC 2 Long and Short Fuel llandling Tools NNS NS RC/NA 3 Upper Guide Structure Lifting Rig NNS 11 [1 1] RC 2 l Core Barrel Lifting Rig NNS 11 [1 1] RC 2 Spent Fuel llandling Machine NNS 11 NA 2 New Fuel Elevator NNS 11 NA 2 Underwater Television NNS NS RC/NA 3 Refueling Pool Seal NNS NS RC 2 In-Core Instrumentation and CEA NNS NS RC 3 Cutter Extension Shaft Uncoupling Tool NNS NS RC 3 Fuel Transfer Tube Quick Closure 2 I RC 2 Approwd Desierr Mater 6al Des / pre of SSC (2/95) Page 3.2-6
-_-.~,
' l i i ! System 80+ Denlan ControlDocarmrt I . l Table 3.2-1 Classification of Structures, Systems, and Components (Cont'd.) ! e i i Safety - Seismic Quality Component hwication Class ' Category LocationIM ClassIM l l Feel Handling System (Cont'd.). , CEA Handling Tools NNS NS RC -3 : ICI Insertion and Removal Tools NNS .NS RC 3 ; Spent Fuel Racks 3 1 NA 1 j New Fuel Racks 3 I NA 1 l l Coodsmante and Feedwater Systesa ! Condensate Pumps NNS NS TB 2- -l Feedwater Pumps NNS NS TB 2 l Feedwater Pump Controllers NNS NS TB 2 l Feedwater Booster Pumps NNS NS TB 2 ! Startup Feedwater Pump NNS NS TB 2 i Imv Pressure Feedwater Heaters NNS NS TB 2 l High Pressure Feedwater Heaters NNS 'NS TB 2 i ~ Demerator NNS NS ,TB 2 ; Piping (13) 2/NNS 1/NS TB/NA/RC/MS 1/2/3 ; Valves (13) 2/NNS 1/NS TB/NA/RC/MS IfF3 l Main Condenser System Main Condenser NNS NS TB 2 ! Che Storage e Systent l Condensate Storage Tanks NNS- NS YA 2 {' Condensate Storage Tank Recycle 11NS NS SB 2 Pumps ! Piping NNS NS YA/SB/TB 2/3 , Valves NNS NS YA/SBfrB 2/3 ! Ca=d====*e Cleanup System } 4 Piping . NNS NS TB 2/3 ! Polishers /Demmeralizers NNS NS TB 2 , < Resin Traps NNS NS TB 2 l Valves NNS NS TB 2/3 i Main Condenser Evacuation Systeen i Vacuum Pumps NNS NS TB 2 ; Piping NNS NS TB 2/3 l Valves NNS NS TB- 2/3 ; i Domineralised Water Makeup i Systeen (DWMS) i Dermneralizer Makeup Water Pumps NNS NS SB 3 ; Demineralizers NNS NS SB 3 l Vacuum Degasifier NNS NS SB 3 i Demineralized Water Storage Tank NNS NS YA 3 .
- Vacuum Pumps . NNS NS SB 3 !
>C Demineralizer Recycle Pump NNS NS SB 3'
- Vacuum Degasifier Transfer Pumps - NNS NS SB 3 !
Dominerclind Water Transfer Pumps NNS NS SB 3 Regenerant Waste Neutralizanon Tank - NNS NS SB 3 l 1. ( Anoment Dee@n aseenoder coe@n er ssC 42251 anpe 3.2-7
System 80+ Design ControlDocument Table 3.2-1 Classification of Structures, Systems, and Components (Cont'd.) Safety Seismic Quality Component identification Class Categan Locationr2sj Classt2 ,j l D%%1S (Cont'd.) Piping [27] 2/NNS I/NS A1: 1/3 Valves [27] 2/NNS I/NS All 1/3 Extraction Steam System Piping NNS NS TB 2 Valves NNS NS TB 2 IIcater Vcn;s Piping NNS NS TB 2 VMyes NNS NS TB 2 Turbine Generator System Turbine Generator High Pressure Turbine NNS NS TB 2 Low Pressure Turbines NNS NS TB 2 Generator NNS NS TB 2 Moisture Separators NNS NS TB 2 Steam Reheaters NNS NS TB 2 Stop Valves NNS NS TB 2 Control Valves NNS NS TB 2 Reheat Stop Valves NNS NS TB 2 Intercept Valves NNS NS TB 2 Valves, other NNS NS TB 2/3 Piping NNS NS TB 2/3 Turbine Bypass System Turbine Bypass Valves NNS NS TB 2 Valves, other NNS NS TB 2 Piping NNS NS TB 2 Turbine Gland Sealing System Gland Seal Condenser NNS NS TB 2 Gland Seal Regulator NNS NS TB 2 Piping NNS NS TB 2 Valves NNS NS TB 2 Turbine Lube Oil System Putups NNS NS TB 2 OilTank NNS NS T6 2 Oil Turbine NNS NS TB 2 Oil Coolers NNS NS TB 2 Oil Filters NNS NS TB 2 Piping NNS NS TB 2/3 Valves NNS NS TB 2/3 Turbine Control System EllC Pumps NNS NS TB 2 EllC Coolers NNS NS TB 2 EHC Sumps NNS NS TB 2 Approwd Desierr MaterW - Desigre of SSC (2/95) Page 3.2-8
System 80+ ossion contro: Document I Table 3.2-1 Classification of Structures, Systems, and Components (Cont'd.) L Safety Seismic Quality Component Identification Class Category LocationIM ClassIM l Turbine Control System (Cont'd.) Pipinr, NNS NS TB 2 Valves NNS NS TB 2 Turbine Generator Cooling System Hydrogen Coolers NNS NS TB 2 Piping NNS NS TB 2 . Valves NNS NS TB 2 Liquid Waste Management System Waste Collection Tanks NNS NS RW 2 Waste Sample Tanks NNS NS RW 2 Process Pumps NNS NS RW 2 NNS NS RW 2 Process Deminer& rs Process Filters NN9 NS RW 2 Piping [27] 2/NINS 1/NS TB/NA/RW 1/2 RC/RB Valves [27] 2/NNS I/NS TB/NA/RW 1/2 RC/RB O) (V Gaseous Waste Management System Gas Coolers / Condenser NNS NS NA 2 ] I Guard / Charcoal Beds NNS NS NA 2 Piping [27] 2/NNS I/NS NA/RC 1/2 l 2/NNS 1/NS NA/RC 1/2 Valves [27] l Solid Waste Management System Spent Resin Transfer Pumps NNS NS NA/RW 2 Spent Resin Tanks NNS NS NA/RW 2 filC Fill / Dewatering Head NNS NS RW 2 Resin Forwarding Pumps NNS NS RW 2 Dry Solids Compactor NNS NS RW 2 Piping NNS NS NA/RW 2 Valves NNS NS NA/RW 2 Heater Drain System Reheater Drain Tanks NNS NS TB 2 Moisture Separator Drain Tanks NNS NS TB 2 lleater Drain Tank NNS NS TB 2 Ileater Drain Pumps NNS NS TB 2 Piping NNS NS TB 2/3 l Valves NNS NS TB 2/3 Process and Emuent Radiation Monitoring System (PERMS) i Gaseous Process and Effluent CN Monitors V Unit Vent Waste Gas NNS NNS NS NS NA RW 2 2 j (2as) rege 12-s L..:: ouien neenwset outen of ssc l l l
System 80+ Design controlDocument Table 3.2-1 Classification of Structures, Systems, and Cornponents (Cont'd.) Safety Seismic Quality Component Identification Class Category Locationt2s Classt2n l , PERMS (Cont'd.) Unit Vent Post-Accident NNS N3 NA 2 Containment Purge Exhaust NNS NS NA 2 Condenser Air Ejector NNS NS TB 2 Liquid Process and Effluent Monitors Component Cooling Water NNS NS NA 2 Liquid Waste Discharge NNS NS RW 2 Plant Discharge Line NNS NS RW 2 Station Service Water NNS NS CX 2 Reactor Coolant Gross Activity NNS NS NA 2 Turbine Building Drains NNS NS TB 2 Steam Generator Blowdown NNS NS TB 2 Airborne Radiation Monitors Containment Atmosphere 3 I NA 1 Nuclear Annex NNS NS NA 2 Radwaste Building NNS NS RW 2 Fuel Building NNS NS NA 2 Ventilation Systems Multisampler NNS NS NA 2 Control Room Intake (A&B) 3 I NA 1 Reactor Building Annulus NNS NS NA 2 Subsphere Ventilation NNS NS NA 2 Area Radiation Monitoni NNS NS RC/NA/RW 2 Special Purpose Area Monitors Main Steam Line NNS NS NA 2 Purification Filter NNS NS NA 2 Contamment Arca liigh Radiation 3 I RC 1 Primary Coolant 3 I RC 1 Containment Isolation System Piping 2 I RC/RB 1 Valves 2 I RC/RB 1 Component Cooling Water System [14] lient Exchangers 3 I CX 1 Ngs 3 I NA 1 Surge Tanks 3 1 NA 1 Sump Pumps NNS NS NA 3 Chemical AdditionTank NNS NS NA 3 Ileat Exchanger Building Sump Pumps NNS NS CX 3 Piping [27] 2/3/NNS I/NS CX/YA/NA 1/2/3 RB/RC Valves 2/3/NNS 1/NS CX/YA/NA 1/2/3 RB/RC Approved Design atatorial Design of SSC (2/95) Page 3.2-10
' System 80 + ^ cosion controlDocument j
'\ Table 3.2-1 Classification of Structures, Systems, and Components (Cont'd.) I Safety Seismic Quality Component Identification Class Category Locationt251 Classt2 ,j l Spent Fuel Pool Cooling System Pumps 3 I NA 1 Exchangers 3 I NA 1 Piping 3/NNS I/NS NA 1/3 Valves 3/NNS I/NS NA 1/3 Pool Purificetion System Pumps NNS NS NA 2 Strainers NNS NS NA 3 Demineralizen, NNS NS NA 2 Filters NNS NS NA 2 Skimmer NNS NS NA 3 Piping [27] 2/3/NNS 1/NS NA/RC 1/2 , Valves [27] 2/3/NNS 1/NS NA/RC 1/2 Primary Sampling System Pump NNS NS NA 2 Heat Exchangers NNS NS NA 2 Sample Vessels NNS NS NA 2 N Piping [27] 2/3/NNS I/NS NA/RC 1/2 Valves [27] 2/3/NNS I/NS NA/RC 1/2 Sink NNS NS NA 3 Boronometer NNS NS NA 2 Process Radiation Monitor NNS NS NA 2 Secondary Chemistry Control Sampling System Heat Exchangers NNS NS NA 2/3 Strainers NNS NS NA 2/3 Monitors NNS NS NA 2/3 Piping [27] 2/NNS I/NS NA/RC 1/3 Valves [27] 2/NNS I/NS NA/RC 1/3 Station Service Water System Pumps 3 I SP 1 Strainers 3 I SP 1 Sump Pumps NNS NS SP 3 Traveling Screens 3 I YA 1 Piping 3/NNS 1/NS SP/CX 1/3 l Valves 3/NNS I/NS SP/CX 1/3 Turbine Building Service Water System Mp'mg NAS NS YA 2/3 Valves NNS NS YA 2/3 s Pumps NNS NS YA 2
). Strainers NNS NS YA 2 v
i
.% 2 % neeserw-Deeen of SSC W95) Page 3.2-11
i I l System 80+ Design contro/ Document l Table 3.2-1 Classification of Structures, Systems, and Components (Cont'd.) Safety Scismic Quality Component identification Class Category Locationt2sl Classt29) l Turbine Building Cooling Water System Piping NNS NS TB/YA 2/3 Valves NNS NS TB/YA 2/3 Heat Exchangers NNS NS YA 2/3 Pumps NNS NS TB 2/3 Surge Tank NNS NS TB 2/3 Chemical AdditionTank NNS NS TB 3 Essential Chilled Water System Refrigeration Units 3 1 NA 1 Pumps 3 I NA 1 Compression Tanks 3 I NA 1 Chemical AdditionTanks NNS NS NA 3 Essential / Normal Heat Exchangers 3/NNS [1] 1 NA 1/2 Piping [27] 2/3/NNS 1/NS NA/RC/RB 1/2/3 2/3/NNS 1/NS NA/RC/RB 1/2/3 Valves [27] Strainers 3/NNS I/NS NA 1/3 Normal Chilled Water System [IS) Refrigeration Units NNS NS NA 2 Pumps NNS NS NA 2 Compression Tanks NNS NS NA 3 Air Separators NNS NS NA 3 Chemical AdditionTanks NNS NS NA 3 Piping [27] 2/NNS 1/NS NA/RC 1/3 Valves [27] 2/NNS 1/NS NA/RC 1/3 Strainers NNS NS NA 3 Condenser Circulating Water System Pumps NNS NS YA 2 Cooling Towers (mechanical portion) NNS NS YA 2 Piping NNS NS YAfrB 2/3 Valves NNS NS YA/TB 2/3 Strainers NNS NS YAfrB 2 Traveling Screens NNS NS YA 2 Instrument Air System Air Compressors NNS NS NA 2 Piping [27] 2/NNS 1/NS All 1/3 Valves [27] 2/NNS 1/NS All 1/3 Air Receivers NNS NS NA 3 Desiccant Air Dryers / Filters NNS NS NA 2 O Approved Design Materief - Design of SSC (2/951 Page 3.2-12
f i System 80+ Desian contro/ Docarmut I- -i j. Table 3.2-1 < Classification of Structures, Systems, and Components (Cont'd.) j Safety Seismic Quality j < Cosopomaat Identification Class . Category I.mcationt2s Classt2n l
- St=*la= Air System Air Compressors NNS NS SB 3 +
Air Dryers / Filters NNS NS SB 3 Air Receivers NNS. NS SB 3 l 7 Piping [27] . 2/NNS I/NS All 1/3 j Valves [27] 2/NNS I/NS All 1/3 I Breathing Air Systent :* Air Compressors NNS NS SB 3 Piping [27] 2/NNS I/NS All 1/3 3' Valves [27] 2/NNS I/NS All 1/3 ) Air Receivers NNS NS SB 3 Air Dryer / Filters NNS NS SB 3 ! ] Casopressed Gas Systeses High Preuure Gas Cylinders NNS NS YA- 3 l 4 Pressure Regulators NNS NS YA 3 Leak Detection Systems NNS NS All 3 ! Liquid Nitrogen Evaporators NNS NS YA 3 i Piping [26, 27] 2/NNS I/NS All 1/3 l ; O Valves [27] Fire Protection System 2/NNS I/NS All 1/3 ! t Jockey Pump NNS NS FP 2 Backup Storage Tank NNS I NA 1 l Fire Pumps NNS NS FP 2 Backup Fire Pump NNS I NA 1 i l Storage Tanks NNS NS FB 2 Water Spray Systems (Deluge and 2/NNS I/II/NS TB/NA/RC/RB/ 1/2 Sprinkler) Piping, Valves [16,27] DG/SB : Hose Systems / Standpipes [16,27] 2/NNS I/NS All 1/2 ; Portable Fire Extinguishers [16] NNS NS All 2 Exterior Distribution System Piping NNS NS YA 2 : Valves NNS NS YA 2 L Strainers NNS NS YA 2 Akernate AC Source /Coimbustion NNS NS YA 2 i Turbine-Generator Esmergency Diesel Generator System Diesel Generators 3 I DG 1 DG Engine Fuel Oil Systen: [17] Fuel Oil Storage Tanka 3 I. DF 1 Recirculation Pumps NNS NS DF 3 O Booster Pumps Fuel Oil Day Te.aks 3 3 I I DG DG 1 1 , 4presed coetyn aseeww.see(pr er asc (rf/ss/ pope 3.2-f3
System 80+ Design ControlDocument Table 3.2-1 Classification of Structures, Systems, and Components (Cont'd.) Safety Seismic Quality Class Category Locationt25) Classt2sj l Component Identification DG Engine Fuel Oil System [17J Fuel Oil Transfer Pumps 3 I DG 1 Strainers 3/NNS I/NS DG/YA 1/3 Filters 3/NNS I/NS DG 1/3 Piping 3/NNS I/NS DG/DF/YA 1/3 Valves 3/NNS I/NS DG/DF 1/3 DG Engine Cooling Water System Circulation Pumps 3 I DG 1 Keep Warm Pumps 3 I DG 1 Jacket Water Coolers 3 I DG 1 Jacket Water Standpipes 3 I DG 1 Chemical Pot Feeders 3 I DG 1 Piping 3 I DG 1 Valves 3 I DG 1 DG Engine Starting Air System [18] Compressors NNS NS DG 2 Aftercoolers NNS NS DG 3 Moisture Separators NNS NS DG 3 Filter / Dryer Units NNS NS DG 3 Air Receivers 3 I DG 1 Strainers 3/NNS I/NS DG 1/3 Traps NNS NS DG 3 Filters 3/NNS 1/NS DG 1/3 Piping 3/NNS I/NS DG 1/3 Valves 3/NNS I/NS DG 1/3 DG Engine Lube Oil System [19] ! Lube Oil Sump Tanks 3 1 DG 1 Lube Oil Coolers 3 I DG 1 l Oil Transfer Pumps NNS NS DG/YA 3 Prelube Oil Pumps 3 I DG 1 l Clean and Used Lube Oil Storage NNS NS YA 3 Tanks Filters 3 I DG 1 ( Strainers 3/NNS I/NS DG 1/3 Piping 3/NNS I/NS DG/YA 1/3 Valves 3/NNS 1/NS DG/YA 1/3 l DG Engine Air Intake and Exhaust System Turbochargers 3 I DG 1 1 Aftercoolers 3 I DG 1 Silencers and Air Filters 3 I DG 1 Piping 3 I DG 1 Aptwoved Design Motoria! Design of SSC (2/9S) Page 3.2-14
l 1
- System 80+ Des /en controlDocument l l
~
i i
'v -
Table 3.2-1 Classification of Structures, Systems, and Components (Cont'd.) Safety Seismic Quality Component Identification Class Category Locationr251 Classt291 l Equipment and Floor Drainage j System : Reactor Building Subsphere Sump 3 I RB 1 l Pumps ! Other Sump Pumps NNS NS 3 l 1 Piping [27} 2/3/NNS I/NS All 1/3 Valves [27] 2/3/NNS I/NS All 1/3 I Diesel Generator Building Sump I Pump System Sump Pumps 3 I DG 1
]
Piping 3/NNS I/NS DG/NA/RW 1/3 j Valves 3/NNS I/NS DG/NA/RW 1/3 j Control Complex Ventilation System Main Control Room Air Conditioning System Air Conditioning Units w/ Filters 3 i NA 1
,q Fans, Ductwork [31] 3/NNS I/II NA 1/2 Q Water-cooling Coils Heating Coils 3
3 I I NA NA 1 1 Dampers 3 I NA 1 Technical Support Center Air Conditioning System Air Conditioning Units w/ Filters NNS 11 NA 2 Fans, Ductwork NNS 11 NA 2 Dampers NNS Il NA 2 Computer Room Air Conditioning l , System l Air Conditioning Units w/ Filters NNS 11 NA 2 l ) Fans, Ductwork NNS 11 NA 2 i Dampers NNS II NA 2 l Essential Electrical Rooms and VitalInstrumentation and Equipment Rooms l (inc. Battery Rooms) Air Conditioning Units w/ Filters 3 I NA 1 l Fans, Ductwork 3 I NA 1 Dampers 3 I NA 1 Balance of Building' Air Conditioning l System Filters NNS NS NA 3 l (]) C Water Cooling Coils Fans, Ductwork . NNS NNS NS NS NA NA 3 3 j Dampers NNS NS NA 3 ; I
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System 80+ Design Contro/ Document Table 3.2-1 Classification of Structures, Systems, and Components (Cont'd.) Safety Seismic Quality Component Identification Class Category Locationr25) Classt291 l Fuel Building Ventilation System Cooling Coil NNS NS NA 3 Heating Coil, Supply NNS NS NA 3 Air Handling Unit w/ Filter NNS 11 NA 2 Ductwork, Supply NNS 11 NA 2 Exhaust System Filter Train 3 I NA 1 Exhaust System Fans 3 1 NA 1 Exhaust System Dampers 3 I NA 1 Ductwork, Exhaust 3 I NA 1 Dampers, Supply NNS 11 NA 2 Nuclear Annex Ventilation System [20] Supply Units NNS 11 NA 2 l Ductwork, Supply NNS 11 NA 2 Cooling Coils NNS 11 NA 3 Particulate Exhaust Filter Units NNS 11 NA 2 Fans, Ductwork NNS 11 NA 2 Dampers NNS 11 NA 2 Radwaste Building Ventilation System Supply Air Handling Units NNS NS RW 2 Cooling Coils NNS NS RW 3 Exhaust Filter Units NNS NS RW 2 Fus NNS NS RW 2 Ductwork NNS NS RW/NA 2 Dampers NNS NS RW 2 Reactor Building Subsphere Ventilation System Individual Cooling Units 3/NNS I/II RB 1/2 Exhaust Fans 3 I NA 1 Cooling Coils and Heating Coils 3 I NA 1 Exhaust System Filter Train 3 I NA 1 Ductwork. Exhaust 3 i NA/RB 1 , Supply Fans NNS 11 NA 2 ] Supply Air Handling Units NNS 11 NA 2 1 Ductwork, Supply NNS 11 NA/RB 2 l Dampers, Exhaust 3 I NA 1 l l Dampers, Supply NNS 11 NA 2 Diesel Building Ventilatioa System j Space Heater 3 I DG 1 Emergency / Normal Fans 3/NNS I/II DG 1/2 Ductwork 3/NNS 1/11 DG 1/2 Dampers 3/NNS 1/11 DG 1/2 Filter, Normal Supply NNS NS DG 2 i i I A)4woved DesAgru Meterial Desiger of SSC (2/95) Page 3.2-16
Sy tem 80+ Design ControlDocument n b Table 3.2-1 Classification of Structures, Systems, and Components (Cont'd.) Safety Seismic Quality Component Identification Class Category Locationtm ClassIM l Annulus Ventilation System Filter Trains 3 I NA 1 Fans 3 I NA 1 Dampers 3 I NA 1 Ductwork 3 I NA/RB 1 Containment Purge Ventilation System Water Cooling Coil NNS NS TA 3 ' Heating Coil NNS NS dA 3 Supply and Exhaust Fans NNS II NA 2 Valves [27] 2/NNS I/II NA/RC 1/2 Filter Trains NNS [28] II NA 2 Ductwork [27, 30] 2/NNS I/II NA/RC 1/2 l Containment Cooling and Ventilation System Containment Cooling Subsystem NNS II RC 2 Control Element Drive Mechanism NNS II RC 2 ( Cooling Subsystem Containment Air Cleanup System NNS 11 RC 2 Cavity Cooling Subsystem NNS 11 RC 2 Ductwork NNS II RC 2 Dampers NNS 11 RC 2 Turbine Building Ventilation System Fans NNS NS TB 3 Dampers NNS NS TB 3 Exhausters NNS NS TB 3 Ductwork NNS NS TB 3 Station Service Water Pump Structure Ventilation System Fans 3 I SP 1 Dampers 3 i SP 1 Ductwork 3 i SP 1 Component Cooling Water IIcat Exchanger Structure (s) Ventilation Systems . Fans NNS II CX 3 Dampers NNS II CX 3 l Space Heaters NNS Il CX 3 Ductwork NNS II CX 3 O O I l l Anaroved ceMon aereerw.w or ssc (2as) rare 2.2-17 j i
Sy3 tem 80+ Design ControlDocument Table 3.2-1 Classification of Structures, Systems, and Coinponents (Cont'd.) Safety Seismic Quality l Component Identification Class Category Locationt251 Classt291 Main Steam Supply System Piping [21] Steam Generator to MSIV's 2 I RC/MS 1 Other NNS NS MS/NA/TB 3 l Main Steam Supply System Valves [21] Safety Valves 2 I MS 1 MSIV's, MSIV Bypass Valves 2 I MS 1 Atmospheric Dump Valves 2 1 MS 1 Valves - 2/NNS 1/NS NA/MS/TB 1/3 Containment flydrogen Recombiner System Ilydrogen Recombiners 2 I NA 1 Hydrogen Analyzers 2 I NA 1 liydrogen Recombiner Control Panel 3 I NA 1 l Piping [27] 2 I NA/RC 1 Valves [27] 2 i NA/RC 1 Steam Generator Blowdown System [22] Flash Tank NNS NS TB 2 11 eat Exchanger NNS NS TB 2 Filter NNS NS TB 2 Demineralizers NNS NS TB 2 l Piping [27] 7/NNS 1/NS RC/TB/MS 1/2 Valves [27] 2/NSS I/NS RC/TB/MS 1/2 Steam Generator Wet Layup Recirculation System [22] Piping [27] 2/NNS 1/NS RC/TB/MS 1/3 Valves [27] 2/NSS I/NS RC/TB/MS 1/3 Ilydrogen Mitigation System liydrogen Igniters NNS I RC 2 l Potable and Sanitary Water Systems NNS NS YA 3 Instrumentation and Control Systems Plant Protection System (PPS) The PPS includes the electrical and mechanical devices and circuitry (from sensors to actuation device input terminals) involved in generating the signals associated with the two protective functions defined below: ApprenniDesign Matwiel Design of SSC (11/96) Page 3,2-18
Sy~ tem 80+ Design Contro/ Document (d Table 3.2-1 Classification of Structures, Systems, and Components (Cont'd.) > Safety Seismic Quality Component Identification Class Category Locationt2sj Class!"3 l . Instrumentation and Control Systems (Cont'd.) Reactor Protective System (RPS) 3 I NA/RC 1 That ponion of the PPS which generates signals that actuate reactor trip Engineered Safety Features 3 I NA/RC 1 Actuation System (ESP) That portion of the PPS which generates signals that actuate engineered safety features Safe Shutdown Systems 3 I DG/NA/CX 1 The safe shutdown systems include SP/MS/ those systems required to secure and RB/RC maintain the reactor in a safe shutdown condition All other systems required for safety 3 1 NA/DG/CX 1 SP/MS/ RB/RC Equipment required to comply with NNS 2 NA/RC 2 10CFR50.62 Equipment specified in Section NNS NS All 2/3 3.3.1.4 of ANSI /ANS-51.1 Contrcl systems not required for safety NNS NS All 2/3 Control Room Panels (safety-related) 3 I NA 1 Control Room Panels (other) NNS 11 NA 1 Instrument Valves and Piping ! Downstream of Safety Class 2 or 3 Root Valves (For l Safety-Related Instruments) Piping, tubing, and fittings 2/3 I All 1 Instrument valves NNS NS All 3 Electric Systems Class IE AC Equipment (includes associated transformers, protective relays, instrumentation and control devices: 4.16 kV Buses 3 I NA 1 480V Load Centers 3 I NA 1 480V Motor Control Centers 3 I NA/CX/DG/SP Class IE DC Equipment: 125V Station Batteries and Racks 3 I NA 1
." ;.:: Design neenwint- Denign of ssa (2/ssi rose 2.21s
System 80+ Des /gn Contro/ Document Table 3.2-1 Classification of Structures, Systems, and Components (Cont'd.) Safety Seismic Quality l Component Identification Class Category Locationt251 Classt29 Electric Systems (Cont'd.) Battery Chargers 3 I NA 1 125V Switchgear and Distribution 3 1 NA 1 Panels l 120V Vital AC System Equipment 3 I NA 1 Inveners 120V Distribution Panels 3 I NA 1 Electrical Cables for Class IE Systems 125V DC Cables (including cable 3 I NA 1 splices, connectors, and terminal blocks) 5 kV Power Cables (including cable 3 I NA/DG/CX/ 1 splices, connectors, and SP terminal blocks) 600V Power Cables (including cable 3 I NA/DG/CX/ splices, connectors, and SP/MS/ terminal blocks) RB/RC Control and Instrumentation Cables 3 I DG/CX/NA 1 (including cable splices, SP/MS/RB connectors, and terminal blocks) Conduit and cable trays and their 3 1 DG/CX/NA/ 1 suppons containing Class IE SP/MS/RB cables and those whose failure RC during a seismic event may damage other safety-related items l Miscellaneous Class IE Electrical Systems Containment building electrical 3 I RC 1 penetration assemblies Non-Class IE Electrical Systems NNS II/NS All 2/3 i instrumention and Display Systems not NNS NS All 2/3 required for safety [32] l Reactor Building Structure Containment Shield Building Steel Containment Vessel 3 I RB 1 Internal Structure 2 I RB 1 Equipment Hatch 3 I RC 1 l Personnel Airlocks 2 I RC 1 Subsphere (including Containment 2 I RC 1 Support Dish) 3 I RB 1 Approved Desier, Material- Design of SSC (2/95) Page 3.2-20 ; I
1 I System 80+ conian contro/ occament
),
- \ Tatde 3.2-1 Classification of Structures, Systems, and Counponents (Cont'd.) !
6 ( Safety Seismic Quality ! Component identificatica Class Category LocaticeW3 ClassW3 l l Nuclear Annex Structure l Control Area 3 I NA 1
; EFW Tank / Main Steam Valve House 3. I NA 1 r
Area
- Emergency Diesel Generator Areas 3 I NA 1 CVCS/ Maintenance Area 3 I NA 1 Fuel Handling Area 3 I NA 1 l Other Structures l :
Unit Vent NNS II NA/RB '2 ; Turbine Iluilding NNS II TB 2 i Radwaste Building [28] NNS II RW 2 l l 4 Station Service Water Pump / Intake 3 I SP 1 l Structure : . Ces-+=st Cooling Water Heat 3 I CX/YD 1 Exchanger Structures and Pipe Tunnels y Diesel Fuel Storage Structure 3 I DF I _ i Station Services Building / Auxiliary NNS NS SB 3 { [ Boiler Structure Administration Building NNS NS ADB 3 Warehouse NNS NS WH 3 + Fire Pump House NNS 'NS FP 3 Alternate AC Source / Combustion NNS NS YA 2 [ Turbine-Generator Structure and Fuel I Tank M Dike (Holdup, Boric Acid Storage and NNS II YA 2 Reactor Makeup Water Tanks) [28] , Dike (Condensate Storage Tank) [28] NNS II YA 2 j Cranes ; 4 Polar Cranc ' NNS II RC 2
)
I- Cask Handling Hoist NNS 11 NA 2 New Fuel Handling Hoist NNS II NA 2 Component Supports [23] I/2/3/NNS 1/NS All 1/2/3 l l r s j i I 4provedessenassemW Desenersac tr r/sst . page 3.2-2r i , . - . - , -- -
System 80+ Design ControlDocument Table 3.2-1 Classification of Structures, Systems, and Components (Cont'd.) Notes: [1] Two safety classes are used for heat exchangers to distinguish primary and secondasy sides where they are different. [2] Loss of cooling water and/or seal water service to the reactor coolant pumps (RCPs) may require stopping the pumps. However, the continuous operation of the pumps is not required during or following an SSE. The auxiliaries are therefore not necessarily Safety Class 3 or Seismic Category I. Provision for cooling water to the pump beanng oil cooler and pump motor air cooler will not comply with the requirements of Regulatory Guide 1.29 (see Section 5.4.1.3). [3] Only those structural ponions of the RCPs which are necessary to assure the integrity of the reactor coolant pressure boundary are Safety Class 1. [4] Safety class of piping within the reactor coolant pressure boundary (as defined in 10 CFR 50) is selected in accordance with the ANSI /ANS 51.1 criteria identified in Section 3.2.2. For purposes of CESSAR, Safety Class 1,2,3, and NNS of ANSI /ANS 51.1 are equivalent to Quality Groups A, B, C, and D of Regulatory Guide 1.26. [5] Flow restrictmg orifices are provided in the nozzles for RCS sampling lines, pressurizer level and pressure instruments, RCP differential pressure instrument lines, SIS pressure instrument lines, RCP seal pressure instrument lines, the charging line differential pressure instrument line, and the SIS hot leg injection pressure instrument lines, to limit flow in the event of a break downstream of the nozzle. The orifice size,7/32-inch diameter and 1-inch long, precludes exceeding fuel design limits while utilizing minimum makeup rates. This permits an orderly shutdown in the event of a downstream break in accordance with General Design Criterion 33 (see Section 3.1.29). A reduction may, therefore, be made in the safety classification of lines downstream of the orifice. [6] The pressure boundary housing for this component is a reactor vessel appunenance and is Safety Class 1 and Seismic Category I, as described in Section 3.9.4.3. [7] Core suppon structures and internals structures are designed to the criteria described in Section 3.9.5.4. [8] CEA and fuel assemblies are designed to the criteria described in Section 4.2. [9] Reactor coolant pump auxiliary components required for lubrication and cooling of pump seals and thrust bearings are not subject to the quality assurance requirements of 10CFR50, Appendix B. [10] Except Lifting Frame Assembly, which is NS. [11] During normal plant operation only. [12] Safety Class I for pressure boundary: Safety Class 3 for electrical ponion of system. [13] The piping, valves, and associated suppons/ restraints of the Main Feedwater System from (and including) the Main Feedwater Isolation Valves to the steam generator feed nozzles are Safety Class 2, Seismic Category 1, and Quality Class I; the remainder is Safety Class NNS. [14] Non-safety Cooling Ileaders are Safety Class NNS, Seismic Category II, Quality Class 2. [15] The Normal Chilled Water System serves no safety function. Ponions of the system which are located in non-safety related areas are classed as non-seismic. Approwd Design hinterial- Design of SSC Pope .T.2 22
f System 80+- . Desian controlDocument O Table 3.2-1 Classification of Structures, Systems, and Components (Cont'd.) Notes: (Cont'd.) [16] Portions of the Fire Protection System piping, valves, and extinguishers which are not in safety-related areas of the plant are designed as non-seismic. [17] Fuel Oil Recirculation System and storage tank fill line strainer are Saiety Gass NNS. [18] The Staning Air System is Safety Class NNS from the starting air compressor through the desiccant drying towers, and Safety Class 3 from the staning air receiver tank inlet check valve to the engine connections. [19] The Clean and Used Oil Transfer System is Safety Class NNS. [20] Mechanical Equipment Room cooling components are Safety Class 3, Seismic Category I, and Quality Class 1. [21] The piping, valves, and associated suppons/ restraints of the Main Steam System from each steam generator to (and including) the Main Steam Isolation Valves are Safety Class 2, Seismic
. Category 1, and Quality Class 1; the remainder is Safety Class NNS.
[22] Piping is Safety Class 2 from the Steam Generators through the Containment isolation Valves. [23] Component suppons are designed to the criteria described in Section 3.9.3.4. [24] Safety injection drain and vent piping is Safety Class NNS, Seismic Category NS and Qaality O. Class 3. V Locations: [25] CX = Component Cooling Water Heat Exchanger Structure DG = Emergency Diesel Generator Area FP = Fire Pump House MS = Main Steam Valve House Area RW = Radwaste Building RB = Reactor Building RC = Steel Containment SP = Station Service Water Pump Structure SB = Station Services Building TB = Turbine Building NA = Nuclear Annex YA = Yard SI = Station Service Water Intake Structure DF = Diesel Fuel Storage Structures ALL Throughout Plant ~
=
[26] Hydrogen lines in safety-related areas are either designed to Seismic Category I requirements, or sleeved with the outer pipe vented to the outside, or equipped with excess flow check valves so that in case of a line break, the hydrogen concentration in the affected area will not exceed 2%. [27] Containment isolation valves and containment penetration piping are Safety Class 2, Seismic Category 1, ana Quality Class 1. z) f Anrewed Denker nsesenie!- Desipur of SSC Page 3.2-23
System 80+ Design ControlDocument Table 3.2-1 Classification of Structures, Systems, and Components (Cont'd.) Notes: (Cont'd.) [28) The foundations / dikes enclosures of these stmetures are designed such that if a Safe Shutdown Earthquake (SSE) occurs, the majority of the liquid inventory expected to be in the building / tank will be contained. it is assumed that the cotrrete would develop cracks and some liquid would be released. This event is bounded by the at alysis in Section 15.7.3. [29] The QA program provides a graded approach w the assurance of quality of work performed by and for ABB-CE by the use of quality class designations to describe the various levels of controls as follows:
- 1) QC-1 is the highest leve! quality class and embodies all necessary controls for items and/or services which are required to meet 10 CFR 50 Appendix B requirements.
- 2) QC-2 is an intermediate level quality class which is used for items or services which require a moderate level of control of activities affecting quality, but which are neither Nuclear Safety-Related nor required to meet the requirements of 10 CFR 50 Appendb.
B. Circumstances appropriate for QC-2 designation include non-standard, corr,Jex items, or those which must perform reliably, in a harsh environment or dn less than normal operator attention or maintenance.
- 3) QC-3 is the quality class which applies to all items or services which are not assigned to another quality class. Quality requirements may be specified in quality plans, procurement documents and/or special procedures if deemed necessary.
[30] The containment low and high purge exhaust ductwork up to the flEPA filters is Seismic Category I. [31] Smoke fan is Safety Class NNS, Seismic Category II, and Quality Class 2. [32] The ALMS is Quality Class 2. The ALMS pressurizer safety valve discharge sensors and signal processing equipment are Seismic Category I. All of the remaining NIMS components are qualified to remain operable following seismic events which do not requite plant shutdown. [33] The boric acid storage tank is classified Seismic Category I but is not designed for tornado wind and wind pressures or tornado generated missiles because it is not required for safe shutdown or accident mitigation. [34] These CVCS components will be constructed in accordtace with ASME Boiler and Pressure Vessel Code, Section III, Class 3. [35] Some CVCS piping and valves designated Safety Class NNS wit be conatructed in accordance with ASME Boiler and Pressure Vessel Code Section Ill, Class 3, as shown on Figure 9.3.4-1. Piping and valves in this category are Seismic Category I, and Quality Class 2. O Approwd Design Matenal Design of SSC page 3.2 24
System 80+ oesign controlDocument (3 ( v/ Table 3,2-2 Safety Class 1,2 & 3 Valves Component Location / Safety Seismic Quality Identiftetion Description Class Category Class Reactor Cool at Eystem (RCS) (1) RC-212 Reactor vessel vent 1 I 1
~
RC-214 Refueling level indicator (Hot leg) 1 1 1 RC-215, 216, 232, 332, 233, RCS drains 1 I 1 333, 234, 334, 235, 335, 310, 311, 312, 313, 314, 315, 316, 317 RC-248, 249, 252, 253, 256, Reactor coolant pump (RCP) pressure 2 1 1 257, 260, 261 differential RC-208, 209, 218, 219, 220 Pressurizer (Pzr) level indicator 2 I 1 RC-204, 205, 206, 207 Pzr pressure indicator 2 1 1 RC-239 Pressurizer refueling 1 1 1 RC-200, 201, 202, 203 Pressurizer safety 1 1 1 RC-240, 241, 442, 443, 236, 237 Pressurizer spray line 1 I 1 p RC-100E,100F Pzt spray line control 1 1 1
'd RC-244 Pzr spray line check 1 1 1 1 RC-210, 213, 238 Sample system 2 1 1 l
l RC-211, 403 Reactor vessel closure head leakoff 2 1 1 l RC-217 O-ring leakoff pressure indicator 2 1 1 RC-265, 266 Mid-loop operating connection to 2 I 1 IRWST RC 268, 269 Mid-loop operating connection to pzr 1 1 1 RC-270, 271, 272, 273, 274, Steam generator differential pressure 2 I 1 I 275, 276, 277, 278, 279, 280, 281, 282, 283, 284, 285 RC-292, 293, 294, 295, 296, RCS pressure differential 2 I 1 l 297, 298, 299 I RC-752, 753, 754, 755 RCP seal housing drain 1 1 1 RC-712, 713, 714, 715 RCP vent 2 1 1 RC-446, 447, 448, 449, 450, RCP HP cooler 1 I 1 l 451,452,453 1 1 RC-868, 869, 870, 871, 700, RCP filter drain 1 I 1 701,702,703 7 RC-724, 725, 726, 727, 736, RCP seal pressure 2 I i 1 ( 737, 738, 739 Approved Design W.:, rial Desiprr of SSC Page 3.2-25 j l
System 80+ Design ControlDocument Table 3.2-2 Safety Class 1,2 & 3 Valves (Cont'd.) Component Location / Safety Seismic Quality Description Class Category Class Identification Reactor Coolant System (RCS) [1] (Cont'd.) .- _ RC-430, 431, 432, 433, 344, RCP controlled bleedoff 2 I ! 345,346,347 - RC-380, 381, 382, 383 RCP vapor seal pressure indicator 2 I 1 Main Steam & Feedwater System (MS&FW) [1] SG-105,106,107,108 ADV block valve 2 I 1 SG-130,135,172,175 Downcomer isolation 2 I 1 C
- 12,137,174,177 Economizer isolation 2 I 1 S(., 140, 141, 150, 151 Main steam isolation 2 I I SG-168,182 Main steam isolation valve bypass 2 I 1 valve SG-169,183 Main steam isolation valve bypass 2 I I valve SG-178,179,184,185 Atmospheric dump valve 2 I 1 SG-552, 553 SG test connection 2 I 1 SG-554, 555, 556, 557, 558, Main steam safety valve 2 I 1 559, 560, 561, 572, 573, 574, 575, 576, 577, 578, 579, 691, 692, 694, 695 SG-567, 598, 599, 612, 650, Economizer check valve 2 I 1 651 SG-586, 587, 605, 609 Downcomer drain valve 2 I 1 SG-603, 611, 661, 655 SG purge connection 2 I 1
~
SG-608, 644 Economizer drain valve 2 I 1 SG-613, 614, 615, 616, 617, SG level indication 2 1 1 618, 621, 622, 623, 624, 625, 626, 627, 628, 619, 630, 631, 632, 635, 636, 637, 638, 639, 640 SG-619, 620, 633,534, 658 Main steam flow indication 7 I 1 659, 662, 663 SG-642, 643, 652, 653 Downcomer check valve 2 I 1 SG-684, 685, 686, 687 Main steam purge l 2 I 1 Component Cooling Water System (CCW) [1] CC-100 CCW HX 1A bypass control 3 I 1 CC-101 CCW HX IB bypass control 3 I 1 CC-102 Non-essential supply header 1 isolation 3 1 1 Approved Des / pro Material Dew!po of SSC Page 3.2 26
System 80+ oesian controlDocument (' Table 3.2-2 Safety Class 1,2 & 3 Valves (Cont'd.) Component Location / Safety Seismic Quali" Identification Description Class Category Clau Component Cooling Water System (CCW) [1] (Cont'd.) CC 103 Non-essential return header 1 isolation 3 1 1 CC-106 CCW HX 1 A inlet isolation 3 I 1 CC 107 CCW HX IB inlet isolation 3 I 1 CC 108 CCW HX 1 A outlet isolation 3 1 1 CC 109 CCW HX IB outlet isolation 3 1 1 CC 110 SCS HX 1 control 3 I 1 CC111 SCS HX 1 outlet isolation 3 I 1 CC-112 SFP HX 1 control 3 1 1 CC-Il3 SFP HX 1 outlet isolation 3 I 1 CC.ll4 CS HX 1 outlet isolation 3 I 1 CC 122 Non-essential supply header 1 isolation 3 I 1 CC-123 Non-essential return header 1 isolation 3 I 1 O CC-130.131,1507 CCW supply to RCP 1A, IB 2 1 1 CC 1302 CCW pump 1 A discharge 3 1 1 CC 1303 CCW pump 1B discharge 3 1 1 CC 1306 CCW pump 1 A surge tank sparger 3 1 1 valve CC-1307 CCW pump 1B surge tank sparger 3 1 1 valve CC 1328 Makeup to CCW surge tank I from 3 I 1 CCWS CC-1331 SC HX 1 header relief 3 1 1 CC-1337 Si pump motor cooler I header relief 3 1 1 CC-1344 SC miniflow HX 1 header relief 3 1 1 CC-1350 SC pump motor cooler I header relief 3 I 1 CC 1356 EFW pump motor cooler 1 header 3 I 1 f relief j CC-136,137,1548 CCW return from RCP 1 A, IB 2 1 1 CC-1362 CS pump motor cooler I header relief 3 I 1 CC-1368 CS miniflow HX 1 header relief 3 I 1 CC-1374 SI pump motor cooler 3 header relief 3 I 1 [ Q} CC-1380 SFP cooling HX 1 header relief 3 1 1 Anwe mt onelen neoneaar cons o n or ssc pope 2.2 27
l l
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System 20+ Design Contro1 Document Table 3.2-2 Safety Class 1,2 & 3 Valves (Cont'd.) Component Location / Safety Seismic Quality Description Class Category Class Identification Component Cooling Water System (CCW) 11] (Cont'd.) CC-1384 SFP cooling pump motor cooler 1 3 I 1 header relief CC-1390 CS HX 1 header relief 3 I 1 CC-1591 CCW pump motor cooler I A header 3 I 1 relief CCW pump motor cooler IB header 3 I 1 CC-1597 relief DGE jacket water cooler I header 3 I 1 CC-1603 relief ECW condenser I header relief 3 I 1 CC-1609 CC-1637 CHG pump motor cooler 1 header 3 I I relief Charge pump miniflow HX 1 header 3 I I CC-1643 relief CC-1851 DGE start air aftercooler I A header 3 I 1 relief DGE start air aftercooler IB header 3 I I CC-1857 relief CC-XXXX CCW surge tank I vacuum breaker 3 I 1 CC-200 CCW HX 2A bypass control 3 I 1 CC-201 CCW HX 2B bypass control 3 I 1 CC-202 Non-essential supply header 2 isolation 3 I 1 CC-203 Non-essential return header 2 isolation 3 1 1 CC-206 CCW HX 2A inlet isolation 3 1 1 CC-207 CCW HX 2B inlet isolation 3 I 1 CC-208 CCW HX 2A outlet isolation 3 I 1 CC-209 CCW HX 2B outlet isolation 3 I 1 CC 210 SCS HX 2 control 3 I 1 CC-211 SCS HX 2 outlet isolation 3 I 1 CC-212 SFP HX 2 control 3 1 1 CC-213 SFP HX 2 outlet isolation 3 1 1 CC-214 CS HX 2 outlet isolation 3 I 1 CC-222 Non-essential supply header 2 isolation 3 I 1 CC-223 Non essential return header 2 isolation 3 I 1 Approvent Design Materia!* Design of GSC Page 3.2-28
l System 80+ Design ControlDocument m
) Table 3.2 2 Safety Class 1,2 & 3 Valves pont'd.) l Component Location / Safety Seismic Quality Identification Description Class Category Class Componert Cooling Water System (CCW) [1] (Cont'd.)
CC-230, 231, 2507 CCW supply to RCP 2A,2B 3 1 1 CC-2303 CCW pump 2A discharge 3 1 1 CC-2303 CCW pump 2B discharge 3 1 1 i
\
CC-2306 CCW pump 2A surge tank sparger 3 1 1 l valve i CC-2307 CCW pump 2B surge tank sparger 3 1 1 valve CC-2328 Makeup to CCW surge tank from 3 1 1 CCWS , 3 ! CC-2331 SC HX 2 header relief 1 1 CC-2337 SI pump motor cooler 2 header relief 3 I 1 CC-2344 SC miniflow HX 2 header relief 3 1 1 1 CC-2350 SC pump motor cooler 2 header relief 3 I 1 O EFW pump motor cooler 2 header 3 1 1 l CC-2356 ' relief CC-236, 237, 2548 CCW return from RCP 2A,2B 2 1 1 CC-2362 CS pump motor cooler 2 header relief 3 I 1 CC-2368 CS miniflow HX 2 header relief 3 1 1 CC-2374 Si pump motor cooler 4 header relief 3 1 1 CC-2380 SFP cooling HX 2 header relief 3 1 1 CC-2384 SFP cooling pump motor cooler 2 3 1 1 header relief CC-2390 CS HX 2 header relief 3 1 1 CC-2591 CCW pump motor cooler 2A header 3 1 1 relief CC-2597 CCW pump motor cooler 2B header 3 1 1 relief CC 2603 DGE jacket water cooler 2 header 3 1 1 relief CC-2609 ECW condenser 2 header relief 3 I 1 CC-2637 CHG pump motor cooler 2 header 3 1 1 relief C' CC-2643 CHG pump miniflow MX 2 header 3 1 1 relief Approved Deelen Meterial Design of SSC Page 3.2 2g
System 80+ Design ControlDocument Table 3.2-2 Safety Class 1,2 & 3 Valves (Cont'd.) Component Location / Safety Seismic Quality Identification Description Class Category Class Component Cooling Water System (CCW) [1] (Cont'd.) CC-2851 DGE start air aftercooler 2A header 3 I 1 relief CC-2857 DGE stan air aftercooler 2B header 3 1 1 relief CC-XXXX CCW surge tank 2 vacuum breaker 3 I 1 CC-240, 241, 2622 CCW supply to letdown HX 2 I 1 CC-242, 243, 2628 CCW return from letdown HX 2 I 1 Chemical and Volume Control System (CVCS) [1] CH-189 CVCS makeup to IRWST check 2 1 1 CH-199 RCP controlled bleedoff to RDT relief 2 1 1 CH-205 Auxiliary spray control 2 I 1 CH 208 Charging backpressure 2 I 1 CH 241, 242, 243, 244 Seal injection flow control 2 I 1 CH-255 Seal injection isolation / I 1 CH-300 RCP bleedoff pressure indicator 2 1 1 isolation CH-301 letdown orifice bypass 2 I 1 CH-304 SCS Purification check 2 I 1 CH-307 SCS Purification isolation 2 1 1 CH-393 Regenerative HX vent isolation 2 1 1 CH-431 Auxiliary spray check 1 1 1 CH-432 Charging line bypass relief 2 1 1 CH-433 Charging line check 1 I 1 CH-447 Auxiliary spray check 1 1 1 CH-448 Charging line check 1 1 1 CH-494 RSSH and RDP to RDH Check 2 1 1 CH-505, 506 RCP CBO containment isolation 2 1 1 CH-507 RCP bleedoff relief isolation 2 I 1 CH 509 IRWST makeup isolation 2 I 1 CH-515 Letdown isolation 1 I 1 Cil-516 Ietdown isolation 1 1 1 CH-523 Letdown containment isolation 2 1 1 CH-524 Charging line isolation 2 I 1 Approved Design AtatorW Dessen of SSC Page 3.2-30
l l 1 System 80+ Design ControlDocument m 1
)
( Table 3.2-2 Safety Class 1, 2 & 3 Valves (Cont'd.) s/ Component location / Safety Seismic Quality Identification Description Class Category Class Chesnical and Volume Control System (CVCS) [1] (Cont'd.) CH-560 RDT suction contamment isolation 2 1 1 CH-561 RDT suction contamment isolation 2 I 1 CH 570,571 Ixtdown orifice bypass 2 I 1 CH-572, 573, 574 letdown orifice isolation 2 I 1 CH-575 letdown contamment isolation 2 I 1 CH 580 RMWS to RDT containment isolation 2 I 1 CH-740 Controlled bleedoff test connection 2 I 1 isolation CH-747 RHX charging inlet line check 2 I 1 CH-748 Charging lire test connection isolation 2 I 1 CH-751 Regen. HX charging isolation 2 1 1 CH-787 Seal injection check 1 I 1 CH-789, 800 Seal injection flow indicator isolation 2 I 1 O Seal injection check I 1 \j CH-802 1 CH-804, 805 Seal injection flow indicator isolation 2 I 1 l CH-807 Seal injection check 1 1 1 i l CH-809, 810 Seal injection flow indicator isolation 2 I 1 j CH-812 Seal injection check 1 1 1 l CH 814,815 Seal injection flow indicator isolation 2 I 1 l CH-833 Seal injection test connection isolation 2 I 1 I CH-835 Seal injection check 2 I 1 l CH-848, 849 Seal injection drain test connection 1 I 1 isolation ) CH-853 letdown line test connection isolation 1 1 1 CH-854 Charging line test connection isolation 2 1 1 1 CH-859, 860 Seal injection test connection isolation 1 1 1
]
CH-866, 867, 868, 869 Seal injection check 1 I 1 l Pool Cooling and Purification System (PCPS) [1] j PC-200, 210 Cooling HX inlet pressure 3 I 1 l PC-201, 293 Cooling HX cross-connect 3 1 1 PC-202, 203 Cooling pump suction isolation 3 I 1 PC-204, 205 Cooling pump discharge pressure 3 1 1 PC-206, 207 Cooling pump discharge check 3 I 1 I
.% ..J Deelyn histend Doespr of SSC Page 3.2-31 } .~. ___ _~ . _ _ . _ _ - - _ _ _ _ _ _ _ - . _ _ _ _ - - _ _ _ _ _ - _ _ _
System 80+ Design ControlDocument 1 Table 3.2-2 Safety Class 1,2 & 3 Valves (Cont'd.) Component Iecation/ Safety Seismic Quality Identification Description Class Category Class Pool Cooling and Purification System (PCPS) [1] (Cont'd.) PC-208, 209 Cooling pump discharge isolation 3 I 1 PC-211, 212 Cooling HX inlet isolation 3 1 1 PC-213, 214 Cooling HX outlet isolation 3 1 1 PC-249 IRWST return line isolation 3 1 1 PC-257, 258 Refueling pool discharge isolation 2 I 1 PC-291, 292 Refueling pool inlet isolation 2 1 1 PC-300, 301, 302, 303 Cooling f ow indication isolation 3 1 1 PC-320, 321 Cooling pump suction pressure 3 1 1 Safety Depressurization System (SDS) RC-406, 407, 408, 409 Rapid depressurization 1 I 1 RC-410, 411, 412, 413 Pressurizer vent 1 I 1 RC-414, 415, 416, 417 Reactor vessel vent 1 I 1 RC-418 RCGVS vent to RDT 2 1 1 RC419 RCGVS vent to IRWST 2 I 1 RC-263, 264 RD pressure indication 2 I 1 RC-267 RCGVS pressure indication 2 1 1 l Safety Injection System (SIS) [1] SI-100,101 IRWST return check valve 2 I 1 SI-102,103 IRWST isolation valve test 2 I 1 SI-104,105 CS pump suction isolation 2 I 1 SI-106,107 SCS pump suction isolation 2 I 1 SI-108,109 SCS pump suction pressure indication 2 I 1 isolation SI-113, 123, 133, 143 Safety injection containment check 2 1 1 S1-115, 125, 135, 145 51 flow indication isolation 2 I 1 S1-116, 126, 136, 146 SI flow indication isolation 2 I 1 S1-117, 127, 137, 147 SIT pressure indication isolation 2 I 1 SI-Il9,129,139,149 SIT pressure indication isolation 2 I 1 9 Approved Design Material Desbyn of SSC (11/96) Page 3.2-32
System 80+ Design ControlDocument o
,m
'O Table 3.2-2 Safety Class 1,2 & 3 Valves (Cont'd.) Component Location / Safety Seismic Quality Identification Description Class Category Class Safety Injection System (SIS) [1] (Cont'd.) 51-130, 131 Safety injection pump isolation (Pumps 2 I 1 3 and 4) S1-157, 158 Containment spray pump check 2 I 1 SI 161,193 CSS IRWST recire line relief 2 I 1 SI-164,165 Containment spray check 2 I 1 SI166 SI HL 2 relief to EDT 2 I 1 SI-168,178 SCS check valve 2 I 1 S1-169 SCS 2 suction thermal relief 1 I 1 SI-170,180 SDCHX vent valve 2 I 1 SI-172,182 SDCHX drain valve 2 I 1 SI-174, 175, 176, 177 CS pump discharge flow indication 2 1 1 SI-179,189 SCS suction line relief valve 2 I 1 SI-187,188 SCS IRWST recirc line relief 2 p \ 1 1 v/ SI-191,194 CS HX relief to EDT 2 I 1 SI-196,197 SCS pump discharge pressure 2 I 1 indication isolation SI-198,199 CS pump suction pressure indication 2 I 1- ; isolation SI-207, 208 IRWST isolation valve test 2 1 1 SI-210, 220, 230, 240 SIT fill and drain isolation 2 1 1 SI-211, 221, 231, 241 SIT relief valve 2 1 1 SI-212, 222, 232, 242 SIT level indication isolation 2 1 1 S1-213, 223, 233, 243 SIT level indication isolation 2 I 1 SI-214, 224, 234, 244 SIT local sample isolation 2 I 1 SI-215, 225, 235, 245 SIT check valve 1 I 1 l S1-216, 226, 236, 246 Injection line pressure indication 2 I 1 isolation SI-217, 227, 237, 247 Safety injection line check 2 1 1 SI-228, 238, 248, 258 SIT level indication isolation 2 I 1 l 51-229, 239, 249, 259 SIT level indication isolation 2 I 1 SI-250, 251 CS miniflow sample isolation 2 1 1
! ) SI-252, 253 SCS miniflow sample isolation 2 I 1 SI-260, 264 SDCHX vent valve 2 I 1 l
Approved Des)po A4sterial Des}pn of SSC Page 3.2-33 l
System 80 + Destga controlDocument Table 3.2-2 Safety Class 1,2 & 3 Valves (Cont'd.) Component Location / Safety Seismic Quality Identification Description Class Category Class Safety IqJection System (SIS) [1] (Cont'd.) SI-261, 267 CS pump miniflow isolation 2 1 1 SI-262, 266 SDCHX drain valve 2 I 1 SI 265, 269 SCS pump miniflow isolation 2 I 1 SI 285, 286 SI 1RWST recire line relief to EDT 2 1 1 SI-287, 289 SCS IRWST recirculation line relief 2 I 1 SI-290 SIT fill line isolation 2 I 1 SI-292 SIT fillline relief to EDT 3 I 1 S1-293 SIT fill line containment isolation 2 1 1 SI-294 SIT fill line local sample 3 1 1 SI-300, 301 CS/SCS 1RWST recirculation isolation 2 1 1 S1-302, 303 SI IRWST recirculation line isolation 2 1 1 SI-304, 305, 308, 309 IRWST isolation valve 2 1 1 SI-310, 311 SDCHX flow control valve 2 I 1 SI-312, 313 SDCHX bypass flow control 2 I 1 SI-314, 315 SCS IRWST recirculation flow control 2 I 1 S1 321, 331 Hot leg injection loop isolation 2 I 1 SI-322, 332 Hot leg check valve leak isolation 1 1 1 SI-340, 342 SCS/CS pump suction cross-connect 2 1 1 SI-341, 343 SCS/CS pump discharge cross-connect 2 I 1 SI-390, 391, 392, 393 Holdup volume tank spillway 2 I 1 SI-394, 395 Reactor cavity spillway 2 I 1 SI 396, 397 IRWST suction BAMP isolation 2 I 1 SI-402, 470 Safety injection pump isolation (Pumps 2 I 1 l 1 and 2) S1-404, 405, 434, 446 Si pump discharge check 2 I 1 SI-408, 416, 433, 436 Pressure gauge isolation 2 I 1 SI-409, 417, 439, 449 SI pump discharge relief to EDT 2 I I ) S1-410, 411, 412, 413 Si pump miniflow isolation 2 I 1 l S1 420, 421 Shutdown purification isolation 2 I 1 SI-422, 423 SDCHX relief to EDT 2 1 1 l SI-424, 426, 448, 451 SI pump bypass check valve 2 I 1 SI-427, 465 Si effluent sampling valve 2 1 1 I I Approved Desiptr Meterial- Desigrr of SSC Page 3.2 34 i
System 80+ oesign controlDocument O V Table 3.2-2 Safety Class 1,2 & 3 Valves (Cont'd.) r Component Location / Safety Seismic Quality Identification Description Class Category Class Safety Igjection System (SIS) [1] (Cont'd.) SI-429, 445 SCS suction sample isolation 2 1 1 SI-435, 447, 476, 478 Si pump discharge isolation 2 1 1 SI-450, 454 IRWST recirculation line refueling 2 1 1 pool isolation S1-466, 467 SCS relief to EDT 2 I 1 SI-46ft SI HL 1 relief to EDT 2 I 1 SI46) SCS 1 suction thermal relief 1 I 1 S1473 SIT fill line thermal relief 2 1 1 SI 474 SIT drain line thermal relief 2 I 1 SI482, 483 CS pump discharge pressure indication 2 I 1 isolation SI-484, 485 CS pump discharge check 2 I 1 SI-488, 489 CS pump discharge isolation 2 I 1 SI-490, 491, 492, 493 SCS flow indication isolation 2 1 1 SI-500, 501, 510, 511 CS header test connection 2 I 1 SI506,516 SI HL p: essure indication 2 1 1 SI-522, 532 Si hot __. ' g l,injection check I i 1 SI 523, 533 S1 HL containment check 1 1 1 SI-525, 526 St HL 1 flow indication 2 1 1 SI-535, 536 SI HL 2 flow indication 2 1 1 SI-540, 541, 542, 543 SI line check 1 1 1 SI550,552,553,555 SI pump test isolation valve 2 I 1 SI558,559 IRWST Fill / CSS Header Tee Isolation 2 1 1 S1-560, 561, 562, 563 CS miniflow HX vent valve 2 1 1 SI-564, 565, 566, 567 CS miniflow HX drain valve 2 1 1 SI-568, 569 SCS pump outlet check valve 2 I 1 SI-570, 571, 572, 573 SCS miniflow HX vent valve 2 1 1 SI 574, 575. 576, 577 SCS miniflow HX drain valve 2 1 1 SI-578, 579 SCS pump outlet isolation 2 1 1 SI-580, 581, 582, 583 CS heat exchanger vent valve 2 1 1 (, ~~) S1 584, 585, 586, 587 CS heat exchanger drain valve 2 1 1
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SI-600, 601 SCS train isolation valve 2 1 1 SI 602, 603 SI low flow control valve 2 I 1 Approved Destgrs Metenial Des # of SSC Page 3.2-35
i System 80+ Design ControlDocument l 1 Table 3.2-2 Safety Class 1,2 & 3 Valves (Cont'd.) Location / Safety Seismic Quality Component Description Class Category Class Identification Safety Indection System (SIS) [1] (Cont'd.) l SI-604, 609 S1 hot leg injection isolation 2 I 1 SI-605, 606, 607, 608 SIT atmospheric vent isolation 2 I 1 SI-611, 621, 631. 641 SIT fill and drain isolation 2 I 1 SI-612, 622, 632, 642 Nitrogen pressure control 2 I 1 SI-613, 623, 633, 643 SIT atmospheric vent isolation 2 I 1 SI-614, 624, 634, 644 SIT discharge isolation valve 1 I 1 SI-616, 626, 636, 646 Injection line isolation 2 I I SI-618, 628, 638, 648 Check valve leakage isolation ! I 1 Nitrogen pressure control 2 I I SI-619, 629, 639, 649 SI-651, 652, 653, 654 SCS suction line isolation 1 I 1 SI-655, 656 SCS suction line isolation 2 I 1
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SI-657, 658 CSS IRWST recirculation flow control 2 I 1 SIT drain to RDT isolation 2 I 1 SI-661 SIT drain to IRWST isolation 2 I 1 SI-670 SI-671, 672 Containment spray header isolation 2 I I SIT fill line isolation 2 I 1 t S1482 SI-686, 696 CS HX to IRWST isolation 2 1 1 SI-687, 695 CS header block valve 2 I 1 SI-688, 693 SCS IRWST recirculation isolation 2 I 1 SI-690, 691 SCS train warm-up flow control 2 I 1 l l SI-700, 701, 702, 703 IRWST level indication 2 I 1 i SI705,707 HVT level indication 2 1 1 l SI-709, 711 Reactor cavity level indication 2 1 1 l 1 SI-720 IRWST retum header isolation 2 1 1 SI-721, 722, 723, 724 Holdup volume tank spillway isolation 2 1 1 SI-725, 726 1RWST pressure indication 2 I 1 l SI727,728 IRWST level indication (local) 2 I 1 S1-730, 731 HVT and RC sump level indication 2 I 1 Emergency Feedwater System (EFW) [1] EF-100,101 Steam-Driven Pump Isolation 2 1 1 EF 102,103 Motor-Driven Pump Isolation 2 1 1 i EF-104,105 Steam-Driven Pump Flow Control 3 1 1 j , Approwd Desipt Material . Desigre of SSC Page 3.2-36 l
System 80 + Design controlDocument 5 Table 3.2-2 Safety Class 1,2 & 3 Valves (Cont'd.) Component Location / Safety Seismic Quality Identification Description Class Category Class Emergency Feedwater System (EFW) [1] (Cont'd.) EF-106,107 Motor-Driven Pump Flow Control 3 I 1 EF-108,109 EFW Pump Turbine Steam Supply 2 I 1 Isolation EF-110,111 Steam Supply Drain Isolation 2 I 1 EF-ll2,113 EFW Pump Turbine Steam Supply 2 I 1 Bypass EF-200, 201, 202, 203 SG isolation Check Valve 2 I 1 EF-204, 205, 206, 207 EFW Pump Discharge Check 3 I 1 EF-208, 209, 210, 211 EFW Pump Suction Isolation 3 I 1 EF-212, 213 EFWST Crossover Isolation 3 I 1 EF-214, 215 Non-Safety Condensate Source 3 1 1 Isolation Check Valve EF-216, 217 EFWST Drain Isolation Valve 3 1 1 EF 220,221,222,223 EFW Pump Minimum Flow Isolation 3 I 1 ( EF-224, 225, 226, 227 Full Flow Test Bypass Isolation 3 I 1 EF 228,229,230,231 Full Flow Test Flow Control 3 I 1 EF232, 213, 234, 235 Full Flow Test Bypass Isolation 3 I 1 l EF-236, 237 Steam-Driven Pump Turbine Bearing 3 I I Oil Cooler Return Isolation EF-238, 239 Steam Supply Maintenance Isolation 2 I 1 EF-240, 241, 244, 245, 246, Steam Supply Drain Isolation 2 I 1 247,248,249 EF-250, 251, 252, 253, 254, Steam Exhaust Drain Isolation 3 I 1 255 EF-256, 257, 258, 259, 260, 3 Flow Indicator Isolation 3 I i 261, 262, 263 EF-264, 265, 266, 267, 268, Pressure Indicator Isolation 3 I 1 269,270,271 EF-272, 273, 274, 275, 276, Flow Indicator Isolation 3 1 1 j 277,278,279 i EF 280,281,282,283 Pressure Test Isolation 3 I 1 EF-284, 285, 286, 287 Level Indication Isolation 3 1 1 l /O \ ) - EF-288, 289, 290, 291 EFW Pump Discharge Crossover 3 I 1
EF 292,293 Pressure Indication Isolation 2 I 1 EF-294, 295 Pressure Test Isolation 2 I 1 l
Approved Design Material Design of SSC Page 3.2-37
System 80+ Design ControlDocument Table 3.2-2 Safety Class 1,2 & 3 Valves (Cont'd.) Component Location / Safety Seismic Quality Identification Description Class Category Class Emergency Feedwater System (EFW) [1] (Cont'd.) EF 296, 297 Pressure Test Isolation 3 1 1 EF-298, 299 Level Switch Isolation 2 I 1 EF-300, 301 Level Switch Isolation 3 I I EF-310, 311 EFWST Cleanup Isolation 3 I I EF-316, 317, 318, 319, 320, Steam Supply Drain Isolation 2 I 1 321, 322, 323, 324, 325, 326, 327 EF 328,329 Turbine Case Drain Isolation 3 I I EF-330, 331 Steam Supply Bypass Maintenance 2 I 1 EF-334, 335, 336, 337 Level Indication Isolation 3 1 1 EF 338, 339, 340, 341 EFW Pump Discharge Maintenance 3 I 1 EF-XXX Steam driven EFW pump I turbine trip 2 I 1 and throttle valve EF-XXX Steam driven EFW pump 2 turbine trip 2 I 1 and throttle valve EF-XXX Steam driven EFW pump I turbine 3 I 1 govemor valve EF-XXX Steam driven EFW pump 2 turbine 3 1 1 govemor valve Diesel Generator Engine Fuel Oil System DF-130 FO day tank 1 level control 3 1 1 DF-230 FO day tank 2 level control 3 I 1 Diesel Generator Engine Starting Air System DS-Il0 Stcc air receiver I A inlet 3 1 1 DS-112 Start air receiver I A outlet 3 1 1 DS-113,123 Start air supply to engine control panel 3 I 1 1 DS-115,116,117,118 DGE 1 start air left bank inlet 3 I 1 DS-120 Start air receiver IB inlet 3 I 1 DS-122 Start air receiver IB outlet 3 I 1 . DS-125,126,127,128 DGE 1 start air right bank inlet 3 1 > 1 DS-210 Stan air receiver 2A inlet 3 I 1 l DS-212 Stan air receiver 2A outlet 3 I 1 l Approved Design Material. Design of SSC Page 3.2.5 ,
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1 System 80+ Design ControlDocument a b Table 3.2-2 Safety Class 1, 2 & 3 Valves (Cont'd.) Component Location / Safety Seisinic Quality Identification Description Class Category Class Diesel Generator Engine Starting Air System (Cont'd.) DS-213, 223 Start air supply to engine control panel 3 1 1 2 DS-215, 216, 217, 218 DGE 2 start air left bank inlet 3 1 1 DS-220 Start air receiver 2B inlet 3 1 1 DS-222 Start air receiver 2B outlet 3 1 1 DS-225, 226, 227, 228 DGE 2 start air right bank inlet 3 1 1 Secondary Sampling System (SC) [1] SC-204, 219 SG 1 cold leg sample 2 1 1 SC-211, 228 SG 1 hot leg sample 2 1 1 SC-220, 221 SG 1 downcomer sample 2 1 1 SC-222, 223 SG 2 cold leg sample 2 1 1 SC-224, 225 SG 1 hot leg sample 2 1 1 SC-226, 227 SG 1 downcomer sample 2 1 1 Primary Sampling System (SS) [1] SS-200, 203 Hot leg sample 2 1 1 SS-201, 2M Pressurizer liquid sample 2 1 1 SS-202, 205 Pressurizer steam space sample 2 1 1 1 SS-208, 210, 211 Holdup volume tank sample 2 1 1 Station Service Water (SSW) , SW-100,102, IN,106,108, SSW strainer IA backwash 3 1 1 110 SW-101,103,105,107,109, SSW strainer IB backwash 3 I I 111 SW-120 CCW HX 1 A inlet isolation 3 1 1 SW 121 CCW HX IB inlet isolation 3 I 1 SW-122 CCW HX 1 A outlet isolation 3 1 1 SW-123 CCW HX IB outlet isolation 3 1 1 SW-1302 SSW pump 1A discharge 3 1 1 SW 1303 - SSW pump IB discharge 3 1 1 SW-1350 CCW HX 1A header relief 3 I 1 O SW-1351 CCW HX IB header relief 3 1 1 SW-200, 202, 202, 206, 208, SSW strainer 2A backwash 3 1 1 210 4 prevent Deekn Ainamiel* Design of SSC Page 3.2-39
System 80+ oesign controt Document Table 3.2-2 Safety Class 1,2 & 3 Valves (Cont'd.) Component Location / Safety Seismic Quality Description Class Category Class Identification Station Service Water (SSW) (Cont'd.) SW-201, 203, 205, 207, 209, SSW strainer 2B backwash 3 I 1 211 SW-220 CCW HX 2A inlet isolation 3 1 1 SW-221 CCW HX 2B inlet isolation 3 I 1 SW-222 CCW HX 2A outlet isolation 3 I 1 SW-223 CCW HX 2B outlet isolation 3 I 1 SW-2302 SSW purap 2A discharge 3 I 1 SW-2303 SSW pump 2B discharge 3 1 1 SW-2350 CCW HX 2A header relief 3 I 1 SW-2351 CCW HX 2B header relief 3 I 1 Other Systems XX-001, 002 Breathing air supply 2 1 1 XX-003, 004 Station air supply 2 1 1 XX-005, 006 Division 1 instrumentation air supply 2 I 1 XX-007, 008 Division 2 instrumentation air supply 2 1 1 XX 010, 011 High wlume containment purge supply 2 1 1 1 XX-012, 013 High volume containment purge supply 2 I I 2 XX-014, 015 High volume containment purge 2 1 1 exhaust I XX-016, 017 High volume containment purge 2 1 1 exhaust 2 XX-018, 019 Low volume containment purge supply 2 1 1 XX-020, 021 1.ow volume containment purge 2 I 1 exhaust XX-030, 031, 032 SG 1 combined blowdown 2 1 1 XX-033, 034, 035 SG 2 combined blowdown 2 1 1 XX-040, 041 Fire water supply 1 2 1 1 XX442, 043 Fire water supply 2 2 1 1 XX-050, 051 Containment radiation monitor sample 2 I 1 inlet XX452, 053 Containment radiation monitor sample 2 1 1 outlet Appwved Des @n neatuw Des 4n or ssc rege 2.2-40
System 80 + Design controlDocument
,y i 'v'\ Table 3.2-2 Safety Class 1,2 & 3 Valves (Cont'd.)
Component Location / Safety Seismic Quality Identification Description Class Category Class Other Systems (Cont'd.) XX-060, 061 Integrated leak rate testing pressure ?. I 1 sensing XX-070, 071 Demineralization water supply 2 I 1 XX-080, 081 Nitrogen supply 2 I 1 XX-090 Integrated leak rate testing 2 I 1 pressurization line XX-100,101 RCP oil fill line 2 I 1 XX-Il0,111 Containment sump pumps' discharge 2 I 1 XX-120,121 Reactor drain tank gas space to GWMS 2 I 1 XX-130,131 Decontamination line 2 1 1 XX-150 SG 1 wet layup recirculation 2 I 1 XX-151 SG 2 wet layup recirculation 2 I 1 XX-160,161,162 Containment vent units drain header 2 I 1
/
v) XX-170,171 XX-172,173 Personnel air lock I equalization line Personnel air lock 2 equalization line 2 2 I I 1 1 XX-180,181 ECW tank 1 N2 supply 3 I 1 XX-182,183 ECW expansion tank 1 SSWS make-up 3 1 1 XX-154,185 ECW expansion tank 1 DMWS make- 3 I 1 up XX-186 ECW pump 1 A discharge 3 I 1 XX-187 ECW pump IB discharge 3 I 1 XX-188,189 ECW tank 2 nitrogen supply 3 I 1 XX-190,191 ECW expansion tank 2 SSWS make-up 3 I 1 XX-192,193 ECW expansion tank 2 DMWS make- 3 I 1 up XX-194 ECW pump 2A discharge 3 I 1 XX-195 ECW pump 2B discharge 3 I 1 XX-196,197 NCW containment supply Division 1 2 I 1 XX 198.199 NCW containment return Division 1 2 I 1 XX-200, 201 NCW containment return Division 2 2 I 1 XX-2040 Channel electrical equipment 3 I 1 .( g) recirculation air handling unit I A V control valve Approntf Design Material Design of SSC Page 3.241
System 80 + Design ControlDocument Table 3.2 2 Safety Class 1,2 & 3 Valves (Cont'd.) Component Location / Safety Seismic Quality Identification Description , Class Category Class Other Systems (Cont'd.) XX-2041 Channel electrical equipment 3 1 1 recirculation air handling unit IB control valve Channel electrical equipment 3 1 1 XX-2042 recirculation air handling unit IC control valve XX-2043 Channel electrical equipment 3 I 1 recirculation air handling unit ID control valve XX-2044 Channel electrical equipment 3 1 1 recirculation air handling unit 2A control valve XX-2045 Channel electrical equipment 3 1 1 recirculation air handling unit 2B control valve XX-2046 Chnt electrical equipment 3 1 1 recirculation air handling unit 2C control valve XX-2047 Channel electrical equipment 3 1 1 recirculation air handling unit 2D control valve XX-2048 Division 1 essential chilled water room 3 1 1 recirculation air handling unit control valve XX-2049 Division 2 essential chilled water room 3 1 1 recirculation air handling unit control valve XX-2050 Remote shutdown panel room 3 I 1 recirculation air handling unit control valve XX-2051 Division 1 motor-driven EFW pump 3 1 1 room recirculation air handling unit control valve XX-2052 Division 2 motor-driven EFW 3 I 1 pump room recirculationair handling j unit control valve ; XX-2053 Division I steam 4 riven EFW pump 3 I 1 l room recirculation air handling unit control valve Approved Deshm Materhel Design of SSC Page 3.2-42
System 80+ oesign controlDocument l 'v' Table 3.2-2 Safety Class 1,2 & 3 Valves (Cont'd.) l 1 Component Location / Safety Seismic Quality i Identification Description Class Category Class Other Systems (Cont'd.) XX-2054 Division 2 steam-driven EFW pump 3 I I room recirculation air handling unit ; control valve XX-2055 Division 1 CS HX room recirculation 3 I 1 air handling unit control valve ] XX-2056 Division 2 CS HX room recirculation 3 I 1 air handling unit control valve XX-2057 Division 1 SCS HX room recirculation 3 I 1 air handling unit control valve XX-2058 Division 2 SCS HX room recirculation 3 1 1 air handling unit control valve XX-2059 SI pump room I recirculation air 3 I 1 handling unit control valve XX-2060 SI pump room 2 recirculation air 3 I 1 , handling unit control valve XX-2061 SI pump room 3 recirculation air 3 1 1 l' handling unit control valve XX-2062 SI pump room 4 recirculation air 3 1 1 handling unit control valve l XX-2063 Division 1 CS pump room recirculation 3 I 1 ) air handling unit control valve XX-2064 Division 2 CS pump room recirculation 3 I 1 air handling unit control valve XX-2065 Division 1 SCS pump room 3 I 1 recirculation air handling unit control valve XX-2066 Division 2 SCS pump room 3 I 1 recirculation air handling unit control valve XX-2067 CCW pump 1 A room recircu' x 3 I 1 handling unit control valve XX-2068 CCW pump IB room recirculation air 3 I 1 ) handling unit control valve l XX-2069 CCW pump 2A room recirculation air 3 1 1 handling unit control valve ,e y XX-2070 CCW pump 2B room recirculation air 3 1 1 ( ) handling unit control valve
%)
4prowd Design Material- Design of SSC Page 3.2-43
i System 80 + Design ControlDocument Table 3.2-2 Safety Class 1, 2 & 3 Valves (Cont'd.) Location / Safety Seismic Quality Component Description Class Category Class Identification Other Systems (Cont'd.) Vital electrical and instrumentation 3 I I XX 2071 room Channel A recirculation air handling unit I control valve Vital electrical and instrumentation 3 I 1 XX-2072 room Channel A recirculation air handling unit 2 control valve Vital electrical and instrumentation 3 I 1 XX-2073 room Channel B recirculation air handling unit I control valve Vital electrical and instrumentation 3 I 1 XX-2074 room Channel B recirculation air handling unit 2 control valve Vital electrical and instrumentation 3 I I XX-2075 room Channel C recirculation air handling unit I control valve Vital electrical and instrumentation 3 I 1 XX 2076 room Channel C recirculation air handlingunit 2 control valve Vital electrical and instrumentation 3 I I XX-2077 room Channel D recirculation air handling unit I control valve XX-2078 Vital electrical and instrumentation 3 1 1 room Channel D recirculation air handling unit 2 control valve XX-2079 Penetration Room A recirculation air 3 I 1 handling unit I control valve XX-2080 Penetration Room A recirculation air 3 1 1 handling unit 2 control valve XX-2081 Penetration Room B recirculation air 3 1 1 handling unit I control valve XX-2082 Penetration Room B recirculation air 3 1 1 handling unit 2 control valve XX-2083 Penetration room C recirculation air 3 I 1 handling unit I control valve XX-2084 Penetration Room C recirculation air 3 I 1 handling unit 2 control valve XX-2085 Penetration Room D recirculation air 3 I 1 handling unit I control valve XX-2086 Penetration Room D recirculation air 3 1 1 handling unit 2 control valve Approved Desigre acetorial- Design of SSC Page 3.244
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System 80+ Design ControlDocument q . I U Table 3.2-2 Safety Class 1,2 & 3 Valves (Cont'd.) Component Location / Safety Seismic Quality Identification Description Class Category Class Other Systems (Cont'd.) XX-2087 Division 1 fuel pool HX room 3 I 1 recirculation air handling unit control valve XX-2088 Division 2 fuel pool HX room 3 I 1 recirculation air handling unit control valve XX-2089 Division 1 mechanical equipment room 3 I I recirculation air handling unit control valve XX-2090 Division 2 mechanical equipment room 3 I 1 recirculation air handling unit control valve , XX-2091 Division I control room recirculation 3 I 1 air handling tmit control valve t XX-2092 Division 2 control room recirculation 3 1 1 ('s air handling unit control valve I XX-210 CS pump Room I backwater valve 3 1 XX-2 t l CS HX Room I backwater valve 3 I 1 XX-212 SI pump Room I backwater valve 3 I 1 XX-213 SC HX Room i backwater valve 3 1 1 XX 214 SC pump Room I backwater valve 3 I 1 XX-215 SI pump Room 3 backwater valve 3 I 1 XX-216 CS pump Room 2 backwater valve 3 I 1 ! i XX-217 CS HX Room 2 backwater valve 3 I 1 XX-218 SI pump Room 2 backwater valve 3 I 1 XX-219 SC HX Room 2 backwater valve 3 I 1 l i XX-220 SC pump Room 2 backwater valve 3 1 1 l XX-221 SI pump Room 4 backwater valve 3 I 1 XX-222 Division 1 EFW pump rooms 3 I 1 backwater valve XX 223 Division 1 CCW pump rooms 3 I 1 backwater valve XX-224 Division 2 EFW pump rooms 3 1 1
- backwater valve I
V XX-225 Division 2 CCW pump rooms 3 1 l backwater valve
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i Apiproved Desips Meterie!- Desips of SSC - Pope 3.2 45
System 80+ _ Design controlDocument Table 3.2-2 Safety Class 1,2 & 3 Valves (Cont'd.) Component Location / Safety Seismic Quality Identification Description Class Caegory Class Other Systems (Cont'd.) 3 1 1 XX 230 RB subsphere Quadrant A sump pump 1 discharge RB subsphere Quadrant A sump pump 3 I I XX-231 2 discharge 3 1 1 XX-232 RB subsphere Quadrant B sump pump 1 discharge RB subsphere Quadrant B sump pump 3 I I XX-233 2 discharge RB subsphere Quadrant C sump pump 3 I i XX-234 1 discharge RB subsphere Quadrant C sump pump 3 I 1 XX-235 2 discharge RB subsphere Quadrant D sump pump 3 1 1 XX-236 1 discharge RB subsphere Quadrant D sump pump 3 I 1 XX-237 2 discharge XX-240 DG building sump pump 1 A discharge 3 1 1 DG building sump pump IB discharge 3 I 1 XX-241 3 1 XX-242 DG building sump pump 2A discharge 1 XX-243 DG building sump pump 2B discharge 3 I 1 XX-250, 251 Division 1 CHRS suction 2 1 1 XX-252, 253 Division 2 CHRS suction 2 I 1 XX-254, 255 Division 1 CHRS discharge 2 I 1 XX-256, 257 Division 2 CHRS discharge 2 I 1 XX-258, 259, 260 Division 1 CHRS individual suction 2 1 1 XX-261, 262. 263 Division 2 CHRS individual suction 2 1 1 XX-264, 265 Division 1 recombiner inlet isolation 2 I 1 XX-266, . Division 2 recombiner inlet isolation 2 1 1 XX-268, 269 Division I recombiner outlet valve 2 1 1 XX-270, 271 Division 2 recombiner outlet valve 2 I 1 l XX-272 Division I hydrogen calibration supply 2 I 1 j XX-273 Division 2 hydrogen calibration supply 2 1 1 XX-274 Division 1 nitrogen supply 2 1 1 XX-275 Division 2 nitrogen supply 2 1 1 Approved Design Material Design of SsC Page 3.2-46
System 80 + Design controlDocument
'd Table 3.2 Safety Class 1,2 & 3 Valves (Cont'd.)
Component Location / Safety Seismic Quality Identification Description Class l Category Class Other Systems (Cont'd.) XX-276 Division 1 analyzer inlet recombiner 2 I 1 supply from XX 277 Division 2 analyzer inlet recombiner 2 I 1 supply from XX-278 Division 1 analyzer inlet from 2 I 1 recombiner discharge XX-279 Division 2 analyzer inlet from 2 I 1 recombiner discharge XX-280 Division 1 analyzer outlet 2 I 1 XX 281 Division 2 analyzer outlet 2 I 1 XX-282 Division 1 CliRS purge to annulus 2 I 1 XX 283 Division 2 CHRS purge to annulus 2 I 1 XX-284 Division 1 Cl " bypass isolation 2 1 1 XX-285 Division 2 CHRS bypass isolation 2 I 1 C)N
\ ' XX-286 Prelube oil pump 1 discharge relief 3 1 1 XX-287 Prelube oil pump 2 discharge relief 3 I 1 XX-288 Air receiver I A relief 3 I 1 XX 289 Air receiver IB relief 3 I 1 XX-290 Air receiver 2A relief 3 1 1 XX-291 Air receiver 2B relief 3 I 1 XX-292 ECW expansion tank 1 relief 3 I 1 XX-293 ECW expansion tank 2 relief 3 I 1 XX 294 Engine-driven FO pump 1 discharge 3 I 1 relief XX 295 Engine driven FO pump 2 discharge 3 1 1 relief XX-296 Motor-driven FO booster pump 1 3 I 1 discharge relief XX-297 Motor-driven FO booster pump 2 3 1 1 discharge relief XX-298 DG engine DFO relief 1 3 1 1
- , XX-299 DG engine DFO relief 2 3 1 1 l \ l Approwd Design Material- Design of Ssc Page 3.247
System 80+ Design ControlDocument Table 3.2-2 Safety Class 1,2 & 3 Valves (Cont'd.) Note: (1) All contamment isolation valves and their operators, including manual valves, check valves, and relief valves which also serve as isolation valves, are subject to the pertinent requirements of the Quality Assurance Program as given in Chapter 17. Table 3.2-3 Relationship of Safety Class to Code Class Code Class Safety Class (ASME Section III) 1 SC-1 SC-2 for reactor containment components MC SC-2 for fluid system components 2 SC-3 for core suppon structures CS 3 SC-3 (otherwise) NNS Industry Standards O O Aymd Design Material- Design of SSC Page 3.2M
System 80+ Desian controlDocument V 3.3 Wind and Tornado Loadings All Seismic Category I structures, including accesses, except those not exposed to wind, are designed for wind and tornado loadings. 3.3.1 Wind Loadings The design for wind loading is in accordance with ANSI /ASCE 7, " Minimum Design Loads for Buildings and Other Structures" (Reference 1). Structural geometries not addressed in ANSI /ASCE 7 shall be evaluated using ASCE Paper 3269, " Wind Forces on Structures" (Reference 2), and ASCE Paper 4933,
" Wind Loads on Dome-Cylinder and Dome-Cone Shapes" (Reference 3).
3.3.1.1 Design Wind Velocity A design wind velocity of 110 mph, at a height of 33 feet above nominal ground elevation is used as the i maximum wind speed for a 50 year recurrence period. Velocity profiles and associated effective pressures for winds with a 100 year recurrence period are calculated in accordance with Section 6 of Reference 1 utilizing an Importance Factor, I, of 1.11 and Exposure C. Gust response factors are dependent on height above grade level and are in accordance with Table 8 of Reference 1 for Exposure C.
'v 3.3.1.2 Determination of Applied Forces i
Based on structure geometry and physical configuration, the effective pressure distribution is transformed ; into applied equivalent static building forces utilizing appropriate shape coefficients given in Reference
- 3. l Wind pressure distribution curves for the containment shield building are shown in Figure 3.3-1. The maximum height of the shield building above grade is approximately 173 feet 3 inches. ,
3.3.2 Tornado Loadings i All Seismic Category I structures that perform a safe shutdown or accident mitigation function, except those structures not exposed to wind, are designed for tornado loadings. 3.3.2.1 Applicable Design Parameters Tornado effects are in accordance with Interim Regulatory Guide 1.76 (Reference 4). The following ; parameters are applicable to the design basis tornado: i Maximum wind speed: 330 mph Rotational speed: 260 mph Translational velocity: 70 mph p Radius: 150 feet Q Maximum pressure differential: Rate of pressure drop: 2.4 psid 1.7 psi /second t
- Missile Spectra: See Table 3.5-2 ;
w= ou+ neeww- ou+ or ssc rees 2.2.s I P
1 I System 80 + Design ControlDocument 3.3.2.2 Determination of Forces on Structures The forces on Seismic Category I structures due to tornado wind loadings are obtained using methods outlined in Section 3.3.1.2, with a wind velocity of 330 mph (vector sum of all component velocities - assumed constant with height). Velocity profiles are determined as outlined in Section 3.3.1.1. Effective pressure distribution loads are transformed into equivalent static building forces as outlined in Section 3.3.1.2. In determining tornado wind loadings, both the importance factor and gust factors are taken as unity. Tornado loadings include tornado wind pressure, internal pressure due to tornado created atmospheric pressure drop, and forces generated due to the impact of credible tornado missiles. These loadings are combined with other loads as described in Section 3.8. 3.3.3 Effect of Failure of Structures or Components Not Designed for Wind and Tornado Loads Structures, systems and components that are adjacent to Seismic Category I structures, systems and components and that are exposed to wind and tornado loads will not be permitted to affect or degrade the capability of Seismic Category I structures, systems and components to perform their intended safety functions. This is accomplished by one of the following methods:
- Designing the structure, system or component adjacent to Seismic Category I structures, systems or components to wind and tornado loadings.
- Investigating the effect of adjacent structural failure on Seismic Category I structures, systems and components to determine that no impairment of function results.
- Designing a structural barrier to protect Seismic Category I structures, systems and components from adjacent structural failure.
References for Section 3.3
- 1. " Minimum Design Loads for Buildings and Other Structures," ANSI /ASCE 7.
- 2. " Wind Forces on Structures," ASCE Paper No. 3269, Transactions, ASCE, Vol.126, Part II, 1961, p.1124.
- 3. " Wind Loads on Dome-Cylinder and Dome-Cone Shapes," ASCE Paper No. 4933, Journal of the Structural Division - Proceedings of the American Society of Civil Engineers, Vol. 92, No.
ST5, October 1966.
- 4. Safety Evaluation by the Office of Nuclear Reactor Regulation of Recommended Modification to the R.G.1.76 Tornado Design Basis for the ALWR, attached to a March 25,1988 NRC letter to the ALWR Utility Steering Committee.
l O Approved Design Material. Design of SSC Page 3.3-2
System 80 + Design ControlDocument _, o V Distance from Top of Dome Along Arc 1.cngth Cacasmeevenne et WBucJi Dome Meets Cylinsee at Springinne - ' ' J 'g , 'g
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s __ __ System 80+ Design ControlDocument 3,4 Water Level (Flood) Design All Seismic Category I structures, components and equipment are designed for applicable loadings caused by postulated floods. Section 2.4 of the site-specific SAR describes, in detail, the relationship of the site-specific flood levels to safety-related buildings and facilities. ((The COL applicant referencing the System 80+ Standard Design will provide a site-specific flood analysis to demonstrate safe shutdown of the plant. This analysis will include a site description and elevations of safety-related structures and equipment; evaluations of penetrations in Seismic Category I structures; and effects of flooding due to postulated pipe breaks.))I 3.4.1 Flood Elevations The elevation level for floods at the reactor site is determined in accordance with Regulatory Guide 1.59,
' Design Basis Floods for Nuclear Power Plants," and ANSI /ANS 2.8, " Determining Design Basis Flooding at Power Reactor Sites." The design basis level for the System 80+ Standard Design is I foot below plant finished yard grade. Flood level values in excess of this I foot level are site-specific and protection measures for that flood level are described in Section 2.4 of the site-specific SAR.
3.4.2 Phenomena Considered in Design Load Calculation All Seismic Category I structures are designed to withstand the static and dynamic forces of the design basis flood level. O V Site-specific information will include a specific description of the site and elevation for all safety-related structures, exterior accesses, equipment and systems. 3.4.3 Flood Source Application The design flood is used in determining the applicable water level for design of all Seismic Category I structures in accordance with the load combinations discussed in Section 3.8.4. The forces acting on those structures are determined on the basis of full external hydrostatic pressure corresponding to that flood level. All Seismic Category I structures will be in a stable condition due to both moment and uplift , forces resulting from the proper load combinations, including design basis flood levels. 3.4.4 Flood Protection 3.4.4.1 Flood Protection Measures for Seismic Category I Structures i The flood protection measures for Seismic Category I structures, systems and components are designed in accordance with Regulatory Guide 1.102, " Flood Protection for Nuclear Power Plants." Seismic Category I structures identified in Table 3.2-1 are designed for flood protection. These Seismic Category I structures are designed to protect safety-related equipment from floods by i incorporating the following safeguards into their construction: i
%J ,
8 COL information itern; see DCD Introduction Section 3.2. 4 prow onyn meerw. oeekn orssc rope 14-r
System 80+ Design ControlDocument e No exterior access openings will be lower than 1 foot above plant grade (yard grade) elevation.
- The finished yard grade adjacent to the safety-related structures d'i be maintained at least 1 foot below the ground floor elevation, except where ramps or steps ans pavided for access.
- Waterstops are used in all horizontal and vertical construction joints in a'l exterior walls up to flood level elevation.
- Water seals are provided for all penetrations in exterior walls up to flood level elevation. The water seals are designed for the static pressure of water at the flood elevation. Water seals at the interface with safety related structures are designed to maintain integrity in the event of a Safe Shutdown Earthquake. In the event of seal failure, any credible leakage is limited to the capacity of the sump pumps, or associated flood effects are shown to be acceptable.
For other Seismic Category I structures where flood protection measures are required (e.g. pumping systems, stoplogs, watertight doors, dikes, retaining walls and drainage systems) the design of means for providing such protection will be described in Section 2.4 of the site-specific SAR. Penetrations located below the external flood level in the external walls of the Nuclear Annex include Component Cooling Water, Radwaste, and Diesel Fuel Oil System piping and cable penetrations. Additional penetrations may be identified once layouts are finalized for systems such as sewage, demineralized water, station air, and security. All penetrations are sealed on the inside of the penetration to eliminate the potential of flooding through the penetration. Non-safety related piping that penetrates an exterior wall flood barrier of a Seismic Category I ometure is designed to Seismic Category I criteria from the wall through an interior isolation valve. For piping in which flow is exiting the structure, the interior valve is a reverse flow check valve. For piping in which flow is entering the structure, the valve is an isolation valve which can be manually isolated to terminate a break in the non-seismic portion of the interior piping to prevent flooding. A site-specific evaluation will be performed to ensure that all penetrations in Seismic Category I structures are properly sealed to protect safety related equipment from flooding. External flooding as a result of secondary flooding sources located in the Turbine Building are addressed in Section 10.4.1.3. Entrances to the Nuclear Annex from the Turbine Building are elevated above plant grade to prevent flood propagation. Internal flood protection in the System 80+ design minimizes possible flood sources. The station service water system is located outside the Nuclear Annex to eliminate a significant source of water. The component cooling water and emergency feedwater systems are fully separated by division, thus eliminating the possibility of a single flood source within these systems impacting both divisions. Lengths of high energy and moderate energy piping have been minimized by equipment location. Equipment in the Reactor Building (RB) Subsphere is located in quadrants to minimize the lengths of piping runs. The RB Subsphere also provides for close proximity of equipment to reduce piping runs from containment. Flood barriers have been integrated into the design to provide further flood protection while minimizing the impact on maintenance accessibility. The primary means of flood control in the Nuclear Annex and RB Subsphere is provided by the divisional well which serves as a barrier between redundant trains of safe shutdown systems and components. Each half of the Subsphere is further divided into two quadrants Approvmf Desip Meteria! Desip of SSC Page 3.4-2
System 80+ Design controlDocument () to sepeate redundant safe shutdown components to the extent practical. Flood barriers provide separation between Subsphere quadrants, while maintaining equipment removal capability. Emergency Feedwater pumps are located in separate compartments within the quadrants with each compartment protected by flood barriers. Penetrations are sealed and no doors are provide.d up to EL. 70+0, the maximum internal flood in the divisional wall that separates the Nuclear Annex and the Reactor Building Subsphere. Where flood doors are provided, open and close sensors are also provided with status indication provided at a central fire alarm station. Flood barriers also provide separation between electrical equipment and fluid mechanical systems at the lowest elevation within the Nuclear Annex. At higher elevations, safety-related electrical components are elevated above the floor so that flooding events will not affect components. Additional barriers (e.g., curbs, sealed penetrations) are provided or safety-related electrical components are elevated, as necessary, to mitigate the effects of postulated pipe rupture addressed in Section 3.6. Flood doors will be specified to withstand the static pressure from the maximum flood elevation as determined in the flood analysis. Flood protection is also integrated into the floor drainage system. The floor drainage systems are separated by division and Safety Class 3 valves are provided to prevent backflow of water to areas containing safety-related equipment. Each subsphere quadrant is provided with redundtnt Safety Class 3 sump pumps and associated instrumentation, which are powered from the diesel generators in the event of loss of offsite power.
,\
The Nuclear Annex floor drainage system is divisionally separated, with no common drain lines between ! ( )
'v' divisions. Floors are gently sloped to allow good drainage to the divisional sumps.
No water lines are routed above or through the control room or the computer room. HVAC water lines contained in rooms around the control room are located in rooms with raised curbs to prevent leakage l l from entering the control room. l Flood protection is incorporated into the Component Cooling Water Heat Exchanger Structure. This structure is divisionally separated by a wall such that a flood in one division can not flood the other division. The Diesel Generator Building floor drain sump pumps and associated instrumentation are Safety Class 3 to prevent flooding of the diesel generators. These pumps are also powered from the diesel generator in the event of loss of offsite power. 3.4.5 Analytical and Test Procedures A description of the methods and test procedures by which static and dynamic effects of the design basis flood conditions or design basis groundwater conditions are applied is detailed in Section 2.4 of the site-specific SAR. A site-specific flood analysis that ensures safe shutdown of the plant shall be performed. This analysis shall include the effects of flooding due to high and moderate energy pipe breaks as described in Section n 3.6.1.
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Approved Design htatorial Design of SSC Page 3.4-3
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