ML19312A207

Chapter 3 of S&W SWESSAR-P1, Design of Structures, Components,Equipment & Sys.

ML19312A207
Person / Time
Site:	05000495
Issue date:	11/29/1978
From:	NEW YORK STATE ELECTRIC & GAS CORP., STONE & WEBSTER, INC.
To:
References
NUDOCS 7909050285
Download: ML19312A207 (527)

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SWESSPR-P1 CHAPTER 3 DESIGN OF STRUCTURES, COMPONENTS, EQUIPMZNT, AND SYSTEMS LIST OF EFFECTIVE PAGES Page, Table (T), Amendment Page, Table (T) , Amendment or Fiqure (F) No. or Figure (F) No.

3-a thru c 39 T3.2.5-1 (sheet 11a) 17 3-1 2 3.3-1 Orig 3-ii 5 3.3-2 1 3-dii 1 3.3-3 7 3-iv 8 3.3-4,4A 16 3-v thru viA 21 3.3-5 21 3-vii/viii 14 T3.3.:-1 thru 7 Orig 3-viiiA 8 3.4-1 7 3-ix 4 3.4-2 21 3-x 7 3.5-1 32 3-xi thru xiv 24 3.5-2 & 2A 35 3-xv 7 3.5-3 14 3-xvi 35 3.5-4 25 3-xvii 18 3.5-5 35 3.1-1/2 8 T3.5.1-1 Orig 3.1-3 thru 5 Oria~

T3.5.2-1 8 3.1-6, 6A 1 T3 . 5 . 3- 1 19 3.1-7/8 Orig T3.5.4-1 14 3.1-9 11 F3.5.4-1 7 3.1-10, 10A 2 F3.5.4-2 14 3.1-11 Orig F3.5.4-3 7 3.1-12 8 3.6-1 20 3.1-13 2 o W 3.6-2 & 2A 35 3.1-14 thru 18 Orig 3.6-3/4 8 3.1-19 thru 20A 5 3.6-4A thru 4G 24 3.1-21 thru 26 Orig 3.6-5 thru 7 Orig 3.1-27 8 3.6-8,8A 8 3.1-28 thru 30 Orig 3.6-9/10 2 3.1-31 8 # 3.6-11 35 3.1-32 thru 35 Orig T3.6-1 (2 sheets) 7 3.1-36 8 T3.6-2 7 3.1-37 11 W T3.6-3 35 3.1-38,39 8 3.7-1 11 3.2-1 Orig 3.7-2 7

3. 2 -2, 2A 5 3.7-3 16 3.2-3 7 3.7-4 thru 4C 15 3.2-4, 5 4 3.7-5 thru 7 7 T3.2.2-1 (2 sheets) 17 3.7-8, 8A 16 T3.2.5-1 (sheet 1) 2 3.7-9/10 Orig T3.2.5-1 (sheet 2) 24 3.7-11 16 T3.2.5-1 (sheets 364) 2 3.7-12 thrc 14B 5 T3.2.5-1 (cheets 5 4 3.7-15 thru 16A 7 thru 10) 3.7-17 16 T3.2.5-1 (eheet 11) 7 3.7-18 thru 18B 7 3-a Amendment 39 f6l 7

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SWESSAR-P1 LIST OF EFFECTIVE PAGES (CONT)

Page, Table (T) , Amendment Page, Table (T) , Amendment or Figure (F) No. or Figure (F) No.

3.7-19 2 3.8-40/40A 32 3.7-20 5 3.8-41 thru 43 4 3.7-20A 14 3.8-44,45 8 3.7-20B 15 3.8-46 24 3.7-21 2 3.8-47 7 3.7-22 thru 27 Orig 3.8-48 24 3.7-28 4 3.8-49 25 3.7-29 thru 30A 7 T3.8.1-1 4 3.7-31 Orig T3.8.1-2 (2 sheets) 4 3.7-32,32A 7 T3 . 8 .1-3 4 3.7-33 4 T3.8.1-4 thru 7 5 3.7-34 thru 38 Orig T3.8.3-1 (6 sheets) 7 3.7-39 25 T3.8. 6- 1 (W-3S) 24 3.7-40 7 (3 sheets)

T3.7.1-1 Orig F3.8.1-1 8 T3.7.1-2 29 F3.8.1-2 thru 5 8 T3.7.2-1 c F3.8.1-6 thru 8 4 T3.7.2-2 5 F3.8.1-9 6 T3.7.2-3 16 F3.8.1-10 3 T3.7.3-1 thru 5 Orig F3 . 8 .1 -11 Orig T3.7.3-6 5 F3.8.1-12 6 13 4 T3.7.6-1 (W) 25 F3.8.1-14 Orig T3.7.6-1 (W-3S) 24 F3.8.1-15 9

'13.7.6-1 (BSW) (sheet 1) 30 F3.8.1-16 & 17 Orig T3.7.6-1 (BSW) (Sheet 2) 34 F3.8.1-18 (2 sheets) 8 F3.7.1-1 thru 12 Orig F3.8.1-19 (sheet 1) Orig F3.7.2-1 7 F3.8.1-19 (sheet 2) 8 F3.7.2-2 thru 4 Orig F3.8.1-20 33 F3.7.2-566 5 F3.8.1-21 8 F3.7.2-768 7 F3.8.3-1 8 F3.7.3-1 thru 5 Orig F3.3.3-2 6 3 4 3.8-1 9 F3.8.3-4 8 3.8-2 13 F3.8.3-5 (BSW, C-E) 23 3.8-3 thru 6 4 F3.8.5-162 8 3.8-7 13 F3.8.5-3 Orig 3.8-8,9 4 F3.8.5-4 8 3.8-10 thru 13 13 3.9-1 16 3.8-14 9 3.9-2,2A 20 3.8-15 4 3.9-3/4 5 3.8-16 8 3.9-4A 6 4B 20 3.8-17/18 9 3.9-5 17 3.8-19 thru 22 4 3.9-6,7 16 3.8-23 13 3.9-8, 9 Orig 3.8-24,25 4 3.9-10 30 3.8-26 thru 28 13 3.9-11 Orig 3.8-29 thru 34 4 3.9-12 30 3.8-35 thru 36A 26 3.9-13 thru 14A 20 3.8-37 4 3.9-15 Orig 3.8-38,38A 7 3.9-16 28 3.8-39 18 T3.9.1-1 Orig 3-b Amendment 39

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SWESSAR-P1 LIST OF EFFECTIVE PAGES (CONT)

Page, Table (T) , Amendment Page, Table (T) , Amendment or Figure (F) No. or Fiqure (F) No.

T3.9.1-2 (2 sheets) Orig T3.9.2-1 (2 sheets) 17 T3.9.2-2 (sheet 1) 16 T3.9.2-2 (sheet 2) Orig T3.9.2-3 Orig 3.10-1 16 3.10-2 4 3.11-1 32 3.11-2,2A 19 3.11-3/4 11 3.11-4A 4 3.11-5/6 4 T3.11-1 29 661 279 3-c Amendment 39 7/14/78

SWESSAR-P1 CHAPTER 3 DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS TABLE OF CONTENTS Section Page 3.1 CONFORMANCE WITII AEC GENERAL DESIGN CRITERIA 3.1-1 3.1.1 Quality Standards and Records (Criterion 1) 3.1-1 3.1.2 Design Bases for Protection Against Natural Phanomena (Criterion 2) 3.1-2 3.1.3 Fire Protection (Criterion 3) 3.1-4 3.1.4 Environmental and Missile Design Bases (Criterion 4) 3.1-4 3.1.5 Sharing of Structures, Systems, and Components (Criterion 5) 3.1-5 3.1.6 Not Assigned 3.1-6 3.1.7 Not Assigned 3.1-6 3.1.8 Not Assigned 3.1-6 3.1.9 Not Assigned 3.1,6 3.1.10 Reactor Design (Criterion 10) 3.1-6 3.1.11 Reactor Inherent Protection (Criterion 11) 3.1-6 3.1.12 Suppression of Reactor Power Oscillations (Criterion 12) 3.1-6 A l2 3.1.13 Instrumentation and Control (Criterion 13) 3.1-7 3.1.14 Reactor Coolant Pressure Boundary (Criterion 14) 3.1-7 3.1.15 Reactor Coolant System Design (Criterion 15) 3.1-7 3.1.16 Containment Design (Criterion 16) 3.1-8 3.1.17 Electric Power Systems (Criterion 17) 3.1-8 3.1.18 Inspection and Testing of Electric Power Systems i (Criterion 18) 3.1-10 2 3-i Amendment 2

. oPD 8/30/74

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SNESSAR-P1 TABLE OF CONTENTS (CONT)

Section Page 3.1.19 Control Room (Criterion 19) 3 1-11 3.1.20 Protection System Functions (Criterion 20) 3.1-12 3.1.21 Protection System Reliability ar.d Testability (Criterion 21) 3.1-13 3.1.22 Protection System Independence (Criterion 22) 3.1-13 3.1.23 Protection System Failure Modes (Criterion 23) 3.1-14 3.1.24 Separation of Protection and Control Systems (Criterion 24) 3.1-15 3.1.25 Protection System Requirements for Reactivity Control Malfunctions (Criterion 25) 3.1-16 3.1.26 Reactivity Control System Redundancy and Capability (Criterion 26) 3.1-16 3.1.27 Combined Reactivity Control Systems Capability (Criterion 27) 3.1-16 3.1.28 Reactivity Limits (Criterion 28) 3.1-17 3.1.29 Protection against Anticipated Operational Occurrences (Criterion 29) 3.1-17 3.1.30 Quality of Reactor Coolant Pressure Boundary (Criterion 30) 3.1-17 3.1.31 Practure Prevention of Reactor Coolant Pressure Boundary (Criterion 31) 3.1-18 3.1.32 Inspection of Reactor Coolant Pressure Boundary (Criterion 32) 3.1-18 3.1.33 Reactor Coolant Makeup (Criterion 33) 3.1-18 3.1.34 Residual Heat Removal (Criterion 34) 3.1-19 3.1.35 Emergency Core Cooling (Criterion 35) 3.1-19 3.1.36 Inspection of Emergency Core Cooling System (Criterion 36) 3.1-20 3.1.37 Testing of Emergency Core Cooling System (Criterion 37) 3.1-20 1

5 3.1.38 Containment Heat Removal (Criterion 38) 3.1-20A 3-ii Amendment 5 12/2/74 g/. .n:

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SWESSAR-P1 TABLE OF CONTENTS 3.1.39 Inspection of Containment Heat Removal System (Criterion 39) 3.1-21 3.1.40 Testing of Containment Heat Removal System (Criterion 40) 3.1-21 3.1.41 Containment Atmosphere Cleanup (Criterion 41) 3.1-22 3.1.42 Inspectior. of Containment Atmosphere Cleanup Systems (Criterion 42) 3.1-23 3.1.43 Testing of Containment Atmosphere Cleanup Systems (Criterion 43) 3.1-23 3.1.44 Cooling Water (Criterion 44) 3.1-24 3.1.45 Inspection of Cooling Water System (Criterion 45) 3.1-25 3.1.46 Testing of Cooling Water System (Criterion 46) 3.1-25 3.1.47 Not Assigned 3.1-25 3.~1.48 Not Assigned 3.1-26 3.1.49 Not Assigned 3.1-26 3.1.50 Containment Design Basis (Criterion 50) 3.1-26 3.1.51 Fracture Prevention of Containment Pressure Boundary (Criterion 51) 3.1-27 3.1.52 Capability for Containment Leakage Rate Testing (Criterion 52) 3.1-27 3.1.53 Provisions for Containment Testing and Inspection (Criterion 53) 3.1-28 3.1.54 Systems Penetrating Containment (Criterion 54) 3.1-28 3.1.55 Reactor Coolant Pressure Boundary Penetrating Containment (Criterion 55) 3.1-29 3.1.56 Primary Containment Isolation (Criterion 56) 3.1-30 3.1.57 Closed Systen Isolation Valves (Criterion 57) 3.1-31 3.1.58 Not Assigned 3.1-31 3.1.59 Not Assigned 3.1-32 3.1.60 Control of Releases of Radioactive Materials 3-iii Amendment 1

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SWESSAR-P1 TABLE OF CONTENTS (CONT)

Section Page to che Environment Criterion 60) 3.1-32 3.1.61 Fuel Storage and Handling and Radioactivity Control (Criterion 61) 3.1-33 3.1.62 Prevention of Criticality in Fuel Storage and Handling (Criterion  ; 3.1-34 3.1.63 Monitoring Fuel and Waste Storage (Criterion 63) 3.1-35 3.1.64 Monitoring Radioactivity Releases (Criterion 64) 3.1-36 3.1.65 Single Fa ilure Criterion 3.1-35 8

3.1.65.1 Definitions 3.1-36 3.1.65.2 Application of Single Failure Criterion 3.1-37 3.2 CLASSII CATION OF STRUCTURES, SYSTEMS, AND COMPONEUTS 3.2-1 3.2.1 Seismic Classification 3.2-1 3.2.2 System Ouality Group Classification (ANSI System Safety Classification) 3.2-1 3.2.3 Quality Assurance Categories 3.2-3 3.2.4 Other Classification Systems 3.2-4 3.2.5 Tabulation of Codes and Classifications 3.2-5 Ref erences for Section 3.2 3.2-5 3.3 WIND AND TORNADO LOADINGS 3.3-1 3.3.1 Wind Loadings 3.3-1 3.3.1.1 Design Wind Velocity 3.3-1 3.3.1.2 Basis for Wind Velocity Selection 3.3-1 3.3.1.3 Vertical Velocity Distribution and Gust Factor 3.3-1 3.3.1.4 Determination of Applied Forces 3.3-1 3-iv Amendment 8 9

3/28/75 00! on:

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SWESSAR-P1 TABLE OF CONTENTS (CONT)

Section Page 3.3.2 Tornado Loadings 3.3-3 3.3.2.1 Applicable Design Parameters 3.3-3 3.3.2.2 Determination of Forces on Structures 3.3-3 3.3.2.3 Ability of Seismic Category I Structures to Perform Despite Failure of Structures Not Designed for Tornado Loads 3. 3 -5 3.3.3 Interface Requirements 3.3-5 21 References for Section 3.3 3.3-5 3.4 WATER LEVEL (FLOOD) DESIGN 3.4-1 3.4.1 Flood Protection 3.4-1 3.4.2 Analysis Procedures 3.4-1 3.4.3 Interface Requirements 3.4-2 21 3.5 MISSILE PROTECTION 3.5-1 3.5.1 Missile Barriers and Loadings 3.5-1 3.5.2 Missile Selection 3.5-1 3.5.2.1 Internal Missiles 3.5-1

.~,.5.2.2 External Missiles 3.5-2 3.5.3 Selected Missiles 3.5-2 3.5.3.1 Selected Internal Missiles 3.5-2 3.5.3.2 Selected External Missiles 3.5-2 3.5.4 Barrier Design Procedures 3.5-2 3.5.5 Missile Barrier Features 3.5-5 3.5.6 Interface Requirements 3.5-S I 21 i

References for Section 3.5 3.5-5 9

3-v Amendment 21 2/29/76 4s

SWESSAR-P1 TABLE OF CONTENTS (CONT)

Section Page 3.6 PROTECTION AGAINST DYNAMIC EFFECTS ASSOCIATED WITH THE POSTULATED RUPTURE OF PIPING 3.6-1 3.6.1 Systems in which Design Basis Piping Breaks Occur 3.6-1 3.6.2 Design Basis Piping Break Criteria 3.6-1 3.6.2.1 Criteria for Inside the Containment 3.6-1 3.6.2.1.1 Break Locations - Class 1 Piping 3.6-2 3.6.2.1.2 Break Locations - Class 2 and 3 Piping 3.6-3 3.6.2.2 C> M eria for Outside the Containment 3.6-4 3.6.2.2.1 High Energy Fluid Systems 3.6-4 3.6.2.2,2 Moderate Energy Fluid Systems 3.6-4C 3.6.2.3 Design Basis Break / Crack Types and Orientation 3.6-4D 3.6.2.3.1 Circumferential Pipe Bmaks 3.6-4D 3.6.2.3.2 Iongitudinal Pipe Breaks 3.6-4E 3.6.2.3.3 Through Wall Leakage Cracks 3.6-4F 3.6.3 Design Loading Combinations 3.6-4G 3.6.4 Dynamic Analysis 3.6-5 3,6.5 Protective Measures 3.6 21 3.6.6 Interface Requirements 3.6-11 3.7 SEISMIC DESIGN 3.7-1 3.7.1 Seismic Input 3.7-1 3.7.1.1 Design Response Spectra 3.7-1 3.7.1.2 Design Response Spectra Derivation 3.7-1 3.7.1.3 Critical Damping Val.ues 3.7-2 9

3-vi Amendment 21 2/20/76 6i 135

SWESSAR-P1 TABLE OF CONTENTS (CONT)

Section Paqe 3.7.1.4 Bases for Site-Dependent Analysis 3.7-3 3.7.1.5 Soil-Supported Category I Structures 3.; -3 3.7.1.6 Soil-Structure Interaction 3.7-3 3.7.2 Seismic System Analysis 3.7-3 3.7.2.1 Seismic Analysis Methods 3.7-3 3.7.2.1.1 Seismic Analysis of Structures 3.7-3

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SWESSAR-P1 TABLE OF CONTENTS (CONT)

Section Page 3.7.2.1.2 Seismic Analysis Methods (Components) 3.7-8 3.7.2.1.3 Seismic Analysis of Piping Systems 3.7-16 f14 3.7.2.2 Natural Frequencies and Response Loads 3.7-16A 3.7.2.3 Procedures Used to Lump Masses 3.7-16A 3.7.2.4 Rocking and Translational Response Summary 3.7-17 3.7.2.5 Methods Used to Couple Soil _with Seismic System Structures 3,7-17 3.7.2.6 Developnent of Floor Response Spectra 3.7-17 3.7.2.7 Differential Seismic Movement of Inter-connected Components 3.7-18A 3.7.2.8 Effects of Variations on Floor Response Spectra 3.7-18A 94 3.7.2.9 Use of Constant Vertical Load Factors 3.7-18A 3.7.2.9.1 Structures, Equipment, and Components 3.7-18A 3.7.2.9.2 Piping System 3.7-18A 3.7.2.10 Method Used to Account for Torsional Effects 3.7-191 3.7.2.11 Ccxnparison of Responses  ?.1-18B s

3.7.2.12 Methods for Seismic Analysis of Dams 3.7-19 3.7.2.13 Methods to Determine Category I Structure Overturning Moments 3.7-19 347.2.14 7atalysis Procedure for Damping 3.7-19 3.7.3 Seismic Subsystem Analysis 3.7-19 3.7.3.1 Determination of Number of Earthquake cycles 3.7-19 3.7.3.1.1 Equipment 3.7-19 is 3.7.3.1.2 Piping Systems 3.7-20 3.7.3.2 Basis for Selection of Forcing Frequencies 3.7-20A 3-vii Amendment 14 4./4

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7/18/75

SWESSAR-P1 .

TABLE OF CONTENTS (CONT)

Page Section 3.7.3.2.1 Equipment 3.7-20A 3.7.3.2.2 Piping Systems 3.7-20A 3.7.3.3 Root Mean Square Basis 3.7-20A 3.7.3.4 Procedure for Combining Modal Responses 3.7-20A 3.7.3.5 Significant Dynamic Response Modes 3.7-21 3.7.3.5.1 Equipment 3.7-21 3.7.3.5.2 Piping Systems 3.7-25 3.7.3.6 Design Criteria and Analytical Procedures for Piping 3.7-25 3.7.3.7 Basis for Computing Combined Response 3.7-30 14 3.7.3.1 Equipment 3.7-30 3.7.3.7.2 Piping Systems 3.7-30 3.7.3.8 Amplified Seismic Responses 3.7-30 3.7.3.8.1 Equipment 3.7-30 3.7.3.8.2 Piping Syste=s 3.7-30 14l 3.7.3.9 Use of Simplified Dynamic Analysis 3.7-30 3.7.3.10 Modal Period Variation 3.7-31 3.7.3.11 Torsional Effects of Eccentric Masses 3.7-31 3.7.3.12 Piping Outside Containment Structure and Annulus Building 3.7-32 3.7.3.13 Interaction of Cther Piping with Category I Piping 3.7-32A 14

. 3.7.3.14 Field Incation of Supports .nd Restraints 3.7-32A 3.7.4 Seismic Instrumentation Program 3.7-33 3.7.4.1 cmparison with Regulatory Guide 1.12 3.7-33 3.7.4.2 Incation and Description of Instrumentation 3.7-33 3-viii Amendment 14 7/18/75 b6l 200

SWESSAR-P1 TABLE OF CONTENTS (CONT)

Section Page 3.7.4.3 Control Room Operator Notification 3.7-36 3.7.4.4 Ccuparison of Measured and Predicted Responses 3.7-36 3.7.5 Seismic Design Control 3.7-37 3.7.5.1 Data Origination 3.7-38 3.7.5.2 Components and Equipnerat 3.7-38 3.7.5.3 Piping 3.7-39 3.7.6 Interface Information 3.7-39 8 l

References for Section 3.7 3.7-39 3.8 DESIGN OF CATEGORY I STRUCTURES 3.8-1 3.8.1 Concrete Containment Structure 3.8-1 3.8.1.1 Description of Containment 3.8-1 bb1 29 4

3-viiiA Amendment 8 3/28/75

SWESSAR-P1 TABLE OF CONTENTS (CONT)

Section Page 3.8.1.1.1 General Description 3.8-1 3.8.1.1.2 Reinforcing Steel Arrangement 3.8-2 4 3.8.1.1.3 Steel Liner and Penetrations 3.8-3 3.8.1.2 Applicable Codes, Standards, and Speci- 3.8-9 fications 3.8.1.2.1 Concrete Structure 3.8-9 3.8.1.2.2 Steel Liner, Penetrations, and Access 3.8-13 Openings 3.8.1.3 Loads and Loading Combinations 3.8-13 3.8.1.3.1 Concrete Structure 3.8-14 3.8.1.3.2 Steel Liner and Penetrations 3.8-15 3.8.1.4 Design and Analysis Procedures 3.8-16 3.8.1.4.1 Concrete Structure 3.8-16 3.8.1.4.2 Steel Liner and Penetrations 3.8-21 3.8.1.5 Structural Acceptance Criteria 3.8-23 3.8.1.5.1 Concrete Structure 3.8-23 3.8.1.5.2 Steel Liner and Penetrations 3.8-24 4

3.8.1.6 Materials, Quality Control, and Special 3.8-26 Construction Techniques 3.8.1.6.1 Concrete 3.8-26 3.8.1.6.2 Reinforcing Steel 3.8-27 3.8.1.6.3 Steel Liner and Penetrations 3.8-30 3.8.1.7 Testing and Inservice Surveillance 3.8-33 Requirements 3 . 8 .1. 7 .1 Concrete Structure 3.8-33 3.8.2 Steel Containment System 3.8-34 3.8.3 Concrete and Structural Steel Internal 3.8-34 3-ix Amendment 4 661 290 11/1/74

SWESSAR-P1 TELE OF CONTENTS (CONT)

Section Page Structures of Concrete Containment 3.8.3.1 Description of Internal Structures 3.8-34 3.8.3.2 Applicable Codes, Standards, and 3.8-35 Specifications 3.8.3.3 Loads and Ioading Combinations 3.8-36 3.8.3.4 Design and Analysis Procedures 3.8-37 3.8.3.5 Structural Acceptance Criteria 3.8-37 3.8.3.6 Materials, Quality Control, and Special 3.8-38 Construction Techniques 3.8.3.7 Testing and Inservice Surveillance 3.8-39 Requirements 3.9.4 Other Category I Structures 3.8-39 3.8.4.1 Description of the Structures 3.8-39 3.8.4.2 Applicable Codes, Standards, and Speci- 3.8-42 fications 3.8.4.3 Loads and Loading Combinations 3.8-42 O 3.8.4.4 Design and Analysis Procedures 3.8-43 3.8.4.5 Structural Acceptance Criteria 3.8-43 3.8 4.6 Materials, Quality Control, and Special 3.8-44 Construction Techniques 3.8.3.7 Testing and Inservice Surveillance 3.8-44 Requirements 3.8.5 Foundations and Concrete Supports 3.8-44 3.8.5.1 Description of the Foundations and Supports 3.8-44 3.8.5.2 Applicable Codes, Standards, and Speci- 3.8-46 fications 3.8.5.3 Loads and Loading Combinations 3.8-46 3.8.5.4 Design and Analysis Procedures 3.8-46 3.8.5.5 Structural Acceptance Criteria 3.8-47 9

3-x Amendment 7 2/28/75

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SWESSAR-P1 TABLE OF COITTENTS (COtTT)

Section Page 3.8.5.6 Materials, Quality Control, and Special 3.8-48 Construction Techniques 3.8.5.7 Testing and Inservice Surveillance 3.8-48 24 Requirements 3.8.6 Structural Interfaces 3.8-48 References for Section 3.8 3.8-48 3.9 MECHANICAL SYSTEMS AND COMPONENTS 3.9-1 3.9.1 Dynamic System Analysis and Testing 3.9-1 3.9.1.1 Vibration Operational Test Program 3.9-1 3.9.1.2 Dynamic Testing Procedures 3.9-1 3.9.1.3 Dynamic System Analysis Methods for Reactor 3.9-1 Internals 3.9.1.4 Correlation of Test and Analytical Results 3.9-1 3.9.1.5 Analysis Methods under LOCA Loadings 3.9-2 3.9.1.6 Analytical Methods for ASME Code Class 1 3.9-2 Components 3.9.1.6.1 Equipment 3.9-2 3.9.1.6.2 Piping System 3.9-4 3.9.2 ASME Code Class 2 and 3 Component's 3.9-4A 3.9.2.1 Plant Conditions and Design Loading 3.9-4A Combinations 3.9.2.1.1 Components 3.9-4A 3.9.2.1.2 Piping Systems 3.9-6 3.9.2.2 Design Loading Combinations 3.9-6 3.9.2.2.1 Components 3.9-6 3.9.2.2.2 Piping Systems 3.9-7 3.9.2.3 Design Stress Limits 3.9-7 3-xi Amendment 24 4/23/76

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SWESSAR-P1 TABLE OF CONTENTS (CONT)

Section Page 3.9.2.3.1 Components 3.9-7 3.9.2.3.2 Piping Systems 3.9-8 3.9.2.4 Analytical and Empirical Methods for Design 3.9-8 of Pumps and Valves 3.9.2.5 Design and Installation Criteria, Pressure 3.9-14 Relieving Devices 3.9.2.5.1 Open Relief System 3.9-14 3.9.2.5.2 Closed Relief System 3.9-14A 3.9.2.6 Stress Levels for Category I Components 3.9-15 3.9.2.7 Field Run Piping Systems 3.9-15 3.9.3 Components Not Covered by ASME Code 3.9-15 3.9.3.1 Cumponents 3.9-15 3.9.3.2 Piping Systems 3.9-16 g 3.9.4 Structural and Seismic Interface 3.9-16 Requirements 3.10 SEISMIC DESIGN OF CATEGORY I INSTRUMENTATION 3.10-1 AND LLECTRICAL EQUIPfENT 3.10.1 Seismic Design Criteria 3.10-1 3.10.2 Seismic Analysis, Testing Procedures, and 3.10-1 Restra.'.nt Measures 3.11 ENVIROh.' ENTAL DESIGN OF MECHANICAL AND 3.11-1 ELECTRICAL EQUIPMENT 3.11.1 Equipment Identification 3.11-1 3.11.2 Qualification Tests and Analyses 3.11-2 3.11.3 Qualification Test Results 3.11-5 3.11.4 Loss of Ventilation 3.11-5 bb 93 3-xii Amendment 24 4/23/76

SWESSAR-P1 LIST OF TABLES Table 3.2.2-1 Nonseismic Category I Components Inside the Containment Structure 3.2.5-1 List of QA and Seismic Category I Structures, Systems, and Components 3.3.1-1 Ef.fective Velcaity Pressures 3.3.1-2 Wind Pressure C3efficients for Rectangular Structures 3.3.1-3 Wind Pressure C.)efficients for Cylindrical Structures (HeiJht ' Diameter = 1) 3 . 3 .1 -4 External Wind Pressure Coefficients (C ) ,

Containment Dome 3.3.1-5 Resultant Design Wind Pressure (W) for Rectangular Structures 3.3.1-6 Resultant Design Wind Pressure (W) for Cylindrical Structures (Areas >

1,000 sq ft)

?.3.1-7 Result ant Design Wind Pressure (W) , Containment Dome 3.5.1-1 External Missile Barriers 3.5.2-1 Tornado Generated External 1:4.ssiles 3 . 5 . 3 -1 Kinetic Energy of Potential MissD es 3.5.4-1 Deleted 3.6-1 Modcrate and High Energy Pipe - Outside Containment 3.6-2 High Energy Pipe - Inside Containment 3.6-3 NSSS Break Criteria 24 3.7.1-1 Damping Factors 3.7.1-2 Acceptable Methods for Soil - Structure Interaction Analysis 3.7.2-1 Base Shear & Overturning Moment for Containment Wall 3-xiii Amendment 24

(, b \ ,, n ',b L/ 4/23/76

SWESSAR-P1 LIST OF TABLES (CONT)

Table p 3 . 7 . 2 -2 Comparison of Natural Frequencies for Containment Wall 3.7.2-3 Equipment Seismic Qualification 3.7.3-1 1G Flat Response 3.7.3-2 Modal Density, n 3 . 7 .3 -3 Amplified Response Dynamic Factor Study 3.7.3-4 Piping System Seismic Design and Analysis Criteria 3.7.3-5 Design Loading Combinations and Stress Limits for Seismic Category I Piping System 3 . 7 . 3 -6 Cycles of Motion for Large Earthquakes 3.7.6-1 Seismic Design Interface Requirements 3.8.1-1 Ioad Combinations for Containment Liner Insert and Overlay Plates, Brackets, and Attachments 3.8.1-2 Load Combination and Allowables 3.8.1-3 Loading Ccxnbination and Limits for Pipe Rupture 3.8.3-1 Structures Loading Criteria 3.8.6-1 Structural Interfaces 3.9.1-1 Stress Limits for ASME III Class 1 (NB) Seismic Category I Components (Elastic Analysis) 3.9.1-2 Summary of Regulatory Guide 1.48 Requirements -

ASME Code Class 1 Components 3.9.2-1 Stress Limits for ASME III Class 2 and 3 S&W Components and Component Supports (Elastic Analysis) 3.9.2-2 Comparison of Class 7 and 3 Requirements -

Regulatory Guide 1.48 vs SAR Table 3.9.2-1 3.9.2-3 Design Criteria for ASME III Class 2 and 3 Active Pumps 3.11-1 Environmental Conditions 3-xiv /Ki 'O' Amendmel.t 24 OOi c/J 4/23/76

SWESSAR-P1 9 LIST OF FIGURES Figure 3.5.4-1 Force Time History for Missile and Barrier 3.5.4-2 Time History for Missile Compression, Barrier Deflection, and Barrier Ductility 3.5.4-3 Time History for Energy Distribution 3.7.1-1 Response Spectra, Horizontal .30g 3.7.1-2 Response spectra, Vertical .30g 3.7.1-3 Horizontal .30g (10.0 Percent Damping) 3.7.1-4 Horizontal .30g (7.0 Percent Damping) 3.7.1-5 Horizontal .30g (5.0 Percent Damping) 3.7.1-6 Hcrizontal .30g (2.0 Percent Damping) 3.7.1-7 Horizontal .30g (1.0 Percent Damping) 3.7.1-8 Vertical .30g (10.0 Percent Damping) 3.7.1-9 Vertical .30g (7.0 Percent Damping) 3.7.1-10 Vertical .30g (5.0 Percent Damping) 3.7.1-11 Vertical .30g (2.0 Percent Damping) 3.7.1-12 Vertical .30g (1.0 Percent Damping) 3.7.2-1 Dynamic Model 3.7.2-2 Design Response Spectra for Standard Site Characteristics, Umbrella Spectrum 3.7.2-3 Alternative Equipment and Piping Standard Plant Design Spectra, Maximum Modal Response 3.7.2-4 Alternative Equipment and Piping Standard Plant Design Spectra, Maximum Site Response 3.7.2-5 Idealized Containment Structure 3.7.2-6 Containment Lumped Mass Models 3.7.2-7 Correlation Factor Between Two Horizontal Earthquakes

(> f3 $ 9D 3-xv Amendment 7 2/28/75

SWESSAR-P1 LIST OF FIGURES (CONT)

Fiqure 3.7.2-8 Correlation Factor Between Vertical and Horizontal harthquakes 3.7.3-1 hepresentation of Family of Peak Response Curves within Broadened Resonant Peak 3.7.3-2 Hypothetical vs Actual Response of Multiple Modes within Broadened Response Peak 3.7.3-3 Justification of Static Load Factor 3.7.3-4 Model Beams 3.7.3-5 Typical Amplified Response Spectra 35 3.8.1-1 Radial Shears Bars in the Containment Wall 3.8.1-2 Dlagonal Rein 1orcing, Containment Structure 3.8 1-3 Containment Wall-Foundation Juncture, Typical Detail 3.8.1-4 Typical Detall of Dome-Cylinder Junction 3.8.1-5 Typical Detall of Concrete to Steel Ring at Apex of Dome

3. 8 .1 - t> Corner Transition Section and Bridging Plates 3.8.1-7 Section - Typical Bridging Bar 3.8.1-8 Section - Typical Bridging Plate 3.8.1-9 Typical L1ner Details 3.8.1-10 Piping System Penetrations 3.8.1-11 Fuel Transfer Tube Enclosure 3.B.1-12 Typical Electrical Penetration Sleeve 3.8.1-13 Equipment Hatch 3.8.1-14 Personnel Hatch 3.8.1-15 study for 312-f t Diameter Mat 3.6.1-16 Comparison of SHELL 1 vs Hand Calculations, Moment and Shear 3-xvi i.b I 'lb Amendment 35 10/6/77

SWESSAR-P1 LIST Ol' FIGURES (CONT)

Fiqure 3.8.1-17 Comparison of SHELL 1 vs Hand Calculations 3.8.1-18 Reinf'>rcing Details, Equipment Access Hatch Opening, Sheet 1 3.8.1-18 Reinforcing Details, Sections through Ring Beam to Equipment Access Hatch, Sheet 2 3.8.1-19 Reinforcing Detail, Personnel Hatch Opening, Sheet 1 3.8.1-19 Reinforcing Details, Sections through Ring Beam of Personnel Access Lock, Sheet 2 3.8.1-20 Radial Deflection Measurements for Thickened Ring Beam 3.8.1-21 Typical Letail of Embedment for Polar Crane Lateral Supports 3.8.3-1 Structural Load Transfer Mechanisms - 1 3.8.3-2 Structural Load Transfer Mechanisms - 2 3 . 8 . 3 -3 Structural Load Transfer Mechanisms - 3 3.8.3-4 Seismic Restraints for Polar Crane 3.8.3-5 Typical Guard Pipe Details, BSW, C-E 18 3.8.5-1 Mat Reinforcing Patterns 3.8.5-2 Groundwater Drainage Under Containment S'..cucture 3.8.5-3 Typical Detail Wall / Column to Foundation Mat Junction 3.8.5-4 Structural Interfaces with Large Equipnent Supports 4 661 297 3-xvii Amendment 18 10/30/75

N O

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66\

SWESSAR-P1 CHAPTER 3 DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENP, AND SYSTEMS 3.1 CONFORMANCE WITH AEC GENERAL DESIGN CRITERIA The design of the PWR Standard Plant conforms as outlined below to Appendix A, 10CFR50, General Design Criteria for Nuclear Power Plants, as published in the Federal Register on February 20, 1971, and as amended in the Federal Register on July 7, 1971.

Conformance with the single failure criterion is included in Section 3.1.65. The conformance with AEC Regulatory Guides is 8 discussed in Appendix 3A to this chapter.

3.1.1 Quality Standards and Records (Criterion 1)

Structures, systems, and components important to safety shall be designed, f abricated, 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 canponents important to safety shall be maintained by or under the control of the nuclear power unit licensee throughout the life of the unit.

Discussion Structures, systems, and components important to safety are designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed. Discussions relating to the design approach to the important safety related features of the plant are presented in the following sections.

Quality standards applicable to safety related stru ctures ,

systems, and components are generally contained in codes such as the ASME Boiler and Pressure Vessel Code. The applicability of these codes is specifically identified throughout this report and is summarized in Section 3.2.5. The procedures for identifying codes for applicability are described in Chapter 17.

Chapter 17 describes the Stone S Webster Qu11ity Assurance Program established to provide assurance that safety related systems, and components satisfactorily perform their 9 structures, intended safety f unctions . That chapter also describes the 3.1-1 Amendment 8

<j 90C C// 3/28/75 (10 t

SWESSAR-P1 procedures for generating and maintaining appropriate design, f abrication, erection, and testing records.

Reference Sections Title No.

Classification of Structures, Systems, and Components 3.2 Quality Assurance 17 3.1.2 Design Bases for Protection Against Natural Phenomena (Criterion 2)

Structures , systems, and components important to saf ety shall be designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches without loss of capability to perform their rafety fun ctions .

The design bases for these structures, systemc, and components shall reflect: (1) appropriate consideration of the most severe of the natural phenomena that have been historica~ 1y reported for the site and surrounding area, with sufficient margin for the limited accuracy, quantity, and period of time in which the historical data have been accumulated, (2) appropriate combinations 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.

Discussion Those features of plant facilities that are essential to the prevention of accidents that could affect the public health and safety or to the mitigation of their consequences are designed to:

1. Quality standards that reflect the importance of the function to be performed. Approved design codes are used when appropriate to the nuclear application.

2. Performance standards that enable the facility to withstand, without loss of the capability to protect the public, the additional forces imposed by the most severe earthquake, flooding condition, wind, ice, or other natural phenomena characteristic of the proposed site, and credible conbinations of the ef fects of normal and accident conditions with the effects of the natural phenomena.

Features of the facility essential to accident prevention and mitigation of accident consequences, which are designed to withstand the ef fects of natural phenomena, are:

1. The reactor coolant pressure boundt ry and containment barriers

~

bbl )bU /5

SWESSAR-P1

2. The controls and emergency cooling systems whose functions are to maintain the integrity of these barriers i

Systems which depressurize the containment following a LOCA

4. Power supply and essential services

5. The components employed to retain and store high level radioactive wastes which if released would result in a site boundary dose exceeding 0.5 Rem whole body.

6. Reactivity systems, monitoring systems, and fuel systems

7. The components used to store und cool spent reactor fuel All piping, components, and supporting structures of the reactor and safety related systems are designed to withstand a specified seismic disturbance and credible combinations of effects of normal and accident conditions with the effects of natural phenomena. Plant design criteria specify that there shall be no loss of function of such equipment in the event of the SSE ground acceleration acting in the horizontal and vertical directions simultaneously. The dynamic response of Seismic Category I structures to ground acceleration, based on an envelope of characteristics of the site foundation soils and on the critical damping of the foundation and structures, is included in the design analysis.

The containment structure is defined as a Seismic Category I structure. Structural members have sufficient capacity to accept a combination of normal operating loads, functional loads due to the design basis accident (DBA) , and the loadings imposed by the maximum wind velocity, or those due to the SSE, whichever is the larger.

The emergency onsite power sources are not subject to interruption due to earthquake, windstorm, floods, or to disturbances on the external power transmission system.

Power cabling, motors, and other equipment required for operation of the engineered safety features are suitably protected against the effects of the design basis accident (DBA) and from severe external weather conditions, as applicable.

3.1-3 661 301

SWESSAR-P1 References Sections Title No.

Classification of Structures, Systems, and Components 3.2 Wind and Tornado Loadings 3.3 Water Level (Flood) Design 3.4 Seismic Design 3.7 3.1.3 Fire Protection (Criterion 3)

Structures, systems, and components important 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 cap. city 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.

Discussion The PWR Standard Plant is designed on the basis of minimizing the use of combustible or explosive materials and maximizing the use of fire resistant materials to the greatest extent passible. The fire protection system is designed to detect, annunciate, and extinguish any probable fire which might occur.

Reference Section Title No.

Fire Protection System 9.5.1 3.1.4 Environmental and Missile Design Bases (Criterion 4)

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 los s--o f--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 equip;aent failures and from events and conditions outside the nuclear power unit.

3.1-4 (> b ; j]}

SWESSAR-P1 Discuasion Structures such as the containment structure, fuel pool, and other systems and components important to cafety are designed in accordance with the codes and classifications indicated in Section 3.1 and are shielded from or designed to withstand the impact of missiles generated from accidents or natural phenomena.

Instrumentation, motors, cables, etc are selected to meet the most adverse accident conditions to which they may be subjected.

These items are either protected from accident conditions or designed to withstand, without failure, exposure to the worst combination of temperature, pressure, and humidity expected during the required operational period.

An investigation of excessive pipe movement will be made and pipe restraints included in the design where required.

Reference Sections Title No.

Classifications of Structures, Systems, 3.2 and Components Wind and Tornado Ioadings 3.3 Missile Protection 3.5 Protection Against Dynamic Effects 3.6 Associate (. with the Postulated Rupture of Piping Seismic Design 3.7 Design of Category I Structures 3.8 Mechanical Systems and Components 3.9 Seismic Design of Category I 3.10 Instrumentation and Electrical Equipment Environmental Design of Mechanical 3.11 and Electrical Equipment 3.1.5 Sharing of Structures, Systems, and Components (Criterion 5)

Structures, systems, and components important to safety shall not be shared among nuclear power units unless it can be shown that such sharing does 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.

3.1-5 bb

SWESSAR-P1 Discussion All structures, systems, and components shared among plants are listed in Section 1.2.1. The only shared items which are safety related are the hydrogen recombiners in the combustible gas control system. As discussed in Section 6.2.5, there are two 100 percent capacity skid-mounted hydrogen recombiner subsystems located in accessible locations and able to be moved to any one of the plants following a LOCA. Failure of one of these subsystems will not prevent the combustible gas control system from performing its safety f unction. No".e of the other shared structures, systems, and components are safety related and no failure of a shared structure, system, or component can affect any system important to safety.

3.1.6 This criterion has not been promulgated by the AEC.

3.1.7 This criterion has not been promulgated by the AEC.

3.1.8 This criterion has not been promulgated by the AEC.

3.1.9 This criterion has not been promulgated by the AEC.

3.1.10 Reactor Design (Criterion 10)

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.

Discussion This criterion is within the NSSS Vendor's scope and is discussed in Section 3.1 of the NSSS Vendor's SAR.

3.1.11 Repctor Inherent Protection (Criterion 11)

The reactor core and associated coolant systems shall be designed so that in the power operating range the net effect of the promnt inherent nuclear feedback characteristics tends to compensate for a rapid increase in reactivity.

/ 7 3 l, e n , u9 3.1-6 Amendment 1 7/30/74

SWESSAL-P1 Discussion The criterion is within the NSSS Vendor's scope and is discussed in Section 3.1 of the NSSS Vendor's SAP.

3.1.12 Suppression of Reactor Power Oscillations (Criterion 12)

The reactor core and associated coolant, control, and protection systems shall be designed to assure that power oscillaticns which can result in conditions exceeding specified acceptable fuel design limits are not possible or can be reliably and readily detected and suppressed.

9 3.1-6A _. Amendment 1 bb4l 505 7/30/7a

SWESSAR-P1 Dis cussion This criterion is within the NSSS Vendor's scope and is discussed in Section 3.1 of the OTSSS Vendor's SAR.

3.1.13 Instrumentation and Control (Criterion _13)

Instrumentation and control shall be provided to monitor variables and syste s over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fismion process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems. Appropriate controls shall be pro"ided to naintain these variables and systems within prescribed oparating ranges.

Discussion Stone & Webster supplied instrumentation and control systems interface with the appropriate USSS Vendor's instrumentation and control systems to provide monitoring of the process variables of the reactor core, reactor coolant system, containment structure, and associated safety related systems for all normal operating, transient, and accident conditions. The instrumentation provides continuous monitoring, warning, and initiation of required safety functions. Indication and associated controls are provided in the control room for maintaining system variables within their prescribed ranges.

Reference Section Title No.

Instrumentation and Control 7 3.1.14 Reactor Coolant Pressure Boundary (Criterion 14)

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.

Discussion This criterion is within the NSSS Vendor's scope and is discussed in Section 1.1 of the NSSS Vendor's SAR.

3.1.15 Reactor Coolant System Design (Criterion 15)

The reactor coolant system and associated auxiliary, control, and protection systems shall be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure 3.1-7 Ibl SOh

SWESSAR-P1 boundary are not exceeded during any condition of normal operation, including anticipated operational occurrences.

Discussion This criterion is within the NSSS Vendor's scope and is discussed in Section 3.1 of the NSSS Vendor's SAR.

3.1.16 Containment Design (Criterion 16)

Reactor containment and associated systems shall be provided to establish an essentially leaktight 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 require.

Disc'tssion A steel-lined, reinforced concrete containment structure encloses the entire reactor coolant system and provides an essentially leaktight barrier. Following a DBA, the containment heat removal systems return tha containment atmosphere to a low pressure, thereby reducing the driving force for the release of radioactivity. These systems maintain a low pressure inside the containment structure for as long as -equired. During the DBA, the size of the containment ensures that the containment pressure does not exceed 48 psig , the design pressure, and that the containment atmospheric temperature does not exceed 280 F, the design temperature.

Reference Sections Title No.

Containment Functional Design 6.2.1 Containment Heat Removal Systems 6.2.2 3.1.17 Electric Power Systems (Criterion 17)

An onsite electric power system and an offsite electric power system shall be provided to permit functioning of structures, systems and components unportant to safety. The safety function for each system (assuming the other system is not functionir.g) shall be to provide sufficient capacity and capability to assure that (1) specified acceptable fuel design limits and dusign conditions of the reactor coolant pressure boundary are not exceeded as a result of anticipated operational occurrences and (2) Ue core is cooled and containment integrity and other vital functions are maintained in the event of postulated accidents.

The onsite electric power supplies. including the batteries, and the onsite electric distribution system, shall have sufficient 9

3.1-8

/,

D O ;i  ?

Un ~/.'

SWESSAR-P1 independence, redundancy, and testability to perform their safety functions assuming a single failure.

Electric power from the transmission network to the onsite electric distribution system shall be supplied by two physically independent circuits (not necessarily on separate rights of way) designed and located so as to minimize to the extent practical the 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 of f site 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 within a few seconds following a loss-of-coolant accident to assure that core cooling, containment integri ty , and other vital safety functions are maintained.

Provisions shall be inclufad to minimize the probability of losing electric power tror ar. 'if 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 tranand ssion network, or the locs of power f- a the onsite electric power sourcea.

@ Discussion Three independ<at effsite power connections, normally energized (see Section 8.3.1.1.2) and two or three redundant onsite power ,I systems (deper. ding on NSSS requirements) are provided (see Section 8.3.1.1, T) . As these sections indicate, each connection provides sufficient capability for operating all equipment that must be operated in the event of a loss of coolant accident.

The emergency trains are physically independent as shown on Fig. 8.3-2, 8.3-4, and 8.3-17 and described in Section S .3.1.4.

The electric systems are tested as the requirements for each system demand. Chapters s,6,7,9, and 10 provide individual system testing dis cussions; also see " Tests and Inspections" of Section 8.3.1.1.3.

Reference Section Title No.

Electric Power 8 3.1-9 11 3 0,eAmendnent s/30/2s 61

SWESSAR-P1 3.1.18 Inspection and Testing of Electric Power Sys tems (Criterion 18)

Electric power systems important to saf ety shall be 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 the condition of their components. The systems shall be designed with a capability to test periodically (1) the operability and functional performance of the components of the systems, such as onsite power sources, relays, switches, and buses, and (2) the operability of the systems as a whole and, under conditions as close to design as practical, the full operation sequence that brings the systems 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.

Discussion Preservice Electrical equipment is specified for manuf acture and test in strict accordance with the latest requirements of NEMA (National Electrical Manufacturers' Association), IEEE (Institute of Electrical and Electronic Engineers), or ANSI (American National Standards Institute, Inc.) standards, where applicable. ll Electrical equipment is properly protected during shipment and storage.

Installation of ?ll equipment is under the supervision of a qualified electrical construction engineer. Special attention is given to mechanical alignment and electrical ground connections.

The dielectric of insulation is measured and corrected , if necessary, before equipment is energized.

Tests and inspections will be made to ensure that all components are correct and properly mounted, connections are correct, circuits continuous, and components are operational.

Tests will be made to determine that emergency loads do not exceed the diesel generator rating and that each diesel generator is suitable for starting, and for accepting and operating the required loads.

Protective relays are set and calibrated by trained personnel of the Utility-Applicant, and metering devices are properly calibruted.

fi1 7nr DDi 0U7 3.1-10 Amendment 2 8/30/74

SWESSAR-P1 Inservice The availability and proper action of emergency equipment are tested periodically while the plant is in operation.

Testing of automatic operation of the voltage tranefer system at the 13,800 V level is performed before initial startup.

Successful operation of the 13,800 V transfer scheme does not prevent a plant shutdown but is designed to provide station auxiliary power automatically when the main generator is out of service.

Automatic starting and loading of the emergency diesel generators are an essential part of the engineered safety features and are arranged for periodic tccting.

3.1-10A Amendment 2 8/30/74

SWESSAR-P1 During power operation , the s tation batteries will be periodically checked for specific gravity and individual cell voltages. An equalizing, or overvoltage, charge is applied long enough to bring all cells up to an equal voltage. Over a period of time, these tests will reveal a weak cell or a weakening trend in any cell, and replacement will be made, as necessary. A disconnected battery or a broken cell connector would be revealed during these equalizing charges. Periodically, the battery charger will be disconnected and the ability of the battery to maintain voltage and assume the direct current load verified.

This test uncovers any high resistance connections or cell internal malfunctione.

Reference Sections Title No.

A-c Power Systems 8.3.1 D-c Power Systems 8.3.2 Emergency Power System Tests 16.4.6 3.1.19 Control Room (Criterion 19)

A control room shall be provided from which actions can be taken to operate the nuclear power 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 exposures 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 shutdown of the reactor through the use of suitable procedures.

Dis cussion A control room is provided and equipped to operate the plant safely under normal and accident conditions. Control room shielding and ventilation are designed to permit continuous occupancy of the control room for the duration of any accident including loss-of-coolant accidents without the dose to personnel exceeding 5 Rem whole body, or its equivalent to any part of the body.

An auxiliary shutdown panel with equipment, controls, and instrumentation is provided to accomplish in conjunction with 9 adjacent controls a prompt hot shutdown in a safe manner. The auxiliary shutdown panel and adjacent controls are located in the 3.1-11 _

!bf f

SWESSAR-P1 energency switchgeur area which is physically isolated f rom the control room so that any occurrence which could cause the control roou to become uninhabitable has no effect on the availability of the auxillary shutdown panel and adjacent controls. Also, equipaent , controls, and instrumentation are locatec throughout the plant to provide capability for a subsequent cold snutdown of the reactor.

The design of the control building, which houses the control room and the auxiliary shutdown panel area, conforms to the above criterion.

Reference Sections Title No.

Habitability Systems 6.4 8 Hot Shutdown Provisions 7.4.3.1 Control builcing Ventilation Systems 9.4.1 3.1.20 Protection System Functions (Critorion 20)

The protection syste.a shall be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable tuel 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 anc components important to safety.

Discussion Stone & Webster supplies protection system inputs and actuated devices which interf ace with the NSSS logic devices such that the protection system (1) automatically initiates the operation of appropriate systems, including the reactivity control system, to ensure that specified acceptable fuel design limits are not exceeded as e result of anticipated operational occurrences and (2) senses accioent conditione and initiates the operation of systems and canponents important to safety.

Stone 6 Webster suppliet inputs to the reactor protection system automatically initiate a protective function in the event a parameter exceeds a Irmit dictated by the USSS Vendor .

Stone & Webster supplied engineered saf ety f eatures actuation system inputs sense the onset of an accident condition and initiate the operation of appropriate systems that support emergency core cooling and protect containment structure integrity.

The protection system loalc is within the NSSS Vendor's scope and is discussed in Section 3.1 or the USSS Vendor's SAR.

3.1-12 Amendment 8 3/28/75 bb 3$2

SWESSAR-P1 Reference Section Title No.

Instrumentation and Control 7 3.1.21 Protection System Reliability and Testability (Criterion 21)

The protection system shall be designed for high functional reliability and inservice testability 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) remaval from service of any component or channel does not result in loss of the required minimum redundancy unless the acceptable reliability of operation of the protection system can be otherwise demonstrated. The protection system shall be designed to permit periodic testing of its functioning when the rea ctor is in operation, including a capability to test channels independently to determine failures and losses of redundancy that may have occurred.

Discussion The Stone & Webster supplied protection system inputs and actuated devices which interface with the USSS logic devices are designed for high functional reliability and inservice testability commensurate with the safety function to be performed. The design meets IEEE 279-1971, " Criteria for Protection Systems for Duclear Power Generating Stations."

Component redundancy and physical and electrical isolation satisf y the single failure criterion with respect to channel independence.

Periodic testing of protection system input sensors, logic trains, and protection system actuated devices will determine f ailures and losses of redundancy that may have occurred.

The reliability and testing of the protection system logic are 2

within the NSSS Vendor's scope and are discussed in Section 3.1 of the USSS Vendor's SAR. Section 7.2 discusses compliance with appropriate codes and standards.

Reference Section Title No.

Instrumentation and Control 7 3.1.22 Protection System Independence (Criterion 22}

The protection system shall be designed to assure that the 9~ effects of natural phenomena, and of normal maintenance, testing, and postulated accident conditions on operating, 3.1-13 ,, ,,_

Amendment 2 bOl ,;j 8/30/74

SWESSAR-P1 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.

Discussion Those portions of the protection system supplied by Stone &

Webster are designed and arranged such that the environment accompanying any emergency condition in which the components are required to function does not result in the loss of the intended safety function.

The possibility of loss of the protection function through improper testing or maintenance is minimized by the usa of administrative controls such as testing procedures and maintenance record requirements.

Safety related sensors and actuated devices are redundant and physically separated. Physical separation of channel cable tray, conduit, and penetrations ensures independence of electrical cable.

Safety related devices will be tested and qualified to demonstrate their operation under normal and accident environmental conditions.

Independence of the protection system logic is within the NSSS Vendor's scope and is discussed in Section 3.1 of the NSSS Vendor's SAR.

Reference Section Title No.

Instrumentation and Control 7 3.1.23 Protection System Failure Modes (Criterion 23)

The protection system shall be designed to fall 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 adverso environments (e .g . , extreme heat or cold, fire, pressure, steam, water, and radiation) are experienced.

Dis cus sion Stone & Webster supplied protection system input sensors and actuated devices are designed with due consideration of the most probable failure modes of the components under various perturbations of energy sources and the environment.

3.1-1tl iu1 ji4

SWESSAR-P1 Fail safe design is incorporated in systems where it can be shown that the failure of a component or system does not generate a

, plant condition which would require additional plant protective functions.

In the event a component or systen. fails to another defined state (non-fail safe) , plant protection is provided by redundancy, independence, and physical separation. Redundant and independent power sources supply energy to the protection system.

Failure modes of the protection system logic are within the NSSS Vendor's scope and are discussed in Section 3.1 of the NSSS Vendor 's SAR.

Reference Sections Title No.

Engineered Saf ety Features 6 Instrumentation and Control 7 Electric Power 8 3.1.24 Separation of Protection and control Systens (Criterion 24)

Tae protection system shall be separated from :,ontrol systems to the extent that failure of any single control ystem component or channel, or failure or removal from serv ice of any single protection system component or channel which is commun co the control and protection systems leaves intact a system satisfying all reliability, redundancy, and independence requirements of tl:e protection system. Interconnection of the protection and control systems shall be limited so as to assure that safety is not significantly impaired.

Discussion Stone & Webster supplied control devices which derive a signal f rom protection system inputs are isolated such that a failure of a control device does not preve.,t the protection system from performing its safety function. This isolation may be accomplished through the use of devices including relays, optical devices, or isolation amplifiers. Failure or removal from service of the control or isolation device leaves intact a system satisfying the requirements of the protection system.

Separation of the protection system logic is within the NSSS Vendor's scope and is discussed in Section 3.1 of the NSSS Vendor's SAR.

661 3 a.

3.1-15

SWESSAR-P1 Reference Section Title No.

Instrumentation and Control 7 3.1.25 Protection System Requirements for Reactivity Control Malfunctions (Criterion 25)

The protection system shall be designed to assure that specified acceptable fuel design limits are not exceeded for any single malfunction of the reactivity contro t systems, such as accidental withdrawal (not ejection or dropout) of control rods.

Dis cussion Thia criterion is within the NSSS Vendor's scope and is discussed in 3ection 3.1 of the NSSS Vendor's SAR.

3.1.26 Reactivity Control System Redundancy and Capability (Criterion 26)

Two indepandent reactivity control systems of different design principles shall be provided. One of the systems shall use control rrids, 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 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 suberitical w br cold conditions.

Discussion This criterion is within the NSSS Vendor's scope and is discussed in Section 3.1 of the NSSS Vendor's SAR.

3.1.27 Combined Reactivity Control Systems Capability (Criterion 27)

The reactivity control systems shall be designed to have 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 appropriate margin for stuck rods the capability to cool the core is maintained.

3.1-16 i , ,,

l

SWESSAR-P1 Discussion h This criterion is within the NSSS Vendor's scope and is discussed la S9ction 3.1 of the USSS Vendor's SAR.

3.1.28 Reactivity Limits (Criterion 28)

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 shall include 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.

Discussion This criterion is within the NSSS Vendor's scope and is discussed in Section 3.1 of the NSSS Vendor's SAR.

3.1.29 Prot ection Against Anticipated Operational Occurrences (Criterion 29)

The protection and reactivity control systems shall be designed to assure an extremely high probability of accomplishing their safety f unctions in the event of anticipated operational occurrences.

Discussion This criterion is within the USSS Vendor's scope and is discussed in Section 3.1 of the NSSS Vendor's SAR.

3.1.30 Quality of Reactor Coolant Pressure Boundary (Criterion 30)

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 practical, identifying the location of the source of reactor coolant leakage.

Discussion This criterion except for detection is within the NSSS Vendor's scope and is discussed in Section 3.1 of the NSSS Vendor's SAR.

Stone & Webster provides a reactor coolant pressure boundary

(:RCPB) leakage detection system.

3.1-17 b6j 7*7 Jl/

SWESSAR-P1 Reference Section Title NO.

Reactor Coolant Pressure Boundary 5.2.7 Leakage Detection System 3.1.31 Fracture Prevention of Reactor Coolant Pressure Boundary (Criterion _31 The reactor coolant pressure boundary shall be designed with suf ficient margin to assure that when stressed under operating, m'intenance, testing, and postulated accident conditions (1) the b andary behaves in a nonbrittle manner and (2) the probability of rapidly propagating fracture is minimized. The design shall reflect consideration of service temperature and other conditions of the boundary material under operating, maintenance, testing, and postulated accident conditions and the uncertainties in determining (1) material properties, (2) the effects of irradiation on material properties, (3) residual, steady-state and transient stresses, and (4) size of flaws.

Discussion This criterion is within the NSSS Vendor's scope and is discussed in Section 3.1 of the USSS Vendor's SAR.

3.1.32 Inspection of Reactor Coolant Pressure Boundary (Criterion 32)

Components which are part of the reactor coolant pressure boundary shall be designed to permit (1) periodic inspection and testing of hnportant areas and features to assess their structural and leaktight integrity, and (2) an appropriate material surveillance program for the reactor pressure vessel.

Discussion This criterion is within the NSSS Vendor 's scope and is discussed in Section 3.1 of the NSSS Vendor's SAR.

3.1.33 Reactor Coolant Makeup (Criterion 33)

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 acceptable fuel design limits are not exceeded as a result of reactor coolant loss due to leakage fron the eactor coolant pressure boundary and rupture of small pipidi or other small components which are part of the boundary. The systems shall be designed 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) the system safety function can be accomplished 3.1-18 e 4 71 o 00l JlO

SWESSAR-P1 using the piping, pumps, and valves used to maintain coolant inventory during normal reactor operation.

Discussion This criterion is within the NSSS Vendor's scope and is discussed in Section 3.1 of the USSS Vendor's SAP.

3.1.34 Residual Heat kemoval (Criterion 34)

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 electric power syston operation (assuming of f site power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single f ailure.

Discussion The main steam system removes residual heat from the reactor coolant system following a suddon load rejection or a turbine trip by typassing steam to the condenser throuch the turbine bypass system or to the atmosphere through the main steam a bnospheric dump valves or the main steam safety valves. The auxiliary f eedwater system provides f eedwater for residual heat removal when the main feedwater pumps are not available. The residual heat removal system, which continues the cooldown after the above systems have sufficiently cooled the reactor coolant 5 system, is within the NSSS Vendor's scope and is discussed in Section 3.1 of the NSSS Vendor's SAR.

Ref erence Sections Title No.

Main Steam System 10.3 Turbine Bypass System 10.4.4 Auxiliary Feedwater System 10.4.10 3.1.35 Emeroency Core Coolina (Criterion 35)

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 da' mage that eculd interfere with

, , -t

' r 3.1-19 ggj ' ; / Amendment 5 12/2/74

SWESSAR-P1 continued effective core cooling is prevented and (2) clad -

metal-watar 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 electric power system operation (assuming of fsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.

Discussion This criterion is within the NSSS Vendor's scopa and is discussed in Section 3.1 of the NSSS Vendor's SAR.

3.1.36 Inspection of Emergency Core Cooling System (Criterion 36)

The emergency core cooling system shall be designed to permit appropriate periodic inspection of important components, such as soray rings in the reactor pressure vessel, water in jection nozzles, and piping, to assure the integrity and capability of the system.

Discussion This criterion is within the NSSS Vendor's scope and is discussed in Section 3.1 of the NSSS '.*endor's SAR.

3.1.37 Testing of Emergency Core Cooling System (Criterion 37)

The emergency core cooling system shall be designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leaktight 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 applicable portions of the protection system, the transfer between normal and emergency power sources, and the operation of the associated cooling water system.

Discussion This criterion is within the NSSS Vendor's scope and is diccussed in Section 3.1 of the NSSS Vendor's SAR.

s ' ;) . '

g 3.1-20 Amendment 5 12/2/7'i

SWESSAR-P1 3.1.38 Containment Heat Removal (Criterion 38)

A system to remove heat from the reactor containment shall be provided. The system safety f unction shall be to reduce rapidly, consistent with the functioning of other associated systems, the containment pressure and temperature following any loss-of-coolant accident and maintain them at acceptably low levels.

Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that for onsite electric power system operation (assuming of fsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplj shed, assuming a single f ailure.

3.1-20A ODl f) c j Amen dment 5 12/2/74

SWESSAR-P1 Discussion Heat is removed from the containment structure following a loss-of-coolant accident (LOCA) by (1) the containment atmosphere recirculation system, (2) the containment spray system, and (3) the residual heat removal (RHR) heat exchangers.

The containment atmosphere recirculation system together with the RHR heat exchangers and the containmeat spray system are designed to reduce the containment temperature and pressure to acceptably low levels and to maintain these low levels even if a single failure is assumed.

Reference Sections Title No.

Containment Heat Removal Systems 6.2.2 Containment Atmosphere Recirculation 9.4.5.1 System 3.1.39 Inspection of Containment Heat Removal System (Criterion 39)

The containment heat removal system shall be designed to permit appropriate periodic inspection of bnportant components, such as the torus, sumps, spray nozzles, and piping to assure the integrity and capability of the system.

Discussion The containment heat removal systems are designed to pemmit appropriate periodic inspection of important components.

Rpference Sections Title No.

Containment Heat Removal Systems 6.2.2 Containment Atmosphere hacirculatier System 9.4.5.1 3.1.40 Testing of Containnant Heat Removal System (Criterion 40)

The containment heat removal system shall be designeZ ta permit appropriate periodic pressure and functional testing to assure (1) the structural and leaktight 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 9 system into operation, including operation of applicable portions 3.1-21

((} }22

SWESSAR-P1 of the protection system, the transfer between normal and emergency power sources, and the operation of the associated cooling water system.

Discussion The containment heat removal systems are designed to permit periodic pressure and functional testing.

Reference Sections Title No.

Containment Heat Removal Systems 6.2.2 Containment-Atmosphere Recirculation System 9.4.5.1 3.1.u1 Containment Atmosphere Cleanup (Criterion 41)

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 quality 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 powe> system cperation (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, assuming a single failure.

Discussion The containment spray system reduces fission product concentration in the containment atmosphere following a DBA by '

spraying borated water containing NaOH (sodium hydroxide) into the containment atmosphere. The combustible gas control system maintains the hydrogen concentration in the containment atmosphere at a safe level following a DBA. If the combustible gas control dilution air subsystem is used to reduce the hydrogen concentration in the containment atmosphere, the discharge air to the atmosphere is sent through the radioactive gaseous waste system to reduce the concentration of fission products released to the environment.

9 3.1-22 bb )

SWESSAR-P1 Reference Sections Title No.

Containment Heat Removal Systems 6.2.2 Combustible Gas Control System 6.2.5 3.1.42 Inspection of Containment Atmosphere Cleanup Systems (Criterion 42)

The containment atmosphere cleanup systems shall be designed to permit appropriate periodic inspection of important components, such as filter frames, ducts, and piping to assure the integrity and capability of the systems.

Discussion The combustible gas control and containment spray systems are designed to permit appropriate periodic inspection of the important components.

Reference Sections Title No.

Containment Heat Removal Systems 6.2.2 Combustible Gas Control System 6.2.5 3.1.43 Testing of Containment Atmosphere Cleanup Systems (Criterion 43)

The containment atmosphere cleanup systems shall be designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leaktight integrity of its components, (2) the operability and performance of the active components of the systems such as fans, filters, dampers, pumps, and 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.

Djscussion The combustible gas control and containment spray systems are designed to permit periodic pressure and functional testing of their components.

3.1-23 7g

, , , au -

SWESSAR-P1 Reference Sections, Title No.

Containment Heat Removal Systems 6.2.2 Combustible Gas Control System 6.2.5 3.1.44 Cooling Water (Criterion 44)

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 component s under normal operating and accident conditions.

Suitable redundancy in components and features, and suitable interconnections leak detection, and isolation capabilities shall be provided to assure that for onsite electric power system operation (assuming of fsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.

Discussion The reactor plant component cooling water system, an intermediate cooling system, transfers heat from systems containing reactor coolant to the reactor plant service water system. Together these systems transfer heat to +he ultimate heat sink from structures, systems, and components important to safety during normal and accident conditions.

Both systems are designed with suitable redundancy in components, with leak protection, and with the capability to isolate redundant components. Both systems are designed to satisfy the cooling water requirements assuming both a loss of onsite or offsite power and a single failure.

Reference Sections Title No.

A-C Power Supply System 8.3.1 Reactor Plant Service Water System 9.2.1 Reactor Plant Component Cooling Water System 9.2.2 Ultimate Heat Sink 9.2.5 O

3.1-24 7nr

{gj JLJ

SWESSAR-P1 3.1.45 Inspection of Cooling Water System (Criterion 45)

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.

Discussion The reactor plant service water system and the reactor plant component cooling water system are designed to permit appropriate periodic inspection to ensure the integrity of the components and the systems as a whole, Reference Sections Title No.

Reactor Plant Service Water System 9.2.1 Reactor Plant Component Cooling Water System 9.2.2 3.1.46 Testing of Cooling Water Systen (Criterion 46)

The cooling water system shall be designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leaktight 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 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.

Dis cussion The designs of the reactor plant service water system and the reactor plant component cooling water system permit, to the extent practicable, periodic pressure and functional testing.

Reference Sections Title No.

Reactor Plant Service Water System 9.2.1 Reactor Plant Component Cooling Water System 9.2.2 3.1.47 This criterion has not been promulgated by the AEC.

((\

3.1-25

SWESSAR-P1 3.1.48 This criterion has not been promulgated by the AEC.

3.1.49 This criterion has not been promulgated by the AEC.

3.1.50 Containment Design Basis (Criterion 50)

The reactor containment structure, including access openings, penetrations, ar.d 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 suf ficient margin, the calculated pressure and temperature conditions resulting from any los s-of-coolant accident. This margin shall effects of potential energy sourcesreflect which consideration have not been of (1) the 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 f rom degraded emergency core cooling functioning, (2) the limited experience and experimental data available for defining accident phenomena and containment responses, and (3) the conservatism of the calculational model and input parameters.

Discussion The containment structure, including personnel and equipment hatches, piping, and electrical penetrations, the containment spray system, and the containment atmosphere recirculation system are designed such that the containment structure leaks at a rate less than the design leakage rate of 0.2 percent by volume of its contents per day under post-DBA conditions. In addition, the containment structure is designed to withstand, by a sufficient

margin, a DBA.

those pressure and temperature conditions resulting from This margin reflects all potential energy sources not considered in the conservative calculation of peak conditions and also takes into consideration the limited experience and experimental data available for defining accident phenomena and containment responses.

Reference Sections Title No.

Concrete Containment Structure 3.8.1 Containment Functional Design 6.2.1 Containment Heat Removal Systems 6.2.2 Containment Atmosphere Recirculation System 9.4.5.1 O

3.1-26 4 q~

66I stl

SWESSAR-P1 3.1.51 Fracture Prevention of Containment Pressure Boundary (Criterion 51)

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.

Discussion Ferritic materials in the containment boundary will be toughness tested in accordance with the provisions of ASME Boiler and Pressure Vessel Code,Section III, Division 2 (ACI 359-Proposed Standard Code for Concrete Reactor Vessels and Containment, Nov. 8 1974) , with the following additions:

1) Notch toughness testing is required for materials with a nominal thickness of 5/8 in, and greater, rather than greater than 5/8 in.

2) Plates thinner than 5/8 in. are toughness tested by full or subsized Charpy V-notch impact tests. The acceptance criteria for full size specimens are the same as for materials with nominal thickness of 5/8 in. and greater.

The acceptance criteria for subsized specimens are as shown in Table 16 of SA-20, ASME Boiler and Pressure Vessel Code,Section II - P a rt A .

3) The test temperature is 60 F below either the lowest service metal temperature or the pneumatic test metal temperature, whichever is lower.

Weld procedure qualification serves to demonstrate that the toughness of the weld metal and heat affected zones follow the same criteria as the base metal.

3.1.52 Capability for Containment Leakage Rate Testina

{ Criterion 52)

The reactor containment and other equipment which may be subjected to containment test conditions shall be designed so that periodic integrated leakage rate testing can be conducted at containment design pressure. ,

,qg Db OLO 3.1-27 Amendment 8 3/28/75

SWESSAR-P1 Discussion The containment structure and related equipment, which will be O subjected to the containment structure test conditions, are designed so that the periodic integrated leakage rate testing can be conducted in accordance with Appendix J of 10CFR50.

Reference Section Title No.

Containment Punctional Design 6.2.1 3.1.53 Provisions for Containment Testing and Inspection (Criterion 53)

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 leaktightness of penetrations which have resilient seals and expansion bellows.

Discussion The containment structure design includes provisions for testing the leaktightness of all penetrations. g Leak test channels are placed over liner seam welds and penetration to liner welds to permit testing the leaktightness of the welds during construction. These channels are capable of being periodically pressurized to test the leaktightness of the welds.

Electrical penetrations will be periodically tested for leaktightness following installation in the containment structural wall.

Reference Sections Title No.

Containment Functional Design 6.2.1 Containment Leakage Monitoring System 6.2.6 3.1.54 Systems Penetrating Containment (Criterion 54)

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 3.1-28 g4 7 ': G

[(i v } JL /

SWESSAR-P1 isolation valves and associated apparatus and to determine if valve leakage is within acceptable limits.

Discussion Piping systems th,tt penetrate the containment structure do so through penetrations that are designed to minimize leakage.

Containment isolation valves provide the capability to seal most penetrations redundantly; those few exceptions are described in detail in Section 6.2.4. Pressure taps provide the capability to test for leakage those containment isolation valves required to be tested by a Type C isolation valve test.

Reference Section Title No.

Containment Isolation System 6.2.4 3.1.55 Reactor Coolant Pressure Boundary Penetrating Containment (Criterion SSJ, Each line that is part of the reactor coolant pressure boundary and that penetrates primary reactor containment shall be provided with containment isolation valves as follows, unless it can be demonstrated that the containment isolation provis".ons for a specific class of lines, such as instrument lines, are acceptable on some other defined basis:

(1) One locked closed isolation valve inside and one locked closed isolation valve outside containment; or (2) One automatic isolation valve inside and one locked closed isolation valve outside containment; or (3) 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 (4) 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 cutside 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 these lines or of lines connected to them chall be provided as necessary to assure adequate safety. Determination of the appropriateness of these bbl JJ 3.1-29

SWESSAR-P1 requirements such as hijher quality in design, fabrication, and testing, additional provisions for inservice inspection, protection against more severe natural phenomena, and additional isolation valves and containment, shall include consideration of the population density, use characteristics, and physical characteristics of the site environs.

Discussion Pipes that are part of the reactor coolant pressure boundary and penetrate the containment structure are provided with containment isolation valves in accordance with the above criterion. Valves outside the containment structure are located as close as practical to that structure; valves on the incide are missile protected. The isolation valves are subject to periodic Type C tests and, upon loss of power, take the position that provides the greatest safety.

Reference Sections Title No.

Containment Functional Design 6.2.1 Containment Isolation System 6.2.4 3.1.56 Primary Containment Isolation (Criterion 56) ,

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 instrument lines, are acceptable on some other defined basis:

(1) One locked closed isolation valve inside and one locked closed isolation valve outside containment; or (2) One automatic isolation valve inside and one locked closed isolation valve outside containment; or (3) 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 (4) 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 containment shall be located as close to the containment as practical and upon loss of actuating power, 9

3.1-30 ti 8 7 y ,I 00i JJ

SNESSAR-P1 automatic isolation valves shall be designed to take the position that provides greater safety.

Discussion Pipes that connect to the containment atmosphere are provided with containment isolation valves in accordance with the above criterion with the following exception: suction piping connectina the engineered safety features (ESF) sumps to the containment spray and high and low pressure safety injection pumps are embedded in concrete over their entire length and thus have no valves inside the containment structure. The isolation valves in this piping, which are outside the containment structure, are located as close as practical to that structure. The piping from the ESF sump to the containment isolation valves is designed as l8 described in Section 3.6.2.2.1. Instrument piping penetrations, otuer than those for the containment pressure monitorina system, are in accordance with Regulatory Guide 1.11 and have remote manual isolation valves outside the containment stru cture . See Section 7.3.3.9 for a discussion of the containment pressure monitoring system penetrations.

Reference Section Title No.

Con tainment Isolation System 6.2.4 3.1.57 Closed System Isolation Valves (Criterion 57)

Each line that penetrates primary reactor containment and is neither part of the reactor coolant pressure boundary nor connected directly to the containment atmosphere shall have at least one containment isolation valve which shall be either automatic, or locked closed, or capable of remote manual operation. This valve shall be outside containment and located as close to the containment as practical. A simple check valve may not be used as the automatic isolation valve.

Discussion Pipes connected to closed systems are provided with at least one containment isolation valve located as close as possible to the outside of the containment.

Reference Section Title No.

Containment Isolation System 6.2.4 3.1.58 This criterion has not been promulgat ed by the NRC. B 3.1-31 ,

Smendment 8 hb! JJL 3/28/75

SWESSAR-P1 3.1.59 This criterion has not been promulgated by the NRC. @

3.1.60 Control of Releases of Radioactive Materials to the Environment (Criterion 60)

The nuclear power unit design shall include means to control suitably the release of radioactive m2.terials in gaseous and liquid effluents and to handle radioactive solid wastes produced during normal reactor operation, including anticipated operational occurrences. Sufficient holdup capacity shall be provided for retention of gaseous and liquid effluents containing radioactive materials, particularly where unfavorable site environmental conditions can be expected to impose unusual operational limitations upon the release of such ef fluents to the environment.

Discussion In all cases, the design for radioactivity control is on the basis of the requirements of 10CFR20 and 10CFR50 for normal operations and for any transient situation that might reasonably be anticipated to occur, and (2) on the basis of 10CrR100 dosage level guidelines for potenzial accidents of exceedingly low probability of occurrence.

The activity level of waste gas effluents is substantially reduced by differential holdup of noble gases in charcoal decay beds and subsequent release through the ventilation vent. Under conditions of concurrent fuel failure and steam generator tube leakage, some radioa ctive gas is present in the steam jet air ejector discharge. The steam jet air ejector discharge is directed to the radioactive gaseous waste system and is discharged from the ventilation vent. Control of liquid waste effluents is maintained by batch processing of all liq ids, sampling before discharge, and controlled rate of release.

Liquid effluents are monitored for radioactivity and rate of flow. Radioactive liquid waste system tankage and evaporator capacity are sufficient to handle any expected transient in the processing of liquid waste volume.

Solid wastes are prepared for offsite disposal by approved procedures. Solid wastes are prepared for shipment by placement in shielded and reinforced containers that - .et applicable AEC and Department of Transportation requirements.

O 3.1-32 r- 7,-

(10 l JJ)

SWESSAR-P1 Reference Sections Title No.

Radioactive Liquid Waste System 11.2 Radioactive Gaseous Waste System 11.3 Process and Effluent Radiation Monitoring 11.4 Systems Radioactive Solid Warte System 11.5 Effluent Release Limits 16.3.8 3.1.61 Fuel Storage and Handling and Radioactivity Control (Crit ( rion 61)

The fuel storage and handling, radioactive waste, and other systems which may contain radioactivity shall be designed to assure adequate safety under normal and postulated accident conditions. These systems shall be designed (1) with a capability to permit appropriate periodic inspection and testing of components 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 testability that reflects the importance to safety of decay heat and other residual heat removal, and (5) to prevent significant reduction in fuel storage coolant under accident conditionc.

Discussion Safety related components of the radioactive waste and fuel storage systems are designed to allow appronriate periodic inspection and testing.

The fuel pool design meets the requirem<r.its of 10CFR20 in providing radiation shielding for operating rarsonnel during fuel transfer and during storage of spent fuel. The fuel transfer canal wall thickness is sufficient to shield work areas adjacent to the canal for personnel access during actual fuel transfer.

Waste storage and processing facilities in the annulus building and the solid waste and decontamination building have shielding to meet the requirements of 10CFR20 for operating personnel.

Periodic surveys by health physics personnel ensure that radiation design levels are not exceeded during plant operating lifetime.

Spent fuel handling systems are designed to preclude gross mechanical failures that could lead to significant radioactivity releases. Floor trench drain systems collect leakage that might 3.1-33 .

bbk

SWESSAR-P1 occur from valve stem leak-offs, heat exchanger and pump drains, and transfer the leakage to the radioactive liquid waste system.

Normally, radioactive gases and particulates released from components are colleuted by the aerated portion of the reactor plant vents and drains system. Uncontrolled laakage of radioactive gases and particulates which may leak from spent fuel or from components containing radioactive fluids is collected.

Discharges from these systems are monitored.

Decay heat from spent fuel is transferred to the fuel pool water and subsequently removed by the cooling portion of the fuel pool cooling and purification system. Redundancy of fuel pool cooling components is provided to increase reliability of heat removal from the r'uel pool.

Design of the fuel pool ensures that there is no significant loss of fuel pool cooling capability under accident conditions.

The piping connected to the fuel pool is designed so that no significant loss of fuel pool water can occur in the event of a pipe rupture. Instrumentation is provided to indicate any significant reduction in fuel pool water level. Redundant sources of fuel pool makeup ensure availability of makeup to the fuel pool even under loss of normal power.

Reference Sections Title No.

Design of Category I Structures 3.8 Puel Pool Cooling and Purification System 9.1.3 Reactor Plant Vents and Drains Systems 9.3.3 Radioactive Waste Management 11 Radiation Protection 12 3.1.62 Prevention of Criticality in Fuel Storage and Handling (Criterion 62)

Criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations.

Discussion Criticality is prevented in the new fuel storage area by physical separation. Crit: 2ality is prevented in the fuel pool by the physical separation of fuel assemblies and by the presence of borated water in the pool.

3.1-34 -, ;

h\ 'E

SWESSAR-P1 Reference Sections Title No.

New Fuel Storage 9.1.1 Spent Fuel Storage 9.1.2 3.1.63 Monitorina Fuel and Waste Storage (Criterion 63)

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.

Discussion The fuel pool water temperature is continuously monitored. The temperature is displayed in the control room where audible and visual alarms annunciate should the water temperature increase above a preset level. In the event of a high temperature alarm, administrative procedures provide for checking the cooling water flow to the fuel pool coolers, the operating status of the fuel pool cooling pumps and the integrity of the fuel pool cooling and purification system, and ensuring that corrective measures are taken to restore fuel pool cooling.

The radiation level above the fuel pool is continually monitored by a radiation detector mounted on the fuel pool bridge.

Continuous surveillance of radiation levels in the solid waste and decontamination building is maintained by appropriately located area radiation detectors. Radiation levels in excess of preset levels for both the fuel building and solid waste and decontamination building initiate audible and visual alarms both locally and in the control room.

In the event of high airborne gaseous or particulate activity, the fuel building ventilation system exhaust is diverted through the supplementary leal collection and release system.

Reference Sections Title No.

Supplementary Leak Collection and Release 6.2.3.1 System Puel Pool Cooling and Purification System 9.1.3 Fuel Building Ventilation 9.4.6 Padiation Protection 12 3.1-35

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SNESSAR-P1 3.1.64 Monitorina Radioactivity Releases (Criterion 64) ll Means shall be provided for monitoring the reactor containment a tmosphere , spaces containing components for recirculation of los s-of-coolant accident fluids, e.ffluent discharge paths, and the plant environs for radioa ctivity that may be released fron nornal operations, including anticipated operational occurrences, and f rom postulated accidents .

Dis cussion The containment atmosphere is monitored continually during normal and transient plant operations using the conta inment atmosphere rad iation monitoring syst em. The annulus building which completely surrounds the containment structure contains a series of airborne radiation monitors to detect and register radiation in its various levels . In addition, the reactor plant component cooling water system is monitored to ensure detection of any leakage of radioactive fluids from primary components into the rea ctor plant component cooling water system. Radioactivity levels in the normal plant effluent discharge paths and in the environs are continually monitored during normal and accident conditions by the various radiation monitoring systems and by the of f site radiological monitoring program.

Reference Sections Title No.

Process and Ef fluent Radiation ?!onitoring 11.4 Systems Of f site Padiological Monitoring System 11.6 Al 'orne Radiation Monitoring 12.2.4

? .65 Single Failure Criterion a

3.1.65.1 Definitions Single Failure Single failure is defined in Appendix A to 10CFR50 as publishea 8 in the Federal Register on February 20, 1971, and as arended on July 7, 1971, as follows:

"A single failure means an occurrence which results in the loss of capability of a component to perform its intended safety

'tnctions. Multiple f ailures resulting from a single occurrence are considered to be a sin gle failure. Fluid and electric sys tems are considered to be designed against an assured single failure if neither (1) a single failure of any active component (a s suming passive components function properly) nor (2) a single 3.1-36 Amendment 8 3

, . 7 ,, , /2 8/7 5

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SWESSAR-P1

, failure of a passive component (assuming active components 0RM f unction properly) results in a loss of the capability of the system to perform its safety functions."

Single failures of passive components including spurious signals in electric systems are assumed in designing against a single failure.

Active Component Failure An active component is one in which mechanical movement must occur in order to complete the component's intended function. An active component failure is its failure to complete its intended function upon demand.

Examples of active component failures include the failure of a powered or check valve to move to its correct position, the f ailure of an electrical breaker or relay to respond, the failure of a pump, fan, or emergency generator to start, and mechanical motion of a passive component of a system due to an electrical malfunction. ,,

Passive Component Failure A passive component is one in which mechanical movement does not occur in order for the component to perform its intended f unction. A passive componer.t failure is the structural failure of a passive component so that it does not perform its intended function. For fluid pressure boundary components, a passive component failure results in a crack in the pressure boundary. A valve which is not required to operate to perform a safety related function is considered a passive component.

Other passive components include piping, cables, valve bodies, etc.

Short Term The short term is defined as the first 24 hr following the start of an incident.

Long Term The long term is defined as the period following the short term during which the system safety function is still required.

3.1.65.2 Application of a Single Failure Q2finition

1. With respect to fluid systems , the single failure definition applies only to safety related fluid systems or portion s thereof. These fluid systems are designed to ensure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power operation (assuming onsite power is not available) , they are able to accomplish 3.1-37 Amendmer.t 11

,,. ;9 00l J J I; 5/30/75

SWESSAR-P1 their safety f unctions assuming that a sing le failure occurs within the total system.

The systems to which the single failure definition applies, as listed and defined in Appendix A to 10CFR50, are (with the appropriate General Design Criterion (GDC) in parentheses) :

Residual Heat Pemoval (NSSS Vendor) (GDC-34)

Emergency Core Cooling (NSSS Vendor) (GDC-35 thru 37)

Containment Heat Removal (GDC-38 thru 40)

Containment Atmosphere Cleanup (GDC-41 thru 43)

Cooling Water (GDC-44 thru 46)

Containment Isolation (GDC-54 thru 57)

Systems other than those listed above may also ne sub-ject to the single failure definition if they are required to perform a safety function (i .e . , mitigate the effects of the incident being analyzed) . In addition, the onsite electrical power sources (both ac and de) and their associated distribution systems are designed to have sufficient independence, redundancy, and test ability to perform their safety functions assuming a single f ailure, as defined in GDC-17. The protection systems, as defined in GDC-20 with regard to 8

failure is designed in accordance with Regulatory Guide 1.53, as discussed in Section 3A.1-1.53. High energy pipe ruptures and crack-initiating incidents are discussed in Section 3.6.

2. Each of the above safety related systems or combination of systems is designed to accept only a single active component failure in the sho rt term and to accept a single active compcnent failure or a single passive component failure in the long term following an accident condition. A passive component failure in the short term is not considered.

3. For the purposes of the single failure definition, the system subject to it includes the safety related system itself and its related service systems (such as component cooling water, service rater, electric power, etc) required for the safety related system to perform its safety function. Even though each system indicated above is designed to the single failure definition, only one single failure and its consequences are assumed to take place in tha aggregate of the safety related systems and related service systems in the unit.

Non-safety related systems are designed so that their failure will not cause safety related systems to lose their safety fimetion.

3.1-38 Amendment 8

, 7 ,p/28/75 0oi JJ/

SWESSAR-P1

4. Where the proper active function of a check valve can be demonstrated despite any reasanable postulated condition (e .g . , a jmumed hinge pin) , the check valve is considered to perf orm its initially intended function free of active component failure.

5. The passive component failure assumed in the failure analysis of a Safety Class 2 or 3 system is discussed in Section 3.6. In designing for a passive component failure, very low stressed components , such as the ductwork of a ventilation system, are not considered to f ail . Also, a valve body is not considered to fail in such a manner that it would result in two open pipes.

6. For design and for of fsite dose calculations, the mass of fluid discharged through the crack prior to effective isolation is considered. Duration of the leak is conservatively determined consistent with the means of 8 leakage detection, location of the leak, and the means of leak isolation. As a guide, a nominal 30 minute leakage duration is considered conservative for manual operator action.

7. The unit design is such that all active components of the designated safety related systems and related service systems can be proved operational by scheduled periodic operational tests or operational status indications.

8. All safetv rela ted electrical control systems are designed to ensure that a single failure (as defined above) cannot cause mechanical motion of a passive component of a fluid system in such a way that its motion results in total loss of the system safety function. Mechanical fluid systems which include passive mechanical components are provided with sufficient redundancy to meet this requirement.

661 340 3.1-39 Amendment 8 3/28/75

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SWESSAR-P1 3.2 CLASSIFICATION OF STRUCTURES, SYSTEMS, AND COMPONENTS 3.2.1 Seismic Classification Seismic Category I structures, systems, and components are those designed to withstand the safe shutdown earthquake (SSE) and operating basis earthquake (OBE) as defined by Appendix A to 10CFR100 dated November 13, 1973 and Section 2.5. The seismic design of Seismic Category I systems and components is described in Sections 3.7 and 3.10 and of structures in Section 3.8.

Seismic Category I structures, systems, and components as defined by Regulatory Guide 1.29 (Section 3A.1-1.29) and ANSI Stand ard N18.2(1J are listed in Table 3.2.5-1.

3.2.2 System Quality Group Cls.ssification (ANSI System Saf ety Classification)

The classification system promulgated by ANSI Standard N18.2, as modified by the proposed revisions dated January 1973 by the ANS Standards Sub-Committee 21, and Regulatory Guide 1.26 (Section 3 A.1 - 1. 2 6 ) is used for classifying the containment structure and safety related fluid systems.

The containment structure and fluid systems are classified according to the classes given below. Sepports are in the same safety class as the supported components if failure of the support could cause a loss of a safety function associated with the supported component.

Table 3.2.5-1 lists the fluid systems that are included in Safety Classes 1, 2, and 3. In addition, the safety class boundaries are shown on the various flow diagrams throughout this SAR.

Safety Class 1 (SC-1)

Safety Class 1 applies to reactor coolant pressure boundary com-ponents whose failure during normal reactor operation would prevent an orderly reactor shutdown and cooldown assuming makeup is provided by normal makeup systems only.

Shutdown and cooldown in an orderly manner means the pressurizer level is above the minimum level indication (other than during an initial transient), normal makeup is provided, the fuel limits and the conditions of the reactor coolant pressure boundary are within technical specification limits and the emergency core cooling system (ECCS) is not automatically actuated. Normal makeup is makeup in which only the system (s) used to maintain reactor coolant inventory during normal operation is (are) operating and in which only the number of pumps normally used are operating.

)

3.2-1

SWESSAR-P1 structure. Internal wind pressure oceft cients are based on ANSI A58.1-1972 for all Seismic Category I structures. The external wind pressure coefficients used for rectangular shaped structures are based on ANSI A58.1-1972 and are shown in Table 3.3.1-2. The external wind pressure coefficients used for cylindrical shaped structures are based on Table 4 (f) ASCE Paper No. 3269 (Ref. 2) and are shown in Table 3.3.1-3. The external wind pressure coefficients used for the containment dome are based on " Wind Stresses in Domes" (Ref. 3) and are shown in Table 3.3.1-4. Use of local pressure coefficients is in accordance with ANSI A58.1-1972.

3. The resultant design wind pressures pi) applied to Seismic Category I structures are determined as:

W = qC p q Cp where:

q=q or y for enclosed structures (see ANSI A58.1-1972). A negative value for W indicates that the resultant pressure acts outward. The external and internal pressures are combined so as to yield the maximum stresses.

The resultant design wind pressures and distributions for Seismic Category I structures are shown in Table 3.3.1-5 for rectangular shaped structures, Table 3.3.1-6 for cylindrical shaped structures, and Table 3.3.1-7 for the containment dome.

A step function of pressure with height is used; the specified resultant design wind pressure at a given height is applied over a height zone defined by one-half the difference in adjacent heights for which the design wind pressures are specified. The resultant design wind pressure acts normal to the surface of the structure being considered.

I 3.3-2 Amendment 1 7/30/74 661 343

SWESSAR-P1

3. Portions of the component and process cooling systems that cool other safety systems, the control room, or safety related electrical components

()b l [e k k 3 . 2 -2 A Amendment 5 12/2/7:4

SWESSAR-P1

4. Components, the failure of which would result in release to the environment of radioactive gases required to be held for decay of the following:

am Auxiliary systems that are required for reactor coolant letdown and makeup not covered by Safety Class 2

b. Radioactive waste systems

5. Fuel pool cooling

6. Onsite emergency power supply support systems external to the emergency diesel generators (the emergency diesel generators are defined in IEEE-STD-387-1972)

7. Portions of the main steam system that supply steam to the turbine drive of the auxiliary feedwater pump

8. Containment atmosphere purification and cleanup systems used to clean up the containment atmosphere after its leakage from the containment structure and other air purification and cleanup systems used after accidents.

Nonnuclear Safety Class (NNS)

This class applies to portions of the plant not covered in Safety Classes 1, 2, or 3. This class includes n.ast of the steam and power conversion systems, radioactive liquid waste system, and boron recovery system.

Nonsafety related systems having components in areas where safety related systems are located will be designed to preclude damage to any safety related components. For example, nonsafety related piping that is in areas of safety related piping will be restrained with Seismic Category I restraints as listed in Table 3.2.5-1. Table 3.2.2-1 lists all components inside the containment structure which are not Seismic Category I but which 1 are designed not to fail and affect safety related equipment.

3.2.3 Ouality Assurance Categories The structures, systems, and components are classified either as QA Category I or not applicable (NA) . The requirements of Appendix B, 10CFR50, apply only to those items classified as QA Category I due to their relationship to public safety. The structures and equipnent which are classified NA are not safety related in the context of 10CFR50 or 10CFR100. However, such structures and equipment may require a quality assurance program incorporating one or more of the requirenents applicable to the safety related items in QA Category I.

7r:

661 N

' .2 _i Amendment 7 2/28/75

SWESSAR-P1 QA Category I is defined as follows:

Plant structures, systems, and components whose failure or malfunction could cause a release of radioactivity that would endanger public safety. This category also includes structures, systems, and components which are vital to a safe shutdown of the plant and the removal of decay and sensible heat, or equipment which is necessary to mitigate consequences to the public of a postulated accident.

9A Category I includes Seismic Category I and Safety Classes 1, 2, and 3; however, QA Category I does not imply Seismic Category I. QA Category I items which are not Seismic Category I are indicated in Table 3.2.5-1.

The term " safety related" used throughout this SAR is synonymous with QA Category I.

Plant structures, systems, and components classified QA Category I are listed in Table 3.2.5-1.

3.2.4 Other Classification Systems ASME Codo Cla:fs, AShh Code Classes 1, 2, 3, and MC are used in the ASME Boiler and Pressure Vessel Code,Section III, Nuclear Power Plant Components and are referren to in this SAR only as ASME III, Class 1, 2, 3 or MC. Components are purchased with the objective of full compliance with the requircsnents of A6ME III and its addenda in accordance with Title 10 at the Code of Federal Regulations, Part 50.55a. With regard to pumps, valves, piping, tanks, and pressure vessels, there is a direct one-for-one correlation between Code Classes 1, 2, and 3, anc ANS Safety Classes, SC-1, 2, and 3 (Section 3.2.2) . The Codes for the concrete portions of the containment structure, classiried as SC-2, are specified 4.n Section 3.8. Metal containment systems, such as the personnel access hatches, are classified as Code Class MC. Paragraphs NE-110 and Ia,-1140 of ASME III delineate the portions of the containment structure classified as Code Class MC.

The ASME coce classes of structures, systems, and components are given in Table 3.2.5-1.

IEEE Classification Systems 1EEE Std-308-1971(2) defines Class IE electric systems as't. hose systems that provide the electric power used to shut down the reactor and limit the release of radioactive material following a DBA (i.e., postulated events used in the design to establish the performance requirements of the structures and systems) . All Class IE systems are QA Category I.

3.2-4 Amendment 4 11/1/74

66) 340

SWESSAR-P1 3.2.5 Tabulation of Codes and Classifications Table 3.2.5-1 identifies all QA and Seismic Category I structures, systems, and canponents.

Table 3.2.5-1 also provides the following information for these structures, systens , and components: -

1. Safety classes

2. Codes and c. ode classes

3. Tornado design criteria

4. Flood design criteria References for Section 3.2

1. ANSI Standard N18.2, " Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants" 1973.

2. IEEE-308, "IEEE Criteria for Class IE Electric Systems for Nuclear Power Generating Stations," 1971.

3. IEEE-279, " Criteria for Protection Systers for Nuclear Power Generating Stations," 1971.

4. IEEE-336, " Installation, Inspection, and Testing Requirements for Instrumentation and Electric Equignent During the Construction of Nuclear Power Generating Stations," 1971.

5. IEEE-317, " Electrical Penetration Assemblies in Containment Strucuures for Nuclear Fueled Power Generating Sta tions ,"

1971.

6. ShACNA, " Duct Manual and Sheet Metal Construction - High Velocity System," 1969.

7. AMCA, All Air Moving and Conditioning dLandards.

8. ARI, American Ref rigeration Institute Standard 401.

9. NFPA, National Fire Protection Association Standard No. 37.

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' 't 3.2-5 Amendment 4 11/1/74

SWESSAR-P1 TABLE 3.2.2-1 NONSEISMIC CATEGORY I COMPONEITI'S IhSIDE THE CONTAINMENT STRUCTURE Piping and Ventilation Systems II containment leakage monitoring (Section 6.2.6)

Fuel pool purification (Section 9.1.3)

NNS portion of the reactor plant component cooling wattr (Section 9.2.2)

Demineralized water (Section 9.2.3)

Primary grade water (Section 9.2.7)

Chilled water (Section 9.2.8)

Instrument and service air (Section 9.3.1)

Containment instrument air (Section 9.3.1)

Aerated vent and drain (Section 9.3.3)

Gaseous vent and drain (Section 9.3.3)

Containment atmosphere recirculation (Section 9.4. 5.1) 17 Containment purge air (Section 9.4.5.2)

Containment atmosphere filtration (Section 9.4.5.3)

Control rod drive mechanism cooling (Section 9.4.8) 17 Water fire prot'ction (Section 9.5.1)

Reactor plant gas supply cystem (Section 9.5.8)

Emergency core cooling system test line* (Section 6.3)

Boron injection surge tank overflow line* (Section 6.5)

Components Located at Mat Level Lower Than and Remote from Related Components Gaseous vent and drain system (Section 9.3.3)

Reuctor coolunt drain tank 1 7

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SWESSAR-P1 TABLE 3.2.2-1 rONT)

Reactor coolant drain tank cooler Reactor coolant drain tank pump Aerated vent sad drain system (Section 9.3.3)

Containment sump pumps Components Separated from Safety Related Components 14 Concrete Walls or Floors Reactor plant component cooling water system (Section 9.2.2)

Reactor vessel support shield tank surge tank keactor vessel support shield tank cooler Containment atmosphere filtration system (Section 9.4.5.3)

Filter units Fans

• Component supplied by NSSS Vendor.

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TABLE 3.2.5-1 (CDNT)

ANS Safety Code Tornado Flood Class Code Class Incation criterion Criterion Notes SWUCTURES Containment Structure Reiniorced concrete Exterior 2 + NA CS D NA Reinforced Concrete Interior 2 + NA CS P NA Contaisunent Structure Liner 2 + NA CS P NA Piping and Duct N netrations 2 ASME III 2 CS P PF/FS Electrical Penetration / 2 IEEE-317 NA CS P PF/AG Assesnblies Persatnel Access Incks 2 ASME III MC CS P NA Equipnent !!atch 2 + NA CS P NA Cable Tunnel from Containment NA + NA CS/CB/AD D PF/AG Structure to Control Building Annulus Building Retnforced Concrete Exterior NA + NA AB D NA Reinforced Concrete Intcrior NA + NA AB P NA Main Steam and Feedwater NA + NA AB P/D PF/AG Valve Areas Engineered Safety Features . + NA AB P PF/FS Areas Fuel Pool NA + NA AB P NA New Fuel Storage Area NA v NA AB P NA 24 Reactor Plant Tank Area Reinforced Concrete Founda- NA + NA TA D NA C' tion Dike Solid Waste And Decontamination Building

Partial Reinforced Concrete NA + NA WB D NA Foundation

+See Section 3.8 for applicable codes Amendment 24 2 of 11 4/23/76

SWESSI TABLE 3.2.5-1 (CONT)

ANS Safety Code Tornado Flood Class Ccde _ Class Location Criterion Cr it erion Notes Control Buildinq NA + N4 CB D PF/AG Diesel Generator Building NA + NA DGB D PF/AG Diesel Generator Fuel Oil NA + NA DGB D PF/FS Pump House Reactor Vessel Support Shield Tank 1 ASME III 1 CS P ta Classified as SC-1 because (Section 5.5.14) supports reactor vessel SYSTEMS Containment Spray System (Section 6.2.2)

Refueling Water Storage Tank (RPST) 2 ASME III 2 AB P PF/FS Chemical Addition 2 ASME III 2 AB P PF/PS Tank ; CAT)

Containment Spray Pumps 2 ASME III 2 ESFA P PF/FS Piping and Valves, excluding RWST 2 ASME III 2 ESFA/CS P PFfFS and CAT Recirculation Lines and 2 Test Piping (NNS)

Instrumentation and Controls NA IEEE-279 NA AB P PF/FS Required to Perform a Safety IRR/ESR/

Function in QA Cat. I Portions of IEEE-336 CR System Supports for QA Cat. I (Note) NA NA (Note) P NA Same as component being Components supported Supplementary Leak Collection and Relvase System (Section 6.2.3.1)

& Supplementary Leak Collection and 3 SMACNA NA AB P PF/AG c' s Release System Dampers, Ductwork ARI, AMCA ANSI-N101

+See Section 3.6 for applicable codes t s)

L7 N

3 of 11 Amendment 2 8/30/74

SWESS.

TABLE 3.2.5-1 (CONT)

ANS Satety Cbde Tornado Flood Class _ Code Class Location Criterion Cr it erion fotes Containment Isolation System (Section 6.2.4)

Containment Isolation Valves and 2 ASME III 2 CS/AB P PF/PS See individual fluid system Associated Piping for Systems figures for delineation of Penetrating Containment Structure SC boundaries Instrumentation and Controls NA IEEE-279 NA CS/ P PF/FS Required to Perform a Safety AB/

Function in QA Cat. I Portions IEEE-336 CR/IRR of System Supports for QA Category I (Flote) NA NA (Note) P NA Same as component being Components supported Combustible Gas Control System (Section 6.2.5)

DBA Hydrogen Recombiner 2 ASME III 2 AH P PF/AG 2

Piping and Valves 2 ASME III 2 AB P DF/AG Instruntentation and Controls NA IEEE -279 NA P PF/AG Required to Perform a Safety Function in QA Cat. I Portions IEEE-336 of System Supports for QA Category I (!bte) NA NA (Note) P NA Same as component Deing Components supported Electrical Systems (Chapter 8)

Diesel Generators IE IEEE-308 IE DGB P PF/AG IEEE-323 IEEE-344 IEEE-387 Unit Batteries and Chargers IE IEEE-308 IE BR P PF/AG C. IEEE-323 C% IEEE-344 Vital Bus and Inverters IE IEEE-308 IE ESR P PF/AG IEEE-323 IEEE-344 L_7 VJ 4 of 11 Amendner.t 2 8/30/74

O SWESSAR-P1 TABLE 3.2.5-1 (Cof4T)

AnS Safety Code Tornado F1ond Class Code class Incat ion cr it erion Cr it erion tantes 12nergency Unit Substations IE IEEE-308 IE ESR P PF/AG IEEE-323 IEEE-344 kanergency Station Service IE IEEE-308 IE ESR P PF/AG Switchgear IEEE-323 IEEE-344 mergencv Motor Control Centers IE IEEE-308 IE tSR P PF/AG IEEE-323 IEEE-344 Q)ntrol Paneltoards: IE IEEE-308 IE P PF/AG IEEE-279 Main Control board and Panels CR with Satety Related Functions Diesel Generator Panel ESR Radiation Monitor Panel CR Auxiliary Shutdown Panel ESR Control Room Air Conditioning CR Control Panel Combustible Gas Control Panel AB Class IE Systems Cable IE IEEE-308 IE All P PF/AG Raceway Sup; orts IEEE-323 Locations IEEE-344 Cables and Connections IE IEEE-308 IE All P PF/AG IEEE-323 Locations IEEE-344 IEEE-383 Fuel Pool Cooling and Purification System lSection 9.1.3)

Fuel Pool Cooling Pumps 3 ASME III 3 AB P PF/FS g

O Fuel Pool Coolets 3 ASFE III 3 AB P PF/FS Piping and Valves Required f or 3 ASME III 3 AE P PF/FS Cooling Ud NA (Note) P t.A Sane as cor'ponent being sum orts for VA Cat. I (Note) NA supported (J1 Otagorents and component s which p, could attect satety related 4

components 5 of 11 Amendment 4 11/1/74

SWESSAR-P1 TABLE 3.2.5-1 (CONT)

ANS Safety Code Tornado Fl(od Class Code Class Incat ion Cr it erion Criterion totes Reactor Plant Service Water Systen (Section 9. 2.1)

Reactor Plant Service Water Pumps 3 A132 III 3 -

P SR Reactor Plant Service Water 3 ASME III 3 -

P NA Strainer Piping ard Valves Supplying 3 ASME III 3 AB/ P SR Cooling Water to QA Cat. I OY(burled)

Equipnent Instrumentation and Controls NA IEEE-279 NA ESA/ P SF kequired to Perform a Safety & DGB/CB Mznction in QA Cat. I IEEE-336 Iortions of System Sup[ orts for VA Cat. I (Not e) NA NA (Note) P NA Same as comIonents lxting tbmponents supported Reactor Plant Component Coolinq Water System (Section 9.2.2)

Reactor Plant Component Cooling 3 ASME III 3 AB P PF/FS Pumps heactor Plant Counionent Cooling 3 ASME III 3 AB P PF/AG 1,' urge Tanks Reactor Plant Component Cooling 3 ASME III 3 AB P PF/FS Heat Exchangers Chemical Addition Tank 3 ASME III 3 AB P PF/AG Piping 4..nd Valves Supplying 3 ASME III 3 AB/(3 P PF/FS Cbcling Water to QA Cat. I Iquipnent Instrumentation and Controls NA IEEE-279 NA U3/CS/ P PF/FS kequired to Perform a Safety & CB

& Function in QA Cat. I IEEE-336 lurtions of Sy. item s

Supports for VA Cat. 1 (Note) NA NA (Note) P NA Same as component being tbmponents and comgonents which supported could af fect saf ety related 4 U cxanyments.

LD U1 6 of 11 Amendment 4 11/1/74

SWESSAR-P1 TABLE 3.2.5-1 (CONT)

ANS Safety Code TOITiado F1 cod Class Code Cla s s Location Criterion Cr it erion Notes Dmineralized Water System (Section 9. 2. 3)

Supports for components which could NA NA HA AB P NA attect safety related mmponents Primary Grade Water System (Section 9.2. 7)

Supports for congenents whi h could NA NA NA AB P NA 4 af f ect safety related compou 2nts C1illed Water System (Section 9.2.8)

Supports for components which could NA NA NA AB P NA attect safety related components Q>mpressed Air Systetr (Saction 9.3.1)

Supports for comionents which could NA NA NA A3 P NA affect saiety related components Reactor Plant Samplino Systems (Section 9.3.2)

Piping and Valves from Reactor 1 ASME III 1 CS/AB P NA Coolant loop and Pressurizer Sampling Up To and Including kesnotely Operated Sample Selection Valve Instrumentation and Controls NA IELE-279 CS/AB P NA kequired to Perf orln a Saf ety f. CR/IRR Function in QA Cat. I fortions IEEE-336 of System Sample Lines Originating Prom 2/3 ASME III 2/3 AB P NA Safety Related Ccunponents, Up 1b and Including the First Isolation

& Valve Outside Containmtsit Supports for VA Cat. I (Note) NA NA (Note) P NA Sare as component being

" supported Gamponents L4 LD 7 of 11 Amendment 4

& 11/1/74

SWESSAR-P1 TAILE 3.2.5-1 (CO!4T)

ANS Safety Ccde Tornado Fl col Class Code Class Incation Criterion Cr it er ion intes Vent and Drain System (Section 9. 3. 3)

Supports for components which could NA NA NA AB P NA attect saf ety related cxxnponents loron Recovery System (Section 9.3.b)

Piping Connecting CVCS and Radimetive Gaseous Waste System 3 ASME III 3 AD P PF/FS Supports for com;onents which could NA NA NA AB P ta dttect satety related components Air Conditioning, Heating, Coolino, and Ventilation Syst nas (Section 9.4)

Cbntainment lieat Removal Fan 2 SMACNA NA CS P FF/AG Cuolers and Associated Equigsnent ARI, AMCA Excluling Chillers Annulus builcag Supply and 3 SMACNA NA AB/ESA PF/AG Exhaust Isolation Dampers ARI, AMCA Diesel Generator Building 3 SMACNA NA DGB P PF/AG Ventilation ARI, AMCA Control Building Heating and 3 SMACNA, NA CB P PF/AG Ventilation, including control ARI, AMCA Building Chilled Water System Engineered Safety Features Area 3 SMACNA NA AB P PF/FS Unit Cooler and Ductwork ARI, AMCA Instrumentation and Control NA IEEE-270 NA ESA/ P PF /AG Only lortions rquired to Required to Fertorm a Safety IEEE-336 DEGB/ mitigate ef fect s of IDCA Function for VA Cat. I Air CB/SWP are Seismic Category I.

O Conditioning, 0001ing, AB/CS p and Ventilation Systems Supports for QA Cat. I Ccanponents (Note) NA NA (Note) P NA Same as conronent being supported ty3 Fire Protection Syste_m (Section 9.5.1)

NI Supports f or cx>mlonent s which could NA NA NA AB P !M 8 of 11 Amendment 4 11/1/74

O SWESSAR-P1 TABLE 3.2.5- (CONT)

A'JS Saiety Code Tornado F1ond Class Code class Location Cr it erion Cr it erion ra>t es a

atfect saf ety related courtenents Diesel Generator Fuel Oil Storage and Transter Systen (Section 9.5.4)

Diesel Generator Fuel Oil Day 3 ASME III 3 DGB P PF/AG Tanks NFPA-37 Diesel Generator Fuel Oil 3 ASME III 3 OY P PF/FS Incated underground in Transf er Purrps diesel generator f uel oil t ransf er purep house.

Diesel Generator Fuel Oil 3 ASME III 3 OY P PF/VS Underground Storage Tanks NFPA-37 Piping and Valves 3 ASME III 3 OY/DGB P PF/FS Instrumentation and Controls NA IEEE-279 NA DGB/0Y P PF/AG Required to Perf orm a Saf ety & IRR/ESR Function in QA Cat. I lortions IEEE-336 of System Supports for QA Category I (Note) NA NA (Note) (Note) NA Same as com[onent tving Cwaponents supported Feactor Plant Gas Supply System (Section 9.S. 8) 4 Supports for components which could NA NA NA AB P NA attect saf ety related cumionents Main Steam System (Section 10.3)

Hsin Steam Piping and Valves f rera 2 ASME III 2 CS/P.SV P PF/AG Steam Generators Up To and Including Main Steam Isolation Valves Main Steam Piping and Valves f rurt 3 ASME III 3 MSV/tds P PF/FS Rain Steam Lines to Auriliary Feedwater Pump Turl;ine D (It Applicable)

D Instrumentation and Cantrols KA IEEE-279 NA PSVjfR/ P PF/FS Re%uired to Perform a Saf et t & IRR Function in QA Cat. I IEEE-336 Initions of System L -.

L7 9 of 11 Amenomm t 4 CD 11/1/74

SWESSAR-P1 TABLE 3.2.5-1 (CONT)

ANS Satety Code Tornado Flcod Class Cc.fe Class Incation Criterion Cr it er ion totes Supports for QA Cat. I (Note) NA NA (Note) P NA Same as corp >nent being Components and congenents which su m rted could at f ect safey related a cucqonents Feedwater SYstMR (Section 10.4.7)

Feedwater Piping and valves Inside 2 ASME III 2 CS/AB P PF /AG Containment Structure Up To and Incitaling Pirst Isolation Valve outside Containcent Structure Instrumentation and Controls NA IEEE-279 NA AB/CR P PF/AG Required to Ferf orm a Saf ety &

} unction in QA Cat. I PLrtions Ik_EE-33 6 of Systems Supports for QA Cat. 1 (Note) NA NA (Note) P hA Sam as emitonent being Qanponents and components which su;Wrt( d I could affect safety related a components St eam Generat or Blowdown Syst em (Section 10.4.8)

Feedwater Piping and Valves Inside 2 ASME III 2 CS/AB P FF/FS Containment Structure I@ To And incitating First Isolation Valve outside Conuinment Structure Instrumentation and Controls NA IET.E-2 79 NA AB/t3R P T F/FS kequired to Perf orm a Saf ety G Function in QA Cat. I Fortions IEEE-336 of Systems Supports for QA Cat. I C40te) NA NA (Note) P hA Same as ox.gonent being Components supported Auxiliary Feedwater System (section 10.4.v.)

CN Piping and valves supplying auxil- 2 ASPI III 2 AL/CS P PF/7J

_ _ , lary teedwater from and including containment isolation valves to cunnections with f eedwater lines or with steam generatcrs L4 g Auxiliary Feedwater Storage Tank 3 ASME III 3 AB P FF/PS

• m

• 10of 11 A;aendmen t 4 11/1/74

SWESSAR-P1 TAPLE 3.2.5-1 (CONT) -

ANS Safety Code 'Ibrnado Flood -

Class Code Class Incation Criterion Cr it er ion Notes (AFSD Auxilj ary Feedwater Pumps (Turbine- 3 ASME III 3 AB P PF/PS and Motor-Driven)

Piping and Valves Supplying Auxil- 3 ASME III 3 AB P PF/FS iary Feedwater frun AFST to but not including containment isolatim valves Instrumentatim and Controls NA IEEE-279 NA AB/CR/ P PF/FS Required to Perform a Saf ety 6 IRR Function in QA Cat. I IEEE-336 Portions of System Suppo. 3 for QA Cat. I (Not e) NA NA (Note) P NA Same as component being Cbmpa ents supported luxiliary Steam and CoMensat e (Section 10.4.12)

Emport s for components which cryuld NA NA NA AB P NA affect safety related cunponents Radimetive Litruid Waste System (Section 11.2)

Supports for components which could NA NA NA AB P NA affect safety related componenta

_ Rag nactive GaseouJ Waste System (Section 1*,3)

Proceas Gas Compressors 3 ASME III 3 AB P NA Procest. Gas C n u s.or Aftercooler 3 ASME III 3 AB P NA Procer s , av s3ar Profilter 1 ASME III 3 AB P NA Process A . al Bed Adsorber 3 ASME III 3 AB P NA D Proces s L. , ,erant Dryer 3 ASME III 3 AB P NA

> Degasifier Recevery Exchanger s 3 ASME III 3 AB P NA Degasifier Condenser 3 ASME III 3 AB P NA fj g 11 of 11 Amendment 7 0 2/28/75

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O SWESSAR-P1 TABLE 3.2.5-1 (COh"r)

ANS Safety Code Tornado F1ood Class Code Class Location Criterion Criterion Not es Degasifier Feed Preheater 3 ASME III 3 AB P 14A Degasifier 3 ASME III 3 Ab P NA Degasifier hecirculation Purp 3 ASME III 3 AB P NA Piping and Valves in Process Gas 3 ASME III 3 AB P NA

?ortion of System Supports for QA Cat. I (Note) NA NA (Note) P NA Same as component being Components supported 0

Fuel Building Monitors 3 IEEE 279 IE AB P PF/AG Miscellaneous Containment Structure Polar Crane NA NA NA CS P NA Designed for earthquake in unloaded condition 130 Ton Crane NA NA NA AB P NA Designed for earthquake in unloaded condition D

N C7 C3 m

11a of 11 Azaendn ent 17 9/30/75

(

.d' F*

(

L 662 002

SWESSAR-P1 3.3 WIND AND TORNADO LOADINGS 3.3.1 Wind Loadings 3.3.1.1 Design Wind Velocity A 120 mph basic wind speed is used as a design wind velocity to determine wind loadings on Seismic Category I structures. Tha basic wind speed has a 100 year mean recurrence interval.

3.3.1.2 Basis for Wind Velocity Selection The basic wind speed of 120 mph corresponds to the fastest mile of sustained wind for a 100 year mean recurrence interval and is based on observed air flow in open level country at a hei:T ht of 30 feet above ground. The selection of 120 mph is based on a maximum value applicable to approximately 90 percent of the continental United States.

The basic wind spaed of 120 mph was obtained from ANSI AS8.1 -1972 (Re f . 1) , page l l, , Figure 2. Additional meteorological data are presented in Seccion 2.3.

3.3.1.3 Vertical Velocity Distribution and Gust Factor The vertical telocity distribution and ef fect of gust factors are in accordance with ANSI A58.1-1972. The gust factors are included in the determination of effective velocity pressures as described in ANSI A58.1-1972. The site is conservatively assumed to be flat, open country which results in the use of the 1/7 power law for vertical velocity distribution.

3.3.1.4 Determination of Applied Forces The resultant design wind pressures pq applied to Seismic Category I structures are determined in the following manner:

1. The effective velocity pressures are determined from Tables 5, 6, and 12 of ANSI A58.1-1972 for various elevations above grade using the 120 mph basic wind speed and exposure C for flat open country as criteria.

The external effective velocity pressures ) for structures and (q p ) for parts of structures (q an, d the internal effective velocity pressure (qm) for structures are shown in Table 3.3.1-1. The effect of gust factors is included in the determination of effective velocity pressures as provided by ANSI A58.1-1972.

2. The external (Cp ) or internal (Cpi ) wind pressure co-efficient is determined for the structure being considered. The wind pressure coefficient- define the pressure acting at local positions on the s ' ace of the 3.3-1 662 003

SWESSAR-P1 structure. Internal wind pressure coefficients are based on ANSI A58.1-1972 for all Seismic Category I structures. The external wind pressure coefficients used for rectangular shaped structures are based on ANSI AS8.1-1972 and are shown in Table 3.3.1-2. The external wind pressure coefficients used for cylindrical shaped structures are based on Table 4 (f) ASCE Paper No. 3269 (Ref. 2) and are shown in Table 3.3.1-3. The external wind pressure coefficients used for the containment dome are based on " Wind Stresses in Domes" (Ref. 3) and are shown in Table 3.3.1-4. Use of local pressure coefficients is in accordance with ANSI ASB.1-1972.

3. The resultant design wind pressures (W) applied to Seismic Category I structures are determined as:

W = qC p -g m Cp where:

q=q or q tor enclosed struc+"res (see ANSI AS8.1-1972). A negative value fc .. indicates that the resultant pressure acts oncward. The external and internal pressures are combined so as to yield the maximum stresses.

The resultant design wind pressures and distributions for Saismic Category I structures are shown in Table 3.3.1-5 for rectangular shaped structure: , Table 3.3.1-6 for cylindrical shaped structures, and Table 3.3.1-7 for the containment dome.

A step function of pressure wit'.4 height is used; the specf Med resultant design wind pressure at a given heig'..c is applied over a height zone defined by one-half the difference in adjacent heights for which the design wind pressures are specified. The resultant design wind pressure acts normal to the surface of the structure being considered.

I 3.3-2 Amendment 1 7/30/74

/r' oum 004

SWESSAR-P1 3.3.2 Tornado Loadings 3.3.2.1 Applicable Design Parameters Maximum tornado design parameters are:

1. Maximum wind speed: 360 mph

2. Rotational wind speed: 290 mph

3. Translational speed: 70 mph

4. Radius of maximum rotational speed: 150 ft

5. Pressure drop: 3 psi l7

6. hate of pressure drop: 2 psi /sec The maximum wind speed is the sum of the rotational speed camponent , 29') n:ph, and the tornado translation (70 mph) .

l7 3.3.2.2 Determiration of Fbrces on Structures Tornado forces M, ) applied to Seismic Category I structures consist of wind load (Wy ), differential pressur "f p ), ai d structural response to missile impact (W* ) . Ttm ou .orces are determined in the following manner:

7 Wind Load W W 7 The external effective velocity pressure, q, is determined as:

q = 0.00256 V2 Where: q = external effective velocity pressure in psf V = applicable design wind speed in mph No gust factors or variations of pressure with height are applied to the tornado wind loading. The external effective velocity pressure is applied to structures uniformly vith respect to height and normal to the surface upon which it actn.

The resultant tornado design wind pressure, pW , is determined as:

Pg ,= qcp Where: Cp = external wind pressure coefficient The external wind pressure coefficient, C p , is determined for rectangular structures, cylindrical structures, and the containment dome as described in Section 3.3.1.4. The wind pressure coefficients define the distribution of pressures acting on the surface of the structure.

The wind load W, is the area integral of the wind pressure, p acting over the exposed surface of the structure of component. 7 3.3-3 Amendment 7 2/28/75 ff-D O ._

rnr be-

SWESSAR-P1 Dif ferential Pressure W p

The tornado induced atmospheric pressure drop, p , is determined C

from the cyclostrophic wind equation.

R a= p dr O

Where:

0 = mass density of air V = rotational wind speed r = radius R = one-half of damage path width The differential pressure load W is the area integral of the tornado induced pressure drop P c abing on the structure or component.

atructural Response to Missile Impact W m

The structural response to missile impact, W*, is determined as described in Section 3.5.

Resultant Tornado Design Load W t

Seismic Category I structures are designed to withstand tornado load combinations as follows:

(i) W t =Ww (ii) W t =Wp (iii)W, =W m (iv) W t =Ww + .5 W P

(v) Wt =Ww +W m 16 (vi) W t =Ww + .5W p + Wm Where: W ..... total tornado load S

3.3-4 Amendment 16 8/29/75 e o UD4 p) O.n ,

SWESSAR-P1 W , ..... tornado wind load W ..... tornado differential pressure load, and P

W m ..... tornado missile load 3.3-4A Amendment 16 6 6 r, n- 8/29/75 2 1.r U /

SWESSAR-I1 For each particular structure or portion thereof, the most adverse of the above combinations are used, as appropriate.

Equation (iii) which consists of missile impingement effects, W ma is calculated on a time history basis. A description of this calculation, including an example, is given in Section 3.5.4.

3.3.2.3 Ability of Seismic Category I Structures to Perform Despite Failure of Structures Not Designed For Tornado Loads All Seismic Category I structures are designed to withstand tornado wind loads to ensure the integrity of safety related systcas . The design of all other structures allows a partial or ccxnplete structural f ailure under tornado conditions, when safety related structures, systems, or components are not endangered.

3.3.3 Interface Requirements

'Ihis design meets the wind and tornado protection requirements of the NSSS Vendors

• Section 3.3, and individual system requirements 21 listed in the following SWESSAR-P1 tables: 5.1-1, 5.5.7-1, 6 . 3 -3 ,

and 9.3.4-1.

References for Section 3.3.

1. "American National Standard Building Code Requirements for Minimum Design Loads in Buildings and Other Structures," ANSI A58.1-1972.

2. " hind Forces on Structures," Transactions of the American Society of Civil Engineers (ASCE) , Paper do. 3269, Volume 126, Part II.

3. "dind Stresses in Domes," P. Gondikas, M. C. Salvadori, Engineering Mechanics Division, ASCE, October 1960.

4. " Wind Speed ar.d Air Flow Patterns in the Dallas Tornado of April 12, 1957," Monthly Weather Review, May 1960.

3.3-5 [1(,2 CCCAmendment21 2/20/76

SWESSAR-P1 TABLE 3.3.1-1 EFFECTIVE VELOCITY PRESSURES (1)

Effective Velocity Pressure ipsf)

Internal

-External l'ressures- - -

P res sure -

Elevation Overall Structures Parts of Structures Above Grade Areas >1,000 sq ft Areas $1,000 sq ft Structures

( f t) qf-(psf) ap-(psf) qm { psf) 30 or less 48 55 37 50 54 61 43 100 63 70 52 150 69 77 58 200 74 82 63 (1) Based on Tables 5, 6, and 12 of ANSI A58.1-1972, using a 120 mph basic wind speed end exposure C as 'riteria.

Gust factors are included.

The effective velocity pressure acts uniformly over a height zone defined by one-half the dif #erence in adjacent heights f or which the effective pressures are specified.

1 of 1

SWESSAR-P1 TABLE 3. 3.1-2 WIND PRESSURE COEFFICIENTS FOR RECTANGULAR STRUCTURESC1)

Wind Pre ssure - Coe1lficient External Internal Location- -C p C pi Windward Wall +0.8 10.3 Leeward Wall (2) --0 . 5 10.3 Side Walls -0.7 10.3 Flat Roof (2) -0.7 10.3

+ Indicates positive pressure

- Indicates suction pressure (1) Based on ANSI A58.1-1972, Tables 7 and 11 for italls, and Section 6.5.3. 2.1 for roofs.

(2) Height /least width of structure <2.5.

1 of 1 n"]

SWESSAR-P1 TABLE 3.3.1-3 WIND PRESSURE COEFFICIENTS FOR CYLINDRICAL STRUCTURES (HEIGHT / DIAMETER ? 1) a WIND PIAN OF STRUCTURE Wind Pressure Coefficients-Horizontal Angle External (1) Inte rnalC 2 )

c -Co- C pi 00 +1.0 10.3 150 +0.8 10.3 300 +0.1 10.3 450 -0.7 10.3 600 -1.2 10.3 750 -1.6 10.3 900 -1.7 10.3 105 -1.2 10.3 1200 -0.7 10.3 1350 -0.5 10.3 150 -0.4 10.3 1650 -0.4 10.3 1800 -0.4 10.3

+ Indicates poeitive pressure

- Indicates suction pressure (1) Based on ASCE Paper No. 3269, Table 4 (f)

(2) Based on ANSI A58.1-1972 1 of 1

f. 3 U' 0\\

SWESSAR-P1 TABLE 3. 3.1-4 EXTERNAL WIND PRESSURE COEFFICIENTS (Cp )

CONTAINMENT DOME JL PH I

• O' P,

WIND '

. _ r w e ra .o*

! ELEVATION hyg7 ABOVE GRADE Cp= 1/2 sin (phi) cos (theta) Ref. 3 where: qf = external effective velocity pressure at spring line elevation above grade phi, theta = angles that define any position on con-tainment dome surface External- Wind Pressure CoefficientsC 1) .

theta Phi 00 +300 +600 +900 +1200 +1500 +1800 900 .5 .43 .25 0 600 .43 .375 .21 0 300 .25 .21 .125 0 00 0 0 0 0 0 0 0

-300 0 .125 .21 . 25

-600 0 .21 .375 . 43

-900 0 .25 .43 .5

+ Indicates pcsitive pressure

- Indicates suction pressure (1) Based on " Wind Stresses in Domes."

(Internal pressure coefficient, Cpi = 10.3.

Based on ANSI A58.1- 197 2) n -

o . t ~p \W 1 of 1 OUL

SWESSAR-P1 TABLE 3.3.1-5 RESULTANT DESIGN WIND PRESSURE (W)

FOR RECTANGUIM STRUCTURES Resultant Design Wind Pressure (W) (psf)

Overall Structures - Areas >1,000 sq ft Parts of Structures - Areas 51,000 sq ft Internal Underpressare Internal overpressure Internal Underpressure Internal overpressure Elevation Wind- Lee- Side Roof Wind- Lee- Side Roof Wind- Lee- Side Roof Wind- Lee- Side Roof Above Grade w ard ward Walls ward ward Walls ward ward Walls ward ward Walls (ft) Wall waQ wall . Wall Wall Wall Wall Wall 30 or less +50 -13 -23 -2; +27 -35 -45 -45 +55 -16 -27 - 27 +33 -39 -50 -50 50 +56 -14 -25 -25 +30 -40 -51 -51 +62 -18 -30 -30 +36 -43 -56 -56 100 +66 -16 -29 -29 +35 -47 -60 -60 +72 -19 -33 -33 +40 -51 -65 -65 150 +73 -17 -31 -31 +38 -52 -66 -66 +79 -21 -37 -37 +44 -56 -71 -71 200 +78 -18 -33 -33 +40 -56 -71 -71 +84 -22 -39 -39 +47 -60 -76 -76 A negative value f or W indicates that the resultant pressure acts outward.

7s ON rJ .

1 of 1 0

SWESSAR-P1 TABLE 3.3.1-6 RESULT / tfr DESIGN WIND PR ECSURE (W)

FOR CYLINDRICAL S1HUCTURES (AREAS >1,000 sq f t)

N a

\

WIND _

PLAN OF STRUCTUR E Resul tan t Design Wind Pressure (W) (psi)

Elevation. for Internal Underpressure Above Gra3e (ft) a (Horizontal Angle) 0* 158 308 45* 60* 758 90* 105* 120* 135* 150* 165* 1808

+59 +50 +16 -23 -47 -66 -71 -47 -23 -13 -8 -8 -8 30 or less 50 +67 +56 +18 -25 -52 -74 -79 -52 -25 -14 -9 -9 -9 100 +79 +b6 +22 -29 -65 -85 -92 -60 -29 -16 -10 -10 -10 150 +86 +73 +24 -31 -65 -93 -100 -65 -31 -17 -10 -10 -10 200 +93 +78 +26 -33 -70 -100 -107 -70 -33 -18 -11 -11 -11 Resultant Design Wind Pressure (W) (psf) for Interr.nl overpressure a (Horizontal Angle) 30 or less +37 +27 -6 -45 -69 -88 -93 -69 -45 -35 -30 -30 -30

+41 +30 -8 -51 -78 -99 -105 -78 -51 -40 -35 -35 -35 50 100 +47 +35 -9 -60 -81 -116 -128 -81 -60 -47 -41 -4a -41

+52 +38 -11 -66 -100 -128 -135 -100 -66 -52 -45 -45 -45 150 200 +55 +40 -12 -71 -108 -137 -145 -108 -71 -56 -49 -49 -49 A negative value f or W indicates that the resultant pressure acts outward.

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SWESSAR-P1 3.4 WATER LEVEL (FLOOD) DESIGN All Seismic Category I structures and equipment are designed for buoyancy and static water forces from flood water elevation up to the station yard grade which is the design flood level. Dikes or other structures are provided external to the Seismic Category I buildings as required to protect them from the si.te probable maximum flood. These dikes or other structures and t le probable maximum flood levels are presented in Sections 3.8.4 and 2.4 respectively of the Utility-Applicant's SAR.

3.4.1 Flood Protection All Seismic Category I buildings are protected trom the damaging effects of flood water to the station yard grade. Flood protection above yard grade is site related and is presented by the Utility-Applicant, if required. Dynamic water forces are 7 considered only for the dikes or other external flood protection structures a.t are evaluated in accordance with the U.S. Army Coastal Engineering Research Center, Technical Report No. 4 or similar procedures as discussed in the Utility-Applicant's SAR.

Satety related systems and components requiring flood protection are protected by locating them in the Seismic Category I buildings. These structures are described in Sections 3.8.1.1.1 and 3.8.4.1. All construction joints in the exterior walls and mat that are required to resist water pressure include water stops.

Access to these structures is located above grade. The only penetrations below grade, or the design flood level, are the service water lines in the annulus building.

Potential inleakage from drainage lines discharging near grade through such phenomena as cracks in the annulus building walls or leaking water stops is collected in sumps at the lower building level and pumped out.

3.4.2 Analysis Procedures All Seismic Category I buildings are protected from the damaging ef f ects of flood water to the station yard grade. Dynamic water forces are considered only on the dikes or other external flood protection structures when required for a specific site.

Hydrostatic forces associated with the above flood level are included in the fluid pressure loads to be imposed on the structure as identified in Sections 3.8.1 and 3.8.4. Forces acting on those structures are calculated on the basis of full external hydrostatic pressure corresponding to this flood level.

Na Seismic Category I structures are in an unstable condition due to either moment or uplift forces resulting from the load combinations which include the design flood.

3.4-1 Amendment '/

2/28/75 662 0i_/

SW.JSAR-P1 3.4.3 Interface Requirements This design meets the flood protection requirements of the NSSS 21 Vendors' Section 3.4, and indivi 'ual system requirensmts listed in the following SWESSAR-P1 tables: 5.1-1, 5.5.7-1, 6.3-3, and 9.3.4-1.

O in n , .

(UL J Uid 3.4-2 Amendment 21 2/20/76

M b

re w

")

j

.as 662 0.-'t'l

SWESSAR-P1 3.5 MISSILE PROTECTION The general design objectives governing the extent of permissible damage resulting from hypothetical dynamic effects of missiles are as follows:

1. The integrity of the containment system must be maintained.

2. The capability ror shutdown of the pactnr and maintenance of core cooling capability must be ensured.

3. A missile accident which is not a LOCA shall not initiate a loss of coolant The design of the containment and piping systems considers the possibility of missiles being generated from pressurized piping and vessels, rotating equipment, and tornadoes. Tne design ensures that none of the hypothesized missiles violate the design objectives.

'Ihe most positive method of achieving missile protection is by arranging the components so that the direction of missile flight is away frm critical structures and components . This method is one of the principal bases for arranging cmponents . In areas where this arrangement ca.:uiot be achieved, protection is obtained

@- by providing suitable narrier or energy absorbing materials.

Secondary missiles created frm these barricrs wnich have the potential for f ailing safety systems inside containment will be precluded from causing da.nage to safety systes either by 32 thickening of the concrete, the barrier, or by installing spall plates. Engineered safety systems are separated in such a manner that the failure of one cannot cause the fai)ure of the other; or that the failure of any plant cmponent which brings about the need for these engineered safety systems does not render the safety system inoperable.

3.S.1 Missile Barriers and Loadings The structures, shields, and barriers that are designed to withstand external missile effects are listed in Taule 3.5.1-1.

3.S.2 Missile Selection 3.5.2.1 intsrnal Missiles Pressurized components and rotating machinery are the most credible srarces of internal missiles. Valve bonnets and valve stems can be ejected by the fluid jet in the event of a fluid pressure butndary failure. Retaining bolts are potential missiles necause of their stored strain energies. The control rod drive closure cap has been analyzed as a piston type missile acted upon by the 2,250 psia normal reactor vessel pressure. The 662 020 ^* "" M ,3

SWESSAR-P1 control rod drive assembly has been postulated to become a jet propelled missile. The effects of these missiles are analyzed.

Protective measures, such as shields and barriers, are provided if necessary to meet the design objectives.

3.5.2.2 External Missiles Tornado generated external missiles selected for all Seismic Category I structures are listed in Table 3.5.2-1.

The bases for selection of these missiles are that the aerodynamic characteristics necessary for flight in a tornado are present and that they are listed as AEC required missiles. A discussion of potential turbine-generated missiles is presented in Section 10.2.3.

3.5.3 Selected Missiles 3.5.3.1 Selected Internal Missiles The characteristics of the selected internal missiles are given in Table 3.5.3-1. The velocity and kinetic energy of the internal missiles have been calculated by the methods given in ORNL-NSIC-22 ( 2 ) . Additional internal missiles will be identified and analyzed during the course of the detailed plant design. A listing of the missiles used for the design of Seismic Category I structures will be presented later.

3.5.3.2 Selected External Missiles The characteristics of the selected external missiles are given in Table 3.5.2-1.

3.5.4 Barrier Desian Procedures SWECO-7703 (a) presents methods for the design of both local and 35 overall structural effects of missile impact on reinforced concrete barriers.

Missile barriers are designed to def eat the missiles described in Section 3.5.3. Defeat of the missile is achieved if:

1. Perforation of the barrier is precluded.

2. Structural collapse of the barrier is precluded.

The barrier thickness required to prevent missile perforation is determined as follows:

. . .-n 3.5-2 Amendment 35 10/6/77

SWESSAR-P1 For reinforced concrete barriers:

1. Petry forIrulaC*).

2. Ballistics Research Lab Formula (1).

9

-n 3.5-2A Ame.dment 35

,(, (, p- '

10/6/77

SWESSAR-P1

3. Combination of Ammann and Whitney formula (3) and National Defense Research Committee formula in Gwaltney(1).

The concrete wall thickness to prevent perforation using the combine ' Ammann and Whitney and National Defense Research Committs formulas is calculated as follows:

1. Determine penetre lon into an infinite barrier by Eq. 4.1.14 and 4.1.15 f rom Ammann and Whitney Report (3 ) .

2. Determine thickness of concrete to just prevent missile perforation by Eq. 30 from R. Gwaltney(1).

The results of these three methods are compared and the largest wall thickness is used based on all missiles which might strike the barrier.

For steel barriers, the Stanford Research formula (1) is used.

The overall structural capacity of both concrete and steel barriers is determined to preclude structural collapse of the barrier under missile impact loading. Structural response to nissile impact is based on a time history solution, by numerical integration, of the dynamic equations of equilibrium for both missile and barrier from the time of impact through the time of 9 the event. The missile and barrier are modeled as equivalent single degree-of-freedom systems from the general method of analysis presented in Chapter 5 of Reference S. The resistance functions describing the missile and barrier are nonlinear springs. For concrete barriers, the ultimate load capacity is based on yield line theory of reinforced concrete slabs. The effects of static and live loads are included in the determination of structural capacity as described in Sections 3.8.1, 3.8.3, and 3.8.4. Adequate resistance is provided to ensure against an edge shear type of failure as provided in References 6 and 7.

The structural capacity of both concrete and steel missile barriers is determined to preclude structural collapse of the barrier under missile impact loading. Structural response to missile impact is based on time-history calculations for elasto-plastic behavior of the missile and the barrier. The general method of analysis is described in Biggs (Reference 5) . The ultimate load capacity of concrete barri ers is determined by yield line theory. Where penetration does not govern, adequate 14 shear resistance is prcvided based on ACI-318-71 (Reference 6) .

The barriers are designed so that the required duc tility of the '

barrier for a load combination time history is less than the maximum allowable ductility.

The maximum allowable ductility factors will be those given in the " Air Force Design Manual, Principles and Practices for Design 3.5-3 Amendment 14 bb2 {]{ j 7/18/75

SWESSAR-P1 of Hardened Structures, AFSWC-TDR-62-138, December 1962." For concrete barriers, if use of a ductility ratio greater than 10 is g required to demonstrate the design adequacy against missile impact, such usr.ye will be identified and justification provided in the Utility-Applicant application for a Construction Permit.

For a steel barrier, the maximum allowable ductility is the extreme fiber strain at the onse' of strain hardening divided by the extreme fiber strain at the yield.

A consistent measure of ductility for the time history analysis is the maximum deflection of the barrier for a load combination equation divided by the barrier deflection when the tension steel begins to yield for a concrete barrier and when first yield begins for a steel barrier. The deflection of the barrier is limited by the maximum allowable ductility, and the barrier p;aximity no safety related equipment.

Fig. 3.5.4-1 to 3.5.4-3 show a sampic missile barrier time history analysis for the load combination equation W =W The missile is the 4 x 12 in. x 12 ft wood plank at 280 mph The barrier is 2 ft of concrete, 10 ft square, with number 11 rebar each way, each face at 10 inch spacing.

Fig. 3.5.4-1 gives the time history of the force at the barrier support, the force decelerating missile, and the force of the missile on the barrier.

Fig. 3.5.4-2 shows a time history for missile compression, barrier def lection, and barrier ductility.

Fig. 3.5.4-3 shows the time history of energy distribution. Most of the kinetic energy of the missile goes into destroying the missile itself because the barrier is stronger than the missile in punching shear, penetrability, and bending resistance.

Any equipment near a barrier is located so that barrier deflection does not af fect the equipment. For this particular case, the maximum barrier deflection is .0035 ft which is negligible.

3.5.5 Missile Earrier Features Protection from missile threats for the plant is provided in accordance with the criteria of Sectic 4 3.5.1. In the course of detailed plant layout, postulated missile paths will be reviewed and the requirement for either supplemental barrier structures (not associated with basic structural design) , relocation of Seismic Category I components, or redirection of the missile path will be assessed and bnplemented to ensura compliance with the design criteria.

v -

o o' m~ (;3 ' r ic4 3.5-4 Amendment 25 4/30/76

SWESSAR-P1 3.5.6 Interface Requirements This design meets the missile protection requirements or the NSSS Vendors' Section 3.5, and individual system requirements listed in the following SWESSAR-P1 tables: 5.1-1, 5.5.7-1, 6.3-3, and 9.3.4-1.

The Utility-Applicant will identify and justify in his application for a Construction Permit if a ductility ratio greater than 10 for concrete barriers is required to demonstrate the design adequacy against missile impact.

References for Section 3.5

1. Gwaltney, R.C., Missile Generation and Protection in Light Water-Cooled Power Reactor Plants, ORNL-NSIC-22, September 1968.

2. Miller, D.R. and Williams, W.A., Tornado Protection for the Spent Fuel Storage Pool, General Electric APED-5696, Class I, November 1968.

3. Ammann & Whitney Consulting Engineers, Report on " Industrial Engineering Study to Establish Safety Design Criteria for Use in Engineering of Explosive Facilities and Operations Wall Response."

4. Amirikian, A., Design of Protective Structures, Bureau of Yards and Docks, Dept of Navy, August 1950

5. Biggs, J.M., Introduction to Structural Dynamics, McGraw-Hill, New York, Naw York, 1964

6. ACI-318-1971 Building Code Requirements for Reinforced Concrete

7. ACI-359, November 1974 Draft, Concrete Reactor Vessels and Containments

8. Jankov, J.D., Shanahan, J.A., and White, M.P., " Missile Barrier Interaction," SWECO-7 iu3, Stone & Webster Engineering g Corporation Topical Report, September 1977 p) n L'LJ 3.5-5 hbd Amendment 35 10/6/77

SWESSAR-P1 TABLE 3.5.1-1 EXTERNAL MISSILE BARRIERS Item Protected Structures, Shields, Reference No. Enclosure and Barriers Figures 1 Containment Circumferential exterior 1.2-3 Structure wall and dome 2 Annulus Building Circumferential exterior wall 1.2-4 and roof 3 Control Building Exterior walls and roof 1.2-5 4 Diesel Generator Exterior walls and roof 1.2-6 Building 5 HVAC Equipment Roof and exterior walls 1.2-5 Room 6 Electrical Cable Roof and exposed walls 1.2-5 Tunnel 7 Reactor protection Roof and walls 1.2-5 motor generator set in emergency switch-gear room 8 Fuel building Roof and walls 1.2-8 1 of 1 - .9' I r, /

V

[J t 0

SWESSAR-P1 TABLE 3.5.2-1 TORNADC GENERATED EXTERNAL MISSILES Type of Missile Weight (lb) Velocity (fps / mph)

1) Wood plank 4" x 12w x'12' 200 420/286

2) Steel pipe 3" diameter, 78 210/143 10' long schedule 40

3) Steel rod 1" diameter J 310/211 x 3' long

4) Steel pipe 6" diameter, 285 210/143 15' long, schedule 40 8

5) Steel pipe 12" diameter, 743 210/143 15' long schedule 40

6) Utility 13.5" diameter 1,490 210/143 pole x 35' long

7) Automobile 20 ft2 frontal 4,000 100/68 area Note:

The above missiles are considered as striking a structure from any direction, and at any elevation.

.3 m' UUC (j, t /

1 of 1 Amendment 8 3/28/75

SNESSAR-P1 TABLE 3.5.3-1 KINETIC ENERGY OF POTENTIAL MISSILES The kinetic energy of potential missiles for each NSSS Vendor may be found in their respective Safety Analysis Report according to the following schedule.

W RESAR-41 Tables 3.5-1, 3.5-2, 3.5-3, 3.5-4 W-3 S RESAR-3S Tables 3.5-1, 3.5-2, 3.5c'3, 3.5-4 19 BSW B-SAR-205 Table 3.5-1 C-E CESSAR Table 3.5.3-1

/ ir Q DOL ULU 1 of 1 Amend:nent 19 12/12/75

SWESSAR-P1 Table 3.5.4-1 is deleted. 14 b U ': 0 '2?/

1 of 1 Amendment 14 7/18/75

TYPICAL HISSILE BARRIER DESIGN -- WOOD PLANK AT 286 HPH .. 2 FT CONCRETE BARRIER PLANK 4IN. X 12IN. X 12FT. = 200 LB. CONCRETE FC = 3000. NO. 11 REBAR AT 10 IN 8- X FORCE AT BARRIER SUPPORT 8 0 FORCE OECELERATING MISSILE O FORCE OF MISSILE ON BARRIER e-8 m _

N = _

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AMENDMENT 7 2/28/75

e @

TTPICAL MISSILE BARRIER DESIGN -- WOOD PLANK AT 286 MPH .. 2 FT CONCRETE BARRIER PLANK 4IN. X 12IN. X 12FT. = 200 LB. CONCRETE FC 3000. NO. 11 REBAR AT 10 IN m 8 o

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_g MAX ALLOWABLE CUCTILITY = 10.0 m s- PER g -

mo 7Ng _ ULTIMATE STRENGTH DESIGN C- ACI CODE 318 m,_ cm ______________

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mmmo- s o.

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g.00 0'.04 W8 0 0'.16 $.00 0'.04 0'.08 0 0 16 o TIME (SEC) =10-[.12 TIME (SEC) = 10-[.12 CN CN N FIG 3.54-2 TIME HISTORY FOR MISSILE COMPRESSION, BARRIER DEFLECTION AND BARRIER DUCTILITY C PWR STANDARD PL ANT L4 SAF E T Y ANALYSIS REPORT SWE SSAR - P l AMENDMENT 14 7/18/7S

9 TYPICAL MISSILE BARRIER DESIGN -- WOOD PLANK AT 286 MPH . . 2 FT CONCRE TE ElRRIER PLANK 4IN. X 12IN. X 12FT. : 200 LB. CONCRETE FC 3000 NO. 11 REBAR AT 10 IN IN!IAL ENERGY FlhAL ENERGY A B C oistRisurroN oisrRieurroN

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KINETIC ENERoY BARRIER BARRIER EL Asto-PL ASTIC total ENER0Y oEFLECTION ENERGY A*s*C+D+E*F+G+M Q' FIG. 3. 5. 4 - 3 TIME HISTORY FOR N ENERGY DISTRIBUTION PWR STANDARD PLANT SAFETY ANALYSIS REPORT SWESSAR-PI AMENOMENTT 2/28/7S

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GWESSAR-P1 3.6 PROTECTION AGAINST DYNAMIC EFFECTS ASSOCIATED WITH THE POSTULATED RUPTURE OF PIPING The containment boundary equipment and systems esnential for a safe shutdown are protected against the effects of nlowdown jet forces and pipe whip resulting from a postulated hich energy pipe rupture. The criteria for protection against pipe whip within the containme.it conform to Regulatory Guide 1.46 (Section 3A.1-1.4 6) as qualified by the detailed criteria stated in Sections 3.6.2.1 and 3.6.2.3. The criteria for protection against pipe break outside the containment conform to the guidelines contained in the branch technical positions APCSB-3-1 and MEB-3-1 (see Standard Review Plans 3.6.1 and 3.6.2, dated March 1975) as discussed in Sections 3.6.2.2 and 3.6.2.3.

3.6.1 Systems in which Design Basis P3 ping Breaks Occur The design basis piping break is postulated to occur in high energy piping systems (or portions of systems) in accordance with 20 the criteria stated in Sections 3.6.2.1 and 3.6.2.2. High energy piping systems are defined as systems that have either a maximum operating temperature exceeding 200 F or a maximum operating pressure exceeding 275 psig. Maximum operating temperature and pressure are defined as the maximum temperature and pressure in the piping systems during occurrences which are expected frequently or regularly in the course of power operation, refueling, or maintenance of the plant. High energy systems inside the containment are identified in Table 3.6-2. High energy systems outside the containment are identified in Table 3.6-1.

Through-wall leakage cracks are postulated in accordance with the criteria stated in Section 3.6.2.2 for pioing systems outside the containment with both a maximum operating temperature and pressure equal to or below 200 F and 275 psig, respectively.

These piping systems are defined as moderate energy systems.

Moderate ener;f systems outside the containment are identified in Table 3.6-1.

The division of responsibility between the NSSS Vendors and Stone & Webster for protection aaainst dynamic effects associated with the postulated rupture of RCS piping is defined in 20 Table 1.8-1.

3.6.2 Desigp Basis Piping Break Criteria 3.6.2.1 Criteria for Inside the Containment Wherever possible high energy piping is separated from systems required for sate shutdown by remote location or structural enclosures. All postulated breaks are systematically analyzed for potential damage to systems and structures required for safe shutdown resulting from pipe whip, jet impingement, and cubicle 3.6-1 - - , Amendment 20 Ubd bJ A 1/23/76

SWESSAR-P1 pressurization. If the damage is unacceptable in terms of relevant protection criteria, protective measures are taken. If rerouting of piping, relocation of equipment, or providing additional enclosures is not feasible, piping whip restraints and/or jet impingement shields are installed to protect the structures, com e aents, and systems required for safe shutdown.

Table 1.8-1 defines the responsibilities of the NSSS Vendor and Stone & Webster with regard to the reactor coolant system. For the reactor coolant loop, the NSSS Vendor is responsible for the design ano analysis of the components, component supports, pipina layout, piping stress analysis, and piping supports and restraints, with the exceptions of the reactor vessel support structure, prassurizer surge line layout, and brandi piping 35l connecting to tre reacter coolant primary loop. In accordance with this responsibility breakdown, the NSSS Vendor is responsible for the design basis break locations, break flow areas, and types postulated for the primary coolant system piping. The final design of the primary system results from an interaction between NSSS piping and equipment supports and balance of plant structures.

The NSSS Vendors have provided break type (s) , location (s) ,

area (s) , and blowdown data in their safety analysis reports referenced in Section 1.6. The worst cases (double ended ruptures and one-cross sectior.a1 area splits) are presented to demonstrate the methodtlogy. The NSSS Vendors have supplemented the information presented in their SARs with topical reports, as listed in Table 3.6-3. They have also specifically provided that breaks limited by mechanical restraints can be used in the analysis. These criteria and methodology have been reviewed by the NRC. For B-SAR-205, the Utility-Applicant shall provide 35 justification for use of the 2.5 ft2 pump discharge and hot leg breaks in the reactor cavity.

The NSSS Vendors supply to S&W their criteria and design requirements, as discussed above. SSW then accommodates these design requirements in the design of the balance of plant structures. The resultant breaks, as used in the design of the containment structure subcompartments , are identified in Table 6.2.1-18, along with the NSSS Vendor reference for each break.

The detailed design relating the criteria and design requirements to the break analysis in a piping system incorporating a specific restraint scheme will be provided in the SWESSAR-P1 application fur Final Design Approval or the first application for an operating license by a Utility-Applicant referencing a SWESSAR-P1 Preliminary Design Approval, whichever comen first.

It is possible that the NSSS Vendor criteria and design require-men +s may require modification during the detailed design of the primary coolant system and its st_ ,3rting structures . If so, the modified critoria and design requirements with the resultant pipe break locations, types, and flow areas will be submitted to the Amendment 35 3.6-2 6'J

- 'J

?

10/6/77

SWESSAR-P1 NRC for review. This modification will be in the first document associated with SWESSAR-P1 submitted to the NRC following the identification of the need to modify the criteria, i.e, the applica tion for a construction permit by a Utility-Applicant referencing SWESSAR-P1 Preliminary Design Approval or the SWESSAR-P1 application for Final Design Approval.

Design basis break locations and types in the balance of plant piping are postulated in accordance with Regulatory positions in Regulatory Guide 1.46 as modified by Section 3.6.2.

3.6.2.1.1 Break Locations - Class 1 Piping ASME Section III Code Class 1 piping breaks are postulated to oc^ tr at the following locations in each piping or branch run:

a. At terminal ends of the pressurized portions of the runs, and

b. At intermediate 1ccations between terminal ends selected by either of the following criteria:

1. At each fitting (e .g . , elbow, tee, cross, flange, and non-standard fittines) , welded attachment, and valve. Where the piping contains no fittings, welded attachments, or valves at ona location at each extreme of the piping run adjacent to or within the protective structure; or

2. Any intermediate locations between terminal ends where the primary plus secondary stress intensities (circumferential or longitudinal) derived on an elastically calculated basis under loadings associated with OBE and operational plant conditions exceed 2.4 S3 ; and ill 1 S* is the design stress intensity as specified in Section III of ASME Boiler and Pressure Vessel Code.

2 U is the cumulative usage factor as specified in Section III of the ASME Boiler and Pressure Vessel Code.

3.6-2A f ,

Amendment 35 OO/ {lg g 10/6/77

SNESSAR-P1 At any intermediate locations between terminal ends where the cumulative usace factor (U 2) , de::ived from the piping f atigue analysis under the loadings associated with OBE and operational plant conditions , exceeds 0.1 but at not less than two separated locations selected on the basis of highes t cumulative usage factor or stress intensity.

3.6.2.1.2 Break Locations - Class 2 and 3 Piping ASME Section III Code Class 2 and 3 piping breaks are postulated to occur at the following locations in each piping run or branch run:

a. At terminal ends of the pressurized portions of the run, and

b. 3t intermediate locations selected by either of the following criteria:

1. At ea ch pipe fitting (e .g . , elbow, tee, cross, flange, and non-standard fitting), welded atta chment , and valve. Where the piping contains no fittings, welded attachments, or valves at one location at each extreme of ae piping run adjacent to or within the protective structure; or

2. At each location where the stresses associated with upset plant conditions and an OBE event, calculated by Eq (9) and (10), Para NC-3652 of the ASME Code,Section III, exceed 0.8 (1.2 Sh +SA) but not less than two separated locations chosen on the basis of highest stress. Where the piping consists of a straight run without fittings, welded attachments, or valves, and all stresses are below

.8 (1.2 Sh +S), A a minimum of one location chosen on the basis of highest stress.

The criteria stated in Section 3.6,2.~ provide supplementary information on the types and orientations of breaks at each location. The criteria allow for selection of only a circumferential break at a given location, provided that detailed stress analysis shows that the axial stresses are significantly higher than the hoop stresses; or only longitudinal breaks, if the hoop stresses are significantly higher. For longitudinal breaks at fittings, the criteria allow for selection of two diametrically oppos ed points on the piping circumference, such 3 S3 and Sh are stress limits associated with normal and upset plant conditions, and an OBE event as calculated by Eq (9) and (10), Para NC-3652 of the ASME Code,Section III, for Class 2 and 3 piping.

3.6-3 Amendment 8 662 037 3/28fv5

SWESSAR-P1 that the jet reacti .n w 1ses r it-of plane bending. In accordance with these crit u.a, lo..gitadinal breaks are not required at t ermi; .1 points - seumless piping, nor at intermedic.te locations where stresses e low but a m#nimum number of break locations must -> ratisfi 4. These dr.ta i led provisions complement Regulatory Guide 1.46 and are consistent with the regulatory positions of the guide.

Sections 3.6.2.1.1 and 3.6.2.1.2 allow ir postulation of intern ediate breaks at each pipe fitting, as an alternative to selection on the basis of stress and camtlative usage. This method, although not included in the regulatory guide, provides a comparable or greater degree of protection. The stress limits for break postulation differ somewhat from those in the regula tory guide. For Class 1 piping, 2.4 S m is specified for both austenitic and ferritic steel, compared to 2.4 43 for austenitic and 2.0 S n3 for territic in the guide. For Class 2 and 3 piping, the stress limit of 0.8 (1.2 Sh +S'g) replaces the limit of 0.8 (S h *SA) in the guide.

8 Exception is taken to Footnotes 10 and 11 of Regulatory Guide 1.46 which deal with longitudinal break area and motion of a whipping pipe. Footnote 10 specifies that the break area of a longitudinal break is equal to the sum of the effective cross sectional flow areas upstream and. downstream of the break location; whereas Section 3.6.2. 3.2 (4) states that the dynamic force of the fluid jet discharge is based on circular break area equal to the ef f ective cross sectional flow area of pipe at the break Jocation. Footnote 11 of the guide states that for circumferential break the pipe can whip in any direction normal to the pipe axis; whereas Section 3.6. 2.3.1 (5) assumes pipe whip to occur in a direction defined by the piping geometry, stiffness and fluid forces. These exceptions provide no diminution of the safety criteria. Finally, the application of these specific positions results in a consistent criteria for break postulation inside and outside the containment.

3.6.2.2 Criteria for Outside the Containment 3.6.2.2.1 High Energy Fluid Syatems A. Fluid Systems Separated f rom Essential Structures , Systems, S 8f Cocoonents Recuired for Safe Shutt.own The primary objective in the piping layout and plant arrangement is to satisfy the separation criteria so that the ef f ects of postulated piping breaks at any location are isolated or physically remote f om structures, systems, and components required for safe shutdown. Pipe breaks are not postulated in 8 these separated high energy piping systems. , . .

J b h.

3.6-4 Amendment 8 3/28/75

SWESSAR-P1 B. Fluid System Piping Between Isolation Valve and Containment Portions of high energy fluid system piping between the isolation valve and the containment penetration are protected by pipe whip \24 restraints that are capable of resisting bending and torsional moments produced by a postulated piping failure outboard of the containment isolation valves. The restraints are located reasonably clcre to the outboard side of the isolation valves and are designed to withstand the containment loadings l24 vesulting from a postulated piping failure beyond these portions of piping s; that neither valve operability nor the leaktight integrity of the containment is inpaired.

The pipe whip restraints for the main steam and feedwater pipes are located on the outside wall of the main steam and feedwater valve area of the annulus building. No breaks are postulated inside the valve area. The first location where a Dreak is 24 postulated in either system is at the terminal point on the outside of the outside wall of the valve area.

Terminal ends of the piping runs extending beyond these portions of hig'a energy fluid system piping are considered to originate at a p=>in t adjacent to these required pipe whip restraints located outside containment.

1. Breaks are not postulated in these portions of piping that are designed to meet the requirements of ASME Code,Section III, Subarticle NE-1110, and the following additional requirements for Class 2 piping:

(a) The maximum stress ranges, as calculated by Eq (9) and (10) in Para NC-3652, ASME Code,Section III considering upset plant conditions (i .e., sustained loads, occasional loads, and thermal expansion) and an OBE event, do not exceed 0.8 (1.2S h *OI*

A (b) The mavimum stress ranges, as calculated by Eq (9) in Fara NC-3652 under the loadings resulting from a postulated piping failure of fluid system piping beyond these portions of piping, do not exceed 1.8S.

h

2. Welded attachments, for pipe supports or other purposes, to these portions of piping are avoided.

3. The number of circumferential and longitudinal piping welds and branc h connections are minimized.

4. The length of these portions of piping are reduced to the minimum length practical.

5. The design of pipe anchors or restraints (e .g . ,

connections to containment penetrations and pipe whip restraints) do not require welding directly to the outer 3.6-4A , o r J Amendment 24 U U /- UJl 4/23/76

SWESSAR-P1 surface of the piping (e .g . , flued integrally forged pipe fittings may be used) except where such welds are inspected in accordance with the requirements of Section 16.4.2 and a detailed stress analysis is performed to demonstrate compliance with the limits.

6. For these portions of high energy fluid system piping, inservice examinations will be performed in accordance with the requirements described in Section 16.4.2.

C. Fluid Systems Enclosed Within Protective Structures

1. Breaks in ASME Code,Section III, Class 2 and 3 piping are postulated at the following locations in each piping and branch run (except those portions of fluid system piping identified in 3.6.2.2.1.B) within a protective 24 structure containing systems and components required for safe shutdown:

a. At terminal ends of the pressurized portions of the run if located with the protective structure.

b. At intermediate locations selected by either of the follownig criteria:

1. At each pipe fitting (e.g . , elbow, tee, cross, and nonstandard fitting) , welded attachment, and valve. If the run contains no fittings, at one location at each extreme of the run (a terminal end, if located within the protective structure, may substitute for one intermediate break).

2. At each location where the stresses

• exceed 0.8 (1.2Sh *SA ) but not less than two separated locations chosen on the basis of highest stress. In the case of a straight pipe run without any pipe fittings, welded attachments, or valves with stresses below 0.8 (1.2S h+S A) , a minimum of one location chosen on the basis of highest stress.

2. Breaks in nonnuclear class piping are postulated at the following locations in each piping or branch run:

a. At terminal ends of the pressurized portions of the run if located within the protective structure.

b. At each intermediate pipe fitting and welded attachment and valve.

• Stresses associated with normal and upset plant conditions, and an OBE event as calculated by Eg (9) and (10) , Par NC-3652 of the ASME Code,Section III, for Class 2 and 3 piping.

3.6-4B Amendment 24 4/23/76

, a Ob$ (m 17 e n

SWESSAR-P1 D. Fluid Systems Not Enclosed Within Protective Structures

1. Breaks in ASME Code,Section III, Class 2 and 3 piping, are postulated at the following locations in each piping and branch run (except those portions of fluid system piping identified in 3 . 6 . 2 . 2 .1.B) outside but routed adjacent to a protective structure containing systems 1 and components required for safe shutdown. 24

a. At terminal ends of pressurized portions of the run if located adjacent to the protective structure.

b. At intermediate locations selected by either of the following 7riteria:

1. At each pipe fitting (e . g . , elbow, tee , cross, and nonstandard fitting) .

2. At each location where the stresses

• exceed 0.8 (1.2Sh *SA ) but at not less than two separated locations chosen on the basis of highest stress. In the case of a straight pipe run without any pipe fittings or welded attachments and valves and stresses below 0.8 (1.2S h*S A) , a minimum of one location chosen on the basis of highest stress,

2. Breaks in nonnuclear class piping are postulated at the following locations in each piping or branch run:

a. At terminal ends of pressurized portions of the run if located adjacent to the protective structure.

b. At each intermediate pipe fitting, welded attachment, and valve.

3.6.2.2.2 Moderate Energy Fluid Systems

1. For the purpose of satisfying the separation provisions of plant arrangement, a review of the piping layout and plant arrangement drawings is conducted to show that the effects of through-wall leakage cracks at any ation are isolated or physically remote from safe , tdown systems.

2. Leakage cracks are not postulated in those portions of piping between isolation valve and containment provided they meet the requirements of ASME Code,Section III, Subarticle NF,-1120, and are designed such that the

• Stresses associated with normal and upset plant conditions, and an OBE event as calculated by Eq (9) and (1) , Par NC-3652 of the ASME Code,Section III, for Class 2 and 3 piping.

3.6-4C Amendment 24 bh2 Ci[ j 4/23/76

SWESSAR-P1 maximum stress range does not exceed 0.4 (1.2S +q ) for &

ASME Code,Section III, Class 2 piping. W

3. Through-wall leakage cracks are postulated in 'luid stem piping located within, or outside and adjacent to, protective structures except where exempted by Sections 3.6.2.2.2 (2) and 3.6. 2.2.7 (4) or where the maximum stress range in these portions of Class 2 and 3 piping (ASME Code,Section III) , or nonnuclear piping ia less than 0.4 (1.2Sp +S ) . The cracks are postulated to occur individually d t 'glocations that result in the maximum effects from fluid spraying and flooding, with the consequent hazards or environmental conditions developed.

4. Cracks are not postulated in moderate energy fluid system piping located in an area in which a break in high energy fluid system is postulated, provided such cracks would not result in more limiting environmental conditions than the high energy piping break. Where a postulated leakage crack in the moderate energy fluid system piping results in more limiting environmental conditions than the break in proximate high energy fluid system piping, the provisions of (3) are applied.

5. Through-wall leakage cracks instead of breaks are postulated in in the piping of those fluid systems that qualify as high energy fluid systems for only short operational periods, but qualify as moderate energy fluid systems for the major operational period.

An operational period is considered short if the fraction of time that the system operates within the pressure-temperature conditions specified for high energy fluid systems is less than 2 percent of the time that the system operates as a moderate energy fluid system (e.g . , systems such as the reactor decay heat renoval systems qualify as moderate energy fluid syntams; however, systems such as auxiliary feedwater systems operated during reactor startup, hot standby, or shutdown qualify as high energy fluid systems) .

3.6.2.3 Design Basis Break / Crack 7Ypes and Orientation 3.6.2.3.1 Circumferential Pipe Breaks The following circumferential breaks are postulated in high energy fluid system piping at the locations specified in Sections 3.6.2.1 and 3.6.2.2:

1. Circumferential breaks are postulated in fluid system piping and branch runs exceeding a nominal nipe size of 1 inch, except where the maximum stress -ange or usage 3.6-4D Amendment 24 4/23/76

'<? 042

SWESSAR-P1 exceeds the limits specified for break postulation, and the maximum stress range in the circumferential direction is at least 1.5 times that in the axial direction, only a longitudinal break need be postulated.

2. Where break locations are selected at pipe fittings without the benefit of stress calculations, breaks are postulated at each pipe-to-fitting valve or welded attachment. If detailed stress analyses (e .g . , finite element analyses) or tests are performed, the maximum stressed location in the fitting may be selected instead of the pipe--to-fitting weld.

3. Circumferential breaks are assumed to result in pipe severance and separation amounting to a one-diameter lateral displacement of the ruptured piping sections unless physically limited by piping restraints, structural members, or piping stiffness as may be demonstrated by inelastic analysis (e .g . , a plastic hinge in the piping is not developed under loading) .

4. The dynamic force of the jet discharge at the break location is based on effective cross-sectional flow area of the pipe and on a calculated fluid pressure as modified by an analytically or experimentally determined thrust coefficient. Limited pipe displacement at the break location, line restrictions, flow limiters, positive pump-controlled flow, and the absence of enoray reservoirs are taken into account, as applicable, in the reduction of jet discharge.

5. Pipe whipping is assumed to occur in the plane defined by the piping geometry and configuration, and to cause pipe movement in the direction of the jet reaction.

3.6.2.3.2 Longitudinal Pipe Breaks The following longitudinal breaks are postulated in high energy fluid system piping at the locations of each circumferential break specified in Section 3.6.2.3.A:

1. Longitudinal breaks in fluid system piping and branch runs are postulated in nominal pipe sizes 4 inch and larger, except where the maximum stress range or usage exceeds the limits specified for break postulation and the maximum stress range in the axial direction is at least 1.5 times that in the circumferential direction ,

only a circumferential break need be postulated.

2. Longitudinal breaks are not postulated at: a) terminal ends if the piping at the terminal ends contains no longitudinal pipe welds (if longitudinal welds aru used, the requirements of (1) above apply). b) At

3. 6-4 E Amendment 24 662 043 "/23 26

SWESSAR-P1 intermediate locations where the criterion for a minimum number of break locations must be specified.

3. Iongitudinal breaks are assumed to result in an axial split without pipe severance. Splits are oriented (but not concurrently) at two diametrically opposed points on the piping circumference such that a je ?. reaction causing out-of plane bending of -he piping conidquration results. Alternatively, a single split may be ussumed at the section of highest stress as determined by detailed stress analysis (e . g . , finite element analysis).

4. The dynamic force of the fluid jet discharge is based on a circular or elliptical (2D x 1/2D) break area equal to the effe(tive cross-sectional flow area of the pipe at the break location and on a calculated fluid pressure modified by an analytically or experimentally determined thrust coefficient as determined for a circumferential break at the same location. Line restrictions, flow limiters, positive pump-controlled flow, and the absence of energy reservoirs shall be taken into account, as applicable, in the reduction of jet discharge.

5. Piping movement is assumed to occur in the direction of the jet reaction unless limited by structural members, piping restraints, or piping stiffness as demonstrated by inelastic limit analysis.

3.6.2.3.3 Through-Wall Leakage Cracks (outside of containment only)

The following through wall leakage cracks are postulated in moderate-energy fluid system piping at the specified locations:

1. Cracks are postulated in moderate energy fluid system piping and branch runs exceeding a nominal pipe size of 1 inch.

2. 'luid flow from a crack is based on a circular opening of area equal to that of a rectangle one-half pipe diameter in length and one-half pipe wall thickness in width.

3. The flow from the crack is assumed to result in an environment that wets all unprotected components within the compartment, with consequent flooding in the compartment and communicating compartments. Flooding effects are determined on the basis of a conservatively estimated time period required to effect corrective actions.

b05 O 'r i

3. 6-4 F Amendment 24 4/23/76

SWESSAR-P1 3.6.3 Design Loading Combinations For the purposes of design, unless otherwise stated, the pipe break event is considered a faulted condition and the pipe whip restraint or barrier, and the structure to which it is attached, are designed accordingly. Analyses of components important to sate shutdown comply with the requirements of Appendix F, ASME 3.6-4G Amendment 24

/. Otc 4/23/76 OOJ Uit J

SwESSAR-P1 III. The allowable strain limits for pipe whip restraints and steel barriers are given in Section 3.6.4.

3.6.4 Dynamic Analysis A. Pipe Whip Dynamic Analysis Criteria

1. An analysis of the pipe run or branch is performed for each longitudinal and circumferential postulated rupture at the design basis break locations.

2. The loading condition of a pipe run or branch prior to postulated rupture in terms of internal pressure, temperature, and stress state are those conditions associated with reactor operating condition (normal and upset) .

3. For a circumferential rupture, pipe whip dynamic analysis is only performed for that end (or ends) of the pipe or branch which is connected to a contained fluid energy reservoir having a sufficient capacity to develop a jet stream. 711 cases are evaluated to determine whether a jet stream can develop.

4. Dynamic ar.alysis methGds used for calculating the piping or piping / restraint system response to the jet thrust developed following postulated rupture adequately account for the effects oft

a. Mass inertia and stiffness properties of the system,

b. Impact and rebound (if any) effects as permitted by gaps between piping and restraint,

c. Elastic and inelastic deformation of piping and/or restraint, and

d. Limiting boundary conditions.

5. The allowable design strain limit for the restraint or steel barrier does not exceed 0.5 ultimate uniform strain of the materials of the restraints. The method of dynamic analyris used is capable of determining the inelastic behavior of piping restraint system response within these design limits.

6. A 10 percent increase of minimum specified design yield strength (Sy) may be used in the analysis to account for strain rate effects.

7. Dynamic analysis methods and procedures consist of:

6-"

gO 0 /,

SWESSAR-P1

a. A representative mathematical model of the piping system or piping / restraint system,

b. The analytical method of solution selected,

c. Solutions for the most severe response among the design basis breaks analyzed, and

d. Solutions with demonstrable accuracy or justifiable conservatiam,

8. The extent of mathematical modeling and analysis is governed by the method of analysis selected among those specified by these criteria.

B. Dynamic Analysis for Restrained Piping Systems One of the following analysis models is used for analysis of each ASME Class 1, 2, and 3 piping system:

1. Lumped Parameter Analysis Model - Lumped mass points are interconnected by springs to take into account inertia and stiffness effects of the system, and time historie s of responses are computed by numerical integration to account for gaps and inelastic effects.

2. Energy Balance Analysis Model - Kinetic energy generated during the first quarter cycle movement of the ruptured pipe as imparted to the piping / restraint system through impact is converted into equivalent strain energy.

Deformations of the pipe and the restraint are compatible with the level of absorbed energy. For applications where pipe rebound may occur upon 1:rpact on the restraint, an additional amplification factor of 1.5 is used to establish the magnitude of the forcing function in order to determine the maximum reaction force of the restraint after the first quarter cycle of response. Amplification factors other than 1.5 may be used if justified by more detailed dynamic analysis.

3. Static Analysis Model - The jet thrust force is repre-sented by a conservatively amplified static loading, and the ruptured system is analyzed statically. The amplification factor used to establish the magnitude of the forcing function is based on selection of a conservative value as obtained by comparison with the factors derived from detailed dynamic analysis performed on comparable systems.

C. Dynamic Analysis for Unrestrained Pipe Whip

1. A lumped parameter analysis model as stated in B.1 is used or 3.6-6 6(( b

SNESSAR-P1

2. An energy balance analysis model as stated in B.2 is used. The energy absorbed by the pipe deformation may be deducted from the total energy imparted to the system.

3. The assumptions used to guide the mechanism of pipe movtsnent are justified to be conservative.

4. The results of analysis are expressed in terms compatible with the approach used for verifying the design adequacy of the impacted structure.

D. Flow Thrust Force for Pipe Whip Dynamics

1. Time-dependent forcing functions are applied to the mathematical model to represent the effects of wave thrust and blowdown thrust. These forces are applied at all locations where there is a change in flow direction.

The time functions consider the initial effects and all subsequent effects within the time span of interest.

2. A steady-state forcing function may be used when condi-tions as specified in D.5 below are met.. The function has a magnitude of T = KpA Where:

p = System pressure prior to pipe break A = Pipe break area K = Thrust coefficient K values are as follows:

a. 1.26 for saturated steam, water, and steam / water mixture

b. 2.00 for subcooled water nonflashing

3. A pulse rise time not exceeding 1 millisecond may be used for the initial pulse, unless longer crack propagation times or rupture opening times can be substantiated by experimental data or analytical theory.

4. The transient function is provided and justified. The shape of the transient function D 1 above is related to the capacity of the upstream energy reservoir, including source pressure, fluid enthalpy, and the capability of the reservoir to supply a high energy flow stream to the break area for a significant interval. The shape of the transient function may be modified by considering the break area and the system flow conditions, the piping U<. O1u 3.6-7

SWESSAR-P1 friction losses, the flow directional changes, and the application of flow limiting devices.

Following a postulated pipe rupture, support reaction forces, moments, and displacements significantly alter the dynamic response of piping runs with low length to diameter ratios. These, effects become negligible in g long, flexible, piping runs. Pipe rupture analyses adequately consider these support reactions at the terminal point (s) of the mathematical model. However, the analyses do not recognize support rea ction s from attachments not designed to pipe rupture loads such as deadweight hangers and seismic snubbers.

5. "ihe jet thrust force may be reprerented by a steady-state function, D.2 above, provided the following conditions are met:

a. The transient function, D.1 above, is monotonically diminishing.

b. The energy balance model or the static model is used in the analysis. In the former case, a step function amplified to the magnitude as indicated in C.2 is used.

c. The energy approach is used for the impact effects of the unrestrained piping.

E. Jet Impingement Load

1. Jet forces are represented by time-dependent forcing functions . The effects of the piping geometry, capacity of the upstream energy reservoir, source pressure, and fluid enthalpy are considered in these forcing functions .

2. The steady-state jet force has a magnitude of:

T. = K.pA I I Where:

p = System pressure prior to pipe break A = Pipe break area K = Jet coefficient The following K values are used whenever the reservoir pressure is constant, pipe friction is negligible, and there are no upstream flev restrictions:

a. 1.26 for saturated steam, saturated water, and steam / water mixture.

/<7

' O (- p A O}

G4 3.6-8 Amen dment 8 3/28/75

SWESSAR-P1

b. 2.00 for nonflashing subcooled water.

3. In calculating the jet impingement load on a distant object, the retarding action of the surrounding air along the jet path is neglected. The jet impingement pressure on a distant object is calculated by assuming the jet force diverges conica11y at a solid angle of 20 degrees from the break location.

4. For a target with its area normal to the jet, the impingement load is the product of the jet pressure and 3.6-8 A /- ' (j j h Amendment 8 UU/L 3/28/75

SWESSAR-P1 the intercepted jet area. If the target area is not 9 normal to the jet, the intercepted jet s tream is deflected rather than totally stopped. Therefore, a shape factor less than unity is included in calculating the impingement load.

3.6.5 Protective Measures Based upon the criteria presented in Section 3.6.1, piping systems are classified as high energy systems subje ct to break / crack postulation and/or moderate energy systems subject to crack postulation only. The specific systems which require break and/or crack postulation outside the containment are listed in Table 3.6-1. Pipe breaks only, being postulated in the specific systems inside the containment, are listed in Table 3.6-2.

It is a primary design objective to separate high energy lines outside the containment by remote location from safety related components. Many large high energy pipes in the turbine building, including main steam and feedwater lines, are remotely located from any safety related components, control room, auxiliary diesels, and auxiliary feed pumps. Where separation by distance is not feasible, enclosures are installed between the piping and the safety related components. If large, high pressure piping is involved, the enclosure (i.e. ., pipe enclosure 2 or equipment enclosure) is coupled with pipe whip restraints to allow for a reasonable wall thickness. if separation or enclosure is not practical, pipe whip restraints are installed.

The main steam and feedwater isolation valves are protected by a combination of valve house enclosure and pipe whip restraints to protect the isolation valves and containment penetration from pipe rupture outboard of the valve house enclosure.

Shields, barriers, or walls are installed, as required , to prevent jet impingement damage resulting from postulated pipe break. Consideration of the environmental effects of cracks in moderate energy lines may result in protective measures, such as optimized routing of piping, relocation of equipment, waterproofing of instrumentation, cabling, and installation of flood barriers.

Inside the containment, piping is routed for maximum separation from safety related components, within allowable spatial limit ations . Pipe whip restraints, impingement shields, and barrier walls are installed, as required, to prot ect the containment and provide sufficient protection to the safety related components to allow for safe shutdown. Protective measures are taken to ensure that damage cannot propagate to the unaffected reactor coolant loops and that a rupture wh ich does not constitute a loss of coolant cannot propagate into a LOCA.

Implementation of the criteria presented in Section 3.6.2 involves the systematic and detailed study of piping, me unical 3.6-9 Amendment 2 8/30/74 662 051

SWESSAR-P1 equipment, electrical, and composite drawings to establish the pipe whip and jet impingement targets. If the target is required to remain operable or intact for the specific rupture event, 2 analysis described in Section 3.6.4 is performed. If it is established that the combined loading on the target exceeds the faulted condition allowables, and separation or enclosure is not feasible, restraints and/or shields for protection of the targets are designed in accordance with Section 3.6.4.

The routing of pipe and the placement of components minimize the possibility of damage to the containment structure and equipment essential for a safe shutdown. The proximity of essential equipment, including the reactor coolant pressure boundary (RCPB) , is considered. The effects of pipe whip are minimized, when possible, by proper routing. Multiple loops are physically separated.,

In regions near postulated break locations where the careful layout of piping and components cannot offer adequate protection against the dynamic effects associated with a postulated pipe rupture, restraints are provided to prevent excessive pipe movement or special barriers are provided. No suppcmacntal protection is provided for components (e .g . , equipment supports) which are designed to accept pipe rupture forces.

Barriers are of two types: energy absorbing materials at d impact (or jet deflection) plates. Energy absorbing materials allow large deceleration distances during pipe impact and thus reduce the impact effect. Plates may distribute a pipe impact load or prevent the impingement of a jet against critical equipment.

In some locations, a noncritical component may be placed to shield a critical component if the noncritical component can act as a barrier.

Restraints to prevent excessive pipe movement are an alternative

o barriers. Judiciously located restraints are the best solution to pipe impact, since pipe displacements are minimized, and large kinetic energies are prevented, Plastic deformation is allowed since the restraint is designed for a single usage. The design af the restraint is based on a strain limit not to exceed 50 percent of the ultimate strain of the material.

Piping and other components of redundant engineered safety features are physically separated to prevent loss of function due to a single failure. Each independent loop of a redundant system enters the containment structure in a different region and is routed to ensure separation from other portions of the same system. In addition to maximizing the distance betwaen loops, the routing takes advantage of existing structural barriers (floors and walls) and internediate components to ensure effective separation. The relative position of these systems to possible sources of damage is also considered.

3.6-10 -

Amendment 2 s' porv,7 8/30/74

SWESSAR-P1 A description of the typical pipe whip restraints and a summary of the number and location of all restraints in each system will be provided when the information is available.

3.6.6 Interface Requirements This design meets the pipe rupture and whip protection, and separation requirements of the NSSS Vendors' Section 3.6; and individual system requirements listed in the following SWESSAR-P1 tables: 5.1-1, 5.5.7-1, 6.3-3, 9.3.4-1, and Section 7.8.

As stated in Section 3.6.2-1 for B-SAR-205, the Utility-Applicar.t shall provide justification for use of the 2.5 f t2 pump 35 discharga and hot leg breaks in the reactor cavity.

3.6-11 Amendment 35

//- , 10/6/71 O O t' ( ) b. ..j

SWESSAR-P1 TABLE 3.6-1 MODERATE AND HIGH ENERGY PITE-OLTPSIDE GNTAIhHENT Energy SYst m Moderate High locat ion Annulus Eldg l'

Engineered Safety Features Systm X (Saf ety Injection)

Fuel Pool Ccoling Systm X Fuel Bldg Fuel Pool Purification Systm X Fuel Bldg Reactor riant Service Water System X Ultimate Beat Sink, Yard, Annulus Bldg, Control Bldg Diesel Generator Bldg Reactor Plant Component Cooling Water Systm X Annulus Bldg Demineralized Water Makeup Systm X Yard, Turbine Bldg Annulus Bldg, Fuel Bldg, Solid Waste and Decontamination Bldg Primary Grade Water System X Solid Waste and Decontamination Bldg, Annulm Bldt , Reactor Plant Tank Area Chilled Water System X All Bldg Instrument and Service Air Systm X Turbine Bldg, Annulus Bldg O Fuel Bldg, Solid Waste and Decontamination Bldg O'

I N.' Aerated Vent and Drain Systm X Turbine Bldg, Annulus Bldg, Solid Waste ard Decmtamination Bldg C X Annulus Bldg

(;- Gaseous Vent and Drain Systm Boron Recovery System X X Annulus Bldg control Room Air Conditioning - Chilled X control Bldg Water Piping Reactor Plant Gas Supply Systm X Solid Waste and Decmtamination Bldq, Annulus Bldq Turbine Steam Systm X Turbine Bldg Main Steam System X Annulus Bldg, Turbine Bldg 1 of 2 Amendnent 7 2/28/75

SWESSAR-P1 TABLE 3.6-1 (CONT)

P.nercy System Pederate Iliqh Inca ' 1[on Turbine Bypass Systm X Turbine Bldg Condensate P211shing System X Turbine Bldg Condm sate System X X Turbine Bldg Feeduster System X Turbine Bldq, Annulus Bldg Steam Generator Blowdown Syst.se (C-E, W only) X X Annulus Bldg Turbine Plant Cmponent Cooling System X Turbine Bldg i Auxiliary Feedwater Syrtem X X 7:.aulus Bldg Turbine Plant Service Water System X Yard, Turbine Bldg Auxiliary Steam and Condensate System I X Auxiliary Boiler B1 4 , Yard, Annulus Bldg, Turbine Bldg, Solid Waste and Decontamination Bldg Radioactive Liquid Waste Systm X X Annulus Bldg, Solid Waste and Decontamination Bldg Radimetive Gaseous Waste System X X Annulus Bldg Radioactive Solid Waste System X Annulus Bldg, solid Waste and Decontaminatism Bldg Chmical and Volume Control System X X Annulus Bldg l'

O Q

FN C -,

y 2 of 2 Amendrnent 7 2/28/75

SWESSAR-P1 TABLE 3.6-2 HIGH ENERGY PIPE - INSIDE CONTAIMICNT System Engineered Safety Features System (Accunulators discharge piping) 7 Reactor Plant Gas Supply System Main Steam System Auxiliary Feedwater System Chemical and Volume Control System li Reactor Coolant System including RHR S SIS branch lines l7 Feedwater System Steam Generator Blowdown System (1 at applicable to BSW) 1 of 1 Amendment 7 2/28/75 rs? 056

SWESSAR-P1 TABLE 3.6-3 NSSS BREAK CRITERIA NSSS Vendor Criteria Westinghouse WCAP 8082-P-A, January 1975 Combustion - Engineering CENPD 168, July 1975 Babcock & Wilcox BAW-10127, December 1976 35 1 of 1 6 c,'z O S ~/ Amenamene 35 10/6/77 A

VOLUME 3 PRESSURIZED WATER REACTOR REFERENCE NUCLEAR POWER PLANT SAFETY ANALYSIS REPORT SWESSAR-P1 STONE & WEBSTER ENGINEERING CORPORATION P. O. BOX 232S BOSTON, MASSACHUSEITS 02107 Copynght 1974 by Stone & Webster Engineering Corporation. All matenal herein is the property of sa.d corporation under which all copy and other nghts have been reserved and no such rights have been granted to others. Stone & Webster Engineering Corporation willin allinstances take such steps as are necessary for the preservation ofits rights and the enforcement of applicable lam.

0 0(g o o u 66L

SWESSAR-P1 TABLE OF CONTENTS Section Volume CHAPTER 1 IffrRODUCTION AND GENERAL DESCRIPTION OF PLANT

1.1 INTRODUCTION

1 1.2 GENERAL PLANT DESCRIPTION 1 1.3 COMPARISON TABLES 2 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS 2 1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION 2 1.6 MATERIAL INCORPORATED BY REFERENCE 2 1.7 TERMINOIDGY AND FLOW DIAGRAM SYMBOLS 2 1.8 INTERFACE WITH NSSS VENDOR AND UTILITY-APPLICA NT SAR 2 CHAPTER 2 13 SITE CHARACTERISTICS 2.1 GEOGRAPHY AND DEMOGRAPHY 2 2.2 NEARBY INDUSTRIAL, TRANSPORTATION, AND MILITARY FACILITIES 2 2.3 METEOROIDGY 2 2.4 HYDROLOGIC EIGINEERIIU 2 2.5 GEOLOGY AND SEISMOLOGY 2 CHAPTER 3 DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS 3.2 CONFORMANCE WITH NRC GENERAL DESIGN CRITERIA 2 3.2 CLASSIFICATION OF STRUCTURES, SYSTEMS, AIO COMPONENTS 2 i Amendment 13

_, . . , 6/30/75 J L L. .; /

SWESSAR-P1 Section Volume CHAPTER 3 (CONP) 3.3 WIND AND TORNADO LOADINGS 2 3.4 WATER LEVEL (FLOOD) DESIGN 2 3.5 MISSILE PROTECTION 2 3.5 PROTECTION AGAINST DYNAMIC EFFECTS ASSOCIATED 2 WITH THE POSTULATED RUPTURE OF PIPING 3.7 SEISMIC DESIGN 3 3.8 DESIGN OF CATEGORY I STRUCTURES 3 3.9 MECHANICAL SYSTEMS AND COMPONENTS 3 3.10 SEISMIC DESIGN OF CATEGORY I INSTRUMENTATION 3 AND ELECTRICAL EQUIPMENT 3.11 ENVIRONMENTAL DESIGN OF MECHANICAL AND 3 ELECTRICAL EQUIPMENT APPENDIX 3A CONFORMANCE WITH NRC REGULATORY GUIDES 3A.1 DIVISION I REGULATORY GUIDES, POWER REACTORS 3 3A.2 OTHER DIVISION REGULA'IORY GUIDES 3 APPENDIX 3B 20l COMPUTER PROGRAMS FOR ANALYSIS OF 3 THE CONTAINMENT STRUCTURE CHAPTER 4 2D REACTOR 3

t. ; ,  !!OU e

ii Amendment 20 1/23/76

SWESSAR-P1 TABLE OF CONTENTS (CONT)

Section Volume CHAPTER 5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS

5.1 INTRODUCTION

3 5.2 INTEGRITY OF REACTOR COOLANT PRESSURE BOUNDARY 3 5.3 THERMAL HYDRAULIC SYSTEM DESIGN 3 5.4 REACTOR VESSEL AND APPURTENANCES 3 5.5 COMPONENT AND SUBSYSTEM DESIGN 3 CHAPTER 6 ENGINEERED SAFETY FEATURES 6.1 GENERAL 4 6.2 CONTAINMENT SYSTEMS 4 6.3 EMERGENCY CORE COJLI?C SYSTEM 4 6.4 HABITABILITY SYSTEMS 4 APPENDIX 6A DATA FOR DETERMINING THE IODINE REMOVAL EFFECTIVENESS FDR THE CONTAINMENT A'IVOSPHERE 6A.1 THE SPRAY DROP DISTRIBUTION AND CHARACTERISTIC SPRAY DROP DIAMETERS FOR THE SPRAY HEADERS 4 6A.2 THE SPRAY COVERAGE OF THE CONTAINMENT 4 ATMOSPHERE 6A.3 ANALYSIS OF SLCRS PERFORMANCE 4 APPENDIX 6B LOCTIC INTERFACE WITH NSSS SUPPLIED DATA 20 6B.1 POST - REFLOOD PERIOD 4 6B.2 LONG TERM MASS - ENERGY RELEASES 4 iii Amendment 20 hhk  !)b ] 1/23/76

SWESSAR-P1 TABLE OF CONTENTS (CONT)

Section Volume CHAPTER 7 INSTRUMENTATION AND CONTROLS

7.1 INTRODUCTION

4 7.2 REACTOR TRIP SYSTEM 4 7.3 ENGINEERED SAFETY FEATURES SYSTEM 4 7.4 SYSTEMS REQUIRED FOR SAFE SHUTDOWN 4 7.S SAFETY RELATED DISPLAY INSTRUMENTATION 4 7.6 ALL OTHER INSTRUMENTATION SYSTEMS REQUIRED FOR 4 SAFETY 7.7 CONTROL SYSTEMS NOT REQUIRED FOR SAFETY 4 7.8 INTERFACE REQUIREMENTS S CHAPTER 8 ELECTRIC POWER

8.1 INTRODUCTION

5 8.2 OFFSITE POWER SYSTEM S 8.3 ONSITE POWER SYSTEM S 8.4 INTERFACE DESIGN INFORMATION 5 20 8.5 ELECTRIC HEAT TRACING 5 CHAPTER 9 AUXILIARY sidTrMS 9.1 FUEL STORAGE AND HANDLING 6 9

iv Amendment 20 f ,

,,1/23/76

() i_ l' [) G l

SWESSAR-P1 TABLE OF CONTENTS (CONT)

Seetlon Volume CHAPTER 9 (CONT) 9.2 WATER SYSTEMS 6 9.3 PROCESS AUXILIARIES 6 9.4 AIR CONDITIONING, HEATING, COOLING, AND 6 VENTILATION SYSTEMS 9.5 OTHER AUXILIARY SYSTEMS 6 CHAPTER 10 STEAM AND POWER COINERSION SYSTEM 10.1

SUMMARY

DESCRIPTION 7 10.2 TURBINE-GENERATOR AND TURBINE STEAM SYSTEM 7 10.3 MAIN STEAM SYSTEM 7 10.4 OTHER FEATURES OF STEAM AND POWER CONVERSION 7 SYSTEM CHAPTER 11 RADIOACTIVE WASTE MANAGEMENT 11.1 SOURCE ITEMS 7 11.2 RADIOACTIVE LIQUID WASTE SYSTEM 7 11.3 RADIOACTIVE GASEOUS WASTE SYSTEM 8 11.4 PROCESS AND EFFLUENT RADIATION MONITORING 8 SYSTEM 11.5 RADIOACTIVE SOLID WASTE SYSTEM S 11.6 OFFSITE RADIOLOGICAL MONITORING PROGRAM 8 v Amendment 20 g, 1/23/76 002 g,J uG

SWESSAR-P1 TABLE OF CONTENTS (COtW)

Section Voltrne CHAPTER 12 RADIATION PROTECTION 12.1 SHIELDING 8 12.2 VENTILATION 8 12.3 HEALTH PHYSICS PROGRAM 8 12.4 RADIOACTIVE MATERIALS SAFETY (FSAP) 8 CHAPTER 13 CONDUCT OF OPERATIONS 13.1 ORGANIZATION STRUCTURE 9 13.2 TRAINING PROGRAM 9 13.3 EMERGENCY PLANNING 9 13.4 REVIEW AND AUDIT 9 13.5 PLANT PROCEDURES 9 13 6 PLANT RECORDS 9 13.7 INDUSTRIAL SECURITY 9 CHAPTER 14 INITIAL TESTS AND OPERATIONS 9 CHAPTER 15 ACCIDENT ANALYSIS 15.1 GENERAL 9 CHAPTER 16 TECHNICAL SPECIFICATIONS 16.1 DEFINITIONS 9 16.2 SAFETY LIMITS AND LIMITIWC SAFETY SYSTEM SETTINGS 9 vi Amendment 20

}/23/76 6bl {.) b 4

SWESSAR-P1 TABLE OF CONTENTS (CONT)

Section Volume CHAPTER 16 (CONT) 16.3 LIMITING CONDITIONS FOR OPERATION 9 16.4 SURVEILLANCE REQUIREMENTS 9 16.5 DESIGN FEATURES 9 16.6 ADMINISTRATIVE CONTROLS 9 CHAPTER 17 QUALITY ASSURANCE 17.1 QUALITY ASSURANCE DURING DESIGN AND 9 CONSTRUCTION 17.2 QUALITY ASSURANCE FOR STATION OPERATION 9 APPENDIX A ENCLOSURE BUILDING NITHOUT MIXING 9 APPENDIX B ENCLOSURE BUILDING WITH MIXING 9 vii / n,c Amendment 20 0 0.4'., UUJ 1/23/76

3. DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS (CONT'D) m O'

rx C

C' G'

/

3.7 t

\

Ruhm 9

1W S

e f,U1 U

L') (. /.

SWESSAP-P1 3.7 SEISMIC DESIGN 3.7.1 Seismic Input 3.7.1.1 Design Response Spectra The design response spectra are developed in accordance with the procedures given in Reference 8 and in accordance with Regulatory Guide 1.60, Rev. 1, dated December 1973 (Section 3A.1-1.60) . In

'chis method the critical parameter is the maximum expected ground accelerat. ion. Associated with the maximum ground acceleration, the maximum ground displacement is determined by prorating to standard vaJues. Detailed smooth spectra for any given value of damping are obtained by locating critical points and joining them by straight lines in a triparti te logarithmic plot. These critical points are obtained from given amplification factors and cutoff frequencies. The design value of the maximum ground acceleration is 0.3 g for the safe shutdown earthquake (SSE) and 0.15 g for the operating basis earthquake (OBE). Fig. 3.7.1-1 and 3.7.1-2 show the smooth design response spectra for horizontal and vertical earthquakes associated with the SSE.

Thus the design response spectra are not related to any site-dependent ground motion time history.

3.7.1.2 Design Pesponse Spectra Derivation An artificial earthquake is gener ated to give response spectra enveloping the design response spectra. The artificial earthquake is generated and checked by comparing spectral values at 250 periods between 0.02 and 5 sec. The periods are spaced logarithmically according to the rule:

Tn = T n-1 Tg 1/(N-l) j,

\ i/

1 =

1.0224 where T. = initial period = .02 see T l = final period = 5.0 sec f

N = total number of periods = 250 T = nth period in spectrum computation n

The acceleration time history yields ground response spectra at damping values of 1, 2, 5, 7, and 10 percent that envelop the smoothed site design ground response spectra for those damping values os shown in Fig. 3.7.1-3 through 3.7.1-12.

3.7-1 6? 0 u3 Amendment 11 5/30/75

SWESSAR-P1 Details of the artificial acceleration record and its development are presented in Section 3.7.2.6.

3 . 7 .1. 3 Critical Damping Values Structural Seismic analysis is performed using total syr, tem damping char-acterized by modal damping . The modal damping value is calculated as a ratio of the sum of the energy dissipated in each component element (based upon the assigned da cping ratio of each element) to the total available modal energy. Further discussion of modal damping appears in Sections 3.7.2.6 and 3.7.2.14.

In determining the modal damping ratios, component damping values consistent with the stress intensities are used. For example, ccxnponent damping for welded structural steel is assigned a value of 2 percent for OBE and 4 percent for SSE. The subgrade canponent damping ratios are taken as 10 percent of critical for translation and rotation for both the OBE and SSE. The damping ratio in any mode, however, is limited to a maximum value of 10 7

percent. The soil internal damping ratio used in finite element soil structure analysis is described in Section 3.7.2.5.

The damping ratios as indicated in Regulatory Guide 1.61,

" Damping Values for Seismic Design of Nuclear Power Plants,"

issued October 19 73, are given in Table 3.7.1-1 for various compon ents. The damping ratios of Table 3.7.1-1 are used in the SWESSAR design.

Equipmant and Components Equipment and components specified or designed for seimaic quali-fication are assumed to exhibit the following percentage of critical damping values:

Operating basis earthquake 2.0 percent Safe shutdown earthquake 3.0 percent Where justified by detailed study (such as testing of strese levels) , hicher d amping values are utilized. These higher damping values and their technical justifications, when utilized, will be submitted for acceptance review in the Utility-Applicant 's SAR. The percent of critical damping incorporated into equipment and component analysis is shown in Table 3.7.1-1.

These damping ratios may be modified by those values developed and justified by the NSSS Vendor for the reactor coolant loop.

. "I ()

. O r.

3.7-2 Amendment 7 2/28/75

SWESSAR-P1 3.7.1.4 Bases for Site-Dependent Analysis Bases for site dependent analysis, if required, will be included in the Utility-Applicant's SAR.

3.7.1.5 Soil-Supported Category I Structures A list of all soil-supported Category I structures will be pro-vided on a site by site basis in the Utility-Applic at's SAR.

3.7.1.6 Soil-Structure Interaction All Category I structures are analyzed by the methods indicated in Table 3.7.1-2 to determine the effects of subgrade structure interaction.

At a soil site finite element representation of the subgrade with

& solution in the frequency domain, as described in Section 3.7.2.5, is used to evaluate the effects of soil-structure interaction. The finite element solution accounts for structural embedment, foundation medium depth and layering (local soil strata) , water table level, density, and other parameters which may af fe ct the system seismic response. The input parameters are determined from actual site subgrade characteristics and verification of plant design adequacy presented in the Utility - Applicant's SAR. The finite element approach is also used to evaluate structure-to-structure interaction ef fects.

3.7.2 Seismic System Analysis 3.7.2.1 Seismic Analysis Methods 3.7.2.1.1 Seismic Analysis of Structures The stresses in Seismic Category I structures associated with the response to the simultaneous application of horizontal and vertical earthquake ground motions are deternuned by the modal analysis response spectra method using the dynamic analysis Cdpabilities of computer program STRUDL II. A description of STRUDL II is given in Appendix 3B. The derivation of the spectra for the SSE and the OBE is discussed in Section 3.7.1. The allowable stresses and design loading conditions are stated in Sections 3.8.1, 3.8.3, 3.8.4, and 3.8.5.

Major Seismic Category I structures that are considered in conjunction with foundation media in forming a soil structure interaction model are defined as " seismic systems." Other Seismic Category I structures, systems and components that are not designated as " seismic systems' are considered as " seismic subsystems."

In general, the absolute frequencies of systems and subsystems have negligible effect on the error due to decoupling. Thus, the 3.7-3 Amendment 16 n,-

U,U 8/29/75 6ed

SWESSAR-P1 mass ratio, Rm, and the frequency ratio, R7, govern the results where R and Rg are defined as:

Total mass of the supported subsystem Rg ,= Mass that supports the subsystem Fundamental frequency of the supported subsystem R = Frequency of the dominant support motion g

The following criteria are used.

1. If r <0.01, decoupling is acceptable for any R .

15

2. If 0.01 s R m 5 0.1, decoupling is acceptable if 0.8 2 R 2 1.25.

3. If R, >0.1, an approximate model of the subsystem is included u the primary system model.

If the subsystem is comparatively rigid and also rigidly connected with the primary system, only the mass of the subsystem is included at the support point in the primary system model. On the other hand, in case of a subsystem supported by very flexible connections, e .g . , pipe supported by hangers, the system is not included in the primary model. In most cases the equipment and components, which come under the definition of subsystems, are analyzed as a decoupled system from the primary structure and the seismic input for the former is obtained by the analysis of the latter. One important exception to this procedure is the reactor coolant system, which is considered a subsystem but is analyzed using a coupled model of the reactor coolant system and primary structure with ground motion as the seismic input.

The containment structure is modeled by a generalized system of lumped masses, each with six degrees of freedom, connected by flexible members. The lumped mass model of the structure, as shown in Fig. 3.7.2-1, is supported on a deformable subgrade represented by equivalent springs whose stif fness is determined as indicated in Section 3.7.1.6. The model is constructed so that it properly represents the free vibration of a cantilevered structure in shear and flexure. Mass locations are selected to give the structural and equipment response at critical locations.

Masses Mg through Mg in Fig. 3.7.2-1 represent the dome and seven cylindrical sections of the containment shell. M 2 consists of the mat and small sections of the cubicle wall, containment shell, and annulus building wall. The internal structure, consisting of equipment, the primary shield wall, and reactor cavity walls interconnected by floors and radial walls, is modeled by masses M g through M The annulus building is 21 represented by masses M3 through M7 The member between joints 15 and 21 represents the horizontal truss connecting the crane girder to the containment shell.

7 s c, ' i; i 3.7 f4 Amendment 15 8/8/75

SWESSAR-P1 Where the subgrade springs are computed from consideration of a rigid body on elastic half-space, they are computed as follows (Ref 1) :

Translational = 32 (1 -U) GR 7-8U Rocking = 8 GRa 3 (1-U)

Vertical , 4 GR 1-U Torsion = 16 GR3 3

where:

G = Shear modulus of subgrade R = Radius of foundation mat U = Poisson's ratio of the subgrade The dynamic models of the other Seismic Category I structures consist of generalized systems of lumped masses connected by flexible members and coupled to the subgrade by springs derived fran the soil stiffness. The masses consist of floors, tributary walls and columns, equipment, and piping. Horizontal, vertical, rocking, and torsional spring constants are included to represent the subgrade. Depending on the type of subgrade, these constants are determined fran consideration of the theory of elasticity relating to rigid plates on an elastic half space or from a finite element analysis of subgrade. The type of analysis is performed for different types of subgrades as described in Section 3.7.1.6.

The floors are treated as rigid plates or diaphragms that transfer earthquake inertia forces to frames and diaphragm walls, which in turn transfer the loads to the foundation mat and subgrade. Beam theory, combining the effects of shear, flexure, torsion, and axial deformation, is used to establish the stiffness characteristics of the frame-wall systems.

The criteria used to determine an adequate number of masses in dynamic modeling of all Category I structures are, in general, dictated by the number of floor elevations and the roof elevation of a structure. It is at these points that masses are lumped and include half the walls above and below the floor, the floor itself, and major pieces of equipment resting on the floor or supported from the walls. This is done because this mass distribution closely approximates the real mass distribution of the actual stru cture . Additional mass points may also be included where mass distribution dictates this be done.

c ,q 3.7-4A bb/ i2' i Amendment 15 8/8/75

SWESSAR-P1 The containment wall of the containment structure does not in general have attached floors. For this shell structure, a study has been performed to determine how many masses are necessary to determine its structural response and provide sufficient accuracy of its natural frequencies for the development of amplified response spectra. This was done by comparing the results of the response spectra method for a 3 mass, 5 mass, 10 mass, and 15 mass fixed cylindrical containment under a base excitation of

.3g. The structure is shown in Fig. 3.7.2-5 and the models in Fig. 3.7.2-6. The results compared are total base shear and overturning shown in Table 3.7.2-1 and horizontal structural frequencies shown in Table 3.7.2-2. In Table 3.7.2-1, the difference between 3-mass and 15-mass models is less than 3 percent for base shear and less than 2 percent for base overturning. On the basis of Table 3.7.2-1, sufficient accuracy in structural response can be obtained with 3 masses. In Table 3.7.2-2, the difference in natural frequencies for the first two modes is less than 1 percent for all models. For the 5, 10, and 15 mass mdels , this difference is less than 8.5 percent through the first four modes. For the 10 and 15 mass models, this difference is less than 7 percent for the first eight modes. A reasonable criterion for establishing the minimum number of masses is one which provides sufficient accuracy for natural frequencies below 33 cps, since above this value there is little amplification of earthquake excitation; that is, equipment stiffer than 33 cps responds at the mass response, and larger errors in structural frequencies can be tolerated. Based on the results of Tables 3.7.2-1 and 3.7.2-2, the containment wall of the reference plant has been modeled by the use of 10 masses.

Inclusion of subgrade deformations in the test problem would not alter the conclusion that 10 masses are adequate to represent the containment shell. Inclusion of subgrade springs in the model increases the total number of modes in the problem. If the subgrade is relatively stiff, the additional modes will have high frequencies and they will not alter the results. The cantilever will essentially behave as a fixed base structure. If the subgrade is relatively flexible, the additional mdes will have low frequencies and these modes will essentially represent rigid body translation and rotation, in which case the number of masses chosen to represent the cantilever will not af fect the response significantly.

The equations of motion for a multi-degree lumped mass system is written in matrix notation as follows:

~"

[M] {x) + [c] [x} + [K] [x] = -kig (t)[ M ][e}

Where

[M] = The mass matrix of the system

[C] = The system damping matrix - , ,

[K] = The system stif fness matrix bO( bi j W 3.7-4B Amendment 15 8/8/75

SWESSAR-P1 (e) = Excitation vector consisting of 'O' and 81.' The 80's are associated with the degrees of freedom that are not parallel to the direction of excitation Ug (t) = Scalar time function of ground acceleration The above system of equations are uncoupled by using undamped eigenvalues and eigenvectors of the system as follows:

Let (4) = eigenvectors of the undamped system normalized with respect to the mass mar.rix such that

[w2 ] = diagonal matrix of eigenvalues of the undamped system Introducing a linear transformation {n} =[c]{xl and premulti-plying by the transpose of [c] matrix the resulting equations of motion are

[7]T[M][C]{n} + [C]T[c][?]{h} + [$lT [g][4]{n} = -Ug (t) [? ] T [g] { e }

Utilizing orthogonality property of eigenvectors,

[c]T[M][c]= I = identity matrix

[c]T[gj[e]=[g 2 Fdiagonal matrix of eigenvalues f', '?i

('i b L U! t 3.7-4C Amendment 15 8/8/75

SWESSAR-P1

[o]T[C][C] is in general not diagonal but the off-diagonal con-tributions are ignored. This does not cause any significant loss of accuracy since the assumed' component damping values given in Section 3.7.1.3 are small. [o]T[C][c] is replaced by a diagonal damping matrix as follows:

[28.] = diagonal damping matrix Where d = the modal damping ratio w = circular frequency of the mode Further discussion on the modal damping appears in Section 3.7.2.14. Hence the uncoupled equations of motion in the generalized coordinate system can be written as

( fi } + [25u]{f} + [w2]{n} = -Og ( t) { F }

i Where iTI =

T :x:te ? = a column vector and each component is called the modal participation factor A time domain solution is made by solving the decoupled equations in generalized coordinate by treating the time history of ground motion as a piece-wise linear function. An explicit solution scheme for piece-wise linear time functions is available where complete solution is obtained by successive application of initial conditions starting from 7=n = 0 at t = 0 (Ref erence 7) .

The solution in the original coordinate system is given by x =

, ti. =  : ' ff } and tR} = .:j{H}

hN I 3.7-5 Amendment 7 2/28/75

SWESSAR-P1 In the spectrum method of analysis, only the maximum values of acceleration, velocity, and displacement are known and these are identified as spectral acceleration, velocity, and displacement, respectively, corresponding to the model frequency and damping.

Paying attention to the ith degree of freedom then, x ir " # ir N ir 8

(xir) max *#r i r dr 1

9 = modal displacemer ts xir- displacement for the ith degree of freedom (dof) at the rth mode at time 't.'

c"ir component of eigenvector for the ith dof and rth mode r,

• participation factor for the rth mode s-dr spectral displacement for the rth mode and damping Since only the modal maximums are known, a method is required to combine these model responses. If the maximum responses are considered independent and equally likely to occur, the method to combine them is given by

• F 6 Xi r=i \

ir r dr ) 2 7f where xi = probable maximum response of the ith decree of freedom Another method of combining modal responses is indicated in Reference 9. This method is described in the following equation:

-n n n g

A i IXir) +

2[ [ (lx ik ) Il*1E !) (1 + E 3 -l

_r=1 k=1 E=k+1 ( _

where:

Xi r , xika Xi1 are responses from the rth, kth, and the fth modes respectively for t.ie ith degree of freecom.

Il --

3.7-6 Amendment 7 2/28/75

SWESSAR-P1 x

ir " # ir r dr8 Typically =

w '

-wg

' k l " '6' gug + Vg Wz O

k k+w kd ,

=

w[ wk b1 ~ IOk}

t = duration of motion, assumed equal d

to 15 seconds Taking absolute values of x. and x. ensures that these respmses are in phase. In u the sped:,ial case of repeated frequencies, the above equation results in adding the responses absolutely. This method yields results that are in good agreement with time history results when modes are closely spaced. Closely spaced modes are defined as those that lie within 110 percent of each other. This second method is used to combine the rrodal responses in all structural seismic analysis.

The number of modes to be used in combining modal responses for the ith degree of freedom, in an n degree-of-freedom system, is determined by or ering the absolute value of the participation factortimestheigt component of the eigenvector for the n modes in decreasing numerical order. A summation of these values is then initiated starting with the largest value. Each succeeding I'r e i r value is checked against the present sum. When the rth value is less than 1 percent of the present sum, the r th through n modes for that particular degree of freedom are considered to be noncontributing.

In cases where the subgrade stiffness matrix is obtained in a frequency dependent form, as from finite element soil structure analysis, the equations of motion of the multi-degree lumped mass system are also solved in the frequency domain using the ccxnputer program FRIDAY described in Appendix 3B. I The spatial components from seismic response analysis are com-bined in accordance with Regulatory Guide 1.92 (Section 3A1-1.92). All Seismic Category I btructures are analyzed from the three orthogonal component motions (two horizontal and one vertical) of the prescribed earthquake. In cases where response spectra analysis is performed, the representative maximum value .

of a particular response of interest for design (e .g. , stress, strain, moment, shear, or displacement) of a given element of a structure, rfstem, or component subjected to a simultaneous action of the three components of the earthquake, is obtained by taking the square root of the sum of the squares (SRSS) of corresponding representative maximum values of the spectrum response to each of the three components calculated independently.

662 077 3.7-7 Amendment 7 2/28/75

SWESSAR-P1 In cases where time history dynamic analysis is used, three statistically independent (maximum correlation factor of 0.2) orthogonal ground accelerations (two horizcntal and one vertical) of the prescribed earthquake are input simultaneously. The correlation factors (Ref 11) for two horizontal earthquekes and for a vertical and horizontal earthquake are shown in Fig. 3.7.2-7 and 3.7.2-8. The particular response in a direction of interest is obtained by algebraic summation of the response in that direction at each time intervel due to each of the three ground accelerations.

A list of specific structural systems designated as Seismic Category I is provided in Table 3.2.5-1.

3.7.2.1.2 Seismic Analysis Methods (Components)

Seismic Category I equipment is documented for seismic adequacy.

Depending upon equipment location, the basic source of seismic design data is either the ground response spectra or the amplified response spectra, derived through a dynamic analysis of the relevant structure as described in Sections 3.7.2.6 and 3.7.2.8.

Three principal methods of documenting adequacy for Seismic Category I components are:

1. Static analysis

2. Dynamic analysis

3. Testing Table 3.7.2-3 provides a list of equipment types and specifically employed metho ds of seismic qualification. Other types of 16 equipment are qualified by the analysis and testing methods outlined herein or by combinations of these methods as applicable.

Static Analysis Static analysis is utilized for equipment and components that can be characterized as relatively simple structures. This type analysis involves the multiplication of the equipment or component total weight by the specified seismic acceleration (direction dependent loading) to produce forces that are applied at the center of gravity in the horizontal and vertical directiona. A stress analysis of equipment components, such as feet, holddown bolts, and other structural members, is performed to determine their adequacy.

In the specification of equipment for static analysis, two or more sets of acceleration data are provided; the choice of which set to use is dependent upon the equipnent's fundamental natural f re quency . The relevant respoose curves are reviewed to determine a " cutoff f requency" which bounds the rigid range f rom the resonance range ot the response curves. Equiament and 3.7-8 Amendment 16 8/29/75 L' b

SWESSAR-P1 components having fundamental natural frequencies above the cutoff frequency are analyzed to rigid range response accelerations.

For components or equipment having a fundamental natural frequency below the cotoff frequency, analysis is based on response accelerations that are not less chan those indicated by the amplified response curves over the full frequency range of the component. If the fundamental mode of the component falls within any of the rasonant response peaks, and if the component cannot be characterized as a single degree of freedom system, the resonant response acceleration is increased by 30 percent as a justified factor for conservatism in order to account for all significant dynamic rt. odes under a resonant situation. (See Section 3.7.3.5.)

Each of the three defined directions of earthquake input (two horizontal and one vertical taken orthogonally) is evaluated separately. The calculated stress result of the three stress analyses is superimposed on a square root of the sum of the squares (SRSS) basis.

r p ,,

) IL YI 7.7-8A Amendment 16 8/29/75

SWESSAR-11 Dynamic Analysis A detailed dynamic analysis is performed when equipnent com-plexity or dynamic interaction precludes static analysis, or when static analysis is too conservative.

Modeling To describe fully the behavior of a component subjected to dynamic loads, infinite numbers of coordinates are required.

Since calculation at every point of a complex model is impractical, the analysis is simplified by a selection of a limited number of mass points. The " lumped mass" approach is employed in dynamic analysis. In the lumped approach, the main structure is divided into substructures and the masses of these substructures are concentrated at a number of discrete points.

The nature of these substructures and the stiffness properties of the corresponding modeling elements determine the minimum spacing of the mass points and the degrees of freedom to associate to each point. In accordance with spacing rem 2irements, the analyst can then choose, for the model, particular mass points that reflect predominant masses of components which are believed to give significant contribution to the total response.

in cases for which some dynamic degrees of freedom do not contribute to the total response, static or kinematic condensation is employed in the analysis.

Method of Analysis The normal mode approach is employed for seismic analysis of equipment and components. Natural frequencies, eigenvectors, participation factors, and modal member-end forces and moments of the undamped structure are calculatet The system of equations which describes the free vibrations of an n-degree of freedom, undamped structure is

[M] {X"] + [K] [X) =0 (1)

Where

[M] = mass matrix

[K] = stiffness matrix

[X) , [X"] = displacement, acceleration vectors The mode shapes and frequencies are solved in accordance with:

[K - w n2 Mi{0}n = 0

.s c r Qd b' {v -

3.7-9

SNESSAR-P1 Where wa= frequency of the nth mode

{c i n = mode shape vector for the nth mode Eigenvector-eigenvalue extraction routines, such as Householder-QR, Jacobi Reduction, and Inverse Iteration are used, depending upon the total number of dynamic degrees of freedom and the number of modes desired.

For each mode, the participation factor for the specific direction "i" is defined by

{c}T[M] {0}i ni 10f T M] t@}

L Where

{ c }T = transpose of mode shape vector for the nth mode

{Dl = earthquake direction vector referring to direction "i" The modal member-end forces and moments are determined by

{F } =

[Km] Ei}n where

[Km] = member stiffness matrix For each modal frequency, the corresponding response acceleration is determined for a given level of equipment damping (see Section 3.7.2.14) from the applicable response curve. Modes within the broadened response peak are assigned the peak resonant response value.

The maximum response for each mode is found by computing:

O

'{

3.7-10

SWESSAR-P1

{x"1 =f ni ni n l

{x'} =

- {x"}n

{x} =-

{x"}n '

{F}n = {F mn n

Where

{x"},, {x'}n' E* }n , {F}n are the modal acceleration, velocity, displacement, and member-end force and moment vectors, respectively. R is the spectrdl acceleration for the nth mode in the ith direction.

The basis for combination of modal responses is discussed in Section 3.7.3.4.

Each of the three defined directions of earthquake input (two horizontal and one vertical taken orthogonally) is evaluated separately. The calculated maximum stress result of the three stress analyses is superimposed on an SRSS basis.

Testing Equipment and components that are tested are seismically qualified in accordance with the f ollowing general instructions for earthquake testing. For tested equipment, these requirements conform with other applicable indastry standards (such as "IEEE Recommended Practices for Seismic Qualification of Class IE Equipment for Nuclear Power Generating Stations," STD-34 4-197 5, g see Section 3.10), or provide guidance for testing where no such standards are available. Equipment packages or components are shown to be seicmically adequate either by being tested individually, as part of a simulated structural section, or part of an assembled module or unit. In any case, the minimum acceptance criteria must include:

l' " '.

a( {v !', UUL 3.7-11 Amendment 16 8/29/75

SWESSAR-P1

a. No loss of safety function, or ability to function, bef ore, during, or af ter the proposed test

b. No structural / electrical failure (i .e . , connections and anchorages) that would compromise safety relat ed component integrity

c. No adverse or maloperation ' efore, during, or af ter the proposed test that could result in an improper safety action Equipment vendors and suppliers are required to formulate programs for qualif ying the equipment in accordance with the conditions specified in the earthquake requirements contained in the equipment specifications. The vendor must submit a sumnery of the proposed effort for review and approval.

The characteristics of the testing input at the equipment mcanting locations are defined by amplified response spectrum curves, "zero period" response curve levels, time history motions, power spectral density functions, or combinations of these as applicable.

The use of single and multifrequency input testing are accepted as methods of seismic qualification based upon the particular plant site, strL ture, and floor response characteris*ics.

Structures, particularly at lower elevations, exhibit a broad frequency range response similar to the ground motion during an earthquake. This broad range frequency motion is filiered at higher structural elevations and response becomes more sinusoidal in nature. Knowledge of the floor response characteristics of the structure and response characteristics of the equipment generally dictate the requirements for testing. Peliodic testing is applicable where periodic floor motion is indicated and, conversely, rar. dom input testing ir most applicable fo? broad frequency range input to components. Periodic testing can be used to envelop multiple peak floor responses- as well as single peak, providing sufficiently high forcing 2 3 used. For equipment exhibiting multiple response modes, single frequency input may be used providing the input has sufficient intensity to envelop the floor response spectra of the individual modes of the equipment.

The input motion f or testing normally f alls into four categories, single or multifrequency and single or multidirectionol.

Multif requency input applied biaxially is the prefe~ red method of qualification. In this program, input motion is applied to the vertical and one principal (or two ot 'hogonal) horizontal axes simultaneously, t7less it is demonstrated that the equipment response along the vertical direction is no? sensitive (coupled) to the vibratory motion along the horizontal direction and vice versa. The time phasing of the inputs in the vertical and horizontal directions should be phased incoherent. so that purely rectilinear resultant input is avoided. An alternative is to 3.7-12 Amendent 5 12/2/74 bO2 003

SWESSAR-P1 have vertical and horizontal inputs in phase, and then repeated with in, ts 180 deg out of phase. In addition, the test must be repeated, with the equipment rotated 90 deg horizontally. Single axis testing may be amployed as justified. Included in the scope of applicable single axis testing programs is equipment with single mode response characteristics.

The testing machine (fixture) setup is arranged so that the equipment tested is mounted to simulate to the extent possible the actual service mounting and so that there is no dynamic coupling to the test item. Equipment is tested in the operating condition wherever possible and functions are monitored and verified both during and after testing. It should be recognized 5 that a true operating environment is sometimes not obtainable for equipnent such as pumps etc.

The in situ application of vibratory devices to superimpose the seismic vibratory loadings on the complex active device for operability testing is acceptable when application is justifiable.

The test program may be based upon selectively testing a representative number of mechanical components according to type, load level, size, etc, on a prototype basis.

It is a general guideline that the ef fects of supports are reflected in the input to seismically qualified equipment and components. General stress limits are given in Section 3 9.3.1.

Specific component stress allowables are given in Tables 3.9.2-1, 3.9.2-2, and 3.9.2-3. Such limits, where jurisdictionally comparable, either conform to or are more conservative than those of ASME Section III, subsection NF -

"Corponent Support Structures."

General testing guidance criteria specified for equipment include the following:

Single Frecuency Testing Single f requency testing -ust produce a response curve based on test table input that envelogs the applicable amplified response curve. Testing is performed for as much of the range between 1 and 40 Hz as practicable or justified. Input for qualification should, an a minimum, equal the "zero period" response curve level.

Sinusoidal Inout

a. A frequency scan (two octaves per minute maximum) at a constant acceleration level is performed over the frequency range of interest. The objective of this test is to determine the natural frequencies and anmlification tactors of the tested equipment and its critical compon ento or 3.7-13 knendment S

(~(, ,

n 12/2/74

SWESSAR-P1 appurtenances and to ensure general seismic adequacy over the g full frequency range of interest. The acceleration inputs W-used are the maximum rigid range accelerations indica ted by the relevant response spectrum curves (damping independent) .

b. A " dwell test" of the equipment at its fundamental natural treguency is included at the acceleration values specified in GO above. Additionally, other frequencies are selected if amplification f actors of 2.0 or more are indicated. A minimum 20 sccond duration is considered acceptable for each dwell.

c. Other methods of sinusoidal testing may be employed as justified. Included are exploratory tests per 90 above, which employ a low acceleration level input to identify equipment response characteristics and to aid in selecting 5 requirements for further testing. A geometrically spaced constant f requency input may be employed for f urther testing.

Intervals of one-half octave or less are employed for this spacing.

Sine Beat Input A sine beat test may be performed in conjunction with a sine scan and is an alternative to the dwell portion of the progran outlined in (b) above. The sine beat test is performed at natural frequencies and bands of large amplification identified during the sine scan. The duration and peak amplitude of the beat for each particular test frequency are chosen to most nearly produce a magnitude of equipment response equivalent to that produced by the particular floor response spectrum at justc.fiable damping levels.

Current pra ctice indicates that a minimum of 10 cycles per beat should be used unless it can be shown thc_ a lower number of cycles is suf ficient to duplicate or exceed the response spectra for the equipment at the appropriate location. Five sine beats with a time delay between beats are commonly used.

An alternative qualification procram consists of applying a series of sine beats at geometrically spaced frequency intervals of one-half octave or less over the frequency range of interest.

The peak amplitude of the beat employed is, as a minimum, the maximum rigid range acceleration indicated by the relevant resconse spectrum curve.

5 Multifrequency Testing Multifrequency testing is applicable as a general qualification method. Input excitation in this category includes time history, random, power spectral density, complex waveshapes, and others as justified. For the type of input applied , the testing machine input must, as a minimum, equal the "zero period" acceleration of 3.7-14 Amen dment 5 f pn; 12/2/74 U br /- dUJ

SWESSAR-P1 the applicable response curve. A frequency range of 1 to 40 Hz is normally considered.

Time History Input A time history of the equipment support location or one based on a synthesized response curve may be used as testing machine input. A 15 to 30 second time history input is normally employed. The test table input must develop a response curve which envelops the relevant response spectrum curve when a synthe-ized record is used.

Pandom Motion Input Random input testing is performed so that the applicable response spectrum curve is enveloped by that produced by the table motion.

The input is controlled by one-third octave (or less) bandwidth filters over the frequency range of interest with a 15 second or greater test duration. Normally, random tests are performed to produce a response curve based on test machine input which envelops the relevant response spectrum curve. A special case random test may be performed when a power spectral density equivalent of the applicable response curve is specified.

Tests combining random input in conjunction with other waveforms may be employed as justified.

Complex Wave Test A complex wave test may be performed by subjecting the equipment to an input motion, generated by summing a group of decaying sinusoids steced at one-third octave or narrower frequency intervals over the frequency range of interest. Individual decay rate controls of from 0.5 to 10 percent are used. Response curves based on test table input must be shown to envelop the relevant response curve.

In addition to the single and multifrequency testing programs outlined, also given consideration are laboratory shock results, in-shipment shock data, or adequate historical dynamic adequacy data (i .e . , previous relevant test or environmental data). The method of test selected must demonstrate the adequacy of principal structural and functional capability of the equipment.

Use of Amplified Response Spectra (Umbrella)

Three alternative methods of using amplified response spectra are considered for use in multiple site or standard plant system dynamic analysis. These are termed:

1. " Umbrella" spectrum

2. Maximum modal response , ,

gg

3. Manmum site response (, [, f (; o O 3.7-14A Amen dment 5 12/2/74

SWESSAR-P1 These approaches provide a conservative basis for the design of equipment for use in multiple site or standard plants.

The umbrella spectrum is, ir fact, representative of a family of possible response spectra for various site characteristics. The present umbrella spectra ars based on dynamic analysis performed for structures founded on fou different subgrades. In addition, a fifth spectrum is develo,ed for a structure with a cracked containment founded on the softest subgrade. The umbrella is formed by connecting the primary resonant peaks for each condition as shown in Fig. 3.7.2-2. In addition, the primary pea k rerresenting the lowest period is broadened by -15 percent before connecting it with the floor acceleration value at zero period. The primary peak representing the highest period is also broadened by +25 percent to ensure additional conservativeness.

For complex systems, analysis to this umbrella spectrum may be too limiting as a qualification method for standard plant application. Where possible, its use as an analysis method is 5

conservative and no alternative qualification option need be considered.

The maximum modal response is a method which may be used when the conservatism in the umbrella spectrum method presents a difficult design situation. The maximum nodal response method is still a conservative approach but requires considerably more a nalysis effort. It is most effective when the modal density in the resonant lange is sparse. The response spectra are comprised of the individual spectra for each of the five conditions considered. In addition, each of the nominal peaks is br oadened by a percentage relating to the fundamental f requency of the system to be analyzed. Systems with fundamental frequencies below 4 Hz will have peaks broadened 125 percent. For systems with fundamental frequency greater than 4 Hz, a broadening of 115 percent is assumed. Sta rting with the first mode of the system to be analyzed, the amplified response spectrum curve, representing one of the five conditions, whose dominant response is closest to the first system mode, is shif ted so as to make the peak response coincide with the first system mode frequency. All other modes are assumed to be subresonant. The system is analyzed for the assumed response spectrum curve and stresses are dete rmined . Subsequent system modes are treated in a similar fashion (see Fig. 3. 7 . 2-3 ) and maximum system stresses are determined for each analysis. Suf ficient modes of the system are bb- bb 3.7-14B Amendment 5 12/2/74

SWESSAR-P1 analyzed to ensure that all parts of the system have been loaded conservatively.

Maximum site response (Fig . 3.7.2-4) is an approach more useful when high modal densities exist in the resonant range. Instead of centering maximum response on system modal frequencies, nominal resonant peaks are centered upon most likely site response frequencies. These curves are broadened by percentages associated to site conditions (soil 125 percent, rock 115 percent, or as further justified) . These site curves may or may not overlap.

The information provided is based on past experience such as Millstone 3 tor a rock site and Gulf States for a soil site. The observations are not documented but the peak broadenina as indicated (125 percent soil site, 115 percent rocksite) is cited in the relevant SARs.

The amount of spectrum peak broadening is associated with the subgrade properties of the site. The spectrum peaks below 4 Hz are bradened by t25 percent consistant with prior practice involving dynamic analysis performed for structures founded on soil sites (e .g . , Docket 50-4 58) . Similarly, spectrum peaks above 4 Hz are broadened by 115 percent consistent with prior practice involving dynamic analysis performed for structures founded on rock sites (e . g . , Docket 50-423). Based on the five analyses performed to cenerate the individual broadened spectrum 7 (see Fig. 3.7.2-2), 4 Hz is chosen as the controlling boundary.

The maxinum modal response method is most effective for systems with low modal density. Under such a situation, peak broadening by any percentage would have no effect since adjacent modes would not f all under broadened peaks. Two further options are possible given high modal density that meet the intent of the analysis philosophy. These would exist for a modal density situation in which more than one mode fell within the peak broadening percentaae criteria :(1 ) If each mode, recardless of density, is considered individually (one by one) as a separate analysis as described , no peak broadening is required. (2 ) If aroups of modes fall within a de fined broadened " nominal" amplified response spectrum (Fig . 3.7.2-3 Curve N ) analysis is done on a modal group basis (group by group). In this case all modes within the broadened peak are assigned resonant response. All other modes are than assigned appropriate subresonant responses.

Industry practice is currently standardizino to a broadening criteria of t10 percent. The proposed criteria reflect conservative past practice which will be revised as necessary to meet industry developments.

Peak spreading of 115 percent for rock sites and 125 percent for soil sites reflects the possibility of variations in structural and subarade propertie s . These values have been determined in 3.7-15 PG Amendment 7 OU 2/28/75

SWESSAR-P1 studies that show, for rock sites, variations of subgrade modulus and structural properties result in negligib! e ef fects in nominal resonant periods while variations of 11/3 in the soil shear modulus cause a spread of 125 percent in nominal resonant periods.

For as many site assumptions as required, analyses are performed and maximum loads or stresses for all parts of the system determined. To the extent that the assumed curves enveloped each site-dependent actual amplified response spectra , the method is correct and conservative. If actual site amplified response spectra are developed which are not enveloped by those used for system design, the analyses must be redone.

The methods described above are an effort to provide maximum latitude of analysis qualification methods for standardized plant application. Choice of one method versus another is based upon

• technical and economical considerations.

Applicability of Umbrella Spectra to Test Methods In order to meet seismic testing criteria which require enveloping of the amplified response spectra, it is important to note that the three methods outlined above are applicable in this context. The test input motion and amplitude may be chosen to either:

1. Envelop the envelcped (multi-site) amplified response O spectra,

2. Envelop a nominal curve at each device natura]

frequency, or

3. Envelop a nominal curve at each defined site frequency.

3.7.2.1.3 Seismic Analysis of Piping Systems Analyses of S & W supplied Seismic Category I and some Non-seismic Category piping (including all ASME Code Classes 1, 2, and 3 piping systems) are performed by the nodel analysis response spectra method. Some Nonseismic Category piping is seismically analyzed when its failure could affect a Seismic Category I system. Each piping system is idealized mathematically as an elastically coupled dynamic structural model in three-dimensional space. Inertial characteristics of the piping system &re simulated by discrete masses of piping components, including all eccentric masses, such as valves, and valve operators lunped at selected nodes. The stiffness matrix of the piping system is calculated based upon the elastic properties of the pipe to include the effects of bending, shear, exial and torsional deformations and a3 co changes in flexibility due to curved members. Model seismic responses at each node of the piping system, due to amplified response spectra excitation applied at its support points, are calculated by a computer procrum. The rodal analysis technique used in the progr am

!( " '

f')

3.7-16 Amendment 7 2/28/75

SWrSSAF-P1 computes the peak response quantities for each mode. Uornal mode, linear elastic, and small displacerent theory are incorporated in the computer prooram.

Ftructural response spectra, consistino of peak responses of a family of neismic loadings for the piping systems, are the ampliiled response spectra, obtained for discrete locations in the structure where the pipinq system is restrained. Damnino f act ors used for pipino systems of sizes equal to or less than 12 inches are 1 percent for the OBE and 2 percent for the er r. :

for pipe sizes arcater than 12 inches are 2 percent for unh and 3 percent for SSF. Detailed description of analytical procedures and denian criteria for foismic Cuteoory I pipino can be found in Section 3.7.3.d.

For plants of standard desion l oca t ed at different sit e s ,

pl.y s ica l models of the standard pipino systens and the displacement / force lounda ry conditions (defined by equipment layouts and pipe support location) will be the same for each st andard plant . The applicable amplified floor response spectra, urM as input seisnic excitation for these at andard pipi 7 synt ems , will he deve? oped for each dynamic structural model as de scribed in Section 3.7.2.6. Seismic responses of pipino systems (including seismic stresses of piping systems, reaction at pipe supports and equipment) will be computed independently tor each case. An envelope of seismic responses from the analyses of these cases will be used us the basis to analyze seismic stresses and transmitted reaction loads in a s ta nda rd plant. Since each standard pipino system in desioned to meet the requirements of enveloped maximum responses in all site location cases considered, this standard desion will meet the seismic analysis reouirement for any site.

3.7.2.2 Natural Frequencies and Pesponse Loads This inf ormation is developed durina detailed desian und will be I

included in EWESSAR for final desion approval (FDA).

3.7.2.3 Procedures Used to Lump ?' asses Tbc dynamic model or a Seismic Category I structure is constructed so as to obtain a satisf actory representation of the dynamic behavior of the actual structure. In ceneral, masses are lumped at floor levels and incorporate the masses of the floors, tributary walls and columns, cauipment, and pipino. The I

eouipnent and pipino contribute less than 10 percent of the total lumned mass; thus, the dynamic responses of the equipment and piping mounted on the floors do not have a significant effect on the overall dynamic response of the structure. The containment shell which coes not have any floors attached is divided into several segments as described in Section 3.7.2.1.1.

@ 0,U 66z 3.7- 16 A Amendment 7 2/28/75

SWESSAR-P1 3.7.2.4 Rocking and Translational Response Summary This information is developed during detailed design and will be included in SWESSAR for final design approval (FDA) .

3.7.2.5 Methods Used to Couple Soil with Seismic System Structures The method used to couple subgrade with structure, for different type subgrades, is described in Sec.: ion 3.7.1.6. When using the finite element method for soil sites, the soil structure interaction for axisymmetric structures is determined with the TRIAX program and for rectangular structures with the PAXLY 4 program.

The PLAXLY 4 and TRIAX programs utilize a finite element discretization of the foundation and the surrounding soil in the near field of the structure, and an expansion into eigenfunctions in the far field. The site amplification and soil-structure interaction are computed sinultaneously. The analysis is performed in the frequency domain and time histories of accelerations and displacements of the structure and the surrounding soil are obtained using a Fast Fourier Transformation technique. In the finite element analysis, appropriate nonlinear g stress-strain and damping relationships for soil are considered.

The selected moduli and damping ratios of soil are based on expected strain anplitudes. The relationships among soil moduli, damping ra tios , and strain levels are based on the relations in Ref 13. The maximum internal soil damping is 10 percent. In the analysis, the soil is treated as a viscoelastic material with hysteretic material damping. Programs TRIAX and PLAXLY 4 are described in Appendix 3b.

3.7.2.6 Development of F.ior Response Spectra Amplified response spectra are plots of maximum response vs natural period of single degree of freedom systems for a given damping at various locations on the structure subjected to dynamic loading. The loading on the equipment is detined by the amplified response spectra of the attachment points of the equipment.

The time history method of analysis is used to obtain the time-dependent floor response. The floor response is used to generate the floor response spe ctra . The equations of motion and their solutions in time domain from a modified decoupled set of equations are described in Section 3.7.2.1.1.

- rni bbd b'I 3.7-17 Amendment 16

~

8/29/75

SWESSAR-P1 The scalar time function that is used for time history analysis is an artificial time history with a total duration of 15 and 2 seconds each of rise and fall tLme whose ground response spectrum is forced to fit the s pecified site spe ctrum. An artificial accelerogram of ground excitation, which reproduces the frecuency content displayed either in a response spectrum or in a power s pectral density f unction, is simulated statistically by using a multi-P stochastic model as described in Ref erence 4. In this model, the earthquake motion is considered tc be a wide-band stationary process whose spectral density function, duration, and maximum acceleration are specified. The artificial motion is generated by matching the target or site spectrum for several specified percentages of critical damping at 250 oscillator periods distributed logarithmically from 0.02 (50 Hz) to 5 (0.2 Hz) seconds. For a detailed treatment of the modeling procedure, see References 5 and 6.

l The component floor response spectra are obtained from the simultaneous input of three statistically independent orthogonal -

I ground accelerations, described in Section 3.7.2.1.1, to the s tructure . The time history of acceleration in a particular direction at the elevation of interest is obtained by algebraic summation of the acceleration response, at that elevation on the 7 s tructure , at each time interval due to each of the three orthogonal ground motions. The amplified response spectra are then obtained from the above described time history of acceleration at the elevation of interest.

D}mamic analyses are perfo rmed for structures founded on subgrades with an envelope of shear modulus values as follows:

G = 6,000 psi G = 24,000 psi G = 300,000 psi G= 1,000,000 psi The 7tructural design is based on the envelope of the dynamic responses obtained for the four sites.

7 This will ensure the adequacy of the structure for any other site with shear modules falling within 6,000 and 1,000,000 psi.

For the containment structure it is further assumed that the containment concrete shell may be cracked due to internal pressure. This condition is combined with the softest subarade property. Thus five dynamic models - one each for each value of subgrade property, another with cracked concrete shell and softest subgrade - are used to obtain five different floor response spectra associated with each direction of excitation and each value of damping. The design floor response spectra are obtained from an envelope of the five floor spectra for each direction of excitation and dampina values.

6o2 c7: G 3.7-18 Amendment 7 2/28/75

SWESSAR-P1 3.7.2.7 Differential Seismic Movement of Interconnected Cbmpon en ts Criteria for differential seismic movements between intercon-nected piping and components are discussed in Section 3.7.3.6.

The loading combinations and stress limits are given in Tables 3.7.3-4, 3.7.3-5, 3.9.1-1, and 3.9.2-1.

Differential seismic movement between Category I structures and between Category I and Noncategory I structures does not cause structure interaction. Sufficient rattlespace is provided 7 between such structures to prevent movement f rom seismic events up to and including the SSE from causing structure interaction.

3.7.2.8 Effects of Variations on Floor Response Spectra In a concrete st ructure , the amount of cracking affects the stiffness and damping and, hence, the response of the system.

Cracking of the containment concrete shell is expected, due to internal pressure. The cracked concrete shell mod el is consistent with the lower bound values of the stif fness properties of the structural elements, while the uncracked shell model is consistent with the upper bound values. In order to obtain the lowest fundamental frequency, the cracked concrete shell is considered with the softest subgrade propert y. In Section 3.7.2.6 it is indicated that amplified response spectra are obtained from five different dynamic models for the containment structure. Seismic Category I structures other than the containment structure are not subjected to internal pressure, dnd Cracking is therefore assumed to be minimal. For other Seismic Catecory I structures, four dynamic models associated with the four subgrade properties are used and design mnplified response spectra are enveloped from those four separate spectra.

The enveloped spectra are further broadened by peak spreading +25 percent for the one extreme of G = 6,000 psi and -15 percent for the other extreme of G = 106 psi. This is done to account for the effects of expected variations in structural properties, damping, and other modeling parameters.

3.7.2.9 Use of Constant Vertical Load Factors 3.7.2.9.1 Structures, Equipment, und Components Constant vertical load factors are not utilized in component unolysis. (See Section 3 . 7 . 3 . 8 .1. )

3.7.2.9.2 Piping System The use of constant load factors based on the peaks of applicable amplified response spectra for simplified dynamic analysis of SSW 9 supplied 3.7.3.9.

small size Category I pipina can be found in Section 3.7-18 A 6h2 0 ') ) Amendment 7 2/28/75

SWESSAR-P1 3.7.2.10 Method Used to 7.ccount for Tbrsional Ef fects h

Seismic Category I structures may have natural torsional modes of vibration due to eccentricities between the centers of rigidity and centers of mass of the structural elements. The presence of eccentricities generates coupling between translational directions of motion, resulting in torsion. Thus, general three-dimensional models are z.e t up, followed by complete dynamic analyses as described previously for the containment structure.

The results of these ana.iyses include torsional ef fects.

3.7.2.11 Comparison of Responses To provide the basis for checking the seismic system analysis, the responses at variour. coordinates of the dynamic model for the containment structure are obtained from (1) modal analysis by the response spectrum method, and (2) modal analysis by time history m ethods . Similar analyses are performed for other Seismic Category I structures.

O

//- oe C D /_ -

3. 7- 18 B Amendment 7 2/28/75

SWESSAR-P1 3.7.2.12 Methods for Seismic Analysis of Dams Method of analysis of dams, if needed, is covered in the Utility

- Applicant's SAR.

3.7.2.13 Methods to Determine Category I Structure Overturning Moments The overturning moments induced by seismic loading are computed by spectrum method of analysis described in Section 3.7.2.1.1 for each direction of excitation separately. The applied overturning m' ment is computed from the SRSS of maximum responses from three 2

in'ividual directions of excitation. This nethod is described in F_ation 3.7.3.7. The vertical inertia force conbined by the above method is assumed to act upward, reducing the effective downward weight of the structure.

3.7.2.14 Analysis Procedure for Damping In order to use modal analysis, dampina values in different elements of a coupled system are accounted for by usina weighted modal damping values as described in Reference 3. According to this nethod, for a system vibrating in its ith node, the ith modal damping can be estinated by evaluating the ratio of the total energy dissipated due to the presence of danping in 2 different elements of the system to the total strain energy stored in the system in its ith mode. The equivalent damping is composed of a viscous damping term and a hysteretic dampir.g term.

Since the viscous damping tern is frequency dependent, i t. can result in unconservative damping values for higher modes.

Consequently, the wg/w term in equation 34 of Reference 3 is limited to a maxim ,

value of 1. The assumption of viscous damping, however, results in conservative values for lower modes.

The damping ratio in any mode is restricted to the liniting value of 10 percent.

3.7.3 Seismic subsystem Analysis 3.7.3.1 Determination of Number of Earthquake Cycles 3.7.3.1.1 Equipment ASME III (NB-3112. 3b) requires that the nunber of earthquake cycles to be used in the analysis of ASME Code Class 1 components be specified as part of the design rechanical loads. The following criteria are used for all components within the jurisdiction of this code:

1. A total of five OBEs and one SSE is assumed.

2 For conservative component design, structures are assumed to cycle (full sian reversal) 20 tines per earthquake.

3.7 19 Anendment 2

((} n 8/30/74

SWESSAR-P1 ASME III (NB-3112.3b) requires that the number of earthquake cycles to be used in the analysis of ASME Code Class 1 components be specified as part of the design mechanical loads. The following criteria are used for all components within the jurisdiction of this code:

1. A total of five OBEs and one SSE is assumed.

2. For conservative component design, structures are assumed to cycle (full sign reversal) 20 times per earthquake .

3. System and components classified as relatively rigid with respect to local structural response " ride" with the structure and are thus assigned 20 stress cycles per earthquake.

4. If the system and/or component is relatively flexible (fundamental frequency equal to or less than 50 percent of structural fundamental frequency), a 20 cycle per earthquake criterion pertains.

5. Systems and components with fundamental modes of vibration falling within the resonant range of the applicable response curves are assigned a value of 100 stress cycles per earthquake. This value a ccounts for additional significant stress cycles involved as the system or component decays in the manner of a damped vibration.

The Applicant's criteria assumes 20 cycles of maximum response for each postulated earthquake. This conservative criterion is bas ed on an examination of the cycles of intense motion which occur in strong, long duration earthquakes. The number of cycles of motion for a number of large ea rthauakes , in which the measured acceleration equaled or exceeded one--half of the maximum acceleration, is shown on Table 3.7.3-6. The number of complete cycles shown is taken as one-half the sum of positive and negative peaks satisfying the above requirements, which is slightly conservative because maximum negative and positive peaks do not occur sequentially. The number of complete cycles ranges from 3 to 12 with only the E-W component of El Centro exceeding 10 cycles.

The cyclic criteria are conservative with respect to AEC criteria and are proposed on an interim basis until the ASME Task Group on Dynamic Analysis recommendations are pranulgated.

3.7.3.1.2 Piping Systems A total of five OBEs and one SSE is assumed. All piping systems are considered to have significant dynamic response modes within the resonant ranges of the applicable amplified response spectra, 3.7-20 Amen dment 5 iu (!' [,

bU/ 12/2/74

SWESSAR-P1 and are designed for 100 stress cycles per seismic event in the analysis.

3.7.3.2 Basis for Selection of Forcing Frecuencies 3.7.3.2.1 Equipment kaplified response spectra (floor) developed for horizontal (two directions) and vertical direction earthquakes are the basic source of seismic design accelerations. As noted in Section 3.7.2.1.2, seismic accelerations are selected from the amplified response spectra based on natural frequency calculations for the component.

3.7.3.2.2 Piping Systems In the seismic design and multi-mass modal analysis of Seismic Category I piping systems (Section 3.7.3.6) , the practice of selecting forcing frequencies to preclude resonance is not used.

In the simplified analysis for small size Saismic Category I piping (Section 3.7.3.9) , the design approach is to preclude the significant normal mode frequencies of the piping system from the highest peak resonant frequency of the structure, as determined from the applicable amplified response spectra.

3.7.3.3 Root Mean Square Basis

'Ihe phrase, square root of the sum of the squares (SRSS) , is used to describe the method of combining modal responses when used herein.

3.7.3.4 Procedure for Combining Modal Responses The SRSS method employed in the ccxnbination of maximum seismic modal responses is supplemented by a scanning routine, in order to accommodate effects of any closely spaced modal responses The procedure to compute maximum seismic response (R) in response spectrum modal analysis is in precise conformance to Position C.2 g of Regulatory Guide 1.92 (Section 3A.1-1.92) as follows:

N number of natural frequencies of structural model (for structures, equipnent, and piping systems) are calculated (f ,

i=1,2,....N), which are grouped into clusters as following:

Cluster Frequencies Scanning Criterion p f; , f ******f P fp /f 1 <1.10, (fpy /f3 21.10) q f ,2 'f+ **f 4 (f 4 /f DH 21'10) g D

. . . . . . . . 2. .' . . . f f.......

4

/f *1<1~10',

D 1 /f w 21.10) r 1.10 (f r+H s f +1 , f r+2 7

..fN (f =N f ) fN /f r+1 s

<1.10

/. ( , r bb L '/

3.7-20A Amendment 14 7/18/75

SWESSAR-P1 Each modal contribution R = 1,2, . . . . N) associated with frequency f is calculated. ben (ithe resultant seismic response R at any node of the structural model is given by:

R= (lRi l+lR2 l+ +lEp l} +lEq l}

+ (l @ ll+l @ 2l+

15

+ ( . . . . . ) 2 + ( l rR +1l + l Rr+2 l '++ l EN! )

x

(' ; - r40

-l0 3.7-20B -Amendment 15 8/8/75

SWESSAR-P1 3.7.3.5 Significant Dyn=mic Resoonse Modes 3.7.3.5.1 Equipment Those components which are considered relatively simple or rigid are designed, by virtue of natural frequency calculations, to withstand the effects of amplified seismic acceleration values dependent upon frequency and amplitude ranges associated with the installation location and corresponding relevant amplified response spectrum. Analysis of components to the peak value of resonant response is considered conservative, since fundamental natural frequencies do not generally coincide with the frequency at resonance of the relevant response curve. Components havinc fundamental natural frequencies within the broadened response peak are designed to peak acceleration values, increased by a factor of 1.3 or as justified, to account for the contribution of all significant dynamic modes under a resonane condition.

Generally, the vibratory characteristics of the components qualified by " resonant static analysis" (relatively simple) is such that no possibility exists for adjacent or multiple modes to exist within the relatively narrow peak of a typical response s pectrum. The procedure for seismic analysis usina an amplified response spectra (unbrella) is given in Section 3.7.2.1.2.

The discussion which follows justifies the use of a factor of 1.3 as a conservative multiple to be applied to single or multiple d egree-o f-f reedom systems having fundamental frequencies within the broadened resonant response peak. Multiple supported or continuous type span components are not cart of this proof and, when such cases arise, a factor of 1.5 times the peak resonant response will be used.

Single Degree-Of-Freedom Systems Peak broadening is intended to reflect a range of uncertainty in the precise location of the resonant peak of the response curve and not to indicate that the multiole peak resonant response is possible within this broadened range. What is concluded is that there is a fairly equal chance that the peak of the curve (singular) would fall in the specified range and thus what, in fact, exists is a " family" of resonant response curves, each 9 3.7-21 Amendment 2 8/30/74 b02 0

SWESSAR-P1 having only one point of peak resonant response (see Fig. 3. 7. 3-1) . If more than one system or component mode of vibration falls within the broadened peak, one and only one mode (a presumed " worst case") can be presumed at an actual response peak value (see Fig. 3 . 7 . 3-2 ) . All other possible modes would realistically respond to lower values. Using the simple vibration theory and some simplifying assumptions, it is shown that a factor of 1.3 is conservative.

A simple damped oscillator responds with a transmissibility:

1 + (2dw/wg)

TR =

2  : -

1- + 2r,w/w g (w/wg) _ - .

The value of TR is sensitive to both the damping value (J) and the assumed placement and spacing of the number of modes around the peak considered.

For instance, if only one mode is considered, the value of TR is as defined above and is equal to the value of the peak of the amplified response spectrum curve (single degree-of-freedom system) .

As more modes are added around the nominal W n mode, TP s r s s increases. Thus the most conservative placement of assumed modes is with one mode on the " peak" and others centered around this peak.

Data are presented in Fig. 3.7.3-3 which prove that the factor 1.3 is conservative for all potential equipment applications.

The curves are developed for two " planes" representing five modes and nine modes assumed acting within the broadened resonant peak.

These numbers are intended to show an upper bound for general equipment application. Equipment damping values of 2.0 and 3.0 percent are used for static analysis. Higher damping is shown to indicate the trend and the conservatism of this me th od .

As further conservatism, all modes are considered equally participating. This is never the case in dynamic analysis. The higher frequencies of the component are given equal weight to the fundamental resonant freqiency and the modos are centered on the

" nominal" response curvt. If the fundamental frequency were placed on the peak of the nominal curve, the results would show even lower transmissibilities.

3.7-22

SWESSAR-P1 The factor 1.3 is applicable only for those components whose fundamental natural frequency f alls within the broadened response peak.

Peak broadening generally spans ranges of from 110 to 125 percent of a nominal amplified response spectra.

It has been shovn that, for the range of values associated with component and system static analysis, use of the 1.3 factor is conservative. In fact, for a predominant number of likely cases, a value far less than this could be justified on the basis of the data.

For example, a value of 1.1 could easily be justified for most components which present only a few significant modes of vibration within the broadened response peak. It is further emphasized that, in reaching these conclusions, the most conservative (and generally improbable) assumptions regarding 1) location of the nominal response curve and 2) the placement of response modes for the arbitrary component have been made.

Multi-Degree ot Freedom Systems As a conclusive supplement to the previous discussion, a study was performed utilizing rigorous dynamic analysis of models closely representative of typical components. The procedures and O results of the study are as follows.

computing the ratio of maximum This investigation consists of dynamic stress to maximum static stress; i.e., the factor denoted by K, for several model beams subject to a flat response and typical amplified response spectra. Since bending stress is dominant for frame / equipment constructions, the actual ratio employed equals K= max dynamic moment max static moment Both SRSS and absolute (ABS) moments are computed for comparison purposes, but conclusions are based solely on SRSS moments because they most closely represent actual dynamic stress.

Maximum static moment co-responds, in the case of the 1 g flat response, to a 1g static load. In the case of a typical amplified response, the maximum static load is based upon the following frequency relationships (ref er to Fig. 3.7.3-5) :

662 i01 3.7-23

SWESSAR-P1 fa sf,p g = g max (peak acceleration) fo >f, p g = acceleration at f o where fo = the fundamental frequency of the model beam and f p =

the peak frequency; i.e., the frequency at which the peak acceleration occurs.

The effect of " peak spreading" is investigated by using a flat response, thus giving all modes the same acceleration. This is equivalent to infinite peak spreading. The importance of the uncertainty in the location of the peak acceleration with respect to the fundamental mode of the model beams is examined by adjusting the fundamental frequency from well below to well above the peak resonant frequency of a typical response spectrum.

The model beams selected for this study are shown in Fig. 3.7.3-4. These beams are typical of the frames and equipment combinations used in nuclear power plants. All dynamic analyses were conducted using the STRUDL computer program.

Static analyses were carried out by hand, except for the simple / fixed beam with overhang. Consistent with design practice, all mountings in this study are assumed rigid.

Results for Flat Response Table 3.7.3-1 summarizes the results for a 1 g flat response applied to the model beams of Fig. 3.7.3-4. Three K factors were computed for comparison purposes:

Max SRSS dynamic moment s/c Max static moment from concentrated load Max SRSS dynamic moment gs /u _-

Max static moment from uniform load Max ABS dynamic moment afu Max static moment from uniform load All conclusions in this report are based on Ks/u because it most closely represents the actual ratio of dynamic moment to static moment. Ka not so chosen because, as can be seen in Table 3.7.3 3, wasmodes are so widely spaced that no more than one modal frequency lies within a 110 percent frequency band. K sg is shown since this is the K factor which represents a typical simplification used in component analysis (concentrated static loads at component center of gravity) .

The 1 g flat response was selected to give infinite peak spreading. As can be seen, K,u s was never greater than unity.

3.7-24 L l0i

SWESSAR-P1 Results for kmplified Response Table 3.7.3-3 presents the results for the simply supported / fixed model beam with 33 percent overhang subjected to the response The colizmn entitled 1st Mode in spectra of Fig. 3.7.3-5.

Table 3.7.3-3 gives the fundamental frequency (fo) and response acceleration (go) at fo . Note that fo was adjusted par density variation) from well below to well above the peak frequency (fo) of the response spectra to determine the effect on K of the uncertainty in the location of the peak frequency with respect to the fundamental frequency of the model beam. Since all values of K Sy were less than unity, it is concluded that this uncertainty ha.s no important ef f ects on the K factor.

Conclusions

1. Peak acceleration times 1.3 applied as a static load to equipment whose fundamental natural frequency is within the broadened peak of the amplified response spectra curve is conservative.

2. No amount of peak spreading can itself result in a Kuu f actor significantly greater than unity.

3. Uncertainty in the frequency at which the peak response acceleration occurs itself has no important effects on the K factor.

4. Multiple supported continuous spans are not included in the scope of this study. Components or equignent which make up a system of continuous multiple span supports will utilize a factor no less than 1.5 times peak acceleration as in 1 above if applicable.

3.7.3.5.2 Piping Systems _

All significant dynamic modes of responses under seismic excitation are included in the dynamic analysis outlined in Section 3.7.3.6. For simplified dynamic analysis, constant load factors based on the peaks of applicable amplified response spectra are used for small size Seismic Category I piping systems, as described in Section 3.7.3.9.

3.7.3.6 Design Criteria and Analytical Procedures for Piping The general analytical procedure of the modal analysis response spectra method for SSW supplied piping systems is described in Section 3.7.2.1.3. Basic steps and equations used in the analytical procedure are described below.

bbb i 3.7-25

SWESSAR-P1 ,

Mathematical Model For the dynamic analysis, the mathematical model is described as a lurrped mass, multi-degree-of-freedom model. The distributed piping mass is lumped at the system nodal points. For the dynamic analysis, the equation of equilibrium for the system is:

MU + Ch + KU = F where M = Mass matrix for assembled system C = Damping matrix for assembled system U = Nodal acceleration vector = U (t)

O = Nodal velocity vector = U (t)

F = Applied dynamic forces = F (t) and F = Mdg for earthquake, where UX = acceleration = Ug (t)

This equation is solved for the system dynamic response as follows: First, the f requency equation, obtained by removing the forcing and damping terms from the abova equation, is solved for the system natural frequencies and mode shapes. Next, the natural mode shapes are used to effect an orthogonal transformation of the equation, yielding a series of independent equations of motion uncoupled in the system modes. Then, the uncoupled equatione are solved by either the step-by-step integration or the response spectrum method to obtain system h

response in each mode, and the individual modal results are combined to determine the total system dynamic response. The mathematical formulation of these steps is as follows:

Natural Frequencies and Mode Shapes The eigenvalues (natural angular frequencies u n) and the eigenvectors (mode shapes on ) for each of the natural modes are calculated by solving the frequency equation:

2 K -u n fin) = f0}

where en = Natural f requency in nth mode K = Stiffness matrix M = Mass matrix

$#n = Mode shape vector in nth mode 0 = Null vector The eigenvalues and eigenvectors are obtained using the Householder-QR algorithm.

G 3.7-26 (,('t', 4 n i U4

SNESSAR-P1 Dynamic Response Pre- and post-multiplication of equation of equilibrium by [y],

the square matrix of mode shape vectors, constitute an orthogonal transformation f rom which the uncoupled equations of motion shown below are obtained.

Yn + 2wn An n+w Yn = Pn where Yn = Generalized (modal) displacement coordinate for the nth mode (U = pn Y) n A = Damping ratio for the nth mode expressed as percent n

of critical damping Pn = Generalized force for the nth mode

=

%n l

F/M n Mn = Generalized mass for nth mode = g[ Mpn Solution to these differential equations may be obtained by direct integration, or by the method of amplified floor response spectrum superposition.

Response Spectrum Superposition System response to seismic ground motions can also be obtained using the method of response spectrum superposition. Based on this method, the maximum generalized acceleration for each mode is given by:

Yn = Rn San M n where Y

n max = Maximum generalized coordinate acceleration response Rn = Participation factor for nu mode S.i n = Spectral acceleration for n* mode (from response spectrum data input to analysis)

M n = Generalized mass for n* mode and the maximum internal inertia forces are given by F. = M n

Y p. = Maximum inertia force at in max n max in nodal mass point i in the nth mode These inertia forces are calculated for each of the system natural modes, and applied as static forces in the same manner as the weight or equivalent thermal forces, to find generalized internal forces in each mode.

nC I

() b2 3.7-27

SWESSAR -P 1 Where a piping system is subjected to more than one amplified response spectrum, such as support points located in different structures or different parts of the same structure, the envelope of all the amplified response spectra is applied to the system.

The procedures used for developing the X, Y, and Z direction

[j input spectra are described in Sections 3.7.2.6 and 3.7.2.1.1.

To predict maximum responses due to seismic excitation, modal responses are combined by the following statistical procedure:

Compute " modal internal moments" due to dynamic responses of inertial forces by the X-direction input spectrum for each mode. This computation is repeated for the Y and Z-direction input spectrum, respectively. The notation of the internal moment of i mass around the x-axis due to jth mode dynamic responses of system by X-direction earthquake spectrum is

@ ix I jX where:

i = number of masses, i= 1, 2, 3.........N j = number of modes employed, j=1, 2, 3.........d A

ss 9 M,  %

An array of internal moments has thus been computed:

(M.1x) j X (M.ly)j X (M.12 ) j X (M.lx) j Y (M. )jY (M.12 ) j Y ly (M.1x )1Z

" (M-ly)jZ (M12)jZ Internal moments at cach i mass are combined as follows:

d -

IT

) jX + (M it,)2 jZ

~

M. =

f +  ! (Mix) j Y I 1x g -

j (M =1x "iy "

I"iy) $X +

17' i

• I"iy)jY 51 =

(Mj7) ~3 X + (M1=) jr +

(M1=)j y 2

e 1= , { , (_, ,.-

\]g

] .

3.7-28 Amendment 4 11/1/74

SWESSAR-P1 In a response spectrum modal dynamic analysis, Reference 10 shows r that S&W's method is more conservative than the SRSS method. For piping analysis, the SSW method described this section is supplemented by the procedure described in Section 3.7.3.4 for closely spaced modes. The result shown in Eeference 10 is applicable to closely spaced frequency modal responses.

The computation of internal moments at each mass node represents maximum inertial seismic reponses due to simultaneous excitations 7 of the vertical amplified response spectrum and horizontal amplified response spectrum applicable to the piping system The calculated primary stress range due to seismic inertial responses for ASME Code Class I piping components are added absolutely to the secondary stresses due to seismic anchor displacements calculated in the following manner:

Maximum relative displacements in two horizontal and the vertical direction between piping supports and anchor points (i .e . ,

between floor penetrations and equipment supports at different elevations within a building, and also between buildings) are used as equivalen t static displacement boundary conditions in order to calculate the secondary stresses of the piping system.

Relative seismic displacements used are obtained from a dynamic analysis of the structures, and are always considered to be out-of phase between different buildings to obtain the most conservative piping responses.

Internal morents and forces as the seismic responses of the piping system are then combined with loads from deadweight, pressure, thermal, and other mechanical loads to complete the stress anulysis of all Seismic Category I and some Nonseismic Category piping. For ASME Code Class 1 piping, streso intensities and cumulative usage factors of the piping system are computed based on the formulation specified in Subarticle NB-3600, ASME III; for ASME Code Class 2 and 3 piping, the formulations in Subarticle NC-3600 are used.

Dynamic force loadings, resulting from sudden closure of an isolation valve or a turbine throttle valve on the piping system (for example, transient loading on steam line due to turbine trip) , are included as occasional mechanical loads in piping analysis. Constraints or hydraulic snubbers are used as required to control excessive displacements or moments due to these transient loadings.

The design criteria, loading combination, and stress limits for S &W supplied Seismic Category I piping systems are tabulated in Tables 3 . 7. 3-4 and 3.7.3-5.

i

.7 .id n 'I

()(J L 3.7-29 Amendment 7 2/28/75

SWESSAR-P1 3.7.3.7 Ba sis for Computing Combined Response 3.7.3.7.1 Equipment The basis for computing combined response is presented in Sectf2n 3.7.3.4 for equipment and components.

3.7.3.7.2 Piping Systems The procedure des cribed in Section 3.7.3.6 for S&W supplied Seismic Category I piping analysis in combining the dynamic responses from two horizontal and vertical amplified response loadings is a conservative interpretation of the regulatory position verified in Ref 10. Section 7.6 of Ref 9 states that, for single degree-of-freedom linear systems oriented arbitrarily with respect to horizontal directions, the vectorial sum of responses, by considering both horizontal components of the spe ctra , is an upper bound of the response of the system.

Piping systems are inherently of complex geometrical configuration with random orientations in the structural global coordinates system. To compute the maximum seismic response (internal forces / moments and reactions at supports and equipment) at each point due to strong motion seismic excitation at any arbitrary horizontal direction, it is necessary to use this upper bound approach. Since the seismic analysis of the piping system is based on an eigenvalue solution of its structural model, this method is conservative in cxxnputing both the piping responses and seismic reactions in equipment and supports.

3.7.3.8 Amplified Seismic Responses 3.7.3.8.1 Equipment Constant load factors are not utilized for vertical 21oor response in the seismic design of Seismic Category I equipment and components. As described in Section 3.7.2.6, amplified response spectra (floor) are developed for horizontal (two directions) and vertical seismic excitation. Components are designed for the combination of operating loads acting simultaneously with horizontal and vertical seismic loads based on these response spectra. Each of the three defined directions of earthquake input (two horizontal and one vertical taken ortrogonally) is evaluated separately. The calculated stress resulms of the three stress analyses are superimposed on an SRSS basis.

3.7.3.8.2 Piping Systems In the simplified dynamic analysis described in Section 3.7.3.9 for S&W supplied Seismic Category I piping, a constant load factor is used as the vertical and horizontal amplified floor response loadings.

3.7-30 Amendment 7 2/28/75 66c_  : ag

SWESSAR-P1 3.7.3.9 Use of Simplified Dynamic Analynii, ASME Code Class 1 piping systems with 1-inch diameter NPS or smaller, such as sample, drain, and instrument lines, and ASME Code Class or 3 systems, with 6-inch diameter hPS or smaller, are subjected to simplified dynamic analyses using acceleration values from the amplified response spectra. Where possible, the

, c r, in \NI bub 3.7-30A Amendment 7 2/28/75

SWESSAR-P1 length of span between supports is selected such that the fundamental frequency is removed from the resonant envelope of the amplified response spectra.

The basic approach to the seismic design of small size piping is to make the system rigid whenever good engineering design practice allows. Stress calculations are performed for small diameter piping in a sectionalized 'between supports

• manner without using computer analysis. This is justifiable because a rigid system with sufficient pipe supports closely spaced represents many one-dimensional straight beam problems, wherein the coupling effects of three-dimensional piping systems are eliminated by placing constraints near all .lbows, tees, and concentrated masses (such as valves, etc). These hand calculations provide sufficient and conservative data to satisfy requirements of NC-3600 ASME III.

An equivalent static load f actor of 1.5 time = Sie peak of the floor response spectra for design of piping supported between two points is us ed for simplified seismic analysis. The SRSS of response due to inertial loads in two horizontal and one vertical direction is then used to obtain the seismic responses. Deadload and thermal responses are also calculated using the length of the predetermined span.

If the stresses calculated by the sbnplified procedure exceed stress allowables, multi-mass dynamic modal analysis is then performed.

3.7.3.10 Modal Period Variation The irifluence of variation of critical parameters is represented by the different mathematical models used in the development of enveloped floor response spectra. Parameters used for the shear modulus of the subgrade are indicated in Section 2.5.2. Effect of concrete cracking as it influences the mathematical model is discussed in Section 3.7.2.8.

3.7.3.11 Torsional Effects of Eccentric Masses If the torsional effect of the valve operator or other eccentric masses is likely to have a significant effect on the results of analysis described in Section 3.7.3.6 for Seismic Category I piping systems, the eccentric mass and its moment arm are included in the mathematical model described in Section 3.7.2.1.3. However, if the pipe stress due to the torsional effect is expected to be less than 500 psi, the offset moment due to the eccentric mass is neglected.

..r

,;o

)

3.7-31

SWESSAR-P1 3.7.3.12 Piping Outside Containment Structure and Annulus Building Responses of buried Seismic Category I piping to differential ground motion, due to particle motions caused by seismic wave propagations, are calculated by a method reported in Sections 10.6 and 16.5 of Ref 9.

Soil effect on the buried Seismic Category I piping systems due to dif ferential seismic motions between structures and soil at penetration is simulated by a series of elastic springs attached to the piping in the mathematical model, including straioht und bent configuration, to calculate the reactions and moments. The maximum expected seismic displacements at the structural penetration and the maximum modulus of the soil foundation are used in the calculation.

The nubgrade reaction approach is used to simulate the ef fect of soil on the deformation and stress of the buried piping due to seismic anchor movements. The approach is based on K. Terzaghi's theory (Ref. 11) that soil subjected to pressure behaves like a syste, of uniformly spaced elastic springs with predetermined stiffness. The soil spring constant, L , and the coefficient of subgrad.' reaction, K, are related by:

k = KA (1) h where A is defined by an area equal to the product of pipe diameter and the spacing between two consecutive soil springs.

The soil spring constant k has unit of force per length, whereas K has unit of force per (length) 3 k and K apply to both horizontal and 'rertical directions as follows:

K H

= C Z/D g I

(2)

K =C (Do + 1)2/02a (3) s 2 where =K =

coefficient of subgrade reaction in the H horizontal direction , tons /ft3 K =

coefficient of subgrade reaction in the vertical direction, tons /ft3 Z depth below ground surf ace, ft Du =

outside diameter of pipe, ft C = coefficient of horizontal subgrade reaction tor I

a pipe with Z/D g = 1, tons /ft3 C,

=

coefficient of=vertical subgrade reaction f or$ a pipe with D 1 ft, tons /ft3 [ (, , i i!

3.7-32 Amen dment 7 2/28/75

EWESSAP-P1 Coerficients C, and C 2 are dependent upon the soil properties at  ;

the site.

The results are superimposed with axial-tension and conpression stress to meet the requirements defined in Subarticles NC-3600 und ND-3600, ASME III. All Seismic Category I buried piping l, systems are either ASME Code Class 2 or 3. If these stresses are l' tound to be excessive, the underground piping is placed wittin concrete or steel conduits which are unattacbed to any structure.

For all other Seismic Category I piping systems located outside the containment structure and annulus building, the analytical procedure and design criteria are described in Section 3.7.3.6.

3.7.3.13 Interaction of Other Pioino with Catecory I Piping If a possibild ly exists that a Nonseismic Category pipino system (which is attached to Seismic Category I piping) could f a ll and by that failure could adversely af f ect Category I piping , each Nonselsmic Category pipino system is designed to be isoluted Irom any Seismic Category I pioing system by either a constraint or barrier, or it is remotely removed from the location of the Seismic Category I piping syste.a. If it is not feasible or practical to isolate the Seismic Category I piping system from the Uonseismic, the 2d jacent Nonseismic Category piping is then seismically designed according to the same criteria as the Seismic Category I p.i ping system. For the Nonseismic Category piping systems attached to Seismic Category I piping systens, the dynamic effects of the Nonseismic Category piping are simulated in the analysis modelino of the Seismic Category I pipina. The Nonseismic Category piping is simulated in a manner that does not degrade the accuracy of the Category I piping analysis. The attached Nonseismic Cateoory piping is also designed in such a munner thut, during an earthquake of SSE intensity, it does not cause a failure of the Seismic Category I piping.

3.7.3.14 Pield Location of Supports and Restraints Field location of seismic supports and restraints for SSW supplied Seismic Category I piping systems, including snubbers and dampers, is shown on approved construction drawings based on seismic analyses. Inspections by qualified seismic pipe stress unulysts are conducted at the construction site to verify that

. ~ ")

, ,\u hbd 3.7-32A Amen dment 7 2/28/75

SWESSAR-P1 Inspections by qualified seismic pipe stress analysts are conducted at the construction site to verify that these seismic restraints are fabricated and located in accordance with the approved construction drawings.

3.7.4 Seismic Instrumentation Program 3.7.4.1 Comparison with AEC Regulatory Guide 1.12 In accordance with AE ' Regulatory Guide 1.12, " Instrumentation for Earthque cs," issued April 1974, seismic instrumentation is provided to monitn and record input motion and behavior of the sites, plant in the event of an earthquake. At multi-unit additional instrumentation beyond that installed for a single unit is provided if essentially different seismic responses are expected at the other units.

3.7.4.2 Iocation and Description of Instrumentation Two strong motion accelerographs are installed in the containment structure of the nuclear island as required by Regulatory Guide 1.12. One is located on the basement floor of the containment structure and the other is located vertically over the first on the operating floor. These locations are selected' to give intormation regarding the motion of the base mat of the nuclear island as near its center as practical and to give information regaroing motion of the superstructure which supports principal equipment other than the reactor. A third strong motion accelerograph is on the top floor of the control building.

Inis location is chosen because the control building has dynamic response characteristics different from the nuclear island.

Except for crystalline rock founded sites, a fourth strong motion to 4

triaxial accelograph is installed in the free field in order obtain more detailed knowledge of soil structure interaction during an earthquake. The strorg motion triaxial accelerographs have the following physical characteristics:

A. Accelerometers are of the force-ba)ance type with the capability of recording a maximum of 1.0 g at full scale.

B. Accelerometers are sensitive to f requencies in the range of 0.1 to 33 Hz.

C. The seismic instrumentation and recording system is in a quiescent state until activated by seismic triggers which are set at 0.01 g. These seismic triggers 3)oth horizontal and vertical) activate che recording system in less than 100 msec and trip an annunciator in the control room to alert the operator. Recording continues until the level ot motion drops below 0.01 g. This trigger value is aelected for sensitivity to record any 9 significant seismic event. while, at the same time, eliminating instrumentation system response by nonseismic background noise.

3.7-33 Amendment 4

.- 11/1/74 bbi l-1)

SNESSAR-P1 D. The recording system is powered by internal batteries with trickle charge from a 120 V a-c power supply.

Recording of the electrical signals from the accelerometers is by magnetic tape with simultaneous time signal recording.

E. Each sensor package contains three mutually orthogonal dCCelerometers.

All strong motion sensor packages are oriented to the same azimuths and also aligned with the major ax7s of the building used in the mathematical model to permit direct use of the data with the model. All strong motion sensor packages are located in areas where they can be serviced during periods of chutdown.

Triaxial peak recording accelerographs are installed at the following locations:

A. One on reactor piping B. One on reactor equipment C. One on Category I equipment or piping outside the containment structure D. One on Category I equipment on the top floor of the control building )

These instruments detect and record peak amplitudes of low frequency acceleration and are used to verify the seismic response determined analytically from the traces recorded by the strong motion accelerographs. Peak recording accelerographs have the following physical characteristics:

A. Accelerometers are of the short period torsional type with a sensitivity of 1/4, 1/2, 1, 2, or 5 g full scale.

B. The accelerometers record by erasure of prerecorded lines on magnetic tape clips, one for each of the three orthogonal axes. The lines are erased by a noncontact stylus.

C. No power is required to operate the instrument.

D. Damning is electromagnetic to greater than 0.5 critical.

E. Operating temperature range is -20 to 150 F.

Seismic instrumentation that provides a spectrum of measured responses is provided. This instrument, called a peak shock recorder, senses and permanently records the information defining a response spectra. It is a completely passive device covering the range of 2 to 33 Hz and 1/3 octave increments. Twelve reeds of different lengths and weights, one for each frequency, have 3.7-34 bb/ t i$

SWESSAR-P1 attached to their free end a diamond-tipped stylus which inscribes a permanent record of its deflection on one of 12 record plates. A calibration sheet lists the resonant frequency and g sensitivity of each reed and allows a plot of acceleration versus frequency to be made. These instruments have the following physical characteristics:

A. Damping for the oscillators is 2 percent.

B. Weight of one peak shock recorder is 34 lb .

C. Operating tenperature range is -67 to 185 F.

D.* Accuracy: frequency is 11 percent, acceleration is 13 percent of full scale.

Triaxial configurations of peak shock recorders (that is three mutually orthogonal peak shock recorders) are installed at the following locations:

A. One on the basement floor of the containment structure B. One on reactor equipment or piping support location C. One on the most pertinent location of each of the following outside the containment structure:

1) Seismic Category I equipment support or appropriate floor location

2) Seismic Category I piping support or appropriate floor location D. One on the top floor of the control building The basis for selection of the above structures and equipment is to provide some measure of redundancy to the strong motion accelerographs and also to provide additional data to verify the seismic response determined analytically from the traces recorded by the strong motion accelerographs.

To provide an immediate signal to the control room to indicate if specified design accelerations have been exceeded, triaxial seismic switches are installed at the following locations:

A. In the basement of the containment structure.

B. At a selected location on reactor equipment or piping supports.

These seismic switches have the following physical characteristics :

3.7-35 , , i;5 ooL

SWESSAR-P1 A. The package is composed of three orthogonal acceleration transducers.

B. Acceleration response is flat from 1 to 10 Hz.

C. Switch closure remains closed for 6 to 20 sec (adjustable) after detection of an acceleration over the preset value.

D. Operating temperature range is from 0 to 130 F.

3.7.4.3 Control Room Operator Notification Recording equipment is located in the control room. The recording system generates acceleration-time history records in the form of strip charts. In order to evaluate the acceleration intensities experienced by the plant, templates made of clear plastic with maximum allowable accelerations shown as upper bounds are placed over the strip charts.

The acceleration recording system remains in a quiescent state until motion of one of its triggers (caused by an acceleration of 0.01 g or greater) causes the system to turn on and record motion. The strip chart readout is not activated by a predetermined value be is produced manually by quick playback _of the recorded signals.

To provide additional immediate information on which the operator h can act, the control room is provided with a peak shock annunciator which monitors the triaxial configuration of peak shock recorders on the basement floor of the containment structure. Dual contacts (on the record plates) which close when the reeds of the peak shock recorders deflect through the predetermined distance provide the capability for giving instantaneous alarm signals when preset accelerations at selected frequencies are exceeded. Three banks of indicator lamps (one bank for each of the three mutually perpendicular axes) light up to indicate that (1) accelerations ~ have approached a given percentage of design limits, or (2) accelerations have exceeded design limits.

3.7.4.4 Comparison of Measured and Predicted Responses In order to make detailed comparisons between measured seismic responses of Seismic Category I structures and equipment with calculated accelerations determined from dynamic analysis, the following procedure is implemented:

A. The magnetic tape records are digitized and corrected for time signal variations and baseline deviations.

B. For soil-founded sites:

O

• /

3.7-36 4 bbd * '

SWESSAR-P1

1. The time-history records from the triaxial sensors on the operating and basement floors of the containment structure are used for direct calculation of amplified response spectra at appropriate critical damping.

2. The time-histoiy-records from the free field triaxial sensor are used as input ground motion for the dynamic model of the nuclear island configuration. Amplified response spectra are then calculated at the locations of the other two sensors in the containment structure for comparison and correlation with the response spectra determined as in B.1 above. Reasonable correlation between the spectra is accomp1.ished on an iterative basis by varying the physical properties of the models (stiffness and damping characteristics) to

" calibrate" the dynamic model. Once the dynamic model has been " calibrated," additional verification of its correctness is made by use of the acceleration readings from the peak recording accelerographs.

C. For rock-founded sites:

1. The time-history records from the triaxial sensor 9 packages at the operating floor level of the containment structure are used for calculation of actual amplified response spectra at direct appropriate critical damping.

2. The time-history records from the basement floor of the containment are used as input ground motion for the dynamic model of the nuclear island configuration. Amplified response spectra are then calculated at the operating floor level of the containment structure. Correlation between actual and predicted responses 3 then obtained on an iterative basis by varying che physical properties of the models to " calibrate" the dynamic model.

D. Structural responses and amplified response spectra are calculated with the " calibrated" dynamic model for comparison with the original plant design parameters.

This comparison permits the evaluation of seismic effects on structures and equipment and forms the basis for detailed analyses and physical inspection.

3.7.5 Seismic Design Control The Seismic Category I components and equipment listed in Table 3.2.5-1 are subject to a comprehensive program of data handling, review, and approval to assure compliance with the provisions of 10CFR50, Appendix B.

.4 7 3.7-37 bb

SWESSAR-P1 3.7.5.1 Data Oricination The basis for all seismic subsystem and component analysis is the amplified (floor) response spectra, developed by structural system analysis. These data, and any revisions, are forwarded to the NSSS Vendor and necessary specialist groups. Coordination to assure that all necessary data requirements are fulfilled and that all affected parties are cognizant of any data revisions is described below.

3.7.5.2 Components and Equipment A. Data Handling One specialist group is responsible for coordination and implementation of data for component and equipment design and/or specification. This group assures consistent application of the latest data to all component applications and acceptance criteria.

B. Data Use (1) Components and Equipment Designed by Applicant - These components are designed and documented in accordance with applicable parts of Section 3.7.2.1.2. Vendors are required only to fabricate and supply.

(2) Components and Equipment Requiring Vendor Documentation - These components are specified for design and documentation in accordance with applicable parts of Section 3.7.2.1.2. Because of the variety of situations presented, a series of review / approval steps are implemented. These steps ensure that at both the bid and purchase stages the data and acceptance requirements are clearly stated and understood. All bidders are required to submit plans for analysis or test. These are reviewed by the specialist group for responsiveness to specification requirements by the cognizant engineer and others as appropriate. In this way, assurance is obtained that the Vendor will be totally responsive to seismic requirements prior to issue of the purchase specification. Prior to release for shipment, vendor documentation of seismic adequacy is revi wed and, in certain cases, independently veriff ay the same specialist group responsible for data and criteria control, in addition to normal purchase considerations. A potential Vendor cannot be approved unless the seismic documentation program is acceptable, and the actual Vendor will not be allowed to ship a component unless the seismic adequacy is reviewed and accepted.

3.7-38 ,g bOL '

SWESSAR-P1 The specialist group responsible for review and control coordinates with the various equipment specialist and project cognizant engineers. This is necessary to ensure that acceptance criteria reflect any special requirements of the particular system involved and that acceptance criteria beyond the expertise of this specialist group (e. g . ,

electrical circuitry) are properly specified. A representative of this group is generally present to witness any testing relevant to earthquake qualification to ensure compliance with the specification.

3.7.5.3 Piping The Pipe Stress Analysis Engineering Section is the responsible group that verifies the adequacy and validity of the analyses and tests employed by constructors of the Seismic Category I piping systems and pressure relieving piping systems listed in Table 3.2.5-1. The appropriate seismic input data for piping system analyses, derived from structural seismic system analyses, are specified in the pipe stress analysis reports included within the scope of Design Control in the Stone & Webster Quality Assurance Program. All ASME III Code Class 1 pipe stress reports, prepared by the responsible engineer, are reviewed, checked for conformance to design specifications as required by ASME III, and certified by the Pipe Stress Analysis Engineering S3ction. ASME III Code Class 2 and 3 pipe stress reports are reviewed, checked for conformance to design specifications, and approved by the Pipe Stress Analysis Engineering Section.

Finally, pipe stress reports are approved by the Stone & Tfebster Project Engineer and then transmitted to the Field Quality Assurance Engineer.

3.7.6 Interface Requirements Interface requirements when imposed on the SWESSAR-P1 design by Section 3.7 of the NSSS Vendor's SAR are addressed in Table 3.7.6-1. This interface section is referenced in SWESSAR-P1 Tables 5.1-1 and 5.5.7-1.

Seismic qualification of all seismic Category I equipment is required. The NSSS Vendor is required to qualify all equipment within h % scope of supply. For the equipment within the SSW and Utility-Applicant's scope of 25 responsibility, the seismic qualification program is defined in Section 3.7.2.1.2. This program definition is sufficient for the construction permit applica tion . 'Ihe Utility-Applicant's operating license application shall describe the detaile of the seismic qualification program.

References for Section 3.7 (1) Whitman, R.V., Design Procedures for Dynamically Lor.Jed Foundations, University of Michigan, 1967.

3.7-39 Amendment 25 r

(m.3 \h0 4/30/76

SWESSAR-P1 (2) Milos Nowk, Vibrations o_f_ f Embedded Footings and Structures, ASCE National Structural Engineering Meeting, April 1973.

(3) Roesset, J .M . , Dobry, R., and Whitnan, R.V., Modal Analysis for Structures with Foundation Interaction, Journal of th e Structural Dtvision, ASCE, March 1973.

(4) Hou, S.N., Earthquake Simulation Models and Their Applications, Research Report R68-17, Department of Civil Engineering, MIT, 1968.

(5) Rascon, O.A. and Cornell, C.A., Strong Motion Earthquake Simulation, Research Report R68-15, Department of Civil Engineering, MIT, 1968.

(6) Tsai, N.C., Spectrum Compatible Motions for Design Purposes, Journal of Engineering Mechanics Division, ASCE, Vol. 98, No.

EM2 Rev. Paper 8807, April 1972, pp 345-356.

(7) Nigam, Navin C. and Jennings, Pacul C., Digital Calculation of Response Spectrum f rom Strong-Mo_ tion Earthquake Records, National Science Foundation, June 1968.

(8) Newmark, Nathan M., Blume, John A., and Kapur, Kmwar K.,

Design Response Spectra for Nuclear Power Plants, presented at ASCE Naticnal Structural Engineering Meeting, April 1973.

(9) Newmark, N M., and Rosenbleuth, E., Fundamentals o Earthquake Engineering, Prentice Hall, 1971.

f (10) Chang, T. Y., " Comparison of Seismic Response Combination Procedures for Piping Systems," Power Division Report, Stone & Webster Engineering Corporation, October 1973.

(11) Terzaghi, Karl, " Evaluation of Coefficients of Subgrade Reaction," Geotechnique, Vol 57 1955 (pp 297-325) .

(12) Bendat , Julius S., and Piersol. Allan G., Random Data:

Analysis and Measurement Procedures, Wiley -

Interscience, I 1971.

(13) Seed, H. Iblton , Idriss, I.M., Soil Moduli and Damping Factors for Dynamic Response Analysis, Earthquake Engineering Center, University of Calif ornia , Berkeley, California.

Report No. EERC 70-10, December 1970.

(: v

.e i _' o 3.7-40 Amendmen: 7 2/28/75

SWESSAR-P1 TABLE 3.7.1-1 DAMPING FACTORS Damping (Percent Critical)

Item, Equipment, or Structure OBE SSE Equipment and large diameter piping systems, pipe diameter greater than 12 in. 2 3 Small diameter piping systems, diameter less than or equal to 12 in. 1 2 Welded steel structures 2 4 Bolted steel structures 4 7 Prestressed concrete structures 2 5 Reinforced concrete structures 4 7 Soil translation and rotation 10 10

.n.

of 1 (, gLk

SWESSAR-P1 TABLE 3.7.1-2 ACCEPTABLE MiTHODS FOR SOIL-STRUCTURE INTERACTION ANALYSIS Soil Foundation **

Shallow Embedment Case Method of Deep Soil Deep Soil Soil- Foundation Foundation Structure Deeply with with Shallow Interitction Rock

• Embedded Uniform Layered Soil Analy: sis Fbundation Case *** Properties Properties Foundation Single Lumped Ma s s-- X X Spring Approach Finite Elenent X X X X X 29 Approach

• A medium for which soil-structure interaction ef1ect is neg-ligible or alternatively a medium with a shear wave velocity greater than or equal to 3500 fps. 29

• • Soil foundation means the depth of soil between the cottom of the foundation slab and base rock.

• • • Actual embedment >15 percent of the least base width.

f /n (y L. /

i7) t "-

1 of 1 Amendment 29 10/29/76

SWESSAR-P.1 Table 3.7.2-1 BASE SHEAR & OVERTURNING MOMENT FOR CONTAINME?TT WALL Model Masses Base Shear (KIPS) Base Ove rturn inc [ KIP-Ft) 5 3 41,239.43 6,449,794.0 5 40,684.5 6,433,781.0 10 40,360.5 6,412,842.0 15 40,055.8 6,351,544.0

. n ',

1 of 1  ;',

, iL J Amen dment 5 12/2/74

SWESSAR-P1 TABLE 3.7.2-2 COMPARISON OF NA'IURAL FREQUENCIES FOR CONTAINMENT UALL HORIZONTAL MODAL FREQUENCIES (CPS)

Model Masses 1 2 3 4 5 6 7 8 5

3 3.76 10.39 16.87 18.56 28.65 37.63 - -

5 3.76 10.70 18.53 22.38 28.89 31.96 33.55 48.89 10 3.76 l0.84 19.28 23.97 33.44 35.65 42.48 49.45 15 3.79 *0.92 19.50 24.35 34.33

. 36.36 44.55 52.84 1 of 1 ,

,,sAmendment 5

L" 12/2/74

SWESSAR-P1 -

TABLE 3.7.2-3 EQUIPMENT SEISMIC QUALIFICATION Type of Equipment Qualification Method Valve Operators (Motor) Testing for Active Valves HVAC Motor Operators Testing Batteries, Battery Chargers, Testing 16 and Inverters Switchgear, Load Centers, Testing Motor Control Centers Relays, Instrumentation, and Testing Control Devices Cable Tray Systems Dynamic Analysis Cranes Dynamic Analysis 1 of 1 Amendment 16

.nCJ 8/29/75

, iL

/ i', L

U 9 7 7 7 3 5 1

/ 9 0 9 9 0 1 0 A .

1 1 1 1 U 9 3 9 7 5 4 7

/ 8 0 8 8 7 7 8 fI S 1

C 9 1 9 8 1 4 7

/ 8 5 5 5 5 7 8 S

n o

i t

a F M F F F F S c

I n

)

b 00 00 00 00 00 00 00 tl 00 00 00 00 00 00 00 n- 0, 0, 0, 0, 0, 0, 0, 0, 0, 0, 4, 2, 0, 0, e.

c min 00 84 46 14 21 77 44 E i ( 00 47 71 67 61 77 77 S m Mf 77 31 1 1 21 1 1 11 1 N a 1 - O n P 3 P y

- . S D 1 R 7 E de ct cf et cf cf cf cf A R x ap ni ni ni ni ni ni ni f S 1 a oy on on on on on on oi o S

E T MlT CU cU CU CU cU cU cth E A 1 w W L L S B F A

T G 1 )

b 00 00 00 00 0 0 00 tl 00 00 00 00 00 00 00 i

c er.n.n 0, 0, 09 0, 0, 96 0, 0, 32 0,0, 29 00, 00, 2, 4 0, 9 0, 0, 26 m ui 24 78 01 56 31 78 57 a M( 66 1 1 1 1 11 81 5 11 ny D

S S S S S S S x SS SS SS SS SS SS 3 S Sy SA a RB RB RB RB RB RB B M SA SA SA SA SA .A l 4 4 2 a

t 1

0 6 na ee 1 1 1 1 1 1 mr aF di W

F N O

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O L p g g g g d u n n n n F n S a a a a O En h h h h ae r r r r N d pl m r e e e e O esp a e v v v v I xdm e v o O o O T iii B e A FMS l o 1  % 1 N l i N 6 3 0 A -

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SWESSAR-P1 TABLE 3.7.3-2 MODAL DENSITY, n*

Cantilever SF y m 3' Mode Freq'n FF' SF SS 331 OH

~~ J ;,' < j No. (Hz) Freqen Freq'n Freq'n Freq'n

.y r,' ' ,

1. 1.0 1 1.0 1 1.0 1 1.0 1 1.0 1

2. 5.8 1 2.7 1 3.2 1 3. 8 1 2.9 1

3. 15.3 1 4.9 1 6.3 1 8.2 1 6.5 1

4. 28.0 1 7.5 1 10.2 1 13.6 1 8.4 1

5. 43.2 1 10.2 1 14.0 1 19.5 1 13.3 1

' 59.6

6. 1

• Modal density is based on a 1101 criterion.

t U-

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.. -.:3 L

~

b i

1 of 1

SWESSAR-P1 TABLE 3.7.3-3 AMPLIFIED RESPONSE DYtRMIC FACTOR STUDY

' " ']l Mar Dynamic Moment Max sitatic Moment K__

Model 1st Dynamic Load Sy . Moment Location Unif orm Load (U) , in.-lb S/U 3

f Beam Mod e Hign M (S) (in.-lb) (Note 1) A/U

) sF-331 9 g 9 overhang f f f Model 6 A .10 2.87 .33 SRSS 20,000 F 222,000 s .09

.70 3-4 20 ABS 30,000 F .13 Model 6B . 10 2.87 .33 SRSS 148,000 F 222,000 s .67 1.0 3-4 20 ABS 157,000 F 71 1 Model 6C 2.87 2.87 .33 SRSS 102,000 S 222,000 s .45 3.3 3-4 20 ABS 118,000 S .53 Model 6D .40 2.87 .33 SRSS 22,000 F .71 10.0 3-4 20 ABS 32,000 F 31,000 s 1.03 4

Model 6E .33 2.87 .33 SRSS 20,000 F 25,700 s .78 20.0 3-4 20 ABS 27,000 F 1.05 Model 6F . 30 2.87 .33 SRSS 18,000 F 23,400 s .77 33.0 3-4 20 ABS 25,000 F 1.07 3

- -. 22 yoge:

1. 9 max if fo 5 fp , g at to if fo > fp

(~;~

Cr p.

CD 1 of I

b SWESSAR-P1 TABLE 3.7.3-4 PIPING SYSTDt SEISMIC DESIGN AND ANALYSIS CRITERIA ASME Sect ion III Code Ty pe of Type of Combined Stress Calculations Class Ea rt tuvua ke Seismi c Arhalyses and Stress criteria 1 SSE Dynamic response spectra ASME Code,Section III, (Siz es 1-1/4 in. N PS Subarticle NB-3600 and larger)

OBE Dynamic response spectra ASME Code,Section III, Subarticle NB-3600 4 1 SSE Simplified dynamic ASME Code,Section III, (Siz es 1 in. NPS and below) analyses Subarticle NC-3600

' OBE Simplified dynamic ASME Code,Section III, analyses Subarticle NC-3600 2 ani 3 SSE Dynamic response gectra ASME Code.Section III, (Sizes 8 in. NPS and larq. c) Subarticle NC-3600 s

OBE Dynamic response spectra ASME Code,Section III, Subarticle NC-3600

+ -

2 and 3 SSE Simplified dynamic ASME Code,Section III, analyses Subarticle 1C-3600

, (Sizes 6 in. NPS and below)

OBE Simplified dynamic ASME Code,Section III, analyses Subarticle NC-3600 ma d

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s

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F a d d

• e n en n e Q ga i d s ah a d s+ sT t A e e P U an rP 2 s ao h t

S+ V es ve si s

t i

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w k b = t ea aP ar hp I

e l o t x I

P+ t ut oE I l mc0 h

) a ua A A ,l S L S CF1 N N ta A A n P hm + N N o

( gr i e ie A t P eh S c M wT e E S T ,d E

S en ra M Y S S Q u ss A G s+ sd h

N s ea t I

P e e rP ro PI i I yt w P rS+ n a lo e I myn m ai c irP S ms n Y ra A A rn a R Pd+ 3 N N ea A A A d O n hp S N N r G o) Tx o E ct E c T eP c A S( l a C n a h h n 5P n h C os S S S i I id M sa 2 0 4 e S

I s an cI 1 1 2 r

a E s c s S e O t r m R tQ n S S

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- F + 5 e i 1 3 y 5 2 0 ns l P S r) id e

- 7 T a ; 1 2 3 aa h v 1

R I mP to o A 3 M i( sL S A A f S I r y u N N b o S E L PP S a 1 E L W B S e S A S h T E d t R h,ct e T t s f S r i niat r o et womin et d n D hn t iaca hn t N n st a dn dtnol n st a dn d n o A o gil ea ei ys p o gil ea e amd i i nepE tl tdds i n e pE tl tt ee t S t i B ap an ad t i D ap an,ttds a N a dhtO i io e a dhtO i iaEsaen c O n ate cy ccegt n ate cy c l S yi t o i oisd oc o hnl i oisd oc l o ,cpS s wli I i T b l wpn sn stti u b lwpn sn s ut p p

A m ua se sn da m ua seag sl ai a

N o td ag aaldo af o td nern rn aasisfd mnma n I C nern rn n C d B etoo seo spnl eo g

et oo seogmisroa eo M q ra i gmi g a hi ra i g o i n gh c n net nnt ynt a

)

n rilt net nl mtt n rilt T

' i ucai i i ia e i i ucai i i i idi t d comd dhd dm,thd d coml k lhd dad l dhn ,

G a nsrn at n arEst n a nsrr at n at neat a s N o osoo oio o o S yi o o osoo oio oiohoil n I L Canc Lwc LnSswc L Canc Lwc Lwctlwp o D i A t a n y n y i I o & c o 6 c d i n d i n d n N t e e t l e e o G

I i

d

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r l t

d i at mae g

r l t c S n r s e u n rs e u ,

E o op m a o op m a e D C NU E F C NU E F r u

t a

l c

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ns ns n os ) os ))

00 e ia 0 ia tl 0 tl 00 h cC 6 cC 66 T e 3 e 33 Se - Se - -

d H d CD .

Eo N Eo NN e MC ( MC (( t o

S S A 1 A 23 N I

' .LC C'c.p

TABLE 3.7.3-6 CYCLES OF MOTION FOR LARGE EARTFOUAKES No. of Cycles in Which Accelera-tion Equals or Exceeds 1/2 of Max Earthquake Acceleration Taft (1952) S59E 9 N21E 9 5

El Centro (1940) NS 10 EW 12 Golden Gate (1957) NE 3 S80E 5 Olympia (1949) M86E 7 San Fernando:

Pacoimia S74W 7 Dam Record (1971) S16E 7

, , , \J

( > .l !

1 of 1 Amendment 5 12/2/74

SWESSAR-P1 TABLE 3.7.6-1 SEISMIC DESIGN INTERFACE REQUIREMENTS Westinghouse Requirements SWESSI.R Design The zero period acceleration of the The zero period containment operating floor does not acceleration ot exceed 5 times tha maximum ground the containment Oceleration value or 2.0 g operating floor (RESAR 41, Section 3.7.1) will not exceed 1.5 g (S times the SSE of 0.3 9).

The procedure used to compute this acceleration is described in Section 3.7.2.1.1.

For structures other than the con- For structures 25' tainment, the maximum floor accelera- other than the tion, at zero period, and at the highest containment, the floor elevation at which Westinghouse maximum floor equipment is located, shall not exceed acceleration at the allowable acceleration of Fig. 3.7-4 zero period and (RESAR-41, Section 3.7.1) . at the highest iloor elevation at which Westing-house equiptaent is located will not exceed the allowable accel-6 erotion of RESAE-

! 41, Fig. 3.7-4.

The procedure used to compute these accelera-tions is described in Section 3.7.2.1.1.

t y.

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(* W_ 1 of 1 Amendment 25 4/30/76 y

w

SWESSAR-P1 ,

TABLE 3.7.6-1 SEISMIC DESIGN INTERFACE REQUIREMENTS Westinghouse Requirement SWESSAR Design The zero period acceleration of the 1. SWESSAR satisfies containment operating floor does not this requirement.

exceed 5 times the maximum ground acceleration value or 2.0 g.

(RESAR-3S, Section 3.7.1)

For structures other than the con- 2. SWESSAR-satisfies tainment, the maximum floor accelera- this requirement.

tion, at zero period, and at the highest floor elevation at which Westinghouse equipment is located, shall not exceed the allowable accelera-tion of Fig. 3.7-2 (RESAR-3S)

The seismic envelope response spectra 3. These equipnent for RESAR-3S primary equipment response spectra, (reactor coolant pump, steam generator, when available, pressurizer) will be provided/ described will be evaluated. 24 to the balance of plant esigner Westinghouse and (RESAR-3S, Section 1.7.1, p 1.7-8, S&W will develop Amendment 4) a design to ensure that RESAR-3S and SWESSAR are com-patible.

Westinghouse will provide the balance 4. The equipment of plant designer with the NSSS equip- properties will ment properties needed to perform the be used in the building analysis. (RESAR-3S, Section analysis of the 1.7.1, p 1.7-9, Amendment 6 ) building. Details of building analysis are given in Section 3.7.

Westinghouse will evaluate the balance 5. Seismic response of plant data for acceptance (RESAR-3 S , spectra and rela-Section 1.7.1, p 1.7-8, Amendment 4.) tive displacement will be provided to the NSSS Vendor.,

W-3S 1 of 1 Amendment 24

, 1 bO2 1J}4/23/76

SWESSAR-P1 TABLE 3.7.6-1 SEISMIC DESIGN INTERFACE REQUIREMENTS Babcock & Wilcox Requirement SWESSAR Design Babcock & Wilcox will provide the following information to the balance-of plant designer.

Seismic response spectra of critical 1.The seismic response spectra support points in the RCS interface loadings, nozzle loads, and displacements at critical support points when provided by the NSSS Vendor will be evaluated. S&W will develop a design to ensure that BSAR-205 and SWESSAR are compatible.

Normal, thermal, and seismic absolute Same as Item 1 30 displacements of attachment interface points in the RCS Sufficient data for the balance of Same as Item 1 plant (BOP) designer to perform piping analysis for piping attached to the RCS Laterface loadings for all supports Same as Item 1 and restraints attached to the RCS Allowable design loads for nozzles Same as Item 1 in the RCS attached by the BOP designer Response spectra at critical plant Same as Item 1 equipment elevations and points of support Design response spectra for auxiliary Same as Itern 1 equipment Design response spectra for Category I Same as Item 1 control and instrumentation equipnent Design "g" value for remote sensors Same as Item 1 ,

bb/ 'J B&W 1 of 2 Amendment 30 1/28/77

SWESSAR-P1 TART E 3.7.6-1 (CONP)

Babcock & Wilcor Requirement SWESSAR Design Design response spectra for Category I Same as Item 1 fuel handling equipment Available correcting piping loads for Same as Item 1 active class 2 i.nd 3 pumps and valves Areas that require detailed analytical methods, either linear or nonlinear, will be coordinated with the BOP de- 34 signer to ensure cocupatibility (B-SAR-205, Section 3.9.1.6, Amendment 2)

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SWESSAR-P1 3.8 DESIGN OF CATEGORY I STRUCTURES 3.8.1 Concrete containment Structure 3.8.1.1 Description of Containment 3.8.1.1.1 General Description The containment st ructure, both concrete and steel liter, is sinilar in design and construction to that of the Millstone Nuclear Power Station Unit 3 of the Northeast Utilities Service Company in Waterford, Connecticut (Docket No. 50-423) . It is a steel lined, heavily reinforced concrete structure with a cylindrical wall and hemispherical dome supported on a reinforced concrete base mat. No new or unique features of containment are utilized in the design of the containment structure for this plant. General arrangement of the containment structure is shown in Fig. 1.2-1 and 1.2-3. The cylindrical portion of the containment s tructure is 150 ft inside diameter, and rises 170 ft from the top of the mat to the bend line. The wall is 5 1/4 f t thick, and the dome is 2 1/2 ft thick. The nuclear island foundation is a single continuous reinforced concrete mat resting on the subgrade. The mat is 10 ft thick with a base elevation typically 61 ft below finished ground grade, actual foundation elevation is dependent on the NSSS, and g is indicated on Fig. 1. 2-1. At some sites, the mat elevation can be raised approximately one floor spacing to better in terface with actual site conditions. The selection of founding elevation does not affect the functional design or methods of structural analysis for the nuclear island. The steel liner for the walls is 3/8 in. thick. A 1/4 in . plate is used over the base mat and the steel liner for the dome is 1/2 in . thick. Section 3.8.5.1 cantains a discussion of the protection of the liner from groundwater. The containment structure is provided with two 7 ft-0 in. ID personnel hatches, and a 21 ft ID equipment hatch. Other containment structure penetrations consist of thermally hot and cold process pipes, including the main steam and feedwater lines, the fuel transter tube, and electrical conductors . e .g*", 44 b b' a> 3.8-1 Amendment 9 4/30/75

SWESSAR-P1 3.8.1.1.2 Reinforcing Steel Arrangement The reinforced concrete structure is designed to withstand the loadings and stresses anticipated during the operating life of the plant, as defined in Section 3.8.1.3. The steel liner is attached to, and supported by, the concrete. The liner runctions primarily as a gas-tight membrane and also transmits loads to the concrete. During construction, the steel liner serves as the inside form for the concrete wall and dome. The containment structure does not require the participation of the liner as a structural component. Hoop tension in the cylindrical concrete wall is resisted by horizontal reinforcing bars near both the outer and inner surfaces or the wall. Horizontal circumferential bars, inclucing those in the dcne , have their splices staggered wherever possible. Iongitudinal tension in the cylindrical wall is resisted by rows of vertical reinforcing bars placed near the interior and exterior faces of the wall. Vertical bars are placed in groups of acoroximately 20 bars of equal length. The groups have a U dif; rential joint height of not less than 4 ft. These groups are arranged so that no adjacent group in the same or opposite face of the wall has splices closer than 4 ft apart vertically. The dome reinforcement consists of layers of reinforcing steel placed meridionally, extending from the vertical reinforcing of the cylindrical wall, and horizontal layers of circumferential bars. Layers are located near both the inner and outer f aces of the concrete. 'Ihe radial pattern of the meridional reinforcing steel terminating in the containment dome results in a hith degree of redundancy of reinforcing steel in the dome. Bars are terminated beyond a point where there is more than twice the amount of steel required for the design purposes. The rate of convergence of these bars, and low stress requirements dictated by this arrangement, results in a satisfactory development length or the meridional reinforcing bars. Near the crown of the dome, the meridional reinforcing bars are welded to a steel ring cast in the concrete concentric with the dome centerline. Radial shear in the containment structure wall resulting from the DBA varies f rom a Inavimum at the base of the wall to zero at some level above the mat. To resist the large radial shear near the base at the wall, flat steel bars, inclined at approximately 45 degrees with the horizontal, are welded to the vertical rein :orcing as shown in Fig. 3.8.1-1. The welded flat bars are ten aated at a level above the mat where the radial shear load has iaduced to a value that can be resisted by radial shear reintorcing bars. These radial reinforcing bars are continued to a level at which the shear stress in concrete reduces to a value the permitted by aCI-359 Code. In the lower portion of containment structure wall, supplementary reinforcing 3.8-2 Amendment 13 o 6/30/751

                                                                         ,,a

SWESSAR-P1 steel, normal to the face of the wall, is provided to resist splitting of the concrete in the plane of the vertical reinforcing bars. The tangential shears resulting from earthquake loading are resisted by concrete and diagonal reinforcing as shown in Fig. 3.8.1-2 and discussed in Sections 3.8.1.4 and 3.8.1.5. Minimum concrete cover f or all principal reinforcing steel of the containment structure equals or exceeds the requirements of ACI 318. The largest and principal reinforcing bar is N18, 4 therefore, the code requires a minimum concrete cover of 2 in. . Fig. 3.8.1-3 shows a typical detail of reinforcing steel in the founaation mat and base of wall. Fig. 3.8.1-4 shows a typical detail of the dome-cylinder junction. Fig. 3.8.1-5 shows the detail of the concentric steel ring embedded in the conert*e at the apex of the dome. Section 3.8.1.4 describes, and Fig. 3.8.1-18 and 3.8.1-19 show, the reinforcing steel arrangement at hatch openings. 3.8.1.1.3 Steel Liner and Penetrations The steel liner plate, penetrations, and access openings are described below. Liner Plate The liner plate consists of a steel cylindrical portion, closed at the top by a hemispherical dome and at the bottom by a floor liner plate. The liner plate includes enbedments, insert plates, overlay plates, and anchors. The liner plate acts as a gas-tight membrane under any one of the conditions that can be encountered throughout the operating life of the plant. The liner plate is anchored to the concrete containment at sufficiently close intervals so that the overall deformation of the liner is essentially the same as that of the concrete containment. It is protected from potential interior missilcs by interior concrete shield walls. The liner plate consists of the following:

a. Cylindrical Portion and Dome The cylindrical portion of the liner plate is in 'he c form of a vertical circular cylinder attached to a skirt anchored to the foundation mat at its base, see Fig. 3.8.1-6. The top of the cylindrical port ion is closed by a hemispherical dome. The liner plate has the dimensions as given in Section 3.8.1.1.1.

3.8-3 Amendment 4

                                              * [ h, 11/1/74

SWESSAR-P1 Lir a plate seams ar e covorod with small steel channels welded continuously along tne edges of their flanges to the liner plate. Theac channe2.s are zoned into test areas by dams welded to the ends of the sections of the channels. The channe l s are ased to check the leaktightness of welds during lint.r erection and to ensure that the overall lea: rcce test requirement is 4

    ' met on completion. Access   to    all test channels is permanently     installed       to     allow   testing     after construction. Should the overall leak rate exceed the
    ' specification in initial or periodic tests, the channels are used to assist in locating the source of increased leakage. Typical liner plate details are shown in Fig. 3.8.1-6 through 3.8.1-9.

b. Floor Liner Plate The bottom of the liner plate consists of flat steel floor liner plates welded together and anchored to the top of the mat concrete. The floor liner plate is anchored to the concrete base mat by continuously welding the plates at their periphery to cruciform steel inserts cast in the reinforced concrete base nat, Fig. 3.8.1-8.

The floor liner plate is 1/4 in. thick with the exception of areas where the transfer of loads through it requires a reinforced thickness. Floor liner plate seams are covered with small steel channels as described for the wall liner plate. Protection against missiles and heat is afforded by a layer of concrete over the floor liner plate.

c. Embedments Three dif ferent types of enbedments are used to maintain the leaktightness of the steel membrane while transferring loads across the floor liner plate to the concrete mat.

1. Corner Transition Section The cylindrical portion of the liner plate is welded to a skirt ring that is, in turn, welded to a skirt anchor as shown in Fig. 3.8.1-6. Eoth the skirt ring and skirt anchor are embedd ed in concrete.

9 3.8-4 Amendment 4 11/1/74 bb2 ,I//

SWESSAR-P1

2. Bridging Plates A concrete slab covering the floor liner plate is anchored through the steel liner plate by 7x 1/2 in. bridging plates and reinforcing bars, ds shown in Fig. 3.8.1-6 and 3.8.1-8. These plates form an integral part of the steel liner and contorm to the material and worknanship specifications of the steel liner.

3. Eridging Bars In those areas within the containment where there is major equipment or interior reinforced concrete walls to be anchored to the concrete mat, bridging bars are used. The bridging bars are used to pass the reinforcing bars through the floor liner plate.

These reinforcing bridging bars are welded to the top and bottom surface of the bars as shown in Fig. 3.8.1-7, thus providing reinforcing l'ar continuity without creating multiple penetrations through the floor liner plate.

d. Insert and Overlay Plates The loads derived from the support of piping such as the spray headers or other misce'laneous equipment are trans ferred to the containment concrete wall through insert plates and their anchors. The anchors are designed in number and size for each insert plate to be within the limits specified for the anchor studs for the liner as shown in Table CC-3700-2 of ACI 359. 4 The thickness of each insert plate is designed to provide a rigid base for the attached anchor stuas and pipe supports.

Sufficient anchorage is provided such that the liner plate adjacent to the insert plates is isolated from loads applied to the pipe supports. Overlay plates are welded to the liner plate for the attachment of the supports for small piping, electric conduit, and cable trays.

e. Anchors The steel containment liner and insert plates are anchored to the concrete wall and dome with steel anchor studs and deformed anchor bars. The anchorage layout is in a diamond pattern. The location tolerance of each anchor is 1 1/2 in. in any direction from its

{) theoretical location as dimensioned on the erection o 3.8-5 bb.- I' Amendment 4 11/1/74

SWESSAR-P1 dravings in order to clear possible interferences with reiatorcement bars or other embedded parts. The insert pli.te anchor layout is based on the applied pipe support loads. To verify the capabilities of the anchor studs, tests sere conducted at Northeastern University, Boston , Massachusetts, using 5/8 in. diameter studs and 3/8 in. thick plate. These tests showed that shear failure occurred in the stud adjacent to the weld connecting the stud to the plate. In no instance was the plate damaged. Tests conducted by one stud manufacturer indicate that, with proper spacing and the manufacturer's recommended depth of embedrent of the stud in concrete, the ultimate strength of the stud material can be developed in direct tension. Penetrations Penetrations are ustd to carry piping, mechanical systene, and electrical services through the containment walls. Containment penetrations are anchored to the reinforced concrete containment wall so that loads can be transf erred from the plaing or sleeve to the reinforced concrete. These penetrations can be classified as follows:

a. Piping System Penetrations Two basic types or penetrations are used for piping systems:

1. Unsleeved -

These penetrations consist of piping installed through the containment wall without a sleeve around the outside of the process piping. Unsleeved penetrations are used for thermally cold piping systems (temperature of the fluid in the piping less than 150 F) when only one pipe is passing through the penetration. The process piping is welded to reinforcement plates, as sho'm in Fig. 3.8.1-10, that will be anchored to the concrete wall.

2. Sleeved -

These penetrations have a sleeve around the outside of the piping. Sleeved penetrations are used for all multiple piping systems passing through one penetration and for all thermally hot (temperature of the fluid in the piping more than 150 F) piping systems, both single and multiple 4 The sleeve is welded to reinforcement plates

               ' anchored to the containment reinforced concrete so 3.8-6               27    iI#' C;   Amendment 4 b'U'             11/1/74

SWESSAR-P1 that the piping loads can be :.ransferred to the containment wall. Thermally hot piping is insulated to prevent the temperature of the concrete adjacent to the sleeve from exceeding 200 F. A water-cooled cooling unit l13 is also installed as shown in Fig. 3.8.1-10 for thermally hot penetrations in which the insulation alone would not be sufficient to maintain the concrete within the allowable temperature limit. The cooling unit is located on the inside of the penetration encompassing the full width of the concrete wall. The cooling water circulation pipes do not require any secondary penetration of the containment structure. Thermally hot penetrations using water-cooled jackets are designed with an air space between the insulation and the cooling unit to min.unize heat transfer to the concrete. The cooling unit limits the radial heat flow resulting from convection and radiation from the thermally hot pipe penetration and keeps the temperature of the concrete in contact with the sleeve at or below 200 F. In addition, it controls l13 the longitudinal heat flow resulting from conduction from the same heat source. This limits the temperature of the liner and temperature gradient along the sleeve. Consequently, values of thermal stress in the liner plate and sleeve are kept within allowable limits as shown in Tables CC-3700-1 of ACI-359 and 3.8.1-2 respectively.

b. Mechanical System Penetrations

1. Fuel Transfer Tube Enclosure - A fuel transfer tube penetration is provided for fuel transfer between the refueling canal in the containment structure and the fuel pool in the fuel building. The penetration consists of a stainless steel pipe installed inside an enclosure as shown in Fig. 3.8.1-11. The inner pipe acts as the transfer tube and connects the containment refueling canal with the spent fuel pool. The enclosure is welded to the containment liner and provision is made for Freal gas leak testing of all welds essential to the integrity of the penetration. Bellows expansion joints are provided on the enclosure to compensate for any differential movement.

3.8-7 {U'

                                        <0     i e' ^

3Ud Amendment 13 6/30/75

SWESSAR-P1 A blind flange with double gaskets and leakage test tap are bolted on the containment end of the enclosure during reactor operation. There are two main requirements for the fuel transfer tube enclosure:

1. The bellows must accommodate the maximum deflections, including rotation and offset, between the fuel pool and the containment refueling canal.

2. One end of the enclosure is inside the containment and must withstand external pressure and temperature during the test and LOCA conditions.

The bellows is selected on the basis of deflections caused by thermal expansion, seismic motions, and radial movement of the containment wall due to internal pressure and temperature.

c. Electrical Service Penetrations Electrical penetrations are used to carry electric cables and instrumentation leads.

Electrical conductors penetrating the containment structure range in size frcxn No. 20 AWG thermocouple leads to 1 in. diam solid copper rods for 1,800 V power circuits. Each penetra' ion assesnbly .sses through either 12 or 18 in. diam steel pipe slet ,es as shown in Fig. 3.8.1-12. The sleeves are welded into the cor' tainment liner reinforcement plates with a leak test angle around the inner reinforcement plate seal weld and a leak test channel around the flange seal weld. The electrical leads are installed in penetration assenblies which are bolted to the electrical penetration sleeve as shown in Fig. 3.8.1-12. The penetration assembly is mounted to the pipe sleeve ,l by a bolted or welded flange connection. The assenblies I also have provisions for leak testing. Each electrical assembly is periodically tested for leaktightness. Access Openings The containment contains the following access openings:

1. Equipment Hatch The equipment hatch is a single closure hatch with an inside diameter of 21 ft as shown in Fig. 3.8.1-13.

3.8-8 Amendment 4 662  ; c f '"

SWESSAR-P1 This hatch is equipped with one hatch cover mounted on the inside of the containment structure. l4 The hatch cover is double gasketed with a leakage test tap between the O-rings. The enclosed space between the O-rings can be pressurized to containment design pressure to test for leakage through the access door when it is bolted in place.

2. Personnel Air Locks Two personnel air locks are installed for entry into the reactor containment structure with hatch covers at both ends as shown in Fig. 3.8.1-14. Beih personnel air locks are double closure penetrations. Each closure head is hinged and double gasketed with a leakage test tap between the 0-rings. The enclosed space between the O-rings can be pressurized to containment design pressure to test for leakage through the access door when it is locked in place. For testing, the personnel air locks can be independently pressurized up to containment design pressure.

Both doors are hydraulically latched, but manually swung after the latch is released. The doors are interlocked so that, in the event one door is open, the other cannot be actuated. Each door is furnished with a pressure equalizing connection. The equalizing valves are manually or pushbutton operated by the person entering or leaving the personnel air lock. 3.8.1.2 Applicable Codes, Standards, and Specifications Structural design, materials, and materials quality control conform to the following codes, standards, and specifications unless otherwise s tated. In all cases the requirements of ACI-359 are met wi h the exceptions noted in Sections 3.8.1.5.1 4 and 3.8.1.6.1. 3.8.1.2.1 Concrete Structure

a. ACI 318-71 " Building Code Requirements for Rein-forced Concrete"

b. ACI 301-72 " Specification for Structural Concrete for Buildings" 4

c. ACI 211.1-70 " Recommended Practice for Selecting Proportions for Normal Weight Concrete"

d. ACI 214-65 " Recommended Practice for Evaluation of Compression Test Results of Field Concrete" 3.8-9

                                                   '    eP) tU-       Amendment 4

[. [ 11/1/74

SWESSAR-P1

e. ACI 304-73 " Recommended Practice for Measuring, Mixing, transporting, and Placing Concrete"

f. ACI 305-72 " Recommended Practice for Bot Weather Concreting"

g. ACI 306-66 " Recommended Practice for Cold Weather Concreting"

h. ACI 309-72 " Recommended Practice for Consolidation of Concrete"

i. ACI 347-68 " Recommended Practice for Concrete Form-work"

i. ACI 359-74 " Code for Concrete Reactor Vessels and Containment - Subsection CC-1000 to 13 CC-6000," January 1, 1975 (ASME Boiler and Pressure Vessel Code, Section III, Division 2-1975 Edition

k. AISC " Specification for the Design, Fabrica-Specification tion, and Erection of Structural Steel" including Supplements 1, 2, and 3 of the American Institute of Steel Construction.

1. AISI A numbered material Standard of the Specification American Iron and Steel Institute

m. AWS D1.1-72 "The Structural Welding Code"

n. AWS D12.1-75 " Recommended Practices for Welding Rein-13l forcing Steel, Metal Inserts, and Connections in Reinforced Concrete Construction"

o. Occupational Safety and Health Standards October 18, 1972 U.S. and Department of Labor, 13 Occupaticnal Safety and Health Administratic n. Proposed Walking-Wrking Surfaces Sept. 6, 1973.

p. NRC Regulatory Guides as qualified in Appendix 3A on the following topics:

Cadweld Splices (1.10) Section 3A.1-1.10 Reinforcing Bar Testing (1.15) Section 3A.1-1.15 Structural Acceptance Testing (1.18) Section 3A.1-1.18 g Examination of Liner Welds (1.19) Section 3A.1-1.19 Placement of Concrete (1.55) Section 3A.1-1.55 Design Response Spectra (1.60) Section 3A.1-1.60 Seismic Damping Values (1.61) Section 3A.1-1.61 Concrete Radiation Shields (1. 6 9) Section 3A.1-1.69 3.8-10 Amendment 13 6/30/75 662 ;c

SWESSAR-P1 Design Basis Ternado (1.7 6) Section 3A.1-1.76 l13

q. State and local building codes as required.

Section 3.8.1.6 describes materials, special construction techniques, procedures, and the exceptions taken to the above listed codes. Structural specifications are written to comply with the applicable codes, standards, specifications, and procedures given in this SAR. The following is a summary of the principal plant structural specifications that are prepared for Seismic Category I materials. Furnish Reinforcing Steel Reinforcing Bars - ASTM A 615, Grade 40 and 60 - Supplement S1 dated December 1972 - See Section 3.8.1.6 13 Reinforcing Bars - ASTM A 706 dated November 1974, see Section 3.8.1.6 Welded Wire Fabric - ASTM A 185 Bar and Rod Mats - ASTM A 184 Detailing and Fabrication Preparation of Ends of Rebars for Welding Quality Assurance Program - testing, inspection, and documentation Furnish Radial Shear Bar Assemblies Inclined Flat Shear Bars - ASTM A 242 *ype 2, ASTM A 441, ASTM A 572, or ASTM A 588 having a minimum yield strength of 50 kai Feinforcing Bars - See Section 3.8.1.6 f13 Filler metal for welding - AWS D1.1 Table 4.1-1, low hydrogen Cadweld sleeves - see Section 3.8.1.6 Fabrication - AWS D1.1 and AWS D12.1 Rebuilding of shear assemblies Quality Assurance Program - testing, inspection, and documentation 3.8-11 (, f' I

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                                                     )bu     Amendment 13 6/30/75

SWESSAR-P1 Mixing and Delivering Concrete Materials Portland cement - ASTM C 150, Type II Air mtraining agent - ACI 301, ASTM C 260 Aggregates - ACI 301, ACI 359, ASTM C 33 Aggregate storage - ACI 359 Water and ice - ACI 301, ACI 359 Chemical admixtures - ACI 301, ASTM C 494 Proportioning concrete - ACI 301, ACI 304, ACI 211.1 Batching and nixing concrete - ACI 301, ACI 304, ACI 359 Delivery - ACI 301, ACI 304 Hot weather requirements - ACI 305 Cold weather requirements - ACI 306 Quality assurance program - tests, inspections, and documentation concrete Testina Services Gather aggregate samples - ASTM D 75 Test water and ice - ACI 359 Test fine aggregates - ACI 359 Test coarse aggregates - ACI 359 Petrographic examination of aggregates - ASTM C 295 Test cement - ASTM C 150 Design concrete mixes - ACI 211.1, ACI 301 Test concrete (See Section 3.8.1.6) Compressive strength - ACI 301, ASTM C 31, ASTM C 39 Slump - ACI 301, ASm C 143 Air content - ACI 301, ASTM 231, or ASm C 173 Unit weight - ASTM C 138 Air and concrete temperatures - ACI 301, ACI 305, ACI 306 Density (specific gravity) ASTM C 642 Hardened conrete - ACI 301, ASTM C 42 Test grout - ASTM C 109 Evaluation of concrete strength - ACI 301, ACI 214 (See Section 3.8.1.6.1) Quality Assurance Program - tests, inspections, and documentation Placing Concrete and Reinforcing Steel Cadweld splices - see Section 3.8.1.6 Welded splices - see Section 3.8.1.6 Storage, handling, and use of electrodes onsite-AWS D1.1 Formwork - ACI 347 and ACI 301 Placing reinf orcement - ACI 301 and ACI 318 Concrete protection for reinforcement - ACI 318 Concrete construction, expansion and control joints - , , , - ACI 301 {gg iUs Compressible material for rattlespaces Water stops 3.8-12 Amendment 13 6/30/75

SWESSAR-P1 Anchor bolts and miscellaneous steel - ASTM A 307, AISI 1035 Grade A Carbon Steel, AISC Specification Inserts, sleeves, and pipes - ACI 301, ACI 318, and AIFC Specification Concrete placing - ACI 301, ACI 304, ACI 305, ACI 306, ACI 318 Cold weather requirements - ACI 306 Hot weather recuirements - ACI 305 Concolidation of concrete - ACI 301, ACI 309 Finishing of concrete lift surfaces - ACI 301 Concrete in blockouts Depositing underwater - ACI 301 Watertight concrete - ACI 301 Concrete in seawater - ACI 301 Routine tests of concrete - see Section 3.8.1.6 Repair of surface defects - ACI 301 Finishing of formed and flat surf aces - ACI 301 Curing and protection - ACI 301 Damp proofing Grouting Quality Assurance Program - tests, inspection, and documentation Structural Steel Shop detail drawings e Inspection and tests Calculation of weights Steel - ASTM A36, AS'IM A440, ASTM A441, ASTM A572, ASTM A58 8, ASTM A24 2 Bolts - ASTM A325, ASTM A490, ASTM A307 , ASTM A193 l,3 3.8.1.2.2 Steel Liner, Penetrations, and Access Openings

a. ASME Boiler and Pressure Vessel Code, Sections II and IX.

b. ASME Boiler and Pressure Vessel Code Section III Division 2 (referenced as ACI 359-74, January 1, 1975) and Division 1 as 13 required by Division 2.

c. AISC Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings, February 12, 1969, including Supplements No. 1, November 1, 1970, and No. 2 ,

December 8, 1971.

d. AEC Regulatory Guide 1.19 as qualified in Appendix 3A on the examination of liner welds - Section 3A.1-1.19.

e. 10CFR50 Proposed Appendix J " Reactor Containment Leakage Testing for Water Cooled Power Reactors" as published in the Federal Registe r , Vol. 36, No. 167, August 27, 1971, Pages 17,053 to 17,056. Details of Type A test (performed on liner plate) and Type D (performed on penetration and access openings) are covered in Section 16.4.4.

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                                          /'n iOb Amendment 13 6/30/75

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f. ASME Code Stamping - the piping penetrations are Code stamped Class 1 for Class 1 piping systems and Class 2 for Class 2 piping systems. The personnel air locks are code stamped Class MC. The stamps are applied in accordance with ASME III .

3.8.1.3 Loads and Loading Combinations 3.8.1.3.1 Concrete Structure Design of the concrete structure complies with ACI 359 Article CC-3000 with the exceptions noted in Sections 3.8.1.5.1 and 3.8.1.6.1. The design is based on:

a. The temperature and pressure generated by the DBA, Section 6.2.2.

b. The SSE and OBE discussed in Sections 2.5 and 3.7

c. Severe weather phenomena, Section 3.3

d. Biological shielding requirements The containment operates at atmospheric pressure. The temperature of the containment air fluctuates between a normal maximum temperature of 105 F and a minimum of 60 F during shutdown, depending upon the ambient temperature of available service water.

Design load criteria based on ACI 359, and others given below, conform to current containment design practice. The combination of dead, pressure, temperature, and earthquake or tornado loading expressed in the criteria contains varying load factors for pressure, temperature, and earthquake forces. The total loading resulting fram the sunmation of any one of the combinations may cause a maximum stress condition depending on the type of stress and member under ccmsideration. DBA pressure loads, and the resulting rise in liner temperature, are described in Sections 6.2.1 and 6.2.2. Missile loads are described in Section 3.5. Normal operating temperatures are described in Section 9.4.5. Earthquake loads are described in Section 3.7. Wind loads, including tornado, are described in Section 3.3. The earthquake loadings include consideration of simultaneous excitation from two horizontal and one vertical set of axes that a re orthogonal to each other. This is deceribed in 9 Section 3.7. 2.1.1. gr9 .n; U O, L iU. 3.8-14 Amendment 9 4/30/75

SWESSAR-P1 3.8.1.3.2 Steel Liner and Penetrations Liner Plate Load combinations for the design of the liner pl.,te, anchors, and embedments are given in Table CC-3200-1 of ACl-35:*. The loads in Table 3.8.1-1 apply to the design of insert and overlay plates. Fatigue effects are considered, where necessary, using the 4 methods of ASME III. Penetrations

a. Piping System Penetrations Table 3.7.3-5 for Seismic Category I piping applies to the following:

               - unsleeved pipe penetrations
               - integrally forged process pipe and flued head of        4 sleeved penetrations.

Sleeves are considered part of the containment boundary as shcun in Figure NE-1132-1 of ASME III. As such, they are desianed for the load combinations of Table 3.8.1-2 l4 which is based on Article NE-3000.

b. Mechanical Systen Penetration The fuel transfer tube enclosure is part of the containment boundary. The expansion bellows are designed for the most severe combinations of seismic deflections plus containment pressurization and temperatures. The bellows are connected by a cylinder which is welded to the liner. The cylinder is designed for the loads listed in Table 3.8.1-2.

l4

c. Electrical Service Penetrations Where electrical penetrations form a part ot the containment botmdary, they are designed for the load combinations in Table 3.8.1-2. These loads conform to l4 the requirements of ASME III, Article NE-3000.

As required by Regulatory Guide 1.63, Position C.2, the electrical penetrations are designed for the containment design pressure, Pd. This pressure includes a safety margin over the maximum pressure, P", as required by footnote 1 to Article NE-3000. Access Openings 662 100 3.8-15 Amendment 4 11/1/74

SWESSAR-P1 Load combinations for access openings are listed in Table 3.8.1-2. As required by ACI 359, par CC-3031.1, these cafainations comply with the requirements of ASME III, Article NF -3 00 0. The attached liner is designed for the same loading combinations as the liner plate, Table CC-.3200-1 of ACI-359. 3 . 8 .1. 4 Desion and Analysis Procedures 3.8.1.4.1 Concrete Structure The design, analysis, and construction of the concrete containment structure are similar to the following plan ts designed and constructed by Stone & Webster Engineerina Corporation: Connecticut Yankee Atomic Power Company, Connecticut Yankee Atomic Power Plant, Unit No. 1 (Docket No. 50-213) Virginia Electric and Power Company, Surry Power Plant, Units 1 and 2 (Dockets No. 50-280 and 50-281) Maine Yankee Atomic Power Company, Maine Yankee Atomic Power Station (Docket No. 50-309) Duquesne Light Company, Beaver Valley Power Station, Unit No. 1 (Docket No. 50-334) Virginia Electric and Power Company, North Anna Power Station , Units 1 and 2 (Dockets No. 50-338 and 50-339) Virginia Electric and Power Company, North Anna Power Station, Units 3 and 4 (Docke ts No . 50-404 and 50-40 5) Northeast Utilities Service Company, Millstone Nuclear Power Station, Unit 3 (Docket No . 5 0-4 2 3) The containment st ructure consists of a hemispherical dome, a cylindrical shell, and a mat supported on an elastic subgrade. The containment ma t is part of the nuclear island foundation. 3 The reactor vessel is surror.nded by the primary shield wall. Secondary shield walls form the concrete cubicle walls that enclose the steam generators and reactor coolant pumps. Outside the containment st ructure is the concentric annulus building wall. *his structural arrangement consists of a set of four axisymmetric shells supported on a circular foundation slab as shown in Fig. 3.8.1-15. This structural system is analyzed and designed for the loading conditions given in Section 3.8.1.3. The dynamic loads are represented by their static equivalent effect. The effect of pressure loading on the containment shell is to alter the shell stiffness due to cracking of concrete. The UUc ;Ul 3.8-16 Amendment 8 3/28/75

SWESSAR-P1 change in the shell stiffness is taken into account by introducing equivalent shell thickness and modulus of elasticity. Analysis is performed using the computer programs SHELL 1 and MAT

6. The SHELL 1 program is a further development of a cmputer program written at AVCO Corporation. It is based upon the general numerical procedure proposed by B. Budiansky and P. P. Radkowski (References 2, 3) to analyze a shell of revolution subjected to arbitrary loadings. The analysis is based upon the (general first-order) linear theory of shells by J. L. Sanders, Jr. (Reference 4).

The results for pressure loads on a shell of revolution have been checked aga. inst hand calculations based upon " Theory of Plates and Shells" by Timoshenko and Woinowksky-Krieger, Second Edition, 1959 (Reference 1) . Actual containment shell problems were used for the purpose of validation of the SHELL 1 program. Plots of the meridional bending moment, meridional shear, and radial deflection of the containment wall showing the comparison between SHELL 1 and hand calculations based on Reference 1 are included for a containrr 7nt structure subjected to uniform internal pressure (Fig. 3 . 8 .1 -16 and 3 . 8 .1-17) . Further information on SHELL 1 is given in Appendix 3B. The MAT 6 program analyzes circular plates, supported on elastic half-space, under axisymmetric loading. The progran is described g in Appendix 3B. MAT 6 is used to analyze the containment mat under axisymmetric loading. The subgrade ie represented by springs attached to the surface of the shell or plate. The vertical subgrade repring stiffness is calculated frm the elastic half-space theory. The horizontal subgrade stiffness due to embedment is accounted 9 for by representing the adjacent soil by equivalent springs. The elastic parameters of the subgrade considered for the design are not a ssociated with any particular site; these are indicated in Section 2.5.2. Typically, the following values of the subgrade shear modulus are used: G= 6,000 psi G = 24,000 psi G = 300,000 psi G= 1,000,000 psi The range of variation of the subgrade elastic parameters is so wide that any fine adjustments to the assumptions involved in the elastic half-space Cwory associated with a particular set of foundation parameters do not alter the design values. Two types of temperature mnditions are considered: 3.8-17 Amendment 9 4/30/75 4 o (a, ,. tig

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a. Temperature under normal operating conditions.

b. Temperature associated with the DBA. This is the transient temperature which, when combined with the coincident internal pressure, produces the most adverse effects on the containment structure.

Under normal operating conditions, the temperature gradient producing the worst stress resultant is used. Under DBA conditions it is assumed that the liner temperature increases while the concrete remains at its ambient temperature. Since expansion of the liner is limited by the concrete shell, pressure develops between the shell and the liner. The equivalent pressure that is exerted by the liner on the concrete shell is given by p, LH L,_ c R L where: Pc = equivalent pressure h t = thickness of liner R g = Radius of liner a LH = s U where: Es = Young's modulus for the steel liner a = coefficient of thermal expansion of the liner 6T = change in liner temperature due to the DBA A short distance above the mat, radial notion is not completely constrained and the effect of the mat-to-shell discontinuity is n egligible. For these conditions, expression for the stresses in the shell and liner are written in terms of the meridional and circumferential liner strains. Conditions of static equilibrium then yield an explicit solution for the liner strains, from which a g is determined. hh2 , 3.8-18 Amendment 9 4/30/75

SWESSAR-P1 The seismic analysis of the containment structure described in Section 3.7.2 yields the static equivalent forces which are imposed on the overall structure foundation arrangement shown in Fig. 3.8.1-15. The effect of each direction of seismic excitation is analyzed separately using the SHELL 1 program. The effects of these separate earthquake loadings are combined by taking the square root of the sum of the squares of indiviaual etfects. The tangential shears caused by the seismic loading are resisted by both the concrete and a system of diagonal reinforcing barn. The allowable tangenti.al shear stress in concrete is described in Section 3.8.1.5. Diagonal reinforcement, as shown in Fig. 3.8.1-2, is provided for tangential shear force in excess of that allowed in the concrete. In calculating the steel requirement, compatibility of strains is considered so that the effect of stresses induced by other loads is included. It is assumed that the liner provides no shear resistance. The overall structural response of a portion of the containment shell subject to both LOCA and Y r+ Y m +Y j is determined by means of a elastic analysis through the use of the Shell-1 computer 4 program with appropriate Fourier coefficients to represent the applied general and local static equivalant loads. The containment shell is designed to ensure its integrity as a leak tight barrier under the ef f ects of the above forces. Tornado wind loading of the containment structure is described in Section 3.3.2. Penetrations through the containment structure are divided into three categories as follows: ar Pipe penetrations 12 in. in diameter, or less. No special concrete reinforcing is provided. Penetrations in this category are located to avoid interference with the reinforcing steel.

b. Pipe ,enetrations greater than 12 in. and up to 4 ft in diameter. Supplementary reinforcement is provided in amount and distribution such that the area of steel requirements for concrete reinforcement is adequately satisfied.

For this size penetration, reinforcing steel interrupted by the opening is terminated at each side of the opening. Supplementary reinforcement is placed parallel to the bars which are interrupted. Horizontal, diagonal, and vertical bars are used to frame the opening. The total area of reinforcement provided in any plane is not less than twice the area of steel which is interrupted by the opening, with one-half of this @ placed on each side of the opening.

                                                  ,.      nD 3.8-19        hb[       t          Amendment 4 11/1/74

SWESSAR-P1 Anchorage of the additional reinforcement is determined 4 in accordance with Subarticle CC-3531.1.2 of ACI 359. This design approach is consistent with the practice used and pressure tested in the Connecticut Yankee containment structure at Haddam Neck, Connecticut, and the Surry Power Station Unit 1 containment structure at Surry, Virginia.

c. Openings larger than 4 ft 0 in. diameter. The openinos, in this category are the 7 f t diameter personnel hatches dnd the 21 ft diameter equipment hatch. Details of the reinforcement provided around these hatch openings are shown in Fig. 3.8.1-18 and 3.8.1-19.

These penetrations are analyzed by means of the three-dimensional finite element capability of the computer program known as STRUDL II (5) described in Appendix 3B. Because the hatches are sufficiently removed from discontinuity points such as the mat and dome, the finite element discretizations selected for these analyses are a quadrant. This quadrant has a maximum radius approximately equal to 2.5 times the radius to the outer edge of the thickened ring beam in order to eliminate the effects of the hatch opening and ring beam at the quadrant's outer boundary. A discretization of quadrilateral shapes emanating radially from the center of the opening is used to provide a fine arid in the meridional and circumferential shell directions. Three elements are taken through the typical wall sections, and additional elements are added on each side of the typical wall section to account for the thickened ring beam section. The STRUDL II computer program allows the modeling of the structural characteristics of the wall (cracked versus uncracked) and the inclusion of the liner for the various load ing s , including temperature. The liner, however, is assumed not to con tribute to the structural capacity of the ring beam or wall for any loading condition other than the containment structure structural acceptance test. During an accident, the liner is normally in a state of compression due to the sharp temperature rise inside the containment structure and, therefore, adds load to the ring beam. In order to obtain a more realistic assessment of the strains, displaconents, and stresses in the area of the hatches, the ring beam and cylindrical wall are assumed to be fully cracked and to have the extensional stiffness of only the reinforcing bars. The STRUDL II program (5) yields the discontinuity effects between the cylinder wall and the thickened ring beam and the pattern of the membrane forces in the region of the hatch 3.8-20 tD i)3 Amendment 4 bUL ' 11/1/74

SWESSAR-P1 openings. Additional reinforcement :.__:cumferential, meridional, and diagonal) is provided in auch regions where a significant increase over the typical membrane forces Un ridional and circumferential) occurs. The principal circumferential and meridional reinforcing bars are extended to the inner face of the ring beam and either bent at right angles and Cadwelded to each other, or attached to flat plates, thereby providing additional shear resistance which is not considered in the design. Construction procedure and pour sequence are arranged to prevent accumulation of shrinkage in the containment structure. However, any shrinkage tends to produce stresses in a reverse direction compared to the DbA condition. Hence, shrinkage effects are ignored as inconsequential to the design condition. Material quality control procedures, as described in Section 3.8.1.6, assure that minimum strength material requirements are achieved. 3.8.1.4.2 Steel Liner and Penetrations Computer programs mentioned in this subsection are fully described in Appendix 3B. Liner Plato

c. Cylinder Portion and Dome The liner plate in the cylinder and dome portions of the containment wall is analyzed to obtain strains using a omputer program such as " Stress Analysis of Shells of Revolution" which treats orthotropic, axisymmetrically loaded c' ells. The orthotropic capability is used to model the reinforcing bar array as an equivalent orthotropic shell.

b. Floor Liner Plate The lower portion of the wall liner and the attached portion of the floor liner plate are analyzed together using a finite element cwaputer program such as "ASAAS" (Asymmetric Stress Analysis of Axisymmetric Solids) .

This program permits a detailed evaluation of this discontinuity including the ef fects af the surrounding concrete under the symmetric and asymmetric loadings 4 given in Table CC-3200-1 of ACI 359. Bounda ry conditions for the wall liner portion are taken f rom the calculations of the cylindrical portion described in (a) above. a n f: 3.8-21 (3f,n/

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c. Unbedments The corner transition section embedment is included in the analysis of the floor liner plate (b).

d. Insert and Overlay Plates, Brackets, and Attachments Manual calculations are employed in the analyses of insert and overlay plates. Brackets and atta ch nents connec&ed to the containment liner are analyzed by technig4es such as those illustrated in AISC-1969,

        " Specification for the Design, Fabrication, and Erection of Structural Steel for Euilding."

e. Anchors Liner plate anchor studs are arranged in a diamond pattern and spaced to prevent buckling of the liner plate or tearing of the anchor-to-plate weld. A procedure such as that set forth in " Theory of Llastic Stability," 2nd Edition by Timmshenko and Gere, is used to evaluate the liner for buckl. .g due to combined axial and circumferential compressive stresses.

Anchors for insert plates are designed primarily using manual calculations. In highly stressed anchors and in designs where the anchor manufacturer's data are not sufficient, the interaction of the anchor, concrete wall section, and the plate is analyzed using a 2-D finite element program such as "SAAS III" (Stress Analysis of Axisymmetric Solids) . Penetrations

a. Piping System Penetrations Pipe penetration assemblies are analyzed for stress using a finite element computer code such as "ASAAS" (Asymmetric Stress Analysis of Axisymmetric Solids).

This program is capable of evaluating the effects of asymmetric loads as well as pressure and temperature loads. Temperature distributions at discontinuity areas exposed to operating conditions are evaluated using a finite difference computer code such as " TAC-2D" (Thermal Analysis Code - 2 Dimensional) . The temperatures as determined by TAC-2D are used as input for ASAAS to determine thermal stresses.

b. Mechanical System Penetration <

1 02 UUt i /J 3.8-22 Amendment 4 11/1/74

SWESSAR-P1 The cylindrical fuel transfer tube enclosure is analyzed using the finite element method. A program such as SAAS III can include the surrounding concrete plus symmetric loads such as pressure and tmperature. Asynnetric loads, such as dead load and earthquake, are analyzed using a program such as "SFET.T. 1" ('Itlin Shell of Revolution Under Asymmetric Ioading) .

c. Electriaal Service Penetrations Electrical service penetrations are analyzed using the same procedure described for the mechanicd service penetration.

Access openings Access openings are analyzed using the procedure described for the mechanical service penetration. In areas where a discontinuity requires detailed stress and strain calculations, a program such as ASAAS is used to evaluate symmetric and asymmetric loads. 3.8.1.5 Structural Acceptance Criteria 3.8.1.5.1 Concrete Structure The containment structure is checked by calculating stress leve.1s in order to confirm the damping factors used in the seismic analysis (see Section 3.7.1.3) . Tangential shear, V , which results from earthquake loading, is resisted by the concrete and diagonal reinforcing steel (see Fig. 3. 8.1-2) . The tangential shear stress carried by the concrete, V , is equal to: 12,000 p for p 5 0.01, r_nd 93 + 2,700 p for 0.01 s p 5 0.025 where v is given in psi and p is the lesser of the steel ratios in the vertical or circumferential directions. The maximum allowable tangential shear stress carried by the concrete, vc , does not exceed 40 psi for the operating basis earthquake and 60 psi for the safe shutdown earthquake. The excess of vu minus v c is resisted by diagonal reinforcement. This is an exception to ACI 359 subsection CC-34 21. 5. b Principal membrane reinforcement is proportioned to resist the active membrane stresses (meridional, circumferential, and excess shear corresponding to v minus v ) plus additional meridional and circumferential tension stresses, each corresponding to 1.5 times the shear associated with v (7)(a), 662 196 3.8-23 Amendment 13 6/30/75

SWESSAR-P1 Fatigue is not a significant factor in design because the number of maximum stress cycles in the life of the structure is small. Repeated reactor shutdowns and startups have no effect on the plant's life or margin of safety because of the small temperature range involved. The containment structure is, therefore, as capable of resisting design loads at the end of its service life as at the beginning. 3.8.1.5.2 Steel Liner and Penetrations Liner Plate

a. Cylindrical Portion and Dome Design limitations for the liner plEtc are given in 4 Table CC-3700-1 of ACI 359. These are applicable to the liner plate which is welded to penetrations and access openings, as well as plates without discontinuities.

b. Floor Liner Plate Strain allowables for the floor liner plate are givan in Table CC-3700-1 of ACI 359.

4I

c. Embedments Allowable strains for the corner transition area embedment are given in Table CC-3700-1 of hCI 339.

d. Insert and Overlay Plates, Brackets, and Attachments The strain limits of Table CC-3700-1 of ACI 359 are used as the criteria in the analysis of the insert plates.

For the Design I and II load cases of Table 3.8.1-1, the 4 normal and abnormal limits of Table CC-3700-1 of ACI 359 are used respectively. The allowables used for the Design I load case for overlay plates, brackets, and attachments are per AISC. The allowables us ed with the Design II load case are 1.5 times those used with the Design I load case. These strength considerations are based on paragraph CC-3750 4 of ACI 359.

e. Anchors Containment liner anchor allowables are given in Table 4l CC-3700-2 of ACI 359. These allowables are applied also to the anchors which are used to secure insert plates to -

the containment wall. () O L

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SWESSAR-P1 Penetrations

a. Pipe Penetrations Allowables for sleeved pipe penetrations are listed in Table 3.8.1-2. As noted in the table, these apply to g4 the sleeve and are derived from the requirements of Article Nb3000, ASME III. The flued head and the pipe, which are an integral forging, are designed to the for piping l4 criteria and by the analytical procedures described in Section 3.7.3.6. The applicable table for loading combinations and stress limits for Seismic Category I piping systems is Table 3.7.3-5. This includes the dynamic system loads associated with the DBA, except where rupture of the pipe line attached to the penetration is postulated.

Unsleeved penetrations are designed for the criteria listed in Table 3.7.3-5. The design criteria for pipe rupture plus SSE loads on sleeved penetra tion assemblies are listed in Table 3.8.1-3. The table is based on 85 percent of the l4 stress levels given in Appendix F of ASME III as discussed in paragraph NE-3131.2 for metal containments.

b. Mechanical System Penetrations Fuel Transfer Tube Enclosure - The stress allowables listed in Table 3.8.1-2 apply to the cylindrical section l4 of the tube enclosure for the area where it forms a portion of the containment boundary.

c. Electrical Service Penetrations Allowables for electrical service penetrations are given in Tuble 3.8.1-2 which is based on Article NE-300 0, l4 ASME III. Regulatory Guide 1.63 references IEEE Std 317-1972 as being an acceptable criteria.

It is Stone and Webster's position to comply with the intent of both the Regulatory Guide and the IE3E Standard. Accordingly, the criteria list ed in Table 3.8.1-2 reflect the requirements of Article l4 2.E-3000, particularly NE-3300 and NE-3700. Access Openings Design conditions for access openings are given in Table 3.8.1-2 and are based on Article NE-3000, ASFI III. O m iGO 3.8-25 b I. d I# Amendment 4 11/1/74

SWESSAR-P1 3.8.1.6 Materials, Quality Control, and Special Construction Techniques For applicable codes, standards, and specifications, refer to Section 3.8.1.2. 3.8.1.6.1 Concrete Materials and Quality Control ACI 301, ACI 318, ACI 347, and ACI 359 form the general bases for the concrete specification. 13 The concrete mixes used in the containment structure yielc a unit weight of at least 140 lb per cu ft, air and oven dried in 13 accordance with ASTM C642. Concrete protection for reinforcement, preparation, and cleanalg of construction joints, concrete mixing, delivering, placing, and curing, equals or exceeds the requirements of Regulatory Guide 1.55 with the exceptions given in Section 3A.1-1.55. All cement cc.nforms to requirements of portland cement of ASTM C 150, Type II, except that calcium aluminate (high-alumina) cement is used f or porous concrete enclosed within the waterproof monhrane under ttle reactor containment mat. This substitution is made to reduce the clogging of voids by continued hydration that occurs with portland cement, and to remove the corrosive effect of free lime that occurs in the early stages of hydration of portland cement. bb2 , ) ~) 3.8-26 Amendment 13 6/30/75

SWESSAR-P1 All concrete batching and placing equipnent is subject to the approval of the Engineers. 'tll batching and placing equipment and methods are reviewed for compliance with the specifications by the Senior Field Quality Control Engineer. Curing and protection of freshly deposited concrete conform to ACI 301, Chapter 12 except that curing compounds are not used on surfaces to which additional concrete is to be bonded. Should core tests be inconclusive or impracticable to obtain and structural analysis does not confirm the safety of the structure, load tests are performed. Concrete work judged inadequate by structural analysis or by load tests is reinforced with additional construction if so directed by SSW, or is replaced entirely. Statistical quality control of the concrete is maintained by a computer program. The program is based on an article in ACI Publication SP-16, " Computer Applications in Concrete Design and Technology." This program analyzes compression test results by the testing laboratory in accordance with methods established by ACI 214. Concrete Construction In general, concrete in the wall and dczne of the containment structures is placed in approximately 6 ft high lifts around the entire circumference. Three foot lifts may be used for two or more pours in the base of the wall. Each lift is constructed in not more than 18 in. layers placed at such a rate that concrete surfaces do not reach their initial set before additional concrete is placed. Past experience indicates that the use of properly controlled concrete mixes, and pours not exceeding 18 in. as described above, followed by careful concrete curing of each lift controls shrinkage sufficiently to provide the necessary quality in the finished concrete. Forms are used on the exterior of the concrete dome to a line about 50 degrees above the horizontal. The permanent steel liner serves as the inner form for concrete. For the area where exterior forms are used, the concrete joints lie in horizontal planes. Above the 50 degree line, the remainder of the dome concrete is cast in one lift without any joints. 3.8-27 Amendment 13

                                          //^    ^OO          6/30/75

SWESSAR-P1 3.8.1.6.2 Reinforcing Steel Materials, keinforcing bars conform to the requirements of ASTM A615 13 Grade 40, A615 Grade 60, or A706. Cadweld T--Series reinforcing steel splices are full tension splices manuf actured by Erico Products, Inc., Cleveland, Ohio, and are used to splice N14 and N18 reinforcing bars. In restricted areas, reinforcing bars are butt welded in a manner conforming to the requirements of AWS D12.1. Cadweld splices are made in accordance with the instructions for their use issued by the manuf acturer, Erico Products, Inc. O bb_ Oi g 3.8-28 Amendment 13 6/30/75

SNESSAR-P1 Construction Techniques

a. General Placing of reinforcino steel conforms to the requirements of Chapter'S of ACI 301, "S tructural 4

Concrete for Buildings ," and Chapter 7 of ACI 318,

            " Building Code Requirements for Reinforced Concrete."

Tack welding of designed reinforcing steel is not permitted. Structural ductility is maintained by staggering critical splices wherever' possible to assure that small adverse effects of multiple splices in the same plane co not occur. Full scale pressure tests, conducted on completed concrete containment structures in which Cadweld splices and welded splices were used in a similar manner to that proposed here, showed no stress concentrations or lack of structural ductility. Locations of splice groups were not discernible from inspection of the test crack patterns.

b. Preparation of Rebar for Cadweld Splices e The ends of the reinforcing steel bars to be joined by the Cadweld process are saw cut or flame cut. The ends of the bars are thoroughly cleaned of all rust, scale, grease, oil, water, or other foreign matter before the joints are made.

Quality Control

a. Reinforcing Bars For the N14 and N18 bars used in the containment structure, ingots and billets are traced with 4 identifying heat numbers. Bundles of bars are tagaed with a heat number as they come of f the rolling mill. A special mark is rolled into bars conforming to this special chemistry to identify them as possessing the chemical and mechanical qualities specified.

S&W inspectors witness, on a random basis, the pouring of the heats and the physical and chemical tests performed by the manuf acturer of the special chemistry reinforcina bars. Bars containing unacceptable inclusions or failing to conform to the required chemistry and physical requirements are rejected. n oc :oz 3.8-29 Amendment 4 11/1/74

SWESSAR-P1 Mill test reports showing actual chemical and physical properties are furnished to the Utility and SSW for each heat of steel used in making all reinforcing steel furnished.

   ,  3.8.1.6.3       Steel Liner and Penetrations Materials

a. Liner Plate All steel materials used in the f abrication of the liner plate are in accordance with approved materials as 4 listed in ASME II.

l Material for the liner plates, insert, and overlay plates is SA-537, Grade B, quenched and tempered, that has a specified minimum tensile strength of 80,000 psi, a minimum guaranteed yield strength of 60,000 psi, and a guaranteed minimum elongation of 22 percent in a standard 2 in. specimen. These plates are ordered to conform with SA-20 with regard to thickness tolerances. All plate materials are quenched and tempered. In addition, they are impact tested as required by ACI 359 to determine 4l embrittlement characteristics. g All steel materials as stated above are tested and certified to verify that their mechanical properties meet the requirements specified in the ASME Code, Section II. The steel liner will be subjected to a maximum neutron flux of 1.4 x 1017 nyt (E 20.55MeV) ; therefore no radiation uamage is expected. Stud welding material conforms with AWSD 1.1 which requires one prequalification test of each size stud. The duration or prequalification is indefinite so long as materials, fluxes, are shields, and stud base geometry are unchanged. In ad dition , manuf acturer 's certification that studs conform to the 4 prequalitication requirements is obtained. This is an exception to ACI 359 subsection CC-2620 which requires that qualification tests shall be conducted for each lot of studs.

b. Pen etra tion s The materials f or the dif f erent components for the pene-trations are listed below:

Materials for Penetrations Carbon steel plates SAS37 GR B quenched and tempered Carbon steel forgings SA350 GR LF2 or SA105 GR II 662 203 3.8-30 Amendment 4 11/1/74

SWESSAR-P1 Carbon steel pipe SA333 GR 6 or SA155 GR KCF70 Carbon steel pipe SA333 GR 6 or AST!i A516-GR 60 sleeves normalized Stainless sta 1 SA182 P304 forgings Stainless ste'( pipe SA312, Type 304/316 or SA240, I Type 304/316 4 Bellows expansion joint SA240, Type 304 or 321 Steel items, such as reinforcement plates around 4 containment openings, etc, except backing plates and anchors, gas testing channels, equipment hatch bolts, and equipment hatch nuts, are quenched and tempered.

c. Access Ooenings The materials listed below are used for the equipment hatch and personnel air locks:

Cylindrical shells ) Reinforcement plates } SA537 GR B or Hatch covers ) SA516 GR 60 or 70 Hatch cover flanges ) Steel materials as stated above are tested and certified to verify that their mechanical properties meet the re-guirements as specified in ASME II. Plate materials are quenched and tempered. In addition, the materials are impact mested as required by ACI 359 4 to determine embrittlement characteristics. These plates are ordered to conform with SA-20 with regard to thicknees tolerances. Quality Control Quality control procedures for the liner plate are describec in Section 17.1.2.2. Special Construction Techniques

a. Liner Plate Erection of the cylindrical portion of the liner plate follows completion of the concrete mat. The 3/8 in.

thick cylindrical portions are erected to approximately 9 202 ft above the mat. The liner plates serve as the 3.8-31 ,f9 4 () O L 7]dAmendment

                                                                 '11 /1/74

SNESSAR-P1 internal form for the concrete containment during construction. All liner seams are double butt welded, except for the lower 30 ft of the cylindrical shell liner where the liner plates are welded using backing plates. Details are shown in Figure 3.8.1-9. The liner plate is continuously anchored to the concrete shell with Lteel anchor studs. The 1/4 in. thick floor liner plates are installed on top of the concrete foundation mat during this period. The cylindrical portion is then completed, finishing with construction of the 1/2 in. thick steel dome, weld inspection , and gas testing. The liner is plumb within 3 in. at any height of the liner, measured from a theoretical vertical line extending up from the base ot the liner. The maximum plus or minus deviation from the true circular or hemispherical form is 1/2 percent of the nominal radius. The maximum misalignment between liner plates is in 4 accordance with paragraph CC-4523 of ACI 359. All measurements are taken on parent metal and not at welds. Flat spots or sharp angles are not allowed. Careful attention is given to the actual circumference of the shell to ensure that all shell rings match properly. The allowable deviation from true circular form does not arfect the elastic stability of the containment liner because of tha O restraint provided by the anchor studs tying it to the reinforced concrete shell.

b. Penetrations Unaleeved Piping Penetrations are shop fabricated and positioned in the containment wall before the con crete is poured.

Sleeved Piping Penetrations - The sleeve with shear lugs and reinforcement plates is shop fabricated and positioned in the containment wall before the concrete is poured. The Puel Transfer Tube Enclosure consists of a stainless steel pipe installed inside a sleeve (Fig. 3.8.1-11) . The outer pipe is welded to the containment liner and provision is made for leak testing of all welds essential to the integrity of the pen etration . Bellows expansion joints are provided on the outer pipe to compensate for any dif f erential movment between the two pipes. 7o mqE b02 uJ 3.8-32 Amendment 4 11/1/74

SWESSAR-P1 Electrical Service penetrations are shop fabricated and positioned in the containment wall before the concrete is pour ed .

c. Access Onenings The equipment hatch and personnel air locks are shop fabricated and placed in position in the containment wall before the concrete is poured.

4 3.8.1.7 Testina and Inservice Surveillance Requirements 3.8.1.7.1 Concrete Structure The pressure test to ensure the structural integrity of the con-tainment is performed after the liner is completed, the last concrete po ured , and all penetrations, sleeves, and hatches installed . The test equals or exceeds the requirements of Regulatory Guide. 1.18 and ACI 359 Article CC-6000. Minor 4 changes at which points are located for measuring deflections at the largest opening with thickened ring beam are established to account for the thickened ring as permitted by paragraph 3 of the Guide. (See Fig, 3.8.1-20.) Radial deflection is measured along six meridians at 13 ft 6 in. above the top of mat, at mid-height between mat and sprinaline, dnd at the springline of the dome. Vertical deflections are measured at the springline of the dome and the apex. Radial measurements are made using linear variable differential 4 transducers (LVDTs) supported from the containment internal s tructures . Vertical measurement is made using Invar tapes. Radial deflections for the largest opening with a thickened ring edge beam are measured as shown in Fig. 3.8.1-20. The change in diameter of the thickened ring is measured on the inside and outside edges of the horizontal and vertical diameters of the clear opening. Mapping of cracks is on exterior surf aces of the containrent at locations selected prior to start of the pressure application. Mapping is on one meridian line at three locations, and one location around the largest opening. Testing is not conducted under extreme weather conditions. The environmental conditions are measured and monitored to permit the evaluation of their contribution to the response of the containment. The testing sequence is repeated if the test pressure drops for unexpected reasons to or below the next lower 3.8-33 662 206 Amendment a 11/1/74

SWESSAR-P1 pressure level, or if significant modification or repcirs are made to the containment following the test. Following are the predicted deformations for the test of the con tainment structure at 1.15 times the containment design pressure:

a. Maximum vertical elongation at the crown of the dome:

1.03 in.

b. Increase in the containment diameter at 40 ft above the top of the mat: 1.22 in.

c. The maximum width of new cracks or increase in existing cracks: not more than 0.02 in. for horizontal c.acks and 0.026 in. for vertical cracks.

d. After containment pressure is reduced to atmospheric, the residual width of new cracks or increased width of existing cracks: less than 0.01 in.

The predicted deformations and crack widths are based on the integrated and compatible interaction between the concrete, the different layers of rebars, and the steel liner plate. Tensile strength of the concrete is ignored. A tolerance of 30 percent is applied to the predicted numbers. It is anticipated that the containment structure remains in the elastic range during the air pressure test and that no permanent distortion exists in the liner or in the concrete once the pressure is released. It is fully expected that there will be small shrinkage cracks in the concrete. In addition to the structural acceptance test for the containment structure, various leakage rate tests are performed. These 4 leakag e rate tests and inservice surveillance requirements are discussed in Section 16.4.4. 3.8.2 Steel Containment System A steel containment system is not used. The concrete containment structure is discussed in Section 3.8.1. 3.8.3 Concrete and Structural Steel Internal Structures of Concrete Containment 3.8.3.1 Description of Internal Structures The containment structure interior shown in Fig. 1.2-3 consists of heavily reinforced concrete walls and slabs designed to support the principal nuclear steam supply system equipment. Fig. 3.8.3-2 and 3.8.3-3 show typical load transfer mechanisms 4 used within the containment structure. The interior concrete 3.8-34 Amendment 4 b' U< t S 7)f/l/74

SWESSAR-P1 also provides biological shielding for operating personnel and protection from missiles resulting from postulated component failure. The primary shield wall surrounds the reactor vessel support shield tank and the reactor vessel above the containment mat. Besides providing shielding, the wall laterally supports the shield tank and the reactor vessel. Longitudinal splits are postulated in the hot and/or cold legs attachments to the reactor pressure vessel (RPV) by several of the NSSS Vendorc for the SWESSAR-P1 design, as discussed in Section 6.2.1.2.2. C-E postulates splits in both hot and cold legs. B&W postulates splits in the cold leg only. The SWESSAR-P1 design incorporates guard pipes in the primary shield wall for the above pipe penetrations to mitigate certain consequences of the postulated breaks. A " typical" guard pipe concept is shown in Fig. 3.8.3-5. The final design of the guard pipe, if used, RPV supports, and reactor vessel support tank will be consistent with the nodal volumes and vent areas used in the reactor cavity 26 analysis described in Section 6.2.1. The final determination of the need for guard pipes for the C-E NSSS will be submitted in the application for a construction permit by a Utility-Applicant referencing SWESSAR-Pl prior to a Final Design Approval and in the SWESSAR-Pl application for Final Design Approval. The secondary shield walls on the steam generator cubicles surround the primary loops and provide shielding between steam generator cubicles and between the steam generators and other areas of the containment. These walls are designed to support the steam generators and the operating floor. Fig. 3.8.3-1 shows , a detail of the intersection of a steam generator cubicle wall with the mat. An overhead polar crane with an approximate capacity of 500 tons is supported by structural steel framework which also contributes to the floor framing system within the containment structure. The steel box girder supporting the crane rails is attached to columns founded at the top of the mat. The crane is restrained horizontally and vertically from seismic dislodgement. Although actual details of the specific crane design are the responsibility of the crane vendor, an acceptable restraining method is typified by Fig. 3.8.3-4. The crane spans approximately 141 feet and is located directly over the centerline of the containment. The crane can reach any location 3.8-35 Amendment 26 6/2/76 47 ncu n bi TUL

SWESSAR-P1 on the operating floor within a radius of approximately 60 feet. For a description of the crane, see Section 9.1.4. The reactor vessel support shield tank is located within the concrete reactor cavity wall and primary shield wall. Further in fomation and description regarding this structure appear in Section 5.5.14. A detailed description of the principal features of the steam generator and reactor coolant pump supports is also presented in Section 5.5.14. . 3.8.3.2 Applicable Codes, Standards, and Specifications The design equals or exceeds the requirements of the ACI 318 Code and the AISC Manual of Steel Construction. Quality assurance requirements equal or exceed ANSI N45.2.5 " Supplementary Quality Assurance Requirements for the Installation, Inspection and Testing of Structural Concrete and Instructural Steel during the Cbnstruction Phase of Nuclear Power Plants." Other codes, standards, specifications, and NRC regulations and regulatory guides used in establishing design methods and material properties are: Codes, Specifications, Design Methods, and Material Properties with the exception of ACI 359 which does not apply. Section 3.8.1.2 Regulatory Guides Appendix 3A ASME Section III, Division 1, Section 5.5.14.1, Subsection NF Table 5.5.14-1 Section 3.8.3.6 describes other materials and quality control procedures used for containment interior structures. 3.8.3.3 Ioads and Ioading Ccnbinations Interior co14 crete structures within the containment are designed to withstand the pressure buildup resulting from the design basis accident (DEA) dis cussed in Section 6. 2.2. The blowdown of a postulated rupture of the reactor coolant pipe occurs in the steam generator cubicles or within the reactor cavity wall. If the break occurs in a cubicle, initial differential pressures develop until the energy passes through cubicle vent spaces to the remaining volume of the structure. All structural components, walls, floors, and beams enclosing these cubicles are designed to withstand this dif ferential pressure. 9 3.8-36 Amendment 26 6/2/76

                                                            /<~      mnn GUl      L 0 ')

SWESSAR-p1 Postulated pipe rupture may also result in blowdown jet forces. The magnitude of a blowdown jet forcing function resulting from a pipe rupture is dependent upon the geometry and distance of the target f rom its source. Appropriate safety related structures are protected from, or designed to resist, the effects of these torces. Section 3.6 describes the protection provided against the dynamic effects associated with a postulated rupture. The interior concrete structures protect the containment liner from potential missiles. Safety related equipment is physically protected from potential sources of missiles either by physical separation, or by providing restraints on the potential missile. Potential missiles and design of missile barriers are described in Section 3.5. Containment internal concrete and steel structures are designed using the loading criteria tabulated in Table 3.8.3-1. All the major loads encountered and/or postulated in a nuclear power plant are listed in this table. All the loads listed, however, are not necessarily applicable to all the containment internal structures and their elements. Loads and the applicable load combinations for which each structure will be designed will depend on the conditions to which that particular structure could be subjected. Earthquake loading is considered to be comprised of three simultaneous and independent components of excitation along the three orthogonal directions of a cartesian coordinate system. These directions are two horizontal and one vertical axes. Section 3.7.1 describes the seismic input. The ef f ects of the simultaneous excitation at any point in the structure are combined by taking the square root of the sum of the squares of the individual responses. This method is described in Se ction 3.7.3.7. A load factor of 1.0 is assigned in Table 3.8.3-1 to loads caused by the SSE (Section 3.7.1) since the SSE acceleration is at least 2 times the largest acceleration expected at the site. 3.8-36A h Amendment 26 6/2/76

SWESSAR-P1 3.8.3.4 Design and Analysis Procedures The interior structures generally cmprise a series of frames, box type structures, and assemblies of slabs. Their design is generally based on a linear elastic response. In walls and slab type structures, limit analysis is used in some instances (for example, yield line method). Material quality control procedures, as described in Section 3.8.3.6, assure that minimum material strength - requirements are met. Overstrength of materials is not a factor because the strength method of design, as described in Section 3.8.3.5, is used in proportioning and rein forcing concrete sections. Procedures used to determine reinforcing steel ratios are outlined in ACI 318, and the principal reinforcement patterns are located in the direction of tensile stresses. Bond and anchorage requirements comply with ACI 318 and, where biaxial tensile field  : exists, the develognent lengths required by Section 12.5 of ACI 318 are increased by a minimum of 25 percent. Structural steel is designed in accordance with the procedures outlined in the AISC " Specifications for the Design, Fabrication, and Erection of Structural Steel for Buildings" (February 12, 1969), Supplement No. 1 (November 1, 1970); Supplement No. 2 (December 8, 1971); and with the Structural Welding Code AWS D1.1. 4 Seismic loading on the interior structures is determined from the dynamic analysis of the containment structure, as discussed in Section 3.8.1.4. The containment analysis yields the actual loads, shears, and moments acting at the interface between internal structures and the foundation mat. The heat generated in the reactor cavity due to gamma rays enunating from the reactor core and arising from neutron interactions in the reactor cavity produce a negligible heat r.se i in the reactor cavity wall. The heat flux generation falls off approximately exponentially with a maximum value of 6.5 x 10 6 Watts per cc at the inside surface of the concrete shield wall and 9x 10-10 Watts per cc at the outside surface. This heating effect causes no detrimental thermal stress or material effects on the reactor cavity wall. Applicable equipment e nonent supports are designed and analyzed in full compliance with subsection NF of the ASME III Code as 4 defined in Section 5.5.14.1. 3.8.3.5 Structural Acceptance Criteria The basic structural acceptance criterion for concrete strength design is expressed as: 3.8-37 bO2 2$ Amendment 4 11/1/74

SWESSAR-P1 Fequired Strength 5 Calculated Strength The required strength is expressed in terms of design loads (loads multiplied by their appropriate load, or safety, f actors) . Calculated strength is computed using ACI 318-71, including the appropriate cape. city reduction f actors given in Section 9.2 of ACI 318. Design of interior steel structures for all loading conditions, exc ept the SSE, is proportioned by allowable stress design using norral working stress levels given in the AISC Specification. The one-third increase in allowable stresses for the OBE, as allowed by the AISC Specification , is not used for this loading condition. Design for loading conditions which include the S SE is proportioned by allowable stress design using 90 percent of the guaranteed minimum yield stress of the steel. Section 3.7.2.8 de scribes the variations of critical parameters incorporated into the seismic analysis to account for cracking of concrete and the range of elastic properties of different founding media. The PWR Standard Plant is designed for a variety of site conditions, for which the bounding parameters used in seismic analysis are given in Section 3.7.2.1. Design of the internal structures is based upon the most conservative values resulting from these variations in assumptions and design parameters. Section 3.7.3.6 discusses differential seismic movement relating to interconnected components, systems, and equipment. Design for horizontal shear forces in the plane of internal structure walls complies with the requirements of Section 11.16 of ACI 3 18 which incorporates the combined effects of shear and tensile stresses in the nominal pe rmis sible shear stre s s , v , allowed to be carried by the concrete. 3.8.3.6 Ma terials, Quality Control, and Special Construction Techniques Material and quality controls used for the containment interior structures are described in Sections 3.8.1.2 and 3.8.1.6 except tha t the requirements of ACI 359 do not apply. In cases where space is not available, the use of heavy aggregate concrete having a density of 200 lb per cu ft or greater is used to provide biolooical shielding. In addition to the periodic inspections of aggregates , a daily inspection program is ca rried out during concrete production to determine cradation, water content, and the quantity of material 1iner than the No. 200 sieve. Concrete radiation damage is instonificant for conditions of tluence up to 2 x 101' nyt and temperatures up to 120 C(10). The highest ne u tro n fluence (E 2 0.5 5 MeV) for concrete inte rnal 3.8-38 h [,[ ']" Amendment 7 2/28/75

SWESSAR-P1 structures is 1.4 x 1017 nvt; therefore no material damage due to radiation is expected. 9 3.8-38A Amendment 7

                                                  ,.-       2/28/75 b (. ,/ L. b

SWESSAR-P1 3.8.3.7 Testing and Incervice Surveillance Requirements No full scale structural testing or inservice surveillance is anticipated for the reactor containment interior structure. Materials used in the construction are tested as described in Sections 3.8.1.2 and 3.8.1.6. 3.8.4 Other Cateoory I Structures An optional Seismic Category I structure, the enclosure building, is discussed in Sections A3.8.4 and B3.8.4 of Appendixes A and B. 18 3.8.4.1 Description of the Structures The plot plan and the general arrangement of Seismic Category I structures and equipe nt within these structures, other than the containment structure, are shown in the following figures: Drawing Figure No. Plot Plan Sh. 1 and 2 1.2-1 1s Annulus Building Sh. 1 to 19 1.2-4 Control Building Sh. 1 to 4 1.2-5 Diesel Generator Building Sh. 1 and 2 1.2-6 With the exception of structural elements supported on continuous structural foundations of the nuclear island, a seismic rattlespace is provided between adjacent structures to provide for movement of structures under earthquake loading. The above referenced figures show the locations of these separations. Nonseismic structures are designed to prevent damage to adjacent Seismic Category I structures by one of the following methods-

1. They are designed to withstand SSE loads or tornado loads to the extent required to prevent damage to Seismic Category I structures, or

2. They are physically separated from Seismic Category I structures.

Seismic Category I structures and structures designed to resist tornado loads are listed in Table 3.2.5-1. Areas of structures, which are protected from tornadoes and tornado missile barriers, are given in Section 3.5. The structural features in the design and construction of Seismic Category I structures are similar to those of Millstone Nuclear Power Station Unit 3, Docket No. 50-423. 3.8-39 Amendment 18 2 <1

                                                                   '25

(>6.2

SWESSAR-P1 Annulus Building he annulus building is a circular shaped building surrounding the containment structure. W e building is concentric with the containment and extends outward from the outside face of the containment, to radius 150 f t 0 in. from the reactor centerline. he basement level is approximately 30 f t below station yard grade and the roof approximately 60 ft above grade. The annulus building is supported on a ecx:tmon reinf orced concrete foundation mat, which also supports the containment structure. he superstructure and exterior walls are mainly reinforced concrete. Interior floors are mainly reinforced concrete decks supported on steel or concrete framing. Grating platforms on steel framing are used for equipment access. Reinforced coucrete walls separate equipment and piping systems to provide biological shielding and protective missile barriers. We annulus building contains the engineered safety features compartments. Flood protection of the engineered safety features equipment from possible water spills within the annulus building is provided by excluding openings other than watertight personnel access doors in the low areas, and sealing penetrations. Wese design features prevent flooding of the engineered safety teatures areas from the base mat to elevation (-) 25 8 0" due to passive failures of flood systems outside the engineered safety 32 features areas. In addition, each engineered safety features area is separate and an internal flood in one area does not jeopardize the other engineered safety features areas. We annulus building contains the following systems: New and spent fuel handling and storage, fuel pool cooling and purification system Boron recovery system (partial) Radioactive licuid waste system (partial) Radioactive gaseous waste system Chemical and volume control systml Auxiliary feedwater system Steam generator blowdown system Reactor plant co::ponent cooling water system Combustible gas control system -

                                                                           ^

L v, ? . .J 3.8-40 Amendment 32 5/11/77

SWESSAR-P1 Seismic Category I ventilation systems Portions of reactor plant service water system m e fuel pool is a reinforced concrete structure approximately 35 by 45 by 43 f t high with 6 f t thick concrete walls. The fuel 32 1x>ol cooling system heat exchangers are located adjacent to the pool. The spent fuel shipping cask storage area is separated fr m the fuel pool by a reinforced cancrete wa11 and the fuel transfer canal. The fuel cask cannot be moved over tie spent fuel because the crane which lifts the cask is prevented by location frm traveling over the spent f uel . 3.8-40A , , ,.- Amendment 32 () (, /_ iiU S/11/77

SWESSAR-P1 The steam and feedwater valve areas are compartments within the upper portion of the annulus building. Separate electrical and piping tunnels are provided throughout the building. The tunnel walls, roof, and floor provide protection for the electrical cables and containment penetrations. Two containm nt personnel hatches and one equipment hatch exit into the annulus buildino. The building walls and roof provide tornado protection for these hatches. The refueling water storage tank and the. auxiliary f eedwater storage tank are supported on the containment-annulus buildi ng foundation mat and enclosed within the building for tornado protection . Control Building The control building is a reintorced concrete three sto ry structure, 122 by 122 by 60 ft high. The upper floors are reinforced concrete supported on concre&e exterior walls and the interior steel framing system. The lowest floor is at yard grade. The control building houses the control room, the battery room, and the emergency switchgear relay rooms, and related ITVAC equipment. 4 Diesel Generator building The diesel generator building houses the energency diesel generators and is located adjacent to the rontrol building (see Fig. 1. 2-6) . This portion of the building is approriraately 122 wide by 100 ft long. The roof is approximately 40 ft above fard grade and the floor is approximately at yard grade. Exterior walls and roof of the building including the wall between the diesel generator building and the control building are reinforced concrete. Fuel Oil Storage Area The fuel oil storage area consists of underground f uel oil storage tanks and fuel oil pump house. The fuel oil pump house is constructed of reinf orced concrete. The oil storage tanks and pumps supplying the emergency diesel generators are tornado protecte d . Electrical Tunnels and Piping Tunnels Electrical tunnels and piping tunnels containing Seismic Category I systems are constructed of reinforced concrete. The tunnel walls and roof are of sufficient thickness to resist penetration by tornado generated missiles. 3.8-41 /. n ' Amendment 4 {) U ' - 11/1/74

Sh'ESSAR-P 1 Tunnels are isolated f rom the structures which they adjoin by a seismic rattlespace. . Peactor Plant Tank Area The reactor plant tank area rests on a Seismic Category I foundation as shown in Fig. 1.2-1. The reinforced concrete structure forms a series of dikes capable of containing the contents of the tanks should they break or leak. Solid Waste anci Decontamination Buildim The solid waste and decontamination building shown in Fig. 1.2-9 is a nonseismic Category I structure with the exception of the reinforced concrete substructure outlined in Fig. 1.2-1. Dams and Dikes Dams and other structures necessary to protect the site from the probable maximum flood will be discussed in the Utility-Applicant's SAR, if required. Ultimate Heat Sink Structure The description and discussion of the structure required for the ultimate heat sink will be included in the Utility-Applicant 's SAR. 3.8.4.2 Applicable Codes, Standards, and Specifications Codes, standards, specifications, and AEC regulations and regulatory Juides used in establishing design methods and 4 material properties for other Category I concrete and steel structures are given in Section 3.8.3.2. 3.8.4.3 Loads and Loading Combinations Seismic Category I etructures other than the containment are designed for the applicable loading criteria given in 4 Section 3.8.3.3 and ta bulated in Table 3.8.3-1. All the major loads encountered and/or postulated in a nuclear power plant are listed in this table. All the loads listed, however, are not necessarily applicable to all the structures and their eles,ents. Loads and the applicable load combinations for which each structure will be designed will depend on the conditions to which that particular structure could be subjected. Tornado loads are described in Sections 3.3.2 and 3.5. Flood loads are described in Section 3.4. The lateral earth pressure load on an underground structure includes the horizontal component of the weight of the soil 3.0-42 ,n .n Amendment 4 bbl _iJ T1/1/74

SWESSAR-P1 against the structure and the lateral pressure resulting from surcharge loads applied at the ground surface or at the bottom of adjacent foundations. 3.8.4.4 Design and Analysis Procedures The loads in structural components are determined by usir.g elastic theory. Steel structures are considered to be supported by the concrete structures or foundations to which they are anchored. Using this assumption, the reactions obtained from the analysis of a steel structure are applied as loads to the supporting concrete. Steel structures b'have elastically under the applied loads. Concrete structures are supported by their foundations. The reactions obtained f rom tae analysis of these structures are applied as loads to the foundations. The mechanism of load transfer to the foundations is in accordance with Chapter 15 of ACI 318 (see Fig. 3.8.5-4). The expected behavior of all structures under load is such that they meet the requirements for 4 control of deflections of Section 9.5 of ACT 318. The quantity and placement of reinforcement are designed to meet the crack control provisions of ACI 318. Material quality control procedures, as referred to in Section 3.8.4.6, assure that minimum strength material requiremerts are achieved. Overstrength of material is not a f actor in calculating the loads in structural members because the loads in structural components are determined by using elastic theory. Earthquake forces on the structures are determined by dynamic analysis and then applied statically in the design of the structures. The analytical techniques used to determine the torces are given in Section 3.7.2. 3.8.4.5 Structural Acceptance Criteria The allowable stresses and tactors of safety for steel structures are in accordance with the AISC specification with the following exceptions: 4

d. Seismic CdtegOly I structures, as identlfied in Table 3.2.5-1, are capable of withstanding the SSE loads in combination with applicable dead and live loads. To resist the SSE, the allowable structural steel s tresses are taken as 90 percent of the specified minimum yield stress of the steel.

b. These structures are also checked usino ObE loads in I t

combination with applicable dead and live ~ loads. For  ! this loading condition, allowable structural steel 3* r 3.8-43 bn L' ', L "i Amendnent 4 11/1/7L

SWESSAR-P1 stresses are the tormal working stresses; and proportioning members for 33 1/3 percent increase, as allowed by the AISC Specification, is not done. Concrete structures are designed by the strength design method of ACI 318. Load f actors are as given in Section 3.8.4.3. The basic criterion for strength design is expressed as: Required Strength 5 Calculated Strength Concrete members and sections of concrete members are proportioned to meet this criterion. The required stren gth is expressed in terms of design loads, or their related internal moments and forces. Design loads are defined as loads that have been multiplied by their appropriate load factors. Calculated strength is that computed by the provisions of ACI 318, includina the appropriate capacity reduction factors. Capacity reduction f actors are taken as given in Section 9.2 of ACI 318. 3.8.4.6 Materials, Quality Control, and Soecial Construction Technicues Materials and quality controls used for o ther Catecory I structures are described in Sections 3.8.1.2 and 3.8.1.6 with the exception of the requirements of ACI 359 which do not apply. In areas where space is not available, the use of heavy aggregate - concrete having a density of 200 lb per cu ft or greater is used to provide biological shielding. 3.8.4.7 Testing and Inservice surveillance Requirements No full scale structural testing or inservice surveillance is anticipated for the structures described in Section 3.8.4.1. Materials used in the construction are tested as described in Section 3.8.4.6. 3.8.5 Foundations and Concrete Supports 3.8.5.1 Descriotion of the Foundations and Sucoorts The foundations f or Seismic Category I structures consist of re-inforced concrete mats desianed for service within an envelope of underlying subgrade conditions ranging from soil to rock. The coological and seicmic design parameters are defined in Section 2.5. There are no unique foundations or concrete support features used in the design of this plant. Nuclear Island Foundation The nuclear island consists of the contcinment and several other Seismic Category I structural elements suoported on a common B foundation mat. The mat is a 312 ft diameter reinforced concrete 3.8 4u Nmendment 8 ec2 20 3/28/vs

SWESSAR-P1 structure witn a thickness of 10 ft. General features of the mat 9 appear in Fig. 1.2-1 and 1.2-3. 8 The nuclear island foundation mat is reinforced with both top and bottom layers of reinforcing steel arranged in a concentric circular and radial pattern as shown in Fig. 3.8.5-1. Splices in adjacent parallel reinforcing bars in the mat are staggered so that these splices are not less than 4 ft apart. Reinforcing steel bars are detailed for maximum length. Mat reinforcing bar splices are not anticipated at the junction of the mat and the containment shell. Fig. 3.8.1-7 shows a typical section through a bridging bar that is used to provide reinforcing steel continuity through the mat liner plate. Fig. 3 . 8 .1 -3 shows a detail of t he mat - containment wall juncture. The liner is protected from hydrostatic uplift forces by channeling ground water migration in the mat into a collection , n etwork . Fig . 3 . 8 . 5-2 illustrates the groundwater drainage system; it also shows the use of a waterproof membrane as an added protection for sites with a foundation mat below the water table. Below the mat, the drainage system consists of a layer of porous concrete. Within the mat itself, all vertical joints lg contain a continuous waterstop, and a permeable half pipe section. The membrane, which is used at sitea with a high water table elevation, encloses the bottom of the porous concrete layer, and seals to the outside face of the z.nnulus building wall at an elevation 5 ft above maximum ground water level. 8 No cathodic protection is provided because adequate corrosion protection of the embedded reinforcing steel is furnished by concrete cover. Research by the National Bureau of Standards (11) and others (12), (13) indicates that cathodic currents damage the bond between the reinforcing steel and concrete. Foundations for Cther Structures Foundations for all other Seismic Category I strdctures consist of conventionally designed and reinforced concrete mat foundations. Fig. 3.8.5-3 shows a typical reinforcing pattern at the junction of the reinforced concrete vertical structural elements and a foundation mat. Ecuipment Supports Section 5.5.14 describes and illustrates the manner in which the major reactor coolant system components are supported. Fig. 3 . 8 . 5-4 illustrates the manner in which these supports 3.8-45 /. ;

                                                               cc' }        Amendment 8 h' t . 'c 3/28/75

SWESSAR-P1 interact with the concrete structure and is typical for support of large equipment in the containiaent structure. The structural 24 - interface requirements are given in Table 3.8.6-1. 3.8.5.2 Applicable Codes, Standard- and Specifications The design codes, s tandard s , specifications, and regulations that are used for the design and construction of the foundation mat and other Category I foundations are given in the sections listed below. For t nuclear island foundation: Section Codes, Specifications, Design Methods, and Material Properties 3.8.1.2 NRC Regulatory Guides 3.8.1.2 Appendix 3A For all other Seismic Category I foundations: 3.8.3.2 3.8.5.3 Loads and Loading Combinations Foundation plan size is established based on the envelope of site psrameters given in Section 3.8.5.4, using unfactored loadings in whu+ever combination governs the design. The loading combinations stated in Sections 3.8.1.3 and 3.8.3.3 are used for the nuclear island foundation design. The combinations stated in Section 3.8.4.3 are used for the design of all other Seismic Category I foundations. In addition to the load combinations referenced above, the following load combinations are used to check against sliding and overturning due to earthquakes, winds, and tornadoes, and against flotation due to floods.

1) D+H+E

2) D + H + W

3) D+H +El

4) D + H + Wt

5) D + F1 Where D, E, W, E1, and Wt are as defined in Table 3.8,3-1, and F1 is the buoyant force of the probable maximum flood. Stability f actors f or these conditions are given in Section 3.8.5.5.

Where Seismic Category 1 structures extend below the surf ace of the finished ground grade, their external walls are designed for the active and passive earth pressures resulting from surcharge loads applied either at the surf ace or f rom foundations for the horizontal component of the weight of the soil, for hydrostatic loads, and for seismic loads. UU _m u 3.8-46 Amendment 24 4/23/76

SWESSAR-P1 3.8.5.4 Design and Analysis Procedures Analynia of the containment structure foundation mat is performe d as a part of the analysis of the entire containment structure for a range of values of subgrade shear modulus. This is described in Section 3.8.1.4. For Seismic Category I structures other than the containment s tructure , the analysis of the foundation mat includes consideration of a range of values of subgrade stear modulus as indicated in Section 3.8.1.4. 3.8.5.5 Structural Acceptance Criteria Structural design of the nuclear island foundation is in accordance with ACI 359. Str uctural design of all other Seismic Category I foundations is in accoriaore with ACI 318 using strength design. The basic criterion for strength design is expressed as: Required Strength 5 Calculated Strength All members and all sections of members are proportioned to meet this criterion. The required strength is expressed in terms of c.esign loads, or their related internal moments and forces. Design loads are defined as loads that are multiplied by their appropriate load factors (safety factors) . Calculated strength is that computed by the provisions of ACI 318, including the appropriate capacity reduction fa ctors . Capacity reduction factors are given in Section 9.2 of ACI 318. Stability f actors against sliding, overturning, and flotation are as follows: Stability Factors Loading Condition Overturning Sliding Flotation 7 Operating basis earthquake 1.5 1.5 - Normal wind 1.5 1.5 - Safe shutdown earthauake 1.1 1.1 - Tornado 1.1 1.1 - Flood flotation - - 1.5 The nuclear island foundation eliminates differential settlement between the annulus building and the containment structure.

                                                          ,n; 3.8-47       f)b d        J     Amendment 7 2/28/75

SWESSAR-P1 3.8.5.6 Materials, Quality Control, and Special Construction Techniquas Materials, quality control, and specifications used for Seismic Category I foundations and supports are described in Section 3.8.5.2. There are c.o special construction techniques required by the design of this plant. Any site-generated variations from routine construction methods will be discussed in the Utility-Applicant's SAR. 3.8.5.7 Tes t ing and Inservice Surveillance Requirements The entire containment structure including its foundation undergoes structural acceptance testing as described in Section 3.8.1.7. Except for this test, no other testing or insarvice surveillance of foundation systems is planned. 3.8.6 Structural Interf aces 24 The interfi.ce information is given in Table 3.8.6-1. References for Section 3.8 (1) Timoshenko ind Woinowsky-Krieger, Theory o_f Plates and Shells, Second Edition, 1957. (2) Eudiansky, B. and Radkowski, P.P., " Numerical Analysis of Unsymmetrical Bending of Shells of Revolution," AIAA Journal, August 1963. (3) Greenbaum, G.A., Comments on " Numerical Analysis of Unsymmetrical Bending of Shells Of Revolution," AIAA Journal, October 1963. (4) Sanders, Jr., J,L., "An Improved First Approximation Theory for Thin Shell," Technical Report R-24 'JAS A . (5) "I CES STRUDL II, The Structural Dasign L 2 age Engineering Users Manual," Vol. I, Frame Anu ysis, November 1968; Vol. II. (6) Newmark, N.M. and Rosenbleuth, E., Pundamentals or Earthquake Engin e ering , Prentice Hall, 1971. (7) Stone & Webster Engineering Corp. Report SWND-5, dated November, 1969, and entitled " Design of Orthogonally-Reinforced Concrete Nuclear Containment Shells for Local Membrane Shear Resistance," by M.J. Holley, Jr., of Hansen, Holley, and Biona. and "Pehavior of Precracked Concrete Subjected to R ng Shearing Stresses " Ly R.N. White, Associated F_ .sor, Cornell University. b i. 3.8-48 Amendment 24 4/23/76

SWESSAR-P 1 (8) Stone & Webster Engineering Corp. Report SWND-SS. Report SWND-SS is a supplementary report of SWND-5, dated March, 1970, and entitled "Effect of Cycled Flexure on Tensioned Rebars," by M.J. Holley, Jr., of Hansen, Holley, and Biggs, and " Behavior of No. 18 Reinforcing Bars Under Combined Axial Load and Plastic Bending," by h.N. White, Associate Professor, Cornell University. (9) Clark, R.G., " Radiation Damage to Concrete," USAEC Report H.W.-56195 Hanford Atomic Products Operation. (10) Clark, R.G., " Radiation Dan. age to Concrete," U .S. AEC Report H.W.-56195, Hanford Atomic Products Operation. 3 (11) Underground Corrosion, U.S. Department of Commerce, National Bureau of Standards, Circular 579, April 1957. (12) Rosa, E.B., McCollum, B., and Peters, O.S., Electrolysis In Concrete, Tech. Paper BS T18 (1913). (13) McCollum, B., Ahlborn, G.H., Special Studies in Electrolysis Mitigation - II, Tech. Paper BS T54 (1916). 9 3.8-49 Amendment 25 f

                                                   ^^r          4/30/76 6o,&     ~~'

SWESSAR-P1 TART 1. 3.8.1-1 4 LOAD COMBINATIONS FOR CONTAINMENT LINER INSERT AND OVERLAY PLATES, BRACKETS, AND ATTACID1ENTS Category Load Combination Design I D+L+R g

                                                         + OBE Design II                          D + L + SSE lbte:

Where: D = Dead load of electrical tray assemblies or other miscellaneous equipment. L = Live loads, i.e., scaffolding, equipment being lifted. OBE = Effects of operating basis earthquake, pipe loads. SSE = Effects of safe shutdown earthquake, pipe loads. Rg = Operating pipe loads, thermal expansion, water hammer.

                                                 ,n  .

1 of 1 b b ,' _ c. b Amendment 4 11/1/74

G SWESSAR-P1 TABLE 3.8.1-2 4 LOAD COMBINATION AND ALLOWABLES far 1) Containment Access openings

2) Containment Electrical Penetration Sleeves

3) Contairacent Piping Penetration Sleeves e

4) Puel Transfer Tube Enclosure Cateqory Itud Ortinations Applicability Str ess Allowables ik? sign I D + L + Ig + OBE + Pg

• R,n h tal structure Pm IS m D+ L + P,,+ OhE
• Pg
• R m P, SI SU m P, (Pf)+P 51.5qn l4 Dasign II D+L+P, + SSE
• Pg + R, Where structure Pm 51.2S m D+ L+P + SSE
• Pg
• R m is integral and P, or1.d

51. 8 S,h,or (S g continuous. P, (P, f + P,, 51. S or 1.5S Operating I D+ L*P + T" + OBE

• P+ T*R Total structure P + Q $3S live usag"e f actor <1.0 f4

                                                    '                      '    '   "                             P' dan +114 Operating II         D+L+P                 + T " + SSE                  Where structure     P

• P + 0 53S
• is not integral /

l4 and continuous Test D+P

• T Mtal structure P 50.9q P {P ) +P 51.25S3 4 3

r,* ' Pipe ruptur e (See Table 3.8.1-7) [V Note: The loads P , T, R m, h,, ,g P and gT do not apply to access opmings or electrical pene,tration sleeves. V) rJ

    %)

y

,f

• f*s

?                  9 Q       t
                         ,s -
                              ~
         ~.

d-. N

                                           %/.      q%

L' ' , '

                                                            !>p
                                         .,-4,,
                                                                 ,My
                                                      -l         'i
                                                             ' k:Qsy 1 of 2                                          Amendment 4 11/1/74

SWESSAR -}- 1 TABLE 3.8.1-2 (Coffr) where: D = Dead load of structur e and contents L = Live load of structures and contents P, = Containment internal design pressure P,.,= Containment extArnal design pressure T3 = Containment design temperature P. = Range or operating pressure in containment including anximum DBA pressure T, = Tempe r a t ur e- associated with ( P, = Pipe design pressure T g = Pipe design temperature Pi = Containment test pressure P, = Range of operating pressure of the pipe T, = Tesuperatures associated with P, R m= Mechanical pipe loads ko = Mechanical and thermal pi l e loads OBE = Effects of operating basis earthquake SSE = Effects of saf e shutdown earthquake S,,, = Ikssic stress intensity allowable, ASME III, Div 1, Table I-10 S, = Miniatus yield stress, ASME III, Div 1, Table I-10 P,n , P , Pg, Q = Per ASME III nomenclature. f4 DBA =[ loads due to design basis accident, except for pipe rupture of the attached line. (See Table 3.8.1-7.) O N. r) [' J CO 2 of 2 Amendment 4 11/1/74

SWESSAR-P1 TA nT.E 3.8.1-3 4 LOADING COMBINATION AND LIMITS FOR PIPE RUPTURE

                                                                        /

Applicability Inads Stress Allowable Pipe Penetration SSE + pipe rupture 85% of Appendix F, Assemblies mechanical loads ASME III, Allowables O 1 of 1 Amendment 4 11/1/74

SWESSAR-P1 TABLES 3.8.1-4, -5, -6, and -7 have been deleted. lof1 Amendment 5

                                            , , - , ,r     12/2 /74 bbd     LJU

SWESSAR-P1 TABLE 3.8.3-1 STRUCTURES LOADING CRITERIA The following nomenclature ar.d definition of terms apply to the criteria in this table. Loads, Definition of Terns, and Nomenclature All the najor loads to be encountered and/or to be postulated in a nuclear power plant are listed. All the loads listed, however, are not necessarily applicable to all the structures and their elements in a plant. Loads and the applicable load combinations for which each structure has to be designed depends on the conditions to which that particular structure could be subjected.

1. Normal Loads Normal loads are those loads to be encountered during normal plant operation and shutdown. They include the following:

D --- Dead loads or their related internal moments and forces, including any permanent ecuipment loads. 7 L --- Live loads or their related internal moments and forces, including any movable equipment loads and other loads that vary with intensity and occurrence. 7 Tg --- Thermal effects and loads during normal operating or shutdown conditions, based on the most critical transient or steady state condition, and R --- Pipe reactions during normal operating or shutdown conditions, based on the most critical transient or steady state condition. F --- Lateral and vertical pressure of liquids, or their related internal moments and forces. H --- Lateral earth pressure, or its related internal moments and forces.

2. Severe Environmental Loads Severe environmental loads are those loads that could infrequently be encountered during the plant life. Included in this category are:

E --- Loads generated by the operating basis earthquake, and W --- Loads generated by the design wind., 1 of 6 Amendment 7

                                                        . .,    2/28/75

() b l. LJ\

SWESSAR-P1 TABLE 3.8.3-1 (CONT)

3. Extreme Environnental Loads Extreme environmental loads are those loads which are credible but are highly improbable. They include:

E --- Loads generated by the safe shutdown earthquake, and W, --- Loads generated by the design basis tornado specified for the plant. They include loads due to the tornado wind pressure, loads due to the tornado-created dif-ferential pressures, and loads due to the tornado-generated missiles.

4. Abnormal Loads Abnormal loads are those loads generated by a postulated high energy pipe break accident within a building and/or compa rtment th ereo f . Includea in this category are the following:

P --- Pressure equivalent static load within or across a ccanpartment and/or building, generated by the postu-lated break, and including an appropriate dynamic load O factor to account for the dynamic nature of the load, P is equal to Pd (from Table 6 2.1-3) times the dy-namic load factor. T, ' Thermal loads under thermal conditions generated by the postulated break and including Tg . R." Pipe reactions under thermal conditions generated by the postulated break end including R . Y --- Equivalent static load on the structure generated by the reaction on the broken high energy pipe during the postulated break, and including an appropriate dynamic load factor to account for the dynamic nature of the load. Y --- Jet impingement equivalent static load on a structure I generated by the postulated break, and including an appropriate dynamic load factor to account for the dynamic nature of the load. Ym --- Missile impact equivalent static load on a structure generated by or during the postulated break, such as pipe whipping, and including an appropriate dynamic load factor to account for the dynamic nature of the load. 2 of 6 < > ?

                                                          'J" Amendment 7 b b. L-           2/28/75

SWESSAR-P1 TABLE 3.8.3-1 (CONT) In determining an appropriate equivalent static load for Y , Yj , and Ym, elastoplastic behavior may be assumed with appropriate ductility ratios and as long as excessive deflections do not result in loss of function of any safety related system.

5. Other Definitions S --- For structural steel, S is the required section strength based on the elastic design methods and the allowable stresses defined in Part 1 of the AISC

           " Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings,"

February 12, 1969. The 33 percent increase in allowable stresses for concrete and steel due to seismic or wind loadings is not permitted. U --- For concrete structures, U is the section strength required to resist design loads and based on methods described in ACI 318-71. Y --- For structural steel, Y is the section strength re-quired to resist design loads and based on plastic design methods described in Part 2 of AISC "Specifi-cation f or the Design, Fabrication, and Erection of Structural Steel for Buildings," February 12, 1969. 3 of 6 Amendment 7

                                            -   o"i LJ' 2/28/75

(> b l

SWESSAR-P1 TABLE 3.8.3-1 (CONT) Concrete Structures

1) U= 1.4D + 1.7 L 4 1.7Ro 7

2) U = 1.4D + 1.4F + 1.7L + 1.7H + 1.9E + 1.7R o

3) U= 1.4 D + 1.4F + 1.7L + 1.7H + 1.7 W + 1.7Rg

4) U=D+F+L+H+T, +R g + E'

5) U=D+ F + L + H + 7 'o +R g +W.

6) U=D+ F+L+H+T+ a R+

a 1.2 6 Pg

7) U=D+ 'F+L+H+Tu + Ra +1,15 P g

                + 1.0 (Yr+ Y 3+ X) + 1.15E

8) U = D + F + L + H + T+ + 1.0 P " +

1.0 O( + Y +Y R'.0E8

9) U= .75 [ 1.5D1.+d 1.7L"f + 1

                                  + 1.4 T + 1.7R ]

4 F + 1.7L I 1.7H + 1.9E +

10) U= .75 [ 1.4D + 1.4T,+ 1.7RJ

11) U= .75 [ 1.4D + 1.4F + 1.7L + 1.7H + 1.7W + 1.4T +

1.7R} For these load combinations, where D or L reduce the effect of other loads, the corresponding coefficients shall be taken as 0.90 for D and zero for L. The vertical pressures of liquids shall be considered as dead load, with due regard to variation in liquid depth. Notes - Concrete Structures In combinations (6) , (7), and (8), the maximum values of P , T 3, R a* Y i,Y r, and Y , including an appropriate dynamic lodd f actor, shall be used unless a time-history analysis is performed to justity otherwise. Combinations (5) , (7) , and (8) shall be satisfied first without the tornado missile load in (5) and without Y r' Y i, and Y3 in ('7) and (8). When considering these loads, however, local section strength capacities may be exceeded under the effect of these concentrated loads, provided there is no loss of function of any safety related system. Both cases of L having its full value or being completely absent should be checked. geel Structures - Ioadi:.a Combinations Either the elastic working stress design methods of Part 1 of AISC, or the plastic design methods of Part 2 of AISC, may be used. 4 of 6

                                                           . Amendment 7
                                                ,,-     G}R        2/28/75

SWESSAR-P1 TABLE 3.8.3-1 (CONT) Elastic Working Stress Design - Service Load Conditions

1) S=D+L

2) S=D+L+E

3) S=D+L+W If thermal stresses due to T and R are present, the follow-ing combinations should also be satisfied:

1a) 1.5 S = D + L + Tg +R g 2a) 1.5 S =D+L+T g +R g +E 3a) 1.5 S = D + L + T,, +R g +W both cases of L having its f ull value or being completely absent should be checked. Elastic Working Stress Design - Factored Load Conditions The following load combinations should be satisfied: If elastic working stress design metheds are used:

4) 1.6S=D+L+T g

                                +P g
                                     + E3

5) 1.6S=D+L+T 0

                                +R a
                                     +Wt                                    7

6) 1.6S=D+L+T g +R y + F

7) 1.6S=D+L+T a

                                +R   + P, + 1.0 O(        +Y r +Y)m   +E

8) 1.7S=D+L+T +R g

                                     + P,,   + 1.0 Of     +Y r +YIm +

Notes - Elastic Working Stress Design For combinations (7) and (8) in computing the required section strength, S, the plastic section modulus of steel shapes may be used. Plastic Design - Service Load Conditions

1) Y = 1.7 D + 1.7 L

2) Y= 1.7 D + 1.7 L + 1.7 E

3) y = 1.7 D + 1.7 L + 1.7 W 5 of 6 Amendment 7

[3 (,2 {}} 2/28/75

SWESSAR-P1 TABLE 3.8.3-1 (CONT) If thermal stresses due to T and k are present, the follow-ing combinations should also be satisfied: lb) Y= 1.3 03 + L+T +R)g 2b) Y= 1.3 p) + L+E+T, +R)g 3b) Y= 1.3 (D + L + W + T o +R)o Both cases of L having its full value or being completely absent should be checked. Plastic Design - Factored Load Conditions

4) .90 Y = D + L + Tg +R g

                                          + E1

5) .90 Y = D + L + Tc +R g

                                          + W,

6) .90 * =D+L+T g +R + 1.5 Pa

7) .90 Y = D + L + Tg + "a + 1.25 Pa + 1.0 Of) +Y r

• Y) m

                       + 1.25 E

8) .90 Y =D+L+T g + R a + 1.0 P g + 1.0 pr +Y r + Y) 7

                       + 1.0 E2 IJotes - Steel Structures In factored load combinations, thermal loads can be neglected when it can be shown that they are secondary and self-lbmiting in nature and where the material is ductile.

In combinations (6) , (7) , and (8), the maximum values of Pg , T a a R ,, , Y Y , and Y including an appropriate dynamic load factor, shallr,be used m, unless a time-history analysis is performed to justify otherwise. Combinations (5) , (7) , and (8) shall be first satisfied without the tornado missile load in (5) and without K , Ym, and Y in (7) and (8). When ceasidering these loads, however, local section strengths may be exceeded under the effect of these concentrated loads, provided there is no loss of function of any safety related system.

                                                    ,,    o  -

6 of 6 Ubd LJU Amendment 7 2/28/75

SWESSAR-P1 "ABLE 3.8.6-1 STELcTURAL INTERFACES RESAR-3S SWESSAR

1. Westinghouse-will review 1. Tne loads exerted on all the nozzle loads for ac- reactor coolant piping ceptance (RESAR-3S Sec- branch nozzles will be pro-tion 1.7.1, p 1.7-8, vided to the NSSS Vendor.

Amendment 4).

2. Westinghouse will review 2. Design inforration and the interface data for drawings affecting NSSS compliance with Westing- components will be provided house design criteria to the NSSS Vendor.

(RESAR-3S, Section 1.7.1, p 1.7-8, Amendment 4) .

3. Westinghouse will pro- 3. See Note 1. 24 vide the balance of plant designers with allowable loads at those locations where NSSS equipment interf aces with balance of plant equip-ment (RESAR-3S, Section 1.7.1, p 1.7-8, Amend-ment 4) .

4. Westinghouse will provide 4. See Note 1 to the balance of plant designers the differen-tial displacements and rotations due to all loads (normal, thermal, seismic, etc) at points of the NSSS that will interface with the balance of plant struc-tures. (RESAR-3S , Sec-tion 1.7.1, p 1.7-9, Amandment 6) n 7 O l. LJl N-3S Amendment 24 1 of 3 4/23/76

SWESSAR-P1 TABLE 3.8.6-1 (CONT) RESAR-3S SWESSAR

5. Westinghouse will provide 5. See Note 1 the balance of plant de-signers with prelim 2. nary structural properties (i.e., structural stiff-ness ranges) of support-ing balance of plant structures representative of those encountered in previous Westinghouse support designs. (RESAR

      -3S, Section 1.7.1, designs p 1.7-9, Amendment 6) .

6. Westinghouse will evaluate 6. Structural stiffnesses the structural stiffnesses will be provided to the (RESAR-3S, Section 1.7.1, NSSS Vendor for their 24 p 1.7-9, Amendment 6) . analysis.

7. Westinghouse will provide 7. See Note 1.

the balance of plant de-signer with loads that are transmitted from the NSSS equipment to the supporting balance of plant structures. (RESAR-3S, Section 1.7.1, p 1.7-9, Amendment 6) .

8. Westinghouse will eval- 8. Structural displacements uate the structural dis- affecting NSSS equipment placements (RESAR-3S, Section will be provided to the Section 1.7.1, p 1.7-9, USSS Vendor.

Amendment 6) .

9. Inelastic analysis, if 9. See Note 1.

use by Westinghouse, will be identified and the results of such analysis will be pro-vided to the balance of plant designer when necessary for hin design. (RESAR-3S Sectiren 1.7.1, p 1.7-9, Amenda.ent 6) . G

                                                         )

W-3S 2 of 3 Amendment 24 4/23/76

SWESSAR-P1 TABLE 3.8.6-1 (CONT) _RESAR -3S SWESSAR

10. Westinghouse will assess 10. Where inelastic analysis the effects o.f the in- is used such that NSSS com-elastic analysis on NSSS ponents may be affected, components (RESAR-3S , the NSSS Vendor vill be pro-Section 1.7.1, p 1.7-9, vided with the results of Amendment 6) . the analysis.

11. Westinghouse will provide 11. Sce Note 1.

the balance of plant de-signer with design enve-lope loads for NSSS equip-ment interfacing with balance of plcnt equipment for all operating condi-tions. (RESAR-3S , Sec-tion 1.7.1, p 1.7-10, Amendment 2).

12. For auxiliary equipment 12. See Note 1. 24 the balance of plant de-signer is provided with loads and criteria that must be met (RE-SAR-3S, Section 1.7.1, p 1.7-10, Amendment 2).

13. Westinghouse will provide 13. See Note 1.

the balance of plant de-signer with loads and criteria that must be met to assure opera-bility of active com-ponents (RESAR-3S, Section 1.7.1, p 1.7-10, Amendment 21: . NOTE 1 When all interface data to be provided by Westinghouse becomes available, it will be checked for conformance. Stone and Webster and Westinghouse will develop a design to ensure that RESAR-3S and SWESSAR are compatible. All interface data will be documented at periodic intervals during the design of a plant by the NSSS Vendor and SSW and their formal acceptance will be indicated by SSW and the NSSS Vendor.

                                                      -    -    a   n
                                                        ._      c/

W-3S 3 of 3 Amendment 24 4/23/76

I 50 No.18 1

                          /,W                                                  /

u 3 [9,'/ / l-(( ~ FULL PENETRATION WELD N o.18 " , N o.18 BOTH SIDES EL -3000 FT TOP OF TAL OF WED M SHEAR BAR ASSEMBLY VERTICALS - No.18

                                     /    ,
                                               ,                    SHEAR BAR ASSEMBLY
                                   /                N o.18 DI AGON A LS - 4"x 1" FLAT        ,/

BARS , , , VERTICALS CADWELDED o- - DIAGONAL BARS TO No.18 SHELL REINF. , N o.18 N o.18 ~ ~ , , +-CONTAINMENT WALL A s l l s l l

                                     /

LINER - EL.(-) 51.00 FT.

                                 ,/       ,

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