ML20247G714
| ML20247G714 | |
| Person / Time | |
|---|---|
| Site: | 05200002, 05000470 |
| Issue date: | 03/30/1989 |
| From: | ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY |
| To: | |
| Shared Package | |
| ML20247G537 | List:
|
| References | |
| NUDOCS 8904040304 | |
| Download: ML20247G714 (154) | |
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CESSAR 8lnincums (Sheet 1 of 8)
EFFECTIVE PAGE LISTING CHAPTER 3 Table of Contents Pagg Amendment i D ii lii D iv D v D vi D vii E viii E ix E x
xi xii E xiii E xiv E xv E xvi E xvii E xviii D xix D xx E xx1 E xxii E xxiii E Text
.P_a gg Amendment 3.1-1 D 3.1-2 D 3.1-3 D 3.1-4 D 3.1-5 D 3.1-6 D 3.1-7 D 3.1-8 D 3.1-9 D 3.1-10 D 3.1-11 D 3.1-12 D 3.1-13 D Amendment E 8904040304f{fj70 DR pyocg ** ' ' PDC
4 CESSAR 88Wncm:u m eet 2 or o ) ( l i O
. EFFECTIVE PAGE LISTING (Cont'd) l Clib.P_T E R 3 Text (Cont'd) page Amendment 3.1-14 D 3.1-15 D 3.1-16 D 3.1-17 D l 3.1-18 D 3.1-19 3.1-20 D )
3.1-21 D j 3.1-22 D j 3.1-23 D l i 3.1-24 D 3.1-25 D 3.1-26 D 3.1-27 D 3.1-28 D 3.1-29 D 3.1-30 D 3.1-31 D 3.1-32 D l 3.1-33 D l 3.1-34 D 1 3.1-35 D j 1.1-36 D ) 3.1-37 D ! 3.1-38 D l 3.1-39 D 3.1-40 D 3.2-1 D 3.2-2 E 3.2-3 D j 3.2-4 D 3.2-5 D 3.3-1 D 3.3-2 , D 3.3-3 D 3.4-1 D 3.4-2 D 3.5-1 D 3.5-2 D 3.5-3 D 3.5-4 D 3.5-5 D 3.5-6 D Amendment E December 30, 1988 l _m ____
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EFFECTIVE PAGE LIS_RN_G (Cont'd) CHAPTER 3 Text (Cont'd) P_ age Amendment 3.5-7 D 3.6-1 E 3.6-2 E 3.6-3 E 3.6-4 E 3.6-5 E 3.6-6 E 3.6-7 E l 3.6-8 E 3.6-9 E 3.6-10 E 3.6-11 E 3.6-12 E 3.6-13 E j'^N 3.6-14 E i l
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3.6-16 E 3.6-17 E 3.6-18 E 3.6-19 E 3.6-20 E 3.6-21 E 3.6-22 E 3.6-23 E 3.6-24 E 3.6-25 E 3.6-26 E 3.6-27 E 3.6-28 E 3.6-29 E 3.6-30 E l 3.7-1 1 3.7-2 3.7-3 1 3.7-4 3.7-5 3.7-6 3 7-7 , 3.7-8 l p, 3.7-9 1 1 [w. ] 3.7-9(a) 1 Amendment E December .30, 1988
CESSAR Einincui:n (S'" *t 4 f 8) O E.FFECTIVE PAGE LI_ STING (Cont'd) CHAPTER 3 Text (Cont'd) ] pge, Amendment ) I 3 7-10 1 l 3.7-11 6 ) 3.7-11(a) 6 . ' 3.7-12 6 3.7-12(a) 6 3.7-13 6 3.7-14 6 3.7-15 3.7-16 3.7-17 6 3.7-18 3.7-19 6
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3.7-20 3.7-21 3.8-1 3.9-1 E ! 3.9-2 E 3.9-3 E , 3.9-4 E j 3.9-5 E j 3.9-6 1: ) 3.9-7 E I 3.9-8 E ) 3.9-9 E l 3.9-10 E 3.9-11 ) 3.9-12 3.9-13 3.9-14 3.9-15 3.9-16 3.9-17 E 3.9-18 E 3.9-19 E 3.9-20 E 3.9-21 E J 3.9-22 E 3.9-23 E 3.9-24 3.9-25 9 Amendment E December 30, 1988 l
CESSAR !!sWicamu (Sheet 5 of 8) I EFFECTIVE PAGE 7;ISTING (Conted) CHAPTER 3 , l TSXt (Cont'd) i Eage Amendment l
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3.9-32 E 3.9-33 E 3.9-34 E 1 3.9-35 E j 3.9-36 E ) i 3.9-37 E 3.9-38 E - 3.9-39 E l
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E E j 3.9-43 E 3.9-44 E 3.9-45 E 3.9-46 E 3.9-47 E 3.9-48 E 3.9-48a E l 3.9-49 E { 3 9-50 l 3.9-51 l 3.9-52 E l 3.9-53 E j 3.9-54 E 3.9-55 E 3.9-56 E 1 3.9-57 E I 3.9-58 3.0-59 3.9-60 E 3.9-61 j 3.9-62 E l 3.9-63 E n 3.9-63a E 3.9-64 (k] 3.9-65 E E I Amendment E December 30, 1988 1
g g g g igN , (Sheet 6 of 8) O EFFECTIVE PAGE LISTING (Cont'd) CHAPTER 3 lexj; (Cont'd) Page Anendment 3.9-66 E 3.9-67 E 3.9-68 E 3.10-1 3.10-2 D D g 3.10-3 D 3.10-4 D 3.10-5 D 3.10-6 D 3.11-1 D 3.11-2 D 3.11-3 3.11-4 D 3.11-5 D 3.11-6 D 3.11-7 D 3.11-8 3.11-9 D Ta.blas. Amendment 3.2-1 (Sheet 1) D l 3.2-1 (Sheet 2) E 3.2-1 (Sheet 3) E 3.2-1 (Sheet 4) D 3.2-1 (Sheet 5) D 3.2-1 (Sheet 6) D 3.2-2 (Sheet 1) D 3.2-2 (Sheet 2) D 3.2-2 (Sheet 3) D 3.2-2 (Sheet 4) D 3.2-2 (Sheet 5) D 3.2-2 (Sheet 6) D 3.2-2 (Sheet 7) D i 3.2-2 (Sheet 8) D l 3 . 2 -- 2 (Sheet 9) D l 3.2-2 (Sheet 10) D 3.2-2 (Sheet 11) D 9 Amendment E December 30, 1988
CESSAR nancu ou (snoet 7 or 8)
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EFFECTIVE PAGE LISTING (Cont'd) j l CHAPTER 3 ) l I Tables (Cont'd) Amendment _ 3.2-2 (Sheet .12) D l 3.2-2 (Sheet 13) C j 3.2-2 (Sheet 14) D { 3.2-2 (Sheet 15) D 3.2-2 (Sheet 16) D 3.2-2 (Sheet 17) D 3.2-2 (Sheet 18) D 3.2-2 (Sheet 19) D 3.2-2 (Sheet 20) D 3.2-2 (Sheet 21) D 3.2-3 D 3.2-4 D 3.5-1 (Sheet 1) E 3.5-1 (Sheet 2) E 3.6-1 E (N 3.6-2 E ( ,) 3.6-3 'E 3.6-4 E 3.7.2-1 1 3.7.2-2 (Sheets 1 and 2) 3.7.2-3 3.9.1-1 (Sheet 1) 3.9.1-1 (Sheet 2) 3.9.1-1 (Sheet 3) E 3.9.2 E 3.9.3 E 3.9-4 (Sheet 1) E 3.9-4 (Sheet 2) E 3.9-4 { Sheet 3) E 3.9-d (Sheet 4) E 3.9-4 (Sheet 5) E 3.9-4 (Sheet 6) E 3.9-4 (Sheet 7) E 3.9-5 E 3.9-6 E 3.9-7 E 3.9-8 (Sheet 1) E 3.9-8 (Sheet 2) E 3.9-9 E 3.9-10 E LJ Amendment E December 30, 1988
CESSAR E!a%uion (Sheet 8 of 8) l O EffElTIVE PAGE LISTING (Cont'd) CHAPTER 3 Tables (Cont'd) Abendment 3.9-11 E 3.9-12 E 3.9-13 E 3.9-14 E 3.9-15 E 3.9-16 E Fiqures Amendaent 3.6-1 E 3.7.1-1 1 3.7.1-2 1 3.7.1-3 1 3.7.1-4 1 3.7.2-1 1.7.2-2 3.7.2-3 3.7.3-1 13 3.7.3-2 3.7.3-3 3.7.3-4 10 3.7.3-5 3.7.3-6 3.7.3-7 3.7.3-8 6 3.9-1 3.9-2 3.9-3 3.9-4 3.9-5 3.9-6 3.9-7 3.9-8 3.9-9 3.9-10 3.3-11 3 5-12 3.9-13 j 3.9-14 I: 1 2.9-15 E 9 Amendment E December 30, 1988 l
CESSAR. EHLuoy Il l TABLE OF CONTEN18 I i
'CKAPTER 3 l 1
Section Bubiect Pace _No. 3.0 DESIGN _OF STRUCTURES. COMPONENTS, 3.1-1 E,UIPMENT Q AND_ SYSTEMS i 3.1 CONFORMANCE WITH NRC GENERAL DESIGN 3.1-1 ! CRITERIA 3.1.1 CRITERION 1 - QUALITY STANDARDS.AND 3.1-1 ! RECORDS i 3.1.2 CRITERION 2 - DESIGN BASES FOR 3.1-2 PROTECTION AGAINST NATURAL PHENOMENA 1 3.1.3 CRITERION 3 - FIRE PROTECTION 3.1-2 : I 3.1.4 CRITERION 4 - ENVIRONMENTAL AND 3.1-3 O. MISSILE DESIGN BASES ! 3,1.5 CRITERION 5 - SHARING OF STRUCTURES, 3.1-4 ~ SYSTEMS, AND COMPONENTS I l 3.1.6 CRITERION 10 - REACTOR DESIGN 3.1-4 l l 3.1.7 CRITERION 11 - REACTOR INHEDENT 3.1-5 PROTECTION 3.1.8 CRITERION 12 - SUPPRESSION OF REACTOR 3.1-5 POWER OSCILLATIONS { 3.3.9 CRITERION 13 - INSTRUMENTATION AND 3.1-6 CONTROL 3.1.10 CRITERION 14 - REACTOR COOLANT PRESSURE 3.1-8 BOUNDARY 1.1.11 CRITERION 15 - REACTOR COOLANT SYSTEM 3.1-9 DESIGN , 3.1.12 CRITERION 16 - CONTAINMENT DESIGN 3.1-9 0 I' 3.1.13 CRITERION 17 - ELECTRICAL POWER 3.1-10 ( j' SYSTEMS i I l Amendment D i e Sept'mber 30, 1988 L '._
CESSAR nnincanou O TADLE, OF CONTENTS (Cont'd) CHAPTER 3 i section S3biect Ea9Le_lo_t. 3.1.14 CRITERION 18 - INSPECTION AND TESTINO 3.'l-11 OF ELECTRICAL POWER ] SYSTEMS 3,1.?5 CRITERION 19 - CONTROL ROOM 3.1-12 3.1.16 CRITERION 00 - PROTECTION SYSTEM 3.1-13 FUNCTIONS 3.1.17 CRITERION 21 - PROTECTION SYSTEM 3.1-14 RELIABILITY AND TESTABILITY 3.1.18 CRITERION ?2 - PROTECTION SYSTEM 3.1-15 INDEPENDENCE 3.1.19 CRITERION 23 - PROTECTION SYSTEM 3.1-16 FAILURE MODES 3.1.20 CRITER1CN 24 - SEPARATION OF 3.1 PROTECTION AND CONTROL SYSTEMS 3.1.23 CRITERION 25 - PROTECTION SYSTEM 3.1-18 REQUIREMENTS FOR REACTIVITY CONTROL MALFUNCTIONS 3.1.22 CRITERION 26 - REACTIVITY CONTROL 3.1-18 I SYSTEM REDUNDANCE AND CAPABILITY ! 3.1.23 CRITERION 27 - COMBINED REACTIVITY 3 1-19 CONTROL SYSTEMS CAPABILITY I 3.1.24 CRITERION 28 - REACTIVITY LIMITS 3.1-19 3.1.25 CRITERION 29 - PROTECTION AGAINST, 3.1-20 ANTICIPATED OPERA-TIONAL OCCURRENCES O ii
CESSAR !!!Mication i O l TABLE OF CONTENTS (Cont'd) CHAPTER 3 Section subiect Pace No. 3.1.26 CRITERION 30 - QUALITY OF REACTOR 3.1-21 COOLANT PRESSURE BOUNDARY j l 3.1.27 CRITERION 31'- FRACTURE PREVENTION. 3.1-21 j l OF REACTOR COOLANT i PRESSURE BOUNDARY 3.1.28 CRITERION 32 - INSPECTION OF REACTOR 3.1-23 i COOLANT PRESSURE- I BOUNDARY l l 3.1.29 CRITERION 33 - REACTOR COOLANT MAKEUP 3.1-24 3.1.30 CRITERION 34 - RESIDUAL HEAT REMOVAL 3.1-25 3.1.31 CRITERION 35 - EMERGENCY CORE COOLING 3.3-26 3.1.32 CRITERION 36 - INSPECTION OF EMERGENCY 3.1-27 CORE COOLING SYSTEM 3.1.33 CRITERION 37 - TESTING OF EMERGENCY 3.1-28 j CORE COOLING SYSTEM 3.1.34 CRITERION 38 - CONTAINMENT HEAT 3.1-28 REMOVAL 3.1.35 CRITERION 39 - INSPECTION OF 3.1-29 CONTAINMENT HEAT REMOVAL SYSTEM 3.1.36 CRITERION 40 - TESTING OF CONTAINMENT 3.1-29 HEAT REMOVAL SYSTEMS 3.1.37 CRITERION 41 - CONTAINMENT ATMOSPHERE 3.1-30 CLEANUP D 3.1.38 CRITERION 42 - INSPECTION OF 3.1-31 CONTAINMENT ATMOSPHERE CLEANUP SYSTEMS O 3.1.39 CRITERION 43 - TESTING OF CONTAINMENT ATMOSPHERE CLEANUP 3.1-31 Amendment D iii September 30, 1988
CESSAR EHhou O TABLE OF CONTENTS (Cont'd) CHAPTER 3 Section Bubiect Page No. 3.1.40 CRITERION 44 - COOLING WATER 3.1-31 3.1.41 CRITERION 45 - INSPECTION OF COOLING 3.1-32 WATER SYSTEM 3.1.47 CRITERION 46 - TESTING OF COOLING 3.1-32 WATER SYSTEM 3.1.43 CRITERION 50 - CONTAINMENT DESIGN 3.1-33 BASIS 3.1.44 CRITERION 51 - FRACTURE PREVENTION 3.1-34 OF CONTAINMENT D PRESSURE BOUNDARY 3.1.45 CRITERION 52 - CAPABILITY FOR 3.1-34 CONTAINMENT LEAKAGE RATE TESTING 3.1.46 CRITERION 53 - PROVISIONS FOR 3.1-35 CONTAINMENT TESTING AND INSPECTION 3.1.47 CRITERION 54 - PIPING SYSTEMS 3.1-35 PENETRATING CONTAINMENT CRITERION 55 - REACTOR COOLANT 3.1-36 3.1.48 PRESSURE BOUNDARY PENETRATING CONTAINMENT 3.1.49 CRITERION 56 - PRIMARY CONTAINMENT 3.1-37 ISOLATION 3.1.50 CRITERION 57 - CLOSED SYSTEM 3.1-38 ISOLATION VALVES l 3.1.51 CRITERION 60 - CONTROL OF RELEASES OF 3.1-38 RADIOACTIVE MATERIAL TO THE ENVIRONMENT O Amendment D iv September 30, 1968
CESSAR E!nincmou ((() TABLE OF CONTENTS (Cont'd) CHAPTER 3 i l Section Subiect Pace No. j 3.1.52 CRITERION 61 - FUEL STORAGE AND 3.1-39 HANDLING AND RADIO- { ACTIVITY CONTROL l 3.1.53 CRITERION 62 - PREVENTION OF 3.1-39 CRITICALITY IN FUEL STORAGE AND HANDLING 3.1.54 CRITERION 63 - MONITORING FUEL AND 3.1-39 O WASTE STORAGE 3.1.55 CRITERION 64 - MONITORING RADIO- 3.1-40 ACTIVITY RELEASES j 3.2 CLASSIFICATION OF STRUCTURES. 3.2-1 COMPONENTS, AND SYSTEMS 3.2.1 SEISMIC CLASSIFICATION 3.2-1 3.2.2 SYSTEM QUALITY GROUP CLASSIFICATIONS 3.2-2 (SAFETY CLASS) l 3.3 WIND AND TORNADO IDADINGS 3.3-1 3.3.1 WIND LOADINGS 3.3-1 3.3.1.1 Desian Wind Velocity 3.3-1 1 3.3.1.2 Determination of ADDlied Forces 3.3-1 3.3.2 TORNADO LOADINGS 3.3-1 3 3.2.1 Applicable Desian Parameters 3.3-1 D 3.3.2.2 Determination of Forces on Structures 3.3-2 3.3.2.3 Effect of Failure of Structures or 3.3-2 Components not Desianed for Tornado j Loads : 3.4 WATER LEVEL (FLOOD) DESIGN 3.4-1 3.4.1 FLOOD ELEVATIONS 3.4-1 Amendment D 1 v September 30, 1988 i i 1 o J
C E S S A R EH L uou O TABLE OF CONTENTS (Cont'd) CHAPTER 3 Section 8_11bject Page No. 3.4.2 PHENOMENA CONSIDERED IN DESIGN 3.4-1 LOAD CALCULATION 3.4.3 FLOOD FORCE APPLICATION 3.4-1 3.4.4 FLOOD PROTECTION 3.4-1 D 3.4.4.1 Flood Protection Measures for Seismic 3.4-1 Catecory I Structures 3.4.4.2 Permanent Dewaterina System 3.4-2 3.4.5 ANALYTICAL AND TEST PROCEDURES 3.4-2 3.5 MISSIIJ PROT _ECTION 3.5-1 3.5.1 MISSILE SELECTION AND DESCRIPTION 3.5-1 3.5.1.1 IIlternally Generated Missiles 3.5-2 (Outside_ Containment) 3.5.1.1.1 Auxiliary Pumps and Motors 3.5-2 3.5.1.1.2 Valves 3.5-3 D 3.5.1.1.3 Pressure Vessels 3.5-3 3.5.1.2 1pternally Generatf_d Missiles 3.5-3 LJnside [lontaiyment;_)_ 3.5.1.3 Turbine Missiles 3.5-4 3.5.1.4 Missiles Generated by_1[a_t_ ural 3.5-4 Phenomena 3.5.1.5 Mjssiles Generated bl Fvents 3.5-4 Near the Site 3.5.1.6 Aircraft Hazard 3.5-4 3.5.2 STRUCTURES, SYSTF.MS, AND COMPONENTS TO 3.5-4 D BE PROTECTED FROM EXTERNALLY GENERATED MISSILES O Amendment D vi September 30, 1988
CESSARna h u I G TABLE OF CONTJNTS (Cont'd) CHAPTER 3 Section Suhject Pajus No. 3.5.3 BARRIER DESIGN PROCEDURES 3.5-4 3.5.3.1 Local Damane Prediction 3.5-5 3.5.3.1.1 Concrete Structures and Barriers 3.5-5 0 3.5.3.1.2 Steel Structures and Barriers 3.5-5 3.5.3.2 Overall Damage Prediction 3.5-5 3.5.4 INTERFACE REQUIREMENTS 3.5-5 3.6 PROTECTION AGAINST DYNAMIC EFFECTS 3.6-1 A_SSOCIATED WITH THE POSTULATED RUPTURE OF PIPING g- 3.6.1 POSTULATED PIPING FAILURES IN 3.6-1 ( FLUID SYSTEMS 3.6.1.1 Desian Bases 3.6-1 E 3.6.1.1.1 High-Energy Piping Systems 3.6-2 3.6.1.1.2 Moderate-Energy Piping Systems 3.6-3 3.6.1.2 Description 3.6-4 3.6.1.3 Safety Evaluation 3.6-8 lE 3.6.2 DETERMINATION OF BREAK LOCATIONS AND 3.6-12 DYNAMIC EFFECTS ASSOCIATED WITH THE POSTULATED RUPTURE OF PIPING 3.6.2.1 Criteria Used to Define Break and 3.6-12 Crack Locations and Configurations 3.6.2.1.1 General Requirements 3.6-12 3.6.2.1.2 Postulated Rupture Descriptions 3.6-13 E 3.6.2.1.3 Piping Evaluated for Leak- 3.6-14 Before-Break 3.6.2.1.4 Piping Other Than Piping Evaluated 3.6-14 for Leak-Before-Break O 3.6.2.1.4.1 Postulated Rupture Locations 3.6-14 (_s/ 3.6.2.1.4.2 Postulated Rupture Configurations 3.6-18 Amendment E vii December 30, 1988 l =
CESSARnah . O TABLE OF CONTENTS (Cont'd) CHAPTER 3 Dection Subipet Pace No. 3.6.2.1.5 Details of Containment Penetrations 3.6-19 3.6.2.2 Analytical Methods to Define Forcina 3.6-20 Eunctions and Response Models 3.6.2.2.1 Piping Evaluated for Leak-Before-Break 3.6-20 3.6.2.2.2 Analytical Methods to Define Forcing 3.6-20 Functions and Responso Models for Piping Excluding That Evaluated for Leak-Before-Break 3.6.2.2.2.1 Determination of Pipe Thrust 3.6-20 and Jet Loads 3.6.2.2.2.2 Methods for the Dynamic Analysis 3.6-21 of Pipe Whip 3.6.2.2.2.3 Method of Dynamic Analysis of 3.6-22 Unrestricted Pipes 3.6.2.3 Dynamic Analysis Methods to Verify 3.6-22 Intecrity and Operability 3.6.2.3.1 Pipe Whip Restraints and Jet 3.6-22 Deflectors for Piping Evaluated for E Leak-Before-Break 3.6.2.3.2 Pipe Whip Restraints and Jet 3.6-22 Deflectors for Piping Other than that Evaluated for Leak-Before-Break 3.6.2.3.2.1 General Description of Pipe 3.6-22 Whip Restraints 3.6.2.3.2.2 Pipe Whip Restraint Componer;ts 3.6-23 3.6.2.3.2.3 Design Loads 3.6-23 3.6.2.3.2.4 Allowable Stresses 3.6-24 3.6.2.3.2.5 Design Criteria 3.6-24 3.6.2.3.2.6 Materials 3.6-25 3.6.2.3.2.7 Jet Impingement Shields 3.6-25 3.6.2.4 Guard Pipe Assembly Desian Criteria 3.6-25 3.6.3 LEAK-BEFORE-BREAK EVALUATION PROCEDURE 3.6-25 3.6.3.1 Applicability of LBB 3.6-25 Amendment E viii December 30, 1988
l CESSAR E!nificuiu n V TABLE OF CONTENTS (Cont'd) CHAPTER 3 Section Subiect Pace No. 3.6.3.2 Leakace Crack Location 3.6-26 3.6.3.3 Leak Detection 3.6-26 3.6.3.3.1 Leak Detection System 3.6-26 3.6.3.3.2 Flow Rate Correlation 3.6-26 i 3.6.3.4 Screenina of Leakage Crack Sizes 3.6-27 Usino EPRI/GE Estimation Scheme 3.6.3.5 Material Properties 3.6-27 3.6.3.6 Leakage Crack Size Determination 3.6-28 3.6.3.7 Computation of J-Intearal Values 3.6-28 O.
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3.6.3.7.1 Range of Crack Sizes 3.6-28 3.6.3.7.2 J-Integral 3.6-28 E 3.6.3.8 Stability Evaluation 3.6-29 3.6.3.9 Results 3.6-29 l APPENDIX 3.6A' DISCUSSION OF FINITE DIFFERENCE ANALYSIE 3.6A-1 FOR ANALYSIS OF PIPE WHIP 3.7 SEISMIC DESIGN 3.7-1 3.7.1 SEISMIC INPUT 3.7-1 3.7.1.1 Desian Fesponse Spectra 3.7-1 3.7.1.2 Desian Time History 3.7-1 3.7.1.3 Critical Dannina Values 3.7-1 3.7.1.4 Supportina Media for Catecorv I Structures 3.7-1 3.7.2 SEISMIC SYSTEM ANALYSIS 3.7-2 3.7.2.1 Reactor Coolant System 3.7-2
'(\v) 3.7.2.1.1 Introduction 3.7-2 Amendment E ix December 30, 1988 l
1 CESSAR naincau:u i O TABLE OF CONTENT 8 (Cont'd) CHAPTER 3 Section Subiect Page No. 3.7.2.1.2 Mathematical Models 3.7-3 3.7.2.1.3 Calculations 3.7-4 3.7.2.1.4 Results 3.7-9 3.7.2.1.5 Conclusion 3.7-9 3.7.2.2 Natural Frequencies and Response Loads 3.7-9 3.7.2.3 Procedure Used For Mede]ing 3.7-9 3.7.2.4 Soil / Structure Interaction 3.7-9 3.7.2.5 Development of Floor Response Spectra 3.7-9 l 3.7.2.6 Three Components of Earthauake Motion 3.7-9a 3.7.2.7 Procedure for Combinina Modal Re_sponses 3.7-9a 3.7.2.8 Interaction of Non-Catecory I 3.7-10 Structures with Seismic Catecorv I Structures 3.7.2.9 Effects of Parameter Variations on 3.7-10 floor Response Sp_egtra 3.7.2.10 Use of Constant Vertical Static Factors 3.7-10 3.7.2.11 Torsional Effects of Eccentric Masses 3.7-10 3.7.2.12 Comparison of Responses 3.7-10 3.7.2.13 Methods for Seismic Analysis of Dams 3.7-10 3.7.2.14 Determination of Spismic Cateaorv I 3.7-10 Structure Overturning Moments 3.7.2.15 Analysis Procedurg__f_or Dampina 3.7-10 3.7.3 SEISMIC SUBSYSTEM ANALYSIS 3.7-11 3.7.3.1 Eg;ismic Analysis Methods 3.7-11 3.7.3.2 Determination of Number of Earthauake 3.7-11 Cycles X
CESSAR 8Hnfic 1cn
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TABLE OF CONTENTS (Cont'd) CHAPTER 3 Section Subiect Pace No. 3.7.3.7 Combination of Modal Response 3.7-12 3.7.3.8 Analytical Procedures for Pinina 3.7-12 3.7.3.9 Multiply Suororted Eauipment Components 3.7-12 With Distinct Inputs l 3.7.3.10 Use of Constant Vertical Static Factors 3.7-12 3.7.3.11 Torsional Effects of Eccentric Masses 3.7-12 3.7.3.12 Duripd Seismic _Ceceaorv 1 Pipina 3.7-12 Systems and Tunnels 3.7.3.13 Interaction of Other Pinina With 3.7-12 ("Sg Cateaory I Pipina V 3.7.3.14 Seismic Analysis of Reactor Internals. 3.7-12a Core and CEDMs l 3.7.3.14.1 Reactor Internals and Core 3.7-12a 3.7.3.14.1.1 Mathematical Models 3.7-13 3.7.3.14.1.2 Analytical Techniques 3.7-15 3.7.3.14.1.3 Results 3.7-18 3.7.3.14.2 Control Element Drive Mechanisms 3.7-18 , (CEDM) l 3.7.3.14.2.1 Input Excitation Data 3.7-18 3.7.3.14.2.2 Analysis 3.7 3.7.3.14.2.3 Tests 3.7-18 3.7.3.15 Analysis Procedure for Damoina 3.7-19 3.7.4 SEISMIC INSTRUMENTATION PROGRAM 3.7 3.8 DESIGN OF CATEGORY I STRUCTURES 3.8-1 3.8.1 CONCRETE CONTAINMENT 3.8-1 l lV 3.8.2 STEEL CONTAINMENT SYSTEMS 3.8-1 xi I
CESSAR nnincuiu 9 TABLE OF CONTENTS (Cont'd) CHAPTER 3 i Srsction S.u_bject P_ age No. 3.8.3 CONCRETE AND STEEL INTERNAL STRUCTURES OF 3.8-1 STEEL OR CONCRETE CONTAINMENTS 3.8.3.1 Description of the Internal Structures 3.8-1 3.8.3.2 Applicable Codes, Standards, And ] Specifications 3.8-1 ' 3.8.3.3 Loads and Load Combinations 3.8-1 3.8.3.4 Desian and Analysis Procedurch 3.8-1 3.8.3.5 Structural Acceptance Criteria 3.8-1 3.8.4 OTHER SEISMIC CATEGORY I STRUCTURES 3.8-1 3.8.5 FOUNDATIONS 3.8-1 3.9 MECHANICAL SYSTEMS AND COMPONENTS 3.9-1 3.9.1 SPECIAL TOPICS FOR MECHANICAL COMPONENTS 3.9-1 3.9.1.1 Desian Transients 3.9-1 3.9.1.2 Computer Procrams Used in Stress Analysis 3.9-3 3.9.1.2.1 Reactor Coolant System 3.9-3 3.9.1.2.1.1 MDC STRUDL 3.9-3 3.9.1.2.1.2 C-E MARC 3.9-4 3.9.1.2.1.3 JEST 3.9-4 3.9.1.2.1.4 SUPERPIPE 3.9-5 3.9.1.2.1.5 DFORCE 3.9-5 3.9.1.2.1.6 SG LINK 3.9-6 3.9.1.2.1.7 CEDAGS 3.9-6 3.9.1.2.1.8 CE177, Head Penetration 3.9-6 Reinforcement Program 3.9.1.2.1.9 CE102, Flange Fatigue Program 3.9-7 3.9.1.2.1.10 CE105, Nozzle Fatigue Program 3.9-7 E 3.9.1.2.1.11 CEC 26, Edge Coefficients Program 3.9-7 3.9.1.2.1.12 CE124, Generalized 4 x4 Program 3.9-7 3.9.1.2.1.13 SEC 11 3.9-8 Amendment E xii December 30, 1988
1 (# h k!k bb $ ICATICN l I i s-i TABLE OF CONTENTS (Cont'd) q l CHAPTER 3 q l 1 Secti_on Subiect Pace.No. 3.9.1.2.1.14 ANGYS 3.9-8 I J' 3.9.1.2.1.15 CE301, The Structural Analysis 3.9-8 for Partial Penetration Nozzles, l Heater Tube Plug Welds, and the Water Level Boundary of the Pressurizer Shell Program l 1 3.9.1.2.1.16 CE223, Primary Structure 3.9-8 1 Interaction Program 3.9.1.2.1.17 CE362, Tube-To-Tubesheet Weld 3.9-9 ) Program j 3.9.1.2.1.18 CE286, Support Skirt Loading 3.9-9 { Program { 3.9.1.2.1.19 CE210, Principal Stress Program 3.9-9 3.9.1.2.1.20 CE211, Nozzle Load Resolution 3.9-9 Program g- 3.9.1.2.1.21 KINI 2100 Program 3.9-9 ] t 3.9.1.2.1.22 CEFLASH-4A 3.9-10 1 3.9.1.2.1.23 CRIBE 3.9-10 3.9.1.2.2 Code Class CS Internals, Fuel and CEDMs 3.9-10 E 3.9.1.2.2.1 MRI/STARDYNE 3.9-10 i 3.9.1.2.2.2 ANSYS 3.9-12 3.9.1.2.2.3 ASHSD 3.9-12 3.9.1.2.2.4 CESHOCK 3.9-13 3.9.1.2.2.5 SAMMSOR/DYNASOR 3.9-14 3.9.1.2.2.6 MODSK 3.9-15 3.5.1.2.2.7 SAPIV 3.9-16 3.9.1.2.2.8 CEFLASH-4B 3.9-16 1 3.9.1.2.2.9 LOAD 3.9-17 l 3.9.1.2.3 Non-NSSS Structures and Components 3.9-17 3.9.1.3 Experimental Stress Analyses 3.9-17 1 3 . '9 .1. 4 Considerations for the Evaluation of 3.9-17 the Faulted Condition 3.9.1.4.1 Seismic Category I.RCS Items 3.9-17 1 3.9.1.4.1.1 Reactor Internals and CEDMs 3.9-19 3.9.1.4.1.2 Non-Code Items 3.9-19 l l Amendment E l xiii December 30, 1988 1 L-
i (#! h h khk bb ICATl3N O' TABLE OF CONTENTS (Cont'd) I CHAPTER 3 i l Bection Bubiect Pace No. i 3.9.1.4.2 Seismic Category I Non-NSSS Items 3.9-19 3.9.2 DYNAMIC SYSTEM ANALYSIS AND TESTING 3.9-19 3.9.2.1 Pipina Vibrations. Thermal Expansion 3.9-19 and Dynamic Effects 3.9.2.1.1 Steady-State Vibration 3.9-20 l 3.9.2.1.2 Transient Vibration 3.9-20 3.9.2.1.3 Thermal Expansion 3.9-21 3.9.2.2 Seismic F:1alification Testina of 3.9-21 Safetv-ienated Mechanical Eauipment 3.9.2.2.1 Nuclear Steam Supply System 3.9-21 3.9.2.2.2 Non-NSSS Iteins 3.9-21 3.9.2.2.2.1 Seismic Testing and Analysis 1.9-22 3.9.2.2.2.2 Seismic Analysis 3.9-22 3.9.2.2.2.3 Basis for Test Input Motion 3.9-22 3.9.2.2.2.4 Random Vibration Input 3.9-22 3.9.2.2.2.5 Input Motion 2.9-22 E 3.9.2.2.2.6 Fixture Design 3.9-23 3.9.2.2.2.7 Equipment Teat'ig 3.9-23 3.9.2.3 Dynamic System Ar._1< sis Methods for 3.9-23 Reactor Vessel Core Support and Internal Structures j l 3.9.2.3.1 Introduction 3.9-23 ) 3.9.2.3.2 Periodic Forcing Function 3.9-24 ! f 3.9.2.3.2.1 Core Support Barrel Assembly 3.9-24 3.9.2.3.2.2 Upper Guide Structure 3.9-24 3.9.2.3.2.3 Lower Support Structure 3.9-24 . Assembly i 3.9.2.3.3 Random Forcing Function 3.9-25 i 3.9.2.3.3.1 Core Support Barrel Assembly 3.9-25 3.9.2.3.3.2 Upper Guide Structure 3.9-2S 3.9.2.3.3.3 Lower Support Structure 3.9-26 Assembly 1 i Amendment E i xiv December 30, 1938 I J
CESSAR EHMncma
/ \s TABLE OF CONTENTS (Cont'd)
CHAPTER 3 Section Bubiect Pace No 3.9.2.3.4 Mathematical Models 3.9-26 3.9.2.3.5 Response Analysis 3.9-27 3.9.2.3.5.1 Deterministic Response 3.9-27 l 3.9.2.3.5.2 Random Response 3.9-27 3.9.2.4 Comprehensive Vibration Assessment 3.9-28 Procram (CVAPl 3.9.2.5 Dynamic System Analysis of the Reactor 3.9 E and CEDMs Under Faulted Conditions 3.9.2.6 Correlation of Test and Analytical 3.9-29 Results 3.9.3 ASME CODE CLASS 1, 2 and 3 COMPONENTS, 3.9-30 5 COMPONENT SUPPORTS AND CLASS CS CORE SUPPORT STRUC'iURES 3.9.3.1 Loading Combinations. Desian 3.9-30 Transients and Stress Limits 3.9.3.1.1 ASME Code Class 1 Components and 3.9-31 Supports 3.9.3.1.2 Core Support Structures (Class CS) 3.9-31 E and Internal Structures (Class IS) l 3.9.3.1,3 ASME Code Class 2 and 3 Components 3.9-32 and Supports 3.9.3.1.3.1 Tanks, Heat Exchangers, and 3.9-32 Filters 3.9.3.1.3.2 Valves 3.9-33 3.9.3.1.3.3 Pumps 3.9-33 3.9.3.1.4 Piping and Piping Supports 3.9-35 < 3.9.3.1.4.1 ASME Code Class 1 3.9-35 3.9.3.1.4.2 ACME Code Class 2-and 3 3.9-35 3.9.3.2 Pump and Valve Operability Assurance 3.9-36 E b, ) 3.9.3.2.1 Active ASME Code Class 2 and 3 3.9-36 Pumps and Class 1, 2 and 3 Valves Furnished with the NSSS Amendment E xv December 30, 1988 l
hhk E ICATitN O I TABLE OF CONTENTS (Cont'd) j CHAPTER 3 { l i Subiect Page No. I Sec. tion 3.9.3.2.1.1 Operability Assurance Program 3.9-36 3.9.3.2.1.2 Operability Assurance Program Results 3.9-37 ' for Active Pumps 3.9.3.2.1.3 Operability Assurance Program for 3.9-37 Active Valves 3.9.3.2.1.3.1 Pneumatically Operated Valves 3.9-39 3.9.3.2.1.3.2 Motor Operated Valves 3.9-40 3.9.3.2.1.3.3 Pressurizer Safety Valves 3.9-41 3.9.3.2.1.3.4 Check Valves 3.9-42 3.9.3.2.2 Non-NSSS Active ASME Code Class 3.9-43 E 2 and 3 Pumps and Class 1, 2 and 3 Valves 3.9.3.2.2.1 Pumps 3.9-43 3.9.3.2.2.2 Valves 3.9-45 3.9.3.3 Desian and Installation Details for 3.9-47 Mountina of Pressure Relief Devices 3.9.3.4 Component Supports 3.9-47 3.9.4 CONTROL ELEMENT DRIVE MECHANISMS 3.9-48a 3.9.4.1 Descriptive Information of CEDM 3.9-48a I 3.9.4.1.1 Control Element Drive Mechanism 3.9-49 Design Description 3.9.4.1.1.1 CEDM Pressure Housing 3.9-50 3.9.4.1.1.2 Motor Assembly 3.9-50 l 3.9.4.1.1.3 Coil Stack Assembly 3.9-50 ' 3.9.4.1.1.4 Reed Switch Assembly 3.9-51 3.9.4.1.1.5 Extension Shaft Assembly 3.9-51 I 3.9.4.1.2 Description of the CEDM Motor 3.9-51 Operation 3.9.4.1.2.1 Operating Sequence for the 3.9-51 Double Stepping Mechanism O Amendment E xvi December 30, 1988 f i l
CESSAR nnikmon
- = _ _ . ,
O TU1I&_,OF CONTENTS (Cent'd) CHAPTER 3 Section gubj eri Pace No. 3.9.4.2 &!( Cylde_SEDlLDS&;m:Jpegifications 3.9-53 3.9,4.3 [82eran Lgads Strfepy _LimLts hnd 3.9-53 Ellowable Deformation 3.9.4.4 CEDM Perforg;3cgfwiprance Proaram 3.9-54 3.9.4.4.1 CEDM Testing 3.9-54 3.9.4.4.1.1 Prototype Accelerated Life Tests 3.9-54 3.9.4.4.1.2 First Froduction Test 3.9-56 3.9.4.4.1.3 Operating Experience at the Palo 3.9-57 Verde Nuclear Generating Station E 3.9.5 REACTOR VESSEL CORE SUPPORT AND INTERNALS 3.9-57 STRUCTURES 3.9.5.1 Desian Arrangements 3.9-57 3.9.5.1.1 Core Support Structure 3.9-57 3.9.5.1.1.1 Core Support Barrel 3.9-57 3.9.5.1.1.2 Lower Support Structure and 3.9-58 Instrument Nozzle Assembly 3.9.5.1.1.3 Core Shroud 3.9-59 3.9.5.1.2 Upper Guide Structure Assembly 3.9-59 3.9.5.1.3 Flow Skirt 3.9-60 3.9.5.1.4 In-Core Instrumentation Support 3.9-60 System 3.9.5.2 Desian Loadina Conditions 3.9-62 3.9.5.3 Desian Loadina Cateaories 3.9-62 3.9.5.3.1 Level A and Level B Service Loadings 3.9-62 3.9.5.3.2 Level D Service Loadings 3.9-63 E 3.9.5.4 Desian Bases for Reactor Internals 3.9-63 3.9.6 IN-SERVICE TESTING OF PUMPS AND VALVES 3.9-64 0 3.9.6.1 In-Service Testina of Pumps 3.9-64 , Amendment E xvii September 30, 1988
CESSAR En!?nCATION e TABLE OF CONTENTS (Cont'd) CHAPTER 3 Eection Bubiect Pace No. 3.9.6.2 In-Service Testina of Valves 3.9-64 E 3.10 SEISMIC DESIGN OF SEISMIC CATEGORY I 3.10-1 INSTRUMENTATION AND ELECTRICAL EOUIPMENT 3.10.1 SEISMIC QUALIFICATION CRITERIA 3.10-1 3.10.2 METHODS AND PROCEDURES FOR QUALIFYING 3.10-1 ELECTRICAL EQUIPMENT AND INSTRUMENTATION 3.10.3 METHODS AND PROCEDURES OF ANALYSIS OR 3.10-3 TESTING OF SUPPORTS OF ELECTRICAL EQUIPMENT AND INSTRUMENTATION 3.11 ENVIRONMENTAL DESIGN OF MECHANICAL AND 3.11-1 ELECTRICAL EOUIPMENT 3.11.1 EQUIPMENT IDENTIFICATION AND ENVIRONMENTAL 3.11-2 CONDITIONS 3.11.2 QUALIFICATION TESTS AND ANALYSES 3.11-2 3.11.2.1 Mechanical and Electrical Component 3.11-2 Environmental Desian and Qualification for Normal OperatioD 3.11.2.2 Mechanical and Electrical Component 3.11-3 Environmental Desian and Qualification for Operation After a Desian Basis Event 3.11.3 QUALIFICATION TEST RESULTS 3.11-6 3.11.3.1 NPM Instrumentation and Electrical 3.11-6 Ecuipment D 3.11.3.2 NPM Mechanigal Ecuipment 3.11-6 3.11.4 CLASS 1E INSTRUMENTATION LOSS OF 3.11-6 VENTILATION EFFECTS 3.11.5 CHEMICAL SPRAY, RADIATION, HUMIDITY, DUST, 3.11-7 SUBMERGENCE AND POWER SUPPLY VOLTAGE AND FREQUENCY VARIATION Amendment E xviii December 30, 1988
CESSAR nairlCATION O TABLE OF CONTENTS (Cont'd) CHAPTER 3 Section Subject Pace No. 3.11.5.1 Chemical Environment 3.11-7 3.11.5.2 Radiation Environment 3.11-7 3.11.5.3 Humidity 3.11-8 3.11.5.4 Dust 3.11-8 3.11.5.5 Submergence 3.11-8 D 3.11.5.6 Power Supply Voltace and Frequelc_y 3.11-8 Variation APPENDIX TYPICAL ENVIRONMENTAL CONDITIONS AND TEST 3.11A-1 3.11A PROFILES FOR STRUCTURES AND COMPONENTS APPENDIX IDENTIFICATION AND LOCATION OF MECHANICAL 3.11B-1 3.11B AND ELECTRICAL SAFETY-RELATED SYSTEMS AND COMPONENTS E O Amendment E xix December 30, 1988
CESSAR8!ninc-O LIST OF TABLES CHAPTER 3 Table Bubiect
- 3.2-1 Classification of Structures, Systems, and Components D
3.2-2 Safety Class 1, 2 and 3 Valves 3.2-3 Relationship of Safety Class to Code Class 3.2-4 Summary of Criteria - Structures D 3.5-1 Kinetic Energy of Potential Missiles 3.6-1 High- and Moderate-Energy Fluid Systems 3.6-2 Systems Required for Safe Shutdown and/or to Mitigate the Consequences of a Design-Basis E Accident 3.6-3 High-Energy Lines Within Containment 3.6-4 High-Energy Lines Outside Containment 3.7.2-1 Damping Ratios Used In Analysis of Category 1 Structures, Systems, and Components
> 3.7.2-2 Natural Frequencies and Dominant Degrees of Freedom Reactor Coolant System 3.7.2-3 Natural Frequencies and Dominant Degrees of Freedom Pressurizer and Surge Line s
3.9-1 Transients Used in Stress Analysis of Code Class 1 Components 3.9-2 Loading Combinations ASME Code Class 1, 2, and 3 Components E 3.9-3 Stress Limits for ASME Code Class 1 Components, Piping, and Component Supports 3.9-4 Seismic I Active Valves 3.9-5 Stress Criteria for Safety-Related ASME Class 2 and Class 3 Vessels Amendment E xx December 30, 1988
CESSAR Enlincmou O LIST OF TABLES (Cont'd) CHAPTER 3 Iable Subiect 3.9-6 Stress Critoria for ASME Code Class 2 and Class 3 Inactive Pumps and Pump Supports 3.9-7 Design Criteria for Active Pumps and Pump Supports 3.9-8 Stress Criteria for Safety-Related ASME Code Class 2 and Class 3 Active and Inactive NSSS and Inactive BOP Valves 3.9-9 BOP Design Criteria for Active Valves 3.9-10 Loading Combinations for ASME Section III Class 1 Piping 3.9-11 Load Combinations and Acceptance Criteria for 9 Pressurizer Safety Valve Piping and Supports ASME Class 1 Portion 3.9-12 Loading Combinations for ASME Section III Classes 2 and 3 Piping E 3.9-13 Load Combinations for Safety Valve Piping ASME Class 2 and 3 Piping 3.9-14 Design Loading Combinations for ASME Code, Class 1, 2, and 3 Piping Supports 3.9-15 Stress Limits for CEDM Pressure Housings 3.9-16 Stress Limits for Design and Service Loads O Amendment E xxi December 30, 1988
)
CESSAR 88Hrbm,. O LIST OF FIGURES CHAPTER 3 Fiqures Subiect Variatic.. of J-Integral with Loads for a Typical E l 3.6-1 Case 1 3.7.1-1 Response Spectrum for Upper Reactor Vessel i Supports ! l 3.7.1-2 Response Spectrum for Upper Steam Generator Supports 3.7.1-3 Response Spectrum for Upper Reactor Coolant Pump l Supports ! 3.7.1-4 Response Spectrum for Vertical Supports for All Components I l i 3.7.2-1 Typical Reactor Coolant System Seismic Analysis ' Model 3.7.2-2 Typical Pressurizer Seismic Analysis Model 1 3.7.2-3 Typical Surge Line Seismic Analysis Model 1 i 3.7.3-1 Reactor Internals Horizontal Seismic Model 3.7.3-2 Reactor Internals Linear Horizontal Seismic Model 1 3.7.3-3 Reactor Internals Linear Vertical Seismic Model 1 3.7.3-4 Reactor Iliternals Nonlinear Horizontal Seismic Model 3.7.3-5 System 80 Core Seismic Model - One Row of 17 Fuel Assemblies 3.7.3-6 Core-Support Barrel Upper Flange Finite-Element Model 3.7.3-7 Lower Support Structure Finite-Element Model 3.7.3-8 Reactor Internals Non-Linear Vertical Seismic Model 3.9-1 Reactor Coolant System Supports Diagram Amendment E xxii December 30, 1988
CESSAR Ennnema O LIST OF FIGURES (Cont'd) CHAPTER 3 Fiqures Subiect
-3.9-2 Summary of Analytical. Methodology 3.9-3 ASHSD Finite Element Model of the CSB System l 3.9-4 Control Element Shroud Tube Finite Element Model 3.9-5 Lower Support S' -" +"ve Instrument Nozzle Assembly Finite Element Model 3.9-6 ICI Support Tube; Outer Position Finite Element Model 3.9-7 Skewed Beam Support Columns Finite Element Model 3.9-8 Control Element Drive Mechanism (Magnetic Jack) 3.9-9 Reactor Vertical Arrangement O 3.9-10 Core Support Barrel Assembly l 3.9-11 Reactor Vessel Core Support Barrel Snubber Assembly 3.9-12 In-core Instrument Support Structure 3.9-13 Core Shroud Assembly 3.9-14 Upper Guide Structure Assembly 3.9-15 In-core Instrument System F
i O Amendment E. ' xxiii December 30, 1988 ( Ll.--__--.--------___.______________-_______________-. . . . _ _ - -- . _ -.
CESSARnn%ma a hb i 3.0 DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND J EYSTEMR I 3.1 CONFORMANCE WITH NRC GENERAL DESIGN CRITERIA In this section, brief discussions are presented in response to l the current General Design Criteria for Nuclear Power Plants, 1 Appendix A to 10 CFR 50. These discussions summarize the manner I in which the principal design featurea meet the individual I criteria and . include references to sections of the safety ; analysis report where more detailed information is given. ' 3.1.1 CRITERION 1 - QUALITY '8TANDARDS AND RECORDS { Structures, systems, and components important to safety shall be i designed, fabricated, erected, and tested to quality standards j 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 i applicability, adequacy, and sufficiency and shall be supple- l mented or modified as necessary to assure a quality product in keeping with the required safety function. A quality assurance program shall be established and implemented .in order to provide adequate assurance that these structures, systems, and components will satisfactorily perform their safety functions. Appropriate records of the design, fabrication, erection, and testing of structures, systems, and components important to safe &y shall be maintained by or under the control of the nuclear power unit licensee throughout the life of the unit. RE6PONSE: l The structures, systems, and components described in CESSAR are classified according to their importance in the prevention and mitigation of accidents using the classification system described in ANSI /ANS 51.1. Each safety-related component is given a D safety class designation. The codes, standards, and quality control applicable to each component and its safety class designation are identified in Section 3.2, Where applicable, D design and fabrication are in accordance with the codes required in 10 CFR 50.55a. The quality assurance program conforms with l the requirements of 10 CFR 50, Appendix B, " Quality Assurance , Criteria for Nuclear Power Plants," and is presented in Chapter
- 17. Chapter 14 describes initial tests and operations to assure performance of installed equipment commensurate with the importance of the safety function.
The design, fabrication, and quality programs for components not p\ r included in the ANSI classification system are governed by U Amendment D 3.1-1 September 30, 1988 ___________-_-_______-O
CESSAR !!nhuou l industry codes appropriate to the application. Details of conformance to these codes are found in the appropriate CESSAR sections. 3.1.2 CRITERION 2 - DESIGN BASES FOR PROTECTION AGAINST NATURAL PHENOMENA Structures, systems, and components important to safety shall bo designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seichen without loss of capability to perform their safety ' functions. The design bases for these structures, systems, and components shall reflect: (1) Appropriate consideration of the most severe of the natural phenomena that have been historically reported for the site and surrounding area, with sufficient margin for the litlited accuracy, quantity, and period of time in which the historical data have been accumulated; (2) appropriate combinations of the effects of normal and accident nonditions with the effects of the natural phenomena; and, (3) the importance of the safety functions to be performed.
RESPONSE
The structures, systems, and components designated Seismic Category I are designed to withstand, without loss of function, the effects of any one of the most severe r.atural phenomena r D including flooding, hurricanes, tornadoes, and the Safe Shutdown Earthquake (SSE) (refer to Chapter 2). Design criteria for wind and tornado, flood and earthquake are discussed in Sections 3.3, 3.4, and 3.7, respectively. The seismic design of safety-related structures, systems, and components is consistent with conservative structural envelopes. These " envelopes" have been selected based on the design basis earthquakes at the majority of potential plant sites in the continental U.S., using current containment structure designs. In the design stage, normal operating and accident loads are appropriately combined with the seismic loads and allowable stress limits and deformations are defined so that safety functions are not jeopardized. 3.1.3 CRITERION 3 - FIRE PROTECTION : Structures, systems, and components important to safety shall bc ; designed and located to minimize, consistent with other safety ) requirements, the probability and effect of fires and explosions. j Noncombustible and heat resistant materials shall be used I wherever practical throughout the unit, particularly in locutions i O I t
)
Amendment D q J.1-2 september 30, 1988 l
CESSAR nai"icmou 1 l l v ' such as the containment and control room. Fire detection and
- l fighting systems of appropriate capacity and capability shall be l provided and designed to minimize the adverse effects of fires on i structures, systems, and components important to safety. .
Firefighting systems shall be designed to assure that their l rupture or inadvertent operation does not significantly impair the safety capability of these structures, systems, and components. , 1 EEJPONSE l l All pressure boundary components and structures and the attendant I auxiliary systems in System 80+ design scope are designed to minimize the probability and effects of fires and explosions. High grade noncombustible and fire resistant materials are used for components located in the containment, components of engineered safety feature system., and throughout the unit I wherever practical. A detailed functional description of 'che Fire Protection System is provided in Section 9.5.1. D l l 3.1.4 CRITERION 4 - ENVIRONMENTAL AND MTBSILE D2 SIGN BASES l l p structures, systems and components important to safety shall be
'j s designed to accommodate the effects of and to be compatible with I the environmental conditions associated with normal operation, j maintenance, testing, and postulated accidents, including loss of j 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 mLy result from equipment failures and from events and conditions outside the nuclear power unit. However, dynamic 4 effects associated with postulated pipe ruptures in nuclear power units may be excluded from the design basis when analyse, D reviewed and approved by the Commission demonstrate that the j probability of fluid system piping rupture is extremely low under i' conditions consistent with the design basis for the piping. EESPONS_Et C-E supplied structures, syster.s, and components important to safety are designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, l including loss-of-coolant-accidents (see Section 3.11).
]i Where appropriate, Standardized Functional Descriptions will D
include design requirements to ensure that these structures, I nystems, and component.s will be appropriately prctected against I oh (d dyr.amic effects (including the effects of missiles, pipe l
)
I Amendment D j 3.1-3 September 30, 1988 i l
CESSARnn%ma whipping, and discharge of fluids) that may result from equipment failures, postulated accidents, and from events and conditions outside the nuclear power unit. I The reactor building is capable of withstanding the effects of missiles originating outside the containment such that n D credible missile can result in a LOCA. The control room is designed to withstand such missiles as may bo directed toward it and still maintain the capability o.f controlling the plant. l 3.1.5 CRITFRION 5 - SHARING OF STRUCTURES, SYSTEMS, AND COMPONENTS Structures, systems, and components important to safety shall not be shared among nuclear power units unless it can be shown that such sharing will not significantly impair their ability to perform their safety functions, including, in the event of an accident in one unit, an orderly shutdown and cooldown of the remaining units. BEf@H8 The System 80+ design is based on non-shared systens. 3.1.6 CRITERION 10 - REACTOR DESIGN The reactor core and associated coolant, control and protection I systems shall be designed with appropriate margiri to assure that specified acceptable fuel design limits are not axceeded during ' any condition of normal operation, including the effects of anticipated operational occurrences. l BERMHFE I 1 Specified Acceptable Fuel Design Limits (SAFDLs) are stated in Section 4.4.1. Operation within the operating limits (Limiting i conditions for Operation) specified by the Technical i Specifications will keep the reactor f9e1 within the SAFDLs for normal operation and during any Anticipated Operational g j Occurrence. j 1 The plant is designed such i: hat operation within Limiting ( Conditions for Operation with safety system settings not less j conservative than the Limiting Safety System Settings prescribeQ j in the Technical Specifications reeults in confidence that SAFDLs will not be exceeded as a result of any Anticipated Operational D i occurrence. Operator action, aided by the control systeris and monitored by plant instrumentation, maintains the plant within Limiting Conditions fer Operation during normal operation. Amendment D 3.1-4 September 30, 1928
CESSAR Mnine m. - - - O V l See the following sections for additional information: l A. Fuel System Design, Section 4.2 ) i I B. Reactor Coolant System, Chapter 5 C. Shutdown Cooling System, Section 5.4.7 0 D. Runctor Protective System, Section 7.2 i E. Analysis of Anticipated Operational Occurrences, Ch. apter 15 F, Technical Specifications, Chapter 16 1 3.1.7 CRITERION 11 - REACTOR INHERENT PROTECTION The reactor core and associated coolant systems 6 hall be designed so that in the power ciperating range the nefe. effect of the prompt inherent nuclear feedback characteristics tends to compensate for j a rapid increase in reactivity. 1 TJSPONH: In the power operating range, the combined response'cf the fuel b' temperature coefficient, the moderator temperature coefficient, the moderator void coefficient, and the moderator pressure coefficient to an increase in reactor power will be a decrease in reactivity; i.e., the inherent nuclear feedback characteristics will not be positive. The reactivity coefff cients for this reactor are discussed in detail in Section 4.3. 3.1.8 CRITERION 12 - SUPPRESSION OF REACTOR POWER t 05CILLATIOND The reactor core and associated coolant, control, and protection [ systems shall be designed to assure that power oscillations which l can result in conditions exceeding specified . acceptable fuel 1 design limits are not possible or can be reliably and readily f detected and suppressed. RESPOH H: Power level oscillations do not occur. The effect of the ; negative 1 power <:oef ficient of reactivity (see GDC ~ 11, Section 1 3.ls7), together with the coolant te:mperature program maintained l by control of regulating rods and soluble boron, provides ) y fundamental mode stability. Power level is continuously j ( monitored by neutron flux detectors (Chapter 7). l l l Amendment D l 2.1-5 September 30, 1988 i
.- _ _ _ __ - - __ A
l l CESSAR nainCATION P9wer distribution oscillations are detected by neutron flux detectors. Axial mode oscillations are suppressed by means of part-strength or full-strength neutron absorber rods. All other D l modes of oscillation are expected to be convergent. Monitoring' and protective requirements imposed by Criteria 10 and 20 are j discussed in Sections 3.1.6, 3.1.16 and in Chapter 4. l 3.1.9 CRITERION 13 - INSTRUMENTATION AND CONTROL Instrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrence, and for accident conditions as appropriate to assure adoquate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure bcundary, and the containment .and its associated systems. Appropriate controls shall be provided to maintain these , variables and systems within prescribed operating ranges.
RESPONSE
Instrumentation is provided to monitor significant process variables which can affect the fission process, the integrity of the reactor core, the Reactor Coolant Pressure Boundary (RCPB) ) and their associated systems. Controls are provided for the ; purpose of maintaining these variables within the limits prescribed for safe operation. Instrumentation for the containment and its associated systems can be found in the appropriate CESSAR chapters and in the site-specific SAR. The D i principal process variables to be monitored and controlled arm A. Neutron flux level (reactor power) B. CEA positions C. Neutron flux distribution (at various axial positions) D. Recctor coolant temperature and pressure E. Reactor pump speed F. Pressurizer level ) G. Steam generator level and pressure In adtlition, Departure from Nucleate Boiling Ratio (DNBR) margin and Local Power Density (LPD} margin, in kW/ft, are also ! monitored. f O\ Amendment D 1 3.1-6 September 30, 1988 )
CESSAR 8HMemon f l k The Plant Protection System (PPS) consists of the Reactor , Protective System (RPS) and the Engineered Safety Features i Actuation System (ESFAS). The RPS is designed to monitor NSSS operating conditions and to initiate reliable . and rapid reactor shutdown if monitored variables or combinations of monitored variables deviate from the permissible operating range to a degree that a safety limit may be reached. j The ESFAS is designed to monitor plant variables and to actuate Engineered Safety Feature (ESF) systems during a design basis event. The following are provided to monitor and maintain control over the fission process during transient and steady state periods over the lifetime of the core: A. Redundant channels of ex-core nuclear instrumentation, which ; constitute the primary means of monitoring the fission I process for protection, control and low power operation. D I B. Redundant and diverse CEA position indicating systums for each CEA. C. Manual and automatic control of reactor power by mans of CEAs. J i D. Manual regulation of coolant boron concentration. j I E. A Boronometer, which determines ' the boron concentration in { the reactor coolant by neutron absorption, provided as a ! backup to the primary method of determining soluble poison concentration by routine sampling and analysis of reactor coolant.
]
F. In-core instrumentation, provided to supplement information on core power distribution and to enable calibration of ex-core flux detectors. ) The non-nuclear instrumentation measures temperatures, pressures, I flows and levels in the Reactor Coolant System and main steam and i auxiliary systems and is used to maintain these variables within the prescribed limits. The instrumentation and control systems : are described in detail in Chapter 7. The Doronometer- is i discussed in Sections 7.7.1.1.7 and 9.3.2 while the process ! radiation monitor is discussed in Section 9.3.2* D i When it is required that a variable be monitored during and after a Design Basis Event (DBE), in addition to normal operation, the i Amendment D 3.1-7 September'30, 1988
CESSARnnhnw O results of analysis of the course of the event are used to ensure that the instruments provided will cover the range anticipated for the event conditions. J 3.1.10 CRITERION 14 - FEACTOR COOLANT PRE 8; DURE BOUNDARY The reactor coolant pressure boundary shall be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating fai?.ure, and of gross rupture. BESPOESD The reactor coolant pressure boundary it. defined in accordance with 10 CFR 50.2(v) and ANSI /ANS S1.1 (see response tc GDC 55, O Section 3.1.48). Reactor Coolant System components are desig,ned to meet the requirements of the ASMF. Code, Section III. To establish operating pressure and temperature limitations during startup and shutdown of the Reactor Coolant System, the fracture toughness rules defined in the ASME Code, Section III, are followed. Qualjty ceintrol, inspection, and testing are periormed as required by ASME Section III and allowable reactor pressure-temperature operations are specified to ensure the integrity of the Reactor Coolant System. ! The reactor coolant pressure boundary is designed to accommodate the system pressures and temperatures attained under all expected j mades of unit operation including $11 anticipated transients, and maintain the stresses within applicable limits. Piping and equipment pressure parts of the reactor coolant pressure boundary are assembled and erected by welding unless applicable codes permit flanged or screwed joints. Helding procedures are employed which produce welds of complete fusion and free of unacceptable defec*.s. All welding precedures, , welders, and welding machine operators tre qualified in j accordance with the requirements of Section IX cf the ASE Boiler ) and Pressure Vecscl Code for the raterisls to be welded. , Qualification records, including the results of the procedure and performance qualification tests and identification symbols ! assigned to each velder are maintained. ) J The pressure bounchry has provisions for lyservice inspection in i accordance with Section XI of the ASME Boiler and Pressure Vessel l Code, to ensure continuance of the structural and leak-tight ) integrity of the boundary (see response to GDC 22, Section 3.1.28). For the reactor vessol, a material surveillance program I conforming with the requiremonte of Appendix H to 10 CFR 50 is I provided. ; Amendment D 3.1-8 September 30, IS88 m - j
l CESSAR!ancm. j 1 l b CRITERION 15 - REACTOR COOLANT SYSTEM DESIGN i 3.1.11 1 The Reactor Coolant System (RCS) and associated auxiliary, i control, and protection systems shall be designed with sufficient nargin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal Operational occurrences. REFJONSEs - 1 The design criteria and bases for the reactor coolant pressure boundary are described in the response to Criterion 14. I The operating conditions for. normal steady state and transient I plant operations are cetablished conservatively. Normal 4 operating li'mits are selected so that an adequate 7.argin exists 3 between ' chem and the design limits. .The plant control systems q are designed to ensure that plant variables are maintained well ; within the establishe1 operating limits. 'The plant transient 1 response characteristics and pressure and temperature distributions during normal operations are considered in the design as well as the accuracy and response of.the instruments ! p and controls. These design techniques ensure that a satisfactory { margin is maintained between the plant's normal operating condi- ] tions, including design transients, and the design limits for the 1 reactor coolant pressure boundary. I i Plant control systems function to minimize the deviations from I normal operating limits in the event of most Anticipated i Operational Occurrences. Where control systems responso would M ] inadequate or fail upon demand, the Plant Protection System ; functions to mitigate the consequences of such events. I The Plant Protection System functions to mitigate the consequences in the eve.nt of accidents. Analyses show that the ! design limits for the 7/eactor coolant pressure boundary are not exceeded in the event of any ANST/ANS SA.1 conditions,, D 3.1.12 CRITERION 16 - CONTAINMENT DESIGi{ I Reactor containment and associc.ted systems shall be provided to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environm6nt and to assure that the containment design conditions important to safety are not exceeded for as long as postulated accident conditions D require. I
\
Amendment D 3.1-9 September 30, 1988
.m-..f CESSARE L a O
RESPONB3 ) The containment system is designed to protect the public from the f consequences of a LOCA, based on the equivalent energy release of a postulated break of reactor coolant piping up to and including a double-ended break of the largest reactor coolant pipe. i The containment vessel, shield building, and the associated D Engineered Safety Feature systems are designed to safely withstand all internal and external environmental conditions that may reasonably be expected to occur during the life of the plant, including both short- and long-term effects following a LOCA. Leak-tightness of the containment system and short- and long-term l performance following a LOCA are analyzed in Section 6.2. 2.1.13 CRITERION 17 - ELECTRICAL P0WER SYSTEMS An onsite electric power system shall be provided to permit functioning of structures, systems and components important to safety. The safety function for each system (assuming the other system is not functioning) shall be to provide sufficient capacity and capability to assure that (1) specified acceptable ) fuel design limits and design conditions of the reactor coolant pressare boundary are not exceeded as a result of anticipated operational occurrences and (2) the core is cooled and containment integrity and other vital functions are maintained in ; the event of postulated accidents. The onsite electric power supplies, including batteries, and the onsite electric distribution system, shall have sufficient independence, redundancy and testability to perform their safety functions assuming a single failure. Electrical power from the transmission network to the onsite electric distribution system shall be supplied by two physically independent circuits (not necessarily on separate right of ways) designed and located so as to minimize to the extent practical any likelihood of their simultaneous failure under operating und postulated accident and environmental conditions. A switchyard common to both circuits is acceptable. Each of these circuits shall be designed to be available in sufficient time following a loss of all onsite alternating current power supplies and the other offsite electric power circuit,, to assure that specified i acceptable fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded. One of these , ) circuits shall be dessgned to be available within a few seconds following a loss-of-coolant-accident to assure that the core cooling, containment integrity and other vital safety functions are maintained. ,
)
Amendment D 3.1-10 September 30, 1988
CESSAR n=nemon j l l lO Provisions shall be included to minimize the probability of losing electric power from any of the remaining supplies as a 4 result of, or coincident with, the loss of power generated by the nuclear power unit, the loss of power from the transmission I network or the loss of power from the onsite electric power supplies. RESPONJJ5: The System 80+ f,tandard Design is previ.ded with an- onsite electric power sistem and an offsite electric power system to , permit functioning of structures, sysr. ems and components important to safety in full compliance with the requirements of this criterion as described in Chapter 8. I l The onsite electric power system consists of separate, redundant I and independent distribution cystems and dedicated power supplies with sufficient capacity, capability, and testability to perform
)
l their safety functions assuming a single failure. J The offsite electric power system. consists of two physically ) independent circuits from the station switchyard. Each circuit D 1 is immediately available and has sufficient capacity and j j capability to perform its safety function. w ' Provisions are made to minimize the probability of losing I electric power from any of the remaining supplies an a result of, or coincident with, the loss of power generated by the nuclear power unit. ) 3.1.14 CRITERION 18 - INSPECTION AND TESTING O!? ELECT'RICAL I POWER SYSTEMS s Electrical power systems important to safety 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 systemn 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 system such as 3 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 system into operation, including operation of
)
j applicable portions of the protection system, and the transfer of j power among the nuclear power unit, the offsite power. system and ! the onsite power system. l O I l Amendment D
'3.1-11 September 30, 1988
I CESSARM$L mu i O RESPONSN: Electrical power systems important to safety are designed to permit appropriate periodic inspection and testing of important areas and features, such as wiring, insulation, connections, and 1 switchboards, "to assess the continuity of the systems and to detect deterioration, if any, of their components. Capability is provided to periodically test the operability and functional performance of the system components. The diesel generator sets will be started and loaded periodically on a routine basis, and D relays, switches, and buses will be inspected and tested for operation and availability on an individual basis. Transfer from normal to emergency sources of power will be made to check the operability of the systems and the full operational sequence that brings the systems into operation. Refer to Section 8.3.1, 8.3.2 and 16.4.8 for more detailed information. j 3.1.15 CRITERION 19 - CONTROL ROOM A control room shall be provided from which actions can be taken ' to operate the nuclear unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant-accidents. Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposure in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident. Equipment at appropriate locations outside the control room shall be provided (1) with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown. (2) with a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures.
RESPONSE
All control stations, switches, controllers, and indicators necessary to operate or shut the unit down and maintain safe control of the facility are located in the control room. The design of the control room permits safe occupancy during i abnormal conditions. The employment of non-combustible and fire I retardant materials in the construction of the control room, the D limitation of combustible supplies, the location of fire fighting equipment, and the continuous presence of a highly trained Amendment D 3.1-12 September 30, 1988 i
CESSARina mu I (O '%.) operator will minimize the possibility that the control room will become uninhabitable. Shielding is designed to maintain l tolerable radiation exposure levels following design basis l accidents. The control room will be isolated from the outside atmosphere during the initial period following the occurrence of ' an accident. The Control Room Ventilation System is designed to recirculate cool control room air as discussed in Sections 9.4.1 and 12.2. Radiation detectors and alarms are provided. 0 Emergency lighting is provided as discussed in Section 9.5.3. , Alternate local controls and instruments are available for equipment required to bring the plant to and maintain a hot standby condition. It is also possible to attain a cold shutdown condition from locations outside of the control room through the use of suitable procedures. Refer to Section 7.4.1.1.10. A discussion of the unit's control room is provided in Section 7.7.1.3 with human factors issues discussed in Chapter 18. A discussion of the hot and cold shutdown capability is provided in Section 7.4 for the systems required for safe shutdown. Discussion regarding adequate radiation protection for the unit's p control facilities is provided in Section 6.4 and in Chapter 12. D l 3.1.16 CRITERION 20 - PROTECTION SYSTEM FUNCTIONS The protection system shall be designed (1) to initiate automatically the operation of appropriate systeus including the l reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety. RESPONS8: A Plant Protection System (PPS), consisting of a Reactor Protective System (RPS) and an Engineered Safety Features Actuation System (ESFAS), is provided. The RPS automatically initiates a reactor trip when any of the monitored process variables reach a trip setpoint. The ESFAS automatically actuates Engineered Safety Feature (ESP) systems and their 0 support systems when any of the monitored process variables reach a predetermined setpoint. The trip setpoints of the RPS are selected to ensure that Design ; Basis Events (DBEs) which are expected to occur once or more
,s during the life of the nuclear generating station do not cause
( l the violation of SAFDLs. The reactor trips also help the ESF V systems in mitigating the consequences of DBEs which are expected Amendment D 3.1-13 September 30, 1988 ;
CESSAR nainemen O\ to occur once during the life of several plants as well as arbi-trary combinations of unplanned events and degraded systems that are never expected to occur, to within acceptable limits. Reactor trip is accomplished by de-energizing the Control Element Drive Mechaninm (CEDM) coils through the interruption of the CEDM j power supply either automatically or manually. The CEDM power supply is a pair of full capacity motor-generator sets. The patn Trip to the CEDMs is interrupted by opening the Reactor Switchgear. With the CEDM coils de-energized, the CEAs are released to drop into the core by gravity, rapidly inserting negative reactivity to shut the reactor down. The CEDMs are i described in Section 4.2, the specific reactor trips used are described in Section 7.2. The ESF systems are actuated to minimize the effects of incidents ! which could occur. Controls are provided for manual actuation of the ESF system. The process variables which automatically lD actuate the ESF system and the circuitry arrangements for the ESFAS are discussed in Section 7.3. The ESF systems are discussed in Chapter 6. The SAFDL on linear heat rate and DNBR are intended to enforce the principal thermal hydraulic design basis given in Section 4.4.1 (i.e., the avoidance of thermally induced fuel damage during normal steady state operation and during Anticipated D Operational Occurrences). 3.1.17 CRITERION 21 - PROTECTION SYSTEM RELIABILITY AND l TESTABILITY The protection system shall be designed for high functional reliability and in-service testability commensurate with the safety functions to be performed. Redundancy and independence } designed into the protection system shall be sufficient to assure I that (1) no single failure results in loss of the protection l function and (2) removal from service of any component or channel does not result in loss of the required minimum redundancy unless the acceptable reliability of protection system operation can be j otherwise demonstrated. The protection system shall be designed I to permit periodic testing of its functioning when the reactor is I in operation, including a capability to test channels 1 independently to determine failures and losses of redundancy that may have occurred. i
RESPONSE
The PPS is designed to provide high functional reliability and in-service testability. The protection system is designed to comply with the requirements of IEEE 279-1971, " Criteria for , Protection Systems for Nuclear Power Generating Stations," and D i Amendment D 3.1-14 September 30, 1988 i
CESSARnn% - 1 l IEEE 603-1980, " Criteria for Safety Systems for Nuclear Power Generating Stations," and other standards as noted in Section D 7.1.2. No credible single failure will result in loss of the protection function. The protection channels are independent I with respect to wire routing, sensor mounting, and supply of power. Each channel of the protection system, including the sensors, up j to the RTSS and ESFAS actuation devices, is capable of being i checked during reactor operation. Process sensors of each I channel in the protection systems are checked by comparison of the redundant process sensor values using the discrete indications and alarms on the control room panels as described in D Section 7.7.1.3.1. Discrepancies among redundant channel sensors beyond specified limits are alarmed as described in Section 7.7.1.4.3 and Chapter 18. l The RTSS and ESFAS are described in Chapter 7. To minimize inadvertent actuation of an ESF system or an inadvertent reactor trip, the protection systems utilize a coincidence of two logics to operate. In addition, the channel being tested is bypassed so that the protection system converts O' to a two-out-of-three logic while maintaining the coincidence of two. This allows periodic testing and operation of the various protective functions without reducing the availability of the protection systems. 3.1.18 CRITERION 22 - PROTECTION SYSTEM INDEPENDENCE The protection system shall be designed to assure that the effects of natural phenomena, and of normal operating, maintenance, testing and postulated accident conditions on redundant channels do not result in loss of the protection function or shall be demonstrated to be acceptable on some other defined basis. Design techniques, such as functional diversity or diversity in component design and principles of operation, shall be used to the extent practical to prevent loss of the protection function.
RESPONSE
The protection systems conform to the independence requirements 1 of IEEE 279-1971. Four independent measurement channels, complete with sensors, sensor power supplies, signal conditioning units, and bistable trip functions are provided for each protective parameter monitored by the protection systems except i s for the CEA position sensors which are two-fold redundant. The l measurement channels are provided with a high degree of ' Amendment D 3.1-15 September 30, 1988
CESSARnennem - - . , O independence by separate connection of the channel sensors to the process systems. Refer to Chapter 7 for a more detailed discussion of the protection systems. Power to the protection system channels is provided by independent vital power supply buses. The power supply systems D are discussed in Chapter 8. Functional diversity is incorporated into the system design, to the extent practical, to prevent loss of the protective function. Whenever an RPS trip function is requir6d it is frequently backed up by other trip functions. The ESFAS actuation signals are used to actuate two independent ESF trains. Where it is practical, an ESFAS can be generated by more than one parameter. The Alternate Protection System augments reactor trip and emergency feedwater actuation by using separate and diverse D non-1E trip logic from that used by the Plant Protection System. i The design goals are accomplished without excessive complexity by using only four channels for each parameter. This allows for testing and maintenance of a channel without reducing the system to a single channel for trip, which would make the system susceptible to spurious trip or actuation signals. The protection systems are each functionally tested to ensure satisfactory operation prior to installation in the plant. Environmental and seismic qualifications are also performed utilizing type tests, specific equipment tests, appropriate analyses, or prior operating experience. For further information, refer to Sections 3.10 and 3.11. 3.1.19 CRITERION 23 - PROTECT. TON S'.8 TEM FAILURE MODES The protection system shall be designed to fail into a safe state or into a state demonstrated to be acceptable on some other defined basis if conditions such as disconnection of the syntem, loss of energy (e.g., electric powar, instrument air) or ; postulated adverse environments (e.g., extreme heat or cold, fire, pressure, steam, water, and radiation) are experienced. i ItESPONSE The Plant Protection System trip channels are designed to fail lD ' into a safe state or into a state established as acceptable in the event of loss of power supply. A failure is assumed to occur in only one channel (i.e., a single failure). This channel can be placed into bypass which places the RPS/ESFAS local D coincidence logic into a two-out-of-three configuration which ! Amendment. D 3.1-16 September 30, 1988
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CESSAR nai"lCATION O retains the coincidence of two for trip initiation. Refer to j Sections 7.2 and 7.3 for Failure Modes and Effects Analysis D ' information. I A loss of power to CEDM coils will cause the CEAs to insert into l the core. Redundancy, channel independence and separation are ) incorporated into the protection system design to minimize the J possibility of the loss of a protective function. The loss of I offsite power will cause the standby diesel electric generators i to start and energize the ESF trains which have actuation signals l present. \ l 3.1.20 CRITERION 24 - BEPARATION OF PROTECTION AND CONTROL BYBTEMB , f The protection system shall be separated from control systems to the extent that failure of any single control system component or channel, or failure or removal from service of any single J i protection system component or channel which is common to the l control and protection systems leaves intact a system satisfying all reliability, redundancy, and independence requirements of the , protection system. Interconnection of the protection and control l systems shall be limited so as to assure that safety is not significantly impaired. BESPOF8E: Protection system components and control system components are electrically and functionally isolated from each other. See Sections 7.2, 7.3 and 7.7.1.1.13 for details. The protection systems are designed so that they can sustain one channel in a tripped condition and one channel bypassed indefinitely and still provide their safety function. , Where control and protection systems have identical sensor input requirements, redundant Class 1E sensors that are used 2 independently by each channel of the protection system may also be used by the control system. For each sensed parameter, the control system monitors all four redundant instrument channels, which are interfaced to the control system via fiber-optic interfaces to ensure electrical independence. Within the control system, signal validation logic is used to detect bypassed or D failed sensors, thereby ensuring that they cause no erroneous control system actions. The control system signal validation I logic is described in Section 7.7.1.1.13. The design ensures that with a sensor or channel in bypass, another sensor can fail with no resulting control system action. Therefore, with one ,O channel in bypass, the protection system remains in an effective two-out-of-three configuration, meeting the required single failure criteria. Amendment D 3.1-17 September 30, 1988
CESSAR En&"ic 12u __ O l 3.1.21 CRITERION 25 - PROTECTION SYSTEM REQUIREMENTS FOR REACTIVITY CONTROL MALFUNCTIONS The protection system shall be designed to assure that specified acceptable fuel design limits are not exceeded for any single { malfunction of the reactivity control systems such as accidental withdrawal (not ejection or dropout) ci' control rods.
RESPONSE
Shutdown of the reactor is accomplished by the opening of the RTSS circuit breakers which interrupts power to the CEDM coils. Actuation of the circuit breakers is independent of any existing control signals. The protection systems are designed such that SAFDLs are not exceeded for any single malfunction of the reactivity control systems, including the withdrawal of a single full- or reactivity control 0 part strength CEA. Analyses of possible system malfunctions are discussed in Chapter 15. The various CEA related DBEs for which the protection systems are designed are discussed in Section 7.2. > 3.1.22 CRITERION 26 - REACTIVITY CONTROL SYSTEM REDUNDANCY AND CAPABILITY Two independent reactivity control systems of different design principles shall be provided. One of the systems shall use control rods, preferably including a positive means for inserting the rods, and shall be capable of : 11 ably controlling reactivity changes to assure that under cc- i.tions of norraal operation, including anticipated operatical 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 changcs resulting from planned, normal power changes (including xenon burnout) to assure acceptable fuel design limits are not exceeded. One of the systems shall be capable of holding the reactor core subcritical under cold conditions. D RESPCNSE: Two independent reactivity control systems of different design principles are provided. The first system, using Control Element Assemblies (CEAs), includes a positive means (gravity) for inserting CEAs and is capable of reliably controlling reactivity changes to assure that under conditions of normal operation, i including Anticipated Operational Occurrences, SAFDLs are not exceeded. The CEAs can be mechanically driven into the core. j Amendment D 3.1-18 September 30, 1988
l CESSAR inacma l ( 1 The appropriate margin for stuck rods is provided by assuming in ] the analyses of anticipated operational occurrences that the- l highest worth CEA does not fall into the core. 1 The second system, using neutron absorbing soluble boron, is capable of reliably compensating for the rate of reactivity changes resulting from planned normal power changes (including Xenon burnup) such that SAFDLs are not exceeded. This system is capable of holding the reactor subcritical under cold conditions. Either system is capable of making the core subcritical from a j hot operating condition and holding it - subcritical in the hot ' standby condition. i I Either system is able t< insert negative reactivity at a rate -1 sufficient to prevent exceeding SAFDLs as the result of a power J change (i.e., the positive reactivity added by Xenon burnup). 3.1.23 CRITERION 27 - COMBINED REACTIVITY CONTROL SYSTEMS CAPABILITY The reactivity control systems shall be designed to have a (~ f 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.
RESPONSE
Dissolved boron addition capability provided by the Safety Injection System (Chapter 6) in concert with the control rod (CEA) system will be such that under postulated accident conditions (Chapter 15), even with the CEA of highest worth stuck out of the core, adequate reactivity control is available to maintain short- and long-term capability to cool the core. 3.1.24 CRITERION 28 - REACTIVITY LIMITS The reactivity control systems shall be' designed with appropriate limits on the potential amount and rate of reactivity increase to assure that the effects of postulated reactivity accidents can neither (1) result in damage. to the reactor coolant pressure boundary greater than limited local yielding nor (2) sufficiently disturb the' core, .ts support structures or other reactor pressure vessel internals to impair significantly the capability to ecol the core. These postulated reactivity accidents shall include consideration of rod ejection (unless prevented by f, 3.1-19
CESSARn % mw O'l positive means), rod dropout, steam line rupture, changes in l reactor coolant temperature and pressure, and cold water addition.
RESPONSE
The bases for Control Element Assembly (CEA) design include ensuring that the reactivity worth of any one CEA is not greater than a preselected maximum value. The CEAs are divided into two sets, a shutdown set and a regulating set, further subdivided into groups as necessary. Administrative procedures and interlocks assure that only one group is withdrawn at a time, and that the regulating groups are withdrawn only after the shutdown groups are fully withdrawn. The regulating groups are programmed to move in sequence and within limits which prevent the rate of reactivity addition and the worth of individual CEAs from exceeding limiting values. The maximum rate of reactivity addition which may be produced by the Chemical and Volume Control System is too low to induce any significant pressure forces which might rupture the reactor coolant pressure boundary or disturb the reactor vessel internals. The reactor coolant pressure boundary (Chapter 5) and the reactor internals (Chapter 4) are designed to appropriate codes (refer for instance, to the response to Criterion 14) and will I accommodate the static and dynamic loads associated with an inadvertent, sudden release of energy, such as that resulting from a CEA ejection or steam line break (Chapter 15), without rupture and with limited deformation which will not impair the capability of cooling the core. 3.1.25 CRITERION 29 - PROTECTION AGAINST ANTICIPATED OPERATIONAL OCCURRENCES The protection and reactivity control systems she.1 : be designed to assure an extremely high probability of accomplishing their i safety functions in the event of anticipated operational ! occurrences.
RESPONSE
Plant events, designated in ANSI /ANS 51.1, " Nuclear Safety D Criteria for the Design of Stationary Pressurized Water Reactor , plants," have been carefully considered in the design of the ' protection and reactivity control systems. Consideration of redundancy, independence and testability in the design, coupled with careful component selection, overall system testing, and Amendment D 3.1-20 Suptember 30, 19 F 53
~ CESSAR EE'ancm. t \ adherence to detailed quality assurance . requirements, assure an , extremely high probability that safety functions are accomplished I in the event of Design Basis Events (DBEs). ] l Detailed discussions of the protection systems are provided in l Chapter 7. Design quality assurance is discussed in Chapter 17. The analysis of DBEs is contained in Chapter 15. 3.1.26 CRITERION 30 - QUALITY OF REACTOR COOLANT PRESSURE I BOUNDARY Components which are part of the reactor coolant pressure boundary shall be designed, fabricated, erected, and tested to the highest quality standards practical. Means shall be provided for detecting and, to the extent practicable, identifying the location of the source of reactor coolant leakage. . I
RESPONSE
The reactor coolant pressure boundary components are designed, fabricated, erected and tested in accordance with the ASME Code Section III. All components are classified Safety Class 1 or 2, in accordance with the ANSI /ANS 51.1, " Nuclear Safety Criteria D g' for the Design of Stationary PWR Plants," definitions for safety classes and the reactor coolant pressure boundary. Accordingly, they receive all of the quality measures appropriate to that classification. Means are provided for the identification of the source of reactor coolant leakage. These include the detection of leakage to systems connected to the reactor coolant pressure boundary as j well as leakage from the boundary into the containment. Instrumentation is provided to indicate and record makeup flow rate and integrated makeup flow to the primary water system. This instrumentation permits detection of suddenly occurring leaks and those which are gradually increasing. D 3.1.27 CRITERION 31 - FRACTURE PREVENTION OF REACTOR COOLANT PRESSURE BOUNDARY The reactor coolant pressure boundary shall be designed with sufficient margin to assure that when stressed under operating, maintenance, testing, and postulated accident conditions: (1) the boundary behaves in a nonbrittle manner; and, (2) The probability of rapidly propagating fracture is minimized. The design shall reflect consideration of service temperatures and other conditions of the boundary material under operating, maintenance, testing, and postulated accident conditions and the o Amendment D ; l 3.1-21 September 30, 1988
CESSARHKinc-O 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. BESPONDE: All the reactor coolant pressure boundary components are designed and constructed in accordance with ASME Section III and comply with the test and inspection requirements of these codes. The test and inspection requirements assure that flaw sizes are limited so that the probability of failure by rapid propagation is extremely remote. Particular emphasis is placed en the quality control applied to the reactor vessel on which teses and inspections exceeding ASME code requirements are performed. The tests and inspections performed on the reactor vessel are summarized in Section 5.2.4.1. Carbon and low alloy steel materials which form part of the pressure boundary are tested in accordance with the requirements of the fracture toughness requirements for materials, ASME Code Section III. Nonductile failure prevention will be ensured by utilizing the appropriate sections of the ASME Code. Excessive embrittlement of the reactor vessel material due to neutron radiation is prevented by providing an annulus of coolant water betwoon the reactor core and the vessel. In addition, to minimize the effects of irradiation on material toughness properties of core beltline materials, restrictions on upper limits for residual elements that directly influence the RT shift are required by the design specification. Specificalk D upper limits are placed on copper, nickel, phosphorous, sulfur, and vanadium. Further, the reactor vessel is forged such that no welds occur in the active core region. The maximum integrated fast neutron flux exposure of the reactor vogelnyt. wall opposite the midplane of the core is less than 6.0 x Thic value assumes a sixty-year vessel design life, D 10 with the plant at the design power level eighty percent of the time. The maximum expected increase in transition temperature is l about 140*F. The actual change in material toughness properties due to irradiation will be verified periodically during plant i lifetime by a material surveillance program. Based on an initial l RT of -20*F, operating restrictions will be applied aslD ncEEIsary to limit vessel stresses. The thermal stresses induced by the injection of cold water into the veccel, following a LOCA, have been examined. Analyses have D shown that there is no gross yielding across the vessel wall when using tht* minimum specified yio}d strength in the ASME Boiler and PraMre Vessel Code, Section III. l Amendment D 3.1-22 SeptenLor 30, 1988 l ____.___.__________w
~ CESSARn % mu (- ( ) G/ 3.1.28 CRITERION 32 - INSPECTION OF REACTOR COOLANT PRESBURE BOUNDARY Components which are part of the reactor coolant pressure i boundary shall be designed to permit: A. Periodic inspection and testing of important areas and features to assess their structural and leak-tight integrity; and B. An appropriate material surveillance program for the reactor pressure vessel. j l RESPONSEt j Provisions have been made in the design for inspection, testing, and surveillance of the Reactor Coolant System boundary as required by ASME Boiler and Pressure Vessel Code Section XI. The site operator is required to install the system so that the D required in-service inspections per Section XI can be performed. C-E recommends a reactor vessel surveillance program to the owner. ('N) y The reactor vessel surveillance program capability provided to v the site operator conforms with ASTM-E-185-73, " Practice for Conducting Surveillance Tests for Light Water Cooled Nuclear D Reactor Vessels," as revised in 1982. Sample pieces taken from the same material used in fabrication of the reactor vessel are installed between the core and the vessel inside wall. These samples will be removed and tested by the site operator at intervals during vessel life to provide an indication of the extent of the neutron embrittlement of the vessel wall. Charpy tests will be performed on the samples to develop a Charpy transition curve. By comparison of this curve with the Charpy curve and drop weight tests for specimens taken at the beginning of the vessel life, the change of RT will be determined and operating procedures adjusted as req E ed. See Chapter 5 for further details. The surveillance program capability provided to the site operator D has provisions which comply with the NRC regulation, " Reactor Vessel Material Surveillance Program Requirements," 10 CFR 50, Appendix H, published in the Federal Register in July 1983. The only exception between the recommended surveillance program and the requirements presented in Appendix H is the following: g ; i LI Amendment D 3.1-23 September 30, 1988 )
CESSAR Mie"icari:n O A. Appendix H, Section II.C.2 - Attachments to the reactor vessel. In adhering to the requirement of placing the surveillance specimens as close as possible to the reactor vessel wall, the capsule holders are attached to the cladding of the reactor vessel and are not major load-bearing components. By such placement, temperature, flux spectra, and fluence differences between the surveillance specimens and the reactor vessel are minimized, thereby permitting more accurate assessment of the changes in the reactor vessel properties. 3.1.29 CRITERION 33 - REACTOR COOLANT MAKEUP 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 from the reactor coolant pressure boundary and rupture of small piping or other small components which are part of the boundary. The system shall be designed to assure that for onsite electrical power system operation (assuming offsite power is not available) and for offsite electrical power system operation (assuming onsite power is not available) the system safety function can be accomplished using the piping, pumps and valves used to maintain coolant inventory during normal reactor operation.
RESPONSE
Reactor coolant makeup during normal operation is provided by the Chemical and Volume Control System (CVCS). The design incorporates a high degree of functional reliability by provision of redundant components and an alternate path for charging. The charging pumps can be powered from either onsite or offsite power sources, including the emergency diesel generators. The system is described in Section 9.3.4. The CVCS has the capability of replacingtheflowlosstothecontainmentduetoleaksinsmalllD reactor coolant lines such as instrument and sample lines. These lines have 7/32 inch diameter by 1 inch long flow restricting devices. The CVCS is not required to perform any safety related function, such as accident mitigation, or be required to perform a safe shutdown. This does not, however, compromise the " defense in D depth" provided by the system as the normal means of maintaining RCS inventory and primary water chemistry. In designing the CVCS as non-safety grade, the following safety functions are performed by dedicated safety systems. Boration and makeup for design Amendment D 3.1-24 September 30, 1988
~ CESSAR W h o O basis events will be provided by the Safety Injection System. Pressure control will be provided by the Safety Depressurization System. The Safety Injection System and the Safety Depressurization System are described in further detail in D Sections 6.3 and 6.7, resp $.ctively. All portions of the CVCS outside of containment have been designed as non-nuclear safety. However, portions of the CVCS which are inside the containment will retain their safety class designation to ensure the integrity of the reactor coolant pressure boundary. 3.1.30 CRITERION 34 - RESIDUAL HEAT REMOVAL A system to remove residual heat shall be provided. The system i safety function shall be to transfer fission product decay heat I and other residual heat from the reactor core at a rate such that specified acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary are not exceeded. Suitable redundancy in components and features, and suitable interconnections, leak detection and isolation capabilities shall be provided to assure that for onsite electrical power system operation (asnuming offsite power is not available) and for offsite electrical power system operation (assuming onsite power i' is not available) the system safety function can be accomplished, assuming a single failure.
RESPONSE
Residual heat removal capability is provided by the Shutdown Cooling System for reactor coolant temperatures less than 350*F. For temperatures greater than 350*F, this function is provided by the steam generators. The Emergency Feedwater (EFW) System provides a dedicated, independent, safety-related means of supplying secondary side, D quality feedwater to the steam generator (s) for removal of heat and prevention of reactor core uncovery. The design incorporates sufficient redundancy, interconnections, leak detection, and isolation capability to ensure that the residual heat removal function can be accomplished, assuming a single active failure. Within i appropriate design limits, either system will remove fission J product decay heat at a rate such that SAFDLs and the design conditions of the reactor coolant pressure boundary will not be ) exceeded. The Shutdown Cooling System and the steam generator auxiliaries are designed to operate either from offsite power or from onsite ) power sources.
]
} Further discussion is included in Section 5.4.7 for the Shutdown Cooling System and in Chapter 10 for the Steam and Power Conversion System.
Amendment D 3.1-25 September 30, 1988 j
l CESSAREnib - I l Ol 3.1.31 CRITERION 35 - EMER3ENCY CORE COOLING A system to provide abundant emergency core cooling shall be provided. The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with ( continued effective core cooling is prevented and (2) clad metal-water reaction is limited to negligible amounts. Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation and containment capabilities shall be provided to assure that for onsite electrical power system operation (assuming offsite power is not available) and for offsite electrical power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.
RESPONSE
Emergency core cooling is provided by the Safety Injection System (SIS) (described in Section 6.3). The system is designed to provide cooling water to remove heat at a rate sufficient to maintain the fuel in a coolable geometry and to assure that zirconium-water reaction is limited to a negligible amount (less than one percent). Detailed analysis has been performed, utilizing models complying with 10 CFR 50, Appendix K, "ECCS Evaluation Models," to verify that the system performance is adequate to meet the intent of the " Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors" of 10 CFR 50.46(b). The system design includes provisions to assure that the required safety functions are accomplished with either onsite or offsite electrical power, assuming a single failure of any component (qualified as described below). The single failure may be an D active failure
- during the short-term cooling phase of safety
- An active failure is a malfunction, excluding passive ,
failure, of a component which relies on mechanical movement to complete its intended function upon demand. Check valves which receive regular exercise to ensure operability are treated as passive coraponents. Examples of active failures include the failure of a valve to move to its correct position, or the failure of a pump, fan, or diesel generator , to start. Spurious action of a powered component j originating within the actuation system or its supporting j systems shall be regarded as an active failure, unless I specific design features or operating restrictions preclude ; such spurious action, j i Amendment D 3.1-26 September 30, 1988 1 _ _ - _ _ - _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ . _. b
.CESSAR MAWNumu injection or an active or limited leakage passive failure
- during the long-term cooling phase of safety injection.
Though the SIS is designed to accommodate a limited leakage passive failure during the long-term cooling phase, it does not accommodate arbitrary large leakage passive failures, such as the 9 complete . double-ended severance of piping, which are extremely low probability events. The site-spcM fic layout and arrangement . will be such that the limited leakage passive failure does non preclude minimum acceptable long-term cooling capability. Where building design is . not relied upon to mitigate and contain leakage from the SIS passive failure, suitable automatic isolation and auxiliary equipment must be provided by the site operator, as necessary. 3.1.32 CRITERION 36 - INSPECTION OF EhERGENCY CORE COOLING SYSTEM The emergency core cooling system shall be designed to permit appropriate periodic inspection of important components, such as spray rings in the reactor pressure vessel, water injection nozzles, and piping to ausure the integrity and capability of the system. E38PONSE: The SIS is designed to facilitate access to all critical D components. All pumps, heat exchangers, valves and piping external to the containment structure are readily accessible for periodic inspection to ensure system leak-tight integrity. Valves, piping and tanks inside the containment may be inspected for leak-tightness during plant shutdowns for refueling and maintenance. Reactor vessel internal structures, reactor coolant piping and water injection nozzles are designed to permit visual inspection for wear due to erosion, corrosion or vibration, and nondestructive inspection techniques where - these are applicable and desirable. ' Details of the inspection program are cescribed in Chapters 5, 6, and 16. l J
- A passive failure is defined as the blockage of a process !
flow path or a breach in the integrity of a component or piping (e.g., a piping-failure). Amendment D 3.1-27 September 30, 1988
CESSAR HEncanow O 3.1.33 CRITERION 37 - TESTII!G OF EMERGENCY CORE COOLING SYSTEM The emergency core cooling system shall be designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leak-tight integrity of its components, (2) the operability and performance of the active components of the system, and (3) the operability of the system as a whole and, under conditions as close to design as practical, the performance of the full operational sequence that brings the system into operation, including operation of the applicable portions of the protection system, the transfer between normal and emergency power sources, and the operation of the associated cooling water system. RESPONSES The SIS is provided with testing capability to demonstrate system D and component operability. Testing can be conducted during normal plant operation with the test facilities arranged not :.o interfere with the performance of the systems or with the initiation of control circuits, as described in Sc< tion 6. 3 and Chapter 14. 3.1.34 CRITERION 38 - CONTAINMENT HEAT REMOVAL A system to remove heat from the reactor containment shall be provided. The system function shall be to reduce rapidly, consistent with the functioning of other associated systems, the containment pressure and 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 electrical power system operation (assuming offsite power is not available) and for offsite electrical power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.
RESPONSE
The Containment Spray System consists of two completely independent subsystems. The heat removal capacity of the flow from either containment spray subsystem is adequate to keep the D 4 containment pressure and temperature below design conditions for I any size break in the RCS piping up to and including a double-ended break of the largest reactor coolant pipe, with an j unobstructed discharge from both ends. l l Amendment D l 3.1-28 September 30, 1988 O
CESSAR ETnc-. r~N l (v) i l Borated water is sprayed downward by the system from the upper I regions of the containment to cool the atmosphere. Cooling reduces the containment pressure and temperature following a major loss-of-coolant-accident. D l Suitable redundancy in components and features is designed into the Containment spray System to maintain the pressure and temperature conditions below containment design even in the event ) of a single failure, including the loss of onsite or offsite / clectrical power. 3.1.35 CRITERION 39 - INSPECTION OF CONTAINMENT HEAT REMOVAL SYGTEM The containment heat removal system shall be designed to permit appropriate periodic inspection of important components, such as the torus, sumps, spray nozzles, and piping to assure the integrity and capability of the system. RESPONSE 1 All essential equipment of the Containment Spray System is jN located outside the containment, except for spray headers, t, v
) nozzles, containment sump, In-containment Refueling Water Storage
'Jank and associated piping. These components include two j containment spray pumps, two shutdown cooling heat exchangers and 0 4 independent containment spray headers.
The detailed arrangement and layout of system piping, pumps, heat i exchangers, and valves will provide the separation, availability, and accessibility required for periodic inspection. Nozzle inspection capability will be provided as well. 3.1.36 CRITERION 40 - TESTING OF CONTAINMENT HEAT REMOVAL l BYSTEM j ( The containment heat removal system shall be designed to permit j appropriate periodic pressure and functional testing to assure ] (1) the structural and leak-tight integrity of its components, I (2) the operability and performance of the active components of I the system, and (3) the operability of the system as a whole, and, under conditions as close to the design as practical, the 1 performance of the full operational sequence that brings the l system into operation, including operation of applicable portions of the protection system, the transfer between normel and emergency power sources, and the operation of the associated cooling water system. r'% j
! ) l 'Q l i
Amendment D 3.1-29 September 30, 1988 i (
CESSAR Un% mon O
}lEEPON3E t System piping, valves, pumps, heat exchangers, and other compcnents of the containment heat remov31 system are arranged so that each component can be tested periodically for operability.
Teating can be conducted during normal plant operation with the test facilities arranged not to interfere with the performance of the system or with the initiation of control circuits, as described in Section 6.2. The performance testing of containment spray pumps is conducted at some time other than refueling. The pumps are aligned to take suction from and return flow to the In-containment Refueling Fater Storage Tank (IRWST). Flcw and head are recorded by the installed instrumentation. O i Heat exchanger operation may be verified during any operating mode by circulating water through the containment spray heat exchanger and back to the IRWST. Actuator-operated valves can be cycled from the control room, and operation verified by observing control room indication. Check valves will be tested to ensure that the valves operate properly. These valves include the IRWST check valves and the valves on the inlets and outlets of the containment spray pumps. 3.1.37 CRITERION 41 - CONTAINIdENT ATMOSPHERE CLEANUP Systems to control fission products, hydrogen, oxygen, and other substances which may be released into the reactor containment shall be provided as necessary to reduce, consistent with the I functioning of other associated systems, the concentration and quantity of fission products released to the er.vironment following postulated accidents, and to control the concentration cf hydrogen or oxygen and other substances in the containment atmosphere following postulated accidents to assure that l containment integrity is maintained. { Each system shall have suitable redundancy in compononts and features, and suitable interconnections, leak detection, isolation, and containment capabilities to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) its safety function can be accomplished, assuming a single failure. Amendment D 3.1-30 September 30, 1988
CESSAR !ahm c9) RESPONSE 2 U (IATER) l 3.1.34 CRITERION 42 - INSPECTION OF CONTAINMENT ATNOSPEERE' l CLEANUP SYSTEMS . The containment. atmosphere cleanup systems shall be -designed to I permit appropriate periodic inspection of imporetant components, such as filter frames, ducts, and piping.to assure the integrity and capability of the systems. MSB2MMt )- (LATER) 3.1.39 CRITERION 43 - TESTING OF CONTAINMENT ATMOSPHERE CLKANUP j The containment atmosphere cleanup systems shall be designed to - l permit appropriate periodic pressure and functional testing to d unsure (1) tt2e structural and leak-tight integrity of' its p
- components, (2) the operability and performance of the active ,
- t. components of the systems such as fans,. filters, ditmpers, pumps, !
'd and valves and (2) the operability.of the. systems as a whole and, j under conditions as close'to design as practical, the performance J of the full operational sequence that brings the ayatoms into operation, including operation of. applicabl'e' portions of the )
protection system, the - transfer between normal and emergency power sources, and the operation of' associated systems. I i MED2EHR (LATER) 3.1.40 CRITERION 44 - COOLIMO WATER 1 1 A system to transfer heat from structures, systems, and componento important to safety, to an ultimate heat-sink'shall be i provided. The system safety function shall be to transfer the j combined heat load of these structures, systems, and components under normal operating and accident conditions. Sultable redundancy in components and features, and suitable interconnections, leak detection, and isolation' capabilities shall be provided to assure that for onsite electrical power system operation (assuming of fsite power is not available) and for offsite electrical power system operation (assuming onsite power is not available) the system safety function can be 5 accomplished, assuming a single failure. Amendment D 3.1-31 September 30, 1988
~
CESSAR EnLra _ O P&D.29EDRt The cooling water systeme which function to renove the combined heat load from structures, systams, and components important to safety under normal operating and' accident. conditions are the Component Cooling Water System and the Station Service Water D System. The Component Cooling Water System is a closed loop { system _ which removec haut from nuclear safety related and potentially radioactive systems. The Station Service Water System removes heat from ther Component cooling Water System and e transfers it to the atmosphere through cooling ponds. The Station Servies Water S,ystem is described in Section 9.2.1 and f the Component Cooling Water System is described in Section 9.2.2. 3.1.41 CRITERION 45 ~ INSPECTION OF COOLING WATER BYSTEM The cooling water system shall be designed to permit appropriate periodic luspection of important components, such as heat exchangers and piping, to assure the integrity and capability of the system. EfdE9ESEI The important components are located in accessible areas. These components have suitable manholes, handholes, inspection ports, D or other appropriate design and layout features to allow periodic inspection. See Sections 9.2.1 and 9.2.2 for details. ) 3.1.42 CRITERION 46 - TESTING OF COOLING WATER BYSTEM The cooling water system shall be designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leak-tight integrity of its components, (2) the operability and the performance of the active components of the system, and (3) the operability of the system as a whole, and under conditions as close to design as practical, the performance of the full operational sequence that brings the system into operation for reactor shutdown and for loss-of-coolant-accidents, including operation of applicable portions of the protection system and the transfer between norral and emergency power sourcee;. RESPONSES The design provides for periodic testing of active components of the cooling water systems for operability and functional D performance. ;
)
@\
Amendment D 3.1-32 September 30, 1988
CESSAR unhion O i l Preoperational performance tests of the components are made by J the inanufacturer. An initial system flow test demonstrates j proper functioning of the system. Thereafter, periodic tests ) ensure that components are functioning properly. Cooling water system valves may be connected to the preferred power source at any time during reactor operation to de.monstrate operability. Many active components are operated normally, thereby demonstrating operability. Remotely operated valves are exercised and actuation circuits tested. Tne automatic actuation D ) J circuitry, valves, and pump . breakers _ also may be checked when integrated system tests are performed during a planned cooldown of the Reactor Coolant System. Provisions have been made to l permit periodic leakage tests to verify the continued leak-tight i j integrity of the systems. Refer to Sections 9.2.1 and 9.2.2 for j details. j l 3.1.43 CRITERION 50 - CONTAINMENT DESIGN BASID l l The reactor containment structure, including access openings, l penetrations, and the containment heat removal system shall be 1 designed so that the corttainment structure and. its internal : compartments can accomodate, { without exceeding the design leakage rate and, with sufficient margin, the calculated pressure
^
and temperature conditions resulting from any LOCA. This margin shall reflect consideration of (1) the effects of rotential energy sources which have not been included in the determination of the peak conditions, such as energy in steam generators 'and energy from metal-water and other chemical reactions that may result from degradation but not total failure of emergency core cooling functioning, (2) the limiced experience and experimental data available for defining accident phenomena and containment responses, and (3) the conservatism of- the calculational model j and input parameters. EMPM!BS The containment structure, including access openings- and penetrations, is designed to accommodate, without exceeding : the design leak rate, the transient peak pressure and temperature associated with a LOCA up to and including a double-ended rupture of the larepst reactor coolant pipe. D l The containment structurs and Engineered Safety Feature systems have been evaluated for various combinations, of energy. release. . Tne analysis accounts for system thermal and chemical energy, and I for nuclear decav heat. The Safety Injection System is designed ; such that no single active failure could result in significant q metal-water reaction (see Section 6.2.1). l 1 Amendment D 3.1-33 September 30, 1988 j
CESSAR nMncwon O 3.1.44 CRITERION S1 - FRACTURE PREVENTION OF CONTAINMENT PRESBURE BOUNDARY The reactor containment boundary shall be designed with aufficient margin to assure that under operating, maintenance, testing, and postulated accident conditions (1) its ferritic materials behave in a nonbrittle manner, ar.d (2) the probability of ropidly propagating fracture is mini 31ced. The design shall reflect consideration of service temperatures and other conditions of the containment boundary material during operation, maintenance, t<wting and postulated accident conditions, at4d the uncertainties in determining (1) material preperties (2) residual, steady-state, and transient stresses, and (3) size of flaws. RMf9El%: The material se?.ected for the containment vessel is UAr' con steel l normalized to refine the grain which results in improved l ductility. In addition, the actual mechanical and chemical propertjes of the material are documented and are within the D limito of minimum ductility defined in (LATER). The containment vessel is built to Subsection NE of the (LATER) l edition of Section III of the ASME Boiler and Pressure Vessel Code. l The design of the vessel reflects consideration of all ranges of temperature and loading conditions which apply to the vessel during oneration; maintenance, testing and postulated accident I conditions. All scam welds in the vessel are 100 percent radiographer, and I the acceptance standards of the radiographs ensure that flaws in welds do not exceed the maximum allowed by the ASME Code.. Steady state and transient strcsses are calculated in accordance with accepted metnods (see Section 3.8). 3.1.45 CRITERION 52 - CAPADILITY FOR CONTAINMENT LEAKAGE RATE TESTING 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 D containment design pressure. O Amendment D 3.1-34 September 30, 1988
C E S S A R 8n!?N.u m (Dq
\s. ;! !
211DEMK: The containment vessel is designed so that integrated leak rate D testing can be performed at design pressure after completion and I I installation of penetrations and equipment in accordence with the requirement of Appendix J of 10 CFR 50 (see Section 6.2.6). 3.1.46 CRITENION 53 - PROVISTONS FOR CONTAINMENT TESTING AND INSPECTION The reactor containment shall be designed to peruit (1) l ' l appropriate periodic inspection of all important areas, such as penetrations, (2) an appropriate surveillance program, and (3) , periodic teuting at containment design pressure of the ' leak-tightness of penetrations which have resilient seals and expansion bellows. MOREDA: , O I The absence of insulation on the containment vessel permits ! periodic inspection of the exposed surfaces of the vessel. The Ivwer portions of the contaiorent vessel are totally encased in , concrete and will not be accessible for inspection. It la j f3 no need any l,V) contemplated that there will be for in-service surveillance program due to tha rigorous design, special i ( fabrication, inspection and pressure testing the containment J vessel receives prior to operation, i Provisions are made to permit. periodic testing at containment ! design pressure o,f penetrati'ons which have resilient seals or expansion bellows to allow leak-tightness to be demonstrated (refer to Section 6.2.6).. I 1 3.1.47 CRITERION 54 - PIPING GYSTEMS PENETRATING CONTAINMENT 4 Piping systems penetrating primary reactor containment shall be ! provided with leak detection, isolation,- and containment j capabilities having redundancy, reliability, and porform&nce capabilities which reflect the importanca to safety of isolating these piping systems. Such piping systeme shall be designed with a capability to test periodically the operability of the isola-tion valves end associated appri.tatus and to determine if valve ! leakage in within acceptable limits. 1 Ef&P91!Et l g Piping systens described in CESffAR which penetrate containment designed to provide the required isolation and tenting b)
! are I
Amendment D ! 3.1-35 September 30, 1988 1 \ .__. . _ _ _ _ _ _ _ _ - - _ _ -
1 CESSAREinh a I
._. = _ )
capabilities. These piping systems are connections to allow periodic leak detection tests pr,oviGed with test to be ei , performed, in accordance with 10 CFR 50, A,Dpendix J. j The Engineered Safety Features Actuation System circuitry provides the means for testing isolation valve operability. I 0 For a discussion of penetration design, refer to Section 6.2.4, ! Containment Isolation Sy6 tem. For additional related discussion, sots the responses to General i Design Criteria 55, 55, and 57 (Sections 3.1.48 through 3.1.50), 3.1.48 CRITERION 55 - REACTOR COOLANT PRESSURE BOUNDARY PENETRATING CONTAINMENT Each line that is part of the reactor coolant pressure boundary and that penetrates primary reactor containment shall be provided wi th 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 acceptabic . on some other defined basis: l A. One locked closed isolation valve incide and one locked closed isolation valve outside containment; or B. One automatic isolation valve inside and one locked closed isolation valve outside containment; or C. One locked closed isolation valve inside and one automatit: isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outcide containment; or l i D. One autonatic isolation valve inside and one automatic isolati:on valve outside containment. A simple check valve 3 may not be used as the automatic isolation valve outside l containment. Isolation valves outside containment shall be located as close to l containment as practical and upon loss of actuating power, I automatic inlation valves shall be designed to take the position i 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 shall be provided an necesstry to assure s adequate safety. De~ termination of the appropriateness of these requirements, such as higher quality in design, fabrication, and I Amendment D I ) 3.1-36 September 30, 1988 j ) L----
CESSAR8HWace j (~ ; V l testi ng, additional provin tons for in-service inspection, : protection against more severe natural phenomena, and additional isolatien valves and containment, shall include consideration of ) the populati1n density and physical characteristics of the site j environo, j i hESPONGE , 1 ! The reactor coolant system pressure boundary for CESSAR is j I lefined in accordance with ANSI /ANS 51.1 arid 10 CFR 50, Section D j 50.2(v). All reactor coolant pressure boundary lines penetrating { containment meet the isolation criteria of GDC 55 psing the j following basis for specific lines in addition to those noted ! above. I A. Safety injection lines, as shown on Figure 6.3.2-1A, are D used to mitigate the consequences of accidents an.1 therefore do not receive an automatic closure signal and are not I locked closed. l B. When in the shutdown cooling mode of operation the Shutdown 1 Cooling System is an extension of the rocctor coolant l pressure boundary. In this mode the systent is isolated from . (v ) the environment by two isolation valves in series. ] C. The charging and seal injection lines shownonFigure9.3-4l'D have automatic valves out's ide the containtient which do not l receive a closure signal (CIAS). This is because it is j desirable to maintain charging and seal injection flow as long as the charging pumps are in operation. l 1 3.1.49 CRITERLON 56 - PRIMARY CONTAINME11T ISOLATION Each line that connects directly to the containment atmospnere and penetrates primary reactor containment shall be provided with I -containment isolatien valves as follows, unicas it can be demonstrated that the containment isolation provisions for a . specific class of lines, such as instrument.11nec, are acceptable ' on some other defined basis: A. One locked closed isolation valve inside and one lockm* closed isolation valve outside containment; or B. One automatic. i ation valve inside and one locked closed f isolation valve , Aside containment; or ( Cm One locked closed isolation valve inside and one autstatic isolation valve outaide containment. A simple check valve h[ may not be used as tne automatic isolation valve outside containment, or Amendment D 3 .1 *T, Geptember 30, 19 %
CESSAR EnWication Ol D. 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 I the containment as practical and upon loss of actuating power, j automatic isolation valves shall be designed to take the position j that provides greater sefety. ) i FE8P_QEyJ: I CESSAR fluid systems Comply with the requirements of GDC 56 with the following exceptions: Lines which connect directly to the containment atmosphere and are used for mitigating the effects of accidents are connected to . a closed piping system outside containment, which is isolated I from the environment in accordance with the requirements of GDC
- 55. In addition, the capability for remote double isolation at the containment bcundary is provided in accordance with GDC 56. 1 3.1.50 CRITERION 57 - CLOSED BrBTEM ISOLATION VALVEB 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 ured as the automatic isolation valve.
MDlQ1LS3t The systems that fall into the category described in GDC 57 0 comply with containment isolation requirements as specified in the containment isolation system sections of CESSAR. 3.1.51 CRITERION 60 - CONTROL OF RELEADES OF RADIOACTIVE MATERIAL TO THE ENVIRONMENT The nuclear power unit design shall include means to control suitably the release of radioactive materials in gaseous and liquid effluents and to handle radioactive solid wastes produced during normal reactor operation, including anticipated operational occurrences. Sufficient holdup capacity shall be provided for retention of gaseous and liquid effluents containing 4 radioactive materials, particularly where unfavorable site ' Amendment D 3.1.-38 September 30, 1988
. i
c~ - - l 1 CESSAR !!Nncmou l I i i environmental conditions can be expected to impose unusual operational limitations upon the release of such effluents to the l environment ' RED 9MR The sources and expected quantities of radioactive materials i preduced dur.ing normal reactor operation, including anticipated I operational occurrences, is presented in Chapter 11. The radioactive waste systems to suitably _ control the release of D these materials in gaseous and liquid ef fluents . and to handle ; radioactive solid wastes are described in Section 11.9. i l I 3.1.52 CRITERION 61 - FUEL 8TORhGE AND HANDLING AND ! RADIOACTIVITY CONTROL l l Tne 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 l D for radiation protection, (3) with appropiinte containment, I t confinement, and filtering systems, (4) with a residual heat removal capability having rollability and testability that reflects the itoportance to safety of' decay heat and other residual heat removnl, and (5) to prevent significant reduction lj in fuel storage coolant inventory under accident conditions. BESPONSE: D (LATER) j { 3.1.53 CRITERION 62 - PREVENTION OF CRITICALITY IN FDEL f BTORAGE AND HANDLING l Criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of 1 geometrically safe configurations. ) l RESPONSE: )
.D 1 (LATER) l 3.1 54 CRITERION 63 - MONITORING FUL'L AND WASTE STORAGE Appropriate systems shall be provided in fuel stcrage and l radioactive waste systems and associated handling areas (1) to Amendment D 3.1-39 September 30, 1988 l
L ,
CESSARn!L mu O detect conditions that may result in loss of residual heat removal capability and excessive radiation levels and (2) to initiate appropriate safety actions. R10PONSE 0 (LATER) 3,1.55 CRITERION 64 - MONITORING RADIOACTIVITY RELEASES Means shall be provided for monitoring the reactor containment atmosphere, spaces containing components for recirculation of loss-of-coolant-accident fluids, effluent discharge paths, and the plant environs for radioactivity that may be released from normal operations, including anticipated operational occurrences, and from postulated accidents. ErdE9FJJ: (LATER) 0 O
)
I Amendment D 3.1-40 September 30, 1988 l
CESSAR !!n%uin 1 3.2 CLASSIFICATION OF FTRUCTURES, C,OMPONENTS, AND SYSTEMS 3.2.1 SEIBMIC CLASSIFICATION Structures, systems, electrical equipment and components which I are important to safety and designed to remain functional in the event of a Safe Shutdown Earthquake (SSE) are designated as Seismic Category I. , l Seismic Category I structures, systems, and components are those necessary to ensure: l A. The integrity of the reactor coolant pressure boundary. B. The capability to achieve safe shutdown of the reactor and keep it in a safe shutdown condition. C. The capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures in excess of 10 CFR 100 guidelines. The selection cf Category I structures, systems, and components lp is in accordance with the definition above and the guidance provided by Regulatory Guide 1.29. Individual components in 'G ! Category I systems are classified by reference to the safety I classes assigned in accordance with ANSI /ANS 51.1 (see Section l 3.2.2). All components in Safety Classes 1, 2, and3areSeismiclD i Category I. Structures, systems and components which do not perform a nuclear safety related function and whose continued function is not required are classified Non-Nuclear Safety (NNS) (see Section NNS structures, systems and components whose structural D 3.2.2). failure or interaction could degrade the functioning of a seismic Category I structure, system, or component to an unacceptable safety level or could result in an incapacitating injury to an j occupant of the control room are designated as Seismic Category l II and are designed and constructed so that the SSE will not I cause such failure in a manner that would adversely affect a safety system. Structures, systems, and equipment which have no enhanced seismic design requirements in addition to those imposed by building codes are designated Non-Seismic (NS). N Amendment D 3.2-1 September 30, 1988 1
l CESSAREn h ou 1 l l O The seismic category and safety and quality classification of j mechanicalcomponentswithintheSystem80+NuclearPowerModulel0 scope are listed in Table 3.2-1 and on the P& ids (Chapters 5, 6, l and 9). Seismic Category I includes all mechanical components within the safety class boundaries and extende to the first seismic restraint beyond the boundary. Structures or supports essential to the performance of a safety function by a mechanical component or capable of disabling interaction with it are D designed to Seismic Category II requirements for structural! l integrity only. This ensures that any structures, systems, or j components that could potentially have a disabling interaction lE ' with Seismic Category I mechanical structures, systems, or components are either prevented from doing so or are designed to O meet Seismic Category I or II structural integrity requirements, depending on the function of the component. The listing of major electrical components is found in Section 3.11, which also includes safety and quality classifications. Electrical structures, systems, and components not classified as Seismic Category I but whose failure could represent a hazard to the operator or could interfere with the performance of required safety functions of electrical structures, systems and components, are classified as Seismic Category II (Reference 1). Any electrical system or structure or component not in seismic Category I or II is considered Non-Seismic (see Section 3.10). D The use of the Seismic Category II designation for electrical components is limited to non-safety control system components which are designed and documented to maintain structural integrity during an SSE. For purposes of this discussion, the motors and solenoids used to provide notive power to mechanical components are treated as part of the mechanical component. 3.2.2 SYSTEM QUALITY GROUP CLASSIFICATIONS (SAFETY CLASS) In general, fluid system components important to safety are classified in accordance with ANSI /ANS 51.1 (Reference 2). For D purposes of CESSAR, Safety Class 1, 2, 3 and NNS of ANSI /ANS 51.1 are equivalent to Quality Groups A, B, C and D of Regulatory Guide 1.26. The criteria establish safety classes which are used as a guide to the selection of codes, standards, and quality assurance provisions for the design and construction of the components. The safety class designations are also used as a guide to those fluid system components to be classified as O Seismic Category I and II (see Section 3. 2.1) . The Safety Class definitions in ANSI /ANS 51.1 are summarized as follows: O Amendment E 3.2-2 December 30, 1988
~ CESSAR8Ennc-
!\
A. Safety Class 1 (SC-1) applies to pressure-retaining portions and supports of mechanical equipment that form part of the RCPB whose failure could cause a loss of reactor coolant in excess of the reactor coolant normal makeup capability and whose requirements are within the scope of the ASME Boiler and Pressure Vessel Code, Section III. B. Safety Class 2 (SC-2) applies to pressure
- retaining portions and supports of primary containment and other mechanical equipment, requirements for which are within the scope of the ASME Boiler and Pressure Vessel Code, Section III, that D are not included in SC-1 and are designed and relied upon to accomplish the nuclear safety functions defined in ANSI /ANS 51.1, Section 3.3.1.2.
C. Safety Class 3 (SC-3) applies to equipment, not included in SC-1 or -2, that is designed and relied upon to accomplish the nuclear safety functions defined in ANSI /ANS 51.1, Section 3.3.1.3. D. Non-Nuclear Safety (NNS) applies to equipment that is not in Safety Class 1, 2, or 3. This equipment is not relied upon p to perform a nuclear safety function. The safety classifications of major components which are in the System 80+ design scope are listed in Table 3.2-1 and Section D 3.11. Seismic category designations and quality assurance requirements are also included. Small components, such as piping and strainers, are not listed; they may be found by reference to the P& ids (Chapters 5, 6, and 9) where the exact boundaries are indicated. All pressure containing components in Safety Classes 1, 2, and 3 are designed, manufactured, and tested in accordance with the rules of the ASME Boiler and Pressure. Vessel Code, Snction III. Components designated NNS are designed and constructed with appropriate consideration of the intended service usinglD applicable industry codes and standards. The relationship between safety class and code class is shown in Table 3.2-2. A higher code class may be used for a component without changing the safety class or affecting the balance of the system in which it is located. Fracture toughness requirements are imposed on materials for pressure retaining parts of ASME Class 2 and 3 System 80+ Nuclear D Power Module components. Test methods, acceptance, and exemption criteria are in conformance with the ASME Code, Section III. l The safety classification system is also used to identify those ; components to which the requirements of 10 CFR 50, Appendix B,
]
Amendment D 3.2-3 September 30, 1988 j
CESSAREnMc-O are applicable. Components in Safety Classes 1, 2, and 3* are D designed and manufactured under a rigorous quality assurance program reflecting the requirements of Appendix B, and are designated Quality Class 1. The Quality Class 1 qualjty assurance program is described in Chapter 17. Components which do not serve a safety related function are designated Quality Class 2. Quality Class 2 components will be designed and manufactured or procured in accordance with the pertinent D requirements of the Quality Assurance Program as given in Chapter 17. The quality class designations of major mechanical and electrical components are shown in Table 3.2-1 and Section 3.11, respectively, in conjunction with the safety and seismic classifications. The use of the above outlined safety and quality classification systems meets the intent of Regulatory Guide 1.26 and the requirements of 10 CFR 50.55a. O
- With the following exception: the CVCS gas stripper is Safety Class 3, Quality Class 2, however, pressure retaining D
portions meet rules applicable to ASME Code Class 3 components. See Table 3.2-1. l l 4 l I Amendment D j 3.2-4 September 30, 1988 ! l l
CESSAR !!!Mncuia i
~-wt J
O u REFEREFCES. FOR SECTIOW_JJ i i
- 1. " Seismic Qualification of C-E Instrumentation Equipment," I Combustion Engineering, Inc., CENPD-182, Revision 1, May 1977.
D i
- 2. " Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants," ANSI /ANS 51.1, 1983.
1 1 l 1 f . t
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\ l Amendment D I 3.2-5 September 30, 1988
-v i CESSAR En=cma J l O l TABLE 3.2-1 (Sheet 1 of 6) 1 CLASSIFICATION OF STRUCTURES. SYSTEMS. AND COMPONENTS Safety Seismic Quality Comconent Identification Class Cateaory Class Reactor Coolant System
- Reactor Vessel 1 I 1
- Steam Generators (primary / secondary) 1/2 (1) I 1 ;
- Pressurizer 1 I 1 1 l
- Reactor Coolant Pumps (2) (3) (9) 1 I 1 Piping within Reactor Coolant Pressure Boundary (S) 1/2 (4) I 1 Control Element Drive Mechanisms (6) (6) I l Core Support Structures (7) 3 I 1 Fuel Assemblies (8) 2 I 1 Control Element Assemblies (8) 3 I 1 jq Closure Head Lift Rig NNS II (10) 2 Q Heated Junction Thermocouple Probe Assembly 1/3 (12) I 1 9
)
HJTC Pressure Housing 1 I 1 ; J Safety Injection System l
- Safety Injection Pumps 2 I 1
)
- Shutdown Cooling Heat Exchangers 2/3 (1) I 1
- Safety Injection Tanks 2 I 1
- Shutdown Cooling Pumps 2 I i
- Containment Spray Pumps 2 I 1 !'
- Containment Spray Heat Exchangers 2/3 (1) I 1
- IRWST 2 I 1 ;
- Shutdown Cooling Mini-flow Heat 2/3 (1) I 1 Exchanger D l
- Containment Spray Mini-flow Heat 2/3 (1) I 1 Exchanger Chemical and Volume Control System
- Regenerative Heat Exchanger NNS NS 2 '!
- Letdown Heat Exchanger NNS NS 2
- Seal Injection Heat Exchanger NNS NS 7.
- Purification lon Exchangers NNS NS 2
( Footnotes to this table are given at the end of the table.
- Including component supports down to (but not including) embedments.
Awndment D I September-30, 1988
CESSAR nai"icmou 9 J_ABLE 3.2-1 (Cont'd) (Sheet 2 of 6) CLASS 1FICATION OF SlMpTURES. SYSTEMS. AND COMPONENTS Safety Seismic Quality Comppnent Identification Class Category Class Chemical and Volume Control System (Cont'd) i
- Deborating lon Exchanger NNS NS 2
- Volume Control Tank NNS NS 2
- Chemical Addition Package NNS NS 2
- Boric Acid Batching Tank NNS NS 2
- Charging Pumps NNS NS 2
- Boric Acid Makeup Pumps NNS NS 2
- Reactor Makeup Water Pumps NNS NS 2 D
- Boric Acid Concentrator NNS NS 2
- Pre-holdup lon Exchanger NNS NS 2
- Mini-flow Heat Exchanger NNS NS 2
- Boric Acid Condensate Ion Exchanger NNS NS 2 .
- Reactor Drain Pumps NNS NS 2
- Holdup Pumps NNS NS 2
- Reactor Drain Tank NNS NS 2 lE
- Holdup Tank NNS NS 2
- Equipment Drain Tank NNS NS 2 {
- Reactor Makeup Water Tank NHS NS 2
- Gas Stripper NNS NS 2
- Purification Filters NNS NS 2
- Reactor Drain Filter NNS NS 2 0
- Seal Injection Filttes NNS NS 2 ,
- Reactor Makeup Filter NNS NS 2 l
- Boric Acid Filter NNS NS 2
- Letdown Strainer NNS NS 2
- Pre-holdup Strainer NNS NS 2 I
- Boric Acid Condensate Ion NNS NS 2 Exchanger Strainer
- lon Exchanger Drain Header Strainer NNS NS 2
- Boric Acid Batching Strainer NHS NS 2
- Chemical Addition Strainer NNS NS 2
- Boric Acid Storage Tank NNS NS 2
- Boric Acid Batching Pump NNS NS 2 Safety Depressurization System (SDS)
(LATER) O I Amendment E December 30, 1988 i
CESSAR innncu,= n b TABLE 3,2-1 (Cont'd) (Sheet 3 of 6) l CLASSIFICATION OF STRUCTURES. SYSTEMS. AND COMPONENTS Safety Seismic Quality l Qomponent m Identification Class Cateaory C' ass Emergency Feedwater System
- Motor-Driven Emergency Feedwater Pump 1 3 I 1 l
- Motor-Driven Emergency Feedwater Pump 2 3 I I j
- Steam-Driven Emergency Feedwater Pump 1 3 1 1 3
- Steam-Driven Emergency Feedwater Pump 2 3 I 1
- Emergency Feedwater Pump Turbine 1 3 I 1 1
- Emergency Feedwater Pump Turbine 2 3 I 1
- Emergency Feedwater Storage Tank 1 3 I 1 !
- Emergency Feedwater Storage Tank 2 3 I 1 i Fuel Handling System )
Refueling Machine NNS 11 2 Fuel Transfer System NNS 11 2 D
- 1. Transfer Carriage NNS II 2 ;
- 2. Upcnding Machine NNS II 2
- 3. Hydraulic Power Unit NNS II 2 Fuel Transfer Tube, Valve, Stand NNS 11 2 CEA Change Platform NNS II 2 Long and Short Fuel Handling Tools NNS NS 2 Upper Guide Structure Lifting Rig NNS II (11) 2 Core Barrel Lifting Rig NNS II (11) 2 Spent Fuel Handling Machine NNS 11 2 New Fuel Elevator NNS 11 2 Underwater Television NNS NS 2 Refueling Pool Seal NNS NS 2 In-Core Instrumentation and CEA Cutter NNS NS 2 Extension Shaft Uncoupling Tool NNS NS 2 Fuel Transfer Tube Blind Flange 2 I 1 CEA Handling Tools NNS NS 2 ICI Cable Tray Support Frame 3 I 1 ICI Holding Frame NNS NS 2 ICI Guide Tubes 1 I 1 ICI Guide Tube Supports 1 1 1 101 Insertion and Removal Tools NNS NS 2 ILI Seal Housing 1' I 1
, ICI Seal Table 1 I 1 Spent Fuel Racks NNS I 2
% New Fuel Racks NNS I 2 E Amendment E December 30, 1988
~
\
i CESSAR E!!'Jricui, j i O IABLE 3.2-1 (Cont'd) (Sheet 4 of 6) I CLASSIFICATION OF STRUCTURES. SYSTE_MS. AND COMPONENTS { Safety Seismic Quality Component Identification Class Category Class , i
)
Component Coeling Water System l (LATER) l Spent Fuel Pool Cooling and Cleanup System (LATER) Sampling System D (LATER) 1 Station Service Water System (safety-related portion only) l (LATER) i Equipment l (LATER) Structures 1 (LATER) O Amendment D September 30, 1988
CESSAR nMacma f (g I IABLE 3.2-1 (Cont'd) (Sheet 5 of 6) l CLASSIFICATION OF STRUCTURES. SYSTEMS. AM CONPONENTS NOTES: (1) Two safety classes are used for heat exchangers to distinguish primary and secondary sides where they are different. (2) Loss of cooling water and/or seal water service to the reactor coolant pumps (RCPs) may require stopping the pumps. However, the continuous operation of the pumps is not-required during or following an SSE. The auxiliaries are therefore not necessarily Safety Class 3 or Seismic Category I. Provision for cooling water to the pump bearing oil cooler and pump motor air cooler will not comply with the requirements of Regulatory Guide 1.29 (see Subsection 5.4.1.3). (3) Only those structural portions of the RCPs which are necessary to assure the integrity of the reactor coolant pressure boundary are Safety Class 1. (4) Safety class of piping within the reactor coolant pressure boundary (as defined in 10 CFR 50) is selected in accordance with the ANSI /ANS 51.1 criteria identified in Section 3.2.2. For purposts of CESSAR, Safety Class 1, 2, 3, and NNS of D ANSI /ANS 51.1 are equivalent to Quality Groups A, B, C, and D of Raulatory Guide 1.26. (5) Flow restricting orifices are provided in the nozzles for the RCS sampling lines, the pressurizer level and pressure instruments, the RCP differential pressure instrument lines, the common SIS header pressure instrument lines, the RCP seal pressure instrument lines, the charging line differential pressure instrument line, and the SIS hot leg injection pressure instrument lines, to limit flow in the event of a break downstream of a nozzle. The orifice size, 7/32 inch diameter and 1 inch long, precludes exceeding fuel design limits while utilizing ntnimum makeup rates. This permits an orderly shutdown in the event of a downstream break in accordance with General Design Criterion 33 (see Section 3.1.29). A reduction may, therefore, be made in the safety classification of lines downstream of the orifice. (6) The pressure boundary housing for this component is a reactor vessel appurtenance and is Safety Class 1 and Seismic Category O I, as described in 3.9.4.3. Amendment D September 30, 1988
CESSAR 8!'Enceu O TABLE 3.2-1 (Cont'd) (Sheet 6 of 6) CLASSIFICATION OF
}J M PTURES. SYSTEMS. AND COMPONENTS (7) Core support structures are designed to the criteria described in 3.9.5.4.
(8) CEA and fuel assemblies are designed to the criteria described in 4.2. (9) Reactor coolant pump auxiliary components required for lubrica-tion and cooling of pump seals and thrust bearings are Quality Class 2. (10) Except Lifting Frame Assembly, which is NS. 0 (11) During normal plant operation only. (12) Safety Class 1 for pressure boundary; Safety Class 3 for electrical portion of system. I l O' Amendment D September 30, 1988
CESSAR HMincamu i G b TABLE 3.2-2 (Sheet 1 of 21) SAFETY CLASS 1. 2 & 3 VALVES Component Location / Safety Seismic Quality ! Identification Description Class Cateaor_Y Class Reactor Coolant System (RCS) (1) D 1 RC-212 Reactor vessel vent 1 I 1 RC-214 Refueling level 1 1 1 indicator RC-215, 216, 232, 332, RCS drains 1 1 1 233, 333, 234, 334, 235, 335 RC-248, 249, 252, 253, Reactor coolant pump (RCP) 2 I 1 256, 257, 260, 261 RC-208, 209, 218, 219, Pressurizer level 2 I 1 220 indicator RC-204, 205, 206, 207 Pressurizer pressure 2 1 1 indicator Pressurizer vent
/O RC-239 RC-200, 201, 202, 203 Pressurizer safety 1
1 I 1 1 1 RC-240, 241, 242, 243, Pressurizer spray line 1 I 1 236, 237 RC-100E, 100F Pressurizer spray line 1 1 1 control RC-244 Pressurizer spray line 1 1 1 check RC-210, 213, 238 Sample system 2 I 1 RC-211 Reactor vessel closure 2 I 1 head leakoff RC-292, 293, 294, 295, RCS pressure differential 2 1 1 296, 297, 298, 299 RC-7b2, 753, 754, 755 RCP teal housing drain 1 I 1 RC-7L2, 713, 714, 715 RCP vent 1 I 1 RC-446, 447, 448, 449, RCP HP cooler 1 1 1 450, 451, 452, 453 RC-772, 773, 774, 775 RCP HP cooler vent 1 I 1 RC-868, 869, 870, 871, RCP filter drain 1 I 1 700, 701, 702, 703 i RC-724, 725, 726, 727, RCP seal cooler pressure 2 1 1 ) 736, 737, 738, 739 ) RC-430, 431, 432, 433, RCP controlled bleedoff 2 I 1 344, 345, 346, 347 RC-380, 381, 382, 383 RCP vapor seal pressure 2 I 1 indicator Amendment D September 30, 1988
CESSAR EL%"lCATl2N i O. TABLE 3.2-2 (Cont'd) (Sheet 2 of 21) SAFETY CLASS 1. 2 1 3 VALVES Component Location / Safety Seismic Quality Identification Description Class .C31egoty Class Chemical and Volume Control System (CVCS) (1) CH-205 Auxiliary spray 1 I 1 CH-208 Charging line back- 1 I 1 pressure CH-209 Charging bypass line 1 I 1 CH-241 Seal injection flow 2 I 1 control (RCP 1A) CH-242 Seal injection flow 2 I 1 control (RCP 18) CH-243 Seal injection flow 2 I 1 control (RCP 2A) CH-244 Seal injection flow 2 1 1 control (RCP 26) CH-255 Seal injection contain. 2 I 1 isol. D CH-303 Charging line isolation 2 I 1 check CH-304 SCS Purification isol. 2 I 1 check CH-307 SCS Purification contain. 2 I 1 isol. CH-431 Auxiliary spray check 1 I 1 CH-433 Charging .ine check 1 1 1 CH-447 Auxiliary spray check 1 I 1 CH-448 Charging line check 1 I 1 CH-494 RSSH and RDP to RDH Check 2 I 1 CH-505 RCP CB0 contain. isol. 2 I 1 CH-506 RCP CB0 contain. isol. 2 I 1 CH-515 Letdown isolation 1 I 1 CH-516 Letdown backup isolation 1 1 1 CH-517 RHX isolation 2 I 1 CH-523 Letdown contain, isol. 2 I 1 CH-524 Charging line contain. 2 1 1 isol. , CH-560 RDT suction isolation 2 I 1 { CH-561 RDT isolatic., 2 I 1 1 CH-570 Letdown contain. isol. 2 I 1 ) CH-580 RMWS to RDT isolation 2 I 1 CH-787 Seal injection check 1 I 1 4 (RCP 1A) A2nendment D September 30, 1988 j l j
CESSAR Elaincam,. l{ ( TABLE 3.2-2 (Cont'd) (Sheet 3 of 21) EfETY CLASS 1. 2 & 3 VALVES Component Location / Safety Seismic QL!ality Identification Description Class Cateaory Class CH-802 Seal injection check 1 I 1 (RCP18) CH-807 Seal injection check 1 1 1 (RCP2A) CH-812 Seal injection check 1 I 1 (RCP 28) CH 835 Seal injection check 2 I 1 contain, isol. CH-866 Seal injection check 1 I 1 (RCP1A) CH-867 Seal injection check 1 1 1 (RCP1B) CH-868 Seal injection check 1 I 1 (RCP 2A) e CH-869 Seal injection check 1 1 1 k (RCP28) D Safety Injection System (SIS) and Shutdown Cooling System (SCS) (1) SI-100 IRWST level indication 2 1 1 isolation SI-101 IRWST level indication 2 I 1 isolation SI-102 IRWST level indication 2 I 1 isolation SI-103 IRWST level indication 2 I 1 isolation SI-104 IRWST return check 2 I 1 SI-105 IRWST return check 2 I 1 SI-106 Safety injection pump 2 I 1 isolation SI-107 Safety injection pump 2 I 1 isolation SI-108 EDT (CVCS) isolation 2 1 1 SI-110 EDT (CVCS) isolation 2 1 1 SI-lll Safety injection pump 1 2 I 1 isolation SI-112 Safety injection pump 2 2 I 1 isolation O SI-113 Safety injection pump 3 isolation 2 1 1 Amendment D September 30, 1988
I cESSAR nuir"icari:n l J IDE_ ?,2-2 (Cont'd) (Sheet 4 of 21) SAFETY CLASS 1. 2 & 3 val.VES , Component Location / Safety Seismic Quality Identification Description Class Cateaory Class SI-114 Safety injection pump 4 2 I 1 isolation SI-ll5 Safety injection pump 1 test isolation 2 1 1 S1-116 Safety injection pump 2 test isolation 2 I 1 SI-117 Safety injection pump 3 test isolation 2 I 1 SI-ll8 Safety injection pump 4 test isolation 2 I 1 SI-119 Safety injection pump 1 pressure indication 2 I 1 SI-120 Safety injection pump 2 pressure indication 2 I 1 SI-121 Safety injection pump 3 pressure indication 2 1 1 SI-122 Safety injection pump 4 pressure indication 2 I 1 SI-123 Safety injection pump 1 4 orifice bypass 2 1 1 SI-124 Safety injection pump 2 D r orifice bypass 2 I 1 SI-125 Safety injection pump 3 orifice bypass 2 I 1 SI-126 Safety injection pump 4 1 orifice bypass 2 I 1 SI-127 Safety injection pump 1 ; bypass check 2 I 1 l SI-128 Safety injection pump 2 i bypass check 2 I i SI-129 Safety injection pump 3 bypass check 2 1 1 i SI-130 Safety injection pump 4 bypass check 2 I 1 . S1-131 Safety injection pump 1 2 I 1 isolation i Amendment D September 30, 1988
i
-CESSAR !!nine-O -
TABLE 3.2-2(Cont'd) l (Sheet 5 of 21) ] l SAFETY CLASS 1. 2 & 3 VALVES j Component Location / Safety Seismic Quality l Identification Description Class Cateaory Class SI-132- Safety injection pump 2 isolation 2 I 1 SI-133 Safety injection pump 3 isolation 2 1 1 SI-134 Safety injection pump 4 isolation 2 I 1 i SI-135 Safety injection to EDT (CVCS) relief 2 I 1 1 SI-136 Safety injection to EDT (CVCS) relief-- 2 I 1 SI-137 Safety injection effluent sampling 2 I 1 SI-138 Safety injection effluent sampling I ( s SI-139 Safety injection pump 1 2 1 l l A discharge chech 2 I 1 SI-140 Safety injection pump 2 D discharge check 2 I 1 SI-141 Safety injection pump 3 . discharge check 2 I 1 SI-142 Safety injection pump 4 discharge check 2 I 1 SI-143 Safety injection pump 1 , discharge isolation 2 1 1 ! SI-144 Safety injection pump 2 ! discharge isolation 2 I 1 SI-145 Safety injection pump 3 discharge isolation 2 1 1 SI-146 5afety injection pump 4 discharge isolation 2 1 1 SI-147 Safety injection pump 1 pressure indication isolation 2 I 1 SI-148 Safety injection pump 2 pressure indication isolation 2 I 1 SI-149 Safety injection pump 3 pressure indication- . isolation 2 I 1 l Amendment D
l C E S S A R En nneuit,. l e; IbBLE 3.2-2 (Cont'd) (Sheet 6 of 21) SAFETY CLASS 1. 2 & 3 VALVES Component Location / Safety Seismic Quality Identification D_gteriDtion Class CateQorY Class SI-150 Safety injection pump 4 pressure indication isolation 2 I 1 SI-151 Safety injection pump 1 discharge to EDT relief 2 I 1 SI-152 Safety injection pump 2 discharge to EDT relief 2 I 1 SI-153 Safety injection pump 3 discharge to EDT relief 2 I 1 SI-154 Safety injection pump 4 discharge to EDT relief 2 1 1 SI-155 SI HL 1 flow indication 2 I 1 isolation SI-156 SI HL 2 flow indication isolation 2 I 1 S1-157 SI HL 1 flow indication isolation 2 1 1 Sf-158 SI HL 2 flow indication isolation 2 I 1 SI-159 Charging pump isolation 2 I 1 SI-160 Charging pump isolation 2 I 1 0 SI-161 Charging pump isolation 2 I 1 SI-162 Charging pump isolation 2 I 1 SI-163 SI HL 1 to EDT relief 2 ! 1 SI-164 SI HL 2 to EDT relief 2 I 1 SI-165 SI HL 1 check 2 I 1 SI-166 SI HL 2 check 2 I 1 SI-167 SI HL 1 pressure indication isolation 2 I 1 SI-168 SI HL 2 pressure indication isolation 2 1 1 SI-169 SI HL 1 containment check 2 I 1 SI-170 SI HL 2 containment check 2 1 1 SI-171 RDT relief 2 1 1 SI-172 EDT/ SIT relief 2 1 1 SI-174 SIT isolation 2 I 1 SI-176 EDT/ SIT local sample 2 I 1 SI-180 Charging pump isolation 2 I 1 SI-181 Charging pump isolation 2 I 1 SI-182 Charging pump isolation 2 I 1 SI-183 Charging pump isolation 2 I 1 Amendment D September 30, 1988 1
i CESSAR nainemow O IbS1LLh2 (Cont'd)- (Sheet 7 of 21) .,. SAFETY CLASS 1. 2 & 3 VALVES Component Location / Safety Seismic Quality _ldflitification -Description Class Cateaory Class SI-184 Charging pump isolation 2 1 1 SI-185 Charging pump isolation 2 I 1 SI-186 Charging pump isolation 2 1 1 SI-187 Charging pump isolation 2 I- 1-SI-188 SI 4 flow indication isolation . 2 1 1 SI-189 SI 3 flow indication isolation. 2 I 1. SI-190 SI 2 flow indication isolation 2 I 1 SI-191 SI 1 flow indication < isolation 2 1 1 SI-192 SI 4 flow indication isolation 2 I 1 SI-193 SI 3 flow indication isolation 2 I 1 SI-194 SI'2 flow indication D isolation 2 I 1 SI-19S SI 1 flow indication
' isolation 2 1 1 SI-196 SI 4 containment check 1 1 1 SI-197 SI 3 containment check 1 I 1 SI-198 SI 2 containment check 1 I 1 SI-199 SI 1 containment check 1 I 1 SI-200 SIT 4 vent 2 I 1 SI-201 SIT 3 vent 2 1 1 SI-202 SIT 2 vent 2 I 1 SI-203 filT 1 vent 2 I 1 SI-204 FIT 4 pressure indication isolation 2 I 1 SI-20S SIT 3 pressure indication I isolation 2 ~I 1 SI-206 UT 2 pressure indication isolation 2 I 1 SI-207 SIT 1 pressure indication isolation 2 I 1-SI-208 SIT 4 pressure indication isolation 2 1 1 51 '209 SET 3 pressure indicatio-isolation 2 I 1 \
Amendment D September 30, 1988 b _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ ____ __
CESSAR ML"lCATION TABLE 3.2-2 (Cont'd) el (Sheet 8 of 21) SAFETY CLASS 1. 2 & 3 VALVES Component Location / Safety Seismic Quality __ M nLiLigg ign _ Description Class .Qat_qsgry ' _ Class j SI-210 SIT 2 pressure indication isolation 2 I 1 SI-211 SIT 1 pressure indication isolation 2 I 1 SI-212 SIT 4 level indication isolation 2 1 1 SI-213 SIT 3 level indication isolation 2 I 1 SI-214 SIT 2 level indication isolation 2 I 1 SI-215 SIT 1 level indication isolation 2 I 1 SI-216 SIT 4 level indication isolation 2 I 1 SI-217 SIT 3 level indication isolation 2 I 1 SI-218 SIT 2 level indication isolation 2 1 1 SI-219 SIT 1 level indication D , isolation 2 1 1 l SI-220 SIT 4 level indication l isolation 2 1 1 1 SI-22i SIT 3 level indication ) isolation 2 I 1 1 SI-222 SIT 2 level indication I isolation '2 I 1 J SI-223 SIT 1 level indication I isalation 2 1 1 ; SI-224 SIT 4 level indication l isolation 2 I 1 , SI-225 SIT 3 level indication i isolation 2 1 1 l SIe/26 SIT 2 level indication - isolation 2 I 1 SI-227 SIT 1 level indication isolation 2 1 1 SI-228 SIT 4 fill and drain isolatier: 2 I 1 SI-229 SIT 3 fill and drain isolation 2 I 1 Amendment D September 30, 1988 l - _ - _ _ _ _ - _ _
~
1 4 CESSARnn% =,. I i 1 j V I 1ARLLL2-2 (Cont'd) (Sheet 9 of 21) l 4
~
}ffETY CLASS 1. 2 & 3 VALVES Component Location / Safety Seismic Quality dditatification Description Class Cottap_ty Class SI-230 SIT 2 fill and drain isolation 2 I 1 l SI-231 SIT 1 fill and drain '
isolation 2 I 1 SI-232 SIT 4 local sample isolation 2 I 1 1 SI-233 SIT 3 local sample ' isolation 2 I 1 SI-234 SIT 2 local sample l isolation 2 I 1 j SI 23S SIT 1 local sample ! isolation 2 1 1 SI-236 SIT 4 check 1 I 1 SI-237 SIT 3 check O SI-238 SI-239 SI-240 SIT 2 check SIT 1 check Injection line 4 pressure 1 1 1 I I I 1 1 1 ! i indication isolation 2 I 1 SI-241 Injection line 3 pressure ( indication isolation 2 I 1 51-242 Injection line 2 pressure ; irdication isolation 2 I I D j SI-243 Injection line 1 pressure indication isolation 2 I 1 j SI-244 Safety injection line 4 i check 1 I 1 SI-245 Safety injection line 3 check 1 I 1 i 51-246 Safety injection line 2 { check 1 I 1 j SI-247 Safety injection line 1 j check 1 I i i SI-300 IRWST containment isolation 2 I 1 j SI-301 IRWST containment isolation 2 I 1 i 51-302 IRWST containment isolation 2 I 1 l SI-303 IRWST containment isolation 2 1 1 S.!-304 SI HL 2 isolation 2 1 1 SI-305 SI HL 1 isolation 2 I 1 i SI-307 IRWST return isolation 2 I 1 l 4 SI-309 RDT isolation 3 I 1 I SI-310 SI 2 bypass isolation 2 I 1 Amendment D September 30, 1988 l j
CESSAR8Becmu O 'l IMLE 3.2-2 (Cont'd) (Sheet 10 of 21)
$_AFETY CLM$_L 2 & 3 VALVEj Component Location / Safety Seismic (juality
_ MantMigation Description Class 93tecory Class SI-311 S( l bypass isolation. 2 1 1
$1-312 SI '4 t.ontrinaent isolation 1 I 1 SI-313 SI 3 containment isolation- 1 1 1 SI-314 SI 2 containment isolation 1 I :
SI-315 SI 1 containment isolation 1 I 1 SI-316 SI til 2 containment isolation 2 I 1 SI-317 SI NL 2 containment isolation 2 1 1 SI-318 SCLL 2 bypass isolation 2 1 1 a SI-319 SCLL 3 bypass isolation 2 I 1 51-320 SIT 4 atmospheric rent isolation 2 I 1 51-321 SIT 3 atmospheric vent . isolation 2 I 1 SI-322 SIT 2 atmospheric vent isolation 2 I 1 D SI-323 SIT 1 atmospheric vent t isolation 2 I ! , 4 SI-324 SIT 4 atmospheric vent i:olation 2 1 1 S1-325 SIT 3 stuospheric vent isolatien 2 f 1 SI-326 SIT 2 atmospheric vent isolation 2 I 1
)
S1-327 SIT A attoospheric vent ! isolation 2 I 1 SI-328 Nitrogen pressure control 2 I 1 SI-329 Nitrogan pressure control 2 I 1 SI-330 Nitrogen pressure control 2 I 1 ! ' SI-331 Nitrogen pressure control 2 I 1 SI-332 Nitrogen pressure control 2 1 1 SI-333 Nitrogen pressure control 2 I 1 SI-334 Nitrogen pressure control 2 I 1 l SI-335 Nitrogen pressure control 2 1 1 j SI-336 SIT 4 fill and drain isolatios 2 I 1 1 SI-337 SIT S fill and drain isolation 2 1 1 SI-338 SIT 2 fill and drain i sol at' ion 2 I 1 l Amendment 0 l September 30, 1988 j l 1 :
CESSAR neinema i s ID LE 3.2-2 (Cont'd) (Sheet' 11 of 21) SAFETY CLASS 1. 2 & 3 VALVES i Component location / Safety Seismic Quality l Identifiution Description , qui Category Class
]
SI-339 SIT 1 fill and drain l isolation 2 I 1 SI-340 SIT 4 discharge isolation 1 2- 1 S(-341 SIT 3 discharge isolation 1 I 1 1 57-342 SIT 2 discherge isolation 1 1 1 j SI-343 SIT 1 discharge isolation 1 .I 1 , S1-345 Check valve leakage line i isolati6n 2 I 1 SI-346 Check valvs leakage line isolation 2 .! . 1- 1 SI-347 Check valve leakage lhe l isoldion - 2 1 1 i SI-387 Check valve leakage line l isolation 2 I 1 l
\ CS-500 r2P 2 . suction isol.ation .I 2 I CS -501 - CSP 1 suction isolation 2 l- 1 i
CS-502 CSP 2 check 2 ! I CS-503 CSP 1 check 2 1 1 C5-504 CSP 2 suction isolation 2 1 1 CS.505 CSP 1 suction isolation 2 I 1 CS-506 CSP 2 suction test 2 I 1 D CS-507 CSP 1 suction test 2 I 1 CS-508 CS miniflow HX 2 vent Z I 1 CS-509 CS miniflow HX 1 vont 2 I 1 C$-510 CS minifled HX 2 vent 2 I 1 i CS-511 CS miniflow HX 1 vent 2 I 1 ! CS-512 CS m!niflow HX 2 drain 2 I 1 CS-513 CS ininiflow HX 1 dr.ain 2 1 J CS-514 CS miniflow HX 2 drain 2 I 1 CS-515 CS miniflow HX 1 drain 2 I 1 CS-516 CSP 2 discharge pressure indication isolation 2 I 1 CC-5J7 CSP 1 discharge pressure indication isolation ~2 I 1 CS 518 CSP 2 discharge flow , indication isolation 2 I 1
)
l d.O j 4 Amendment D ' Septerber 30, 1988
- _ - _ _ _____ O
CESSARnnLue __ l LAJLE 3.2-2 (Cont'd) , I 1 (Sheet 12 of 21) S FETY CLASS 1. 2 1 3 VALyf} - Component Location / Safety Seismic Quality L l.19ntititgipf1_ _ __Q.UJ;dpiion _ C1 ass ERqgply jag _ C5-519 CSP 1 discharge flow indication isolation 2 I 1 CS-520 CSP 2 discharge flow indication isolation 2 I 1 CS 521 CSP 1 discharge flow - Ardication isoistion 2 I 1 k CS-5?2 CSP 2 discharge check 2 I 1 1 CS-523 CSP 1 discharge check 2 I .1 l CS-524 CSP 2 discharge isolation 2 I i ~ CS-525 CSP : discharge isolation 2 I I CS'525 CS6 2 discharge isolation 2 I 1 , CS-527 CSP J discharge isolation 2 1 R CS-528 CS HX 2 inlet pressure f indication isolatiu 2 I 1 CS-529 CS HX i !nlet pressur.e , indir.ation isolation 2 I 1 f CS-530 CS HX 2 drain 1 I 1 CS-531 CS RX 1 draits 2 I 1 CS-532 CS HX 2 Grain 2 I 1 CS-533 CS HX 1 drain 2 1 1 CS-534 CS MX 2 vent 2 I 1 g CS-$35 CS HX 1 vent ~2 1 1 l CS-536 CS HX 2 rent 2 .I 1 l CS 537 CS HX 1 vent E I i i CS-V38 CS HX 2 to EDT rolief 2 i 1 - C1-539 CS HX 1 to EDT relief 2 I 1 ) CS-540 CS HX 2 isolation 2 I 1 ! l CS-541 CS HX 1 i' solation 2 1 1 CS-542 CS HX 'I to IRWST isolation 2 I 1 CS-543 CS HX 1 to IRWST isolation 2 I 1 CS-544 Containment spray outside ) l containment test 2 I 1 1 CS-545 Containment spray outside containment test 2 I 1 ! CS-546 Containment spray check 2 1 1 l CS 547 Containment . spray check 2 1 1 l CS-548 Containment spray j in-containtsent test 2 I 1 i CS-549 Containment spray i in-containment test 2 I 1 l l Amendment D I September 30, 1988 ,
i i CESSARn h ou l l _ __ _ . _ m_ j n l l LJ l IA4LE.1&2 (conted) (Sht:#t 13 of 21) l l SAffJL;te'SS.It.LJ 3 VALVES
]
Component Location / Safety Seismic Quality 149M10fiation pnqil;ttiqn _,_ {ing_ c_aiespf,y _Q.1 m _ CS 600 Containnient spre hardn J 2 isolation 2 I i ! C3-601 Cor,tainwent spre, header ) 1 i nlation 2 1 1 SD-650 SCSH1. 2 bypass isclation 2 1 1 50-651 SCSHX 1 bypass isoldion 2 I 1 S0-652 SCSE 2 outlet isolatic,n 2 I 1 S0-653 SCSKX 1 uutlet isolation 2 1 1 5D-654 SCS 2 cf>ntainneent isolation 2 1 1 i S0-655 SCS 1 containment isolation 2 1 1 SD-656 Safsty injection warmup itolation 2 i 1 SD-667 Safety injection warmup (p) SR 658 isol,a tion SCP 2 containment isolation 2 2 I I 1 1 t kJ SD-659 SCP 1 containment isolation 2 1 1 ) SD-670 SCLL 2 backup isola' lion 2 I 1 l SD G71 SClit 1 backup isolation 2 1 1 l SD 671 SCLL 2 1 solation 2 I 1 l SD-673 SCLL 1 holatica 2 I I SDt700 PCPS isolation 2 I 1
$0-701 PCPS isolation 2 I 1 0
50-702 Shutdown purification tsolation 2 I 1 SD-703 Shutdown purification i sol atior. 2 I 1 SD-704 PCPS/EST is61ation 2 I 1 q
$D-705 PCPS/EDT isolation 2 1 1 SD-706 EDT relief 2 I 1 50-707 EDT relief 2 1 1 l 50-708 SCLL 2 local sample !
isolation 2 1 1 l SD-709 SCLL 2 local sample isolation 2 1 1 SD-710 Containment spray pump 2 isolation { 2 1 1 ! SD-711 Containment spray pump 1 ' isolation 2 I 1
. SD-712 SCP 2 suction isolation 2 I 1 1 :
i'O SD-713 SCP 1 suction isolation (). SD-714 SCP 2 suction test isolation 2 2 1 I 1 1 l. i Ame.ndment D ! Sept ember 30, 1988 _____ - - - - - - - l
I CESSARHi!% mt .. O1 TABLE 3.2_-2(Cont'd) (Sheat 14 of 21) , SffiY CLA$} 1. 2 1 3 VALVES Component . i.ocatlon/ Safety Seismic Qualtty ) IGitt.liinfli2D__. - D913r.10110A_ Gla11 U19.S9.tX Class 1
.S0 715 SCP 1 suction test isolation 2 I 1 SD-716 SCE mtnificw HX ? Vent 2 i 1
$D-717 SCS miniflow HX 1 vent 2 I 1 S0-718 SCS miniflow HX 2 tunt 2 1 1 i SD-71t SCS miniflow HX 1 vent 2 'I 1 SD-776 SCS mlniflow HX 2 drain '2 I J SD-721 SCS miniflow HX 1 drain 2 I i S3-722 SCS miniflow HX 2 drain 2 I 1 SD-723 SCS :ciniflow HX 1 drain 2 I 1 3
SD 724 SCP 2 discharge pressure { indication isolation 2 I 1 I SD-726 SCP 1 discharge pressure indication isolation 2 1 1 50 716 SCP 2 discharge flow ' indication isolation 2 I 1 l S0-727 SCP 1 discharge flow a' indication is,olation 2 1 1 D SD-728 SCP 2 discharge flow indication isolation 2 I 1 ) 30-729 SCP 1 discharge flow ind ham on 1s<4ation ? 1 1 SD-730 SCP 2 oui.let check 2 1 1 SD-731 SCP 1 discharge check 2 I 1 50-732 SCP 2 outlet isolation 2 I 1 SD-733 SCP 1 discharge isolation 2 1 1 4 SD-T34 SCSHX 2 bypass 2 I 1 50-735 SCSHX 1 bypass 2 J 1 f SD-736 SCP 2 inlet isolation 2 I 1 i S0-737 SCSHX 1 inlet isolation 2 'I 1 l SD-738 SCSHX 2 inlet pressure l indication isolation 2 I 1 SD T39 !'SHX 1 inlet pressure indication isolation 2 I 1 i S0440 SCSHX 2 vent 2 I 1 l SD-141 SCSHX 1 vent 2 I 1 l SD-742 SCSHX 2 vent 2 1 1 l SD-743 SCSHX 1 vent 2 I 1 , i S0 744 SCSHX 2 drain 2 I 1 i 50 745 SCSHX 1 drain 2 I 1 1 S0 746 SCSHX 2 drain I i 1 SD1747 SCSHX 1 drain 2 1 1 . Amendment D September 30, 1988
CESSAR ER' ricamu A l \ l TABLE 3.2-2 I, Cont'd) l lL (Sheet 15 of 21) SAFETY CLASS 1. 2 & 3 VALVES i Component Location / Safety Seismic Quality l I(Ltatificat19_0__ _ _ Description Class Cateoory Clast ! t i l S0-748 SDCHX 2 to EDT relief 2 I l- ! SD-749 SDCHX 1 to EDT relief 2 I 1 SD-750 Safety injection. isolation 2 I 'l 50-751 Safety injection isolation 2 I 1 SD-752 Safety injection to EDT-relief 2 I 1 50-753 Safety injection to EDT relief 2 1 1 SR 754 Safety injection to shutdown purification isolation 2 I 1 5D-755 Safety injection to shutdown purification isolation 2 I 1 SD-766 SCS 2 check 2 I 1 f S0-767 3CS 1 check- 2 I 1 I Safety Depressurization System (SDS) J l (LATER) 9 1 Emergency Feedwater System (EFW) (1) EF-100 Steam-Driven EFW Pump I 2 I 1 Steam Generator Isolation EF-101 Steam-Driven EFW Pump 2 j Steam Generator Isolation 2 I 1 q l EF-102 Motor-Driven EFW Pump 1 1 Steam Generator.1sv ation 2 1 1 EF-103 Motor-Driven EFW Pump 2 i Steam Generator Isolation 2 1 .1 ! EF-104 Steam-Driven EFW Pump 1 Flow j Control 3 1 1 d ! EF-105 Steam-Driven EFW Pump 2 M ow l l Control 3 I 1 ! l EF-106 Motor-Driven EFW Fuap.1 Flow ! l Control' 3 I 1 EF-107 Hotor-Driven EFW Pump 2 Flow j Control :3 1 1 i EF-108 Steam-Driven EFW Pump ; Turbine 1 Steam Supply ;
/~ Isolation 2 I- 1 Amendment D September 30, 1988 1
CESSAR !a'Jnema eij MSLE 3J-R (Cont'd) (Sheet 16 of 21) JAFETY CM$$ 1. mP LLY8LYfJ Component Location / Safety Seismic Quality _ldsIltiLLGation Desg.tjpt19J1 __ G_hui., 031eggn ,,S1 ass EF-109 Steam-Driven EFW Pump Turbine 2 Steac Supply Isolation 2 I 1 EF-110 Steam Supply Drain 1 Isolation 2 I 1 EF-lli Steam Supply Drain Isolation 2 I 1 EF-ll2 Steam Supply Bypass Isolation 2 I 1 l EF ll3 Steam Supply Bypass . Isolation 2 I 1 EF-200 Steam Generator Isolation Check Valve 2 I 1 EF 201 Steam Generator Isolation Check Valve 2 I 1 EF-202 Steam Generator Isolation Check Valve 2 I 1 . EF-203 Steam Generator Isolation ) Check Valve 2 I 1 0 j i EF-204 EFW Pump Discharge Check 3 I 1 EF 205 EFW Pump Discharge Check 3 I 1 1 EF-206 EFW Pump Discharge Check 3 I 1 j EF 207 EFW Pump Discharge Check 3 I 1 1 EF-200 EFH Pump Suction Isolation 3 1 1 EF-209 EFW Pump Suction 1sulation 3 I 1 EF-210 EFW Pump Suction Isolation 3 1 1 EF-211 EFW Pump Suction Isolation 3 I 1 r EF 212 EFWST Crossover Isolation 3 I 1 EF-213 EFWST Crossover Isolation 3 1 1 EF 214 Non-Safety Condensate Source Isolation Check 3 I 1 j EF-215 Non-Safety Condensate Source Isolation Check 3 I 1 EF-216 EFWST-1 Drain Isolatf on 3 I 1 EF-217 EfWST-2 Drain Isolation 3 I 1 EF-220 EFW Pump Minimum flow , I. solation ? I 1 EF-221 EFW Pump Minimum Flcw Isolation 3 I 1 EF-222 EFW rump Minimum Flow Isolation 3 I 1 l EF-223 EFW Punp Minimum Flow Isolation 3 I 1 l Amendment D l September 30, 1985 l _ _ i
CESSAR Einincam. I l TABLE 3.2-2 (Cont'd) J (Sheet 17 of 21)
]
SAFETY CtASS 1. 2 & 3 VALVES component Location / Safety Seismic Quality Identification .. Description Class Catecory Class EF-224 Full Flow Test Bypass l Isolation -3 I 1 EF-225 Full Flow Test Bypass Isolation 3 I 1 l EF-226 Full Flow Test Bypass j Isolation 3 I 1 l EF-227 Full Flow Test Bypass j Isolation 3 I 1 EF-228 Full Flow Test Flow Control 3 1 1 EF-229 Full Flow Test Flow Control 3 1 1 EF-230 Full Flow Test Flow Control 3 I 1 : EF-231 Full Flow Test Flow Control 3 I 1 i EF-232 Full Flow Test Bypass Isolation 3 I :1 [ EF-233 Full Flow Test Bypass Isolation s, 3 I 1 EF.234 Full Flow Test Bypass D Isolation 3 I 1 EF-235 Full Flow Test Bypass Isolation 3 I 1 EF-236 Steam-Driven EFW Pump Turbine Bearing 011 Cooler Return Isolation 3 I 1 EF-237 Steam-Driven EFW Pump Turbine Bearing 011 Cooler Return Is.olation 3 I 1 EF-238 Steam Supply Maintenance
. Isolation 2 I 1 EF-239 Steam Supply Maintenance
' Isolation 2 I 1 EF-240 Steam Supply Drain Isolation 2 1 1 1 EF-241 Steam Supply Drain Isolation 2 I 1 EF-244 Steam Supply Drain Isolation 2 1 1 EF-245 Steam Supply Drain Isolation 2 1 -1 EF-246 Steam Supply Drain Isolation 2 I 1 EF-247 Steam Supply Drain Isolation 2 I 1 EF-248 Steam Supply Drain Isolation 2 I 1 EF-249 Steam Supply Drain Isolation 2 1 1
)
Amendment D September 30, 1988
CESSAR E!Ence i l
@1i IfBLE 3.2-2 (Cont'd) j (Sheet 18 of 21) 3MIII CLASS 1. 2 & 3 VALVES Con 1ponent Location / Safety Soismic Quality Identification heriotion plass - Cateaory C1 ass EF-250 Steam Exhaust Orain Isolation 3 I 1 EF-251 Steam Exhaust Drain Isolation 3 I 1 l EF-252 Steam Exhaust Drain Isolation 3 I 1 EF-253 Steam Exhaust Drain Isolation 3 I 1 EF-254 Steam Exhaust Drain Isolation 3 I 1 EF-255 Steam Exhaust Drain Isolation 3 I 1 L EF-256 Flow Indicator Isolation 3 I 1 EF-257 Flow Indicator Isolation 3 I 1 EF-258 Flow Indicator Isolation 3 I 1 EF-259 Flow Indicator Isolation 3 I 1 EF-260 Flow Indicator Isolation 3 I 1 EF-261 Flow Indicator Isolation 3 I 1 EF-262 Flow Indicator Isolation 3 I 1 D
EF-263 Flow Indicator Isolation 3 I 1 EF-264 Pressure Indicator Isolation 3 I 1 EF-265 Pressure Indicator Isolation 3 1 1 EF-266 Pressure Indicator Isolation 3 I 1 EF-267 Pressure Indicator Isolation 3 I 1 EF-268 Pressure Indicator Isolation 3 I 1 EF-269 Pressure Indicator Isolation 3 I 1 EF-270 Pressure Indicator Isolation 3 I 1 EF-271 Pressure Indicator Isolatlon 3 I 1 EF-272 Flow Indicator Isolation 3 I 1 EF-273 Flow Indicator Isolation 3 I 1 EF-274 Flow Indicator Isolation 3 I .1 EF-275 Flow Indicator Isolation 3 I 1 EF-276 Flow Indicator Isolation 3 I 1 EF-277 Flow Indicator Isolation 3 I 1 ! EF-278 Flow Indicator Isolation 3 I 1 l EF-279 Flow Indicator Isolation 3 I 1 l EF-280 Pressure Test Isolation 3 I 1 1 EF-281 Pressure Test Isolation 3 I 1 1 EF-282 Pressure Test Isolation 3 I 1 EF-283 Pressure Test Isolation 3 I 1 EF-284 Level Indication Isolation 3 I 1 l k Amendment D l September 30, 1988
CESSARUninc-O %) TABLE 3.2-2 (Cont'd) (Sheet 19 of 21) SAFETY CLASS 1. 2 & 3 VALVES Component Location / Safety Seismic Quality Identification Description Class Cateaory Class EF-285 Level Indication Isolation 3 I 1 EF-286 Level Indication Isolation 3 I 1 EF-287 Level Indication Isolation 3 I 1 EF-288 EFW Pump Discharge Crossover Isolation 3 I 1 EF-289 EFW Pump Discharge Crossover Isolation 3 I 1 EF-290 EFW Pump Discharge Crossover Isolation 3 1 1 EF-291 EFW Pump Discharge Crossover Isolation 3 I I EF-292 Pressure Indication Isolation 2 I 1 . EF-293 Pressure Indication I Isolation I iO EF-294 Pressure Test Isolation 2 2 I 1 1 EF-295 Pressure Test Isolation 2 I 1 EF-296 Pressure Test Isolation 3 1 1 EF-297 Pressure Test Isolation 3 I 1 EF-298 Level Switch Isolation 2 1 1 EF-299 level Switch Isolation 2 I 1 EF-300 Level Switch Isolation 3 I 1 D EF-301 Level Switch Isolation 3 1 1 l EF-310 EFWST Cleanup Isolation 3 I 1 EF-311 EFWST Cleanup Isolation 3 I 1 EF-312 EFWST Sample Isolation NNS NS 1 ; EF-313 EFWST Sample Isolation NNS NS 1 1 EF-314 EFWST Cleanup Isolation NNS NS 2 EF-315 EFWST Cleanup Isolation NNS NS 2 EF-316 Steam Supply Drain Isolation 2 I 1 EF-317 Steam Supply Drain Isolation 2 1 1 EF-318 Steam Supply D ain Isolation 2 I 1 EF-319 Steam Supply Drain Isolation 2 I 1 EF-320 Steam Supply Drain Isolation 2 I 1 EF-321 Steam Supply Drain Isolation 2 1 1 EF-322 Steam Supply Drain Isolation 2 I 1-EF-323 Steam Supply Drain Isolation 2 I .1 EF-324 Steam Supply Drain Iso htion 2 I 1 EF-325 Steam Supply Drain Isolation 2 I 1 l EF-326 Steam Supply Drain Isolation 2 I 1 Amendment D September 30, 1988
CESSAR EEGcuen O TABLE 3.2-2 (Cont'd) (Sheet 20of21) 1AEEH_CJA1512 & 3 VALVES Component Location / Safety Seismic Quality Identification Description Class Eftecory _ Class EF-327 Steam Supply Orain Isolation 2 I 1 EF-328 Turbine Case Drain Isolation 3 1 1 EF-329 Turbine Case Drain Isolation 3 I 1 EF-330 Steam Supply Bypass Maintenar,ce Isolation 2 I 1 EF-331 Steam Supply Bypass Maintenance Isolation 2 I 1 EF-334 Level Indication Isolation 3 I 1 D EF-335 Level Indication Isolation 3 I 1 EF-336 Level Indication Isolation 3 I 1 EF-337 Level Indication Isolation 3 I 1 i EF-338 EFW Pump Discht.rge Maintenance Isolation 3 I 1 EF-339 EFW Pump Discharge , Maintenance Isolation 3 I I EF-340 EFW Pump Discharge Maintenance Isolation 3 I 1 EF-341 EFW Pump Discharge i Maintenance Isolation 3 7 1 O Amendment D September 30, 1988 l l
CESSAR nainc nou O IAEIJ E (Cont'd) (Sheet 21 of 21) SAFETY CLASS 1. 2 & 3 VALVES 0 NOTE: (1) All containment isolation valves (and their , operators) within C-E's scope of design - including manual valves, check valves, and relief valves which also serve as isolation valves will be subject to the pertinent requirements of the Quality Assurance Program as given in Chapter 17. l O Amendment D September 30, 1988
CESSAR annncmon O TABLE 3.2-3 RELATIONSHIP 0F SAFETY CLASS TO CODE CLASS Code Class Safety Class (ASME Section IIII k SC-1 1 ! i 50-2 for reactor MC containment components SC-2 for fluid system 2 components SC-3 for core support CS structures D SC-3 (otherwise) 3 NNS Industry Standards O i 1 l l O l Amendment D September 30, 1988 l
i 1 CESSAR !!ntricuin O TABLE 3.2-4 1 SUMARY OF CRITERIA - STR'jCTURES (LATER) O O Amendment D September 30, 1988
CESSAR nuincariou l 3.3 WIND AND TORNADO LOADINGS All Seismic category I structures, except those not exposed to wind, are designed for wind and tornado loadings. 3.3.1 WIND LOADINGS The design for wind loading is in accordance with ANSI A58.1,
" Building Code Requirements for Minimum Design Ioads in Buildings and other Structures" (Reference 1), and ASCE Paper 3269, " Wind Forces on Structures" (Reference 2).
3.3.1.1 Desian Wind Velocity A design wind velocity of 130 mph, at a height of 30 feet above nominal ground elevation is used as the fastest mile of wind for a 100 year recurrence period. Vertical wind velocity profiles and associated effective pressures are calculated in accordance with Appendix A6 of Reference 1 utilizing an Importance Factor, 1, of 1.07 and Exposure C. In accordance with Reference 2 and due to the massive size and rigidity of all Seismic Category I structures, a gust factor equal to unity is used for the wind load design of these structures. D 3.3.1.2 Determination of ADolied Forces Based on structure geometry and physical configuration, the effective pressure distribution is transformed into applied equivalent static building forces utilizing appropriate shape coefficients given in References 1 and 2. Wind pressure distribution curves for the containment shield buildMg are shown in Figure (LATER). The maximum height of the shield building above grade is approximately 173 feet 3 inches. 3.3.2 TORNADO LOADINGS All Seismic Category I structures, except those structures not exposed to wind, are designed for tornado loadings. 3.3.2.1 ADDlicable Desian Parameters Tornado effects are in acccrdance with ANSI /ANS-2.3, "Standa.d for Estimating Tornado and Extreme Wind Characteristics at: O Nuclear P per year.ower basis tornado. TheSites" (Reference 3), using a probability of 10 ' ' following parameters are applicable to the design l Amendment D 3.3-1 September 30, 1988
1 CESSAR EHL"icari u O Transnational velocity: 57 mph Paximum wind speed: 260 mph Radius: 453 feet Maximum pressure differential: 1.46 paid Missile Spectra: Per INSl/ANS-2.3 3.3.2.2 Determination of Forces on Structures The forces on Seismic Category I structures due to tornado wind loadings are obtained using methods outlined in Section 3.3.1.2, with a wind velocity of 260 mph (vector sum of all component velocities). Velocity profiles are determined as outlined in Section 3.3.1.1. Tornado loadings include tornado wind pressure, internal pressure due to tornado created atmospheric pressure drop, and forces generated due to the impact of credible tornado missiles. These loadings are combined with other loads as described in Section 3.8. 3.3.2.3 Effect of Failure of Structures or Components Not Desicned for Torn 3do Loads Adjacent structures 'll not be permitted to affect or degrade the capability of f,e.. lic Category I structures to perform their intended safety fr.nctions as a result of tornado loadings. This is accomplished !/ one of the following methods: A. Designing the adjacent structure to Seismic Category I torne.,to loadings. B. Tilvestigating the effect of adjacent structural failure on Seismic Category I structures to determine that no I impairment of function results. I C. Designing a structural barrier to protect Seismic Category I structures from adjacent structural failure. O Amendment D 3.3-2 September 30, 1988 i -
CESSAR ' an"icuion 6 O REFERENCES FOR SECTION 3.3
- 1. " Minimum Design Loads for Buildings and Other Structures,"
ANSI A58.1-1982.
- 2. " Wind Forces on Structures," ASCE Paper No. 3269, Transactions, ASCE, Vol. 126, Part II, 1961, p. 1124.
3 4 " Standard for Estimating Tornado and Extreme Wind Characteristics at Nuclear Power Sites," ANSI /ANS-2.3-1983. O O Amendment D 3.3-3 September 30, 1988
CESSAR EEnnCATCH o 3.4 WATER LEVEL (FLOOD) DESIGN All Seismic Category I structures, components and equipment are designed for applicable loadings caused by postulated floods. Section 2.4 of the site-specific SAR describes, in detail, the relationship of the site-specific flood levels to safety-related buildings and facilities. 3.4.1 FLOOD ELEVATIONS The elevation level for floods at the reactor site is determined in accordance with Regulatory Guide 1.59, " Design Basis Floods i for Nuclear Power Plants," and ANSI /ANS 2.8-1983, " Determining l Design Basis Flooding at Power Reactor Sites." The design basis level for the System 80+ Standard Design is limited to 1 foot below plant finished yard grade as the minimum flood level value. The maximum flood level value is site-specific and protection measures for that flood level are described in Section 2.4 of the site-specific SAR. 3.4.2 PHENOMENA CONSIDERED IN DESIGN LOAD CALCULATION All safety-related structures of the reactor building complex are D designed to withstand the static and dynamic forces of the plant flood level. Other safety-related structures or systems essential for plant operation are designed for the site-related flood level as described in Section 2.4 of the site-specific SAR. 4 3.4.3 FLOOD FORCE APPLICATION The design flood is used in determinirq the applicable water level for design of all Seismic Cr;agory I structures in accordance with the load combinations discussed in Section 3.8.4. The forces acting on those structures are determined on the basis of full external hydrostatic pressure corresponding to that flood level. All Seismic Category I structures will be in a stable i condition due to both moment and uplift forces resulting from the ! proper load combinations, including design basis flood levels. ], 3.4.4 FLOOD PROTECTION ; 3.4.4.1 Flood Protection Nessures for Seismic Catecory I Structures The flood protection measures for Seismic Category I structures, systems and components are designed in accordance with Regulatory Guide 1.102, " Flood Protection for Nuclear Power Plants. " The following structures and systems in the reactor complex area are [Si designed for flood level protection: V Amendment D 3.4-1 September 30, 1988
CESSAR EHL"icarin. O
- Reactor Building
- Diesel Generator Buildings
- Control Building
- Fuel Building
- Auxiliary Building These safety-related structures are designed to maintain a dry environment during all floods by incorporating the following safeguards into their construction:
A. No exterior access openings will be lower than 1 foot above plant grade elevation. B. The finished yard grade adjacent to the safety-related structures will be maintained at least 1 foot below the ground floor elevation. C. Waterstops are used in all horizontal and vertical construction joints in all ex. erior walls up to flood level elevation. D. Water seals are provided for all penetrations in exterior walls up to flood level elevation. D For other safety related structures where flood protection measures are required (e.g. pumping systems, stoplogs, watertight doors, dikes, retaining walls and drainage systems) the design of means for providing such protection will be described in Section 2.4 of the site-specific SAR. 3.4.4.2 Permanent Dewaterina System The System 80+ design does not incorporate a permanent dewatering system to control the level of ground water. All Seismic Category I structures of the reactor building complex are designed for full hydrostatic water pressure to within 2 feet below yard grade elevation. 3.4.5 ANALYTICAL AND TEST PROCEDURES A description of the methods and test procedures by which static and dynamic effects of the design basis flood conditions or design basis groundwater conditions are applied is detailed in Section 2.4 of the site-specific SAR. O Amendment D 3.4-2 September 30, 1988
CESSAR nnincyI,. 3.5 MISSILE PROTECTION The missile protection design for Seismic category I stMctures, systems and components is described in this section. Missile protection or redundancy is provided for Seismic Category I equipment and components such that internal and external missiles will not cause the. release of significant amounts of radioactivity or prevent the safe and orderly shutdown of the reactor. The protection of essential structures, systems and components will be accomplished by one or more of the following: ! A. Minimizing the sources of missiles by equipment design features that prevent missile generation. j B. Orientation or physical separation of potential missile ] sources away from safety-related equipment and components. C. Containment of potential missiles through the use of I protective shields and barriers-near the source. O D. The hardening of safety-related equipment and components to withstand missile impact, where such impacts cannot be reasonably avoided by the methods above. 0 i l ANSI /ANS-58.1, " Plant Design Against Missiles," is used as a , guide for missile protection. I 3.5.1 MISSILE SELECTION AND DESCRIPTION
'I Potential missiles are identified and characterized by type and .)
source and their probability of occurrence, retention and impact. I For equipment with energy sources capable of creating a missile, the selection is based on the application of. a single-failure criterion to the retention, features- of the component. Where sufficient retention redundancy is provided in the event of a failure, no missile is postulated. Internally generated missiles can be generated potentially from two types of equipment: rotating components and pressurized components. Rotating components include turbine wheels, fans, auxiliary pumps and their . associated motors. Pressurized components include valves, heat exchangers, vessels ~ and their associated components. O Amendment D 3.5-1 September 30, 1988
CESSAR8 mince O The types of missiles considered and/or not considered in the design of Seismic Category I structures, systems, and components are discussed in the following sections: i A. Internally Generated Missiles (Outside Containment), described in Section 3.5.1.1. B. Containment Internal Missiles, defined in Table 3.5-1 and Section 3.5.1.2. C. Turbine Missiles, described in Section 3.5.1.3. D. Natural Phenomena (Tornado) Missiles, described in Section 3.5.1.4. E. Site Proximity Missiles (Except Aircraft), described in Section 3.5.1.5. F. Aircraft Hazards, described in Section 3.5.1.6. 3.5.1.1 Internally Generated Missiles (Outside Containment) i Internally generated missiles (outside containment) from rotating and pressurized components are not considered credible for the reasons discussed below. 3.5.1.1.1 Auxiliary Pumps and Motors There are no postulated missiles originating from auxiliary pumps and associated motors outside containment for the following reasons: A. The pump motors are synchronous induction type which have D relatively slow running speeds and are not prone to overspeed. The motors are all pretested at full running speed by the motor vendor prior to installation. B. In addition to the low likelihood of missiles due to motor overspeed as discussed in A. above, the actor stator would tend to serve as a natural container of rotor missiles if there were to be any. 1 C. The pumps all have relatively low suction pressures and, therefore, would not tend to be driven to overspeed due to a pipe break in the discharge line. In addition, the synchronous motor would tend to act as a brake to prevent pump overspeed. O' Amendment D , 3.5-2 September 30, 1988 1
CESSAR8HMc-1 (m)
%/
D. Industry pump designs are such that ( and service history shows) no occurrences of impeller pieces penetrating pump casings. 3.5.1.1.2 Valves l There are no missiles postulated from valves for the following reasons: A. All valve stems are provided with a backseat or shoulder larger than the valve bonnet opening. B. Motor operated and manual valve stems are restrained by stem threads. D C. Operators on motor, hydraulic and pneumatic operated valves prevent stem ejection. 1 D. Pneumatic operated diaphragms and safety valve stems are restrained by spring force. E. All valve bonnets are either pressure sealed, threaded or i
- bolted such that there is redundant retention for prevention of missile generation.
3.5.1.1.3 Pressure Vessels All pressurized vessels are considered moderate energy (275 psig) or less and are designed and constructed to the standards of the ASME Code. In addition to the ASME Code examination and testing requirements, all vessels will receive periodic in-service ir.epections . Where appropriate, these components are provided with pressure relief devices to ensure that no pressure buildup will exceed material design limits. On this basis, moderate energy pressure vessels are not considered credible missile sources. 3.5.1.2 Internally Generated Missiles (Inside Containment) Table 3.5-1 lists postulated missiles from equipment inside ; containment, and summarizes their characteristics. Included are j major pretensioned studs and nuts, instruments, and the CEDM i missile. Other 1tems which were considered and specifically excluded because of redundant retention features are valve stems, valve bonnets and pressurized cover plates. n V i Amendment D 3.5-3 September 30, 1988
CESSAR raa CERTIFICATION O 3.5.1.3 Turbine Missiles The probability of turbine missile generation and adverse impact effects on Seismic Category I systems and components is assured to be acceptably low by a combination of the following measures: A. Reliable turbine overspeed protection provisions. B. Adequate assurance of turbine disc integrity. C. Placement and orientation of the turbine generator. D. Consideration of the protection provided by plant structures not explicitly designed as barriers that may limit missile . penetrating capabilities to less than the capability of Seismic Category I structures. , 3.5.1.4 Missiles Generated by Natural PhenortgAt Tornado-generated missiles are the limiting natural hazard and, as such, are a part of the design basis for Seismic Category I structures and components. Table 3.2-4 (LATER) lists those structures, shields and barriers that will be designed for tornado missile effects. The missiles considered in the design are given in Section 3.3.2.1 and Table 3.4-1 of ANSI /ANS 2.3-1983. D 3.5.1.5 Missiles Generated by Events Near the Site Justification will be provided in the site-specific SAR. 3.5.1.6 Aircraft Hazards Justification will be provided in the site-specific SAR. 3.5.2 STRUCTURES, SYSTEMS, AND COMPONENTS TO BE PROTECTED l FROM EXTERNALLY GENERATED MISSILES j Tornado missiles are the design basis missiles from external sources. All safety related systems, equipment and components required to safely shut the reactor down and maintain it in a safe condition are housed in Category I structures designed as tornado resistant (see Section 3.5.1.4) and as such are considered to be adequately protected. 3.5.3 BARRIER DESIGN PROCEDURES Missile barriers, whether steel or concrete, are designed with sufficient strength and thickness to stop postulated mistiles and Amendme:;t D l 3.5-4 September 30, 1988 l _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . i
1 l CESSAR EnHrbo i p) ( v to prevent overall damage to Seismic Category I structures. The procedures by which structures and barriers are designed to perform this function are presented in this section. 3.5.3.1 Local Damace Prediction The prediction of local damage in the immediate vicinity of an impacted area depends on the basic material of construction of l the barrier itself (i.e. either concrete or steel). Corresponding procedures are discussed separately below. 3.5.3.1.1 Concrete Structures and Barriers I Local damage prediction for concrete structures includes the j estimation of the depth of missile penetration and an assessment of whether secondary missiles might be generated by spalling. l Generally, the Modified Petry Formula or the Modified NDRC { Formula (References 2 and 3) is used to estimate missile l penetration with appropriate constants taken from available test j data. To insure that no secondary missiles (due to spalling) are j generated, a minimum barrier thickness of 3 times the penetration , depth is provided. Depending on certain missile characteristics, ! r~N additional penetration formulas may be employed as justified by (v) full scale impact tests (References 3 and 4). D
)
i i 3.5.3.1.2 Steel Structures and Barriers l The Stanford equation (Reference 5) is used as the basis for the design and analysis of steel structures and barriers. 3.5.3.2 Overall Damace Prediction i The overall response of a structure or barrier to missile impact depends largely on the location of impact (e.g. near mid-span or . I near a support), the dynamic and deformation properties of the barrier and the missile, and the kinetic energy of the missile itself. Depending on the deformation characteristics of both the barrier and the missile, an impact force time history can be developed using either work-kinetic energy principles or conservation I of momentum. The structural response to this impulse loading, in conjunction with other appropriate design loads, is evaluated by the procedures given in References 3 and 6. 3.5.4 INTERFACE REQUIREMENTS ) g' Protection for all Seismic category I structures, systems and { components shall be provided by one of the following: [ Amendment D 3.5-5 September 30, 1988
- _ _ _ _ _ _ - _ _ _ ~
CESSAR n!Mneww A. For systems and parts of systems located inside the e containment (RCS and connected systems, Engineered Safety , Feature systems), appropriate missile barrier design ] procedures are used to ensure that the impact of any potential missile will not lead to a loss-of-coolant-accident or preclude the systems from carrying out their specified safety functions. B. For systems and equipment outside containment, appropriate legign procedures (e.g., proper turbine orientation, natural separation, or missile barriers) are used to ensure that the D impact of any potential missile does not prevent the system or equipment from carrying out its specified safety function. C. For all systems and equipment, 8ppropriate design procedures are used to ensure that the impact of any pote.ntial missile does not prevent the conduct of a safe plant shutdown, or prevent the plant from remaining in a safe shutdown condition. D. Refer to Sections (LATER) for rs.ated Standardized Functional Descriptitans. , I l l O\ Amendment D 3.5-6 September 30, 1988 i
i; CESSAR8!=nemu
- )
__ . __ . j O 1 MEFR811Eqtfl.J.QR SEQTl&2.al. j 1
- 1. " Plant Design Against MissiAes," ' ANSI /ANS-58.1. (DRAFT -
Formerly ANSI N177-1974.) j
- 2. A. Amirikan, " Design of Protective Structuree " Report ,No.
NT-3726, Bureau of Yhrds and Docks, Dept < of the Navy, August 1950. 3 .. " Structural Analysis and Design of Nuclear Power Plant l Facilitiese " Menual No. 58,,7' Chapter 6, American Society of Civil.Enginuers, 1980.
- 4. Stephenson, A. E., " Full Scale Torr. ado-Missile Impact D (
Tests," EPRI NP-440, July: 1977, Prepared for the Electric Power ReeGarch Instit.ute by Sandia tiationa). Laboratories. S. Cottrell, W. 9 and A.- W. Savolainen, "U.S. Reactor Containment Technology," ORNL-NSIC-5, Vol. 1, Chapter 6, Oak Ridge National lanoratory.
- 6. Williamson, R. A. and R. R. Alvy, " Impact Effects of f~ Fragments Striking Structural Elements," Holmes and Narver, I Inc. , Revised !!ovember, 1973. ;
}
)
J \ Amendment D 3.5-7 September 30, 1988
- - - - _ _ _ - - - - - - - _ _ _ - - - _ - - - - - - - - - - - - - - - - - - - __ - - - - - - - - - ----.--___-------J
~ .
5 -f CESSdRMuir: cum-l i c IABLE_3.5-1 l (Sheet 1 of 2) ) K]NETIC ENERGY CF POTENTIAL. M1551LES II) Initial Kinetic W'eight l Item (E) Energy (ft-lb)_ (1 b) _ Impact Section i
- 1. Reactor Vessel Closure Head Nut 1,706 100 Annular Ring, 00 = 10.125" ,
ID = 6.9" i Closure Head Nut ; and Stud 5,226 577 Solid Circle, 6.75" Diameter j Control Rod Drive 57,600 1100 1.875" Diameter Solid Circle Assembly within a Concentric 7" E i Diameter by .109" Wall Shroud
- 2. Steam Generator l b Primary Manway 71 4.25 Solid Circle, 1.5" Diameter Stud and Nut Secondary Handhole 7 1.15 Solid Circle,- 0.75" Diameter Stud and Nut Secondary Manway 7 3.36 Solid Circle,1.25" Diameter Stud i
L 3. Pressurizer Safety Valve 15 3.7 Solid Circle, 1.25" Diameter Flange Bolt Lower Temperature 288 3 Edge of Solid Disk 2.75" Element Diameter and 0.5" Thick Manway Stud and Nut 71 4.25 Solid' Circle, 1.5" Diameter
- 4. Reactor Coolant Pump and Piping Temperature Nozzle 1,095 8 Edge'of Solid Disk 2.75" With RTD Assembly Diameter and 0.5" Thick Amendment E December 30, 1988
7 I CESSAR naincma l 9 TABLE 3.5-1 (Cont'd) , l (Shcet 2 of 2) KINETIC ENERGY OF POTENTIAL MISSILES fI) 1 l Initial Kinetic Weight item ( ) Energy (f t lb)_ _(lbj_ Impact Section Surge and Spray 277 3.75 Edge of Solid Disc 2.75" Piping Thermowells Diameter and 0.5" Thick l with RTD Assembly l Reactor Coolant Pump Edge of Solid Disk 2.75" Thermowell with RTD 1,095 8 Diameter and 0.5" Thick l E O I I i I i II6TES: (1) All dimensions, weights and kinetic energies are typical values. (2) All materials are steel. Amendment E December 30, 1988
1 CESSAR nainemou l
.j
( 1 3.6 ?_Rj2TECTION AGAINST DYNAMIC EFFECTS ASS _OCIATED_WITH KIE POSTULATED RUPTURE 0F PIPING l Protection of vital equipment is achieved primarily by separatien J of redundant safe shutdown systems and by separation' of j high-energy pipe lines from safe shutdcwn systems, which are required to be functional following specific pipe rupture evcnts.' This redundancy and separation results in a design which requires { very few special protective features (such an whip restraints and. :] jet deflectors) to ensure safe . shutdown capability -following a ! postulated high-energy line break. ]
, J Separation is maintained by barriers such as the containment 1 I l
secondary shield wall, refueling cavity 9/all' and certain primary 1 l auxiliary building walls and tunnels or by physical distance. J Loadings and jet zonca of influence are calculated using i methodology described in Section 3.6.2. .l 3.6.1 POSTULATED PIPING FAILURES IN FLUID SYSTEMS I 3.6.2. 1 Desi m Basis j I Most systems and components outside. containment requi' red for safe ( plant shutdown are located in the primary auxiliary building. E The primary auxiliary building is divided by a structural wall ' which serves as a barrier between redundant trains of safe shutdown systems and components. Each half of the- p'imary r auxiliary building is compartmentalized to sep' a rate redundant safe shutdown components to the extent practical. High-energy l p.1 ping systems located in the auxiliary building, which'are not required to be functional for safe shutdown such as cteam i generator blowdown downstream of containment isolation and i auxiliary steam, are routed primarily in designated pipe tunnels i in the secondary auxiliary building to provide s e p a r a t i o n ~l r c m j safe shutdown systems and components. The primary and seco d ry auxiliary buildings are separated by structural walls that provide physical barriers. l Systems and components.inside containment, which are required to I be functional for safe plant shutdown, are protected from postulated pipe failure dynamic effects primarily by ..c'eparation and barriers. The secondary shield wall serves as a barrier, between the reactor coolant loops and the containment liner. Thef refueling cavity walls, the operating floor, and the secondary. shield wall provide separation betwe ' the reactor coolant loops. 1 The steam generators and pressurize._ are cnclosed in cavities which also provide separation. q O Amendment 'E 3.6-1 December 30, 1988
1 CESSAR naiemou , I 9 lI l Main steam and main feedwater (downcomer and economizer) linesq outside containment are separated from essential systems end j ; components by virtue of the plant arrangement that p3?ces these' i 1 lines in a special anclosure constructed along the roof of the primary auxiliary building. The two sides; of this chase adjacent to the primary auxiliary building (floor and one walli are 5 Seismic Category I concrete walls. The 0ther twu sides of this chase are metallic siding adjacent to the environment. In the event of a high-energy line break in this enclosure, the metallic' ciding wi,11 blow out and no significant pressurization will i result. The essential portions of these systems ( containihent isolation valves) are located in the MSIV rocms. These rooms are I separated from all c>ther essential systems and components by Srismic tategory I concrete slabs 7-d walls. The MSSS design cont,ists of two steam generators per unit, which facilitated separation of redt.adant systems and components inside containment. Otter than for the safety injection system ' components, which must circulate ::o' sling water to the vessel, the cngineered safety features are generally located outside the accendary shield wall. The safety injection system pipes and j cables, which terminate inside the secondary shield wall, are routed outside the secondary chield wall to the extent practical ! to avoid postulated hazards. Most of the main steam and l feedwater piping insido containment is , located at higher i olevations, and the postulated dynamic effects are separated from I safe shutdown systens and components by distance and configuration. Table 3.6-1 provides a list of plant fluid systems that contain high- and moderate-energy piping in the auxiliary and containment buildings. j 1 Table 3.6-2 provides a list of the systems that a m required for E I safe shutdown or to support safe shutdown. High- and moderate-energy pipe failure locations are postulated ; as described in Section 3.6.2. Each postulated rupture location ! is evaluated for its effect on safe shutdown systems and'
.:oinpenents required follot, Jing the specific pipe failure event.
3.6.1.1.1 High-Energy Piping Systems 1 A high-energy pipe failure is postulated in branches or piping ) runs larger than one inch nominal diameter and which operate j during normal plant conditions with high energy fluid. j Included in this category are fluid systems or portions of fluid - systems which are pressurized above atmospheric pressure during j 9' 4 Amendment E ! 3.6-2 December 30, 1988
CES$AR!a h m (
)
normal plant operation and which, in addition, operate during normal plant conditions and where either or both of the following . are met: l A. Maxima?a operating teinperature exceeds 200 *F, or ' B. Maximum operating pressure exceeda 275 psig. i Fluid piping systems that qualify as high-energy for only short f portigos or their operational period are considered moderate- i energy systems if the portion of their operational period within i the pressure temperature specified above for high energy fluid systems is less than cither of the following: A. One percent of the nornal operating life span of the plant, l or E , l B. Two percent of the time period required to accomplish its system design function. In analyzing the effects of a high-energy pipe failure, the consequences of pipe whip, water _ spray, jet impingement, I flooding, compartment pressurization, and environmental l conditions are considered. 1 3.6.1.1.2 Moderate-Energy Piping Systems ] l l A moderate-energy pipe failure is postulated .in branches or t piping runs larger than one inch nominal diareter and which operato during normal plant conditions with moderate-energy ! fluid. Included in this category are fluid' systems or portions of fluid syctems which are pressurized above atmospheric pressure during normal plant operation and which, in additions operate during> ; normal plant conditions and where both'of the'following are met: t A. Maximum operating temperature is 200*F or less, and B. Maximum operating pressure is 275 psig ot'less. In analyzing the effects of a moderate-energy failure, the consequences of water spray, j,et impingement, flooding, compartment pree,surization, and environmental ' conditions are considered. l l I i Ameqdment E 3.6-3 December 30, 1988 v _ .
y . - - _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ I l i CESSAR naincrnou 1 ei Descrjati on ! 2.6.1.2 A listing of the high-energy lines inside the containment is i given in Table 3.6-3. A listi.ng of high-energy lines outside the j , c o ntai ntr ent is given in Table 3.6-4. Since the Turbine and I Fadwaste Buildingn contain no safety-related equipment, high- l energy line breakn in those bufldings are generally excluded from I this table I 1 j Essential Eystems are those systems that .a r e needed to safely snut down the reacter or mitigate the consequences of a pipe break for a given postulated piping failure. However, depending ] upon the type and location of a postulated pipe break, certain j safety equipment may not be classified as essential for that j particular event. q l The essential systems which are to be protected from the effects j cf postulated piping failures are identified below. These ecsontial cystems were selected for each postulated break to satisfy the protection criteria given in the introduction to i Section 3.6.
- n. The following systems, or portions of these systems, are required to mitigate the consequences of postulated breaks of high-energy reactor coolant pressure boundery piping that result in a loss-of-coolant-accident (LOCA) assuming a loss l of offsite power. [
- 1. Reactor Protective System.
- 2. Engineered Safety Features Actuation System.
- 3. Safety Injection System.
- 4. Containment Spray System.
- 5. Class lE Electrical Systems, AC and DC (including !
switchgear, batteries, and distribution systems). I
- 6. Diesel Generator Systems, including Diesel Generator Starting, Lubrication, and Combustion Air Intake and I Exhaust Systems.
- 7. Diesel Fuel Oil Storage and Transfer System.
l l 8. Hydrogen RecoJnbiner Gystems, i
- 9. Control Building HVAC Syste'm.
i l l i Amendment E ! 3.6-4 December 30, 1988 1 1
CESSAR 8laibion O(,/
- 10. component Cooling Water System (portions required for operation of other listed systems). l i
- 11. Essential Spray Pond System. !
1
- 12. l'uel Building HVAC System. f i
l
- 13. Diesel Generator Building HVAC Systen'. l
- 14. Main Control Board (See Tables 7 3-2 and 7.3-14 for I systems required). l
- 15. Containment Isolation Systems: i l
- a. Penetration assemblier. j i
- b. Isolation valves
- c. Equipment hatch
- d. Luergency personnni hatch j
- e. Personnel lock j D f. Liner plate l
- g. Test connections 1 I
- h. Piping betwcen penetration assemblies and isolation valves,. E
- 16. Ex-core Neutron Monitoring System i
- 17. Safety-related Radiation Mon'itors (refer to S'cttion e )
11.5).
- 18. Shutdown Cooling Systou.
- 19. Essential Chilled Water System. l
- 20. Safety Depressurization System.
- 21. Emergency Feedwater System.
- 22. Air Coolers
- 23. Station Service Water System n
't
\ Amendment E 3.6-5 December 30, 1988'-
CESSAR Rfei?icariou B. The foll-owing systems, or portions of these systems are required to mitigate the consequences of postulate:d breaks in high-energy secondary pressure boundary piping (main cteam, main feedwater, blowdown, or emergency feedwater) assuming a loss of offsite power.
- 1. Reactor Protective System. l
- 2. Engineered Safety Features Actuation System.
- 3. Safety Injection System.
1
- 4. Containment Spray System (for breaks inside the {'
containment only).
- 5. Main Steam and Feedwater System (from unaf fected steam 1 i generator out to the containment isolation valt es . {
including the atmospheric steam dump, steam supply tu ! the turbine-driven emergency feedwater pump, and the steam generator blowdown line).
- 6. Shutdown Cooling System.
- 7. Class 1E Electrical Systems, AC and DC (including Switchgear, Batteries and Distribution Systems). j
- 8. Diesel Generator System, including Diesel Generator (
Starting, Lubrication and Combustion Air Intake and Exhaust Systems. , l
- 9. Diesel Fuel Oil Storage and Transfer System. E
- 10. Component Cooling Water System (portions required for operation of other listed systems).
- 11. Essential Spray Pond System.
- 12. Control Building HVAC System.
- 13. Fuel Building HVAC System.
- 14. Maiti Control Board (Sea Tables 7.3-2 and 7.3-14 for systeins required) .
- 15. Essential CDilled Water System.
- 16. Containment Isolation Systems:
O Amendment E 3.6-6 December 30, 1988
CESSAR8Hibia rm l ?
- a. Penetration assemblies l
b .. Isolation valves
- c. Equipment hatch
- d. Emergency personnel hatch l
- e. Personnel lock l
- f. Liner plate !
! I
- g. Test connections {
i
- h. Piping betweEn penetration assemblies and I isolation valves.
- 17. Diesel Generator Building HVAC System.
- 18. Emergency Feedwater System.
l l 19. Condensate Storage System. ( 20. Ex-core Neutron Monitoring.
- 21. Station Service Water System.
E
- 22. Air Coolers.
C. For other postulated breaks not included in items A and B above, systems must not be affected such that any break, evaluated on a case-by-case basis, violates the following criteria: l
- 1. The pipe break must not cause a reactor coolant, steam, l 'or feedwater line break.
l
- 2. The function of safety systems required to perform protective actions to mitigate the consequences of the postulated break must be maintained.
- 3. The ability to place the plant in a safe shutdown condition must be maintained.
A systematic approach of multidiscipline analyses of safety-related and associated systems was initiated to verify cempliance with design criteria, interface requirements, and safety design bases. On-going reviews of the model identified
, potential hazards and highlighted susceptibility of essential s
l Amendment E ' 3.6-7 December 30, 1988 w-__--___--__-___-_-_-_______ _ _ _ _ _ _ _ _ _ _ _ _ _ - _
CESSAR 8lninCAT10N O equipment from common modo failure, as well as provided an independent method of verification of ths availability of essential equipment required to mitigate the consequences of postulated accident scenarios. The resolution of comments raised . during these reviews resulted in changes to equipment layout, design of pipe whip and jet impingement restraints, upgrading some non-seismic supports to seismic, and the addition of curbs, drains, and other flood mitigation measures. The potential effects of flooding as a consequence of a pipe break, or leakage or through-wall cracks (as defined in Sections 3,6-.2.1.2.C and 3 . 6. 2.1. 2 . D) were analyzed on a case-by-case basis to ensure that the operability of safety-related equipment would not be impaired. An analysis of the potent.ial effects of missiles is discussed in Section 3.5. ] The potential environmental e f fects of steam on essential systems are discussed in Section 3 11. In general, because of the protective measures of redundancy and separation between systems and trains, the consequential effect of the transport of steam I will not be sufficient to impair the ability of the essential E system to shut down the plant and/or mitigate the consequences of the given accident of interest. , ( There are no high-energy lines in the vicinity of the control room. As such, there are no effects upon the habitability of the control room by pipe break either from pipe whip, jet impingement, or transport of uteam. Further discussion on control room habitability systems is provided in Section 6.4. 3.6.1.3 Safety Evaluation By means of design features such as separation, barriers, and pipe whip and jet impingement restraints, all of which are discussed below, the effects of pipe break will not damage essential systems to an extent that would impair their design function nor affect necessary cc gonent operability. The ability of specific safety-related systems to withstand a single active failure concurrent with a postulated event is discussed in the failure modes and ef f ects analyses provided in Sections 5.4.7, 6.2, 6.3, 6.5, 7.2, 7.3, 8.3, 9.2 and 10.4. A. Separation The plant arrangement provides separation to the extent practical between redundant safety systems in order to Amendment E 3.6-8 December 30, 1988 [ -
CESSARnnhow O prevent loss of safety function as a result of hazards different from those for which the system is required to function, as well as for the specific event for which the system is required to be functional. Separation between redundant safety systems with their related auxiliary supporting features is the basic protective measure. In general, layout of the facility followed a multistep process to ensure adequate separation.
- 1. Safety-related systems are located away from most high-energy piping.
- 2. Redundant (e.g., "A" and "B" trains) safety systems and q subsystems are located in separate compartments. l l
- 3. As necessary, specific components are enclosed to maintain the redundancy required for those systen.s that must function as a consequence of specific piping failure events. I I
B. Barriers-Shields and Enclosures Protection requirements are met through the protection afforded by the walls, floors, columns, abutments, and foundations in many cases. Where adequate protection does not already exist due to separation, additional barriers, deflectors, or shields are provided as necessary to meet the functional protection requirements. Where compartments, barriers, and structures are required to provide the necessary protection, they are designed to withstand the combined effects of the postulatcd failure plus normal E operating loads plus earthquake leadings. C. Piping Restraint Protection Where adequate protection does not already exist due to separation, barriers or shields, piping restraints are provided as necessary to - meet the functional protection requirements. Restraints are not provided when it can be shown that the pipe break would not cause unacceptable damage to essential systems or components. The design criteria for pipe whip restraints are given in Section 3.6.2.3.3. D. Facility Response Analyses l An analysis of postulated pipe break events was performed to
- identify those safety-related systems and components that Amendment E 3.6-9 December 30, 1988
CESSARE!aiNma O, provide protective actions required to mitigate, to acceptable limits, the consequences of the postulated pipe l break event. l Whenever the separation inhcrent in the plant design is { shown to assure the functional capability of the safety 1 systems required following a postulated pipe break event, no l additional protective measures are required for that event, an additional considerations of break type, location, orientation, restraints, an other protective measures are not required. When necessary, additional protective ) measures are incorporated into the design, as required, to j assure the functional capability of safety systems required i following the postulated pipe break event. l l In conducting the facility response analyses, the following criteria are utilized to establish the integrity of systems j and components necessary for safe reactor shutdown and j maintenance of the shutdown condition: j 1
- 1. Offsite power is assumed to be unavailable if an .
automatic turbine generator trip or automatic reactor trip is a direct consequence of a postulated piping j failure. i i
- 2. In addition to the postulated pipe failure and its !
accompanying effects, a single active component failure is assumed in the systems required to mitigate the consequences of the postulated piping failure. The single active component failure is assumed, except as noted in Section 3.6.1.3.D.4.
- 3. Each high- or moderate-energy fluid system pipe failure E is considered separately as a single postulated initiating event occurring during normal plant conditions.
- 4. Where a postulated piping failure is assumed in one of two redundant trains of a system that is required to operate during normal plant conditions as well as to shut down the reactor, single failures that prevent the functioning of the other train or trains of that system are not assumed, provided the system is designed to Seismic Category I standards, is powered from offsite and onsite sources, and is designed, constructed, operated, and inspected to quality assurance, testing, and inservice inspection standards appropriate for nuclear safety class systems.
Amendment E 3.6-10 December 30, 1988
af CESSAR E!airicuiu 3, - 0 l l S. All available systeins and compone;nts , including I I noh-Seismic Category I and those actuated by operator actions, may be employed to mitigate the consequences of a postulated piping failure. In judging the availability of such systems and components, account is taken of the postulated failure and its direct ! consequences, such as unit trip and loss of offsite j power, and of the aasumed single active component j failure and its direct consequences. The feasibility 1 of carrying out operator actions is based on a minimum j of 30 minutes delay responding to alarm indication and { adequate access to equipment being available for the 1 proposed actions. (Access to the containment post-LOCA is not assumed.) ,
- 6. Piping systems containing high-energy flutds are i designed so thal the effects of a single postulated pipe break cannot, in turn, 'cause failures of other pipes or components with unacceptable consequences. )
- 7. For a postulated pipe failure, the escape of steam, water, and heat from structures enclosing the i d
f3 high-energy fluid containing piping does not preclude: I U a. Accessibility to surrounding areas important to the safe control of reactor operations. ; I'
- b. Habitability of the control room.
- j. c. Ability of instrumentation, electric power supplies, and component s and controls to initiate, actuate, and complete a safety action. (A loss of E redundancy is permissible, but not the loss of i function.) I
\
The design criteria define acceptable types of isolation for I safety-related elements and for high-energy line from similar elements of the redundant train. Separation is accomplished by: A. Routing the two groups through separate compartments, or B. Physically separating the two groups by a specified minimum i distance, or C. Separating the two groups by structural barriers. The design criteila assure that a postulated failure of a high-energy line or a safety-related element cannot take more : , O than one safety-related train out of service. The failure of a U component or subsystem of one train may cause failure of another Amendment E 3.6-11 December 30, 1988
l CESSARnnLmw l [ O portior, of the same train; for example, a "B" train high-energy ) pipe may cause failure of a "B" train electrical tray, but not l l failure of any "A" train component. The capability to shut the l plant down safely under such a failure will therefore remain j intact. Given the separation criteria above., and the pipe break criteria in Section 3.6.2.1.2, the effects of high-energy pipe breaks are not analyzed where it is determined that all essential systems, components, and structures are sufficiently physically remote from a postulated break in that piping run. 3.6.2 DETERMINATION OF BREAK LOCATIONS AND DYNAMIC EFFECTS ABSOCIATED WITH THE POSTULATED RUPTURE OF PIPING l l Described herein are the design bases for locating breaks and cracks in piping inside and outside containment, the procedure used to define the thrust at the break location, the jet impingement loading criterits, ad the dynamic response models and results. 3.6.2.1 Criteria Used to Define Break and Crack Locations and Conficturations 3.6.2.1.1 General Requirements O Postulated pipe ruptures are considered in all plant piping systems and the associated potential for damage to required systems and components is evaluated on the basis of the energy in the system. System piping is classified as high-energy or moderatt.-energy, and postulated ruptures are classified as i circumfarential breaks, longitudinal breaks, leakage cracks, or ) througn-wall cracks. Each postulated rupture is considered separately as a single postulated initiating event. For each postulated circumferential and longitudinal break, an ! evaluation is made of the effects of pipe whip, jet impingement, compartment pressurization, environmental conditions, and flooding. Also, if required to demonstrate safe plant shutdown, i an internul fluid system load evaluation is performed on the effects of fluid forces on components within or bounding the fluid system. For each postulated leakage crack, an evaluation is made of the effects of compartment pressurization, environmental conditions and flooding. For each postulated through-wall crack, an evaluation is made of the effects of environmental conditions and flooding. The evaluation of the requirea systems and components demonstrate that the protection requirements of Section 3.6.1 are met. O Amendment E 3.6-12 December 30, 1988 l d
CESSAR EMWicarieu i O Irrespective of the fact that the criteria in Section 3.6.2 may not require specific breaks, if a structure outside containment separates a high-energy line from an essential component, that ! separating structure is designed to withstand the consequences of the pipe break in the high-energy line that produces the greatest effect at the structure.
]
3.6.2.1.2 Postulated Rupture Descriptions A. Circumferential Break A circumferential break is assumed to result in pipe severance with full separation of the two severed pipe ends unless the extent of separation is limited by consideration of physical means. The break plane area (A is assumed perpendicular to the longitudinal axis of the*) pipe , and is assumed to be the cross-sectional flow area of the pipe at . the break location. The break flow area (A from each of 1 the broken pipe segments for a circamferentk)al break, with I full separation of the two broken pipe segments, is equal to j the break plane area (A . The break flow area, discharge ! coefficient and dischaig)e correlation are substantiated l analytically or experimentally. tO B. Longitudinal Break A longitudinal break is assumed to result in a split of the pipe wall along the pipe longitudinal axis, but without l severance. The break plane area (A g ) is assumed parallel to l the longitudinal axis of the pipe and equal to the cross-sectional flow area of the pipe at the break location. E The break flow area (A ) is equal to the break plane area q (A ). The break is absumed to be circular in shape or elfiptical (2D x D/2) with its long axis parallel to the axis. The discharge coefficient and any other values used for the area or shape associated with a longitudinal break
.are substantiated analytically or experimentally.
C. Leakage Crack A leakage crack is assumed to be a crack through the pipe wall where the size of the crack and corresponding flow rate are determined by analysis and a leak detection system, y as described in Section 3.6.3. I 1 D. Through-Wall Crack A through-wall crack is assumed to be a circular orifice g through the pipe wall of cross-sectional flow area equal to N the product of one-half the pipe inside diameter and 1 ona-half the pipe wall thickness. I Amendment E .I 3.6-13 December 30, 1988 ): l
CESSAREnacum O 4 3.6.2.1.3 Piping Evaluated for Leak-Before-Break A leak-before-break evaluation is performed for Class 1 piping with a diameter of ten inches or greater (i.e., the reactor , coolant system (RCS) main loop piping, surge line, shutdown l cooling and safety injection lines) and for the main steam line in order to eliminate the dynamic effects of pipe rupture from the design basis. The' evaluation is intended to meet the requirements of 10 CFR 50, Appendix A, General Design Criterion , (GDC) 4. The evaluation is performed using the guidelines of ( NUREG 1061, Vol. 3 (Reference 1) as described in Section 3.6.3. I 3.6.2.1.4 Piping Other than Piping Evaluated for Leak-Before-Break l i This section applies to all high- and moderate-energy piping other than that whose dynamic effects due to pipe breaks are eliminated from the design basis by leak-before-break evaluation, i as identified in Section 3.6.2.1.3. 3.6.2.1.4.1 Postulated Rupture Locations A. Class 1 Piping Ruptures, as specified in Items D and E below, are postulated to occur at the following locations in each piping network designed in accordance with the rules of the ASME Boiler and Pressure Vessel Code, Section III (Reference E
- 2) for Class 1 piping:
- 1. the terminal ends of the pressurized portions of the l network. l
- 2. intermediate locations where S from equation (10) exceeds 2.4S m' 3 .. intermediate locations where U exceeds 0.1.
where, as defined in Subarticle NB-3650, O j Amendment E 3.6-14 December 30, 1988
CESSAR !!nh.ou O S = primary-plus-secondary stress-intensity range under the combination of loadings for which either iAvel A or Level B service limits have been specified, as calculated from equF ~ . - (10) . S = allowable Stress-intensity value. U = the cumulative usage factor. B. Class 2, Class 3, or Seismically Anal, zed ANSI B11.1 Piping Ruptures, as specified in Items D and E below, are postulated to occur at the following locations in each piping network designed in accordance with the rules of the ASME Boiler and Pressure Vessel Code, Section III, (Reference 2) for Class 2 and Class 3 piping, or with the rules of the ASME Code for Pressure Piping, B31, Power Piping, ANSI /ASME B31.1-1983 (Reference 3) for seismically analyzed ANSI B31.1 piping:
- 1. the terminal ends of the pressurized portion of the network, and 9 2. either E
- a. intermedie.e tocations of potential high stress or fatigue such as pipe fittings, valves, flanges and welded-on attachments, or
- b. intermediate locations where the stress, S, exceeds 0.8(X + Y).
Where, as defined in Subarticle NC-3650, S = stresses under the combination of loadings for which either Level A or Level B service limits have been specified, as calculated from the sum of equations (9) and (10). X = equation (9) Service Level B allowable stress. 1 = equation (10) allowable stress. O Amendment E 3.6-15 December 30, 1988 __ f
CESSARR%Aum O C. Non-Seismically Analyzed ANSI B31.1 Piping l Ruptures, as specified in Items D and E, are postulated to occur at the following locations in each ASME Code for Pressure Piping, B31, Power Piping, ANSI /ASME B31.1-1983 (Reference '3 ) piping network that is not seismically analyzed.
- 1. at terminal ends of the pressurized portions of the network, and
- 2. at each intermediate location of potential high stress or fatigue, such as pipe fittings, valves, flanges, and welded-on attachments.
D. Break Locations Both circumferential and longitudinal breaks are post"11ated to occur, but not concurrently, in all high-energy piping systems at the locations specified In Items A, B, or C, except as follows: 1, Circumferential breaks are not postulated in piping E runs of a nominal dianeter equal to or less than 1 inch.
- 2. Longitudinal breaks are not postulated in piping runs of a nominal diameter less than 4 inches.
- 3. Longitudinal breaks are not postulated at terminal [
ends. l 4 I Amendment E 3.6-16 December 30, 1988 k
CESSARnahms O
- 4. Only one type of break !n postulated at locations where, from a detailed stress analysis, such as finite-element analysis, the state of stress can be used to identify the most probable type. If the primary plus secondary stress in the axial direction is found to be at least 1.5 times that in the circumferential direction for the most severe loading combination association with Level A and Level B service limits, then only a circumferential break is postulated.
Conversely, if the primary plus secondary stress in the circumferential direction is found to be at least 1.5 t'mes that in the axial direction for the most severe
.toading combination associated with Level A and Level B service limits, then only a longitudinal break is postulated.
- 5. Circumferential and longitudinal breaks are not postulated at locations where the requirements of Item F are satisfied.
- 6. Circumferential and longitudinal breaks are not postulated at locations where the criterion in Item E.2 is used.
E. Crack Locations
- 1. Through-Wall Cracks Through-wall cracks are post:ulated in all high-energy and moderate-energy piping systems having a nominal diameter greater than 1 inch at the locations specified E in A, B or C, except that through-wall cracks are not postulated at locations where:
- a. For Class 1 piping, the calculated value of S, as defined in Item A, is less than one-half the limits of Item A.2.b.
- b. For Class 2, Class 3 or seismically analyzed ANSI B31.1 piping, the calculated values of S a;s defined in Item b is less than one-half the limits of Item B.2.b.
- c. The requirements of Item F are satisfied,
- d. The criterion in 2. below is used.
O l Amendment E 3.6-17 December 30, 1988
~
CESSARnasem O
- 2. Leakage Cracks A leakage crack is postulated in place of a circumferential break, or longitudinal break, or through-wall crack, if justified by an analysis performed on the pipeline in accordance with the requirements of Section 3.6.3.
F. Piping Near Containment Isolation Valves 1 Ruptures are not postulated between the containment wall and I the inboard or outboard isolation valves in piping, which is l designed in accordance with the rules of the ASME Boiler and Pressure Vessel Code, Sectiort III (Reference 2), and which meets the fol' lowing additional requirements:
- 1. The limits for postulating intermediate rupture locations, as specified in Item A.2.b for Class 1 piping and Item B.2.b for Class 2 and 3 piping, are not exceeded in that portion of piping.
- 2. Following a postulated pipe break of high-energy piping beyond either isolation valve, the stresses in the piping from the containment wall, to and including the length of the isolation valve, are maintained within Levol C Service Limits as specified in the ASME Boiler aMd Pressure Vessel Code, Section III, (Reference 2).
- 3. The design and in-seNice inspection requirements, as specified in MEB 3-1 (Reference 4}, are satisified.
E -
- 4. The containment isolation valves arc appropriately qualified to assure that operability and leak tightness are maintained when subjected to ~ any combination of Imdings, which may be transmitted to the valves from p otuleted pipe breaks beyond the valves.
3.6.2.1.4.2 Postulated hupture Configurations A. Break Configurations Where the postulated break location is at a tee, elbow, or the following pipe locations, the configurations and types of breaks are determined as follows:
- 1. Without the benefit of a detailed stress analysis, the following are assumed:
O Amendment E 3.6-18 December 30, 1988
CESSAREnnne- {
]
- a. Circumferential breaks are postulated to occur individually at each tee or elbow pipe-to-fitting ,
weld where the criteria in Section 3.6.2.1.4.1, ! Item C are exceeded, and longitudinal breaks postulated to occur individually on each side of the tee or elbow at its center and oriented i perpendicular to the plane of the fitting.
- b. At a branch run connection, a circumferential break is postulated at the branch run-to-main run weld, or the branch run-to-fitting weld, and the break plane area (A- is assumed to be the 1 cross-sectional flow a9e)a of the branch.
- c. At a welded attachment (lug, stanchion, etc.) a longitudinal break is postulated at the centerline of the walded attachment.with an area equal to the 4 pipe sarface area that is bounded by the attachment weld.
- d. At an axisymmetric pipe location, such as a .
reducer, circumferential and longitudinal breaks 1 (g are postulated at each pipe-to-fitting weld where I Q the criteria in Section 3.5 2.1.4.1, Item C are exceeded. Longitudinal breaks are oriented. to produce out-of-plane bending of the piping ! configuration. !
- 2. Alternatively, where a detailed stress analysis or test is performed, the results are used to predict the most probable rupture location (s) and type of break.
B. Crack Configurations E At a postulated leakage crack or through-wall crack location, the orifice is assumed to be located concurrently at each and every point about the circumference of the pipe, unless otherwise substantiated. 3.6.2.1.5 Details of Containment Penetrations Details of containment penetrations are discussed in Sections 3.8.1 and 3.8.2. l l Amendment E 3.6-19 December 30, 1988
E i CESSAR8!ainem O 3.6.2.2 Analytical Methods to Define Forcing Functions and Response Models 3.6.2.2.1 Piping Evaluated for Leak-Before-Break There are no forcing functions or response models for the reactor coolant loop, surge line, shutdown cooling line, safet" injection line and main steam line based upon elimination t_ dynamic offects by leak-before-break evaluation. 3.6.2.2.2 Analytical Methods to Define Forcing Functions and Responso Models for Piping Excluding that Evaluated for Leak-Before-Break This section applies to all high-energy piping other than that whose dynamic effects due to pipe breaks are eliminated from the design basis by leak-before-break evaluation. 3.6.2.2.2.1 Determination of Pipe Thrust and Jet Loads l A. Circumferential Breaks Circumferential breaks are assumed to result in pipe severance and separation amounting to at least a one-diameter lateral displacement of the ruptured piping sections, unless physically limited by piping restraints, structural members, or piping stiffness. The dynamic force of the jet discharge at the break location is based on the effective cross-sectional flow area of the pipe and on a calculated fluid pressure as modified by an analytically determined thrust coefficient. Limited pipe displacement at l the break locations, line restriction flou limiters, i positive pump controlled flow, and the absence of energy E reservoirs are taken into account, as applicab?e, in the reduction of the jet discharge. Pipe whip 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. B. Dynamic Force of the Fluid Jet Discharge The dynamic force of the fluid jet discharge is based on a circular break area equal to the cross-sectional flow area of the pipe at the break location and on a calculated fluid pressure modified by an analytically determined thrust coefficient, as determined for a circumferential break at the same location. Line restrictions, flow limiters, O Amendment E 3.6-20 December 30, 1988
CESSAREMec-r) U positive pump-controlled fl ow , and the absence of energy reservoirs are taken into account, as applicable, in the reduction of jet discharge. Piping movement is assumed to occur in the direction of the jet reaction, unless limited by structural members, piping restraints, or piping stiff. ness. C. Pipe Blowdown Force and Wave Force The fluid thrust forces that result from either postulated circumferential or longitudinal breaks, are calculated using a simplified one step forcing function methodology. This methodology is based on the simplified methods described in ANSI 58.2 (Reference 5) and in Reference 6. When the simplified method discussed above leads to impractical protective measures, then a more detailed computer solution which more accurately reflects the postulated pipe rupture event is used. The computer solution is based on the NRC's computer program developed for calculating two-phase blowdown forces (Reference 7). O D. Evaluation of Jet Impingement Effects Q E Jet impingement force calculations are performed only if structures or components are located near postulated high energy line breaks and it cannot be demonstrated that failure of the structure or component will not adversely affect safe shutdown capability. 3.6.2.2.2.2 Methods for the Dynamic Analysis of Pipe Whip Pipe whip restraints usually provide clearance for thermal expansion during normal operation. If a break occurs, the restraints or anchors nearest the break are designed to prevent unlimited movement at the point of break (pipe whip) . A finite difference model will be used to analyze simplified models of the local region near the break. Displacements and strains of the pipe and restraint will be estimated using a power law moment curvature relationship. A. Finite Difference Analysis A 'inite difference formulation specialized to the case of a straight beam and neglecting axial inertia and large deflection ef fects is used for the analysis of pipe whip. The dynamic analysis is performed by direct numerical time integration of the equations of motion presented in Appendix Os 3.6A. Amendment E 3.6-21 December 30, 1988
I CESSAR8!nh . O 3.6.2.2.2.3 Method of Dynamic Analysis of Unrestricted Pipes The impact velocity and kinetic energy of unrestricted pipes is calculated on the basis of the assumption that the segments at each side of the break act as rigid-plastic cantilever beams , subject to piecewise constant blowdown forces. The hinge l location is fixed either at the nearest restraint or at a point ) determined by the requirement that the shear at an interior l plastic hinge is zero. The kinetic energy of an accelerating I cantilever segment is equal to the difference between the work ( done by the blowdown force and that done on the plastic hinge. l The impact velocity V is found from the expression for the j kinetic energy: j KE = (1/2) M egYI where M is the mass of the single degree of freedom dynamic model of9the cantilever. The impacting mass is assumed equal to j "eg* f 3.6.2.3 Dynamic Analysis Methods to Verify Integrity and i Op_erability 3.6.2.3.1 Pipe Whip Restraints and Jet Deflectors for G) Piping Evaluated for Leak-Before-Break E i There are no pipe whip restraints and jet deflector for the j reactor coolant loop, surge line, shutdown cooling line, safety l injection line and main steam line based upon elimination of l dynamic effects due to pipe breaks by leak-before-break evaluation. 3.6.2.3.2 Pipe Whip Restraints and Jet Deflectors for Piping Other than that Evaluated for Leak-Before-Break This section applies to pipe whip restraints for all piping other than that whose dynamic effects due to pipe breaks are eliminated from the des'.gn Sasis by leak-before-break evaluation. 3.6.2.3.2.1 General Description of Pipe Whip Restraints Pipe whip restraints are provided to protect the plant against the effects of whipping during postulated pipe break. The design of pipe whip restraints is governed not only by the pipe break blowdown thrust, but also by functional requirements, deformation limitations, properties of whipping pipe and the capacity of the support structure. Typically, a pipe whip restraint consists of a ring around the pipe and components supporting the ring from Amendment E 3.6-22 December 30, 1988
CESSAR unirlCATION the supportiw structure. The diameter of the ring is established s easidering the pipe diameter, maximum thermal movement of pipe, thickness of insulation and an additional 1/2-inch for installation tolerance. The restraint is designed for the impact force induced by _he maximum possible initial gap between the ring and the pipe. This impact energy is usually too high for an elastic restraint system or support structure to absorb. Therefore, energy absorbing measures designed by the energy balance approach (impact energy + external work = internal energy of pipe restraint system), are provided. 3.6.2.3.2.2 Pipe Whip Restraint Components Pipe whip restraints typically consist of the following components: A. Energy Absorbing Members Members that are under the influence of impacting pipes I (pipe whip) absorb energy by significant plastic deformations (e.g., rods, and crushable honeycomb material). B. Non-Energy Absorbing Members Those components which form a direct link between the pipe i and the structure (e.g., ring and components other than i energy absorbing members). 1 C. Structural Attachments E Those fasteners which provide the method of attaching connecting members to the structure or the ring (e.g., weld attachments). D. Structural Components Steel and concrete support structures which ultimately carry the restraint load. Design criteria are specified in Section 3.8. 3.6.2.3.2.3 Design Loads Restraint design loads, the reactions, and the corresponding deflections are established using the criteria delineated in Section 3.6.2.2.2. n v Amendment E 3.6-23 December 30, 1988
CESSAR 88Wnce O 3.6.2.3.2.4 Allowable Stresses The allowable stresses are as follows: A. For energy absorbing members: 0.95 F with 0.5 cu strain for steel in tension, where the dynam5N yield strength, F is considered 15% higher than the static yield strength F Yd and e is the ASTM specified minimum elongation for the rEE1 given as a percentage with 50% usable strain for crushable energy absorption material in compression as determined by dynamic testing. B. For non-energy absorbing member: 1.6 times the AISC - allowable stress, but not to exceed 0.95 F . for bending and 0.55 F for shear where F d is considereY 10% higher than l F foY compression membeds, the allowable stress is 0.9 } t5faes the buckling stress F CRC as follows: ' F =
/ xF
- CRC a where 5/3 = r, awe r bound factor of safety in AISC for compression stress F
a
= AISC allowable compression stress DIF = Dynamic increase factor = 1.1 C. For structural attachments and structural components -
allowable stresses will be the same as described in Sections 3.8.3 and 3.8.4. 3.6.2.3.2.5 Design Criteria E The unique features in the design of pipe whip restraint components relative to the structural steel design are geared to the loads used and the allowable stresses. These are ao follows: A. Energy-absorbing members are designed for the restraint reaction and the corresponding deflection established according to the pipe size and material and the blowdown force using the criteria delineated ic Section 3.6.2.2. B. Non-ennrgy-absorbing members and their attachments are designed for 1.25 times the restraint reaction to ensure that the required deflection occurs in the energy absorbing members and that the connecting members remain elastic. O Amendment E 3.6-24 December 30, 1988
CESSAR En@ication ( C. Structural components and their attachment to the building structure are designed for the restraint reaction with an appropriate dynanic load factor. The dynamic load factor is calculated from the resistance forcing function provided by the (LATER) computer program as described in Appendix 3D. All essential components are evaluated for jet impingement and pipe whip effects using a dynamic or an equivalent static analysis of testi g to demonstrate either the functional capability and/or 'erability in addition to the structural integrity of the co ' ent. 3.6.2.3.2.6 Mater dis The materials used are as follows: i A. For energy-absorbing members: ASTM A-1093 Grade B7 or l equivalent for tension rods, and crushable honeycomb made of stainless steel for compression. B. For other components: ASTM A-538, ASTM A-572 Grade 50, and ASTM A-36. Charpy tests will be performed on steels (~% subjected to impact loads and lamination tests are performed on steels subjected to through thickness tension. 1.6.2.3.2.7 Jet Impingement Shields Protection from jets is provided by using separation and redundancy, as described in Section 3.6.1, in lieu of jet i shields. E 3.6.2.4 Guard Pipe Assembly Desian Criteria Guard pipes to limit pressurization effects in the containment penetration area will not be used. 3.6.3 LEAR-BEFORE-BREAK EVALUATION PROCEDURE l This section describes Leak-Before-Break (LBB) analysis to all applicable piping. LBB analysis is used to eliminate, from the l structural design bases the dynamic effects of double-ended I guillotine breaks and equivalent- longitudinal breaks for an applicable piping system. 3.6.3.1 Applicability of LDB Pipina evaluated for LBB is first shown to meet the applicability requirements for NUREG 1061, Volume 3. The piping is designed to O meet the requirement to be not particularly susceptible to Amendment E 3.6-25 December 30, 1988
CESSAR nahnos 9 failure from the effects of corrasion, water hammer or low- and high-cycle fatigue, or degradation or failure of the piping from indirect causes. 3.6.3.2 Leakage Crack Location A survey of the piping is performed to determine the locations of highest stress loading and coincident poorest material properties. All base metal, weld materials, heat affected zones in the vicinity of the terminal ends, and all intermediate elbow locations are considered. 3.6.3.3 Leak Detection There are two major aspects to leak rate based on crack detection in addition to the crack opening size; leak detection capabili.ty, and flow rate correlation for leakage through a crack. 3 6.3.3.1 Leak Detection System A leak detection system is required by Regulatory Guide 1.45, Reference 8, capable of detecting a leakage rate of less than 1.0 gpm from the primary system. NUREG 1061, Volume 3, recommends a . safety margin of ten on the leak detection system. Diverse measurement means are required, including water inventory monitoring, sump level and flow monitoring, and measurement of airborne radioactive particulate or gases. Leak detection system requirements to support the LBB analysis for main steam line piping are (IATER) . 3.6.3.3.2 Flow Rate Correlation The other major aspect of crack detection based on the leak rate,I namely the flow rate correlation for leakage through a given crack size, cannot be predicted precisely. Variables such as E surface roughness of the side walls of the crack, the nonparallel relationship of the side walls due to th elongated crack shape, and possibly zigzag tearing of the material during crack formation all introduce uncertainties in defining an exact flow rate correlation. The leakage rate required to be detectable is 1.0 gpm. The licensing guidelines (NUREG 1061, Volume 3) recommend a factor of 10 on that leakage rate, for conservatism; therefore, a crack length which leaks 10 gpm at normal operating conditions is ! selected as the design leakage crack. Using recent work by EPRI I (Reference 10), the leakage rate per square inch of leak area in the 10 gpm leakage rate range is computed to be approximately 250 0 Amendment E 3.6-26 December 30, 1988 )
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CESSAR 8a!hiou O gpn per square jnch for ths primary system for the range of pip 1 Lizes of interest. The crack opening area corresponding to the 10 gpm rate is found to be'O.4 square inch. This crack opening area is used to' determine the length of the detectable crack for stability evaluation. 3.6.3.4 pereeninc of Leukace Crack Bizes Using EPRI/GE Estimation Schenc Prior to detailed calculations of through-wall leakage cracks and corresponding margins on loads and crack sizes, a preliminary scoping evaluation is performed. In this part, all possible lonations in the pipling evaluated are screened to identify the most critical candidates for detailed study. The screening study is performed using the EPRI/CE estimation scheme (Reference 11) for the determination of crack opening areas using elastic plastic fracture mechanics methods, and thn C-E developed JEST computer program fc, the leckage rates through cracks. l This estimation procedure ,a used to compare the severity of hypothesized flaws in all piping locations in order to reduce the number of cases to be subjected to detailed analysis. The l p procedure also provides an estimate of the leakage crack length. l t l for input to the detailed finita-alement &nalysis discussed in l Section 3.6.3.1.6. 3.6.3.5 Material Properties For the main coolant loop, the hot and cold leg piping,raterial is SA-516 GR, 70 7 All hot- and cold-leg pipe-to-pipe welds and the pipe-to-reactor vessel, steam generator and reactor coolant pump safe end welds are carbon steel. .E All main loop component nozzles are SA508 CL 2 or 3 carbon steel or SAS41 CL1, 2 or 3. l The surge line is SA351 GR CF8M stainless steel, resulting in l bimetallic safe end welds. The shutdown cooling line is (LATER). 1 The direct vessel safety injection line is (LATER). The main { steca line is (LATER). The detailed analysis of cracks in pipe welds requires l consideration of the properties of the pipe and the weld materials. Previous work by C-E has shown that a conservative bounding analysis results when the . material stress-strain properties of the base metal (lower yield) and the fracture properties of the weld (lower toughness) are used for the entire structure, (Reference 12). This material representation is used for all analyses. The tensile (stress-strain) curves and the J D vs. aa curves are requircG for each material type. l Amendment. E 3.6-27 December 30, 1988 ____-- - ____- - ______ _ a
! CESSAR nai"icariou s 3.6.3.6 Leakage crack Size Determinate _o_Il j It is necessary that hypothesized through-wall cracks open significantly to allow detection by normal leakage monitoring ; under normal full power loadings. The crack opening area is i calculated at all The chosen locations for normal operating l conditions, since *ecse are the 7st prevelent conditions during ] plant operating itte, when leakage is to be detected. A range of j i crack sizes is considered to evaluate the sensitivity to leakage rate assumptions. The result of the analysis of several assumed l crack sizes is then used to estimate the leakage length. Under i normal operation loading, almost all the material in the pipe is in the clastic range. With a small amount of plasticity, this method is very accurate in estimating the crack size. The crack j opening area is computed on the inside and outside of the pipe { and the smaller area is used. j The finite-element analysis is performed for those cracks determined . by the estimation scheme to be of the assured leak j detection cize. The analysis is also performed for a crack twice 1 that length subjected to the same loading conditions. These two analysis results enable a curve of crack opening area vs. crack length to be drawn. From this curve, the length of the assured leak detectable crack is determined. 3.6.3.7 C_ogtputation of J-Intecral Values 3.6.3.7.1 Range of Crack Sizes i l The crack lengths estimated using the EPRI/GE estimating scheme E : (Reference 11) are input to the detailed stability analysis of i the through-wall cracks in the piping evaluated. The finite- ] clement analysis is carried out for the estimated leakage crack j size and twice that length. This procedure, therefore, considers the stability of a range of crack lengths for all locations selected for the analysis. 3.6.3.7.2 J-Integral J I The stability of through-wall cracks is evaluated using the J-integral technique. The J-integral is determined in the finite-element analysis for pressure, normal operation, and safe ! i shutdown earthquake loadings for two different crack lengths for each geometric model. For the margin on loads evaluation, the j i J-integral for the estimated leakage crack size is also evaluated for /2 x (Pressurc+NOP+SSE) loads. A typical J-integral vs. load ) increment curve is shown in Figure 3.6-1. Ol I Amendment E 3.6-28 December 30, 1988 l I l j
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O 3.6.3.8 . Stab.ility Evaluation The stability of the cracked pipes ~is assessed by comparing the J-integral value due to the applied loads on the pipe to the material ci'ack resistance. The stability cri?arion for.ductila crack extension employed is: , if J-applied < JIC "" * "I' # p h applied < hmaterial then crack stability is assured. The slope of the J-integral vs. a (crack length) curve for each location is obtained by fitting a polynominal curve through the computed J-integral values. By plotting J-integral vs. dJ/da,, then both parts of the stability criterion can be evaluated simultaneously. The evaluations are performed for the locations chosen to envelop , a ?.1 limiting cases. The pipes with the leakage crack length ! subject to loads of ff~ x (P+NOP+SSE) and the pipes with crack ( length twice the leakage crack length with loads of (P+NOP+SSE) are demonstrated to have significant margln between the material curve and the loading curve, indicating that all pipe locations E satisfy the LBB crack stability criteria. 3.6.3.9 Results Class 1 pipitig with a diameter of ten inches or greater and main steam line piping meet all the criteria for application of the leak-before-break according to NUREG 1061, Volume 3. Specifically, these criteria require that: A. Cracks which are assumed to grow through the pipe wall leak significantly while remaining stable. The amount of leakage is detectable with a safety margin of at least a factor of
- 10. u B. Cracks of the length that leak at the rate in A. can withstand normal or.eration plus safe shutdown earthquake loads with a safety factor of at least /2.
C. Cracks twice as long as those addressed in B. will remain stable when subjected to normal operation plus safe shutdown earthquake loads. Is Amendment E 3.6-29 December 30, 1988
1 l C E S S A R n aiMe m on It_EjgRENCEB E FOR SEC.TJON 3.6
- 1. " Evaluation of Potential for Pipe Breaks," NUREGo1061, ]
Vol. 3. I
- 2. ASME Boiler and Pressure Vessel Code, Section III, Nuclear Power Plant Components, Class 1, 2 or 3, 1986 Ettition. l 1
- 3. ASME Code for Pressure Piping, B31, Power Piping, ANSI /ASME B31.1-1986 Edition.
- 4. USNRC Branch Technical Position MEB 3 Postulated Rupture' Locations in Fluid System Piping Inside and Outside j Containment, attached to Standard Review Plan 3.6.2, July j 1981.
- 5. Americ6n National Standard Design Basis for Protection of Light Water Nuclear Power Plants Against the Effects of ,
Postulated Pipe Rupture, ANSI /ANS 58.2-1988, h i A
- 6. R. T. Lahey, Jr. and F. J. Moody., " Pipe Thrust and Jet Loads," The Thermal Hydraulics of aj3cilina Water NilcJ ear Reactor, Section 9.2.3, pp. 375-409, Published by American ,
Nuclear Society, Prepared by the Division of Technical Information United States Energy Research and Development Administration, 1977.
- 7. RELAP 4/ MOD 5, Computer Program User's Manual 098. 02G-5.5.
- 8. USNRC Regulatory Guide 1.45 " Reactor Coolant Pressure I Boundary Leakage Detection Systems," May 1973. 1
- 9. NUREG/CR-1319, " Cold Leg Integrity Evaluation," Battelle ;
Columbus Laboratories, February 1980. j l
- 10. PICEP: Pipe Crack Evaluation Program, EPRI NP 3596-SR, l August, 1984. ]
- 11. "An Engineering Approach for Elastic-Plastic Fracture l Analysis," EPRI N P2 9'31, by V. Kumar, M. D. German, C. F.
Shih, July 1981.
- 12. " Analysis of Cracked Pipe Weldments," EPRI NP-LOS7, February !
1987. O! Amendment E 3.6-30 December 30, 1988
e , z
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1 ICATION O' i i TABLE 3 5-1 f HIGH- AND H0DERATE-ENERGY FLUh1 SYSTEMS l ((i_sh-Energy Fluid Systems Main Steam Main feedwater Steam Generator Blowdown Auxiliary Steam j Reactor Coolant Surge Line Safety Injection { Chemical and Volume Control 1 1 I Moderate-Energy FLuldJstems i Chemical and Volume Control . Component Cooling l Steam Benerator Blowdown j Safety Injection /Snutdown Cooling i Contaircent Air Monitoring / Sampling j O. Diesel Fuel .0il Station Service Mater Fire Protection E H,, N,, and CO i Sbrvite Water 2 ( Spent Fuel Pool Cooling and Cleanup- ) Station Heating i Containment Spray i Drain Chilled Water ] ' Breathir>" air Instrunien+ Air I Service A. 1 I l O
- Systems shown are either totally or partially high-energy.
Amendment E December 30, 1988 1 E_ ___ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ - . _ _ _ . _ - - - _ _ __ _ .--
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4 1 TABLE L_6d { SYSlEMS REqqlR_fD FOR SAFE SHUTDOWN ANDf0R 70 NITIGATE THE CONSEQUENCES-l ! 0F A DESLQXrBASIS ACCIDENT
- j 1
Reactor Protective Reactor Coolant .
.. i Safety injection / Shutdown Coolinrj j Containment Spray .
Emergency Feedwater- l Component ~ Cooling Water ! l Station Service Water Auxiliary Power j Area Radiatior. ' Containment Isolation i Condensate Storage Battery and DC Distribution ; Diesel Generator j Diesel fuel Oil i Engineered Safety _ Features l Combustible Gas Control i Instrument and Control Power {~j , In-core Thermocouple , , l Ex-core Neutron Monitoring l Main Steam i Main feedwater Control Element Assembly Drive i i Control Room HVAC .; Diesel Generator Room HVAC ESF Switchgear Room HVAC E Station Service Water P ap' House HVAC Primary Containment Ventilation Safety Injection System fquipment Room HVAC , Essential Chilled Water ' Essential Spray Pond Main Control Board Safety Depressurization [ l '
- Systems listed are either totally or partially required for safe-shutdcwn.
Amendment E December 30, 1988
r DESIGN ' C Cllus C'@O*dM A E) ERCERTIFICATION ( i
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(v / TAB._LE 3.6-3 E HIGH-ENERGY LINES WITHIN CONTAINFiENT (LATER) i i i i l 1 l j% 5 ( 1
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Amendment E December 30, 1938
h O < 1500 - I 3 ( g 1000 - 5 4 E l E cc 8
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PR ESS. NO. OP. SSE / _,41 (PR EGS+NO.OP.+SSEJ
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i Amendment E i O) December 30, 1988 vu Figur.e - VARIATION OF J-INTEGRAL jfd / u[ WITH LOADS FOR A TYPICAL CASE 3.6- 1 i i}}