ML20090D289

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Nonproprietary RESAR-SP/90 Westinghouse Advanced PWR Module 4, Rcs
ML20090D289
Person / Time
Site: 05000601
Issue date: 06/30/1984
From:
WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML19273A237 List:
References
NUDOCS 8407180180
Download: ML20090D289 (400)
v • d • e


Text

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RESAR-SP/90 REACTOR COOLANT SYSTEM WESTINGHOUSE

_. ADVANCED PRESSURIZED WATER REACTOR I -

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WAPWR-RCS i JUNE, 1984 i 1393e:1d  ;

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TABLE OF CONTENTS 1;

Reference SAR Section Section Title Page Status

1.0 INTRODUCTION

AND GENERAL DESCRIPTION OF PLANT 1.1 -1 II

1.1 INTRODUCTION

1.1-1 II 1.2 GENERAL PLANT DESCRIPTION 1.2-1 II 1.2.2 Principal Design Criteria 1.2-1 II .

1.2.3 Plant Description 1. 2-1 II ~

1.2.3.2 Reactor Coolant System 1. 2-1 1 1.3 COMPARISON TABLES 1. 3-1 II

1. 3.1 Comparison with Similar Facility Designs 1. 3-1 II

.1.6 MATERIAL INCORPORATED BY REFERENCE 1.6-1 II 1.7 DRAWINGS AND OTHER DETAILED INFORMATION 1.7-1 II 1.7.1 Piping and Instrumentation Diagnms 1. 7-1 11 1.8 CONFORMANCE WITH THE STANDARD REVIEW PLAN 1. 8-1 II 2.0 SITE CHARACTERISTICS 2. 0-1 N/A 3.0 DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS 3.1-1 II 3.1 CONFORMANCE WITH NRC GENERAL DESIGN CRITERIA 3.1 -1 II 3.2 CLASSIFICATION OF STRUCTURES, COMPONENTS, AND SYSTEMS 3.2-1 II 3.2.1 Seismic Classification 3.2-2 II 3.2.2 System Quality Group Classification 3.2-2 II 3.2.3 Safety Classes 3.2-2 II 3.2.4 References 3.2-2 II 4.0 REACTOR 4.4-1 II 4.4 THERMAL AND HYDRAULIC DESIGN 4. 4-1 II 4.4.6.4 Digital Metal Impact Monitoring System 4. 4-1 1 4.4.6.5 Inadequate Core Cooling Instrumentation 4.4-2 I O

WAPWR-RCS 11 JUNE, 1984 1393e:1d

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TABLE OF CONTENTS (Cont) f i  :

Reference l SAR Section f i

O Section Title Pige. Status 5.0 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 5.1 -1 II  !

5.1

SUMMARY

DESCRIPTION 5.1-1 I [

b 5.1.1 Design Bases 5.1-1 I i

5.1.2 Design Description 5.1 -3 I .

5.1.3 System Components 5.1 -4 I  !

5.1.4 System Performance Characteristics S.1-7 I [

5.1.4.1 Reactor Coolant Flows 5.1 -7 I I 5.2 INTEGRITY OF REACTOR COOLANT PRESSURE BOUNDARY 5.2-1 I f 5.2.1 Compliance with Code and Code Cases 5.2-2 I f

, 5.2.1.1 Co;np11ance with 10CFR50.55a 5.2-2 I l

5.2.1.2 Applicable Code Cases 5.2-2 I O 5.2.2 Overpressure Protection 5.2-3 I 5.2.2.1 Design Bases 5.2-3 I  !

5.2.2.2 Design Evaluation 5.2-5 I 5.2.2.3 Piping and Instrumentation Diagrams 5.2-5 I i 5.2.2.4 Equipment and Component Description 5.2-5 I 5.2.2.5 Mounting of Pressure Relief Devices 5.2-6 I [

5.2.2.6 Applicable Codes and Classification 5.2-6 1 5.2.2.7 Material Specifications 5.2-6 I I 5.2.2.8 Process Instrumentation 5.2-6 I l 5.2.2.9 System Reliability 5.2-7 I  !

5.2.2.10 RCS Pressure Control During Low-Temperature 5.2-7 I l

Operation j 5.2.2.10.1 System Operation 5.2-7 I l 5.2.2.10.2 Evaluation of Low-Temperature Overpressure 5.2-8 1 j Transients - Pressure Transient Analysis 5.2.2.10.3 Operating Basis Earthquake Evaluation 5.2-8 I l 5.2.2.10.4 Administrative Controls 5.2-9 I l

WAPWR-RCS

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TABLE OF COMTENTS (Cont)

I i i j Reference f

(. SAR Section f L Section Title Pace Status f

i 5.2.3 Reactor Coolant Pressure Boundary Materials 5.2-12 I 5.2.3.1 Material Specifications 5.2-12 I 5.2.3.2 Compatibility with Reactor Coolant 5.2-13 I 5.2.3.2.1 Chemistry of Reactor Coolant 5.2-13 I .

l' 5.2.3.2.2 Compatibility of Construction Materials with 5.2-14 I Reactor Coolant 5.2.3.2.3 Compatibility with External Insulation and 5.2-15 I Environmental Atmosphere l 5.2.3.3 Fabrication and Processing of Ferritic 5.2-16 1 l Materials l ' 5.2.3.3.1 Fracture Toughness 5.2-16 I 5.2.3.3.2 Control of Welding 5.2-17 I [

5.2.3.4 Fabrication and Processing of Austenitic 5.2-17 I f Stainless Steel j 5.2.3.4.1 Cleaning and Contamination Protection 5.2-18 I l Procedures f

. I 5.2.3.4.2 Solution Heat Treatment Requirements 5.2-19 I

^

i l

5.2.3.4.3 Material Testing Program 5.2-19 I }

5.2.3.4.4 Prevention of Intergranular Attack of 5.2-19 I  !

Unstabilized Austenitic Stainless Steels f 5.2.3.4.5 Retesting Unstabilized Austenitic Stainless 5.2-23 I (

Steel Exposed to Sensitization Temperatures  !

i . 5.2.3.4.6 Control of Welding 5.2-24 I  !

5.2.4 Inservice Inspection and Testing of Reactor 5.2-26 I l Coolant Pressure Boundary 5.2.4.1 System Boundary Subject to Inspection 5.2-27 I 5.2.4.2 Arrangement and Accessibility 5.2-27 I j O  !

t MAPWR-RCS iV JUNE, 1984 f 13g3e:1d f i

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h TA8LE OF CONTENTS.(Cont)

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l 1 Reference SAR Section Section Title PAge. Status f i

5.2.4.3 Examination Techniques and Procedures 5.2-27 I l 5.2.4.4 Inspection Intervals 5.2-28 I  !

5.2.4.5 Examination Categories and Requirements 5.2-28 I

' 5.2.4.6 Evaluation of Examination Results 5.2-28 I  !

5.2.4.7 System Leakage and Hydrostatic Pressure Tests 5.2-28 I f

5.2.5- Detection of Leakage Through Reactor Coolant 5.2-28 I Pressure Soundary 5.2.5.1 Design Bases 5.2-29 I ,

5.2.5.1.1 Leakage Classification 5.2-29 I i

5.2.5.1.2 Limits for Reactor Coolant Leakage 5.2-29 I [>

t l 5.2.5.2 Identified Intersystem Leakage Detection 5.2-30 I  !

l 5.2.5.2.1 Description and Operation of Identified 5.2-31 I i

Leakage Detection System ,

'5.2.5.3 Unidentified Leakage Detection 5.2-33 I f l 5.2.5.3.1 Description and Operation of Main Unidentified 5.2-34 I j i Leak Detection Systems l l

e 5.2.5.3.2 Additional Unidentified Leakage Detection 5.2-38 I }

Methods l 5.2.5.4 Safety Evaluation 5.2-39 I  !

5.2.5.5 Tests and Inspections 5.2-39 I f 5.2.5.6 Instrumentation Applications 5.2-40 I 5.2.6 References 5.2-40 1  ;

5.3 REACTOR VESSEL 5.3-1 I i I

l-5.3.1 Reactor Vessel Materials 5. 3-1 I l

5. 3.1.1 Material Specifications 5.3-1 1 l

5.3.1.2 Special Processes Used for Manufacturing and 5.3-1 I l

Fabrication 5.3.1.3 Special Methods for Nondestructive Examination 5.3-2 1 5.3.1.3.1 Ultrasonic Examination 5.3-2 I t

WAPWR-RCS y JUNE, 1984 1393e:1d l

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TABLE OF CONTENTS (Cont) i r

Reference [

SAR Section j l

Section Title PaSe Status ,

l 5.3.1.3.2 ' Liquid Penetrant Examination 5.3-3 I l 5.3.1.3.3 Magnetic Particle Examination 5.3-3 I l c

O 5.3.1.4 Special Controls for Ferrit'e and Austenitic 5.3-4 I I l

l- Stainless Steels .

I 5.3.1.5 Fracture Toughness 5.3-4 I l 5.3.1.6 Material Surveillance 5.3-5 I I

5. 3.1. 6.1 Measurement of Integrated Fast Neutron 5.3-7 I f (E > 1.0 MeV) Flux at the Irradiation Samples ,

5.3.1.6.2 Calculation of Integrated Feit Neutron 5.3-11 I l (E > 1.0 MeV) Flux at the Irradiation [

Samples j 5.3.1.7 Reactor Vessel Fasteners 5.3-13 I j 5.3.2 Pressure-Temperature Limits 5.3-13 1  !

5.3.2.1 Limit Curves 5.3-13 I I 5.3.3 Reactor vessel Integrity 5.3-14 I 5.3.3.1 Design 5.3-14 I  :

5.3.3.2 Materials of Construction 5.3-16 I j 5.3.3.3 Fabrication Methods 5.3-17 I l 5.3.3.4 Inspection Requirements 5.3-17 I 5.3.3.5 Shipment and Installation 5.3-17 I  ;

5.3.3.6 Operating Conditions 5.3-17 I I ..3.3.7 Inservice Surveillance 5.3-20 I 5 5.3.4 References 5.3-21 I

5.4-1 II 5.4 COMPONENT AND SUBSYSTEM DESIGN l 5.4.1 Reactor Coolant Pump Assembly 5. 4-1 I f

5.4.1.1 Design Bases 5. 4-1 I i O  !

vi JUNE,1984 MAPWR-CCS {

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l TABLE OF CONTENTS (Cont) y r Reference j SAR Section Section Title Pace Status .

! 5.4.1.2 Pump Assembly Description 5.4-1 I  !

5.4.1.2.1 Design Description 5.4-1 I f

l 5.4.1.2.2 Description of Operation 5.4-2 1  !

5.4.1.2.3 Loss of Seal Injection 5.4-4 I .

l l- 5.4.1.2.4 Loss of Component Cooling Water 5.4-5 I l- 5.4.1.2.5 Backup Seal Injection Capability 5.4-5 I .

5.4.1.3 Design Evaluation 5.4-6 I l

5.4.1.3.1 Pump Performance 5.4-6 I 5.4.1.3.2 Coastdown Capability 5.4-8 I  !

5.4.1.3.3 Bearing Integrity 5.4-8 I I 5.4.1.3.4 Locked Rotor 5.4-9 I 5.4.1.3.5 Critical Speed 5.4-10 I ,

5.4.1.3.6 Missile Generation 5.4-10 1 5.4.1.3.7 Pump Cavitation 5.4-10 I  !

5.4.1.3.8 Pump Overspeed Considerations 5.4-10 I 5.4.1.3.9 - Antireverse Rotation Device 5.4-11 I .

l 5.4.1.3.10 Shaft Seal Leakage 5.4-12 I i 5.4.1.3.11 Seal Discharge Piping 5.4-12 I  !

5.4.1.4 Tests and Inspections 5.4-13 I l 5.4.1.5 Pump Flywheel 5.4-13 I l 5.4.1.5.1 Design Basis 5.4-13 I l I

5.4.1.5.2 Fabrication and Inspection 5.4-13 I 5.4.1.5.3 Material Acceptance Criteria 5.4-14 I  !

5.4.2 Steam Generators 5.4-15 I 5.4.2.1 Design 8ases 5.4-15 I 5.4.2.2 Design Description 5.4-16 1 {

5.4.2.3 Design Evaluation 5.4-18 I  !

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t TABLE OF CONTENTS (Cont) 1; Reference i SAR Section l Section Title Page, Status l [

5.4.2.3.1 Forced Convection 5.4-18 I  ;

5.4.2.3.2 Natural Circulation Elow 5.4-18 I i I

5.4.2.3.3 Mechanical and Flow-Induced Vibration 5.4-19 I Under Nonnal Operating Conditions  ;

5.4.2.3.4 ' Allowable Tube Wall Thinning Under Accident 5.4-20 I .

Conditions  !

1 5.4.2.4 Steam Generator Materials 5.4-21 I i 5.4.2.4.1 Selection and Fabrication of Materials 5.4-21 1  !

5.4.2.4.2 Steam Ger.arator Design Effects on Materials 5.4-23 I l

5.4.2.4.3 Compatibility of Steam Generator Tubing with 5.4-23 I Primary and Secondary Coolants i 5.4.2.4.4 Secondary Side Cleaning Provisions 5.4-25 1 5.4.2.5 Steam Generator Inservice Inspection 2.4-25 I [

5.4.2.6 Quality Assurance 5.4-26 I  ;

5.4.3 Reactor Coolant Piping 5.4-27 I f 5.4.3.1 Design Bases 5.4-27 I l 5.4.3.2~ Design Description 5 4-28 I  ;

5.4.3.3 Design Evaluation 5.4 32 I  !

5.4.3.3.1 Material Corrosion / Erosion Evaluation 5.4-32 1 5.4.3.3.2 Sensitized Stainless Steel 5.4-32 I j 5.4.3.3.3 Containment Control 5.4-32 I i 5.4.3.4 Tests and Inspections 5.4-33 I i 5.4.4 Main Steam Line Flow Restrictors 5.4-33 I

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5.4.4.1 Design Basis 5.4-33 I i 5.4.4.2 Design Description 5.4-34 I f 5.4.4.3 Design Evaluation 5.4-34 I {

5.4.4.4 Inspections 5.4-34 I j 5.4.10 Pressurizer 5.4-35 I O r l

JUNE,1984 l WAPWR-RCS viii 1393e:1d l

e TABLE OF CONTENTS (Cont) 1; ,

Reference l G '

SAR Section i Section Title Pg.gi Status 5.4.10.1 Design Bases 5.4-35 I l 5.4.10.2 Design Description 5.4-36 I v 5.4.10.2.1 Pressurizer and Connected Piping 5.4-36 I 5.4.10.2.2 Pressurizer Spray and Relief Lit ? Instrumen- 5.4-37 I .

tation 5.4.10.3 Design Evaluation 5.4-38 I l 5.4.10.3.1 System Pressure Control 5.4-38 I

! 5.4.10.3.2 Pressurizer Level Control 5.4-38 I i 5.4.10.3.3 Pressure Setpoints 5.4-39 I l l

l 5.4.10.3.4 Pressurizer Spray 5.4-39 I  !

5.4.10.4 Tests and Inspections 5.4-40 I  ;

j. 5.4.11 Pressurizer Relief Discharge System 5.4-41 I {

5.4.11.1 Doign Bases 5.4-41 I J

5.4.11.2 tesign Description 5.4-42 I l 5.4.11.3 Design Evaluation 5.4-45 I .

5.4.11.4 Instrumentation Requirements 5.4-47 I  !

5.4.11.5 Inspection and Testing Requirements 5.4-47 I 5.4.12 Valves 5.4-48 I I 5.4.12.1 Design Bases 5.4-48 I 5.4.12.2 Design Description 5.4-48 I f

5.4.12.3 Design Evaluation 5.4-49 I  ;

5.4.12.4 Tests and Inspections 5.4-49 I 5.4.13 Safety and Relief Valves 5.4-50 I I

5.4.13.1 Design Bases 5.4-50 I I 5.4.13.2 Design Description 5.4-50 I l 5.4.13.3 Design Evaluation 5.4-52 I "

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O l-WAPWR-RCS ix . LUNE, 1984 1393e:1d

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TA8LE OF CONTENTS (Cont)

O  ;

\  !

Reference l D

V Section Title Page SAR Section Status  !

5.4.13.4 Tests and Inspections 5.4-52 I q '5.4.14 Component Supports 5.4-53 I l 5.4.14.1 Design Bases 5.4-53 I I 5.4.14.2 Description 5.4-53 1 5.4.14.2.1 Reactor Pressure Vessel 5.4-54 I t 5.4.14.2.2 Steam Generator 5.4-54 I  ;

5.4.14.2.3 Reactor Coolant Pump 5.4-55 I {

5.4.14.2.4 Pressurizer 5.4-55 I  ;

5.4.14.2.5 Control Rod Drive Mechanism (CRDM) Supports 5.4-56 I l 5.4.14.2.6 Displacer Rod Drive Mechanisms (DRDM) Supports 5.4-57 I 5.4.14.3 Evaluation 5.4-57 I  !

b 5.4.14.4 Tests and Inspections 5.4-58 I  ;

5.4.15 Reactor Vessel Head Vent System 5.4-58 I l 5.4.15.1 Design Basis 5.4-58 I ,

l 5.4.15.2 Design Description 5.4-59 I  ;

5.4.15.3 Design Evaluation 5.4-60 I 5.4.15.4 Inspection and Testing Requirements 5.4-61 I ,

5.4.15.5 Instrumentation Requirements 5.4-61 I r 5.4.16 References 5.4-61 I  ;

6.0 ENGINEERED SAFETY FEATURE MATERIALS 6. 0-1 N/A  !

7.0 INSTRUMENTATION AND CONTROLS 7 . 0-1 N/A t 8.0 ELECTRIC POWER 8.0-1 N/A  !

9.0 AUXILIARY SYSTEMS 9 . 0-1 N/A 10.0 STEAM AND POWER CONVERSION SYSTEM 10.0-1 N/A [

11.0 RADIDACTIVE WASTE MANAGEMENT 11.0-1 N/A 12.0 RADIATION PROTECTION 12.0-1 N/A l 13.0 CONDUCT OF OPERATIONS 13.0-1 N/A O '

MAPWR-RCS x JUNE,1984 '

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I TABLE OF CONTENTS (Cont)  !

Reference SAR Section l Section Title Pgit Status i

14.0 INITIAL TEST PROGRAM 14.0-1 N/A l

, 15.0 ACCIDENT ANALYSES 15.0-1 11 l 15.0.1 General 15.0-1 II 15.0.2 Classification of Plant Conditions 15.0-1 II .

j i 15.0.2.1 Condition I - Normal Operation and Operational 15.0-2 II

['

Transients i 15.0.2.2 Condition II - Faults of Moderate Frequency 15.0-4 II 15.0.2.3 Condition III - Infrequent Faults 15.0-6 II 15.0.2.4 Condition IV - Limiting Faults 15.0-7 II I 15.0.3 Optimization of Control Systems 15.0-8 II I 15.0.4 Plant Characteristics and Initial Conditions 15.0-9 II I Assumed in the Accident Analysis

- 15.0.4.1 Design Plant Conditions '5.0-9

. 11 l

15.0.4.2 Initial Conditions 15.0-9 II

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15.0.4.3 Power Distribution 15.0-10 II l

15.0.5 Reactivity Coefficients Assumed in the 15.0-11 II Accident Analyses j l

15.0.6 Rod Cluster Control Assembly Insertion 15.0-12 II i Characteristics f 15.0.7 Trip Points and Time Delays to Trip Assumed 15.0-13 II [

in Accident Analyses i I

15.0.8 Instrumentation Drif t and Calorimetric 15.0-14 II  !

Errors - Power Range Neutron Flux f 15.0.9 Plant Systems and Components Available for 15.0-15 II Mitigation of Accident Effects f 15.0.10 Fission Product Inventories 15.0-16 11 i

15.0.10.1 Inventory in the Core 15.0-16 II f

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O I i xj .lVNE, 1984 {

WAPWR-RCS 1393e:1d

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l l

i TABLE OF CONTENTS (Cont) j O, i i

Reference SAR Section O Section Title Pace Status i

15.0.10.2 Inventory in the Fuel Pellet Clad Gap 15.0-16 II l 15.0.10.3 Inventory in the Reactor Coolant 15.0-16 II  !

'O 15.0.11 Residual Decay Heat 15.0-17 II 15.0.11.1 Total Residual Heat 15.0-17 II .

i 15.0.12 Computer Codes Utilized 15.0-17 , II 15.0.12.1 FACTRAN 15.0-17 II j 15.0.12.2 LOFTRAN 15.0-18 II l 15.0.12.3 TWINKLE 15.0-19 II f

15.0.12.4 THINC 15.0-19 II l 15.0.13 References 15.0-19 II 15.3' OECREASE IN REACTOR COOLANT SYSTEM FLOWRATE 15.3-1 I (

d 15.3.1 Partial Loss of Forced Reactor Coolant 15.3-1 I Flow [

15.3.1.1 Identification of Causes and Accident 15.3-1 I l Description f 15.3.1.2 Analysis of Effects and Consequences 15.3-2 I  ;

15.3.1.3 Conclusions 15.3-4 I [

15.3.2 Complete Loss of Forced Reactor Coolant' Flow 15.3-5 I

, l 15.3.2.1 Identification of Causes and Accident 15.3-5 I l 1 Description l 15.3.2.2 Analysis and Effects of Consequences 15.3-6' I

{

15.3.2.3 Conclusions 15.3-7 I j 15.3.3 Reactor Coolant Pump Shaft Seizure (Locked 15.3-7 I l Rotor)

  • O' 15.3.3.1 Identification of Causes and Accident 15.3-7 I Description 15.3.3.2 Analysis of Effects and Consequences 15.3-8 I i O
  • l r
WAPWR-RCS xii JUNE,1984 f 1393e
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!. I TABLE OF CONTENTS (Cont) l ,

i i; Reference l l

SAR Section i l Status Section Title Pg_qe, l 15.3.3.2.1 Method of Analysis 15.3-8 I l 15.3.3.2.2 Evaluation of the Pressure Transient 15.3-9 I 15.3.3.2.3 Evaluation of Departure from Nucleate 15.3-10 1  !

Boiling (DNB) in the Core During the .

Accident i 15.3.3.2.4 Film Boiling Coefficient 15.3-10 I 15.3.3.2.5 Fuel Clad Gap Coefficient 15.3-10 I l 15.3.3.2.6 Zirconium - Steam Reaction 15.3-11 I f 15.3.3.2.7 Results 15.3-12 1 15.3.3.3 Radiological Consequences 15.3-12 1  !

15.3.3.3.1 Analytical Assumptions 15.3-13 I 15.3.3.3.1.1 Source Term Calculations 15.3-13 I i 15.3.3.3.1.2 General Parameters Used in the Analysis 15.3-14 1 15.3.3.3.1.3 Identification of Leakage Pathways and 15.3-14 I Resultant Leakage Activity 15.3.3.3.2 Identification of Uncertainties and Con- 15.3-14 I

)

servative Elements in the Analysis  ;

15.3.3.3.3 Conclusions 15.3-15 I f 15.3.3.3.3.1 Filter Loadings 15.3-15 I i l

15.3.3.3.3.2 Doses to Receptor at the Exclusion 15.3-15 I Area Boundary and Low Population Zone j Outer Boundary 1 15.3.4 Reactor Coolant Pump Shaft Break 15.3-16 1 15.3.4.1 Identification of Causes and Accident 15.3-16 I  !

l Description f i

15.3.4.2 Conclusion 15.3-17 I 15.3.5 References 15.3-17 I (

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TABLE OF CONTENTS (Cont)

N i

Reference SAR Section O Section Title Pace Status l

l 15.4 REACTIVITY AND POWER DISTRIBUTION ANOMALIES 15.4-1 II l

15.4.4 Startup of an Inactive Reactor Coolant Pump 15.4-1 I  ;

at an Incorrect Temperature i 15.4.4.1 Identification of Causes and Accident 15.4-1 I -

l

! Description ,

15.4.4.2 Analysis of Effects and Consequences 15.4-2 I (

l 15.4.4.2.1 Method of Analysis 15.4-2 1 15.4.4.2.2 Results 15.4-3 1 15.4.4.3 Conclusions 15.4-4 1 15.4.5 References 15.4-4 I ,

15.5 INCREASE IN REACTOR COOLANT INVENTORY 15.5-1 I l 15.5.1 Inadvertent Operation of ECCS During Power 15.5-1 I l Operation 15.5.1.1 Identification of Causes and Accident 15.5-1 I ,

Description ,

15.5.1.2 Conclusions 15.5-2 1 15.5.2 Chemical and Volume Control System Mal- 15.5-2 I function that Increases Reactor Coolant Inventory 3 l

15.6 DECREASE IN REACTOR COOLANT INVENTORY 15.6-1 II  :

15.6.1 Inadvertent fspening of a Pressurizer Safety 15.6-1 I or Relief Valve l 15.6.1.1 Identification of Causes and Accident 15.6-1 I Description 15.6.1.2 Analysis of Effects and Conclusions 15.6-2 I 15.6.1.2.1 Method of Analysis 15.6-2 I l 15.6.1.2.2 Results 15.6-3 I O  :

xiv JUNE, 1984  !

MAPWR-RCS 1393e:Id l

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TABLE OF CONTC. HTS (Cont) x.'

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, y SAR Section I leslig m

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Title it

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15.6.1.3 % Conclusion s

'15.6-4 I 15.6'.2 =Pefarences.  : x,, . 10.6-4 I '

4 15A.1 s EENERAL ACCIDENT PARANETERS \\ 15.A-1 II  !

r - -

-1 .'15A.2 .; ' 0FFSITE RADIOLOGICAL CONSEQUENCES 3 15.A-1 II - l CALCULATIONAL MODELS ., j f 5 I 9

15A.2.4 .

Accident Release Pathways 15.A-2 II

'k 15A.2f.2 Single Megion Relear.e Model s 15.A-2 s II  ;

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15A.'2.3 ' yTwo-Region Spray Mode'l in x s 15.A-4 II  :

1 ~Q Containment ((004) ,

j 15A.2.4 > 15.A-5

} OffsiteThyr,oictosyCaiculationModel \"

II l

l s15A.2.5 Offsite Beta-Skir. Oose Calcula'tio'nal Model 15 '. A-6 II 15A.2.6 Of fsite hanna-BodV Dose Calt.alational Model 15.A-6 II l

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15A.3 Control Room Radiclogical Consequences 15.A-7 11  !

Calculational Models y g l, 4

L, 15A.3.1 s Integ' rated, Activi{.y in Corttrol Room s 15.'A-7 11 i 15A.3.2 . Integrated Activity Concentration in' 15.A-8 II lL ' " s n. .i l

hControlRoomfromSingle-RegionSystem ,

t Con.colRoomThyroid9csecaiculational

! , 15A.2.3 s 15.A-9 II i Model

\

15A.3.4 Control Room Beta-Skin. Dose Calculational 15. A-10 II

,- s 'Model U 15A.3.5 Control Rocm Gamma-Body Dose Calculation 15.A-11 II' f 15A.3.5.1 Model for Radiological Consequences Due to e 15.A-11 II  :

RadioactiveCloudExternqlt.othe ,

f Control Room {

15A.4 References 15. A-12 II i g

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Reference

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SAR Section O Section Title PLqe Status 16.0 TECHNICAL SPECIFICATIONS 16.0-1 N/A  !

! 17.0 QUALITY ASSURANCE 17.1-1 11 i l 17.1 QUALITY ASSURANCE DURING DESIGN AND 17.1-1 11 I

CONSTRUCTION t

17.1.1 References 17.1-1 II i I

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TABLE OF CONTENTS (Cont)

LIST OF TABLES i Title Page Number O 1.3-1 Design Comparisons 1.3-2

1. 6-1 Material Incorporated by Reference 1.6-2 1.7-1 Piping and histrumentation Diagrams 1.7-2 1.8-1 Stand trd Review Plan Deviations 1.8-2 1.8-2 CoMormance to US NRC Tiegulatory Guides Applicuble 1.0-3 to tie WAPWR RCS 3.1-1 GDC Apt li;able to RCS 3.1-2 3.2-1 Classif cation cf Structures, Systems and 3.2-3 Compoents for the Reactor Ccolant System 5.1-1 Notes to RCS Process Flow Diagram (Figure 5.1-1) 5.1-9 5.1-2 System Design and Operating Parar:eters 5.1-11 5.2-1 Primary and Auxiliary Components Typical Material 5.2-42 Specifications m 5.2-2 Reactor vessel Internals Material SDecifications 5.2-47 5.2-3 Recommended Reactor Coolant Water Chemistry 5.2-48 Swcifications 5.3-1 Reactor Vessel Quality Assurance Program 5.3-22 5.3-2 Reactor Vesul Design Parameters 5.3-24 5.4-1 Reactor Loalant Pumo Design Pa " meters, Model 100A 5.4-62 5.4-2 Reactor Coolant Pump Quality Assurance Program 5.4-64 5.4-3 Steam Generator Design Parameters 5.4-65 5.4-4 Steam Generator Quality Assurance Program 5.4-66 5.4-5 Reactor Coolant Piping Design Parameters 5.4-68 5.4-6 Reactor Coolant Piping Quality Assurance Program 5.4-69 5.4-7 Pressuri:er Design Parameters 5.4-70 5.4-8 Reactor Coolant System Design Pressure Settings 5.4-71 5.4-9 Pressurize- Relief Tank Design Parameters 5.4-72

, 5.4-10 Pressurizer Relief Discharge System Nondestructive 5.4-73 Testing Program WAPWR-RCS JUNE, 1984 1393e:1d xvii

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'l TABLE OF CONTENTS (Cont) l L LIST OF TABLES i

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i Number Title Pace O Reactor Coolant System Valve Design Parameters 5.4-74 i

5.4-11 l l 5.4-12 Reactor Coolant System Valves Nondestructive 5.4-75 l l

Examination Program '

5.4-13 Pressurizer Safety and Relief Valves Design 5.4-76 f Parameters .

f 5.4-14 Reactor Vessel Head Vent System Equipment 5.4-77 f

Design Parameters 15.0-1 Nuclear Steam Supply System Power Ratings 15.0-21 i l 15.0-2 Values of Pertinent Plant Parameters Utilized in 15.0-22 i Accident Analysis (ITDP) [

l 15.0-2a Values of Pertinent Plant Parameters Utilized in 15.0-23 l' l' Accident Analysis (Non-ITDP) l 15.0-3 Summary of Initial Conditions and Computer Codes 15.0-24 l Used 15.0-4 Trip Points and Time Delays to Trip Assumed in 15.0-27 j Accident Analyses I

'15.0-5 Determination of Maximum Overpower Trip 15.0-28 l Point - Power Range Neutron Flux Channel - l Based on Nominal Setpoint Considering Inherent f

Instrument Errors j

.15.0-6 Plant Systems and Equir:nent Available for 15.0-30 l

~ Transient and Accident conditions l 15.0-7 Fuel and Rod Gap Inventories, Core (Ci) 15.0-34 .

15.0-8 Reactor Coolant Iodine Concentrations for 15.0-35 i 1pci/ gram and 60pCi/ gram of Dose l

l~ Equivalent I-131  !

! 15.0-9 Reactor Coolant Noble Gas Specific Activity 15.0-36  !

Based on One Percent Defective Fuel l i I  !

MNE, N  ;

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LIST OF TABLES T EW!!!h.tt TitIe PD.11 r

15.0-10 Iodine Appearance Rates in the Reactor 15.0-37 l Coolant (Curies /Sec) l 15.3-1 Time Sequence of Events for Incidents which 15.3-18 f Result in a Decrease in Reactor Coolant j System Flowrate -

15.3-2 Sunraary of Results for Locked Rotor 15.3-20 1 Transients (Four Loops Operating Initially) 15.3-3 Parameters used in Evaluating the Radio- 1 5.3-21 ,

logical Consequences of a Locked Rotor [

, 1 Accident j 15.3-4 Radiological Consequences of a Locked 15.3-23 Rotor Accident t 15.4-1 Time Sequence of Events for Incidents which 15.4-5 j cause Reactivity and Power Distribution Anomalies l 15.6-1 . Time Sequence of Events for Incidents which 15.6-5 Cause a Decrease in Reactor Coolant f

Inventory- t 15A-1 Parameters Used in Accident Analysis 15. A-13  !

15A-2 Limiting Short-Term Atmospheric Dispersion Factors 15. A-14 f 3

-for Accident Analysis (s/m ),

15A-3 Dose Conversion Factors Used in Accident Analysis 15. A-15  :

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TABLE OF CONTENTS (Cont)

LIST OF FIGURES g i

Number Title  ;

1.2-1 Reactor Coolant System l 1.7-1 Flow Diagram Legend 5.1-1 Reactor Coolant System Process Flow Diagram ,

5.1-2 Reactor Coolant System Piping and Instrumentation l Diagram . [

5.3-1 WAPWR Reactor Vessel \

5.4-1 Model 100A Reactor Coolant Pump l 5.4-2 Reactor Coolant Pump Curve

-5.4-3 Steam Generator 5.4-4 Structural Broach Support Configuration .

I 5.4-5 Pressurizer 5.4-6 Pressurizer Relief Tank  ;

5.4-7 Pressurizer Safety and Relief Valve Piping ano i i Support Arrangement [

5.4-8 Reactor Vessel Supports [

5.4-9 Steam Generator Supports -

' i 5.4-10 Reactor Coolant Pump Supports j 5.4-11 Pressurizer Supports I t

15.0-1 Illustration of Core Thermal Limits and DNB l Protection (N Loop Operation)

[3 15.0-2 Doppler Power Coefficient Used in Accident V Analysis h i

15.0-3 RCCA Position vs. Time to Dashpot i

! 15.0-4 Normalized RCCA Reactivity Worth vs. Fraction l Insertion l L 15.0 Nonnalized RCCA Bank Reactivity Worth vs. Normalized ,

I Drop Time WAPWR-RCS xx JUNE,1984 l

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LIST OF FIGURES z i- ,

Number Title O ,

l-15.3-1 ' Flow Transients for 4 Loops in Operation, 2 Pumps  !

Coasting Down  !

15.3-2 Nuclear Power and Pressurizer Pressure Transients  !

for 4 Loops in Operation, 2 Pumps Coasting down ,

15.3-3 Average and Hot Channel Heat Flux Transients for 4 -

l Loops in Operation, 2 Pumps Coasting Down  !

15.3-4 DN8R Versus Time for 4 Loops in Operation, 2 Pumps j Coasting Down 15.3-5. Flow Transients for 4 Loops in Operation, 4 Pumps  ;

Coasting Down f l- 15.3-6 Nuclear Power and Pressurizer Pressure Transients for [

4 Loops in Operation, 4 Pumps Coastirg Down 15.3-7 Average and Hot Channel Heat Flux Transients for _;

4 Loops in Operation, 4 Pumps Coasting Down  !

15.3-8 DNBR Versus Time for 4 Loops in Operation, 4 Pumps Coasting down i 15.3-9 Flow Transients for 4 Loops in Operation, 1 Locked Rotor l 15.3-10 Peak Reactor Coolant Pressure for 4 Loops in Operation, [

i 1 Locked Rotor  ;

! 15.3-11 Average and Hot Channel Heat Flux Transients for 4 Loops l in Operation, 1 Locked Rotor f 15.3-12 Nuclear Power and Maximum Clad Temperature at Hot Spot  !

! Transients for 4 Loops in Open2 tion, 1 Locked' Rotor i 15.4-1 Improper Startup of an Inactive Reactor Coolant Pump i 15.4-2 Improper Startup of an Inactive Reactor Coolant Pump l

.15.4-3 Improper Startup of an Inactive Reactor Coolant Pump j f

(' 15.4-4 Improper Startup of an Inactive Reactor Coolant Pump  !

15.4-5 Improper Startup of an Inactive Reactor Coolant Pump lO  !

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TABLE OF CONTENTS (Cont) i O  !

I LIST OF FIGURES 1 i

i Number Title 15.6-1 Nuclear Power and DNBR Transients for Inadvertent Opening l of a Pressurizer Safety valve  :

15.6-2 Pressurizer Pressure 'ransients and Core Avg. Temp.  !

l ~ Transient for Inadvertent Opening of a Pressurizer l r

Safety Valve . I i

15.A-1 Release Pathways  !

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