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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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

i I

i L

TABLE OF CONTENTS (Cont) f i

Reference l

SAR Section f

O Section i

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 f

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

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 iii JUNE, 1984 1393e:1d

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f 1

TABLE OF COMTENTS (Cont)

I i

i j

Reference f

(.

SAR Section f

f L

Section Title Pace Status 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

l 5.2.3.2.1 Chemistry of Reactor Coolant 5.2-13 I

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

[

f 5.2.3.4 Fabrication and Processing of Austenitic 5.2-17 I

Stainless Steel j

5.2.3.4.1 Cleaning and Contamination Protection 5.2-18 I

l f

Procedures 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

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5.2.3.4.4 Prevention of Intergranular Attack of 5.2-19 I

f Unstabilized Austenitic Stainless Steels 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

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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

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

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

5.2.5.1.2 Limits for Reactor Coolant Leakage 5.2-29 I

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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

Leakage Detection System i

f

'5.2.5.3 Unidentified Leakage Detection 5.2-33 I

l 5.2.5.3.1 Description and Operation of Main Unidentified 5.2-34 I

j Leak Detection Systems l

i l

e 5.2.5.3.2 Additional Unidentified Leakage Detection 5.2-38 I

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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 l

5. 3.1.1 Material Specifications 5.3-1 1

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

l 5.3.1.3.2

' Liquid Penetrant Examination 5.3-3 I

Magnetic Particle Examination 5.3-3 I

l O

5.3.1.3.3 I

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

l c

l-Stainless Steels I

5.3.1.5 Fracture Toughness 5.3-4 I

I l

5.3.1.6 Material Surveillance 5.3-5 I

f

5. 3.1. 6.1 Measurement of Integrated Fast Neutron 5.3-7 I

(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 5.3.3.6 Operating Conditions 5.3-17 I

I

..3.3.7 Inservice Surveillance 5.3-20 I

5.3.4 References 5.3-21 I

5.4 COMPONENT AND SUBSYSTEM DESIGN 5.4-1 II l

f 5.4.1 Reactor Coolant Pump Assembly

5. 4-1 I

5.4.1.1 Design Bases

5. 4-1 I

i O

MAPWR-CCS vi JUNE,1984

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l l

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

f 5.4.1.2.1 Design Description 5.4-1 I

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

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5.4.2.3 Design Evaluation 5.4-18 I

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WAPWR-RCS yij JUNE, 1984 1393e:1d l1.-

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

f 5.4.3 Reactor Coolant Piping 5.4-27 I

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

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5.4.4 Main Steam Line Flow Restrictors 5.4-33 I

5.4.4.1 Design Basis 5.4-33 I

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5.4.4.2 Design Description 5.4-34 I

5.4.4.3 Design Evaluation 5.4-34 I

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5.4.4.4 Inspections 5.4-34 I

j 5.4.10 Pressurizer 5.4-35 I

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l WAPWR-RCS viii JUNE,1984 l

l 1393e:1d

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 l

5.4.10.3.3 Pressure Setpoints 5.4-39 I

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

I 5.4.12 Valves 5.4-48 I

5.4.12.1 Design Bases 5.4-48 I

f 5.4.12.2 Design Description 5.4-48 I

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

5.4.13.1 Design Bases 5.4-50 I

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

I i

TA8LE OF CONTENTS (Cont)

O

\\

Reference l

SAR Section D

V Section Title Page Status 5.4.13.4 Tests and Inspections 5.4-52 I

q

'5.4.14 Component Supports 5.4-53 I

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 1393e:1d

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

(

l 15.0.4.3 Power Distribution 15.0-10 II 15.0.5 Reactivity Coefficients Assumed in the 15.0-11 II l

Accident Analyses j

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 f

i 15.0.10.1 Inventory in the Core 15.0-16 II I

i O I

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TABLE OF CONTENTS (Cont) j O,

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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

f 15.0.12.3 TWINKLE 15.0-19 II 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

15.3.2.1 Identification of Causes and Accident 15.3-5 I

l Description l

1

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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 O'

Rotor) 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

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TABLE OF CONTENTS (Cont) i i;

Reference l

SAR Section i

l Section Title

Pg_qe, Status 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

f l

15.3.3.2.6 Zirconium - Steam Reaction 15.3-11 I

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

Description f

l i

15.3.4.2 Conclusion 15.3-17 I

15.3.5 References 15.3-17 I

(

O MAPWR-RCS xiii JUNE, 1984

[

1393e:1d l

l

.- - -. a

TABLE OF CONTENTS (Cont)

N i

Reference SAR Section O

Section Title Pace Status l

15.4 REACTIVITY AND POWER DISTRIBUTION ANOMALIES 15.4-1 II 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 l

Inventory 3

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 MAPWR-RCS xiv JUNE, 1984 1393e:Id l

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~

y 15.6.1.3 Conclusion

'15.6-4 I

s 10.6-4 I

15.6'.2

=Pefarences.

x,,

4 15A.1 EENERAL ACCIDENT PARANETERS

\\\\

15.A-1 II s

r

-1

.'15A.2.; '

0FFSITE RADIOLOGICAL CONSEQUENCES 15.A-1 II l

3 CALCULATIONAL MODELS

., j f

15A.2.4 Accident Release Pathways 15.A-2 II I

5 9

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

i

s 4

i J ;f,j 15A.'2.3 ' yTwo-Region Spray Mode'l in x

15.A-4 II s

~Q Containment ((004) j 1

\\

II l

}

15A.2.4 OffsiteThyr,oictosyCaiculationModel 15.A-5 l

15A.2.5 Offsite Beta-Skir. Oose Calcula'tio'nal Model 15 '. A-6 II s

15A.2.6 Of fsite hanna-BodV Dose Calt.alational Model 15.A-6 II l

\\

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 15.'A-7 11 i

s lL ' "

15A.3.2

. Integrated Activity Concentration in' 15.A-8 II s

n.

.i l

hControlRoomfromSingle-RegionSystem t

15A.2.3 Con.colRoomThyroid9csecaiculational 15.A-9 II i

s Model

\\

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

15. A-10 II

'Model s

f U

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

f RadioactiveCloudExternqlt.othe Control Room

{

15A.4 References

15. A-12 II i

g l

JUNE, 1984

{

MAPWR-RCS

.}

xy 1393e:Id i

._.._,.- _ + _._ _._,_ _ ___

l i

c TABLE OF CONTENTS (Cont) i e

i

(

Reference 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 CONSTRUCTION I

t 17.1.1 References 17.1-1 II i

I I

I i

f O

i I

i l

O i

I i

t i

i WAPWR-RCS xyj JUNE, 1984

[

1393e:1d l

- - - -,. ~. _.

TABLE OF CONTENTS (Cont)

LIST OF TABLES i

Number Title Page 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 5.2-2 Reactor vessel Internals Material SDecifications 5.2-47 m

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

i

'l TABLE OF CONTENTS (Cont) l L

LIST OF TABLES

-s i

i Number Title Pace O

i 5.4-11 Reactor Coolant System Valve Design Parameters 5.4-74 l

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

l Examination Program f

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

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

l Accident Analysis (ITDP)

[

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

j Used 15.0-4 Trip Points and Time Delays to Trip Assumed in 15.0-27 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 WAPWR-RCS.

xyjjj L

1393e:1d i

'I t

i f

i TABLE OF CONTENTS (Cont)

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) f 15.3-1 Time Sequence of Events for Incidents which 15.3-18 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 f

15A-2 Limiting Short-Term Atmospheric Dispersion Factors

15. A-14 3

-for Accident Analysis (s/m ),

15A-3 Dose Conversion Factors Used in Accident Analysis

15. A-15 l

l i

i l

l r

i I

O l

l WAPWR-RCS xix JUNE,1984

(

1393e:1d

i I

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

h Analysis 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 JUNE,1984 l

WAPWR-RCS xx l

1393e:1d

7 i

t TABLE OF CONTENTS (Cont)

O.

l t

LIST OF FIGURES z

i-Title O

Number 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 i

f WAPWR-RCS JUNE, 1984 t

t

    • I T393e:1d

I i

TABLE OF CONTENTS (Cont) i O

I LIST OF FIGURES 1

i Number Title i

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 I

r t

lO i

l l

i I

O l

I l

' O i,

t i

JUNE

  • 1984 s

MAPWR-RCS l

1393e:1d Xxii i

_ _,. _. _,.

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