ML20094D180

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Nonproprietary Vols 1 & 2 of RESAR-SP/90 Westinghouse Advanced Pwr,Module 5, Reactor Sys
ML20094D180
Person / Time
Site: 05000601
Issue date: 07/31/1984
From:
WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML19273A435 List:
References
NUDOCS 8408080289
Download: ML20094D180 (562)
v • d • e


Text

O RESAR-SP/90 0 REACTOR SYSTEM WESTINGHOUSE ADVANCED PRESSURIZED '

O WATER REACTOR O

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STANDARD PLANT DESIGN O

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TABLE OF CONTENTS Reference SAR Section O Section Title Page Status a

1.0 INTRODUCTION

AND GENERAL DESCRIPTION OF 1.1-1 11 PLANT 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 Plar.t Description 1. 2-1 II 1.2.3.1 Reactor System 1. 2-1 II 1.3 COMPARISON TABLES 1. 3-1 II 1.3.1 Comparison With Similar Facility Designs 1. 3-1 II 1.5 REQUIREMENTS FOR FURTHER TECHNICAL 1.5-1 II INFORMATION O 1.5.1 Fuel System Tests Fuel Assembly Tests

1. 5-1
1. 5-1 1

'I 1.5.1.1 1.5.1.1.1 Fuel Assembly Structural Tests 1. 5-1 1 1.5.1.1.2 Fuel Assembly Hydraulic Flow Tests 1.5-2 1 1.5.1.2 Core Components Tests 1.5-3 I

1. 5.1. 2.1 Rod Cluster Control and Gray Rod 1.5-3 I Assembly Tests 1.5.1.2.2 Water Displacer Rod Assembly Tests 1.5-3 1 1.5.1.3 Orive Mechanism Tests 1.5-4 I
1. 5 .1. 3 .1 Control Rod Drive Mechanism Tests 1.5-4 I 1.5.1.3.2 Water Displacer Rod Drive Mechanism 1.5-5 I Tests 1.5.2 Reactor Internals Design Verification 1.5-6 I Tests 1.6 MATERIAL INCORPORATED BY REFERENCE 1.6-1 II 1.8 CONFORMANCE WITH THE STANDARD REVIEW 1.8-1 II PLAN O

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TABLE OF CONTENTS (cont)

Reference SAR Section O Section Title Page Status e

2.0 SITE CHARACTERISTICS 2.0-1 NA 3.0 DESIGN OF STRUCTURES, COMPONENTS, 3.1 -1 II O- EQUIPMENT AND SYSTEMS 3.1 CONFORMANCE WITH NRC GENERAL DESIGN 3.1 -1 II CRITERIA ,

3.2 CLASSIFICATION OF STRUCTURES, COMPONENTS, 3. 2-1 II SYSTEMS 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 O- 3.9 MECHANICAL SYSTEMS AND COMPONENTS 3. 9-1 II 3.9.2 Dynamic Testing and Analysis 3.9-1 ~II 3.9.2.3 Oynamic Response Analysis of Reactor 3.9-1 I Internals Under Operational Flow Transients and Steady-State Conditions 3.9.2.4 Preoperational Flow-Induced Vibration 3.9-1 I Testing of Reactor Internals 3.9.2.5 Dynamic System Analysis of the Reactor 3.9-2 I Internals Under Faulted Conditions 3.9.2.6 Correlations of Reactor Internals 3.9-3 I Vibration Tests With Analytical Results 3.9.4 Rod Drive Systems 3.9-3 I 3.9.4.1 Descriptive Information of the Control and 3.9-3 I O Gray Rod Drive Systems 3.9-3 I 3.9.4.1.1 Control Rod Drive Mechanism (CROM) and Gray Rod Drive Mechanism (GROM) 3.9.4.1.1.1 RCCA and GRA Withdrawal 3.9-7 I 3 . 9 . 4 .1.1. 2 RCCA and GRA Insertion 3.9-9 I WAPWR-RS ..

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TABLE OF CONTENTS (cont)

Reference SAR Section O Section Title Page Status a

3.9.4.1.1.3 Holding and Tripping of the RCCAs and 3.9-10 I GRAS 3.9.4.1.2 Applicable Control Rod and Gray Rod Drive 3.9-11 I System Design Specifications (CROS/ GROS) 3.9.4.1.2.1 Design Bases 3.9-11 I 3.9.4.1.2.2 Design Stresses 3.9-11 I 3.9.4.1.2.2.1 Allowable Stresses 3.9-12 1 3.9.4.1.2.2.2 Dynamic Analysis 3.9-12 I 3.9.4.1.2.3 Control Rod Drive Mechanisms (CROMs)/ Gray 3.9-12 I Rod Drive Mechanisms (GRDMs) ,

3.9.4.1.2.4 CROM/GRDM Operational Requirements 3.9-12 I 3.9.4.1.3 CROM/GRDM Design Loads, Stress Limits, 3.9-13 I and Allowable Deformations 3 . 9 . 4 .1. 3.1 Pressure Retaining Components 3.9-13 I 3.9.4.1.3.2 Orive Rod Assembly and Hub Extension 3.9-14 I ,

Assembly 3.9.4.1.3.3 Latch Assembly and Coil Stack Assembly 3.9-15 I 3.9.4.1.3.3.1 Results of Dimensional and Tolerance 3.9-15 I Analysis 3.9.4.1.4 CROM/GRDM Performance Assurance 3.9-16 I Program 3.9.4.2 Displacer Rod Drive System 3.9-18 I 3.9.4.2.1 Descriptive Information of the Displacer 3.9-18 I Rod Drive System 3.9.4.2.2 Displacer Rod Drive Mechanism (DROM) 3.9-18 I 3.9.4.2.3 Displacer Rod Drive Operation 3.9-21 I 3.9.4.2.3.1 Design Bases 3.9-22 I 3.9.4.1.3.2 Design Stresses 3.9-22 I

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TABLE OF CONTENTS (cont)

Reference SAR Section Section Title Page Status 3.9.4.2.3.3 Allowable Stresses 3.9-23 I

,s 3.9.4.2.3.4 Dynamic Stresses 3.9-23 I 3.9.4.2.4 Displacer Rod Drive Operational 3.9-23 I Requirements 3.9.5 Reactor Internals 3.9-23 I 3.9.5.1 Description 3.9-23 I 3.9.5.1.1 - Calandria 3.9-24 I 3.9.5.1.2 Upper Internals Assembly 3.9-25 I 3.9.5.1.2.1 Water Displacer Rod Guide Structure 3.9-26 I 3.9.5.1.2.2 RCCA Guide Structure 3.9-26 I 3.9.5.1.3 Lower Internals 3.9-26 I 3.9.5.1.3.1 Core Barrel Assembly 3.9-27 I 3.9.5.1.3.2 Radial Core Support Keys 3.9-27 1 3.9.5.1.3.3 Secondary Core Support 3.9-27 I 3.9.5.1.3.4 Bottom Mounted Instrumentation 3.9-28 I 3.9.5.1.3.5 Head and Vessel Alignment Pins 3.9-28 I 3.9.5.1.3.6 Hold Down Spring 3.9-29 I 3.9.5.1.3.7 Upper Core Plate Guide Pins 3.9-29 I 3.9.5.1.3.8 Irradiation Specimen Guides 3.9-30 I 3.9.5.1.3.9 Safety Injection Nozzle Deflectors 3.9-30 1 3.9.5.1.3.10 Radial Reflector 3.9-30 I 3.9.5.2 Design Loading Conditions 3.9-31 I 3.9.5.2.1 Normal Conditions Transients (Operating 3.9-31 I Condition I) 3.9.5.2.2 Upset Ccnditions Transients (Operating 3.9-31 I

, Condition II) 3.9.5.2.3 Emergency Conditions Transients 3.9-32 I (Operating Condition III)

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TABLE OF CONTENTS (cont)

Reference SAR Section Section Page

( Title Status 3.9.5.2.4 Faulted Conditions Transients 3.9-32 I (Operating Condition IV)

O 3.9.5.3 3.9.5.3.1 Design Bases Allowable Stresses 3.9-32 3.9-33 I

I 3.9.6 References 3.9-34 II 4.0 REACTOR 4 .1 -1 I 4.1

SUMMARY

DESCRIPTION 4.1-1 I 4.2 FUEL SYSTEM DESIGN 4.2-1 1 4.2.1 Design Bases 4.2-2 I 4.2.1.1 Cladding (Zircaloy-4) 4.2-3 I 4.2.1.2 Fuel Material 4.2-4 I O 4.2.1.3 4.2.1.4 Fuel Rod Performance Spacer Grids 4.2-S 4.2-6 I

I 4.2.1.5 Fuel Assembly 4.2-6 I 4.2.1.6 Core Components 4.2-9 I 4.2.1.7 Testing, Irradiation Demonstration 4.2-11 I and Surveillance 4.2.2 Design Description 4.2-12 I 4.2.2.1 Fuel Rods 4.2-14 I 4.2.2.2 Fuel Assembly Structure 4.2-14 I

% 4.2.2.2.1 Bottom Nozzle 4.2-14 I 4.2.2.2.2 Top Nozzle 4.2-15 I 4.2.2.2.3 Guide and Instrumentation Thimbles 4.2-16 I l 4.2.2.2.4 Grid Assemblies 4.2-18 I e-'s 4.2.2.3 Core Components 4.2-19 I 4.2.2.3.1 Roa Cluster Control Assemblies 4.2-19 I 4.2.2.3.2 Gray Rods 4.2-20 I O

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TABLE OF CONTENTS (cont)

Reference SAR Section p

b Section Title Page Status 4.2.2.3.3 Neutron Source Assembly 4.2-21 I 4.2.2.3.4 Water Displacer Rods 4.2-21 I 4.2.3 Design Evaluation 4.2-22 I 4.2.3.1 Cladding 4.2-23 I 4.2.3.2 Fuel Material Consideration 4.2-28 I 4.2.3.3 Fuel Rod Performance 4.2-29 I 4.2.3.4 Spacer Grids 4.2-38 I 4.2.3.5 Fuel Assembly 4.2-38 I 4.2.3.6 Reactivity Control Assemblies and 4.2-39 I Source Rods 4.2.4 Testing and Inspection Plan 4.2-41 I 4.2.4.1 Quality Assurance Program 4.2-41 I 4.2.4.2 Quality Control 4.2-42 'I 4.2.4.3 Core Component Testing and Inspection 4.2-46 I 4.2.4.4 Tests and Inspections by Others 4.2-48 I 4.2.4.5 Onsite Inspection 4.2-48 I 4.2.5 References 4.2-48 I 4.3 NUCLEAR DESIGN 4.3-1 I 4.3.1 Design Basis 4.3-1 I 4.3.1,.1 Fuel Burnup 4.3-2 I 4.3.1.2 Negative Reactivity Feedbacks 4.3-3 I (Reactivity Coefficient) 4.3.1.3 Control of Power Distribution 4.3-4 I 4.3.1.4 Maximum Controlled Reactivity Insertion 4.3-5 I

, Rate 4.3.1.5 Shutdown Margins 4.3-6 I 4.3.1.6 Stability 4.3-8 I 4.3.1.7 Anticipated Transients Without Trip 4.3-9 I 4.3.2 Description 4.3-9 I WAPWR-RS Vii ' JULY, 1984 1476e:1d i l

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TABLE OF CONTENTS (cont)

Reference SAR Section Section Title Page Status 4.3.2.1 Nuclear Design Description 4.3-9 I 4.3.2.2 Power Distributions 4.3-13 ,I 4.3.2.2.1 Definitions 4.3-13 I 4.3.2.2.2 Radial Power Distributions 4.3-16 I 4.3.2.2.3 Assembly Power Distributions 4.3-17 ,1 4.3.2.2.4 Axial Power Distributions 4.3-17 I 4.3.2.2.5 Local Power Peaking 4.3-19 I 4.3.2.2.6 Limiting Power Distributions 4.3-20 I 4.3.2.2.7 Experimental Verification of Power 4.3-29 I Distribution Analysis 4.3.2.2.8 Testing 4.3-33 I 4.3.2.2.9 Monitoring Instrumentation 4.3-33 I 4.3.2.3 Reactivity Coefficients 4.3-33 I 4.3.2.3.1 Fuel Temperature (Doppler) Coefficient 4.3-35 I 4.3.2.3.2 Moderator Coefficients 4.3-36 -I 4.3.2.3.3 Power Coef ficient 4.3-38 I 4.3.2.3.4 Comparison of Calculated and Experimental 4.3-39 I Reactivity Coefficients 4.3.2.3.5 Reactivity Coefficients Used In Transient 4.3-39 I Analysis 4.3.2.4 Control Requirements 4.3-40 I 4.3.2.4.1 Doppler 4.3-41 I I

4.3.2.'4.2 Variable Average Moderator Temperature 4.3-41 I 4.3.2.4.3 Redistribution 4.3-41 1 4.3.2.4.A Void Content 4.3-42 I 4.3.2.4.5 Rod Insertion Allowance 4.3-42 I

. 4.3.2.4.6 Burnup 4.3-42 I 4.3.2.4.7 Xenon and Samarium Poisoning 4.3-42 I 4.3.2.4.8 pH Effects 4.3-43 I 4.3.2.4.9 Experimental Confirmation 4.3-43 I O

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TABLE OF CONTENTS (cont)

Reference q SAR Section Section Title Pace Status 4.3.2.4.10 Control 4.3-43 I 4.3.2.4.11 Chemical Poison 4.3-43 I 4.3.2.4.12 Rod Cluster Control Assemblies 4.3-44 I 4.3.2.4.13 Water Displacer Rod Assemblies 4.3-45 I 4.3.2.4.14 Gray Rod Assemblies and Load Following 4.3-46 I 4.3.2.4.15 Reactor Coolant Temperature 4.3-47 I 4.3.2.4.16 Integral Fuel Burnable Absorbers (IFBA) 4.3-48 I 4.3.2.4.17 Peak Xenon Startup 4.3-48 I 4.3.2.4.18 Load Follow Control and Xenon Control 4.3-48 I 4.3.2.4.19 Burnup 4.3-49 I 4.3.2.5 Control Rod. Gray Rod, Water Displacer 4.3-49 I Rod Patterns and Reactivity Worths 4.3.2.6 Criticality of the Reactor During 4.3-52 I Ref ueling and Criticality of Fuel Assemblies 4.3.2.7 Stability 4.3-55 I 4.3.2.7.1 Introduction 4.3-55 I 4.3.2.7.2 Stability Index 4.3-56 I 4.3.2.7.3 Prediction of the Core Stability 4.3-56 I 4.3.2.7.4 Stability Measurements 4.3-57 I 4.3.2.7.5 Comparison of Calculations with 4.3-59 I l Measurements Stability Control and Protection 4.3-60 I I 4.3.2.7.6 l

4.3.2.8 Vessel Irradiation 4.3-61 I 4.3.3 Analytical Methods 4.3-62 I 4.3.3.1 Fuel Temperature (Doppler) Calculations 4.3-63 I 4.3.3.2 Macroscopic Group Constants 4.3-64 I 4.3.3.3 Spatial Few-Group Diffusion Calculations 4.3-66 I O

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TABLE OF CONTENTS (cont)

Reference SAR Section Section Title Page Status 4.3.4 Changes 4.3-67 I 4.3-68 I

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4.3.5 References 4.4 THERMAL AND HYORAULIC DESIGN 4 . 4-1 II 4.4.1 Design Basis 4.4-1 1 4.4.1.1 Departure from Nucleate Boiling Design 4.4-2 I Basis 4.4.1.2 Fuel Temperature Design Basis 4.4-4 I 4.4.1.3 Core Flow Design Basis 4.4-4 I 4.4.1.4 Hydrodynamic Stability Design Basis 4.4-5 I 4.4.1.5 Other Considerations 4.4-5 I 4.4.2 Description 4.4-6 I 4.4.2.1 Sununary Comparison 4.4-6 I 4.4.2.2 Critical Heat Flux Ratio or Departure 4.4-7 1 from Nucleate Boiling Ratio and Mixing Technology 4.4.2.2.1 Departure from Nucleate Boiling 4.4-7 I Technology 4.4.2.2.2 Definition of Departure from Nucleate 4.4-8 I Boiling Ratio 4.4.2.2.3 Mixing Technology 4.4-9 I 4.4.2.2.4 Hot Channel Factors 4.4-11 I 4.4.2.2.5 Effects of Rod Bow on DNBR- 4.4-13 I 4.4.2.3 Linear Heat Generation Rate (LHGR) 4.4-14 I 4.4.2.4 Void Fraction Distribution 4.4-14 I 4.4.2.5 Core Coolant Flow Distribution 4.4-14 I 4.4.2.6 Core Pressure Drops and Hydraulic Loads 4.4-14 I 4.4.2.6.1 Core Pressure Drops 4.4-14 I 4.4.2.6.2 Hydraulic Loads 4.4-15 I O

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TABLE OF CONTENTS (cont)

Reference

"% SAR Section Sectior Title Page Statys 9

4.4.2.7 Correlation and Physical Data 4.4-16 I s 4.4.2.7.1 Surface Heat Transfer Coefficients 4.4-16 I 4.4.2.7.2 Total Core and Vessel Pressure Orop 4.4-17 I 4.4.2.7.3 Void Fraction Correlation 4.4-19 I 4.4.2.8 Thermal Effects of Operational Transients 4.4-20 1 4.4.2.9 Uncertainties in Estimates 4.4-20 1 4.4.2.9.1 Uncertainties in Fuel and Cladding 4.4-20 I -

Temperatures 4.4.2.9.2 Uncertainties in Pressure Drops 4.4-21 I 4.4.2.9.3 Uncertainties Due to Inlet Flow 4.4-21 I Ma1 distribution s 4.4.2.9.4 Uncertainty in DNB Correlation 4.4-22 1 4.4.2.9.5 Uncertainties in DNBR Calculations 4.4-22 "I 4.4.2.9.6 Uncertainties in Flow Rates 4.4-23 1 4.4.2.9.7 Uncertainties in Hydraulic Loads 4.4-23 I 4.4.2.9.8 Uncertainty in Mixing Coefficient 4.4-23 1 4.4.2.10 Flux Tilt Consideration 4.4-24 I 4.4.2.11 Fuel and Cladding Temperatures 4.4-25 I 4.4.2.11.1 UO Thermal Conductivity 4.4-26 I 2

4.4.2.11.2 Radial Power Distribution in UO 2 4.4-27 I

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Fuel Rods 4.4.2.11.3 Gap Conducta'nce 4.4-27 I 4.4.2.11.4 Surface Heat Transfer Coefficients 4.4-29 I l 4.4.2.11.5 Fuel Clad Temperatures 4.4-29 I 4.4.2.11.6 Treatment of Peaking Factors 4.4-29 I 4.4.3 Description of the Thermal and Hydraulic 4.4-30 I Design of the Reactor Coolant System 4.4.3.1 Plant Configuration Data 4.4-30 I f

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TABLE OF CONTENTS (cont)

Reference SAR Section Title Page Status Section a

Operating Restrictions on Pumps 4.4-31 1 4.4.3.2 Power-Flow Operating Map (BWR) 4.4-31 I 4.4.3.3 Temperature - Power Operating Map 4.4-31 I 4.4.3.4 Load Following Characteristics 4.4-32 I

' 4.4.3.5 Thermal and Hydraulic Characteristics 4.4-32 I 4.4.3.6 Summary Table 4.4-32 I 4.4.4 Evaluation 4.4-32 I 4.4.4.1 Critical Heat Flux 4.4-32 I 4.4.4.2 Core Hydraulics 4.4-32 I 4.4.4.2.1 Flow Paths Considered in Core Pressure Drop and Thermal Design 4.4-33 I 4.4.4.2.2 Inlet Flow Distributions 4.4-34 I 4.4.4.2.3 Empirical Friction Factor Correlations 4.4-35 I 4.4.4.3 Influence of Power Distribution Nuclear Enthalpy Rise Hot Channel 4.4-35 I 4.4.4.3.1 Factor, F H

4.4-37 I 4.4.4.3.2 Axial Heat Flux Distributions 4.4-38 I 4.4.4.4 Core Thermal Response 4.4-39 I 4.4.4.5 Analytical Techniques 4.4-39 I 4.4.4.5.1 Core Analysis 4.4-39 I i

4.4.4.5.2 Steady-State Analysis 4.4-40 I 4.4.4.5.3 Experimental verification 4.4-40 I 4.4.4.5.4 Transient Analysis 4.4-41 I 4.4.4.6 Hydrodynamic and Flow Power Coupled O Instability 4.4-44 I 4.4.4.7 Fuel Rod Behavior Effects from Coolant Flow Blockage 4.4-46 I 4.4.5 Testing and verification

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TABLE OF CONTENTS (cont)

Reference 2 SAR Section Section Title .PJ,ge, Status a

.4.5.1 Tests Prior to Initial Criticality 4.4-46 I 4.4.5.2 Initial Power and Plant Operation 4.4-46 I O 4.4.5.3 Component and Fuel Inspections Instrumentation Requirements 4.4-46 4.4-46 I

I 4.4.6 4.4.6.1 Incore Instrumentation 4.4-47 I 4.4.6.2 Overtemperature and Overpower AT 4.4-47 I Instrumentation 4.4.6.3 Instrumentation to Limit Maximum Power 4.4-48 Output 1 4.4.7 References 4.4-49 I 4.5 REACTOR MATERIALS 4.5-1 I 4.5.1 Control Rod Drive System Structural 4.5-1 I Materials 4.5.1.1 Control Red Drive Mechanism (CROM) 4.5-1 I and Gray Rod Drive Mechanism (GRDM)

Materials Specifications 4.5.1.2 Fabrication and Processing of Austenitic 4.5-2 I Stainless Steel Components 4.5.1.3 Contamination Protection and Cleaning of 4.5-3 I 1 Austenitic Stainless Steel ,

4.5.1.4 Other Materials 4.5-3 1 4.6 FUNCTIONAL DESIGN OF REACTIVITY CONTROL 4.6-1 I SYSTEMS 4.6.1 Information for the Control Rod Drive 4.6-1 I System (CROS) 4.6.2 Evaluation of the CROS 4.6-1 I

! 4.6.3 Testing and Verification of the CROS 4.6-2 I O

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TABLE OF CONTENTS (cont)

Reference SAR Section Section Title Page Status 4.6.4 Information for Combined Performance 4.6-2 I of Reactivity Systems O 4.6.5 Evaluation of Combined Performance 4.6-3 4.6-4 I

I 4.6.6 References 5.0 REACTOR COOLANT SYSTEM AND CONNECTED 5.0-1 NA SYSTEMS 6.0 ENGINEERED SAFETY FEATURES 6.0-1 NA 7.0 INSTRUMENTATION AND CONTROLS 7 . 0-1 NA 8.0 ELECTRIC POWER 8. 0-1 NA

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9.0 AUXILIARY SYSTEMS 9.0-1 NA 10.0 STEAM AND POWER CONVERSION SYSTEM 10.0-1 NA

! Os 11.0 RADI0 ACTIVE WASTE MANAGEMENT 11.0-1 NA 12.0 RADIATION PROTECTION 12.0-1 HA 13.0 CONDUCT OF OPERATIONS 13.0-1 NA 14.0 INITIAL TEST PROGRAM 14.0-1 NA 15.0 ACCIDENT ANALYSES 15.0-1 II 15.0.1 General 15.0-1 II 15.0.2 Classification of Plant Conditions 15.0-1 II l

15.0.2.1 Condition I - Normal Operation and 15.0-2 II l

Operational Transients O 15.0.2.2 Condition II - Faults of Moderate Frquency 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 15.0.3 Optimization of Control Systems. 15.0-8 II 15.0.4 Plant Characteristics and Initial 15.0-9 II Conditions Assumed in the Accident Analyses 15.0.4.1 Design Plant Conditions 15.0-9 II l

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TABLE OF CONTENTS (cont)

Reference SAR Section O Section Title Pace Status 15.0.4.2 Initial Conditions 15.0-9 II 15.0.4.3 Power Distribution 15.0-10 II 15.0.5 Reactivity Coefficients Assumed 15.0-11 II in the Accident Analyses 15.0.6 Rod Cluster Control Assembly 15.0-12 II Insertion Characteristics 15.0.7 Trip Points and Time Delays 15.0-13 II to Trip Assumed in Accident Analyses 15.0.8 Instrumentation Drif t and Calorimetric 15.0-14 II Errors - Power Range Neutron Flux 15.0.9 Plant Systems and Components Available 15.0-15 II

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for Mitigation of Accident Effects 15.0.10 Fission Product Inventories 15.0-16 II 15.0.10.1 Inventory in the Core 15.0-16 II 15.0.10.2 Inventory in the Fuel Pellet 15.0-16 II Clad Gap 15.0.10.3 Inventory in the Reactor Coolant 15.0-16 II 15.0.11 Residual Decay Heat 15.0-17 II 15.0.11.1 Total Residual Heat 15.0-17 II 15.0.12 Computer Codes Utilized 15.0-17 II 15.0.12.1 FACTRAN 15.0-17 II 15.0.12.2 LOFTRAN 15.0-18 II 15.0.12.3 TWINKLE 15.0-19 II 15.0.12.4 THINC 15.0-19 II 15.0.13 References 15.0-19 II i

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l TABLE OF CONTENTS (cont)

Reference SAR Section i O Section Title Page Status l 15.4 REACTIVITY AND POWER DISTRIBUTION 15.4-1 II ANOMALIES-15.4.1 Uncontrolled Rod Cluster Control Assembly 15.4-2 I Bank Withdrawal f rom a Subcritical or )

Low-Power Startup Condition l 15.4.1.1 Identification of Causes and Accident 15.4-2 I I Description 15.4.1.2 Analysis of Effects and Consequences 15.4-4 I Method of Analysis 15.4-4 I  !

15.4.1.2.1 15.4.1.2.2 Results 15.4-7 I 15.4.1.3 Conclusions 15.4-7 I 15.4.2 Uncontrolled Rod Cluster Control A:sembly 15.4-8 I Bank Withdrawal at Power 15.4.2.1 Identification of Causes and Accident 15.4-8 I Description 15.4.2.2 Analysis of Effects and Consequences 15.4-10 1 15.4.2.2.1 Method of Analysis 15.4-10 I l 15.4.2.2.2 Results 15.4-12 1 15.4.2.3 Radiological Consequences 15.4-14 I

, 15.4.2.4 Conclusions 15.4-14 I 15.4.3 Rod Cluster Control Assembly Misoperation 15.4-15 II (System Malfunction or Operator Error) 15.4.3.1 Identification of Causes and Accident 15.4-15 II Description 15.4.3.2 Analysis of Effects and Consequences 15.4-17 II l

l 15.4.3.2 1 Method of Analysis for Dropped or 15.4-18 II Misaligned RCCA 15.4.3.2.2 Statically Misaligned RCCA Results 15.4-18 I 15.4.3.2.3 Single RCCA Withdrawal Method of Analysis 15.4-19 I WAPWR-RS xvi JULY, 1984 1476e:1d .

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m TABLE OF CONTENTS (cont)

Reference SAR Section

'/ Section Title Pace Status 15.4.3.2.4 Single RCCA Withdrawal Results 15.4-20 I

N 15.4.3.3 Radiological Consequences 15.4-20 I 15.4.3.4 Conclusions 15.4-21 1 15.4.7 Inadvertent loading and Operation of a 15.4-21 I Fuel Assembly in an Improper Position 15.4.7.1 Identification of Causes and Accident 15.4-21 I Description 15.4.7.2 Analysis of Effects and Consequences 15.4-22 I 15.4.7.2.1 Method of Analysis 15.4-22 I 15.4.7.2.2 Results 15.4-23 I 15.4.7.3 Radiological Consequences 15.4-24 I O 15.4.7.4 Conclusions 15.4-24 I 15.4.8 Spectrum of Rod Cluster Control Assembly 15.4-25 'I Ejection Accident 15.4.8.1 Identification of Causes and Accident 15.4-25 I Description 15.4.8.1.1 Design Precautions and Protection 15.4-25 I 15.4.8.1.1.1 Mechanical Design 15.4-25 I 15.4.8.1.1.2 Nuclear Design 15.4-26 I 15.4.8.1.1.3 Reactor Protection 15.4-27 I O 15.4.8.1.1.4 Effects on Adjacent Housings Effects of Rod Travel Housing Longitudinal 15.4-27 15.4-27 I

I 15.4.8.1.1.5 Failures 15.4.8.1.1.6 Effects of Rod Travel Housing 15.4-28 I O Circumferential Failures Possible Consequences 15.4-28 I 15.4.8.1.1.7 15.4.8.1.1.8 Summary 15.4-28 I 15.4.8.1.2 Limiting Criteria 15.4-29 I WAPWR-RS xvii JULY, 1984 1476e:1d

TABLE OF CONTENTS (cont)

Reference SAR Section v Section Title Pace Status 15.4.8.2 Analysis of Effects and Consequences 15.4-30 I 15.4.8.2.1 Calculation of Basic Parameters 15.4-32 I O 15.4.8.2.1.1 Ejected Rod Worths and Hct Channel Factors 15.4-32 I 15.4.8.2.1.2 Reactivity Feedback Weighting Factors 1.5.4-33 I

, 15.4.8.2.1.3 Moderator and Doppler Coef ficients 15.4-34 I 15.4.8.2.1.4 Delayed Neutron Fraction, 8,ff 15.4-34 I 15.4.8.2.1.5 Trip Reactivity Insertion 15.4-34 I l

15.4.8.2.1.6 Reactor Protection 15.4-35 I 15.4.8.2.1.7 Results 15.4-36 I l 15.4.8.2.1.8 Fission Product Release 15.4-39 I O 15.4.8.2.1.9 Pressure Surge 15.4-39 15.4-39 I

'I 15.4.8.2.1.10 Lattice Deformation '

15.4.8.3 Radiological Conse,quences 15.4-40 I 14.4.8.3.1 Analytical Assumptions 15.4-40 1 15.4.8.3.1.1 Source Term Calculations 15.4-40 I  !

15.4.8.3.1.2 Mathematical Models Used in the Analysis 15.4-41 1 15.4.8.3.1.3 Identification of Leakage Pathways and 15.4-41 I Resultant Leakage Activity 15.4.8.3.2 Identification of Uncertainties and 15.4-42 I O. Conservative Elements in the Analysis I

15.4.8.3.3 Conclusions 15.4-43 15.4.8.3.3.1 Filter Loadings 15.4-43 I 15.4.8.3.3.2 Dose to Receptor at the Exclusion Area 15.4-43 I Boundary and Low Population Zone Outer Boundary 15.4.9 References 15.4-44 II w

i WAPWR-RS xviii JULY, 1984 1476e:1d

O TABLE OF CONTENTS (cont)

Reference SAR Section O Section Title P64e Status 15A ACCIDENT ANALYSIS RADIOLOGICAL 15.A-1 II CONSEQUENCES EVALUATION MODELS AND O PARAMETERS 15A.1 General Accident Parameters 15.A-1 II 15A.2 Offsite Radiological Consequences 15.A-1 11 Calculational Models 15A.2.1 Accident Release Pathways 15.A-2 11 15A.2.2 Single-Region Release Model 15.A-2 11 15A.2.3 Two-Region Spray Model in Containment 15.A-4 II (LOCA) 15A.2.4 Offsite Thyroid Dose Calculation Model 15.A-5 II O 15A.2.5 Offsite Beta - Skin Dose Calculational 15.A-6 II Model 15A.2.6 Offsite Gamma-Body Dose Calculational 15.A-6 II Model 15A.3 Control Room Radiological Consequences 15.A-7 II Calculational Models 15A.3.1 Integrated Activity in Control Room 15.A-7 II 15A.3.2 Integrated Activity Concentration in 15.A-8 II Control Room From Single-Region O System Control Room Thyroid Dose Calculational 15.A-9 11 15A.3.3 Model 15A.3.4 Control Room Beta - Skin Dose 15. A-10 II O Calculational Model Control Room Gamma - Body Dose 15. A-11 II 15A.3.5 Calculation 15 A . 3. 5.1 Model for Radiological Consequences Due 15. A-11 II l () to Radioactive Cloud External to the Control Room WAPWR-RS xjx JULY, 1984 1476e:1d

, i l TABLE OF CONTENTS (cont) i Reference i SAR Section i

, Section Title PLge. Status i

15A.4 References 15. A-12 II 16.0 TECHNICAL SPECIFICATIONS 16.0-1 N/A 17.0 QUALITY ASSURANCE 17.1-1 II

17.1 Quality Assurance During Design and 17.1-1 II Construction 17.1.1 References 17.1-1 II 1

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O TABLE OF CONTENTS (cont) ,

l KEY TO " REFERENCE SAR SECTION STATUS" COLUMN:

Category I Those sections .which are complete and for which no additional inf ormation is to be provided for the PDA application.

Category II Those sections which are complete insof ar as providing material relevant to

this system module but for which additional information will be provided in support of subsequent modules.

I Category III O Those sections for which information on interfacing systems will be provided at a later date.

NA.

Those sections for which categorization is not applicable. Only the section titles are included for clarity.

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xxj JULY, 1984 WAPWR-RS 1476e:1d

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N s TABLE OF CONTENTS (cont)

LIST OF TABLES Number Title Page o

1. 3-1 Design Comparison 1.3-2 1.6-1 Material Incorporated by Reference 1.6-2

]

1.8-1 Standard Review Plan Deviations 1.8-2 1.8-2 Conformance to US NRC Regulatory Guides 1.8-3 Applicable to the WAPWR Reactor System 3.1-1 GDC Applicable to the Reactor System 3.1-2 3.2-1 Classification of Structures, Systems, and 3.2-3 Components for the Reactor System 4.1-1 Reactor Design Comparison Table 4.1-4 4.1-2 Analytical Techniques In Core Design 4.1-9 p 4 .1 -3 Design Loading Conditions Considered for 4.1-11 Reactor Core Components .

4.3-1 Reactor Core Description 4.3-73 4.3-2 Nuclear Design Parameters 4.3-77 4.3-3 Reactivity itaquirements for Rod Cluster 4.3-80 Control Assemblies 4.3-4 Benchmark Critical Experiments 4.3-82 4.3-5 Axial Stability Index for Pressurized Water 4.3-83 Reactor Core With a 12-Foot Height 4.3-6 Typical Neutron Flux Levels (n/cm -sec) 4.3-84 ON At Full Power 4.3-7 Comparison of Measured and Calculated 4.3-85 Doppler Defects m 4.3-8 Saxton Core II Isotopics Rod MY, Axial Zone 6 4.3-86 4.3-9 Critical Boron Concentrations, HZP, BOL 4.3-87 4.3-10 Benchmark Critical Experiments, B4C Conuol 03-88 j Rod Worth O

WAPWR-RS xxfi JULY, 1984 1476e:ld

O TABLE OF CONTENTS (cont)

LIST OF TABLES O Number Title a

Page 4.3-11 Comparison of Measured and Calculated Moderator 4.3-89

's Coefficients at HZP, BOL

\ 4.4-1 Thermal and Hydraulic Comparison Table 4.4-58 15.0-1 Nuclear Steam Supply System Power Ratings 15.0-21 15.0-2 Values of Pertinent Plant Parameters Utilized 15.0-22 In Accident Analyses (ITOP) 15.0-2a Values of Pertinent Plant Parameters Utilized 15.0-23 In Accident Analyses (non-ITDP) i 15.0-3 Summary of Initial Conditions and Computer 15.0-24 Codes Used 15.0-4 Trip Points and Time Delays to Trip Assumed 15.0-27 O In Accident Analyses Determination of Maximum Overpower Trip Point - 15.0-28 15.0-5 Power Range Neutron Flux Channel-Based on

! Nominal Setpoint Considering Inherent Instrument Errors 15.0-6 Plant Systems and Equipment Available for 15.0-30 Transient and Accident Conditions l Fuel and Rod Gap Inventories, Core (Ci) 15.0-34 i 15.0-7 15.0-0 Reactor Coolant ludine Concentrations for 15.0-35 O 1 uCi/ gram and 60 pCi/ gram of Dose Equivalent I-131 15.0-9 Reactor Coolant Noble Gas Specific Activity 15.0-36 Based on One Percent Def ective Fuel

(

j O 15.0-10 Iodine Appearance Rates In the Reactor Coolant 15.0-37 I (Curies /sec.)

I xxjjj JULY, 1984 WAPWR-RS 1476e:1d

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O TABLE OF CONTENTS (cont)

LIST OF TABLES O Number Title Page 15.4-1 Time Sequence of Events for Incidents Which 15.4-46 Result in Rer'.;ivity and . Power Distribution Anomalies 15.4-2 Minimum Calculated DNBR for Rod Cluster 15.4-50 Control Assembly Misalignment f 15.4-3 Parameters Used in the Analysis of the Rod 1 5.4 -51 Cluster Control Assembly Ejection Accident 15.4-4 Parameters Used in Evaluating the Radiological 15.4-52 Consequences of a Control Rod Ejection Accident

..(For a Typical Four-Loop Westinghouse PWR) 15.4-5 Radiological Consequences of a Control Rod 15.4-55 l

Ejection Accident (For a Typical Four-Loop .

Westinghouse PWR) l

15A-1 Parameters Used in Accident Analysis 15. A-13 15A-2 Limiting Short-Term Atmospheric Dispersion 15.A-14 Factors for Accident Analysis (S/M )*

15A-3 Dose Conversion Factors used in Accident Analysis 15. A-15 0 .

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WAPWR-RS xxiy JULY, 1984 l

1476e:ld

TABLE OF CONTENTS (cont)

LIST OF FIGURES Number Title i

1. 2-1 Fuel Assembly Outline 1.2-2 Reactor Vessel 1.2-3 Integrated Head Paci ge 1.2-4 Displacer Rod Drive Mechanism 1.2-5 Comparison of 414 and WAPWR 3 . 9-1 Typical Full-length Control Rod Drive f Mechanism 3.9-2 Typical Full-Length Control Rod Drive Mechanism Schematic 3.9-3 Nominal Latch Clearance at Minimum and Maximum Temperatures 3.9-4 Nominal Control Rod Drive Mechanism I

Latch Clearance Thermal Effect 3.9-5 DRDM Vent System 3.9-6 Displacer Rod Drive Mechanism 3.9-7 WAPWR Reactor Internals (General Assembly Layout) 3.9-8 WAPWR Calandria Conceptual Design Layout 3.9-9 WAPWR, WDRC, RCCA Cros: Section 3.9-10 WAPWR Inner Barrel Conceptual Design Layout 3.9-11 Radial Reflector Overall Plan view l

(Quadrant) 3.9-12 WAPWR Radial Reflector Module General Assembly (Typical) 4 . 2-1 19x19 Fuel Assembly With 16 Guide Thimbles 4.2-2 WAPWR 19x19 Fuel Assembly f WAPWR-RS xxy JULY, 1984 1476e:1d

TABLE OF CONTENTS (cont)

LIST OF FIGURES Number Title a

4.2-3 Fuel Rod Schematic 4.2-4 Top and Bottom Nozzles

e. 2-5 Guide Thimble to Bottom Grid and Nozzle Joint 4.2-6 Plan View of Mid-Grid to Guide Thimble Joint (Bottom View) 4.2-7 Elevation View of Mid-Grid to Guide Thimble Joint 4.2-8 Top Grid to Guide Thimble and Top Nozzle ,

Attachment 4.2-9 Rod Cluster Control and Drive Rod 4.2-10 RCCA and Gray Rod Assembly O 4.2-11 Gray and Absorber Rodlet Schematic d 4.2-12 Secondary Source Rod Schematic 4.2-13 Water Displacer Rodlet Schematic 4.2-14 Water Displacer Assembly 4.2-15 Typical Rod Cluster Arrangement - 19x19 Fuel Assembly Array,16 Guide Thimbles Per Assembly 4.3-1 First Core Loading Pattern 4.3-2 Production and Consumption of Higher Isotopes ,

4.3-3 Boron Concentration Versus First Cycle Burnup 4.3-4 Fuel Assembly Cross-Section 4.3-5 Normalized Power Density Distribution Near Beginning of Life, Unrodded Core, Hot Full Power, No Xenon, WDR's and GR's Inserted 4.3-6 Normalized Power Density Distribution Near Beginning of Life, Unrodded Core, Hot Full Power, Equilibrium Xenon, WDR's and GR's Inserted xxvi JULY , 1984 WAPWR-RS 1476e:1d

TABLE OF CONTENTS (cont)

LIST OF FIGURES Number Title 4.3-7 Normalized Power Density Distribution Near Beginning of Life, Group D at 18% Inserted, Q Hot Full Power, Equilibrium Xenon, WDR's and GR's Inserted 4.3-8 Normalized Power Density Distribution Near Middle of Life, Unrodded Core, Hot Full Power, Equilibrium Xenon, WDR's and GR's Inserted 4.3-9 Normalized Power Density Distribution, Unrodded Core, Hot Full Power, Equilibrium Xenon, WDR's and GR's Inserted 4.3-10 Normalized Power Density Distribution, Group D at 18% Inserted, Hot Full Power, Equilibrium .

Xenon, WDR's and GR's Inserted 4.3-11 Normalized Power Density Distribution, Unrodded Core, Hot Full Power, Equilibrium Xenon, following Withdrawal of WDR's and GR's 4.3-12 Normalized Power Density Distribution, GrouD D at 18% Inserted, Hot Full Power, Equilibrium Xenon, Following Withdrawal of WDR's and GR's

. 4.3-13 Normalized Power Density Distribution Near End of Life, Unrodded Core, Hot Full Power, Equilibrium Xenon With WDR's and GR's Withdrawn 4.3-14 Normalized Power Density Distribution Near End of Life, Group D at 18% Inserted, Hot Full Power, Equilibrium Xenon 4.3-15 Rodwise Power Distribution in a Typical Assembly ( Assembly F-12), Near Beginning of 1.ife, Hot Full Power, No Xenon, Unrodded Core, WDR's and GR's Inserted WAPWR-RS XXVii JULY, 1984 1476e:ld

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TABLE OF CONTENTS (cont)

LIST OF FIGURES Number Title a

4.3-16 Rodwise Power Distribution in a Typical Assembly (Assembly F-11), Hot Full Power, Equilibrium Xenon Unrodded Core, WDR's and GR's Inserted 4.3-17 Rodwise Power Distribution in a Typical l

Assembly ( Assembly F-13), Near Beginning of Life, Hot Full Power, No Xenon, Unrodded Core, WDR's and GR's Inserted 4.3-18 Rodwise Power Distribution in a Typical Assembly ( Assembly F-13), Hot Full Power, Equilibrium Xenon, Unrodded Core, WDR's and GR's Inserted 4.3-19 Rodwise Power Distribution in a Typical

! Assembly ( Assembly F-12), Near End of 4

Life, Hot Full Power, Equilibrium Xenon, i Unrodded Core, WOR's and GR's Withdrawn 4.3-20 Rodwise Power Distribution in a Typical Assembly ( Assembly F-13), Near End of Lif e, Hot Full Power, Equilibrium Xenon, l

Unrodded Core, WOR's and GR's Withdrawn 4.3-21 Typical Axial Power Shapes Occurring at Beginning of Life, WDR's and GR's inserted 4.3-22 Typical Axial Power Shapes Occurring at Middle of Life, WOR's and GR's Inserted 4.3-23 Typical Axial Power Shapes Occurring at End of Life, WDR's and GR's Inserted 4 4.3-24 Typical Axial Power Shapes Occurring at

) End of Lif e, WOR's and GR's Withdrawn WAPWR-RS xxviii JULY, 1984 1476e:1d v--- , -

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s TABLE OF CONTENTS (cent)

LIST OF FIGURES I Number Title _

! 4.3-25 Comprison of Assembly Axial Power Distribution With' Core Average Axial v Distribution Bank Slightly Inserted  ;

4.3-26 Flow Chart for Determining Spike Model j 4.3-27 Predicted Power Spike Due to Single j Non-flattened Gap in the Adjacent Fuel 4.3-28 Power Spike Factor as a Function of Axial Power f 4.3-29 Maximum F x Power Versus Axial Height q

During Normal Operation (Typical Envelope) ,

4.3-30 Peak Linear Power During Control Rod  !

( Malfunction Overpower Transient j 4.3-31 Peak Linear Power During Boration/Deboration .

l Overpower Transients l 4.3-32 Typical Comparison Between Calculated and Measured Relative Fuel Assembly Power Distribution l )

4.3-33 Comparison of Calculated and Measured Axial Shape 1 l 4.3-34 Comparison of Calculated and Measured Peaking l Factor, [F xPREL] g Max Envelope as a Function of )

Core Height j 4.3-35 Doppler Temperature Coefficient at BOL and EOL, Cycle 1 )

4.3-36 Doppler Only Power W

  • M nt at BOL and EOL, Cycle 1 4.3-37 Doppler Only Power tief ect at 30L and EOL, v Cycle 1 I 4.3-38 Moderator Temperature Coefficient at SQL, Cycle 1, No Rods, WDR's and GR's Inserted O

WAPWR-RS XXiX JULY, 1984 1476e:1d

O TABLE OF CONTENTS (cont)

LIST OF FIGURES O idumber Title a

4.3-39 Moderator Temperature Coefficient Near EOL, Cycle 1 WOR's and GR's Inserted O- 4.3-40 Moderator Temperature Coefficient at EOL, Cycle 1, WOR's and GR's Withdrawn 4.3-41 Moderator Temperature Coefficient as a Function of Boron Concentration, BOL, Cycle I, No Rods, WOR's and GR's Inserted 4.3-42 Hot Full Power Temperature Coefficient During Cycle 1 for the Critical Boron Concentration 4.3-43 Total Power Coefficient, BOL, EOL, Cycle 1 4.3-44 Total Power Defect, BOL, EOL, Cycle 1 O 4.3-45a Rod Cluster Control Assembly and Gray Rod Pattern DROM/WORC Group Pattern 4.3-45b 4.3-46 Accidental Simultaneous Withdrawal of Two Control Banks, EOL, HZP, Banks C and B Moving in Same Plane 4.3-47 Typical Design Trip Curve-4.3-48 Normalized Rod Worth Versus Percent Insertion, All Rods But One 4.3-49 Axial Offset Versus Time, PWR Core With a 12'ft.

O Height and 121 Assemblies XY Xenon Test Thermocouple Response Quandrant 4.3-50 Tilt Difference Versus Time 4.3-51 Calculated and Measured Doppler Defect and O Coefficient at BOL, 2-Loop Plant, 121 Assemblies, 12 ft. Core 4.3-52 Comparison of Calculated and Measured Boron Concentration f or 2-Loop Plant,121 Assemblies,

( 12 ft. Core xxx JULY, 1984 WAPWR-RS 1476e:ld l

A TABLE OF CONTENTS (cont) ,

LIST OF FIGURES O Number Mtle 4.3-53 Comparison of Calculated and Measured CB , 3-Loop Plant, 157 Assemblies, 12 ft. Core O 4.3-54 Comparison of Calculated and Measured C Plant, 193 Assemblies, 12 ft. Core B

, 4-L p 4.4-1 Improved Thermal Design Procedure Illustration 4.4-2 Measured Versus Predicted Critical Heat Flux, WRB-2 Correlation 4.4-3 TDC versus Reynolds Number for 26 inch Grid Spacing 4.4-4 Normalized Radial Flow and Enthalpy Rise Distribution at Elevation of 1/3 of Core Height 4.4-5 Normalized Radial Flow and Enthalpy Rise O Distribution at Elevation of 2/3 of Core Height Normalized Radial Flow and Enthalpy Rise 4.4-6 Distribution at Core Exit Elevation 4.4-7 Void Fraction Versus Thermodynamic Quality

"-" SAT /Hg-HSAT 4.4-B Thermal Conductivity of UO2 (Data Corrected to 95% Theoretical Density) 4.4-9 Reactor Coolant System Temperature Percent Power Map V 4.4-10 Distribution of Incore Instrumentation 15.0-1 Illustration of Core Thermal Limits and DNB Protection (N-Loop Operation) 15.0-2 Doppler Power Coefficient Used In Accident O, Analysis 15.0-3 RCCA Position vs. Time to Dashpot 15.0-4 Normalized RCCA Reactivity Worth vs. Fraction Insertion

  • JULY, 1984 WAPWR-RS 1476e:1d 1

TABLE OF CONTENTS (cont)

LIST OF FIGURES O Number Title a

15.0-5 Normalized RCCA Bank Reactivity Worth vs.

Normalized Drop Time 15.4-1 Neutron Flux Transient for Uncontrolled Rod Withdrawal f rom a Subtritical Condition 15.4-2 Thermal Flux Transient for Uncontrolled Red Withdrawal f rom a Subcritical Condition 15.4-3 Fuel and Clad Temperature for Uncontrolled Rod Withdrawal from a Subtritical Condition 15.4-4 Nuclear Power Transient and Heat Flux Transient for Uncontrolled Rod Withdrawal from Full Power with Minimum Feedback and 75 pcm/sec Withdrawal Rate 15.4-5 Pressurizer Pressure and Water Volurae Transients for Uncontrolled Rod Withdrawal from Full Power with Minimum Feedback and 75 pcm/sec Withdrawal Rate 15.4-6 Core Average Temperature Transient and DNBR vs. Time for Uncontrolled Rod Withdrawal from Full Power with Minimum Feedback and 75 pcm/sec Withdrawal Rate 15.4-7 Nuclear Power Transient and Heat Flux Transients for Uncontrolled Rod Withdrawal from Full Power with Minimu.a Feedback and 1 pcm/sec Withdrawal Rate 15.4-8 Pressurizer Pressure and Water Volume Transients for Uncontrolled Rod Withdrawal from Full Power with Minimum Feedback and 1 pcm/sec Withdrawal Rate O

xxxij JULY, 1984 WAPWR-RS 1476e:ld

f TABLE OF CONTENTS (cont)

LIST OF FIGURES Number Title 15.4-9 Core Average Tempe,ature Transient and DNBR vs. Time for Uncontrolled Rod Withdrawal from O Full Power with Minimum Feedback and 1 pcm/sec Withdrawal Rate 15.4-10 Rod Withdrawal at Power (100% Power) 15.4-11 Rod Withdrawal at Power (60% Power) 15.4-12 Representative % Change in Local Assy. Avg.

Power for Interchange Between Region 1 and Region 3 Assy.

15.4-13 Representative % Lhange in Local Assy. Avg.

Power for Interchange Between Region 1 and O Region 2 Assy. with BP Rods Retained by the Region 2 Assy.

15.4-14 Representative % Change in Local Assy. Avg.

Power for Interchange Between Region 1 and Region 2 Assy. with the BP Rods fransferred to Region 1 Assy.

15.4-15 Representative % Change in Local Assy. Avg.

Power for Enrichment Error (Region 2 Assy.

Loaded into Core Central Position) 15.4-16 Representative % Change in Local Assy. Avg.

Power for Loading Region 2 Assy. into Region 1 Position Near Core Periphery 15.4-17 Nuclear Power Transient BOL, Voided, Full Power 15.4-1B Hot Spot Fuel and Clad Temperature vs. Time BOL, Voided, Full Power O

WAPWR-RS xxxijj JULY, 1984 1476e:id

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1 TABLE OF CONTENTS (cont) l

LIST OF FIGURES l

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15.4-19 Nuclear Power Transient, EOL, Flooded, i Zero Power 15.4-20 Hot Spot Fuel and Clad Temperatures vs. ,

Time EOL, Flooded, Zero Power 15.A-1 Release Pathways I ,

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