ML20129E715
| ML20129E715 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 10/14/1996 |
| From: | SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | |
| Shared Package | |
| ML20129E384 | List: |
| References | |
| SAR-961014, NUDOCS 9610280144 | |
| Download: ML20129E715 (185) | |
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I a
DEFUELED SAFETY ANALYSIS REPORT TABLE OF CONTENTS CHAPTER 1 INTRODUCTION AND
SUMMARY
! Section Ii11g page 4
1.1 INTRODUCTION
1.1-1 1.2 DESIGN HIGHLIGHTS 1.2-1 1
1.2.1 SITE CHARACTERISTICS 1,2-1 i
1.2.2 POWER LEVEL 1.2 1 1.2.3 REACTOR BUILDING 1.2-1 1.2.4 SAFETY FEATURES 1.2-1 4 1.2.5 ELECTRICAL SYSTEMS AND EMERGENCY POWER 1.2-1 l 1 !
1.2.6 SEISMIC DESIGN QUALITY ASSURANCE 1.2-2
- 1.2.7 FUEL STORAGE BUILDING 1.2-7 j i
1.3 DESIGN CHAR ACTERISTICS 1.3-1
- 1.3.1 DESIGN PARAMETERS 1.3-1 l 1.3.2 SIGNIFICANT DESIGN REVISIONS 1.3-1 i l.4 PRINCIPAL DESIGN CRITERIA - ORIGINAL 1.4-1 2
4 1.5 PRINCIPAL DESIGN CRITERIA - PERMANENTLY 1.5-1 4
DEFUELED MODE ;
l i 1.5.1 CRITERION 1 - QUALITY STANDARDS AND RECORDS 1.5-1
! 1.5.2 CRITERION 2 - DESIGN BASIS FOR PROTECTION 1.5-2 AGAINST NATURAL PHENOMENA ,
1 1.5.3 CRITERION 3 - FIRE PROTECTION 1.5-3 l.5.4 CRITERION 4 - ENVIRONMENTAL AND MISSILE 1.5-5 1 DESIGN BASES 1.5.5 CRITERION 5 - SHARING OF STRUCTURES, SYSTEMS, 1.5-5 AND COMPONENTS 1.5.6 CRITERION 10- REACTOR DESIGN 1.5-5 4
Amendment 1 i 9610200144 961014 a PDR ADOCK 05000312 i K PDR i
DEFUELED SAFETY ANALYSIS REPORT TABLE OF CONTENTS l CHAPTER 1. INTRODUCTION AND SUMM ARY (Continued) l klion M Eagg j 1.5.7 CRITERION 11 - REACTOR INHERENT PROTECTION 1.5-6 l l 1.5.8 CRITERION 12 - SUPPRESSION OF REACTOR POWER 1.5-6 l OSCILLATIONS 1.5.9 CRITERION 13 - INSTRUMENTATION AND CONTROL 1.5-6 l 1.5.10 CRITERION 14 - REACTOR COOLANT PRESSURE 1.5-6 BOUNDARY l 1.5.11 CRITERION 15 - REACTOR COOLANT SYSTEM DESIGN 1.5-6 !
1.5.12 CRITERION 16 - CONTAINMENT DESIGN 1.5-7 l ;
1.5.13 CRITERION 17 - ELECTRIC POWER SYSTEMS 1.5-7 ;
1.5.14 CRITERION 18 - INSPECTION AND TESTING OF 1.5-8 ELECTRICAL POWER SYSTEMS ;
1.5.15 CRITERION 19 - CONTROL ROOM 1.5-9 >
1.5.16 CRITERION 20 - PROTECTION SYSTEM FUNCTIONS 1.5-10 )
1.5.17 CRITERION 21 - PROTECTION SYSTEM RELIABILITY 1.5-10 i AND TESTABILITY l 1.5.18 CRITERION 22 - PROTECTION SYSTEM INDEPENDENCE 1.5-10 f 1.5.19 CRITERION 23 - PROTECTION SYSTEM FAILURE MODES 1.5-10 l l
1.5.20 CRITERION 24 - SEPARATION OF PROTECTION AND 1.5-10 CONTROL SYSTEMS 1.5.21 CRITERION 25 - PROTECTION SYSTEM REQUIREMENTS 1.5-10 ,
FOR REACTIVITY CONTROL MALFUNCTIONS i 1.5.22 CRITERION 26 - REACTIVITY CONTROL SYSTEM 1.5-11 i REDUNDANCY AND CAPABILITY 1.5.23 CRITERION 27 - COMBINED REACTIVITY CONTROL 1.5-11 l l SYSTEMS CAPABILITY ;
1.5.24 CRITERION 28 - REACTIVITY LIMITS 1.5-11 l{
l.5.25 CRITERION 29 - PROTECTION AGAINST ANTICIPATED 1.5-11 l OPERATIONAL OCCURRENCES i Amendment 2 -
I 1
il l
DEFUELED SAFETY ANALYSIS REPORT TABLE OF CONTENTS CHAPTER L INTRODUCTION AND
SUMMARY
(Continued)
Section Iltic Ease 1.5.26 CRITERION 30 - QUALITY OF REACTOR COOLANT 1.5-11 PRESSURE BOUNDARY 1.5.27 CRITERION 31 - FRACTURE PREVENTION OF REACTOR 1.5-11 COOLANT PRESSURE BOUNDARY 1.5.28 CRITERION 32 - INSPECTION OF REACTOR COOLANT 1.5-11 PRESSURE BOUNDARY 1.5.29 CRITERION 33 - REACTOR COOLANT MAKEUP 1.5-12 l 1.5.30 CRITERION 34- RESIDUAL HEAT REMOVAL 1.5-12 1.5.31 CRITERION 35 - EMERGENCY CORE COOLING 1.5-12 l ;
1.5.32 CRITERION 36 - INSPECTION OF EMERGENCY CORE 1.5-12 l COOLING SYSTEM 1.5.33 CRITERION 37 - TESTING OF EMERGENCY CORE 1.5-12 COOLING SYSTEM 1.5.34 CRITERION 38 - CONTAINMENT HEAT REMOVAL 1.5-12 1.5.35 CRITERION 39 - INSPECTION OF CONTAINMENT HEAT 1.5-12 i
REMOVAL SYSTEM 1.5.36 CRITERION 40 - TESTING OF CONTAINMENT HEAT 1.5-12 REMOVAL SYSTEM 1.5.37 CRITERION 41 - CONTAINMENT ATMOSPHERE CLEANUP 1'5-13 .
l 1.5.38 CRITERION 42 - INSPECTION OF CONTAINMENT 1.5-13 l ATMOSPHERE CLEANUP 1.5.39 CRITERION 43 - TESTING OF CONTAINMENT 1.5-13 l ATMOSPHERE CLEANUP SYSTEMS 1.5.40 CRITERION 44 - COOLING WATER 1.5-13 l 1.5.41 CRITERION 45 - INSPECTION OF COOLING WATER 1.5-14 l SYSTEM 1.5.42 CRITERION 46 - TESTING OF COOLING WATER SYSTEM 1.5-14 Arc v.dT =r t 2 iii
DEFUELED SAFETY ANALYSIS REPORT TABLE OF CONTENTS CHAPTER 1. INTRODUCTION AND
SUMMARY
(Continued)
Section 'Ihic Eage 1.5.43 CRITERION 50 - CONTAINMENT DESIGN BASIS 1.5-14 1.5.44 CRITERION 51 - FRACTURE PREVENTION OF 1.5-15 l
CONTAINMENT PRESSURE BOUNDARY 1.5.45 CRITERION 52 - CAPABILITY FOR CONTAINMENT 1.5-15 l LEAKAGE RATE TESTING 1.5.46 CRITERION 53 - PROVISIONS FOR CONTAINMENT 1.5-15 l TESTING AND INSPECTION 1.5.47 CRITERION 54 - PIPING SYSTEMS PENETRATING 1.5-15 CONTAINMENT 1.5.48 CRITERION 55 - REACTOR COOLANT PRESSURE 1.5-15 BOUNDARY PENETRATING CONTAINMENT 1.5.49 CRITERION 56 - PRIMARY CONTAINMENT ISOLATION 1.5-15 1.5.50 CRITERION 57 - CLOSED SYSTEM ISOLATION VALVES 1.5-15 1.5.51 CRITERION 60 - CONTROL OF RELEASES OF 1.5-16 RADIOACTIVE MATERIALS TO THE ENVIRONMENT 1.5.52 CRITERION 61 - FUEL STORAGE AND HANDLING AND 1.5-17 RADIOACTIVITY CONTROL 1.5.53 CRITERION 62 - PREVENTION OF CRITICALITY IN 1.5-18 l FUEL STORAGE AND HANDLING 1.5.54 CRITERION 63 - MONITORING FUEL AND WASTE 1.5-18 STORAGE 1.5.55 CRITERION 64 - MONITORING RADIOACTIVITY 1.5-19 RELEASES 1.6 COMPARISON WITH SAFETY GUIDES 1.6-1 1.6.1 SAFETY GUIDE 1 - NET POSITIVE SUCTION HEAD FOR 1.6-1 EMERGENCY CORE COOLING AND CONTAINMENT HEAT REMOVAL SYSTEM PUMPS 1.6.2 SAFETY GUIDE 2 - THERM AL SHOCK TO REACTOR 1.6-1 PRESSURE VESSELS Amendnet 2 iv
DEFUELED SAFETY ANALYSIS REPORT TABLE OF CONTENTS CHAPTER 1. INTRODUCTION AND
SUMMARY
(Continued)
Section TILin Eage 1.6.3 SAFETY GUIDE 3 - ASSUMPTIONS USED FOR 1.6-1 EVALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF A LOSS-OF-COOLANT ACCIDENT FOR BOILING WATER REACTORS 1.6.4 SAFETY GUIDE 4- ASSUMPTIONS USED FOR 1.6-1 EVALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF A LOSS-OF-COOLANT ACCIDENT FOR PRESSURIZED WATER REACTORS 1.6.5 SAFETY GUIDE 5 - ASSUMPTIONS USED FOR 1.6-1 EVALUATING THE POTENTIAL CONSEQUENCES OF A STEAM LINE BREAK ACCIDENT FOR BOILING WATER REACTORS 1.6.6 S AFETY GUIDE 6 - INDEPENDENCE BETWEEN 1.6-1 REDUNDANT STANDBY (ON-SITE) POWER SOURCES AND THEIR DISTRIBUTION SYSTEMS 1.6.7 SAFETY GUIDE 7 - CONTROL OF COMBUSTIBLE GAS 1.6-2 CONCENTRATIONS IN CONTAINMENT FOLLOWING A LOSS-OF-COOLANT ACCIDENT 1.6.8 SAFETY GUIDE 8 - PERSONNEL SELECTION AND 1.6-2 TRAINING 1.6.9 SAFETY GUIDE 9 - SELECTION OF DIESEL GENERATOR 1.6-2 SET CAPACITY FOR STANDBY POWER SUPPLIES 1.6.10 SAFETY GUIDE 10 - MECHANICAL (CADWELD) SPLICES 1.6-2 IN REINFORCING BARS OF CONCRETE CONTAINMENT 1.6.11 S AFETY GUIDE I 1 - INSTRUMENT LINES PENETRATING 1.6-2 PRIMARY REACTOR CONTAINMENT 1.6.12 SAFETY GUIDE 12 - INSTRUMENTATION FOR 1.6-2 EARTHQUAKES 1.6.13 S AFETY GUIDE 13 - FUEL STORAGE FACILITY 1.6-3 DESIGN BASIS 1.6.14 SAFETY GUIDE 14 - REACTOR COOLANT PUMP FLY- 1.6-4 WHEEL INTEGRITY v
i.
I DEFUELED SAFETY ANALYSIS REPORT TABLE OF CONTENTS
- CHAIYTER L IN1'RODUCTION AND SUMM ARY.(Continued) ;
Section Ii11c Eage i
- 1.6-5 J 1.6.15 SAFETY GUIDE 15 - TESTING OF REINFORCING BARS j FOR CONCRETE STRUCTURES ,
h 1.6-5 l.6.16 SAFETY GUIDE 16 - REPORTING OF OPERATING t
- INFORMATION 1.6.17 SAFETY GUIDE 17 - PROTECTION AGAINST INDUSTRIAL 1.6-5 i
- SABOTAGE >
t
< l.6.18 SAFETY GUIDE 18 - STRUCTURAL ACCEPTANCE TEST 1.6-7
- FOR CONCRETE PRIMARY REACTOR CONTAINMENTS 1.6.19 S AFETY GUIDE 19 - NONDESTRUCTIVE TESTING OF 1.6-7 l F i PRIMARY CONTAINMENT LINERS
! 1.6.20 S AFETY GUIDE 20 - VIBRATION MEASUREMENTS ON 1.6-7 i REACTOR INTERNALS ;
l 1.6.21 S AFETY GUIDE 21 - MEASURING AND REPORTING OF 1.6-7 i EFFLUENTS FROM NUCLEAR POWER PLANTS 1.7 RESEARCH AND DEVFLOPMENT 1.7-I 1.8 IDENTIFICATION OF AGENTS AND CONTRACTORS 1.8 1 I
- i
1.9 CONCLUSION
S 1.91 1
1.10 REFERENCES 1.10-1 l
! CHAlrrER 2. SITE AND ENVIRONMENT
. 2.1
SUMMARY
2.1-1 :
\
2.2 SITE AND ADJACENT AREAS 2.2-1 i 2.2.1 SITE LOCATION 2.2-1 SITE OWNERSHIP 2.2-1 l 2.2.2 I
2.2.3 SITE ACTIVITIES 2.2-2 l 2.2.4 POPULATION 2.2-2 i
2.2.5 LAND USE 2.2-6 l Annu. sat 2 l Vi 1
l DEFUELED SAFETY ANALYSIS REPORT TABLE OF CONTENTS CHAPTER 2 SITE AND ENVIRONMENT (Continued) l Section lille Eage 2.2.5.1 Current T and Use 2.2-6 2.2.5.2 Future Iand Use 2.2-6 2.2.6 ACCESS AND EGRESS 2.2-11 2.2.7 AIRFIELDS AND MISSILE SITES 2.2-11 2.3 METEOROLOGY 2.3-1 2.
3.1 INTRODUCTION
2.3-1 2.3.1.1 Onerational Meteorolooieni Program 2.3-1 l 2.3.2 CLIMATOLOGICAL STATISTICS 2.3-1 1 2.3.3 ATMOSPHERIC DISPERSION FACTOR 2.3-1 2.4 HYDROLOGY 2.4-1 2.4.1 CHARACTERISTICS OF STREAMS AND LAKES IN 2.4-1 VICINITY 2.4.2 TOPOGRAPHY 2.4-1 2.4.3 TERMINAL DISPOSAL OF STORM RUNOFF 2.4-1 2.4.4 HISTORICAL FLOODING 2.4-1 2.4.5 PREDICTION OF LAND URBANIZATION 2.4-3 2.4.6 GROUNDWATER 2.4-3 2.4.6.1 Occurrence and Movement 2.4-3 2.4.6.2 Water Supoly 2.4-4 2.4.6.3 Onality 2.4-4 2.4.7 WELLS AND BORINGS 2.4-4 2.5 GEOLOGY 2.5-1 2.6 SEISMOLOGY 2.6-1 Amendment 2 vii
DEFUELED SAFETY ANALYSIS REPORT TABLE OF CONTENTS CHAPTER 1 SITE AND ENVIRONMENT (Continued)
Section Iille Eage I
2.7 SOILS 2.7-1 2.8 ENVIRONMENTAL MONITORING PROGR AM 2.8-1
2.9 REFERENCES
2.9-1 CHAPTER 3. REACTOR 3.1 DESIGN BASES 3.1-1 3.1.1 PERFORMANCE OBJECTIVES 3.1-1 3.1.2 LIMITS 3.1-1 3.1.2.1 Core Comnonents 3.1-1 3.2 REACTOR DESIGN 3.2-1 3.3 TESTS ANDINSPECTIONS 3.3-1
3.4 REFERENCES
3.4-1 CHAPTER 4. REACTOR COOLANT SYSTEM
{
4.1 DESIGN BASES 4.1-1 4.1.1 PERFORMANCE OBJECTIVES 4.1-1 4.2 SYSTEM DESCRIPTION AND OPERATION 4.2-1 4.2.1 LEAK DETECTION 4.2-1 4.2.2 COMPONENT FOUNDATIONS AND SUPPORTS 4.2-1 4.2.2.1 Reactor Veel 4.2 1 4.2.2.2 Pressurim 4.2-1 4.2.2.3 Steam Generator 4.2-1 l
4.2.2.4 Piping 4.2-2 4.2.2.5 Pumn and Motor 4.2-2 viii
DEFUELED SAFETY ANALYSIS REPORT TABLE OF CONTENTS l CHAPTER 4. REACTOR COOLANT SYSTEM (Continued)
Section Iilla Eage 4.3 SYSTEM DESIGN EVALUATION 4.31 7 4 1
4.3.1 REACTOR VESSEL CLOSURE 4.3-1 4.4 TESTS ANDINSPECTIONS 4.4-1
4.5 REFERENCES
4.5-1 CHAPTER 5. STRUCTURES AND CONTAINMENT SYSTEM l 5.1 GENERAL 5.1-1 5.1.1 CLASSES OF STRUCTURES 5.1-1 5.1.1.1 Class I 5.1-1 5.1.1.2 Clam II 5.1-1 5.1.1.3 Clam HI .i.1-2 5.1.2 DESIGN LOADS AND STRUCTURAL BEHAVIOR 5.1-2 5.1.2.1 C1na I Structums 5.1-2 5.1.2.1.1 Normal Operations 5.1-2 5.1.2.1.2 Accident and Seismic Conditions 5.1-2 5.1.2.1.3 Missiles 5.1-2 5.1.2.1.4 Separation of Structures and Components 5.1-3 5.1.2.1.5 Seismic Design of Structures 5.1-3 5.1.2.1.6 Wind Loads 5.1-6 5.1.2.1.7 Tornado Loads 5.1-7 5.1.2.1.8 Seismic Design of Equipment, Structures, and Supports 5.1-10 5.1.2.1.9 Design of Foundations and Subgrade Walls 5.1-12 5.1.2.1.10 Buried Tunnels, Piping, and Cables 5.1-13 Amendment 2 ix
l l
DEFUELED SAFETY ANALYSIS REPORT TABLE OF CONTENTS CHAPTER 5. STRUCTURES AND CONTAINMENT SYSTEM (Continued)
Section Tillet P_agn 5.1.2.2 Clau II Sinv tures 5.1-14 5.1.2.3 Cina III Stn=-tures 5.1-14 5.1.3 GOVERNING CODES AND SPECIFICATIONS 5.1-15 1
5.1.4 LOAD COMBINATIONS CRITERIA AND STRUCTURAL 5.1-15 ANALYSIS 5.1.4.1 At Deden Inn <k 5.1-17 5.1.4.2 At Factored Imnda 5.1-17 l 5.2 REACTOR BUTI DING 5.2-1 5.2.1 CONTAINMENT STRUCTURE 5.2-1 5.2.2 INTERIOR CONTAINMENT STRUCTURE 5.2-1 5.3 AUXILIARY BUIT DING 5.3-1 5.3.1 GENERAL DESCRIPTION 5.3-1 5.3.2 DESIGN BASES 5.3-1 5.3.2.1 Design Loads 5.3-1 5.3.2.1.1 Dead Loads 5.3-1 5.3.2.1.2 Live Loads 5.3-2 5.3.2.1.3 Earthquake Loads 5.3-2 5.3.2.1.4 Wind Loads 5.3-2 5.3.2.1.5 Rain Loads 5.3-2 5.3.2.2 Dedgn Criteria 5.3-3 5.3.2.3 Structural Design Analysis 5.3-3 5.4 FUEL STORAGE BUTI DING 5.4-1 5.4.1 GENERAL DESCRIPTION 5.4-1 5.4.2 DESIGN BASES 5.4-1 Amendment 2 X
-. . - . - _ . _~ _. _. ~. ._. - - .
1 1
DEFUELED SAFETY ANALYSIS REPORT l TABLE OF CONTENTS l l
CH APTER 5. STRUCTURES AND CONTAINMENT SYSTEM (Continued) l Section Iille Eage 5.4.2.1 Desien Londa 5.4-1 1
5.4.2.1.1 Dead Loads 5.4-1 l l
5.4.2.1.2 Live Loads 5.4-2 1 5.4.2.1.3 Earthquake Loads 5.4-2 5.4.2.1.4 Wind Loads 5.4-2 i
5.4.2.1.5 Thermal Stresses 5.4-2 1 5.4.2.2 Design Criteria 5.4-3 5.4.2.3 Structural Design Analysis 5.4-4 5.5 OTHER STRUCTURES 5.5-1 5.5.1 'A' NUCLEAR SERVICE SPRAY POND AND PIPE LINES 5.5-1 5.5.2 TURBINE BUILDING 5.5-1 5.5.3 COOLING TOWERS 5.5-1 5.5.4 STORAGE RESERVOIR 5.5-3 5.5.5 NUCLEAR SERVICES ELECTRICAL BUILDING 5.5-3 5.5.5.1 General Descrintion 5.5-3 5.5.6 TRAINING AND RECORDS BUILDING 5.5-4 5.5.7 INTERIM ON-SITE STORAGE BUILDING FOR LOW 5.5-4 LEVEL RADWASTE 5.5.7.1 General Descrintion 5.5-4 5.5.7.2 Desian Basis 5.5-5 5.5.7.2.1 Codes, Standards, and Regulatory Requirements 5.5-6 5.5.7.3 Material Reauirements 5.5-7 5.5.7.4 Structural Reanirements 5.5-8 5.5.8 SOLIDIFICATION BUILDING 5.5-9 Xi
I DEFUELED SAFETY ANALYSIS REPORT TABLE OF CONTENTS CHAPTER 5. STRUCTURER AND CONTAINMENT SYSTEM (Continued)
Rection IiLig pagg 5.5.8.1 General D=vintion 5.5-9 :
5.5.8.2 Design Basis 5.5-9 5.6 MATERIAI 9 AND CONSTRUCTION PRACTICES 5.6-1 ,
5.6.1 CONSTRUCTION ORGANIZATION 5.6-1 5.6.2 CONSTRUCTION SPECIFICATIONS 5.6-1
?
5.6.3 CONSTRUCTION MATERIALS 5.6-2 5.6.3.1 Concmte 5.6-2 5.6.3.1.1 Aggregates 5.6-2 5.6.3.1.2 Cement 5.6-4 5.6.3.1.3 Pozzolan 5.6-4 -
i 5.6.3.1.4 Water and Ice 5.6-4 5.6.3.1.5 Admixtures 5.6-5 5.6.3.1.6 Concrete Mix Design and Testing 5.6-5 5.6.3.1.7 Concrete Production and Testing 5.6-6 5.6.3.2 Reinforcine Steel 5.6-7 5.6.3.2.1 Materials 5.6-7 5.6.3.2.2 Mechanical Splices 5.6-7 5.6.3.2.3 Fabrication and Placement 5.6-10 5.6.3.2.4 Inspection and Testing of Reinforcement 5.6-11 5.6.3.2.5 Inspection and Testing of Cadweld Splices 5.6-11 i 5.6.3.3 Staal Pre-strenine Tendons 5.6-13 5.6.3.4 I iner Plate and Penetration Sleeves 5.6-14 j 5.6.3.5 Penetrations 5.6-14 i xii -
DEFUELEDSAFETY ANALYSIS REPORT TABLE OF CONTENTS CHAPTER 5. STRUCTURES AND CONTAINMENT SYSTEM (Continued)
Section Tillt Eage 5.6.3.6 Structural and MiwHaneous Steel 5.6-14 5.6.3.6.1 Materials 5.6-14 5.6.3.6.2 Fabrication and Emetion 5.6-15 5.6.3.6.3 Inspection and Testing 5.6-15 5.6.3.7 Welder OnnliGcations and Tnenection of Field Welding 5.6-15 5.6.3.7.1 Welding Procedures 5.6'-15 5.6.3.7.2 Welder Qualification 5.6-15 5.6.3.7.3 Welding Inspector Qualifications 5.6-15 5.6.3.7.4 GeneralInspection Procedures 5.6-16 5.6.3.7.5 Inspection of Post Weld Heat Tmatment 5.6-17 5.6.3.7.6 VisualInspection ofWelds 5.6-17 5.6.3.7.7 Nondestructive Testing 5.6-18 5.6.3.7.8 Repairs 5.6-19 5.6.3.7.9 Records 5.6-19 5.7 SEISMIC INSTRUMENTATION 5.7-1
5.8 REFERENCES
5.8-1 CHAPTER 6. SAFETY FEATURES 6.1 GENERAL 6.1-1 6.2 SAFETY FEATURES SYSTEMS I FAKAGE AND 6.2-1 RADIATION CONSIDERATIONS 6.3 POST LOSS-OF-COOLANT ACCIDENT HYDROGEN 6.3-1 CONTROL
6.4 REFERENCES
6.4 1 xiii
DEFUELEDSAFETY ANAI-YSIS REPORT TABLE OF CONTENTS CHAPTER 7. INSTRUMENTATION AND CONTROL j Section litic Eage 7.1 PROTECTION SYSTEMS 7.1-1 7.2 REGULATION SYSTEMS 7.2-1 7.3 INSTRUMENTATION 7.3-1 7.3.1 NON-NUCLEAR PROCESS INSTRUMENTATION 7.3-1 7.3.1.1 System D-sinn 7.3-1 7.3.2 IN-CORE MONITORING SYSTEM 7.3-1 7.4 OPERATING CONTROL STATIONS 7.4-1 7.4.1 GENERAL LAYOIJr 7.4-1 7.4.2 INFORMATION DISPLAY AND CONTROL FUNCTION 7.4-2 7.4.2.1 Console and Panel 12y-out 7.4-7 7.4.3
SUMMARY
OF ALARMS 7.4-11 7.4.4 COMMUNICATION 7.4-11 7.4.5 OCCUPANCY 7.4-11 7.4.6 AUXILIARY CONTROL STATIONS 7.4-12 7.4.7 SAFETY CONSIDERATIONS 7.4-12 7.4.8 SYSTEM EVALUATION 7.4-12 7.4.8.1 Control Room Av=Hability 7.4-12
7.5 REFERENCES
7.5-1 CHAMER R. EI ECTRICAL SYSTEMS 8.1 DESIGN BASES 8.1-1 8.2 FI ECTRICAL SYSTEM DESIGN 8.2-1 8.2.1 NETWORK INTERCONNECTIONS 8.2-1 Amendment 1 xiv
DEFUELED SAFETY ANALYSIS REPORT TABLE OF CONTENTS CHAPTER 8. FI.FCTRICAL SYSTEMS (Continued)
Section Iille P_ age 1
8.2.1.1 Single T.ine Dineram 8.2-1 8.2.1.2 Reliability Considerations 8.2-1 8.2.2 STATION DISTRIBUTION SYSTEM 8.2-2 8.2.2.1 Auxiliary Trnnsformers 8.2-2 8.2.2.2 6900-Volt Anriliarv Svstem 8.2-3 l 8.2.2.3 4160-Volt Auxiliary System 8.2-3 l 8.2.2.4 480-Volt Anviliarv System 8.2-3 8.2.2.5 125-Volt d-c Svstem 8.2-3 8.2.2.6 120. Volt a-c Unregulated Power System 8.2-4 8.2.2.7 120-hlt a-c Uninterruptible Power Supply System 8.2-4 8.2.2.8 i ighting 8.2-5 8.2.2.9 Evaluation of the Phveical Layout of Electrical Distribution 8.2-5 System Eauipment 8.
2.3 DESCRIPTION
OF POWER SOURCES 8.2-7 8.2.3.1 Off-site Power 8.2-7 l 8.2.3.2 On-site Power 8.2-7 8.2.3.3 220/230-kV Switchvard Control Power 8.2-7 8.3 TESTS AND INSPECTIONS 8.3-1 8.3.1 IN-SERVICETESTS ANDINSPECTIONS 8.3-1
8.4 REFERENCES
8.4-1 CHAPTER 9. AUXILIARY AND EMERGENCY SYSTEMS 9.1 GENERAL 9.1-1 9.2 COOLING WATER SYSTEMS 9.2-1 Amendment 1 xv
DEFUELED SAFETY ANALYSIS REPORT TABLE OF CONTENTS CHAM'ER 9. AUXII TARY AND EMERGENCY SYSTEMS (Continued)
Section M Pagg CONDENSER CIRCULATING WATER SYSTEM 9.2-1 9.2.1 9.2.1.1 System Dem rintion 9.2-1 9.2.1.2 Desip Dat2 9.2-2 9.2.2 PLANT COOLING WATER SYSTEM 9.2-2 9.2.2.1 System Deacrintion 9.2-2 9.2.2.2 Design Data 9.2-3 9.2.3 COMPONENT AND TURBINE PLANT COOLING WATER 9.2-3 SYSTEM 9.2.3.1 Svstem Daa rintion 9.2-3 9.2.3.2 Design Data 9.2-5 9.3 DECAY HEAT REMOVAL SYSTEM 9.3-1 9.3.1 Borated Water Storage Tank 9.3-1 9.4 SPENT FUEL COOLING SYSTEM 9.4 1 9.4.1 DESIGN BASES 9.4-1 9.4.2 SYSTEM DESCRIFFION 9.4-4 9.4.2.1 Codes and standards 9.4-5 9.4.2.2 Material Camnatihility 9.4-5 9.4.2.3 Comnonent Desip 9.4-5 9.4.2.3.1 Piping and Valves 9.4-5 9.4.2.3.2 Pumps 9.4-5 9.4.2.3.3 Heat Exchanger 9.4-5 9.4.2.3.4 Filters and Ion Exchanger 9.4-6 9.4.2.4 1 eakage cansideranians 9.4-6 Amendment!
xvi
DEFUELED SAFETY ANALYSIS REPORT TABLE OF CONTENTS l CHAPTER 9. AUXII.IARY AND EMERGENCY SYSTEMS (Continued) l Section Ilile Eage
- 9.4.2.5 Failure Condderations 9.4-6 9.4.2.6 Onerating Conditions 9.4-6 l
9.5 STATION VENTTI ATION SYSTEMS 9.5-1 9.5.1 DESIGN BASES 9.5-1
- 9.5.1.1 Reactor Building 9.5-1 9.5.1.1.1 General Conditions 9.5-1 j 9.5.1.1.2 Sizing 9.5-1 9.5.1.2 Auxiliary Building 9.5-1 9.5.1.3 Nuclear Service Electrient Building (NSEB) 9.5-2 9.5.2 SYSTEM DESCRIFFION 9.5-2
^
9.5.2.1 Reactor Building 9.5-2 9.5.2.1.1 General 9.5-2 9.5.2.1.2 Purge System 9.5-2 9.5.2.1.3 Codes, Standards, Tests 9.5-7 9.5.2.2 Auxiliarv Building 9.5-7 1
9.5.2.2.1 General 9.5-7 f
9.5.2.2.2 Control Room and Technical Support Center 9.5-8 9.5.2.2.3 Radiochemical and Service Areas 9.5-9 9.5.2.2.4 Radwaste and Fuel Storage Areas 9.5-9 9.5.2.2.5 Electrical Equipment, Switchgear, and AC/DC Panel Rooms 9.5-10 9.5.2.2.6 Coolant and Miscellaneous Waste Tanks 9.5-11 9.5.2.2.7 Emergency Pump Rooms 9.5-11 9.5.2.2.8 Auxiliary Building Exhaust Air Filtration System 9.5-11 Amendment 2 xvii
l I
DEFUELEDSAFETY ANALYSIS REPORT l
TABLE OF CONTENTS i CHAPTER 9. AUXII.IARY AND EMERGENCY SYSTEMS (Continued)
Section Title Eage 9.5.2.2.9 Communication Room 9.5-13 l 9.5.2.2.10 - Chilled Water System 9.5-13 l 9.5.2.3 Nucient Service Flacerient Buildina (NSFBI 9.5-13 9.5.2.4 Batterv Buildina 9.5-14 l 1
l l 9.5.2.5 Interim On-nite Stor==e Buildia! GOSB1 9.5-14 )
i 9.5.2.6 Switchyard control Buildia! 9.5-14 9.5.2.7 Codes Standards. and Te=t= 9.5-15 l
9.6 FUEL HANDIJNG SYSTEM 9.6-1 9.6.1 DESIGN BASES 9.6-1 !
i 9.6.1.1 General Svetem Function 9.6-1 9.6.1.2 Snent Fuel Stormon Pool 9.6-1 ,
i 9.6.1.3 Snent Fuel Pool Water Chemintrv 9.6-2 9.6.1.4 Fuel Transfer Tuk 9.6-2 9.6.1.5 Fuel Handline Enninment 9.6-3 9.6.2 SYSTEM DESCRIPTION AND EVALUATION 9.6-3 9.6.2.1 Handlina Snent Fuel Anaemblies 9.6-3 9.6.2.2 Handling and Londine Snent Fuel Casks 9.6-3 9.6.2.3 Crane Use in Fuel Handlina 9.6-4 9.6.2.3.1 Design 9.6-4 9.6.2.3.2 Evaluation 9.6-6 9.6.2.4 Safety Provisions 9.6-9 9.7 OTHER AUXILIARY SYSTEMS 9.7-1 9.7.1 FIRE PROTECTION SYSTEM 9.7-1 Amendment 2 xviii
DEFUELED SAFETY ANALYSIS REPORT TABLE OF CONTENTS CHAPTER 9. AUXILIARY AND EMERGENCY SYSTEMS (Continued)
Section T111C EsLgC 9.7.1.1 Design h= 9.7-1 9.8 PLANT COMPRESRFn SERVICE GAS SYSTEM 9.8-1
9.9 REFERENCES
9.9-1
- 10. STEAM AND POWER CONVERSION SYSTEM 10.1 DESIGN BASES 10.1-1 4
CHANER 11. RADIOACITVE WASTE AND RADIATION PROTECTION 11.1 SOURCE TERM 11.1-1 11.1.1 RADIONUCLIDEINVENTORY 11.1-1 11.1.1.1 Snent Fuel Assemblies 11.1-1 11.1.1.2 Reactor Vennel and Internals and Concrete Primary Shield 11.1-2 11.1.1.3 Pinnt Svetems 11.1-2 11.2 LIOUID WASTE TREATMENT SYSTEMS 11.2-1 11.2.1 COOLANT RADWASTE AND REACTOR COOLANT DRAIN 11.2-1 SYSTEM 11.2.1.1 Functions 11.2-1 11.2.1.2 System Derription 11.2-2 11.2.2 MISCELLANEOUS LIQUID RADWASTE SYSTEM 11.2-2 11.2.2.1 Function 11.2-2 11.2.2.2 System Derription 11.2-2 11.2.3 WASTE WATER DISPOSAL 11.2-12 11.2.3.1 Plant Elbent 11.2-12 11.2.3.2 Normal Radioactive Discharge 11.2-13 1 Amendment 2 XiX
i l I I
{ DEFUELED SAFETY ANALYSIS REPORT TABLE OF CONTENTS CHAPTER 11. RADIOACTIVE WASTE AND RADIATION PROTECTION (Continued)
Section Titic Ease 11.2.3.3 Ofr-Normal R ndinactive Di= hnree 11.2-14 11.2.3.4 Non-radioactive Wnete Water 11.2-15 11.2.4 OPERATION, TESTING, AND INSPECTION 11.2-15 11.2.5 SYSTEM EVALUATION 11.2-16 11.2.6 PROCESSING WET RADIOACTIVE WASTES INTO SOLID 11.2-17 RADIOACTIVE WASTE 11.2.6.1 Solidification and Dewntering of Wet Radioactive Wastes 11.2-17 11.2.6.2 Drving of Wet R ndianctive Wastes 11.2-19 11.2.6.2.1 Design Basis 11.2-19 l 11.3 GASEOUS WASTE MANAGEMENT SYSTEM 11.3-1 11.3.1 DESIGN BASIS 11.3-1 11.3.2 SYSTEM DESCRIPTION 11.3-1 11.3.3 HYDROGEN GAS MIXTURES 11.3-2 11.3.4 OPERATION,1 tsiING, AND INSPECTION 11.3-2 11.3.5 RADIOACTIVE RELEASES 11.3-3 11.3.5.1 Pathways 11.3-3 11.3.5.2 Secondarv Plant Contamination 11.3-3 11.3.5.3 Interim On-site Storage Building (IOSB) 11.3-3 11.3.6 METHOD OF ASSESSMENT 11.3-4 11.3.6.1 Plume Ernosure (Noble Gnei 11.3-4 11.3.6.2 Food Pathway 11.3-5 l 11.3.7 EVALUATION OF WASTE DISCHARGE 11.3-5 11.4 SOLID WASTE MANAGEMENT SYSTEM 11.4-1 l Amendment!
xx 1
DEFUELED SAFETY ANALYSIS REPORT TABLE OF CONTENTS CHAFTER 11. RADIOACITVE WASTE AND RADIATION PROTECTION (Continued)
Section Tida h 11.4.1 DESIGN BASIS 11.4-1 11.4.1.1 Renorts 11.4-2 11.4.2 SYSTEM DESCRIPTION 11.4-2 11.4.2.1 Drv Solid Waste Disnosal Svetem/Prnem 11.4-2 11.4.2.2 Concentrated Lianid Wnste Disnosal System /Proceaa 11.4-3 l 11.4.2.3 Snent Resin Disnosal Svstem/ Proc ~ l1.4-3 11.4.2.4 Filter Disnosal Proem 11.4-3 11.4.2.5 Solid Radwaste Storage 11.4-4 11.5 R ADIOAcrIVE WASTE. EFFL UENT CONTROL. AND 11.5-1 ENVIRONMENTAL MONITORING PROGRAMS 11.5.1 DESIGN BASIS 11.5-1 i 11.5.2 OFF-SITE DOSE CALCULATION MANUAL (ODCM) 11.5-1 11.5.2.1 Lianid Discharge Pathway 11.5-1 l 11.5.2.2 Gneous Discharge Pathway 11.5-2 11.5.3 PROCESS CONTROL PROGRAM (PCP) 11.5-2 11.5.4 RADIOLOGICAL ENVIRONMENTAL MONITORING 11.5-2 11.5.4.1 Pre-oneration21 REMP 11.5-2 11.5.4.1.1 Pre-onerational Ernosure Fatimation 11.5-8 11.5.4.2 oft-site Post-onerationn1 REMP 11.5-8 11.5.4.2.1 Post-operational REMP Sampling Frequency 11.5-10 11.5.4.2.2 REMP Sample Types 11.5-10 11.5.4.2.3 REMP Sample Statistical Analysis 11.5-11
'11.5.4.3 Emuent and Waste Disnosal Environmental Renorts 11.5-11 Amendment 1 xxi
DEFUELED SAFETY ANALYSIS REPORT TABLE OF CONTENTS CHAMER 11. RADIOACTIVE WASTE AND RADIATION PROTECTION (Continued)
Section Title Eage 11.6 ENSURING THAT OCCUPATIONAL RADIATION 11.6-1 EXPOSURR9 ARE AS LOW AS IS REASONABLY ACHIEVABI E (ALAR A) 11.6.1 ALARA POLICY CONSIDERATIONS 11.6-1 11.6.2 ALARA DESIGN CONSIDERATIONS 11.6-3 11.6.3 ALARA OPERATIONAL CONSIDERATIONS 11.6-3 11.7 RADIATION SOURCES 11.7 1 11.7.1 CONTAINED SOURCES 11.7-1 11.8 R ADIATION PROTECTION DESIGN FEATUR ES 11.8 1 11.8.1 FACILITY DESIGN FEATURES 11.8-1 11.8.2 SHIELDING 11.8-1 11.8.2.1 Design Criteria 11.8-2 11.8.2.2 Rndintion Zone Clansifications 11.8-2 11.3.2.3 Descrintion of Shielding 11.8-2 11.8.2.3.1 Primary Shield 11.8-2 11.8.2.3.2 Secondary Shield 11.8-3 11.8.2.3.3 Reactor Building Shield 11.8-3 11.8.2.3.4 Control Room Shield 11.8-3 11.8.2.3.5 Auxiliary Shield 11.8-3 11.8.2.3.6 Spent Fuel Shielding 11.8-3 11.8.2.4 Shielding Materinta 11.8-5 1
11.8.3 VENTILATION 11.8-5 11.8.4 RADIATION MONITORING SYSTEM 11.8-5 i
Amendment 1 xxii l l
DEFUELED SAFETY ANALYSIS REPORT TABLE OF CONTENTS CHAPTER 11. RADIOACTIVE WASTE AND RADIATION PROTECTION (Continued)
Section Iillg- Eage 11.8.4.1 Desien Criteria 11.8 5 11.8.4.2 System Deacrintion 11.8-6 11.8.4.2.1 Area Radiation Monitors 11.8-7 11.8.4.2.2 Process /Efiluent Radiation Monitors 11.8-7 ,
1 11.9 DOSE ASSFSRMENT 11.9-1 ,
i 11.9.1 PERSONNEL MONITORING 11.9-1 11.9.2 PERSONNEL EXPOSURE RECORD SYSTEM 11.9-2 l 11.9.3 MEDICAL EXAMINATION PROGRAM 11.9-2 11.10 RADIATION PROTECTION PROGRAM 11.10-1 11.10.1 ORGANIZATION 11.10-1 11.10.2 EQUIPMENT, INSTRUMENTATION, AND FACILITIES 11.10-3 l 11.10.2.1 Personnel Protective Eauinment 11.10-3 11.10.2.2 Radiation Protection Instrumentation 11.10-3 11.10.2.3 Facilitles 11.10-4 11.10.3 RADIATION PROTECTION PROCEDURES 11.10-5 11.10.3.1 Procedures 11.10-5 11.10.3.2 Radiation Work Permit Pomdure 11.10-6 l
11.10.4 PERIODIC PERSONNEL EXPOSURE REPORTING 11.10-6 11.11 REFERENCES 11.11-1 CHAPTER 12. CONDUCT OF OPERATIONS 12.1.1 NUCLEAR ORGANIZATION 12.1-1 12.1.2 PLANT PERSONNEL RESPONSIBILITIES AND 12.1-1 AUTHORITIES Amendment 1 xxiii
4 DEFUELED SAFETY ANALYSIS REPORT TABLE OF CONTENTS CHAPTER 12. CONDUCT OF OPERATIONS (Continued)
Section lille East 12.1.2.1 DAGM Nuclear 12.1-2
- 12.1.2.2 Onerating Shill Crews 12.1-6
- 12.1.2.3 Succminn of Rennnnaihility 12.1-8 l l 12.1.3 QUALIFICATIONS OF NUCLEAR PLANT PERSONNEL 12.1-8 -
l 12.2 PERSONNET. TR AINING 12.2-1 12.2.1 TRAINING PROGRAMS 12.2-1 12.2.2 CERTIFIED FUEL HANDLER TRAINING PROGRAMS 12.2-2 12.2.2.1 Structure ofInitini(htification Procram 12.2-3 12.2.2.2 Evaluation 12.2-3 12.2.3 STRUCTURE OF CERTIFIED FUEL HANDLER 12.2-4 PROFICIENCY PROGRAM 12.2.3.1 Evaluatinn 12.2-5 12.2.4 LICENSING TRAINING PROGRAM 12.2-5 12.2.5 MAINTENANCE TRAINING PROGRAM 12.2-6 12.2.5.1 Initial Training 12.2-7 12.2.5.2 Continuine Trainine
. . 12.2-7 12.2.6 SITE SUPPORT TRAINING PROGRAMS 12.2-7 !
12.2.6.1 Initial Training 12.2-7 12.2.6.2 Continuing Trainine 12.2-8 12.3 EMERGENCY PLANNING 12.3-1 12.4 REVIEW AND AUDITOFOPERATIONS 12.4-1 12.5 PLANTPROCEDURES AND PROCESS STANDARDS 12.5-1 12.5.1 PROCEDURES 12.5-1 Amendment 1 1
Xilv l
l
____ ~ _ ___ ._._.__ _ _ _ _ _ .. _ . ._ _ _ _ . . _ ._ _. __ _ . _ _ . . ._ _ . _ _ _ . - .
DEFUELED SAFETY ANALYSIS REPORT TABLE OF CONTENTS CHAPTER 11 CONDUCT OF OPERATIONS (Continued)
Section Tgle Eage 12.5.1.1 Conformaw with surety Guide 33 12.5-1 12.5.1.2 Prennration of Prncedum 12.5-1 12.5.1.2.1 Procedure Changes 12.5-2 12.5.1.3 Conduct of Oneratiana 12.5-2 12.5.1.3.1 Shut down Control Room Operator Authority 12.5-2 12.5.1.3.2 Certified Fuel Handler Authority 12.5-3 l
12.5.1.3.3 Activities AITecting Plant Operations or Indications 12.5-3 During the PDM 12.5.1.3.4 Manipulation of Facility Controls 12.5-4 12.5.1.3.5 Responsibility for Fuel Handling Operations 12.5-4 12.5.1.3.6 ReliefofDuties 12.5-4 12.5.1.3.7 Equipment Control 12.5-4 12.5.1.3.8 Surveillance Testing Schedule 12.5-4 12.5.1.3.9 Log Books 12.5-4 12.5.2 OPERATING AND OTHER PROCEDURES 12.5-5 12.5.2.1 Onerating Praceduren 12.'5-5 12.5.2.1.1 System Procedmes 12.5-5 12.5.2.1.2 Special Test Procedures 12.5-6 12.5.2.1.3 Annunciator Alarm Response Procedures 12.5-6 12.5.2.1.4 Casualty Procedures 12.5-6 12.5.2.2 Other Procedum 12.5-6 12.5.2.2.1 Maintenance Procedures 12.5-6 12.5.2.2.2 Instrument and Control (I&C) Procedures 12.5-7 Amendment 1 xxv
DEFUELED SAFETY ANALYSIS REPORT TABLE OF CONTENTS CHAPTER 12. CONDUCT OF OPERATIONS (Continued)
Section Iitic Eage 12.5.2.2.3 Surveillance Procedures 12.5-7 12.5.2.2.4 Chemistry Procedums 12.5-7 12.5.2.2.5 Radioactive Waste Management Procedures 12.5-8 12.5.2.2.6 Radiation Protection Procedures 12.5-8 l
12.5.2.2.7 Security Plan 12.5-8 12.5.2.2.8 Emergency Plan and Implementing Procedures 12.5-8 12.5.2.2.9 Fire Protection Procedums 12.5-9 12.5.2.2.10 Quality Assurance 12.5-9 12.5.2.2.11 Certified Fuel Handler and Non-Certified Operator 12.5-9 Training Programs l
12.5.3 PROCESS STANDARDS 12.5-9 12.6 INDUSTRIAL SECURITY 12.6-1 12.7 RECORDS 12.7-1 1
12.7.1 OPERATING RECORDS 12.7-1 l 12.7.2 ADMINISTRATIVE RECORDS 12.7-1 12.7.3 MAINTENANCE RECORDS 12.7-2 l 12.7.4 !
HEALTH PHYSICS RECORDS 12.7-2 '
12.7.5 OTHER RECORDS 12.7-2
12.8 REFERENCES
12.8-1 CHAPTER 13. INITIAL TESTS AND OPERATIONS
13.1 INTRODUCTION
13.1-1 Amendment 1 XXVI
DEFUELED SAFETY ANALYSIS REPORT TABLE OF CONTENTS CHAPTER 14. SAFETY ANALYSIR Section lilla Eagt 3 14.1 ACCIDENTS CONSIDERED CREDIBT E IN THE
' 14.1 1 PERMANENTLY DEFUEI FD MODE (PDM) j 14.1.1 FUEL HANDLING ACCIDENT 14.1-1 14.1.1.1 Annlysis and Results 14.1 1 14.1.2 LOSS OF OFF-SITE (A-C) POWER
- 14.1-2 14.2 R EFERENCES
- 14.2-1 l t
4 1
't I
i i
i s
Amendment 2 xxvii
TABLE OF CONTENTS Section Title g
- 1. INTRODUCTION AND
SUMMARY
1.1-1 1
1.1 INTRODUCTION
1.1-1 1 1.2 DESIGN HIGHLTCHTS 1.2-1 I
i 1.2.1 SITE CHARACTERISTICS 1.2-1 1.2.2 POWER LEVEL 1.2-1 1.2.3 REACTOR BUILDING 1.2-1 l
1.2.4 SAFETY FEATURES 1.2-1 1.2.5 ELECTRICAL SYSTEMS AND EMERGENCY POWER 1.2-1 1.2.6 SEISMIC DESIGN QUALITY ASSURANCE 1,2-2
)
1.2.7 FUEL STORAGE DUILDING 1.2-7 1.3 DESIGN CHARACTERISTICS 1.3-1 )
l 1.3.1 DESIGN PARAMETERS 1. ,,;-l 1.3.2 SIGNIFICANT DESIGN REVISIONS 1.3-1 1.4 PRINCIPAL DESIGN CRITEk12 - ORIGINAL 1.4-1 1.5 PRINCIPAL DESIGN CRITERIA - PERMANENTLY DEFUELED 1.5-1 EDDE 3.5.1 CRITERION 1 - QUALITY STANDARDS AND RECORDS 1.5-1 1.5.2 CRITERION 2 - DESIGN BASIS FOR PROTECTION AGAINST 1.5-2
- NATURAL PHENOMENA 1.5.3 CRITERION 3 - FIRE PROTECTION 1.5-3 1.5.4 CRITERION 4 - ENVIRONMENTAL AND MISSILE DESIGN BASES 1.5-5 1.5.5 CRITERION 5 - SHARING OF STRUCTURES, SYSTEMS, AND 1.5-5 COMPONENTS 1.5.4 CRITERION 10 - REACTOR DESIGN 1.5-5 i
1.5.7 CRITERION 11 - REACTOR INHERENT PROTECTION 1.5-6 l Amenchment 2 1-1
TABLE OF CONTENTS (Continued) section Title Page 1.5.8 CRITERION 12 - SUPPRESSION OF REACTOR POWER 1.5-6 f OSCILLATIONS 1.5.9 CRITERION 13 - INSTRUMENTATION AND CONTROL 1.5-6 l 1.5.10 CRITERION 14 - REACTOR COOLANT PRESSURE BOUNDARY 1.5-6 1.5.11 CRITERION 15 - REACTOR COOLANT SYSTEM DESIGN 1.5-6 1.5.12 CRITERION 16 - CONTAINMENT DESIGN 1.5-7 l 1.5.13 CRITERION 17 - ELECTRIC POWER SYSTEMS 1.5-7 l
1.5.14 CRITERION 18 - INSPECTION AND TESTING OF ELECTRICAL 1.5-8 POWER SYSTEMS 1.5.15 CRITERION 19 - CONTROL ROOM 1.5-9 1.5.16 CRITERION 20 - PROTECTION SYSTEM FUNCTIONS 1.5-10 l 1.5.17 CRITERION 21 - PROTECTION SYSTEM RELIABILITY AND 1.5-10 TESTABILITY l l
1.5.18 CRITERION 22 - PROTECTION SYSTEM INDEPENDENCE 1.5-10 1.5.19 CRITERION 23 - PROTECTION SYSTEM FAILURE MODES 1.5-10 1.5.20 CRITERION 24 - SEPARATION OF PROTECTION AND CONTROL 1.5-10 SYSTEMS 1.5.21 CRITERION 25 - PROTECTION SYSTEM REQUIREMENTS FOR 1.5-10
, REACTIVITY CONTROL MALFUNCTIONS 1.5.22 CRITERION 26 - REACTIVITY CONTROL SYSTEM REDUNDANCY 1.5-11 l AND CAPABILITY l
1.5.23 CRITERION 27 - COMBINED REACTIVITY CONTROL SYSTEMS 1.5-11 I l
CAPABILITY 1.5.24 CRITERION 28 - REACTIVITY LIMITS 1.5-11 l 1.5.25 CRITERION 29 - PROTECTION AGAINST ANTICIPATFD 1.5-11 OPERATIONAL OCCURRENCES 1.5.26 CRITERION 30 - QUALITY OF REACTOR COOLANT PRESSURE 1.5-11 PRESSURE BOUNDARY Amendment 2 1-11
TABLE OF CONTENTS (Continued)
Section Title Rag.a 1.5.27 CRITERION 31 - FRACTURE PREVENTION OF REACTOR 1.5-11 COOLANT PRESSURE BOUNDARY t
1.5.28 CRITERION 32 - INSPECTION OF REACTOR COOLANT 1.5-11 i PRESSURE BOUNDARY 1.5.29 CRITERION 33 - REACTOR COOLANT MAKEUP 1.5-12 l 1.5.30 CRITERION 34 - RESIDUAL HEAT REMOVAL 1.5-12 l ,
1.5.31 CRITERION 35 - EMERGENCY CORE COOLING 1.5-12 l 1.5.32 CRITERION 36 - INSPECTION OF EMERGENCY CORE 1.5-12 l COOLING SYSTEM I 1.5.33 CRITERION 37 - TESTING OF EMERGENCY CORE COOLING 1.5-12 SYSTEM l
1.5.34 CRITERION 38 - CONTAINMENT HEAT REMOVAL 1.5-12 l
1.5.35 CRITERION 39 - INSPECTION OF CONTAINMENT HEAT 1.5-12 '
REMOVAL SYSTEM {
1 1.5.36 CRITERION 40 - TESTING OF CONTAINMENT HEAT REMOVAL 1.5-12 l SYSTEM I 1.5.37 CRITERION 41 - CONTAINMENT ATMOSPHERE CLEANUP 1.5-13 l 1.5.38 CRITERION 42 - INSPECTION OF CONTAINMENT ATMOSPHERE 1.5-13 l CLEANUP 1.5.39 CRITERION 43 - TES' TING OF CONTAINMENT ATMOSPHERE 1.5-13 l CLEANUP SYSTEMS 1.5.40 CRITERION 44 - COOLING WATER 1.5-13 l 1.5.41 CRITERION 45 - INSPECTION OF COOLING WATER SYSTEM 1.5-14 l 1.5.42 CRITERION 46 - TESTING OF COOLING WATER SYSTEM 1.5-14 1.5.43 CRITERION 50 - CONTAINMENT DESIGN BASIS 1.5-14 1.5.44 CRITERION 51 - FRACTURE PREVENTION CF CONTAINMENT 1.5-15 l '
PRESSURE BOUNDARY 1.5.45 CRITERION 52 - CAPABILITY FOR CONTAINMENT LEAKAGE 1.5-15 l )
RATE TESTING Amendment 2 l
1-iii
a TABLE OF CONTENTS (Continusd)
$ Section Title Eggg l 1.5.46 CRITERION 53 - PROVISIONS FOR CONTAINMENT TESTING 1.5-15 l AND INSPECTION 1.5.47 CRITERION 54 - PIPING SYSTEMS PENETRATING 1.5-15 CONTAINMENT 1.5.48 CRITERION 55 - REACTOR COOLANT PRESSURE BOUNDARY 1.5-15
- PENETRATING CONTAINMENT
- 1.5.49 CRITERION 56 - PRIMARY CONTAINMENT ISOLATION 1.5-15
^
1.5.50 CRITERION 57 - CLOSED SYSTEM ISO'ATION VALVES 1.5-15 1.5.51 CRITERION 60 - CONTROL OF RELEASCS OF RADIOACTIVE 1.5-16 l MATERIALS TO THE ENVIRONMENT 1.5.52 CRITERION 61 - FUEL STORAGE AND HANDLING AND 1.5-17 RADIOACTIVITY CONTROL 1.5.53 CRITERION 62 - PREVENTION OF CRITICALITY IN FUEL 1.5-18 l STORAGE AND HANDLING 1.5.54 CRITERION 63 - MONITORING FUEL AND WASTE STORAGE 1.5-18 ,
1.5.55 CRITERION 64 - MONITORING RADIOACTIVITY RELEASES 1.5-19 1.6 COMPARISON WITH SAFETY GUIDES 1.6-1 1.6.1 SAFETY GUIDE 1 - NET POSITIVE SUCTION HEAD FOR 1.6-1 EMERGENCY CORE COOLING AND CONTAINMENT HEAT REMOVAL SYSTEM PUMPS 1.6.2 SAFETY GUIDE 2 - THERMAL SHOCK TO REACTOR 1.6-1 PRESSURE VESSELS i
1.6.3 SAFETY GUIDE 3 - ASSUMPTIONS USED FOR EVALUATING 1.6-1 THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF A ,
LOSS-OF-COOLANT ACCIDENT FOR BOILING WATER REACTORS 1.6.4 SAFETY GUIDE 4 - ASSUMPTIONS USED FOR EVALUATING 1.6-1 l THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF A LOSS-OF-COOLANT ACCIDENT FOR PRESSURIZED ,
WATER REACTORS 1.6.5 SAFETY GUIDE 5 - ASSUMPTIONS USED FOR EVALUATING 1.6-1 THE POTENTIAL CONSEQUENCES OF A STEAM LINE BREAK ACCIDENT FOR BOILING WATER REACTORS Amendment 2 1-iv
TABLE OF CONTENTS (Continued)
Section Title Engg 1.6.6 SAFETY GUIDE 6 - INDEPENDENCE BETWEEN REDUNDANT 1.6-1 STANDBY (ON-SITE) POWER SOURCES AND THEIR DISTRIBUTION SYSTEMS 1.6.7 SAFETY GUIDE 7 - CONTROL OF COMBUSTIBLE GAS 1.6-2 CONCENTRATIONS IN CONTAINMENT FOLLOWING A LOSS-OF-COOLANT ACCIDENT 1.6.8 SAFETY GUIDE 8 - PERSONNEL SELECTION AND 1.6-2 TRAINING 1.6.9 SAFETY GUIDE 9 - SELECTION OF DIESEL GENERATOR 1.6-2 SET CAPACITY FOR STANDBY POWER SUPPLIES 1.6.10 SAFETY GUIDE 10 - MECHANICAL (CADWELD) SPLICES 1.6-2 IN REINFORCING BARS OF CONCRETE CONTAINMENT 1.6.11 SAFETY GUIDE 11 - INSTRUMENT LINES PENETRATING 1.6-2 PRIMARY REACTOR CONTAINMENT 1.6.12 SAFETY GUIDE 12 - INSTRUMENTATION FOR 1.6-2 EARTHQUAKES 1.6.13 SAFETY GUIDE 13 - FUEL STORAGE FACILITY 1.6-3 DESIGN BASIS 1.6.14 SAFETY GUIDE 14 - REACTOR COOLANT PUMP FLY- 1.6-4 NHEEL INTEGRITY 1.6.15 SAFETY GUIDE 15 - TESTING OF REINFORCING 1.6-5 BARS FOR CONCRETE STRUCTURES 1.6.16 SAFETY GUIDE 16 - REPORTING OF OPERATING 1.6-5 l INFORMATION 1.6.17 SAFFTY GUIDE 17 - PROTECTION AGAINST INDUSTRIAL 1.6-5 SABOTAGE 1.6.18 SAFETY GUIDE 18 - STRUCTURAL ACCEPTANCE TEST 1.6-7 FOR CONCRETE PRIMARY REACTOR CONTAINMENTS 1.6.19 SAFETY GUIDE 19 - NONDESTRUCTIVE TESTING OF 1.6-7 PRIMARY CONTAINMENT LINERS 1.6.20 SAFETY GUIDE 20 - VIBRATION MEASUREMENTS ON REACTOR 1.6-7 INTERNALS Amendment 1 1-v
i
, l I
i TABLE OF CONTENTS (Continuad) I 4
i Section Title Eagg )
i 1.6.21 SAFETY GUIDE 21 - MEASURING AND REPORTING OF 1.6-7 !
EFFLUENTS FROM NUCLEAR POWER PLANTS
, 1.7 RESEARCH AND DEVELOPMENT 1.7-1 3
1.8 IDENTIFICATION OF ACRNTS AND CONTRACTORS 1.8-1
1.9 CONCLUSION
S 1.9-1 a
j 1.10 REFERENCES 1.10-1 I
1-vi
l LIST OF FIGURES Flouren Title 1.1-1 Sacramento Municipal Utility District Service Area Boundary 1.1-2 Rancho Seco General Plant Arrangement l
l l
l Amendment 2 1-vii
-. . - - - . . . - _ . - -. . - . - ~ - - -_- ..--~ - ~ . _ . - - - - . - . - . . _ ~ -
RANCHO SECO DEFUELED SAFETY ANALYSIS REPORT i
DSAR CHAPTER 1. INTRODUCTION l
1.1 INTRODUCTION
r l The Defueled Safety Analysis Report (DSAR) represents the licensing
} basis for the operation of the shut down Rancho Seco nuclear facility in l the Permanently Defueled Mode (PDM). The DSAR reflects:
{, !
i 1. The Possession-Only License (POL) status of the Rancho Seco ,
I nuclear facility, i .
1 NRC approved license amendments, exemptions, and waivers that 2.
were granted based on the permanently defueled condition of ,
the Rancho Seco nuclear reactor, and j 3. The NRC order approving the Rancho Seco Decommissioning Plan j (RSDP) and authorizing decommissioning of the Rancho Seco nuclear facility.
a 4
Amendment No. 8) as the primary licensing basis document applicable to q
the operation of Rancho Seco in the PDM.
l This DSAR contains the changes to the licensing basis information and l analyses submitted to the NRC in the original Final Safety Analysis j
l Report (FSAR) in 1971, and reflects the changes made since Rancho Seco
- permanently shut down reactor operations on June 7, 1989. The i Sacramento Municipal Utility District (hereinaf ter referred to as SMUD '
- or the District) will update the DSAR in accordance with the methodology l specified in 10 CFR 50.71(e), except that the frequency of updates shall be at least every two years. Revisions to the DSAR will be numbered, i starting with Amendment 1. DSAR revisions shall include a page change identification table and change indicator lines in the right margin in ;
. the area where a change occurred. !
).
The purpose of the DSAR is to provide a Safety Analysis Report that
- supports the long-term safe storage and handling of irradiated fuel ;
j assemblies. The DSAR provides assurance that, based on the spent fuel 1 assembly storage and handling systems and components described herein, i and the administrative controls and programs required in the PDM, no j undue risk to the public health and safety will occur during normal PDM ,
! operations and postulated accident conditions.
4 i
f Amendment 2 1.1-1 i
I l
l
1.1 INTRODUCTION
(Continued)
)
The SMUD service area is shown on Figure 1.1-1. The Rancho Seco general plant arrangement, with the major nuclear facility structures indicated, )
is shown on Figure 1.1-2.
l To update the history of major plant operation and licensing related '
actionsi the District provides the following information: l
- l. Rancho Seco initially went critical on September 16, 1974, -
and began commercial operation on April 18, 1975.
- 2. Following approval of a public referendum on June 6, 1989, SMUD permanently shut down Rancho Seco on June 7, 1989.
- 3. The District completely defueled the Rancho Seco reactor on December 8, 1989. l
- 4. The Nuclear Regulatory Commission (NRC) issued an Order and license condition on May 2, 1990, that prevented SMUD from ,
moving fuel into the Rancho Seco reactor building without
- prior NRC approval.
- 5. SMUD submitted a proposed Decommissioning Plan for Rancho Seco on May 20, 1991.
- 6. The POL and Permanently Defueled Technical Specifications (PDTS) for Rancho Seco became effective on April 28, 1992. ;
- 7. The NRC issued a decommissioning order and approved the Rancho Seco decommissioning funding plan on March 20, 1995.
The DSAR assumes the 493 irradiated fuel assemblies stored at Rancho Seco are stored under water in the Spent Fuel Pool (SFP). The SFP is a l reinforced concrete, stainless steel lined pool and is housed in the l Fuel Storage Building (FSB). The irradiated fuel assemblies are stored i in free standing, high density storage racks that are in the SFP. The j high density storage racks contain a neutron absorbing material i (Boraflex) within the rack plates. The racks are designed to hold 1,080 t fuel assemblies and ensure a K,n <0. 95 with un-borated water in the SFP. "
i I
DSAR Chapter 14 evaluates the facility licensing design basis accidents considered credible in the PDM. The credible defueled condition design ,
basis accidents are the Fuel Handling Accident (FHA) and the Loss Of i Off-site Power (LOOP) event.
The calculated dose consequences are extremely small for the defueled condition FRA (9.9 mrem) and negligible for the LOOP.
Amendment 2 I
1.1-2
, - - ~. - - ~ ..- - --~ _ - - - - - . . . . . - - - . . - _.-... ..-
1 I
)
1.5.3 CRITERION 3 - FIRE PROTECTION (Continued)
PDM, including the organizational responsibilities for fire protection, l fire protection features, and fire protection program implementation requirements. The Rancho Seco fire protection program provides a defense-in-depth approach to fire protection in order to:
! 1. Prevent fires from starting,
- 2. Promptly detect, appropriately respond to, and eventually extinguish a fire, if one should occur, and l
- 3. Limit fire damage.
i In addition, the RSDFPP contains fire protection system operational l requirements, compensatory measures, test requirements, fire brigade staffing requirements, and off-site fire fighting assistance !
coordination requirements. !
I The RSDFPP addresses the active and passive fire protection design features required during the PDM to ensure safe storage of irradiated fuel. These fire protection design features include:
- 1. Fire barriers and fire breaks,
- 2. Fire detection and alarm systems, and 4 3. Fire suppression systems.
A description of the plant fire suppression and detection system arrangements is included in the RSDFPP.
Rupture or inadvertent actuation of fire suppression systems will not significantly impair the operation of any equipment important to safety.
Inadvertent operation of carbon dioxide systems will not prevent the operation of any equipment required to safely store irradiated fuel in the spent fuel pool. Inadvertent operation of a section of a wet-pipe sprinkler system is not expected to result in impairment of equipment important to safety, primarily because these systems are excluded from rooms containing electrical switchgear and pumps or motors that are important to safety. In those cases where water suppression is provided in areas containing equipment important to safety, that equipment is adequately protected to function properly while exposed to water, or, inadvertent operation is precluded by use of a pre-action system. Areas protected by water suppression systems are reviewed to ensure adequate drainage is provided for fire water.
Amendment 2 1.5-4
t 1.5.3 CRITERION 3 - FIRE PROTECTION (Continued)
The District maintains a fire protection program for Rancho Seco that addresses the potential for fires which could result in a nuclear hazard. The objectives of the fire protection program are to: (1)
Reasonably prevent such fires from occurring; (2) rapidly detect, ,
control, and extinguish those fires which do occur; and (3) Ensure that the potential hazard due to fire to the public, environment, and plant personnel is small. The District assesses the Rancho Seco fire protection program on a regular basis and revises the program, as appropriate. The District makes changes to the Rancho Seco fire protection program without NRC approval provided that the changes to not reduce the effectiveness of fire protection measures needed to prevent a nuclear hazard, taking into account the decommissioning plant conditions and activities.
s 1.5.4 CRITERION 4 - ENVIRONMENTAL AND MISSILE DESIGN BASES Structures, systems, and components important to safety shall be designed to accommodate the effects of, and to be compatible with, the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents.
These structures, systems, and components shall be appropriately ;
i protected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids, that may result from equipment failures and from events and conditions outside the nuclear power unit.
i Discussion:
The basic design criterion for Rancho Seco structures, systems, and components important to safety is that there should be no loss of j function associated with environmental conditions during normal j operation, maintenance, testing, and postulated accidents during the PDM.
1.5.5 CRITERION 5 - SHARING OF STRUCTURES, SYSTEMS, AND COMPONENTS Since the Rancho Seco site was a single-unit nuclear power plant facility, this General Design Criterion never applied to Rancho Seco.
1.5.6 CRITERION 10 - REACTOR DESIGN This General Design Criterion is not applicable to Rancho Seco in the PDM.
Amendment 2 1.5-5
. -. . . - . . ~ -.- - - -- - - . _
I l
i l
1.5.7 CRITERION 11 - REACTOR INHERENT PROTECTION i
This General Design Criterion is not applicable to Rancho Seco in the PDM.
1.5.8 CRITERION 12 - SUPPRESSION OF REACTOR POWER OSCILLATIONS l
This General Design Criterion is not applicable to Rancho Seco in the I PDM.
1.5.9 CRITERION 13 - INSTRUMENTATION AND CONTROL Instrumentation and control shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for j anticipated operational occurrences, and for accident conditions as l
appropriate to assure adequate safety, including those variables and l systems that can affect the fission process, the integrity of the l
reactor core, the reactor coolant pressure boundary, and the containment and its associated systems. Appropriate controls shall be provided to l
maintain these variables and systems within prescribed operating ranges.
I Discussion:
The instrumentation and controls described above are not required in the PDM. But, the instrumentation necessary to monitor the safe storage of spent fuel during PDM operations is maintained to assure adequate safety. DSAR Chapter 7 contains more specific instrumentation and l control information.
Instrumentation is provided to monitor spent fuel pool level and temperature. Radiation monitoring instrumentation is provided to ensure radiation levels are maintained within prescribed limits.
1.5.10 CRITERION 14 - REACTOR COOLANT PRESSURE BOUNDARY This General Design Criterion is not applicable to Rancho Seco in the PDM.
1.5.11 CRITERION 15 - REACTOR COOLANT SYSTEM DESIGN This General Design Criterion is not applicable to Rancho Seco in the PDM.
1.5-6
1.5.12 CRITERION 16 - CONTAINMENT DESIGN This General Design Criterion is not applicable to Rancho Seco in the PDM.
1.5.13 CRITERION 17 - ELECTRICAL POWER SYSTEMS An on-site electrical power system and an off-site electrical power system shall be provided to permit functioning of structures, systems, and components important to safety. The safety function for each system (assuming the other system is not functioning) shall be to provide sufficient capacity and capability to assure that (1) specified acceptable fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded as a result of anticipated operational occurrences and (2) the core is cooled and containment integrity and other vital functions are maintained in the event of postulated accidents.
The on-site electric power sources, including the batteries and the on-site electric distribution system, shall have sufficient independence, redundancy, and testability to perform their safety functions assuring a single failure.
Electric power from the transmission network to the on-site electric distribution system shall be supplied by two physically independent circuits (not necessarily on separate rights of way) designed and located so as to minimize to the extent practical the likelihood of their simultaneous failure under operating and postulated accident and ;
environmental conditions. A switchyard common to both circuits is l acceptable. Each of these circuits shall be designed to be available in
]
sufficient time following a loss of all on-site alternating current l power supplies and the other off-site electrical power circuit, to assure that specified acceptable fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded. I One of these circuits shall be designed to be available within a few seconds following a loss-of-coolant accident to assure that core cooling, containment integrity, and other vital safety functions are maintained.
Provisions shall be included to minimize the probability of losing electrical power from any of the remaining sources as a result of, or coincident with, the loss of power generated by the nuclear power unit, the loss of power from the transmission network, or the loss of power l
from the on-site electric power sources. !
l 1.5-7
-. ~ . - - - . , ~ - _ - - - . - _ . . . - - . . . - - - . - - . .- - - . -.-.
l l
1 1
1.5.13 CRITERION 17 - ELECTRICAL POWER SYSTEMS (Continued) ]
piscussion:
In the PDM, on-site emergency or auxiliary electrical power sources are not required, otherwise, the site electrical power system conforms to the applicable portions of this General Design Criterion through an off-site 220-kV transmission system. This off-site a-c power source provides sufficient capacity and capability to assure operation of necessary safety' functions within the time required during anticipated operational occurrences and postulated accidents in the PDM. j l
Electrical power is supplied to the Rancho Seco electrical distribution system switchyard by six transmission lines from four separate transmission network switchyards (Bellota, Elk Grove, Pocket, and Hedge). Four circuits are installed on two double-circuit towers that follow the same route. The two other circuits are installed on a double-circuit tower following a separate route. Two physically independent circuits with start-up transformers sized to carry full l plant auxiliary loads are provided from the switchyard to the on-site !
electrical distribution system. The electrical distribution system arrangement minimizes circuit vulnerability to physical damage. DSAR Section 8.2.2 addresses the Rancho Seco electrical distribution system )
in more detail.
1 Two independent power sources, through their corresponding start-up transformers, feed the entire plant a-e power.needs during the PDM. To ensure continuity of power, the main and auxiliary transformers can be ;
used as a back feed to supply the loads normally supplied by start-up transformer No. 2. If all the off-site power sources are lost, sufficient time is available to re-establish an off-site power source to ensure the spent fuel pool temperature is maintained below 180'F, and that no damage to fuel assemblies results. The off-site power sources ;
are designed for non-interrupted availability. Administrative controls are used to supplement the reliability of the plant electrical power system design under the condition of electrical system degradation.
The switching arrangement in the 220-kV switchyard includes two full capacity main buses. Primary and backup relaying are provided for each circuit along with circuit failure backup protection.
1.5.14 CRITERION 18 - INSPECTION AND TESTING OF ELECTRICAL POWER SYSTEMS Electrical power systems important to safety shall be designed to permit periodic inspection and testing of important areas and features, such as wiring, insulation, connectors, and switchboards, to assess the 1.5-8
1.5.14 CRITERION 18 - INSPECTION AND TESTING OF ELECTRICAL POWER SYSTEMS (Continued) j l
continuity of the systems and the condition of their components. The systems shall be designed with a capability to test periodically (1) the operability and functional performance of the components of the systems, such as on-site power sources, relays, switches, and buses, and (2) the operability of the systems as a whole and, under conditions as close to design as practical, the full operation sequence that brings the systems into operation, including operation of applicable portions of the protection system and the transfer of power among the nuclear power unit, the off-site power system, and the on-site power system.
Discussion:
The provisions made for testing and inspecting the operation and functional performance of electrical power system components important to safety are discussed in DSAR Section 8.3.
1.5.15 CRITERION 19 - CONTROL ROOM A control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents. Adequate radiation protection shall be ,
provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body or its equivalent to any part of the body for the duration of the accident.
l Equipment at appropriate locations outside the control room shall be l provided (1) with a design capability for prompt hot shutdown of the l
reactor, including necessary instrumentation and controls to maintain l
the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactor through
{ the use of suitable procedures. ,
l t Discussion:
A control room is provided and equipped to safely monitor and control Rancho Seco in the PDM under normal and accident conditions. Adequate radiation protection is provided to ensure radiation exposures to personnel occupying the control room are minimized. The consequences of j a credible accident in the PDM would expose control room personnel to much less than the 5 rem whole body habitability limit.
I 1.5-9
1.5.15 CRITERION 19 - CONTROL ROOM (Continued)
Considering the limited activities in the PDM, the reduced scope of credible accidents, and the design features provided to ensure continuous control room access, it is unlikely that any necessity could arise that would require evacuation of the control room. DSAR Section 7.4 addresses control room occupancy and availability in greater detail.
1.5.16 CRITERION 20 - PROTECTION SYSTEM FUNCTIONS This General Design Criterion is not applicable to Rancho Seco in the PDM.
1.5.17 CRITERION 21 - PROTECTION SYSTEM RELIABILITY AND TESTABILITY This General Design Criterion is not applicable to Rancho Seco in the PDM.
1.5.18 CRITERION 22 - PROTECTION SYSTEM INDEPENDENCE This General Design Criterion is not applicable to Rancho Seco in the PDM.
1.5.19 CRITERION 23 - PROTECTION SYSTEM FAILURE MODES This General Design Criterion is not applicable to Rancho Seco in the PDM.
1.5.20 CRITERION 24 - SEPARATION OF PROTECTION AND CONTROL SYSTEMS l
This General Design Criterion is not applicable to Rancho Seco in the PDM.
{
1.5.21 CRITERION 25 - PROTECTION SYSTEM REQUIREMENTS FOR REACTIVITY CONTROL MALFUNCTIONS This General Design Criterion is not applicable to Rancho Seco in the PDM.
l 1
1.5-10
1.5.22 CRITERION 26 - REACTIVITY CONTROL SYSTEM REDUNDANCY AND CAPABILITY This General Design Criterion is not applicable to Rancho Seco in the PDM.
1.5.23 CRITERION 27 - COMBINED REACTIVITY CONTROL SYSTEMS CAPABILITY This General Design Criterion is not applicable to Rancho Seco in the PDM. i 1.5.24 CRITERION 28 - REACTIVITY LIMITS This General' Design Criterion is not applicable to Rancho Seco in the PDN.
1.5.25 CRITERION 29 - PROTECTION AGAINST ANTICIPATED OPERATIONAL OCCURRENCES This General Design Criterion is not applicable to Rancho Seco in the PDM. ;
1.5.26 CRITERION 30 - QUALITY OF REACTOR COOLANT PRESSURE BOUNDARY ,
This. General Design Criterion is not applicable to Rancho Seco in the PDN.
1.5.27 CRITERION 31 - FRACTURE PREVENTION OF REACTOR COOLANT PRESSURE BOUNDARY This General Design Criterion is not applicable to Rancho Seco in the i PDM.
l 1.5.28 CRITERION 32 - INSPECTION OF REACTOR COOLANT PRESSURE BOUNDARY ]
1 This General Design Criterion is not applicable to Rancho Seco in the l I
PDM.
1.5-11
1.5.29 CRITERION 33 - REACTOR COOLANT MAKEUP This General Design Criterion is not applicable to Rancho Seco in the PDM.
1.5.30 CRITERION 34 - RESIDUAL HEAT REMOVAL This General Design Criterion is not applicable to Rancho Seco in the PDM.
1.5.31 CRITERION 35 - EMERGENCY CORE COOLING This General Design Criterion is not applicable to Rancho Seco in the PDM.
1.5.32 CRITERION 36 - INSPECTION OF EMERGENCY CORE COOLING SYSTEM This General Design Criterion is not applicable to Rancho Seco in the PDN.
1.5.33 CRITERION 37 - TESTING OF EMERGENCY CORE COOLING SYSTEM This General Design Criterion is not applicable to Rancho Seco in the PDM.
1.5.34 CRITERION 38 - CONTAINMENT HEAT REMOVAL This General Design Criterion is not applicable to Rancho Seco in the PDM.
1.5.35 CRITERION 39 - INSPECTION OF CONTAINMENT HEAT REMOVAL SYSTEM This General Design Criterion is not applicable to Rancho Seco in the PDM.
1.5.36 CRITERION 40 - TESTING OF CONTAINMENT HEAT REMOVAL SYSTEM This General Design Criterion is not applicable to Rancho Seco in the PDM.
1.5-12
1.5.37 CRITERION 41 - CONTAINMENT ATMOSPHERE CLEANUP This General Design Criterion is not applicable to Rancho Seco in the PDM.
1.5.38 CRITERION 42 - INSPECTION OF' CONTAINMENT ATMOSPHERE CLEANUP This General Design Criterion is not applicable to Rancho Seco in the PDM.
1.5.39 CRITERION 43 - TESTING OF CONTAINMENT ATMOSPHERE CLEANUP SYSTEMS This General Design Criterion is not applicable to Rancho Seco in the PDM.
1.5.40 CRITERION 44 - COOLING WATER A system to transfer heat from structures, systems, and components important to safety to an ultimate heat sink shall be provided. The system safety function shall be to transfer the combined heat load of these structures systems, and components under normal operating and accident conditions.
Suitable redundancy in components and features, and suitable interconnections, leak detection, and isolation capabilities shall be provided to assure that for on-site electrical power system operation (assuming off-site power is not available) and for off-site electrical power system operation (assuming on-site power is not available) the system safety function can be accomplished, assuming a single failure.
Discussion:
On-site emergency or auxiliary electrical power sources are not required to support plant operations in the PDM. The primary spent fuel pool cooling system is the Spent Fuel Cooling system (SFC). SFC is cooled by the Component Cooling Water system (CCW), which is cooled by the Plant Cooling Water system (PCW). No other cooling water system is required during the PDM to help ensure the SFP water maintains within its design basis temperature. SFC is capable of providing the cooling required to maintain the spent fuel pool well below 14 0*F . The back-up spent fuel pool cooling system is the Radwaste and Fuel Storage Area HVAC system, which is a ventilation system and not a cooling water system.
Amendment 1 1.5-13
. .__ _ _ _ _ _ . _ . . m > _ _ _ .. . _ __ m . ..... . _ . _ . _
l.5.40 CRITERION 44 - COOLING WATER (Continued)
CCW is periodically sampled for radioactivity. For additional details on these cooling water systems, see DSAR Sections 9.2 and 9.4.
1.5.41 CRITERION 45 - INSPECTION OF COOLING WATER SYSTEM The cooling water system shall be designed to permit appropriate periodic inspection of important components, such as heat exchangers and piping, to assure the integrity and capabilities of the system.
Discunnion:
SFC, CCW, and PCW are designed to permit the required periodic inspections in accordance with the In-service Confirmation Program as described in DSAR Section 9.1.
1.5.42 CRITERION 46 - TESTING OF COOLING WATER SYSTEM The cooling water system shall be designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leak-tight integrity of its components, (2) the operability and the performance of the active components of the system, and (3) the l operability of the system as a whole and, under conditions as close to design as practical, the performance of the full operational sequence that brings the system into operation for reactor shutdown and for loss-of-coolant accident, including operation of applicable portions of the protection system and the transfer between normal and emergency power sources.
Discussion:
)
1 SFC, CCW, and PCW are designed to permit the testing required in the PDM in accordance with the In-service Confirmation Program as described in DSAR Section 9.1.
1.5.43 CRITERION 50 - CONTAINMENT DESIGN BASIS This General Design Criterion is not applicable to Rancho Seco in the PDH.
4 9
Amendment 1 1.5-14
1
)
1.5.44 CRITERION 51 - FRACTURE PREVENTION OF CONTAINMENT PRESSURE BOUNDARY This General Design Criterion is not applicable to Rancho Seco in the PDM.
P 1.5.45 CRITERION 52 - CAPABILITY FOR CONTAINMENT LEAKAGE RATE TESTING This General Design Criterion is not applicable to Rancho Seco in the PDH.
1.5.46 CRITERION 53 - PROVISIONS FOR CONTAINMENT TESTING AND INSPECTION This General Design Criterion is not applicable to Rancho Seco in the PDM.
1.5.47 CRITERION 54 - PIPING SYSTEMS PENETRATING CONTAINMENT This General Design Criterion is not applicable to Rancho Seco in the PDM.
1.5.48 CRITERION 55 - REACTOR COOLANT PRESSURE BOUNDARY PENETRATING CONTAINMENT This General Design Criterion is not applicable to Rancho Seco in the PDM.
1.5.49 CRITERION 56 - PRIMARY CONTAINMENT ISOLATION This General Design Criterion is not applicable to Rancho Seco in the PDM.
1.5.50 CRITERION 57 - CLOSED SYSTEM ISOLATION VALVES This General Design Criterion is not applicable to Rancho Seco in the PDM.
1.5-15
.. . - _ . -. - --_ ~. - _ _ _ -
1.5.51 CRITERION 60 - CONTROL OF RELEASES OF RADIOACTIVE MATERIALS TO THE ENVIRONMENT i
The nuclear power unit design shall include means to control suitably the release of radioactive materials in gaseous and liquid effluent and
, to handle radioactive solid wastes produced during normal reactor operations, including anticipated operational occurrences. Sufficient holdup capacity shall be provided for retention of gaseous and liquid effluent containing radioactive materials, particularly where unfavorable site environmental conditions can be expected to impose unusual operation limitations upon the release of such effluent to the environment.
Discussion:
i The radioactive waste systems and radiological administrative control programs at Rancho Seco effectively control the collection, segregation, processing, packaging, and discharge or disposal of radioactive solids, liquids, and gases. These systems and programs function to control releases to as low as is reasonably achievable levels and are designed to give reasonable assurance that the numerical guidelines defined in 10 CFR 50.34a, 50.36a, and 10 CFR 50, Appendix I, and contained in the Off-site Dose Calculation Manual (ODCM) will be met.
The Interim On-site Storage Building (IOSB) provides an area of shielded, safe retrievable storage for packaged low level radioactive waste. Potential release pathways of radionuclides in particulate form are controlled and monitored to verify compliance with the dose guidelines of 10 CFR 50, Appendix I.
Liquid and solid wastes are normally processed in batches for off-site disposal. Gaseous effluent is released in a continuous mode and is monitored to assure the releases are maintained within the numerical dose objectives defined in 10 CFR 50, Appendix I.
A permanently installed spent fuel pool area radiation monitor and the plant vent radiation detectors monitor gaseous effluent discharges. In addition, portable monitors are available on site for supplemental surveys, if necessary.
Leakage of radioactive liquid into cooling water systems required to support safe storage of spent fuel in the PDM is detected by radiation monitors. These monitors are used to identify contamination of systems that are not normally radioactive.
1.5-16
f 1.5.52 CRITERION 61 - FUEL STORAGE AND HANDLING AND RADIOACTIVITY i
CONTROL j The fuel storage and handling, radioactive waste, and other systems which may contain radioactivity shall be designed to assure adequate safety under normal and postulated accident conditions. These systems ,
I
] shall be designed (1) with a capability to permit inspection and testing j of components important to safety, (2) with suitable shielding for radiation protection, (3) with appropriate containment, confinement, and f filtering systems, (4) with a residual heat removal capability having
[
- reliability and testability that reflects the importance to safety of
{ decay heat and other residual heat removal, and (5) to prevent ,
7 significant reduction in fuel storage coolant inventory under accident conditions. >
l 1
j Dimeussion:
t The fuel storage and handling, radioactive waste, and other systems ,
! which may contain radioactivity are designed to assure adequate safety under normal ar.d postulated accident conditions. The systems have the l
j capability for periodic inspection and testing of components important 1 to safety. Suitable shielding for plant personnel radiation protection ,
i .
i
- The spent fuel pool is located in the Fuel' Storage Building.
g Radioactive gaseous waste ventilation equipment is contained in the i Auxiliary Building. Gaseous waste from the Fuel Storage Building and j the Auxiliary Building is continuously discharged through the Auxiliary ,
i Building vents. Also, administrative controls are in place to ensure l that an Auxiliary Building radioactive gaseous effluent ventilation i system takes suction on the Solidification Building and removes any airborne contaminants generated from operation of the Blender / Dryer radwaste processing unit.
I j Radioactive liquid leakage into the component cooling water system is
! ~ determined by a radiation monitor in the component cooling water pump l
!- suction header. Any accidental leakage from liquid waste storage tanks i in the Auxiliary Building is collected in sumps and transferred to f radioactive liquid waste tanks in the Auxiliary Building to prevent I i release to the environment. Any accidental leakage from the Blender /
- Dryer radwaste processing unit in the Solidification Building is
- contained in a bermed area and can either be directed to Auxiliary
- Building sumps with manual actions or cleaned up in accordance with Radiation Protection procedures.
k The spent fuel pool cooling system contains a small purification loop that removes fission products and other contaminants contained in the j spent fuel storage pool water. The design bases of the spent fuel pool j 1.5-17 l
. -.. - . -. 1
. _ _ . - . _ . . . - - - - - . __ - _ - . _ _ - - - . _ . . _ _ _ _ . _ _ . _ . _ _ . . ~ . . - _ . _ _ _ . - ,
l 1.5.52 CRITERION 61 - FUEL STORAGE AND HANDLING AND RADIOACTIVITY l CONTROL (Continued) cooling system reflects the safety significance of this system. See DSAR Sections 9.4 for additional spent fuel pool cooling system design information. The capability for appropriate testing has been provided.
For information on the cooling water systems that support spent fuel {
pool cooling, see DSAR Section 1.5.40, CRITERION 44 - COOLING WATER. ]
The Fuel Storage Building and the spent fuel pool liner are designed as Seismic Category I and can withstand all credible conditions of loading, including normal loads and loads from the design basis earthquake. The spent fuel pool cannot be gravity drained below the top of active fuel without a gross failure of the Fuel Storage Building, the spent fuel pool, and the liner, because the lowest penetration in the spent fuel pool is above active fuel. Spent fuel pool design precludes siphoning below safe levels.
1.5.53 CRITERION 62 - PREVENTION OF CRITICALITY IN FUEL STORAGE AND HANDLING Criticality in the fuel storage and handling system shall be prevented l by physical systems or processes, preferably by use of geometrically l safe configuration. I l
Discussion:
Criticality in the spent fuel storage pool is prevented by the use of a hautron absorbing material (Boraflex) which is sandwiched in the middle of the plates that comprise the high density spent fuel storage rack.
The racks are designed to hold 1080 fuel assemblies at 4.0 weight j percent enrichment with un-borated water in the spent fuel pool while l maintaining K,n less than 0.95. Rancho Seco has 493 irradiated fuel assemblies in the spent fuel pool. The highest calculated synthesized- ;
enrichment (remaining U-235 and fissile Pu) for any assembly in the i spent fuel pool is 2.573 weight percent.
1.5.54 CRITERION 63 - MONITORING FUEL AND WASTE STORAGE Appropriate systems shall be provided in fuel storage and radioactive waste systems and associated handling areas (1) to detect conditions that may result in. loss of residual heat removal capability and excessive radiation levels and (2) to initiate appropriate safety actions.
Amendment 2 1.5-18
~ _ _ . _ . _ - - . _ _ _ _ _ , __ ,,
_ . _ _ . _ _ _ . . . . _ . . . . _m . _. _ _._ _._ _ ____ _ _._ __... __ _
l.5.54 CRITERION 63 - MONITORING FUEL AND WASTE STORAGE (Continued)
Discussion:
The spent fuel storage pool is provided with a high temperature and high and low level alarms. Also, an area radiation monitor near the spent fuel pool provides. warning of an excessive spent fuel pool area l radiation level condition. This instrumentation is capable of detecting l conditions which indicate a loss of residual' heat removal, a loss of !
spent fuel pool water inventory, and excessive radiation levels in the i spent fuel storage area.
It is not necessary to automatically initiate backup spent fuel pool ,
cooling when normal cooling is lost because the back-up spent fuel pool ,
cooling system (the Radwaste and Fuel Storage Building HVAC system) 1.s '
designed to run continuously. If the primary spent fuel pool cooling ,
system is lost, greater than 20 days is available before the spent fuel pool temperature can reach its maximum equilibrium temperature of ,
approximately 145*F with the backup cooling system operating.
DSAR Section 11.8 contains details of the radiation monitoring system, including the radwaste area and process monitors that measure, indicate, annunciate, and record the radiation levels at selected areas.
Radiation monitor alarms are annunciated at the annunciator panel in the control room. .Also, the radwaste panel multi-point recorder and the ;
high radiation level radwaste panel annunciator provide indication of l abnormal system radiation levels. .
1.5.55 CRITERION 64 - MONITORING RADIOACTIVITY RELEASES Means shall be provided for monitoring the reactor containment atmosphere, spaces containing components for recirculation of loss-of-coolant accident fluids, effluent discharge paths, and the plant environs for radioactivity that may be released from normal operations, including anticipated operational occurrences, and from postulated accidents.
ll Discunnion:
The Rancho Seco radiation' monitoring system-meets the portions of this criterion that is applicable to a plant in the PDM. The monitoring of radioactivity releases is addressed and described in DSAR Section 11.8.
Amendment 1 1.5-19
_ _ _ _ _ . _ _ _ . _ ...___m _ _ _ _ _ _ _ _ . _ _ _ _ _ - . . _ . _ _ _ . . _ . . _ _ _ . _ _ . _
l.5.18 SAFETY GUIDE 18 - STRUCTURAL ACCEPTANCE TEST FOR CONCRETE PRIMARY FACTOR CONTAINMENTS This Safety Guide is not applicable to Rancho Seco in the PDM.
1.6.19 SAFETY GUIDE 19 - NONDESTRUCTIVE TESTING OF PRIMARY CONTAINMENT LINERS This Safety Guide is not applicable to Rancho Seco in the PDM.
1.6.20 SAFETY GUIDE 20 - VIBRATION MEASUREMENTS ON REACTOR INTERNALS This Safety Guide is not applicable to Rancho Seco in the PDM.
1.6.21 SAFETY GUIDE 21 - MEASURING AND REPORTING OF EFFLUENT FROM NUCLEAR POWER PLANTS The measuring and reporting of radiological effluent from Rancho Seco complies with the requirements of Safety Guide 21.
Annual reports summarizing the quantity of radionuclides released from the site are submitted to the NRC in accordance with 10 CFR 50.36a (a) (2) . This summary data is comprised of radioactive release information collected in accordance with the Off-site Dose Calculation Manual (ODCM).
Normal liquid radioactive waste water discharges are made from the Retention Basins. Prior to a release of radioactive liquid waste water, grab samples are taken and analyzed for fission product and activation product radionuclides. During normal discharges to the environment, the radioactive liquid effluent is continuously monitored in the discharge stream.
Gaseous effluent discharged to the environment is evaluated by continuously operating radiation monitors located in the plant gaseous effluent discharge stacks. Monthly samples of gases and particulates are taken and evaluated for fission product and/or activation product radionuclides. Similar samples are taken during Reactor Building ventilation discharges.
Default plant vent stack flow rates and meteorological data are used in ODCM evaluations. The results of these evaluations are reported in the Amendment 2 1.6-7
1.6.21 SAFETY GUIDE 21 - MEASURING AND REPORTING OF EFFLUENT FROM NUCLEAR POWER PLANTS (Continued)
Annual Radioactive Effluent Release Report. This report contains summary data of radionuclides found in routine and abnormal radioactive l
liquid and gas releases during the applicable period. Each report also includes total curies of each radionuclide released and the fraction of the 10 CFR 50, Appendix I dose guidelines released from the site. i l
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i 1.7 REMRAR(*H AND DEVELOPMENT The'only area of research and development peri ^ormed for the original licensing of Rancho Seco that is applicable in the PDW is Bechtel Topical Report No. BC-TOP-4, dated April 30, 1971, Seismic Analysis of Structures and Equipment for Nuclear Power Plants.
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1.8 IDRMTIFICATION OF AGENTS AND CONTRACTORE !
SMUD is responsible for the design, purchasing, construction, testing, maintenance, safe storage of spent fuel, and eventual decommissioning of the Rancho Seco nuclear facility.
1.8-1
- - - . - . - . - - . - . - . . . ~ - . . _ . _ _ . - . - . - - . - - . . . . . . . - . - - . . - . . .-
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1.9 CONCLUSION
S The personnel assembled to operate, maintain, and safely store spent nuclear fuel at Rancho Seco in the PDM are capable of performing their required project function. The health and safety of the public and plant personnel, and the reliability of equipment and systems important to safety are among the primary concerns during the PDM.
The site was examined and found to be suitable for operation of a nuclear plant.
The Rancho Seco plant is compatible with the surrounding population and land uses, present and expected. Site characteristics of meteorology, hydrology, geology, and seismology are favorable.
Accordingly, SMUD has concluded that the Rancho Seco nuclear facility was designed and constructed safely and can be operated in a safe manner in the PDM; that the design provides adequate protection to the public from any sequence of events resulting in disablement of equipment from causes, natural or mechanical; and that the District is qualified to operate and maintain this nuclear facility in accordance with all applicable laws and regulations and in a manner satisfactory to the NRC, the public interests, and to itself.
l.S-1
i 1.10 REFERENCES
- 1. License Amendment No. 119, dated March 19, 1992, Permanently Defueled Technical Specifications
- 2. Safety Analysis and No Significant Hazards Consideration (Log No.
1091, Revision 3) for Proposed Amendment 182, Revision 3, Permanently Defueled Technical Specifications
- 3. License Amendment No. 117, dated March 17, 1992, Possession-Only License
- 4. Ranr'.o Seco Updated Safety Analysis Report (USAR), Amendment No. 8
- 5. 10 CFR 50.59 evaluation for procedure B.10, Revision 9, "Defueled Condition" l 6. SMUD Calculation Z-SFC-M2555, " Peak Spent Fuel Pool Temperature Versus Calendar Time."
l 7. SMUD Calculation Z-SFC-M2557, " Spent Fuel Decay Heat Based on ORIGEN2 Computer Code." >
l
- 8. NRC Order Approving the Rancho Seco Decommissioning Plan, dated March 20, 1996.
- 9. Rancho Seco License Amendment No. 122, dated July 19, 1995.
- 10. SMUD Calculation Z-SFC-N0049, Revision 3, " Maximum Predicted whole Body and Skin Doses and Dose Rates at the Site Boundary from Postulated Accidents During Plant Shutdown."
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I l 1 TABLE OF CONTENTS l 1 i Section Title Page i I
- 2. SITE AND ENVIRONMENT 2.1-1 2.1
SUMMARY
2.1-1 2.2 SITE ann ADJACENT AREhn 2.2-1 l 2.2.1 SITE LOCATION 2.2-1 2.2.2 SITE OWNERSHIP 2.2-1 2.2.3 SITE ACTIVITIES 2.2-2 l 2.2.4 POPULATION 2.2-2 2.2.5 LAND USE 2.2-6 l 2.2.5.1 current Land Use 2.2-6 l 2.2.5.2 Future Land Use 2.2-6 2.2.6 ACCESS AND EGRESS 2.2-11 l 2.2.7 AIRFIELDS AND MISSILE SITES 2.2-11 l 2.3 METEOROLOGY 2.3-1 2.
3.1 INTRODUCTION
2.3-1 2.3.1.1 Ooerational Meteoroloaical Program 2.3-1 2.3.2 CLIMATOLOGICAL STATISTICS 2.3-1 1 2.3.3 ATMOSPHERIC DISPERSION FACTOR 2.3-1 2.4 HYDROLOGY 2.4-1 2.4.1 CHARACTERISTICS OF STREAMS AND LAKES IN VICINITY 2.4-1 2.4.2 TOPOGRAPHY 2.4-1 2.4.3 TERMINAL DISPOSAL OF STORM RUNOFF 2.4-1 2.4.4 HISTORICAL FLOODING 2.4-1 Amendment 2 2-L
TABLE OF CONTENTS (Continued) Section Title Page 2.4.5 PREDICTION OF LAND URBANIZATION 2.4-3 2.4.6 GROUNDWATER 2.4-3 2.4.6.1 occurrence and Movement 2.4-3 2.4. 6.2 vater sunnly 2.4-4 2.4.6.3 ouality 2.4-4 2.4.7 WELLS AND BORINGS 2.4-4 2.5 GEOLOGY 2.5-1 2.6 SEISMOLOGY 2.6-1 2.7 SOILS 2.7-1 2.8 ENVIRONMENTAL MONITORING PROGRAM 2.8-1
2.9 REFERENCES
2.9-1 I l i i t 2-11
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LIST OF TABLES Table Title Enga 2.2-1 Projected Population Within a 50-Mile Radius of 2.2-3 Rancho Seco Site 2.2-2 Permanent Population Distribution Within a 10-Mile 2.2-4 Radius of the Rancho Seco Site 2.2-3 Institutions and Facilities Within a 10-M11e Radius 2.2-5 of Rancho Seco Site 2.4-1 Data on Reservoirs and Lakes Within a 50-Mile Radius 2.4-2 Radius I l i 1 2-111
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i LIST OF FIGURES 2 4
! riaures Title 2.2-1 General Area Map 2.2-2 SMUD Owned Land and Rancho Seco Site Layout l
4 2.2-3 Permanent Population Within a 10-Mile Radius of the Rancho i Seco Site 4 2.2-4 Land Use within a 5-Mile Radius J j 2.4-1 Streams and Lakes j 4 Amendment 2 2-iv
DSAR CHAPTER 2. SITE AND ENVIRONMENT 2.1
SUMMARY
The Rancho Seco nuclear facility is located approximately 26 miles north- northeast of Stockton and 25 miles southeast of the city of Sacramento in Sacramento County, California. The Rancho Seco nuclear facility is approximately 87 acres and sits within a 2480 acre plot of land that is owned by the District. The nuclear facility boundary is defined by a double chain link, eight foot high fence that surrounds the plant. This nuclear facility boundary is also known as the Industrial Area boundary. The nearest population center of 25,000 or more is the city of Lodi, which is about 17 miles southwest of the plant. The area around the site is almost exclusively agricultural and is presently used as grazing land. There are no commercial dairy cattle within a five-mile radius of the Rancho Seco nuclear facility. The climatology of the site is typical of the Great Central Valley of California. Cloudless skies prevail during summer and much of the spring and fall seasons due to the Pacific anticyclone off the California coast that prevents Pacific storms from reaching inland. The rainy season usually extends from December through March. Dilution water for the very low level radioactive waste water periodically discharged from the plant is supplied from the Folsom South Canal, which is a feature of the Central Valley Water Project. The canal was constructed by the Bureau of Reclamation. A pipeline and pumping station are located between the plant and the Folsom South Canal. Groundwater in the site area occurs under free or semi-confined conditions. Groundwater is stored mainly in the alluvium, the older alluvial type deposits, and the Mehrten Formation. Groundwater movement in the area is to the southwest with a slope of about ten feet per mile. There is no indication of faulting beneath the site. The nearest fault system, the Foothill Fault System, is about ten miles east of the site and has been inactive since the Jurassic Period, approximately 135 million years ago. Ground accelerations of no greater than 0.05g are anticipated at the site during the life of the plant. The soils at the Rancho Seco site are sufficiently strong to safely support the Fuel Storage Building and spent fuel pool and other facilities necessary for PDM operations. These soils can be categorized as hard to very hard silts and silty clays with dense to very dense sands and gravels. 2.1-1
_ _ _ . . _ _ . _ . - . _ . - - _ _ _ _ . _ _ _ - _ . __ _ . ~ _ _ > _ . ~ . 2.2 SITE AND ADJAPRNT AREAR 2.2.1 SITE LOCATION The District owned land that the Rancho Seco nuclear facility is located i on is in the southeast part of Sacramento County, California. The ! District owned land lies either wholly or partly in Sections 27, 28, 29, l 32, 33, and 34 of Township 6 North, Range 8E. The Rancho Seco site is , ! approximately 26 miles north-northeast of Stockton and 25 miles ' southeast of Sacramento. The general location of the Rancho Seco site is shown on Figure 2.2-1. Figure 2.2-2 provides a more specific layout of the SMUD owned land and a general layout of the site, including a description and location of major site structures and facilities. l l The Rancho Seco Independent Spent Fuel Storage Installation (ISFSI) is located on District owned land approximately 600 feet west of the Rancho [ Seco Interim On-site Storage Building and within a security fence. The j District owned land boundary is 1,200 feet to the west and 1,500 feet to the north of the ISFSI site. Twenty-two Horizontal Storage Modules sit on a concrete slab that is approximately 225 feet long, 170 feet wide, I and 2 feet thick. The ISFSI electrical building is located on the ISFSI slab. Also, two storage casks can be accommodated on the ISFSI pad, if necessary. ! l The approximate coordinate location of the Rancho Seco facility is 38*-20'-40.44" north latitude and 121*-07'-09.94" west longitude. The Universal Transverse Mercator coordinates are 4245540 m N and 664360 m E. The containment structure is located within Zone II of the California plane coordinate system, Lambert projection. The site is located between the Sierra Nevada mountain range to the east and the Pacific Coastal Range to the west in an area of flat to lightly rolling terrain at an elevation that averages approximately 200 feet mean sea level (msl). To the east of the site the land becomes more rolling, rising to an elevation of 600 feet msl at a distance of about seven miles, and increasing in elevation thereafter across the Sierra Nevada foothills to the peaks of the Sierra Nevada mountain range. 2.2.2 SITE OWNERSHIP The site, totaling 2,480 acres, is wholly owned and controlled by the District, including the land the Rancho Seco Reservoir and Recreation Area occupies. The recreation area was developed in accordance with the l Rancho Seco Recreation Area Development Plan, dated December 1969, which l follows the guidelines provided in the Davis-Grunsky Act. Access for ; transmission lines and water lines is from the west side of the property. Amendment 2 2.2-1
__ _ _ _ _ _ _ _ ~ _ _ . _ _ _ .__ ___ .. _ .._ _._ _-. 2.2.3 SITE ACTIVITIES 1 The land in the general vicinity around tt.e Rancho Seco site is j presently undeveloped and is used primarily for grazing beef cattle and other agricultural activities. There is no development projected to the north, east, and south sides of the site. There may be some additional subdivision of the land to the west of the site in the future, with subsequent development of new residences. One to ten acre plots are projected, thus the buildup will be relatively sparse. 2.2.4 POPULATION j The most recent population distribution estimates are contained in the
" Evacuation Time Estimate for the Rancho Seco Plume Exposure Pathway Emergency Planning Zone" prepared by HMM Associates, Inc., in Dacember 1989. The State of California Department of Finance demographic report, dated February 1989, provides population projections for a 50-mile radius around Rancho Seco through 2020.
These Department of Finance population projections are tabulated on Table 2.2-1. Figure 2.2-3 shows the 1989 permanent population for 22-1/2 degree sectors for one-mile wide segments out to approximately a 10-mile radius from the Rancho Seco site. Permanent residents are defined as those persons having year-round residences within the described area. Table 2.2-2 presents, in tabular form, the permanent population distribution for one-mile segments out to 13 miles from the center of the Rancho Seco site. The nearest population center of 25,000 or more is Lodi, 17 miles south-southwest of the site. Other population centers of 25,000 or more within a 50-mile radius include Sacramento at 25 miles, Stockton at 26 miles, and Modesto at 50 miles. The five-mile radius area surrounding the Rancho Seco facility is a low population area. This area is primarily farm land, with few tourist attractions. State Route 104 runs along the northern boundary of the site and connects with State Route 99 and Interstate Route 5 to the west and State Route 88 to the east. A summary of institutional facilities located within approximately a 10-mile radius of the, Rancho Seco site, showing functional title, approximate peak population, and location from Rancho Saco, is shown in Table 2.2-3. Also on the District owned property is the Rancho Seco Reservoir and Recreation Area and a District owned solar power (photo-voltaic) electrical generating station. 2.2-2
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I [ TABLE 2.2-1 > PROJECTED POPULATION WITHIN A 50-MILE RADIUS OF RANCHO SECO SITE *- COUNTY 1.12A 1.9.81 2AQA 2.Qla 2.Q2A Alameda 100 300 350 375 395 l Amador 11,807 18,200 26,000 34,500 40,000 Calaveras 13,456 16,100 25,900 38,700 46,000 t Contra Costa 69,301 168,900 206,000 224,500 242,450 El Dorado 28,567 47,100 72,120 88,200 88,200 i Nevada 330 400 500 575 630 [ Placer 66,943 113,100 162,000 201,800 236,000 Sacramento 630,189 952,500 1,161,300 1,324,700 1,482,200 l San Joaquin 630,194 355,900 528,000 624,600 722,000 , P Solano 53,646 104,400 148,500 173,600 197,300 Stanislaus 111,201 128,000 176,000 206,000 237,000 Sutter 3,050 5,000 6,200 6,800 7,400 Tuolumne 12,509 24,100 38,600 44,800 51,200 Yolo 88,548 120,600 144,300 151,900 172,500 Yuba 1,735 2,100 2,480 2,610 2,720 i TOTALS 1,381,582 2,056,700 2,698,250 3,123,660 3,541,720 l
- Source of population information is the State Department of Finance's demographic report dated February 1989 2.2-3
. - - - -- - . - .-- .- . _~ . . - . _ . ~ ~ . . , _ ._ - . . - - . ~ . - - - - . - l l l 2.2.5 LAND USE 2.2.5.1 current tand use The land area in the vicinity of the Rancho Seco site is almost exclusively agricultural. Figure 2.2-4 provides the location and a description of the agricultural and residential activities within a five-mile radius of Rancho Seco. There are three large-scale commercial dairies in the vicinity of the Rancho Seco site, each with over 200 cows. The closest dairy is i approximately eight miles northwest of the site. A ranch 1 mile east of Rancho Seco has dairy cows for domestic use only. l l l 2.2.5.2 Future Land Use Activities in the area immediately surrounding the site are not expected to change extensively. Proposed land use for the southeast section of Sacramento County as adopted by the Sacramento Planning Department is predominantly (70 percent) agricultural and is expected to remain predominantly agricultural. l The closest significant future land use activity is a proposed development project known as Clay Station 1200, which is awaiting approval by the Sacramento County Planning Board. This proposed development project is located approximately 5 miles west and a little north of the plant. If approved, this development project would call for 222 residential lots, I school, 1 equestrian arena, 1 fire station, and 4 wetland conservation areas. Within a 15-mile radius of Rancho Seco there are five counties (Amador, San Joaquin, Sacramento, El dorado, and Calaveras). Only very small portions of El 'Jorado and Calaveras counties are within a 15-mile radius of Rancho Seco. There is no significant projected growth or developments within these small portions of El Dorado and Calaveras counties. The projected development within Amador, Sacramento, and San Joaquin counties is as follows: A. Amador county Amador County is considering the following six planned developments:
- 1. East Lambert Mine Reactivation and Clay Industrial Tile Plant - This will be a strip-mining plant for stone and gravel. There will be no explosive type mining. The facility will be located approximately 11 miles 2.2-6
4 I 2.2.5.2 Future Land Use (Continued) ; northeast of Rancho Seco. The plant will employ ; approximately 109 employees. The proposed project will have a public hearing, and a decision is expected in ) l August 1992.
- 2. Buena Vista Meadows - This is a newly proposed housing development still in the preliminary stages. The
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proposal calls for 125 residential lots. It is located ! approximately 13 miles east-southeast of Rancho Seco. The project proponents have no yet completed an Environmental Impact Report (EIR) . j
- 3. Comanche Oaks - Located 12 miles southeast of Rancho )
Seco, this housing proposal calls for 131 residential I lots. Amador County has been waiting approxisately 1 1/2 years for a completed EIR. The county believes this ] project will not go forward, j
- 4. Comanche Greens - Located approximately 9 miles southeast of Rancho Seco, this housing project includes 683 residential lots and an 18-hole golf course. Amador County has been waiting approximately 1 1/2 years for a completed EIR. The county believes this project will not go forward.
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- 5. Goose Hill - This is a new housing proposal in the preliminary stages. It is located approximately 14 miles east-southeast of Rancho Seco with plans for 25 residential lots. An EIR has not been completed yet.
- 6. Murieta Ranches - Located approximately 14 miles east-northeast of Rancho Seco. This development will have 13 j residential lots. An EIR has not been completed yet.
The decision on this project is expected in the third quarter of 1992. B. citv of Tone fin Amador countv) Ione is located 11 miles east of Rancho Seco. Within the next ten years, the City of Ione plans to develop most of the I area. Upon completion of these developments there will be a total of 252 multiple family units, and 1,861 single family units, for a total of 2,113 units. The average single family in Ione consists of 2.66 people. This gives a predicted growth of approximately 5,621 people over the next ten years. 2.2-7 4
_. ~ . _ . . - -- -.. . -. . . _ _- l 2.2.5.2 Future Land Use (Continued) I Currently, there are fourteen planned housing developments in the City of Ione. All developments are expected to be within a two mile area of each other. These housing developments are as follows:
- 1. Castle oaks - This housing development will have 96
{ multiple family (MF) units and 823 single family (SF) j units for a total of 919 units. It is under construction and its estimated completion time is 10 years. Also, an 18-hole golf course is planned. .
- 2. Country Club Place - This development has plans for 160 SF units. It is pending approval. IF approved, this j project is estimated to be completed in 10 years.
- 3. Edgebrook Estates Units 1 & 2 - This development will have 83 SF units. It is currently under construction with an estimated completion time of 2 years.
]
- 4. Edgebrook Estates Units 3 through 5 - This project is a )
housing development with 109 SF units planned. It has been approved but is not yet under construction. The project is expected to be completed in 5 years.
- 5. Golden Gate Estates - This development has plans for 121 SF units. It is currently under construction. The estimated completion time is 2 years.
- 6. Harvest Townhouses - 14 MF units planned. The development project has been approved but is not yet under construction. The estimated completion time is 2 years.
- 7. Ione Oaks - This is a 52 SF unit development project.
It has been approved but is not yet under construction. The estimated completion time is 2 years.
- 8. Q Ranch - This housing development has plans for 142 MF units and 474 SF units for a total of 616 units. An 18-hole golf course is also being considered. The Q Ranch is currently in the environmental review stage.
- 9. Sutter Place - 12 SF units planned. This development has been approved but is not yet under construction.
The estimated completion time is 2 years. 2.2-8
2.2.5.2 ' Future Tmnd Umm (Continued)
- 10. Vimini Estates - 27 units planned. This development has been approved but is not yet under construction. The estimated completion time is 2 years.
The project proponents for the Castle Oaks and Q Ranch ' completed EIRs, which'found no hazard and no significant impact due to Rancho Seco. All of the other projects filed negative declarations. C. SACRAMENTO COUNTY The Planning Department for the City of Sacramento intends to leave the majority of the area included in the 15-mile radius [ around Rancho Seco as agricultural and grazing land. But, Sacramento Sounty projects the following development in this area:
- 1. Elk Grove / Vineyard - Located 14 miles west-northwest of Rancho Seco. This project is currently under construction. This development has planned building .
approximately 40,000 housing units over the next 10 years. The boundaries for the development project are Gerber Road to the north, Grant Line Road to the south, ! Excelsior Road to the east, and Highway 99 to the west. The EIR makes no reference to Rancho Seco. No major highways are planned, nor are any airports.
- 2. _
Galt - There is currently construction underway on a , 4,200 unit housing development. No new highways are planned. The EIR found there is no hazard or significant impact due to Rancho Seco.
- 3. Clay Station 1200 - This development is described above.
D. SAN JOAOUIN COUNTY l l The following areas are currently planned for development:
- 1. Rama Ranch - This housing developoent is located 14 ;
miles south-southwest of Rancho Seco There are 12 lots : planned for development. The schedule for this ; development is still unknown. The developer filed a l negative declaration for the project. l
- 2. River Oaks - This housing development is located.
approximately 13 miles south of Rancho Seco. The project includes construction of 300 single-family homes on 124.7 2.2-9
2.2.5.2 ruture tand use (Continued) l acres. The project includes a man-made lake on the l southern end of the project site, a park, library, and fire station. The estimated time to completion is 6 years. The EIR found there is no hazard or significant impact due to Rancho Seco.
- 3. Buckeye Ranch - This project is projected to be a 26-unit development with an 18-hole private golf course, a private equestrian center, enhancement of two lakes, and j a 720-acre nature preserve. The outer perimeter of the site is located approximately 14 miles _ southwest of j Rancho Seco. The project is currently in the Draft EIR stage. If the plan is approved (probably late August or early September) the development should take 2 to 3 years to complete.
- 4. Dry Creek Village - This project is located approximately 12 miles southwest of Rancho Seco. This l
housing development calls for 187 units. This project is presently in the Initial Study and Notice of Preparation phase. The schedule for this project has not yet been determined.
- 5. Liberty - This project is located approximately 10 miles south of Rancho Seco. The propesal consists of three residential components:
- a. A retirement village (3,000 units),
- b. The Borden community (3,000 units), and
- c. The Forester Community (700 units).
The Liberty project proposal.results in a projected total of 6,700 new residences. The county projects an additional 1,361 residences in the area, giving an { overall total of 8,061 residential units on 7,960 acres. Two golf courses and two parks are planned. Also, the Liberty project calls for a town hall facility, a full , service library, and a child care facility to I accommodate the projected needs of the first 2000 family homes. Provisions for commercial support include five acres of office / commercial land, 50 acres of neighborhood and commercial land, and 20 acres of community land. To provide on-site employment opportunities for residents, a total of 90 acres of research park land are proposed. 2.2-10
2.2.5.2 Future Land Use (Continued) The Liberty project would require construction of a network of expressway, arterial, and collector roads. No mass transit facilities are proposed. To date, the county. supervisors have rejected this project, but the project is currently under appeal. 2.2.6 ACCESS AND EGRESS ; As shown in Figure 2.2-1, State Route 104 runs just north of the site in a general east-west direction and connects with US Route 5 and State Route 99 to the west and State Route 88 to the east. The Rancho Seco Access Road from Twin Cities Road is identified in Fig'ure 2.2-2. This road is the main access road to the plant and to the nearby Rancho Seco Reservoir and Recreation Area. This road is not a through road, and is designed to handle heavy construction vehicles. Rail access to the Rancho Seco facility is available via a rail spur from the existing Southern Pacific Railroad line that runs roughly parallel to State Route 104 adjacent to the site. The routing of the rail spur is shown in Figure 2.2-2. 2.2.7 AIRFIELDS AND MISSILE SITE The nearest major airfield to the Rancho Seco site is the Mather Air Force Base, 18 miles northwest of the site. There are no military flight patterns in the vicinity of the site; however, some military training flights may occasionally come within a 10-mile radius of the site. The nearest defensive missile site is more than 45 miles from the site. I l 1 2.2-11
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_____.,__m. _ . _ _ . . . _ _ _ _ _ . _ _ . . . _ . _ _ _ _ _ _ _ _ _ . . _ _ _ _ . . _ . . .__ _ ._. ___ TABLE OF CONTENTS Section Title Eagg
- 5. STRUCTURES AND CONTAINMENT SYSTEM 5.1-1 5.1 GENERAL 5.1-1 5.1.1 CLASSES OF STRUCTURES 5.1-1 5.1.1.1 clama I 5.1-1 5.1.1.2 clama II 5.1-1 5.1.1.3 clans III 5.1-2 5.1.2 DESIGN LOADS AND STRUCTURAL BEHAVIOR 5.1-2 5.1.2.1 clans I structures 5.1-2 5.1.2.1.1 Normal Operations 5.1-2 5.1.2.1.2 Accident and Seismic conditions 5.1-2 q.1.2.1.3 Missiles 5.1-2 1
_ 5.1.2.1.4 Separation of Structures and Components. 5.1-3 I l 5.1.2.1.5 Seismic Design of Structures 5.1-3 5.1.2.1.6 Wind Loads 5.1-6 5.1.2.1.7 Tornado Loads 5.1-7 5.1.2.1.8 Seismic Design of Equipment, Structures, and 5.1-10 Supports 5.1.2.1.9 Design of Foundations and Subgrade Walls 5.1-12 5.1.2.1.10 Buried Tunnels, Piping, and Cables 5.1-13 5.1.2.2 clans II structures 5.1-14 5.1.2.3 clama III structures 5.1-14 l 5.1.3 GOVERNING CODES AND SPECIFICATIONS 5.1-15 Amesiament 2 5-1
4 4 } TABLE OF CONTENTS (Continued) ) , SecttQn Title Eggg 5.1.4 LOAD COMBINATIONS CRITERIA AND STRUCTURAL 5.1-15 ANALYSIS 1 5.1.4.1 At Design Loads 5.1-17 5.1.4.2 At Factored Loads 5.1-17 5.2 REACTOR BUILDING 5.2-1 5.2.1 CONTAINMENT STRUCTURE 5.2-1 s 4 5.2.2 INTERIOR CONTAINMENT STRUCTURE 5.2-1 i l 5.3 AUXILIARY BUILDING 5.3-1
- 5.3.1 GENERAL DESCRIPTION 5.3-1 5.3.2 DESIGN BASES 5.3-1 i 5.3.2.1 Design Loads 5.3-1 i
5.3.2.1.1 Dead Loads 5.3-1 i 5.3.2.1.2 Live Loads 5.3-2 l 5.3.2.1.3 Earthquake Loads 5.3-2 5.3.2.1.4 Wind Loads 5.3-2 5.3.2.1.5 Rain Loads 5.3-2 i , 5.3.2.2 Desian criteria 5.3-3 ! + 5.3.2.3 structural Desian Analysis 5.3-3 5.4 FUEL STORAGE BUILDING 5.4-1 5.4.1 GENERAL DESCRIPTION 5.4-1 5.4.2 DESIGN BASES 5.4-1 1 5.4.2.1 Desian Loads s.4-1 l Amendr. eat 2 5-11
l TABLE OF CONTENTS (Continued) Section Title Eagg 5.4.2.1.1 Dead Loads 5.4-1 5.4.2.1.2 Live Loads 5.4-2 5.4.2.1.3 Earthquake Loads 5.4-2 5.4.2.1.4 Wind Loads 5.4-2 5.4.2.1.5 Thermal Stresses 5.4-2 5.4.2.2 Desian Criteria 5.4-3 i 1 5.4.2.3 structural Demian Analysis 5.4-4 i 5.5 OTHER STRUCTURES 5.5-1 5.5.1 'A' NUCLEAR SERVICE SPRAY POND AND PIPE LINES 5.5-1 1 5.5.2 TURBINE BUILDING 5.5-1 5.5.3 COOLING TOWERS 5.5-1 5.5.4 STORAGE RESERVOIR 5.5-3 5.5.5 NUCLEAR SERVICES ELECTRICAL BUILDING 5.5-3 5.5.5.1 ceneral Descriotion 5.5-3 5.5.6 TRAINING AND RECORDS BUILDING 5.5-4 5.5.7 INTERIM ON-SITE STORAGE BUILDING FOR LOW 5.5-4 LEVEL RADWASTE 5.5.7.1 ceneral Description 5.5-4 5.5.7.2 Desian Basis 5.5-5 5.5.7.2.1 Codes, Standards, and Regulatory 5.5-6 Requirements 5.5.7.3 Material Reauirements 5.5-7 5.5.7.4 structural Reauirements 5.5-8 5.5.8 SOLIDIFICATION BUILDING 5.5-9 5-111
TABLE OF CONTENTS (Continued) l l Section Title Page 5.5.8.1 General Descrintion 5.5-9 5.5.8.2 Demian Basis 5.5-9 5.6 MATERIAY5 AND CONSTRUCTION PRACTICES 5.6-1 , 5.6.1 CONSTRUCTION ORGANIZATION 5.6-1 5.6.2 CONSTRUCTION SPECIFICATIONS 5.6-1 5.6.3 CONSTRUCTION MATERIALS 5.6-2 5.6.3.1 Concrete 5.6-2 5.6.3.1.1 Aggregates 5.6-2 5.6.3.1.2 Cement 5.6-4 5.6.3.1.3 Pozzolan 5.6-4 5.6.3.1.4 Water and Ice 5.6-4 5.6.3.1.5 Admixtures 5.6 5.6.3.1.6 Concrete Mix Design and Testing 5.6-5 5.6.3.1.7 Concrete Production and Testing 5.6-6 5.6.3.2 Reinforcina Steel 5.6-7 5.6.3.2.1 Materials 5.6-7 I 5.6.3.2.2 Mechanical Splices 5.6-7 5.6.3.2.3 Fabrication and Placement 5.6-10 l 5.6.3.2.4 Inspection and Testing of Reinforcement 5.6-11 5.6.3.2.5 Inspection and Testing of Cadweld Splices 5.6-11 5.6.3.3 Steel Pre-stressino Tendons 5.6-13 5.6.3.4 Liner Plate and Penetration Sleeves 5.6-14 1 l 5.6.3.5 Penetrations 5.6-14 l 5-iv
TABLE OF CONTENTS (Continued) i Section Title Eggg 5.6.3.6 structural and Miscellaneous steel 5.6-14 5.6.3.6.1 Materials 5.6-14 5.6.3.6.2 Fabrication and Erection 5.6-15 5.6.3.6.3 Inspection and Testing 5.6-15 5.6.3.7 Welder cualifications and Inspection of 5.6-15 Field Weld 1na 5.6.3.7.1 Welding Procedures 5.6-15 5.6.3.7.2 Welder Qualification 5.6-15 5.6.3.7.3 Welding Inspector Qualifications 5.6-15 5.6.3.7.4 General Inspection Procedures 5.6-16 5.6.3.7.5 Inspection of Post Weld Heat Treatment 5.6-17 5.6.3.7.6 Visual Inspection of Welds 5.6-17 5.6.3.7.7 Nondestructive Testing 5.6-18 5.6.3.7.8 Repairs 5.6-19 5.6.3.7.9 Records 5.6-19 5.7 SEISMIC INSTRUMENTATION 5.7-1
5.8 REFERENCES
5.8-1 5-v
_ _ . ___ . . = . _ . . . = - _ . . _ . _ _ _ . _ _ _ _ _ _ _ . . _ ._. . _ . - _ _ _ . - - _ _ _ .. _ . ._ . _ - _ _ . . . l LIST OF TABLES 4 f Table Title Eng.R 5.1-1 Seismic Criteria Summary 5.1-4 5.1-2 Hypothetical Wind Borne Missiles 5.1-8 5.1-3 Maximum Safe Wind Velocities for Class I 5.1-9 Structures 5.6-1 Aggregate User Tests 5.6-3 I l l l a I i 1 5-vi
D5AR CRAPTER 5. SiAUCTURES AND CONTAINMENT SYSTEM 5.1 GENERAL The information regarding the structural design bases for normal , operating conditions are governed by the applicable building design , codes. The basic design criterion for the design basis accident and seismic condition is that there should be no loss of function if that function is related to public safety. For Rancho Seco in the PDM and with a POL, only those structures which house and/or contain fuel or radioactive material are considered important to safety. i 5.1.1 CLASSES OF STRUCTURES , I Structures are placed in various classes, depending on how their ] function relates to the safe storage of spent fuel and protection of public health and safety. 5.1.1.1 class I l Class I structures are defined as those structures whose loss of j function could: 1 A. Cause or increase the severity of a design basis accident, or B. Result in a release of radioactivity to the site boundary in excess of the 1 percent of 10 CFR 100 guidelines. i Class I structures are designed to withstand the design basis seismic loads and other applicable loads without loss of function. Class I-structures are sufficiently isolated or protected from other structures to ensure that the integrity cf class I structures is maintained at all times. The following structures are required to remain Class I structures in r the PDM: 1 i A. Ausiliary Building, housing radioactive waste treatment systems, and B. Fuel Storage 3uilding, holding the spent fuel and radioactive core componer:a. 5.1.1.2 class II Class II structures are defined as those structures whose loss of function could interrupt spent fuel pool cooling or result in loss of control over normal releases of radioactivity. 5.1-1
! l i-i 5.1.1.3 clans III Class III structures are defined as those structures whose failure could ,
inconvenience normal plant operation but are not required for safe l storage of spent fuel and safe plant operations in the PDM. t j 5.1.2 DESIGN LOADS AND STRUCTURAL BERAVIOR ' 5.1.2.1 class I structures 9 E . ] 5.1.2.1.1 Normal Operations 5 Structural loads encountered during normal plant operations in the PDM are resisted by the structure through design methods that conform to the applicable codes and standards. , e 5.1.2.1.2 Accident and Seismic conditions 1 1 Class I structures are proportioned to maintain elastic behavior when 4 subjected to various combinations of dead loads, accident loads, thermal , , loads, and earthquake loads. There is one exception to this statement. ! The cask washdown structure is designed to go into plastic deformation
, under accident load conditions. Concrete structures are designed for .
ductile behavior to control stresses wherever possible. This stress l controlling design employs re-enforcing steel. I i The cask washdown structure is designed for unlimited use as a decontamination platform for spent fuel casks. The cask washdown structure is designed to go into plastic deformation, but maintain stability, if a loaded cask is ocopped onto the structure. The cask l
- washdown structure is designed to support only one accidental cask drop. i In addition, seismic stops are installed on the cask washdown structure j to prevent horizontal accelerations generated from a design basis earthquake from moving a loaded cask off the end of the structure and potentially causing the cask to tip over. I 5.1.2.1.2 Missiles li 4
Class I auxiliary structures are enclosed by reinforced concrete walls, with a minimum thickness of 10 inches, to withstand any impact from missiles that could be generated from the maximum design basis wind. l
)
l l i Amendment 2 5.1-2
5.1.2.1.4 Separation of Structures and Components The maximum displacement of structures, systems, and components was considered to ensure that major components have adequate structural separation. Structures designated as Class I as part of their original design are separated by six inches to keep them from contacting each other, assuming no interaction in the soil. Based upon the analyzed response of the structures, the designed separation is more than adequate to preclude one building from hitting the other. Interaction with other structures is also considered in the design. The foundation material is uniform and strong enough that no permanent settlement or tilting will result from seismic loads. 5.1.2.1.5 Seismic Design of Structures The operating basis earthquake (OBE) for Rancho Seco is a maximum ground acceleration of 0.13g horizontally and 0.09g vertically. The design basis earthquake (DBE) is a maximum ground acceleration of 0.25g horizontally and 0.17g vertically. The Nuclear Services Electrical Building (NSEB), which contains some electrical equipment that supports some Auxiliary Building loads, was designed for a three component OBE having a maximum ground acceleration horizontally and vertically of 0.13g. The damping effect of the soil was considered in the seismic analysis of the NSEB. Seismic loads on structures are determined by realistic evaluation of dynamic properties and the acceleration from response spectrum. The structures are designed for the horizontal and vertical seismic ground acceleration components acting separately, simultaneously, or in combination with each other. A summary of seismic criteria for Class 1, II, and III structures is shown in Table 5.1-1. The seismic criteria and additional seismic design information for the entire Rancho Seco facility is contained in the historical records of Appendix 5B of the USAR, Amendment No. 8. Class I structures which extend well below finished grade were analyzed as free-standing structures with the appropriate earthquake ground motion applied at their bases. No credit was taken for the damping effect of the soil on building seismic responses for the Reactor Building, Auxiliary Building, and the Fuel Storage Building. This approach yields higher design stresses in the structures than would actually be experienced during the specified seismic events. Torsional effects during the OBE and DBE were considered in the design of Class I structures. Torsional loads due to rotational components of ground motion, torsional loads due to eccentricities of mass and stiffness, and torsional loads due to differences between actual and chance eccentricities of masses and stiffness were distinguished from each other. Eccentricities due to yielding of parts of the structure 5.1-3 i
. _ . . _ . _ _ . . _ __ _. - _ _ _ _ _ _ _ . ~ . . . _ _ ._ ___ _ _ __
i J T
! i l
TABLE 5.1-1 i SEISMIC CRITERIA
SUMMARY
J , Response Class of Earthquake Intensity of Spectrum Allowab1<a Structure Condition Earthquake Method Used Stresse) i j I DBE 3.25g hor. Yes' Less than minimum guaranteed j 0.l'ig vert. Yes* yield stress of the material, or ~ the capacity of structure or component to resist the earth-1 OBE 0.13g hor. Ye:s' quake forces. Within allowable 0.00g vert. res' working stress limits as set , , forth in design standards. 2 II Equivalent 0.20g hor.' N o* Allowable working stress range i static load 0.10g vert.' N o* for materials but not increased for seismic loading. 9 III Equivalent 0.10g hor. No* Allowable working stress range ; static load 0.05g vert. N o* may be increased for seismic f loading. NOTES: (1) Minimum values used. Where applicable, Uniform Building Code (1967 edition), Zone 2 or greater load was used. (2) Indicates that equivalent static loading was used. t (3) For Class I structures, the " Peak Value Approach" and " Testing Approach" of Seismic Design Criteria may be substituted for the " Response Spectrum Method." 5.1-4
I l
)
i 1 5.1.2.1.5 Seismic Design of Structures (continued) 1 were not considered, si.nas the structures are designed to withstand , these loadings within the yield strain values. I The torsional loads due to actual and chance eccentricities for the Auxiliary Building was evaluated as follows. The initial design of this i structure regarding seismic loading was done using a one dimensional l lumped mass model. This design was later checked using a three-dimensional model that automatically considered the actual eccentricities. All stiffness and masses were included in this three-dimensional model at their proper geometrical location. 1 Because of the complexity of the model, the NASTRAN program is used to accommodate all the static and dynamic degrees of freedom. It is believed that the actual eccentricities in this structure were handled as well as possible with the most recent computer capabilities. The resulting stresses were so low that chance eccentricities would not have significantly changed the comparatively low stresses considering the fact the Auxiliary Building is so asymmetrical. Regardless of the signs of the stresses obtained, they were added to maximize the stresses from the other loadings on the structure. The response spectra seismic analysis method was used as the basis for the dynamic analysis of the Reactor Building, Auxiliary Building, and the Spent Fuel Building. The mathematical models developed were in accordance with Bechtel Corporation Topical Report BC-TOP-4, " Seismic Analysis of Structures and Equipment for Nuclear Power Plants." In particular, the report was applied as a general method of approaching three principal elements of the analysis: modeling techniques, damping, and the computation of structural response. Sections 3.0, 4.0 and 5.0 of the Topical Report were applied in total for the three structures listed above. The techniques and methods of analysis included therein were then combined with additional analyses inherent to the design of the individual structures (i.e., the hydrodynamic analysis of the Fuel Storage Building, the three ; dimensional finite element dynamic analysis of the Auxiliary Building, and the three dimensional lumped mass dynamic analysis of the Turbine Building) for a more complete and exact analysis of the respective structures. I BC-TOP-4 presents specific techniques that were used for the modeling of I the Reactor Building as well as general approaches for modeling other structures. It evaluates and establishes the damping coefficients that were used in conjunction with the modal represea ?. tion of Class I structures, and it establishes techniques for 40<c. ting the response of the various structures for the specified ground ex;ttation. ! l l I 5.1-5
5.1.2.1.5 Seismic Design of Structures (Continued) A lumped mass model was developed using tributary sections for the analysis of the Reactor Building. For a typical sketch of the model used see Figure 3.4 in BC-TOP-4. Stiffness coefficients were determined assuming beam behavior of the structure. The soil-structure interaction affecting the Reactor Building was based on Section 3.3 of BC-TOP-4. The equipment hatch, personnel lock, and escape lock were designed as Class I components. Stresses within the locks and hatch due to seismic loads were also incorporated. Seisaic forces due to supporting these components were considered in the design of the Reattor Ge1.lding and secondary reinforcement was provided to withstand lcpsl ccscentrated loads. Seismic forces in the Auxiliary Building were determined using a three-dimensional modal analys.is. A plan stress finite element model was developed which included both walls and floors. For typical elevation and plan views of the model used see Figures 5.1-1 through 5.1-4. A Guyon reduction was applied to the modal analysis in order to reduce the structure to 135 degrees of freedom. The seismic stresses developed were found to be within the limits allowed by the applicable codes. A lumped mass model using tributary sections was also used for the seismic analysis of the Fuel Storage Building. This is depicted in Figure 5.1-5. The stiffness coefficients were determined assuming beam behavior. Hydrodynamic forces were considered in accordance with TID-7024, Nuclear Reactors and Earthquakes. The nuclear service spray , ponds were also analyzed for the effects of hydrodynamic forces per TID-7024, and sufficient freeboard was provided. The mathematical model for the Nuclear Services Electrical Building was developed from Bechtel Corporation Topical Report BC-TOP-4A, " Seismic Analysis of Structures and Equipment for Nuclear Power Plan'.s". 5.1.2.1.6 Wind Loads Wind loading for the original plant structures are based on Figure 1(b) of ASCE Paper 3269, " Wind Forces on Structures," using the highest wind speed for a 100-year recurrence period. ASCE Paper 3269 was also used to determine shape factors, gust factors, and variation of wind velocity with height. Based upon the site location and inland classification, the design wind velocity for these structures is 90 mph at a reference 30 feet above ground level. In addition to design for normal wind loadings, Class I structures have been analyzed to assess their capability to withstand short-term extreme wind loads. It was determined that all Class I structures are capable of withstanding extreme wind loads of approximately twice the stated ! design wind load. I 5.1-6
- - . - - - . . - - . . . . - - . . . - . . - . - - . . - . . . - - - - . ~ - . ~ 5.1.2.1.7 Tornado Loads The probability of a tornado or winds in excess of 101 mph in the vicinity of the Rancho Seco site is considered to be low; however, Class ; 1 I structures have been investigated to determine the maximum wind forces I and wind-generated missile impacts which could be sustained without loss of function. Rancho Seco established a 101 miles per hour wind velocity for its wind design criteria, which is a conservative figure for this-location. Structural capabilities under wind loadings are similar to seismic loadings and are generally in excess of the design criteria since earthquake loadings have governed the design. The number and size of wind-borne missiles which could be experienced during a windstorm are a function of the maximum wind velocity and do not constitute a significant hazard at lower velocities. For a conservative estimate, it was assumed that design basis missiles would ! be generated at velocities in excess of 101 miles per hour. i To determine that maximum wind velocities under which the function of Class I structures would not be affected, ASCE Paper No. 3269 was used to analyze wind forces on structures. Various reports on missile ; impact, including Stanford Research Institute SRIA-ll3, NAVDOCKS NT-3726 I and U.S. Army TM5-855-1, were used to determine impact hazards. The wind-borne missiles studied in the analysis provide a conservative representation of the missiles which can be expected. These include a ! 4" x 12" x 10' board weighing 108 pounds, a 3-inch schedule 40 pipe weighing 76 pounds, and a 4000-pound automobile at heights to 25 feet. ! The plank is assumed to move at the velocity of the wind and the pipe I and automobile at reduced velocities. The relationship between wind ! velocity and missile velocity is shown in Table 5.1-2. Table 5.1-3 shows the results of the investigation, listing the maximum safe wind velocities for forces and missile impact. In all cases, the worst missile conditions are listed. Where calculated wind velocities fall between the listed levels, the lower level of velocity is given. The same element of conservatism was used in the selection of wind force shape factor coefficients. The study of missile impact effects is limited by the available technology and the complexity and diversity of the possible modes of failure. Extrapolation of a laboratory study of impact'on flat plates to actual field conditions requires extensive study and a high level of judgment and experience. The tabulated values are believed to be conservative in that all foreseeable modes of failure have been studied and conservative assumpt. cons made. Failure of the hyperbolic cooling towers would not affect Class I structures. Another area of concern is the overturning of the turbine gantry crane. With no consideration of the gantry crane tie-downs, a wind force in excess of 135 miles per hour would be required to overturn 5.1-7
. . _ - . _ _ . . . m_ . . . - . . __ _ - _ _ _ _ . _ _ _ _ __ . . _ . . . . _ _ _ _ . . , _ . _ _ . _ . . . __.__
TABLE 5.1-2 HYPOTHETICAL WIND BORNE MISSILES Missiles: Type 1. 4" x 12" x 10' Board 9108 pounds Type 2. 3" diameter Schedule 40 pipe 976 pounds Type 3. 4000 pound Automobile lifted up to 25' in height Wind Velocity (mph) 300 250 200 175 101 Missile 1 Velocity 300 250 200 175 101 Missile 2 Velocity 100 85 67 60 33 Missile 3 velocity 50 42 33 30 NA 5.1-8
. . . ._ - . . . . . _ _ . . . _ . _ . _ . . . _ . _ . . ~ _ . . _ . _ . . . . . . . . - . . . . _ . .___...___..___.__._._.m.. _ l TABLE 5.1-3 MAXIMUM SAFE WIND VELOCITIES FOR CLASS 1 STRUCTURES 200 mph 175 mph Structure Wind Missile Wind Missile l j Fuel Storage Building X X Auxiliary Building X X Auxiliary Building doors X X' I at 40' Elevation (Exterior). f NOTES: (1) The corridor door will not withstand 175 mph missile, but missile entry into the control room is not possible. ! i 1 5.1-9
5.1.2.1.7 Tornado Loads (Continued) the crane. With the tie-downs, the crane and its supporting structure (the Turbine Building) are safe at winds up to 250 miles per hour. When not in use, the turbine gantry crane will be tied down. The two exterior doors to the Auxiliary Building at elevation 40 ft were analyzed for the 175 mph wind and a 108-pound plank missile having a velocity of 175 mph. These doors are the double doors on the north wall leading to the east-west corridor and the control room emergency exit door on the west wall at the southwest corner. The control room emergency door will withstand the pressure loading from a 175 mph wind. A concrete missile shield prevents the plank missile from impacting the emergency door. The corridor double doors will withstand the pressure loading from a 175 mph wind. In the analysis of potential missile entry, it was found that missiles entering the corridor could not enter the control room area by either a direct hit or a ricochet. 5.1.2.1.8 Seismic Design of Equipment, Structures, and Supports class I Eauinment and comnonents Sunnlied by the District The only piece of equipment or component supplied by the District that is required to be maintained as Class I in the PDM is the spent fuel pool liner. This component was analyzed and/or evaluated based on the modal response spectra developed for various structure elevations. The time-history analysis was used only as a means of obtaining in-structure response spectra for equipment inside structures. These modal spectra curves were developed using dynamic models excited by an accelerogram from the 1952 Taft earthquake. The following is a description of the methods and criteria used for the analysis. All Class i equipment and components were designed according to the seismic specifications prepared by the engineer. The seismic specification contains response spectrum curves at the floor elevations for both the OBE and DBE generated from the floor acceleration time history for the Class I structures. The seismic analysis for Class I equipment and components are conducted by the manufacturer. The manufacturer is given two ways by which the equipment can be qualified, i.e., by dynamic analysis and/or by testing. Dynamic tests are normally accomplished on a shaker system with base connections simulating the actual installation configurations. The input excitation equals or exceeds the specified DBE acceleration levels. 5.1-10
l l l l 5.1.2.1.8 Seismic Design of Equipment, Structures, and Supports (Continued) Criteria for Lumnina Mannes For structure modeling, supported elements are lumped with the structure i masses. Under these circumstances the effect of the supported element ! stiffness is neglected. However, subsequent response motions used in l the component seismic analysis are not directly applied to the ' equipment. Consideration is given to the immediate supporting system. Generally these supports are included in the modeling of the equipment. l Thus, a series of subsystems are analyzed whereby the initial structural response is applied to the first subsystem. From the analysis of the l first subsystem, a second motion is obtained to be used as input to the i second subsystem and so forth. Thus, a flexible slab lumped into the overall model because of its relatively small mass ratio is modeled with its mass and stiffness as the support for the component. The gross structured response is then applied with the proper boundary conditions and the response of the component obtained. Criteria for Comnutina Stresses and combinina Modal Frecuencies I The shears, moments, stresses, deflections and accelerations were { computed separately for each mode of vibration. The resultant vibration I for all considered modes is obtained by taking the " square root of the sum of the squares." In the structures considered there were no closely spaced frequencies. The structures were analyzed for the independent input motions in both vertical and horizontal directions. The structural members were 4 designed to the maximum combination of these forces and displacements, I where the sign of one or the other direction was used to maximize the stresses. With respect to the floor response spectra, the response spectra in each direction due to both motions applied independently are additive and this occurs only for three dimensional solutions. l Criteria for De te rmi ni na Nn=her of Modes Considered l A. District Sunplied Eauipment For the response spectrum modal analysis of items other than piping systems, all modes less than or equal to 20 cycles per s<scond were considered. However, if the number of modes below 20 cycles per second was less than or equal to three, the modes were combined by summing the absolute value of the response. Otherwise, the " square root of the sum of squares" method was used. 5.1-11
5.1.2.1.8 Seismic Design of Equipment, Structures, and Supports (Continued) The effects on floor response spectra (e.g., peak width and period coordinates) of expected variations of structural properties and soil structure interactions were considered. The most important characteristic of the floor response spectra is the building frequency. Because of the very high magnification of motion at that frequency, the calculated frequencies of the buildings were carefully determined. Considering all the variables of the structure and soil (elastic modulus, stiffness properties, mass distribution, discretization of the continuous properties and Poisson's ratio) the development of a reasonable probabilistic model would be extremely difficult. Therefore, the possible variations of the building frequencies were determined by an evaluation of each individual property and engineering judgment. For the Auxiliary Building a frequency shift to 120 percent is used. Depending on the final shape of the curves, further conservative smoothing of the curves were made. The properties that contribute to the frequency shift, including the soil, wore evaluated at the same time. At the time of design, no averaging of peak response was made because the state-of-the-art was not developed sufficiently to justify any reductions. ' 5.1.2.1.9 Design of Foundations and Subgrade Walls
]
Soil interaction between adjacent structures was not considered in the design of the foundations for Class I structure foundations. When the l Auxiliary Building and the Fuel Storage Building design was completed, l this phenomena was not considered significant because of the l substantially different characteristics of the dynamic seismic response for the Class I structures. Soil interaction between the Auxiliary Building and the Fuel Storage Building during an OBE or DBE disturbance was considered minimal and therefore neglected. However, relative displacements between structures affecting interconnecting elements were considered in the design. The Auxilicry Building is separated from the Reactor Building by a six inch gep extending down to elevation (-) 20 feet. Between the tendon gallery and Auxiliary Building sub-basement this gap extends to elevatjer (-)47 feet. For the seismic analysis, the building is considered fixed at these levels and at grade level on the west side. The rigadity and size of the sub-basement makes this a valid and conservative assumption. Soil loadings on the walls of the Auxiliary Building below grade are quite complex. Surcharge loadings from the adjacent Fuel Storage Building are substantial and have been included in the analysis. Other walls, which are not affected by this building, have been designed to sustain an arbitrary 500 psf surcharge in addition 5.1-12
-- -. - . . . . _ _ _ . - - - - . . - . - - . . - - _ _ _ , . . - ~ _ . . ~ . - -
1 a 1 1
- 5.1.2.1.9 Design of Foundations and Subgrade Walls (Continued)
I to the passive soil pressure. In modeling the structure for dynamic analysis as a three-dimensional system it was found that the soil-structure displacements were insignificant and that soil-structure interaction could be neglected. It should be noted that the cellular configuration of the Auxiliary Building substructure creates a situation i in which exterior walls transverse to the loading will have negligible 1 stiffness in relation to the supporting interior walls acting as shear j diaphragas, i
- The Fuel Storage Building is embedded six feet in the ground. This is
] the thickness of the base mat. No soil loading exists on the. walls. 1 l The overturning moments for the Auxiliary Building and the Fuel Storage l l Building were calculated by the modal response spectrum method using the
" square root of the sum of squares" method whenever the number of modes j considered was more than three. The stresses due to the overturning j moment and vertical accelerations were superimposed to the stresses i obtained from the other loadings. In the containment structure, the
( seismic stresses were added to the other stresses obtained from the l cracked section analysis. I i j For the Auxiliary Building, the seismic soil stresses, both for j horizontal and vertical motions, were obtained from the three i dimensional finite-element dynamic analysis program, NASTRAN. These j were extremely small and in most cases the dead and live load l combination governed. I I I 4 j The dynamic soil stresses under the Auxiliary Building are less than 1.6 ! kips per square foot and provides for a safety factor of 12 based on 1 allowable dynamic soil pressure. i ' s I i j 5.1.2.1.10 Buried Tunnels, Piping, and Cables l j The seismic response of buried elements, such as tunnels, underground ' j piping, and underground cables is different than the response of main j structures. The stresses in the buried elements at their junctions with . the main structures are controlled by a variety of means. l A mathematical model for the DBE was used for a dynamic analysis of the
- Auxiliary Building. The results of the NASTRAN computer program output j indicated displacement of the structure to be less than 1/16 inch in the l region below grade. The Rancho seco soil is quite stiff and will not I allow much relative motion between the structure and the soil as indicated by the small differential motion determined by the analysis.
- Due to the small amount'of differential movement, bending stress introduced in the piping and total stress will remain within allowable i stress limits. Piping that penetrates the Auxiliary Building was 1
I 5.1-13 i i J
5.1.2.1.10 Buried Tunnels, Piping, and Cables (Continued) wrapped with two layers of polyethylene tape, which is an effective bond breaker between the pipe and concrete. This condition will prevent the piping from developing any undue axial stresses. Electrical concrete encased duct banks are separated all around by 1 inch of compressible material. This effectively isolates the concrete encased electrical duct runs from the structure. In conjunction with the isolation of the electrical duct runs, cables are looped inside the building to provide sufficient slack to prevent severance. There are no underground tunnels, piping, or electrical connections to the spent fuel pool. 5.1.2.2 class II structures , Structures designated as Class II are designed in accordance with standard practice. However, in no case is the design criteria in this classification of structures less restrictive than that required by applicable codes and standards or the requirements of th= Uniform Building Code. See Table 5.1-1 for class II structure seismic criteria. Adequate separation (minimum of 1 1/2 inches) between Class I and Class II structures is provided to preclude interference under a seismic event. This minimum separation occurs at elevation 40 ft above grade between the Turbine Building, the Auxiliary Building, and the Fuel , Storage Building. The maximum calculated movement of the Auxiliary Building is 1/4 inch and for the Turbine Building it is 3/8 inch. The maximum movement of the buildings relative to each other is 5/8 inch. This would occur only if the building motions were out of phase. Other Class II structures are located such that their failure will not endanger Class I structures. Where Class II elements are supported by Class I structures and/or are in the vicinity of Class I components, the design is based on precluding damage to the Class I components or structures from failure of the Class II elements. This was accomplished by increasing the physical separation distances to satisfy Class I criteria, by designing supports for Class II elements to meet Class I requirements, or by providing barriers to ensure that failure of Class II elements will not result in damage to adjacent ~'$ ss I components or structures. 5.1.2.3 class 211_ Structures Class III structures are designed in accordance with the requirements of the Uniform Building Code. See Table 5.1-1 for Class III structure seismic criteria. 5.1-14
1 l l I 5.1.3 GOVERNING CODES AND SPECIFICATIONS i The following codes and specifications are the basis for the design and { construction of all original plant structures. Modifications to these codes and specifications are made where necessary to meet the specific requirements of the structure. These modifications are noted in the description of the structures.
- 1. Uniform Building Code (1967 Edition) l
- 2. Building Code Requirements for Reinforced Concrete (ACI 318-63) !
i
- 3. Specification for Structural Concrete for Buildings (ACI 301-66)
- 4. Manual of Steel Construction (AISC Sixth Edition) {
- 5. ASME Boiler and Pressure Vessel Code (1968 Edition)
- 6. ASTM Standards - Materials and Testing Procedures used are j referenced in the appropriate sections. Nhenever possible ASTM or !
ASME and testing procedures were used. New structures are designed to the codes and specifications identified in the description for the new structure. l 5.1.4 LOAD COMBINATIONS CRITERIA AND STRUCTURAL ANALYSIS l 1 h Specific load combinations, criteria, and a description of the methods of analysis are included in the following sections for the major l structures. In general, two types of loading conditions, criteria, and ) analysis are used for the design of Class I structures: A. The design load case, where normal working stress design methods are used. l B. The factored load case, where the capacity of the structure is used to verify its ability to withstand load combinations in excess of the maximum that could be expected under accident conditions. The load factors and load combinations in the design criteria represent a consensus of opinion among the Bechtel engineers and consultants experienced in both structural and nuclear power plant design. Their judgment was influenced by current and past practice, by the degree of ) conservatism inherent in the basic loads, and particularly by the probabilities of coincident occurrences in the case of accident, wind, and seismic loads. The following is the justification for the individual factors. 5.1-15
5.1.4 LOAD COMBINATIONS CRITERIA AND STRUCTURAL ANALYSIS (Continued) A. Dead Load Dead load in a large structure such as this is easily identified, and its effect can be accurately determined at each point in the structure. For dead load in combination with accident and seismic or wind loads, a load factor representing a tolerance of 5 percent was chosen to account for dead load inaccuracies. The ACI Code allows-a tolerance of +25 percent and -10 percent, but the code was written to cover a variety of conditions where weights and configurations of materials in and on the structure may not be clearly defined and are subject to change during the life of the structure. B. Live Lnad The live load that would be present along with accident, i seismic and wind -loads would produce a very small portion of the stress at any point. Also, it is extremely unlikely that
- the full live load would be present over a large area at the 4
time of an unusual occurrence. For these reasons, a low load
- factor is felt to be justified and live load will be considered together with dead load at a load factor of 1.00 1.05.
C. Sei ==i c The operating base earthquake (OBE) is considered to be the strongest probable earthquake which could occur during the life of the plant. In addition to the operating base earthquake, a design base earthquake (DBE) which defines the maximum credible earthquake that could occur at the site was considered in the design. Structures originally designed as Class I are designed so that no loss of function would result from a DBE. Consequently the probability of an earthquake causing a design basis accident is very small. For this reason, the two events, seismic and accident, were considered together, but at much lower load factors than those applied to the events separately. The earthquake load-factors of 1.25 and 1.0 used in the containment design are conservative for the OBE and DBE. D. Mind Loads are determined from the highest wind speed for a 100-year occurrence as shown in Figure 1 (b) , ASCE Paper No. 5.1-16
~ - _ - _ , _ . _ _ . . _. - . ~ . _ . . --- . . _ -
l 5.1.4 LOAD COMBINATIONS CRITERIA AND STRUCTURAL ANALYSIS (Continued) 3269, " Wind Forces on Structures." With the containment 4 structure designed for this extreme wind, it is inconceivable that the wind would cause an accident. However, as with seismic loads, wind loads will be considered with accident loads, but at reduced load factors to reflect the remote i chance of simultaneous occurrence of both extreme load conditions. E. Accident l The probability of a design basis accident occurring simultaneously with a maximum wind or seismic disturbance is very small; therefore, a reduced load factor of 1.25 was used for the combination of events. i The Auxiliary Building was designed based on the working stress design method using ACI 318-63, ar d the design was verified using factored loads and the ultiwate strength design method. The Fuel Storage Building was designed, based on the working stress design method using ACI 318-63, and checked by factored loads using the ultimate strength design method. 5.1.4.1 At Desian Loads The structures are designed for all normal conditions of loading, including dead load, live load, temperature and wind. They are also designed for the operating basis earthquake. 5.1.4.2 At Factored Loads The structures are checked for various combinations of factored loads. The load factors are the ratio by which loads are multiplied for design purposes to assure that the load / deformation behavior of the structure is one of low strain behavior. The load factor approach is used to make a rational evaluation of the isolated factors that must be considered to assure an adequate safety margin for the structure. This approach permits the designer to place the greatest conservatism on those loads most subject to variation and that most directly control the overall safety of the structure. It also places minimum emphasis of the fixed gravity loads and maximum emphasis on accident and earthquake or wind loads. 5.1-17
5.2 REACTOR BUILDING Since there are no design basis events associated with the Reactor Building that can cause or increase the severity of a design basis accident, or result in a release of radioactivity in excess of one percent of the 10 CFR 100 guidelines, the Reactor Building is not required to be maintained as a Class I structure during the PDM. The Reactor Building is not required to function in the PDM, except for ALARA considerations to minimize potential occupational personnel exposures and for safe storage of irradiated and contaminated core ' components and reactor systems. The following is provided for information only. 5.2.1 CONTAINMENT STRUCTURE , The containment structure is a fully continuous reinforced concrete structure in the shape of a cylinder with a shallow domed roof and a flat foundation slab. The cylindrical portion was prestressed by a post-tensioning system consisting of horizontal and vertical tendons. The dome has a three-way post-tensioning system. The foundation slab is reinforced with conventional reinforcing steel. A welded steel liner is attached to the inside face of the concrete shell. The dimensions of the building are: A. Inside diameter; 130 feet, B. Inside height; 185 feet, C. Vertical wall thickness; 3 3/4 feet, D. Dome Thickness; 3 1/2 feet, E. Foundation slab; 8 feet, and F. Internal net free volume; 1,980,000 cubic feet. 5.2.2 INTERIOR CONTAINMENT STRUCTURE The Reactor Building interior structure consists of a concrete primary shield wall surrounding the reactor vessel, secondary concrete shields surrounding the remainder of the reactor systems, equipment accesses, and a maintenance floor and platforms. The reactor vessel is supported by an integral support skirt. The two steam generators are supported by integral support skirts and lateral seismic restraints located at the upper tube sheet. The pressurizer is mounted on a framed steel support structure in one of the steam generator compartments. The four reactor coolant pumps and motors are ; supported by constant support spring hangers and lateral seismic shock l suppressors. l 5.2-1 j
- . _ . . _ ...m . . _ _ .. _ _ _ __, . _ . . . . _ . . _ _ _ _ _ _ . __ _ _ _ . _ , . . . - - i I 5.3 AffYTLTARY RUILDING l 5.3.1 GENERAL DESCRIPTION The Auxiliary Building is a three-story reinforced concrete structure with a basement and sub-basement located adjacent to both the Reactor Building and Turbine Building. The basement houses the radioactive waste treatment systems and equipment. The Auxiliary Building also contains the control room and most of the switchgear required to operate the plant during the PDN. The Auxiliary Building is designed and constructed as a Class I structure. The structure is designed with a vertical load carrying reinforced concrete frame and internal reinforced concrete shear walls to carry the lateral loads. Figure 5.3-1 shows typical details of the l Auxiliary Building. ~ 5.3.2 DESIGN BASES l 5.3.2.1 Demian rm.a. The Auxiliary Building is designed for all credible conditions of loading, including normal loads and loads during an earthquake. The following loads are considered: A. Structure and equipment dead loads, I I B. Live loads, ' C. Earthquake loads, D. Wind loads, and j E. Rain loads. The critical loading conditions are those caused by an earthquake. 5.3.2,1.1 Dead Loads Dead loads consist of the weight of the concrete, structural steel, and interior partitions, equipment, piping, and electrical conductors. 5.3-1
e 5.3.2.1.2 Live Loads Live loads for the design of the structural framing members are consistent with the intended use of the structure and the recommendations of the Uniform Building Code (UBC). l I 5.3.2.1.3 Earthquake Loads l 1 Earthquake loading is predicated upon an OBE at the site having a ', horizontal ground acceleration of 0.13g and .a vertical ground acceleration of 0.09g. In addition, the DBE with a horizontal ground acceleration of 0.25g and a vertical acceleration ground of 0.17g is used to check the design to ensure no loss of function. Seismic response spectrum curves for both horizontal and vertical ground motion are contained in the historical records of Appendix 5B of the USAR, .Tmendment No. 8. A dynamic analysis was used to arrive at equivalent static loads for Auxiliary Building design. 5.3.2.1.4 Wind Loads Wind loading is based on Figure 1(b) of ASCE paper 3269, " Wind Forces on Structures" using the highest wind speed for a 100-year recurrence period or the recommendations of the UBC, whichever is greater. 5.3.2.1.5 Rain Imads Specific rainfall loads were not incorporated in the design of the Auxiliary Building roof, since the annual rainfall in the Rancho Seco site area is small. However, the Auxiliary Building roof can support 13 inches of water. The roof drain system is sized to remove 2 inches of water per hour. A rainfall of this intensity has a duration of less than 30 minutes with a 100-year frequency of recurrence. The roof is cloped towards each of four 5-inch drains located along a line parallel to the east-west axis, and towards a single 3-inch drain located in the southwest corner. If the normal drains were to clog, adequate drainage would be provided by an 8-inch wide scupper located 3 inches above the lowest point of the roof. In addition, there is an 8-foot wide pipe opening located 10 1/2 inches above the lowest point in the roof. I i 5.3-2
5.3.2.2 Desian criteria The main considerations in establishing the structural design criteria for the Auxiliary Building is to provide a structure which will withstand the normal operating loads and the loads from an earthquake. Except as noted in the UBO, design methods and allowable stresses r.re for the design of reinforced concrete and steel. The strength of the structure at working stress and ultimate capacity is compared to various loading combinations to ensure safety. The structure is designed to meet the performance and strength requirements under both the design loads and factored loads. 5.3.2.3 Structural Desian analvsis The static analysis of the Auxiliary Building is moment distribution using skip loadings where applicable for design of the slabs and beams. No reduction in live or dead loads was taken for the column and wall loadings. A minimum 0.25 percent reinforcing steel in both directions is provided in all the shear walls with additional steel provided at all openings. The stresses due to seismic forces in concrete and reinforcing steel are checked using the values obtained from the dynamic analysis of the building. In critical areas the structure is analyzed by both the working stress and ultimate strength methods. 5.3-3
l 5.4 FUEL STORACE BUILDING 5.4.1 GENERAL DESCRIPTION The Fuel Storage Building is a Class I reinforced concrete rectangular tank with a stainless steel liner and a super structure with concrete j walls and steel roof system. l The spent fuel pool and storage racks are designed to hold up to 1080 spent fuel assemblies and to withstand all external influences, particularly seismic conditions. The design of the walls takes into account the hydrodynamic effects of the water including both impulsive and convective forces generated by the inertial characteristics of the mass of water during earth movement. The building is structurally separated from the Auxiliary Building, and bellows joints are provided in the fuel transfer system to allow for differential movement between the Reactor Building and the spent fuel pool due to thermal or earthquake loads. Figure 5.4-1 shows typical details of the Fuel Storage Building. i 5.4.2 DESIGN BASES 5.4.2.1 Desian Loads The Fuel Storage Building is designed for all credible conditions of loading, including normal loads and loads from a design basis earthquake. The following loads are considered: A. Structure dead loads B. Live loads C. Earthquake loads D. Wind loads The critical loading conditions are those caused by the hydrostatic effects of the spent ivel pool water, a spent fuel shipping cask, and an earthquake. 1 5.4.2.1.1 Dead Loads t Dead loads consist of the weight of the concrete, structural steel, equipment, and cask washdown structure. Amendment 2 5.4-1 1
l 5.4.2.1.2 Live Loads The following live loads are considered in the design: A. Roof loads of 20 pounds per square foot t B. Hydrostatic loads from the spent fuel pool filled with water C. Loads from the spent fuel D. Loads from the spent fuel cask 5.4.2.1.3 Earthquake Loads Earthquake loading is predicated on both an OBE at the Rancho Seco site having a horizontal ground acceleration of 0.13g and a vertical ground acceleration of 0.09g. The DBE having a horizontal ground acceleration of 0.25g and a vertical ground acceleration 0.17g is used to check the ' design to ensure no loss of function. Seismic response spectrum curves for both horizontal and vertical ground motion are located in USAR Amendment No. 8, Appendix 5B for historical reference. A dynamic analysis which includes the hydrodynamic effect of the water is used to arrive at equivalent static loads for ruel Storage Building design. j
\
5.4.2.1.4 Wind Loads Wind loading is based on Figure 1(b) of ASCE Paper 3269, " Find Forces on Structures" using the highest wind speed for a 100-year recurrence ' period or the recommendation of the UBC, whichever is greater. 5.4.2.1.5 Thermal Stresses Reinforcement for crack control for the spent fuel pool is in accordance with ACI-318-63, as a minimum. In the PDM, abnormal thermal stresses are not a concern for the design of the fuel pool. The heat-up analysis for the spent fuel pool during ~ the PDM indicates the maximum temperature the spent fuel pool could reach with a complete loss of cooling and no make-up water is 185'F. Two spent fuel pool cooling systems and several make-up water methods are available to the spent fuel pool. In case of the functional loss of these systems, sufficient time is available to re-establish spent fuel cooling and make-up water to the spent fuel pool to ensure the spent fuel . pool temperature does not exceed 180*F. The spent fuel pool can withstand a water temperature up to 212*F. 5.4-2
.. _. . . . _ . . . . . __. - . . -- = _ ._=_ ._- - _.- - - - - -
1 5.4.2.2 Desian criteria The main consideration in the structural design criteria for the Fuel Storage Building was to provide a leak-tight pool to contain spent fuel under all conditions of loading, including earthquakes. The spent fuel pool is designed with a leak-chase system that collects and conveys any l j leakage to the radioactive waste treatment systems in the Auxiliary l j Building. A leak-chase monitoring system keeps track of the leak rate l volume from the spent fuel pool. ! l Except as noted in these criteria, the ACI 318-63 and AISC, Sixth 5 Edition, design methods and allowable stresses were used for the design ! of reinforced concrete and steel, respectively. I I ] The strength of the structure at working stress and over-all yielding 1 was compared to various loading combinations to ensure safety. The structure is designed to meet the performance and strength requirements ! under the following conditions: I A. At design loads B. At factored loads C. Loads from spent fuel D. Loads from the spent fuel cask Within the Fuel Storage Building, the greatest height to which a spent l fuel cask can be lifted is 40 feet immediately above the floor of the , spent fuel loading pit. This maximum lift is established by limitations i of crane hook travel, cask bail, and long sling. The floor of the spent j fuel loading pit is depressed 4'-6" below the main floor of the pool. l The pit is 10'-6" by 12'-6" in plan and 6 feet thick. It is continuous I with the 6-foot-thick main floor. The 3/16-inch-thick stainless steel liner plate is continuous over the entire inside pool surface. The i floor of the spent fuel loading pit is backed by an embedded stainless steel plate 8'-0" by 10'-4" by 1" thick to which the liner plate is welded. In evaluating the functional capability of the spent fuel loading area slab, a maximum anticipated cask size was assumed. This was a conceptual cask, having a gross weight of the cask, spent fuel, basket, l and water of 195 kips. It had a base diameter of 6'-8" and a height of ) approximately 17 feet. l The thickness of the concrete slab is adequate for the hypothetical impact evaluated in the following analysis. j The drop of 40 feet was through water which results in an effective buoyant weight of 155.6 kips. The cask was assumed to be a cylindrical, 5.4-3
5.4.2.2 Denian cri teria (Continued) smooth sided vessel with blunt, plane ends. No consideration was made for the rugosity of the sides nor for the increased drag contributed by the energy absorbing fins at both ends. The velocity at the point of impact in the spent fuel pool was found to be 34.05 ft/sec. Since this velocity is less than the 43.8 ft/sec velocity that corresponds to a 30-foot free fall in air, cask integrity will be unaffected by the evaluated cask drop in water. To simplify the analysis of the effect on the spent fuel pool base slab under this cask impact loading, the bottom of the pit was assumed to be an isolated footing 22'-6" by 24'-6". The closest analogy for the dropped cask loading is a drop-forge foundation design using the soil parameters for the Rancho Seco site. This method was used for the soil / slab analysis. Maximum deflection was found to be 0.67 inches or 1/400 of the span. Actual deflection will be less since the affected foundation mass is substantially larger. No account was taken of the energy absorbing fins on the cask, which will increase the deceleration distance to approximately six inches, making the calculated deflection amplitude almost an order of magnitude lower. Since this analysis did not consider the impacting area and that scabbing of the top surface is inhibited by the 1 3/16-inch stainless steel facing, the slab is considered adequate for any angle of incidence, particularly since the energy-absorbing fins will tend to spread the loading over larger areas. No experimental development work has been performed on this project, and no analogous unclassified test data is available. The method of analysis, however, is based on well-substantiated technical procedures. The fuel storage pool would not be functionally damaged by the maximum anticipated spent fuel cask drop. 5.4.2.3 Structural nesian An=1 vain The fuel storage pool was analyzed using coefficients obtained from PCA ST 63. The hydrodynamic forces were computed by the method outlined in TID 7024. The superstructure was analyzed by conventional UBC methods. Attention was given in the analysis and design details to providing a structure that would yield limited deformations during all loading conditions, thereby lis.iting the strain on the stainless steel pool liner. 5.4-4
5.4.2.3 structurni nemian Analvnin (Continued) ! The pool is supported on a soil foundation; thus, the effects of dropping a spent fuel cask are minimized, since the loads would be transmitted directly from the slab to the foundation material. The . liner plate in the area where the cask is set down is thickened to further protect against the possibility of damaging the liner. The , layout of the spent fuel pool is such that at no time is a spent fuel l cask lifted over spent fuel. The superstructure of the fuel storage building was analyzed for loads from a 200 mph wind. t 6 5.4-5 l l j
- . _ , - ._, .. __ . .. . = -
Page 2 of 2 TABLE 7.4-1 (Continued) REGULATORY GUIDE 1.97 PARAMETERS MONITORED BY SPDS, IDADS, AND DRMS Variable Variables per SMUD SMUD Computer Number R.G. 1.97, R3 Tag No. Cat. Display Range TYPE E VARTARTER (Continued)
- 77. All Identified R-15017A 3 IDADS 3.4E-7 to Plant Release R-15017B DRMS 3.4E-1 uCi/cc Points R-15044 Kr-85 (Category 3) R-15045 0-110% design R-15546A flow FIRQ-95108
- 78. Airborne Portable 3 10 to Radioactivity Samplers & 10~' uCi/cc (Category 3) Analyzers
- 79. Plant & Environs Portable 3 10 to Radiation Analyzers 10' R/hr (Category 3) (photons and beta)
- 80. Plant & Environs Portable 3 Isotopic Radioactivity Samplers & Analysis (Category 3) Analyzers Amendment 2 7.4-5
7.4.2 INFORMATION DISPLAY AND CONTROL FUNCTION (Continued) IDADS consists of a central computer system located in the Computer Room. It has various types of input multiplexing equipment, display terminals, and hard-copy equipment in the control room and TSC. Plant inputs enter the system from the following sources:
- 1. The Bailey multiplexer,
- 2. The MODCOMP multiplexer, and
- 3. The Anatec remote multiplexer (REMUX) system.
IDADS provides the following general features for operations in the PDM:
- 1. Data Acquisition,
- 2. Man-Machine Interface,
- 3. Alarm Annunciation,
- 4. Radioactive Release Monitoring, and
- 5. On-Line Trending.
The DRMS provides on-line information concerning radiation levels of selected plant processes and dose rate information for various areas within the plant. Process measurements provide diagnostic or status information for a particular portion of the plant and monitor releases of radioactive material from the plant. DRMS consists of subsystem elements including: 1
- 1. The RM-ll central control and display system, ,
l
- 2. Gaseous effluent and area process monitors,
- 3. Liquid effluent process monitors,
- 4. Strip chart recorders, and
- 5. The RM-23 remote control and display units.
The RM-ll computer communicates with all its monitors using redundant communication line loops. Communication between each monitor and the RM-23 module is accomplished over a single dedicated line separate from the RM-11 loops. 7.4-6
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TilIS FIGURE IS BASED ON SMUD DRWG L - m --- MMN ONE-UNE DIAGRAM E-101 REVISION 30. MORE CURRENT VERSIONS OF TilIS DRAWING MAY BE Amendment 2 AVAILABLE AND SHOULD BE USED TO l MAKE ENGINEERING CllANGES. SMUD SACRAME,gTO MUMCTPAL UTILITY DISTRICT
TABLE OF CONTENTS (Continued) I Section Title Eggg 9.5.2.2.3 Radiochemical and Service Areas 9.5-9 9.5.2.2.4 Radwaste and Ftiel Storage Areas 9.5-9 9.5.2.2.5 Electricial Equipment, Switchgear, and AC/DC 9.5-10 Panel Rooms 9.5.2.2.6 Coolant and Miscellaneous Waste Tanks 9.5-11 9.5.2.2.7 Emergency Pump Rooms 9.5-11 9.5.2.2.8 Auxiliary Building Exhaust Air Filtration System 9.5-11 i i 9.5.2.2.9 Communication Room 9.5-13 9.5.2.2.10 Chilled Water System 9.5-13 : l 9.5.2.3 Nuclear Service Electrical Buildina (NSEni 9.5-13 i 9.5.2.4 natterv nulldina 9.5-14 l 9.5.2.5 Interim on-site Storane Buildina (IoSn) 9.5-14 9.5.2.6 Switchvard control Buildina 9.5-14 9.5.2.7 codes. Standards. and Tests 9.5-15 9.6 FUEL HANDLING SYSTEM 9.6-1 9.6.1 DESIGN BASES 9.6-1 9.6.1.1 general Svstem Function 9.6-1 9.6.1.2 Snent Fuel Storane Pool 9.6-1 9.6.1.3 Snent Fuel Pool Water chemistrv 9.6-2 9.6.1.4 Fuel Transfer Tube 9.6-2 9.6.1.5 Fuel Handij aa Eaulnment 9.6-3 Amendment 2 9-111
1
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1 I TABLE OF CONTENTS (Continued) l Section Title Egg.g l 9.6.2 SYSTEM DESCRIPTION AND EVALUATION 9.6-3 9.6.2.1 Handlina Snent ruel Assemblies 9.6-3 9.6.2.2 nandline. - R Loadina Snent ruel canks 9.6-3 9.6.2.3 crano une in ruel mandlina 9.6-4 . i 9.6.2.3.1 Design 9.6-4 { 9.6.2.3.2 Evaluation 9 . '6 - 6 9.6.2.4 Safetv Provisione 9.6-9 9.7 OTHER AUXILIARY SYSTEMS 9.7-1 9.7.1 FIRE PROTECTION SYSTEM 9.7-1 9.7.1.1 nemian n==en 9.7-1 - 9.8 PLANT CQHPRESSED SERVICE CAS SYSTEM 9.8-1
9.9 REFERENCES
9.9-1 i Amer, hent 1 9-iv ;
. ._ ._ _ _ _ __ _ _ - _ = _ _ .
'f 9.5.2.2.1 General (Continued)
- 3. Radwaste and Fuel Storage Areas ,
1
- 4. Electrical Equipment, Switchgear, and AC/DC Panels l
S. Coolant and Miscellaneour Waste Tanks i i l
- 6. Auxiliary Building Exhaust Air Filtration System !
I
- 7. Communication Room l Flow diagrams of the Auxiliary Building HVAC systems are shown in '
Figures 9.5-2 through 9.5-11. Design data for major components of the ) ventilation systems are given in Table 9.5-1 above. 9.5.2.2.2 Control Room and Technical Support Center The control room is served by its normal low pressure HVAC system during the PDM. This system is shown in Figure 9.5-2. The Technical Support Center (TSC) is adjacent to the control room and is ventilated by the Radiochemical and Service Area HVAC system during the PDM. See DSAR Section 9.5.2.2.3 for a description of the RVAC l system that services the Technical Support Center. During the PDM, the HVAC system for the control room and associated computer and office spaces is provided by a low pressure, dual duct HVAC system that includes a pre-filter, fan, cooling coils and electric duct heating coils, and air mixing boxes. The fan discharges a mixture of ] f resh and recirculated air into the cooling and heating coil ducts. The air streams are then supplied separately to the mixing boxes serving different temperature zones of the control room and associated comput'er l rooms and office spaces. The air mixing ratio is automatically j controlled to maintain the preset temperatures within each zone. The j system also maintains positive pressure in the rooms, forcing the excess air to leak to the adjoining areas. 4 The normal control room air-handling unit is autematically de-energized when a loss of airflow is sensed in the normal FVAC duct or when the a system isolation dampers close. To ensure control room operators will not become incapacitated following an accidental release of chlorine during the PDM, the total quantity of gaseous chlorine allowed within the Industrial Area is administratively limited to 100 pounds or less. Amendment 1 9.5-8 4
9.5.2.2.2 Control Room and Technical Support Center (Continued) Fire and smoke dampers are installed in the TSC supply and return duct-work so that the control room is isolated from the TSC in case of fire and/or smoke, thus ensuring the safe habitability of the control room in the event of a fire in the TSC. 9.5.2.2.3 Radiochemical and Service Areas The ventilation system for the radicchemical and service area, which includes the TSC, is a low-pressure, dual duct, multi-zone system that supplies a mixture of fresh and recirculated air to different temperature zones through mixing boxes. The system is shown in Figure 9.5-3. Electric duct heating coils are provides for the Radio-Chemistry Lab Count Room to protect temperature sensitive equipment. , A loss of air flow causes the TSC isolation dampers to close, but the normal TSC air handling unit continues to supply conditioned air to other service areas. Excess air supplied to this area is exhausted into the atmosphere through the plant vent as described in DSAR Section 9.5.2.2.4. The points of exhaust are located such that the air movement within the area is from points of lower to higher chemical activity and radioactivity. A separate fan supplies air to chemical fume hoods, which exhaust to the plant vent.
)
9.5.2.2.4 Radwaste and Fuel Storage Areas
, The radwaste and fuel storage areas are served by two separate ventilation units, each containing a fan, heating coil, evaporative l cooler, and a pre-filter. This system is shown in Figure 9.5-4.
l The exhaust system for the radiochemical and service areas and the i radwaste and fuel storage areas consists of two units, each sized to handle 100% of the ventilation load. Each unit contains fan and filter l sections, including a pre-filter and a HEPA filter. The cleaned and l decontaminated air is discharged to the atmosphere through the plant vent. The radwaste and fuel storage area ventilation system is a I Seismic Category II system, is normally functioning, and acts as the l back-up spent fuel pool cooling system during the PDM as described in I DSAR Section 9.2. I I Amendment 2 9.5-9
9.5.2.2.4 Radwaste and Fuel Storage Areas (Continued) The supply system for the radwaste and fuel storage areas consists of air supply units, filters, and appurtenant duct-work. The air supply to the fuel storage area is provided by an air-handling unit located on the roof of the Fuel Storage Building. To prevent discharging contaminated air directly to the atmosphere, this unit trips if the Fuel Storage Building is less than a vacuum pressure of 0.03 inches of water. P The fuel storage area air exhaust is directed to the intake plenum of the fuel storage area and radwaste exhaur.c system. A ventilation exhaust unit failure, which is detected by pressure sensors, results in an auto-start signal to the other unit. ' The exhaust fan motors can be stopped or started remotely in the control room. Also, high air flow rates are alarmed in the control room. During fuel handling operations, the doors to the Fuel Storage Building are kept closed, except to allow passage of plant personnel. If the ventilation system becomes inoperable, the primary mechanism for , dispersal of radioisotopes from the Fuel Storage Building following the drop of a fuel assembly is diffusion. This diffusion may be aided by a slight pressure differential across the building walls induced by wind movements. There is no data available to accurately quantify this leakage, but it will be small, certainly less than one building volume percent per day. 9.5.2.2.5 Electrical Equipment, Switchgear, and AC/DC Panel Rooms The electrical equipment, ac/dc panel, and switchgear rooms are served > by a single ventilation system. This ventilation system is shown in Figure 9.5-5. Since emergency back-up power is not required in the PDN, the battery room ventilation system is not required to function and is not described in the DSAR. The electrical equipment, ac/dc panel, and switchgear rooms are served by four separate air-conditioning units, each containing a fan, a pre-filter, and heating and cooling coils. Amendment 1 9.5-10
9.5.2.2.6 Coolant and Miscellaneous Waste Tanks The vapor spaces of the coolant waste receiver tanks, coolant waste holdup tanks, and miscellaneous waste tanks are ventilated. Air from the radwaste area is drawn through the tanks and discharged into the radwaste area exhaust duct. The operating radwaste area exhaust fan provides the suction to ventilate the tanks. Figure 9.5-6 shows this radwaste area ventilation system. The radwaste area exhaust plenum has two redundant, full-capacity exhaust fans. Only one operates at a time. Should the operating fan fail, pressure sensors automatically start the other fan and close the motorized damper at the discharge of the faulted component. If the filters become dirty, the increase in pressure at the filter inlet is alarmed in the control room, and the operator can start the other fan. A single failure analysis for this portion of the ventilation system is provided in Table 9.5-2. 9.5.2.2.7 Emergency Pump Rooms
- The emergency pump rooms are ventilated by part of the radwaste area ventilation system (see Figure 9.5-6). This is a Seismic Category II ventilation system. The cystem exhausts to the fuel and radwaste area exhaust system where the effluent is passed through HEPA filters and monitored prior to discharge through the plant vent. The radiation monitor alarms in the control room on high level. This ventilation system is not required to function following an accident. The failure 4 of this system would result in any released radioisotopes being confined
- to the lower level of the Auxiliary Building. l 9.5.2.2.8 Auxiliary Building Exhaust Air Filtration System
- The Auxiliary Building Exhaust Air Filtration system is Quality Class 2
) and is located on the mezzanine roof of the Auxiliary Building. I Figure 9.5-7 shows the flow diagram for this system. The system consists of an air-handling unit with pre-filter and HEPA filter designed to reduce the level of radioactivity in the exhausted air from both the ventilation equipment room and the electrical penetration room, which are located at the 20-foot level and grade level of the Auxiliary Building, respectively. After filtration, the air-handling unit discharges the air through the Auxiliary Building Grade Level Vent. Also, this system has a radiation monitor that samples and monitors the exhaust air from the ventilation equipment room and the electrical penetration room. The monitor employs an iso-kinetic probe which obtains its sample distribution ceross the grade level vent duct. 9.5-11
-.t-TABLE 9.5-2 SINGLE FAILURE ANALYSIS FOR THE RADWASTE TANKS VENTILATION AND SAMPLING SYSTEMS Component Malfunction Comments and Consequences A. VENTILATING SYSTEM
- 1. Plant Vent Fan Fails to operate Pressure switch at fan inlet starts other fan.
Ventilation is not interrupted.
- 2. Motorized Fails closed Pressure switch at fan Isolation Damper inlet starts other fan.
Ventilation is not interrupted.
- 3. Plant Vent Become dirty or Increase in pressure in Filters clogged upstream of filters is alarmed in Control Room and operator starts other fan. Ventilation rate is only briefly slowed prior to start of alternate fan.
9.5-12
9.5.2.2.9 Communication Room The Communication Room HVAC system consists of two units. The air-handling unit, which gets its make-up air from the outside environment, operates continuously. A recirculating packaged air-conditioning unit with direct expansion cooling coils operates when supplementary cooling is required. The system is shown in Figure 9.5-8. 9.5.2.2.10 Chilled Water System Chilled water for cooling the Auxiliary Building is provided by two refrigeration units, each sized to carry 50 percent of the design cooling load. Each of the units has four compressors so that the failure of any one compressor will not significantly affect system operation. The chilled water is circulated by two pumps, with one additional pump serving as a standby. The system is shown in Figure 9.5-9. 9.5.2.3 Nuclear Service Electrical Buildina (NSEB) Both the ' A' and 'B' side normal NSEB HVAC system is required during the PDM to support the electrical equipment that functions to supply normal e power to some of the Auxiliary Building equipment required to function during the PDH. The ' A' and ' B' side NSEB norm 1 HVAC system is shown in Figure 9.5-10, sheets 1 and 2. Only portions of the normal NSEB HVAC system are maintained functional during the PDM. The essential NSEB . HVAC system is not required. Each side of the NSEB normal HVAC system consists of one air-conditioning unit that serves half of the NSEB. This air-conditioning unit consists of an air-handling unit, an exhaust / return fan, and a condensing unit. The air-handling unit consists of a medium efficiency filter, a direct expansion cooling coil, and a supply fan. The condensing unit consists of two compressors, condensing coils, and condenser fans. In addition, there are two roof ventilators and a unit heater to serve the NSEB area. The normal temperature specified for equipment located in the NSEB is 50*F to 80*F. The abnormal high temperature specification is 102*F, which is based on 10 events of 8 hours duration each. These specifications will not be exceeded if the ' A' and ' B' side normal HVAC system operates. l i I i 9.5-13 l
l 9.5.2.4 Batterv nulldina I The Battery Building includes the old Central Alarm System (CAS) room, two charger rooms, two battery rooms, and two diesel generator rooms. In accordance with the approved defueled condition Security Plan, no personnel are required to occupy the old Battery Building Central Alarm System room because the Central Alarm System is in the control room during the PDM. Also, only one train of Battery Building ventilation l equipment is required to function during the PDM in accordance with the I Security Plan. Therefore, the Battery Building ventilation system is only required to provide temperature control for one train of equipment in the Battery Building during the PDM. The charger room is served by a thermostatically controlled air-conditioning unit with an air-cooled condenser on the roof and an exhaust fan. The battery room is served by a common air-conditioning ! unit and a common exhaust fan. The diesel generator room is ventilated ) by a self-contained diesel generator shaf t-mounted fan. 9.5.2.5 Interim on-site stornae nulldina (Toss) I The IOSB radioactively clean areas (control room, records room, frisk ! area, and count room) are served by a thermestatically controlled roof-mounted heat pump unit with a supply and return duct and outside air make-up. Figure 9.5-11 shows the flow path for the IOSB HVAC system. The contaminated areas (upper cell storage, dry active waste (DAW) storage, DAW handling, and truck bay) are served by two supply air-handling units with electric heaters and an exhaust unit. The units are designed to maintain a negative pressure in the contaminated areas. All three units are interlocked to trip off-line to prevent pressurization of the building if the exhaust unit fails to operate. The units are also tripped on detection of high radiation in the exhaust duct. 9.5.2.6 Switchvard control Buildina The Switchyard Control Building consists of a control room and two battery rooms. The control room is served by a split system air-conditioning unit with an air-cooled condenser located outside. The battery rooms have natural draft ventilation to prevent hydrogen accumulation. 9.5-14
9.5.2.7 coden. standards. and Tests a The work, equipment and materials conform to the following codes and standards as applicable: A. American Society of Heating, Refrigerating and Air Conditioning Engineers (ASHRAE) Handbook of Fundamentals and Guides B. Air Moving and Conditioning Association (AMCA) j C. Air Conditioning and Refrigeration Institute (ARI) D. National Fire Protection Association Pamphlet 90A i . The HVAC equipment is accessible for applicable periodic testing, maintenance, and servicing during the PDM. Where redundant equipment is provided, it is operated alternatively. The normal and other HVAC systems were designed as Quality Class 3. 3 a Amendment 1 9.5-15
- . - ~ - . - _ . . _ - . _ - ... - ,
I 9.6.1.4 Fuel Tran=fer vnha (Continued) upender pits. The fuel transfer tubes are not required to function and remain closed during the PDM. 9.6.1.5 Fuel unndli na y,anin-.nt l This equipment consists of a fuel handling bridge, fuel handling tools, spent fuel storage racks, fuel elevator, failed fuel transfer containers, control rod handling tools, viewing equipment, fuel transfer mechanisms, and shipping casks. 9.6.2 SYSTEM DESCRIPTION AND EVALUATION 5 9.6.2.1 Handlina snant ruel an===hlima The spent fuel assemblies are handled by a fuel handling bridge equipped with a fuel handling mechanism and fuel grapple. This bridge spans the fuel storage pool and permits the fuel handling crew to handle a spent fuel assembly in any one of the rack positions. 9.6.2.2 Handlina and ta=dina an.nt ruel canka Space is provided in the spent fuel pool to receive a spent fuel shipping cask and to provide for long term storage of the cask, if necessary. Spent fuel shipping casks can be handled by the easterly cantilever of the 185-ton capacity Turbine Building gantry crane. Figure 9.6-1 shows the configuration of the gantry crane and its relation to the Fuel Storage Building. The hoist trolley may be run out on the cantilever extension of the bridge a distance of 22 feet from the column / rail centerline (G-line) where it is stopped by limit switches and prevented from further travel by rail stops and bumpers. This location centers the main hook above the centerline of the north roof hatch and the centerline of travel of the cask as it is being transported in and out of the Fuel Storage Building. Once in this position, there is no further need to traverse the hoist trolley during fuel cask shipment, so the hoist traverse controls will be locked out during these operations as an administrative safety feature. All master switches are equipped with off-position latches. From a point directly above the fuel loading pit the gantry crane can only be moved in a northerly direction when the hoist is on the 9.6-3
9.6.2.2 unndlina and ta=dina sn.nt ruel canks (Continued) cantilever. This is accomplished by automatic interlocks to prevent the hoist being positioned over the spent fuel pool. The crane operator's cab is positioned on the southeast leg of the gantry. This cab is furnished with full height windows front and sides providing direct observation of all areas exterior to the Fuel Storage Building. Observation within the building from the cab is necessarily limited by the superstructure above the storage pool, but, this restriction is obviated by the administrative requirement of the crane operator to follow a precise and repeated pattern when moving the shipping cask from location to location. This permits the use of indexing markers for both hoist elevation and gantry position to supplement direct observation and control by personnel within the building. The gantry crane is equipped with precise inching control for all modes of travel. It is anticipated that no special handling fixtures will be required for use with the crane other than the two slings described in DSAR Section 9.6.2.3.2. The special tools and devices used with the shipping cask are not used with the gantry crane. When the head or cover of the shipping cask is removed by disconnecting the bail after the cask is placed in the loading pit, the head and bail may be moved to the intermediate ledge or the cask wash-down area. The gantry crane is not used for loading spent fuel and is in stand-by status until the cask is ready to be moved. A specially designed cask washdown structure, located in the Fuel Storage Building, is used to facilitate decontamination of the outside surfaces of a cask af ter the cask has been removed from the spent fuel pool. 9.6.2.3 crane use in ruel nandiina 9.6.2.3.1 Design The structural design of the spent fuel shipping cask crane (another name for the Turbine Building gantry crane) conforms to the applicable requirements of the following specifications:
- 1. American Institute of Steel Construction specification for the Design, Fabrication and Erection of Structural Steel.
Amendment 2 9.6-4
9.6.2.3.1 Design (Continued)
- 2. Specifications for Overhead Traveling Cranes published by the Electric Overhead Crane Institute.
- 3. The American Welding Society Specifications for Welding of Highway and Railroad Bridges.
- 4. All of the requirements of State of California, Department of Industrial Relations, Sub-Chapter 7, General Industry Safety Orders, Group 7, Cranes and Other Hoisting Equipment.
- 5. USAS B30.20 Overhead and Gantry Cranes.
The design of structural members other than the bridge girders is in accordance with the AISC Specifications except that the allowable unit stresses were reduced by ten percent. The design of the bridge girders is in accordance with the Electric Overhead Crane Institute Specifications except that stresses were proportionally increased to conform to a stress level equal to 90 percent of the AISC allowable basic stress for the material used. All work performed was in accordance with the most advanced practice for this class of equipment. All materials furnished and work performed were in accordance with the requirements of the latest ANS, IEEE, ASTM, ASME, HEI, and NEMA Specifications. In addition, the completed equipment complies with all applicable requirements of Federal and National board codes, California Codes, and the State of California, Division of Industrial Safety, General Industry Safety Orders. The gantry crane was designed to maintain complete stability under the loaded and unloaded condition. The truck and trolley frame for the I gantry crane was designed to resist vertical, lateral and torsional strains. All hooks were annealed and each hook was shop tested at 150 percent of j its rated load. All hooks were completely radiographed in accordance ) with ASTM E-94 and ultrasonically tested for internal defects prior to I l'oad testing. l The hoist is provided with three means of braking, two mechanical and one electrical. There is one mechanical brake located on the motor j shaf ts and one on the high speed shaf t of the hoist gear boxes. These brakes are capable of overcoming at least 150 percent of the full load torque exerted by the motors. The electrical braking system has a slow l lowering feature in the event of a power failure and the simultaneous : failure of both mechanical brakes. This feature causes the hoist motor i i l 9.6-5
9.6.2.3.1 Design (Continued) to be automatically placed in a generator mode, and the hoist motor absorbs the power produced in a resistor bank. The gantry crane is equal to the Electric Overhead Institute Class A for standby service. The rated capacity of the gantry crane is 185 tons. The rated capacity of the main hoist is 185 tons. The capacity of the auxiliary hoist is
- 35 tons.
1 9.6.2.3.2 Evaluation The gantry crane was operated with a field test load of 125 percent of its rated capacity. Demonstration that the gantry crane can raise, lower, hold in any position, and transport its load without excessive deflections in the crane parts was demonstrated during the field test program. The gantry crane was detailed and fabricated by a reputable and experienced crane manufacturer under rigid engineering control. Features of the design were developed to minimize the impact of historical failure modes which have been experienced during the development of crane technology. Perhaps the most important consideration in appraising safety and reliability of the gantry crane system is the fact that operation of the gantry crane is expected to be well below the gantry crane's rated system capacities. Design features which are intended to help avoid foreseeable crane accidents include: A. The use of field loss relays to shut down the gantry crane system in the event of motor failure for bridge movement, trolley movement, or hoist movement, B. Redundant fail-safe mechanical and electrical brake systems rated at 150 percent of full motor torque, C. Dead-man switches, D. Wheel stops and bumpers designed for stopping gantry crane travel at rated speed with the power off, E. limit switches which reset automatically, F. Provisions that an axle break shall not drop the bridges more than one inch, 9.6-6
. - _ . .... ...- . . . - - . - - _ - -- -- ._ . ~ . _ - - .. - - . - . - .-..
l i 9.6.2.3.2 Evaluation (Continued) I G. The' righting moment shall exceed overturning moment by at least a ratio of 1.5/1.0 under the most extreme loading conditions, ' H. The worst stress on any mechanical part shall not exceed 90 percent of yield stress under breakdown conditions or a locked rotor torque, and I. Collector shoes to clear tracks of obstructions. All gantry crane mechanical parts are designed, as a minimum, with a safety factor of 2.5 on yield strength.- ! I Plant configuration and administrative controls preclude a cask drop from exceeding the 30 foot maximum drop (through air) allowed by 10 CFR I 71, Appendix B. 1 Figure 9.6-2 shows a schematic cross section of the Fuel Storage , Building and the yard area adjacent to the railroad siding. Possible l drop areas and maximum possible drop heights are listed in Table 9.6-1. I I In establishing these maximum foreseeable drops, consideration was given i to the constraints imposed by the equipment that will be used for l transferring spent fuel casks from the cask loading pit to a ' railroad flatcar. The concept is based on the General Electric IF-400 cask, conservatively assumed to measure 20 feet from the bottom of the cask to the attachment point on the lifting bail. The maximum vertical travel of the gantry crane hook is to the 80 foot elevation. This is a limiting value, since the hook cannot be raised above this elevation. Two different length slings must be used between the bail and the crane hook. The shorter length sling is required to clear structures at the l 40 foot elevation. The longer sling is required to avoid immersing the crane hook in the spent fuel pool water. The long sling is used from the loading pit to the intermediate ledge in I the spent fuel pool. The short sling is used from the intermediate ledge to the flatcar. Positioning the cask over the railway spur without first lowering the hook to elevation 68 feet, 6 inches (outside the Spent Fuel Bcilding) is precluded by a hook-height / crane position interlock. Therefore, the cask cannot be put in a configuration at any time' during the on-site transport process where it could be dropped more than 26 feet in air (or 33 feet, 6 inches in water) . It was established l in DSAR Section 5.4 that the terminal velocity for a 40-foot drop through water is less than that for a 30-foot drop through air. I 9.6-7
TABLE 9.6-1
]
I FUEL CASK DROP HEIGHTS Location unwinn= Drop , l Above the loading pit 33 ft.-6 in.* l 1 Above the intermediate ledge 26 ft.-6 in.* Above the cask washdown structure 4 inches l l Above the main steam lines immediately outside the 1 ft.-6 in. building i Above the remaining yard area to the railroad siding 16 ft.-6 in, i Above the flatcar on the railroad siditg. 26 ft.-0 in. 1
- Through water ,
I 1 l Amendment 2 ) 9.6-8 i
_ . _ _ . _ _ _ . _ _ _ . _ _ _ _ _ _ . _ _ . . - . _ _ . . .. <__._.__-mm._ ___m _ . . 9.6.2.4 safetv proviniana (Continued) 1 NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants", July 1980, j provided seven guidelines and interin measGres which are to be followed- , by all overhead handling systems and programs used to handle heavy loads near spent fuel stored in the spent fuel pool. These guidelines are identified in NUREG-0612 as follows. , i Guideline 1 - Safe Load Paths Guideline 2 - Load Handling Procedures i Guideline 3 - Crane Operator Training Guideline 4 - Special Lif ting Devices I Guideline 5 - Lifting Devices (Not Specially Designed) i Guideline 6 - Cranes (Inspection, Testing, and Maintenance) Guideline 7 - Crane Design An analysis of the overhead handling systems at Rancho Seco was performed by SMUD. The NRC and its consultant, the Franklin Research Center, reviewed the SMUD analysis and concluded that the above guidelines had been satisfied. The NRC has established. six interim protection measures that were intended to provide reasonable assurance that no heavy loads will be handled over spent fuel and that measures exist to reduce the potential for accidental load drops that could impact spent fuel. Four of the six interim measures of the report are captured by general Guideline 1, Safe Load Paths and general Guideline 6, Cranes (Inspection, Testing, and Maintenance). The two remaining interim measures cover the following criteria: l l Interim Measure 1: Heavy load technical specifications, and l I Interim Measure 2: Special review for ' heavy' loads handled over j spent fuel. ' i Rancho Seco complied with Interin Measure 2 based on the District conducting a special review of loads handled over spent fuel. ' 1 Permanently Defueled Technical Specification (PDTS) D3/4.3, " Fuel Storage ! Building Load Handling Limits," satisfies the requirements of Interim Measure 1 (heavy load technical specifications). A ' heavy' load is Amendment 2 9.6-11
~. - - -. .- _ - - - - - - .
l 9.6.2.4 Safety Provisinna (Continued) defined as any load in excess of the design basis load. In accordance with PDTS D3/4.3, no load in excess of the combined weight of a fuel assembly, its control component, and associated handling tool (i.e., the design basis load) is allowed to be handled over spent fuel assemblies stored in the spent fuel pool. One exception to this ' heavy' load limit I is described below. i
- PDTS D3/4.3 specifies one exception to this heavy load ILuit. During spent fuel assembly off-load activities to the ISFSI, the dry shielded canister (DSC) top shield plug, the cask lifting yoke and yoke extension, and the Gantry Crane lower load block may be handled with the Gantry Crane over irradiated fuel assemblies that have been placed in a DSC in the' spent fuel pool. This exception is based on the specified lifting .
t components being designed and tested in accordance with ANSI N14.6-1986. Also, the Gantry Crane is designed such that it can only handle loads over
- the cask loading pit area of the spent fuel pool and can not move a load over the spent fuel pool storage racks.
i 4 l l l I l I Amendment 2 ; l 9.6-12 l l
9.7 OTHPR AUYTLTARY SYSTEMM 9.7.1 FIRE PROTECTION SYSTEM 9.7.1.1 nemign n== m The Rancho Seco Decommissioning Fire Protection Plan (RSDFPP) provides a defense-in-depth approach to fire protection that is designed to: A. Minimize the probability of fires and explosions and the potential effects of such events on structures, systems, and components important to safety during the PDM, B. Promptly detect, appropriately respond to, and eventually extinguish a fire, if one should occur, and C. Minimize the potential consequences of a fire. Active and passive design features are provided to detect, contain, and suppress fires. The Rancho Seco fire protection program includes the following fire protection features: A. Fire suppression systems, B. Fire detection and alarm systems, and C. Fire barriers and fire breaks. The RSDFPP provides a detailed description of the organizational requirements, design features, operational requirements, compensatory measures, testing requirements, fire brigade staffing requirements, and off-site fire fighting assistance coordination requirements that collectively define Rancho Seco's fire protection program. During the PDM, fire loading, ignition sources, and combustible materials are significantly reduced and the possibility for a major fire is greatly diminished. By reference in the DSAR, the RSDFPP is a licensing basis document. The District maintains a fire protection program for Rancho Seco that addresses the potential for fires which could result in a nuclear hazard. The objectives of the fire protection program are to: (1) Reasonably prevent such fires from occurring; (2) rapidly detect, control, and extinguish those fires which do occur; and (3) Ensure that the potential hazard due to fire to the public, environment, and plant Amendment 2 9.7-1
d 9.7.1.1 nanian a==am (Continued) personnel is small. The District assesses the Rancho Seco fire protection program on a regular basis and revises the program, as appropriate. The District makes changes to the Rancho Seco fire protection program without NRC approval provided that the changes to not reduce the effectiveness of fire protection measures needed to prevent a nuclear hazard, taking into account the decommissioning plant conditions and activities. l t
)
Jueenh;st 2 9.7-2
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9.8 PLANT COMPREEMED SERVICE CER SYSTEMR l The plant compressed air system provides service and/or instrument air to the following areas during the PDM: A. Turbine Building and Yard, B. Auxiliary Building, C. Reactor Building and Yard, and D. Nuclear Service Electrical Building. Oil-free air at 100 psig is provided in quantities up to 900 scfm by two single stage oil-free air compressors. This system supplies air for ! both instrument and service air usage. Approximately 100 scfm required ; for instrument use is passed through a dual tower desiccant-type air I l dryer and air filter before it is piped to the instrument required during the PDM throughout the plant. This entire system was designed and constructed to Seismic Category II requirements. Table 9.8-1 lists those storage vessels which contain compressed service gases that continue to be used in the PDM. Also, Table 9.8-1 contains the applicable vessels' design pressure, operating pressure, potential energy, number, and location information. The location of service gas vessels were selected to avoid heavily travelled areas. Protective guards are provided to minimize the l possibility of vessel damage from external sources. Further, since I these vessels and their hold-downs are designed, manufactured, tested, inspected, and maintained in accordance to applicable regulatory standards in the PDN, the likelihood of vessel failure from inherent defect or misuse is unlikely. The vessels, except the compressed air cylinders, are filled on site. Operating procedures protect the on-site vessels from over-pressurization during filling. In addition, each vessel is provided with relief valves which prevent exceeding design pressure during filling and normal operations. The compressed air cylinders have rupture disks rather than relief valves. 1 Amendment 1 9.8-1
0 TABLE 9.8-1 COMPRESSED SERVICE GAS VESSELS i Pressure - Psig Potential Energy Number vessel ----------------- Per Vessel: of Location h Content Demian onoratina Foot-poundm* Vessels of vesselm 1 Air-gas 125 110 1.9 x 10' 3 Grade level in Turbine Bldg.; 20 ft. west'of column row G; 6 ft. south of column row 2. 2 CO,-liq. 363 300 260 x 10' 1 Grade level; north of RB; 6 ft. west of RB center-line; 1 ft. south of column 1.4. f Conservatively assumed to be the total energy of compression contained in the vessel. Amendment 2 9.8-2 j
9.9 REFERENCES
i l '
- 1. License Amendment No. 119, dated March 19, 1992, Permanently l Defueled Technical Specifications
- 2. Safety Analysis and No Significant Hazards Consideration (Log No. !
1091, Revision 3) for Proposed Amendment 182, Revision 3, l Permanently Defueled Technical Specifications {
- 3. License Amendment No. 117, dated March 17, 1992, Possession-Only License l
- 4. D. Brock (SMUD) to S. Weiss (NRC) letter DAGM NUC 91-183, dated November 19, 1991, Proposed Amendment No. 182, Revision 3 - Permanently Defueled Technical Specifications
- 5. Letter, J. F. Stolz (NRC) to R. J. Rodriguez (SMUD), Control of Heavy Loads, Phase 1, October 25, 1983, transmitting Technical Evaluation Report, " Control of Heavy Loads Rancho Seco Nuclear Generating Station", Franklin Research Center, September 21, 1983.
- 6. Letter, J. J. Mattimoe (SMUD) to Director of Reactor Licensing (NRC), Proposed Amendment No. 84, transmitting " Licensing Report for High Density Spent Fuel Storage Racks for Rancho Seco Nuclear ;
Generating Station", September 28, 1982. I l
- 7. Letter, J. F. Stolz (NRC) ta J. J. Mattimoe (SMUD),
Subject:
I Amendment No. 52 to Rancho Seco Technical Specifications, January 20, 1984. l
- 8. 10 CFR 50.59 evaluation for procedure B.10, Revision 9, "Defueled Condition"
- 9. SMUD Calculation Z-SFC-M2555, " Peak Spent Fuel Pool Temperature Versus Calendar Time."
- 10. SMUD Calculation Z-SFC-M2557, " Spent Fuel Decay Heat Based on ORIGEN2 Computer Code."
- 11. SMUD Calculation Z-SFC-M2560, " Spent Fuel Pool Heat-Up During LOOP with Pool at 23.25 feet."
- 12. NRC Order Approving the Rancho Seco Decommissioning Plan, dated i March 20, 1996.
l l j Amendment 2 9.9-1 1 1
TABLE OF CONTENTS Section Title Page
- 11. RADIOACTIVE WASTE Al".s RADIATION PROTECTION 11.1-1 11.1 SOURCE TERM 11.1-1
. 11.1.1 RADIONUCLIDE INVENTORY 11.1-1 11.1.1.1 Snent Fuel hamm=hlien 11.1-1 11.1.1.2 Reactor Vemmel and Internals and Concrete 11.1-2 P ri ma ry Shield 11.1.1.3 Plant svat=== 11.1-2 11.2 LIOUID WASTE TREATMENT SYSTEMS 11.2-1 11.2.1 COOLANT RADWASTE AND REACTOR COOLANT DRAIN 11.2-1 SYSTEMS 11.2.1.1 Functions 11.2-1 11.2.1.2 Svstem Descrintion 11.2-2 11.2.2 NISCELLANEOUS LIQUID RADWASTE SYSTEM 11.2-2 11.2.2.1 Function 11.2-2 11.2.2.2 avstem noneription 11.2-2 11.2.3 WASTE WATER DISPOSAL 11.2-12 11.2.3.1 Plant Effinant 11.2-12 11.2.3.2 Normal Radioactive Discharge 11.2-13 11.2.3.3 Off-Normal andicactive Discharae 11.2-14 11.2.3.4 Enn-. radioactive mante water 11.2-15 11.2.4 OPERATION, TESTING, AND INS *ECTION 11.2-15 11.2.5 SYSTEM EVALUATION 11.2-16 11.2.6 PROCESSING WET RADIOACTIVE WASTES INTO 11.2-17 SOLID RADIOACTIVE WASTE 11.2.6.1 Solidification and Dewaterina of wet 11.2-17 Radioactive manten Amendment 2 11-1
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l l l TABLE OF CONTENTS (Continued) Section Title Eggg 11.2.6.2 Drvina of wet nadioactive vastes 11.2-19 l 11.2'.6.2.1 Design Basis 11.2-19 l 11.3 GASEOUS BASTE MANAGEMENT SYSTEM 11.3-1 11.3.1 DESIGN BASIS 11.3-1 11.3.2 SYSTEM DESCRIPTION 11.3-1 11.3.3 HYDROGEN GAS MIXTURES 11.3-2 11.3.4 OPERATION, TESTING, AND INSPECTION 11.3-2 11.3.5 RADIOACTIVE RELEASES 11.3-3 11.3.5.1 Pathways 11.3-3 11.3.5.2 secondarv Plant Contamination 11.3-3 11.3.5.3 Interin on-mite storaae nuildina (Tosni 11.3-3 11.3.6 METHOD OF ASSESSMENT 11.3-4 11.3.6.1 Pin =a Runosure (Noble Cases) 11.3-4 11.3.6.2 Food Pathway 11.3-5 11.3.*1 EVALUATION OF WASTE DISCRARGE 11.3-5 11.4 SOLID WASTE MAMAGEMENT SYSTEM 11.4-1 11.4.1 DESIGN BASIS 11.4-1 , l 11.4.1.1 Renorts 11.4-2 l l 11.4.2 SYSTEM DESCRIPTION 11.4-2 11.4.2.1 Dry solid unste Dinnosal system / Process 11.4-2 11.4.2.2 concentrated Liania vaste Disnosal system / 11.4-3 Process . 11.4.2.3 Snent Resin Discosal System / Process 11.4-3 L f 11.4.2.4 Filter Dinnosal Process 11.4-3 ! i Amendment 1 11-11
TABLE OF CONTENTS (Continued) Section Title g 11.10.1 ORGANIZATION 11.10-1 11.10.2 EQUIPMENT, INSTRUMENTATION, AND FACILITIES 11.10-3 11.10.2.1 Personnel Protective Emt innan n t 11.10-3 11.10.2.2 Radiation Protection Instrumentation 11.10-3 11.10.2.3 racilitima 11.10-4 11.10.3 RADIATION PROTECTION PROCEDURES 11.10-5 11.10.3.1 Procedures 11.10-5 11.10.3.2 Radtation 1Eork Permit Procedure 11.10-6 11.10.4 PERIODIC PERSONNEL EXPOSURE REPORTING 11.10-6 11.11 REFERENCES 11.11-1 Amendment 1 11-v
, LIST OF TABLES Table Tit 1e Enga b 11.2-1 Liquid Wastes Disposal System Component Data 11.2-3 - 11.2-2 Integrated , ses at the Site Boundary Resulting 11.?-1R from various Tr.nk Ruptures , t 11.5-1 Suunnary of fadiation Background for Sacramento, 11.5-3 r California 11.5-2 Results of the Rancho Seco Pre-operational 11.5-5 Surveillance Program 11.8-1 Principal Shielding 11,8-4 11.8-2 Area Radiation Monitors 11.8-8
, 11.8-3 Process Radiation Monitors 11.8-10 L
i i. 1 4 l l
\
Amendment 2 l I 11-vi
DSAR CHAPTER 11. RADIOACTIVE WASTE AND RADIATION PROTECTION-11.1 S.OURCE TERM The only. Rancho Seco design basis accident considered credible during the PDM that results in off-site dose consequences is the Fuel Handling Accident (FHA). This accident is postulated to result in the release to the environment the gap activity from all 208 fuel rods of the hottest l spent fuel assembly. A District calculation' determined the source term for the FRA, decay corrected to five years after final reactor shutdown. The 1 only gap gas radionuclide that is (1) still present in a measurable quantity and (2) an off-site exposure concern is Krypton 85 (Kr-85). The activity projected to be released to the environment from a FRA during the Permanently Defueled Mode (PDH) is 5.41E+03 Curies. This design basis accident analysis source term is conservative since i Rancho Seco permanently shut down reactor operations on June 7,1989. Dose contributions following a FRA from other radionuclides, such as I-131, Xe-131m, and Xe-133, are insignificant during the PDM. Kr-85 is the predominant nuclide for a gap gaseous release and dominates the FHA accident analysis source term. The maximum dose to a member of the public at the Industrial Area boundary following a FRA is 9.9 mrem to the total body. The maximum total skin dose from a FRA is 1,116 area. 11.1.1 RADIONUCLIDE INVENTORY 1 l The largest fraction of the on-site radionuclide inventory is contained in the spent fuel, with the reactor vessel and internals containing the next largest fraction. Radionuclides are also present in corrosion films within , various plant systems. Based on the permanently shutdown and non-operating I plant status, most radionuclide sources are not readily dispersible during the PDN. j 11.1.1.1 Enent Fuel Asammhlies ] I The Rancho Seco spent fuel pool cor tains 493 spent fuel assemblies, j Uranium burn-ups range from 8,214 to 39,073 megawatt-days per metric ton l 1 of Uranium. Each fuel assembly originally contained approximately 0.464 metric tons of Uranium. The total activity inventory after shutdown was approximately 1.41E+08 Curies, with Ce-144, Pr-144, Cs-137, Ba-137m, Sr-90, Y-90, Pm-147, and Pu-241 comprising nearly 70% of the activity. By the year 2010, the total activity will decay to approximately 3.96E+07 Curies, ! with Cs-137, Ba-137m, Sr-90, Y-90, and Pu-241 comprising over 97t of the l activity. Amendment 2 11.1-1
11.1.1.2 Reactor Vennel and Internals and Concrete Primary Shield The District performed an activation analysis of the reactor vessel and internals and the concrete primary shield using the ORIGEN2 computer code. The resulting total activation products inventory is approximately 8.94E+05 Curies, with Fe-55 and co-60 comprising over 82% of the activity. 11.1.1.3 Plant S vn t-e The District characterized the internal contamination of plant systems from operational s ample analysis results, post-operating characterization efforts, and radiation monitoring survey results. The District also incorporated corrosion film radionuclide inventory estimates. The total plant systems activity inventory was estimated to be 4,490 Curies, with Fe-55, co-58, Ni-63, and Co-60 comprising over 88% of the activity. The District will update this data during the planning phase for the final decontamination and dismantlement decommissioning work. Amendment 2 11.1-2
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l
, i 11.2.3.1 Plant Effluent (Continued) automatically stopped, but dilution water entering the plant effluent stream downstream of these valves continues.
11.2.3.2 Mormal Radioactive Discharge One source of slightly contaminated liquid that is discharged from the alte originates from the draining of secondary plant systems not required to function in the PDN. Also, liquid collection systems collect liquids in sumps that are directed to the ' A' or 'B' Regenerant Hold-Up Tank (RHUT) or the Spent Regenerant Tank. In addition, the RHUTs can receive water from secondary plant system sources not located within the turbine building, including: 1
- 1. Transfers of water from the Demineralized Reactor Coolant Storage Tank (DRCST),
- 2. Some Tank Farm sumps, and
- 3. Auxiliary boiler blow-down and area drains.
- The capacity of the 'A' and 'B' RHUT is 100,000 gallons and 200,000 l gallons, respectively. Normally, one RHUT is lined up to receive plant I wastes while the other is out-of-service during processing, sampling, analysis, and transfer to a Retention Basin. The tanks have plastic l liners for chemical resistance. Each has an agitator and a sample pump to ensure representative sampling.
Prior to transfer of an RHUT's contents to a Retention Basin, the RHUT is isolated, recirculated, sampled, and analy.ted for radioactivity. Dose projections are made in accordance with the Off-site Dose Calculation Manual (ODCM) to determine compliance with the 10 CFR 50, Appendix I dose guidelines. Actual off-sitt releases are made from the Retention Basins. But, because the RHUTs are a more concentrated waste stream than the Retention Basins, dose accounting is done at the RHUTs to give greater assurance that radioactivity is detected in waste water, if present. If the measured radioactivity content in an RRUT is high relative to the 10 CFR 50, Appendix I dose guidelines, the RHUT may be processed with the Liquid Effluent Radwaste Treatment System (LERTS) to reduce th.e radioactivity content in the RHUT. The LERTS consists of a skid-mounted demineralizer system. The system contains a mechanical filter, a charcoal pre-filter, and four ion exchange vessels. The process flow rate is approximately 50 gallons per minute. The LERTS configuration 11.2-13
11.2.3.2 varumi nadiometive nischarae (Continued) allows adaptation to portable domineralizing equipment to assist in j processing a RHUT batch, if necessary. ! l When analysis for chemical and radiological quality of an REUT's f' contents is complete, the REUT water is directed to a Retention Basin via the RHUT transfer pumps. The REUTs may also be gravity drained. An effluent strainer is installed to prevent domineralizer/ ion exchange : resins, which can reach the RHUTs from the LERTS, from being pumped to the Retention Basins and released off-site. The strainer consists of ; six parallel mechanical filters. If a strainer saturates, an automatic backwash is initiated. Water is pumped through the filter in the i reverse direction and the trapped resin material is removed on fine mesh ! socks. The filters can remove 95 percent of the particles greater than [ 140 microns in size. The Retention Basins are also sampled for radioactivity. Radiological analysis of a Basin is used to determine the required release rate and/or dilution water flow rate necessary to ensure compliance with the i liquid effluent 10 CFR 20 radioactivity concentration limits. In addition, the Basin effluent is monitored by a process radiation j monitor, which- is interlocked to terminate a release if preset ! concentration limits are exceeded. The monitor setpoints are calculated [ in accordance with the ODCM- Radioactive releases are not made from the ? Retention Basins if dilution water is not available. [ t 11.2.3.3 off-norami nadioactive nischara. Radioactively contaminated water is contained within plant equipment located in the Tank Farm. Occasional, small-volume leakage from the tanks, pipes, valves, instrument lines, and other components is possible. -Because there are stora drains in the area that are routed directly to the plant effluent stream, trace amounts of radioactivity may-occasionally be released via this pathway, f Off-normal releases are accounted for on a case-by-case basis. This includes tracking the activity and calculating the dose impact. , Releases are not evaluated individually if it can be demonstrated that the doses are less than 1% of the 10 CFR 50, Appendix I annual whole l body dose guidelines. Dose calculations are performed in accordance with the ODCM. The source term is based on grab sample analysis of the most reprecentative process stream. Amendment 2 11.2-14
11.5.4.2.2 REMP Sample Types (Continued) pathways to man, the annual Land Use Census, and regulatory requirements and guidance. Details of the post-operational REMP sampling types are described in the REMP Manual and implementing procedures. I 11.5.4.2.3 REMP Sample Statistical Analysis j Data is assessed statistically from measurements of counter backgrounds, instrument response, and radioactive decay principles. The quality control program provides the data for the analysis of analytical errors. , Statistical uncertainty is determined using standard statistical j methods. The minimum detectable limit varies for each sample and sample l type and is calculated in accordance with the Lower Limit of Detection l equation specified in the REMP Manual. The values determined are j reported for each sample. I I Quarterly, the results of the surveillance program are reviewed with respect to historical values and their mean concentrations. Standard statistical tests are performed to identify outlyers as well as changes in observed values. i 11.5.4.3 Effluent and waste nianomal Environmental Renorts Periodic reports are submitted to the NRC to meet reporting requirements specified in the Rancho Seco Permanently Defueled Technical Specifications (PDTS). The reports contain information for the preceding calendar year in the f ollowing areas:
- 1. Environmental protection programs that monitor non-radiological and radiological effects upon the environment.
Also, a report submitted to the NRC that contains radioactive effluent information and shows compliance with discharge limits.
- 2. Analysis results of air, water, soil, vegetation, milk and animal life samples taken each year in accordance with REMP.
REMP reports are submitted to the California Department of Health Services on a regular basis. Amendment 2 11.5-11
l 1 4 11.6 ENSURING TRAT OCCUPATIONAL RADTATION RYPORUREM ARE AS LOW AS I IS REASONABLY ACHIEVABLF f ATARA) l 11.6.1 ALARA POLICY CONSIDERATIONS In accordance with 10 CFR 20.1101, SMUD will make every reasonable ! effort to maintain individual and collective occupational radiation exposures at Rancho Seco "As Low As Reasonably Achievable" (ALARA). l This operating philosophy also applies to radiation exposures to the l general population resulting from the conduct of activities at Rancho l Seco during the PDM. i j The ALARA Policy applies to all SMUD and contract personnel who require ! access to the Radiologically Controlled Area of the plant; who are i involved in the operation, maintenance, or modification of systems containing radioactive material during the PDH; or who are responsible for monitoring plant effluent. Rancho Seco shall be operated, maintained, modified, and decommissioned with the following considerations: A. Ensure radiation doses to employees, contractors, and the general public (including visitors) are maintained at a minimum, while promoting efficient conduct of operations during the PDM. B. Ensure a radiologically safe working environment for employees, contractors, and the general public (including visitors). C. Ensure no significant environmental impact results from activities performed at Rancho Seco during the PDM. The Rancho Seco Radiation Protection program provides reasonable assurance that extetnal dose to personnel from ionizing radiation will be maintained within administrative as well as NRC regulatory limits. Plant controls (i.e., procedures, policy, and planning) provide reasonable assurance internal uptakes will be maintained less than that allowed by NRC regulations. Engineering controls, containment, and respiratory protection devices shall be used to the extent possible to maintain internal uptakes ALARA. Amendment 1 11.6-1
11.8.4.2 svatem nenerintion (Continued) , The radiation monitor subsystems, together with their control and ' display equipment, annunciator, computer system interfaces, and ; auxiliary support equipment, comprises the complete Radiation Monitoring } System. l 11.8.4.2.1 Area Radiation Monitors The area radiation monitors warn personnel of excessive gamma radiation at selected areas throughout the Industrial Area. The Fuel Storage Building Area Radiation Monitors detect radiation conditions in the area
- of spent fuel storage. These monitors can detect an adverse ;
radiological condition in the spent fuel storage area (i.e., an
~
excessive loss of spent fuel pool water level or a fuel handling accident) and be used to project the off-site dose to the public resulting from the design basis fuel handling accident. The Perimeter Area Radiation Monitors detect abnormal releases in the unlikely event of a fuel handling accident and are used as an indicator to assist in dete rmining if site evacuation is necessary and through what site exit i it should take place. Table 11.8-2 describes the various area radiation monitors. Monitored points within the plant are in areas where personnel exposure to radiation is most likely. Monitor alarm setpoints are based on the ; normal background radiation at the detector location and the calculated : levels for abnormal conditions. 11.8.4.2.2 Process / Effluent Radiation Monitors The process radiation monitors in the PDM are mostly effluent monitors which monitor the plants gaseous and liquid release pathways. The operability and surveillance requirements and alarm / trip set-point calculations for the effluent monitors are described in the Of fside Dose Calculation Manual (ODCM). Table 11.8-3 describes the various process radiation monitors. t The gaseous effluent radiation monitors are R-15044, R-15045, and R-15546A. R-15044 monitors the Reactor Building Stack for noble gases (Kr-85) and provides isokinetic grab sampling capability for particulate matter. Interlocks are provided to automatically terminate releases,by tripping the Reactor Building Exhaust and Supply Fans and closing the Equalization Block Valve upon a high radiation or monitor failure signal. R-15045 and R-15546A monitor the Auxiliary Building Stack and Auxiliary Building Grade Level Vent release paths, respectively. These monitors are part of the digital radiation monitoring system, 11.8-7
_- -...- - _ - - . - . -.-. -- - ._ .. - - - . - ~ - TABLE 11.8-2 AREA RADIATION HONITORS Arsa Detector / Readout Sensitivity Alarm / Control / Radiation Location Equipment Computer Input M:nitor Parsonnel Ionization Radiation 0.1 aren/hr Visual and audible eccess Chamber / monitors level, 0.1- high rad alarm at hatch area access hatch 10' ares /hr detector; R-15025 area inside in access indication also Reactor hatch and outside Reactor Building at ARMS Building and cabinet in the control room
'r Spent fuel Ionization Radiation 0.1 mren/hr Visual and audible area chamber / monitors . level, 0.1- high radiation R-15028 spent fuel pool 10' aren/hr alarm at detector area at detector and in the control '
and ARMS room. Computer cabinet input to IDADS. IOSB Low-range GM Radit. tion 0.1 area /hr Visual and audible Extended- detector and level, 0.1- high radiation .rcnge area high-range ion 10' arem/hr alarm at detector manitors chamber detector / at detector and IOSB and main R-15110 sump area, and ARMS control rooms , R-15111 east cell area, cabinet R-15112 loading dock, R-15113 west cell area, R-15114 DAW storage area, R-15115 DAW handling area Control room Ionization FWdiation 0.1 mrem /hr Visual and audible crea chamber / monitors level, 0.1- high radiation R-15030 control room It' aren/hr alarm at detector at detector -and in the control and ARMS room cabinet Amendment 2 11.8-8 1
Page 2 of 2 TABLE 11.8-2 (Continued) AREA RADIATION MONITORS Arsa Detector / Readout Sensitivity Alarm / Control / R-diation Location Equipment Computer Input M:nitor Radiochem Ionization Radiation 0.1 aren/hr Visual and audible lab area chamber / monitors level, 0.1- high radiation R-15031 radiochemical 10' ares /hr alarm at detector laboratory at detector and in the control and ARMS room cabinet Drum decon Ionization Radiation 0.1 arem/hr Visual and audible end loading chamber / monitors level, 0.1- high radiation crea drum decon and 10' mres/hr alarm at detector R-15033 loading area at detector and in the control and ARMS room cabinet Redwaste Ionization Radiation 0.1 mrem /hr Visual and audible sump pump chamber / monitors level, 0.1- high radiation r ma radwaste sump 10' aren/hr alarm at detector is+15034 pump area at detector and in the control and ARMS room cabinet Source Ionization Radiation 0.1 arem/hr Visual and audible calibration chamber / monitors level, 0.1- high radiation room source cal- 10' area /hr alarm at detector R-15039 ibration room at detector and in the control and ARMS room cabinet Perimeter Ionization Radiation 0.1 mrem /hr Visual and audible conitors chamber / monitors level, 0.1- high radiation R-15040 the north, south, 10' aren/hr alarm in the R-15041 east, and west at the ARMS control room. R-15042 perimeters of cabinet Computer input to R-15043 the plant IDADS. Amendment 2 11.8-9
Page 1 of 2 TABLE 11.8-3 PROCESS RADIATION MONITORS Process Sampler-Detector Readent Sensitivity Alarm / Control / M:nitor Equipment Computer Input Racctor Fixed removable CRT display Range: Alarms in control guilding isokinetic sampler. in control 3.4E-7 thru room. Visual alarm Stack One low range room with 3.4E-1 in TSC. Auto secure R-15044 channel of noble digital pCi/cc purge vent system. gas detection. Low display and Kr-85 Computer inputs to range detector type recorder in IDADS and RM-11. A~ scintillation. the TSC. Auxiliary Fixed removable CRT display Range: Alarms in control Building isokinetic sampler. in control 3.4E-7 thru room. Visual alarm Stock One low range room with 3.4E-1 in TSC. Computer R-15045 channel of noble digital pCi/cc inputs to IDADS and gas detection. Low display and Kr-85 RM-11. range detector type recorder in 8- scintillation. the TSC. Auxiliary Fixed removable CRT display Range: Alarms in control Building isokinetic sampler. in control 3.4E-7 thru room. Visual alarm Grede One low range room with 3.4E-1 in TSC. Computer L2 vel channel of noble digital PC1/cc inputs to IDADS and Vsnt gas detection. Low display and Kr-85 RM-11. R-15546A range detector type recorder in S- scintillation. the TSC. l Component Off-line liquid Log count SE-7 pCi/cc Alarm on high cooling sampler with rate meter Cs-137 radiation signal or water scintillation 1El to IE6 channel failure. total detector. Samples counts per flow CCW system flow minute R-15008 continuously Amendment 1 11.8-10 l
. . - - . . -- . - - - -~-_~ . - . . _ _ - . - . . . - . . _ _ . . - . .
Page 2 of 2 TABLE 11.8-3 (Continued) PROCESS RADIATION MONITORS l Process Sampler-Detector Readout Sensitivity Alarm / Control / Monitor Equipment Computer Input I i l l R3tention Off-line liquid CRT display MDC is Alarm on high rad B00in sampler with in control 3E-8 pCi/cc signal or monitor ; effluent scintillation room. Local Cs-137 failure. Auto I diccharge detector. Samples digital secure liquid R-15017A liquid discharges { readout at release on high rad l from the plant. the monitor, signal or channel and a re- failure. Computer corder in the input to IDADS and computer room. RM-11. Rstention Off-line liquid CRT display MDC is Alarm on high rad Basin sampler with in control 3E-8 pCi/cc signal or monitor Inlet scintillation room. Local Cs-137 failure. Auto R-15017B detector. Samples digital secure liquid mixture of Folsom readout at release on high rad canal and plant the monitor, signal or channel waste water, and a re- failure. Computer corder in the input to IDADS and computer room. RM-11. IOSB vent Off-line CRT display MDC is Alarms in control gts particulate in control 4E-11 pCi/cc room. Visual alarm R-15106 continuous air room. Local DAW in IOSB. Auto monitor with display in isotopic secure IOSB exhaust scintillation the IOSB. mix fan on high rad ; detector and alarm. isokinetic sample system. l 11.8-11
_ ..__ _ _. __- _ . _ _ . . _ . . . . . m.._. _ _ _ . _ _ _ . _ _ _ _ . . _ _ _ . . _ . . - _ . _ . . . . . . __ _._ _ ___ __..____.__ _ 11.8.4.2.2 Process / Effluent Radiation Monitors (Continued) Liquid effluent radiation monitor R-15017A monitors and controls Retention Basin Discharges. Another radiation monitor monitors the plant effluent waste water stream. The Retention Basins are the radioactive liquid effluent discharge point for 10 CFR 20 compliance. Automatic termination of a Retention Basin discharge and diversion of the plant effluent stream to a preselected Retention Basin occurs upon high radiation indication or monitor failure by the appropriate monitor.
- The liquid effluent monitors have a fail safe feature that provides an alarm / trip signal upon any of the following conditions:
A. Circuit failure, B. Down-scale failure, or ' C. Controls not set in the operate mode. l l l 11.8-12
~. .- . -_ . . - - _ - -.- . .--- -- _ ~ -.- . . .- _ . - - .
11.9 DOSE ASSESSMENT
- 11.9.1 PERSONNEL MONITORING The personnel monitoring program monitors internal and external radiation exposure to plant personnel who require access to Radiologically Controlled Areas to ensure that administrative and regulatory limits are met. To measure external radiation exposure, radiation workers entering a Radiologically Controlled Area wear a l
thermoluminescent dosimeter (TLD) badge and direct reading dosimetry. Personnel not normally working in or frequenting Radiologically Controlled Areas (i.e., visitors) may wear a TLD or 2 low-range dosimeters in Radiologically Controlled Areas in accordance with Radiation Protection procedures. l Individual exposures are maintained within the limits established in 10 CFR 20. Exposure records are kept current through recording the direct reading dosimeters results and then updating the individual exposure information when the TLD analysis results become available. In addition to TLD badges and direct reading dosimeters, persons requiring more extensive radiation exposure monitoring are supplied with one or more of the following: A. Extremity dosimeters, B. Alarming dosimeters, or C. Multiple whole-body TLDs. Radiation exposure monitoring is also provided for persons, visitors, and company employees not assigned to the plant, but who have occasion to enter Radiologically Controlled Areas or perform work involving possible radiation exposure. Whole body counting and/or radio-bioassay are used to detect radioactive material within the body and to determine internal exposure levels of personnel. Initial baseline measurements are taken for all plant personnel prior to their engaging in radiation work at Rancho Seco. Annual measurements are made throughout a radiation workers employment at Rancho Seco, with additional measurements taken whenever ingestion or inhalation of radioactive material is suspected to have occurred. When possible, a final measurement is taken when the radiation worker terminates radiation work at the plant. Amendment I 11.9-1
l 11.9.1 PERSONNEL MONITORING (Continued) i
- Radiation exposure histories and current occupational exposure records are maintained on all personnel who enter Controlled Areas.
11.9.2 PERSONNEL EXPOSURE RECORD SYSTEM The Radiation Protection department uses a computer data system to process information and perform various record keeping functions. The Radiation Protection (RP) department reports personnel exposure information in accordance with the Radiation Protection program and applicable regulatory requirements. The RP department maintains a personnel exposure history file on the computerized data system. This history file includes: A. Personnel information files, B. Current exposure tracking forms (NRC Form 5 equivalents), and C. Past exposure tracking forms (NRC Form 4 equivalents). The computer data system also tracks daily exposures and provides Radiologically Controlled Area access control outputs, termination exposure reports, bioassay exposure reports, and the annual NRC Regulatory Guide 1.16 and 10 CFR 20.2206(b) reports, 11.3.3 MEDICAL EXAMINATION PROGRAM No special medical examination is considered necessary for radiation workers. SMUD employees receive a pre-employment physical to determine health status and ability to perform their job. Radiation workers are given a pulmonary tunctions test before they are allowed to wear respiratory equipment. Radiation workers are required to take the pulmonary functions test annually in order to maintain their respiratory equipment qualification. Physicians trained in the care and treatment of injuries involving personnel over-exposed to radiation or contaminated with radioactive material are available in a Sacramento area hospital. Radiation Protection personnel will assist the physicians and the hospital to implement proper radiological controls for the treatment of these injured personnel. Amen h nt I 11.9-2
.. . ..- - - . . . . . . . . - - . - - . - - - . . - . - - - . - . . _ _ . - _ .~ - . . _ . - - . .. 1 l 11.10 RADTATTON PROTECTTON PROCRAM l 11.10.1 ORGANIZATION The Plant Manager is responsible for directing the conduct of radiological monitoring and radiation protection control measures at Rancho Seco during the PDM. The Radiation Protection / Chemistry (RP/ Chem) . Superintendent is ! responsible for monitoring radiological conditions within the Industrial l Area and the surrounding environs and specifying radiation protection i requirements for work activities in Radiologically Controlled Areas. The RP/Chen Superintendent reports directly to the Plant Manager. Functional Radiation Protection responsibilities include: A. Handling, receiving, storing, and shipping radioactive materials, B. Monitoring personnel exposure to radioactivity, l C. Maintaining personnel exposure records, D. Reporting exposure histories and abnormal exposure results,
)
I E. Developing and implementing the radiological respiratory protection program, F. Developing and implementing programs for waste minimization and segregation, G. Implementing decontamination procedures to control the spread of contamination i H. Developing and implementing a program to calibrate equipment used in monitoring exposure and radiological conditions I. Implementing and maintaining the Radiological Effluent Control Program, J. Implementing the ALARA Program, K. Developing and implementing the. Hot Particle Control Program, and Amendment I 11.10-1
1 l l 11.10.1 ORGANIZATION (Continued) I L. Solving progrm==mtic problems related to health physics and radiation protection issues to ensure employee and public ) radiation exposure risks are maintained As Low As Reasonably ! Achievable (ALARA). Personnel assigned to work at Rancho Seco and visitors are required to l follow established administrative controls for protection against radiation exposure and contamination. The RP/ Chem Superintendent directs radiation control activities, determines acceptable personnel exposures, maintains accurate dose records, and enforces observance of Radiation Protection Standards. Administrative controls for radiation protection are subject to the same review and approval as those that govern other facility procedures. These procedures include a Radiation Work Permit system, radioactive effluent controls, and the control of radioactive waste packaging, shipping, and disposal. A member of the RP/ Chem department reviews the release permits for 1 continuous and batch releases of radioactive materials in gaseous and l liquid effluent to ensure conformance with the limits of 10 CFR 20 and I the guidelines of 10 CFR 50, Appendix I. Also, a member of the RP/ Chem department reviews proposed shipments of radioactive materials to ensure conformance with the radwaste disposal requirements of 10 CFR 71, 10 CFR 61, and 49 CFR. The RP/ Chem department maintains records of radioactive shipments, stored wastes, and radioactive effluent releases and submits reports to the NRC, as required. Personnel without unrestricted site access badges (visitors) requiring access to the Industrial Area are escorted in accordance with plant administrative site access procedures. These personnel are provided dosimetry in accordance with RP/Chen Department administrative procedures, if necessary. 1 A program to keep levels of radioactive materials in liquid and gaseous effluent to unrestricted areas as low as is reasonably achievable has
]
been implemented pursuant to 10 CFR 50.36a and 10 CFR 50, Appendix I. ' i I Amendment 1 11.10-2
11.10.2 EQUIPMENT, INSTRUMENTATION, AND FACILITIES 11.10.2.1 Perannnel Protective Enuin==nt Special protective or anti-contamination clothing is furnished and worn as necessary to protect personnel against contact with radioactive contamination. This clothing may consist of coveralls, lab coats, , hoods, gloves, and shoe covers. A change room is provided. I Radiological respiratory protective equipment is available for the protection of personnel against airborne radioactive contamination and the possibility of internal radiation exposure. This equipment consists of full-face, air-supplied respirators and self-contained breathing apparatus. Other types of respirators, hoods, plastic suits, and protective clothing are provided as necessary. The protective equipment is maintained in accordance with the manu facturer's recommendations and rules of good practice such as those published by the American Industrial Hygiene Association. The use and
. maintenance of protective clothing and radiological respiratory protective equipment are under the direct control of the Radiation Protection group. Plant personnel must meet training standards for l respiratory protection equipment before they use the equipment in their i work. The use of respiratory equipment is in accordance with 10 CFR 20, Subpart H and NUREG-0041.
Special articles of safety apparel, such as face shields, ear i protectors, and insulated gloves, are available for protection during ' unusual plant activities in the PDN. 11.10.2.2 Radiation Protection Ins t rumen ta tion A variety of instruments are used to cover the entire spectrum of l radiation measurements at Rancho- Seco. These include instruments to detect and measure alpha, beta, gamma, and neutron radiation. Calibration sources are available to allow for instrument calibration, response checks, maintenance, and repair. Portable radiation survey and monitoring instrumentation is available for routine use and is the responsibility of the Radiation Protection group. These instruments include: A. Low and high-range beta-gamma survey meters, B. Neutron survey meters, and Amendment I 11.10-3
11.10.2.2 Radiation Protection I ns t remen ta t i on (Continued) C. Alpha survey instruments. Portable particulate samplers are available for routine and emergency use. In addition to fixed radiation monitoring equipment, portable personnel contamination radiation monitoring instruments may be installed at exits from Radiologically Controlled Areas. These instruments detect i contamination on personnel, materials, or equipment, and help prevent I contamination from being spread into uncontrolled areas. Appropriate monitoring instruments are available within the Radiologically Controlled Areas. A portal monitor checks personnel as they leave the Industrial Area. Rancho Seco has permanently installed area and process radiation monitoring systems. These systems monitor or sample airborne particulate and gaseous radioactivity, including external radiation levels. The systems often provide an audible alarm and radiation level indication in the areas of concern, and read out in the control room. The Radiation Protection group operates and maintains a counting room with instruments for radioactivity measurements. These instruments include: A. Multi-channel gamma spectrometer, B. Liquid scintillation counter for tritium monitoring, C. Alpha-beta flow proportional counter with scaler, and D. Low background beta-gamma counter with scaler. 11.10.2.3 racilities A change room facility is provided where personnel can obtain clean protective clothing and other equipment required for work in radiation and contaminated areas. The change room is located at the +40 foot level of the Auxiliary Building and is considered a buffer zone for the Radiologically Controlled Area. The buffer zone is used for the removal and handling of protective clothing. Showers, sinks, and monitoring equipment are provided in or near the Auxiliary Building change room. f Amendment 1 11.10-4
- -- . . . - . . - . _ . . . - - . _ - - . - . - - . ~ .. . - . - . - . - 11.10.2.3 racili tian (Continued) Equipment decontamination facilities are provided for large and small items of plant equipment and components. A cask decontamination area is adjacent to the spent fuel pool. To protect personnel from radiation exposure, the Radiologically Controlled Areas on site are divided into areas of increasingly controlled access, depending upon the radiation levels. Control of ' personnel access to contaminated and radiation areas is accomplished by appropriate radiation or contamination caution signs, barricades, locked doors or gates, and audible and/or visual indicators. The Radiation Work Permit system and other administrative controls control access to the Radiologically Controlled Areas. l Instrument facilities consists of a calibration room and an instrument repair area. The maintenance and use of radiation protection facilities and equipment is the responsibility of the Radiation Protection group. 11.10.3 RADIATION PROTECTION PROCEDURES 11.10.3.1 Procedures The RP/ Chem department has prepared and implemented procedures governing personnel radiation protection. The Radiation Protection group maintains these procedures consistent with the requirements of 10 CFR
- 20. The procedures cover the scope of procedures recommended in Appendix A of Safety Guide 33, November 1972. Radiation protection procedures and required safety evaluations are reviewed and approved by the Plant Review Committee.
Chemistry procedures address the requirements governing the release of radioactive liquid and gaseous effluent to the environment. Chemistry procedures and their required safety evaluations are reviewed and approved by the Plant Review Committee. Radiation protection procedures also address the requirements governing the disposal of solid radioactive waste. These procedures comprise the Process Control Program. The philosophies, policies, and objectives of Radiation Protection group procedures implement the federal regulations contained in the Code of Federal Regulations (CFR) and are designed to maintain doses to workers and the general public ALARA. Amendment 1 11.10-5
. _. -.____._.._____.-_._-m._...._ . . . _ _ _ _
h t 11.10.3.2 Ratll ati on Work Permit Procedure The Radiation Work Permit (RWP) is an administrative tool used to inform workers of the radiological conditions in the work area and the . I requirements - for protective clothing, respiratory protective equipment, ; dosimetry, and engineering controls. The RWP may be used to delineate job prerequisites, radiological safety practices, or additional requirements as needed. As an exposure tracking device, the RWP , provides information necessary to ensure that exposures are kept ALARA. An RWP is required for:
- 1. Work in a Radiation Area, i
- 2. Work in or entry to areas of potential neutron exposure,
- 3. Work in or entry to Contaminated Areas or equipment, 4.
Work in or entry to High Radiation Areas, Secured-High Radiation Areas, Very High Radiation Areas, or Airborne Radioactivity Areas, 5. Work involving opening, cutting, or welding on systems which contain radioactive contamination.
- 6. Work as deemed necessary by Radiation Protection supervision. l
- 7. Work in a Hot Particle Zone.
- 8. Entry into Radiologically Controlled Areas.
11.10.4 PERIODIC PERSONNEL EXPOSURE REPORTING Two reports are issued annually to the NRC to meet reporting requirements under the Possession-Only License (POL) for Rancho Seco. These reports contain information for the preceding year in the following areas: 1
- 1. Tabulation of personnel receiving exposures greater than 100 l mrem during the preceding year according to work and job functions. This requirement is addressed in plant Technical Specification Section D6 and follows the format of NRC Regulatory Guide 1.16.
I
- 2. Tabulation of' numbers of personnel for whom exposure monitoring was provided in accordance with 10 CFR 20.2206 (b) and the Rancho Seco Permanently Defueled Technical Specifications.
Amendment 1 11.10-6 l _ ._ __ ________________________________________________i
I i 11.11 REFERENCES j
- 1. l License Amendment No. 119, dated March 19, 1992, Permanently Defueled Technical Specifications 2.
Safety Analysis and No Significant Hazards Consideration (Log No. 1091, Revision 3) for Proposed Amendment 182, Revision 3, Permanently Defueled Technical Specifications 3. License Amendment No. 117, dated March 17, 1992, Possession-Only License
- 4. 10 CFR 20 and 10 CFR 50
-5. Rancho Seco Process Control Program (PCP) 6.
Rancho Seco Off-site Dose Calculation Manual (ODCM) 7. Rancho Seco Radiological Environmental Monitoring Program (REMP)
- 8. SMUD Calculation E-SFC-N0049, Revision 3, " Maximum Predicted Whole Body and Skin Doses and Dose Rates at the Site Boundary from Postulated Accidents During Plant Shutdown."
i Amendment 2 11.11-1
TABLE OF CONTENTS (Continued) Section Title Page 12.5.2.1.4 Casualty Procedures 12.5-6 12.5.2.2 Other Proemdures 12.5-6 12.5.2.2.1 Naintenance Procedures 12.5-6 12.5.2.2.2 Instrument and Control (IEC) Procedures 12.5-7 l 12.5.2.2.3 Surveillance Procedures 12.5-7 12.5.2.2.4 Chemistry Procedures 12.5-7 12.5.2.2.5 Radioactive Waste Management Procedures 12.5-8 l 12.5.2.2.6 Radiation Protection Procedures 12.5-8 l 12.5.2.2.7 Security Plan 12.5-8 12.5.2.2.8 Emergency Plan and Implementing Procedures 12.5-8 12.5.2.2.9 Fire Protection Procedures 12.5-9 l 12.5.2.2.10 Quality Assurance 12.5-9 l 12.5.2.2.11 Certified Fuel Handler and Non-Certified Operator 12.5-9 Training Programs l 12.5.3 PROCESS STANDARDS 12.5-9 12.6 INDUSTRIAL SECURITY 12.6-1 12.7 RECORDS 12.7-1 12.7.1 OPERATING RECORDS 12.7-1 12.7.2 ADMINISTRATIVE RECORDS 12.7-1 12.7.3 MAINTENANCE RECORDS 12.7-2 12.7.4 HEALTH PHYSICS RECORDS 12.7-2 12.7.5 OTHER RECORDS 12.7-2
12.8 REFERENCES
12.8-1 Amendment 1 12-111
LIST OF FIGURES E1211IA Title 12.1-1 SMUD Nuclear Organization j I l i I Amendment 2 12-iv i l I
l DSAR CHAPTER 12. CONDUCT OF OPERATIONS l This chapter describes the organization and general plans for conducting activities at Rancho Seco during the Permanently Defueled Mode (PDM). Plant organization is included with brief descriptions of the i responsibilities of Managers, Superintendents, Supervisors, and other key personnel. The training program for the plant staff is described, along with a more general discussion of the replacement and retraining program. The standards and procedures that govern daily plant activities, the records developed as a result of these activities, and the controls used that promote plant safety and assure compliance with the facility license and federal, state, and local regulations applicable during the PDM are discussed. l 12.1 ORGANIZATIONAL STRUCTURE OF SMUD 12.1.1 NUCLEAR ORGANIZATION The SMUD organization that oversees the activities at the Rancho Seco ) nuclear facility is presented in Figure 12.1-1. The Manager, Plant Closure & Decommissioning (Plant Manager) heads the on-site Rancho Seco nuclear organization and directs the activities of the functional on-site departments. The Plant Manager is responsible for the overall, day-to-day safe operation and maintenance of Rancho during the PDM. The Plant Manager reports directly to the Director, Power Generation. The Director, Power Generations reports to the Assistant General Manager (AGM) Energy Supply & Chief Engineer, who reports to the SMUD General Manager, who reports to the SMUD Board of Directors. The SMUD General Manager, through the AGM Energy Supply & Chief Engineer and the Director, Power Generation, has corporate responsibility for the overall safe operation of Rancho Seco and ensures acceptable performance of the staff in operating, maintaining, and providing technical support to the facility during the PDM. 12.1.2 PLANT PERSONNEL RESPONSIBILITIES AND AUTHORITIES The responsibilities and authority of major plant positions are summarized below. All plant personnel are selected and trained for their assigned duties, with particular emphasis on the supervisory, technical, and operating staffs to assure safe and efficient operation of the plant during the PDM. In addition to the responsibilities and Amendment 2 12.1-1
1 l I 12.1.2 PLANT PERSONNEL RESPONSIBILITIES AND AUTHORITIES (Continued) , authorities stated below, each department head is responsible for conducting a departmental training program which meets the applicable requirements and standards. 12.1.2.1 Plant Manmaer The Plant Manager is responsible to the SNUD General Manager, through the Director, Power Generation and AGM Energy Supply & Chief Engineer, l for the safe and reliable operation of Rancho Seco. The Plant Manager assures the safety of plant personnel and the general public, approves and administers the nuclear organization budget, and approves overall scheduling of plant activities and expenditures associated with those activities. The Plant Manager is also responsible for providing management oversight of nuclear plant administrative functions such as: 1 1
- 1. Cost control, l
)
- 2. Commitment Management, l
- 3. Audit / Inquiry Responses,
- 4. District Representative / Negotiator,
- 5. Resource / Budget Management,
- 6. Procedure Management,
- 7. Plant Closure and Decommissioning Project Management Oversight,
- 8. Nuclear Fuels Management, and
- 9. Plant Personnel and Operator Training i
1 The Plant Manager has the authority to establish nuclear organization policy and make commitments to the NRC and is supported by the following personnel: ; Amendment 2 12.1-2
.-. . . . - . -_ . - _ . ...~ - . _ . - . -.- . - - _ - __~ . _ .
l l 12.1.2.1 Plant unnmaer (Continued) { 1 A. Quality Annurance/Licanmina/adelnistration Sunerintendant
)
The Quality Assurance / Licensing / Administration Superintendent l ensures that the Quality Assurance, Licensing, and l Administrative programs are implemented in accordance with the applicable regulatory requirements. The Superintendent (1) is independent of the pressures of plant operations during the PDM, (2) has sufficient authority and organizational freedom to identify problems that affect j quality, recommend solutions, and verify implementation of solutions, and (3) has the authority to take any issues concerning the quality of operations at Rancho Seco to the Director, Power Generations. The Quality Assurance / Licensing / Administration Superintendent l is responsible for the overall administration of the Rancho Seco Quality Assurance Programs. Areas of functional responsibility include quality auditing, quality control, and the Corrective Action Program. l In the Licensing area, the Quality Assurance / Licensing / Administration Superintendent provides regulatory guidance, compliance, and licensing services and maintains the Rancho Seco facility license. The Superintendent ensures compliance l with regulatory requirements and commitments, controls and coordinates correspondence and interface with the NRC, and is responsible for updating licensing basis documents, maintaining the Permanently Defueled Technical Specifications (PDTS), and coordinating and managing District commitments l with the NRC and other federal, State, and local regulatory agencies. In the Administrative area, the Quality Assurance / Licensing / Administration Superintendent provides centralized administrative services and support, including document control, records management, and office services, for the entire nuclear organization. The Superintendent also establishes policies and direction within the Administrative area to support the goals and objectives established by the Plant Manager for Rancho Seco during the PDH. Amendment 2 12.1-3
- - . - . .- . - . _ . - - -. . . . . ~ - - - - _ ~ _ - - ..
l d 4 i 12.1.2.1 Elant unnmaer (Continued) , 4 B. Oneratiana sunarintendent ! 1 The Operations Superintendent ensures plant activities and , operations during the PDM are conducted in accordance with l the requirements of the facility license, the rDTS, plant l , ] operating procedures, and applicable local, state, and ! federal regulations. Details regarding operator shift crews i j are discussed in Section 12.1.2.2. The Superintendent i manages the activities associated with safe storage of spent fuel during the PDM. C. security onerattana sunervisor The Security Operations Supervisor manages the activities of , the Rancho Seco Security group during routine, emergency, and , contingency PDM conditions. The Supervisor develops and maintains the security training program and accurity related licensing basis documents and implementing procedures to ! ensure compliance with site, local, state, and federal ! security related requirements and regulations. i D. site Train.ina senervinar l The Site Training Supervisor develops and implements l programs, plans, policies, and procedures which ensure nuclear organization personnel are trained in accordance with District standards and applicable regulatory standards and commitments. Also, the Supervisor directs the preparation, ! scheduling, and conduct of the Certified Fuel Handler, Shut- i down Control Room Operator, and Non-Certified Operator Training Programs. ' The Site Training Supervisor reports to l the Plant Manager. E. Radiation P ro t ec tion / chm = 1 s t rv sunerintendent i The Radiation Protection / Chemistry (RP/ Chem) Superintendent is responsible for:
- 1. Minimizing employee and public exposure to radiation and .
- radioactive material, j
- 2. Maintaining a personnel radiation exposure and ,
monitoring record keeping program, ) 1 Amendment 2 12.1-4
12.1.2.1 Plant Managgg (Continued)
- 3. Ensuring compliance with regulatory requirements regarding radiation protection and radwaste management,
- 4. Establishing and evaluating the content and effectiveness of radiation technician training, '
- 5. Developing, maintaining, and implementing the ALARA l progrsa, !
i l, , 6. Ensuring that the Radiological Environmental Monitoring
- Program is implemented in accordance with regulatory requirements and commitments,
- 7. Developing, implementing, and maintaining the Rancho Seco Emergency Plan, i
l
- 8. Managing the plant chemistry program, which establishes chemistry and radiochemistry controls and surveillances !
for plant systems through the Off-site Dose Calculation l Manual and implementing procedures and Chemistry administrativa procedures, i 9. Controlling and monitoring radioactive liquid and gaseous releases, i 10. Establishing and evaluating the content and effectiveness of chemistry technician training, and f
- 11. Providing General Employee Treining (GET) and Fire a Brigade training.
- 12. Carrying out health physics functions. This provides sufficient organizational freedom to ensure health j physics functions will be performed independent of !
operating pressures. l F. Maintenance sunerintendant a The Maintenance Superintendent (1) maintains the physical condition of the plant, through the performance of in-service . confirmations and preventive and corrective maintenance, to optimize reliability of systems end components, and (2) l directs modification activitis.33. The Superintendent also l provides nuclear organization support in the area of scheduling. Amendment 2 12.1-5
1 J-i I i ] 12.1.2.1 Plant Manmaar (Continued) i
- G. Technical sersicem Runerintandant i.
The Technical Services Superintendent is responsible for operations and maintenance support, . support for and technical j direction in plant design, developing design modifications to j j plant systems, design specification development, design j
- j. change control, configuration management, and discipline ;
j engineering. Also, the Superintendent is responsible for l l j system engineering, performance monitoring, surveillance j testing, the In-service Confirmation Program, reactor j engineering, welding, fire protection, and instrumentation ' 3 and controls engineering. In addition, the Superintendent l l (1) is responsible for the Asset Recovery Program for the f i systems, structures, and components not required to function ! during the PDM, and (2) maintains baseline design documents, I the Master Equipment List (NEL), and other documents defining I technical requirements for systems, structures, and } components required to function during the PDN. t i r i 12.1.2.2 onerator shift crews 3 j The minimum shift crew composition consists of two Operations personnel, l at least one of which must be a Certified Fuel Handler. The other member of the shift crew may be a non-certified operator. A minimum of , i j one member of the shift crew must be in the control room when fuel is j stored in the spent fuel pool. The person that stands watch in the ; control room must be qualified to do so in accordance with the ; l l Operations Department training procedures. This individual must be l l either a Shut-down Control Room Operator (either certified or non-l certified) or the Shift Supervisor. If a member of the minimum shift
- crew is absent or incapacitated due to illness or injury, a qualified replacement must be designated to report to the site within two hours.
l i l Fuel handling operations conducted during the PDM must be directly j ! supervised by a Certified Fuel Handler. ; l j The duties of the Operations Superintendent are discussed in Section ; , 12.1.2.1. The responsibilities and authorities of shif t operations i personnel are described below-i ' < E s i Amendment 2 4 12.1-6 ;
t i 12.1.2.2 Onerator shi f t crews (Continued) A. shift S une rvis o r The Shif t Supervisor is a Certified Fuel Handler and is accountable for safe and efficient plant operation during the ( PDM in accordance with the Permanently Defueled Technical Specifications, federal, state, and local regulations, and plant procedures. The Shif t Supervisor has the authority to terminate any site activity judged to be a public, personnel, or station hazard and is generally present in the control room during major plant evolutions. l B. Shut-down control Room Onerator The Shut-down Control Room Operator is either a Certified Fuel Handler or a non-certified operator, who is trained and qualified to stand watch in the control room in accordance with Operations Department procedures. C. Non-Certified Onerator The non-certified operator monitors and operates plant equipment and systems in support of station operation during the PDN. The non-certified operator also checks, analyzes, and logs equipment / system parameters and initiates corrective action when abnormal conditions exist. A non-certified operator may be trained and qualified in accordance with Operation Department training procedures as a Shut-down Control Room Operator. A site Fire Brigade, which includes plant operators and security personnel, is maintained on-site at all times as required by the Fire Protection Plan. Operations personnel are on shift who are trained and qualified as RP Responders to perform the following functions:
- 1. Initial response to emergencies with known or previously evaluated radiological conditions,
- 2. Routine, informational radiological area monitoring to verify existing conditions are comparable to the most recent RP Technician performed radiological surveys, and Amendment 1 12.1-7
._ _. _ _ ___ ~ ._. _ ___ _ _ _ .
I 12.1.2.2 onerator shift crews (Continued)
- 3. Entry into radiologically controlled areas under existing
+ Radiation Work Permits (RWPs) to perform routine Operator j activities such as surveillances. ] RP Responders are not trained or permitted to perform tasks that require j an ANSI qualified RP Technician. Plant operation evolutions that j require RP Technician support, such as fuel handling operations, are j scheduled and conducted when RP Technicians are available. f The RP Responder program requires RP Supervision to be on-call so that j RP Responders may contact an individual qualified to perform RP j Technician activities whenever (1) additional instruction or guidance is i needed, or (2) RP Responders encounter radiological conditions beyond ! the scope of their training and qualifications. The RP Responder i program requires a maximum two hour response time for RP Supervision or l an RP Techelcian to report to the site, when necessary, to perform any ! required RP Technician functions. I 12.1.2.3 succenaton of mannannihility i 3 The succession to responsibility for operation of the plant in the event of absences, incapacitation of personnel, or other emergencies is as
- follows:
! 1. Plant Manager i i l 2. Operations Superintendent i i
- 3. Shift Supervisor a
12.1.3 OUALIFICATIONS OF NUCLEAR PLANT PERSONNEL l l Each member of the plant staff meets or exceeds the minimum i qualifications of ANSI N18.1-1971 for ccmparable positions, except for l the RP/ Chem Superintendent, who meets or exceeds the recommendations and l 4 qualifications of Regulatory Guide 1.8, September 1975, for the Radiation Protection Manager. Plant personnel are selected and trained . for their assigned duties, with particular emphasis on the supervisory,
, technical, and operating staffs, to assure activities are conducted in a safe and efficient manner at Rancho Seco during the PDM.
Amendment 1 12.1-8
b 4 5 12.1.3 QUALIFICATIONS OF NUCLEAR PLANT PERSONNEL (Continued) Personnel selection policy for the Rancho Seco staff includes reference checks, motor vehicle driver's record check, and a review of each application for employment with particular emphasis on arrest record, previous employment and reason for leaving, and military service record. l Amendment 2 12.1-9
12.2 PERSONNEL TRAINING Retraining and Replacement Training Programs for the plant operating staff are maintained under the direction of the Plant Manager and are conducted in accordance with plant procedures. Operations staff training is conducted by the Site Training Supervisor. Retraining and Replacement Training meets or exceeds the requirements and recommendations of ANSI N18.1-1973, for non Operations staff and ANSI 3.1-1981, for Operations staff. Each department head is responsible for conducting a departmental training program that meets the applicable requirements and standards, including testing individuals as appropriate and maintaining training documentation within areas of responsibility. Rancho Seco security force training is the responsibility of the Security Operations Supervisor. 12.2.1 TRAINING PROGRAMS The following descriptions outline the training program guidelines that I govern personnel training at Rancho Seco. The following is a list of training programs that comprise most of the discipline related training programs at Rancho Seco: 1 1
- 1. Certified Fuel Handler Training
- 2. Non-Certified Operator Training
- 3. Shutdown Control Room Operator Training
- 4. RP Responder Training
- 5. Chem / Rad Decosuaissioning Technician Training
- 6. Maintenance Training
- 7. Licensing Training In addition, training programs required by the Emergency Plan, Physical Security Plan, Fire Protection Plan, Permanently Defueled Technical Specifications, administrative requirements, and applicable state and federal regulations include the follows:
Amendment 2 12.2-1
}
l 12.2.1 TRAINING PROGRAMS (Continued)
- 1. General Employee Training (GET)
- 2. First Aid
- 3. Fire Brigade, Fire Protection
- 4. Emergency Plan
- 5. Security
- 6. Quality Assurance 1
- 7. Radwaste Handler 1
- 8. Dosimetry Technician l
- 9. ALARA j i
- 10. Safety
- 11. PRC and MSRC 1
- 12. Qualified Reviewer (10 CFR 50.59 Safety Evaluations)
Personnel working at Rancho Seco participate in the training programs required for their job position. Training is conducted and documented in accordance with departmental training procedures. 12.2.2 CERTIFIED FUEL RANDLER TRAINING PROGRAMS Training of personnel for certified Fuel Handler is conducted in accordance with Operations Department Training Procedures. This training program is designed to ensure the trainee is prepared to safely and efficiently operate the plant during the PDM with the fuel stored in the spent fuel pool. Instructors who teach the Certified Fuel Handler Training Program shall be certified to teach in accordance with Operations Department training procedures. The instructors shall participate in the Proficiency Training Program. 12.2-2
l 12.2.2.1 Structure of Ynitial Certification Proaram The Initial Certification Training Program is divided into three basic phases. These phases are the classroom phase, the self study phase, and i the job training phase. l i
- 1. Classroom Phase The classroom phase of the training program consists of formal classroom presentations in fundamentals, related theoretical subjects, plant systems, and procedures.
- 2. Self Study Phame The self study phase of the training program is designed to allow students to gain an in-depth knowledge of systems and procedures. Each student is issued a training manual, which contains study references and check-out sheets for self-evaluation. This manual ensures the student will i accomplish the study objectives and provides the documentation of successful completion of check-outs. When a )
trainee completes all the topics on a particular checklist, a ; certified training instructor or Certified Fuel Handler administers an Operating examination / quiz. Each student is required to satisfactorily complete all checklists. The self study phase contains three modules (1) Systems Training, 1 1 (2) Procedure Training, and (3) Miscellaneous Materials Training.
- 3. Job Trainina Phase 1
i The job training phase of the training program is designed to provide the opportunity for the trainee to gain a " feel" for system controls through manipulation of controls and develop a greater sense of expected equipment response under varying conditions. It also allows the trainee to develop an understanding of surveillance requirements, procedures, and testing requirements, and gain additional operating experiences. The job training phase consists of an On-The-Job Training module. I 12.2.2.2 Evaluation End of course evaluation for the Certified Fuel Handler initial training consists of two separate examinations. j i 12.2-3 I
)
-_ - - ..-_ - .- - . . . . . . .. _ ~ _ - - - _ - ..~ - . .- .--- _...-.- . . - - - 12.2.2.2 Evaluation (Continued)
- 1. Written examination
- 2. Oral / walk-through Following successful completion of the training program, candidates are '
eligible for certification. 12.2.3 STRUCTURE OF CERTIFIED FUEL HANDLER PROFICIENCY PROGRAM The Certified Fuel Handler Proficiency Training Program is conducted biennially
- with successive programs using the same format and schedule.
This program is divided into three basic phases: (1) the classroom phase, (2) the self study phase, and (3) the job training phase. The components of each phase and a brief description are provided below: A. clamaroom Phame The Classroom Phase contains the following courses:
- 1. Fundamentals Review Consists of classroom instruction on radiation control and safety, reactor theory, and heat transfer and fluid flow associated with spent fuel storage.
- 2. System and Procedures Review Consists of classroom instruction on plant system design, construction, and operation. The instruction also reviews major plant systems, operating and casualty procedures, and the Emergency Plan.
- 3. Plant Modification Consists of classroom instruction on recent changes in plant construction and operation.
In addition, specific topical areas will, in part, be determined by the results of the annual requalification examinations and plant operating history.
- Biennial is defined as 24 calendar months in the requalification procedures.
12.2-4
I 12.2.3 STRUCTURE OF CERTIFIED FUEL HANDLER PROFICIENCY PROGRAM (Continued) B. Self Study Phase The self study phase consists of reading assignments. Plant modifications, procedure changes, operational assessments, and operational events may be contained in the reading assignments. C. Job Trainina Phase The job training phase consists of system operation rsview, conducted by the Shift Supervisor, a certified instructor, or an Operations staff supervisor. 12.2.3.1 Evaluation A. Written Rummination The proficiency program includes an evaluation and observation system to obtain maximum benefits from the retraining program and provide a method for determining the areas in which retraining is needed. Written examinations are administered in accordance with the Operations Department Training Procedures. Written exams are used as a method to determine Certified Fuel Handler knowledge of subjects covered in the proficiency program. B. Ooeratina Rummination The Site Training Supervisor or his/her designee administers l an operating examination to each certified Fuel Handler annually. 12.2.4 LICENSING TRAINING PROGRAM The intent of the Licensing Training Program is to assist personnel in adapting their technical expertise to the performance of various tasks and administrative processes to ensure compliance with the facility License and applicable regulatory requirements, such as the Permanently Defueled Technical Specifications and 10 CFR 50.59. The goal of the program is to assure quality performance of processes and tasks. This Amendment 2 12.2-5
12.2.4 LICENSING TRAINING PROGRAM (Continued) is accomplished by cross-training to ensure personnel have the requisite knowledge and skill to perform satisfactorily. Each participating department determines positions requiring training under this Program. The Quality Assurance / Licensing / Administrative Superintendent is responsible for administering the Licensing Training Program to the designated individuals. 12.2.5 MAINTENANCE TRAINING PROGRAM Training of personnel in the Maintenance disciplines is conducted in accordance with Maintenance Department Training Procedures. This program is designed to prepare the trainee for safe and efficient maintenance / operation of Rancho Seco in the PDN. The Maintenance Superintendent is responsible for implementing the Maintenance Training Program. The Maintenance Training Program consists j of Administrative Process Training. ' Administrative Process Training is based on developing and maintaining the prerequisite skills and knowledge required by maintenance workers to j accomplish specific maintenance tasks. These tasks include, but are not ' limited to, initiating, implementing, adhering to and/or processing:
- 1. Work Requests (WRs),
- 2. Potential Deviations from Quality (PDQs),
- 3. Ignition source permits,
- 4. Clearance / test tags, and
- 5. Scaffolding requests, Also, maintenance personnel are trained in the maintenance of fork lifts and plant cranes, and they should have a working knowledge of the plant drawing system and the vendor manual system.
Amendment 2 12.2-6
i l 12.2.5.1 Initini Trainina Initial training is designed to provide the trainee with knowledge and
- t. kills necessary to function as part of the Maintenance Department during the PDM. This training is typically completed in 2 years.
~ 12.2.5.2 continuina Trainina Continuing training is designed to reinforce and improve knowledge and < skills of Maintenance personnel. Continuing training is conducted on a 2 year cycle. 12.2.6 SITE SUPPORT TRAINING PROGRAMS Site support training is conducted by the department whose area of responsibility covers a particular site support function. Training of personnel in the site support disciplines is conducted in accordance with the applicable department's training procedures. These programs are designed to prepare the trainee for safe and efficient maintenance / operation of Rar.cho Seco during the PDM. Site support training includes training programs for the following areas:
- 1. Chem / Rad Decommissioning Technician
- 2. RP Responder
- 3. General Employee Training
- 4. Fire Protection /First Aid 12.2.6.1 Initial Trainina The responsibility for each site support training program is assigned to the applicable department manager. Classroom and laboratory training i are provided by the responsible department when appropriate or ;
necessary. On the Job Training (DJT) is provided within most ' disciplines. OJT consists of, but is not limited to, task training and evaluation, procedure training, and specific discipline-related training requirements. 1 l l Amendment 1 l l 12.2-7 l i
12.2.6.2 continnino Trainina Continuing training programs are designed to meet the specific needs of the participating departments and may include plant change review, procedure change review, administrative training commitments, OJT training review, and material from the initial training program. ; i l 1 I i l Amendment 1 12.2-8
12.3 EMERGENCY PLAM The Emergency Plan in effect during the PDM provides a description of the organization, equipment, and preparations made to enable appropriate and effective response to postulated emergency situations that may arise at Rancho Seco during the PDN. The focus of concern for the Emergency Plan is the protection of plant personnel and the surrounding population. The Emergency Planning Ione for Mancho Seco during the PDM is the Industrial Area. The calculated exposures at the Industrial Area i boundary for postulated emergency planning accident scenarios in the PDM do not result in a radiological release that could exceed the EPA plume exposure Protective Action Guidelines. Emergency response activities performed at Rancho Seco are the responsibility of SMUD management. Off-site emergency response activities are under the authority of public agencies, with SMUD , providing the appropriate information to these agencies dvring the course of an emergency as directed by the SMUD Emergency C.sardinator. l The Emergency Plan Implementing Procedures (EPIPs) address specific actions that should be taken when responding to various types of. emergencies. The EPIPs also provide data, instructions, personnel assignments, criteria for site evacuation, other specific information , needed during an emergency, and instructions to obtain names and telephone numbers for emergency call-out, if necessary. The Emergency Plan and EPIPs form a complete and detailed program which aids Rancho Seco personnel and affected off-site agencies in the safe and efficient handling of emergency conditions. The emergency conditions considered in the development of the defueled Emergency Plan include the design basis accidents or conditions considered credible during the PDM as well as other postulated accidents. The design basis accidents or conditions considered are the Fuel Handling Accident and the Loss Of Off-site Power (LOOP) condition. These design basis events are further evaluated in DSAR Chapter 14. The Emergency Plan implemented in the PDM reflects many NRC granted exemptions from the emergency preparedness requirements specified in 10 CFR 50.47(b), 10 CFR 50, Appendix E, and 10 CFR 50.54 (q) . The defueled condition Emergency Plan conforms as fully as practicable to the emergency planning requirements applicable to Rancho Seco in the PDM. ! I 1 i i k I f L
- 12.3-1
3 12.4 REVIEW AND AUDIT OF OPERATIONS Administrative controls are in place in the form of approved written procedures to ensure the safe conduct of activities, testing, and 4 response to emergency situations during the PDM. Plant management holds frequent meetings to keep all staff areas informed of the status of plant activities during the PDM. Two committees are responsible for review and audit of plant activities during the PDM: (1) the Plant Review Committee (PRC) and (2) the Management Safety Review Committee (MSRC) . The PRC advises the i Director, Power Generation and the MSRC on the required safety evaluations of activities proposed during the PDM and other matters related to Nuclear Safety. The MSRC provides independent review and audit of designated activities in the areas of:
- 1. Facility Operation, 4
- 2. Engineering / Nuclear Engineering,
- 3. Chemistry and Radiochemistry, l 4. Radiological Safety, and
- 5. Quality Assurance Practices, '
and advises the AGM Energy Supply & Chief Engineer in these areas. PRC and MSRC requirements and responsibilities are detailed in the l Permanently Defueled Technical Specifications and Rancho Seco { Administrative Procedures. 4 i i Amendment 2 12.4-1
NUCLEAR ORGANIZATION BOARD OF DIRECTORS i GENERAL MANAGER I AGM, ENERGY SUPPLY
& CIIIEF ENGINEER I
------------------- DIRECTOR, POWER GENERATION I
MANAGEMENT SAFE'IT l j REVIEW COMMITTEE MANAGER, g PLANT CLOSURE & I DECOMMISSIONING l (Quality l PIANT REVIEW COMMITTEE issues) l I I I l NUCLEAR QUALITY / NUCLEAR TECIINICAL NUCLEAR NUCLEAR L_ LICENSING / ADMIN MAINTENANCE SERVICES RP & CIIEMISTRY OPERATIONS SUPERINTENDENT SUPERINTENDENT SUPERINTENDENT SUPERINTENDENT SUPERINTENDENT
- ShiftSupervisor SECURITY SITE OPERATIONS TRAINING
- Certified FuelIIandler FIGURE 12.1-1 SUPERVISOR SUPERVISOR
- Non-Certified Operator SMUD NUCLEAR ORGANIZATION Anaendment 2 JSMUD M Af'n AMrNTO MilNICIPAt tfTilITY DISTRICT
l l TABLE OF CONTENTS i section Title Eaga 1 1
- 14. SAFETY ANALYSIS 14.1-1 14.1 AccInENTs CONSIDERED CREDIBLE IN THE 14.1-1 PERMANENTLY DEFURY.Rn MODE fPDM) 14.1.1 FUEL HANDLING ACCIDENT 14.1-1 k
1 14.1.1.1 Analysis and Results ' 14.1-1 14.1.2 LOSS OF OFF-SITE (A-C) POWER 14.1-2
14.2 REFERENCES
14.2-1 1 I I i Amenchment 2 14-1
I LIST OF TABLES Table Title Page i 14.1-1 Fuel Handling Accident Analysis Assumptions 14.1-3 I i
}
14-11
DSAR CHAPTER 14. SAFETY ANALYSIS 14.1 ACCTDENTS CONSTDERED CREnTRLE IN THE PERMANFMTLY DEFUELED MODE During the Permanently Defueled Mode (PDM), the number and consequences f of design basis accidents or conditions considered credible are greatly reduced from the accidents evaluated for the initial licensing of the Rancho Seco nuclear facility. The Rancho Seco design basis accidents and/or conditions that are considered credible during the PDM are: A. The Fuel Handling Accident (FRA); and B. A Loss Of Off-site (a-c) Power (LOOP) condition. 14.1.1 FUEL HANDLING ACCIDENT The spent fuel assemblies stored at Rancho Seco during the PDM are stored in the spent fuel pool and handled entirely under water. The spent fuel assemblies rest in fuel storage racks that contain a neutron absorbing material (Boraflex) and have an ever-safe geometric array. Under this storage condition, no boron is required to be maintained in the spent fuel pool water to ensure K,,, is maintained below 0.95; therefore, a criticality accident during fuel storage or handling is not considered credible. But, mechanical damage to the fuel assemblies during fuel handling operations in the PDM is possible. The mechanical damage type of accident (a dropped fuel assembly) is considered the maximum potential source of activity release during the PDM. The load that can be transported over spent fuel during the PDM is limited in the Permanently Defueled Technical Specifications to one fuel assembly, control component, and associated handling tool, which together weigh less than 2,000 pounds. A fuel handling accident would result in, at most, the release of gap activity from one fuel assembly, due to the limitation of available impact kinetic energy (4). 14.1.1.1 Annivsis and Results The FRA is the worst case design basis accident postulated to occur l during the PDM. The FRA is assumed to release the gap gas activity from all 208 fuel rods in the hottest fuel assembly directly to the spent fuel pool water. The FRA off-site dose consequences evaluation takes no credit for retention of noble gases or iodine in the spent fuel pool water. The reactor is assumed to have been shut down for five years. The assumptions made for the District's THA analysis are shown in Table ( 14.1-1. Amendment 2 ) 14.1-1
_ _. . . _ . __ m_ 14.1.1.1 Anm1 vain and n.= nits (Continued) As a minimum, the gap gases will pass through 10 feet of water. Although there is experimental evidence that a portion of the noble gases will remain in the water, no retention of noble gases is assumed. The noble gas activity released to the spent fuel pool water at five years after final reactor shutdown is 5.41E+03 Curies of Krypton 85 (Kr-85). The FRA analysis is conservative since the reactor was actually shut down on June 7,1989. Due to the short half-life (8.05 days) and the absence of a radioiodine production mechanism during the PDN, radiciodine is not present and the calculated thyroid dose resulting from the FRA is zero. The primary isotope of concern following a FRA is Kr-85 (see DSAR Section 11.1). Results of the FRA analysis show that the two-hour integrated total body dose at the Industrial Area boundary to the maximum exposed individual is 9.9 mrem. This dose is: A. An extremely small fraction of the 10 CFR 100 design basis ' accident dose limits, B. Significantly less than tho 10 CFR 20 normal operational annual dose limit, and t C. Significantly less than the EPA plume exposure Protective Action Guidelines. 14.1.2 LOSS OF OFF-SITE (A-C) POWER During the PDM, a loss of off-site a-c power (LOOP) incident would result in the loss of all spent fuel pool cooling systems. However, based on the reduced level of decay heat emanating from the fuel stored in the spent fuel pool (see DSAR Table 9.4-2) and conservative District heat-up calculations, several days are available to restore a-c power or take alternative actions to ensure the spent fuel pool temperature and level are maintained within the Technical Specification limits. Amendment 2 14.1-2
_ ._ ._. .m . _ . . _ _ _ . . _ _ _ . _ . . . .- . . . _.-._m... . . _ . _ . _ . .- __ . . _ _ . _ _ , _ . . i - TABLE 14.1-1 FUEL HANDLING ACCIDENT ANALYSIS ASSUNPTIONS i
)
l
- 1. Period of continuous operation prior to final 1017 days shut-down -{
1 l
- 2. Power level of hottest assembly during reactor 26.6 MWt operation
- 3. Length of decay time after final shut-down 5 years when FHA occurs
- 4. Water to air partition factor None
- 5. Exhaust air filters decontamination factor None
- 6. Distance from release to the maximum exposed 100 meters individual
- 7. Atmospheric dispersion coefficient, X/Q 3.61E-3 s/m*
- 8. Fuel assembly gap gas source tern 5.41E+03 Curies-Kr-85 (see DSAR Section 11.1)
Amendment 2 14.1-3
.- . .. . . - - . _ . - . = . . - . .. . - . . - . . . . . - . _ . - _ . .. . _ _ . - .
i 14.1.2 LOSS OF OFF-SITE (A-C) POWER (Continued) The minimum time available before the spent fuel pool could heat up from 70*F to 140*F following the occurrence of a LOOP condition is 6.4 days. ! The additional time required to reach the licensing design basis spent fuel pool temperature (180*F) is 3.8 days. Therefore, the LOOP heat-up ! analysis concludes that a minimum of approximately 10 days are available, if no actions are taken, before the spent fuel pool temperature could reach 180*F. l F Amendment 1 14.1-4
I l
14.2 REFERENCES
- 1. License Amendment No. 119, dated March 19, 1992, Permanently Defueled Technical Specifications
- 2. Safety Analysis and No Significant Hazards Consideration (Log No.
1091, Revision 3) for Proposed Amendment 182, Revision 3, Permanently Defueled Technical Specifications
- 3. License Amendment No. 117, dated March 17, 1992, Possession-Only License
- 4. J. F. Stolz (NRC) to J. J. Mattimoe (SMUD) letter dated January 20, 1984, Rancho Seco Licenre Amendment No. 52
- 5. SMUD Calculation Z-SFC-M2560, " Spent Fuel Pool Heat-Up During LOOP with Pool at 23.25 feet."
- 6. SMUD Calculation Z-SFC-M2557, " Spent Fuel Decay Heat Based on ORIGEN2 Computer Code."
l
- 7. SMUD Calculation Z-SFC-N004 9, Revision 3, " Maximum Predicted Whole Body and Skin Doses and Dose Rates at the Site Boundary from Postulated Accidents During Plant Shutdown."
l l l l l l l l I I Amendment 2 14.2-1}}