ML20155D480
| ML20155D480 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 10/27/1998 |
| From: | SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | |
| Shared Package | |
| ML20155D447 | List: |
| References | |
| NUDOCS 9811030227 | |
| Download: ML20155D480 (192) | |
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DEFUELED SAFETY ANALYSIS REPORT TABLE OP CONI'lWIS CHAPTER 5. STRUCTURES AND CONTAINMENT SYSTEM (Continued)
Section Title Page 5.4.2.1.5 Thermal Stresses 5.4-2 5.4.2.2 Desien Criteria 5.4-3 5.4.2.3 Structural Desien 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 5.5-2 l 5.5.5 NUCLEAR SERVICES ELECTRICAL BUILDING 5.5-2 l 5.5.5.1 General Description 5.5-2 l 5.5.6 TRAINING AND RECORDS BUILDING 5.5-3 l 5.5.7 INTERIM ON SITE STORAGE BUILDING FOR LOW 5.5-3 l LEVEL RADWASTE '
5.5.7.1 General Description 5.5-3 l i 5.5.7.2 Desien Basis 5.5-4 l 5.5.7.2.1 Codes, Standards, and Regulatory Requirements 5.5-5 l 5.5.7.3 Material Reauirements 5.5-6 l 5.5.7,4 Structural Reauirements 5.5-6 l 5.5.8 SOLIDIFICATION BUILDING 5.5-7 5.5.8.1 General Descriotion 5.5-7 l 5.5.8.2 Design Ba. sis 5.5-8 l 5.6 MATERIALS AND CONSTRUCTION PRACTICES 5.6 1 5.6.1 CONSTRUCTION ORGANIZATION 5.6-1 Amendment 3 9911030227 901027 F y PDR ADOCK 05000312 y W PDR g
DEFUELED SAITrY ANALYSIS REPOItT TABLE OF COVfEVIS CHAPTER 5. STRUCTURES AND CONTAINMENT SYSTEM (Continued)
Section Title Page 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 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
- o.3.3
. Steel Pre-stressinn Tendons 5.6 13 5.6.3.4 Liner Plate and Penetration Sleeves 5.6 14 5.6.3.5 Penetrations 5.6-14 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 xii
DEITELED SAFI7PY ANALYSIS REPOlff TABLE OF CONTENIS CHAITER 5. STRUCTURES AND CONTAINMENT SYSTEM (Continued)
Section Title Eage 5.6.3.7 - Welder Oualifications and Inspection of Field Weldine 5.6-15 5.6.3.7.1- WeIding 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 CHAPTER 6. SAFETY FEATURES 6.1 GENERAL 6.1-1 6.2 SAFETY FEATURES SYSTEMS LEAKAGE AND RADIATION 6.2-1 CONSIDERATIONS 6.3 EQST LOSS-OF-COOLANT ACCIDENT HYDROGEN CONTROL 6.3-1
6.4 REFERENCES
6.4- l CHAfiER 7. INSTRUMENTATION AND CONTROL Section Title Page 7.1 PROTECTION SYSTEMS 7.1-1 7.2 REGULATION SYSTEMS 7.2 1 7.3 INSTRUMENTATION 7.3- 1
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DEFUELED SAFETPY ANALYSIS REPOIrr TABLE OF CONTEVIE CHAPTER 7. INSTRUMENTATION AND CONTROL (Continued) 7.4 OPER ATING CONTROL STATIONS 7.4- 1 7.4.1 GENERAL LAYOUT 7.4- 1 7.4.2 INFORMATION DISPLAY AND CONTROL FUNCTION 7.4-2 7.4.2.1 Console and Panel Lav-out 7.4-6 l 7.4.3
SUMMARY
OF ALARMS 7.4-7 l 7.4.4 COMMUNICATION 7.4-7 l 7.4.5 OCCUPANCY 7.4-8 l 7.4.6 AUXILIARY CONTROL STATIONS 7.4-9 !
7.4.7 SAFETY CONSIDERATIONS 7.4-9 l 7.4.8 SYSTEM EVALUATION 7.4-9 l 7.4.8.1 Control Room Availability 7.4-9 l
7.5 REFERENCES
7.5-1 CHAPTER 8. ELECTRICALSYSTEMS 8.1 DESIGN BASES 8.1-1 8.2 ELECTRICAL SYSTEM DESIGN 8.21 8.2.1 NETWORK INTERCONNECTIONS 8.2-1 8.2.1.1 Single Line Diaeram 8.2-1 8.2.1.2 Reliability Considerations 8.2-1 y 8.2.2 STATION DISTRIBUTION SYSTEM 8.2-2 8.2.2.1 Auxiliary Transformers 8.2 2 8.2.2.2 6900 Volt Auxiliary System 8.2-3 8.2.2.3 4160-Volt Auxiliary System 8.2-3 8.2.2.4 480-Volt Auxiliary System 8.2-3 Amendment 3 xiv
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DEFL'ELED SAFETI'Y ANALYSIS REPOIYr TABLE OF CONTENIS CHAPTER 8. ELECTRICAL SYSTFMS (Continued)
Section Title Eage 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-Volt a-c Unintermotible Power Sunolv System 8.2-4 8.2.2.8 Lighting 8.2-5 1 8.2.2.9 Evaluation of the Physical Layout of Electrical Distribution System 8.2-5 Eauioment 8.
2.3 DESCRIPTION
OF POWER SOURCES 8.2-7 8.2.3.1 Off-site Power 8.2-7 8.2.3.2 On-site Power 8.2-7 8.2.3.3 220/230-kV Switchyard Control Power 8.2-7 8.3 TESTS AND INSPECTIONS 8.3-1 8.3.1 IN-SERVICE TESTS AND INSPECTIONS 8.3-1 l
8.4 REFERENCES
8.4- 1 CHAITER 9. AUXILIARY AND EMERGENCY SYSTEM _S 9.1 GENERAL 9.1-1 I 9.2 COOLING WATER SYSTEMS 9.2-1 9.2.1 CONDENSER CIRCULATING WATER SYSTEM 9.2-1 l 9.2.1.1 System Description 9.2 1 9.2.1.2 Design Data 9.2-1 l 9.2.2 PLANT COOLING WATER SYSTEM 9.2-2 9.2.2.1 System Description 9.2 2 L 9.2.2.2 Desien Data 9.2-2 1
Amendment 3 xv
l-DEFUELED SAFETY ANALYSIS REPOIYr TAllLE OF CONTENIS CHAPTER 9. AUXILIARY AND EMERGENCY SYSTEMS (Continued)
Page Section Title 9.2.3 COMPONENT AND TURBINE PLANT COOLING WATER 9.2-2 SYSTEM System Descriotion 9.2-2 l 9.2.3.1 9.2.3.2 Desien Data 9.2-4 l 9.3 DECAY HEAT REMOVAL SYSTEM 9.3-1 9.3.1 Borated Water Storage Tank 9.3-1 SPENT FUEL COOLING SYSTEM 9.4- 1 9.4 9.4.1 DESIGN BASES 9.4-1 9.4.2 SYSTEM DESCRIPTION 9.4-4 9.4.2.1 Codes and Standards 9.4-4 l 9.4.2.2 Material Comoatibility 9.4-4 l 9.4.2.3 Component Design 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 9.4-5 9.4.2.4. Leakaec Considerations 9.4-5 9.4.2.5 Failure Considerations 9.4-6 9.4.2.6' Operatine Conditions 9.4-6 9.5 STATION VENTILATION SYSTEMS 9.5 1 9.5.1 DESIGN B ASES 9.5-l 9.5.1.1 Reactor Buildine 9.5-1 9.5.1.1.1 General Conditions 9.5-1 Amendment 3 l
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DEFUELED SAFETY ANALYSIS REPORf TABLE OF CONTENTS CHAMER 9. AUXILIARY AND EMERGENCY SYSTEMS (Continued) i'
- Section Title Eage 1
i 9.5.1.1.2 . Sizing .9.5 1 9.5.1.2 Auxiliary Building 9.5-1 9.5.1.3 Nuclear Service Electrical Buildine (NSEB) 9.5 ' I l
9.5.2 SYSTEM DESCRIPTION 9.5 ' i 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 l 9.5.2.1.3 Codes, Standards, Tests 9.5-7 9.5.2.2 Auxiliary Building 9.5-7 9.5.2.2.1 General 9.5-7 9.5.2.2.2 Control Room and Technical Suppen Center 9.5 7 9.5.2.2.3 Radiochemical and Service Areas 9.5-8 l )
9.5.2.2.4 Radwaste and Fuel Storage Areas 9.5-8 l-9.5.2.2.5 Electrical Equipment, Switchgear, and AC/DC Panel Rooms 9.5-9 l 9.5.2.2.6 Coolant and Miscellaneous Waste Tanks 9.5-9 l L
9.5.2.2.7 Emergency Pump Rooms 9.5 10 l l.
9.5.2.2.8 Auxiliary Building Exhaust Air Filtration System 9.5 10 l 9.5.2.2.9 Communication Room 9.5 10 l 9.5.2.2.10 Chilled Water System 9.5 10 l 9.5.2.3 Nuclear Service Electrical Buildine (NSEB) 9.5-12 l 9.5.2.4 Battery Building 9.5-12 l 9.5.2.5 Interim On-site Storage Buildine (IOSB) 9.5-12 l Amendment 3 xvii i
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f DEFUELED SAFETY ANALYSIS REPOllT TABLE OF CONTENlE 4
CHAPTER 9. AUXILIARY AND EMERGENCY SYSTEMS (Continued)
Sectioq Title hge 9.5.2.6 Switchyard Control Buildine 9.5 13 9.5.2,7 Codes. Standards. and Tests 9.5-13 9.6 FUEL HANDLING SYSTEM 9.6-1 9.6.1 ' DESIGN BASES 9.6 1 9.6.1.1 General System Function 9.6-1 9.6.1.2 Soent Fuel Storace Pool 9.6 1 9.6.1.3 Soent Fuel Pool Water Chemistry 9.6-2 9.6.1.4 Fuel Transfer Tube 9.6-2 9.6.1.5 Fuel Handline Eauipment 9.6-2
-9.6.2 SYSTEM DESCRIPTION AND EVALUATION 9.6-2 l 9.6.2.1 Handling Soent Fuel A'ssemblies 9.6-2 l 9.6.2.2 Handline and Loadine Spent Fuel Casks 9.6-3 9.6.2.3 Crane Use in Fuel Handline 9.6-4 9.6.2.3.1 Design 9.6-4 9.6.2.3.2 Evaluation 9.6-5 9.6.2.4 Safety Provisions 9.6-7 l 9.7 OTHER AUXILIARY SYSTEMS 9.7- 1 9.7.1 FIRE PROTECTION SYSTEM 9.7-1 9.7.1.1 Design Bases 9.7-1 9.8 PLANT COMPRESSED SERVICE GAS SYSTEM 9.8 1
9.9 REFERENCES
9.9-1
- 10. STEAM AND POWER CONVERSION SYSTEM 10.1 DESIGN B ASES Amendment 3 xviii
DEFUELED SAFI7fY ANALYSIS REPORT TABLE OF CO.VrE.VIS CHAPTER 11. RADIOACTIVE WASTE AND RADIATION FROTECTION Section Title Page i 1.1 SOURCE TERM 11.1-1 11.1.1. RADIONUCLIDE INVENTORY l1.1-1
'11.1.1.1 Spent Fuel Assemblies 11.1-1
-11.1.1.2 Reactor Vessel and Intemals and Concrete Primary Shield i1.1-2 11.1.1.3 Plant Systems 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 i1.2.1.1 Functions 11.2-1 11.2.1.2 System Description 11.2-1 l 11.2.2 MISCELLANEOUS LIQUID RADWASTE SYSTEM 11.2-2 11.2.2.1 Funetion 11.2-2 11.2.2.2 System Description 11.2-2 11.2.3 WASTE WATER DISPOSAL 11.2-11 !
11.2.3.1 . Plant Effluent i1,2-11 !
!1.2.3.2 Normal Radioactive Discharge 11.2-12 l 11.2.3.3 Off-Normal Radioactive Discharge 11.2 14 11.2.3.4 Non radioactive Waste Water 11.2 l 11.2.4 OPERATION. TESTING, AND INSPECTION 11.2-14 l 11.2.5. SYSTEM EVALUATION 11.2-15 11.2.6 PROCESSING WET RADIOACTIVE WASTES INTO SOLID 11.2-16 l RADIOACTIVE WASTE 11.2.6.1 Solidification and De-waterine of Wet Radioactive Wastes 11.2 16 l 11.2.6.2 Drvine of Wet Radioactive Wastes 11.2-18 l Amendment 3 xix
DEFUELED SAFETY ANALYSIS REPOlYr TABLE OF CO.NTE. V IS
- CHAPTER 11. RADIOACTIVE WASTE AND RADIATION PROTECTION (Continued)'
e S_ec.tioa Tille Page 11.2.6.2.1 Design Basis 11.2-18 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, TESTING, ANDINSPECTION 11.3-2 11.3.5 RADIOACTIVE RELEASES 11.3-3 11.3.5.1 Pathways 11.3 3 11.3.5.2 Secondary Plant Contamination 11.3-3 ,
11.3.5.3 Interim On-site Storage Building (IOSB) 11.3-3 11.3.6 METHOD OF ASSESSMENT 11.3-3 l 11.3.6.1 Plume Exposure (Nobie Gases) 11.3-3 l 11.3.6.2 Food Pathway 11.3-4 11.3.7 EVALUATION OF WASTE DISCHARGE I1.3-4 l 11.4 SOLID WASTE MANAGEMENT SYSTEM 11.4-1 .
I1.4.1 DESIGN BASIS 11.4-1 11.4.1.1 Recons 11.4-2 11.4.2 SYSTEM DESCRINION 11.4-2 11.4.2.1 Dry Solid Waste Disposal System / Process 11.4-2 11.4.2.2 Concentrated Liauid Waste Disposal System / Process Il.4-2 l
!!.4.2.3 Snent Resin Disposal System / Process 11.4-3 11.4.2.4 Filter Disposal Process 11.4-3 Amendment 3 n
DEFUELED SAFWIT ANALYSIS REPOIrl'
'I'.WLE OF CONTENIE CHANER I1.- RADIOACTIVE WASTE AND RADIATION PROTECTION (Continued)
Section Title Pace 11.4.2.5 Solid Radwaste Storage 11,4-3 11.5 RADIOACTIVE WASTE. EFFLUENT CONTROL. AND 11.5-1 ENVIRONMENTAL MONITORING PROGRAMS 11.5.1 DESIGN BASIS 11.5-1 11.5.2 OFF-SITE DOSE CALCULATION MANIJAL (ODCM) 11.5-1 11.5.2.1 Liauid Discharge Pathway 11.5-1 11.5.2.2 Gaseous 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-operational REMP 11.5-2 l 11.5.4.1.1 Pre-operational Exposure Estimation 11.5-4 l
11.5.4.2 Off-site Post-operational REMP 11.5-8 11.5.4.2.1 Post-operational REMP Sampling Frequency 11.5-9 l 11.5.4.2.2 REMP Sample Types i1.5-10 11.5.4.2.3 REMP Sampic Statistical Analysis 11.5-10 l 11.5.4.3 Effluent and Waste Disposal Environmental Reports 11.5-10 11.6 ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES 11.6-1 ARE AS LOW AS iS REASONABLY ACHIEVABLE (ALARA) 11.6.1 'ALARA POLICY CONSIDERATIONS 11.6 1 11.6.2 ALARA DESIGN CONSIDERATIONS 11.6-2 l 11.6.3 ALARA OPERATIONAL CONSIDERATIONS 11.6-3 l
l 11.7 R ADIATION SOURCES 11.7-1 11.7.1 CONTAINED SOURCES 11.7-1 Amendment 3 l-ni I
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DEFUELED SAFICTY AXiLYSIS REPOITI' TABLE OF CONTEVIS CHAPTER 11. RADIOACTIVE WASTE AND RADIATION PROTECTION (Continued)
Page Section Title 11.8 R ADIATION PROTECTION DESIGN FEATURES 11.8-1 11.8.1 FACILITY DESIGN FEATURES 11.8-1 11.8.2 SHIELDING 11.8-1 11.8.2.1 Desien Criteria 11.8-1 11.8.2.2 Radiation Zone Classifications 11.8-2 11,8.2.3 Description of Shielding 11.8-2 11.8.2.3.1 Primary Shield 11.8-2 11.8.2.3.2 Secondary Shield 11.8-2 l 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 1l.8.2.3.6 Spent Fuel Shielding i1.8-3
!!.8.2.4 Shieldine Materials 11.8-3 1l.8.3 VENTILATION 11.8-5 11.8.4 RADIATION MONITORING SYSTEM 11.8-5 11.8.4.1 Desien Criteria 11.8-5 l1.8.4.2 Svstem Descriotion i1.8-5 11.8.4.2.1 Area Radiation Monitors 11.8-6 l 11.8.4.2.2 Process / Effluent Radiation Monitors 11.8-6 l 11.9 DOSE ASSESSMENT I1,9 1 I
!1.9.1 PERSONNEL MONITORING 11.9.I 11.9.2 PERSONNEL EXPOSURE RECORD SYSTEM i1.91 l Amendment 3 l .uii l .
DEFUELED SAFIITY ANALYSIS REPOIrr l 1
TABLE OF CONTENIE CHAPTER 11. RADIOACTIVE WASTE AND RADIATION PROTECTION (Continued) l Section Title Page 11.9.3' . MEDICAL EXAMINATION PROGRAM i1.9-2 11.10 RADIATION PROTECTION PROGRAM 11.10 1 11.10.I' ' ORGANIZATION 11.10-1 11.10.2 EQUIPMENT, INSTRUMENTATION, AND FACILITIES 11.10-3 I 11.10.2.1 Personnel Protective Eauipment 11.10-3 11.10.2.2 Radiation Protection Instrumentation 11.10-3 11.10.2.3 Eacilities ti,io.4 I 1
11.10.3 RADIATION PROTECTION PROCEDURES 11.10-5
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l 11.10.3.1 ' Procedures - 11.10-5 l 1
! !.10.3.2 Radiation Work Permit Procedure 11.10-6 l
11.10.4 PERIODIC PERSONNEL EXPOSURE REPORTING 11.10-6 11.11 REFERENCES 11.11-1 CHAfrTER 12. CONDUCT OF OPERATIONS 12.1.1 NUCLEAR ORGANIZATION 12.1-1 12.1.2 PLANT PERSONNEL RESPONSIBILITIES AND AUTHORITIES 12.1 ~1 12.1.2.1 Plant Manager 12.1-1 12.1.2.2 Operatine Shift Crews 12.1-5 l 12.1.2.3 Succession of Responsibility 12.1-7 l 12.1.3 QUALIFICATIONS OF NUCLEAR PLANT PERSONNEL 12.1 7 12.2 PERSONNEL TR AINING 12.2-1 12.2.1 TRAINING PROGRAMS 12.2-1 Amendment 3 niii l
DEFUELEDSAFITY ANALYSIS REPORT TAILLE OF CO.VI'E. VIE CHANER 12. CONDUCT OF OPERATIONS (Continued)
Section Title Pace 12.2.2 CERTIFIED FUEL HANDLER TRAINING PROGRAMS 12.2-2 12.2.2.1. Structure of Initial Certification Program 12.2-2 l 12.2.2.2 Evaluation 12.2-3 12.2.3 STRUCTURE OF CERTIFIED FUEL HANDLER PROFICIENCY 12.2 4 PROGRAM 12.2.3.1 Evaluation 12.2 5 12.2.4 LICENSING TRAINING PROGRAM 12.2-5 12.2.5 MAINTENANCE TRAINING PROGRAM 12.2-5 l 12.2.5.1 Initial Trainine 12.2-6 l I
12.2.5.2 Continuing Training 12.2-6 12.2.6 SITE SUPPORT TRAINING PROGRAMS 12.2-6 l 12.2.6.1 Initial Trainine 12.2-7 12.2.6.2 Continuine Training 12.2-7 12.3 EMERGENCY PLANNING 12.3-1 12.4 REVIEW AND AUDIT OFOPERATIONS 12.4 1 12.5 ELANT PROCEDURES AND PROCESS STANDARDS 12.5-1 12.5.1 PROCEDURES 12.5-1 i
12.5.1.1 Conformance with Safety Guide 33 12.5-1 12.5.1.2 Preparation of Procedures 12.5-1 12.5.1.2.1 Procedure Changes 12.5 2 12.5.1.3 Conduct of Operations 12.5-2 12.5.1.3.1 Shut-down Control Room Operator Authority 12.5-3 Amendment 3 niv
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l- DEFUIsLED SAFliTIT ANALYSIS REPOIrf TABLE OF COVfEVIS
! CHAIYTER 12. CONDUCT OF OPER ATIONS (Continued) l Section Title Pace 12.5.1.3.2 Certified Fuel Handler Authority 12.5-3
- 12.5.1.3.3 Activities Affecting Plant Operations or Indications During the PDM 12.5-3 12.5.1.3.4 Manipulation of Facility Controls 12.5-3 l 12.5.1.3.5 Responsibility for Fuel Handling Operations 12.5-3 l 12.5.1.3.6 Relief of Duties 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 12.5-4 l 12.5.2.1 Onerating Procedures 12.5-4. l 12.5.2.1.1 System Procedures 12.5-5 12.5.2.1.2 Special Test Procedures 12.5-5 12.5.2.1.3 Annunciator Alarm Response Procedures 12.5-5 12.5.2.1.4 Casualty Procedures 12.5-6 12.5.2.2 Other Procedures 12.5-6 12.5.2.2.1 Maintenance Procedures 12.5-6 12.5.2.2.2 Instrument and Control (I&C) Procedures 12.5-6 l 12.5.2.2.3 Surveillance Procedures 12.5-6 l 12.5.2.2.4 Chemistry Procedures 12.5-7 12.5.2.2.5 Radioactive Waste Management Procedures 12.5 7 l 12.5.2.2.6 Radiation Protection Procedures 12.5-7 l 12.5.2.2.7 Security Plan 12.5-7 l Amendment 3
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DEFTELED SAFliTrY ANALYSIS REPOIrr TADLE OF CONTENIS CHAPTER 12. CONDUCT OF OPERATIONS (Continued)
Page Section . Title 12.5.2.2.8 Emergency Plan and implementing Procedures 12.5-7 l Fire Protection Procedures 12.5-8 l 12.5.2.2.9 Quality Assurance 12.5-8 l 12.5.2.2.10 12.5.2.2.11 Certified Fuel Handler and Non-Certified Operator Training Programs 12.5-8 12.5.2.2.12 Decommissioning License Basis Documents 12.5-8 12.5.3 PROCESS STANDARDS 12.5-8 12.6 INDUSTRIAL SECURITY 12.6-1 12.7 RECORDS 12.7-1 12.7.1 OPERATINO RECORDS 12.7 1 i
12.7.2 ADMINISTRATIVE RECORDS 12.7-1 12.7.3 MAINTENANCE RECORDS 12.7-1 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. INITIALTESTS AND OPERATIONS
13.1 INTRODUCTION
13.1-1 CHAPTER 14. SAFETY ANALYSIS 14.1 ACCIDENTS CONSIDERED CREDIBLE IN THE PERM ANENTLY 14.1-1 DEFUELED MODE (PDM)
'14.1.1 FUEL HANDI.INO ACCIDENT 14.1-1 l 14.1.1.1 Analysis and Results 14.1 1 Amendment 3 nvi
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! f DEFUELED SAFIfrY ANALYSIS REPOIYl' F
TABLE OF CONTENIS CHAPTER 14. SAFETY ANALYSIS (Continued) 14.1.2 LOSS OF OFF-SITE (A-C) POWER 14.1-2 i
14.2 DECOMMISSIONING ACCIDENT ANALYSIS 14.2-1 i
14.2.1 PREPARATION FOR SAFSTOR 14.2-1 l l
14.2.2 ACCIDENTS DURING DECON 14.2 2 l
14.3 - REFERENCES 14.3 1 i
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! TABLE OF CONTENTS Section Title Eage
- 5. STRUCTURES AND CONTAINMENT SYSTEM 5.1 1 5.1 GENERAL 5.1-1 5.1.1 CLASSES OF STRUCTURES 5.1-1 i 5.1.1.1 ClassI 5.1-1 5.1.1.2 Class II 5.1-1 5.1.1.3 Class III 5.1-2 5.1.2 DESIGN LOADS AND STRUCTURAL BEHAVIOR 5.1-2 5.1.2.1 Class 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 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 l 5.1.2.1.8 Seismic Design of Equipment, Structures, and Supports 5.1 10 5.1.2.1.9 Design of Foundations and Sub grade Walls 5.1-12 5.1.2.1.10 Buried Tunnels, Piping, and Cables 5.1-13 5.1.2.2 Class 11 Structures 5.1 l4 5.1.2.3 Class III Structures 5.1-14 5.1.3 GOVERNING CODES AND SPECIFICATIONS 5.1-15
( 5.1.4 LOAD COMBINATIONS CRITERIA AND STRUCTURAL 5.1-15 ANALYSIS Amendment 2 5-1
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r TABLE OF CONTENTS (Continued)
Title Page Section 5.1.4.1 At Design Loads 5.1-17 5.1.4.2 At Factored Loads 5.1-17 5.2- 1 5.2 REACTOR BUILDING CONTAINMENT STRUCTURE 5.2-1 5.2.1 5.2.2 INTERIOR CONTAINMENT STRUCTURE 5.2-1 AUXILI ARY BUILDING 5.3- 1 5.3 GENERAL DESCRIPTION 5.3- 1 5.3.1 5.3.2 DESIGN BASES 5.3-1 5.3.2.1 Design Loads .
5.3.2.1.1 Dead Loads 5.31 5.3.2.1.2 Live Loads 5.3-2 I 5.3.2.1.3 Eanhquake 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 Design Criteria 5.3-3 5.3.2.3 Structural Design Analvsis 5.3-3 FUEL STOR AGE BUILDING 5.4- 1 5.4 5.4.1 GENERAL DESCRIPTION 5.4- 1 5.4.2 DESIGN BASES 5.4-1 5.4.2.1 Desien Loads 5.4- 1 5.4.2.1.1 Dead Loads 5.4- 1 5.4.2.1.2 Live Loads 5.4-2 l
Amendment 2 1 5-ii
TABLE OF CONTENTS (Continued)
Section Title Eage 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 Suesses 5.4-2 5.4.2.2 Desien 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 5.5-2 l 5.5.5 NUCLEAR SERVICES ELECTRICAL BUILDING 5.5-2 l 5.5.5.1 General Descrintion 5.5-2 l 5.5.6 TRAINING AND RECORDS BUILDING 5.5-3 l 5.5.7 INTERIM ON SITE STORAGE BUILDING FOR LOW LEVEL 5.5-3 l RADWASTE 5.5.7.1 General Descriotion 5.5-3 l 5.57.2 Desien Basis 5.5-4 l 5.5.7.2.1 Codes, Standards, and Regulatory Requirements 5.5-5 l 5.5.7.3 Material Reouirements 5.5-6 l 5.5.7.4 Structural Reauirements 5.5-6 l 5.5.8 SOLIDIFICATION BUILDING 5.5 7 l 5.5.8.1 General Descriotion 5.5-7 l 5.5.8.2 Desien Basis 5.5-8 Amendment 3 5-iii
TABLE OF CONTENTS (Continued)
Secti9.0 Title Eage 5.6 MATERIALS 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 i 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-l l 5.6.3.2.5 Inspection and Testing of Cadweld Splices 5.6- l 1 5.6.3.3 Steel Pre-stressine Tendons 5.6-13 1 5.6.3.4 Liner Plate and Penetration Sleeves 5.6-14 5.6.3.5 Penetrations 5.6-14 5.6.3.6 Structural and Miscellaneous Steel 5.6-14 5.6.3.6.1 Materials 5.6-14 1
5-iv '
TABLE OF CONTENTS (Continued)
Section Title Page 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 Oualifications and Inspection 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 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 l 5.6.3.7.9 Records 5.6-19 5.7 SEISMIC INSTRUMENTATION 5.7-1
5.8 REFERENCES
5.8-1 i
i I
5-v
LIST OF TABLES Table Tills Eagg a
- 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 Stmetures 5.1-9 5.6-1 ,
Aggregate User Tests 5.6-3 t
a 1
4 4
4 4
3 4
5-vi
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5.4.2.1.2- - Live Loads The following live loads are considered in the design:
A. Roofloads of 20 pounds per square foot B. ~ Hydrostatic loads from the spent fuel pool filled with vs.< r 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 Fuel Storage Building design.
5.4.2.1.4 - 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 recommendation of the UBC, 4 whichever is greater. j l
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.
i During the PDM, abnormal thermal stresses are not a concern for the design of the spent fuel pool. The spent fuel pool heat-up analyses indicate the maximum temperature the pool can reach with a loss of primary spent fuel pool cooling is approximately 147 F. With the functional loss of spent fuel pool cooling and no make-up water, the heat-up analyses indicate the spent fuel pool temperature could reach 212 F. But, sufficient time is available to re-establish spent fuel cooling and establish make-up water to ensure the spent fuel pool temperature does not exceed 180 F (i.e., the Technical Specification limit). The spent fuel pool can withstand a water temperature of 212 F.
Amendment 3 5.4-2
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l 5.4.2.2 Desian Criteria l l
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 leakage to the radioactive waste treatment systems in the Auxiliary Building. A leak-chase l monitoring system keeps track of the leak rate volume from the spent fuel pool.
Except as noted in these criteria, the ACI 318-63 and AISC, Sixth Edition, design methods and allowable stresses were used for the design of reinforced concrete and steel, respectively.
The strength of the structure at working stress and over-all yielding was compared to various loading combinations to ensure safety. The structure is designed to meet the performance and strength requirements under the following conditions:
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 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 of crane hook travel, cask bail, and long sling. The floor of the spent fuel loading pit is depressed 4t6" below the main floor of the pool. The pit is 10t6" by 12'-6" in plan and 6 feet thick. It is continuous 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 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, and water of 195 kips. It had a base diameter of 6'-8" and a height of approximately 17 feet.
The thickness of the concrete slab is adequate for the hypothetical impact evaluated in the following analysis.
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, smooth sided vessel with blunt, plane ends. No consideration was made for the nigosity 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.
5.4-3.
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i 5.4.2.2 Desien Criteria (Continued)
To simplify the analysis of the effect on the spent fuel pool base slab under this cask impact l
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 i 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 13/16-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.
I 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 1 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 Design Analysis
'F 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 limiting the strain on the stainless steel pool liner.
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 cask lifted over spent fuel.
The superstructure of the fuel storage building was analyzed for loads from a 200 mph wind.
5.4-4 i
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INTENTIONALLY LEFT BLANK Amendment 3
l 5.5 OTHER STRUCTURES j 1
5.5.1 NUCLEAR SERVICE SPRAY PONDS AND PIPE LINES The Nuclear Service Cooling Water System is not required to function and is not part of the l licensing design basis for operation of Rancho Seco during the PDM. Therefore, a description of l the Nuclear Service spray ponds and pipe lines are not included in the DSAR. i 5.5.2 TURBINE BUILDING The Turbine Building was designed as a Class 11 structure consisting of two parts: (1) the turbine pedestal and (2) the support structure for power conversion system equipment. The pedestal is a massive reinforced concrete structure, and the suppon structure is a braced steel frame with a concrete deck and braced steel floor system with grating at the mezzanine level.
The turbine building helps support the turbine gantry crane during normal operations. The crane load bearing members are Class I and the crane girders and columns are capable of withstanding loads imposed from a design base earthquake, with the crane handling the analyzed spent fuel cask.
5.5.3 COOLING TOWERS The cooling towers have no Class I characteristics and are designed by the supplier to withstand a wind pressure based upon Figure 1(b) of ASCE Paper 3269 " Wind Forces on Structures" using the fastest wind speed for a 100-year occurrence period, but not less than 30 lb/ft2 . The towers have the capability to resist both the dynamic and static effects of the wind loading. Wind loads rather than Class II seismic requirements govern the cooling tower design. An independent cheek of the design indicates that the towers can withstand the operating base earthquake.
The physical layout of the facility and the potential modes of failure for the type of hyperbolic cooling towers employed at Rancho Seco preclude damage to any critical components resulting from a tower collapse. The towers are thin-shell structures of reinforced concrete. They are approximately 300 feet in diameter at the base, narrowing to 169 feet at the throat,195 feet across at the top, and 425 feet tall. The wall thickness is 26 inches at the bottom tapering to a minimum of seven inches at the throat.
The west tower is remote from any critical plant components. The diesel fuel oil storage tank, which is nearby,is used only to store fuel for an auxiliary boiler and has no safety function. In the vicinity of the east tower are the borated water storage tank, the demineralized reactor coolant storage tank and the condensate storage tank. The shortest distance from any of these tanks to the outer edge of the east tower is approximately 160 feet. Even funher away is the Fuel Storage Building. This places these tanks and the Fuel Storage Building at a safe distance in the unlikely event the east tower should collapse. based on the evaluation below.
Amendment 3 S.5-1
5.5.3 COOLING TOWERS (Continued)
Critical stress points occur at the throat and at thejuncture of the shell with the ring girder. The towers are frangible once failure occurs and will break into small pieces during collapse. These structures could not overturn as a unit under wind or seismic loading. These statements are strongly supported by experiences with cooling tower failures at Ferrybridge, England. The English towers, with the exception of having thinner shells and less reinforcing, are virtually identical to the towers at Rancho Seco. The Ferrybridge failures resulted from high wind loads complicated by vortexing. Seismic loads would result in a similar stress distribution. The following excerpts from " Report of the Committee of Inquiry into Collapse of Cooling Towers at Ferrybridge,1 November 1965" are significant:
A. In all three collapses the tower debris fell almost exclusively within the area of the pond. (Photographs show no debris beyond the pond sill.)
B. Failures resulted in complete fragmentation.
C. Failures initiated just below the throat.
D. Breaks were due to vertical tension with bending.
With particular reference to the similarity of the Ferrybridge towers and the Rancho Seco towers and to statements I and 2 of the report,in the unlikely event a tower at Rancho Seco should fail, no damage to important plant components will result.
5.5.4 STORAGE RESERVOIR A storage reservoir (known as Rancho Seco Lake) can be used as an attemate supply of make-up l
, and dilntion water for the plant if the Folsom South Canal is out of service. The reservoir is
- located approximately 2 miles southeast of the plant. Water from this reservoir is not required for safe operation of Rancho Seco during the PDM.
The dam is under the jurisdiction of the State of Califomia, Division of Dam Safety, and as such it was designed and constructed to standards established by the State of California, which include consideration for earthquake.
The effects on the plant of a dam failure or other sudden release of water were investigated. An instantaneous break 50 feet wide and the full height of the dam occurring simultaneously with the peak flow from a design storm would not flood the plant site.
5.5.5 NUCLEAR SERVICES ELECTRICAL BUILDING 5.5.5.1 General Description The Nuclear Services Electrical Building (NSEB) is a three-story reinforced concrete structure that is 62 feet wide,74 feet long and 60 feet high with a 3 foot parapet at the roof perimeter. It contains two 10 feet square cable shafts that extend the full height of the building terminating at
! 5.5-2 l
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5.5.5.1 General Description (Continued) the bottom into a two compartment below grade tunnel to the Auxiliary Building cable shafts. At the 40 foot elevation a passageway extends to the Technical Support Center in the Auxiliary Building. The NSEB houses electrical buses and inverters that are used to supply electrical power to required loads in the Auxiliary Building. Also, the NSEB normal HVAC equipment functions to support this functional electrical equipment. The remaining equipment housed in the NSEB is not required to function during the PDM.
The NSEB was built as a shear wall structure. The floor system consists of concrete slabs on steel decking and steel beams and steel girders. The NSEB was designed to withstand all known credible abnormal environmental loads i.e., seismic, tornado winds, and tomado-generated missiles. The NSEB was designed as a Seismic Category I structure.
5.5.6 TRAINING AND RECORDS BUILDING The Training and Records (T & R) Building is a five story structure that is 62 feet wide,157 feet long, and 65 feet high that was designed as a Class II structure. Because the T & R Building is close to a Scismic Category I structure (the Auxiliary Building), a seismic analysis was performed on the initial design to assure that the building would not impair Seismic Category I systems due to structural failure during the occurrence of a seismic event. The Training and Records Building is primarily a reinforced concrete structure consisting of two-way slabs, floor beams, columns, and shear walls. The T & R Building is not required in the PDM.
i !
5.5.7 INTERIM ON-SITE STORAGE BUILDING FOR LOW LEVEL RADWASTE l 5.5.7.1 General Description The purpose of the Interim On site Storage Building (IOSB)is to house packaged low level radioactive waste (radwaste)in a retrievable mode. The structure is modular in design and allows expansion for udded storage volume.
The IOSB radwaste storage area was designed with a holding capacity equivalent to
" approximately 2-1/2 years of radwaste estimated to be produced during normal nuclear power reactor operations. IOSB features and systems, such as sump and drains, electrical, fire protection, etc., were sized to suppon expansion of the IOSB to a facility that could store up to 10 years of generated radwaste. Radwaste production during the PDM will be significantly less than the amount estimated for normal nuclear power reactor operations.
The IOSB is designed to store radwaste in two basic configurations, dependent on activity level.
High activity waste can be stored in a shielded, covered cell arrangement designed to accommodate a range of waste containers from 55 gallon drums to 300-cubic foot disposable liners. The cells have individual shield covers, with cell cover and waste container handling
- accomplished by an overhead remotely operated bridge crane system.
The low activity radwaste is stored in a shielded open floor warehouse arrangement designed to accommodate a range of waste containers from 55 gallon drums to 120-cubic-foot metal bins.
5.5-3
5.5.7.1 . General Description (Continued)
Waste that is less than 10 mR/hr contact dose represents the majority of the total quantity of low activity waste. This waste is stored on the outside face of the stacking volume. The remaining low activity waste is stacked in the center of the stacking volume and is partially shielded by the other radwaste. Stacking height for the containers is not to exceed 18 feet. Materials for the open warehouse are handled by a manually operated forklift truck. At no time will the total container surface dose rate source term in the IOSB exceed 146,300 mrad /hr.
The IOSB storage areas and radwaste material handling systems are designed to provide easy container retrievability. The ability to quickly and remotely remove and examine a container for
! integrity was considered in the design.
5.5.7.2 Design Basis Loads are applied in accordance with the requirements of UBC (82) Sections 2303 through 2312, except as modified in the governing design codes as set forth in DS AR Section 5.5.7.3.
Wind loads are determined in accordance with UBC (82) Section 2311. Wind forces are based on a Basic Wind Speed of 110 mph, Exposure C, and an occupancy importance factor of 1.0.
l The IOSB is designed to accommodate 35-ton crane loads, although the actual crane used is a 25-ton crane. Special loads are determined in accordance with UBC (82) Section 2308.
Structural elements of the IOSB are proportioned to resist the most severe combination of structural effects resulting from the following combinations of seismic load and roof load:
- 1. A roofload of 20 psf combined with Seismic Zone 3 seismic loads.
- 2. A snow load of 40 psf combined with Seismic Zone 2 seismic loads.
Seismic loads are determined and applied in accordance with the provisions of UBC (82) Section 2312 based on a horizontal force factor of 1.33, occupancy importance factor of 1.0, and a site period determined per UBC standard 23,1.
A. Floor Live Loads Cell Floor - 2.000 psf (based on stacking twelve 32,300 lb,300-cu ft liners three high plus platforms).
Cell Cover - 350 psf (based on stacking one,1 ft cell cover plus 200 psf live load).
Dry Active Waste (DAW) Area - 1,000 psf (based on portable shield units, 200-cu-ft capacity, and 66,400-lb loaded weight).
Truck Bay - HS 20-44 truck loading or 105,000 lb truck (single axle load not to !
exceed 18.000 lb.) applied in accordance with UBC (82) Section 2304 (c); 1,000 l
psf for lay down areas (based on lay down of 32,300-Ib,300-cu ft liners).
5.5-4
5.5.7.2 Design Basis (Continued)
Other floor and platform live loads were determined in accordance with UBC (82), Section 2304, with consideration given to equipment maintenance.
5.5.7.2.1 Codes, Standards, and Regulatory Requirements The following codes, standards, and regulatory requirements were used, where applicable, in the design of the IOSB.
A. Eederal Regulations and Guides I Title 10 CFR Part 20 - Standards for Protection against Radiation. l l
Title 10 CFR Part 50 - Domestic Licensing of Production and Utilization Facilities. !
Title 10 CFR Part 50, Appendix A - General Design Criteria 60,63 and 64.
Title 40 Part CFR 190 - Environmental Radiation Protection Standards for Nuclear Power Operations.
Title 10 CFR Part 71 - Packaging of Radioactive Material for Transport and Transportation of Radioactive Material under Certain Conditions.
1
]
NRC Regulatory Guide 8.8, Rev. 3. June 1978 - Information relevant to ensuring j that occupational radiation exposures at nuclear power stations will be as low as i reasonably achievable (ALARA).
NRC Regulatory Guide 8.10 - ALARA Design. :
1 NRC NUREG 0800 - Standard Review Plan Section 11.4, Appendix A (July l 1981), Design Guidance for Temporary On-site Storage of Low Level Radioactive 1 Waste. I 1
OSHA - Occupational Safety and Health Administration Standards (or appropriate !
local governing codes). j i
NRC Generic Letter 81 Storage of Low Level Radioactive Wastes at Power Reactor Sites (November 10,1981).
1 NRC Regulatory Guide 1.143, July 1978. Rev.1. October 1979. !
D. Other Codes and Standards (as applicable) 4 Uniform Building Code (UBC)- 1982.
i American Concrete Institute (ACl) Standard 318 Building Code Requirements for Reinforced Concrete.
5.5-5
5.5.7.2.1 Codes, Standards, and Regulatory Requirements (Continued)
American Society of Heating, Refrigeration and Air Conditioning Engineers (ASHRAE) 1981 - Handbook of Fundamentals and Guides.
American Institute for Steel Construction (AISC)- Manual of Steel Construction, Eighth Edition.
American Nuclear Standards Institute (ANSI) A58.1 Building Code Requirements for Loads, Minimum Design in Buildings and Other Structures.
ANSI B31.1-1981 - Power Piping.
National Electric Code,1981.
National Fire Code - NFPA,1981.
ANSI- Standard N13.1-1969 - Guide to Sampling Airborne Radioactive Materials in Nuclear Facilities.
i National Environmental Policy Act of 1969.
ACI-301-72 (Revised 1981)- Specification for Structural Concrete for Buildings.
5.5.7.3 Material Reauirements l
A.
Structural Steel Structural steel conforms to the material specifications listed in Section 1.4 of the AISC code, Eighth Edition. Building framing steel is ASTM A36.
B. Concrete 3,000 psi day compressive strength.
l 4.000 psi 28 day compressive strength (cell cover support structure).
l l 2.000 psi day compressive strength (lean concrete fill).
C. Reinforcing Steel Deformed billet ASTM A615 - Grade 60.
5.5.7.4 Structural Reauirements Cell Area and Truck Bay - The area below the cell covers and truck bay is a reinforced cast-in-place concrete design.
l l
l S.5-6
1 l
l 1
l 5.5.7.4 Structural Reauirements (Continued) l Cell Covers - The cell covers are a two-section (1-ft thick each) concrete construction. Covers are supported by a beam and column system; interior cell partitions are provided.
Cell Area (Above Shield Covers)- The area above the cell covers is structural steel framing with pre-cast concrete panel walls.
DAW Storage Area - The DAW storage is structural steel framing with pre-cast concrete panel l walls.
1 Service Head - The service head is structural steel framing with pre-cast concrete wall panels. '
Floor construction for the upper floor is concrete placed on a self-supporting metal form deck.
l Foundation System - The foundation for the facility consists of a concrete mat under the cell and truck bay areas and slabs-on-grade with spread footings, as required, elsewhere.
l Wall thickness is as follows:
Exterior walls below cell covers: 36 inches Exterior walls above cell covers: 12 inches Exterior walls for DAW storage area: 12 inches Cell cover thickness is 24 inches in two 12-inch sections.
Cell Dimensions are as follows:
Width: 14 ft.10 in. clear opening Length: 15 ft. 6 in. clear opening Effective Depth: 36-ft. clear height (top to bottom, below cell covers)
Roof System:
Concrete thickness is 2 inches in cell area and 6 inches in the 9AW area.
An earthen berm is provided around the facility to minimize the shielding requirements of the building.
5.5.8 SOLIDIFICATION BUI.LDING 5.5.8.1 General Description The Solidification Building is a metal building that is designed to enclose radioactive waste processing activities. The building provides weather protection and containment of possible 5.5-7
1 5.5.8.1 General Description (Coatinued) radioactive waste spills during transfer of radioactive wastes and operation of radioactive waste processing equipment.
The Solidification Building is located along the full length of the east end of the Auxiliary Building (120 feet),is 30 feet wide, t.nd wraps around the northeast corner of the Auxiliary Building with an additional 36 foot by 24 foot section. The radioactive waste-processing portion of the Solidification Building is constructed on perimeter curbs to contain possible spills. The building includes:
A. Roll-up doors for vehicle access, B. Doors for personnel access, C. Automatic sprinklers for fire protection, l
l D. Ventilation provided by the Auxiliary Building HVAC system,
! E. Emergency lighting, anu F. Service Air System supply.
l 5.5.8.2 Desien Basis l
The Solidification Building is designed in accordance with NRC Standard Review Plan (NUREG-0800) section i1.4, Solid Waste Management Systems, for structures which house l
portable radioactive waste processing systems.
I l
l A District calculation concluded that the structural failure of the building would not adversely affect the Auxiliary Building or equipment inside the Auxiliary Building. This calculation also determined that the Solidification Building design is adequate to withstand 101 mile per hour winds and is structurally designed for Zone 3 seismic loads with a 1.5 Importance Factor per the Uniform Building Code.
The structure and building piping installed is designed and constructed to the following codes and regulatory guidance:
A. Quality Class 2 B. Seismic Category 11 C. Standard Review Plan Section i1.4 i
D. Uniform Building Code i 5,5-8 l
5.8 REFERENCES
- 1. License Amendment No. I19, 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. I17, dated March 17,1992, Possession Only Licente
- 4. BC-TOP-4: Seismic Analyses of Structures and Equipment for Nuclear Power Plants.
April 30,1971
- 5. BC-TOP-5: Prestressed Concrete Nuclear Reactor Containment Structures, August 23, 1971
- 6. Miller, D. R. and Williams, W. A., Tornado Protection for the Spent Fuel Storage Pool, General Electric Company, APED-5696, San Jose, California, November,1968
- 7. D. A. Nesterenko and H. G. Wagner, Consulting Engineering, August 1963
- 9. The Nastran User's Manual NAS A SP-222, C. W. McCormick, Editor, October 1969
- 10. Refer T. C. Waters and N. T, Barrett, Prestressed Concrete Pressure Vessels for Nuclear Reactor, J. Brit. Nucl. Soc. 2,1963
- 11. " Torsion in Symmetrical Buildings," Fourth World Conference on Earthquake Engineering, Santiago, Chile,1969
- 12. Updated Safety Analysis Report (USAR) Amendment No. 8
- 13. SMUD Calculation Z-SFC-M2555, " Peak Spent Fuel Pool Temperature Versus Calendar Time."
- 14. SMUD Calculation Z-SFC-M2560, " Spent Fuel Pool Heat Up During LOOP with Pool at 23.25 feet."
Amendment 3 5.8-1
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l i TABLE OF CONTENTS l l
' Section Title Page
+
- 7. INSTRUNIENTATION AND CONTROL 7.1-1 i
7.1 PROTECTION SYSTENIS 7.1-1 7.2 REGULATION SYSTENIS 7.2 1 l 7.3 INSTRUNIENTATION 7.3- 1 7.4 OPER ATING CONTROL STATIONS 7.4- 1 l'
7.4.1 GENERAL LAYOUT 7.4- 1 I
7.4.2 INFORNIATION DISPLAY AND CONTROL FUNCTION 7.4-2 7.4.2.1 Console and Panel Lav-out 7.4-6 l 7.4.3 SUMNIARY OF ALARN!S 7.4-7 l 7,4.4 CONINIUNICATION 7.4-7 l )
7.4.5 OCCUPANCY . 7.4-8
'7.4.6 AUXILIARY CONTROL STATIONS 7.4-9 l 7.4.7 SAFETY CONSIDERATIONS 7.4-9 l 7.4.8 SYSTEN1 EVALUATION 7.4-9 l 7.4.8.1 Control Room Availability 7.4-9 l
7.5 REFERENCES
7.5 1 Amendment 3 7-1 4
LIST OF TABLES Table Title Page 7.41 Regulatory Guide 1.97 Parameters Monitored by 7.4-3 PICS and DRMS Amendment 3 l l
l 7-11 l
\
l l
7.4 OPER ATINO CONTROL STATIONS 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 is minimized. Since the ,
l consequences of a credible accident in the PDM would expose control room personnel to much I less than the 5 rem whole body habitability limit, special radiation protection equipment is not required to protect control room personnel from excessive radiation exposure. Also, non-radiological control room habitability following a chlorine leak is assured based on an administrative limit for chlorine of no more than 100 pounds allowed within the Industrial Area at any time.
Considering the limited plant activities and reduced scope of credible accidents in the PDM, it is unlikely that a situation could arise that would necessitate evacuation of the control room.
The control room is manned at all times during the PDM, as required per the Permanently Defueled Technical Specifications, by a Certified Fuel Handler (CFH) or a Shutdown Control Room Operator (SCRO), as a minimum. ((The CFH and SCRO are trained and qualified in the monitoring activities and response functions required to stand watch in the control room and '
safely store spent fuel in accordance with the District maintained CFH Training Program (initially NRC approved) and the Non-Certified Operator Training Program, respectively.)]
l The Technical Support Center (TSC), located near the control room, is provided to enable the plant management personnel that comprise the Emergency Response Organization (ERO) to relieve control room personnel during emergency conditions. The TSC will be maintained during the PDM in accordance with the latest approved Emergency Plan.
7.4.l GENER AL LAYOUT Control room design allows one qualified individual to supervise the operation of Rancho Seco l during the PDM. Figure 7.4-l shows the lay-out of the control room. The instrumentation and l control devices labeled in Figure 7.4-1 that are for reactor start up, shutdown, and normal and emergency reactor operations are not required during the PDM.
The TSC has facilities to support the plant management and technical personnel, who comprise the ERO, during an emergency. Once activated, the TSC becomes the primary on-site communication center for the plant during an emergency.
From the control room it is possible to monitor plant status and plant radiation via perimeter monitoring. If meteorological data is needed in conjunction with an actual or potential airbome off-site radiological release, it may be obtained from the National Weather Service in Sacramento. Communication systems, both hard-line and radio frequency, assure that Rancho Seco will have contact with off-site agencies for informational notification purposes.
Amendment 3 i
7.4-1 l .
1
7.4.2 INFORMATION DISPLAY AND CONTROL FUNCTION The information necessary for routine monitoring of the plant is displayed on one console and several vertical panels in the control room. Information display and control equipment employed l on a routine or emergency basis is located on console H2TV. Less frequently used equipment is mounted on vertical panels visible to control room personnel. A plant computer, with monitors )
located on console H2TV,is available for alarm monitoring, performance monitoring, radiation ,
monitoring, and data loggmg.
I 1
Information displays are designed to provide the operator with sufficient information to make l proper evaluations under the full range of PDM operating conditions. The displays are arranged to facilitate evaluation and to avoid the possibility of confusing the operator.
Regulatory Guide 1.97 describes a method acceptable to the NRC for complying with regulations requiring instrumentation to monitor Type A, B, C, D, or E plant variables during and following an accident. A variable included as Type A does not preclude it from being included as Type B,
( C, D, or E or vice versa.
Type A Variables are those variables to be monitored that provide the primary information required to permit control room personnel to take specific manually controlled actions for which no automatic control is provided and that is required to accomplish their safety functions for design basis accident events. There are no Type A Variables required during the PDM.
Type B Variables are those variables that provide information to indicate whether plant safety l functions are being accomplished. Plant safety functions are (1) reactivity control,(2) core cooling, (3) maintaining reactor coolant system integrity, and (4) maintaining containment integrity (including radioactive effluent control). There are no Type B Variables required in the PDN1.
l l
Type C Variables are those variables that provide information to indicate the potential or actual breach of fission product release barriers. The barriers are (1) fuel cladding,(2) primary coolant pressure boundary, and (3) containment _ In the PDM, only the fuel cladding barrier is the applicable concern, and only the Auxiliary Building Stack radiation monitor is required to be l maintained as a Type C Variable. Table 7.4-1 lists the Type C Variables applicable tc Rancho Seco in the PDM.
Type D Variables are those variables that provide information to indicate the operation of individual safety systems and other systems important to safety. These variables are to help the operator make appropriate decisions in using the individual systems important to safety in mitigating the consequences of an accident. There are no Type D Variables required in the PDM.
Type E Variables are those variables to be monitored as required for use in determining the magnitude of the release of radioactive materials and continually assessing such releases.
Several plant effluent process and area radiation monitors and portable radiation analyzers and ,
samplers are required to be maintained as Type E Variables in the PDM. Table 7.4-1 lists the 1 Type E Variables applicable to Rancho Seco in the PDM. l l
Amendment 3 7.4-2 1
4 1
Page 1 of 2 TABLE 7.4-1 REGULATORY GUIDE 1.97 PARAMETERS
' MONITORED BY PICS AND DRMS Variable Variables per SMUD SMUD Computer Number R.G.1.97, R3 Tag No. Cat. Display Range TYPE A VARIABLES l
- None TYPE B VARIABLES
~None TYPE C VARIABLES '
- 33. Effluent R-15045 2 PICS 3.4E-7 to Radioactivity 3.4E-1 uCi/cc (Category 2) Kr-85 TYPE D VARIABLES i
None !
TYPE E VARIABLES l
- 69. Radiation R-15025 3 IE-4 to Exposure Rate R 15030 lE+4 R/hr (Category 3) R-15031 R-15033 R-15034 Portable Analyzers
- 72. Auxiliary Bldg. R-15045 N/A PICS 3.4E-7 to Effluent R-15546A 3.4E-1 uCi/cc Discharge . Kr-85 (Category 2) l-Amendment 3 l
7.4-3 l'
l
? - -_- -
Page 2 of 2
- TABLE 7.4 1 (Continued)
REGULATORY GUIDE 1.97 PARAMETERS MONITORED BY PICS AND DRMS Variable Variables per SMUD= SMUD Computer Number R.G.1.97, R3 Tag No. Cat. Display Range TYPE E VARIABLES (Continued)
- 77. -AllIdentified R-15017A . 3. _ PICS.. 3.4E-7 to
' Plant Release R-150178 . 3.4E-1 uCi/cc Points R-15044 Kr 85
,, (Category 3) R-15045 0-110% design l R-15546A flow FIRQ-95108
. 78. Airborne Portable 3 lE-9 to Radioactivity Samplers & IE-3 uCi/cc l.
(Category 3) Analyzers l
. 79. Plant & Environs Portable 3 IE-3 to ,
Radiation Analyzers lE+4 IUhr (Category 3) (photons and ,
l.
l- beta)
- 80. Plant & Environs Portable 3 Isotopic Radioactivity Samplers & Analysis
, (Category 3) Analyzers i
Amendment 3 l i
7.4-4
. _ . ~ - - - - - - _ = _ - - . _ - .-..- - .-- - .- . _ . _ .
7.4.2 INFORMATION DISPLAY AND CONTROL FUNCTION (Continued)
Hardwired indicators and recorders, computer driven displays, and other means are utilized to present Regulatory Guide 1.97 variables in the control room. This approach is necessary due to space limitations and provides the opportunity to incorporate favorable human factors into the presentation of the information.
The Plant Integrated Computer System (PICS) replaced the Interim Data Acquisition and Display System (IDADS), Bailey computer, Control Room annunciator panels, and the Control Room l console for the Digital Radiation Monitoring System (DRMS). PICS is designed to monitor the !
Regulatory Guide 1.97 type C and E variables as specified in Table 7.4-1.
l PICS consists of two computers located on console H2TV, Remote Transmit-receive Units (RTUs), two display monitors, an alarm printer, and a report printer. During the Permanently Defueled Mode (PDM), appropriate field signals are inputted into the PICS computers and I displayed on the Control Room monitors. Alarms are displayed on a monitor and printed out on the alarm printer. Parameter trends are displayed on a monitor and printed out on the report i pnnter, l l
l Input signals from plant equipment enter PICS from the following sources-
- 2. RTU 1 (H4CDAR2)
- 5. RTU 4 (H4CDAR10) t
- 6. RTU 5 (H4DOEl) -
! 7. RTU 6 (H2WW)
- 8. RTU 7 (HSSFSI) l
[ l l
PICS provides the following general features for operation during the PDM:
l
- 1. Data Acquisition, i
- 2. Man Machine Interface,
- 3. Alarm Annunciation, i
Amendment 3 7.4-5
_ _ _ _ _ . _ . . - _ _ . . _ _.___ ___m __ _ _ . _ . _ __
)
7.4.2 INFORMATION DISPLAY AND CONTROL FUNCTION (Continued)
- 4. Radioactive Release Monitoring, and
- 5. . On line Trending.
The DRMS, through PICS, provides on line information conceming radiation levels of selected plant processes and dose rates for various areas within the plant. Process measurements (1) provide diagnostic or status information on particular parts of the plant and (2) monitor releases of radioactive material from the plant.
DRMS consists of subsystem elements, including:
- 1. Gaseous effluent snd area process monitors,
- 2. Liquid effluent process monitors,
- 3. Strip chart ircorders, and i
- 4. The RM-23 remote control and display units.
i PICS communicates with selected DRMS monitors. Communication between each DRMS monitor and the RM-23 module is accomplished over a single dedicated line. The DRMS design, in conjunction with PICS, incorporates the following features:
- 1. Distributed data processing,
- 2. Expansion capability,
- 3. Centralization of control and display, and L
- 4. Reliability.
7.4.2.1 .C_onsole and Panel Layout One control console and several vertical panels are required to function during the PDM as described below.
A. Security Alarm Station Console The Security alarm station console (H2TV) contains the necessary phone, radio communications, plant computer, radiation monitoring, and alarm acknowledgement capability to permit a control room operator to act as both the Control Room and Alarm Station Operator.
Amendment 3 ;
1 7.4-6
7.4.2.1 Console and Panel Lav-out (Continued)
B. Vertical Panels !
Only the plant cooling water panel, the auxiliaries panel, tM electrical switching panels, and a small portion of the safety features panel are needed during the PDN! and are summarized below.
1
- 1. The plant cooling water panel (Il2WC) contains instruments end controls for the !
plant cooling water system and site water supply. l l
- 2. The auxiliaries panel (H2X) contains instruments and controls for auxiliary )
equipment including the ventilation system. !
l
- 3. The back-fitted electrical switching panel (H2EW) contains controls and I indicators for the expanded power distribution system. A mimic bus is integrated l with that of the original electrical switching panel (H2ES),
l
- 4. The original electrical switching panel (H2ES) contains controls and indicators for :
the switchyard and power distribution breakers. A mimic bus layout is mounted l on the front of the panel, which proves a pictorial display of the electrical circuit l breaker status. l
- 5. The safety features panel (H2SF) contains some controls associated with the Reactor Building purge system, which is required to function in the PDNI.
C. Annunciators No station annunciators are required to function in the Control Room during the PDNI.
Several station annunciator panels remain in services that are located locally in the plant.
7.4.3 SUNIN!ARY OF ALARN!S Visible and audible alarm units are used in the control room to wa'm the operator of abnormal conditions in any system. Audible alarms are sounded in the control room and in the appropriate areas throughout the plant if high radiation conditions occur.
7.4.4 CONINIUNICATION in addition to the information contained herein, refer to the Emergency Plan for a further description of on-site communication systems.
Amendment 3 7.4-7
7.4.4 CONINIUNICATION (Continued)
There is a paging system within the power block buildings and a station telephone and paging system. Acoustic booths and noise canceling transmitters are used where the background noise levelis high. Communications outside the station are through the local telephone network.
Radio voice communications capabilities also exist.
7.4.5 OCCUPANCY The Auxiliary Building is designed to ensure safe occupancy of the control room and TSC during the PDN!. The control room and TSC are designed with adequate shielding to ensure radiation levels will be maintained at low levels during accident conditions in the PDN1. Also, the control room has a radiation detector with appropriate alarms. Because high levels of airborne radioactivity or chlorine cannot occur in the PDNt (see DSAR Section 7.4), the control room and TSC air is not required to be automatically recirculated. Backup lighting is provided in the control room and TSC.
The potential magnitude of a fire in the control room is limited by the following factors:
A. Control Room construction is of noncombustible materials.
B. Furniture in the Control Room is of metal construction.
C. All areas of the Control Room are accessible and adequate fire extinguishers are provided.
D. The Control Room is occupied at all times by a qualified person trained in fire extinguishing techniques.
E. The vertical control panels form a barrier between the cabinet area and the Control Room.
The only flammable materials inside the Control Room are:
A. Paper in the form of logs, records, procedures, manuals, and diagrams, as required for station operation.
B. Small amounts of combustible materials used in the manufacture of various electronic equipment.
As indicated by the above list, flammable materials have been minimized to the extent that a fire is not likely to spread. Therefore, if a fire is started,it will be so small that it can be extinguished using a hand. held fire extinguisher. The resulting smoke and vapors would be removed by the normal ventilation system.
7.4-8
7.4.6 AUXILIARY COhTROL STATIONS Auxiliary control stations are provided where their use simplifies the control of auxiliary systems equipment. Sufficient indicators and alarms are provided to inform control room personnel of abnormal conditions at remote control stations.
7.4.7 SAFETY CONSIDERATIONS The primary objectives of the control room layout and design was to provide necessary controls to start, operate, and shut down the nuclear power plant with sufficient controls, displays, and alarms to ensure safe and reliable operation under normal and accident conditions. Special emphasis was given to maintaining control integrity during accident conditions. The design of the control room is more than adequate to ensure safe operation of the facility during the PDM.
Deviations from predetermined conditions are annunciated so that the operator may take appropriate corrective action.
7.4.8 SYSTEM EVALUATION 7.4.8.1 Control Room Availability Safe operation and shutdown of the power plant was conducted from the control room. The control room is specifically designed to permit the operator to perform his duties under all credible accident conditions.
Forced abandonment of the control room is deemed highly unlikely and is not credible in the PDM for the following reasons:
A. The control room has been given the highest priority for shielding from external radiation of any area in the plant.
B. Non-flammable construction materials were used for all interior components, i.e.,
control boards, fumiture, etc.
C. Adequate fire-fighting equipment is available in the control room and operators have fire-fighting training.
D. Cables and switchboard wiring have passed flame tests as described in IPCEA publications S-61-402 and NEMA WC 5-1961.
E. Combustible materials in the control room are kept to the minimum required.
Permanent plant records and non-essential reference materials are stored elsewhere.
l 7.4-9
4 7.4.8.1 Control Room ^vailability (Continued)
F. The'venical control boards provide a fire and smoke barrier between most of the electrical devices that could generate large amounts of smoke.
'O. Fireproof or fire-resistant doors are installed on all rooms adjoining the control room where significant amounts of combustible materials are stored.
7.4-10 ,
i
7.5 REFERENCES
- 1. License Amendment No. I19, 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
- 1
- 3. License Amendment No. I17, dated March 17,1992, Possession-Only License l 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 I X
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l TABLE OF CONTENTS L
i
- Section Title Pace 1
- 9. AUXILIARY AND EMERGENCY SYSTEMS 9.1-1 9.1 GENERAL 9.1-1 9.2 COOLING WATER SYSTEMS 9.2-1
.9.2.1 CONDENSER CIRCULATING ' WATER SYSTEM 9.2-1 9.2.1.1 S,ystem Description 9.2-1 9.2.1.2 Design Data 9.2-1 9.2.2 PLANT COOLING WATER SYSTEM 9.2-2 9.2.2.1 System Descriotion 9.2-2 9.2.2.2 Design Data 9.2-2 9.2.3 COMPONENT AND TURBINE PLANT COOLING WATER 9.2-2 l SYSTEM 9.2.3.1 Svstem Description 9.2-2 9.2.3.2 Design Data 9.2-4 l
. 9.3 ' DECAY HEAT REMOVAL SYSTEM 9.3-1 9.3.1 Borated Water Storage Tank 9.3 1 i 9.4 SPENT FUEL COOLING SYSTEM 9.4- 1 9.4.1 DESIGN BASES 9.41 9.4.2 SYSTEM DESCRIPTION - 9.44 9.4.2.1. Codes and Standards 9.4-4 l 9.4.2.2 Material Compatibility 9.4-4 l i 9.4.2.3 Component Desien 9.4-5 9.4.2.3.1 Piping and Valves 9.4-5 Amendment 3 9-1 5
TABLE OF CONTENTS (Continued)
Section. Title bgg P
9.4.2.3.2 Pumps 9.4-5 9.4.2.3.3 Heat Exchanger 9.4-5 i'
9.4.2.3.4 Filters and Ion Exchanger 9.4-5 9.4.2.4 Leakage Considerations 9.4-5 ' l 9.4.2.5 Failure Considerations . 9.4-6 9.4.2.6 - - Ooeratine Conditions . -
~9.4-6 9.5 STATION VENTILATION SYSTEMS 9.5- 1 9.5.1 DESIGN BASES 9.5-1 i 9.5.1.1 Reactor Buildine 9.5-1 9.5.1.1.1 General Conditions 9.5-1
--9.5.1.1.2 Sizing 9.5-1 9.5.1.2 Auxiliarv Buildina 9.5-1 9.5.1.3 Nuclear Service Electrical Buildine (NSEB) 9.5-2 9.5.2 SYSTEM DESCRIPTION 9.5-2 9.5.2.1 Reactor Buildine 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.l.3 Codes, Standards. Tests 9.5-7 9.5.2.2- Auxiliary Buildine 9.5-7 9.5.2.2.1 General 9.5-7 9.5.2.2.2 Control Room and Technical Support Center 9.5-7 Amendment 3 9-ii
TABLE OF CONTENTS (Continued)
SKtio.q Iille Page 9.5.2.2.3 Radiochemical and Service Areas 9.5-8 l 9.5.2.2.4 Radwaste and Fuel Storage Areas 9.5-8 l 9.5.2.2.5 Electrical Equipment, Switchgear, and AC/DC Panel Rooms 9.5-9 l 9.5.2.2.6 Coolant and Miscellaneous Waste Tanks 9.5-9 l 9.5.2.2.7 Emergency Pump Rooms 9.5-10 l 9.5.2.2.8- Auxiliary Building Exhaust Air Filtration System 9.5-10 l 9.5.2.2.9 Communication Room 9.5-10 l 9.5.2.2.10. Chilled Water System 9.5-10 l 9.5.2.3 Nuclear Service Electrical Buildine (NSEB) 9.5-12 l 9.5.2.4 Batterv Building 9.5-12 l 9.5.2.5 Interim On-site Storace Buildine (IOSB) 9.5-12 l 9.5.2.6 Switchvard Control Buildine 9.5-13 l 9.5.2.7 Codes. Standards and Tests 9.5-13 l 9.6 FUEL HANDLING SYSTEM 9.6-1 9.6.1 DESIGN B ASES 9.6 1 9.6.1.1 General System Function 9.6- 1 9.6.1.2 Spent Fuel Storace Pool 9.6-1 9.6.1.3 Spent Fuel Pool Water Chemistry 9.6-2 9.6.1.4 Fuel Transfer Tube 9.6-2 9.6.1.5 Fuel Handling Eauioment 9.6-2 l
9.6.2 SYSTEM DESCRINION AND EVALUATION 9.6-2 Amendment 3 l
9-iii i.. ....;...,-
TABLE OF CONTENTS (Continued)
Section Title Page. 1 9.6.2.1 Handling Soent Fuel Assemblies 9.6-2 l 9.6.2.2 Handling and Loading Spent Fuel Casks 9.6-3 9.6.2.3 Crane Use in Fuel Handline 9.6-4 9.6.2.3.1 Design 9.6-4 9.6.2.3.2 Evaluation 9.6-5
-+
9.6:2.4-- - -Safety Provisions - - -- - --
- 9.6 l 9.7 OTIER AUXILIARY SYSTEMS 9.7-1 9.7.1 - FIRE PROTECTION SYSTEM 9.7-1 9.7.1.1 Desien Bases 9.7-1 9.8 PLANT COMPRESSED SERVICE GAS SYSTEM 9.8-1
9.9 REFERENCES
' 9.9-1 Amendment 3 9-iv
- . . . ~ . ... .. . _- . . . . . . . . .- - . - - . _ ~ . . . . . _. .- . - ..
LIST OF TABLES Table Title Page 9.2-1 Plant Cooling Water Pumps Performance Data 9.2-3 l 9.2-2 Component Cooling Water System Data 9.2-5 l 9.4- 1 Spent Fuel Cooling System and Equipment Performance 9.4-2 9.4-2 Spent Fuel Pool Heat Load and Heat-Up Analysis 9.4-3 9.5-1 Station Ventilation System Major Component Data 9.5-3 9.5-2 Single Failure Analysis for the Radwaste Tanks Ventilation and 9.5-11 l Sampling Systems 9.6 1 Fuel Cask Drop Heights 9.6-8 9.8-1 Compressed Service Gas Vessels 9.8-2 I
i l
l.
. Amendment 3 t-l 9-v e - . -
ll -
I LIST OF FIGURES l Figure Title 9.1-1 Flow Diagram Identifications 9.2-1 Removed ,
1 9.2-2 Removed 9.2-3 Removed 9.2-4 Plant Cooling Water System 9.2-5 Component Cooling Water /rurbine Plant Cooling Water System 9.2-6 Turbine Plant Cooling Water System Figure was Combined with Figure 5.2-5 9.3- 1 Removed
.9.4- 1 Spent Fuel Cooling System 9.5-1 Reactor Building Purge System 1
9.5 2 Control Room & TSC Normal HVAC System L
9.5-3 . Radio-Chem Lab & Service Area HVAC System 9.5-4 Radwaste & Fuel Storage Area HVAC System 9.5-5 Electrical Equipment & Switchgear Rooms HVAC System 9.6-6 Auxiliary Building Radwaste Area HVAC System 9.5-7 Auxiliary Building Exhaust Air Filtration Unit 9.5-8 Auxiliary Building Communication Room HVAC System 9.5-9 Auxiliary Building Normal Chilled Water System 9.5-10 Nuclear Service Electrical Building HVAC System 9.5-11 Interim On-site Storage Building HVAC System 9.6-1 Plant Cross Section Showing Crane Configuration and Function 9.6-2 Cask Transport Profile Amendment 3 9-vi
9.2 COOLING WATER SYSTEMS The Spent Fuel Pool Cooling system (SFC)is the primary spent fuel pool water cooling system.
This system is addressed in DSAR Section 9.4.
The back-up spent fuel pool cooling system is the Radwaste and Fuel Storage Area Ventilation System. Due to the reduced level of decay heat emanating from the spent fuel, relying on evaporative cooling as the back-up spent fuel pool cooling mechanism is sufficient to ensure the spent fuel pool does not exceed its temperature lim.its. This system is addressed in DSAR Section 9.5.2.2.4.
The following cooling water systems are maintained functional in the PDM to support operation of the SFC system:
A. Plant Cooling Water system, and B. Component Cooling Water system.
These SFC support systems are addressed in more detail below.
9.2.1 CONDENSER CIRCULATING WATER SYSTEM 9.2.1.1 System Description Since no steam is discharged into the dual pressure condenser and the cooling towers are not in operation during the PDM, the majority of the Condenser Circulating Water system is not required in the PDM and is not described in the DS AR. Only the circulating water intake canal is used during the PDM.
The circulating water intake canal, also known as the circulating water basin, receives make-up water from the Folsom South Canal via an off-site pumping station. The make up line from the canal pumping station discharges into the circulating water basin make-up fiume. Also, the Rancho Seco reservoir may be lined up to supply make-up water to the circulating water basin.
The flow of make-up water is controlled by the circulating water basin intake structure water level control.
The circulating water basin and intake structure is the primary supply of water for the plant fire water supply system. The Rancho Seco Fire Protection Plan provides the fire water supply system details.
9.2.1.2 Design Data The circulating water basin and intake structure was designed as Quality Class 2 and Seismic Category II.
Amendment 1 9.2-1
9.2.2 PLANT COOLING WATER SYSTEM 9.2.2.1 Svstem Description The Plant Cooling Water system (PCW) is required to function in the PDM. A simplified flow diagram of the PCW system is shown in Figure 9.2-4. Of the various site cooling water systems that the PCW system cooled during power plant operations, only the Component Cooling Water system (CCW)is required to function in the PDM.
During the PDM, the important-to-safety function of PCW is to remove heat from CCW, while the important-to-safety function of CCW is to remove heat from SFC, which removes the decay heat generated by the spent fuel in the spent fuel pool. Because the heat removal loads and cooling requirements are greatly reduced from that of an operating nuclear power plant, two vertical, wet pit PCW pumps are maintained during the PDM. One PCW pump is normally continually operating while the second pump serves as an alternate. The pumps take suction from the make-up fiume of the circulating water basin. After passing through the PCW heat exchangers, the water is returned to the basin at a different location.
I PCW relies on make-up water supply from either the Folsom South Canal or Rancho Seco reservoir. The flow of make-up water is controlled by the circulating water basin intake structure level control system. The make-up water will always be cooler than the water in the intake canal, and, except for short periods during summer, this will result in a make-up flume water temperature that is satisfactory for use in PCW When the make-up flume water temperature exceeds 80 F, more of the cooler makeup water will be allowed into the fiume than is required by the level control. This is accomplished when the flume water temperature control overrides the level control, and attempts to maintain the PCW water temperature no greater than 80 F.
PCW water flow through the heat exchangers is controlled by control valves installed at each heat exchanger's outlet in accordance with the temperature of the water in the secondary loop.
9.2.2.2 Design Data PCW is designed as quality Class 2 and seismic Category II. Design data for the PCW pumps are shown in DSAR Table 9.2-1.
9.2.3 COMPONENT AND TURBINE PLANT COOLING WATER SYSTEM 9.2.3.1 Svstem Description A simplified flow diagram of the combined Component Cooling Water (CCW) and Turbine Plant Cooling Water (TPCW) system is shown in Figure 9.2-5.
Amendment 3 9.2-2
l TABLE 9.2-1 PLANT COOLING WATER PUMPS PERFORMANCE DATA i
COMPONENT DATA (on a per pump basis)
Plant Cooling Water Pumos l
- Quantity 2 l l
Type Vertical, wet pit I Design flow 12,000 gpm Design head 75 feet Motor horsepower 300 horsepower Material Mi resistant / Stainless Steel / bronze Design pressure 53 psig Amendment i 9.2-3
_ _ _ _ . , _ . , _ _ _ ,, ..--w.-
i I
9.2.3.1 System Description (Continued)
During the PDM, the imponant-to-safety function of CCW is to provide cooling water to the spent fuel pool cooler / heat exchanger. Also, CCW and a portion of the turbine plant cooling water system provide coolin'g to the plant air compressor after-coolers and the various coolers in the radioactive waste disposal system. Only that portion of TPCW that cools the air compressor after-coolers and instrument air dryers is operated during the PDM.
CCW consists of two CCW pumps, two full-capacity CCW heat exchangers, a surge tank, a chemical addition tank, piping, and valves, all arranged in a closed loop. The one additional closed cooling loop, provided between the control rod drive cooling water heat exchangers and l the control rod drives,is not required in the PDM. Heat removed from the various in-service
- heat exchangers is rejected to PCW at the CCW heat exchangers. Indication of the CCW surge tank water level is shown locally via a sight level gauge.
Make-up water to the CCW surge tank is added manually. .CCW contains a radiation monitor to detect excessive levels of radiation. CCW is monitored for radioactivity, and annunciation is i provided to the control room. The sensitivity of the radiation monitor is given in DS AR Section 11.8, Table 11.8 3. The alarm is set to the minimum setting above background that will avoid spurious alarms.
When a radiation alarm is received and a leak is confirmed, the leak will be located and isolated as soon as possible.
- 9.2.3.2 Desien Data l The majority of CCW is designed as Quality Class 2 and Seismic Category II. Drains and other
- auxiliary lines are designed as Quality Class 3. Design data for CCW is shown in DSAR Table 9.2-2.
If CCW fails and the spent fuel pool cooling system is not able to remove enough heat to maintain the spent fuel pool coolant within its temperature limits, the Radwaste and Fuel Storage 1 j Area Ventilation System is relied on to ensure the spent fuel pool temperature does not exceed its ;
limits.
j i
i l
Amendment 3 l~
9.2-4
- . _- .- - . - - --.= - . - .- . - - -- . - . . . - . . -
TABLE 9.2-2 COMPONENT COOLING WATER SYSTEM DATA i l
COMPONENT DATA (on a per unit basis)
Component Cooline Water Pumps Quantity 2 l Type Horizontal, centrifugal
- Design flow - 11,000 gpm Design head 150 feet Motor horsepower 600 horsepower Material Cast Iron / Bronze Design Pressure 110 psig l Component Cooline Water Heat Exchangers Quantity 2 Type Tube and shell Material, shell/ tube Carbon Steel / Admiralty Metal 6
Cooling capacity 110 x 10 BTU /hr 6
Cooling water flow 6 x 10 lbs/hr Code ASME Section VIII:TEMA Class "R" Surge Tank Volume 46.8 cubic feet Operating pressure Static Head Code ASME Section VIII t
i Amendment 3 9.2-5
- - , , -. -. ,.,,-.,-..n -
- - . . - ~ - _ . . ~ . - . . . - . - . . . . - - . . - - - - . - .. ..- ..
9.3 DECAY HEAT REMOVAL SYSTEM Since the Radwaste and Fuel Storage Area Ventilation System is now the back-up spent fuel pool
- cooling system, the Decay Heat Removal System is not required to function during the PDM.
Information that addresses the design bases for this system is not required and is not included in the DSAR. The only information previously contained in DS AR Section 9.3 that continues to apply during the PDM is presented below.
9.3.1 . Borated Water Storage Tank The Borated Water Storage Tank (BWST) is located in the tank farm outside the Reactor 4 Building and the Auxiliary Building. The BWST will normally contain a small amount of radioactive water during the PDM, but, the BWST is not required to perform its original safety function of providing borated water for emergency core injection or other reactor control purposes. The BWST may be used to assist in processing spent fuel pool water once the spent fuel is removed from the spent fuel pool,if necessary.
Amendment 1 9.3-1
1 9.4 SPENT FUEL COOLING SYSTEM l
l 9.4.1 DESIGN BASES A central objective of the high-density fuel rack design installed in 1984 was to ensure the spent fuel poo! cooling system (SFC) could provide adequate spent fuel cooling to protect fuel l assembly cladding during long-term storage of spent fuel. SFC heat removal design capacity is large enough to handle a full spent fuel pool (1080 fuel assemblies), which includes a freshly discharged core (177 assemblies decayed 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> with a 9 day core discharge time). Rancho Seco is storing only 493 spent fuel assemblies in the PDM, and the reactor has not operated since June 7,1989. The spent fuel pool heat load is well within the heat removal design capacity of !
SFC. For more details on the high-density spent fuel storage rack design, see DS AR Section 9.6 j or References 2 and 3 listed in DS AR Section 9.9. Performance data for the spent fuel pool heat exchangers is presented in Table 9.4-1. l The District calculations that determined the amount of decay heat generated in the spent fuel l pool by the stored spent fuel assemblies during the PDM were performed using the ORIGEN2 l computer code. The results of these calculations are presented in Table 9.4-2 for various dates after permanent plant shut down. l The District also evaluated heat-up of the spent fuel pool water following a loss of the primary spent fuel pool cooling system. DS AR Table 9.4-2 presents the results of this evaluation for two initial spent fuel pool levels (the minimum allowed Permanently Defueled Technical 1 Specification levels for when spent fuel handling operations are or are not in progress), for various initial spent fuel pool water temperatures, and for the calculated spent fuel pool decay heat load as of March 1,1992 (0.95E6 BTU /HR). The maximum, calculated steady state spent fuel pool coolant temperature is approximately 147 F. Thus, the spent fuel pool will not reach 180"F, and pool boiling will not occur if the primary spent fuel pool cooling system is lost.
The quickest heat-up time to 140"F following a loss of primary spent fuel pool cooling, starting from an initial spent fuel pool coolant temperature of 90"F, is 12 days. Therefore, plant operators have a significant amount of time to take action to restore the primary spent fuel pool cooling system if this system stops operating. Also, this amount of time could be extended by adding make-up water to the spent fuel pool.
Makeup water to the spent fuel pool may be provided by any of the following methods:
A. Pumping water from the Miscellaneous Water Holdup Tank directly to the spent fuel pool through a valve and pipe.
B. Pumping water from the Demineralized Water Header, which gets its water from l the Condensate Storage Tank, to the spent fuel pool through a hose.
C. Pumping water from the Folsom South Canal or Rancho Seco reservoir to the spent fuel pool via the diesel or electric fire pump and a fire hose station.
Amendment 1 9.4-1
TABLE 9.4-1 SPENT FUEL COOLING SYSTEM AND EQUIPMENT PERFORMANCE '
COMPONENT OUANTITIES AND CAPACITIES 6
Normal SFC system cooline capacity 10.3 x 10 BTU /hr SFC system design pressure 75 psig
'SFC system desien temocrature 212 F
' SFC system cooler data
- Quantity 1 Type Tube and shell Material, shell tube Carbon Steel / Stainless Steel 6
Duty 8.76 x 10 BTU /hr Cooling water flow' 5.0 X 10' lbs/hr i SFC system nump data Quantity 1 Type Horizontal, centrifugal Material Stainless steel (wetted parts)
Flow 1,000 gpm Head 25 feet Motor horsepower 15 horsepower
-SFC system coolant ion exchanger Flow rate 160 gpm 3
i Bed volume 50 feet Type Non regenerative mixed bed Vessel material Stainless Steel l Design pressure 150 psig Design temperature 150 F SFC system coolant filter Flow rate 160 gpm Type Disposable cartridge Vessel material Stainless Steel Design pressure 200 psig Design temperature 212 F Soent fuel cool water i Volume 81,000 feet' L 9.4-2 ,
I l
.,-,,,,-1
l 1
l TABLE 9.4 2 SPENT FUEL POOL HEAT LOAD AND HEAT-UP ANALYSIS SPENT FUEL POOL DECAY HEAT LOAD VS. CALENDAR DATE DATE 03/01/92 12/31/92 12/31/93 12/31/94 12/31/95 12/31/98 DECAY HEAT 0.95E6 0.83E6 0.75E6 0.70E6 0.66E6 0.60E6 l (BTU /HR) !
^
1 TIME FOR SPENT FUEL POOL TEMPERATURE TO REACH STEADY STATE VS. INITIAL l
SPENT FUEL POOLTEMPERATURE AS OF MARCH 1,1992 FOR A 23.25 FEET l INITIAL SPENT FUEL POOL WATER LEVEL (Spent Fuel Pool steady state temperature is approximately 143 F)
INITIAL POOL TEMP 90 100 110 120 130 140
( F)
! TIME TO REACH l
STEADY STATE 27 25 23 20 17 8 (DAYS)
TIME FOR SPENT FUEL POOL TEMPERATURE TO REACH STEADY STATE VS. INITIAL SPENT FUEL POOL TEMPERATURE AS OF MARCH 1,1992 FOR A 37 FEET INITIAL SPENT FUEL POOL WATER LEVEL (Spent Fuel Pool steady state temperature is approximately 147 F)
INITIAL POOL TEMP 90 100 110 120 130 140
("F)
TIME TO REACH STEADY STATE 27 26 24 22 19 15 (DAYS)
Amendment 1 9.4-3
,, - - . a - - - - , -- . - . , - ,
l l
l 9.4.1 DESIGN B ASES (Continued)
D. Supplying water from the pressurized Service Water system to the spent fuel pool through a hose.
9.4.2 SYSTEM DESCRIPTION A simplified flow diagram of the spent fuel cooling system is shown in Figure 9.4-1. The system consists of the (a) cooling loop, including the spent fuel storage pool, pump, cooler, piping and valves,(b) purification loop, including the spent fuel coolant demineralizer pump, filter, ion exchanger, piping and valves, and (c) skimmer loop, which is valved into the normal purification loop.
l The removal of the decay heat generated in the spent fuel storage pool is accomplished by continuous recirculation of the coolant through the spent fuel cooler. The purification system is l
tied into the cooling loop downstream of the cooler. The purification sy' stern keeps the spent fuel pool water clean and removes residualionic radionuclides present in the pool. A fraction of the total cooling system flow is diverted through the purification system, passed through the filter and ion exchanger, and returned back to the primary flow path. In addition, a skimmer is installed to keep the pool surface clean from dust. The skimmer takes suction from the pool surface and discharges through the filter and ion exchanger in the purification loap and returns through a set of nozzles back into the pool. The nozzles are arranged to cause movement of water in the pool and facilitate cleaning.
The cooling capacity of the spent fuel cooling system is normally supplemented and can be supplanted,if necessary, by the Radwaste and Fuel Storage Area Ventilation System. This back-up spent fuel pool cooling system normally operates in a continuous mode.
9.4.2.1 Codes and Standas SFC is designated as Quality Class 2 and Seismic Category II, with the exception of the cooling loop pipe segment between the spent fuel storage pool and the first shut-off valve, which was designed as Quality Class I and Seismic Category I. The complete skimmer loop is Quality Class
- 3. Piping from the spent fuel cooler to the spent fuel pool was upgraded to Seismic Category I.
9.4.2.2 Material Compatibilitv SFC system components in contact with water containing boric acid are made of stainless steel or stainless clad carbon steel for protection against corrosion and deterioration.
l l Amendment 1 l ,
l l 9.4-4
l t
1 L 9.4.2.3 Component Desien l
1 l
9.4.2.3.1 Piping and Valves Piping and valves are made of stainless steel. With the exception of a few lines supplying air, demineralized water, and new resin slurry to the ion exchanger, piping was designed to USAS Code B31.7 Class 3 and valves are ASME Code for Pumps and Valves for Nuclear Power Class 3.
l l
9.4.2.3.2 Pumps The SFC pumps are of proven design and made of stainless steel, with all materials and non-destructive testing in accordance with the ASME Code for Pumps and Valves for Nuclear i
Power Class 3. The pumps were subjected to a hydrostatic test, withstanding pressures of at least 1.5 times the design pressure. Factory performance tests were conducted to demonstrate adherence to the design specifications.
9.4.2.3.3 Heat Exchanger The spent fuel cooler is designed and manufactured to the ASME Boiler and Pressure Vessel Code Section ill-C. General requirements conform to TEMA Class R standards. Nondestructive testing including ultrasonic and eddy-current testing of tubes conforms to the specifications of the above code. Both the shell and the tube side of the cooler were hydrostatically tested to 1.5 times the design pressure.
9.4.2.3.4 Filters and Ion Exchanger The spent fuel coolant filter and ion exchanger are made of stainless steel and designed, manufactured, and tested in accordance with the ASME Boiler and Pressure Vessel Code.
9.4.2.4 Leakace Considerations If a fuel assembly leaks in the spent fuel pool, a small quantity of fission products may enter the spent fuel pool cooling water. A purification loop is provided for removing these fission products and other contaminants from the water.
The air inside the Fuel Storage Building is ventilated in a controlled manner to the outside, via the Radwaste and Fuel Storage Area Ventilation System, through the Auxiliary Building Stack.
Any SFC leakage is routed to radwaste collection systems.
The fuel transfer tubes that penetrate the Fuel Storage Building have valve and blind flange ends that isolate the fuel transfer tubes.
9.4-5
9.4.2.5 Failure Considerations The most serious type of failure associated with the spent fuel pool would be a complete loss of spent fuel pool water. To protect against the possibility of a severe loss of spent fuel pool water
-inventory due to a SFC breach, SFC piping penetrates the spent fuel pool at or above approximately the 23 foot level; therefore, the spent fuel pool cannot be gravity-drained even close to the top of active fuel. Also, the fuel transfer tubes were designed per ASNE Section - ;
IIIB with a blank flange on the Reactor Building end and a gate valve on the Fuel Storage Building end.
9.4.2.6 Operatine Conditions .
A high temperature alarm annunciates if the spent fuel pool water temperature exceeds 108 F,
.No minimum Boron concentration is required in the spent fuel pool for reactivity control. The Boraflex fuel storage racks provide sufficient reactivity control during the PDM.
i i
i e
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9.4-6' 4
ev --- --. - - -------- ------- ___ _ _- - - - _ . . - - - - _ _ - - - - - - - - - . - - - - - - - - - ,
1 l
9.5 STATION VENTILATION SYSTEMS l
9.5.1 DESIGN BASES 9.5.1.1 Reactor Buildine -
9.5.1.1.1 General Condition 5 -
The Reactor Building emergency ventilation system is not required to perform its safety function during the PDM and is not described in the DSAR. Also, only the portion of the normal Reactor Building ventilation system that is used to purge the Reactor Building with fresh air is required to function in the PDM.
9.5.1.1.2 Sizing The purge system is sized to provide two air changes per hour in the Reactor Building.
9.5.1.2 Auxiliary Buildin_g The station ventilation systems are designed to provide a suitable environment and maximum safety for operating personnel and equipment. Where possible, ventilated space is divided into zones separating the potentially contaminated areas from the clean areas. The path of ventilating air in the potentially radioactive contaminated areas is from the areas of low activity toward the areas of progressively higher activity.
The design is based upon outside conditions of 101 F Dry Bulb and 72 F Wet Bulb temperature in summer and 30 F Dry Bulb in winter in accordance with ASHRAE Fundamentals Handbook recommendations. The Control Roomfrechnical Support Center (TSC) Essential HVAC system is not required to perform its safety function during the PDM and is not described in the DS AR.
Inside conditions in the Auxiliary Building are listed as follows:
! Auxiliary Building Inside Temperatu.c (#F)
(Maximum Desien) l l Control Room (normal HVAC) 78 i Radio-chemical and service area 78
- Radwaste and fuel handling area 90 4
Electrical equipment switchgear, and 78
- ac/dc panel rooms (normal HVAC) 9.5-1 l.
.9.5.1.3 Nuclear Service Electrical Building (NSEB)
Both the 'A'and B' sides of the NSEB normal Heating Ventilating and Air-Conditioning (HVAC) system is maintained functional during the PDM to support electrical equipment that provides normal power to some Auxiliary Building loads required during the PDM. Both the 'A' and B' sides of the essential (emergency) NSEB HVAC system are not required to support plant operations during the PDM and are not discussed in the DSAR. The normal NSEB HVAC system is designed to provide a suitable environment for operating personnel and equipment under normal conditions.
9.5.2 . SYSTEM DESCRIPTION 9.5.2.1 Reactor Building 9.5.2.1.1 General The only portion of the Reactor Building ventilation system required to function during the PDM is the Reactor Building purge system. A simplified flow diagram of the Reactor Building purge system is shown in Figure 9.51. Design data for major components of the Reactor Building purge system, as well as other plant ventilation systems, is given in Table 9.5-1.
9.5.2.1.2 Purge System i
The purge system consists of the Reactor Building purge air supply and exhaust units, installed outside of the Reactor Building, with penetration ducts, isolation valves and controls. The
, supply unit consists of a fan, electric heating coil, pre filter, and high efficiency filter. The exhaust unit consists of a fan, a pre-filter, and HEPA filter. Carbon adsorber filters are not required on the exhaust unit due to an absence ofiodine and an iodine production mechanism in the PDM.
Exhaust air is discharged to the atmosphere through the Reactor Building Stack. The supply and
! exhaust units are equipped with vane control and dilution dampers.
The gaseous discharge is monitored for radioactivity. Upon detection of an excessively high radiation level, the Reactor Building Stack radiation monitor activates an alarm and causes the l purge to automatically terminate. The supply and exhaust units and dampers are operated remotely from a control panel in the control room.
The purge system is provided with motor-operated isolation valves inside the Reactor Building and pneumatically actuated salves outside the Reactor Building, in both the supply and discharge ducts. A high radiation level signal from the Reactor Building Stack radiation monitor automatically trips the Reactor Building purge supply and exhaust fans but does not trip the Reactor Building purge isolation valves. After assessing the problem, the operators can close the j valves remotely from the control room, if necessary.
l 9.5-2 l .
l.
l -
1 1
l Page 1 of 4 TABLE 9.5-1 i STATION VENTILATION SYSTEMS MAJOR COMPONENT DATA Unit Nun ber l . Equipment . Capacity Installed A. REACTOR BUILDING -
- 1. PURGE SYSTEM
- a. Purge Supply Unit 66,700/ 1 1
Fan 16,670 scfm Heating Coil 210 Kilowatts - 1 ;
1
- b. Purge Exhaust Unit 74,000/ 1 l
18,500 scfm l l ..
l l- B. AUXILIARY BUILDING l j l
- 1. CONTROL ROOM l
- a. Normal HVAC Units '
1 Fan 12,800 scfm 1
- j. Cooling Coil 385,000 BTU /hr i j - Heating Coil 35 Kilowatts -1 l
- 2. RADIOCHEMICAL AND SERVICE AREA i
- a. Air Conditioning and Heating Units 1
L Fan 9,890 scfm 1 Cooling Coil 475,000 BTU /hr i Heating Coil 6 Kilowatts !
. l
! b. Exhaust Hoods 3,200 scfm 1 I Make-up Fan Amendment 1 9.5-3
+ --
s .-, , .- ,- -
, , , - ,,n-- , , - , - - . . , , - - - , . , - - - -
Page 2 of 4 TABLE 9.5-1 (Continued)
STATION VENTILATION SYSTEMS MAJOR COMPONENT DATA Unit Number Equipment Capacity Installed
- 3. R.ADWASTE AREA
- 4. FUEL HANDLING AREA p a. Air Handling Unit i
Fan 6,500 scfm 1 Evaporative Cooler 6,500 scfm 1
- 5. RADWASTE, FUEL, AND RADIOCHEMICAL LAB
- a. Exhaust Unit 55,000 scfm 2 l
- 6. ELECTRICAL EQUIPMENT, SWITCHGEAR, AND AC/DC PANEL
- a. Air Conditioning Unit
, Fan 9,400 scfm 1 l
Cooling Coil 242,000 BTU /hr I l
l i
Amendment 3 9.5-4
Page 3 of 4 TABLE 9.5-1 (Continued)
STATION VENTILATION SYSTEMS MAJOR COMPONENT DATA Unit Number Equipment Capacity Installed
- b. Air Conditioning Unit Fan 6,400 scfm 1 Cooling Coil 165,000 BTU /hr 1 l
- c. Air Conditioning Unit Fan 5,600 scfm 2 i
Cooling Coil 145,000 BTU /hr .2
- d. Air Handling Unit i Fan 1,200 scfm 2 l
\
l.
- e. Air Handling Unit l l l Fan 1,350 scfm i 1 l
- f. Exhaust Fan 650 scfm 4 l~
Exhaust Fan 1,300 scfm 1 l
- 7. EMERGENCY PUMP ROOMS Fan 8,000 scfm 5 l
l Cooling Coil 170,000 BTU /hr 5 l
- 8. AUXILIARY BUILDING EXHAUST l AIR FILTRATION UNIT Fan 31,000 scfm i Amendment 3 9.5-5
- -- , ., . -- - - - , ,,, , - - . - , . - - . - , , , , - . ~ . -,- ---
. _ . . _ __ . . _ _ . _ . . . _ - _ _ _ _ _ _ _ _ . . _ _ . - . ~ . _ _ _ _ . .
Page 4 of 4 TABLE 9.51 (Continued)
STATION VENTILATION SYSTEMS MAJOR COMPONENT DATA
- Unit Number Equipment Capacity Installed
- 9. COMMUNICATION ROOM
- a. Air Handling Unit Fan 3,100 scfm 1 Cooling Coil 80,000 BTU /hr I
- b. Air Conditioning Unit Fan 3,500 scfm 1
! Cooling Coil 102.000 BTU /hr 1 1
- 10. SERVING ALL AREAS
( a. Packaged \yater Chiller Unit - 80 tons 2 C. NUCLEAR SERVICE ELECTRICAL BUILDING
- 1. NORMAL AIR CONDITIONING UNIT 2 l
Fan 23,700 scfm 2 Retum Fan 22,400 scfm 2 i Refrigeration Unit 48.9 tons 2
- 2. EXHAUST FANS I
l Fan 240 scfm 4 i Fan 500 sefm 2 9.5-6
1 1
1 9.5.2.1.3 Codes, Standards, Tests The purge system isolation valves and penetration piping were designed and constructed to the ASME Code for Pumps and Valves for Nuclear Power Plants and ANSI B31.7, respectively.
. 9.5.2.2 Auxiliary Building 9.5.2.2.1 General Each of the following areas in the Auxiliary Building that contain equipment required to support PDM operations is served by a sepatate HVAC or ventilation system: 'i
- 1. Control Room and Technical Support Center
- 2. Radiochemical and Service Areas
- 3. Radwaste and Fuel Storage Areas
- 4. Electrical Equipment, Switchgear, and AC/DC Panels
- 5. Coolant and Miscellaneous Waste Tanks
- 6. Auxiliary Building Exhaust Air Filtration System
- 7. Communication Room Flow diagrams of the Auxiliary Build 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 Suppon 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 Suppon 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 HVAC 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 fresh and re-circulated 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 l i
Amendment 1 9.5-7 4
9.5.2.2.2 Control Room and Technical Support Center (Continued) and associated computer rooms and office spaces. The air / mixing ratio is automatically controlled to maintain the preset temperatures within each zone. The system also maintains positive pressure in the rooms, forcing the excess air to leak to the adjoining areas.
The normal control room air-handling unit is automatically de-energized when a loss of airflow is sensed in the normal HVAC duct or when the 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.
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 radiochemical 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 i 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 DS AR 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 cooler, and a pre-filter. This system is shown in Figure 9.5-4.
The exhaust system for the radiochemical and service areas and the radwaste and fuel storage areas consists of two units, each sized to handle 1007c of the ventilation load. Each unit contains fan and filter sections, including a pre filter and a HEPA filter. The cleaned and decontaminated air is discharged to the atmosphere through the plant vent. The radwaste and fuel storage area ventilation system is a Seismic Category II system, is normally functioning, and acts as the back-up spent fuel pool cooling system during the PDM as described in DS AR Section 9.2.
Amendment 2 9.5-8
I 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 iocated 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.
The fuel storage area air exhaust is directed to the intake plenum of the fuel storage area and radwaste exhaust system. A ventilation exhaust unit failure, which is detected by pressure sensors, results in an auto-stan 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 poweris not required in the PDM, the battery room ventilation system is not required to function and is not described in the DS AR.
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.
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 Amendment 1 9.5-9
9.5.2.2.6 Coolant and Miscellaneous Waste Tanks (Continued) 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 Il ventilation system. The system 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 of this system would result in any released radioisotopes being confined to the lower level of the Auxiliary Building.- -
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. 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 across the grade level vent duct.
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.
9.5-10
.. ._ _ . .. _.. _ _ _ ---_ _ _ _ _.-. _ _ . _ . . _ . _ _._. ~ . _ ._. _- _ __ _ . - _ ..- ... _ --.
4 TABLE 9.5-2 SINGLE FAILURE ANALYSIS FOR THE RADWASTE TANKS l P
VENTILATION AND SAMPLING SYSTEMS 1
Component Malfunction Comments and Consequences ;
l l
1 A. VENTILATING SYSTEM
- 1. Plant Vent Fan Fails to operate Pressure switch at fan inlet starts I other fan. Ventilation is not I interrupted. l l
l l
' 1
- 2. Motorized Fails closed Pressure switch at fan inlet starts Isolation Damper other fan. Ventilation is not interrupted.
1' 3. Plant Vent Become diny or Increase in pressure in upstream Filters clogged of filters is alarmed in Control Room and operator starts other fan. Ventilation rate is only briefly slowed prior to stan of 2'
attemate fan.
1 3-9.5-11
_ - _ , _ . . . _ - _ . . ., ~ _ . _ _ . - - _ . . ~-
9.5.2.3 Nuclear Service Electrical Buildine (NSEB)
Both the 'A'and B' side normal NSEB HVAC system is required during the PDNI to support the electrical equipment that functions to supply normal power to some of the Auxiliary Building equipment required to function during the PDNI. The 'A'and 'B' side NSEB normal HVAC system is shown in Figure 9.510, sheets 1 and 2. Only portions of the normal NSEB HVAC system are maintained functional during the PDN1. 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 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> duration each. These specifications will not be exceeded if the 'A'and 'B' side normal HVAC system operates.
9.5.2.4 Batterv Building 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 PDNI.
Also, only one train of Battery Building ventilation equipment is required to function during the PDN1 in accordance with the 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 PDNI.
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 shaft-mounted fan.
9.5.2.5 Interim On-site Storage Buildine (IOSB)
The IOSB radioactively clean areas (control room, records room, frisk area, and count room) are I served by a thermostatically controlled roof-mounted heat pump unit with a supply and retum duct and outside air make-up. Figure 9.5-1I shows the flow path for the IOSB HVAC system.
l i
9.5-12 l
9.5.2.5 Interim On-site Storage Buildine (IOSB)(Continued)
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 Building 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.
9.5.2.7 Codes. Standards. and Tests The work, equipment and materials confomi to the following codes and standards as applicable:
A. American Society of Heating, Refrigerating and Air Conditioning Engineers (ASHRAE) Handbook of Fundamentals and Guides !
l B. Air Moving and Conditioning Association (AMCA)
)
1 C. Air Conditioning and Refrigeration Institute (ARI) 1 1
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 attematively. The normal and other HVAC systems were designed as Quality Class 3.
9.5-13
9.6 FUEL HANDLING SYSTEM 9.6.1 DESIGN B ASES 9.6.1.1 General System Function Only the Fuel Storage Building fuel handling equipment is required to function during the PDM.
This equipment is designed to safely and effectively handle and move fuel within the spent fuel pool. The system is designed to minimize the possibility ofimproper operation that can cause fuel assembly damage and/or potential fission product release. The system was designed in accordance with NRC Safety Guide 13 as discussed in DSAR Section 1.6.13.
The fuel handling equipment is designed to handle the spent fuel assemblies under water, even when the spent fuel is eventually placed in a cask for removal from the Fuel Storage Building.
Underwater transfer of spent fuel assemblies provides an effective, economical, and transparent radiation shield. The spent fuel pool water also serves as a reliable cooling medium for removal of decay heat emanating from the spent fuel assemblies. Borated spent fuel pool water is not required to ensure suberitical conditions during fuel storage or movement in the spent fuel pool.
i 9.6.1.2 Spent Fuel Storage Pool The spent fuel pool is a reinforced concrete pool,is located in the Fuel Storage Building, and is lined with stainless steel. The pool is sized to accommodate 1080 spent fuel assemblies in high density storage racks. Fuel Assembly control components are also stored in the spent fuel pool until their eventual disposal. Additional spaces are provided for the storage of four failed fuel containers in the spent fuel pool.
The high density spent fuel racks consist of individual cells that have an approximately 9" X 9" square cross section. Each cell can accommodate one fuel assembly with or without an inserted control component. The cells are arranged within rack modules that contain a varying number of cells, which have 10.50 inch center to center spacing. A total of 1080 cells are arranged in 11 distinct modules. Thea high density spent fuel storage racks employ a free-standing and self-supporting rack design. A borated flexible polymeric neutron absorber (Boraflex)is sandwiched between double stainless steel sections that comprise the rack walls.
The high density racks are engineered to achieve the dual objective of maximum protection against structural loadings (arising from ground motion, thermal stresses, etc.) and the maximization of available storage locations. In general, the modules were made as wide as possible, within the constraints of initial delivery to Rancho Seco and site handling capabilities, to provide as great a margin as possible against rigid body tipping.
The modules are not anchored to the pool floor, to each other, or to the pool walls. A minimum gap of 2.0 inches is provided between the modules to ensure that kinematic movements of the modules during the Design Basis Earthquake will not cause inter-module impact, or violate the 9.6-1
9.6.1.2 Spent Fuel Storage Pool (Continued) minimum distance required to ensure adequate margins for nuclear sub-criticality. Adequate clearance with other pool hardware, e.g. cask catchers, pool elevator, etc. is also provided.
In accordance with NRC acceptance criteria, the Rancho Seco high density spent fuel storage racks are designed to assure that a Ker equal to or less than 0.95 is maintained with the racks fully loaded with fuel of the highest anticipated reactivity and the spent fuel pool flooded with
- un-borated water at a temperature corresponding to the highest reactivity. The maximum calculated reactivity includes a margin for uncertainty in reactivity calculations and in mechanical tolerances, statistically combined, such that the true kar will be equal to or less than 0.95 with a 95% probability at a 95% confidence level.
l 9.6.1.3 Spent Fuel Pool Water Chemistry During the PDM, Rancho Seco implements a chemistry control program for the spent fuel pool in accordance with the Permanently Defueled Technical Specifications (PDTS). The basis for the Chloride and Fluoride chemistry limits specified in the PDTS is derived from the B & W Water Chemistry Manual - Dual Type Plants, B AW-1385, April 24,1974. The PDTS limits for Chloride and Fluorirle ensure the potential for degradation of fuel assemblies, storage racks, and the spent fuel pool liner is minimized during long-term storage of spent fuel.
9.6.1.4 Fuel Transfer Tube Two horizontal tubes connect the Fuel Storage Building to the Reactor Building. These tubes have gate valves on the Fuel Storage Building side and a flanged closure on the Reactor Building side. The fuel transfer carriages are stored on the Fuel Storage Building side in the upender pits.
l The fuel transfer tubes are not required to function and remain closed during the PDM.
9.6.1.5 Fuel Handling Equipment 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 l transfer mechanisms, and shipping casks.
9.6.2 SYSTEM DESCRIPTION AND EVALUATION 9.6.2.1 Handline Spent Fuel Assemblies 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
9.6.2.2 Handline and Loadine Spent Fuel Casks l
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.
l l
Figure 9.6-1 shows the configuration of the gantry crane and its relation to the Fuel Storage Building. When handing the cask, the gantry crane hoist trolley is positioned on the east cantilever extension near the end of travel. Easterly travel over spent fuel storage racks is prevented first by automatic interlocks and then by rail stops and bumpers.
From a point directly above the fuel loading pit the gantry crane can only be moved in a nonherly .
direction when the crane trolley is on the cantilever. This control feature prevents the hoists from i being positioned over the spent fuel storage racks south of the fuel loading pit. This travel limit is accomplished by an automatic interlock. A single axis interlock permits crane motion in only one direction at a time when the cask is being moved within the Fuel Storage Building.
The crane trolley and bridge are in a mandatory slow zone whenever the trolley is located on the east cantilever. The slow zone restricts the maximum speed to approximately one-quarter the designed full speed. Limit switches that track the height of the main sister hook serve as a
, backup to administrative limits to ensure analyzed cask drop heights are not exceeded.
i The gantry crane can be operated from a cab positioned on the southeast leg of the gantry or with a portable radio control unit. Radio control is the preferred method of crane operation during the fuel casking effort because it provides direct observation of load movements inside the Fuel
' Storage Building. The radio control design is narrow-band digital frequency modulation. This design effectively blocks spurious signals from external sources that could result in non-commanded motion. The crane is shut down if the receiver cannot discern a proper digital command from the radio transmitter. Comprehensive testing of the radio control system was conducted along the cask load path to prove the functionality of the radio control system.
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 after the cask has been removed froin the spent fuel pool.
Amendment 3 9.6-3
1 9.6.2.3' Crane Use in Fuel Handline 1
1 l
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: l
- 1. American Institute of Steel Construction specification for the Design, Fabrication I
and Erection of Structural Steel.
)
- 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 ofIndustrial Relations, Sub-Chapter 7, General Industry Safety Orders, Group 13, Cranes and Other Hoisting Equipment.
- 5. US AS 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 pet formed was in accordance with the most advanced practice for this class of equipment. All materials fumished and work performed were in accordance with the requirements of the latest ANS,IEEE, ASTNI, ASNIE, HEI, and NENIA Specifications. In addition, the completed equipment complies with all applicable requirements of Federal and National board codes, Califomia 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 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 its rated load. All hooks were completely radiographed in accordance with ASTN! E-94 and ultrasonically tested for intemal defects prior to load testing.
Amendment 3 9.6-4
4 9.6.2.3.1 Design (Continued)
The hoist is provided with three means of braking, two mechanical and one electrical. There is one mechanical brake located on the motor shafts and one on the high speed shaft 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 lowering feature in the event of i
a power failure and the simultaneous failure of both mechanical brakes. This feature causes the hoist motor to be automatically placed in a generator mode, and a resistor rack absorbs the power produced by the hoist motor. 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 rated capacity of the east cantilever is 130 tons. The capacity of the l auxiliary hoist is 35 tons.
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.
i 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. ;
l Design features that 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 I hoist motor failure.
B. Redundant fail safe mechanical and electrical brake systems rated at 150 percent of full motor torque, C. Dead-man switches, l
- D. Wheel stops and bumpers designed for stopping gantry crane travel at rated speed L 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, 1
l Amendment 3 9.6-5 l
l f 9.6.2.3.2 Evaluation (Continued)
G. The righting moment shall exceed overturning moment by at least a ratio of l 1.5/1.0 under the most extreme loading conditions, l
H. The worst stress on any mechanical part shall not exceed 90 percent of yield stress under breakdown conditions or a locked rotor torque, I. Collector shoes to clear tracks of obstructions, and J. Temperature sensors on each hoist motor that shut down the crane if the mo or temperature rises above design parameters.
All gantry crane mechanical parts are designed, as a minimum, with a safety factor of 2.5 on yield strength.
In preparation for fuel casking, the District renovated and re-inspected the Turbine Building
- gantry crane.' The renovations and inspections included the following:
- 1. Replaced the drive controls for the gantry, trolley, and both hoists.
- 2. Added a gantry crane radio controller.
l 3. Installed a new digital programmable limit switch to precisely measure and limit i l main hook height. l l
- 4. Added mandatory bridge and trolley slow zones that are active when the trolley is on the east cantilever.
- 5. Modified the gantry crane control logic to include a single-axis interlock. This interlock prohibits moving loads inside the Fuel Storage Building in more than one direction at a time.
l 6. Replaced the main hoist wire rope.
- 7. Destructively tested a piece of the new wire rope to ten times the load of the heaviest loaded cask.
- 8. Inspected the main hoist upper and lower sheaves, main hook. and main hook nut using NDE techniques.
- 9. Load tested the gantry crane and both hoists to 125% of their rated capacities in accordance with Title 8 of the Califomia Code of Regulations.
Amendment 3 l
9.6-6 L
t
.yp w+-r-
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- 9.6.2.3.2 Evaluation (Continued)
Figure 9.6 2 shows a schematic cross section of the Fuel Storage Building and the yard area adjacent to the railroad siding. Possible drop areas and maximum possible drop heights are listed
! in Table 9.6-1.
. In establishing these maximum foreseeable drops, consideration was given to the constraints
- imposed by the equipment that will be used for transfening spent fuel casks from the cask loading pit to a railroad flatcar. The concept is based on the General Electric IF-400 cask, j conservatively assumed to measure 20 feet from the bottom of the cask to the attachment point
- on the lifting bail. The maximum vertical trasel 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 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 the spent fuel pool. The short sling is used from the intermediate ledge to the flatcar. Positioning the cask over the j railway spur without first lowering the hook to elevation 68 feet,6 inches (outside the Spent Fuel
- Building)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 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.
9.6.2.4 Safetv Provisions i Safety provisions are designed into the fuel handling system to prevent the development of hazardous conditions in the event of component malfunctions, accidental damage, or operational i
and administrative failures during fuel movement activities.
- The fuel storage racks are designed so that it is impossible to insert fuel assemblies in other than 4
the prescribed locations, thereby insuring the necessary spacing between assemblies. During the PDM, a criticality accident during fuel storage or movement is not considered a credible event.
Fuel handling equipment is designed to minimize the possibility of mechanical damage to the 4
fuel assemblies during fuel movement operations. If fuel damage should occur, the amount of radioactivity reaching the environment will present no hazard. The fuel handling accident is analyzed in DS AR Chapter 14.
All spent fuel assembly transfer operations are conducted underwater. The water level in the spent fuel pool is maintained at a minimum level to ensure at least 9 feet of water is maintained over the active fuel line of the spent fuel assemblies during either storage or movement of spent fuel assemblies.
9.6-7 e.- , ~- . ~ . . . _ . . _ , , , _ . _ . . - - - - . - . e
k TABLE 9.6-1 FUEL CASK DROP FEIGHTS Location Maximum Drop r
l Above the loading pit 33 ft. 6 in.'
l Above the intermediate ledge 26 ft.-6 in.'
l-i f'
i Above the cask washdown structure 4 inches Above the main steam lines immediately outside the building i ft.-6 in.
Above the remaining yard area to the railroad siding 16 ft.-6 in.
Above the flatcar on the railroad siding 26 ft.-0 in.-
l l \
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- Through water Amendment 2 L
I: l l
9.6-8
_ - . _ . _ _ ~ . . , . , , . . -
9.6.2.4 Safety Provisions (Continued)
Ntaintaining a minimum of 9 feet of water shielding over the spent fuel during all modes of PDN1 operations ensures adequate shielding is provided to protect operators or other personnel in the Fuel Storage Building.
The spent fuel pool water is normally cooled by the Spent Fuel Cooling system as described in DSAR Section 9.4. The Radwaste and Fuel Storage Building Ventilation system is the back-up spent fuel pool cooling system, is designed to run continuously, and is described in DSAR Section 9.5.2.2.4. If the normal cooling system should fail, greater than 15 days is available before the spent fuel pool temperature could reach its maximum equilibrium temperature of approximately l47*F. l A power failure during the PDN1 will create no immediate hazard situation, because, with no spent fuel pool cooling, a minimum of approximately 10 days are available for operators to
-restore cooling before the spent fuel pool temperature could reach its design basis limit of 180 F.
The spent fuel pool is completely lined with stainless steel plates for leak tightness and for case of decontamination. The fuel transfer tubes are attached to these liners to maintain leak integrity.
Administrative controls ensure (1) the fuel transfer tubes remain closed during the PDN! and (2) the spent fuel pool will not drain via this pathway. The spent fuel pool cannot be inadvertently gravity drained, since the water must be pumped out.
The fuel handling mechanism is designed so that the fuel and control component assemblics are drawn up into the fuel handling bridge mast tube for protection prior to transfer. Interlocks l prevent operating the fuel handling bridge or trolley until an assembly has been hoisted to the
- upper limit in the mast tube. NIandatory slow zones are provided for the hoisting mechanisms
! during insertion of fuel and rod assemblies. The slow zones are in effect during entry into a spent fuel storage rack position and just before and during bottoming of the fuel and rod assembly onto the spent fuel pool floor. The controls are appropriately interlocked to prevent simultaneous movement of the bridge. trolley, or hoist. The grapple mechanisms are interlocked with the hoists to prevent vertical movement unless the grapples are either fully opened or fully closed. The fuel grapple is so designed that when loaded with the fuel assembly, the fuel grapple cannot be opened as a result of operator error, or pneumatic system, electrical, or hydraulic
! failure. Hard stops prevent raising an assembly above minimum shielding depth in the event of an up limit failure.
The fuel handling system operating mechanisms are stored in the fuel handling and storage area for ease of maintenance and accessibility for inspection prior to start of fuel movement i
operations. All electrical equipment, with the exception of the position switches on the fuel I transfer tube gate valves, is located above the water level for greater integrity and ease of maintenance.
Suspected defective fuel may be tested for leakage. Leakage testing is performed in a container i that can be sealed and sampled for fission products.
1 Amendment 3 l
9.6-9
i 9.6.2.4 Safety Provisions (Continued) i The fuel handling bridge is limited to handling the fuel and rod assemblies only. Travel speeds t for the fuel handling bridge and hoist are mechanically controlled to ensure safe fuel handling l conditions, i
The 185 ton capacity turbine building gantry crane is electrically interlocked to prevent l movement of the trolley over the fuel storage rack area. These electrical interlocks, plus l administrative controls, prevent movement of the spent fuel shipping casks or other heavy loads over stored spent fuel assemblies.
NUREG-0612. " Control of Heavy Loads at Nuclear Power Plants", July 1980, provided seven guidelines and interim measures which are to be followed by all overhead handling systems and
- l. programs used to handle heavy loads near spent fuel stored in the spent fuel pool. These guidelines are identified in NUREG-0612 as follows:
Guideline 1 - Safe Load Paths Guideline 2 - Load Handling Procedures Guideline 3 Crane OperatorTraining Guideline 4 - Special Lifting Devices Guideline 5 - Lifting Devices (Not Specially Designed)
Guideline 6 - Crancs (Inspection, Testing, and Ntaintenance) i l l Guideline 7 - Crane Design l
An analysis of the overnead nardling systems at Rancho Seco was performed by SNIUD. The l
NRC and its consultant, the Franklin Research Center, reviewed the SNIUD analysis and l concluded that the above guidehnes 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 l (Inspection. Testing, and N!aintenance). The two remaining interim measures cover the following i criteria:
Interim N!easure 1: Heavy load technical specifications, and Interim Nieasure 2: Special review for heavy' loads handled over spent fuel.
l Rancho Seco complied with Interim N!easure 2 based on the District conducting a special review of loads handled over spent fuel.
9.6-10
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9.6.2.4 Safety Provisions (Continued)
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 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 is described below.
PDTS D3/4.3 specifies one exception to this heavy load limit. 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 components being designed and tested in accordance with-ANSI N 14.6-1986. 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 i
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1 Amendment 2 9.6-11
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- _X_ _X_ _X FIGURE 9.2-5 Amendment 3 SACRAMENTO MUNICIPAL UTILITY DISTRICT
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FIGURE 9.2-5, Component Cooling Water System !
4 I
Amendment 3
- TABLE OF CONTENTS Section Title Eagg IL RADIOACTIVE WASTE AND RADIATION PROTECTION 11.1-1 11.1 SOURCE TERM 11.1 1 11.1.1 RADIONUCLIDE INVENTORY I1.1-1
!!.l.1.1 Spent Fuel Assemblies 11.1-1
- 11.1.1.2 Reactor Vessel and Intemals and Concrete Piimary Shield 11.1-2 4
11.1.1.3 Elant Systems 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 Description 11.2-1 11.2.2 MISCELLANEOUS LIQUID RADWASTE SYSTEM i1.2 2 11.2.2.1 Funetion i1.2-2 11.2.2.2 System Description 11.2-2 l l
11.2.3 WASTE WATER DISPOSAL 11.2-11 l 11.2.3.1 Plant Efnuent 11.2 11 l 11.2.3.2 Normal Radioactive Discharee 11.2-12 l 11.2.3.3 Off-Normal Radioactive Discharge 11.2-14 11.2.3.4 Non-radioactive Waste Water 11.2-14 l 11.2.4 OPERATION, TESTING. AND INSPECTION 11.2 14 l 11.2.5 SYSTEM EVALUATION 11.2-15 l i1.2.6 PROCESSING WET RADIOACTIVE WASTES INTO SOLID 11.2-16 l RADIOACTIVE WASTE 11.2.6.1 Solidification and Dewaterine of Wet Radioactive Wastes 11.2-16 l Amendment 3 11-1
TABLE OF CONTENTS (Continued)
.Section Iit.Le hage 11.2.6.2 Drving of Wet Radioactive Waste 1 11.2-18 l
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11.2.6.2.1 Design Basis 11.2 18 l 11.3. GASEOUS WASTE MANAGEMENT SYSTEM 11.3-1
'l1.3.1 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 RADIOAGTIVE 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 Buildine (IOSB) 11.3-3 11.3.6 METHOD OF ASSESSMENT 11.3-3 11.3.6.1 Plume Exposure (Noble Gases) 11.3-3 l i- 'i1.3.6.2 Food Pathwav i 1.3-4 l l1.3.7 EVALUATION OF WASTE DISCHARGE 11.3-4 l 11.4 SOLID WASTE MANAGEMENT SYSTEM 11.4-1 11.4.1 DESIGN BASIS 11.4 1 1
11.4.1.1 8.coorts 11.4-2 l1.4.2 SYSTEM DESCRIPTION 11.4 2
[
11.4.2.1 Dry Solid Waste Disposal System / Process 11.4 2 11.4.2.2 Concentrated Liquid Waste Disposal System / Process 11.4-2 l 11.4.2.3 Spent Resin Disposal System / Process 11.4-3 11.4.2.4 Filter Disposal Process 11.4-3 l
Amendment 3 11-ii
- _ m __ m - -- - - .-
TABLE OFCONTENTS (Continued)
Section Title Eagg a
11.4.2.5 Solid Radwaste Storagg 11.4-3 11.5 RADIOACTIVE WASTE. EFFLUENT CONTROL. AND 11.5-1 ENVIRONMENTAL MONITORING PROGRAMS 11.5.1 DESIGN BASIS 11.5-1 11.5.2 OFF-SITE DOSE CALCULATION MANUAL (ODCM) 11.5-1 11.5.2.1 Liould Discharee Pathwav 11.5-1 11.5.2.2 Gaseous Discharge Pathway 11.5-2 11.5.3 PROCESS CONTROL PROGRAM (PCP) 11.5-2 l
11.5.4 R ADIOLOGICAL ENVIRONMENTAL MONITORING 1l'.5-2 11.5.4.1 Pre-operational REMP 11.5-2
!!.5.4.1.1 Pre-operational Exposure Estimation 11.5-4 11.5.4.2 Off-site Post-onerational REMP 11.5-8 11.5.4.2.1 Post-operational REMP Sampling Frequency 11.5-9 11.5.4.2.2 REMP Sample Types 11.5-10 l 11.5.4.2.3 REMP Sample Statistical Analysis 11.5 10 l 11.5.4.3 Effluent and Waste Disposal Environmental Reports 11.5-10 l 11.6 ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES 11.61 ARE AS LOW AS IS REASONABLY ACHIEVABLE (ALARA) j
-11.6.1 ALARA POLICY CONSIDERATIONS 11.6 1 '
11.6.2 ALARA DESIGN CONSIDERATIONS 11.6-2 11.6.3 ALARA OPERATIONAL CONSIDERATIONS 11.6 3 11.7. R ADIATION SOURCES 11.7-1 11.7.1 CONTAINED SOURCES 11.7-1 11.8 R ADIATION PROTECTION DESIGN FEATURES i1.8-1 11.8.1 FACILITY DESIGN FEATURES 11.8 1 4
4 Amendment 3 11-111
~ . . . . - . - . - - _ . - - . . - . - ~ . . - - - - _ _ . . . - . . - - ..
TABLE OF CONTENTS (Continued)
Section Iillo Pagg 11.8.2 SHIELDING 11.8-1 11.8.2.1 Design Criteria 11.8 l 11.8.2.2 Radiation Zone Classificatio_mi 11.8-2 11.8.2.3 Description of Shielding i1.8 2 11.8.2.3.1 Primary Shield 11.8 2 11.8.2.3.2 Secondary Shield ' 11.8 2 l 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 l1.8.2.3.6 Spent Fuel Shielding 11.8-3 11.8.2.4 Shielding Materials 11.8 3 l l 11.8.3 VENTILATION 11.8-5 I
(. 11.8.4 RADIATION c.IONITORING SYSTEM 11.8 5
. I1.8.4.l Desien Criteria 11.8 5 1t.8.4.2 Svstem Descriotion 11.8-5 l
.I1.8.4.2.1 Area Radiation Monitors 11.8-6 l 11.8.4.2.2 Process / Effluent Radiation Monitors 11.8-6 l 11,9 DOSE ASSESSMENT 11.9 1 11.9.1 PERSONNEL MONITORLNG l1.91 11.9.2 PERSONNEL EXPOSURE RECORD SYSTEM i1.9-1 l 11.9.3' MEDICAL EXAMINATION PROGRAM i 1.9-2 11.10 R ADIATION PROTECTION PROGR AM i1.10-1 Amendment 3 I-11-iv
. ~ - * .. , , - .---m-a . ~% e --m---.~.- . - - - - - - - -+ --------a- - - -
TABLE OF CONTENTS (Continued)
Section litls
_ Eags 11.10.1. OROANIZATION 11.10-1 11.10.2 EQUIPMENT, INSTRUMENTATION, AND FACILITIES 11.10-3 l 11.10.2.1 fersonnel Protective Eauipment 11.10-3 l 11.10.2.2 Radiation Protection instrumentation i1.10-3 11.10.2.3 Facilitiej 11.10-1 11.10.3 RADIATION PROTECTION PROCEDURES 11.10-5 11.10.3.1 Procedures _
11.10-5 11.10.3.2 Radiation Work Permit Procedus 11.10-6 11.10.4 PERIODIC PERSONNEL EXPOSURE REPORTING 11.10-6 11.11 REFERENCES 11.11 1 Amendment 1 1
11-v
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UST OFTABLES t'
Iills Table E;agg 11.2 1 Liquid Wastes Disposal System Component Data 11.2-3 11.2 2 Integrated Doses at the Site Boundary Resulting from Various 11.2 17 Tank Ruptures 11.5 1- Summary of Radiation Background for Sacramento, 11.5 3 Califomia 11.5-2 Results of the Rancho Seco Pre-operational 11.5-5 Surveillance Program 11.8-1 Principal Shielding i1.8-4 11.8 2' Area Radiation Monitors -11.8 7 4
11.8-3 Procers Radiation Monitors 11.8 9 l
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Amendment 3 11-vi l
i
! 1.2 1.IOUID WASTE TREATNIENT SYSTENIS The radioactive liquid waste treatment, disposal, and piping systems are housed within Seismic Category I structures, except for the Blender / Dryer radwaste processing unit, which is housed in the Solidification Building. The Sohdification Building meets the Standard Review Plan Section 11.4 seismic design requirements for portable solid waste systems. The seismic design criteria and analytical procedures for structures housing the liquid radwaste components, other than the Solidification Building, are provided in DS AR Section 5.3. Design information for the Solidification Building is contained in DSAR Section 5.5.8. hlajor components of the liquid waste system are tanks, pumps, ion exchangers, evaporators, and floor and equipment drains with associated pumps.
All components that could present a radiation hazard to plant personnel are contained within rooms or cubicles that are separated from operating areas and adjacent equipment by concrete walls. The liquid waste processing systems, except for the Blender / Dryer unit and associated transfer piping, are contained within the Auxiliary Building; therefore, leaks within the Auxiliary Building will be collected in sumps, and reprocessed through the hiiscellaneous Liquid Radwaste System. 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 remosed in accordance with Radiation Protection procedures.
. Radiation monitors are strategically located to warn of radiation within radwaste areas. A portable radiation monitor that locally alarms on high radiation is required in the Blender / Dryer
- unit area during Blender / Dryer operation, j The wastewater disposal system includes all tanks, pumps, piping, processing equipment, and associated instrumentation used to process waste water prior to discharge off site, ll.2.1 COOLANT RADWASTE AND REACTOR COOLANT DRAIN SYSTENIS 11.2.1.1 Functions l
The portions of the Coolant Radwaste System (CRS) and the Reactor Coolant Drain System ;
(RCDS) that remain in service during the PDNI are used to support radioactive wastewater l processing and convey radioactive wastewater from the non operating reactor coolant systems to !
the hiiscellaneous Liquid Radwaste System (NILRS), respectively. 1 11.2.1.2 System Description A drain path from the Reactor Coolant System (RCS) low point drains to the NILRS is provided to supply a means of monitoring evaporation in the RCS. Drainage from Reactor Building piping systems is conveyed from the RCDS to the NILRS for processing.
- Amendment I
- 11,2-1
-v re.mm ,-=-- - r- ., , - - . , , -
11.2.1.2 System Description (Continued)
Data for the various major radioactive liquid waste processing system components that are functional in the PDM are provided in Table i1.2-1.
I1.2.2 MISCELLANEOUS L:OUID RADWASTE SYSTEM 11.2.2.1 Function The Miscellaneous Liquid Radwaste System (MLRS) receives liquid of low quality from various tanks and systems. The MLRS is used to process and concentrate radioactive liquid in preparation for disposal. During the PDM, liquid concentrates processed in the MLRS are typically dried, using the Blender / Dryer radwaste processing unit, and disposed of as solid radwaste. The distillate from processed MLRS liquid is either recycled for plant use via the Demineralized Reactor Coolant Storage Tank or transferred to a Regenerant Hold-up Tank for eventual discharge off-site.
I1.2.2.2 System Description Miscellaneous waste liquids generated during the PDM are ultimately collected and processed in the MLRS. The major sources of liquid waste are:
- 1. Auxiliary Buildirig sumps,
- 2. Radiochemical laboratory drains,
- 3. Coolant Radwaste System,
- 4. Floor drains,
- 5. Systems and tanks containing radioactive liquid that are not required to function in the PDM,
- 6. Spent fuel pool liner leakage, and
- 7. Reactor Coolant Drain System.
The radwaste sump and the east and west decay heat room sumps have automatic level control.
As necessary, their contents are pumped to the Miscellaneous Weste Tank. This tank and the acid sump are directed to the Spent Regenerant Tank. Water in the Miscellaneous Waste Tank l can also be sent through the miscellaneous waste demineralizer to the Miscellaneous Waste Evaporator (MWE). l Amendment 1 11.2-2 L-- _
f l
Page 1 of 8 ,
TABLE 11.2-1 LIQUID WASTES DISPOSAL SYSTEM COMPONENT DATA Item No. Component Capacity Desian Material Vent Design Seismic l Name. (each) Temp, Press Body, Lining to Code Cat. ;
gal *F psig I l IANKS l
T-607A Coolant waste 60,000 200 14.9 55 None RD None I ;
T-6078 receiver tanks T-610 Coolant waste 60,000 200 14.9 SS None RD None I l
holdup tank i l l i T-621 Demin reactor 450,000 100 2 SS None Atm API-620 I l
coolant storage i l tank
'T-667 Miscellaneous 6,500 150 14.9 CS Amer- RD None II wastes tank coat 75 l T-669 Anti-foam tank 100 150 14.9 CS None AB ASME Sec III VIII T-674A Misc wastes 15,500 200 14.9 CS Plasite RD None II T-6748 condensate 7155 storage tanks T-679A Misc wastes 3,000 200 14.9 CS Plasite RD None II concentrate 7122 storage tank
-T-679C Misc wastes 60,000 200 14.9 SS None RD None I concentrate storage tank MT = Miscellaneous waste tank GH = Gas collection header RD = Radwaste area exhaust duct Atm = Atmosphere or atmospheric SS = Stainless Steel CS = Carbon Steel
( AB = Auxiliary Building RCS = Reactor Coolant System Amendment 1 11.2-3
i Page 2 of 8 TABLE 11.2-1 (Continued)
LIQUID WASTES DISPOSAL SYSTEM COMPONENT DATA Ittm No. Component Capacity Design Material Vent Design Seismic Name (each) Temp, Press Body, Lining to Code Cat.
gal *F psig LAMS (Continued)
V-681 -Spent resin 5,200 200 .150. SS None MT ASME..Sec II tank III/C T-6898 Spent 26,500 200 14.9 SS None RD None I regenerant tank T-704 Concentrated 60,000 150 14.9 SS None MT None I boric acid storage tank T-950A Regenerant 100,000 200 Atm CS Poly- Atm API-650 III hold-up tank ethyl-ene bag liner T-950B Regenerant 200,000 200 Atm CS Epoxy Atm API-650 III hold-up tank Amendment 3 .
1 11.2-4
11.2.2.2 System Descriotion (Continued) .
l Wastes in the Spent Regenerant Tank are processed through a combination filter-oil absorber prior to entering the MWE. The MWE is constructed of 316 stainless steel for corrosion 1
-resistance.
Miscellaneous wastes fed to the MWE pass through the integral gas stripper to remove any entrained gases, which are then released through the Auxiliary Building ventilation system. The CRS is cross-connected to the MWE so that the MWE may process reactor coolant radwaste.
- Also, radioactive wastewater collected in the RCDS is conveyed to the MLRS for eventual processing through the MWE. l Concentrated waste bottoms from the evaporator are pumped to one of two Miscellaneous i Wastes Concentrate Storage Tanks or the Concentrated Boric Acid Storage Tank prior to further l processing into solid waste in accordance with the Process Control Program. Solid radioactive i waste is stored in the IOSB and/or shipped to a NRC-licensed disposal facility. I J
Condensate from the evaporator is pumped through miscellaneous waste condensate !
demineralizers to one of two Condensate Storage Tanks.- MWE condensate can be pumped to the l Miscellaneous Water Hold-up Tank, the Demineralized Reactor Coolant Storage Tank, or the Spent Regenerant Tank.
i Figure 11.2-1 shows the simplified flow diagrams for the Miscellaneous Liquid Radwaste l System. Data on individual components in the system are provided in Table 11.2-1 above.
l The Borated Water System (BWS)is cross-tied to the MLRS to allow part of the BWS to be used to support MLRS operation. During the PDM, the Concentrated Boric Acid Storage Tank,its
. pump, and associated heat-traced piping is used to help process and store liquid radwaste. The heat tracing allows the Concentrated Boric Acid Storage Tank to receive and store highly 1 concentrated liquid radwaste similar to the Miscellaneous Wastes Concentrate Storage Tanks.
11.2.3 WASTE WATER DISPOSAL i1.2.3.1 Plant Effluent The plant liquid effluent consists of many sources of non-radioactive wastewater. Typical sources are:
- 1. Backwash from Service Water System (SWS) filters,
- 2. Storm drains,
- 3. Service Water System pump drains.
Amendment 1 11.2-11 4
.e -.. n -- - , -----,.._, ~,, . .... , , - - _. , < , < , n ,-,., g - -- , --
11.2.3.1 Plant Effluent (Continued)
- 4. Building area drains and roof drains,
- 5. Oily water separator,
- 6. Sewage treatment system,
- 7. Backwash from the reservoir strainer, and
- 8. ' Circulating water basin blow-down. 1 These waste streams converge at a mixing box located at the southwest comer of the Industrial Area. At this point the plant ef0uent is normally diluted with water supplied from Folsom South
! Canal. Dilution water may also be obtained from the Rancho Seco reservoir, which can be gravity drained to the plant effluent discharge stream downstream of the mixing box.
The total flow rate in the plant effluent stream, including the dilution water,is maintained at or i above the minimum annual average flow rate determined to give reasonable assurance of l compliance with the 10 CFR 50, Appendix I, dose guidelines. Various plant evolutions such as maintenance or testing may require the dilution water flow rate to be temporarily reduced.
However, the plant effluent system is operated to meet o.r exceed the minimum annual average i dilution flow rate. This plant effluent discharge forms the anificial headwater of Clay Creek.
i l The plant effluent stream contains instrumentation to monitor conductivity, pH, and l radioactivity. If any of these parameters are out of specification, the plant wastewater process l stream is automatically or manually diverted into one of two Retention Basins, which have a
. capacity of 500,000 gallons each. If the plant process stream flow is diverted, dilution water l
supplied upstream of the isolation valves is automatically stopped, but dilution water entering the plant effluent stream downstream of these valves continues.
11.2.3.2 Normal Radioactive Discharge l One source of slightly contaminated liquid that is discharged from the site originates from the
! draining of secondary plant systems not required to function in the PDM. Also, liquid collection i
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:
l 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.
11,2-12
. j
11.2.3.2 Normal Radioactive Discharge (Continued)
The capacity of the 'A'and 'B'RHUT is 100,000 gallons and 200,000 gallons, respectively.
,. Normally, the 'B' RHUT is lined up to receive plant wastes. The 'A' RHUT is maintained as a back-up wastewater collection source. The 'A' RHUT has a plastic liner and the 'B' RHUT has an epoxy coating for chemical resistance. Each has an agitator and a sample pump to ensure l representative sampling. )
Prior to transferring an RHUT's contents to a Retention Basin, the RHUT is isolated, re-circulated, sampled, and analyzed for chemical and radiological quality. Dose projections are made in accordance with the Off-site Dose Calculation Manual (ODCM) to determine compliance with the 10 CFR 50, Appendix ! dose guidelines. Actual off-site 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.
1 If the measured radioactivity content in an RHUT is high relative to the 10 CFR 50, Appendix I l dose guidelines, the RHUT may be processed with the Liquid Effluent Radwaste Treatment System (LERTS) to reduce the 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 allows adaptation to portable demineralizing equipment to ,
assist in processing a RHUT batch, if necessary.
When analysis for chemical and radiological quality of an RHUT's contents is complete, the RHUT water is directed to a Retention Basin via the RHUT transfer pumps. The RHUTs may also be gravity drained.
An effluent strainer prevents demineralizer/ 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 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 Retention Basin is used to determine the required release rate and/or dilution water flow rate necessary to ensure compliance with the liquid effluent 10 CFR 20 radioactivity concentration limits. In addition, the Retention Basin effluent is monitored by a process radiation 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.
Amendment 3 11.2-13
11.2.3.3 Off-Normal Radioactive Discharge Radioactit ely contaminated water is contained within plant equipment located in the Tank Farm.
Occasion.al, small-volume leakage from the tanks, pipes, valves, instrument lines, and other components is possible. Because there are storm drains in the area that are routed directly to the plant effluent stream, trace amounts of radioactivity may occasionally be released via this pathway.
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 ifit can be demonstrated that the doses are less than 1% of the 10 CFR 50, Appendix I annual whole body dose guidelines. Dose calculations are performed in accordance with the ODCM. The source term is based on grab sample analysis of the most representative process stream.
I1.2.3.4 Non-radioactive Waste Water The sources of non-radioactive wastewater, which require treatment prior to entry into the plant effluent stream, are the site sewage treatment system and storm drains.
Sanitary waste from the buildings in use at the Rancho Seco site are either gravity drained or pumped to the sanitary waste sump. Here, two sump pumps car. send the waste to a packaged digestion sewage treatment plant (Aer-O-Flo), where the waste can be treated and pumped to a 20,000 gallon aeration pond. Also, an overland sewage treatment system is available as an additional sewage treatment method. The overland sewage treatment system offers a nominal 36,000 gallons per day treatment capacity, including three separate overland flow terraces, flow control facilities, and related appurtenances.
The three processes described above are available for treating sewage generated at the plant during the PDM. Operating these systems in the desired combination provides flexibility and a high degree of reliability that Rancho Seco will produce good quality effluent from its sewage treatment process.
Storm drains in the transformer yard, transformer alley, and various roof drains are directed to the oily water separator prior to discharge to the plant effluent.
I1.2.4 OPERATION. TESTING, AND INSPECTION Operating procedures govem the. valve alignment and general operation of liquid waste systems.
Maintenance on components and plant operating needs dictate bypassing redundant processes or system trains.
The operation of systems and equipment during waste processing demonstrates the integrity of the systems. Inspections of the systems are conducted periodically to ensure that system performance is consistent. Routine inspections or checks are specified in plant procedures.
Amendment 1 11.2-14
11.2.4 : OPERATION, TESTING, AND INSPECTION (Continued)
Detailed test procedures and the schedule for tests are also included in plant procedures.
' Examples of checks and inspections include, but are not limited to:
- 1. Operation of automatic starting pumps
- 2. Closing or opening of automatic valves on alarm signals 1
- 3. Calibration of gauges and meters
- 4. Operation of alarms
- 5. Inspection of systems for leaks -
Reliable operation of the radwaste systems is assured by a design that uses components whose reliability has been demonstrated in prior commercial service. The selection of pumps, compressors, valves, etc., that are commercially available also ensures the availability of-maintenance parts and replacement units from multiple sources on short notice. These factors are typical of the many features that are incorporated into the radwaste systems to enhance their reliability and availability.
Piping to major process pumps that handle liquids from waste tanks is arranged so that failure of a single pump can not prevent pumping out the tank.
I1.2.5 SYSTEM EVALUATION The MWE is used to concentrate waste water collected in the MLRS. This concentrated waste water is then stored until a sufficient quantity is available for additional processing and eventual disposal as low-level solid radwaste in accordance with the Process Control Program. Distillate from the MWE is sampled and analyzed prior to transferring to the appropriate storage tank. The distillate from the MWE may be released to the environment through the liquid effluent pathway.
Before this may be accomplished, strict compliance with 10 CFR 50 and 10 CFR 20 is verified at the RHUTs and Retention Basins, respectively.
The liquid waste processing systems, except for the Blender / Dryer unit and associated transfer piping, are contained within the Auxiliary Building; therefore, leaks within the Auxiliary Building will be collected in sumps, and reprocessed through the Miscellaneous Liquid Radwaste System. 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 Auxihary Building sumps with manual actions or cleaned up in accordance with Radiation Protection procedures.
11.2-15
11.2.5 SYSTEM EVALUATION (Continued)
The radioactive waste disposal systems provide for the controlled handling and disposal of liquid, gaseous, and solid wastes. The systems are designed to ensure that plant personnel and the general public are protected against excessive exposure to radiation from radioactive wastes processed at the plant and discharged from Rancho Seco during the PDM. The systems are designed to minimize the discharge to the environment radioactivity that is of station origin.
The analysis of the environmental consequences at the site boundary for liquid efflue'n t due to component failure in the liquid waste systems is shown in Table 11.2-2. This analysis is bounding for the PDM. The potential source terms during the PDM for the listed tanks are significantly lower than those assumed in the tank rupture analysis for the whole body dose values presented in Table 11.2-2.
All known normal low-level radioactively contaminated liquid waste streams are directed to the RHUTs for analysis and to account for any radioactivity prior to transfer to a Retention Basin.
Prior to release off-site from a Retention Basin, effluent release rates and dilution now rates are determined in accordance with the ODCM to ensure compliance with 10 CFR 20. The Retention Basin release pathway contains a process radiation instrument that monitors the concentration of the final efnuent and can terminate the release if concentrations exceed the setpoint determined in accordance with the ODCM. All dose, MPC fraction, and monitor set-point calculations are performed in accordance with the ODCM.
11.2.6 PROCESSING WET R ADIOACTIVE WASTES INTO SOLID RADIOACTIVE WASTE Wet radioactive wastes generated at Rancho Seco that are to be disposed of at a licensed disposal facility is processed into an acceptable form by de-watering, drying, absorption, or solidification.
Processing is performed using equipment operated in accordance with the Process Control Program (PCP) and applicable implementing procedures. The PCP provides reasonable assurance that all processed wet radioactive waste will meet the applicable packaging, transportation, and licensed disposal site requirements.
11.2.6.1 Solidification and Dewatering of Wet Radioactive Wastes Solidification and de-watering are normally performed by a contractor, which has a NRC-submitted topical report w hich is either pending approval or has been approved which describes their solidification process. Site personnel may also perform solidification and de-watering activities in accordance with the PCP and applicable implementing procedures.
Cement, wax, bitumen, or polymers may be used in the solidification process. The solidification unit is specifically designed to optimize solidification of radioactive wastes, evaporator bottoms, ion exchange resin slurries, and sludges. The process results in stable waste forms in accordance with 10 CFR 61. Process parameters are established by the contractor in a PCP.
11.2-16
TABLE I1.2 2 INTEGRATED DOSES AT THE SITE BOUNDARY RESULTING FROM VARIOUS TANK RUPTURES Whole Body Tank Description Tank No. Dose (Rem)
Coolant was'te receiver tanks T-607 A or B 0.362 Coolant waste holdup tank T-610 0.313 Demineralized R.C. storage tank T-621 Negligible Spent regenerant tank T-689B Negligible Miscellaneous waste tank T-667 Negligible .
Miscellaneous waste evaporator V-696 Negligible Misc. waste evaporator condensate tank T-699 Negligible Misc. waste condensate storage tanks T-674A or B Negligible Misc. waste concentrate storage tank T-679A Negligible l
Concentrated boric acid storage tank T-704 Negligible
- Negligible designates a whole body dose I millirem.
Amendment 1 11.2-17
11.2.6.1 Solidification and De-waterine of Wet Radioactive Wastes (Continued)
The solidification unit is a portable system containing all piping, support, control, and monitoring equipment necessary to solidify or de-water radioactive liquids.
The unit is composed of several processing subsystems, each controlling a specific function of the process. These subsystems may include waste transfer, chemical addition, conveyor, and vent and de watering systems. Control functions for the unit are incorporated into the pneumatic and main control panels. Service supplies are distributed through the service air, water, and
- electrical distribution systems.
l l
Typically, most of the mobile unit components are a: Tanged in mobile trailers (skids) to provide flexibility of operations.
A closed-circuit television system is an integral part of the mobile unit and allows the operator to monitor the solidification process.
11.2.6.2 Drvine of Wet Radioactive Wastes A portable, skid mounted evaporator, known as the Blender / Dryer, takes MWE evaporator concentrates and liquid wastes from other sources (i.e., the Borated Water Storage Tank, other miscellaneous liquid radwaste tanks, and the spent fuel pool) and reduces the liquid wastes to dried salts. The distillate from the drying process is returned to the MLRS for additional processing in accordance with the Radioactive Effluent Control Program. A wax binder can be added to the dried salts, and the resulting solid waste is packaged in drums for eventual shipment to a licensed disposal facility.
I1.2.6.2 I Design Basis The seismic design of the Solidification Building for Blender / Dryer operation meets the design criteria of Standard Review Plan (SRP) Section 11.4, Solid Waste Management Systems, for portable solid waste systems. Also, the ventilation design provisions for operation of the Blender / Dryer in the Solidification Building meet the design criteria of SRP Section 11.4 for portable solid waste systems, as long as specified administrative controls are met.
The Blender / Dryer design incorporates several administrative controls into plant procedures to meet radioactive waste system design requirements and mitigate potential safety concerns associated with Blender / Dryer operation. Operating procedures contain the safety and design related administrative controls. Processing wet radioactive wastes by drying liquid wastes is addressed in the PCP.
The list of administrative controls that are considered part of the Blender / Dryer design and implemented to mitigate potential safety concerns are as follows:
11.2-18 l
L____ -
11.2.6.2.1 Design Basis (Continued)
A. The Blender / Dryer television camera must be operating during transfers of liquid waste or flush water into the Blender / Dryer.
B. Operators must monitor all liquid waste transi'ers to and from the Blender / Dryer so immediate action may be taken to mitigate the consequences of a leak or pipe rupture. ,
C. The activity contained in a batch of wet radioactive waste to be transferred and processed in the Blender / Dryer must not exceed a total of 10 Curies or a concentration of 0.5 Ci/mi, excluding tritium. At a concentration of 0.5 Ci/ml, the volume of one batch of wet radioactive waste must not exceed 5280 gallons to ensure the 10 Curie total activity limit, excluding tritium, is met.
D. The external doors to the Solidification Building must be closed during -
Blender / Dryer operation; except, the external doors may be opened for short periods of time, when necessary, to facilitate ongoing controlled area activities. j Also, the interior roll.up doors must be kept open, and at least one operating l Auxiliary Building ventilation system must be properly aligned to provide i ventilation to the Solidification Building during Blender / Dryer operation. These l ventilation configuration administrative controls are necessary to ensure the Blender / Dryer ventilation design requirements are met during Blender / Dryer operation.
E. A portable radiation monitor that alarms on high radiation is required in the Blender / Dryer area of the Solidification Building to meet radiation monitoring requirements for Blender / Dryer operation.
i F. The start up and operating procedure for the Blender / Dryer Chiller system must j ensure the Chiller is filled and re filled with non radioactive demineralized water.
G. The Blender / Dryer operating procedure that addresses transferring Blender / Dryer .
distillate to a demineralizer must address the valve line-up that is required before l l Blender / Dryer distillate discharge pump may be started. !
H. No more than one B/D batch worth of binder wax may be stored in the ,
Solidification Building at one time, because of fire protection concerns.
The Blender / Dryer design considers the simultaneous rupture of all tanks in the Blender / Dryer l j system with release of the entire liquid inventory to the floor of the Solidification Building. A !
bermed area around the Blender / Dryer is of sufficient size to contain the entire liquid inventory 1 of the Blender / Dryer and connecting pipes. Liquid collected in the bermed area can either be I l directed to Auxiliary Building sumps with manual actions or cleaned up in accordance with Radiation Protection procedures. l l
i I
l 11.2-19
. - - . , , , . , . ~ - - . , , , - - - .
11.2.6.2.1 ' Design Basis (Continued)
SMUD calculations document the dose consequences of a catastrophic failure during a radwaste system resin transfer and an evaporator bottoms transfer. The consequences of a resin transfer accident bounds the evaporator bottoms transfer accident by approximately three orders of magnitude. A Blender / Dryer liquid radioactive waste transfer accident is comparable to the evaporator bottoms transfer accident. Therefore, SMUD calculations bound the worst case accident for Blender / Dryer operation.
The following are codes, standards, and regulatory design guides and requirements that were incorporated in the design of the Blender / Dryer:
A. Design Codes and Standards
- 1. ASTM Standards 1 ' 1NSI B31.1-198'6 code for pressure piping
- 3. AWS DI.1-1972 code for structural welding
- 4. NEC-1987 code for electrical design B. Regulatory Design Guides and Requirements
- 1. Standard Review Plan Section 11.4
- 3. 10 CFR 20 i
- 4. 10 CFR 61 S. 49 CFR l
- 6. NRC Information Notice 7919
+
1
?
l 11.2-20 l
1 l
e
1 11.3 GASEOUS WASTE MANAGEMENT SYSTEAj 11.3.1 DESIGN BASIS l
The gaseous waste management system at Rancho Seco includes the administrative controls, equipment, and ducting necessary to control, mechanically filter, monitor, and convey waste gases evolved from the plant during the PDM. The predominant radioactive isotope discharged I from the plant in gaseous effluent during the PDM is Tritium. The only other isotope of significant gaseous effluent concern during ,ormal or accident release conditions in the PDM is Krypton-85.
Gaseous waste management system hard me includes the plant vents and their associated Ventilation Exhaust Treatment Systerr p a. riltration units) and radiation monitoring instrumentation. The plant vents incluce . ' % : tor Building Stack (RBS), the Auxiliary Building Stack (ABS), and the Auxiliary Emng Grade Level Vent (ABGLV). Normal gaseous effluent evolves from the various tanks and equipment that are either in use or are just storing radioactive liquid during the PDM. An accidental release of gaseous effluent is postulated from ,
a dropped fuel assembly in the Fuel Storage Building, which would release Krypton-85 gas from
{
the fuel rod gaps to the Auxiliary Building Stack.
l The Auxiliary Building Grade Level Vent mechanically filters, monitors, and discharges air from the grade level of the Auxiliary Building. Also, this ventilation system may be lined up to take suction on the Solidification Building, via openings between the two buildings at the grade level, to ensure gaseous efnuent evolved from concentrated wet radioactive waste processing in the i Solidification Building is controlled, monitored, and mechanically filtered prior to release.
The Reactor Building Stack mechanically filters, monitors, and discharges Reactor Building air during the PDM. All planned radioactive gaseous discharges are administratively controlled through the Off site Dose Calculation Manual (ODCM) and implementing procedures.
1 11.3.2 SYSTEM DESCRIPTION I Low activity gaseous effluent is conveyed from the radwaste area exhaust duct and discharged directly into the continuous exhaust plenum of the Auxiliary Building Stack ventilation exhaust treatment system. This system contains a HEPA filter, which filters the exhaust for particulates prior to release into the ABS vent. This gaseous effluent is monitored by the ABS radiation monitor, which annunciates in the control room on a high particulate radioactivity indication.
The low activity gaseous effluent sources are:
A. Vent gases from spent regenerant tanks B. Vent gases from miscellaneous waste tanks C. Vent gases from coolant waste holdup tanks 11.3-1
11.3.2 SYSTEN! DESCRIPTION (Continued)
D. Vent gases from coolant waste receiver tanks E. Niiscellaneous Waste Evaporator (N1WE) strip gas F. Blender / Dryer processing unit strip gas During the PDN1, there are no normal sources of high activity gaseous effluent. Only a dropped fuel assembly accident could result in the release of high activity gaseous effluent.
The allowed gaseous efnuent release rates from the plant vents are determined in accordance with the Off-site Dose Calculation Nianual to assure compliance with the 10 CFR 20 gaseous effluent dose limits. The gaseous effluent dose rate limits provide reasonable assurance that radioactive material discharged in gaseous effluent will not exceed the annual average concentrations specified in 10 CFR 20, Appendix B.
I1.3.3 HYDROGEN GAS NilXTURES
- Since hydrogen gas is not used in the PDNI, control of hydrogen gas mixtures in plant systems is not necessary and is not addressed in the DSAR.
l 11.3.4 OPERATION, TESTING, AND INSPECTION l
l System and equipment operation during gaseous effluent discharge demonstrates the integrity of the systems. Testing of the systems is conducted periodically in accordance with the ODCN1 to l
ensure that system performance is consistent with performance objectives.
Routine tests, checks, and periodic evaluations of gaseous effluent discharged from the plant are specified in the ODChi and plant procedures. Detailed test procedures and the schedule for tests are also included in the ODCN! and plant procedures. Examples of checks, tests, and evaluations include, but are not limited to:
A. Closing or opening of automatic valves on alarm signals B. Calibration of gauges and meters l C, 31-day dose projections D. Operation of all alarms E. Calibration of radiation monitors 1
11,3-2
11.3..' LADIOACTIVE RELEASES tl.3.5.1 Pathways i
The pathways for the release of radioactive gaseous effluent are the following: l A. Reactor Building purges l B. Auxiliary Building ventilation via the Auxiliary Building Stack C. Auxiliary Building ventilation via the Auxiliary Building Grade Level Vent l
11.3.5.2 Secondary Plant Contamination l
Because the secondary plant is mostly laid up in the PDM, miscellaneous gaseous releases are not expected from this source.
I1.3.5.3 Interim On-site Storace Buildine (IOSB)
The IOSB is another possible source of a miscellaneous gaseous release. Because low-level l radioactive solid waste is handled in the IOSB, a possibility exists for a release of particulates to the atmosphere. For this reason, the IOSB exhaust is normally continuously monitored. The IOSB particulate radiation monitor has control features to isolate an airbome releases if radioactivity being released exceeds the monitor set-point. If the IOSB particulate radiation monitor is out-of-service, activities which could lead to a gaseous release are ceased or a particulate grab sampler is put into service to monitor the IOSB release path. No routine releases of gaseous effluent are expected. Particulate monitoring instrumentation surveillance, operating, and sample analysis requirements are specified in the ODCM. A detected IOSB gaseous release (particulate) would be evaluawd on a case-by-case basis as a miscellaneous release in accordance with the ODCM.
I1.3.6 METHOD OF ASSESSMENT 11.3.6.1 Plume Exposure (Noble Gases)
The design basis mathematical model used for external whole body dose assessment is based on l " Semi Infinite Cloud" assumptions, using the following conventional formula:
gD',(x,y,0.t) = 0.25 Eg X Qb Q
where:
1 11.3-3 l
I1.3.6.1 Plume Exoosure (Noble Gases)(Continued)
- D',,(x,y,0,t) = Gamma dose rate (rad /sec) to a person at location (x, y,0) at time t from I a semi-infinite cloud Es = Average gamma energy per disintegration (Mev/ dis)
X = Default atmospheric dispersion factor (m/sec')
Q Q'o = Release rate of gamma emitting isotope (Ci/sec)
The dispersion factor used to calculate the expected whole body dose from normal gaseous effluent discharged from the plant vents during the PDM is a conservative default value set at 1.00 x 10" m/sec 3. For releases of gaseous effluent following an accident, a dispersion factor of
,3.6,1 x,10 m/sec' is used. To be. conservative, cloud depletion rates are assumed to be zero.- -
For noble gases, the models are based on the dose a person would receive if they were surrounded by a semi-infinite cloud of radioactive gas. A much higher external dose will be delivered from a cloud of radioactive noble gas rather than from noble gas that is held in the lungs or other body organs.
l1.3.6.2 . Food Pathway
[
l Since Krypton 85 is not known to react chemically with other substances, Krypton 85 is not expected to be present in the food chain. The particulate daughter product of Krypton-85 is , >
Rubidium-85, which is stable.
11.3.7 EVALUATION OF WASTE DISCHARGE The release points for gaseous effluent conveyed to the RBS and ABS are at the top and the northwest side of the Reactor Building, approximately 145 feet above grade level. The RBS, which is used for Reactor Building purging during the PDM,is adjacent to the ABS. The discharge of the ABGLV is located above the 20 foot level of the Auxiliary Building and is used to vent Auxiliary Building grade level areas which may occasionally contain airborne radioactivity and which may not be vented through the ABS.
l L
11.3-4 l .*
11.4 SOLID WASTE NIANAGENIENT SYSTENt i1.4.1 DESIGN BASIS )
Solid radioactive wastes are collected and processed on a batch basis in accordance with an ,
approved Process Control Program (PCP) and applicable government regulations. Solid wastes !
are packaged in containers which conform to Department Of Transportation (DOT) requirements (49 CFR) for disposal at a licensed disposal facility.
Solid waste disposal is handled in a manner that minimizes exposure to site personnel during collection and packaging and protects the public health and safety during transport of the solid wastes to a licensed disposal site. The amount of solid waste produced during the PDNI is much less than that produced during power plant operation. The majority of the solid waste generated during the PDNI is from processing wet radioactive wastes into solid wastes and the decontamination work related to preparing contaminated areas and system around the plant for a long-term safe storage condition. Solid wastes shipped off-site will be below maximum allowable radiation levels for transportation of radioactive materials as specified in 10 CFR 71 j and 49 CFR part 170 through 199.
Niost of the solid waste disposal system equipment and piping is housed within the Seismic :
Category I Auxiliary Building. Of the wet radioactive waste processing methods (i.e., 1 dewatering, drying, absorption, and solidification), drying and solidification are conducted in the Seismic category II Solidification Building.
The major components of the solid waste disposal system that are housed in the Auxiliary Building include a compactor with associated controls, and various piping for the transfer of wet radioactive wastes for processing in the Solidification Building. The compactor is contained within a room which is separated from adjacent equipment and nomially occupied areas by concrete walls.
The solid waste disposal system is operated in accordance with the PCP and applicable Radiation Protection and Operations procedures. The solid waste processing program provides for the transfer of wet radioactive wastes to the Solidification Building, the packaging of radioactive solid wastes in accordance with the applicable governing regulations, and the shipment of solid radioactive wastes to a licensed disposal facility. The solid waste packaging process is performed in a manner which minimizes the radiation exposure to individuals and minimizes the volume of packaged waste.
To facilitate the shipping process for solid radioactive waste, interim on-site storage is provided in a building away from the location where wastes are processed. The Interim On-site Storage Building (IOSB) allows for storage, segregation, consolidation, decontamination, encapsulation, and preparation for shipment of low level waste in an area that minimizes operator exposure while providing for efficient and safe handling of radioactive materials.
11,4-1
11.4.1.1 Renons The total curie content and major radionuclide composition by waste type are reported in the Radioactive Effluent Release Report required pursuant to 10 CFR 50.36a and the plant Technical Specifications.
11.4.2 SYSTEM DESCRIPTION The solid waste disposal system is divided into four subsystems:
- 1. The dry solid waste disposal system in which dry active waste (DAW) is either compacted into 55 gallon drums or packaged into drums or boxes as appropriate.
- 2. The concentrated liquid waste disposal system which provides for the transport of concentrated radioactive wastes to the Solidification Building, where the3yet .
wastes are processed into solid waste in accordance with the PCP.
- 3. The spent resin waste disposal system which provides for the transport of spent resins to the Solidification Building.
- 4. The filter disposal process which provides for the disposal of radioactive filters used in various systems within the plant.
11.4.2.1 Dry Solid Waste Disposal Svstem/ Process Compactable DAW is normally packaged and compacted into 55 gallon drums. Any DAW which cannot be compacted is packaged into an appropriate container Waste packaging is performed in a manner which minimizes radiation exposure to the workers and maximizes the available container usable volume. The compactor is available in the compactor room, which is located in the Auxiliary Building at grade level. Compaction of waste is performed on a batch basis. As soon as a sufficient amount of DAW has accumulated from plant operations during the PDM, the solid waste is processed and compacted into 55 gallon drums. Compaction may also be performed off-site by a solid radwaste volume reduction vendor.
I1.4.2.2 Concentrated Liauid Waste Disposal System / Process The concentrated liquid waste disposal system / process consists of piping and the associated pumps, valves, and controls necessary to transport wet radioactive wastes stored in various liquid radwaste tanks to the Solidification Building for processing into solid waste. Wet radioactive wastes are processed into solid wastes in the Solidification Building using the Blender / Dryer processing unit or vendor supplied solidification equipment in accordance with the PCP.
Adherence to the PCP will ensure compliance with the applicable regulations for packaging, transportation, and disposal of the solid waste.
11.4-2
11.4.2.3 Spent Resin Disposal System / Process The spent resin waste disposal system / process consists of a piping manifold and the associated valves and controls needed to transport spent resin stored in the spent resin tank to the Solidification Building. These resin wastes originate in various systems that used resin beds to maintain the appropriate chemical environment in systems that operated during nuclear power plant operations. Spent resin wastes are transferred to the solidification Building where they are processed to comply with the applicable regulations for disposal and transportation. The I processing of spent resin wastes is performed in accordance with the approved PCP, which ensures the solidification / dewatering of spent resin wastes will comply with packaging, transportation and disposal facility requirements.
I1.4.2.4 Filter Disposal Process Mechanical filters of varying types are used in several plant systems that handle radioactive materials. The majority of these filters are located in the -20 foot elevation of the Auxiliary Building. Filters that become loaded with corrosion products (crud) are occasionally changed out due to excessive differential pressure across the filter or high radiation levels. These filters are placed into an appropriate storage container (normally a 55-gallon drum) with appropriate I shielding or a shield assembly depending upon the radiation level. The filters are packaged for disposal depending upon their construction and/or radiation level. The packaged filters are
(
prepared for disposal in a manner that ensures compliance with transportation and disposal site regulations and requirements.
I1.4.2.5 Solid Radwaste Storace The IOSB is capable of storing at least 21/2 years of accumulated waste consisting of appropriately processed wet radioactive wastes, compactable trash, and resins. In addition to
- storage, the following activities may be performed in the IOSB
! A. The sorting and segregation of DAW, l B. The consolidation of un-compactable waste prior to shipment, C. Laundry inspection, survey, sorting, and storage, D. Respirator unit cleaning and filter testing operations, l E. Decontamination of equipment and/or volume reduction of fixed contaminated equipment and gear, and k F. Encapsulation of solid radioactive materials.
i Amendment 3 11.4-3 l
11.4.2.5 Jolid Radwaste Storage (Continued)
' The IOSB materials handling system is a multifaceted system designed to receive containerized radwaste from the plant radwaste facilities and to place containers in appropriate storage locations within the IOSB for eventual retrieval and off site shipment of the stored waste. The system was designed to keep operator exposures ALARA while providing for efficient and safe handling of radioactive materials.
The IOSB allows for storage of waste in two basic configurations, depending on the activity level. The higher activity. waste may be stored in a shielded, covered cell arrangement designed to accommodate a range of waste containers from 300 cubic foot disposable liners to small drums. The storage cells have individual shield covers. Storage cell cover and waste container handling is accomplished by an overhead, remotely operated bridge crane system. The storage cell area and open floor warehouse storage area are covered by a roof to protect the facility and solid radwaste from environmental effects.
Low activity waste is stored in a shielded, open floor warehouse arrangement designed to accommodate a range of waste containers from 55 gallon drums to 120 cubic foot metal bins.
The overall facility shielding design allows for normal radioactive materials handling. Shielding takes into account expansion of the facility to accommodate storage of the equivalent of up to 10 years of solid waste that could be accumulated. Under maximum design radioactivity loading, IOSB shielding is designed to ensure the radiation dose rate from the facility at the Industrial Area is maintained within the limits set forth in 10 CFR 20 and 40 CFR 190. Administrative controls are incorporated into p! ant procedures that address placement of solid waste within the IOSB in such a manner as to minimize shielding requirements.
The storage facility is designed as an unmanned facility, except during waste storage, processing, or shipping operations. The facility storage areas and the waste materials handling systems are designed to provide for case of container retrievability. The ability to remotely remove and examine a container for integrity is considered in the design.
The IOSB cell area is provided with a gravity drainage and sump system consisting of trench drains to a closed sump. The sump has manual sampling capabilities and a pump for recirculating and discharging the sump contents. Processing the sump contents is accomplished by plant liquid radwaste treatment systems.
The DAW storage area and the DAW handling area drains are routed to the cell drain collection sump.
The IOSB has a building ventilation system, which provides protection against the release of radioactive airborne material in particulate form. The IOSB is constructed of non-combustible materials and has a fire protection system that is designed to be expandable to accommodate increased facility sizing for up to the 10 years worth of solid waste accumulation as described above.
11.4-4
l 11.5 R ADIOACTIVE WASTE. EFFLUENT CONTROL. AND ENVIRONNIENTAL NIONITORING PROGRANis 11.5.1 DESIGN B ASIS l
The radioactive waste and effluent control programs provide the administrative controls 1 necessary to ensure the disposal of solid, liquid, and gaseous radioactive wastes is appropriately controlled during the PDNI. Also, an environmental monitoring program has been and continues to be in place since the initial construction phase of the Rancho Seco nuclear facility. These programs are designed to ensure that plant personnel and the general public are protected against excessise radiation exposure from radioactive wastes, in accordance with 10 CFR 20,10 CFR 50, and 40 CFR 190. Requirements for programs that administratively control the discharge of i radioactive solids. liquids, and gaseous wastes and address radiological monitoring of the I environment are contained in the Permanently Defueled Technical Specifications (PDTS). These PDTS program element requirements are implemented through the Off-site Dose Calculation N!anual (ODCNI), the Process Control Program (PCP), and the Radiological Environmental Nfonitoring Program (RENIP), and their implementing procedures.
l 11.5.2 OFF-SITE DOSE CALCULATION NiANUAL l The ODCN1 provides the information and methodologies used to evaluate the impact of radiological liquid and gaseous eftluent discharged from the plant. The ODCN1 is used to demonstrate that the plant complies with the requirements of 40 CFR 190 and 10 CFR 20, and ;
the dose guidelines of 10 CFR 50, Appendix I. Calculations for continuous airborne releases use l default atmospheric diffusion coefficients and gaseous effluent flow rates. Exposures due to plant operation are estimated by calculational methods specified in the ODCN1 using plant l effluent data obtained from plant effluent radiation monitors and/or manual sampling and '
analyses.
I1.5.2.l Liquid Discharge Pathway The primary liquid discharge source of potentially contaminated water is from the 'A'or B' regenerant hold-up tank (RHUT). RilUTs A and B can receive water from miscellaneous secondary sumps and equipment drains, the auxiliary boiler blow-down and area sump, and the demineralized reactor coolant storage tank.
Dose accountability for normal radioactive liquid releases is performed at the RHUTs. Waste water is transferred from the A or B RHUT to the north or south Retention Basin. Radioactive liquid releases into the environment are made from the Retention Basins.
Waste water collected in the RHUTs is sampled, analyzed, and monitored to ensure compliance with 10 CFR 50 Appendix 1, prior to transfer to a Retention Basin. An effluent strainer downstream of the RHUTs prevents excess resin and resin fines from entering process streams and carrying contamination off-site. There are no direct connections from the radioactive liquid radwaste treatment systems to the environment.
11.5-1
11.5.2.2 Gaseous Discharge Pathway The principal discharge sources of normal radioactive gaseous effluent during the PLM are the Reactor Building purge and Auxiliary Building exhaust pathways. During the PDM, releases from these pathways have been and are expected to continue to be very small; well within the 10 CFR 50, Appendix I dose guidelines and 10 CFR 20 concentration limits.
I1.5.3 PROCESS CONTROL PROGRAM The Process Control Program, which provides the administrative controls for the solid waste generation, packaging, shipping, and disposal program at Rancho Seco during the PDM, is addressed in detail in DS AR Section 11.4.
I1.5.4 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM (REMP) 11.5.4.1 Pre-operational REMP The State of California Radiological Surveillance Network was used to determine background radiation levels in the Sacramento Area. Data for 5 years, starting in January 1967 was compiled and evaluated. This program established the background radioactivity levels in the Sacramento area for man-made and naturally occurring radionuclides. Table 11.5-1 presents the means and standard deviations for specific sample types that were detected at statistically significant levels.
The Table 11.51 data presents the sample analysis results for gross beta on air filters and Sr-90 in milk. Although the frequency of the sampling may have varied, the California surveillance network sampled the same pathways. The results of the pre-operational Radiological Environmental Monitoring Program (REMP) program are presented in Table 11.5-2. These data summaries do not include sample analysis results that were not statistically different from zero.
The appearance of high radioactivil'y levels in the samples is attributed to radioactive fallout produced by nuclear weapons tests. Because of weather patterns in the state, increases at the Sacramento area were correlated with increases at Berkeley to the west, Redding to the north, and Fresno to the south. When nuclear weapons fallout is the source of radionuclides in the area,it is expected that increases in activity in the Rancho Seco area will be observed at other state sampling stations. Reactor emissions. on the other hand, would be characterized by unequal rises in preference to the prevailing wind direction with levels decreasing with distance. As shown previously, levels due to expected releases cannot, in general, be distinguished from variations in background. The expected and measured background levels were therefore essentially the same.
An analysis of human exposure to specific nuclides expected to be released during normal plant operations was carried out. Exposures due to these releases are calculated to be well below the numerical guidelines of Appendix I to 10 CFR 50. Therefore, these expected releases constitute only a small fraction of anticipated doses from natural (background) radiation. On this basis, the magnitude of the increase in environmental radiation levels due to airborne releases from the Rancho Seco Nuclear Generating Station is expected to be insignificant.
11.5-2
I i l l
l l
i TABLE I1.5-1 l l
~
SUMMARY
OF RADIATION BACKGROUND FOR SACRAMENTO, CALIFORNIA l l
1 Sacramento Air Filters Sr-90 Milk 3
(oCi/m ) (oCi/em Ca)
Standard Standard Quarter Mean Deviation Mean Deviation 1
1-67 2.2 11.3 2.7 0.5 2-67 0.15 0.15 3.7 0.4 3-67. 0.26 0.18 1.7 0.8 4-67 0.26 0.38 1.8 1.1 1-68 0.56 0.47 1.7 0.1 2-68 0.49 0.31 2.5 0.1 3-68 0.33 0.28 2.1 0.4 4-68 0.20 0.16 3.0 1.0 1-69 0.15 0.10 2.1 0.7 2 69- 0.41 0.33 2.7 0.3 i d-3-69 0.65 0.28 2.5 0.6 4-69 0.33 0.22 1.6 0.2 1-70 0.34 0.27 2.1 0.5 2-70 0.58 0.27 2.1 0.4 3 70 0.49 0.22 2.4 0.7
'4-70 0.31 0.38 2.0 0.3 1 71 0.32 0.29 3.5 2.5
'2 71- 0.63 0.29 3.3 0.5 3-71 0.44 0.25 2.0 0.2 4-71 0.24 0.83 1.5 0.1 1-72 0.23 0.26 1.6 0.2 2-72 0.25 0.32 --- ---
i.
11.5-3
11.5.4.1 Pre-operational REMP (Continued)
The samples collected in the pre-operational surveillance program are listed in Table 11.5 2.
Detection sensitivity depends on many factors, among which are detector efficiency and4 background radiation levels, the decay scheme of the radionuclide under consideration, the statistical nature of the radioactive decay process, and practical counting times (a maximum of 1,000 minutes). For beta radiation counters, the counter background level depends only on the success of shielding the detector from cosmic radiation or correcting by coincidence systems.
However, for gamma scans, counter background levels depend not only on the aforementioned factors, but also on the Compton scatter contribution arising from other radionuclides in the sample. In Table 11.5 2, gamma scan sensitivities are for the individual radionuclides only. The minimum detectable limits are derived from the analysis of experiments performed to assess system reliability. The sensitivities may change when additional nuclides are present in the sample.
. Quarterly reports were submitted to the Califomia Department of Public Health. In addition, results for 1 year (four quarterly reports) were submitted to the Atomic Energy Commission on February 28,1972.
11,5.4.1.1 Pre-operational Exposure Estimation
(
- The sources of radioactivity which are observed in the pre-operational surveillance program are natural radioactivity from cosmic radiation, naturally occurring radionuclides, and man-made radionuclides arising from nuclear weapons tests, etc. Each factor has a variability caused by
- natural phenomena, which is on the order of 50 percent. These uncertainties limit the certainty l with which the exposure can be determined. Design calculations performed on the expected releases indicated that emissions are so low that it is doubtful that the plant emissions will ever l
be detected by the environmental surveillance network.
l When beta activity is detected in a sample, an activity balance can be performed to account for the activity. Such a program has two goals: first, the accountability of the radioactivity from various sources; and second, quality control aspects. Gamma pulse height analysis was used instead of gross gamma counts because this analysis method should account for the activity found by gross beta scans. Because of the lower efficiencies encountered by gamma detectors, such accountability becomes practical only when beta activities are 100 counts per minute or more; therefore, such procedures were applied to samples with these activity levels.
The total background radiation doses were measured using thermoluminescent dosimeters with calcium sulfate-dysprosium phosphor discs. The dosimeter is preheated and annealed to reduce fading rate of the stored signal.
l l
Total background doses was measured monthly instead of quarterly to improve statistical sensitivity of the sample analysis data.
1 l
11.5-4
- - _ _ _ . - - . -------a _ - _ _ - - - - _ _ _ - - - - - _ _ - - . . - - - - - _ _ - - ,
nee- . . -. --.
_ _ . ~ _ - --_ _- ... ...
L Page 1 of 3 l TABLE 11.5-2 l RESULTS OF THE RANCHO SECO PRE-OPERATIONAL SURVEILLANCE PROGRAM Air Filters 3
(pCi/m ) .
Collection Location RAA RAB RAC RAD RAE RAF RAG ist Quarter 0.10 0.10 0.09 0.11 0.11 (F c) 0.06 0.06 0.06 0.06 0.07 1
2nd Quarter 0.18 0.15 0.19 0.14 0.19 (M c) 0.14 0.12 0.36 0.10 0.14 3rd Quarter 0.40 0.31 0.38 0.40 I
(P. c) 0.20 0.19 0.21 0.14 4th Quarter 0.27 0.28 0.25 0.25 0.33 0.26 0.21 (P. c) 0.14 0.08 0.14 0.11 0.16 0.13 0.06 5th Quarter 0.16 0.13 0.11 0.03 0.10 0.07 0.10 1
' (9. c) 0.07 0.05 0.03 0.05 0.05 0.02 0.05 l
11.5-5 l
[ -
- . - . . . - - - - . . . . . . . - - . . - . - . - . . . . . - . . . . ~ . - . . . - . . . - . - - .-.--
Page 2 of 3
, TABLE 11.5-2 RESULTS OF THE RANCHO SECO PRE OPERATIONAL SURVEILLANCE PROGRAM Water - Gross B (pCi/l) a- Standard Mean Deviatioa -
RRWA- Ground runoff Filtrate i1.6 3.0 Crud 9.2 9.9 1
RRWB- . Plant effluent Filtrate 4.3 3.7 Crud 7.7 5.9 RRWC- Site boundarv Filtrate 7.6 5.7 Crud 13.0 19.0 RWWA- On-site 5.9 0.4 RWWB - Clay well 2.4 0.5 RWWC- Drinking fountain 5.5 1.0 i
i 11.5-6 l
. ~.- -. . -. - . . = . _ . - . . - . -- .. ... - - . _._. - - --
Page 3 of 3 TABLE 11.5 2 RESULTS OFTHE RANCHO SECO PRE OPERATIONAL SURVEILLANCE PROGRAM Quarter 1 2 3 4 5 Average Milk - Sr-90 1.6 1.9 3.1 2.0 1.3 (pCi/gm Ca) 2.0 (p, c) 0.3 0.5 1.0 0.6 0.3 0.3 Rabbit '- Sr-90 ' 11.9 23.3 10.9 (pCi/gm) 16.4 ;
(p. c) 5.1 16 7.9 10.5 Sr-90 208 145 175 72 57 (pCi/kgm) 131 (p,0) 55 104 89 19 49 85 RSWA Comanche Res.
filtrate - B 3.1 1.9 - 3.1 1.7 2.2 (pCill) 23 1.1 0.6 10.2 0.2 0.1 0.7 l Solids 1.2 0.8 2.7 1.4 l.2 1.4 0.5 0.4 2.0 0.4 0.5 1.0 i
?
l RSWB 1 Folsom Res.
l filtrate - B 1.4 1.5 1.7 1.4 1.8 1.6 l +0.8 0.7 0.2 0.1 0.3 0.4 Solids 1.3 1.1 1.8 1.9 1.2 1.5
+0.4 +0. 5 +0.4 + 1.0 +0.3 +0.6 l
l RSWC 1 l
Rancho Seco Res.
l filtrate - B 5.1 5.5 8.9 10.7 11.7 8.4 0.1 0.6 1.0 0.9 1.4 2.8 Solids 0.1 3.7 5.7 3.7 2.6 4.2
} 32 1.1 0.2 0.1 0.9 1.7 i
11.5-7 I
r ,r-+, -w - , - , -- -
11.5.4.1.1 Pre-operational Exposure Estimation (Continued)
, Milk samples were taken and analyzed monthly. Because of fallout, detectable levels of Sr-90 were observed. These values have leveled off to a small and constant level. Sr-90 levels in milk samples are not expected to decline much below the historical values because of the long half-life of Sr-90. More frequent sampling of the milk chain is not necessary unless there is a deposition of fresh fission products in the Sacramento area that show up in milk at detectable levels.
11.5.4.2 Off-site Post-operational REMP Gaseous radioactive effluent from the plant consists primarily of tritium during the PDM, with the possibility of Krypton 85 being released in the normal plant effluent. Liquid radioactive effluent from_the plant consists primarily of very low levels of fission and activation products.
. Tn e following pathways of human exposure that may result from gaseous radioactive discharges .
during the PDM are considered:
A. Whole body extemal exposure B. Inhalation exposure C. Deposition on grass - cattle - milk - man D. Deposition on leafy vegetables - man E. Deposition on grass - cattle - beef- man F. Deposition on soil - plants - man The following pathways of human exposure that may result from liquid radioactive discharges during the PDM are considered:
A. Whole body external exposure via submersion B. Forage irrigation - cattle - milk - man C. Vegetables irrigation - man D. Forage irrigation - cattle - beef - man E. Water - fish - man F. Water - drinking water - man G. Water - invertebrates - man H. - Water - shoreline deposition - man 11.5-8
11.5.4.2 Off-site Post-operational REMP (Continued)
The purpose of the off-site post-operational REMP is to monitor radiation levels in the environment and to provide a basis for identification of changes in background levels.
Measurements are made during PDM operations to determine the radiation levels and the radioactive materials in the exposure pathways which lead to the highest potential radiation j exposures to the public. Efforts are made to correlate any significant changes in background I levels with events such as fallout from nuclear weapons testing, volcano eruptions, the Chernobyl i accident, natural phenomena, or changes in plant operation. The REMP can be used to verify estimates of exposure from radioactive effluent to a real member of the public as defined in 40 CFR 190.
l Plant personnel perform dose calculations in accordance with the ODCM, using plant effluent data collected from plant effluent monitors and/or manual sampling and analyses, to estimate radiation exposures that result from plant operations in the PDM. The estimated annual exposure t_o an individual (whole body dose) living near the plant effluent boundaries from normal plant I operations during the PDM is less than 1 mrem per year. This level of exposure is I indistinguishable from natural background exposures due to the variation in natural background radiation.
, The REMP supplements the radiological effluent control program (the ODCM) by verifying that ,
the measurable concentrations of radioactive materials and levels of radiation in the environs !
around the plant are not higher than expected on the basis of the effluent measurements and l modeling of the environmental exposure pathways. The post-operational REMP is described in and implemented by the REMP Manual and its implementing procedures.
1 11.5.4.2.1 Post-operational REMP Sampling Frequency ;
i The frequency of sampling given in the REMP is based on:
i A. The frequency and amount of effluent releases in the PDM, B. The effective biological half-life of the radionuclides involved in the biosphere, C. The potential pathways, and D. Regulatory requirements and guidance.
Because Rancho Seco has been shut down since June 1989, the radionuclides of interest for REMP evaluations are limited to those nuclides with half-lives greater than a few months.
The radionuclides that are of most concem are those that are longer-lised and may accumulate in the various food chains. Of major importance are Cs-134, Cs-137 and Co-60.
Details of the REMP sampling frequency are described in the REMP Manual and implementing procedures.
11.5-9
11.5.4.2.2 REMP Sample Types The sample types included in the post-operational REMP are based on operational and pre-operational REMP experience, the potential dose pathways to man, the 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.
I1.5.4.2.3 REMP Sample Statistical Analysis 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 uncenn.inty is determined using standard statistical methods. The minimum detectable limit varies for each sample and sample type and is calculated in accordance with the Lower Limit of Detection equation specified in the REMP Manual. The values determined are reported for each sample.
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 l well as changes in observed values.
11.5.4.3 Effluent and Waste Disposal Environmental Reports 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 following areas:
- 1. Environmental protection programs that monitor non-radiological and radiological i effects upon the environment. Also, a report submitted to the NRC that contains l 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.
l REMP reports are submitted to the California Depanment of Health Services on a regular basis.
I 1
Amendment 3 l
11,5-10 1
l
l l l l l 11.6 ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS IS REASONABLY ACHIEVABLE (ALARA) 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). This operating philosophy also applies to radiation
[ exposures to the general population resulting from the conduct of activities at Rancho Seco during the PDM.
The ALARA Policy applies to all SMUD and contract personnel who require access to the Radiologically Controlled Area of the plant; who are involved in the operation, maintenance, or modification of systems containing radioactive material during the PDM; or who are responsible for monitoring plant effitient.
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 external 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 intemal ,
uptakes will be maintained less than that allowed by NRC regulations. Engineering controls, i containment, and respiratory protection devices shall be used to the extent possible to maintain internal uptakes ALARA.
l Plant controls also work to minimize loose surface contamination, and thus, sources of internal l uptakes. Additionally, these plant controls shall minimize the number and size of contaminated l areas, the amount of accumulated radioactive waste, the potential for cross-contamination of tools and equipment, and minimize the need for protective clothing.
f Amendment 1 11.6-1 l
e
11.6.1 ALARA POLICY CONSIDERATIONS (Continued)
The Radiation Protection program controls radioactive materials ;o provide reasonable assurance that radioactive material is not lost or misplaced, accidental exposure or contamination of '
personnel will not occur, and accidental release of radioactive matvial into the environment wills not occur.
Radioactive liquids and gases released to the environment as a result of plant activities are maintained as low as is reasonably achievable in accordance with the Radiation Protection program and the applicable NRC and Environmental Protection Agency (EPA) regulations.
Plant personnel conduct comprehensive environmental monitoring and assessment programs to determine environmental radioactivity levels and the significance of these levels.
Procedures for work in radiological environments include applicable provisions and requirements that are commensurate with the severity of the radiological environment to ensure exposure is maintained ALARA.
The design process included appropriate measures to ensure that ALARA considerations are factored into all phases of a design.
> Personnel training is conducted to ensure that all personnel are adequately prepared to discharge their responsibilities as well as the requirements of the ALARA Policy, and that they have adequate training for working in a radiological environment.
The Quality Audit group performs periodic audits of the ALARA program to ensure the ALARA policy is properly implemented during the PDM.
The Plant hianager administers the ALARA Policy. The h!aintenance Superintendent, Technical Services Supervisor, Operations Superintendent, Operations Training Supervisor, and Quality Assurance / Licensing / Administration Supervisor ensure that the ALARA Policy is implemented.
The Radiation Protection / Chemistry Superintendent is responsible for overseeing the ALARA program and its implementation.
! The Plant Review Committee and Management Safety Review Committee (MSRC) review i ALARA reports, plant personnel exposure reports, and off-site radioactive effluent dose reports.
Also, the MSRC establishes annual ALARA goals.
I1.6.2 ALARA DESIGN CONSIDERATIONS Because the original design of the plant preceded the regulatory implementation of the ALARA concept, design considerations were adopted into various Rancho Seco documents to supplement the original facility design and to ensure that occupational doses to plant personnel are maintained ALARA. The ALARA Manual establishes the overall program to implement the Rancho Seco ALARA Policy. Design considerations are contained in Technical Services procedures in the form of design criteria and ALARA design reviews and cost / benefit evaluations.
Amendment 1 11.6-2 l .~. .. . _ _ _ _ _ . . _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _
_ _ _ _ _ _ _ - ~ _ _ _ . . _ . _ _ _ _._._______ _.___ _.. __ .
11.6.2 ALARA DESIGN CONSIDERATIONS (Continued)
Each plant modification or design change receives an ALARA evaluation, which takes into account the:
, 1. Radioactive hazards of the modification, I
- 2. Estimated personnel exposure for installation, L 3. Effects on the radiation area,
- 4. Reason for the modification,
- 5. - Shielding,
- 6. Maintenance and testing,
- 7. Control of contamination and radiation exposure,
- 8. Leakage
- 9. Crud traps, and
! 10. Cost / benefit of resultant reduction of personnel exposure. ,
i ALARA design features are described further in DSAR Section 11.8.1.
I1.63 ALAR A OPERATIONAL CONSIDERATIONS Delivery personnel and other visitors allowed within the Industrial Area (which is also the Restricted Area) are controlled in accordance with Security department administrative site access procedures. Visitors are provided with dosimetry,if necessary,in accordance with Radiation Protection procedures.
The following plant area definitions are provided to facilitate a better understanding of the ALARA and access control procedures:
- 1. Industrial Area - Operations area, protected by a double chain link fence and guarded entry points. This area is the same as the Restricted Area.
- 2. Radiation Area - An area, accessible to individuals,in which radiation levels could result in an individual receiving a dose equivalent in excess of 5 mrem in l
l one hour at 30 centimeters from the radiation source or from any surface that the j radiation penetrates.
l
! Amendment 3 l
11.6-3
11.6.3 ALARA OPERATIONAL CONSIDERATIONS (Continued)
- 3. High Radiation Area - An area, accessible to individuals,in which radiation levels could result in an individual receiving a dose equivalent in excess of 0.1 rem in one hour at 30 centimeters from the radiation source or from any surface that the radiation penetrates.
- 4. Secured-High Radiation Area - Any area accessible to personnel in which the level of radiation is such that a major ponion of an individual's body could receive a dose of 1000 mrem in any I hour.
- 5. Radiologically Controlled Area - Any area within the Industrial Area to which access is controlled for the purpose of protecting individuals from radiation or radioactive materials.
- 6. Hot Particle Zone - An area known or suspected to contain a discrete particle of contamination which has an activity greater than 5000 counts per minute at 0.5 inches with an RM-14 radiation monitor equipped with an HP-260 probe or equivalent.
- 7. Very High Radiation Area - An area, accessible to individuals,in which radiation levels could result in an individual receiving an absorbed dose in excess of 500 rads in one hour at one meter from a radiation source or from any surface that the radiation penetrates.
- 8. Restricted Area - An area, access to which is limited by the licensee for the purpose of protecting individuals against undue risks from exposure to radiation and radioactive materials. Restricted Area does not include areas used as residential quarters, but separate rooms in a residential building may be set apart as a Restricted Area.
l Amendment 1 i
11.6-4
- - - , - . _ . , _ . . . _ . , - y _ ,_
l l
5 11.7 RADIATION SOURCES
. I1.7.1 CONTAINED SOURCES The radiation sources that provide the basis for shield design are divided into five categories according to origin or location, as follows:
^ 1. The reactor vessel and internals,
- 2. Spent fuel pool cooling equipment,
- 3. Radioactive materials handling and processing equipment, and ;
l
' 4. Spent fuel elements and control components !
Radiation emanating from the reactor vessel consists primarily of gamma rays emitted from the residual itctivation of the reactor vessel and intemal core components.
- Shut down radiation levels in the Reactor Coolant System are a result of the buildup of activated corrosion products in the piping system, mainly in the crud traps.
11.7-1
11.8 RADIATION PROTECTION DESIGN FEATURES 11.8.1 . FACILITY. DESIGN FEATURES Per the general design philosophy of Rancho Seco, the layout for personnel access, routing of piping, and 'ocation of components are designed to minimize personnel radiation exposure.
Access control and traffic patterns are evaluated to assure that radiation exposure is maintained ALARA. In general, access to a given radiation zone is arranged so as not to require passing through a higher radiation zone. Also, consideration is given to lay-down areas for maintenance operations, provisions for rigging equipment, and platforms for easy access. Radioactive filter areas, ion exchangers, and spent resin collection system components are located as close to the solid radwaste processing areas as practicable. Radioactive wastes are stored in shielded enclosures that are separated from normally accessible areas, when appropriate.
Equipment requiring low background radiation levels is located in areas such as the counting room, away from highly radioactive components.
Equipment requiring frequent access, such as control panels, readout devices, and transmitters, is located in low radiation zones to minimize operator exposure.
To reduce radiation exposure due to sampling operations, sample stations are isolated as much as practicable from other radioactive equipment. Exposed sample piping is minimized. Sample stations consider the location of sample coolers, sinks, drain lines, hoods, ventilation and shielding.
i i
Shielding in radiation areas is provided to reduce the radiation from surrounding sources based upon the expected occupancy of each area. Valve galleries are provided for valves servicing equipment containing or processing highly radioactive material.
ll.8.2 SHIELDING Radiation shielding for Rancho Seco is provided principally by concrete walls, floors, and ceilings or roofs. The shielding generally functions as a structural member of buildings and is designed to meet both structural strength and shielding criteria.
I1.8.2.1 Design Criteria Rancho Seco operating personnel are protected by radiation shielding wherever a potential radiation hazard may exist. The shield;ng performs four primary functions:
- 1. Ensures that under normal PDM activities, the potential radiation dose to operating personnel and the general public is within the limits of 10 CFR 20 and
- 10 CFR 50, and that the potential radiation dose to maintenance and operating personnel, contractors, and visitors on-site is maintained ALARA.
I L
i 11.8-1 l
11.8.2.1 Desien Criteria (Continued)
- 2. Ensures that operating personnel and the general public are adequately protected l from radiation exposure following an accident during the PDM without undue hazard.
- 3. Protects certain components from excessive radiation exposure.
- 4. Facilitates access for maintenance of certain components.
11.8.2.2 Radiation Zone Classifications Radiation zones are classified in accordance with 10 CFR 20 and plant Radiation Protection procedures.
To prevent inadvertent entry of personnel into high radiation areas, each high radiation area in which the intensity of radiation is greater than 100 mrem /hr shall be barricaded and posted. Each ,
high radiation area in which radiation intensity is greater than 1,000 mrem /hr are controlled in l accordance with the requirements of the plant Technical Specifications. l l
11.8.2.3 Description of Shielding Radiation shielding is divided into the categories described below.
I1.8.2.3.1 Primary Shield l 1
l The primary shield is a large mass of reinforce I concrete surrounding the reactor vessel and l extending upward from the Reactor Building floor to form the walls of the fuel transfer canal. i The shield thickness is 5 feet up to the height of the reactor vessel flange, where it narrows to 4.5 l feet. The primary shield is designed to reduce, in conjunction with the secondary shield, the !
radiation level from sources within the reactor vessel and reactor coolant system.
I1.8.2.3.2 Secondary Shield The secondary shield is a reinforced concrete structure surrounding the reactor coolant equipment, including piping, pumps, and steam generators. This shield protects personnel from direct gamma radiation emanating from reactor coolant activation products and fission products.
The secondary shield is 4 feet thick.
l 11.8-2
11.8.2.3.3 Reactor Building Shield The Reactor Building shield is a reinforced, pre-stressed concrete containment structure that completely surrounds the nuclear steam supply system components. The Reactor Building wall and dome are 3.75 feet and 3.50 feet thick, respectively. Other significant radiation shielding inside the Reactor Building is listed in Table 11.8-1.
I1.8.2.3.4 Control Room Shield The thickness of the control room shielding is I foot.
I1.8.2.3.5 Auxiliary Shield Auxiliary shielding includes all concrete walls, covers, and removable blocks that shield the numerous radiation sources in the radioactive waste disposal systems and other contaminated systems and components remaining on-site, functional in the PDM or otherwise. Typical components that require shielding include waste holdup tanks, miscellaneous waste evaporator, demineralizers, filters, pumps, waste handling areas, and the Reactor Coolant System drain tank.
The major components and their shielding thickness are given in Table 11.8-1.
I1.8.2.3.6 Spent Fuel Shielding Shielding provides protection during all phases of spent fuel assembly movement and storage in the PDM. Spent fuel handling operations are performed under water to provide radiation protection. Maintaining a minimum water level of 37 feet in the spent fuel pool during spent fuel handling operations ensures a minimum of 9 feet of water is maintained over active fuel. This limits the radiation at the surface of the spent fuel pool to less than 10 mrem /hr. When fuel handling operations are not in progress, the spent fuel pool water level is normally maintained around 37 feet with a minimum allowed level of 23.25 feet. Again, this ensures that a minimum of 9 feet of water shielding is maintained over spent fuel during the PDM.
The spent fuel pool water also shields personnel from activated fuel assembly control components that are stored or may be handled during the PDM. Dose rates are generally less than 2.5 mrem /hr in working areas inside the Fuel Storage Building, however, certain manipulations of fuel assemblies and control components may produce short term exposures in excess of 2.5 mrem /hr. Radiation levels in the working areas are monitored during fuel handling operations to ensure that exposures to plant operating personnel do not exceed administrative or the applicable regulatory limits.
11.8.2.4 Shielding Materials The material used for the primary, secondary, Reactor Building, and auxiliary shields is ordinary concrete with a density of approximately 140 lb/ cubic foot.
11,8-3
. ._- . . . . -. . - - - -. - . . - .. _ . .-.-.- . - - ~ - _ - _ - . . . - - -
i
! TABLE 11.8-1 l PRINCIPAL SHIELDING l
l l
REACTOR BUILDING Comoonent Concrete Thickness (ft) l' Primary shield (below flange) 5 Primary shield (above flange) 4.5 l Secondary shield .
4 Reactor Building vertical walls 3.75 Reactor Building dome 3.50 Side walls of fuel transfer canal 5 End walls of fuel transfer canal 4.5 Floor of fuel transfer canal 4.5 AUXILIARY BUILDING Reactor coolant system drain tank 2.5 Miscellaneous waste tank 2 Spent regenerant tank 2 i Coolant waste holdup tank 2 l Miscellaneous waste evaporator 2 j Boric acid concentrator 1.5 -
Decay heat removal cooler 2.5 Makeup pump 2 High pressure injection pump 2 Concentrated boric acid storage tank 2 Coolant waste receiver tank 2 Flash tank 4.5 Miscellaneous waste condensate storage tank 1 ,
Spent resin storage tank 3.5 l Makeup tank Primary ion exchanger 3
)
3.75 Miscellaneous waste ion exchanger 3.75 Secondary ion exchanger 3.75 i Spent fuel coolant ion exchanger 3.75 l Spent fuel coolant filter i
~
Seal retum coolers 2 Waste drumming area 1 FUEL STOR AGE BUILDING
< i l
Side walls of spent fuel pool 5 End walls of spent fuel pool 5 Bottom of spent fuel pool 5 Spent fuel pool water gate wall 5.5 11.8-4 n- +--m , --
e --
11.8.3 VENTILATION
, The station ventilation system is described in DS AR Section 9.5.
I1.8.4 RADIATION NIONITORING SYSTEN!
11.8.4.1 Desien Criteria l
The Radiation Ntonitoring System measures, indicates, and/or records the radiation levels at l selected areas within the Industrial Area and in the normal radioactive liquid and gaseous effluent ;
i pathways. Annunciation occurs upon detection of abnormal radiation levels. On a high radiation signal, radiation monitor interlocks function to automatically terminate liquid and Reactor )
Building gaseous effluent releases to ensure the 10 CFR 20 concentration release limits for radioactive material to unrestricted areas are not exceeded. The system also provides particulate sampling of gaseous effluent for assessing compliance with the dose guidelines of 10 CFR 50, Appendix I.
The system is designed to operate during normal plant conditions as well as during and following accidents. Some radiation monitors meet the guidance contained in Regulatory Guide 1.97, revision 3. The radiological information that the radiation monitoring system provides is used to assess plant releases and plant conditions during the PDN1, personnel protection, and protection of the general public. This information is also used in support of the Emergency Plan.
In addition to the criteria stated above, the system also meets the requirements of 10 CFR 50, Appendix A, General Design Criterion 13,60,63, and 64.
I1.S.4.2 System Description The Radiation Nfonitoring System employs a digital and analog system manufactured by General Atomies and Victorcen, respectively. The digital monitors communicate with a centralized data acquisition and display computer. A control and display console for these monitors provides monitor status information, annunciation, and monitor controls, and is located in the control room. The analog monitors have control and display cabinets in the control room. Annunciation for these monitors is provided on a control room annunciator panel.
In general, the Radiation Nionitoring System is comprised of many subsystems called radiation monitors, each of which hase their own detector channels, sampling systems (if process type),
alarms, measurement electronics, check sources, readout devices, and other instrumentation and controls, Additionally, some monitors have recorders, computer outputs, and process control interlocks. Radiation monitors are categorized into two types: Area Radiation N1onitors and Process Radiation N!onitors. Tables 11.8-2 and 11.8 3 contains the following information for each area and process radiation monitor, respectively, that is required to function in the PDNt:
11,8-5
11.8.4.2 System Description (Continued) 1
- 1. The type of radiation detector,
- 2. Location and identification number of the radiation monitor, i
- 3. Range and location of the readout equipment,
- 4. Sensitivity of the detector channel (s),
l lJ 5. Type and location of alarms / indications, controls, and computer inputs.
The radiation monitor subsystems, together with their control and display equipment, j 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 wam 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 determining if site evacuation is necessary and through what site exit it should take place. Table 11.S 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 Offside Dose Calculation Manual (ODCM). Table 11.8-3 describes the various process radiation monitors.
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 l 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 8-6
, , . , , , - , ,- - , - - -, -m - ...,
Page 1 of 2 TABLE 11.8 2 AREA RADIATION MONITORS Area Detector / Readout Sensitivity Alarm / Control /
Radiation Location Equipment Computer input Monitor Personnel lonization Radiation 0.1 mrem /hr Visual and audible
-access Chamber / monitors level 0.1- high rad alarm at hatch area access hatch 10' mrem /hr detector:
R-15025 area inside in access indication also
- Reactor ' hatch and'* outside Reactor -
Building at A R M S Building and cabinet in the control room Spent fuel Ionization Radiation 0.1 mrem /hr Visual and audible j- Area chamber / monitors level,0.1- high radiation
!. R-15028 spent fuel pool 10' mrem /hr alarm at detector l area at detector and in the control and ARMS room j cabinet IOSB Low range GM Radiation 0.1 mrem /hr Visual and audible Extended- detector and level,0.1- high radiation range area high range ion 10' mrem /hr alarm at detector monitors chamber detector / at detector and lOSB and main
. R-15110 sump area, and ARMS control rooms R-15l11 cast cell area, cabinet R-15112 loading dock, R-15113 west cell area, R-15114 DAW storage area, R-15tl5 DAW handling area Control room lonization Radiation 0.1 mrem /hr Visual and audible
' Area - chamber / monitors le vel, 0.1- high radiation R-15030 control room 10' mrem /hr alarm at detector at detector and in the control and ARMS room cabinet Amendment 3 11.8-7 ve- r-- - v. o-,-- - ~ - . , . - - - - - - . -,- - - - - - - - - - _ . . - - - - .__------__-s ~ - - - -- -
I t
Page 2 of 2 TABLE I1.8 2 (Continued)
AREA RADIATION MONITORS Area Detector / Readout Sensitivity Alarm / Control /
s Radiation Location Equipment Computer input Monitor
' Radiochem Ionization Radiation 0.1 mrem /hr Visual and audible lab area chamber / monitors level,0.1- high radiation l R-15031 radiochemical 10' mrem /hr alarm at detector l laboratory . at detector and in the control and ARMS room cabinet 1
Drum decon' Ionization Radiation 0.1 mrem /hr Visual and audible I and loading chamber / monitors level,0.1- high radiation
. area drum decon and 10' mrem /hr alarm at detector ,
R-15033 loading area at detector and in the control l and ARMS room cabinet Radwaste Ionization Radiation 0.1 mrem /hr Visual and audible sump pump chamber / monitors level, 0.1- high radiation area radwaste sump 10' mrem /hr alarm at detector R 15034 pump area at detector and in the control and ARMS room cabinet Source Ionization Radiation 0.1 mrem /hr Visual and audible calibration chamber / monitors level, 0.1- high radiation room source cal- 10' mrem /hr alarm at detector
! R l5039 ibration room at detector and in the control and ARMS room cabinet l
Perimeter Ionization Radiation 0.1 mrem /hr Visual and audible monitors chamber / monitors level,0.1- high radiation l
R-15040 the north, south, 10' mrem /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.
t P
Amendment 2 11.8-8
< c,wo- - - , - . -- - - - - - - -, --ww -
Page 1 of 2 TABLE 11.8-3 PROCESS RADIATION MONITORS Process Sampler-Detector Readout Sensitivity Alarm / Control /
g Monitor Equipment Computer input Reactor Fixed removable CRT display Range: Alarms in control Building 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-II.
S' scintillation. the TSC.
Auxiliary Fixed removable CRT display Range: Alarms in control Building isokinetic sampler. in control 3.4E 7 thru room. Visual alarm Stack One low range room with 3.4E- 1 in TSC. Computer R-15045 channel of noble digital Ci/cc inputs to IDADS and gas detection. Low display and Kr-85 R M II.
range detectoc type recorder in 1
B- scintillation, the TSC.
Auxiliary Fixed removable CRT display Range: Alarms in control Building isokinetic sampler. in control 3.4E-7 thru room. Visual alarm
- Grade One low range room with 3.4E 1 in TSC. Computer I.evel channel of noble digital pCi/cc inputs to IDADS and Vent gas detection. Low display and Kr 85 RM-II.
R-15546A range detector type recorder in l
B. 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 lEl to IE6 channel failure.
! total detector. Samples counts per l
flow CCW system flow minute R-15008 continuously Amendment 1 i
11.8-9
I l l
l l
Page 2 of 2 TABLE 11.8-3 (Continued) l PROCESS RADIATION MONITORS l
l Process Sampler-Detector Readout- Sensitivity Alarm / Control /
Monitor Equipment Computer Input ,
I l
l l Retention Off line liquid CRY display M D C is Alarm on high rad ,
l Basin - sampler with in control 3E-8 pCi/cc signal or monitor - l l effluent scintillation room and local Cs-137 failure. Auto discharge detector. Samples digital secure liquid R-15017A liquid discharges readout at release on high rad from the plant, the monitor, signal or channel
( failure. Computer input to RM 11.
Retention Off line liquid CRT display MDC is Alarm on high rad Basin ' sampler with in control 3E-8 pCi/cc signal or monitor Inlet sciatillation room and 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. failure. Computer input to RM 11.
IOSB vent Off line CRT display MDC is Alarms in control gas particulate in control 4E 11 Ci/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.
i Amendment 3 11.8-10
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I1.8.4.2.2 Process / Effluent Radiation Monitors (Continued)
Liquid effluent radiation monitor R-15017A monitors and controls Retention Basin Discharges.
Another radiation monitor monitors the plant effluent wastewater 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 pre-selected Retention Basin occurs upon high radiation indication or monitor failure by the i
appropriate monitor, i
- The liquid effluent monitors have a fail safe feature that provides an alarm / trip signal upon any l of the following conditions:
A. Circuit failure, B. Down-scale failure, or C. Controla not set in the operate mode.
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f e 1 4
J 11,8-11
,w e .*m v - . - - - - . * - - . . - . . . - - - . - - - . -
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l 11.9 P_Q1E ASSESSN1ENT 11.9.1 PERSONNEL MONITORING l
The personnel monitoring program monitors intemal 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 l
entering a Radiologically Controlled Area wear a thermoluminescert dosimeter (TLD) badge and direct reading dosimetry. Personnel not normally working in or frequenting Radiologically ,
Controlled Areas (i.e., visitors) wear dosimetry in Radiologically Controlled Areas in accordance l with Radiation Protection procedures.
Individual exposures are maintained within the limits established in 10 CFR 20. Exposure l records are kept current through recording the direct reading dosimeter 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 l C. N!ultiple 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.
Radiation exposure histories and current occupational exposure records are maintained on all personnel who enter Controlled Areas.
I1.9.2 PERSONNEL EXPOSURE RECORD SYSTEN!
The Radiation Protection department uses a computer data system to process information and perform various record keeping functions.
l Amendment 3 11.9-1
1 i
< 11,9.2 PERSONNEL EXPOSURE RECORD SYSTEM (Continued) l l
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.
I1.9.3 MEDICAL EXAMINATION PROGRAM No special medical examination is considered necessary for radiation workers. SMUD i employees receive a pre-employment physical to determine health status and ability to perform theirjob. Radiation workers are given a pulmonary function test before they are allowed to wear respiratory equipment. Radiation workers are required to take the pulmonary function 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.
Amendment 1 11.9-2
~, - - . . - - - , - , - - - - . . - . , - , , . . , - , , ,,.e- -- -.-. - -- -
v- ~-- e
- TABLE OF CONTENTS Section Iillq Page
- 12. CONDUCT OF OPERATIONS 12.1-1 12.1 ORGANIZATIONAL STRUCTURE OF SMUD 12.1-1 12.1.1 NUCLEAR ORGANIZATION 12.1-1 12.1.2 PLANT PERSONNEL RESPONSIBILITIES AND AUTHORITIES 12.1-1 12.1.2.1 Plant Manager 12.1-1 l 12.1.2.2 . Operating Shift Crews 12.1-5 l 12.1.2.3 Succession of Responsibility 12.1-7 l 12.1.3 QUALIFICATIONS OF NUCLEAR PLANT PERSONNEL 12.1-7 l 12.2 PERSONNEL 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 of Initial Certification Program 12.2-2 12.2.2.2 Evaluation 12.2 3 12.2.3 STRUCTURE OF CERTIFIED FUEL HANDLER PROFICIENCY 12.2-4 PROGRAM 12.2.3.1 Evaluation 12.2 5 12.2.4 LICENSING TRAINING PROGRAM 12.2 5 12.2.5 MAINTENANCE TRAINING PROGRAM 12.2 5 l 12.2.5.1 Initial Training - 12.2-6 l 12.2.5.2 Continuing Training 12.2-6 l 12.2.6 SITE SUPPORT TRAINING PROGRAMS 12.2-6 l 12.2.6.1 Initial Training 12.2 7 Amendment 3 12-1 2 ..
4 TABLE OF CONTENTS (Continued)-
Section Title f.;Lu
'12.2.6.2 Continuine Trainine 12.2-7 l 12.3 EMERGENCY PLANNING 12.3-1 12.4 REVIEW AND AUDIT OFOPERATIONS 12.4-1 12.5 PLANT PROCEDURES AND PROCESS STANDARDS 12.5-1 12.5.i PROCEDURES 12.5-1 12.5.1.1 Conformance with Safety Guide 33 12.5-1
. .12.5.1.2 Preparation of Procedures . . --
12.5-1 -
12.5.1.2.1 Procedure Changes 12.5 2 12.5.1.3 Conduct of Operations 12.5-2
- 12.5.1.3.1 Shut down Control Room Operator Authority 12.5-3 12.5.1.3.2 Certified Fuel Handler Authority 12.5-3 12.5.1.3.3 Activities Affecting Plant Operations or Indications during the PDM 12.5-3 12.5.1.3.4 Manipulation of Facility Controls 12.5-3 l 12.5.1.3.5 Responsibility for Feel Handling Operations 12.5-3 12.5.1.3.6 Relief of Duties 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 s
12.5.2 OPERATING AND OTHER PROCEDURES 12.5-4 l 12.5.2.1 Operatine Procedures 12.5-4 l 12.5.2.1.1 System Procedures 12.5-5 12.5.2.1.2- Special Test Procedures 12.5 5 l 12.5.2.1.3 Annunciator Alarm Response Procedures 12.5-5 l Amendment 3 12-11
TABLE OF CONTENTS (Continued) 1 Section lills P.agg 12.5.2.1.4 Casualty Procedures 12.5-6 12.5.2.2 Other Procedures 12.5-6 12.5.2.2.1 Maintenance Procedures 12.5-6 12.5.2.2.2 Instrument and Control (I&C) Procedures 12.5-6 l 12.5.2.2.3 Surveillance Procedures 12.5-6 l 12.5.2.2.4 Chemistry Procedures 12.5-7 12.5.2.2.5 Radioactive Waste Management Procedures - 12.5-7 l 12.5.2.2.6 Radiation Protection Procedures 12.5-7 l 12.5.2.2.7 Security Plan 12.5-7 12.5.2.2.8 Emergency Plan and Implementing Procedures 12.5-7 l 12.5.2.2.9 Fire Protection Procedures 12.5-8 l 12.5.2.2.10 Quality Assurance 12.5-8 l 12.5.2.2.11 Certified Fuel Handler and Non-Certified Operator Training Programs 12.5-8 l 12.5.2.2.12 Decommissioning License Basis Documents 12.5-8 l 12.5.3 PROCESS STANDARDS 12.5 8 l 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 12.7-1 l
- 12.7.4 HEALTH PHYSICS RECORDS 12.7-2 12.7.5 OTHER RECORDS 12.7-2
12.8 REFERENCES
12.8 1 Amendment 3 12-111
LIST OF FIGURES Figure Iills 12.1-1 Nuclear Organization 12.5 1 Decommissioning License Basis Documents l
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Amendment 3 l
l 12-iv 1
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DSAR CHAPTER 12. CONDUCT OF OPERNIIONS 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 responsibilities of N!anagers, 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.
12.! 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. 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 authorities stated below, each department head is responsible for conducting a departmental training program which meets the applicable requirements and standards.
I 12.1.2.1 Plant Manager The Plant Manager is responsible to the SMUD General Manager, through the Director, Power Generation and AGM Energy Supply & Chief Engineer, for the safe and reliable operation of Amendment 2 12.1-1
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\ l 12.1.2.1 Plant Manager (Continued) 1 I
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. Cost Control.
- 2. Commitment Management,
- 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 The Plant Manager has the authority to establish nuclear organization policy and make commitments to the NRC and is supported by the following personnel:
A. Ouality Assurance /Licensine/ Administration Suncrintendent l The Quality Assurance / Licensing / Administration Superintendent ensures that the l Quality Assurance, Licensing, and Administrative programs are implemented in accordance with the applicable regulatory requirements. The Superintendent (1) l is independent of the pressures of plant operations during the PDM,(2) has sufficient authority and organizational freedom to identify problems that affect quality, recommend solutions, and verify implementation of solutions, and (3) has the authority to take any issues conceming the quality of operations at Rancho Seco to the Director, Power Generations.
The Quality Assurance / Licensing / Administration Superintendent is responsible l 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.
Amendment 2 12.1-2
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! 12.1.2.1 Mant Niananer (Continued)
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 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 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 PDM.
B. Oncrations Superintendent The Operations Superintendent ensures plant activities and operations during the PDM are conducted in accordance with the requirements of the facility license, the PDTS, plant operating procedures, and applicable local, state, and federal regulations. Details regarding operator shift crews are discussed in Section 12.1.2.2. The Superintendent manages the activities associated with sale storage of spent fuel during the PDM.
C. District Supervisor, Security Operations The District Supervisor, Security Operations reparts to the District Manager, ,
General Services and is responsible for the Rancho Seco Security group dunng l routine, emergency, and contingency PDM conditions. The District Supervisor, l
Security Operations develops and maintains the security training program and security related licensing basis documents and implementing procedures to ensure compliance with site, local, state, and federal security related requirements and i regulations. The Plant Manager approves nuclear security policy, program i changes, and security program implementing procedures.
D. Site Training Supervisor The Site Training Supervisor develops and implements 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 Amendment 3 12.1-3
12.1.2.1 Plant Manaser (Continued) conduct of the Certified Fuel Handler, Shut-down Control Room Operator, and Non-Certified Operator Training Programs. The Site Training Supervisor reports to the Plant Manager.
E. Radiation Protection / Chemistry Superintendent The Radiation Protection / Chemistry (RP/ Chem) Superintendent is responsible for:
- 1. Minimizing employee and public exposure to radiation and radioactive material,
- 2. Maintaining a personnel radiation exposure and monitoring record keeping program,
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' 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 program, d 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,
- 8. Managing the plant chemistry program, which establishes chemistry and radiochemistry controls and surveillances for plant systems through the Off-site Dose Calculation Manual and implementing procedures and Chemistry administrative procedures,
- 9. Controlling and monitoring radioactive liquid and gaseous releases, l0. Establishing and evaluating the content and effectiveness of chemistry technician training, and
- 11. Providing General Employee Training (GET) and Fire Brigade training.
- 12. Carrying out health physics functions. This provides sufficient organizational freedom to ensure health physics functions will be performed independent of operating pressures.
Amendment 2 12.1-4
m_ _. _ .._. __ __ . - _ _ _ . ...- _. .. __ , _. . . _ . ._ __ . . _ . _ . _ .
i 12.1.2.1 Plant Manacer (Continued)
F. Maintenance Sunerintendent 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 and components, and (2) directs modification activities. The Superintendent also provides nuclear organization suppon in the area of scheduling.
G. Technical Services Superintendent The Technical Services Superintendent is responsible for operations and l maintenance support, suppon for and technical direction in plant design,.
developing design modifications to plant systems, design specification development, design change control, configuration management, and discipline engineering. Also, the Superintendent is responsible for system engineering, performance monitoring, surveillance testing, the In-service Confirmation Program, reactor engineering, welding, fire protection, and instrumentation and I controls engineering. In addition, the Superintendent (1) is responsible for the l Asset Recovery Program for the systems, structures, and components not required to function during the PDM, and (2) maintains baseline design documents, the Master Equipment List (MEL), and other documents defining technical requirements for systems, structures, and components required to function during the PDM.
12.1.2.2 Operator Shift Crews The minimum shift crew composition consists of two Operations personnel, 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 one member of the shift crew must be in the control room when fuel is 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 Operations Department training procedures. This individual must be either a Shut-down Control Room Operator (either certified or non-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.
Fuel handling operations conducted during the PDM must be directly supervised by a Cenified Fuel Handler.
The duties of the Operations Superintendent are discussed in Section 12.1.2.1. The responsibilities and authorities of shift operations personnel are described below:
Amendment 2 12.1-5
._-,,. .,- --.% _ ,_-m e. _i_ .-._+.o.m.. - -_. . . , . .--._ - - , . . -m ,--- ei.-- .
12.1.2.2 Operator Shift Crews (Continued)
> A. Shift Supervisor The Shift 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 Shift 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.
! B. Shut-down Control Room Operator 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 i
room in accordance with Operations Department procedures. ,
C. Non-Certified Operator The non-certified operator monitors and operates plant equipment and systems in support of station operation during the PDM. 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 l
and qualified in accordance with Operation Department training procedures as a Shut-down Control Room Operator.
1 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:
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- 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
- 3. Entry into radiologically controlled areas under existing Radiation Work Permits (RWPs) to perform routine Operator activities such as surveillances.
RP Responders are not trained or permitted to perform tasks that require an ANSI qualified RP Technician. Plant operation evolutions that require RP Technician support, such as fuel handling operations, are scheduled and conducted when RP Technicians are available.
Amendment 1 12.1-6
12.1.2.2 Operator Shift Crews (Continued)
The RP Responder program requires RP Supervision to be on-call so that RP Responders may contact an individual qualified to perform RP Technician activities whenever (1) additional ,
instmetion or guidance is needed, or (2) RP Responders encounter radiological conditions beyond the scope of their training and qualifications. The RP Responder program requires a maximum two hour response time for RP Supervision or an RP Technician to report to the site, when necessary, to perform any required RP Technician functions.
12.1.2.3 Succession of Responsibility
- 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
- 2. Operations Superintendent
- 3. Shift Supervisor 12.1.3 QUALIFICATIONS OF NUCLEAR PLANT rERSONNEL Each member of the plant staff meets or exceeds the minimum qualifications of ANSI N 18.1-1971 for comparable positions, except for the RP/ Chem Superintendent, who meets or exceeds the recommendations and 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.
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 panicular emphasis on arrest record, previous employment and reason for leaving, and military service record.
Amendment i 12.1-7
12.2 PERSONNEL TR AINING 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 N18.1-1971, 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 District Supervisor, Security Operations.
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l 12.2.1 TR AINING PROGRAMS l
l The following descriptions outline the training program guidelines that govem 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. Certified Fuel Handler Training
- 2. Non-Certified Operator Training
- 3. Shutdown Control Room Operator Training
- 4. RP Responder Training
- 5. Chem / Rad Decommissioning Technician Training
- 6. Maintenance Training
- 7. Licensing Training I
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:
- 1. General Employee Training (GET)
- 2. First Aid Amendment 3 12.2-1
.12.2.1 TRAINING PROGRAMS (Continued)
' 3. Fire Brigade, Fire Protection
- 5. Security
.6. Quality Assurance
- 7. Radwaste Handler
- 8. Dosimetry Technician
- 9. ALARA 1 - 10. Safety
, i 1. PRC and MSRC
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- 12. Qualified Reviewer (10 CFR 50.59 Safety Evaluations) 4 Personnel working at Rancho Seco participate in the training programs required for theirjob j position.- Training is conducted and documented in accordance with departmental training i proceduresc 12.2.2 CERTIFIED FUEL HANDLER TRAINING PROGRAMS I 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.
12.2.2.1 Structure ofinitial Certification Procram The Initial Certification Training Program is divided into three basic phases. These phases are the classroom phase, the self study phase, and the job training phase.
a 12.2-2
12.2.2.1 Structure of Initial Cenification Program (Continued)
- 1. Classroom Phase The classroom phase of the training program consists of formal classroom presentations in fundamentals, related theoretical subjects, plant systems, and procedures.
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- 2. Self Study Phase ;
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 accomplish the study objectives and provides the documentation ~of successful completion of :
check-outs. When a trainee completes all the topics on a panicular checklist, a -
cenified training instructor or Cenified Fuel Handler administers an Operating examination / quiz. Each student is required to satisfactorily complete all l
checklists. The self study phase contains three modules: (1) Systems Training, (2) Procedure Training, and (3) Miscellaneous Materials Training.
- 3. Job Trainine Phase l The job training phase of the training program is designed to provide the I
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.
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12.2.2.2 Evaluation End of course evaluation for the Certified Fuel Handler initial training consists of two separate examinations.
- 1. Written examination
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- 2. Oral / walk-through Following successful completion of the training program, candidates are eligible for certification.
12.2-3
, ,, ----.-,,n--_ -. -,- -. - - - - - - - - . - - - , - - , _ - - - - - - - - - - - - _ _ - - - - - - - _ _ . -
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l 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. Classroom Phase 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.
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B. Self Study Phase l
The self study phase consists of reading assignments. Plant modifications, l procedure changes, operational assessments, and operational events may be contained in the reading assignments.
C. Job Trainine Phase The job training phase consists of system operation review, conducted by the Shift Supervisor, a certified instructor, or an Operations staff supervisor.
Biennial is defined as 24 calendar months in the requalification procedures.
12.2-4
12.2.3.1 Evaluation A. Written Examination 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. Oncratine Examination i
y, The Site Training Supervisor or his/her designee administers an operating examination to each Certified Fuel Handler annually. l 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 l 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 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 PDM.
The Maintenance Superintendent is responsible for implementing the Maintenance Training l Program. The Maintenance Training Program consists of Administrative Process Training.
Administrative Process Training is based on developing and maintaining the prerequisite skills and knowledge required by maintenance workers to accomplish specific maintenance tasks.
These tasks include, but are not limited to, initiating, implementing, adhering to and/or processing:
Amendment 2
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12.2-5 t v -4 - - - -<- - .- _ _ , ___ ____ - .
12.2.5-- MAINTENANCE TRAINING PROGRAM (Continued)
- 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 !
l they should have a working knowledge of the plant drawing system and vendor manual system.
12.2.5.1 Initial Trainine Initial training is designed to provide the trainee with knowledge and skills necessary to function as part of the Maintenance Department during the PDM. This training is typically completed in 2 years, i l
12.2.5.2 Cor.tinuine Trainine 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 I particular site support function. Training of personnelin 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 Rancho Seco during the PDM.
Site support training includes training programs for the following areas:
- 1. Chem / Rad Decomrnissioning Technician
- 2. RP Responder
- 3. General Employee Training
- 4. Fire Protection /First Aid Amendment 1 12.2-6
..- -- +s., - + J rha Wa-J,4- -4 1- -+A4 *O E
- E 4 -4e- 4 a .kW-- -4 3 A ,..'M-12.2.6.1 Initial Training The responsibility for each site support training program is assigned to the applicable department manager. Classroom and laboratory training are provided by the responsible department when appropriate or necessary. On the Job Training (OJT)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.
12.2.6.2 Continuine Trainine Continuing training programs are designed to meet the specific needs of the panicipating departments and may include plant change review, procedure change review, administrative training commitments, OJT training review, and material from the initial training program.
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12.2-7
-+e + ., , .- y
__4 _ ...-- , _- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ __ , , _ _ _ _ _ _ _ _ . . , _ . _ ____ _ . _ _ _ _ . _ _ _ _ _ _____. __._.
12.3 EMERGENCY PLAN i
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 l
emergency situations that may arise at Rancho Seco during the PDM. The focus of concern for i the Emergency Plan is the protection of plant personnel and the surrounding population. The l Emergency Planning Zone for Rancho Seco during the PDM is the Industnal Area. The calculated exposures at the Industrial Area 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, l with SMUD providing the appropriate information to these agencies during the course of an I emergency as directed by the SMUD Emergency Coordinator.
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 I needed during an emergency, and instructions to obtain names and telephone numbers for l
emergency call-out, if necessary.
l The Emergency Plan and EPIPs form a complete and detailed program which aids Rancho Seco I 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. l 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. l l
12.3-1
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 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 Plant Manager 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;
- 2. Engineering / Nuclear Engineering;
- 3. Chemistry and Radiochemistry;
- 4. Radiological Safety; and
- 5. Quality Assurance Practices, l
i and advises the Director, Power Generation in these areas. PRC and MSRC requirements and responsibilities are detailed in the Permanently Defueled Technical Specifications and Rancho Seco Administrative Procedures.
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l 1
l l
l Amendment 3 12.4-1
12.5 PLANT PROCEDURES AND PROCESS STANDARDS 12.5.1 PROCEDURES The performance of work and the conduct of activities at Rancho Seco are guided by procedures.
Activities important to safety are conducted in accordance with detailed written procedures.
12.5.1.1 Conformance with Safety Guide 33 The procedures discussed in DSAR Section 12.5 cover the activities referenced in the following documents:
A. Applicable procedures recommended in Appendix A of Safety Guide 33 November 1972 -
B. Permanently Defueled Technical Specifications (PDTS)
C. Surveillance and test activities of equipment important to safety D. Security Plan E. Emergency Plan 1
F. Fire Protection Plan G. Off-site Dose Calculation Manual (ODCM) l l
Fl. Radiological Environmental Monitoring Program (REMP) MANUAL
- 1. Process Control Program (PCP) Manual J. Rancho Seco Quality Manual (RSQM)
K. Radiation Protection Manual L. Certified Fuel Handler and Non Certified Operator Training Programs 12.5.1.2 Preparation of Procedures Procedures are divided into hierarchy levels. These include:
A. The Rancho Seco Quality Manual B. Rancho Seco Administrative Procedures 12.5-1
12.5.1.2 Pren;1 ration of Procedures (Continued)
C. Departmental Administrative Procedures D. Technical Procedures The Rancho Seco Administrative Procedures (RSAPs) define and implement administrative requirements or activities involving inter-department processes and administrative responsibilities.
.i Departmental Administrative Procedures define administrative requirements, activities, or actions specific to one department within the nuclear organization. In addition, Departmental Administrative Procedures may direct activities of other Departments when necessary for support.
Procedure hierarchy, preparation, review, approval, revision, and control is established in accordance with the applicable RS APs.
Those plant procedures that are considered important to safety or required by the PDTS are reviewed by the Plant Review Committee (PRC) and approved by Plant Management in accordance with Rancho Seco Administrative Procedures and the PDTS.
12.5.1.2.1 Procedure Changes Plant procedures are reviewed periodically as required by the PDTS and RS APs. Permanent procedure changes are reviewed and approved in accordance with RS APs. Temporary procedure changes may be made provided:
- 1. The intent of the original procedure is not altered.
- 2. The change is approved by two members of plant management staff, at least one of whom is a Certified Fuel Handler.
- 3. The change is documented, reviewed, and approved as required by PDTS.
In cases of emergency and in accordance with 10 CFR 50.54(x), plant personnel may depart from approved procedures that implement a license condition or PDTS requirement if the action is needed to protect public health and safety. Records of such departures will be logged, indicating the prevailing conditions / situation and the reason for the action (s) taken.
12.5-2
12.5.1.3 Conduct of Operations i
12.5.1.3.1 Shut-down Control Room Operator Authority The Shut-down Control Room Operator (SDCO) can be either a non-certified operator, trained and qualified in accordance with the Non-Certified Operator Training Progran to stand watch in the control room, or a Certified Fuel Handler (CFH). The SDCO, who is not a CFH, does nat have the authority to manipulate the controls of the ntclear storage facility during the PDM l
unless directed by a CFH. The SDCO monitors and operates control room instrumentation and I equipment, and responds to control room alarms and the direction of the Shift Supervisor or other CFHs. .
l 1
12.5.1.3.2 Certified Fuel Handler Authority l l
The Certified Fuel Handler (CFH) has all the authorities of the SDCO, plus the authority to direct ;
the activities of non-certified operators and other CFHs. Also, a CFH has the authority to manipulate the controls of the nuclear storage facility and to conduct and/or directly supervise fuel handling operations. While a CFil is the Shift Supervisor, the CFH is ultimately responsible for the conduct of the day to day activities at the plant during the PDM that are within the CFH's l
control. A CFH is deemed to manipulate a control if the CFH directs an operator to manipulate a control.
12.5.1.3.3 Activities Affecting Plant Operations or Indications During the PDM 1
Plant personnel performing activities during the PDM that may affect plant operations or control l
room indications are required to notify the control room (either the Shut-down Control Room l
Operator or the Shift Supervisor) prior to initiating such actions. Removal from service of an '
instrument or component that is required to function during the PDM requires the permission of an on-shift CFH.
l 12.5.1.3.4 Manipulation of Facility Controls ;
Only Certified Fuel Handlers and non-certified operators and trainees, who are enrolled in l approved training programs and operating under the direction of a Certified Fuel Handler, are permitted to manipulate nuclear storage facility controls. Controls are defined in Operations Department piocedures.
12.5.1.3.5 Responsibility for Fuel Handling Operations The responsibility for directly supervising fuel handling operations is assigned to Certified Fuel Handlers.
12.5-3
_- _ _. _. . _ _ _ _ . __ .. _ _ ~ . __. _ . _ _ _ -
12.5.1.3.6 Relief of Duties Relief of control room shift duties is accomplished through a checklist of appropriate items for shift turnover,
, 12.5.1,3.7 Equipment Control Equipment control is maintained and documented through the use of tags, labels, status logs, or other suitable means.
12.5.1.3.8 Surveillance Testing Schedule Approved procedures establish a surveillance testing schedule to assure that required testing is performed and evaluated in accordance with the requirements of the PDTS and other licensing .
basis documents. Surveillance testing is scheduled and performed such that the safety of the station is not dependent on the performance of a structure, system, or component which has not been tested within its specified testing interval.
i j 12.5.1.3.9 Log Books
- Log books are maintained and periodically reviewed. Log books document significant events that occurred during a shift. Logs maintained in the control room include information such as significant o' arms received, abnormal operating conditions, and releases of radioactive effluent.
l l
12.5.2 OPERATING AND OTHER PROCEDURES 12.5.2.1 Operating Procedures Operating procedures include start up, operation, and shut-down of systems. These procedures may include but are not limited to procedures covering the following types of operations:
- 1. Electrical
- 2. Instrumentation and control
- 3. Normal and back up spent fuel pool cooling
- 4. Spent fuel pool make up water methods
- 5. Radwaste and radioactive effluent processing systems
- 6. Site water and air supply 12.5-4 l
L __ _____ _ _ ._, - --- - - _ . - . - -
l 12.5.2.1 Operatine Procedures (Continued) l 1
- 7. Auxiliary steam boiler system ;
l
- 8. Heating, ventilation, and air conditioning l 12.5.2.1.1 System Procedures l Operating activities which affect the proper functioning of the plant systems and components considered important to safety during the PDM are perfonned in accordance with approved, written procedures.
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System operating procedures provide the necessary instructions for the integrated operation of plant systems. The procedures include Check-off lists to ensure the necessary operating instructions, tests, and calibrations are or were completed. Checklists are also used to ensure major procedure steps are performed in the proper sequence.
These procedures provide a pre-planned method of conducting system operations during the l PDM. Operating procedures are sufficiently detailed that qualified Operations Department l individuals can perform the required functions without direct supervision. Written procedures cannot address all contingencies; therefore, operating procedures contain a degree of flexibility appropriate for the activity to which they apply. l l
l 12.5.2.1.2 Special Test Procedures I Special test procedures are written, plant management approved procedures issued for operating activities which are not of a recurring nature. Exarnples of special test procedures are those that address special testing or provide guidance for special operation of a system. These procedures are sufficiently detailed that qualified individuals can perform the required functions without direct supervision. Written procedures cannot address all contingencies; therefore, special test procedures contain a degree of nexibility appropriate for the activity to which they apply.
12.5.2.1.3 Annunciator Alarm Response Procedures !
Annunciator alarm response procedures specify operator actions necessary to respond to an l off-normal condition as indicated by an alarm. The annunciator response procedures include an appropriate combination of the following features: (1) probable cause, (2) automatic actions, (3) immediate operator actions, (4) supplementary actions, and (5) applicable references.
Each annunciator panel and annunciator window within a panel is designated by unique identification symbols. The annunciator response procedures are grouped by panel, then subdivided by annunciator so the operator can quickly find and refer to the appropriate response procedure for any annunciator.
12.5-5
l l
l 12.5.2.1.4 Casualty Procedures Casualty procedures have been prepared that specify operator actions for restoring an operating variable to its normal controlled value when it departs from its range, or to restore normal operating conditions following a perturbation.
12.5.2.2 Other Procedures 12.5.2.2.1 Maintenance Procedures Maintenance of plant structures, systems, and components important to safety is performed in accordance with written procedures or drawmgs appropriate to the circumstances (e. g., skills normally possessed by qualified maintenance personnel do not require detailed step-by step delineation in a written procedure). When appropriate, the procedures refer to vendor manuals, -
instructions, and approved drawings. When vendor manuals, instructions, and approved drawings do not provide adequate guidance to assure the required quality of work, an approved, written procedure is provided. These procedures are sufficiently detailed so that qualified maintenance workers can perform the required functions without direct supervision. Written procedures cannot address all contingencies; therefore, procedures contain a degree of flexibility appropriate to the activity to which they apply.
Maintenance procedures describe the methods of equipment clearance, surveillance, preventive maintenance, repair, and calibration.
12.5.2.2.2 Instrument and Control (l&C) Procedures Maintenance, testing, and calibration of plant instruments important to safety is performed in accordance with written procedures or drawings appropriate to the circumstances (i.e., skills normally possessed by qualified I&C personnel do not require detailed step-by-step delineation in a written procedure). When appropriate, these procedures refer to vendor manuals, instructions, and approved drawings. When vendor manuals, instructions, and approved drawings do not provide adequate guidance to assure the required quality of work, an approved, written procedure is provided. I&C procedures are sufficiently detailed so that qualified I&C personnel can perform the required functions without direct supervision. Written procedures cannot address all contingencies; therefore, procedures contain a degree of flexibility appropriate to the activity to which they apply.
12.5.2.2.3 Surveillance Procedures Periodic testing is conducted in accordance with approved, written procedures to determine various plant parameters and to verify the continuing capability of systems, components, and instruments important to safety to meet performance requirements . These procedures are sufficiently detailed so that qualified personnel can perform the required functions without direct supervision.
12.5-6
12.5.2.2.4 Chemistry Procedures Chemical and radio-chemical sampling, analyses, and other activities associated with chemical control of plant systems and components are conducted in accordance with approved, written procedures to meet performance requirements. These procedures are sufficient y detailed that l qualified personnel can perform the required functions without direct supervision. Written '
procedures cannot address all contingencies; therefore, procedures contain a degree of flexibility appropriate to the activity to which they apply.
12.5.2.2.5 Radioactive Waste Management Procedures l
Radioactive waste management activities associated with the plant's liquid, gaseous, and solid waste systems are performed in accordance with approved, written procedures. These procedures are sufficiently detailed that qualified personnel can perform the required functions without direct supervision. Written procedures cannot address all contingencies: therefore, procedures contain a degree of flexibility appropriate to the activity to which they apply.
Written Radiation Protection, Environmental Monitoring, and Chemistry Department procedures cover PCP, REMP, ODCM requirernents as well as other radioactive waste related regulatory requirements and commitments applicable to Rancho Seco in the PDM.
I 12.5.2.2.6 Radiation Protection Procedures Radiation protection procedures address radiation exposure policies, limits, and the ALARA l program. These procedures specify the limits for exposure to radiation and radioactive materials l and ensure compliance with state and federal regulations. Information concerning these i procedures is contained in DSAR Chapter i1. l l
12.5.2.2.7 Security Plan The Security Plan describes protection of the plant and the spent fuel during the PDM. The Security Plan is described in DS AR Section 12.6 and is submitted separately to the NRC.
12.5.2.2.8 Emergency Plan and implementing Procedures The Emergency Plan provides a description of the organization, equipment, and preparations available to rapidly and effectively respond to an emergency situation that may arise at Rancho l Seco during the PDM. The primary purpose of the Emergency Plan and the associated l implementing procedures is the protection of public health and safety and the protection of plant personnel from excessive exposure to radiation following an accident. These procedures are discussed in DSAR Section 12.3.
, 12.5-7 l
l
12.5.2.2.9 Fire Protection Procedures The Fire Protection Plan and implementing procedures address topics such as fire brigade training, reporting of fires, posting of fire watches, and fire alarm, detection, and suppression systems. l 1
l 12.5.2.2.10 Quality Assurance The Quality Manual has been separately approved by the NRC. The manual describes the j 10 CFR 50, Appendix B required Quality Assurance Program applicable to Rancho Seco in the PDM. Changes to the Quality Manual are made and submitted to the NRC in accordance with 10 CFR 50.59 and 10 CFR 50.54(a).
i l
l 12.5.2.2.11 Cenified Fuel Handler and Non-Certified Operator Training Programs l
l Certified Fuel Handlers are trained and qualified in accordance with the District administered and I maintained Certified Fuel Handler Training Program. This program was initially approved by the NRC. Subsequent changes to this program are governed by 10 CFR 50.59 and do not require i NRC approval prior to implementation as long as a PDTS change is not required and an I
Unreviewed Safety Question is not involved.
Non-certified operators, including the Shut-down Control Room Operator, are trained and .
l qualified in accordance with the District approved, administered, and maintained Non-Certified I Operator Training Program. Changes to this program are conducted in accordance with 10 CFR I l
50.59.
12.5.2.2.12 Decommissioning License Basis Documents !
I The hierarchy and relationship between the various documents that define the licensing basis for I I
operation of Rancho Seco during the decommissioning phase are presented in Figure 12.51.
12.5.3 PROCESS STANDARDS l The Process Standards applicable in the PDM are the operating requirements established to promote efficient and safe operation of the plant in the defueled condition. These standards clearly define the limits of the measurable parameters which, when adhered to, assure that plant operations in the PDM are carried out within the restrictions to which the District is legally bound by the facility license, state, and federal regulations. The Process Standards are composed of two parts: the Process Standards File (a computerized database, which is a subset of NUCLEIS), and appropriate Operating Procedures.
1 Amendment 3 12.5-8 I 1
l
1 12.5.3 PROCESS STANDARDS (Continued)
The computerized database is related to the Master Equipment L;st (MEL) and contains plant setpoints. The Process Standards File is a subset of the MEL program. This relationship allows setpoints to be found by using the search and sort functions of the MEL program, because the setpoints are filed by the associated Equipment Identification Number (EIN).
The Operating Procedures contain limits and precautions for the operation of plant systems required to function in the PDM, They also contain graphs and operating curves for the applicable systems and components.
The Process Standards File contains a listing of the system and component setpoints applicable in the PDM.
Changes to either part of the Process Standards must be made in accordance with established procedures.
1 1
12.5-9
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l 12.6 INDUSTRIAL SECURITY The Security Plan describes the appropriate protection measures taken at Rancho Seco during the PDM to ensure continued safe storage of spent fuel and other highly radioactive materials on-site. The Rancho Seco Physical Security Plan in effect during the PDM provides guidance to all personnel responsible for or directly concerned with the security of Rancho Seco. The Plan prescribes responsibilities, policies, procedures, and standards penaining to the overall security program for Rancho Seco. It provides information on all aspects of physical security for static, mobile, normal, and ernergency conditions as they apply to Rancho Seco during the PDM.
The Security Plan details the physical measures designed to safeguard personnel and the spent fuel; to prevent unauthorized access to equipment, facilities, material, property, or special nuclear material; and to safeguard them against sabotage, damage, and theft.
The Security Plan implemented during the PDM reflects specific exemptions obtained from and implementation of the applicable requirements of 10 CFR 73. Submission of the Plan and amendments are made to the NRC as required by 10 CFR 50.54(p),10 CFR 50.59, and 10 CFR 50.90.
i 12.6-1
______ _ _ _.y l
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4 12.7 RECORDS 12.7.1 OPERATING RECORDS L
The following records are produced and kept on file to document plant operations during the i
PDM:
A. Shiftly Log l
A log of plant parameters is prepared each shift by on-shift operator (s). This log includes a record of spent fuel pool level, temperature, and cooling parameters :
, and other selected plant conditions considered significant during the PDM. )
B. Control Room Loe i
This log contains a summary of plant operations for each operator shift. The log is normally prepared by shift operators and contains chronological notations of pertinent events. Also, notation of any significant conditions, both normal and
)
abnormal, is maintained in this log. This includes notation of gaseous and liquid 1
- radioactive effluent released from the plant. l
.12.7.2 ADMINISTRATIVE RECORDS The following administrative records are retained as required:
I. Special nuclear material inventory records (to satisfy the requirements of 10 CFR j 70).
- 2. By-product material inventory records (to satisfy the requirements of 10 CFR 30).
- 3. Incident report records (security).
- 4. Records required by the Permanently Defueled Technical Specifications.
- 5. Plant design and construction records.
t
- 6. Records of Permanently Defueled Technical Specification required surveillance tests.
) 12.7.3 MAINTENANCE RECORDS l
l The following maintenance records are retained as required:
1 1. Major equipment history, including work requests and preventive maintenance.
I 12.7-1
12.7 3 MAINTENANCE RECORDS (Continued)
- 2. Calibration, preventive maintenance, and work requests of station instrumentation, radiation survey meters, and environmental radiation monitors.
12.7.4 HEALTH PHYSICS RECORDS The following health physics records are maintained as required:
A. Personnel Exposure
- 1. Dosimeter readings (daily).
- 2. Film badge - TLD record (quarterly or more frequently as required).
- 3. Radio bioassay records (as required). . 4
- 4. Radiation exposure history and current exposure status (as required by 10 c CFR 20).
! B. Survey. Monitoring. and Waste Discosal l
l l 1. Routine station surveys and monitoring reports.
- 2. Environmental monitoring records.
- 3. Solid and liquid radioactive material shipment records (each shipment).
- 4. Liquid and gaseous waste discharge (each permit).
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l 12.7.5 OTHER RECORDS l
l The following additional records are maintained as required:
- 1. Plant chemistry records for systems and components required in the PDM.
- 2. Plant recorder charts.
- 3. Special tests or operating data sheets.
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l 12.7-2 n..
12.8 REFERENCES
1, License Amendment No. I19, 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. I17, dated March 17,1992, Possession-Only License l
l 12.8-1
NUCLEAR ORGANIZATION BOARD OF DIRECTORS i
GENERAL MANAGER AGM. ENERGY SUPPLY
& CHIEF ENGINEER DIRECTOR, I____________________ POWER GENERATION i i MANAGEMEN! SAFETY 1 l
REVIEW COMMITTEE j l
l MANAGER, l PLANT CLOSURE & l sa DECOMMISSIONING (n lit)- l j Tssues ! SITE TRAINING PLANT REVIEW I SUPERVISOR COMMITTEE I
I l l l l I NUCLEAR QUALITY / NUCLEAR TECHNICAL NUCLEAR NUCLEAR I-- LICENSING / ADMIN MAINTENANCE SERVICES RP & CHEMISTRY OPERATIONS SUPERINTENDENT SUPERINTENDENT SUPERINTENDENT SUPERINTENDENT SUPERINTENDENT l FIGURE 12.1-1 l
($suun
~
"e""e"'
SACRAMENTO MUNICIPAL UTILITY DISTRICT 3
DECOMMISSIONING LICENSE i- BASIS DOCUMENTS HIERARCHY l
l 10 CFR ISFSI SER - - - - - - - - - LICENSE - - - - - - - - - -
POL SER 10 CFR 50 (POL) 10 CFR 72 &
-- TECH SPECU - - - - - - - - -
TECH SPEC SERs i
DECOMM. ENVIRONMENTAL _ _ _ _ DECOMMISSIONING ___.
DECOMM. ORDER SER ASSESSMENT ORDER DSAR ISFSI SAR
+ Approved 10 CFR + Approved 10 CFR 50.59 Evoluotions 72.48 Evoluotions SECURITY PLAN ODCM REMP CFH TRNG QA FIRE PROT EMERG PCP PSDAR PROGRAM MANUAL PLAN PLAN i
FIGURE 12.5-1
- Amendment 3 .
SACRAMENTO MUNICIPAL UTILITY DISTRICT
TABLE OF CONTENTS f
Section Title Page a
- 14. SAFETY ANALYSIS 14.1-1 14.1 ACCIDENTS CONSIDERED CREDIBLE IN THE 14.1-1 PERMANENTLY DEFUELED MODE (PDM) 14.1.1' FUEL HANDLINO ACCIDENT 14.1-1 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 DECOMMISSIONINO ACCIDENT ANALYSIS 14.2-1 14.2.1 PREPARATION FOR SAFSTOR 14.2-1 14.2.2 ACCIDENTS DURING DECON 14.2-2 14.3 '
REFERENCES 14.3-1 Amendment 3 s 14-1
i LIST OF TABLES Table Title Page 14.1 1 Fuel Handling Accident Analysis Assumptions 14.1-3 14-ii
f TABLE 14.1-1 l
FUELIIANDLING ACCIDENT ANALYSIS ASSUMPTIONS
- 1. Period of continuous operation prior to final 1017 days shut-down
- 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 3
- 7. Atmospheric dispersion coefficient X/Q 3.61E-3 s/m
l Amendment 2 1
14.1-3
14.2 DECOMMISSIONING ACCIDENT ANALYSIS The Rancho Seco Decommissioning Accident Analysis (RSDAA)is part of the Rancho Seco licensing design basis for operation'during the PDM. The District updated the original RSDAA when the District converted the Rancho Seco Decommissioning Plan to the Post Shutdown Decommissioning Activities Report (PSDAR, Revision 1). In the 10 CFR 50.59 evaluation for PSDAR, Revision 1, the District committed to including an updated version of the RSDAA in the DSAR. The following is an updated version of the RSDAA.
While decommissioning radioactively contaminated structures, systems, and components at Rancho Seco, it is necessary to assure the safety of the public in the surrounding area and workers. Worker safety is addressed in the Rancho Seco Radiation Protection Program, which relies on ALARA principles, and the Rancho Seco Safety Program, which is defined in the Rancho Seco Safety Manual. The safety of the public is principally related to potential hazards associated with an airborne release of radioactive materials from Rancho Seco during decommissioning operations.
14.2.1 PREPARATION FOR SAFSTOR Immobilizing radioactivity and reducing radioactively contaminated areas is a major operation performed during Preparation for SAFSTOR. Preparation for S AFSTOR includes:
- 1. Facility decontamination, including decontamination or disposal of excess tools,
- 2. Component and system deactivation and residual contamination immobilization,
- 3. Dismantlement of non radioactive and low level radioactive components and systems (Incremental Dismantlement),
- 4. Maintenance, -
- 5. Waste management, and
- 6. Surveillance activities.
NUREG/CR-0130,' Technology, Safety, and Cost of Decommissioning a Reference PWR Power Station," identified and evaluated public and occupational safety impacts from decommissioning operations at the reference PWR nuclear power plant. The safety evaluation considers impacts for public radiation exposure, occupational radiation exposure, industrial accidents, and chemical pollutants. The NUREG takes a conservative approach using parameters that tend to maximize the projected consequences and safety impacts of each decommissioning operation.
Amendment 3 14.2-1 u_
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14.2.1 PREPARATION FOR SAFSTOR (Continued) l Potential accidents associated with Preparation for S AFSTOR activities at Rancho Seco are bounded by the accidents postulated in NUREG/CR-0130. As discussed in PSDAR Section 4.0,
, " Environmental Review," Rancho Seco's decommissioning attributes are within the attributes associated with the reference plant. The actual radionuclide inventory at Rancho Seco is less than the inventory assumed in NUREG/CR-0130. This fact supports the conclusion that Preparation for S AFSTOR accidents would result in minimal doses to workers and the public.
f L 14.2.2 ACCIDENTS DURING DECON l
i DECON will begin after an extended SAFSTOR period. During DECON, the District will perform decontamination and dismantlement of the remaining structures, systems, and components in addition to maintenance, waste management, and surveillance. The accidents discussed in NUREG/CR 0130 associated with immediate dismantlement y ould be applicable !
during DECON at Rancho Seco. However, the potential consequences asscciated with these accidents would be less because of a reduction in the Rancho Seco radionuclide inventory due to:
- 1. Decontamination efforts made before DECON,
- 2. Prior radioactive waste shipments, and
- 3. Radioactive decay.
Therefore, the potential DECON accidents at Rancho Seco are bounded by the accident evaluation specified in NUREG/CR-0130.
Operational accidents during DECON could result from equipment failure, human error, and service conditions. With spent fuel removed from the plant before the DECON decommissioning phase, operational accidents during DECON may be categorized as follows:
- 1. Radioactive waste transportation accidents.
- 2. Explosions and/or fires associated with explosive and/or combustible materials,
- 3. Loss of contamination control.
- 4. Natural phenomena, and
- 5. Human caused events external to Rancho Seco.
Amendment 3 14.2-2 l 1
'14.2.2 ACCIDENTS DURING DECON (Continued)
These potential operational accidents during DECON are addressed in NUREG/CR-0130 for immediate dismantlement. Therefore, for DECON operations at Rancho Seco, the associated, potential accidents are bounded by the NUREG/CR-0130 evaluation.
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i Amendment 3
- 14.2-3
I
14.3 REFERENCES
- 1. License Amendment No. I19, dated March 19,1992, Permanently Defueled Technical l Specifications l
- 2. Safety Analysis and No Significant Hazards Consideration (Log No.1091, Revision 3) for Proposed Amendment 182, Revision 3, Permanently Defueled Technical l Specifications I
- 3. License Amendment No. I17, dated March 17,1992, Possession-Only License
- 4. J. F. Stolz (NRC) to J. J. Mattimoe (SMUD) letter dated January 20,1984, Rancho Seco I License Amendment No. 52 l
- 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 l
L 7. 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"
- 8. NUREG/CR-0130," Technology, Safety, and Cost of Decommissioning a Reference PWR Power Station"
- 9. NUREG-0586," Final Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities" l
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