ML20206K958
| ML20206K958 | |
| Person / Time | |
|---|---|
| Site: | 07200011 |
| Issue date: | 06/29/2020 |
| From: | Gacke B Sacramento Municipal Utility District (SMUD) |
| To: | Andrea Kock Document Control Desk, Office of Nuclear Material Safety and Safeguards |
| References | |
| DPG 20-082 | |
| Download: ML20206K958 (60) | |
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- June 29, 2020 DPG 20-082 ATTN: Document Control Desk Director, Division of Fuel M<!llagement Office of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Rancho Seco Independent Spent Fuel Storage Installation Renewed Materials License No. SNM-2510 Docket 72-11 RANCHO SECO ISFSI FINAL SAFETY ANALYSIS REPORT, REVISION 9 Attention: Andrea Kock Attached is Revision 9 to _the Rancho Seco Independent Spent Fuel Storage Installation (ISFSI) Final Safety Analysis Report (FSAR). The updated ISFSI FSAR _reflects items that were identified while performing safety reviews in accordance with 10 CFR 72.48 reflecting termination of the Rancho Seco 10CFR50 license. Safety evaluations performed while voiding the Offsite Dose Calculation Manual (ODCM), Post-Shutdown Decommissioning Activities Report (PSDAR), Fire Protection Plan, and the Radioactive Material Storage and Decommissioning Safety Analysis Report (RADSAR) each identified required updates to the ISFSI FSAR.
The enclosed attachments include removal/insertion instructions for the changed pages, a List of Effective Pages, and the affected FSARpages. Vertical lines in the left-hand margin of the affected FSAR pages indicate the area of changed text. "Revision 9" and "June 2020" is also typed at the bottom right of each changed page.
By signature below, the Sacramento Municipal Utility District certifies that the above is true and correct. If you, or members of your staff, have questions requiring additional information or clarification you may contact me at (916) 732-4812.
Sincerely, ;4-os-3
~ tJ/V/5SZD Brad Gacke
. Manager, Rancho Seco Assets
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- NRC, Region IV William Allen: Program Manager Part 72 RIC lF.099 Rancho Seco Nuclear Generating Station I 14440 Twin Cities Road I Herald, CA 95638-9799 I 916.452.3211 I smud.org
ISFSI FSAR REVISION 9 REMOVAL/INSERTION INSTRUCTIONS Remove Insert List of Effective Pages (pp.1-21) List of Effective Pages (pp.1-21)
Volume 1, Table of Contents (pp. i-xvi) Volume 1, Table of Contents (pp. i-xvi)
Volume 1, Page 1.1-1 Volume 1, Page 1.1-1 Volume 1, Page 2.1-2 Volume 1, Page 2.1-2 Volume 1, Page 2.1-3 Volume 1, Page 2.1-3 Volume 1, Page 2.7-1 Volume 1, Page 2.7-1 Volume 1, Page 3.1-1 Volume 1, Page 3.1-1 Volume 1, Table 3-3 Volume 1, Table 3-3 Volume 1, Page 4.4-1 Volume 1, Page 4.4-1 Volume 1, Page 4.7-2 Volume 1, Page 4.7-2 Volume 1, Page 4.7-3 Volume 1, Page 4.7-3 Volume 1, Page 7 .2-1 Volume 1, Page 7.2-1 Volume 1, Page 7.5-1 Volume 1, Page 7.5-1 Volume 1, Page 7.5-2 Volume 1, Page 7.5-2 Volume 1, Page 7 .6-1 Volume 1, Page 7.6-1 Volume 1, Page 8.2-1 Volume 1, Page 8.2-1 Volume 1, Page 9.1-1 Volume 1, Page 9.1-1 Volume 1, Page 9.1-2 Volume 1, Page 9.1-2 Volume 1, Page 9.1-3 Volume 1, Page 9.1-3 Volume 1, Page 9.2-1 Volume 1, Page 9 .2-1 Volume 1, Page 9.6-2 Volume 1, Page 9.6-2 Volume 1, Page 9.7-1 Volume 1, Page 9.7-1 Volume 1, Page 11.1-1 Volume 1, Page 11.1-1
LIST OF EFFECTIVE PAGES Docket No. 72-11 June2020 Revision 9 Page or Figure Number ISFSI FSAR Revision No.
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8.1-55 Revision4A 8.1-56 Revision4A 8.1-57 Revision 4A 8.1-58 Revision 4A 8.2-13 Revision4A 8.2-14 Revision4A 8.2-15 Revision4A 8.2-16 Revision 4A 8.2-17 Revision4A 8.2-18 Revision4A 8.2-19 Revision 4A 8.2-20 Revision4A 8.2-21 Revision4A 8.2-22 Revision 4A 8.2-23 Revision4A 8.2-24 Revision 4A 8.2-25 Revision 4A 8.2-26 Revision 4A 8.2-27 Revision4A 8.2-28 Revision4A 8.2-29 Revision 4A 8.2-30 Revision4A 8.2-31 Revision4A 8.2-32 Revision4A 8.2-33 Revision4A 8.2-34 Revision4A 8.2-35 Revision 4A 8.2-36 Revision4A 8.2-37 Revision 4A 8.2-38 Revision 4A 8.2-39 Revision4A 8.2-40 Revision4A 8.2-41 Revision4A 8.2-42 Revision4A 8.2-43 Revision4A 8.2-44 Revision4A 8.2-45 Revision4A APPENDIXC 1-1 Revision 0 1-2 Revision 0 1-3 Revision 0 2-1 Revision 0 3-1 Revision 0 LIST OF*;EFFECTIVE PAGES Docket.No. 72-11 June 2020 Revision 9 Page or Figure Number ISFSI FSAR Revision No.
3-2 Revision 0 3-3 Revision 0 3-4 Revision 0 3-5 Revision 0 4-1 Revision 0 4-2 Revision 0 4-3 Revision 0 4-4 Revision 0 4-5 Revision 0 4-6 Revision 0 5-1 Revision 0 5-2 Revision 0 6-1 Revision 0 7-1 Revision 0 7-2 Revision 0 7-3 Revision 0 7-4 Revision 0 7-5 Revision 0 7-6 Revision 0 8-1 Revision 0 8-2 Revision 0 9-1 Revision 0 9-2 Revision 0 10-1 Revision 0 10-2 Revision 0 10-3 Revision 0 10-4 Revision 0 10-5 Revision 0 10-6 Revision 0 11-1 Revision 0 TABLE OF CONTENTS
- 1. Introduction and General Description of Installation 1.1-1 1.1 Introduction 1.1-1 1.1.1 Principal Function of the Installation 1.1-1 1.1.2 Proposed Location of the ISFSI 1.1-1 1.2 General Description of the Installation 1.2-1 1.2.1 Horizontal Storage Module 1.2-1 1.2.2 Dry Shielded Canister (DSC) 1.2-2 1.2.3 NUHOMS-MP187 Cask 1.2-3 1.3 General Systems Description 1.3-1 1.3 .1 Storage Systems 1.3-1 1.3 .2 Transfer System 1.3-1 1.3.3 Auxiliary Systems 1.3-1 1.3.3.1 The Vacuum Drying System (VDS) 1.3-1 1.3 .3 .2 The Welding System 1.3-1 1.3.3.3 The Waste Processing System 1.3-2 1.3.3.4 The Security System 1.3-2 1.3.3.5 The Temperature and Pressure Monitoring System 1.3-2 1.4 Identification of Agents and Contractors 1.4-1 1.5 Material Incorporated by Reference 1.5-1 1.6 References 1.6-1
- 2. Site and Environment 2.1-1 2.1 Geography and Demography 2.1-1 2.1.1 Site Location 2.1-1 2.1.2 Site Description 2.1-1 2.1.2.l Other Activities Within the Site Boundary 2.1-2 2.1.2.2 Boundaries for Establishing Effluent Release Limits 2.1-2 2.1.3 Population Distribution and Trends 2.1-2 2.1.4 Uses ofNearby Land and Waters 2.1-3 2.1.4.l Land Use 2.1-3 2.1.4.2 Access and Egress 2.1-4 2.1.4.3 Water Supply 2.1-4 2.2 Nearby Industrial, Transportation, and Military Facilities 2.2-1 2.2.1 Industrial 2.2-1 2.2.2 Transportation 2.2-1 2.2.3 Military 2.2-2 2.3 Meteorology 2.3-1 2.3.1 Regional Climatology 2.3-1 2.3.1.1 General Climate 2.3-1 2.3.1.2 Severe Weather 2.3-2 Volume I Revision 9 Rancho Seco ISFSI FSAR June 2020 I
I --
TABLEd~ tONTENTS 0
Page 2.3 .2 Local Meteorology . 2.3-3 2.3 .2.1 Data Sources 2.3-3 2.3.3 On-Site Meteorological Measurements Program 2.3-4 2j,4 Diffus1on Estimat~s . , 2.3-4 2.4 Hydrology 2.4-1 2.4.1 Characteristics of Streams and,L:akes in Vicinity 2.4-1 2.4.2 Topography 2.4-1 2.4.3 Terminal Disposal of s.treamRunoff 2.4-1 2.4.4 Historical Flooding *.. . _
- 2.4-1 2.4.4.1 Floods From Frontal Storms . 2.4-2 2.4.4.2 Floods From'.Thunderstorms 2.4-2 2.4.4.3 Spillway Capacity 2.4-2 2.4.5 Prediction of Land Urbanization 2.4-3 2.4.6 Groundwater 2.4-3 2.4.6.1 Occurrence and Movement 2.4-3
. 2:4,6.2 W~t~r Supply 2.4-4 2.(6.3Quality .. " 2.4-5 2.4.7,_. \Yells . ,,- ..,. : :-.l:L. ,, *,... 2.4-5 2.5 Geology and Seismology 2.5-1 2.5.1 Geology 2.5-1 2.5.2 Seismology 2.5-1 2.6
- Soils 2.6-1 2.6.1 Rancho Seco Site *. *.
- 2.6-1 2.6.2 Rancho Seco ISFSl Site. 2.6-1
- 2. 7 References 2.7-1
- 3. Principal Design Criteria 3.1-1 3 .1 Purpose of Installation 3.1-1 3 .1.1 Material to be Stored 3.1-1 3 .1.1.1 Physical Characteristics . 3.1-1 3 .1.1.2 Thermal Characteristics . 3.1-1 3 .1.1.3 Radiological Characteristics 3.1-2 3 .1.2 General Operating Functions 3.1-3 3.1.2.1 Handling and Transfer Equipment 3.1-3 3.1.2.2 Waste Pro~essing, Paqkaging and Storage Areas 3.1-3 3.2 Structural and Mechanical Safety Criteria 3.2-1 3 .2.1 ., Tornado and :Wind Loadings. 3.2-1 3 .:i. '1 .1 Applicable *Design Parameters . , . 3.2-1 3.2.1.2 Determination.of Forces on Structures 3.2-1 3 ;2.13 AbVity of' sini~tures -to Perform Despite Failure
.. of Structures N9t_bes\gned for Tornado Loads 3.2-2 3.2.1.4 Tornado Missiles *
TABLE OF. CONTENTS 3.2.2 Water Level (Flood) Design 3.2-2 3.2.2.1 Flood Elevations 3.2-3 3.2.2:2 Phenomena Considered in Design Load Calculations 3.2-3 3.2.2.3 Flood Force Application *
- 3.2-3 3.2.2.4 Flood Protection * ' 3.2-3 3.2.3
- Seismic Design , .* ., , .. ,. , 3.2-3 Seismic-System Analyses **., 1,; 3.2-4 3.2.4 Snow and Ice Loadings ' .. ;.',., 3.2-5 3.2.5 Load Combination Criteria: . . - 3.2-6 3.2.5.1 Horizontal Storage,Module0>* 3.2-6 3.2.5.2 Dry Shielded Canister *, ,* 3.2-6 3.2.5.3 NUHOMS ..MP187,Cask 3.2-7 3.3 Safety Protection System 3.3-1 3.3.1 General 3.3-1 3.3.2 Protection by Multiple Cblifinerri~ri.t Barriers & Systems 3.3-1 3 .3 .2.1 Confinement Barriers and Syste1Usi .. 3.3-1 3 .3 .2.2 Ventilation - Offgas * * * *
- _* _7 ;_: .. . 3.3-1 3.3.3 Protection by Equipment and Instrumentati~h Selection* 3.3-2 3.3.3.1 Equipment 3.3-2 3.3.3.2 Instrumentation 3.3-2 3.3.4 Nuclear Criticality Safety 3.3-2 3 .3 .4.1 Control Methods for the Prevention of Criticality . 3.3-3 3 .3 .4.2 Spent Fuel Loading. ,. , * **'
- 3.3-5 3 .3 .4.3 Model Specifica~ion _: . 3.3-5 3 .3 .4.4 Criticality Calculation 3.3-7 3.3.4.5 Error Contingency Criteria 3.3-11 3.3.4.6 Verification Analysis 3.3-11 3.3.5 Radiological Protection 3.3-13 3.3.5.1 Access Control 3.3-13 3.3.5.2 Shielding 3.3-14 3.3.5.3 Radiological Alarm Systems* 3.3-14 3.3.6 Fire and Explosion Protection
- 3.3-14 3 .3. 7 Materials Handling and Storage 3.3-15 3.3.7.1 Spent Fuel Handling and Sfora:ge 3.3-15 3.3).2 Radioactive Waste Treatment 3.3-15 3.3.7.3 Waste Storage Facilities
- 3.3-15 3.3.8 Industrial and Chemical Safety*
- 3.3-15 3.4 Classification of Structures; Components, arid _Sy~tems 3.4-1 3.4.1 Major ISFSI Components : : : .. . 3.4-1 3 .4.2 Geological 'and Seismofogtdaf 'Char~cteristics 3.4-1 3.4.2.1 Soil ~h<;ll"acteri~tfos atthe'I_~FSI Pad"* 3.4-1 3 .4.2.2 'Soil Liquefaction_ : ' 3.4-2 1
Volume I Revision 9
.:. . Rancho Seco ISFSI FSAR 111 June 2020
TABLE d{CONTENTS 3.4.2.3 Soil Amplification Due to Soil-Structure Interaction 3.4-3 3.5 Decommissioning Considerations 3.5-1 3.6 Summary of ISFSI Design Ctiteda .
3.6-1 3.7 References 3.7-1
'4': Installation Design 4.1-1 4.1 Summary Description , 4.1-1 4.1.1 Location and Layout of Installation 4.1-1 4.1.2 Principal Features.. ;.; 4.1-1 4.2 Storage Structures 4.2-1 4.2.1 Design Bases and Safety Assurance .' . 4.2-1 4.2.2 Compliance with General Design Criteria **
- 4.2-2 4.2.2.1 10 CFR 72.122 Overall Requirement~. . 4.2-2 4.2.2.2 10 CFR 72.124 - Criteria for Nuclear Criticality Safety 4.2-3 4.2.2.3 10 CFR 72.126 Criteria for Radiological Protection' 4'.2-4 4.2.2.4 10 CFR 72.128 Criteria.for Spent Fuel, High-level Radioactive Waste, and other Radioactiye Waste Handling and Storage 4.2-4 42.2.5 10 GFll: 72.130 Criteria for Decommissioning 4.2-5 4.2.3 Structural Specificatiqns 4.2-5 4.2.4 Installation Layout _ 4.2-5 4.2.5 Individual Unit.I>escription 4.2-5 4.2.5.1 Horizontal Storage Module (HSM)* 4.2-5 4.2.5.2 Dry Shieldeq.C~ster .. 4.2-5 4.2.5.3 The NUHOMS-MP187 Cask 4.2-8 4.3 Auxiliary Systems - * . 4.3-1 4.3.1 Ventilation and Offgas Requirements 4.3-1 4.3 .2 Electrical System Requirel]lents. 4.3-1 4.3 .3 Air Supply System 4.3-2 4.3.4 Steam Supply and Distribution System 4.3-2 4.3.5 Water Supply System 4.3-2 4.3.6 Sewage Treatment System 4.3-2 4.3.7 Communication and Alarm Systems 4.3-2 4.3.8 Fire Protection*system
- 4.3-3 4.3.9 Cold Chemical System 4.3-3 4.3.10 Air Sampling System 4.3-3 4.4 Decontamination System . 4.4-1 4.4.1 Equipment Decontamination 4.4-1 4.4.2 Personnel Decontamination 4.4-1 4.5 Repair and Maintenance 4.5-1 Volume I Revision 9 Rancho Seco ISFSI FSAR IV June*i'o20
TABLE Of CONTENTS Page 4.5.1 Repair 4.5-1 4.52,. Maintenance 4.5-1 4.6 Cathodic Protection 4.6-1 4.7 Fuel Handling Operation Systems 4.7-1
- 4. 7.1 Individual Unit Description* 4.7-1 4.7.1.1 Function 4.7-1 4.7.1.2 Lifting Yoke and Extension 4.7-2
- 4. 7 .2 Design Bases and Safety Assurance 4.7-2 4.7.3 Structural SpecificatiQns * *. 4.7-2 4.7.4 Installation*Layotlt . , . 4.7-2 4.7.5 Individual Unit Descriptions** 4.7-3 4.7.5.1 Function 4.7-3 4.7.5.2 Components
- 4.7-3 4.7.6 Design Basis and Safety Assurance .
- 4.7-3 4.8 References ;* 4.8-1 Operation Systems .. _.. 5.1-1 5.1 Operation Desctfption '* :, } * . 5.1-1 5.1.1 Narrative Description ** ' ,, *.,.
- 5.1-1 5.1.2 Process Flow Diagram . -- 5.1-1
- 5. l .3 Ideritification of Subjects for Safety Analysis 5.1-2 5.1.3.l Criticality Control * '* * ** 5.1-2 5.1.3.2 Chemical Safety 5.1-2 5.1.3.3 Operation Shutdown Modes; 5.1-2
- 5. l .3A Instriunentation 5.1-2 5.1.3.5 Maintenance"{echhiques* ** 5.1-2
, I / *1 5.2 Fuel Handling Systems 5.2-1 5.2.1 Spent Fuel Handling and Transfer 5.2-1 5 .2.1.1 Function Description 5.2-1 5.2.1.2 Safety Features 5.2-2 5.2.2 Spent Fuel Storage 5.2-2 5.2.2.1 Safety Features 5.2-2 5.3 Other Operating Systems . 5.3-1 5.3.l Operating System* . . 5.3-1 5.3.2 Component/Equipment Spares 5.3-1 5.4 Operation Support System 5.4-1 5.4.1 Instrumentation and Control Sys:tem.s
- 5.4-1 5.4.2 System and Component Spares. 5.4-1 5.5 Control Room and/or Control-Areas ; .: .. 5.5-1 5.6 Analytical Sampling 5.6-1
- . Volume I Revision 9
TABLE OF CONTENTS 5.7 References 5.7-1
- 6. Waste Confinement and Management 6.1-1 6.1 Waste Sources 6.1-1 6.2 Off-gas Treatment and Ventilation. *, '* *: *
- d 6.2-1 6.3 Liquid Waste Treatment and Retention*
- 6.3-1
.; . ;:~ .-
6.4 Solid Wastes 6.4-1
~ *\ .:; . ** ,*
6.5 RadiologicalJmp~ct pf NQrm~l Operations :-. Summary 6.5-1 6.6 References 6.6-1 7..,., Radiation Protection 7.1-1 7.1 Ensuring that Occupational.' Radiation Exposures Are As Low As Is Reasonably Achievable (ALAR.A) ' 7.1-1 7 .1.1 Policy Considerations 7.1-1 7 .1.2 Design Considerations . 7.1-1 7 .1.3 Operational Consideratiqns 7.1-3 7.2 Radiation Sources : *-- .* *-,. ** '. . . . .; 7.2-1 7.2.1 Characterization of Sources 7.2-1 7.2.2 Airborne Radioactive Material Sources 7.2-1 7.2.3 Sealed Sources 7.2-2 7.3 Radiation Protection D_esign Features_
- 7.3-1 7.3.1 Installation Design Features . : 7.3-1 7.3.2 SNeldjng . .
- 7.3-1 7.3.2.1 Radiation Shieiding Design Features 7.3-1 7.3.2.2 Shielding Analysis 7.3-1 7.3.3 Ventilation *
- 7.3-1 7.3.4 Area Radiation.an<l ,A,i~bome Radipactivhy Monitoring Instrum~ntation 7.3-2 7.4 Estimated Onsite Collective Dose. Assessment .
- 7.4-1 7.4.1 Operational Dose Assessment 7.4-1 7.4.2 Site Dose Assessment 7.4-2 7.5 Health Physics Program 7.5-1 7.5.1 Organization .. 7.5-1 7.5.2 Equipment, Instrumentation, and Facilities.** 7.5-1 7.5 .3 Procedures 7.5-2 7.6 Estimated Offsite Collective Dose Assessment 7.6-1 7.6.1 pffluent a11,d Envirpnmental Monitoring Program 7.6-1 7.6.2 Analysis ofMultiple,,Contribu,tion 7.6-1 7.6.3 Estimated Dose Equivalents 7.6-1 7 .6.4 Liquid Release 7.6-1 Volume I Revision 9 Rancho Seco ISFSI FSAR vi June;2020
TABLE OF CONTENTS 7.7 References 7.7-1
- 8. Analysis of Design Events 8.1-1 8.1 Normal and Off-Normal Operations 8.1-1 8.1.1.1 Dead Weight i , : *, 8.1-1 8.1.1.2 Design Basis Internal Pressure 8.1-6 8.1.1.3 Design Basis Thermal Loads 8.1-7 8.1.1.4 Normal Operatioria'.lHandliiig Load~* - 8.1-9 8.1.1.5 Off-Normal Handling Loads 8.1-12 8.1.1.6 Design Basis Live Loads 8.1-12 8.1.1.7 DSC Fatigue Evaluation . . 'J 8:1-13 8J.l.8 Cask Fatigue Evaluation, ,. 8.1-13 8.1.1.9 Thermal Cycling of the Cas~ 8.1-14 8.1.2 Horizontal Storage Module 8.1-14 8.2 Accident Analyses for the ISFSI *! , 8.2-1 8.2.1 Accidental Cask Drop ' . -: . : *- 8.2-1 8.2.1.1 Postulated Cause of Event ;,.;;;' ;, , .. 1.:'..
- 8.2-2 8.2.1.2 Detection ofEv.ent.: .* _ 8.2-2 8.2.1.3 Analysis of Effects and Consequences 8.2-2 8.2.1.4 Corrective Actions 8.2-9 8.2.2 DSC Leakage . . .. __ 8.2-10 8.2.2.1 Postulated* Cause ofE'7ent 8.2-10 8.2.2.2 Detection ofEverif' .. :. . ._, 8.2-10 8.2.2.3 Analysis of Effects and C::onsequences 8.2-10 8'.2.2.4 Corrective Actions 8.2-11 8.2.3 Accident Pressurization 8.2-11 8.2.3.1 Postulated Cause of Event' .
- 8.2-11
- 8.2.3.2 D~tection of Event' ., ' 8.2-11 8.2.3.3 Analysis of Effects and Con~equences 8.2-11 8.2.3.4 Corrective Actions 8.2-12 8.2.4 Earthquake 8.2-12 8.2.4.1 Postulated Cause of Event 8.2-12 8.2.4.2 Detection of Event 8.2-12 8.2.4.3 Analysis of Effects and Consequences* 8.2-12 8.2.4A Corrective Actio.ns 8.2-13 8.2.5 Fire 8.2-13 8.2.5.1 Postulated Cause of Event 8.2-13 8.2.5.2 Dectection 'of Event * - *. 8.2-13 8.2.5.3 Analysis of Effects *and Consequences 8.2.14 8.2.5.4 Corrective Action. *: : ; -' 8.2-15
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. -*-Rancho Seco ISFSI FSAR Vll June 2020
TABLE-Of CONTENTS 8.3 Load Combination Evaluation 8.3-1 8.3 .1 DSC Load Combination Evaluation 8.3-1 8.3.2 Cask Load Combination Evaluation 8.3-2 8.3.3 Summary of Design Requirements Met 8.3-2 8._4 Site Characteristics Affecting Safety Analysis 8.4-1 8.5 References 8.5-1 Conduct of Operations 9.1-1
.-i 9.1 Organizational Structure* * . 9.1-1 9 .1.1 Corporate Organizatfon ,. - . _ 9.1-1
- 9. l .1.1 Corporate Functions~* Responsibilitie~, and Authorities 9.1-1 9J.Li in-House*6rganiz~tfon - * -_ * : * . 9.1-2 9.1.1.3 Interrelationship with Contractors and Suppliers 9.1-2
- 9.1.1.4 TechriicalStaff * '
- 9.1-2 9 .1.2 Operaiing Organization, Management and
,-'Administrative Controi~ystem * . '* 9.1-2 9 .1.2._1 Onsite Orga11iz~iion ' _ 9.1-2
... -~J. L2.2 PersoAfi~f Functions, :Responsibilities and Authorities 9.1-2 9.1.3 Personnel QualificatfonR.equirenients 9.1-4 9 .1.4 Liaison with Ot~ei: Organizations 9.1-4 9.2 Pre-Operational Testi11g and Operation , 9.2-1 9), 1 Administrative Propedures for Conducting Test Program 9.2-1
- 9.2.2 . -. Test P,rogrfil11 I)es~ripti~n 9.2-1 9 .2.2.1 Phy~ical Facilities and Operations 9.2-1 9.2.3 Test Discussion: , . 9.2-2 9.3 Training Program 9.3-1 9.3. I Program Description 9.3-1 9.3-.-1.1 Scope of Training 9.3-1 9.3.1.2 Radiation Protection Technician Training 9.3-2 9.3.1.3 Dry Fuel Storage Equipment Operator Training 9.3-2 9.3.1.4 Maintenance Training 9.3-3 9.3.1.5 Trainer Qualifications 9.3-3 9.3 .2 Administration and Records 9.3-3 9.4 Normal Operations 9.4-1 9 .4.1 Procedures
- 9.4-1 9 .4 .1.1 :Administrative Procedures 9.4-1
_:9 .4.1.2 R;idiatio.p: Pl}otection Procedures ** 9.4-1 9.4.1.3 Maintenance.Procedures.* 9.4-1 9 .4 .1.4 Op~r_atiµg _Propedures 9.4-1 9.4.1.5 Test Procedures 9.4-2
'Volunie I Revision 9
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TABLE OP.CQNTENTS 9 .4 .1.6 Pre-operationar Test Procedures 9.4-2 9.4.2 Records * *. 9.4-2 9.5 Emergency Plai:nnng .* *. 9.5-1 9.6 Decommissioning Pl~ 9.6-1 9.7 References 9.7-1 9.8 Aging Management ':.;**-.*;* ,:": .... **: ... ._.
9.8-1 9.8.1 Scoping Evaluation Methodol9gy, , . . 9.8-1
- 9. 8.2 Results of Scoping Evalua,tjon * *. * .., ,, *:' * * * / .,._ 9.8-2 9.8.3 Aging.Management Review,* '. '*': :. . ... 9.8-3 pf
.. 9:8.3.1 Results Aging M~ge~entReview DSC 9.8-5
.9.,8.3.2 Rest1lts of Aging.Management.Review HSM 9.8-6 9:g.3.3 *Results of Aging :tvfanageni~nt Review-
. ... Concrete Basemat : . ** 9.8-7 9.8.3.4 Results of Aging MaiJage.tl}ent R..e1tlew ~
Transfer Cask . - ., . *.. .. . ....... , ,
9.8.3.5 Results ofAgi~g'Mauage~en,t Re;i~~
Spent Fuel Assemblies . . ., .. 9.8-9 9.8.4 Summary ot":nme-Lunited Aging Analyses . 9.8-10 9.8.4.l DSC Time-Limited Aging.Analyses . 9.8-10 9.8.4.2 HSM Time-Limited Aging Analyses 9.8-11 9.8.4.3 Transfer CaskTime"Litnited*A:ging Analyses 9.8-11 9.8.4.4 ISFSI BasematTime-Lirriited Aging*Analyses 9.8-11 9.8.5 Summary of Aging Management Programs** 9.8-11 9.8.5.1 DSC Aging Management Program*
- 9.8-12 9.8.5.2 HSM Aging Management Progrfilll 9.8-12 9.8.5.3 TC Aging Managem~nt Program 9.8-12 9.8.5.4 ISFSI Basemat Aging Management Program 9.8-12 9.8.6 f\ging Managell_lent Tollgates . 9.8-12
- 10. Operating Controls and Limits 10.1-1 10.1 Proposed Operating Controls and Limits 10.1-1 10.2 Development of Operating Controls, arid Limits
- 10.2-1 10.2.1 Functional and Operating.Limits;:Monitoririg Instruments, and Limiting Control :Settings ; *. * **
- 1 , 10.2-1 10.2.2 Limiting Conditions forOpetation: 10.2-1 10 .2.2.1 Equipment 10.2-1
- ' , .V.qlurne I . . Revision 9
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TABLE'bF-CONTENTS Page
- 10.2.2.2Technical Conditions and Characteristics *, '10.2-1 10.2.3 Verification Requirements .10.2-3 10.2.4 Design Features - - f0.2-3 10.2.5 Administrative Controls
- 10.2-3 10.2.5 .1 Qualification of Spent Fuel 10.2-4 10.2.5.2 Spent Fuel Identification - 10.2-5 10.3 Operating Control and Limit Specifications 10.3-1 10.3.1 Spent Fuel Speci:tfcations - -- 10.3-1 10.3.1.lFO and FC-DSCFuel Specifications 10.3-1 10.3.l.2FF-DSC Fu~l Specific~tions J0.3-1 10.3.2 DSC Vacuum Pre~sur~ During Drying _ 10.3-2 10.3.3 DSC Helillill Ba'.ckfill*Pressure - * - 10.3-3 10.3.4 DSC Helium Leakage Rate*oflhner Seal*W~ld . -
- 10.3-4 10.3.5 DSC Dye Penetrant Test of Closure Welds . . lQ.3-5 10.3:6 DSC Surface Co~ta.minati~i'i. _- . - . :. :.' i0.3-6
-- **. t. *103-.7 ISFSl Security Area Dose-Rate
- 10.3-7 10.3.8 Cask, DSC, and Fuel Assembly Inspection Following ,_
- Accidental Cask Drop - )0.3-8 10.3.9 Pos! Fire Recovery Plan - . 10.3-9 10:3.lODSCLiftirrg Heights-: - - - rn.3-10 10.3.l lDSC Top Eng Dose Rates :10.3-11 10.3.12HSM Dose Rates 10.3-12 10.3.13Transfer Cask Dose Rates 10.3-13
.* 10_.3 .14 DSC- Re-flpog, Flow Rate 10.3.14 10.3.15 Heatup Duration of a Loaded DSC Filled with Water
' * * ' * ** r 10.3.15 10 .4 References 10.4-1
- 11. Quality Assurance 11.1-1 11.1 Sacramento Municipal Utility :Oistrict Quality Assurance Program 11.1-1 11.2 Quality Assurance Program - Contractors _ 11.2-1 11.2.l Architect-Engineer . 11.2-1 11.2.2 Storage System Supplier * * * . 11.2-1 11.3
References:
11.3-1 Appendix A ASME Code Exception List Appendix B Standardifed S.i'\R _llefer~11~es Appendix C Greater than Class C Waste Volume I .Revision 9 Rancho Seco ISFSI FSAR X June 2020
LIST OF TABLES
. . Table 2-1 Permanent .Populatio:n. Distribution Withi,n J 3 .Miles of RSN GS
- T,cible 2-2 Highest One-Minute Average Windspeeds:
.. Table 2-3 Precipitation Climatology
- Table 2-4 Precipitation Intensity
-Table 2-5 Mean Number of Days of Thunderstorms .*; *
.. *' ~' ; + ~ .. . .. '* .
. Table 3-1 Rancho Seco Fuel Ch~acteristic~. * *: . *
-Table 3-2 Rancho Seco Control Element Characted~tics
- Table 3-3 DSC Loading Summary for CaskHa11dHn~:*Cqnditio°'s
- Table 3-4 Cask Loading Summary fot Cask Handling Conditions*
.* '.:Table 3-5 Summary of Dry Shielded Canister Capa~ify.
and. Internal Pressure; Design Criteria
- .Table 3-6 DSC Load Combinations and Service:LevelsforCask*Storage a:nd Handling Modes -,. ::* .. *
- : Tc).ble 3-7 Structural Design Criteria for DS~ ** *: / :\: :. ,
- L : ' - ,.
- Table 3-8 Cask Load Combinations and Service Lev~ls {or Cask Handling Modes Table 3-9 Structural Design Criteria for On-Site Transfer Cask .
Table 3-10 Structural Design Criteria for Bolts Table 3-11 Rancho Seco ISFSI Major Componenfs 1and Safety Classifications
. Table 3-12 Summary oflSFSI Design Criteria .
Table 3-13 Maximum Fuel Loading Parameters Table 3-14 Design Basis Fuel Parameters for Criticality Analysis Table 3-15 KENO Model Atom Densities Table 3-16 KENO Model Moderator Atoin Densities--
Table 3-17 Summary of KENO Parametric Studies.
Table 3-18 MP187/FO-DSC KENO Results (Guide Sleeve Deformation)
-Table 3-19 MP187/FO-DSC KENO Results (Cask Layer Removal)
Table 3-20 MP187/FO-DSC KENO Results * .. '.' ...~
Table 3-21 MP187/FF-DSC KENO Results Table 3-22 MP187/FF-DSC KENO Results (Cask L~yer Retnoval)
Table 3-23 Benchmark Calculation Results
.Volunie I Revision 9
-Rancho Seco ISFSI FSAR Xl June.2020
Table 5-1 Inst~entation Used Duririg NUHOMS System Loading Operation~
Table 7-1 Design Basis Neutron Source Per Assembly and Energy Spectl;11Ill
, ' . . t -* .. ** ' _*,_,
Table 7-2 Design Basis Gamma-Ray Source per Assembly and Energy Spectrum Table 7-3 Estimated-Occµpational Ex:posµre for One HSM _Load _. --
Table 7-4 Rancho Seco ISFSI Area Dose Rates Table :8-1 NUHOMSISFSI Normal and Off-Normal Operating Loading-Summary:
Table 8-2a FO/FC/FF DSC Cavity Normal, Off-Normal, and Accident Pressures Without Control Components* ' - . - - ' - .
Table 8~2b Pres*shles',fo the FC/FF DSC with Control Components Table 8-3 Cask ISFSI Normal and Off-Normal Operating Condition Stresses Table 8-4 FO-DSC ISFSI Norm~l anddi(No~al Op.erati~g Conditi.on: Stre'sse~
Table 8-5 FC-DSC ISFSI Normal and Off-No~al c'onditio-~ Stresses -. , _*.
., ' . ; . . '. . . . . l
- Table 8-6 FF-DSC ISFSI :Normal and Off-Normal Condition Stresses Table 8-8 Cask ISFSI Accident Condition Stresses Table 8-9 FO-DSC ISFS.I Accide~t C~ndition shesses Table 8-10 FC-DSC ISFSI Accident Condition Stresses Table 8-11 FF-DSC ISFSI Accident Condition Stresses Table 8-12 ~ask Enveloping Lo~d Combiriation Results for Normal and Off-Normal Loads
-- :(ASME Service Leveh;: A and B)
Table 8-13 Cask Enveloping Load Combination Results for Accident Loads (ASME Service Level C)
Table 8-14 Cask Enveloping Load Combination Results for Accide.nt Loads (ASME S.ervice Level D)
Table 8-15 FO-DSC Enveloping Load Combination Results for Normal and Off-Normal Loads (ASME Service Levels A and B)
Table 8-16 FO-DSC Enveloping Load Combination Results for Accident Loads (ASME Service Level C)
Table 8-17 FO-DSC Enveloping Load Combination Results for Accident Loads (ASME Service Level D)
Table 8-18 FC-DSC Enveloping Load Combination Results for Normal and Off-Normal Loads (ASME Service Levels A and B)
Table 8-19 FC-DSC Enveloping Load Combination Results for Accident Loads (ASME Service Level C)
Volume I Rev'ision 9
_LIST OF TABLES Table 8-20 FC-DSC Enveloping Load Comb.inati~~ Results for Accident Loads (ASME
. Service Level' D) . . ., . .
Table 8-21 FF-DSC. Envei~ping Load Combination Results for Normal and Off-Normal Loads (ASME-Service Levels A and B)
Table 8-22 FF-DSC Envelbping Load Combination Results for Accident Loads (ASME Service Level C)
Table 8-2~;FF-I)SC Enveloping Load GombinatiQn,Results for-A~cidentLoad~ (ASME .
Service Level D) * * * * * * ** * * - * * * * * * ** * *
- l .-* : ' '* .~ ***
Table 8-24 Maximum Pressure Differential Across DSC.Shell . '
~ . .
~ '
Table 8-25 Cask Cavity J:lres~ure A,ssurajµg D,S.~-_Lr;,aj(age:After_I_>lacem~n..t inS!~r-~ge,,; : ,
(Delet~d)
- > *\ * * * ' ... * * * * *, ' ii: : '. .. :: . ~- ..*. :" .
Table 9.8-1 Scoping Evaluation of Rancho Seco ISFSI SSCs
~ . *
- i , * * ; :, '
- I , ,. I' :
Table 9.8-2 Scoping Evaluation Results for SF~s Table 9 .8-3 Rancho Seco FO~DSC Intended F~~iions and J\MR Results ..
Table 9.8-4 Rancho Seco FC-DSC Intended Functions and AMR Results ,*:.
r : .._.
Table 9.8-5 Rancho Seco FF-DSC Intend~d F~cti~ns' .. ~1iAMR°:R~s~lts** ,.
Table 9.8-6 Rancho Seco GTCC-DSC Intend~ci F~~ctions and AMR Results Table 9.8-7 Rancho Seco HSM lritended Functions .
and. . AMR.Results Table 9.8-8 MP187 TC Intended F~ctions a~d AMR Results .
. ' ~ .. ..
Table 9.8-9 DSC External Surfaces Aging Management Program
. : . ' *:.*.. -! . ' ,- *. , ' .,1 **
Table 9.8-10 HSM Aging Management Progr~ for ~x~emal and Inte111al Surfaces
-Table 9.8-11 TC Aging~Management Program .
- Table 9.8-12 ISFSI Basemat Aging Management Program Table 9.8~13 Rancho 8eco ISFSI'Tollgates
- Table 10-1 Area Where Controls and Limits Are
. \ .
Specified
,,i
LIST OF FIGURES Figure 1-1 Rancho Seco ISFSI Location Figure 1-2 ISFSI Layout Figure 1-3 Overview of the Horizontal Storage Module Figure 1-4 Overview of the Dry Shielded Canister Figure 1-5 Overview of the Cask Figure 1-6 Rancho Seco ISFSI Storage System Flowchart Figure 1-7 General Arrangement of the Transfer System Figure 2-1 Regional Map ofRSNGS Figure 2-2 RSNGS Site Figure 2-3 Rancho Seco ISFSI Site Figure 2-4 Permanent Population Surrounding RSNGS (deleted)
Figure 2-5 Wind Trajectories for RSNGS Figure 2-6 Wind Trajectories at RSNGS Figure 2-7 Boring Location Map Figure 2-8 Subsurface Exploration Log B-1 Figure 2-9 Subsurface Exploration Log B-2 Figure 3-1 KENO Model and DSC Basket Figure 3-2 Exploded View of KENO Model Figure 3-3 Structure of KENO Model UNITS 33 and 34 Figure 3-4 Keno Model UNITS 1-8 Figure 3-5 KENO Model of a Design Basis Fuel Assembly Figure 3-6 K vs. Guide Sleeve Deformation Figure 3-7 FF-DSC Broken Fuel Rod Models Figure 3-8 FF-DSC Double-Ended Rod Break Models Figure 3-9 NUHOMS-MP 187 Cask/PO-DSC Criticality Results Figure 3-10 NUHOMS-MP187 Cask/FF-DSC Criticality Results Figure 3-11 Critical Benchmark Results Figure 5-1 DSC Loading Operations Flow Chart Figure 5-2 DSC Sealing, Draining and Drying Operations Flow Chart Figure 5-3 Primary Operations for DSC Fuel Handling Volume I
Figure 5-4 Primary Operations for DSC Closure .
Figure 7-1 Occupational Exposure Contribution from Each DSC Loading Operation Figure 7-2 Rancho Seco ISFSI On-Site Dose Rates Figure 7-3 Dose Versus Distance from the Rancho Seco ISFSI
. Volume I
,.*.. .Reyision 9
LIST.OF ACRONYMS ACI American Concrete Institute .
AGM. Assistant General .Manager . . . .
AISC .of
- . Amerfcan Institute Steel Construction.
ALAR.A As Lo~ As Reasonably .Achie~able
, ANS . Amer,ican Nuclear Sodety.
- ,. ANSL .. .
_ .~erfoan.Natiop.~l ~tandards Institute ASME. American Society of, Mechanical Engineers
- ASTM American s,ociety fortesting and. Materials
- n&w**_, Babcock & WifooX *. . . .
- .:DOE U;S~ Departmerit. of~µeigy '* .
. FC-DSC
. FO-DSC * "* Fuel-Only DSC
- GM * . * 'denei-al Manager, *,.: .. :
. HSM. * *Horizontal.Storage Module .
Ib~B. . Inte;im .011site Storage Building ISFSI . , Independent $p~~t Fuel Storage:Installatio,n . _
. MP187 NUHOMS Multi-:-Purpose (Transfer ap.d Transportation) Cask
- NDRC National Defi:mse.ResearGh Committee.
NFPA National Fire Protection Association NRC
- U.S. Nuclear R.egiilatory Commission
- *NUHOMS *NUTECH Hori~ontal*Modular Storage NUREG Nuclear Regulatory Guide OSHA Occupational Safety and Health Administration PWR Pressurized Water Reactor RSNGS R~cho Seco N:u~lear Generating Station SAR . Safety Analys1s Report .
SFA :'!"' Spent Fuel Assembly SMUD~ Sacramento Municipal Utility District VDS Vacuum Drying System Volume I Revision 9 Rancho Seco ISFSI FSAR xvi* June2020
- 1. INTRODUCTION AND GENERAL DESCRIPTION OF INSTALLATION 1.1 Introduction The Nuclear Regulatory Commissibn (NRC) iss~ed-an operating license, DPR-54, for Rancho Seco Nuclear Generating Station (RSNGS) in August 1974, and the plant began commercial operation in April 1975. However, as a result of a public referendum of Sacramento Municipal Utility District (SMUD) \roters on June 6, 1989, RSNQS has ceased operation, and the reactor has been permanently defueled. *Accordingly, on May 20, 1991, SMUD submitted its Proposed Decommissioning Plan [l.6J] to the NRC discussing the method to be used to decommission RSNGS. "Tills plan*was approved by*an NRC order dated March 20, 1995. SMUD subsequently '.revised the decommissioning plan a Post :to Shutdown Decommissioning Activities Report.(PSDAR)to*meet the requirenients of the revised regulations (10 CFR 50.82) regarding decommissioning.. * . '
The Independent Spent F~ei Storage Installation. (ISFSI}-is intencled to provide* dry storage capacity for Rancho Seco spent nuclear fuel_ and. Greater than Class C (GTCC) radioactive waste. The storage system was:d~signe,d for 50,-year s~:rvice, and _initially licensed for 20 years in accordance with 10 CFR 72. On March 9, 2020, theNRC approved renewal of the ISFSI license (SNM-2510) for an additional40 .years. The aging management activities associated with this renewal applies to Amendmt?nt 4. AnY futur~ amendments will include an aging management review (AMR) and any associated reqµir~d aging management m
activities. The current aging management results ru:-_:l detaileq Ch.apter 9, Sectiop. 9.8.
The original ISFSI FSAR chapters indicate design life and' se~fo~ life values of 50 years 1.
The new design life is 60 years. Time-limited-aging-analyses (TLAA.s) to assess SSCs that have a time-dependent operating life to demonstrate that the existing licensing basis remains valid and that the intended functions of the SSCs iri scope ofrenewal are maintained during the period of extended operation (PEO) to 60 years are detailed in Chapter 9, Section 9.8.4.
Construction of the Rancho Seco ISFSI was co~p,l~t~d during. i 996 and the initial license was received on June 30, 2000. All fuel was in dry storage at the ISFSI in August 2002 and the single GTCC waste canister was loaded-at the 1s;FSiin: August 2006.
- 1.1.1 Principal Function of the Installation -
The Rancho Seco ISFSI design provides temporary dry storage *for* 100% of the Rancho Seco spent fuel assemblies (SF As) and GTCC waste in order to complete full plant dismantlement.
It is designed with safety features that eliminate the need for an operable spent fuel pool to recover from certain unlikely accident scenarios. The spent fuel will be stored in this manner until it is accepted by the Department ofEri.~rgy (DOE).Location of the ISFSI.
1.1.2 Location of the ISFSI The Rancho Seco ISFSI is located within the Owner Controlled Area of the Rancho Seco site which is owned and operated by SMUD. The Rancho Seco site comprises approximately 2,480 acres in Sacramento County, California. It is characterized by isolation from population centers, a sound foundation for structures, and favorable conditions of meteorology, seismology, and hydrology.
1 The terms design life and service life are equivalent and interchangeable.
_;, ,Volume I
- _Revision 9
.**. * .Rancho Seco ISFSI FSAR 1.1-.1 June 2020
The soils at the Rancho Seco site are sufficiently strong to safely support the Rancho Seco ISFSI structure and appurtenant facilities. These soils can be categorized as hard to very hard silts and silty clays with dense to very dense sands and gravels.
2.1.2.1 Other Activities Within the Site Boundary The Rancho Seco ISFSI lies wholly within the 2,480 acre Rancho Seco Nuclear Generating Station site. This site is owned and controlled by SMUD, who has full authority to determine all activities within the site including the exclusion and removal of individuals and property.
The Rancho Seco ISFSI Protected Area is approximately 225 feet X 170 feet in size. The Protected Area is located within licensed boundary denoted by the 100 meter fence surrounding the Protected Area. Also within the licensed boundary of the ISFSI lies the Fuel Transfer Equipment Storage Building (FTESB), a 40 foot X I 00 foot enclosure to store contaminated fuel handling and transportation support equipment while the spent nuclear fuel remains in storage.
SMUD has completed construction of a 500-MW natural gas fired power plant located approximately1/2 mile south of the Industrial Area boundary.
Access for transmission lines and water lines is :from the west and south sides of the property.
2.1.2.2 Boundaries for Establishing Effluent Release Limits .
There are no radioactive effluent releases associated with the Rancho Seco ISFSI.
2.1.3 Population Distribution and Trends The land surrounding the site is presently undeveloped and is used primarily for grazing beef cattle and other agricultural activities. The most recent population distribution estimates are contained in the "Evacuation Time Estimate for the Rancho Seco Plume Exposure Pathway Emergency Planning Zone" [2.2.2].
There are five counties (Amador, San Joaquin, Sacramento, El Dorado, and Calaveras) within a 15-mile radius of Rancho Seco. Only very small portions of El Dorado and Calaveras counties are within the 15-mile radius of Rancho Seco. There is no significant projected growth within these portions of these two counties. The projected development within Amador, Sacramento, and San Joaquin counties is discussed in Section 4.2 of the Rancho Seco ISFSI Environmental Report, Revision 1 [2.2.3]. A five-mile radius area surrounding the Rancho Seco facility is defined as the low population zone. This area is primarily farm land and vineyards, with few tourist attractions and little seasonal variation in the population.
The Rancho Seco Reservoir and Recreation Area (Rancho Seco Park) attracts a number of day visitors to the area. The average annual number of visitor days at the park for the last four years is 114,860 visitor days.
Volume I Revision 9 Rancho Seco ISFSI FSAR 2.1-2 June 2020
Additionally, a wildlife sanctuary has been built at Rancho Seco Park. It is estimated that an additional 625 cars could visit the park during special functions at this facility.
A survey of the area beyond the 5-mile radius shows that the nearest population concentration is approximately 6.5 miles from the plant site. The nearest population center of25,000 or more is Lodi, 17 miles south-southwest of the site. Other population centers of greater than 25,000 people include Sacramento at 25 miles, Stockton at 26 miles, and Modesto at 50 miles.
There are 16 special facilities in Amador and Sacramento Counties within a 1O1/2 mile radius of Rancho Seco. They consist of five public schools (one high school and four elementary schools), one private elementary school, one treatment center for TB and alcoholic patients, four residential care homes, an adult training center for developmentally disabled, a California Department of Forestry Fire Academy, the Preston School of Industry, a nudist ranch, and Mule Creek State Prison. A summary of these facilities is shown in Table 2.2-3 of Rancho Seco Defueled Safety Analysis Report (DSAR), Amendment 4 [2.2.4]1.
2.1.4 Uses ofNearby Land and Waters 2.1.4.1 Land Use The* site area is almost exclusively agricultural. DSAR, Amendment 4 Figure 2.2-4 1 provides a description of agriculture and residential activities within a 5-mile radius of the site. There are no commercial dairy farms within this 5-mile radius.
There are at present three large-scale commercial dairies in the vicinity, each with over 200 cows. The closest dairy is approximately 8 miles northwest of the site. A ranch 1 mile east of the site has dairy cows for domestic use only.
Proposed land use for the southeast section of Sacramento County as adopted by the Sacramento Planning Department is predominantly (70 percent) agricultural and is expected to remain agricultural. Approximately 2000 acres of vineyards are being developed on land in proximity to the Rancho Seco site.
1 Current license environmental information is contained in Environmental Assessment for the Proposed Renewal of the U.S. Nuclear Regulatory Commission License Number SNM-2510 for the Rancho Seco Independent Spent Fuel Storage Installation in Sacramento County, California (ADAMS, ML19241A378). References to the Defueled Safety Analysis Report (DSAR) retained for historical licensing basis context.
Volume I Revision 9 Rancho Seco ISFSI FSAR 2.1-3 April 2020
- 2. 7 References 2.1 Deleted 2.2 HMM Associates, Inc., Evacuation Time Estimate for the Rancho Seco Plume Exposure Pathway Emergency Planning Zone, December 1989.
2.3 Rancho Seco ISFSI Environmental Report, Revision 1, June 1993.
2.4 Rancho Seco Nuclear Generating Station Defueled Safety Analysis Report, Docket No. 50-312.
2.5 Regulatory Guide 1.91 "Evaluations of Explosions Postulated to Occur on Transportation Routes Near Nuclear Power Plants," Revision 1, February 1978.
2.6 Rancho Seco Nuclear Generating Station Updated Safety Analysis Report, Docket No.50-312.
- 2. 7 Rancho Seco Nuclear Generating Station Preliminary Safety Analysis Report, Docket No. 50-312 .
2.8 Environmental Geotechnical Consultants, Inc., "Geotechnical Study for Proposed Independent Spent Fuel Storage Installation Rancho* Seco Nuclear Generating Station Sacramento County California," EE-519/E306-01, June 1, 1993.
Volume I Revision 9 Rancho Seco ISFSI FSAR 2.7-1 June 2020
- 3. PRINCIPAL DESIGN CRITERIA This Section establishes the design criteria for the Rancho Seco ISFSI. These include environmental parameters which the facility must withstand, fuel clad temperature limits, DSC design criteria, etc. Design criteria unique to HSM storage are addressed separately in Volume II.
3 .1 Purpose of Installation The Rancho Seco ISFSI is designed to provide interim storage for 100% of the RSNGS spent fuel assemblies and control components. The facility must store 100% of the spent fuel assemblies and control components since the RSNGS spent fuel pool will be decommissioned as a part of the overall plant decommissioning effort.
3 .1.1 Material to be Stored RSNGS fuel is Babcock & Wilcox 15X15 Mark B PWR fuel. The fuel will be stored as non-consolidated fuel assemblies both with and without non-fuel hardware/control components. Since this is a 100% fuel storage campaign, provisions are made to store assemblies with cladding degradation in a specifically designated DSC.
The total amount of uranium to be stored at the ISFSI is approximately 220.32 metric tons of intact and damaged fuel assemblies.
The Rancho Seco ISFSI is also designed to store Rancho Seco GTCC radioactive waste.
Appendix C discusses the storage of GTCC waste.
3 .1.1.1 Physical Characteristics
'The physical characteristics of the fuel to be stored are described in detail in Section 3 .2 of DSAR, Amendment 4 [3.3.1] 1 and are summarized in Table 3-1. The characteristics of the control components are also described in detail in Section 3.2 ofDSAR, Amendment 4 1 and are summarized in Table 3-2.
3 .1.1.2 Thermal Characteristics Since Rancho Seco is in a permanently defueled configuration, the heat load for all 493 fuel assemblies has been quantified prior to ISFSI design and operation.
The ISFSI is designed to store the hottest 24 (or 13 in the case of the FF-DSC) RSNGS fuel assemblies in any single DSC assuming storage campaign initiation after June 1996. Actual heat loads should be much less since many RSNGS fuel assemblies have only a fraction of the design basis thermal power.
The maximum single assembly decay heat power, including control components, is less than 0.679 + 0.085 = 0.764 kW where 0.679 kW is the bounding decay heat from the fuel assembly only and 0.085 kW the bounding decay heat from the control component.
1 References to the I OCFR Part 50 Defueled Safety Analysis Report (DSAR) retained for historical licensing basis context.
Volume I Revision 9 Rancho Seco ISFSI FSAR 3.1-1 June 2020
Table 3-3 DSC Loading Summary for Cask Handling Conditions Design Load Type Section Design Parameter Flood 3.2.2 Maximum water height: 50 ft.
Seismic 3.2.3 Peak Ground Accelerations:
Horizontal: 0.25g (both directions)
Vertical: 0.17g Dead Loads 8.1.1.1 Weight of loaded DSC Normal and 8.1.1.2 Maximum Internal Pressure:
Off-Normal Pressure Normal Conditions: 10 psig Off-Normal Conditions: 10 psig Test Pressure 8.1.1.2 Enveloping internal pressure of 20.0 psig< 1) applied w/o DSC outer top cover plate and w/ strongback.
Normal and 8.1.1.3 DSC with spent fuel rejecting 13.5 kW (FC and Off-Normal Operating FO) or 9.93 kW (FF) of decay heat. Ambient air Temperature temperature range of 0°F to 101 °F (normal) and -20° to 1 l 7°F (off-normal).
Normal Handling 8.1.1.4 Deadweight +/- l .0g in vertical direction Loads Deadweight +/- l .0g in radial direction Deadweight +/- l .0g in axial direction Deadweight +/- 0.5g simultaneously in vertical, radial and axial directions Hydraulic ram load of 60,000 lb.
Off-Normal Handling 8.1.1.5 Hydraulic ram load of 80,000 lb.
Loads Accidental Cask Drop 8.2.1 Equivalent static decelerations:
Loads<2) Vertical end drop: 75 g Horizontal side drop: 7 5g Oblique comer drop: 25g Accident Internal 8.2.3 Enveloping internal pressure of 50 psig based on Pressure 100% fuel cladding rupture and fill gas release, 30% fission gas release, ambient air temperature of 1 l 7°F, and blocked HSM vents.
Notes:
- 1. Envelops the following pressures:
- a) 11 +3/-0 psig test pressure, applied to the shell only, during fabrication.
b) 20 psig blowdown pressure to evacuate water after fuel loading, prior to installation of the outer top cover plate.
- 2. These decelerations bound the Rancho Seco DSAR evaluations 1*
1 References to the I 0CFR Part 50 Defueled Safety Analysis Report (DSAR) retained for historical licensing basis context.
Volume I Revision 9 Rancho Seco ISFSI FSAR April 2020
4.4 Decontamination System 4.4.1 Equipment Decontamination No decontamination equipment is required at the ISFSI.
The principal decontamination activity performed in the fuel storage building is the removal of contamination from the outside surfaces of the cask, lifting yoke, and upper end of the DSC shell. Such contamination is due to immersion in the spent fuel pool. To prevent contamination of the DSC exterior surface and the cask cavity by pool water, the annulus between the DSC and cask is filled with clean demineralized water prior to insertion into the pool. The annulus is then sealed closed with an inflatable seal.
Upon withdrawal from the fuel pool, the exterior surfaces of the cask, lifting yoke, and upper end of the DSC are decontaminated prior to proceeding with transfer operations to the ISFSI. Decontamination operations are generally performed in the Fuel Storage Building.
As part of DSC closure operations, the seal is removed and the water in the cask/DSC annulus drained by means of the cask drain. The DSC exterior surface is checked for smearable contamination to a depth of about one foot below the top surface to verify that neither the exterior of the DSC nor the cask cavity has become contaminated. If no smearable contamination has penetrated to this depth, the DSC exterior is presumed to be clean throughout its length. If smearable contamination exceeds administrative limits, then the annulus is flushed with clean demineralized water until acceptable smearable contamination levels are obtained.
Decontaminating the casks after loading fuel is discussed in Section 9.6.2.2 ofDSAR, Amendment 4 [4.4.3] 1* After the cask leaves the Fuel Storage Building, there are no credible mechanisms that could result in contamination of the outside surface of the DSCs, other ISFSI components, or individuals. Therefore, the Rancho Seco ISFSI does not require provisions for decontamination.
4.4.2 Personnel Decontamination No personnel decontamination facilities are needed at the ISFSI.
Personnel decontamination will be conducted, if necessary, using existing plant equipment and procedures.
1 References to the 10CFR Part 50 Defueled Safety Analysis Report (DSAR) retained for historical licensing basis context."
Volume I Revision 9 Rancho Seco ISFSI FSAR 4.4-1 June 2020
body top lids following placement of the failed fuel assemblies into the fuel cans. All other loading and unloading operations for the FF-DSC are similar to those described in the Standardized NUHOMS SAR.
See Appendix B for Standardized SAR, Section 4. 7 (pages 4. 7-1 to 4. 7-17).
4.7.1.2 Lifting Yoke and Extension The cask typically uses the standard NUHOMS transfer system as described by the Standardized NUHOMS SAR [4.4.1]. The main difference between the certified Standardized NUHOMS SAR transfer equipment and the Rancho Seco cask transfer equipment is the design of the lifting yoke.
The lifting yoke and extensions provide the means for performing all cask handling operations within, and outside, the Fuel Storage Building. The lifting yoke and extension have a lifting capacity of 130 tons versus 100 tons for the Standardized NUHOMS SAR yoke and extension. A lifting pin connects the gantry crane hook and the lifting yoke and extension.
The codes and standards used to design and fabricate the lifting yoke are presented in Section 4.7._4 of the Standardized NUHOMS SAR [4.4.1].
See Appendix B for Standardized SAR, Section 4.7.4 (pages 4.7-10 to 4.7-11).
4.7.2 Design Bases and Safety Assurance The Standardized NUHOMS SAR [4.4.1] provides a list of the codes and standards to which the transfer system equipment is fabricated.
See Appendix B for Standardized SAR, Section 4. 7.4 (pages 4. 7-10 to 4. 7-11 ).
4.7.3 Structural Specifications The gantry crane and spent fuel handling machine are described in DSAR, Amendment 4
[4.4.3 ]1. The codes and standards for the transfer equipment are described in the StandardizedNUHOMS SAR [4.4.1] Section4.7.4 and Volume ID.
See AppendixB for Standardized SAR, Section 4.7.4 (pages 4.7-lOto 4.7-11).
4.7.4 Installation Layout The Rancho Seco fuel building is shown in DSAR, Amendment 4 [4.4.3] 1* The layout of the Rancho Seco ISFSI is discussed in Section 4.1.
1 References to the 10CFR Part 50 Defueled Safety Analysis Report (DSAR) retained for historical licensing basis context."
Volume I Revision9 Rancho Seco ISFSI FSAR 4.7-2 June2020
General layout criteria for the Rancho Seco ISFSI site (including HSM location, multi-purpose cask location, fence location, distance to site boundary, distance to personnel, work areas, etc.) which have radiological dose impact are addressed in Chapter 8.
The propane tank along the transfer route from the fuel building to the ISFSI has been removed. The caustic and acid tanks have been removed. The liquid nitrogen tank and bottles have been removed, and the hydrogen bottles have been disconnected, vented, depressurized, and abandoned.
4.7.5 Individual Unit Descriptions 4.7.5.1 Function The transport system used to move the loaded DSCs from the Fuel Storage Building to the ISFSI includes the cask (refer to Volume III), Turbine Building Gantry Crane, and transport trailer. All fuel movement is conducted onsite, thus precluding any licensing activities related to 10 CFR 71.
4.7.5.2 Components The transfer equipment components are described in Section 4.7 of the Standardized NUHOMS SAR [4.4.1]. The Turbine Building gantry crane is described in DSAR.,
Amendment 4 [4.4.3] 1* The casks will be transported from the Fuel Building to the ISFSI using a transfer trailer designed specifically for the storage system.
See Appendix B for Standardized SAR, Section 4.7 (pages 4.7-1 to 4.7-17).
4.7.6 Design Basis and Safety Assurance All fuel handling equipment is designed in accordance with the codes and standards as required by the RSNGS 10 CFR 50 license. No unique transfer operations are required for the cask loaded with a DSC. All fuel and cask handling operations will be conducted in accordance with approved procedures.
The transfer trailer and associated transfer equipment will have dimensions that will allow cask movement along the transfer path from the Fuel Building to the ISFSI. The transfer trailer is designed to handle a loaded cask.
1 References to the 10CFR Part 50 Defueled Safety Analysis Report (DSAR) retained for historical licensing basis context."
Volume I Revision 9 Rancho Seco ISFSI FSAR 4.7-3 June2020
7.2 Radiation Sources 7.2.1 Characterization of Sources The radioactive material to be stored in the Rancho Seco ISFSI consists of the RSNGS inventory ofB&W 15xl5 fuel assemblies, the associated control components, internal startup sources, and miscellaneous fuel structures. The ORIGEN2 computer code [7.7.4] was used to calculate [7.7.8] the worst case neutron and gamma-ray source terms for any assembly and control component in the Rancho Seco fuel pool, assuming that the ISFSI becomes operational after June 1996 The fuel assembly with the largest neutron source term is a 3.18 weight percent U-235 initial enrichment, 38,268 MWd/MTU burnup assembly cooled for 13 years. The fuel assembly with the largest gamma-ray source term is a 3.21 weight percent U-235 initial enrichment, 34,143 MWd/MTU burnup assembly cooled for 7 years. The control component with the largest gamma-ray source term is an axial power shaping rod assembly. These maximum neutron and gamma ray source terms were combined to form a composite design basis fuel assembly for use in all shielding calculations. The neutron and gamma-ray source strengths and spectra are given in Table 7-1 and Table 7-2, respectively.
The primary neutron source in the spent fuel assemblies is due to the spontaneous fission of Cm-244. The neutron spectrum shown in Table 7-1 is therefore the Cm-244 spontaneous fission spectrum [7.7.5].
In addition to. the radioactive material associated with the RSNGS spent nuclear fuel, a sealed source of Sr-90 will be stored in the radioactive materials storage building located within the ISFSI controlled area. The source consists of 200µCi of SrO in ceramic.
7.2.2 Airborne Radioactive Material Sources The release of airborne radioactive material is addressed for the following operations:
- 1. Fuel handling in the spent fuel pool
- 2. Drying and sealing of the DSC
- 3. DSC transfer and storage
- 4. Removing fuel from a loaded DSC Potential airborne releases from irradiated fuel assemblies in the spent fuel pool are discussed in DSAR, Amendment 4 1.
1 References to the 10CFR Part 50 Defueled Safety Analysis Report (DSAR) retained for historical licensing basis context.
Volume I Revision 9 Rancho Seco ISFSI FSAR 7.2-1 June 2020
7 .5 Health Physics Program 7.5.l Organization The Radiation Protection organization is described in Section 11.10 of DSAR, Amendment 4 1* Qualified individuals will perform radiological surveillance, radioactive waste packaging and shipping, emergency planning, and environmental monitoring. The radiation protection functional responsibilities are:
- 1. Handling, receiving, storing, and shipping radioactive materials.
- 2. Monitoring personnel exposure to radioactivity.
- 3. Maintaining personnel exposure records, reporting exposure histories, and reporting abnormal exposure results.
- 4. Developing and implementing a program to calibrate equipment used in monitoring exposure and radiological conditions.
- 5. Implementing the ALARA program.
- 6. Solving programmatic issues related to operational health physics and radiation protection programs to assure employee and public radiation exposures are maintained ALARA.
As discussed in DSAR, Amendment 4, Section 11.10 1, all individuals assigned to the Rancho Seco site and all visitors are required to follow established administrative controls for protection against radiation and contamination. Delivery personnel and other visitors (non-badged) requiring access to the radiation controlled area are escorted and provided dosimetry, as required.
The qualifications and experience of Rancho Seco personnel are considered more than sufficient for the operation of the ISFSI because these individuals have gained considerable experience at RSNGS.
7.5.2 Equipment, Instrumentation. and Facilities The radiation control equipment, instrumentation, and facilities for the Rancho Seco ISFSI will be those ofRSNGS. Qualified technicians will conduct radiation surveys with portable instruments during activities at the ISFSI.
As indicated in Section 7.2.2, respiratory protection equipment will not be needed at the ISFSI. Similarly, protective clothing will also not be needed.
A variety of instruments are used to cover the entire spectrum of radiation measurements at RSNGS. These include instruments to detect and measure alpha, beta, gamma, and neutron radiation. Calibration sources, or other appropriate methods, are available to allow for instrument calibration, response checks, maintenance, and repair.
1 Current nuclear and RP organization and responsibilities are contained in RSNAP-010 and RSNAP-305 respectively."
Volume I Revision 9 Rancho Seco ISFSI FSAR 7.5-1 June 2020
Portable radiation survey and monitoring instruments for routine use are the responsibility of the Radiation Protection Group. These instruments include:
- 1. Low and high-range beta-gamma survey meters
- 2. Neutron survey meters
- 3. Alpha survey instruments Dosimetry procedures for the ISFSI will be the same as those used for RSNGS and will comply with appropriate regulatory guidance.
7.5 .3 Procedures The methods and procedures for conducting radiation surveys at the ISFSI will be those used at RSNGS. These procedures are maintained consistent with the requirements of 10 CFR 20 and 10 CFR 72, and are adhered to for all operations involving personnel radiation exposure.
Radiation Protection procedures and any required safety evaluations are reviewed and approved in accordance with plant administrative procedures.
The philosophies, policies, and objectives of Radiation Protection procedures are based on, and implement, Federal regulations and associated Regulatory Guides to maintain doses to workers and the public ALARA..
Administrative controls for radiation protection are subject to the same review and approval as those that govern other RSNGS procedures. These procedures include Radiation Work
- Permits (RWP), control of waste shipment and disposal, and access control. The RWP is an administrative tool used in the Radiation Protection Program at RSNGS and Rancho Seco ISFSI. The RWPs issued for w9rk at the ISFSI will be used to inform workers of the radiological conditions in the area and the requirements for dosimetry and engineering controls. The RWP may be used to delineate job prerequisites, radiological safety practices, or additional requirements as needed. As an exposure tracking device, the RWP provides information necessary to ensure that exposures are kept ALARA. After the fuel has been placed at the ISFSI, the RWP and other administrative controls will be used to control access to the ISFSI.
Section 7.1 describes the radiation protection and ALARA procedures and planning that will be used for the ISFSI. Complete details are in the Rancho Seco Radiation Control Manual, ALARA Manual, and associated implementing procedures.
Access control will be accomplished by means of a fence with a locked gate surrounding the ISFSI. Control of the keys will be in accordance with appropriate administrative procedures.
Volume I Revision 9 Rancho Seco ISFSI FSAR 7.5-2 June2020
7.6 Estimated Offsite Collective Dose Assessment 7.6.1 Effluent and Environmental Monitoring Program No effluents are released from the ISFSI during operation. Effluents released during DSC loading are treated using existing RSNGS systems as described in Chapter 6. Since no effluents are released from the Rancho Seco ISFSI site, no effluent monitoring program is required. Direct radiation monitoring is discussed in Section 7.3.4.
A radiological release due to a cask drop in the Fuel Storage Building is discussed in DSAR, Amendment 4 1.
7.6.2 Analysis of Multiple Contribution As shown in Table 7-4, the predicted annual dose equivalent at both the Rancho Seco site boundary and at the nearest residence (assuming 2080 hours0.0241 days <br />0.578 hours <br />0.00344 weeks <br />7.9144e-4 months <br /> per year at the boundary and 100% occupancy at the nearest residence) is well below the 10 CFR 72. 104 and 40 CFR 190 limits of 25 mrem.
In accordance with NRC guidance, a normal operation confinement evaluation was performed assuming that fission products and actinides escape all 21 DSCs at the leak rate specified in Section 10.3.4. The calculated annual exposure, 2.3 mrem, is well below the 10CFR72.104 limit, even when added to the annual dose equivalent due to direct and air scattered radiation from Table 7-4. The doses to the thyroid and other critical organs are also below the 10CFR72.104 limits. It is therefore concluded that the radiation exposure due to the Rancho Seco ISFSI coupled with all other fuel cycle operations will not exceed the regulatory requirements of 10 CFR 72 and 40 CFR 190.
As discussed in Chapter 2, there are approximately 77 permanent residents within a two mile radius of the Rancho Seco site. The collective annual dose due to the ISFSI for this population is conservatively calculated by assuming that all of these persons are located at the closest residence to the ISFSI, 1000 meters away. The collective annual dose assuming 100% occupancy is then less than 15 person-mrem spread over 77 people. Considering the conservatisms in this calculation and the rapid attenuation of neutrons and gamma-rays with distance, the collective dose for the more distant population would be negligible.
The ISFSI restricted area fence will be approximately 350 feet from the edge of the ISFSI pad. The dose rate at this distance will be less than 0.1 mrem/hr. Assuming a conservative occupancy factor of 500 hr/yr, the annual dose to an individual member of the public would be 50 mrem/yr. This dose is below the regulatory limit of 100 mrem/yr in 10 CFR 20.1301.
7.6.3 Estimated Dose Equivalents Since no airborne effluents are postulated to emanate from the ISFSI, the direct and air-scattered radiation exposure discussed in previous chapters comprises the total radiation exposure to the public. No estimation of effluent dose equivalents is necessary.
7.6.4 Liquid Release No liquids are released from the Rancho Seco ISFSI.
1 References to the 10CFR Part 50 Defueled Safety Analysis Report (DSAR) retained for historical licensing basis context."
Volume! Revision 9 Rancho Seco ISFSI FSAR 7.6-1 June2020
8.2 Accident Analyses for the ISFSI Section 8.2 analyzes postulated ISFSI accidents to demonstrate that adequate safety margin exists and that radiological consequences are within regulatory limits. The postulated accidents addressed in this SAR section include:
- 1. Accidental cask drop
- 2. DSC leakage
- 3. Accident pressurization
- 4. Earthquake
- 5. Fire For each postulated accident, the SAR discusses the postulated cause of the event, detection of the event, analysis of the effects and consequences of the event, and appropriate corrective actions. In addition, Section 3 .2 discusses the safety criteria for natural phenomena events, including:
- 1. Tornado wind loadings and tornado generated missiles
- 2. Flood
- 3. Seismic design The results of the analyses discussed above show that adequate safety margin exists for all postulated accidents and natural phenomena events. The only event with radiological consequences is the postulated DSC leakage event. While this is a non-credible event, it provides the bounding case for radiological consequences, and demonstrates that the radiation dose from an accident or natural phenomena event does not exceed the limits in 10 CPR 72.106(b). A summary of affected components for each load type is presented in Table 8-7.
8.2.1 Accidental Cask Drop For cask-handling activities conducted under the 10 CPR 72 license, the postulated accidental cask drop event is discussed in Section 8.2.5 of the Standardized NUHOMS SAR
[8.8.5]. The results of a cask drop event for activities conducted under the IO CPR 50 license are discussed in DSAR, Amendment 4 1.
See Appendix B for Standardized SAR, Section 8.2.5 (pages 8.2-26 to 8.2-42).
1 References to the I 0CFR Part 50 license and Defueled Safety Analysis Report (DSAR) retained for historical licensing basis context.
Volume I Revision 9 I Rancho Seco ISFSI FSAR 8.2-1 June 2020
- 9. CONDUCT OF OPERATIONS This chapter describes the organization and general plans for operating the Rancho Seco ISFSI. The organization section includes a brief description of the responsibilities of managers, supervisors, and other key personnel. The training program for the plant staff is described, along with a more general discussion of replacement and retraining plans.
Standards and procedures that govern daily operations and the records developed as a result of these operations are also discussed, as are the controls used to promote safety and ensure compliance with the license and the regulations under which the facility operates.
Initially, the managerial and administrative controls for the conduct of operations at the Rancho Seco ISFSI will be built upon the existing organization under the 10 CFR 50 license1* The Superintendent, Rancho Seco Assets is currently responsible for oversight of the Rancho Seco facility and for ensuring the safe storage of the spent nuclear fuel and irradiated core components. This individual will continue to be responsible for safe storage of the fuel and will be responsible for the safe management of the Rancho Seco ISFSI.
The administrative and procedural controls applicable to the 10 CFR 50 license1 have been expanded to include the requirements of the IO CFR 72 license. Programs such as radiation protection, environmental monitoring, emergency preparedness, quality assurance, and training will be adapted as necessary to ensure the safe management of the ISFSI. SMUD has submitted and the NRC has approved the ISFSI security program which addresses the specific requirem~nts for ISFSI security..
Upon termination of the IO CFR 50 license1, those license requirements will be removed from the procedures. Appropriate 10 CFR 72.48 reviews will be conducted to ensure continued compliance with ISFSI license requirements. This process will result in stand-alone ISFSI.programs that implement the 10 CFR 72 license. SMUD will maintain the appropriate administrative and managerial controls at the Rancho Seco ISFSI until the DOE takes title to the fuel.
9.1 Organizational Structure 9.1.1 Corporate Organization SMUD's organization and its relationship to the nuclear organization is presented in the Rancho Seco Defueled Safety Analysis Report (DSAR) [9. 7.1] 1* Both Rancho Seco licensed facilities (ISFSI and Interim Onsite Storage Building) are managed under the same organization.
9.1.1.1 Corporate Functions, Responsibilities. and Authorities SMUD's Board of Directors is the policy-making body which has ultimate responsibility for the Rancho Seco ISFSI license. The Chief Executive Officer & General Manager (GM) reports directly to the Board, of Directors.
1 References to the IOCFR Part 50 license and Defueled Safety Analysis Report (DSAR) retained for historical licensing basis context. Current nuclear organization and responsibilities are contained in RSNAP-010.
Volume I Revision 9 Rancho Seco ISFSI FSAR 9.1-1 June2020
The GM, through the Chief Energy Delivery Officer, and Director, Power Generation has corporate responsibility for overall safety and management of the facility and shall take any measures needed to ensure acceptable performance of the staff in managing, maintaining, and providing technical support to the facility to ensure nuclear safety.
9.1.1.2 In-House Organization The facility organization is described in the DSAR[9. 7.1] 1*
9.1.1.3 Interrelationship with Contractors and Sup_pliers The prime contractor for design and analysis of the Rancho Seco ISFSI dry shielded canisters, horizontal storage modules, auxiliary and transfer equipment and casks is Transnuclear West, Inc. of Fremont, California. The prime contractor for the design of the Rancho Seco ISFSI civil facilities, including the storage pad, fencing and lighting system, etc. was Im.pell Corporation of San Ramon, California. Construction of the Rancho Seco ISFSI was the responsibility ofBRCO Constructors, Inc. of Loomis, California. The Rancho Seco ISFSI is owned and operated by SMUD.
9.1.1.4 Technical Staff The Corporate technical staff supporting the Rancho Seco ISFSI is described in the DSAR
[9.7.1]1. .
9.1.2 Operating Organization, Management and Administrative Control System 9.1.2.1 Onsite Organization The RSNGS organization is responsible for management of the Rancho Seco ISFSI. This organization is described in DSAR [9. 7.1] 1*
9.1.2.2 Personnel Functions, Responsibilities and Authorities The responsibilities and authority of major RSNGS positions or departments are summarized below. RSNGS personnel are selected and trained for their assigned duties, with particular emphasis on the supervisory and technical staffs to assure safe and efficient management of the Rancho Seco facilities.
Chief Energy Delivery Officer The Chief Energy Delivery Officer is responsible for the overall Rancho Seco facility and the Rancho Seco organization. This includes ensuring the safe storage of irradiated core components, ensuring effective day-to-day management, and maximizing the effectiveness of nuclear policies and procedures.
Director, Power Generation The Director, Power Generation is responsible for ensuring effective management of the licensed facilities and ensuring the safe storage of irradiated core components.
1 References to the 10CFR Part 50 license and Defueled Safety Analysis Report (DSAR) retained for historical licensing basis context. Current nuclear organization and responsibilities are contained in RSNAP-010.
Volume! Revision9 Rancho Seco ISFSI FSAR 9.1-2 June2020
Manager, Rancho Seco Assets The Manager, Rancho Seco Assets (MRSA) is the lead SMUD representative at the Rancho Seco site and is responsible for all facets of day-to-day management of the licensed facilities.
The MRSA is responsible for site security during routine, emergency, and contingency operations. The MRSA is also responsible for the implementation and maintenance of the Physical Protection Plan.
The MRSA meets all qualifications for and is the Radiation Protection Manager and implements the Radiation Protection program. The MRSA is responsible for health physics surveillance, personnel monitoring and record keeping, radwaste management, emergency preparedness and environmental monitoring.
The MRSA utilizes available SMUD and contract personnel to resolve engineering, design, and other technical issues required to support the 10 CFR 72 ISFSI licensing process in accordance with applicable regulations as well as similar issues conducted under the 10 CFR 50 license 1.
The MRSA is responsible for ensuring that management of the Rancho Seco ISFSI is
- conducted in accordance with Technical Specifications, federal and state regulations, **
Physical Protection Plan, and plant procedures and has the primary responsibility for cask and canister handling operations.
- Staff under the direction of the MRSA is engaged in a continual retraining program, as described in Section 9.3, to ensure that ISFSI operations are conducted in a safe and efficient manner.
Personnel under the direction of the MRSA as designated by site procedures check, analyze, and log system parameters, and initiate corrective actions when abnormal conditions exist.
These personnel perform initial fire response and notifications in accordance with the fire protection program.
Individuals on shift are trained and qualified to implement appropriate radiation protection procedures.
Supporting Organizations outside Generation and Grid Assets Audit & Quality Services is responsible for ensuring that the quality assurance program is implemented in accordance with regulatory requirements. The Audit & Quality Services organization has the authority to take any issue regarding the quality of program management at Rancho Seco to the General Manager and the Chief Energy Delivery Officer.
Emergency Preparedness is responsible for maintaining and administering the Emergency Plan under the direction of the Manager, Rancho Seco Assets. The Emergency Preparedness staff trains all personnel implementing the Emergency Plan as well as directing drills and other activities necessary to maintain regulatory compliance.
1 References to the 10CFR Part 50 license and Defueled Safety Analysis Report (DSAR) retained for historical licensing basis context. Current nuclear organization and responsibilities are contained in RSNAP-010.
Volume I Revision 9 Rancho Seco ISFSI FSAR 9.1-3 June 2020
9.2 Pre-Operational Testing and Operation Before the operation of the Rancho Seco ISFSI, the electrical system, communications system, and transportable storage system will be tested to ensure their proper functioning.
The electrical system will be tested to ensure that power is available for lighting, security systems, and general service receptacles. The communications system will be tested to ensure that all ISFSI telephones are properly connected into the station phone system.
To the extent practicable, functional tests of the in-plant operations, transfer operations, and HSM loading and retrieval will be performed to verify that the storage system components (e.g., DSC, cask, transfer trailer, etc.) can be operated safely and effectively. Pre-operational testing may be performed using the actual cask and canister or a training cask and canister with test weights, as appropriate. The training cask and canister were designed and fabricated to approximate the size, weight,- and behavior of an MPl 87 cask and canister.
9.2.1 Administrative Procedures for Conducting Test Program The system for preparing, reviewing, approving, and implementing testing procedures, and instructions for the Rancho Seco ISFSI will be the same as those used for RSNGS. Any changes to, or deviations from, these procedures and instructions will be reviewed and approved in accordance with Technical Specification requirements of the 10 CFR 72 license.
9.2.2 Test Program Description The objectives of the pre-operational testing program are to ensure that the storage system performs its intended safety functions and meets the operating controls and limits proposed in Chapter 10.
9.2.2.1 Physical Facilities and Operations 9.2.2.1.1 DSC and Associated Equipment An actual DSC and a full and part-length mock-up of a DSC will be obtained for pre-operational testing. A DSC will be loaded into the cask to verify fit and suitability of the DSC lift rig. Additionally, a DSC will be used in pre-operational testing of the transfer equipment and RSM.
The part-length mock-up will be used for checkout of the automated welding and cutting equipment including actual welding and removal of the top cover plates. Emphasis will be placed on acceptability of the weld, as well as compliance with approved ALARA practices.
9.2.2.1.2 Cask and Handling Equipment Functional testing will be performed with the cask and lifting yoke. These tests will ensure that the cask can be safely transported from the trailer loading area to the cask washdown area. From there, it will be placed into the spent fuel pool to verify clearances and travel path.
9.2.2.1.3 Off-Normal Testing of the DSC and Cask In the unlikely event that a problem arises during actual loading of the spent fuel assemblies (SFAs) into the DSC, seal welding of the DSC, or emplacement of a loaded DSC into an Volume I Revision 9 Rancho Seco ISFSI FSAR 9.2-1 June 2020
HSMs The HSMs may become slightly radioactive due to neutron activation. District calculation Z-XXX-N0057, Revision 1, estimates the amount of activation that may occur within the NUHOMS HSMs. Components evaluated in the calculation include the concrete, heat shield, and canister support structure. Calculation Z-XX:X-N0057 was provided as Appendix A to the Decommissioning Plan.
After DOE has removed the DSCs from the Rancho Seco ISFSI, the associated HSMs can be made available to DOE, or others, who will provide handling and transportation costs. Some support eq1,tlpment may be reusable. If the HSMs cannot be sold, they will remain at the Rancho Seco ISFSI site until any activated material has decay to below releasable levels (approximately 1-2 years after removal of the loaded DSCs). The internal metal structures will be removed and recycled. The concrete will be demolished and buried.
MP-187Cask The MP-187 cask may also become slightly activated, and may have some internal and/or extemal contamination. After DOE has accepted the fuel, the cask may be made available to DOE, who will ultimately be responsible for cask decommissioning. If DOE has no use for the cask, it may be made available to others, who will provide handling and transportation costs, etc. If the MP-187 cask cannot be sold, it will remain in storage at the Rancho Seco site until it is free releasable, and can be disposed of. ISFSI concrete pad and remaining support equipment will not be activated or contaminated. These components will be demolished and disposed of.
Based on the above, the DSCs, HSMs, and MP-187 cask can be disposed of without the need for low level waste disposal. Therefore, the only funding required for ISFSI decommissioning is that already provided for in the site restoration phase of decommissioning Rancho Seco.
S11.UIYs Board of Directors has agreed, by resolution, to fund Rancho Seco site restoration and ISFSI decommissioning, and will begin funding after decommissioning. The funding program will be in accordance with 10 CFR 72.30(c).
Volume I Revision 9 Rancho Seco ISFSI FSAR 9.6-2 June2020
- 9. 7 References 9.7.1 Rancho Seco Nuclear'Generating Station Defueled Safety Analysis Report (DSAR),
Docket No. 50-312 1*
- 9. 7.2 Rancho Seco Independent Spent Fuel Storage Installation Materials License SNM-2510 Amendment 4, November 2017 (Docket 72-11).
9.7.3 U.S. Nuclear Regulatory Commission, NUREG-1927, Revision 1, "Standard Review Plan for Renewal of Specific Licenses and Certificates of Compliance for Dry Storage of Spent Nuclear Fuel," June 2016.
9.7.4 ASJ.VlE B&PV Code,Section III, Division 1, 1992 Edition, with Addenda through 1993.
9.7.5 Nuclear Energy Institute, "Format, Content and Implementation Guidance for Dry Storage Operations-Based Aging Management," NEI 14-03 Revision 2, 2016.
9.7.6 Enclosure 4 to Letter from Dan Tallman to NRC Document Control Desk, "RESPONSE TO REQUEST FOR CLARIFICATION OF RESPONSE TO ADDITIONAL INFORMATION FOR THE TECHNICAL REVIEW OF THE APPLICATION FOR RENEWAL OF THE RANCHO SECO INDEPENDENT SPENT FUEL STORAGE INSTALLATION LICENSE NO. SNM-2510 (CAC/EPID NOS. 001028/1-2018-RNW-0005; 0009~3/L-2018-LNE-0004)," dated July 12, 2019 (ADAMS ML19204A239) 9.7.7 U.S. Nuclear Regulatory Commission, "Safety Evaluation Report for the Rancho Seco Independent Spent Fuel Storage Installation", J1.µ1e 30, 2000. (ADAMS ML003729758) 1 References to the 10CFR Part 50 license and Defueled Safety Analysis Report (DSAR) retained for historical licensing basis context.
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- 11. QUALITY ASSURANCE 11.1 . Sacramento Municipal Utility District Quality Assurance Program 10 CFR 72.140 requires that licensees establish, maintain, and execute a quality assurance (QA) program satisfying each of the applicable criteria in 10 CFR 72, Subpart G.
10 CFR 72.140(d) states that an NRC-approved QA program that satisfies the criteria of 10 CFR 50, Appendix B, and that is established, maintained, and executed with regard to an ISFSI is acceptable for satisfying the QA program requirements.
SMUD has established and implemented a QA program based on the criteria in 10 CFR 50, Appendix B for the RSNGS. This program will be implemented for the structures, systems, and*components-oftheRancho Seco ISFSI that are important to safety.
In conjunction with requests to terminate the Rancho Seco Part 50 license ("Sacramento Municipal Utility District (SMUD), Termination of the Rancho Seco Nuclear Generating Station 10 CFR Part 50 License, Number DPR-54") ADAMS Accession No. MLl 7313A481 and renew the Rancho Seco ISFSI license ("Rancho Seco Independent Spent Fuel Storage Installation - Submittal of Application for the License Renewal") ADAMS Accession No. ML18101A020, SMUD submitted "Revised Rancho Seco Quality Manual" ADAMS Accession No. MLl 7095A980. RSQM, Revision 4, reflects the termination of the 10 CFR Part 50 license and demonstrates compliance with the criteria in 10 CFR 72, Subpart G.
On August 31, 2018, the NRC terminated the Rancho Seco Nuclear Generating Station's Part 50 License ("Termination of Rancho Seco Nuclear Generating Station Operating License DPR-054") ADAMS Accession No. ML18082B076. As stated in the" Summary ofNRC Staff Review, Conclusion Re: Application for Termination of Rancho Nuclear Generating Station Operating License" (Encl. 1) "Issuance ofRSQM Revision 4 will satisfy the quality assurance requirements of 10 CFR Part 71, Subpart H, and 10 CFR Part 72, Subpart G. The NRC has reviewed proposed RSQM, Revision 4, and determined that the QA Program, as described in the RSQM, Revision 4 is in conformance with the applicable portions of Subpart H to 10 CFR Part 71 and Subpart G to 1 0 CFR Part 72 and is, therefore, acceptable" The Manager, Rancho Seco Assets is responsible for the safe and reliable operation of Rancho Seco. The Manager, Rancho Seco Assets has the responsibility and authority to implement the Rancho Seco Quality Assurance Program and ensure optimum quality performance of Rancho Seco.
The governing document for this program is the Rancho Seco Quality Manual (RSQM)
[11.1] which has been reviewed and approved by the NRC. The program is implemented through the RSQM and appropriate administrative procedures. The Rancho Seco RSQM will be applied to those activities associated with the Rancho Seco ISFSI that are important to safety.