RA-23-0110, 8 to Independent Spent Fuel Storage Installation Safety Analysis Report
| ML23145A164 | |
| Person / Time | |
|---|---|
| Site: | Robinson, 07200003  |
| Issue date: | 05/25/2023 |
| From: | Duke Energy, Duke Energy Progress |
| To: | Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML23145A158 | List: |
| References | |
| RA-23-0110 | |
| Download: ML23145A164 (1) | |
Text
INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT H. B. Robinson Steam Electric Plant Revision No. 22
HBRSEP ISFSI SAR EFFECTIVE PAGE LIST Page Revision Page Revision Title Page 22 Fig. 1.1-1 5 Fig. 1.1-2 28 EPL 1 28 Fig. 1.2-1 1 EPL 2 22 Fig. 1.3-1 5 EPL 3 27 Fig. 1.3-2 1 Fig. 1.3-3 1 Fig. 1.3-4 1 Table of Contents 22 Fig. 1.3-5 5 i 22 Chapter 2 Cover 22 ii 22 iii 22 2.1-1 22 iv 22 2.2-1 22 v 22 2.3-1 22 vi 22 2.4-1 22 vii 22 2.5-1 22 viii 22 2.6-1 22 ix 22 2.7-1 22 x 22 xi 22 Chapter 3 Cover 22 xii 22 xiii 22 3.1-1 22 xiv 22 3.1-2 22 xv 22 3.1-3 22 xvi 22 3.1-4 22 3.1-5 22 Chapter 1 Cover 22 3.2-1 22 3.2-2 22 1.1-1 22 3.2-3 22 1.1-2 22 3.3-1 22 1.1-3 22 3.3-2 22 1.2-1 26 3.3-3 22 1.2-2 22 3.3-4 22 1.2-3 22 3.3-5 22 1.2-4 22 3.3-6 22 1.2-5 22 3.4-1 22 1.3-1 22 3.5-1 22 1.3-2 22 3.R-1 22 1.3-3 22 1.3-4 22 1.3-5 22 1.4-1 22 1.5-1 22 1.R-1 22 EPL-1 Revision 28
HBRSEP ISFSI SAR EFFECTIVE PAGE LIST Page Revision Page Revision Chapter 4 Cover 22 Chapter 6 Cover 22 4.1-1 22 6.1-1 22 4.2-1 22 6.2-1 22 4.2-2 22 6.3-1 22 4.3-1 22 6.4-1 22 4.3-2 22 6.5-1 22 4.4-1 22 6.R-1 22 4.5-1 22 4.6-1 22 Chapter 7 Cover 22 4.7-1 22 4.7-2 22 7.1-1 22 4.R-1 22 7.1-2 22 7.2-1 22 Fig. 4.2-1 Sh. 1 5 7.3-1 22 Fig. 4.2-1 Sh. 2 1 7.3-2 22 Fig. 4.2-2 1 7.3-3 22 Fig. 4.5-1 20 7.3-4 22 7.3-5 22 7.4-1 22 Chapter 5 Cover 22 7.4-2 22 7.4-3 22 5.1-1 22 7.4-4 22 5.1-2 22 7.5-1 22 5.1-3 22 7.6-1 22 5.1-4 22 7.R-1 22 5.2-1 22 5.2-2 22 Fig. 7.3-1 0 5.2-3 22 Fig. 7.4-1 0 5.3-1 22 Fig. 7.4-2 0 5.4-1 22 Fig. 7.4-3 0 5.5-1 22 Fig. 7.6-1 1 5.6-1 22 5.R-1 22 Chapter 8 Cover 22 Fig. 5.1-1 0 8.0-1 22 Fig. 5.1-2 5 8.1-1 22 Fig. 5.1-3 Sh. 1 5 8.1-2 22 Fig. 5.1-3 Sh. 2 5 8.1-3 22 Fig. 5.1-3 Sh. 3 0 8.1-4 22 Fig. 5.1-3 Sh. 4 5 8.1-5 22 Fig. 5.2-1 1 8.2-1 22 Fig. 5.2-2 1 8.2-2 22 Fig. 5.4-1 1 8.2-3 22 8.2-4 22 8.2-5 22 EPL-2 Revision 22
HBRSEP ISFSI SAR EFFECTIVE PAGE LIST Page Revision Page Revision 8.2-6 22 Chapter 10 Cover 22 8.2-7 22 8.2-8 22 10.0-1 22 8.2-9 22 10.1-1 22 8.3-1 22 10.1-2 22 8.3-2 22 10.2-1 22 8.3-3 22 10.3-1 22 8.3-4 22 10.4-1 22 8.3-5 22 8.3-6 22 Chapter 11 Cover 22 8.4-1 22 8.4-2 22 11.0-1 24 8.5-1 27 11.1-1 24 8.R-1 22 11.2-1 22 11.3-1 22 Fig. 8.2-1 1 11.R-1 22 Fig. 8.2-2 1 Fig. 8.2-3 1 Fig. 8.2-4 1 Fig. 8.3-1 1 Fig. 8.3-2 1 Fig. 8.4-1 1 Chapter 9 Cover 22 9.1-1 22 9.1-2 22 9.2-1 22 9.3-1 22 9.4-1 22 9.5-1 22 9.5-2 22 9.6-1 22 9.7-1 22 9.7-2 22 9.R-1 22 EPL-3 Revision 27
H. B. ROBINSON INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT TABLE OF CONTENTS SECTION TITLE PAGE
1.0 INTRODUCTION
AND GENERAL DESCRIPTION OF INSTALLATION 1.1-1
1.1 INTRODUCTION
1.1-1 1.2 GENERAL DESCRIPTION OF INSTALLATION 1.2-1 1.2.1 GENERAL DESCRIPTION 1.2-1 1.2.2 PRINCIPAL SITE CHARACTERISTICS 1.2-1 1.2.3 PRINCIPAL DESIGN CRITERIA 1.2-1 1.2.3.1 Structural Features 1.2-1 1.2.3.2 Decay Heat Dissipation 1.2-2 1.2.4 OPERATING AND FUEL HANDLING SYSTEMS 1.2-2 1.2.5 SAFETY FEATURES 1.2-2 1.2.6 RADIOACTIVE WASTE AND AUXILIARY SYSTEMS 1.2-2 1.3 GENERAL SYSTEMS DESCRIPTIONS 1.3-1 1.3.1 SYSTEMS DESCRIPTIONS 1.3-1 1.3.1.1 Canister Design 1.3-1 1.3.1.2 Horizontal Storage Module 1.3-1 1.3.1.3 Transfer Cask 1.3-1 1.3.1.4 Transporter 1.3-1 1.3.1.5 Skid 1.3-2 1.3.1.6 Horizontal Hydraulic Ram 1.3-2 1.3.1.7 System Operation 1.3-2 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS 1.4-1 1.5 MATERIAL INCORPORATED BY REFERENCE 1.5-1 i Revision No. 22
H. B. ROBINSON INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT TABLE OF CONTENTS (Contd)
SECTION TITLE PAGE 2.0 SITE CHARACTERISTICS 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 2.1.2.1 Other Activities Within the Site Boundary 2.1-1 2.1.2.2 Boundaries for Establishing Effluent Release Limits 2.1-1 2.1.3 POPULATION DISTRIBUTION AND TRENDS 2.1-1 2.1.3.1 Population Within 10 Miles 2.1-1 2.1.3.2 Population Between 10 and 50 Miles 2.1-1 2.1.3.3 Transient Population 2.1-1 2.1.4 USES OF NEARBY LAND AND WATERS 2.1-1 2.2 NEARBY INDUSTRIAL, TRANSPORTATION, AND MILITARY FACILITIES 2.2-1 2.3 METEOROLOGY 2.3-1 2.4 SURFACE HYDROLOGY 2.4-1 2.4.1 FLOODS 2.4-1 2.4.2 POTENTIAL DAM FAILURES 2.4-1 2.4.3 PROBABLE MAXIMUM SURGE AND SEICHE FLOODING 2.4-1 2.4.4 PROBABLE MAXIMUM TSUNAMI FLOODING 2.4-1 2.4.5 ICE FLOODING 2.4-1 2.4.6 FLOODING PROTECTION REQUIREMENTS 2.4-1 2.4.7 ENVIRONMENTAL ACCEPTANCE OF EFFLUENTS 2.4-1 2.5 SUBSURFACE HYDROLOGY 2.5-1 ii Revision No. 22
H. B. ROBINSON INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT TABLE OF CONTENTS (Contd)
SECTION TITLE PAGE 2.6 GEOLOGY AND SEISMOLOGY 2.6-1 2.6.1 ISFSI FOUNDATION 2.6-1 2.6.2 SLOPE STABILITY 2.6-1 2.7
SUMMARY
OF SITE CONDITIONS AFFECTING CONSTRUCTION AND OPERATING REQUIREMENTS 2.7-1 3.0 PRINCIPAL DESIGN CRITERIA 3.1-1 3.1 PURPOSE OF THE 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 3.1-1 3.1.2 GENERAL OPERATING FUNCTIONS 3.1-1 3.1.2.1 Overall Functions of the Facility 3.1-1 3.1.2.2 Handling and Transfer Equipment 3.1-2 3.2 STRUCTURAL AND MECHANICAL SAFETY CRITERIA 3.2-1 3.2.1 TORNADO AND WIND LOADINGS 3.2-1 3.2.1.1 Applicable Design Parameters 3.2-1 3.2.1.2 Determination of Forces on the Structures 3.2-2 3.2.1.3 Ability of Structures to Perform 3.2-2 3.2.2 WATER LEVEL (FLOOD) DESIGN 3.2-2 3.2.3 SEISMIC DESIGN 3.2-2 iii Revision No. 22
H. B. ROBINSON INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT TABLE OF CONTENTS (Contd)
SECTION TITLE PAGE 3.2.3.1 Input Criteria 3.2-2 3.2.3.2 Seismic-System Analysis 3.2-2 3.2.4 SNOW AND ICE LOADS 3.2-3 3.2.5 COMBINED LOAD CRITERIA 3.2-3 3.3 SAFETY PROTECTION SYSTEM 3.3-1 3.3.1 GENERAL 3.3-1 3.3.2 PROTECTION BY MULTIPLE CONFINEMENT BARRIERS AND SYSTEMS 3.3-1 3.3.2.1 Confinement Barriers and Systems 3.3-1 3.3.2.2 Ventilation - Offgas 3.3-1 3.3.3 PROTECTION BY EQUIPMENT AND INSTRUMENTATION SELECTION 3.3-1 3.3.3.1 Equipment 3.3-1 3.3.3.2 Instrumentation 3.3-2 3.3.4 NUCLEAR CRITICALITY SAFETY 3.3-2 3.3.5 RADIOLOGICAL PROTECTION 3.3-2 3.3.5.1 Access Control 3.3-2 3.3.5.2 Shielding 3.3-2 3.3.5.3 Radiological Alarm System 3.3-2 3.3.6 FIRE AND EXPLOSIVE PROTECTION 3.3-3 3.3.6.1 Fire Protection 3.3-3 3.3.6.2 Explosive Protection 3.3-3 3.3.7 MATERIALS HANDLING AND STORAGE 3.3-3 3.3.7.1 Irradiated Fuel Handling and Storage 3.3-3 3.3.7.2 Radioactive Waste Treatment 3.3-3 iv Revision No. 22
H. B. ROBINSON INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT TABLE OF CONTENTS (Contd)
SECTION TITLE PAGE 3.3.7.3 Waste Storage Facilities 3.3-4 3.3.8 INDUSTRIAL AND CHEMICAL SAFETY 3.3-4 3.4 CLASSIFICATION OF STRUCTURES, COMPONENTS, AND SYSTEMS 3.4-1 3.5 DECOMMISSIONING CONSIDERATIONS 3.5-1 4.0 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 PRINCIPLE FEATURES 4.1-1 4.1.2.1 Site Boundary 4.1-1 4.1.2.2 Controlled Area 4.1-1 4.1.2.3 Emergency Planning Zone 4.1-1 4.1.2.4 Site Utility Supplies and Systems 4.1-1 4.1.2.5 Storage Facilities 4.1-1 4.1.2.6 Stack 4.1-1 4.2 STORAGE STRUCTURES 4.2-1 4.2.1 STRUCTURAL SPECIFICATIONS 4.2-1 4.2.1.1 Design Basis 4.2-1 4.2.1.2 Construction, Fabrication, and Inspection 4.2-1 4.2.2 INSTALLATION LAYOUT 4.2-1 4.2.2.1 Building Plans 4.2-1 4.2.2.2 Confinement Features 4.2-1 4.2.3 INDIVIDUAL UNIT DESCRIPTION 4.2-1 v Revision No. 22
H. B. ROBINSON INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT TABLE OF CONTENTS (Contd)
SECTION TITLE PAGE 4.3 AUXILIARY SYSTEMS 4.3-1 4.3.1 VENTILATION AND OFFGAS SYSTEM 4.3-1 4.3.1.1 Ventilation System 4.3-1 4.3.1.2 Offgas System 4.3-1 4.3.2 ELECTRICAL SYSTEM 4.3-1 4.3.3 AIR SUPPLY SYSTEM 4.3-1 4.3.3.1 Compressed Air 4.3-1 4.3.3.2 Breathing Air 4.3-1 4.3.4 STEAM SUPPLY AND DISTRIBUTION SYSTEM 4.3-1 4.3.5 WATER SUPPLY SYSTEM 4.3-1 4.3.5.1 Major Components and Operating Characteristics 4.3-1 4.3.5.2 Safety Consideration and Controls 4.3-2 4.3.6 SEWAGE TREATMENT SYSTEM 4.3-2 4.3.6.1 Sanitary Sewage 4.3-2 4.3.6.2 Chemical Sewage 4.3-2 4.3.7 COMMUNICATIONS AND ALARM SYSTEM 4.3-2 4.3.8 FIRE PROTECTION SYSTEM 4.3-2 4.3.9 MAINTENANCE SYSTEMS 4.3-2 4.3.10 COLD CHEMICAL SYSTEMS 4.3-2 4.3.11 AIR SAMPLING SYSTEM 4.3-2 4.4 DECONTAMINATION SYSTEMS 4.4-1 4.5 SHIPPING CASK REPAIR AND MAINTENANCE 4.5-1 4.6 CATHODIC PROTECTION 4.6-1 vi Revision No. 22
H. B. ROBINSON INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT TABLE OF CONTENTS (Contd)
SECTION TITLE PAGE 4.7 FUEL HANDLING OPERATION SYSTEM 4.7-1 4.7.1 STRUCTURAL SPECIFICATIONS 4.7-1 4.7.2 INSTALLATION LAYOUT 4.7-1 4.7.2.1 Building Plans 4.7-1 4.7.2.2 Confinement Features 4.7-1 4.7.3 INDIVIDUAL UNIT DESCRIPTION 4.7-1 4.7.3.1 Shipping Cask Preparation 4.7-1 4.7.3.2 Spent Fuel Loading 4.7-1 4.7.3.3 DSC Drying, Backfilling, and Sealing 4.7-2 5.0 OPERATION SYSTEMS 5.1-1 5.1 OPERATION DESCRIPTION 5.1-1 5.1.1 NARRATIVE DESCRIPTION 5.1-1 5.1.1.1 Preparation of the Transfer Cask and Canister 5.1-1 5.1.1.2 Fuel Loading 5.1-1 5.1.1.3 Cask Drying Process 5.1-1 5.1.1.4 DSC Sealing Operations 5.1-2 5.1.1.5 Transport of the Cask to the Horizontal Storage Module (HSM) 5.1-2 5.1.1.6 Loading of the Canister into the HSM 5.1-2 5.1.1.7 Monitoring Operations 5.1-3 5.1.1.8 Unloading the DSC from the HSM 5.1-3 5.1.2 FLOW SHEET 5.1-4 5.1.3 IDENTIFICATION OF SUBJECTS FOR SAFETY ANALYSIS 5.1-4 vii Revision No. 22
H. B. ROBINSON INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT TABLE OF CONTENTS (Contd)
SECTION TITLE PAGE 5.1.3.1 Criticality Prevention 5.1-4 5.1.3.2 Chemical Safety 5.1-4 5.1.3.3 Operation Shutdown Modes 5.1-4 5.1.3.4 Instrumentation 5.1-4 5.1.3.5 Maintenance Techniques 5.1-4 5.2 FUEL HANDLING SYSTEMS 5.2-1 5.2.1 SPENT FUEL HANDLING AND TRANSFER 5.2-1 5.2.1.1 Functional 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 SYSTEM 5.3-1 5.3.1 OPERATING SYSTEM 5.3-1 5.3.2 COMPONENTS/EQUIPMENT SPARES 5.3-1 5.4 OPERATION SUPPORT SYSTEM 5.4-1 5.5 CONTROL ROOM AND/OR CONTROL AREAS 5.5-1 5.6 ANALYTICAL SAMPLING 5.6-1 6.0 WASTE CONFINEMENT AND MANAGEMENT 6.1-1 6.1 WASTE SOURCES 6.1-1 6.2 OFFGAS TREATMENT AND VENTILATION 6.2-1 6.3 LIQUID WASTE TREATMENT AND RETENTION 6.3-1 6.4 SOLID WASTES 6.4-1 6.5 RADIOLOGICAL IMPACT OF NORMAL OPERATIONS-
SUMMARY
6.5-1 7.0 RADIATION PROTECTION 7.1-1 7.1 ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES viii Revision No. 22
H. B. ROBINSON INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT TABLE OF CONTENTS (Contd)
SECTION TITLE PAGE ARE ALARA 7.1-1 7.1.1 POLICY CONSIDERATIONS 7.1-1 7.1.2 DESIGN CONSIDERATION 7.1-1 7.1.3 OPERATIONAL CONSIDERATION 7.1-2 7.2 RADIATION PROTECTION 7.2-1 7.2.1 CHARACTERIZATION OF SOURCES 7.2-1 7.2.2 AIRBORNE SOURCES 7.2-1 7.3 RADIATION PROTECTION DESIGN FEATURES 7.3-1 7.3.1 INSTALLATION DESIGN FEATURES 7.3-1 7.3.2 SHIELDING 7.3-1 7.3.2.1 Radiation Shielding Design Features 7.3-1 7.3.2.2 Shielding Analysis 7.3-1 7.3.3 VENTILATION 7.3-3 7.3.4 RADIATION MONITORING INSTRUMENTATION 7.3-3 7.4 ESTIMATED ONSITE COLLECTIVE DOSE ASSESSMENT 7.4-1 7.4.1 OPERATIONAL DOSE ASSESSMENT 7.4-1 7.4.2 STORAGE TERM DOSE ASSESSMENT 7.4-1 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 7.5-1 7.6 ESTIMATED OFFSITE COLLECTIVE DOSE ASSESSMENT 7.6-1 7.6.1 EFFLUENT AND ENVIRONMENTAL MONITORING PROGRAM 7.6-1 7.6.2 ANALYSIS OF MULTIPLE CONTRIBUTION 7.6-1 7.6.3 ESTIMATED DOSE EQUIVALENTS 7.6-1 ix Revision No. 22
H. B. ROBINSON INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT TABLE OF CONTENTS (Contd)
SECTION TITLE PAGE 8.0 ANALYSIS OF DESIGN EVENTS 8.0-1 8.1 NORMAL AND OFF-NORMAL OPERATIONS 8.1-1 8.1.1 NORMAL OPERATION ANALYSIS 8.1-1 8.1.2 OFF-NORMAL OPERATION ANALYSIS 8.1-3 8.1.2.1 Transport 8.1-3 8.1.2.2 Air Flow Blockage 8.1-3 8.1.3 RADIOLOGICAL IMPACT FROM OFF-NORMAL OPERATIONS 8.1-3 8.2 ACCIDENT ANALYSIS 8.2-1 8.2.1 LOSS OF AIR OUTLET SHIELDING 8.2-1 8.2.2 TORNADO/TORNADO GENERATED MISSILE 8.2-1 8.2.3 EARTHQUAKE 8.2-2 8.2.3.1 Accident Analysis 8.2-2 8.2.3.2 Accident Dose Calculation 8.2-2 8.2.4 DROP ACCIDENT 8.2-3 8.2.4.1 Postulated Cause of Events 8.2-3 8.2.4.2 Drop Accident Analysis 8.2-3 8.2.5 LIGHTNING 8.2-6 8.2.5.1 Postulated Cause of Events 8.2-6 8.2.5.2 Analysis of Effects and Consequences 8.2-6 8.2.6 BLOCKAGE OF AIR INLETS AND OUTLETS 8.2-6 8.2.7 ACCIDENT PRESSURIZATION OF DSC 8.2-6 8.2.8 FIRE 8.2-7 8.2.9 DRY STORAGE CANISTER LEAKAGE 8.2-7 8.2.10 LOAD COMBINATION 8.2-7 8.3 FOUNDATION DESIGN 8.3-1 8.4 DSC INSTRUMENTATION PENETRATION DESIGN 8.4-1 x Revision No. 22
H. B. ROBINSON INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT TABLE OF CONTENTS (Contd)
SECTION TITLE PAGE 8.5 TRAIN DERAILMENT 8.5-1 9.0 CONDUCT OF OPERATIONS 9.1-1 9.1 ORGANIZATIONAL STRUCTURE 9.1-1 9.1.1 CORPORATE ORGANIZATION 9.1-1 9.1.1.1 Corporate Functions, Responsibilities, and Authorities 9.1-1 9.1.1.2 Applicants In-House Organization 9.1-1 9.1.1.3 Interrelationships with Contractors and Suppliers 9.1-1 9.1.1.4 Applicants Technical Staff 9.1-1 9.1.2 OPERATING ORGANIZATION, MANAGEMENT, AND ADMINISTRATIVE CONTROLS SYSTEM 9.1-1 9.1.2.1 Onsite Organization 9.1-1 9.1.3 PERSONNEL QUALIFICATION REQUIREMENTS 9.1-2 9.1.3.1 Minimum Qualification Requirements 9.1-2 9.1.4 LIAISON WITH OUTSIDE ORGANIZATIONS 9.1-2 9.2 PRE-OPERATIONAL TESTING AND OPERATION 9.2-1 9.2.1 ADMINISTRATIVE PROCEDURES FOR CONDUCTING TEST PROGRAM 9.2-1 9.2.2 CP&L TEST PROGRAM DESCRIPTION 9.2-1 9.2.3 TEST DISCUSSION 9.2-1 9.2.3.1 Physical Facilities Testing (Thermal Testing of HSM and DSC) 9.2-1 9.2.3.2 Operations Testing (Handling Rests) 9.2-1 9.3 TRAINING PROGRAM 9.3-1 9.3.1 PLANT STAFF TRAINING PROGRAM 9.3-1 9.3.2 REPLACEMENT AND RETRAINING PROGRAM 9.3-1 9.4 NORMAL OPERATIONS 9.4-1 9.4.1 PROCEDURES 9.4-1 9.4.2 RECORDS 9.4-1 xi Revision No. 22
H. B. ROBINSON INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT TABLE OF CONTENTS (Contd)
SECTION TITLE PAGE 9.5 EMERGENCY PLANNING 9.5-1 9.6 DECOMMISSIONING PLAN 9.6-1 9.7 LICENSE RENEWAL ACTIVITIES 9.7-1 9.7.1 AGING MANAGEMENT PROGRAMS 9.7-1 9.7.1.1 ISFSI Aging Management Program 9.7-1 9.7.1.2 Transfer Cask Aging Management Program 9.7-1 9.7.2 TIME-LIMITED AGING ANALYSIS 9.7-1 9.7.2.1 DSC Shell Cracking Due to Fatigue 9.7-1 9.7.2.2 DSC Penetration Assembly Epoxylite Seal Change in Material Properties 9.7-1 Due to Ionizing Radiation 9.7.2.3 DSC Poison Plate Depletion of Boron 9.7-1 9.7.2.4 5% Boron-Polyethylene Front Access Cover Plate Cracking and Change in 9.7-2 Material Properties Due to Ionizing Radiation 10.0 OPERATING CONTROLS AND LIMITS 10.0-1 10.1 FUEL SPECIFICATIONS 10.1-1 10.2 LIMITS FOR THE SURFACE DOSE RATE OF THE HSM WHILE THE DSC IS IN STORAGE 10.2-1 10.3 LIMITS FOR THE MAXIMUM AIR TEMPERATURE RISE AFTER STORAGE 10.3-1 10.4 SURVEILLANCE OF THE HSM AIR INLETS 10.4-1 11.0 QUALITY ASSURANCE 11.0-1 11.1 CORPORATE QUALITY ASSURANCE 11.1-1 11.2 H. B. ROBINSON QUALITY ASSURANCE PROGRAM 11.2-1 11.3 NUTECH QUALITY ASSURANCE 11.3-1 xii Revision No. 22
H. B. ROBINSON INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT LIST OF TABLES TABLE TITLE PAGE 1.1-1 ACRONYMS 1.1-2 1.1-2 ABBREVIATIONS 1.1-3 1.2-1 DESIGN PARAMETERS FOR THE HBR ISFSI 1.2-3 1.2-2
SUMMARY
OF ISFSI FUEL HANDLING OPERATIONS 1.2-4 1.2-3 PRIMARY DESIGN PARAMETERS FOR THE ISFSI OPERATING SYSTEMS 1.2-5 1.3-1 MAJOR SYSTEMS, SUBSYSTEMS, AND COMPONENTS OF THE H. B. ROBINSON ISFSI 1.3-5 3.1-1 PHYSICAL CHARACTERISTICS OF PWR FUEL ASSEMBLIES BASED ON NOMINAL DESIGN 3.1-4 3.1-2 ACCEPTABLE RADIOLOGICAL CRITERIA FOR STORAGE OF MATERIAL IN THE HBR ISFSI 3.1-5 3.3-1 H. B. ROBINSON ISFSI IMPORTANT TO SAFETY (SAFETY RELATED) FEATURES 3.3-5 3.3-2 RADIOACTIVITY CONFINEMENT BARRIERS AND SYSTEM OF THE ISFSI 3.3-6 5.2-1 TRANSFER SYSTEM COMPONENT DESCRIPTION 5.2-3 7.3-1 DSC END SHIELDING MATERIAL THICKNESSES 7.3-4 7.3-2 SHIELDING ANALYSIS RESULTS 7.3-5 7.4-1
SUMMARY
OF ESTIMATED ONSITE DOSES DURING FUEL HANDLING OPERATIONS 7.4-2 7.4-2 ESTIMATED ANNUAL ONSITE DOSES DURING STORAGE PHASE 7.4-4 8.1-1 DRY SHIELDED CANISTER AND HORIZONTAL STORAGE MODULE COMPONENT WEIGHTS 8.1-4 8.1-2 MAXIMUM DRY STORAGE CANISTER SHELL STRESSES FOR NORMAL OPERATING LOADS 8.1-5 8.2-1 MAXIMUM DSC STRESSES FOR 8-FOOT BOTTOM END DROP ACCIDENT 8.2-8 xiii Revision No. 22
H. B. ROBINSON INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT LIST OF TABLES (Contd)
TABLE TITLE PAGE 8.2-2 DSC ENVELOPING LOAD COMBINATION 8.2-9 8.3-1 FOUNDATION BEARING STRESS 8.3-3 8.3-2 FOUNDATION SLAB-MAXIMUM BENDING MOMENTS 8.3-4 8.3-3 FOUNDATION ANCHOR LOADS 8.3-6 10.1-1 ACCEPTABLE RADIOLOGICAL CRITERIA FOR 10.1-2 STORAGE OF MATERIAL IN THE HBR ISFSI xiv Revision No. 22
H. B. ROBINSON INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT LIST OF FIGURES FIGURE TITLE 1.1-1 Primary Components of the ISFSI 1.1-2 Plot Plan 1.2-1 Horizontal Storage Module 1.3-1 Dry Shielded Canister and Internal Basket 1.3-2 HSM Air Flow Diagram 1.3-3 Skid Features 1.3-4 Hydraulic Ram 1.3-5 Primary Canister Handling Operations 4.2-1 Sh. 1 Dry Shielded Canister 4.2-1 Sh. 2 Dry Shielded Canister 4.2-2 Horizontal Storage Module 4.5-1 General Arrangement - HBR2 Fuel Handling Building 5.1-1 Cask Extension for GE IF-300 Cask 5.1-2 Cask Liner for GE IF-300 Cask 5.1-3 Sh. 1 Handling Operations Flow Sheet 5.1-3 Sh. 2 Handling Operations Flow Sheet 5.1-3 Sh. 3 Handling Operations Flow Sheet 5.1-3 Sh. 4 Handling Operations Flow Sheet 5.2-1 Hydraulic Ram System 5.2-2 DSC Grappling System 5.4-1 DSC Instrument Locations 7.3-1 Location of Reported Dose Rates (Table 7.3-2) 7.4-1 Annual Dose (mrem/yr) from 3 HSMs (Assuming 2080 hour0.0241 days <br />0.578 hours <br />0.00344 weeks <br />7.9144e-4 months <br />s/yr) 7.4-2 Dose Rate vs. Distance from Surface of HSM (Assuming 3 Modules) 7.4-3 Radiation Zone Map of Module Surface Dose Rates xv Revision No. 22
H. B. ROBINSON INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT LIST OF FIGURES FIGURE TITLE 7.6-1 Annual Offsite Dose (mrem per year) from 3 HSMs (Based on 24 hrs/day, 365 days/yr) 8.2-1 Cask Drop Height Criteria 8.2-2 Cask Deceleration vs. Time, 8-foot Drop 8.2-3 DSC Bottom Region ANSYS Model 8.2-4 DSC Top Region ANSYS Model 8.3-1 Mat Foundation STARDYNE Model 8.3-2 Foundation Uplift Model 8.4-1 Penetration Model for Horizontal Drop Analysis xvi Revision No. 22
HBRSEP ISFSI SAR CHAPTER 1 INTRODUCTION AND GENERAL DESCRIPTION OF INSTALLATION Revision No. 22
1.0 INTRODUCTION
AND GENERAL DESCRIPTION OF INSTALLATION
1.1 INTRODUCTION
Carolina Power & Light Company (CP&L) entered into an agreement with the U. S. Department of Energy (DOE) to conduct a licensed at-reactor dry storage demonstration program for spent nuclear fuel to be located at the H. B. Robinson Steam Electric Plant Unit No. 2 (HBR2). This document provides the safety analysis report (SAR) required as part of the license application under 10 CFR Part 72, "Licensing Requirements for the Storage of Spent Fuel in an Independent Spent Fuel Storage Installation (ISFSI)." This SAR is organized in accordance with the guidelines contained in Regulatory Guide 3.48. If information intended to be included in this SAR is redundant to information already included in the Updated Final Safety Analysis Report (UFSAR) for the HBR2 operating license, then such information will be included by reference to the UFSAR, Reference 1.4. Table 1.1-1 lists the acronyms and Table 1.1-2 lists the abbreviations used throughout this document.
The Nuclear Waste Policy Act of 1982 (NWPA) established 1998 as an operational date for a spent fuel/high level radioactive waste repository. To assist utilities in providing for spent fuel storage until a repository is operational, the Nuclear Waste Policy Act requires DOE to ". . . establish a demonstration program in cooperation with the private sector, for the dry storage of spent nuclear fuel at civilian nuclear power reactor sites, with the objective of establishing one or more technologies that the Commission [NRC] may by rule approve for use at the sites of civilian nuclear power reactors without, to the maximum extent practicable, the need for additional site-specific approvals by the Commission." Accordingly, on May 9, 1983, the DOE issued its Solicitation for Cooperation Agreement Proposal
(#DE-SC06-83RL10432) for a Licensed, At-Reactor, Dry Storage Demonstration Program. Carolina Power & Light Company, in response to the DOE solicitation, submitted a proposal on August 23, 1983 to DOE to demonstrate the NUTECH Engineers, Inc. (NUTECH) Horizontal Modular Storage (NUHOMS) system at the site of the H. B.
Robinson Steam Electric Plant Unit No. 2. DOE accepted CP&L's proposal in October 1983, and the contract was signed in March 1984 (Reference 1.1). The Electric Power Research Institute (EPRI) also entered into the program as a participant. EPRI, in particular, was involved in the research and development aspects of the program.
The NUHOMS system is the dry storage design used for the HBR2 ISFSI, designated as the 7P-ISFSI. In addition to this SAR, Revision 1 of the generic Topical Report for the NUHOMS system, submitted by NUTECH in November 1985 (Reference 1.2), provided the details of the system to be utilized at HBR2. Revision 2 of the NUHOMS Topical Report was issued in March 1990 (Reference 1.6). Figure 1.1-1 shows the primary components of the HBR 7P-ISFSI. The location of the 7P-ISFSI on the Robinson site is shown on Figure 1.1-2.
The NUHOMS system provides long-term interim storage for irradiated fuel assemblies. The fuel assemblies are confined in a helium atmosphere by a stainless steel canister. The canister is protected and shielded by a massive concrete module. Decay heat is removed by thermal radiation, conduction and convection from the canister to an air plenum inside the concrete module. Air flows through this internal plenum by natural draft convection. The canister containing seven irradiated fuel assemblies is transferred from the reactor fuel pool to the concrete module in a transfer cask. The cask is precisely aligned and the canister is inserted into the module by means of a hydraulic ram.
The NUHOMS system is a totally passive installation that is designed by analysis to provide shielding and safe confinement of irradiated fuel. The dry shielded canister and horizontal storage module have been designed to function during and withstand certain accidents as described in this SAR.
The fuel assemblies stored in the ISFSI were previously located in the HBR2 spent fuel pool and were irradiated in the HBR2 reactor. Seven fuel assemblies are stored in each dry shielded canister. One dry shielded canister is stored in each concrete module. A total of eight modules have been constructed and filled with a total of 56 fuel assemblies.
In addition to the 7P-ISFSI, a separate dry fuel storage facility, designated as the 24P-ISFSI, was constructed on-site and placed into operation under the general license provisions of 10 CFR 72. The 24P-ISFSI utilizes the NUHOMS-24PTH design, a higher capacity, higher heat load version of the 7P-ISFSI. The 24P-ISFSI was designed to support a total of 66 horizontal storage modules and is located northwest of the HBR2 containment as shown in Figure 1.1-2.
1.1-1 Revision No. 22
HBRSEP ISFSI SAR TABLE 1.1-1 ACRONYMS ACI American Concrete Institute ALARA as low as reasonably achievable ANSI American National Standards Institute ASTM American Society of Testing and Materials CP&L Carolina Power & Light Company DBT Design Basis Tornado DOE U. S. Department of Energy DSC dry shielded canister EOC Emergency Operations Center EPRI Electric Power Research Institute FSAR Final Safety Analysis Report GE General Electric Company HBR H. B. Robinson Steam Electric Plant HBR2 H. B. Robinson Steam Electric Plant Unit No. 2 HSM horizontal storage module IFA irradiated fuel assembly ISFSI independent spent fuel storage installation NED Nuclear Engineering Department NFS Nuclear Fuel Section NRC Nuclear Regulatory Commission NUHOMS NUTECH Horizontal Modular Storage NUTECH NUTECH Engineers, Inc.
NWPA Nuclear Waste Policy Act of 1982 PWR pressurized water reactor QA quality assurance QC quality control RG Regulatory Guide SAR Safety Analysis Report SC South Carolina SSE safe shutdown earthquake UFSAR Updated Final Safety Analysis Report for HBR2 Operating License - Reference 1.4 10 CFR Code of Federal Regulations, Title 10 1.1-2 Revision No. 22
HBRSEP ISFSI SAR TABLE 1.1-2 ABBREVIATIONS atm atmosphere bar bar cm centimeter
° C degrees Centigrade
° F degrees Fahrenheit fps feet per second ft/s feet per second ft foot ft-lb foot pounds He helium kg kilogram kw kilowatt k-in kip inch ksi kips per square inch Kr-85 Krypton 85 MWd/MT megawatt days per metric ton MWe megawatts electric MWt megawatts thermal Hg Mercury m meter Ci/cm2 microcuries per square centimeter mph miles per hour mm millimeter mrem/hr millirem per hour mR/hr milliroentgen per hour k neutron multiplication factor, effective eff N Newton Pu-239 Plutonium-239 Pu-241 Plutonium-241 lb pound lbf pounds-force psf pounds per square foot psi pounds per square inch psia pounds per square inch, atmospheric psig pounds per square inch, gauge sec second sq. mi. square mile kips thousand pounds U-235 Uranium 235 1.1-3 Revision No. 22
HBRSEP ISFSI SAR 1.2 GENERAL DESCRIPTION OF INSTALLATION 1.2.1 GENERAL DESCRIPTION The ISFSI provides for the horizontal, dry storage of irradiated fuel assemblies (IFAs) in a concrete module. The principal components are a concrete horizontal storage module (HSM) and a steel dry shielded canister (DSC) with an internal basket which holds the IFAs.
Each HSM contains one DSC and each DSC contains seven fuel assemblies. The modules are constructed on a common foundation and are interconnected. The outer, exposed walls are 3 1/2 feet thick concrete to provide the necessary shielding. The initial phase of construction included three modules. A total of eight modules have been built and operated at the Robinson site. The second foundation for the additional five modules is constructed nearby, but separate from the initial three. It was determined that construction of the additional five modules would not have any impact on continued operation of the initial three modules. Figure 1.2-1 shows the configuration of the HBR ISFSI.
In addition to these primary components, the HBR ISFSI also requires transfer equipment to move the DSCs from the irradiated fuel pool (where they are loaded with the IFAs) to the HSMs where they are stored. This transfer system consists of a transfer cask, a hydraulic ram, a tow vehicle, a trailer and a cask skid. This transfer system interfaces with the existing HBR2 irradiated fuel pool, the cask crane, the site layout (i.e., roads and topography) and other procedural requirements.
1.2.2 PRINCIPAL SITE CHARACTERISTICS The ISFSI is located on the H. B. Robinson Steam Electric Plant site near Hartsville, South Carolina. Duke Energy Progress owns and operates a 2339 MWt nuclear generating unit (Unit 2) on the Robinson site. The ISFSI is located within the Unit 2 protected area approximately 600 ft. west of the HBR2 containment building.
1.2.3 PRINCIPAL DESIGN CRITERIA The principal design criteria and parameters for the HBR ISFSI are shown in Table 1.2-1. The radiation sources are for the maximum burnup fuel. For the fuel to be stored, the radiation sources shall be less than or equal to the sources described in Table 1.2-1.
1.2.3.1 Structural Features The HSM is a low profile reinforced concrete structure designed to withstand normal operating loads, the abnormal loads created by seismic activity, tornados and other natural events and the postulated accidental loads which may occur during operation.
The structural features of the DSC are defined, to a large extent, by the cask drop accident. The operational procedures for the transfer of the cask from fuel pool to the module site are such that the maximum height at which a credible cask drop can occur is limited to 2.44 m (8 ft). The detail description of the cask handling procedure and the transfer operation is presented in Sections 1.3.1.7 and 8.2.4 of this report. The canister body, the double containment welds on each end, and the DSC internals are designed to provide their intended safety functions after a 2.44 m (8 ft) drop. In fact, the original design of the DSC and its internals, as presented in the NUHOMS topical report, has been modified to withstand decelerations associated with drop heights significantly higher than 2.44 m (8 ft) limit. However, the 2.44 m (8 ft) drop is the minimum requirement used as a design criteria, since it envelopes any actual drop accident that could occur at the Robinson site. The details of the cask drop accident are contained in Section 8.2.4 of this report.
1.2-1 Revision No. 26
HBRSEP ISFSI SAR 1.2.3.2 Decay Heat Dissipation The decay heat of the IFAs is removed from the DSC by natural draft convection. Air enters the lower part of the HSM, rises around the DSC and exits through the top shielding slab. The flow cross-sectional area is designed to provide adequate air flow from the draft height of the HSM and the inlet and outlet air temperature differences for the hottest day conditions (i.e., 51.7oC (125oF) inlet and 98.9oC (210oF) outlet).
1.2.4 OPERATING AND FUEL HANDLING SYSTEMS The major operating systems of the ISFSI are those required for fuel handling and transport of the fuel from the spent fuel pool to the ISFSI and back, if necessary. General operations are outlined in Table 1.2-2 and the primary design parameters of the required systems are listed in Table 1.2-3. The fuel handling operations involving the cask (i.e., fuel loading, drying, trailer loading, etc.) and the remaining operations (cask-HSM alignment and DSC transfer) are unique to the ISFSI. Procedures for these activities were developed using HBR fuel shipment procedures and experience gained during testing of the ISFSI.
1.2.5 SAFETY FEATURES The principal safety feature of the ISFSI is the containment provided by the DSC and the concrete shielding of the HSM. This shielding reduces the gamma and neutron flux emanating from the IFAs inside a DSC so that the average outside surface dose rate on the HSM is less than 20 mrem/hr. Additional ISFSI safety features include:
a) Filling the DSC and cask annulus with demineralized water and providing a seal prior to lowering them into the spent fuel pool - Minimizes contamination of the DSC exterior by pool water.
b) Internal shield blocks inside the HSM - Reduces scatter dose out of the air inlet.
c) External shield blocks on the HSM - Reduces scatter dose out of the air outlet.
d) Shield plugs on the DSC - Reduces dose during DSC drying, helium filling and seal welding.
e) Double containment closure welds on each end of the DSC - Prevent leakage of radioactive gases or particulates if the fuel rods should fail.
1.2.6 RADIOACTIVE WASTE AND AUXILIARY SYSTEMS Because of the passive nature of the ISFSI, there are no radioactive waste or auxiliary systems required during normal storage operations. There are, however, some waste and auxiliary systems required during DSC loading, drying and transfer into the module. The HBR2 waste systems handle the fuel pool water and air and inert gas which are vented from the DSC and cask during drying. Auxiliary handling systems (such as hydraulic pressure control, alignment, crane, etc.) are also required during the loading and transfer operation.
1.2-2 Revision No. 22
HBRSEP ISFSI SAR TABLE 1.2-1 DESIGN PARAMETERS FOR THE HBR ISFSI Category Criterion or Parameter Value Fuel Acceptance Fissile Content 3.5% Fissile (U-235 Equivalent)
Criteria Radiation Source Gamma 5.73 x 1015 photons/sec/assembly(1)
Neutron 1.67 x 108 neutron/sec/assembly Heat Load 1 KW/Assembly Dry Shielded Capacity per canister 7 PWR Fuel Assemblies Canister Size Length (typical) 4.56m (179.5 in)
Diameter 0.94m (36.9 in)
Temperature (max. fuel rod clad) 380°C (716°F)
Cooling Natural Convection Design Life 60 Years(2)
Material 304 Stainless Steel with Lead End-Shields Internal Helium Pressure 0.0 psig +/- 0.5 psig Horizontal Capacity 1 Dry Shielded Canister per Module Storage Module Unit Size 3 modules per Unit Length 6.71m (22.00 ft)
Height 3.81m (12.50 ft)
Width 7.54m (24.75 ft)
Unit Size 5 modules per Unit Length 6.71m (22.00 ft)
Height 3.81m (12.50 ft)
Width 11.86m (38.92 ft)
Surface Radiation Dose 20 mrem/hr Rate (average on contact)
Material Reinforced Concrete Design Life 60 years(2)
(1) Actual design limits are for seven assemblies in the DSC with source rates of 1.17 x 109 neutrons/sec/DSC and 4.01 x 1016 photons/sec/DSC.
(2) Expected life is much longer (hundreds of years); however, initial license application was for 20 years only. Renewed license is for 60 years.
1.2-3 Revision No. 22
SUMMARY
OF ISFSI FUEL HANDLING OPERATIONS
- 1. Clean the DSC and load it into the transfer cask.
- 4. Lift the cask containing the DSC into the spent fuel pool.
- 5. Remove the top lead plug.
- 6. Load the fuel into the DSC.
- 8. Install the cask collar lid.
- 9. Lift the cask containing the filled DSC out of the spent fuel pool and move it to the cask decontamination facility.
- 10. Drain water from the cask and DSC to a level approximately two inches below the top surface of the lead plug (approximately 15 gallons).
- 12. Perform liquid penetrant examination of top lead plug seal weld.
- 13. Drain the water from the cask annulus.
- 14. Drain, evacuate, and dry the DSC interior.
- 15. Backfill the DSC with helium.
- 16. Perform helium leak test.
- 17. Seal weld plugs in the drain and vent lines of the DSC.
- 18. Perform liquid penetrant examination of drain and vent line plug welds.
- 19. Perform helium leak test of drain and vent line plug welds.
- 20. Place and seal weld the top cover plate.
- 21. Perform liquid penetrant test on top cover plate weld.
- 22. Install the cask collar lid.
- 23. Lift the cask onto the trailer and lower it into the horizontal position.
- 24. Tow the trailer to the HSM.
- 25. Remove the HSM front access cover.
- 26. Remove the cask collar lid.
- 27. Align the cask and the HSM.
- 28. Insert the hydraulic ram.
- 29. Pull the DSC into the HSM.
- 30. Move the cask and the trailer away from the front of the HSM.
- 31. Replace and tack weld the HSM front access cover.
- 32. Disassemble the hydraulic ram from the rear of the HSM.
- 33. Install the seismic retainer and rear cover plate.
1.2-4 Revision No. 22
HBRSEP ISFSI SAR TABLE 1.2-3 PRIMARY DESIGN PARAMETERS FOR THE ISFSI OPERATING SYSTEMS System Parameters Value Cask Cavity Diameter 0.953m (37.5 in.)
Cavity Length 4.572m (180 in.)
Payload Capacity 9524 kg (21,000 lb)
Heat Rating > 7kw Shielding (Surface Dose) 200 mrem/hr Cask Movement Liftable by Crane N/A Rotatable by Crane from N/A Vertical to Horizontal Cask Lid Removable in Horizontal Position N/A Trailer and Truck Transportable N/A Skid Cask Lid Must Protrude Past 15.25cm (6 in.)
End of Trailer and Skid Capacity (Trailer) 73,000kg (80 tons)
(Skid) 100,000kg (110 tons)
Positioning Capability 6 in. Vertically 5 in. Towards Module 3 in. Parallel to Module 1.2-5 Revision No. 22
HBRSEP ISFSI SAR 1.3 GENERAL SYSTEMS DESCRIPTIONS The major systems, subsystems, and components of the HBR ISFSI are shown in Table 1.3-1. The following subsections briefly describe the principal systems and components and their operation.
1.3.1 SYSTEMS DESCRIPTIONS 1.3.1.1 Canister Design Figure 1.3-1 shows the dry shielded canister. The DSC is sized to hold seven irradiated pressurized water reactor (PWR) fuel assemblies. The main component of construction is a stainless steel cylinder with a 0.5 inch wall and 0.94m (36.9 in.) outside diameter. The overall length is 179.5 in. The general system description of the canister design is available in the NUTECH Topical Report.
1.3.1.2 Horizontal Storage Module An isometric view of a unit of three HSMs is shown in Figure 1.2-1. The HSM provides a unitized modular storage location for irradiated fuel. The HSM is constructed from reinforced concrete and structural steel. The modules were constructed in place at the storage location. The thick concrete walls and roof of the HSM provide adequate neutron and gamma shielding. The general systems description of the HSM is provided in the NUTECH Topical Report.
The HSMs are placed in service on a load bearing foundation. Certain civil work was required to prepare the storage site for a level foundation and access area. This work included the relocation of any existing underground utilities, excavation, backfill, compaction and leveling. Also, a 4 inch mud slab was placed on the leveled subgrade to provide smooth working surface for the placement of the foundation.
1.3.1.3 Transfer Cask The transfer cask used with the ISFSI provides shielding during the DSC drying operation and during the transfer to and from the HSM. For the HBR ISFSI, the IF-300 cask (which CP&L owns) licensed under 10 CFR 71 as a transportation cask was used (Reference 1.3). In order to meet the cask cavity minimum length requirement and the criteria for cask collar lid removal in the horizontal position, the IF-300 cask requires an addition. The addition includes an extension collar (12.5 inches long and approximately 6 inches thick) with the inside diameter the same as that of the cask, and a 1 inch thick cask collar lid. In this modified configuration the energy absorbing properties of the cask is significantly reduced. However, as described in detail in Section 8.2.4 of this report there is no credible condition during the cask handling and transfer operation in which the cask could be dropped on its head. Hence, the removal of the cask head with its radial impact limiting fins does not affect the safety features of the ISFSI transfer operation.
1.3.1.4 Transporter The transporter consists of a trailer with a capacity of 80 tons. The trailer carries the cask skid and the loaded transfer cask. The trailer is designed to ride as low to the ground as possible to minimize the HSM height. Four hydraulic jacks are placed under the trailer to provide vertical movement for alignment of the cask and HSM. The trailer is pulled by a conventional tractor.
1.3-1 Revision No. 22
HBRSEP ISFSI SAR 1.3.1.5 Skid The cask positioning skid is a welded wide-flange frame assembly, which houses the cask cradle support and the cask saddle assemblies. These components are welded to the skid frame. The cask cradle and its support provide the rotational capabilities required to orient the cask from vertical to horizontal position. Once in the horizontal position the cask will be seated on the saddle and will be prevented from further movement. The skid is seated on 75 ton (minimum) capacity guided Hilman Rollers at each corner. There are four hydraulic positioner cylinders mounted on the skid frame and the trailer bed. The rollers and the cylinders will be used for the final alignment of the cask and the HSM. The entire skid assembly is seated on the trailer bed. During towing of the trailer, the skid is secured to the trailer bed by means of tie down brackets.
The skid and its various components are designed to withstand the inertia forces associated with transportation shocks. The features of the skid described above are shown in Figure 1.3-3.
1.3.1.6 Horizontal Hydraulic Ram The horizontal hydraulic ram is a telescopic, hydraulic boom with a minimum capacity of 22,000 lbf and a reach of 25 ft. The ram will be mounted on the concrete foundation and wall of the HSM on the opposite side from the loading position. Figure 1.3-4 shows the hydraulic ram.
1.3.1.7 System Operation The primary operations (in sequence of occurrence) for the HBR system are shown schematically in Figure 1.3-5 and are described below:
a) Cask Preparation - Cask preparation includes exterior washdown and interior decontamination. These operations are done in the HBR2 cask decontamination facility outside the spent fuel pool area. The operations are standard cask operations and have been previously performed by CP&L personnel. Detailed procedures for these operations are described in Chapter 5.
b) Canister Preparation - The canister exterior is wiped down with demineralized water. The interior is cleaned of any debris as required. This ensures that the newly fabricated canister will meet existing HBR2-specific criteria for placement in the spent fuel pool.
c) Cask-Canister Loading - The empty canister is inserted into the cask. Proper alignment is assured by visual inspection.
d) Cask Lifting and Placement in the Pool - The cask annulus and DSC inside the cask are filled with demineralized water. A seal is then installed. This prevents an inrush of pool water when the cask with the DSC is placed in the spent fuel pool. This will also reduce (if not prevent) contamination of the DSC outer surface by the pool water. The cask with the DSC inside is then lifted into the fuel pool.
e) Canister-Assembly Loading - Seven irradiated fuel assemblies are placed into the canister basket. This operation is performed using approved ISFSI procedures.
f) Cask Collar Lid Placement - The cask collar lid placement operation consists of placing the DSC upper end-shield plug inside the DSC using the overhead crane. The cask collar lid is then placed on the cask using temporary guide pins and it is raised to the surface where the cask collar lid bolts are installed. The cask collar lid firmly holds the DSC in place while in the cask.
1.3-2 Revision No. 22
HBRSEP ISFSI SAR g) Cask Lifting out of the Pool - The filled and closed cask is lifted out of the spent fuel pool and placed (in the vertical position) on the drying pad inside the decontamination area. This operation is performed using ISFSI procedures. During this operation the overhead crane is equipped with a redundant yoke and as such is operating in a single failure proof mode. The use of the redundant yoke eliminates the possibility of any drop accident at this stage of operation.
h) Canister Sealing - The seal is removed from the cask/DSC annulus. Using the cask drain valve, a sufficient amount of water is drained from the annulus so as to enable a contamination survey to be performed approximately 12 inches below the top of the DSC shell. Should the contamination values from the annulus survey be higher than acceptable, the annulus may be flushed using demineralized water to reduce the contamination levels in this area.
Using a pump, the water level in the canister is reduced to approximately 2 inches below the bottom surface of the top lead shield plug. A seal weld is applied to the interface of the DSC shell and the lead shield plug. This provides the primary seal for the DSC. After completion of this weld, the remaining water is removed from the DSC interior in preparation of backfilling with helium to perform a helium leak test of the weld.
i) Cask-Canister Drying - Initial removal of water from the DSC is accomplished by connecting a hose from the instrument air supply to the DSC vent connection. A second hose is then connected to the DSC siphon connection and routed to the drain manifold at CD-7. The instrument air valve is opened and the air flow recalculated. When the DSC is drained, the instrument air valve is closed. Following evacuation of the water from the DSC using compressed air, the vacuum drying system is connected to the DSC and activated to facilitate DSC drying until the water content meets the design criteria.
j) Helium Filling - In order to ensure that no fuel and/or cladding oxidation occurs during storage, the canister is filled with helium (He). To accomplish this, a portable He gas bottle is connected to the piping/tubing of the vacuum drying system.
The canister is filled with He gas (by activating valves) to a minimum pressure of 25 psig and a helium leak test is performed. The pressure is then reduced to 0.0 psig. After the canister is filled with the inert gas, the vent and drain lines are removed and the DSC vent and drain line connectors are plugged and welded closed. These welds are then checked for inert gas leakage. When the steel cover plate is then welded in place, the integrity of the penetrations is assured.
k) Final Canister Sealing - After the inert gas filling, the steel cover plate is positioned and seal welded. This provides a redundant seal at the upper end of the DSC. The lower end also has redundant seal welds, which were made and tested during fabrication. This operation provides the double seal integrity of the DSC.
l) Transporter-Skid Loading - The cask collar lid is positioned and bolted in place to close the cask for transport to the HSM. The cask is then lifted from the decontamination area by the overhead crane with its redundant yoke attached, and is placed on a concrete pad adjacent to the skid/trailer. Once the cask is on the concrete pad the lower half of the redundant yoke is removed to allow the cask to fit into the cask cradle. The maximum height required to raise the cask above the cradle is 2.44 m (8 ft). The cask is then raised just above the cradle atop the skid assembly which is held in the horizontal position. At this juncture the crane is operating without the redundant yoke. The cask is then lowered into the cradle until it is firmly seated. Next, the cask is tilted from the vertical to the horizontal position until the top region of the cask is firmly positioned on the saddle atop the skid. The lifting and tilting procedures used are identical to those used for loading the IF-300 shipping cask onto its rail car skid (Reference 1.5).
The crane yoke is then removed and the cask is secured to the skid frame. At no time during the tilting operation is the bottom of the cask more than 2.44 m (8 ft) off the ground. Furthermore, since the cask is always lifted from the cask trunnions located on the upper region of the cask, the failure of the yoke will not cause the cask to drop on its head; hence, no possibility of a cask top end drop. If the yoke fails during tilting operation, the cask would land on its steel ring fins located near the top of the cask's outer shell. The cask drop accident analysis and further discussion on the postulated drop heights and orientations are presented in Section 8.2.4.
1.3-3 Revision No. 22
HBRSEP ISFSI SAR m) Transfer - Once loaded and secured, the transfer trailer is towed to the HSM. This movement is completely within the HBR2 plant site and protected area. The skid assembly is designed such that none of the cask redundant tie downs and support mechanism can fail due to the inertia forces associated with the transportation shocks and vibrations. Additionally, the skid is secured to the trailer bed by means of tie down brackets, designed to withstand the same forces. The possibility of cask dropping from the skid, or the cask/skid/trailer tipping or rolling over is extremely remote. The impact decelerations generated by such unlikely events are enveloped by the 2.44 m (8 ft) horizontal and vertical drop criteria due to the fact that the center of gravity of the cask is less than 2.44 m (8 ft) from the ground level during the transport operation.
n) Cask-Module Preparation - At the HSM storage area, the transfer trailer is backed into position and the HSM front access cover is raised and removed. Next, the cask collar lid is removed. The rear access cover plate is also removed. An alignment system and the hydraulic skid positioners are used for the final alignment of the canister and module.
o) Module Loading - After final alignment, the canister is pulled into the HSM by the hydraulic ram (located at the rear of the HSM).
p) Storage - After the DSC is inside the HSM, the hydraulic ram is released from the DSC. The transfer trailer is pulled away and the HSM front access cover plate is closed. The seismic retainer is installed in the rear opening.
The rear access cover plate is also installed and secured in place. The DSC is now in storage within the HSM.
q) Retrieval- For retrieval, the cask is positioned as previously described and the hydraulic ram is used to push the DSC into the cask. All coupling, attachment, alignment, and closure operations are done in the same manner as previously described, but in reverse order. Once back in the cask, the DSC and its cargo of irradiated fuel assemblies are ready for shipment to a permanent repository or other storage location. Provisions have been made to return the canister to the HBR2 spent fuel pool if necessary.
1.3-4 Revision No. 22
HBRSEP ISFSI SAR TABLE 1.3-1 MAJOR SYSTEMS, SUBSYSTEMS AND COMPONENTS OF THE H. B. ROBINSON ISFSI Dry Shielded Canister Canister Basket Square Cells Spacer Disk Support Rods Canister Body Shielded End Plugs Top Cover Steel Plate Horizontal Storage Module Concrete Module Precast Outlet Shielding Blocks Dry Shielded Canister Support Assembly DSC Seismic Retaining Assembly Alignment System Front Access Cover Plate Rear Access Cover Plate Air Flow Penetrations Trailer Cask Positioning Skid Skid Positioning System (Vertical and Horizontal)
Transfer Cask Cask Body Cask Lids Cask Drains Cask Extension Collar See Figures 1.1-1, 1.2-1, and 1.3-1.
1.3-5 Revision No. 22
HBRSEP ISFSI SAR 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS The prime contractor for design and analysis of the HBR ISFSI was NUTECH, Inc. of San Jose, California. The construction was the responsibility of the CP&L onsite construction organization. Carolina Power & Light Company Nuclear Engineering Department, along with the Modification Implementation Section, were responsible for material procurement as well as contract administration.
1.4-1 Revision No. 22
HBRSEP ISFSI SAR 1.5 MATERIAL INCORPORATED BY REFERENCE The Topical Report for the NUTECH Horizontal Modular Storage (NUHOMS) System for Irradiated Nuclear Fuel (NUH-001) is hereby incorporated into this SAR by reference. The NUHOMS topical is referenced in Chapters 1, 3, 4, 5, 7, 8, 10, and 11. At the time of licensing of the HBR ISFSI, NUH-001, Revision 1, ADV001.0100, as submitted to the Nuclear Regulatory Commission by NUTECH Engineers Inc. in November 1985, was the applicable revision. The current version is Revision 2 as submitted by Pacific Nuclear Fuel Services, Inc. in March 1990.
The General Electric Company Safety Analysis Report for the IF-300 Shipping Container is also incorporated by reference. The IF-300 SAR is referenced in Chapters 3, 5, and 8. At the time of licensing of the HBR ISFSI, Revision 2 (NEDO-10084-2) was the applicable revision. The current version is Revision 5 (NEDO-10084-5) and was issued by Duratek.
1.5-1 Revision No.22
REFERENCES:
CHAPTER 1 1.1 CP&L/DOE Licensed At-Reactor Dry Storage Demonstration Program, Cooperative Agreement No.
DE-FC06-84RL10532, Amendment No. M005, October 1988.
1.2 NUTECH Engineers, Inc., "Topical Report for the NUTECH Horizontal Modular Storage System for Irradiated Nuclear Fuel," NUH-001, Revision 1, November 1985.
1.3 Docket Number 71-9001, Certificate of Compliance Number 9001 for General Electric Model No.
IF-300 Shipping Container, Package Identification No. USA/9001/B( )F.
1.4 Carolina Power and Light Company, "H. B. Robinson Steam Electric Plant Unit No. 2 Updated Final Safety Analysis Report," Docket No. 50-261, License No. DPR-23.
1.5 Carolina Power and Light Company, H. B. Robinson Steam Electric Plant, Plant Operating Manual, "Refueling Instruction Spent Fuel Cask Handling Instructions for Loading and Shipping of Power Fuel,"
FHP-034. Note - This was the applicable procedure at the time of loading of the ISFSI. This procedure has been deleted since the last loading in 1989.
1.6 Pacific Nuclear Fuel Services, Inc., Topical Report for the NUTECH Horizontal Modular Storage System for Irradiated Nuclear Fuel, NUH-001, Revision 2, March 1990.
1.R-1 Revision No. 22
Q x1-300 7Mmr8 CASU rr CASE SKID (HILUUN ROLLERS ASSEMBLY)
CASK/SKID TIEDOWN BRACKETS AMENDMENTNO. 5 H. B. ROBINSON INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT PRIMARY COMPONENTS OF THE ISFSI Figure 1.1-1
-5 LEGEND P
P P P X P X P X PROTECTED AREA P
CA P
P P P NA L X R X R X RADIATION BOUNDARY P
P DI P
P SC X V X V X VISITOR CENTER FENCE HA P
P P P P
RG P
E P P P P
P FUEL N
P P
458 P
P OIL BLDG P NOTES:
P DEEPWELL DEEPWELL PUMP "C" P REACTOR HEAD P P PUMP SPENT FUEL STORAGE BLDG. ENCLOSURE DIESEL GENERATOR P 370 P
- 1. FOR SERVICE WATER CABLE TRENCH P
DRY STORAGE BRE P P DPW-PMP-C-GEN 24P ISFSI -6 OIL P
LAYOUT SEE HBR2-12393.
P P
ISFSI P P DISPENSING SERVICE BLDG.
34.00' 165.00' P P IC TURBINE T BLDG.
RA BUILDING
- 2. FOR LOCATION OF PFSB (BLDG. #468)
BRE R ST NSF -7 R R R 135 PLANT OR O 457 R P AG RME P 220 E
PA R REFER TO HBR2-13720 AND HBR2-13723.
D FUEL P
E & RC BLDG. HANDLING P WELL #6 225 PUMP HOUSE BLDG 375 BLDG
- 3. FOR SOCA FENCING REFER TO HBR2-12020 HOT 215 473 AND HBR2-14405 THROUGH HBR2-14438 P P SHOP 461 FUEL BUILDING R
P BRE P 150 SPENT FUEL R -9 DRY STORAGE 250 230 FIRE 463A MAINTENANCE 466 P 7P ISFSI R CONT. CONT.
TRAINING FACILITY SHOP P 390 RCA DECON/
STOR. 205 STOR.
FACILITY TOOL ISSUE P REACTOR P 463B B.5.B. STORAGE R 210 UNIT 2 AUXILIARY RADWASTE 380 SECURITY CONTAINMENT BUILDING P BUILDING PAINT/FACILITIES 463C BLDG. BUILDING WAREHOUSE SHOP P CV 200 9
40 T ES TECH. TRAINING P
I ACCESS LI R RCA CI BRE BLDG CONTROL FA 325 -4 P PROCESSING ROOM 455 O&M FACILITY P
BLDG R
P P R R P 110 400 R RCA 385 330 P CHEM. STOR. BLDG. SECONDARY ENTRY STEAM SAMPLING SW PIPE P GEN. 331 WORK 335 350 ENCLOSURE P TOMB CONTROL 115 TURBINE BUILDING P
355 PAP EAST 401 P TURBINE 340 P BRE PROJECT EMER. -8 OFFICE DIESEL GEN. 315 AFW PUMP C GEN. 345 P 105 OUTAGE 318 TRANSFORMERS SWGR.
M ANAGEM ENT 405 DIESEL GENERATOR 310 MANAGEMENT BLDG TRANSFORMERS 300 BUILDING AFW-PMP-C-GEN COND.
BULK STORAGE WAREHOUSE PLANT POLISHER PAP WEST BRE TRANSFORMERS 469 P UNIT 1
-1 BLDG.
P P P 320 INTAKE P P P P P P OFFICE F-AFW-PMP-1 & 2 BUILDING BRE DEEPWELL PUMP "A" BRE E DEEPWELL REMOTE -3 P P P
-2 P P DIESEL GENERATOR P CONTROL PANEL PARKING DPW-PMP-A-GEN ROBINSON P P P P P P P P P P P P P P P P P P P P LAKE 406 P UNIT 2 P
LP TURBINE STORAGE WALK IN INTAKE WAREHOUSE COOLER WALK IN P FREEZER 230KV P P P G.T.S. SUT 115kV TRUCK PARKING 408 TRASH BLDG SWITCHYARD PARKING ADMIN. DUMPSTER 410 435 PARKING SLAB FACILITY FIRE WALL PARKING TSC TRAINING E DEEPWELL 175 170 BUILDING DG 230kV PARKING PARKING E DEEPWELL SWITCHYARD SECURITY BLDG.
PARKING 430 SIMULATOR BUILDING X
EPABX V WELD BLDG MAINTENANCE TEST OUTSIDE FAB X SHOP 420 425 SHOP/
415 SG CHANNEL PARKING V HEAD MOCK-UP X V X V X V X V SPARE RCP X
V MOTOR STORAGE PARKING X V
X V
X V ION X ETENT V
DEEPWELL PUMP "B" STE R X DIESEL GENERATOR WA SIN BA V
X DPW-PMP-B-GEN ION 0 45 V ETENT VI X E R ENTER EXIT SI T V WAST N CE OR
'S X BASI NT ER V X
V RE ST X
. V X
V PARKING PA RK ING LOT PICNIC AREA A
ENTER EXIT PARKING EAST ENTRANCE ACCESS CONTROL POINT GUARD HOUSE 456 LOT WEST F
ENTRANCE 3
.2 NO Y
WA IGH REF. HBR2-09800 H
C.
S.
H.B. ROBINSON INDEPENDENT SPENT FUEL STORAGE INSTALLATION LOT C
SAFETY ANALYSIS REPORT PLOT PLAN FIGURE 1.1-2 REVISION NO. 28
AMENDMENT 1 I H. B. ROBINSON INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT HORlZONTAL STORAGEMODULE Figure 1.2-1
7 BOTTOMEND PLUG LEAD SHIELD PLUG END COVERPLATe
Air Temp. , .
Amendment No. 1 H. B. ROBINSON INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT HSM AIR FLOW DIAGRAM Figure 1.3-2
F300 SHIPPING CASK CRADLE SKID ASSEMBLY P
c3 3
-I4 l
HORIZONTAL STORAGE 7 . . . . .
MODULE . .- l
- HYDRAULIC CYLINDER -..
. GRAPPLE ASSEMBLY
\ .;' (SEE DETAIL 1)
AR RAM SPIDER HSM FOUNDATION WEbGE LINK,ARM r
HYDRAULIC CYLINDER 1
L GRAPPLE ARM AMENDMENT HYDRAULIC RAM Figure 1.3-4
- 1. TRANSPORTCASK
- 2. DRY SHIELDEDCANISTER
- 3. IRRADIATEDFUEL ASSEMBLY
- 4. OVERHEADCRANE
- 5. IRRADIATEDFUEL STORAGEPOOL
- 6. TRANSFERTRAILER
- 7. SKID
- 8. TOWVEHICLE
- 9. HYDRAULICPOSITIONERS
- 10. HYDRAULICRAM
- 11. HORIZONTALSTORAGE MODULE
- 12. HORIZONTALROLLERS MODULE LOADING MODULE UNLOADING (RETRIEVAL)
HBRSEP ISFSI SAR CHAPTER 2 SITE CHARACTERISTICS Revision No. 22
HBRSEP ISFSI SAR 2.0 SITE CHARACTERISTICS 2.1 GEOGRAPHY AND DEMOGRAPHY 2.1.1 SITE LOCATION The Independent Spent Fuel Storage Installation (ISFSI) is located on the H. B. Robinson (HBR) Plant site. The site location is described in Section 2.1.1.1 of the UFSAR.
2.1.2 SITE DESCRIPTION A description of the site is provided in Section 2.1.1.2 of the UFSAR.
2.1.2.1 Other Activities Within the Site Boundary Carolina Power & Light Company owns and operates a 2339 MWt nuclear generating plant (Unit 2) on the HBR Plant site. The ISFSI is located within the protected area for the nuclear unit. Unit 2 received an operating license from the U. S. Atomic Energy Commission in 1970 (Docket No. 50-261/License No. DPR-23).
Other activities within the site boundary are described in Section 2.1.1.2 of the UFSAR.
A spur track of a commercial railroad branches from a mainline at McBee, South Carolina, and passes 1600 ft. west of the plant. Extensions of this spur enter the immediate plant area and run both north and south of the ISFSI. The maximum speed for locomotives on the tracks near the ISFSI is 5 mph.
Various chemicals, gas bottles, and fire sources such as oil are stored onsite, but are not located in the immediate vicinity of the ISFSI and hence do not represent a hazard to the facility.
2.1.2.2 Boundaries for Establishing Effluent Release Limits The exclusion zone is defined as the 1400 ft. radial area surrounding the plant. There are no residences or agricultural activities inside of the exclusion zone.
The controlled area for the ISFSI is defined as being contained within the HBR Unit 2 exclusion area. The protective area is within the Plant's exclusion zone. Exclusion area authority and control is discussed in Section 2.1.2 of the UFSAR.
2.1.3 POPULATION DISTRIBUTION AND TRENDS 2.1.3.1 Population Within 10 Miles The estimated resident population between zero and ten miles of the HBR Plant site is presented in Section 2.1.3 of the UFSAR.
2.1.3.2 Population Between 10 and 50 Miles The estimated resident population between ten and fifty miles is presented in Section 2.1.3 of the UFSAR.
2.1.3.3 Transient Population The transient population within 10 miles of the HBR Plant site is composed of four major components: the industrial labor force, seasonal population variation, school population, and hospital/nursing home populations. These are discussed in Section 2.1.3 of the UFSAR.
2.1.4 USES OF NEARBY LAND AND WATERS Uses of land and water are discussed in Sections 2.1 and 2.2 of the UFSAR.
2.1-1 Revision No. 22
HBRSEP ISFSI SAR 2.2 NEARBY INDUSTRIAL, TRANSPORTATION, AND MILITARY FACILITIES This information is provided in Section 2.2 of the UFSAR.
2.2-1 Revision No. 22
HBRSEP ISFSI SAR 2.3 METEOROLOGY Meteorological information, including regional climatology, local meteorology, onsite meteorological measurement programs, and diffusion estimates is provided in Section 2.3 of the UFSAR. The topography of the site is presented in Figure 2.1.1-2 of the UFSAR.
2.3-1 Revision No. 22
HBRSEP ISFSI SAR 2.4 SURFACE HYDROLOGY Surface hydrology information is presented in Section 2.4 of the UFSAR.
2.4.1 FLOODS Flooding of the HBR site, on which the ISFSI is located, including information on the probable maximum flood and probable maximum precipitation at the site, is addressed in Section 2.4.4 of the UFSAR. Flooding at the ISFSI will not occur because the facility grade is above the maximum lake level which can be maintained by the dam and appurtenant structures. Two different peak flows were calculated using the design unit hydrograph for the drainage area above the Lake Robinson Dam. Using these two peak flows, the lake level would not exceed 222 ft. during high flow conditions. Grade elevation at the location of the ISFSI is approximately 235 ft.
2.4.2 POTENTIAL DAM FAILURES The dams for both Lake Robinson and Prestwood Lake are downstream of the ISFSI; therefore, a dam failure would not flood the site, and since the ISFSI requires no water from the lakes for operation, a potential dam failure would not affect operation of the ISFSI.
2.4.3 PROBABLE MAXIMUM SURGE AND SEICHE FLOODING The meteorology of the area is discussed in Section 2.3. Lake Robinson is a long, narrow lake and meteorological and astronomical events do not cause significant effects. The maximum expected lake level during high flow conditions is 222 ft. The ISFSI is approximately 1400 ft. away from the lake and the grade elevation is 235 ft.
Therefore, surge and seiche flooding at the ISFSI will not occur.
2.4.4 PROBABLE MAXIMUM TSUNAMI FLOODING The ISFSI is not adjacent to coastal areas; therefore, tsunami flooding is not a credible event.
2.4.5 ICE FLOODING Due to the regional meteorological conditions (see Section 2.3) and the characteristics of the area water bodies (see Sections 2.4.1 and 2.4.2), ice flooding will not occur at the ISFSI.
2.4.6 FLOODING PROTECTION REQUIREMENTS Flooding of the ISFSI is not a credible event; therefore, no flood protection requirements are necessary.
2.4.7 ENVIRONMENTAL ACCEPTANCE OF EFFLUENTS The only liquid used for the ISFSI is during preparation of the spent fuel assemblies for loading into the GE IF-300 shipping canister is water. No liquids are used during the actual operation of the ISFSI.
Dispersion, dilution, and travel times of accidental releases of liquid effluents in surface water from the HBR Plant, discussed in Section 2.4.6 of the UFSAR, will remain unaffected by the ISFSI.
2.4-1 Revision No. 22
HBRSEP ISFSI SAR 2.5 SUBSURFACE HYDROLOGY The Independent Spent Fuel Storage Installation provides for the storage of spent nuclear fuel in a dry condition.
Therefore, there will be no consumption of groundwater or impact to the groundwater system as a result of installing the ISFSI at the HBR Plant site. Information on groundwater is presented in Section 2.4.7 of the UFSAR.
2.5-1 Revision No. 22
HBRSEP ISFSI SAR 2.6 GEOLOGY AND SEISMOLOGY Specific soil testing was performed at the designated location for the ISFSI. The data obtained from this testing was utilized in the foundation design for the ISFSI. As part of the foundation analysis/design, the subject of soil liquefaction was addressed. The following sections discuss the HBR Plant site geology and seismology. The foundation analysis is presented in Section 8.3 of this report. Geologic and seismic information for the site is presented in Section 2.5 of the UFSAR.
2.6.1 ISFSI Foundation A specific soil testing and foundation evaluations was performed at the designated location for the ISFSI. The data obtained from this testing was utilized in the foundation design for the ISFSI. The subject of soil liquefaction has been addressed as follows:
The analysis of the subsurface profile at the site of the ISFSI confirms previous findings made from other subsurface investigations carried out on the HBR2 site. The analysis of the data obtained from the ISFSI site borings used the procedures recommended by Seed, Idriss, and Arango for estimating liquefaction susceptibility using standard penetration resistances (N-values).
Data collected from the borings indicates potential liquefaction; however, this sample represents the soil conditions at a depth of approximately 100 feet. This loose soil zone is less than 5 feet thick, is located beneath approximately 23 feet of the very hard silty clay of the Middendorf Formation and is surrounded by approximately 25 feet of firm to dense sand. Hard silty clay continues below the sand stratum at a depth of approximately 112 feet.
Based on a review of the previous explorations performed at the HBR Plant site, there is no indication of a similar zone of loose sand located at the depths encountered by this boring. The 5-foot zone of loose sand is believed to be an isolated pocket or lens not representative of an area-wide layer in the Middendorf Formation. Based on this analysis, the ISFSI site profile is considered to have a very low to no likelihood of liquefaction during a design basis earthquake.
Because liquefaction of the loose sand pocket at 100 feet cannot be completely discounted, the effects of liquefaction were examined. In the case of a confined layer of potentially liquefiable saturated sands at large depths, as in this boring for the ISFSI, the paramount problem is not one of bearing capacity or landslide susceptibility as in surficial sands, but one of surface settlement following complete liquefaction. It can be expected that the hard silty clay and dense sands surrounding the loose sand layer would bridge any effects from the liquefaction of the loose sands and no surface settlements would occur.
2.6.2 SLOPE STABILITY The failure of any slopes at the HBR Plant site will not adversely affect the ISFSI. A discussion of slope stability of the earth dam and appurtenances at the HBR Plant site is provided in Section 2.5.5 of the UFSAR.
2.6-1 Revision No. 22
SUMMARY
OF SITE CONDITIONS AFFECTING CONSTRUCTION AND OPERATING REQUIREMENTS The HBR ISFSI is a totally passive installation designed by analysis to provide shielding and containment of irradiated fuel. The ISFSI is located within the protected area of the nuclear unit.
There are no residences or agricultural activities inside of the 1400 ft. radial area exclusion zone. The plant site has a temperate climate with the ISFSI facility grade above the maximum lake level which can be maintained by the dam and appurtenant structures.
The ISFSI will not affect any natural drainage features. A dam failure will also have no affect on the ISFSI as the dams for Lake Robinson and Lake Prestwood are downstream of the ISFSI.
On the basis of historical data, it is expected that the plant site area could experience a shock on the order of the 1959 McBee shock once during the life of the ISFSI facility. The design basis for the ISFSI is consistent with the HBR2 earthquake design basis of 0.2g horizontal acceleration.
A study of the possibility of the existence of faults in the area indicates that no active faulting was apparent. The sediments underlying the site are quite thick and apparently undisturbed. The surface of the buried crystallines is an ancient eroded one and active faulting is unknown in the vicinity of the site.
The failure of any slopes at the HBR site will not adversely affect the ISFSI.
2.7-1 Revision No. 22
HBRSEP ISFSI SAR CHAPTER 3 PRINCIPAL DESIGN CRITERIA Revision No. 22
HBRSEP ISFSI SAR 3.0 PRINCIPAL DESIGN CRITERIA 3.1 PURPOSE OF THE INSTALLATION The purpose of the H. B. Robinson (HBR) Independent Spent Fuel Storage Installation (ISFSI) is to provide the long-term storage of irradiated fuel assemblies (IFAs) in a dry environment. The HBR ISFSI is based on the NUTECH Horizontal Modular Storage (NUHOMS) System and is composed of a series of reinforced concrete horizontal storage modules (HSM). Each HSM will house a steel, helium filled, dry shielded canister (DSC) containing seven IFAs. The double weld sealed DSC serves as the confinement vessel for the IFAs while the HSM provides the biological shielding as well as a passive heat removal system for the decay heat of the IFAs.
3.1.1 MATERIAL TO BE STORED Each modular unit of the ISFSI is capable of storing up to seven pressurized water reactor (PWR) fuel assemblies as specified in Section 3.1 of the NUHOMS Topical Report (Reference 3.1). The following subsections describe the physical, thermal, and radiological parameters of the fuel to be stored at the HBR ISFSI.
3.1.1.1 Physical Characteristics The mechanical and structural design of the DSC is based on the physical characteristics of the PWR IFAs to be stored within the DSC. The physical characteristics of these IFAs are presented in Table 3.1-1. The overall length of the IFAs in Table 3.1-1 includes an allowance for 3/4 inch irradiation growth. Additional information on the physical characteristics of these fuel assemblies is contained in Section 4.2 of the HBR Updated Final Safety Analysis Report (UFSAR) (Reference 3.2).
3.1.1.2 Thermal Characteristics The heat generation per fuel assembly is limited to one kilowatt per assembly. This results in a maximum of seven kilowatts per DSC. Fuel irradiated to less than 35,000 MWd/MT and cooled for five years will meet this criteria (based on ORIGEN2 (Reference 3.3) calculations). Other combinations of burnup and cooling time may also be acceptable upon further analysis.
3.1.1.3 Radiological Characteristics The principal design criteria for acceptable radiological characteristics are shown in Table 3.1-2. The decay heat generation and the radiological characteristics described above bound fuel that has been irradiated to 35,000 MWd/MT and cooled for five years. This is considered to be the maximum radiation source fuel. The radiation source from the majority of the fuel to be stored will be less than or equal to that described in the NUHOMS Topical Report. ORIGEN2 calculations were used to determine this maximum burnup and minimum cooling time. Other combinations of decay time and burnup may also be acceptable if additional calculations are performed. However, for fuel which can be shown to have a burnup of less than or equal to 35,000 MWd/MT and been cooled for at least five years, no further fuel specific analysis is necessary.
3.1.2 GENERAL OPERATING FUNCTIONS 3.1.2.1 Overall Functions of the Facility The ISFSI is designed to be totally passive, requiring no utilities or waste processing system and to utilize HBR2's existing cask handling, fuel handling and associated auxiliary equipment in preparing the IFAs for dry storage. The cask to be used for the onsite transfer operation is the GE IF-300 shipping cask which is fully documented in Reference 3.5.
3.1-1 Revision No. 22
HBRSEP ISFSI SAR The DSC will be placed into the cavity of the GE IF-300 shipping cask. The shipping cask will then be lowered into the existing HBR2 spent fuel pool where seven fuel assemblies will be placed into the DSC. Once the top lead shield plug is placed onto the loaded DSC, the cask containing the loaded DSC will be raised out of the pool and the water drained partially from the DSC cavity. At this juncture, the top lead shield plug is welded into place. The DSC cavity will then be drained and vacuum dried of all water and backfilled with helium through the vent and siphon tube penetrations. After backfilling the DSC with helium, the penetrations will be seal welded and the top cover plate will be welded into place. With the sealed DSC still within the confines of the shipping cask, the shipping cask will be transported to the HSM and aligned with the front access of the HSM. A hydraulic ram will then extend from the rear access of the HSM through the HSM and attach onto the DSC grappling plate. The hydraulic ram will be retracted through the HSM, pulling the DSC into the HSM. Once the DSC is properly positioned within the HSM, the front and rear accesses of the HSM will be closed.
The HSMs which house the DSCs are located on a level, reinforced concrete, load bearing slab. The slab is designed for normal and postulated accident conditions. The HSM is also designed to maintain its dimensional and structural integrity during postulated environmental and geological events.
Safe storage in the HSM is provided by: (1) a natural convection heat removal path, (2) the concrete radiation shielding, and (3) the double closure welds of the DSC. The operation of the HSMs and DSCs is totally passive. No active systems are required.
The first three units of the ISFSI facility were used as part of the demonstration program. Therefore, two of the HSMs and the DSCs were initially instrumented for the purpose of collecting data. The instrumentation was limited to placement of a number of thermocouples in these components. The placement of thermocouples inside the HSMs does not effect its structural and mechanical integrity. The instrumentation of the DSC, however, requires a feed through penetration at its bottom cover plate. This penetration is designed such that the confinement integrity of the DSC is not compromised under both normal operating and accident conditions. Furthermore the instrumentation of these components does not change the total passive nature of the system. Details of the instrument penetration analysis are provided in Chapter 8 of this report. The instrumentation is no longer in service.
A more detailed description of each component's functions is located in the following subsections.
3.1.2.2 Handling and Transfer Equipment The ISFSI is designed to utilize the existing HBR2 fuel handling equipment and the GE IF-300 shipping cask. The function of the various HBR2 fuel handling equipment which are employed in loading and handling the IF-300 shipping cask are described in the plant's existing procedures and Section 9.1.4 of the HBR2 Updated FSAR (Reference 3.2).
The systems and equipment which are unique to the ISFSI for handling and transfer of IFAs are the DSC, the docking collar, the transfer trailer, the cask skid, the hydraulic ram, and the HSM. The function of each transfer and handling system or piece of equipment along with the waste processing system are briefly described in the following paragraphs.
a) DSC - The DSC will serve as the confinement vessel during transport of the IFAs to and from the HSM as well as during storage of the IFAs in the HSM. The shielded end plugs will provide biological shielding during transport of the fuel assemblies and also provide shielding for the front and rear accesses of the HSM.
b) Cask - The GE IF-300 shipping cask is used to transport the DSC to and from the HSM. The function of the shipping cask is to provide biological shielding along the axial length of the IFAs and a means for removing a sufficient quantity of decay heat so that the mechanical integrity of the DSC or IFAs is not jeopardized. The GE IF-300 is fully documented in Reference 3.5. As stated in Section 1.3.1.3 a new cask collar and lid are used on the IF-300 cask. This addition will provide the minimum cask cavity length requirement.
3.1-2 Revision No. 22
HBRSEP ISFSI SAR c) Cask Positioning Skid - The function of the cask positioning skid is to provide a means by which the final alignment of the cask with respect to the HSM can be achieved, and to restrain the shipping cask in the horizontal position during the transfer of the DSC to and from the HSM. The skid and the cask will be transported from the fuel building to the HSMs by a trailer. Both the skid and the DSC are designed to withstand the inertia forces associated with the transportation shock loads.
d) Hydraulic Ram - The hydraulic ram is used to move the DSC between the cask and the HSM. The hydraulic ram is a long hydraulic cylinder with a grapple device mounted at the cylinder's head. The hydraulic ram will be positioned at the rear access of the HSM. The ram will then be extended through the rear access and through the entire length of the HSM. Once the grapple is seated within the grappling plate of the DSC, the arms of the grapple will be extended so that it is securely in place between the DSC and the grapple plate. The ram will then be retracted, pulling the DSC out of the cask cavity and into the HSM.
e) Horizontal Storage Module- The function of the HSM is to provide protection for the DSC against geological and environmental events as specified in Section 3.2, and serve as the principle biological shield for irradiated fuel during storage. The HSM contains shielded air ducts near the bottom of the structure to admit ambient air around the DSC for cooling purposes. The air, warmed by the canister, is exhausted through shielded vents at the top of the HSM by natural draft convection. The HSM also provides support for the DSC. The DSC rests on a support rail assembly which is anchored to the walls of the HSM. The rear end of the DSC support rails are equipped with stopping blocks. The purpose of these stopping blocks is to establish the final axial position of the DSC inside the HSM. The final positioning is achieved when the DSC, being pulled into the HSM by the ram, makes contact with these stopping blocks. After the DSC insertion into the HSM is completed, a seismic retaining assembly will be attached to the grappling plate of the DSC top cover plate. This will prevent any possible axial movement of the DSC during any postulated accident such as an earthquake. The front access of the HSM is covered by a plate. The air inlets and outlets are covered with stainless steel wire bird screens to prevent foreign objects from entering the HSM.
f) Waste Processing - During the normal storage of IFA's in the ISFSI, no waste will be generated. However, contaminated water and possibly contaminated gases will be removed from the DSC cavity during the cask drying operation. The cask drying operation will take place in the HBR2 decontamination facility. The HBR2 utilities and radioactive waste processing system are described in Section 9.1.4. and Chapter 11.0 of the HBR2 Updated FSAR (Reference 3.2).
3.1-3 Revision No. 22
HBRSEP ISFSI SAR TABLE 3.1-1 PHYSICAL CHARACTERISTICS OF PWR FUEL ASSEMBLIES1 BASED ON NOMINAL DESIGN Array 15 x 15 Envelope (in) 8.426 Overall Length2 (in) 162.05 Weight (lbs) 1466 Fuel Rod Number 204 Fuel Rod Length (in) 152.0 Active Fuel Length (in) 144.0 Maximum Distance Between 26.19 Grid Straps (in) 1 See Reference 3.4.
2 Additional 3/4 inch added to overall length to allow for irradiation growth.
3.1-4 Revision No. 22
HBRSEP ISFSI SAR TABLE 3.1-2 ACCEPTABLE RADIOLOGICAL CRITERIA FOR STORAGE OF MATERIAL IN THE HBR ISFSI CRITERIA VALUE 16 GAMMA SOURCE PER CANISTER (total) 1.48 x 10 Mev/sec Fractional Breakdown*
Above 1.3 Mev 0.004 Between 1.3 Mev and 0.8 Mev 0.114 Between 0.8 Mev and 0.4 Mev 0.808 Below 0.4 Mev 0.074 NEUTRON SOURCE PER CANISTER (total)** 1.17 x 109 n/sec Fractional Breakdown Above 5 Mev 5.40 x 107 n/sec = 5.41%
Between 2.5 and 5 Mev 2.43 x 108 n/sec = 24.32%
Between 1 and 2.5 Mev 4.56 x 108 n/sec = 45.67%
Below 1 Mev 2.45 x 108 n/sec = 24.53%
- Fractional breakdown based on isotopic composition and resulting gamma spectrum calculated by ORIGEN2 analysis.
- Spectrum from U-235 fission, total number of neutrons per second from ORIGEN2 analysis.
3.1-5 Revision No. 22
HBRSEP ISFSI SAR 3.2 STRUCTURAL AND MECHANICAL SAFETY CRITERIA The HBR ISFSI is designed to perform its intended function under extreme environmental and geological hazards as specified in 10CFR, Part 72.122(a). The HSMs are installed on a monolithically placed, reinforced concrete mat foundation. The mat foundation is also designed to resist forces generated by extreme environmental and geological conditions. Specifics of the foundation design are reported in Section 8.3.
The environmental features at the ISFSI site, which are used to define the normal operating design basis for the DSCs and HSMs, are such that they are the same as or enveloped by those specified in Table 3.2-1 of the NUHOMS Topical Report (Reference 3.1). Specifically, the highest recorded ambient temperature of 107oF, and the lowest recorded temperature of -5oF, are well within the extreme ambient range of 125oF to -40oF specified in the NUHOMS Topical Report. The maximum diurnal temperature range for the ISFSI site is 25oF. This range is also lower than the 45oF diurnal temperature range specified in the above referenced report. The site area design basis solar radiation value of 188 Btu/hr-(ft2) is the same as that reported in the above referenced report.
In general, the structural and mechanical safety criteria of the ISFSI are the same as or enveloped by the criteria specified in Section 3.2 of the NUHOMS Topical Report.
As stated earlier in this report, some design features of the HBR ISFSI differ from those of the NUHOMS generic concept. In particular, the NUHOMS Topical Report addresses an eight module unit, whereas the HBR ISFSI consists of a three-module unit and a five-module unit. Each unit will be a monolithically poured in place unit, with rear access penetrations for each HSM which allows for the hydraulic ram to be operated from the back of the modules. These unique features of the ISFSI reinforced concrete modules have no impact on the structural evaluation presented throughout chapter 8 of the NUHOMS Topical Report. This is due to the fact that the NUHOMS structural analysis of the HSM utilizes the frame action of the roof slab and the walls of only a single module in the transverse direction. This single module approach conservatively envelopes the structural response of any multi-module units including the three or five-module concept which is the subject of this report. The rear access penetration, however, require additional shielding evaluation which is presented in chapter 7 of this report.
Some of the design features of the HBR DSCs also differ from those of the NUHOMS Topical Report. In particular, the bottom region of the DSC has been redesigned to fit into the IF-300 cask. Furthermore, both the top and bottom regions, the spacer disk and the support rods of the DSC have been redesigned to withstand the inertia forces associated with cask drop accidents in which the drop heights are significantly greater than the minimum 8 foot drop criteria established earlier in this report. This was done for compatibility with future shipping options. Another unique feature of the HBR DSCs is the instrument penetration at the bottom region of two of the initial three DSC assemblies. The instrumentation of the DSC was for the purpose of collecting temperature data during the first year.
The penetration assembly is designed to maintain the confinement integrity of the DSC during both normal operating and accident conditions. The structural evaluation of the instrument penetration under various normal operating and accident conditions is addressed in chapter 8 of this report.
3.2.1 TORNADO AND WIND LOADINGS 3.2.1.1 Applicable Design Parameters The ISFSI is constructed within the existing boundaries of the HBR Steam Electric Plant which is located within Region I of the NRC Regulatory Guide 1.76 regionalization, and as such, the intensity of the design basis tornado for this region is the same as that assumed in the NUHOMS Topical Report (Reference 3.1).
The design basis tornado (DBT) intensities were obtained from NRC Regulatory Guide 1.76. Region I intensities were considered since it has the most severe parameters. For this region, the maximum wind speed is 360 miles per hour, the rotational speed is 290 miles per hour, the maximum translational speed is 70 miles per hour, the radius or maximum rotational speed is 150 feet, the pressure drop across the tornado is 3.0 psi, and the rate of pressure drop is 2.0 psi per second. The maximum transit time based on the specified 5 miles per hour minimum translational speed was not used since the transit time is conservatively assumed to be infinite.
3.2-1 Revision No. 22
HBRSEP ISFSI SAR 3.2.1.2 Determination of Forces on the Structures The forces due to the design basis tornado and tornado generated missiles are enveloped by those reported in the NUHOMS Topical Report.
The method of analysis for overall and local damage prediction due to a design basis tornado and tornado generated missiles is discussed in Section 3.2.1 of the NUHOMS Topical Report, and is fully applicable to the HBR site specific analysis.
3.2.1.3 Ability of Structures to Perform The ISFSI is designed to withstand the design basis tornado wind loads. Furthermore, all components of the ISFSI with the exception of the air outlet shielding block are designed to withstand the tornado generated missile forces.
The loss of an air outlet shielding block is addressed in Section 8.2.1 of this report.
Since the ISFSI is not housed in any storage building, there is no possibility of any roof collapse on the facility.
However, the possibility of total air inlet and outlet blockage by foreign objects or burial under debris during a tornado event is considered. The effect of facility burial under debris is presented in Section 8.2.
3.2.2 WATER LEVEL (FLOOD) DESIGN The maximum flood water level at the HBR site is 222 ft. elevation (see Section 2.4 of the UFSAR). The grade level of the ISFSI foundation is at the 234 ft. elevation. Therefore, there is no possibility of flooding within the ISFSI.
3.2.3 SEISMIC DESIGN 3.2.3.1 Input Criteria The maximum horizontal ground acceleration specified for HBR2 is 0.20g for safe shutdown earthquake (SSE) (see Section 2.6.2.3 and Reference 3.2). The maximum vertical ground acceleration is specified at two thirds of the hori-zontal component, or 0.133g. These horizontal and vertical component values are less than the values of 0.25g and 0.17g, respectively, specified in the NUHOMS Topical Report (reference 3.1). Hence, the seismic evaluation con-tained in Section 8.2.3 of the referenced report is fully applicable to those components of the H. B. Robinson ISFSI which have design features similar to those of the NUHOMS generic concept. It is also applicable to those that can be conservatively enveloped by the NUHOMS generic assumptions, such as the single module approach to HSM evaluation discussed earlier in this section. In this manner, the seismic response of the HSM and the DSC support assemblies of the HBR ISFSI are enveloped by the responses reported in the NUHOMS Topical Report for these components. The DSC, however, which has some unique features, is analyzed for the site specific seismic event, utilizing the methodology and the analytical approach of the NUHOMS Topical Report. The seismic evaluation of the DSC is contained in Section 8.2 of this SAR.
3.2.3.2 Seismic-System Analysis The stresses in the HSM and the DSC support assembly due to the 0.20g horizontal and 0.133g vertical acceleration are enveloped by the results of the generic seismic analysis reported in the NUHOMS Topical Report. The stresses in the DSC due to the horizontal and vertical seismic acceleration specified above are evaluated and reported along with the DSC rollover evaluation in Section 8.2 of this report.
The ISFSI foundation and the HSMs tie down system are also designed to withstand the forces generated by the SSE. The details of the foundation design are provided in Section 8.3 of this report.
3.2-2 Revision No. 22
HBRSEP ISFSI SAR 3.2.4 SNOW AND ICE LOADS The NUHOMS Topical Report specified a postulated live load of 200 pounds per square foot which conservatively envelopes the maximum snow loads for the HBR site (see Section 2.3 of the UFSAR for meteorology of the site area).
3.2.5 COMBINED LOAD CRITERIA Load combination criteria established in the NUHOMS Topical Report (Reference 3.1) for the HSM, DSC and DSC support assembly are also applicable to the HBR ISFSI. The specific load combination evaluation of the DSC, utilizing the NUHOMS criteria, is reported in Section 8.2 of this report.
The facility's mat foundation is designed to meet the requirements of ACI 349-80 (3.6). The ultimate strength method of analysis was utilized with appropriate strength reduction factors. The load combination procedure of Section 9.2.1 of the ACI 349.80 was used in combining normal operating loads (i.e. dead loads and live loads) with severe and extreme loads (i.e., seismic and tornado loads). The details of the load combination procedure are described in the NUHOMS Topical Report. Specific foundation analyses including load combination and anchorage analysis are presented in Section 8.3 of this report.
3.2-3 Revision No. 22
HBRSEP ISFSI SAR 3.3 SAFETY PROTECTION SYSTEM 3.3.1 GENERAL The HBR Independent Spent Fuel Storage Installation is designed for safe and secure, long-term containment and storage of IFAs. The equipment which must be designed to assure that the safety objectives are met are shown in Table 3.3-1.
The major features which require special design consideration are:
a) Double Closure Seal Welds on DSC Upper End b) Radiation Exposure During DSC Drying and Closure c) Design of DSC Body and Internals for a Cask Drop Event During the Transfer Operation d) Minimization of Contamination of the DSC Exterior by the Spent Fuel Pool Water e) Minimization of Radiation Shine During Transfer of the DSC from the Cask to the HSM These items are addressed in the following subsections.
3.3.2 PROTECTION BY MULTIPLE CONFINEMENT BARRIERS AND SYSTEMS 3.3.2.1 Confinement Barriers and Systems The ISFSI relies on a system of multiple confinement barriers during all handling and storage operations. Table 3.3-2 from the NUHOMS Topical Report (Reference 3.1) has been included and summarizes the radioactive confinement barriers and systems employed in the design of the ISFSI.
During transport and storage operations, the IFAs are confined within the DSC. The DSC consists of a cylindrical shell and multiple end plates. Each end plate will be seal welded to the canister in order to provide redundant seals for the DSC. These redundant seals minimize the likelihood of an uncontrollable release of radioactivity. Detailed discussion of the DSC confinement integrity, including the discussion on helium confinement, is presented in Section 3.3.2.1 of the NUHOMS Topical Report and is fully applicable here.
The criteria for protection against any postulated internal or external natural phenomena are discussed in Section 3.2 and Chapter 8 of this report.
3.3.2.2 Ventilation - Offgas During the normal storage operations of the ISFSI, there will be essentially no release of radioactive material.
Additionally, as discussed in Chapter 8 of the NUHOMS Topical Report (Reference 3.1), there are no credible accidents which could cause a release of radioactivity. Therefore, the HBR ISFSI does not require an offgas system.
During the cask drying operation, water and gas will be removed from the cavity of the DSC. This operation will take place in the HBR2 decontamination facility and the water and gas will be routed through Unit 2's existing radioactive waste processing system.
3.3.3 PROTECTION BY EQUIPMENT AND INSTRUMENTATION SELECTION 3.3.3.1 Equipment The DSC and the GE IF-300 shipping cask are the only equipment that specifically provide protection during normal and off-normal operations of the ISFSI. The design criteria for the DSC are provided in Section 3.2 of this SAR.
The design criteria for the cask are listed in the GE IF-300 Safety Analysis Report (Reference 3.5).
3.3-1 Revision No. 22
HBRSEP ISFSI SAR 3.3.3.2 Instrumentation The HBR ISFSI is designed to be totally passive and therefore, no safety related instrumentation is required for operation of the facility. However, two of the DSCs and HSMs were instrumented for experimental purposes only for a one year test period (Agreements with DOE and EPRI).
The instrumentation was limited to placement of a number of thermocouples within these components.
Instrumentation of the HSM does not affect its structural and mechanical properties. The placement of thermocouples in the DSC, however, requires a feed-through penetration at the DSC bottom region. This feed-through incorporates the same redundant seal philosophy used in the DSC design. The penetration is also designed such that confinement integrity of the DSC is not compromised under both normal operating and accident conditions.
The instrument penetration analysis is provided in Section 8.4 of this report.
3.3.4 NUCLEAR CRITICALITY SAFETY The DSC internals are designed to provide nuclear criticality safety during wet loading operations. A combination of administrative procedures, materials properties, geometry, and neutron poisons are used to assure that subcritical conditions exist at all times.
The DSC internals were analyzed for criticality safety using the KENO-IV Monte Carlo criticality code. The calculational procedures, cross-section sets, biasing techniques, and computer hardware used to perform the criticality analyses were identical to the methods used in the NUHOMS generic design (refer to the NUHOMS Topical Report - Reference 3.1).
Since the DSC for the HBR ISFSI design uses a different boron content in the aluminum boron alloy material on the fuel guide sleeves, additional calculations were performed to assure criticality safety. The 70° H2O case (Case No. 1, Table 3.3-4 of the NUHOMS Topical Report) was executed using the minimum specified boron content (4.5%). Twenty-one thousand neutron histories were executed to obtain the maximum k .
eff Max K = K + 2 sigma + K eff eff bias
= 0.91960 + 2(0.00537) + 0.0174
= 0.94774 For details of the analysis methodology, and a discussion of the computational bias, see Section 3.3.4 of the NUHOMS Topical Report (Reference 3.1).
3.3.5 RADIOLOGICAL PROTECTION 3.3.5.1 Access Control A fence with a locked gate is installed partially around the ISFSI for the purpose of designating the area a radiation control area. The key is controlled by the HBR2 Radiation Control unit. Access to the ISFSI is on an as needed basis.
3.3.5.2 Shielding Estimates of collective onsite and offsite doses during operations and around the ISFSI are presented in Chapter 7.
3.3.5.3 Radiological Alarm System There are no radiological alarms required.
3.3-2 Revision No. 22
HBRSEP ISFSI SAR 3.3.6 FIRE AND EXPLOSION PROTECTION 3.3.6.1 Fire Protection The degree of fire protection a structure requires is based on a number of factors: type and location of combustible materials and their proximity to or location within the ISFSI, type of construction and its fire resistance characteristics, fire barriers, and the ability of the plant's fire brigade to reach and effectively extinguish a credible fire.
No combustible materials are stored within the ISFSI or within the ISFSI's boundaries. There is no fixed fire suppression system within the boundaries of the ISFSI. The facility is, however, located outside the confines of any building and is directly accessible to HBR2's fire brigade. The fire brigade has access to HBR2's existing portable fire suppression equipment or the site's water fire protection system, as described in Section 9.5.1 of the HBR2 FSAR (Reference 3.2).
3.3.6.2 Explosion Protection The DSC and HSM contain no volatile materials and therefore, no credible internal explosion is possible. Internal explosions are not considered as part of the design criteria. The design basis for explosions away from the HSM is bounded by the design basis tornado missile described in Section 3.2 of this report and of the NUHOMS Topical Report (Reference 3.1).
3.3.7 MATERIALS HANDLING AND STORAGE 3.3.7.1 Irradiated Fuel Handling and Storage The fuel handling systems used in loading the IFAs into the DSC are presented in Section 9.1.4 of the HBR2 UFSAR (Reference 3.2). Irradiated fuel handling outside the spent fuel storage pool will be done with the fuel assemblies enclosed in the canister. Criticality safety during handling and storage is discussed in Section 3.3.4. The criterion for safe configuration is an effective mean plus two-sigma neutron multiplication factor (keff) of 0.95.
Calculations have shown that the expected keff value is well below this limit.
The basic criterion for the cooling of irradiated fuel during storage is a maximum cladding temperature of 380oC (716oF). Higher temperatures may be sustained for brief periods without endangering cladding integrity. During canister drying and other normal and abnormal transients, the criterion is a cladding temperature of 570oC for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. For further details, see Section 3.3.7.1 of the NUHOMS Topical Report (Reference 3.1).
The canister external contamination limits are listed below:
-3 Beta/Gamma Emitters 10 Ci/Cm2 or 220,000 dis/min/100 cm2
-5 Alpha Emitters 10 Ci/Cm2 or 2,200 dis/min/100 cm2 The canister is sealed by double welds prior to storage so that any contamination of the canister interior or its contents will remain confined during transfer and storage.
3.3.7.2 Radioactive Waste Treatment The contaminated water removed from the cask annulus and cavity of the loaded DSC will be handled by HBR2's radioactive waste treatment system. The site's radioactive waste treatment system is described in Chapter 11 of the UFSAR.
3.3-3 Revision No. 22
HBRSEP ISFSI SAR 3.3.7.3 Waste Storage Facilities No radioactive wastes will be generated during the life of the ISFSI. The contaminated water removed from the cask annulus and cavity of the loaded DSC will be handled by HBR2's radioactive waste treatment system. The waste storage facility associated with the HBR2 radioactive waste treatment system is described in Chapter 11 of the UFSAR.
3.3.8 INDUSTRIAL AND CHEMICAL SAFETY No hazardous or volatile chemicals or chemical reactions are involved in the operation of the ISFSI and therefore, were not considered in any of the facility's design criteria.
3.3-4 Revision No. 22
HBRSEP ISFSI SAR TABLE 3.3-1 H. B. ROBINSON ISFSI IMPORTANT TO SAFETY (SAFETY RELATED) FEATURES I. Transfer Cask (IF-300) Safety Related1 (Q-List)
II. Dry Storage Canister (DSC) Safety Related2 (Q-List)
II.a. Basket II.b. Stiffener Plates II.c. Support Rods II.d. Lead Plug/Support II.e. Canister Body II.f. End Closure Plates III. Horizontal Storage Module (HSM) Non-Safety Related3 III.a. Concrete Shielding III.b. DSC Support Assembly IV. Foundation Non-Safety Related3 V. Transfer Components Non-Safety Related V.a. Transport Vehicle V.b. Hydraulic Ram VI. Instrumentation Non-Safety Related Notes:
- 1. As defined by 10 CFR 71 for a licensed transportation cask.
- 2. For the purposes of this license application, CP&L considers the Dry Storage Canister Safety Related as defined by 10 CFR 50-Appendix B.
- 3. CP&L applied a "Radwaste Related" QA program to these components for the procurement, construction and testing phases of the project. The 10 CFR 50 Appendix B program as described in RNP UFSAR Section 17.3 shall be applied to the operational phase of the project.
3.3-5 Revision No. 22
HBRSEP ISFSI SAR TABLE 3.3-2 RADIOACTIVITY CONFINEMENT BARRIERS AND SYSTEMS OF THE ISFSI Radioactivity Confinement Barriers and Systems Contaminated Spent Fuel Pool Water 1. Demineralized Water in Cask
- 2. Cask/DSC Annulus Seal Irradiated Fuel Assemblies 1. Fuel Cladding
- 2. DSC Body
- 3. Seal Welded Primary Closure
- 4. Seal Welded Secondary Closure 3.3-6 Revision No. 22
HBRSEP ISFSI SAR 3.4 CLASSIFICATION OF STRUCTURES, COMPONENTS, AND SYSTEMS A classification of the ISFSI's various structures, components, and systems is listed in Table 3.3-1.
3.4-1 Revision No. 22
HBRSEP ISFSI SAR 3.5 DECOMMISSIONING CONSIDERATIONS The dry storage canisters are intended to be transferred to a federal repository when such a facility is operational.
The concrete module is designed so that the canister can be safely returned to a shipping cask and transported offsite to the federal repository.
Shipping cask design and transportation requirements will depend on the regulations in effect at the time when the federal repository begins receiving spent fuel.
Contamination on the canister exterior will be very small (see Section 3.3.7 of the NUHOMS Topical Report (Reference 3.1)). The resulting contamination of the module internals and air passages will also be minimal. This level of contamination may be removed by manual methods so that the reinforced concrete module can be broken-up and removed using conventional methods.
The canister itself may be contaminated internally by crud from the irradiated fuel and will be slightly activated by spontaneous neutron emissions from the irradiated fuel. The canister is designed to be used in the repository for final disposal; however, if the fuel is removed from the canister, the canister could be disposed of as low-level waste.
The exact decommissioning plan to be applied will depend on the status of the U.S. waste repository program at the time of decommissioning.
3.5-1 Revision No. 22
REFERENCES:
CHAPTER 3 3.1 NUTECH Engineers, Inc. "Topical Report for the NUTECH Horizontal Modular Storage System For Irradiated Nuclear Fuel, NUH-001, Revision 1, November 1985. (Note - currently NUH-001, Rev. 2) 3.2 Carolina Power and Light Company, "H. B. Robinson Steam Electric Plant Unit No. 2 Updated Final Safety Analysis Report," Docket No. 50-261, License No. DPR-23.
3.3 Oak Ridge National Laboratory, "ORIGEN2 - Isotope Generation and Depletion Code - Matrix Exponential Method," CCC-371, 1982.
3.4 Exxon Nuclear Company, Inc., Drawing NX-302.517 (NP), Exxon Nuclear Company, Inc., Richland, Washington, 1975.
3.5 General Electric Co., "IF-300 Shipping Cask Consolidated Safety Analysis Report," NEDO-10084-2, Nuclear Fuel and Special Products Division, March l983. (Note - currently NEDO-10084-5 issued by Duratek) 3.6 American Concrete Institute, Code Requirements for Nuclear Safety Related Concrete Structures and Commentary, ACI 349-80 and ACI 349R-80, American Concrete Institute, Detroit, Michigan, 1980.
3.R-1 Revision No. 22
HBRSEP ISFSI SAR CHAPTER 4 INSTALLATION DESIGN Revision No. 22
HBRSEP ISFSI SAR 4.0 INSTALLATION DESIGN 4.1
SUMMARY
DESCRIPTION 4.1.1 LOCATION AND LAYOUT OF INSTALLATION The H. B. Robinson (HBR) Independent Spent Fuel Storage Installation (ISFSI) is located within the existing site boundary of the H. B. Robinson Steam Electric Plant, Unit No. 2 (HBR2). The design of the ISFSI is based on the NUTECH Horizontal Modular Storage (NUHOMS) System for irradiated nuclear fuel. The ISFSI is composed of a series of reinforced concrete horizontal storage modules (HSM). Each HSM contains one dry shielded canister (DSC) which serves as the confinement barrier for the irradiated fuel assemblies (IFAs). The following sections describe the ISFSI, its components and their interrelationship with HBR2's existing facilities. Figure 1.2.2-1 of the UFSAR shows the location of the ISFSI at HBR2.
4.1.2 PRINCIPLE FEATURES 4.1.2.1 Site Boundary The site boundary is shown on Figure 2.1.1-4 of the UFSAR.
4.1.2.2 Controlled Area The controlled area is shown on Figure 1.2.2-1 of the UFSAR.
4.1.2.3 Emergency Planning Zone The exclusion zone is shown on Figure 2.1.1-4 of the UFSAR.
4.1.2.4 Site Utility Supplies and Systems No utility supplies and systems are required for the ISFSI.
4.1.2.5 Storage Facilities HBR2 facilities are shown on Figure 1.2.2-1 of the UFSAR.
4.1.2.6 Stack The ISFSI has no stack.
4.1-1 Revision No. 22
HBRSEP ISFSI SAR 4.2 STORAGE STRUCTURES 4.2.1 STRUCTURAL SPECIFICATIONS 4.2.1.1 Design Basis The ISFSI and all its components are designed in accordance with the requirements of 10CFR Part 72. The components are designed to maintain their dimensional and structural integrity and safely perform their intended functions under normal and off-normal operating conditions and during postulated geological or environmental events. The loading conditions associated with normal and accident events are specified in Section 3.2 and Chapter 8 of this report.
4.2.1.2 Construction, Fabrication, and Inspection a) Horizontal Storage Module - The HSM is constructed of reinforced concrete. The HSM is designed in accordance with the requirements of the ACI 349-80. The applicable American Society of Testing and Material (ASTM) standards referenced in Section 3.8 of the ACI code was used as part of the fabrication and construction requirements of the HSM. Construction of the HSM is in accordance with ACI 301-84, Specification for Structural Concrete for Buildings.
The concrete materials used for construction of the HSM and the foundation consist of: Type II cement conforming to ASTM C150, fine and coarse aggregates conforming to ASTM C33, concrete air-entraining admixture conforming to ASTM C-260, and reinforcing bars conforming to ASTM A615 Grade 60. The concrete compressive strength specified is 4000 psi at 28 days. Minimum specified density of concrete is 145 pounds per cubic foot. The HSM walls are tied to the foundation by reinforcing dowels to prevent possible overturning or sliding during any accident condition specified in Section 8.2 of this report.
b) Dry Shielded Canister - The DSC is designed and fabricated in accordance with the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section III, Division 1, Subsection NB, 1983 Edition. Material selection, welding, and inspection of the DSC was per requirement of this code and the applicable ASTM, ANSI, and other codes and standards invoked by the ASME code.
Details of the material and construction of the DSC and its internals are contained in Section 4.2.3.1 of the NUHOMS Topical Report (Reference 4.1).
4.2.2 INSTALLATION LAYOUT 4.2.2.1 Building Plans There is no building associated with the HBR2 ISFSI other than the horizontal storage module which is discussed in Section 1.3.1.2.
4.2.2.2 Confinement Features The confinement features for each unit of the ISFSI are provided in Section 3.3.2.1.
4.2.3 INDIVIDUAL UNIT DESCRIPTION Each unit of the ISFSI is composed of two components; a dry shielded canister and a horizontal storage module.
The DSC is a weld-sealed stainless steel container that provides confinement of contaminants associated with the irradiated fuel assemblies, encloses the fuel basket in an inert atmosphere and provides biological shielding at the ends of the canister. The HSM is a reinforced concrete structure that serves as the primary biological shield for the irradiated fuel assemblies along with providing protection for the DSC against environmental and geological hazards. A detailed discussion and description of the function, design basis, and safety assurance considerations of 4.2-1 Revision No. 22
HBRSEP ISFSI SAR each component are provided in Section 4.2.3.1 of the NUHOMS Topical Report (Reference 4.1). The engineering drawings of the DSC and the HSM, showing plan views, sections, and elevations of these components are provided in Figures 4.2-1 and 4.2-2.
Two DSCs and their corresponding HSMs were instrumented for the purpose of collecting data during the first year of storage. The instrumentation was limited to placement of thermocouples in the DSC and the HSM. The HSM thermocouples are cast in place or attached to the concrete surfaces, and as such, have no impact on the integrity of the structure. The DSC thermocouples were connected to an external cable by means of a specially designed feed-through. This feed-through penetration incorporates the same redundant seal philosophy used in the DSC containment design. After the penetration plug assembly has been welded to the bottom of the DSC cover plate, a sleeve was welded over the plug, forming a redundant seal. Thermocouple sheaths were brazed to the plug assembly at inner and outer surfaces of the penetrations. To ensure that possible leakage through the aluminum oxide insulation is precluded, each end of the brazed thermocouple was sealed with an environmentally qualified resin.
4.2-2 Revision No. 22
HBRSEP ISFSI SAR 4.3 AUXILIARY SYSTEMS The design of the HBR ISFSI is based on the NUHOMS system for irradiated fuel. Each unit of this dry storage system is totally passive and self contained, requiring no auxiliary systems other than a transfer cask for transport operation.
4.3.1 VENTILATION AND OFFGAS SYSTEM 4.3.1.1 Ventilation System The decay heat rejection system for each module is based on natural circulation convection cooling. The IFAs are confined within a double weld sealed DSC. The decay heat from the IFAs is transferred by radiation, conduction and convection to the surface of the canister. Air inlets near the bottom and air outlets at the top of each HSM allow air to circulate around the DSC. Decay heat is removed by convection and radiation from the surface of the DSC.
The driving force for circulation of the air is thermal buoyancy. An analysis of the HSM ventilation system is described in Section 8.1.3 of the NUHOMS Topical Report (Reference 4.1). Since the IFAs are confined within the double weld sealed DSC and the contamination on the external surface of the DSC results in essentially no release of radioactive material, no filtration system for contamination is required.
4.3.1.2 Offgas System The ventilation system used to process the offgassing during the DSC drying and backfilling operations is the same as that described in Chapter 9 of the UFSAR.
4.3.2 ELECTRICAL SYSTEM The ISFSI is totally passive and requires no electrical system. The HBR2 electrical system associated with the fuel handling area is utilized during DSC drying and backfill operations.
4.3.3 AIR SUPPLY SYSTEM 4.3.3.1 Compressed Air The ISFSI requires no compressed air supply system. The HBR2 compressed air supply system will be utilized during the DSC drying operation.
4.3.3.2 Breathing Air The ISFSI is located in an outside environment and requires no breathing air supply. Provisions for breathing air supply during an emergency situation exist at HBR2.
4.3.4 STEAM SUPPLY AND DISTRIBUTION SYSTEM The ISFSI requires no steam supply.
4.3.5 WATER SUPPLY SYSTEM The ISFSI requires no water supply system. The HBR2 water supply system associated with the fuel handling area will be used during cask preparation and decontamination operations.
4.3.5.1 Major Components and Operating Characteristics The existing water supply system of HBR2 will be utilized during the cask preparation and decontamination operations and for fire protection purposes.
4.3-1 Revision No. 22
HBRSEP ISFSI SAR 4.3.5.2 Safety Consideration and Controls A loss of water supply would have no effect on the operation of the ISFSI. The water is used for removing contamination from the surface of the cask and DSC. Any loss of water during the cask wash down and loading operations would only extend the period of time which the DSC would be contained within the IF-300 shipping cask.
4.3.6 SEWAGE TREATMENT SYSTEM 4.3.6.1 Sanitary Sewage The ISFSI requires no sanitary sewage system.
4.3.6.2 Chemical Sewage The ISFSI requires no chemical sewage treatment system.
4.3.7 COMMUNICATIONS AND ALARM SYSTEM The ISFSI is totally passive and requires no communication or alarm systems.
4.3.8 FIRE PROTECTION SYSTEM No flammable or combustible substances are stored within the ISFSI or in its immediate vicinity. The ISFSI is constructed of noncombustible heat-resistant materials (concrete and steel). Since no combustible materials are stored within the ISFSI, the ISFSI requires no fire extinguishing system. In the unlikely event of a fire near the HSMs, the existing HBR2 fire extinguishing system will be utilized. This system and procedures for fire extinguishing are described in Section 9.5.1 of the HBR2 Updated FSAR (Reference 4.2).
4.3.9 MAINTENANCE SYSTEMS The ISFSI is totally passive and requires no maintenance other than periodic inspection of the air inlets and outlets and possible removal of debris from the air inlets (see Chapter 10).
4.3.10 COLD CHEMICAL SYSTEMS The ISFSI has no cold chemical system.
4.3.11 AIR SAMPLING SYSTEM No air samples are required at any time during operation. Therefore, the ISFSI has no air sampling system.
4.3-2 Revision No. 22
HBRSEP ISFSI SAR 4.4 DECONTAMINATION SYSTEMS The ISFSI has no decontamination system; however, the existing HBR2 decontamination facility will be used to decontaminate the spent fuel cask.
The decontamination facility has been abandoned in place. Portable equipment is used to aid in the decontamination process.
4.4-1 Revision No. 22
HBRSEP ISFSI SAR 4.5 SHIPPING CASK REPAIR AND MAINTENANCE Existing HBR2 procedures are followed for maintenance and repair of CP&L's GE IF-300 shipping container and its auxiliary systems. Routine inspections of the cask system and components are accomplished in accordance with existing procedures. Where applicable, General Electric's recommended inspection intervals are followed.
Corrective maintenance is planned and accomplished in a controlled manner as needed. Minor repairs can be accomplished onsite with the cask in the decontamination stand using existing HBR2 facilities (see Figure 4.5-1).
However, if major repairs should be required on any large or heavy portion of the cask system, it may require offsite work. Such work can be accomplished for example at General Electric's Morris Operations facility located in Morris, Illinois.
During all phases of repair or maintenance on the cask, existing HBR2 health physics procedures are followed.
These procedures address contamination control and emphasize occupational exposure reduction.
All necessary records generated during the maintenance, repair, or operation of the cask system will be retained in a controlled manner for the time period required by appropriate QA procedures.
4.5-1 Revision No. 22
HBRSEP ISFSI SAR 4.6 CATHODIC PROTECTION The ISFSI is dry and above ground so that cathodic protection in the form of impressed current is not required. The normal operating environment for all metallic components is well above ambient air temperatures so that there is no opportunity for condensation on those surfaces.
The austenitic canister body requires no corrosion protection for any foreseeable event. The DSC support assembly components are Type A36 carbon steel. The top surface of the DSC rail is to be machined and plated with a high-phosphorus electroless nickel finish having a minimum thickness of .001 inch. The remaining surfaces of the rail and other components of the support assembly shall be painted with Carbo-Zinc 11 (Reference 4.3). Consequently, the A36 structural steel is protected against corrosion.
4.6-1 Revision No. 22
HBRSEP ISFSI SAR 4.7 FUEL HANDLING OPERATION SYSTEM The HBR ISFSI is based on the NUHOMS system for storage of irradiated nuclear fuel. The basis and engineering design for the various fuel handling systems used during the operation of the ISFSI are described in the following sections.
4.7.1 STRUCTURAL SPECIFICATIONS The bases and engineering design of HBR2's fuel handling systems are described in Section 9.1.4 of the HBR2 Updated FSAR (Reference 4.2). Other handling and transport operation equipment (cask positioning skid, hydraulic ram, and trailer) are designed to meet the criteria established in Chapter 3 of the NUHOMS Topical Report (Reference 4.2). This equipment is designed to safely perform its intended functions under both normal and off-normal operating conditions. However, this equipment does not impact the safety features of the ISFSI facility and as such is not safety related.
4.7.2 INSTALLATION LAYOUT 4.7.2.1 Building Plans Fuel handling operations will occur within the existing HBR2 Fuel Handling Building (see Figure 4.5-1).
4.7.2.2 Confinement Features The irradiated fuel assemblies will only be handled while in the spent fuel pool or in the confines of the DSC which is placed inside the cavity of the shipping cask or the HSM. The confinement features of the HBR2 spent fuel pool are provided in Section 9.1 of the HBR2 Updated FSAR (Reference 4.2) and the confinement features of the NUHOMS system are discussed in Section 3.3.2 of the NUHOMS Topical Report.
4.7.3 INDIVIDUAL UNIT DESCRIPTION 4.7.3.1 Shipping Cask Preparation During the preparation, the DSC will be placed into the shipping cask cavity. This operation will take place in the HBR2 decontamination facility (see Figure 4.5-1). After loading, the cask annulus and the DSC will be filled with demineralized water and then lifted into the spent fuel pool.
The following components will be used for this operation:
a) Spent Fuel Cask Handling Crane - The spent fuel cask handling crane is used to place the DSC into the shipping cask cavity. The design basis and safety assurance features of the crane are discussed in Section 9.1.4 of the HBR2 Updated FSAR (Reference 4.2).
b) Spent Fuel Cask Lifting Yoke - The spent fuel cask lifting yoke is used in transporting and handling the shipping cask while in the plant and loading the shipping cask onto the transport skid. The design basis and safety features of the lifting yoke are provided in Section 9.1.4 of the HBR2 Updated FSAR (Reference 4.2).
4.7.3.2 Spent Fuel Loading Loading of the IFAs into the DSC takes place in the existing spent fuel pool. The components which are used are described below.
a) Spent Fuel Cask Handling Crane - The spent fuel cask handling crane is used to transport the shipping cask to and from the spent fuel pool. The design basis and safety assurance features of the crane are discussed in Section 9.1.4 of the HBR2 Updated FSAR (Reference 4.2).
4.7-1 Revision No. 22
HBRSEP ISFSI SAR b) Spent Fuel Pit Bridge - The spent fuel pit bridge is used to move the fuel assemblies within the spent fuel pool.
The design basis and safety features are described in Section 9.1.4 of the HBR2 Updated FSAR (Reference 4.2).
4.7.3.3 DSC Drying, Backfilling and Sealing Once the IFAs have been placed into the DSC, the top lead shield plug is placed on the DSC. The cask collar cover plate is then installed. Using the spent fuel cask handling crane, the loaded DSC, sitting in the spent fuel cask cavity, will be removed from the spent fuel pool and moved to the decontamination facility where the cask will be drained. The DSC will be drained, vacuum dried, and backfilled with helium. The top lead shield plug and cover plates will be seal welded to the DSC body. After these operations, the cask will be placed onto the transport skid and taken to the ISFSI site where the DSC will be loaded into the HSM. Throughout all of the above operations, the fuel will be confined within the DSC and the DSC will be seated within the shipping cask. The design basis and safety assurance features of the DSC are discussed in Sections 3.2 and 3.3 of the NUHOMS Topical Report (Reference 4.1). More details on the drying and sealing operations are provided in Chapters 4 and 5 of Reference 4.1.
4.7-2 Revision No. 22
REFERENCES:
CHAPTER 4 4.1 NUTECH Engineers, Inc., "Topical Report for the NUTECH Horizontal Modular Storage System for Irradiated Nuclear Fuel," NUH-001, Revision 1, November 1985.
4.2 Carolina Power and Light Company, "H. B. Robinson Steam Electric Plant Unit No. 2 Updated Final Safety Analysis Report," Docket No. 50-261, License No. DPR-23.
4.3 Carboline Company, St. Louis, MO.
4.R-1 Revision No. 22
. Amendment No. 5 H.B. ROBINSON INDEPENDENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT
I . mi.zs I A:
I ALI
7m I I E
HBRSEP ISFSI SAR CHAPTER 5 OPERATION SYSTEMS Revision No. 22
HBRSEP ISFSI SAR 5.0 OPERATION SYSTEMS 5.1 OPERATION DESCRIPTION The following sections describe the operating procedures, which are unique to the operations of the H. B.
Robinson (HBR) Independent Spent Fuel Storage Installation (ISFSI) such as loading, unloading, and surveillance. Existing fuel and cask handling operations, which are currently being employed at fuel and cask handling operations, which are currently being employed at H.B. Robinson Steam Electric Plant Unit No. 2 (HBR2) will be incorporated with these procedures.
5.1.1 NARRATIVE DESCRIPTION The following steps describe the operating procedures for the ISFSI.
5.1.1.1 Preparation of the Transfer Cask and Canister a) Prior to the start of the operation, the fuel assemblies to be place in dry storage will be visually examined (e.g., by television cameras, binoculars, or other means, etc.) to insure that no visible defects exist and that the assembly structure is intact. The assemblies will also be checked (by analysis or by examination of appropriate records) to verify that they meet the physical, thermal and radiological criteria described in Chapter 3. Measures will be taken to ensure that no known failed fuel will be placed in dry storage. This process will be independently verified to ensure that fuel assemblies meeting the fuel specifications (see chapter 10) are selected for storage.
b) Place the cask in the vertical position in the decontamination facility.
c) Prior to the loading of fuel, the fuel basket used during the transportation operation will be removed from the GE IF-300 shipping cask and the cask cavity will be cleaned or decontaminated as necessary.
d) Using the spent fuel cask handling crane, lower the docking collar (Figure 5.1-1) onto the cask.
Once the docking collar (cask extension) is properly oriented onto the transportation cask, bolt the docking collar into place and tighten.
e) Using the crane, lower the dry shielded canister (DSC) into the cask cavity.
f) Fill the DSC and the cask-canister annulus with clean, demineralized water.
g) Seal the top of the gap between the DSC exterior and cask interior.
h) Place the lid on the cask and lift the cask into the spent fuel pool.
5.1.1.2 Fuel Loading a) Remove the cask lid and lead plug and place the irradiated fuel assemblies (IFAs) in the DSC (which is inside the shipping cask) using the existing HBR2 fuel handling equipment and procedures.
b) When all seven of the IFAs have been loaded into the DSC and the lead plug and cask lid have been secured, the cask will be moved to the decontamination facility.
5.1.1.3 Cask Drying Process a) Place the cask in a vertical position in the decontamination stand.
b) Remove the cask lid.
5.1-1 Revision No. 22
HBRSEP ISFSI SAR c) Lower the water level about 2 in. in the cask canister annulus by removing water from the cask.
Lower the water level in the DSC by removing approximately 15 gallons from the DSC.
d) Seal weld the upper steel cladding plates of the top lead plug to the canister body.
e) Connect a compressed air supply to the vent tube and another hose from the siphon connection to the decontamination facilitys radioactive waste system. Activate the air supply forcing the remaining water out of the DSC cavity. (See figure 4.7-1 of the NUHOMS Topical Report (Reference 5.1) for a schematic of the piping system).
f) Once the water stops flowing from the DSC, remove the siphon hose from the DSC.
g) Connect the vent tube piping system to the intake of the vacuum pump. A hose should be connected from the discharge outlet of the vacuum system to the site radioactive waste system.
h) Start the vacuum system and draw a vacuum of 3 mm Hg within the DSC cavity.
i) Once a vacuum of 3mm Hg has developed in the DSC cavity, disengage the vacuum pump, connect the helium source, and backfill the DSC with 1.5 atm (22 psig) of helium. Verify with pressure gauge that pressure is holding and helium leak test the entire length of the primary end plug closure weld.
j) Seal weld the prefabricated plug over the vent and siphon tube connection.
k) Seal weld the prefabrication plug over the vent tube connection.
l) Perform helium leak test to ensure weld tightness.
5.1.1.4 DSC Sealing Operations a) Place the top cover plant onto the DSC and weld the cover plate to the body of the DSC.
b) Perform dye penetrant test on seal weld.
c) Drain water from annulus and place the cask lid onto the cask and bolt the lid into place.
5.1.1.5 Transportation of the Cask to the Horizontal Storage Module (HSM) a) Use redundant yoke to lift cask out of decontamination pit.
b) Disengage lower portion of redundant yoke and position cask in tilting cradle.
c) Lower cask to horizontal position on trailer and secure.
5.1.1.6 Loading of the Canister into the HSM a) Inspect all air inlets and outlets on the HSM to ensure that they are clear of debris. Inspect all screens on the air inlets and outlets for damage. Replace screens if necessary. Leave the front access of the HSM open.
b) Using the optical alignment system, align the ram with the cask centerline.
c) Position the cask so that the docking collar is within 1 foot of the HSM.
d) Remove the cask collar lid.
5.1-2 Revision No. 22
HBRSEP ISFSI SAR e) Using an optical alignment system and targets on the cask, skid, and HSM as necessary, adjust the position of the cask until the cask us properly positioned with respect to the HSM.
f) Extend the hydraulic ram and activate the grapple to grab the canister.
g) Move the cask against the HSM, so that the docking collar is positioned in the HSM recess.
h) Retract the piston of the hydraulic ram. If the ram fails to retract when the load on the hydraulic system exceeds 6,200 lbs, stop the ram. Check the orientation of the cask with respect to the HSM and reorient the cask if necessary. Continue this step until the DSC contacts the stopping blocks mounted on the top of the DSC support rails at the rear end. These blocks will assure correct axial positioning of the DSC within the HSM.
i) Collapse the grapple arms and withdraw the hydraulic ram.
j) Pull the skid away from the HSM.
k) Install the plate over the front access of the HSM.
l) Insert through the rear access opening the seismic retainer assembly until it couples with the DSC grapple assembly. Bolt down the retainer assembly to the rear access cover plate. Bolt down the cover plate to the rear access embedded plate.
5.1.1.7 Monitoring Operations On a daily basis, site personnel will walk around the perimeter of the HSMs to visually inspect the air inlets and outlets to ensure that they remain unblocked and the integrity of the screens remains intact. Any debris present will be removed. If damage is evident, appropriate remedial action will be taken. The level of action will be responsive to the observation and will be sufficient to ensure safe operation of the facility.
Response to any finding potentially affecting the safe operation of the facility will be appropriate to the finding. If a generic problem is indicated, the remaining modules will be inspected and appropriate remedial action taken. It is anticipated that the many conservatisms and safeguards inherent in this design will ensure safe spent fuel storage over the lifetime of the facility.
5.1.1.8 Unloading the DSC from the HSM a) Using the optical alignment system, align the cask and ram with respect to the HSM.
b) Remove front access cover and rear cover plate and seismic retainer of the HSM and install the ram and grapple.
c) Align the ram with the cask centerline. Extend the ram through the rear access of the HSM until it contacts the DSC.
d) Activate the grapple of the ram.
e) Activate the ram and push the DSC into the cask.
f) Retract the ram piston out of the cask and HSM.
g) Slowly pull the cask forward 1 foot away from the HSM.
h) Place the cask lid onto the cask and bolt the lid into place.
5.1-3 Revision No. 22
HBRSEP ISFSI SAR i) Close the front and rear accesses to the HSM.
The fuel is in a safe configuration in the DSC within the IF-300 shipping cask.
5.1.2 FLOW SHEET A flow sheet for the handling operations is presented in Figure 5.1-3.
5.1.3 IDENTIFICATION OF SUBJECTS FOR SAFETY ANALYSIS 5.1.3.1 Criticality Prevention Criticality is prevented by geometrical separation of the guide sleeves and by boron poison contained in the boral guide sleeves of the canister basket. All DSC baskets will include seven boral guide sleeves.
5.1.3.2 Chemical Safety There are no chemicals used during the operation of the ISFSI that require special precautions.
5.1.3.3 Operation Shutdown Modes The ISFSI is a totally passive system and therefore this section is not applicable.
5.1.3.4 Instrumentation The ISFSI is a totally passive system requiring no instrumentation. However, some of the units were temporarily instrumented for experimental purposes only. The description of the instrumentation is provided in section 4 of this report.
5.1.3.5 Maintenance Techniques The ISFSI is a totally passive system and therefore will not require maintenance. However, to insure that the airflow is not interrupted, the module will be periodically inspected to insure that no debris is in the airflow inlet or outlet. This inspection will be performed daily (see chapter 10).
5.1-4 Revision No. 22
HBRSEP ISFSI SAR 5.2 FUEL HANDLING SYSTEMS 5.2.1 SPENT FUEL HANDLING AND TRANSFER The ISFSI is a modular storage system which provides for the dry storage of irradiated fuel in a horizontal position with natural draft cooling of the dry storage canister. The ISFSI is located within the HBR2 protected area and utilizes HBR2's existing system for handling the irradiated fuel and irradiated fuel cask.
The DSC is designed to be used for transporting the spent fuel to a federal repository and can be removed from the HSM as described in Section 5.1.1.8.
5.2.1.1 Functional Description Figure 5.1-3 presents the flow diagrams for the transfer, loading and retrieval operations. The transfer system is composed of the HBR2 fuel handling system, the GE IF-300 irradiated fuel cask, a transport skid, an optical alignment system, the hydraulic ram, and the HSM T-section guides. Table 5.2-l lists these major systems and their important subsystems.
a) HBR2 Fuel Handling System - The ISFSI is designed to utilize the existing HBR2 fuel handling system. The major components of this system that will be employed during the cask loading operation are the spent fuel pit bridge, the spent fuel cask handling crane, and the spent fuel cask lifting yoke. A description of these components is provided in Section 9.1.4 of the HBR2 Updated FSAR (Reference 5.2).
b) HBR2 Decontamination Facility - The HBR2 cask decontamination facility will be used to decontaminate the irradiated fuel cask. It also provides the location for preparing the DSC.
c) Irradiated Fuel Cask - The General Electric IF-300 Shipping Cask is used to transfer the loaded DSC to and from the HSM. The cask provides shielding along the axial length of the fuel during the transfer, loading, and retrieval operations. A description of the cask's cooling and shielding capabilities is provided in the IF-300 Shipping Cask Safety Analysis Report (Reference 5.3).
The DSC surfaces will be treated with a lubricant that is compatible with the spent fuel pool chemistry. The cask docking collar is a circular ring of steel which is bolted to the top of the cask.
The top six inches of the cask docking collar are seated inside of the HSM walls. The cask, HSM, and docking collar serve as the shield for radiation during the transfer operation.
The cask auxiliary components described above aid in the horizontal transfer of the DSC and were previously described in sections of this document and the NUHOMS Topical Report (Reference 5.1).
d) Cask Positioning Skid - The purpose of the skid is to transport the cask in a horizontal position to the HSM and to maintain the cask in the properly aligned position during the loading and retrieval operations.
e) Optical Alignment System - Once the loaded skid has been positioned at the HSM front access, the cask will be aligned with the HSM. The alignment system consists of a precision transit and targets installed on the cask, skid, and HSM as required. Once the cask is aligned with the HSM, the jack system and cask clamping system will insure that the alignment is maintained throughout the transfer or retrieval operation.
The cask position control system physically moves the cask into precise alignment with the HSM.
It consists of a group of hydraulic jacks to adjust vertical position and a set of hydraulic cylinders to control horizontal position.
f) Ram and Grappling Apparatus - The ram is a telescopic hydraulic cylinder which extends from the back of the HSM through the length of the HSM. The grappling apparatus is mounted on the front 5.2-1 Revision No. 22
HBRSEP ISFSI SAR of the piston. Figures 5.2-l and 5.2-2 show drawings of the hydraulic ram and the grappling apparatus, respectively. The hydraulics for the grappling apparatus are activated and the arms move out between the cover plate and grappling plate. Once the arms are extended, they are locked into position, the ram is retracted, pulling the DSC out of the cask and into the HSM. For retrieval of the DSC, the process is reversed.
g) HSM T-Section Guide - During the transfer operation, the DSC will slide out of the cask and onto the T-section guides, which are within the cavity of the HSM. The T-section guides serve as both the sliding surfaces during the transfer operation as well as supports during storage of the DSC.
5.2.1.2 Safety Features Except for the transfer of the DSC from the cask to the HSM, the loaded DSC will always be seated inside the cask cavity until it is inside the HSM. The safety features of the HBR2 fuel handling systems are described in Section 9.1.4 of the HBR2 Updated FSAR (Reference 5.2). The safety features of the shipping cask are described in the GE IF-300 SAR (Reference 5.3).
To ensure that the minimum amount of force is applied to the DSC during the transfer operation, the surfaces of the DSC may be treated with a solid film lubricant. A low coefficient of friction will minimize the amount of force applied to the DSC, thus minimizing the possibility of damage to the DSC.
The maximum force which may be exerted by the ram is 22,000 lbs. All components of the DSC, ram, grappling assembly and DSC supports are designed to withstand this force. Materials and lubricants are specified so that an operating force of 6200 pounds should easily accelerate the DSC from rest and move it into the HSM. A pressure limitation device in the hydraulic pump will limit the ram force to less than 6200 pounds. The operator can increase the ram force to the 22,000 pound maximum design pressure only by stopping the ram and resetting the limit. Operating procedures will establish the methods for resetting or adjusting the ram speed during travel on extend and retract. It should be noted that it is not expected that a 22,000 pound force will ever be required. However, the system was designed to take such a force if it is ever needed.
5.2.2 SPENT FUEL STORAGE A description of the operations involved in the transfer and retrieval of the DSC to and from the HSM are presented in Section 5.1. During storage, the ISFSI area will be patrolled and the HSM will be visually inspected once per day. The removal of the DSC from storage was described in Section 5.1.1.8 of this document and the NUHOMS Topical Report (Reference 5.1).
5.2.2.1 Safety Features The features, systems and special techniques, which provide for the safe loading and retrieval operations are described in Section 5.2.1.2.
5.2-2 Revision No. 22
HBRSEP ISFSI SAR TABLE 5.2-l TRANSFER SYSTEM COMPONENT LIST HBR2 Fuel Handling System HBR2 Cask Handling System HBR2 Decontamination Facility Irradiated Fuel Shipping Cask Cask Docking Collar Cask Lid Cask Positioning Skid Tilting Cradle Skid Body (with rollers)
Transport Trailer Optical Alignment System Precision Transit Optical Targets Cask Position Control System Hydraulic Jacks Hydraulic Cylinders Control Unit Cask Clamping System Hydraulic Ram Grappling Device Control Unit 5.2-3 Revision No. 22
HBRSEP ISFSI SAR 5.3 OTHER OPERATING SYSTEMS The ISFSI is a totally passive storage system, which requires no additional operating systems other than those systems associated with the loading and retrieval of the DSC.
5.3.1 OPERATING SYSTEM No operating systems are required other than those used in transferring the DSC to and from the HSM.
5.3.2 COMPONENTS/EQUIPMENT SPARES The only component postulated to be damaged during the life of the installation is the air outlet shielding block. As described in Section 8.2.1, a tornado induced missile could damage or knock off the shielding blocks. Consequently, two additional shielding blocks were precast during construction and are maintained as spares at the site. The screens on the air inlets and outlets will be inspected periodically for damage or blockage by debris. If the screens appear to be damaged they will be replaced. Additional or alternate responses to any event affecting the integrity of the screens will be appropriate to the level of damage or disturbance observed.
5.3-1 Revision No. 22
HBRSEP ISFSI SAR 5.4 OPERATION SUPPORT SYSTEM The ISFSI is a self-contained system and requires no instrumentation and control systems to monitor any of the safety-related variables. For research purposes, however, some of the DSCs and the HSMs to be installed at the HBR facility were designed to accept instrumentation. Instrumentation was included as part of an agreement between CP&L, EPRI and the DOE to augment the U.S. data base on LWR fuel rods in dry storage.
The instrumentation of these components was limited to placement of the thermocouples. The DSC thermocouples were connected to an external cable by means of a specially designed feed-through. This feed-through incorporates the same redundant seal philosophy used in the DSC containment design. Details of the feed-through are shown in Figure 5.4-l. After the penetration plug assembly was welded to the bottom of the DSC cover plate, a sleeve was welded over the plug, forming a redundant seal.
Thermocouple sheaths were brazed to the plug assembly at inner and outer surfaces of the penetrations. To ensure that possible leakage through the aluminum oxide insulation is precluded, each end of the sheathed thermocouple is sealed with an environmentally qualified resin.
HSM instrumentation will consist of thermocouples cast in place in the concrete and others attached to the surface and at various locations on the heat shield.
5.4-1 Revision No. 22
HBRSEP ISFSI SAR 5.5 CONTROL ROOM AND/OR CONTROL AREAS The ISFSI requires no control room or control area to safely operate under both normal and off-normal conditions.
5.5-1 Revision No. 22
HBRSEP ISFSI SAR 5.6 ANALYTICAL SAMPLING The ISFSI requires no analytical sampling. Any analytical sampling such as DSC surface contamination levels will utilize the existing HBR2 analytical equipment.
5.6-1 Revision No. 22
REFERENCES:
CHAPTER 5 5.1 NUTECH Engineers, Inc., Topical Report for the NUTECH Horizontal Modular Storage System For Irradiated Nuclear Fuel, NUH-001, Revision 1, November 1985. (Note - currently NUH-001, Rev. 2) 5.2 Carolina Power and Light Company, H. B. Robinson Steam Electric Plant Unit No. 2 Updated Final Safety Analysis Report, Docket No. 50-261, License No. DPR-23.
5.3 General Electric Co., IF-300 Shipping Cask Consolidated Safety Analysis Report, "NEDO-10048-2, Nuclear Fuel and Special Products Division, March 1983. (Note - currently NEDO-10084-5 issued by Duratek) 5.R-1 Revision No. 22
Cask Lid Existing IF-300 BWR Head Cask (Provides a Extension 180 in. cavity)
New Cask Head for Existing BWR Head on site use for off-site use H. B. ROBINSON INDEPENDENT SPENT FUEL STORAGE INSTALLATlON SAFETY ANALYSIS REPORT CASK EXTENSION FOR THE GE IF-300 CASK Figure 5.1-1
THIS FIGUREDELETEDBY AMENDMENT NO. 5 AMENDMENT NO. 5 I H. B. ROBINSON INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT CASK LINER FOR THE GE IF-300 CASK
CASK HANDLING AREA FUEL POOL TRAILER AREA HSM 01 03 LOWER THE CLEAN AND CASK WITH THE LOAD CANISTER CANISTER INTO INTO CASK FUEL POOL 2
FILL CANISTER CAVITY AND CASK/
CANISTER ANNULUS WITH DEMINERALIZED WATER 08 I PLACE LEAD I
PLUG ONTO CANISTER I 1 I
REMOVE I
APPROXIMATELY 4 RAISE CASK TO INCHES OF WATER FROM THE POOL SURFACE CANISTER CAVITY 07 LIFT CASK WITH FILLED DSC FROM SPENT FUEL POOL AND PLACE ON DRYING PAD SHEET 1 OF 4 H. B. ROBINSON INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT HANDLING OPERATIONS FLOW SHEET FIGURE Amendment No.5 5.1-3
CASK HANDLING AREA FUEL POOL TRAILER AREA HSM 0
10 0
17 PERFORM DYE SEAL WELD PENETRANT TEST LEAD PLUG ON TOP COVER PLATE WELD DRAIN AND INSTALL CASK VACUUM DRY CANISTER LID AND BOLT INT0 PLACE BACK FILL DSC LOWER CASK CAVITY WITH HELIUM ONTO SK IO PERFORM MOVE LOAOEO TRAILER TO HSM HELIUM LEAK TEST VENT OPENINGS INSTALL TOP COVER PLATE L 4 I
SHEET 2 OF 4 16 H. B. ROBlNSON INDEPENDENT SPENT FUEL SEAL WELD TOP STORAGE INSTALLATION COVER PLATE TO SAFETY ANALYSIS REPORT CANISTER BODY Amendment No. 5 HANDLING OPERATIONS FLOW SHEET FIGURE 5.1-3
Cask Handling Area Fuel Pool Trailer Area HSM (Loading) 022 028 Inspect HSM Pull DSC into Interior for
- HSM with ram Debris l
- 023 l 0 29
+
Remove Cask Pull Cask out Lid of HSM Opening I
024
- 0 30 Align Cask with HSM Close Front Access Door Opening
, 1 I
025 031 cc Back the Cask Pull ram away into HSM from rear access Opening of HSM 026 w Install Jack Close Rear Support System and Access Door Cask Restraint Extend Ram Through HSM '-I SHEET 3 OF 4 H. B. ROBINSON INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT HANDLING OPERATIONS FLOW SHEET Figure 5.1-3
CASK HANDLING AREA FUEL POOL TRAILER AREA HSM (UNLOADING) 0 1 0 7 ALIGN TRAILER PULL CASK OUT OF HSM WITH HSM FRONT OOOR OPENING CLOSE DOORS ON HSM REMOVE FRONT DOOR AN0 REAR CONCRETE PLUG ALIGN CASK WITH OPENING PLACE CASK LID ONTO CASK MOVE CASK INTO TRANSPORT CASK HSM OPENING TO NEXT LOCATION ON SITE EXTENO RAM THROUGH BACK OF HSM PUSH DSC OUT OF HSM INTO CASK WITH RAM SHEET 4 OF 4 H. B. ROBINSON INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT HANDLING OPERATIONS FLOW SHEET FIGURE Amendment No. 5 5.1-3
.a I
AMENDMENT 1 HYDRAULIC RAM SYSTEM FIGURE 5.2-1
t n
f AMENDMENT 1 DSC GRAFFELING SYSTEM FIGURE 5.2-2
AMENDMENT 1 H. B. ROBINSON lNDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Figure 5.4-1 DSC INSTRUMENT LOCATION
HBRSEP ISFSI SAR CHAPTER 6 WASTE CONFINEMENT AND MANAGEMENT Revision No. 22
HBRSEP ISFSI SAR 6.0 WASTE CONFINEMENT AND MANAGEMENT 6.1 WASTE SOURCES No contaminated wastes are generated during the storage of irradiated fuel in the H. B. Robinson (HBR)
Independent Spent Fuel Storage Installation (ISFSI). The contaminated waste generated as a result of loading the spent fuel into the dry shielded canister (DSC) and into the IF-300 shipping container (see Chapter 5 for discussion of operations) is handled using existing H. B. Robinson Steam Electric Plant Unit No. 2 (HBR2) systems and procedures. The small amounts of contaminated waste which may be generated consist of the spent fuel pool water and air and inert gas which are vented from the DSC and the cask during the drying operation. The handling of spent fuel and loading into the canisters occurs in the existing cask handling area within the Fuel Handling Building of HBR2.
6.1-1 Revision No. 22
HBRSEP ISFSI SAR 6.2 OFFGAS TREATMENT AND VENTILATION Small amounts of slightly contaminated air and inert gas may be generated during loading of the spent fuel into the dry shielded canister and then the cask. Since the spent fuel will be handled as if it is being prepared for offsite shipment using the IF-300, no new procedures are required. The HBR2 ventilation systems will be used to control any gaseous wastes generated. The ventilation systems are described in Chapter 9 of the UFSAR.
6.2-1 Revision No. 22
HBRSEP ISFSI SAR 6.3 LIQUID WASTE TREATMENT AND RETENTION The only liquid waste generated from the ISFSI results from loading the spent fuel assemblies into the dry shielded canister and into the IF-300 shipping container. This liquid waste is handled using existing HBR2 systems and procedures. The liquid waste system is described in Section 11.2 of the UFSAR.
6.3-1 Revision No. 22
HBRSEP ISFSI SAR 6.4 SOLID WASTES The ISFSI is not expected to generate any solid waste. Should any solid waste be generated, however, the existing HBR2 procedures will be used to process and dispose of the waste. Section 11.4 of the UFSAR describes the HBR2 solid waste processing system, the design basis, and an identification of the normal types of solid waste generated by HBR2 operations.
6.4-1 Revision No. 22
HBRSEP ISFSI SAR 6.5 RADIOLOGICAL IMPACT OF NORMAL OPERATIONS -
SUMMARY
The HBR ISFSI is a passive and independent system which provides for the dry storage of irradiated fuel. The ISFSI will not cause a significant radiological impact; the ISFSI has no requirement for radwaste or auxiliary systems for normal operation. Liquid and gaseous effluents utilized in the loading and handling of the Dry Shielded Canisters will be handled using the plant's existing process systems. There will be essentially no solid, liquid, or gaseous wastes generated during normal operation of the ISFSI.
The existing HBR2 health physics program and procedures are followed to maintain exposures ALARA during routine surveillances of the ISFSI.
6.5-1 Revision No. 22
REFERENCE:
CHAPTER 6 6.1 Carolina Power & Light Company, "H. B. Robinson Steam Electric Plant Unit No. 2 Updated Final Safety Analysis Report," Docket No. 50-261, License No. DPR-23.
6.R-1 Revision No. 22
HBRSEP ISFSI SAR CHAPTER 7 RADIATION PROTECTION Revision No. 22
HBRSEP ISFSI SAR 7.0 RADIATION PROTECTION 7.1 ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE ALARA 7.1.1 POLICY CONSIDERATIONS The Carolina Power & Light Company (CP&L) corporate and facility (H. B. Robinson) health physics policies, described in Section 12.1.1 of Reference 7.1, are applicable to the Independent Spent Fuel Storage Installation (ISFSI). Carolina Power & Light Company is committed to a program of keeping occupational radiation exposure as low as reasonably achievable (ALARA). The Company follows the general guidance of Regulatory Guides 1.8, 8.8, 8.10, and publications that deal with ALARA concepts and practices, including Title 10, Code of Federal Regulations, Part 20.
The goals and objectives of the health physics programs are to maintain the annual dose to individual facility personnel to as low as reasonably achievable and to maintain the annual integrated dose to facility personnel; i.e., the sum of annual doses (expressed in man-rem) to all facility personnel, as low as reasonably achievable. The health physics programs identify the organizations participating in the programs, the positions involved, and the responsibilities and functions of the various positions in conducting the programs.
Adequate trained personnel are provided to develop and conduct all necessary health physics programs. The health physics personnel possess the necessary training and expertise to carry out the health physics programs in an efficient manner to assure that Company and regulatory requirements are met.
Appropriate training programs in the fundamentals of radiation protection and facility exposure control procedures are established to provide instructions to all facility personnel including contractors whose duties require working in radiation areas. Training programs for health physics personnel are provided to improve their performance in the health physics programs.
7.1.2 DESIGN CONSIDERATIONS The designs of the dry shielded canister (DSC) and horizontal storage module (HSM) comply with 10 CFR 72 concerning ALARA considerations. Specific considerations that are directed toward ensuring ALARA are:
a) Thick concrete walls on the HSM to reduce the surface dose to an average of less than 20 mrem/hr.
b) Lead shield plugs on the ends of the canister to reduce the dose to workers performing the drying and sealing operations.
c) Use of a shielded transfer cask for handling and transportation operations involving loaded DSCs.
7.1-1 Revision No. 22
HBRSEP ISFSI SAR d) Fuel loading procedures which follow accepted practice and build on existing experience.
e) Recess in HSM front for cask docking to reduce scattered radiation during transfer.
f) Double seal welds on each end of DSC to provide redundant radioactive material containment.
g) Placing demineralized water in the cask annulus and DSC and sealing the DSC-cask gap to minimize contamination of the DSC exterior during loading.
h) Placing external shielding blocks over the HSM air outlets to reduce direct and streaming doses.
i) Passive system design that requires minimum maintenance.
j) Insertion of internal shielding blocks around air inlets to reduce direct and streaming doses.
k) Development of shipping procedures based upon previously used procedures and experience to control contamination during handling and transfer of fuel.
l) Use of additional shielding in front access cover plates.
7.1.3 OPERATIONAL CONSIDERATIONS Operational considerations at HBR2 that promote the ALARA philosophy include the determination of the origins of radiation exposures, the proper training of personnel, the preparation of radiation protection procedures, the development of conditions for implementing these procedures, and the formation of a review system to assess the effectiveness of the ALARA philosophy.
Operational radiation protection objectives deal with access to radiation areas, exposure to personnel, and decontamination. Working at or near highly radioactive components requires planning, special methods, and criteria directed toward keeping occupational radiation exposure ALARA. Job training and debriefing following selected high exposure jobs contribute toward reduced exposures. Decontamination also helps to reduce exposure.
Procedures and techniques are based upon operational criteria and experience that have worked to keep radiation exposure ALARA.
Procedures for the ISFSI were integrated into existing HBR2 procedures and incorporate the same ALARA philosophy.
The ISFSI is considered a radiation control area. Therefore, a fence with a locked gate is located partially around the ISFSI. The HBR2 Radiation Control unit controls the key to this area.
7.1-2 Revision No. 22
HBRSEP ISFSI SAR 7.2 RADIATION SOURCES 7.2.1 CHARACTERIZATION OF SOURCES Table 3.1-2 lists the radiation sources to be stored at the HBR ISFSI. Due to the lower initial enrichment and the longer burnup, the HBR fuel has slightly different neutron and gamma sources compared to the NUTECH Topical Report (Reference 7.2). The neutron source is slightly larger and the gamma source is slightly smaller. However, based on the analysis described in Section 7.3, the shielding of the DSC and the HSM are sufficient to ensure that the dose limits are not exceeded.
7.2.2 AIRBORNE SOURCES Loading the irradiated fuel into the DSC is carried out under water and is identical to the existing HBR2 loading procedures for the IF-300 cask. Airborne sources are processed by the existing fuel building ventilation system.
The sealing and drying of the DSC are performed under procedures to prevent any gaseous leakage. All vent lines (liquid and gaseous) are routed to the existing HBR2 radioactive waste processing systems. Once the DSC is dried and seal welded, there are no design basis accidents which can cause breaching of the DSC and the airborne release of radioactivity. However, to demonstrate ultimate safety, a postulated leak is analyzed in Section 8.2.8 of Reference 7.2.
At the storage location, the only potential for airborne radioactivity is from any potential contamination of the DSC exterior. However, as described in Section 5.1 of the NUHOMS Topical Report, procedures are used to minimize this contamination and should limit any possible DSC surface contamination to below the limits specified in the Technical Specifications. There is essentially no onsite release of radioactivity from the DSC exterior and therefore does not present any health hazard.
7.2-1 Revision No. 22
HBRSEP ISFSI SAR 7.3 RADIATION PROTECTION DESIGN FEATURES 7.3.1 INSTALLATION DESIGN FEATURES The ISFSI is a passive outdoor storage system. Each HSM is capable of providing sufficient ventilation to ensure adequate cooling of the DSC and its contents. The convective cooling system is completely passive and requires no filtration system.
7.3.2 SHIELDING 7.3.2.1 Radiation Shielding Design Features Radiation shielding is an integral part of both the DSC and HSM designs. The features described in this section assure that doses to personnel and the public are ALARA.
The DSC body is a section of 0.5 inch thick, 36 inch inside diameter stainless steel pipe. Two lead-filled end plugs and three steel plates provide shielding at the ends of the DSC. During handling operations, shielding in the radial direction is provided by the IF-300 shipping cask.
Two penetrations in the top lead plug allow water draining, vacuum drying and helium backfilling of the DSC. The penetrations are located away from fuel assemblies and contain sharp, non-coplanar bends to reduce radiation streaming. Table 7.3-1 lists relevant dimensions of the shielding materials present at the ends of the canister.
The HSM provides shielding in both the radial and axial directions during the storage phase. Forty-two inch thick, portland-cement concrete walls and roofs provide the shielding. The module's front access is covered by a two-inch thick composite plate which has additional shielding.
Four penetrations in the module allow convective air cooling of the DSC and module internals. Two identical intake vents at the bottom of the front HSM wall draw air into a shielded box inside the module. The exit vents are placed at both ends of the module roof. Openings to the HSM interior are placed above the end shield regions and not directly over the active fuel region. Sharp duct bends and precast concrete shielding caps over the exhaust exits assure that radiation streaming is reduced to a minimum. Figure 4.2-2 shows details of the module penetrations.
Further details of the radiation shielding design features are presented in Section 7.3.2 of Reference 7.2.
7.3.2.2 Shielding Analysis The shielding analysis methodology for the NUHOMS generic design (Section 7.3 of Reference 7.2) is applicable to the HBR ISFSI as described in this section. However, due to the increased burnup and decreased enrichment of the HBR fuel, the neutron and gamma source terms are slightly different from those used for the generic design. Source terms are 11.4% lower for gamma rays and 17.2% higher for neutrons. This causes the dose rates calculated in Reference 7.2 to require scaling for use in this SAR. Also, because of the revised structure criteria and the necessity of fitting into an existing shipping cask, some steel and lead plate thickness on the DSC have changed. The HSM also has a rear ram access. These design changes required complete analyses of some of the HBR ISFSI shielding.
Figure 7.3-1 shows the locations at which dose rates are presented. Table 7.3-2 shows the resulting surface doses for HBR fuel. HBR ISFSI-unique shielding calculations have been performed for points shown in Figure 7.3-1. Dose rates at other points were scaled from the values in the Topical Report. The following paragraphs provide a brief description of the analysis at each point that was reanalyzed.
7.3-1 Revision No. 22
HBRSEP ISFSI SAR Due to the different design of the bottom shield plug on the DSC the shielding analysis of the HSM front air outlet was redone (points 2.1 and 3.1). This analysis used the DOT code to calculate the neutron and gamma dose rates at point 3.1. Hand calculation was then used to calculate the attenuation by the shield cap of the radiation beam coming out of the air outlet slot. The results are shown in Table 7.3-2.
Points 4, 5, and 7 provide estimates of the dose rate at the front of the module with the door open or closed. The dose rate at point 5 was obtained with the DOT analysis of the front half of the module (same analysis used for point 3.1). Dose rates of point 7 (4.5 feet away from the open door) and point 4 (surface of closed door) were obtained by hand calculations.
Because the dose at point 6 is primarily due to radiation leaving the canister longitudinal surface (and the design for the topical and the HBR canisters are identical for the canister body), the dose rate here was obtained by scaling the topical dose rates by the factors of the source strengths. The dose rates at points 8 and 9.1 were obtained by using the dose rate on the top cover plate surface of the DSC from the DOT analysis and a hand calculation to determine the attenuation due to the 3 1/2 feet distance to the outside HSM surface and the attenuation of the above plate (point 8). The dose rate at points 8 and 9.2 were obtained from a DOT analysis of the rear of the HSM with the canister half way in the HSM (a scoping study was done by hand albedo methods to determine that this was the case for the worse dose at the uncovered rear ram exit.
The dose rates for all points on the DSC top and transfer cask were calculated using the QADMOD (gamma only),
DOT and the MORSE (Monte Carlo) programs. The three different codes were used to assure that the radiation streaming through the cask-canister gap was adequately modeled and estimated. The dose rates from all three
-3 methods yielded similar results (reduction factors of 10 ). The QAD and DOT analysis methodologies were described in the Topical Report. The MORSE calculations are described below.
MORSE, a three-dimensional Monte Carlo shielding code (Reference 7.5), was used to assess the severity of the neutron and gamma ray streaming which occurs when the loaded DSC is inside the transport cask. The shielding calculations were performed in 22 neutron groups and 18 gamma ray groups with a P3 expansion of the angular distributions using a coupled cross section set. Thus, secondary gamma rays are included when primary neutrons are taken as the source.
The MORSE model was constructed in three dimensions using the MARS geometry package. The PICTURE code was used to verify the model. One octant of the cask/canister system was modeled with reflective boundaries at symmetric planes. Fuel assemblies were modeled discretely as two-zone regions containing an outer layer of (homogenized) 0.125 inch thick Boral and a homogenized interior composed of irradiated fuel and cladding. The upper two spacer disks were modeled discretely due to their significant effect of local dose rates in the area of interest. Additional spacer disks were omitted to reduce computational time and ensure conservatism. The lead region in the top shield plug was modeled as a disk with reduced diameter to account for the steel siphon line region.
The cask/canister system was modeled with the nominal 0.25 inch annular gap. A boundary crossing routine was employed to determine the average dose rate in the annular gap containing air only. The choice of boundary crossing, rather than point detector collision estimating, was made in order to allow octant modeling and to improve the statistical deviation.
The source terms were obtained in a fashion similar to the NUHOMS Topical Report and will not be discussed here.
Russian roulette, path length stretching, and source energy biasing were all used to minimize statistical deviation in the area of interest. Russian roulette weighting parameters were established based on the number of mean free paths from significant sources to the cask/canister gap. Path length stretching was in the forward direction. Adjoint XSDRNPM-S (Reference 7.5) calculations were used to determine source energy biasing parameters.
Dose rates were calculated from the fluxes by using the Snyder-Neufeld factors for neutrons and the Henderson factors for gamma rays (Reference 7.8).
7.3-2 Revision No. 22
HBRSEP ISFSI SAR The application of MORSE to the cask/canister streaming problem represents a refinement in the albedo technique used in the NUHOMS Topical Report. Sixty-four thousand, eight hundred neutron histories and 1,920,000 primary gamma ray histories were executed to obtain streaming dose rates. The neutron dose rate (which includes a negligible contribution due to secondary gamma rays) was calculated to a one-sigma deviation of 6.6%. The primary gamma ray dose rate was calculated to a one-sigma deviation of 11.9%
The dose rates reported for the DSC in the cask include numerous combinations of the presence of water in the DSC and DSC-cask gap (the gap hereafter). These combinations are present due to the operational procedures. The various cases are described below.
Point 1.1, lead plug on DSC with water in the DSC and gap is the condition during the welding of the lead plug assembly to the DSC. After welding the DSC will be drained, dried and backfilled with helium. The calculations for point 1.2 reflect the estimates for the dose rate after the DSC is drained. However, it should be noted that personnel will not be required to be directly over the canister during this time.
All operations will be done from the side of the cask and only the forearms and hands of the personnel will be over the cask for a short time during the connecting and disconnecting (with "quick connect" Swagelok fittings) of water, air and He lines. The dose rate in the gap during this initial welding and drying is shown for point 3.1. Point 2 gives the dose rates on the top cover plate during the welding of that plate to the DSC body. The dose rate in the gap is given for point 3.2.
After the top cover plate is welded on the DSC, the cask lid will be lowered into position, bolted on, and then the water drained from the gaps. Point 5 gives the estimated dose rate for the top of cask lid.
Point 4 gives the dose rate at the cask side where the operating personnel will be working.
Point 6 gives the dose rate at the cask collar side during the insertion of the DSC. While the DSC is being seal welded, the lead plug is next to the inner surface of the cask collar and, hence, the dose at the outside of the collar is small. The large dose rate shown for point 6 is only present during the loading of the DSC into the HSM. This dose rate is present over a 6-inch wide ring of the collar which is not inserted inside the HSM cask docking recess.
Although no personnel will be within 20 feet of this area during loading, it would be prudent, from an ALARA standpoint, to use portable lead/ polyethylene shielding to reduce the dose rate in the vicinity of the HSM/cask interface.
7.3.3 VENTILATION The HSM has a ventilation system to provide for natural draft cooling of the DSC. However, no off-gas or filtered treatment system is required due to the low exterior contamination level of the DSC.
The ISFSI is designed for essentially no release of radioactive material during normal storage of the DSC in the HSM. Therefore, additional design work and equipment would not result in a reduction of radioactive materials released. Furthermore, no credible site accident will result in a radioactive leak because of the integrity of the double seal welds at each end of the DSC, the passive nature of the system, the operational limits and controls used during handling and the integrity of the stainless steel body of the canister.
7.3.4 RADIATION MONITORING INSTRUMENTATION The operation of the ISFSI will be monitored under the HBR2 radioactivity monitoring program. No additional radiation monitoring instrumentation is required.
7.3-3 Revision No. 22
HBRSEP ISFSI SAR TABLE 7.3-1 DSC END SHIELDING MATERIAL THICKNESSES DSC Top End Shields Lead Shield Plug 0.50 in. Steel + 3.5 in. Lead + 1.75 in. Steel Cover Plate 1.25 in. Steel DSC Bottom End Shields Pressure Plate 2.00 in. Steel Lead Shield Plug 0.25 in. Steel + 4.75 in. Lead "Top" and "Bottom" refer to the top and bottom ends of the irradiated fuel assemblies 7.3-4 Revision No. 22
HBRSEP ISFSI SAR TABLE 7.3-2 SHIELDING ANALYSIS RESULTS Nominal Surface Dose Rates (mrem/hr)
ISFSI Location Neutron Gamma Total DSC in HSM
- 1. HSM Wall or Roof 0.03 2.5 2.5
- 2. HSM Air Outlet Shielding Cap 0.03 10 10 2.1 Front HSM Shield Cap 0.06 103 103 2.2 Rear HSM Shield Cap 0.06 10 10
- 3. HSM Air Outlet (no shielding cap) 3.1 Front HSM Air Outlet (no shielding cap) 3.0 4450 4450 3.2 Rear HSM Air Outlet (no shielding cap) 1.6 440 440
- 4. Center of Door 35 81 116
- 5. Center of Door Opening 50 378 428
- 6. Center of Air Inlets 27 29 56
- 7. 4.5 Ft. from HSM Door 7.1 17 24
- 8. Ram Opening with Access Plate 2.9 0.37 3.3 (fully inserted DSC)
- 9. Ram Opening Without Access Plate 9.1 Fully Inserted DSC 3.8 0.92 4.7 9.2 Half Inserted DSC 1.2 605 606 DSC in Cask
- 1. Centerline Top of DSC Plug 1.1 No Water in DSC, Water in Gap 296 92 390 1.2 Water in DSC and Gap 0.83 36 37
- 3. Cask/Canister Annular Gap 3.1 Water in Gap and DSC 0.34 390 390 3.2 Water in Gap (no water in DSC) 190 365 555
- 4. Transfer Cask 3.3 2.8 6.1 Side Surface (no water in DSC)
- 5. Transfer Cask Top Cover Surface 194 24 218
- 6. Cask Collar during DSC 37 620 657 Transfer from Cask to HSM Note: These values for worst case situation where no water is present in the gap. For ALARA purposes, water should be present in the gap which will reduce streaming exposures by approximately 1/20.
7.3-5 Revision No. 22
HBRSEP ISFSI SAR 7.4 ESTIMATED ONSITE COLLECTIVE DOSE ASSESSMENT Estimated doses for the fuel loading, drying, sealing and transfer are provided in Section 7.4.1 of the NUHOMS Topical Report (Reference 7.2). Onsite radiation dose rates due to the storage of the fuel in the first three HSMs are shown in Figure 7.4-1. Design dose rates for eight modules being filled will be 8/3 times the values listed in Figure 7.4-1. Based on these design values, the resulting dose for a person located at the ISFSI fence for eight hours a day for 250 days per year would be approximately 10 mrem. For a person located at the offices, the yearly dose would be less than 1 mrem. (Actual dose inside the office is less due to shielding from the building).
7.4.1 OPERATIONAL DOSE ASSESSMENT This section establishes the expected cumulative dose delivered to site personnel during the fuel handling and transfer activities associated with one ISFSI module. Chapter 5 describes in detail the ISFSI operational procedures, a number of which involve radiation exposure to personnel.
A summary of the operational procedures which result in radiation exposure to personnel is given in Table 7.4-1.
The cumulative dose is calculated by a four step process. First, the number of personnel required to perform the task is estimated. Then the time required to perform the task is estimated based on the operational guidelines presented in Chapter 1, engineering judgment, and previous experience with similar or identical operations. An ambient dose rate is obtained for the operation. It is based on an estimated distance from the individual's trunk to the most significant radiation source. The dose rate is conservatively calculated by modeling radiation sources as line, cylinder, or plane sources, as is appropriate to the particular geometry of the operation. With the number of personnel, the time required, and the local dose rate, individual and collective exposures may be calculated.
7.4.2 STORAGE TERM DOSE ASSESSMENT Figure 7.4-2 is a graph of the dose rate versus distance from the face of a filled storage array of three ISFSI modules.
Direct neutron and gamma flux, as well as the air-scattered radiation from the module surfaces, are considered.
Air-scattered dose rates are determined with the computer code SKYSHINE-II (Reference 7.3). Initial loading of all modules with the NUHOMS Topical Report design basis five-year post irradiated fuel is assumed. If five additional modules are added, the dose rate will be 8/3 times that shown in Figure 7.4-2.
Estimates of cumulative doses to site personnel from three filled ISFSI modules are given in Table 7.4-2. The dose rates are based on data shown in Figure 7.4-2. Occupancy information for number of personnel, location, and time is estimated based on the five-year plan for facilities' layout at the HBR2 site. Because of the very rapid decrease of dose rate with distance from the storage facility, a maximum distance of 600 feet was used in these analyses. No credit is taken for shielding of personnel by buildings, or for radioactive decay.
7.4-1 Revision No. 22
SUMMARY
OF ESTIMATED ONSITE DOSES DURING FUEL HANDLING OPERATIONS Average Distance Total From Source Dose Dose Per Personnel Number of Time Surface Rate Person Dose Operation Personnel (Hours) (Feet) (mrem/hr) ( mrem) (mrem)
Location: Fuel Pool Load fuel into canister 2 8.0 - 5 40 80 Bolt lid assembly onto cask 2 0.5 1.5 22.3 11.2 22.3 Location: Cask Handling Area (2)
Decontaminate outer surface of cask 3 8.0 - 6.1 48.8 146.4 Place scaffolding around cask 4 0.8 4.0 9.5 7.6 30.4 Unbolt lid, remove lid and spacer 2 0.75 1.5 22.3 16.73 33.5 Remove approximately 15 gal. of 2 0.5 1.5 22.3 11.2 22.3 water from DSC and lower water level in canister cask gap Weld lead plug to canister 2 3.0 4.0 3.5 10.5 21.0 Hydrotest canister 2 2.0 1.5 22.3 44.6 89.2 (1)
Remove water from canister 2 2.3 4.0 29.8 68.5 137.0 cavity Seal weld prefabricated plug to 2 0.5 1.5 210 105 210 siphon tube connection (1)
Vacuum dry canister and 2 4.0 4.0 29.5 118.0 236 backfill with Helium Helium leak test weld 2 1.0 1.5 210 210 420 Seal weld prefabricated plug to 2 0.5 1.5 210 105 210 vent tube Perform NDE (PT) 1 1.0 1.5 210 210 210 Install end cap 2 0.5 1.5 210 105 210 Weld end cap to canister 2 2.3 4.0 29.5 67.9 135.8 Install cask lid and bolt into place 2 0.5 1.5 118.3 59.2 118.3 7.4-2 Revision No. 22
HBRSEP ISFSI SAR TABLE 7.4-1 (Continued)
Remove scaffolding from around cask 4 0.8 4.0 9.5 7.6 30.4 Transport cask to skid and trailer 2 0.5 - 2 1 2 Location: Trailer (2)
Attach skid tie down to cask 2 0.5 1.0 6.1 3.1 6.1 (2)
Transport cask to HSM 5 0.5 5 6.1 3.1 15.5 Remove cask lid 2 0.5 1.5 118.3 59.2 118.3 (2)
Install cask jacking system and 4 1.5 3.0 6.1 9.3 37.2 align cask with HSM (2)
Transfer canister from cask to HSM 4 0.5 3.0 6.1 3.1 12.4 (2)
Install steel plate over front 2 0.5 3.0 2.5 1.3 2.6 access of HSM Tack weld front access door 2 0.5 1.5 70.2 35.1 70.2 (2)
Install seismic retainer assembly 2 0.5 1.0 4.7 2.4 4.8 (2)
Install cover plate to rear access 2 0.5 1.5 2.5 1.65 3.3 TOTAL 42.95 1366 2635 (1) Monitoring operation - personnel could leave radiation field.
(2) Conservatively assumed to be surface dose rate.
7.4-3 Revision No. 22
HBRSEP ISFSI SAR TABLE 7.4-2 ESTIMATED ANNUAL ONSITE DOSES DURING STORAGE PHASE Average Distance Total From Dose Dose Per Personnel Number of Time Facility Rate Person Dose Area Personnel (Hours/yr) (Feet) (mrem/hr) (mrem/yr) (mrem/yr)
-3 E&RC Building 301 2080 560 1.8 x 10 3.7 110
-3 Operations and Maintenance 2251 2080 360 4.3 x 10 8.9 2000 (O&M) Building
-2 Yard Work Area11 131 2080 120 4.9 x 10 100.0 1300
-3 Yard Work Area2 121 2080 350 4.4 x 10 9.2 110
-3 TSC/EOF Building 251 2080 580 1.6 x 10 3.3 83
-3 Primary Access Point 102 8760 290 7.5 x 10 66.0 660
-3 Bulk Storage Warehouse 251 2080 360 4.3 x 10 8.9 220 Daily Visual Inspection 1 60 4 26 1560.0 1560 HSM Interior Inspection 1 0.5 1.5 4.0 2.0 6045 1
One 8-hour shift per day, 5 days per week.
2 Continuous occupancy (44 people at one 8-hour shift per day, 5 days per week, 50 weeks per year).
7.4-4 Revision No. 22
HBRSEP ISFSI SAR 7.5 HEALTH PHYSICS PROGRAM Appropriate health physics programs are established for all Company operations which deal with radiation. The programs are consistent with the corporate health physics policy and all applicable regulations. The Nuclear Assessment Section periodically evaluates the various health physics programs and other Company activities which have impacts on the programs and reports to senior management regarding the effectiveness and adequacy of the programs. RNP Nuclear Oversight Section makes recommendations to senior management, as necessary, to maintain effective overall health physics programs.
The existing HBR2 health physics program is applicable to the ISFSI. The HBR2 program has been established to provide an effective means of radiation protection for plant personnel, visitors, and the general public. Section 12.5 of the UFSAR describes the HBR2 health physics program.
7.5.1 ORGANIZATION In the HBR2 organization, the Superintendent - Radiation Control reports to the Plant General Manager. This organization allows the Plant General Manager to be involved in the review and approval of specific ALARA goals and objectives as well as review of data and dissemination of information related to the ALARA program.
The organization also provides the ALARA Analyst, who is normally free from routine health physics activities, to implement the plant's ALARA program. This individual is primarily responsible for coordination of plant ALARA activities and routinely interfaces with first line supervision in radiation work planning and post-job review.
7.5.2 EQUIPMENT, INSTRUMENTATION, AND FACILITIES The ISFSI will utilize the equipment, instrumentation, and facilities of HBR2 as necessary. Section 12.5 of the UFSAR describes the HBR2 equipment, instrumentation, and facilities.
7.5.3 PROCEDURES The ISFSI will utilize existing HBR2 health physics procedures. These procedures are discussed in Section 12.5 of the UFSAR. Health physics procedures specific to the ISFSI were incorporated into the existing HBR2 procedures.
In particular, radiation surveys of the ISFSI will be conducted on an annual basis in accordance with plant procedures. Surveys in accordance with Technical Specification 4.2 will also be performed during maintenance on HSMs or when routine surveys identify an increase in radiation levels that warrant further evaluation. Additional health physics/radiation protection procedures required for use during the ISFSI test program will be developed as needed and will comply with the existing HBR2 program.
7.5-1 Revision No. 22
HBRSEP ISFSI SAR 7.6 ESTIMATED OFFSITE COLLECTIVE DOSE ASSESSMENT Because the ISFSI provides containment yielding essentially no radioactive gaseous or liquid effluents, assessment of offsite collective dose is limited to one of direct and reflected radiation to the nearest residence.
7.6.1 EFFLUENT AND ENVIRONMENTAL MONITORING PROGRAM The ISFSI is located within the protected area of HBR2. The HBR2 environmental program is described in the Offsite Dose Calculation Manual, Section 4.0 (Reference 7.4).
7.6.2 ANALYSIS OF MULTIPLE CONTRIBUTION An analysis of multiple contribution was performed in order to determine the radiological impact the ISFSI will impose on the population surrounding the HBR plant. This impact added to contributions made by other uranium cycle facilities were compared to the natural background radiation and the regulatory requirements of 40 CFR 190.
The maximally exposed member of the public would receive approximately 1.6 mrem per year from an ISFSI made up of a three-unit HSM (reference Figure 7.6.1). An ISFSI consisting of an eight-unit HSM would contribute approximately 4.3 mrem per year. This is a result of external radiation only; there are no gaseous, particulate, or liquid effluents associated with the normal operation of the ISFSI. It can be concluded that the actual exposure contribution from the ISFSI along with the total of all other uranium fuel cycle activities is within the regulatory limits set forth in 40CFR190.
7.6.3 ESTIMATED DOSE EQUIVALENTS During the normal operation of the ISFSI, there are essentially no effluents released.
The cumulative dose for the population (due to the ISFSI) integrated over 10 regions out to 50 miles radially from the ISFSI, is less than 2.0 man-rem per year. The currently accepted national average background radiation level is approximately 300 mrem/year per person which includes 200 mrem/year per person for radon.
7.6-1 Revision No. 22
REFERENCES:
CHAPTER 7 7.1 Carolina Power and Light Company, "H. B. Robinson Steam Electric Plant Unit No. 2 Updated Final Safety Analysis Report," Docket No. 50-261, License No. DPR-23.
7.2 NUTECH Engineers, Inc., "Topical Report for the NUTECH Horizontal Modular Storage System for Irradiated Nuclear Fuel," NUH-001, Revision 1, November 1985.
7.3 C. M. Lampley, "The SKYSHINE-II Procedure: Calculation of the Effects of Structure Design on Neutron, Primary Gamma-Ray and Secondary Gamma-Ray Dose Rates in Air," NUREG/CR-0781, RRA-T7901, USNRC, 1979.
7.4 Carolina Power & Light Company, H. B. Robinson Steam Electric Plant, Unit No. 2 Offsite Dose Calculation Manual (ODCM),@ Docket No. 50-261.
7.5 Oak Ridge National Laboratory, "SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation," ORNL/NUREG/CSD-3.
7.6 Oak Ridge National Laboratory, "ANISN-ORNL Multigroup One Dimensional Discrete Ordinates Transport Code with Anisotropic Scattering," RSIC CCC-254, 1973.
7.7 J. H. Price, W.G.M. Blattner, "Utilization Instructions for QADMOD-G," RRA-N7914, RSIC CCC-396, 1979.
7.8 Oak Ridge National Laboratory, "Cask 40 Group Coupled Neutron and Gamma-Ray Cross-Section Data," RSIC DLC-23, 1978.
7.9 Grove Engineering, "MICRO SKYSHINE User's Manual," July 15, 1987, Grove Engineering, Inc.,
Rockville, MD.
7.10 Grove Engineering, "MICROSHIELD User's Manual," Version 3, August 1988, Grove Engineering, Inc., Rockville, MD.
7.R-1 Revision No. 22
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HBRSEP ISFSI SAR CHAPTER 8 ACCIDENT ANALYSES Revision No. 22
HBRSEP ISFSI SAR 8.0 ANALYSIS OF DESIGN EVENTS The purpose of this section is to evaluate the safety of the H. B. Robinson (HBR) Independent Spent Fuel Storage Installation (ISFSI). The safety evaluation is accomplished by analyzing the response of the various components of the ISFSI to normal and off-normal conditions and a range of credible and hypothetical accident conditions.
In accordance with NRC Regulatory Guide 3.48, design events identified by ANSI/ANS 57.7-1981 are used in the safety evaluation of the ISFSI. In ANSI/ANS 57.7-1981, four categories of design events are defined. Design events of the first and second type are addressed in Section 8.1, and design events of the third and fourth type are addressed in Section 8.2 of this report.
Many of the design events in the above four categories have been addressed in the NUTECH Horizontal Modular Storage (NUHOMS) System Topical Report (Reference 8.1) using enveloping criteria. Whenever the site specific load is enveloped by that of the NUHOMS Topical Report, it will be noted and will reference the appropriate section of the Topical Report. Additional site specific analysis which has not been covered in the NUHOMS Topical Report will be discussed in detail in the following sections.
As discussed in Section 3.2 of this Safety Analysis Report (SAR), some design features of the HBR ISFSI are unique and differ from those of the NUHOMS generic concept. In particular, the HSM has a rear access penetration, whereas the generic concept is without any rear access. However, as discussed earlier the methodology of the structural evaluation of the HSM under the above categories of design events as utilized by the referenced report is such that it will conservatively envelop any modular stacking arrangement. Hence, the stress evaluation and the analytical results presented in Chapter 8 of the referenced report for the NUHOMS modules are fully applicable to the site specific HSM.
Some design features of the HBR DSC are also different than those of the NUHOMS generic concept. Specifically the DSC is designed to withstand inertia forces associated with cask drop accidents in which the drop height is significantly higher than the soft drop criteria established earlier in this report. Because of these design features, additional structural evaluation of the DSC is required. The method of analysis, however, for many of the design event cases is the same as the methodology utilized in the NUHOMS Topical Report. For these cases the appropriate sections of the referenced report containing the applicable methodology will be referenced. In other cases where a new methodology is utilized, such as the drop accident case, the analytical approach will be presented.
In either case the resulting DSC stress evaluation will be tabulated and reported throughout this chapter.
The design of the DSC support assembly for the HBR ISFSI is identical to the NUHOMS generic concept and as such the stress evaluation presented in the referenced report is fully applicable to this component.
Since a foundation design was not included in the NUHOMS Topical Report, Section 8.3 is included in this Safety Analysis Report (SAR) to describe the foundation design and analysis using the four categories described above.
As described earlier in this report, two of the DSCs will be instrumented for the purpose of collecting data. Section 8.4 of this report addresses the safety features of the instrument penetration.
8.0-1 Revision No. 22
HBRSEP ISFSI SAR 8.1 NORMAL AND OFF-NORMAL OPERATIONS Design events of the first type consist of a set of events that occur regularly in the course of normal operation of the ISFSI. These events are addressed in Section 8.1.1 of this report. Design events of the second type consist of events that might occur with moderate frequency (on the order of once during any calendar year of operation). These off-normal events are addressed in Section 8.1.2 of this report.
8.1.1 NORMAL OPERATION ANALYSIS The loads associated with the normal operating condition of the ISFSI are as follows: dead weight loads, design basis internal pressure loads, design basis operating temperature loads, operation handling loads, and design basis live loads. The structural components effected by these loads are the dry shielded canister (DSC), DSC internals, horizontal storage module (HSM), DSC support assembly and the foundation. The following paragraphs discuss these loads and compare them to the generic assumptions reported in Section 8.1.1 of the NUHOMS Topical Report (Reference 8.1).
a) Dead Weight Loads - Dead weight analysis contained in Chapter 8 of the NUHOMS Topical Report for the HSM and the DSC support assembly envelops the HBR ISFSI analysis. Hence, the analysis of dead weight in the Topical Report is applicable to the HBR ISFSI analysis for these components.
The DSC component weights are tabulated in Table 8.1-1. The dead weight analysis of the DSC shell is based on the same analytical approach specified in Section 8.1.1.2, Page 8.1-17 of the referenced report. Furthermore, since the total weight of the DSC is approximately the same as that of the NUHOMS generic DSC, the resulting DSC shell stresses are the same. For the dead weight analysis of the spacer disk the results of the finite element analysis reported in Section 8.1.1.3, page 8.1-32 of the referenced report can be directly ratioed for the effect of the weight redistribution and the change in the spacer disk thickness. The site specific spacer disks are 2 inch thick compared to the 1.25 inch of the NUHOMS spacers. Also, the maximum total weight distributed on one spacer is 2034 pounds compared to the 1834 pounds for the NUHOMS. Based on these differences the maximum resulting stress reported in Table 8.1-7 of the referenced report can be ratioed by the relation:
SSP = SNU x (WSP/WNU) x (tNU/ tSP)
Where:
SSP = ksi, the site specific spacer disk membrane stress SNU = 1.58 ksi, the NUHOMS spacer disk membrane stress WSP = 2,034 lb, weight per site specif. spacer disk WNU = 1,823 lb, weight per NUHOMS spacer disk tNU = 1.25 in, NUHOMS spacer disk thickness tSP = 2.0 in, site specific spacer disk thickness Therefore SSP = 1.10 ksi The results of the above analyses are tabulated in Table 8.1-2 of this report. The stresses caused by the weight of other components of the DSC and its internals are insignificant and do not warrant extra analysis.
8.1-1 Revision No. 22
HBRSEP ISFSI SAR b) Design Basis Internal Pressure Loads - The HBR DSCs are operated with 0.0 psig pressure. However, the DSC is designed for 25.0 psig operating pressure at off-normal conditions. This pressure is the same as that specified in the referenced report. Since the HBR DSC has the same shell thickness as the NUHOMS, the resulting primary membrane stress will remain unaffected. However, for the secondary stresses at the discontinuities the analysis reported in Section 8.1.1.2, Pages 8.1-21 through 8.1-24 of the referenced report is reworked to incorporate the change in the effective thickness of the cover plates. The NUHOMS analysis is based on an effective thickness of 1.5 inch, whereas the minimum available cover plate thickness of the site specific DSC is 1.75 inch. With all other conditions and assumptions being identical, the analysis yields a maximum secondary membrane plus bending stress of 7.64 ksi.
For the bending stress on the cover plate itself, the result of the analysis contained in Pages 8.1-24 and 8.1-25 of the referenced report is multiplied by the square of the ratio of the thicknesses. In this manner, the maximum bending stress on the 1.75 inch thick cover plate is 3.27 ksi.
The results of the above pressure analysis of the DSC and comparison against code allowables are contained in Table 8.1-2 of this report. The maximum DSC internal pressure under accident conditions is 39.7 psig, which is the same as that specified in the NUHOMS Topical Report.
c) Design Basis Operating Temperature Loads - The extreme range of ambient temperature at the HBR site is -5oF to 105oF. For the NUHOMS Topical Report design, a range of -40oF to 125oF was assumed. Consequently the thermal analyses of the HSM and the DSC support assembly reported in Sections 8.1.4 through 8.1.5 of the NUHOMS Topical Report conservatively envelopes those of the HBR ISFSI.
The DSC thermal analysis contained in Section 8.1.1.2 of the referenced report conservatively envelopes the site specific DSC thermal analysis. This is due to the fact that the maximum shell bending stress reported in that report is based on the generic assumption that no gaps exist between the spacer disk and the inside cavity of the DSC (Section 8.1.1.2, Pages 8.1-26 through 8.1-28). The HBR DSC, however allows for a nominal radial gap of 0.13 inch. This amount of gap is larger than the differential thermal expansion of the disk. Other thermal stress evaluations of the NUHOMS canister, such as the shell stress evaluation due to temperature variation in circumferential direction, and due to dissimilar material, indicated stresses far below the 20.9 ksi obtained for the case discussed above. Hence, for the sake of conservatism and in order to envelope the actual state of thermal stress in the HBR DSC, the thermal stress obtained from the differential expansion of the spacer disk will be reported herein and is tabulated in Table 8.1.2 of this report.
For the thermal expansion evaluation of the DSC internals, the evaluation reported in Section 8.1.1.3, Pages 8.1-33 and 8.1-34 of the referenced report is also fully applicable. This is due to the fact that the gap existing between the top of the fuel region and the bottom of the HBR DSC lead plug is the same as that reported in the referenced report.
d) Operation Handling Loads - The handling loads on the DSC, DSC support assembly, and the HSM are based on the maximum capacity of the hydraulic ram of 22000 pounds. This capacity is the same as that specified in the NUHOMS Topical Report. Therefore, the handling load analysis of the Topical Report, Section 8.1.1, covers the site specific design. Since the ram mounting plate assembly at the rear access of the HSM is site specific, the loading from the ram on this assembly was investigated. The ram loads are transferred to the wall through the embedded pipe and plate which have welded stud anchors. The 22000 pound loading was found to have a negligible effect on the HSM rear wall. The net effect of the tornado-generated missile impact considered in the topical report is to load the side wall with over 1000 kips. The much narrower end wall, during operational loading, is easily enveloped by the previous analysis. The results of the operational handling load analysis of 22000 pounds are tabulated in Table 8.1-2 of this report.
e) Design Basis Live Loads - The maximum snow load (or other live loads) for the HBR site as derived from the Updated FSAR (Reference 8.2) is bounded by the NUHOMS Topical Report which assumes a live load of 200 psf.
8.1-2 Revision No. 22
HBRSEP ISFSI SAR 8.1.2 OFF-NORMAL OPERATION ANALYSIS This section describes the design basis off-normal events associated with the operation of the HBR ISFSI. The events which are considered here are expected to occur on a moderate frequency.
8.1.2.1 Transport Off-normal events associated with the transport operation of the DSC may occur due to malfunctioning of the auxiliary components (i.e., crane, transporter ram, etc.), or by misalignment of the DSC with respect to the HSM.
Malfunctioning of the auxiliary components does not relate to the safe functioning of the DSC and can be rectified without any impact to the operation of the system. As described in Section 1.3.1.7 of this report the only time the cask crane is operating without the redundant yoke is during the cask lowering on the skid assembly. A postulated malfunction or more specifically a yoke failure during this operation is considered as part of the cask drop accident which is reported in Section 8.2.4 of this report. The DSC and the ram grappling assembly are designed to the maximum ram capacity loading of 22,000 pounds. Hence, any off-normal event such as misalignment or ram malfunctioning will not cause any damage to any component of the ISFSI. The misalignment of the DSC may also cause jamming or binding of the canister casing. The analysis of the DSC under assumed jamming and binding conditions is covered in Section 8.1.2 of the NUHOMS Topical Report (Reference 8.1), and is applicable to HBR ISFSI operation.
All auxiliary components used during the transport operation (i.e., the cask positioning skid, the cask tie-down system, the cradle support, the saddle and the transporter) are designed to withstand the inertia forces associated with transport shock loadings. The DSC and the cask are designed for the postulated drop accident. The inertia forces of a drop accident is significantly greater than the transport shock forces and, hence, inertia forces associated with transportation shock for these components are enveloped by the 8 ft drop accident.
8.1.2.2 Air Flow Blockage Another off-normal event that may occur is the possibility of air inlet blockage. Because the air inlets are close to the ground, there is a chance that they could become blocked with blowing paper, dirt, snow or other debris. Due to the height of the air outlets, their separation and since hot air is blowing out of the exits, it is less likely that both the exits would become blocked. Furthermore, blockage of one exit alone would not be as severe as blockage of both inlets. Therefore, this off-normal event is defined as complete blockage of the HSM inlets. Blockage of all inlets and outlets is considered highly unlikely and is presented in Section 8.2 of this SAR and in Section 8.2 of the NUHOMS Topical Report.
The blockage of the air inlets has been addressed in the NUHOMS Topical Report in Section 8.1.2. The results of this analysis indicate that the rise in temperature and pressure in various components of the storage system is well within the acceptance limits. The blockage of the air inlets would be discovered during the normal surveillance of the modules. As the analysis shows, excessive temperatures are not reached and, hence, if the blockage were to occur just after one inspection and not be discovered until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> later, no threat to the public health and safety would result. Once detected, the air inlets will be cleared of the blockage.
8.1.3 RADIOLOGICAL IMPACT FROM OFF-NORMAL OPERATIONS Based on the off-normal operations described in Section 8.1.2, there is no additional radiological impact from the ISFSI beyond what is described in Chapter 7.
8.1.3 Revision No. 22
HBRSEP ISFSI SAR TABLE 8.1-1 DRY SHIELDED CANISTER AND HORIZONTAL STORAGE MODULE COMPONENT WEIGHTS COMPONENT DESCRIPTION CALCULATED WEIGHT (Pounds)
- 1. Dry Shielded Canister:
Casing 2849 Top Grapple Assembly 49 Top Cover Plate 357 Top Lead Casing 617 Top Lead Plug 1307 Top Ring Plate 60 Bottom Cover Plate 605 Bottom Lead Casing 152 Bottom Lead Plug 1555 Total 7551
- 2. Canister Internals:
Spacer Disks 1591 4 x 2 1/2" 0/ Support Rods 897 7 x Boral Tubes 918 Total 3406
- 3. 15 x 15 PWR Spent Fuel Assembly 9975 Total Three Loaded Canisters Weight 62796
- 4. 3 Canister Support Assemblies 4725
- 5. 3-Bay Reinforced Concrete Module 800770
- 6. 3 x 2" Steel Door 12528
- 7. 6 x Shielding Blocks 10556 Total (3 Bay HSM Weight Loaded) 892236 8.1-4 Revision No. 22
HBRSEP ISFSI SAR TABLE 8.1-2 MAXIMUM DRY STORAGE CANISTER SHELL STRESSES FOR NORMAL OPERATING LOADS LOAD TYPE STRESS (ksi) (1)
DCS ASME COMPONENTS DESIGN DESIGN OPERATION CODE DEAD BASIS BASIS HANDLING ALLOWABLES STRESS WEIGHT PRESSURE TEMPERATURE (3) (ksi) (2)
TYPE Primary 0.21 0.91 7.4 0.38 18.7 Membrane Local Canister Primary N/A 1.38 7.4 N/A 28.05 Shell Membrane Primary Membrane +
11.55 7.64 20.90 10.99 56.10 Secondary Bending Primary N/A N/A N/A N/A 18.70 Membrane Cover Plate Primary Membrane + N/A 3.27 0.45 13.36 28.05 Bending Spacer Primary 1.10 N/A N/A N/A 18.70 Disk Membrane Notes:
- 1. Values shown are maximums irrespective of location.
- 2. Allowable stresses are conservatively taken at 400°F.
- 3. Values are based on ram capacity load of 22,000 lb.
8.1-5 Revision No. 22
HBRSEP ISFSI SAR 8.2 ACCIDENT ANALYSIS This section addresses design events of the third and fourth types specified by ANSI/ANS 57.7-1981 and any other credible accident that could affect safe operation of the HBR ISFSI. The postulated accidents are:
o Loss of air outlet shielding blocks o Tornado and tornado generated missiles o Earthquake o Eight foot drop o Lightning o Blockage of air inlets and outlets o Accident pressurization of the DSC o Fire o Leakage of the DSC o Load Combination o Train Derailment In the following paragraphs, the accident analyses for various components of the ISFSI are described. When the accident loads or conditions are the same as (or enveloped by) those addressed in the NUHOMS Topical Report (Reference 8.1), reference will be made to the appropriate section of that report.
8.2.1 LOSS OF AIR OUTLET SHIELDING This postulated accident assumes the loss of both air outlet shielding blocks from the top of the horizontal storage module. All other components of the ISFSI are assumed to be in normal condition. The air outlet shielding blocks are designed to remain in place and remain completely functional for all postulated accidents except tornado generated missiles. There are no structural or thermal consequences to the ISFSI as a result of the loss of the shielding blocks; however, there are radiological consequences which have been addressed and analyzed in the NUHOMS Topical Report, Section 8.2.1. The resulting increase in air scattered (sky shine) doses or direct radiation as reported in the Topical Report are within 10CFR100 dose limits.
Recovery To recover from a lost or damaged shielding block caused by a tornado projectile, one of the spare blocks is transferred to the HSM. After the shield block is transferred to the HSM, a crane is used to lift the block into position. The block is then bolted in place. The entire remounting operation should take less than 30 minutes, during which a mechanic will be on the HSM roof for approximately 15 minutes. During this time he will receive less than 50 mrem. The dose to the crane operator and the mechanic on the ground while putting the shield block in place will be approximately 20 mrem each (assuming an average distance of 15 ft from the center of the module roof). Note: The times listed are only to provide estimates of radiation dose to workers. There are no commitments to ensure the shield blocks would be replaced within 30 minutes.
8.2.2 TORNADO/TORNADO GENERATED MISSILE The most severe tornado wind loadings as specified by NUREG 0800 (Reference 8.3) and NRC Regulatory Guide 1.76 (1974) are selected as a design basis for this accident condition. The applicable design parameters of the design basis tornado (DBT) are the same as those specified in Section 3.2.1 and 3.2.2 of the NUHOMS Topical Report.
The accident analysis of the ISFSI under the DBT is covered by the analysis presented in Section 8.2.2 of the referenced report. Given the fact that the HSM method of structural analysis as utilized by the referenced report conservatively envelopes any stacking arrangement of the modules, including the three modular concept at the HBR ISFSI, the maximum moment and shear for the design basis wind pressure and missiles are also enveloped by the values given in Table 8.2-3 of this referenced report. Furthermore, the walls of the horizontal storage modules are anchored into the concrete foundation and as such, there is no possibility of overturning or sliding of the 8.2-1 Revision No. 22
HBRSEP ISFSI SAR modules due to the impact of a massive high kinetic energy missile. The uplift forces generated by the impact of the massive missile and tornado wind loads are included in the foundation design presented in Section 8.3 of this report.
The design and analysis of the anchorage system is also presented in Section 8.3.
The result of this accident analysis indicates that all components of the ISFSI are capable of withstanding the tornado wind loads and tornado generated missiles with the exception of the air outlet shielding blocks. The loss of the shielding blocks is addressed in Section 8.2.1 of this report.
8.2.3 EARTHQUAKE 8.2.3.1 Accident Analysis As specified in Section 3.2 of this report, the maximum ground horizontal acceleration is 0.20g and the maximum ground vertical acceleration is 0.133g. The NUHOMS Topical Report assumes a value of 0.25g for maximum horizontal acceleration and 0.17g for maximum vertical acceleration. In the Topical Report, for the seismic stress analysis of various components, a multiplier of 2 is used to account for multimode excitations. Since the values of the vertical and horizontal acceleration of the referenced report are higher than the HBR site accelerations, the seismic analysis for the HSM and the DSC support assembly presented in Section 8.2.3 of this referenced report is fully applicable and the results of these analyses envelop the site specific design. To establish the actual seismic response of the HBR DSC additional analysis is performed. However, the methodology is the same as that reported in Section 8.2.3.2, Pages 8.2-15 through 8.2-19 of the referenced report. Since the site specific design is different than that of the Topical Report, the longitudinal horizontal seismic loading on the HSM was reviewed. First of all, the DSC loads are transferred to the seismic retainer during a seismic event. The retainer is connected to the ram mounting assembly plate through tiedown bolts in the two inch cover plate. Consequently the loading is transferred to the embedded pipe and plate which are anchored into the HSM rear wall. The maximum loading of 8800 pounds generated by this event was found to have a negligible effect on the HSM rear wall.
In the referenced report, Section 8.2.3.2, the DSC shell ovaling mode was found to yield the lowest natural frequency. Since the HBR DSC shell parameters (i.e., the thickness and nominal diameter) have not been changed the lowest natural frequency remains the same at 37.2 Hz. The stresses induced on the canister casing and the basket due to the 0.20g horizontal and 0.133g vertical seismic accelerations are calculated on the basis of equivalent static method. The static stresses obtained are increased by a factor of 2.0 to account for multimode excitation. To obtain the DSC stresses due to the vertical component of the seismic load, the bending stresses calculated for the dead weight analysis can be factored directly by 0.266. The maximum stress obtained in this manner is 3.07 ksi. For the horizontal seismic analysis both the longitudinal and the transverse directions are considered. For the horizontal acceleration in the transverse direction, the method of analysis presented in Page 8.2-16 of the referenced report was employed and a bending stress intensity of 7.5 ksi was obtained. The stresses in the DSC shell and outer top plate due to the restraining action of the seismic restraint assembly under the longitudinal seismic loading was also investigated and found to have negligible effect. The shell stresses obtained for the vertical and horizontal cases were summed absolutely and a combined stress of 9.32 ksi was obtained.
Additionally, using the same methodology as that presented in Section 8.2.3.2, Page 8.2-17 of the referenced report, a margin of safety against a DSC roll over during a seismic event was established. A value of 2.5 was obtained for this margin of safety against the DSC roll over.
In summary, the ISFSI seismic analysis using site specific accelerations is enveloped by that reported in the NUHOMS Topical Report. Furthermore, the HSMs are anchored to the foundation and as such, no overturning or sliding of the modules is possible. However, the overturning effects on the foundation are included in the foundation design which is presented in Section 8.3 of this report. The anchorage design is also presented in Section 8.3 of this report.
8.2.3.2 Accident Dose Calculation The major components of the HBR ISFSI are designed and analyzed to withstand the forces generated by the safe shutdown earthquake; hence there are no dose consequences.
8.2-2 Revision No. 22
HBRSEP ISFSI SAR 8.2.4 DROP ACCIDENT 8.2.4.1 Postulated Cause of Events As described in Section 1.3.1.7 of this report, the only time during the transfer operation that the IF-300 cask is operating without its redundant yoke is during the cask lowering into the cradle of the skid assembly. As shown in Figure 8.2-1 the maximum height that the cask is raised during this operation is 8.0 feet. Hence, the maximum height of a postulated drop accident is limited to this value. Furthermore, since the cask is always lifted from the trunnions located at the upper regions of the cask, the postulate failure of the single yoke can only cause a cask bottom end or a corner drop. Consequently, if the yoke fails during the tilting operation the cask will either land on the bottom end fins or on the side steel rings located near the upper and lower regions of the cask outer shell.
Based on the above discussion, an 8-foot drop criteria in either horizontal or vertical bottom end orientation will bound any possible drop orientation during the transfer operation, including a corner drop orientation. The skid assembly and the cask/skid/trailer tie down systems are designed to withstand the inertia forces associated with the transportation shock loads, and as such there is no possibility of a cask drop during the transport operation from the decon area to the HSM site. Even if such unlikely event occurs or the cask/skid/trailer tip over as a unit, the height of this drop condition is enveloped by the 8 foot drop height criteria.
8.2.4.2 Drop Accident Analysis As stated earlier in Section 1.3.1.3 of this report, the IF-300 cask requires an additional extension collar and a new cask lid, in order to meet the cask cavity minimum length requirement and meet the criteria for cask lid removal in horizontal orientation. In this modified configuration the cask's impact limiters which are the radial fins attached to the cask's original head are removed. Hence, the energy absorbing properties of the cask is significantly reduced at upper regions. However, as discussed earlier the cask is always handled in upright position and no postulated failure mechanism can produce a top end drop. Additionally, the 8 feet rise of the cask is not sufficient for the cask to rotate 180 degrees in mid air to land on its head or upper corner. The remaining part of the cask impact limiters, i.e., the bottom radial fins and the ring and both ends are not altered and will provide the energy absorption mechanism needed for the vertical bottom end and the horizontal drop.
The IF-300 cask energy absorbing properties are contained in the cask Safety Analysis Report (Reference 8.4). This SAR contains extensive data concerning a 30-foot drop accident.
The latest deceleration time history development work of the IF-300 cask is contained in Appendix V-1 of the above referenced document. These particular impact time histories contain peak deceleration values, at early time of impact. These peak acceleration values are associated with the dynamic yield stress characteristic of the stainless steel fins (strain rate dependency). These time histories which envelop the previous histories reported in the referenced document, include 3 horizontal and 2 vertical drop orientations. These selected time histories were modified to reflect the 8 ft drop criteria described earlier. Since the overall geometry and the weight of the loaded cask are not significantly changed, these deceleration time histories were linearly scaled to reflect the 8 ft drop criteria. Figure 8.2-2 shows the modified deceleration time histories used in the DSC drop analyses.
Horizontal Drop Principle structures effected by the horizontal drop are the spacer disk and the boral tubes. The boral tubes serve only as a guide for the fuel assemblies and are not considered load bearing members, except for their own weight. In the NUHOMS Topical Report, Section 8.2.9, Page 8.2-35, the stresses in the boral tube under the inertia forces of a 34g drop criteria were evaluated by a finite analysis technique. Since the boral tube design is not changed, the result of this analysis can be directly ratioed for the higher deceleration value of 54.4g. In this manner a maximum stress of 4.24 ksi is obtained.
The DSC basket is designed such that the locations of the spacer disk coincide with the fuel assembly grid strap.
Therefore the weight of the fuel assemblies is directly transmitted to the disk. For the analysis of this 8.2-3 Revision No. 22
HBRSEP ISFSI SAR member, the finite elements analysis reported in Section 8.2.4, pages 8.2-31 through 8.2-34 of the referenced report can be utilized directly. This is due to the fact that the overall configuration of this member has not been changed from that of the NUHOMS generic concept, with the exception of thicker disks, and the analysis is linear elastic.
Consequently, the results of the referenced analysis can be factored to include the effect of the mass, thickness, length and deceleration value changes. Additionally, a factor was added to include the additional weight of the support rods in relation to the mass used in the STARDYNE Model.
Ssp = Snu (Msp/Mnu)(tnu/tsp)(gsp/gnu)(Ia/Iu)(Mnsr/Mm) where:
Ssp = Maximum Stress, ksi Snu = 38.52 ksi (from NUHOMS)
Msp = 2034 lb Mass, CP&L without rods Mnu = 1818 lb Mass, NUHOMS without rods gsp = 54.4g, site deceleration value gnu = 34g, NUHOMS drop value Ia = 26.19 in, actual cell length Iu = 26.00 in, NUHOMS length Mnsr = 1962.1 lbs Mass, NUHOMS with lids Mm = 2133.1 lbs Mass, STARDYNE Model therefore:
Ssp = 39.93 ksi It must be noted that this stress intensity is mainly due to the shear force developed near the imposed artificial support boundary, and as such is not representative of the actual stress of the disk. A more critical stress location of the disk is at the spacer beams adjacent to the fuel assemblies. The maximum membrane stress intensities is 23.5 ksi which is obtained by the same ratioing technique discussed above. The results of the horizontal drop analysis are contained in Table 8.2-1 of this report.
Vertical Bottom End Drop The components of the DSC that are critically effected during a vertical bottom end drop are the DSC shell, the top and bottom DSC regions, the support rods and related DSC welds. The vertical drop analyses utilize both hand calculations and finite element technique. For the DSC shell and the end regions ANSYS program was employed for its axisymmetric and linear or nonlinear features. Other components of the DSC are analyzed by hand calculation techniques.
As stated earlier in this report, the HBR DSC configuration is different from that of the NUHOMS generic concept.
The DSC has been redesigned to fit into the IF300 cask, and also is designed to withstand a drop accident in which the height of the drop is significantly greater than the 8-foot criteria. This is done for compatibility with future shipping options.
For the DSC bottom region analysis a model consisting of 131 elements and 183 nodes was developed. The model is shown in Figure 8.2-3. Both lead and steel are modeled as 2-D, 4-node isoparametric axisymmetric finite elements (STIF42). The interface of the lead and steel is modeled with coincident nodes which are coupled in vertical direction only. In this manner, only normal forces are transmitted between the two surfaces, and the shear and friction forces are conservatively released. Both physical and symmetrical boundary conditions are imposed at appropriate locations. The material properties are conservatively taken at 400oF to envelop peak temperatures of the 8.2-4 Revision No. 22
HBRSEP ISFSI SAR DSC shell. The entire weight of the basket and the fuel assemblies are included as added mass elements along the top surface of the 2 inch cover plate. The weight of the top region of the DSC and that portion of the DSC shell that is not included in the model, is also included as added mass at the appropriate location on the shell. The response of the DSC bottom region under the drop impact time history was conservatively approximated by an equivalent static analysis. The impact time history has a very short duration and essentially behaves like a very short triangular impulse. Frequency analyses performed on various DSC components indicated that the longest natural period was much greater than the duration of the impulse. Thus, the dynamic impact loads cannot produce a response that exceeds the static response, and as such the dynamic amplification factor is less than unity. Therefore, the static analysis performed is more conservative than a dynamic analysis. An acceleration value of 76.5g was imposed statically on the model. Both membrane and extreme fiber stress intensities at critical locations including the weld elements are reported in Table 8.2-1 of this report.
For the top region of the DSC, another ANSYS finite element model as shown in Figure 8.2-4 was developed. This model consists of 298 isoparametric STIF42 elements and 409 nodes. Similar assumptions and modeling technique as discussed for the bottom region model were employed. Static acceleration of 76.5g was applied to the model to obtain the membrane and bending stresses for various components and welds. The results of this analysis are also included in Table 8.2-1 of this report along with the ASME code allowables.
The 2 1/2 inch diameter steel support rods were analyzed under the postulated vertical drop. These rods extend the entire length inside cavity of the DSC. The main function of these rods is to provide resistance to axial loads for the spacer disks.
Each of the seven spacer disks is welded to these rods by means of fillet welds. One inch clearance is provided between the support rods and the top lead plug of the DSC. This clearance is provided so that thermal expansion of the components and deflection of components during accident loading conditions, such as a drop accident, will not cause interference.
The 2 1/2 inch diameter support rods are designed so that they will resist the weight of the spacer disks under the postulated drop. The most critical segment of the support rod is between the two bottom spacer disks. For this analysis, the weight imposed on a single rod at this critical location was the weight of six spacer disks divided by 4 plus the self weight of 1 rod. The axial stress at this 26 inch segment of the rod was found from the following relationship:
Smx = (W x a) / A where: Smx = Axial stress W = 651.7 lb, total weight imposed on the rod a = 76.5g, peak vertical deceleration A = 4.91 in2, cross sectional area of the rod therefore:
Smx = 10.15 ksi The support rod material has been changed from SA304 stainless steel to SA-479 Type XM19 material which has a yield stress of 40.8 ksi at 400°F. The allowable compressive stress for this material is established by the rules of the ASME Appendix XVII, and Appendix F, which include the effect of slenderness ratio.
The results of the support rod analysis along with the compressive stress allowable are tabulated in Table 8.2-1 of this report.
The results of the horizontal and vertical drop analysis as shown in Table 8.2-1 indicate that the stresses in all components of the DSC and its internals are within the ASME acceptance limits and are capable to withstand inertia forces associated with the 8 foot drop accident condition. A corner drop accident was also considered. However the deceleration values as established by the IF-300 cask SAR are significantly lower than the values of either the 8.2-5 Revision No. 22
HBRSEP ISFSI SAR horizontal or the vertical deceleration components. Therefore the stresses for corner drop analysis are bounded by the analyses presented above.
8.2.5 LIGHTNING 8.2.5.1 Postulated Cause of Events Since the ISFSI is outdoors, there is a likelihood that lightning could strike the ISFSI. Section 2.3.1 of the HBR2 Updated FSAR (Reference 8.2) provides information on the frequency of cloud-to-ground lightning strokes (the only type of lightning stroke which poses a hazard to the ISFSI) at the site. In order to protect the ISFSI from any damage which could be caused by a lightning discharge, a lightning protection system is installed on the ISFSI. The lightning protection system is designed in accordance with NFPA No. 78-1979 Lightning Protection Code (Reference 8.5). This system will prevent any damage to the HSM and its internals. Therefore, lightning striking the HSM and causing an off-normal condition is not a credible accident.
8.2.5.2 Analysis of Effects and Consequences Lightning protection systems have proven to be an effective means of protecting a structure and its contents from the effects of a lightning discharge. The lightning protection system does not prevent the occurrence of a lightning discharge; however, the system does intercept the lightning discharge before it can strike the HSM and provides a continuous path for the discharge to the earth. In the event of lightning striking the HSM, the air terminal located on the HSM roof slab would intercept the lightning discharge. The current will follow the low impedance path of the air terminal, conductors, and ground terminals to the earth. Since the system diverts the current, the HSM and its contents will not be damaged by the heat or mechanical forces generated by the current passing through the HSM. In addition, since the ISFSI requires no electrical system for its continuous operation, the resulting current discharge will have no effect on the operation of the ISFSI.
8.2.6 BLOCKAGE OF AIR INLETS AND OUTLETS This accident is the complete and total blockage of the air inlets and outlets of the horizontal storage module. Since the ISFSI is located outdoors, it can be postulated that the module is totally covered by debris from such an unlikely event as a tornado. The ISFSI's design features, such as a perimeter fence and separation of air inlets and outlets, minimize the probability of such an accident occurring under normal conditions. Nevertheless, such an accident is postulated and analyzed.
There are no structural consequences under this event. The thermal consequence of this accident results from heating of the DSC and HSM due to the blockage of air flow. Section 8.2.7 of the NUHOMS Topical Report addresses this accident condition. The results of the analysis indicate that there is no structural or dose consequence if the air inlets and outlets are cleared within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. This 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> time limit for clearing the air inlets and outlets is specified in the HBR2 ISFSI operation and limits criteria (See Chapter 10).
8.2.7 ACCIDENT PRESSURIZATION OF DSC Internal pressurization of the DSC results from fuel cladding failure and the subsequent release of fuel rod fill gas and free fission gas. To establish the maximum accident pressurization, it is assumed that all fuel rods in the DSC are ruptured and that the fission gas release fraction is 25%, and the original fuel rod fill pressure is 500 psig. (HBR fuel actually has a fill pressure of 300 psig.) The resulting internal pressures at HBR's maximum ambient temperature of 105oF and at the minimum ambient temperature of -5oF are below the accident pressures reported in Section 8.2.9 of the NUHOMS Topical Report (for temperature extremes of 125oF and -40oF). The limiting accident for canister pressurization is the blockage of air flow to the DSC. Under these conditions, the gas temperatures in the DSC will rise to 413oC (775oF) producing a DSC internal gauge pressure of 2.76 bar (39.7 psig). The canister shell stresses due to accident pressurization are enveloped by those reported in the Topical Report.
The DSC has a safety margin of greater than 3 under this accident condition and as such, there are no dose consequences.
8.2-6 Revision No. 22
HBRSEP ISFSI SAR 8.2.8 FIRE No flammable or combustible substances are stored within the ISFSI or within the ISFSI's radiation control area.
Additionally, the ISFSI is constructed of non-flammable heat-resistant materials (concrete and steel). The only credible accident which could expose the ISFSI to a flammable substance would be the accidental spillage of a flammable liquid, either through human error or equipment malfunction, at the perimeter of the ISFSI. However, the sandy soil between the sides of the ISFSI's perimeter fence and the HSM, is highly porous. Most of the flammable liquid would be absorbed by the soil, greatly reducing the intensity or duration of the fire.
The only other time in which a component of the ISFSI would be exposed to a potential fire hazard would be during the DSC drying and transport operations. Throughout these operations, the DSC is located within the cavity of the GE IF-300 shipping cask.
Based on the above discussion, exposure of the ISFSI to a long or intense fire is not considered a credible accident.
8.2.9 DRY STORAGE CANISTER LEAKAGE The DSC is designed for no leakage under any normal or credible accident conditions. The accident analyses in previous sections show that none of the events could breach the canister body. However, to show the ultimate safety of the ISFSI, a total and instantaneous leak was postulated. The postulated accident assumes that one DSC ruptured and all fuel rod claddings failed simultaneously such that 25% of all fission gases in the irradiated fuel assemblies (mainly Kr-85) are instantaneously released to the atmosphere. The dose consequences from the leaking DSC are evaluated in the NUHOMS Topical Report, Section 8.2.8, and the resulting accident dose is found to be well below the 10 CFR Part 72.68 acceptable limit of 5.0 rem.
8.2.10 LOAD COMBINATION Normal operating and postulated accident loads associated with various components of the ISFSI are either the same as or are enveloped by those reported in the NUHOMS Topical Report, except for the DSC and the foundation.
Hence, the combined effect of various accident and normal operating loads for the DSC support assembly and the HSM are enveloped by the load combination results presented in Section 8.2.10 of the Topical Report. The methodology used in combining normal operating and accident loads and their associated over load factors for various components of the ISFSI, with the exception of the foundation, is presented in the aforementioned report.
Load combination procedures for the foundation are addressed in Section 8.3 of this report. The DSC analysis load combination utilizes the same methodology as in the Topical Report, but due to design differences the results are changed slightly. The results of the DSC load combination for the worst case, i.e., drop accident, are contained in Table 8.2-2 of this report. Furthermore, the DSC fatigue analysis due to normal operating pressure loads, accident pressure loads, seismic loads, seasonal temperature loads, and daily temperature cycling as presented in Section 8.2.10 of the Topical Report, envelops the HBR site specific analysis. This is because the extreme ambient temperature selected for generic design of the DSC (-40oF to 125oF) envelops the HBR ambient temperature range
(-5oF to 105oF) and the HBR has a lower seismic acceleration.
8.2-7 Revision No. 22
HBRSEP ISFSI SAR Table 8.2-1 MAXIMUM DSC STRESSES FOR 8-FOOT BOTTOM END DROP ACCIDENT DSC STRESS STRESS (ksi)
COMPONENTS(2) TYPE CALCULATED ALLOWABLE(1)
Primary Membrane 10.11 44.88 Canister Shell Primary Membrane 16.56 64.40
+ Bending Bottom Primary Membrane 6.09 44.88 Cover Primary Membrane Plate 13.40 64.40
+ Bending Top Primary Membrane 2.77 44.88 Cover Primary Membrane Plates 5.47 64.40
+ Bending Support Ring Primary Membrane 1.71 44.88 For Top Lead Plug Primary Membrane 5.43 64.40
+ Bending Lead Compressive 1.65 6.80 Lead Casing Primary Membrane 17.23 44.88 Spacer Disk Primary Membrane 39.93 44.88 Primary Membrane Boral Tubes 4.24 64.40
+ Bending 2 1/2 in. diam.
Compression 10.16 21.13(3)
Support Rods 1/4 in. Fillet Primary 11.64 22.44 Weld J-Weld Primary 6.16 29.20 Notes:
- 1. Allowable stresses shown correspond to service Level D limits, unless noted otherwise.
- 2. Material properties taken at 400°F design temperature.
- 3. Compressive stress allowable of the support rods is based on Appendices XVII and F rules and for Level A limits.
8.2-8 Revision No. 22
HBRSEP ISFSI SAR Table 8.2-2 DSC ENVELOPING LOAD COMBINATION(1)
DSC STRESS STRESS (ksi)
COMPONENTS(4) TYPE(5) COMBINED(2) ALLOWABLE(3)
Primary Membrane 11.23 44.88 Canister Shell Primary Membrane 35.75 64.40
+ Bending Bottom Primary Membrane 6.09 44.88 Cover Primary Membrane Plate 15.91 64.40
+ Bending Top Primary Membrane 2.77 44.88 Cover Primary Membrane Plates 8.74 64.40
+ Bending Support Ring Primary Membrane 1.71 44.88 For Top Lead Plug Primary Membrane 5.44 64.40
+ Bending Lead Compressive 1.86 6.80 Lead Casing Primary Membrane 17.29 44.88 Spacer Disk Primary Membrane 41.03 44.88 Primary Membrane Boral Tubes 4.36 64.40
+ Bending 2 1/2 in. diam.
Compression 10.19 21.13(6)
Support Rods 1/4 in. Fillet Primary 11.68 22.44 Weld J-Weld Primary 6.18 29.20 Notes:
- 1. When applicable, stresses due to the drop accidents are combined with that of pressure and dead weight.
- 2. Stresses for each DSC components are conservatively combined irrespective of location.
- 3. Allowable stresses shown correspond to service Level D limits, unless noted otherwise.
- 4. Material properties taken at 400°F design temperature.
- 5. Thermal stresses need not be included under service Level D limits.
- 6. Compressive stress allowable of the support rods is based on Appendices XVII and F rules for Level A limits.
8.2-9 Revision No. 22
HBRSEP ISFSI SAR 8.3 FOUNDATION DESIGN To provide a means of transmitting the reaction loads of the ISFSI modules to the ground, a rectangular, flat plate type, mat foundation was selected. The mat foundation is ideally suited for the ISFSI since it spreads out the loadings and consequently reduces the soil bearing pressure and at the same time minimizes the differential settlements.
To accommodate the ISFSI modules, the front cask unloading area and the hydraulic ram area behind the modules, an overall foundation size of 28'-9" by 60'-0" was selected. The HSM foundation slab is 3 feet thick. A construction joint connects this slab to the cask unloading slab which is 2 feet thick starting from a point 5 feet from the module front. The ram mounting slab at the rear of the modules is 8 inches thick and connects to the 3 foot foundation by an expansion joint. The foundation concrete is 4000 psi normal weight concrete poured on a 4 inch mud slab. The HSM foundation and the cask unloading slab are interlaced with continuous two-way reinforcing top and bottom.
Number 9 bars are used for tensile reinforcement and as dowels to anchor the HSM walls to the foundation. The ram mounting slab has a number 5 bar continuous two-way reinforcing at the bottom only. Welded wire fabric is placed at the top of the 8 inch slab.
For analysis purposes, a STARDYNE rectangular plate finite element model as shown in Figure 8.3-1 was developed. The model consists of 255 nodes and 224 plate elements. At each node, a ground support spring was added to simulate the soil elastic properties. The elastic soil spring is obtained by modifying the experimental modulus of subgrade reaction by an appropriate size factor of the foundation. The modified modulus of subgrade reaction is then multiplied by the tributary area associated with each node. The resulting values of the spring stiffnesses were used as input to the finite element model as a ground stiffness matrix. The method for finding the stiffness K is shown below (Reference 8.7):
K = Kv [(B+1)/2B]2 A Where:
Kv = experimental modulus of subgrade reaction
= 100#/in3 (for granular soil)
B = foundation width = 28.75 feet A = nodal tributary area (varies)
Five separate load cases were considered in the foundation design:
- 1) Center module loading
- 2) Outside module loading
- 3) Dead Weight + Live Load
- 4) Dead Weight + Tornado Wind/Impact (lengthwise)
- 5) Dead Weight + Tornado Wind/Impact (widthwise)
Since cask unloading and ram mounting slabs are cast in place after HSM construction, the differential settlement due to HSM dead weight will not be experienced by the cask unloading slab. Consequently, for load cases 1 and 2 the dead weight was not included. For load cases 1 and 2 the total trailer loading of 175 k, which includes the saddle, canister, cask, skid, trailer, rollers, trunnion, and cradle is applied as concentrated loads at nodal locations in the unloading areas. Since only one loading or unloading operation will occur at a time, the two load cases were evaluated independently. Load case 3 consists of the dead weight of the three modules containing the DSCs. This total 3 bay module weight of 800.8 k is added to the weight of three DSCs. The total loading is divided by the surface area in contact with the foundation to get an equivalent pressure load. Additionally, a live load of 200 psf is postulated for the HSM roof. This total load is also divided by the contact area to get a pressure loading on the foundation. These loads are applied as pressure loads on the appropriate plate elements of the STARDYNE model.
Load cases 4 and 5 are the maximum uplift load combinations caused by tornado loadings in the two horizontal directions. Using a conservative wind pressure of 400 psf applied on the module walls and roof plus the reaction 8.3-1 Revision No. 22
HBRSEP ISFSI SAR load of 458.2 k caused by a 3967 lb. automobile traveling at 184.8 ft./s applied to the top of the module in the same direction combined with the module dead weight, the maximum uplift forces were calculated. Live loads are excluded since they would reduce uplift loads. A simple frame model was used to calculate uplift forces as shown in Figure 8.3-2. The maximum uplift force of 38.80 k is converted to a pressure using the contact area of one wall.
This yields a maximum uplift pressure of 3.1 psi which is used in the analysis. Comparison of tornado loads and HSM seismic loadings shows that tornado loads are much more severe. Consequently, seismic loads will not be included in the foundation analysis. Once the uplift forces are calculated they are applied as negative pressures on the appropriate plate elements corresponding to the HSM/foundation connection surface. Tornado wind and impact loads are evaluated for both directions.
Since an uplift force is created by the tornado loads the foundation itself will have a negative bearing or uplift along the edge of the module. The soil itself does not resist uplift. Results from load cases 4 and 5 were reviewed for the effects of the uplift along the foundation edge. Minimal uplift was experienced in low stressed areas. Therefore, results from load cases 4 and 5 are not significantly affected by the uplift since the high stressed areas are not in that vicinity.
The maximum calculated bearing stress is shown in Table 8.3-1. For sandy soils present at the bearing level of the mat foundation, allowable soil bearing pressures in the range of 3000 to 4000 pounds per square foot are recommended by the Southern Standard Building Code per the geotechnical exploration performed at the HBR site by Law Engineering Testing Company. Since the maximum soil contact pressure produced by the HSM foundation analysis is 2210 psf the bearing strength is sufficient. For normal dead weight and live loads the bearing pressure is only 1605 psf.
The reinforcement design was based on the element bending moment results from the finite element analysis. Using the ultimate design method, the reinforcement was designed to withstand all postulated load combination bending moments with a conservative load factor of 1.7 applied to envelope all load combination factors specified in Section 9.2 of ACI 349-80. A tabulation of the results for the HSM Foundation and the cask unloading slab is presented in Table 8.3-2.
A license amendment issued on March 23, 1989, authorized construction of a five module foundation. Supporting information regarding the five module foundation is found in CP&L letters dated January 7, 1989, and February 1, 1989, and in the NRC letter dated March 23, 1989.
8.3-2 Revision No. 22
HBRSEP ISFSI SAR TABLE 8.3-1 FOUNDATION BEARING STRESS LOAD LOAD BEARING CASE DESCRIPTION STRESS (KSF) 1 Center Module 0.247 Loading 2 Outside Module 0.463 Loading 3 Dead Weight + 1.605 Live Load 4 Tornado Wind + 2.210 Impact (Widthwise) 5 Tornado Wind + 0.671 Impact (Lengthwise)
Note:
- 1. Dead weight of module not included in load cases 1 and 2.
8.3-3 Revision No. 22
HBRSEP ISFSI SAR TABLE 8.3-2 FOUNDATION SLAB MAXIMUM BENDING MOMENTS MAXIMUM ALLOWABLE LOAD LOAD SLAB MOMENT MOMENT CASE DESCRIPTION THICKNESS (K-in./in.) (K-in./in.)
Center 2'-0" 13.8 87.0 1 Module Loading 3'-0" 28.9 186.0 Outside 2'-0" 13.1 87.0 2 Module Loading 3'-0" 24.5 186.0 Dead Weight 2'-0" N/A 87.0 3
+ Live Load 3'-0" 46.4 186.0 Tornado Wind 2'-0" N/A 87.0 4 + Impact (Widthwise) 3'-0" 80.1 186.0 Tornado Wind 2'-0" 57.6 87.0 5 + Impact (Lengthwise) 3'-0" 155.8 186.0 Notes:
- 1. Dead weight of module not included in load cases 1 and 2.
- 2. All moments conservatively factored by 1.7 to envelop all ACI 349 load combination factors.
8.3-4 Revision No. 22
HBRSEP ISFSI SAR The moment capacity of the sections are calculated per methods identical to the NUHOMS Topical Section 8.1.1.5, Equation 8-1-32. The 3'-0" slab with number 9 bars at 9 inches yields an ultimate strength of 186 k.in per inch section. The 2'-0" slab with number 9 bars at 12 inches yields an ultimate strength of 87 k.in per inch section.
Therefore all bending moments experienced by the foundation are below ultimate capacity.
As calculated before, the maximum uplift pressure exerted by the module wall is 3.1 psi for load case 5. For a 1'-0" section of module the resulting uplift is 1.56 k/ft. section. The dowel area required can be calculated by:
A = (UPLIFT) (1.7) / (0/)(fy)
Where:
UPLIFT = 1.56 K 0/ = 0.9 = Factor for Tension fy = 60 ksi Therefore, Area A = 0.05 in2. Conservatively 2 number 6 bars with an area of 0.88 in2 will be used every 12 inches to prevent uplift. Keyways will be used between the module foundation interface to prevent sliding. Assuming the maximum horizontal tornado loads are shared by the walls perpendicular to load yields a maximum shear force of 24 k/ft including the load factor of 1.7. The nominal shear strength of the keyway and dowels can be found from:
Vn = (Vs + Vc) 0/
Where:
Vc = 2 ( f'c)1/2 bw d = concrete shear strength (k) f'c = 4000 psi bw = 9 inches d = 12 inches Vs = (Av) (fy) = steel reinforcement shear strength (k)
Av = .88 in2 fy = 60 ksi 0/ = .85 = shear factor Consequently, Vn = 56.49 k which exceeds the maximum factored shear force of 24 k. Thus, the module will neither slide nor overturn. Table 8.3-3 presents foundation anchor loads and capacities for load cases 4 and 5.
The 8 inch ram mounting slab was designed by hand calculations suggested by Teng (Reference 8.7) and Bowles (Reference 8.8). By applying the maximum factored spider leg loadings from the hydraulic ram a simple span is approximated by treating the soil as a uniform load and the spider leg as reaction points. A maximum factored moment of 32.3 k-in/ft is calculated. Using the ultimate strength method with number 5 bars at 12 inches, the ultimate strength of the 8 inch slab is 64.2 k-in/ft. Welded wire fabric was placed at the top of the slab as shrinkage and temperature reinforcement. Additionally all punching shear from the ram supports were found to be negligible.
Furthermore, the cask unloading slab was analyzed for bearing and punching shear due to the hydraulic cylinder. A maximum bearing stress of 2.34 ksi was calculated which is less than the allowable of 4.76 ksi calculated from ACI 349-80 Section 10.16. The maximum punching shear was also found to be under code allowables.
8.3-5 Revision No. 22
HBRSEP ISFSI SAR TABLE 8.3-3 FOUNDATION ANCHOR LOADS LOAD LOADING LOAD CAPACITY CASE DESCRIPTION TYPE (K/FT) (K/FT)
Tornado Wind Shear 13.00 56.5 4 + Impact (Widthwise) Uplift 0 28.0 Tornado Wind Shear 24.00 56.5 5 + Impact (Lengthwise) Uplift 1.56 28.0 Notes:
- 1. All Shear and Uplift loads factored by 1.7 to envelop all ACI 349 load combination factors.
- 2. Shear capacity based on concrete keyway plus embedded dowels.
- 3. Uplift capacity calculated from embedded dowel area.
8.3-6 Revision No. 22
HBRSEP ISFSI SAR 8.4 DSC INSTRUMENTATION PENETRATION DESIGN Instrumentation is not required to support the operation of the ISFSI. However, for research purposes two of the DSCs at the HBR facility have been designed to accept instrumentation. Instrumentation was included as part of an agreement between CP&L, EPRI and DOE to augment the U.S. database on LWR fuel rods in dry storage.
The DSC thermocouples were connected to an external cable by means of a specially designed feed-through. This feed-through incorporates the same redundant seal philosophy used in the DSC containment design. After the penetration plug assembly was welded to the bottom of the DSC cover plate, a sleeve was welded over the plug, forming a redundant seal. Thermocouple sheaths were brazed to the plug assembly at inner and outer penetrations.
To preclude possible leakage through the aluminum oxide insulation, each end of the sheathed thermocouples was sealed with an environmentally qualified resin.
The instrumentation penetration described above was analyzed for the maximum of the three load combinations described below:
- 1) DW + Accident Pressure + Thermal + 8 Ft. Vertical Drop
- 2) DW + Accident Pressure + Thermal + 8 Ft. Horizontal Drop
- 3) DW + Accident Pressure + Thermal + Seismic The dead weight of the instrumentation penetration including the sleeve, the lead, the junction box and miscellaneous fittings was approximately 11 pounds. A 1g acceleration was added on top of the peak deceleration for the 8 foot drop to account for dead weight stresses.
An accident pressure of 39.7 psi was applied to the external surface of the stainless steel tubing sleeve. Using the formula from Roark for a thick-walled vessel (Reference 8.9, Table 32, Case 1d) the maximum stresses were calculated for the accident pressure load as shown below.
qa 2 s1 =
a2 b2 2qa 2 s2 =
a2 b2 s2 t MAX =
2 where:
S1= longitudinal stress, psi S2 = circumferential stress, psi tMAX = maximum shear stress, psi a = .8125 in., outside radius b = .625 in., inside radius q = 39.7 psi, accident pressure S1 = 97 psi S2 = 195 psi tMAX = 97 psi Consequently the maximum stress intensity for the accident pressure case is 0.25 ksi.
8.4-1 Revision No. 22
HBRSEP ISFSI SAR The thermal expansion of the tubing sleeve between the 2 in. plate and the outer 1/4 in. lead casing plate was examined. After comparison of the tubing axial stiffness in relation to the 1/4 in. plate stiffness it was concluded that the sleeve is essentially free to grow since the plate is approximately 72 times as flexible as the tubing. Consequently thermal stresses are considered negligible for the penetration analysis.
The maximum seismic ground accelerations in the horizontal and vertical directions are .2g and .133g, respectively.
They are enveloped by the 8 foot drop peak decelerations.
For the 8 foot drop analysis both vertical and horizontal drops were considered. The peak deceleration for the horizontal drop of 55.4g (1g added for dead weight) was applied to the junction box and tubing extending internally from the weld on the two inch plate. Assuming a cantilever type beam as shown in Figure 8.4-1, the maximum stress intensity of the tubing is 8.19 ksi which is located near the weld. Applying the horizontal drop load to the weld shows a maximum stress of 0.89 ksi. Additionally, the effect of the 2 1/4 in. lead plug pressure loading on the outside surface of the tubing was checked. Results from this analysis indicated a stress much less than the 8.19 ksi previously calculated. Therefore this was not a critical area.
The peak vertical bottom end drop deceleration of 77.5g (1g added for dead weight) was applied to the tubing penetration in the axial direction. The corresponding axial stress was 1.01 ksi, which is considerably less than the horizontal drop case. The weld stress for the vertical drop was calculated as 0.17 ksi. Consequently, the 8 foot horizontal drop load case will be used as the governing load combination.
Combining the dead weight, accident pressure and horizontal drop stresses absolutely for the tubing penetration indicates a maximum stress intensity of 8.44 ksi. The maximum calculated weld stress was 0.89 ksi. These stresses are far below allowable stresses for both components. Clearly then, the confinement integrity of the instrumented DSC will not be jeopardized.
The method for sealing the thermocouple sheaths has been changed from the originally planned sealing by metalizing the aluminum oxide insulation to the use of an environmentally qualified resin. The thermocouple system with the resin sealant has been analyzed for drop (up to 15 inches) and cooling-related accidents. In addition, the integrity of the epoxy seal was reviewed against effects of the following:
- environmental conditions (thermal, radiation) inside and outside of the DSC;
- potential changes in epoxy characteristics over time under expected environmental conditions;
- evaluation of permeation rates; and
- responses to accidents including overpressurization Analyses have concluded that under normal operating conditions seal leakage will be insignificantly small. Also, the structural capability of the material is such that possible accidents will not compromise performance. A detailed discussion of the analyses may be found in References 8.10, 8.11, 8.12, and 8.13.
8.4-2 Revision No. 22
HBRSEP ISFSI SAR 8.5 TRAIN DERAILMENT The railroad in the area of the ISFSI pad was removed as part of EC 287959. Train derailment is not a credible hazard.
8.5-1 Revision No. 27
REFERENCES:
CHAPTER 8 8.1 NUTECH Engineers, Inc., "Topical Report for the NUTECH Horizontal Modular Storage System for Irradiated Nuclear Fuel," NUH-001, Revision 1, November 1985. (Note - currently NUH-001, Rev. 2) 8.2 Carolina Power and Light Company, "H. B. Robinson Steam Electric Plant Unit No. 2 Updated Final Safety Analysis Report," Docket No. 50-261, License No. DPR-23.
8.3 U.S. Nuclear Regulatory Commission, "Missiles Generated by Natural Phenomena," Standard Review Plan NUREG-0800, 3.5.1.4, Revision 2, July 1981.
8.4 General Electric Company, "IF-300 Shipping Cask Consolidated Safety Analysis Report," NEDO-10048-2, Nuclear Fuel and Special Products Division. (Note - currently NEDO-10084-5 issued by Duratek) 8.5 National Fire Protection Association, National Fire Codes, No. 78, 1979 Edition.
8.6 Cybernet Services, STARDYNE User Information Manual, Control Data Corporation, Minneapolis, Minnesota, Revision C, April 1980.
8.7 W. C. Teng, "Foundation Design," Prentice-Hall, Inc., Englewood Cliffs, N.J., 1962.
8.8 J. E. Bowles, "Foundation Analysis and Design," McGraw-Hill, New York, N.Y., 1977.
8.9 R. J. Roark and W. C. Young, "Formulas for Stress and Strain," Fifth Edition, McGraw-Hill, New York, N.Y.,
1975.
8.10 Letter M. A. McDuffie, CP&L to NRC dated January 11, 1989, NLS-89-002.
8.11 Letter L. I. Loflin, CP&L to NRC dated April 28, 1989, NLS-89-117.
8.12 Letter L. I. Loflin, CP&L to NRC dated June 2, 1989, NLS-89-164.
8.13 Letter L. C. Rouse, NRC to CP&L dated June 22, 1989, Amendment to Materials License No. SNM 2502, Amendment No. 7 8.R-1 Revision No. 22
1% MAX. CLEARANCE
.. . A MAXIMUM HEIGHT OF -
BOY-IOM OF CASK .--
TI E-DOWN 8 RACKE TRAILER BED CASK DROP HEIGHT CRITERIA FIGURE 8.2-1
I-NO.
1 2
CASK ORIENTATION 0 OEGREES 48 DEGREES PEAK Gs S4.4 40.8 PLATEAU Gs 243 10.1 3 90 DEGREES 36s IS.7 4 BOTTOM END 7&S 333 80 80 40 20 0 1 I I I 0 1 2 3 4 CASK DECELERATION vs. TIME FOOT DROP FIGURE 8.2-2
AMENDMENT 1 H. B. ROBINSON INDEPENDENT SPENT FUEL STORAGE INSTALLATION DSC BOTTOM REGION ANSYS MODEL FIGURE 8.2-3
I b-l 13" Ammdorat No. 1 DSC TOP REGION ANSYS MODEL FIGURE 8.2-4
H. B. ROBINSON INDEPENDENT SPENT FUEL STORAGE INSTALLATION MAT FOUNDATION STARDYNE MODEL FIGURE 8.3-1
800.8K (DEAD WEIGHT) 0.4 ksf (WIND) s e
a 8 a H
Av ev AV- 31~30~ wfum 13p62215QK FOUNDATION UPLIFT MODEL FIGURE 8.3-2
6LB.
(JUNCTION 80X) 0.75 lb./ IN (TUBING) 3w l o 111) mlrl) 2 0 I I I PENETRATION MODEL FOR 8-FOOT HORIZONTAL DROP ANALYSIS FIGURE 8.4-1
HBRSEP ISFSI SAR CHAPTER 9 CONDUCT OF OPERATIONS Revision No. 22
HBRSEP ISFSI SAR 9.0 CONDUCT OF OPERATIONS 9.1 ORGANIZATIONAL STRUCTURE The development of the Independent Spent Fuel Storage Installation (ISFSI) was managed by Carolina Power &
Light (CP&L) Company with support from NUTECH Engineers, Inc. (NUTECH), the U. S. Department of Energy (DOE), and the Electric Power Research Institute (EPRI). Final responsibility for construction, preoperational testing, startup, and operation of the ISFSI remains with CP&L. Therefore, CP&L's organization and its interfaces with outside support organizations are described below.
9.1.1 CORPORATE ORGANIZATION CP&L is engaged in the production, transmission, distribution, and sale of electric energy to residential, commercial, and industrial customers spread over a service area of 30,000 sq. mi. in North and South Carolina. The Company has extensive experience in the design, construction, startup, testing, operating, and staffing of modern generating facilities. The Company's management of nuclear activities is described in Reference 9.1.
9.1.1.1 Corporate Functions, Responsibilities, and Authorities CP&Ls corporate staff was responsible for installation engineering and design, quality assurance, and acceptance of testing results for the ISFSI. The Nuclear Fuels Section (NFS) and the Nuclear Engineering Department (NED) were jointly responsible for the engineering and design of the ISFSI and the acceptance of test results prior to operation.
An overall Program Manager within the NED was responsible for execution of the program until completion of fuel loading.
9.1.1.2 Applicant's In-House Organization The corporate organization is shown in Reference 9.2, Figure 13.1.1-1. Ultimate responsibility for operation of the ISFSI rests with the Senior Vice President and Chief Nuclear Officer - Nuclear Generation Group reporting to the Group President - Energy Supply. The details of the responsibilities of the CP&L Groups and Departments are described in Section 13.1 of the HBR2 Updated Final Safety Analysis Report (UFSAR) (Reference 9.2).
9.1.1.3 Interrelationships with Contractors and Suppliers The development of the ISFSI was managed by CP&L with support from DOE and EPRI. NUTECH, CP&L's subcontractor, provides certain engineering, technical support, and other services for the program relating primarily to the NUTECH Horizontal Modular Storage (NUHOMS) system design. CP&L and other subcontractors provide the remaining engineering, technical support, and services.
9.1.1.4 Applicant's Technical Staff The Corporate technical staff supporting the ISFSI is described in Section 13.1.1 of Reference 9.2.
9.1.2 OPERATING ORGANIZATION, MANAGEMENT, AND ADMINISTRATIVE CONTROLS SYSTEM 9.1.2.1 Onsite Organization Operation of the Facility is the responsibility of CP&L, primarily by the H. B. Robinson Steam Electric Plant, Unit No. 2 (HBR2), Operations Unit, with support by the Robinson Engineering Section. The RNP Nuclear Oversight Section is responsible for performing periodic assessments of ISFSI activities. The on-site organization is described in Reference 9.2, Chapter 13.1.2.
9.1-1 Revision No. 22
HBRSEP ISFSI SAR 9.1.3 PERSONNEL QUALIFICATION REQUIREMENTS 9.1.3.1 Minimum Qualification Requirements Minimum qualification requirements are discussed in Reference 9.2 Section 13.1.3.
9.1.4 LIAISON WITH OUTSIDE ORGANIZATIONS The development of the ISFSI was managed by CP&L with support from NUTECH, DOE, and EPRI. NUTECH, CP&L's subcontractor, provided certain engineering, technical support, and other services for the program relating primarily to the NUHOMS system design. The overall program was conducted under a Program Management Plan which had been approved under the DOE/CP&L Cooperative Agreement (Reference 9.4). This Program Management Plan provided for an oversight committee which includes representatives of CP&L, DOE, and EPRI.
The oversight committee was responsible for input to, and approval of, the schedular, technical, procedural, and administrative details of the program. EPRI, acting through and in cooperation with CP&L (Reference 9.5),
participated in the formulation and formal definition of the research and development aspects of the program. DOE participated in the program through financial and/or in-kind services (DOE/CP&L Cooperative Agreement).
Program review and guidance was provided to CP&L by all participants. The design, procurement, construction, and operation of the HBR ISFSI are controlled through existing CP&L procedures and programs. Operation of the Facility is the sole responsibility of CP&L.
9.1-2 Revision No. 22
HBRSEP ISFSI SAR 9.2 PRE-OPERATIONAL TESTING AND OPERATION Two types of tests were planned prior to loading of spent fuel. The first is thermal testing of the DSC/HSM system to verify its heat removal capacity; the second is testing of the DSC transfer system to ensure a safe, smooth DSC insertion into the HSM and back to the cask.
It was only necessary to perform the thermal tests only upon the first HSM and DSC (or prototype) constructed.
Subsequent units built to the same requirements may be safely presumed to operate in a similar manner. Likewise, handling tests need only be performed once, since all future operations will use equipment proven during pre-operational testing or equipment fabricated from those plans.
9.2.1 ADMINISTRATIVE PROCEDURES FOR CONDUCTING TEST PROGRAM Existing HBR2 modification procedures will be followed for the ISFSI test program.
9.2.2 CP&L TEST PROGRAM DESCRIPTION Thermal testing was planned to be performed using one of the completed HSMs. The DSC need not be an exact replica of the actual stainless steel canister, but may be simply an appropriately sized closed-end pipe. Resistance-type electric heating elements within the DSC replica will simulate decay heat. The purpose of this test is to verify the thermal-buoyancy driven convective air flow through the HSM.
Handling testing was planned to be performed using the actual HSM, a DSC replica and the associated transfer handling equipment. The purpose of this series of tests is to ensure that all DSC handling operations will be performed safely. The tests simulate, as nearly as possible, actual handling operations with spent fuel including mock-ups for performing the DSC welding operations. Off-normal handling conditions need not be addressed, since the DSC can remain in the cask indefinitely or be returned to the spent fuel pool, if proper alignment cannot be achieved.
9.2.3 TEST DISCUSSION 9.2.3.1 Physical Facilities Testing (Thermal Testing of HSM and DSC)
The DSC replica will be placed in the HSM and internally heated by electric resistance heaters. Heat flux will simulate the decay heat value of seven assemblies (7 kw). The purpose of the test is to verify the passive heat removal capacity of the HSM/DSC. The test will measure the difference between entering and exiting air temperatures. The expected response is that exit air will be 100°F above entering air temperature. The air flow volume will be measured to allow computation of heat flow by convection through the module. This test will be repeated with partial and complete blockage of the HSM air passages.
If the results of the tests showed that insufficient convective cooling occurs, resulting in elevated concrete and fuel cladding temperatures, corrective measures could have been taken and the test repeated.
9.2.3.2 Operations Testing (Handling Tests)
The purpose of the handling tests is to verify that the DSC handling system is adequate to achieve alignment of the cask with the HSM and insertion of the DSC into the HSM.
The expected response of the system testing is that alignment and insertion may be achieved. A successful test is one in which alignment and insertion is achieved without damage to system components.
Corrective action, in the event of an unsuccessful test, would have been to modify system components as appropriate.
Handling tests were also planned to be conducted to ensure that the DSC can be removed from the HSM, returned to the IF-300 shipping cask, and returned to the HBR2 spent fuel pool.
9.2-1 Revision No. 22
HBRSEP ISFSI SAR 9.3 TRAINING PROGRAM 9.3.1 PLANT STAFF TRAINING PROGRAM The training programs for HBR2 were modified to incorporate the operation of the Independent Spent Fuel Storage Installation. Since the ISFSI is essentially a passive structure with no equipment or instrumentation required for normal operation, the operator training program will concentrate on the accident training. In addition to the operator training, the radiation protection and fire brigade training programs have been expanded to cover the ISFSI.
The current HBR2 plant staff training program is described in Section 13.2 of the HBR2 Updated FSAR (Reference 9.2).
9.3.2 REPLACEMENT AND RETRAINING PROGRAM The HBR2 replacement training and retraining programs have been updated to include the ISFSI.
The current HBR2 replacement and retraining programs are described in Section 13.2 of the HBR2 Updated FSAR (Reference 9.2).
9.3-1 Revision No. 22
HBRSEP ISFSI SAR 9.4 NORMAL OPERATIONS The HBR ISFSI provides an independent and passive system for the dry storage of irradiated fuel. No radioactive waste or auxiliary systems are required during normal storage operations.
9.4.1 PROCEDURES Although not considered normal operations, the processes for transfer of the fuel from the spent fuel pool to the horizontal storage modules (HSMs) are briefly discussed below. These processes will be conducted only when loading and unloading the ISFSI. Specific procedures were developed for transporting, loading into the HSM, and retrieving operations involving the dry shielded canister (DSC). The existing HBR2 procedures for handling irradiated fuel in the spent fuel pool, loading of the GE IF-300 shipping cask, and handling by the spent fuel crane will be used for these ISFSI operations. As discussed in Chapter 6, some waste and auxiliary systems are required during the DSC loading, drying and transfer into the module. The waste systems handle the spent fuel pool water, air, and inert gas which are vented from the DSC and cask during drying. Auxiliary handling systems (such as hydraulic pressure, control, alignment, crane, etc.) are also required during the loading and transfer operation.
9.4.2 RECORDS The ISFSI records will be maintained in accordance with existing HBR2 procedures.
9.4-1 Revision No. 22
HBRSEP ISFSI SAR 9.5 EMERGENCY PLANNING The Emergency Program for HBR2 has been determined to be adequate for events which might occur involving the Independent Spent Fuel Storage Installation. The Emergency Program for HBR2 consists of the Robinson Emergency Plan and its implementing Procedures. Also included are related radiological emergency plans and procedures of state and local organizations. The purpose of these programs is to provide protection of plant personnel and the general public and to prevent or mitigate property damage that could result from an emergency at the HBR2 Plant. The combined emergency preparedness programs have the following objectives:
a) Effective coordination of emergency activities among all organizations having a response role.
b) Early warning and clear instructions to the population-at-risk in the event of a serious radiological emergency.
c) Continued assessment of actual or potential consequences both on-site and off-site.
d) Effective and timely implementation of emergency measures.
e) Continued maintenance of an adequate state of emergency preparedness.
The HBR2 Emergency Plan has been prepared in accordance with Section 50.47 and Appendix E of 10 CFR Part 50.
The Plan shall be implemented whenever an emergency situation is indicated. Radiological emergencies can vary in severity from the occurrence of an abnormal event, such as a minor fire with no radiological health consequences, to nuclear accidents having substantial onsite and/or offsite consequences.
In addition to emergencies involving a release of radioactive materials, events such as security threats or breaches, fires, electrical system disturbances, and natural phenomena that have the potential for involving radioactive materials are included in the Plan. Other types of emergencies that do not have a potential for involving radioactive materials are not included in the Plan.
The activities and responsibilities of outside agencies providing an emergency response role at the HBR Plant site are summarized in the Plan (for completeness) and detailed in the State Emergency Plans.
The Plan provides the basis for performing advance planning and for defining specific requirements and commitments to be implemented by other documents and procedures. The HBR2 Plant site procedures provide the detailed actions and instructions that will be required to implement the Plan in the event of an emergency.
The emergency response resources available to respond to an emergency consist of the personnel at Corporate Headquarters, at other CP&L facilities, and, in the longer term, at organizations involved in the nuclear industry.
The first line of defense in responding to an emergency lies with the normal operating shift on duty when the emergency begins. Therefore, members of the HBR2 staff are assigned defined emergency response roles that are to be assumed whenever an emergency is declared. The overall management of the emergency is normally performed by plant management. Onsite personnel have preassigned roles to support the Site Emergency Coordinator/Emergency Response Manager and to implement their directives. These roles, for the purpose of emergency planning, are cast in terms of emergency teams and assignments, each having a designated person assigned to it.
Special provisions have been made to assure that ample space and proper equipment are available to effectively respond to the full range of possible emergencies.
The emergency facilities available include the Robinson Plant Control Room, Operational Support Center, Technical Support Center, Emergency Operations Facility, Joint Information Center, and Harris Energy & Environmental Center. Descriptions of these facilities as well as the South Carolina EOC, the Darlington County EOC, the Lee County EOC, and the Chesterfield County EOC, are described in the Plan.
9.5-1 Revision No. 22
HBRSEP ISFSI SAR Emergency plan implementing procedures define the specific (i.e., step-by-step) actions to be followed in order to recognize, assess, and correct an emergency condition and to mitigate its consequences. Procedures to implement the Plan provide the following information:
a) Specific instructions to the plant operating staff for the implementation of the Plan.
b) Specific authorities and responsibilities of plant operating personnel.
c) A source of pertinent information, forms, and data to ensure prompt actions are taken and that proper notifications and communications are carried out.
d) A record of the completed actions.
e) The mechanism by which emergency preparedness will be maintained at all times.
9.5-2 Revision No. 22
HBRSEP ISFSI SAR 9.6 DECOMMISSIONING PLAN The HBR ISFSI makes provisions for the dry storage canister to be transferred to a federal repository when such a facility becomes operational. The concrete storage module is designed so that the canister can be safely returned to a shipping cask and transported offsite to the federal facility.
The shipping cask design and transportation requirements will depend on the regulations in effect at the time when the federal facility begins receiving spent fuel. In the absence of new regulations, it is likely that the existing GE IF-300 shipping cask would be used to transport the canisters.
Upon removal of the canisters, the level of contamination within the concrete module is expected to be extremely low and would be removed so that the concrete could be broken-up and removed by conventional methods.
Section 3.5 of this report discusses decommissioning considerations for the HBR ISFSI.
9.6-1 Revision No. 22
HBRSEP ISFSI SAR 9.7 LICENSE RENEWAL ACTIVITIES 9.7.1 AGING MANAGEMENT PROGRAMS An assessment of the RNP ISFSI and Transfer Cask inspection and monitoring activities identified existing activities necessary to provide reasonable assurance that ISFSI and Transfer Cask components within the scope of license renewal will continue to perform their intended functions consistent with the current licensing basis (CLB) for the renewal period.
This section describes these aging management programs.
9.7.1.1 ISFSI AGING MANAGEMENT PROGRAM The RNP ISFSI Aging Management Program involves monitoring the exterior surface of the horizontal storage module. It includes visual inspection of the accessible concrete (including below grade concrete, if exposed during excavation) and any exposed steel embedments and attachments. It also includes monitoring the area radiation levels, airborne contamination, and smearable contamination at selected areas of the ISFSI. This is primarily a condition monitoring program, however, preventive actions include a daily surveillance to ensure horizontal storage module air inlets and outlets are not obstructed.
9.7.1.2 TRANSFER CASK AGING MANAGEMENT PROGRAM The Transfer Cask Aging Management Program performs visual inspections of the exterior surfaces of the IF-300 Transfer Cask, cask collar, and lid. The program also monitors the water chemistry of the cask neutron shield water jacket to prevent the corrosion of exposed internal surfaces.
Note: If the environmental conditions are changed at a later date, such as by moving the cask to indoor storage or draining the neutron shield water jacket fluid, there will be no need to continue visual inspections of the external stainless steel surfaces or to continue monitoring the chemistry of the neutron shield water jacket.
9.7.2 TIME-LIMITED AGING ANALYSIS This section discusses the results for each of the time-limited aging analyses (TLAAs) evaluated for license renewal.
The evaluations have demonstrated that the analyses are valid through the end of the renewed license period.
9.7.2.1 DSC SHELL CRACKING DUE TO FATIGUE The Dry Shielded Canister (DSC) shell fatigue cumulative usage factor (CUF) was calculated to be 0.21 for the first 50 years of service. The additional 10 years of service was estimated to increase the usage factor to 0.25. This value is less than the allowable 1.0. Therefore, the CUF has been reanalyzed and projected to be valid for the renewed license period.
9.7.2.2 DSC PENETRATION ASSEMBLY EPOXYLITE SEAL CHANGE IN MATERIAL PROPERTIES DUE TO IONIZING RADIATION Two of the DSCs contain sheathed thermocouple feeds through a penetration plug assembly in the bottom end plug, which were sealed (wire to sheathing) using a heat-cured epoxy-resin material. The integrated exposure for epoxy-resin after 40 years was calculated to be 2.21E11 ergs/g, and after 50 years calculated to be 2.56E11 ergs/g. The integrated exposure after 60 years of service was projected to be less than 2.91E11 ergs/g. This is well below the acceptable level of 1E12 ergs/g for this type of resin. In addition, the validity of the assumed negligible effects of neutron radiation was confirmed for the renewed license period. Therefore, the effects of ionizing radiation on the epoxy-resin have been reanalyzed and projected to be valid for the renewed license period.
9.7.2.3 DSC POISON PLATE DEPLETION OF BORON Boron depletion can result in reduced neutron absorption capability for neutron poison materials, such as the aluminum boron metal used in the DSC baskets. The total flux was estimated to be 4.1E8 n/cm2/sec. After 60 years of service, the total flux (fluence) is estimated to be 7.8E17 n/cm2. Conservatively assuming that the total flux is 9.7-1 Revision No. 22
HBRSEP ISFSI SAR thermal, the depletion in Boron B-10 level will be negligible (less than 0.3 %). Therefore, DSC Poison Plate Depletion of Boron has been reanalyzed, boron depletion has been determined to be negligible, and this has been projected to be valid for the renewed license period.
9.7.2.4 5% BORON-POLYETHYLENE FRONT ACCESS COVER PLATE CRACKING AND CHANGE IN MATERIAL PROPERTIES DUE TO IONIZING RADIATION The front access cover plate on each horizontal storage module is a carbon steel plate and frame that encloses a lead plate and a 2-inch thick boron-polyethylene sheet. Degradation of the boron-polyethylene sheet due to ionizing radiation may result in reduced shielding capability. The gamma dose rate on the inside face of the access cover plate was calculated to be 23 mR/hr, which results in a gamma radiation dose of less than the allowable 5E8 Rads over a life of 60 years. The neutron dose on the inside face of the front access cover plate was calculated to be 1.14E12 n/cm2, which is less than the allowable 2.5E13 n/cm2 over a life of 60 years. Therefore, degradation of the boron-polyethylene sheets due to ionizing radiation has been reanalyzed and projected to be valid for the renewed license period.
9.7-2 Revision No. 22
REFERENCES:
CHAPTER 9 9.1 Carolina Power & Light Company, Harris Nuclear Project, "Management Capability Report," letter from E. E. Utley (CP&L) to H. R. Denton (NRC) dated January 10, 1984.
9.2 Carolina Power & Light Company, "H. B. Robinson Steam Electric Plant Unit No. 2 Updated Final Safety Analysis Report," Docket No. 50-261, License No. DPR-23.
9.3 Carolina Power & Light Company, "Technical Specifications and Bases for H. B. Robinson Unit No. 2,"
Appendix A to Facility Operating License DPR-23, Docket No. 50-261, Darlington County, SC.
9.4 CP&L/DOE Licensed At-Reactor Dry Storage Demonstration Program, Cooperative Agreement No.
DE-FC06-84RL10532, Amendment No. A000, March 1984.
9.5 Associate Party Agreement Between CP&L and EPRI, Inc., RP2566-1, July 1984.
9.R-1 Revision No. 22
HBRSEP ISFSI SAR CHAPTER 10 OPERATING CONTROLS AND LIMITS Revision No. 22
HBRSEP ISFSI SAR 10.0 OPERATING CONTROLS AND LIMITS The H. B. Robinson (HBR) Independent Spent Fuel Storage Installation (ISFSI) is a totally passive system which requires minimum operating controls. In the following sections, the operating controls and limits that are pertinent to the ISFSI are specified. The conditions and other items to be controlled are based on the safety assessments for normal and postulated accident conditions provided in Chapter 8 of this report.
The following operating controls and limits are specified:
10.1 Fuel Specifications 10.2 Limits for the Surface Dose Rate of the HSM While the DSC is in Storage 10.3 Limits for the Maximum Air Temperature Rise After Storage 10.4 Surveillance of the HSM Air Inlets 10.5 Surveillance of the HSM Inside Cavity 10.0-1 Revision No. 22
HBRSEP ISFSI SAR 10.1 FUEL SPECIFICATIONS
1.1 Title
Fuel Specifications
1.2 Specifications
Type 15 x 15 PWR Fuel Burnup <
_ 35,000 MWd/MT Initial (Beginning of Life)
Enrichment _ 3.5% U-235 Post Irradiation Time _ 5 years Weight Per Distance Between Adjacent Spacers Per Assembly _ 106.56 kg Distance Between Spacers <
_ 0.65 m Any fuel not specifically filling the above requirements may still be stored in the ISFSI if all the following requirements are met:
Decay Power < 1 kw/assembly Neutron Source 1.67 x 108 n/sec/assembly 15 Gamma Source 5.73 x 10 photons/sec/canister With spectrum bounded by that shown in Table 10.1-1 End of Life 0.8% U-235 Fissile Content 0.5% Pu-239 0.1% Pu-241
1.3 Applicability
This specification is applicable to all fuel to be stored in the ISFSI.
1.4 Objective
This specification was derived to insure that the peak fuel rod temperatures, surface doses and nuclear subcriticality are below the design values.
1.5 Action
If this specification is not met, additional analysis and/or data must be presented before the fuel can be placed in the DSC.
1.6 Surveillance The fuel selected for storage must have the parameter values specified in 1.2 Requirements: above verified prior to fuel loading. No other surveillance is required.
1.7 Basis
The fuel parameters specified in this operating control and limit were selected to bound the types of PWR fuel which were in use at the time the HBR ISFSI was installed and loaded with spent fuel.
10.1-1 Revision No. 22
HBRSEP ISFSI SAR TABLE 10.1-1 ACCEPTABLE RADIOLOGICAL CRITERIA FOR STORAGE OF MATERIAL IN THE HBR ISFSI CRITERION VALUE 16 GAMMA SOURCE PER CANISTER (total) 1.48 x 10 Mev/sec Fractional Breakdown1 Above 1.3 Mev 0.004 Between 1.3 Mev and 0.8 Mev 0.114 Between 0.8 Mev and 0.4 Mev 0.808 Below 0.4 Mev 0.074 NEUTRON SOURCE PER CANISTER (total)2 1.17 x 109 n/sec Fractional Breakdown Above 5 Mev 5.40 x 107 n/sec = 5.41%
Between 2.5 and 5 Mev 2.43 x 108 n/sec = 24.32%
Between 1 and 2.5 Mev 4.56 x 108 n/sec = 45.67%
Below 1 Mev 2.45 x 108 n/sec = 24.53%
1 Fractional breakdown is based on isotopic composition and resulting gamma spectrum calculated by ORIGEN2 analysis.
2 Spectrum from U-235 fission, total number of neutrons per second from ORIGEN2 analysis.
10.1-2 Revision No. 22
HBRSEP ISFSI SAR 10.2 LIMITS FOR THE SURFACE DOSE RATE OF THE HSM WHILE THE DSC IS IN STORAGE 2.l
Title:
Surface Dose Rates on the HSM while the DSC is in Storage
2.2 Specification
Surface dose rates at the following locations
- 1) Outside of HSM door on centerline of DSC 200 mrem/hr
- 2) Center of air inlets 200 mrem/hr
- 3) Center of air outlets 200 mrem/hr Average Dose rates for the following surfaces
- 1) Roof 50 mrem/hr
- 2) Front/Back 50 mrem/hr
- 3) Side 50 mrem/hr Dose rates one meter from the center of the following surfaces of a unit of modules
- 1) Front/Back 20 mrem/hr
- 2) Side 20 mrem/hr
2.3 Applicability
This specification is applicable to the ISFSI.
2.4 Objective
The objective of this specification is to maintain as-low-as-reasonably-achievable dose rates on the modules.
2.5 Action
If the dose rates are exceeded, temporary shielding must be placed so as to reduce the dose rates to the specified levels. When temporary shielding is used, the outlet air temperature must be measured after the shielding is installed to verify that the air flow has not been restricted.
2.6 Surveillance
The HSM shall be monitored to verify that this specification has been met immediately after the DSC is placed in storage and the HSM front and rear accesses are closed.
2.7 Basis
The dose rates stated in this specification were selected to maintain as-low-as-reasonably-achievable exposures to personnel performing air duct clearing on the HSM. These dose rates are within industry accepted standards for contact handling, operation and maintenance of radioactive material. Maintenance personnel will be required to remove any potential air blockage. At 200 mrem/hr the dose for a one hour job of unblocking the air inlets (or outlets) would be less than 200 mrem (whole body) and hence would be only 4% of the total yearly burden. Furthermore, analysis provided in Chapter 7 of the HBR ISFSI SAR shows that the expected dose rates around the HSM surface will be well below the specifications listed above.
10.2-1 Revision No. 22
HBRSEP ISFSI SAR 10.3 LIMITS FOR THE MAXIMUM AIR TEMPERATURE RISE AFTER STORAGE
3.1 Title
Maximum Air Temperature Rise from HSM Inlet to Outlet
3.2 Specification
Maximum air temperature rise l00°F (55.6°C)
3.3 Applicability
This specification is applicable to the ISFSI.
3.4 Objective
To limit the maximum air temperature around the DSC.
3.5 Action
If the temperature rise is greater than l00°F (55.6°C), the air inlets and exits should be checked for blockage. 10.3-1 If the blockage is cleared and the temperature still exceeds this limit, the DSC must be removed from the module or additional information and analysis shall be provided that will prove the existing condition does not represent an unsafe condition.
3.6 Surveillance
The temperature rise shall be checked at the time the DSC is stored in the HSM, again 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> later, and again after 7 days.
3.7 Basis
The l00°F (55.6°C) temperature rise was selected to limit the hottest rod in the DSC to below 716°F (380°C). If this temperature rise is maintained, then the hottest rod will be below the 716°F (380°C) limit even on the hottest day conditions of 125°F (5l.7°C). The expected temperature rise is less than 100°F (i.e., 82°F (45.5°C) per NUHOMS Topical Report (Reference 8.1), Section 8.1.3) and hence, the current design provides adequate margin for this specification. If the temperature rise is within the specifications, then the HSM and DSC are performing as designed and no further temperature measurements are required during normal surveillance.
10.3-1 Revision No. 22
HBRSEP ISFSI SAR 10.4 SURVEILLANCE OF THE HSM AIR INLETS AND OUTLETS
4.1 Title
Surveillance of the HSM Air Inlets and Outlets
4.2 Specifications
Normal visual inspection frequency: Daily Accident visual inspection frequency: Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after an accident
4.3 Applicability
This specification is applicable to the ISFSI.
4.4 Objective
To assure that no HSM air inlets or outlets are plugged for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> and to assure that complete blockage of all inlets and exits due to an accident will be removed in less than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
4.5 Action
If the air inlets or outlets are plugged, they should be cleared. If the screen is damaged, it should be replaced.
4.6 Surveillance
The HSM shall be inspected to verify that the air inlets are free from obstructions.
4.7 Basis
Analysis in Chapter 8 of the HBR ISFSI SAR showed that no temperature limits are exceeded if a module is completely plugged for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. Furthermore, analysis showed that blockage of the air inlet alone did not result in unacceptable temperatures. Therefore, for normal operations, an inspection of the inlets once per day will assure that any local obstructions can be removed.
Likewise, after an accident (such as those described in Chapter 8) the HSMs should be examined within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to assure that if a module is completely buried, flow can be restored within another 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
10.4-1 Revision No. 22
HBRSEP ISFSI SAR CHAPTER 11 QUALITY ASSURANCE Revision No. 22
HBRSEP ISFSI SAR 11.0 QUALITY ASSURANCE The activities associated with the Independent Spent Fuel Storage Installation (ISFSI) are performed under the guidance of applicable portions of the CP&L Quality Assurance (QA) Program delineated in Reference 11.1.
Controls are established for the applicable activities which CP&L performs as well as for those appropriate activities which subcontractors perform.
Nuclear Oversight personnel perform oversight to assure that appropriate quality requirements are being met in the normal implementation of technical programs. Work associated with this project will be performed in the same way as other nuclear plant related activities, and will, therefore, be covered by the normal audit process.
11.0-1 Revision No. 24
HBRSEP ISFSI SAR 11.1 QUALITY ASSURANCE It is the policy of CP&L to design, construct, and operate nuclear power plants without jeopardy to its employees or to the public health and safety. The QA programs shall be developed, implemented, and updated as necessary to assure that the Company's nuclear facilities will be managed such that all systems used to produce, convey, or use nuclear generated steam and all systems used to treat, store, or convey waste produced by the generation of nuclear steam will be designed, constructed, and operated in a safe manner. Deviations from this program shall be permitted only upon written authority from the corporate management position originally approving the program or implementing procedures.
The design, construction, and operation of nuclear facilities shall be accomplished in accordance with the U. S.
Nuclear Regulatory Commission (NRC) Regulations specified in Title 10 of the U.S. Code of Federal Regulations.
All commitments to the NRC Regulatory Guides and to engineering and construction codes shall be carried out.
The operation of the Company's ISFSI shall be in accordance with the terms and conditions of the ISFSI material license issued by the NRC (10 CFR Part 72). Changes in operating procedures, experiments at the ISFSI, modifications to the ISFSI hardware or systems, shall be made in accordance with the terms and conditions of the ISFSI material license.
The Dry Storage Canister and Transfer Cask are considered safety-related and are subject to a QA program in conformance with the requirements of 10 CFR 50, Appendix B and the QA Program described in Reference 11.1.
This QA program will be applied to those structures, systems, and components of the Horizontal Storage Module (HSM) and foundation that are important to safety. The program will apply for the continued inspection, testing, operation, maintenance, repair, and modification of these HSM structures, systems, and components which are important to safety.
A Radwaste QA Program was applied to the HSM and foundation during the procurement, construction, and testing phases of the project. After these phases were completed and the initial loading of the Dry Shielded Canisters (DSCs) was completed, the HBR ISFSI deemed "operational." Keeping in mind that the NUHOMS system being utilized for the HBR ISFSI is a totally passive installation, there are no pieces of equipment which require operation nor a requirement for data collection and reporting.
However, as stated in Section 5.1.1.7 of this document, a daily visual inspection of air inlets and outlets will be made to insure that they remain unblocked and the integrity of the screens remains intact. In addition, it is stated in Section 7.3.4 of this document that "The operation of the ISFSI will be monitored under the HBR Unit 2 Radioactivity Monitoring Program. No additional radiation monitoring instrumentation is required." Operation of the HBR ISFSI will be as established by the requirements of the Technical Specifications contained in the Safety Analysis Report and will be carried out in accordance with the QA program.
Audits will be performed by Nuclear Oversight at least every two years. Audits and other management oversight activities shall be performed in accordance with the QA Program for safety-related activities.
11.1-1 Revision No. 24
HBRSEP ISFSI SAR 11.2 H. B. ROBINSON QUALITY ASSURANCE PROGRAM The HBR QA Program is controlled by the policies and requirements of the QA Program. These policies and requirements are implemented through the Plant Operating Manual and other approved procedures. The program is designed to ensure compliance with the applicable NRC Regulatory Guides and ANSI Standards.
Details of the HBR QA program are contained in Reference 11.1.
11.2-1 Revision No. 22
HBRSEP ISFSI SAR 11.3 NUTECH QUALITY ASSURANCE The HBR ISFSI is designed by NUTECH Engineers, Inc. The quality assurance followed by NUTECH is described in Reference 11.2 and has been approved by CP&L Quality Assurance.
11.3-1 Revision No. 22
REFERENCES:
CHAPTER 11 11.1 Carolina Power & Light Company, "H. B. Robinson Steam Electric Plant Unit No. 2 Updated Final Safety Analysis Report," Docket No. 50-261, License No. DPR-23.
11.2 NUTECH Engineers, Inc., "Topical Report for the NUTECH Horizontal Modular Storage System for Irradiated Nuclear Fuel," NUH-001, Revision 1, November 1985.
11.R-1 Revision No. 22