ML20125D226
| ML20125D226 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 11/06/1992 |
| From: | HOLTEC INTERNATIONAL |
| To: | |
| Shared Package | |
| ML20125D184 | List: |
| References | |
| HI-92828, HI-92828-R05, HI-92828-R5, NUDOCS 9212150058 | |
| Download: ML20125D226 (327) | |
Text
._--_- _ - . - - - - - __ .
I i
l, HOLTEC INTERNATIONAL i
i I LICENSING REPORT l FOR i
1 SPENT FUEL STORAGE 4
CAPACITY EXPANSION
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i i
FORT CALHOUN STATION i
i DOCKET 50-285 i
l l Holtec Report HI-92828
- Omaha Public Power District Omaha, Nebraska
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9212150058 921207 PDR ADDCK 05000285 P PDR i
1 HOLTEC lNTERNATIONAL REVIEW AND CERTIFICATION LOG DOCUMENT NAME: LICENSING REPORT FOR SPENT FUEL STORAGE CAPACITY EXPANSION HI- 2828 HOLTEC DOCUMENT I.D. #
HOLTEC PROJECT NUMBER _20330 CUSTOMER / CLIENT Omaha Public Power District 4
REVISION BLOCK ISSUE QUALITY PROJECT NO. AUTHOR REVIEWER ASSURANCE MANAGER
& DATE & DATE & DATE & DATE ORIGINAL WA % 9lWl9t b
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nahlY1- l9'2-REV. 6 Must be Project Manager or his Designee.
NOTE: Signatures and printed names are required in the review block.
This document conforms to the requirements of the design specification and the applicable sections of the governing codes.
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I J
SUMMARY
OF REVISIONS LOG Report HI-92828 Preliminary Issue: May 14, 1992.
Revision 0: Issued August 21, 1992.
Revision 1: Issued October 5, 1992.
Contains the following number of pages:
Title Page 1 Review and Certification Log 1 Table of Contents 4 List of Tables 4 List of Figures 5 Section 1 9 Section 2 21 Section 3 18 Section 4 29 Section 5 34 Section 6 129 Note: Section 6 includes a page 35a Section 7 7 Section 8 15 Section 9 15 Section 10 6 Section 11 6 Revision 2: Issued October 8, 1992.
Revision 2 incorporates OPPD's comments.
Revision 2 contains the same number of pages as Revision 1 with the exceptions of Section 9 (now 14 pages) and Section 11 (now 7 pages).
Revision 3 incorporates OPPD's comments, and revises the-following pages: 4-4, 4-11, 4-22, 5-1, 8-8, 9-2, 9-5, 9-9, and 9-12.
l Revision 3 contains the saae number of pages as Revision 2.
Revision 4 incorporates OPPD's comments and revises the following number of pages: 1, ii, v, vii, viii, and 7.3.
. Section 4 is revised to add calculations for considering-CEA.
Contains same number of pages except Section 4 now contains 34 pages.
Revision 5' incorporates OPPD's comments and revises'the I following pages: 1-4, 1-7,2-5, 2-7,_2-8, 3-3, 3-5, 3-6, 4-1, 4-4, 4-5, 4-12, A-2, A-12, 5-1, 5-2, 5-4,~5-9, 5-15, 6-2, 6-3, L 6-4, 6-5, 6-6, 6-7,-6-9,-6-11, 6-15, 6-19, 6-23, 6-26, 6-30, l 6-35a, 6-46, 7-2, 8-6, 9-3, 9-4,_9-5, 9-11, 10-3,-and 10-6.
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l TABLE OF CONTENTS
1.0 INTRODUCTION
] 2.0 MODULE LAYOUT FOR INCREASED STORAGE
! 2.1 Background 2-1 j 2.2 Multi-Region Storage 2-2 l 2.3 Material Considerations 2-3
- 2.3.1 Introduction 2-3 j 2.3.2 Structural Materials 2-3
- 2.3.3 Poison Material 2-4 j 2.3.4 Compatibility with Coolant 2-6
- 2.4 Existing Rack Modules and- 2-6 Proposed Raracking Operation j 2.5 Heavy Load Considerations for the 2-6 Proposed Reracking Operation j 3.0 RACK FABRICATION AND APPLICABLE CODES i 3.1 Fabrication objective 3-1
,' -3.2 Rack Module-for Region I 3-1 l 3.3 Rack Module for Region II 3-3
- 3.4 Codes, Standards, and Practices 3-5 for the-Fort Calhoun Station Spent Fuel Racks 4.0 CRITICALITY SAFETY CONSIDERATIONS 4.1 Design Basis 4-1 2
4.2 Summary of Criticality Analyses 4-4 i 4.2.1 Normal Operating Conditions - 4-4 4.2.2 Abnormal and Accident Conditions 4-6 l 4.3 Reference Fuel Storage Cells 4-7 j
4.3.1 -Reference Fuel Assembly 4-7 4.3.2 Region 1 Fuel Storage Calls 4-7 4.3.3. Region 2 Fuel Storage Cells -
4-8
- 4.4 Analytical Methodology 4-9
- 4.4.1 Reference Design Calculations 4-9
! 4.4.2 Fuel'Burnup Calculations and 4-10 Uncertainties 4.4.3 Effect of Axial Burnup 4-11 Distribution
, 4.4.4 Long-term Changes in Reactivity 4-12 4.5 Region I Criticality Analyses and 4-13 i Tolerances
- 4.5.1 Nominal' Design 4-13 4.5.2 Uncertainties due to Tolerances- 4-13 4.5.2.1 Boron Loading Tolerances 4-13 4.5.2.2 Boral Width Tolerance 4-14 4
4.5.2.3 Tolerances in Cell 4-14 Lattice Spacing a
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TABLE OF CONTENTS (continued) 4.5.2.4 Stainless Steel Thickness 4-14 Tolerances
- 4.5.2.5 Fuel Entichment and 4-15 Density Tolerances 4.6 Region 2 Criticality Analyses 4-16 4.6.1 Nominal Design Case 4-16 4.6.2 Boundary Cells (Region 3) 4-17
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4.6.3 Uncertainties due to Tolerances 4-18 4.6.3.1 Boron Loading Tolerances 4-18
, 4.6.3.2 Boral Width Tolerance 4-18 i 4.6.3.3 Tolerance in Cell Lattice 4-18 Spacing 4.6.3.4 Stainless Steel Thickness 4-19 Tolerance 4.6.3.5 Fuel Enrichment and 4-19 Density Tolerances 4.7 Fuel Currently in Storage 4-19 4.8 Abaormal and Accident Conditions 4-20 4.8.1 Temperature and Water Density 4-20
, Effects 4.8.2 Eccentric Fuel Positioning 4-20 4.8.3 Dropped Fuel Assembly 4-21
! 4.8.4 Lateral Rack Movement 4-21 4.8.5 Abnormal Location of a Fuel 4-21 Assembly 4.9 References for Section 4 4-23 Appendix A to Section 4 5.0 THERMAL-HYDRAULIC CONSIDERATIONS 4
, 5.1 Introduction 5-1 d
5.2 Spent Fuel cooling System and Cleanup 5-2
System Description
5.3 Decay Heat Load Calculations 5-3 5.4 Discharge Scenarios 5-3 5.5 Bulk Pool Temperatures 5-4 5.6 Local Pool Water Temperature 5-8 5.6.1 Basis 5-7 5.6.2 Model Description 5-8 5.7 Cladding Temperature 5-10 5.8 Results 5-12 5.9 References for Section 5 5-13 6.0 SEISMIC / STRUCTURAL CONSIDERATIONS 6.1 Introduction 6-1 6.2 Analysis Outline 6-1 6.3 Artificial Time-Histories 6-6 11
i TABLE OF CONTENTS (continued) i 1
6.4 Rack Modeling for Dynamic Simulations 6-10 4
6.4.1 General Remarks 6-10 i
6.4.2 The 3-D 22 DOF Model for Single Rack Module 6-12 Assumptions 6.4.2.1 6-12 6.4.2.2 Model Details 6-13 6.4.2.3 Fluid Coupling Details 6-14 j 6.4.2.4 Stiffness Element Details 6-15 6.4.3 Whole Pool Multi-Rack (WPMR) 6-17 Model 6.4.3.1 General Remarks 6-17
! 6.4.3.2 Whole Pool Fluid Coupling 6-17 t
6.4.3.3 Coefficients of Friction 6-18 6.4.3.4 Modeling Details 6-18 6.5 Acceptance Criteria, Stress Limits and 6-19 Material Properties -
- 6.5.1 Acceptance Criteria 6-19 j 6.5e2 Stress Limits for Various Conditions 6-21
, 6.5.2.1 Normal and Upset Conditions 6-21 (Level A or Level B) i 6.5.2.2 Level D Service Limits 6-23 6.5.2.3 Dimensionless Stress Factors 6-23 6.5.3 Mat 6 rial Properties 6-24 4 6.6 Governing Equations of Motion 6-24 6.7 Results of 3-D Nonlinear Analyses of 6-26 4
Single Racks 6.7.1 Impact Analyses 6-28
! 6.7.2 Weld Stresses 6-29 6.8 Results from Whole Pool Multi-Rack Analyses 6-30 6.9 Bearing Pad Analyses 6-32 6.10 References for Section 6 6-33 7.0 ACCIDENT ANALYSIS AND MISCELLANEOUS STRUCTURAL EVALUATIONS 7.1 Introduction 7-1 7.2 Refueling Accidents 7-1 1 7.2.1 Dropped Fuel Assembly 7-1 7.2.2 Gate Drop onto the Top of a Rack 7-2
- 7.3 Local Buckling of Fuel Cell Walla 7-3 7.4 Analysis of Welded Joints in Rack 7-4 due to Isolated Hot Cell 7.5 References for Section 7 7-5 8.0 FUEL POOL STRUCTURE INTEGRITY CONSIDERATIONS 8.1 Introduction 8-1 8.2 Description of Spent Fuel Pool Structure 8-1 iii
l TABLE OF CONTENTS (continued)
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8.3 Definition of Loads 8-2 8.3.1 Static Loading 8-2
- 8.3.2 Dynamic Loading 8-2 8.3.3 Thermal Loading 8-3 8.3.4 Loads from Adjacent Structure 8-3 4 8.4 Analysis Procedures 8-3 8.4.1 Finite Element Analysis Model 8-3 8.4.2 Analysis Methodology 8-4 8.4.3 Load Combinations 8-5 j 6.5 Results of Analyses 8-6 i 8.6 Pool Liner 8-8 8.7 Conclusions 8-8 8.8 References for Section 8 8-8 5 9.0 RADIOLOGICAL EVALUATION 1
9.1 Radiological Consequences of Accidents 9-1 9.1.1 Fuel Handling Accident 9-1 9.1.2 Cask Drop Accident 9-3 9.1.3 Spent Fuel Pool Oste Drop 9-4 Accident 9.1.4 Seismic Event in the Spent 9-4 Fuel Pool d
9.2 Solid Radwaste 9-5
, 9.3 Gaseous Releases 9-6 j 9.4 Personnel Exposures 9-7 9.5 Anticipated Exposure During Reracking 9-8 9.6 Conclusions 9-10 9.7 References for Section 9 9-11 10.0 BORAL SURVEILLANCE PROGRAM 10.1 Purpose 10-1 10.2 Coupon Surveillance Program 10-2 10.2.1 Coupon Description 10-2 10.2.2 Surveillance Coupon Testing Schedule 10-3 10.2.3 Measurement Program 10-4 10.2.4 Surveillance Coupon Acceptance 10-4 Criteria 10.3 References for Section 10 10-5 11.0 ENVIRONMENTAL COST-BENEFIT ASSESSMENT 11.1 Introduction 11-1 11.2 Project Cost Assessment 11-3 11.3 Environmental Assessment 11-5 11.4 Conclusions 11-6 iv
LIST OF TABLES 1.1.1A Existing Discharge History (Total of 529 Assemblies as of Spring, 1992) 1.1.1B Pro]ected Discharge Schedule 1.1.2 Available Storage in the FCS Pool 1.1.3 Rack Module Data, Existing and Proposed Racks 2.1.1 Module Data for FCS Maximum Density Racks 2.2.1 Total Cell Data 2.2.2 Common Module Data 2.3.1 Boral Experience List (Domestic and Foreign) 2.3.2 1100 Alloy Aluminum Physical Properties 2.3.3 Chemical Composition - Aluminum (1100 Alloy) 2.3.4 Boron Carbide Chemical Composition, Weight %
Boron Carbide Physical Properties 2.5.1 Heavy Load Handling Compliance Matrix (Nurag-0612) 4.2.1 Summary of Criticality Safety Analyses 4.2.2 Reactivity Effects of Abnormal and Accident Conditions 4.3.1 Design Basis Fuel Assembly Specifications 4.6.1 Design Parameters for CEA Rod Assemblies 4.6.2 Summary of Criticality Analyses with CEA Assembly Installed 4.8.1 Effect of Temperature and Void on Calculated Reactivity of Storage Rack 5.4.1A Existing Discharge History (Total of 529 Assemblies) 5.4.1B Projected Discharge Schedule 5.4.2 Data for Discharge Scenarios 5.6.1 Data for Local Temperature
-5.7.1 Peaking Factors 5.8.1 Fuel Specific Power and Pool Capacity Data 5.8.2 SFP Bulk Pool Temperature
.5,.8.3 Time-to-Boil Results 5.8.4 Maximum Local Pool Water and Fuel Cladding Temperature for the Limiting Case (Case 1) v
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1 LIST OF TABLES (continued) 6.1.1 Listing of Pl*ats Where DYNARACK Was Applied 6.3.1 Four Sets of Time-Histories Generated from Maximum Hypothetical Earthquake (MHE) and Their Cross-Correlation Coefficients 6.3.2 Four Sets of Time-Histories Generated from Design Earthquake Response Spectra (DES) and Their Cross-Correlation coefficients 6.4.1 Degrees-of-Freedom 6.4.2 Numbering System for Gap Elements and Friction Elements 6.4.3 Spent Fuel Pool Loading 6.5.1 Rack Material Data (200*F)
- Support Material Data (200'F) 6.7.1 Results of Single Rack Analyses (List of All Runs)
! 6.7.2 Summary of Worst Results from 24 Runs of Single Rack Analysis 6.7.3 Summary Results of 3-D Single Rack Analysis for Rack Module: A 8x10 (Run I.D.: dralmheo.rf8) 6.7.4 Summary Results of 3-D Single Rack Analysis for Rack l Module: A 8x10 (Run I.D.: dralmheo.rf2)
! 6.7.5 Summary Results of 3-D Single Rack Analysis for Rack Module: A 8x10 (Run I.D.: dralmheo.hx8) 6.7.6 Summary Results of 3-D Single Rack Analysis for Rhck Module: A 8x10 (Run I.D.: dralmheo.hx2) 6.7.7 Summary Results of 3-D Single Rack Analysis for Rack Module: A 8x10 (Run I.D.:-dralmheo.hy8) 6.7.8 Summary Results of 3-D Single Rack Analysis for Rack Module: A 8x10 (Run I.D.: dralmheo.hy2) 6.7.9 Summary Results of 3-D Single Rack - Analysis for Rack Module: A 8x10 (Run I.D.: dralmheo.re8) 6.7.10 Summary Results of 3-D Single Rack Analysis for Rack Module: A 8x10 (Run I.D.: dralmheo.re2) 6.7.11 Summary Results of 3-D Single Rack Analysis for Rack Module: D 8x11 (Run I.D.: drdmheo,rf8) l vi
I LIST OF TABLES (continued) 6.7.12 Summary Results of 3-D Single Rack Analysis for Rack Module: D 8x11 (Run I.D.: drdahoo.rf2) 6.7.13 Summary Results of 3-D Single Rack Analysis for Rack Module: D 8x11 (Run 1.D.: drdmheo.hx8) 6.7.14 Summary Results of 3-D Single Rack Analysis for Rack #
Module: D 8x11 (Run I.D.: drdmheo.hx2) 6.7.15 Summary Results of 3-D Single Rack Analysis for Rack Module: D 8x11 (Run I.D.: drdmheo.hy8) 6.7.16 Summary Results of 3-D Single Rack Analysis for Rack Module: D 8x11 (Run I.D.: drdmhoo.hy2) 6.7.17 Summary Results of 3-D Single Rack Analysis for Rack Module: D 8x11 (Run I.D.: drdmheo.re8) 6.7.18 Summary Results of 3-D Single Rack Analysis for Rack Module: D 8x11 (Run I.D.: drdmheo.re2) 6.7.19 Summary Results of 3-D Single Rack Analysis for Rack Module: F 120 (Run I.D.: drfimhoo.rf8) 6.7.20 Summary Results of 3-D Single Rack Analysis for Rack Module: F 120 (Run I.D.: drfimheo.rf2) 6.7.21 Summary Results of 3-D Single Rack Analysis for Rack Module: F 120 (Run I.D.: drfimhea.hx8) 6.7.22 Summary Results of 3-D Single Rack Analysis for Rack Module: F 120 (Run I.D.: drfimheo.hx2) r 6.7.23 Summary Results of 3-D Single Rack Analysis fog Rack Module: F 120 (Run I.D.: drfimheo.hya) 6.7.24 Summary Results of 3-D Single Rack Analysis for Rack Module: F 120 (Run I.D.: drfimheo.hy2) 6.7.35 Summary Results of 3-D Single Rack Analysis for Rack Module: F 120 (Run I.D.: drfimheo.re8) 6.7.26 Summary Results of 3-D Single Rack Analysis for Rack Module: F 120 (Run I.D.: drfimheo.re2) 6.7.27 Comparison of Calculated and Allowable Loads / Stresses at impact Locations and at Welds vil
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l LIST OF TABLES (continued) 6.8.1 Maximum Absolute Displacements of Rack Corners at Both the Top and Bottom of Each Rack in Global X and Y Directions, in.
6.6.2 Maximum Absolute Displacements of Rack Corners at Both the Top and Bottom of Each Rack in Global X and Y Directions, in.
6.8.3 Maximum Rack Pedestal Vertical Loads from Whole Pool Multi-Rack Analysis (MHE)
! 6.8.4 Maximum Rack Pedestal Vertical Loads from Whole Pool Multi-Rack Analysis (DE) l 6.8.5 Ccmparison of Results from Single Rack Analysis and Whole Vool Multi-Rack Analysis (WPMR!
I 6.9.1 AVCrege Bearing Pad Pressure - Comparison of Calculated orid Allowable Streases (f, = 4000 psi) 8.5.1 Calculation of Equivalent 21ab Vertical Pressures 4
8.5:2 Bending Strength Evaluation 8.5.3 Shear Strength Evaluation 8.5.4 Results for Pile Analysis 9.1.1 Fort Calhoun Fuel Assembly Gas Gap Activity 9.1.2 Radiological Consequences of a Postulated Fuel Handling Accident 9.4.1 Typical Concentrations of Radionuclides in the Spent Fuel Pool Water 9.5.1 Preliminary Estimate o.f Person-Rem Exposures During.
Reracking 10.1 Coupon Measurement Schedule viii
LIST OF FIGURES 2.1.1 Layout of the Existing Racks in the FCS Pool 2.1.2 Proposed Modular Layout 3.2.1 Seam Welding Precision Formed Channels 3.2.2 Lead-In for Region 1 Modules 3.2.3 Composite Box Assembly 3.2.4 Assembling of Region I Boxes 3.2.5 Elevation view of a Region 1 Rack Showing Two Storage Cells 3.2.6 Adjustable Support 3.3.1 Elevation View of-Region II Rack Module 3.3.2 Typical Array of Region II Cells (Non-Flux Trap construction) 4.2.1 Limiting Burnup criteria for Acceptable Storage -in Region 2
4.3.1 Cross-Section of Region 1 Cells 4.3.2 Cross-Section of Region 2 Cells 4.7.1 Region 2 Burnup Limits Showing Fuel Currently in Storage 5.4.1 Ft. Calhoun Fuel Pool Discharge Scenario One 5.4.2 Ft. Calhoun Fuel Pool Discharge Scenario Two 5.5.1 Bulk Pool Analysis Model 5.6.1 Idealization of Rack Assembly 5.6.2 Thermal chimney Flow Model-5.6.3 Convection Currents in the Pool 5.8.1- Bulk Pool Temperature - Case 1 5.8.2 Bulk Pool Temperature - Case 2 5.8.3 Heat Load -Case 1 5.8.4 Heat. Load - Case 2:
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LIST OF FIGURES (continued) 5.8.5 Water Elevation After Loss-of-Cooling - Case 1 5.8.6 Water Elevation After Loss-of-Cooling - Case 2 6.3.1 Horizontal Response Spectrum (Internal Structure within Containment end Mat (Masa #5) 6.3.2 Vertical Response Spectrum (Containment & Auxiliary Building at Any Elevation) 6.3.3 Horizontal Response Spectrum (Internal Structure within Containment and Mat (Mass #5) 6.3.4 Vertical Response Spectrum (Containment & Auxiliary Building at Any Elevation) 6.3.5 Maximum Hypothetical Seismic Synthetical Acceleration Time-History: Set-1-H1 6.3.6 Maximum Hypothetical Seismic Synthetical Acceleration Time-History: Set-1-H2 6.3.7 Maximum Hypothetical Seismic Synthetical Acceleration Time-History: Set-1-VT 6.3.8 Maximum Hypothetical Seismic Synthetical Acceleration Time-History: Set-2-H1 6.3.9 Maximum Hypothetical Seismic Synthetical Acceleration Time-History: Set-2-H2 6.3.10 Maximum Hypothetical Seismic Synthetical Acceleration Time-History: Set-2-VT 6.3.11 Maximum Hypothetical Seismic Synthetical Acceleration Time-History: Set-3-H1 6.3.12 Maximum Hypothetical Seismic Synthetical Acceleration Time-History: Set-3-H2 6.3.13 Maximum Hypothetical Seismic Synthetical Acceleration Time-History: Set-3-VT 6.3.14 Maximum Hypothetical Seismic Synthetical Acceleration Time-History: Set-4-H1 6.3.15 Maximum Hypothetical Seismic Synthetical Acceleration Time-History: Set-4-H2 6.3.16 Maximum Hypothetical Seismic Synthetical Acceleration
-Time-History: Set-4-VT X
_ - . - _ _ _ _ _ . = _ _ _ _ _ . _ _ _ _ _ _ - _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ ___ _ -
i LIST OF FIGURES (continued) l 6.3.17 Average Response Spectrum from 4 Synthetic Time Histories and the Response Spectrum of Maximum Hypothetical Seismic, Horizontal H1, 0.5 percent equipment damping 6.3.18 Average Response Spectrum from 4 Synthetic Time Histories and the Response Spectrum of Maximum Hypothetical Seismic, Horizontal H2, 0.5 percent equipment damping 6.3.19 Average Response Spectrum from 4 Synthetic Time Histories '
and the Response Spectrum of Maximum Hypothetical Seismic, Vertical VT, 0.5 percent equipment damping 6.3.20 Design Earthquake, Synthetica1 Acceleration Time-History:
Set-1-H1 6.3.21 Design Earthquake, Synthetical Acceleration Time-History:
Set-1-H2 6.3.22 Design Earthquake, Synthetical Acceleration Time-History:
Set-1-VT 6.3.23 Design Earthquake, Synthetical Acceleration Time-History:
Set-2-H1 i 6.3.24 Design Earthquake, Synthetical Acceleration Time-History:
Set-2-H2 6.3.25 Design Earthquake, Synthetica1 Acceleration Time-History:
Set-2-VT I
6.3.26 Design Earthquake, Synthetical Acceleration Time-History:
i Set-3-H1 6.3.27 Design Earthquake, Synthetical Acceleration Time-History:
Set-3-H2 f
- 6.3.28 Design Earthquake, Synthetical Acceleration Time-History
j Set-3-VT ,
j 6.3.29 Design Earthquake, Synthetical Acceleration Time-History:-
L Set-4-H1 i 1
6.3.30 Design Earthquake, Synthetical Acceleration Time-History:
1 Set-4-H2 6.3.31 Design Earthquake, Synthetical Acceleration Time-History:
! Set-4-VT .
r f XI-v,.m-+-..e- ,- ,, , - - .
,-,%- -~,-...M , . _
- [. w , -...~.,w n.- o - - . .+-4.r..- . - - - + e- ,- - , . . - =
j i
I j LIST OF FIGURES (continued) ,
I 6.3.32 Average Response Spectrum from 4 Synthetic Time-Histories and Design Response Spectrum, Horizontal H1, 0.5 percent j equipment damping i
6.3.33 Average Response Spectrum from 4 Synthetic Time-Histories 4
and Design Response Spectrum, Horizontal H2, 0.5 percent ;
equipment damping 4
j 6.3.34 Average Response Spectrum from 4 Synthetic Time-Histories i and Design Response Spectrum, Vertical VT, 0.5 percent
- equipment damping 1
6.3.35 Induced Horizontal-and Vertical Accelerations (Maximum l Hypothetical Earthquake) i
- 6.3.36 Induced Horizontal and Vertical Accelerations (Design i Earthquake)
} 6.4.1 Pictorial-View of Rack Structure l-6.4.2 Schematic Model for DYNARACK 6.4.3 Rack-to-Rack Impact Springs i
6.4.4 Fuel-to-Rack Impact Springs 6.4.5 Degrees-of-Freedom Modeling Rack Motion j 6.4.6 Rack Degree-of-Freedom for Y-Z ? lane. Bending i
j 6.4.7 Rack Degree-of-Freedom for X-Z Plane Bending 6.4.8 2-D View of Rack Module
- 6.4.9 Pool Layout and Rack Pedestal Numbering I
, 6.8.1 Gap Time-History (Gap between Rack G1 and East. Wall, j
South Side, Top, MHE-Set-4
) 6.8.2 Gap Time-History (Gap between Rack G1 and Rack G-2, East 4
Side, Top, MHE-Set-4 6.8.3 Gap Time-History (Gap between Rack G1 Rack.E, South Side, Top, MHE-Set-4 i
j 6.8.4- Gap Time-History (Gap between Rack E and West Wall, South Side, Top, MHE-Set-4 7.3.1 Loading on Lower Portion of Fuel Cell Wall i
xii
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LIST OF FIGURES (continued) i j
7.4.1 Welded Joint in Rack 8.2.1 FCS SFP FE Model - View from Southeast 8.2.2 FCS SFP FE Model - View from Northwest 11.1 Expected Net Present Value Cost Analysis ($1000) - Base Case i
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4 f
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1.O INTRODUCTION The Fort Calhoun Station (FCS) is a pressurized light water nuclear power reactor installation owned by the Omaha Public Power District (OPPD). FCS is located on the banks of the Missouri River at a distance of approximately 19 miles north of the city of Omaha, Nebraska. FCS received its construction permit from the NRC (formerly AEC) in June, 1968, and went critical for the first time in August, 1973. Shortly thereafter, in September, 1973, FCS went into commercial operation. The FCS fuel storage system is made up of a spent fuel pool 399 inches long and 247 inches wide with an integral cask laydown area. The pool presently contains 729 spent fuel storage locations. After the Spring, 1992 refueling, a total of 529 fuel bundles will be stored in the FCS pool. Since the full core consists of 133 fuel assemblies, maintaining full core offload capability implies that 596 storage cells (729 minus 133) are available for normal offload storage. Table 1.1.1A provides the data on previous fuel assembly discharges and Table 1.1.1B provides the projected fuel assembly discharges in the FCS spent fuel pool.
Table 1.1.2, constructed from Tables 1.1.1A and 1.1.1B data, indicates that FCS will lose full core discharge capability after the scheduled refueling in 1995. This projected loss of full core discharge capability prompted the present undertaking to increase spent fuel storage capability in the FCS pool.
The purpose of this submittal is to request the authorization to rerack the FCS pool and equip it with new poisoned high density storage racks containing 1083 storage cells.
The principal owner and operator of FCS, the Omaha Public Power District, entered into a contract with Holtec International of Cherry Hill, New Jersey, in February, 1992 to design, procure material, fabricate, deliver and install high density spent fuel racks in the FCS fuel pool. This licensing document has been prepared by the Omaha Public Power District and its contractor, Holtec International of Cherry Hill, New Jersey.
1-1
After the proposed reracking, eleven free-standing poisoned rack modules positioned in the spent fuel pool with a prescribed and geometrically controlled gap between them will contain a total of 1083 storage cells. The design and construction of the storage cells is described in Section 3 of this licensing document. Out of the total of 1083 cells, 160 cells in two rack modules are flux-trap cells *, and the remaining cells are of the so-called non-flux trap type. The storage modules suitable for storing fresh fuel (up to 4.2% enrichment) are located adjacent to the refueling canal to facilitate the transfer of fuel between the pool and the reactor core. Consistent with the, concept of Discrete Zone Two Region storage, the placement of fuel with a given burnup in the allowable location is administrative 1y controlled. No credit is taken for soluble boron in normal refueling and in full core offload storage conditions.
It is noted that the proposed reracking effort in 1994 will increase the number of licensed storage locations to 1083 and, as indicated in Table 1.1.2, will extend the date of loss of full core discharge capability (133 bundles) through the year 2007. The batch size projections in Table 1.1.2 were estimated conservatively (45 bundles 0 18 month cycles).
Table 1.1.3 presents key comparison data for existing and proposed maximum density rack modules for FCS.
The new spent fuel storage racks are free-standing and self supporting. The principal construction materials are SA240-Type 304 stainless steel sheet and plate stock for the new racks and SA564-630 (precipitation hardened stainless steel) for the adjustable support spindlea. The only non-stainless material A flux-trap construction implies that there is a water gap between adjacent storage cells such that the neutrons emanating from a fuel assembly are thermalized before reaching an adjacent fuel assembly.
1-2
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, utilized in the rack is the neutron absorber material which is boron carbide and aluminum-composite sandwich available under the patented product name "Boral".
j The new racks are designed and analyzed in accordance with Section i
III, Divicion 1, Subsection NF of the ASME Boiler and Pressure Vessel (B&PV) Code. The meterial procurement, analysis, and fabrication of the rack modules conform to 10CFR 50 Appendix B requirements.
i This Licensing Report documents the design and analyses performed to demonstrate chat the new spent fuel racks satisfy all governing l requirements of the applicable codes and standards, in particular,
, "OT Position for Review and Acceptance of Spent Fuel Storage and i Handling Applications", USNRC (1978) and 1979 Addendum thereto.
4 The safety assessment of tha proposed rack modules involved demonstration of their thermal-hydraulic, structural, and
- criticality adequacy. Hydrothermal adequacy requires that fuel cladding will not fail due to excessive thermal stress, and that the steady state pool bulk temperature will remain within the limits prescribed for the spent fuel pool to satisfy the pool structural strength constraints. Demonstration of structural adequacy primarily involves analyses showing that the primary I
stresses in the rack module structure will remain below the ASME B&PV Code (subsection NF) allowables under the postulated seismic events. The structural qualification also includes analytical demonstration that the subcriticality of the stored fuei will be maintained under accident scenarios such as fuel assembly drop, accidental misplacement of a fuel assembly outside a rack, drop of a gate, etc. Structural consequences of these postulated accidents are evaluated and presented in Section 7 of this report.
l 1-3
. . .. - . . . . - _ . - .. -. .- -. . .- ~- -. .
The criticality safety analysis shows that the neutron multiplication factor for the stored fuel array'is bounded by_the USNRC limit of 0.95 (OT Position Paper) under assumptions of 95%
probability and 95% confidence. Consequences of the inadvertent placement of a fuel assembly are also evaluated as part of the criticality analysis. The criticality safety analysis synopsized in Section 4 sets the requirements on the length of the Boral panel and the areal B-10 density.
This Licensing Report contains documentation of the analyses performed to demon 9trate the-large margins of safety with respect to all'USNRC specified criteria. This report also contains the results of the analysis performed _to demonstrate the integrity of the fuel pool reinforced concrete structure, and an appraisal of radiological- considerations. A summary of the cost / benefit consideration demonstrating reracking'as the mor' cost effective approach to increase the on-site storage capacity- _of the Fort Calhoun Station is also included in this report.
This rerack application is limited to_ storing regular (unconsolidated) f ut:1. The District, however, has performed analyses to confirm the structural adequacy of the new racks assuming that the fuel assembly weight is - 2480 lbs. ,-- simulating
-the condition of consolidated storage.
All computer programs utilized in performing the- analyses documented in this licensing -report are identified _in -the appropriate sections.- All computer codes are _ benchmarked and verified in accordance with Holtec International's nuclear Quality Program.
)
l 1-4 l
f i
.i . ' .
The analyses presented herein clearly demonstrate that the rack
! module arrays possess wide margins of-safety from-all four vantage points: thermal-hydraulic, -criticality, structural. and ,
j radiological. The No-Significant-Hazard Considerations submitted
- l to the Commission along with this Licensing Report is based on the descriptions and analyses synopsized in the subsequent sections of-3 this report.
i f
i i
l 1
l l
i I
I 4
4 t
i k
l I
E i
i t
t 4
(
?
5 4
4 l 1-5 L ;
, , , . ,.c s ,,.,,,7,,.,,,.,,,mv. ,.,-,g 9 , 9. w ,9 3 .n y
, . - - - . - . - . . - . . - .- . . - . . . - - ~ _.- .-- . .. - . - -- . - . . . ..
4 4
l J .
j i Table 1.1.1A- ,
EXISTING DISCHARGE HISTORY _l
}
(Total of 529 Assemblies as of Spring, 1992) f i
Bounding Full Months ;
s- Power . After l Cycle Number of Operation Discharge Previous
- j Number Assemblies Days Date .Discharae j 1 25 1575 - 04/10/75 0 *
! 2 36 1575 - 12/08/76 19.9 i
i 3 52 1575 11/16/77 11.3
! 4 44 1575- 12/10/78 - 12.8 4
i
- 5 40 1575 04/16/80 16.2
} 6 40 1575 - 12/01/81~ 19.5
- 7 20 1575 02/18/83 14.6 i
! 8 2f 1575 06/18/84 R1'6 . 0 T
l 9 65 1575 12/16/85 _ 18.0 i- 10 45 1575 03/23/87 . 15.2
) 11 44 1575 10/30/88 - 19.3 U 12 40 1575. 02/15/90 15.5 13 52 1575- 02/15/92 24.0 l
5 i
i 6
- -- .-
- r .,~, , , , - -mw -' e,-e....., e - ,, ,,-y--w w ,>mmr.r-,-.e.-:w., -+re .-,,.E-,,ry,,,y or -rvsy,.,p.-
I 5-i
! Table 1.1.1B
- PROJECTED DISCHARGE SCHEDULE t
! Assumed l= Full i Month / Power. Number'of Cumulative l Cycle Year of Operation Assemblies Assemblies i Number Discharae Days Discharced Discharced
} 14 8/1993 1575 45 574 i
?
15 2/1995 1575 45 619 16 8/1996 1575 45 664 !
17 2/1998 1575 45 709 18 8/1999 1575 45 754 19 2/2001 1575 45 799 20 8/2002 1575 45 844 21 2/2004 1575 45 889 22 8/2005 1575 45 934 23 2/2007 1575 45 979 24 8/2008 1575 45. 1024 25 2/2010 1575 45 1069-26 8/2011 .-1575- 45 1114--
27 2/2013 1575- -45 1159 28 8/2014- 1575 45 -1204-4 1 -!
i l 1
l Table 1.1.2 AVAILABLE STORAGE IN THE FCS POOL l WITH PRESENT LICENSED CAPACITY FUEL EOC AFTER RERACKING 4 CYCLE YEAR (729 CELLS) (1083 CELLS) e l 13 1992 200 559 14 1993 155 514 15 1995 110 469
, 16 1996 424 17 1998 379 4 18 1999 334 19 2001 ?89 20 2002 244 21 2004 199 22 2005 154 23 2007 -109 t
Loss of full core reserves.
1-8
! l j 1 I
. \
l k
i i
4 5-
- 1 g
1 i
i Table 1.1. 3 :
RACK MODULE DATA, EXISTING AND PROPOSED RACKS i
! PROPOSED. MAXIMUM l' ITEM EXISTING RACKS DENSITY' RACKS l-1 Number of cells 729' - 1083 i
Number ~of modules- 11- 12:
, Neutron Absorber- Boraflex- Boral-i Approximate, cell 9.935 ' 8.652 (non-flux trap) i
- pitch, inches . 9.821 ' (flux trap. in .
- E-W directions)-
.- . 10.'363-(flux trap in
. N-S direction).
4 p-i f
P t
a P
t
--- l --
4 s
t F - -m -. * -* < - * * -v
l 1
2.0 MODULE DATA 2.1 Backaround 1
The Fort Calhoun Station pool (FCS) is of rectangular planform section, approximately 399 inches x 247 inches with a slightly recessed integral cask pit of 8 feet square cross-section located in the northwest corner of the pool. The nominal elevation of the fuel pool liner is 995'-6" and that of the cask pit liner is 993'-
6". The existing arrangement of the rack modules in the FCS pool is illustrated in Figure 2.1.1. Figure 2.1.2 shows the proposed layout. This layout features a total of 1083 storage cells in eleven free-standing rack modules. The new densified storage in the FCS pool calls for no major physical alterations. However, the SFP cooling suction and sparger pipes will be truncated to the elevation of the top of racks. Two major tool brackets will be relocated from their existing locations on the south wall to the cask pit. Table 2.1.1 provides the cell count and module dimension data.
It is noted from Table 2.1.1 that the rack modules for the FCS pool have been proportioned to have an aspect ratio (length to width ratio) in the range of 1 to 2, which ensures a large margin of safety against kinematic instability under seismic events.
The hollow circles in the planview of the modules in Figura 2.1.2 shows the location of supports. The support pedestals adjacent to the cask region are deliberately set one cell inboard to maximize the distance from the cask pit edge.
f OPPD does not contemplate permanently installing any modules in the cask pit at the present time.
2-1 i
j i
The existing module layout in the FCS pool in Figure 2.1.1 illustrates the utilization of the FCS spent fuel pool space l typical of the rack installations in the late 1970s. Unfortunately, the unutilized pool floor space, while large and extant, is not
- available in a form such that additional modules can be added to the pool.
2.2 Multi-Reaion Storac_q i The high density spent fuel storage racks in the FCS pool will provide storage locations for up to 1083 fuel assemblies and will l be designed to maintain the stored fuel, having an initial enrichment of up to 4.2 wt% U-235, in a safe, coolable, and subcritical configuration during normal and abnormal conditions.
l All rack modules for the FCS spent fuel pool are of the " free-standing" type such that the modules are not attached to the pool floor nor do they require a.1y lateral braces or restraints. These rack modules will be placed in the pool in their designated i locations, and the support legs remotely leveled (using a telescopic removable handling tool) by an operator on the fuel handling bridge- No additional lifting equipment is needed to carry the weight of a rack while leveling is being performed.
The racks will be arranged in two regions in-the spent fuel pool.
Region I will have 160 storage locations and capable of storing
, unirradiated fuel of up to 4.2 wt% U-235 initial enrichment.
4 Region I has locations in excess of what is required to store a full core discharge (133 assemblies) from the reactor plus a normal discharge. Table 2.2.1 provides cell data-for Region I and Region II racks. Region II will have 923 locations for storage of-fuel which meets enrichment and burnup criteria developed as part of the rack design. Certain locations in Region II can also be used to j 2-2
store fresh fuel so long as sufficient number of fuel assemblies
, with a prescribed minimum burnup surrcund it. Table 2.1.1 provides key module data for both Region I and Region II racks. Table 2.2.2 contains design data common to both Regions I and II modules.
Each FCS rrck module is supported by a minimum of four legs which are remotely adjustable. Thus, the racks can be made vertical and
] the top of the racks can easily be made co-planar with each other.
The rack module support legs are engineered to accommodate variations of the pool floor. The placement of the racks in the spent fuel pool has been designed to preclude any support legs from being located over existing obstructions on the pool floor.
2.3 Material Considerations 2.3.1 Introduction 1
Safe storage of nuclear fuel in the FCS spent fuel pool requires that the materials - utilized in the fabrication of racks be of proven durability and be compatible with the pool water
, environment. This section provides the necessary information on this subject.
2.3.2 Structural Materials
^
The following structural materials are utilized in the fabrication of the spent fuel racks:
- a. ASME SA240-304 for all sheet metal stock.
- b. Internally threaded support legs: ASME SA240-304.
i c. Externally threaded support spindle: ASME SA564-630 precipitation hardened stainless steel (heat- treated to 1100 F).
.d. Weld caterial - per the following ASME specification: SFA 5.9 R308L.
2-3
4 i
2.3.3 Poison Material
! 1 i
j_ In addition to the structural and non-structural stainless-l material, the racks employ Boral*, a patented product - of AAR l Brooks and Perkins, as the neutron absorber material. A brief i description of Boral, and its fuel pool experience list follows.
i j Boral is a thermal neutron poison material composed of boron carbide and 1100 alloy aluminum. Boron carbide is a compound having
- a high boron content in a physically stable and chemically inert.
l form. The 1100-alloy aluminum'is a light-weight metal with high-2 tensile strength which is protected from corrosion by a highly I
i resistant oxide film. The two materials, boron carbide and j aluminum, are ch'emically compatible and ideally' suited for-long-term use in the radiation, thermal and chemical-environment'of a l
j nuclear reactor or the spent. fuel pool.
i 1
I l Boral's use in the spent- fuel pools as the neutron absorbing material can be attributed to the following reasons:
i
' (i) The content and placement of boron carbide provides a very high removal cross - = section for - thermal neutrons.
(ii) Boron carbide, in the form of fine particles, is homogeneously dispersed throughout the central layer i of-the Boral panels.
f (iii) The-boron carbide and aluminum ~materialo in Boral j are totally unaf fected by long-term exposure to
~
radiation.
{- (iv). - The neutron absorbing central- layer of Boral is clad
}; with permanently bonded surfaces of aluminum.
, (v) "B'o ral' -
is stable, strong, durable, and corrosion resistant.-
3
(- 4 2
1-2-4 4
_. m .. _ .- _ . _ .. . - _ . _ . _ :. - . _
I 4 Holtec International's Q.A. program ensures that Boral is 1 j manufactured by AAR Brooks is Perkins under the control- and i surveillance of.a Quality Assurance / Quality Control Program that
! conforms to the requirements of 10CFR50 - Appendix B, " Quality
! Assurance Criteria for Nuclear Power Plants".
i l As indicated in Table 2.3.1, Boral has been licensed by the USNRC for use in numerous BWR and PWR spent fuel storage racks and has been extensively used in overseas nuclear installations.
i Boral Materia 1 Characteristics 4
Aluminum: Aluminum is a silvery-white,- ductile metallic element-
] that is the most abundant in the earth's crust . . The_1100 alloy .
aluminum is used extensively in heat exchangers, pressure and
{ storage tanks, chemical equipment, reflectors and sheet metal work.
! It has high resistance to corrosion in industrial and marine
- atmospheres. Aluminum has atomic number of 13, atomic weight of
- 26.98, specific gravity of 2.69 and valence of 3. . The physical, f mechanical and chemical properties of the 1100 alloy Sluminum are listed in Table 2.3.2 and 2.3.3.
The excellent corrosion resistance of the 1100. alloy aluminum is-provided by the protective' oxide film that develops on-its surface f from exposure to the= atmosphere or water. 'This film. prevents the i- loss-of metal from general corrosion or pitting corrosion.
i 1
i i.
Boron . Carbide : The boron carbide contained in Boral is a fine <
l' . granulated powder that conforms-to ASTM C-750-80-nuclear grade Type:
III. The particles range in size between 60 and:200 mesh and the-j- material conforms to the chemical composition and properties listed j in Table 2.3.4.
+
i 2-5 a
l i
L
, , , v w-, - < e- - . , # ._.,--..-,,i..-.,o. _ . . , , , ,~...~,w - -
,,.%., y .*wn-- ,wr. .c-+ ,y-.-
1 I
2.3.4 Comoatibility with Coolant
]
4 All materials used in the construction.of the FCS racks have an established history'of-in-pool usage. Their physical, chemical 3 and radiological compatibility with the pool environment is an.
l established fact at this time. As noted in Table 2.3.1, Boral has
- been used in both vented and unvented configurations in fuel pools j with equal success. Austenitic stainless steel is'perhaps the most I
widely used stainless alloy in nuclear power plants.
4 2.4 Existino Rack Modules and Procosed Rerackina Operation
{
i j The reracking of the FCS pool will be carried out in the February,
! 1994 - August, 1994 time ~ frame. At the time of the reracking l operation, there will-be 574 spent fuel bundles in the FCS pool i leaving only 155 open locations. -A rack change-out and fuel I
reshuffle -scheme is being developed by . the Omaha Public Power-l District -(OPPD) which eliminates the potential of. any damage to the l- stored fuel during the handling operations: associated with the l- rerack work effort. The details of the " defense-in-depth" approach
- are provided in-the following.
l It is OPPD's intention'to perform all field operations remotely .
l insofar as it is consistent with_the"ALARA objectives. In other l words, divers will be' utilized to carry out an in-pool operation f only if such an approach-demonstrably minimizes the total radiation :
exposure-to the personnel.
! 2.5 Heavy Load Considerations for - the - Procosed Rerackinc i operation-i The Auxiliary Building Crane 1will be utilized. for handling all~
- heavy - loads -in ' the reracking - operation.: . The Auxiliary i Building -
Cranelis?a double, girder overhead bridge crane With a.75 ton main 2-G 4
i l4 hook which is qualified as single failure proof in accordance with the provisions of NUREG-0612. This crane-is also equipped with a 10 ton auxiliary hook which is not' single failure proof.
l A remotely engagable lift rig, meeting NUREG-0612 stress criteria, will be used to lift the empty _ modules (old and new). The rig
- designed for handling the FCS racks is identical in its physical attributes to the rigs utilized to rerack Millstone Unit One
- (198P), Vogtle Unit Two (1989), Indian Point Unit Two (1990),
Ulchin Unit Two (1990), Hope Creek-(1990), Laguna Verde Unit One 1
(1990), Kuosheng (1991), Three Mile Island Unit 1 (1992) ,- Zion (1993), and J.A. FitzPatrick (1992). The rig consists of independently loaded lif t rods with a " cam type" lif t configuration which ensures that failure of one traction rod will-not result in
- uncontrolled lowering of the load being carried by the rig (which l complies with the duality feature - called ' for in Section 5.1.6 (1) of NUREG 0612). The rig has the following additional attributes
- a. The stresses in the lift rods are self limiting inasmuch l as an increase in the magnitude of the load-reduces the eccentricity between the upward force and downward reaction (moment arm).
- b. It is impossible for a traction rod to lose its engagement with the rig in locked position. Moreover, the locked configuration can be directly verified from above the pool water without the aid of an underwater camera.
- c. The stress analysis of the rig isL carried out using a finite element code, and the primary stress -limits postulated in ANSI 14.6 (1978) are shown to be met.
i
- d. The rig is load tested with-150% of the maximum weight to be lifted. The test weight is maintained.in the air l- for one hour. All critical weld joints are liquid
, penetrant examined, after the load test,: to establish the soundness of all' critical joints.
i 2-7 4
- -- . - - .,w . ,c. , - --
Pursuant to the defense-in-depth approach of NUREG-0612, the
! following additional measures of safety will be undertaken for the raracking operation.
(i) The cranes used in the project will be given a preventive maintenance checkup and inspection per the FCS maintenance procedures before the beginning of the reracking operation.
(ii) The cranes used will lift no more than 50% of their
, rated capacity at any time during the reracking operation. Table 2.1.1 provides dimensions and weight data on the new FCS modules.
(iii) The old fuel racks will be lifted no more than 6 inches above the pool floor and held in that elevation .for approximately 10 minutes before beginning the vertical lift.
I (iv) Safe load paths will be developed for moving the old and new racks in the Auxiliary Building. The
, "old" or "new" racks will not be carried over any region of the pool containing fuel.
(v) The rack upending or laying down will be carried out in an area which is not overlapping to any safety related component.
(vi) All crew members involved in the use of the lifting and upending equipment will be given training.
1 The fuel reshuffle scheme developed for the spent fuel pool corresponding to the rack change-out presented in the preceding section is predicated on the following criteria:
(1) No heavy load (rack or rig) with a potential to drop on a rack has less than 3 feet lateral free zone clearance from active fuel.
(2) All heavy loads are lifted in such a manner that the C.G.
of the lift point is aligned with the C.G. of the-load being lifted.
2-8
4 l l l 4
)
- - l I -
l (3) Turnbuckles are utilized to " fine tune" the verticality of the rack being lifted.
All phases of the reracking activity will' be conducted- in '
l accordance with written procedures which will- be reviewed and f approved by OPPD.
i j Our proposed compliance with the objectives of NUREG-0612 follows
! the guidelines contained in section 5 of that _ document. The l guidelines of NUREG-0612 call for measures to " provide an adequate defense-in-depth for handling of heavy loads n' ear spent fuel...".
f _
The NUREG-0612 guidelines cite four. major causes of load handling l
accidents, namely 4
! i. operator errors i 11. rigging failure 4
111. lack-of adequate inspection j iv. inadequate procedures The FCS rerack program ensures maximum emphasis to mitigate the potential load drop accidents by implementing measures to eliminate shortcomings in all aspects of the operation including-.the.four aforementioned areas. A summary of the measures - specifically planned to deal with the major causes is provided below.
Operator errord: As mentioned above, OPPD plans to provide-comprehensive training to the installation crew.
Ricaina failure:- The lif ting device designed for handling an:1 installation of the racks in the FCS fuel = pool has-redundancies:in the lift legs,.and lift eyes such that there are four independent load members. Failure of any one load bearing member would not lead
-to uncontrolled lowering of the' load. ~The rig-complies with all provisions of ANSI 14.6 -
1978, including compliance with the primary stress criteria', load testing at 150% of maximum lif t load, and. dye' examination of critical welds;-
2-9
d l
The FCS rig design is similar to the rigs used in the rerack of l numerous other plants, such as Hope Creek, Millstone Unit 1, Indian Point Unit Two, Ulchin II, Laguna Verde, J. A. FitzPatrick and Three Mile Island Unit 1.
Lack of adecuate insoection: The designer of the racks will develop a set of inspection points which have proven to have eliminated any incidence of rework or erroneous installation in numerous prior rerack projects.
i Inadeauate orocedures: FCS plans over twenty operating procedures to cover the entire gamut of operations pertaining to the rerack 4 effort, such as mobilization, rack handling, upending, lifting,
- installation, verticality, alignment, dummy gage testing, site safety, and ALARA compliance. Procedures for both old racks and new racks will be developed.
The series of operating procedures planned for FCS rerack are the successor of the procedures implemented successfully in other i
projects in the past.
In addition to the above, a complete inspection of the fuel handling bridge and the Auxiliary Building Crane used in the project and relubrication of their moving parts prior to the start of reracking are planned. Safe load paths have been developed as
]
- required by NUREG-0612.
Table 2.5.1 provides a synopsis of the requirements delineated in i NUREG-0612, and our intended compliance.
In summary, the measures implemented in FCS reracking are identical to the those utilized in the most recent successful rerack projects (such as Indian Point Unit 2, concluded in October, 1990, Hope Creek, completed in March, 1992, and Three Mile Island, Unit 1 completed in September, 1992.
2-10
i
- i. '
j i- ,
j .t i-s r
- =
2 h
i .
1 Table 2.1.1 j -MODULE DATA FOR FCS MAXIMUM DENSITY RACKS.
I 1 >
i' g MODULE ENVELOPE
- NUM8ER OF SIZE (inches). SHIPPING MODULE N-S EW CELLS- . WEIGHT-l 1.0. DIRECTION DIRECTION. PER RACK - NS. E-W . (Lbs.) .
MODULE TYPE.
{;- A1. . 10 8 ' 80 ' 102.06 .77.4_ 16000' Region 1-1
{
A2 10 8 - 80_ 102.06 .s 77.4 - 16000- Reglon I i
[ st 12 9 108- 104.16 78.12 15200-- - segfon Il y 82 - 12 9 108 104.16 78.12 15200' Region II
-n j G1 10. 9 90 86.8 78.12 12600 Region 11
~
j . G2 10, 9_ 90 86.8 78.12 12600' - Region 11 f_ C 11 -' 9 99 95.48 78.12 13900 .. Region 11
! O- 11- 8 88 95.48 69.44 12400 Region II-E 10- -10 100 86.8 ~86.8 14000 Region II: '
~
4
- F1 12 10 120- 104.16 - 86;8 - 16800 . Region II 4
F2 12 - 10 120 104.16 56.8 ' c16800 Reglon II-i J
a 1-i 4
2-11 4
A
' ii,
. . . . . . . _ .. , , _ , , , , . _ _ _ _ - ,,-,.-.,.;,;.,-.e..,J-,--_,,
i 4
j 1
1 4
i Table 2.2.1 TOTAL CELL DATA l TOTAL NUMBER j REGION NUMBER OF RACKS 'OF CELLS i
I (Flux-Trap) 2 160 923 II (Non-Flux Trap) 9 i
GRAND TOTAL: 1083 4
j 1
4 2-12 1
w - ,- --
y
i a
1 j.
4 i
i i *
) Table 2.2.2
! COMMON MODULE DATA 4 i-
!. . Storage cell inside
. dimension (nominal): 18.46 inch l ' Storage cell heigt? 161 inch (above'the baseplate):
a j Baseplate thickness:: H0.75~ inch i
Support leg height: 1/4=inchL(ncminal) b-Support leg type: Remotely adjustable 11egs
. with lateral gussets'.
' Number of support legs: 4-(minimum) f Remote lifting and:
- < . handling provision
- Yes
, Poison material: Boral Poison' length:. . 128-inch- i i
F -
Poison width: 7.25 inch Cell Pitch: -8.652.inchL(Region II) i _9.821 inchE(Region 1I, in t- E-W-direction)- ,
' 10.363 inch (Region I, in i-N-S direction)'
4 i
4 i
- f. .
t
- 2-13 i-3 f,
t v V + w w -rm- - -- --s o rw r r--see s ---#r-a < - rv e . * - -+*4
Table 2.3.1 BORAL EXPERIENCE LIST (Domestic and Foreign)
Pressurized Water Reactors Vented Construc- Mfg.
Plant Utility tion Year Bellefonte 1,2 Tennessee Valley Authority No 1981 Donald C. Cook Indiana & Michigan Electric No 1979 Indian Point 3 NV Pcuer Authority Yes 1987 Maine Yankee Maine Yankee A'omic Power Yes 1977 Salem 1, 2 Public Service Elec & Gas No 1980 Sequoyah 1,2 Tennessee Valley Authority No 1979 Yankee Rowe Yankee Atomic Power Yes 1964/1933 Zion 1,2 Commonwealth Edison Co. Yes 1980 Byron 1,2 Commonwealth Edison Co. Yes 1988 Braidwood 1,2 Commonwealth Edison Co. Yes 1988 Yankee Rowe Yankee Atomic Electric Yes 1988 Three Mile Island I GPU Nuclear Yes 1990 4 Boiling Water Reactors Browns Ferry 1,2,3 Tennessee Valley Authority Yes 1980 Brunswick 1,2 Carolina Power & Light Yes 1981 Clinton Illinois Power Yes 1981 Cooper Nebraska Public Power Yes 1979 j Dresden 2,3 Commonwealth Edison Co. Yes 1981 Duane Arnold Iowa Elec. Light & Power No 1979 J.A. Fitzpatrich NY Power Authority No 1978 E.I. Hatch 1,2 Georgia Power Yes 1981 Hope Creek Public Service Elec & Gas Yes 1985 Humboldt Bay Pacific Gas & Electric Yes 1986 Lacrosse Dairyland Power Yes 1976 Limerick 1,2 Philadelphia Electric No 1980 Monticello Northern States Power Yes 1978 Peachbottom 2,3 Philadelphia Electric No 1980 Perry, 1,2 Cleveland Elec. Illuminating No 1979 Pilgrim Boston Edison No 1978 Susquehanna 1,2 Pennsylvania Power & Light No 1979 Vermont Yankee -Vermont Yankee Atomic Power Yes 1978/1986 Hope Creek Public Service Elec & Gas Yes 1989 Shearon Harris Carolina Power & Light Yes 1991 Pool B 2-14
Table 2.3.1 (continued)
)
Foreign Installations Using Boral France 12 PWR Plants Electricite de France South-Africa Koeberg 1,2 _ESCOM Switzerland Beznau 1,2 .Nordostschweizerische-Kraftwerke AG -
Gosgen -Kernkraftwerk:Gosgen-Daniken AG Taiwan Chin-Shan 1,2 Taiwan Power Company Kuoshengfl,2 Taiwan Power Company.
Mexico Laguna Verde Comision Federal de---Electricidad:
Units 1 & 2
-i
'2-15
l l
l Table 2.3.2 l
1100 ALLOY ALUMINUM PHYSICAL PROPERTIES Density 0.098 lb/cu. in.
2.713 gm/cc Melting Range 1190-1215--deg. F 643-657 deg. C Thermal Conductivity 128 BTU /hr/sq ft/deg..F/ft (77 deg. F) 0.53 cal /sec/sq.cm/deg. C/cm Coef. of Thermal 13.1 x 10 4 in/in., *F Expansion 23.6 x 10 4 cm/cm, C (68-212 deg. F)
Specific heat 0.22 BTU /lb/deg. F-(221 deg. F) 0.23 cal /gm/deg. C Modulus of 10x106 psi L Elasticity i
Tensile Strength 13,000 psi annealed (75 deg. F) 18,000 psi as rolled l Yield Strength 5,000 psi annealed (75 deg. F) 17,000 psi as rolled Elongation 35-45% annealed (75 deg. F) 20% as rolled l
Hardness (Brinell) 23 annealed 32 as rolled-Annealing Temperature 650 deg. F-l 343 deg. C l
2-16
i 4
s E t i
Table 2.3.3 CHEMICAL COMPOSITION - ALUMINUM (1100 ALLOY) i
^
99.00% min. -Aluminum 1.00% max. Silicone and Iron-l 0.05-0.20% max. Copper
.05% max.- Manganese
.10% max. Zinc l .15% max. others each J
4.
a i
b i
2-17 e
,,. , -,-c-'-- y
?
4
?
! 1 l
2 1
4 T
?
! Table 2.3.4 i
l ?
j BORON CARBIDE CHEMICAL COMPOSITION. WEIGHT %
e i
j- Total-boron -70.0 min.
! B isotopic content in 18.0 natural boron -
e Boric oxide 3.0 max.
4 l-4 Iron l2.0-max.
t.
l Total boron plus 94.0 min..
3 total carbon i
! BORON-CARBIDE PHYSICAL PROPERTIES j Chemical formula- .BC 4 Boron content-(weight) 78.28%
Carbon content (weight) 21.72%
1.
Crystal Structure rombohedral l Density. 2. 51: gm. /cc-0. 09 07;-lb/cu. l in, .
d
[
s-Melting Point- 24500C.. - 4442 0F1 Boiling Point. 0 0 3500 C-6332 Fl
! Microscopic. Capture 600 barn.
-cross-section 1
1 2-18 i
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l Table 2.5.1 HEAVY LOAD HANDLING COMPLIANCE MATRIX (NUREG-0612)
Criterion comoliance
- 1. Are safe load paths defined for Yes the movement of heavy loads to minimize the potential of impact, if dropped on irradiated fuel?
- 2. Will procedures be developed to Yes 4 cover: identification of required i equipment, inspection and acceptance criteria required before movement l of load, steps and proper sequence for handling the load, defining the safe load paths, and special precautions?
- 3. Will crane operators be trained Yes and qualified?
- 4. Will special lifting devices meet Yes the guidelines of ANSI 14.6-1978?
~
- 5. Will non-custom lifting devices Yes be installed and used in accordance with ANSI B30.9-1971?
- 6. Will the cranes be inspected and Yes
, tested prior to use in rerack?
- 7. Does the crane meet the intent of Yes ANSI B30.2-1976 and CMMA-70?
- 8. Is the crane and its main hook single failure proof? Yes i
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i 3.0 RACK FABRICATION AND APPLICABLE CODES The object of this section is to provide a self-contained description of rack module construction for the FCS fuel pool to j enable an independent appraisal of the adequacy of design. A list of applicable codes and standards is also presented.
3.1 Fabrication Obiective The requirements in manufacturing the high density storage racks for the FCS pool may be stated in four interrelated points:
, (1) The rack module is fabricated in such a manner that there
, is no weld splatter on the storage cell surfaces which
, would come in contact with the fuel assembly.
4 (2) The storage locations are constructed so that redundant flow paths for the coolant are available.
(3) The fabrication process involves operational sequences which permit immediate verification by the inspection staff.
(4) The storage cells are connected to each other by austenitic stainless steel corner welds which leads to a honeycomb lattice construction. The extent of welding is selected to "detune" the racks from the postulated seismic input motion.
3.2 Rack Module for Recion I This section describes the constituent elements of the FCS Region I rack modules in the fabrication sequence.
l The rack module manufacturing begins with fabrication of the box.
The " boxes" arc fabricated from two precision formed channels.by
- seam welding in a machine equipped with copper chill bars and pneumatic clamps to minimize distortion due to welding heat input.
Figure 3.2.1 shows the box.
3-1 3
-ws'y
The minimum wold penetration will be 80% of the box metal gage l which is 0.075" (14 gage). The boxes are manuf actured to 8.46" I.D. (nominal insido diameter).
A die is used to flare out one end of the box to provide the 300 tapered lead-in (Figure 3.2.2). One inch diameter holes are punched on two sides near the other end of the box to provide the ,
requisite auxiliary flow holes.
Each box constitutes a storage location Each side of a box facing another box is equipped with a narrow rectangular cavity which houses one integral Boral sheet (poison material).
The design objective calls for installing Boral with minimal surface loading.-This is accomplished by die forming a " picture
- frame sheathing" as shown in Figure 3.2.3. This sheathing is made to precise dimensions such that the offset is .010 to .005 inches greater than the poison material thickness.
The poison material is placed in the customized flat depression region of the sPeathing, which is next laid on a side of the " box".
The precision of the shape of the sheathing obtained by die forming guarantees that the poison sheet installed in it will not be subject to surface compression. The flanges of the sheathing (on all four sides) are attached to.the box using skip welds. The-sheathing serves to locate and posit.. ion the poison sheet accurately, and to preclude its movement under sv.smic conditions.
P After fabricating-the required number of composite box assemblies, the boxes are joined together in a fixture usir.g connector elements in the manner shown in Figure 3.2.4. The pitch be a the box centerlines is p, in one principal _ direction and py An the other principal direction, p, and py are unequal to : agree with the j dimensions of the FCS pool. The fabrication _ procedure in either l
3-2 1
i direction is identical, since the channels are fillet welded to make the inter-box connection. Figure 3.2.5 shows an elevation view of two storage cells of a Region I rack module.
l Joining the cells by connector elements results in a well defined i
shear flow path, and essentially makes the box assemblage into a multi-flanged beam type structure, i
In the next step of manufacture, the " base plate" is attached to 3
the bottom edge of the boxes. The base plate is a 3/4" thick
] austenitic stainless steel plate stock which has 5" holes burned out in a pitch identical to the box pitch. The base plate is
! attached to the cell assemblage by fillet welding the box edge to the plate.
In the final step, adjustable leg supports (shown in Figure 3.2.6) are welded to the underside of the base plate. The adjustable legs provide a i 1/2" vertical height adjustment at each leg location.
1 The manufacturing of the Region I rack modules culminates with appropriate NDE of welds, which includes visual examination of cell
, longitudinal seam welds and cell-to-cell connection welds and liquid dye penetrant examination of support velds, in accordance
, with the design drawings.
3.3 Rack Module for Reaion II Region II storage cell locations have a single poison panel between
- adjacent austenitic stainless steel surfaces. The significant components (discussed below) of the Region II racks are
- (1) the
$ otorage box subassembly the base plate, (2) (3) the neutron absorber material, (4) picture frame sheathing, and (5) support legs.
2 2
i 3-3
j (1) Storace cell box subassembiv: As described for Region I, the " boxes" are fabricated from two precision formed channels by seam welding in a machine equipped with
! copper chill bars and pneumatic clamps to minimize distortion due to welding heat input. Figure 3.2.1 shows
, the " box".
j The minimum weld penetration will be 80% of the box metal gage which is 0.075" (le gage). The boxes are i
manufactured to 8.46-inch inside dimension (nominal).
As shown in Figure 3.3.1, each box has two lateral holes punched near its be.ctom edge to provide auxiliary flow holes. A " picture frame sheathing" is attached to each side of the box with the poison material installed in the i sheathing cavity. The edges of the sheathing and the box are wolded togeth r to form a smooth edge. The box, with integrally connected sheathing, is referred to as the
" composite box".
The " composite boxes" -are arranged in a checkerboard array to form an assemblage of storage cell locations (Figure 3.3.2). Flat plates are welded to the edges of
, the boxes at the outside boundary of the rack to make the periphery cells. The inter-box welding and pitch adjustment are accomplished by small longitudinal
, connectors.
- This assemblage of box assemblies is welded edge-to-edge as shown in Figure 3.3.2, resulting in a honeycomb structure with axial, flexural and torsional rigidity
. depending on the extent of intercell welding provided.
It can be seen from Figure 3.3.? that two edges of each interior box are connected to the contiguous boxes resulting in a well defined path for " shear flow".
(2) Base Plate: The base plate provides a continuous horizontal surface f':r supporting the-fuel assemblies.
The base plate is attached to the cell assemblage by fillet welds. The baseplate in each storage cell has a i
5" diameter flow hole.
I (3)' The neutron absorber material: As mentioned in the preceding section, Boral is used as the neutron absorber-material.
(4) Picture Frame Sheathi_ng: -As described earlier, the sheathing- serves as the locator' and retainer of the i poison material. Figure 3.2.3 shows a schematic of the i sheathing.
i 3-4
3 1
1 (5) Succort Least As stated earlier, all support legs are 4 the adjustable type (Figure 3.2.6). The top position is j made of austenitic stainless steel material. The bottom
- part is made of 17
- 4 Ph series stainless steel to avoid galling problems.
l Each support leg is equipped with a readily accessible socket to enable remote leveling of the rack after its
- placement in the pool. Lateral holes in the support leg
. provide the requisite coolant flow path.
i
! 3.4 Codes. Standards, and Practices for the Fort Calhoun
{ Station Soent Fuel Pool Racks
}
! The fabrication of the rack modules for FCS is performed under a i
i strict quality assurance system suitable for manufacturing and
- complying with the provisions of 10CFR50 Appendix B.
l
- The following codes, standards and practices are used as applicable j for the design, construction, and assembly of the spent fuel j storage racks. Additional specific references related to detailed j analyses are given in each section.
- s
! a. Desian Codes
! (1) AISC Manual of Steel Construction, 8th Edition,1980 1
(provides detailed structural criteria for linear j type supports).
(2) ANSI N210-1976, " Design Objectives for Light Water i Reactor Spent Fuel Storage Facilities at Nuclear
! Power Stations" (contains guidelines for fuel rack i design).
f (3). American Society of Mechanical Engineers (ASME),
, Boiler and Pressure Vessel Code,Section III, 1986 Edition, including -up to 1988 Addenda (governing i material procurement, fabrication and NDE).
} (4) ASNT-TC-1A June, 1980 American Society for
- Nondestructive Testing (Recommended Practice - for Personnel Qualifications).
(5) Building Code Requirements for Reinforced Concrete, ACI318-89.
(
i 3-5
I 1
I i i l (6) Code Requirements for Nuclear Safety Related j Concrete Structures, ACI349-85/ACI349R-85, and i
! ACI349.1R-80. ,
(7) ASME NQA-2-1989, Quality Assurance Requirements for l Nuclear Facility Applications.
- b. Material Codes - Standards of ASTM l (1) E165 - Standard Methods for Liquid Penetrant
- Inspection j (2) A240 -
Standard Specification for Heat-Resisting i Chromium and Chromium-Nickel Stainless Steel Plate, j Sheet and Strip for Fusion-Welded Unfired Pressure l Vessels-i I A262 - Detecting Susceptibility to Intergranular (3) i Attack in Austenitic Stainicas Steel i !
(4) A276 -
Standard _ Specification - for Stainless and '
- Heat-Resisting Steel Bars and Shapes j (5) A479 - Steel Bars for Boilers & Pressuro Vessels l (6) C750 -
Standard Specification for Nuclear-Grade i Boron Carbide Powder i
i (7) C992 -
Standard Specification for Boron-Based l Neutron Absorbing Material Systems for Use in j Nuclear Spent Fuel Storage Racks (8) American Society of Mechanical Engineers ( ASME) ,
- Boiler and-Pressure Vessel Code, Section II-Parts
. A and C,. 1986 Edition, up to and including 1988 i Addenda.
- c. Weldino Codes ASME Boiler and Pressure Vessel Code,Section IX -Welding and Brazing Qualifications,- 1986 Edition up to and s
including 1988-Addenda.
- d. Quality Assurance, Cleanliness, Packaging, Shipping, Receiving, Storage, and Handling Requirements
! (1) ANSI 45.2.1 -
Cleaning of Fluid ~ Systems. and Associated Components'during-Construction Phase of.
Nuclear-Power Plants.-
1 4
d 3-6 i
. . . -_ ,- .,-,.-.,,., - . , - , , ,. -...-.- .-- . _ ,- , ..-. - , - ~ - . . ~
I l I
i 1
i (2) ANSI N45.2.2 -
Packaging, Shipping, Receiving, Storage and Handling of Items for Nuclear Power
) Plants.
i
- (3) ANSI - N45.2.6 -
Qualifications of Inspection,
, Examination, and Testing Personnel for Nuclear Power i Plants (Regulatory Guide 1.58).
1 l (4) ANSI-N45.2.8, Supplementary Quality Assurance
- Requirements for Installation, Inspection and 2 Testing of Mechanical Equipment and Systems for the
- Construction Phase of Nuclear Plants, 1975.
i (5) ANSI - N4 5. 2.11,1974 Quality Assurance Requirements for the Design of Nuclear Power Plants.
l (6) ANSI-N45.2.12, Requirements for Auditiiig of _ Quality
- Assurance Programs for Nuclear Power Plants, 1977.
4 (7) ANSI N45.2.13 - Quality Assurance Requirements for
( Control of Procurement of Equipment Materials and Services for Nuclear Power Plants (Regulatory Guide 1.123).
I (8) ANSI N45.2.23 - Qualification of Quality Assurance
- Program Audit Personnel for Nuclear Power Plants q (Regulatory Guide 1.146).
(9) ASME Boiler and Pressure Vessel, Section V,
, Nondestructive Examination,1983 Edition, including Summer and Winter 1983.
(10) ANSI -
N16.1-75 Nuclear Criticality Safety Operations with Fissionable Materials outside i Reactors.
- (11) ANSI - N16.9-75 Validation of Calculation Methods i for Nuclear Criticality Safety.
- e. Governina NRC Desian Documents j (1) NUREG-08_0 0, Standard Review Plan (1981).
l (2) "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," dated April 14, 1978, and the - modifications to this document of-January.18, 1979.
(3) NUREG 0612, " Control of Heavy Loads on Nuclear Power Plants",_USNRC, Washington, D.C.
i 3-7
1
]
- f. Qther ANSI Standards (not listed in the crecedinal (1) N16.1 -
Nuclear Criticality Safety in operations with Fissionable Materials outside Reactors (2) N45.2 - Quality Assurance Program Requirements for Nuclear Facilities - 1971
! (3) N45.2.9 - Requirements for Collection, Storage and Maintenance of Quality Assurance Records for Nuclear l Power Plants - 1974 (4) N45.2.10 - Quality Assurance Terms and Definitions -
j 1973
- (5) N210 -
Design objective for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power i Plants (6) N14.6-78, American National Standard for Special
, Lifting Devices for Shipping Containers Weighing l 10,000 pounds (4500 kg) or mora for Nuclear Materials a g. Code-of-Federal Reculations 4
(1) 10CFR21 - Reporting of Defects and Non-compliance 1
(2) 10CFR50 - Appendix A - General Design Criteria for Nuclear Power Plants (3) 10CFR50 - Appendix B - Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing i Plants
Assumptions Used for Evaluating the
! Potential Radiological Consequences- of a Fuel Handling Accident in the Fuel Handling and Storage Facility of Boiling and Pressurized Water Reactors (3) RG 1.28 - (ANSI N45.2) - Quality Assurance Program Requirements, June, 1972 (4) RG 1.29 - Seismic Design Classification 3-8 4
(5) RG 1.38 -
(ANSI N45.2.2) Quality Assurance Requirements for Packaging, Shipping, Receiving, Storage and Handling of Items for Water-cooled Nuclear Power Plants, March, 1973. .
(6) RG 1.44 - Control of the Use of Sensitized Stainless Steel (7) RG 1.58 -- ( ANSI N4 5. 2. 6) Qualification of Nuclear Power Plant Inspection, Examination, and Testing Personnel.- Rev. 1, September, 1980 (8) RG 1.64 -
(ANSI N45.2.11) Quality Assurance Requirements for the Design of Nuclear Power Plants, October, 1973.
(9) RG 1.74 -
(ANSI N45.2.10) Qua'.ity Assurance Terms and Defin$ tions, February, 19' 4.
(10) RG 1.88 -
(ANSI- N4 5. 2.9) Collection, Storage-and Maintenance of Nuclear Power Plant Quality Assurance Records. Rev. 2, October,_1976.
l (11) RG 1.92 -
Combining Modal Responses and Spatial '
Components in Seismic Response Analysis (12) RG 1.123 -
(ANSI N45.2.13) Quality Assurance Requirements for Control of Procurement of Items and Services for Nuclear. Power Plants.
(13) RG 8.8 -
"Information Relevant to Ensuring that Occupational Radiation Exposures at Nuclear Stations Will be as Low-as is Reasonably Achievable", Rev.
3, Jun. 1978, Rev. 2, Mar. 1977, Rev. 1,-Sep. 1975, Base Jul. 1973,
- i. Branch Technical Position' (1) CPB 9.1 Criticality in Fuel _ Storage Facilities (2) ASB 9-2 -
Residual Decay Energy _ for_ . Light-Water Reactors for Long-Term Cooling.
-j. Standard Review Plan (1) SRP 3.7.1 - Seismic Design Parameters (2) SRP 3.7.'2 - Seismic: System Analysis-(3) SRP 3.7.2 -' Seismic-Subsystem Analysis 3-9
k i
I (4) SRP 3.8.4 - Other Seismic Category I i Structures (including Appendix D) !
i' -
(5) SRP 9.1.2 - Spent Fuel Storage I
J (6) SEP 9.1.3 - Spent Fuel Pool Cooling and Cleanup System i
I
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) l FIGURE 3.2.5 ELEVATION VIEW 0F A REC 10N I RACK SH0 TING TWO STORACE CELLS 3-15
Re Baseplata I l ks N s s s s s s si i % s s xs ssss sxs s s, 1 1 i r F I
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FIGURE 3.2.6 ADJUSTABLE SUPPORT 3-16
4,_ _ _ _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ - .
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y ELEVATION'VIEif 0F REGION R RACK TiODULE.
4 I 3-17
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FIGURE 3.3.2 TYPICAL ARRAY OF REC 10N 11 CELLS:
- INON-FLUX TRAP CONSTRUCTION) 3-18
. - - a,-_ , .-.a,.-, -, . .-, -
4.0 CRITICALITY SAFETY ANALYSES 4.1 DESIGN BASES The high density spent fuel storage racks for the Ft. Calhoun Station are dealgned to assure that the effective neutron multiplication factor (k,y) is equal to or less than 0.95 with the racks fully loaded with fuel of the highest anticipated reactivity, and flooded with unborated water at the temperature corresponding to the highest reactivity. The maximum calculated reactivity includes a margin for uncertainty in reactivity calculations including mechanical tolerances. All uncertainties are statistically combined, such that the final k,, will be equal to or less than 0.95 with a 95%-probability at a'95% confidence level.
Applicable codes, standards, and- regulations or pertinent sections thereof, include the following:
o General Design Criteria 62, Prevention of Criticality in Fuel Storage and Handling.
o USNRC Standard Review Plan, NUREG-0800, Section 9.1. 2, Spent Fuel Storage, Rev. 3 - July 1981.
o USNRC letter of April 14, 1978, to all Power Reactor Licensees --0T Position for Review and Acceptance of Spent Fuel Storage and Handling Applications, including modification letter dated January 18, 1979.
o USNRC Regulatory Guide 1.13,- Spent -Fuel Storage-l Facility Design Basis, Rev. 2 (proposed), December 1981.
t o_ ANSI ANS-8.17-1984, Criticality. Safety Criteria for the j Handling, Storage. and Transportation of LWR Fuel Outside Reactors.
l 4 '. - l' i .
l
)
_ _ . . . .- _ .= __ _ _ _ _ _ _ .._ _ . _ . . . . _- __
4
$ o ANSI N210-1976, Design Objectives for Light Water i Reactor Spent Fuel Storage Facilities at Nuclaar Power i Plants.
t
- USNRC guidelines and the applicable ANSI standards specify that the maximum effective multiplication f actor, "k,,,", including j uncertainties, shall be less than or equal to 0.95. The infinite multiplication f actor, "k,", is calculated f or an infinite array, neglecting neutron losses due to leakage from the actual storage j rack, and thorefore results in a higher and more conservative value. In the present evaluation of criticality safety in the Pt. Calhoun storage racks, the limiting design basis criterion was assumed to be a "k," o f 0.95, which is more conservative i than the limit specified in the regulatory guidelines.
] To assure the true reactivity will always be less than the calculated reactivity, the following conservative assumptions
! Were made l
o Moderator is unborated water at a temperature that results in the highest reactivity (4'C), corresponding
- to the maximum possible moderator density).
l o In all cases (except for the assessment of peripheral
?
effects and certain abnormal / accident conditions where neutron leakt.go is inherent), the infinite multiplica-tion factor, K, was used rather than the effective multiplication factor, k neutron loss from radialandaxialleakaga,l,an(i.e., eglected).
o Neutron absorption in minor structural members is neglected, i.e. , spacer grids are analytically replaced by water.
The design basis fuel assembly is a 14 x 14 fuel assembly containing Uoa at a maximum initial-enrichment of 4.2 wtt U-235 by weight, corresponding to approximately 48.2 grams U-235 per. axial centimeter of fuel assembly.
4-2
Two separate storage regions are provided in the spent fuel storage pool, with independent criteria defining the highest potential reactivity in each of the two regions as follows:
o Region 1 is designed to accommodate new fuel with a maximum enrichment of 4.2 wt% U-235, or spent fuel regardless of the discharge fuel burnup.
o Region 2 is designed to accommodate fuel of various initial enrichments which have accumulated a minimum burnup within an acceptable barnup domain. The peripheral cells of Region 2, because they -are in a l l high neutron leakage area, effectively constitute a i Region 3 zone which can safely accommodate fuel of i lower burnups, i
i i
The water in the spent fuel storage pool normally _ contains soluble boron which would result in large suberiticality_ margins
- t l
under actual operating conditions. However, the NFC guidelines, j based upon the accident condition in which all soluble poison-is-j assumed to have been lost, specify that the limiting k,,, of 0.95
} be evaluated in the absence of soluble boron. The double
- j. contingency principle of ANSI N-16.1-1975 and of the April 1978 j NRC letter allows credit for soluble boron under other abnormal j or accident conditions since only a single accident- need be l considered at one time. Consequences of abnormal and accident j conditions have also been evaluated, where " abnormal" refers to l
conditions (such as higher water temperatures resulting from full-core discharge) which may reasonably be expected to occur .
I during the lifetime of the plant and " accident" refers to
! conditions which are not expected to occur but nevertheless must I be protected against.
1 4-3 f
4 2
y , , p . y- y .,37,y.,, +-,,,v. . - , - ,-m-- , .y. . ..--%3v.,,-,y,y-, ,,,,u m . - .___.,.,v y-ye-,-.%-- ,--,r-,-.- , r p- , --
4.2
SUMMARY
OF CRITICALITY ANALYSES 4.2.1 Normal Operatina Conditions The criticality analyses of each of the two separate regions of the spent fuel storage pool are summarized in Table 4.2.1 for the design basis storage conditions which assumes the design basis single accident condition of the loss of all soluble boron. For Region 1 with 4.2% enriched fuel, the maximum k,, is 0.928, including uncertainties. With 4.2% fuel burned to 32,000 Mwd /MtU, the corresponding maximum reactivity in Pegion 2 is 0.935 including uncertainties. The calculated maximum reactivity in Region 2 includes a burnup-dependent allowance for uncertainty in depletion calculations and, furthermore, provides a substan-tial margin below the limiting infinite multiplication factor (k2 ) of 0.95. As cooling time increases in long-term storage, decay of pu-241 results in a significant decrease in reactivity, which will provide an increasing subcriticality margin and tends to further compensate for any uncertainty in depletion calcula-tions.
Region 2 can safely accommodate fuel of various initial enrich-ments and discharge fuel burnupe, provided the combination falls within the acceptable domain illustrated by the solid line in Figure 4.2.1. For convenience, the minimum (limiting) burnup data in Figure 4.2.1 for unrestricted storage in Region 2 can be described as a function of the initial enrichment, E, in weight percent U-235 by a fitted polynomial expression as follows:
Minimum Burnup in MWD /MtU =
-30240 + 21810 E - 2360 E2 + 166 E 3 (for initial. enrichments up to 4.5 wt% U-235) 4-4 i
{
i i
Because they are located in an area of high neutron leakage, the
- peripheral cells of Region 2 can safely accommodate fuel of i higher reactivity (lower burnup) . Calculations identified the
, limiting burnup for those peripheral cells (called Region 3), and ;
I the limit is shown by the lower curve in Figure 4.2.1. The Region 3 limiting burnups have been fitted to a polynomial j expression as follows: ,
i i .
, i j Minimum Burnup in MWD /MtU =
a i -33270 + 23411 E - 3308 Et + 274 E 3
- (for-initial enrichments up to 4.5 wtt U-235) 1 i
]! The burnup criteria for acceptable storage in Region 2 will be l implemented ir. appropriate administrative procedures to assure
- verified burnup as specified in the proposed Regulatory Guide
- i. 1.13, Revision 2. Administrative procedures will also be j employed to confirm and' assure the presence of soluble poison in j the pool water during fuel handling operations, as a further 1 cement during ual handling oper tions.
1 In addition, a number of used CEA rod assemblies are available and may be inserted into fuel assemblies that do not otherwise meet the burnuip requirements for unrestricted storage in Region 2. -(The USNRC has previously approved such use of CEA rod ,
j assemblies). With .a full-length CEA assembly installed and latched in a fuel assembly of 4.2% enrichment (no burnup), the
-maximum k , in Region 2 is 0.923, including uncertainties. Thus,
- with 'a-- CEA installed, any fuel of 4.2%' enrichment ~ or less and
, with any burnup may ba_ safely stored in'any Region 2 cell.
i-
- 4 -5 s
k
,.'+.
. - . + , - - r. ..m., . , . , , , , ,m -
. . . , . . , . , - - , . . . - . . . . , _ , . . .k,m-.., , , _ . . ,,,-..,-..r ..% , -
I l
4.2.2 Abnormal and Accident Conditions l
Although credit for the soluble poison normally present in the spent fuel pool water is permitted under abnormal or accident l conditions, most abnormal or accident conditions will not result in exceeding the limiting reactivity (k,,, of 0.95) even in the absence of soluble poison. The effects on reactivity of credible abnormal and accident conditions are presented in detail in Section 4.8 and briefly summarized in Table 4.2.2.
of the abnormal / accident conditions identified, only two have the potential for a more than negligible positive reactivity effect.
9 The inadvertent misplacement of a fresh fuel assembly without a CEA installed either (into a Region 2 storage cell or outside and adjacent to a rack module) has the potential for exceeding the
- limiting reactivity, should there be a concurrent and independent j accident condition resulting in the loss of all soluble poison.
Administrative procedures to assure the presence of soluble poison during fuel handling operations will preclude the possibility of the simultaneous occurrence of the two independent
, accident conditions.
The largest reactivity increase would occur if a new fuel assembly of the highest permissible reactivity were--to be
, positioned in an otherwise fully loaded Region 2 storage rack
- module. Under this accident condition, credit for the presence
, of soluble poison is permitted by NRC guidelines *, KENO-Sa i
calculations indicate that a minimum boron concentration of 75 ppm boron would be adequate to assure that the limiting k,,, of 4
0.95 is not exceeded.
- Double contingency principle of ANSI N16.1-1975, as specified in the April 14, 1978 NRC letter (Section 1.2) and implied in the proposed revision to Reg. Guide 1.13 (Section 1.4, Appendix A).
4-6
,,, _ . - - , - - - - , ~ . - . ,
l I
I 4.3 REFERENCE FUEL STORAGE CELLS l l
4.3.1 Reference Fuel Assembly The design basis fuel assembly, illustrated in Figures 4.3.1 and
_ 4.3.2, is a Westinghouse 14 x 14 array of fuel rods with 20 rods replaced by 5 control rod guide tubes. Table 4.3.1 summarizes 2
the fuel assembly design specifications (4.3.1, 4.3.2) and the expected range of the significant manufacturing tolerances.
Initially, calculations were also made for fuel assemblies of Combustion Engineering and ANF designs. CE and Westinghouse fuel were comparable in reactivity, as indicated below, for fuel of 4.2% enrichment, although ANF fuel was lower in reactivity.
Westinghouse fuel k, (CASMO-3 ) = 0. 9173 i
Combustion Engineering fuel k, (CASMO-3 ) = 0.914 8 ANF fue1 k, (CASMO-3 ) = 0. 9114 Consequently, the Westinghouse fuel design was used for the remainder of the calculations.
4.3.2 Reaion 1 Fuel Storace Cells The nominal spent fuel storage cell used for the criticality l analyses of Region 1 storage cells is shown in Figure 4.3.1. The j rack is composed of Boral absorber material outboard of an 8.46- .
inch I.D. , 0.075-inch thick inner stainless steel box. The fuel assemblies are centrally located in each storage cell on a
- nominal lattice spacing of 10.363 1 0.080 inches in one direction and 9.821 0.080 inches in the other direction. Stainless steel channels connect one storage cell box to another in a' rigid structure and define an outer water space-between boxes. This outer water space constitutes a flux-trap between the two Boral 1
- 4-7 i
4 1
l i
absorber panels that are each effectively opaque to thermal neutrons. The Boral absorber has a thickness of 0.075 1 0.004 I
inch and a nominal B-10 areal density of 0.0151 g/cm2 (minimum of ,
j 0.014 g/cm )i .
I I
i 4.3.3 Reaion 2 Puel Storace Cells
! The design basis for Region 2 storage cells is fuel of 4.2 wt% U-235 initial enrichment burned to 32,000 Mwd /MtU. In Region 2, j the storage cells are composed of a single Boral absorber panel j between the stainless steel walls of adjacent storage cells.
) These cells, shown in Figure 4. 3. 2, are located on a lattice h spacing of 8.652 0.040 inches. The Boral absorber has a l thickness of 0.075 1 0.004 inch and a nominal B-10 areal density of 0.0151 g/cm 2 (minimum of 0.014 g/cm 2),
i N
b i
e f
f i
t 4
i 4-8 l
I
- . , , -w . - . , , -. ~~* - - . , , . - ~ ~
4.4 ANALYTICAL METHODOLOGY 4.4.1 Reference Desian CalculatioDa In the fuel rack analyses, the primary criticality analyses of the high density spent fuel storage racks were performed with a two-dimensional multi-group transport theory technique, using the CASMO-3 ( 4. 4.1) computer code. Independent verification calcula-tions were made with a Monte Carlo technique utilizing the AMPX-KENO-5a computer package (4.4.2), with the 27-group SCALE
- cross-section library (4.4.3) and the NITAWL subroutine for U-238 resonance shielding effects (Nordheim integral treatment).
Benchmark calculations, presented in Appendix A, indicate a bias of 0.0000 with an uncertainty of i 0.0024 for CASMO and 0.0101 1 0.0018 (95%/95%) for NITAWL-KENO-Sa. CASMO was also used both for burnup calculations and for evaluating the small reactivity increments associated with manufacturing tolerances.
In tracking long-term (30-year) reactivity effects of spent fuel stored in Region 2 of the fuel storage rack, previous calcula- !
tions (4.4.4,4.4.5) have confirmed a continuous reduction in reactivity with time (after Xe decay) due primarily to pu-241 decay and Am-241 growth.
In the geometric model used in the calculations, each fuel rod and its cladding were described explicitly and reflecting boundary conditions (zero neutron current) were used in the axial direction. In Region 1, reflecting boundary conditions were also used at the centerline through the water gap and, in Region 2, at the centerline of the Boral and steel plates between storage
- " SCALE" is an acronym for Etandardized _Q.omputer Analysis for Licensing Evaluation, a standard cross-section set developed by ORNL for the USNRC.
4-9
_ _ _ . = _ _ _ . _ . _ . . . _ _ _ _ _ _ . _ _ _ _ . _ _ _ . . _ _ . _ _ . _ . . _. _ . . . _ _
4 l calls. These boundary conditions have the effect of creating an l infinite array of storage cells in the radial directions. A 30 cm water reflector was used in the axial direction of the KENO-Sa
- calculational model. In calculating the boundary cells (i.e.,
I i
Region 3 storage cells), a five assembly array with a 1.5 inch water-24 inch concrete reflector in one radial direction was .
l assumed.
) NITAWL-KZNo-Sa Monte Carlo calculations inherently include a statistical uncertainty due to the random nature of neutron tracking. To minimize the statistical uncertainty of the KENO-f l calculated reactivity, 1,250,000 neutron histories in 2500 j generations of 500 neutrons each, are accumulated in each
, calculation.
i 4.4.2 Fuel Burnuo Calculations and Uncertainties j CASMO-3 was used for burnup calculhtions in the hot operating .
f' condition. CASMO has been extensively benchmarked (Appendix A j and references 4.4.1 and 4.4.6) against' cold, clean, critical experiments (including plutonium-bearing fuel), Monte-Carlo calculations, reactor operations, and heavy-element concentra-l- tions in irradiated fuel. ;
l i
l Since. critical experiment data with spent fuel is not ava'ilable for determining the uncertainty in depletion calculations, an j allowance for uncertainty in reactivity was assigned based upon i
other considerations. -Over a_ considerable portion of the burnup -
I history in PWRs, the reactivity loss rate'is approximately'O'.01
} Sk for each 1000 Mwd /MtU burnup, becoming smaller at the-higher burnups. It was assumed that the uncertainty in depletion- -
1-4 l P
i i
--m. - - - - .- ..m - . _ . ..--._ ,.. , - - . . . . , , . _ . - . . - - . - , ~ . . . - - -- - - - -
_ . . _ _ _ . . . . . . .__. _ . - _ . _ _ _ _ . . _ _ _ _ _ _ . . . _ . . _ _ . . _ _ _ _ _. ..._m._ __ -. .
l 1
i calculations
- is equal to or less than 5% of the total reactivity
! decrement from the beginning of life to the burnup of interest.
l For the Fort Calhoun storage racks at the design basis burnup of
[ 32,000 Mad /MtU, (4.2% initial enrichment), the reactivity allowance for uncertainty is 0.0133 8k. For additional
$ conserve.tism, the uncertainty in depletion was treated as an additive term rather than being combined statistically with other j uncertainties. This allowance for uncertainty in depletion j calculations is believed to be a conservative estimate, particu-larly in view of the substantial reactivity decrease as the fuel i ages.
(
4.4.3 Effect of Axial Burnuo Distribution Initially, fuel loaded into the reactor will burn with a slightly
{ skewed cosine power distribution. As burnup progresses, the l burnup distribution will tend to flatten, becoming more highly l burned in the central regions than in the upper and lower ends.
i At high burnup, the more reactive fuel near the ends of the fuel l assembly (less than average burnup) occurs in regions of lower *
[ reactivity worth due to neutron leakage, consequently, it would be expected that over most of the burnup history, distributed 1
burnup fuel assemblies would exhibit a slightly lower reactivity than that calculated for the average burnup. As burnup progress-l es, the distribution, to some extent, tends to be self-regulating ;
- as controlled by the axial power distribution', precluding the existence of large regions of significantly reduced burnup.
Among others, Turner (4.4.4) has provided_ generic analyses of the axial burnup effect based upon calculated -and measured . axial burnup distributions. These analyses confirm the minor and-1 L
only that portion of the uncertainty due_to burnup. Other
- uncertainties are accounted'for elsewhere.
4 - 11 e
y ,. w-- , , ,,, -- v. * + ,r y+ , , - -c 4, 4=y- g-, y,--
y y,,wpi--, s- -,w -g,e-y ++%g ,,,-r , a- y .,,aav. r -ei,v-we+v- e = - - -wywy v er-*g g -
generally negative reactivity effect of the axially distributed burnups at values less than about 30,000 Mwd /M' J 'ibe trends observed, however, suggest the possibility of a saal; positive reactivity effect at higher burnup values. KENO-Sa calculations were made, based on a representative PWR axial'burnup distribu-tion in 10 zones and burnup-equivalent enrichments in each of the axial zones. Results of these calculations indicate the effect of the axial distribution in burnup could be as much as 0.0038 Sk at the design basis burnup of 32,000 Mwd /MtU.
i
] 4.4.4 Lona-term Chances-in Reactivity Since the fuel racks in Region 2 are intended to contain spent fuel for long periods of time, consideration was given to the r long-term changes in reactivity of spent fuel. Published data .\
(4.4.4,4.4.5) confirm that reactivity continuously decreases as the spent fuel in storage ages. Early in the decay period, Xenon grows from Iodine docay (reducing reactivity) and~ subsequently decays, with the reactivity reaching a maximum at 100-200 hours.
The decay of Pu-241 (13-year half-life) and growth of Am-241 substantially reduce the reactivity during long term storage.
The design of the Ft. Calhoun racks do not take credit for this long-term reduction in reactivity, other than to indicate an increasing subcriticality. margin in Region - 2 of the-spent fuel storage pool.
4 - 12
4 4.5 Region 1 CRITICALITY ANALYSES AND TOLERANCES 4.5.1 Uominal Design y For the nominal storage cell design in Region 1, the CASMO cslculation resulted in a k, of 0.9173 1 0.0024, which, when combined with all known uncertainties, results in a maximum k, of 0.928. Independent calculations with AMPX-KENO gave a bias -
corrected k, of 0.9195 i 0.0022, including a one-sided tolerance factor (4.5.1) for 95% probability at a 95% confidence level.
, combining all known uncertainty factors, the calculated k, becomes 0.9195 1 0.0103 or a maximum k, value of 0.930, confirming the reference CASMO calculation, within statistical uncertainty.
l 4.5.2 Uncertairties Due to Tolerances J
The uncertainties due to manufacturing tolerances are listed in i.
Table 4.2.1 and discussed below.
4.5.2.1 Boron Loadina Tolerances Boral a'osorber panels used in the . storage cells are nominally 2
'5 inG thick, 7.25-inches wide and 128 inches long, with a r 9.nal B-10 areal density of 0.0151 g/cm2 . The vendors manufac-3 Lu' ling tolerance limit is t 0.0011 g/cm in B-10 content which assures that at any point, the minimum B-10. areal density will not be less than 0.014 g/cm 2. Differential CASMO calculations indfcata that these tolerance limits result in incremental reactivity uncertainties of i 0.0030 8k.
[ 4 - 13 i
l l
1
4.5.2.2 Boral Width Toleranca The reference storage cell design uses a Boral panel with an initial width of 7.25 t 1/16 inches. For the maximum tolerance, the calculated reactivity uncertainty is 1 0.0007.
4.5.2.3 Tolerances in Cell Lattice Spacina The design storage cell lattice spacing between fuel assemblies is 9.821 1 0.080 in one direction and 10.363 1 0.080 in the other direction. Decreasing the water spacing in the flux-trap results in a small increase in reactivity. The average flux-trap water thickness is 1.271 0.080 inches, which results in an uncertain-ty of 0.0090 6k due to the tolerance in flux-trap water
- thickness, assuming the water thickness is simultaneously reduced on all four sides. Since the manufacturing tolerances on each of the four sides are statistically -independent, the actual reactivity uncertainties would be less than 0.0090 6k, although the more conservative value has been used in the criticality evaluation.
4.5.2.4 Stainless Steel Thickness Tolerances The nominal stainless steel thickness is 0.075 0.005 inch for the stainless steel box. The maximum positive reactivity effect of the expected stainless steel thickness tolerance variations, was calculated (differential CASMO runs) to be 0.0004 6k.
4 - 14 r
l 1
i l 4.5.2.5 Fuel Enrichment and Density Tolerances I-l The design' maximum enrichment is 4.20 1 0.05 wt% U-235.
! Calculations of the sensitivity to small enrichment variations'by
] CASMO yielded a coefficient of 0.0043 8k per 0.1 wt% U-235 at the l design enrichment. For the. assumed _ tolerance on U-235 enrichment j of i 0.05 wt%, the uncertainty on k, is' i 0.0022 8k.
l Calculations were also made with the UO 2fuel density increased-l to the maximum ~ expected - value of 10.49 g/cm 3 (stack' density).
For the reference design calculations, the uncertainty in.--
l reactivity is 1 0.0022 8k over the maximum expected range of UO 2-
! density.
1 i
3 l
N
.~
l h
I I'
i I
- t l- 4 - 152- '
i 7.
l l
l
, , , , _ _ ~ - . . , - - . . .. ,- ..,,.,.........,..,,,,,.-,m.,_,,-m,, ,. , ,.-
i 1
- l 1
j- 4.6 REGION 2 CRITICALITY ANALYSES i
i 4.6.1 Nominal Desian Case i
i The principal method of analysis in Region 2 was the CASMO-3 code, using the restart option to analytically transfer fuel of
- a specified burnup into the storage rack configuration at a j reference temperature of 4 'C. Calculations were made for fuel of several different initial enrichments and, at each enrichment, a limiting k, value was established which included an additional factor for uncertainty in the burnup analyses and for the; axial l
burnup distribution. The restart CASMO calculations (cold, no-i Xenon, rack geometry) were then interpolated to define the burnup
- value yielding the limiting ~ k,, value for each enrichment, as indicated in Table 4.4.1. These converged burnup values define
- the boundary of the acceptable domains shown in Figure 4.2.1 for Region 2 and the peripheral cells (Region 3). '
l 4
L. Minimum =Burnup in MWD /MtU ='
f D-30240 +;21'10 8 E ^2360'E2 '+ L 16 61. E3 '
! (for:initialJenrichments:.upfto 4.U wt%;U-235) z
-At a burnup of 32,000 Mwd /MtU, the sensitivity tc burnup .was
( calculated to be 0.00785 6k per 1000 Mwd /MtU. During long-term storage, the k,, values of the Region 2 - fuel rack will ' decrease continuously as indicated in Section 4.4.4.
1 An independent AMPX-KENO calculation was used to provide l additional confidence -in the reference Region - 2 criticality l 4 - 16 i
i l
i l
I
-- - , w w ,+,e- - - + , , ,
I i
analyses. Fuel of 1.651 wt% initial enrichment (equivalent to I the reference rack design for burned fuel without burnup uncertainty) was analyzed by NITAWL-KENO-Sa and by . the CASMO model used for the Region 2 rack analysis. For this case, the CASMO k,, 0.9126 at 1.651 wt% enrichment was essentially the same as the bias-corrected KENO value 0.9123 1 0.0005 obtained in the NITAWL-KENO-Sa cell calculations, confirming the CASMO calcula-tion.
A special case exists in the use of full-length CEA at,semblies installed and latched into a fuel assembly. Calculations for this case assumed that the CEA was depleted to 76% of its initial Boron-10 loading (85% depletion originally estimated, reduced to 4
75% for conservatism (4.6.1)). The CEA assembly design parame-ters are (4.6.1, 4.6.2) shown in Table 4.6.1. Calculations, i
summarized in Table 4.6.2, indicate a k,,, of 0.9174 (CASMO3),
which with uncertainties added, results in a maximum k, , , of I 0.923, which is well within the USNRC limit and therefore acceptable. The corresponding KENO-Sa value for the maximum k,g is 0.922 which confirms the CASMO3 value. ,
4.6.2 Boundary Cells (Rection 3 )
The boundary cells facing the pool walls are an area of high neutron leakage and may be used to store fuel assemblies that have not achieved a burnup adequate for unrestricted storage in
~
Region 2. Calculations with KENO-Sa for the boundary cells, using a 1.5 inch water-24 inch concrete reflector, show that these cells can safely accommodate fuel of 4.2 % initial-enrichment burned to 27000 Mwd /MtU. With the boundary cells containing fuel of 27000 Mwd /MtU burnup and the remainder of the racks filled with fuel of 32000 Mwd /MtU burnup, . the KENO-Sa calculated reactivity was 0.9070 i 0.0005 (1 o) . This reactivity 4 - 17
is less than the corresponding k of 0.9099 1 0.005 (1 o) for 2
Region 2 filled with fuel of 4.2% enrichment burned to 32000 Mwd /MtU. Both of these calculations include axial leakage.
, Thus, the boundary cell can safely accommodate fuel of the enrichment-burnup combination indicated by the lower curve in Figure 4.2.1.
4.6.3 Uncertainties Due to Tolerances The uncertainties due to manufacturing tolerances are listed in Table 4.2.1 and discussed below.
I
, 4.6.3.1 Boron Loadina Tolerances i
4 The Boral absorber panels used in the Region'2 storage cells are 4
0.075 inch thick with a nominal B-10 areal density of 0.0151 i 2 g/cm. The manufacturing limit of 0.0011 g/cm3 in B-10 loading assures that at any point the minimum B-10 areal density will not be less than 0.014 g/cm. 2 Differential CASMO calculations indicate that this tolerance limit results in an incremental reactivity uncertainty of 0.0032 6k.
4.6.3.2 Boral Width Tolerance The reference storage cell design for Region 2 (Figure 4.3.2) uses a Boral absorber width of 7.25 i 1/16 inches. This tolerance results in a reactivity uncertainty of 0.0005 6k.
4 4.6.3.3 Tolerance in Cell Lattice Scacina The manufacturing tolerance on inner box dimension affects the storage cell lattice spacing between fuel assemblies in Region 2 is 0.04 inches. This corresponds to an uncertainty in reactivity of' O.0010 6k.
4 4 4 - 18
'T
4.6.3.4 Stainless Steel Thickness Tolerance The-nominal thickness of the stainless steel box wall is 0.075' inch with a tolerance of 10.005 inch, resulting in an uncertainty in reactivity of i 0.0003 Sk.
4.6.3.5 Fuel Enrichment and Density Tolerances Uncertainties in reactivity due to tolerances on fuel enrichment and UO density in Region 2 are assumed to be the.same'as those determined for Region 1.
)
l 4.7 .Puel Currently in Storace The enrichment and burnup characteristics of fuel now in storage in the Fort Calhoun spent fuel pool (as of September 1992) are shown - as data points- on Figure 4.7.1.- 'Almost all of the 529 .
i existing fuel assemblies would be acceptable for unrestricted i storage in-Region 2. However, a few assemblies fall below the bounding curve in Figure 4.7.1 for unrestricted storage 'in Region 2.- These assemblies may be safely stored-in the outer
-peripheral cells of Region 2 (i.e.,-a Region 3) as indicated _by.
the lower bounding curve in_ Figure 4.7.1.
l i
.l 1
i
.i 4 -19 i
4 4.8 ABNORMAL AND ACCIDENT CONDITIONS 4.8.1 Temoerature and Water Density Effects The moderator temperature coefficient of reactivity is negative; a moderator temperature of 4*C was assumed for the reference designs, which assures that the true reactivity will always be lower over the expected range of water temperatures. Temperature effects on reactivity have been calculated and the results are shown in Table 4. 8.1. Introducing voids in the water internal to the storage cell (to simulate boiling) decreased reactivity, as shown in the table.
With soluble poison present, the temperature coefficients of 4
reactivity would differ from those inferred from the data in
. Table 4.8.1. However, the reactivities would also be substan-tially lower at all temperatures with soluble boron present, and the data in Table 4.8.1 is pertinent to the higher-reactivity unborated case.
4.8.2 Eccentric Fuel Positionina The fuel assembly is assumed to be normally located in the center of the storage rack cell. Calculations were made for both Regions 1 and 2 with the fuel-assemblies assumed to be in the corner of the storage rack cell (fcur-assembly cluster at closest approach). These calculations indicated that eccentric fuel positioning.results in a small decrease in reactivity as deter-
! mined by KENO-Sa calculations. The highest reactivity, therefore, corresponds to the reference design with the fuel assemblies positioned in the center of the-storage cells.
4 - 20
4.8.3 Droceed Fuel Assembly i
For a drop on top of the rack, the fuel assembly will come to rest horizontally on top of the rack with a minimum separation distance from the fuel in the rack of more than 12 inches, including any deformation under seismic or accident conditions,
, At thic separation distance, the effect on reactivity would be
, insignificant (<0.0001 8k). Furthermore, soluble boron in the pool water would substantially reduce the reactivity and assure that the true reactivity is always less than the limiting value for any conceivable dropped fuel accident.
2 4.8.4 Lateral Rack Movement i Lateral motion of the rack modules under seismic conditions could potentially alter the spacing between rack modules. Region 2 storage cells do not use a finx-trap and the reactivity is j therefore insensitive to the spacing between modules. Soluble
- poison is present in the pool water and, in event of lateral rack movement in Region 1, would assure that a reactivity less than the regulatory limit is maintained under all conditions.
4.8.5 Abnormal Location of a Fuel Assembly The abnormal location of a fresh unirradiated fuel assembly of l 4.2 wt% enrichment could, in the absence of soluble poison, result in exceeding the design reactivity limitation ( k,, of 0.95). This could occur if a fresh fuel assembly of the highest permissible enrichment were to be inadvertently loaded into a Region 2 storage-cell. Soluble boron in the spent fuel pool
~
water, for which credit is permitted under these accident 4 - 21
i i
'l conditions, would assure that the reactivity is maintained.
I substantially less than the design limitation. KENO-Sa calcula-l tions indicate that a -concentration- of 80 ppm ; boron would be
{' adequate to maintain the k,, less than'O.95.
1 4'- 22
4.9 References for Section 4 (4.3.1] Private Communication, Jan Uden to S.E.Tu-rner, dated 12/4/90.
[4.3.2] OPPD, Technical Specification Spent Fuel Storage Rack, Table Al-3, 12/31/91.
(4.4.1] A. Ahlin, M. Edenius, H. Haggblom, "CASMO
.l - A Fuel Assembly Burnup Program," AE-RF-76-4158, Studsvik report (proprietary).
A. Ahlin and M. Edenius, "CASMO - A Fast Transport Theory Depletion Code for LWR Analysis," ANS Transactions, Vol. 26, p.
604, 1977.
M. Edenius et al., "CASMO Benchmark Re-port," Studsvik/ RF-78-6293, Aktiebol-aget Atomenergi, March 1978.
[4.4.2] Green, Lucious, Petrie, Ford, White, Wright, "PSR-63 /AMPX-1 (code package) , AMPX Modular Code System for Generating Coupled Multigroup Neutron-Gamma Libraritss from ENDF/B," ORNL-TM-3706, Oak Ridge National Laboratory, March 1976.
[4.4.3] R.M. Westfall et al., " SCALE: A Modular Code System for performing Standardized
, Computer Analyses for Licensing Evalua-i tion," NUREG/CR-0200, 1979.
[4.4.4] -S. E. Turner, " Uncertainty Analysis -
Burnup Distributions", in Proceedinas of a Workshoo on the use of Burnuo Credit in Soent Fuel Transport Casks, Sandia Report SAND-89-0018, October 1989.
(4.4.5) C.V. Parks, " Parametric Neutronic Analyses Related to Burnup Credit Cask Design" in Proceedinas of a Workshoo on the use of Burnuo Credit in Soent Fuel Transoort i
Casks, Sandia Report SAND-89-0018, October 1989.
i 4 - 23
l 1
i 1
I
[4.4.6) E.E. Pilat, " Methods for the Analysis of-Boiling Water Reactors (Lattice Physics),"
4 YAEC-1232, Yankee Atomic Electric Co.,
i 4 December 1980. .
4 j
(4.4.7) H. Richings, Some Notes on PWR (H) Power i Distribution Probabilities for LOCA Proba-
! bilistic Analyses, NRC Memorandum to P.S.
?
Check, dated July 5, 1977.
[4.5.1) M.G. Natrella, Experimental Statistics 4
National Bureau of Standards, Handbook 91, i August 1963.
l [-4 . 6.1) -Omaha.Public Power District, Amendment No.
133 to License DPR-40, NRC Docket 50-285, January 26, 1990.
} [4.6.2] Private Communication dated October 15, 1991, Jan Bostelman and Joe Willett to S.
{ E. Turner, Holtec International.-
i s
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!=
, 4 - 24 1
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(
i i
Table 4.2.1 I
SUMMARY
OF CRITICALITY SAFETY ANALYSES i
k j Region 1. Region 2 4
1 l Design Basis 4.2% enrichment 4.2% enrichment
- at 32000 Mwd /MtU i
Temperature for analysis 4*C 4'C
]
Reference k7 (CASMO-3) 0.9173 0.9124 4
l Uncertainties y
-In Bias 0.0024m t 0.0024*
B-10 loading i C.0030m 1 0.0032*-
Boral width i 0.0007m t 0.0005*
Inner box dimension 1 0.0010m 0.0010*
Water-gap thickness m 0.0090*L NA SS thickness i 0.0004Q t 0.0003*
Fuel enrichment 0.0022m t-0.0022*
Fuel density 1 0.0022* ! 0.0022*
i Eccentric position Negative *- Negative
- l Statistical combination 110.0103 t 0.0052 of uncottaintiesm 4
1
- Burnup Uncertainty NA 0.0133 .
i j Axial Burnup Distribution NA + 0.0038*
f Total 0.9173 1 0.0103 0.9295 0.0052
{
l- Maximum Reactivity (39)- 0.928 0.935-i
- Section 4.4.1 m Section 4.5.2 t
m Section 4.6.3-
- LFor fuel-tolerances, uncertainties.lin Region-2 assumed to be the same as1those for Region 1.-
L
!~
- Section 4.8.2 l
- Section'4.4.2~
P m Square root of sum _of squares.
{ Section 4.4.3 4
- 4 ,
l
- __ . . , . . . . . . , - - - , _ . ..,_r_ , , , . - , . . . . , , . , , _ , . _
i l
- l. I o ,
s
's j-
'I
) ,
l l Table 4.2.2
- i 3 REACTIVITY EFFECTS OF ABNORMAL AND ACCIDENT CONDITIONS i
j Accident / Abnormal Conditions Reactivity Effect Temperature increase (above 4*C) Negative- (Table-4.8.1) f Void (boiling) Negative (Table 4.8.1)_
{ Assembly dropped on top of rack Negligible'(<0.0001-8k) i i
{ Lateral rack module movement Positive (controlled by in seismic event soluble boron)-
Misplacement of a fuel assembly Positive,(controlled by-soluble boron) s f
2 l
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i i
1 3
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4
)
i-f 1
l Table 4.3.1
- j DESIGN BASIS FUEL ASSEMBLY SPECIFICATIONS ,
4 .
l- FUEL ROD DATA W CE ANF i
l
- Outside diameter, in. 0.440 0.440 0.442 f Cladding inside diameter, in. 0.384 0.388 0.378 9
Cladding material Zr-4 Zr-4 Zr-4 Pellet density, % T.D. 95 95 95 Stack density, g UO 2/cc (i 0.20) 10.29 10.05 10.19'
! Pellet diameter, in. 0.3765 0.3785 0.370 1
Enrichment, wt %.U-235 (10.05) 4.20 4.20 4.20 ft-i' 3
FUEL ASSEMBLY DATA Fuel rod array- 14x14 l
'! Number of fuel rods 176-Fuel rod pitch, in. 0.580 j
Number of control-rod guide and 5 '
instrument thimbles
[ Thimble'O.D., in. (nominal) '1.115 l Thimble I.D., in. (nominal) 1.035 l-i-
i r
I i
i.
1:
j t i
3
.j' I 4 - 27 i --
I' l-
l 2
Table 4.6.1 DESIGN PARAMETERS FOR CEA ROD ASSEMBLIES Number rods per Assembly 5 Poison Material Bg2 Poison Length, in. 127 Poison Material Density, g/cc 1.38*
Poison Material OD, in 0.86 Sheath Material Inconel 625 i
Sheath OD, in. 0.948 Sheath ID, in. 0.868
- Assumed to be 73% of theoretical density (2.52 g/cc) and depleted to 75% of initial density during in-core operation.
a 4 - 28
4 i
a Table 4.6.2
SUMMARY
OF CRITICALITY ANALYSES ,
WITH CEA ASSEMBLY INSTALLED Design Basis 4.2% enrichment Temperature for analysis 4*C t
Reference k, (CASMO-3) 0.9174 l Uncertaintiesd)
In Bias 0.0024 B-10 loading i 0.0032 Boral width 0.0005 Inner box dimension i 0.0010 SS thickness i 0.0003 l Fuel enrichment 0.0022 Fuel density i 0.0022 Eccentric position Negative Statistical combination 1 0.0052 of uncertainties Burnup Uncertainty NA j Axial Burnup Distribution NA Total 0.9174 1 0.0052 Maximum Reactivity (k.) 0.923(2) 1 (u Manufacturing uncertainties assumed same as Region 2.
. (2) Corresponding k, from NITAWL-KEN 05a calculation is 0.9167- i 0.0051 or a maximum k,,, of 0.922.
I 4 - 29 4
. . . . - _ . - . . . . . . - . - . ~ - . . . . _ ~. - .. .-.. . . . . . . . . . - . . . . - . . . . . . . . - . = - . . . . . .-
4 i
i l' ;
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4
- Table 4.8.1.
4 i EFFECT OF TEMPERATURE ' AND VOID ' ON CALCUIATED
- ' REACTIVITY OF STORAGE RACK i
i I
Case Incremental Reactivity Change,.bk l
i Region 1- Region 2-4 1
- 4*C (39'F) . Reference Reference 20'C (68'F)- -0.0011 -0.0014
- 40'C-(104*F) -0.0035- -0.0034 80*C (176'F) -0.0103 -0.0087 t
120'C (248'F) -0.0194 -0.0157 5
l -120'C (248'F) + 20% void -0.0421' -0.0333 i
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j 1
j APPENDIX A I
e BENCHMARK CALCULATIONS 1
I by l
6 Stanley E. Turner, PhD, PE HOLTEC INTERNATIONAL
}
August, 1992 l
1 a
Y
1.0 INTRODUCTION
AND
SUMMARY
i-
-The objective of this benchmarking study is to verify both the NITAWL-KEN 05 a(1,2) methodology with the 27-group SCALE cross-section library and the CASMO-3 codeU) for use in criticality l.
safety calculations of high density spent fuel storage racks. Both l
d calculational methods 'are based upon transport theory and have been l benchmarked against critical experiments that simulate typical spent fuel storage rack designs as realistically as possible.
Results of these benchmark calculations with both methodologies are l consistent with corresponding.. calculations reported in the j literature.
l Results of the benchmark - calculations show that the l 27-group (SCALE) NITAWL-KEN 05a calculations consistently under-l predict the critical eigenvalue by 0.0101 i 0.0018 8k (with a.95%.
probability at a 95% confidence level) for critical experiments")
that are as representative as possible of realistic spent fuel
- storage rack configurations and poison worths.
Extensive benchmarking calculations of critical experi-I ments with CASMO3-have also been reported d ), giving'a mean kg , of
- 1.0004 0.0011 for 37 cases. With a K-factor of 2.144) for 95%
! probability at a 95% confidence level, and conservatively neglect-
, ing the small overprediction, the CASMO3 bias then becomes 0.0000 0.0024. CASMO3 and NITAWL-KEN 05a intercomparison calculations of infinite arrays of poisoned cell configurations (representative of typical spent fuel storage rack designs) show very good agreement, I confirming that O.OOOO i O.0024 is a reasonable bias and uncertain-ty for CASMO3 calculations. Reference 5 also documents good a
agreement of heavy nuclide concentrations for the Yankee core
-isotopics, agreeing with the measured values within experimental error.
A-1 1
e p- s$ +-
re m- tv + v w-
k i
] The benchmark calculations reported here confirm that
] either the 27-group (SCALE) NITAWL-KENO or CASMO3 calculations are l acceptable for criticality analysis of high-density spent fuel storage racks. Where possible, reference calculations for storage rack designs should be performed with both code packages to provide
! independent verification. CASMO3, however, is not reliable when large water gaps ( > 2 or 3 inches) are present.
2.0 NITAWL-KENO Sa BENCHMARK CALCULATIONS Analysis of a series of Babcock & Wilcox critical l
j experiments *, including some with absorber panels typical of a i poisoned spent fuel rack, is summarized in Table 1, as calculated with NITAWL-KEN 05a using the 27-group SCALE cross-section library l and the Nordheim resonance integral treatment in NITAWL. Dancoff f factors for input to NITAWL were calculated with the Oak Ridge
{ SUPERDAN routine (from the SCALES system of cr 'es) . The mean for these calculations is 0.9899 0.0028 (1 a standard deviation of
- the population). With a one-sided tolerance factor = corresponding to 95% probability at a 95% confidence level *, the calculational bias is + 0.0101 with an uncertainty of the mean of 0.0018 for the sixteen critical experiments analyzed.
l
- Similar calculational deviations have been reported by ORNL* for some 54 critical experiments (mostly clean criticals l
without strong absorbers), obtaining a mean bias of 0.0100 0.0013 j (95%/95%). These published results are in good agreement with the results obtained'in the present-analysis and lend further credence j to the validity of the 27-group NITAWL-KEN 05a calculational model-1 for use in criticality analysis of high density' spent fuel storage racks. No trends in k m with intra-assembly water gap, with absorber panel reactivity worth, with enrichment or- with poison concentration were identified, comparable ~ to those previously -
[ observed
- with the 123-group GAM-THERMOS cross-section library.
A-2
(
J Additional benchmarking calculations were also made for a series of French critical experiments
- at 4.75% enrichment and for several of the BNWL criticals with 4.26% enriched fuel.
Analysis of the French criticals (Table 2) showed a tendency to overpredict the reactivity, a result also obtained by O E d1M. The calculated k,,, values showed a trend toward higher values with decreasing core size. In the absence of a significant enrichment effect (see Section 3 below), this trend and the overprediction is attributed to a small inadequacy in NITAWL-KENO 5a in calculating neutron leakage from very small assemblies.
Similar overprediction was alt:0 observed for the BNWL 4
series of critical experimentsOU, which also are small assemblies
- (although significantly larger than the French criticals) . In this case (Table 2), the overprediction appears to be small, giving a mean k,,, of 0. 9 959 i O . 0013 (1 o population standard deviation).
Because of the small size of the BNWL critical experiments and the absence of any significant enrichment effect, the overprediction is also attributed to the f ailure of NITAWL-KEN 05a to adequately treat neutron leakage in very small assemblies.
s Since the analysis of high-density spent fuel storage racks generally does not entail neutron leakage, the observed inadequacy of NITAWL-KENO 5a is not significant. Furthermore, omitting results of the French and BNWL critical experiment 1
analyses from the determination of bias is conservative since any-leakage that might enter into the analysis would tend to result in overprediction of the. reactivity.
?
2 A-3
~
-w- - -
- - -- - . ~ . ~ . , -- . -
- 3.0 WORKER ROUTINE The WORKER routine was obtained from ORNL and is intended to interpolate tha hydrogen scattering matrices for temperature in
- order to correct for the deficiency noted in NRC Information Notice 91-66 (October 18, 1991). Benchmark calculations were made against CASMO3, based on the assumption that two independent methods of analysis would not exhibit the same error. Results of these
- calculations, shown in Table 4, confirm that the trend with
! temperature obtained by both codes are comparable. This agreement i establishes the validity of the WORKER routine, in conjunction with
! NITAWL-KEN 05a, in calculating reactivities at temperatures other than 20*C (the reference library temperature).
{
- The deficiency in the hydrogen scattering matrix does not
[ appear except in the presence of a large water gap where the
, scattering matrix is important. However, the absolute -value of the km from CASMO3 is not reliable in the presence of a large water gap, although the relative values should be accurate. In the l calculations shown in Table 4 and in Figure 1, the absolute i reactivity values differ somewhat but the trends with temperature 4
are sufficiently in agreement to lend credibility to the WORKER
! routine, 1
l 4.O CLOSE-PACKED ARRAYS i
l The BAW close-packed series of critical experimentso2) intended l to simulate consolidated fuel, were analyzed with NITAWL-KENO 5a.
Results of these analyses, shown in Table 5, suggest a slightly
, higher bias than that for fuel with normal lattice spacings.
! Because there are so few cases available-for analysis, it is l recommended that the maximum - bias for close-packed lattices be 4 taken as 0.0155, including uncertainty. This would conservatively encompass all but one of the cases measured.
3 Similar-results were obtained by ORNLN.
s.
d A-4 i
- .- . ,, . w e r e-e v <: w ' ,-w- n w w
l l
5.0 CASMO3 BENCHMARK CALCULATIONS The CASMO3 code is a multigroup transport . theory code utilizing transmission probabilities to accomplish two-dimensional calculations of reactivity and depletion for BWR and PWR fuel assemblies. As such, CASMO3 is well-suited to the criticality analysis of spent fuel storage racks, since general practice is to treat the racks as in infinite medium of storage cells, neglecting leakage effects.
CASMO3 is a modification of the CASMO-2E code and has been extensively benchmarked against both mixed oxide and hot and cold critical experiments by Studsvik Energiteknik s)t . Reported ana-lysea ts) of 37 critical experiments indicate a mean k,,, of 1.0004 i O.0011 (10). To independently confirm the validity of CASMO3 (and to investigate any effect of enrichment), a series of calculations were made with CASMO3 and with NITAWL-KEN 05a on identical poisoned storage cella representative of high-density spent fuel storage racks. Results of these intercomparison calculations * (shown in Table 3 ar. in Figure 1) are within the normal statistical variation of KENO calculations and confirm the bias of 0.0000 i O.0024 (95%/95%) for CASMO3.
Since two independent methods of analysis would not be expected to have the same error function with enrichment, results i of the intercompadson analyses (Table 3) indicate that there is no significant effect of fuel enrichment over the range of enrich-ments involved in power reactor fuel. Furthermore, neglecting the French and BNWL critical benchmarking in the determination of bias is a conservative approach.
i
- Intercomparison between analytical methods is a technique endorsed by Reg. Guide 5.14, " Validation of Calculational litfub for Nuclear Criticality Safety".
A-5
t l
1
, A second serier of CASMO3-KEN 05a intercomparison calculations j l consisting of five cases from the BAW critical experiments analyzed )
i for the central cell only. The calculated results, also shown in
- Table 3, indicato a mean difference within the 95% confidence limit
- of the KEN 05a calculations. This lends further credonce to the i recommended bias for CASMO3.
1 a
h i
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3 I
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)
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1 4
A-6
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6.0 REFERENCES
TO APPENDIX A >
- 1. Green, Lucious, Petrie, Ford, White, and Wright, "PSR /NITAWL-1 (code package) NITAWL Modular Code System For Generating Coupled Multigroup Neutron-Gamma Libraries from ENDF/B", ORNL-TM-3706, Oak Ridge National Laboratory, November 1975.
- 2. R.M. Westf all et. al. , " SCALE: A Modular System for Performing Standardized Computer Analysis for Licensing Evaluation",
NUREG/CR-02OO, 1979.
- 3. A. Ahlin, M. Edenius, and H. Haggblom, "CASMO - A Fuel Assembly Burnup Program", AE-RF-76-4158, Studsvik report.
A. Ahlin and M. Edenius, "CASMO - A Fast Transport Theory Depletion Code for LWR Analysis", ANS Transactions, iol. 26,
- p. 604, 1977.
"CASMO3 A Fuel Assembly Burnup Program, Users Manual",
Studsvik/NFA-87/7, Studsvik Energitechnik AB, November 1986
- 4. M.N. Baldwin et al., " Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel", BAW-1484-7, The Babcock & Wilcox Co., July 1979.
- 5. M. Edenius and A. Ahlin, "CASMO3: New Features, Benchmarking, and Advanced Applications", Nuclear Science and Fncineerina, 100, 342-351, (1988)
- 6. M.G. Natrella, -Excerimental Statistics, National Bureau of Standards, Handbook 91, August 1963.
l
- 7. R.W. Westfall and J. H. Knight, " SCALE System Cross-section Validation with Shipping-cask Critical -Experiments", ALS.
Transactions, Vol. 33, p. 368, November 1979
- 8. S.E. Turner and M.K. Gurley, " Evaluation of NITAWL-KENO Benchmark Calculations for High Density Spent Fuel' Storage Racks", Nuclear Science and Enaineering, 80(2):230-237, February 1982.
A-7 e
m- y w- -,w-%
'I f
r j
- 9. J.C. Manaranche, et. al. , " Dissolution and Storage Experiment I
with 4.75% U-235 Enriched UO 2 Rods", Nuclear Technoloov, Vol.
I 50, pp 148, September 1980 f 10. A.M. Hathout, et. al., "Validat.on of Three Cross-section Libraries Used with the SCALE System for Criticality Analy-sis", Oak Ridge National Laboratory, NUREG/CR-1917, 1981.
- 11. S.R. Bierman, et. al., " Critical Separation between Sub-critical Clusters of 4.29 Wt. % 235U Enriched UO2 Rods in Water with Fixed Neutron Poisons", Battelle Pacific Northwest Laboratories, NUREG/CR/OO73, May 1978 (with August 1979
, errata).
)
- 12. G.S. Hoovler, et al., " Critical Experiments Supporting i Underwater Storage of Lightly Packed Configurations of Spent i
Fuel Pins", BAW-1645-4, Babcock & Wilcox Company (1981).
I 1 13. R.M. Westf all, et al. , " Assessment of Criticality Computation-i al Software for the U.S. Department of Energy Office of Civilian Radioactive Waste Management Applications", Section 6, Fuel Consolidation Applications, ORNL/CSD/TM-247 (undated) .
i 4
)
l A-8
i 1
1 Table 1 i
i RESULTS OF 27-GROUP (SCALE) NITAWL-KEN 05a CALCULATIONS OF B&W CRITICAL EXPERIMENTS Experiment Calculated o
- Number k,,,
I 0.9922 1 0.0006 i
II 0.9917 1 0.0005 l III 0.9931 1 0.0005 i
- IX 0.9915 1 0.0006
)
X 0.9903 i 0.0006 l
XI 0.9919 1 0.0005 l
XII 0.9915 1 0.0006
- XIII 0.9945 i 0.0006 XIV 0.9902 1 0.0006 i'
XV 0.9836 1 0.0006 XVI 0.9863 1 0.0006 XVII 0.9875 1 0.0006 XVIII 0.9880 1 0.0006
, XIX 0.9882 1 0.0005 3 XX 0.9885 i 0.0006 l XXI 0.9890 i 0.0006 l Mean 0.9899 1 0.000703 i Bias (95%/95%) 0.0101 1 0.0018
)
Standard Deviation of the Mean, calculated from the k,,, values.
. A-9
Table 2 RESULTS OF 27-GROUP (SCALE) NITAWL-KEN 05a CALCULATIONS OF FRENCH and DNWL CRITICAL EXPERIMENTS French Experiments Separation critical calculated Distance, cm Height, cm k,, ,
O 23.8 1.0302 i O.0008 2.5 24.48 1.0278 i O.0007 5.0 31.47 1.0168 i O.0007 10.0 64.34 0.9998 i O.0007 (2};i ', r;pers 'nts Calculated Case Myt. No. k,, ,
1 No Absorber 004/032 0.9942 0.0007 SS Plates (1.05 B) 009 0.9946 i O.0007 SS Plates (1.62 B) 011 0.9979 i O.0007 SS Plates (1.62 B) 012 0.9968 i O.0007 SS Plates 013 0.9956 i O.0007 SS Plates 014 0.9967 0.0007 Zr Plates 030 0.9955 i O.0007 Mean 0.9959 i O.0013 A - 10
. _. . __ . _ - - _ _ _ - - . -. . = _ -. _ . .- -. . . _ . -
4 Table 3 RESULTS OF CASMO3 AND NITAWL-KEN 05a i DENCHMARK (INTERCOMPARISON) CALCULATIONS Enrichment")
j Wt. % U-235 NITAWL-KEN 05aG) k"CASMO3 l8kl 2.5 0.8376 i O.0010 0.8386 0.0010 3.0 0.8773 i O.0010 0.8783 0.0010 3.5 0.9106 i O.0010 0.9097 0.0009 4.0 0.9367 i O.0011 0.9352 0.0015 j 4.5 0.9563 i O.0011 0.9565 0.0002
! 5.0 0.9744 i O.0011 0.9746 0.0002 Mean 0.0008 l Expt. No.03 XIII 1.1021 1 0.0009 1.1008 0.0013
- XIV 1.0997 i 0.0008 1.1011 0.0014 j XV 1.1086 i 0.0008 1.1087 0.0001 4 XVII 1.1158 i 0.0007 1.1168 0.0010 XIX 1.1215 1 0.0007 1.1237 0.0022 Mean 0.0012 i
U) Infinite array of assemblies typical of high-density spent fuel storage racks.
") k, from NITAWL-KEN 05a corrected for bias.
03 Contral Cell from BAW Critical Experiments A - 11
- r- + - - e 3 - - - , 4,--- . - . - + +
Table 4 Intercomparison of WORKER-NITAWL-KEN 05a and CASM03 Calculations at Various Temperatures TemDerature CASMO 3 W-N-KEN 05aM 4'c 1.2276 1.2345 t 0.0014 17.5'C 1.2322 1.2328 t 0.0015 25'C 1.2347 1.2360 2 0.0013 50'C 1.2432 1.2475 t 0.0014 75'C 1.2519 1.2569 i 0.0015 120'C 1.2701 1.2746 1 0.0014
- Corrected for bias i
l l
4 1
l l
4 5
A - 12
_ e _ __ - . . _ - . _ _ _ _ _ . _ _ - . - . _-. . _ _ . .- .=-
1 l
i j Table 5 Reactivity Calculations for Close-Packed Critical Experiments
) Calc. BAW Pin Module Boron Calculated l No. Expt. Pitch Spacing Conc. k, g
- No. cm cm ppm j)
KS01 2500 Square 1.792 1156 0.9891 0.0005 1.4097 KS02 2505 Square 1.792 1068 0.9910 i 0.0005
- 1.4097 1
KS1 2485 Square 1.778 886 0.9845 v.0005 Touching 1 KS2 2491 Square 1.778 746 0.9849 i 0.0005 "
I Touching KT1 2452 Triang. 1.86 435 0.9845 i 0.0006 Touching KT1A 2457 Triang. 1.86 335 0.9865 0.0006 Touching KT2 2464 Triang. 2.62 361 0.9827 i 0.0006 Touching f
KT3 2472 Triang. 13.39 121 1.0034 1 0.0006 Touching A - 13
i 5.0 THERMAL-HYDRAULIC CONSIDERATIONS 5.1 Introduction i
4 This section provides a summary of the methods, models, analyses j and numerical results to demonstrate the compliance of the reracked j Fort Calhoun Station (FCS) spent fuel pool with the provisions of Section III of the USNRC "OT position Paper for Review and t
j Acceptance of Spent Fuel Storage and 11andling Applications", (April i 14, 1978).
j Similar methods of thermal-hydraulic analysis have been used in
] previous licensing efforts on high density spent fuel racks for Fermi 2 (Docket No. 50-341), Quad Cities 1 and 2 (Docket Nos.
50-254 and 50-265) , Rancho Seco (Docket No. 50-312), Grand Gulf l
Unit 1 (Docket No. 50-416), Oyster Creek (Docket No. 50-219),
l i Virgil C. Summer (Docket No. 50-395), Diablo Canyon 1 and 2 (Docket l Nos. 50-275 and 50-323), Byron Units 1 and 2 (Docket Nos. 50-454
{ and 455), St. Lucie Unit One (Docket No. 50-335), Millstone Unit
{ One (Docket No. 50-245), Vogtle Unit 2 (Docket No. 50-425),
j Kuosheng Units 1 & 2 (Taiwan Power Company), Ulchin Unit 2 (Korea l Electric Power Company) , and J. A. FitzPatrick (Docket No. 50-333),
j D.C. Cook (Docket Nos. 50-315 and 50-316), Zion (Docket-Nos. 50-l 295 and 50-304), Sequoyah (Docket Nos. 50-327 ' and 50-328), and i Three Mile Island Unit One (Docket No. 50-289), among others.
4 l
The analyses to be carried out. for the thermal-hydraulic l
i qualification of the rack array may be broken down into the following categories:
{
! (1) Pool decay heat evaluation and pool bulk temperature
- variation.with. time.
(ii) Determination of the maximum pool local temperature
[ at the instant when the bulk temperature reaches
- its maximum value.
(iii) Evaluation of the maximum fuel-cladding temperature
!' to establish that bulk nucleate boiling at any location resulting in two . phase . conditions environment around the fuel is not possible.
l 5-?
(iv) Evaluation of the time-to-boil if all heat rejection ;
paths from the cooler are lost.
(v) Compute the effect of a blocked fuel cell opening on the local water and maximum cladding temperature. '
The following sections present a synopsis of the taethods employed to perform such analyses and final results.
5.2 Soent Fuel Pool Coolina System and Cleanun System DescriDtion The spent fuel pool cooling system consists of two 900 GPM storage pool circulation pumps, a 9x10 8 Btu /hr. storage pool heat exchanger, a demineralizer and- filter, two fuel transfer canal drain pumps, piping, valves and instrumentation. The spent-fuel pool cooling system is a seismic class I system.
The storage pool pumps circulate borated water through the storage pool heat exchanger and return it to the pool. Cooling water to the heat exchanger is provided by - the component cooling water system, which is also a seismic class I system. The purity and clarity is maintained by diverting a portion of the circulated water through the domineralizer and the filter. The fuel transfer canal drain pumps are utilized to:
- a. provide pool make-up water from the safety injection and
- refueling water (SIRW) tank;
- b. Drain the fuel transfer canal and return the refueling water to the SIRW tank or the radioactive waste disposal system (WD).
During refueling periods the domineralizer and filter provide a purification system for the refueling. water in the containment refueling - cavity. The reactor coolant drain tank puraps take suction from the containment refueling cavity and circulate the borated water through the demineralizer and filter and return it to the spent fuel pool.
5-2
i t
To ensure availability of adequate heat removal capacity following
{ unloading of a full core, blanked-off connections are provided in the pool cooling piping for a tie to the shutdown cooling system.
) 5.3 Decay Heat Load Calculations i
The decay heat load calculation is performed in accordance with the provisions of "USNRC Branch Technical Position ASB9-2, " Residual Decay Energy for Light Water Reactors for Long Term Cooling", Rev.
2, July, 1981.
5.4 Discharae Scenarios i
The ongoing reracking will generate a total of 1083 fuel storage cells. The thermal-hydraulic analysis considered a maximum of 1337 regular fuel bundles stored in the pool for the heat generation.
More fuel assemblies imply higher decay heat load. Therefore, the maximum heat load in the pool will be calculated conservatively in the analysis. The decay heat calculations are based on the previous
! and projected cycles as shown in Tables 5 . 4 .1A and 5.4.18. The
] following discharge scenarios are considered 15 months after the projected refueling for cycle 27, which generates a total of 1159 fuel assemblies in the pool.
. Case one
, Full core of 133 fuel assemblies are discharged to the pool for the scheduled normal refueling batch 28. The core is assumed to have 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in-core decay and is transferred to the pool at a rate of 3 assemblies / hour (Figure 5.4.1) . 45 assemblies of the core are
, assumed to have 1575 days of operation, 44 assemblies with 1050 4
days of operation, and 44 assemblies with 525 days of operation.
Case Two The reactor begins to operate after 56 days outage for the Case one discharge, and experiences an emergency shutdown after 30 days of operation. The core is transferred to the pool at a rate of 3 assemblies per hour after 144 hours0.00167 days <br />0.04 hours <br />2.380952e-4 weeks <br />5.4792e-5 months <br /> decay in the reactor (Figure 5.4.2). The last batch (45 assemblies) in the core has 30 days operation; 44 assemblics have 555 days of operatica, and the last 44 assemblies have 1080 days of operation.
5-3
1 Table 5.4.2 provides the major input for the scenarios. !
i j 5.5 Bulk Pool Temoeratures
,' In order to perform the analysis conservatively, the heat exchangers are assumed to be fouled to their design maximum and 5%
j of the heat exchanger tubes are assumed to be plugged. Thus, the temperature effectiveness, p, for the heat exchanger utilized in the analysis is the lowest postulated value calculated from heat exchanger thermal-hydraulic codes. The value of p is assumed to j remain constant in the calculation. It is noted that the FCS spent fuel cooling heat exchanger has not had any tubes plugged in more than 18 years of operation.
The mathematical formulation can be explained with reference to the I
simplified heat exchanger alignment of Figure 5.5.1.
1, Referring to the Spent Fuel Pool Cooling System, the governing
- differential equation can be written by utilizing conservation of energy
dT C =Qt - Qgx (5-1) dr QL= Pw , + Q (r) - Qgy (T, t) where:
C: Thermal capacitance of the pool (net water volume times water density and times heat capacity), Btu /*F.
1 Q:t Heat load to the heat exchanger, Btu /hr.
Q(r): Heat generation rate from recently discharged fuel, which is a specified function of time, r, Btu /hr.
5-4
j P =8Pt e Heat generation rate from "old" fuel, Btu /hr.
4 (P = average assembly operating power, Btu /hr) 0,g : Heat removal rate by the heat exchanger, Btu /hr.
- Quv (T, t.)
- Heat loss to the surroundings, which is a function of pool temperature T and ambient temperature t., Btu /hr.
Q,g is a non-linear function of time if we assume the temperature effectiveness p is constant during the calculation. 0g 3 can, however, be written in terms of effectiveness p as follows:
I Qit x = W C, p (T - t i) (5-2) 1 p"
0 ~b T-t where:
W: Coolant flow rate, Ib./hr.
j C,: Coolant specific heat, Btu /lb.*F.
4 p Temperature effectiveness of heat exchanger.
T: Pool water temperature, 'F 4
t:i Coolant inlet temperature, 'F i
t:o Coolant outlet temperature, 'F p is determined by the heat exchanger design basis performance.
Q(r) is specified according to the provisions of "USNRC Branch Technical Position ASB9-2, " Residual Decay Energy for Light Water Reactors for Long Term Cooling", Rev. 2, July, 1981. Q(r) is a function of decay time, number of assemblies, and in-core exposure time. During the fuel transfer, the heat load in the pool will increase with respect to the rate of fuel transfer and equals Q(r)
. after the fuel transfer.
5-5
. i e
i
' t i
Ogy is a non-linear function of pool temperature and ambient temperature. Ogy contains the heat evaporation loss through the-i pool surface, natural convection from the pool surface and heat i conduction through the pool walls and slab. Experiments show that
(
- the heat conduction takes only about 4% of the total heat loss l
l (5.5.1), therefore, it can be neglected. The evaporation heat and I
3 natural convection heat loss can be expressed ast
)
- Osv = a r A, + he A, e (5-3) .
1 i where:
m Mass evaporation rate, ib./hr. ft.2 P Latent heat of pool water, Btu /lb.
l A, Pool surface area, f t.2 i
1 h, convection heat transfer coefficient at pool d surface, Btu /ft.2 hr. 'F 0 = T-t : The temperature-difference between pool water and i ambient air, 'F j
The mass evaporation rate a can be obtained as a non-linear ,
i function of 9. We, therefore, have t 4
1
- a = ho (e) (W, - W.) (5-4) i
! where:
i
- W
- p Humidity ratio of saturated moist air at pool water l surface temperature T.
s l W.: Humidity ratio of saturated moist air at ambient j, temperature t, ho(0) : Diffusion coefficient at pool water surface. ho is
! a non-linear function of 9, lb./hr. ft. 'F a.
[ The non-linear single order differential equation -(5-1) is solved
- using Holtec's Q. A. validated numerical-integration code "ONEPOOL".
2 1
1- 5-6 l
we -r+. .m. - me,, w,n<-...--e w ~- , ,y,. .m..- . . , -
- < - , - , ~ , - .r...,v-.x.vw +, . , ~ ~ , , - , , - - - . - -
The next step in the analysis is to determine the temperature rise profile of the pool water if all forced indirect cooling modes are suddenly lost.
If the cooling makeup water is added at the rate of G lb/hr and the cooling water is at temperature, tw, the governing enthalpy balance equation for this condition can be written as dT (C + G(C ) (f - f,) ] = P. + Q(f + tw) +G ( C,) (tg - T) dr where water is assumed to have specific heat of unity - (C, = 1.0 Btu /lb.*F), and the time coordinate r is measured from the instant of loss-of-cooling. r, is the time coordinate when the direct addition (fire hose) cooling water application is begun. tw is the time coordinate measured from the instant of reactor shutdown to the instant of loss-of-cooling. T is tr.e dependent variable (pool water temperature). For conservatism, Ogy is conservatively assumed to remain constant af ter pool water temperature reaches and rises above 170*F.
l A Q. A. validated numerical quadrature code is used to integrate the foregoing equation. The pool water heat up rate, time-to-boil, and subsequent water evaporation-time profile are generated and l compiled for safety _ evaluation.
5.6 Local Pool Water Temoerature In this section, a summary of the methodology, calculations and L results for local pool vater temperature is presented.
5-7
5.6 1 Basis In order to determine an upper bound on the maximum. fuel cladding temperature, a series of conservative assumptions are made. The most important assumptions are listed below:
The fuel pool will contain spent fuel with varying time-af ter-shutdown (t,) . Since the heat emission falls off rapidly with increasing r,, it is conservative to assume that all fuel assemblies are from the latest batch discharged simultaneously in the shortest possible time and they all have had the maximum postulated years of operating time in the reactor. The heat emission rate of each fuel assembly is assumed to be equal and maximum.
As shown in the pool layout drawings, the modules occupy an irregular floor space in the pool. For the hydrothermal analysis, a circle circumscribing the actual rack floor space is drawn (Fig. 5.6.1). It is further assumed that the cylinder with this circle as its base is packed with fuel assemblies at the nominal layout-pitch.
The actual downcomer space around the rack module group varies. The nominal downcomer gap available in the pool is assumed to be the total gap available around the idealized cylindrical rack; thus, the maximum resistance to downward flow is incorporated into the analysis (Figs.
5.6.2 and 5.6.3) (i.e. minimum gap between the pool wall and rack module, including seismic kinematic effect).
No downcomer flow is assumed to exist between the rack modules.
No heat transfer is assumed to occur between pool water and ' he surroundings (wall, etc.)
The effect of the truncation of the sparger and suction line in the fuel pool is appropriately accounted for by setting the bottom plenum temperature equal to the spatial average temperature of the pool.
5.6.2 d2dpl Description In this manner, a conservative idealized model for the rack assemblage is obtained. The water flow is axisymmetric about the vertical axis of the circular rack assemblage, and thus, the flow is two-dimensional (axisymmetric three-dimensional) . Fig. 5.6.2 5-8
shows a typical " flow chimney" rendering of the thermal-hydraulics model. The governing equation to characterize the flow field in the pool can now be written. The resulting integral equation can be solved for the lower plenum velocity field (in the radial direction) and axial velocity (in-cell velocity field), by using the method of collocation. The hydrodynamic loss coefficients which enter ihco the formulation of the integral equation are also taken from well-recognized sources [5.6.1) and wherever discrepancies in reported values exist, the conservative values are consistently used. Reference (5.6.2) gives the details of mathematical analysis used in this solution process.
After tr ?xial velocity field is evaluated, it is a straight-forward , .er to compute the fuel assembly cladding temperature.
The knowl.dge of the overall flow field enables pinpointing of the storage location with the minimum axial flow (i.e, maximum water outlet temperatures). This is called the most " choked" location.
In order to find an upper bound on the temperature in a typical cell, it is assumed that it is located at the most choked location.
Knowing the global plenum velocity field, the revised axial flow through this choked cell can be calculated by solving the Bernoulli's equation for the flow circuit through this cell. Thus, an absolute upper bound on the water exit temperature and maximum fuel cladding temperature is obtained. In view of the aforementioned assumptions, the temperatures calculated in this manner overestimate the temperature rise that will actually occur in the pool. Holtec's computer code THERPOOL , based on the theory of (5.6.2), automates this calculation. The analysis 1.ocedure embodied in THERPOOL has been accepted by the Nuclear Regulatory THERPOOL has been used in qual!fying the spent fuel pool reracking of over 20 projects, including Enrico Fermi Unit 2 (1980), Quad Cities 1 and 2 (1981), Oyster Creek (1984), V.C.
Summer (1984), Rancho Seco (1983), Grand Gulf 1 (1985), Diablo Canyon 1 and 2 (1986), St. Lucie Unit-One (1987), Millstone Unit One (1989), Hope Creek (1990), Three Mile Island Unit 1 (1991), and
- others, 5-9
Commission on - several dockets. The code THERPOOL for local temperature analyses includes the calculation of void generations.
The effect of void on the conservation equation, crud layer in the clad, and the clad stress calculation when a void exists, are all incorporated in THERPOOL. The major inputs for the local temperature analysis are given in Table 5.6.1.
5.7 Claddina Temoerature The maximum specific power of a fuel array qr can be given by gr=qP n (5-5) where:
-F, = radial peaking factor q = average fuel assembly specific power The maximum temperature rise of~ pool- water in the most disadvantageously placed fuel assembly is = computed for all loading cases. Having determined the maximum local water temperature in -
the pool, it is now possible to determine the maximum fuel cladding temperature. A fuel' rod can produce F, times ' the average heat
- emission rate over a small length, where F, is the axial rod peaking j factor. The axial heat distribution' in a rod is generally a maximum in the central region, and tapers off at its two extremities. The 5
peaking factors used for the FCS spent fuel pool are shown in Table 5.7.1.
It can be shown that the power distribution corresponding to the-l'
- chopped cosine power emission rate is given by w (a + x)
. q(x) = qA 810 1 + 2a
, 5-10 1
+,,,e y-- wr-- <e,r-www-=,<m-. -~ --r"- **" 7
where:
1: active fuel length k .
as chopped length at both extremities in the power curve
, x: axial coordinate with origin at the bottom of the active fuel region The value of a is given by 1z a=
1-22 where:
1 1 1 1 2 z= - - +
! n F, n2 F,2 , p, g2 i
where F, is the axial peaking factor. I l
- The cladding temperature T, is governed by a third order j differential equation which has the form of j d3 T d3 T dT t
3
+ a3 2 a2 "f (X) dx dx dx 4
where og, a2 and f(x) are functions of x, and fuel assembly geometric properties. The solution of this differential equation with appropriate boundary conditions provides the fuel cladding j temperature and local water temperature profile.
.J In order to introduce some additional conservatism in the analysis, we assume that the fuel cladding has a uniform crud deposit over the entire surface developing a thermal resistance of .005 *F-sq.
ft-hr/ Btu.
5-11
5.8 Results It is shown in the calculations that the decay heat load of the "old" fuel assemblies (background heat load) in the pool (27 refueling cycles of 1159 assemblies) is 3.60 x 106 Btu /hr during the final discharges specified in Section 5.4. Table 5.8.1 gives the general input for the bulk pool temperature analyses. The maximum bu,P pool temperature results are presented in Table 5.8.2.
The time va. ing bulk pool temperatures and heat load in the pool are plotted vs. time-af ter-shutdown in Figures 5.8.1 to 5.8.4. It is shown from the analyses that the maximum bulk pool temperature resulted from the limiting full core offload is 135'F at 122 hours0.00141 days <br />0.0339 hours <br />2.017196e-4 weeks <br />4.6421e-5 months <br /> after-reactor-shutdown. The coincident heat load to the cooling system is 20.72 x 10' Btu /hr (excluding 0.25 x 10' Btu /hr evaporation heat losses). The maximum calculated temperature is well below the temperature guidelines for both normal and abnormal conditions specified in the NUREG-0800 Standard Review Plan, Section 9.1.3. The temperature limit of 140*F prescribed in the FCS USAR is also maintained.
The loss-of-cooling events have been considered for the specified discharge scenarios. The loss of all forced cooling is assumed to occur at the instants of peak pool temperature. Table 5.8.3 summarizes the results of the time-to-boil and maximum evaporation rate. The calculated minimum time from the loss of pool cooling until the pool boils is 9.90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> and the maximum boil-off rate is 33.4 gpm. The water elevations in the pool are plotted vs. time af ter loss-of-cooling in Figures 5.8.5 and 5.8.6 assuming no makeup water is added to the pool.
Table 5.8.4 gives the results of the maximum local water temperature and maximum local fuel cladding temperature for the limiting discharge scenario (Case 1) . Calculations are performed assuming non-blockage and 50% blockage, respectively. The blockage is assumed to occur in the thermally limiting storage cell by a 5-12
- m. . .. ._._.. _.___ _._.- - - . . _ . _ _ . _ _ _ _ _ . _ _ _ . _ _ . _ . . . .__. _ -. ._
i
! i i :
i j horizontally placed (misplaced) fuel assembly. It is shown that j the calculated maximum local temperatures will not cause local 5
1 nucleate boiling at any location in the racks. .
5.9 References for section s J !
(5.5.1) Wang, Yu, " Heat Loss to the Ambient From Spent
} Fuel Pools: Correlation of Theory with 1 Experiment", Holtec Report HI-90477, Rev. O, April 3, 1990.
4 (5.6.1) General Electric-Corporation, R&D Data Books, 4 " Heat Transfer and Fluid Flow", 1974 and j updates.
i l (5.6.2) Singh, K.P. et al., " Method for Computing the Maximum Water Temperature in a Fuel Pool 4 Containing Spent Nuclear Fuel", Heat Transfer l Engineering, Vol. _7, No.1-2, pp. 72-82 (1986) .
i i
1 s
i.
4 i
j t
)
l i
r t
a I a
i I.
4-
[. ~ 5-13 1
j ii i
+-cr, r.~w,-w. , ,+.-q~r-wu- w., -, .-.we, , , - - - , _ , . - ,. ,-.~..y ..,p.+, ,.w., ,.*,,;<_. ,,.,,-,y re.:y,... .-.=ce t-c-w.++-= ev ' W' v
l l
i
$ i 1
I j Table 5.4.1A i i j EXISTING DISCHARGE HISTORY '
a j (Total of 529 Assemblies) -t 1
a
! Bounding j Full Months
- Power After i Cycle Number of operation Discharge Previous. '
! Number Assemblies J1gy.g Date Discharae
]
1 25 1575 04/10/75 0 2 36 1575 12/08/76 19.9 j- 3 52 1575 11/16/77 11.3 l 4 44 1575 12/10/78 12.8 .
l 5 40 1575 04/16/80 16.2 j 6 40 1575 12/01/81 19.5 l 7 20 1575 02/18/83 14.6 -
i
! 8 26 1575 06/18/84 16.0 9 65 1575 12/16/85 18.0
- j. 10 45 1575 03/23/87 15.2 l- 11 44 1575 10/30/88 19.3 i '
- 12 40 -1575 -02/15/90 15.5-l l 13 52 1575 02/15/92 -24.0' i
I
.{
i t
i
- . 5-14
?
j
, ,__. ___.._..___-.-.a.-
.-. -- ..,-..- .. . ,,_-..,_...--.,_.,,,,,w.-_....
, ,,.. ~ _ -. , . . . . . . .
Tablo 5.4.1D PROJECTED DISCHARGE SCHEDULE Assumed Full Month / Power Number of Cumulative Cycle Year of Operation Assemblies Assemblies Hymber Discharae Dagg Discharced Discharced 14 8/1993 1575 45 574 15 2/1995 1575 45 619 16 8/1996 1575 45 664 17 2/1998 1575 45 709 18 8/1999 1575 45 754 19 2/2001 1575 45 799 20 8/2002 1575 45 844 21 2/2004 1575 45 889 22 8/2005 1575 45 934 23 2/2007 1575 45 979 24 8/2008 1575 45 1024 25 2/2010 1575 45 1069 26 8/2011 1575 45 1114 27 2/2013 1575 45 1159 28 8/2014 1575 45 1204 5-15
t i
i 4
1 Table 5.4.2 l
i DATA FOR DISC 11ARGE SCENARIOS Number of assemblies in refueling batch: 44 or 45 i
humber of assemblies in full core: 133 Number of fuel pool coolers in the SFPCS: 1 Fuel normal exposure time, hrs.: 37800 Fuel transfer rate, assemblies /hr.: 3 3
J I
t 5-16 t.
. . . .- _ . . . _. . . - - - _ . - . - - - - - . - - . . . _ - _ _ = . - . - - . . . . - . . . _ _
l I 4 .
i i
4 i b
l '
i f >
l Table 5.6.1 j DATA FOR LOCAL TEMPERATURE f Type of fuel assembly PRR 14x14 l Fuel cladding outer diameter, inches 0.44 1
i Fuel cladding inside diameter, inches 0.388 ,
l storage cell inside dimension, inches 8.46 j Active fuel length, inches 128 l
!. Number of fuel rods / assembly 176 i
j operatiyg power per fuel assembly 38.47 Po x 10 , Btu /hr i
l ' Cell pitch, inches' -8.652 .
t i
! Cell height, inches -161 3 Bettom height, inches 5.25 i
i . Plenum radius, feet 19.55 t i
i Peripheral average rack-to-wall gap, inches 2.29 i
4 i
I i,
i 5-17 r
- _, --, _ - . _ .- . _ , _ _ _ _ , - _ . _ _ _ ._,_._-...._,;..._;._,.~--. _. ,_ . . - . .
Table 5.7.1 PEAKING FACTORS Factor Value Radial 1.955 Axial 1.4605 Total 3.067' l
5-18 l
l l
..._ _ . . . ....___ . . _ . - . _ . . . _ _ _ . . _ _ _ . . _ _ . _ _ _ _ ~ . _ _ . _ _ . . . _ _ _ _ . . - _ . . . . _ . . .
i J !
l 2
i i
4 1
i I
4 i ,
d j i 1
Table 5.8.1 i
i FUEL SPECIFIC POWER AND POOL CAPACITY DATA i
j 5 j- Net water volume of pool, gal. 164,000-
, Fuel pool thermal capacity,10' Btu /'F 1.34 4
Avprage operating power of a fuel assembly, l 10 Btu /hr 38.47
}' SFP Cooler Coolant-inlet temperature, F 90 t
f- SFP Cooler Coolant flow rate, ~10' lb/hr - : 0.50 t
I J
i i
i 1
f a
o i
l r
4 r a
i i
4 a
' 5-19 a
gTg - * + eq9 N+7 *(rr'~ e-iw--- y =t--+e- e y e' --y -a *y-*----teru.tt"'~.v49 * - 8'
_ _ _ . _ . _ _ . . _ ~ . . . _ . . _ --_ . . _ - - . . . - _ . . . _ _ . .
e E
4 l
6 i
i i
1
! 4 l
r L
Table-5.8.2 r
SFP BULK POOL TEMPERATURE -
V
+
i Coincident! Coincident- Coincident Maximum Time After -Heat Load Evaporation
- Pool Reactor Shut- .to Hept Losses
-10,SFP . BtuHxs Temp., 'T down, hrs._- /hr- 10 Btu /hr Discharge 134.88 122 20'.72 0.253 I in Case-1 '
H -Discharge 128.03 195 _17.~ 5 8 0.162 in Case 2 -s f
l I-t-
1 l
5-20_
~_ . . _ . . _ , . - , . . -, ..2 - . . . - _ _ __ 1,.,.
1, 4
1 4
i P ,
i- .
1 l-1 a ,.
4 i
4 i
i I
4 Table 5'.8.3~
i i
i TIME-TO-BOIL RESULTS 7
l Time-to-Boil-(Hours)_ -Maximum-i (Without Make-Up- Evaporation 4
Case No. Water) - Rate (GPM)
J 4
['
1 9.90- 33;4 i 2 10.82 - 32.6 i-i i:
i' 4
i 5
1 4
k I
i
, y Y
I
- k. _.
I 1,.'
, r.,, , ~ . , - , ,,vy..--er , ,y -
-r--*sm * - -v v. w, + - o w k - r-r. *v-
1 i
i i
1 5
4 1
l i
4 1
i i.
i k
i ,
4 Table 5.8.4' l.
MAXIMUM LOCAL POOL WATER AND FUEL CLADDING TEMPERATURE l
FOR THE LIMITING CASE (Case 1) i i
i l l Maximum- Maximum-Local Local
! Pool Fuel.
Water Cladding-~
Temo..
- F *F-Temo..
- .No Blockage 207.9- 254.9 l.
}- 50% Blockage 221.2 - 264.9-i b '
i e
1..
l i
L l4-i 1
~5-22
= * . -. .- ,e'= m e a a .vi '- e y . U r ,- , , ,y- y e v - < v - g , o- y , 3 y-(
HOLTEC INTERNATIONAL o
ONE NORMAL REFUELING BATCH o
O POOL FrLLED EXCEPT w STORAGE CAPACITY FOR b 1 NORMAL BATCH E AND 1 FULL CORE REMA!NtNG RELOAD 88 FAS BACK TO REACTOR
~
) V s ,
W 72 HOURS / -
s = -
/g i /
7 ) FULL CORE (133 FAS) OFFLOAD AT 3 FAS/HR 5
__ __ _ . . _ . _ _ . -__.___.-_.__-_._--_-.-_.-a--- ----
REACTOR SHUTDOWN END OF OUTAGE FIGURE 5.4.1 FT. CALHOUN FUEL POOL DISCHARGE SCENARIO ONE
, . _ - . . . . . - - ~ ~ - _ _ . - - . _ . - _ . . - - . . _ , _ _ . . _ . . _ _ _ _ _ . . . _ . _ . - _ . . _ . . . . _ _ _ _ . . . . . ~ . _ _ _ _ . . _ _ ._...-_m.~_____._.
h HOLTEC INTERNATIONAL a
NORMAL BATCH IN SCENARIO ONE EMERGENCY FULL CORE OFFLOAD 8
O (L
5 g
30 DAG RELOAD 88 FAS BACK TO REACTOR E
(
j e
') ) REACTOR ~144 HOURS
.$ ' OPERATING
.f 72 HOURS _ -
g N ( (
.a
. FULL CORE OFFLOAD FULL CORE (133 FAS) OFFLOAD AT 3 FAS/HR
?
fo 56 DAYS 30 DAYS
. _ . . _ _ . . . _ . . _ _ , i REACTOR SHUTDOWN - END OF OUTAGE REACTOR SHUTDOWN i '
FIGURE 5.4.2
, FT. CALHOUN FUEL POOL DISCHARGE SCENARIO TWO t
s i
4
1 4
k 4
3 1
i >
i, I
k EVAPORATIONHEATLOSS i
i
! - SPENT FUEL FOOL =
i I
t HEATEXCHANGER P-Tg
.e m.,.
i .
1 1I
.~ COOLANT tg W, .
4 4
i FIGURE 5.5.1.
BULK POOL ANALYSIS MODEL-.
i.
- - 5-25
~
1 1-I id e all: e d O u tlin e "*'
j Of Ra c k .A s s embly Cut!!ne of i
Rack Assembly _
- /
Y
\
/
/ ///dm/ RACK ASSEMBLY s
!~
l-
[
~
_i.
~ - Actual Outline of Pool Ideall:e d -
Outline of .
Assumed Added i- Pool Boundary Fuel Assemblies
- j. FIGURE' 5.6.1 i_
IDEAllZATION OF -R ACK ASSEMBLY -
.5-26
. . . . - . __ , , . ~ . . .
t j
J J-
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l
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l.li ,- ,
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- FIGURE 5.6.2 i
THERMAL CHIMNEY FLOW MODEL S-27 u, _. -g- -w. <,_._n-
< -yy.,,,m.,w.,_w, e ., r e y, wy ,,w..,
,,-.,w -,a,,,,,,m3, .3-.,w f;wre- ,+, .w., , , . , ,
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4 i o l A 4 Y ; i i i 4-i .
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'i -
4
! 4 4 COMER 4
4 .
.a 4
% ' s 4
,.4 44 1 . .
- m. . . o. 6 .
g,3 ,
..a . . . .- ...,.
e PL ENU M FIGURE 5.6.3 CONVECTION CURRENTS IN THE POOL.
5-28
- .. . . _ - _ .. _ . , . . - . . _ . . - . . ~
HOLTEC INTERNATIONAL FCS. RERELING RLL CORE OFFLOAD - CASE 1 i
REACTOR SHLITDOWN 140 _
it : --
130 _
en -
1 2 0 _-
i w -
4 -
110 _
w :
i .-
100
/ -
i 9e ,,,,,,,,,ii..,,,,,,,,,,,,,,,,,,,,,,,,,,,
0 100 200 300 400 TIPE AFTER REACTOR FIRST SHJTDOWN, HRS
- FIGURE 5.8.1 DULK POOL TEMPERATURE - CASE 1
. _ . . . ~ _ _ _ - . _ _ _ _ _ . . _ _ . _ . _ . , _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ . _ _ . . . _ . . . . . . . _ . _ . . . . . _ _ _ _ _ _ _ . _ _ . . , _ . _ . . . . _ . - - ._.. _._ _ . _ _.
I i
.HCLTEC INTERNATIONAL :
FCS. EPER0ENCY FtLL CORE OFFLOAD - CASE' 2 :
l REACTOR SHLEXXN
'130 1
.g _
i m _
t 120 l
T u
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l 1 u 11e -
r
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f.
i- 100 iii,,,iii_ .ii. iiiiiiii..i . ,,. .
! 1344- 1444 1544 1644 1744 ,
I' TITE AFTER REACTOR ' FIRST SHJTDOWN, HRS -
i FIGUR E - S.8.2 BULK POOL. TEMPERATURE - CASE 2 :
s
p l
l l
HOLTEC INTERNATIONAL FCS, REFLELING FlLL CORE OFFLOAD - CASE 1 REACTOR SERJTD06N 2.50E+7 -
Sui Suu Ans Gud emu
- ET FEAT LOAD 2.00E+7 _
b m
W8' e
s -
m u, .
1.50E+7 _ %
u o -
~ 4 O 2
) -
b- 2 to 1.00E+7 I .m emu eux eul aus W
en 5.00E+6 I
eum W
ens
~
EVAPORATION FEAT LOSSES
~
0.00E+0 ,,,iii,iiii ,iiii,i iiiiii,iiiii,i,,iii
-0 100 200 300 400 TITE AFTER REACTOR FIRST SHJTDOLN, FRS
}
FIGURE 5.8.3 HEAT LOAD - CASE 1
HOLTEC INTERNATIONAL FCS, EPERGENCY FtLL CORE OFFLOAD - CASE 2 REACTOR SHJTD06N 2.00E+7 _
ET FEAT LOAD g 1.80E+7 2
m d -
O w J :
a F n < 1. 00E+ 7 _
w _
I .
5.00E+6 _
EVAPORATION FEAT LOSSES 0.00E+0 i,,,,,,,,,,iiiiiii,iii . ii..,,, .
,iii 1344 1444 1544 '1644 1744 TIPE AFTER REACTOR FIRST SHJTDOLN, FRS FIGURE 5.8.4 HEAT LOAD - C ASE 2
HOLTEC INTERNATIONAL FCS SFP LDSS OF COOLING - CASE 1 (LAST REFLELING FlLl_ CORE OFFLOMD i i
LOSS OF COOLING 4e.e t- _
as.e
\
y : N
- . w o
y
~
\
u 6 -
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1
.o- 2e 4e de se ses
! TIPE AFTER LOSS OF COOLING. FRS FIGURE 5.'8.5 - WATER ELEVATION AFTER LOSS-OF-COOLING - C ASE 1
. . . . , . - _ . . _ . . . . . . - . .. . - _ . - - .- . - - . - . . _ . . - . - . - . . . - . . - - - . - . . . = . _ = . .
'HOLTEC' INTERNATIONAL FCS SFP LDBS OF COOLING - CASE 2 (EPER0ENCY FU_L CORE OFFLOAD $ ,
LOSS OF COOLING 4e.e N
. I g m.e :
x y i5 ,
- s. :
~
\
. as.e I
. g
- N '
s :
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N\
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,,,, ,,,, ,,,, ,,,, ,,,, ,,,, ,,,, ,,,, iii, ,,,,
! e' 2e 4e 'de se les TIPE AFTER LOSS OF COOLING FRS FIGURE 5.8.6 WATER ELEVATION ' AFTER LOSS-OF-COOLING - C ASE 2
- _ _ . . _ . _ . _ _ _ _ _ _ _ _ . _ _ _ _ . _ _ _________________--m
6.0 STRUCTURAL / SEISMIC CONSIDERATIONS 6.1 Introduction This section contains analyses to demonstrate structural adequacy of the high density spent fuel rack design under seismic loadings postulated for the plant spent fuel pool. Analyses and subsequent evaluations are in compliance with the requirements of tL OT Position Paper,Section IV [6.1.1], and follow the USNRC Standard Review Plan (SRP) [6.1.2]. The dynamic analyses employ a time-history simulation code used in previous licensing efforts listed in Table 6.1.1. This section provides details of the method of analysis, modeling assumptions, numerical convergence studies and parametric evaluations performed to establish the required margins of safety.
Results reported herein show that the high density spent fuel racks are structurally and kinematically adequate to meet requirements defined in references [6.1.1], [6.1.2], and.in the ASME Code (6.1.3] with large margins of safety.
6.2 Analysis outline 1
a The spent fuel rack is a seismic category I structure (6.2.1]. It is a free-standing structure consisting of discrete storage cells which are loaded with free-standing fuel assemblies. The response of a rack module to seismic inputs is highly nonlinear involving a complex combination of motions (sliding, rocking, twisting, and turning), resulting in impacts and friction effects. Linear methods such as modal analysis and response spectrum techniques cannot accurately simulate the structural response of such a highly nonlinear structure to seismic excitation. A correct simulation is obtained only by direct integration of-the nonlinear equations of motion using actual pool slab acceleration time-histories to provide the loading. Therefore, as an initial step in
, spent fuel rack qualification, synthetic time-histories for three l 6-1 i
orthogonal directions are developed in compliance with the guidelines of USNRC SRP (6.1.2]. In particular, the synthetic time-histories must meet the criteria of statistical independence and enveloping of the design response spectra.
As stated above, a free-standing spent fuel rack, subject to a seismic loading, executes nonlinear motions - even when isolated.
The motion of an array of closely spaced racks in the spent fuel pool involves additional interactions due to fluid coupling between adjacent racks and between racks and adjacent walls.
Further mechanical interactions between racks occur if rack-to-rack impacts take place during the event. To demonstrate structural qualification, it is required to show that stresses are within allowable limits and that displacements remain within the constraints of the contemplated design layout for the pool. This implies that impacts between rack modules, if they occur, must be confined to locations engineered for this purpose, such as the baseplate edge and possibly the upper region of the rack above the active fuel region. Similarly, rack-to-pool wall impacts, if engineered into the rack design (not contemplated for these racks), must be within stipulated limits. Impact loads between pedestal and liner must be assessed to assure liner integrity.
Accurate and reliable assessment of the stress' field and kinematic behavior of the rack modules calls for a comprehensive and conservative dynamic model which incorporates all key attributes of the actual structure. This means that the model must feature the ability to execute concurrent sliding, rocking, bending, twisting and other motion forms available to the rack modules.
- Furthermore, the model must possess the capability to effect momentum transfers which occur due to rattling of the fuel assemblies inside the storage cells and due to impacts of support pedestals on bearing pads. Finally, the contribution of the water mass in the interstitial spaces around the rack modules and within storage cells must be modeled in an accurate manner because erring in the quantification of fluid coupling on either side of the 6-2
actual value is no guarantee of conservatism. Similarly, the coulomb friction coefficient at the pedestal-to-pool liner (or bearing pad) interface may lie in a rather wide range and a conservative value of friction cannot be prescribed a priori. In fact, a perusal of results of rack dynamic analyses in numerous dockets (Table 6.1.1) indicate that an upper bound value of the coefficient of friction, p, often maximizes the computed rack displacements as well as the equivalent elastostatic stresses.
Further, the analysis must consider that a rack module may be fully or partially loaded with fuel assemblies or entirely empty.
The pattern of loading in a partially loaded rack may also have innumerable combinations. In short, there are a large number of parameters with potential influence on the rack motion. A comprehensive structural evaluation should deal with all of these without sacrificing conservatism.
The 3-D single rack dynamic model introduced by Holtec I
International in the Enrico Fermi Unit Two rack project (ca. 1980) and used in some twenty rerack projects since that time (Table
, 6.1.1) considers the above mentioned array of parameters in a most appropriate manner. The details of this classical methodology are published in the permanent literature (6.2.2) and have been widely replicated by other industry groups in recent years. Briefly speaking, the single rack 3-D model handles the array of variables as follows:
i Interface Coefficient of Friction Parametric runs are made with upper bound and lower bound values of the coefficient of friction. The-limiting values are based on experimental-data [6.4.1).
Imoact Phenomena Compression-only gap elements are used to provide for opening and closing of interfaces such as the pedestal-to-bearing pad interface.
6-3 b
l Fuel Loadine Scenarios The fuel assemblies are conservatively assumed to rattle in unison which obviously exaggerates the contribution of impact against the 4
cell wall. The different patterns of possible fuel assembly loadings in the rack are simulated by orienting the center of gravity column of the assemblage of fuel assemblies with respect to the module geometric centerline in an appropriate manner.
Fluid Coucling The contribution of fluid coupling forces is ascertained by prescribing the motion of the racks (adjacent to the one being analyzed). Generally, it is assumed that for single rack analysis, the adjacent racks vibrate out-of-phase with respect to the rack being analyzed.
Despite the above simplifying assumptions, targeted for accuracy and conservatism, a large menu of cases is run to foster confidence in the calculated safety margins. Most of the safety analyses reported in the previous dockets (Table 6.1.1) over the past decade have relied on single rack 3-D model. From a conceptual standpoint, all aspeu a of the 3-D single rack model are satisfactory except for the fluid coupling effect. One intuitively expects the relative motion of the free-standing racks in the pool to be poorly correlated, given the random harmonics in the impressed slab motion. Single rack analyses cannot model this interactive behavior between racks. However, as described later, analytical and experimental research in this field has permitted rack analyses to be extended to all racks in the pool
- simultaneously. Holtec International had successfully extended Fritz's classical two-body fluid coupling model to multiple bodies and utilized it to perform the first two-dimensional multi-rack analysis (Diablo Canyon, ca. 1987). Subsequently, laboratory experiments were conducted to validate the multi-rack fluid coupling theory. This technology was incorporated in the computer
' code DYNARACK which now could handle simultaneous simulation of all racks in the pool. This development marked a pivotal expansion in the rack structural modeling capability and was first utilized in Chin Shan, Oyster Creek and Shearon Harris plants
! [6.2.3]. The Whole Pool Multi-Rack (WPMR) 3-D analyses have corroborated the uncanny accuracy of the single rack 3-D solutions in predicting the maximum structural stresses. The multi-rack analyses also serve to improve predictions of rack kinematics. No i
assumptions need be made concerning how the racks are moving with respect to one another.
t I
6-4
i In order to ensure utmost confidence in the results of structural analyses, results for both single rack 3-D and Whole Pool Multi-Rack (WPMR) 3-D analyses are obtained. The intent of this parallel approach is to foster added confidence and .to uncover any peculiarities in the dynamic response which are germane to the structural safety of the storage system.
The following summarizes the sequence of model developmat and analysis steps that are undertaken. Subsequent subsections provide model detail, limiting criteria for stress and displacement, and results of the analyses.
- a. Prepare three-dimensional dynamic models of individual fuel racks which embody all elastostatic characteristics and structural nonlinearities of the free-standing rack modules.
- b. Perform 3-D dynamic analyses of single racks on limiting module geometry types (from all those present .n i the spent fuel pool) and include various physical conditions (such as coefficient of friction, extent of cells containing fuel assemblies, and proximity of other racks).
- c. Perform detailed stress analysis for the limiting case of all the dynamic analysis runs made in the foregoing steps.
Demonstrate compliance with ASME Code Section III, subsection NF [6.1.3) limits on stress and displacement.
t
- d. Perform a degree-of-freedom (DOF) reduction procedure on the single rack 3-D model and evolve a Reduced DOF model for each rack such that kinematic responses calculated by the RDOFM are in general agreement with' responses obtained using the baseline single rack models of step (b). The RDOFM is also truly three-dimensional.
- e. Prepare a whole pool multi-rack dynamic model which includes the RDOFM's of A1.1, rack modules in the pool, and includes all fluid coupling interactions among them, as well as fluid coupling interactions between racks and pool walls. This 3-D simulation is referred to as a Whole Pool Multi-Rack (WPMR) model,
- f. Perform 3-D Whole Pool Multi-Rack (WPMR) analyses to demonstrate that all kinematic criteria for the spent fuel rack modules are satisfied, and that resultant structure loads confirm the validity of the structural qualification.
The principal kinematic criteria - are (i) no rack-to-pool wall impact, and (ii) rack-to-rack impacts, if they occur, are confined to hardened regions above and below the active fuel region of the racks.
6-5
6.3 Artificial Time-Histories Section 3.7.1 of the SRP (6.1.2) provides guidelines for establishing seismic time-histories. Subsection 3.7.1.II.1.b gives applicable criteria for generation of time-histories from design response spectra.
There are two options for generating seismic time-histories:
Option 1, Single Time-History and Option 2, Multiple Time-Histories. For both horizontal and vertical input motions, either a single time-history or multiple time-histories can be used.
Option 1 requires consideration of both Response Spectra and Power Spectral Densities and pre-supposes that a target PSD is available. Option 2 requires only that-target response spectra be available. For rack analysis purposes, Option 2 is used in generations of seismic time-histories from Design Earthquake response spectra (DE) and from Maximum Hypothetical Earthquake (MHE) response spectra.
Option 2 requires, as a minimum, that four time-histories should be considered for analyses. The response spectra calculated for each individual time-history need not envelope the design response spectra. However, the multiple time-histories are acceptable if the average calculated response spectra generated from these time-histories envelop the design response-spectra.
The acceptance criterion for spectrum enveloping (either for individual or for average values) is that no more than five points of the spectrum obtained from the time-history f all below, and no more than 10% below, the design response spectrum. The SRP states that an acceptable method of comparison is to choose a set of frequencies such that each frequency is within 10% of the previous one for each of the four time-histories. The nature of the spent 6-6 .
I
i 4
fuel rack structure is such that primary response- is to 4 excitations above 5-8 HZ. Within the 5-33 BZ range, discrete check points are established from the above 10% criterion.
l i'
Generated artificial time-histories- must also be statistically j independent. Any two time-histories 'are considered to be statistically independent if their normalized correlation coefficient is less than 0.15.
i l Figures 6.3.1-6.3.4 give the governing response spectra applicable to the spent fuel pool at the Fort Calhoun Station. The response f spectra are taken from (6.3.1) and are for the Maximum
! Hypothetical Earthquake (MHE) and for the Design Earthquake (DE).
j The horizontal spectra are at elevation 989', while the vertical l spectra are valid for any elevation. The response spectra _were
- originally derived by others from a building structural model'
- incorporating 5% and 2% structural damping to' obtain time-
' histories. These two histories were then_ used to generate equipment response spectra using 0.5% equipment damping. For the I
current spent fuel rack analyses, suitable time-histories from the four given spectra provided by Omaha Public Power District are
- j. developed. For conservatism, all subsequent fuel rack analyses using the developed time-histories assume 0.5% structural damping l
i and are scaled up as needed to reflect zero period. accelerations-l at higher elevations [6.3.1).
I The Holtec Proprietary- program GENEQ (6.3.2) generates L four sets l of synthetic- time-histories for the 'MHE 'and fori the DE, respectively. Each set consists of three, statistically I independent time-histories for-two horizontal, . and the vertical- -
i -directions, respectively. .
l l
l l I
f
, 6-7
i i
! Figures 6.3.5-6.3.16 show four - sets of generated time-histories-1
- for the MHE over the specified event time of'20 seconds-(Figures l 6.3.5-6.3.7
- Set 1; Figures 6.3.8-6.3.10: Set-2; . Figures 6.3.11-
! 6.3.13:-Set-3; Figures 6.3.14-6.3.16: Set-4).
The average calculated response spectra for 0.5% ~ damping from 1-j these time-histories are obtained by Holtec Proprietary program
!f AVESPC (6.3.3] and are shown in ' Figures 6.3.17-6.3.19. The i original spectra for MBE are also shown in these- figures for 4
l comparison.
- In calculating the response spectra of the four sets of MBE time-l histories, the total numbers of periods - at equal intervals on a
! logarithmic scale are 830,-920,_870 for the two horizontal and the vertical directions, respectively. The_ intervals are much smaller l
, than required by the acceptable-set provided by-Table 3.7.1-1 in i
l reference (6.1.2)..
1 i: It is clear from Figures 6.3.17-6.3.19 that the average response spectra for the four sets of. synthetic time-histories. meet the r requirement of enveloping the-corresponding' original spectra.
i l
! The normalized correlation coefficients-~pij between any two time-l histories i, j in each set are provided in: Table 6.3.1 for the: MBE l event. It shows that the generated time-histories meet the requirements of the statistical independence.
I
{ ; Figures 6.3.20-6.3.31 show the -four sets 'of generated time-
[ histories for the'DE over the event time of 20 - seconds -- (Figures -
Figures Figures j 6.3.20-6'.3.22: Set-1;' 6.3.23-6.3.25: ' Set-2;-
l- 6. 3. 2 6-6'. 3 2 8 : Set-3 ; l Figures 6. 3. 2 9-6'. 3. 31 : - Set-4 ) . - The average i calculated response ~ spectra for 0 5% t damping ' from- ;these: time-histories - and ' thel ' corresponding ; original- response spectra Lare
, shown in Figures'6.3.32-6'.3.34.
h h
. 6-8
4
- Figures 6.3.32-6.3.34 show that the average response spectra for the four sets of time-histories meet the requirement of enveloping i
the corresponding original spectra. In calculating the regenerated response spectra of the four sets of time-histories, the total numbers of periods at equal intervals on a logarithmic scale are 920, 827, 913 for the two horizontal and the vertical directions, respectively. The intervals are much smaller than required by the acceptable set provided in Table 3.7.1-1 in reference [6.1.2]. The normalized cross-correlation coefficients pij between any two time-histories i and j in each set are j provided in Table 6.3.2 for the DE event. Table 6.3.2 shows that the generated time-histories meet the requirement of statistical 4
independence.
Note that the response spectra are at elevation 989', while the
, top of the pool slab is at elevation 995'6". The induced
, horizontal accelerations vary with the elevation as shown in Figures 6.3.35 and 6.3.36, while the vertical acceleration is
] valid for any elevation.
The generated acceleration time-histories for the MBE and for the DE are for the elevation 989'. To obtain horizontal acceleration time-histories for the elevation of the spent fuel pool slab top surface, for the MBE and for the DE events, the corresponding generated time-histories are multiplied by amplifiers for MBE and for DE, respectively.
j From Figure 6.3.35, the amplifier for pool slab time-histories in horizontal directions for MBE is calculated as AMHE = 1.021435.
From Figure 6.3.36, the ampli.fier for pool slab time-histories in horizontal directions for the DE is calculated as ADE = 1.0229565.
6 s
1 6.4 Rack Modeline for Dynamic Simulations
)
6.4.1 General Remarks
\
l Spent fuel storage racks are Seismic Class I equipment. They are j required to remain functional during-and after an SSE event. The
.; racks are free-standing; they are neither anchored to the pool
[ floor nor attached to the sidewalls. Individual. rack - modules are not interconnected. Figure 6.4.1- shows ' a pictorial view of a l typical module. The baseplate extends beyond the cellular region
- envelope ensuring'that inter-rack impacts, if any, occur first at i
{ the baseplate elevation; this area is structurally qualifiable to
} withstand any large in-plane impact loads.
1 l A rack may be completely loaded with fuel assemblies (which
- corresponds to greatest total mass), or it may . be completely
- empty. The coefficient of friction, y, between pedestal supports
! and' pool floor is indeterminate. According to Rabinowicz (6.4.1], results _ of 199 tests performed on austenitic stainless
- steel plates submerged in water show a mean value of p to be 0.503
] with standard deviation of 0.125. Upper and lower bounds (based on twice- standard: deviation) are- 0.753 and 0.253,- respectively.
Analyses are therefore performed for coefficient- of friction values of 0.2 (lower limit) and for 0.8 (upper limit), and-for random friction values clustered.about a mean of 0.5. The bounding j values of p = 0.2 and .0. 8 ' have been found to _ bracket the - upper-
- limit of module response in previous rerack projects._
Since. free-standing racks are not anchored to the-pool slab,.not l attached to the pool walls, and not- interconnected, they can l execute a wide variety of motions. Racks may _ slide on the pool floor, one or more rack support pedestals may momentarily tip and lose contact with the iloor slab- liner, or - racks may exhibit a
, combination of . sliding and tipping. The structural models developed permit simulation of these -kinematic events with -
inherent built-in conservatisms. The rack models also include l
'f 6-10
components for simulation of potential inter-rack and rack-to-wall impact phenomena. Lift-off of support pedestals and subsequent liner impacts are modeled using impact (gap) elements, and Coulomb friction between rack and pool liner is simulated by piecewise linear (friction) elements. Rack elasticity, relative to the rack base, is included in the model with linear springs representing a beam like action. These special attributes of rack dynamics require strong emphasis on modeling of linear and nonlinear springs, dampers, and compression only gap elements. The term
- " nonlinear spring" is a generic term to denote the mathematical element representing the case where restoring force is not linearly proportional to displacement. In the fuel rack simulations, the coulomb friction interface between rack support pedestal and liner is typical of a nonlinear spring.
1 3-D dynamic analyses of single rack modules require a key modeling assumption. This relates to location and relative motion of neighboring racks. The gap between a peripheral rack and adjacent pool wall is known, with motion of the wall prescribed. However, another rack, adjacent to the rack being analyzed, is also free-standing and subject to motion during a seismic event. To conduct the seismic analysis of a given rack, its physical interface with neighboring modules must be specified. The standard procedure in analysis of a single rack module is to assume that neighboring racks move 180' out-of-phase in relation to the subject rack.
I Thus, the available gap before inter-rack impact occurs is 50% of
- the physical gap. This " opposed-phase motion" assumption increases likelihood of intra-rack impacts and is thus conservative. However, it also increases the relative i contribution of fluid coupling, which depends on fluid gaps and relative movements of bodies, making overall conservatism a less certain assertion. Whole Pool Multi-Rack 3-D analyses carried out for Taiwan Power Company's Chin Shan Station, and for GPU Nuclear's Oyster Creek Nuclear Station' demonstrate that single rack simulations predict smaller rack displacement during seismic 6-11
responses. Nevertheless, 3-D analyses of single rack modules permit detailed evaluation of stress fields, and serve as a benchmark check for the much more involved, WPMR analysis.
Particulars of modeling details and assumptions for 3-D Single Rack analysis and for Whole Pool Multi-Rack analysis are given in the following subsections.
6.4.2 The 3-D 22 DOF Model for Sincle Rack Module 6.4.2.1 Assumotions
- a. The fuel rack structure is very rigid; motion is captured by modeling the rack as a twelve degree-of-freedom structure. Movement of the rack cross-section at any height is described by six degrees-of-freedom of the rack base and six degrees-of-freedom at the rack top. Rattling fuel
- assemblies within the rack are modeled by five lumped masses located at H, .75H, .5H, .25H, and at the rack base (H is the rack height measured above the baseplate). Each lumped fuel mass has two horizontal displacement' degrees-of-freedom.
Vertical motion of the fuel assembly mass is assumed equal to rack vertical motion at the
, baseplate level. The centroid of each fuel assembly
- mass can be located off center, relative to the rack structure centroid at that level, to simulate a partially loaded rack.
- b. Seismic motion of a fuel rack is characterized by
! random rattling of fuel assemblies in their individual storage locations. All fuel assemblies are assumed to move in-phase within a rack. This exaggerates computed dynamic loading on the rack
- structure and therefore yields conservative
- results,
- c. Fluid coupling between rack and fuel assemblies, and between rack and wall, is simulated by appropriate inertial coupling in the system kinetic
-energy. Inclusion of these effects uses the methods of [6.4.2] and [6.4.3] for rack / assembly coupling and' for rack-to-rack coupling, respectively. Fluid coupling terms for rack-to-rack coupling are based -on opposed-phase motion of adjacent modules.
(
6-12
- d. Fluid damping and form drag is conservatively neglected.
- e. Sloshing is negligible at the top.of the rack and is neglected in the analysis of the rack.
- f. Potential impacts between rack and fuel assemblies are accounted for by appropriate " compression only" gap elements between masses involved. The possible incidence of rack-to-wall or rack-to-rack impact is simulated by gap elements at top and bottom of the rack in two horizontal directions. Bottom elements are located at the baseplate elevation.
- g. Pedestals are modeled by gap elements in the vertical direction and as " rigid links" for
, transferring horizontal stress. Each pedestal j support is linked to the pool liner by two friction springs. Local pedestal spring stiffness accounts
, for floor elasticity and for local rack elasticity just above the pedestal,
- h. Rattling of fuel assemblies inside the storage locations causes the gap between fuel assemblies and cell wall to change from a maximum of twice the nominal gap to a theoretical zero gap. Fluid coupling coefficients are based on the nominal gap.
6.4.2.2 Model Details Figure 6.4.2 shows a schematic of the model. Si (i = 1,...,4) represent support locations, pi represent absolute degrees-of-freedom, and gi represent degrees-of-freedom relative to the slab.
H is the height of the rack above the baseplate. Not shown in Fig. 6.4.2 are gap elements used to model pedestal / liner impact locations and impact locations with adjacent racks.
Table 6.4.1 lists the degrees-of-freedom for the single rack model. Translational and rotational degrees-of-freedom 1-6 and 17-22 describe the rack motion; rattling fuel masses (nodes 1*, 2*,
3*, 4*, 5* in Fig. 6.4.2) are described by translational' degrees-of-freedom 7-16. Ui(t) represents pool floor slab displacement seismic time-history.
6-13
I Figures 6.4.3 and 6.4.4, respectively, show inter-rack impact springs (to track potential for impact between racks or between rack and wall), and fuel assembly / storage cell impact springs at one location of rattling fuel assen61y mass.
Figures 6.4.5, 6.4.6, and 6.4.7 show the modeling technique and degrees-of-freedom associated with rack elasticity. In each bending pitae a shear and bending spring simulate elastic effects
[6.4.4). Linear elastic springs coupling rack vertical and torsional degrees-of-freedom are also included in the model.
Additional details concerning fluid coupling and determination of stiffness elements are provided below.
6.4.2.3 Fluid Couplina Details The " fluid coupling effect" [6.4.2),[6.4.3) is described as follows: If one body (mass mi) vibrates adjacent to a second body
' mass m2), and both bodien are submerged in frictionless fluid, Men Newton's equations of motion for the two bodies ares (mi + Mll) X1+M12 X2 = applied forces on mass mi + 0 (X12)
M21 X1 + (m2 + H22) X2 = applied forces on mass m2 + 0 (X2; 2 N N X2 denote absolute accelerations of masses mi and m2, res, *ively,
. and the notation O(X 2 ) denotes nonlinear terms.
M11, M12s M21, and H22 are fluid coupling coefficients which depend on body ch'pe, relative disposition, etc. Fritz (6.4.3) gives data for Mij for various body shapes and arrangements. The fluid adds mass to the body (M11 to mass mi), and an external force proportional to acceleration of the adjacent body (mass m2)-
Thus, acceleration of one body affects the force field on another.
This force field is a function of interbody gap, reaching large values for small gaps. Lateral motion of a fuel assembly inside a storege location encounters this effect. For example, fluid 6-14
l coupling is between nodes 2 and 2* in Figure 6.4.2. The rack analysis also contains inertial fluid coupling terms which model the effect of fluid in the gaps between adjacent racks. Terms modeling effects of fluid flowing between adjacent racks are computed assuming that all racks adjacent to the rack being analyzed are vibrating 100' out of phase from the rack being analyzed. Thus, the modeled rack is enclosed by a hydrodynamic mass computed as if there were a plane of symmetry located in the middle of the gap region. Rack-to-rack gap elements (Figure 6.4.3) have initial gaps set to 50% of the physical gap to reflect this symmetry.
6.4.2.4 Stiffness Elgment Details The cartesian coordinate system for the sirgle rack model has the following nomenclature:
x = Horizontal coordinate along the short direction of rack rectangular planform (N-S or E-Wj y = Horizontal coordinate along the long direction of the rack rectangular planform (N-S or E-W) z = Vertical coordinate upward from the rack base Note that this coordinate systema is specific to the comuputer model and associated progran code. The coordinate system is not associated with any defined x, y, x coordinate systems of the plant.
Table 6.4.2 lists cll spring elements used in the 3-D 22 DOF single rack model.
If the simulation model is restricted to two dimensions (one horizontal motion plus vertical motion, for example), for the purooses of model clarification only, then a descriptive model of 6-15 i
_ _ . . _ _ _ _ _ _ _ _ l
b 1
i I l l
the simulated structure which includes gap and friction elements i is shown in Figure 6.4.8. This simpler model is used to elaborate on the various stiffness modeling elements.
j Gap elements modeling impacts between fuel assemblies and rack have local stiffness XI in Figure 6.4.8. In Table 6.4.2, for example, gap elements 5 through 8 act on the rattling fuel mass at I the rack top. Support pedestal spring rates Ks are modeled by I elements 1 through 4 in Table 6.4.2. Local compliance of the i
concrete floor is included in Ks. Friction elements 2 plus 8 and 4 i plus 6 in Table 6.4.2 are shown in Figure 6.4.8. Friction at support / liner interface is modeled by the piecewise linear l friction springs with suitably large stiffness Kf up to the limiting lateral load, pN, where N is the current compression load at the interface between support and liner. At every time step during transient analysis, the current value of N (either zero if
- the pedestal has lifted off the liner, or a compressive finite
- value) is computed. Finally, support rotational friction springs Kg reflect any rotational restraint that may be offered by the
] foundation. The rotational friction spring rate is calculated using a modified Bousinesq equation [6.4.4] and is included to simulate resistive moment by the slab to counteract rotation of i
the rack pedestal in a vertical plane. The nonlinearity of these springs (friction elements 9,. 11, 13, and 15 in Table 6.4.2) reflects the edging limitation imposed on the base of the rack support pedestals and the shift in location of slab resistive load as the rack pedestal rotates.
The gap element KS, modeling the offective compression stiffness of the structure in the vicinity of. the support, includes
- stiffness of the pedestal, local stiffness of the underlying pool slab, and local st:2fness of the rack cellular structure above the pedestal.
6-16
The previous discussion is limited to a 2-D model solely for simplicity. Actual analyses incorporate 3-D motions and include all stiffness elemants listed in Table 6.4.2. .
6.4.3 Whole Pool Multi-Rack (WPMRI Model 6.4.3.1 General Remarks The single rack 3-D (22 DOF) model outlined in the preceding subsection is used to evaluate structural integrity, physical stability, ud to initially assess kinematic compliance (no rack-to-rack iqact in the cellular region) of the rack modules.
Prescribing the motion of the racks adjacent to the module being analyzed is an assumption in the single rack simulations. For closely spaced racks, demonstration of kinematic compliance is further confirmed by modeling all modules in one comprehensive simulation using a Whole Pool Multi-Pack (WPMR) model. In WPMR analysis, all racks are modeled, and their correct fluid interaction is included in the model.
6.4.3.2 Whole Pool Fluid Coucline The presence of fluid moving in the-narrow gaps between racks and between racks and pool walls-causes-both near and far field fluid l coupling effects. A single rack simulation can effectively include j only hydrodynamic effects due to contiguous racks when a certain f set of assumptions is used for the motion of contiguous racks. In I a WPMR analysis, far field fluid coupling effects.of all racks are
{ accounted f or - using - the correct model of: pool fluid mechanics.
[ The external hydrodynamic- mass due to the presence of . walls or
{ adjacent racks-is computedEin a manner consistent with fundamental fluid mechanics- principles '[6.4.5) :using .conservativo nominal-
-fluid gaps in the pool at~ the beginning _ of the - seismic . event.
Verification of the computed hydrodynamic ef fect E by . comparison
~
- with experiments'is also provided-in-[6.4.5). This formulation has
[ ' been reviewed and approved- by the -Nuclear Regulatory Commission l-i l 6 r
during post-licensing multi-rack analyses for the Diablo Canyon Unit I and II reracking project. The fluid flow model used to obtain the whole pool hydrodynamic effect reflects actual gaps and rack locations.
6.4.3.3 Coefficients of Friction To eliminate the last significant element of uncertainty in rack dynamic analyses, the friction coefficient is ascribed to the support pedestal / pool bearing pad interface consistent with Rabinowicz's data (6.4.1). Friction coef ficients, developed by a random number generator with Gaussian normal distribution characteristics, are imposed on each pedestal of each rack in the pool. The assigned values are then held constant during the entire simulation in order to obtain reproducible results.* Thus, the WPMR analysis can simulate the effect of different coefficients of friction at adjacent rack pedestals.
6.4.3.4 M2delina Details Figure 6.4.9 shows a planform view of the spent fuel pool which includes rack and pedestal numbering scheme and- the global coordinate system used for the WPMR analysis. Table 6.4.3 gives details on number of cells per rack, and on rack and fuel weights for Holtec new racks. In Whole Pool Multi-Rack analysis, a reduced degree-of-freedom (RDOF) set is used to model each rack plus contained fuel. The rack structure is modeled by six degrees-of-freedom. A portion of contained fuel assemblies is assumed to rattle at the top of trae rack, while the remainder of the
- Note that DYNARACh has the capability to change the coefficient of friction at .11ty pedestal at each instant of contact based on a random readine of the PC-clock cycle. However, exercising this option would yield results that could not be reproduced. Therefore, the rar.fom choice of coefficients-is made only once per run.
6-18
l contained fuel is assumed as a distributed mass attached to the rack. The rattling portion of the contained fuel is modeled by two horizontal degrees-of-freedom.
Thus, the WPMR model involves all racks in the spent fuel poo) with each individual rack modeled as an 8 degree of freedom structure. The rattling portion of fuel mass, within each rack, is chosen to ensure comparable results from displacement predictions from single rack analysis using a 22 DOF model and predictions from 8 DOF analysis under the'same conditions.
The Whole Pool Multi-Rack model includes gap elements representing compression only pedestals, representing impact potential at fuel assembly-fuel rack interfaces, and at rack-to-rack or rack-to-wall locations at top and bottom corners of each rack module. Each pedestal has two friction elements associated with force in the s vertical compression element. Values used for spring constants for i the various stitfness elements reflect the values used in the 22 1
l DOF model.
i
! 6.5 Mgsotance Criteria, Stress-Limits,-and Material P_rsoerties 6.5.1 M :entance Criteria
! There are two sets of criteria to be satisfied by the rack l modules:
i
- a. Kinematic Criteria l
! *he- rack must be a physically stable structure and 'it l' must be demonstrated that there are no -inter-rack j impacts in the cellular region.- The- criteria for i- physical- stability is that. -an isolated rack in water ,
! exhibit no overturning tendency when a seismic event of: .
magnitude 1.1 x SSE is applied [6.1.2). _t
- b. Stress Limit Criteria-
, Stress . limits must-not.be exceeded under certain load j
combinations. The following_ _ loading combinations; _are applicable (6.1.3) and only those -load elements .are-l included which pertain to spent fuel racks. !
l i
l 6-19 [
L Y v4,4,94'=-S S4 v4r ~- 4 '--m~' w is e ws" -wm e m * ~ = we- e-~N=-~ ~4-m'~~~- -e v w- *---w e- 'e- * ' wa s ' - + * "v***-'vemv'
}
.i 1
- Loadina Combination Service Level D+L Level A D + L + To l D + L + To + E D + L + Ta + E Level B D + L + To + Pg D + L + Ta + E' Level D D+L+Fd The functional capability i' of the fuel racks should be demonstrated.
Abbreviations are those used in Section 3.8.4 of the
, Standard Review Plan and the " Review and Acceptance of Spent Fuel Storage and Handling Applications" section:
l D = Dead weight-induced internal moments (including fuel assembly weight)
, L = Live Load (not applicable for the fuel rack, since there are no moving objects in the rack load path).
Fd = Force caused by the accidental drop of the heaviest load from the maximum possible height
. (see Chapter 7 of this report).
Pf = Upward force on the racks caused by postulated stuck fuel assembly (see Section 7)
E = Operating Basis Earthquake (DE for Fort Calhoun)
E' = Safe Shutdown Earthquake (MHE for Fort Calhoun)
= Differential temperature induced loads (normal To operating or shutdown condition based on the most critical transient or steady state condition)
= Differential temperature induced' loads Ta (the a
highest temperature associated with the postulated abnormal design conditions)
Ta and To cause local thermal stresses to be produced. For fuel rack analysis, only one bounding scenario need be examined. The worst situation is obtained when an isolated storage location has 6-20
a fuel assembly generating heat at maximum postulated rate and surrounding storage locations contain no fuel. Heated water makes unobstructed contact with the inside of the storage walls, thereby producing maximum possible temperature difference between adjacent cells. Secondary stresses produced are limited to the body of the rack; that is, support pedestals do not experience secondary l (thermal) stresses. For rack qualification, Te, Ta are the same.
6.5.2 Stress Limits for various conditions Stress limits are derived from the ASME Code, Section III, Subsection NF [6.1.3). Parameters and terminology are in accordance with the ASME Code.
6.5.2.1 Normal and Uoset Conditions (Level A or Level B)
- a. Allowable stress in tension on a net section ist Ft = 0.6 Sy (Sy = yield stress at temperature)
(Ft is equivalent to primary membrane stress)
- b. Allowable stress in shear on a net section ist Fy = .4 S y
- c. Allowable stress in compression on a net section (kl)2 2
[1 -
2
/2Ce Sy Fa "
5 kl kl 3 3
{( ) + [3 ( ) /8Cc) - [( ) /8C e ))
3 r r where:
(2n 2 E) 12
/
Ce = [ ]
Sy 6-21 t-
1 = unsupported length of component k = length coef ficient which gives influence of boundary conditions; e.g.
k =1 (simple support both ends)
= 2 (cantilever beam)
= 0.65 (clamped at both ends)
E = Young's Modulus r = radius of gyration of component kl/r for the main rack body is based on the full height and cross-section of the honeycomb region,
- d. Maximum allowable bending stress at the outermost fiber of a net section, due to flexure about one plane of symmetry ist Fb = 0.60 Sy (yield stress at temperature)
(equivalent to primary bending)
- e. Combined flexure and compression on a net section satisfies:
fa Cmx fbx Cmyf by
+ + <1 Fa DxFbx DFy by where fa = Direct compressive stress in the section fxb = Maximum flexural stress along x-axis fby = Maximum flexural stress along y-axis C mx =
Cmy = 0.85 for members with restrained ends.
= 1.00 for members whose ends are unrestrained.
fa Dx=1-F'ex fa Dy=1-F'ey 6-22 1
. . . . . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ - . - - _ -- J
l
'f 4 12 n2 E P'ex,ey = 2 1
23 ( )
r x,y and subscripts x,y reflect the particular bending plane.
- f. Combined flexure and compression (or tension) on a j not sections fa fx b fby
+ + < 1.0 l 0.6S y Fx b Fby The above requirements are to be met for both
- direct tension or compression.
4 6.5.2.2 Level D Service Limits Section F-1370 (ASME Section III, Appendix F), states that limits for the Level D condition are the minimum of 1.2 (Sy /Ft) or 4 (0.7Su/Ft) times the corresponding limits for the Level A condition. Su is ultimate tensile stress at the specified re ck design temperature. For example, if the material is such that J 1.2S y is less than 0.7Su, then the multiplier on the Level A limits, to obtain Level D limits, is 2.0.
6.5.2.3 Dimensionless Stress Factors Stress results are presented in dimensionless form. Dimensionless stress factors are defined as the ratio of the actual developed stress to the specified limiting value. Stress factors are only developed for the single rack analyses. The limiting value of each stress factor is 1.0 for the DE and 2.0 for the MHE condition.
Stress factors reported are:
= Ratio of direct tensile or compressive stress on a R1 net section to its allowable value (note pedestals only resist compression) 6-23
1 i
i k
i R2
= Ratio of gross shear on a net section in the x-j direction to its allowable value I = Ratio of maximum bending stress . due to bending R3 4
about the x-axis to its allowable value for the section R4
= Ratio of maximum bending stress . due to bending i about the y-axis to its allowable value for the j section R5
= Combined flexure and compressive factor (as defined I in 6.5.2.le above)
{ R6
= Combined flexure and tension (or compression)
- factor (as defined in 6.5.2.lf) i j R7
= Ratio of gross shear on a net section in the y-( direction to its allowable.value i
1 6.5.3 Material Procerties Physical properties of. the rack and support materials, obtained l
from the ASME Boiler & Pressure Vessel Code,Section III, 3
appendices, are listed in Table 6.5.1. Maximum pool bulk 1 temperature is less than 200 F; this is used as the reference s
l design temperature for evaluation of material properties. Stress i
limits for Level A,D, corresponding to conditions in Section 6.5.2 i above, are evaluated using given yield strength data.
l
, 6.6 Governina Ecuations of Motion I
Using the structural model for either 22 DOF single rack analysis,
{ or the set of s'implified 8 DOF models that' comprise a whole Pool ,
i Multi-Rack model, equations of motion corresponding-to each degreo.
1
- of freedom are obtained using Lagrange's Formulation [6.6.1]. The
- -system-kinetica energy-includes-contributicns from solid structures and from - trapped and surrounding fluid. The final system of
- equations obtained have the matrix form:
{q"}'= {Q} + {G)
['M ]
r F
6-24 4
_ . . _ . ,,_ _ _ _ _ _ __ a _ _ .. u _ _ _ __ .,_..._._n.~
where:
[M) -
total mass matrix (including structural and i fluid mass contribution 2;
{q} -
the nodal displacement vector relative to the pool slab displacement; (double primo stands for second derivatives with respect to time)
(C} -
a vector dependent on the given ground acceleration
{Q) - a vector dependent on the spring forces j (linear and nonlinear) and the coupling l between degrees-of-freedom The equations can be rewritten as
{q") = (M1-1 {Q) + (M1-1 (G)
- This equation set is mass uncoupled, displacement coupled at each instant in time; numerical solution uses a central difference scheme built into the proprietary, computer program "DYNARACK"
, (6.6.2 - 6.6.5). As indicated earlier, this program has been used in the licensing effort for a considerable number of reracking
- projects.
, DYNARACK has been validated against exact solutions, experimental data, and solutions obtained using alternate numerical schemes
[6.6.5). These solutions are chosen to exercise all features of DYNARACK. It is demonstrated there that well-known classical nonlinear phenomena (subharmonic resonance, bifurcation, stick-slip) can be reproduced using DYNARACK.
The application of DYNARACK to the spent fuel rack analysis requires the establishment of a time step to ensure convergence and stability of the results. DYNARACK utilizes the classical central difference algorithm (6.4.4). Stability of the results is assured as long as. the time step is significantly below the smallest period of the equivalent linear problem. Convergence is obtained by performing a series of rack analyses with different l time steps to ascertain the upper limit on time step that will i
l l
l 6-25 l
provide converged results. This is done by taking a typical rack i
module and subjecting it to the given time-histories using different integration time steps. Once an appropriate time step is determined, it is used in subsequent simulations.
1 i
Results of the dynamic simulations are time-history response of all degrees-of-freedom of the particular model, and of all forces and moments at important sections of the structure. From these results, maximum movements and stresses can be ascertained for the event, and appropriate structural qualifications can be carried a out. Where required, DYNARACK automatically tracks maximum values
! of dimensionless factors R1 to R7 defined above in Section 6.5,
- and reports results for the rack cross section just above the baseplate and for each pedestal cross section just below the-baseplate. These are the critical sections which develop the highest stresses due to the geometry of a fuel rack structure.
From the archived results, time-histories of all rack-to-rack fluid gaps, all rack-to-wall fluid gaps, and motion of any point on any rack can be generated. Sections 6.7 and 6.8 present results obtained from single and multi-rack analyses, respectively. The results demonstrate satisfaction of all requirements on structure and kinematic integrity.
6.7 Results of 3-D Nonlinear Analyses of Sincle Racks This section focuses on results from all 3-D single rack analyses.
In the following section, we present results from the whole pool multi-rack analysis and discuss the similarities and differences between single and multi-rack analysis.
A summary of results of all analyses performed for racks in the pool, using a single rack model, is presented in summary Tables 6.7.1 - 6.7.26. Table 6.7.1 lists all runs carried out. Table 6.7.2 presents the bounding results f rom all runs , and Tables 6-26
l 6.7.3 - 6.7.26 give details for each run. Analyses are carried out for different coefficients of friction, rack geometries, and fuel loading patterns. The heaviest rack (Rack F1 at a corner), the rack with the maximum ratio of side lenr~'. (Rack D), and a Region 1 rack (Rack A1) are chosen for single rack analysis. For single rack analysis, opposed-phase motions of adjacent racks is assumed to emphasize impact potential between racks. Analyses are performed assuming a fuel weight of 1380 lbs (dry). The racks have also been qualified for a heavier fuel weight of 2480 lbs, but these results are not germane to this license effort. The tabular results for each run give maximax (maximum in time and in space) values of stress factors at important locations in the rack. Results are given for maximum rack displacements (see Section 6.4.2.4 for x,y orientation), maximum impact forces at pedestal-liner interface, and rack cell-to-fuel, rack-to-rack, and rack-to-wall impact forces. The single rack analyses results show that no rack-to-rack or rack-to-wall impacts are likely to occur in the cellular region of the racks. In the single rack analysis, kinematic criteria are checked by confirming that no inter-rack gap elements at the top of the rack close (see Figure 6.4.3). By virtue of the symmetry assumption discussed in subsection 6.4.2.3, impact is likely to occur if the local horizontal displacement exceeds 50% of the actual rack-to-rack gap in the case of opposed-phase motion.
Structural integrity at various rack sections is considered by
- computing the appropriate stress factors Ri. Results corresponding to the MHE event yield the highest stress factors. Limiting stress factors for pedestals are at the upper section of the support and are to be compared with the bounding value of 1.0 (DE) or 2.0 (MHE). Stress factors for the. lower portion of the support are not limiting and are not reported. From the summary Table-6.7.2, all stress factors are below the allowable limits. All the runs in Table 6.7.1 are for the MHE event. Note that no runs for the DE i
l l
l 6-27
event are needed for single rack analysis because all the stress factors for the HHE event are lower than 1.0, thus meeting the allowable limits for the DE event.
Overturning has also been considered. A multiplier of 1.1 on MHE horizontal earthouakes is applied to an isolated rack and the predicted displacements examined. Horizontal displacements do not grow to auch an extent as to imply any possibility for overturning. In Section 6.8 va report results from Whole Pool Multi-Rack arsalysis which complement and reinforce the results obtained using the single rack model.
Additional investigation of important structural items is carried out and results are summarized in Table 6.7.27. The results are based on limiting loads obtained from either single or multi-rack analyses, whichever governs.
6.7.1 Imoact Analyses
- a. Imoact Load 1nc Between Fuel Assembly and Cell Wall Local cell wall integrity is conservatively estimated from peak impact loads. Plastic analysis is used to obtain the limiting impact load. Table 6.7.27 gives the limiting impact load and compares the limit with the-highest value obtained from any of the single rack analyses. The limiting load is much greater than the load obtained from any of the simulations reported in Tables 6.7.3 - 6.7.26. This limiting load is based on the cell wall. The actual impact loads, when considered as loads applied to the fuel assembly structure, are much lower than the load limits imposed by the fuel manufacturer.
- b. Imoacts Between Racks and Wall and Between Adjacent Racks No impacts are found between rack and walls. This is confirmed by the Whole Pool Multi-Rack analysis results in Section 6.8. Because of the closely spaced racks, impact protection is provided at the top corners of racks at potential impact sites.
While the nominal rack-to-rack gap is used for 6-28 j
calculation of hydrodynamic effects, the gap i elements at the corners of the rack reflect the
- actual smaller gaps that are present at the top corners and at the baseplate level. These impact i
protection hard points ensure that impacts, if they 4 occur, will be above or below the active fuel region. The results of the single rack analyses do
- not include any rack-to-rack impact hardening locations. The whole pool analysis indicates that some rack-to-rack impact during a seismic event is
. likely to occur. The design of these impact 4
protection sites is based on the highest anticipated impact force from present and future fuel loading scenarios and limits the local stress in the cell wall near the impact protection site to prevent buckling, and ensures that impact loads are appropriately distributed into the cell structure.
6.7.2 Weld Stresses Weld locations subjected to significant seismic loading are at the bottom of the rack at the baseplate-to-cell connection, at the top of the pedestal support at the baseplate connection, and at cell-to-cell connections. Results from dynamic analyses of single racks and whole pool analyses are surveyed and loading used to qualify the welds,
- a. Baseolate-to-Rack Cell Welds and Baseolate-to-Pedestal Welds Reference [6.1.3) (ASME Code Section III, Subsection NF) permits, for the MIIE event, an allowable weld stress r =
.42 Su. A comparison of this allowable value with the highest weld stress predicted is given in Table 6.7.27.
The highest predicted weld stress is less than the allowable weld stress value.
The weld between baseplate and support pedestal is checked using limit analysis techniques (6.7.1]. The structural weld at that location is considered safe if the interaction curve between net force and moment is such that G = Function (F/Fy ,M/My) < 1.0 6-29
1 y My are the limit load and moment under direct load F,
only and direct moment only. These values depend on the configuration and on material yield strengths. F and M are absolute values of actual force and moments applied to the weld section. The calculated value of G for the pedestal / baseplate weld is presented in Table 6.7.27 and is less than the limit value of 1.0. This calculated value is conservatively based on instantaneous peak loading. '
- b. Cell-to-Cell Welds cell-to-cell connections are by a series of welds along I l the cell height. Stresses in storage cell to storage cell welds develop along the length due to fuel assembly impact with the cell wall. This occurs if fuel l assemblies in adjacent cells are moving out of phase l with one another so that impact loads in two adjacent '
cells are in opposite directions; this tends to separate the two cells from each other at the weld. Table 6.7.27
- gives results for the maximum allowable load that can be transferred by these welds based on the available weld area. An upper bound on the load required to be transferred is also given in Table 6.7.27 and is less than the allowable load. This upper bound value is obtained by using the highest rack-to-fuel impact load from any rack analysis, and multiplying the result by 2 (assuming that two impact locat
- .ons are supported by every weld connection).
6.8 Results from Whole Pool Multi-Rack Analyses The Whole Pool Multi-Rack Analysis is carried out using realistic assumptions for friction coef ficients. A random set of friction coefficients in the range of 0.2-0.8 with mean value of 0.5 are used for pedestal-liner friction. Each pedestal has a constant coefficient-of-friction chosen randomly within the above guidelines. The input seismic loadings are the governing earthquake time-histories for the MHE and the DE, respectively.
Two whole pool multi-rack runs are carried out, r
t i 6-30
i
~
Tables 6.8.1 and 6.8.2 show the maximum corner absolute horizontal displacements at both the top and bottom of each rack in global X and Y directions. Tables 6.8.1 and 6.8.2 reflect analyses per the configuration in Figure 6.4.9 with global X direction corresponding to the plant " north".
Tables 6.8.3 and 6.8.4 present pot, pedestal compressive loads for all pedestals in the pool for each of the WPMR analyses (refer to Figure 6.4.9 for pedestal locations for Tables 6.8.3 and 6.8.4.
In addition to a report of maximum pedestal loads, the time-history of each pedestal load vector for each rack is archived for ll future use.
In Table 6.8.5, the maximum displacement and pedestal vertical loads obtained from the free-standing racks for MBE multi-rack simulation are compared with the limiting single rack analyses.
The absolute displacement values are higher than those obtained from any of the single rack analyses. Thus, it appears essential to perform a Whole Pool Multi-rack analysis to verify the behavior of all racks in the pool.
Rack-to-rack impacts are predicted by the whole pool multi-rack analyses; as noted previously, impact protection locations above
, the active fuel region are provided for sufficient redistribution of any rack-to-rack impact load.
- Figures 6. 8.1 - 6. 8. 4 show the time-histories of rack-to-rack and j rack-to-wall gaps on corners of rack top at typical locations (see Figure 6.4.9 for locations). A survey of all of the rack-to-rack and rack-to-wall impact elements confirms that there are no rack-to-wall impacts by any-rack in the spent fuel pool. All rack-to-rack impacts are above or below the region of active fuel.
6-31
l The Whole Pool Multi-Rack analyses confirms that no new concerns are identified; overall structural integrity conclusions are confirmed by both single and multi-rack analyses. While peak l pedestal vertical loads may be higher than predictions of the single rack analysis, a simple scaling of the stress factors shows that there are no breaches of the structural integrity limit.
6.9 Bearina Pad Analysis To protect the slab from high localized dynamic loadings, bearing pads are placed between the pedestal base and the slab. Fuel rack pedestals impact on these bearing pads during a seismic event and pedestal loading is transferred to the liner. Bearing pad dimensions are set to ensure that the average pressure on the slab surface due to a static load plus a dynamic impact load does not exceed the American Concrete Institute (6.9.1) limit on bearing pressures. Pedestal 1; cations are set generally to avoid overloading of any leak chase regions under the slab. However, to provide additional conservatism for structural integrity predictions, analyses have been performed assuming that a " worst case" pad placement, directly over a leak chase, is possible. All bearing pads are designed to provide proper load distribution in this event. Time-history results from dynamic simulations for each
, pedestal are used to generate appropriate static and dynamic pedestal loads which are then used to develop the bearing pad size.
Section 10 of [6.9.1) gives the design bearing strength as fb = p (.85 fe') G where p = .7 and fe' is the specified concrete strength for the spent fuel pool. E = 1 except when the supporting surface is wider on all sides than the loaded area. In that case, G =
(A2 /A1 ).5, but not more than 2. Al is the actual loaded area, and 6-32
) A2 is an area greater than At and is defined in [6.9.1]. Using a value of E > 1 includes credit for the confining effect of the
} surrounding concrete.
Bearing pads are sized so as to provide sufficient margin on average beariers pressure. Table 6.9.1 summarizes the limiting result.
6.10 References for Section 6
[6.1.1) "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications", dated April 14, 1978, and January 18, 1979 amendment thereto.
1
- [6.1.2) USNRC Standard Review Plan, NUREG-0800 (1981),
a
[6.1.3) ASME Boiler & Pressure Vessel Code,Section III, Subsection NF, appendices (1986).
l (6.2.1) USNRC Regulatory Guide 1.29, " Seismic Design
! Classification," Rev. 3, 1978.
[6.2.2) Soler, A.I. and Singh, K.P., " Seismic Responses of Free Standing Fuel Rack Constructions to 3-D Motions", Nuclear
- Engineering and Design, Vol. 80, pp. 315-329 (1984).
[6.2.3) Singh, K.P. and Soler, A.I., " Seismic Qualification of Free Standing Nuclear Fuel Storage Racks - the Chin Shan Experience",
Nuclear Engineering International, UK (March 1991).
[6.3.1) OPPD Fort Calhoun Document PLDBD-CS-51.
(6.3.2)
Holtec Proprietary Report - Verification and User's Manual for Computer Code GENEQ, Report HI-89364, January, 1990.
[6.3.3) Holtec Proprietary report HI-92871, " User's Manual and Validation of Program AVESPC.FOR",
1992.
6-33
i 4
4 i
t 1
[6.4.1) Rabinowicz, E., " Friction Coefficients of Water Lubricated Stainless Steels for a Spent
] Puel Rack Facility," MIT, a report for Boston i Edison Company, 1976.
1 [6.4.2) Singh, K.P. and Soler, A.I., " Dynamic
] Coupling in a Closely Spaced Two-Body System
. Vibrating in Liquid Med;.um: The Case of Fuel j Racks," 3rd International Conf ere nce on i Nuclear Power Safety, Keswick, England, May
- 1982.
- l. [6.4.3) Fritz, R.J., "The Effects of Liquids on the ,
i Dynamic. Motions of Immersed Solids," Journal of Engineering for. Industry, Trans. of the ASME, February 1972, pp 167- 172.
[6.4.4) Levy, S. and Wilkinson, J.P.D., "The j Component Element Method in Dynamics with
- Application to Earthquake and Vehicle Engineering," McGraw Hill, 1976.
[6.4.5) P3ul, B., " Fluid Coupling in Fuel Racks
. Correlation of Theory and Experiment", Holtec
! -Proprietary Report HI-88243.
l [6.6.1) " Dynamics of Structures," R.W. Clough and J.
j Penzien, McGraw Hill (1975).
I
[6.6.2) Soler, A.I.,. " User Guide for PREDYNA1 and DYNAMO", Holtec- Proprietary Report HI-89343, Rev. 2, March, 1990.
l [6.6.3) Soler, .A.I., " Theoretical Background for l Single and Multiple Rack Analysis", Holtec
! Proprietary Report ~HI-90439, Rev. O, j February, 1990.
[6.6.4) Soler, A.I., DYNARACK Theoretical. Manual",
- .Holtec Proprietary Report HI-87162, Rev. 1,
, January, 1988.
! [6.6.5) Soler, A.I., "DYNARACK Validation Manual, l
Holtec. Proprietary Report HI-91700, Rev. O, October, 1991.
[6.7.1) Singh, K.P., Soler, A.I., and Bhattacharya,
~
S., " Design Strength of Primary Structural Welds in Free Standing Structures", ASME, Journ. of Pressure Vessel Technolocy, August, 1991.
l r
6-34 f 1
, - .-. _.- __ _ _ . _ -. ._,___.._. . . _., _ . , . _ _ _ _ _ _ _ _.,._ _, ~ _ . . , _ .
[6.9.1) ACI 349-85, Code Requirements for Nuclear Saf ety Related Concrete Structures, American Concrete Institute, Detroit, Michigan, 1985.
l l
6-35
Table 6.1.1 LISTING OF PLANTS WHERE DYNARACK WAS APPLIED PLANT DOCKET NO.
Enrico Fermi Unit 2 USNRC 50-341 Quad Cities 1 and 2 USNRC 50-254, 50-265 Rancho Seco USNRC 50-312 Grand Gulf Unit 1 USNRC 50-416 Oyster Creek USNRC 50-219 Pilgrim Unit I USNRC 50-293 V.C. Summer USNRC 50-395 Diablo Canyon Units 1 and 2 USNRC 50-275, 50-323 Byron Units 1 & 2 USNRC 50-454, 50-455 Braidwood Units 1 & 2 USNRC 50-456, 50-457 Vogtle Unit 2 USNRC 50-425 St. Lucio Unit 1 USNRC 50-335 Millstone Unit 1 USNRC 50-245 D.C. Cook Unita 1 & 2 USNRC 50-315, 50-316 Indian Point Unit 2 USNRC 50-247 Three Mile Island Unit 1 USNRC 50-289 J.A. FitzPatrick USNRC 50-333 Shearon Harris Unit 2 USNRC 50-401 Kuosheng Units 1 & 2 Taiwan Power Company Chin Shan Units 1 & 2 Taiwan Power Company Ulchin Unit 2 Korea Electric Power Laguna Verde Units 1 & 2 Comision Federal de Electricidad Zion Station Units 1 & 2 USNRC 50-295, 50-304 6-35a
TABLE 6.3.1 FOUR SETS OF TIME HISTORIES GENERATED FROM MAXIMUM HYPOTHETICAL EARTHQUAKE ( MHE )
AND THEIR CROSS-CORRELATION COEFFICIENTS Time-history file name Cross Correlation Coeff.
SET NO. X-Seismic Y-Seismic Z-Seismic E-W
- N-S
- VT X-Y X-Z Y-Z 1 a-tmhe.h11 a-tmhe.h21 a-tmhe.vt1 .049658 .047539 .029510 2 a-tmho.h12 a-tmhe.h22 a-tmhe.vt2 .012533 .020900 .001380 3 a-tmhe.h13 a-tmhe.h23 a-tmho.vt3 .055372 .007897 .014012 4 a-tmhe.h14 a-tmho.h24 a-tmhe.vt4 .033636 .022693 .005253 The directions are defined for single rack analysis.
6-36 1
- ___ _ __ _ . - _ _ _ . _ . _ . . _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ __ ~ _ . _ . _ _ . _ _
l 1
I i
i f
1 TABLE 6.3.2 l FOUR SETS OF TIME HISTORIES GENERATED
, FROM DESIGN EARTHQUAKE RESPONSE SPECTRA ( DES )
j AND THEIR CROSS-CORRELATION COEFFICIENTS l
i.
Time-history file name cross correlation coeff.
, SET NO. X-Seismic Y-Seismic Z-Seismic
. E-W
- N-S
- VT X-Y X-Z Y-Z l
1 1 a-tdes.h11 a-tdes.h21 a-tdes.vtl .026136 .008124 .015733 i
i 2 a-tdes.h12 a-tdes.h22 a-tdes.vt2 .129344 .013434 .005146 1
j 3 a-tdes.h13 a-tdes.h23 a-tdes.vt3 .003801 .004435 .007472 4
- 4 a-tdes.h14 a-tdes.h24 a-tdes.vt4 .105971 .012635 .014274 The directions are defined for single rack analysis.
i A
i i
1
(- 6-37
. - . . . _ . .. -. . . - . - _. -. - - =
l
! l j
t i
1 Table 6.4.1 i
DEGREES-OF-FREEDOM j Displacement Rotation
! Location U, Uy U, 0, 07 0, (Node) 4 1 P1 P2 P3 94 45 96 l 2 P7 Pts P19 q20 9 21 9 22 Point 2 is assumed attached to rigid rack at i the top most point.
, e 4
2 p7 pa 4 3 P9 Pio 4 P11 P12 I
S' p3 p34 1 Pts Pts where:
pi = qi(t) + Ul (t) i = 1,7,9,11,13,15,17
= qi(t) + U2 (t) i = 2,8,10,12,14,16,18
= qi(t) + U3 (t) i = 3,19 Ui (t) are the 3 known earthquake displacements.
i 6-38
.~.
l j
i j i i
Table 6.4.2 1
- NUMBERING SYSTEM FOR GAP ELEMENTS AND FRICTION, ELEMENTS i
i I. Nonlinear Sorinas (Gao Elements) (64 Total) f Number Node Location Descriotion 1 Support S1 Z compression only element 2 Support S2 Z compression only element 1 3 Support S3 Z compression only element ;
j 4 Support p4 Z compression only element l 5 2,2 X rack / fuel-assembly impact 3
element i <
l 6 2,2 X rack / fuel assembly impact l element .
! 7 2,2 =Y : rack / fuel assembly impact-
! element j 8 2,2 Y rack / fuel assembly- impact-
} element '
l 9-24 Other rattling masses for nodes 1 , 3, 4 and 5 I
, ?
25 Bottom cross- Inter-rack impact elements j section of rack .
, (around edge) 1 Inter-rack impact elements-l . Inter-rack impact elements
. Inter-rack impact elements
. Inter-rack impact elements
, . -Inter-rack impact elements Inter-rack impact elements
- 44 Inter-rack impact elements
! 45 Top cross-eection Inter-rack impact elements i . -of rack Inter-rack impact elements
} . '(around-edge) Inter-rack impact-elements-L . Inter-rack impact elements
[ . ~ Inter-rack impact' elements i
.. Inter-rack impact' elements- .
. Inter-rack impact elements 64 Inter-rack 11mpact elements
- 4 9
s 6-39
~
?
d Table 6.4.2 (continued)
)
NUMBERING SYSTEM FOR GAP ELEMENTS AND FRICTION ELEMENTS i
II. Friction Elements (16 total) !
Number Node Location Descrintion 1 Support S1 X direction friction 2 Support S1 Y direction friction 3 Support S2 X direction friction
, 4 Support S2 Y direction friction 5 Support S3 X direction friction 6 Support S3 Y direction friction 7 Support S4 X direction friction 8 Support S4 Y direction friction
~
9 S1 X Slab moment i
10 S1 Y Slab moment 11 S2 X Slab moment 12 S2 Y Slab moment 13 S3 X Slab moment 14 S3 Y Slab moment 15 S4 X Slab moment 16 S4 Y Slab moment l
.q 6-40
_ . . . . . . . . _ _ . . _ . . . . . . _ . _ . . _ . . . _ . . _ . . _. . _ . _ _ _ . - _. . . . . . _~ __ _
I
- I t l
- \
, J l ',
i s
1
. Table 6.4.3 3
1
- -SPENT FUEL POOL LOADING i~ Fuel Weight-
- _ Cell Configuration ~' Rack _ _ _(lbs) i (No, of Cells in . Weight- (Channelled
- Rack NS x EW Direction)
.(lbs) Fuel Assembly; I
A1 (Region 1) 10x8'. 16000 13PO ,
i
- A2 (Region-1) 10x8 16000 1380
. B1 (Region.2) 12x9 15200- -1380 j B2 (Region.2)I 12'X9 15200 '1380' C (Region 2) 11x9- 13900 -1380 F
3 __
D (Region 2) '11x8 ~12400- =1380 E (Region.2) 10x10 -14000- 1380'
- F1 (Region;2)L 12x10- 16800 1380 i
F2 :(Region :2)
~
t- :12x10- 16800' 1380 i
G1;(Region 2)l .10x9-' 12600- ,1380
{
G2 (Regio.n: 2 ), .
10x9- 12600 1380 1
i' i~
p
'6-411 4
4 -
I. , , . , 4_.-- - . , , . _ . . , . , , . . . . . . . _ . . . , . . . . _ . _ _ , , , . . . . _ . , - . ._ ;
- , - - . - . ~ _-- . - - . . . - . . .-. . - . - - - _ . . . . . - . . . . .-
-4 i
i l ?
Table 6.5.1 RACK MATERIAL DATA - (200'F)
Young's Yield- _
Ultimate Modulus 'Strentith Strength Material E-(psi) Sy (psi) S. (psi):
304 S.S. 27.6 x=10' - 25,000 71,000 ,
Section III- Table Table ' Table Reference' I-6.0. I-2.2' .__I-3.2 <
i.
SUPPORT MATERIAL DATA (200*F)=
+
Young's. Yield' Ultimate' Modulus- -Strength.- Strength' Material E .-- (psi) Sy - (psi) - S, :(psi)
~
1 ASTM-240,: 27.6x10' l25,000 -71,000 Type 304 (upper.part-ofl support-
-feet) 8 ASTM'564-630'. 27.6x10' -1'06,300
-140,000_- '
_(lower part _
--- i
.ofJsupport feet;-
j' i-age hardened _- at'
' 1110.* F) l-l' l-
'6-42~
l .,
- .. ,. ._ . . _ . _ , , . ~ . _ . - . - _ _ _ _ . , - _ , _ . , , , . . . .. - _- - -.. , . _
d Table 6.7.1 a
RESULTS OF SINGLE RACK ANALYSES I
List of All Runs Holtec Rack Fuel Fuel Leading Seismic Coefficient Motien i '
Run I.D. I.D. I.D. Ccndition Loading of Friction Mcde 4
dralmheo.rf8 Al regular Fully Loaded MHE-SET-3 0.8 opp-phase Region-I 1380# 80 cells x1.25 1-dralmheo.rf2 Al regular Fully Loaded MHE-SET-3 0.2 opp-phase Region-I 1380# 80 cells x1.25 dralmheo.hx8 Al regular Half Loaded MHE-SET-3 0.8 opp-phase Region-I 1380# in X,4C cells x1.25 i
dralmheo.hx2 Al regular Half Loaded MHE-SET-3 0.2 opp-phase i Region-I 1380# in X,40 cells x1.25 dralmhee.hy8 Al regular Half Loaded MHE-SET-3 0.8 opp-phase Region-I 1380# in Y,40 cells x1.25
?
dralmheo.hy2 Al regular Half Loaded MHE-SET-3 0.2 opp-phase Region-I 1380# in Y,40 cells x1.25
, dralmheo.rea Al regular Symmetrically MHE-SET-3 0.8 opp-phase Region-I 1380# 8 cells x1.25 dralmhea.re2 Al- regular Symmetrically MHE-SET-3 0.2 opp-phase Region-I 1380# 8 cells x1.25 i
( to be continued )
6-43 4
r
i Table 6.7.1 ( continued )
Holtec Rack Fuel Fuel Loading Seismic Coefficient Motion Run I.D. I.D. I.D. Condition Loading of Friction Mcde drdmheo.rf8 D regular Fully Loaded MHE-SET-3 0.8 opp-phase
- Region-II 1380# 96-cells x1.25 drdmheo.rf2 D regular Fully Loaded MHE-SET-3 0.2 opp-phase a Region-II 1380# 96 cells x1.25
, drdmheo.hx8 D regular Half Loaded MHE-SET-3 0.8 opp-phase Region-II 1380# in X,48 cells x1.25 4
drdmhee.hx2 D regular- Half Loaded MHE-SET-3 0.2- opp-phase Region-II 1380# in X,48 cells- x1.25 drdmheo.hy8 D regular Half Loaded MHE-SET-3 0.8 opp-phase Region-II 1380# in Y,48 cells x1.25 l drdmheo.hy2 D regular Half Loaded MHE-SET-3 0.2 opp-phase Region-II 1380#. in Y,48 cells x1.25 drdmheo.re8 D- regular Symmetrically MHE-SET-3 -0.8 opp-phase Region-II 1380# 8 cells x1.25 drdmhec.re2- D regular ~ Symmetrically MHE-SET-3 0.2 opp-phase Region-II 1380# 8 cells x1.25-4 a
a'
( to be continued )
4
~
) 6-44 1
4 1
a 4
Table 6.7.1 ( continued )
} Holtec Rack Fuel Fuel Loading Seismic Coefficient Motion Run I.D. I.D. I.D. Condition Loading of Friction Mcde d'efimheo.rf8 F1 regular Fully Loaded MHE-SET-3 0.8 opp-phase Region-II 1380# 120 cells x1.25 drfimheo.rf2 F1 regular Fully Loaded MHE-SET-3 0.2 opp-phase Region-II 1380# 120 cells x1.25 drfimheo.hx8 F1 regular Half Loaded' MHE-SET-3 0.8 opp-phase Region-II 1380# in X,60 cells x1.25 i
i drfimheo.hx2 _ F1 regular Half Loaded MHE-SET-3 0.2 opp-phase Region-II 1380# in X,60 cells x1.25 drfimheo.hy8 F1 regular' Half Loaded -- MHE-SET-3 0,8 opp-phase
, Region-II 1380# in Y,60 cells x1.25 drfimheo.hy2 F1 regular Half Loaded- MHE-SET-3 0.2 opp-phase Region-II 1380# in Y,60 cells :x1.25 drfimheo.re8 F1 regular Symmetrically MHE-SET-3 0 ~. 8 ' opp-phase Region-II 1380#. '12 cells x1.25 drfimheo.re2 _ F1
_ ' regular Symmetrically MHE-SET-3 -0.2 opp-phase Regl'on-II 1380# 12. cells x1.25 t
6 4
6-45 4
-.m.-
- i-1
?
i i Table 6.7.2 j
SUMMARY
OF WORST RESULTS l FROM 24 RUNS OF SINGLE RACK ANALYSIS ,
l
( LOADED WITH 1380# REGULAR FUEL ASSEMBLIES; MAXIMUM HYPOTHETICAL-EARTHQUAKE TIME-HISTORY SET-3 x1.25-)
i l- Item Value Run I.D.
l 1. Maximum total vertical pedestal load: 426,890 lbs. drfimheo.rf2 j 2. Maximum vertical load.
[ in any single pedestal: 124,680 lbs. .drfimheo.rf8 i- 3. Maximum shear load in any single pedestal:- 48,758.lbs. dralmheo.rf8-
- 4. Maximum fuel assembly-to-cell wall i impact load at one local position: 774 lbs. drfimheo.re2 i
1 5. Maximum rack-to-wall-impact load at baseplate level: O lbs.
l
- 6. Maximum rack-to-wall j impact load at the top of rack: ' O lbs. -
- 7. Maximum rack-to-rack j
impact load at baseplate level: O lbs.
i
{ 8. Maximum rack-to-rack .
j- impact load at the top-of rack: 0 lbs.-
I.
I
- 9. Maximum corner displacements Top corner in x. direction:' O.1566; fin. drdmheo.hy8 in y direction: 0.24811 in, dralmheo.rf8~
Baseplate corner-in1x direction:- 0.0216 in. drdmheo.re2 ,
in y direction: -0.0490 in, dralmheo.rf2
- 10. Maximum stress factors Above_ baseplate:- '053211(R6) dralmheo.rf8 Support pedestals:' O.192J(R6) Ld ralmheo.rf8, 6-46
. . . _ . ~ . _.~... _ ,_ _
Table 6.7.3
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: A 8x10 Holtec Run I.D.: dralmneo.rf8 Seismic Loading: MHESET3x1.25 Fuel Assembly I.D. and Weight: 1380# reg. ; 1380.0 (lbs.)
Fuel Loading: 80 cells loaded; Fuel centroid X,Y: .0, .0 (in.)
Coefficient of friction at the bottom of support pedestal: 0.8
$ Revision: 3.46 S SLogfile: C:/ racks /dynam0/ dynamo.fov S
, SRevision: 2.5 S
$Logfile: C:/ racks /dynam0/dynasi.fov S SRevision: 3.36 S SLogfile: C:/ racks /dynam0/dynas2.fov $
DYNAMIC IMPACT LOADS (lbs.)
(1) Maximum total vertical pedestal load: 274474.0 (2) Maximum vertical load in any single pedestal: 120768.1 (3) Maximum shear load in any single pedestal: 48758.3 (4) Maximum fuel-cell impact at one local position: 507.4 (5) Maximum rack-to-wall impact at baseplate: .0 (6) Maximum rack-to-wall impact at rack top: .0
, (7) Maximum rack-to-rack impact at baseplate: .0 l (8) Maximum rack-to-rack impact at rack top: .0 MAXIMUM CORNER DISPLACEMENTS (in.)
l Location: X-direction Y-direction l Top corner: .1113 .2481 Baseplate corner: .0031 .0075 MAXIMUM STRESS FACTORS
- Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .072 .025 .211 .090 .281 .321 .042 Support pedestal: .168 .055 .155 .084 .264 .282 .101
- See Section 6.5.2.3 of the Licensing Report for definitions.
6-47
a I
Table 6.7.4
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: A 8x10 Holtec Run I.D.: dralmheo.rf2 Seismic Loading: MHESET3x1.25 1380# reg. ; 1380.0 (lbs.)
Fuel Assembly I.D. and Weight:
Fuel Loading: 80 cells loaded; r uel centroid X,Y: .0, .0 (in.)
Coefficient of friction at the bottom of support pedestal: 0.2 SRevision: 3.46 S
$Logfile: C:/ racks / dynamo /dynano.fov 5
$ Revision: 2.5 S SLogfile: C:/ racks /dynam0/dynasi.fov $
$ Revision: 3.36 $
$Logfile: C:/ racks /dynam0/dynas2.fov S DYNAMIC IMPACT LOADS (lbs.)
(1) Maximum total vertical pedestal load: 252048.5 (2) Maximum vertical load in any single pedestal: 105852.8 (3) Maximum shear load in any single pedestal: 20726.6
) (4) Maximum fuel-cell impact at one local position: 515.8 (5) Maximum rack-to-wall impact at baseplate: .0 (6) Maximum rack-to-wall impact at rack top: .0 (7) Maximum rack-to-rack impact at baseplate: .0 (8) Maximum rack-to-rack impact at rack top: .0 MAXIMUM CORNER DISPLACEMENTS (in.)
Location: X-direction Y-direction Top corner: .1114 .2124 Baseplate corner: .0132 .0490 HAXIMUM STRESS FACTORS
- Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .066 .024 .177 .087 .226 .259 .032 Support pedestal: .147 .030 .065 .046 .200 .210 .043
- See-Section 6.5.2.3-of the Licensing Report for definitions.
t 6-48
~.
Table 6.7.5
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: A 8x10 Holtec Run I.D.: dralmheo.hx8 Seismic Loading: MHESET3x1.25 Fuel Assembly I.D. and Weight: 1380/ reg. ; 1380.0 (lbs.)
Fuel Loading: 40 cells loaded; Fuel centroid X,Y: 19.6, .0 (in.)
Coefficient of friction at the bottom of support pedestal: 0.8 4
SRevision: 3.46 S
$Logfile: C:/ racks /dynam0/dynam0.fov $
SRevision: 2.5 $
SLogfile: C:/ racks /dynam0/dynasi.fov $
$ Revision: 3.36 $
$Logfile: C:/ racks /dynam0/dynas2.fov $
DYNAMIC IMPACT LOADS (lbs.)
(1) Maximum total vertical pedestal load: 117257.1 (2) Maximum vertical load in any single pedestal: 75611.3 (3) Maximum shear load in any single pedestal: 20075.0 4
(4) Maximum fuel-cell impact at one local position: 275.4 (5) Maximum rack-to-wall impact at baseplate: .0 (6) Maximum rack-to-wall impact at rack top: .0 (7) Maximum rack-to-rack impact-at baseplate: .0 (8) Maximum rack-to-rack impact at rack top: .0 MAXIMUM CORNER-DISPLACEMENTS (in.)
Location: X-direction Y-direction Top corner: .1085 .1160 Baseplate corner: .0024 .0042 MAXIMUM STRESS FACTORS
- Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .029 .011 .086 .036 .109 .126 .022 Support pedestal: .104 .028 .064 .042 .137 .144 .042
- See Section 6.5.2.3 of the Licensing Report for definitions.
6-49
Table 6.7.6
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: A 8x10 HClr.cc Run I.D.: dralmheo.hx2 Seismic Loading: MHESET3x1.25 Iael assembly I.D. and Weight: 1380# reg. ; 1380.0 (lbs.)
Fuel Loading: 40 cells loaded; Fuel centroid X,Y: 19.6, .0 (in.)
Coefficient of friction at the bottom of support pedestal: 0.2
$ Revision: 3.46 S
$Logfile: C:/ racks /dynam0/ dynamo.fov S SRevision: 2.5 S
$Logfile: C:/ racks /dynam0/dynasi.fov $
SRevision: 3.36 $
SLogfile: C:/ racks /dynam0/dynas2.fov S DYNAMIC IMPACT LOADS (lbs.)
(1) Maximum total vertical pedestal load: 111515.0 (2) Maximum vertical load in any single pedestal: 61763.7 (3) Maximum shear load in any single pedestal: 12320.4 (4) Maximum fuel-cell impact at one local position: 356.4 (5) Maximum rack-to-wall impact at baseplate: .0 (6) Maximum rack-to-wall impact at rack top: .0 (7) Maximum rack-to-rack impact at baseplate: .0 1
(8) Maximum rack-to-rack impact at rack top: .0 MAXIMUM CORNER DISPLACEMENTS (in.)
Location: X-direction Y-direction Top corner: .0899 .0905 Baseplate corner: .0118 .0149 MAXIMUM STRESS FACIORS
- Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .025 .011 .087 .036 .105 .121 .015 Support pedestal: .086' .022 .038 .034 .119 .125 .025
- See Section 6.5.2.3 of the Licensing Report for definitions.
6-50
l l
1 Table 6.7.7
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: A 8x10 Holtec Run I.D.: dralmheo.hy8 Selsmic Loading: MHESET3x1.25 Fuel Assembly I.D. and Weight: 1380# reg. ; 1380.0 (1bs.)
Fuel Loading: 40 cells loaded; Fuel centroid X,Y: .0, 25.9 (in.)
Coefficient of friction at the bottom of support pedestal: 0.8 SRevision: 3.46 S
$Logfile: C:/ racks /dynam0/dynam0.fov $
$ Revision: 2.5 $
$Logfile: C:/ racks /dynam0/dynasi.fov $
SRevision: 3.36 $
- SLogfile
- C:/ racks /dynam0/dynas2.fov $
DYNAMIC IMPACT LOADS (lbs.)
(1) Maximum total vertical pedestal load: 118557.6 4
- (2) Maximum vertical load in any single pedestal: 59995.7 (3) Maximum shear load in any single pedestal: 29111.2 (4) Maximum fuel-cell impact at one local position: 247.0 (5) Maximum rack-to-wall impact at baseplate: .0 (6) Maximum rack-to-wall impact at rack top: .0 (7) Maximum rack-to-rack impact at baseplate: .0 $
(8) Maximum rack-to-rack impact at rack top:
, .0 i
MAXDIUM CORNER DISPLACEMENTS (in.)
Location: X-direction Y-direction Top corner: .0791 .1702 Baseplate corner: .0021 .0050 MAXIMUM STRESS FACTORS
- Stress factor: R1 - R2 R3 R4 R5 R6 R7 i
Above baseplate: .029 .013 .106 .053 .113 .129 .039 Support pedestal: .083 .032 .091 .048 .150 .162 .060
- See Section 6.5.2.3 of the Licensing Report for definitions.
6-51
Table 6.7.8
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: A 8x10 )
Holtec Run I.D.: dralmuco.hy2 Seismic Loading: MHESET3x1.25 Fuel Assembly I.D. and Weight: 1380# reg. ; 1380.0 (lbs.)
- Fuel Leading
- 40 cells loaded; Fuel centroid X,Y: .0, 25.9 (in.)
Coefficient of friction at the bottom of support pedestal: 0.2
, SRevision: 3.46 $
$Logfile: C:/ racks /dynam0/ dyna =0.fov S
$ Revision: 2.5 S
$Logfile: C:/ racks /dynam0/dynasi.fov $
$ Revision: 3.36 S SLogfile: C:/ racks /dynam0/dynas2.fov $
DYNAMIC IMPACT LOADS (lbs.)
(1) Maximum total vertical pedestal load: 120925.1 (2) Maximum vertical load in any single pedestal: 64454.3 (3) Maximum shear load in any single pedestal: 12781.7 (4) Maximum fuel-cell impact at one local position: 272.5 (5) Maximum rack-to-wall impact at baseplate: .0 (6) Maximum rack-to-wall impact at rack top: .0 (7) Maximum rack-to-rack impact at baseplate: .0 (8) Maximum rack-to-rack impact at rack top: .0 MAXIMUM CORNER DISPLACEMENTS (in.)
Location: X-direction Y-direction Top corner: .0859 .1047 Baseplate corner: .0149 .0784 MAXIMUM STRESS FACTORS
- Stress facter: -R1 R2 R3 R4- R5 R6 R7 Above baseplate: .030 .013 .089 .049 .111 .126 .016 Support pedestal: .089 .018 .040 .028 .123 .129 .026
- See Section 6.5.2.3 of the Licensing Report for definitions.
6-52
l Table 6.7.9
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: A 8x10 Holtec Run I.D.: dralmheo.re8 Selsmic Loading: MHESET3x1.25 Fuel Assembly I.D. and Weight: 1380# reg. ; 1380.0 (lbs.)
Fuel Loading: 8 cells loaded; Fuel centroid X,Y: .0, .0 (in.)
Coefficient of friction at the bottom of support pedestal: 0.8
~ '~
SRevision: 3.46 S SLogfile: C:/ racks /dynam0/dynam0.fov S SRevision: 2.5 S
$Logfile: C:/ racks /dynam0/dynasi.fov $
, $ Revision: 3.36 S
$Logfile: C:/ racks /dynam0/dynas2.fov S DYNAMIC IMPACT LOADS (lbs.)
(1) Maximum total vertical pedestal load: 34461.3 (2) Maximum vertical load in any single pedestal: 25094.5 (3) Maximum shear load in any single pedestal: 10045.7 (4) Maxim 1m fuel-cell impact at one local position: 425.1 (5) Maximum rack-to-wall impact at baseplate: .0 (6) Maximum rack-to-wall impact at rack top: .0
- (7) Maximum rack-to-rack impact at baseplate
- .0 (8) Maximum rack-to-rack impact at rack top: .0 MAXIMUM CORNER DISPLACEMENTS (in.)
Location: X-direction Y-direction Top corner: .0408 .0520 Baseplate corner: .0015 .0020 MAXIMUM STRESS FACTORS
- Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .010 .006 .035 .030 .049 .056 .011 Support pedestal: .035 .009 .029 .014 .059 .063 .019 i
- See Section 6.5.2.3 of the Licensing Report for definitions.
6-53
Table 6.7.10
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: A 8x10
. Holtec Run I.D.: dralmheo.re2 Seismic Loading: MHESET3x1.25 Fuel Assembly I.D. and Weight: 1380/ reg. ; 1380.0 (lbs.)
Fuel Loading: 8 cells loaded; Fuel centroid X,Y: .0, .0 (in.)
Coefficient of friction at the bottom of support pedestal: 0.2 SRevision: 3.46 S SLogfile: C:/ racks / dynamo / dynamo.fov $
$ Revision: 2.5 S i SLogfile: C:/ racks /dynam0/dynac1.fov $
$ Revision: 3.36 $
i $Logfile: C:/ racks /dynam0/dynas2.fov $
i DYNAMIC IMPACT LOADS (lbs.)
(1) Maximum total vertical pedestal load: 34136.5 (2) Maximum vertical load in any single pedestal: 18299.2 (3) Maximum shear load in any single pedestal: 3656.1 (4) Maximum fuel-cell impact at one local position: 415.5 (5) Maximum rack-to-wall impact at baseplate: 0 (6) Maximum rack-to-wall impact at rack top: .0 (7) Maximum rack-to-rack impact at baseplate: .0 ,
(8) Maximum rack-to-rack impact at rack top: .0 MAXIMUM CORNER DISPLACEMENTS (in.)
-Location: X-direction Y-direction-Top corner: .0327 .0547 Baseplate corner: .0153 .0376 MAXIMUM STRESS FACTORS
- Stress factor: R1 R2 R3 R4 R5 R6 R7
, Above. baseplate: .010 .004 .034 .029 .042 .048 .005 Support pedestal: .025 .006 .012 .010 .035 .037 .008
- See Section 6.5.2.3 of the Licensing Report for definitions.
a 6-54
l
?
l i
Table 6.7.11
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: D 8x11 Holtec Run I.D.: drdmheo.rf8 Seismic Loading: MHESET3x1.25 Fuel Assembly I.D. and Weight: 1380 preg. ; 1380.0 (lbs.)
Fuel Loading: 88 cells loaded; Fuel centroid X,Y: .0, .0 (in.)
Coefficient of friction at the bottom of support pedestal: 0.8 SRevision: 3.46 S SLogfile:- C:/ racks /dynam0/ dynamo.fov S
$ Revision: 2.5 S
, SLogfile: C:/ racks /dynam0/dynasi.fov S
$ Revision: 3.36 S SLogfile: C:/ racks /dynam0/dynas2.fov $
~ ~ '
, DYNAMIC IMPACT LOADS (lbs.)
(1) Maximum total vertical pedestal load: 280067.9 (2) Maximum vertical load in any single pedestal: 107038.7 (3) Maximum shear load in any single pedestal: 46956.7 (4) Maximum fuel-cell impact at one local position: 416.4 (5) Maximum rack-to-wall impact at baseplate: .0 (6) Maximum rack-to-wall impact at rack top: . -0 4
(7) Maximum rack-to-rack impact at baseplate: .0 (8) Maximum rack-to-rack-impact at rack top: .0 MAXIMUM CORNER DISPLACEMENTS (in.)
! Location: X-direction Y-direction Top corner: .1024 .1404 j Baseplate corner: .0020 .0048 MAXIMUM STRESS FACTORS
- Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .092 .022 .147 .079 .241 .270 .054 Support pedestal: .147 .059 .147 .091 .241 .263 .097
- See Section 6.5.2.3 of the Licensing Report for definitions.
6 , . _ J
I i
Table 6.7.12
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: D 8x11 3
Holtec Run I.D.: drdmneo.rf2 Seismic Leading: MHESET3x1.25 Fuel Assembly I.D. and Weight: 1380# reg. ; 1380.0 (lbs.)
i Fuel Leading: 88 cells loaded; Fuel centroid X,Y: .0, .0 (in.)
4 Coefficient of friction at the bottom of support pedestal: 0.2
^
$ Revision: 3.46 S SLogfile: C:/ racks /dynam0/dynam0.fov $
$ Revision: 2.5 $
SLogfile: C:/ racks /dynam0/dynasi.fov $
$ Revision: 3.36 $
$Logfile: C:/ racks /dynam0/dynas2.fov $
DYNAMIC IMPACT LCADS (lbs.)
(1) . Maximum total vertical pedestal load: 280094.7' (2) Maximum vertical-load in any single pedestal: 106301.9 (3) Maximum shear load in any single pedestal: 16960.5 (4) Maximum fuel-cell impact at one local position: 482.9 (5) Maximum rack-to-wall impact at baseplate: .0 (6) Maximum rack-to-wall impact at rack top: .0 (7) Maximum rack-to-rack impact at baseplate: .0 (8) Maximum rack-to-rack impact at rack top: .0 MAXIMUM CORNER DISPLACEMENTS (in.)
! Location: X-direction Y-direction
( fop corner: .0986 .1342 Baseplate corner: .0083 .0164
, MAXIMUM STRESS FACTORS
- Stress factor: R1 R2 R3 - R4 R5 R6 R7
[ Above baseplate: .092 .035 .148 .074 .237 .266 .036
- Support pedestal
- .146 .031 .054 .047 .167 .175 .035
- See Section 6.5.2.3 of the Licensing Report for definitions.
l 1
l l
6-56 !
l
i l
l Table 6.7.13
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: D 8x11 Holtec Run I.D.: drdmheo.hx8 Seismic Loading: MHESET3x1.25 Fuel Assembly I.D. and Weight: 1380# reg. ; 1380.0 (lbs.)
Fuel Loading: 44 cells loaded; Fuel centroid X,Y: 17.4, .0 (in.)
Coefficient of friction at the bottom of support pedestal: 0.8
$ Revision: 3.46 $
SLogfile: C:/ racks /dynam0/ dynamo.fov $
SRevision: 2.5 $
SLogfile: C:/ racks /dynam0/dynasi.fov $
SRevision: 3.36 S SLogfile: C:/ racks /dynam0/dynas2.fov $
DYNAMIC IMPACT LOADS (lbs.)
(1) Maxiraum total vertical pedestal load: 107318.8 (2) Maximum vertical load in any single pedestal: 59936.5 (3) Maximum shear load in any single pedestal: 16075.1 (4) Maximum fuel-cell impact at one local position: 302.2 (5) Maximum rack-to-wall impact at baseplate: .0 d
(6) Maximum rack-to-wall impact at rack top: .0 (7) Maximum rack-to-rack impact at baseplate: .0 (8) Maximum rack-to-rack impact at rack top: .0 MAXIMUM CORNER DISPLACEMENTS (in.)
Location: X-direction Y-direction Top corner: .1006 .1188 Baseplate corner: .0019 . .0035 i
MAXIMUM STRESS FACTORS
- Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .025 .011 .104 .038 .126 .144 .026 Support pedestal: .083 .012 .051 .018 .107 .111 .033
- See Section 6.5.2.3 of the Licensing Report for definitions.
6-57
, _ ._. _ . . . . - _ _ _ , . . ~ _ . - _ - -
3
, Table-6.7.14
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: D 8x11 L Holtec Run I.D.: _drdmheo.hx2 Seismic Leading: MHESET3x1.25 l Fuel Assembly I.D. and Weight: 1380# reg. ; 1380.0 (lbs.)
8-j Fuel Loading: 44 cells loaded; Fuel centroid X,Y: 17.4, .0 (in.)
Coefficient of friction atJthe bottom of support pedestal: 0.2 l SRevision: 3.46 $
j SLogfile: C:/ racks /dynam0/ dynamo.fov $
j $ Revision: 2.5 S j $Logfile: C:/ racks /dynam0/dynasi.fov $
$ Revision: 3.36 S i $Logfile: C:/ racks /dynam0/dynas2.fov S
) DYNAMIC IMPACT LOADS (lbs.)
a (1) Maximum total vertical pedestal load: 109003.4 i
3 (2) Maximum vertical load in any single pedestal: 59496,1-1
- j. (3) Maximum shear lead in any-single pedestal:' 11870.4
~
- (4) Maximum fuel-cell impact _at one local position
- 491.0 1
l- (5) Maximum rack-to-wall impact at-baseplate: .0 i
(6) Maximum rack-to-wall impact at rack top: _ : .~ 0 :
j .( 7 ) Maximum rack-to-rack impact at-baseplate:- .0
?
(8) Maximum rack-to-rack impact-at-rack top: .0-
- MAXIMUM CORNER DISPLACEMENTS (in.)
Location: X-directioni Y-direction-l' -
[ Top' corner:- .1012 '. 1170
[ Baseplate corner: .0136 .0101 MAXIMUM STRESS-FACTORS *-
Stress factor:- _-- R1 'R2- R35 _R4~ 'R5 R6 R7' Above' baseplate: .026 .011; -.099 .038- .124 .141. .017
. -Support pedestal: .082 .013 .036 .020- .115 .120 .024-1
- ~See Section 6.5.2.3 of the Licensing-Report:for. definitions.
58
Table 6.7.15
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: D 8x11 Holtec Run I.D.: drdmheo.hy8 Seismic Loading: MHESET3x1.25 Fuel Assembly I.D. and Weight: 1380# reg. ; 1380.0 (lbs.)
Fuel Leading: 44 cells loaded; Fuel centroid X,Y: .0, 23.9 (in.)
Coefficient of friction at the bottom of support pedestal: 0.3
$ Revision: 3.46 S SLogfile: C:/ racks /dynam0/dynam0.fov $
$ Revision: 2.5 S SLogfile: C:/ racks /dynam0/dynasi.fov $
- $ Revision
- 3.36 S
$Logfile: C:/ racks /dynam0/dynas2.fov $
DYNAMIC IMPACT LOADS (lbs.)
(1) Maximum total vertical pedestal load: 132496.4 (2) Maximum vertical load in any single pedestal: 61829.6 (3) Maximum shear lert in any single pedestal:- 34668.1 (4) Maximum fuel-cell impact at one local position: 245.9 (5) Maximum rack-to-wall impact at baseplate: .0 (6) Maximum rack-to-wall impact at rack top: .0 (7) Maximum rack-to-rack impact at baseplate: .0 l
(8) Maximum rack-to-rack iapact at rack top: .0 MAXIMUM CORNER DISPLACEMENTS (in.)
Location: X-direction Y-direction Top corner: .1566 .1838 Baseplate corner: .0043 .0045 MAXIMUM STRESS FACTORS
- Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .041 .015 .091 .095 .133 .150 .035 Support pedestal: .084 .031 .107 .048 .167 .182 .070
- See Section 6.5.2.3 of the Licensing Report for definitions.
6-59.
i l
i i
i- ,
Table 6.7.16 i-
SUMMARY
RESULTS OF'3-D SINGLE RACK ANALYSIS FOR RACK MODULE: D 8x11-Holtec Run I.D.: _drdmheo.hy2 Seismic Loading: MHESET3x1.25 Fuel Assembly I.D. and Weight: 1380# reg. ; 1380.0'(lbs.)
{
Fuel Loading: 44 cells loaded; Fuel centroid X,Y: .0, 23.9 (in.) .
Coefficient of friction at-the bottom of support pedestal: 0.2 f
4 SRevision: 3.46 S ;
l $Logfile: C:/ racks /dynam0/ dynamo.fov $
i $ Revision: 2.5 $
SLogfile: C:/ racks /dynam0/dynasi.fov $
- $ Revision: 3.36 $-
l $Logfile: -C:/ racks /dynam0/dynas2.fov- S i
l=
1 DYNAMIC IMPACT LOADS (lbs.).
1 j' (1) Maximum total vertical pedestal load:_ 130847.3 1
l (2) Maximum vertical load in any single pedestal: 63304.3 (3) Maximum shear load in any single pedestal:. f12648.9 l
?
(4) Maximum fuel-cell impact:at one local position:- 367.2-l (5) Maximum rack-to-wall impact at baseplate: .0-1 (6) Maximum rack-to-wall impact at rack top: .0
[ (7) Maximum rack-to-rack impact at baseplate: .0 :
(8)-Maximum rack-to-rack impact at rack top:
_ _.0 b MAXIMUM-CORNER DISPLACEMENTS-(in.)
Location: X-direction: -Y-direction =
A Top _ corner: .1093 .1681
- Baseplate corner
- .0072- .0115' MAXIMUM STRESS-FACTORS *-
Stress factor: .R1 R2 'R3 1R4- RS- R6 :R7-Above-baseplate: :.039 . 015.- .076 '071-
. .126 . 142 .019 i 'Supporttpedestal: .088: .018 .040 .028 .122- .128: .026 d
- See Section 6.5.2.3 of the Licensing Report for definitions.-
\
io ,
b 6-60
_ - -- . . _ _ _ _ _ _ __ - . . _ _ , _ _ - _ . ~ ~ _ . _ _
i Table 6.7.17
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: -D 8x11 l Holtec Run I.D.: drdmheo.re8 Seismic-Loading: MHESET3x1.25 l
j Fuel Assembly I.D. and Weight: 1380# reg. ;
1380.0 (lbs.)
l- Fuel Leading: 8 cells loaded; Fuel centroid X,Y: .0, .0 (in.)
5 Coefficient of friction at the bottom of support pedestal: 0.8 1
{ - SRevision: 3.46 S i $Logfile: C:/ racks /dynam0/ dynamo.fov S-i $ Revision: 2.5 $
SLogfile: C:/ racks /dynam0/dynasi.fov $
- $ Revision
- 3.36 $
i $Logfile: C:/ racks / dynamo /dynas2.fov $
l l DYNAMIC IMPACT LOADS (lbs.)
} (1) Maximum total vertical pedestal load: 29200.9 (2) Maximum vertical load in any single pedestal: 16589.6 i
- - (3) Maximum shear load in any single pedestal: 6596.1 4
j 4
(4) Maximum fuel-cell impact at one local position: 478.9 (5) Maximum rack-to-wall impact at baseplate: .0 O'
j (6) Maximum rack-to-wall impact at rack top: .0 5
{.
(7) Maximum rack-to-rack impact at--baseplate: .0 s
(8) Maximum rack-to-rack-impact at rack top: .0 1
- MAXIMUM CORNER DISPLACEMENTS (in.)
- j. Location: X-direction Y-direction 1
j Top corner: .0429 .0500--
l Baseplate corner: .0006 .0013 4
] MAXIMUM-STRESS FACTORS-:*
4 Stress factor: .R1 R2 R3 R4 R5 R6- R7
.Above: baseplate: .009- .003 .034 .013 .'044 .051- .010
- . Support pedestal
- .023 .005 .021. .007 ,. 03 7- .040 .014 i
- See Section 6.5.2.3 of the Licensing Report-for definitions.-
611 a
. . , - - . c..,.- .- _ -
, _, .m,. - ,, ,- . , _ , , , . - --
Table 6.7.18
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: D 8x11 Holtec Run I.D.: drdmheo.re2 Seismic Leading: MHESET3x1.25 Fuel Assembly I.D. and Weight: 1380# reg. ; 1380.0 (lbs.)
Fuel Loading: 8 cells loaded; Fuel centroid X,Y: .0, .0 (in.)
Coefficient of friction .c the bottom of support pedestal: 0.2
$ Revision: 3.46 $
$Logfile: C:/ racks /dynam0/dynam0.fov S
$ Revision: 2.5 $
$Logfile: C:/ racks /dynam0/dynasi.fov $
SRevision: 3.36 $
SLogfile: C:/ racks /dynam0/dynas2.fov $
DYNAMIC IMPACT LOADS (lbs.)
(1) Maximum total vertical pedestal load: 29148.7 (2) Maximum vertical load in any single pedestal: 15375.8 j (3) Maximum shear load in any_ single pedestal: 3055.1 (4) Maximum fuel-cell impact at one local position: 685.7 (5) Maximum rack-to-wall impact at baseplate: .0 (6) Maximum rack-to-wall impact at rack top: .0 (7) Maximum rack-to-rack impact at baseplate: .0 1
(8) Maximum rack-to-rack impact at rack top: .0 MAXIMUM CORNER DISPLACEMENTS (in.)
Location: X-direction Y-direction Top corner: .0393 .0485 Baseplate corner: .0216 .0209 MAXIMUM STRESS FACTORS
- Stress factor: R1 R2 R3 R4 R5- R6 R7 Above baseplate: .009 .004 .034 .012 .045 .051- .004 i_ Support pedestal: .021 .005 .010 .008 .029 .031 .006
- See Section 6.5.2.3 of the Licensing Report for definitions.
6-62
Table 6.7.19
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: F1 120 Holtec Run I.D.: drfimheo.rf3 Seismic Loading: MHESET3x1.25 Fuel Assembly I.D. and Waight: 1380/ reg. ; 1380.0 (1bs.)
Fuel Leading: 120 cells loaded; Fuel centroid X,Y: .0, .0 (in.)
Coefficient of friction at the bottom of support pedestal: 0.8 SRevision: 3.46 S SLogfile: C:/ racks /dynam0/ dynamo.fov $
$ Revision: 2.5 S
$Logfile: C:/ racks /dynam0/dynasi.fov S
$ Revision: 3.36 $
SLogfile: C:/ racks /dynam0/dynas2.fov $
DYNAMIC IMPACT LOADS (lbs.)
(1) Maximum total vertical-pedestal load: 426324.0 (2) Maximum vertical load in any single pedestal: 124680.1 (3) Maximum shear load in any single pedestal: 38629.1 (4) Maximum fuel-cell impact at one local position: 306.8 (5) Maximum rack-to-vall impact at baseplate: .0 (6) Maximum rack-to-wall impact at rack top: .0 (7) Maximum rack-to-rack impact at baseplate: .0 (8) Maximum rack-to-rack impact at rack top: .0 MAXIMUM CORNER DISPLACEMENTS (in.)
Location: X-direction Y-direction Top corner: .0739 .0786 Baseplate corner: .0019 .0036 MAXIMUM STRESS FACTORS
- Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .110 .055 .088 .050 .183 .197- .051 Support pedestal: .173 .058 .121 .088 .247 .262 .079
- See Section 6.5.2.3 of the Licensing Report for definitions.
6-63
i Table 6.7.20
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: F1 120 Holtec Run I.D.: drf1mheo.rf2 Seismic Loading: MHESET3x1.25 Fuel Assembly I.D. and Weight: 1380# reg. ; 1380.0 (lbs.)
4 Fuel Loading: 120 cells loaded; Fuel centroid X,Y: .0, .0 (in.)
Coefficient of friction at the bettem of support pedestal: 0.2 SRevision: 3.46 S 1 SLogfile: C:/ racks /dynam0/ dynamo.fov $
SRevision: 2.5 S SLogfile: C:/ racks / dynamo /dynasi.fov $
$ Revision: 3.36 $
- SLogfile
- C:/ racks /dynam0/dynas2.fov $
DYNAMIC IMPACT LOADS (lbs.)
(1) Maximum total vertical pedestal load: 426890.9 (2) Maximum vertical load in any single pedestal: 124S38.0 (3) Maximum shear load in any single pedestal: 21086.7 (4) Maximum fuel-cell impact at one local position: 315.2 (5) Maximum rack-to-wall impact at baseplate: .0 (6) Maximum rack-to-wall impact at rack top: .0 (7) Maximum rack-to-rack impact at baseplate: .0 (8) Maximum rack-to-rack impact at rack top: .0 MAXIMUM CORNER DISPLACEMENTS (in.)
l Location: X-direction Y-direction Top corner: .0716 .0778 Baseplate corner: .0041 .0164 MAXIMUM STRESS FACTORS
- Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .111 .041 .086 .050 .184 .198 .046 Support pedestal: .173 .041 .066 .062 .223 .232 .043
- See Section 6.5.2.3 of the Licensing Report for definitions.
6-64
Table 6.7.21
SUMMARY
RESULTS OF 3-D SINGLE RACX ANALYSIS FOR RACX MODULE: F1 120 Holtoc Run I.D.: drf1mheo.hx8 Seismic Loading: MHESET3x1.25 Fuel Assembly I.D. and Weight: 1380# reg. ; 1380.0 (lbs.)
Fuel Loading: 60 cells loaded; Fuel centroid X,Y: 21.7, .0 (in.)
Coefficient of friction at the bottom of support pedestal: 0.8 SRevision: 3.46 S
$Logfile: C:/ racks /dynam0/dynam0.fov S
$ Revision: 2.5 $
SLogfile: C:/ racks /dynam0/dynasi.fov $
SRevision: 3.36 $
SLogfile: C:/ racks /dynam0/dynas2.fov $
DYNAMIC IMPACT LOADS (lbs.)
d (1) Maximum total vertical pedestal load: 197006.5 (2) Maximum vertical load in any single pedestal: 70927.8 (3) Maximum shear load in any single pedestal: 15943.7 (4) Maximum fuel-cell impact at one local position: 600.7 (5) Maximum rack-to-wall impact at baseplate: .0 (6) Maximum rack-to-wall impact at rack top: .0-(7) Maximum rack-to-rack impact at baseplate: .0 (8)-Maximum rack-to-rack impact at-rack top: .0 MAXIMUM CORNER DISPLACEMENTS (in.)
-Location: X-direction Y-direction Top corner: .0644 .0620 j Baseplate corner: .0009 .0019 S
MAXIMUM STRESS FACTORS
- Stress factor: R1 R2 R3 R4 R5 R6 R7 Above_ baseplate: .048 .010 .057 .033 .097 .106 .014 Support pedestal: .099 .025 .048 .038 .118 .122 .031
- See Section 6.5.2.3 of the Licensing Report for definitions.
6-65
= _ _ - - -
Table 6.7.22 4
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: F1 120 9
Holtec Run I.D.: drf1mheo.hx2 Seismic Loading: MHESET3x1.25 Fuel Assembly I.D. and Weight: 1380# reg. ; 1380.0 (lbs.)
Fuel Loading: 60 cells loaded; Fuel centroid X,Y: 21.7, .0 (in.)
Coefficient of friction at the bottom of support pedestal: 0.2 SRevision: 3.46 S SLogfile: C:/ racks /dynam0/ dyna =0.fov S
$ Revision: 2.5 S SLogfile: C:/ racks /dynam0/dynasi.fov $
$ Revision: 3.36 S
$Logfile: C:/ racks /dynam0/dynas2.fov S
! DYNAMIC IMPACT LOACS (lbs.)
(1) Maximum total vertical pedestal load: 196773.4 (2) Maximum vertical load in any single pedestal: 70878.8 (3) Maximum shear load in any single pedestal: 12104.4 (4) Maximum fuel-cell impact at one local position: 603.1 (5) Maximum rack-to-wall impact at baseplate: .0 (6) Maximum rack-to-wall impact at rack top: .0 (7) Maximum rack-to-rack impact at baseplate: .0 (8) Maximum rack-to-rack impact at rack top: .0 4
MAXIMUM CORNER DISPLACEMENTS (in.)
Location: X-direction Y-direction Top corner: .0644 .0617 Baseplate corner: .0015 .0024 4
MAXIMUM STRESS FACTORS
- Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .048 .012 .058~ .033 .098 .107 .018 Support pedestal: .098 . 024 .034 .036 .120 .126 .022
- See Section 6.5.2.3 of the Licensing Report for definitions.
6-66
Table 6.7.23 1
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: F1 120 Holtec Run I.D.: drf1mheo.hy8 Seismic Loading: MHESET3x1.25 Fuel Assembly I.D. and Weight: 1380# reg. ; 1380.0 (1bs.)
Fuel Loading: 60 cells loaded; Fuel centroid X,Y: .0, 26.0 (in.)
Coefficient of friction at the bottom of support pedestal: 0.8
$ Revision: 3.46 S SLogfile: C:/ racks /dynam0/ dynamo.fov $
SRevision: 2.5 $
i $Logfile: C:/ racks /dynam0/dynasi.fov $
SRevision: 3.36 $
SLogfile: C:/ racks /dynam0/dynas2.fov $
DYNAMIC IMPACT LOADS (lbs.)
(1) Maximum total vertical pedestal load: 222690.4 (2) Maximum vertical load in any single pedestal: 82444.2 (3) Maximum shear load in any single pedestal: 35540.4 (4) Maximum fuel-cell impact at one local position: 351.2 (5) Maximum rack-to-wall impact at baseplate: .0 (6) Maximum rack-to-wall impact at rack top: .0 (7) Maximum rack-to-rack impact at baseplate: .0 (8) Maximum rack-to-rack impact at rack top: .0 i
MAXLWM CORNER DISPLACEMENTS (in.)
i Location: X-direction Y-direction
, Top corner: .0703 .0981 Baseplate corner: .0018 .0021 MAXIMUM STRESS FACTORS
- Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .059 .016- .051 .033 .100 .107 .022 support pedestal: .114 .045 .113 .069 .166 .177_ .074
- See Section 6.5.2.3 of the Licensing Report for definitions.
6-67
. . . . _ . _ _ _ _ . . _ _ . _ . . _ _ _ _ _._ . . - . - . _ - . _ . _ _ . _ _ _ _ _ _ _ . ~ _ . . _
l i
i l Table 6.7.24 5
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK-MODULE: F1 120 4
j Holtec Run I.D.: drfimheo.hy2 Seismic Loading: MHESET3x1.25 Fuel Assembly I.D. and Weight: 1380# reg. ; 1380.0 (lbs.)
! Fuel Loading: 60 cells loaded; Fuel centroid X,Y: .0, 26.0 (in. ) '
)
j Coefficient of friction at the bottom of support pedestal: 0.2
} SRevision: 3.4L $ _.
$Logfile: C:/ racks /dynam0/dynam0.fov $
l $ Revision: 2.5 $
- SLogfile
- C:/ racks /dynam0/dynasi.fov $
[ $ Revision: 3.36 _S.
! $Logfile: C:/ racks /dynam0/dynas2.fov S f
DYNAMIC IMPACT LOADS (lbs.)
(1) Maximum total vertical-pedestal load: 224038.0 i
i (2) Maximum vertical load in any single pedestal: 81504.5 i
(3) Maximum shear load in any single pedestal: _ 15862.7 i
(4) Maximum fuel-cell impact..at_one local' position: 296.5 (5) Maximum. rack-to-wall impact at baseplate: .0 (6) Maximum rack-to-wall impact-at rack ~ top: .0 (7) Maximumorack-to-rack: impact at baseplate: .0 (8) Maximum rack-to-rack' impact at rack top:- _.0-MAXIMUM CORNER -DISPLACEMENTS - (in. ),
l Location: =X-direction: 'Y-direction >
l Top corner:- .068'4- .0940l I
l Baseplate corner: .0028 .0094
.RAXIMUM STRESS FACTORS *'
' Stress factor: R1 R2 R31 R4 RS. R6- R7; i- Above baseplate: .060' -
.015 -.050 .033 .106~ .114 .022 l
Support pedestal: .113 .020; .050 .031 .153 . 161 . 0 r1 l
.* See-Section_6.5.2.3 of the Licensing Report for definitions.
I I
e
, :6-68
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- - _ _ = . - . .- . _ __ . . . _ _ _ - . . _ _ ..
Table 6.7.25
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: F1 120 Holtec Run I.D.: drfimheo.re8 Seismic Loading: MHESET3x1.25 Fuel Assembly I.D. and Weight: 1380# reg. ; 1380.0 (lbs.)
Fuel Loading: 12 cells loa?r.., Fuel centroid X,Y: .0, .0 (in.)
Coefficient of friction at the bottom of support pedestal: 0.8 SRevision: 3.46 S
$Logfile: C:/ racks /dynam0/dynam0.fov $
SRevision: 2.5 $
SLogfile: C:/ racks /dynam0/dynasi.fov $
SRevision: 3.36 S SLogfile: C: / racks /dyr.3m0 /dynas 2. f ov $
DYNAMF1 IMPACT LOADS (1bs.)
(1) Maximum total verti x- pedestal loodt 215379.5 (2) Maximum vertical load in any siagle pedestal: 59930.3 (3) Maximum shear load in any single pedestal: 25053.4 (4) Maximum fuel-cell impact at one local position: 474.2 (5) Maximum rack-to-wall impact at baseplate: .0 (6) Maximum rack-to-wall impact at rack top: .0 (7) Maximum rack-to-rack impact at baseplate: .0 (8) Maximum rack-to-rack impact at rack top: .0 MAXIMUM CORNER DISPLACEMENTS (in.)
l Location: X-direction Y-direction Top corner: .0191 .0274 Baseplate corner: .0103 .0110 MAXIMUM STRESS FACTORS
- Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .084 .049 .029 .032 .119 .126 .062 Support pedestal: .082 .044 .078 .068 .140 .152 .051
- See Section 6.5.2.3 of the Licensing Report for definitions.
6-69
l I
l Table 6.7.26
> SUMM' RY RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: F1 120 holtec Run I.D.: drf1 neo.re2 Seismic Loading: MHESET3x1.25 Fuel Assembly I.D. and Weight: 1380/ reg. ; 1380.0 (lbs.)
Fuel Leading: 12 cells loaded; Fuel centroid X,Y: . 0, .0 (in.)
Coefficient of friction at the bottem of support pedestal: 0.2 SRevision: 3.46 S
}. $Logfile: C:/ racks /dynam0/ dynamo.fov S 1
$ Revision: 2.5 S a
$Logfile: C:/ racks /dynam0/dynasi.fov S
- $ Revision
- 3.36 $
l $Logfile: C:/ racks /dynam0/dynas2.fov $
DYNAMIC IMPACT LOADS (lbs.)
(1) Maximum total vertical pedestal load: 215605.6 (2) Maximum vertical load in any single pedestal: 57462.5 4 (3) Maximum shear load in any single pedestal: 8558.3 (4) Maximum fuel-cell impact at one local position: 773.5 (5) Maximum rack-to-wall impact at baseplate: .0 (6) Maximum rack-to-wall impact at rack top: .0 (7) Maximum rack-to-rack impact at baseplate: .0 (8) Maximum rack-to-rack impact at rack top: .0 MAXIMUM CORIER DISPLACEMENTS (in.)
Location: X-direction Y-direction Top corner: .0181 .0129 Baseplate corner: .0135 .0071 MAXIMUM STRESS FACTORS
- Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .084 .018 .013 .028 .105 .110 .019 Support _ pedestal: .079 .017 .026 .026 .100 .104 .017
- See Section 6.5.2.3 of the Licensing Report for definitions.
~
~-
1 l
I l
6-70
- _ . . . . . - ~.. .... - . .. - ....-. - .. _. -..- - - - - - - . - - .-
a t i
~
- I d
2 i
d b
I .!
Table 6.7.27 i l
t' j COMPARISON OF CALCULATED AND ALLOWABLE LOADC/ STRESSES j AT IMPACT LOCATIONS AND-AT WELDS i
4 i
i VALUE :
i- .
Item / Location
! Calculated Allowable .
3.
i l Fuel assembly / cell wall 852 ' (multi-rack). 5409
- impact, lbs.
i i Rack / baseplate weld, psi. 9876 ; 29820 i
l' Pedestal / baseplate weld .
t I
(dimensionless limit load .529 (multi-rack) 1.0 -i j ratio - conservatively =
i neglects gussets in-
- calculation)- _
- Cell / cell welds-(lbs) 3079- . 7906- I 1
i a
i 5
- j. .
4 1
l- Where -indicatedi . calculated results are baset- on bounding loads obtained from Whole Pool. Multi-Rack analyses.
).
i:
i 5
4
.I b t:
) 6-71 -
2 I
t e
,s..w.~,.,_mw,-.sd,. .;.-em,A,,'w,-, 'm
, ,, ,.-y - , . , . -' e - w 'l, , .. - ,#....- , .r s -p - ,. w . . w e. . 4 , , ,
Table 6.8.1 MAXIMUM ABSOLUTE DISPLACEMENTS OF RACK CORNERS AT BOTH THE TOP MID BOTTOM OF EACH RACK IN GLOBAL X AND Y DIRECTIONS, in.
FROM WHOLE POOL MULTI-RACK ANALYSIS FORT CALHOUN, OPPD, 11 Holtec Racks; 3/16" Bumper Bars Added to Rack Top Corners; Fully Loaded with 1380/ Reg. Fuel; Random Friction coeff. with mean = 0.5; Seismic MHE SET-4 x 1.02; dt=0.00005 sec.
Rack uxt uyt uxb uyb No. (in.) (in.) (in.) (in.)
1 .7424E+00 .7921E+00 .2669E+00 .2691E+00 2 .3339E+00 .3708E+00 .8369E-01 .9452E-01 3 .5823E+00 .2989E+00 .2632E-01 .2826E-01 4 .9939E+00 .5306E+00 .2613E+00 .2574E+00 5 .5509E+00 .6770E+00 .1537E+00 .2221E+00 6 .6328E+00 .3697E+00 .5281E-01 .4829E-01 7 .2910E+00 .2588E+00 .1394E-01 .1302E-01 8 .5671E+00 .3926E+00 .1794E+00 .2209E+00 9 .5719E+00 .3622E+00 .5524E-01 .9727E-01 10 .6145E+00 .2649E+00 .1267E+00 .4699E-01 11 .6337E+00 .4236E+00 .1815E+00 .1832E+00
$ Revision: 1.8 $
$Logfile: C / racks /multirac/maxdisp.fov $
l 6-72
Table 6.8.2 MAXIMUM ABSOLUTE DISPLACEMENTS OF RACK CORNERS AT BOTH THE TOP AND BOTTOM OF EACH RACK IN GLOBAL X AND Y DIRECTIONS, in.
FROM WHOLE POOL MULTI-RACK ANALYSIS FORT CALHOUN, OPPD, 11 Holtec Racks; 3/16" Bumper Bars Added to Rack Top Corners; Fully Loaded with 1380# Reg. Fuel; Random Friction Coeff. With mean = 0.5; Seismic DE SET-4 x 1.02; dt=0.00005 sec.
Rack uxt uyt uxb 11yb No. (in.) (in.) (in.) (in.)
1 .1994E+00 .1609E+00 .6601E-02 .5102E-02 2 .1337E+00 .1887E+00 .4301E-02 .6004E-02 3 .6820E-01 .9680E-01 .2200E-02 .3000E-02 4 .1154E+00 .1302E+00 .3647E-02 .4163E-02 5 .1149E+00 .2190E+00 .3701E-02 .7002E-02 6 .8570E-01 .9110E-01 .2700E-02 .2900E-02 7 .1058E+00 .1129E+00 .3402E-02 .3602E-02 8 .9250E-01 .1063E+00 .2901E-02 .3301E-02 9 .1257E+00 .1877E+00 .4000E-02 .5900E-02 10 .1175E+00 .1297E+00 .6104E-02 .4102E-02 11 .9460E-01 .1526E+00 .3002E-02 .4804E-02
$ Revision: 1.8 $
$Logfile C:/ racks /multirac/maxdisp.fov $
s 6-73
l 4
l 4
Table 6.8.3 I
MAXIMUM RACK PEDESTAL VERTICAL LOADS FROM WHOLE POOI MULTI-RACK ANALYSIS FORT CALHOUN, OPPD, 11 Holtec Racks; 3/16" Bumper Bars Added to Rack Top Corners; i Fully Loaded with 1380/ Reg. Fuel; Random friction coeff. With mean =0.5; Seismic: MHE SET-4 x 1.02; dt=0.00005 sec.
4 PEDESTAL MAX. FORCE TIME
! No. (Ibs.) (sec.)
1 RACK-1 1
1.491E+05 1.041E+01 1 2 1.242E+05 9.637E+00 3 1.689E+05 9.985E+00 4 1.191E+05 1.020E+01 RACK-2:
, 1 9.356E+04 1.122E+01 i 2 1.105E+05 5.604E+00 3 1.078E+05 1.167E+01
, 4 8.518E+04 1.106E+01 i RACK-3:
1 1.120E+05 1.499E+01 2 1.303E+05 1.030E+01 3 1.290E+05 6.947E+00 4 1.123E+05 1.178E+01 RACK-4:
1 1.172E+05 9.963E+00
, 2 1.263E+05 9.409E+00 3 1.226E+05 1.029E+01 4 1.017E+05 1.472E+01 RACK-5:
1 1.367E+05 7.439E+00 2 9.330E+04 1.105E+01 4 3 1.396E+05 9.977E+00 4 4 1.288E+05 1.167E+01 3 RACK-6:
1 1.204E+05 9.315E+00 2 1.083E+05 1.440E+01 3 1.379E+05 1.086E+01 4
4 1.165E+05 1.402E+01 RACK-7:
1 1.181E+05 8.432E+00 2 1.073E+05 5.412E+00 3 1.047E+05 1.273E+01 i 4 1.133E+05 1.106E+01
( to be continued )
6-74
, ( Table 6.8.3 continued )
4 i .
RACK-8:
- 1 1.103E+05 9.626E+00 2 9.189E+04 9.250E+00
} 3 1.013E+05 9.118E+00 4 8.589E+04 1.084E+01
]
RACK-9:
4 1 1.120E+05 9.849E+00
- 2 1.283E+05 9.418E+00 1 3 1.214E+05 1.039E+01 l 4 1.457E+05 1.006E+01 RACK-10
l 1 1.412E+05 1.040E+01 l 2 1.253E+05 1.245E+01 3 1.196E+05 1.274E+01 1 4 1.329E+05 1.189E+01 RACK-11:
1 1.597E+05 1.030E+01 2 1.199E+05 5.321E+00
, 3 1.711E+05 1.086E+01 4 1.357E+05 9.048E+00 1
i k
i 6-75 l
l i
i Table 6.8.4 MAXIMUM RACK PEDESTAL VERTICAL LOADS j FROM WHOLE POOL MULTI-RACK ANALYSIS i FORT CALHOUN, OPPD, 11 Holtec Racks;
- 3/16" Bumper Bars Added to Rack Top Corners; J Fully Loaded with 1380/ Reg. Fuel; i Random friction coeff. With mean =0.5; i Seismic
- DE SET-4 x 1.02; dt=0.00005 sec.
5 PEDESTAL MAX. FORCE TIME j No. (lbs.) (sec.)
l 1 RACK-1:
1 1 7.706E+04 1.743E+01 2 7.776E+04 1.422E+01
. 3 8.851E+04 8.461E+00 e
4 7.559E+04 8.366E+00 RACK-2:
1 7.076E+04 1.318E+01 i 2 7.773E+04 8.996E+00 l 3 7.206E+04 1.743E+01 4 6.379E+04 8.452E+00 RACK-3:
1 6.877E+04 6.082E+00 2 7.628E+04 1.429E+01 3 7.333E+04 1.409E+01 4 6.864E+04 1.573E+01 RACK-4:
1 5.849E+04 6.858E+00 2 7.142E+04 9.374E+00
. 3 5.947E+04 1.421E+01 4 6.473E+04 1.463E+01 RACK-5:
1 8.016E+04 1.596E+01 1
2 5.927E+04 1.533E+01 3 7.048E+04 1.623E+01 4 6.943E+04 1.335E+01 RACK-6:
1 5.269E+04 8.652E+00 2 5.184E+04 1.542E+01 3 6.591E+04 1.616E+01 4 6.043E+04 1.752E+01 RACK-7:
1 7.361E+04 6.080E+00 2 7.576E+04 1.437E+01 3 7.293E+04 5.599E+00 3
4 8.024E+04 1.573E+03
( to be' continued )
6-76
?
i i ( Table 6.8.4 continued )
i l
RACK-8:
1 6.821E+04 7.509E+00 l 2 5.837E+04 1.493E+01 3 3 6.076E+04 6.343E+00 j 4 5.957E+04 1.232E+01 RACK-9:
i 1 9.187E+04 8.364E+00 2 9.164E+04 1.366E+01 3 7.39aE+04 7.920E+00
{ 4 8.140E+04 1.289E+01 l RACK-10:
- 1 9.978E+04 6.721E+00 1 2 7.822E+04 1.430E+01 3 8.474E+04 7.835E+00
! 4 1.010E+05 8.455E+00 RACK-11:
I 1 9.338E+04 8.661E+00 2 7.452E+04 7.010E+00 3 7.282E+04 8.449E+00 i 4 8.506E+04 1.566E+01 i-I 4
+
T 1
1 6 2
+ . - . .,-, --
i Table 6.8.5 j COMPARISON OF RESULTS FROM SINGLE RACK ANALYSIS j AND WHOLE POOL MULTI-RACK ANALYSIS (WPMR)
(MHE; Fully Loaded with 1380 lb fuel)
Sj.nale Rack Analysis WPMR Analysis Maximum absolute 0.2481 0.9939 displacement, in.
4 at top of rack Maximum pedestal 124,680 171,100 vertical load, lb.
1 4
6-78
. - , . . ~ , , -, ,
1 Table 6.9.1 i AVERAGE BEARING PAD PRESSURE - COMPARISON OF CALCULATED
, AND ALLOWABLE STRESSES (f, = 4000 psi)
! Level-D a Controlling AVERAGE BEARING PAD PRESSURE
- Pedeptal (psi)
Pad Size / Location (lb)
Calculated Allowable 12" x 12" x 1.5" 134,674 1338 2380 (No Leak Chase) 12" x 12" x 1.5" 134,674 1463 2380 j (over a Leak Chase) i i
a i
This &nalysis carried out using a higher total pedestal load to reflect future fuel load scenarios.
6-79
M0289031 -
o EQUIPMENT FIIEQUENCY-HZ 25 20 m1 - 10 0 6 23 ~ 50 333-302 16F 1 43
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m HESPONSE 10 EL CEN T(10-lime IllSIOilY ANALYSIS 3 m f- I
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HESPONSE TO EL CENTHO-ilMElllSTOllY ANALYSIS 53 7
o NOllMAtlZED GROUND flESPOtJSE SPECIHUM h.
a am O I c d 4 O o z g i ti m
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, H0289034 -
CAD HEPHODUCTION OF FT. CAulOUN USAR FIGtNE F-22
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o O-rd - . GAP TIME HISTORY, FORT CAU10UN STATION, OPPD. Gap between Rock-G1 and . East Wall. South Side. Top, MHE-SET 4, Fully Loaded with 1380# Re Friction coefficient = random ( mean =g. Fuel Assemblies.0.5 ). File:g l-w-4. dot. 'i i b e.
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i i o . 1 O- - ,1 : GAP TIME filSTORY. FORT CALHOUN STATION, OPPD.
, : Gap between Rock-G1 and. Rack-G2, East Side. Top. ,
i . idl1E-SET 4, Fully Loaded with 1380g Reg. Fuel Assemblies. Friction coefficient = random ( rneon = 0.5 ). File: 9 1-g2-22. dot . , o .- co. .. , o- _ o-to ..
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o ' o-J GAP TIME illSTORY, FORT CAlliOUN STATION, OPPD.
' : Gap between Rack-G2. and. Rack-E, South Side, Top, Mile-SET 4, Fully Loaded with 1380// Reg. Fuel Assemblies, Friction coefficient = random ( rneon = 0.5 ). File: 92-e-4 0. dot.
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l . 1 4 ! 7.0 ACOIDENT ANALYSIS AND MISCELLANEOUS STRUCTURAL EVALUATIONS 7.1 Jntroductign l This section provides results of accident analyses and j miscellaneous evaluations performed to demonstrate regulatory l compliance of the new fuel racks. 1 l i The following limiting accident and miscellaneous structural f evaluations are considered: 4 i + Refueling Accidents i ! a. Drop of a fuel assembly plus handling tools weighing l 2780 lbs. (dry) - which enters _ an empty cell and j impacts the baseplate. ; f b. Drop of a fuel assembly plus handling tools weighing
- 2780 lbs. which impacts the top of a rack.
l c. Drop of the spent fuel pool gate (11,522 lb.-dry 1 weight) . (between the spent fuel pool and the fuel transfer area) - with rotation of gate onto top of , fuel rack. The above drop accidents are evaluated to ensure that the ' new fuel racks are capable of withstanding the event to the extent that no encroachment on the current radiological safety envelope occurs _and that loads transmitted -to the spent fuel pool slab are below the bounding loads frva- the postulated bounding - seismic . event. 4 Local cell wall buckling Analysis of welded joints due to isolated but cell 7.2 =Refuelina Accidents 7.2.1 Droceed Fuel Assembly The consequences of dropping a fuel assembly as it is being moved over stored fuel is discussed below; Based on the highest lift of j a new fuel assembly plus associated handling equipment, the maximum-distance from the bottom of a fuel _ assembly, travelling over fuel l l l 7-1 ! l
1 i i i i racks, to the top of the rack is assumed as 15" in water. To bound l the solution and to account for future fuel loads, we evaluate the accidents for a fuel assembly plus associated handling equipment weight of 2780 lbs. (higher than the 1680 lb. dry weight for the present fuel plus handling equipment). l
- a. Droceed Fuel Assembly Accident (Deen Dron Scenariol i
1 A 2780 lb weight (fuel assembly plus handling tools) is i dropped from 15" above the top of a storage location and impacts the base of the module. Local failure of the I baseplate is acceptable; however, the rack design should .* ensure that gross structural failure does not occur and the subcriticality of the adjacent fuel assemblies is not violated. Calculated results show that there will be no ) change in the spacing between cells. Local deformation
)
, of the baseplate in the neighborhood of the impact will 3 occur, but the dropped assembly will be contained and :' not impact the liner. We show that the maximum movement of the baseplate toward the liner after the impact is
! less than 2.615". The load transmitted to the liner through the support by such an accident is demonstrated i to be well below the loads caused by seismic events j (given in Section 6).
- b. Droceed Fuel Assemb1v Accident (Shallow Droo-Scenario) one fuel assembly plus handling equipment is dropped from l 15" above the top of the rack and impacts the top of tl.e i rack. Bounding kinetic energy to be absorbed (neglecting any fluid drag) is 3023 f t.-lb. permanent deformation of the rack is acceptable, but is required to be limited to i
the top region such that the rack cross-sectional l geometry at the level of the top of the active iuel I bundle (and below) is not altered. A conservative assessment of energy absorption capabilities of the top i ' structure (only one cell wall absorbing energy) indicates that damage will be restricted to a depth of between 4.4" j and 7.3" below the top of the rack. This is above the j top most projection of a fuel assembly. 7.2.2 Gate Dron onto the Too of a Rack.
-The 11,522 lb. (dry weight) . spent fuel pool gate is assumed to be dropped from a height of 4.5 feet, impacts the lower edge _of-the spent fuel pool gate opening and rotates over to impact a rack. The
[ analysis used to model this accident is based on the gate starting l: 7-2 i
i l from rest and in the vertical orientation and then rotating over l to the horizontal position and impeting the top of the rack. It I is shown that damage to the rack is confined.to the region above
- I l the active fuel area with maximum depth of damaged area 1.17" based on an assumption that only 6 cells absorb the kinetic energy of rotation. Loads transmitted to the spent fuel pool liner are i enveloped by the loads from the bounding seismic event.
1 l 7.3 Local Bucklina of Fuel Cell Walls S This subsection presents details on secondary stresses produced by ; local buckling. L The allowable local buckling stresses in the fuel cell walls are ,
-obtained by using classical plate buckling analysis. The following formula for the critical stress, which is appropriate for one wall-of a rectangular cell, has been used based.on a width of cell "b"
, (7.3.1) (see Figure 7.3.1 where q-is the applied stress, a is the length of the section and ' b is the width. t is the plate thickness). l . 2 2
. 8. 7 Et l ca =
12 b2 (y _ p 2) The above result assumes simple support conditions on all sides of the buckling plate. o,is the limiting vertical compressive stress in'the tube, E = 27.6 x 106 psi, p =L0.3, (Poison?s-ratio), t=
.075", b = 8.46"..The factor.8 is a coefficient depending on a/b and has a minimum value of.4.0 for'a/b = 1-[7.3.1).-For buckling below;the end of the sheathing a/b = 0.4, 8 = 8.41.
For.the-given data, o,=-16487-psi t 7 ..-,a..--. . . - ..-.;-.--.. - ,. , - . - - , . - . . , - _ .
.:.- - w - .z.,,-
It should be noted that this elastic stability calculation is based on the applied stress being uniform along the entire length of the cell wall. In the actual fuel rack, the compressive stress comes from consideration of overall bending of the rack structure during a seismic event and as such is negligible at the rack top and
- maximum at the rack bottum. It is very conservative to apply the above equation to the rack cell wall if we coeparc n,, with the j maximum compressive stress anywhere in the cell wall. As shown in Section 6, the local buckling stress limit is not violated anywhere in the body of the rack modules. The maximum compressive stress in the outermost cell is obtained by multiplying the limiting value j of the stress factor R6 (for the cell cross-section just above the baseplate) by the allowable stress. Thus, from Table 6.7.2, o=
R6 x allowable stress =
.321 x (.6 x 25000) = 4815 psi under faulted conditions.
7.4 Analysis of Welded Joints in Rack due to Isolated Hot Cell In this subsection, in-rack welded joints are examined under the loading conditions arising from thermal effects due to an isolated hot cell. A maximum thermal gradient between cells will develop when an isolated storage location contains a fuel assembly emitting maximum postulated heat, while the surrounding locations are empty. We can obtain a conservative estimate of weld stresses along the length of an isolated hot cell by considering a beam strip (a cell wall) uniformly heated and restrained from growth along one long edge. The strip is subject to a uniform temperature rise AT = '86.5*F. The temperature rise has been calculated from the difference of the maximum local water temperature and bulk water temperature in the spent fuel pool (see Tables 5.8.2 and 5.8.4). Then, using a shear beam theory, we can calculate an estimate of the anximum value of j e 1 7-4
l 1 l the average shear stress in the strip (see Figure 7.4.1, which ! shows a " hot" strip section welded to a " cold" aase along a length i l L of area tH). . The final result for wall maximum shear stress, under conservative restraint assumptions is given as (7.5.1): Ea T
.931 l
Where a = 9.5 x 10 4 in/in 'F. Therefore, we obtain an estimate of maximum weld shear stress in
- an isolatnd hot cell as f.= 24626 psi Since this is a secondary thermal stress, it is appropriate to compare this to the allowable weld shear stress for a faulted event f < .42S, = 29820 psi. In the fuel rack, this maximum stress occurs near the top of the rack and does not interact with any other critical stress.
7.5 References for Section 7 (7.3.1) " Strength of Materials", S.P. Timoshenko, 3rd Edition, Part II, pp 194-197 (1956). (7.5.1) " Seismic Analysis of High Density Fuel Racks, Part III -Structural Design Calculations - Theory", HI-89330, Revision 1, 1989. 7-5
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1 i 1 l j 1 j 8.0 FUEL POOL STRUCTURE INTEGRITY CONSIDERATIONS 8.1 Introduction l l The Fort Calhoun Station (FCS) spent fuel pool is a safety related, l seismic category I, reinforced concrete structure. In this section, the analysis to demonstrate structural adequacy of the pool structure, ag rpquiz9d by Section IV of the USNRC OT Position-j Paper (8.1.1), is abstracted. j The FCS spent fuel pool region is analyzed using the _ finite element i method. Results for individual load components are combined using l- factored load combinations mandated by SRP 3.8.4 (8.1.2) based on the " ultimate strength" design method. These load combinations are j_ generally more conservative than those in the ACI Code [8.1.3). It 1s demonstrated that for the critical bounding factored load l j combinations, structural integrity is maintained when the fuel pool . ! is assumed to be fully loaded with new, free-standing, high density fuel racks with all storage locations occupied by fuel assemblies. i The regions examined in the fuel pool are the_ slab, the highly l loaded wall sections adjoining the pool slab and_the foundation ! pilings. Both moment and shear capacities are checked for concrete structural integrity. Local punching and bearing integrity of the
- slab in the vicinity of a rack module support pedestal pad is evaluated. All- structural capacity calculations are made -using design formulas meeting the requirements _of the American Concrete 1
Institute (ACI). 8.2 Descriotion of Soent Fuel Pool Structure
- The FCS spent fuel pool is a rectangular _ reinforced concrete _(RC) j structure with'a stainless steel liner and is supported by the slab -
foundation mat. The-spent fuel-pool (SFP) slab is 12 ft, thick. { The top of the - SFP floor slab is at EL. 995'-6". ,The mat is-l continuous under the Containment end htxiliary. Buildings ~ and is 8-1
supported by foundation piles. The RC walls, 43 feet high above j the mat, on the north, east and west sides of the SFP, are 5'-6" thick and also provide support to segments of intermediate floors of the Auxiliary Building. The south wall of the pool separates the SFP from the fuel transfer canal. This wall is 4 '-0" thick and j is notched near the southeast corner of the pool to allow for the transfer of fuel into the pool. A gate is installed in this notch during normal operation. The walls of the SFP are supported on the foundation mat. Some lateral support to the spent fuel pool wall is provided by adjacent intermediate floors. Adjacent to the SFP, to the south, is the fuel transfer canal. This canal is assumed to be dry for this evaluation. This ,' assumption will provide the most limiting load condition for the l pool south wall. Figures 8.2.1 and 8.2.2 show a finite element grid i representation of the model to be considered. 8.3 Definition of Loads ] Pool structural loading involves the following discrete components: 8.3.1 Static Load.ing ] 1) Dead weight of pool structure plus 43 feet (high water limit) of water (including hydrostatic
- pressure on the pool walls). Combining the hydrostatic pressure and structure dead weight is in conformance with (8.1.3].
- 2) Dead weight of rack modules and fuel assemblies stored in the modules.
8.3.2 Dynamic Loadino
- 1) Vertical loads transmitted by the rack support pedestals to the slab during a MHE or DE seismic event.
- 2) Inertia loads due to the slab, pool walls and contained water mass and sloshing loads (considered in accordance with (8.3.1)) which.arise during a seismic event.
- f 8-2
l
- 3) Hydrodynamic loads caused by rack motion in the pool '
during a seismic event. l 8.3.3 Thermal Loadina Mean temperature rise and temperature gradient across the pool slab and the pool walls due to temperature differential between the pool water and the atmosphere externcl to the slab and walls. 8.3.4 Loads from Adjacent Structure As noted previously, the walls of the spent fuel pool act as support to certain floor structures external to the pool. The external floors are included in the finite element model as necessary to simulate the mass and stiffness effect on the spent fuel pool walls. 8.4 Analysis Procedures 8.4.1 Finite Element Analysis Model The gridwork in different regions of Figures 8.2.1 and 8.2.2 shows the totality of elements used. The detailed finite element model is only of the walls, the slab, and the main vertical pilings. The structure external to the spent fuel pool is modelled only to the extent that the appropriate interactions are accounted for. The finite element code used is ANSYS [8.4.1). The three-dimensional finite element model is constructed using STIF45 solid elements for the main pool wall and slab, and STIF63 shell element and STIF14 spring elements to model the external structure and imbedded pilings. Tt:s element thickness in the various regions of the structure is the actual thickness of the st'ucture at the location. The finite element model is prepared for the analysis of both mechanical load and thermal load. The effect of reinforceLant and concrete cracking is reflected in element properties assigned to various locations during the simulations. The general reinforcement pattern of the structure is as follows: l l 8-3
]
l i l The slab has #14 bars at 12" spacing each way on the top face of the slab and #9 bars at 12" spacing in the north-south direction and #11 bars at 12" spacing in the east-west direction on the bottom face of the slab. The walls have reinforcement on each f ace in each direction. The walls use #6 through #11 bars for reinforcement. The intent of the model is to provide a conservative evaluation of the moment distribution in the slab and in the surrounding pool walls. Towards that end, any rotational resistance from any lateral external supporting member not explicitly modelled by solid elements is neglected. 8.4.2 Analysis Methodolony In Section 6 of this document, Whole Pool Multi-Rack analyses have been carried out. The results of these analyses (for MHE and DE seismic events) establish pedestal load time histories on the pool slab and hydrodynamic pressure-time histories for the wall structure adjacent to the racks. For spent fuel pool re-qualification, the pool is assumed to ( contain eleven free-standing, fully loaded, spent fuel racks. A total of 1083 cells are assumed loaded conservatively with fuel having dry weight 2480 lbs per assembly. This loading bounds the actual dry weight of regular fuel (1380 lb per assembly). 8.4.2.1 The major structure loadings discussed in Section 8.3 are imposed on the finite element model in the following manner:
- a. Dead weight of the structure simulated by a 1-g vertical gravitational load imposed on the model,
- b. Loads (static and dynamic) from the racks plus the fuel loads are imposed on the slab as effective uniform pressure loads. Effective dynamic adders are obtained from the results of the Whole Pool analyses. The dynamic adders are defined as 8-4
-.-- . - . - - - . - - . ~ . _ - - - - - - - - - _ - - - . - . - . -
l. I j equivalent static loads giving the same total j impulso as the actual timo varying dynamic loads i during the time period when the total load is j greater than the static load. i
- c. Water loads imposed on walls and slab as a l hydrostatic pressure and as effective uniform 4
j horizontal pressures to simulate the effect of various hydrodynamic effects. i d. The pool water temperatures (section 8.3.3. ) are j assumed uniform in space. Heat conduction / convection analyses on representative wall and slab sections
- are performed to obtain actual concrete surface ,
i temperatures which are then imposed on the elements ' { to simulate a mean temperature plus a temperature j gradient. The maximum pool water temperature is-i 140*F. Ambient air temperature is assumed to be 70*F i and the steady state' grade _ temperature is assumed to be 50*F twelve feet below bottom of the slab for-the purpose-of establishing concrete temperature distribution through.the concrete. The temperatures present a limiting. thermal- load under both normal , and abnormal operating conditions. j 8.4.2.2 The effect of concrete cracking (permitted in analyses ! per the ACI Code) is simulated by appropriate reduction of . the element Young's Moduli in appropriate regions. In this manner load re-distribution that will occur in the walls. and slab is accounted for in the model. Concrete cracking is only considered under the thermal load condition. 1 i
- l. 8.4.3 Load combinations-l i For the pool structural analysis, cert'ain factored ' load
! combinations are required in-accordance with_(8.1.2).- Out of all 4 the mandated _ f actored loading . combinations, the following are l
. potentia)1y' limiting (af ter deleting those loads which are not
.. applicable =to the FCS spent fuel pool). i
- 1) 1.4D + 1.9E 3
- 2) . 7 5 - (1. 4 D + 1. 7To . t 1. 9E)
)
'3) _D + T. _E'
, 4) D ~ + T, . l1.25E 4 _5? e 8-5 e.-,.m.--- , ., u,.% , -- -c,,y',, - , . . , , . , ,_,,,_.,_,,n.,.. ,-.y ,,-w.E,. .,,%--,.g. . . ,urc,__-.-r
- . . - . . . . - - - - . - . - ~ - = . - . - . . . - - . . - - - . - ~ . . - . - _ . - - - - . .
i I 4 1 i 4 1 In the above defined combinations, the notation of [8.1.2) is used
} D = dead loads and hydrostatic loads l j E = DE resultant loads
- E' = MHE resultant loads I
To = thermal loads due to postulated normal thermal condition; to be conservative, we assume T, = T . T= thermal loads due to postulated abnormal thermal , i condition , i
! In the load combinations involving seismic combinations, note that i
the seismic effects act with "plus" or "minus" signs on the I ) structural and hydrodynamic components to reflect the arbitrary l directional nature of the seismic loading. Both directions are l evaluated to establish the limits at various structural locations. t i i i l The dead load and the thermal ~1oad tend to reduce _ each other in 4 l some of the areas. In accordance with SRP _3.8.4 mandated j procedures, adjustments to load _ coefficients are made and l additional combinations performed when a load acts in a direction so as to reduce the effect of other loads. For considerations
- involving dead and thermal loadings in conjunction with the seismic -
l load, additional combinations are carried out with reduced l coefficients on dead load and/or thermal load. ! 8.5 Results of Analyses l I j The ANSYS postprocessing capability is used to form the appropriate load combinations and to establish the limiting bending moments and l ! shear forces in various sections'of the pool structure.- Section f l limit strength formulas for bending loading are computed using appropriate concrete and reinforcement strengths. For FCS, the
- _ concrete and reinforcement allowable strengths are
- .
i concrete f,' =- 4000 psi reinforcement fy = 40000 psi 8-6 _- _ __,,& ,_,.__-.u_...n_- . _ . _ . , . _ . - _ - . .a,,._.---,
_.__m.___.___ . _ . _ _ _ _ . - _ _ _ _ . _ _ _ . ~ . _ _ _ _ . _ _ _ . _ _ _ . _ _ . _ . _ _ l i 1 i i
) To assess the potential importance of cartain load components on l the FCS pool structure, a calculation is made to evolve an j equivalent static uniform pressure on the pool slab from each of i the major distributed mechanical loads. Table 8.5.1 shows that the
! major contributions are from the water and the racks plus fuel, and that the dead load contribution is significantly higher than the j seismic contribution. Table 8.5.2 shows results from potentially l limiting load combinations for the bending strength of the slab and walls. For each section, we define the limiting safety margins as ; l the limited strength bending moment or shear force defined by ACI I for that structural section divided by the calculated bending l moment or shear force (from the finite element analyses) . The major ! regions of the pool structure consist of 13 different areas ) corresponding to different reinforcements and section thickness. Each area is searched independently for the maximum bending-moments in different bending directions and for the maximum shear forces. j safety margins are determined from the calculated maximum bending moments and shear forces based on the local strengths. The procedures are repeated for all the potential limiting load' combinations. Therefore, limiting safety margins are determined. l l Table 8.5.2 demonstrates'that the limiting safety margins for all I sections are above 1.0 as required.- Table 8.5.3 shows results of-' shear capacity calculations _for the slab and walls. Calculated margins are again to be compared with an allowable margin of 1.0. l l Table 8.5.4 presents results for the pilings. Margins. must - be greater or equal to 1.0. 1 i I L '8-7
. . _. -_. ,. w ,_ . . - _ . . , _ . _ _ _ _ . _ _ _ _ , _ _ . . _ . . , _ . , _ _ , _ . -
. . - . _ . . _ - _ . - - - = . . - . . . - . - . _ . .. -. _.
1 1 ) 8.6 Pool Liner i The pool liner is subject to in-plane strains due to movement of the rack support feet during the seismic event. Analyses are l performed to establish that the liner will not tear or rupture under limiting loading conditions in the pool, and that there is l no fatigue problem under the condition of 1 MHE event plus 5 DE ! events. These analyses are based on loadings imparted from the most 4 j highly loaded pedestal in the pool assumed to be positioned in the j most unfavorable position. Bearing strength requirements on the I most highly loaded pedestal are shown to be satisfied even in the event that the bearing pad is located over a leak chase. i 8.7 Conclusions
- Regions affected by loading the fuel pool completely with high density racks are examined for structural integrity under bending and shearing action. It is determined that adequate safety margins l exist assuming that all racks are fully loaded with a bounding fuel
! weight and that the factored load combinations are checked against the appropriate structural design strengths. It is also shown that i local loading on the liner does not compromise liner integrity 4 under a postulated fatigue condition and that: concrete bearing strength limits are not exceeded. 8.8 References for Section 8 [8.1.1) OT Po'sition for Review and Acceptance of Spent Fuel Handling Applications, by B.K. Grimes,' USNRC, Washington, D.C., April'14, 1978. l (8.1.2) NUREG-0800, SRP-3.8.4, Rev. 1, July 1981. l (8.1.3) ACI 349-85, Code Requirements for Concrete Nuclear Safety Related Concrete Structures, American Concrete Institute, Detroit, Michigan. 8-8 l l - ,
(8.3.1] " Nuclear Reactors and Earthquakes, U.S. Department of Commerce, National Bureau of Standards, National Technical Information Service, Springfield, Virginia (TID 7024). , (8.4.1) ANSYS User's Manual, Swanson Analysis Rev. 4.4A, 1990. i l l l l l 8-9
i i 1 : i I 1 4 , I !
\ r i ;
4, l I f Table 8.5.1 i ! CALCULATION OF EQUIVALENT SLAB VERTICAL PRESSURES i , 1 i i- ' I . t-l- Slab:. Weight 12.5 psi i (average 1
- i. '
thickness-12'). ! Hydrostatic Load 18.63 psi i t (43 Water) i a- Rack + Fuel:. .26.2 ps i- - i j Dynamic-Adder j (Water, Racks + and Fuels) (NHE) 5.42 psi +
^
- Dynamic Adder l (Water,-Racks
} -and Fuels) '(DE) 4.79 psi I l i. t f. 4 I 6 f i l i 8 . . r . . . _ ~. A v ! - .. , , .L,-, ,,,,'+,
~
. , ..Jm . # . =,_ ,, ,,..s.._a L,m,,,, ,r, n. - . . . _ _ _ _ . , , .f#4 , , , ..,#, ....~,1_s-
Table 8.5.2 BENDING STRENGTH EVALUATION Critical Limiting Load Combination Location Safety Marain (see Section 8.4.3) East wall 1.61 2 West wall 1.57 2 North wall 1.79 2 Intermediate wall between fuel pool 1.61 2 and fuel transfer canal South wall 2.48 2 Slab 2.31 1 8-11
l 1, Table 8.5.3 4 SHEAR STRENGTH EVALUATION 4 i Critical Limiting Load Combination Location Safety Marcin (see Section 8.4.3) l East wall 1.14 2 7 West Wall 2.01 2 North wall 1.74 2 4 i Intermediate wall 1.03 1 South wall 2.46 2 Slab 2.86 1 O i e 8-12
4 J f
! Table 8.5.4 RESULTS FOR PILE ANALYSIS 1
Limiting Safety Marain Average strength .1.17 Local strength 1.07 4 1' 1 4 1 8-13
1 4: -
't q'):.
i 4, . FCS SFP FE MODEL - UIDI FROM SOUTH ERST' FIGURE 8.2.1 8-14 l
l \ l ( i s
-s i ~
I KS SFP FE MODEL - UIBl FROM NORTH llEST FIGURE 8.2.2 8-15
1
~
4s
';t
~
x
~
l[
,,.w...
#4 FCS SFP FE HODEL - UfEH FROM SOUM ERST FIGURE 8.2.1 8-14
l 1 l r l ' l i ; i i i
~
L l *
* . i
~ _
, I !;
~ .
V
* = .,
FCS SFP FE410 DEL - UIB1 FROM NORTH llESE .[ , FIGURE 8.2.2 l l l l t 8-15 ; I
.,,,.. -_.-_ -.. m ,_... ..-_..-_.._., . , _ . - _ . _ , _ _ _ _ . . . . _ . . _ . . . -
9.O RADIOLOGICAL EVALUATION The radiological evaluation summarized in this section considers th9 potential fo't increases in various radioactivity levels based u ..n possible changes in fission and activation product inventories
,f associated with the projected increase in spent fuel storage in the y spent fuel pool.
3 4.1 Radfolooical Consecuences of Accidg]lta 4 stential radiological consequences of accidents around the SFP ,, the Fort Calh. sun Station Auxiliary Building have been k_i -rmined. 9- . Puel Handlino Accident Assumotions and Scsurce Term Criculations ,, Evaluations of the accidents were based on fuel burne.d to 36,000 Mvd/mtU (Bounding Value) . The reactor was assumed to have been aperating at 1500 Mwt prior to shute wn. The assumptions used in the Fuel Handling Accident analysj. Je been previously reviewed end accepe by the U. S. Nuclear Regulatory Commission-(9.1.1). As in the USAR evalaation, the fuel handling accident was conservatively assumed to result in the release of the gaseous fission pre: acts contained in the fuel / cladding gaps of all the rods in the peak-power fuel ossembly at the time of the accident. Gap inventories of fission products available for release were-estimated using the release fractions identified in Table 14.18-7 of the USAR (plant specific) . The NRC staff performed a independent fuel handling accident dose calculations using essumptions
- 1. (a) through C. 1. (k) of Regulatory I'containedinPo 9-1
Guide 1.25 and the procedures specified in Standard Review Plan (SRP) Section 15.7.4. Dose calculations were performed for a fuel cooling time of 72 hours. The gaseous fission products that have significant impacts cn the off-site doses following short fuel cooling periods are the short-lived nuclides of iodine, xenon and krypton, which have saturation inventories during in-core operation. These inventories depend primarily on the fuel specific power over the few months immediately preceding reactor shutdown. In the highest power assembly, the specific power and hence the inventory of iodine, xenon and krypton will be proportional to the peaking factor (1 1.65) per Reg. Guide 1.25. At the conservative cooling time of 72 hours used in the Fort Calhoun calculations, most of the thyroid dose comes from Iodine-131, while most of the whole-body dose comes from Xenon and i Krypton. The present evaluation uses values for atmospheric diffusion factor , (X/Q) and for filter efficiencies that have previously been reviewed and accepted by the NRC in Ref. [9.1.1]. Release fractions under steady state conditions (Curies) of fission products were estimated using ANSI /ANS-5.4, 1982, "American National Standard Method for Calculation of the Fraction Release of Volatile Fission Products from Oxide Fuel", based upon parameters stated earlier (specific power of 36,000 Mwd /mtU and cooling time of 72 hours). Ths results of the calculations for isotopes of interest are given in Tat,le 9.1.1, along with the percentages of the core inventories released from the fuel to the fuel rod gaps. Table 14.18-9 of the Fort Calhoun Station USAR identified' key parrtAters used in evaluating the Radiological Consequencas of a Fuel Handling Accident. 9-2
4
! l
+ 2 4 o 1 Radiological' effects of a dropped fuel assembly were considered. To determine the extent-of damage possible to a fuel assembly during handling, it is assumed that the fuel assembly is dropped during handling. The dose calculation presented here assumed all 176 rods l j in a single fuel assembly would fail' and contribute to ti,e . gas f release. The failure of 176 rods (full assembly) _ is a NRC review requirement that is consistent for all power plants. i Fuel Handlina Accident Results ! The Fort Calhoun Station = site boundary doses from the speu.:.fied; fuel handling accident are tabulated in Table 14.18-8 of the-USAR. l- The doses, which are based =on the release of all gaseous fission product activity in-the gaps of all the fuel rods in the highest f l power assembly, are given and reproduced as Table 9.1.2. i 0.1.2 Cask Dron Accident OPPD does not presently utilize or own a spent fuel sample cask or ! spent fuel shipping cask. If such casks were to be utilized, the single-failure-proof. main hook of-the Auxiliary Building Crane
- ould -be -used -to move these casks. In addition,- electrical l' interlocks are installed which preclude travel of the main. hook and auxiliary hook over the ~ spent. fuel pool (NUREG-0612). .The l likelihood-of occurrence of a cask drop is greatly reduced through
! these considerations. - Heavy Load . paths have been proposed to , circumvent carrying a cask over spent fuel. _Thus the ' need for consideration' of a- radiological . release . impact of a cask . drop L accident -is obviated.- OPPD concluder4; that an ' analysis ' of the ! radiological ~ consequences of a cask drop accident is'not required. I I i-i [ 9-3
- y. ., ,c..- ,. -
.m., ,., , ,.r.%
9.1.3 Scent Fuel Pool Gate Droo Accident l The spent fuel pool gate is moved by using the 10 ton auxiliary hook (which is subject to single failure) of the Auxiliary Building Crane. The gate is moved directly from its notch in the spent fuel pool wall to storage in the fuel transfer canal without moving over spent fuel. If the gate was dropped and was to undergo a highly unlikely rotational motion following swinging of the supporting wire rope and hook, it could impact the tops of the spent fuel racks. The impact kinetic energy transmitted to a particular cell is no more than that transmitted in a fuel handling accident; therefore, offsite radiological consequences would not exceed those from such an accident. 9.1.4 Seismic Event in the Scent Fuel Pool e It will be verified that the seismic loads imposed on the fuel pool liner walls will not result in any damage to the liner such as to cause (1) significant releases of radioactivity due to mechanical damage to the fuel, (2) significant loss of water from the pool which could uncover the fuel and lead to release of radioactivity due to heatup, (3) loss of ability to cool the fuel due to flow blockage caused by a portion or one complete section of the liner plate f alling on top of the fuel rack, (4) damage to safety related-equipment as the result of pool leakage, and (5) uncontrolled release of significant quantities of radioactive fluids to the environs. The SFP liner is designed to support all postulated or effective dead and live loads, hydrostatic loads, temperatures to 200'F, and the effects of seismic events. Drainage grooves are provided behind
- the stainless steel liner which permit detection of any liner leakage. The liner is designed to withstand seismic loads without failing, and thus, no postulated accident could damage, uncover the fuel or result in radioactive releases in excess of the 10CFR100 limits.
9-4 4
.y - , . . . , -.
m _ . _ _ . . _ _ . . ___ _ _ .__ _ _ _ _ _ __ ._
- 'I 4
1 . Thus, the seismic loads imposed on the fuel pool liner surfaces 3 1 i will not result in damage to the liner, and there would be no f-l consequential-releases of radioactivity from-the pool, liner due to j mechanical damage to the fuel. There would be no significant losses [ of water from the pool which could uncover the fuel, since the l liner is designed to withstand an earthquake. No safety related 4-equipment is within the postulated leakage area. The uncontrolled j releases of significant quantities of radioactive fluids to the-i environment would not occur, since the-liner plate is-designed to l prevent leakage during the most severe accidents. ! 9.2 -Solid Radwaste The only solid radioactive waste associated with operation of the l j spent fuel pool purification system is - in the filters and l demineralizers, which have solid waste volumes-per change-out of 5 , ft3 and 45 ft', respectively. The necessity for-resin replacement 1 l is determined primarily by.the requirement for water clarity, and the resin is normally changed about once per operating cycle. No ! significant-increase'in the volume of solid radioactive wastes is j expected - with the expanded storage capacity. During reracking h operations, -a small amount of additional resins may-be generated by the pool cleanup system on a one-time basis.- l-l-
' The existing spent fuel storage racks willLbe released as low level
- radioactive waste. The spent fuel storage racks may be required-to
- be stored temporarily onsite due to the uncertainty of oper'ation of j the Central States Compact low level radioactive' waste facility, f - Nebraska is'a member of the' Central States Compact which has' sited i
a Low Level Radioactive Waste facility in-Nebraska which-according ! to the Waste' Policy Act is' required to be-operational by Jan.-~1,- , 11993. The old spent fuel-racks will be stored on site if OPPD is l denied access to an existing low level radioactive: facility i i I L L , 9-5 i 8 , , . . , ,-,-.+-.--..<-m 4 ,' + - . .-w . ., w a e w i , ...w. .mmn .,,...,_-~,.,.En,.,.~r4 ...,,,v.....+,.r,-y,, ,h bw , ., y, , why, ,
- - - - . - . . - ~ . _. . .- . -- . .. . - -- - - . . -.-
s ,i i~ (Hanford, WA, Barnwell, SC, or Beatty, NV). Fort Calhoun Station } has the capacity to store spent fuel storage racks on site on a
- . temporary basis.
i i l 9.3 Gaseous Releases l l Gaseous releases from the fuel storage area are combined with other l plant exhausts. All effluents from this area are discharged through l the auxiliary building stack, via the auxiliary building HVAC l system and are monitored through gaseous _ effluent and particulate j monitors. In the event that these gaseous affluent monitors [ indicate activity levels are in excess of the alarm limits,.the j auxiliary building flow paths . are manually closed. The exhaust j ventilation ductwork from the spent fuel storage = area is equipped ] with iodine absorbers - which are manually brought into operation j whenever irradiated fuel is- being- handled. 1Normally, the j- contribution from-the fuel storage area-is negligible compared to -- the other releases and no significant increases are expected as a l result of the expanded storage capacity. - F e l_ The spent fuel pool area is monitored by an area monitoring system. i Actual' dose rates can be read locally and in the control room. The f monitor will alert personnel in the SFP area when dose rates exceed-the setpoints. .This monitor is located outside the SFP wall. Protection-- personnel carrying portable monitoring'_ Radiation l instruments will also provide a further means to -identify increases in radiation levels. Additionally, a radiation monitor-is located
-on-the spent-fuel handling machine to indicate dose rates in this
[ area. The anticipated- consequences of expanded . storage capacity
- will not require any modification of the area monitoring system.
{. t f. } 9-6
,, . . _ _ . _ _ . . . . , , . . _ _ _ ..,. _,, a , . _ _ , . , . . . . . . _ . _ . ,,,-.;.-_..........~.,,.~...;-.,,,
9.4 Personnel ExDosures During normal operations, personnel working in the fuel storage area are exposed to radiation from the spent fuel pool. Operating experience has shown that the area radiation dose rates, which originate primarily from radionuclides in the pool water, are generally 1-3 mrem /hr, with an occasional reading of 5 mrem /hr. Dose rates on the pool bridge platform are 4 to 5 mrem /hr. These doses may temporarily increase clifaely during refueling operations. Radiation levels in zonen surrvinding the pool are not expected to be significantly affected. Existing 'hielding around the pool (water depth and concrete walls) provids more than adequate protection, despite the slightly closer approach of the new racks to the walls of the pool. A recent analysis of the spent fuel pool water indicates that the evaporated activity is 3.19 x 10'3 pCi/ml [9.4.1]. During rerac. king operations, the activity concentration might be expect.ed to increase due to crud deposits spalling from spent fuel assemblies. While these effects may increase the concentrations, the pool cleanup system soon reduces the concentrations to the normal operating range. Experience to date has not indicated a major increase as a cca::equence of reracking. Fort Calhoun Station went through a reracking the last quarter of 1983. Operating experience has shown that there have been negligible concentrations of airborne radioactivity and no increases are expected as a result of the expanded storage capacity. Area monitors for airborne activities are available in the immediate vicinity of the spent fuel pool. In the pool, concentrations of 8 x 104 to 3 x 10-5 pCi/ml are typical [9.4.1]. Table 9.4.1 provides typical radionuclide concentrations in the pool water. There has been no significant increase in potential dose due to the buildup of crud along the 9-7
l l l sides of the pool. If crud buildup eventually becomes a major contributor to pool doses, measures will be taken to reduce such doses to As Low as Reasonably Achievable (ALARA). f No increase in radiation exposure to operating personnel is expected; therefore neither the current Radiation Protection program nor the area monitoring system needs to be modified. 9.5 Anticioated Exoosure Durino Rerackina All of the operations involved in reracking will utilize detailed procedures prepared with full consideration of ALARA principles. Similar operations have been performed in a number of facilities in i the past, which helps assure that reracking can be cafely and efficiently accomplished at the Fort Calhoun Station with minimum radiation exposure to personnel. The pool modification consists of replacing, in a predetermined sequence, the presently installed spent fuel storage racks with the higher density racks by simply lifting out the present ones and
- lowering the new ones in place.
f Any significant increase in personnel dosage due to- this modification beyond that incurred during regular refueling outages would be from the rack washing and diving processes. Underwater appurtenances shall be removed using remote handling tools to the-greatest extent possible. Diving operations may be required to remove certain underwater appurtenances. The dose rate to divers j would be minimized by placing fuel at a distance from cutting areas. Careful monitoring and adherence to pre-prepared procedures would assure that the radiation dose to the divers is maintained-ALARA. i 9-8
Below is an outline of the actions that will be taken to assure that occupational doses during each task of the pool modification will be ALARA: ,
- 1. A Radiation Protection technician will be present at all times during rack movements to monitor for excessive airborne or high radiation by utilizing portable radiation monitoring instruments.
- 2. Area Radiation Monitors will be used to alarm on a high radiation signal.
- 3. Personnel shall be required to wear appropriate clothing as determined by the Radiation Protection technician to preclude contamination.
- 4. As the racks are pulled out of the water, they will be washed (hydrolased).
- 5. All rack decontamination areas will be enclosed by a cover to reduce airborne contamination.
Total occupational exposure for the reracking operation is estimated to be between 5 and 10 person-rem (no dive operations), as indicated in Table 9.5.1. While individual task efforts and exposures may differ from those in Table 9.5.1, the total is believed to be a reasonable estimate for planning purposes. It may be necessary to use divers to remove certain underwater appurtenances. Careful monitoring and adherence to procedures will assure that the radiation dose to the divers will be maintained ALARA. The existing radiation protection program at Fort Calhoun is I adequate for the reracking operations. Where there is a. potential l for significant airborne activity, continuous air samplers will be l in operation. Personnel will wear protective clothing and, if necessary, respiratory protective equipment. Activities will be governed by a Radiation Work Permit, and personnel monitoring equipment will be assigned to each individual. As a minimum, this 9-9
i 4 i will include thermoluminescent - and pocket dosimeter. Additional ! personnel monitoring equipment (i. e. , extremity badges or alarming { dosimeters) may be utilized as required. i Work, personnel traffic, and the movement of equipment will be , monitored and controlled to minimize contamination and to assure that exposures are maintained ALARA. In reracking, the existing ! storage racks will' be' removed and temporarily stored (covered) ! prior to shipment as Low Level Radioactive Waste. i , 9.6 Conclusions l I Based upon the above evaluations, OPPD concludes that the likelihood of a cask drop accident resulting -in radionuclide !. releases is sufficiently small that this accident need not be; I considered. Also, if a highly unlikely SFP gate drop accident i
-should occur, the radionuclide releases no greater than those i postulated in a fuel handling accident would-result. Additionally, l a fuel handling accident would not result in radionuclide releases leading to offsite radiological consequences exceeding those of the fuel handling accident in the OPPD Final Environmental Statements.
Since there_ will- be a negligible change in radiological conditions due to the increased storage capacity of the spent fuel pool, no [ change is anticipated in the radiation protection program. In i' addition, the- environmental consequences of a fuel handling-L accident in the spent fuel pool, described in the-USAR, Section ! 14.18 remain unchanged. Therefore, there will be no change or-f impact to any previous - determinations . of OPPD . Final - Environmental-
- -Statements, 1972. Therefore
- OPPD- concludes that - the proposed l . modification is acceptable.
E t l l i h i' 9-10 b
- ,. . ..~ _ . __..._a,a_,,,-.,.....,..
_. _.. -.. ___ . . _ . . . . _ . . . . __ _ _ . ~ . . _ _ . _ . _ _ . _ . _ _ . _ . _ . .. _ . _ . ... .. i i j 9.7 References-for Sectior. 9 i^ (9.1.1) Letter - dated August 14, 1991, T. R. . Quay ! (NRC-NRR) . to W. - G. Gates- (OPPD) ,, " Review of i Fort Calhoun Revised Fuel Handling Accident in j- the Fuel Pool (TAC No.-80635)". i i (9.4.1) Fort Calhoun Station Chemistry Forms, FC-254, } " Monthly Chemical Analysis _Nov. 91-Apr. 92".' (9.5.1) Holtec Proposal, " Storage Capacity _ Expansion - Fort Calhoun Station", Vol. 1. Oct. 1991. e
,e 9-11 1
--e ww *-ms,w-n,e e m -- o-wr w r-wsr'
I 1 i I t Table 9.1.1 ] FORT CALHOUN FUEL ASSEMBLY GAS GAP ACTIVITY i Release Calculated 4 Isotope Fraction Activity (Ci) I-131 0.122 4.53 E 04 Xe-131m 0.064 9.41 E 02 Xe-133 0.029 2.00 E 04 , Kr-85 0.075 1.58 E 03 i Y 9-12
1 4 i i b Table 9.1.2 RADIOLOGICAL CONSEQUENCES OF A POSTULATED FUEL { HANDLING ACCIDENT 1 i Dose Type i At the EAB: Dose (Rem) ~ 10CFR100 Limit (Rem) 3 Thyroid 5.94< -300 i Skin- 0.191 --- Whole Body 0.072 25 !- At the LPZ: Thyroid 0.'106: -300 Skin 0.0034- ----
- Whole Body. 'O.0013 25 4
i These potential. doses are well within the exposure guideline
- values of 10CFR- Part 100,_ paragraph _ 11.
_ Independent
- calculations performed- by -the NRC- concurred- that the-radiological consequences were -- well within the~ guidelines -
specified in SRP 15.7.4. t i
'9-13
- -- ..,
- r 9 e M r.--y- .*y'tv
i i-i, a 4 k 4 i 4 I -Table 9.4.1 a e TYPICAL CONCENTRATIONS OF RADIONUCLIDES
~
. IN THE SPENT FUEL POOL WATER ~ i . Nuclide pCi/ml + l Ag-110M 4.63E-05 Co-58 6.87E-04
- Co-60 3.45E-05 l Cs-134 -9.24E-05 Cs-137 -5.76E-04' i-1 I
4 e i l1 4 1 !~ i 4 - L i 9-14
4 i l l 4 4 Table 9.5.1 (Ref. [9.5.1]) PRELIMINARY ESTIMATE OF PERSON-REM EXPOSURES DURING RERACKING a t I Estimate Step Number of Personnel Person-Rem Exp.(D i 4 Remove empty racks 5 0.5 to 1.0 [ Wash dacks 3- 0.08 to 0.2 Clean and Vacuum Pool 3 0.3 to 0.6 Remove. Underwater Obstructions (Diving). Partial installation j .new rack modules 5 0.25 to 0.5 l Move fuel to new re ks 2 0.8 to 1.5
- Remove remaining racks- 5 1.5 to 3.0 i
( Wash Racks 3 0.2.to 0.4 Install remaining new 5 ~ 0.4-to 0.8 racks: , Prepare old racks for 1.0 to'2.0 Q storage 4 l 4 Total Exposure,. i person-rem -5 to:10: t i
- Assumes minimum dose : rate of : 2 1/2 mR/hr - (expected) to a
, maximum of 5 mR/hr, except for pool vacuuming operations which assumes 4 to.8 mR/hr. m Estimated exposure, although- details of1 preparation and , packaging of old racks for shipment have'not been determined. d i. l f 9-15
--. , . -. . . - . - . - , . - . . . . .-....-.-n , .
- 10.0 BORAL* SURVEILLANCE PROGRAM 10.1 Purpose Boral*, the neutron absorbing material incorporated in the spent fuel storage rack design to assist in controlling system reactivity, consists of finely divided particles of boron carbide I
(B4 C) uniformly distributed in type 1100 aluminum powder, clad in type 1100 aluminum and pressed and sintered in a hot-rolling process. Tests simulating the radiation, thermal and chemical environment of the spent fuel pool have demonstrated the stability
- and chemical inertness of Boral [10.1.1-10.1.3). The accumulated 3 dose to the Boral over the expected rack lifetime is estimated to 4 be about 3 x 10" to 1 x 10" rads depending upon how the racks are used and the number of full core offloads that may be necessary.
Based upon the accelerated test programs, Boral is considered a j satisfactory material for reactivity control in spent fuel storage . racks and is fully expected to fulfil its design function over the j lifetime of the racks. Neverthelers, it is prudent to establish a surveillance program to monitor the integrity and performance of Boral on a continuing basis and to assure that slow, long-term synergistic effects, if any, do not become significant. Furthermore, the April 14, 1978 USNRC letter to all power reactor
- licensees (10.1.4], specifies that i
" Methods for verification of long-term material s
stability and mechanical integrity of special poison materials utilized for neutron
- absorption should include actual tests."
The purpose of the surveillance program is to characterize certain properties of the Boral with the objective of providing- data necessary to assess the capability of the Boral pancis in the racks to continue to perform their intended function. The surveillance program is also capable of detecting the onset of any significant degradation with ample time to take such corrective action as may be necessary. 10-1
4 In response to the need for a comprehensive Boral' surveillance program to assure that the suberiticality requirements of the
, stored fuel array are safely maintained, a surveillance progran has been developed incorporating certain basic tests and acceptance criteria. The Boral surveillance program depends primarily on representative coupon samples to monitor performance of the absorber material without disrupting the integrity of the storage
] system. The principal parameters to be measured are the thickness (to monitor for svelling) anc neutron attenuation (to monitor for
- the continued presence of boron in the Boral) .
10.2 Coupon Surveillance Procram 10.2.1 Coupon Descriotion q ) The coupon measurment program includes coupons susranded on a mounting (called a " tree"), placed in a designated cell, and l surrounded by spent fuel. Coupons will be removed from the array on a prescribed schedule and certain physical and chemical properties measured from which the stability and integrity of the Boral in the storage cells may be inferred. The coupon surveillance program will use one tree, with a total of ten test coupons.- Each coupon is mounted in - a stainless steel jacket, simulating as nearly as possible, the actual in-service geometry, physical mounting, materials, and flow conditions of the l Boral in the storage racks. The jacket (of the-same alloy used in manufacture of the racks) will be closed by screws to allow easy opening with minimum possibility of mechanical damage to the Boral specimen inside. In mounting the coupons on the tree, the coupons will be positioned axially within the central 8 feet of the fuel zone where the gamma flux is expected to be reasonably uniform. The coupons will be taken from the same lot as that used for construction of the racks. Each coupon will be carefully precharac-l terized prior to insertion in the pool to provide reference initial j values for comparison with measurements made af ter _ irradiation.
- As a minimum, the surveillance coupons will be precharacterized for
! 10-2 l l 1
i weight, dimensions (especially thickness) and neutron attenuation. In addition, two coupons (which need not be jacketed) wil] be ~ preserved as archive samples for comparison with subseifuent t.est coupon measurements. Wet chemical analyses of samples from the same lot of Boral will be available from the vendor for comparison. 10.2.2 Surveillance Coucon Testina Schedule To assure that the coupons will have experienced a slightly higher radiation dose than the Boral in the racks, the coupon tree is surrounded by freshly-discharged fuel assemblics at each of the first seven refuelings following installation of the racks. Beginning with the seventh load of spent fuel, the fuel assemblies will remain in place for the remaining lifetime of the racks. A sample coupon measurement schedule is shown in Table 10.1. At the time of the first fuel offload following installation of the coupon tree, the four storage cells surrounding the tree shall be loaded with freshly-discharged fuel assemblies that had been among the higher specific power assemblies in the core. At the scheduled test date, the coupon tree is removed and a coupon removed for evaluation. The coupon tree is then reinstalled to the same cell and, at teload, the surrounding cells are replaced with freshly-discharged fuel assemblies. This procedure is continued for the third, fourth, fifth, sixth and seventh offloading of spent fuel. From the seventh cycle on, the fuel assemblies ir. the fcur
- surrounding cells remain in place.
t Evaluation of the coupons removed will provide information of the effects of the radiation, thermal and chemical environment of the pool and by inference, comparable information on the Boral panels in the racks. Over the duration of the coupon testing program, the coupons will have accumulated more radiation dose than the expected lifetime dose for normal storage cells. l l 10-3 l I
Coupons which have not been destructively analyzed by wet-chemical processes, may optionally be returned to the storage pool and i remounted on the tree. The-/ will then be available for subsequent investigati a of defects, should any be found. 10.2.3 Measurement Procram
- The coupon measurement program is intended to monitor changes in physical properties of the Boral absorber material by performing the following measurements on the preplanned schedule:
- Visual Observation and Photography Neutron Attenuation
+ Dimensional Measurements (length, width and thickness) a Weight and Specific Gravity P.
The most significant measurements are thickness and neutron attenuation'. In the event loss of boron is observed or suspected, ' ! the data may be augmented by wet-chemical analysis (a destructive grav; metric technique for total boron only). 10.2.4 Surveillance Coupon Accentance criteria
- Of the measurements to be performed on the Boral surveillance coupons, the most important are '1) the neutron attenuation measurements (to verify the continued presence of the baron) and f (2) the thickness measurement (as a monitor of potential swelling) .
! Acceptance criteria for these measurements are as follows: , Neutron attenuation measurements are a precise instrumental l method of chemical analysis for Boron-10 -content using a ! nondestructive technique in which the percentage of thermal r neutrons uransmitted through the panal is measured and compared l with predetermined calibration data. Boron-10 is tne nuclide of ! principal interest since it is the isotope responsible for neutron absorption in the Boral panel. 10-4
i I A decrease of no more than 5% in Boron-10 i content, as determined by neutron attenuation, is acceptable. (This is tantamount to a
- requirement for no loss in boron within the j accuracy of the measurement.)
i An increase in thickness at any point should not exceed 10% of the initial thickness at that point. 1 Changes in excess of either of these two criteria requires investigation and engineering evaluation which may include early retrieval and measurement of one or more of the remaining coupons to provide corroborative evidence that the indicated change (s) is real. If the deviation is determined to be real, an engineering evaluation shall be performed to identify further testing or any corrective action that may be necessary. The remaining measurement parameters serve a supporting role and should be examined for early indications of the potential onset of Boral degradation that would suggest a need for further
- attentuation and possibly a change in enasurement schedule. These include (1) visual or photographic evidence of unusual surf 7e pitting, corrosion or edge deterioration, or (2) unaccountable weight loss in excess of the measurement accuracy.
l 10.2 References to Section 10 i (10.1.1] " Spent Fuel Storage Module Corrosion Report",
- Brooks & Perkins Report 554, June 1, 1977.
, [10.1.2] " Suitability of Brooks & Perkins Spent Fuel i Storage Module for Use in PWR Storage Pools", l Brooks & Perkins Report 57' July 7, 1978. [10.1.3] "Boral Neutron Absorbing / Shielding Material - Product Performance Report", Brooks & Perkins Report 624, July 20, 1982. (10.1.4] USNRC Letter to All Power Reactor Licensees, transmitting the "OT Position for Review and ! Acceptance of Spent Fuel Storage and Handling ! Applications", April 14,_1978. l l 10-5 i
N 4 i-4 i~ f i 1: l Table-10.1 j COUPON MEASUREMENT SCHEDULE l Couoon. Years (D
- l' 1 ,
4 i 21 2-i 3 4 i -4 -7 i 5 10 I j 6 15-7 20-I- 8 -25 i i 9- 30' i 10 35 I f r. l' (D The years - pertain .Ito those after - the . -installation of -: new i< racks. . 1 I i l j -- I l t- -
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11.0 ENVIRONMENTAL COST-BENEFIT ASSESSMENT 11.1 Introduction In early 1977, the Federal Government announced the indefinite suspension of the reprocessing of spent nuclear fuel to recover the unused fissile material. In 1977, the government also announced plans to acquire facilities for receipt and. storage of spent fuel-assemblies from light water reactors, but such plans have neither been implemented nor authorized. Therefore,- utilities with operating reactors have had to make other provisions for storing-spent fuel discharged from such reactors. In 1975, OPPD responded to this apparent shortfall in reprocessing with a decision to replace the original fuel storage racks in the spent fuel pool. _ In 1976, the original spent fuel racks were replaced with higher density storage racks which would provide adequate storage space on-site-in the spent fuel pool for all fuel to be discharged through 1985. Providing racks with storage capacity to handle fuel until 1985 appeared to be a prudent decision consistent with-industry practice at that time. In 1980 studies were conducted to find- available means for providing storage for spent fuel ~beyond 1985, since the Federal Government provided no means for implementation- of- assuming responsibility for spent fuel storage. Based on- the 1980 evaluation, it was concluded that reracking- the Fort Calhoun Station Unit No. 1 spent fuel-pool-again could; provide a maximum capacity to accommodate-spent fuel assemblies discharged'through 1994. The 1980 studies concluded that,- if available, rod consolidation could provide sufficient storage capacity-beyond.1994 to accommodate. all the spent- fuel likely to be discharged from Fort
-Calhoun-Station Unit No.-1 during'its useful operating lifetime.
However, since the Federal Government has still failed:to assume v 11-1
responsibility for spent fuel storage in lieu of reprocessing and acceptable commercial fuel consolidation capability is not yet demonstrated to be a viable technology, it was necessary for OPPD to once again consider storage options for spent fuel to be dischreged from Fort Calhoun Station Unit No. 1 beyond 1994. The current need to increase the existing storage capacity for the spent fuel at Fort Calhoun Station Unit No. 1 by 1994 is based on the regulatory requirement to maintain full-core offload capacity through the end of licensed plant life. OPPD conducted a study in 1990 for available spent fuel storage options. A solution was sought which would accommodate all the spent fuel likely to be discharged from Fort Calhoun until the expiration of its operating license in 2008, and through a license extension application request, if granted, to 2013, and which would provide the lowest overall cost to OPPD. Several onsite storage options which did not involve modifications to the present spent fuel pool or racks were considered during the evaluation. Use of portable casks to store fuel on site was considered, as well as use of non-portable casks or other above ground structures for dry storage. In the study, fuel rod consolidation was considered as both an immediate and longer term option. Fuel rod consolidation was not viewed by OPPD as a technology that can be relied upon for commercial use by 1994. Anticipated development.al work is expected to make fuel rod consolidation commercially acceptable to OPPD by the end of the 1990's, thus consolidation can be considered as a supplement to long range planning. I i l 11-2
l l 1 I other options considered in the study were reracking the spent fuel pool and use of the M uitored Retrievable Storage Facility (MRS). 4 The MRS is a Federally owned, off site storage facility planned by Dv6 to be available by 1998. Spent fuel would be shipped to the MRS for interim storage before final disposal. The MRS, if it were available, could satisfy storage requirements until the decommissioning of Fort Calhoun Station Unit No. 1, but does not address OPPD's storage needs after 1994. Reracking was viewed by OPPD as the best immediate solution for fuel storage which would have to be followed or augmented by either rod consolidation or dry , storage if the MRS is not available. Other fuel storage options that were considered but disqualified as impractical or excessively expensive included adding another spent fuel pool, shipping to another utility, and shipping to another country for storage and/or reprocessing. The 1990 study recommended proceeding with reracking the spent fuel pool based on lowest overall expected lifetime costs, and this recommendation was accepted by OPPD. The spent fuel pool will be reracked with free-standing, high density, poisoned fuel racks. The engineering design and licensing will be completed for a full reracking of the pool. The design will be able to accommodate consolidated fu n in sufficient quantity to reach end of life. Licensing requirements specific to actual storage of consolidated fuel will, however, be pursued at the time consolidation is chosen for future storage needs. 11.2 Proiect Cost Assessment The total capital cost for the rerack project is estimated'to be approximately $4.7 million. 11-3
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All the options previously descr %ed in the referenced study were evaluated for technical and ecc amic merit prior to proceeding with the raracking option. The rarack option had a slightly lower r expected net present value litatime cost than the dry storage E option. Although the options were considered relatively close in cost, the 1990 study concluded there was also uncertainty risk associated with licensing dry storage for Fort Calhoun Station Unit No.1 based upon current state regulatory issues. Figure 11.1 shows a decision tree of the scenarios evaluated with corresponding probab!11ty of success weighting. Figure 11.1 illustrates that there were two initial paths that could be pursued with future choices of varying uncertainty. The Rerack option with possible future alternatives (K tS , Fuel Rod Consolidation, Dry Storage) had an expected net present value cost of $26.1 million ($1990, 950 fuel stcrage cells). Proceeding directly to dry cask storage options yielded an expected net present value cost of $27.0 million ($1990, 950 fuel storage cells). The scenario which 4 considered 1050 fuel storage cells which is near the proposed design had an expected net present value cost of $21.8 million versus $23.9 million for dry storage. On a per kilogram basis, the expected cost of reracking first is
$5 5/FgU and that of dry cask storage first is $69/KgU ($1990, based upon 950 fuel storage cells).
The key variable was the number of storage spaces that could be provided by the rerack. The break even point was between 900 and 950 fuel storage spaces. The design exceeded the break even point by a significant margin (1083 fuel storage spaces). 11-4 l i............ . _ _ _ _ -..J
There are no acceptable alternatives for offsite spent fuel storage capacity for the Fort Calhoun Station. Currently there are no commercial independent spent fuel storage facilities operating in the U. S. The adoption of the Nuclear Waste Policy Act (NWPA) created a de facto throw-away nuclear fuel cycle. Since the cost of spent fuel reprocessing is not offset by the salvage value of the residual uranium reprocessing represents an added cost for the nuclear fuel cycle which already includes the NWPA Nuclear Waste Fund fees. In any event, there are no domestic reprocessing facilities. OPPD has no other operating nuclear power plant f acilities; therefore, shipment of spent fuel from the Fort Calhoun Station to other company nuclear power plants is not possible. Therefore, the selected alternative provides the flexibility to respond to changes in government or industry programs by providing for fuel storage in the conventional manner. 11.3 Environmental Assessment The anticipated maximum bulk pool temperature will not increase from the previously licensed 140*F expected limit, as detailed in the calculations described in Section 5.0 of this report. The resultant total heat-load (worst case) is 20.7 million BTU /HR. The net result of the increased heat loss and water vapor emission (due to increawed evaporation) to the environment is negligible and bound within the previous licensed limits. The rerack modification will be contained within the Auxiliary Building wt'.h houses the spent fuei pool. The rerack modification meets all current design criteria. Therefore, no adverse environmental impact is expected. 11-5
11.4 Conclusions As discussed earlier in this document, this facility modification will not have any significant impact on the environment, or the public health and safety In view of this and the favorable cost benefits described in this section, it is clear that the proposed modification is the most prudent and cost effective alternative available. , 11-6 _a
Fgure 11.1 Expected Net Present Value Cost Analysis ($1000)
- Base Case Rerack/MRS $4,326 ,
10% Probability Rerack/ Consolidate Reracking Attemative Drv Cask Storace $26,382
$26,059
! 40% Probability Rorack/ Dry Cask $30,148 l 50% Probability - e i Dry Cask /MRS $12,784
- 10% Probability I
$27,014 Dry Cask Storage Attemative f
1 i ' Dry Cask Storage $28,595
- 90% Probability i
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